ML17212A042

Final Safety Analysis Report, Rev. 35, Chapter 1, Introduction and Summary

ML17212A042
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Site:	Millstone
Issue date:	06/29/2017
From:	Dominion Nuclear Connecticut
To:	Office of Nuclear Reactor Regulation
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Text

Millstone Power Station Unit 2 Safety Analysis Report Chapter 1

Table of Contents tion Title Page INTRODUCTION ............................................................................................... 1.1-1

SUMMARY

DESCRIPTION.............................................................................. 1.2-1 1 General........................................................................................................ 1.2-1 2 Site ............................................................................................................. 1.2-1 3 Arrangement .............................................................................................. 1.2-2 4 Reactor ........................................................................................................ 1.2-2 5 Reactor Coolant System.............................................................................. 1.2-3 6 Containment System ................................................................................... 1.2-4 7 Engineered Safety Features Systems .......................................................... 1.2-4 8 Protection, Control and Monitoring Instrumentation ................................. 1.2-7 9 Electrical Systems....................................................................................... 1.2-7 10 Auxiliary Systems....................................................................................... 1.2-8 10.1 Chemical and Volume Control System ...................................................... 1.2-8 10.2 Shutdown Cooling System.......................................................................... 1.2-9 10.3 Reactor Building Closed Cooling Water System ....................................... 1.2-9 10.4 Fuel Handling and Storage ....................................................................... 1.2-11 10.5 Sampling System ...................................................................................... 1.2-11 10.6 Cooling Water Systems ............................................................................ 1.2-11 10.7 Ventilation Systems .................................................................................. 1.2-12 10.8 Fire Protection System.............................................................................. 1.2-13 10.9 Compressed Air Systems .......................................................................... 1.2-13 11 Steam and Power Conversion System ...................................................... 1.2-14 12 Radioactive Waste Processing System ..................................................... 1.2-14 13 Interrelation With Millstone Units 1 and 3 ............................................... 1.2-15 14 Summary of Codes and Standards ............................................................ 1.2-17 COMPARISON WITH OTHER PLANTS ......................................................... 1.3-1 PRINCIPAL ARCHITECTURAL AND ENGINEERING CRITERIA FOR DESIGN............................................................................................................... 1.4-1 1 Plant Design ................................................................................................ 1.4-1 2 Reactor ........................................................................................................ 1.4-1 3 Reactor Coolant and Auxiliary Systems ..................................................... 1.4-2 3.1 Reactor Coolant System.............................................................................. 1.4-2 3.2 Chemical and Volume Control System ...................................................... 1.4-4 3.3 Shutdown Cooling System.......................................................................... 1.4-5 4 Containment System ................................................................................... 1.4-5 1-i Rev. 35

tion Title Page 5 Engineered Safety Features Systems .......................................................... 1.4-6 6 Protection, Control and Instrumentation System ........................................ 1.4-6 7 Electrical Systems....................................................................................... 1.4-7 8 Radioactive Waste Processing System ....................................................... 1.4-7 9 Radiation Protection ................................................................................... 1.4-7 10 Fuel Handling and Storage ......................................................................... 1.4-7 RESEARCH AND DEVELOPMENT REQUIREMENTS ................................ 1.5-1 1 General........................................................................................................ 1.5-1 2 Fuel Assembly Flow Mixing Tests ............................................................. 1.5-1 3 Control Element Assembly Drop Tests ...................................................... 1.5-2 4 Control Element Drive Assembly Performance Tests ................................ 1.5-2 5 Fuel Assembly Flow Tests.......................................................................... 1.5-3 6 Reactor Vessel Flow Tests.......................................................................... 1.5-4 7 In-core Instrumentation Tests ..................................................................... 1.5-4 8 Materials Irradiation Surveillance .............................................................. 1.5-5 9 References................................................................................................... 1.5-5 IDENTIFICATION OF CONTRACTORS ......................................................... 1.6-1 1 References................................................................................................... 1.6-1 GENERAL DESIGN CHANGES SINCE ISSUANCE OF PRELIMINARY SAFETY ANALYSIS REPORT ......................................................................... 1.7-1 1 General........................................................................................................ 1.7-1 2 Control Element Drive Mechanisms........................................................... 1.7-1 3 Radioactive Waste Processing System ....................................................... 1.7-1 3.1 Clean Liquid Waste Processing System ..................................................... 1.7-1 3.2 Gaseous Waste Processing System............................................................. 1.7-1 4 Vital Component Closed Cooling Water System ....................................... 1.7-2 5 Electrical ..................................................................................................... 1.7-2 5.1 AC Power.................................................................................................... 1.7-2 5.2 Diesel Generators........................................................................................ 1.7-2 5.3 DC Supply................................................................................................... 1.7-2 5.4 Instrument Power ........................................................................................ 1.7-3 6 Axial Xenon Oscillation Protection ............................................................ 1.7-3 7 Number of Control Element Assemblies and Drive Mechanisms .............. 1.7-3 8 Burnable Poison Shims ............................................................................... 1.7-3 9 Structures .................................................................................................... 1.7-3 1-ii Rev. 35

tion Title Page 10 High Pressure Safety Injection Pumps........................................................ 1.7-4 11 Containment Purge Valve Isolation Actuation System .............................. 1.7-4 12 Control Element Drive System ................................................................... 1.7-4 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS SPECIAL INTEREST ITEMS [THIS SECTION PROVIDES HISTORICAL INFORMATION PROVIDED TO THE ACRS AT THE TIME OF INITIAL LICENSING AND WAS NOT INTENDED TO BE UPDATED.].............................................................. 1.8-1 1 General........................................................................................................ 1.8-1 1.1 Ability of Fuel to Withstand Transients at End of Life and Experimental Verification of Maximum Linear Heat Generation Rate ............................ 1.8-1 1.2 Fuel Integrity Following a Loss-of-Coolant Accident................................ 1.8-1 1.3 Primary System Quality Assurance and In-Service Inspectability ............. 1.8-2 1.4 Separation of Control and Protective Instrumentation ............................... 1.8-3 1.5 Instrumentation for Detection of Failed Fuel ............................................. 1.8-3 1.6 Effects of Blowdown Forces on Core and Primary System Components .. 1.8-4 1.7 Reactor Vessel Thermal Shock................................................................... 1.8-4 1.8 Effect of Fuel Rod Failure on the Capability of the Safety Injection System .....

1.8-5 1.9 Preoperational Vibration Monitoring Program........................................... 1.8-5 1.9.1 Basis of Program......................................................................................... 1.8-5 1.9.2 Millstone Unit 2 Program ........................................................................... 1.8-6 2 Special for Millstone Unit 2........................................................................ 1.8-7 2.1 Release of Radioactivity in Case of Damaged Fuel Assemblies in Spent Fuel Pool ............................................................................................................. 1.8-7 2.2 Hydrogen Control ....................................................................................... 1.8-7 2.3 Common Mode Failures and Anticipated Transients Without Scram ........ 1.8-7 3 References................................................................................................... 1.8-8 TOPICAL REPORTS .......................................................................................... 1.9-1 MATERIAL INCORPORATED BY REFERENCE ........................................ 1.10-1 AEC GENERAL DESIGN CRITERIA FOR NUCLEAR POWER PLANTS .. 1.A-1 1-iii Rev. 35

List of Tables mber Title 1 Licensing History 1 Summary of Codes and Standards for Components of Water-Cooled Nuclear Power Units (1) 1 Comparison with Other Plants 1 Seismic Class I Systems and Components 1 Comparison of Preoperational Vibration Monitoring Program Design Parameters 1 Topical Reports 1-iv Rev. 35

List of Figures mber Title 1 Site Layout

-2 Plot Plan 3 General Arrangement, Turbine Building Plan at Operating Floor Elevation 54 Feet 6 Inches 4 General Arrangement, Turbine Building Plan at Mezzanine Floor Elevation 31 Feet 6 Inches 5 General Arrangement, Turbine Building Plan at Ground Floor Elevation 14 Feet 6 Inches

-6 General Arrangement Containment Plan at Floor Elevation 14 feet 6 inches and Elevation 36 feet 6 inches 7 General Arrangement Auxiliary Building Plan at Elevation 36 feet 6 inches and Elevation 38 feet 6 inches

-8 General Arrangement Auxiliary Building Sections G-G and H-H

-9 General Arrangement Auxiliary Building Ground Floor Elevation 14 feet 6 inches and Cable Vault Elevation 25 feet 6 inches

-10 General Arrangement Containment and Auxiliary Building Plan at Elevation (-)5 feet 0 inches and Elevation (-)3 feet 6 inches 11 General Arrangement Containment and Auxiliary Building Plan at Elevation (-)25 feet 6 inches and Elevation (-)22 feet 6 inches 12 General Arrangement Containment and Auxiliary Building Plan at Elevation (-)45 feet 6 inches 13 General Arrangement Containment and Auxiliary Building Section A-A 14 General Arrangement Containment and Auxiliary Building Section B-B 15 General Arrangement Turbine Building Sections C-C and E-E 16 General Arrangement Turbine Building Sections D-D and F-F 17 General Arrangement Intake Structure Auxiliary Steam Boiler Room Plan and Section 1-v Rev. 35

INTRODUCTION s Final Safety Analysis Report (FSAR) was initially submitted in support of the application of Connecticut Light and Power Company (CL&P), The Hartford Electric Light Company LCO), Western Massachusetts Electric Company (WMECO), and Northeast Nuclear Energy mpany (NNECO), for a license to operate the second nuclear powered generating unit at the of the Millstone Power Station. Since the initial licensing of the unit, unless otherwise cated, the FSAR has been updated a number of times to reflect current design and analysis rmation. On the basis of the information presented in the FSAR and referenced material at the e of application for operating license, the applicants concluded that Millstone Unit 2 is gned and constructed and will be operated without undue risk to the health and safety of the lic.

struction of Millstone Unit 2 was authorized by the United States Atomic Energy mmission (AEC) when it issued Provisional Construction Permit CPPR-76 on December 11,

0. Commercial operation of Millstone Unit 2 commenced in December 1975 at a gross trical output of 865 megawatts.

lstone Unit 2 is located Millstone Point in the Town of Waterford, Connecticut. It is located ediately to the north of the first unit (Millstone Unit 1) and south of the third unit (Millstone t 3). Commercial operation of Millstone Unit 1 was authorized by the AEC by issuing visional Operating License DPR-21 on October 7, 1970. Commercial operation of Millstone t 1 commenced in December, 1970. Commercial operation of Millstone Unit 3 was authorized he United States Nuclear Regulatory Commission (NRC) (formerly the AEC) by issuing the Power License on November 25, 1985, and the Full Power License on January 31, 1986.

mmercial operation of Millstone Unit 3 commenced in April 1986. A licensing history for the lstone Unit 2 plant is presented in Table 1.1-1.

lstone Unit 2 utilizes a pressurized water nuclear steam supply system (NSSS). The unit is ilar, in this respect, to the former Yankee Atomic Electric Company generating plant in Rowe, ssachusetts, (NRC Docket Number 50-29), the former Haddam Neck Plant operated by the necticut Yankee Atomic Power Company on the Connecticut River at Haddam, Connecticut C Docket Number 50-213), and the Maine Yankee Atomic Power Company plant at casset, Maine (NRC Docket Number 50-309). The NSSS for Millstone Unit 2 is supplied by mbustion Engineering, Inc. (CE) which also supplied the steam supply system for the Maine kee plant. The Millstone Unit 2 NSSS is similar to the systems supplied by CE for the initial units of the Baltimore Gas and Electric Calvert Cliffs Nuclear Power Plant (NRC Docket mbers. 50-317 and 50-318).

lstone Unit 2 has been designed to operate safely under all normal operating conditions and cipated transients. Although the unit produces small amounts of radioactive waste, the offsite osal of these wastes is rigidly controlled and maintained below established limits.

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C is an indirect wholly-owned subsidiary of Dominion Energy, which is in turn owned by minion Resources, Inc. (DRI). Virginia Power, which is the licensed owner and operator of the th Anna and Surry nuclear stations, is also a subsidiary of DRI.

transmission and distribution assets on the site will continue to be owned by Connecticut ht and Power (CL&P) and will be operated under an Interconnection Agreement between

&P and DNC.

FSAR will retain references to Northeast Utilities and Northeast Nuclear Energy Company uments/activities when they are used in a historic context and are required to support the plant nsing bases.

n license transfer, all records and design documents necessary for operation, maintenance, decommissioning were transferred to DNC. Some of these drawings are included (or renced) in this FSAR. These drawings often have title blocks (or drawing numbers) which list theast Nuclear Energy Company (et. al) or Northeast Utilities Service Company (et. al). In eral, no changes to these title blocks will be made at this time. Based on this general note, e drawings shall be read as if the title blocks list Dominion Nuclear Connecticut, Inc.

lstone Unit 2 has been designed to operate reliably without accident. Nevertheless, to ensure no reasonably credible accident could result in dangerous releases of radioactive material, the incorporates a number of features designed to minimize the effects of such an accident. The quacy of these safety features under the conditions of various postulated accidents is discussed hapter 14.

initial license to operate Millstone Unit 2 was at a full power core thermal output of 2560 awatts. This corresponded to a NSSS thermal rating, which includes core power and other tor coolant heat sources such as reactor coolant pumps and pressurizer heaters, of 2570 MWt.

lstone Unit 2 is currently licensed for a steady state reactor core power level of 2700 MWt, esponding to a NSSS rating of 2715 MWt. All Chapter 14 analyses have been evaluated on basis of these current values.

ce the construction permit was issued, and during the design and construction of the unit, there e been no major deviations from the information supplied in the Preliminary Safety Analysis ort (PSAR). However, changes in various specific design features have been found desirable these are covered in the appropriate sections of this report. A summary of the more significant gn changes incorporated in the plant since the issuance of the PSAR up to the time of lication for an operating license is provided in Section 1.7.

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TABLE 1.1-1 LICENSING HISTORY EVENT DATE nstruction Permit Issued December 11, 1970 al Safety Analysis Report Filed August 15, 1972 ll Term Operating Licensing Issued September 26, 1975 ll Power License September 26, 1975 tial Criticality October 17, 1975 0% Power March 20, 1976 mmercial Operation December 26, 1975 retch Power June 25, 1979 erating License Extension Requested December 22, 1986 erating License Extension Issued January 12, 1988 ll Term Operating License Expires December 11, 2010 erating License Expires July 31, 2035 1.1-3 Rev. 35

1 GENERAL ummary description of Millstone Unit 2 of the Millstone Nuclear Power Station is provided in section. The description includes the following:

a. Site

b. Arrangement

c. Reactor

d. Reactor coolant system

e. Containment system

f. Engineered safety features systems

g. Protection, control and instrumentation system

h. Electrical systems

i. Auxiliary systems

j. Steam and power conversion system

k. Radioactive waste processing system

l. Interrelation with Millstone Units 1 and 3

m. Summary of Codes and Standards ithheld under 10 CFR 2.390 (d) (1) 2.2 1.2-1 Rev. 35

ithheld under 10 CFR 2.390 (d) (1) 2.3 containment houses the NSSS, consisting of the reactor, steam generators, reactor coolant ps, pressurizer, and some of the reactor auxiliaries. The containment is equipped with a polar e.

enclosure building completely envelopes the containment and provides a filtration region ween the containment and the environment.

turbine building houses the turbine generator, condenser, feedwater heaters, condensate and water pumps, turbine auxiliaries and certain of the switchgear assemblies.

ithheld under 10 CFR 2.390 (d) (1) 4 REACTOR reactor is a pressurized light water cooled and moderated type fueled by slightly enriched nium dioxide. The uranium dioxide is in the form of pellets and is contained in pressurized aloy-4 tubes fitted with welded end caps. These rods are arranged into fuel assemblies each sisting of 176 fuel rods arranged on a 14 rod square matrix. Space is left in the fuel rod array to w for the installation of five guide tubes. These guide tubes provide for the smooth motion of trol element assembly fingers. The assembly is fitted with end fittings and spacer grids to ntain fuel rod alignment and to provide structural support. The end fittings are also drilled h flow holes to provide for the flow of cooling water past the fuel tubes.

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ier elements had used stainless steel as the absorber material. Five absorber elements are nected together by a spider yoke in a square matrix with a center element. The five elements stitute a control element assembly (CEA). The 73 CEAs are connected, either singly or dually, ugh extension shafts, to 61 magnetic jack type control element drive mechanisms (CEDMs) ch are mounted on nozzles on the reactor vessel head. Each CEA is aligned with and can be rted into the guide tubes of fuel assemblies. The dual CEAs are utilized for shutdown rods.

single CEAs are divided into regulating groups. The eight part length control rods of Cycle were replaced by dummy flow plugs. Two of the flow plugs were replaced by reactor vessel l indication system detectors, then in Cycle Twelve, the last six remaining flow plugs were oved. The resulting increase in core bypass flow has been accounted for in the safety analysis.

replacement head has a total of 78 nozzle penetrations. 67 of these nozzles are suitable for porting control element drive mechanisms (61 are in use, while the other 6 nozzles are capped h nozzle adapters). Two nozzles are used for heated junction thermocouples, which enable nitoring reactor vessel between the top of the vessel dome and the area directly above the fuel dles. Eight nozzles are used for nuclear instrumentation and one nozzle is used for the reactor sel head vent. The location, size and the number of nozzles on the replacement reactor vessel ure head are maintained in the same configuration as before (prior to cycle 16).

mical shim control is provided by boric acid dissolved in the coolant water. The concentration oric acid is maintained and controlled as required by the chemical and volume control system.

reactor core rests on the core support plate assembly which is supported by the core support el. The core support barrel is a right circular cylinder supported from a machined ledge on the de surface of the vessel flange forging. The support plate assembly transmits the entire weight he core to the core support barrel through a structure made of beams and vertical columns.

rounding the core is a shroud which serves to limit the coolant which bypasses the core. An er guide structure, consisting of upper support structure, control element assembly shrouds, a alignment plate and a spacer ring, serves to support and align the upper ends of the fuel mblies, prevents lifting of the fuel assemblies in the event of a loss-of-coolant accident CA) and maintains spacing of the CEAs. Chapter 3 contains more detailed information on the tor.

5 REACTOR COOLANT SYSTEM reactor coolant system consists of two closed heat transfer loops in parallel with the reactor sel. Each loop contains one steam generator and two pumps to circulate coolant. An trically heated pressurizer is connected to one loop hot leg. The coolant system is designed to rate at a thermal power level of 2715 MWt to produce steam at a nominal pressure of 880 psia.

reactor vessel, loop piping, pressurizer and steam generator plenums are fabricated of low y steel, clad internally with austenitic stainless steel. The pressurizer surge line and coolant ps are fabricated from stainless steel and the steam generator tubes are fabricated from onel.

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quench tank where the steam discharge is condensed.

two steam generators are vertical shell and U-tube steam generators each of which produces x 106 lb/hr of steam. Steam is generated in the shell side of the steam generator and flows ard through moisture separators. Steam outlet moisture content is less than 0.2 percent.

reactor coolant is circulated by four electric motor-driven, single-suction, centrifugal pumps.

h pump motor is equipped with a non reverse mechanism to prevent reverse rotation of any p that is not being used during operation with less than four pumps energized. Chapter 4 tains more detailed information on the reactor coolant system.

6 CONTAINMENT SYSTEM ouble containment system is used for Unit 2. The containment system consists of a prestressed crete cylindrical structure referred to as the containment, which is completely enclosed by the losure building (EB). The enclosure building filtration region (EBFR) includes the region ween the containment and the enclosure building, the penetration rooms and engineered safety ure equipment rooms. In the unlikely event of a LOCA the EBFR is maintained at a slightly ative pressure by the enclosure building filtration system (EBFS). Air in the EBFR would be cessed through charcoal filters and released through the 375 foot Millstone stack during a CA.

containment uses a prestressed post-tensioned concrete design. The containment is a vertical t cylindrical structure with a dome and a flat base. The interior is lined with carbon steel plate urther ensure leak tightness.

de the containment, the reactor and other NSSS components are shielded with concrete.

ess to portions of the containment during power operation is permissible.

containment, in conjunction with the engineered safety features, is designed to withstand the hest internal pressure and coincident temperature resulting from the main steam line break dent (Section 14.8.2). The structural design conditions are for an internal pressure of 54 psig a coincident equilibrium temperature of 289°F.

enclosure building is a limited leakage steel framed structure partially supported off the tainment and auxiliary building with uninsulated metal siding and an insulated metal roof k.

