View source for NL-15-169, Browns Ferry, Units 1, 2, and 3, Template of NRC Review Standard for Extended Power Uprates, RS-001 Safety Evaluation Template GDC Markup (with Redline/Strike Out)

{{Adams
| number = ML15282A260
| issue date = 09/21/2015
| title = Browns Ferry, Units 1, 2, and 3, Template of NRC Review Standard for Extended Power Uprates, RS-001 Safety Evaluation Template GDC Markup (with Redline/Strike Out)
| author name = 
| author affiliation = Tennessee Valley Authority
| addressee name = 
| addressee affiliation = NRC/NRR
| docket = 05000259, 05000260, 05000296
| license number = DPR-033, DPR-052, DPR-068
| contact person = 
| case reference number = CNL-15-169
| package number = ML15282A154
| document type = Safety Evaluation
| page count = 88
}}

=Text=
{{#Wiki_filter:ATTACHMENT 48 RS-OO1 SE Template GDC Markup (with redlinelstrike out)
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. xxx TO FACILITY OPERATING LICENSE NO. XXX-xx[LICENSEE]
[PLANT NAME, UNIT NO.]DOCKET NO. 5O-XXX NOTE: This document uses unqeHeading styles for each SE section so that a table of contents can be generated automatically.
The Heading styles appear on the Word ribbon (Home tab) in the"Styles" gallery (to see Heading 4 and Heading 5, click on the drop-down arrow next to the gallery).SE Sectin 1.1 1.2 1.1.1 1.1.1.1 1.1.1.1.1 2.0 Apply Style Called...Heading I Heading 2 Heading 3 Heading 4 Heading 5 Heading 1, etc.If you need to add a new numbered SE section, type the text only and then apply the appropriate Heading style fmom the gallery. When you generate a new TOO, the new SE section will be added.To update the TOC for page numbering or if you've added new SE sections (and applied appropriate Heading styles): 1. Click anywhere in the TOO which will appear greyed out.2. Then right-click and select Update Field... Update entire table.1 
[PLANT NAME, UNIT NO.]SAFETY EVALUATION FOR EXTENDED POWER UPRATE TABLE OF CONTENTS 

==1.0 INTRODUCTION==

............................................................................ 1.1 A~lcto ....................................................... 
............................ 1.2 Backgqround.................................................................................. 1.3 Licensee's Approach........................................................................ 1.4 Plant Modifications............................................................................ 1.5 Method of NRC Staff Review............................................................... 2.0 EVALUATION 
............................................................................... 2.1 Materials and Chemical Engineering 
......................................-
3-2.1.1 Reactor Vessel Material Surveillance Program ........................................... 2.1.2 Pressure-Temperature Limits and Upper-Shelf Energy .................................. 2.1.3 Reactor Internal and Core Support Materials 
............................................. 2.1.4 Reactor Coolant Pressure Boundary Materials 
........................................... 2.1.5 Protective Coating Systems (Paints) -Organic Materials................................ 2.1.6 Flow-Accelerated Corrosion................................................................ 2.1.7 Reactor Water Cleanup System............................................................ 2.1.8 [Additional Review Areas (Materials and Chemical Engineering)]....................... 2.2 Mechanical and Civil Engineeringq.......................................................... 2.2.1 Pipe Rupture Locations and Associated Dynamic Effects ............................... 2.2.2 Pressure-Retaining Components and Component Supports...........................-
10-2.2.3 Reactor Pressure Vessel Internals and Core Supports................................. 2.2.4 Safety-Related Valves and Pumps....................................................... 2.2.5 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment 
.................................................................................. 2.2.6 [Additional Review Areas (Mechanical and Civil Engineering)]......................... 2.3 Electrical Engineering 
.................................................-
16 -2.3.1 Environmental Qualification of Electrical Equipment 
.................................... 2.3.2 Offsite Power System...................................................................... 2.3.3 AC Onsite Power System ................................................................. 2 2.3.4 DC Onsite Power System................................................................. 2.3.5 Station Blackout ........................................................................... 2.3.6 [Additional Review Areas (Electrical Engineering)]...................................... 2.4 Instrumentation and Controls .................. 
.......................................... 2.4.1 Reactor Protection, Safety Features Actuation, and Control Systems................. 2.4.2 [Additional ReviewAreas (Instrumentation and Controls)].............................. 2.5 Plant Systems.............................................................................. 2.5.1 InternaIHazards 
........................................................................... 2.5.1.1 Flooding ................................................................................ 2.5.1.1.1 Flood Protection 
....................................................................... 2.5.1.1.2 Equipment and FlocrDrains 
.......................................................... 2.5.1.1.3 Circulating Water System ............................................................. 2.5.1.2 Missile Protection...................................................................... 2.5.1.2.1 Internally Generated Missiles......................................................... 2.5.1.2.2 Turbine Generator..................................................................... 2.5.1.3 Pipe Failures........................................................................... 2.5.1.4 Fire Protection 
......................................................................... 2.5.2 Fission Product Control.................................................................... 2.5.2.1 Fission Product Control Systems and Structures 
................................... 2.5.2.2 Main Condenser Evacuation System ................................................ 2.5.2.3 Turbine Gland Sealing System ....................................................... 2.5.2.4 Main Steam Isolation Valve Leakage Control System.............................. 2.5.3 Component Cooling and Decay Heat Removal ......................................... 2.5.3.1 Spent Fuel Pool Cooling and Cleanup System...................................... 2.5.3.2 Station Service Water System ........................................................ 2.5.3.3 Reactor Auxiliary Cooling Water Systems ........................................... 2.5.3.4 Ultimate Heat Sink..................................................................... 2.5.4 Balance-of-Plant Systems................................................................. 2.5.4.1 Main Steam ............................................................................ 2.5.4.2 Main Condenser 
....................................................................... 2.5.4.3 Turbine Bypass........................................................................ 2.5.4.4 Condensate and Feedwater 
............................................................ 2.5.5 Waste Management Systems ............................................................ 2.5.5.1 Gaseous Waste Management Systems ............................................. 2.5.5.2 Liquid Waste Management Systems................................................. 3 iii 2.5.5.3 Solid Waste Management Systems.................................................. 2.5.6 AdditionalIConsiderations................................................................. 2.5.6.1 Emergency Diesel Engine Fuel Oil Storage and Transfer System................. 2.5.6.2 Light Load Handling System (Related to Refueling)................................ 2.5.7 [Additional Review Areas (Plant Systems)] 
.............................................. 2.6 Containment Review Considerations......................-39 
-2.6.1 Primary Containment Functional Design ................................................. 2.6.2 SubcompartmnentAnalyses 
............................................................... 2.6.3 Mass and Energy Release ................................................................ 2.6.3.1 Mass and Energy Release Analysis for Postulated Loss of Coolant............... 2.6.4 Combustible Gas Control in Containment................................................ 2.6.5 Containment Heat Removal....:........................................................... 2.6.6 Secondary Containment Functional Design ............................................. 2.6.7 [Additional Review Areas (Containment Review Considerations)] 
..................... 2.7 Habitability.
Niltration, and Ventilation.................................................... 2.7.1 Control Room Habitability System........................................................ 2.7.2 Engineered Safety Feature Atmosphere Cleanup ...................................... 2.7.3 Control Room Area Ventilation System .................................................. 2.7.4 Spent Fuel Pool Area Ventilation System................................................ 2.7.5 Auxiliary and Radwaste Area and Turbine Areas Ventilation Systems ................ 2.7.6 Engineered Safety Feature Ventilation System ......................................... 2.7.7 [Additional Review Areas (Habitability, Filtration, and Ventilation)] 
.................... 2.8 Reactor Systems............................................ 2.8.1 Fuel System Design ....................................................................... 2.8.2 Nuclear Design............................................................................. 2.8.3 Thermal and Hydraulic Design............................................................ 2.8.4 Emergency Systems ...................................................................... 2.8.4.1 Functional Design of Control Rod Drive System .................................... 2.8.4.2 Overpressure Protection During Power Operation 
.................................. 2.8.4.3 Reactor Core Isolation Cooling System.............................................. 2.8.4.4 Residual Heat Removal System...................................................... 2.8.4.5 Standby Liquid Control System....................................................... 2.8.5 Accident and Transient Analyses......................................................... 4 E~KJ a-.- ~puI-.iv 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 ............
,......................................................-5 2.8.5.2 Decrease in Heat Removal by the Secondary System ............................. 2.8.5.2.1 Loss of External Load; Turbine Trip; Loss of Condenser Vacuum;Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed)............................................................ 2.8.5.2.2 Loss of Nonemergency AC Power to the Station Auxiliaries 
....................... 2.8.5.2.3 Loss of Normal Feedwater Flow......................................................  -2.8.5.3 Decrease in Reactor Coolant System Flow.......................................... 2.8.5.3.1 Loss of Forced Reactor Coolant Flow................................................ 2.8.5.3.2 Reactor Recirculation Pump Rotor Seizure and Reactor Recirculation Pump Shaft Break ..................................................................... 2.8.5.4 Reactivity and Power Distribution Anomalies........................................ 2.8.5.4.1 Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition......................................................... 2.8.5.4.2 Uncontrolled Control Rod Assembly Withdrawal at Power......................... 2.8.5.4.3 Startup of a Recirculation Loop at an Incorrect Temperature and Flow Controller Malfunction Causing an Increase in Core Flow Rate.................... 2.8.5.4.4 Spectrum of Rod Drop Accidents 
..................................................... 2.8.5.5 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 Pressure Relief Valve .................................... 2.8.5.6.2 Emergency Core Cooling System and Loss-of-Coolant Accidents 
................ 2.8.5.7 Anticipated Transients Without Scrams ..............................................-
73 -2.8.6 FuelStorage................................................................ 
'................ 2.8.6.1 New Fuel Storage...................................................................... 2.8.6.2 Spent Fuel Storage.................................................................... 2.8.7 [Additional Review Areas (Reactor Systems)] 
........................................... 2.9 Source Terms and Radiologqical Consequences Analyses ............................. 2.9.1 Source Terms for Radwaste Systems Analyses ........................................ 2.9.2 Radiological Consequences of Control Rod Drop Accident ............................ 2.9.3 Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment 
................................................... 2.9.4 Radiological Consequences of Main Steamline Failure Outside Containment................................................................................ 2.9.5 Radiological Consequences of a Design-Basis Loss-of-Coolant Accident............ 5 V 2.9.6 Radiological Consequences of Fuel Handling Accidents 
............................... 2.9.7 Radiological Consequences of Spent Fuel Cask Drop Accidents...................... 2.9.8 [Additional Review Areas (Source Terms and Radiological Consequences Analyses)].................................................................................. 2.10 Health Physics............................................................................. 2.10.1 Occupational and Public Radiation Doses............................................... 2.10.2 [Additional Review Areas (Health Physics)] 
............................................. 2.11 Human Performance 
...................................................................... 2.11.1 Human Factors............................................................................. 2.11.2 [Additional Review Areas (Human Performance)] 
....................................... 2.12 Power Ascension and Testing Plan ...................................................... 2.12.1 Approach to EPU Power Level and Test Plan........................................... 2.12.2 [Additional Review Areas (Power Ascension and Testing Plan)]....................... 2.13 Risk Evaluation 
............................................................................ 2.13.1 Risk Evaluation of EPU.................................................................... 2.13.2 [Additional Review Areas (Risk Evaluation)]............................................. 3.0 FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATION CHANGES ................................................................................. 4.0 REGULATORY COMMITMENTS 
........................................................ 5.0 RECOMMENDED AREAS FOR INSPECTION..........................................  

==6.0 STATE CONSULTATION==
................................................................. 7.0 ENVIRONMENTAL CONSIDERATION 
.... .............................................  

==8.0 CONCLUSION==
.............................................................................  

==9.0 REFERENCES==
.............................................................................
88-.6 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. xxx TO FACILITY OPERATING LICENSE NO. XXX-xx[LICENSE El[PLANT NAME, UNIT NO.1 DOCKET NO. 50-xxx 

==1.0 INTRODUCTION==

1.1 AApllication.
By application dated , as supplemented by letters dated_________________________________________, [Licensee] 
([Licensee Abbreviation], the licensee) requested changes to Facility Operating License No. NPF-029 and the Technical Specifications (TSs) for [Plant Name, Unit No.] ([Plant Abbreviation)).
Portions of the letters dated contain sensitive unclassified non-safeguards information and, accordingly, have been withheld from public disclosure.
The supplemental letters dated , provided additional clarifying information that did not expand the scope of the initial application and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal/Register on [date] (XX FR XXXX).The proposed changes would increase the maximum steady-state reactor tore power level from[current licensed power level] megawatts thermal (MWt) to [power level proposed by the licensee]
MWt, which is an increase of approximately 
[##] percent. The proposed increase in power level is considered an extended power uprate (EPU).1.2 Backgqround
[Plant Name] is a boiling-water reactor (BWR) plant of the BWRI[#] design with a Mark-[#]containment. 
[Plant Name] has the following special features/unique designs:[Insert any special features/unique designs]The NRC originally licensed [Plant Name] on [date] for operation at [original licensed power level] MWt. [By Amendment No. [W#] dated [ ], the NRC granted a power uprate to[Plant Name] of [##] percent, allowing the plant to be operated at [current licensed power level] MWt.] Therefore, the proposed EPU would result in an increase of approximately 7 
[##] percent over the original licensed power level [and [##] percent over the current licensed power level] for [Plant Name].] '1.3 L'icensee's Approach The licensee's application for the proposed EPU follows the guidance in the Office of Nuclear Reactor Regulation's (NRR's) Review Standard (RS)-OO1, "Review Standard for Extended Power Uprates," to the extent that the review standard is consistent with the design basis of the plant. Where differences exist between the plant-specific design basis and RS-OO1, the licensee described the differences and provided evaluations consistent with the design basis of the plant. The licensee also used [Identify topical reports or other documents used by the licensee for guidance related to the scope of the proposed EPU; NRC staff approvals, ranges of applicability, any limitations/restrictions associated with the documents; and consistency of the licensee's application with the ranges of applicability and limitations/restrictions.
The discussion in this section is to cover topical reports and other documents referenced for the overall power uprate process. It is not intended to cover topical reports and other documents for specific methods of analyses.
Topical reports and other documents referenced for specific methods of analyses are to be covered in the applicable technical evaluation section of this safety evaluation].
Insert this sentence if the licensee is planning to implement the EPU in one stage.[The licensee plans to implement the EPU in one step. The licensee plans to make the modifications necessary to implement the EPU during the refueling outage in[season year (e.g., fall 2003)]. Subsequently, the plant will be operated at [##] MWt starting in Cycle [##~].]Insert this paragraph if the licensee is planning to implement the EPU in stages:[The licensee plans to implement the EPU in E#J steps of [## and ##] percent. The licensee plans to make modifications necessary to implement the first step during the refueling outage in Eseason year (e.g., fall 2003)]. Subsequently, the plant will be operated at [It#] MWt during Cycle [fl]. The remainder of the modifications will be completed during the refueling outage in [season year (e.g., fall 2003)], with subsequent operation at [##] MWt starting in Cycle [##].]1.4 Plant Modifications The licensee has determined that several plant modifications are necessary to implement the proposed EPU. The following is a list of these modifications and the licensee's proposed schedule for completing them.[Provide a list of plant modifications.]
The NRC staffs evaluation of the licensee's proposed plant modifications is provided in Section 2.0 of this safety evaluation.
a 1.5 Method of NRC Staff Review The NRC staff reviewed the licensee's application to ensure that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) activities proposed will be conducted in compliance with the Commissjon's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public. The purpose of the NRC staff's review is to evaluate the licensee's assessment of the impact of the proposed EPU on design-basis analyses.
The NRC staff evaluated the licensee's application and supplements.
The NRC staff also evaluated 
[Include additional review items, as necessary (e.g., audits of certain information at the plant and vendor sites, and independent analyses), for areas where such analyses were deemed appropriate by the NRC staff].In areas where the licensee and its contractors used NRC-approved or widely accepted methods in performing analyses related to the proposed EPU, the NRC staff reviewed relevant material to ensure that the licensee/contractor used the methods consistent with the limitations and restrictions placed on the methods. In addition, the NRC staff considered the eaffects of the changes in plant operating conditions on the use of these methods to ensure that the methods are appropriate for use at the proposed EPU conditions.
Details of the NRC staff's review are provided in Section 2.0 of this safety evaluation.
Audits of analyses supporting the EPU were conducted in relation to the following topics:[Provide a list of areas for which audits were performed.]
The results of the audits are discussed in Section 2.0 of this safety evaluation.
Independent NRC staff calculations were performed in relation to the following topics:[Provide a list of areas for which independent NRC staff calculations were performed.]
The results of the calculations are discussed in Section 2.0 of this safety evaluation.

