ML16188A333
| ML16188A333 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 07/01/2016 |
| From: | Arizona Public Service Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 102-07277-MLL/GWA | |
| Download: ML16188A333 (180) | |
Text
{{#Wiki_filter:Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 7 Technical Analysis [NON-PROPRIETARY VERSION]
Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 7 TECHNICAL ANALYSIS TABLE OF CONTENTS
- 1. INTRODUCTION AND SUM MARY ........................................................................................ 4 1.1. Description of the NG F Design ......................................................................................... 4 1.2. NG F Qualification Testing ................................................................................................. 4 1.3. NG F Lead Fue I Assemblies ................................................................................................ 4 1.4. References ........................................................................................................................ S
- 2. M EC HAN ICAL DESIGN ....................................................................................................... lS 2.1. Fuel Assembly Structural Analyses (Non-Seismic/LOCA) ............................................... lS 2.2. Fuel Assembly Structural Analyses (Se ism ic/LOCA) ....................................................... 17 2.3. CEA Scram Time Analyses .............................................................................................. 17 2.4. Fuel Rod Analyses ........................................................................................................... 17 2.S. References ...................................................................................................................... 18
- 3. CORE DESIGN .................................................................................................................... 19 3.1. Introd u ctio n .................................................................................................................... 19 3.2. Method of Analysis ......................................................................................................... 19 3.3. Results ............................................................................................................................ 19 3.4. Summary I Conclusions .................................................................................................. 20 3.S. References ...................................................................................................................... 21
- 4. FUEL PE RFO RMAN CE ................................................................; ....................................... 32 4.1. References ...................................................................................................................... 33
- s. CORE TH ER MAL HYDRAU UC DESIGN ................................................................................ 3S S.l. Introd u ct ion .................................................................................................................... 3S S.2. Description of T-H Methods ........................................................................................... 3S S.3. NG F Assembly ................................................................................................................. 40 S.4. Hydra u lie Compatibility .................................................................................................. 41 S.S. Transition Core D NB Effect. ............................................................................................ 42 S.6. Effects of Fuel Rod Bow on DNBR .................................................................................. 43 S.7. Rods in DNB .................................................................................................................... 44 S.8. DNBR Calculations at Low Pressure Conditions ............................................................. 44 S.9. Conclusion ...................................................................................................................... 44 S.10. References ...................................................................................................................... 44 ATTACHMENT 7, Page 1
Enclosure Description and Assessment of Proposed License Amendment
- 6. FUEL ROD CORROSION ANALYSIS .................................*................................................... 47 6.1. References ...................................................................................................................... 47
- 7. NON-LOCA SAFETY ANALYSIS ............................................................................................ 48 7.1. UFSAR Section 15.1- Increase in Heat Removal by the Secondary System .................. 50 7.2. UFSAR Section 15.2 - Decrease in Heat Removal by the Secondary System ................. 53 7.3. UFSAR Section 15.3 - Decrease in Reactor Coolant System Flowrate ........................... 55 7.4. UFSAR Section 15.4- Reactivity and Power Distribution Anomalies ............................. 57 7.5. UFSAR Section 15.5 - Increase in Reactor Coolant System Inventory ........................... 61 7.6. UFSAR Section 15.6 - Decrease in Reactor Coolant System Inventory .......................... 61 7.7. UFSAR Section 15.7 - Radioactive Material Release from a Subsystem or Component 62 7.8. UFSAR Appendix 15.E - Limiting Infrequent Events (Loss of Flow from the Specified Acceptable Fuel Design Limit) ........................................................................................ 63 7.9. References ...................................................................................................................... 63
- 8. ECCS PERFORMANCE LOCA ACCIDENTS ................................................... ,....................... 64 8.1. Introduction ...................................................................................................................... 64 8.2. Objective ................................................................................................, ....................... 64 8.3. Regulatory Basis ...... ,....................................................................................................... 65 8.4. Method of Analysis ......................................................................................................... 65 8.5. References .................................................................................................................... 161
- 9. CONTAINMENT RESPONSE ANALYSIS ............................................................................. 165 9.1. Mass and Energy Release Analysis for Postulated Loss-of-Coolant Accidents ............ 165 9.2. Mass and Energy Release Analysis for Postulated Secondary .System Pipe Ruptures Inside Containment ....................................................................................................... 165 9.3. Mass and Energy Release for Containment Subcompartments .............................. ,... 165
- 10. RADIOLOGICAL SOURCE TERM EVALUATIONS ............................................................... 166
- 11. RADIOLOGICAL ACCIDENT EVALUATIONS ...................................................................... 167 11.1. Compliance with Regulatory Guide 1.25 ...................................................................... 167 11.2. References .................................................................................................................... 167
- 12. SETPOINTS ANALYSIS ...................................................................................................... 168' 12.1. References .................. :................................................................................................. 168
- 13. STRUCTURAL ANALYSIS ..................................................... .- ............................................. 169
- 14. DESIGN, SYSTEMS, AND COMPONENTS ANALYSIS ............. :........................................... 170 14.1. Fluid Systems Analysis .................................................................................................. 170 ATTACHMENT 7, Page 2
Enclosure Description and Assessment of Proposed License Amendment
- 15. OTHER ISSUES ................................................................................................................. 171 15.1. Other Technical Specification Considerations ............................................................. 171 15.2. End-of-Life Grid Crush Strength for NGF ...................................................................... 172 15.3. Spent Fuel Pool Criticality ............................................................................................ 173 15.4. Fukushima Orders ........................................................................................................ 173 15.5. References .................................................................................................................... 173
- 16. ACRONYMS ..................................................................................................................... 174 ATTACHMENT 7, Page 3
Enclosure Description and Assessment of Proposed License Amendment
- 1. INTRODUCTION AND
SUMMARY
The Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3 cores currently consist of 241 System 80 fuel assemblies. The current Westinghouse supplied PVNGS fuel design (Value Added Fuel, STD; referred to as Standard (STD fuel throughout), consists of 236 fuel rods, four outer guide tubes, one center/instrument guide tube, an lnconel top grid, and lnconel bottom grid, 9 Zircaloy mid grids, and upper and lower end fittings. The rods are arranged in a square 16 x 16 array. The proposed change will allow for the implementation of Combustion Engineering 16x16 Next Generation Fuel (CE 16x16 NGF; described as NGF throughout), including the use of Optimized ZIRLO' fuel rod cladding material. This change is to be implemented commencing with the fall 2018 refueling outage (2R21). . 1.1. Description of the NGF Design The key features of the NGF design are sh.own in Figure 1-1, with planned grid design changes for use at PVNGS. Table 1-1 provides a descriptive comparison of the current PVNGS STD fuel design versus the PVNGS NGF design. Tables 1-2 and 1-3 provide additional comparison data for the STD and NGF designs. Figures 1-2 and 1-3 compare key axial dimensions of the NGF fuel assembly and fuel rod design with the STD fuel assembly and fuel rod design, which will be co-resident in the PVNGS Units 1.* 2 and 3 cores during fuel transition. 1.2. NGF Qualification Testing A fuel design must meet a significant number of design and performance criteria. Comprehensive testing is a vital part of confirming that the design meets or exceeds design criteria. A series of tests was conducted to verify that the NGF design's mechanical performance and compatibility, as well as its thermal-hydraulic performance, are acceptable. These tests were performed for introduction of NGF at Waterford 3 and Arkansas Nuclear One - Unit 2 (which are also CE NSSS 16x16 plants) and remain applicable to PVNGS. In addition to the testing performed for previous applications of NGF at the other CE-NSSS 16x16 plants, specific testing was performed to provide data to be used to verify that NGF with mid and intermediate flow mixing (IFM) grids fabricated with modified outer straps (MOS) [ ]avwith added outer strap tabs (OST) met established criteria and to ensure safe and reliable fuel performance. 1.3. NGF Lead Fuel Assemblies In support of the proposed change, PVNGS undertook an evaluation of the NGF design by installing NGF lead fuel assemblies (LFA). Eight NGF LFAs were fabricated and introduced at PVNGS Unit 3 in Cycle 16 (References 1.1, 1.2 and 1.3). LFA inspections were performed at PVNGS Unit 3 at end-of-cycle 18 (EOC-18) after completing their third cycle of operation. Inspection parameters included fuer assembly visual examinations, assembly length measurements, shoulder gap measurements, rod bow measurements, guide tube eddy current testing and fiberscope visual inspection and peripheral rod oxide measurements. Fuel assembly visual examinations showed no mechanical or structural anomalies. Shoulder gaps were normal with very uniform positioning of the fuel rods. Clad corrosion appearance ATTACHMENT 7, Page 4
Enclosure Description and Assessment of Proposed License Amendment was normal. As planned, fuel assembly length measurements were performed on four of the eight LFAs. The maximum assembly growth was approximately [ ]a,c inches. This agreed very well with assembly growth data collected at the other PVNGS units on non-NGF fuel. Shoulder gap measurements were performed on four LFAs and showed that the average rod growth of the Optimized ZIRLO' fuel rods agreed very well with the ZIRLO rod growth database and trended well with previous Optimized ZIRLO' rod growth data. Rod bow measurements were performed on one LFA and results compared favorably to channel closure and standard deviation limits. Guide tube eddy current testing (ECT) on four LFAs showed insignificant wall wear. Peripheral rod oxide thickness measurements showed considerable margin to the limit. 1.4. References 1.1. APS Letter to NRC, Palo Verde Nuclear Generating Station (PVNGS) Unit 3; Docket No. STN 50-530; Request for Temporary Exemption from the Provisions of 10 CFR 50.46 and 10 CFR 50, Appendix K for Lead Fuel Assemblies, November 2, 2009 (ADAMS Accession No. ML093160596) 1.2. APS Letter to NRC, Palo Verde Nuclear Generating Station (PVNGS) Unit 3; Docket No. STN 50-530; Response to Request for Additional Information Regarding the Request for Temporary Exemption from the Provisions of 10 CFR 50.46 and 10 CFR 50, Appendix K for Lead Fuel Assemblies, May 12, 2010 (ADAMS Accession No. ML101410262) 1.3. NRC Letter to APS, Palo Verde Nuclear Generating Station, Unit 3; Temporary Exemption from the Requirements of 10 CFR Part 50, Section 50.46 and Appendix K (TAC No. ME2590}, August 26, 2010 (ADAMS Accession No. ML101900254) ATTACHMENT 7, Page 5
Enclosure Description and Assessment of Proposed License Amendment Figure 1-1: Key NGF Design Features Antl-RotaUan Gulde Tube Jofnt Axial BlanketJ
- Improved fuel cycle ccoocmic1 by reducing ncutroo leakage from the top Reduced Rod Dem Incanel Top Grid and bottan of core.
- Reduced spring forces to minimize potential for fuel rod bow. Optimized ZIRLQTll daddlna
- Improved fuel ITlllIJJin* in corrosioo, rod I-Sprln1 Mid Grid Deslp intc:mal presrurc, post-LOCA clad oxidation
- Advanced grid design fir improved grid-
- Leads to increase peaking factors , less feed to-rod fretting rcsistmu:c. assemblies, enables plant uprating, extended
- Sidc-rupportcd wnes fer thc:rmal margin cycle length.
md fuel rod loading aid Intermediate Flow Mlxln1 (IF'M) Grids InteKral Fuel Burnable Absorber (IFBA) (not shown)
- ZrB 1 burnable absorber for no residual absorber
- Two (2) IFM grids for additiooal DNB penalty leading to improved fuel cycle md LOCA margin md improved crud and ccooanics. Significant cycle 1avings vcnu1 corrosioo resislllru:c. Gadolinium md F.rbia.
Lem Tbt ZIRLQTllMJd Grids GuardlanTM Debris Resistant Bott0111---~
- Lower structural ccrrosioo md hydrogen Grid pickup.
- Proven qicrating history of excellent
- Lower fuel assembly growth at high bum-up; debris protection in CE md System80 lower grid growth.
rcact<n.
- Lawer control clement insmioo drag fcrcc at high bum-up STD (HID1 L) versus NGF Grid Comparison HID-1L MOS/OST/[ ]3*c Vaned ATTACHMENT 7, Page 6
Enclosure Description and Assessment of Proposed License Amendment Figure 1-1: Key NGF Design Features (continued) Notable Design Features on the NGF Grids Dimples replaced by I-spring Current Outer Strap Design Modified Outer Strap Design
+
Current NGF Outer Strap and Modified Outer Strap (MOS) Designs Additional Outer Strap Tab (OST) ATTACHMENT 7, Page 7
Enclosure Description and Assessment of Proposed License Amendment Figure 1-2: 16x16 STD versus NGF Design for PVNGS Units 1, 2 and 3 a.c ATTACHMENT 7, Page 8
Enclosure Description and Assessment of Proposed License Amendment Figure 1-3: Comparison of 16x16 Current STD and NGF Fuel Rod Designs a,c ATIACHMENT 7, Page 9
Enclosure Description and Assessment of Proposed License Amendment Table 1-1: Comparison of PVNGS 16x16 STD and NGF Designs PVNGS 16x16 STD PVNGS NGF Feature Feature Description Feature Description Top Grid lnconel 625 Straight Strip lnconel 718, Plus 7 design grid with L-shaped outer with corner weld outer strap strap Top Nozzle PVNGS STD Nozzle See next item Top Nozzle N/A Ears added to guide tube Anti-Rotation Joint flange & keyways in flow plate Top Guide Thimble (GT) Zr-4 Welded Zr-4 bulged Flange Joint Mid Grids Zr-4 Wavy grid, no mixing Low-Tin ZlRLO' "I" spring vanes, alternating rod grid with Side Supported supports Vanes on selected grids, alternating rod supports & ZlRLO' sleeves lFM Grids Not Used Low-Tin ZlRLO' non-contacting dimples grid with ZI RLO' sleeves Mid & IFM Grid Outer Strap Zr-4 L-shaped Low-Tin ZIRLO' L-shaped Design (IFM N/A) Top, Mid & lFM Grid to GT Welded Sleeves Bulged to GT Joints Outer GTs SRA Zr-4 GTs with dashpot SRA ZlRLO' GTs with and provision for Inner Wear dashpot, but no *provision for Sleeve & welded to grids Inner Wear Sleeve. Bulged to grid sleeves Instrument Tube Zr-4 GTs with dimples for SRA ZlRLO' GTs with ICI centering & welded to dimples for ICl centering & grids bulged to top grid sle_eve only Guardian' Grid and Joint lnconel 625 grid with skirt Same, but use perimeter with Bottom Nozzle strap scalloped on both ends, no welding to lower nozzle and add sleeves in GT holes Bottom Nozzle PVNGS STD Nozzle Same, except features for welding Guardian' grid not required, and bosses added
/
for thimble crimp screws Fuel Rod 0.382 in. ZIRLO rod with W 0.374 in. Optimized STD Plenum Spring and ZI RLO TM rod with low Guardian' solid end cap volume plenum spring and Guardian' solid end cap ATTACHMENT 7, Page 10
Enclosure Description and Assessment of Proposed License Amendment Table 1-2: Comparison of PVNGS NGF and STD and NGF Design Data a,c ATTACHMENT 7, Page 11
Enclosure Description and Assessment of Proposed License Amendment Table 1-2: Comparison of PVNGS NGF and STD and NGF Design Data a,c ATTACHMENT 7, Page 12
Enclosure Description and Assessment of Proposed License Amendment Table 1-2: Comparison of PVNGS NGF and STD and NGF Design Data a,c ATTACHMENT 7, Page 13
Enclosure Description and Assessment of Proposed License Amendment Table 1-3: Comparison of 16x16 STD and NGF Fuel Rod Design Data a,c ATTACHMENT 7, Page 14
Enclosure Description and Assessment of Proposed License Amendment
- 2. MECHANICAL DESIGN The NGF mechanical design for PVNGS Units 1, 2 and 3 was assessed against specific design criteria to verify that it is within the design bases for the original cores. This assessment took into consideration changes to regulatory requirements, the intended operating conditions of the fuel, and whether analytical models or material correlations had changed from the original core design. The NGF design is developed in accordance with Reference 2.1 and Reference 2.2.
NGF analyses were performed to verify that all applicable design criteria were satisfied for the LFA irradiation project. With the introduction of the Modified Outer Strap (MOS), [ ]8 'c, and Outer Strap Tab (OST) NGF grid design features, as well as now considering full PVNGS cores of NGF, several analyses required updating, including:
- The fuel assembly structural integrity analyses during non-seismic/LOCA events, due to the changes in full region effects that were not addressed in PVNGS LFA analyses, grid design changes (including MOS/OST/[ ]8 'c) and core operating conditions;
- The fuel assembly structural integrity analyses during seismic/LOCA events, due to the changes in full region effects that were not addressed in PVNGS LFA analyses, grid design changes (including MOS/OST/[ ]8 'c), changes in fuel assembly properties that occur during operation, and core operating conditions;
- The CEA scram time and CEA structural analyses, due to the change in scram loads resulting predominantly from the NGF assembly's increased pressure drop;
- The fuel rod analyses (stress, strain, fatigue, clad flattening, maximum internal pressure), due to changes in the operating conditions.
2.1. Fuel Assembly Structural Analyses (Non-Seismic/LOCA) All design criteria are satisfied for full region implementation of NGF. Evaluations were conducted to ensure applicability to full regions by making appropriate adjustments to the previous LFA analyses to account for differences between the operating conditions and grid design changes (MOS/OST/[ ]8 'c) of the PVNGS LFAs and planned full regions. The evaluations of the criteria are summarized below.
- Fuel Rod Growth: Since the LFA growth prediction remains applicable and the fuel rod growth does not change, the minimum shoulder gap is unchanged. Therefore, this criterion is met for the full region NGF with grid design changes (MOS/OST/[ ]8 'c).
- Stress Limits: The limiting loadings (5000 lbf tensile load and reconstitution) are non-operational and unaffected by full region NGF implementation. Additionally, the only non-seismic/LOCA operational load (holddown spring force) calculated for LFAs was shown to remain applicable for full region NGF application. Therefore, this criterion is met for the full region with grid design changes (MOS/OST/[ ]8 'c) as the stress margins remain unchanged from LFA results.
- Tensile Load Requirement: The 5000 lbf tensile load is non-operational and unaffected by full region NGF implementation. Therefore, this criterion is met for the full region NGF with grid design changes (MOS/OST/[ ]a,c).
ATTACHMENT 7, Page 15
Enclosure Description and Assessment of Proposed License Amendment
- Interface Criteria - Reactor Internals:
- Axial Engagement of Nozzles: The only change due to the full region implementation of NGF with the grid design changes is a [ ]a,e in differential thermal expansion between the fuel and the internals. Since the core insert pin height of 3.375 inches is greater than the maximum possible axial movement of [ ]a,e inches, the fuel assembly remains engaged with the core structures for all operating conditions. Therefore, the design criterion is satisfied for full region implementation of NGF with grid design changes (MOS/OST/[ ]a,e). - Provision for Fuel Assembly Growth: The PVNGS LFA growth prediction remains applicable and the differential thermal expansion between the fuel and the reactor internals increases clearance to the upper guide structure and holddown spring solid height. Therefore, the design criteria are satisfied for full region implementation of NGF with grid design changes (MOS/OST/[ ]a,e). - Loads on Reactor Internals: Holddown spring forces calculated for the PVNGS LFAs remain applicable. Additionally, fuel uplift margins calculated for the PVNGS LFAs also remain applicable. Therefore, the design criterion is satisfied for full region implementation of NGF with grid design changes (MOS/OST/[ ]a,e).
- Interface Criteria - Compatibility with Co-resident Fuel Assemblies: The pressure drop associated with full region NGF is between those of the STD fuel and the PVNGS LFAs. As a result, the design criteria are satisfied for full region implementation of NGF with grid design changes (MOS/OST/[ re), since the higher pressure drop difference evaluated for the PVNGS LFAs is conservative for these criteria.
- Holddown Force: The holddown spring forces are unchanged, and the uplift forces are slightly conservative when compared to LFA results. Therefore, this criterion is satisfied.
- Spring Stress: The stresses are unaffected by any assembly growth differences since the PVNGS LFA calculations evaluated the stress with the spring compressed to its solid height. Therefore, the design criterion is satisfied for full region implementation of NGF.
- CEA Impact Velocity Interface: The CEAs impact on the Fuel Alignment Tube Sheet, not the fuel assembly, so no CEA impact forces are transmitted to the NGF or STD assembly. Therefore, this criterion is satisfied.
- Instrument Tube/Top and Bottom Nozzle Interface: The [ re clearance calculated for the PVNGS LFAs is [ re in assembly growth calculated for full region implementation of NGF with grid design changes (MOS/OST/[ ]a*°). Thus, the gap calculated in the analysis of record is adjusted by
[ re. The criterion for this interface is only that a gap be shown to exist. As such, this criterion is satisfied.
- Top Nozzle Alignment and Engagement: Because the assembly growth remains applicable, the only adjustment to the PVNGS LFA engagements is a [
]a,e due to the [ ]a,e differential thermal expansion between the fuel and the reactor internals. The adjusted engagements are [ re, which exceed their respective required engagements of [
re. Therefore, the criterion is met for full region implementation of NGF with grid design changes (MOS/OST/[ re). ATTACHMENT 7, Page 16
Enclosure Description and Assessment of Proposed License Amendment 2.2. Fuel Assembly Structural Analyses (Seismic/LOCA) The methodology of CENPD-178-P, Revision 1-P (Reference 2.3) was used to calculate the seismic and Branch Line Pipe Break (BLPB) response loads on the NGF assemblies and spacer grids. Analysis demonstrates that the effect of operation on spacer grid strengths during seismic/LOCA events is satisfied for all the transition cores and a full core of NGF. This analysis considers the effects of shroud-adjacent residence on fuel assembly properties. Spacer grid strengths are greater than the predicted impact loads that occur during a combined seismic and LOCA event except some peripheral STD assemblies in transition cores, which are justified via coolability and insertability analyses. The coolability analysis considered maximum hypothetical deformation and determined that with no reduction of the assembly flow area due to grid deformation and hot channel flow area blockages due to grid deformation ranging from 0% blockage to a maximum of [ ]% blockage, there was no penalty on the ECCS performance analysis results for PVNGS Units 1, 2 and 3 for NGF and NGF transition. Therefore, the grid strengths of the NGF and STD fuel are acceptable when evaluated against the seismic/LOCA loads predicted for PVNGS Units 1, 2 and 3. The NGF assemblies were evaluated with the co-resident STD fuel assemblies and the CEAs, under transition core seismic and LOCA conditions to demonstrate compliance with the applicable design criteria. The overall conclusions are that the loadings and associated stresses for the fuel designs and the CEAs satisfy their respective design criteria. The evaluation of fuel assembly structural response to externally applied forces (i.e., End-of-Life [EOL]) grid crush strength for NGF) is addressed .in Section 15.2. 2.3. CEA Scram Time Analyses Scram times are acceptable for NGF in PVNGS Units 1, 2, and 3. Scram times were calc.ulated for each unit for a full core of both STD and NGF. The scram times for the full NGF cores are slightly [ ]a,c than for the full cores of STD fuel, [ re within the limits. 2.4. Fuel Rod Analyses The fuel rod design analyses are broken up into two separate sets of calculations; the Mechanical Performance Analysis, and the FATES3B-based Fuel Performance Analysis discussed in Section 4 of this Technical Analysis. The Mechanical Performance analysis demonstrates the NGF's ability to meet the cladding collapse, cladding fatigue, cladding stress and cladding strain criteria and has no downstream impacts. The FATES3B analysis confirms the rod internal pressure, fuel melt and spent fuel pool rod internal pressure criteria are satisfied while also providing a number of downstream groups with inputs. The Mechanical Fuel performance analysis was performed for NGF and STD designs to show that the cladding collapse, cladding fatigue, cladding stress and cladding strain licensing criteria were satisfied. The calculations in this analysis are performed on a bounding basis using conservative inputs related to core physics and plant conditions, while explicitly accounting for the geometry and material properties of the fuel design. Prior to NGF, the fuel cladding burst acceptance criteria were met for all Combustion Engineering Fleet ZrB 2 fuel types in Reference 2.4. This analysis determined that cladding burst is precluded if peak cladding temperature and engineering hoop stress remain below the values for cladding rupture depicted in Section 4.4.2.1 of Reference 2.4. To support the ATTACHMENT 7, Page 17
Enclosure Description and Assessment of Proposed License Amendment implementation of the NGF design an analysis was performed to demonstrate that the Reference 2.4 conclusions remained valid and bounding (i.e., that the cladding temperature and engineering hoop stress remain below the burst limit for the implementation of NGF). The revised analysis started with the analysis of record and revised the model to support the implementation of NGF. The fuel cladding burst analysis results are well below the limits and demonstrate that the acceptance criteria are satisfied. 2.5. References 2.1. WCAP-16500-P-A, Revision 0, CE 16x16 Next Generation Fuel Core Reference Report,August2007 2.2. WCAP-16500-P-A, Supplement 2, Evolutionary Design Changes to CE 16x16 Next Generation Fuel and Method for Addressing the Effects of End-of-Life Properties on Seismic and Loss of Coolant Accident Analyses, June 2016 2.3. Combustion Engineering Topical Report CENPD-178-P, Revision 1-P, Structural Analysis of Fuel Assemblies for Seismic and Loss of Coolant Accident Loading, August 1981 2.4. CENPD-404-P-A, Revision 0, Implementation of ZIRLO Cladding Material in CE Nuclear Power Fuel Assembly Designs, November 2001 ATTACHMENT 7, Page 18
Enclosure Description and Assessment of Proposed License Amendment
- 3. CORE DESIGN 3.1. Introduction This section describes the neutronics modeling of NGF implementation at PVNGS. The immediate preceding fuel cycle is denoted as Cycle N-1. Cycle N-1 consists entirely of the CE 16x16 standard fuel design with erbia burnable absorbers. The Cycle N-1 fuel management scheme is presented in Figure 3-1. Four representative cycles were developed to evaluate the transition to a full core of NGF: a first transition cycle (Cycle N) where the feed fuel is all NGF; a second transition cycle (Cycle N+1) where the core contains two Batches of NGF fuel with the center assembly and core periphery being standard fuel; a full core NGF cycle (Cycle N+2) where the core is all NGF except the center assembly; and a "near equilibrium" cycle (Cycle N+3) where the core is all NGF except the center assembly.
3.2. Method of Analysis The evaluations and assessments of the PVNGS core design entailed the development of explicit neutronics models for the representative NGF fuel cycles. The presence of the NGF fuel in the PVNGS core was explicitly incorporated into these models, including the specific NGF geometry and associated nuclear cross sections, and the use of Zirconium Diboride (ZrB 2 ) integral fuel burnable absorber (IFBA). The calculations supporting the implementation of NGF were performed using NRC approved codes and nuclear design methodology (References 3.1, 3.2 and 3.3). The features of the NGF design are within the demonstrated range of those methodologies. Typical PVNGS core reloads are 92 to 108 fresh feed assembly reloads. The PVNGS reactor core has an odd number of fuel assemblies (i.e., 241) and on occasion the center assembly is re-inserted from the spent fuel pool. After transition to NGF fuel, the need to use an assembly from the pre-NGF design for the core center assembly may be necessary or desired. This assembly is typically high burnup and not limiting with respect to power peaking and thermal performance. Use of this type of center assembly, i.e., pre-NGF, will not be analyzed as a mixed core. For all non-physics analyses and for all administrative purposes, this type of core will be considered a full core of the new fuel type. However, the core physics analysis will specifically model this center assembly to ensure that the fuel pin burnup and fiuence limitations are not exceeded and the peak integrated radial peaking factor for this assembly
- will be maintained at 0.95 or less of the core maximum integrated radial peaking factor at all times in core life to ensure that this assembly does not become limiting during cycle operation.
3.3. Results The representative fuel management schemes for Cycles N, N+1, N+2, and N+3 are shown in Figures 3-2, 3-3, 3-4, and 3-5, respectively. Note that in each of these cycles, as a calculational convenience, the center assembly is assumed to be a twice-burned standard design discharged Region H assembly reinserted from the spent fuel pool (SFP). In reality, after the supply of such assemblies is exhausted, the center assembly would be either a different discharged (standard design) assembly from the SFP, a twice-burned assembly from the previous cycle (provided the peak pin burnup limit was met) or an extra feed assembly for use as the center assembly for two cycles. None of the*se realistic alternatives to the notional Region H assembly would invalidate any of the analyses and conclusions documented here and elsewhere in this report. ATTACHMENT 7, Page 19
Enclosure Description and Assessment of Proposed License Amendment From a neutronics modeling perspective, the primary differences between the NGF and standard fuel assembly are the incorporation of the smaller diameter fuel pellet and rod, a higher poisoned pellet stack density, and the use of IFBA as the burnable absorber. Due to the smaller diameter fuel pellet and rod, the water-to-fuel ratio of the NGF fuel is slightly greater than that of the standard fuel assembly, making the NGF fuel more reactive, all other things being equal. However, this effect is minor compared to the reactivity differences between fresh and burned fuel, thus it has little impact on the fuel management process. Another consequence of the softer neutron spectrum of the NGF design is, all other things being equal, a slight increase in MTC. Again, this does not fundamentally alter the fuel management process since the MTC is readily controlled by use of the appropriate number of burnable absorber pins. Due to the use of ZrB 2 burnable absorber, the RCS soluble boron concentration increases during approximately the first 100 EFPD of a fuel cycle's initial operation. This is due to the ZrB2 burnable absorber depleting faster than the fuel depletes. Consequently, the fuel cycle's most positive, or least negative, MTC will not occur at beginning-of-cycle, but at the point of maximum RCS soluble boron concentration, or approximately 100EFPD. *The Nuclear Design Process will confirm that the COLR MTC Limits are validated over the entire fuel cycle.
- The Nuclear Design Process assures meeting the fundamental design limits such as most-positive MTC, power peaking factors, and 1-pin burnup. These limits are achieved by a combination of the number of feed assemblies, the use of zone-enrichment within the assembly, and the use of spatially-varying the amount of ZrB 2 burnable absorber. Figure 3-6 compares the boron versus cycle length (letdowri) curves for the representative NGF fuel cycles and compares them to the standard fuel pre-NGF fuel cycle. The NGF fuel cycles display the typical boron increase during the initial 100 EFPD associated with ZrB2 burnout.