7 ENGINEERED SAFETY FEATURES SYSTEMS engineered safety features systems (ESFS) provide protection for the public and plant onnel against the incidental release of radioactive products from the reactor system, icularly as a result of postulated LOCA. These safety features localize, control, mitigate and 1.2-4 Rev. 35

engineered safety features consist of the following systems:

a. Safety injection

b. Containment spray

c. Containment air recirculation and cooling

d. Enclosure building filtration

e. Hydrogen control

f. Auxiliary feedwater automatic initiation system h of these systems is divided into two redundant independent subsystems which in turn are ered by the associated redundant independent emergency electrical subsystem (see tion 1.2.9). The first three are cooled by the associated redundant independent reactor building ed cooling water headers (see Section 1.2.10.3).

owing a postulated LOCA, borated water is injected into the reactor coolant system by either h and/or low pressure safety injection pumps and safety injection tanks. This provides cooling mit core damage and fission product release, and assures an adequate shutdown margin. The ty injection system also provides continuous long term post-accident cooling of the core by rculating borated water from the containment sump through shutdown cooling heat hangers and back to the reactor core (see Section 6.2).

r safety injection tanks are provided, each connected to one of the four reactor inlet lines. The ume of each tank is 2019 cubic feet. Each tank contains about 1100 cubic feet of borated water efueling concentration and is pressurized with nitrogen at 200 psig. In the event of a LOCA, borated water is forced into the reactor coolant system by the expansion of the nitrogen. The er from three tanks adequately cools the entire core. Borated water is injected into the same zles by two low pressure and three high pressure injection pumps taking suction from the eling water storage tank (RWST). For maximum reliability, the design capacity from the bined operation one high pressure and one low pressure pump provides adequate injection for any LOCA; in the event of a design basis accident (DBA), at least one high pressure and low pressure pump will receive power from the emergency power sources if preferred power ost and one of the emergency diesel generators is assumed to fail. When the refueling water age tank supply is nearly depleted, the high pressure pump suctions automatically transfer to containment sump and the low pressure pumps are shut down. One high pressure pump has icient capacity to cool the core adequately at the start of recirculation. During recirculation, t in the recirculating water is removed through the shutdown cooling heat exchangers via er the low pressure injection pumps or containment spray pumps.

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tainment. Test lines are provided to permit running the pumps for test purposes during plant ration.

safety injection system is designed in accordance with AEC General Design Criteria 35, 36, 37 in Appendix A to 10CFR50 and General Criteria as described in Section 6.1. An analysis he performance of the safety injection system (emergency core cooling system) following a tulated LOCA is given in Section 14.6.

o independent, full capacity systems are provided to remove heat from the containment osphere by containment sprays and/or air recirculation and cooling after the postulated CA.

a. The containment spray system supplies borated water to cool the containment atmosphere. The spray system is sized to provide adequate cooling with two containment spray pumps. The pumps take suction from the refueling water storage tank. When this supply is nearly depleted, the pump suction is transferred automatically to the containment sump (see Section 6.4).

b. The containment air recirculation and cooling system is designed to cool the containment atmosphere. The cooling coils and fans are sized to provide adequate containment cooling with three of the four units in service (see Section 6.5).

c. A combination of one containment spray pump aligned with the shutdown cooling heat exchanger and two containment air recirculation units provides adequate cooling of the containment. Each spray pump and two associated containment air recirculation units are cooled by one of two associated redundant reactor building cooling water and service water subsystems. They are powered by the associated emergency electrical subsystem.

enclosure building filtration system would collect and filter all potential containment leakage minimize environmental radioactivity levels resulting from the discharge of all sources of tainment leakage into the enclosure building filtration region in the unlikely event of a LOCA.

enclosure building filtration system would also collect and filter any radioactive releases in unlikely event of a fuel handling accident inside the containment or spent fuel pool areas (see tion 6.7).

hydrogen control system is provided to mix and monitor the concentration of hydrogen gas hin the containment. This system consists of the post-accident recirculation system for mixing containment environment and the hydrogen monitoring system for continuous monitoring of post-accident containment atmosphere. The hydrogen purge system and hydrogen mbiners which are not credited in accident analyses are provided for reducing containment rogen concentrations.

1.2-6 Rev. 35

matically actuates two motor driven auxiliary feedwater pumps (see Section 10.4.5.3), and ns the two auxiliary feedwater flow control valves via the automatic initiation control circuitry Section 7.3.2.2.h). The AFAIS is actuated upon completion of a 2-out-of-4 logic matrix ated by a low steam generator level. Upon receipt of an actuation signal both pumps are ted and the flow control valves to both steam generators are opened (see Section 7.3).

8 PROTECTION, CONTROL AND MONITORING INSTRUMENTATION ious instrumentation systems provide protection, control, and monitoring functions for the and efficient operation of Millstone Unit 2.

tection instrumentation systems function to shut down the reactor and activate safety systems ontinuously monitored key plant process parameters exceed predetermined limits. Specific ection instrumentation systems include the Reactor Protective System (RPS) and the ineered Safety Features Actuation System (ESFAS). The RPS functions to shut down or trip reactor if any two of four safety channels generate coincident trip signals. An RPS trip oves power from the reactor control rods, allowing them to drop into the reactor, and shut it

n. The ESFAS functions to actuate the engineered safety features systems described in FSAR tion 1.2.7. The exception to this is the containment purge valve isolation where one of four tainment air radiation detectors can generate a trip signal. Actuation of the ESFS occurs if any of four safety channels generate coincident trip signals.

trol instrumentation systems function to maintain plant parameters within operational limits ng both steady state and normal operating transients. Major control systems include the trol Element Drive System (CEDS), the Reactor Regulating System (RRS), Pressurizer Level ulating System (PLRS), Reactor Coolant Pressure Regulating System (RCPRS), Feed Water ulating System (FWRS), and Turbine Generator Control System (TGCS).

cations are provided to monitor normal and abnormal plant operation. Indicators are located hin the control room and throughout the plant. The indicators are used to monitor the status operation of the protective and control systems, and the status of other support systems.

or indication systems include the Control Element Assembly (CEA) Position Indication, lear Instrumentation (NI), In-Core Instrumentation (ICI), Radioactivity Monitoring System S), Integrated Computer System (ICS), Control Room Annunciation, and Post Accident nitoring Instrumentation (PAMI).

ails of the above and other protective, control, and monitoring instrumentation systems are vided in Chapter 7.

9 ELECTRICAL SYSTEMS Millstone Nuclear Power Station consists of Millstone Unit 1 which is no longer generating er, Millstone Unit 2 with a 1011-MVA, 0.90 power factor generator, and Millstone Unit 3 h a 1354.7-MVA, 0.925 power factor generator (see Chapter 8).

1.2-7 Rev. 35

kV transmission lines. The switchyard, in addition to carrying the electrical output of the ion, also provides a means of supplying power to the units from external sources. Startup er and reserve auxiliary power for Millstone Unit 2 are taken from the 345 kV switchyard ugh the reserve station service transformer. Normal station service power is taken from the erator main leads through the normal station service transformer. A second source of off site er for the engineered safety features is provided from normal station service transformer

-3SA or reserve station service transformer 15G-23SA, both associated with Millstone Unit 3 a 4160V crosstie connection. Two diesel generators provide the on site emergency power for lstone Unit 2. The 4160V crosstie from Unit 3 can also be configured (by operator action) to ply power directly from the Unit 3 Alternate AC (SBO) diesel generator to provide an rnate AC source for Unit 2 Appendix R and Station Blackout requirements.

iliary power for Millstone Unit 2 is provided at 6900, 4160, 480, and 120/208 volts. Direct ent 125 volt systems are also available for emergency power, engineered safety feature trol, and essential nuclear instrumentation, control and relaying.

preferred and on site emergency sources of electrical power are each adequate to permit mpt shutdown and maintain safe conditions under all credible circumstances. The on site rgency power source consists of two separate and redundant diesel generators. Each diesel is able of carrying all required auxiliary loads following postulated LOCA without exceeding its tinuous rating.

h of the two separate and redundant station batteries is capable of carrying essential 125 volt and 120 volt AC inverter loads associated with a postulated LOCA.

redundant channel wiring associated with these emergency electrical sources is physically arated.

10 AUXILIARY SYSTEMS 10.1 Chemical and Volume Control System chemistry of the reactor coolant is controlled by purification of a regulated letdown stream of tor coolant. Water removed from the reactor coolant system is cooled in the regenerative heat hanger. The fluid pressure is then reduced and flow is regulated by the letdown control valves.

perature is reduced further in the letdown heat exchanger. From there, the flow passes ugh a filter and a purification ion exchanger to remove corrosion and fission products. A ll fraction of the flow is diverted prior to entering the ion exchanger. This stream of coolant s through a process radiation monitor. Upon leaving the ion exchanger, the coolant flows ugh a strainer and another filter and is then sprayed into the volume control tank.

lant is returned to the reactor coolant system by the charging pumps, through the regenerative t exchanger. Prior to entering the charging pumps, the coolant boron concentration is adjusted 1.2-8 Rev. 35

volume control system automatically controls the rate at which coolant must be removed m the reactor coolant system to maintain the pressurizer level within the prescribed control d, thereby compensating for changes in volume due to coolant temperature changes. Using the ume control tank as a surge tank decreases the quantity of liquid and gaseous wastes which ld otherwise be generated.

ctor coolant system makeup water is taken from the primary water storage tank and the two centrated boric acid storage tanks. The boric acid solution is maintained at a temperature ch prevents crystallization. The makeup water is pumped through the regenerative heat hanger into the reactor coolant loop by the charging pumps.

on concentration in the reactor coolant system can be reduced by diverting the letdown flow y from the volume control tank to the radioactive waste processing system. Demineralized er is then used for makeup.

en the boron concentration in the reactor coolant system is low, the feed and bleed procedure viously described would generate excessive volumes of waste to be processed. Therefore, the mical and volume control system is equipped with a deborating ion exchanger which reduces on concentration late in cycle life. A complete description is given in Section 9.2.

10.2 Shutdown Cooling System shutdown cooling system (see Section 9.3) is used to reduce the reactor coolant temperature, controlled rate, from 300°F to a refueling temperature of approximately 130°F. It also ntains the proper reactor coolant temperature during refueling. Once entry conditions are met, shutdown cooling system can provide long term cooling capability in the event of a LOCA r the reactor coolant system has refilled (see Section 14.6.5.3).

shutdown cooling system utilizes the low pressure safety injection pumps to circulate the tor coolant through two shutdown cooling heat exchangers. It is returned to the reactor lant system through the low pressure safety injection header.

reactor building closed cooling water system (RBCCW) supplies cooling water for the tdown heat exchangers.

10.3 Reactor Building Closed Cooling Water System RBCCW system consists of two separate independent headers, each of which includes a CCW pump, a service water (seawater)-cooled RBCCW heat exchanger, interconnecting ng, valves and controls. A third RBCCW pump and a third RBCCW heat exchanger are vided as installed spares. The corrosion inhibited, demineralized water in this closed system is ulated through the RBCCW heat exchanger where it is cooled to 85°F by seawater which has aximum design inlet temperature of 80°F (see Section 9.4).

1.2-9 Rev. 35

he RBCCW system include:

Containment air recirculation and cooling unit Reactor vessel support concrete cooling coils Containment spray pump seal coolers High and low pressure safety injection pump seal coolers Shutdown cooling heat exchangers Engineered safety feature room air recirculation coils Reactor coolant pump thermal barrier and oil coolers Primary drain and quench tanks heat exchanger CEDM coolers Letdown heat exchanger Degasifier effluent cooler Degasifier vent condenser Sample coolers Spent fuel pool heat exchangers Waste gas compressor aftercoolers Steam generator blowdown quench heat exchanger h of the independent headers supply cooling water to components in the associated redundant ty related sub-systems (see Section 1.2.7). The RBCCW heat exchangers, connected to each pendent RBCCW headers, are cooled by the associated independent service water header (see tion 1.2.10.6). Components in each independent RBCCW header, the associated safety related systems, and the associated service water header are powered from the associated redundant pendent emergency electrical power subsystem (see Section 1.2.9).

ote manually operated valves allow the spare RBCCW pump and/or heat exchanger to be rated with either of the two independent headers. The RBCCW surge tank absorbs the umetric changes caused by temperature changes of the water within the RBCCW headers.

hemical addition system is provided for the RBCCW system to maintain the corrosion bitor concentration as required.

ing normal plant operation and normal shutdown, both of the independent RBCCW headers in service.

owing a postulated LOCA, each of the RBCCW headers, in conjunction with the associated ice water header and electrical subsystem, would provide the necessary cooling capacity to associated engineered safety feature subsystems.

1.2-10 Rev. 35

fuel handling systems provide for the safe handling of fuel assemblies and control element mblies and for the required assembly, disassembly, and storage of the reactor vessel head and rnals. These systems include a refueling machine located inside the containment above the eling pool, the fuel transfer carriage, the upending machines, the fuel transfer tube, a fuel dling machine over the spent fuel pool, a new fuel elevator in the spent fuel pool, a spent fuel k crane, a new fuel inspection machine in the fuel handling area of the auxiliary building, and ous devices used for handling the reactor vessel head and internals (see Section 9.8).

w fuel is stored dry in vertical racks within a storage vault near the spent fuel pool in the iliary building. Storage space is provided for approximately one-third of a core.

vault is designed to avoid criticality by spacing fuel assemblies at 20.5 inches, center to ter. The spent fuel pool, located in the auxiliary building, is constructed of reinforced concrete d with stainless steel. The spent fuel storage racks are separated into four regions, designated ions 1, 2, 3, and 4. Section 9.8.2.1 contains a detailed description of spent fuel storage design components.

ling and purification equipment is provided for the spent fuel pool water (see Section 9.5).

s equipment can also be used to clean up the refueling water during and after its use in the eling pool. Backup cooling methods are also available.

10.5 Sampling System sampling system consists of Sampling Stations 1 and 2, the Post Accident Sampling System SS), the Corrosion Monitoring Sample Station, and the Waste Gas Sample Sink. These vide the means for determining physical, chemical and radioactive conditions of process ds, waste gas and containment air. The system is supplemented by independent sampling of radioactive fluids in numerous locations within the unit, including sampling of the chlorinated er. (See Section 9.6.)

10.6 Cooling Water Systems exhaust steam from the main turbine and steam generator feedwater pump turbines is densed in the condenser, which is cooled, in turn, by circulating water flowing through the denser tubes, (see Section 9.7.1).

r circulating water pumps, with 548,800 gpm total capacity, take suction from and discharge Long Island Sound. The circulating water system is designed to maintain condenser back sure at 2 inches Hg absolute with a 60.8°F inlet circulating water temperature.

service water system (see Section 9.7.2) provides cooling water to the RBCCW, TBCCW, el engine cooling water, chilled water system heat exchangers, vital switchgear room cooling s and the circulating water pump bearings. Three vertical, centrifugal, half capacity service 1.2-11 Rev. 35

service water system consists of two redundant, independent cross-connected supply headers h isolation valves to all heat exchangers and two discharge headers for the RBCCW heat hangers. Two discharge headers exist for the emergency diesel generator cooling water; once erground these headers combine prior to entering the discharge canal. Service Water discharge m the TBCCW, chilled water system and vital switchgear room cooling heat exchangers bine into a common header prior to entering the discharge canal. Each of the supply headers upplied by one of the service water pumps. During normal operation and shutdown and owing a postulated LOCA, the two pumps connected to the two redundant supply headers are ervice. However, only one service water pump and header is required to provide cooling of the CCW and diesel following a LOCA or for unit shutdown. Remote manually operated valves w the third service water pump to be connected to either of the redundant headers.

intake structure consists of four independent bays. The intake structure is equipped with a rination system, consisting of two 1800 gallon sodium hypochlorite storage tanks and two ction systems with one supplying sodium hypochlorite to the service water system and the r to the circulating water intake.

10.7 Ventilation Systems mally the containment environment is cooled by the containment air recirculation and cooling em. Following a postulated LOCA, these units reduce the temperature and pressure of the tainment atmosphere to a safe level (see Sections 1.2.7, 6.5 and 9.9.1). The containment iliary circulation fans maintain uniform containment environmental temperature by mixing air. Normally, the environment for the control element drive mechanisms is maintain by the DM fan-coil units. A forced outside air purge system is provided to maintain a suitable ironment within the containment whenever access is desired. The exhaust of this containment purge system is monitored to assure that radioactive effluents are maintained within acceptable ts.

auxiliary building is served by separate ventilation systems in the fuel handling area, the oactive waste area and for the nonradioactive waste area. Each area is provided with a heating ventilating supply unit and separate exhaust fans. Exhausts from the potentially contaminated s are filtered through high efficiency particulate air (HEPA) filters, monitored, and discharged ugh the Unit 2 stack. Exhaust from clean areas is discharged directly to the atmosphere (see tion 9.9.6).

dling of irradiated fuel or moving a cask over the spent fuel pool does not require fuel dling area integrity or ventilation but it may be desirable to use the main exhaust or EBF ems, if available, as the exhaust discharge paths. If boundary integrity is set then these harge paths provide a monitored radiological release pathway. If boundary integrity is not red then suitable radiological monitoring is recommended per the Millstone Effluent Control gram.

1.2-12 Rev. 35

ilable, releases from the fuel handling area are monitored per the Millstone Effluent Control gram to ensure appropriate radiological effluent controls are in place.

o full capacity and redundant air conditioning systems are provided for the control room. In the nt of an accident, a bypass through either of the two full capacity and redundant control room ation systems, which contain charcoal filters, is provided to protect control room operating onnel from exposure to high radiation levels.

turbine building is equipped with supply and exhaust fans for year round ventilation.

access control area is air conditioned for year-round comfort. All other areas are provided h ventilation for cooling during summer and unit heaters for heating during the winter.

10.8 Fire Protection System fire protection systems' (see Section 9.10) function to protect personnel, structures, and ipment from fire and smoke. The fire protection systems have been designed in accordance h the applicable National Fire Protection Association (NFPA) Codes and Standards, regulatory uirements, industry standards, and approved procedures. The design of the various fire ection systems has been reviewed by American Nuclear Insurers (ANI).

fire detection and protection systems are designed such that a fire will be detected, contained,

/or extinguished. This is accomplished through the use of noncombustible construction, ipment separation, fire walls, stops and seals, fire detection systems, and automatic and ual water suppression systems. As a minimum, portable extinguishers, hose stations, and fire rants are available for all areas to control or extinguish a fire.

10.9 Compressed Air Systems instrument air system consists of one 640 scfm and two 237 scfm (each) instrument air pressors, receivers, dryers, and after-filters to provide a reliable supply of clean, oil free dry for the unit pneumatic instrumentation and valves. Station air for normal unit maintenance is vided by a separate 630 scfm station air compressor. Operating pressures for both systems ge between 80 to 120 psig depending on how the compressors are aligned and how the systems interconnected.

station air is used as a backup to the instrument air with tie-in points at the receiver inlets and de the containment. The compressed air systems for Units 3 and 2 are interconnected by ng and manually operated valves.

criptions of the compressed air systems are given in Section 9.11.

1.2-13 Rev. 35

turbine generator for Unit 2 is furnished by General Electric Company. It is an 1800 rpm em compound, four flow exhaust, indoor unit designed for saturated steam conditions.

er nominal steam conditions of 870 psia and 528°F at the turbine stop valve inlets and with ines exhausting against a condenser pressure of 2 inches Hg absolute, the gross electrical put is 935 MWe. Turbine output corresponds to a NSSS thermal power level of approximately 5 MWt.

condensate and feedwater system consists of three condensate pumps, one steam packing auster, two steam jet air ejectors, two external drain coolers, two trains each having five stages ow pressure feedwater heaters, two turbine-driven steam generator feedwater pumps, two high sure feedwater heaters as well as the associated piping, valves and instrumentation.

mally, the steam generator feedwater pump turbines are driven by extraction steam. At low s, main steam is used to drive the steam generator feedwater pump turbines.

omplete description of the steam and power conversion system is given in Section 10.

12 RADIOACTIVE WASTE PROCESSING SYSTEM radioactive waste processing system provides controlled handling and disposal of liquid, eous and solid waste from Unit 2 (see Section 11.1). Gaseous and liquid wastes discharged to environment are controlled to comply with the limits given in the Technical Specifications and blished to meet the requirements of 10 CFR Part 20 Sections 1301 and 1302 and Appendix B the as low as reasonably achievable (ALARA) requirement of 10 CFR Part 50, Appendix I.

radioactive waste processing system consists of the following parts.

a. Clean Liquid Waste Processing System The clean liquid waste processing system collects and processes reactor coolant wastes from the chemical and volume control system, primary drain tank and the closed drains system. The system is comprised of pumps, filters, degasifier, demineralizers, receiver tanks, monitor tanks and the necessary instrumentation, piping, controls and accessories.

