ML12286A015
| ML12286A015 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 09/30/2012 |
| From: | GE-Hitachi Nuclear Energy Americas |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| DRF 0000-0107-2302 NEDO-33566, Rev 0 | |
| Download: ML12286A015 (129) | |
Text
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 2.8 Reactor Systems 2.8.1 Fuel System Design Regulatory Evaluation The fuel system consists of arrays of fuel rods, burnable poison rods, spacer grids and springs, end plates, channel boxes, and reactivity control rods. Exelon reviewed the fuel system to ensure that:
(1) the fuel system is not damaged as a result of normal operation and AOOs; (2) fuel system damage is never so severe as to prevent control rod insertion when it is required; (3) the number of fuel rod failures is not underestimated for postulated accidents; and (4) coolability is always maintained. The review covered fuel system damage mechanisms, limiting values for important parameters, and performance of the fuel system during normal operation, AOOs, and postulated accidents. The regulatory acceptance criteria are based on: (1) 10 CFR 50.46, insofar as it establishes standards for the calculation of ECCS performance and acceptance criteria for that calculated performance; (2) GDC-10, insofar as it requires that the reactor core be designed with appropriate margin to assure that SAFDLs are not exceeded during any condition of normal operation, including the effects of AOOs; (3) GDC-27, insofar as it requires that the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the ECCS, of reliably controlling reactivity changes under postulated accident conditions, with appropriate margin for stuck rods, to assure the capability to cool the core is maintained; and (4)
GDC-35, insofar as it requires that a system to provide abundant emergency core cooling be provided to transfer heat from the reactor core following any LOCA.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-27, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:
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Draft GDC-6, Draft GDC-37, Draft GDC-41, and Draft GDC-44. There is no Draft GDC directly associated with current GDC-27.
The Fuel System Design is described in PBAPS UFSAR Section 3, "Reactor."
Technical Evaluation Both PBAPS units plan to transition to GNF2 fuel in Cycle 19 and will continue to use only GEH/GNF fuel types through EPU implementation. Because PBAPS uses GNF2 fuel, the CLTR (Reference 1) is not applicable for fuel design dependent evaluations. Therefore, the fuel product line design is evaluated on a plant-specific basis in accordance with ELTRI (Reference 2).
The EPU evaluations assume a reference equilibrium core of GNF2 fuel. GNF2 fuel is resident in the PBAPS core. The fuel design limits are established for all new fuel product line designs as a part of the fuel introduction and reload analyses. ((
At the CLTP as well as at the EPU RTP conditions, all fuel design limits will be met through fuel bundle and core design combined with plant operational strategies. However, revised loading patterns, larger batch sizes and potentially new fuel designs may be used to provide additional operating flexibility and maintain fuel cycle length.
Therefore, because the fuel design limits are evaluated in accordance with approved methodology for each core reload, the assessment of the PBAPS fuel product line design is acceptable.
Conclusion The effects of the proposed EPU on the fuel system design of the fuel assemblies, control systems, and reactor core have been reviewed. Exelon concludes the review has adequately accounted for the effects of the proposed EPU on the fuel system and demonstrated: (1) the fuel system will not be damaged as a result of normal operation and AOOs; (2) the fuel system damage will never be so severe as to prevent control rod insertion when it is required; (3) the number of fuel rod failures will not be underestimated for postulated accidents; and (4) the fuel is adequately cooled during all operational modes. Based on this, Exelon concludes the fuel system and associated analyses will continue to meet the requirements of 10 CFR 50.46 and the current licensing basis following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to the fuel system design.
2.8.2 Nuclear Design Regulatory Evaluation Exelon reviewed the nuclear design of the fuel assemblies, control systems, and reactor core to ensure fuel design limits will not be exceeded during normal operation and anticipated operational transients, and the effects of postulated reactivity accidents will not cause significant damage to the RCPB or impair the capability to cool the core. The review covered core power distribution, reactivity coefficients, reactivity control requirements and control provisions, control rod patterns 2-360
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) and reactivity worths, criticality, burnup, and vessel irradiation. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires the reactor core be designed with appropriate margin to assure that SAFDLs are not exceeded during any condition of normal operation, including the effects of AOOs; (2) GDC- 11, insofar as it requires the reactor core be designed so that the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity; (3) GDC-12, insofar as it requires the reactor core be designed to assure that power oscillations, which can result in conditions exceeding SAFDLs, are not possible or can be reliably and readily detected and suppressed; (4) GDC-13, insofar as it requires I&C be provided to monitor variables and systems affecting the fission process over anticipated ranges for normal operation, AOOs and accident conditions, and to maintain the variables and systems within prescribed operating ranges; (5) GDC-20, insofar as it requires the protection system be designed to initiate the reactivity control systems automatically to assure that acceptable fuel design limits are not exceeded as a result of AOOs and to automatically initiate operation of systems and components important to safety under accident conditions; (6) GDC-25, insofar as it requires the protection system be designed to assure that SAFDLs are not exceeded for any single malfunction of the reactivity control systems; (7) GDC-26, insofar as it requires two independent reactivity control systems be provided, with both systems capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes; (8) GDC-27, insofar as it requires the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the ECCS, of reliably controlling reactivity changes under postulated accident conditions, with appropriate margin for stuck rods, to assure the capability to cool the core is maintained; and (9) GDC-28, insofar as it requires the reactivity control systems be designed to assure that the effects of postulated reactivity accidents can neither result in damage to the RCPB greater than limited local yielding, nor disturb the core, its support structures, or other reactor vessel internals so as to significantly impair the capability to cool the core.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is Subject matter relating to the intent of that particular criteria.
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For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-27, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:
Draft GDC-6, Draft GDC-7, Draft GDC-8, Draft GDC-12, Draft GDC-13, Draft GDC-14, Draft GDC-15, Draft GDC-16, Draft GDC-27, Draft GDC-29, Draft GDC-30, Draft GDC-31, and Draft GDC-32. There is no Draft GDC directly associated with current GDC-27. Current GDC-12 listed in the Regulatory Evaluation above is applicable to PBAPS as described in UFSAR Section 7.17.4, "Description and Performance Analysis."
Nuclear design is described in PBAPS UFSAR Section 3.6, "Nuclear Design."
Technical Evaluation 2.8.2.1 Core Operation EPU increases the average power density proportional to the power increase and has some effects on operating flexibility, reactivity characteristics and energy requirements. The additional energy requirements for EPU are met by an increase in bundle enrichment, an increase in the reload fuel batch size, and/or changes in fuel loading pattern to maintain the desired plant operating cycle length.
Because PBAPS uses GNF2 fuel, the CLTR (Reference 1) is not applicable for fuel design dependent evaluations. Therefore, the core design and fuel thermal margin monitoring threshold are evaluated on a plant-specific basis in accordance with ELTRI (Reference 2).
2.8.2.1.1 Core Design PBAPS is currently licensed with an average bundle power of 4.60 MW/bundle. The average bundle power for EPU is 5.17 MW/bundle. The EPU average bundle power is comparable to the range of other operating BWRs.
The maximum allowable peak bundle power is not increased by power uprate. The additional energy requirements for power uprate are met by an increase in bundle enrichment, an increase in the reload fuel batch size, and/or changes in fuel loading pattern to maintain the desired plant operating cycle length. The power distribution in the core is changed to achieve increased core power, while limiting the minimum critical power ratio (MCPR), maximum linear heat generation rate (MLHGR), and maximum average planar linear heat generation rate (MAPLHGR) in any individual fuel bundle to be within limits as defined in the COLR.
The reactor core design power distribution represents a limiting thermal operating state at design conditions. It includes allowances for the combined effects on the fuel heat flux and temperature of the gross and local power density distributions, control rod pattern, and reactor power level adjustments during plant operation. NRC-approved core design methods were used to analyze core performance at the EPU RTP level. Detailed fuel cycle calculations of a representative equilibrium core design for this plant demonstrate the feasibility of EPU RTP operation while maintaining fuel design limits. Thermal-hydraulic design and operating limits ensure an acceptably low probability of boiling transition-induced fuel cladding failure occurring in the core, even for the most severe 27362
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) postulated operational transients. As needed, limits are also placed on fuel APLHGR and/or fuel rod LHGRs in order to meet both PCT limits for the limiting LOCA and fuel mechanical design bases.
The subsequent reload core designs for operation at the EPU RTP level will take into account the above limits to ensure acceptable differences between the licensing limits and their corresponding operating values. EPU may result in a small change in fuel burnup, the amount of fuel to be used, and isotopic concentrations of the radionuclides in the irradiated fuel relative to the original level of burnup. NRC-approved limits for burnup on the GNF2 fuel designs are not exceeded (Reference 83). For an example CLTP condition, the PBAPS Unit 2 EOC 18 peak bundle average discharge exposure is predicted to be (( )) MWd/MT. The EPU peak bundle average discharge exposure is predicted to be (( )) MWd/MT. GNF2 fuel is required to have Bundle Average Discharge Exposure less than (( )) MWd/MT. This is in compliance with the fuel-dependent limitations on discharge burnup. Also, due to the higher steady-state operating power associated with the EPU, the short-term curie content of the reactor fuel increases. The PBAPS Unit 2 Cycle 19 weighted average fresh bundle enrichment is predicted to be (( )). The PBAPS Unit 2 Cycle 18 weighted average fresh bundle enrichment was (( )). The EPU core weighted average fresh bundle enrichment is predicted to be (( )). There is no maximum licensed GNF2 bundle enrichment. The maximum licensed pellet enrichment is E[ )). This is in compliance with fuel-dependent and TSs limits on U-235 isotopic enrichment. The effects of higher power operation on radiation sources and DBA doses are discussed in Section 2.9.
Therefore, because the core design is established in accordance with approved methodology for each core reload, the assessment of this topic for PBAPS is acceptable.
2.8.2.1.2 Fuel Thermal Margin MonitoringThreshold The CLTR states that the percent power level above which fuel thermal margin monitoring is required may change with EPU. The original plant operating licenses set this monitoring threshold at a typical value of 25% of rated thermal power (RTP). E[
For EPU, as specified in the CLTR, the fuel thermal margin monitoring threshold is scaled down, if necessary, to ensure that monitoring is initiated ((
)) then the existing power threshold value must be lowered by a factor of 1.2/P 25.
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For PBAPS, the fuel thermal monitoring threshold is established at 23% of EPU RTP ((1.2 MWt /
bundle) * (764 bundles / 3951 MWt)) = 23%). A change in the fuel thermal monitoring threshold also requires a corresponding change to the TS reactor core safety limit for reduced pressure or low core flow.
The basis for not monitoring thermal limits below this threshold is the large margin to critical power as described in the TS Bases, Section 2.0 Safety Limits. Therefore, with these large margins, there are no transients that have limiting consequences when initiated from the 0 - 23 percent power range.
Therefore, the fuel thermal margin monitoring threshold meets all CLTR dispositions.
2.8.2.2 Thermal Limits Assessment The effect of EPU on the MCPR safety and operating limits and on the MAPLHGR and Linear Heat Generation Rate (LHGR) limits for PBAPS varies from no effect to a very slight effect.
Operating limits ensure that regulatory and/or safety limits are not exceeded for a range of postulated events (e.g., transients, LOCA). This section addresses the effects of EPU on thermal limits. A representative equilibrium core is used for the EPU evaluation. Cycle-specific core configurations, evaluated for each reload, confirm EPU capability, and establish or confirm cycle-specific limits, as is currently the practice.
For the topics to be addressed in Sections 2.8.2.2.2, 2.8.2.2.3, and 2.8.2.2.4, PBAPS is consistent with Sections 5.3.2 and 5.7.2 of ELTRI (Reference 2).
2.8.2.2.1 Safety Limit MCPR The Safety Limit MCPR (SLMCPR) can be affected slightly by EPU due to the flatter power distribution inherent in the increased power level, per the response to RAI 14 of ELTRI (Reference 2).
Experience has shown that the power uprate flatter power distribution results in an increase in the SLMCPR of less than 0.02 per Reference 2. A SLMCPR corresponding to the representative equilibrium core is used for the EPU evaluation. Cycle-specific SLMCPR calculations are evaluated for each reload, and establish or confirm cycle-specific limits, as is currently the practice.
The SLMCPR for SLO will normally be 0.01 or 0.02 greater than the SLMCPR for two-loop operation. The SLMCPR will be evaluated for the uprated reload core prior to EPU implementation. The calculated values will be reported in the SRLR for the EPU core.
Therefore, because the SLMCPR is established in accordance with approved methodology for each core reload, as required by ELTRI (Reference 2), the assessment of this topic for PBAPS is acceptable.
2.8.2.2.2 MCPR OperatingLimit Consistent with Sections 5.3.2 and 5.7.2.1 of ELTRI (Reference 2) and Section 3.4 of ELTR2 (Reference 4), the OLMCPR is calculated by adding the change in MCPR due to the limiting AOO event to the SLMCPR and is determined on a cycle-specific basis from the results of the reload 2-364
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) transient analysis. This approach does not change for EPU. The effect of EPU on the AOO events is addressed in Section 2.8.5. The required OLMCPR is not expected to significantly change
(<0.03) as shown in Table 3-1 of ELTRI (Reference 2) and Figure 5-3 of ELTR2 (Reference 4) and from experience with other uprated BWRs.
This small effect is due to the small changes in transient void and scram reactivity response and the flatter radial power distribution at EPU RTP. GEH BWR experience to date for power uprates up to 120% of OLTP confirms this assessment with changes in the operating limit MCPR of +0.018 to
-0.013. Limitations and Condition 9.19 of Reference 7 that requires a 0.01 OLMCPR adder is well within the experience base for the effect EPU has on the OLMCPR. For the reference, equilibrium core of GNF2 fuel, the OLMCPR for EPU RTP operation is discussed in Section 2.8.2.
Therefore, because the OLMCPR is established in accordance with approved methodology for each core reload, as required by ELTR1 (Reference 2), the assessment of this topic for PBAPS is acceptable.
2.8.2.2.3 MAPLHGR Limit Consistent with Section 5.7.2.2 of ELTRI (Reference 2), EPU operating conditions do not usually affect the MAPLHGR Operating Limit.
The MAPLHGR Operating Limit ensures that the plant does not exceed regulatory limits established in 10 CFR 50.46 or by the fuel design limits. The MAPLHGR Operating Limit is determined by analyzing the limiting LOCA for the plant. No significant change in operation is anticipated due to the EPU based on experience from other BWR uprates.
The ECCS performance is addressed in Section 2.8.5.6.2, and uses a reference equilibrium core of GNF2 fuel for EPU. Compared to CLTP, this evaluation shows that no change in the MAPLHGR limit is required for EPU for SLO or Dual Loop Operation (DLO) recirculation system operation.
Therefore, because the MAPLHGR Operating Limit is established in accordance with approved methodology for each core reload, as required by ELTRI (Reference 2), the assessment of this topic for PBAPS is acceptable.
2.8.2.2.4 LHGR OperatingLimit Consistent with Section 5.7.2.3 of ELTRI, EPU operating conditions do not usually affect the LHGR Operating Limit.
The LHGR Operating Limit is determined by the fuel rod thermal-mechanical design and is not affected by EPU.
Therefore, because the Maximum LHGR Operating Limit is established in accordance with approved methodology for each core reload, as required by Section 5.7 of ELTRI (Reference 2),
the assessment of this topic for PBAPS is acceptable.
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NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 2.8.2.2.5 Power and Flow Dependent Limits NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 9.1.2 of the CLTR addresses the effect of CPPU on Power and Flow Dependent Limits. The results of this evaluation are described below.
PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:
'K ~TopC'LTR P 132kPS ,
Dispo*osition Result Meets CLTR Power and Flow Dependent Limits R[ )) Disposition The CLTR states that power and flow dependent limits are not affected by EPU.
The operating MCPR and LHGR thermal limits are modified when the plant is operating at reduced core flow. This modification is primarily based upon an evaluation of the slow recirculation increase event. The current PBAPS analysis is based upon a conservative flow runup rod line that bounds operation to the rod line documented in Section 1.2. ((
Similarly, the thermal limits are modified by a power factor when the plant is operating at less than 100% power.
Er The power and flow dependent limits at PBAPS meet all CLTR dispositions.
2.8.2.3 Reactivity Characteristics As noted in Section 2.3 of the CLTR (Reference 1), the higher core energy requirements of power uprate may reduce the hot excess reactivity and reduce operating shutdown margin. Because PBAPS uses GNF2 fuel, the effect of power uprate on core reactivity is evaluated as described in Section 5.7.1 of ELTR1 (Reference 2).
The higher core energy requirements of power uprate may reduce the hot excess reactivity and; reduce operating shutdown margin during the cycle.
Based on experience with previous plant-specific power uprate submittals, the required hot excess reactivity and shutdown margin can be achieved for EPU through appropriate fuel and core design.
These parameters must meet the approved limits established in GESTAR II (Reference 5) on a cycle-specific basis.
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Therefore, because plant reactivity margins are established in accordance with approved methodology for each core reload, the assessment of these topics for PBAPS is acceptable.
2.8.2.4 Additional Topics from GEH Licensing Topical Report NEDC-33173P, "Applicability of GE Methods to Expanded Operating Domains" GEH Licensing Topical Report NEDC-33173P, "Applicability of GE Methods to Expanded Operating Domains," IMLTR was initially approved by the NRC in July 2009. GEH subsequently provided supplemental information to modify the original limitations and conditions stated in Section 9.0 of the NRC Safety Evaluation concerning additional topics to be addressed in EPU applications. Reference 7 is the current NRC-approved revision to the IMLTR. Appendix A provides a summary disposition of these limitations and conditions. Additional information for those topics related to Nuclear Design is provided below.
The Reference 7 topical report is applicable to fuel designs through GNF2. The PBAPS EPU core design includes the GNF2 fuel product line.
2.8.2.4.1 Steady-State 5 Percent Bypass Voiding Evaluations Limitation and Condition 9.17 of IMLTR (Reference 7) requires the bypass voiding to be evaluated on a cycle-specific basis to confirm that the void fraction remains below 5 percent. Limitation and Condition 9.17 is applicable to EPU conditions consistent with Reference 85.
The best-estimate means of determining 4-channel bypass void fraction is with TRACG. TRACG was applied in response to Methods RAI 14 (Reference 86). TRACG is capable of accurately modeling bypass heating and cross flow.
A conservative approach (ISCOR) was discussed in References 87 and 88. ISCOR conservatively calculates hot bypass channel voiding using its direct moderator-heating model and providing no credit for cross flow while applying additional conservatism with bounding 4-bundle peaking. The use of ISCOR is a more simplified and efficient process to implement compared to the use of TRACG and typically demonstrates margin to the 5% bypass void fraction requirement at the LPRM D Level.
For PBAPS reload core prior to EPU implementation, a calculation will be performed with the conservative ISCOR process at licensed EPU core power and minimum core flow (e.g. 120%
OLTP, 99% flow). The purpose of the calculation is to confirm that the bypass void fraction remains below 5 percent at all LPRM levels when operating at steady-state conditions within the licensed operating domain consistent with Reference 85.
If the resulting bypass void fraction is found to exceed the 5% requirement, it is acceptable to relax the conservative ISCOR input assumptions as long as the overall approach can be demonstrated to remain conservative relative to TRACG. It is also acceptable to perform a cycle-specific TRACG analysis with consideration of assumptions that will tend to maximize bypass void fraction (e.g.
bypass flow and 4-bundle peaking).
The highest calculated bypass voiding at any LPRM level will be provided with the plant-specific SRLR.