==2.0 EVALUATION==
2.1 Materials and Chemical Engqineering 2.1.1 Reactor Vessel Material Surveillance Program Regulatory Evaluation The reactor vessel material surveillance program provides a means for determining and monitoring the fracture toughness of the reactor vessel beltline materials to support analyses for ensuring the structural integrity of the ferritic components of the reactor vessel. The NRC staff's review primarily focused on the effects of the proposed EPU on the licensee's reactor vessel surveillance capsule withdrawal schedule.
The NRC's acceptance criteria are based on (1)-.draft General Design Criterion (GDC)-9, insofar as it requires that the reactor coolant pressure boundary (RCPS) be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage; Critcrizon (ODO) 11, 9 fratte (2) draft GDC-33, insofar as it requires that the RCPB be capable of accommodating without rupture, onfwijrnited allowance for energy absorption through plastic deformation, the stati6 and dynamic lbads imeposed on any boundary component as a result of any inadvertent and sudden release of energy to the coolant;final -GDC-31, insofar as it requires that the RCPB be designed with margin sufficient to assure that, under specified conditions, it wilt behave in a nonbrittle manner and the probability of a rapidly propagating fracture is minimized; (34) 10 CFR Part 50, Appendix H, which provides for monitoring changes in the fracture toughness properties of materials in the reactor vessel beltline region; and (45) 10 CFR 50.60, which requires compliance with the requirements of 10 CFR Part 50, Appendix H. Specific review criteria are contained in Standard Review Plan (SRP) Section 5.3.1 and other guidance provided in Matrix 1 of RS-001.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed EPU on the reactor vessel surveillance withdrawal schedule and concludes that the licensee has adequately addressed changes in neutron fluence and their effects on the schedule.
The NRC staff further concludes that the reactor vessel capsule withdrawal schedule is appropriate to ensure that the material surveillance program will continue to meet the requirements of 10 CFR Part 50, Appendix H, and 10 CFR 50.60, and will provide the licensee with information to ensure continued compliance with draft GDCs-9 and 33, and finalGQ44ad GDC-31 in this respect following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the reactor vessel material surveillance program.2.1.2 Pressure-Temperature Limits and Upper-Shelf Energy Reoulatorv Evaluation Pressure-temrperature (P-T) limits are established to ensure the structural integrity of the ferritic components of the RCPB during any condition of normal operation, including anticipated operational occurrences and hydrostatic tests. The NRC staff's review of P-T limits covered the P-T limits methodology and the calculations for the number of effective full power years specified for the proposed EPU, considering neutron embrittlement effects and using linear elastic fracture mechanics.
The NRC's acceptance criteria for P-T limits are based on (1) draft GDC-9, insofar as it requires that the RCPB be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage; (4-GQ-+an-ekernly~o-p~oebi 
....o rapidly propgatng racur, (2) final 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 nonbrittle manner and the probability of a rapidly propagating fracture is minimized; (3) 10 CFR Part 50, Appendix G, which specifies fracture toughness requirements for ferritic components of the RCPB; and (4) 10 CFR 50.60, which requires compliance with the requirements of 10 CFR Part 50, Appendix G. Specific review criteria are contained in SRP Section 5.3.2 and other guidance provided in Matrix I of RS-001.10 Technical Evaluation ....[Insert technical evaluation.
The technical evaluation should (1) ciearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a ciear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed EPU on the P-T limits for the plant and concludes that the licensee has adequately addressed changes in neutron fluence and their effects on the P-T limits. The NRC staff further concludes that the licensee has demonstrated the validity of the proposed P-T limits for operation under the proposed EPU conditions.
Based on this, the NRC staff concludes that the proposed P-T limits will continue to meet the requirements of 10 CFR Part 50, Appendix G, and 10 CFR 50.60 and will enable the licensee to comply with draft GDC-9,GDG-4 and final GDC-31 in this respect following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the proposed P-T limits.2.1.3 Reactor Internal and Core Support Materials Regqulatory Evaluation The reactor internals and core supports include structures, systems, and components (SSCs)that perform safety functions or whose failure could affect safety functions performed by other SSCs. These safety functions include reactivity monitoring and control, core cooling, and fission product confinement (within both the fuel cladding and the reactor coolant system (RCS)). The NRC staffs review covered the materials' specifications and mechanical properties, welds, weld controls, nondestructive examination procedures, corrosion resistance, and susceptibility to degradation.
The NRC's acceptance criteria for reactor internal and core support materials are based on draft GDC-1 GDC-4-and 10 CFR 50.55a for material specifications, controls on welding, and inspection of reactor internals and core supports.
Specific review criteria are contained in SRP Section 4.5.2 and Boiling Water Reactor Vessel and Internals Project (BWRVIP)-26.
Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conciusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed EPU on the susceptibility of reactor internal and core support materials to known degradation mechanisms and concludes that the licensee has identified appropriate degradation management programs to address the effects of changes in operating temperature and neutron fluence on the integrity of reactor internal and core support materials.
The NRC staff further concludes that the licensee has demonstrated that the reactor internal and core support materials will continue to be acceptable and will continue to meet the requirements of GDQ-4-draft GDC-1 and 10 CFR 50.55a with respect to material specifications, welding controls, and inspection following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to reactor internal and core support materials.
11 2.1.4 Reactor Coolant Pressure Boundary Materials Regulatory Evaluation The RCPB defines the boundary of systems and components containing the high-pressure fluids produced in the reactor. The NRC staffs review of RCPB materials covered their specifications, compatibility with the reactor coolant, fabrication and processing, susceptibility to degradation, and degradation managembnt programs.
The NRC's acceptance criteria for RCPB materials are based on (1) 10 CFR 50.55a and draft GDC-1,GQG=-I-, insofar as they require that those systems and components which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed; 
.. as .....reqir SSr imp etesafey-edeinedi-febr 
.............
edronkutdtstd d pete~ted~-
ef n...........at.d with-nF~maJ-eperatien-,mainteilRenee 7
(2) draft GDC-2, insofar as those systems and components which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects; (3) draft GDC-9 insofar, as it requires that the RCPB be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage; -GQ- s~eofaf-asit-fequirestatte rapidl-pepag-atinj-raeturej-(4,3) final 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 nonbrittle manner and the probability of a rapidly propagating fracture is minimized; and (45) 10 CFR Part 50, Appendix G, which specifies fracture toughness requirements for ferritic components of the RCPB. Specific review criteria are contained in SRP Section 5.2.3 and other guidance provided in Matrix 1 of RS-O01. Additional review guidance for primary water stress-corrosion cracking (PWSCC) of dissimilar metal welds and associated inspection programs is contained in Generic Letter (GL) 97-01, Information Notice (IN) 00-17, Bulletin (BL) 01-01, BL 02-01, and BL 02-02, Additional review guidance for thermal embrittlement of cast austenitic stain~less steel components is contained in a letter from C. Grimes, NRC, to D. Walters, Nuclear Energy Institute (NEI), dated May 19, 2000.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed EPU on the susceptibility of RCPB materials to known degradation mechanisms and concludes that the licen~see has identified appropriate degradation management programs to address the effects of changes in system operating temperature on the integrity of RCPB materials.
The NRC staff further concludes that the licensee has demonstrated that the RCPB materials will continue to 12 be acceptable following implementation of the proposed EPU and will continue to meet the requirements of draft GDCs-1, 2, and 9, CDC 1, GCO l, GOC 11,,final GDC-31, 10OCFR Part 50, Appendix G, and 10 CFR 50.55a. Therefore, the NRC staff finds the proposed EPU acceptable with respect to RCPB materials.
2.1.6 Protective Coating Systems (Paints) -Organic Materials Regulatory Evaluation Protective coating systems (paints) provide a means for protecting the surfaces of facilities and equipment from corrosion and contamination from radionuclides and also provide wear protection during plant operation and maintenance activities.
The NRC staff's review covered protective coating systems used inside the containment for their suitability for and stability under design-basis loss-of-coolant accident (DBLOCA) conditions, considering radiation and chemical effects. The NRC's acceptance criteria for protective coating systems are based on (1) 10 CFR Part 50, Appendix B, which states quality assurance requirements for the design, fabrication, and construction of safety-related SSCs and (2) Regulatory Guide 1.54, Revision 1, for guidance on application and performance monitoring of coatings in nuclear power plants.Specific review criteria are contained in SRP Section 6.1.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed EPU on protective coating systems and concludes that the licensee has appropriately addressed the impact of changes in conditions following a DELOCA and their effects on the protective coatings.The NRC staff further concludes that the licensee has demonstrated that the protective coatings will continue to be acceptable following implementation of the proposed EPU and will continue to meet the requirements of 10 CFR Part 50, Appendix B. Therefore, the NRC staff finds the proposed EPU acceptable with respect to protective coatings systems.2.1.6 Flow-Accelerated Corrosion Regqulatory Evaluation Flow-accelerated corrosion (FAC) is a corrosion mechanism occurring in carbon steel components exposed to flowing single- or two-phase water. Components made from stainless steel are immune to FAG, and FAC is significantly reduced in components containing small amounts of chromium or molybdenum.
The rates of material loss due to FAC depend on velocity of flow, fluid temperature, steam quality, oxygen content, and pH. During plant operation, control of these parameters is limited and the optimum conditior~s for minimizing FAC effects, in most cases, cannot be achieved.
Loss of material by FAC will, therefore, occur. The N RC staff has reviewed the effects of the proposed EPU on FAC and the adequacy of the licensee's FAC program to predict the rate of loss so that repair or replacement of damaged components could be made before they reach critical thickness.
The licensee's FAG program is based on NUREG-1344, GL 89-08, and the guidelines in Electric Power Research Institute (EPRI) Report NSAC-202L-R2.
It consists of predicting loss of material using the CHECWORKS computer code, and visual inspection and volumetric examination of the affected components.
13 The NRC's acceptance criteria are baatoon of the minimum acceptable wall thickness for the components undergoing degradation by FAC.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluation of the' effect of the proposed EPU on the FAC analysis for the plant and concludes that the licensee has adequately addressed changes in the plant operating conditions on the FAC analysis.
The NRC staff further concludes that the licensee has demonstrated that the updated analyses will predict the loss of material by FAC and will ensure timely repair or replacement of degraded components following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to FAC.2.1.7 ReactorWater Cleanup System Regqulatory Evaluation The reactor water cleanup system (RWCS) provides a means for maintaining reactor water quality by filtration and ion exchange and a path for removal of reactor coolant when necessary.
Portions of the RWCS comprise the RCPB. The NRC staff's review of the RWCS included component design parameters for flow, temperature, pressure, heat removal capability, and impurity removal capability; and the instrumentation and process controls for proper system operation and isolation.
The review consisted of evaluating the adequacy of the plant's TSs in these areas under the proposed EPU conditions.
The NRC's acceptance criteria for the RWCS are based on (1) draft GDCs-S and 34, insofar as they require that the RCPB be designed and constructed so as to have an exceedingly low probability of gross rupture, significant leakage, or rapidly propagating type failures; (2) draft GDC-70, insofar as it requires that the plant design include means necessary to maintain control over the plant -radioactive effluents; and (3) draft GDC-51, insofar as it requires that systems that parts of the RCPB outside containment have appropriate features necessary to protect the health and safety of the public in case of an accidental rupture in that part-.-G04 that-syetemct t ..............
rad ......
.......fn~ementk Specific review criteria are conteined in SRP Section 5.4.8.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed EPU on the RWCS and concludes that the licensee has adequately addressed changes in impurity levels 14 and pressure and their effects on the RWCS. The NRC stfffurther concludes that the licensee has demonstrated that the RWCS will continue to be acceptable following implementation of the proposed EPU and will continue to meet the requirements of draft GDCs-9, 34, 51, and 70. GDC-41,CC 0 nd D-6Therefore, the NRC staff finds the proposed EPU acceptable with respect to the RWCS.[AdiioalRview-Ar-eae-{-aterladChmca l Engineering}
2.2 Mechanical and Civil Engineering 2.2.1 Pipe Rupture Locations and Associated Dynamic Effects Reqiulatorv Evaluation SSCs important to safety could be impacted by the pipe-whip dynamic effects of a pipe rupture.The NRC staff conducted a review of pipe rupture analyses to ensure that SSCs important to safety are adequately protected from the effects of pipe ruptures.
The NRC staff's review covered (1) the implementation of criteria for defining pipe break and crack locations and configurations, (2) the implementation of criteria dealing with special features, such as augmented inservice inspection (I61) programs or the use of special protective devices such as pipe-whip restraints, (3) pipe-whip dynamic analyses and results, including the jet thrust and impingement forcing functions and pipe-whip dynamic effects, and (4) the design adequacy of supports for SSCs provided to ensure that the intended design functions of the SSCs will not be impaired to an unacceptable level as a result of pipetwhip or jet impingement loadings.
The NRC staff's review focused on the effects that the proposed EPU may have on items (1) thru (4)above. The NRC's acceptance criteria are based on draft GDC-40 insofar as it requires that protection be provided for engineered safety features (ESFs) against the dynamic effects and missiles that might result from plant equipment failures.
GOG=-4,-whieh-equk~ee-SSG=s-imp.anteaft 
*bedsiged,-aemm at4 ..... m" ffects of-apostulated-ipe-r-uptufre-.Specific review criteria are contained in SRP Section 3.6.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluations related to determinations of rupture locations and associated dynamic effects and concludes that the licensee has adequately addressed the effects of the proposed EPU on them. The NRC staff further concludes that the licensee has demonstrated that SSCs important to safety will continue to meet the requirements of draft GDC-4OGG-following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the determination of rupture locations and dynamic effects associated with the postulated rupture of piping. , 2.2.2 Pressure-Retaining Components and Component Supports Regulatory Evaluation is st!yprsereann The NRC staff has reviewed the (and their supports) designed in accordance with' the American Society df Mechanical Engineers (ASME)Boiler and Pressure Vessel Code (B&PV Code), Section III, Division 1, and draft GDCs-1, 2, 9, 33,3440 an 42GDc 1, 1, an 15 The NRC staff's review focused on the effects of the proposed EPU on the design input pgarameters and thd design-basis loads and load combinations for normal operating, upset, emergency, and faulted conditions.
The NRC staff's review covered (1) the analyses of flow-induced vibration and (2)the analytical methodologies, assumptions, ASME Code editions, and computer programs used for these analyses.
The NRC staff's review also included a comparison of the resulting stresses and cumulative fatigue usage factors (CUFs) against the code-allowable limits. The NRC's acceptance criteria are based on (1) 10 CER 50.55a and draft GDC-1, insofar as they, require that those systems and components which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed; ODC 1, insofa ..s they r..quire that c to safety" b comneuat wthth ipoanceef-the-osfoty-functin to.... bc-........ 
(2) draft GDC-2, insofar as those systems and components which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects of earthquae 
... with the effects of normal or acciden condition;.. 
(3) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA;(4.) draft GDCs-9 and 33, insofar as they require that the ROPE be designed and constructed so as to have an exceedingly low probability of RCPB gross rupture or significant leakage; and (5) draft GDC-34 insofar as it requires that the RCPB be designed to minimize the probability of rapidly propagating type failures. 
&G-C4,-insofar-ae-4-Tequilres4hat-S$s-tmpeoat4eofet marnen........e,...... 
-,an.d-p.....tated-aooident,. ) DC1, insof......r....s it ; ,qu s-habt-he-ef-rapidWy-prepagati, ,4 .ae,,u.e,, ......)-,, Q , ..... eefar-as itrqie"htteRSb e-ee~~gay6niine-e-aeeaim.Seii review criteria are contained in SRP Sections 3.9.1, 3.9.2, 3.9.3, and 5.2.1.1; and other guidance provided in Matrix 2 of RS-001.Technical Evaluation Nuclear Steam Supply System Piping.q Components, and Supports[Insert technical evaluation for nuclear steam supply system (NSSS) piping, components, and supports.
Include an intermediate conclusion in the form of "Because [summarize reasons], the NSSS piping, components, and supports are adequate under the proposed EPU conditions."]
Balance-of-Plant Piping.q Components, and Supports[Insert technical evaluation for balance-of-plant piping, components, and supports.Include an intermediate conclusion in the form of "Because [summarize reasons], the balance-of-plant piping, components, and supports are adequate under the proposed 16 EPU conditions."]-
Reactor Vessel and Supports[Insert technical evaluation for reactor vessel and supports.
Include an intermediate conclusion in the form of "Because [summarize reasons], the reactor vessei and supports are adequate under the proposed EPU conditions."]
Control Rod Drive Mechanism[Insert technical evaluation for control rod drive mechanism.
Include an intermediate conclusion in the form of "Because [summarize reasons], the control rod drive mechanism is adequate under the proposed EPU conditions."]
Recirculation Pumos and Supports[Insert technical evaluation for reactor coolant pumps and supports.
Include an intermediate conclusion in the form of "Because [summarize reasons], the recirculation pumps and supports are adequate under the proposed EPU conditions."]
Conclusion The NRC staff has reviewed the licensee's evaluations related to the structural integrity of pressure-retaining components and their supports.
For the reasons set forth above, the NRC staff concludes that the licensee has adequately addressed the effects of the proposed EPU on these components and their supports.
Based on the above, the NRC staff further concludes that the licensee has demonstrated that pressure-retaining components and their supports will continue to meet the requirements of 10 CFR 5O.55a, CDC i, SOC 2, Soc '1 GDC !4i, ond GD-6draft GDCs-1, 2, 9, 33, 34, 40, and 42 following implementation of the proposed EPU.Therefore, the NRC staff finds the proposed EPU acceptable with respect to the structural integrity of the pressure-retaining components and their supports.2.2.3 'Reactor Pressure Vessel Internals and Core Supports Regqulatory Evaluation Reactor pressure vessel internals consist of all the structural and mechanical elements inside the reactor vessel, including core support structures.
The NRC staff reviewed the effects of the proposed EPU on the design input parameters and the design-basis loads and load combinations for the reactor internals for normal operation, upset, emergency, and faulted conditions.
These include pressure differences and thermal effects for normal operation, transient pressure loads associated with loss-of-coolant accidents (LOCAs), and the identification of design transient occurrences.
The NRC staff's review covered (1) the analyses of flow-induced vibration for safety-related and non-safety-related reactor internal components and (2) the analytical methodologies, assumptions, ASME Code editions, and computer programs used for these analyses.
The NRC staff's review also included a comparison of the resulting stresses and CUFs against the corresponding Code-allowable limits. The NRC's acceptance criteria are based on (1) i0 CFR 5O.55a and draft GDC-1 insofar as they require that those systems and components which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences be designed, fabricated, erected, tested, and inspected to quality standards that reflect the importance of the safety function to be performed;-GQC-1+sfes-as-hey-euike-hat-toqualitystandards-c-ommensurate-with-The4mpe~t-noe-ef-the-safety-funeiesto-be-eem 17 (2 rf D-,insofar as those sy on s which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to performance standards that will enable the facjlity to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects;GDC 2, insofra t eur-stht S mprtn t e-safety be designed-to-withstand the e -etso a,,t a obndwthe s-eff+t-fnemlo acidn 'odtiensj (3) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects and missiles that might result from plant equipment failures, as well as the effects of a loss of coolant accident; 
&G4-noa si eqie ht5C te4he-ef feetc o-f-and to-bo compail-ihte evrne nt!oniton associated with normaloeation, maintenance, testing, and (4) final GDC-1O, insofar as it requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
Specific review criteria are contained in SRP Sections 3.93.1, 3.9.2, 3.9.3, and 3.9.5; and other guidance provided in Matrix 2 of RS-O01..Technical Evaluation Elnsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link td the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluations related to the structural integrity of reactor internals and core supports and concludes that the licensee has adequately addressed the effects of the proposed EPU on the reactor internals and core supports.
The NRC staff further concludes that the licensee has demonstrated that the reactor internals and core supports will continue to meet the requirements of 10 CFR 50.55a, final GDC-1 0 and draft GDCs-1, 2, 40 and 42 GD , GD , CO , an-&G4following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the design of the reactor internal and core supports.2.2.4 -Safety-Related Valves and Pumps Regulatory Evaluation The NRC's staff's review included certain safety-related pumps and valves typically designated as Class 1, 2, or 3 under Section III of the ASME B&PV Code and within the scope of Section Xl of the ASME B&PV Code and the ASME Operations and Maintenance (O&M) Code, as applicable.
The NRC staff's review focused on the effects of the proposed EPU on the required functional performance of the valves and pumps. The review also covered any impacts that the proposed EPU may have on the licensee's motor-operated valve (MOV) programs related to GL 89-10, GL 96-05, and GL 95-07. The NRC staff also evaluated the licensee's consideration of lessons learned from the MOV program and the application of those lessons learned to other safety-related power-operated valves. The NRC's acceptance criteria are based on (1) draft GOC-1, insofar as it requires that those systems and components which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to quality standards that reflect the importance of the safety functions to be performed;GD-3 inta as it rsquires that "5cmontosft be dsge, farcae, erce, adte 18 (2) draft GDCs-38, 46, 47, 48, 59, 60, 61, 83, 64, and 65G- ,GC4O-,G ,-n-GG 46, insofar as they require that the emergency core cooling system (ECCS), the containment heat removal system, the containment atmospheric cleanup systems, and the cooling water systemdeepectively, be designed to permit appropriate periodic testing to ensure the leak-tight integrity and performance of their active components; (3) draft GDC-57,GL4G-44T, insofar as it requires that capability shall be provided for testing functional operability of valves and associated apparatus essential to the containment function for establishing that no failure has occurred and for determining that valve leakage does not exceed acceptable limitst4pieai systems penetrating containment be designdwth capbiit"t priedieally-testqhe-erera+lI~tt--o-he-ieelaton-~veaodetenmne-4-valve-4ea kege-s-wvith~aGGeptable-tmit; and (4) 10 CFR 50.55a(f), insofar as it requires that pumps and valves subject to that section must meet the inservice testing program requirements identified in that section. Specific review criteria are contained in SRP Sections 3.9.3 and 3.9.6; and other guidance provided in Matrix 2 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessments related to the functional performance of safety-related valves and pumps and concludes that the licensee has adequately addressed the effects of the proposed EPU on safety-related pumps and valves. The NRC staff further concludes that the licensee has adequately evaluated the effects of the proposed EPU on its MOV programs related to GL 89-10, GL 96-05, and.,GL 95-07, and the lessons learned from those programs to other safety-related, power-operated valves. Based on this, the NRC staff concludes that the licensee has demonstrated that safety-related valves and pumps will continue to meet the requirements of draft GDCs-1, 38, 46, 47, 48, 57, 59, 60, 61, 63,64, 65, GDC 1, GDC 37, GDC 10, GDC 13, GDC -!16, CDC $1, and 10 CFR 50.55a(f) following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to safety-related valves and pumps.2.2.5 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment Regulatory Evaluation Mechanical and electrical equipment covered by this section includes equipment associated with systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal. Equipment associated with systems essential to preventing significant releases of radioactive materials to the environment are also covered by this section. The NRC staff's review focused on the effects of the proposed EPU on the qualification of the equipment to withstand seismic events and the dynamic effects associated pipe-whip and jet impingement forces. The primary input motions due to the safe shutdown earthquake (SSE) are not affected by an EPU. The NRC's acceptance criteria are based on (1) draft GDC-1, insofar as it requires that those systems and components which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences be designed, fabricated, and erected to quality standards that reflect the importance of the safety functions to be performed;GDC4-1,asofat-aesbreqiresFhaSS~smpgant-e-19 
.'. *'~ -- -___ -_______________
thtcmoot ha r oto heRP edserjae ..bnc ............
d, an.d-tested-to theC-higheesalty-sttandar~s-p~aet~aIl-draft GDC-2, insofar as those systems and components which are essentila tothe prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the, public, the additional forces that might be imposed by natural phenomena such as'earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effectsGD 2,.. in.of.r.a..it eufr c -tht SSc ;important-t-aeybedcge to .ih..d.h..fetso ealqa esembinc ihteefcso norml o acien -odiin; (34) 10 CFR Part 100, Appendix A, which sets forth the principal seismic and geologic considerations for the evaluation of the suitability of plant design bases established in consideration of the seismic and geologic characteristics of the plant site; (45) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects and missiles that might result from plant equipment failures, as well as the effects of a LOCA; G-C4-nsofar-as-it-iureehat-SSGe4mpoen~t-tes-efety-bedsgndo acemnmdeate-the-effects-of-andoe4e-eompatiblawith-theernvkenmentakedens-aeseite-wt~
oia4op ....e...................estin 
.............
oodentsi-(56) draft GDCs-9 and 33, insofar as they require that the RCPB be designed and constructed so as to have an exceedingly low probability of RCPB gross rupture or significant leakage; and (6) draft GDC-34, insofar as it requires that the RCPB be designed to minimize the probability of rapidly propagating type failures; GDC444sofeast-extiemelyIewgrebabtit-y-ef-rapidly-prepagat4ngj-raeturet-;and (7) 10 CFR Part 50, Appendix B, which sets quality assurance requirements for safety-related equipment.
Specific review criteria are contained in SRP Section 3.10.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion.
The NRC staff has reviewed the licensee's evaluations of the effects of the proposed EPU on the qualification of mechanical and electrical equipment and concludes that the licensee has (1)adequately addressed the effects of the proposed EPU on this equipment and (2) demonstrated that the equipment will continue to meet the requirements of draft GDCs-1, 2, 9, 33, 34, 40 and 42;..... 1, 2,4, 1, an... 30; 10 CFR Part 100, Appendix A; and 10 CFR Part 50, Appendix B, following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the qualification of the mechanical and electrical equipment. 
-
1 ,-and-Cenqu sionsotn-ae 2%62.3 Electrical En~qineerinaq Environmental Qualification of Electrical 20 Equipment Reg ulatorv Evaluation Environmental qualificatiort(EQ) of electr ca equ pment nvolves demonstrating that the equipment is capable of performing its safety function under significant environmental stresses which could result from DBAs. The NRC staffs review focused on the effects of the proposed EPU on the environmental conditions that the electrical equipment will be exposed to during normal operation, anticipated operational occurrences, and accidents.
The NRC staffs review was conducted to ensure that the electrical equipment will continue to be capable of performing its safety functions following implementation of the proposed EPU. The NRC's acceptance criteria for EQ of electrical equipment are based on 10 CFR 50.49, which sets forth requirements for the qualification of electrical equipment important to safety that is located in a harsh environment.
Specific review criteria are contained in SRP Section 3.11.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the EQ of electrical equipment and concludes that the licensee has adequately addressed the effects of the proposed EPU on the environmental conditions for and the qualification of electrical equipment.
The NRC staff further concludes that the electrical equipment will continue to meet the relevant requirements of 10 CFR 50.49 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the EQ of electrical equipment.
Offsite Power System Regulatory Evaluation The offsite power system includes two or more physically independent circuits capable of operating independently of the onsite standby power sources. The NRC staff's review covered the descriptive information, analyses, and referenced documents for the offsite power system;and the stability studies for the electrical transmission grid. The NRC staffs review focused on whether the loss of the nuclear unit, the largest operating unit on the grid, or the most critical transmission line will result in the loss of offsite power (LOOP) to the plant following implementation of the proposed EPU. The NRC's acceptance criteria for offsite power systems are based on final GOC-17. Specific review criteria are contained in SRP Sections 8.1 and 8.2, Appendix A to SRP Section 8.2, and Branch Technical Positions (BTPs) PSB-i and ICSB-1 1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the offsite power system and concludes that the offsite power system will continue to meet the 21 
'., t -,, requirements of final GDC-17 follo w~e ~ tin proposed EPU. Adequate physical and electrical separation exists and the offsite power system has the capacity and capability to supply power to all safety loads and other required equipment.
The NRC staff further concludes that the impact of the proposed EPU on grid, stability is insignificant.
Therefore, the NRC staff finds the p'roposed EPU acceptable With respect to the offsite power system.
AC Onsite Power , System Regqulatory Evaluation The alternating current (ac) onsite power system includes those standby power sources, distribution systems, and auxiliary supporting systems provided to supply power to safety-related equipment.
The NRC staff's review covered the descriptive information, analyses, and referenced documents for the ac onsite power system. The NRC's acceptance criteria for the ac onsite power system are based on final GDC-17, insofar as it requires the system to the capacity and capability to perform its intended functions during anticipated operational occurrences and accident conditions.
Specific review criteria are contained in SRP Sections 8.1 and 8.3.1.Technical Evaluation EInsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the ac onsite power system and concludes that the licensee has adequately accounted for the effects of the proposed EPU on the system's functional design. The NRC staff further concludes that the ac onsite power system will continue to meet the requirements of final GDC-17 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the ac onsite power system.2,6:42.3.4 DC Onsite Power System Regulatory Evaluation The direct current (dc) onsite power system includes the do power sources and their distribution and auxiliary supporting systems that are provided to supply motive or control power to safety-related equipment.
The NRC staff's review covered the information, analyses, and referenced documents for the dc onsite power system. The NRC's acceptance criteria for the dc onsite power system are based on (1) draft GDC-24, insofar as it requires that in the event of loss of all offsite power, sufficient alternate sources of power shall be provided to permit the required functioning of the protection systems; and (2) draft GDC-39, insofar as it requires that alternate power systems shall be provided and designed with adequate independency, redundancy, capacity, and testability to permit the functioning required of the engineered safety features.thc system to have the itnddfnci during-antieipated-eper-atiobe~al-surrenaes-and-aeeident-eendition&s Specific review criteria are contained in SRP Sections 8.1 and 8.3.2.Technical Evaluation 22 
[Insert technical evaluation.
The technical evaluationW'i'ould (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the dc onsite power system and concludes that the licensee has adequately accounted for the effects of the proposed EPU on the system's functional design. The NRC staff further concludes that the dc onsite power system will continue to meet the requirements of draft GDCs-24 and 39G04 following implementation of the proposed EPU. Adequate physical and electrical separation exists and the system has the capacity and capability to supply power to all safety loads and other required equipment.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to the dc onsite power system.2,042.3.5 Station Blackout Reaulatorv Evaluation Station blackout (SBO) refers to a complete loss of ac electric power to the essential and nonessential switchgear buses in a nuclear power plant. SBO involves the LOOP concurrent with a turbine trip and failure of the onsite emergency ac power system. SBO does not include the loss of available ac power to buses fed by station batteries through inverters or the loss of power from "alternate ac sources" (AACs). The NRC staff~s review focused on the impact of the proposed EPU on the plant's ability to cope with and recover from an SBO event for the period of time established in the plant's licensing basis. The NRC's acceptance criteria for SBO are based on 10 CFR 50.63. Specific review criteria are contained in SRP Sections 8.1 and Appendix B to SRP Section 8.2; and other guidance provided in Matrix 3 of RS-00i.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the plant's ability to cope with and recover from arn S80 event for the period of time established in the plant's licensing basis. The NRC staff concludes that the licensee has adequately evaluated the effects of the proposed EPU on 880 and demonstrated that the plant will continue to meet the requirements of 10 CFR 50.63 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to SEO.
onc4usion -sectiens-s 2472.4 Instrumentation and Controls 24412.4.1 Reactor Protection, Safety Features Actuation, and Control 23 Systems Reaulatorv Evaluation Instrumentation and control systems are provided (1) to control plant processes having a significant impact on plant safety, (2) to initiate the reactivity control system (including control rods), (3) to initiate the engineered safety features (ESF) systfems and essential auxiliary supporting systems, and (4) for use to achieve and maintain a safe shutdown condition of the plant. Diverse instrumentation and control systems and equipment are provided for the express purpose of protecting against potential common-mode failures of instrumentation and control protection systems. The NRC staff condlucted a review of the reactor trip systerr, engineered safety feature actuation system (ESFAS), safe shutdown systems, control systems, and diverse instrumentation and control systems for the proposed EPU to ensure that the systems and any changes necessary for the proposed EPU are adequately designed such that the systems continue to meet their safety functions.
The NRC staff's review was also conducted to ensure that failures of the systems do not affect safety functions.
The NRC's acceptance criteria related to the quality of design of protection and control systems are based on 10 CFR 5O.55a(a)(1), 10 CFR 5O.55a(h), and final GDC-19 and draft GDCs-1, 12, 13, 14, 15,19, 20, 22, 23, 25, 26, 40, and 42...... 1,1,13......2..2.
23, .nd..2.. Specific review criteria are contained in SRP Sections 7.0, 7.2, 7.3, 7.4, 7.7, and 7.8.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's application related to the effects of the proposed EPU on the functional design of the reactor trip system, ESFAS, safe shutdown system, and control systems. The NRC staff concludes that the licensee has adequately addressed the effects of the proposed EPU on these systems and that the changes that are necessary to achieve the proposed EPU are consistent with the plant's design basis. The NRC staff further concludes that the systems will continue to meet the requirements of 10 CFR 50.55a(a)(1), 10 CFR 50.55(a)(h), and final GDC-19 and draft GDCs-1, 12, 13, 14, 15, 19, 20, 22, 23, 26, 26, 40, and 42. GQC the NRC staff finds the licensee's proposed EPU acceptable with respect to instrumentation and controls.2=82.S Plant Systems 2=842.5.1 "Internal Hazards 2=84-2.5.1.1 Flooding 2444-_-.2.8.1.1.1 Flood Protection Reaqulatory Evaluation The NRC staff conducted a review in the area of flood protection to ensure that SSCs important to safety are protected from flooding.
The NRC staffs review covered flooding of SSCs important to safety from internal sources, such as those caused by failures of tanks and vessels.24 The NRC staffs review focused ntak and vesselsassumed in flooding analyses to assess the impact of any additional fluid on the flooding protection that is provided.
The NRC's acceptance criteria for flood protection are based on draft GDC-2.GDC-2% Specific review criteria are contained in SRP Section 3.4.1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the proposed changes in fluid volumes in tanks and vessels for the proposed EPU. The NRC staff concludes that SSCs important to safety will continue to be protected from flooding and will continue to meet the requirements of draft G DG-2GDC, following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to flood protection.
24,-s4-.2-2.5.1.i.2 Equipment and Floor Drains Regqulatory Evaluation The function of the equipment and floor drainage system (EFDS) is to assure that waste liquids, valve and pump leak-offs, and tank drains are directed to the proper area for processing or disposal.
The EFDS is designed to handle volume of leakage expected, prevent a backflow of water that might result from maximum flood levels to areas of the plant containing safety-related equipment, and protect against the potential for inadvertent transfer of contaminated fluids to an uncontaminated drainage system. The NRC staff's review of the EFDS included the collection and disposal of liquid effluents outside containment.
The NRC staffs review focused on any changes in fluid volumes or pump capacities that are necessary for the proposed EPU and are not consistent with previous assumptions with respect to floor drainage considerations.
The NRC's acceptance criteria for the EFD$ are based on draft GDC-2 GDl~2an4insofar as itthey requires the EFDS to be designed'to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects.the-ru..tur.c..
Specific review criteria are contained in SRP Section 9.3.3.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.].
Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the EFOS and concludes that the licensee has adequately accounted for the plant changes resulting in increased water volumes and larger capacity pumps or piping systems. The NRC staff concludes that the EFDS has sufficient capacity to (1) handle the additional expected leakage resulting from the plant changes, (2) prevent the backflow of water to areas with safety-25 related equipment, and (3) enurotat ~atransferred to non-contaminated drainage systems. Based on this, the NRC staff concludes that the EFOS will continue to meet the requirements of draft GDC-2 GD~e-2-a.nd-4-following implementation of the proposed E~PU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the EFDS. '"
.1.3 Circulating r ystem Regulatory Evaluation The circulating water system (CWS) provides a continuous supply of cooling water to the main condenser to remove the heat rejected by the turbine cycle and auxiliary systems. The NRC staff's review of the CWS focused on changes in flooding analyses that are necessary due to increases in fluid volumes or installation of larger capacity pumps or piping needed to accommodate the proposed EPU. The. NDRO' .c........
criteria, forth.....c...
oe pedes~-opblties-of-sefety-re4ated4-SSG--Specific review criteria are contained in SRP Section 10.4.5.Technical Evaluation Elnsert technic:al evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the modifications to the CWS and concludes that the licensee has adequately evaluated these modifications.
The NRC staff coclde.tat con.isten 
,,,th th.... r .....ent of GD , the increased volumes of fluid leakage that could potentially result from these modifications would not result in the failure of safety-related SSCs following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the CWS.2.5.1.2 Missile Protection 2.5.1.2.1 Internally Generated Missiles Regqulatory Evaluation The NRC staff's review concerns missiles that could result from in-plant component overspeed failures and high-pressure system ruptures.
The NRC staff's review of potential missile sources covered pressurized components and systems, and high-speed rotating machinery.
The NRC staff's review was conducted to ensure that safety-related SSCs are adequately protected from internally generated missiles.
In addition, for cases where safety-related SSCs are located in areas containing non-safety-related SSCs, the NRC staff reviewed the non-safety-related SSCs to ensure that their failure will not preclude the intended safety function of the safety- related SSCs.The NRC staff's review focused on any increases in system pressures or component overspeed conditions that could result during plant operation, anticii~iated operational opcurrences, or changes in existing system configurations such that missile barrier considerations could be affected.
The NRC's acceptance criteria for the protection of SSCs important to SafetY against the effects of internally generated missiles that may result from equipment failures are based on draft 26 GDC-40.GD;G-.4.
Specific review criteria are contained in SRP Sections 3.5.1.1 and 3.5.1.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion T'he NRC staff has reviewed the changes in system pressures and configurations that are required for the proposed EPU and concludes that SSCs important to safety will continue to be protected from internally generated missiles and will continue to meet the requirements of draft GDC-40,GDG-4 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to internally generated missiles.2.5.1.2.2 Turbine Generator Regulatory Evaluation The turbine control system, steam inlet stop and control valves, low pressure turbine steam intercept and inlet control valves, and extraction steam control valves control the speed of the turbine under normal and abnormal conditions, and are thus related to the overall safe operation of the plant. The NRC staff's review of the turbine generator focused on the effects of the proposed EPU on the turbine overspeed protection features to ensure that a turbine overspeed condition above the design overspeed is very unlikely.
The NRC's acceptance criteria for the turbine generator are based on draft GDC-40G90;-4, and relates to protection of SSCs important to safety from the effects of turbine missiles by providing a turbine overspeed protection system (with suitable redundancy) to minimize the probability of generating turbine missiles.
Specific review criteria are contained in SRP Section 10.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the turbine generator and concludes that the licensee has adequately accounted for the effects of changes in plant conditions on turbine overspeed.
The NRC staff concludes that the turbine generator will continue to provide adequate turbine overspeed protection to minimize the probability of generating turbine missiles and will continue to meet the requirements of draft GDC-40-GG-following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the turbine generator.
2.5.1.3 Pipe Failures Regulatory Evaluation 27 The NRC staff conducted a review JAtd -0 p6~to from piping failures outside containment to ensure that (1) such falrswould not cause the loss of needed functions of safety-related systems and (2) the plant coiuldbe safel y shut down in the event of such failures.The NRC staff's review of pipe failures included high and moderate energy fluid system piping located outside of containment.
TheNRC staff's review focused on the effects of pipe failures on plant environmental conditions, control room habitability, and access to areas important to safe control of post-accident operations where the consequences are not bounded by previous analyses.
The NRC's acceptance criteria for pipe failuresyare based on draft GDC-40, insofar as it requires that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures.GDC 
.1, r.....c in. pad, that, c to.
et -ef-eepoeuiate-p tusesrinldlnig-the-effeots of-pi~pe-whlpplnga+4dseharging-fkiId&
Specific review criteria are contained in SRP Section 3.6.1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the changes that are necessary for the proposed EPU and the licensee's proposed operation of the plant, and concludes that SSCs important to safety will continue to be protected from the dynamic effects of postulated piping failures in fluid systems outside containment and will continue to meet the requirements of draft GDC-40-GQG-4 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to protection against postulated piping failures in fluid systems outside containment.
2.5.1.4 Fire Protection Regqulatory Evaluation The purpose of the fire protection program (FPP) is to provide assurance, through a defense-in-depth design, that a fire will not prevent the performance of necessary safe plant shutdown functions and will not significantly increase the risk of radioactive releases to the environment.
The NRC staffs review focused on the effects of the increased decay heat on the plant's safe shutdown analysis to ensure that SSCs required for the safe shutdown of the plant are protected from the effects of the fire and will continue to be able to achieve and maintain safe shutdown following a fire. The NRC's acceptance criteria for the FPP are based on (1) 10 CFR 50.48,-and-eisseeieted-Appedi-R--to-4OGF insofar as ittkey requires the development of an FPP to ensure, among other things, the capability to safely shut down the plant; (2) final GDC-3, insofar as it requires that (a) SSCs important to safety be designed and located to minimize the probability and effect of fires, (b) noncombustible and heat resistant materials be used, and (c) fire detection and fighting systems be provided and designed to minimize the adverse effects of fires on SSCs important to safety; (3) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing. D "" .. .. 
'- mortn o sfe, no b shared ~m ng-uolear-power-Untt, it les ant-:-be-showP,-thae-Whnig-wilkot imp.. their.......
ablt to par.r their.......
st fun....o...
Secific review criteria are contained in SRP Section 9.5.1, as supplemented by the guidance provided in Attachment I to Matrix 5 of Section 2.1 of RS-001.28 T_.echnical Evaluation"[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's fire-related safe shutdown assessment and concludes that the licensee has adequately accounted for the effects of the increased decay heat on the ability of the required systems to achieve and maintain safe shutdown conditions.
The NRC staff further concludes that the FPP will continue to meet the requirements of 103 CFR 50.48, Appendix Rto ICFR Part 50, and final GDC-3, and draft GDC-'1Dc n 6following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to fire protection.
2.5.2 Fission Product Control 2.5.2.1 Fission Product Control Systems and Structures Reoulatory Evaluation The NRC staff's review for fission product control systems and structures covered the basis for developing the mathematical model for OBLOCA dose computations, the values of key parameters, the applicability of important modeling assumptions, and the functional capability of ventilation systems used to control fission product releases.
The NRC staff's review primarily focused on any adverse effects that the proposed EPU may have on the assumptions used in the analyses for control of fission products.
The NRC's acceptance criteria are based on draft GDC-70, insofar as it requires that the plant design include means to control the release of radioactive effluents.GDC- 
!1, inoofar as it requires that the containment atmosphere eleaeup-syetcm bc provideto rdcet eoontration of enirnmn fllwng postulated cidns Specific review criteria are contained in SRP Section 6.5.3._Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a ciear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on fission product control systems and structures.
The NRC staff concludes that the licensee has adequately accounted for the increase in fission products and changes in expected environmental conditions that would result from the proposed EPU. The NRC staff further concludes that the fission product control systems and structures will continue to provide adequate fission product removal in pest-post-accident environments following implementation of the proposed EPU. Based on this, the NRC staff also concludes that the fission product control systems and structures will continue to meet the requirements of draft Therefore, the NRC staff finds the proposed EPU acceptable with respect to the fission product control systems and structures.
29 
-- 9&;Q Zr 2.5.2.2 Main Condenser Evacuation System Reaqulatorv Evaluation The main condenser evacuation system (MCES) generally consists of two subsystems:
the"hogging" or startup system which initially establishes main condenser vacuum and the system which maintains condenser vacuum once it has been established.
The NRC staff's review focused on modifications to the sYstem that may affect gaseous radioactive material handling and release assumptions, arid design features to preclude the possibility of an explosion (if the potential for explosive mixtures exists). The NRC's acceptance criteria for the MCES are based on (1) draft GOC-7OGDG-60, insofar as it requires that the plant design include means to control the release of radioactive effluents; and (2) draft GDC-17G90-&sect;e4, insofar as it requires that means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, inefuding-from anticipated transients, and from accident conditionsoperationab-osur-rencs enpstulated-odents.
Specific review criteria are contained in SRP Section 10.4.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.J Conclusion The NRC staff has reviewed the licensee's assessment of required changes to the MCES and concludes that the licensee has adequately evaluated these changes. The NRC staff concludes that the MCES will continue to maintain its ability to control and provide monitoring for releases of radioactive materials to the environment following implementation of the proposed EPU. The NRC also concludes that the MCES will continue meet~the requirements of draft GDCs-1 7 and 7OGDC 60 and 61. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the MCES.2.5.2.3 Turbine Gland Sealing System Repulatorv Evaluation The turbine gland sealing system is provided to control the release of radioactive material from steam in the turbine to the environment.
The NRC staff reviewed changes to the turbine gland sealing system with respect to factors that may affect gaseous radioactive material handling (e.g., source of sealing steam, system interfaces, and potential leakage paths). The NRC's acceptance criteria for the turbine gland sealing system are based on (1) draft GDC-7OGDQ-6O, insofar as it requires that the plant design include means to control the release of radioactive effluents; and (2) draft GDC-17GDC-64, insofar as it requires that means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, including anticipated transients, and from accident conditionsepaier~~nabaeuncs-and poct-ulatod accidenqts.
Specific review criteria are contained in SRP Section 10.4.3.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and 30 (2) provide a clear link to the conclusionsi l i reached by' th NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of required changes to the turbine gland sealing system and concludes that the licensee has adequately evaluated these changes. The NRC staff concludes that the turbine gland sealing system will continue to maintain its ability to control and provide monitoring for releases of radioactive materials to the environment consistent with draft GDCs-17 and 70GDG-60-and-64.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to the turbine gland sealing system.2.5.2.4 Main Steam Isolation Valve Leakage Control System[Not Applicable.
BFN does not have a MSIV leakage control system.]Redunan,-euiekatin -selatio-,al es-arprcvidedo e "c manseal. Th eaae.... Me..........
44............e-amouP~.ef-direet 
,-et.feate 
...... fr. --
the mnale--aedenainmentis-
~eur-ed....
The NRC staff's rev-iew,'
of the ,MSIV/,akage-eete6 seem.... -...feeI3sed the.eff.ct..a.
theprooe d P on........
the- amouto eakage-s sumed toi ccr.Te,-C-c.panecrtei systems-aenetring contateinmentsee provded det~etinfadfioltionef-phe-tleeser-Sp-oif-2.5.3.1o SpeeafontaFuel Pool Cooling and~CInertu Sytehia Revclulationy TEvautehiaonla~nshu-4}lalyeps-h~e Threpospetfelpoolaprovidestwet setoraeeofspent fe ssemlie* he s~w a f'ety funtion of th s-)pentifue aslearblink tovhered nui onwaer dringache bsthrae NCondti ffons* do NCumstaf eview frthe propuso~sedEUfcusedn nteefcsojtepooe P nth aaiiyo h system-tonproindeaeqate-coolingeeo theaspaenutfely-deurntgdaelle operatingfandpaccident EPUonditioshe assume lepakage throuhtheri for the spCen taffloo coulng condcluestatu sythe arebas4eed nt(1) draft GDC-4,einsofar as r-eator facidiies satl ntshae syk~e-stemsi orby cDm5poTereforness ith C ston afft fins nth ipropoed EPU acetabe whaith; ranec to) dath GDC-%&7., Componha rliben Colndn ecay Heat Reoalssemovaledsge oprvn aaet 2~&92.5.A Spnt Fel Pol Colin an 
* .nv-n*,L the fuel in storage facilities' thauc to plant operating areas a..................n.
be pr.ovded and. (3) draft GDC-69, insofar as containment of fuel shall be provided if accidents could lead to release of undue a'ni6unts of radioactivity to the public environs.
CCC 61, insofar as.t.equre tha fu...... strg"ytm eindwt in&sect;4heqmpeanee4e-safety-of-deeay-heat-reamvalradmaue+s-te-feven-e-sig4icn oso ulsoagfoln"netr ndracdn odtos Speic review criteria are contained in SRP Section 9.1.3, as supplemented by the guidance provided in Attachment i to Matrix 5 of Section 2.1 of RS-OO1.Technical Evaluation EInsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section. ]I Conclusion The NRC staff has reviewed the licensee's assessment related to the spent fuel pool cooling and cleanup system and concludes that the licensee has adequately accounted for the effects of the proposed EPU on the spent fuel pool cooling function of the system. Based on this review, the NRC staff concludes that the spent fuel pool cooling and cleanup system will continue to provide sufficient cooling capability to cool the spent fuel pool following implementation of the proposed EPU and will continue to meet the requirements of draft67 and 69. GDGs-6-A%4r,-at-Therefore, the NRC staff finds the proposed EPU acceptable with respect to the spent fuel pool cooling and cieanup system.2-44422.5.3.2 Station Service Water System Regqulatory Evaluation The station service water system (SWS) provides essential cooling to safety-related equipment and may also provide cooling to non-safety-related auxiliary components that are used for normal plant operation.
The SWS includes the Emergency Equipment Cooling Water (EECW) and the Residual Heat Removal Service Water (RHRSW) systems. The NRC staff's review covered the characteristics of the station SWS (i.e., EECW and RHRSW systems)components with respect to their functional performance as affected by adverse operational (i.e., water hammer) conditions, abnormal operational conditions, and accident conditions (e.g., a LOCA with the LOOP). The NRC staff's review focused on the additional heat load that would result from the proposed EPU. The NRC's acceptance criteria are based on (1) draft GD~s-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as theeffects of a LOCA;oefetatoR 
(..wtr..rn0 matenanoe.4eetn-ad (2) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing. CDC 5, in"sofar as-be~sha ed-amo eng-cear-power-unit~u~s-u ea be-shewn-that-shanng-w4noet-sgfn~fieantty-impafi 
.....effo~m4heiw-safety-fumeioensi-and43)-GDA-44nsoefas~r-a eqtuis-tat-a-system-w~h-the-eapabiity-o-ttasfr-eat~ad em previ-de-.-Specific review criteria are contained in SRP Section 9.2.1, as supplemented by GL 32 89-13 and GL 96-06.Y..UA. ?,2, .i.q.F,-OHI Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment related to the effects of the proposed EPU on the station SWL EECW and RHRSW systems and concludes that the licensee has adequately accounted for the increased heat loads on system performance that would result from the proposed EPU. The NRC staff concludes that the station-SWS EECW and RHRSW systems will continue to be protected from the dynamic effects associated with flow instabilities and provide sufficient cooling for SSCs important to safety following implementation of the proposed EPU. Therefore, the NRC staff has determined that the station-SWS EECW and RHRSW systems will continue to meet the requirements of draft GDCs-4, 40, and 42. GOGs-4,-7 andA44.Based on the above, the NRC staff finds the proposed EPU acceptable with respect to the station SWS EECW and RHRSW systems.2,&082.5.3.3 Reactor Auxiliary Cooling Water Systems Requlatorv Evaluation The NRC staff's review covered reactor auxiliary cooling water systems that are required for (1) safe shutdown during normal operations, anticipated operational occurrences, and mitigating the consequences of accident conditions, or (2) preventing the occurrence of an accident.These systems include non-safety related-eosed-Jee auxiliary cooling water systems, Reactor Building Closed Cooling Water (RBCCW) system and Raw Cooling Water (RCW)system, for reactor system components, reactor shutdown equipment, ventilation equipment, and components of the ECCS. The NRC staff's review covered the capability of the auxiliary cooling water systems to provide adequate cooling water to safety-related ECCS components and reactor auxiliary equipment for all planned operating conditions.
Emphasis was placed on the cooling water systems for safety-related components (e.g., ECCS equipment, ventilation equipment, and reactor shutdown equipment).
The NRC staff's review focused on the additional heat load that would result from the proposed EPU. The NRC's acceptance criteria for the reactor auxiliary cooling water system are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA;GDG-4-, postulated..accidents. 
(2) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing; and (3) draft GDC-41, insofar that the Reactor Auxiliary Cooling Water Systems are relied upon by engineered safety features for performing their safety functions.
GDC 5, insofar as it be shown tha s.haring...
wil net. significantlyH, ability to, perfor thei safety+,
provided-.Specific review criteria are contained in SRP Section 9.2.2, as supplemented by GL 89-13 and GL 96-06.:33 
&.;A;O "h2UL JAjI-'U~l Technical EvaluationS
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conciusion section.]Conclusion.
The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the reactor auxiliary cooling water systems and concludes that the licensee has adequately accounted for the increased heat loads from the proposed EPU on system performance.
The NRC staff concludes that the reactor auxiliary cooling water systems will continue to be protected from the dynamic effects ass~ociated with flow instabilities and provide sufficient cooling for SSCs important to safety following irnplementation of the proposed EPU. Therefore, the NRC staff has determined that the reactor auxiliary cooling water systems will continue to meet the requirements of draft GDCs-4, 40, 41, and Based on the above, the NRC staff finds the proposed EPU acceptable with respect to the reactor auxiliary cooling water systems.2J442.43.4 Ultim ate Heat SinkReuatr Evaluation The ultimate heat sink (UHS) is the source of cooling water provided to dissipate reactor decay heat and essential cooling system heat loads after a normal reactor shutdown or a shutdown following an accident.
The NRC staff's review focused on the impact that the proposed EPU has on the decay heat removal capability of the UHS. Additionally, the NRC staff's review included evaluation of the design-basis UHS temperature limit determination to confirm that post-licensing data trends (e.g., air and water temperatures, humidity, wind speed, water volume) do not establish more severe conditions than previously assumed. The NRC's acceptance criteria for the UHS are based on (1) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is not impaired byte sharing; GDC 5, incofar os-i-eue-htS4-m~ratt ayntb shae-ae&On-nu~ear-peweURmts-Wn~ess~t~-eabe-shRheaPhaRR-Whar4OgSIiR44oGft~t-impafrtheie-ability4e-prfecm-hei-s~afetyj-and-(2) draft GDC-41, insofar that the UHS is relied upon by engineered safety features for performing their safety functions; and (3) draft GDC-52, insofar that the UHS is relied upon by containment heat removal systems for performing their safety functions.
CDCr 11, icofar a-it rngui... that ...t.. with.. tho-oapabitity ta an -normal-oper-atil&sect;and 0ident-eenditiene-be-frevided-.Specific review criteria are contained in SRP Section 9.2.5.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusio.0n The NRC staff has reviewed the information that was provided by the licensee for addressing 34 
?32U JAIOHHCIO the effects that the proposed EPU would have on the UHS &Tafetyfncin including the licensee's validation of the design-basis UHS temperature limit based on post-licensing data.Based on the information that was provided, the NRC staff concludes that the proposed EPU will not compromise the design-basis safety function of the UHS, and that the UHS will continue to satisfy the requirements of draft GDCs-4, 41, and 5..and...
following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the UHS.245-1-2.5.4 Balance-of-Plant Systems Z-6-A0-.2.5.4.1 M amn Steam Rgltr Evaluation The main steam supply system (MSSS) transports steam from the NSSS to the power conversion system and various safety-related and non-safety-related auxiliaries.
The NRC staff's review focused on the effects of the proposed EPU on the system's capability to transport steam to the power conversion system, provide heat sink capacity, supply steam to drive safety system pumps, and withstand adverse dynamic loads (e.g., water steam hammer resulting from rapid valve closure and relief valve fluid discharge loads). The NRC's acceptance criteria for the MSSS are based on (1) draft GDC-40 insofar as it requires that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures;GDC..1, insof....r ac it .re rc htS.,ipran-osaeyb poetd.gi dyna n~et-impingernentes-and (2) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing. G-DC-5,-unessit-enabeshown+
hat-shaing-w~l-net-signifieently-ip.4heirabiilbityt-eflmtb review criteria are contained in SRP Section 10.3.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the MSSS and concludes that the licensee has adequately accounted for the effects of changes in plant conditions on the design of the MSSS. The NRC staff concludes that the MSSS will maintain its ability to transport steam to the power conversion system, provide heat sink capacity, supply steam to steam-driven safety pumps, and withstand steam hammer. The NRC staff further concludes that the MSSS will continue to meet the requirements of draft GDCs-4 and 40.GDCs -'. and 5. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the MSSS.2449_422.5.4.2 Ma in Condenser Regulatory Evaluation 35 The main condenser (MC) system issle~WC di d deaer:ate the exhaust steam from the main turbine and provide a heat sink'f~rdhe tLbinebypass system (TBS). F-o&BW~e-Because BFN does not havewfthaut an MSIV leakage control system, the MC system may-also serves an accident mitigation function to act as a holdup volume for the plate out of fission products leaking through the MSIVs following core damage. The NRC staff's review focused on the effects of the proposed EPU on the steam bypass capability with respect to load rejection assumptions, and on the ability of the MC system to withstand the blowdown effects of steam from the TBS. The NRC's acceptance criteria for the MC system are based on draft GDC-70-GD-6, insofar as it requires that the plant design include' means to control the release of radioactive effluents.
Specific review criteria are contained in SRP Section 10.4.1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the MC system and concludes that the licensee has adequately accounted for the effects of changes in plant conditions on the design of the MC system. The NRC staff concludes that the MC system will continue to maintain its ability to withstand the blowclown effects of the steam from the TBS and thereby continue to meet draft GDC-70&D-4 with respect to controlling releases of radioactive effluents.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to the MC system.24-_10a2.5.4.3 T urbine Bypass Regqulatory Evaluation The TBS is designed to discharge a stated percentage of rated main steam flow directly to the MC system, bypassing the turbine. This steam bypass enables the plant to take step-load reductions up to the TBS capacity without the reactor or turbine tripping.
The system is also used during startup and shutdown to control reactor pressure.
For a BWR without an MSIV leakage control system, the TBS could also provide an accident mitigation function.
A TBS, along with the MSSS and MC system, may be credited for mitigating the effects of MSIV leakage during a LOCA by the holdup and plate out of fission products.
The NRC staffs review for the TBS focused on the effects that the proposed EPU have on load rejection capability, analysis of postulated system piping failures, and the consequences of inadvertent TBS operation.
The NRC's acceptance criteria for the TBS are based on (4-1draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA; GQC-4T,.. ..be dsgncd toccmotchefecsf
..... ...,m atil- with-th...................
condition a...ociat.d "h noma op....ie"-
maineee-tesaF-nd-esulat-edla~cci, t (including......
pipe. brak or..malfunctions oasf he TBSraed (2- leC.-hea ineafa asitr-eq~uir-est r--that r aRHR-sysem e povde t raofefss and .. thc ; des ...n conditnso te "R.... am ,,xcoodo~d.--Specific review criteria are contained in SRP Section 10.4.4.Technical 36 
, .A [Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the TBS. The NRC staff concludes that the licensee has adequately accounted for the effects of changes in plant conditions on the design of the TBS. The NRC staff concludes that the TBS will continue to mitigate the effects of MSIV leakage during a LOCA and provide a means for shutting down the plant during normal operations.
The NRC staff further concludes that TBS failures will not adversely affect essential SSCs. Based on this, the NRC staff concludes that the TBS will continue to meet draft GDCs-40 and 42.GD the NRC staff finds the proposed EPU acceptable with respect to the TBS.%2=5r1442.5.4.4 Condensate and Feedwater Regiulatory Evaluation The condensate and feedwater system (CFS) provides feedwater at a particular temperature, pressure, and flow rate to the reactor. -The only part of the CFS classified as safety-related is the feedwater piping from the NSSS up to and including the outermost containment isolation valve. The NRC staff's review focused on how the proposed EPU affects previous analyses and considerations with respect to the capability of the CFS to supply adequate feedwater during plant operation and shutdown, and isolate components, subsystems, and piping in order to preserve the system's safety function.
The NRC's acceptance criteria for the CFS are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA;GDQ4 7 nofa si euires that SS~csmpotankt to safety be designeado-t a~aemnmodate-theaeffects-ef-and-te-be-oempatible-with-t4e environmen tab- ced einetuding posil flui flwi -bltiee-waeham maintenanc, tosin, an otted-aojns (2) draft GDC-4, insofar as reactor facilities shall not share systems or components unless itis shown safety is not impaired by the s har ing.
tye--ehreamn-a nuc.Jear-power-un454&n~lessA-oae-e-shown that-sharnlnotsgeiieetymaiphefrab4it caiity tto heatlas-rmsafety-relatd SSot- a-eat winkudw ot ora avial frmol tecste-system-oreonly the ofeit syte, asuin sig~ uwe--Specific review criteria are contained in SRP Section 10.4.7.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the CFS and concludes that the licensee has adequately accounted for the effects of changes in 37 plant conditions on the design 6ff~t NClf ocludes that the CFS will continue to maintain its ability to satisfy feedwater requirements for normal operation and shutdown, withstand water hammer, maintain isolation capability in order to preserve the system safety function, and not cause failure of safety-related SSCs. The NRC staff further concludes that the CFS will continue to meet the requirements of draft GDCs-4, 40 iind 42. GD..,. ad1.Therefore, the NRC staff finds the proposed EPU acceptable with respect to the CFS.
Waste Management Systqtms,.! 44-142.5.5.1 Gaseous Waste Management Systems Recqulatory Evaluation The gaseous waste management systems involve the gaseous radwaste system, which deals with the management of radioactive gases collected in the offgas system or the waste gas storage and decay tanks. In addition, it involves the management of the condenser air removal system; the gland seal exhaust and the mechanical vacuum pump operation exhaust; and the building ventilation system exhausts.
The NRC staff's review focused on the effects that the proposed EPU may have on (1) the design criteria of the gaseous waste management systems, (2) methods of treatment, (3) expected releases, (4) principal parameters used in calculating the releases of radioactive materials in gaseous effluents, and (5) design features for precluding the possibility of an explosion if the potential for explosive mixtures exists. The NRC's acceptance criteria for gaseous waste management systems are based on (1) 10 CFR 20.1302, insofar as it provides for demonstrating that annual average concentrations of radioactive materials released at the boundary of the unrestricted area do not exceed specified values; (2) final GDC-3, insofar as it requires that (a) SSCs important to safety be designed and located to minimize the probability and effect of fires, (b) noncombustible and heat resistant materials be used, and (c) fire detection and fighting systems be provided and designed to minimize the adverse effects of fires on SSCs important to safety; (3) draft GDC-7OGDQ-;-60, insofar as it requires that the plant design include means to control the release of radioactive effluents; (4) draft GOC-69GQ0-64, insofar as it requires that containment of fuel and waste storage shall be provided if accidents could lead to release of undue amounts of radioactivity to the public.
and (5) 10 CFR Part 50, Appendix I, Sections 11.8, IIC, and 11.0, which set numerical guides for design objectives and limiting conditions for operation to meet the "as low as is reasonably achievable" (ALARA) criterion.
Specific review criteria are contained in SRP Section 11.3.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment related to the gaseous waste management systems. The NRC staff concludes that the licensee has adequately accounted for the effects of the increase in fission product and amount of gaseous waste on the abilities of the systems to control releases of radioactive materials and preclude the possibility of an explosion if the potential for explosive mixtures exists. The NRC staff finds that the gaseous waste management systems will continue to meet their design functions following implementation of the proposed EPU. The N RC staff further concludes that the licensee has demonstrated that the gaseous waste management systems will continue to meet the requirements of 10 CFR 20.1302;final GQ~e-3GDC-3, draft GDCs-69 and 7060, 5-a-d1t; and 10 CER Part 50, Appendix I, 38 Sections 1I.8, I.C, and 11.0. Thrfrthe NR safin. rpedEPU acceptable with respect to the gaseous waste management systems.2.-4A-42.5.5.2 Liquid Waste Management SystemsReuatr Evaluation The NRC staff's review for liquid waste management systems focused on the effects that the proposed EPU may have on previous analyses and considerations related to the liquid waste management systems' design, design objectives, design criteria, methods of treatment, expected releases, and principal parameters used in calculating the releases of radioactive materials in liquid effluents.
The NRC's acceptance criteria for the liquid waste management systems are based on (1) 10 CFR 20.1302, insofar as it provides for demonstrating that annual average concentrations of radioactive materials released at the boundary of the unrestricted area do not exceed specified values; (2) draft GDC-70GDG-4O, insofar as it requires that the plant design include means to control the release of radioactive effluents; and (3)-GOC-&64-draft GDC-69, insofar as it requires that containment of fuel and waste storage shall be provided if accidents could lead to release of undue amounts of radioactivity to the public environsinsofaaas-it-
+equesht-r e
and (4) 10 CFR Part 50, Appendix I, Sections lI.A and i1.D, which set numerical guides for dose design objectives and limiting conditions for operation to meet the ALARA criterion.
Specific review criteria are contained in SRP Section 11.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment related to the liquid waste management systems. The NRC staff concludes that the licensee has adequately accounted for the effects of the increase in fission product and amount of liquid waste on the ability of the liquid waste management systems to control releases of radioactive materials.
The NRC staff finds that the liquid waste management systems will continue to meet their design functions following implementation of the proposed EPU. The NRC staff further concludes that the licensee has demonstrated that the liquid waste management systems will continue to meet the requirements of 10 CFR 20.1302; draft GDCs-69 and 70GQs-& -an6; and 10 CFR Part 50, Appendix I, Sections lI.A and lI.D. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the liquid waste management systems.-2A-_t32.5.5.3 Solid Waste Management Systems Regjulatorv Evaluation The NRC staff's review for the solid waste management systems :(SWMS) focused on the effects that the proposed EPU may have on previous analyses and considerations related to th~e design objectives in terms pf expected volumes of waste to be processed and handled, the wet and dry types of waste to be processed, the activity and expected radionuclide distribution contained in the waste, equipment design capacities, and the principal parameters emploYed in 39 r r the design of the SWMS. The NRC's paSWMS are based on (1) 10 CFR 20.1302, insofar as it provides for demonstrating that annual average concentrations of radioactive materials released at the boutidary of the unrestricted area do not exceed specified values; (2) draft GDC-70G4C-6O, insofar as it requires that the plant design include means to control the release of radioactive effluents; (3) draft;GDC-18GD-6, insofar as it requires that monitoring and alarm instrumentation shall be provided for fuel and waste storage and handling areas for conditions that might contribute to loss of continuity in decay heat removal and to radiation exposureseystenebeprviedinwete-bantin-a~eae~e-deteet 6eondtiosht--may-eeti-xeeve-diatioemvele, (4) draft GDC-1 7GQG-64, insofar as it requires that means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, from anticipated transients, and from accident conditionsielnitg AQOc,,,''', and postulte ...c...nt.; and (5) 10 CFR Part 71, which states requirements for radioactive material packaging.
Specific review criteria are contained in SRP Section 11.4.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment related to the SWMS. The NRC staff concludes that the licensee has adequately accounted for the effects of the increase in fission product and amount of solid waste on the ability of the SWMS to process the waste. The NRC staff finds that the SWMS will continue to meet its design functions following implementation of the proposed EPU. The NRC staff further concludes that the licensee has demonstrated that the SWMS will continue to meet the requirements of 10 CFR 20.1302, draft GDCs-17, 18, and 7OG s6, 6, an ', and 10 CFR Part 71. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the SWMS.%&42.5.6 Additional Considerations 2.-4-4 --4-2,5.6.1 Emergency Diesel Engine Fuel Oil Storage and Transfer System Recqulatorv Evaluation Nuclear power plants are required to have redundant onsite emergency power supplies of sufficient capacity to perform their safety functions (e.g., power diesel engine-driven generator sets), assuming a single failure. The NRC staff's review focused on increases in emergency diesel generator electrical demand and the resulting increase in the amount of fuel oil necessary for the system to perform its safety function.
The NRC's acceptance criteria for the emergency diesel engine fuel oil storage and transfer system are based on (1) draft GDC-40 insofar as it requires that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures; GD ............
r... it re..re that-39Cc mp44at-9 feree coiao wit piebok; (2) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing;GQC-,, ,iteqtfe4hat-33Cc mportant-to sa fetnebesadam g nucioer pwe units uessita-eshewnmtat-shar4&sect;-i4-ne (3) final GDC-17, insofar as it requires onsite power supplies to have sufficient independence and redundancy to perform their safety functions, assuming a single 40 faiure Seciicreview criteria are contained in SRP Sectin954 Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section,]The NRC staff has reviewed the licensee's assessment related to the amount of required fuel oil for the emergency diesel generators and concludes that the licensee has adequately accounted for the effects of the increased electrical demand on fuel oil consumption.
The NRC staff concludes that the fuel oil storage and transfer system will continue to provide an adequate amount of fuel oil to allow the diesel generators to meet the onsite power requirements of final GDC-17 and draft GDCs-4, and ,-4rran Therefore, the NRC staff finds the proposed EPU acceptable with respect to the fuel oil storage and transfer system.%.&4242.5.6.2 Light Load Handling System (Related to Refueling)
Reciulatorv Evaluation The light load handling system (LLHS) includes components and equipment used in handling new fuel at the receiving station and the loading of spent fuel into shipping casks. The NRC staff's review covered the avoidance of criticality accidents, radioactivity releases resulting from damage to irradiated fuel, and unacceptable personnel radiation exposures.
The NRC staff's review focused on the effects of the new fuel on system performance and related analyses.
The NRC's acceptance criteria for the LLHS are based on (1) draft GDCs-68 and 69gc6, insofar as theylt requires that systems that contain radioactivity be designed with appropriate oonfinemeit-containment and with suitable shielding for radiation protection; and (2) draft GDC-660G62, insofar as it requires that criticality be prevented.
Specific review criteria are contained in SRP Section 9.1.4.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the new fuel on the ability of the LLHS to avoid criticality accidentsaand concludes that the licensee has adequately incorporated the effects of the new fuel in the analyses.