The processed clean liquid wastes are collected in monitor tanks, sampled, and monitored prior to discharge to the circulating water system after ensuring that the predetermined limits for release are not exceeded.

b. Aerated Liquid Waste Processing System Aerated liquid wastes, consisting of radioactive liquid wastes exposed to the atmosphere, are collected in drain tanks and processed through filters, and demineralizers. The processed wastes are collected in a monitor tank, sampled, and 1.2-14 Rev. 35

c. Gaseous Waste Processing System Radioactive waste gases are collected through the waste gas header into the waste gas surge tank. These gases are drawn from the surge tank by one of two compressors and are pumped into a waste gas decay tank for storage to allow radioactive decay. After decay, the tank contents are sampled and monitored prior to discharge and released through a particulate filter, at a predetermined controlled rate, into the Millstone stack. The discharge is monitored prior to its entering the stack and while in the stack, thus ensuring that the predetermined limits for release are not exceeded. The six waste gas decay tanks which are provided allow a minimum of 60 days storage capacity prior to release.

d. Solid Waste Processing System Radioactive solid wastes are collected and placed in suitable containers for off site disposal. Spent demineralizer resins are held for radioactive decay prior to being dewatered and placed in a shielded cask for removal. Contaminated filter elements are placed in shielded drums for subsequent storage and off site disposal.

Low activity compactible solid wastes such as contaminated rags, paper, etc., are compacted at the Millstone Radwaste Reduction Facility prior to being shipped for disposal. Noncompactible solid wastes may be shipped to an off site processor for volume reduction prior to disposal.

13 INTERRELATION WITH MILLSTONE UNITS 1 AND 3 umber of the facilities of the Millstone Nuclear Power Station are common to Millstone Units

, and 3. The safe shutdown of any unit will not be impaired by the failure of the facilities and ems which are shared. A list of these facilities and systems follows:

a. Facilities Radiochemistry laboratory Radioactive and clean change facilities, including showers, lockers, clothing storage, and toilets Radiation Protection offices Instrument repair room Warehouse machine shop Millstone stack (for Unit 2 waste gas), main condenser air ejector and enclosure building filtration system discharge General offices 1.2-15 Rev. 35

Lunch room Visitors gallery 345 kV switch yard Millstone Unit 3 normal station service transformer 15G-3SA Millstone Unit 3 reserve station service transformer 15G-23SA Millstone Unit 3 SBO diesel generator system Makeup water treatment (Millstone Units 2 and 3 only)

Bulk storage chemical ton (Millstone Units 2 and 3 only)

Millstone Unit 2 Control Room (for monitoring and controlling Millstone Unit 1 systems)

b. Systems Low pressure nitrogen storage Fire protection (water supply and fire detection)

Auxiliary steam Makeup water treatment Building heating Sanitary sewers Plant water Communications Station Air (A system cross-tie between Unit 3 service air and Unit 2 station air headers is provided) rating and maintenance personnel are employed in all three units as described in Section 12.1.

h units have a double containment system with rectangular outer envelopes.

40 CFR 190 off site radiation dose limits will not be exceeded by simultaneous operation of lstone Units 1, 2, and 3.

Millstone Station Physical Security Plan has been implemented in accordance with CFR 73.55 Requirements for Physical Protection of Licensed Activities in Nuclear Power ctors Against Industrial Sabotage to prohibit unauthorized access to vital areas.

s plan includes measures to deter or prevent malicious actions that could result in the release radioactive materials into the environment through sabotage. Section 12.7 contains a cription of the Security Plan.

1.2-16 Rev. 35

ensure the integrity and operability of pressure-containing components important to safety, blished codes and standards are used in the design, fabrication and testing. Table 1.2-1 lists e codes and standards for components relied upon to prevent or mitigate the consequences of dents and malfunctions originating within the reactor coolant pressure boundary, to permit tdown of the reactor, and to maintain the reactor in a safe shutdown condition.

1.2-17 Rev. 35

MPS2 UFSAR TABLE 1.2-1

SUMMARY

OF CODES AND STANDARDS FOR COMPONENTS OF WATER-COOLED NUCLEAR POWER UNITS (1)

CODE CLASSIFICATION Component Group A Group B Group C Group D Pressure Vessels ASME Boiler and Pressure Vessel Code, ASME Boiler and Pressure Vessel Code,Section III, ASME Boiler and Pressure Vessel Code, ASME Boiler and Pressure Vessel Code,Section III, Class A, 1968 Edition, Addenda Class C (1968 Edition including Addenda through Section VIII, Division 1 Section VIII, Division 1 or Equivalent through Summer 1969 Summer 1969)

Reactor Vessel (2) Safety Injection Tanks (4) Reactor Building Closed Cooling Water Heat Service Water Strainers (3)

Exchangers (3)

Pressurizer (2) Reactor Building Closed Cooling Water Surge Vital Chilled Water System Condensers/

Tank Evaporators Steam Generators (3) Shutdown Heat Exchangers (2)

Concentrated Boric Acid Storage Tanks (2)

Refueling Water Storage Tank (4) 0-15 psig Storage Tanks API-620 with the NDT Examination Requirements in API-620 with the NDT Examination API-620 or Equivalent Condenser Table NST-1, Class 2 Requirements in Table NST-1, Class 3 Storage Tank Atmospheric Storage Applicable Storage Tank Codes such as API-650, Applicable Storage Tank Codes Such as API- API-650, AWWAD100 or ANSI B 96.1 Tanks AWWAD100 or ANSI B96.1 With the NDT 650 AWWAD100 or ANSI B 96.1 with the or Equivalent Diesel Oil Supply Tanks Examination Requirements in Table NST-1, Class 2 NDT Examination Requirements in Table NST-1, Class 3 Rev. 35 1.2-18

MPS2 UFSAR TABLE 1.2-1 CONTINUED TABLE 1.2-1

SUMMARY

OF CODES AND STANDARDS FOR COMPONENTS OF WATER-COOLED NUCLEAR POWER UNITS (1)

CODE CLASSIFICATION Component Group A Group B Group C Group D Pumps and Valves 1. ASME Standard Code for Pumps and Draft ASME Code for Pumps and Valves, Class II, Draft ASME Code for Pumps and Valves Class Valves - ANSI B 31.1.0 or Equivalent Valves for Nuclear Power, Class 1, March November 1968. See Footnote (5). III Pumps - Draft ASME Code for Pumps 1970 Draft and Valves Class III or Equivalent

2. ASME Section III, Paragraph N153 in Summer 1969 Addenda

3. ASME Section III, Appendix IX Reactor High Pressure Safety Injection Pumps and Valves Vital Chilled Water Pump Vital Chilled Water Valves Coolant Pumps and Valves Low Pressure Safety Injection Pumps and Valves Service Water Pumps and Valves ASME Section III 1971 Edition, 1971 Winter Addenda Standards of the Hydraulic Institute, ANSI G16.5 Class 1 Reactor Coolant System Branch Connection Valves RBCCW Pumps and Valves Standards of beyond Second Isolation Valves ASME Standard Code the Hydraulic Institute, ANSI B16.1, for Pumps and Valves, Class 2, March 1970 draft ANSI B31.1 All Containment Penetration Isolation Valves ASME Auxiliary Feedwater Pumps ASME Code Section III, 1971; Draft ASME Pump and Valve Code, for Pumps and Valves for Nuclear Power, 1980, 1983 Class II NEMA Standard SM20-1958 Hydraulic Institute Chemical and Volume Control System-Concentrated Boric Acid Service-Pumps Acid Service-Pumps and Valves Draft ASME Code for Pump and Valves, Class II, November 1968 Containment Spray Pumps and Valves Pressurizer Safety Valves 1. ASME Section III, Class A, 1968 Edition, Addenda through summer of 1970. Code Case 1344-1.

Pressurizer Power ASME Section III Class 1, 1977 Edition Operated Relief Valves through winter 1979 Addenda Rev. 35 1.2-19

MPS2 UFSAR TABLE 1.2-1 CONTINUED TABLE 1.2-1

SUMMARY

OF CODES AND STANDARDS FOR COMPONENTS OF WATER-COOLED NUCLEAR POWER UNITS (1)

CODE CLASSIFICATION Component Group A Group B Group C Group D Piping 1. ANSI B 31.7 Class I, 1969 Edition ANSI B 31.7, Class II 1969 Edition ANSI B 31.7, Class III 1969 Edition ANSI B 31.1.0 or Equivalent

2. ASME Section III, Paragraph N153 in Summer 1969 Addenda

3. Code Case 70 to B31.7 Primary Coolant Piping and Surge Line High Pressure Safety Injection Piping Low Pressure Safety Injection Piping

4. Other Reactor Coolant Pressure Pressure Reactor Coolant System Branch Piping beyond Second Service Water Piping RBCCW Piping Boundary Class I Isolation Valves Piping-ASME Section III Code - 1971 Chemical and Volume Control System Concentrated Edition, Class I. Boric Acid Service Piping ANSI B31.1.0 modified (inside Containment) Containment Spray Piping All Containment Piping Penetrations

1. ANSI B-31.1, Piping Code, ANSI B31.7 Nuclear Piping Code, Class I or II as a minimum, 1969 Edition.

3. ASME Section III, Class 1 or 2, 1971 Edition 1 This table summarizes the Codes and Standards used for major pressure retaining components. Not all components are listed. Later codes and standards may be employed for plant modifications if permitted by applicable design and regulatory requirements in effect at the time of the modification.

2 The reactor vessel head and the replacement pressurizer are constructed in accordance with ASME Boiler & Pressure Vessel Code,Section III, Subsection NB 1998 Edition, through 2000 Addenda.

3 Including ASME Code Case N-416.

4 1971 ASME Boiler and Pressure Code,Section III, Class 3.

5 All pressure-retaining cast parts shall be radiographed (or ultrasonically tested to equivalent standards). Where size or configuration does not permit effective volumetric examination, magnetic particle or liquid penetrant examination may be substituted. Examination procedures and acceptance standards shall be at least equivalent to those specified in the applicable class in the code.

Rev. 35 1.2-20

Withheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-1 SITE LAYOUT

ithheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-2 PLOT PLAN 1.2-22 Rev. 35

ithheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-3 GENERAL ARRANGEMENT, TURBINE BUILDING PLAN AT OPERATING FLOOR ELEVATION 54 FEET 6 INCHES 1.2-23 Rev. 35

ithheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-4 GENERAL ARRANGEMENT, TURBINE BUILDING PLAN AT MEZZANINE FLOOR ELEVATION 31 FEET 6 INCHES 1.2-24 Rev. 35

ithheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-5 GENERAL ARRANGEMENT, TURBINE BUILDING PLAN AT GROUND FLOOR ELEVATION 14 FEET 6 INCHES 1.2-25 Rev. 35

ithheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-6 GENERAL ARRANGEMENT CONTAINMENT PLAN AT FLOOR ELEVATION 14 FEET 6 INCHES AND ELEVATION 36 FEET 6 INCHES 1.2-26 Rev. 35

ithheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-7 GENERAL ARRANGEMENT AUXILIARY BUILDING PLAN AT ELEVATION 36 FEET 6 INCHES AND ELEVATION 38 FEET 6 INCHES 1.2-27 Rev. 35

ithheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-8 GENERAL ARRANGEMENT AUXILIARY BUILDING SECTIONS G-G AND H-H 1.2-28 Rev. 35

ithheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-9 GENERAL ARRANGEMENT AUXILIARY BUILDING GROUND FLOOR ELEVATION 14 FEET 6 INCHES AND CABLE VAULT ELEVATION 25 FEET 6 INCHES 1.2-29 Rev. 35

ithheld under 10 CFR 2.390 (d) (1)

IGURE 1.2-10 GENERAL ARRANGEMENT CONTAINMENT AND AUXILIARY BUILDING PLAN AT ELEVATION (-)5 FEET 0 INCHES AND ELEVATION (-)3 FEET 6 INCHES 1.2-30 Rev. 35

ithheld under 10 CFR 2.390 (d) (1)

IGURE 1.2-11 GENERAL ARRANGEMENT CONTAINMENT AND AUXILIARY BUILDING PLAN AT ELEVATION (-)25 FEET 6 INCHES AND ELEVATION (-)22 FEET 6 INCHES 1.2-31 Rev. 35

ithheld under 10 CFR 2.390 (d) (1)

IGURE 1.2-12 GENERAL ARRANGEMENT CONTAINMENT AND AUXILIARY BUILDING PLAN AT ELEVATION (-)45 FEET 6 INCHES 1.2-32 Rev. 35

ithheld under 10 CFR 2.390 (d) (1)

IGURE 1.2-13 GENERAL ARRANGEMENT CONTAINMENT AND AUXILIARY BUILDING SECTION A-A 1.2-33 Rev. 35

ithheld under 10 CFR 2.390 (d) (1)

IGURE 1.2-14 GENERAL ARRANGEMENT CONTAINMENT AND AUXILIARY BUILDING SECTION B-B 1.2-34 Rev. 35

ithheld under 10 CFR 2.390 (d) (1)

IGURE 1.2-15 GENERAL ARRANGEMENT TURBINE BUILDING SECTIONS C-C AND E-E 1.2-35 Rev. 35

ithheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-16 GENERAL ARRANGEMENT TURBINE BUILDING SECTIONS D-D AND F-F 1.2-36 Rev. 35

ithheld under 10 CFR 2.390 (d) (1)

IGURE 1.2-17 GENERAL ARRANGEMENT INTAKE STRUCTURE AUXILIARY STEAM BOILER ROOM PLAN AND SECTION 1.2-37 Rev. 35

le 1.3-1 presents a summary of the characteristics of the Millstone Unit 2 Nuclear Power Plant he time of application for operating license. The table includes similar data for Calvert Cliffs ts 1 and 2, Maine Yankee Unit Number 1, Turkey Point Units Numbers. 3 and 4 and Palisades t Number 1. Bechtel Corporation and Combustion Engineering (CE), Inc. are identified as tractors in Section 1.6. The Palisades plant is included in the table because its coolant system milar to that of Millstone Unit 2, because both Bechtel Corporation and CE, Inc. are Palisades tractors and because it is an example of a CE, Inc. nuclear steam supply system which is rating. Calvert Cliffs and Maine Yankee were selected because their cores are similar to that of lstone Unit 2 and the most contemporaneous plants for which operating licenses have been ed with which CE is associated. Turkey Point is included because it is another comparable t with which Bechtel Corporation is associated.

1.3-1 Rev. 35

MPS2 UFSAR TABLE 1.3-1 COMPARISON WITH OTHER PLANTS HYDRAULIC and THERMAL DESIGN PARAMETERS REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Total Core Heat Output, MWt 3.5 2,560 2,200 2,200 2,560 2,440 Total Core Heat Output, Btu/hr 3.5 8,737 x 106 7,479 x 106 7,509 x 106 8,740 x 106 8,328 x 106 Heat Generated in Fuel, % 3.5 97.5 97.4 97.5 97.5 97.5 Maximum Overpower, % 3.5 12 12 12 12 12 System Pressure, Nominal, psia 3.5 2,250 2,250 2,100 2,250 2,250 System Pressure, Minimum Steady State, psia 3.5 2,200 2,200 2,050 2,200 2,200 Hot Channel Factors, Overall Heat Flux, Fq 3.5 3.00 3.23 3.8 3.00 2.89 Hot Channel Factors, Enthalpy Rise, F H 3.5 1.65 1.77 2.51 1.65 1.62 DNB Ratio at Nominal Conditions 3.5 2.30 1.81 2.00 2.18 2.45 Coolant Flow: Total Flow Rate, lb/hr 3.5 134 x 106 101.5 x 106 125 x 106 122 x 106 122 x 106 Coolant Flow: Effective Flow Rate for Head 3.5 130 x 106 97.0 x 106 121.25 x 106 117.5 x 106 117.5 x 106 Transfer, lb/hr Coolant Flow: Effective Flow Area for Heat 3.5 53.5 41.8 58.7 53.5 53.5 Transfer, ft2 Coolant Flow: Average Velocity along Fuel 3.5 16 14.3 12.7 13.6 13.9 Rods, ft/sec Coolant Flow: Average Mass Velocity, lb/hr-ft2 3.5 2.4 x 106 2.32 x 106 2.07 x 106 2.20 x 106 2.29 x 106 Coolant Temperatures, °F: Nominal Inlet 3.5 542 546.2 545 543.4 538.9 Coolant Temperatures, °F: Maximum Inlet due 3.5 544 550.2 548 548 546 to Instrumentation Error and Deadband, °F Coolant Temperatures, °F: Average Rise in 3.5 45 55.9 46 52 51.1 Vessel, °F Coolant Temperatures, °F: Average Rise in Core, 3.5 46 58.3 47 54 53.1

°F Coolant Temperatures, °F: Average in Core, °F 3.5 565 575.4 568.5 570.4 565.4 Coolant Temperatures, °F: Average in Vessel 3.5 564 574.2 568 569.5 564.4 Rev. 35 1.3-2

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONTD)

HYDRAULIC and THERMAL DESIGN PARAMETERS REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Coolant Temperatures, °F: Nominal Outlet of 3.5 640 642 642.8 643 636 Hot Channel Average Film Coefficient, Btu/hr-ft2-F 3.5 5270 5400 4860 5240 5300 Average Film Temperature Difference, °F 3.5 34.5 31.8 30 33.5 33 Heat Transfer at 100% Power: Active Heat 3.5 48,400 42,460 51,400 48,416 47,000 Transfer Surface Area, ft2 Heat Transfer at 100% Power: Average Heat 3.5 176,600 171,600 142,400 176,000 170,200 Flux, Btu/hr-ft2 Heat Transfer at 100% Power: Maximum Heat 3.5 527,800 554,200 541,200 527,900 502,300 Flux, Btu/hr-ft2 Heat Transfer at 100% Power: Average Thermal 3.5 5.94 5.5 4.63 5.94 5.74 Output, kw/ft Heat Transfer at 100% Power: Maximum 3.5 16.6 17.6 (2) 17.6 (2) 17.8 16.9 Thermal Output, kw/ft Maximum Clad Surface Temperature at Nominal 3.5 657 657 648 657 657 Pressure, °F Fuel Center Temperature, °F: Maximum at 100% 3.5 3,780 4,030 4,040 3,780 3,640 Power Fuel Center Temperature, °F: Maximum at Over 3.5 4,070 4,300 4,350 4,070 3,940 Power Thermal Output, kw/ft at Maximum Over Power 3.5 19.6 20.0 19.7 (2) 20.0 19.0 CORE MECHANICAL DESIGN PARAMETERS REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Fuel Assemblies: Design 3.3 CEA RCC Cruciform CEA CEA Fuel Assemblies: Rod Pitch, inches 3.3 0.58 0.563 0.550 0.58 0.580 Fuel Assemblies: Cross-Section Dimensions, 3.3 7.98 x 7.98 8.426 x 8.426 8.1135 x 8.1135 7.98 x 7.98 7.98 x 7.98 inches 1.3-3 Rev. 35

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONTD)

CORE MECHANICAL DESIGN PARAMETERS REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Fuel Assemblies: Fuel Weight (as UO2), pounds 3.3 207,035 176,200 210,524 207,269 203,934 Fuel Assemblies: Total Weight, pounds 3.3 282,500 226,200 295,800 282,570 279,235 Fuel Assemblies: Number of Grids per 3.3 8 7 8 8 8 Assembly Fuel Rods: Number 3.3 36,896 32,028 43,168 36,896 36,352 Fuel Rods: Outside Diameter, inches 3.3 0.44 0.422 0.4135 0.44 0.440 Fuel Rods: Diametral Gap, inches 3.3 0.0085 0.0065 0.0065 0.0085 0.0085 Fuel Rods: Clad Thickness, inches 3.3 0.026 0.0243 0.022 0.026 0.026 Fuel Rods: Clad Material 3.3 Zircaloy Zircaloy Zircaloy Zircaloy Zircaloy Fuel Pellets: Material 3.3 UO2 Sintered UO2 Sintered UO2 Sintered UO2 Sintered UO2 Sintered Fuel Pellets: Diameter, inches 3.3 0.3795 0.367 0.359 0.3795 0.3795 Fuel Pellets: Length, inches 3.3 0.650 0.600 0.600 0.650 0.650 Control Assemblies: Neutron Absorber 3.3 B4C / S.S. Cd-In-Ag Cd-In-Ag (5-15-80%) Cd-In-Ag (5-15-80%) Cruciform B4C / S.S. / Cd-In-Ag B4C / S.S. / Cd-In-Ag Control Assemblies: Cladding Material 3.3 NiCrFe Alloy 304 SS-Cold Worked 304 SS Tubes, E.B. NiCrFe Alloy NiCrFe Alloy (Inconel 625) Welded to 13.5 inch span Control Assemblies: Clad Thickness 3.3 0.040 0.109 0.016 0.040 0.040 Control Assemblies: Number of Assembly, full / 3.3 73 53 41 / 4 Cruciform Rods 77 / 8 77 / 8 part length Control Assemblies: Number of Rods per 3.3 5 20 117 Tubes per Rod 5 5 Assembly Core Structure: Core Barrel ID / OD, inches 3.3.2.2 148 / 151.5 133.875 / 137.875 149.75 / 152.5 148 / 149.75 148 / 149.75 Core Structure: Thermal Shield ID / OD, inches 3.3.2.5 156.75 / 162.75 142.625 / 148.0 None None 156 / 162 1.3-4 Rev. 35