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For the representative equilibrium core used in EPU evaluation, the steady-state 5% bypass voiding evaluation is provided in the following table:
%oatd(oeRatated ilr Ho hn1"iiaf~n~
- n. ss' 100 100 0.000 100 99 0.000 2.8.2.4.2 Power-to-FlowRatio Limitation and Condition 9.3 of IMLTR (Reference 7) and Reference 85 require plant-specific EPU applications to confirm that the core thermal power to core flow ratio will not exceed 50 MWt/Mlbm/hr at the low flow point at rated power (e.g., EPU: 100% Power / 99.0% Flow) statepoint in the allowed operating domain. The core thermal power to total core flow ratio is reported in the following table:
fO/ ýed t d. pre~ 4 of Rated'Core<
ýiPower to-Flowý Ratio NNN;/X1m/r 100 99.0 38.9 The 120% OLTP power (i.e. the EPU power) is 3951 MWt and the minimum flow at this power is 99.0% rated flow, or 101.475 Mlbm/hr. Therefore, the resulting core thermal power to total core flow ratio at the 100% EPU RTP is 38.9 MWt/Mlbm/hr and does not exceed 50 MWt/Mlbm/hr.
This power-to-flow ratio analysis is consistent with the GEH letter discussing the implementation of methods limitations from NEDC-33173P (Reference 85). The power-flow map is independent of fuel design and does not change from cycle to cycle. Therefore, the power-to-flow ratio for PBAPS's future EPU cycles will also remain below 50 MWt/Mlbm/hr at this statepoint.
2.8.2.4.3 R-Factor Limitation and Condition 9.6 of IMLTR (Reference 7) requires the plant-specific R-factor calculation at a bundle level be consistent with lattice axial void conditions expected for the hot channel operating state.
The GNF2 bundle R-factors generated for this evaluation are consistent with GNF standard design procedures which use an axial void profile shape with 60% average in-channel voids. This is consistent with lattice axial void conditions expected for the hot channel operating state.
Figure 2.8-19 shows bundle average void history corresponding to hot channels with the limiting low critical power ratios (MCPRs) from each cycle exposure statepoint of the PBAPS EPU core.
The figure demonstrates that the generic R-factor profile, with an average void fraction of 0.60, is representative of the MCPR-limiting void conditions predicted by PANAC 11.
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NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 2.8.2.4.4 Plant-SpecificApplication Limitation and Condition 9.24 of IMLTR (Reference 7) requires plant-specific EPU applications to provide a prediction of key parameters for cycle exposures for operation at EPU. The following parameters: (1) Maximum Bundle Power; (2) Flow for Peak Bundle Power; (3) Exit Void Fraction for Peak Power Bundle; (4) Maximum Channel Exit Void Fraction; (5) Core Average Exit Void Fraction; (6) Peak LHGR; and (7) Peak Nodal Exposure are shown in Figures'2.8-1 through 2.8-6 and Table 2.8-1. The PBAPS data are plotted with the available EPU experiencebase as required by Limitation and Condition 9.24.
Quarter core maps with mirror symmetry are plotted in Figures 2.8-7 through 2.8-18 showing bundle power, bundle operating MCPR, and LHGR for beginning of cycle (BOC), middle of cycle (MOC), and EOC. Because the minimum margins to specific limits occur at exposures other than the traditional BOC, MOC, and EOC, the data are provided at these other exposures as applicable (Figures 2.8-16 through 2.8-18). Note that the bundle power in Figures 2.8-7 through 2.8-9 and Figure 2.8-16 is dimensionless. Obtain the bundle power in MWt by multiplying each number by 5.17. This number is equal to 3951/764, where 3951 MWt is the EPU RTP and 764 is the total number of bundles in the core.
2.8.2.4.5 Application of 10 Weight Percent Gd Limitation and Condition 9.13 of IMLTR (Reference 7) requires review and approval of 10 weight percent Gd to EPU applications.
For PBAPS, the maximum burnable poison concentration used is 8.0 weight percent Gd 20 3 ;
therefore, Limitation and Condition 9.13 is not applicable.
2.8.2.4.6 Mixed Core Method ]
Limitation and Condition 9.21 of IMLTR (Reference 7) requires that plants implementing EPU or MELLLA+ with mixed fuel vendor cores provide plant-specific justification for extension of GE's analytical methods or codes. The content of the plant-specific application will cover the topics addressed in this SE as well as subjects relevant to application of GE's methods to legacy fuel.
Alternatively, GE may supplement or revise LTR NEDC-33173P for mixed core application.
For PBAPS, there is no mixed core; therefore, the mixed core is not evaluated and Limitation and Condition 9.21 is not applicable.
2.8.2.4.7 Mixed Core Method 2 Limitation and Condition 9.22 of IMLTR (Reference 7) requires that for any plant-specific applications of TGBLA06 with fuel type characteristics not covered in the Interim Methods review, GE needs to provide assessment data similar to that provided for the GEH/GNF fuels. The Interim Methods review is applicable to all GEH/GNF lattices up to GNF2. Fuel lattice designs, other than GEH/GNF lattices up to GNF2, with the following characteristics are not covered by this review:
6 [
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The acceptability of the modified epithermal slowing down models in TGBLA06 has not been demonstrated for application to these or other geometries for expanded operating domains.
Significant changes in the Gd rod optical thickness will require an evaluation of the TGBLA06 radial flux and Gd depletion modeling before being applied. Increases in the lattice Gd loading that result in nodal reactivity biases beyond those previously established will require review before the GE methods may be applied.
For PBAPS, there is no mixed core; therefore, the mixed core is not evaluated and Limitation and Condition 9.22 is not applicable.
Conclusion The effects of the proposed EPU on the nuclear design of the fuel assemblies, control systems, and reactor core have been reviewed. Exelon concludes the review has adequately accounted for the effects of the proposed EPU on the nuclear design and has demonstrated the fuel design limits will not be exceeded during normal or anticipated operational transients, and the effects of postulated reactivity accidents will not cause significant damage to the RCPB or impair the capability to cool the core. Based on this evaluation and in coordination with the reviews of the fuel system design, thermal and hydraulic design, and transient and accident analyses, Exelon concludes the nuclear design of the fuel assemblies, control systems, and reactor core will continue to meet the applicable requirements of the current licensing basis. Therefore, Exelon finds the proposed EPU acceptable with respect to the nuclear design.
2.8.3 Thermal and Hydraulic Design Regulatory Evaluation Exelon reviewed the thermal and hydraulic design of the core and the RCS to confirm the design:
(1) has been accomplished using acceptable analytical methods; (2) is equivalent to or a justified extrapolation from proven designs; (3) provides acceptable margins of safety from conditions which would lead to fuel damage during normal reactor operation and AOOs; and (4) is not susceptible to thermal-hydraulic instability. The review also covered hydraulic loads on the core and RCS components during normal operation and DBA conditions and core thermal-hydraulic stability under normal operation and ATWS events. The regulatory acceptance criteria are based on: (1)
GDC-10, insofar as it requires the reactor core be designed with appropriate margin to assure that SAFDLs are not exceeded during any condition of normal operation, including the effects of AOOs; and (2) GDC-12, insofar as it requires the reactor core and associated coolant, control, and protection systems be designed to assure power oscillations, which can result in conditions exceeding SAFDLs, are not possible or can reliably and readily be detected and suppressed.
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Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-6 and Draft GDC-7. Current GDC-12 listed in the Regulatory Evaluation above is applicable to PBAPS as described in UFSAR Section 7.17.4, "Description and Performance Analysis."
The Thermal and Hydraulic Design is described in PBAPS UFSAR Sections 3.7, "Thermal and Hydraulic Design," 7.17, "Nuclear Stability Analysis," and 14, "Plant Safety Analysis."
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class 111, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 2.4.4 of the CLTR addresses the effect of CPPU on Plants with Option III.
Section 9.3.3 of the CLTR addresses the effect of CPPU on ATWS with Core Instability. The results of this evaluation are described below.
Section 3.2 of ELTR2 documents interim corrective actions (ICAs) and four stability Long-Term Solutions (LTSs). PBAPS has adopted the Option III solution. A generic evaluation was performed for the ICAs in Section 3.2.1 of Reference 4. This generic evaluation continues to be applicable for EPU. ICA stability boundaries are kept the same in terms of absolute core power and flow; power levels, reported as a percentage of rated power, are scaled based on the new uprated power. For the LTSs, evaluations are core reload dependent and are performed for each reload fuel cycle. The analyses of Option III are addressed below.
PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:
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Option III Meets CLTR (Oscillation Power Range Monitor (OPRM) Trip- Disposition Enabled Region and Trip Setpoint)
Option III Meets CLTR (Hot Channel Oscillation Magnitude) Disposition Meets CLTR ATWS with Core Instability Disposition 2.8.3.1 Option III 2.8.3.1.1 Option 111.- OPRM Trip EnabledRegion and Trip Setpoint The CLTR states that the Option III trip setpoint may be affected by EPU operating conditions.
The OPRM trip-enabled region will be rescaled with EPU.
Option III is a detect-and-suppress solution that combines closely spaced LPRM detectors into "cells" to effectively detect any mode of reactor instability. Plants implementing Option III must demonstrate that the Option III trip setpoint is adequate to provide SLMCPR protection for anticipated reactor instability. This evaluation is dependent upon the core and fuel design and is performed for each reload. ((
PBAPS has adopted the stability Option III LTS (Reference 89). Option III evaluations are core reload dependent and are performed for each fuel cycle reload. No changes to the PBAPS currently licensed Option III stability solution hardware and software algorithms are required for EPU operation. In the event that the OPRM system is declared inoperable, PBAPS will continue to operate under the BWR Owners Group (BWROG) Guidelines for Backup Stability Protection (BSP) as described in Reference 90. After the EPU is implemented, cycle-specific setpoints and BSP regions will continue to be determined and documented in the cycle-specific SRLR.
The Option III solution combines closely spaced LPRM detectors into "cells" to effectively detect either core-wide or regional modes of reactor instability. These cells are termed OPRM cells and are configured to provide local area coverage with multiple channels. The PBAPS Option III hardware combines the LPRM signals and evaluates the cell signals with instability detection algorithms. The Period Based Detection Algorithm (PBDA) is the only algorithm credited in the Option III licensing basis. Two defense-in-depth algorithms, referred to as the Amplitude Based Algorithm (ABA) and the Growth Rate Based Algorithm (GRA), offer a higher degree of assurance that fuel failure will not occur as a consequence of stability related oscillations.
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The OPRM trip is armed only when plant operation is within the OPRM trip-enabled region. The current PBAPS OPRM trip-enabled region is defined as the, region on the power/flow map with power Ž 29.5% of RTP and core flow < 60% of rated core flow. For EPU, the PBAPS OPRM trip-enabled region is rescaled to maintain the same absolute power/flow region boundaries. Because the rated core flow is not changed, the 60% core flow boundary is not rescaled. The 29.5% CLTP boundary is rescaled to the 26.2% EPU thermal power limit using the CLTP/EPU ratio.
The PBAPS OPRM trip-enabled region is shown in Figure 2.8-20. The BSP evaluation described in Section 2.8.3.1.3 shows the current Option III trip-enabled region is adequate after scaling to EPU. The adequacy of the OPRM trip-enabled region will be confirmed for each fuel reload.
Stability Option III provides SLMCPR protection by generating a reactor scram if a reactor instability that exceeds the specified OPRM trip setpoints is detected. The OPRM setpoints are determined per an NRC-approved methodology (References 89 and 91).
The Option III stability reload licensing basis calculates the limiting OLMCPRs required to protect the SLMCPR for both steady-state and transient stability events as specified in the Option III methodology. These OLMCPRs are calculated for a range of OPRM setpoints for EPU operation.
Selection of appropriate OPRM trip setpoints is then based upon the OLMCPRs required to provide adequate SLMCPR protection. This determination relies on the DIVOM curve to determine the OPRM Setpoints that protect the SLMCPR during an anticipated instability event. The DIVOM slope was developed based on a TRACG evaluation in accordance with the BWROG Regional Mode DIVOM Guideline (Reference 91).
Option III solution for PBAPS EPU (OPRM trip-enabled region and trip setpoints) meets all CLTR dispositions.
2.8.3.1.2 Option III - Hot Channel Oscillation Magnitude The CLTR states that the Option III trip setpoint may be affected by EPU operating conditions.
The OPRM trip-enabled region will be rescaled with EPU.
The Option III automatic scram is provided by the OPRM system. The generic analyses for the Option III hot channel oscillation magnitude and the OPRM hardware were designed to be independent of core power. ((
Although the Option III solution requires cycle-specific evaluations, a demonstration analysis was performed based on an equilibrium GNF2 core design. For this analysis, the calculated DIVOM slope was used. As shown in Table 2.8-2, with an assumed SLMCPR of 1.09, and an assumed rated OLMCPR of 1.36, an OPRM amplitude setpoint of 1.09 is the highest setpoint that may be used without stability setting the OLMCPR. The actual setpoint will be established at the time of each fuel reload based on cycle-specific core design.
Consistent with Limitation 9.18 per Reference 7, the OPRM system will incorporate a 5%
calibration error on the OPRM setpoints to address the bypass voiding uncertainty at low-flow 2-373
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) conditions. This calibration error has been included in the OPRM amplitude setpoints shown in Table 2.8-2. In addition, Table 2.8-2 includes a 0.01 adder so that the bypass voiding penalty incorporated in the transient-based OLMCPR is not used as a credit in stability analysis, consistent with Reference 85. Because PBAPS utilizes the Option III stability solution, the APRM calibration error required by Limitation 9.18 is not applicable (see Section 6.2 of Reference 7).
Option III (hot channel oscillation magnitude) meets all CLTR dispositions.
2.8.3.1.3 BSP Evaluation PBAPS implements BSP (Reference 90) as the stability licensing basis should the Option III OPRM system be declared inoperable. The BSP evolved from the stability ICAs (Reference 92), which restrict plant operation in the high power, low core flow region of the BWR power/flow operating map. The ICAs provide guidance that reduces the likelihood of an instability event by limiting the period of operation in regions of the power/flow map most susceptible to thermal hydraulic instability.
If the Option III OPRM system is declared inoperable, implementation of the associated BSP regions will constitute the stability licensing basis for PBAPS (Reference 90). The BSP regions consist of two regions (I-Scram and II-Controlled Entry), which are reduced from the three ICA regions (I-Scram, II-Exit and III-Controlled Entry) in Reference 92. The standard ICA region endpoints on the High Flow Control Line (HFCL) and on the Natural Circulation Line (NCL) define the base BSP region endpoints on the HFCL and on the NCL. The bounding plant-and-cycle-specific BSP region endpoints must enclose the corresponding base BSP region endpoints on the HFCL and the NCL. If a calculated BSP region endpoint is located inside the corresponding base BSP region endpoint, the corresponding base BSP region endpoint must replace it. That is, the selected points will result in the largest, or most conservative, region sizes. The proposed BSP Scram and Controlled Entry region boundaries may be constructed by connecting the corresponding bounding endpoints on the HFCL and the NCL using the Modified Shape Function (MSF)
(Reference 93).
The GNF2 equilibrium demonstration analysis was used to determine the ODYSY calculated BSP endpoints for nominal FW temperature and minimum FW temperature as shown in Tables 2.8-3 and 2.8-4, respectively. The limiting endpoints between the calculated and the Base BSP endpoints are used along with the MSF to construct the BSP regions for nominal FW temperature and minimum FW temperature as shown in Figures 2.8-21 and 2.8-22, respectively.
2.8.3.2 ATWS with Core Instability The CLTR states that the ATWS with core instability event occurs at natural circulation following a RPT. Therefore, it is initiated at approximately the same power level as a result of EPU operation because the MELLLA upper boundary is not increased. The core design necessary to achieve EPU operations may affect the susceptibility to coupled thermal-hydraulic/neutronic core oscillations at the natural circulation condition, but would not significantly affect the event progression. The analysis and results of the limiting ATWS events are presented in Section 2.8.5.7.
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Several factors affect the response of an ATWS instability event, including operating power and flow conditions and core design. The limiting ATWS core instability evaluation presented in References 94 and 95 was performed for an assumed plant initially operating at OLTP and the MELLLA minimum flow point. ((
EPU allows plants to increase their operating thermal power but does not allow an increase in control rod line. ((
1))
FWHOOS and FFWTR are operational flexibility options that allow continued operation with reduced FW temperature.
Initial operating conditions of FWHOOS and FFWTR do not significantly affect the ATWS instability response reported in References 94 and 95. The limiting ATWS evaluation assumes that all FW heating is lost during the event and the injected FW temperature approaches the lowest achievable MC hot well temperature. ((
)) Initial power is unchanged for both the FWHOOS and FFWTR conditions - the additional reactivity associated with the reduced FW temperature is typically offset with control rods, as needed. For both the FWHOOS and FFWTR conditions, an ATWS event analysis would be initiated from the same limiting power/flow statepoint assumed for the normal FW temperature case (statepoint 'D' in PUSAR Figure 1-1) and transition to essentially the same natural circulation statepoint (statepoint 'A' in PUSAR Figure 1-1) prior to the onset of power oscillations. ((
Operator actions will mitigate an ATWS instability event. The actions contained in References 94 and 95 bound the entire BWR fleet and are applicable to PBAPS. The conclusion of Reference 95 and the associated NRC SER that the analyzed operator actions effectively mitigate an ATWS instability event are applicable to the operating conditions expected for EPU at PBAPS. Therefore, the EPU effect on ATWS with core instability at PBAPS meets all CLTR dispositions.
Conclusion The effects of the proposed EPU on the thermal and hydraulic design of the core and the RCS have been reviewed. Exelon concludes the review has adequately accounted for the effects of the proposed EPU on the thermal and hydraulic design and demonstrated the design (1) has been accomplished using acceptable analytical methods; (2) is a proven design; (3) provides acceptable margins of safety from conditions that would lead to fuel damage during normal reactor operation and AOOs; and (4) is not susceptible to thermal-hydraulic instability. Exelon further concludes it 2-375
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) has adequately accounted for the effects of the proposed EPU on the hydraulic loads on the core and RCS components. Based on this, Exelon concludes the thermal and hydraulic design will continue to meet the requirements of the current licensing basis, following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to thermal and hydraulic design.
2.8.4 Emergency Systems 2.8.4.1 Functional Design of Control Rod Drive System Regulatory Evaluation Exelon reviewed the functional performance of the CRD system to confirm the system can effect a safe shutdown, respond within acceptable limits during AOOs, and prevent or mitigate the consequences of postulated accidents. The review also covered the CRD system cooling system to ensure that it will continue to meet its design requirements. The regulatory acceptance criteria are based on: (1) GDC-4, insofar as it requires SSCs important to safety be designed to accommodate, the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents; (2) GDC-23, insofar as it requires the protection system be designed to fail into a safe state; (3) GDC-25, insofar as it requires the protection system be designed to assure that SAFDLs are not exceeded for any single malfunction of the reactivity control systems; (4) GDC-26, insofar as it requires two independent reactivity control systems be provided, with both systems capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes; (5) GDC-27, insofar as it requires the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the ECCS, of reliably controlling reactivity changes under postulated accident conditions, with appropriate margin for stuck rods, to assure the capability to cool the core is maintained; (6) GDC-28, insofar as it requires the reactivity control systems be designed to assure that the effects of postulated reactivity accidents can neither result in damage to the RCPB greater than limited local yielding, nor disturb the core, its support structures, or other reactor vessel internals so as to significantly impair the capability to cool the core; (7) GDC-29, insofar as it requires the protection and reactivity control systems be designed to assure an extremely high probability of accomplishing their safety functions in event of AOOs; and (8) 10 CFR 50.62(c)(3),
insofar as it requires BWRs have an alternate rod injection (ARI) system diverse from the reactor trip system, and the ARI system have redundant scram air header exhaust valves.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design 2-376
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For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-27 and GDC-29, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-26, Draft GDC-27, Draft GDC-29, Draft GDC-30, Draft GDC-31, Draft GDC-32, Draft GDC-40, and Draft GDC-42. There is no Draft GDC directly associated with current GDC-27 or current GDC-29.
The design of the CRD system is described in PBAPS UFSAR Section 3.4, "Reactivity Control Mechanical Design."
In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10).