Based on this review, the NRC staff further concludes that the LLHS will continue to meet the requirements of draft GDCs-66, 68, ad6G s 61 a nd6 for radioactivity releases and prevention of criticality accidents.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to the LLHS.
f4nsert-Requlate'yE~auatoie-,Te~hnie4lEvcaluatienran-Gesusioseostiens-s nesessarA 41 2.6 Containment Review Conside :tons .A i , 2.6.1 Primary Containment Functional r -. Regqulatory Evaluation The containment encloses the reactor system and is the final barrier against the release of significant amounts of radioactive fission products in the event of an accident.
The NRC staffs review for the primary containment functional design covered (1) the temperature and pressure conditions in the drywell and wetwell due to a spectrum of postulated LOCAs, (2) the differential pressure across the operating deck for a spectrum of LOCAs (Mark II containments only), (3)suppression pool dynamic effects during a LOCA or following the actuation of one or more RCS safety/relief valves, (4) the consequences of a LOCA occurring within the containment (wetwell), (5) the capability of the containment to withstand the effects of steam bypassing the suppression pool, (6) the suppression pool temperature limit during ROS safety/relief valve operation, and (7)the analytical models used for containment analysis.
The NRC's acceptance criteria for the primary containment functional design are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA; GQ I noa as " "requre with-the.-envir.............
dit...... 
.......ted-w..............pe 
.......,,,,, i't an.. c , to s,.,ng, and ,.'pastulated-eeidentsrandth at-suehWSSs-be-proteeted 
*against-dlynamrnmefeots} 
(2) draft G DC-I0, insofar as it requires that reactor containment be designed to sustain the initial effects of gross equipment failures, such as a large coolant boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain functional capability for as long as the situation requires; GD-,&
to. ct a ..i.. n...... en......... , ....-tlght-barrier-against-the-uno-entrolled-elease ef-radloaatvityo-te-ewrnent (3) draft GDC)-49, insofar as it requires that the containment and its associated heat removal systems be designed so that the containment structure can accommodate, without exceeding the design leakage rate, the pressures and temperatures resulting from the largest credible energy release following a LOCA, including considerable margin for effects from metal-water or other chemical reactions that could occur as a consequence of failure of emergency t;ore cooling systems; GDC50 ins=Rofar" ; ...asi&#xf7;q4fehet ts asseeatadha 
...... o..a.. sy..tem..bc d...n. so that.......
conta.. in. nt...t.u.t..r..c.n a~eomredater-witheu-x~eeedinff-the-desigrnleakegc rate anrd wi!th cuffiientmargiPrthe-
.... (4) draft GDnC-I2 insofar as it requires that instrumentation and controls be provided as required to monitor and maintain variables within prescribed operating ranges;CDC 13-, nsofar as' it. requf ,.rcz" (5) draft GDC-.17GD-6, insofar as it requires that means be provided to monitor the reactor containment atmosphere for radioactivity that may be released from normal operations, from anticipated transients, and from accident conditionspeetdaed-aodet.
Specific review criteria are contained in SRP Section 6.2.1.1.C.
Technical Evaluation EInsert technical evaluation.
The technical evaluation should (I) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion 42 The NRC staff has reviewed the licensesassessmeri of the containment temperature and pressure transient and concludes that the licensee has adequately accounted for the increase of mass and energy resulting from the proposed EPU. The NRC staff further concludes that containment systems will continue to provide sufficient pressure and temperature mitigation capability to ensure that containment integrity is maintained.
The NRC staff also concludes that containment systems and instrumentation will continue to be adequate for monitoring containment parameters and release of radioactivity during normal and accident conditions and the containment and associated systems will continue to meet the requirements of draft GDCs-10, 12, 17, 40, 42, and 49 GDC.__ .1, 13., 50,n 61! following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to primary containment functional design.2.6.2 Subcompartment Analyses Reaqulatorv Evaluation A subcompartment is defined as any fully or partially enclosed volume within the primary containment that houses high-energy piping and would limit the flow of fluid to the main containment volume in the event of a postulated pipe rupture within the volume. The NRC staff's review for subcompartment analyses covered the determination of the design differential pressure values for containment subcompartments.
The NRC staff's review focused on the effects of the increase in mass and energy release into the containment due to operation at EPU conditions, and the resulting increase in pressurization.
The NRC's acceptance criteria for subcompartment analyses are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA; GDC-4, insofar-ae-it-eukshat-SS-Gpe-ipo~nt-e-safemye-designed-te-a~cemmedate-the effe~ts-of the-ernvkonentak-eeadit ens- ssooeedith-nerma~eperatioen'maintena eetesting~and GDC-49, insofar as it requires that the containment be designed so that the containment structure can accommodate, without exceeding the design leakage rate, the pressures and temperatures resulting from the largest credible energy release following a LOCA.across the... , walls , of .th subcom.rtmnt Specific review criteria are contained in SRP Section 6.2.1.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the subcornpartment assessment performed by the licensee and the change in predicted pressurization resulting from the increased mass and energy release.The NRC staff concludes that containment SSCs important to safety will continue to be protected from the dynamic effects resulting from pipe breaks and that the subccmpartments will continue to have sufficient margins to prevent fracture of the structure due to pressure difference across the walls following implementation of the proposed EPU. Based on this, the NRC staff concludes that the plant will dontinue to meet draft GDCs-40, 42 and 49 GDs' and......for the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to subcompartment analyses.43 2.6.3 Mass and Energy Release 2.6.3.1 Mass and Energy Release Analysis for Postulated Loss of Coolant Regqulatory Evaluation 
,' 'The release of high-energy fluid into containment from pipe breaks could challenge the structural integrity of the containment, including subcompartments and systems within the containment.
The NRC staff's review covered the energy sources that are available for release to the containment and the mass and energy release rate calculations for the initial blowdown phase of the accident.
The NRC's acceptance criteria for mass and energy release analyses for postulated LUCAs are based on (1) draft GDC-49, insofar as it requires that the containment and its associated heat removal systems be designed so that the containment structure can accommodate, without exceeding the design leakage rate, the pressures and temperatures resulting from the largest credible energy release following a L O CA;-GDC , -the-mass--
and-energy-release-,na~aysi~s-tassare~hateentainment-designagn~ateined and (2) 10 CFR Part 50, Appendix K, insofar as it identifies sources of energy during a LOCA. Specific review criteria are contained in SRP Section 6.2.1.3.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's mass and energy release assessment and concludes that the licensee has adequately addressed the effects of the proposed EPU and appropriately accounts for the sources of energy identified in 10 CFR Part 50, Appendix K.Based on this, the NRC staff finds that the mass and energy release analysis meets the requirements in draft GDC-49 ensuring that the analysis is conservative.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to mass and energy release for postulated LOCA.2.6.4 Combustible Gas Control in Containment Regulatory Evaluation Following a LOCA, hydrogen and oxygen may accumulate inside the containment due to chemical reactions between the fuel rod cladding and steam, corrosion of aluminum and other materials, and radiolytic decomposition of water. If excessive hydrogen is generated, it may form a combustible mixture in the containment atmosphere.
The NRC staff's review covered (1) the production and accumulation of combustible gases, (2) the capability to prevent high concentrations of combustible gases in local areas, (3) the capability to monitor combustible gas concentrations, and (4) the capability to reduce combustible gas concentrations.
The NRC staff's review primarily focused on any impact that the proposed EPU may have on hydrogen release assumptions, and how increases in hydrogen release are mitigated.
The NRC's acceptance criteria for combustible gas control in containment are based on (1) 10 CFR 50.44, insofar as it requires that plants be provided with the capability for controlling combustible gas concentrations in the containment atmosphere; and (2) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is 44 not impaired by the sharing.CD5, insofr a- t re be ..hared am ........e pewer-N niteuantess hteh n-44nt-tt~eeAy-sj-4GDC*-4-1inse fs-4r-a qu#restIa-into the reactor containment-fellewi~ng-posuaeds cdente teensure 
... i ,,e.t4ntegty-.
D-(42A-G oar-ase4t-equ~es to-permitapp.epoiateperodis n
requaire yOO1 esi de ~ed -pamit-a Jdi~estig -flndd t-he-fellewig sentenc.e for-BWAswith Mark Il conami" nt: ,e conaimnttat dotrl ona-nertd tmshoo o otrolhdoe nie-the o-entafnment]-Specific review criteria are contained in SRP Section 6.2.5.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment related to combustible gas and concludes that the plant will continue to have sufficient capabilities consistent with the requirements in 10 CFR 50.44 and draft GDCA4, G-DP=s-&4, 42-1,4,an~4-4,-as discussed above.Therefore, the NRC staff finds the proposed EPU acceptable with respect to combustible gas control in containment.
2.6.5 Containment Heat Removal Requlatorv Evaluation Fan cooler systems, spray systems, and residual heat removal (RHR) systems are provided to remove heat from the containment atmosphere and from the water in the containment wetwell.The NRC staff's review in this area focused on (1) the effects of the proposed EPU on the analyses of the available net positive suctjon head (NPSH) to the containment heat removal system pumps and (2) the analyses of the heat removal capabilities of the spray water system and the fan cooler heat exchangers.
The N RC's acceptance criteria for containment heat removal are based on draft GDCs-41 and 52, insofar as they require that a containment heat removal system be provided, and that its function shall be to prevent exceeding containment design pressure under accident conditions.
e a cnanetha-emoval-system-be-pr-ovide t he qta-cmain-on rcc e lew4eve Specific review criteria are contained in SRP Section 6.2.2, as supplemented by Draft Guide (DG) 1107.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion 45 
* ;' * ; , *.- -..The NRC staff has reviewed the imv ssesaesenprvddb the licensee and concludes that the licensee has adequately addressed the effects of the proposed EPU. The NRC staff finds that the systems will continue to meet draft GDCs-41 and 52 GD4&=-3with respect to rapidly reducing the containment pressure and temperature following a LOCA and maintaining them at acceptably low levels. Therefore, the NRC staff finds the proposed EPU acceptable with respect to containment heat remd6val systems.2.6.6 Secondary Containment Functional Design Regulatory Evaluation The secondary containment structure and supporting systems of dual containment plants are provided to cpllect and process radioactive material that may leak from the primary containment following an accident.
The supporting systems maintain a negative pressure within the secondary containment and process this leakage. The NRC staff's review covered (1) analyses of the pressure and temperature response of the secondary containment following accidents within the primary and secondary containments; (2) analyses of the effects of openings in the secondary containment on the capability of the depressurization and filtration system to establish a negative pressure in a prescribed time; (3) analyses of any primary containment leakage paths that bypass the secondary containment; (4) analyses of the pressure response of the secondary containment resulting from inadvertent depressurization of the primary containment when there is vacuum relief from the secondary containment; and (5) the acceptability of the mass and energy release data used in the analysis.
The NRC staff's review primarily focused on the effects that the proposed EPU may have on the pressure and temperature response and drawdown time of the secondary containment, and the impact this may have on offsite dose. The NRC's acceptance criteria for secondary containment functional design are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCACDC- .,.in..far....
reqio th" S " imporant te-safety-be-designed4e-aeeemwnoate-the-ef rnermab-eperatonermintenaaee 7 4estiag~aandpstttd-a reteerdmm dynamie.effet~~effe -ef uds}4hat-may resu~t-frram-equipment4ailur-esj-and (2) draft GDC-1O, insofar as it requires that reactor containment be designed to sustain the initial effects pf gross equipment failures, such as a large coolant boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain functional capability for as long as the situation requires.
GD 6 zfrasi eurstat ractor-containment-an,-
uneeflt-role~~se~ff-adieaetivty~etee4wimnament---Specific review criteria are contained; in SRP Section 6.2.3.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment related to the secondary containment pressure and temperature transient and the ability of the secondary containment to provide an essentially leak-tight barrier against uncontrolled release of radioactivity to the environment.
The NRC staff concludes that the licensee has adequately accounted for the increase of mass and energy that would result from the proposed EPU and further concludes that the secondary 46 containment and associated systems will continue to po'de an esnilyLa-ih are against the uncontrolled release of radioactivity to the environment following implementation of the proposed EPU. Based on this, the NRC staff also concludes that the secondary containment and associated systems will eontinue to meet the requirements of draft GDCs-1O, 40 and 42.GDc' ad6 Therefore, the NRC staff finds the proposed EPU acceptable with respect to secondary containment functional design..[Adiio aluRc rat*ConTaine" "viw Gned "t es 2.7 Habitability, Filtration, and Ventilation 2.7.1 Control Room Habitability System Regqulatory Evaluation The NRC staff reviewed the control room habitability system and control building layout and structures to ensure that plant operators are adequately protected from the effects of accidental releases of toxic and radioactive gases. A further objective of the NRC staff's review was to ensure that the control room can be maintained as the backup center from which technical support center personnel can safely operate in the case of an accident.
The NRC staff's review focused on the effects of the proposed EPU on radiation doses, toxic gas concentrations, and estimates of dispersion of airborne contamination.
The NRC's acceptance criteria for the control room habitability system are based on (1) draft GDC-40, insofar as it requires that, protection for e'ngineered safety features shall be provided againstdynamic effects and missiles that might result from plant equipment failures-SS~s-importan4t-e-safety-be-dcndtoaccommodato tho offocts of and to bo comptbl wit th ovrnmontal conditiors assesiated-wfth-postutated-eeeiden~tirncludin&sect;Ahe-effests-oef-h~e~ese-ef4oxis ases; and (2)final GDC-1 9 and 10 CFR 50.67, insofar as it requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent, to any part of the body, for the duration of the accident.
Specific review criteria are contained in SRP Section 6.4 and other guidance provided in Matrix 7 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment related to the effects of the proposed EPU on the ability of the control room habitability system to protect plant operators against the effects of accidental releases of toxic and radioactive gases. The NRC staff concludes that the licensee has adequately accounted for the increase of toxic and radioactive gases that would result from the proposed EPU. The NRC staff further concludes that the control room habitability system will continue to provide the required protection following implementation of the proposed EPU. Based on this, the NRC staff concludes that the control room habitability system will continue to meet the requirements of draft GDC-40 and final GDJC-1 9 and 10 CFR 50:67. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the control room habitability system.47 2.7.2 Engineered Safety Feature Atmosphere Cleanup Recqulatorv Evaluation 
, ESE atmosphere cleanup systems 'are designed for fissiqn prbduct removal in post-accident environments.
These systems generally include primary systems (e.g., in-containment recirculation) and secondary systems (eg., standby gas treatment systems and emergency or post-accident air-cleaning systems) for the fuel-handling building, control room, shield buIlding, and areas containing ESF components.
For each ESF atmosphere cleanup system, the NRC staff's review focused on the effects of the proposed EPU on system functional design, environmental design, and provisions to preclude temperatures in the adsorber section from exceeding design limits. The NRC's acceptance criteria for ESF atmosphere cleanup systems are based on (1) final GDC-19 and 10 CFR 50.67, insofar as it requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent, to any part of the body, for the duration of the accident; (2) ODO -411,insofai-as4t-rerues4hat-system~s4ooeto-eeflssien prmded4e-reuee-the-eecn " io-n --uliisiea-pfduets-F4eesede4-thenevironment GDC-70, insofar as it requires that the plant maintain control over the radioactive effluents during normal operation and for any transient situation-;
and GD 1,isfar as reqire that systms that maysonai radioact;vit b desigedt-assure--adequate-safet y une'omladpcuae acien od in; od-(34)GDG-64draft GDC-17, insofar as it requires that means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences (AOOs), and postulated accidents.
Specific review criteria are contained in SRP Section 6.5.1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the ESE atmosphere cleanup systems. The NRC staff concludes that the licensee has adequately accounted for the increase of fission products and changes in expected environmental conditions that would result from the proposed EPU, and the NRC staff further concludes that the ESF atmosphere cleanup systems will continue to provide adequate fission product removal in postaccident environments following implementation of the proposed EPU.Based on this, the NRC staff concludes that the ESF atmosphere cleanup systems will continue to meet the requirements of 10 CFR 50.67, final GDC-19 and draft GDCs-17 and 70. GD~s-4-t4-,44 -t--an464-Therefore, the NRC staff finds the proposed EPU acceptable with respect to the ESF atmosphere cleanup systems.2.7.3 Control Room Area Ventilation System Regqulatory Evaluation The function of the control room area ventilation system (CRAVS) is to provide a controlled environment for the comfort and safety of control room personnel and to support the operability of control room components during normal operation, AOOs, and OBA conditions.
The NRC's 48 review of the CRAVS focused on the effects EPU will have on the functional performance of safety-related portions of the system. The review included the effects of radiation, combustion, and other toxic products; and the expected environmental conditions in areas served by the CRAVS. The NRC's acceptance criteria for the CRAVS are based on'(1) draft GDC..40GG, insofar as it requires that protection for engineered safety features be provided against dynamic effects and missiles that might result from plant equipment failu fety-be-designe44e-aeeemmoedat ......c..f.an to4 be" , eide~ts;(2) final GDC-19, insofar as it requires that adequate radiation protection be provided to permit access and' occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body; or its equivalent to any part of the body, for the duration of the accident; and (3) draft GDC-70, insofar as it requires that the plant design include means to control the release of radioactive effluents.
Specific review criteria are contained in SRP Section 9.4.1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the ability of the CRAVS to provide a controlled environment for the comfort and safety of control room personnel and to support the operability of control room components.
The NRC staff concludes that the licensee has adequately accounted for the increase of toxic and radioactive gases that would result from a DBA under the conditions of the proposed EPU, and associated changes to parameters affecting environmental conditions for control room personnel and equipment.
Accordingly, the NRC staff concludes that the CRAVS will continue to provide an acceptable control room environment for safe operation of the plant following implementation of the proposed EPU. The NRC staff also concludes that the system will continue to suitably control the release of gaseous radioactive effluents to the environment."Based on this, the NRC staff concludes that the CRAVS will continue to meet the requirements of final GDC-19 and draft GDCs-40 and 70s-4,-44,and4-8.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to the CRAVS.2.7.4 Spent Fuel Pool Area Ventilation System R ePuatr--valuat4ie
[Section 2.7.4 is not applicable to BFN]-The4 a n 0 Abeef~epent-fuet-pee-ea-venti~atien-system-QSF PA h ffes fth esed-EP-on4hJete-funetieefa--
pefrac ftesft eltdproso -The NRC's accapacocie~a4ar4he-which conai eadoaetity-b~e-de.
riate-seefinemenbd-enament peslfie-review 49 a.-r EliInert technical evaluation.
The technical evaluation should (1) clearly explain why the-peposed-ohangcn satisfy each ofthe requirements in the-wgulato~y-evatuation-and Conclusion r..quirement.
of GD. c and 61. There..ore,.
the. NRC&#xa2; st-ff finds the proposed-EPU-aeetble-2.7.5 yen-Rdaste.Are-and Turbine...re..Reactor, Turbine, and Radwaste Building Ventilation Systems 245-Regqulatory Evaluation The function of the ........, ...Turbine and Radwaste Building vVentilation eSystem to maintain ventilation in the uiir ndrdat eqimn an turbine..a...sreactor, turbine, and radwaste buildings to permit personnel access, and control the concentration of airborne radioactive material in these areas during normal operation, during AQOs, and after postulated accidents.
The NRC staff's review focused on the effects of the proposed EPU on the functional performance of the safety-related portions of these systems. The NRC's acceptance criteria for the ARV ndTvsystems are based on draft GDC-7OGQ-6, insofar as it requires that the plant design include means to control the release of radioactive effluents.
Specific review criteria are contained in SRP Sections 9.4.3 and 9.4.4.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the ARAVS-and-T4ASReactor, Turbine, and Radwaste Building Ventilation System. The NRC staff concludes that the licensee has adequately accounted for the effects of the proposed EPU on the capability of these systems to maintain ventilation in the aux4+iar-y-aend-adweste-eqimn areas. and. in.. the tu.. +,e-a,. turbine, and radwaste buildings to permit personnel access, control the concentration of airborne radioactive material in these areas, and control release of gaseous radioactive effluents to the environment.
Based on this, the NRC staff concludes that the ARA-V-S-an4-T-AVSsystems will continue to meet the requirements of draft Therefore, the NRC staff finds the proposed EPU acceptable with respect to the Reactor, Turbine, and Radwaste Building Ventilation SystemARAVS and the:TAVS.50 247142.7.6 Engineered Safety Feature Ventilation System Requlatory Evaluation The function of the engineered safety feature ventilation system (ESFVS) is to provide a suitable and controlled environment for ESF components following certain anticipated transients and DBAs. The NRC staff's review for the ESFVS focused on the effects of the proposed EPU on the functional performance of the safety-related portions of the system. The NRC staff's review also covered (1) the ability of the ESF equipment in the areas being serviced by the ventilation system to function under degraded ESFVS performance; (2) the capability of the ESFVS to circulate sufficient air to prevent accumulation of flammable or explosive gas or fuel-vapor mixtures from components (e.g., storage batteries and stored fuel); and (3) the capability of the ESFVS to control airborne particulate material (dust) accumulation.
The NRC's acceptance criteria for the ESFVS are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA;GDC 1, incofnr as-it-iequies-that fety-e-desi44ed4e~aeomed e-e he-effeot-fade-e oomat~~wh~eevkonmenat dteeseeatedwit h-oe~a~oel~atieo,-maPitenanee, testin, an'otltdacdente;-(2) final GDC-17, insofar as it requires onsite and offsite electric power systems be provided to permit functioning of SSCs important to safety; and (3)draft insofar as it requires that the plant design include means to control the release of radioactive effluents.
Specific review criteria are contained in SRP Section 9.4.5.Technical Evaluation Elnsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section,]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the ESFVS. The NRC staff concludes that the licensee has adequately accounted for the effects of the proposed EPU on the ability of the ESFVS to provide a suitable and controlled environment for ESF components.
The NRC staff further concludes that the ESFVS will continue to assure a suitable environment for the ESF components following implementation of the proposed EPU. The NRC staff also concludes that the ESFVS will continue to suitably control the release of gaseous radioactive effluents to the environment following implementation of the proposed E2PU. Based on this, the NRC staff concludes that the ESFVS will continue to meet the requirements of final GDC-1 7 and draft G DCs-40, 42 and 70.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to the ESFVS.[tnsertRegu late~y-Evahuatiern-Teohn~iaa-EvatuatieoP-ad-Gel usi 2.8 Reactor Systems 2.8.1 Fuel System Design Regulatory Evaluation S51 
-- ------ ---The fuel system consists of arrays of fuel rods, .burnable poison rods,.spacer grids and springs, end plates, channel boxes, and reactivity control rods. The NRC staff reviewed the fuel system to ensure that (1) the fuel system is not damaged. s~eulo~~~
operation and AOOs, (2) fuel system damage is never so severe as to prevent conftrl r~d 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 NRC staff's 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 NRC's acceptance criteria are based on (1) 10 CFR 50.46, insofar as it establishes standards for the calculation of emergency core cooling system (ECCS) performance and acceptance criteria for that calculated performance: 
(2) final GDC-10, insofar as it requires that the reactor core be designed with appropriate margin to assure that SAFOLs are not exceeded during any condition of normal operation, including the effects of AOOs; (3) final G DC-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; end (4) final 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.Specific review criteria are contained in SRP Section 4.2 and other guidance provided in Matrix 8 of RS-001.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the fuel system design of the fuel assemblies, control systems, and reactor core. The NRC staff concludes that the licensee has adequately accounted for the effects of the proposed EPU on the fuel system and demonstrated that (1) the fuel system will not be damaged as a result of normal operation and AQOs, (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) coolability will always be maintained.
Based on this, the NRC staff concludes that the fuel system and associated analyses will continue to meet the requirements of 10 CFR 50.46, final GDC-1O, GDC-27, and GDC-35 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the fuel system design.2.8.2 Nuclear Design Regulatory Evaluation The NRC staff reviewed the nuclear design of the fuel assemblies, control systems, and reactor core to ensure that fuel design limits will not be exceeded during normal operation and anticipated operational transients, and that the effects of postulated reactivity accidents will not cause significant damage to the RCPB or impair the capability to cool the core. The NRC staffs review covered core power distribution, reactivity coefficients, reactivity control requirements and control provisions, control rod patterns and reactivity worths, criticality, burnup, and vessel irradiation.
The NRC's acceptance criteria are based on (1) final GDC-10, insofar as it requires 52 that the reactor core be to assure that SAFDLs are not exceeded during any condijion of normal operatio~n, including the effects of AOOs; (2) draft GDC-8, insofar as it requires that the reactor core be designed so that the overall power coefficient in the power operating range shall not be positive; G004-14,-nsofar~as4t-requies.tht.thereactr.cor.be.dsigne..o.tat.th nt..effect..of the..........
prmt-(3 inhrent nucla 7GQG.:-12, insofar as it requires that the core design shall ensure that power oscillations which could cause damage in excess of acceptable fuel damage limits are not possible or can be readily suppressed; inscore....be .designedh 
....to..as.sucre that-pewe-eeeiloins hch can... reul in conito. , xceed.. ,ing, SOrs, are. ..... not pas~l or... ;""^oawbe-eably-an teadtlydeteeteedand-suppressedi-(4) d raft GD~s -12 a nd 13 in sofrar as they require that instrumentation and controls be provided as required to monitor and maintain variables within prescribed operating ranges through the core life;G-DC-.3-, rangesj (6) draft GDCs-14 and 15, insofar as they require that the protection system be designed to initiate the reactivity control systems automatically to prevent or suppress conditions thait could result in exceeding acceptable fuel damage limits and to initiate operation of ESFs under accident situations;GDG=-:20,-
eofar-as4tequires~hat-the-that-aeetbe fue deig liit ar not..exceeded as a result of draft GDC-31, insofar as it requires that the reactivity control systems be capable of sustaining any single malfunction without causing a reactivity transient which could result in exceeding acceptable fuel damage limits; preetinssembe-designed.to assur tha SAOsa o xeeded feeany-ing-mafnto-fthe-reactivity-e~nt491-systemsj; (7) draft GDCs-27 and 28 insofar as they require that at least two independent reactivity control systems be provided, with both systems capable of making and holding the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits; (8) draft GIJCs-29 and 30, insofar as they require that at least one of the reactivity control systems be capable of making and holding the core subcritical under any condition sufficiently fast to prevent exceeding acceptable fuel damage limits; GD-265nsofeeet-r-equk~e-that-final 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 ECOS, 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 (910) draft GDC-32, insofar as it requires that limits, which include considerable margin, be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling. GC2," inoa "si reufres-hath reactivity..
cotrlsytmsb stouresoreother-eie ceore-Specific review criteria are contained in SRP Section 4.3 and other guidance provided in Matrix 8 of RS-0O1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the 53 proposed changes satisfy each oft e requ emrregulatory evaluation and (2)provide a clear link to the conclusibns reached by the NRC staff, as documented in the conclusion section.]C.onclusion 2 ."i The NRC staff has reviewed the licensee's analyses related to the effect of the proposed EPU on the nuclear design of the fuel assemblies, control systems, and reactor core. The NRC staff concludes that the licensee has adequately accounted forethe effects of the proposed EPU on the nuclear design and has demonstrated that the fuel design Jimits wilJ not be exceeded during normal or anticipated operational transients, and that 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, the NRC staff concludes that the nuclear design of the fuel assemblies, control systems, and reactor core will continue to meet the applicable requirements of final GDCs-l0 and 27, and draft GDCs-7, 8, 12, 13, 14, 15, 27, 23, 29, 30, 31 and 32.. .. 00c1,1,1,1,2,2-26,2,ed2.Therefore, the NRC staff finds the proposed EPU acceptable with respect to the nuclear design.2.8.3 Thermal and Hydraulic Design Regulatory Evaluation The NRC staff reviewed-the thermal and hydraulic design of the core and the RCS to confirm that 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 AQOs, 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 anticipated transients without scram (ATWS) events. The NRC's acceptance criteria are based on (1) final GDC-1 0, 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 AQOs; and (2) draft GDC-7, insofar as it requires that the core design shall ensure that power oscillations which could cause damage in excess of acceptable fuel damage limits are not possible or can be readily aemcatcd cclntreonro, 'n rteto sytm b -~g oascurehat-powee-eolnhtiens-u-wia-h F-DL-s-,areoet-possible-er-an-reliably-and readily be detected and cupprecsed.
Specific review criteria are contained in SRP Section 4.4 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation Elnsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the thermal and hydraulic design of the core and the RCS. The NRC staff concludes that the licensee has adequately accounted for the effects of the proposed EPU on the thermal and hydraulic design and demonstrated that the design (1) has been accomplished using acceptable 54 analytical methods, (2) is [equivalenfi1btio from] proven designs, (3)provides acceptable margins of safety, from conditions thatF~ould lead to fuel damage during normal reactor operation and AOOs, and (4) is not susceptible to thermal-hydraulic instability.
The NRC staff further concludes that the licensee has adequately accounted for the effects of the proposed EPU on the hydraulic loads onthe core and RCS components.
Based on this, the NRC staff concludes that the thermal and hydraulic design will continue to meet the requirements of final GDCs--1O and draft GDC-7t2 following implementation of the proposed EPU. Therefore, the NRC staff 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 The NRC staff's review covered the functional performance of the control rod drive system (CRDS) to confirm that the system can affecteffest a safe shutdown, respond within acceptable limits during AOOs, and prevent or mitigate the consequences of postulated accidents.
The review also covered the CRDS coolingsystem to ensure that it will continue to meet its design requirements.
The NRC's acceptance criteria are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ES~s against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA;GDC-4--
inoao -i-ourc htS~ impotan to...... saot b .. do e to.. accorrmeda~te-hefet-fh nvr 
+er44ateperatioa-, maintnnoretin&sect;anpestuated-aoodentsi (2) draft G.DC-26, insofar as it requires that the protection system be designed to fail into a safe state; GD-22-,-ieofar-os-Ueruies-that-the-proteetion-systern-be-designed~o~aNl-nto-a-safe-state,-(3) d raft G DC-3 1, insofar as it requires that the reactivity control systems be capable of sustaining any single malfunction without causing a reactivity transient which could result in exceeding acceptable fuel damage limits ;GlC-, -Thefeae-as qufres-kt-hae-pr-oteotion-system-be-for-an y-singjmaune of-4he-reaetity o.on-toksystemsj (4) draft GDCs-27 and 28 insofar as they require that at least two independent reactivity control systems be provided, with both systems capable of making and holding the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits; (5) draft GDCs-29 and 30, insofar as they require that at least one of the reactivity control systems be capable of making and holding the core subcritical under any condition sufficiently fast to prevent exceeding acceptable fuel damage limits; GDGC2,ieofa....i -reouirocsthat-twa-iFdependent-reaotivt y-so14 "" .. ,,q, Gfermal-powee0-4angese
(,5) GDG27,-nsofar-as~~ure~thathe~eeativty-ent-lsy sterns-te-desigd 4e-ae rer4ty-eharnjes-under-pstu s- ppop iate-marg-n efr-stucWo toassure-he-aapabiitye-to-oe4he-eo~e~s-mait(nd6) draft GDC-32, insofar as it requires that limits, which include considerable margin, be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a)rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling; C C2,insf -ast-iequifesthat~h-eaoftviy-sent~o4-systeme-be-desg ed-ta ass 4aheefe~sef-postulat edreaetivyaeiet-an-netthereastlt--dama~e-to-4he-reaeto*-ve~ssel-inte~oals-soe .s-.te-signif4oantly--mpaia~thecapabi~it-y.to-ceotbe-oer-e;-{-7..- Fneofa t t-- --ad aetv 55 
* * *1. .cxtrome= high eetofA~ and (87) 10 CFR 50.62(c)(3), insofar as it requires that all BWRs have an alternate rod injection (ARI)system diverse from the reactor trip system, and that the ARI system have redundant scram air header exhaust valves. Specific review criteria are contained in SRP Section 4.6.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the functional design of the CR0S. The NRC staff concludes that the licensee has adequately accounted for the effects of the proposed EPU on the system and demonstrated that the system's ability to affecteffeot a safe shutdown, respond within acceptable limits, and prevent or mitigate the consequences of postulated accidents will be maintained following the implementation of the proposed EPU. The NRC staff further concludes that the licensee has demonstrated that sufficient cooling exists to ensure the system's design bases will continue to be followed upon implementation of the proposed EPU. Based on this, the NRC staff concludes that the fuel system and associated analyses will continue to meet the requirements of draft GDCs-26, 27, 28, 29, 30, 31, 32, 40 and 42, 25, 26,27,-2, nd2, and 10 CFR 50.62(c)(3) following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the functional design of the ORDS.2.8.4.20Overpressure Protection During Power Operation Regqulatory Evaluation Overpressure protection for the RCPB during power operation is provided by relief and safety valves and the reactor protection system. The NRC staff's review covered relief and safety valves on the main steamlines and piping from these valves to the suppression pool. The NRC's acceptance criteria are based on (1) draft GDC-9, insofar as it requires that the RCPB be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime; in.....ofar,,., as. it reqfuires-that-the-RCSadaooite-wlia%
t-ooa nd-pet eetien-ysem b dsine wit h-suffi~eiet-marg~nto -assur that th-einsniin-f~e-RPB ae nt-excedddrn n eend4Ie ........maloperatio, incldn A~;and (2) draft GDCs-33, 34, and 35G10,C-3t, insofar as itthey requires that the RCPB be designed with sufficient margin to assure that it behaves in a nonbrittle manner and that the probability of rapidly propagating fracture is minimized.
Specific review criteria are contained in SRP Section 5.2.2.Technical Evaluation Elnsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the overpressure protection capability of the plant during power operation.
The NRC staff 56 concludes that the licensee has (1) adequately accounted fo the effects of the proposed EPU on pressurization events and overpressure protection features and (2) demonstrated that the plant will continue to have sufficient pressure relief capacity to ensure that pressure limits are not exceeded.
Based on this, the NRC staff concludes that the overpressure protection features will continue to meet draft GDCs-9, 33, 34, and 35 ODes 15 and 31-following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to overpressure protection during power operation.
2.8.4.3 Reactor Core Isolation Cooling System Regqulatory Evaluation The reactor core isolation cooling (RCIC) system serves as a standby source of cooling water to provide a limited decay heat removal capability whenever the main feedwater 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 water supply for the RCIC system comes from the condensate storage tank, with a secondary supply from the suppression pool. The NRC staffs review covered the effect of the proposed EPU on the functional capability of the system. The NRC's acceptance criteria are based on (1) draft GDC-40 insofar as it requires that protection be provided for ESFs against dynamic effects; GD,4,-4nsefar--a4Ptejukree-..p.. tant to.... cafot et protet ... e .....dyneieef;e (2) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is not impaired by the shari ng;GQ&-5Jneefar-as-it-equiret-aSSGs-impe~ae+t~e-safety-nobbe-
..hared amn n uc!oar-power unt.nl...anb oontao ththrn ............
ability4o-peffomtsafety-funetion,-{3-)
GDG-2Q,-ineefar-ast-r~equJires~hat-the-preteet on-ed-a reastivty 
.-ent high-preobabi~ty-ofceomplishing-their safety* functin in... ovont of ,",,.. (.1) CCC 3, i nsoa,. i rgio ht ytmt provide- eaetoeleeant-mak -e r-pmhteetionagaiast -sma4-beaks-n n-he-RCPB-epovdeso the4uedeig~nimits-.a~re~et-axoceeded+/--5}GD-35nef ar-as4 t~u -a-ros~d uoaheat-uaN.eat *from-tlhe-rfea,-tor-eer-e-Fat-a-tesueh-that-SAF-D~sefld4he-desiwGn~onldltensef~he44GP-Ba4<e-net-(63) draft GDCs-51 and 57, insofar as they require that piping systems penetrating containment be designed with appropriate features as necessary to protect from an accidental rupture outside containment and the capability to periodically test the operability of the isolation valves to determine if valve leakage is within acceptable l im its; GD5-4I4esofar-as~trequi~ee-ha~t-pipig-systems-penetkatin&sect; ontalnmnenat--etlesigle with.... the capailit to teriodecy te...t th operabilit of the. isebatien alvc t detorminc if valve leakg swti ..cp..~ limts;and 
(.74) 10 CFR 50.63, insofar as it requires that the plant withstand and recover from an SBO of a specified duration.
Specific review criteria are contained in SRP Section 5.4.6.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]-
Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the ability of the RCIC system to provide decay heat removal following an isolation of main feedwater event and a station blackout event and the ability of the system to provide makeup to the core following a small break in the RCPB. The NRC staff concludes that the licensee has adequately accounted for the effects of the proposed EPU on these events and demonstrated 57 that the RCIC system will cotnu heat removal and makeup for these events following implementation of the proposed EPU. Based on this, the NRC staff concludes that the RCIC system will continue to meet the requirements of draft GDCs-4, 40, 51, and 57, .........
29r3+3 rd--4, and 10 CER 50.63 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the ROIC system.2.8.4.4 Residual Heat Removal System 'Reaulatorv Evaluation The RHR system is used to cool down the ROS following shutdown.
The RHR system is typically a low pressure system which takes over the shutdown cooling function when the RCS temperature is reduced. The NRC staff's review covered 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. The NRC's acceptance criteria are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against dynamic effects; and (2)draft GDC-4, insofar as reactor facilities shall not share systems or corn ponents unless it is shown safety is not impaired by the sharing. SOC 5, incofa as i..... r.. cqu.. ,i..........t-SS e-,
Specific review criteria are contained in SRP Section 5.4.7 and other guidance provided in Matrix 8 of RS-001.Technical Evaluation Elnsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the RHR system. The NRC staff concludes that the licensee has adequately accounted for the effects of the proposed EPU on the system and demonstrated that the RHR system will maintain its ability to cool the RCS following shutdown and provide decay heat removal. Based on this, the NRC staff concludes that the RHR system will continue to meet the requirements of draft GDCs-4, 40 and 42 SO~t 1., 5, an 31, following implementation of the proposed EPU.Therefore, the NRC staff finds the proposed EPU acceptable with respect to the RHR system.2.8.4.5 Standby Liquid Control System Reciulatory Evaluation The standby liquid control system (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 NRC staff's review covered the effect of the proposed EPU on the functional capability of the system to deliver the required amount of boron solution into the reactor. The NRC's acceptance criteria are based on (1) draft GDCs-27 and 28 insofar as they require that at least two independent reactivity control systems be provided, with both systems capable of making and holding the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits; (2) draft GDCs-29 and 30, insofar as they require that at least one of the reactivity 58 control systems be capable of makn an hlngi&#xb6;6re subcritical under any condition sufficiently fast to prevent exceeding acceptable fuel damage limits;GD-6 inees res4ha~wo4ndependeat-ieao~it f dffoen4t dsign--apb~e-ofaJl-hodifl4hl e4eeetersubedtieal-th cold condit~on; (2) GDC 27v, insofar as it requir. that, he r..c..it control...
cctmc. have,. a..reattivtty-.eheeges-undepetuated-aceiden-edtinsj,-an4-(3) 10 CFR 5O.62(c)(4), insofar as it requires that the SLOS be capable of reliably injecting a borated water solution into the reactor pressure vessel at a boron concentration, boron enrichment, and flow rate that provides aset level of reactivty control, and [DEPENDING...ON..............
PER.MIT DAT O ORIG4NAL-,QESI 
..........-
syeteminitiate-auitematiealiy.
Specific review criteria are contained in SRP Section 9.3.5 and other guidance provided in Matrix 8 of RS-001.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the SLCS and concludes that the licensee has adequately accounted for the effects of the proposed EPU on the system and demonstrated that the system will continue to provide the function of reactivity control independent of the control rod system following implementation of the proposed EPU. Based on this, the NRC staff concludes that the SLCS will continue to meet the requirements of draft GDCs-27, 28, 29 and 30, GDCs 26 and 27, and 10 CFR 50.62(c)(4) following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the SLCS.2.8.5 Accident and Transient Analyses 2.8.5.1.1 Decrease in Feedwater Temperature, Increase in Feedwater Flow, Increase in Steam Flow, and Inadvertent Opening of a Main Steam Relief or Safety Valve Regqulatory 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.
The NRC staff's review covered (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 reactor protection system, (6) operator actions, and (7) the results of the transient analyses.
The NRC's acceptance criteria are based on (1) draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; w~ith-mar-gkvsuffioieet-oeneswreh akthedtesign6eonditieon-of.-heR&P-8are-t-exoeeded-during 
-any-condit4oef-nermjaorationj-(32) draft GDCs-14 and 15, insofar as they require that 59 the core protection system be de prevent or suppress conditions that could result in exceeding acceptable fuel damage limits and that protection systems be provided for sensing accident situations and initiating the operation of necessary ESFs; GDG-2O-0-iseofar-as4t-eqir~es-t~hat-the-eaotor-protection-nermekep.ea tlen 7 (43) draft GDC-29 insofar as they require that a reactivity control system be provided capable of preventtng exceeding acceptable fuel damage limits. soC 26, insofar as it requires that a reactivty control be.. provided-,='''
and'norwal-eper-atien~inek~ng-AOQCSAFOS aeno e~edec-.Specific review criteria are contained in SRP Section 15.1.1-4 and other guidance provided in Matrix 8of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the excess heat removal events described above and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the SAFOLs and the RCPB pressure limits will not be exceeded as a result of these events. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft GDCs-6, 14, 15, and 29 GD , 1 ,152, an 6following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the events stated.2.8.5.2 Decrease in Heet Removal by the Secondary System 2.8.5.2.1 Loss of External Load; Turbine Trip; Loss of Condenser Vacuum;Closure of Main Steam isolation Valve; and Steam Pressure Regulator 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.
The NRC staffs review covered 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 NRC's acceptance criteria are based on (1) draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acc:eptable fuel damage limits; SDC 10, insofar as it that.,= the.,,
draft GDC-29 insofar bs it requires that a reactivity control system be provided capable of making the core subcritical under any conditions (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. GO2~60 mcfar a.. it r..uir. t, h.at a reactivtw otrlcsemb r'ddrand be capable of relia;bly'....... re-n~et-exeeede4,-.Specific review criteria are contained in SRP Section 15.2.1-5 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the decrease in heat removal events described above and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the SAFDLs and the ROPE pressure limits will not be exceeded as a result of these events. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft ,s-GDCs-6 and 29 implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the events stated.2.8.5.2.2 Loss of Nonemergency AC Power to the Station Auxiliaries Regqulatory Evaluation The loss of nonemergency 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.
The NRC staff's review covered (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 NRC's acceptance criteria are based on (1) draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; G&GC Ins......as-tr-a th C edsge-ih app ..........................
AFDL ar ot-excedd uin......pra~es AO',c-. (2 GC 5 t equre that... the.... RCS....and.
it ssociated auxiliar"y ysems-be des ged-i mt4n-ar fir en en i-hat-thedesign-eenditiowfth-G4 are-net-ex-ceeded-durin -any-cond-tierff-nermefo-perationj-and-(32) draft G DC-29 insofar as it requires that a reactivity control system be provided capable of making the core subcritical undler any conditions (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. ~ ie~a-a rcasiiycnrlcetent-pevdedadt-ape e-.of-ealy-onteffin&sect;-he-ra-te-ef-veaot~vty-chnes~teeneur4h r-neeaitenbnso ,eraeperatien44daelnig AQOs, SAF DLc am-e nt-exeeeded.
Specific review criteria are contained in SRP Section 15.2.6 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the 61 
'I t~i .s4gqq&.*
I*. &I0I.IttrIItoI~
conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the loss of nonemergency ac power to station auxiliaries event and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has 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 this event. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft GDCs-6 and !0, 15,an~d-2&following.
implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the loss of nonemergency ac power to station auxiliaries event.2.8.5.2.3 Loss of Normal Feedwater Flow Recqulatory Evaluation A loss of normal feedwater flow could occur from pump failures, valve malfunctions, or a LOOP.Loss of feedwater flow results in an increase in reactor coolant temperature and pressure which eventually requires a reactor trip to prevent fuel damage. Decay heat must be transferred from fuel following a loss of normal feedwater flow. Reactor protection and safety systems are actuated to provide this function and mitigate other aspects of the transient.
The NRC staffs review covered (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 NRC's acceptance criteria are based on (1) draft GDC-6, insofar as It requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; SOC 10 nc;,ofar ac. it r.qu.ro that the, R, S be designed,, ,with aiek4ateg-maO gir;-(2) sOC 5, n-oara-sA it r-suire- that th, ee RSan- t associat-eduxlir nyte (3,2) draft GDC-29 insofar as it requires that a reactivity control system be provided capable of making the core subcritical under any conditions (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits.
syste-.-.o.de,.ndb.cpaleofreial cn.........the rate....of rativity ..h..nge to-eesufe-hat-u~fdeF-eefldjie3o 
!ng,^AO, , SAFO~ r Specific review criteria are contained in SRP Section 15.2.7 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the loss of normal feedwater flow event and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the SAFDLs and the RCPB pressure limits will not be 62 exceeded as a result of the loss of normal ,wY;1 ?as? onthsteNRsaf concludes that the plant will continue to nmeet the requirements d~f draft GDCs-6 and 2943.lG-404an-6following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the loss of normal feedwater flow event.2.8.6.3 Decrease in Reactor Coolant System Flow 2.8.5.3.1 Loss of Forced Reactor Coolant Flow Regulatory Evaluation A decrease in reactor coolant flaw 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.
The NRC staff's review covered (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 reactor protection system, (6) operator actions, and (7) the results of the transient analyses.
The NRC's acceptance criteria are based on (1)draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; G-9)C-,4.0 5 inse fart-asrequres~hathe RCSb eined a~p epfiate-mrin-te-ernsure-hat-AFDS ar o ocoedd u r4ng~nmal-oper-atioas-4noludg-AQOsi(}GG&nasfaF-as 
;t........u...ee"-
that4he-C=GS-andqte-asscated-auwliareystemse-bedesigeditma -sufee-en epeiatienjd(3,2) draft GDC-29 insofar as it requires that a reactivity control system be provided capable of making the core subcritical under any conditions (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage himits,.G-&isofar at it requirc that....r..ct...ty control ....tom" bo p"vdd an'ecpbe of feliab~y-cont'otling*-the*r-ate-ef-reactivty-c.haIn&sect;es-to-en surethatm.
e-ueeenditions-of nerm at-operation, including,,, AC, SAF^ero "... not .exeeded Specific review criteria are contained in SRP Section 15.3.1- 2 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the decrease in reactor coolant flow event and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models.The NRC staff further concludes that the licensee has 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 this event. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft GDCs-6, and 29 GD4Cs 10, 15, anid 26 following.
implementation~of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the decrease in reactor coolant flow event.2.8.5.3.2 Reactor Recirculation Pump Rotor Seizure and Reactor Recirculation Pump Shaft Break 63 Reciulatorv 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 ,floQw while the reactor is at power results in a degradation of core heat transfer which could result in fuiel 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 either case, reactor protection and safety systems are actuated to mitigate the transient.
The NRC staff's review covered (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 reactor protection system, (6) operator actions, and (7) the results of the transient analyses.
The NRC's acceptance criteria are based on (1) final 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 ECOS, 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) draft GDC-32, insofar as it requires that limits, which include considerable margin, be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling; SO;;C ,28, isoar-asit-....qukes hat &#xf7;he, eetesseebeindto sueta h effects of postlt'd reacti.it accident...
can..neithe~-resu~t-iame4e-te4Ae-RG:P-B-re~etere4har4tedkeea4eldner-dstet4he-aore-4tis-eeIte-eee~and (3) draft GDCs-33, 34, and 35, insofar as they require 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 rapidly propagating fractures is minimized.
prepagatm-faoture4s-mniiMz.e4-.Specific review criteria are contained in SRP Section 15.3.3-4 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain whY the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the sudden decrease in core coolant flow events and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated 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 nonbrittle manner, the probability of propagating fracture of the RCPB is minimized, and adequate core cooling will be provided.
Based on this, the NRC staff concludes that the plant will continue to meet the requirements of final GDC-27 and draft GDCs-32, 33, 34, and 35 GOCs 27, 28, and 31 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the sudden decrease in core coolant flow 64 events.2.8.5.4 Reactivity and Power Distribution Anomalies 2.8.5.4.1 Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or 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.
The NRC staffs review covered (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) th6 results of the transient analyses.
The NRC's acceptance criteria are based on (1) final GDC-1O, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations, including AQOs; (2) draft GDCs-14 and 15, insofar as they require that the core protection systems be designed to act automatically to prevent or suppress conditions that could result in exceeding acceptable fuel damage limits and that protection systems be provided for sensing accident situations and initiating the operation of necessary ESFs; dsgae ensar~e-thePA~
s-ere-e-ex~eeeda~sisuW-4fAOOsi-and (3) draft GDC-31, insofar as it requires that the reactivity control systems be capable of sustaining any single malfunction without causing a reactivity transient which could result in exceeding acceptable fuel damage limits. 000 25, incofar aa~-iqueha-te-teotien-system-e Specific review criteria are contained in SRP Section 15.4.1I and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the uncontrolled control rod assembly withdrawal from a subcritical or low power startup condition and concludes that the licensee's analyses have adequately accounted for the changes in core design necessary for operation of the plant at the proposed power level. The NRC staff also concludes that the licensee's analyses were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure the SAFDLs are not exceeded.
Based on this, the NRC staff concludes that the plant will continue to meet the requirements of final GDC-1 0 and draft GDCs-14, 15, and  44, 23, an4.25SfOllowing implementation of the proposed EPU. Therefore, the NRC staff 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 Uncontrolled Control Rod Assembly Withdrawal at Power Repq ulatorv Evaluation 65 
,rJ ,, ir "' ,.An uncontrolled control rod asm *pwraybe caused by a malfunction of the reactor control or rod control systemsyj his wjtl~drawpt.wi l.uncontrollably add positive reactivity to the reactor core, resulting in a power excursion.
The NRC staff's review covered (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 ana y~sis, (4) the analytical methods and computer codes used, and (5) the results of the associated analyses.
The NRC's acceptance criteria are based on (1) final 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; draft GDCs-1 4 and 15, insofar as they require that the core protection systems be designed to act automatically to prevent or suppress conditions that could result in exceeding acceptable fuel damage limits and that protection systems be provided for sensing accident situations and initiating the operation of necessary ESFs; GDO-20O eactinitys control sytestoensr taf Qsi;and (3) draft GDC-31, insofar as it requires that the reactivity control systems be capable of sustaining any single malfunction without causing a reactivity transient which could result in exceeding acceptable fuel damage limits. G.D6-2i-so far-as-itequires~hat-t ef-the pecific review criteria are contained in SRP Section 15.4.2 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff,as~
documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the uncontrolled control rod assembly withdrawal at power event and concludes that the licensee's analyses have adequately accounted for the changes in core design required for operation of the plant at the proposed power level. The NRC staff also concludes that the I~cerisee's analyses were performed using acceptable analytical models. The NRC staff furthet Cbn(ludles that the licensee has demonstrated that the reactor protection and safety systd6ts will continue to ensure the SAFDLs are not exceeded.
Based on this, the NRC staff concludes that the plant will continue to meet the requirements of final GDC-1 0 and draft G DCs-1 4, 15, and 31 GD~s-14-0-ad&
following implementation of the proposed EPU. Therefore, the N RC staff finds the proposed EPU acceptable with respect to the uncontrolled control rod assembly withdrawal at power.2.8.5.4.3 Startup of a Recirculation Loop at an lndodr~ect Temperature and Flow Controller Malfunction Causing an n *Core Flow Rate Regqulatory Evaluation A startup of an inactive loop transient may result in either an increased core flow or the introductior-i of cooler water into the core. This event causes an increase in core reactivity due to decreased moderator temperature and core void fraction.
The NRC staff's review covered (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 NRC's acceptance criteria are based on (1) final GDC-1 0, insofar as it requires that the RCS be designed with appropriate 66 margin to assure that SAFDLs are not exceeded &iring any condition of normal operation, including the effects of ACOs; (2) draft GDCs-14 and 15, insofar as they require that the core protection systems be designed to act automatically to prevent or suppress conditions that could result in exceeding acceptable fuel damage limits and that protection systems be provided for sensing accident situations and initiating the operation of necessary ESFs;GDC20 isofar as it requires that-the-protectionsse m ed indt nitate-autoeaieaflsU-the-per-ati~-apg~ogiate-systems4oe~mure.4hat-SA5 DL-s-rnoqt-exee-as-a-result-ofoertoa ,ocrrences; (3) SOC insoa asi-eur- htte C n t*soitd uiir cycem"b deig -with-magia--uff4icint to ensure that the design-pRCBaentceeded-during AOOsj (43) draft GDC-32, insofar as it requires that limits, which include considerable margin, be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the care, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling; G desined-to sueta teefcsofpsuae ratvy-aeiensn-ehe-sutt4n damage4o4he 7.-its-s uppe~t-structufes temals&-se-eao,-t-saifnieaflIym~pk~
e-eabitt-t-eete GGeej-;afl-(54) draft GDC-29, insofar as it requires that at least one of the reactivity control systems be capable of making the core subcritical under any condition sufficiently fast to prevent exceeding acceptable fuel damage limits. CDC---2&,4resofaras-Pqivres~hate-a reaotiv-ty-cento be provideandtae-apabe-ofreiabIy-eenetrne4ea~teo-rae .i changes to-ensrt~e, .that-.udr-cenditieons ef-exceeded--Specific review criteria are contained in SRP Section 1 5.4.4-5 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the increase in core flow event and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the SAFOLs and the RCPB pressure limits will not be exceeded as a result of this event. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of final GDC-1O and draft GDCs-14, 16, 29, and 32 GC 1, 1, 2, 2, an 8following implementation of the proposed EPU. Therefore, the NRC staff finds the p~roposed EPU acceptable with respect to the increase in core flow event.2.8.5.4.4 Spectrum of Rod Drop Accidents Recqulatorv Evaluation The NRC staff evaluated the consequences of a control rod drop accident in the area of reactor physics. The NRC staff's review covered 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 NRC's acteptance criteria are based on draft GDC-32, insofar as it requires that limits, 67 
-.".'4 .-- '.9"t~I.which include considerablemag': reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary~or (b) disrupt the core,,its support structures, or other vessel internals sufficiently to impair the effectiveness of'emergency core cooling. GG 28, inso.r...
it.... iro that......the 
... reatiity Co...........tems .b .dei e t.....urc.that.th......ct ot-postuat ed-eae~vity-aseIdens- -neithf-rersuU..........e-t...........grtr tha "im~SRP Section 15.4.9 and other guidance provided in Matrix 8 of RS-O01.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the rod drop accident and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated 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, the NRC staff concludes that the plant will continue to meet the requirements of draft GDC-32GQG-24&
following implementation of the EPU. Therefore, the NRC staff 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 Regqulatory 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 ftiel 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. The NRC staff's review covered (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 NRC's acceptance criteria are based on (1) final GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFOLs are not exceeded during normal operations, including AOOs; (2-)-AGOsj;-and (32) draft GDCs-29, insofar as it requires that at least one of the reactivity control systems be capable of making the core subcritical under any condition sufficiently fast to prevent exceeding acceptable fuel damage limits. Gr.C_ 26,' as. ,SAF-DL-s-arenot-e~eeede4.
Specific review criteria are contained in SRP Section 15.5.1-2 and 68 other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[lnaert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the inadvertent operation of EGOS or malfunction that increases reactor coolant inventory and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has 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 this event. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of final GDC-10, and draft implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the inadvertent operation of EGGS 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 Pressure Relief Valve R eculatory 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 turbine control valves (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 feedwater control system using water from the condensate storage tank via the condenser hotwell. The NRC staffs review covered (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 NRC's acceptance criteria are based on (1) final GDC-1 0, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFIJLs are not exceeded during normal operations, including AOOs; 7
iifes-that-the-RCS-and-its-as~seated-,ufLfieie -te-ensure-that~h-deig end.ins-f-Uhe-RCP ar=no ceeedd-uring-AO~sj-and (32) draft 000-29 insofar as it requires that a reactivity control system be provided capable of making the corn subcritical under any conditions (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits.-ODC 26, noofr., %A-eqi that
* r *ciiyente-yt
.....iod *" be capabl of, -reli..bly c...r... ,&sect;-4he-rtc of recivt chneoesure-that-unde .. onditi..n..of..normal op.rati.n,.inc.....n.
AC............
arc not ..........
Specific review criteria are contained in SRP Section 15.6.1 and other guidance provided in Matrix 8 of RS-O01.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the 69 proposed changes satisfy each of tlreqiurementsitl regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion 
.", The NRC staff has reviewed the licensee's analyses of the inadvertent opening of a pressure relief valve event and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the SAFOLs and the RCPB pressure limits will not be exceeded as a result of this event. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of final GDC-1 0 and draft G DC-29 cI 3'-26-following implementation of the proposed EPU. Therefpre, the NRC staff finds the proposed EPU acceptable with respect to the inadvertent opening of a pressure relief valve event.2.8.5.6.2 Emergency Core Cooling System and Loss-of-Coolant Accidents Reaqulatorv 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.
The NRC staff's review covered (1) the licensee's 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 peak cladding temperature, 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 EGGS systems; and (7) operator actions. The NRC's acceptance criteria are based on (1) 10 CFR &sect; 50.46, insofar as it establishes standards for the calculation of ECCS performance and acceptance criteria for that calculated performance; (2) 10 CFR Part 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) draft GDCs-40 and 42, insofar as they require that protection be provided for ES~s against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA;GDC27,insofar-as it-requires that-Sthe reativtacntrolesysems be .... detged- asto h~av~e-e a ombfe-Sapab~lity 7 4-iw-e~junetknWt-~
sen-add~if by teEC, f elia~y-entf-ellng-reastivit 
..... ge u..der. postulte accident...v condtion, with oppro a'emf&sect;1,,,  to assure the capailit to, coo the. core is.. maintain.....d; and (54) final 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.
Specific review criteria are contained in SRP Sections 6.3 and 15.6.5 and other guidance provided in Matrix 8 of RS-001.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evajuation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]70 S Conclusion The NRC staff has reviewed the licensee's analyses of the LOCA events and the ECOS. The NRC staff concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and that the analyses were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection system and the ECCS will continue to ensure that the peak cladding temperature, 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, the NRC staff concludes that the plant will continue to meet the requirements of final GDC-35 and draft GDCs-40 and 42, GDCc '1, 27, 3.7-,and 10 CFR 50.46 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the LOCA.2.8.5.7 Anticipated Transients Without Scrams Regulatory Evaluation ATWS is defined as an AOO followed by the failure of the reactor portion of the protection system specified in draft GDCs-14 and 15.GDG=-20 The regulation at 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 standby liquid control system (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 system initiation must be automatic.
* each BWR have equipment to trip the reactor coolant recirculation pumps automatically under conditions indicative of an ATWS.The NRC staff's review was conducted to ensure that (1) the above requirements are met, (2) sufficient margin is available in the setpoint for the SLCS pump discharge relief valve such that SLOS operability is not affected by the proposed EPU, and (3) operator actions specified in the plant's Emergency Operating Procedures are consistent with the generic emergency procedure guidelines/severe accident guidelines (EPGs/SAGs), insofar as they apply to the plant design. In addition, the NRC staff reviewed the licensee's ATWS analysis to ensure 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 &deg;F; (3) the peak suppression pool temperature is less than the design limit; and (4) the peak containment pressure is less than the containment design pressure.
The NRC staff also evaluated the potential for thermal-hydraulic instability in conjunction with ATWS events using the methods and criteria approved by the NRC staff. For this analysis, the NRC staff reviewed 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.
lnse4-the-the ..........
uetfication of.. th ,. applicbilit of generi .......
e for,,.,";-- propoe EU]Review guidance is provided in Matrix 8 of RS-001.71 V -Technical Evaluation u[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the information submitted by the licensee related to ATWS and concludes that the licensee has adequately accounted for the effects of the proposed EPU on ATWS. The NRC staff concludes that the licensee has demonstrated that ARI, SLCS, and recirculation pump trip 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, the NRC staff finds the proposed EPU acceptable with respect to ATWS.2.8.6 Fuel Storage 2.8.6.1 New Fuel Storage Regqulatory 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. The NRC staffs review covered 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 NRC's acceptance criteria are based on draft GDC-6G9-6, insofar as it requires the prevention of criticality in fuel storage systems by physical systems or processes, preferably utilizing geometrically safe configurations.
Specific review criteria are contained in SRP Section 9.1 .1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effect of the new fuel on the analyses for the new fuel storage facilities and concludes that the new fuel storage facilities will continue to meet the requirements of draft GDC-66 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the new fuel storage.2.8.6.2 Spent Fuel Storage Regqulatory Evaluation Nuclear reactor plants include storage facilities for the wet storage of spent fuel assemblies.
The safety function of the spent fuel pool and storage racks is to maintain the spent fuel assemblies in a safe and subcritical array during all credible storage conditions and to provide a safe means 72 of loading the assemblies into shipping casss t rvecordthefctfte proposed EPU on the criticality analysis (e.g., reactivity/of fW6l spen fulstorage array and boraflex degradation or neutron poison efficacy).
The NRC's acceptance criteria are based on (1) draft GDC-40GD0*-4, insofar as it requires that protection be provided for engineered safety features against the dynamic effects and missiles that might result from plant equipment failu resinsofar-as itrequies thatSS&#xa3;Ce 
,,,pe,'anttsafewtey4eine~o ao it4-thG-eflv4r~nmeflta 
;ni sscae wih ora operation, mainte. nance t....ting, an otltdaedne n 2 raft GDC-66GC6, 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.
Specific review criteria are contained in SRP Section 9.1.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the spent fuel storage capability and concludes that the licensee has adequately accounted for the effects of the proposed EPU on the spent fuel rack temperature and criticality analyses.The NRC staff also concludes that the spent fuel pool design will continue to ensure an acceptably low temperature and an acceptable degree of subcriticality following implementation of the proposed EPU. Based on this, the NRC staff concludes that the spent fuel storage facilities will continue to meet the requirements of draft GDCs-40 and 66GDCs ', and 62 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to spent fuel storage.2.9 Source Terms and Radiologqical Consequences Analyses 2.9.1 Source Terms for Radwaste Systems Analyses Regqulatory Evaluation The NRC staff reviewed the radioactive source term associated with EPUs to ensure the adequacy of the sources of radioactivity used by the licensee as input to calculations to verify that the radioactive waste management systems have adequate capacity for the treatment of radioactive liquid and gaseous wastes. The NRC staff's review included the parameters used to determine (1) the concentration of each radionuclide in the reactor coolant, (2) the fraction of fission product activity released to the reactor coolant, (3) concentrations of all radionuclides other than fission products in the reactor coolant, (4) leakage rates and associated fluid activity of all potentially radioactive water and steam systems, and (5) potential sources of radioactive materials in effluents that are not considered in the plant's Updated Final Safety Analysis Report related to liquid waste management systems and gaseous waste management systems. The NRC's acceptance criteria for source terms are based on (1) 10 CFR Part 20, insofar as it establishes requirements for radioactivity in liquid and gaseous effluents released to unrestricted areas; (2) 10 CFR Part 50, Appendix I, insofar as it establishes numerical guides for design objectives and limiting conditions for operation to meet the "as low as is reasonably 73 achievable" criterion; and (3), draft requires that the plant design include means to control the release of radioactive effluents.
Specific review criteria are contained in SRP Section 11.1.Technical Evaluation 
: Elnsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the .requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the radioactive source term associated with the proposed EPU and concludes that the proposed parameters and resultant composition and quantity of radionuclides are appropriate for the evaluation of the radioactive waste management systems.The NRC staff further concludes that the proposed radioactive source term meets the requirements of 10 CFR Part 20, 10 CFR Part 50, Appendix I, and draft GDC-70GD0;-69.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to source terms.NOTE Us Scttes-% ,2-n below- t4he Iieensee'e-ra~ietegieal-eeesequenses-n-ae-atemative-sour-ee4erm.
2.9.2 Radiological Cosquneso Cnro,,,,l Accide,.;,nt Uin Alternative Source Term Regulatory Evaluation The NRC staff reviewed the DAnlssoth radiological consequences ofnaly control roddropoia drpaccident (CRDA). The NRCn stafreviewincued anrexaiati(LBn ofe (1) sthepant's respwonseac totaccident, (2lssinldd()
the rees ffseinrouecefeetsfo th d core todtels enviro mentvions the vatlh oueisn f pr amee-nususdbbh iensee fAB-a4~w~or hecacuation-zeof ethertotnrand infethve cotole romq uovleto (ThEDrEle) sfotean. The NRC's acceptance criteria for thiloilcn equne sadrsfrradiological consequences of a ponsrtureadro accidentabae ond (1 ) fO 1.inaofaC-1s9,t isfrsitrequires-that-adequateeradiationnprotectionbb provided to permit access andocuayoft-ocuanyoetecntrol room under accident conditions without personnel receiving rdainepsrsi raeditonan(2 10osre CFR P-arts 100 rinsofar, as iteftbished rqinrements, for asrn theuatio ofa thec oe fo osuae accidents ilb cetby12. Specific review criteria-are cnandi R eto 50I Tcntined invaPluations6 n S..,adohrudnepoie nMti fR oQ74 
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has evaluated the licensee's revised accident analyses performed in support of the proposed EPU and concludes that the licensee has adequately accounted for the effects of the proposed EPU. The NRC staff further concludes that the plant site and the dose mitigating ESFs remain acceptable with respect to the radiological consequences of postulated DBAs since, as set forth above, the calculated total effective dose equivalent (TEDE) at the exclusion area boundary (EAB), at the low population zone (LPZ) outer boundary, and in the control room meet the exposure guideline values specified in 10 CFR 50.67 and final GDC-19, as well as applicable acceptance criteria denoted in SRP Section 15.0.1. Therefore, the NRC staff finds the licensee's proposed EPU acceptable with respect to the radiological consequences of DBAs.oeseue cof a control rod ep ace dont-and-oenoludes-tatathe4ioonsee-has-adequately-aeoou44o-fr-the-effeets~f~he-piopesed P4on.thse-ealay~sesr-.-qhe-NRGg-stafffurte-e,,poeu reguideflne-vele n1 CE10.11. The-G-staffalsoe-oelude s-tbat-4he-oee~o-room-meets~he-oese-requ~iremenef-Q-OC-*1Oef-iAs,-..*T-hee 2.0. adiolongicsal Cons-equenesteofehe-Faimale-ef-armt-L-iessue-Garoun&sect;y-{-=-
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nmven  ThenRC saff r~efi-dth nlyif t heimrantdie~qogicalyconsequencesef faiues inhes-n-sapelne) h NCsafsrve icue idmentfct-io-of~smal lnes~-t ee-4nss-{2AaP, urc s moelund a.. sump.....
ns..for...t..
cacltion..
of.the.radiological doesfr h psultdalue and-(4an-evarutioneaf~mhe-fpnaeoRbdmnectvh 
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..the conta:inm -enof-small, le ...n.cct.. to, the prim....ry effest~s*o44herefqosed- -oe-these-naybe&,--The-Ns site-a hAe dose-get -S-aw emain-acesbe4Ath -spect~etedegioai-NRCG-staff-lacnldsta t te -nrol0m-meets-be-dose reurmnsofGC1 o DBAs,<. herefoie;.he.-N R-O-staff -flnds-thelioeneee'-p se4l EPLJ-accep~ta ble-wt-esebo otside he..c............ -e*--matl-in....
connected to the-p4masr~ooant-p~ressu re-bond,-2-_23.0 Railgiaoneuncso Main.........a....in..
reFaiurO tsie-Containment Re gulato-vlaien The .RC .taf revewed the-analyss f h r.di....i con...e..unc..es-of-a MSB cidn spike and (2)a ....... wit th maiu eqiiru cnetain for..... in.........eu-epeatn The....... .RC's-aceptane-criterieo te radi.log.ca.
consequences of an.... MSLB-ousd otim kt-ase-ba n....4).GDC.4-inoferae-t-qure tha adqut raito pfoetectien4e-hpfvd-o permit-eeees.aF-end-o, upaney-oft4he-ob.40r-nee-aeedent-,ondtoswtot p" o t insofarsite~shabnges requirements forthe assuringt that h-redolgia dose,:esaa fro potuete a-olearlink4o~he-conetusiens-Feaohe 4-by-the44RC-stff;-as-documeetedkhe-on~
si ConclMusion ed-th~ie4nsee~srvisedidn-aesei~r-th-radioloioa cosqene f-an MSLB outside cotimn an cnl., ht~eI;e ...... eqetI~acone o h fet ftepoosdEUe h nlc h R tf fudh-;nl~mnb'--l c with respect toa equdi4u~,m-eonc.entation4or-contie full.. pow...r o per-ation 7-T-PeNRC-etffaseoeoudeat--
Phegcont-rol room meets., the, doze- rqueete-Q---orRs--eefe-h-N -ef-76 euteid-ontinm ,t alCnsguccs ES opnnsotiecntimn htcnrbt to th totaLOA doses. The NRC taf reiwas incude ()-the-contribution to thdose due-to leakage frm the-main-steam-isolatan-a~ves-(MS4Vs)-2-)-thereethdoegyn-areeu~ts-ef-oletlattos-the  oreeneeuenses-resu~tin&sect;-frem-eortan meflt-an4E-SF-eompaeflentafd-Ms hypethetical-L-OCA;  )-an-seeseto tecnanmentwt ropGt-oteeumpters andth npt aa meerorte dse-calculaton.
Th R'S caclain er baedo pertnetinemaieat4nmt~
a ted-Safety 
#natyeie-R t-orjpdet~ed Repo~andaon-ee~sieNeARG-st aff-s-e vaI~atieonotdosmifgatng-ESF-s-ThNc CC 1, nsfaran i-requires-4het-ade~qute-eraie-pet~etin prvie toprita/ s
-.
4e r-ass f-evew-ofitei-r4eaeoenaine~nSP-Se4en-64-and-AFppenieoes-ArBEfa Qef SRP.-e$5oa40 
-&6- .other-guid~ance.-provided.4.
Matrix9-o#f-P-D0t-
[ls44eeni-abev 
..u.....n The t e a-NRC t-fre-ourneneAn-h~e-onflusion Cencluiont aff-has-vauat~hl-ense ivad-ee.dintr-anaaysee4erh-mdi4e-gieal-,e-of-a-des ais-L-OCA-andJ-eoncludes-that--helicenseee-has-adequaey aeoeonedehe-effe,-te-ef-the-rpoedEpogeaayse&T--RGstffu he.coneaude that-theplant sits...nd the reai acceptbewihrsp e--the-dloses-at4he-E-AB-andqhe-LP-p--u ter-beu~mary-do-nfet-exedhe-.axpesuare-gudene--vale-of4 Trhe-turoeseo4his-rview-was-4o-evatuate~he-edequaey-of-syatem-desige-eaufe-nd~
nt t-a~a~lagee- .-scemblies.Sc cidents-may-77 licaeneefrthe caclto oferadilogca dura ;(2tioaeqa of the-aESee; providedI fosr-ere-ad ...........
j-....(3--ta cntin" n v"1tl4ten-yt ea.ta~inmont 
.Tahebyhe-NRC'tacceptance criter-iaortheu raioogca protctin bepro~de4opemit-oees-ane.....y.o the+ controltroom under accidente crondtoswtotpronnercevigeraditionte-rpdosuesinees-of 5frmFholbdyorit requrestha sytem tha cotai raieat~tybe-esinedwit appoprateconainent 78 