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONTD)

NUCLEAR DESIGN DATA REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Structural Characteristics: Core Diameter, inches 3.3.1 136 119.5 136.71 136.0 136.0 (Equivalent)

Structural Characteristics: Core Height, inches 3.3.1 136.7 144 132 136.7 136.7 (Active Fuel)

H2O/U, Unit Cell (Cold) 3.4.1 3.50 4.18 3.50 3.44 3.44 Number of Fuel Assemblies 3.3 217 157 204 217 217 UO2 Rods per Assembly, Unshimmed / - 204 212 / 208 - -

Shimmed UO2 Rods per Assembly, Unshimmed / 3.3 176 - - 176 176 Shimmed: Batch A UO2 Rods per Assembly, Unshimmed / 3.3 164 - - 164 160 Shimmed: Batch B UO2 Rods per Assembly, Unshimmed / 3.3 (176 / 164 / 164) - - (176 / 164 / 164) (176 / 164 / 160)

Shimmed: Batch C Performance Characteristics Loading Technique 3.4.1 3 Batch Mixed Central 3 Regions Non-uniform 3 Batch Mixed Central Zone 3 Batch Mixed Central 3 Batch Mixed Central Zone Zone Zone 1.3-5 Rev. 35

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONTD)

NUCLEAR DESIGN DATA REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Fuel Discharge Burnup, Mwd/MTU: Average 3.4.1 12,770 13,000 10,180 13,775 13,795 First Cycle Fuel Discharge Burnup, Mwd/MTU: First Core 3.2.1 22,000 14,500 17,600 22,550 30,000 Average Feed Enrichment (weight percent): Region 1 3.4.1 1.93 1.85 1.65 2.05 2.01 Feed Enrichment (weight percent): Region 2 3.4.1 2.33 2.55 2.08 / 2.54 2.45 2.40 Feed Enrichment (weight percent): Region 3 3.4.1 2.82 3.10 2.54 / 3.20 2.99 2.95 Feed Enrichment (weight percent): Equilibrium - - 2.54 / 3.20 - -

Control Characteristics Effective Multiplication 3.4.1 1.170 1.180 1.212 1.194 1.170 (beginning of life): Cold, No Power, Clean Control Characteristics Effective Multiplication 3.4.1 1.129 1.38 1.175 1.152 1.129 (beginning of life): Hot, No Power, Clean Control Characteristics Effective Multiplication 3.4.1 1.078 1.077 1.111 1.094 1.075 (beginning of life): Hot, Full Power, Xe Equilibrium Control Assemblies: Material 3.3 B4C / S.S. Cd-In-Ag Cd-In-Ag (5-15-80%) Cd-In-Ag (5-15-80%) B4C / S.S.-Cd-In-Ag B4C / S.S.-Cd-In-Ag Control Assemblies: Number of Control 3.4.1 73 53 45 Cruciform 85 85 Assemblies Number of Absorber Rods per RCC (or CEA) 3.3 5 20 117 Tubes Welded to Form 13.5 inches 5 5 Assembly span Total Rod Worth (Hot), % 3.4.1 11.0 7 8.6 9.6 9.9 Boron Concentrations - To shut reactor down 3.4.1 945 / 935 1,250 / 1,210 1,180 / 1,210 1,120 / 1,095 945 / 935 with no rods inserted, clean, ppm: Cold / Hot, ppm Boron Concentrations - To shut reactor down 3.4.1 820 / 590 1,000 / 670 1,070 / 830 960 / 725 820 / 590 with no rods inserted, clean, ppm: To control at power with no rods inserted, clean / equilibrium xenon, ppm Kinetic Characteristics, Range Over Life: 3.4.1 -0.4 x 10-4 to -2.1 x 10-4 +.3 x 10-4 to -1.96 x 10-4 -0.08 x 10-4 to -2.25 x 10-4 -.20 x 10-4 to -1.96 x 10-4 -0.40 x 10-4 to Moderator Temperature Coefficient (3) /°F -3.5 -2.20 x 10-4 1.3-6 Rev. 35

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONTD)

NUCLEAR DESIGN DATA REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Kinetic Characteristics, Range Over Life: 3.4.1 -0.65 x 10-6 to -0.3 x 10-6 to +3.4 x 10-6 +0.10 x 10-6 to +1.7 x 10-6 +0.65 x 10-6 to +0.65 x 10-6 to Moderator Pressure Coefficient (3) /psi +2.39 x 10-6 +2.39 x 10-6 +2.39 x 10-6 Kinetic Characteristics, Range Over Life: 3.4.1 -0.41 x 10-3 to +0.5 x 10-3 to -0.06 x 10-3 to -1.0 x 10-3 -0.41 x 10-3 to -0.41 x 10-3 to Moderator Void Coefficient (3) /% Void -1.43 x 10-3 -2.5 x 10-3 -1.43 x 10-3 -1.43 x 10-3 Kinetic Characteristics, Range Over Life: 3.4.1 -1.45 x 10-5 to -1.0 x 10-5 to -1.6 x 10-5 -1.56 x 10-5 to -1.08 x 10-5 -1.46 x 10-5 to -1.45 x 10-5 to Doppler Coefficient (4) /°F -1.07 x 10-5 -1.06 x 10-5 -1.07 x 10-5 REACTOR COOLANT SYSTEM - CODE REQUIREMENTS REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Reactor Vessel 4.2.2 ASME III Class A ASME III Class A ASME III Class A ASME III Class A ASME III Class A Steam Generator: Tube Side 4.2.2 ASME III Class A ASME III Class A ASME III Class A ASME III Class A ASME III Class A Steam Generator: Shell Side 4.2.2 ASME III Class A ASME III Class C ASME III Class A ASME III Class A ASME III Class A Pressurizer 4.2.2 ASME III Class A ASME III Class A ASME III Class A ASME III Class A ASME III Class A Pressurizer Relief (or Quench) Tank 4.2.2 ASME III Class C ASME III Class C ASME III Class C ASME III Class C ASME III Class C Pressurizer Safety Valves 4.2.2 ASME III ASME III ASME III ASME III ASME III Reactor Coolant Piping 4.2.2 ANSI B 31.7 USAS B 31.1 USAS B 31.1 USAS B 31.7 USAS 31.1 PRINCIPAL DESIGN PARAMETERS OF THE COOLING SYSTEM REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Operating Pressure, psig 4.2.1 2235 2235 2085 2235 2235 Reactor Inlet Temperature, °F 4.2.1 539.7 546.2 545 544.5 540 Reactor Outlet Temperature, °F 4.2.1 595.1 602.1 591.1 599.4 592.8 Number of Loops 4.1 2 3 2 2 3 Design Pressure, psig 4.3.4 2,485 2,485 2,485 2,485 2,485 1.3-7 Rev. 35

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONTD)

PRINCIPAL DESIGN PARAMETERS OF THE COOLING SYSTEM REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Design Temperature, °F 4.3.4 650 650 650 650 650 Hydrostatic Test Pressure (cold), psig 4.2.1 3,110 3,110 3,110 3,110 3,110 Total Coolant Volume, cubic feet 4.2.1 11,101 9,088 10,809 11,101 11,026 PRINCIPAL DESIGN PARAMETERS OF THE REACTOR VESSEL REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Material 4.3.1, 4.5.6 SA-533, Grade B Class I, SA-302, Grade B, low alloy SA-302, Grade B, low alloy steel, SA-533, Grade B, Class I, SA-533, Grade B, low alloy steel plates and steel, internally clad with internally clad with Type 304 austenitic steel, internally clad Type forgings-A-508-64 SA-508-64, Class 2 Type 304 austenitic SS SS 304 austenitic SS Class 2, cladding weld forgings, internally clad deposited 304 SS with Type 304 (5) equivalent austenitic SS Design Pressure, psig 4.3.1 2,485 2,485 2,485 2,485 2,485 Design Temperature, °F 4.3.1 650 650 650 650 650 Operating Pressure, psig 4.2.1 2,235 2,235 2,085 2,235 2,235 Inside Diameter of Shell, inches 4.3.1 172 155.5 172 172 172 Outside Diameter across Nozzles, inches 4.3.1 253 236 254 253 266-5/8 Overall Height of Vessel and Enclosure Head, 4.3.1 41 feet 11.75 inches 41 feet 6 inches 40 feet 1-13/16 inches 41 feet 11.75 inches 42 feet 1-3/8 feet-inches to top of CRD Nozzle inches Minimum Clad Thickness, inches 4.3.1 1/8 5/32 3/16 1/8 1/8 1.3-8 Rev. 35

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONTD)

PRINCIPAL DESIGN PARAMETERS OF THE STEAM GENERATORS REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Number of Units 4.3.2 2 3 2 2 3 Type 4.3.2 Vertical U-Tube with Vertical U-Tube with Vertical U-Tube with integral moisture Vertical U-Tube with Vertical U-tube with integral moisture integral moisture separator separator integral moisture integral moisture separator separator separator Tube Material 4.3.2 Ni-Cr-Fe Alloy Ni-Cr-Fe-Alloy Ni-Cr-Fe Alloy Ni-Cr-Fe Alloy Ni-Cr-Fe Alloy Shell Material 4.3.2 SA-533 Gr. B Class 1 SA-302 Carbon Steel SA-533 Gr. B Class 1 and SA-533 Gr. B Class 1 and SA-516 gr 70 SA-516 gr 70 and SA-516 gr 70 Tube Side Design Pressure, psig 4.3.2 2,485 2,485 2,485 2,485 2,485 Tube Side Design Temperature, °F 4.3.2 650 650 650 650 650 Tube Side Design Flow, lb/hr 4.3.2 61 x 106 33.93 x 106 62.5 x 106 61 x 106 40.67 x 106 Shell Side Design Pressure, psig 4.3.2 1,000 1,085 985 985 985 Shell Side Design Temperature, °F 4.3.2 550 556 550 550 550 Operating Pressure, Tube Side, Nominal, psig 4.3.2 2,235 2,235 2,085 2,235 2,235 Operating Pressure, Shell Side, Maximum, psig 4.3.2 885 1,020 885 885 885 Maximum Moisture at Outlet at Full Load, % 4.3.2 0.2 0.25 0.2 0.2 0.2 Hydrostatic Test Pressure, Tube Side (cold), psig 4.3.2 3,110 3,107 3,110 3,110 3,110 Steam Pressure, psig, at full power 4.3.2 800 730 755 835 800 Steam Temperature, °F, at full power 4.3.2 520.3 510 513.8 525.2 520.3 1.3-9 Rev. 35

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONTD)

PRINCIPAL DESIGN PARAMETERS OF THE REACTOR COOLANT PUMP REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Number of Units 4.3.3 4 3 4 4 3 Type 4.3.3 Vertical, single stage Vertical, single stage radial Vertical, single stage radial flow with Vertical, single stage Vertical, single stage centrifugal with bottom flow with bottom suction bottom suction and horizontal discharge centrifugal with bottom centrifugal with suction and horizontal and horizontal discharge suction and horizontal bottom suction and discharge discharge horizontal discharge Design Pressure, psig 4.3.3 2,485 2,485 2,485 2,485 2,485 Design Temperature, °F 4.3.3 650 650 650 650 650 Operating Pressure, nominal psig 4.3.3 2,235 2,235 2,085 2,235 2,235 Suction Temperature, °F 4.3.3 540 546.5 545 543.4 538.9 Design Capacity, gpm 4.3.3 81,200 89,500 83,000 81,200 108,000 Design Head, feet 4.3.3 243 260 260 300 290 Hydrostatic Test Pressure, (cold), psig 4.3.3 3,110 3,107 3,110 3,110 3,110 Motor Type 4.3.3 AC Induction AC Induction AC Induction AC Induction AC Induction 4.3.3 Single Speed Single Speed Single Speed Single Speed Single Speed Motor Rating, hp 4.3.3 6,500 6,000 6,250 7,200 9,000 PRINCIPAL DESIGN PARAMETERS OF THE REACTOR COOLANT PIPING REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Material 4.3.4 SA516 - GR 70 with Austenitic SS SA212B clad with SS SA516 - gr 70 with SA516 - gr 70 with SS minimum 1/8 304L SS nominal 7/32 SS clad clad clad Hot Leg - ID, inches 4.3.4 42 29 42 42 33.5 Cold Leg - ID, inches 4.3.4 30 27.5 30 30 33.5 Between Pump and Steam Generator - ID, inches 4.3.4 30 31 30 30 33.5 1.3-10 Rev. 35

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONTD)

CONTAINMENT SYSTEM PARAMETERS REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Type 5.2.1 Double containment with Steel lined prestressed post Steel lined prestressed post tensioned Steel lined prestressed Steel lined reinforced steel lined prestressed tensioned concrete cylinder, concrete cylinder, curved dome roof post tensioned concrete concrete flat bottom post tensioned concrete shallow dome roof cylinder, curved dome and hemispherical cylinder, curved dome roof dome roof completely enclosed by Enclosure Building Containment Parameters: Inside Diameter, feet 5.2.1 130 116 116 130 135 Containment Parameters: Height, feet. 5.2.1 175 169 190.5 181-2/3 169.5 Containment Parameters: Free Volume, ft3 5.2.1 1,920,000 (5) 1,550,000 1,640,000 2,000,000 1,855,000 Containment Parameters: Reference Incident 5.2.1 54 59 55 50 55 Pressure, psig Containment Parameters: Concrete Thickness, feet Containment Parameters: Vertical Wall 5.2.1 3.75 3.75 3 3.75 4.5 Containment Parameters: Dome 5.2.1 3.25 3.25 2.5 3.25 2.5 Containment Leak Prevention and Mitigation 6.7.2.1 Completely enclosed Leak tight penetration and Leak tight penetration and continuous Leak tight penetration and Leak tight Systems containment has leaktight continuous steel liner, steel liner, automatic isolation where continuous steel liner, penetration and penetrations and automatic isolation where required automatic isolation where continuous steel liner, continuous steel liner. required required. The exhaust automatic isolation Enclosure Building from penetration rooms to where required Filtration region at small vent.

negative pressure during LCI. Automatic isolation where required. The exhaust from filtration region passed through charcoal filters to 375 feet Millstone stack following incident.

Gaseous Effluent Purge 11.1.2.1.3 Discharge through Unit 2 Through particulate filter & Discharge through stack Discharge through vent Discharge through stack monitors part of main stack exhaust system 1.3-11 Rev. 35

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONTD)

ENGINEERED SAFEGUARDS REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Safety Injection System: Number of High Head 6.3.2.1 3 4 (shared) 3 3 3 (charging)

Pumps Safety Injection System: Number of Low Head 6.3.2.1 2 2 2 2 2 Pumps Safety Injection System: Safety Injection Tank, 6.3.2.1 4 3 4 4 3 number Containment Fan Coolers: Number of Units 6.5.1.2 4 3 4 4 6 Containment Fan Coolers: Air Flow capacity, 6.5.2.2 34,800 65,000 30,000 55,000 N/A each at emergency condition, cfm Post-Incident Filters Inside Containment: None None None None None Number of Units Post-Incident Filters Inside Containment: Type None None None None None Containment Spray Number of Pumps 6.4.2.1 2 2 - 2 3 Emergency Power Diesel Generator Units 8.3.1.1 2 2 total for both units 2 3 total for both units 2 Enclosure Building Filtration System Number of 6.7.2.1 2 - - - 0 Units RADIOACTIVE WASTE PROCESSING SYSTEMS REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Design Failed Fuel, % 11.1.1.1 1 1 1 1 1 Gaseous Waste Processing System 11.1.2.1 11.1.2.1 14,344 (6) 4,539 66,240 (6)

Annual Volume of Gases Discharge, ft3 Annual Activity Discharge, Curies 11.1.2.1 556 14,758 (6) 6) (6)

Decay Storage Time for Gases, Days 11.1.2.1 60 (Minimum) 45 30 (Minimum) 60 (6)

Compressors: Number 2 2 (7) 2 2 (7) 2 1.3-12 Rev. 35

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONTD)

RADIOACTIVE WASTE PROCESSING SYSTEMS REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Compressors: Capacity, each 11.1.2.2 25 SCFM 40 CFM 2.35 SCFM 4 to 7 SCFM (6)

Decay Tanks: Number 6 6 (7) 3 3 (7) 3 Decay Tanks: Capacity, (each), ft3 582 525 100 610 200 LIQUID WASTE PROCESSING SYSTEMS REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Clean Liquid Waste (Reactor Coolant Wastes) 11.1.2.1 Design Volume Wastes per Year 11.1.2.1 14 Reactor Coolant (6) 14 Reactor Coolant (6)

System Volumes System (840,000 Gallons)

Expected Volume of Waste Discharge Per Year, 11.1.2.1 404,234 (Design Incorporates 724,300 805,542 (6)

Gallons Recycle of Waste to R.C.

System Clean Liquid Waste (6) (6) (6)

Annual Expected Activity Discharged, curies 11.1.2.1 286 (includes H3) System Not Compared)

Percentage of 10 CFR Part 20 11.1.4.1 0.6% (6) (6) (6)

Degasifier: Number 11.1.2.2 1 1 2 2 Degasifier: Type 11.1.2.2 Packed Column Utilizing Vacuum Packed Tower Flashing Internal Generated Stripping Steam Degasifier: Design Flow Rate, gpm 11.1.2.2 132 160 120 100 Degasifier: Decontamination Factors 11.1.2.2 1,000 (Kr & Xe) 40 10 (6)

Storage Tanks: Number 11.1.2.2 4 4 4 2 Storage Tanks: Total Capacity 11.1.2.2 3 Reactor Coolant System 200,000 Gallons 6 Reactor Coolant System 250,000 Gallons (180,000 Gallons) Volumes (7)

Storage Tanks: Vent Discharge 11.1.2.2 To Gaseous Waste System To Exhaust Plenum Plant Vent To ventilation System for storage and decay and stack Demineralizers: Number 11.1.2.2 3 3 4 2 1.3-13 Rev. 35

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONTD)

LIQUID WASTE PROCESSING SYSTEMS REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Demineralizers: Type 11.1.2.2 Mixed Bed Non Mixed bed Mixed Bed Non Cesium Removal Regenerative Regenerative Demineralizers: Decontamination 11.1.2.2 1,000 10 100 (6)

Demineralizers: Factors 11.1.2.2 (0 for Y, Mo, H3)

Evaporator (Boron Recovery): Number 11.1.2.2 1 N/A 2 1 Evaporator (Boron Recovery): Type 11.1.2.2 Vacuum, Submerged Horizontal Spray Film Forced Calculating, U-Tube Single Effect Evaporator (Boron Recovery): Capacity, GPM 11.1.2.2 25 20 30 Distillate Evaporator (Boron Recovery): Decontamination 11.1.2.2 (6) 105 (Nonvolatiles)

Evaporator (Boron Recovery): Factors 11.1.2.2 1,000 (Nonvolatiles), 50 104 (Gases)

(Halogens), 100 (Dissolved Gases)

Aerated Liquid Waste Processing System (Miscellaneous Wastes)

REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Design Volume of Waste per year 11.1.2.1 3,639,400 (Gallons) (6) (6) (6) (6)

Expected Volume of Waste Discharged per year, 11.1.2.1 313,000 508,620 (6) 1,330,320 (6)

Gallons Annual Expected Activity Discharged, Curies 11.1.2.1 1.11 (includes H3) 0.077 (6) (6) (6)

Percentage of 10 CFR Part 20 11.1.4.1 Less than 0.1%

Storage Tanks: Number 11.1.2.2 1 2 1 2 2 Storage Tanks: Total Capacity 11.1.2.2 5,000 Gallons 2,000 Gallons 5,500 Gallons 8,000 Gallons 24,800 Gallons Demineralizers: Number 11.1.2.2 1 (6) N/A 1 N/A Demineralizers: Type 11.1.2.2 Mixed Bed Non (6) Mixed Bed Non Regenerative Regenerative Demineralizers: Decontamination Factors 11.1.2.2 500 (6) 100 1.3-14 Rev. 35

MPS2 UFSAR Aerated Liquid Waste Processing System (Miscellaneous Wastes)

REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Evaporator: 11.1.2.2 N/A N/A N/A Evaporator: Number 11.1.2.2 1 (7) 1 Evaporator: Type 11.1.2.2 (6) (6)

Evaporator: Capacity, Distillate GPM 11.1.2.2 (6) (6)

Evaporator: Decontamination Factors 11.1.2.2 (6) (6) 106 Solid Waste Processing System REFERENCE CYCLE 1 MILLSTONE TURKEY POINT (1) CALVERT CLIFFS (1)

<Parameter> SECTION UNIT 2 UNITS 3 AND 4 PALISADES (1) UNIT 1 UNITS 1 AND 2 MAINE YANKEE (1)

Evaporator Concentrates 11.1.2.1 Solidified in Concrete in Solidified in Concrete in 55 N/A Solidified in Concrete in 55 Gallon drums 55 Gallon drums Gallon drums 55 Gallon drums Spent Resins Shipping & Volumes 11.1.2.1 Shipping cask after Dewatered 55 Gallon (6) Solidified in Concrete in Shipping cask dewatering, 225 ft3 Drums 55 Gallon Drums Contaminated Filter Cartridges & Volumes 11.1.2.1 55 Gallon drums 55 Gallon drums 55 Gallon drums Solidified in Concrete in Cask or 55 Gallon 55 Gallon Drums drums Annual Activity Shipped, curies 11.1.2.1 4,250 (6) (6) (6) (6) 1 The values listed for these plants were taken from public documentation.