The license renewal evaluation associated with the CRD system is documented in NUREG-1769, Section 2.3.3.3. Management of aging effects on the CRD system is documented in NUREG-1769, Section 3.3.3.
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 2.5 of the CLTR addresses the effect of CPPU on the functional design of the CRD system. The results of this evaluation are described below.
As stated in Section 2.5 of the CLTR, the CRD system is used to control core reactivity by positioning neutron absorbing control rods within the reactor and to scram the reactor by rapidly inserting withdrawn control rods into the core.
PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:
Meets CLTR Scram Time Response Disposition 2-377
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Meets CLTR CRD Positioning Disposition CRD Cooling Meets CLTR Disposition Meets CLTR Disposti CRD Integrity Disposition 2.8.4.1.1 Scram Time Response The CLTR states that for pre-BWR/6 plants, the scram times are decreased by transient pressure response, and therefore the effect of EPU is bounded by current response.
At normal operating conditions, the CRD Hydraulic Control Unit accumulator supplies the initial scram pressure and, as the scram continues, the reactor becomes the primary source of pressure to complete the scram. Because the normal reactor dome pressure for EPU does not change, the scram time performance relative to current plant operation is the same. Therefore, pre-BWR/6 plants will retain their current TS scram requirement.
((I The CRD system control rod scram at PBAPS is confirmed to be consistent with the generic description provided in the CLTR pre-BWR/6 plants because PBAPS is a BWR/4 plant.
2.8.4.1.2 Control Rod Drive Positioningand Cooling As stated in Section 2.5 of the CLTR, the increase in reactor power at the EPU operating condition results in ((
)) from the CRD System to the CRDs during normal plant operation.
The EPU is evaluated on the basis of operation at the same dome pressure but higher core power and steam flow. ((
2.8.4.1.2.1 Control Rod Drive Positioning The CLTR states that, with reactor dome pressure unchanged, there is ((
)), and the automatic operation of the system flow control valve maintains the required drive water pressure. Therefore, the CRD 2-378
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For PBAPS Unit 2, plant operating data has confirmed that ((
For PBAPS Unit 3, plant operating data has confirmed that ((
1))
Therefore, the CRD system drive positioning meets all CLTR dispositions.
2.8.4.1.2.2 Control Rod Drive Cooling The CLTR states that, with reactor dome pressure unchanged, there is ((
)), and the automatic operation of the system flow control valve maintains the required cooling water flow rate. Therefore, the CRD cooling function is not affected. The CRD cooling function is an operational consideration, not a safety-related function, and is not affected by EPU operating conditions.
For PBAPS Unit 2, plant operating data has confirmed that ((
For PBAPS Unit 3, plant operating data has confirmed that ((
Therefore, the CRD system drive cooling meets all CLTR dispositions.
2.8.4.1.3 Control Rod Drive Integrity Assessment The CLTR states that (( on CRD integrity. The transient pressures due to uprated power may create higher pressure loadings.
The postulated abnormal operating condition for the CRD design assumes a failure of the CRD system pressure-regulating valve that applies the maximum pump discharge pressure to the CRD mechanism internal components. ((
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Er]
)) Other mechanical loadings are addressed in Sections 2.2.2 and 2.2.3 of this report.
Therefore, the CRD system integrity meets all CLTR dispositions.
Conclusion The effects of the proposed EPU on the functional design of the CRD system have been evaluated.
The ability of the CRD system to effect a safe shutdown, respond within acceptable limits, and prevent or mitigate the consequences of postulated accidents following the implementation of the proposed EPU has been demonstrated. In addition, it has been determined that sufficient cooling exists to ensure the system's design bases will continue to be met upon implementation of the proposed EPU. Based on this, Exelon concludes that the fuel system and associated analyses will continue to meet the requirements of the current licensing basis and 10 CFR 50.62(c)(3) following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to the functional design of the CRD system.
2.8.4.2 Overpressure Protection During Power Operation Regulatory Evaluation Relief and safety valves and the RPS provide overpressure protection for the RCPB during power operation. The effect of the proposed EPU on the performance of the relief and safety valves on the MS lines and the piping from these valves to the suppression pool has been reviewed. The regulatory acceptance criteria are based on: (1) GDC-15, insofar as it requires that the RCS and associated auxiliary, control, and protection systems be designed with sufficient margin to assure that the design conditions of the RCPB are not exceeded during any condition of normal operation, including AOOs; and (2) GDC-3 1, insofar as it requires that the RCPB be designed with sufficient margin to assure that it behaves in a non-brittle manner and that the probability of rapidly propagating fracture is minimized.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric 2-380
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Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-15, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:
Draft GDC-33, Draft GDC-34, and Draft GDC-35. There is no Draft GDC directly associated with current GDC- 15.
Overpressure protection during power operation is described in PBAPS UFSAR Section 4.4, "Nuclear System Pressure Relief System."
In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10).
The license renewal evaluation associated with overpressure protection is documented in NUREG-1769, Section 2.3.4.1. Management of aging effects on overpressure protection is documented in NUREG- 1769, Section 3.4.1.
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class 11, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 3.1 of the CLTR addresses the effect of CPPU on Nuclear System Pressure Relief/Overpressure Protection. The results of this evaluation are described below.
As stated in Section 3.1 of the CLTR, the system operating pressure does not change but the steam flow rate increases. The increased steam flow rate associated with uprated power may increase steam line vibration. The increased core steam generation also causes an increase in the pressurization during some transient events.
PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:
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Meets CLTR Overpressure capacity Met Disposition The CLTR states that the increased core steam generation causes an increase in the pressurization during some transient events.
The nuclear system pressure relief system prevents overpressurization of the nuclear system during AOOs, the plant ASME Upset overpressure protection event, and postulated ATWS events. The plant SRVs, along with other functions, provide this protection. An evaluation was performed in order to confirm the adequacy of the pressure relief system for EPU conditions.
The SRV discharge lines were designed and configured so that the discharge backpressure at the valve outlet is not greater than 40% of the inlet pressure. The valves were designed to achieve sonic (choked) flow conditions through the valve up to this backpressure ratio to provide flow independence to the discharge piping losses and backpressure. The backpressure to inlet pressure ratio is a function of discharge line geometry, which will not change with EPU. Therefore, SRV capacity will not be affected by the EPU discharge line backpressure.
The adequacy of the pressure relief system is also demonstrated by the overpressure protection evaluation performed for each reload core and by the ATWS evaluation performed for EPU (Section 2.8.5.7).
((
The design pressure of the reactor vessel and RCPB remains at 1250 psig. The acceptance limit for pressurization events is the ASME code allowable peak pressure of 1375 psig (110% of design value). The overpressure protection analysis description and analysis .method are provided in ELTRI. The MSIVF. event is conservatively analyzed assuming a failure of the valve position scram and event initiation at a reactor dome pressure of 1068 psia (which is the TS 3.4.10 Limiting Condition for Operation (LCO) value). The analysis also conservatively assumes one SRV out-of-service. Starting from 102% of EPU RTP, the calculated peak RPV pressure, located at the bottom of the vessel, is 1335 psig. The corresponding calculated maximum reactor dome pressure is 1314 psig. The peak calculated RPV pressure remains below the 1375 psig ASME limit, and the 2-382
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) maximum calculated dome pressure remains below the TS 1325 psig Safety Limit. Therefore, the results are acceptable and within the applicable limits. The results of the EPU overpressure protection analysis for the PBAPS MSIVF event are ((
)) The PBAPS response to the MSIVF event is provided as Figure 2.8-23.
The MSIVF event is performed using the NRC-approved code ODYN (see Table 1-1).
Therefore, the overpressure capacity meets all CLTR dispositions.
Conclusion The effects of the proposed EPU on the overpressure protection capability of the plant during power operation have been reviewed. The results of that review demonstrate that: (1) pressurization events and overpressure protection features adequately account for the effects of the proposed EPU; and (2) the plant will continue to have sufficient pressure relief capacity to ensure that pressure limits are not exceeded. Based on this, Exelon concludes that the overpressure protection features will continue to meet the current licensing basis following implementation of the proposed EPU.
Therefore, Exelon finds the proposed EPU acceptable with respect to overpressure protection during power operation.
2.8.4.3 Reactor Core Isolation Cooling System Regulatory Evaluation The RCIC system serves as a standby source of cooling water to provide a limited decay heat removal capability whenever the main FW system is isolated from the reactor vessel. In addition, the RCIC system may provide decay heat removal necessary for coping with a station blackout.
The primary water supply for the RCIC system comes from the CST, with a secondary supply from the suppression pool. The effect of the proposed EPU on the functional capability of the system has been reviewed. The regulatory acceptance criteria are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be protected against dynamic effects; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be demonstrated that sharing will not impair its ability to perform its safety function; (3) GDC-29, insofar as it requires that the protection and reactivity control systems be designed to assure an extremely high probability of accomplishing their safety functions in event of AOOs; (4) GDC-33, insofar as it requires that a system to provide reactor coolant makeup for protection against small breaks in the RCPB be provided so the fuel design limits are not exceeded; (5) GDC-34, insofar as it requires that a RHR system be provided to transfer fission product decay heat and other residual heat from the reactor core at a rate such that SAFDLs and the design conditions of the RCPB are not exceeded; (6) GDC-54, insofar as it requires that piping systems penetrating containment be designed with the capability to periodically test the operability of the isolation valves to determine if valve leakage is within acceptable limits; and (7) 10 CFR 50.63, insofar as it requires that the plant withstand and recover from an SBO of a specified duration.
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Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-29 and current GDC-34, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-4, Draft GDC-37, Draft GDC-39, Draft GDC-40, Draft GDC-42, Draft GDC-51, Draft GDC-54, Draft GDC-55, Draft GDC-56, and Draft GDC-57. There is no Draft GDC directly applicable to current GDC-29 or current GDC-34.
The RCIC system is described in PBAPS UFSAR Section 4.7, "Reactor Core Isolation Cooling System."
In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10).
The license renewal evaluation associated with the RCIC system is documented in NUREG- 1769, Section 2.3.2.4. Management of aging effects on the RCIC system is documented in NUREG-1769, Section 3.2.4.
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 3.9 of the CLTR addresses the effect of CPPU on the RCIC system. The results of this evaluation are described below.
The RCIC system evaluation for EPU at PBAPS addressed the following topics:
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" System performance and hardware
" Net positive suction head
" Adequate core cooling for limiting LOFW events (Addressed in Section 2.8.5.2.3)
" Inventory makeup - Operational Level 1 avoidance (Addressed in Section 2.8.5.2.3)
PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:
Meets CLTR System performance and hardware (RCIC) 1[ Disposition Meets CLTR Net positive suction head (RCIC) ]Disposition 2.8.4.3.1 System Performanceand Hardware The CLTR states that there is no effect on RCIC system performance and hardware due to EPU.
The RCIC system is required to maintain sufficient water inventory in the reactor to permit adequate core cooling following a reactor vessel isolation event accompanied by loss of flow from the FW system. The system design injection rate must be sufficient for compliance with the system limiting criteria to maintain the reactor water level above TAF at EPU conditions. The RCIC system is designed to pump water into the reactor vessel over a wide range of operating pressures.
The results of the PBAPS plant-specific evaluation indicate adequate water level margin above TAF at EPU conditions. Thus, the RCIC injection rate is adequate to meet this design basis event.
An operational requirement is that the RCIC system can restore the reactor water level while avoiding ADS timer initiation and MSIV closure activation functions associated with the low-low-low reactor water level setpoint (Level 1). This requirement is intended to avoid unnecessary initiations of safety systems. The results of the PBAPS plant-specific evaluation indicates that the RCIC system is capable of maintaining the water level outside the shroud above nominal Level 1 setpoint through a limiting LOFW event at EPU conditions. Thus, the RCIC injection rate is adequate to meet the requirements for inventory makeup. The reactor system response to a LOFW transient with RCIC is discussed in Section 2.8.5.2.3.
For EPU, there is no change to the normal reactor operating dome pressure (1050 psia for both CLTP and EPU conditions) and the SRV setpoints remain the same. There is no change to the maximum specified reactor pressure for RCIC system operation, ((
)) The PBAPS RCIC pump is adequate to support EPU. ((
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The system performance and hardware for RCIC meets all CLTR dispositions.
2.8.4.3.2 Net Positive Suction Head The CLTR states that there is no effect on RCIC NPSH due to EPU.
The PBAPS minimum NPSH available for the PBAPS RCIC pump does not change because there are no physical changes to the pump suction configuration, and no changes to the system flow rate or minimum atmospheric pressure in the suppression pool or CST. EPU does not affect the capability to transfer the RCIC pump suction on high suppression pool level or low CST level from its normal alignment, the CST, to the suppression pool, and does not change the existing requirements for the transfer. Therefore, the specified operational temperature limit for the process water does not change with EPU. Because PBAPS is not changing the RCIC pump or its operating parameters, the required NPSH does not change. The effect of EPU on the operation of the RCIC system during SBO events is discussed in Section 2.3.5. The effect of EPU on the operation of the RCIC system during a 10 CFR 50 Appendix R Fire Event is discussed in Section 2.5.1.4.
Maximum pump speed and maximum pump flow (600 gpm) are unchanged for EPU and there is no change in maximum normal operating dome pressure (1050 psia at CLTP and EPU conditions).
The SRV setpoints remain the same. No RCIC system power dependent functions or operating requirements (flows, pressure, temperature, and NPSH) are added or changed from the original design or licensing bases.
The RCIC NPSH at PBAPS meets all CLTR dispositions.
Conclusion The effects of the proposed EPU on the ability of the RCIC system to provide decay heat removal following an isolation of main FW event and an SBO event have been analyzed. Accounting for the effects of the proposed EPU on these events, it has been demonstrated that the RCIC system will continue to provide sufficient decay heat removal and makeup for these events following implementation of the proposed EPU. Based on this, Exelon concludes that the RCIC system will continue to meet the requirements of 10 CFR 50.63 and the current licensing basis following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to the RCIC system.
2.8.4.4 Residual Heat Removal System Regulatory Evaluation The RHR system is used to cool down the RCS following shutdown. The RHR system is a low pressure system, which takes over the shutdown cooling function when the RCS temperature is reduced. The effect of the proposed EPU on the functional capability of the RHR system to cool the RCS following shutdown and provide decay heat removal has been reviewed. The regulatory acceptance criteria are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be protected against dynamic effects; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly 2-386
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) impair their ability to perform their safety functions; and (3) GDC-34, which specifies requirements for an RHR system.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-34, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:
Draft GDC-4, Draft GDC-40, and Draft GDC-42. There is no Draft GDC directly associated with current GDC-34.
The RHR system is described in PBAPS UFSAR Section 4.8, "Residual Heat Removal System."
In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10).
The license renewal evaluation associated with the RHR system is documented in NUREG-1769, Section 2.3.2.5. Management of aging effects on the RHR system is documented in NUREG-1769, Section 3.2.5.
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 3.10 of the CLTR addresses the effect of CPPU on the RHR system. The results of this evaluation are described below.
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As explicitly stated in Section 3.10 of the CLTR, the RHR system is designed to restore and maintain the reactor coolant inventory following a LOCA and remove reactor decay heat following reactor shutdown for normal, transient, and accident conditions. The EPU effect on the RHR system is a result of the higher decay heat in the core corresponding to the uprated power and the increased amount of reactor heat discharged into the containment during a LOCA.
For PBAPS, the RHR system is designed to operate in the LPCI mode, Shutdown Cooling mode, SPC mode, CSC mode, and FPC Assist (Supplemental SFP Cooling). PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:
Addressed in Section LPCI mode 2.8.5.6.2 Suppression Pool and Containment Spray Addressed in Section Cooling Modes 2.6.5 Shutdown Cooling Mode Meets CLTR Disposition Addressed in Section Fuel Pool Cooling Assist ]2.5.3.1 2.8.4.4.1 LPCI Mode The CLTR states that there is no change in the reactor pressures at which the LPCI mode of RHR is required. The LPCI mode, as it relates to the LOCA response, is discussed in Section 2.8.5.6.2, which concludes that 10 CFR 50.46 limits are met at EPU conditions. The LPCI system at PBAPS meets all CLTR dispositions.
2.8.4.4.2 Suppression Pool and ContainmentSpray Cooling The CLTR states that the suppression pool temperature increases as a result of the higher decay heat associated with EPU. The SPC mode is manually initiated following isolation transients and a postulated LOCA to maintain the containment pressure and suppression pool temperature within design limits. The CSC mode reduces DW pressure, DW temperature, and suppression chamber pressure following an accident. The adequacy of these operating modes is demonstrated by the containment analysis (Section 2.6.5).
With additional SP heat removal associated with modifications to eliminate credit for CAP, the peak SP temperature at EPU will be lower than that in the current analysis of record. Suppression pool temperatures for evaluated design basis EPU events remain within the design limits.
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Therefore, the suppression pool temperature during a postulated LOCA at EPU conditions does not change the capabilities of RHR system equipment to perform the LPCI, SPC and CSC functions.
Containment pressures for these EPU events increased slightly above the CLTP analyzed pressures, but remained below the existing peak containment internal pressure limit. The slight increase in the predicted containment pressure during a postulated LOCA at EPU conditions (See Table 2.6-1) remains within the equipment design parameters and thus does not adversely affect the hardware capabilities of RHR system equipment to perform the LPCI, SPC and CSC functions. Therefore, the Suppression Pool and CSC modes meet all CLTR dispositions.
2.8.4.4.3 Shutdown Cooling Mode The CLTR states that a longer time is required for reactor cool down as a result of the higher decay heat associated with EPU. The shutdown cooling mode is designed to remove the sensible and decay heat from the reactor primary system during a normal reactor shutdown. This non-safety operational mode allows the reactor to be cooled down within a certain time objective, so that the shutdown cooling mode of operation will not become critical path during refueling operations.
EPU increases the reactor decay heat, which requires a longer time for cooling down the reactor.
The shutdown cooling analysis for the EPU determined that the time needed for cooling the reactor to 125°F during normal reactor shutdown, with two shutdown cooling subsystems in service, is increased to 34.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at EPU conditions. The current UFSAR Section 4.8.6.1 time criterion for shutdown cooling is "approximately 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />." The increase in the normal reactor shutdown time for EPU indicates that a normal reactor shutdown may take longer, which could affect outage schedules. This impact may have an effect on plant availability, but has no effect on plant safety or the design operating margins and therefore, requires no change to the RHR system. Therefore, the shutdown cooling mode meets all CLTR dispositions.
2.8.4.4.4 Fuel Pool Cooling Assist The CLTR states that the SFP heat load increases due to the decay heat generation as a result of the EPU. The FPC Assist (Supplemental SFP Cooling) mode, using existing RHR system heat removal capacity, provides supplemental fuel pool cooling capability in the event that the fuel pool heat load exceeds the heat removal capability of the Fuel Pool Cooling and Cleanup system (FPCCS). The adequacy of fuel pool cooling, including use of the SFP Cooling mode, is discussed in Section 2.5.3.1, which concludes that EPU does not affect this system. Therefore, the FPC Assist mode meets all CLTR dispositions.
Conclusion The effects of the proposed EPU on the RHR system have been reviewed. The results of that review demonstrate that the RHR system will maintain its ability to cool the RCS following shutdown and provide decay heat removal after the proposed EPU. Based on this, Exelon concludes that the RHR system will continue to meet the requirements of the current licensing basis following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to the RHR system.
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NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION -,CLASS I (PUBLIC) 2.8.4.5 Standby Liquid Control System Regulatory Evaluation The SLCS provides backup capability for reactivity control independent of the control rod system.