htbd, orit ouiv ae4en4-an thettcontain radicoctivity be dosigned-Ath-a~ppeprdte ytes ,nd (3) 10 CFR-Part 10OGasefar-asit-eetablhshes-requir-ements-fe-assu,4n&sect;Ahat-SRP-See~tner6A a-4-a 4;- 1 d-et, a er~&sect;uidaneefervdedAMa#xe-R-Q a ci-Ngar lin ..to.th conclusionsmchdb the NRC' ac douotdi teenl......ue.ce of a.. spn fue coc"k drop acci~denad, onluestat nne-lieesee-has.-
adqatl acone o h fat of.......-terepsed4 .PU on..h.. analyses.
TheNRC;-staff-calculated wha-ebedyanlod doses-at-the-EABante-LP-Zueter-boundaryaew acciente 2.5. " UA: quditonM avi........
Aro ..Source ....Terms.. andRadol ..cl onseqen......
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as-neeeesar
*24782.10 Health Physics 24784-t2.10.1 Occupational and Public Radiation Doses Regulatory Evaluation The NRC staff conducted its review in this area to ascertain what overall effects the proposed EPU will have on both occupational and public radiation doses and to determine that the licensee has taken the necessary steps to ensure that any dose increases will be maintained as low as is reasonably achievable.
The NRC staffs review included an evaluation of any increases in radiation sources and how this may affect plant area dose rates, plant radiation zones, and plant area accessibility.
The NRC staff evaluated how personnel doses needed to access plant vital areas following an accident are affected.
The NRC staff considered the effects of the proposed EPU on nitrogen-16 levels in the plant and any effects this increase may have on radiation doses outside the plant and at the site boundary from skyshine.
The NRC staff also considered the effects of the proposed EPU on plant effluent levels and any effect this increase may have on radiation doses at the site boundary.
The NRC's acceptance criteria for occupational and public radiation doses are based on 10 CFR Part 20 and final GDC-I 9. Specific review criteria are contained in SRP Sections 12.2, 12.3, 12.4, and 12.5, and other guidance provided in Matrix 10 of RS-001.Technical Evaluation 79 
[Insert technical evaluation.
The n Qa e ialaton (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion' The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on radiation source terms and plant radiation levels. The NRC staff concludes that the licensee has taken the necessary steps to ensure that any increases in radiation doses will be maintained as low as reasonably achievable.
The NRC staff further concludes that the proposed EPU meets the requirements of 10 CFR Part 20 and final GDC-1 9. Therefore, the NRC staff finds the licensee's proposed EPU acceptable with respect to radiation protection and ensuring that occupational radiation exposures will be maintained as low as reasonably achievable.
......onekeview-Aeas-(-Health-P-hysies)j t.lnse4t-Reg 2492.11 Human Performance 2.4942.11.1 Human Factors Reaulatorv Evaluation The area of human factors deals with programs, procedures, training, and plant design features related to operator performance during normal and accident conditions.
The NRC staff's human factors evaluation was conducted to ensure that operator performance is not adversely affected as a result of system changes made to implemented the proposed EPU. The NRC staff's review covered changes to operator actions, human-system interfaces, and procedures and training needed for the proposed EPU. The NRC's acceptance criteria for human factors are based on final GDC-19, 10 CFR 50.120, 10 CFR Part 55, and the guidance in GL 82-33.Specific review criteria are contained in SRP Sections 13.2.1, 13.2.2, 13.5.2.1, and 18.0.Technical Evaluation The NRC staff has developed a standard set of questions far the review of the human factors area. The licensee has addressed these questions in its application.
Following are the NRC staff's questions, the licensee's responses, and the NRC staff's evaluation of the responses.
: 1. Chanqes in Emergqency and Abnormal Operating Procedures Describe how the proposed EPU will change the plant emergency and abnormal operating procedures. (SRP Section 13.5.2*.1)
[Insert licensee's response followed by NRC staff statement on why the response is acceptable]
: 2. Changes to Operator Actions Sensitive to Power Uprate Describe any new operator actions needed as a result of the proposed EPU. Describe changes to any current operator actions related to emergency or abnormal operating procedures that will occur as a result of the proposed EPU. (SRP Section 18.0)(i.e., Identify and describe operator actions that will involve additional response time or 80 will have reduced time available.
Your response should address any operator workarounds that might affect these response times. Identify any operator actions that are being automated or being changed from automatic to manual as a result of the power uprate. Provide justification for the acceptability of these changes).[Insert licensee's response followed by NRC staff statement on why the response is acceptable]
: 3. Chances to Control Room Controls.
Displays and Alarms Describe any changes the proposed EPU will have on the operator interfaces for control room controls, displays, and alarms. For example, what zone markings (e.g. normal, marginal and out-of-tolerance ranges) on meters will change? What setpoints will change? How will the operators know of the change? Describe any controls, displays, alarms that will be upgraded from analog to digital instruments as a result of the proposed EPU and how operators will be tested to determine they bould use the instruments reliably. (SRP Section 18.0)[Insert licensee's response followed by NRC staff statement on why the response is acceptable]
: 4. Chancqes on the Safety Parameter Displav System Describe any changes to the safety parameter display system resulting from the proposed EPU. How will the operators know of the changes? (SRP Section 18.0)[Insert licensee's response followed by NRC staff statement on why the response is acceptable]
: 5. Changes to the Operator Training Proaram and the Control Room Simulator Describe any changes to the operator training program and the plant referenced control room simulator resulting from the proposed EPU, and provide the implementation schedule for making the changes. (SRP Sections 13.2.1 and 13.2.2)[Insert licensee's response followed by NRC staff statement on why the response is acceptable]
Conclusion The NRC staff has reviewed the changes to operator actions, human-system interfaces, proceduies, and training required for the proposed EPU and concludes that the licensee has (1) appropriately accounted for the effects of the proposed EPU on the available time for operator actions and (2) taken appropriate actions to ensure that operator performance is not adversely affected by the proposed EPU. The NRC staff further concludes that the licensee will continue to meet the requirements of final GDC-1 9, 10 CFR 50.120, and 10 CER Part 55 following implementation of the proposed EPU. Therefore, the NRC staff finds the licensee's proposed EPU acceptable with respect to the human factors aspects of the required system changes.[Aditina 'cio Aro4umnP fp, sert-Regulater-y-E-valuatine,,heioab~ahatien ,andtonetsionseetenaa~e  2,802.1 2 Power Ascension and Testingq Plan*2:8O42.12.1 Approach to EPU Power Level and Test Plan Reglulatory Evaluation 81 The purpose of the EPU test program 1ibdf ifrlf'VSCs will perform satisfactorily in service at the proposed EPU power level. The test program also provides additional assurance that the plant will continue to operate with design criteria at EPU conditions.
The NRC staff's review included an evaluation of: (1) plans for the initial approach to the proposed maximum licensed thermal power level, including verification of adequate plant performance, (2)transient testing necessary to demonstrate that plant equipment will perform satisfactorily at the proposed increased maximum licensed thermal power level, and (3) the test program's conformance with applicable regulations.
The NRC's acceptance criteria for the proposed EPU test program are based on 10 CFR Part 50, Appendix B, Criterion XI, which requires establishment of a test program to demonstrate that SSCs will perform satisfactorily in service.Specific review criteria are contained in SRP Section 14.2.1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The staff has reviewed the EPU test program, including plans for the initial approach to the proposed maximum licensed thermal power level, transient testing necessary to demonstrate that plant equipment will perform satisfactorily at the proposed increased maximum licensed thermal power level, and the test program's conformance with applicable regulations.
The staff concludes that the proposed EPU test program provides adequate assurance that the plant will operate in accordance with design criteria and that SSCs affected by the proposed EPU, or modified to support the proposed EPU, will perform satisfactorily in service. Further, the staff finds that there is reasonable assurance that the EPU testing program satisfies the requirements of 10 CFR Part 50, Appendix B, Criterion XI. Therefore, the NRC staff finds the proposed EPU test program acceptable.-f [#dditionai Review Aroas (PowereAsccs" adTcngPa)[-nser4-Regulatery-~valuat'+er1-,-Teo4eIe-E-valuaticn, and Cono "o cztaca' ncsay 2-41-2.13 Risk Evaluation Risk Evaluation of EPU Recsulatory Evaluation The licensee conducted a risk evaluation to (1) demonstrate that the risks associated with the proposed EPU are acceptable and (2) determine if "special circumstances" are created by the proposed EPU. As described in Appendix D of SRP Chapter 19, special circumstances are present if any issue would potentially rebut the presumption of adequate protection provided by the licensee to meet the deterministic requirements and regulations.
The NRC staff's review covered the impact of the proposed EPU on core damage frequency (ODE) and large early release frequency (LERF) for the plant due to changes in the risks associated with internal events, external events, and shutdown operations.
In addition, the NRC staff's review covered the quality of the risk analyses used by the licensee to support the application for the proposed EPU. This included a review of the licensee's actions to address issues or weaknesses that may have been raised in previous NRC staff reviews of the licensee's individual plant examinations (IPEs) and individual plant examinations of external events (IPEEE), or by an industry peer review. The NRC's risk acceptability guidelines are contained in 82 RG 1.174. Specific review guidance is contained in Matrix 13 of RS-OO1 and its attachments.
Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the risk implications associated with the implementation of the proposed EPU and concludes that the licensee has adequately modeled and/or addressed the potential impacts associated with the implementation of the proposed EPU. The NRC staff further concludes that the results of the licensee's risk analysis indicate that the risks associated with the proposed EPU are acceptable and do not create the"special circumstances" described in Appendix D of SRP Chapter 19. Therefore, the NRC staff finds the risk implications of the proposed EPUI acceptable.
............
Additie nal-Review.Ar-eae-(Risk-E-valuatier)}
fin r t-Reoulatery-Evatuatien-T~ehnioa.E-va ad.C nen4us1n-se~tions-s-nssry}
3.0 FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATION CHANGES To achieve the EPU, the licensee proposed the following changes to the Facility Operating License and TSs for [Plant Name].EProvide a list of license and TSs changes (including license conditions) and an NRC staff evaluation of each.]4.0 REGULATORY COMMITMENTS Insert the following sentence if the licensee has not made any regulatory commitments in support of the EPU.The licensee has made no regulatory commitments in its application for the EPU., Insert the following if the licensee has made regulatory commitments in support of the EPU.The licensee has made the following regulatory commitment(s):
[Provide a summary of each regulatory commitment made by the licensee.]
The NRC staff finds that reasonable controls for the implementation and for subsequent evaluation of proposed changes pertaining to the above regulatory commitment(s) are best provided by the licensee's administrative processes, including its commitment management program. The above regulatory commitments do not warrant the creation of regulatory requirements (items requiring prior NRC approval of subseq uent changes).6.0 RECOMMENDED AREAS FOR INSPECTION" As described above, the NRC staff has conducted an extensive review of the licensee's plans and analyses related to the proposed EPU and concluded that they are acceptable.
The NRC staff's review has identified the following areas for consideration by the NRC inspection staff 83 during the licensee's implementation of thl propos 5 areas are recommended based on past experience with EPUs, the extent and unique nature of modifications necessary to implement the proposed EPU, and new conditions of operation necessary for the proposed EPU.They do not constitute inspection requirements, but are intended to give inspectors insight into important bases for approving the EPU.'"[Provide list of recommended areas for inspection.]