2 Based on total heat output of the core rather than heat generated in the fuel alone.

3 Values shown are for beginning of life full power / end of cycle full power.

4 Values shown are for beginning of life zero power/beginning of life cycle full power.

5 Measured value from pre-operational volume verification test and used for integrated leak rate testing. Includes volume of vented pressurizer, safety injection tanks, and other tanks.

6 Not Specifically Available in Public Documents.

7 Shared by Two (2) Units.

Rev. 35 1.3-15

principal architectural and engineering features used in the design of Unit 2 of the Millstone lear Power Station are summarized in the following material.

1 PLANT DESIGN cipal structures and equipment which may serve either to prevent accidents or to mitigate r consequences have been designed, fabricated and erected in accordance with applicable es so as to withstand the most severe earthquakes, flooding conditions, windstorms, ice ditions, temperature and other deleterious natural phenomena which could be reasonably med to occur at the site during the lifetime of this plant. Systems and components designed Seismic Category I requirements are listed in Table 1.4-1. It should be noted that the terms tegory and Class are used interchangeably throughout the MP2 FSAR in defining seismic gn classifications of Structures, Systems and Components. Unit 2 was designed so that the ty of one unit will not be impaired in the unlikely event of an accident in the other unit.

cipal structures and equipment were sized for the maximum expected nuclear steam supply em (NSSS) and turbine outputs.

undancy is provided in the reactor and safety systems so that the single failure of any active ponent of either system cannot prevent the action necessary to avoid an unsafe condition. The is designed to facilitate inspection and testing of systems and components whose reliabilities important to the protection of the public and plant personnel.

visions have been made to protect against the hazards of such events as fires or explosions.

tems and components which are significant from the standpoint of nuclear safety are designed, icated and erected to quality standards commensurate with the safety function to be ormed. Appendix 1.A of this FSAR addresses the implementation of Atomic Energy mmission (AEC) General Design Criteria for Nuclear Power Plants, 10 CFR Part 50, endix A. Section 12.8 describes the Quality Assurance Program.

2 REACTOR following criteria (see Chapter 3) apply to the reactor:

a. The reactor is of the pressurized water-type, designed to provide heat to steam generators which, in turn, provide steam to drive a turbine generator. The initial full power core thermal output was 2560 megawatts (the NSSS rating was 2570 megawatts) prior to its uprating to the current 2700 megawatts thermal power level (NSSS rating of 2715 megawatts).

b. The reactor is refueled with slightly enriched uranium dioxide contained in zircalloy tubes.

1.4-1 Rev. 35

failure or damage. The maximum fuel centerline temperature evaluated at the design overpower condition will be below that value which could lead to fuel rod failure. The melting point of the UO2 will not be reached during routine operation and anticipated transients.

d. Fuel rod clad is designed to maintain cladding integrity throughout fuel life.

Fission gas release within the rods and other factors affecting design life will be considered for the maximum expected exposures.

e. The reactor and control systems are designed so that any xenon transients can be adequately damped.

f. The reactor is designed to accommodate the anticipated transients safely and without fuel damage.

g. The reactor coolant system (RCS) is designed and constructed to maintain its integrity throughout the expected plant life. Appropriate means of test and inspection are provided.

h. Power excursions which could result from any credible reactivity addition accident will not cause damage, either by deformation or rupture, to the pressure vessel or impair operation of the engineered safety features (ESF).

i. Control element assemblies (CEA) are capable of holding the core subcritical at hot zero power conditions following a trip, and providing a safety margin even with the most reactive CEA stuck in the full, withdrawn position.

j. The chemical and volume control system (CVCS) can add boric acid to the reactor coolant at a sufficient rate to maintain an adequate shutdown margin when the RCS is cooling down following a reactor trip. This is accomplished at a maximum design rate. This system is independent of the CEA system.

k. The combined response of the fuel temperature coefficient, the moderator temperature coefficient, the moderator void coefficient and the moderator pressure coefficient to an increase in reactor thermal power is a decrease in reactivity. In addition, the reactor power transient remains bounded and damped in response to any expected changes in any operating variable.

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3.1 Reactor Coolant System design bases in this section are those used for the integrated design of the RCS or those which ly to all of the system components. The design bases unique to each component are discussed ection 4.3.

RCS is designed for the normal operation of transferring 2715 MWt (9.26 x 10 Btu/hr) from reactor core (2700 MWt) and reactor coolant pumps (15 MWt) to the steam generators. In the m generator, this heat is transferred to the secondary system forming 5.9 x 106 lb/hr of 880 saturated steam per generator with a 0.2 percent maximum moisture content.

e RCS is designed to accommodate the normal design transients listed. These transients ude conservative estimates of the operational requirements of the systems and are used to e the required component fatigue analyses.

a. 500 heatup and cooldown cycles at a maximum heating and cooling rate of 100°F/hr. The pressurizer is designed for a maximum cooldown rate of 200°F/hr.

b. Pressurizer spray piping is limited to 160 plant heatup and cooldown cycles.

Primary manway studs of the replaced steam generators are limited to 200 heatup and cooldown cycles.

c. 15,000 power change cycles in the range between 15 and 100 percent of full load with a ramp load change of 5 percent of full load per minute increasing or decreasing. This will occur without reactor trip.

d. Primary manway studs for the replaced steam generators are limited to 1,000 cycles with a ramp load change of 5% per minute decreasing and 30% per hour increasing (plant loading/unloading).

e. 2,000 step power changes of 10 percent, both increasing and decreasing between 15 and 100 percent of full load. Primary manway studs for the replaced steam generator are limited to 1,500 step power changes.

f. 10 cycles of hydrostatic testing at 3,110 psig and a temperature at least 60°F above the nil ductility transition temperature (NDTT) of the component having the highest NDTT.

g. 200 cycles of leak testing at 2,485 psig and a temperature at least 60°F greater than the NDTT of the component with the highest NDDT.

h. Primary manway studs for the replaced steam generators are limited to 80 cycles of leak testing at 2,485 psig.

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operating pressure and +/-6°F at operating temperature and pressure.

j. 400 reactor trips when at 100 percent power. Primary manway studs for the replaced steam generator are limited to 200 reactor trips when at 100% power.

addition to these normal design transients, the following abnormal transients are also sidered to arrive at a satisfactory usage factor as defined in Section III, Nuclear Vessels, of the ME Boiler and Pressure Vessel Code:

a. 40 cycles of loss of turbine load from 100 percent power.

b. 40 cycles of loss of reactor coolant flow when at 100 percent.

c. 5 cycles of loss of main steam system pressure.

mponents of the RCS are designed and will be operated so that no deleterious pressure or mal stress will be imposed on the structural materials. The necessary consideration has been en to the ductile characteristics of the materials at low temperature.

3.2 Chemical and Volume Control System major functions of the CVCS (see Section 9.2) are to:

a. Maintain the required volume of water in the RCS.

b. Maintain the chemistry and purity of the reactor coolant.

c. Maintain the desired boric acid concentration in the reactor coolant.

d. Provide a controlled path to the waste processing system.

system is designed to accept the discharge when the reactor coolant is heated at the design of 100°F/hr and to provide the required makeup when the reactor coolant is cooled at the gn rate of 100°F/hr. Discharge is automatically diverted to the waste processing system when volume control tank is at its highest permissible level. The system will also supply makeup or ept discharge due to power decreases or increases. The design transients are +/-10 percent of full er step changes and ramp changes of +/-5 percent of full power per minute between 15 to 100 ent power. On power increases, the letdown flow is automatically diverted to the waste cessing system when the volume control tank reaches the highest permissible level. On power reases, sufficient coolant is in the volume control tank to allow a full to zero power decrease hout additional makeup, in the event of a makeup system failure or override.

an assumed 1 percent failed fuel condition, the activity in the reactor coolant does not exceed Ci/cc at 77°F. The system is also designed to maintain the reactor coolant chemistry within limits specified in Section 4.4.3.

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uired boron (as boric acid). The maximum rate at which the reactor coolant boron centration can be reduced must be substantially less than the equivalent maximum rate of tivity insertion by the CEA.

r to refueling, the system is capable of increasing the reactor coolant boron concentration m zero to 1720 ppm by feed and bleed when the reactor coolant is at hot standby operating perature.

visions to facilitate the plant hydrostatic testing and to leak test the RCS are included.

3.3 Shutdown Cooling System shutdown cooling system (see Section 9.3) is designed to cool the RCS from approximately

° to 130°F in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, assuming that the component cooling water inlet temperature is at its imum design value of 95°F. The design RCS cooldown rate is 100°F/hr. A temperature of

°F or less can be achieved 27.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after reactor shutdown, assuming an infinitely exposed

. The maximum allowable pressure for the RCS during shutdown cooling is approximately psig.

4 CONTAINMENT SYSTEM containment (see Sections 5.2 and 14.8), including the associated access openings and etrations, is designed to contain pressures and temperatures resulting from a postulated main mline break (MSLB) in which:

a. A range of power level, break sizes, and single failures are considered.

b. Cases with the loss of offsite power and with AC power available are analyzed to determine which scenario maximizes the energy removal into containment.

c. Safety injection is not assumed since it would tend to reduce the energy released into containment.

d. The containment air recirculation cooling system and the containment spray system are credited to mitigate the containment pressure and temperature consequences.

tainment response to a loss-of-coolant (LOCA) accident was also analyzed. It was found that peak containment pressure and temperature of the MSLB accident bound the LOCA.

containment is designed to assure integrity against postulated missiles from equipment ures and against postulated missiles from external sources.

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ated seals, sealing compounds, expansion bellows, and the interior of the containment.

enclosure building (see Section 5.3) is designed to withstand a wind loading of 115 mph, with ts of 140 mph, snow load of 60 psf and seismic loads. The Enclosure Building is designed so is structural framing will withstand tornado loads, but the siding will be blown away (see tion 5.3.3).

5 ENGINEERED SAFETY FEATURES SYSTEMS design incorporates redundant independent full capacity engineered safety features systems FS). These, in conjunction with the containment, ensure that the release of fission products, owing any postulated occurrence, at least the minimum ESF required to terminate that urrence are operable. The following are required as minimum safety features:

One high pressure safety injection (HPSI) train One low pressure safety injection (LPSI) train Four safety injection tanks (water quantity of three is required to reach the core)

One containment spray and two containment air recirculation and cooling subsystems, or equivalent (Section 6.4)

One hydrogen control subsystem One enclosure building filtration train One auxiliary feedwater trains h of these subsystems is independent of its redundant counterpart with the exception of the ty injection subsystems. The HPSI and LPSI subsystems (Section 6.3) are independent up to common pipe connections to the four reactor coolant cold legs. Remote manually operated es provide appropriate cross-connections between redundant subsystems for backup and to w maintenance. Redundant components are physically separated.

ESFS are designed to perform their functions for all break sizes in the RCS piping up to and uding the double-ended rupture of the largest reactor coolant pipe. The safety injection system ts fuel and cladding damage to an amount which will not interfere with adequate emergency cooling and holds metal-water reactions to minimal amounts. Two full capacity systems, ed on different principles remove heat from the containment to maintain containment integrity, containment spray system (Section 6.4) and the containment air recirculation and cooling em (Section 6.5). The enclosure building filtration system (EBFS) (Section 6.7) maintains the losure building filtration region (EBFR) at a slightly negative pressure and filters the exhaust m this space. The containment postaccident hydrogen control system (Section 6.6) mixes and 1.4-6 Rev. 35

6 PROTECTION, CONTROL AND INSTRUMENTATION SYSTEM eactor protective system (RPS) (see Section 7.2) is provided which initiates reactor trip if the tor approaches an unsafe condition.

rlocks and automatic protective systems are provided along with administrative controls to ure safe operation of the plant.

ficient redundancy is installed to permit periodic testing of the RPS so that failure or removal m service of any one protective system component or portion of the system will not preclude tor trip or other safety action when required.

protective system is isolated from the control instrumentation systems so that failure or oval from service of any control instrumentation system component or channel does not bit the function of the protective system.

7 ELECTRICAL SYSTEMS mal, reserve and emergency sources of auxiliary electrical power are provided to assure safe orderly shutdown of the plant and to maintain a safe shutdown condition under all credible umstances. Onsite electrical power sources and systems are designed to provide dependability, pendence, redundancy and testability in accordance with the requirements of 10 CFR Part 50, endix A. The load-carrying capability and other electrical and mechanical characteristics of rgency power systems are in accordance with the requirements of Safety Guide Number 9.

o redundant, independent, full capacity emergency power sources and distribution subsystems provided. Each of these subsystems powers all equipment in the associated safety related systems as described in Section 1.4.5.

8 RADIOACTIVE WASTE PROCESSING SYSTEM radioactive waste processing system (see Section 11.1) is designed so that discharges of oactivity to the environment are minimized and are in accordance with the requirements of tions 1301 and 1302 and Appendix B of 10 CFR Part 20 and Appendix I of 10 CFR Part 50.

9 RADIATION PROTECTION lstone Unit 2 is provided with a centralized control room which has adequate shielding (see tion 11.2.2.3) and ventilation system features (see Section 9.9.10) to permit occupancy during postulated accidents involving radiation releases.

radiation shielding in Millstone Unit 2 and the radiation control procedures ensure that rating personnel do not receive exposures during normal operation and maintenance in excess he applicable limits of 10 CFR Part 20.

1.4-7 Rev. 35

l handling and storage facilities (see Section 9.8) are provided for the safe handling and age of fuel. The design precludes accidental criticality.

1.4-8 Rev. 35

System Components fety Injection System HPSI pumps and motors LPSI pumps and motors Safety Injection Tanks Refueling Water Storage Tank Piping and supports Valves and valve operators ntainment Spray System Containment spray pumps and motors Shutdown cooling heat exchangers Refueling water storage tank Piping and supports Valves and valve operators Containment sump screen ntainment Air Recirculation and Cooling Fans and motors stem Cooling Coils Housing closure Building Filtration System and Fans and motors ergency Spent Fuel Pool Cleanup Filters and housing Electric heaters Piping, ductwork and supports Dampers and damper operators drogen Control System Hydrogen recombiners PIR fans and motors Piping and supports Hydrogen purge valves and valve operators Hydrogen monitoring system 1.4-9 Rev. 35

System Components ntrol Room Air Conditioning System Fans and motors cluding the control room filtration system) Direct expansion and condenser coils Housings Compressor CRFS Filters Ductwork and supports Dampers and damper operators Refrigeration piping and supports Refrigerant valves and valve operators Temperature control system Control Panels gineered Safety Feature Room Air Fans and motors circulation System Cooling coils Ductwork and supports Dampers and damper operators esel Generator Ventilation System Fans and motors Ductwork and supports Dampers al Switchgear Ventilation System Fans and Motors Cooling Coils Chillers and control panels Pumps and motors Piping; valves and supports Ductwork and supports Dampers and Damper Operators ntainment Isolation System Piping and sleeves Valves and valve operators 1.4-10 Rev. 35

System Components ctrical Power Supply System Station batteries, racks and chargers 125 VDC Switchgear DC/AC Inverters Vital AC and DC distribution panels 4160 Volt Emergency Switchgear 480 Volt Emergency Load Centers 480 Volt Emergency Motor Control Centers ctrical Distribution System Vital tray system and supports Vital underground duct banks Penetration assemblies actor Coolant System Reactor vessel and internals Control element assemblies and drives Pressurizer Reactor coolant pumps and motors Reactor coolant piping Pressurizer surge line and supports Pressurizer safety and relief valves Steam generators Vent, sampling and drain piping, supports and valves up to and including second isolation valve Quench tank

• Pressurizer safety and relief valves piping and supports to quench tank

• Reactor coolant pump supports 1.4-11 Rev. 35

System Components emical and Volume Control System Boric acid storage tanks Boric acid pumps and drivers Boric acid piping supports and valves Charging pumps and drivers Charging line piping, supports, valves and pulsation dampeners Letdown line piping, supports and valves up to and including second isolation valve Regenerative Heat exchanger Letdown heat exchanger

• Letdown line piping, supports, and valves downstream of reactor coolant system isolation valves

• Letdown filters

• Ion exchangers

• Volume control tank

• ent Fuel Pool Cooling System Piping, supports and valves between spent fuel pool and shutdown heat exchangers Spent fuel pool cooling pumps Spent fuel pool heat exchangers Spent fuel pool cooling pump drivers

• Piping, supports, and valves associated with normal spent fuel cooling (up to and including pipe support beyond isolation valve on branch lines)

• seous Waste Processing System Waste gas decay tanks

• Waste gas compressors

• Waste gas filter

• High pressure (150 psig) service piping, supports, and valves

• 1.4-12 Rev. 35

System Components el and Reactor Component Handling Containment polar crane uipment Spent fuel cask crane Spent fuel platform crane

• Refueling machine

• Fuel transfer machine

• Fuel tilting mechanisms

• Fuel transfer tube and isolation valve New and spent fuel storage racks New fuel elevator

• Spent fuel inspection machine

• CCW System RBCCW Pumps and Motors RBCCW Heat Exchangers RBCCW Surge Tank Piping and Supports Expansion Joints Valves and Valve Operators vice Water System Pumps and Drivers Piping and Supports Valves and Valve Operators Service Water Strainers ergency Diesel Generators Diesel Oil System Air Intake and Exhaust Piping Control Panels Diesel Oil Supply Tanks Piping, Valves and Supports be Oil System Pumps and motors Coolers Piping and supports Heaters Piping and supports 1.4-13 Rev. 35

System Components ket Water Cooling System Pumps and motors Coolers Piping and supports Heaters Jacket water expansion tank Valves and valve operators esignated seismic Class II components but designed for Class I earthquake basis.

r Cooling System Pumps Coolers Piping and supports Valve and valve operators rting Air System AC and DC Motor Driven Compressors Starting Air tanks Piping and supports upstream of check valves xiliary Feedwater System Auxiliary. feedwater pumps and drivers Condensate storage tank Piping and supports Valves and valve operators in Steam System Main steam safety relief valves pstream of isolation valves Atmospheric dump valves Main Steam isolation valves Piping and supports Valves and valve operators gineered Safety Actuation System, Status nel actor Protection System smic Measurement Instrumentation in Control Boards in Steam Isolation Panel 1.4-14 Rev. 35

System Components t Shutdown Control Boards ric Acid Heat Tracing Panels diation Monitoring System esignated seismic Class II components but designed for Class I earthquake basis.