The SLCS functions by injecting a boron solution into the reactor to effect shutdown. The effect of the proposed EPU on the functional capability of the system to deliver the required amount of boron solution into the reactor has been reviewed. The regulatory acceptance criteria are based on: (1)
GDC-26, insofar as it requires that two independent reactivity control systems of different design principles be provided, and that one of the systems be capable of holding the reactor subcritical in the cold condition; (2) GDC-27, insofar as it requires that the reactivity control systems have a combined capability, in conjunction with poison addition by the ECCS, to reliably control reactivity changes under postulated accident conditions; and (3) 10 CFR50.62(c)(4), insofar as itrequires that the SLCS be capable of reliably injecting a borated water solution into the RPV at a boron concentration, boron enrichment, and flow rate that provides a set level of reactivity control., As stated in PBAPS TS Bases B 3.1.7, the SLCS is manually initiated from the main CR, as directed by the EOPs, if the operator believes the reactor cannot be shut down, or kept shut down, with the control rods.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups. of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a stateme nt of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-27, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:
Draft GDC-27, Draft GDC-29, and Draft GDC-30. There is no Draft GDC directly associated with current GDC-27.
The SLCS is described in PBAPS UFSAR Section 3.8, "'Standby Liquid Control System."
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In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10).
The license renewal evaluation associated with the SLCS is documented in NUREG- 1769, Section 2.3.3.4. Management of aging effects on the SLCS is documented in NUREG-1769, Section 3.3.4.
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.5 of the CLTR addresses the effect of CPPU on SLCS. The results of this evaluation are described below.
The SLCS is designed to shut down the reactor from rated power conditions to cold shutdown in the postulated situation that some or all of the control rods cannot be inserted. This manually operated system pumps a highly enriched sodium pentaborate solution into the vessel, to provide neutron absorption and achieve a subcritical reactor condition. SLCS is designed to inject over a wide range of reactor operating pressures.
PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:
Meets CLTR Core Shutdown Margin (( Disposition Meets CLTR System Performance and Hardware Disposition Suppression Pool Temperature Following Meets CLTR Limiting ATWS Event Disposition 2.8.4.5.1 Core Shutdown Margin Section 6.5 of the CLTR states that the ability of the SLCS boron solution to achieve and maintain safe shutdown is not a direct function of core thermal power, and therefore, is not affected by EPU.
SLCS shutdown capability (in terms of the required reactor boron concentration) is reevaluated for each fuel reload. The EPU evaluations assumed GNF2 fuel. The boron shutdown concentration of 660 ppm does not change for EPU. No changes are necessary to the solution volume /
concentration or to the boron-10 enrichment for EPU to achieve the required reactor boron concentration for cold shutdown conditions. However, the boron-10 enrichment has been increased 2-391
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Therefore, the SLCS shutdown margin capability meets all CLTR dispositions.
2.8.4.5.2 System Performanceand Hardware As stated in Section 6.5 of the CLTR, the effect of EPU on system performance and hardware is increased heat load and potential increase in transient reactor pressure. The SLCS is designed for injection at a maximum reactor pressure equal to the upper AV for the lowest group of SRVs operating in the safety relief mode. At PBAPS, the nominal reactor dome pressure and the SRV setpoints are unchanged for EPU. Consequently, the capability of the PBAPS SLCS to provide its backup shutdown function is not affected by EPU. The SLCS is not dependent upon any other SRV operating modes.
Based on the results of the PBAPS EPU ATWS analysis, the maximum reactor lower plenum pressure following the limiting ATWS event reaches 1190.3 psig (1205 psia) during the time the SLCS is analyzed to be in operation. Consequently, there is a corresponding increase in the maximum pump discharge pressure to 1265 psig (75 psi pressure drop from the SLCS pump to the reactor vessel) and a decrease in the operating pressure margin for the pump discharge relief valves.
Consideration was given to system flow, head losses for full injection, and cyclic pressure pulsations due to the positive displacement pump operation in determining the pressure margin to the opening set point for the pump discharge relief valves. The relief valve setpoint margin is 185 psi. This margin is based on a SLCS pump relief valve nominal setpoint of 1450 psig. The pump discharge relief valves are periodically tested to confirm the setpoint. The operation of the pump discharge system was analyzed to confirm that the loss of flow through an open relief valve would not compromise the required boron injection function (due to an early SLCS initiation). The evaluation compared the open/close setpoint of the pump discharge relief valves with the calculated maximum SLCS pump discharge pressure expected during the most limiting ATWS transient. It was confirmed that the SLCS relief valves would close prior to analyzed initiation if system initiation were to occur prior to the reactor pressure recovering from the initial transient peak.
Therefore, the current SLCS process parameters associated with the minimum boron injection rate are not changed.
The SLCS ATWS performance is evaluated in Section 2.8.5.7 for a representative core design for EPU. The evaluation confirmed acceptable results and demonstrates that EPU has no adverse effect on the ability of the SLCS to mitigate an ATWS. Therefore, PBAPS SLCS performance and hardware meet all CLTR dispositions.
2.8.4.5.3 Suppression Pool Temperature FollowingATWS Event As stated in Section 6.5 of the CLTR, changes in the fuel design for EPU may require modifications to the SLCS as a result of the increase in the suppression pool temperature for the limiting ATWS event.
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The boron injection rate, specifically the isotropic enrichment of Boron-10, requirement for maintaining the peak suppression pool water temperature limits, following the limiting ATWS event with SLCS injection, is increased for EPU and therefore suppression pool temperature is decreased. The suppression pool temperature following an ATWS event at PBAPS was performed on a (( )) consistent with the CLTR disposition.
Conclusion The effects of the proposed EPU on the SLCS have been reviewed and it was found the SLCS adequately accounts for the EPU. It was demonstrated that the system will continue to provide the function of reactivity control independent of the CRD system following implementation of the proposed EPU. Based on this, Exelon concludes that the SLCS will continue to meet the requirements of 10 CFR 50.62(c)(4) and the current licensing basis, following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to the SLCS.
2.8.4.6 Reactor Recirculation System Performance NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 3.6 of the CLTR addresses the effect of CPPU on the RRS performance. The results of this evaluation are described below. RRS performance is not specifically addressed in NRC "Review Standard for Extended Power Uprates," RS-001.
The EPU power condition is accomplished by operating along extensions of current rod lines on the power/flow map with no increase in the maximum core flow. The core reload analyses are performed with the most conservative allowable core flow. The evaluation of the RRS performance at EPU power determines that adequate core flow can be maintained.
The cavitation protection interlock remains the same in terms of absolute flow rates. This interlock is based on subcooling in the external recirculation loop and thus is a function of absolute FW flow rate and FW temperature at less than full thermal power operating conditions. Therefore, the interlock is not changed by EPU.
PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:
Meets CLTR Net Positive Suction Head (( Disposition Flow Mismatch Meets CLTR Disposition Meets CLTR Single-Loop Operation ]Disposition 2-393
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 2.8.4.6.1 Net Positive Suction Head The CLTR states that increased voids in the core during normal uprated power operation requires a slight increase in the recirculation drive flow to achieve the same core flow.
The CLTR shows that recirculation pump NPSH at full EPU power does not significantly increase the NPSH required or significantly reduce the NPSH margin. There is no change in maximum core flow. The maximum design core flow (110% rated core flow) at CLTP and EPU is 112.8 Mlb/hr.
Based on past uprate analyses, the NPSH required at full power does not significantly increase or reduce the NPSH margin because the required increase in recirculation flow is small.
Therefore, the effects of EPU on NPSH meets all CLTR dispositions.
2.8.4.6.2 Flow Mismatch The PBAPS recirculation loop jet pump flow mismatch TS limits do not change because these limits are based on rated core flow, which is not affected by EPU, and the flow mismatch limits are not affected because a detailed ECCS evaluationwas not required for PBAPS at EPU conditions by the EPU LOCA evaluation.
Therefore, the effect of EPU on flow mismatch meet all CLTR dispositions.
2.8.4.6.3 Single-loop Operation The CLTR states that increased voids in the core during normal uprated power operation requires a slight increase in the recirculation drive flow to achieve the same core flow.
SLO is limited to off-rated conditions and is not affected by EPU. SLO operation at PBAPS is restricted to a reactor power of 2,701 MWt and a flow of 57.4 Mlb/hr. The absolute power limit for SLO stays the same, requiring a proportional reduction in the percent of rated power at the uprate power level.
Therefore, the effects of EPU on SLO meet all CLTR dispositions.
2.8.5 Accident and Transient Analyses 2.8.5.1 Decrease in Feedwater Temperature, Increase in Feedwater Flow, Increase in Steam Flow, and Inadvertent Opening of a Main Steam Relief or Safety Valve Regulatory Evaluation Excessive heat removal causes a decrease in moderator temperature, which increases core reactivity and can lead to a power level increase and a decrease in shutdown margin. Any unplanned power level increase may result in fuel damage or excessive reactor system pressure. Reactor protection and safety systems are actuated to mitigate the transient. A review has been performed of the effects of the proposed EPU on: (1) postulated initial core and reactor conditions; (2) methods of thermal and hydraulic analyses; (3) the sequence of events; (4) assumed reactions of reactor system components; (5) functional and operational characteristics of the RPS; (6) operator actions; and (7) the results of the transient analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are 2-394
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) not exceeded during normal operations including AOOs; (2) GDC-15, insofar as it requires that the RCS and its associated auxiliary systems be designed with margin sufficient to ensure that the design condition of the RCPB are not exceeded during any condition of normal operation; (3)
GDC-20, insofar as it requires that the RPS be designed to initiate automatically the operation of appropriate systems, including the reactivity control systems, to ensure that SAFDLs are not exceeded during any condition of normal operation, including AOOs; and (4) GDC-26, insofar as it requires that a reactivity control system be provided, and be capable of reliably controlling the rate of reactivity changes to ensure that under conditions of normal operation, including AGOs, SAFDLs are not exceeded.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-15, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:
Draft GDC-6, Draft GDC-14, Draft GDC-15, Draft GDC-27, Draft GDC-29, and Draft GDC-30.
There is no Draft GDC directly associated with current GDC-15.
The analysis of a loss of FW heating transient is described in PBAPS UFSAR Section 14.5.2.3, "Loss of Feedwater Heating." The analysis of a FW controller failure with maximum demand is described in PBAPS UFSAR Section 14.5.2.2, "Feedwater Controller Failure - Maximum Demand." The analysis of an inadvertent opening of a relief valve, causing steam flow increase, is described in PBAPS UFSAR Section 14.5.4.2, "Inadvertent Opening of a Relief Valve or Safety Valve." The analysis of increase in steam flow is described in PBAPS UFSAR Section 14.5.4.1, "Pressure Regulator Failure."
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Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel products.
Because PBAPS is based on GNF2, this section will be based on ELTR1. NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTRI) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Although ELTRI is the licensing basis for the PBAPS AOO events, for the events discussed in this section, the sensitivity of the CPR changes provided as part of the transient analysis results in Table 2.8-12 is consistent with the sensitivity that established the scope approved in the CLTR.
((I
)) The following is a summary of the evaluation provided for the excessive heat removal events:
Consistent with the CLTR, the Decrease in FW Temperature limiting events (Loss of FW Heater (LFWH) with manual flow control) and the Increase in FW Flow limiting event (FW 'Controller Failure Maximum Demand (FWCF)) are confirmed to be within the PBAPS reload evaluation scope. The LFWH is performed with the NRC-approved methods described in GESTAR II (Reference 5). The computer code used to evaluate the LFWH is PANACEA. The transient evaluation initial conditions are provided in Table 2.8-11, and the results of the EPU evaluations are reported in Table 2.8-12.
Independent of EPU, PBAPS will be replacing the recirculation system pump motor power supplies from M/G set power supplies to ASDs (see PBAPS EPU LAR Attachment 9, Section 2, for additional information). The FWCF ((
The Increase in Steam Flow event and the Inadvertent Opening of a SRV event are not listed in ((
)) to be analyzed for EPU.
The Increase in Steam Flow event (((
))) is ((
)). This event results ((
The Inadvertent Opening of a SRV event is ((
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Conclusion The analyses of the excess heat removal events described above have been reviewed to ensure they have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. Based on these analyses, it has been demonstrated that the reactor protection and safety systems will continue to ensure that the SAFDLs and the RCPB pressure limits will not be exceeded as a result of these events. Based on this, Exelon concludes that the plant will continue to meet the requirements of the current licensing basis, following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to the events stated.
2.8.5.2 Decrease in Heat Removal by the Secondary System 2.8.5.2.1 Loss of ExternalLoad; Turbine Trip; Loss of Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam PressureRegulator Failure(Closed)
Regulatory Evaluation A number of initiating events may result in unplanned decreases in heat removal by the secondary system. These events result in a sudden reduction in steam flow and, consequently, result in pressurization events. Reactor protection and safety systems are actuated to mitigate the transient.
A review has been performed of the effects of the proposed EPU on the sequence of events, the analytical models used for analyses, the values of parameters used in the analytical models, and the results of the transient analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations, including AOOs; (2) GDC-15, insofar as it requires that the RCS and its associated auxiliary systems be designed with margin sufficient to ensure that the design conditions of the RCPB are not exceeded during any condition of normal operation; and (3)
GDC-26, insofar as it requires that a reactivity control system be provided, and be capable of reliably controlling the rate of reactivity changes to ensure that under conditions of normal operation, including AOOs, SAFDLs are not exceeded.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design 2-397
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-15, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:
Draft GDC-6, Draft GDC-27, Draft GDC-29, and Draft GDC-30. There is no Draft GDC directly associated with current GDC- 15.
The analysis of a generator load rejection is described in PBAPS UFSAR Section 14.5.1.1, "Electrical Load Rejection (Turbine Control Valve Fast Closure) with Bypass Failure." The analysis of a turbine trip without bypass is described in PBAPS UFSAR Section 14.5.1.2, "Turbine Trip." The analysis of an MSIV closure event is described in PBAPS UFSAR Section 14.5.1.3, "Main Steam Line Isolation Valve Closure." The analysis of a loss of condenser vacuum (LOCV) event is described in PBAPS UFSAR Section 14.5.1, "Events Resulting in a Nuclear System Pressure Increase." The analysis of a steam pressure regulator failure (closed) is described in PBAPS UFSAR Section 14.5.4.1, "Pressure Regulator Failure."
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel products.
Because PBAPS is based on GNF2, this section will be based on ELTRI. NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTRI) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Although ELTRI is the licensing basis for the PBAPS AOO events, for the events discussed in this section, the sensitivity of the CPR changes provided as part of the transient analysis results in Table 2.8-12 is consistent with the sensitivity that established the scope approved in the CLTR.
((
)) The following is a summary of the evaluation provided for the decreased heat removal events:
Consistent with the CLTR, the Loss of External Load limiting event (Generator Load Rejection with Steam Bypass Failure (LRNBP)) and the Turbine Trip limiting event (Trip with Steam Bypass Failure (TTNBP)) are confirmed to be within the PBAPS reload evaluation scope. The transient evaluation initial conditions are provided in Table 2.8-11, and the results of the EPU evaluations are reported in Table 2.8-12.
The LRNBP and TTNBP events employ a ((
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The Closure of Main Steam Isolation Valves with Direct Scram (MSIVD) event is within the PBAPS reload evaluation scope. The transient evaluation initial conditions are provided in Table 2.8-11, and the results of the EPU evaluations are reported in Table 2.8-12.
The Pressure Regulator Failure Closed (Pressure Regulator Failure Downscale (PRFDS)) event is
(( )) (See Table 3, Item 4 of the response to NRC RAI Set 9 Number 14 RSXB contained in the CLTR.).
The LOCV event for GEH BWRs is also ((
MSIVF for PBAPS is evaluated in Section 2.8.4.2.
Consistent with Limitations and Conditions 9.9 and 9.11 of Reference 7, acceptable fuel rod thermal-mechanical performance for both U0 2 and GdO 2 fuel rods was demonstrated. Results for all AOO pressurization transient events analyzed, including equipment out-of-service, showed at least 10 percent margin to the fuel centerline melt and at least 10 percent margin to the one percent cladding circumferential plastic strain acceptance criteria. The minimum calculated margin to the fuel centerline melt criterion was 21%. The minimum calculated margin to the cladding strain criterion was 21%. Fuel rod thermal-mechanical performance will be evaluated as part of the RLA performed for the cycle-specific core. Documentation of acceptable fuel rod thermal-mechanical response will be included in the SRLR or COLR consistent with Limitation and Condition 9.10 of Reference 7.
Conclusion The analyses of the decrease in heat removal (i.e., an increase in reactor pressure) events described above have been reviewed to ensure they have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The results of those analyses demonstrate that the reactor protection and safety systems will continue to ensure that the SAFDLs and the RCPB pressure limits will not be exceeded as a result of these events.
Based on this, Exelon concludes that the plant will continue to meet the requirements of the current 2-399
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2.8.5.2.2 Loss of Non-Emergency AC Power to the Station Auxiliaries Regulatory Evaluation The loss of non-emergency AC power is assumed to result in the loss of all power to the station auxiliaries and the simultaneous tripping of all reactor coolant circulation pumps. This causes a flow coastdown as well as a decrease in heat removal by the secondary system, a turbine trip, an increase in pressure and temperature of the coolant, and a reactor trip. Reactor protection and safety systems are actuated to mitigate the transient. A review has been performed of the effects of the proposed EPU on: (1) the sequence of events; (2) the analytical model used for analyses; (3) the values of parameters used in the analytical model; and (4) the results of the transient analyses. The regulatory acceptance criteria are.based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations, including AOOs; (2) GDC-15, insofar as it requires that the RCS and its associated auxiliary systems be designed with margin sufficient to ensure that the design conditions of the RCPB are not exceeded during any condition of normal operation; and (3) GDC-26, insofar as it requires that a reactivity control system be provided, and be capable of reliably controlling the rate of reactivity changes to ensure that under conditions of normal operation, including AOOs, SAFDLs are not exceeded.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-15, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:
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Draft GDC-6, Draft GDC-27, Draft GDC-29, and Draft GDC-30. There is no Draft GDC directly associated with current GDC- 15.
The analysis for loss of non-emergency AC power to the Recirculation Flow Control System is described in PBAPS UFSAR Section 7.9.5, "Safety Evaluation." The analysis for loss of non-emergency AC power to station auxiliaries is described in PBAPS UFSAR Section 14.5.4.4, "Loss of Auxiliary Power."
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. However, the CLTR is limited to application of GEl4 and earlier fuel products.
Because PBAPS is based on GNF2, this section will be based on ELTRI. NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTRI) was approved by the NRC as an acceptable method for evaluating the effects of EPUs.
)) The Loss of Non-Emergency AC Power to the Station Auxiliaries event is not listed in (( )) to be analyzed for EPU. The following is a summary of the evaluation provided for the Loss of Non-Emergency AC Power to the Station Auxiliaries event:
Consistent with ELTRI, the Loss of Non-Emergency AC Power to the Station Auxiliaries event is
]1 Conclusion The analysis of the loss of non-emergency AC power to station auxiliaries event has been reviewed to ensure it adequately accounted for operation of the plant at the proposed power level and was performed using acceptable analytical models. The results of that analysis demonstrate that the reactor protection and safety systems will continue to ensure that the SAFDLs and the RCPB pressure limits will not be exceeded as a result of this event. Based on this, Exelon concludes that the plant will continue to meet the requirements of the current licensing basis, following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to the loss of non-emergency AC power to station auxiliaries event.
2.8.5.2.3 Loss of Normal FeedwaterFlow Regulatory Evaluation A loss of normal FW flow could occur from pump failures, valve malfunctions, or a LOOP. LOFW flow results in an increase in reactor coolant temperature and pressure, and eventually requires a reactor trip to prevent fuel damage. Decay heat must be transferred from fuel following a loss of 2-401
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) normal FW flow. Reactor protection and safety systems are actuated to provide this function and mitigate other aspects of the transient. A review has been performed of the effects of the proposed EPU on: (1) the sequence of events; (2) the analytical model used for analyses; (3) the values of parameters used in the analytical model; and (4) the results of the transient analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations, including AOOs; (2) GDC-15, insofar, as it requires that the RCS and its associated auxiliary systems be designed with margin sufficient to ensure that the design conditions of the RCPB are not exceeded during any condition of normal operation; and (3) GDC-26, insofar as it requires that a reactivity control system be provided, and be capable of reliably controlling the rate of reactivity changes to ensure that under conditions of normal operation, including AOOs, SAFDLs are not exceeded.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-15, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:
Draft GDC-6, Draft GDC-27, Draft GDC-29, and Draft GDC-30. There is no Draft GDC directly associated with current GDC- 15.