==6.0 STATE CONSULTATION==

In accordance with the Commission's regulations, the EName of State] State official was notified of the proposed issuance of the amendment.
The State official had [no] comments.[If comments were received, address them here.]7.0 ENVIRONMENTAL CONSIDERATION Pursuant to 10 CFR 51.21, 51.32, 51.33, and 51.35, a draft Environmental Assessment and finding of no significant impact was prepared and published in the Federal Register on[Date] ( FR ). The draft Environmental Assessment provided a 30-day opportunity for public comment. if no comments were received, use the following sentence: 
[No comments were received on the draft Environmental Assessment.]
If comments were received, use the following sentence: 
[The NRC staff received comments which were addressed in the final environmental assessment.]
The final Environmental Assessment was published in the Federal Register on [Date] ( FR ). Accordingly, based upon the environmental assessment, the Commission has determined that the issuance of this amendment will not have a significant effect on the quality of the human environment.

==8.0 CONCLUSION==

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.9.0 REFRNE 1. RS-001, Revision 0, "Review Standard for Extended Power Uprates," December 2003.2. [Insert additional references as necessary]

==Attachment:==

List of Acronyms Principal Contributors:
Date: 84 aI LIST OF ACRONYMS AAC alternate ac sources ac alternating current ALARA as low as reasonably ach!evable ARAVS auxiliary and radwaste area ventilation system ARI alternate rod insertion ASME American Society of Mechanical Engineers ATWS anticipated transient without scram B&PV boiler and pressure vessel BL bulletin BOP balance-of-plant BTP branch technical position BWR boiling-water reactor BWR VIP Boiling Water Reactor Vessel and Internals Project COF core damage frequency CFR Code of Federal Regulations CFS condensate and feedwater system CRAVS control room area ventilation system CRDA control rod drop accident GRDM control rod drive mechanism CRDS "control rod drive system CUF cumulative usage factor CWS circulating water system OBA design-basis accident DELOCA design-basis loss-of-coolant accident do direct current DG draft guide EAB exclusion area boundary ECCS emergency core cooling system EFDS equipment and floor drainage system EPG emergency procedure guideline EPRI Electric Power Research Institute EPU extended power uprate EO environmental qualification ESF engineered safety feature ESFAS engineered safety feature actuation system ESFVS engineered safety feature ventilation system FAC flow-accelerated corrosion g5 FHA fuel handling accident.,'
FPP fire protectior/program 
"'GDO -general design criterion (or criteria)GL generic letter l&C instrumentation and controls IN information notice IPE ,individual plant examination IPEEE individual plant examination of external events LERF large early release frequency LLHS light load handling system LOCA loss-of-coolant accident LOOP loss of offsite power LPZ low population zone MC main condenser MCES imain condenser evacuation system MlOV motor-operated valve MSIV main steam isolation valve MSIVLCS main steam isolation valve leakage control system MSLB main steamline break MSSS main steam supply system MWt megawatts thermal NEI Nuclear Energy Institute NPSH net positive suction head NRC Nluclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NSSS nuclear steam supply system O&M operations and maintenance P-T pressure-temperature PWSCC3 primary water stress-corrosion cracking RCIC reactor core isolation cooling RCPB reactor coolant pressure boundary RCS reactor coolant system RG regujatory guide RHR residual heat removal RS review standard RWCS reactor water cleanup system SAFDL specified acceptable fuel design limit 86 

.-&#xb8;. ..., SAG severe accident guideline SAR Safety Analysis Report SBO station blackout SFP spent fuel pool SFPAVS spent fuel pool area ventilation system SGTS standby gas treatment system SLCS standby liquid control system'SRP Standard Review Plan SSCs structures, systems, and components SSE safe-shutdown earthquake SWMS solid waste management system SWS service water system TAVS turbine area ventilation system TBS turbine bypass system TCV _ turbine control valve TEDE total effective dose equivalent TS technical specifi cation UHS ultimate heat sink 87 ATTACHMENT 48 RS-OO1 SE Template GDC Markup (with redlinelstrike out)
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. xxx TO FACILITY OPERATING LICENSE NO. XXX-xx[LICENSEE]
[PLANT NAME, UNIT NO.]DOCKET NO. 5O-XXX NOTE: This document uses unqeHeading styles for each SE section so that a table of contents can be generated automatically.
The Heading styles appear on the Word ribbon (Home tab) in the"Styles" gallery (to see Heading 4 and Heading 5, click on the drop-down arrow next to the gallery).SE Sectin 1.1 1.2 1.1.1 1.1.1.1 1.1.1.1.1 2.0 Apply Style Called...Heading I Heading 2 Heading 3 Heading 4 Heading 5 Heading 1, etc.If you need to add a new numbered SE section, type the text only and then apply the appropriate Heading style fmom the gallery. When you generate a new TOO, the new SE section will be added.To update the TOC for page numbering or if you've added new SE sections (and applied appropriate Heading styles): 1. Click anywhere in the TOO which will appear greyed out.2. Then right-click and select Update Field... Update entire table.1 
[PLANT NAME, UNIT NO.]SAFETY EVALUATION FOR EXTENDED POWER UPRATE TABLE OF CONTENTS 

==1.0 INTRODUCTION==

............................................................................ 1.1 A~lcto ....................................................... 
............................ 1.2 Backgqround.................................................................................. 1.3 Licensee's Approach........................................................................ 1.4 Plant Modifications............................................................................ 1.5 Method of NRC Staff Review............................................................... 2.0 EVALUATION 
............................................................................... 2.1 Materials and Chemical Engineering 
......................................-
3-2.1.1 Reactor Vessel Material Surveillance Program ........................................... 2.1.2 Pressure-Temperature Limits and Upper-Shelf Energy .................................. 2.1.3 Reactor Internal and Core Support Materials 
............................................. 2.1.4 Reactor Coolant Pressure Boundary Materials 
........................................... 2.1.5 Protective Coating Systems (Paints) -Organic Materials................................ 2.1.6 Flow-Accelerated Corrosion................................................................ 2.1.7 Reactor Water Cleanup System............................................................ 2.1.8 [Additional Review Areas (Materials and Chemical Engineering)]....................... 2.2 Mechanical and Civil Engineeringq.......................................................... 2.2.1 Pipe Rupture Locations and Associated Dynamic Effects ............................... 2.2.2 Pressure-Retaining Components and Component Supports...........................-
10-2.2.3 Reactor Pressure Vessel Internals and Core Supports................................. 2.2.4 Safety-Related Valves and Pumps....................................................... 2.2.5 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment 
.................................................................................. 2.2.6 [Additional Review Areas (Mechanical and Civil Engineering)]......................... 2.3 Electrical Engineering 
.................................................-
16 -2.3.1 Environmental Qualification of Electrical Equipment 
.................................... 2.3.2 Offsite Power System...................................................................... 2.3.3 AC Onsite Power System ................................................................. 2 2.3.4 DC Onsite Power System................................................................. 2.3.5 Station Blackout ........................................................................... 2.3.6 [Additional Review Areas (Electrical Engineering)]...................................... 2.4 Instrumentation and Controls .................. 
.......................................... 2.4.1 Reactor Protection, Safety Features Actuation, and Control Systems................. 2.4.2 [Additional ReviewAreas (Instrumentation and Controls)].............................. 2.5 Plant Systems.............................................................................. 2.5.1 InternaIHazards 
........................................................................... 2.5.1.1 Flooding ................................................................................ 2.5.1.1.1 Flood Protection 
....................................................................... 2.5.1.1.2 Equipment and FlocrDrains 
.......................................................... 2.5.1.1.3 Circulating Water System ............................................................. 2.5.1.2 Missile Protection...................................................................... 2.5.1.2.1 Internally Generated Missiles......................................................... 2.5.1.2.2 Turbine Generator..................................................................... 2.5.1.3 Pipe Failures........................................................................... 2.5.1.4 Fire Protection 
......................................................................... 2.5.2 Fission Product Control.................................................................... 2.5.2.1 Fission Product Control Systems and Structures 
................................... 2.5.2.2 Main Condenser Evacuation System ................................................ 2.5.2.3 Turbine Gland Sealing System ....................................................... 2.5.2.4 Main Steam Isolation Valve Leakage Control System.............................. 2.5.3 Component Cooling and Decay Heat Removal ......................................... 2.5.3.1 Spent Fuel Pool Cooling and Cleanup System...................................... 2.5.3.2 Station Service Water System ........................................................ 2.5.3.3 Reactor Auxiliary Cooling Water Systems ........................................... 2.5.3.4 Ultimate Heat Sink..................................................................... 2.5.4 Balance-of-Plant Systems................................................................. 2.5.4.1 Main Steam ............................................................................ 2.5.4.2 Main Condenser 
....................................................................... 2.5.4.3 Turbine Bypass........................................................................ 2.5.4.4 Condensate and Feedwater 
............................................................ 2.5.5 Waste Management Systems ............................................................ 2.5.5.1 Gaseous Waste Management Systems ............................................. 2.5.5.2 Liquid Waste Management Systems................................................. 3 iii 2.5.5.3 Solid Waste Management Systems.................................................. 2.5.6 AdditionalIConsiderations................................................................. 2.5.6.1 Emergency Diesel Engine Fuel Oil Storage and Transfer System................. 2.5.6.2 Light Load Handling System (Related to Refueling)................................ 2.5.7 [Additional Review Areas (Plant Systems)] 
.............................................. 2.6 Containment Review Considerations......................-39 
-2.6.1 Primary Containment Functional Design ................................................. 2.6.2 SubcompartmnentAnalyses 
............................................................... 2.6.3 Mass and Energy Release ................................................................ 2.6.3.1 Mass and Energy Release Analysis for Postulated Loss of Coolant............... 2.6.4 Combustible Gas Control in Containment................................................ 2.6.5 Containment Heat Removal....:........................................................... 2.6.6 Secondary Containment Functional Design ............................................. 2.6.7 [Additional Review Areas (Containment Review Considerations)] 
..................... 2.7 Habitability.
Niltration, and Ventilation.................................................... 2.7.1 Control Room Habitability System........................................................ 2.7.2 Engineered Safety Feature Atmosphere Cleanup ...................................... 2.7.3 Control Room Area Ventilation System .................................................. 2.7.4 Spent Fuel Pool Area Ventilation System................................................ 2.7.5 Auxiliary and Radwaste Area and Turbine Areas Ventilation Systems ................ 2.7.6 Engineered Safety Feature Ventilation System ......................................... 2.7.7 [Additional Review Areas (Habitability, Filtration, and Ventilation)] 
.................... 2.8 Reactor Systems............................................ 2.8.1 Fuel System Design ....................................................................... 2.8.2 Nuclear Design............................................................................. 2.8.3 Thermal and Hydraulic Design............................................................ 2.8.4 Emergency Systems ...................................................................... 2.8.4.1 Functional Design of Control Rod Drive System .................................... 2.8.4.2 Overpressure Protection During Power Operation 
.................................. 2.8.4.3 Reactor Core Isolation Cooling System.............................................. 2.8.4.4 Residual Heat Removal System...................................................... 2.8.4.5 Standby Liquid Control System....................................................... 2.8.5 Accident and Transient Analyses......................................................... 4 E~KJ a-.- ~puI-.iv 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 ............
,......................................................-5 2.8.5.2 Decrease in Heat Removal by the Secondary System ............................. 2.8.5.2.1 Loss of External Load; Turbine Trip; Loss of Condenser Vacuum;Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed)............................................................ 2.8.5.2.2 Loss of Nonemergency AC Power to the Station Auxiliaries 
....................... 2.8.5.2.3 Loss of Normal Feedwater Flow......................................................  -2.8.5.3 Decrease in Reactor Coolant System Flow.......................................... 2.8.5.3.1 Loss of Forced Reactor Coolant Flow................................................ 2.8.5.3.2 Reactor Recirculation Pump Rotor Seizure and Reactor Recirculation Pump Shaft Break ..................................................................... 2.8.5.4 Reactivity and Power Distribution Anomalies........................................ 2.8.5.4.1 Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition......................................................... 2.8.5.4.2 Uncontrolled Control Rod Assembly Withdrawal at Power......................... 2.8.5.4.3 Startup of a Recirculation Loop at an Incorrect Temperature and Flow Controller Malfunction Causing an Increase in Core Flow Rate.................... 2.8.5.4.4 Spectrum of Rod Drop Accidents 
..................................................... 2.8.5.5 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 Pressure Relief Valve .................................... 2.8.5.6.2 Emergency Core Cooling System and Loss-of-Coolant Accidents 
................ 2.8.5.7 Anticipated Transients Without Scrams ..............................................-
73 -2.8.6 FuelStorage................................................................ 
'................ 2.8.6.1 New Fuel Storage...................................................................... 2.8.6.2 Spent Fuel Storage.................................................................... 2.8.7 [Additional Review Areas (Reactor Systems)] 
........................................... 2.9 Source Terms and Radiologqical Consequences Analyses ............................. 2.9.1 Source Terms for Radwaste Systems Analyses ........................................ 2.9.2 Radiological Consequences of Control Rod Drop Accident ............................ 2.9.3 Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment 
................................................... 2.9.4 Radiological Consequences of Main Steamline Failure Outside Containment................................................................................ 2.9.5 Radiological Consequences of a Design-Basis Loss-of-Coolant Accident............ 5 V 2.9.6 Radiological Consequences of Fuel Handling Accidents 
............................... 2.9.7 Radiological Consequences of Spent Fuel Cask Drop Accidents...................... 2.9.8 [Additional Review Areas (Source Terms and Radiological Consequences Analyses)].................................................................................. 2.10 Health Physics............................................................................. 2.10.1 Occupational and Public Radiation Doses............................................... 2.10.2 [Additional Review Areas (Health Physics)] 
............................................. 2.11 Human Performance 
...................................................................... 2.11.1 Human Factors............................................................................. 2.11.2 [Additional Review Areas (Human Performance)] 
....................................... 2.12 Power Ascension and Testing Plan ...................................................... 2.12.1 Approach to EPU Power Level and Test Plan........................................... 2.12.2 [Additional Review Areas (Power Ascension and Testing Plan)]....................... 2.13 Risk Evaluation 
............................................................................ 2.13.1 Risk Evaluation of EPU.................................................................... 2.13.2 [Additional Review Areas (Risk Evaluation)]............................................. 3.0 FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATION CHANGES ................................................................................. 4.0 REGULATORY COMMITMENTS 
........................................................ 5.0 RECOMMENDED AREAS FOR INSPECTION..........................................  

==6.0 STATE CONSULTATION==
................................................................. 7.0 ENVIRONMENTAL CONSIDERATION 
.... .............................................  

==8.0 CONCLUSION==
.............................................................................  

==9.0 REFERENCES==
.............................................................................
88-.6 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. xxx TO FACILITY OPERATING LICENSE NO. XXX-xx[LICENSE El[PLANT NAME, UNIT NO.1 DOCKET NO. 50-xxx 

==1.0 INTRODUCTION==

1.1 AApllication.
By application dated , as supplemented by letters dated_________________________________________, [Licensee] 
([Licensee Abbreviation], the licensee) requested changes to Facility Operating License No. NPF-029 and the Technical Specifications (TSs) for [Plant Name, Unit No.] ([Plant Abbreviation)).
Portions of the letters dated contain sensitive unclassified non-safeguards information and, accordingly, have been withheld from public disclosure.
The supplemental letters dated , provided additional clarifying information that did not expand the scope of the initial application and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal/Register on [date] (XX FR XXXX).The proposed changes would increase the maximum steady-state reactor tore power level from[current licensed power level] megawatts thermal (MWt) to [power level proposed by the licensee]
MWt, which is an increase of approximately 
[##] percent. The proposed increase in power level is considered an extended power uprate (EPU).1.2 Backgqround
[Plant Name] is a boiling-water reactor (BWR) plant of the BWRI[#] design with a Mark-[#]containment. 
[Plant Name] has the following special features/unique designs:[Insert any special features/unique designs]The NRC originally licensed [Plant Name] on [date] for operation at [original licensed power level] MWt. [By Amendment No. [W#] dated [ ], the NRC granted a power uprate to[Plant Name] of [##] percent, allowing the plant to be operated at [current licensed power level] MWt.] Therefore, the proposed EPU would result in an increase of approximately 7 
[##] percent over the original licensed power level [and [##] percent over the current licensed power level] for [Plant Name].] '1.3 L'icensee's Approach The licensee's application for the proposed EPU follows the guidance in the Office of Nuclear Reactor Regulation's (NRR's) Review Standard (RS)-OO1, "Review Standard for Extended Power Uprates," to the extent that the review standard is consistent with the design basis of the plant. Where differences exist between the plant-specific design basis and RS-OO1, the licensee described the differences and provided evaluations consistent with the design basis of the plant. The licensee also used [Identify topical reports or other documents used by the licensee for guidance related to the scope of the proposed EPU; NRC staff approvals, ranges of applicability, any limitations/restrictions associated with the documents; and consistency of the licensee's application with the ranges of applicability and limitations/restrictions.
The discussion in this section is to cover topical reports and other documents referenced for the overall power uprate process. It is not intended to cover topical reports and other documents for specific methods of analyses.
Topical reports and other documents referenced for specific methods of analyses are to be covered in the applicable technical evaluation section of this safety evaluation].
Insert this sentence if the licensee is planning to implement the EPU in one stage.[The licensee plans to implement the EPU in one step. The licensee plans to make the modifications necessary to implement the EPU during the refueling outage in[season year (e.g., fall 2003)]. Subsequently, the plant will be operated at [##] MWt starting in Cycle [##~].]Insert this paragraph if the licensee is planning to implement the EPU in stages:[The licensee plans to implement the EPU in E#J steps of [## and ##] percent. The licensee plans to make modifications necessary to implement the first step during the refueling outage in Eseason year (e.g., fall 2003)]. Subsequently, the plant will be operated at [It#] MWt during Cycle [fl]. The remainder of the modifications will be completed during the refueling outage in [season year (e.g., fall 2003)], with subsequent operation at [##] MWt starting in Cycle [##].]1.4 Plant Modifications The licensee has determined that several plant modifications are necessary to implement the proposed EPU. The following is a list of these modifications and the licensee's proposed schedule for completing them.[Provide a list of plant modifications.]
The NRC staffs evaluation of the licensee's proposed plant modifications is provided in Section 2.0 of this safety evaluation.
a 1.5 Method of NRC Staff Review The NRC staff reviewed the licensee's application to ensure that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) activities proposed will be conducted in compliance with the Commissjon's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public. The purpose of the NRC staff's review is to evaluate the licensee's assessment of the impact of the proposed EPU on design-basis analyses.
The NRC staff evaluated the licensee's application and supplements.
The NRC staff also evaluated 
[Include additional review items, as necessary (e.g., audits of certain information at the plant and vendor sites, and independent analyses), for areas where such analyses were deemed appropriate by the NRC staff].In areas where the licensee and its contractors used NRC-approved or widely accepted methods in performing analyses related to the proposed EPU, the NRC staff reviewed relevant material to ensure that the licensee/contractor used the methods consistent with the limitations and restrictions placed on the methods. In addition, the NRC staff considered the eaffects of the changes in plant operating conditions on the use of these methods to ensure that the methods are appropriate for use at the proposed EPU conditions.
Details of the NRC staff's review are provided in Section 2.0 of this safety evaluation.
Audits of analyses supporting the EPU were conducted in relation to the following topics:[Provide a list of areas for which audits were performed.]
The results of the audits are discussed in Section 2.0 of this safety evaluation.
Independent NRC staff calculations were performed in relation to the following topics:[Provide a list of areas for which independent NRC staff calculations were performed.]
The results of the calculations are discussed in Section 2.0 of this safety evaluation.