Figures 3-7 through 3-10 compare the Fr, Fxy, Fq, and Fz peaking factors, respectively, for the representative NGF cycles with those of the standard fuel pre-NGF fuel cycle. Generally, the cycle maximum peaking factors are slightly larger for the NGF cycles as compared to the standard fuel pre-NGF fuel cycle. 3.4. Summary I Conclusions From a core physics viewpoint, the transition from Standard 16x16 to NGF 16x16 fuel can be accomplished, while still meeting the current fundamental design limits such as most-positive MTC, power peaking factors, and 1-pin burnup. Cycles using NGF fuel are expected to
- produce similar physics inputs to Chapter 15 events, so that with few exceptions, the current physics parameter limits can be validated. Detailed physics data characteristic of representative NGF fuel cycles was provided to downstream engineering groups, such as Fuel Performance, LOCA, and non-LOCA Transient Analysis, to assist them in confirming the viability of NGF fuel. The cycle-specific physics characteristics of NGF cycles will be calculated each cycle and validated against the bounding physics data used in the Analyses-of-Record, such as the Fuel Performance, non-LOCA Transient, and LOCA Analyses.
Parameter values that are not validated are resolved in the Nuclear Design Process using NRC approved methods, including, but not limited to performing cycle-specific analyses, or performing a new Analysis-of-Record. ATTACHMENT 7, Page 20
Enclosure Description and Assessment of Proposed License Amendment 3.5. References 3.1. Letter USN RC to Arizona Public Service Company, Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2, and 3 - Issuance of Amendments on CASM0-4/SIMULA TE-3 (TAC Nos. MA9279, MA9280, and MA9281), March 20, 2001 3.2. WCAP-16072-P-A, Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs, August, 2004 3.3. CENPD-266-P-A, The ROCS and D/T Computer Codes for Nuclear Design, April 1983 ATTACHMENT 7, Page 21
Enclosure Description and Assessment of Proposed License Amendment G Figure 3-1: Cycle N-1 Pre-NGF Standard Fuel Design
°"BB =*Fuel Q"'" "'*(QC) Lo~**,
Batch Identifier xx = QC Location (Previous Cycle) yy = Number of 90° Rotations 1 Tl 33 2 22R2 Tl 3 41R2 T2 4 Tl 49 5 36R2 6 28 7 8 9 10 T2 R2 V2 V3 V3 V4 11 61R3 12 13 14 8 15 16 9 17 T3 Vl V2 U2 V6 U2 V6 18 14R2 19 20 21 24 22 23 31 24 25 67R3 T2 Vl vs U4 vs U3 vs U3 26 21 27 28 55 29 30 29R3 31 32 15R3 33 12R2 R2 V2 U4 vs U4 V6 U4 Ul 34 42Rl 35 36 44 37 38 39 39 19R2 40 7R3 41 42 20R2 Tl V2 U2 vs U4 Ul Ul V6 U3 43 37R2 44 45 46 47 47 48 35 Rl 49 59 50 13R2 51 Tl V3 V6 U3 V6 Ul U3 U2 V6 52 57R2 53 54 53 55 56 45Rl 57 58 27R2 59 60 62R3 T2 V3 U2 vs U4 V6 U2 V4 U2 61 49R3 62 63 64 51Rl 65 12Rl 66 20Rl 67 68 lORl 69 17 Tl V4 V6 U3 Ul U3 V6 U2 HS
*** reinsert from EOC N- 11
**** reinsert from EOC N-4 Feed Batch Characteristics Burnable Sub- Number of Absorber Enrichments batch Assembl ies No. Pin s - w/o U-235 Type V1 8 12-erbia 4.63/4.33/4.03 V2 16 28-erbia 4.63/4.33/4.03 V3 16 36-erbia 4.63/4.33/4.03 V4 8 76-erbia 4.63/4.33/4.03 V5 24 84-erbia 4.63/4.33/4.03 V6 32 92-erbia 4.63/4.33/4.03 Total 104 6688 4.23 ATTACHMENT 7, Page 22
Enclosure Description and Assessment of Proposed License Amendment Figure 3-2: Cycle N First NGF Transition Cycle ~ ""
- Fuel Q"'" "" (QC) "'~""
1 14R2 2 40R3 3 17 4 29 BB = xx yy Batch Identifier
= QC Location (Previous Cycle)
= Number of 90° Rotations U2 Ul V6 vs 5 33 6 45R2 7 8 9 10 Ul V6 Wl W2 W2 W4 11 68 12 13 14 7R2 15 16 24R3 17 U2 Wl W2 V2 W4 vs ws 18 39 19 20 21 8R2 22 23 53R3 24 25 59Rl Ul Wl W3 V3 ws V3 ws V4 26 15R2 27 28 44R2 29 30 31 31 32 27R3 33 12R2 V6 W2 V3 ws V6 ws V2 Vl 34 36R2 35 36 35R2 37 38 47 39 19R2 40 37Rl 41 42 20R2 U2 Wl V2 ws V6 Vl vs ws vs 43 48Rl 44 45 46 9Rl 47 48 22R3 49 50 41 51 Ul W2 W4 V3 ws vs W4 V6 W3 52 67 53 54 55Rl 55 56 13Rl 57 58 57 59 60 62R3 V6 W2 vs ws V2 ws V6 W3 V4 61 29R3 62 63 64 59 65 12Rl 66 20Rl 67 68 lORl 69 39 vs W4 ws V4 Vl vs W3 V4 HS
*** reinsert from EOC N-11 Feed Batch Cha racteristics Burnable Sub- Number of Absorber Enrichments batch Assemblies No. Pins - wlo U-235 Type W1 16 32-ifba 4.33/4.03/3 .73 W2 24 60-ifba 4.33/4 .03/3.73 W3 12 80-ifba 4.03/3. 73/3.43 W4 16 88-ifba 4.03/3.73/3.43 W5 40 100-ifba 4.03/3.73/3.43 Total 108 8320 3.90 average ATTACHMENT 7, Page 23
Enclosure Description and Assessment of Proposed License Amendment Figure 3-3: Cycle N+1 Second NGF Transition Cycle ~ oo "Q""'" Co" (QC) U>~t*o 1 6R2 2 4SR3 3 37R2 4 61R3 BB = Fuel Batch Identifier xx = QC Location (Previous Cycle) yy = Number of 90° Rotations V6 W4 ws vs 5 3R2 6 67Rl 7 8 9 10 V6 W3 Xl X2 X2 X3 11 39 12 13 14 44R2 15 16 24Rl 17 Vl Xl X2 W2 X4 ws X4 18 52R2 19 20 21 27R3 22 23 41 24 25 20R3 V6 Xl X4 W2 X4 ws X4 W3 26 59R3 27 28 13Rl 29 30 63 31 32 35 33 19R3 W3 X2 W2 X4 ws X4 Wl Wl 34 26R2 35 36 8R2 37 38 29 39 40 47R3 41 42 49R2 V6 Xl W2 X4 ws X3 ws X4 W4 43 lSRl 44 45 46 57 47 48 31Rl 49 12R2 so 9R2 51 W4 X2 X4 ws X4 ws Wl W2 X4 52 22R2 53 54 SSR3 55 56 7 57 58 53R2 59 60 62R3 ws X2 ws X4 Wl X4 W2 X4 W4 61 61R2 62 63 64 20R2 65 19R2 66 49Rl 67 68 62R2 69 39R2 vs X3 X4 W3 Wl W4 X4 W4 HS
*** reinsert from EOC N-11 Feed Batch Characteristics Burnable Sub- Number of Absorber Enrichments batch Assemblies No. Pins - wlo U-235 Type X1 16 32-ifba 4 .58/4 .28/3 .98 X2 24 60-ifba 4 .58/4.28/3.98 X3 8 88-ifba 4.28/3 .98/3.68 X4 60 100-ifba 4 .28/3 .98/3 .68 Total 108 8656 4.15 averaQe ATTACHMENT 7, Page 24
Enclosu re Description and Assessment of Proposed License Amendment Figure 3-4: Cycle N+2 Full Core NGF Cycle ~ "" BB =* Fuel Q"'" "" (<)CJ "'~"" 1 2R2 2 3R2 3 31 4 63Rl xx yy Batch Identifier
= QC Location (Previous Cycle)
= Number of 90° Rotations W4 ws X4 X4 5 6R2 6 45R2 7 8 9 10 W3 X4 Yl Y2 Y2 Y3 11 49 12 13 14 9Rl 15 16 55 17 19R3 Wl Yl Y2 X2 Y4 X4 Xl 18 26R2 19 20 21 8 22 23 41R3 24 25 29 W3 Yl Y3 X2 Y4 X4 Y4 X4 26 15R2 27 28 44 29 30 13R3 31 32 35 33 12R2 X4 Y2 X2 Y4 X2 Y4 Xl Xl 34 43R2 35 36 53R3 37 38 27Rl 39 62R3 40 22R2 41 42 39Rl W4 Yl X2 Y4 X2 X3 X4 Y4 X3 43 52R2 44 45 46 57Rl 47 48 37R2 49 50 67R2 51 ws Y2 Y4 X4 Y4 X4 Y3 X4 Y4 52 47 53 54 24 55 56 7 57 58 20R2 59 60 59 X4 Y2 X4 Y4 Xl Y4 X4 Y3 X4 61 63 62 63 19R2 64 29R3 65 12Rl 66 39 67 68 59R3 69 39Rl X4 Y3 Xl X4 Xl X3 Y4 X4 HS
*** reinsert from EOC N-11 Feed Batch Characteristics Burn able Sub- Number of Absorber Enrich ments batch Assemblies No. Pins - w/o U-235 Type Y1 16 32-ifba 4.40/4.10/3.80 Y2 24 60-ifba 4.40/4.10/3.80 Y3 16 88-ifba 4. 10/3.80/3 .50 Y4 48 100-ifba 4.10/3 .80/3. 50 Total 104 8160 3.98 averaqe ATTACHMENT 7, Page 25
Enclosure Description and Assessment of Proposed License Amendment Figure 3-5: Cycle N+3 Near Equilibrium Full Core NGF Cycle ~ "' * ""'" core (QQ L<>ratioo BB = Fuel Batch Identifier 1 6R2 2 3R2 3 37R3 4 29Rl xx = QC Location (Previous Cycle) X4 X4 Y4 Y4 yy = Number of 90° Rotations 5 65Rl 6 47Rl 7 8 9 10 Xl Y4 Zl Z2 Z2 Z3 11 61R2 12 13 14 9Rl 15 16 57R2 17 X4 Zl Z2 Y2 Z4 Y4 Z4 18 63 19 20 21 8R2 22 23 45 24 25 49R2 Xl Zl Z4 Y2 Z4 Y4 Z4 Y3 26 31R3 27 28 44R2 29 30 59R2 31 32 35 33 12R2 Y4 Z2 Y2 Z4 Y3 Z4 Yl Yl 34 26R2 35 36 53R3 37 38 20 39 19 40 27R2 41 42 67Rl X4 Zl Y2 Z4 Y3 Yl Y2 Z4 Y4 43 52R2 44 45 46 15 47 48 13R2 49 50 55R3 51 X4 Z2 Z4 Y4 Z4 Y2 Z4 Y4 Z4 52 22Rl 53 54 41R2 55 56 7 57 58 24Rl 59 60 62R3 Y4 Z2 Y4 Z4 Yl Z4 Y4 Z3 Y3 61 29 62 63 64 49Rl 65 12Rl 66 67 67 68 62R2 69 39R3 Y4 Z3 Z4 Y3 Yl Y4 Z4 Y3 HS
*** reinsert from EOC N-11 Feed Batch Characteristics Burnable Sub- Number of Absorber Enrichments batch Assemblies No. Pins - wlo U-235 Type Z1 16 32-ifba 4 .55/4.25/3.95 Z2 24 60-ifba 4 .55/4 .25/3.95 Z3 8 88-ifba 4 .25/3 .95/3 .65 Z4 60 100-ifba 4.25/3.95/3.65 Total 108 8656 4 .12 average ATTACHMENT 7, Page 26
Enclosure Description and Assessment of Proposed License Amendment Figure 3-6: Soluble Boron versus Cycle Length for Representative NGF Cycles
~ ~ 800 +-~~~--+-~~~~t---"~ '-:----t-~~~-+~~~~+-~~~---i ~
0 100 200 300 400 500 600 EA>D Cycle length BOC max rise #fresh efpd ppm ppm ppm shims cyN 494 927 1079 152 8320 cyN+1 514 989 1125 136 8656 cyN+2 489 939 1063 124 8160 cyN +3 514 99 1 1132 141 8656 Note: Cycle N-1 is pre-NGF Standard Fuel Design with Erbia Burnable Absorber fo r comparison . ATTACHMENT 7, Page 27
Enclosure Description and Assessment of Proposed License Amendment Figure 3-7: Maximum Fr Peaking Factor versus Cycle Length for Representative NGF Cycles 0 100 200 300 400 500 600 EfPD Note: Cycle N-1 is pre-NGF Standard Fuel Design with Erbia Burnable Absorber for comparison. ATTACHMENT 7, Page 28
Enclosure Description and Assessment of Proposed License Amendment Figure 3-8: Maximum Fxy Peaking Factor versus Cycle Length for Representative NGF Cycles
~ 1.42 .+---__._.."---->i~---+------4-----+---A.-.:W~'7-l------l 0 100 200 300 400 500 600 EfPD Note: Cycle N-1 is pre-NGF Standard Fuel Design with Erbia Burnable Absorber for comparison .
ATTACHMENT 7, Page 29
Enclosure Description and Assessment of Proposed License Amendment Figure 3-9: Maximum Fq Peaking Factor versus Cycle Length for Representative NGF Cycles 1.72 1.70 cyN-1 1.68 1.66 1.64 C" LL. 1.62 1.60 1.58 1.56 1.54 1.52 0 100 200 300 400 500 600 Eff>D Note: Cycle N-1 is pre-NGF Standard Fuel Design with Erbia Burnable Absorber fo r comparison . ATTACHMENT 7, Page 30
Enclosure Description and Assessment of Proposed License Amendment Figure 3-10: Maximum Fz Peaking Factor versus Cycle Length for Representative NGF Cycles 1.20 ~---~--------~---------~---~ N u.. 1.14 +-il l"-- - - '\--t- ---.li 1.08 ~---~----~---~----~---~~---~ 0 100 200 300 400 500 600 EFPD Note: Cycle N-1 is pre-NGF Standard Fuel Design with Erbia Burnable Absorber for comparison . ATTACHMENT 7, Page 31
Enclosure Description and Assessment of Proposed License Amendment ~ FUELPERFORMANCE Analyses and assessments have been performed to demonstrate acceptable fuel performance (e.g., rod internal pressure, power-to-melt, spent fuel pool rod internal pressures, axial densification factor and engineering factor on linear heat rate) for the NGF design and expected operating conditions for the implementation of the NGF assemblies. These fuel performance results, based on NGF project guidelines, meet criteria using approved methods. A fuel performance analysis was performed for the PVNGS NGF fuel rods to demonstrate acceptable performance relative to fuel criteria over the design lifetime for the NGF rods and to provide information to other functional areas so that the appropriate downstream analyses could be performed by those groups. Similar analyses were also performed for any co-resident STD rods. The fuel performance analysis of the PVNGS NGF and STD fuel rod designs was performed to assess if there is acceptable predicted power fall-off margin relative to what is predicted by Physics for NGF implementation cycles. FATES3B bounding analysis results were obtained using applicable NRC approved methodology such that required Fuel Performance limits are satisfied. The methodology for the fuel performance analysis is obtained from the FATES3B Fuel Evaluation Topical Reports (References 4.1, 4.2, and 4.3), and includes fuel rod maximum pressure technology (Reference 4.4). The FATES3B fuel performance code is used to calculate fuel temperatures and rod internal pressures for the safety analysis and thermal-mechanical calculations for CE-NSSS plants. The FATES3B code was developed to conservatively bound the fuel temperature and fission gas release data measured, and consequently provide a conservative prediction of rod internal pressure. In addition to the conservatisms built into the models, the application methodology of the FATES3B code, especially through the use of bounding power histories, bounding axial power shapes and accounting for peaking factor reduction with burn up compensates for the effects of thermal conductivity degradation (TCD) for the safety and design applications. In addition, the PVNGS Units 1, 2, and 3 core designs shall retain margin to the radial fall-off (RFO) temperatures [ 0
] *c. The fuel centerline temperature allowance is given in Table 4-1.
Table 4-1: Fuel Temperature Allowance a,c ATTACHMENT 7, Page 32
Enclosure Description and Assessment of Proposed License Amendment The methodology for modeling zirconium diboride (ZrB2 ) fuel (i.e., Integral Fuel Burnable Absorber [IFBA] fuel rods) is contained in the ZrB 2 topical report (Reference 4.5). Use of the methodology of Reference 4.5 is consistent with its NRC SER. The methodology for modeling standard ZIRLO and Optimized ZIRLO' cladding is contained in their fuel cladding topical reports (References 4.6 and 4.7, respectively). Use of the methodology of References 4.6 and 4.7 is consistent with their NRC SERs. Reference 4.6 is currently approved for use at PVNGS. Optimized ZIRLO' is described in CENPD-404-P-A, Addendum 1-A "Optimized ZIRLO'" (Reference 4.7). Optimized ZIRLO' fuel cladding is different from standard ZIRLO in two respects: 1) The tin content is lower; and 2) The microstructure is different. This difference in tin content and microstructure can lead to differences in some material properties. Most of the material properties of standard ZIRLO and Optimized ZIRLO' are the same within the uncertainty of the data. The NRC staff approved Optimized ZIRLO' fuel cladding based on:
- 1) Similarities with standard ZIRLO; 2) Demonstrated material performance; and 3) A commitment to provide irradiated data and validate fuel performance models ahead of burnups achieved in batch application. Use of the methodology of Reference 4. 7 is consistent with its SER.
The CE 16x16 Next Generation Fuel Core Reference Report (Reference 4.8) has been reviewed and approved by the NRC. This topical report describes the CE 16x16 Next Generation Fuel assembly design and the methods and models used for evaluating its acceptability. Use of the methodology of Reference 4.8 is consistent with its NRC SER. The PVNGS NGF fuel performance analysis produced bounding radial fall-off curves for both NGF and SID fuel (with and without burnable absorbers). [
]a,c.
The confirmation of the fuel performance results are performed for every reload. These results include the confirmation of fuel rod design criteria and results for use in the mechanical design, the Non-LOCA analysis, the ECCS performance analysis and the digital setpoints functions. 4.1. References 4.1. CEN-161 (B)-P-A, Improvements to Fuel Evaluation Model, August 1989 4.2. CEN-161(8)-P, Supplement 1-P-A, Improvements to Fuel Evaluation Model, January 1992 4.3. CENPD-139-P-A, Fuel Evaluation Model, July 1974 4.4. CEN-372-P-A, Fuel Rod Maximum Allowable Gas Pressure, May 1990 4.5. WCAP-16072-P-A, Revision 0, Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs, August 2004 4.6. CENPD-404-P-A, Implementation of ZIRLO Cladding Material in CE Nuclear Power Fuel Assembly Designs, November 2001 ATTACHMENT 7, Page 33
Enclosure Description and Assessment of Proposed License Amendment 4.7. CENPD-404-P-A Addendum 1-A, Optimized Z/RLO', July 2006 4.8. WCAP-16500-P-A, Revision 0, CE 16x16 Next Generation Fuel Core Reference Reporl,August2007 ATTACHMENT 7, Page 34
Enclosure Description and Assessment of Proposed License Amendment
- 5. CORE THERMAL HYDRAULIC DESIGN 5.1. Introduction This section describes the core thermal-hydraulic (T-H) analyses methodology performed as part of the PVNGS NGF program. The analyses support transition from a full core of STD fuel through a mixed-fuel core to a full core of NGF.
5.2. Description of T-H Methods The current T-H design basis for PVNGS includes the prevention of departure from nucleate boiling (DNB) on the limiting fuel rod with a 95 percent probability at a 95 percent confidence level (95/95) during normal operations and Anticipated Operational Occurrences (AOO) described in the PVNGS UFSAR. The design basis is documented in PVNGS .UFSAR Section 4.4. The thermal-hydraulic design methods remain the same as discussed in the current PVNGS UFSAR, except for the following changes:
- The NRG-approved Westinghouse version of VIPRE-01 subchannel analysis code (Reference 5.1 [referred to as VIPRE-W] is used for Departure from Nucleate Boiling Ratio (DNBR) calculations of CE fuel designs.
- The NRG-approved ABB Non-vane (ABB-NV) critical heat flux (CHF) correlation in References 5.4, 5.5, and 5.18 is used in place of the CE-1 CHF correlation (References 5.12 and 5.13) to predict DNBR in the co-resident STD fuel during transition cores.
- The NRG-approved Westinghouse Side Supported Vane (WSSV, WSSV-T) CHF correlation in Reference 5.7 is applied to mixing vane and IFM grid regions and ABB-NV CHF correlation is applied to non-mixing vane grid regions to predict DNBR in NGF assembly.
- The NRG-approved Statistical Combination of Uncertainties (SCU) methodology in Reference 5.19 was added with no change to the current SCU method described in Reference 5.1 O and supplemented by Reference 5.11 to [
re to calculate an overall uncertainty factor.
- The NRG-approved WLOP CHF correlation in Reference 5.5 can be used to supplement to ABB-NV and WSSV to predict DNBR at low pressure conditions as an alternative to the MacBeth correlation.
* +
5.2.1. Subchannel Analysis Code APS has elected to transition from the currently licensed TORC code (References 5.14 and 5.15) to the VIPRE-W code (Reference 5.1) for most DNB analyses. VIPRE-W is the
- westinghouse version of the VIPRE-01 code (Reference 5.3) developed by Battelle Pacific Northwest Laboratories under the sponsorship of the Electric Power Research Institute (EPRI).
In Reference 5.4, the VIPRE-W code was demonstrated to be equivalent to the TORC code for DNBR calculations, and was approved by the NRC for CE-Nsss* applications. In Reference 5.8, VIPRE-W code was accepted by the US NRC for the NGF design application. DNBR calculations were performed with the VIPRE-W code to develop the SCU DNBR limit and associated probability density function (pdf). Transition core analysis and all other analyses where minimum DNBR and/or DNB overpower for the previous core reloads were predicted ATTACHMENT 7, Page 35
Enclosure Description and Assessment of Proposed License Amendment using the TORC code. CETOP-D code (Reference 5.16) and its applications to digital Setpoints and UFSAR Chapter 15 events remain unchanged, but the CETOP-D correction factors are developed based on results of the VIPRE-W subchannel analysis code. Relative to the existing TORC modeling process, the VIPRE-01 geometric modeling for the PVNGS core includes [ re. This is an acceptable modeling technique approved by the NRG in References 5.1 and 5.4. [ re. The steady state VIPRE-01 one-pass model execution utilizes [
]a,e with appropriate axial noding, fuel rod modeling, power distributions, turbulent mixing, crossflow modeling, axial hydraulic loss, two-phase flow correlations and variation in fuel fabrication and core inlet flow to solve conservation equations of mass, axial and lateral momentum and energy to determine the detailed three-dimensional description of fluid enthalpy, axial flow, .lateral flow and momentum pressure drop.
5.2.2. CHF Correlations and Limits The ABB-NV CHF correlation is applied with the VIPRE-W code to predict DNBR in STD fuel. The ABB-NV and WSSV CHF correlations are applied to NGF with the VIPRE-W code per the NGF topical requirement in Reference 5.8. The ABB-NV CHF correlation was developed for CE 14x14 and 16x16 fuel designs having non-mixing vane grids with the TORC code, based on the CHF data obtained from 5x5 and 6x6 rod assembly tests conducted in the Heat Transfer Research Facility of Columbia University. As indicated in the applicable SER (Reference 5.18), ABB-NV with a 95/95'correlation DNBR safety limit of 1.13 is approved for the CE 14x14 and 16x16 fuel designs as an alternative to the CE-1 CHF correlation. Relative to the CE-1 correlation, ABB-NV includes [
]a,e. Accordingly, the ABB-NV CHF correlation provides a more realistic prediction of DNB margin relative to the CE-1 CHF correlation. The ABB-NV CHF correlation is also qualified to use with the VI PRE-W code in References 5.4, 5.5, and 5.18 for fuel designs or fuel regions containing non-mixing vane grids with the same 95/95 correlation DNBR safety limit of 1.13.
The thermal performance of the NGF design is enhanced by side-supported (SS) mixing vane and IFM grids. CHF tests were performed with the NGF mixing vane grids with different grid spacing at the Columbia University Heat Transfer Research Facility. To reflect the thermal performance of the fuel design, new CHF correlations, WSSVand WSSV-T, were developed with the test data for the NGF design with the VIPRE-W code and the TORC code, respectively, in Reference 5.7. Therefore, the VIPRE-W and TORC codes can be used for thermal-hydraulic analysis of the core loaded with NGF design using the WSSV CHF correlation. Both WSSV and WSSV-T have the same 95/95 correlation limit of 1.12, applicable to the NGF fuel region above the first mixing vane grid. The ABB-NV CHF correlation with a 95/95 limit of 1.13 (References 5.4, 5.5, and 5.18) is applied to the fuel region below the first mixing vane mid grid of the NGF design. For the current PVNGS DNB analysis, the MacBeth CHF correlation (Reference 5.22) is used to supplement the primary CHF correlation where the primary CHF correlation is not applicable as ATTACHMENT 7, Page 36
Enclosure Description and Assessment of Proposed License Amendment discussed in Subsection 15.1.5.2 of the PVNGS UFSAR. Although the current applications of* MacBeth CHF correlation are maintained for the PVNGS program, the WLOP CHF correlation, as approved by the NRC with the VIPRE-W code in Reference 5.5 for the CE fuel designs, can be used as an alternative to the MacBeth correlation for more accurate DNBR predictions. The WLOP CHF correlation has been derived based on low pressure CHF data from the CE 14x14 and 16x16 rod assembly tests similar to those in the ABB-NV database and its wider range of applicability provides more flexibility to DNBR evaluation at the hypothetical hot zero power steam line break (HZPSLB) conditions. The WLOP correlation 95/95 DNBR safety limit with the VIPRE-W code is 1.18. 5.2.~. DNB Methodology The DNB analyses continue to be based on the Modified Statistical Combination of Uncertainties (MSCU) process as approved by the NRC in CEN-356(V)-P-A, Revision 01-P-A (Reference 5.20) and WCAP-16500-P-A Supplement 1, Revision 1 (Reference 5.24) for CE-NSSS's with digital setpoint systems. With the MSCU methodology, uncertainties are treated in two groups. One group combines system parameter uncertainties with CHF correlation uncertainty and subchannel code uncertainty statistically to generate a 95/95 DNBR safety limit and associated probability density function (pdf). The other group uses this pdf and statistically combines it with state parameter uncertainties and uncertainties related to Core Operating Limits Supervisory System (COLSS) and Core Protection Calculator (CPC) algorithm, simulator model, computer processing and startup measurements to determine the COLSS and CPC _overall uncertainty factors. The section herein discusses the system parameter SCU treatment method of the first group. The system -parameters are characterized by the physical system through which the coolant passes. *The parameter and code uncertainties included in the overall system parameter DNB uncertainty factor are: *
- Core inlet flow distribution
- Engineering factor on enthalpy rise
- Systematic fuel rod pitch
- Systematic fuel rod OD
- Engineering factor on heat flux
- Subchannel code uncertainty The PVNGS UFSAR Section 4.4 references Enclosure 1-P to LD-82-054 (Reference 5.10) and Supplement 1-P to Enclosure 1-P to LD-82-054 (Reference 5.11) as the licensing documents for PVNGS-specific SCU system parameter uncertainty treatment method. The PVNGS specific system parameter SCU method involves using the [. ]0 *c, presented in Section 5.3 of CEN-139(A)-P (Reference 5.19) to combine inlet flow factor uncertainty to calculate an overall uncertainty factor. Other system parameter uncertainties are combined using [
0
] *c, as discussed in Reference 5.11. For the applications, the system parameter SCU methodology approved by the NRC (Reference 5.19) was used to combine all system parameter uncertainties, including the uncertainty in the core inlet flow distribution, with the CHF correlation uncertainties [ ]0 *c. The 95/95 probability/confidence DNBR safety limit was determined from the resultant pdf and was further adjusted to account for the rod bow penalty to arrive at the final DNBR safety limit.
With the introduction of the VIPRE-W code, the system parameter SCU process was improved by achieving [ ]0 *c with VIPRE-W runs, instead of through a [ ATTACHMENT 7, Page 37
Enclosure Description and Assessment of Proposed License Amendment
]a,c. The main purpose of the SCU [ ]8*c was to facilitate a large number of [ ]a,c without making excessive detailed subchannel code runs which was impractical with the available technology in the past. With the significant improvements in the computer technology in recent years, it is feasible to perform a large number of [ ]a,c without the need for the [ ]8 'c. The improved SCU process allows running [ ]a,c for robust sampling and calculations with the VI PRE-W code [ ]a,c when linked with an uncertainty analysis code, instead of the uncertainty convolution through the [ ]a,c_ The uncertainty analysis code utilizes [ ]a,c techniques approved by the NRC in CEN-356(V)-P-A, Revision 01-P-A (Reference 5.20). Replacement of the [
]a,c construction with the [ ]a,c is consistent with the existing SCU methodology as approved by the NRC (Reference 5.19).
To perform [ ]a,c with the improved SCU process, data population based on the input of the system parameter uncertainties were generated by [
]8-c. All of the data population for the DNBR distribution was generated at [
]a,c of state parameter conditions. The tmproved SCU process still maintains the same level of conservatism pertaining to the [ ]a,c that is searched at the [
]a,c. Consistent with the existing SCU process, the DNBR distribution was further adjusted to deterministically account for rod bow penalty and to preserve artificial margin to bound future reload designs.
The fuel-related system parameter uncertainties and uncertainties in code and CHF correlation for the current CE-1 based SCU analysis are listed in Table 3-1 of Reference 5.20. The uncertainties in inlet flow distribution are listed in Table 3-2 of Reference 5.11. For SCU DNBR safety limit and pdf development for STD fuel based on ABB-NV CHF correlation, the current parameter uncertainties remain applicable with the exception of the CHF correlation uncertainty. The ABB-NV CHF correlation uncertainty was applied in place of the CE-1 CHF correlation uncertainty. The subchannel code uncertainty in VIPRE-W is the same as that in the TORC code. The parameter uncertainties used to develop the SCU DNBR safety limit for the STD design are presented in Table 5~1. Since no changes have been introduced to the lower core support structure, flow skirt, and In-Core Instrument (ICI) arrangement as part of NGF implementation, the inlet flow factors and uncertainties remain unchanged as reported *in Reference 5.11. The variations and tolerance deviations pertaining to NGF design [ .