The analysis of the loss of normal FW flow transient is described in PBAPS UFSAR Sections 4.7, "Reactor Core Isolation Cooling System," and 14.5.4.3, "Loss of Feedwater Flow."
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 9.1.3 of the CLTR addresses the effect of CPPU on Loss of Water Level Events. The results of this evaluation are described below.
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PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:
Loss of Water Level Events Meets CLTR 1[ Disposition (Loss of Feedwater Flow)
Loss of Water Level Events Meets CLTR (Loss of One Feedwater Pump) Disposition 2.8.5.2.3.1 Loss of FeedwaterFlow Event As stated in the CLTR, higher decay heat results in a lower reactor water level for loss of water level events.
For the LOFW event, adequate transient core cooling is provided by maintaining the water level inside the core shroud above the TAF. A (( )) was performed for PBAPS at EPU conditions. This analysis assumed failure of the HPCI system and used only the RCIC system to restore the reactor water level.
Because of the extra decay heat from EPU, slightly more time is required for the automatic systems to restore water level. Operator action is only needed for long-term plant shutdown. The results of the LOFW analysis for PBAPS show that the minimum water level inside the shroud is 129 inches above the TAF at EPU conditions. After the water level is restored, the operator manually controls the water level, reduces reactor pressure, and initiates RHR shutdown cooling. This sequence of events does not require any new operator actions or shorter operator response times. Therefore, the operator actions for an LOFW transient do not significantly change for EPU.
As described in Table 1-1, for Transient Analysis, the modeling tool used is the SAFER04 model, which is the same model used in the ECCS LOCA analysis. The analysis is consistent with the CLTR. The following is the general sequence of events in the analysis. The reactor is assumed to be at 102% of the EPU power level when the LOFW occurs. The initial level in the model is conservatively set at the low-level scram setpoint and reactor FW is instantaneously isolated at event initiation. Scram is initiated at the start of the event. When the level decreases to the low-low level setpoint, the RCIC system is initiated. The RCIC flow to the vessel begins at 68 seconds into the event, minimum level is reached at 1273 seconds and level is recovered after that point.
Only RCIC flow is credited to recover the reactor water level. There are no additional failures assumed beyond the failure of the HPCI system.
The only other key analysis assumption for the LOFW analysis, discussed in Section 9.1.3 of the CLTR, was the assumed decay heat level of ANS 5.1-1979 with a two-sigma uncertainty. The assumed decay heat level for the EPU analysis was ANS 5.1-1979 decay heat +10%, which bounds ANS 5.1-1979 + two sigma. Thus, the key analytical assumptions are the same or conservative relative to the current licensing basis.
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This LOFW analysis is performed to demonstrate acceptable RCIC system performance. The design basis criterion for the RCIC system is confirmed by demonstrating that it is capable of maintaining the water level inside the shroud above the TAF during the LOFW transient. The minimum level (see Figure 2.8-24) is maintained at least 129 inches above the TAF, thereby demonstrating acceptable RCIC system performance. There are no applicable equipment out of service assumptions for this transient.
As discussed in Section 2.8.4.3, an operational requirement is that the RCIC system restores the reactor water level while avoiding ADS timer initiation and MSIV closure activation functions associated with the low-low-low reactor water level setpoint (Level 1). This requirement is intended to avoid unnecessary initiations of safety systems. This requirement is not a safety-related function. The results of the LOFW analysis for PBAPS show that the nominal Level 1 setpoint trip is avoided.
Therefore, the LOFW event meets all CLTR dispositions.
2.8.5.2.3.2 Loss of One Feedwater Pump As stated in the CLTR, higher decay heat results in a lower reactor water level for loss of water level events.
The Loss of One Feedwater Pump event was included in ELTRI only for operational considerations. As stated in the USNRC Safety Evaluation, Section 4.5, to ELTR2, "A plant-specific analysis of the loss of one FW pump event will be submitted per Appendix E of ELTRI to assess the effect of a higher flow control line on scram avoidance."
The Loss of One FW pump event only addresses operational considerations to avoid reactor scram on low reactor water level (Level 3). This requirement is intended to avoid unnecessary reactor shutdowns. Because the MELLLA region is extended along the existing upper boundary to the EPU RTP, there is no increase in the highest flow control line for the PBAPS EPU.
Therefore, the Loss of One FW Pump event meets all CLTR dispositions.
Conclusion The analysis of the Loss of Normal FW Flow event has been reviewed to ensure it adequately accounted for operation of the plant at the proposed power level and was performed using acceptable analytical modelsk The results of that analysis demonstrate that the reactor protection and safety systems will continue to ensure that the SAFDLs and the RCPB pressure limits will not be exceeded as a result of the loss of normal FW flow. Based on this, Exelon concludes that the plant will continue to meet the requirements of the current licensing basis, following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to the Loss of Normal FW Flow event.
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NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 2.8.5.3 Decrease in Reactor Coolant System Flow 2.8.5.3.1 Loss of ForcedReactor Coolant Flow Regulatory Evaluation A decrease in reactor coolant flow occurring while the plant is at power could result in a degradation of core heat transfer. An increase in fuel temperature and accompanying fuel damage could then result if SAFDLs are exceeded during the transient. Reactor protection and safety systems are actuated to mitigate the transient. A review has been performed of the effects of the proposed EPU on: (1) the postulated initial core and reactor conditions; (2) the methods of thermal and hydraulic analyses; (3) the sequence of events; (4) assumed reactions of reactor systems components; (5) the functional and operational characteristics of the RPS; (6) operator actions; and (7) the results of the transient analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations, including AOOs; (2) GDC-15, insofar as it requires that the RCS and its associated auxiliary systems be designed with margin sufficient to ensure that the design conditions of the RCPB are not exceeded during any condition of normal operation; and (3)
GDC-26, insofar as it requires that a reactivity control system be provided, and be capable of reliably controlling the rate of reactivity changes to ensure that under conditions of normal operation, including AOOs, SAFDLs are not exceeded.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-15, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:
Draft GDC-6, Draft GDC-27, Draft GDC-29, and Draft GDC-30. There is no Draft GDC directly associated with current GDC- 15.
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The analysis of the loss of forced reactor coolant flow is described in PBAPS UFSAR Section 14.5.5, "Events Resulting in a Core Coolant Flow Decrease."
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel products.
Because PBAPS is based on GNF2, this section will be based on ELTRI. NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTRI) was approved by the NRC as an acceptable method for evaluating the effects of EPUs.
I((
)) The Loss of Forced Reactor Coolant Flow event is not listed in ((
)) to be analyzed for EPU. The following is a summary of the evaluation provided for the Loss of Forced Reactor Coolant Flow event:
The Loss of Forced Reactor Coolant Flow event, including the Trip of Pump Motor and Flow Controller Malfunctions events, result in a decrease in reactor power level and increase in margins.
These events are The replacement of the recirculation pump motor power supplies from M/G sets to ASDs does not affect the transient response upon Trip of Pump Motor or Flow Controller Malfunction -
Decreasing Flow events. The ASD modification (see PBAPS EPU LAR Attachment 9, Section 2, for additional information) does not affect the recirculation pump, shaft or motor inertia. Thus, the Trip of Pump Motor transient response will be the same with the ASD modification as with the original M/G sets. The PBAPS ASD design does not incorporate a braking function upon a flow controller malfunction. ((
))for PBAPS.
Conclusion The analyses of the decrease in reactor coolant flow event have been reviewed to ensure they have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. These analyses demonstrate that the reactor protection and safety systems will continue to ensure that the SAFDLs and the RCPB pressure limits will not be exceeded as a result of this event. Based on this, Exelon concludes that the plant will continue to meet the requirements of the current licensing basis, following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to the decrease in reactor coolant flow event.
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NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 2.8.5.3.2 Reactor RecirculationPump Rotor Seizure and Reactor Recirculation Pump Shaft Break Regulatory Evaluation The events postulated are an instantaneous seizure of the rotor or break of the shaft of a reactor recirculation pump. Flow through the affected loop is rapidly reduced, leading to a reactor and turbine trip. The sudden decrease in core coolant flow while the reactor is at power results in a degradation of core heat transfer that could result in fuel damage. The initial rate of reduction of coolant flow is greater for the rotor seizure event. However, the shaft break event permits a greater reverse flow through the affected loop later during the transient and, therefore, results in a lower core flow rate at that time. In both events, reactor protection and safety systems are actuated to mitigate the transient. A review has been performed of the effects of the proposed EPU on: (1) the postulated initial and long-term core and reactor conditions; (2) the methods of thermal and hydraulic analyses; (3) the sequence of events; (4) the assumed reactions of reactor system components; (5) the functional and operational characteristics of the RPS; (6) operator actions; and (7) the results of the transient analyses. The regulatory acceptance criteria are based on: (1)
GDC-27, insofar as it requires that the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the ECCS, of reliably controlling reactivity changes under postulated accident conditions, with appropriate margin for stuck rods, to assure the capability to cool the core is maintained; (2) GDC-28, insofar as it requires that the reactivity control systems be designed to assure that the effects of postulated reactivity accidents can neither result in damage to the RCPB greater than limited local yielding, nor disturb the core, its support structures, or other reactor vessel internals so as to significantly impair the capability to cool the core; and (3) GDC-31, insofar as it requires that the RCPB be designed with margin sufficient to assure that, under specified conditions, it will behave in a non-brittle manner and the probability of a rapidly propagating fracture is minimized.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
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For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-27, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:
Draft GDC-32, Draft GDC-33, Draft GDC-34, and Draft GDC-35. There is no Draft GDC directly associated with current GDC-27.
The analysis of a recirculation pump seizure accident is described in PBAPS UFSAR Section 14.5.5.4, "Recirculation Pump Seizure."
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel products.
Because PBAPS is based on GNF2, this section will be based on ELTRI. NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTRI) was approved by the NRC as an acceptable method for evaluating the effects of EPUs.
)) The following is a summary of the evaluation provided for the Reactor Recirculation Pump Rotor Seizure and Reactor Recirculation Pump Shaft Break events:
The Reactor Recirculation Pump Rotor Seizure event results in a decrease in reactor core coolant flow rate. Events in this category, ((
)) Therefore, the Reactor Recirculation Pump Rotor Seizure event is not analyzed for EPU.
The Reactor Recirculation Pump Shaft Break event results in a decrease in reactor core coolant flow rate. Events in this category, ((
)) Therefore, the Reactor Recirculation Pump Shaft Break event is not analyzed for EPU.
Conclusion The analyses of the sudden decrease in core coolant flow events have been reviewed to ensure they have adequately accounted for operation of the plant at the proposed power level and were performed using, acceptable analytical models. These analyses demonstrate that the reactor protection and safety systems will continue to ensure that the ability to insert control rods is maintained, the RCPB pressure limits will not be exceeded, the RCPB will behave in a non-brittle manner, the probability of propagating fracture of the RCPB is minimized, and adequate core cooling will be provided. Based on this, Exelon concludes that the plant will continue to meet the 2-408
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Therefore, Exelon finds the proposed EPU acceptable with respect to the sudden decrease in core coolant flow events.
2.8.5.4 Reactivity and Power Distribution Anomalies 2.8.5.4.1 UncontrolledControl Rod Assembly Withdrawalfrom a Subcriticalor Low Power Startup Condition Regulatory Evaluation An uncontrolled control rod assembly withdrawal from subcritical or low power startup conditions may be caused by a malfunction of the reactor control or rod control systems. This withdrawal will uncontrollably add positive reactivity to the reactor core, resulting in a power excursion. A review has been performed of the effects of the proposed EPU on: (1) the description of the causes of the transient and the transient itself; (2) the initial conditions; (3) the values of reactor parameters used in the analysis; (4) the analytical methods and computer codes used; and (5) the results of the transient analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations, including AOOs; (2) GDC-20, insofar as it requires that the RPS be designed to initiate automatically the operation of appropriate systems, including the reactivity control systems, to ensure that SAFDLs are not exceeded as a result of AOOs; and (3) GDC-25, insofar as it requires that the protection system be designed to assure that SAFDLs are not exceeded for any single malfunction of the reactivity control systems.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is 2-409
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The analysis of an RWE transient is described in PBAPS UFSAR Sections 14.5.3.2, "Continuous Rod Withdrawal During Reactor Startup," and 14.5.3.3, "Control Rod Removal Error During Refueling."
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 5.1.2 of the CLTR addresses the effect of CPPU on Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition. The results of this evaluation are described below.
The evaluation of the Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low.
Power Startup Condition event for EPU is a comparison of the expected maximum increase in peak fuel enthalpy with the acceptance criterion of 170 cal/gram. The CLTP Uncontrolled Control Rod Assembly Withdrawal analysis is based on Reference 96. The PBAPS EPU core consists only of GE (GNF) fuel assemblies and the EPU is limited to < 120% of OLTP. There is no change to the reactor manual control system or control rod hydraulic control units for EPU. The RWM installed provides the same level of protection for GE (GNF) fuel following EPU provided the power increase is < 120% of OLTP, and BPWS is used at power levels below the lower LPSP AL. The evaluation of this event for the PBAPS EPU considering these features and GE (GNF) fuel demonstrates the CLTR disposition is applicable. No change in peak fuel enthalpy is expected due to EPU because an RWE is a localized low-power event. If the peak fuel rod enthalpy is conservatively assumed to increase by a factor of 1.2, the RWE peak fuel enthalpy at EPU will be 72 cal/gram. This enthalpy is well below the acceptance criterion of 170 cal/gram.
Conclusion The analysis of the uncontrolled control rod assembly withdrawal from a subcritical or low power startup condition was reviewed to ensure it has adequately accounted for the changes in core design necessary for operation of the plant at the proposed power level. The review also confirmed the analysis was performed using acceptable analytical models. The analysis demonstrates that the reactor protection and safety systems will continue to ensure the SAFDLs are not exceeded. Based on this, Exelon concludes that the plant will continue to meet the requirements of the current licensing basis, following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to the uncontrolled control rod assembly withdrawal from a subcritical or low power startup condition.
2.8.5.4.2 UncontrolledControl Rod Assembly Withdrawal at Power Regulatory Evaluation An uncontrolled control rod assembly withdrawal at power may be caused by a malfunction of the reactor control or rod control systems. This withdrawal will uncontrollably add positive reactivity 2-410
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) to the reactor core, resulting in a power excursion. A review has been performed of the effects of the proposed EPU on: (1) the description of the causes of the AOO and the description of the event itself; (2) the initial conditions; (3) the values of reactor parameters used in the analysis; (4) the analytical methods and computer codes used; and (5) the results of the associated analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations, including AOOs; (2) GDC-20, insofar as it requires that the RPS be designed to initiate automatically the operation of appropriate systems, including the reactivity control systems, to ensure that SAFDLs are not exceeded as a result of AOOs; and (3) GDC-25, insofar as it requires that the protection system be designed to assure that SAFDLs are not exceeded for any single malfunction of the reactivity control systems.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-6, Draft GDC-14, Draft GDC-15, and Draft GDC-31.
The analysis of an RWE transient is described in PBAPS UFSAR Section 14.5.3.1, "Continuous Rod Withdrawal During Power Range Operation."
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. However, the CLTR is limited to application of GEl4 and earlier fuel products.
Because PBAPS is based on GNF2, this section will be based on ELTRI. NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, 2-411
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February 1999 (also referred to as ELTRI) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Although ELTRI is the licensing basis for the PBAPS AOO events, for the events discussed in this section, the sensitivity of the CPR changes provided as part of the transient analysis results in Table 2.8-12 is consistent with the sensitivity that established the scope approved in the CLTR.
)) The following is a summary of the evaluation provided for the Uncontrolled Control Rod Assembly Withdrawal at Power event.
Consistent with the CLTR, the Uncontrolled Control Rod Assembly Withdrawal at. Power (RWE) event is confirmed to be within the PBAPS reload evaluation scope. The RWE is performed with the NRC-approved methods described in GESTAR II (Reference 4). The computer code used to evaluate the RWE is PANACEA. The transient evaluation initial conditions are provided in Table 2.8-11, and the result of the EPU evaluation is reported in Table 2.8-12.
Conclusion The analysis of the uncontrolled control rod assembly withdrawal at power event hasbeen reviewed to ensure that the analysis has adequately accounted for the changes in core design required for operation of the plant at the proposed power level and was performed using acceptable analytical models. The analysis demonstrates that the reactor protection and safety systems will continue to ensure the SAFDLs are not exceeded. Based on this, Exelon concludes that the plant will continue to meet the requirements of the current licensing basis, following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to the Uncontrolled Control Rod Assembly Withdrawal at Power event.
2.8.5.4.3 Startup of a Recirculation Loop at an Incorrect Temperature andFlow Controller Malfunction Causing an Increase in Core Flow Rate Regulatory Evaluation A startup of an inactive loop transient may result in either an increased core flow or the introduction of cooler water into the core. This event causes an increase in core reactivity due to decreased moderator temperature and core void fraction. A review has been performed of the effects of the proposed EPU on: (1) the sequence of events; (2) the analytical model; (3) the values of parameters used in the analytical model; and (4) the results of the transient analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to assure that SAFDLs are not exceeded during any condition of normal operation, including the effects of AOOs; (2) GDC-20, insofar as it requires that the protection system be designed to initiate automatically the. operation of appropriate systems to ensure that SAFDLs are not exceeded as a result of operational occurrences; (3) GDC-15, insofar as it requires that the RCS and its associated auxiliary systems be designed with margin sufficient to ensure that the design condition of the RCPB are not exceeded during AOOs; (4) GDC-28, insofar as it requires that the reactivity control systems be designed to assure that the effects of postulated reactivity accidents can neither result in damage to the RCPB greater than limited local yielding, nor disturb the core, 2-412
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) its support structures, or other reactor vessel internals so as to significantly impair the capability to cool the core; and (5) GDC-26, insofar as it requires that a reactivity control system be provided, and be capable of reliably controlling the rate of reactivity changes to ensure that under conditions of normal operation, including AOOs, SAFDLs are not exceeded.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-15, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:
Draft GDC-6, Draft GDC-14, Draft GDC-15, Draft GDC-27, Draft GDC-29, Draft GDC-30, and Draft GDC-32. There is no Draft GDC directly applicable to the current GDC-15.
The analysis of the startup of a recirculation loop at an incorrect temperature or a flow controller malfunction causing an increase in core flow rate is described in PBAPS UFSAR Section 14.5.6, "Events Resulting in a Core Coolant Flow Increase."
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. However, the CLTR is limited to application of GEl4 and earlier fuel products.
Because PBAPS is based on GNF2, this section will be based on ELTR1. NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTR1) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Although ELTRI is the licensing basis for the PBAPS AOO events, for the events discussed in this section, the sensitivity of the CPR changes provided as part of the transient analysis results in Table 2.8-12 is consistent with the sensitivity that established the scope approved in the CLTR.
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)) The following is a summary of the evaluation provided for the Startup of a Recirculation Loop at an Incorrect Temperature and the Flow Controller Malfunction Causing an Increase in Core Flow Rate events:
Consistent with the CLTR, ((
The Failure of the Recirculation Flow Controller can result in either a slow or fast recirculation increase. The disposition of these events for EPU indicates that ((
)) The transient evaluation initial conditions are provided in Table 2.8-11, and the results of the EPU evaluations are reported in Table 2.8-12.
The Failure of the Recirculation Flow Controller with fast recirculation increase was also analyzed for the ASD modification (see PBAPS EPU LAR Attachment 9, Section 2, for additional information) to replace the recirculation system M/G sets. The results of this evaluation are reported in Table 2.8-12.