==2.0 EVALUATION==
2.1 Materials and Chemical Engqineering 2.1.1 Reactor Vessel Material Surveillance Program Regulatory Evaluation The reactor vessel material surveillance program provides a means for determining and monitoring the fracture toughness of the reactor vessel beltline materials to support analyses for ensuring the structural integrity of the ferritic components of the reactor vessel. The NRC staff's review primarily focused on the effects of the proposed EPU on the licensee's reactor vessel surveillance capsule withdrawal schedule.
The NRC's acceptance criteria are based on (1)-.draft General Design Criterion (GDC)-9, insofar as it requires that the reactor coolant pressure boundary (RCPS) be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage; Critcrizon (ODO) 11, 9 fratte (2) draft GDC-33, insofar as it requires that the RCPB be capable of accommodating without rupture, onfwijrnited allowance for energy absorption through plastic deformation, the stati6 and dynamic lbads imeposed on any boundary component as a result of any inadvertent and sudden release of energy to the coolant;final -GDC-31, insofar as it requires that the RCPB be designed with margin sufficient to assure that, under specified conditions, it wilt behave in a nonbrittle manner and the probability of a rapidly propagating fracture is minimized; (34) 10 CFR Part 50, Appendix H, which provides for monitoring changes in the fracture toughness properties of materials in the reactor vessel beltline region; and (45) 10 CFR 50.60, which requires compliance with the requirements of 10 CFR Part 50, Appendix H. Specific review criteria are contained in Standard Review Plan (SRP) Section 5.3.1 and other guidance provided in Matrix 1 of RS-001.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed EPU on the reactor vessel surveillance withdrawal schedule and concludes that the licensee has adequately addressed changes in neutron fluence and their effects on the schedule.
The NRC staff further concludes that the reactor vessel capsule withdrawal schedule is appropriate to ensure that the material surveillance program will continue to meet the requirements of 10 CFR Part 50, Appendix H, and 10 CFR 50.60, and will provide the licensee with information to ensure continued compliance with draft GDCs-9 and 33, and finalGQ44ad GDC-31 in this respect following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the reactor vessel material surveillance program.2.1.2 Pressure-Temperature Limits and Upper-Shelf Energy Reoulatorv Evaluation Pressure-temrperature (P-T) limits are established to ensure the structural integrity of the ferritic components of the RCPB during any condition of normal operation, including anticipated operational occurrences and hydrostatic tests. The NRC staff's review of P-T limits covered the P-T limits methodology and the calculations for the number of effective full power years specified for the proposed EPU, considering neutron embrittlement effects and using linear elastic fracture mechanics.
The NRC's acceptance criteria for P-T limits are based on (1) draft GDC-9, insofar as it requires that the RCPB be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage; (4-GQ-+an-ekernly~o-p~oebi 
....o rapidly propgatng racur, (2) final 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 nonbrittle manner and the probability of a rapidly propagating fracture is minimized; (3) 10 CFR Part 50, Appendix G, which specifies fracture toughness requirements for ferritic components of the RCPB; and (4) 10 CFR 50.60, which requires compliance with the requirements of 10 CFR Part 50, Appendix G. Specific review criteria are contained in SRP Section 5.3.2 and other guidance provided in Matrix I of RS-001.10 Technical Evaluation ....[Insert technical evaluation.
The technical evaluation should (1) ciearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a ciear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed EPU on the P-T limits for the plant and concludes that the licensee has adequately addressed changes in neutron fluence and their effects on the P-T limits. The NRC staff further concludes that the licensee has demonstrated the validity of the proposed P-T limits for operation under the proposed EPU conditions.
Based on this, the NRC staff concludes that the proposed P-T limits will continue to meet the requirements of 10 CFR Part 50, Appendix G, and 10 CFR 50.60 and will enable the licensee to comply with draft GDC-9,GDG-4 and final GDC-31 in this respect following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the proposed P-T limits.2.1.3 Reactor Internal and Core Support Materials Regqulatory Evaluation The reactor internals and core supports include structures, systems, and components (SSCs)that perform safety functions or whose failure could affect safety functions performed by other SSCs. These safety functions include reactivity monitoring and control, core cooling, and fission product confinement (within both the fuel cladding and the reactor coolant system (RCS)). The NRC staffs review covered the materials' specifications and mechanical properties, welds, weld controls, nondestructive examination procedures, corrosion resistance, and susceptibility to degradation.
The NRC's acceptance criteria for reactor internal and core support materials are based on draft GDC-1 GDC-4-and 10 CFR 50.55a for material specifications, controls on welding, and inspection of reactor internals and core supports.
Specific review criteria are contained in SRP Section 4.5.2 and Boiling Water Reactor Vessel and Internals Project (BWRVIP)-26.
Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conciusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed EPU on the susceptibility of reactor internal and core support materials to known degradation mechanisms and concludes that the licensee has identified appropriate degradation management programs to address the effects of changes in operating temperature and neutron fluence on the integrity of reactor internal and core support materials.
The NRC staff further concludes that the licensee has demonstrated that the reactor internal and core support materials will continue to be acceptable and will continue to meet the requirements of GDQ-4-draft GDC-1 and 10 CFR 50.55a with respect to material specifications, welding controls, and inspection following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to reactor internal and core support materials.
11 2.1.4 Reactor Coolant Pressure Boundary Materials Regulatory Evaluation The RCPB defines the boundary of systems and components containing the high-pressure fluids produced in the reactor. The NRC staffs review of RCPB materials covered their specifications, compatibility with the reactor coolant, fabrication and processing, susceptibility to degradation, and degradation managembnt programs.
The NRC's acceptance criteria for RCPB materials are based on (1) 10 CFR 50.55a and draft GDC-1,GQG=-I-, insofar as they require that those systems and components which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed; 
.. as .....reqir SSr imp etesafey-edeinedi-febr 
.............
edronkutdtstd d pete~ted~-
ef n...........at.d with-nF~maJ-eperatien-,mainteilRenee 7
(2) draft GDC-2, insofar as those systems and components which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects; (3) draft GDC-9 insofar, as it requires that the RCPB be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage; -GQ- s~eofaf-asit-fequirestatte rapidl-pepag-atinj-raeturej-(4,3) final 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 nonbrittle manner and the probability of a rapidly propagating fracture is minimized; and (45) 10 CFR Part 50, Appendix G, which specifies fracture toughness requirements for ferritic components of the RCPB. Specific review criteria are contained in SRP Section 5.2.3 and other guidance provided in Matrix 1 of RS-O01. Additional review guidance for primary water stress-corrosion cracking (PWSCC) of dissimilar metal welds and associated inspection programs is contained in Generic Letter (GL) 97-01, Information Notice (IN) 00-17, Bulletin (BL) 01-01, BL 02-01, and BL 02-02, Additional review guidance for thermal embrittlement of cast austenitic stain~less steel components is contained in a letter from C. Grimes, NRC, to D. Walters, Nuclear Energy Institute (NEI), dated May 19, 2000.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed EPU on the susceptibility of RCPB materials to known degradation mechanisms and concludes that the licen~see has identified appropriate degradation management programs to address the effects of changes in system operating temperature on the integrity of RCPB materials.
The NRC staff further concludes that the licensee has demonstrated that the RCPB materials will continue to 12 be acceptable following implementation of the proposed EPU and will continue to meet the requirements of draft GDCs-1, 2, and 9, CDC 1, GCO l, GOC 11,,final GDC-31, 10OCFR Part 50, Appendix G, and 10 CFR 50.55a. Therefore, the NRC staff finds the proposed EPU acceptable with respect to RCPB materials.
2.1.6 Protective Coating Systems (Paints) -Organic Materials Regulatory Evaluation Protective coating systems (paints) provide a means for protecting the surfaces of facilities and equipment from corrosion and contamination from radionuclides and also provide wear protection during plant operation and maintenance activities.
The NRC staff's review covered protective coating systems used inside the containment for their suitability for and stability under design-basis loss-of-coolant accident (DBLOCA) conditions, considering radiation and chemical effects. The NRC's acceptance criteria for protective coating systems are based on (1) 10 CFR Part 50, Appendix B, which states quality assurance requirements for the design, fabrication, and construction of safety-related SSCs and (2) Regulatory Guide 1.54, Revision 1, for guidance on application and performance monitoring of coatings in nuclear power plants.Specific review criteria are contained in SRP Section 6.1.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed EPU on protective coating systems and concludes that the licensee has appropriately addressed the impact of changes in conditions following a DELOCA and their effects on the protective coatings.The NRC staff further concludes that the licensee has demonstrated that the protective coatings will continue to be acceptable following implementation of the proposed EPU and will continue to meet the requirements of 10 CFR Part 50, Appendix B. Therefore, the NRC staff finds the proposed EPU acceptable with respect to protective coatings systems.2.1.6 Flow-Accelerated Corrosion Regqulatory Evaluation Flow-accelerated corrosion (FAC) is a corrosion mechanism occurring in carbon steel components exposed to flowing single- or two-phase water. Components made from stainless steel are immune to FAG, and FAC is significantly reduced in components containing small amounts of chromium or molybdenum.
The rates of material loss due to FAC depend on velocity of flow, fluid temperature, steam quality, oxygen content, and pH. During plant operation, control of these parameters is limited and the optimum conditior~s for minimizing FAC effects, in most cases, cannot be achieved.
Loss of material by FAC will, therefore, occur. The N RC staff has reviewed the effects of the proposed EPU on FAC and the adequacy of the licensee's FAC program to predict the rate of loss so that repair or replacement of damaged components could be made before they reach critical thickness.
The licensee's FAG program is based on NUREG-1344, GL 89-08, and the guidelines in Electric Power Research Institute (EPRI) Report NSAC-202L-R2.
It consists of predicting loss of material using the CHECWORKS computer code, and visual inspection and volumetric examination of the affected components.
13 The NRC's acceptance criteria are baatoon of the minimum acceptable wall thickness for the components undergoing degradation by FAC.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluation of the' effect of the proposed EPU on the FAC analysis for the plant and concludes that the licensee has adequately addressed changes in the plant operating conditions on the FAC analysis.
The NRC staff further concludes that the licensee has demonstrated that the updated analyses will predict the loss of material by FAC and will ensure timely repair or replacement of degraded components following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to FAC.2.1.7 ReactorWater Cleanup System Regqulatory Evaluation The reactor water cleanup system (RWCS) provides a means for maintaining reactor water quality by filtration and ion exchange and a path for removal of reactor coolant when necessary.
Portions of the RWCS comprise the RCPB. The NRC staff's review of the RWCS included component design parameters for flow, temperature, pressure, heat removal capability, and impurity removal capability; and the instrumentation and process controls for proper system operation and isolation.
The review consisted of evaluating the adequacy of the plant's TSs in these areas under the proposed EPU conditions.
The NRC's acceptance criteria for the RWCS are based on (1) draft GDCs-S and 34, insofar as they require that the RCPB be designed and constructed so as to have an exceedingly low probability of gross rupture, significant leakage, or rapidly propagating type failures; (2) draft GDC-70, insofar as it requires that the plant design include means necessary to maintain control over the plant -radioactive effluents; and (3) draft GDC-51, insofar as it requires that systems that parts of the RCPB outside containment have appropriate features necessary to protect the health and safety of the public in case of an accidental rupture in that part-.-G04 that-syetemct t ..............
rad ......
.......fn~ementk Specific review criteria are conteined in SRP Section 5.4.8.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed EPU on the RWCS and concludes that the licensee has adequately addressed changes in impurity levels 14 and pressure and their effects on the RWCS. The NRC stfffurther concludes that the licensee has demonstrated that the RWCS will continue to be acceptable following implementation of the proposed EPU and will continue to meet the requirements of draft GDCs-9, 34, 51, and 70. GDC-41,CC 0 nd D-6Therefore, the NRC staff finds the proposed EPU acceptable with respect to the RWCS.[AdiioalRview-Ar-eae-{-aterladChmca l Engineering}
2.2 Mechanical and Civil Engineering 2.2.1 Pipe Rupture Locations and Associated Dynamic Effects Reqiulatorv Evaluation SSCs important to safety could be impacted by the pipe-whip dynamic effects of a pipe rupture.The NRC staff conducted a review of pipe rupture analyses to ensure that SSCs important to safety are adequately protected from the effects of pipe ruptures.
The NRC staff's review covered (1) the implementation of criteria for defining pipe break and crack locations and configurations, (2) the implementation of criteria dealing with special features, such as augmented inservice inspection (I61) programs or the use of special protective devices such as pipe-whip restraints, (3) pipe-whip dynamic analyses and results, including the jet thrust and impingement forcing functions and pipe-whip dynamic effects, and (4) the design adequacy of supports for SSCs provided to ensure that the intended design functions of the SSCs will not be impaired to an unacceptable level as a result of pipetwhip or jet impingement loadings.
The NRC staff's review focused on the effects that the proposed EPU may have on items (1) thru (4)above. The NRC's acceptance criteria are based on draft GDC-40 insofar as it requires that protection be provided for engineered safety features (ESFs) against the dynamic effects and missiles that might result from plant equipment failures.
GOG=-4,-whieh-equk~ee-SSG=s-imp.anteaft 
*bedsiged,-aemm at4 ..... m" ffects of-apostulated-ipe-r-uptufre-.Specific review criteria are contained in SRP Section 3.6.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluations related to determinations of rupture locations and associated dynamic effects and concludes that the licensee has adequately addressed the effects of the proposed EPU on them. The NRC staff further concludes that the licensee has demonstrated that SSCs important to safety will continue to meet the requirements of draft GDC-4OGG-following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the determination of rupture locations and dynamic effects associated with the postulated rupture of piping. , 2.2.2 Pressure-Retaining Components and Component Supports Regulatory Evaluation is st!yprsereann The NRC staff has reviewed the (and their supports) designed in accordance with' the American Society df Mechanical Engineers (ASME)Boiler and Pressure Vessel Code (B&PV Code), Section III, Division 1, and draft GDCs-1, 2, 9, 33,3440 an 42GDc 1, 1, an 15 The NRC staff's review focused on the effects of the proposed EPU on the design input pgarameters and thd design-basis loads and load combinations for normal operating, upset, emergency, and faulted conditions.
The NRC staff's review covered (1) the analyses of flow-induced vibration and (2)the analytical methodologies, assumptions, ASME Code editions, and computer programs used for these analyses.
The NRC staff's review also included a comparison of the resulting stresses and cumulative fatigue usage factors (CUFs) against the code-allowable limits. The NRC's acceptance criteria are based on (1) 10 CER 50.55a and draft GDC-1, insofar as they, require that those systems and components which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed; ODC 1, insofa ..s they r..quire that c to safety" b comneuat wthth ipoanceef-the-osfoty-functin to.... bc-........ 
(2) draft GDC-2, insofar as those systems and components which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects of earthquae 
... with the effects of normal or acciden condition;.. 
(3) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA;(4.) draft GDCs-9 and 33, insofar as they require that the ROPE be designed and constructed so as to have an exceedingly low probability of RCPB gross rupture or significant leakage; and (5) draft GDC-34 insofar as it requires that the RCPB be designed to minimize the probability of rapidly propagating type failures. 
&G-C4,-insofar-ae-4-Tequilres4hat-S$s-tmpeoat4eofet marnen........e,...... 
-,an.d-p.....tated-aooident,. ) DC1, insof......r....s it ; ,qu s-habt-he-ef-rapidWy-prepagati, ,4 .ae,,u.e,, ......)-,, Q , ..... eefar-as itrqie"htteRSb e-ee~~gay6niine-e-aeeaim.Seii review criteria are contained in SRP Sections 3.9.1, 3.9.2, 3.9.3, and 5.2.1.1; and other guidance provided in Matrix 2 of RS-001.Technical Evaluation Nuclear Steam Supply System Piping.q Components, and Supports[Insert technical evaluation for nuclear steam supply system (NSSS) piping, components, and supports.
Include an intermediate conclusion in the form of "Because [summarize reasons], the NSSS piping, components, and supports are adequate under the proposed EPU conditions."]
Balance-of-Plant Piping.q Components, and Supports[Insert technical evaluation for balance-of-plant piping, components, and supports.Include an intermediate conclusion in the form of "Because [summarize reasons], the balance-of-plant piping, components, and supports are adequate under the proposed 16 EPU conditions."]-
Reactor Vessel and Supports[Insert technical evaluation for reactor vessel and supports.
Include an intermediate conclusion in the form of "Because [summarize reasons], the reactor vessei and supports are adequate under the proposed EPU conditions."]
Control Rod Drive Mechanism[Insert technical evaluation for control rod drive mechanism.
Include an intermediate conclusion in the form of "Because [summarize reasons], the control rod drive mechanism is adequate under the proposed EPU conditions."]
Recirculation Pumos and Supports[Insert technical evaluation for reactor coolant pumps and supports.
Include an intermediate conclusion in the form of "Because [summarize reasons], the recirculation pumps and supports are adequate under the proposed EPU conditions."]
Conclusion The NRC staff has reviewed the licensee's evaluations related to the structural integrity of pressure-retaining components and their supports.
For the reasons set forth above, the NRC staff concludes that the licensee has adequately addressed the effects of the proposed EPU on these components and their supports.
Based on the above, the NRC staff further concludes that the licensee has demonstrated that pressure-retaining components and their supports will continue to meet the requirements of 10 CFR 5O.55a, CDC i, SOC 2, Soc '1 GDC !4i, ond GD-6draft GDCs-1, 2, 9, 33, 34, 40, and 42 following implementation of the proposed EPU.Therefore, the NRC staff finds the proposed EPU acceptable with respect to the structural integrity of the pressure-retaining components and their supports.2.2.3 'Reactor Pressure Vessel Internals and Core Supports Regqulatory Evaluation Reactor pressure vessel internals consist of all the structural and mechanical elements inside the reactor vessel, including core support structures.
The NRC staff reviewed the effects of the proposed EPU on the design input parameters and the design-basis loads and load combinations for the reactor internals for normal operation, upset, emergency, and faulted conditions.
These include pressure differences and thermal effects for normal operation, transient pressure loads associated with loss-of-coolant accidents (LOCAs), and the identification of design transient occurrences.
The NRC staff's review covered (1) the analyses of flow-induced vibration for safety-related and non-safety-related reactor internal components and (2) the analytical methodologies, assumptions, ASME Code editions, and computer programs used for these analyses.
The NRC staff's review also included a comparison of the resulting stresses and CUFs against the corresponding Code-allowable limits. The NRC's acceptance criteria are based on (1) i0 CFR 5O.55a and draft GDC-1 insofar as they require that those systems and components which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences be designed, fabricated, erected, tested, and inspected to quality standards that reflect the importance of the safety function to be performed;-GQC-1+sfes-as-hey-euike-hat-toqualitystandards-c-ommensurate-with-The4mpe~t-noe-ef-the-safety-funeiesto-be-eem 17 (2 rf D-,insofar as those sy on s which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to performance standards that will enable the facjlity to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects;GDC 2, insofra t eur-stht S mprtn t e-safety be designed-to-withstand the e -etso a,,t a obndwthe s-eff+t-fnemlo acidn 'odtiensj (3) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects and missiles that might result from plant equipment failures, as well as the effects of a loss of coolant accident; 
&G4-noa si eqie ht5C te4he-ef feetc o-f-and to-bo compail-ihte evrne nt!oniton associated with normaloeation, maintenance, testing, and (4) final GDC-1O, insofar as it requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
Specific review criteria are contained in SRP Sections 3.93.1, 3.9.2, 3.9.3, and 3.9.5; and other guidance provided in Matrix 2 of RS-O01..Technical Evaluation Elnsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link td the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's evaluations related to the structural integrity of reactor internals and core supports and concludes that the licensee has adequately addressed the effects of the proposed EPU on the reactor internals and core supports.
The NRC staff further concludes that the licensee has demonstrated that the reactor internals and core supports will continue to meet the requirements of 10 CFR 50.55a, final GDC-1 0 and draft GDCs-1, 2, 40 and 42 GD , GD , CO , an-&G4following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the design of the reactor internal and core supports.2.2.4 -Safety-Related Valves and Pumps Regulatory Evaluation The NRC's staff's review included certain safety-related pumps and valves typically designated as Class 1, 2, or 3 under Section III of the ASME B&PV Code and within the scope of Section Xl of the ASME B&PV Code and the ASME Operations and Maintenance (O&M) Code, as applicable.
The NRC staff's review focused on the effects of the proposed EPU on the required functional performance of the valves and pumps. The review also covered any impacts that the proposed EPU may have on the licensee's motor-operated valve (MOV) programs related to GL 89-10, GL 96-05, and GL 95-07. The NRC staff also evaluated the licensee's consideration of lessons learned from the MOV program and the application of those lessons learned to other safety-related power-operated valves. The NRC's acceptance criteria are based on (1) draft GOC-1, insofar as it requires that those systems and components which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to quality standards that reflect the importance of the safety functions to be performed;GD-3 inta as it rsquires that "5cmontosft be dsge, farcae, erce, adte 18 (2) draft GDCs-38, 46, 47, 48, 59, 60, 61, 83, 64, and 65G- ,GC4O-,G ,-n-GG 46, insofar as they require that the emergency core cooling system (ECCS), the containment heat removal system, the containment atmospheric cleanup systems, and the cooling water systemdeepectively, be designed to permit appropriate periodic testing to ensure the leak-tight integrity and performance of their active components; (3) draft GDC-57,GL4G-44T, insofar as it requires that capability shall be provided for testing functional operability of valves and associated apparatus essential to the containment function for establishing that no failure has occurred and for determining that valve leakage does not exceed acceptable limitst4pieai systems penetrating containment be designdwth capbiit"t priedieally-testqhe-erera+lI~tt--o-he-ieelaton-~veaodetenmne-4-valve-4ea kege-s-wvith~aGGeptable-tmit; and (4) 10 CFR 50.55a(f), insofar as it requires that pumps and valves subject to that section must meet the inservice testing program requirements identified in that section. Specific review criteria are contained in SRP Sections 3.9.3 and 3.9.6; and other guidance provided in Matrix 2 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessments related to the functional performance of safety-related valves and pumps and concludes that the licensee has adequately addressed the effects of the proposed EPU on safety-related pumps and valves. The NRC staff further concludes that the licensee has adequately evaluated the effects of the proposed EPU on its MOV programs related to GL 89-10, GL 96-05, and.,GL 95-07, and the lessons learned from those programs to other safety-related, power-operated valves. Based on this, the NRC staff concludes that the licensee has demonstrated that safety-related valves and pumps will continue to meet the requirements of draft GDCs-1, 38, 46, 47, 48, 57, 59, 60, 61, 63,64, 65, GDC 1, GDC 37, GDC 10, GDC 13, GDC -!16, CDC $1, and 10 CFR 50.55a(f) following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to safety-related valves and pumps.2.2.5 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment Regulatory Evaluation Mechanical and electrical equipment covered by this section includes equipment associated with systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal. Equipment associated with systems essential to preventing significant releases of radioactive materials to the environment are also covered by this section. The NRC staff's review focused on the effects of the proposed EPU on the qualification of the equipment to withstand seismic events and the dynamic effects associated pipe-whip and jet impingement forces. The primary input motions due to the safe shutdown earthquake (SSE) are not affected by an EPU. The NRC's acceptance criteria are based on (1) draft GDC-1, insofar as it requires that those systems and components which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences be designed, fabricated, and erected to quality standards that reflect the importance of the safety functions to be performed;GDC4-1,asofat-aesbreqiresFhaSS~smpgant-e-19 
.'. *'~ -- -___ -_______________
thtcmoot ha r oto heRP edserjae ..bnc ............
d, an.d-tested-to theC-higheesalty-sttandar~s-p~aet~aIl-draft GDC-2, insofar as those systems and components which are essentila tothe prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the, public, the additional forces that might be imposed by natural phenomena such as'earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effectsGD 2,.. in.of.r.a..it eufr c -tht SSc ;important-t-aeybedcge to .ih..d.h..fetso ealqa esembinc ihteefcso norml o acien -odiin; (34) 10 CFR Part 100, Appendix A, which sets forth the principal seismic and geologic considerations for the evaluation of the suitability of plant design bases established in consideration of the seismic and geologic characteristics of the plant site; (45) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects and missiles that might result from plant equipment failures, as well as the effects of a LOCA; G-C4-nsofar-as-it-iureehat-SSGe4mpoen~t-tes-efety-bedsgndo acemnmdeate-the-effects-of-andoe4e-eompatiblawith-theernvkenmentakedens-aeseite-wt~
oia4op ....e...................estin 
.............
oodentsi-(56) draft GDCs-9 and 33, insofar as they require that the RCPB be designed and constructed so as to have an exceedingly low probability of RCPB gross rupture or significant leakage; and (6) draft GDC-34, insofar as it requires that the RCPB be designed to minimize the probability of rapidly propagating type failures; GDC444sofeast-extiemelyIewgrebabtit-y-ef-rapidly-prepagat4ngj-raeturet-;and (7) 10 CFR Part 50, Appendix B, which sets quality assurance requirements for safety-related equipment.
Specific review criteria are contained in SRP Section 3.10.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion.
The NRC staff has reviewed the licensee's evaluations of the effects of the proposed EPU on the qualification of mechanical and electrical equipment and concludes that the licensee has (1)adequately addressed the effects of the proposed EPU on this equipment and (2) demonstrated that the equipment will continue to meet the requirements of draft GDCs-1, 2, 9, 33, 34, 40 and 42;..... 1, 2,4, 1, an... 30; 10 CFR Part 100, Appendix A; and 10 CFR Part 50, Appendix B, following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the qualification of the mechanical and electrical equipment. 
-
1 ,-and-Cenqu sionsotn-ae 2%62.3 Electrical En~qineerinaq Environmental Qualification of Electrical 20 Equipment Reg ulatorv Evaluation Environmental qualificatiort(EQ) of electr ca equ pment nvolves demonstrating that the equipment is capable of performing its safety function under significant environmental stresses which could result from DBAs. The NRC staffs review focused on the effects of the proposed EPU on the environmental conditions that the electrical equipment will be exposed to during normal operation, anticipated operational occurrences, and accidents.
The NRC staffs review was conducted to ensure that the electrical equipment will continue to be capable of performing its safety functions following implementation of the proposed EPU. The NRC's acceptance criteria for EQ of electrical equipment are based on 10 CFR 50.49, which sets forth requirements for the qualification of electrical equipment important to safety that is located in a harsh environment.
Specific review criteria are contained in SRP Section 3.11.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the EQ of electrical equipment and concludes that the licensee has adequately addressed the effects of the proposed EPU on the environmental conditions for and the qualification of electrical equipment.
The NRC staff further concludes that the electrical equipment will continue to meet the relevant requirements of 10 CFR 50.49 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the EQ of electrical equipment.
Offsite Power System Regulatory Evaluation The offsite power system includes two or more physically independent circuits capable of operating independently of the onsite standby power sources. The NRC staff's review covered the descriptive information, analyses, and referenced documents for the offsite power system;and the stability studies for the electrical transmission grid. The NRC staffs review focused on whether the loss of the nuclear unit, the largest operating unit on the grid, or the most critical transmission line will result in the loss of offsite power (LOOP) to the plant following implementation of the proposed EPU. The NRC's acceptance criteria for offsite power systems are based on final GOC-17. Specific review criteria are contained in SRP Sections 8.1 and 8.2, Appendix A to SRP Section 8.2, and Branch Technical Positions (BTPs) PSB-i and ICSB-1 1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the offsite power system and concludes that the offsite power system will continue to meet the 21 
'., t -,, requirements of final GDC-17 follo w~e ~ tin proposed EPU. Adequate physical and electrical separation exists and the offsite power system has the capacity and capability to supply power to all safety loads and other required equipment.
The NRC staff further concludes that the impact of the proposed EPU on grid, stability is insignificant.
Therefore, the NRC staff finds the p'roposed EPU acceptable With respect to the offsite power system.
AC Onsite Power , System Regqulatory Evaluation The alternating current (ac) onsite power system includes those standby power sources, distribution systems, and auxiliary supporting systems provided to supply power to safety-related equipment.
The NRC staff's review covered the descriptive information, analyses, and referenced documents for the ac onsite power system. The NRC's acceptance criteria for the ac onsite power system are based on final GDC-17, insofar as it requires the system to the capacity and capability to perform its intended functions during anticipated operational occurrences and accident conditions.
Specific review criteria are contained in SRP Sections 8.1 and 8.3.1.Technical Evaluation EInsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the ac onsite power system and concludes that the licensee has adequately accounted for the effects of the proposed EPU on the system's functional design. The NRC staff further concludes that the ac onsite power system will continue to meet the requirements of final GDC-17 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the ac onsite power system.2,6:42.3.4 DC Onsite Power System Regulatory Evaluation The direct current (dc) onsite power system includes the do power sources and their distribution and auxiliary supporting systems that are provided to supply motive or control power to safety-related equipment.
The NRC staff's review covered the information, analyses, and referenced documents for the dc onsite power system. The NRC's acceptance criteria for the dc onsite power system are based on (1) draft GDC-24, insofar as it requires that in the event of loss of all offsite power, sufficient alternate sources of power shall be provided to permit the required functioning of the protection systems; and (2) draft GDC-39, insofar as it requires that alternate power systems shall be provided and designed with adequate independency, redundancy, capacity, and testability to permit the functioning required of the engineered safety features.thc system to have the itnddfnci during-antieipated-eper-atiobe~al-surrenaes-and-aeeident-eendition&s Specific review criteria are contained in SRP Sections 8.1 and 8.3.2.Technical Evaluation 22 
[Insert technical evaluation.
The technical evaluationW'i'ould (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the dc onsite power system and concludes that the licensee has adequately accounted for the effects of the proposed EPU on the system's functional design. The NRC staff further concludes that the dc onsite power system will continue to meet the requirements of draft GDCs-24 and 39G04 following implementation of the proposed EPU. Adequate physical and electrical separation exists and the system has the capacity and capability to supply power to all safety loads and other required equipment.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to the dc onsite power system.2,042.3.5 Station Blackout Reaulatorv Evaluation Station blackout (SBO) refers to a complete loss of ac electric power to the essential and nonessential switchgear buses in a nuclear power plant. SBO involves the LOOP concurrent with a turbine trip and failure of the onsite emergency ac power system. SBO does not include the loss of available ac power to buses fed by station batteries through inverters or the loss of power from "alternate ac sources" (AACs). The NRC staff~s review focused on the impact of the proposed EPU on the plant's ability to cope with and recover from an SBO event for the period of time established in the plant's licensing basis. The NRC's acceptance criteria for SBO are based on 10 CFR 50.63. Specific review criteria are contained in SRP Sections 8.1 and Appendix B to SRP Section 8.2; and other guidance provided in Matrix 3 of RS-00i.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the plant's ability to cope with and recover from arn S80 event for the period of time established in the plant's licensing basis. The NRC staff concludes that the licensee has adequately evaluated the effects of the proposed EPU on 880 and demonstrated that the plant will continue to meet the requirements of 10 CFR 50.63 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to SEO.
onc4usion -sectiens-s 2472.4 Instrumentation and Controls 24412.4.1 Reactor Protection, Safety Features Actuation, and Control 23 Systems Reaulatorv Evaluation Instrumentation and control systems are provided (1) to control plant processes having a significant impact on plant safety, (2) to initiate the reactivity control system (including control rods), (3) to initiate the engineered safety features (ESF) systfems and essential auxiliary supporting systems, and (4) for use to achieve and maintain a safe shutdown condition of the plant. Diverse instrumentation and control systems and equipment are provided for the express purpose of protecting against potential common-mode failures of instrumentation and control protection systems. The NRC staff condlucted a review of the reactor trip systerr, engineered safety feature actuation system (ESFAS), safe shutdown systems, control systems, and diverse instrumentation and control systems for the proposed EPU to ensure that the systems and any changes necessary for the proposed EPU are adequately designed such that the systems continue to meet their safety functions.
The NRC staff's review was also conducted to ensure that failures of the systems do not affect safety functions.
The NRC's acceptance criteria related to the quality of design of protection and control systems are based on 10 CFR 5O.55a(a)(1), 10 CFR 5O.55a(h), and final GDC-19 and draft GDCs-1, 12, 13, 14, 15,19, 20, 22, 23, 25, 26, 40, and 42...... 1,1,13......2..2.
23, .nd..2.. Specific review criteria are contained in SRP Sections 7.0, 7.2, 7.3, 7.4, 7.7, and 7.8.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's application related to the effects of the proposed EPU on the functional design of the reactor trip system, ESFAS, safe shutdown system, and control systems. The NRC staff concludes that the licensee has adequately addressed the effects of the proposed EPU on these systems and that the changes that are necessary to achieve the proposed EPU are consistent with the plant's design basis. The NRC staff further concludes that the systems will continue to meet the requirements of 10 CFR 50.55a(a)(1), 10 CFR 50.55(a)(h), and final GDC-19 and draft GDCs-1, 12, 13, 14, 15, 19, 20, 22, 23, 26, 26, 40, and 42. GQC the NRC staff finds the licensee's proposed EPU acceptable with respect to instrumentation and controls.2=82.S Plant Systems 2=842.5.1 "Internal Hazards 2=84-2.5.1.1 Flooding 2444-_-.2.8.1.1.1 Flood Protection Reaqulatory Evaluation The NRC staff conducted a review in the area of flood protection to ensure that SSCs important to safety are protected from flooding.
The NRC staffs review covered flooding of SSCs important to safety from internal sources, such as those caused by failures of tanks and vessels.24 The NRC staffs review focused ntak and vesselsassumed in flooding analyses to assess the impact of any additional fluid on the flooding protection that is provided.
The NRC's acceptance criteria for flood protection are based on draft GDC-2.GDC-2% Specific review criteria are contained in SRP Section 3.4.1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the proposed changes in fluid volumes in tanks and vessels for the proposed EPU. The NRC staff concludes that SSCs important to safety will continue to be protected from flooding and will continue to meet the requirements of draft G DG-2GDC, following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to flood protection.
24,-s4-.2-2.5.1.i.2 Equipment and Floor Drains Regqulatory Evaluation The function of the equipment and floor drainage system (EFDS) is to assure that waste liquids, valve and pump leak-offs, and tank drains are directed to the proper area for processing or disposal.
The EFDS is designed to handle volume of leakage expected, prevent a backflow of water that might result from maximum flood levels to areas of the plant containing safety-related equipment, and protect against the potential for inadvertent transfer of contaminated fluids to an uncontaminated drainage system. The NRC staff's review of the EFDS included the collection and disposal of liquid effluents outside containment.
The NRC staffs review focused on any changes in fluid volumes or pump capacities that are necessary for the proposed EPU and are not consistent with previous assumptions with respect to floor drainage considerations.
The NRC's acceptance criteria for the EFD$ are based on draft GDC-2 GDl~2an4insofar as itthey requires the EFDS to be designed'to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects.the-ru..tur.c..
Specific review criteria are contained in SRP Section 9.3.3.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.].
Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the EFOS and concludes that the licensee has adequately accounted for the plant changes resulting in increased water volumes and larger capacity pumps or piping systems. The NRC staff concludes that the EFDS has sufficient capacity to (1) handle the additional expected leakage resulting from the plant changes, (2) prevent the backflow of water to areas with safety-25 related equipment, and (3) enurotat ~atransferred to non-contaminated drainage systems. Based on this, the NRC staff concludes that the EFOS will continue to meet the requirements of draft GDC-2 GD~e-2-a.nd-4-following implementation of the proposed E~PU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the EFDS. '"
.1.3 Circulating r ystem Regulatory Evaluation The circulating water system (CWS) provides a continuous supply of cooling water to the main condenser to remove the heat rejected by the turbine cycle and auxiliary systems. The NRC staff's review of the CWS focused on changes in flooding analyses that are necessary due to increases in fluid volumes or installation of larger capacity pumps or piping needed to accommodate the proposed EPU. The. NDRO' .c........
criteria, forth.....c...
oe pedes~-opblties-of-sefety-re4ated4-SSG--Specific review criteria are contained in SRP Section 10.4.5.Technical Evaluation Elnsert technic:al evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the modifications to the CWS and concludes that the licensee has adequately evaluated these modifications.
The NRC staff coclde.tat con.isten 
,,,th th.... r .....ent of GD , the increased volumes of fluid leakage that could potentially result from these modifications would not result in the failure of safety-related SSCs following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the CWS.2.5.1.2 Missile Protection 2.5.1.2.1 Internally Generated Missiles Regqulatory Evaluation The NRC staff's review concerns missiles that could result from in-plant component overspeed failures and high-pressure system ruptures.
The NRC staff's review of potential missile sources covered pressurized components and systems, and high-speed rotating machinery.
The NRC staff's review was conducted to ensure that safety-related SSCs are adequately protected from internally generated missiles.
In addition, for cases where safety-related SSCs are located in areas containing non-safety-related SSCs, the NRC staff reviewed the non-safety-related SSCs to ensure that their failure will not preclude the intended safety function of the safety- related SSCs.The NRC staff's review focused on any increases in system pressures or component overspeed conditions that could result during plant operation, anticii~iated operational opcurrences, or changes in existing system configurations such that missile barrier considerations could be affected.
The NRC's acceptance criteria for the protection of SSCs important to SafetY against the effects of internally generated missiles that may result from equipment failures are based on draft 26 GDC-40.GD;G-.4.
Specific review criteria are contained in SRP Sections 3.5.1.1 and 3.5.1.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion T'he NRC staff has reviewed the changes in system pressures and configurations that are required for the proposed EPU and concludes that SSCs important to safety will continue to be protected from internally generated missiles and will continue to meet the requirements of draft GDC-40,GDG-4 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to internally generated missiles.2.5.1.2.2 Turbine Generator Regulatory Evaluation The turbine control system, steam inlet stop and control valves, low pressure turbine steam intercept and inlet control valves, and extraction steam control valves control the speed of the turbine under normal and abnormal conditions, and are thus related to the overall safe operation of the plant. The NRC staff's review of the turbine generator focused on the effects of the proposed EPU on the turbine overspeed protection features to ensure that a turbine overspeed condition above the design overspeed is very unlikely.
The NRC's acceptance criteria for the turbine generator are based on draft GDC-40G90;-4, and relates to protection of SSCs important to safety from the effects of turbine missiles by providing a turbine overspeed protection system (with suitable redundancy) to minimize the probability of generating turbine missiles.
Specific review criteria are contained in SRP Section 10.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the turbine generator and concludes that the licensee has adequately accounted for the effects of changes in plant conditions on turbine overspeed.
The NRC staff concludes that the turbine generator will continue to provide adequate turbine overspeed protection to minimize the probability of generating turbine missiles and will continue to meet the requirements of draft GDC-40-GG-following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the turbine generator.
2.5.1.3 Pipe Failures Regulatory Evaluation 27 The NRC staff conducted a review JAtd -0 p6~to from piping failures outside containment to ensure that (1) such falrswould not cause the loss of needed functions of safety-related systems and (2) the plant coiuldbe safel y shut down in the event of such failures.The NRC staff's review of pipe failures included high and moderate energy fluid system piping located outside of containment.
TheNRC staff's review focused on the effects of pipe failures on plant environmental conditions, control room habitability, and access to areas important to safe control of post-accident operations where the consequences are not bounded by previous analyses.
The NRC's acceptance criteria for pipe failuresyare based on draft GDC-40, insofar as it requires that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures.GDC 
.1, r.....c in. pad, that, c to.
et -ef-eepoeuiate-p tusesrinldlnig-the-effeots of-pi~pe-whlpplnga+4dseharging-fkiId&
Specific review criteria are contained in SRP Section 3.6.1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the changes that are necessary for the proposed EPU and the licensee's proposed operation of the plant, and concludes that SSCs important to safety will continue to be protected from the dynamic effects of postulated piping failures in fluid systems outside containment and will continue to meet the requirements of draft GDC-40-GQG-4 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to protection against postulated piping failures in fluid systems outside containment.
2.5.1.4 Fire Protection Regqulatory Evaluation The purpose of the fire protection program (FPP) is to provide assurance, through a defense-in-depth design, that a fire will not prevent the performance of necessary safe plant shutdown functions and will not significantly increase the risk of radioactive releases to the environment.
The NRC staffs review focused on the effects of the increased decay heat on the plant's safe shutdown analysis to ensure that SSCs required for the safe shutdown of the plant are protected from the effects of the fire and will continue to be able to achieve and maintain safe shutdown following a fire. The NRC's acceptance criteria for the FPP are based on (1) 10 CFR 50.48,-and-eisseeieted-Appedi-R--to-4OGF insofar as ittkey requires the development of an FPP to ensure, among other things, the capability to safely shut down the plant; (2) final GDC-3, insofar as it requires that (a) SSCs important to safety be designed and located to minimize the probability and effect of fires, (b) noncombustible and heat resistant materials be used, and (c) fire detection and fighting systems be provided and designed to minimize the adverse effects of fires on SSCs important to safety; (3) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing. D "" .. .. 
'- mortn o sfe, no b shared ~m ng-uolear-power-Untt, it les ant-:-be-showP,-thae-Whnig-wilkot imp.. their.......
ablt to par.r their.......
st fun....o...
Secific review criteria are contained in SRP Section 9.5.1, as supplemented by the guidance provided in Attachment I to Matrix 5 of Section 2.1 of RS-001.28 T_.echnical Evaluation"[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's fire-related safe shutdown assessment and concludes that the licensee has adequately accounted for the effects of the increased decay heat on the ability of the required systems to achieve and maintain safe shutdown conditions.
The NRC staff further concludes that the FPP will continue to meet the requirements of 103 CFR 50.48, Appendix Rto ICFR Part 50, and final GDC-3, and draft GDC-'1Dc n 6following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to fire protection.
2.5.2 Fission Product Control 2.5.2.1 Fission Product Control Systems and Structures Reoulatory Evaluation The NRC staff's review for fission product control systems and structures covered the basis for developing the mathematical model for OBLOCA dose computations, the values of key parameters, the applicability of important modeling assumptions, and the functional capability of ventilation systems used to control fission product releases.
The NRC staff's review primarily focused on any adverse effects that the proposed EPU may have on the assumptions used in the analyses for control of fission products.
The NRC's acceptance criteria are based on draft GDC-70, insofar as it requires that the plant design include means to control the release of radioactive effluents.GDC- 
!1, inoofar as it requires that the containment atmosphere eleaeup-syetcm bc provideto rdcet eoontration of enirnmn fllwng postulated cidns Specific review criteria are contained in SRP Section 6.5.3._Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a ciear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on fission product control systems and structures.
The NRC staff concludes that the licensee has adequately accounted for the increase in fission products and changes in expected environmental conditions that would result from the proposed EPU. The NRC staff further concludes that the fission product control systems and structures will continue to provide adequate fission product removal in pest-post-accident environments following implementation of the proposed EPU. Based on this, the NRC staff also concludes that the fission product control systems and structures will continue to meet the requirements of draft Therefore, the NRC staff finds the proposed EPU acceptable with respect to the fission product control systems and structures.
29 
-- 9&;Q Zr 2.5.2.2 Main Condenser Evacuation System Reaqulatorv Evaluation The main condenser evacuation system (MCES) generally consists of two subsystems:
the"hogging" or startup system which initially establishes main condenser vacuum and the system which maintains condenser vacuum once it has been established.
The NRC staff's review focused on modifications to the sYstem that may affect gaseous radioactive material handling and release assumptions, arid design features to preclude the possibility of an explosion (if the potential for explosive mixtures exists). The NRC's acceptance criteria for the MCES are based on (1) draft GOC-7OGDG-60, insofar as it requires that the plant design include means to control the release of radioactive effluents; and (2) draft GDC-17G90-&sect;e4, insofar as it requires that means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, inefuding-from anticipated transients, and from accident conditionsoperationab-osur-rencs enpstulated-odents.
Specific review criteria are contained in SRP Section 10.4.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.J Conclusion The NRC staff has reviewed the licensee's assessment of required changes to the MCES and concludes that the licensee has adequately evaluated these changes. The NRC staff concludes that the MCES will continue to maintain its ability to control and provide monitoring for releases of radioactive materials to the environment following implementation of the proposed EPU. The NRC also concludes that the MCES will continue meet~the requirements of draft GDCs-1 7 and 7OGDC 60 and 61. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the MCES.2.5.2.3 Turbine Gland Sealing System Repulatorv Evaluation The turbine gland sealing system is provided to control the release of radioactive material from steam in the turbine to the environment.
The NRC staff reviewed changes to the turbine gland sealing system with respect to factors that may affect gaseous radioactive material handling (e.g., source of sealing steam, system interfaces, and potential leakage paths). The NRC's acceptance criteria for the turbine gland sealing system are based on (1) draft GDC-7OGDQ-6O, insofar as it requires that the plant design include means to control the release of radioactive effluents; and (2) draft GDC-17GDC-64, insofar as it requires that means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, including anticipated transients, and from accident conditionsepaier~~nabaeuncs-and poct-ulatod accidenqts.
Specific review criteria are contained in SRP Section 10.4.3.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and 30 (2) provide a clear link to the conclusionsi l i reached by' th NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of required changes to the turbine gland sealing system and concludes that the licensee has adequately evaluated these changes. The NRC staff concludes that the turbine gland sealing system will continue to maintain its ability to control and provide monitoring for releases of radioactive materials to the environment consistent with draft GDCs-17 and 70GDG-60-and-64.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to the turbine gland sealing system.2.5.2.4 Main Steam Isolation Valve Leakage Control System[Not Applicable.
BFN does not have a MSIV leakage control system.]Redunan,-euiekatin -selatio-,al es-arprcvidedo e "c manseal. Th eaae.... Me..........
44............e-amouP~.ef-direet 
,-et.feate 
...... fr. --
the mnale--aedenainmentis-
~eur-ed....
The NRC staff's rev-iew,'
of the ,MSIV/,akage-eete6 seem.... -...feeI3sed the.eff.ct..a.
theprooe d P on........
the- amouto eakage-s sumed toi ccr.Te,-C-c.panecrtei systems-aenetring contateinmentsee provded det~etinfadfioltionef-phe-tleeser-Sp-oif-2.5.3.1o SpeeafontaFuel Pool Cooling and~CInertu Sytehia Revclulationy TEvautehiaonla~nshu-4}lalyeps-h~e Threpospetfelpoolaprovidestwet setoraeeofspent fe ssemlie* he s~w a f'ety funtion of th s-)pentifue aslearblink tovhered nui onwaer dringache bsthrae NCondti ffons* do NCumstaf eview frthe propuso~sedEUfcusedn nteefcsojtepooe P nth aaiiyo h system-tonproindeaeqate-coolingeeo theaspaenutfely-deurntgdaelle operatingfandpaccident EPUonditioshe assume lepakage throuhtheri for the spCen taffloo coulng condcluestatu sythe arebas4eed nt(1) draft GDC-4,einsofar as r-eator facidiies satl ntshae syk~e-stemsi orby cDm5poTereforness ith C ston afft fins nth ipropoed EPU acetabe whaith; ranec to) dath GDC-%&7., Componha rliben Colndn ecay Heat Reoalssemovaledsge oprvn aaet 2~&92.5.A Spnt Fel Pol Colin an 
* .nv-n*,L the fuel in storage facilities' thauc to plant operating areas a..................n.
be pr.ovded and. (3) draft GDC-69, insofar as containment of fuel shall be provided if accidents could lead to release of undue a'ni6unts of radioactivity to the public environs.
CCC 61, insofar as.t.equre tha fu...... strg"ytm eindwt in&sect;4heqmpeanee4e-safety-of-deeay-heat-reamvalradmaue+s-te-feven-e-sig4icn oso ulsoagfoln"netr ndracdn odtos Speic review criteria are contained in SRP Section 9.1.3, as supplemented by the guidance provided in Attachment i to Matrix 5 of Section 2.1 of RS-OO1.Technical Evaluation EInsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section. ]I Conclusion The NRC staff has reviewed the licensee's assessment related to the spent fuel pool cooling and cleanup system and concludes that the licensee has adequately accounted for the effects of the proposed EPU on the spent fuel pool cooling function of the system. Based on this review, the NRC staff concludes that the spent fuel pool cooling and cleanup system will continue to provide sufficient cooling capability to cool the spent fuel pool following implementation of the proposed EPU and will continue to meet the requirements of draft67 and 69. GDGs-6-A%4r,-at-Therefore, the NRC staff finds the proposed EPU acceptable with respect to the spent fuel pool cooling and cieanup system.2-44422.5.3.2 Station Service Water System Regqulatory Evaluation The station service water system (SWS) provides essential cooling to safety-related equipment and may also provide cooling to non-safety-related auxiliary components that are used for normal plant operation.
The SWS includes the Emergency Equipment Cooling Water (EECW) and the Residual Heat Removal Service Water (RHRSW) systems. The NRC staff's review covered the characteristics of the station SWS (i.e., EECW and RHRSW systems)components with respect to their functional performance as affected by adverse operational (i.e., water hammer) conditions, abnormal operational conditions, and accident conditions (e.g., a LOCA with the LOOP). The NRC staff's review focused on the additional heat load that would result from the proposed EPU. The NRC's acceptance criteria are based on (1) draft GD~s-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as theeffects of a LOCA;oefetatoR 
(..wtr..rn0 matenanoe.4eetn-ad (2) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing. CDC 5, in"sofar as-be~sha ed-amo eng-cear-power-unit~u~s-u ea be-shewn-that-shanng-w4noet-sgfn~fieantty-impafi 
.....effo~m4heiw-safety-fumeioensi-and43)-GDA-44nsoefas~r-a eqtuis-tat-a-system-w~h-the-eapabiity-o-ttasfr-eat~ad em previ-de-.-Specific review criteria are contained in SRP Section 9.2.1, as supplemented by GL 32 89-13 and GL 96-06.Y..UA. ?,2, .i.q.F,-OHI Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment related to the effects of the proposed EPU on the station SWL EECW and RHRSW systems and concludes that the licensee has adequately accounted for the increased heat loads on system performance that would result from the proposed EPU. The NRC staff concludes that the station-SWS EECW and RHRSW systems will continue to be protected from the dynamic effects associated with flow instabilities and provide sufficient cooling for SSCs important to safety following implementation of the proposed EPU. Therefore, the NRC staff has determined that the station-SWS EECW and RHRSW systems will continue to meet the requirements of draft GDCs-4, 40, and 42. GOGs-4,-7 andA44.Based on the above, the NRC staff finds the proposed EPU acceptable with respect to the station SWS EECW and RHRSW systems.2,&082.5.3.3 Reactor Auxiliary Cooling Water Systems Requlatorv Evaluation The NRC staff's review covered reactor auxiliary cooling water systems that are required for (1) safe shutdown during normal operations, anticipated operational occurrences, and mitigating the consequences of accident conditions, or (2) preventing the occurrence of an accident.These systems include non-safety related-eosed-Jee auxiliary cooling water systems, Reactor Building Closed Cooling Water (RBCCW) system and Raw Cooling Water (RCW)system, for reactor system components, reactor shutdown equipment, ventilation equipment, and components of the ECCS. The NRC staff's review covered the capability of the auxiliary cooling water systems to provide adequate cooling water to safety-related ECCS components and reactor auxiliary equipment for all planned operating conditions.
Emphasis was placed on the cooling water systems for safety-related components (e.g., ECCS equipment, ventilation equipment, and reactor shutdown equipment).
The NRC staff's review focused on the additional heat load that would result from the proposed EPU. The NRC's acceptance criteria for the reactor auxiliary cooling water system are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA;GDG-4-, postulated..accidents. 
(2) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing; and (3) draft GDC-41, insofar that the Reactor Auxiliary Cooling Water Systems are relied upon by engineered safety features for performing their safety functions.
GDC 5, insofar as it be shown tha s.haring...
wil net. significantlyH, ability to, perfor thei safety+,
provided-.Specific review criteria are contained in SRP Section 9.2.2, as supplemented by GL 89-13 and GL 96-06.:33 
&.;A;O "h2UL JAjI-'U~l Technical EvaluationS
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conciusion section.]Conclusion.
The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the reactor auxiliary cooling water systems and concludes that the licensee has adequately accounted for the increased heat loads from the proposed EPU on system performance.
The NRC staff concludes that the reactor auxiliary cooling water systems will continue to be protected from the dynamic effects ass~ociated with flow instabilities and provide sufficient cooling for SSCs important to safety following irnplementation of the proposed EPU. Therefore, the NRC staff has determined that the reactor auxiliary cooling water systems will continue to meet the requirements of draft GDCs-4, 40, 41, and Based on the above, the NRC staff finds the proposed EPU acceptable with respect to the reactor auxiliary cooling water systems.2J442.43.4 Ultim ate Heat SinkReuatr Evaluation The ultimate heat sink (UHS) is the source of cooling water provided to dissipate reactor decay heat and essential cooling system heat loads after a normal reactor shutdown or a shutdown following an accident.
The NRC staff's review focused on the impact that the proposed EPU has on the decay heat removal capability of the UHS. Additionally, the NRC staff's review included evaluation of the design-basis UHS temperature limit determination to confirm that post-licensing data trends (e.g., air and water temperatures, humidity, wind speed, water volume) do not establish more severe conditions than previously assumed. The NRC's acceptance criteria for the UHS are based on (1) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is not impaired byte sharing; GDC 5, incofar os-i-eue-htS4-m~ratt ayntb shae-ae&On-nu~ear-peweURmts-Wn~ess~t~-eabe-shRheaPhaRR-Whar4OgSIiR44oGft~t-impafrtheie-ability4e-prfecm-hei-s~afetyj-and-(2) draft GDC-41, insofar that the UHS is relied upon by engineered safety features for performing their safety functions; and (3) draft GDC-52, insofar that the UHS is relied upon by containment heat removal systems for performing their safety functions.
CDCr 11, icofar a-it rngui... that ...t.. with.. tho-oapabitity ta an -normal-oper-atil&sect;and 0ident-eenditiene-be-frevided-.Specific review criteria are contained in SRP Section 9.2.5.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusio.0n The NRC staff has reviewed the information that was provided by the licensee for addressing 34 
?32U JAIOHHCIO the effects that the proposed EPU would have on the UHS &Tafetyfncin including the licensee's validation of the design-basis UHS temperature limit based on post-licensing data.Based on the information that was provided, the NRC staff concludes that the proposed EPU will not compromise the design-basis safety function of the UHS, and that the UHS will continue to satisfy the requirements of draft GDCs-4, 41, and 5..and...
following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the UHS.245-1-2.5.4 Balance-of-Plant Systems Z-6-A0-.2.5.4.1 M amn Steam Rgltr Evaluation The main steam supply system (MSSS) transports steam from the NSSS to the power conversion system and various safety-related and non-safety-related auxiliaries.
The NRC staff's review focused on the effects of the proposed EPU on the system's capability to transport steam to the power conversion system, provide heat sink capacity, supply steam to drive safety system pumps, and withstand adverse dynamic loads (e.g., water steam hammer resulting from rapid valve closure and relief valve fluid discharge loads). The NRC's acceptance criteria for the MSSS are based on (1) draft GDC-40 insofar as it requires that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures;GDC..1, insof....r ac it .re rc htS.,ipran-osaeyb poetd.gi dyna n~et-impingernentes-and (2) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing. G-DC-5,-unessit-enabeshown+
hat-shaing-w~l-net-signifieently-ip.4heirabiilbityt-eflmtb review criteria are contained in SRP Section 10.3.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the MSSS and concludes that the licensee has adequately accounted for the effects of changes in plant conditions on the design of the MSSS. The NRC staff concludes that the MSSS will maintain its ability to transport steam to the power conversion system, provide heat sink capacity, supply steam to steam-driven safety pumps, and withstand steam hammer. The NRC staff further concludes that the MSSS will continue to meet the requirements of draft GDCs-4 and 40.GDCs -'. and 5. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the MSSS.2449_422.5.4.2 Ma in Condenser Regulatory Evaluation 35 The main condenser (MC) system issle~WC di d deaer:ate the exhaust steam from the main turbine and provide a heat sink'f~rdhe tLbinebypass system (TBS). F-o&BW~e-Because BFN does not havewfthaut an MSIV leakage control system, the MC system may-also serves an accident mitigation function to act as a holdup volume for the plate out of fission products leaking through the MSIVs following core damage. The NRC staff's review focused on the effects of the proposed EPU on the steam bypass capability with respect to load rejection assumptions, and on the ability of the MC system to withstand the blowdown effects of steam from the TBS. The NRC's acceptance criteria for the MC system are based on draft GDC-70-GD-6, insofar as it requires that the plant design include' means to control the release of radioactive effluents.
Specific review criteria are contained in SRP Section 10.4.1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the MC system and concludes that the licensee has adequately accounted for the effects of changes in plant conditions on the design of the MC system. The NRC staff concludes that the MC system will continue to maintain its ability to withstand the blowclown effects of the steam from the TBS and thereby continue to meet draft GDC-70&D-4 with respect to controlling releases of radioactive effluents.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to the MC system.24-_10a2.5.4.3 T urbine Bypass Regqulatory Evaluation The TBS is designed to discharge a stated percentage of rated main steam flow directly to the MC system, bypassing the turbine. This steam bypass enables the plant to take step-load reductions up to the TBS capacity without the reactor or turbine tripping.
The system is also used during startup and shutdown to control reactor pressure.
For a BWR without an MSIV leakage control system, the TBS could also provide an accident mitigation function.
A TBS, along with the MSSS and MC system, may be credited for mitigating the effects of MSIV leakage during a LOCA by the holdup and plate out of fission products.
The NRC staffs review for the TBS focused on the effects that the proposed EPU have on load rejection capability, analysis of postulated system piping failures, and the consequences of inadvertent TBS operation.
The NRC's acceptance criteria for the TBS are based on (4-1draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA; GQC-4T,.. ..be dsgncd toccmotchefecsf
..... ...,m atil- with-th...................
condition a...ociat.d "h noma op....ie"-
maineee-tesaF-nd-esulat-edla~cci, t (including......
pipe. brak or..malfunctions oasf he TBSraed (2- leC.-hea ineafa asitr-eq~uir-est r--that r aRHR-sysem e povde t raofefss and .. thc ; des ...n conditnso te "R.... am ,,xcoodo~d.--Specific review criteria are contained in SRP Section 10.4.4.Technical 36 
, .A [Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the TBS. The NRC staff concludes that the licensee has adequately accounted for the effects of changes in plant conditions on the design of the TBS. The NRC staff concludes that the TBS will continue to mitigate the effects of MSIV leakage during a LOCA and provide a means for shutting down the plant during normal operations.
The NRC staff further concludes that TBS failures will not adversely affect essential SSCs. Based on this, the NRC staff concludes that the TBS will continue to meet draft GDCs-40 and 42.GD the NRC staff finds the proposed EPU acceptable with respect to the TBS.%2=5r1442.5.4.4 Condensate and Feedwater Regiulatory Evaluation The condensate and feedwater system (CFS) provides feedwater at a particular temperature, pressure, and flow rate to the reactor. -The only part of the CFS classified as safety-related is the feedwater piping from the NSSS up to and including the outermost containment isolation valve. The NRC staff's review focused on how the proposed EPU affects previous analyses and considerations with respect to the capability of the CFS to supply adequate feedwater during plant operation and shutdown, and isolate components, subsystems, and piping in order to preserve the system's safety function.
The NRC's acceptance criteria for the CFS are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA;GDQ4 7 nofa si euires that SS~csmpotankt to safety be designeado-t a~aemnmodate-theaeffects-ef-and-te-be-oempatible-with-t4e environmen tab- ced einetuding posil flui flwi -bltiee-waeham maintenanc, tosin, an otted-aojns (2) draft GDC-4, insofar as reactor facilities shall not share systems or components unless itis shown safety is not impaired by the s har ing.
tye--ehreamn-a nuc.Jear-power-un454&n~lessA-oae-e-shown that-sharnlnotsgeiieetymaiphefrab4it caiity tto heatlas-rmsafety-relatd SSot- a-eat winkudw ot ora avial frmol tecste-system-oreonly the ofeit syte, asuin sig~ uwe--Specific review criteria are contained in SRP Section 10.4.7.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the CFS and concludes that the licensee has adequately accounted for the effects of changes in 37 plant conditions on the design 6ff~t NClf ocludes that the CFS will continue to maintain its ability to satisfy feedwater requirements for normal operation and shutdown, withstand water hammer, maintain isolation capability in order to preserve the system safety function, and not cause failure of safety-related SSCs. The NRC staff further concludes that the CFS will continue to meet the requirements of draft GDCs-4, 40 iind 42. GD..,. ad1.Therefore, the NRC staff finds the proposed EPU acceptable with respect to the CFS.
Waste Management Systqtms,.! 44-142.5.5.1 Gaseous Waste Management Systems Recqulatory Evaluation The gaseous waste management systems involve the gaseous radwaste system, which deals with the management of radioactive gases collected in the offgas system or the waste gas storage and decay tanks. In addition, it involves the management of the condenser air removal system; the gland seal exhaust and the mechanical vacuum pump operation exhaust; and the building ventilation system exhausts.
The NRC staff's review focused on the effects that the proposed EPU may have on (1) the design criteria of the gaseous waste management systems, (2) methods of treatment, (3) expected releases, (4) principal parameters used in calculating the releases of radioactive materials in gaseous effluents, and (5) design features for precluding the possibility of an explosion if the potential for explosive mixtures exists. The NRC's acceptance criteria for gaseous waste management systems are based on (1) 10 CFR 20.1302, insofar as it provides for demonstrating that annual average concentrations of radioactive materials released at the boundary of the unrestricted area do not exceed specified values; (2) final GDC-3, insofar as it requires that (a) SSCs important to safety be designed and located to minimize the probability and effect of fires, (b) noncombustible and heat resistant materials be used, and (c) fire detection and fighting systems be provided and designed to minimize the adverse effects of fires on SSCs important to safety; (3) draft GDC-7OGDQ-;-60, insofar as it requires that the plant design include means to control the release of radioactive effluents; (4) draft GOC-69GQ0-64, insofar as it requires that containment of fuel and waste storage shall be provided if accidents could lead to release of undue amounts of radioactivity to the public.
and (5) 10 CFR Part 50, Appendix I, Sections 11.8, IIC, and 11.0, which set numerical guides for design objectives and limiting conditions for operation to meet the "as low as is reasonably achievable" (ALARA) criterion.
Specific review criteria are contained in SRP Section 11.3.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment related to the gaseous waste management systems. The NRC staff concludes that the licensee has adequately accounted for the effects of the increase in fission product and amount of gaseous waste on the abilities of the systems to control releases of radioactive materials and preclude the possibility of an explosion if the potential for explosive mixtures exists. The NRC staff finds that the gaseous waste management systems will continue to meet their design functions following implementation of the proposed EPU. The N RC staff further concludes that the licensee has demonstrated that the gaseous waste management systems will continue to meet the requirements of 10 CFR 20.1302;final GQ~e-3GDC-3, draft GDCs-69 and 7060, 5-a-d1t; and 10 CER Part 50, Appendix I, 38 Sections 1I.8, I.C, and 11.0. Thrfrthe NR safin. rpedEPU acceptable with respect to the gaseous waste management systems.2.-4A-42.5.5.2 Liquid Waste Management SystemsReuatr Evaluation The NRC staff's review for liquid waste management systems focused on the effects that the proposed EPU may have on previous analyses and considerations related to the liquid waste management systems' design, design objectives, design criteria, methods of treatment, expected releases, and principal parameters used in calculating the releases of radioactive materials in liquid effluents.
The NRC's acceptance criteria for the liquid waste management systems are based on (1) 10 CFR 20.1302, insofar as it provides for demonstrating that annual average concentrations of radioactive materials released at the boundary of the unrestricted area do not exceed specified values; (2) draft GDC-70GDG-4O, insofar as it requires that the plant design include means to control the release of radioactive effluents; and (3)-GOC-&64-draft GDC-69, insofar as it requires that containment of fuel and waste storage shall be provided if accidents could lead to release of undue amounts of radioactivity to the public environsinsofaaas-it-
+equesht-r e
and (4) 10 CFR Part 50, Appendix I, Sections lI.A and i1.D, which set numerical guides for dose design objectives and limiting conditions for operation to meet the ALARA criterion.
Specific review criteria are contained in SRP Section 11.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment related to the liquid waste management systems. The NRC staff concludes that the licensee has adequately accounted for the effects of the increase in fission product and amount of liquid waste on the ability of the liquid waste management systems to control releases of radioactive materials.
The NRC staff finds that the liquid waste management systems will continue to meet their design functions following implementation of the proposed EPU. The NRC staff further concludes that the licensee has demonstrated that the liquid waste management systems will continue to meet the requirements of 10 CFR 20.1302; draft GDCs-69 and 70GQs-& -an6; and 10 CFR Part 50, Appendix I, Sections lI.A and lI.D. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the liquid waste management systems.-2A-_t32.5.5.3 Solid Waste Management Systems Regjulatorv Evaluation The NRC staff's review for the solid waste management systems :(SWMS) focused on the effects that the proposed EPU may have on previous analyses and considerations related to th~e design objectives in terms pf expected volumes of waste to be processed and handled, the wet and dry types of waste to be processed, the activity and expected radionuclide distribution contained in the waste, equipment design capacities, and the principal parameters emploYed in 39 r r the design of the SWMS. The NRC's paSWMS are based on (1) 10 CFR 20.1302, insofar as it provides for demonstrating that annual average concentrations of radioactive materials released at the boutidary of the unrestricted area do not exceed specified values; (2) draft GDC-70G4C-6O, insofar as it requires that the plant design include means to control the release of radioactive effluents; (3) draft;GDC-18GD-6, insofar as it requires that monitoring and alarm instrumentation shall be provided for fuel and waste storage and handling areas for conditions that might contribute to loss of continuity in decay heat removal and to radiation exposureseystenebeprviedinwete-bantin-a~eae~e-deteet 6eondtiosht--may-eeti-xeeve-diatioemvele, (4) draft GDC-1 7GQG-64, insofar as it requires that means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, from anticipated transients, and from accident conditionsielnitg AQOc,,,''', and postulte ...c...nt.; and (5) 10 CFR Part 71, which states requirements for radioactive material packaging.
Specific review criteria are contained in SRP Section 11.4.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment related to the SWMS. The NRC staff concludes that the licensee has adequately accounted for the effects of the increase in fission product and amount of solid waste on the ability of the SWMS to process the waste. The NRC staff finds that the SWMS will continue to meet its design functions following implementation of the proposed EPU. The NRC staff further concludes that the licensee has demonstrated that the SWMS will continue to meet the requirements of 10 CFR 20.1302, draft GDCs-17, 18, and 7OG s6, 6, an ', and 10 CFR Part 71. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the SWMS.%&42.5.6 Additional Considerations 2.-4-4 --4-2,5.6.1 Emergency Diesel Engine Fuel Oil Storage and Transfer System Recqulatorv Evaluation Nuclear power plants are required to have redundant onsite emergency power supplies of sufficient capacity to perform their safety functions (e.g., power diesel engine-driven generator sets), assuming a single failure. The NRC staff's review focused on increases in emergency diesel generator electrical demand and the resulting increase in the amount of fuel oil necessary for the system to perform its safety function.
The NRC's acceptance criteria for the emergency diesel engine fuel oil storage and transfer system are based on (1) draft GDC-40 insofar as it requires that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures; GD ............
r... it re..re that-39Cc mp44at-9 feree coiao wit piebok; (2) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing;GQC-,, ,iteqtfe4hat-33Cc mportant-to sa fetnebesadam g nucioer pwe units uessita-eshewnmtat-shar4&sect;-i4-ne (3) final GDC-17, insofar as it requires onsite power supplies to have sufficient independence and redundancy to perform their safety functions, assuming a single 40 faiure Seciicreview criteria are contained in SRP Sectin954 Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section,]The NRC staff has reviewed the licensee's assessment related to the amount of required fuel oil for the emergency diesel generators and concludes that the licensee has adequately accounted for the effects of the increased electrical demand on fuel oil consumption.
The NRC staff concludes that the fuel oil storage and transfer system will continue to provide an adequate amount of fuel oil to allow the diesel generators to meet the onsite power requirements of final GDC-17 and draft GDCs-4, and ,-4rran Therefore, the NRC staff finds the proposed EPU acceptable with respect to the fuel oil storage and transfer system.%.&4242.5.6.2 Light Load Handling System (Related to Refueling)
Reciulatorv Evaluation The light load handling system (LLHS) includes components and equipment used in handling new fuel at the receiving station and the loading of spent fuel into shipping casks. The NRC staff's review covered the avoidance of criticality accidents, radioactivity releases resulting from damage to irradiated fuel, and unacceptable personnel radiation exposures.
The NRC staff's review focused on the effects of the new fuel on system performance and related analyses.
The NRC's acceptance criteria for the LLHS are based on (1) draft GDCs-68 and 69gc6, insofar as theylt requires that systems that contain radioactivity be designed with appropriate oonfinemeit-containment and with suitable shielding for radiation protection; and (2) draft GDC-660G62, insofar as it requires that criticality be prevented.
Specific review criteria are contained in SRP Section 9.1.4.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the new fuel on the ability of the LLHS to avoid criticality accidentsaand concludes that the licensee has adequately incorporated the effects of the new fuel in the analyses.
Based on this review, the NRC staff further concludes that the LLHS will continue to meet the requirements of draft GDCs-66, 68, ad6G s 61 a nd6 for radioactivity releases and prevention of criticality accidents.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to the LLHS.
f4nsert-Requlate'yE~auatoie-,Te~hnie4lEvcaluatienran-Gesusioseostiens-s nesessarA 41 2.6 Containment Review Conside :tons .A i , 2.6.1 Primary Containment Functional r -. Regqulatory Evaluation The containment encloses the reactor system and is the final barrier against the release of significant amounts of radioactive fission products in the event of an accident.
The NRC staffs review for the primary containment functional design covered (1) the temperature and pressure conditions in the drywell and wetwell due to a spectrum of postulated LOCAs, (2) the differential pressure across the operating deck for a spectrum of LOCAs (Mark II containments only), (3)suppression pool dynamic effects during a LOCA or following the actuation of one or more RCS safety/relief valves, (4) the consequences of a LOCA occurring within the containment (wetwell), (5) the capability of the containment to withstand the effects of steam bypassing the suppression pool, (6) the suppression pool temperature limit during ROS safety/relief valve operation, and (7)the analytical models used for containment analysis.
The NRC's acceptance criteria for the primary containment functional design are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA; GQ I noa as " "requre with-the.-envir.............
dit...... 
.......ted-w..............pe 
.......,,,,, i't an.. c , to s,.,ng, and ,.'pastulated-eeidentsrandth at-suehWSSs-be-proteeted 
*against-dlynamrnmefeots} 
(2) draft G DC-I0, insofar as it requires that reactor containment be designed to sustain the initial effects of gross equipment failures, such as a large coolant boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain functional capability for as long as the situation requires; GD-,&
to. ct a ..i.. n...... en......... , ....-tlght-barrier-against-the-uno-entrolled-elease ef-radloaatvityo-te-ewrnent (3) draft GDC)-49, insofar as it requires that the containment and its associated heat removal systems be designed so that the containment structure can accommodate, without exceeding the design leakage rate, the pressures and temperatures resulting from the largest credible energy release following a LOCA, including considerable margin for effects from metal-water or other chemical reactions that could occur as a consequence of failure of emergency t;ore cooling systems; GDC50 ins=Rofar" ; ...asi&#xf7;q4fehet ts asseeatadha 
...... o..a.. sy..tem..bc d...n. so that.......
conta.. in. nt...t.u.t..r..c.n a~eomredater-witheu-x~eeedinff-the-desigrnleakegc rate anrd wi!th cuffiientmargiPrthe-
.... (4) draft GDnC-I2 insofar as it requires that instrumentation and controls be provided as required to monitor and maintain variables within prescribed operating ranges;CDC 13-, nsofar as' it. requf ,.rcz" (5) draft GDC-.17GD-6, insofar as it requires that means be provided to monitor the reactor containment atmosphere for radioactivity that may be released from normal operations, from anticipated transients, and from accident conditionspeetdaed-aodet.
Specific review criteria are contained in SRP Section 6.2.1.1.C.
Technical Evaluation EInsert technical evaluation.
The technical evaluation should (I) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion 42 The NRC staff has reviewed the licensesassessmeri of the containment temperature and pressure transient and concludes that the licensee has adequately accounted for the increase of mass and energy resulting from the proposed EPU. The NRC staff further concludes that containment systems will continue to provide sufficient pressure and temperature mitigation capability to ensure that containment integrity is maintained.
The NRC staff also concludes that containment systems and instrumentation will continue to be adequate for monitoring containment parameters and release of radioactivity during normal and accident conditions and the containment and associated systems will continue to meet the requirements of draft GDCs-10, 12, 17, 40, 42, and 49 GDC.__ .1, 13., 50,n 61! following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to primary containment functional design.2.6.2 Subcompartment Analyses Reaqulatorv Evaluation A subcompartment is defined as any fully or partially enclosed volume within the primary containment that houses high-energy piping and would limit the flow of fluid to the main containment volume in the event of a postulated pipe rupture within the volume. The NRC staff's review for subcompartment analyses covered the determination of the design differential pressure values for containment subcompartments.
The NRC staff's review focused on the effects of the increase in mass and energy release into the containment due to operation at EPU conditions, and the resulting increase in pressurization.