]a,c were evaluated to confirm that the existing values for heat flux and enthalpy rise engineering factor listed in Table 3-1 of Reference 5.20 continued to be bounding for the NGF design. The uncertainty in systematic rod pitch for the NGF design was derived using the
[ ]a,c for a NGF LFA. Modifications to the outer strap (MOS) and addition of the anti-hangup tabs (OST) at the face central rods have little to no influence on the DNB-limiting interior cell dimensions. A [ ]a,c on the 0.374 inch rod OD was used to derive the uncertainty on systematic rod OD for the NGF design. The parameter uncertainties used to develop SCU DNBR safety limit for NGF design are presented in Table 5-2. Since the uncertainties are considered in determining the SCU DNBR safety limit, the DNBR calculations ATTACHMENT 7, Page 38
Enclosure Description and Assessment of Proposed License Amendment in the plant safety analyses are performed using the nominal values (without the uncertainties) of the system parameters. The current PVNGS SCU DNBR safety limit and pdf based on CE-1 CHF correlation include allowance for NRG-imposed HID-1 (STD) grid DNBR penalty and for rod bow DNBR penalty. Both ABB-NV and WSSV CHF correlations include [
]a,c. Therefore, no HID-1 (STD) DNBR penalty is required to apply on the STD or the NGF fuel due to different grid spacing. With the application of the ABB-NV and WSSV CHF correlations, the allowance for the rod bow DNBR penalty is the only allowance incorporated into the SCU DNBR safety limit and pdf deterministically.
The ABB-NV CHF correlation was applied to the STD design. Based on the applicable system parameter uncertainties, ABB-NV CHF correlation uncertainty, and the VIPRE-W based DNBR sensitivity, an SCU DNBR safety limit value of 1.34 was established for the DNBR analyses using VIPRE-W and the ABB-NV CHF correlation for the STD fuel. The NRG-approved 95/95 CHF correlation DNBR safety limit for the ABB-NV CHF correlation was preserved in the statistical treatment per NRC IN-2014-1 (Reference 5.23). The ABB-NV and WSSV CHF correlations were applied to the NGF design. The ABB-NV CHF correlation was applied to non-mixing vane grid region and the WSSV CHF correlation was applied to the mixing vane grid region of the NGF design. [ Based on the applicable system parameter uncertainties, ABB-NV/WSSV CHF correlation uncertainties and the VIPRE-W-based DNBR sensitivity, an SCU DNBR safety limit value of 1.25 was established for the DNBR analyses using VIPRE-W and the ABB-NV/WSSV CHF correlations. The NRG-approved 95/95 CHF correlation DNBR safety limits for the applied CHF correlations were preserved in the statistical treatment per NRC IN-2014-1 (Reference 5.23). For the PVNGS NGF DNB analysis events where the fluid conditions are outside the ABB-NV and WSSV CHF correlation range of applicability, deterministic method of uncertainty treatment continues to be applicable. The MacBeth CHF correlation with 1.30 DNBR safety limit remains conservative for low pressure events. The NRG-approved WLOP CHF correlation with 1.18 DNBR safety limit is also applicable for low pressure events for use with the VIPRE-W subchannel analysis code, as an alternative to the MacBeth correlation. ATTACHMENT 7, Page 39
Enclosure Description and Assessment of Proposed License Amendment Table 5-1: Components Combined in the DNBR pdf for PVNGS STD Fuel Std. Deviation at Parameters Mean 95% Confidence Inlet flow distribution * *
]a,b,e [ ]a,b,e Enthalpy rise factor [
]a,b,e ]a,b,e Systematic pitch, in. [ [
]a,b,e [ ]a,b,e Systematic clad OD, in. [
Heat flux factor ]a,b,e [ ]a,b,e [
]a,b,e [ rb,e ABB-NV CHF correlation (M/P) [
]a,b,e ]a,b,e VIPRE-W code uncertainty [ [
DNBR pdf [ re [ ]a,e
- Inlet flow distribution uncertainties reported in Reference 5.10 are applied.
** The mean value is conservatively adjusted to preserve the NRG-approved 95/95 DNBR safety limit. *** [ re rod bow penalty multiplied to both the mean and standard deviation. The
[ ]a;e penalty is based on a burn up of [ re. Table 5-2: Components Combined in the DNBR pdf for PVNGS NGF Fuel Std. Deviation at Parameters Mean 95% Confidence Inlet flow distribution * * [ ]a,b,e [ ]a,b,e Enthalpy rise factor Systematic pitch, in. [ ]a,b,e [ ]a,b,e Systematic clad OD, in. [ ]a,b,e [ ]a,b,e
]a,b,e ]a,b,e Heat flux factor [ [
WSSV CHF correlation (M/P) [ ]a,b,e [ ]a,b,e VIPRE-W code uncertainty [ rb,e [ ]a,b,e DNBR pdf [ ]a,e [ re
- Inlet flow distribution uncertainties reported in Reference 5.10 are applied.
** The mean value is conservatively adjusted to preserve the NRG-approved 95/95 DNBR safety limit. *** [ ]a,e rod bow penalty multiplied to both the mean and standard deviation. The
[ ]a,e penalty is based on a burnup of [ re. 5.3. NGF Assembly The NGF design is a significant fuel design upgrade relative to the existing PVNGS STD fuel from a thermal-hydraulic viewpoint due to introduction of flow mixing vanes and IFM grids (References 5.8 and 5.9). It also has a slightly smaller rod outside diameter (0.374 inch OD) as compared to STD fuel (0.382 inch OD), yielding an increased assembly flow area. The PVNGS-specific NGF fuel design is equipped with a Guardian' grid at the bottom, three non-mixing vane mid-grids, six mixing vane mid grids and two IFM grids in the active fuel region, ATTACHMENT 7, Page 40
Enclosure Description and Assessment of Proposed License Amendment and an lnconel grid at the top. The structural mid grids are essentially at the same axial elevations; however, the NGF design incorporates two IFM grids positioned in between the two mixing vane grid spans, relative to the STD fuel assembly. The mixing vane grids of the NGF design contribute to a greater pressure drop relative to STD. However, the mixing vane and IFM grids promote thermal margin gains over the current resident STD fuel due to improved turbulent mixing. The addition of two IFM grids improves fuel rod crud and corrosion resistance for selected grid span locations. The smaller diameter of the advanced 0.374 inch OD NGF rod accommodates the higher pressure drop of the mid and IFM grids. As previously noted, a full scale hydraulic test was conducted of the original NGF design in the Westinghouse FACTS loop to determine the design loss coefficients for the assembly. Loss coefficients are used to define the hydraulic characteristics for the assembly design. The FACTS test loop is capable of testing a full size fuel assembly over a wide range of conditions by varying temperatures and flow rate in order to determine the hydraulic characteristics as a function of Reynolds Number. A direct extrapolation of the loss coefficients from the test data is used to determine the design loss coefficients at actual reactor operating conditions. Due to a small change in grid height for the IFM grid following the testing, the measured IFM grid loss coefficient [
]a,c. Additionally, the [
5.4. Hydraulic Compatibility 5.4.1. Transition Core GTRF Considerations 5A1.1. Hydraulic Testing Performed Due to the higher hydraulic resistances of the mixing vane grids and IFMs, the NGF design will likely cause the neighboring STD fuel assemblies to experience large crossflows which may reduce margin to the STD fuel design criterion with respect to GTRF wear. The fuel assembly hydraulic compatibility is evaluated using vibration [
]a.c. Full scale hydraulic tests were performed on the original NGF assembly in the Westinghouse VIPER test loop to investigate the vibration and GTRF wear characteristics of the NGF design and co-resident STD design containing a top lnconel grid. A 500 hour accelerated wear test was performed to investigate (1) the GTRF wear characteristics of the NGF assembly, (2) the mixed
- core impact on fretting of the NGF design and (3) the mixed core impact on fretting of the STD design.
Additionally, single assembly [ ]a,c tests for the original NGF and STD prototypical assemblies were conducted in the FACTS test loop. Both assemblies were\ tested over a wide range of flow rates to determine [
]a.c. The [ ]a.c was repeated over a wide range of flows prior to the dual assembly VIPER endurance test to confirm that [
]a,c.
The wear and vibration results from testing indicate that the cross-flow between original NGF and STD fuel assemblies satisfy the design criteria. Since the minor grid design changes due to modified outer strap (MOS) and addition of anti-hangup tab (OST) impose only minimal changes to loss coefficients, the test results for the original NGF design are applicable to the MOS/OST design. ATTACHMENT 7, Page 41
Enclosure Description and Assessment of Proposed License Amendment 5.4.1.2. Grid-to-Rod Fretting Wear Calculations The Grid-to-Rod Fretting (GTRF) wear analysis approach for [
]a,c wear depths from post-irradiation examination (PIE) data collected from PVNGS Unit 3 EOC.15 STD fuel.
Results of the GTRF wear calculation show that the maximum wear depth anticipated for the STD design during transition cycles meets Westinghouse's field wear criteria for transition cycles. 5.5. Transition Core DNB Effect Transition core effects on both the NGF and the STD fuels are evaluated with respect to DNB. For each transition cycle, the entire transition core is reviewed and the limiting assembly (STD or NGF) is then analyzed with the respective correlation. Although NGF assemblies will likely experience some flow starvation due to flow diversion in the mixing vane grid region in transition cores, thermal margin gains from WSSV CHF correlation will offset any flow penalty in that region. The VIPRE-W code is used in the transition core analysis for benchmarking the CETOP-D model such that the CETOP-D results are conservative relative to VIPRE-W results. The benchmark process consists of selection of the limiting fuel assembly and determination of the CETOP-D correction factor. The process for selecting the limiting fuel assembly is similar to the existing selection process
- implemented for PVNGS reloads bi;ised on a "Complete Screening" approach. The limiting assembly candidates for CETOPNIPRE-W benchmarking are determined by [
]a,c. Several candidates are typically identified that meet the specified limiting assembly selection criteria by performing [
]a,c runs to compute DNBRs for all fuel assemblies having pin peak [ re of the core maximum peak.
In keeping with the existing TORC-based screening process for selecting the limiting fuel assembly, a [ ]a,c VIPRE-W execution process is developed to screen for limiting ASls and sets of coolant conditions. [ ATTACHMENT 7, Page 42
Enclosure Description and Assessment of Proposed License Amendment *
]a,e.
Based on the results of the [ ]a,e screening, limiting assemblies are used to determine CETOP-D correction factors over a wide range of operating conditions for use in non-LOCA accident and setpoint analyses. The results from the [ ]a,e modeling have been verified and validated with result!? from the [ ]a,e VIPRE-W model. Additionally, the VIPRE-W/CETOF-D benchmark process yields more conservative results (i.e., the larger CETOP-D overpower penalty) than the TORC/CETOP-D benchmark process. Additional detail on the transition cycle benchmark analysis is provided below. 5.5.1. First NGF Cycle The online reactor monitoring and protection systems use a CETOP-D model input for thermal margin calculations that [ ]a,e. Consequently, it is necessary to apply a CETOP-D model and [ re during transition cycles. The STD design is likely to be the limiting fuel during first transition cycle of NGF fuel due to comparable power with fresh NGF fuel, and STD fuel having no mixing vane grids. Therefore, the more limiting DNBR limit of 1.34 for the STD fuel assemblies will be applied to both STD and NGF assemblies during the first transition core. 5.5.2. Second NGF Cycle The CETOPNIPRE-W benchmarking analysis was performed to generate CETOP-D correction factors for a second transition cycle. The results of this analysis could allow taking partial credit of NGF since NGF assemblies will be the most limiting assemblies in the second transition core as the STD fuel are second-burned positioned on the core peripheral locations. The ABB-NV and WSSV CHF correlation will be applied to NGF assemblies in VIPRE-W calculations .. Since STD fuel is no longer a limiting assembly candidate, the CETOP-D model based on the NGF design and WSSV/ABB-NV CHF correlations and SCU DNBR safety limit of 1.25 will be applied to the second transition core. 5.6. . Effects of Fuel Rod Bow on DNBR Rod bow can occur in the spans between the grids, reducing the spacing between adjacent fuel rods and reducing the margin to DNB. Rod bow must be accounted for in the DNB safety analysis of Condition I and Condition II events. The rod bow penalty is a function of the grid span length, the fuel cladding Young's Modulus, the fuel rod moment of inertia and the fuel rod's burnup. The CHF correlation statistics was convoluted with the pdf for rod bow closure data and with the rod bow effect model to determine the rod bow penalty as a function of fuel burnup consistent with the methodology described in Reference 5.17. The [ re correction factor is used for extrapolating the CE-NSSS 14x14 fuel rod bow model to the CE-NSSS 16x16 fuel design. [ re. The CE-NSSS 14x14 fuel model has been approved by NRC to use for CE-NSSS 16x16 fuel. References 5.6 and 5.11 confirmed the rod bow topical report (Reference 5.17) to be applicable for STD and NGF designs with ABB-NV and WSSV CHF correlation applications. The rod bow penalty at [ re was deterministically applied to ABB-NV and WSSV-based derived DNBR safety limit and pdf. No additional rod bow penalty is required for burnups greater than [ ]a,e, since credit is taken for the effect of peaking factor burndown due to the decrease in fissionable isotopes and the buildup of fission products. ATTACHMENT 7, Page 43
Enclosure Description and Assessment of Proposed License Amendment 5.7. Rods in DNB The seized rotor (SR), sheared shaft (SS) and loss of flow (LOF) at DNB Specified Acceptable Fuel Design Limit (SAFDL) events were analyzed to determine the probability of rods in DNB consistent with methodology approved by the NRC (Reference 5.21). The seized rotor and sheared shaft accidents are classified as Condition IV events and LOF at DNB SAFDL is classified as a Condition Ill event. DNBR calculations are performed to quantify the inventory of rods that would undergo DNB and be conservatively presumed to fail. The [ ]a,c approach was applied to execute a steady state core thermal hydraulic VIPRE-W model where the boundary conditions provided by the system transient analyses are [ 5.8. DNBR Calculations at Low Pressure Conditions A DNB evaluation of NGF was modeled in the HRISE code using the MacBeth CHF correlation to analyze post-trip steam line break (SLB) event. This analysis demonstrated that the current DNB acceptance criterion for the SLB event continued to be met in the minimum DNBR predicted by MacBeth CHF correlation during post-trip SLB analysis with the implementation of the NGF design. Additional DNBR calculations for the NGF post-trip SLB event were performed for the limiting case identified by the MacBeth analysis using the VIPRE-W code and the WLOP CHF correlation. The WLOP correlation was used for this application because the system pressure was less than the low-pressure limit of applicability for the primary CHF correlations. Because the SCU DNBR safety limit is not valid at such low pressure conditions, the fuel parameter uncertainties were applied deterministically. The limiting case specific axial and radial power distributions were applied. For this application, the applicable DNBR safety limit is the 95/95 correlation limit of 1.18 for the WLOP CHF correlation. The results of the VIPRE-W calculations for the NGF design confirmed that the WLOP correlation DNBR safety limit of 1.18 was met at the hot channel condition similar to that used in the MacBeth analysis. 5.9. Conclusion The thermal-hydraulic evaluation of the fuel upgrade for PVNGS has shown that STD and NGF assemblies are hydraulically compatible and that the DNB margin gained through use of the new fuel design (NGF) with the WSSV CHF correlation is sufficient to offset flow-related penalties in the mixing vane grid regions. All current thermal-hydraulic design criteria are satisfied. 5.10. References 5.1. WCAP-14565-P-A, VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis, October 1999 5.2. Letter from T. H. Essig (NRC) to H. Sepp (W), Acceptance for Referencing of Licensing Topical Report WCAP-14565, VIPRE-01 Modeling and Qualification for Pressurized ATTACHMENT 7, Page 44
Enclosure Description and Assessment of Proposed License Amendment Water Reactor Non-LOCA Thermal/Hydraulic Safety Analysis (TAC NO. M98666), January 1999 5.3. Letter from C. E. Rossi (NRG) to J. A. Blaisdell (UGRA Executive Committee), Acceptance for Referencing of Licensing Topical Report, EPRl-NP-2511-CCM,
'VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores,' Volumes 1, 2, 3 and 4, May 1986 5.4. WCAP-14565-P-A, Addendum 1-A, Addendum 1 to WCAP-14565-P-A Qualification of ABB Critical Heat Flux Correlations with VIPRE-01 Code, August 2004 5.5. WCAP-14565-P-A, Addendum 2-P-A, Addendum 2 to WCAP-14565-P-A, Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low Pressure Applications, April 2008 5.6. Letter from Ho K. Nieh (NRC) to J. A. Gresham (Westinghouse), Final Safety Evaluation for Westinghouse Electric Company (Westinghouse) Topical Report (TR) WCAP-14565-P, Addendum 2, Revision 0, 'Addendum 2 to WCAP-14565-P-A, Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP
[Westinghouse Low Pressure] for PWR [Pressurized Water Reactor] Low Pressure Applications' (TAC NO. MD3184), February 2008 5.7. WCAP-16523-P-A, Westinghouse Correlations WSSVand WSSV-TforPredicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes, August 2007 5.8. WCAP-16500-P-A, Revision 0, CE 16x16 Next Generation Fuel Core Reference Report,August2007 5.9. WCAP*16500-P-A, Supplement 2, Evolutionary Design Changes to CE 16x16 Next Generation Fuel and Method fat Addressing the Effects of End-of-Life Properties on Seismic and Loss of Coolant Accident Analyses, June 2016
- 5.10. LD-82-054, Enclosure 1-P, Statistical Combination of Uncertainties, Combination of System Parameter Uncertainties in Thermal Margin Analyses for SYSTEM 80, May 1982 5.11. LD-82-054, Supplement 1-P to Enclosure 1-P, System 80 Inlet Flow Distribution Supplement 1-P to Enclosure 1-P to LD-82-054, February 1993 5.12. CENPD-162-P-A, Critical Heat Flux Correlation for CE Fuel Assemblies with Standard Spacer Grids, Part 1, Uniform Axial Power Distribution, September 1976 5.13. CENPD-207-P-A, Critical Heat Flux Correlation for CE Fuel Assemblies with Standard Spacer Grids, Part 2, Non-Uniform Axial Power Distribution, December 1984 5.14. CENPD-161-P-A, TORC Code, A Computer Code for Determining the Thermal Margin of Reactor Core, April 1986 5.15. CENPD-206-P-A, TORC Code, Verification and Simplified Modeling Methods, June 1981 5.16. CEN-160(S)-P, Revision 1-P, CETOP-D Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Units 2 and 3, September 1981 5.17. CENPD-225-P-A, Fuel and Poison Rod Bowing, June 1983 5.18. CENPD-387-P-A, ABB Critical Heat Flux Correlations for PWR Fuel, May 2000 ATTACHMENT 7, Page 45
Enclosure Description and Assessment of Proposed License Amendment 5.19. CEN-139(A)-P, Statistical Combination of Uncertainties; Combination of System Parameter Uncertainties in Thermal Margin Analyses for Arkansas Nuclear One - Unit 2, November 1980 5.20. CEN-356(V)-P-A, Revision 01-P-A, Modified Statistical Combination of Uncertainties, May 1988 5.21. CENPD-183-A, Loss of Flow C-E Methods for Loss of Flow Analysis, June 1984 . 5.22. LD-W0-3900, MacBeth CHF Correlation Approval, August 1983 5.23. NRC IN-2014-1, Fuel Safety Limit Calculation Inputs were Inconsistent with NRC-Approved Correlation Limit Values, February 21, 2014 5.24. WCAP-16500-P-A, Supplement 1, Revision 1, Application of CE Setpoint Methodology for CE 16x16 Next Generation Fuel (NGF), December 2010
\
ATTACHMENT 7, Page 46
Enclosure Description and Assessment of Proposed License Amendment
- 6. FUEL ROD CORROSION ANALYSIS The licensing requirements for fuel rod corrosion analysis of the NGF design are documented in References 6.1 and 6.2. For the Optimized ZIRLO' cladding used in the NGF design (and the ZIRLO cladding used in the STD design), the [
]a,c is limited to a licensed peak value of 100 microns. The clad hydrogen pickup is also limited to a best estimate volumetric averaged value of [ ]a,c at end-of-life. The maximum Thermal Reaction Accumulated Duties (TRDs) in the cladding oxidation models are also limited to values corresponding to a cladding corrosion amount of 100 microns.
Calculations were performed based on projected physics and thermal-hydraulics conditions for a number of PVNGS Unit 2 future cycles. The analysis was performed for the fuel cladding oxide thickness and the fuel cladding hydrogen pickup using the approved TRD-based moqels (Reference 6.2). The analysis of PVNGS NGF implementation shows that the requirements on maximum predicted oxide thickness and maximum predicted volume average hydrogen pickup and maximum predicted TRD are satisfied, with significant margin to the limits. The analyses performed show that the NGF and STD fuel were able to meet the licensing requirement limits for mixed cores at pre-full core NGF conditions. Additionally, the calculations showed that full - cores composed of NGF were able to meet these limits using projected physics and thermal-hydraulic conditions. Based on these results, it can be concluded that the NGF design is expected to meet the corrosion limits in the PVNGS cores. 6.1. References 6.1. WCAP-16500-P-A, Revision 0, CE 16x16 Next Generation Fuel Core Reference Reporl,August2007 6.2. CENPD-404-P-A, Addendum 2-A, Westinghouse Clad Corrosion Model for ZIRLO and Optimized ZIRLO', October 2013 - ATTACHMENT 7, Page 47
Enclosure Description and Assessment of Proposed License Amendment
- 7. NON-LOCA SAFETY ANALYSIS To quantify the effect of introducing the NGF design into the PVNGS safety analysis licensing basis, all Updated Final Safety Analysis Report (UFSAR) Section 15 Non-LOCA transient analyses were evaluated to ensure applicability. All analyses were performed consistent with current Non-LOCA methodology.
All UFSAR Section 15 Non-LOCA transient analyses were performed for PVNGS Units 1, 2 and 3 based on the introduction of the NGF design at a rated thermal power of 3990 MWt (4070 MWt with uncertainty). The Non-LOCA safety analyses were evaluated utilizing an Analytical DNBR limit of 1.38. The Analytical DNBR limit basis is the DNBR safety limit values described in Section 5.2.3 of this Technical Analysis with additional discretionary margin applied. Table 7-1 lists the Non-LOCA safety analyses (transient events) by category of Anticipated Incident of Moderate Frequency (MF), Infrequent Event (IE) or Limiting Fault (LF), and defines the level of evaluation performed. All UFSAR events received the necessary level of evaluation to ensure applicability. The levels of evaluation are:
- The event has been evaluated for PVNGS for the purpose of introduction of NGF and reanalysis was required.
- The event remains "bounded" by the existing PVNGS UFSAR analysis.
- The event was "not impacted" by the introduction of the NGF design.
- The event is not applicable to a CE System 80 plant design (i.e., PVNGS), or the event is only applicable to boiling water reactors and therefore not applicable to the PVNGS pressurized water reactor units.
Radiological accident evaluations are addressed in Section 11 of this Technical Analysis. Table 7-1: Impact of the Use of NGF on UFSAR Chapter 15 Non-LOCA Transient Events "C .....u Q) i<
.....caQ)
"C Q)
"C ca 0.
- c
..... ca ~
UFSAR c: 0 u 0 Section UFSAR Section 15 Event ::i E z ::0. Cl Section iii ::i 0 Q)
> ra 0 0. ca w z ct (.)
7.1 15.1 Increase in Heat Removal by the Secondary System Decrease in Feedwater (FW) 7.1.1 15.1.1 Temperature x MF 7.1.2 15.1.2 Increase in Main FW Flow x MF 7.1.3 15.1.3 Increase in Main Steam Flow (IMSF) x MF Inadvertent Opening of a Steam 7.1.3 15.1.4 Generator Atmospheric Dump Valve x IE (IOSGADV) Steam System Piping Failures Inside and Outside Containment - Operating Modes 1and2 7.1.4 15.1.5 HFP Post-trip MSLB, with and without x LF LOP HZP Post-trip MSLB, with and without LOP ATTACHMENT 7, Page 48
Enclosure Description and Assessment of Proposed License Amendment Table 7-1: Impact of the Use of NGF on UFSAR Chapter 15 Non-LOCA Transient Events Q)
"C "C ti "~
UFSAR
....ca Q) Q)
"C ca c.
1i
..., ca 0 Section UFSAR Section 15 Event ::s c: .§ z0 CJ
- .=
C) Section iii
- s 0 0 c.
c.
....ca Q) w Ill z <( (.)
Steam System Piping Failures Inside and Outside Containment - Operating Modes 7.1.5 15.1.5 1and2 x LF HFP Pre-trip MSLB Steam System Piping Failures Inside and 7.1.6 15.1.6 Outside Containment - Operatino Mode 3 x LF 7.2 15.2 Decrease in Heat Removal by the Secondary S i.'Stem 7.2.1 15.2.1 Loss of External Load x MF 7.2.2 15.2.2 Turbine Trip x MF 7.2.3 15.2.3 Loss of Condenser Vacuum x MF 7.2.4 15.2.4 Main Steam Isolation Valve Closure x MF 7.2.5 15.2.5 Steam Pressure Requlator Failure x N/A Loss of Non-emergency AC Power to the 7.2.6 15.2.6 Station Auxiliaries x MF 7.2.7 15.2.7 Loss of Normal FW Flow x MF 7.2.8 15.2.8 FW System Pipe Breaks x LF 7.3 15.3 Decrease in Reactor Coolant Flow Rate 7.3.1 15.3.1 Total Loss of Reactor Coolant Flow x MF Flow Controller Malfunction Causing a 7.3.2 15.3.2 Flow Coastdown x N/A Single Reactor Coolant Pump (RCP) 7.3.3 15.3.3 Rotor Seizure with Loss of Offsite Power x LF (LOP) 7.3.4 15.3.4 RCP Shaft Break with LOP x LF 7.4 15.4 Reactivity and Power Distribution Anomalies Uncontrolled Control Element Assembly MF 7.4.1 15.4.1 Withdrawal (CEAW) from a Subcritical or x Low (Hot Zero) Power Condition 7.4.2 15.4.2 Uncontrolled CEA Withdrawal at Power x MF 7.4.3 15.4.3 Sinqle Full-Length CEA Drop x MF 7.4.4 15.4.4 Startup of an Inactive RCP x MF Flow Controller Malfunction Causing an N/A 7.4.5 15.4.5 Increase in BWR Core Flow x 7.4.6 15.4.6 Inadvertent Deboration x MF Inadvertent Loading of a Fuel Assembly MF 7.4.7 15.4.7 into the Improper Position x 7.4.8 15.4.8 CEA Ejection x LF 7.5 15.5 Increase in RCS Inventory 7.5.1 15.5.1 Inadvertent Operation of the ECCS x MF Chemical and Volume Control System 7.5.2 15.5.2 Malfunction - Pressurizer Level Control x IE Svstem Malfunction with LOP 7.6 15.6 Decrease in RCS Inventory Inadvertent Opening of a Pressurizer 7.6.1 15.6.1 Safety/Relief Valve x N/A Double-Ended Break of a Letdown Line 7.6.2 15.6.2 Outside Containment x LF ATTACHMENT 7, Page 49
Enclosure Description and Assessment of Proposed License Amendment Table 7-1: Impact of the Use of NGF on UFSAR Chapter 15 Non-LOCA Transient Events Q)
"C Q) "C Q)
+-'
(.) ca ::c "~ UFSAR
+-'
ca "C c.. ., ca 0 c: 0 (.) Section UFSAR Section 15 Event ::I E z :.= C) Section ca> ::I 0 0 c.. c.. Q)
+-'
ca w m z <( (.J 7.6.3 15.6.3 Steam Generator Tube Rupture x LF Radiological Consequences of MSL 7.6.4 15.6.4 Failure Outside Containment (BWR) x N/A Loss-of-Coolant Accidents (see Section 8 7.6.5 15.6.5 N/A of this Technical Analysis) Radioactive Material Release from a 7.7 15.7 Subsystem or Component (see N/A Section 11 of this Technical Analysis) Analysis Methods for Loss of Primary 7.3.1 15.D Coolant Flow x LF Limiting Infrequent Events 7.8 15.E (Loss of Flow from the SAFDL) x LF
- MF - Incident of Moderate Frequency IE - Infrequent Event LF - Limiting Fault N/A - Not Applicable 7.1. UFSAR Section 15.1 - Increase in Heat Removal by toe Secondary System 7.1.1. UFSAR Section 15.1.1 - Decrease in Feedwater Temperature Event is Bounded.