Per NUREG-0800 Section 15.4.4-15.4.5, "Startup of an Inactive Loop or Recirculation Loop at an Incorrect Temperature, and Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate," Revision 2, March 2007 (Reference 61), for a BWR, "Startup of a Recirculation Loop at an Incorrect Temperature event" is called the "Startup of an Idle Recirculation Pump" event.
The Startup of an Idle Recirculation Pump event is ((
)) ARTS was approved for PBAPS by References 97, 98 and 99.
Although the recirculation pump startup sequence is affected by replacing the MG set with ASD, the analysis assumptions that form the bases of the Startup of an Idle Recirculation Pump event are not affected. Therefore, this assures that the ASD installation has no effect on the Startup of an Idle Recirculation Pump analysis.
Conclusion The analyses of the increase in core flow events have been reviewed to ensure they have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. These analyses demonstrate that the reactor protection and safety systems will continue to ensure that the SAFDLs and the RCPB pressure limits will not be exceeded as a result of these events. Based on this, Exelon concludes that the plant will continue to meet the requirements of the current licensing basis, following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to the increase in core flow events.
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NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 2.8.5.4.4 Spectrum of Rod Drop Accidents Regulatory Evaluation Another type of reactivity or power distribution anomaly event is the CRDA. A review has been performed of the effects of the proposed EPU on the occurrences that lead to the accident, safety features designed to limit the amount of reactivity available and the rate at which reactivity can be added to the core, the analytical model used for analyses, and the results of the analyses. The regulatory acceptance criteria are based on GDC-28, insofar as it requires that the reactivity control systems be designed to assure that the effects of postulated reactivity accidents can neither result in damage to the RCPB greater than limited local yielding, nor disturb the core, its support structures, or other reactor vessel internals so as to significantly impair the capability to cool the core.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria,"
contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-32.
The analyses of rod drop accidents are described in PBAPS UFSAR Section 14.6.2, "Control Rod Drop Accident."
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 5.1.2 and 5.3.4 of the CLTR addresses the effect of CPPU on the Rod Control and Information System. The results of this evaluation are described below.
The spectrum of CRDAs does not change with EPU. The evaluation of a CRDA for the PBAPS EPU is a comparison of the expected maximum increase in peak fuel enthalpy with the acceptance criterion 2-415
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) of 280 cal/gram. The CLTP CRDA for PBAPS is based on Reference 52. The PBAPS EPU core consists only of GE (GNF) fuel assemblies and the EPU is limited to < 120% of OLTP. Control Rod Sequencing at PBAPS for CLTP and EPU follows the BPWS. There is no change to the PBAPS reactor manual control system or control rod hydraulic control units for EPU. The RWM installed at PBAPS provides the same level of protection for GE (GNF) fuel following EPU provided the power increase is < 120% of OLTP and BPWS is used at power levels below the lower LPSP AL. The evaluation of this event for the PBAPS EPU considering these features and GE (GNF) fuel demonstrates the CLTR disposition is applicable. No change in peak fuel enthalpy is expected due to EPU because with the rod drop accident is a limiting localized low-power event. If the peak fuel rod enthalpy is conservatively assumed to increase by a factor of 1.2, the CRDA peak fuel enthalpy at EPU will be 162 cal/gram. This enthalpy is well below the acceptance criterion of 280 cal/gram.
Conclusion The analyses of the rod drop accident have been reviewed to ensure they have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The analyses demonstrate that appropriate reactor protection and safety systems will prevent postulated reactivity accidents that could (1) result in damage to the RCPB greater than limited local yielding, or (2) cause sufficient damage that would significantly impair the capability to cool the core. Based on this, Exelon concludes that the plant will continue to meet the requirements of the current licensing basis, following implementation of EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to the rod drop accident.
2.8.5.5 Inadvertent Operation of ECCS or Malfunction that Increases Reactor Coolant Inventory Regulatory Evaluation Equipment malfunctions, operator errors, and abnormal occurrences could cause unplanned increases in reactor coolant inventory. Depending on the temperature of the injected water and the response of the automatic control systems, a power level increase may result and, without adequate controls, could lead to fuel damage or overpressurization of the RCS. Alternatively, a power level decrease and depressurization may result. Reactor protection and safety systems are actuated to mitigate these events. A review has been performed of the effects of the proposed EPU on: (1) the sequence of events; (2) the analytical model used for analyses; (3) the values of parameters used in the analytical model; and (4) the results of the transient analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations, including A0Os; (2) GDC-15, insofar as it requires that the RCS and its associated auxiliary systems be designed with margin sufficient to ensure that the design conditions of the RCPB are not exceeded during AOOs; and (3)
GDC-26, insofar as it requires that a reactivity control system be provided, and be capable of reliably controlling the rate of reactivity changes to ensure that under conditions of normal operation, including AOOs, SAFDLs are not exceeded.
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Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria,"
contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-15, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:
Draft GDC-6, Draft GDC-27, Draft GDC-29, and Draft GDC-30. There is no Draft GDC directly applicable to the current GDC- 15.
The analysis of an inadvertent operation of ECCS or a malfunction that increases reactor coolant inventory is described in PBAPS UFSAR Section 14.5.2.1, "Inadvertent Pump Start (HPCIS)."
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel products.
Because PBAPS is based on GNF2, this section will be based on ELTRI. NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTR1) was approved by the NRC as an acceptable method for evaluating the effects of EPUs.
)) The following is a summary of the evaluation provided for the Inadvertent Operation of ECCS or Malfunction that Increases Reactor Coolant Inventory events:
Consistent with the CLTR, the Inadvertent Operation of ECCS or Malfunction that Increases Reactor Coolant Inventory (the Inadvertent HPCI System Start) is confirmed to be within the PBAPS reload evaluation scope. The transient evaluation initial conditions are provided in Table 2.8-11, and the results of the EPU evaluations are reported in Table 2.8-12.
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((I Conclusion The analyses of the inadvertent operation of ECCS or malfunction that increases reactor coolant inventory have been reviewed to ensure they have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. These analyses demonstrate that the reactor protection and safety systems will continue to ensure that the SAFDLs and the RCPB pressure limits will not be exceeded as a result of this event. Based on this, Exelon concludes that the plant will continue to meet the requirements of the current licensing basis, following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to the Inadvertent Operation of ECCS or Malfunction that Increases Reactor Coolant Inventory.
2.8.5.6 Decrease in Reactor Coolant Inventory 2.8.5.6.1 Inadvertent Opening of a PressureRelief Valve Regiulatory Evaluation The inadvertent opening of a pressure relief valve results in a reactor coolant inventory decrease and a decrease in RCS pressure. The pressure relief valve discharges into the suppression pool.
Normally there is no reactor trip. The pressure regulator senses the RCS pressure decrease and partially closes the TCVs to stabilize the reactor at a lower pressure. The reactor power settles out at nearly the initial power level. The coolant inventory is maintained by the FW control system using water from the CST via the condenser hotwell. A review has been performed of the effects of the proposed EPU on: (1) the sequence of events; (2) the analytical model used for analyses; (3) the values of parameters used in the analytical model; and (4) the results of the transient analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations, including AGOs; (2) GDC-15, insofar as it requires that the RCS and its associated auxiliary systems be designed with margin sufficient to ensure that the design conditions of the RCPB are not exceeded during AOOs; and (3) GDC-26, insofar as it requires that a reactivity control system be provided, and be capable of reliably controlling the rate of reactivity changes to ensure that under conditions of normal operation, including AOOs, SAFDLs are not exceeded.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although 2-418
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria,"
contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-15, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:
Draft GDC-6, Draft GDC-27, Draft GDC-29, and Draft GDC-30. There is no Draft GDC directly applicable to the current GDC- 15.
The analysis of an inadvertent opening of a pressure relief valve is described in PBAPS UFSAR Section 14.5.4.2, "Inadvertent Opening of a Relief Valve or Safety Valve."
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel products.
Because PBAPS is based on GNF2, this section will be based on ELTRI. NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTRI) was approved by the NRC as an acceptable method for evaluating the effects of EPUs.
[II
)) The Inadvertent Opening of a Pressure Relief Valve event is not listed in ((
)) to be analyzed for EPU. The following is a summary of the evaluation provided for the Inadvertent Opening of a Pressure Relief Valve event.
Consistent with ELTR1, the Inadvertent Opening of a Safety Valve event is ((
Conclusion The analysis of the inadvertent opening of a pressure relief valve event has been reviewed to ensure it has adequately accounted for operation of the plant at the proposed power level and was performed using acceptable analytical models. This analysis demonstrates that the reactor protection and safety systems will continue to ensure that the SAFDLs and the RCPB pressure 2-419
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) limits will not be exceeded as a result of this event. Based on this, Exelon concludes that the plant will continue to meet the requirements of the current licensing basis, following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to Inadvertent Opening of a Pressure Relief Valve event.
2.8.5.6.2 Emergency Core Cooling System and Loss-of-Coolant Accidents Regulatory Evaluation LOCAs are postulated accidents that would result in the loss of reactor coolant from piping breaks in the RCPB at a rate in excess of the capability of the normal reactor coolant makeup system to replenish it. Loss of significant quantities of reactor coolant would prevent heat removal from the reactor core, unless the water is replenished. The reactor protection and ECCS systems are provided to mitigate these accidents. A review has been performed of the effects of the proposed EPU on: (1) the determination of break locations and break sizes; (2) postulated initial conditions; (3) the sequence of events; (4) the analytical model used for analyses, and calculations of the reactor power, pressure, flow, and temperature transients; (5) calculations of PCT, total oxidation of the cladding, total hydrogen generation, changes in core geometry, and long-term cooling; (6) functional and operational characteristics of the reactor protection and ECCS systems; and (7) operator actions. The regulatory acceptance criteria are based on: (1) 10 CFR 50.46, insofar as it establishes standards for the calculation of ECCS performance and acceptance criteria for that calculated performance; (2) 10 CFR 50, Appendix K, insofar as it establishes required and acceptable features of evaluation models for heat removal by the ECCS after the blowdown phase of a LOCA; (3) GDC-4, insofar as it requires that SSCs important to safety be protected against dynamic effects associated with flow instabilities and loads such as those resulting from water hammer; (4) GDC-27, insofar as it requires that the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the ECCS, of reliably controlling reactivity changes under postulated accident conditions, with appropriate margin for stuck rods, to assure the capability to cool the core is maintained; and (5) GDC-35, insofar as it requires that a system to provide abundant emergency core cooling be provided to transfer heat from the reactor core following any LOCA at a rate so that fuel clad damage that could interfere with continued effective core cooling will be prevented.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria,"
contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with 2-420
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-27, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:
Draft GDC-37, Draft GDC-40, Draft GDC-41, Draft GDC-42, and Draft GDC-44. There is no Draft GDC directly applicable to current GDC-27.
The analysis of a loss-of-coolant accident is described in PBAPS UFSAR Section 14.6.3, "Loss of Coolant Accident." The description of ECCS Systems is found in PBAPS UFSAR Section 6, "Core Standby Cooling Systems."
In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10). The license renewal evaluations associated with HPCI, CS, RHR, and ADS are located in NUREG-1769 Sections 2.3.2.1, 2.3.2.2, 2.3.2.5, and 2.3.3.13. Management of aging effects associated with HPCI, CS, RHR, and ADS are documented in NUREG-1769 Sections 3.2.1, 3.2.2, 3.2.5, and 3.3.13.
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Sections 4.2.2, 4.2.3, 4.2.4, 4.2.5, and 4.3 of the CLTR address the effect of CPPU on the ECCS and LOCAs. The results of this evaluation are described below.
The ECCS includes the HPCI system, the CS system, the LPCI mode of the RHR system, and the ADS.
The PBAPS EPU LOCA analyses are based on NRC-approved GEH LOCA analysis methods and are in full compliance with 10 CFR 50.46. No new fuel designs are being introduced. No ECCS changes are required to meet LOCA analysis acceptance criteria.
Each ECCS is discussed in the following subsections. The effect on the functional capability of each system due to EPU is addressed. ((
PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:
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Meets CLTR High Pressure Coolant Injection Er Disposition Meets CLTR Core Spray Disposition Meets CLTR Low Pressure Coolant Injection System Disposition Meets CLTR Disposition Automatic Depressurization 2.8.5. 6.2.1 High PressureCoolantInjection The CLTR states that there is no change to the normal reactor operating pressure or the SRV setpoints.
Er )) the increase in decay heat changes the response of the reactor water level following a small break LOCA or a LOFW transient event.
There is no change to the normal reactor operating pressure or the SRV setpoints.
The HPCI system is designed to pump water into the reactor vessel over a wide range of operating pressures. The primary purpose of the HPCI system is to maintain reactor vessel coolant inventory in the event of a small break LOCA that does not immediately depressurize the reactor vessel. In this event, the HPCI system maintains reactor water level and helps depressurize the reactor vessel.
The adequacy of the HPCI system is demonstrated in the ECCS performance discussion at the end of this section.
Er )) for EPU, there is no change to the maximum nominal reactor operating pressure of 1050 psia and the SRV ALs remain the same at 1150 psig. ((
Because the maximum normal operating pressure and the SRV setpoints do not change for EPU, the HPCI System performance requirements do not change. ((
2.8.5.6.2.2 Core Spray The CLTR states that there is no change in the reactor pressures at which the CS is required.
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The CS system is automatically initiated in the event of a LOCA. When operating in conjunction with other ECCS, the CS system is required to provide adequate core cooling for all LOCA events.
There is no change in the reactor pressures at which the CS system is required.
The CS system sprays water into the reactor vessel after it is depressurized. The primary purpose of the CS system is to provide reactor vessel coolant inventory makeup for a large break LOCA and for any small break LOCA after the reactor vessel has depressurized. It also provides long-term core cooling in the event of a LOCA. The CS system meets all applicable safety criteria for the EPU.
The slight change in the system operating condition, such as peak suppression pool temperature and pressure, due to EPU for a postulated LOCA does not affect the hardware capabilities of the CS system. The generic CS distribution assessment provided in ELTR2 (Reference 4), Section 3.3, continues to be valid for EPU. CS distribution is not directly credited in the short-term cooling LOCA analyses. This is consistent with ECCS evaluation models specified in Appendix K to 10 CFR 50. Therefore, the convective heat transfer coefficients used during the short-term spray cooling period are the conservative values specified in Appendix K.
The CS system at PBAPS meets all CLTR dispositions because ((
2.8.5.6. 2.3 Low PressureCoolant Injection The CLTR states that there is no change in the reactor pressures at which the LPCI mode of RHR is required.
The LPCI mode of the RHR system is automatically initiated in the event of a LOCA. The primary purpose of the LPCI mode is to help maintain reactor vessel coolant inventory for a large break LOCA and for any small break LOCA after the reactor vessel has depressurized. The LPCI operating requirements are not affected by EPU and the ECCS performance evaluation demonstrates the adequacy of the LPCI mode core cooling performance.
The LPCI mode at PBAPS meets all CLTR dispositions.
2.8.5.6.2.4 Automatic DepressurizationSystem The CLTR states that EPU does not change the conditions at which the ADS must function.
The ADS uses SRVs to reduce the reactor pressure following a small break LOCA when it is assumed that the high-pressure systems have failed. This allows the CS and LPCI to inject coolant into the reactor vessel. EPU does not change the conditions at which the ADS must function. The ADS initiation logic and valve control are not affected by EPU conditions.
The adequacy of the ADS is demonstrated by the performance evaluation discussed in Section 2.8.5.6.2.5. The ADS at PBAPS meets all CLTR dispositions because the SRV setpoints and functions remain the same, the ADS timers are not changed and the small break LOCA event mitigation is acceptable.
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Because the PBAPS EPU is based on GNF2, this section will be based on ELTRI. NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate,"
Class III, February 1999 (also referred to as ELTRI) was approved by the NRC as an acceptable method for evaluating the effects of EPUs.
The PBAPS ECCS is designed to provide protection against postulated LOCAs caused by ruptures in the primary system piping. The ECCS performance characteristics are not changed for EPU.
The effects of EPU on ECCS-LOCA response is evaluated on a plant-specific basis. ECCS-LOCA performance analyses demonstrate that the 10 CFR 50.46 requirements continue to be met at the EPU rated thermal power conditions.
The basic break spectrum response is not affected by, EPU. There are two limiting points on the break spectrum: the full sized recirculation suction line break and the worst small break under the Battery Failure scenario. Consistent with Limitation and Condition 9.7 of the IMLTR SE (Reference 7), both top and mid peaked power shapes were considered for both large and small break LOCA. ((
)) The break spectrum response is determined by the ECCS network design and is common to all BWRs. GEH BWR power uprate evaluation experience shows that the basic break spectrum response is not affected by changes in core power.
For SLO, a multiplier is applied to the Two-Loop LHGR and MAPLHGR Operation limits. The operating conditions for SLO are not changed with EPU; therefore, the current SLO analysis remains acceptable for EPU. At EPU power condition, the MELLLA core flow extends to approximately 99.0 % of rated core flow. Therefore, the EPU analysis results at rated power and flow are applied to the MELLLA condition. Also, the effect of ICF on PCT is negligible with EPU.
Thus the SLO, MELLLA, and ICF domain remain valid with EPU.
The Licensing Basis PCT is based on the most limiting Appendix K case plus a plant variable uncertainty term that accounts statistically for the uncertainty in parameters that are not specifically addressed by 10 CFR 50 Appendix K. The Appendix K results demonstrate that the limiting LOCA is the small recirculation discharge line break under the limiting single failure of the Battery. The EPU Licensing Basis PCT for GNF2 fuel is less than 1925°F, which represents an increase from the CLTP Licensing Basis PCT of less than 1870'F evaluated at CLTP power and rated core flow. The EPU Licensing Basis PCT incorporates the effects of all identified Evaluation Model changes and errors as noted by the 10 CFR 50.46 reporting process through notification letter 2011-03.
Restrictions imposed by the NRC on Upper Bound PCT have been removed for PBAPS (Reference 100). The Upper Bound PCT has been shown to be bounded by the Licensing Basis PCT, consistent with the previous evaluation (References 101 and 102). The results of these analyses are provided in Table 2.8-6.
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Although no changes to the ECCS performance characteristics are required for EPU, in order to eliminate the need for CAP credit for ECCS NPSH after a DBLOCA, the PBAPS RHR system will be modified as described in EPU LAR Enclosure 9c, to limit the run-out flow of the RHR system in the LPCI mode of operation. The effect of this reduction in LPCI flow is evaluated to determine the effect on the ECCS-LOCA response. Exelon is also implementing a modification to the RRS independent of EPU to replace the current power supply to the reactor recirculation pump motors from motor-generator (M/G) sets to ASDs (see PBAPS EPU LAR Attachment 9, Section 2, for additional information). There is no change to the EPU Licensing Basis PCT for GNF2 fuel (less than 1925-F) with the reduction in LPCI flow and the ASD modification (Table 2.8-6).
2.8.5.6.2.5.1 Large Break Peak Clad Temperature (PCT) - Limiting Case (ECCS-LOCA)
The PBAPS break spectrum response is determined by the ECCS network design that is common to all BWRs. The PCT for the limiting large break LOCA is determined primarily by the hot bundle power, which is unchanged with EPU. In the PBAPS analysis, the hot bundle is assumed to be operating at the thermal limits (MCPR, MAPLHGR, and LHGR); these limits are not changed for EPU. Comparison of the PBAPS PCT results for CLTP and EPU indicate a minimal change, and therefore large break LOCA has a negligible effect on compliance with the other acceptance criteria of 10 CFR 50.46 (local cladding oxidation, core-wide metal-water reaction, coolable geometry and long-term cooling). The local fuel conditions are not significantly changed with EPU, because the hot bundle operation is still constrained by the same operating thermal limits. Because EPU has such a small effect on the PBAPS large break PCT, the system response over the large break spectrum is not affected.