The NRC's acceptance criteria for subcompartment analyses are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA; GDC-4, insofar-ae-it-eukshat-SS-Gpe-ipo~nt-e-safemye-designed-te-a~cemmedate-the effe~ts-of the-ernvkonentak-eeadit ens- ssooeedith-nerma~eperatioen'maintena eetesting~and GDC-49, insofar as it requires that the containment be designed so that the containment structure can accommodate, without exceeding the design leakage rate, the pressures and temperatures resulting from the largest credible energy release following a LOCA.across the... , walls , of .th subcom.rtmnt Specific review criteria are contained in SRP Section 6.2.1.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the subcornpartment assessment performed by the licensee and the change in predicted pressurization resulting from the increased mass and energy release.The NRC staff concludes that containment SSCs important to safety will continue to be protected from the dynamic effects resulting from pipe breaks and that the subccmpartments will continue to have sufficient margins to prevent fracture of the structure due to pressure difference across the walls following implementation of the proposed EPU. Based on this, the NRC staff concludes that the plant will dontinue to meet draft GDCs-40, 42 and 49 GDs' and......for the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to subcompartment analyses.43 2.6.3 Mass and Energy Release 2.6.3.1 Mass and Energy Release Analysis for Postulated Loss of Coolant Regqulatory Evaluation 
,' 'The release of high-energy fluid into containment from pipe breaks could challenge the structural integrity of the containment, including subcompartments and systems within the containment.
The NRC staff's review covered the energy sources that are available for release to the containment and the mass and energy release rate calculations for the initial blowdown phase of the accident.
The NRC's acceptance criteria for mass and energy release analyses for postulated LUCAs are based on (1) draft GDC-49, insofar as it requires that the containment and its associated heat removal systems be designed so that the containment structure can accommodate, without exceeding the design leakage rate, the pressures and temperatures resulting from the largest credible energy release following a L O CA;-GDC , -the-mass--
and-energy-release-,na~aysi~s-tassare~hateentainment-designagn~ateined and (2) 10 CFR Part 50, Appendix K, insofar as it identifies sources of energy during a LOCA. Specific review criteria are contained in SRP Section 6.2.1.3.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's mass and energy release assessment and concludes that the licensee has adequately addressed the effects of the proposed EPU and appropriately accounts for the sources of energy identified in 10 CFR Part 50, Appendix K.Based on this, the NRC staff finds that the mass and energy release analysis meets the requirements in draft GDC-49 ensuring that the analysis is conservative.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to mass and energy release for postulated LOCA.2.6.4 Combustible Gas Control in Containment Regulatory Evaluation Following a LOCA, hydrogen and oxygen may accumulate inside the containment due to chemical reactions between the fuel rod cladding and steam, corrosion of aluminum and other materials, and radiolytic decomposition of water. If excessive hydrogen is generated, it may form a combustible mixture in the containment atmosphere.
The NRC staff's review covered (1) the production and accumulation of combustible gases, (2) the capability to prevent high concentrations of combustible gases in local areas, (3) the capability to monitor combustible gas concentrations, and (4) the capability to reduce combustible gas concentrations.
The NRC staff's review primarily focused on any impact that the proposed EPU may have on hydrogen release assumptions, and how increases in hydrogen release are mitigated.
The NRC's acceptance criteria for combustible gas control in containment are based on (1) 10 CFR 50.44, insofar as it requires that plants be provided with the capability for controlling combustible gas concentrations in the containment atmosphere; and (2) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is 44 not impaired by the sharing.CD5, insofr a- t re be ..hared am ........e pewer-N niteuantess hteh n-44nt-tt~eeAy-sj-4GDC*-4-1inse fs-4r-a qu#restIa-into the reactor containment-fellewi~ng-posuaeds cdente teensure 
... i ,,e.t4ntegty-.
D-(42A-G oar-ase4t-equ~es to-permitapp.epoiateperodis n
requaire yOO1 esi de ~ed -pamit-a Jdi~estig -flndd t-he-fellewig sentenc.e for-BWAswith Mark Il conami" nt: ,e conaimnttat dotrl ona-nertd tmshoo o otrolhdoe nie-the o-entafnment]-Specific review criteria are contained in SRP Section 6.2.5.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment related to combustible gas and concludes that the plant will continue to have sufficient capabilities consistent with the requirements in 10 CFR 50.44 and draft GDCA4, G-DP=s-&4, 42-1,4,an~4-4,-as discussed above.Therefore, the NRC staff finds the proposed EPU acceptable with respect to combustible gas control in containment.
2.6.5 Containment Heat Removal Requlatorv Evaluation Fan cooler systems, spray systems, and residual heat removal (RHR) systems are provided to remove heat from the containment atmosphere and from the water in the containment wetwell.The NRC staff's review in this area focused on (1) the effects of the proposed EPU on the analyses of the available net positive suctjon head (NPSH) to the containment heat removal system pumps and (2) the analyses of the heat removal capabilities of the spray water system and the fan cooler heat exchangers.
The N RC's acceptance criteria for containment heat removal are based on draft GDCs-41 and 52, insofar as they require that a containment heat removal system be provided, and that its function shall be to prevent exceeding containment design pressure under accident conditions.
e a cnanetha-emoval-system-be-pr-ovide t he qta-cmain-on rcc e lew4eve Specific review criteria are contained in SRP Section 6.2.2, as supplemented by Draft Guide (DG) 1107.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion 45 
* ;' * ; , *.- -..The NRC staff has reviewed the imv ssesaesenprvddb the licensee and concludes that the licensee has adequately addressed the effects of the proposed EPU. The NRC staff finds that the systems will continue to meet draft GDCs-41 and 52 GD4&=-3with respect to rapidly reducing the containment pressure and temperature following a LOCA and maintaining them at acceptably low levels. Therefore, the NRC staff finds the proposed EPU acceptable with respect to containment heat remd6val systems.2.6.6 Secondary Containment Functional Design Regulatory Evaluation The secondary containment structure and supporting systems of dual containment plants are provided to cpllect and process radioactive material that may leak from the primary containment following an accident.
The supporting systems maintain a negative pressure within the secondary containment and process this leakage. The NRC staff's review covered (1) analyses of the pressure and temperature response of the secondary containment following accidents within the primary and secondary containments; (2) analyses of the effects of openings in the secondary containment on the capability of the depressurization and filtration system to establish a negative pressure in a prescribed time; (3) analyses of any primary containment leakage paths that bypass the secondary containment; (4) analyses of the pressure response of the secondary containment resulting from inadvertent depressurization of the primary containment when there is vacuum relief from the secondary containment; and (5) the acceptability of the mass and energy release data used in the analysis.
The NRC staff's review primarily focused on the effects that the proposed EPU may have on the pressure and temperature response and drawdown time of the secondary containment, and the impact this may have on offsite dose. The NRC's acceptance criteria for secondary containment functional design are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCACDC- .,.in..far....
reqio th" S " imporant te-safety-be-designed4e-aeeemwnoate-the-ef rnermab-eperatonermintenaaee 7 4estiag~aandpstttd-a reteerdmm dynamie.effet~~effe -ef uds}4hat-may resu~t-frram-equipment4ailur-esj-and (2) draft GDC-1O, insofar as it requires that reactor containment be designed to sustain the initial effects pf gross equipment failures, such as a large coolant boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain functional capability for as long as the situation requires.
GD 6 zfrasi eurstat ractor-containment-an,-
uneeflt-role~~se~ff-adieaetivty~etee4wimnament---Specific review criteria are contained; in SRP Section 6.2.3.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment related to the secondary containment pressure and temperature transient and the ability of the secondary containment to provide an essentially leak-tight barrier against uncontrolled release of radioactivity to the environment.
The NRC staff concludes that the licensee has adequately accounted for the increase of mass and energy that would result from the proposed EPU and further concludes that the secondary 46 containment and associated systems will continue to po'de an esnilyLa-ih are against the uncontrolled release of radioactivity to the environment following implementation of the proposed EPU. Based on this, the NRC staff also concludes that the secondary containment and associated systems will eontinue to meet the requirements of draft GDCs-1O, 40 and 42.GDc' ad6 Therefore, the NRC staff finds the proposed EPU acceptable with respect to secondary containment functional design..[Adiio aluRc rat*ConTaine" "viw Gned "t es 2.7 Habitability, Filtration, and Ventilation 2.7.1 Control Room Habitability System Regqulatory Evaluation The NRC staff reviewed the control room habitability system and control building layout and structures to ensure that plant operators are adequately protected from the effects of accidental releases of toxic and radioactive gases. A further objective of the NRC staff's review was to ensure that the control room can be maintained as the backup center from which technical support center personnel can safely operate in the case of an accident.
The NRC staff's review focused on the effects of the proposed EPU on radiation doses, toxic gas concentrations, and estimates of dispersion of airborne contamination.
The NRC's acceptance criteria for the control room habitability system are based on (1) draft GDC-40, insofar as it requires that, protection for e'ngineered safety features shall be provided againstdynamic effects and missiles that might result from plant equipment failures-SS~s-importan4t-e-safety-be-dcndtoaccommodato tho offocts of and to bo comptbl wit th ovrnmontal conditiors assesiated-wfth-postutated-eeeiden~tirncludin&sect;Ahe-effests-oef-h~e~ese-ef4oxis ases; and (2)final GDC-1 9 and 10 CFR 50.67, insofar as it requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent, to any part of the body, for the duration of the accident.
Specific review criteria are contained in SRP Section 6.4 and other guidance provided in Matrix 7 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment related to the effects of the proposed EPU on the ability of the control room habitability system to protect plant operators against the effects of accidental releases of toxic and radioactive gases. The NRC staff concludes that the licensee has adequately accounted for the increase of toxic and radioactive gases that would result from the proposed EPU. The NRC staff further concludes that the control room habitability system will continue to provide the required protection following implementation of the proposed EPU. Based on this, the NRC staff concludes that the control room habitability system will continue to meet the requirements of draft GDC-40 and final GDJC-1 9 and 10 CFR 50:67. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the control room habitability system.47 2.7.2 Engineered Safety Feature Atmosphere Cleanup Recqulatorv Evaluation 
, ESE atmosphere cleanup systems 'are designed for fissiqn prbduct removal in post-accident environments.
These systems generally include primary systems (e.g., in-containment recirculation) and secondary systems (eg., standby gas treatment systems and emergency or post-accident air-cleaning systems) for the fuel-handling building, control room, shield buIlding, and areas containing ESF components.
For each ESF atmosphere cleanup system, the NRC staff's review focused on the effects of the proposed EPU on system functional design, environmental design, and provisions to preclude temperatures in the adsorber section from exceeding design limits. The NRC's acceptance criteria for ESF atmosphere cleanup systems are based on (1) final GDC-19 and 10 CFR 50.67, insofar as it requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent, to any part of the body, for the duration of the accident; (2) ODO -411,insofai-as4t-rerues4hat-system~s4ooeto-eeflssien prmded4e-reuee-the-eecn " io-n --uliisiea-pfduets-F4eesede4-thenevironment GDC-70, insofar as it requires that the plant maintain control over the radioactive effluents during normal operation and for any transient situation-;
and GD 1,isfar as reqire that systms that maysonai radioact;vit b desigedt-assure--adequate-safet y une'omladpcuae acien od in; od-(34)GDG-64draft GDC-17, insofar as it requires that means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences (AOOs), and postulated accidents.
Specific review criteria are contained in SRP Section 6.5.1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the ESE atmosphere cleanup systems. The NRC staff concludes that the licensee has adequately accounted for the increase of fission products and changes in expected environmental conditions that would result from the proposed EPU, and the NRC staff further concludes that the ESF atmosphere cleanup systems will continue to provide adequate fission product removal in postaccident environments following implementation of the proposed EPU.Based on this, the NRC staff concludes that the ESF atmosphere cleanup systems will continue to meet the requirements of 10 CFR 50.67, final GDC-19 and draft GDCs-17 and 70. GD~s-4-t4-,44 -t--an464-Therefore, the NRC staff finds the proposed EPU acceptable with respect to the ESF atmosphere cleanup systems.2.7.3 Control Room Area Ventilation System Regqulatory Evaluation The function of the control room area ventilation system (CRAVS) is to provide a controlled environment for the comfort and safety of control room personnel and to support the operability of control room components during normal operation, AOOs, and OBA conditions.
The NRC's 48 review of the CRAVS focused on the effects EPU will have on the functional performance of safety-related portions of the system. The review included the effects of radiation, combustion, and other toxic products; and the expected environmental conditions in areas served by the CRAVS. The NRC's acceptance criteria for the CRAVS are based on'(1) draft GDC..40GG, insofar as it requires that protection for engineered safety features be provided against dynamic effects and missiles that might result from plant equipment failu fety-be-designe44e-aeeemmoedat ......c..f.an to4 be" , eide~ts;(2) final GDC-19, insofar as it requires that adequate radiation protection be provided to permit access and' occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body; or its equivalent to any part of the body, for the duration of the accident; and (3) draft GDC-70, insofar as it requires that the plant design include means to control the release of radioactive effluents.
Specific review criteria are contained in SRP Section 9.4.1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the ability of the CRAVS to provide a controlled environment for the comfort and safety of control room personnel and to support the operability of control room components.
The NRC staff concludes that the licensee has adequately accounted for the increase of toxic and radioactive gases that would result from a DBA under the conditions of the proposed EPU, and associated changes to parameters affecting environmental conditions for control room personnel and equipment.
Accordingly, the NRC staff concludes that the CRAVS will continue to provide an acceptable control room environment for safe operation of the plant following implementation of the proposed EPU. The NRC staff also concludes that the system will continue to suitably control the release of gaseous radioactive effluents to the environment."Based on this, the NRC staff concludes that the CRAVS will continue to meet the requirements of final GDC-19 and draft GDCs-40 and 70s-4,-44,and4-8.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to the CRAVS.2.7.4 Spent Fuel Pool Area Ventilation System R ePuatr--valuat4ie
[Section 2.7.4 is not applicable to BFN]-The4 a n 0 Abeef~epent-fuet-pee-ea-venti~atien-system-QSF PA h ffes fth esed-EP-on4hJete-funetieefa--
pefrac ftesft eltdproso -The NRC's accapacocie~a4ar4he-which conai eadoaetity-b~e-de.
riate-seefinemenbd-enament peslfie-review 49 a.-r EliInert technical evaluation.
The technical evaluation should (1) clearly explain why the-peposed-ohangcn satisfy each ofthe requirements in the-wgulato~y-evatuation-and Conclusion r..quirement.
of GD. c and 61. There..ore,.
the. NRC&#xa2; st-ff finds the proposed-EPU-aeetble-2.7.5 yen-Rdaste.Are-and Turbine...re..Reactor, Turbine, and Radwaste Building Ventilation Systems 245-Regqulatory Evaluation The function of the ........, ...Turbine and Radwaste Building vVentilation eSystem to maintain ventilation in the uiir ndrdat eqimn an turbine..a...sreactor, turbine, and radwaste buildings to permit personnel access, and control the concentration of airborne radioactive material in these areas during normal operation, during AQOs, and after postulated accidents.
The NRC staff's review focused on the effects of the proposed EPU on the functional performance of the safety-related portions of these systems. The NRC's acceptance criteria for the ARV ndTvsystems are based on draft GDC-7OGQ-6, insofar as it requires that the plant design include means to control the release of radioactive effluents.
Specific review criteria are contained in SRP Sections 9.4.3 and 9.4.4.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the ARAVS-and-T4ASReactor, Turbine, and Radwaste Building Ventilation System. The NRC staff concludes that the licensee has adequately accounted for the effects of the proposed EPU on the capability of these systems to maintain ventilation in the aux4+iar-y-aend-adweste-eqimn areas. and. in.. the tu.. +,e-a,. turbine, and radwaste buildings to permit personnel access, control the concentration of airborne radioactive material in these areas, and control release of gaseous radioactive effluents to the environment.
Based on this, the NRC staff concludes that the ARA-V-S-an4-T-AVSsystems will continue to meet the requirements of draft Therefore, the NRC staff finds the proposed EPU acceptable with respect to the Reactor, Turbine, and Radwaste Building Ventilation SystemARAVS and the:TAVS.50 247142.7.6 Engineered Safety Feature Ventilation System Requlatory Evaluation The function of the engineered safety feature ventilation system (ESFVS) is to provide a suitable and controlled environment for ESF components following certain anticipated transients and DBAs. The NRC staff's review for the ESFVS focused on the effects of the proposed EPU on the functional performance of the safety-related portions of the system. The NRC staff's review also covered (1) the ability of the ESF equipment in the areas being serviced by the ventilation system to function under degraded ESFVS performance; (2) the capability of the ESFVS to circulate sufficient air to prevent accumulation of flammable or explosive gas or fuel-vapor mixtures from components (e.g., storage batteries and stored fuel); and (3) the capability of the ESFVS to control airborne particulate material (dust) accumulation.
The NRC's acceptance criteria for the ESFVS are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA;GDC 1, incofnr as-it-iequies-that fety-e-desi44ed4e~aeomed e-e he-effeot-fade-e oomat~~wh~eevkonmenat dteeseeatedwit h-oe~a~oel~atieo,-maPitenanee, testin, an'otltdacdente;-(2) final GDC-17, insofar as it requires onsite and offsite electric power systems be provided to permit functioning of SSCs important to safety; and (3)draft insofar as it requires that the plant design include means to control the release of radioactive effluents.
Specific review criteria are contained in SRP Section 9.4.5.Technical Evaluation Elnsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section,]Conclusion The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on the ESFVS. The NRC staff concludes that the licensee has adequately accounted for the effects of the proposed EPU on the ability of the ESFVS to provide a suitable and controlled environment for ESF components.
The NRC staff further concludes that the ESFVS will continue to assure a suitable environment for the ESF components following implementation of the proposed EPU. The NRC staff also concludes that the ESFVS will continue to suitably control the release of gaseous radioactive effluents to the environment following implementation of the proposed E2PU. Based on this, the NRC staff concludes that the ESFVS will continue to meet the requirements of final GDC-1 7 and draft G DCs-40, 42 and 70.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to the ESFVS.[tnsertRegu late~y-Evahuatiern-Teohn~iaa-EvatuatieoP-ad-Gel usi 2.8 Reactor Systems 2.8.1 Fuel System Design Regulatory Evaluation S51 
-- ------ ---The fuel system consists of arrays of fuel rods, .burnable poison rods,.spacer grids and springs, end plates, channel boxes, and reactivity control rods. The NRC staff reviewed the fuel system to ensure that (1) the fuel system is not damaged. s~eulo~~~
operation and AOOs, (2) fuel system damage is never so severe as to prevent conftrl r~d 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 NRC staff's 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 NRC's acceptance criteria are based on (1) 10 CFR 50.46, insofar as it establishes standards for the calculation of emergency core cooling system (ECCS) performance and acceptance criteria for that calculated performance: 
(2) final GDC-10, insofar as it requires that the reactor core be designed with appropriate margin to assure that SAFOLs are not exceeded during any condition of normal operation, including the effects of AOOs; (3) final G DC-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; end (4) final 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.Specific review criteria are contained in SRP Section 4.2 and other guidance provided in Matrix 8 of RS-001.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the fuel system design of the fuel assemblies, control systems, and reactor core. The NRC staff concludes that the licensee has adequately accounted for the effects of the proposed EPU on the fuel system and demonstrated that (1) the fuel system will not be damaged as a result of normal operation and AQOs, (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) coolability will always be maintained.
Based on this, the NRC staff concludes that the fuel system and associated analyses will continue to meet the requirements of 10 CFR 50.46, final GDC-1O, GDC-27, and GDC-35 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the fuel system design.2.8.2 Nuclear Design Regulatory Evaluation The NRC staff reviewed the nuclear design of the fuel assemblies, control systems, and reactor core to ensure that fuel design limits will not be exceeded during normal operation and anticipated operational transients, and that the effects of postulated reactivity accidents will not cause significant damage to the RCPB or impair the capability to cool the core. The NRC staffs review covered core power distribution, reactivity coefficients, reactivity control requirements and control provisions, control rod patterns and reactivity worths, criticality, burnup, and vessel irradiation.
The NRC's acceptance criteria are based on (1) final GDC-10, insofar as it requires 52 that the reactor core be to assure that SAFDLs are not exceeded during any condijion of normal operatio~n, including the effects of AOOs; (2) draft GDC-8, insofar as it requires that the reactor core be designed so that the overall power coefficient in the power operating range shall not be positive; G004-14,-nsofar~as4t-requies.tht.thereactr.cor.be.dsigne..o.tat.th nt..effect..of the..........
prmt-(3 inhrent nucla 7GQG.:-12, insofar as it requires that the core design shall ensure that power oscillations which could cause damage in excess of acceptable fuel damage limits are not possible or can be readily suppressed; inscore....be .designedh 
....to..as.sucre that-pewe-eeeiloins hch can... reul in conito. , xceed.. ,ing, SOrs, are. ..... not pas~l or... ;""^oawbe-eably-an teadtlydeteeteedand-suppressedi-(4) d raft GD~s -12 a nd 13 in sofrar as they require that instrumentation and controls be provided as required to monitor and maintain variables within prescribed operating ranges through the core life;G-DC-.3-, rangesj (6) draft GDCs-14 and 15, insofar as they require that the protection system be designed to initiate the reactivity control systems automatically to prevent or suppress conditions thait could result in exceeding acceptable fuel damage limits and to initiate operation of ESFs under accident situations;GDG=-:20,-
eofar-as4tequires~hat-the-that-aeetbe fue deig liit ar not..exceeded as a result of draft GDC-31, insofar as it requires that the reactivity control systems be capable of sustaining any single malfunction without causing a reactivity transient which could result in exceeding acceptable fuel damage limits; preetinssembe-designed.to assur tha SAOsa o xeeded feeany-ing-mafnto-fthe-reactivity-e~nt491-systemsj; (7) draft GDCs-27 and 28 insofar as they require that at least two independent reactivity control systems be provided, with both systems capable of making and holding the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits; (8) draft GIJCs-29 and 30, insofar as they require that at least one of the reactivity control systems be capable of making and holding the core subcritical under any condition sufficiently fast to prevent exceeding acceptable fuel damage limits; GD-265nsofeeet-r-equk~e-that-final 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 ECOS, 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 (910) draft GDC-32, insofar as it requires that limits, which include considerable margin, be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling. GC2," inoa "si reufres-hath reactivity..
cotrlsytmsb stouresoreother-eie ceore-Specific review criteria are contained in SRP Section 4.3 and other guidance provided in Matrix 8 of RS-0O1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the 53 proposed changes satisfy each oft e requ emrregulatory evaluation and (2)provide a clear link to the conclusibns reached by the NRC staff, as documented in the conclusion section.]C.onclusion 2 ."i The NRC staff has reviewed the licensee's analyses related to the effect of the proposed EPU on the nuclear design of the fuel assemblies, control systems, and reactor core. The NRC staff concludes that the licensee has adequately accounted forethe effects of the proposed EPU on the nuclear design and has demonstrated that the fuel design Jimits wilJ not be exceeded during normal or anticipated operational transients, and that 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, the NRC staff concludes that the nuclear design of the fuel assemblies, control systems, and reactor core will continue to meet the applicable requirements of final GDCs-l0 and 27, and draft GDCs-7, 8, 12, 13, 14, 15, 27, 23, 29, 30, 31 and 32.. .. 00c1,1,1,1,2,2-26,2,ed2.Therefore, the NRC staff finds the proposed EPU acceptable with respect to the nuclear design.2.8.3 Thermal and Hydraulic Design Regulatory Evaluation The NRC staff reviewed-the thermal and hydraulic design of the core and the RCS to confirm that 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 AQOs, 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 anticipated transients without scram (ATWS) events. The NRC's acceptance criteria are based on (1) final GDC-1 0, 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 AQOs; and (2) draft GDC-7, insofar as it requires that the core design shall ensure that power oscillations which could cause damage in excess of acceptable fuel damage limits are not possible or can be readily aemcatcd cclntreonro, 'n rteto sytm b -~g oascurehat-powee-eolnhtiens-u-wia-h F-DL-s-,areoet-possible-er-an-reliably-and readily be detected and cupprecsed.
Specific review criteria are contained in SRP Section 4.4 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation Elnsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the thermal and hydraulic design of the core and the RCS. The NRC staff concludes that the licensee has adequately accounted for the effects of the proposed EPU on the thermal and hydraulic design and demonstrated that the design (1) has been accomplished using acceptable 54 analytical methods, (2) is [equivalenfi1btio from] proven designs, (3)provides acceptable margins of safety, from conditions thatF~ould lead to fuel damage during normal reactor operation and AOOs, and (4) is not susceptible to thermal-hydraulic instability.
The NRC staff further concludes that the licensee has adequately accounted for the effects of the proposed EPU on the hydraulic loads onthe core and RCS components.
Based on this, the NRC staff concludes that the thermal and hydraulic design will continue to meet the requirements of final GDCs--1O and draft GDC-7t2 following implementation of the proposed EPU. Therefore, the NRC staff 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 The NRC staff's review covered the functional performance of the control rod drive system (CRDS) to confirm that the system can affecteffest a safe shutdown, respond within acceptable limits during AOOs, and prevent or mitigate the consequences of postulated accidents.
The review also covered the CRDS coolingsystem to ensure that it will continue to meet its design requirements.
The NRC's acceptance criteria are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ES~s against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA;GDC-4--
inoao -i-ourc htS~ impotan to...... saot b .. do e to.. accorrmeda~te-hefet-fh nvr 
+er44ateperatioa-, maintnnoretin&sect;anpestuated-aoodentsi (2) draft G.DC-26, insofar as it requires that the protection system be designed to fail into a safe state; GD-22-,-ieofar-os-Ueruies-that-the-proteetion-systern-be-designed~o~aNl-nto-a-safe-state,-(3) d raft G DC-3 1, insofar as it requires that the reactivity control systems be capable of sustaining any single malfunction without causing a reactivity transient which could result in exceeding acceptable fuel damage limits ;GlC-, -Thefeae-as qufres-kt-hae-pr-oteotion-system-be-for-an y-singjmaune of-4he-reaetity o.on-toksystemsj (4) draft GDCs-27 and 28 insofar as they require that at least two independent reactivity control systems be provided, with both systems capable of making and holding the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits; (5) draft GDCs-29 and 30, insofar as they require that at least one of the reactivity control systems be capable of making and holding the core subcritical under any condition sufficiently fast to prevent exceeding acceptable fuel damage limits; GDGC2,ieofa....i -reouirocsthat-twa-iFdependent-reaotivt y-so14 "" .. ,,q, Gfermal-powee0-4angese
(,5) GDG27,-nsofar-as~~ure~thathe~eeativty-ent-lsy sterns-te-desigd 4e-ae rer4ty-eharnjes-under-pstu s- ppop iate-marg-n efr-stucWo toassure-he-aapabiitye-to-oe4he-eo~e~s-mait(nd6) draft GDC-32, insofar as it requires that limits, which include considerable margin, be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a)rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling; C C2,insf -ast-iequifesthat~h-eaoftviy-sent~o4-systeme-be-desg ed-ta ass 4aheefe~sef-postulat edreaetivyaeiet-an-netthereastlt--dama~e-to-4he-reaeto*-ve~ssel-inte~oals-soe .s-.te-signif4oantly--mpaia~thecapabi~it-y.to-ceotbe-oer-e;-{-7..- Fneofa t t-- --ad aetv 55 
* * *1. .cxtrome= high eetofA~ and (87) 10 CFR 50.62(c)(3), insofar as it requires that all BWRs have an alternate rod injection (ARI)system diverse from the reactor trip system, and that the ARI system have redundant scram air header exhaust valves. Specific review criteria are contained in SRP Section 4.6.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the functional design of the CR0S. The NRC staff concludes that the licensee has adequately accounted for the effects of the proposed EPU on the system and demonstrated that the system's ability to affecteffeot a safe shutdown, respond within acceptable limits, and prevent or mitigate the consequences of postulated accidents will be maintained following the implementation of the proposed EPU. The NRC staff further concludes that the licensee has demonstrated that sufficient cooling exists to ensure the system's design bases will continue to be followed upon implementation of the proposed EPU. Based on this, the NRC staff concludes that the fuel system and associated analyses will continue to meet the requirements of draft GDCs-26, 27, 28, 29, 30, 31, 32, 40 and 42, 25, 26,27,-2, nd2, and 10 CFR 50.62(c)(3) following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the functional design of the ORDS.2.8.4.20Overpressure Protection During Power Operation Regqulatory Evaluation Overpressure protection for the RCPB during power operation is provided by relief and safety valves and the reactor protection system. The NRC staff's review covered relief and safety valves on the main steamlines and piping from these valves to the suppression pool. The NRC's acceptance criteria are based on (1) draft GDC-9, insofar as it requires that the RCPB be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime; in.....ofar,,., as. it reqfuires-that-the-RCSadaooite-wlia%
t-ooa nd-pet eetien-ysem b dsine wit h-suffi~eiet-marg~nto -assur that th-einsniin-f~e-RPB ae nt-excedddrn n eend4Ie ........maloperatio, incldn A~;and (2) draft GDCs-33, 34, and 35G10,C-3t, insofar as itthey requires that the RCPB be designed with sufficient margin to assure that it behaves in a nonbrittle manner and that the probability of rapidly propagating fracture is minimized.
Specific review criteria are contained in SRP Section 5.2.2.Technical Evaluation Elnsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the overpressure protection capability of the plant during power operation.
The NRC staff 56 concludes that the licensee has (1) adequately accounted fo the effects of the proposed EPU on pressurization events and overpressure protection features and (2) demonstrated that the plant will continue to have sufficient pressure relief capacity to ensure that pressure limits are not exceeded.
Based on this, the NRC staff concludes that the overpressure protection features will continue to meet draft GDCs-9, 33, 34, and 35 ODes 15 and 31-following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to overpressure protection during power operation.
2.8.4.3 Reactor Core Isolation Cooling System Regqulatory Evaluation The reactor core isolation cooling (RCIC) system serves as a standby source of cooling water to provide a limited decay heat removal capability whenever the main feedwater 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 water supply for the RCIC system comes from the condensate storage tank, with a secondary supply from the suppression pool. The NRC staffs review covered the effect of the proposed EPU on the functional capability of the system. The NRC's acceptance criteria are based on (1) draft GDC-40 insofar as it requires that protection be provided for ESFs against dynamic effects; GD,4,-4nsefar--a4Ptejukree-..p.. tant to.... cafot et protet ... e .....dyneieef;e (2) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is not impaired by the shari ng;GQ&-5Jneefar-as-it-equiret-aSSGs-impe~ae+t~e-safety-nobbe-
..hared amn n uc!oar-power unt.nl...anb oontao ththrn ............
ability4o-peffomtsafety-funetion,-{3-)
GDG-2Q,-ineefar-ast-r~equJires~hat-the-preteet on-ed-a reastivty 
.-ent high-preobabi~ty-ofceomplishing-their safety* functin in... ovont of ,",,.. (.1) CCC 3, i nsoa,. i rgio ht ytmt provide- eaetoeleeant-mak -e r-pmhteetionagaiast -sma4-beaks-n n-he-RCPB-epovdeso the4uedeig~nimits-.a~re~et-axoceeded+/--5}GD-35nef ar-as4 t~u -a-ros~d uoaheat-uaN.eat *from-tlhe-rfea,-tor-eer-e-Fat-a-tesueh-that-SAF-D~sefld4he-desiwGn~onldltensef~he44GP-Ba4<e-net-(63) draft GDCs-51 and 57, insofar as they require that piping systems penetrating containment be designed with appropriate features as necessary to protect from an accidental rupture outside containment and the capability to periodically test the operability of the isolation valves to determine if valve leakage is within acceptable l im its; GD5-4I4esofar-as~trequi~ee-ha~t-pipig-systems-penetkatin&sect; ontalnmnenat--etlesigle with.... the capailit to teriodecy te...t th operabilit of the. isebatien alvc t detorminc if valve leakg swti ..cp..~ limts;and 
(.74) 10 CFR 50.63, insofar as it requires that the plant withstand and recover from an SBO of a specified duration.
Specific review criteria are contained in SRP Section 5.4.6.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]-
Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the ability of the RCIC system to provide decay heat removal following an isolation of main feedwater event and a station blackout event and the ability of the system to provide makeup to the core following a small break in the RCPB. The NRC staff concludes that the licensee has adequately accounted for the effects of the proposed EPU on these events and demonstrated 57 that the RCIC system will cotnu heat removal and makeup for these events following implementation of the proposed EPU. Based on this, the NRC staff concludes that the RCIC system will continue to meet the requirements of draft GDCs-4, 40, 51, and 57, .........
29r3+3 rd--4, and 10 CER 50.63 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the ROIC system.2.8.4.4 Residual Heat Removal System 'Reaulatorv Evaluation The RHR system is used to cool down the ROS following shutdown.
The RHR system is typically a low pressure system which takes over the shutdown cooling function when the RCS temperature is reduced. The NRC staff's review covered 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. The NRC's acceptance criteria are based on (1) draft GDCs-40 and 42, insofar as they require that protection be provided for ESFs against dynamic effects; and (2)draft GDC-4, insofar as reactor facilities shall not share systems or corn ponents unless it is shown safety is not impaired by the sharing. SOC 5, incofa as i..... r.. cqu.. ,i..........t-SS e-,
Specific review criteria are contained in SRP Section 5.4.7 and other guidance provided in Matrix 8 of RS-001.Technical Evaluation Elnsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the RHR system. The NRC staff concludes that the licensee has adequately accounted for the effects of the proposed EPU on the system and demonstrated that the RHR system will maintain its ability to cool the RCS following shutdown and provide decay heat removal. Based on this, the NRC staff concludes that the RHR system will continue to meet the requirements of draft GDCs-4, 40 and 42 SO~t 1., 5, an 31, following implementation of the proposed EPU.Therefore, the NRC staff finds the proposed EPU acceptable with respect to the RHR system.2.8.4.5 Standby Liquid Control System Reciulatory Evaluation The standby liquid control system (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 NRC staff's review covered the effect of the proposed EPU on the functional capability of the system to deliver the required amount of boron solution into the reactor. The NRC's acceptance criteria are based on (1) draft GDCs-27 and 28 insofar as they require that at least two independent reactivity control systems be provided, with both systems capable of making and holding the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits; (2) draft GDCs-29 and 30, insofar as they require that at least one of the reactivity 58 control systems be capable of makn an hlngi&#xb6;6re subcritical under any condition sufficiently fast to prevent exceeding acceptable fuel damage limits;GD-6 inees res4ha~wo4ndependeat-ieao~it f dffoen4t dsign--apb~e-ofaJl-hodifl4hl e4eeetersubedtieal-th cold condit~on; (2) GDC 27v, insofar as it requir. that, he r..c..it control...
cctmc. have,. a..reattivtty-.eheeges-undepetuated-aceiden-edtinsj,-an4-(3) 10 CFR 5O.62(c)(4), insofar as it requires that the SLOS be capable of reliably injecting a borated water solution into the reactor pressure vessel at a boron concentration, boron enrichment, and flow rate that provides aset level of reactivty control, and [DEPENDING...ON..............
PER.MIT DAT O ORIG4NAL-,QESI 
..........-
syeteminitiate-auitematiealiy.
Specific review criteria are contained in SRP Section 9.3.5 and other guidance provided in Matrix 8 of RS-001.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the SLCS and concludes that the licensee has adequately accounted for the effects of the proposed EPU on the system and demonstrated that the system will continue to provide the function of reactivity control independent of the control rod system following implementation of the proposed EPU. Based on this, the NRC staff concludes that the SLCS will continue to meet the requirements of draft GDCs-27, 28, 29 and 30, GDCs 26 and 27, and 10 CFR 50.62(c)(4) following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the SLCS.2.8.5 Accident and Transient Analyses 2.8.5.1.1 Decrease in Feedwater Temperature, Increase in Feedwater Flow, Increase in Steam Flow, and Inadvertent Opening of a Main Steam Relief or Safety Valve Regqulatory 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.
The NRC staff's review covered (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 reactor protection system, (6) operator actions, and (7) the results of the transient analyses.
The NRC's acceptance criteria are based on (1) draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; w~ith-mar-gkvsuffioieet-oeneswreh akthedtesign6eonditieon-of.-heR&P-8are-t-exoeeded-during 
-any-condit4oef-nermjaorationj-(32) draft GDCs-14 and 15, insofar as they require that 59 the core protection system be de prevent or suppress conditions that could result in exceeding acceptable fuel damage limits and that protection systems be provided for sensing accident situations and initiating the operation of necessary ESFs; GDG-2O-0-iseofar-as4t-eqir~es-t~hat-the-eaotor-protection-nermekep.ea tlen 7 (43) draft GDC-29 insofar as they require that a reactivity control system be provided capable of preventtng exceeding acceptable fuel damage limits. soC 26, insofar as it requires that a reactivty control be.. provided-,='''
and'norwal-eper-atien~inek~ng-AOQCSAFOS aeno e~edec-.Specific review criteria are contained in SRP Section 15.1.1-4 and other guidance provided in Matrix 8of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the excess heat removal events described above and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the SAFOLs and the RCPB pressure limits will not be exceeded as a result of these events. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft GDCs-6, 14, 15, and 29 GD , 1 ,152, an 6following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the events stated.2.8.5.2 Decrease in Heet Removal by the Secondary System 2.8.5.2.1 Loss of External Load; Turbine Trip; Loss of Condenser Vacuum;Closure of Main Steam isolation Valve; and Steam Pressure Regulator 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.
The NRC staffs review covered 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 NRC's acceptance criteria are based on (1) draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acc:eptable fuel damage limits; SDC 10, insofar as it that.,= the.,,
draft GDC-29 insofar bs it requires that a reactivity control system be provided capable of making the core subcritical under any conditions (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. GO2~60 mcfar a.. it r..uir. t, h.at a reactivtw otrlcsemb r'ddrand be capable of relia;bly'....... re-n~et-exeeede4,-.Specific review criteria are contained in SRP Section 15.2.1-5 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the decrease in heat removal events described above and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the SAFDLs and the ROPE pressure limits will not be exceeded as a result of these events. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft ,s-GDCs-6 and 29 implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the events stated.2.8.5.2.2 Loss of Nonemergency AC Power to the Station Auxiliaries Regqulatory Evaluation The loss of nonemergency 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.
The NRC staff's review covered (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 NRC's acceptance criteria are based on (1) draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; G&GC Ins......as-tr-a th C edsge-ih app ..........................
AFDL ar ot-excedd uin......pra~es AO',c-. (2 GC 5 t equre that... the.... RCS....and.
it ssociated auxiliar"y ysems-be des ged-i mt4n-ar fir en en i-hat-thedesign-eenditiowfth-G4 are-net-ex-ceeded-durin -any-cond-tierff-nermefo-perationj-and-(32) draft G DC-29 insofar as it requires that a reactivity control system be provided capable of making the core subcritical undler any conditions (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. ~ ie~a-a rcasiiycnrlcetent-pevdedadt-ape e-.of-ealy-onteffin&sect;-he-ra-te-ef-veaot~vty-chnes~teeneur4h r-neeaitenbnso ,eraeperatien44daelnig AQOs, SAF DLc am-e nt-exeeeded.
Specific review criteria are contained in SRP Section 15.2.6 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the 61 
'I t~i .s4gqq&.*
I*. &I0I.IttrIItoI~
conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the loss of nonemergency ac power to station auxiliaries event and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has 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 this event. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft GDCs-6 and !0, 15,an~d-2&following.
implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the loss of nonemergency ac power to station auxiliaries event.2.8.5.2.3 Loss of Normal Feedwater Flow Recqulatory Evaluation A loss of normal feedwater flow could occur from pump failures, valve malfunctions, or a LOOP.Loss of feedwater flow results in an increase in reactor coolant temperature and pressure which eventually requires a reactor trip to prevent fuel damage. Decay heat must be transferred from fuel following a loss of normal feedwater flow. Reactor protection and safety systems are actuated to provide this function and mitigate other aspects of the transient.
The NRC staffs review covered (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 NRC's acceptance criteria are based on (1) draft GDC-6, insofar as It requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; SOC 10 nc;,ofar ac. it r.qu.ro that the, R, S be designed,, ,with aiek4ateg-maO gir;-(2) sOC 5, n-oara-sA it r-suire- that th, ee RSan- t associat-eduxlir nyte (3,2) draft GDC-29 insofar as it requires that a reactivity control system be provided capable of making the core subcritical under any conditions (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits.
syste-.-.o.de,.ndb.cpaleofreial cn.........the rate....of rativity ..h..nge to-eesufe-hat-u~fdeF-eefldjie3o 
!ng,^AO, , SAFO~ r Specific review criteria are contained in SRP Section 15.2.7 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the loss of normal feedwater flow event and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the SAFDLs and the RCPB pressure limits will not be 62 exceeded as a result of the loss of normal ,wY;1 ?as? onthsteNRsaf concludes that the plant will continue to nmeet the requirements d~f draft GDCs-6 and 2943.lG-404an-6following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the loss of normal feedwater flow event.2.8.6.3 Decrease in Reactor Coolant System Flow 2.8.5.3.1 Loss of Forced Reactor Coolant Flow Regulatory Evaluation A decrease in reactor coolant flaw 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.
The NRC staff's review covered (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 reactor protection system, (6) operator actions, and (7) the results of the transient analyses.
The NRC's acceptance criteria are based on (1)draft GDC-6, insofar as it requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits; G-9)C-,4.0 5 inse fart-asrequres~hathe RCSb eined a~p epfiate-mrin-te-ernsure-hat-AFDS ar o ocoedd u r4ng~nmal-oper-atioas-4noludg-AQOsi(}GG&nasfaF-as 
;t........u...ee"-
that4he-C=GS-andqte-asscated-auwliareystemse-bedesigeditma -sufee-en epeiatienjd(3,2) draft GDC-29 insofar as it requires that a reactivity control system be provided capable of making the core subcritical under any conditions (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage himits,.G-&isofar at it requirc that....r..ct...ty control ....tom" bo p"vdd an'ecpbe of feliab~y-cont'otling*-the*r-ate-ef-reactivty-c.haIn&sect;es-to-en surethatm.
e-ueeenditions-of nerm at-operation, including,,, AC, SAF^ero "... not .exeeded Specific review criteria are contained in SRP Section 15.3.1- 2 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the decrease in reactor coolant flow event and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models.The NRC staff further concludes that the licensee has 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 this event. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of draft GDCs-6, and 29 GD4Cs 10, 15, anid 26 following.
implementation~of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the decrease in reactor coolant flow event.2.8.5.3.2 Reactor Recirculation Pump Rotor Seizure and Reactor Recirculation Pump Shaft Break 63 Reciulatorv 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 ,floQw while the reactor is at power results in a degradation of core heat transfer which could result in fuiel 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 either case, reactor protection and safety systems are actuated to mitigate the transient.
The NRC staff's review covered (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 reactor protection system, (6) operator actions, and (7) the results of the transient analyses.
The NRC's acceptance criteria are based on (1) final 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 ECOS, 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) draft GDC-32, insofar as it requires that limits, which include considerable margin, be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling; SO;;C ,28, isoar-asit-....qukes hat &#xf7;he, eetesseebeindto sueta h effects of postlt'd reacti.it accident...
can..neithe~-resu~t-iame4e-te4Ae-RG:P-B-re~etere4har4tedkeea4eldner-dstet4he-aore-4tis-eeIte-eee~and (3) draft GDCs-33, 34, and 35, insofar as they require 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 rapidly propagating fractures is minimized.
prepagatm-faoture4s-mniiMz.e4-.Specific review criteria are contained in SRP Section 15.3.3-4 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain whY the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the sudden decrease in core coolant flow events and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated 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 nonbrittle manner, the probability of propagating fracture of the RCPB is minimized, and adequate core cooling will be provided.
Based on this, the NRC staff concludes that the plant will continue to meet the requirements of final GDC-27 and draft GDCs-32, 33, 34, and 35 GOCs 27, 28, and 31 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the sudden decrease in core coolant flow 64 events.2.8.5.4 Reactivity and Power Distribution Anomalies 2.8.5.4.1 Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or 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.
The NRC staffs review covered (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) th6 results of the transient analyses.
The NRC's acceptance criteria are based on (1) final GDC-1O, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations, including AQOs; (2) draft GDCs-14 and 15, insofar as they require that the core protection systems be designed to act automatically to prevent or suppress conditions that could result in exceeding acceptable fuel damage limits and that protection systems be provided for sensing accident situations and initiating the operation of necessary ESFs; dsgae ensar~e-thePA~
s-ere-e-ex~eeeda~sisuW-4fAOOsi-and (3) draft GDC-31, insofar as it requires that the reactivity control systems be capable of sustaining any single malfunction without causing a reactivity transient which could result in exceeding acceptable fuel damage limits. 000 25, incofar aa~-iqueha-te-teotien-system-e Specific review criteria are contained in SRP Section 15.4.1I and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the uncontrolled control rod assembly withdrawal from a subcritical or low power startup condition and concludes that the licensee's analyses have adequately accounted for the changes in core design necessary for operation of the plant at the proposed power level. The NRC staff also concludes that the licensee's analyses were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure the SAFDLs are not exceeded.
Based on this, the NRC staff concludes that the plant will continue to meet the requirements of final GDC-1 0 and draft GDCs-14, 15, and  44, 23, an4.25SfOllowing implementation of the proposed EPU. Therefore, the NRC staff 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 Uncontrolled Control Rod Assembly Withdrawal at Power Repq ulatorv Evaluation 65 
,rJ ,, ir "' ,.An uncontrolled control rod asm *pwraybe caused by a malfunction of the reactor control or rod control systemsyj his wjtl~drawpt.wi l.uncontrollably add positive reactivity to the reactor core, resulting in a power excursion.
The NRC staff's review covered (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 ana y~sis, (4) the analytical methods and computer codes used, and (5) the results of the associated analyses.
The NRC's acceptance criteria are based on (1) final 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; draft GDCs-1 4 and 15, insofar as they require that the core protection systems be designed to act automatically to prevent or suppress conditions that could result in exceeding acceptable fuel damage limits and that protection systems be provided for sensing accident situations and initiating the operation of necessary ESFs; GDO-20O eactinitys control sytestoensr taf Qsi;and (3) draft GDC-31, insofar as it requires that the reactivity control systems be capable of sustaining any single malfunction without causing a reactivity transient which could result in exceeding acceptable fuel damage limits. G.D6-2i-so far-as-itequires~hat-t ef-the pecific review criteria are contained in SRP Section 15.4.2 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff,as~
documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the uncontrolled control rod assembly withdrawal at power event and concludes that the licensee's analyses have adequately accounted for the changes in core design required for operation of the plant at the proposed power level. The NRC staff also concludes that the I~cerisee's analyses were performed using acceptable analytical models. The NRC staff furthet Cbn(ludles that the licensee has demonstrated that the reactor protection and safety systd6ts will continue to ensure the SAFDLs are not exceeded.
Based on this, the NRC staff concludes that the plant will continue to meet the requirements of final GDC-1 0 and draft G DCs-1 4, 15, and 31 GD~s-14-0-ad&
following implementation of the proposed EPU. Therefore, the N RC staff finds the proposed EPU acceptable with respect to the uncontrolled control rod assembly withdrawal at power.2.8.5.4.3 Startup of a Recirculation Loop at an lndodr~ect Temperature and Flow Controller Malfunction Causing an n *Core Flow Rate Regqulatory Evaluation A startup of an inactive loop transient may result in either an increased core flow or the introductior-i of cooler water into the core. This event causes an increase in core reactivity due to decreased moderator temperature and core void fraction.
The NRC staff's review covered (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 NRC's acceptance criteria are based on (1) final GDC-1 0, insofar as it requires that the RCS be designed with appropriate 66 margin to assure that SAFDLs are not exceeded &iring any condition of normal operation, including the effects of ACOs; (2) draft GDCs-14 and 15, insofar as they require that the core protection systems be designed to act automatically to prevent or suppress conditions that could result in exceeding acceptable fuel damage limits and that protection systems be provided for sensing accident situations and initiating the operation of necessary ESFs;GDC20 isofar as it requires that-the-protectionsse m ed indt nitate-autoeaieaflsU-the-per-ati~-apg~ogiate-systems4oe~mure.4hat-SA5 DL-s-rnoqt-exee-as-a-result-ofoertoa ,ocrrences; (3) SOC insoa asi-eur- htte C n t*soitd uiir cycem"b deig -with-magia--uff4icint to ensure that the design-pRCBaentceeded-during AOOsj (43) draft GDC-32, insofar as it requires that limits, which include considerable margin, be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the care, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling; G desined-to sueta teefcsofpsuae ratvy-aeiensn-ehe-sutt4n damage4o4he 7.-its-s uppe~t-structufes temals&-se-eao,-t-saifnieaflIym~pk~
e-eabitt-t-eete GGeej-;afl-(54) draft GDC-29, insofar as it requires that at least one of the reactivity control systems be capable of making the core subcritical under any condition sufficiently fast to prevent exceeding acceptable fuel damage limits. CDC---2&,4resofaras-Pqivres~hate-a reaotiv-ty-cento be provideandtae-apabe-ofreiabIy-eenetrne4ea~teo-rae .i changes to-ensrt~e, .that-.udr-cenditieons ef-exceeded--Specific review criteria are contained in SRP Section 1 5.4.4-5 and other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the increase in core flow event and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the SAFOLs and the RCPB pressure limits will not be exceeded as a result of this event. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of final GDC-1O and draft GDCs-14, 16, 29, and 32 GC 1, 1, 2, 2, an 8following implementation of the proposed EPU. Therefore, the NRC staff finds the p~roposed EPU acceptable with respect to the increase in core flow event.2.8.5.4.4 Spectrum of Rod Drop Accidents Recqulatorv Evaluation The NRC staff evaluated the consequences of a control rod drop accident in the area of reactor physics. The NRC staff's review covered 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 NRC's acteptance criteria are based on draft GDC-32, insofar as it requires that limits, 67 
-.".'4 .-- '.9"t~I.which include considerablemag': reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary~or (b) disrupt the core,,its support structures, or other vessel internals sufficiently to impair the effectiveness of'emergency core cooling. GG 28, inso.r...
it.... iro that......the 
... reatiity Co...........tems .b .dei e t.....urc.that.th......ct ot-postuat ed-eae~vity-aseIdens- -neithf-rersuU..........e-t...........grtr tha "im~SRP Section 15.4.9 and other guidance provided in Matrix 8 of RS-O01.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the rod drop accident and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated 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, the NRC staff concludes that the plant will continue to meet the requirements of draft GDC-32GQG-24&
following implementation of the EPU. Therefore, the NRC staff 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 Regqulatory 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 ftiel 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. The NRC staff's review covered (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 NRC's acceptance criteria are based on (1) final GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFOLs are not exceeded during normal operations, including AOOs; (2-)-AGOsj;-and (32) draft GDCs-29, insofar as it requires that at least one of the reactivity control systems be capable of making the core subcritical under any condition sufficiently fast to prevent exceeding acceptable fuel damage limits. Gr.C_ 26,' as. ,SAF-DL-s-arenot-e~eeede4.
Specific review criteria are contained in SRP Section 15.5.1-2 and 68 other guidance provided in Matrix 8 of RS-OO1.Technical Evaluation
[lnaert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses of the inadvertent operation of EGOS or malfunction that increases reactor coolant inventory and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has 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 this event. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of final GDC-10, and draft implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the inadvertent operation of EGGS 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 Pressure Relief Valve R eculatory 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 turbine control valves (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 feedwater control system using water from the condensate storage tank via the condenser hotwell. The NRC staffs review covered (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 NRC's acceptance criteria are based on (1) final GDC-1 0, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFIJLs are not exceeded during normal operations, including AOOs; 7
iifes-that-the-RCS-and-its-as~seated-,ufLfieie -te-ensure-that~h-deig end.ins-f-Uhe-RCP ar=no ceeedd-uring-AO~sj-and (32) draft 000-29 insofar as it requires that a reactivity control system be provided capable of making the corn subcritical under any conditions (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits.-ODC 26, noofr., %A-eqi that
* r *ciiyente-yt
.....iod *" be capabl of, -reli..bly c...r... ,&sect;-4he-rtc of recivt chneoesure-that-unde .. onditi..n..of..normal op.rati.n,.inc.....n.
AC............
arc not ..........
Specific review criteria are contained in SRP Section 15.6.1 and other guidance provided in Matrix 8 of RS-O01.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the 69 proposed changes satisfy each of tlreqiurementsitl regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion 
.", The NRC staff has reviewed the licensee's analyses of the inadvertent opening of a pressure relief valve event and concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection and safety systems will continue to ensure that the SAFOLs and the RCPB pressure limits will not be exceeded as a result of this event. Based on this, the NRC staff concludes that the plant will continue to meet the requirements of final GDC-1 0 and draft G DC-29 cI 3'-26-following implementation of the proposed EPU. Therefpre, the NRC staff finds the proposed EPU acceptable with respect to the inadvertent opening of a pressure relief valve event.2.8.5.6.2 Emergency Core Cooling System and Loss-of-Coolant Accidents Reaqulatorv 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.
The NRC staff's review covered (1) the licensee's 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 peak cladding temperature, 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 EGGS systems; and (7) operator actions. The NRC's acceptance criteria are based on (1) 10 CFR &sect; 50.46, insofar as it establishes standards for the calculation of ECCS performance and acceptance criteria for that calculated performance; (2) 10 CFR Part 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) draft GDCs-40 and 42, insofar as they require that protection be provided for ES~s against the dynamic effects that might result from plant equipment failures, as well as the effects of a LOCA;GDC27,insofar-as it-requires that-Sthe reativtacntrolesysems be .... detged- asto h~av~e-e a ombfe-Sapab~lity 7 4-iw-e~junetknWt-~
sen-add~if by teEC, f elia~y-entf-ellng-reastivit 
..... ge u..der. postulte accident...v condtion, with oppro a'emf&sect;1,,,  to assure the capailit to, coo the. core is.. maintain.....d; and (54) final 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.
Specific review criteria are contained in SRP Sections 6.3 and 15.6.5 and other guidance provided in Matrix 8 of RS-001.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evajuation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]70 S Conclusion The NRC staff has reviewed the licensee's analyses of the LOCA events and the ECOS. The NRC staff concludes that the licensee's analyses have adequately accounted for operation of the plant at the proposed power level and that the analyses were performed using acceptable analytical models. The NRC staff further concludes that the licensee has demonstrated that the reactor protection system and the ECCS will continue to ensure that the peak cladding temperature, 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, the NRC staff concludes that the plant will continue to meet the requirements of final GDC-35 and draft GDCs-40 and 42, GDCc '1, 27, 3.7-,and 10 CFR 50.46 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the LOCA.2.8.5.7 Anticipated Transients Without Scrams Regulatory Evaluation ATWS is defined as an AOO followed by the failure of the reactor portion of the protection system specified in draft GDCs-14 and 15.GDG=-20 The regulation at 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 standby liquid control system (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 system initiation must be automatic.
* each BWR have equipment to trip the reactor coolant recirculation pumps automatically under conditions indicative of an ATWS.The NRC staff's review was conducted to ensure that (1) the above requirements are met, (2) sufficient margin is available in the setpoint for the SLCS pump discharge relief valve such that SLOS operability is not affected by the proposed EPU, and (3) operator actions specified in the plant's Emergency Operating Procedures are consistent with the generic emergency procedure guidelines/severe accident guidelines (EPGs/SAGs), insofar as they apply to the plant design. In addition, the NRC staff reviewed the licensee's ATWS analysis to ensure 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 &deg;F; (3) the peak suppression pool temperature is less than the design limit; and (4) the peak containment pressure is less than the containment design pressure.
The NRC staff also evaluated the potential for thermal-hydraulic instability in conjunction with ATWS events using the methods and criteria approved by the NRC staff. For this analysis, the NRC staff reviewed 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.
lnse4-the-the ..........
uetfication of.. th ,. applicbilit of generi .......
e for,,.,";-- propoe EU]Review guidance is provided in Matrix 8 of RS-001.71 V -Technical Evaluation u[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the information submitted by the licensee related to ATWS and concludes that the licensee has adequately accounted for the effects of the proposed EPU on ATWS. The NRC staff concludes that the licensee has demonstrated that ARI, SLCS, and recirculation pump trip 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, the NRC staff finds the proposed EPU acceptable with respect to ATWS.2.8.6 Fuel Storage 2.8.6.1 New Fuel Storage Regqulatory 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. The NRC staffs review covered 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 NRC's acceptance criteria are based on draft GDC-6G9-6, insofar as it requires the prevention of criticality in fuel storage systems by physical systems or processes, preferably utilizing geometrically safe configurations.
Specific review criteria are contained in SRP Section 9.1 .1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effect of the new fuel on the analyses for the new fuel storage facilities and concludes that the new fuel storage facilities will continue to meet the requirements of draft GDC-66 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to the new fuel storage.2.8.6.2 Spent Fuel Storage Regqulatory Evaluation Nuclear reactor plants include storage facilities for the wet storage of spent fuel assemblies.
The safety function of the spent fuel pool and storage racks is to maintain the spent fuel assemblies in a safe and subcritical array during all credible storage conditions and to provide a safe means 72 of loading the assemblies into shipping casss t rvecordthefctfte proposed EPU on the criticality analysis (e.g., reactivity/of fW6l spen fulstorage array and boraflex degradation or neutron poison efficacy).
The NRC's acceptance criteria are based on (1) draft GDC-40GD0*-4, insofar as it requires that protection be provided for engineered safety features against the dynamic effects and missiles that might result from plant equipment failu resinsofar-as itrequies thatSS&#xa3;Ce 
,,,pe,'anttsafewtey4eine~o ao it4-thG-eflv4r~nmeflta 
;ni sscae wih ora operation, mainte. nance t....ting, an otltdaedne n 2 raft GDC-66GC6, 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.
Specific review criteria are contained in SRP Section 9.1.2.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's analyses related to the effects of the proposed EPU on the spent fuel storage capability and concludes that the licensee has adequately accounted for the effects of the proposed EPU on the spent fuel rack temperature and criticality analyses.The NRC staff also concludes that the spent fuel pool design will continue to ensure an acceptably low temperature and an acceptable degree of subcriticality following implementation of the proposed EPU. Based on this, the NRC staff concludes that the spent fuel storage facilities will continue to meet the requirements of draft GDCs-40 and 66GDCs ', and 62 following implementation of the proposed EPU. Therefore, the NRC staff finds the proposed EPU acceptable with respect to spent fuel storage.2.9 Source Terms and Radiologqical Consequences Analyses 2.9.1 Source Terms for Radwaste Systems Analyses Regqulatory Evaluation The NRC staff reviewed the radioactive source term associated with EPUs to ensure the adequacy of the sources of radioactivity used by the licensee as input to calculations to verify that the radioactive waste management systems have adequate capacity for the treatment of radioactive liquid and gaseous wastes. The NRC staff's review included the parameters used to determine (1) the concentration of each radionuclide in the reactor coolant, (2) the fraction of fission product activity released to the reactor coolant, (3) concentrations of all radionuclides other than fission products in the reactor coolant, (4) leakage rates and associated fluid activity of all potentially radioactive water and steam systems, and (5) potential sources of radioactive materials in effluents that are not considered in the plant's Updated Final Safety Analysis Report related to liquid waste management systems and gaseous waste management systems. The NRC's acceptance criteria for source terms are based on (1) 10 CFR Part 20, insofar as it establishes requirements for radioactivity in liquid and gaseous effluents released to unrestricted areas; (2) 10 CFR Part 50, Appendix I, insofar as it establishes numerical guides for design objectives and limiting conditions for operation to meet the "as low as is reasonably 73 achievable" criterion; and (3), draft requires that the plant design include means to control the release of radioactive effluents.
Specific review criteria are contained in SRP Section 11.1.Technical Evaluation 
: Elnsert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the .requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the radioactive source term associated with the proposed EPU and concludes that the proposed parameters and resultant composition and quantity of radionuclides are appropriate for the evaluation of the radioactive waste management systems.The NRC staff further concludes that the proposed radioactive source term meets the requirements of 10 CFR Part 20, 10 CFR Part 50, Appendix I, and draft GDC-70GD0;-69.
Therefore, the NRC staff finds the proposed EPU acceptable with respect to source terms.NOTE Us Scttes-% ,2-n below- t4he Iieensee'e-ra~ietegieal-eeesequenses-n-ae-atemative-sour-ee4erm.
2.9.2 Radiological Cosquneso Cnro,,,,l Accide,.;,nt Uin Alternative Source Term Regulatory Evaluation The NRC staff reviewed the DAnlssoth radiological consequences ofnaly control roddropoia drpaccident (CRDA). The NRCn stafreviewincued anrexaiati(LBn ofe (1) sthepant's respwonseac totaccident, (2lssinldd()
the rees ffseinrouecefeetsfo th d core todtels enviro mentvions the vatlh oueisn f pr amee-nususdbbh iensee fAB-a4~w~or hecacuation-zeof ethertotnrand infethve cotole romq uovleto (ThEDrEle) sfotean. The NRC's acceptance criteria for thiloilcn equne sadrsfrradiological consequences of a ponsrtureadro accidentabae ond (1 ) fO 1.inaofaC-1s9,t isfrsitrequires-that-adequateeradiationnprotectionbb provided to permit access andocuayoft-ocuanyoetecntrol room under accident conditions without personnel receiving rdainepsrsi raeditonan(2 10osre CFR P-arts 100 rinsofar, as iteftbished rqinrements, for asrn theuatio ofa thec oe fo osuae accidents ilb cetby12. Specific review criteria-are cnandi R eto 50I Tcntined invaPluations6 n S..,adohrudnepoie nMti fR oQ74 
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2) provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has evaluated the licensee's revised accident analyses performed in support of the proposed EPU and concludes that the licensee has adequately accounted for the effects of the proposed EPU. The NRC staff further concludes that the plant site and the dose mitigating ESFs remain acceptable with respect to the radiological consequences of postulated DBAs since, as set forth above, the calculated total effective dose equivalent (TEDE) at the exclusion area boundary (EAB), at the low population zone (LPZ) outer boundary, and in the control room meet the exposure guideline values specified in 10 CFR 50.67 and final GDC-19, as well as applicable acceptance criteria denoted in SRP Section 15.0.1. Therefore, the NRC staff finds the licensee's proposed EPU acceptable with respect to the radiological consequences of DBAs.oeseue cof a control rod ep ace dont-and-oenoludes-tatathe4ioonsee-has-adequately-aeoou44o-fr-the-effeets~f~he-piopesed P4on.thse-ealay~sesr-.-qhe-NRGg-stafffurte-e,,poeu reguideflne-vele n1 CE10.11. The-G-staffalsoe-oelude s-tbat-4he-oee~o-room-meets~he-oese-requ~iremenef-Q-OC-*1Oef-iAs,-..*T-hee 2.0. adiolongicsal Cons-equenesteofehe-Faimale-ef-armt-L-iessue-Garoun&sect;y-{-=-
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consequences of an.... MSLB-ousd otim kt-ase-ba n....4).GDC.4-inoferae-t-qure tha adqut raito pfoetectien4e-hpfvd-o permit-eeees.aF-end-o, upaney-oft4he-ob.40r-nee-aeedent-,ondtoswtot p" o t insofarsite~shabnges requirements forthe assuringt that h-redolgia dose,:esaa fro potuete a-olearlink4o~he-conetusiens-Feaohe 4-by-the44RC-stff;-as-documeetedkhe-on~
si ConclMusion ed-th~ie4nsee~srvisedidn-aesei~r-th-radioloioa cosqene f-an MSLB outside cotimn an cnl., ht~eI;e ...... eqetI~acone o h fet ftepoosdEUe h nlc h R tf fudh-;nl~mnb'--l c with respect toa equdi4u~,m-eonc.entation4or-contie full.. pow...r o per-ation 7-T-PeNRC-etffaseoeoudeat--
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nt t-a~a~lagee- .-scemblies.Sc cidents-may-77 licaeneefrthe caclto oferadilogca dura ;(2tioaeqa of the-aESee; providedI fosr-ere-ad ...........
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htbd, orit ouiv ae4en4-an thettcontain radicoctivity be dosigned-Ath-a~ppeprdte ytes ,nd (3) 10 CFR-Part 10OGasefar-asit-eetablhshes-requir-ements-fe-assu,4n&sect;Ahat-SRP-See~tner6A a-4-a 4;- 1 d-et, a er~&sect;uidaneefervdedAMa#xe-R-Q a ci-Ngar lin ..to.th conclusionsmchdb the NRC' ac douotdi teenl......ue.ce of a.. spn fue coc"k drop acci~denad, onluestat nne-lieesee-has.-
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*24782.10 Health Physics 24784-t2.10.1 Occupational and Public Radiation Doses Regulatory Evaluation The NRC staff conducted its review in this area to ascertain what overall effects the proposed EPU will have on both occupational and public radiation doses and to determine that the licensee has taken the necessary steps to ensure that any dose increases will be maintained as low as is reasonably achievable.
The NRC staffs review included an evaluation of any increases in radiation sources and how this may affect plant area dose rates, plant radiation zones, and plant area accessibility.
The NRC staff evaluated how personnel doses needed to access plant vital areas following an accident are affected.
The NRC staff considered the effects of the proposed EPU on nitrogen-16 levels in the plant and any effects this increase may have on radiation doses outside the plant and at the site boundary from skyshine.
The NRC staff also considered the effects of the proposed EPU on plant effluent levels and any effect this increase may have on radiation doses at the site boundary.
The NRC's acceptance criteria for occupational and public radiation doses are based on 10 CFR Part 20 and final GDC-I 9. Specific review criteria are contained in SRP Sections 12.2, 12.3, 12.4, and 12.5, and other guidance provided in Matrix 10 of RS-001.Technical Evaluation 79 
[Insert technical evaluation.
The n Qa e ialaton (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion' The NRC staff has reviewed the licensee's assessment of the effects of the proposed EPU on radiation source terms and plant radiation levels. The NRC staff concludes that the licensee has taken the necessary steps to ensure that any increases in radiation doses will be maintained as low as reasonably achievable.
The NRC staff further concludes that the proposed EPU meets the requirements of 10 CFR Part 20 and final GDC-1 9. Therefore, the NRC staff finds the licensee's proposed EPU acceptable with respect to radiation protection and ensuring that occupational radiation exposures will be maintained as low as reasonably achievable.
......onekeview-Aeas-(-Health-P-hysies)j t.lnse4t-Reg 2492.11 Human Performance 2.4942.11.1 Human Factors Reaulatorv Evaluation The area of human factors deals with programs, procedures, training, and plant design features related to operator performance during normal and accident conditions.
The NRC staff's human factors evaluation was conducted to ensure that operator performance is not adversely affected as a result of system changes made to implemented the proposed EPU. The NRC staff's review covered changes to operator actions, human-system interfaces, and procedures and training needed for the proposed EPU. The NRC's acceptance criteria for human factors are based on final GDC-19, 10 CFR 50.120, 10 CFR Part 55, and the guidance in GL 82-33.Specific review criteria are contained in SRP Sections 13.2.1, 13.2.2, 13.5.2.1, and 18.0.Technical Evaluation The NRC staff has developed a standard set of questions far the review of the human factors area. The licensee has addressed these questions in its application.
Following are the NRC staff's questions, the licensee's responses, and the NRC staff's evaluation of the responses.
: 1. Chanqes in Emergqency and Abnormal Operating Procedures Describe how the proposed EPU will change the plant emergency and abnormal operating procedures. (SRP Section 13.5.2*.1)
[Insert licensee's response followed by NRC staff statement on why the response is acceptable]
: 2. Changes to Operator Actions Sensitive to Power Uprate Describe any new operator actions needed as a result of the proposed EPU. Describe changes to any current operator actions related to emergency or abnormal operating procedures that will occur as a result of the proposed EPU. (SRP Section 18.0)(i.e., Identify and describe operator actions that will involve additional response time or 80 will have reduced time available.
Your response should address any operator workarounds that might affect these response times. Identify any operator actions that are being automated or being changed from automatic to manual as a result of the power uprate. Provide justification for the acceptability of these changes).[Insert licensee's response followed by NRC staff statement on why the response is acceptable]
: 3. Chances to Control Room Controls.
Displays and Alarms Describe any changes the proposed EPU will have on the operator interfaces for control room controls, displays, and alarms. For example, what zone markings (e.g. normal, marginal and out-of-tolerance ranges) on meters will change? What setpoints will change? How will the operators know of the change? Describe any controls, displays, alarms that will be upgraded from analog to digital instruments as a result of the proposed EPU and how operators will be tested to determine they bould use the instruments reliably. (SRP Section 18.0)[Insert licensee's response followed by NRC staff statement on why the response is acceptable]
: 4. Chancqes on the Safety Parameter Displav System Describe any changes to the safety parameter display system resulting from the proposed EPU. How will the operators know of the changes? (SRP Section 18.0)[Insert licensee's response followed by NRC staff statement on why the response is acceptable]
: 5. Changes to the Operator Training Proaram and the Control Room Simulator Describe any changes to the operator training program and the plant referenced control room simulator resulting from the proposed EPU, and provide the implementation schedule for making the changes. (SRP Sections 13.2.1 and 13.2.2)[Insert licensee's response followed by NRC staff statement on why the response is acceptable]
Conclusion The NRC staff has reviewed the changes to operator actions, human-system interfaces, proceduies, and training required for the proposed EPU and concludes that the licensee has (1) appropriately accounted for the effects of the proposed EPU on the available time for operator actions and (2) taken appropriate actions to ensure that operator performance is not adversely affected by the proposed EPU. The NRC staff further concludes that the licensee will continue to meet the requirements of final GDC-1 9, 10 CFR 50.120, and 10 CER Part 55 following implementation of the proposed EPU. Therefore, the NRC staff finds the licensee's proposed EPU acceptable with respect to the human factors aspects of the required system changes.[Aditina 'cio Aro4umnP fp, sert-Regulater-y-E-valuatine,,heioab~ahatien ,andtonetsionseetenaa~e  2,802.1 2 Power Ascension and Testingq Plan*2:8O42.12.1 Approach to EPU Power Level and Test Plan Reglulatory Evaluation 81 The purpose of the EPU test program 1ibdf ifrlf'VSCs will perform satisfactorily in service at the proposed EPU power level. The test program also provides additional assurance that the plant will continue to operate with design criteria at EPU conditions.
The NRC staff's review included an evaluation of: (1) plans for the initial approach to the proposed maximum licensed thermal power level, including verification of adequate plant performance, (2)transient testing necessary to demonstrate that plant equipment will perform satisfactorily at the proposed increased maximum licensed thermal power level, and (3) the test program's conformance with applicable regulations.
The NRC's acceptance criteria for the proposed EPU test program are based on 10 CFR Part 50, Appendix B, Criterion XI, which requires establishment of a test program to demonstrate that SSCs will perform satisfactorily in service.Specific review criteria are contained in SRP Section 14.2.1.Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The staff has reviewed the EPU test program, including plans for the initial approach to the proposed maximum licensed thermal power level, transient testing necessary to demonstrate that plant equipment will perform satisfactorily at the proposed increased maximum licensed thermal power level, and the test program's conformance with applicable regulations.
The staff concludes that the proposed EPU test program provides adequate assurance that the plant will operate in accordance with design criteria and that SSCs affected by the proposed EPU, or modified to support the proposed EPU, will perform satisfactorily in service. Further, the staff finds that there is reasonable assurance that the EPU testing program satisfies the requirements of 10 CFR Part 50, Appendix B, Criterion XI. Therefore, the NRC staff finds the proposed EPU test program acceptable.-f [#dditionai Review Aroas (PowereAsccs" adTcngPa)[-nser4-Regulatery-~valuat'+er1-,-Teo4eIe-E-valuaticn, and Cono "o cztaca' ncsay 2-41-2.13 Risk Evaluation Risk Evaluation of EPU Recsulatory Evaluation The licensee conducted a risk evaluation to (1) demonstrate that the risks associated with the proposed EPU are acceptable and (2) determine if "special circumstances" are created by the proposed EPU. As described in Appendix D of SRP Chapter 19, special circumstances are present if any issue would potentially rebut the presumption of adequate protection provided by the licensee to meet the deterministic requirements and regulations.
The NRC staff's review covered the impact of the proposed EPU on core damage frequency (ODE) and large early release frequency (LERF) for the plant due to changes in the risks associated with internal events, external events, and shutdown operations.
In addition, the NRC staff's review covered the quality of the risk analyses used by the licensee to support the application for the proposed EPU. This included a review of the licensee's actions to address issues or weaknesses that may have been raised in previous NRC staff reviews of the licensee's individual plant examinations (IPEs) and individual plant examinations of external events (IPEEE), or by an industry peer review. The NRC's risk acceptability guidelines are contained in 82 RG 1.174. Specific review guidance is contained in Matrix 13 of RS-OO1 and its attachments.
Technical Evaluation
[Insert technical evaluation.
The technical evaluation should (1) clearly explain why the proposed changes satisfy each of the requirements in the regulatory evaluation and (2)provide a clear link to the conclusions reached by the NRC staff, as documented in the conclusion section.]Conclusion The NRC staff has reviewed the licensee's assessment of the risk implications associated with the implementation of the proposed EPU and concludes that the licensee has adequately modeled and/or addressed the potential impacts associated with the implementation of the proposed EPU. The NRC staff further concludes that the results of the licensee's risk analysis indicate that the risks associated with the proposed EPU are acceptable and do not create the"special circumstances" described in Appendix D of SRP Chapter 19. Therefore, the NRC staff finds the risk implications of the proposed EPUI acceptable.
............
Additie nal-Review.Ar-eae-(Risk-E-valuatier)}
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3.0 FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATION CHANGES To achieve the EPU, the licensee proposed the following changes to the Facility Operating License and TSs for [Plant Name].EProvide a list of license and TSs changes (including license conditions) and an NRC staff evaluation of each.]4.0 REGULATORY COMMITMENTS Insert the following sentence if the licensee has not made any regulatory commitments in support of the EPU.The licensee has made no regulatory commitments in its application for the EPU., Insert the following if the licensee has made regulatory commitments in support of the EPU.The licensee has made the following regulatory commitment(s):
[Provide a summary of each regulatory commitment made by the licensee.]
The NRC staff finds that reasonable controls for the implementation and for subsequent evaluation of proposed changes pertaining to the above regulatory commitment(s) are best provided by the licensee's administrative processes, including its commitment management program. The above regulatory commitments do not warrant the creation of regulatory requirements (items requiring prior NRC approval of subseq uent changes).6.0 RECOMMENDED AREAS FOR INSPECTION" As described above, the NRC staff has conducted an extensive review of the licensee's plans and analyses related to the proposed EPU and concluded that they are acceptable.
The NRC staff's review has identified the following areas for consideration by the NRC inspection staff 83 during the licensee's implementation of thl propos 5 areas are recommended based on past experience with EPUs, the extent and unique nature of modifications necessary to implement the proposed EPU, and new conditions of operation necessary for the proposed EPU.They do not constitute inspection requirements, but are intended to give inspectors insight into important bases for approving the EPU.'"[Provide list of recommended areas for inspection.]