The Decrease in Feedwater Temperature (DFWf) event is no more adverse than the Increase in Main Steam Flow event (UFSAR Section 15.1.3) or the IOSGADV event (UFSAR Section 15.1.4). Therefore, the DFWf event (with and without a single failure) remains bounded by these UFSAR events for both transition and full core implementation of the NGF design. 7.1.2. UFSAR Section 15.1.2 - Increase in Main Feedwater Flow Event is Bounded. The Increase in Main Feedwater Flow (IFWF) event is no more adverse than the Increase in Main Steam Flow event (UFSAR Section 15.1.3) or the Inadvertent Opening of an SG ADV event (UFSAR Section 15.1.4). Therefore, the IFWF event (with and without a single failure) remains bounded by these UFSAR events for both transition and full core implementation of the NGF design. 7.1.3. UFSAR Section 15.1.3 - Increase in Main Steam Flow & UFSAR Section 15.1.4 Inadvertent Opening of a Steam Generator Atmospheric Dump Valve and IOSGADV with Loss of AC Power Events are Evaluated and Reanalyzed. UFSAR Sections 15.1.3 and 15.1.4 describe the increase in main steam flow and the inadvertent opening of a steam generator atmospheric dump valve events. The analysis has been updated to determine the impact of the implementation of a transition and a full core of NGF in combination with the first time implementation of the NGF CHF correlations (Reference 7.1). ATTACHMENT 7, Page 50
Enclosure Description and Assessment of Proposed License Amendment The fuel rod and hot channel changes associated with the NGF design together with the use of the CHF correlations result in a revised calculated DNBR value. The methodology is the same as that used in the current analysis presented in UFSAR Sections 15.1.3 and 15.1.4. No CENTS computer code (Reference 7.2) cases were reanalyzed, since the AOR flow coastdown is conservative and the remaining system model input changes due to NGF were small and the impact on the overall transient system response is insignificant. The overall transient system response remains the same as the current UFSAR analyses and only the DNB margin analysis is affected by the implementation of NGF. With the implementation of NGF, the minimum CETOP-D (Reference 7.3) WSSV DNBR remains above the Analytical DNBR limit of 1.38 for the increase in main steam flow (IMSF) event and inadvertent opening of a steam generator atmospheric dump valve (IOSGADV) event; therefore, the DNB acceptance criterion is satisfied. With the implementation of NGF, the minimum CETOP-D WSSV DNBR calculated for either the limiting case of the IMSF with a loss of power (IMSF+LOP) or the IOSGADV with a loss of power (IOSGADV+LOP) is less than the Analytical DNBR limit and therefore fuel failure analyses are performed. The overall time in DNB was evaluated and found to be less than four seconds before returning to a value greater than the DNBR limit and therefore DNB propagation will not occur. The calculated fuel failure for the IMSF+LOP and IOSGADV+LOP events are bounded by the current analyses. The results of the increase in main steam flow and inadvertent opening of a steam generator atmospheric dump valve event support both transition and full core implementation of NGF. 7.1.4. UFSAR Section 15.1.5-Steam System Piping Failures Inside and Outside Containment- (Post-Trip Main Steam Line Break) Operating Modes 1 and 2 Event is Evaluated and Reanalyzed. The Post-trip Main Steam Line Break (MSLB) event analyses for the hot full power (HFP) and hot zero power (HZP) condition were re-evaluated to determine the impact of implementing a transition or a full core of the NGF design. The implementation of the NGF design impacts the fuel rod and hot channel data used in the MacBeth DNBR critical heat flux correlation. These fuel rod and hot channel changes result in an insignificantly lower calculated DNBR value. The NRG-approved WLOP CHF correlation with the VIPRE-W code (see Section 5.2) to predict DNBR at low pressure conditions is an alternative to the current MacBeth correlation. The methodology is the same as that used in the current analysis presented in UFSAR Section . 15.1.5.3.1.2. The implementation of the NGF design did not change the input data delineated in UFSAR Section 15.1.5.3.2.2 and UFSAR Table 15.1.5-4. All four post-trip MSLB events were analyzed, and the hot full power (HFP) with Loss of AC power (LOP) remained limiting. The other three MSLB events; HFP with AC power available, hot zero power (HZP) with LOP and HZP with AC power available had minimum Macbeth DNBR values greater than five. Hence, they had sufficient margin to the 1.30 Macbeth deterministic DNBR safety limit. No CENTS computer code (Reference 7.2) cases were reanalyzed, since system model input changes due to NGF were small and the impact on the overall transient system response is insignificant. As there were no revised CENTS computer cases required, there were no changes to: the HFP with LOP sequence of events and event figures, time of minimum DNBR, time of peak linear heat rate, and the maximum calculated linear heat rate value. As the overall transient system response remains the same as the current UFSAR analyses only the DNB margin analysis is affected. With the implementation of NGF, the minimum MacBeth DNBR decreases from 2.42 to 2.38 for the HFP with LOP. The minimum MacBeth ATTACHMENT 7, Page 51
Enclosure Description and Assessment of Proposed License Amendment DNBR still remains above the MacBeth deterministic SAFDL of 1.30 and therefore, the DNB acceptance criterion is satisfied. The Linear Heat Rate (LHR) remains below the 21 kw/ft limit, thus demonstrating that fuel centerline melting does not occur and therefore, the LHR acceptance criterion is satisfied. RCS and secondary system pressures are not challenged as this is a decreasing pressure event. The results of the modes 1 and 2 post-trip steam line break event support both transition and full core implementation of NGF. 7.1.5. UFSAR Section 15.1.5-Steam System Piping Failures Inside and Outside Containment- (Pre-Trip Main Steam Line Break) Operating Modes 1 and 2 Event is Evaluated and Reanalyzed. UFSAR Section 15.1.5 describes the Pre-Trip Steam Line Break event. The analysis has been updated to determine the impact of the implementation of a transition or a full core of NGF in combination with the first time implementation of the NGF CHF correlations (Reference 7.1). The fuel rod and hot channel changes associated with the NGF design together with the use of the CHF correlations result in a revised calculated DNBR value. The methodology is consistent with that used in the current analysis presented in UFSAR Section 15.1.5.2. Implementation of NGF did not change the input data delineated in UFSAR Section 15.1.5.3 and UFSAR Table 15.5.1-3. No CENTS computer code (Reference 7.2) cases were reanalyzed, since system model input changes due to NGF were small and the impact on the overall transient system response is insignificant. Therefore, the reactor coolant system pressure performance described in UFSAR Section 15.1.5.4 is not impacted and remains applicable for the implementation of NGF at PVNGS. The overall transient system response remains the same as the current UFSAR pre-trip steam line break analyses and only the DNBR margin analysis is affected by the implementation of NGF. The initial margin required as a result of this analysis is pr~served by the limiting condition of operation on DNBR margin. The current Pre-Trip Steam Line Break Outside Containment event analysis results in a minimum CE-1 DNBR calculated by CETOP-D (Reference 7.3) which is less than the DNB SAFDL. In this case the CETOP-D results are analyzed as part of the reload process with a detailed subchannel model to demonstrate that the minimum DNB remains greater than the DNB SAFDL. With the implementation of NGF, the minimum WSSV DNBR calculated by CETOP-D for the Pre-Trip Steam Line Break Outside Containment event is greater than the Analytical DNBR limit of 1.38. Therefore, the DNB acceptance criterion is satisfied. RCS and secondary system pressures are not challenged by the introduction of NGF and therefore, the reported values in UFSAR Section 15.1.5.4.3 remain valid. *** The current and NGF Pre-Trip Steam Line Break Inside Containment event analyses result in minimum DNBR values less than the DNB SAFDL. In these cases, the conservative CETOP-D results are analyzed as part of the reload process with the VIPRE-W detailed subchannel model described to demonstrate that the minimum DNB remains greater than the DNB SAFDL. This conclusion is based on either analyzing the CETOP-D results with the detailed VIPRE-W subchannel model or increasing the initial thermal margin reserved by the LCOs to ensure the DNB SAFDL is not violated. Incorporation of either of these methods will result in the minimum DNBR remaining above the DNB SAFDL, thereby ensuring fuel cladding integrity. The results of the pre-trip steam line break event support both transition and full core implementation of NGF. ATTACHMENT 7, Page 52
Enclosure Description and Assessment of Proposed License Amendment 7.1.6. UFSAR Section 15.t6-Steam System Piping Failures Inside and Outside Containment - (Post-Trip Main Steam Line Break) Operating Mode 3 Event is Evaluated and Reanalyzed. The Post-trip Main Steam Line Break (MSLB) event analysis for the subcritical Mode 3 condition is re-evaluated to determine the impact of implementing a transition or a full core of NGF. The implementation of NGF impacts the fuel rod and hot channel data used in the MacBeth DNBR critical heat flux correlation. These fuel rod and hot channel changes result in an insignificantly lower calculated DNBR value. The NRG-approved WLOP CHF correlation with the VIPRE-W code (see Section 5.2) to predict DNBR at low pressure conditions is an alternative to the current MacBeth correlation. The methodology is the same as the current analysis presented in UFSAR Section 15.1.6.3.1. The implementation of NGF did not change the input data delineated in UFSAR Section 15.1.6.3.2 and UFSAR Table 15.1.6-2. Only the limiting subcriticai MSLB initiated at a RCS cold leg temperature of 572°F with Loss of AC power (LOP) and minimum steam generator tube plugging was analyzed. The remaining subcritical MSLB events were not analyzed as they are bounded by the limiting subcritical MSLB event No CENTS computer code (Reference 7.2) cases were reanalyzed since system model input changes due to NGF were small and the impact on the overall transient system response is insignificant. As the overall transient system response remains the same as the current UFSAR analyses, only the DNB margin analysis is affected. With the implementation of NGF, the minimum calculated MacBeth DNBR decreases' from 1.95 to 1.91 for the subcritical MSLB with LOP. The minimum MacBeth DNBR still remains greater than the deterministic SAFDL value of 1.30 and therefore, the DNB acceptance criterion is satisfied. The Linear Heat Rate remains below the 21 kw/ft limit, thus demonstrating that fuel centerline melting does not occur and therefore, the LHR acceptance criterion is satisfied. RCS and secondary system pressures are not challenged as this is a decreasing pressure event. No fuel failures are reported. The results of the Mode 3 steam line break event support both transition and full core implementation of NGF. 7.2. UFSAR Section 15.2 - Decrease in Heat Removal by the Secondary System 7.2.1. UFSAR Section 15.2.1 - Loss of External Load Event is Bounded. The Loss of External Load event is no more adverse than the Loss of Condenser Vacuum (LOCV) event (UFSAR Section 1.5.2.3). Therefore, the Loss of External Load event (with and without a single failure) remains bounded by the UFSAR LOCV event for both transition and full core implementation of the NGF design. 7.2.2. UFSAR Section 15.2.2 - Turbine Trip Event is Bounded. The Turbine Trip event is no more adverse than the LOCV event (UFSAR Section 15.2.3). Therefore, the Turbine Trip event (with and without a single failure) remains bounded by the UFSAR LOCV event for both transition and full core implementation of the NGF design. 7.2.3. UFSAR Section 15.2.3 - Loss of Condenser Vacuum Event is Evaluated and Reanalyzed. ATTACHMENT 7, Page 53
Enclosure Description and Assessment of Proposed License Amendment UFSAR Section 15.2.3 describes the Loss of Condenser Vacuum event. The analysis has been updated to determine the impact of the implementation of a transition or a full core of NGF in combination with the first time implementation of the NGF CHF correlations (Reference 7.1). The fuel rod and hot channel changes associated with the NGF design together with the use of the CHF correlations result in a revised calculated DNBR value. The methodology is consistent with that used in the current analysis presented in UFSAR Section 15.2.3.2. The implementation of NGF did not change the input data delineated in UFSAR Section 15.2.3.3 and UFSAR Table 15.2.3-3. The loss of condenser vacuum event did not require the use of the CENTS computer code (Reference 7.2) cases to reanalyze the system response as system model input changes due to NGF were small and the impact on the overall transient system response is insignificant. As the overall transient system response remains the same as the current analyses only the DNBR margin analysis is affected. The initial margin required as a result of this analysis is preserved by the limiting condition of operation on DNBR margin. With the implementation of NGF, the minimum WSSV DNBR calculated by CETOP-0 (Reference 7.3) for the LOCV event is >1.80. The minimum WSSV DNBR remains above the Analytical DNBR limit of 1.38 and therefore, the DNB acceptance criterion is satisfied. RCS and secondary system pressures are not challenged by the introduction of NGF and therefore, the reported values in UFSAR Section 15.2.3.4.C remain valid. The minimum DNBR remains above the SAFDL, thereby ensuring fuel cladding integrity. The results of the loss of condenser vacuum event support both transition and full core implementation of NGF. 7.2.4. UFSAR Section 15.2.4- Main Steam Isolation Valve Closure Event is Bounded. The Main Steam Isolation Valve Closure event is no more adverse than the LOCV event (UFSAR Section 15.2.3). Therefore, the Main Steam Isolation Valve Closure event (with and without a single failure) remains bounded by the UFSAR LOCV event for both transition and full core implementation of the NGF design. 7.2.5. UFSAR Section 15.2.5-Steam Pressure Regulator Failure Event is Not Applicable, The Steam Pressure Regulator Failure event does not apply to the Combustion Engineering Standard Safety Analysis Report (CESSAR) System 80 PWR design. 7.2.6. UFSAR Section 15.2.6-Loss of Non-emergency AC Power to Station Auxiliaries Event is Bounded. The Loss of Non-emergency AC Power to Station Auxiliaries event is no more adverse than the LOCV event (UFSAR Section 15.2.3) and the Total Loss of Reactor Coolant Flow event (UFSAR Section 15.3.1 ). Therefore, the Loss of Non-emergency AC Power to Station Auxiliaries event (with and without a single failure) remains bounded by the UFSAR LOCV and Total Loss of Reactor Coolant Flow events for both transition and full core implementation of the NGF design. 7.2.7. UFSAR Section 15.2.7- Loss of Normal Feedwater Flow Event is Bounded. The Loss of Normal Feedwater Flow event is no more adverse than the LOCV event (UFSAR Section 15.2.3). Therefore, the Loss of Normal Feedwater Flow event (with and without a single ATTACHMENT 7, Page 54
Enclosure Description and Assessment of Proposed License Amendment failure) remains bounded by the UFSAR LOCV event for both transition and full core implementation of the NGF design. 7.2.8. UFSAR Section 15.2.8- Feedwater System Pipe Breaks Event is Evaluated and Reanalyzed. UFSAR Section 15.2.8 describes the Feedwater Line Break event. The analysis has been updated to determine the impact of the implementation of a transition or a full core of NGF in combination with the first time implementation of the NGF CHF correlations (Reference 7.1 ). The fuel rod and hot channel changes associated with the NGF design together with the use of the CHF correlations result in a revised calculated DNBR value. The methodology is consistent with that used in the current analysis presented in UFSAR Section 15.2.8.2.2. The implementation of NGF did not change the input data delineated in UFSAR Section 15.2.8.2.3 and UFSAR Table 15.2.8-2. The feedwater line break event did not require the use of the CENTS computer code (Reference 7.2) cases to reanalyze the system response as system model input changes due to NGF were small and the impact on the overall transient system response is insignificant. As the overall transient system response remains the same as the current analyses only the DNBR margin analysis is affected. The initial margin required as a result of this analysis is preserved by .the limiting condition of operation on DNBR margin. With the implementation of NGF, the minimum WSSV DNBR calculated by CETOP-D (Reference 7.3) for the feedwater line break event is >1.40. The minimum WSSV DNBR remains above the Analytical DNBR limit of 1.38 and therefore, the DNB acceptance criterion is satisfied. RCS and secondary system pressures are not challenged by the introduction of NGF and therefore the reported values in UFSAR Section 15.2.8 remain valid. The minimum DNBR remains above the SAFDL, thereby ensuring fuel cladding integrity. The results of the feedwater line break event support both transition and full core implementation of NGF.
- 7.3. UFSAR Section 15.3 - Decrease in Reactor Coolant System Flowrate 7.3.1. UFSAR Section 15.3.1 -Total Loss.of Reactor Coolant Flow and UFSAR Appendix 15.D-Analysis Methods for Loss of Primary Coolant .Flow Event is Evaluated and Reanalyzed.
UFSAR Section 15.3.1 describes the total loss of forced reactor coolant event and UFSAR Appendix 15.D describes the analysis methods for loss of primary coolant flow. The analysis has been updated to determine the impact of the implementation of a transition or a full core of NGF in combination with the first time implementation .of the NGF CHF correlations (Reference 7.1). The fuel rod and hot channel changes associated with the NGF design together with the use of the CHF correlations result in a revised calculated DNBR value. The methodology is consistent with that used in the current analysis presented in UFSAR Section 15.3.1.2. The implementation of the NGF design did not change the input data delineated in UFSAR Section 15.3.1.3 and UFSAR Table 15.3.1-2. No CENTS computer code (Reference 7.2) cases were reanalyzed since model input changes due to NGF were small and the impact on the overall transient system response is insignificant. Therefore, the reactor coolant system pressure performance described in UFSAR Section 15.3.1.4 is not impacted and remains applicable for the implementation of NGF. ATTACHMENT 7, Page 55
Enclosure Description and Assessment of Proposed License Amendment The total loss of flow event was reanalyzed due to NGF-related changes in fuel rod dimensions, fuel assembly pressure drop and the resulting impact on the reactor coolant pump flow c9astdown curve. The initial thermal margin required as a result of this analysis is preserved by the limiting condition of operation on DNBR margin. With the implementation of NGF, the minimum WSSV DNBR calculated by CETOP-D (Reference 7.3) for the total loss of flow event is >1.40. The minimum WSSV DNBR remains above the Analytical DNBR limit of 1.38 and therefore, the DNB acceptance criterion is satisfied. RCS and secondary system pressures are not challenged by the introduction of NGF and therefore, the reported values in UFSAR Section 15.3.1.4.C remain valid. The minimum DNBR remains above the SAFDL, thereby ensuring fuel cladding integrity. The results of the total loss of reactor coolant flow event support both transition and full core implementation of NGF. 7.3.2. UFSAR Section 15.3.2- Flow Controller Malfunction Causing a Flow Coastdown Event is Not Applicable. The Flow Controller Malfunction event is applicable to a BWR plant and does not apply to the CESSAR System 80 PWR design. 7.3.3. UFSAR Section 15.3.3-Single Reactor Coolant Pump Rotor Seizure with Loss of Offsite Power Event is Evaluated and Reanalyzed. UFSAR Section 15.3.3 describes the single reactor coolant pump (RCP) shaft seizure event. The analysis has been updated to determine the impact of the implementation of a transition or a full core of NGF in combination with the first time implementation of the NGF CHF correlations (Reference 7.1 ). Conclusions remain consistent with UFSAR Section 15.3.3 and remain bounded by the current UFSAR Section 15.3.4 analysis. The results of the single reactor coolant pump rotor seizure with loss of offsite power event support both transition and full core implementation of NGF. 7.3.4. UFSAR Section 15.3.4-Reactor Coolant Pump Shaft Break with Loss of Offsite Power
- Event is Evaluated and Reanalyzed.
UFSAR Section 15.3.4 describes the single reactor coolant pump (RCP) sheared shaft events. The analysis has been updated to determine the impact of the implementation of a transition or ,g__ full core of NGF in combination with the first time implementation of the NGF CHF correlations (Reference 7.1). The fuel rod and hot channel changes associated with the NGF design together with the use of the CHF correlations result in a revised calculated DNBR value. The methodology is consistent with that used in the current analysis presented in UFSAR Section 15.3.4.3. The implementation of NGF did not change the input data delineated in UFSAR Section 15.3.4.3 and UFSAR Table 15.3.4~1. No C:::ENTS computer code (Reference 7.2) cases were reanalyzed since system model input changes due to NGF were small and the impact on the overall transient system response is insignificant. Therefore, the reactor coolant system pressure performance described in UFSAR Section 15.3.4.4 is not impacted and remains applicable for the implementation of NGF. The reactor coolant pump shaft break with loss of offsite power event was reanalyzed due to NGF-related changes in fuel rod dimensions, fuel assembly pressure drop, and the resulting ATTACHMENT 7, Page 56
Enclosure Description and Assessment of Proposed License Amendment impact on the reactor coolant pump flow coastdown curve. The overall transient system
- response remains the same as the current analyses and only the DNB margin analysis is affected. The calculated minimum DNBR is below the Analytical DNBR limit of 1.38 and therefore a fuel failure analysis was performed. The calculated fuel failure for the event is bounded by the current analysis. The overall time in DNB was evaluated and found to be less than four seconds before returning to a value greater than the DNBR limit and therefore DNB propagation will not occur. Conclusions remain consistent with UFSAR Section 15.3.4.
The results of the reactor coolant pump shaft break with loss of power event support both transition and full core implementation of NGF. 7.4. UFSAR Section 15.4 - Reactivity and Power Distribution Anomalies 7.4.1. UFSAR Section 15.4.1 - Uncontrolled Control Element Assembly Withdrawal from a Subcritical or Low (Hot Zero) Power Condition Event is Evaluated and Reanalyzed. UFSAR Section 15.4.1 describes the Subcritical I HZP Bank CEA Withdrawal at Power event. The analysis was updated to determine the impact of the implementation of a transition or a full core of NGF in combination with the first time implementation of the NGF CHF correlations (Reference 7.1). The fuel rod and hot channel changes associated with the NGF design together with the use of the CHF correlations result in a revised calculated DNBR value. The methodology is consistent with that used in the current analysis presented in UFSAR Section 15.4.1.2. The implementation of NGF does not change the input data delineated in UFSAR Section 15.4.1.3 and UFSAR Table 15.4.1-2. The subcritical/HZP CEA withdrawal event did not require the use of the CENTS computer code (Reference 7.2) cases to reanalyze the system response as system model input changes due to NGF were small and the impact on the overall transient system response is insignificant. As the overall transient system response remains the same as the current analyses only the DNBR margin analysis is affected. The initial margin required as a result of this analysis is preserved by the limiting condition of operation on DNBR margin. With the implementation of NGF, the minimum WSSV DNBR calculated by CETOP-D (Reference 7.3) for the Subcritical I HZP Bank CEA Withdrawal at Power is >1.40. The minimum WSSV DNBR remains above the Analytical DNBR limit of 1.38 and therefore, the DNB acceptance criterion is satisfied. RCS and secondary system pressures are not challenged by the introduction of NGF and therefore, the reported values in UFSAR Section 15.4.1.4.C remain valid. The minimum DNBR remains above the SAFDL, thereby ensuring fuel cladding integrity. The Subcritical I HZP Bank CEA Withdrawal at Power peak pressure results were obtained from cases in which the initial and transient conditions were selected to maximize heat transfer degradation and fuel centerline temperature. This event is not a peak pressure event because of the small amount of heat transferred to the RCS from fuel during the transient. The results of the subcritical/HZP CEA withdrawal event support both transition and full core implementation of NGF. 7.4.2. UFSAR Section 15.4.2 - Uncontrolled Control Element Assembly Withdrawal at Power Event is Evaluated and Reanalyzed. ATTACHMENT 7, Page 57
Enclosure Description and Assessment of Proposed License Amendment UFSAR Section 15.4.2 describes the Bank CEA Withdrawal at Power event. The analysis has been updated to determine the impact of the implementation of a transition or a full core of NGF in combination with the first time implementation of the NGF CHF correlations (Reference 7.1Error! Reference source not found.). The fuel rod and hot channel changes associated with the NGF design together with the use of the CHF correlations result in a revised calculated DNBR value. The methodology is consistent with that used in the current analysis presented in UFSAR Section 15.4.2.2. The implementation of NGF does not change the input data delineated in UFSAR Section 15.4.2.3 and UFSAR Table 15.4.2-2. The CEA withdrawal at power event did not require the use of the CENTS computer code (Reference 7.2) cases to reanalyze the system response as system model input changes due to NGF were smail and the impact on the overall transient system response is insignificant. As the overall transient system response remains the same as the current analyses only the DNBR margin analysis is affected. The initial margin required as a result of this analysis is preserved by the limiting condition of operation on DNBR margin. With the implementation of the NGF design, the minimum WSSV DNBR calculated by CETOP-D (Reference 7.3) for the CEA Withdrawal at Power event is >1.40. The minimum WSSV DNBR remains above the Analytical DNBR limit of 1.38 and therefore, the DNB acceptance criterion is satisfied. RCS and secondary system pressures are not challenged by the introduction of NGF and therefore, the reported values in UFSAR Section 15.4.2.4.C remain valid. The minimum DNBR remains above the SAFDL, thereby ensuring fuel cladding integrity. The r~sults of the bank CEA withdrawal at power event support both transition and full core implementation of NGF. 7.4.3.
- UFSAR Section 15.4.3-Single Full-Length Control Element Assembly Drop Event is Evaluated and Reanalyzed.
UFSAR Section 15.4.3 describes the single full length CEA drop event. The analysis has been updated to determine the impact of the implementation of a transition or a full core of NGF in combination with the first time implementation of the NGF CHF correlations (Reference 7.1 ). The CEA Drop Analysis was reviewed to determine the impact of the implementation of a full core of NGF in .combination with the first time implementation of the CHF correlations (Reference 7.1). The Core Operating Limits Report (Reference 7.4, COLR Figure 3.1.5-1) allow the plant to maintain the initial power for fifteen minutes. After this initial time period, the plant must reduce power according to the power reduction curve (Reference 7.4, COLR Figure 3.1.5-1). Within the fifteen minutes following the CEA rod drop, thermal margin protection is provided by the initial margin (Required Overpower Margin, ROPM) reserved in the COLSS system. Reduction of power by operator action at 15 minutes assures that the minimum DNBR remains above the DNB SAFDL and all applicable acceptance criteria are satisfied. The CEA drop analysis was performed and demonstrated that the implementation of the new Thermal-Hydraulic models (i.e., ABB-NV and WSSV with STD and NGF, respectively) in combination with the APS AOR Required Overpower Margins (ROPMs) still meets all predetermined acceptance criteria. Correspondingly, the PVNGS ROPMs will remain bounding and/or valid for the implementation of NGF at the current rated power of 3990 MWth, plus 2% uncertainty (4070 MWt total). Specifically*, the CEA drop event radial power peaking distortion factor limits have been determined to assure that the DNBR and LHR SAFDLs are not exceeded. The minimum WSSV DNBR remains above the Analytical DNBR limit of 1.38 and therefore, the DNB acceptance criterion is satisfied. For the implementation cycle and beyond, ATTACHMENT 7, Page 58
Enclosure Description and Assessment of Proposed License Amendment the reload analysis process will confirm that these radial distortion factor limits are not exceeded for the as-built core. The results of the single full length CEA drop event support both transition and full core implementation of NGF. 7.4.4. UFSAR Section 15.4.4-Startup of an Inactive Reactor Coolant Pump Event is Not Impacted. The introduction of NGF does not impact the maximum plant heatup and cooldown limits, the reactivity condition (critical/subcritical) or the number of reactor coolant pumps (RCP) required for power operation. Thus, there are no changes to the maximum RCS heatup limit, the maximum cooldown limit and the reactivity condition for Modes 3 through 5. Operation in Modes 1 and 2 require two reactor loops to be in operation per Technical Specification 3.4.4, RCS Loops Modes 1 and 2. Therefore, all RCPs are in operation and startup of an inactive RCP is precluded for Modes 1 and 2. Thus, there is no impact on the startup of an inactive RCP during any mode of operation due to the introduction of NGF. 7.4.5. UFSAR Section 15.4.5- Flow Controller Malfunction Causing an Increase in BWR Core Flow Event is Not Applicable. The Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate event is applicable to a BWR design and does not apply to the CESSAR System 80 PWR design. 7.4.6. UFSAR Section 15.4.6 - Inadvertent Deboration Event is Not Impacted. The Inadvertent Boron Dilution/ Inadvertent Deboration (IBD) event is analyzed for Modes 3 through 6. Modes 1 and 2 do not have to be analyzed because they are bounded by the Uncontrolled CEA Bank Withdrawal from Subcritical or Low (hot zero) Power Condition event (UFSAR Section 15.4.1) and the Uncontrolled CEA Withdrawal at Power event (UFSAR Section 15.4.2). For Modes 3 through 6, NGF is not impacted since DNBR is not an explicit criterion for an IBD event. The introduction of NGF does not impact the key physics input parameters (e.g., critical boron concentration and inverse boron worth) for the event. Inverse boron worth is a function of critical boron concentration, initial shutdown margin, RCS mass (volume), charging flow and the time interval to criticality. Since all of the IBD input is either not impacted by NGF or the NGF impact is negligible (i.e., RCS volume), the inverse boron worth is not impacted by the introduction of NGF. Hence, there is no impact on the AOR IBD events due to the introduction of NGF. 7.4.7. UFSAR Section 15.4.7- Inadvertent Loading of a Fuel Assembly into the Improper Position Event is Evaluated and Reanalyzed. UFSAR Section 15.4.7 describes the fuel misload event. The analysis has been updated to determine the impact of the implementation of a transition or a full core of NGF. A 30 nuclear design code was used to simulate the interchange of two assemblies and calculate the impact on core power distribution and in-core detector signals. The CECOR core monitoring code (Reference 7.5) was used to simulate the "measured core power distribution" based on the in-core detector signals calculated for the misload. Various configurations of inoperable in-core detectors were considered in the CECOR calculations. The detectability of ATTACHMENT 7, Page 59.
Enclosure Description and Assessment of Proposed License Amendment the misload was determined by applying the startup test criteria to the power distribution predicted by CECOR for beginning of cycle 20% power conditions. The mDNBR associated with the misload at any time during the cycle was calculated by the VIPRE-W code (Reference 7.6). The DNB overpower margin associated with the error between the 30 nuclear design code and CECOR predicted radial power peaking factor at full power conditions was calculated by the CETOP code (Reference 7.3). Since the CECOR measured radial power peaking factors are used by the COLSS operating limit monitoring system, no fuel failure is expected to occur as thermal margin protection is provided by the initial margin (Required Overpower Margin, ROPM) reserved in the COLSS system for the worst undetectable misload. Using the methodology described above, the worst undetectable (during startup testing) fuel assembly misload was identified. The mDNBR at nominal core conditions for the representative worst case undetectable misload including allowance for physics calculational uncertainties is significantly greater than the mDNBR limit of 1.25. The peak linear heat generation rate (PLHGR) for this worst undetectable misload is also well below the LHR SAFDL for fuel centerline melt. The associated minimum COLSS required overpower margin to prevent fuel failure for any allowed operating condition with the worst undetectable misload was calculated as a function of the number of failed in-core detectors and this value will be used by COLSS for cycles containing NGF. The inadvertent misloading of a fuel assembly into an improper position has been analyzed for PVNGS cores containing the NGF assemblies and has* been shown to not result in fuel failure for any allowed operating condition with the representative worst undetectable misload provided that the appropriate ROPM is installed in the COLSS system. This result is consistent with the current PVNGS Analysis of Record. Therefore, the reactor coolant system pressure performance described in U FSAR Section 15.4. 7 is not impacted and remains applicable for the implementation of NGF. 7.4.8. UFSAR Section 15.4.8-Control Element Assembly Ejection Event is Evaluated and Reanalyzed. UFSAR Section 15.4.8 describes the control element assembly (CEA) ejection events. The analysis has been updated to determine the impact of the implementation of a transition or a full core of NGF in combination with the first time implementation of the NGF CHF correlations (Reference 7.1). The methodology is consistent with that used in the current analysis presented in UFSAR Section 15.4.8.2. The implementation of NGF did not change the input data delineated in UFSAR Section 15.4.8.3 and UFSAR Table 15.4.8-2. Sufficient DNB thermal margin is reserved by the limiting conditions for operation to prevent all design basis events from violating the DNB specified acceptable fuel design limit (SAFDL), with the exception for the sheared shaft/seized rotor, the loss of flow from a specified acceptable fuel design limit, the pre-trip steam line break and the CEA ejection events. Of these four events, only the CEA ejection event was analyzed for DNB propagation as it results in the largest calculated fuel failure levels. Of the CEA ejection cases at varying power levels only the 20% rated thermal power case was reanalyzed as it is the limiting of the CEA ejection power levels analyzed and results in the most limiting DNB thermal margin value. DNB propagation is precluded if the maximum clad strain is less than 29.3% (Reference 7.7, Appendix A, Table 3-3). The maximum strain for the 20% power level analyzed is less than the strain limit that induces DNB propagation (29.3%). The maximum strain analysis calculates a cladding strain of <1 % for a power level of 20%. The CEA ejection event did not require the use of the CENTS computer code (Reference 7.2) cases to reanalyze the system response as system model input changes due to NGF were ATTACHMENT 7, Page 60
Enclosure Description and Assessment of Proposed License Amendment small and the impact on the overall transient system response is insignificant. As the overall transient system response remains the same as the current analyses peak reactor coolant system pressures reported remain bounding. The analysis was performed to determine the effects on the NGF design on the radial average and centerline fuel enthalpies as well as to determine the number of fuel rods that experience cladding damage (percent fuel failure). The CEA ejection event results demonstrate that the peak primary pressure remains below the acceptance criterion. The maximum fuel rod radial average and maximum incipient centerline melting enthalpy remain within the acceptance criteria. The total calculated fuel failures are bounded by the current analysis. The results of the CEA ejection event support both transition and full core implementation of NGF. . 7.5. UFSAR Section 15.5 - Increase in Reactor Coolant System Inventory 7.5.1. UFSAR Section 15.5.1-lnadvertent Operation of the ECCS Event is Not Impacted. The Inadvertent Operation of the ECCS event is assumed to actuate the two high pressure safety injection (HPSI) pumps and open the corresponding discharge valves. For this Non-LOCA evaluation, the initial RCS pressure remains above the HPSI pump shutoff head. Hence, there is no impact on the Inadvertent Operation of the ECCS event due to the introduction of NGF. The Inadvertent ECCS event is a low temperature overpressure protection (LTOP) code analysis consideration only. It is discussed in Section 14 of this Technical Analysis and in UFSAR Section 5.2.2. 7.5.2. UFSAR Section 15.5.2- Chemical and Volume Control System (CVCS) Malfunction
- Pressurizer Level Control System Malfunction with Loss of Offsite Power Event is Bounded.