The large break PCT results increased slightly as a result of the LPCI flow reduction but were significantly less than the limiting small break PCT results.
Elimination of the rotating mass associated with the reactor recirculation M/G Set when replaced with the ASD results in a reduction in inertia, or stopping characteristics, and subsequently a faster coastdown of the recirculation pumps. The reduction in inertia for the recirculation pump in the broken loop would have negligible effect, as blowdown is principally a function of the break location and area. However, the reduced inertia of the pump in the intact loop could have a more direct effect on core level and the first heatup in cladding temperature. ((
)) Analysis results for large break PCT with the ASD are shown in Table 2.8-6.
2.8.5.6.2.5.2 Small Break PCT - Break Spectrum (ECCS-LOCA)
The PBAPS break spectrum response is determined by the ECCS network design that is common to all BWRs. For PBAPS, the indicated decay heat for EPU is higher and results in a longer ADS blowdown and a higher PCT for the small break LOCA Appendix K case. Previous analyses (References 101 and 102) demonstrate that PBAPS is a small break Appendix K PCT limited plant.
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The effect of EPU on the calculated small break PCT is acceptable as long as the impact of the results on the Licensing Basis PCT remains below the 10 CFR 50.46 limits. The current TS values for ECCS initiation are bounded by the analysis; no changes to these values were required for EPU.
Plant-specific analyses demonstrate that there is sufficient ADS capacity, with five ADS valves in service and none out of service, at EPU conditions, to remain below these limits. Key input parameters to the SAFER/GESTR LOCA evaluation model are provided in Table 2.8-5. Input parameters are selected as nominal or representative values. For Appendix K calculations, select inputs are chosen so as to set a bounding condition or to assure conservatism.
The small break PCT results did not change as a result of the LPCI flow reduction because it occurs at or before LPCI injection. However, the current TS values for ECCS initiation will change as a result of the LPCI flow reduction.
2.8.5.6.2.5.3 Local Cladding Oxidation (ECCS-LOCA)
EPU has negligible effect on the Local Cladding Oxidation (ECCS LOCA). ((
)) This conclusion on local cladding oxidation is confirmed by the increase in plant Licensing Basis PCT due to EPU, which is sufficiently below the 10 CFR 50.46 limit of 2200'F. The bounding PBAPS EPU Licensing Basis PCT of 1925'F ensures this 10 CFR 50.46 requirement is met.
EPU with the LPCI flow reduction has no effect on the Local Cladding Oxidation (ECCS LOCA).
EPU with the ASD modification has no effect on the Local Cladding Oxidation (ECCS LOCA).
2.8.5.6.2.5.4 Core-Wide Metal-Water Reaction (ECCS-LOCA)
EPU has no effect on the Core-Wide Metal-Water Reaction (ECCS LOCA). ((
)) This conclusion on the core-wide metal-water reaction is confirmed by the increase in plant Licensing Basis PCT due to EPU, which is sufficiently below the 10 CFR 50.46 limit of 2200'F. The bounding PBAPS EPU Licensing Basis PCT of 1925'F ensures this 10 CFR 50.46 requirement is met.
EPU with the LPCI flow reduction has no effect on the Core-Wide Metal-Water Reaction (ECCS LOCA).
EPU with the ASD modification has no effect on the Core-Wide Metal-Water Reaction (ECCS LOCA).
2.8.5.6.2.5.5 Coolable Geometry (ECCS-LOCA)
EPU has no effect on the Coolable Geometry (ECCS LOCA). Coolable geometry has been dispositioned for BWRs per Reference 67. Conformance with coolable geometry requirements is demonstrated by conformance with the 2200'F Licensing Basis PCT limit and local cladding 2-426
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2.8.5.6.2.5.6 Long-Term Cooling (ECCS-LOCA)
EPU has no effect on the Long-Term Cooling (ECCS LOCA). Long-term cooling has been dispositioned for BWRs per Reference 67. Long-term cooling is assured by either: (1) core can be reflooded above TAF or (2) core can be reflooded to the elevation of the jet pump suction and one CS system can be placed in operation at rated flow. The ECCS system design at PBAPS ensures this 10 CFR 50.46 requirement is met.
Conclusion The analyses of the loss of coolant accident have been reviewed to ensure they have adequately accounted for operation of the plant at the proposed power level and that the analyses were performed using acceptable analytical models. The analyses demonstrate that the RPS and the ECCS will continue to ensure that the PCT, total oxidation of the cladding, total hydrogen generation, and changes in core geometry, and long-term cooling will remain within acceptable limits. Based on this, Exelon concludes that the plant will continue to meet the requirements of 10 CFR 50.46, 10 CFR 50 Appendix K and the current licensing basis following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to the LOCA.
2.8.5.7 Anticipated Transients Without Scram Regulatory Evaluation ATWS is defined as an AOO followed by the failure of the reactor portion of the protection system specified in GDC-20. 10 CFR 50.62 requires that:
- Each BWR have an ARI system that is designed to perform its function in a reliable manner and be independent (from the existing reactor trip system) from sensor output to the final actuation device.
- Each BWR have a SLCS with the capability of injecting into the reactor vessel a borated water solution with reactivity control at least equivalent to the control obtained by injecting 86 gpm of a 13 weight-percent sodium pentaborate decahydrate solution at the natural boron-10 isotope abundance into a 251-inch inside diameter reactor vessel. The SLCS initiation must be automatic.
- Each BWR have equipment to trip the reactor coolant recirculation pumps automatically under conditions indicative of an ATWS.
A review has been performed of the effects of the proposed EPU on: (1) the ability to meet the above requirements; (2) sufficient margin available in the setpoint for the SLCS pump discharge relief valve; and (3) operator actions specified in the EOPs (consistent with the generic emergency procedure guidelines/severe accident guidelines (EPGs/SAGs)). The review confirmed that: (1) the peak vessel bottom pressure is less than the ASME Service Level C limit of 1500 psig; (2) the peak clad temperature is within the 10 CFR 50.46 limit of 2200'F; (3) the peak suppression pool 2-427
NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) temperature is less than the design limit; and (4) the peak containment pressure is less than the containment design pressure. In addition, the potential for thermal-hydraulic instability in conjunction with ATWS events was also evaluated using the methods and criteria approved by the NRC staff. This evaluation considered the effects of the proposed EPU on the limiting event determination, the sequence of events, the analytical model and its applicability, the values of parameters used in the analytical model, and the results of the analyses.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
For the current GDC listed in the Regulatory Evaluation above, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-14 and Draft GDC-15.
The description of systems required by 10 CFR 50.62 is described in PBAPS UFSAR Sections 1.6.3.4, "Alternate Rod Insertion," 3.8, "Standby Liquid Control System," and 7.9.4.4.2, "Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT)."
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 9.3.1 of the CLTR addresses the effect of CPPU on ATWS.
Analysis of ATWS events is required for CLTP and for EPU RTP to ensure that the following ATWS acceptance criteria are met:
" Maintain reactor vessel integrity (i.e., peak vessel bottom pressure less than the ASME Service Level C limit of 1500 psig).
" Maintain containment integrity (i.e., maximum containment pressure and temperature less than the design pressure (56 psig) and temperature (180'F) of the containment structure).
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- Maintain coolable core geometry (Coolable core geometry is assured by meeting the 2200'F PCT and the 17% local cladding oxidation acceptance criteria of 10 CFR 50.46).
This evaluation reviewed the results of the ATWS analyses considering the limiting cases for RPV overpressure and for suppression pool temperature / containment pressure. Previous evaluations considered four ATWS events. Based on experience and the generic analyses performed for Reference 4 (ELTR2), only two cases need to be further analyzed for PBAPS: (1) Main Steam Isolation Valve Closure (MSIVC) and (2) PRFO. For PBAPS, a LOOP does not result in a reduction in the RHR pool cooling capability relative to these cases. Thus, with the same RHR pool cooling capability, the containment responses for the MSIVC and PRFO cases bound the LOOP case. These events have been analyzed and the results are presented below in Sections 2.8.5.7.1 through 2.8.5.7.3.
The EPU ATWS analysis is performed using the NRC-approved code ODYN (see Table 1-1). The key inputs to the ATWS analysis are provided in Table 2.8-7. The results of the analysis are provided in Table 2.8-8 and discussed below.
The results of the ATWS analysis meet the above ATWS acceptance criteria. Therefore, the PBAPS response to an ATWS event at EPU is acceptable. The potential for thermal-hydraulic instability in conjunction with ATWS events is evaluated in Section 2.8.3.2.
PBAPS meets the ATWS mitigation requirements defined in 10 CFR 50.62:
- Installation of an ARI system;
- Installation of automatic RPT logic (i.e., ATWS-RPT).
The 86 gpm boron injection equivalency requirement of 10 CFR 50.62 is satisfied via the following relationship:
(Q/86) x (M251/M) x (C/13) x (E/19.8) > 1 where:
Q = Expected SLCS flow rate (gpm)
M251/M = Mass of water in a 251-inch diameter reactor vessel and recirculation system (lbs) / mass of water in the reactor vessel and recirculation system at hot rated condition (lbs)
C= Sodium pentaborate solution concentration (weight percent)
E= Boron- 10 isotope enrichment (atom-percent)
For PBAPS, Q = 49.1 gpm M251/M = 1 (because PBAPS has a 251-inch diameter reactor vessel) 2-429
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92.00 % (i.e., natural boron enrichment)
Therefore, the 86 gpm equivalency requirement is satisfied as follows:
(Q/86) x (M251/M) x (C/13) x (E/19.8) > 1 (49.1/86) x (1) x (8.3/13) x (92.00/19.8) = 1.69 > 1 New operator actions are required to supplement the CST volume from the RWST and are described further in Section 2.11.1.2.3. No other changes to operator actions are required.
PBAPS meets all CLTR dispositions with exception for PCT, a fuel integrity related evaluation, as the CLTR disposition is applicable to GE fuel types up to GE14 and PBAPS EPU assumes a full core of GNF2. The results in this evaluation are described below. The topics addressed in this evaluation are:
Meets CLTR ATWS (Overpressure) - Event Selection (( Disposition Meets CLTR ATWS (Overpressure) - Limiting Events Disposition ATWS (Suppression Pool Temperature) Meets CLTR Event Selection Disposition ATWS (Suppression Pool Temperature) - Meets CLTR Limiting Events Disposition Does Not Meet CLTR ATWS (Peak Cladding Temperature) )) Disposition*
- The CLTR is not applicable to GEH fuel types beyond GE14. Because PBAPS EPU is based on a full core of GNF2 fuel, the PCT is explicitly evaluated for GNF2 fuel. See Section 2.8.5.7.3.
2.8.5.7.1 A TWS (Overpressure)
As stated in Section 9.3.1 of the CLTR, the higher operating steam flow will result in higher peak vessel pressures. The higher power and decay heat will result in higher suppression pool temperatures. The increased core power and reactor steam flow rates, in conjunction with the SRV capacity and response times, could affect the capability of the SLCS to mitigate the consequences of an ATWS event.
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The overpressure evaluation includes a review of the results of the analyses of ATWS events to identify the most limiting RPV overpressure conditions. Two events, MSIVC and PRFO, were further analyzed for PBAPS. The ATWS (Overpressure) - Event Selection meets all CLTR dispositions.
The MSIVC and PRFO sequence of events are given in Tables 2.8-9 and 2.8-10, respectively. The short-term and long-term transient response to the MSIVC and PRFO ATWS events is presented in Figures 2.8-25 through 2.8-40. The key inputs and limiting results are presented in Tables 2.8-7 and 2.8-8. The limiting ATWS event with respect to RPV overpressure for PBAPS is PRFO. The PRFO event produces the highest peak lower plenum pressure at SLCS initiation (1205 psia).
Therefore, ATWS (Overpressure) - Limiting Events meet all CLTR dispositions.
2.8.5.7.2 ATWS (SuppressionPool Temperature)
As stated in Section 9.3.1 of the CLTR, the higher operating steam flow will result in higher peak vessel pressures. The higher power and decay heat will result in higher suppression pool temperatures. The increased core power and reactor steam flow rates, in conjunction with the SRV capacity and response times, could affect the capability of the SLCS to mitigate the consequences of an ATWS event.
The suppression pool temperature evaluation includes a review of the results of the analyses of ATWS events to identify the most limiting containment response. Two events, MSIVC and PRFO, were further analyzed for PBAPS. The ATWS (Suppression Pool Temperature) - Event Selection meets all CLTR dispositions.
The MSIVC and PRFO sequence of events are given in Tables 2.8-9 and 2.8-10, respectively. The short-term and long-term transient responses to these events are presented in Figures 2.8-25 through 2.8-40. The key inputs and limiting results are presented in Tables 2.8-7 and 2.8-8. The limiting ATWS event with respect to containment response for PBAPS is PRFO. Therefore, the ATWS (Suppression Pool Temperature) - Limiting Events meet all CLTR dispositions.
2.8.5.7.3 ATWS (Peak Cladding Temperature)
Section 4.3 of the CLTR states that EPU has a negligible effect on the PCT or local cladding oxidation. ((
For ATWS events, the acceptance criteria for PCT and local cladding oxidation for ECCS, defined in 10 CFR 50.46, are adopted to ensure an ATWS event does not impede core cooling. Coolable core geometry is assured by meeting the 2200'F PCT and the 17% local cladding oxidation acceptance criteria stated in 10 CFR 50.46.
The ATWS analysis results demonstrate significant margin to the PCT acceptance criteria of 2200'F. Two events, MSIVC and PRFO, were further analyzed for PBAPS. The highest calculated 2-431
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PCT for ATWS events is 1342°F, which resulted from the PRFO event. Local cladding oxidation is not explicitly analyzed because, with PCT less than 1600'F, cladding oxidation has been demonstrated to be insignificant compared to the acceptance criteria of 17% of cladding thickness.
Therefore, the local cladding oxidation for the PBAPS ATWS events is qualitatively evaluated to demonstrate compliance with the acceptance criteria of 10 CFR 50.46.
Therefore, ATWS (PCT) is in compliance with the acceptance criteria of 10 CFR 50.46.
Conclusion The analysis of the ATWS event has been reviewed to ensure it has adequately accounted for the effects of the proposed EPU. The analysis demonstrates that ARI, SLCS, and RPT systems have been installed and that they will continue to meet the requirements of 10 CFR 50.62 and the analysis acceptance criteria following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to ATWS.
2.8.6 Fuel Storage 2.8.6.1 New Fuel Storage Regulatory Evaluation Nuclear reactor plants include facilities for the storage of new fuel. The quantity of new fuel to be stored varies from plant to plant, depending upon the specific design of the plant and the individual refueling needs. A review has been performed of the effects of the proposed EPU on the ability of the storage facilities to maintain the new fuel in a subcritical array during all credible storage conditions. The review focused on the effect of changes in fuel design on the analyses for the new fuel storage facilities. The regulatory acceptance criteria are based on GDC-62, insofar as it requires the prevention of criticality in fuel storage systems by physical systems or processes, preferably utilizing geometrically safe configurations.
Peach Bottom Current Licensing Basis Current GDC-62 is applicable to PBAPS as described in the NRC SER for PBAPS Unit 2 and Unit 3 License Amendments 175 and 178 (Reference 62), respectively.
New Fuel Storage is described in PBAPS UFSAR Section 10.2, "New Fuel Storage."
In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10).
The license renewal evaluation associated with the fuel storage is documented in NUREG-1769, Sections 2.3.3.1 and 2.3.3.2. Management of aging effects on fuel storage is documented in NUREG-1769, Section 3.3.
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Technical Evaluation Nuclear reactor plants include facilities for the storage of new fuel. PBAPS has a new fuel storage facility (also referred to as the new fuel storage vault) for each unit. However, this facility is not used and the new fuel is placed directly in the SFP following receipt inspection (PBAPS UFSAR Section 10.2 and PBAPS TS 4.3.1.2). Consequently, the effect of EPU on the new fuel storage facility has not been evaluated.
Conclusion Not applicable.
2.8.6.2 Spent Fuel Storage Regulatory Evaluation Nuclear reactor plants include storage facilities for the wet storage of spent fuel assemblies. The safety function of the SFP and storage racks is to maintain the spent fuel assemblies in a safe and sub-critical array during all credible storage conditions and to provide a safe means of loading the assemblies into shipping casks. A review has been performed of the effects of the proposed EPU on the criticality analysis (e.g., reactivity of the spent fuel storage array and Boraflex degradation or neutron poison efficacy). The regulatory acceptance criteria are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be
,compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents; and (2) GDC-62, insofar as it requires that criticality in the fuel storage systems be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC) Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.
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For the current GDC listed in the Regulatory Evaluation above, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-40, Draft GDC-42, and Draft GDC-66.
Current GDC-62 is applicable to PBAPS as described in the NRC SER for PBAPS Unit 2 and Unit 3 License Amendments 175 and 178 (Reference 62), respectively.
Spent Fuel Storage is described in PBAPS UFSAR Section 10.3, "Spent Fuel Storage."
In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10).
The license renewal evaluation associated with the fuel storage is documented in NUREG-1769, Sections 2.3.3.1 and 2.3.3.2. Management of aging effects on the fuel storage is documented in NUREG-1769, Section 3.3.
Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.3.4 of the CLTR addresses the effect of CPPU on Fuel Racks. The results of this evaluation are described below.
PBAPS meets all CLTR dispositions. The following topics are addressed in this section:
Dpr Result
°<
Meets CLTR Fuel Racks (( ))Disposition As explicitly stated in Section 6.3.4 of the CLTR, the increased decay heat from EPU results in a higher heat load in the racks during long-term storage.
Although there is an increase in the fuel pool heat load due to higher decay heat, the pool temperature continues to remain below the design temperature for all design-basis offload scenarios. The temperature will also remain below the operating limit for normal operating conditions. There is no effect on the spent fuel storage racks from the increased EPU heat load.
The current PBAPS Units 2 and 3 high-density SFP storage racks utilize Boraflex as a neutron absorber material for reactivity control. Due to Boraflex degradation, PBAPS implemented an ongoing Boraflex monitoring program, to include RACKLIFE simulation of the rack degradation and blackness testing using the BADGER B-10 areal density measurement system. The effect of EPU on the Boraflex degradation is accounted for by the Boraflex monitoring program.
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The original Boraflex rack design and PBAPS Licensing Basis ensured that the SFP will remain subcritical with the fuel pool fully loaded with fuel assemblies having a kinf < 1.362, assuming a uniform Boron-10 (B-10) areal density of 0.021 g/ cm2, the minimum certified manufactured B-10 density. A reduction in the amount of Boraflex in the SFP racks will reduce the criticality margin such that actions are required to ensure that the Licensing Basis requirements continue to be met.
To ensure the SFP storage racks can maintain criticality margin in accordance with the PBAPS TS 4.3.1.1.b requirement of 5 percent (Keff < 0.95), a peak in-core kinf limit of 1.2344 is used to support Boraflex degradation beyond the minimum certified areal density of 0.021 g/cm2. This limit is not impacted by EPU implementation. The kinf for the fuel bundles used in the representative equilibrium core design for PBAPS EPU is 1.2095, which is below the peak in-core kinf limit of 1.2344. The spent fuel storage facility will be capable of storing EPU fuel within the TS limits. To provide for future margin, Exelon is planning to install NETCO-SNAP-IN inserts.
The peak in-core kin1f limit is planned to increase to 1.270 following insert installation. Therefore, the Spent Fuel Storage meets all CLTR dispositions.
Conclusion The analyses related to the effects of the proposed EPU on the spent fuel storage capability have been reviewed to ensure they have adequately accounted for the effects of the proposed EPU on the spent fuel rack temperature and criticality analyses. The analyses demonstrate that the SFP design will continue to ensure an acceptably low temperature and an acceptable degree of sub-criticality following implementation of the proposed EPU. Based on this, Exelon concludes that the spent fuel storage facilities will continue to meet the requirements of the current licensing basis, following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to spent fuel storage.