==6.0 STATE CONSULTATION==

In accordance with the Commission's regulations, the EName of State] State official was notified of the proposed issuance of the amendment.
The State official had [no] comments.[If comments were received, address them here.]7.0 ENVIRONMENTAL CONSIDERATION Pursuant to 10 CFR 51.21, 51.32, 51.33, and 51.35, a draft Environmental Assessment and finding of no significant impact was prepared and published in the Federal Register on[Date] ( FR ). The draft Environmental Assessment provided a 30-day opportunity for public comment. if no comments were received, use the following sentence: 
[No comments were received on the draft Environmental Assessment.]
If comments were received, use the following sentence: 
[The NRC staff received comments which were addressed in the final environmental assessment.]
The final Environmental Assessment was published in the Federal Register on [Date] ( FR ). Accordingly, based upon the environmental assessment, the Commission has determined that the issuance of this amendment will not have a significant effect on the quality of the human environment.

==8.0 CONCLUSION==

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.9.0 REFRNE 1. RS-001, Revision 0, "Review Standard for Extended Power Uprates," December 2003.2. [Insert additional references as necessary]

==Attachment:==

List of Acronyms Principal Contributors:
Date: 84 aI LIST OF ACRONYMS AAC alternate ac sources ac alternating current ALARA as low as reasonably ach!evable ARAVS auxiliary and radwaste area ventilation system ARI alternate rod insertion ASME American Society of Mechanical Engineers ATWS anticipated transient without scram B&PV boiler and pressure vessel BL bulletin BOP balance-of-plant BTP branch technical position BWR boiling-water reactor BWR VIP Boiling Water Reactor Vessel and Internals Project COF core damage frequency CFR Code of Federal Regulations CFS condensate and feedwater system CRAVS control room area ventilation system CRDA control rod drop accident GRDM control rod drive mechanism CRDS "control rod drive system CUF cumulative usage factor CWS circulating water system OBA design-basis accident DELOCA design-basis loss-of-coolant accident do direct current DG draft guide EAB exclusion area boundary ECCS emergency core cooling system EFDS equipment and floor drainage system EPG emergency procedure guideline EPRI Electric Power Research Institute EPU extended power uprate EO environmental qualification ESF engineered safety feature ESFAS engineered safety feature actuation system ESFVS engineered safety feature ventilation system FAC flow-accelerated corrosion g5 FHA fuel handling accident.,'
FPP fire protectior/program 
"'GDO -general design criterion (or criteria)GL generic letter l&C instrumentation and controls IN information notice IPE ,individual plant examination IPEEE individual plant examination of external events LERF large early release frequency LLHS light load handling system LOCA loss-of-coolant accident LOOP loss of offsite power LPZ low population zone MC main condenser MCES imain condenser evacuation system MlOV motor-operated valve MSIV main steam isolation valve MSIVLCS main steam isolation valve leakage control system MSLB main steamline break MSSS main steam supply system MWt megawatts thermal NEI Nuclear Energy Institute NPSH net positive suction head NRC Nluclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NSSS nuclear steam supply system O&M operations and maintenance P-T pressure-temperature PWSCC3 primary water stress-corrosion cracking RCIC reactor core isolation cooling RCPB reactor coolant pressure boundary RCS reactor coolant system RG regujatory guide RHR residual heat removal RS review standard RWCS reactor water cleanup system SAFDL specified acceptable fuel design limit 86 

.-&#xb8;. ..., SAG severe accident guideline SAR Safety Analysis Report SBO station blackout SFP spent fuel pool SFPAVS spent fuel pool area ventilation system SGTS standby gas treatment system SLCS standby liquid control system'SRP Standard Review Plan SSCs structures, systems, and components SSE safe-shutdown earthquake SWMS solid waste management system SWS service water system TAVS turbine area ventilation system TBS turbine bypass system TCV _ turbine control valve TEDE total effective dose equivalent TS technical specifi cation UHS ultimate heat sink 87}}