1.4-15 Rev. 35

1 GENERAL design of Millstone Unit 2 is based upon concepts which have been successfully applied in design of other pressurized water reactor power plants. However, certain programs of retical analysis or experimentation (constituting research and development as defined in the mic Energy Act, as amended, and in Nuclear Regulatory Commission (NRC) Regulations) e been undertaken to aid in plant design and to verify the performance characteristics of the t components and systems. This section describes the results and status of these analytical and programs, including experimental production and testing of models, devices, equipment and erials at time of application for an operating license.

mbustion Engineering (CE), Inc., which conducted these programs, had taken into sideration information derived from research and development activities of the NRC and other anizations in the nuclear industry.

CE research and development programs required to justify the design to Millstone Unit 2 were pleted and all test results were factored into design of the plant.

2 FUEL ASSEMBLY FLOW MIXING TESTS 966, a series of single-phase tests on coolant turbulent mixing was run on a prototype fuel mbly which was geometrically similar to the Palisades assembly. The model enabled rmination of flow resistance and vertical subchannel flow rates using pressure rumentation and the average level of eddy flow using dye-injection and sampling equipment.

tests yielded the value of the inverse Peclet number characteristic of eddy flow (0.00366).

ing the course of the tests the value was shown to be insensitive to coolant temperature and to ical coolant mass velocity. The design value of the inverse Peclet Number was established as 35 on the basis of the experimental results.

part of a CE sponsored research and development program, a new series of single-phase dye ction mixing tests were conducted in 1968. The tests were performed on a model of a portion ontrol element assembly (CEA) type fuel assembly which was sufficiently instrumented to ble measurement, via a data reduction computer program, of the individual lateral flows across boundaries of 12 subchannels of the model. Although these tests were not intended for that pose, some of the test results could be used to determine the average level of turbulent mixing he reference design assembly. The inverse Peclet Number calculated from the average of 56 vidual turbulent missing flows (two for each subchannel boundary) obtained from the licable data was 0.0034. With respect to general turbulent mixing, therefore, the more recent y on the CEA verifies the constancy of the inverse Peclet number for moderately different assembly geometries and confirms the design value of that characteristic.

1.5-1 Rev. 35

eries of tests was completed on both single and dual CEAs in a cold water, low pressure lity to satisfy the following objectives:

a. Determine the mechanical and functional feasibility of the CEA type control rod concept.

b. Experimentally determine the relationship between CEA drop time and CEA drop weight, annular clearance between CEA fingers and guide tubes, and coolant flow rate within the guide tube.

c. Experimentally determine the relationship between flow rate and pressure drop within the guide tube as a function of CEA axial position and of finger-to-guide-tube clearance.

d. Determine the effects on drop time of adding a flow restriction or of plugging the flow holes in the lower portion of a guide tube (as might occur under accident conditions).

e. Determine the effects of misalignment within the CEA guide tube system on drop time.

results of these tests were used as the basis for selecting the final CEA and guide tube metrics. The tests have demonstrated that the five-finger CEA concept is mechanically and ctionally feasible and that the CEA design has met the criteria established for drop time under most adverse conditions. The testing has also verified that the analytical model used for dicting the drop times gives uniformly conservative results.

effects on drop time of all possible combinations of frictional restraining forces in the control ment drive mechanism (CEDM), angular and radial misalignment of the CEDM, bowing of the de tubes, and misalignments of the CEA should have been experimentally investigated and ned. The conditions tested simulated all the effects of tolerance buildup, dynamic loadings, thermal effects. The tests demonstrated that misalignments and distortions in excess of those ected from tolerance buildup or any other anticipated cause would still result in acceptable p times.

4 CONTROL ELEMENT DRIVE ASSEMBLY PERFORMANCE TESTS accelerated life test of a magnetic jack coupled to a CEA was completed. This test consisted of tinuous operation of the mechanism for a total accumulated travel of 32,500 feet at conditions ilar to those it will encounter when installed on the operating reactor. The mechanism was rated at a speed of 40 inches per minute. without malfunction or adjustments. In addition, 200 height drops were completed with all drop times less than 2.5 seconds for 90 percent rtion. Subsequent testing at various conditions was conducted to determine maintenance les.

1.5-2 Rev. 35

eference 1.5-2, a magnetic jack type CEDM, similar to that installed at Unit 2 was verified to capable of withstanding a complete loss of air cooling for a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period with the plant at mal operating temperature and pressure (600°F and 2250 psi) without damage to the CEDM holding the CEA. In addition, the coils stacks were later subjected to a steam environment for minutes without affecting their electrical capabilities.

design of the CEDM is such that loss of CEDM cooling will not prevent the CEDM from asing the CEA. The ability of the CEDM to release the rods is not dependent on the cooling provided by the CEDM cooling system. Cooling function is only to ensure reliability of the DM coil stack.

5 FUEL ASSEMBLY FLOW TESTS ocity and static pressure measurements were made in an oversized model of a fuel assembly to rmine the flow distributions present. Effects of the distributions on thermal behavior and gin are to be evaluated, where necessary, with the use of a CE version of the COBRA thermal hydraulic code (Reference 1.5-1). Subjects investigated include the following:

a. Assembly inlet flow distribution as affected by the core support plate and bottom header plate flow hole geometry: Flow distribution was measured and results indicate that uniform nominal value is achieved within 10 percent of core height.

The normal inlet flow distribution arising from the geometric configuration of the core support plate and lower end fitting of the fuel assembly was shown to have an effect on thermal margin which was small enough so that no allowance had to be made in the context of CE current conservative thermal-hydraulic calculational techniques.

b. Assembly inlet flow distribution as affected by a blocked core support plate flow hole: Flow distribution was measured and indicated that flow was recovered to at least 50 percent of the uniform nominal value at an elevation corresponding to 10 percent of core height. Analysis of several of the flow maldistributions arising from the unlikely blockage of a flow hole in the core support plate or from the blockage of one to nine subchannels indicated that flow recovery is rapid enough downstream of the obstruction so that the complete blockage of a core support-plate flow hole or of a single subchannel during 120 percent of full power operation would not result in a W-3 departure from boiling ratio (DNBR) of less than 1.0. The experimental data also indicated that the upstream influence of a subchannel blockage diminished very rapidly in that direction.

c. Flow distribution within the assembly as affected by complete blockage of one to nine subchannels: The flow distributions were measured and indicated very little upstream effect on such blockage, followed by recovery to normal subchannel flow conditions within 10 to 15 percent of core height, depending upon the number of subchannels blocked.

1.5-3 Rev. 35

Measurements of the flow distribution near the top of the active core demonstrated that there was a negligible effect of the fuel assembly end fitting, alignment plate, and CEA shroud on that distribution.

6 REACTOR VESSEL FLOW TESTS ts were conducted with one-fifth scale models of CE reactors to determine hydraulic ormance. The first tests were performed for the Palisades plant which has a reactor coolant em (RCS) similar to that of Millstone Unit 2. The tests investigated flow distribution, pressure p and the tracing of flow paths within the vessel for all four pumps operating and various part-p configurations. Air was used as the test medium. CE has also conducted tests on a one-fourth e model of the Fort Calhoun reactor using air as the test medium.

ilar one-fifth scale model tests have been performed for Maine Yankee, which has a core ilar to that of Millstone Unit 2. These tests were conducted in a cold water loop. All ponents for the model were geometrically similar to those in the reactor except for the core re 217 cylindrical core tubes were substituted for the fuel bundles. The core tubes contained ices to provide the proper axial flow resistance.

w characteristics for Millstone Unit 2 were determined by taking into consideration ilarities between Millstone Unit 2 and other CE reactors in conjunction with the experimental from the flow model programs.

7 IN-CORE INSTRUMENTATION TESTS ts on in-core thermocouples and flux detectors were performed to ensure that the rumentation will perform as expected at the temperatures to be encountered and that it does vibrate excessively and cause excessive wear or fretting. Cold flow testing has been pleted on a similar detector cable; no adverse vibrations or wear effects were encountered.

flow testing is also complete. After 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> at 590°F and 2,100 psig in a test loop, no ch of mechanical integrity was observed.

chanical tests of the insertion and removal equipment and instrumentation were performed on bles of the same approximate configuration as those used on Millstone Unit 2. The top entry ore instrumentation design provides a means of eliminating the need of handling instrument mblies separately, thus, minimizing downtime and personnel exposure. A full-scale mockup built to accommodate three in-core instrumentation thimble assemblies. Major components subassemblies of the mockup included:

a. An in-core instrumentation test assembly, including the upper guide structure support plate, three thimble guide sleeves, fuel alignment plate, three fuel bundle guide tubes, and the core support plate.

1.5-4 Rev. 35

c. An upper guide tube with the guide tube attached to the thimble extension in and the detector cable partially inserted in the guide tube.

rtion and withdrawal tests were performed to determine the frictional forces of a multi-tube rument thimble assembly during insertion and withdrawal from a set of fuel bundles. This test ulated the operation that will be performed during the refueling of the reactor. To determine ther jamming of the thimbles would occur during this operation, bending loads were applied he thimble assembly by tilting the instrument plate in 0.5 degree increments up to a total of degrees from horizontal. Guide tubes were filled with water. The assembly was raised and ered approximately five times for each tilt setting. Results showed no discernible difference in friction forces for the various tilt settings. The tests demonstrated that the repeated insertion withdrawal of in-core instrumentation thimble assemblies into the fuel bundle guides can be omplished with reasonable insertion forces.

cycle tests were performed to determine if the frictional forces increase as a result of 40 rtions and withdrawals. An automatic timer was installed in the crane electrical circuitry to matically cycle the thimble assembly between the fully inserted and withdrawn position. The rument plate was set for five degrees tilt and the assembly was cycled 60 times. The insertion withdrawal forces were measured during the first and last five cycles. No discernible erence was noticed.

off-center lift test was performed to determine if the thimble assembly could be withdrawn m the core region while lifting the assembly from an extreme off center position. For a lifting nt 11 inches off center, insertion was accomplished without incident. The flexibility of the ble is such that jamming of the assembly due to off-center lifting does not occur.

le insertion tests were performed to determine the forces required to completely insert and hdraw a detector cable from the in-core instrumentation thimble assembly. The guide tube ing included typical bends equal to, or worse than, those found in the reactor. The detector le was passed through the guide tubing and into a thimble. In all cases, the insertion and hdrawal forces were reasonable for hand insertion.

8 MATERIALS IRRADIATION SURVEILLANCE veillance specimens of the reactor vessel shell section material are installed on the inside wall he vessel to monitor the change in fracture toughness properties of the material during the tor operating lifetime. Details of the program are given in Section 4.6.

9 REFERENCES 1 Rowe, D. S., Cross-Flow Mixing Between Parallel Flow Channels During Boiling.

COBRA Computer Program for Coolant Boiling in Rod Arrays, Part 1, BNWL-371, March 1967.

1.5-5 Rev. 35

1.5-6 Rev. 35 ginally, The Connecticut Light and Power Company (CL&P), the Hartford Electric Light mpany (HELCO), and Western Massachusetts Electric Company (WMECO) (the Owners),

Northeast Nuclear Energy Company (NNECO) were the applicants for the operating license Millstone Unit 2. At that time NNECO acted as the agent for the owners and was responsible the design, construction and operation of the plant. However, in 2001, the operating license transferred to Dominion Nuclear Connecticut, Inc., at which time they became the sole owner operator of Millstone Unit Number 2.

mbustion Engineering (CE), Inc. was engaged to design, manufacture and deliver the Nuclear m Supply System (NSSS) and nuclear fuel for the first core and the first two core reload hes to the site. The NSSS includes the reactor coolant system, reactor auxiliary system ponents, nuclear and certain process instrumentation, and the reactor control and protective em. In addition, CE furnished technical assistance for erection, initial fuel loading, testing and al startup of the NSSS.

htel Corporation was engaged as the Engineer-Constructor for this project and as such ormed engineering and design work for the balance-of-plant equipment, systems and ctures not included under the CE scope of supply. Bechtel was engaged to perform onsite struction of the entire plant with technical advice for installation of the reactor components vided by CE.

reactor vessel closure head was replaced during refueling outage 16 with a new head mbly fabricated from materials that are less susceptible to Primary Water Stress Corrosion cking (PWSCC). The new head assembly was manufactured by Mitsubishi Heavy Industries.

tinghouse/CE was engaged in the design, installation and testing of the head.

pressurizer assembly was replaced in 2006 with a new assembly fabricated from materials are less susceptible to PWSCC. AREVA was engaged in the design, fabrication, installation testing of the replacement pressurizer.

1 REFERENCES 1 Millstone Unit 3, Final Safety Analysis Report, Section 13.1 - Organizational Structure.

1.6-1 Rev. 35

1 GENERAL ce the issuing of the Preliminary Safety Analysis Report (PSAR), a number of changes were e in the design of Millstone Unit 2. These changes improved the operating characteristics and ance plant safety and reliability. The following reflects changes made up to the time of rating license application.

2 CONTROL ELEMENT DRIVE MECHANISMS gnetic jack drive mechanisms are provided for positioning the control element assemblies A) instead of rack and pinion drive mechanisms. The magnetic jack control element drive hanism (CEDM) is completely sealed by a pressure boundary, eliminating the need for seals.

tion of the control element drive shaft is accomplished by sequencing five solenoid coils ted around the pressure boundary.

mbustion Engineering (CE), Inc., supplied identical CEDMs on previous plants, including ne Yankee (Atomic Energy Commission (AEC) Docket Number 50-309) and Calvert Cliffs ts 1 and 2 (AEC Docket Number 50-317 and 50-318).

3 RADIOACTIVE WASTE PROCESSING SYSTEM 3.1 Clean Liquid Waste Processing System losed drains system and a 700 gallon equipment drain sump tank were included in the system ollect liquids containing dissolved hydrogen and fission gases from equipment drains, valve leakoffs, and relief valve discharges. The liquid wastes are collected in this tank via the ed drains system. This tank was provided to minimize the release of radioactive gases to the osphere without prior processing by the gaseous waste system.

flash tank was replaced by a packed column-type degasifier utilizing internally generated pping steam. The degasifier has a better decontamination factor for xenon and krypton than ld have been possible with the proposed flash tank.

nt space and the necessary piping and valves were provided for incorporating two additional ineralizers into the system, if required, based on operating experience.

3.2 Gaseous Waste Processing System r additional waste gas decay tanks were added to the system to allow for a minimum of 60 day ay of all hydrogenated waste gases, including cover gases, collected by the system prior to ase to the atmosphere through the Millstone stack.

1.7-1 Rev. 35

vital components closed cooling water system was deleted and the components cooled as ows:

Component Cooling System rvice air compressors and instrument air Turbine building closed cooling water mpressors (interconnecting piping provided to reactor building closed cooling water) xiliary feedwater pump turbine oil cooler Water being pumped esel generator Service water ntrol room air conditioners Air 5 ELECTRICAL 5.1 AC Power station service transformers supply power at 6900V and 4160V via their respective station ice busses for large motor loads. Further, the 4160V supplies power to the 480V unit station transformers for smaller loads.

preserve redundancy and separation, each motor control center is fed from only one 480 volt center rather than from two.

5.2 Diesel Generators the change in the diesel engine cooling water supply, see Section 1.7.4.

itional conditions under which the diesel generators will start automatically are noted in tion 8.3.3.1.

5.3 DC Supply hird station battery was added to care for the non safety-related 125 volt DC loads associated h the turbine generator.

h 125 volt DC distribution panel formerly had a feeder from each of the two station batteries, h diodes to prevent tying the battery buses together. To maintain the independence of undant sources, the diodes were removed and the DC distribution panels fed from redundant ery buses.

1.7-2 Rev. 35

o 120 volt regulated AC instrument buses were provided (instead of one) to assure redundant er sources for vital instrumentation.

6 AXIAL XENON OSCILLATION PROTECTION omatic initiation of an appropriate protection system for axial xenon oscillation was rporated into the reactor protective system. This addition provided compliance with the Cs General Design Criterion 20 as published February 20, 1971, in the Federal Register and interpreted for preceeding reactors of similar design (see Calvert Cliffs Units 1 & 2 endment 15, Question 3.14). The basis for this addition was to provide an automatic protective kup to the operator in the unlikely event he should fail to adjust the full length CEA as uired late in core life when axial xenon oscillations may become divergent.

7 NUMBER OF CONTROL ELEMENT ASSEMBLIES AND DRIVE MECHANISMS number of CEAs in the Millstone Unit 2 reactor is 73, compared to 85 CEAs shown in the R design. The number of drive mechanisms was changed from 65 in the PSAR to 69 for le 1. Then, removal of 8 part-length CEAs in 1978 reduced the number of drive mechanisms

1. This resulted in a net increase in the number of single CEAs (37 to 49) and a net reduction he number of dual CEAs (40 to 24), thereby providing greater flexibility for optimization of A programming and fuel management.

8 BURNABLE POISON SHIMS nable poison shims were added to the fuel assemblies in Cycle 1, replacing some fuel. These ms permitted lowering of the initial boric acid concentration in the coolant. This provided itional assurance that the moderator temperature coefficient, at power at beginning of life, ld not be positive.

9 STRUCTURES following changes have been made:

a. The post-tensioning tendons were encased in galvanized rather than ungalvanized semi-rigid sheaths.

b. The bearing plate material was changed from A-36 to VNT steel.

c. The warehouse area and turbine building were designated Class I structures.

d. All concrete reinforcing steel larger than number 11 was mechanically spliced.

e. Dye penetrant and magnetic particle inspection were not used for liner plate weld quality control.

1.7-3 Rev. 35

h Pressure Safety Injection (HPSI) pump P-41B (Figure 6.1-1) (Sheet 2) was connected to h of the two suction headers but is normally isolated by valving. This HPSI pump served as a e and was aligned, process wise and electrically, for operation only when either of the other HPSI pumps is taken out of service. Two operable HPSI pumps satisfy redundancy uirements for core cooling.

11 CONTAINMENT PURGE VALVE ISOLATION ACTUATION SYSTEM tainment Purge Valve Actuation System was changed from two-out-of-four to one-out-of-four

c. See Sections 7.3.2.3 and 7.5.6.3 for details.

12 CONTROL ELEMENT DRIVE SYSTEM Control Element Drive System (CEDS) was modified to include a CEA Motion Inhibit ure which acts to help the operator assure that limits on CEA position are not exceeded. The DS is described in Section 7.4.2.

1.7-4 Rev. 35

THE ACRS AT THE TIME OF INITIAL LICENSING AND WAS NOT INTENDED TO BE UPDATED.]

1 GENERAL s section describes the status of programs initiated to investigate the items which were tified by the Advisory Committee on Reactor Safeguards (ACRS) as being of special interest pertaining to all large water-cooled power reactors up to the time of application for an rating license.

arrying out these programs, information derived from research and development activities of Atomic Energy Commission (AEC) and other organizations in the nuclear power industry e considered.

1.1 Ability of Fuel to Withstand Transients at End of Life and Experimental Verification of Maximum Linear Heat Generation Rate fuel cladding was designed to limit the transient stresses to two-thirds of the unirradiated e of the yield stress even during a depressurization transient near the end of life, when the rnal gas pressure is highest.

erimental verification of the maximum linear heat generation rate employed in the Millstone t 2 design was discussed in the original FSAR submitted at the time of application for an rating License. Numerous irradiation tests, which bracket the design of these units, were ormed, including those in the Westinghouse test reactor, the Shippingport blanket irradiations, mixed oxide irradiations in the Saxton reactor, the zirconium clad UO2 fuel rod evaluations in Vallecitos boiling water reactor, the large speed blanket reactor rod irradiations, the center ting irradiations in Big Rock, Peach Bottom 2 irradiations, and NRX irradiations CL-Canada). In these tests, fuel rods similar to those employed in the design of the Millstone t 2 core were successfully irradiated to fuel burnups varying from very short term tests up to 000 MWD/MTU and at linear heat rates ranging from 5.6 up to 27.0 kW/ft.

1.2 Fuel Integrity Following a Loss-of-Coolant Accident ACRS had asked that information be developed to show that the ...melting and subsequent ntegration of a portion of fuel assembly...will not lead to unacceptable conditions. They rred specifically to the ...effects in terms of fission product release, local high pressure duction, and the possible initiation of failure in adjacent fuel elements....

uiry was made as to whether accident conditions that might occur which cause clad peratures to reach such high temperatures that embrittlement occurs, and whether subsequent nching operations will cause the embrittled portions to disintegrate and thereby prevent a icient flow of emergency core coolant to the remainder of the core.