The Pressurizer Level Control System (PLCS) Malfunction event produces an increasing RCS pressure that compensates for the elevated RCS temperatures, such that the available thermal margin does not degrade before the onset of the LOP. With respect to the DNBR acceptance criteria, this event is no more adverse than the Total Loss of Reactor Coolant Flow event (UFSAR Section 15.3.1 ). Thus, the overall DNBR degradation experienced during a PLCS Malfunction event with LOP remains bounded by the Total Loss of Reactor Coolant Flow event (UFSAR Section 15.5.2.2) for both transition and full core implementation of the NGF design. 7.6. UFSAR Section 15.6 - Decrease in Reactor Coolant System Inventory 7.6.1. UFSAR Section 15.6.1 - Inadvertent Opening of a Pressurizer Safety/Relief Valve The inadvertent opening of a pressurizer safety/relief valve is analyzed as part of the Emergency Core Cooling System Performance LOCA Accidents discussed in Section 8 of this Technical Analysis and in UFSAR Section 6.3.3. 7.6.2. UFSAR Section 15.6.2- Double-Ended Break of a Letdown Line Outside Containment Event is Evaluated and Reanalyzed. ATTACHMENT 7, Page 61
Enclosure Description and Assessment of Proposed License Amendment UFSAR Section 15.6.2 describes the Letdown Line Break event. The analysis has been updated to determine the impact of the implementation of a full core of NGF in combination with the first time implementation of the NGF CHF correlations (Reference 7.1). The fuel rod and hot channel changes associated with the NGF design together with the use of the CHF correlations result in a revised calculated DNBR value. The methodology is consistent with that used in the current analysis presented in UFSAR Section 15.6.2.2. Implementation of NGF did not change the input data delineated in UFSAR Section 15.6.2.3 and UFSAR Table 15.6.2-3. No CENTS computer code (Reference 7.1) cases were reanalyzed since system model input changes due to NGF were small and the impact on the overall transient system response is insignificant. Therefore, the reactor coolant system performance described in UFSAR Section 15.6.2.4 is not impacted, the minimum DNBR remains above the DNB SAFDL and all applicable acceptance criteria are satisfied. Therefore, the letdown line break event as described in UFSAR Section 15.6.2 remains applicable for the implementation of NGF. As the overall transient system response remains the same as the current AOR only the DNBR margin analysis is affected by NGF introduction. The initial margin required as a result of this analysis is preserved by the limiting condition of operation on DNBR margin. With the implementation of NGF,. the minimum WSSV DNBR calculated by CETOP-D (Reference 7.3) for the Letdown Line Break event is >1.50. The minimum WSSV DNBR remains above the Analytical DNBR limit of 1.38 and therefore, the DNB acceptance criterion is satisfied. RCS and secondary system pressures are not challenged by the introduction of NGF and therefore, the reported values in UFSAR Section 15.6.2.4.C remain valid. The results of the letdown line break event support both transition and full core implementation of NGF. 7.6.3. UFSAR Section 15.6.3 - Steam Generator Tube Rupture Event is Not Impacted. The Steam Generator Tube Rupture (SGTR) event (UFSAR Section 15.6.3) is a slow primary system depressurization transient prior to reactor trip. Post-reactor trip, credit is taken for a three-second delay for the LOP and avoids any additional degradation in DNBR. The introduction of NGF does not impact the key SGTR event's input data. Hence, there is no impact on the AOR SGTR event due to the introduction of NGF. 7.6.4. UFSAR Section 15.6.4- Radiological Consequences of Main Steam Line Failure Outside Containment (BWR) Event is Not Applicable. The Radiological Consequence of Main Steam Line (MSL) Failure Outside Containment (i3WR) event is applicable to BWR designs and does not apply to the CESSAR System 80 PWR design. 7.6.5. UFSAR Section 15.6.5 - Loss-of-Coolant Accidents The Loss of Coolant Accident events are outside the scope of the Non-lOCA transient analysis. These events are analyzed separately and discussed in Section 8 of this Technical Analysis. 7.7. UFSAR Section 15.7 - Radioactive Material Release from a Subsystem or Component Radiological accident evalua.tions are addressed in Section 11 of this Technical Analysis. ATTACHMENT 7, Page 62
Enclosure Description and Assessment of Proposed License Amendment 7.8. UFSAR Appendix 15.E- Limiting Infrequent Events (Loss of Flow from the
~pecified Acceptable Fuel Design Limit)
Event is Evaluated and Reanalyzed. UFSAR Appendix 15E describes a composite event that is evaluated to bound all infrequent events, including Anticipated Operational Occurrences (AOOs) in combination with a single active failure, with respect to the degradation in the Departure from Nucleate Boiling Ratio (DNBR). The limiting infrequent event is a loss of flow from a specified acceptable fuel design limit. The analysis has been updated to determine the impact of the implementation of a transition or a full core of NGF in combination with the first time implementation of the NGF CHF correlations (Reference 7.1). The fuel rod and hot channel changes associated with the NGF design together with the use of the CHF correlations result in a revised calculated DNBR value. The methodology is the same as that used in the current analysis presented in UFSAR Section 15E.2. The implementation of NGF had only small changes to the input data delineated in UFSAR Section 15E.3 and UFSAR Table 15E-1. No CENTS computer code (Reference 7.2) cases were reanalyzed since system model input changes due to NGF were small and the impact on the overall transient system response is insignificant. The analysis as described in UFSAR Appendix 15.E remains applicable for the implement~tion of NGF. As the overall transient system response remains the same as the current analyses, only the DNB margin analysis is affected. With the implementation of NGF, the minimum CETOP-D (Reference 7.3) WSSV DNBR calculated for the loss of flow from a specified acceptable fuel design limit is 1.18. The calculated fuel failure for the event is bounded by the current analysis. RCS pressure is not challenged by the introduction of NGF and therefore, the content in UFSAR Section 15E.4 remains valid. The results of the loss of flow from a specified acceptable fuel design limit event support both transition and full core implementation of NGF. 7.9. References 7.1. WCAP-16523-P-A, Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes, August 2007 7.2. WCAP-15996-P-A, Revision 1, Technical Description Manualfor the CENTS Code, November 2005 7 .3. CEN-160(S)-P, Revision 1-P, CETOP-D Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Units 2 and 3, September 1981 7.4. Palo Verde Nuclear Generating Station (PVNGS) Core Operating Limits Report, Unit 1 Revision 24, Unit 2 Revision 20, and Unit 3 Revision 23 7.5. CENPD-153, Revision 1, Evaluation of Uncertainty in the Nuclear Power Peaking Measured by the Self-Powered, Fixed In-Core Detector System, May 1980 7.6. WCAP-14565-P-A, VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis, October 1999 7.7. CEN-372-P-A Task 617, Revision 0, Fuel Rod Maximum Allowable Gas Pressure, May 1990 ATTACHMENT 7, Page 63
Enclosure Description and Assessment of Proposed License Amendment
- 8. ECCS PERFORMANCE LOCA ACCIDENTS 8.1. Introduction This section summarizes the Emergency Core Cooling System (ECCS) performance analyses performed for the implementation of NGF assemblies into the PVNGS units. These ECCS performance analyses were performed to demonstrate conformance to the ECCS acceptance criteria for light water nuclear power reactors, 10 CFR 50.46 (Reference 8:1 ). Analyses were performed for a spectrum of large break (LB) and small break (SB) loss-of-coolant accidents (LOCAs).
The ECCS performance analysis performed for PVNGS supports the implementation of NGF during transition and full cores. The implementation of NGF necessitated LBLOCA and SBLOCA reanalysis for the ECCS Performance AOR. The 4070 MWt analyses for implementation of the Simplified Head Assembly for LBLOCA, ZIRLO cladding implementation for SBLOCA and the ECCS Performance Analysis Comprehensive Checklist were used as a basis for the NGF ECCS performance analysis. The fuel design changes for NGF that are important for ECCS performance analyses are compared to STD assembly characteristics as follows:
- The NGF fuel design contains Optimized ZIRLOTM clad fuel rods. The STD assemblies are comprised of ZIRLO clad fuel rods.
- The NGF fuel rod cladding and fuel pellet radial dimensions are reduced compared to the STD rod design (0.374 in. versus 0.382 in.). This produces an increase in the fuel rod pitch-to-diameter ratio compared to the STD design and an increase in the core cross-sectional area for coolant flow. Also, the NGF fuel rod cladding diameter-to-thickness ratio is increased relative to the STD design. This ratio is used in calculating the engineering hoop stress across the fuel rod cladding for analy.zing any mechanical deformation of the cladding.
- The NGF hydraulic resistance is increased relative to the STD design due to the addition of mixing vane grids. As a result, a transition mixed core assessment for NGF was performed in order to address the impact of co-resident hydraulically dissimilar fuel assemblies (i.e., NGF and STD assemblies) on ECCS performance.
8.2. Objective The objective of the ECCS performance analysis is to demonstrate conformance to the ECCS acceptance criteria of 10 CFR 50.46(b): Criterion 1: Peak Cladding Temperature: The calculated maximum fuel element cladding temperature shall not exceed 2200°F. Criterion 2: Maximum Cladding Oxidation: The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation. Criterion 3: Maximum Hydrogen Generation: The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react. Criterion 4: Coolable Geometry: Calculated changes in core geometry shall be such that the core remains amenable to cooling. ATTACHMENT 7, Page 64
Enclosure Description and Assessment of Proposed License Amendment Criterion 5: Long-Term Cooling: After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long lived radioactivity remaining in the core. 8.3. Regulatory Basis As required by 10 CFR 50.46(a)(1)(i), the ECCS performance analysis must conform to the ECCS acceptance criteria identified in Section 8.2. Additionally, the ECCS performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated LOCAs of different sizes, locations and other properties sufficient to provide assurance that the most severe postulated LOCAs are calculated. The evaluation model may either be a realistic evaluation model as described in 10 CFR 50.46(a)(1 )(i) or must conform to the required and acceptable features of Appendix K ECCS Evaluation Models (Reference 8.2). The evaluation models used to perform the ECCS performance analyses documented herein are Appendix K evaluation models. 8.4. Method of Analysis WCAP-16500-P-A and its Supplement 1 (References 8.30 and 8.31, respectively), are the Core Reference Report for NGF. Section 5.2 of Reference 8.30 documents the ECCS performance methods suitable for use to analyze the implementation of NGF. The methods used for the ECCS performance analyses of PVNGS Units 1, 2 and 3 are summarized in the following sections. - The NGF design utilizes-optimized ZIRLO' (as defined in Section 4 of this Technical Analysis), an advanced cladding alloy. The implementation of Optimized ZIRLO' in CE-NSSS plants is documented in Reference 8.32 and approved by the NRC in Reference 8.33. As required by the SER limitations in Reference 8.33, the ECCS performance analysis computer codes have been updated to include the Optimized ZIRLO' cladding property changes detailed in the topical report. The use of Optimized ZIRLO' cladding in the NGF assemblies requires a cladding exemption from the requirements of 10 CFR 50.46 and 10 CFR Part 50, Appendix K. Topical Report CENPD-404-P-A, Addendum 1-A (Reference 8.32).addresses Optimized ZIRLO' and demonstrates that Optimized ZIRLO' has essentially the same properties as currently licensed ZIRLO. The Topical Report has been approved by the NRC. This license amendment request includes a proposed change to revise the PVNGS Units 1, 2, and 3 Operating Licenses to allow the use of Optimized ZIRLO' fuel rod cladding material. Acceptable fuel rod cladding material is identified in PVNGS Units 1, 2, and 3 Technical Specification (TS) 4.2.1, Fuel Assemblies. The proposed change will add Optimized ZIRLO' fuel rod cladding material as an acceptable material. A permanent exemption from certain requirements of 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors," and 10 CFR 50 Appendix K, "ECCS Evaluation Models" is required to support this change. By letter dated August 26, 2010, the NRC staff approved a temporary exemption from these requirements to support the PVNGS Units 1, 2 and 3 NGF Lead Fuel Assembly (LFA) program (Reference 8.50). The requested permanent exemption will replace the approved temporary exemption. ATTACHMENT 7, Page 65
Enclosure Description and Assessment of Proposed License Amendment 8.4.1. Large-Break LOCA (LBLOCA) The Westinghouse ECCS Performance Appendix K Evaluation Model for CE plants is the 1999 Evaluation Model (1999 EM) for LBLOCA (References 8.3, 8.4, 8.5, 8.6 and 8. 7). The 1999 EM for LBLOCA is augmented by CENPD-404-P-A for analysis of ZIRLO cladding (Reference 8.34) and by Addendum 1 to CENPD-404-P-A for analysis of Optimized ZIRLO' cladding (Reference 8.32). Additionally, the 1999 EM is further supplemented by WCAP-16072-P-A (Reference 8.35), for implementation of ZrB2 (IFBA) fuel assembly designs. The 1999 EM for LBLOCA includes the following computer codes: The CEFLASH-4A computer code (References 8.8, 8.9, 8.1 O and 8.11) is used to perform the blowdown hydraulic analysis of the reactor coolant system (RCS) and the COMPERC-11 computer code (References 8.12, 8.13 and 8.14) is used to perform the RCS refill/reflood hydraulic analysis and to calculate the containment minimum pressure. It is also used in conjunction with the methodology described in Reference 8.36 to calculate the FLECHT-based reflood heat transfer coefficients used in the hot rod heatup analysis. The HCROSS (Reference 8.37) and PARCH (References 8.15, 8.16 and 8.17) computer codes are used to calculate steam cooling heat transfer coefficients. The hot rod heatup analysis, which calculates the peak cladding temperature and maximum cladding oxidation, is performed with the STRI KIN-II computer code (References 8.18, 8.19, 8.20 and 8.21). Core-wide cladding oxidation is calculated using the COMZIRC computer code (Appendix C of Supplement 1 of Reference 8.12). The initial steady state fuel rod conditions used in the analysis are determined using the FATES3B computer code (References 8.22, 8.23 and 8.24). In performing the LBLOCA calculations, conservative assumptions are made concerning the availability of safety injection flow. It is assumed that offsite power is lost and all pumps must await diesel startup before they can begin to deliver flow. Also, it is assumed that all safety injection flow delivered to the broken cold leg is lost directly to the containment. The limiting initial fuel rod conditions used in the LBLOCA analysis (i.e., the conditions that result in the highest calculated peak cladding temperature) were determined by performing burnup dependent calculations with the 1999 EM using initial fuel rod conditions calculated by FATES3B. The LBLOCA.analysis included the analysis of both U0 2 and ZrB2 (IFBA) burnable absorber fuel rods in the NGF rod design and both U02 and Erbia burnable absorber fuel rods in the STD rod design. A study was performed to determine the most limiting single failure of ECCS equipment. The study analyzed no failure, failure of an emergency diesel generator, failure of a high pressure safety injection (HPSI) pump and a failure of a low pressure safety injection (LPSI) pump consistent with NRG-approved topical reports. Maximum safety injection pump flow rates were used in the no failure case; minimum safety injection pump flow rates were used in the emergency diesel generator, HPSI or LPSI pump failure cases. The pumps were actuated on a safety injection actuation signal (SIAS) generated by low pressurizer pressure with appropriate startup delay. Minimum refueling water storage pool temperature was used in all four cases as a result of a sensitivity study of the refueling water storage pool water temperature. The study also investigated the impact of variation in safety injection tank (SIT) pressure, water temperature and water volume on peak cladding temperature and peak local cladding oxidation. In addition, the study investigated the impact of SIT line resistance on maximum cladding temperature. *
- A spectrum of guillotine breaks in the RCP discharge leg was analyzed. As described in Section 3.4 of Reference 8.7, the discharge leg is the most limiting break location and a guillotine break is more limiting than a slot break. In particular, the 0.6, 0.8 and 1.0 Double-Ended Guillotine breaks in the RCP discharge leg (DEG/PD) were analyzed.
ATTACHMENT 7, Page 66
Enclosure Description and Assessment of Proposed License Amendment Because the NGF design has a higher pressure drop than STD fuel, a transition mixed core assessment was performed to address the effect of flow redistribution on NGF assemblies during the transition cycles consisting of co-resident hydraulically dissimilar fuel assembHes. 8.4.1.1. Plant Design Data Important core, RCS, ECCS and containment design data used in the LBLOCA analysis are listed in Table 8-1. The listed fuel rod conditions are for rod average burnup of the hot rod that produced the highest calculated peak cladding temperature. In particular, the results of this ECCS Performance analysis support a peak linear heat generation rate of 13.1 kW/ft. Plant design data for the containment (e.g., data for the containment initial conditions, containment volume, containment heat removal systems and containment passive heat sinks) were selected to minimize the transient containment pressure. The core inlet temperature was the minimum RCS cold leg temperature at full power including uncertainty.
- 8.4.1.2. Large Break LOCA Results The results for full core implementation of NGF demonstrate conformance to the ECCS acceptance criteria as summarized below. The results for the current AOR with 10% Steam Generator Tube Plugging (SGTP) are provided in Table 8-2 for comparison.
The acceptance criterion in 10 CFR 50.46(b)(2) requires, with respect to ECCS performance analyses for postulated LOCAs, that the calculated total oxidation of the cladding shall nowhere exceed 17% of the total cladding thickness before oxidation. The NRC staff position regarding this 17% oxidation limit is that it refers to the total oxidation of the cladding, including both pre-accident oxidation and accident-generated oxidation. This position is documented in a number of references, including NRC Information Notice* No. 98-29, "Predicted Increase in Fuel Rod Cladding Oxidation," dated August 3, 1998; a letter from NRC to the Nuclear Energy Institute (NEI) dated November 8, 1999 [NRC ADAMS Accession No. ML993270252]; NRC Safety Evaluations for PVNGS power uprate license amendments dated September 29, 2003 and November 16, 2005 [NRC ADAMS Accession Nos.* ML032720538 and ML053130275, respectively]; and an NRC Safety Evaluation associated with the Westinghouse Large Break LOCA (LBLOCA) Evaluation Model (EM) topical report (CENPD-132, Supplement 4-P-A, Addendum 1-P) [NRC ADAMS Accession No. ML071730336]. The PVNGS LBLOCA ECCS performance analysis for NGF with Optimized ZIRLO' cladding calculated a bounding Peak Local Oxidation (PLO) of 15. 78% for a 0.6 double-ended guillotine break in the pump discharge leg (DEG/PD) (see Table 8-3). This result corresponds to a time very early in core life, when the hot rod average burnup is approximately 0.1 GWd/MTU and the initial stored energy in the fuel rod is near its maximum SOL value. Figures 2.1.2 through 2.1.4 of Addendum 2-A to CENPD-404-P-A, "Westinghouse Clad Corrosion Model for ZIRLO and Optimized ZIRLO'," dated October 2013, show that the amount of pre-accident oxidation at this time in life would be very small. The bounding PLO analysis explicitly considered a thin layer of pre-accident oxide on the cladding surface that is representative of conditions early in core life. Thus, it is concluded that the bounding PLO analysis shows that the total local oxidation limit of 10 CFR 50.46(b)(2) is not exceeded.
- The 10 CFR 50.46(b)(2) acceptance criterion for total local oxidation would also be met for all other times in core life. Although the figures referenced above show that the thickness of a pre-accident oxide layer may be expected to increase with burnup, this is compensated for by a decrease in accident-generated oxidation, due in part to reduced initial stored energy in the fuei rods. Additionally, for burnups in excess of approximately 35 GWd/MTU, imposition of a radial fall-off curve in core design restricts the operation of fuel rods to conditions that are less limiting than the design Peak Linear Heat Generation Rate (PLHGR).
ATTACHMENT 7, Page 67
Enclosure Description and Assessment of Proposed License Amendment The acceptance criterion in 10 CFR 50.46(b)(3) likewise requires that the calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 1% of the hypothetical amount that would be generated if all of the cladding, excluding the cladding surrounding the plenum volume, were to react. The PVNGS LBLOCA ECCS performance analysis for NGF with Optimized ZIRLOTM cladding calculated a bounding Core-Wide Oxidation (CWO) of 0.813% for a 0.8 DEG/PD break (see Table 8-3). As previously noted in NRC Safety Evaluations for PVNGS power uprate license amendments dated September 29, 2003 and November 16, 2005 [NRC ADAMS Accession Nos. ML032720538 and ML053130275, respectively], pre-accident oxidation of the cladding is not expected to contribute to the LOCA maximum core-wide hydrogen generated. Table 8-3 lists the peak cladding temperature and oxidation percentages for the spectrum of large break LOCAs. Times of interest are listed in Table 8-4. The variables listed in Table 8-5 are plotted as functions of time for the 1.0 DEG/PD break in Figures 8-1 through 8-8. The variables listed in Table 8-5 and Table 8-6 are plotted as functions of time in Figures 8-9 through 8-29 for the 0.8 DEG/PD break, the limiting large break LOCA. The variables listed in Table 8-5 are plotted as functions of time for the 0.6 DEG/PD break in Figures 8-30 through 8-37. ATTACHMENT 7, Page 68
Enclosure Description and Assessment of Proposed License Amendment Table 8-1: Large Break LOCA ECCS Performance Analysis Core and Plant Design Data Quantity Value Reactor power level (102% of 3990 MWt rated power), MWt 4070 Peak linear heat generation rate (PLHGR) of the hot rod, kW/ft 13.1 Average linear heat generation rate (102% of rated), kW/ft 5.7348 Gap conductance at the PLHGR<1>, BTU/hr-ft2 -°F 2552 1 Fuel centerline temperature at the PLHGR< >, °F 3034 Fuel average temperature at the PLHGR< 1>, °F 1894 1 Hot rod gas pressure< >, psia 424 Moderator temperature coefficient at initial density, fip/°F +0.5x104 RCS flow rate, lbm/hr 155.8x106 Core flow rate, lbm/hr 151.1x106 RCS pressure, psia 2250 Cold leg temperature, °F 548.0 Hot leg temperature, °F 613.8 Plugged tubes per steam generator (max per SG) 1258 Low pressurizer pressure SIAS setpoint, psia 1670 Safety injection tank pressure (min/max), psia 602/652 Safety injection tank water volume (min/max), ft3 1750/1950 Containment pressure, psia 13.2 Containment temperature, °F 50 Containment humidity, % 100 Containment net free volume, ft3 3.0x10 6 Containment spray pump flow rate, gpm/pump 5250 Refueling water tank temperature (min/max), °F 50/130 ( 1) These quantities correspond to the rod average burn up of the hot rod (0.1 GWd/MTU) that yields the highest peak cladding temperature. ATTACHMENT 7, Page 69
Enclosure Description and Assessment of Proposed License Amendment Table 8-2: Comparison of NGF to Current AOR Results for LBLOCA CurrentAOR Parameter Criterion NGF Results Results Peak Cladding Temperature, °F 2200 2129.6 <1l 2106 Maximum Cladding Oxidation 17% 15.78% <2 l 11.9% Maximum Core-Wide Oxidation 1% <1% (3) <1% Coolable Geometry Yes Yes Yes (1) The limiting break size for PCT changed from 0.6 DEG/PD (current AOR) to 0.8 DEG/PD (NGF results). (2) The limiting break size for MCO changed from 0.8 DEG/PD (current AOR) to 0.6 DEG/PD (NGF results). (3) The limiting break size for CWO in the NGF results is unchanged from the current AOR (0.8 DEG/PD). ATTACHMENT 7, Page 70
Enclosure Description and Assessment of Proposed License Amendment Table 8-3: PCT and Oxidation Percentages for LBLOCA Break Spectrum Analysis Break Size Peak Cladding Maximum Cladding Maxim um Core-(DEG/PD) Temperature (°F) Oxidation(%) Wide Oxidation (%) Spectrum Results for Peak Cladding Temperature (1l 1.0(1) 2113.4 12.23 0.684 0.8(1) 2129.6 13.13 0.712 0.6(1) 2105.8 12.24 0.627 Case* Results for Peak Local Oxidation (2l 0.6(2) 2108.2 15.78 0.774 Case Results for Maximum Core~Wide Oxidation (3l 0.8(3) 2106.8 15.24 0.813 (1) Results are for NGF ZrB 2 fuel type at beginning of life (corresponding to FATES3B Cycle 4) (2) Results are for NGF U0 2 fuel type at beginning of life (corresponding to FATES3B Cycle 4). (3) Results are for NGF U02 fuel type at beginning of life (corresponding to FATES3B Cycle 4). ATTACHMENT 7, Page 71
Enclosure Description and Assessment of Proposed License Amendment Table 8-4: Times of Interest for LBLOCA ECCS Performance Analysis (Seconds After Break Occurs) Break Spectrum Analysis Break Size End of Start of SITs Sl 16l Hot Rod SITs On (DEG/PD) Slowdown Reflood 15l Empty Pumps on Rupture Spectrum Results for Peak Cladding Temperature 12l 1.011 l 13.55 19.36 25.28 57.87 37.67 94.82 0.8< 1l 14.79 20.61 26.5 59.2 37.77 90.73 0.6< 1l 16.97 22.83 28.68 61.53 37.97 105.82 Case Results for Peak Local Oxidation <3l 0.6< 1l 17.57 25.78 34.00 86.20 37.97 75.67 Case Results for Maximum Core-Wide Oxidation <4l 0.8<1l 15.33 23.37 31.62 83.73 37.77 62.17 (1) DEG/PD: Double Ended Guillotine Break at Pump Discharge Leg. (2) Results are for NGF Zr8 2 fuel type at beginning of life (corresponding to FATES38 Cycle 4). (3) Results are for NGF U02 fuel type at beginning of life (corresponding to FATES3B Cycle 4). (4) Results are for NGF U0 2 fuel type at beginning of life (corresponding to FATES3B Cycle 4). (5) Start of reflood is defined by the Contact Time. (6) Safety Injection ATTACHMENT 7, Page 72
Enclosure Description and Assessment of Proposed License Amendment Table 8-5: LBLOCA Analysis - Variables Plotted as a Function of Time for Each Break Core Power Pressure in Center Hot Assembly Node Leak Flow Rate Hot Assembly Flow Rate (Below and Above Hot Spot) Hot Assembly Quality Containment Pressure Mass Added to Core During Reflood Peak Cladding Temperature Table 8-6: LBLOCA Analysis - Variables Plotted as a Function of Time for Liqliting Break Mid Annulus Flow Rate Quality Above and Below the Core Core Pressure Drop Safety Injection Flow Rate into Intact Discharge Legs Water Level in Downcomer During Reflood Hot Spot Gap Conductance Maximum Local Cladding Oxidation Percentage Fuel Centerline, Fuel Average, Cladding, and Coolant Temperature at the Hot Spot Hot Spot Heat Transfer Coefficient Hot Rod (Pin) Pressure Containment Atmosphere Temperature Containment Sump Temperature Core Bulk Channel Flow Rate ATTACHMENT 7, Page 73
Enclosure Description and Assessment of Proposed License Amendment Figure 8-1: LBLOCA ECCS Performance Analysis - 1.0 DEG/PD Break - Normalized Total Core Power 1.2 I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I 1.0
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-2: LBLOCA ECCS Performance Analysis - 1.0 DEG/PD Break - Pressure in Center Hot Assembly Node 2400 2000 1600 <( CJ) a.. w a: 1200
- J CJ)
CJ) w a: a.. 800 400 0 0 5 10 15 20 25 TIME, SECONDS ATTACHMENT 7, Page 75
Enclosure Description and Assessment of Proposed License Amendment Figure 8-3: LBLOCA ECCS Performance Analysis - 1.0 DEG/PD - Break Leak Flow Rate 120000 I-I I I I I I I I-100000 PUMP Sii )E I-
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-4: LBLOCA ECCS Performance Analysis - 1.0 DEG/PD Break - Hot Assembly Flow Rate 30 20 BELOW OT SPOT
--------- - - . ABOVE OT SPOT 10
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-5: LBLOCA ECCS Performance Analysis - 1.0 DEG/PD Break - Hot Assembly Quality 1.2 1.0 I I I I I I I
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-6: LBLOCA ECCS Performance Analysis - 1.0 DEG/PD Break - Containment Pressure 60 ,_I I I I I I I I I I I I I I I I I I I I I I I I I I I- - I- - I- - I- - I- - I- - I- - I- - 50 I- - I- - I- - I-I- I-I- I- - I-40 I-I- I- <(
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-7: LBLOCA ECCS Performance Analysis - 1.0 DEG/PD Break - Mass Added to Core During Reflood 150000 125000 2 ~ 100000 w a: 0 () 0 6 75000 w 0 0 <( (f) (f) ~ 50000 TIME, SEC REFLOOD ATE, IN./SEC [0.0 - 4.72] [4.1102] [4.72 - 66.84] [1.2551] [66.84 - 500.00 [0.6399] 25000 0 0 100 200 300 400 500 TIME AFTER CONTACT, SECONDS
\
ATTACHMENT 7, Page 80
Enclosure Description and Assessment of Proposed License Amendment Figure a~a: LBLOCA ECCS Performance Analysis - 1.0 DEG/PD Break - Peak Cladding Temperature 2400 2100 u. (!) 1800 UJ 0 UJ a:
~ 1500 a:
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-9: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Normalized Total Core Power 1.2 1.0 v a: I-w ~
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-10: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Pressure in Center Hot Assembly Node 2400 2000 1600 <t: (J) a.. w a: 1200
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-11: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Leak Flow Rate 120000 100000 PUMP Sii DE - VESSEL: ~IDE - 80000 () UJ (J) ~ m _J UJ 60000 I- <( a: "\
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-12: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Hot Assembly Flow Rate 30 I ii 11 11 1\
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-13: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Hot Assembly Quality 1.2 1.0 ,, I I I
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-14: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Containment Pressure 60 50 40 <(
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-15: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Mass Added to Core During Reflood 150000 125000 ~ ~ 100000 w a: 0 () 0 6 75000 w 0 0 <( (/) (/) ~ 50000 TIME, SEC REFLOOD ATE, IN./SEC [0.0 - 4.54] [4.5132] [4.54 - 62.32] [1.2604] [62.32 - 500.00 [0.6388] 25000 0 0 100 200 300 400 500 TIME AFTER CONTACT, SECONDS ATTACHMENT 7, Page 88
Enclosure Description and Assessment of Proposed License Amendment Figure 8-16: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Peak Cladding Temperature 2400 2100 u. (!J 1800 UJ 0 UJ a:
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-17: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Mid Annulus Flow Rate 10000 5000 0 (.) w -(J) 2 ro
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-18: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Quality Above and Below the Core 1.2 1 I I I I
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-19: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Core Pressure Drop 30 20 -20 -30 0 5 10 15 20 25 TIME. SECONDS ATTACHMENT 7, Page 92
Enclosure Description and Assessment of Proposed License Amendment Figure 8-20: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Safety Injection Flow Rate into Intact Discharge Leg 24000 I I I I I I I I I I I I I I I I I I I 20000 0 w (J) 2 OJ _J 16000 ~ 0_J z LL 0 12000 I-I\ 0 w """") z I-w 8000
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-21: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Water Level in Downcomer During Reflood 18 I I I I I I I I I I I I I I
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-22: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Hot Spot Gap Conductance 2400 ~ I I I
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-23: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Local Cladding Oxidation Percentage 18 15 12 -;!2. 0 z 0 0 9 x 0 0 <l'. () 6 3 0 0 200 400 600 800 1000 TIME, SECONDS ATTACHMENT 7, Page 96
Enclosure Description and Assessment of Proposed License Amendment Figure 8-24: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Fuel CL & Ave., Clad, Coolant Temps at the Hot Spot 2400 I I I I I I I I l
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- 800 I
400 0 ' ' 0 100 200 300 400 500 TIME, SECONDS ATTACHMENT 7, Page 97
Enclosure Description and Assessment of Proposed License Amendment Figure 8-25: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Hot Spot Heat Transfer Coefficient 120 100
- u. 80 CJ w
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-26: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Hot Rod (Pin) Pressure 600 I I I I I I I I I I I I I I I-I- I-I- 500 I-I- I-Initial Gap Pres sure 424 psia I-I- Time of Ruptun ~ 90.73 sec I-I- I 400 (J) a.. w
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-27: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Containment Atmosphere Temperature 300 ' ' - 250 200 (\
~r - - -
u. (!:} w 0 w a:
- J I- 150
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-28: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Containment Sump Temperature 300 t- - t- - t- - t-t- - t-t- - t- - t-250 t- -
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-29: LBLOCA ECCS Performance Analysis - 0.8 DEG/PD Break - Core Bulk Channel Flow Rate 30000. 20000 CORE IN ET CORE E IT 10000 I () \ w I I ~ca I I
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-10000
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OF THE T TAL CORE FL W AREA
-30000 0 5 10 15 20 25 TIME, SECONDS ATTACHMENT 7, Page 102
Enclosure Description and Assessment of Proposed License Amendment Figure 8-30: LBLOCA ECCS Performance Analysis - 0.6 DEG/PD Break - Normalized Total Core Power 1.2 ~I I I I I I I I I I I I I I I I I I I I I I I I I I I I I-1.0 v
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-31: LBLOCA ECCS Performance Analysis - 0.6 DEG/PD Break - Pressure in Center Hot Assembly Node 2400 2000 1600 <( (J) a_ UJ a: 1200 (J) (J) UJ a: a_ 800 400 0 0 5 10 15 20 25 TIME, SECONDS ATTACHMENT 7, Page 104
Enclosure Description and Assessment of Proposed License Amendment Figure 8-32: LBLOCA ECCS Performance Analysis - 0.6 DEG/PD Break - Leak Flow Rate 120000 -- L L L L
-I-I-
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--------- - - . VESSEL. SIDE
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0 5 10 15 20 25 TIME, SECONDS ATTACHMENT 7, Page 105
Enclosure Description and Assessment of Proposed License Amendment Figure 8-33: LBLOCA ECCS Performance Analysis - 0.6 DEG/PD Break - Hot Assembly Flow Rate 30 20 I I I I I BELOW OT SPOT I ABOVE OT SPOT 10 () w en 2 lil w 0 ~ a: 0 u..