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Table 2.8-1 Peak Nodal Exposures A 18 38.849 A 19 43.784 B 9 56.359 B 10 51.544 C 7 53.447 C 8 47.766 D 13 56.660 E 11 55.387 F Equilibrium - 120% 51.174 PBAPS EPU Equilibrium - 120% 55.578 2-436
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Table 2.8-2 Option III Setpoints Demonstration 1.04 1.216 1.223 1.05 1.241 1.248 1.06 1.267 1.274 1.07 1.294 1.301 1.08 1.323 1.330 1.09 1.352 1.360 1.10 1.382 1.389 1.11 1.412 1.420 1.12 1.445 1.453 1.13 1.479 1.487 1.14 1.514 1.522 Notes:
- 1. 5% calibration error is included in the OPRM Amplitude Setpoint.
- 2. 0.01 bypass voiding penalty was added to OLMCPR(SS) and OLMCPR(2PT) values. Assumed SLMCPR = 1.09.
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Table 2.8-3 BSP Region Intercepts at Nominal Feedwater Temperature I
Al 58.1 42.:0 0.798 0.256 AI-ICA 56.6 40.0 I t BI 36.7 30.8 0.800 0.162 B 1-ICA 40.7 31.0 I A2 64.6 50.1 0.799 0.226 I
-I
- ~ 4 A2-ICA 64.5 50.0 I 2-438
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Table 2.8-4 BSP Region Intercepts at Minimum Feedwater Temperature Al 68.8 55.5 0.791 0.213 AI-ICA 56.6 40.0 A2 69.1 55.9 0.800 0.197
-i A2-ICA 64.5 50.0 2-439
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Table 2.8-5 Key Input Parameters for SAFER/GESTR LOCA Evaluation Itern~ Pw~ree fi Nominal Value VpedxKNaluie Current Licensed Thermal MWt 3514 3528 Power (CLTP)
Analysis of Record LOCA Analysis Power (1) MWt 3623 3 Extended Power Uprate (EPU) MWt 3951 4030 4 Vessel Steam Dome Pressure psia 1060 1063 5 Rated Core Flow Mlbm/hr 102.5 102.5 Recirculation Suction Line Break 46.
Area (2) 4.168 4.168 Recirculation Discharge Line ft2 Break Area (3) 1.958 1.958 8 GNF2 Number of Fuel Rods per N/A 92 92 Bundle 9 GNF2 PLHGR kW/ft 13.75 14.40 10 GNF2 MAPLHGR kW/ft 13.15 13.78 11 GNF2 Worst Pellet Exposure for MWd/MT 14600 14600 ECCS Evaluation 12 Single Failure Input N/A Battery Battery 13 Limiting Large Break N/A Recirculation Recirculation Suction Location Suction Line Line 14 Limiting Small Break N/A Recirculation Recirculation Location Discharge Line Discharge Line Time constant of Recirculation M/G Set: 5 M/G Set: 5 15 Pump coastdown upon loss of AC Seconds ASD: 3 ASD: 3 input power I
- 2. The maximum recirculation suction line break area is the recirculation suction line break area (4.168 ft2) including the bottom head drain area (0.021 ft2).
- 3. The maximum recirculation discharge line break area is the recirculation discharge line break area (1.958 ft2) including the bottom head drain area (0.021 ft2).
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Table 2.8-6 ECCS Conformance Results Licensing Basis PCT (°F) < 1870 < 1925 < 1925 < 1925 *2200 Cladding Oxidation
(% Original Clad <4 <4 <4 <4 17 Thickness)
Hydrogen Generation, Core-Wide Metal-Water < 0.1 < 0.1 < 0.1 < 0.1 _ 1.0 Reaction (%)
PCT Oxidation
< 2200 °F <17%
and Local Coolable Geometry Acceptable Acceptable Acceptable Acceptable Core flooded to TAF OR Core flooded to jet pump Core Long Term Cooling Acceptable Acceptable Acceptable Acceptable suction elevation and at least one CS system is operating at rated flow.
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Table 2.8-6 ECCS Conformance Results (continued) 2-442
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Table 2.8-7 PBAPS Key Inputs for ATWS Analysis Reactor power (MWt) 3514 3951 Reactor dome pressure (psia) 1050 1050 Each SRV capacity at 1080 psig (Mlbm/hr) 0.8 0.8 Each SSV capacity at 1230 psig (Mlbm/hr) 0.9257 0.9257 High pressure ATWS-RPT setpoint (psig) 1106.0 1106.0 Number of SRVs 11 11 Number of SRVs OOS 1 1 Number of SSVs 2 3 Number of SSVs OOS 0 0 Table 2.8-8 PBAPS Results for ATWS Analysis
,Acceptance Crite~ria CLP [ EPU Peak vessel bottom pressure (psig) 1492 1458 Peak suppression pool temperature (0 F) 188.4 168.3 Peak containment pressure (psig) 11.6 8.3 Peak cladding temperature ( 0F) 1415 1342 Local cladding oxidation (%) < 17 < 17 2-443
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Table 2.8-9 MSIVC Sequence of Events I MSIV Isolation Initiated 0.0 0.0 2 High Pressure ATWS Setpoint 3.8 3.8 3 MSIVs Fully Closed 4.0 4.0 4 Peak Neutron Flux 4.1 4.0 5 Recirculation Pumps Trip 4.5 4.5 6 Opening of the First Relief Valve 4.5 4.5 7 Peak Heat Flux 4.9 4.9 8 Peak Vessel Pressure 9.8 9.5 9 FW Reduction Initiated 30.0 30.0 10 SLCS Pumps Start (1) 123.8 123.8 11 RHR Cooling Initiated 660 660 12 Hot Shutdown Boron Weight Achieved and Initiate Level 1013 1013 Increase 13 Hot Shutdown Achieved (Neutron Flux remains < 0.1%) 1208 1163 14 Peak Suppression Pool Temperature 10706 6738 Notes:
- 1. SLCS injection is the later time of either: 1) two minutes after the high-pressure RPT or 2) when the suppression pool temperature reaches the boron injection initiation temperature (BIIT).
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Table 2.8-10 PRFO Sequence of Events 1 TCVs and Bypass Valves Start Open 0.1 0.1 2 MSIV Closure Initiated by Low Steam Line Pressure 17.9 16.8 3 MSIVs Fully Closed 21.9 20.8 4 Peak Neutron Flux 21.8 20.7 5 High Pressure ATWS Setpoint 23.5 22.5 6 Recirculation Pumps Trip 24.2 23.1 7 Opening of the First Relief Valve 24.2 23.1 8 Peak Heat Flux 24.6 23.4 9 Peak Vessel Pressure 29.7 28.4 10 FW Reduction Initiated 48.7 48.7 11 SLCS Pumps Start (1) 143.5 142.5 12 RHR Cooling Initiated 660 660 13 Hot Shutdown Boron Weight Achieved and Initiate Level Increase 1029 1029 14 Hot Shutdown Achieved (Neutron Flux remains < 0.1%) 1217 1323 15 Peak Suppression Pool Temperature 10378 5331 Notes:
- 1. SLCS injection is the later time of either 1) two minutes after the high-pressure RPT or 2) when the suppression pool temperature reaches the BIIT.
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Table 2.8-11 Parameters Used For Transient Analysis Rated Thermal Power (MWt) 3514 3951 Analysis Power (% EPU Rated) 100 / 102 ' 100 / 102 '
Analysis Dome Pressure (psia) 1050/1068 2 1050/1068 2 Rated Core Flow (Mlbm/hr) 102.5 102.5 Rated Power Core Flow Range (% Rated) 82.8 - 110 99 - 110 Normal FW Temperature (0 F) 381.5 381.5 Change in FW Temperature (Maximum) (worst 100 100 single failure of FW heaters) (AT 'F)
Combined Number of SRVs and SSVs assumed 13 14 3 in the analysis Notes:
- 1. GEMINI MCPR analyses at 100%, non-MCPR analyses at 102%.
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Table 2.8-12 Transient Analysis MCPR Results Load Rejection with Bypass Failure 0.26 0.27 Turbine Trip with Bypass Failure 0.25 0.26 FW Controller Failure Max Demand 0.24 0.23 Inadvertent HPCI with Level 8 Trip Note 1 0.24 Loss of FW Heating 0.13 0.12 Rod Withdrawal Error 0.21 0.27 Slow Recirculation Increase Note 2 MCPRf Fast Recirculation Increase Note 1 0.14 3/0.1778 4 Load Rejection With Bypass Note 1 0.21 MSIV Closure All Valves Note 1 0.00 MSIV Closure 1 Valve Note 1 0.09 Notes:
- 2. Event not analyzed at CLTP.
- 3. Result for Event with M/G set power to recirculation pump motors. The Fast Recirculation Increase event initiates from 54.9% EPU Rated power and 38% rated flow.
- 4. Result for Event with ASD power to recirculation pump motors. The Fast Recirculation Increase event initiates from 54.9% EPU Rated power and 38% rated flow.
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Figure 2.8-1 Power of Peak Bundle versus Cycle Exposure 8.0 7.5
- 7.0
. 6.5 0
10.
5 6.0
= 5.5 E
25.0 4.5 4.0 0 2 4 6 8 10 12 14 16 18 Cycle Exposure (GWD/ST) 2-448
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Figure 2.8-2 Coolant Flow for Peak Bundle versus Cycle Exposure 13- T '
12 L--- --- ---------- --------------------- 1------ -----
olO*- .-.....--
8 h------- ------------------------------.
t----- ------ 4--
0.
, I I I 7 - - - - - - -- - - - - - -- _0 - - - -i-------- - - --------------I- .
I-0 Plant A Cycle 18 U Plant A Cycle 19
, lBy 9 Plant B Cycle 10 6 - Plant C Cycle 7 --- Plant C Cycle 8 Plant D Cycle 13 -- Plant E Cycle 11 5 __ ___Plant F -- PBAPS EPU T0200 5 1 1 1 0 2 4 6 8 10 12 14 16 18 Cycle Exposure (GWD/ST) 2-449
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Figure 2.8-3 Exit Void Fraction for Peak Power Bundle versus Cycle Exposure 0.90
&.0.85 0
0.
0.
- 0.80 U.
70.7 5
- Plant A Cycle 18 E- Plant A Cycle 19
> -*- Plant B Cycle 9 Plant B Cycle 10 x
-4.- Plant C Cycle 7 --- Plant C Cycle 8 Plant D Cycle 13 -Plant E Cycle 11 0.70 -Plant F 4 PBAPS EPU T0200 I I i I Il 0 2 4 6 8 10 12 14 16 18 Cycle Exposure (GWD/ST) 2-450
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Figure 2.8-4 Maximum Channel Exit Void Fraction versus Cycle Exposure 0.90 r- ,-i --
0.88 - ------- 4 ---------- - ---------- - i----
--- I----1------t--
" 0.86 _.-- ----- -----------
0
,0.84 ------ -- ,--
U-
~0.82 - - -- ------ 4 -- - ---. I- 4--------------.-- - -------- j----
~0.80-------...I--.........--÷......... .1.- -I-,--
= 0.78
"~...
(* 07J 0.7 /I I4 I - - ----..-- --- -
x __ __ __
18-Cycle _ , _ i_ --- Planti_A Cycle 19
_ _ I _
-A- Plant B Cycle 9 ---- Plant B Cycle 10 0.72 Plant C Cycle 7 -.- Plant C Cycle 8
--- Plant D Cycle13 - Plant E Cycle 11 0.70 -Plant F -- PBAPS EPU T0200 0 2 4 6 8 10 12 14 16 18 Cycle Exposure (GWD/ST) 2-451
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Figure 2.8-5 Core Average Exit Void Fraction versus Cycle Exposure 2-452
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Figure 2.8-6 Peak LHGR versus Cycle Exposure 16 14 12
.10 0 8 6
4
-4Plant A Cycle 18 = Plant A Cycle 19
-A- Plant B Cycle 9
- Plant B Cycle 10 2 - Plant C Cycle 7 -- Plant C Cycle 8 NEPlant D Cycle 13 Plant E Cycle 11
-Plant F -.-- PBAPS EPU T0200 0
0 2 4 6 8 10 12 14 16 18 Cycle Exposure (GWDIST) 2-453
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Figure 2.8-7 Dimensionless Bundle Power at BOC (200 MWd/ST) 2-454
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Figure 2.8-8 Dimensionless Bundle Power at MOC (9000 MWd/ST) 2-455
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Figure 2.8-9 Dimensionless Bundle Power at EOC (17671 MWd/ST) 2-456
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Figure 2.8-10 Bundle Operating LHGR (kW/ft) at BOC (200 MWd/ST) 2-457
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Figure 2.8-11 Bundle Operating LHGR (kw/ft) at MOC (9000 MWd/ST) 2-458
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Figure 2.8-12 Bundle Operating LHGR (kw/ft) at EOC (17671 MWd/ST) 2-459
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Figure 2.8-13 Bundle Operating MCPR at BOC (200 MWd/ST) 2-460
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Figure 2.8-14 Bundle Operating MCPR at MOC (9000 MWd/ST) 2-461
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Figure 2.8-15 Bundle Operating MCPR at EOC (17671 MWd/ST) 2-462
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Figure 2.8-16 Dimensionless Bundle Power at 11000 MWd/ST (peak dimensionless bundle power point) 2-463
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Figure 2.8-17 Bundle Operating LHGR (kW/ft) at 15250 MWd/ST (peak MFLPD* point)
- MFLPD - Maximum Fraction of Limiting Power Density 2-464
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Figure 2.8-18 Bundle Operating MCPR at 15250 MWd/ST (peak MFLCPR point) 2-465
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Figure 2.8-19 Bundle Average Void History vs. MCPR 1.8 1.75 1.7
---.- N- --
1.65 ---- r-
- ------ --- ---------- --- ------ ---T - --- --- ------ --- ---------- --- ------
.. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .." .. .. .. .. .. .. .. .. . 1 .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .
1.6 nx 0.
Q) 1.55
"= ------------------- Ui ------------ --------- -------------------------
- -------------- i -- ------------ ---------------------a -
i~~~~ ~ ~ ~ ~ ~~~~~~~~~~~~~~~- ..........--......................--...................
......--...- T...............................
1.5 1.45 1.44 0.45 0.5 0.55 0.6 0.66 BundleAverage Void History 2-466
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Figure 2.8-20 Illustration of OPRM Trip-Enabled Region Core Flow (MIb/hr) 0 10 20 30 40 50 60 70 80 90 100 110 120 120 4500 Power Flo W 110 00 EU 100 3t04ý - 32931M E0ý0/ YO u 100%OLTP MELAoudy 00%Core Flow 102. 5MIbfhr 4000 100 3951 M.---.
90 3500 80 3000o 70 1 i Enbe Rgo 2500 60 --------
NatralCirulaio 0 50 2000 i
.5 0 40 1500 U liMinimm Pum Spee111s 30 1000 20 5O00 10 0
0 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 Core Flow (%)
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Figure 2.8-21 BSP Regions at Nominal Feedwater Temperature Core Flow (MIb/hr) 0 10 20 30 40 50 60 70 80 90 100 110 120 120
- I I I I II1 I I Power Flow a . 4500 A: 49.4% 31.3% 100%EPU = 3951 MWt 110 B: 54.9% 38.0% 100%CLTP = 3514MWt 7 I_
C: 100.0% 990% - . a D: 100.0% 100.00/ 100%OLTP MW ,3293 MELLLA Bounda , I I D I 4000 100 E: 100.0% 110. 0%/ 100% Core Flow 102.5 Mlb/hr--------f ~------f ----- -- -3951 M1W 90 ' * ---
I -*-- -t . . .. . --- . ..,- a --I a ..----- ! -- .. . I I---- ,.. i a a, -,-
I .--...---- a i.h -----
+, I --.. .,.I ..4 - - I - -
3500 a a---a-a --- 1--
80 a a ! I ! 4! I ! i I i a --- I I I 3000o In 70 . . .. I ,, ..- a. a! I C.)
I I I I I I I I l ^ i- Ii II --
--IIII------ I .. Ii -
0 a i i a i i iB a . I, I I i I i i i I I i 2500 60 a.
e S! ! ! ! ! i *! ! a a II ! /
50 2000 E I,-
o 40 Natural Circulation ..... T .... -T.---4 1500 U
30 E;-I------ j 44 i, Ia/i - Minimu Pum Speed I ! a a I I I ! I 1000 20
.- a i i-- i im I.--.--.L...-..---.--.----.--.--.J........a.
i iu I a i --- a a
. .J 500 10 0
0 0 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 Core Flow (%)
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Figure 2.8-22 BSP Regions at Minimum Feedwater Temperature Core Flow (MIb/hr) 0 10 20 30 40 50 60 70 80 90 100 110 120 I r4 110 A
A:
B 49.4%
Power 54.9%
31.3%
Flow 38.0%
100% EPU 100%CLTP
= 3951 M Wt
= 3514 MWt 1 . ...
I
.. _*. . . ._1 _ .,1 ..
I I
.... .. *. ...... .... .* . ... .. 4500 C 100.0% 99.0/ l00%OLTP = 3293 MWt D: 1000% 110010/0 OI MELLLA Boundary 4000 100 E: 1000 0 100%CoreFlow = 102.5 Mlb/hr -- -t 90 I i I I 3500 80 -- -- t - - - ---
Ii * !I - 3000 70 i.
-- -- --- - I1- - 1-
, / i i , ---- , ', , ', --------
i - 2500 IL 0
60 a.
..... . 4 . 4. B - 4. -i..-.
50 2000 I-0 40 !---------- i -
C.)
Natural Circulation I , + 1500
-b B1I I ,
3U 1000 20 - - , t 10 -.-... . . .. --- Minimum Pu! Speei i-l- l 7 -7 -, 500 I- -I - --- - -- --
0 0 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 Core Flow (%)
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Figure 2.8-23 Response to MSIV Closure with Flux Scram (102% EPU power, 99% core flow, and 1068 psia initial dome pressure)
M6 '3m Igoe Go l 10 is0 2 0 so to 70 t0 00 ia -0 30 40 so as 70a s rTM,(S4o Timh(%"
C I
U I* 0 i
U
-J do 10 OG 10 0 10 60 70 40 1 0 0 1 1h-O Tiw(904 2-470
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Figure 2.8-24 Loss of Feedwater Flow 600.0 550.0 _ _ __
1 000 450.0 - --------
400.0 Top of Aqtlve Fuel 350.0 0 400 800 1200 1600 2000 2400 2800 3200 3600 Time (sec) 2-471
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Figure 2.8-25 EPU MELLLA BOC MSIVC (Short-Term)
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Figure 2.8-26 EPU MELLLA BOC MSIVC (Long-Term) 2-473
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Figure 2.8-27 EPU MELLLA BOC MSIVC (Long-Term)
))
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Figure 2.8-28 EPU MELLLA BOC MSIVC (Long-Term)
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))
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Figure 2.8-29 EPU MELLLA BOC PRFO (Short-Term)
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Figure 2.8-30 EPU MELLLA BOC PRFO (Long-Term)
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Figure 2.8-31 EPU MELLLA BOC PRFO (Long-Term)
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Figure 2.8-32 EPU MELLLA BOC PRFO (Long-Term) 2-479
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Figure 2.8-33 EPU MELLLA EOC MSIVC (Short-Term)
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Figure 2.8-34 EPU MELLLA EOC MSIVC (Long-Term)
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Figure 2.8-35 EPU MELLLA EOC MSIVC (Long-Term)
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Figure 2.8-36 EPU MELLLA EOC MSIVC (Long-Term)
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Figure 2.8-37 EPU MELLLA EOC PRFO (Short-Term)
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Figure 2.8-38 EPU MELLLA EOC PRFO (Long-Term) 2-485
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Figure 2.8-39 EPU MELLLA EOC PRFO (Long-Term)
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Figure 2.8-40 EPU MELLLA EOC PRFO (Long-Term)
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