1.8-1 Rev. 35

racteristics of the UO2 core and by the provision of engineered safety features (ESF).

h regard to the nonexcursion mechanisms leading to the conditions described by ACRS, the owing two conditions might be conjectured:

Fuel bundle inlet flow blockage during full power operation and subsequent overheating of the coolant-starved fuel, or loss of reactor coolant.

dition A, inlet flow blockage during full-power operation and subsequent overheating and ting of the fuel, is not considered possible because open (nonshrouded) fuel bundles are used, eby providing cross-flow to the flow-starved channel even if some of the inlet holes were ked. Details and conclusions of the tests performed at Combustion Engineering (CE), Inc. on influence of inlet geometry on flow in the entrance region are presented in ASME paper WA/HT-34 delivered at the December 1968 Winter Annual Meeting. Further analysis of these s showed that if a group of four flow holes in the core support plate at the base of the fuel dle were blocked, the subchannels above the blocked region would have an inlet velocity ut 21 percent of the core average bulk inlet velocity. Because of crossflow from the ounding nonblocked regions, the net effect of this flow shortage, using conservative ulations, is to increase the enthalpy rise of the blocked region by a maximum of 35 percent. At inal conditions, the hot channel departure from nucleate boiling ratio (DNBR) would drop m 2.0 to 1.4, assuming that the blockage occurred directly below the design hot channel.

dition B was covered comprehensively in the Statement of Affirmative Testimony and dence of Combustion Engineering in the Matter of Rulemaking Hearing for the Acceptance eria for Emergency Core Cooling System for Light-Water-Cooled Nuclear Power Reactors, ket Number RM-50-1. The emergency core cooling system (ECCS) is designed to remove the ay heat from the core for the necessary period of time following a loss-of-coolant accident CA). Core power distributions and LOCA temperature-time histories indicate that for peak temperatures below 2300°F, the total clad oxidation will be significantly less than 1 percent.

1.3 Primary System Quality Assurance and In-Service Inspectability omprehensive quality assurance program has been established to assure that Millstone Unit 2 esigned, fabricated, and constructed in accordance with the requirements of applicable cifications and codes. The program started with the initial plant design and has continued ugh all phases of equipment procurement, fabrication, erection, construction, and plant ration. The program provides for review of specifications to assure that quality control uirements are included and for surveillance and audits of the manufacturing and construction rts to assure that the specified requirements are met.

ummary description of the Quality Assurance Program (QAP) is included as Section 12.8.

s program fully meets the guidelines established in the former AEC Regulation 10 CFR Part Appendix B entitled Quality Assurance Criteria for Nuclear Power Plants. The quality 1.8-2 Rev. 35

eline inspection and subsequently in-service inspections are performed and are further ussed in Section 4.6.6.

1.4 Separation of Control and Protective Instrumentation addition to any redundancy and separation provided for control or for protective rumentation, the control and protective instrumentation are independent of each other. Control on and protective action derived from the same process variable are generated by separate rumentation loops. Malfunction of a single control instrumentation loop cannot impair the ration of the protective instrumentation loop and conversely malfunction of the protective rumentation loop does not affect operation of the control loop. The instrumentation for a le protective and a single control channel may be located adjacent to one another, and their uits may be routed in the same cable tray, but each is capable of performing its function pendently of the other. Further discussion is provided in Chapters 7 and 8.

1.5 Instrumentation for Detection of Failed Fuel ly detection of the gross failure of fuel elements permits early applications of action necessary mit the consequences.

ed on a study of the expected fission and corrosion product activities in the reactor coolant, it concluded that the gross gamma plus specific isotope monitor provides a simple and reliable ns for early detection fuel failures.

design bases of the detection system include the following:

a. Trends in fission product activity in the reactor coolant system (RCS) (specifically Rb-88) are used as an indication of fuel element cladding failures.

b. There is a time delay of less than five minutes before the activity, emitted from a fuel element cladding failure, is indicated by the instrumentation. This time delay is a function of the location of the monitor.

c. The information obtained from this system will not be used for automatic protective or control functions or detection of the specific fuel assembly (or assemblies) which has failed.

d. The high activity alarm will be supplemented with radiochemical analysis of the reactor coolant for fission products to provide positive identification of a fuel element failure.

location and operation of the detector, designated as a process radiation monitor, is described ections 7.5.6.3 and 9.2.2.

1.8-3 Rev. 35

1.6 Effects of Blowdown Forces on Core and Primary System Components dynamic response of reactor internals resulting from hydrodynamic blowdown forces under a tulated LOCA condition was the subject of a CE topical report which contained a complete cription of the theoretical basis for methods of analysis for the various reactor components, as l as documentation of computer programs and the respective analytical structural models.

ctor vessel internal structures were analyzed to ensure the required structural integrity during ormal operating conditions, including the effects of blowdown, pressure drop and buckling es. For the LOCA, the CEFLASH-4 computer program was used to define the flow transient the WATERHAMMER program determines the corresponding dynamic pressure load ribution. The dynamic response of the reactor vessel internals to the space and time-dependent sure loads were obtained through the use of a number of structural dynamic analysis codes.

eral and vertical dynamic response of the internals were considered, as well as the transient onse and dynamic buckling of a core support barrel in shell modes. Both the CEFLASH-4 and TERHAMMER models were evaluated against the LOFT program results.

loads resulting from the LOCA condition were added to the loads resulting from normal ration and the design basis earthquake (DBE) for each critical component and the component ections and stresses analyzed to ensure compliance with the criteria specified in Section 4.2.

1.7 Reactor Vessel Thermal Shock ficient emergency core cooling water is available to flood the core region in the event of a or LOCA. The Millstone Unit 2 design uses a section of each of the RCS cold legs to conduct water from the safety injection nozzles to the reactor vessel. This water then flows into the ncomer annulus and into the lower plenum of the reactor vessel before flooding the core lf. Analytical investigations were performed to provide assurance that the resultant cooling of irradiated inner surface of the thick-walled reactor vessel will not induce or propagate cracks icient to cause the reactor vessel to fail.

analytical evaluation of pressurized thermal shock effects in CEs NSSS was issued by CE in ember 1981 (CEN-189). The limiting case is a small break LOCA with the assumption of current loss of all feedwater. For Millstone Unit 2, it was found that crack initiation would not ur during this limiting transient throughout the unit's design life (32 EFPY).

sequently, the Pressurized Thermal Shock Rule (10 CFR 50.61, 1986) was used for rittlement shift prediction. The results confirmed that the reactor vessel was fully able to hstand a postulated pressurized thermal shock imposed by the ECCS through the unit's design 1.8-4 Rev. 35

conducted experimental and analytical investigations of fuel-rod failures under simulated CA conditions. The analytical work provided indications of the actual conditions to be ected in the core during a transient, in terms of potential clad heating rates, internal pressures transient duration. The experimental work applied these parameters in various combinations stablish the nature of fuel-rod deformation which might occur under accident conditions. This ject was covered comprehensively in the Statement of Affirmative Testimony and Evidence of mbustion Engineering in the Matter of Rulemaking Hearing for the Acceptance Criteria for ergency Core Cooling Systems for Light-Water-Cooled Nuclear Power Reactors, Docket mber RM-50-1.

1.9 Preoperational Vibration Monitoring Program reoperational vibration monitoring program (PVMP) was completed for the Palisades reactor rnals. Results of this program were submitted to the AEC by CE Report CENPD-36.

itional PVMPs were developed for both the Maine Yankee and Fort Calhoun reactor internals.

eeping with the requirements for prototype vibration test programs, predictions of hydraulic ing functions and structural response were made for the Maine Yankee and Fort Calhoun tor internals and correlated to test program measurements. Vibration test data from all three tors was used in demonstrating the adequacy of the Millstone Unit 2 reactor vessel internals sustain flow-induced vibration effects. The vibration test data available, together with ropriate analyses, permitted the assessment of design or fabrication differences existing ng the subject reactors as they related to the vibrational response characteristics of the lstone Unit 2 reactor internals. A comparison of applicable design parameters for the sades, Fort Calhoun, Maine Yankee and Millstone Unit 2 reactors as of the time of application operating license is presented in Table 1.8-1.

analytical methods which formed the basis for the CE vibration response predictions were vided in the Maine Yankee and Fort Calhoun vibration monitoring programs submittals.

sades, Maine Yankee and Fort Calhoun Flow Model Test reports and a description of the hodology utilized to relate these data to in-reactor forcing functions were provided, as well as scription of the structural response computer code.

1.9.1 Basis of Program suitability of using PVMP data from Palisades, Omaha and Maine Yankee as a composite otype was based on the following:

a. Reactor internals structural response and LOCA hydraulic loadings could be adequately predicted with computer programs available, and the methods and procedures will be provided and justified.

b. The hydraulic forcing function predicting method was provided and justified. The forcing function method was verified by measurements in the prototype(s).

1.8-5 Rev. 35

Safety Guide 20).

The prediction methods and procedures were used to predict the responses (amplitude and frequency) for the Fort Calhoun PVMP.

d. The Maine Yankee and Fort Calhoun PVMP results were satisfactory, satisfying AEC licensing requirements for all CE reactor plants which had either construction or operating permits, providing the configuration and flow modes were similar as specified in Regulatory Guide 1.20 (formerly Safety Guide 20).

e. CE provided predictive methodology and predicted and limiting values of response (acceptance criteria) on the Maine Yankee program. The program results were provided on a timely basis in accordance with the Regulatory Guide 1.20 (formerly Safety Guide 20).

f. CE submitted a report on the LOCA dynamic analysis methods and procedures.

1.9.2 Millstone Unit 2 Program PVMP to be conducted for Millstone Unit 2 reactor internals was consistent with those ions of the former Safety Guide 20 (after replaced by Regulatory Guide 1.20), which ressed nonprototype reactors.

following was the PVMP plan for Millstone Unit 2. As noted above, this program was tingent upon the results to be obtained from Maine Yankee and Fort Calhoun PVMP.

1. The reactor internals important to safety were be subjected during the preoperation functional testing program to all significant flow modes of normal reactor operation and under the same test conditions conducted on the Palisades, Fort Calhoun, and Maine Yankee designs.

The test duration was at least as long as that conducted on the Palisades, Fort Calhoun and Maine Yankee designs.

2. Following completion of the preoperational functional tests, the reactor internals were removed from the reactor vessel and visual and nondestructive examination of the reactor internals was conducted. The areas examined included:

a. All major load bearing elements of the reactor internals relied upon to retain the core structure in place;

b. The lateral, vertical, and torsional restraints provided within the vessel; 1.8-6 Rev. 35

d. Those other locations on the reactor internal components which were examined on the Palisades, Fort Calhoun, and Maine Yankee designs;

e. The interior of the reactor vessel for evidence of loose parts or foreign material.

ummary of the PVMP inspections described above was submitted after the completion of the ection and tests in a report.

hould be pointed out that the reactor thermal shield was removed from the lower internals mbly because of the damage suffered due to excessive vibratory movement. An evaluation performed to assess the effects of thermal shield removal on the vibratory response of the rest eactor internals. It was concluded that the effect would be minimal and that the conclusions of PVMP were still valid.

2 SPECIAL FOR MILLSTONE UNIT 2 2.1 Release of Radioactivity in Case of Damaged Fuel Assemblies in Spent Fuel Pool he event of release or radioactivity resulting from damaged fuel in the spent fuel pool, the iliary exhaust system (AES) which is described in Section 9.9.8, diverts the effluent through enclosure building filtration system (EBFS) charcoal filters prior to release through the lstone stack. The AES maintains the fuel handling area under a negative pressure to limit ontrolled release of radioactivity.

2.2 Hydrogen Control independent systems in the hydrogen control systems monitor and mix hydrogen in the tainment following a LOCA (see Section 6.6). Each is a full capacity, completely redundant, pendent system. Air to operate the hydrogen monitoring system CIVs is provided by the rument air system with a backup air bottle system that is designed to meet single failure eria. Two, full capacity hydrogen purge systems not credited in accident analyses are provided.

hydrogen recombiner system has no mitigating function.

2.3 Common Mode Failures and Anticipated Transients Without Scram analyzed the response of pressurized water reactors which are typical of Millstone Unit 2 to onstrate the diversity of the reactor protective system in mitigating common mode failures the response of the plant to anticipated transients without scram (ATWS). Results of these ies were submitted to the AEC as topical reports.

Report CENPD-11, entitled Reactor Protection System Diversity was submitted on March 971. This report evaluated systematic, nonrandom, concurrent failures, (i.e., common mode 1.8-7 Rev. 35

nnels which measure a given process parameter, the report, nevertheless, addresses this type of ure. Monitoring of the condition by diverse means or principles enables a protection system to hstand common mode failures. The evaluations included the following accidents: control ment assembly (CEA) withdrawal, CEA drop, loss of reactor coolant flow, excess load, loss of and loss of feedwater. The results of the study demonstrated that the diversity of the reactor ective system is such that gross fuel damage or consequential failures in the RCS or in the n steam system will not occur for any of the accidents analyzed.

raft of the CE report, entitled Topical Report on Anticipated Transients Without Scram prietary) was submitted to the AEC on January 10, 1972. Evaluations were performed in this ort based upon the assumption that no CEA are inserted into the core during the course of the owing transients: CEA withdrawal, CEA drop, idle loop startup, loss of flow, boron dilution, ess load, loss of load, loss of feedwater, sample line break, and pressurizer safety valve failure.

transient resulting from loss of normal onsite and offsite power was also analyzed but with a servative one percent negative reactivity insertion assumed following reactor trip signal eration, since for this case the failures which initiate the transient would also remove power m the control element drive mechanism (CEDM), allowing the CEAs to insert. The final ort, with results and their applicability to Millstone Unit 2, was submitted to the AEC.

3 REFERENCES 1 Millstone Unit 3, Final Safety Analysis Report, Section 13.1 - Organizational Structure.

1.8-8 Rev. 35

MONITORING PROGRAM DESIGN PARAMETERS

<Parameter> Palisades Fort Calhoun Maine Yankee Millstone Unit 2 Rmean, inches 75-7/8 61-5/16 75.25 75.25 Upper CSB: t, inches 2 2 2.5 2.5 Upper CSB: L, inches 109.25 101-3/8 135-5/8 141.75 Upper CSB: Rmean, inches 75-5/8 61-1/16 74-7/8 74-7/8 Middle CSB: t, inches 1.5 1.5 1.75 1.75 Middle CSB: L, inches 166.75 166-1/8 144.75 148.75 Middle CSB: Rmean, inches 75-3/8 60-11/16 74-5/8 74-5/8 Lower CSB: t, inches 2 2.25 2.25 2.25 Lower CSB: L, inches 38.5 35-5/8 38 38 Lower Cylinder ID, inches Integral Integral 141 141 Core Cylinder OD, inches Integral Integral 145 145 Support Cylinder L, inches Integral Integral 42 42 Structure Supported Integral Integral CSB Flange CSB Flange Core Shroud Support Bolted to CBS Bolted to CBS Bolted to CBS Bolted to CBS Core Shroud: Rmean, inches 73.5 59-1/16 72-5/8 72-5/8 Core Shroud: Cylinder t, inches 2 1.5 2 2 UGS: L, inches 15 24 24 24 UGS: Beams inches 18 by 1.5 24 by 1.5 24 by 1.5 24 by 1.5 UGS: Plate t, inches 3 3.25 4 4 1.8-9 Rev

<Parameter> Palisades Fort Calhoun Maine Yankee Millstone Unit 2 Thermal Shield No Yes Yes Yes Number of Loops 2 2 3 2 Design Minimum. Flow, 106 lbm/hr 125 71.7 122 139 Inlet Design Temperature, F 548 547 546 544 Inlet ID, inches (a) 35-1/8 28.75 39 35-3/16 Outlet ID, inches (a) 48-5/8 37 40 48-1/8 Inlet Pipe Velocity, ft/sec 37.7 33.7 39.2 41.6 Downcomer Velocity, ft/sec 19.6 25.2 24.9 26.7 Core Inlet Velocity, ft/sec 12.2 12.4 13.0 15.4 Outlet Pipe Velocity, ft/sec 41.4 41.3 42.6 46.5 (a)These IDs are measured at the inside wall of the reactor vessel as shown for the Millstone 2 reactor vessel in Figure 4.3-1.

CSB = Core Support Barrel UGS = Upper Guide Structure Velocity = Design Minimum Velocity 1.8-10 Rev

upport of the Final Safety Analysis Report, various topical reports prepared by Combustion ineering, Inc., and Bechtel Corporation were referenced throughout this document. A list of ical reports as of the time of application for operating license is given in Table 1.9-1.

1.9-1 Rev. 35

mbustion Engineering, Inc.

Millstone Unit 2 Title Original FSAR Section ME paper 68-WA/HT-34, December 1968 Winter Annual Meeting 1.8.1.2 tement of Affirmative Testimony and Evidence of Combustion 1.8.1.2 gineering in the matter of Rulemaking Hearing for the Acceptance 1.8.1.8 teria for Emergency Core Cooling System for Light-Water-Cooled clear Power Reactors, Docket Number RM-50-1 namic Analysis of Reactor Vessel Internals Under Loss of Coolant 1.8.1.6 cident CENPD-42-3 (Submittal to AEC in July 1972) ermal Shock Analysis of Reactor Vessels Due to Emergency Core 1.8.1.7 oling System Operation, A-68-9-1, March 15,1968, submitted as t of Amendment 9 to the Maine Yankee license application perimental Determination of Limiting Heat Transfer Coefficients 1.8.1.7 ring Quenching of Thick Steel Plates in Water, A-68-10-2, cember 13, 1968 ite Element Analysis of Structural Integrity of a Reactor Pressure 1.8.1.7 ssel During Emergency Core Cooling, A-70-19-2, January 1970 isades Precritical Vibration Monitoring Program, CENPD-36 1.8.1.9 critical Vibration Monitoring Program, CENPD-55 1.8.1.9 actor Protective System Diversity, CENPD-11, February 1971 1.8.2.3 pical Report on Anticipated Transients Without Scram, CENPD-41 1.8.2.3 THERMIC, A Computer Code for Analysis of Thermal Mixing, 3.5.3 NPD-8 SMO IV, A Thermal and Hydraulic Steady State Design Code for 3.5.3 ter Cooled Reactors, CENPD-9 smic Qualification of Category I Electric Equipment for Nuclear 7.2.6.3 am Supply Systems, CENPD-61 1.9-2 Rev. 35

chtel Corporation Millstone Unit 2 Original FSAR Title Section Consumer Power Company Palisades Nuclear Power Plant 5.2.4.5 ntainment Building Liner Plate Design Report, B-TOP-1 (submitted to C in October, 1969)

Full-Scale Buttress Test for Prestressed Nuclear Containment 5.2.3.3.3 uctures, BC-TOP-7 Testing Criteria for Integrated Leak Rate Testing of Primary 5.2.9.1 ntainment Structures for Nuclear Power Plants, BN-TOP-1 Design for Pipe Break Effects, BN-TOP-2 (REV. 1) Question 4.16 1.9-3 Rev. 35

following is a list of material incorporated by reference in the Final Safety Analysis ort (1):

1. Millstone Unit 2 Technical Requirements Manual (TRM).

2. As identified in the List of Figures, the engineering controlled plant drawings that are, coincidentally, MPS-2 FSAR Figures.

Information incorporated by reference into the Final Safety Analysis Report is subject to the update and reporting requirements of 10 CFR 50.71(e) and change controls of 10 CFR 50.59 unless separate NRC change control requirements apply (e.g., 10 CFR 50.54(a)).

1.10-1 Rev. 35

10 CFR PART 50 APPENDIX A February 20, 1971, the Atomic Energy Commission published in the Federal Register the eral Design Criteria for Nuclear Power Plants. Prior to this date, proposed General Design eria for Nuclear Power Plants as issued on July 11, 1967, in the Federal Register were in ct. Before issuance of the construction permit for Millstone Unit 2, discussions reflecting the gn intent in consideration of the 1967 proposed criteria were submitted in the PSAR. Design construction was thus initiated and has been completed based upon the 1967 proposed eria.

ce February 20, 1971, the applicants have attempted to comply with the intent of the newer eral Design Criteria to the extent possible, recognizing previous design commitments. The nt to which this has been possible is reflected in the discussions of the 1971 General Design eria which follow.