-10
-20
-30 0 5 10 15 20 25 TIME, SECONDS ATTACHMENT 7, Page 106
Enclosure Description and Assessment of Proposed Licens~ Amendment
*Figure 8-34: LBLOCA ECCS Performance Analysis -
0.6 DEG/PD Break - Hot Assembly Quality 1.2 1.0 ,,
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-35: LBLOCA ECCS Performance Analysis - 0.6 DEG/PD Break - Containment Pressure 60 I I I I I I 50 40 <( CJ) 0.. w a:
- '.) 30
(\ CJ) CJ) w a: 0..
\____ r---.___
20 10 0 0 100 200 300 400 500 TIME AFTER BREAK, SECONDS ATTACHMENT 7, Page 108
Enclosure Description and Assessment of Proposed License Amendment Figure 8-36: LBLOCA ECCS Performance Analysis - 0.6 DEG/PD Break - Mass Added to Core During Reflood 150000 125000
- 2:
~ 100000 W-O: 0 () 0 6 75000 w 0 0 <:( Cl) Cl) ~ 50000 TIME, SEC REFLOOD ATE, IN./SEC [0.0 - 4.77] [4.8231] [4.77 - 73.92] [1.2107] [73.92 - 500.00 [0.6323] 25000 0 0 100 200 300 400 500 TIME AFTER CONTACT, SECONDS ATTACHMENT 7, Page 109
Enclosure Description and Assessment of Proposed License Amendment Figure 8-37: LBLOCA ECCS Performance Analysis - 0.6 DEG/PD Break - Peak Cladding Temperature 2400 2100 u. CJ 1800 LI.I 0 LI.I a:
- J f-
<( a: 1500 LI.I a.. 2 LI.I f-0 <( .....J () 1200 900 600 0 100 200 300 400 500 TIME, SECONDS ATTACHMENT 7, Page 110
Enclosure Description and Assessment of Proposed License Amendment 8.4.2. Small-Break LOCA (SBLOCA) The small break LOCA ECCS performance analysis used the Supplement 2 version (referred to as the S2M or Supplement 2 Model) of the Westinghouse small break LOCA evaluation model for CE-NSSS PWRs (References 8.25, 8.26 and 8.27). The S2M for SBLOCA is augmented by CENPD-404-P-A for analysis of ZIRLO cladding (Reference 8.34) and by Addendum 1 to CENPD-404-P-A for analysis of Optimized ZIRLO' cladding (Reference 8.32). Also, the S2M is supplemented by WCAP-16072-P-A for implementation of ZrB 2 IFBA fuel assembly designs (Reference 8.35). The S2M for SBLOCA uses the following computer codes: The CEFLASH-4AS computer program (References 8.28 and 8.29) is used to perform the hydraulic analysis of the RCS until the time the safety injection tanks (SITs) begin to inject. After injection from the SITs begins, the COMPERC-11 computer program (References 8.12, 8.13 and 8.14) is used to perform the hydraulic analysis. COMPERC-11 is only used in the SBLOCA evaluation model for larger break sizes that exhibit prolonged periods of SIT flow and significant core voiding. The COMPERC-11 computer code was not run for this analysis because the limiting break size did not credit injection from the SITs. *As is typical of S2M analyses, the limiting break size was determined to be the largest small break for which the PCT occurs at approximately the same time that injection from the SITs starts. In this case, the PCT for the limiting break size was calculated to occur approximately 70 seconds after SIT injection would have started had it been credited. The hot rod cladding temperature and maximum cladding oxidation are calculated by the STRIKIN-11 computer program (References 8.18, 8.19, 8.20 and 8.21) during the initial period of forced convection heat transfer and by the PARCH computer program (References 8.15, 8.16 and 8.17) during the subsequent period of pool boiling heat transfer. Core-wide cladding oxidation is conservatively represented as the rod-average cladding oxidation of the hot rod. The initial steady state fuel rod conditions used in the analysis are determined using the FATES3B computer program (References 8.22, 8.23 and 8.24). The SBLOCA analysis was performed for the fuel rod conditions that result in the maximum initial stored energy in the fuel. The calculations included the analysis of both U0 2 and ZrB 2 burnable absorber (IFBA) fuel rods in the NGF rod design and both U02 and Erbia burnable absorber fuel rods in STD rod design. In addition, studies were performed using the PARCH computer code to determine the fuel rod internal pressures that cause cladding rupture to occur at the times that result in the maximum PCT and the maximum cladding oxidation for the spectrum of breaks analyzed. The analysis was performed using the failure of an emergency diesel generator (EOG) as the most limiting single failure of the ECCS. This failure results in a loss of one train of safety injection pumps, namely, a high pressure safety injection (HPSI) pump and a low pressure safety injection (LPSI) pump. This results in a minimum of safety injection water being available to cool the core. Based on the failure of an EOG and the design of the PVNGS ECCS, 73% of the flow from one HPSI pump is credited in the SBLOCA analysis. LPSI pump flow is not explicitly credited in the SBLOCA analysis since the RCS pressure never decreases below the LPSI pump shutoff head during the portion of the transient that is analyzed. However, 50% of the flow from one LPSI pump is available to cool the core given an EOG failure and a break in the RCP discharge leg. A spectrum of three break sizes in the RCP discharge (PD) leg was analyzed to bracket the limiting break size, which for PVNGS was the 0.07 ft2/PD break. The RCP discharge leg is the limiting break location since it maximizes the amount of spillage from the ECCS. The limiting ATTACHMENT 7, Page 111
Enclosure Description and Assessment of Proposed License Amendment SBLOCA is the largest small break for which the hot rod cladding heatup transient is terminated solely by injection from a HPSI pump. No SBLOCA mixed-core analysis is necessary during transition core cycles due to the negligible effect of variations in core hydraulic losses on SBLOCA analysis results. 8.4.2.1. SBLOCA Results The results for the 0.07 ft2/PD break, the limiting small break LOCA, demonstrate conformance to the ECCS acceptance criteria as summarized in Table 8-7. Table 8-8 provides a listing of the key initial PVNGS core and plant design data used in the analysis. Table 8-9 provides the safety injection flow versus pressure modeled in the analysis. Table 8-10 lists the peak cladding temperature and oxidation percentages for the spectrum of small break LOCAs. Times of interest are listed in Table 8-11. The variables listed in Table 8-12 are plotted as a function of time for each break in Figures 8-38 through 8-61. ATTACHMENT 7, Page 112
Enclosure Description and Assessment of Proposed License Amendment Table 8-7: Comparison of NGF to Current AOR Results for SBLOCA Current NGF Parameter Criterion AOR Results Results Peak Cladding Temperature, °F 2200 1678 1618 Maximum Cladding Oxidation 17% 4.5% 1.28% Maximum Core-Wide Oxidation 1% < 0.33% < 0.20% Coolable Geometry *Yes Yes Yes ATTACHMENT 7, Page 113
Enclosure Description and Assessment of Proposed License Amendment Table 8-8: Small Break LOCA ECCS Performance Analysis -- Key Initial Core and Plant Design Data Quantity Value Reactor Power Level (102% of 3990 MWt rated power), MWt 4070 Peak linear heat generation rate (PLHGR), kW/ft 13.1 Nominal RCS Pressure, psia 2250 Minimum RCS Flow rate, lbm/hr 155.8x106 Maximum Cold Leg Temperature, °F 566.0 Hot Leg Temperature, °F 628.6 Maximum Number of Plugged Tubes per Steam Generator 10% 1 Main Steam Safety Valve First Bank Opening Pressure ' ', psia 1311.1 Minimum Low Pressurizer Pressure Reactor Trip Setpoint 1670 (harsh environment), psia Minimum Low Pressurizer Pressure SIAS Setpoint 1670 (harsh environment), psia High Pressure Safety Injection Pump Flow Rate, gpm versus psig Table 8-9 Maximum Time Delay for Actuation of HPSI Flow 30 (with Loss of Offsite Power), sec Minimum Safety Injection Tank Injection Pressure, psia 602.0 Minimum Axial Shape Index, (ASI), ASI Units -0.3 Maximum Moderator Temperature Coefficient, (MTC), ~p/°F O.Ox104 Hot Rod Gas Pressure '2', psia 518.6 Fuel Centerline Temperature at Peak Linear-Heat Generation Rate ' 2', °F 3047 2 Fuel Average Temperature at Peak Linear Heat Generation Rate ' ', °F 1904 (1) Includes accumulation and pressure drop in main steam header. (2) These quantities correspond to the rod average burnup of the hot rod (500 MWD/MTU) that yields the maximum initial stored energy. ATTACHMENT 7, Page 114
Enclosure Description and Assessment of Proposed License Amendment Table 8-9: S8LOCA ECCS Performance Analysis - High Pressure Safety Injection Pump Flow Rate Versus RCS Pressure (with failure of an Emergency Diesel Generator) RCS Injection Injection Injection Injection Pressure PointA1 Point A2 Point 81 Point 82 (psig) (gpm) (gpm) (gpm) (gpm) 1700 0.5 0.5 0.5 0.5 1581 51.25 51.25 51.25 51.25 1483 76.75 76.75 76.75 76.75 1349 101.25 101.25 101.25 101.25 1199 123.50 123.50 123.50 123.50 993 149.25 149.25 149.25 149.25 782 172.00 172.00 172.00 172.00 605 189.00 189.00 189.00 189.QO 310 214.00 214.00 214.00 214.00 200 222.75 222.75 222.75 222.75 130 228.00 228.00 228.00 228.00 100 230.25 230.25 230.25 230.25 50 234.00 234.00 234.00 234.00 0 237.50 237.50 237.50 237.50 Notes: (1) Injection points A1 and A2 are the RCP discharge legs in one* steam generator loop. Injection points 81 and 82 are the RCP discharge legs in the other steam generator loop. (2) 27% of the total flow to the broken discharge leg is modeled to spill out the break. (3) 73% of the total flow is modeled to be split equally to each of the three intact discharge legs for the cases run in the current analysis. (4) Values are for injection from one train of pumps following a loss of an emergency diesel generator. *
- ATTACHMENT 7, Page 115
Enclosure Description and Assessment of Proposed License Amendment Table 8-10: SBLOCA ECCS Performance Spectrum Analysis Results Break Size, ft2 0.065 0.070 0.075 Peak cladding temperature, °F 1632 1678 1514 Maximum cladding oxidation, % 3.7 4.5 1.4 Core-wide cladding oxidation, % < 0.29 < 0.33 < 0.16 Table 8-11: SBLOCA ECCS Performance Analysis Times of Interest (seconds after break) Break Size HPSI Flow LPSI Flow SIT Flow Peak Cladding Delivered fo Delivered to Delivered to Temperature RCS RCS RCS Occurs 0.065 ft2/PD 213 (1) 1574 (2) 1526 0.070 ft2/PD 201 (1) 1417 (2) 1486 0.075 ft2/PD 189 (1) 1357 1387 (1) The calculation completed before LPSI flow delivery to RCS begins. (2) Injection from the SITs is not credited. This value is the time injection would have begun had it been credited. Table 8-12: SBLOCA ECCS Performance Analysis Variables Plotted as a Function of Time for Each Break Core Power Inner Vessel Pressure Break Flow Rate Inner Vessel Inlet Flow Rate Inner Vessel Two-Phase Mixture Level Heat Transfer Coefficient at Hot Spot* Coolant Temperature at Hot Spot Cladding Temperature at Hot Spot ATTACHMENT 7, Page 116
Enclosure Description and Assessment of Proposed License Amendment Figure 8-38: SBLOCA ECCS Performance Analysis - 0.065 ft2/PD Break - Core Power 1.50 t-I I I I I I I I t-t- t-t- t-t- t-1.25 t-t- t-t- t-t- t-a: UJ 1.00
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-39: SBLOCA ECCS Performance Analysis - 0.065 ft2/PD Break - Inner Vessel Pressure 2400 2000 1600 <( en a.. ur a:
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-40: SBLOCA ECCS Performance Analysis - 0.065 ft2/PD Break - Break Flow Rate 1200 L L L L L L L
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-41: SBLOCA ECCS Performance Analysis - 0.065 ft2/PD Break - Inner Vessel Inlet Flow Rate 50000 I I I I I I I I I I I I I I I I I I 40000
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-42: SBLOCA ECCS Performance Analysis - 0.065 ft2/PD Break - Inner Vessel Two-Phase Mixture Level 48 40 I-32 u. _j UJ UJ
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-43: SBLOCA ECCS Performance Analysis - 0.065 ft2/PD Break - Heat Transfer Coefficient at Hot Spot 6 10 I I I I I I I I I 5 10
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-44: SBLOCA ECCS Performance Analysis - 0.065 ft2/PD Break - Coolant Temperature at Hot Spot 1800 1550 1300 u. 0 w-a:
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-45: SBLOCA ECCS Performance Analysis - 0.065 ft2/PD Break - Cladding Temperature at Hot Spot 1800 1550 1300 LL 0 w a: I-
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-46: SBLOCA ECCS Performance Analysis - 0.070 tt2/PD Break - Core Power 1.50 - 1.25 a: LU 1.00
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-47: SBLOCA ECCS Performance Analysis - 0.070 ft2/PD Break - Inner Vessel Pressure 2400 2000 1600 <( (J) a.. w a: 1200
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-48: SBLOCA ECCS Performance Analysis - 0.070 ft2/PD Break - Break Flow Rate 1200 - 1000 I 800 j () -w (f) 2 CJ _J u.f 600
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-49: SBLOCA ECCS Performance Analysis - 0.070 ft2/PD Break - Inner Vessel Inlet Flow Rate 50000
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-50: SBLOCA ECCS Performance Analysis - 0.070 ft2/PD Break - Inner Vessel Two-Phase Mixture Level 48 40 I-32 LL _.s UJ UJ ...J UJ 24 Cf) <( I TOP OF CO E
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-51: SBLOCA ECCS Performance Analysis - 0.070 ft2/PD Break - Heat Transfer Coefficient at Hot Spot 6 10 : 5 10 4 10 :- LL 0 I C\I I-LL 3 I a: 10 I I-cc c.5 I- 2 I 10 -
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-52: SBLOCA ECCS Performance Analysis - 0.070 ft2/PD Break - Coolant Temperature at Hot Spot 1800 1550 1300 LL 0 w~ a:
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-53: SBLOCA ECCS Performance Analysis - 0.070 ft2/PD Break - Cladding Temperature at Hot Spot 1800 1550 1300 LL 0 w a: '.:) I- <( 1050 a: w a.. 2 w I-800 550 300 0 600 1200 1800 2400 3000 TIME, SEC ATTACHMENT 7, Page 132
Enclosure Description and Assessment of Proposed License Amendment Figure 8-54: SBLOCA ECCS Performance Analysis - 0.075 ft2/PD Break - Core Power 1.50 - 1.25 -
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-55: SBLOCA ECCS Performance Analysis - 0.075 ft2/PD Break - Inner Vessel Pressure 2400 2000 1600 <( (J) a. w a: 1200
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-56: SBLOCA ECCS Performance Analysis - 0.075 ft2/PD Break - Break Flow Rate 1200 I I I I I I I I I -
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-57: SBLOCA ECCS Performance Analysis - 0.075 ft2/PD Break - Inner Vessel Inlet Flow Rate 50000 I I I I- -
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-58: SBLOCA ECCS Performance Analysis - 0.075 ft2/PD Break - Inner Vessel Two-Phase Mixture Level 48 40 I-32 u. _j w w __J w 24 (/) <( I a.. ---------- I 0 I-16 BOTTOM 0 CORE 8 ----------- ----------- ----------- ----------- ----------- 0 0 300 600 900 1200 1500 TIME, SEC ATTACHMENT 7, Page 137
Enclosure Description and Assessment of Proposed License Amendment Figure 8-59: SBLOCA ECCS Performance Analysis - 2 0.075 ft /PD Break - Heat Transfer Coefficient at Hot Spot 6 10 I I I I I I I I I I I 5 10 I'--. 4 10 LL 0I C\I I-LLI 3 a: 10 I
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-60: SBLOCA ECCS Performance Analysis - 0.075 ft2/PD Break - Coolant Temperature at Hot Spot 1800 1550 1300 u. 0 w a:
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Enclosure Description and Assessment of Proposed License Amendment Figure 8-61: SBLOCA ECCS Performance Analysis - 0.075 ft2/PD Break - Cladding Temperature at Hot Spot 1800 1550 1300 u. 0 w~ a:
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Enclosure Description and Assessment of Proposed License Amendment 8.4.3. Non-Limiting Small Break Accidents 8.4.3.1. Inadvertent Opening of a Pressurizer Safety Valve There is no significant impact of NGF implementation (transition or full core) on the inadvertent opening of a pressurizer safety valve (IOPSV) analysis results. The results of the AOR for IOPSV continue to apply. 8.4.3.2. Bottom Mounted Instrumentation Break There is no significant impact of NGF implementation (transition or full core) on Bottom Mounted Instrumentation (BMI) break analysis results. The results of the AOR for BMI break continue to apply. 8.4.4. Transition Mixed Core A transition mixed core assessment was performed for NGF in order to address the impact of co-resident hydraulically dissimilar fuel assemblies (i.e., NGF and STD assemblies) on ECCS performance. The NGF core hydraulic resistance is greater than the standard fuel assembly due to the addition of mixing grids. Therefore, adjacent NGF and standard assemblies will experience a net redistribution of fiow from the higher resistance NGF assembly to the lower resistance STD assembly. This flow redistribution in the STD/NGF mixed transition cores produces a slight penalty on the NGF assembly ECCS performance during the LBLOCA. However, a smaller cross-sectional core area for coolant flow (relative to a full core of NGF assemblies) is credited in the transition core assessment to improve the core hydraulics behavior during the bloWdown period. Also, the smaller cross-sectional core area increases the core reflooding rates for the mixed core analysis during the reflood period relative to the bounding full core NGF analysis. The net impact on ECCS performance is a slight reduction in the peak cladding temperature, peak cladding oxidation and core-wide cladding oxidation percentages. Multiple core configurations were examined and bound the likely transition core scenarios and address core loading differences that are expected in the coming PVNGS cycles of operation. The transition mixed core ECCS performance assessment determined that the results were bounded by the results of the full core NGF implementation analysis. The evaluation bounds the likely transition core scenarios. 8.4.5. Post-LOCA Long Term Cooling 8.4.5.1. Introduction This section summarizes the Post-Loss-of-Coolant-Accident (LOCA) Long-Term Cooling (LTC) analyses performed for the implementation of NGF assemblies into the PVNGS Units 1, 2, and 3. The post-LOCA LTC analyses were performed to demonstrate conformance to the Emergency Core Cooling System (ECCS) acceptance criteria for light water nuclear power reactors, 10 CFR 50.46 (Reference 8.1). The analyses support PVNGS operation with both NGF and STD. In parallel, the impact of a decrease in High Pressure Safety Injection (HPSI) performance on the long term decay heat removal capability of the plant was evaluated. The post-LOCA LTC analysis used the Westinghouse evaluation model (EM) for Combustion Engineering (CE) designed pressurized water reactors (PWRs), CENPD-254-P-A. (Reference 8.38) with the exceptions discussed below. This EM is comprised of four computer codes, CEPAC, NATFLOW, CELDA and BORON. However, the Westinghouse computer code, SKBOR, was used in place of the BORON computer code described in Appendix C of Refererwe 8.38. SKBOR is essentially the same as BORON and has been approved by the Nuclear Regulatory Commission (NRC) for the intended application for a number of other ATTACHMENT 7, Page 141
Enclosure Description and Assessment of Proposed License Amendment pressurized water reactors (PWRs) such as South Texas Project, Watts Bar, D.C. Cook, Point Beach, Turkey Point, R.E. Ginna, and Beaver Valley. The conservation of mass (liquid, vapor, and solute) and energy models used in SKBOR are as applicable to PVNGS as they are to the plants listed above. Furthermore, SKBOR has been benchmarked against the LOCADM code in Appendix E.9 of WCAP-16793-NP-A (Reference 8.39). Relative to the BORON code, SKBOR provides much finer detail in the numeric solution during the period prior to sump recirculation which results in improved treatment of time dependent phenomena such as void fraction and mixing volume, two areas of concern identified by the NRC. More specifically, the NRC documented (Reference 8.40) four requirements that need to be addressed by licensees on a plant-specific basis in any submittal regarding post-LOCA LTC:
- 1. The mixing volume must be justified and the void fraction must be taken into account when computing the boric acid concentratio.n.
- 2. The mixing,volume is a variable quantity that increases with time. The analysis to determine boric acid concentration needs to account for the variation in the mixing region while considering the pressure drop in the loop. The resultant limiting boric acid concentration must be shown to remain below the precipitation limit.
- 3. The solubility limit must be justified, especially if containment pressures greater than 14.7 psia are assumed or additives are contained in the sump water.
- 4. If using a Part 50 of Title 10 of the Code of Federal Regulations (10 CFR), Appendix K model (Reference 8.2), the decay heat multiplier must be 1.2 for all times. Paragraph 50.46(b)(5) of 10 CFR (Reference 8.1) states that" ... decay heat shall be removed for an extended period of time required by the long-lived radioactivity remaining in the core."
Section l.A.4 of Appendix K, entitled "Fission Product Decay," states, in part, "The heat
- generation rates from radioactive decay of fission products shall be assumed to be equal to 1.2 times the values for infinite operating time ... " If using a nonMAppendix K model, a realistic decay heat multiplier may be used with sufficient justification.
The LTC EM consists of two separate analyses, a boron precipitation analysis and a decay heat removal analysis. The purpose of the boric acid precipitation analysis is to demonstrate that for the limiting break, i.e., a large cold leg break, the maximum boric acid concentration in the core remains below the solubility limit, thereby preventing the precipitation of boric acid in the reactor vessel. For large hot leg breaks, the short term cold side ECCS injection in excess of core boil-off will provide a substantial flushing flow through the core that will maintain the core boric acid concentration near the RWT concentration. For large cold leg breaks, boric acid could concentrate in the core until the long term simultaneous hot and cold side injection ECCS alignment occurs. The purpose of the decay heat analysis is to prove that, regardless of break size, decay heat can be removed for the long-term and that in doing so, the core remains covered with two-phase liquid, thereby ensuring that core temperatures are maintained at acceptably low values. It is important to recognize the behavioral difference between large and small break LOCAs in the long term. This difference is that the RCS will remain at high pressure for small breaks and the injection flow rate will be too low for effective cooling; thus, small breaks require cooling of the RCS by the SG's until SOC can be initiated. Large breaks, on the other hand, are adequately cooled by the injection flow because this flow is large due to the low RCS pressure; however, large breaks must utilize simultaneous hot and cold side injection to flush boric acid from the vessel. As a consequence, the LTC large break and small break analyses are different. ATTACHMENT 7, Page 142
Enclosure Description and Assessment of Proposed License Amendment The post-LOCA analyses performed for PVNGS Units 1, 2, and 3 supports current operation with STD fuel and the implementation of NGF, and considers the fuel-related geometry input (e.g. elevations and flow areas) changes when transitioning from STD to NGF fuels 1. 8.4.5.2. Objective The objective of the post-LOCA LTC analysis is to demonstrate conformance to the ECCS acceptance criterion of 10 CFR 50.46(b):
- Criterion 5: Long-term cooling. After any calculated successful initial opera.tion of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
8.4.5.3. Regulatory Basis As required by 10 CFR 50.46(a)(1)(i), the post-LOCALTC analysis must conform to the ECCS acceptance criterion identified in Section 8.4.5.2. Additionally, the post-LOCALTC analysis must address requirements listed in Section 8.4.5.1. The evaluation model may either be a realistic evaluation model as described in 10 CFR 50.46(a)(1)(i) or must conform to the required and acceptable features of Appendix K ECCS Evaluation Models (Reference 8.2). The evaluation model and supporting analytical tools used to perform the post-LOCA LTC analysis documented herein conform to the requirements for Appendix K evaluation models. 8.4.5.4. Method. of Analysis As initially described in Section 8.4.5.1, the LTC EM consists of two separate analyses, a post-LOCA boron precipitation analysis and a post-LOCA decay heat removal analysis. Both analyses were performed in accordance with the Westinghouse post-LOCA long term cooling evaluation model for Combustion Engineering (CE)-designed pressurized water reactors (PWRs), CENPD-254-P-A (Reference 8.38) with the following exceptions. The Westinghouse computer code, SKBOR, was used in place of the BORON computer code, and for each analysis, the four NRC requirements (Reference 8.40) to the LTC evaluation model were applied, as appropriate. In summary, the EM, the supporting analytical tools and the application of the methodology are in compliance with the requirements of Appendix K to 10 CFR 50 (Reference 8.2). The product of the Westinghouse LTC EM for CE-designed PWRs, is a plant-specific LTC plan based on and justified by the analyses results. 'During the long term, defined from the time the core is reflooded to when the plant is secured, operator action is needed to assure core cooling is maintained until the plant can be brought to a cold shutdown condition. In general, the Westinghouse EM long term cooling plan, documented in Reference 8.38, is a sequence of automatic and operator actions that result in acceptable post-LOCA long term cooling for all break sizes. The two major components of the plan provide ( 1) actions that ensure decay heat is removed for the long term and (2) actions that ensure the concentration of boric acid in the core remains below the solubility limit, i.e., that boric acid precipitation does not occur. In summary, the LTC analysis and results evaluations were performed to demonstrate that PVNGS Units 1, 2, and 3 will meet the requirements of ECCS Performance Criterion 5 of 10 CFR 50.46, emergency core cooling system (ECCS) Acceptance Criteria (Reference 8.1) The transition from STD to NGF only is not expected to have an effect on the decay heat analysis results. ATTACHMENT 7, Page 143
Enclosure Description and Assessment of Proposed License Amendment with implementation of NGF. In addition, the decay heat portion of the analysis evaluates reduced HPSI pump performance. 8.4.5.4.1. Post-LOCA Boron Precipitation The Westinghouse post-LOCA boron precipitation EM for CE plants is CENPD-254-P-A (Reference 8.38) along with four additional requirements identified by the NRC in Reference 8.40 which are listed in Section 8.4.5.1. As noted in Section 8.4.5.1, the SKBOR computer code was used in place of the BORON computer code described in Appendix C of Reference 8.38. Key elements of the SKBOR calculation are summarized as follows:
- A typical SKBOR calculation considers two volumes, one representing the effective vessel mixing volume (denoted as the CORE) and one representing the remaining system inventory (denoted as the SUMP). The SKBOR code allows for the flexibility to model a user specified boric acid concentration of the make-up coolant. This option was used in the PVNGS analysis to represent the RWT boric acid concentration, 4400 ppm, during the ECCS injection phase then switching to the sump concentration during the recirculation phase. Conservation of mass during the long-term cooling phase is assured by SKBOR calculations. Vapor generated due to boiling exits the CORE with a boric acid concentration of zero and is returned to the SUMP as unborated liquid. In this manner, the CORE boric acid concentration gradually increases toward the boric acid solubility limit, while the SUMP boric acid concentration gradually decreases.
- The containment atmosphere would be expected to become saturated with water vapor very quickly after a LOCA. This effect has been taken into account in the PVNGS boron precipitation analysis. The amount of water vapor that is needed to saturate the containment atmosphere is less than 40,000 lbm, or approximately 6% of the initial RCS liquid mass prior to blowdown. The amount of water vapor that is needed to saturate .
containment is also an even smaller fraction, by about an order of magnitude, of the liquid mass that would be introduced into the SUMP from the RWT and other sources. Hence, containment vapor space water mass is modeled as an initial bias and is not explicitly calculated as a function of time.
- Similarly, condensation.efficiency is not explicitly calculated as a function of time by SKBOR. 100% condensation efficiency is assumed for the vapor that leaves the CORE and enters the saturated containment atmosphere, thus making that inventory available for return to the SUMP. Hot leg switchover time is not sensitive to this SKBOR condensation efficiency model.
The mixing volume includes 50% of the lower plenum volume, which represents 100% of the liquid volume effectively reduced by 50% to account for the boric acid concentration gradient that will exist in the lower plenum. The liquid mixing volume in the core is calculated at each timestep and accounts for the void fraction. The void fraction model used in the PVNGS Units 1, 2, and 3 analysis is the modified form of the Yeh correlation (References 8.41, 8.42 and 8.43). This correlation has been used extensively in boric acid precipitation analyses for Westinghouse-designed PWRs. The liquid mixing volume in regions above the active core is calculated using the core exit void fraction adjusted for differences in flow area. The top of the mixing volume corresponds to the elevation of the top of the hot leg. This elevation is justified by evaluating the hydrostatic balance of the liquid in the reactor vessel downcomer with the inner vessel mixture level and accounting for the effect of loop pressure drop (refer to Section 8.4.5.4.1.2). The boric acid precipitation analysis used a concentration of 32 wt% as the solubility limit which is based on a RCS pressure of 20 psia, with a corresponding saturation temperature of 228°F. ATTACHMENT 7, Page 144
Enclosure Description and Assessment of Proposed License Amendment Containment pressure analysis shows that the containment pressure remains above 20 psia for at least 24 hours following a large break LOCA. 8.4.5.4.1.1. Plant Design Data The important input parameters to the boron precipitation analysis include core power, boric acid concentrations, and component water masses, which are the significant contributors to the containment sump inventory post-LOCA. Plant design data and selected input parameters used in the PVNGS Units 1, 2, and 3 NGF boron precipitation analysis are given in Table 8-13. 8.4.5.4.1.2. Pos1-LOCA Boron Precipitation Analysis Results The results of the post-LOCA boron precipitation analysis for NGF and STD fuel are summarized in Table 8-.14. The boron precipitation analysis demonstrates that a minimum flow rate of 415 gpm to both the hot side and the cold side of the RCS, initiated at three hours post-LOCA, maintains the boric acid concentration in the core below the solubility limit of 32 wt% for either fuel type. With no hot side injection flow, the boric acid concentration was calculated to reach the solubility limit at approximately 3.4 hours for both NGF and STD Fuel. Figure 8-62 shows the results of the post-LOCA boron precipitation analysis for NGF and Figure 8-63 shows the results for STD fuel. The analysis also determined that the potential for entrainment of the hot leg injection by the steam flowing in the hot legs falls below the entrainment onset criterion at approximately one hour post-LOCA using the lshii-Grolmes entrainment onset criterion of Reference 8.44 at the steaming rate of 72.43 lbm/second calculated using Appendix K decay heat with a multiplier of 1.2 for all times. The purpose of this calculation is to determine the earliest acceptable time to . initiate simultaneous hot and cold side Injection based solely on the hot leg' liquid film entrainment threshold. The liquid film entrainment threshold in the hot leg was evaluated by applying both the lshii-Grolmes inception criteria of Reference 8.44 and the Wallis-Steen liquid entrainment onset criterion of Reference 8.45. These entrainment correlations are valid for flow conditions where the liquid phase does not take up a significant volume of the pipe (such as in the hot legs post-LOCA) and viscous effects in the liquid are not dominant, i.e., the liquid phase is in the turbulent regime. The lshii-Grolmes entrainment onset criterion for the rough turbulent regime can be applied irrespective of the liquid flow direction and is; therefore, applicable for countercurrent flow that occurs in the hot leg during simultaneous hot and cold side injection. The loop pressure drop margin for NGF is shown on Figure 8..:64 and for STD fuel on r=:igure 8-65. The mixing volume used in the boron precipitations analysis is acceptable when consic;lering the effect of loop pressure drop since the loop pressure drop margin is greater than zero. The four NRC requirements, beyond those in CENPD-254, are addressed as follows:
- 1. The mixing volume takes into account the void fraction.
- 2. The time-dependent mixing volume accounts for the pressure drop in the loop.
- 3. The boric acid solubility limit, 32 wt%, is based on a containment pressure of 20 psia was credited which is greater than 14.7 psia. This value is justified by containment pressure analysis that shows the containment pressure remains above 20 psia for at
- least 24 hours following a large break LOCA. No credit for additives in the containment sump water was assumed in the analyses.
- 4. An Appendix K model has been used with a decay heat generation rate of 1.2 times the values for infinite operating time.
ATTACHMENT 7, Page 145
Enclosure Description and Assessment of Proposed License Amendment 8.4.5.4.2. Post-LOCA Decay Heat Removal The decay heat removal analysis was performed with the CELDA, NATFLOWand CEPAC computer codes (Reference 8.38, Appendix A, Appendix B, and Appendix D, respectively) using the CENPD-254-P-A post-LOCA decay heat removal approach. The CEPAC computer code calculates the SG coo Id own transient that is input to CELDA and NATFLOW. The amount of emergency feedwater that is used during the cooldown transient is also computed by CEPAC. The NATFLOW computer code calculates the natural circulation flow rate and the RCS temperatures under natural circulation flow conditions following a LOCA. The results of NATFLOW are used to determine when the RCS is cooled down to the shutdown cooling entry temperature. The CELDA computer code calculates the long term thermal hydraulic response of the RCS following a LOCA for a series of cases wherein a spectrum of break sizes is analyzed. The evaluation of RCS inventory from CELDA output is significant in justifying an acceptable LTC plan. In summary, the execution of the CENPD-254-P-A post-LOCA decay heat removal analysis methodology is designed to categorize each break (LOCA) analyzed according to the following two criteria:
- The break is small enough such that the RCS has refilled and the shutdown cooling (SOC) entry temperature has been met, and thus the SOC system is used to remove decay heat and prevent boric acid precipitation (small breaks).
- Breaks are large enough for the break flow and simultaneous hot/cold side injection to remove decay heat and prevent boric acid precipitation in the long-term (large breaks).
In addition, since the LTC methodology implementation defines a subset of the spectrum of break sizes which satisfy bo.th small break and large break LTC strategies, this analytical approach is consistent with LOCA emergency operating procedures and the underlying philosophy of CENPD-254-P-A, namely, that it can be analytically demonstrated that decay heat can be removed in the long-term for any size LOCA and that the operator can correctly identify and initiate an appropriate means of long-term decay heat removal. 8.4.5.4.2.1. Plant Design Data The major input parameters to the PVNGS Units 1, 2, and 3 decay heat removal analysis are determined by the inputs required to execute the CEPAC, NATFLOW, and CELDA computer codes.
- The SG cooldown transient calculation using CEPAC is based on core power, RCS I SG liquid and metal masses, atmospheric dump valve (ADV) area, and RCS I SG cooldown rate.
- The natural circulation calculation using NATFLOW is based on core power, SG tube flow area (reduced by tube plugging), core flow area, geometric and frictional resistances through the RCS, and initial RCS flow rate.
- The long-term thermal hydraulic RCS response calculation using CELDA is based on core power, SG parameters (same as NATFLOW), core flow area, geometric and frictional resistances through the RCS (same as NATFLOW), total RCS volume, the heat transfer coefficient and metal specific heat (sensible heat load), SG heat transfer area (reduced by tube plugging), and minimum HPSI delivery curve.
Plant design data and selected input parameters used in the PVNGS Units 1, 2, and 3 decay heat removal analysis are provided in Table 8-15. Changes in plant configuration due to the NGF and RSGs were incorporated as applicable. ATTACHMENT 7, Page 146
Enclosure Description and Assessment of Proposed License Amendment 8.4.5.4.2.2. Post-LOCA Decay Heat Removal Analysis Results A long-term cooling decay heat removal analysis was performed to create a long term cooling plan for PVNGS Units 1, 2, and 3 with reduced HPSI flow. The following paragraphs describe the CENPD-254-P-A post-LOCA decay heat removal methodology steps for the creation of the long term cooling plan from the analysis results. In the first step using CEPAC, the SG cooldown and associated condensate storage tank (CST) inventory depletion is examined. The basis used for this step is a SG cooldown initiating at two hours post-LOCA, using two SG ADVs (one ADV per SG) and resulting in a cooldown rate of 75 °F/hr. CST depletion, or the time at which the SGs were lost as heat sinks, was calculated as 9.25 hours. In the second step using NATFLOW, 7.7 hours was determined as the time to reach the actual RCS shutdown cooling (SOC) entry temperature, 335 °F. From CEPAC results in Step 1, CST inventory usage was calculated as 232,000 gallons at this time. A comparison of the time (7.7 hours) to the SOC entry temperature to the time (9.25 hours) of CST depletion, proves that the steam generators remain available as heat sinks for greater than one hour past the time in which SOC entry conditions were met. In the third step, the LTC plan decision time was determined from the previous steps. The decision time is the time that the operator decides whether to implement the large break or the small break procedure for decay heat removal (i.e., maintain simultaneous hot and cold side injection or initiate shutdown cooling, respectively). A decision time of 8 hours was selected wherein both SOC entry conditions will be met and the operators are ensured more than one hour to apply shutdown cooling initiation procedures before the steam generators are lost as heat sinks. In the fourth step, using CELDA, the largest small break is determined as the largest break for which the RCS refills before the decision time. For all break sizes up to and including this break size, it was confirmed that the RCS has sufficient inventory to initiate shutdown cooling. In this analysis, the largest small break was determined as the 0.0290 ft2 break. In the fifth step, again using CELDA, the smallest large break is determined as the smallest break for which the break flow rate serves as an adequate means of RCS heat removal after the steam generators are lost as heat sinks. This was confirmed by the core remaining covered with two-phase liquid after the time the steam generators were lost as heat sinks. In more detail, when the steam generators were lost as heat sinks, the RCS pressure will temporarily increase and then decrease. If the break area is sufficiently small, the increase in pressure will increase the break flow rate and decrease the HPSI flow rate such that the RCS inventory can decrease to the point that the core uncovers. The breaks for which this occurs are too small for the large break procedure to be acceptable. Therefore, the large break procedure was acceptable for all breaks for which*the core remained covered after the steam generators were lost as heat sinks. In this analysis, the smallest large break was determined as the 0.011 O ft2 break. In the sixth step, the LTC plan decision pressure was determined from the previous steps. The decision pressure is the value for the indicated RCS (pressurizer) pressure that, in the long-term cooling methodology, the operator uses to determine whether to implement the large break or the small break procedure for long-term decay heat removal. The decision is made at the decision time of 8 hours as documented above. The following test indicates acceptability of the decision pressure: The absolute values of the differences between the decision pressure and the RCS pressure at the decision time for either the largest small break or the smallest large break must be greater than the measurement uncertainty (+/-100) of the indicated RCS ATTACHMENT 7, Page 147
Enclosure Description and Assessment of Proposed License Amendment (pressurizer) pressure. In this analysis, a decision pressure of approximately 200 psia was determined acceptable. In summary, the decay heat removal analysis shows that, regardless of break size, decay heat can be removed for the long term and that in doing so, the core remains covered with two-phase liquid, thereby ensuring that core temperatures are maintained at acceptably low values. The analysis identified a decision time of 8 hours and a decision pressure of approximately 200 psia. At the decision time, for breaks as large as 0.0290 ft2 , the RCS has refilled, and therefore, shutdown cooling may be used as the long-term decay heat removal method up to 0.0290 ft2
- For breaks as small as 0.0110 ft2 the core remains covered with two-phase liquid, and therefore, decay heat may be removed in the long term by simultaneous hot and cold side injection (provided the SGs are maintained as heat sinks for at least 9.25 hours) down to 0.0110 ft2
- The overlap in these two break ranges (and the overlap in the associated RCS pressures) ensures that, regardless of the indicated RCS pressure (versus actual RCS pressure) at the decision time, an appropriate long term decay heat removal method would be selected by the operator.
The decay heat removal analysis also demonstrated that Auxiliary Feedwater (AFW) flowrate is sufficient to support the LTC plan. Selected analysis results are listed in Table 8-16. Figure 8-66 is a plot of break area versus RCS refill time. Figure 8-67 is a plot of RCS pressure versus break area at the decision time. Figure 8-68 tabulates break size and RCS pressure at this decision time. Figure 8-68 also indicates the range of break sizes that are large breaks (i.e., simultaneous hot and cold side injection is acceptable for long-term decay heat removal) and the range of break sizes that are small breaks (i.e., shutdown cooling is acceptable for long term decay heat removal). ATTACHMENT 7, Page 148
Enclosure Description and Assessment of Proposed License Amendment Table 8-13: Post-LOCA Boron Precipitation Analysis Plant Design Data Item Quantity Value 1 Reactor power level (102% of 3990 MWt rated power) 4070 MWt 1.2 x ANS '71 2 Decay heat model with actinides for infinite operation BAMT Parameters 3 BAMT configuration Not applicable RWT Parameters 4 RWT volume, maximum 760,000 gal 5 RWT boron concentration, maximum 4400 ppm 6 RWT temperature, minimum 50 °F SIT Parameters 7 Number of tanks, maximum 4 8 SIT boron concentration, maximum 4400 ppm 9 SIT water deliverable volume, maximum 1950 ft3 10 SIT water/gas temperature, minimum 50 °F 11 SIT gas cover pressure, maximum 651 psia Pump Parameters 12 Minimum number of HPSI pumps 1 13 Minimum number of LPSI pumps 1 4600 gpm 14 Minimum combined HPSI and LPSI flow rate (injection mode) at 20 psia 15 Minimum number of CS pumps 1 16 Minimum CSP flow rate 3000 gpm Minimum simultaneous hot and cold side safety injection flow 17 415 gpm rate RCS Parameters 18 RCS water volume 14,299 ft3 19 RCS boron concentration, maximum 2100 ppm ATTACHMENT 7, Page 149
Enclosure Description and Assessment of Proposed License Amendment Table 8-14: Post-LOCA Boron Precipitation Analysis Results Parameter Result NGF Analysis Boric acid solubility limit 32wt% Initiation time for simultaneous hot and cold side injection 3 hours Maximum core boric acid concentration with 415 gpm of hot 29.8wt% side injection started at 3 hours STD Fuel Analysis Boric acid solubility limit 32wt% Initiation time for simultaneous hot and cold side injection 3 hours Maximum core boric acid concentration with 415 gpm of hot 30.0 wt% side injection started at 3 hours ATTACHMENT 7, Page 150
Enclosure Description and Assessment of Proposed License Amendment Figure 8-62: NGF Analysis - Boric Acid Concentration in the Core Boric Acid Concen t ration (wt.%) No Core Flushin9 Flow
- - - -
- With Core f l ush 1n9 Flow
- - -
- Boric Acid Solub i l ity Limi (32 - 0 wl1.)
Moss F l ow Role (lbm/sec)
- - - Core Boil-off - ----- - -* Hot Side Safety Injection Flow (415 9pm) 40
~ ' ......_, --------.----- I
--,----------r---------
I I 200 - ~ c: 0 30 ' I I I 1
~-----*--------------*
..... I I
't I
I --- ----------- I 0 150
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I t I 1 I 1 I I o-+-__..___.___.__.._,_.___..__.___.,....._.___ ....__._----1Po--......__._ _ _.___,...o 0 2 4 6 8 Time after LOCA (hr) ATTACHMENT 7, Page 151
Enclosure Description and Assessment of Proposed License Amendment Figure 8-63: STD Fuel Analysis - Bor ic Acid Concen tra tion in the Core Bor ic Acid Concentration (wt.%) No Core Flushin9 Flow
- - - -
- Wi t h Core Flushing Flow
- - -
- Boric Acid Solub i l i ty L imit (32 . 0 wtr.)
Moss Flow Role ( l bm/sec)
- - - Core Boil-off - -- -----* Ho t S i de Sofety I njection F l ow (415 gpm) 40 ....------------.------------.-------------~~------~-250 I
I I I I ~ I I 'I
--------,.----- I
----.----------r---------
I I -+-'
~
c: 0 -+-' 30 --------------+------
- I I
I I 1 I I
'i ,
I ..._ I I 200 0
~ 150 c: <l> I I
I
-..... I <..> 20 --- *------ -- ------*--------------, *-------------
- I I ..._
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u \ 100 er -0 <( " .......... 3 c.J)
............ I I I
_______ ..._ __ *-------------- *-------------- *-------------- <b .\:::: 0
-.,.: -- -- ;* -------~
- - - :~ -------------- 50
(') en I 1 I I I - , - - - - - I I I I I I o-+-__..__.....__.__+-_.___...__.______.___.__....._--1__.....___._____--+-o I I 0 2 4 6 8 Time ofter LOCA (hr) ATTACHMENT 7, Page 152
Enclosure Description and Assessment of Proposed License Amendment Figure 8-64: NGF Analysis - Loop P ressure Drop Ma rgin Pressure Drop Ma rgin 3 *- *- *- *- *- *- *+ - *- *- *- *- *- *- *+ - *- *- *- *- *- *- *+ - *- *- *- *- *- *-
- I I. I.
- - - - *- + - *- - - *- - *- *+ - - - - - - - *+ - - *- - - - -
*- *- *- *- *- - *- *+ - *- *- - *- *- *- *+ - *- *- *- *- *- *-
- I 0 1.5 - *- - *- *- - *- +. - *- *- - *- - - - - ~ --=-~
. - - *- - *- *+ - *- *- - *- *- -
- I I
- *- - *- *- - *- *+. - *- *- *- *- - *- *+. - *- *- - *- - *- *+. - *- *- - *- - -
- I 0.5 *- *- *- *- *- *- *- *+ - *- *- *- *- *- *- *+ - *- *- - *- *- *- *+*- *- *- - *- *- *-
- 0 0 2 3 4 Time ofter LOCA (hr)
ATTACHMENT 7, Page 153
Enclosure Description and Assessment of Proposed License Amendment Figure 8-65: STD Fuel Analysis - Loop Pressure Drop Margin Pressure Drop Mor9in 3.5 - 3 *- *- *- *- *- *- *+ - *- *- *- *- *- *- *+ - *- *- *- *- *- *- *+*- *- *- *- *- *- *- * --{/') 0..
<l>
- s
{/') 2.5 I. I.
- - - - - *- + - - - - - - - *+ - - *- - - - - + - - *- - - - -
I I. {/')
<l>
2 *+ - *- - *- *- *- *- *+ - *- *- *- *- *- *- *+*- *- *- *- *- *- *-
- 0... I I 0 1.5 - *- - *- *- - *- + - *- - - *- - *
~ c:
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- *- - *- - - *- +. - *- *- *- *- - *- *+. - *- *- *- *- *- - *+. - - *- - *- *-
Cl I I 0.5 *- *- *- *- *- *- *- *+. - *- *- *- *- *- *- *+. - *- *- *- *- *- *- *+*- *- *- *- *- *- *-
- I 0
0 2 3 4 Time ofter LOCA {hr) ATTACHMENT 7, Page 154
Enclosure Description and Assessment of Proposed License Amendment Table 8-15: Post-LOCA Decay Heat Removal Analysis Input and Plant Design Data Item Quantity Value Reactor Heat Generation 1 Reactor power level (102% of 3990 MWt rated power) 4070 MWt 1.2 x ANS '73 2 with actinides 2 Decay heat model for all analysis times RCS Parameters 3 RCS cold leg temperature 566 °F 4 RCS cooldown rate, minimum 75 °F/hour 5 Time for start of cooldown 2 hours SG Parameters 6 Number of SG ADVs, total 2 (1 per SG) 7 SG tube plugging 10% 8 Total number of tubes I SG 12580 Condensate Storage Tank Parameters 9 Minimum usable liquid volume 257,000 gal 10 Maximum liquid temperature 130°F SOC Parameters Minimum Actual Shutdown Cooling Entry Temperature for 11 Analysis, including Uncertainties (maximum indicated 335 °F temperature 3) 12 Maximum temperature measurement uncertainty 15 °F Significant Pump Parameters 13 Number of HPSI pumps 1 14 Percent reduction in HPSI flow 2% Pressurizer Pressure Measurement Parameter of Interest to LTC Decay Heat Removal Analysis Pressurizer pressure measurement uncertainty, maximum 15 +/-100 psi post-LOCA (harsh environment) 2 The 1973 standard was a re-issuance of the 1971 draft standard, with a table of Fraction of Operating Power versus Time After Shutdown added. [See Attachment 1 to Reference 8.11] 3 The maximum indicated hot leg temperature for entry into shutdown cooling is 350°F. The 15°F difference between actual and indicated hot leg temperature is associated with the maximum temperature measurement uncertainty. ATTACHMENT 7, Page 155
Enclosure Description and Assessment of Proposed License Amendment Table 8-16: Summary of Results for the Decay Heat Removal Analysis Item Quantity Value 1 Minimum time empties/SGs are lost as heat sinks 9.25 hours Maximum time that Shutdown Cooling entry temperature can 7.7 hours 2 be reached CST inventory used at time SOC entry temperature was 232,000 gallons 3 reached 4 Decision time 8 hours 5 Decision pressure =200 psia 6 Smallest large break 0.0110 ft2 7 Largest small break 0.0290 ft 2 ATTACHMENT 7, Page 156
Enclosure Description and Assessment of Proposed License Amendment Figure 8-66: Break Area versus Refill Time
..a.as E CS FLOW= 1 HPSI PUM R S/SG COO DOWN BEGI SAT 2 HO RS 0.04 C\I . 0.03 I-u.
LU' a:
~
LU a: 0.02 CD 0.01 0 0 2 4 6 8 10 , RCS REFILL TIME, HOURS ATTACHMENT 7 *. Page 157
Enclosure Description and Assessment of Proposed License Amendment Figure 8-67: RCS Pressure at the Decision Time versus Break Area 1800 RE POWE = 4070 M TIME = 8 OURS AFT LOCA 1500 1200 <( en a.. w a: en 900 en w a: a.. en (.) a: 600 LB LOCA PROCEDUR 300 0 0 0.01 0.02 0.03 0.04 0.05 BREAK AREA, FT2 ATTACHMENT 7, Page 158
Enclosure Description and Assessment of Proposed License Amendment Figure 8-68: Overlap of Acceptable LTC Procedures In Terms of Cold Leg Break Size RCS Pressure at 8 Break Size Hours Post-LOCA (ft2) (psia) 0.0500 82.9 For break areas 0.0110 fr and larger, 0.0450 82.9 simultaneous hot/cold leg injection cools 0.0400 83.0 the core and flushes bode acid from the core. 0.0350 83.0 0.0300 83.l 0.0290 100.2 0.0280 114.5 0.0270 118.1 0.0250 126.7 0.0200 153.6 0.0150 199.8 2 For break areas 0.0290 ft and smaller, 0.0110 301.9 refill of the RCS disperses boric acid 0.0100 339.0 throughout the RCS and the SGs cool the RCS to the SDC entry tempernnrre. 0.0080 442.4 0.0070 523.7 0.0060 625.2 0.0050 761.8 0.0010 1590.2 0.0005 1651.4 ATTACHMENT 7, Page 159
Enclosure Description and Assessment of Proposed License Amendment 8.4.6. Generic Safety Issue 191 (GSl-191) Following a LOCA, a debris mix could collect on the sump screen and create sufficient resistance to recirculating flow that long-term core cooling might be challenged. There is also concern about the consequences of the debris that may pass through the sump screen. This debris could be ingested into the ECCS and flow into the reactor coolant system (RCS). This passed debris may collect on the fuel. These concerns have been broadly grouped under Generic Safety Issue 191 (GSl-191) (Reference 8.46). Significant work has been performed by the industry to address the issues associated with GSl-191. This has included a PWR Owners Group program which performed testing to assess the effects of the collection of debris and chemical precipitates on core components and on the head loss across the core at flow rates representative of when ECCS is realigned to recirculate coolant from the containment sump. This program utilized a partial length (4% foot) Westinghouse 17x17 OFA assembly as a representative fuel assembly for the testing. The results of this testing program are presented in WCAP-17057-P, Revision 1, "GSl-191 Fuel Assembly Yest Report for PWROG" (Reference 8.47). The use of the Westinghouse fuel assembly for the testing was based on testing of other fuel designs, including a CE Guardian' Grid assembly, as discussed in the WCAP. The WCAP provides a detailed examination of the debris loads, based on plant data, on a fuel assembly at conditions representative of both hot-leg and cold-leg loss of coolant break scenarios. The intent of this report was to encompass all "current" PWR fuel designs and to define a limiting representative fuel assembly for testing, which is the previously mentioned Westinghouse 17x17 OFA fuel assembly. There was not an SER issued for this topical report; however, there is an SER (Reference 8.48) for WCAP-16793-NP-A, Revision 2, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid" (Reference 8.49), which provides a comprehensive evaluation of the GSl-191 issue. This WCAP and the SER (References 8.48 and 8.49) both refer to WCAP-17057-P, Revision 1 (Reference 8.47) for details on the testing protocol and the testing details. The CE 16NGF fuel product to be implemented at the Palo Verde units, as detailed in [RTSR] Section 1, is acceptable with respect to the GSl-191 sump issue and is not as limiting as the fuel assemblies currently tested, as documented in WCAP-17057-P, Revision 1 (Reference 8.47). 8.4.7. Conclusions An ECCS performance analysis was completed for PVNGS at the reactor power level of 4070 MWt (102% of 3990 MWt rated power) for the implementation of NGF. The calculations included the analysis of both U02 and ZrB 2 burnable absorber (IFBA) fuel rods in the NGF rod design and both U02 and Erbia burnable absorber fuel rods in STD rod design, including a mixed core assessment. The analysis included consideration of LBLOCAs and SBLOCAs, as well as post-LOCA boron precipitation and post-LOCA decay heat removal. The limiting break size, i.e., the break size that resulted in the highest peak cladding temperature, was determined to be the 0.8 DEG/PD break. The results of the analysis demonstrate conformance to the ECCS acceptance criteria at a PLHGR of 13.1 kW/ft as follows: Criterion 1: Peak Cladding Temperature: The calculated maximum fuel element cladding temperature shall not exceed 2200 °F. Result: The ECCS performance analysis calculated a peak cladding temperature of 2130°F for the 0.8 DEG/PD break. ATTACHMENT 7, Page 160
Enclosure Description and Assessment of Proposed License Amendment Criterion 2: Maximum Cladding Oxidation: The calculated total oxidation of the cladding shall nowhere exceed 0.17 timi3s the total cladding thickness before oxidation. Result: The ECCS performance analysis calculated a maximum cladding oxidation of 0.1578 times the total cladding thickness before oxidation for the 0.6 DEG/PD break. Criterion 3: Maximum Hydrogen Generation: The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react. Result: The ECCS performance analysis calculated a maximum hydrogen generation of less than 0.01 times the hypothetical amount for the 0.8 DE,G/PD break. Criterion 4: Coolable Geometry: Calculated changes in core geometry shall be such that the core remains amenable to cooling. Result: The cladding swelling and rupture models used in the ECCS performance analysis account for the effects of changes in core geometry that would occur if cladding rupture is calculated to occur. Adequate core cooling was demonstrated for the changes in core geometry that were calculated to occur as a result of cladding rupture. In addition, the transient analysis was performed to a time when cladding temperatures were decreasing and the RCS was depressurized, thereby precluding any further cladding deformation. Therefore, a coolable geometry was demonstrated for both transition and full core NGF core loading.
- Criterion 5: Long-Term Cooling: After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low.
value and decay heat shall be removed for the extended period of time required by the long lived radioactivity remaining in the core. Result: The effects of the transition to NGF on the post-LOCA LTC analyses were analyzed. The post-LOCA boron precipitation analysis demonstrates that a minimum flow rate of 415 gpm to both the hot side and the cold side of the RCS, initiated at three hours post-LOCA, maintains the boric acid concentration in the core below the solubility limit of 32 wt% for both STD and NGF fuel types. The post-LOCA decay heat removal analysis shows that, regardless of break size, decay heat can be removed for the long term and that in doing so, the core remains covered with two-phase liquid, thereby ensuring that core temperatures are maintained at acceptably low values. The incorporation of the RSG configuration and a 2% reduction in HPSI flow were included in the decay heat removal analysis. 8.5. References 8.1. Code of Federal Regulations, Title 10, Part 50, Section 50.46, Acceptance Criteria for Emergency Core Cooling System for Light Water Nuclear Power Reactors 8.2. Code of Federal Regulations, Title 10, Part 50, Appendix K, EGGS Evaluation Models 8.3. CENPD-132P, Calculative Methods for the C-E Large Break LOCA Evaluation Model, August 1974 ATTACHMENT 7, Page 161
Enclosure Description and Assessment of Proposed License Amendment 8.4. CENPD-132P, Supplement 1, Calculational Methods for the C-E Large Break LOCA Evaluation Model, February 1975 8.5. CENPD-132-P, Supplement 2-P, Ca/cu/ational Methods for the C-E Large Break LOCA Evaluation Model, July 1975 8.6. CENPD-132, Supplement 3-P-A, Calculative Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and W Designed NSSS, June 1985 8.7. CENPD-132, Supplement4-P-A, Calculative Methods forthe C-E Nuclear Power Large Break LOCA Evaluation Model, March 2001 8.8. CENPD-133P, CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis, August 1974 8.9. CENPD-133P, Supplement 2, CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis (Modifications), February 1975 8.10. CENPD-133, Supplement 4-P, CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis, April 1977 8.11. CENPD-133, Supplement 5-A, CEFLASH-4A, A FORTRAN77 Digital Computer Program for Reactor Blowdown Analysis, June 1985 8.12. CENPD-134 P, COMPERC-11, A Program for Emergency-Refill-Reflood of the Core, August 1974 8.13. CEN PD-134 P, Supplement 1, COMPERC-11, A Program for Emergency Refill-Re flood of the Core (Modifications), February 1975 8.14. CENPD-134, Supplement 2-A,COMPERC-11, A Program for Emergency Refill- Reflood of the Core, June 1985 8.15. CENPD-138P, PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup, August 1974 8.16. CENPD-138P, Supplement 1, PARCH, A FORTRAN-JV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup (Modifications), February 1975 8.17. CENPD-138, Supplement 2-P, PARCH, A FORTRAN-JV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup, January 1977 8.18. CENPD-135P, STRIKIN-11, A Cylindrical Geometry Fuel Rod Heat Transfer Program, August 1974 8.19. CENPD-135P, Supplement 2, STRIKIN-11, A Cylindrical Geometry Fuel Rod Heat Transfer Program (Modifications), February 1975 8.20. CENPD-135, Supplement 4-P, STRIKIN-11, A Cylindrical Geometry Fuel Rod Heat Transfer Program, August 1976 8.21. CENPD-135-P, Supplement 5, STRIK/N-11, A Cylindrical Geometry Fuel Rod Heat Transfer Program, April, 1977 8.22. CENPD-139-P-A, C-E Fuel Evaluation Model, July 1974 8.23. CEN-161(8)-P-A, Improvements to Fuel Evaluation Model, August 1989 8.24. CEN-161(8)-P, Supplement 1-P-A, Improvements to Fuel Evaluation Model, January 1992 ATTACHMENT 7, Page 162
Enclosure Description and Assessment of Proposed License Amendment 8.25. CENPD-137P, Calculative Methods for the C-E Small Break LOCA Evaluation Model, August 1974 8.26. CENPD-137, Supplement 1-P, Calculative Methods for the C-E Small Break LOCA Evaluation Model, January 1977 8.27. CENPD-137, Supplement 2-P-A, Calculative Methods for the ABB CE Small Break LOCA Evaluation Model, April 1998 8.28. CENPD-133P, Supplement 1, CEFLASH-4AS, A Computer Program for the Reactor B/owdown Analysis of the Small Break Loss of Coolant Accident, August 1974 8.29. CENPD-133, Supplement 3-P, CEFLASH-4AS, A Computer Program for the Reactor Blowdown Analysis of the Small Break Loss of Coolant Accident, January 1977 8.30. WCAP-16500-P-A, Revision 0, CE 16x16 Next Generation Fuel Core Reference Report, August 2007 8.31. WCAP-16500-P-A Supplement 1, Revision 1, Application of CE Setpoint Methodology for CE 16x16 Next generation Fuel (NGF), December 2010 8.32. CENPD-404-P-A Addendum 1-A, Optimized ZIRLO', July 2006 8.33. Letter from H. N. Berkow (NRC) to J. A. Gresham (Westinghouse), Final Safety Evaluation for Addendum 1 to Topical Report WCAP-12610-P-A and CENPD-404-P-A,'Optimized ZIRLO (TAC No. MB8041), June 10, 2005 8.34. CENPD-404-P-A, Implementation of ZIRLO Cladding Material in CE Nuclear Power Fuel Assembly Designs, November 2001 8.35. WCAP-16072-P-A, Revision 0, Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs, August 2004 8.36. CENPD-213-P, Application of FLECHT Reflood Heat Transfer Coefficients to C-E's 16x16 Fuel Bundles, January 1976 8.37. LD-81-095, Enclosure 1-P-A, C-E EGGS Evaluation Model, Flow Blockage Analysis, Decembei1981
- 8.38. CENPD-254-P-A, Post-LOCA Long Term Cooling Evaluation Model, June 1980 8.39. WCAP-16793-NP-A, Revision 2, Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid, July 2013 8.40. ADAMS Accession No. ML053220569, Suspension of NRG Approval for use of Westinghouse Topical Report CENPD-254-P, 'Post-LOCA Long-Term Cooling Model,'
Due to Discovery of Non-Conservative Modeling Assumptions During Calculations Audit (TAC NO. MB1365), November 2005 8.41. J .. P. Cunningham and H. C. Yeh, £xperiments and Void Correlation forPWR Small-Break LOCA Conditions, Trans. ANS 17, p. 369-370 (1973) 8.42. L. E. Hochreiter and H. C. Yeh, Mass Effluence During FLECHT Forced Reflood Experiments, Nuclear Engineering and Design, 60, p. 413-429 (1980) 8.43. H.C. Yeh, Modification of Void Fraction Calculation, Proceedings of the Fourth International Topical Meeting on Nuclear Thermal-Hydraulics, Operations and Safety, Volume 1, Taipei, Taiwan, June 6, 1988 8.44. Ishii, M., Grolmes, M.A., Inception Criteria for Droplet Entrainment in Two-Phase Concurrent Film Flow, AIChE Journal, Vol. 21, No. 2, pp. 308-318, March 1975 ATTACHMENT 7, Page 163
Enclosure Description and Assessment of Proposed License Amendment 8.45. Wallis, G. 8., One-dimensional Two-phase Flow, McGraw-Hill Book Company, 1969 8.46. Generic Safety Issue 191 (GSl-191), Assessment of Debris Accumulation on Pressurized Water Reactor (PWR) Sump Performance 8.47-. WCAP-17057-P, Revision 1, GSl-191 Fuel Assembly Test Report for PWROG, September 2011 8.48. US NRC SER; Final Safety Evaluation for Pressurizer Water Reactor Owners Group Topical Report WCAP-16793-NP, Revision 2, Evaluation of Long-Term Cooling Considering Particulate Fibrous and Chemical Debris in the Recirculating Fluid (TAC No. ME1234), April 8, 2013 (ADAMS Accession No. ML13084A161) 8.49. WCAP-16793-NP-A, Revision 2, Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid, July 2013 (ADAMS Accession Nos. ML13239A114 and ML13239A115) 8.50. Letter from J. R. Hall (USNRC) to R. K. Edington (APS) of August 26, 2010, Palo Verde Nuclear Generating Station, Unit 3 - Temporary Exemption from the Requirements of 10 CFR Part 50, Section 50.46 and Appendix K (TAC No. ME2590) (ADAMS Accession No: ML101900254) ATTACHMENT 7, Page 164
Enclosure Description and Assessment of Proposed License Amendment
- 9. CONTAINMENT RESPONSE ANALYSIS Implementation of NGF will result in an increase to the pressure differential across the reactor core. This will result in a slight reduction to reactor coolant system (RCS) flow rate, which in turn will slightly increase RCS T-hot. The increased T-hot will result in a slight increase to Steam Generator (SG) operating pressure. RCS T-cold will be maintained at 557.7 °F with no change to RCS Operating pressure.
These differences are indiscernible relative to the plant nominal operating point and represent inconsequential changes to system operating and design limits. 9.1. Mass and Energy Release Analysis for Postulated Loss-of-Coolant Accidents An evaluation of the NGF impacts on the LOCA Mass and Energy (M&E) AORs was performed. A comparison of fuel parameters and operating conditions was performed in the evaluation. The change to the fuel geometry and associated pressure drops were determined to be negligible for transition and full core NGF. Fuel parameters such as core average linear heat rate, pellet and cladding geometry, centerline temperature, decay heat and metal/water reaction were determined to either be bounded by the AOR inputs or have a negligible effect on the mass and energy release analysis. The RCS initial conditions also remain bounded by the AOR. The evaluation concludes that the LOCA M&E AORs remain applicable for NGF transition. 9.2. Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures Inside Containment An evaluation of the NGF impact on the MSLB M&E AOR was performed. A comparison of fuel parameters and operating conditions was performed in the evaluation. The M&E source energy based on NGF operating conditions will remain bounded by the AOR MSLB source energy. The rated thermal power, uncertainty and RCP heat are the same. The cold leg temperature program remains bounded by the values used in the AOR. The RSG pressure in the AOR bounds the NGF RSG pressure. The secondary energy remains bounded by the AOR. The RSG inventories used in the AOR remain bounding for the 102% and 75% power cases. The AOR feedwater temperature bounds the NGF feedwater temperature. Maximum RCS flow will maximize the RCS heat transfer to the RSGs. The AOR RCS flow bounds the transition and full core NGF RCS flow rate. The evaluation concludes that the MSLB M&E AOR remains applicable for NGF transition. 9.3. Mass and Energy Release for Containment Subcompartments The tributary line break transients that contribute mass and energy release to the containment subcompartments are of short duration (approximately 1 second). There is not sufficient time for the reactor core and the primary and secondary sides to interact to significantly affect the mass and energy releases. The initial conditions that have an impact on the analysis, such as maximum RCS pressure and temperature, do not change as a result of the transition and full core NGF. The evaluation concludes that the subcompartment M&E AOR remains applicable for NGF transition. ATTACHMENT 7, Page 165
Enclosure Description and Assessment of Proposed License Amendment
- 10. RADIOLOGICAL SOURCE TERM EVALUATIONS Source Terms for evaluating the radiological consequences of postulated accidents (LBLOCA) are based on methodology as described in Chapter 15 of the UFSAR. NGF fuel design parameters (e.g., initial uranium mass, burnup, power factors, and operating histories) are essentially equivalent to those for current CE 16x16 VAF with ZIRLO clad [STD fuel].
Additionally, system model input changes due to NGF are small and the impacts to overall transient system responses are insignificant. Based on these facts NGF has been found to introduce no changes that would affect Chapter 15 source terms. UFSAR Section 11.1 provides Reactor Coolant System specific activities for 1% failed fuel conditions. NGF evaluations similar to those for accident source terms have found that fission product specific activities reported in UFSAR Section 11.1 remain conservative and bounding for NGF. UFSAR Chapter 11 also presents source term data for tritium, nitrogen-16, crud activities, and carbon-14. Production rates for activation products as specified in the UFSAR remain bounding or are not appreciably affected by NGF fuel. ATTACHMENT 7, Page 166
Enclosure Description and Assessment of Proposed License Amendment
- 11. RADIOLOGICAL ACCIDENT EVALUATIONS All limiting offsite and control room dose consequences reported in UFSAR Chapters 15 and 6.4. 7 remain bounding and applicable with NGF fuel. As described in Section 1O of this Technical Analysis, accident source terms do not change and the impacts on the overall transient responses are insignificant.
11.1. Compliance with Regulatory Guide 1.25 The peak NGF maximum fuel rod discharge pressure for NGF is 1500 psig. This is at variance with the maximum fuel rod discharge pressure of 1200 psig specified in Reference 11.1. However, the NGF value of 1500 psig has been reviewed and approved in Reference 11.2 where it was concluded that there is reasonable assurance that fuel rod design pressures of up to 1500 psig will not invalidate analysis assumptions related to iodine decontamination from Reference 11.1. 11.2. References 11.1. NRC Regulatory Guide 1.25, Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors, Revision 0, March 1972 11.2. WCAP-16072-P-A, Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs, August 2004 ATTACHMENT 7, Page 167
Enclosure Description and Assessment of Proposed License Amendment
- 12. SETPOINTS ANALYSIS Setpoints analysis uses the Modified Statistical Combination of Uncertainties (MSCU) methodology (Reference 12.1 ), as the basis for stochastically combining uncertainty terms to calculate Core Operating Limits Supervisory System (COLSS) DNB Power Operating Limit (POL) and Core Protection Calculator (CPC) DNBR addressable uncertainty constants.
The incorporation of NGF at PVNGS Units 1, 2 and 3 requires modifications to the MSCU methodology in order to model the NGF design and to address its thermal-hydraulic characteristics. Therefore, the MSCU methodology is augmented by Reference 12.2 to support the change in fuel to NGF. This methodology is used without exception in support of the transition and full core implementation of NGF at PVNGS Units 1, 2 and 3. The MSCU methodology described and approved in Reference 12.1, as augmented by Reference 12.2, has been approved by the NRC for use in the setpoint analysis of cores that use NGF assemblies (e.g., PVNGS Units 1, 2 and 3). The overall uncertainty factors determined using the MSCU methodology described in Reference 12.1 and the MSCU process steps described in Reference 12.2 ensure that the COLSS DNB POL calculations and the CPCS DNBR calculations will be conservative to at least a 95% probability and a 95% confidence level. 12.1. References 12.1. CEN-356(V)-P-A, Revision 01-P-A, Modified Statistical Combination of Uncertainties, May 1988 12.2. WCAP-16500-P-A, Supplement 1, Revision 1, Application of CE Setpoint Methodology
. for CE 16x16 Next Generation Fuel (NGF), December 201 O ATTACHMENT 7, Page 168
Enclosure Description and Assessment of Proposed License Amendment
- 13. STRUCTURAL ANALYSIS Compliance with ASME Code requirements for Class 1 components and supports located inside Containment have been evaluated with NGF using methodology described in UFSAR Section 3.9. Under the various conditions of design for the Systems, Structures and Components (SSCs), only faulted loads from Branch Line Pipe Break (BLPB) LOCA loads are slightly affected by NGF. Analyses of Record for affected Class 1 SSCs have been evaluated and in all cases, ASME Code allowable stress limits will be satisfied upon implementation of NGF in the Palo Verde units.
ATTACHMENT 7, Page 169
Enclosure Description and Assessment of Proposed License Amendment
- 14. DESIGN, SYSTEMS, AND COMPONENTS ANALYSIS 14.1. Fluid Systems Analysis 14.1.1. RCS Flow One of the impacts of a transition to NGF at the PVNGS units will be an increase in core pressure drop. This local change on RCS loop conditions caused by the transition to NGF was evaluated with respect to RCS flow rate and loop pressure drop values. The Westinghouse hydraulic model AOR was fully evaluated to determine the impact of this increased core pressure drop. The specific changes in fuel pressure loss were developed. The hydraulic model evaluates RCS flow at both at-power conditions and during start up or shut down conditions. Forty different scenarios were evaluated with tube plugging, temperature, RCP performance, dP bias (minimum/maximum bias) and number of operating pumps being varied to establish minimum and maximum local flow rate and pressure drop conditions.
The results of the analyses performed show a decrease in overall RCS flow rate, consistent with expectations for an increase in core pressure drop. The full power decrease in RCS flow rate corresponds to approximately [ ]a,c in reactor vessel flow rate. 14.1.2. Hydraulic Loads As a consequence of the increased core pressure drop for NGF fuel, normal operating design hydraulic loads for the reactor vessel and internals were recalculated for a full core containing NGF assemblies for use in the structural analyses addressed in Section 13 of this Technical Analysis. ATTACHMENT 7, Page 170
Enclosure Description and Assessment of Proposed License Amendment
- 15. OTHER ISSUES 15.1. Other Technical Specification Considerations 15.1.1. TS Safety Limit 2.1.1.1- DNBR Safety Limit 10 CFR 50.36, Technical Specifications, defines a safety limit as a limit upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity. 10 CFR 50.36 also states "A Limiting Safety System Setting is the setting for automatic protective devices related to those variables having significant safety functions. Where a limiting safety, system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded."
10 CFR Part 50 General Design Criterion (GDC) 10, Reactor Design, requires that specified fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95% confidence level (95/95 DNB criterion) that DNB will not occur. No changes are required of the Technical Specification Safety Limit 2.1.1.1 which addresses the Departure from Nucleate Boiling Ratio (DNBR) safety limit of 1.34. This DNBR safety limit is based on use of the CE-1 critical heat flux (CHF) correlation. The WSSV and ABB-NV CHF correlations will be used in the NGF safety and setpoint analyses. However, because of existing hardware limitations, the Core Protection Calculator (CPC) algorithm will retain the CE-1 correlation. Since the CPC thermal-hydraulic algorithm retains the CE-1 correlation, any change to the DNBR-Low trip setpoint and Allowable Value would introduce inconsistency between the trip setpoint and the Control Room monitors. To ensure that the Plant Operators have consistency between the trip setpoint and their Control Room monitors (i.e., a human factors concern), the DNBR-Low trip setpoint and Allowable Value will remain set at 1.34. The TS SL 2.1.1 Bases will be revised to include text to address TS SL 2.1.1.1 both prior to and after NGF implementation: Prior to NGF implementation: The minimum value of the DNBR during normal operation and design basis AOOs is limited to 1.34, based on a statistical combination of CE-1 CHF correlation and engineering factor uncertainties, and is established as an SL. Additional factors such as rod bow and spacer grid size and placement will determine the limiting safety system settings required to ensure that the SL is maintained. After NGF implementation: The minimum value of the DNBR during normal operation and design basis Anticipated Operational Occurrences (AOOs) is limited to 1.34 using the WSSV and ABB-NV correlations for the first NGF transition core. This value is based on a statistical combination of CHF correlation and engineering factor uncertainties, and is established as a SL for the first NGF transition core. For the second NGF transition core and subsequent cores with NGF, the minimum value of the DNBR during normal operation and design basis AOOs is limited to 1.25 using the WSSV and ABB-NV correlations. This value is based on a statistical combination of CHF correlation and engineering factor uncertainties. Additional ATTACHMENT 7, Page 171
Enclosure Description and Assessment of Proposed License Amendment factors such as rod bow and placement will determine the limiting safety system settings required to ensure that the SL is maintained. The WSSV and ABB-NV correlations are used in the NGF safety and setpoint analyses. However, because of existing hardware limitations, the CPC algorithm will retain the CE 1* correlation and the DNBR-Low trip setpoint and Allowable Value of 1.34. To maintain consistency with the CPC setpoint, the safety limit will remain at 1.34 after the first NGF transition core. The adjustment to the lower DNBR limit will be made within the safety and setpoint analyses. The CPC power uncertainty factor used in the DNBR calculations (addressable constant BERR1), is calculated using the WSSV and ABB NV correlations in accordance with the setpoint methodology of Reference 1. The BERR1 constant is calculated such that a CPC trip at a DNBR of 1.34 using the CE-1 CHF correlation in the CPC assures that the bounding DNBR safety limit of 1.25 for the WSSV and ABB-NV correlations will not be violated during normal operations and AOOs to at least a 95/95 probability I confidence level. Thus, the trip setpoint listed in TS SL 2.1.1.1 conservatively remains at 1.34 even though credit is being obtained (via the adjusted BERR1 constant) for the improved WSSV and ABB-NV correlations and the DNBR safety limit of 1.25. Therefore, compliance with 10 CFR 50.36 and GDC 10 is maintained. 15.1.2. TS Safety Limit 2.1.1.2- Peak Fuel Centerline Temperature Safety Limit Technical Specification Safety Limit 2.1.1.2 provides the peak fuel centerline temperature with a statement that it should be adjusted for burnable poisons per CENPD-382-P-A. Westinghouse Topical Report CENPD-382-P-A addresses erbium as a burnable poison. Zirconium Diboride (ZrB2) Integral Fuel Burnable Absorber (IFBA) is used in the NGF fuel assembly design. Topical Report WCAP-16072-P-A, Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs, describes the use of Zirconium Diboride (ZrB2) Integral Fuel Burnable Absorber (IFBA) used in the NGF fuel assembly designs (Reference 4). Neither WCAP 16072-P-A nor its Safety Evaluation requires a burnable poison adjustment to the peak fuel centerline temperature. As such, no change is required for TS SL 2.1.1.2. The TS SL 2.1.1 Bases state that the design melting temperature of new fuel is adjusted downward depending on the amount of burnup and amount and type of burnable poison in the fuel. This statement is generic and remains valid. 15.2. End-of-Life Grid Crush Strength for NGF Appendix A to NUREG-0800 Standard Review Plan (SRP) Section 4.2, Fuel System Design" (Reference 15.1) provides NRC review guidance for the evaluation of fuel assembly structural response to externally applied forces. The review guidance contained in SRP 4.2 indicates that it is acceptable to assume that fuel spacer grid strength at the beginning-of-life is most limiting. However, NRC Information Notice (IN) 2012-09 (Reference 15.2) states that Operating Experience (OE) regarding the effects of in-reactor service on fuel assembly component response to externally applied forces (i.e., earthquakes and postulated pipe breaks in the reactor coolant system) challenge this existing NRC staff guidance. Specifically, OE shows that the crush strength of fuel assembly spacer grids may decrease during the life of a fuel assembly due to the effects of irradiation. The Pressurized Water Reactor Owners Group (PWROG) is proposing a resolution for the IN 2012-09 issue for Westinghouse and Combustion Engineering fuel designs that will be submitted for NRC review and approval. As discussed in the NRC public meeting held with ATTACHMENT 7, Page 172 .
Enclosure Description and Assessment of Proposed License Amendment APS on March 24, 2016 (ADAMS accession number ML16088A060), since a resolution has not yet been approved, APS is requesting a License Condition to address IN 2012-09. This License Condition would require APS to incorporate NRG-approved guidance into the current licensing basis regarding fuel assembly integrity under externally applied forces as described in IN 2012-09 within 2 cycles following Mode 4 entry with the first NGF transition core. APS will notify the NRC when this action is completed. 15.3. Spent Fuel Pool Criticality By letter dated November 25, 2015 (Reference 15.3) as supplemented by letter dated January 29, 2016 (Reference 15.4), APS submitted a license amendment request regarding an updated criticality safety analysis for both new and spent fuel storage that included NGF as a type of fuel that could be present. The analyses supporting implementation of NGF in the core are independent from the analyses completed for the Spent Fuel Pool (SFP) criticality (References 15.3 and 15.4). Approval of both LARs are necessary for use of NGF at PVNGS. 15.4. Fukushima Orders Implementation of NGF at PVNGS will not impact compliance with NRC Orders EA-12-049 (Mitigation Strategies for Beyond-Design-Basis External Events) or EA-12-051 (Reliable Spent Fuel Pool Instrumentation). 15.5. References 15.1. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP), Section 4.2, Fuel System Design, Revision 3, March 2007 (ADAMS Accession No. ML070740002) 15.2.. Information Notice (IN) 2012-09, Irradiation Effects on Fuel Assembly Spacer Grid Crush Strength, June 28, 2012 (ADAMS accession number ML113470490) 15.3. Letter from D. C. Mims (APS) to USN RC of November 25, 2015, Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3; Docket Nos. STN 50-528, 50-529, and 50-530; License Amendment Request to Revise Technical Specifications to Incorporate Updated Criticality Safety Analysis, (ADAMS Accession No. ML071160348) 15.4. Letter from M. L. Lacal (APS) to USNRC of January 29, 2016, Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3; Dock(3t Nos. STN 50-528, 50-529, and 50-530; Supplemental Information Regarding License Amendment Request to Revise Technical Specifications to Incorporate Updated Criticality Safety Analysis, (ADAMS Accession No. ML16043A361) ATTACHMENT 7, Page 173
Enclosure Description and Assessment of Proposed License Amendment
- 16. ACRONYMS Acronym Meaning 2-D Two-Dimensional 3-D Three-Dimensional ABB/CE ASEA Brown-Boveri/Combustion Engineering, now Westinghouse ABB-NV ABB Non-Vane Correlation ADAMS NRC Agencywide Document Access Management System ADV Atmospheric Dump Valve AFW Auxiliary Feedwater ANS American Nuclear Society AOO Anticipated Operational Occurrence AOPM Available Overpower Margin AOR Analysis of Record APS Arizona Public Service ARO All Rods Out ASI Axial Shape Index ASME American Society of Mechanical Engineers AST Alternative Source Term BAMT Boric Acid Makeup Tank BE Best Estimate BLPB Branch Line Pipe Break BMI Bottom Mounted Instrumentation BOC Beginning-of-Cycle BOL Beginning-of-Life BWR Boiling Water Reactor cal/gm Calories per gram CBC Critical Boron Concentration CCFL Counter-Current Flow CCL Comprehensive Checklist CE Combustion Engineering, now Westinghouse CEA Control Element Assembly (Control Rod)
CEAE Control Element Assembly Ejection CEAW Control Element Assembly Withdrawal CEDM Control Element Drive Mechanism CESSAR Combustion Engineering Standard Safety Analysis Report CFR US Code of Federal Regulations CHF Critical Heat Flux COLR Core Operating Limits Report COLSS Core Operating Limits Supervisory System ATTACHMENT 7, Page 174
Enclosure Description and Assessment of Proposed License Amendment Acronym Meaning CPC(S) Core Protection Calculator (System) CPS Core Protection System CST Condensate Storage Tank CyN Cycle N (of operation) OBA Design Basis Accident DBE Design Basis Earthquake DC Downcomer - Down flow coolant pathway outside core barrel DEG Double Ended Guillotine DEGB Double Ended Guillotine Break DEVL Design Verification List DFWT Decrease in Feedwater Temperature DNB Departure from Nucleate Boiling DNBR Departure from Nucleate Boiling Ratio DOR Division of Responsibility ECCS Emergency Core Cooling System EOG Emergency Diesel Generator EFPD Effective Full Power Days EFPH Effective Full Power Hours EM Evaluation Model EOC End-of-Cycle EOL End-of-Life EPAC Equivalence Table for the Physics Assessment Checklist EPRI Electric Power Research Institute F8H Nuclear Enthalpy Rise Hot Channel Factor FACTS Fuel Assembly Compatibility Test System Fe Tong Fe shape factor for power shape correction Fq, Fq Power Distribution Total Peaking Factor Fr Power Distribution Integrated Peaking Factor Fxy Power Distribution Planar Peaking Factor FOi Fuel Duty Index FHA Fuel Handling Accident FHI Fuel Handling Incident FIV Flow-Induced Vibration FP Fuel Performance GDC General Design Criterion (or Criteria) GL Generic Letter GTRF Grid-to-Rod Fretting GWd/MTU Gigawatt-Day(s) per Metric Ton Uranium ATTACHMENT 7, Page 175
Enclosure Description and Assessment of Proposed License Amendment Acronym Meaning HAR Hot Assembly Rod HFP Hot Full Power HPSI High Pressure Safety Injection HR Hot Rod HZP Hot Zero Power ICI In-Core Instrument ID Inside Diameter IFBA Integral Fuel Burnable Absorber IFM Intermediate Flow Mixing IFWF Increase in Main Feedwater Flow IN Information Notice IOSGADV Inadvertent Opening of a Steam Generator Atmospheric Dump Valve kw/ft Kilowatt per foot LAR License Amendment Request LBB Leak Before Break LBLOCA Large Break LOCA LCO Limiting Condition of Operation LFA Lead Fuel Assembly LEF Lower End Fitting LHR Linear Heat Rate LLC Limited Liability Company. LOCA Loss of Coolant Accident LOCV Loss of Condenser Vacuum LOF Loss of Flow LOOP Loss of Offsite Power LOP Loss of AC Power LPD Local Power Density, as in LPD Trip LPSI Low Pressure Safety Injection [ ]a,c [ r*c LTC Long Term Cooling LUA Lead Use Assemblies (referred to as LFAs at PVNGS) M&E Mass and Energy mFDI Modified Fuel Duty Index MOS Modified Outer Strap MSCU Modified Statistical Combination of Uncertainties MSLB Main Steam Line Break MTC Moderator Temperature Coefficient MTU Metric Ton of Uranium ATTACHMENT 7, Page 176
Enclosure Description and Assessment of Proposed License Amendment Acronym Meaning MUR Measurement Uncertainty Recapture MWD/MTU, MWd/MTU Megawatt-Day(s) per Metric Ton Uranium MWt, MWth Megawatt(s) Thermal NCLO No Clad Lift Off NFM PVNGS Nuclear Fuel Management NGF Next Generation Fuel NOp Normal Operation NRC,USNRC United States Nuclear Regulatory Commission NSSS Nuclear Steam Supply System OBE Operating Basis Earthquake OD Outer Diameter OE Operating Experience OST 0 uter Strap Tab PAC Physics Assessment Checklist PCM Percent Mil (Unit of Reactivity= 10-5 Lik/k) PCT Peak Cladding Temperature PD Pump Discharge pdf or p.d.f. or PDF fuel failure Probability Distribution Function PDIL Power Dependent Insertion Limit Pl Potential Issue PIE Post-Irradiation Examination PLCS Pressurizer Level Control System PLHGR Peak Linear Heat Generation Rate PLHR Peak Linear Heat Rate PLO Peak Local Oxidation PMS Plant Monitoring System POL Power Operating Limit PSV Pressurizer Safety Valve PTM power-to-centerline melt PVNGS Palo Verde Nuclear Generating Station PWR Pressurized Water Reactor PWROG Pressurized Water Reactor Owners Group QA Quality Assurance RAI, RAls Request(s) for Additional Information RCP Reactor Coolant Pump RCS Reactor Coolant System RDB Reload Data Block RFO Radial Fall-Off ATTACHMENT 7, Page 177
Enclosure Description and Assessment of Proposed License Amendment Acronym Meaning RIP Rod Internal Pressure ROPM Required Over Power Margin RPS Reactor Protection System RSE Reload Safety Evaluation RSG Replacement Steam Generator RTSR Reload Transition Safety Report RVGVS Reactor Vessel Gas Vent System RVLMS Reactor Vessel Level Monitoring System RWT Refueling Water Tank SAFDL Specified Acceptable Fuel Design Limits SAM Seismic Anchor Motion SBLOCA Small Break LOCA scu Statistical Combination of Uncertainties soc Shutdown Cooling SE US NRC Safety Evaluation SER US NRC Safety Evaluation Report SFP Spent Fuel Pool SG Steam Generator SGTP Steam Generator Tube Plugging SGTR Steam Generator Tube Rupture SHA Simplified Head Assembly SIA Safety Injection SIAS Safety Injection Actuation Signal SIS Safety Injection System SIT Safety Injection Tank SL Safety Limit SLB Steam Line Break SP, SP's Set Point(s) SR Seized Rotor SR Surveillance Requirement SRA Stress Relief Annealed SRP Standard Review Plan SS Sheared Shaft SS Side Supported SSC Systems, Structures and Components SSE Safe Shutdown Earthquake STAR Startup Test Activity Reduction STD Standard Fuel (also known as Value-Added Fuel) ATTACHMENT 7, Page 178
Enclosure Description and Assessment of Proposed License Amendment Acronym Meaning Tavg Average Temperature TBD To Be Determined TCD Thermal Conductivity Degradation TER Technical Evaluation Report T-H Thermal Hydraulic TORC Thermal-Hydraulics of Reactor Core TR Topical Report TRD Thermal Reaction Accumulated Duties TS Technical Specifications TSTF Technical Specification Task Force UFSAR Updated Final Safety Analysis Report VIPER Vibration Investigation and Pressure-Drop Experimental Research VIPRE Versatile Internals and Component Program for Reactors VOPT Variable Over Power Trip WCAP Westinghouse Commercial Atomic Power WLOP Westinghouse Low-Pressure Correlation wssv Westinghouse Side Supported Vane Correlation wt% Weight Percent ATTACHMENT 7, Page 179}}