CRITERION 1 - QUALITY STANDARDS AND RECORDS Structures, systems, and components important to safety are designed, fabricated, erected and tested to quality standards commensurate with the importance of the safety functions performed. Where generally recognized codes and standards are used, they are identified and evaluated to determine their applicability, adequacy, and sufficiency and are supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program has been established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection and testing of structures, systems, and components important to safety are maintained by or under the control of the nuclear power unit licensee throughout the life of the unit.

cussion of the quality standards for those structures and components which are essential to the vention of incidents which would affect the public health and safety or to mitigation of their sequences are presented in appropriate sections of the FSAR. The quality assurance program ffect to assure that these structures, systems, and components will satisfactorily perform their ty functions is discussed in Section 12.8.

example, components of the safety injection and containment cooling systems are designed fabricated in accordance with established codes and/or standards as required to assure that r quality is in keeping with the safety function of the component. It is not intended, however, mit quality standards requirements to this list.

h Pressure Injection, Low Pressure Injection, and Containment Spray Pumps

a. Surfaces of pressure retaining materials for the high and low pressure safety injection pumps were examined by liquid penetrant techniques in accordance with 1.A-1 Rev. 35

penetrant techniques in accordance with the provisions of Draft ASME Code for Pumps and Valves for Nuclear Power, Class II, 1968. Casings for all three types of pumps have been hydrostatically tested to at least 1.5 times the design pressures.

b. Pressure containing butt welds for the safety injection pumps have been radiographed in accordance with Section VIII of the ASME Code, Paragraph UW-51.

c. The pump supplier submitted certified mill test reports of pressure containing materials.

d. At least one pump of each type has been hydraulic-performance tested for capacity and head, in accordance with the requirements of the Hydraulics Institute.

e. The pump seals have been designed to provide a high degree of assurance of their proper operation, including compatibility of seal materials with water chemistry conditions and minimum dependence on externally supplied cooling water.

f. Pump drive motors conform to NEMA Standards, MG-1.

ety Injection Tanks ME Code,Section III, Class C.

ety Injection and Containment Spray System Motor-Operated Valves and Control Valves

a. The design criteria for pressure containing parts is in accordance with ANSI B16.5.

b. Radiographic inspection of pressure containing butt welds has been performed in accordance with the requirements of ASME Code,Section VIII.

c. Certified mill test reports of pressure containing materials were provided by the supplier.

d. Pressure containing parts were hydrostatically tested in accordance with MSS-61.

e. Isolation valves are designed, fabricated, and tested in accordance with Draft ASME Code for Pumps and Valves for Nuclear Power, Class II, 1968. Control valves are designed, fabricated, and tested in accordance with ASME Code Section III, Nuclear Power Plant Components, Class II, 1971.

1.A-2 Rev. 35

a. The cooling coils are similar to a representative section of a coil which was tested under the maximum environmental conditions which would exist following a loss-of-coolant accident (LOCA). The test results demonstrated that the full size coil assembly would be capable of removing the required heat load. These data are filed with the AEC in Topical Report W-CAP-7336-L.

b. The cooling coils are tested in accordance with ASME Code,Section VIII.

c. Air moving equipment, including fan motors, were designed to standards of the Air Moving and Conditioning Association, AMCA-211A.

d. A fan and motor combination were satisfactorily tested to prove their ability to operate under the conditions which would exist within the containment following a LOCA. These data will be presented to the AEC in Topical Report W-CAP-7829.

The motor insulation and internal cable splice are filed in Topical Reports W-CAP-7343-L and W-CAP-9003, respectively.

e. Piping from the fan coolers to the containment penetrations was designed in accordance with the provisions of ANSI B31.1.0. The penetrations piping was designed to ANSI B31.7, Class II and the penetration isolation valves to the ASME Pump and Valve Code, Class II.

f. Valves, other than the penetration isolation valves, were designed in accordance with ANSI B31.1.0 and ANSI B16.5. Manually operated butterfly valves were in accordance with AWWA-C504.

tdown Heat Exchangers

a. Pressure containing materials were tested and examined per ASME Code,Section III, Class C.

b. Heat transfer design and physical design are in accordance with TEMA standards.

c. Certified mill test reports of pressure containing materials were provided by the supplier.

d. Radiographic inspection of pressure containing butt welds was performed in accordance with the requirements of ASME Code,Section III, Class C.

e. Pressure containing parts were hydrostatically tested in accordance with ASME Code,Section III, Class C.

1.A-3 Rev. 35

standards.

appropriate sections in the FSAR discuss the specific codes and standards invoked in icating or erecting the structures, systems, and components important to safety.

ropriate records of the design, fabrication, erection, and testing of structures, systems, and ponents important to safety shall be maintained for the life of the plant. (See Section 12.8).

CRITERION 2 - DESIGN BASES FOR PROTECTION AGAINST NATURAL PHENOMENA Structures, systems, and components important to safety are designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, flood, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components reflect:

(1) Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of the natural phenomena, and (3) the importance of the safety functions to be performed.

structures, systems, and components important to safety have been designed to withstand, hout loss of the capability to protect the public, the additional forces that might be imposed by ral phenomena. The most severe natural phenomena which are considered and discussed in r sections of this FSAR are as follows:

a. Earthquakes / Seismology Section 2.6

b. Wind and Tornadoes / Meteorology Section 2.3

c. Floods / Hydrology Section 2.5.4 ropriate natural phenomena are considered in the designs of structures, systems, and ponents. Accepted standards for the forces imposed by natural phenomena are used in the gn.

general description of the seismic analysis program is found in Section 5.8. Additional rmation on major structure design against the effects of natural phenomena is included in the owing sections:

Containment Structure Section 5.2 Enclosure Building Section 5.3 Auxiliary Building Section 5.4 Turbine Building Section 5.5 1.A-4 Rev. 35

Reactor Vessel Internals Appendix 3.A Reactor Coolant System Appendix 4.A CRITERION 3 - FIRE PROTECTION Structures, systems, and components important to safety are designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and heat resistant materials are used wherever practical throughout the unit, particularly in locations such as the containment and control room.

Fire detection and fighting systems of appropriate capacity and capability are provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Fire fighting systems are designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.

lstone Unit number 2 structures, systems, and components important to safety are designed located to minimize the probability and effects of fires. Fire protection systems (active and sive) have been provided to assure that all possible fires are detected, controlled, and nguished.

protection and detection systems and components are designed and installed in accordance h applicable requirements of the National Fire Protection Association (NFPA). In areas where bustible material may exist, fixed fire detection and suppression are generally provided ction 9.10).

detection and fire suppression systems of appropriate types and capacities are designed to imize the adverse effects of fires on structures, systems, and components important to safety.

ome areas, portable extinguishers are used in lieu of water suppression systems. In areas such he D.C. equipment rooms, a Halon suppression system is used in lieu of fixed water pression to assure that sensitive electronics are not affected by water spray.

fighting systems are designed to assure that their rupture or inadvertent operation does not ificantly impair the capabilities of any structure, system, or component important to safety.

reas where water may cause damage to safety equipment, such as vital electrical panels or the rgency diesel generators, either shielding is provided or the water suppression system is gned such that its actuation does not affect the safety systems it protects (pre-action sprinkler em, manual activation, shielding, etc.).

CRITERION 4 - ENVIRONMENTAL AND MISSILE DESIGN BASES Structures, systems, and components important to safety are designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components are appropriately protected against 1.A-5 Rev. 35

nuclear power unit.

However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses, reviewed and approved by the commission, demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.

structures are designed in accordance with accepted and time proven building codes (as cified in Section 5.1.2) for the loading conditions stated in Sections 5.2.2, 5.3.3, 5.4.3, 5.5.3 5.6.3 which insures that they will operate under normal conditions in a safe manner. In ition, those structures and/or components which could affect public safety were designed to ction safely during an earthquake as discussed in Section 5.8. Wind and tornado storm ection design criteria are discussed in Sections 5.2.2.1.6, 5.3.3.1.4, 5.4.3.1.6, 5.5.3.3.2, 3.1.5, and 5.7.3.1.4. Protection against postulated missiles is discussed in Section 5.2.5.1.

design loads for the containment and major component supports to ensure a safe shutdown r a loss-of-coolant accident are described in Section 5.2.2.1.3.

tems and components important to safety are designed to operate satisfactorily and to be patible with environmental conditions associated with normal operation and postulated dent conditions. Those systems and components located in the containment are designed to rate in an environment of 289°F and 54 psig. Systems and components important to safety are gnated Seismic Class I and designed in accordance with the criteria given in Section 5.2.4.3.

sile protection and pipe whipping protection criteria for these systems and components are n in Sections 5.2.5.1 and 5.4.3.1.

k-before-break (LBB) analyses for the reactor coolant system (RCS) main coolant loops, for pressurizer surge line, and unisolable RCS portions of the safety injection and shutdown ling piping, which demonstrated that the probability of fluid system piping rupture was emely low, were reviewed and approved by the commission. Subsequent to the commission ew and approval, weld overlays were applied to dissimilar metal welds (DMWs) at the tdown cooling, the safety injection and the pressurizer surge nozzles. A revised LBB analysis performed for these nozzles (see Reference A.30). Accordingly, pursuant to GDC 4, 1998 sion, the dynamic effects associated with pipe ruptures in the above piping segments, uding the effects of pipe whipping and discharging fluids have been excluded from the design s of the following components and systems:

Core barrel snubbers, core barrel stabilizer blocks Reactor vessel core support ledge Reactor Cavity Seal, Neutron Shielding Pressurizer Blockhouse Protection of Closed Systems RBCCW piping 1.A-6 Rev. 35

Steam Generator Blow Down sampling piping CRITERION 5 - SHARING OF STRUCTURES, SYSTEMS, AND COMPONENTS Structures, systems, and components important to safety are not shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety function, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

h the auxiliary and the turbine buildings of Millstone Unit 2 are structurally connected to their ective Millstone Unit 1 buildings. The combined buildings are isolated in the lateral direction iscussed in Section 5.4.1 (auxiliary building) and Section 5.5.1 (turbine building). All vertical s which may interact between Millstone Unit 1 and Millstone Unit 2 portions of the buildings e investigated to ensure that they will function safely under all design conditions.

Millstone Unit 2 Condensate Polishing Facility is located in Warehouse Number 5, which is ated North of the Millstone Unit 2 Turbine Building and South of the Millstone Unit 3 densate Polishing Facility and Auxiliary Boiler Building.

st of shared facilities appears in Section 1.2.13.

safe shutdown of any unit will not be impaired by the failure of the facilities and systems ch are shared.

CRITERION 10 - REACTOR DESIGN The reactor core and associated coolant, control and protection systems are designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

nt conditions have been categorized in accordance with their anticipated frequency of urrence and risk to the public, and design requirements are given for each of the four gories. These categories covered by this criterion are Condition I - Normal Operation and dition II - Faults of Moderate Frequency.

design requirement for Condition I is that margin shall be provided between any plant meter and the value of that parameter which would require either automatic or manual ective action; it is met by providing an adequate control system. The design requirement for dition II is that such faults shall be accommodated with, at most, a shutdown of the reactor, h the plant capable of returning to operation after corrective action; it is met by providing an quate protective system. The following design limits apply:

a. The value of the departure from nucleate boiling ratio (DNBR) will not be less than its design limit to ensure that fuel failure does not occur.

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UO2 (considering effects of irradiation on melting point).

c. The maximum primary stresses in the zircaloy fuel clad shall not exceed two-thirds of the minimum yield strength of the material at the operating temperature.

d. Net unrecoverable circumferential strain shall not exceed 1 percent as predicted by computations considering clad creep and fuel-clad interaction effects.

e. Cumulative strain cycling usage, defined as the sum of the ratios of the number of cycles at a given effective strain range (E) to the permitted number (N) at that range shall not exceed 1.0.

f. The fuel rod will be designed to prevent gross clad deformation under the combined effects of external pressure and long term creep.

thermal margins during normal operation ensure that the minimum thermal margins during cipated operational occurrences do not exceed the design basis. The DNBR limit ensures a probability of occurrence of DNB.

occurrence of DNB does not necessarily signify cladding damage; it represents a local ease in temperature which may or may not cause thermal damage, depending upon severity duration.

design is adequate to satisfy the design bases in the event of a reactor coolant system ressurization transient at the end of a fuel cycle.

itation of fuel burnup will be determined by material rather than nuclear considerations. See rences in Chapter 3. Sufficient margin is provided in this core design to allow for the ratio of k-to-average burnup.

CRITERION 11 - REACTOR INHERENT PROTECTION The reactor core and associated coolant systems are designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.

combined response of the fuel temperature coefficient, the moderator temperature coefficient, moderator void coefficient, and the moderator pressure coefficient to an increase in reactor er in the power operating range will be a decrease in reactivity; i.e., the inherent nuclear back characteristics will not be positive.

reactivity coefficients for this reactor are listed in Table 3.4-2 and are discussed in detail in tion 3.4.3.

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The reactor core and associated coolant, control, and protection systems are designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

reactor core is designed not to have sustained power oscillations. If any power oscillations ur, the control system is sufficient to suppress such oscillations.

basic stability of a pressurized water reactor with UO2 fuel is due to the fast acting negative tribution to the power coefficient provided by the Doppler effect.

trend toward xenon oscillations which may occur in the core are controlled and suppressed movement of the control element assemblies (CEAs) so that the thermal design bases are not eeded. Xenon oscillations are characterized by long periods and slow changes in power ribution. The nuclear instrumentation will provide the information necessary to detect these nges.

on stability analysis for Millstone Unit 2 is discussed in Section 3.4.5. The reactor protective em is discussed in Section 7.2.

reactor protective system automatically trips the reactor if axial xenon oscillations are mitted to approach unsafe limits (Sections 7.2.3.3.10 and 1.7.6).

CRITERION 13 - INSTRUMENTATION AND CONTROL Instrumentation are provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls are provided to maintain these variables and systems within prescribed operating ranges.

rumentation is provided, as required, to monitor and maintain significant process variables ch can affect the fission process, the integrity of the reactor core, the reactor coolant pressure ndary, and the containment and its associated systems. Controls are provided for the purpose aintaining these variables within the limits prescribed for safe operation.

principal variables and systems to be monitored include neutron level (reactor power);

tor coolant temperature, flow, and pressure; pressurizer liquid level; steam generator level and sure; and containment pressure and temperature. In addition, instrumentation is provided for tinuous automatic monitoring of process radiation level and boron concentration in the reactor lant system.

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a. Ten independent channels of nuclear instrumentation, which constitute the primary monitor of the fission process. Of these channels, the four wide range channels are used to monitor the reactor from startup through full power; four will monitor the reactor in the power range and are used to initiate a reactor shutdown in the event of overpower; two Reactor Regulating channels will monitor the reactor in the power range.

b. Two independent CEA Position Indicating Systems.

c. Manual control of reactor power by means of CEA's.

d. Manual regulation of coolant boron concentrations.

ore instrumentation is provided to supplement information on core power distribution and to vide for calibration of out-of-core flux detectors.

rumentation measures temperatures, pressures, flows, and levels in the main Steam System Auxiliary Systems and is used to maintain these variables within prescribed limits.

reactor protective system is designed to monitor the reactor operating conditions and to effect able and rapid reactor trip if any one or a combination of conditions deviate from a preselected rating range.

containment pressure and temperature instrumentation is designed to monitor these meters during normal operation and the full range of postulated accidents.

instrumentation and control systems are described in detail in Chapter 7.

CRITERION 14 - REACTOR COOLANT PRESSURE BOUNDARY The reactor coolant pressure boundary is designed, fabricated, erected and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure and of gross rupture.

ctor coolant system components are designed in accordance with the ASME Code, tion III, Pump and Valve Code (reactor coolant system pumps), and ANSI B31.7 (see tion 4 for codes and effective dates). Quality control, inspection, and testing as required by e standards and allowable reactor pressure-temperature operations ensure the integrity of the tor coolant system.

reactor coolant system components are considered Class I for seismic design.

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The reactor coolant system and associated auxiliary, control, and protection system is designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

design criteria and bases for the reactor coolant pressure boundary are described in the onse to Criterion 14.

operating conditions established for the normal operation of the plant are discussed in the R and the control systems are designed to maintain the controlled plant variables within these rating limits, thereby ensuring that a satisfactory margin is maintained between the plant rating conditions and the design limits.

reactor protective system functions to minimize the deviation from normal operating limits in event of certain anticipated operational occurrences. The results of analyses show that the gn limits of the reactor coolant pressure boundary are not exceeded in the event of such urrences.

CRITERION 16 - CONTAINMENT DESIGN Reactor containment and associated systems are provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

reactor containment structure, described in Section 5.2, consists of a prestressed concrete nder and dome with a reinforced concrete base. A one-quarter inch thick welded steel liner e is attached to the inside face of the concrete to provide a high degree of leak tightness.

igned as a pressure vessel, the containment structure is capable of withstanding all design tulated accident conditions including a loss-of-coolant accident (LOCA). All containment etrations are sealed as described in Section 5.2.6. Isolation valves are provided for all piping ems which penetrate the containment, as described in Section 5.2.7.

an extra measure of safety, an enclosure building completely surrounds the containment. In the nt of an accident, the enclosure building filtration region (EBFR), described in Section 6.7.2, aintained at a slightly negative pressure to preclude leakage to the environment. Potential age from the containment is channeled into the enclosure building filtration system as cribed in Section 6.7. Throughline leakage that can bypass the EBFR is discussed in tion 5.3.4.

CRITERION 17 - ELECTRIC POWER SYSTEMS An on site electric power system and an off site electric power system are provided to permit functioning of structures, systems, and components important to safety. The safety 1.A-11 Rev. 35

and design conditions of the reactor coolant pressure boundary (RCPB) are not exceeded as a result of anticipated operational occurrences; and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.

The on site electric power supplies, including the batteries, and the on site electric distribution system, have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.

Electric power from the transmission network to the on site electric distribution system is supplied by two physically independent circuits (not necessarily on separate rights-of-way), designed and located so as to minimize to the extent practical, the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions. A switchyard common to both circuits is acceptable. Each of these circuits is designed to be available in sufficient time following a loss of all on site AC power supplies and the other off site electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the RCPB are not exceeded. One of these circuits is designed so it is available within a few seconds after a loss-of-coolant accident (LOCA) to assure that core cooling, containment integrity, and other vital safety functions are maintained.

Provisions are included to minimize the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the transmission network, or from the on site electric power supplies.

off site power supplies system is described in Sections 8.1 and 8.2. The preferred source of iliary power for unit shutdown is from or through the reserve station service transformers.

tem interconnection is provided by four 345 kV circuits. These transmission lines are on a le right-of-way with each line installed on an independent set of structures. A description of structure routing configuration is described in Section 8.1.2.1.

combination breaker-and-a-half and double breaker-double bus switching arrangement in the kV substation includes two full capacity main buses. Primary and backup relaying are vided for each circuit along with circuit breaker failure backup protection. These provisions mit the following:

a. Any circuit can be switched under normal or fault conditions without affecting another circuit.

b. Any single circuit breaker can be isolated for maintenance without interrupting the power or protection to any circuit.

c. Short circuits on any section of bus are isolated without interrupting service to any element other than those connected to the faulty bus section.

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generator for this contingency condition; however, power can be restored to the good element in less than eight hours by manually isolating the fault with appropriate disconnect switches.

rhead lines from the switchyard to the reserve station service transformers are separated at the tchyard structure and are carried on separate towers. These transformers are located near each t, and are physically isolated from the normal station service transformers and from the main sformers.

he event of loss of power from the normal station service transformer, there is an immediate matic transfer of auxiliary loads to the Unit 2 reserve station service transformer. In the kely event that power is not available from this source, and from the On site Emergency sel mentioned below, the operator can manually connect emergency bus A-5 (24E) to Unit 3 34A or 34B. By means of interlocked circuit breakers, the Unit 2 post accident loads can be from this source.

on site power supply system is described in Sections 8.3 and 8.5. Two full capacity, separate redundant batteries are provided for all DC loads and for 120 volt AC vital instrument loads.

he event that off site power is not available when needed, a start signal is given to both rgency diesel generators (DG). These generators and their auxiliaries are entirely separate and undant, and each generator feeds one 4,160 volt emergency bus. A generator is automatically nected to its bus only if there is no bus voltage and only if the dead bus did not result from ective relay action.

electric power distribution system is described in Section 8.7. The redundancy of the power rces is enhanced by separate and redundant auxiliary power and control distribution systems.

ingle failure and any possible related failures in that channel cannot adversely affect ipment and components on the other redundant channel.

to the redundancy and separation of power supplies, distribution and control required for l functions, all components can be readily inspected and tested. Similarly, most subsystems be tested in their entirety.

CRITERION 18 - INSPECTION AND TESTING OF ELECTRIC POWER SYSTEMS Electric power systems important to safety are designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components. The systems shall be designed with a capability to test periodically (1) the operability and functional performance of the components of the systems, such as on site power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the 1.A-13 Rev. 35

operability and functional performance of the components of these systems are verified by odic inspections and tests as described in Chapter 8.

verify that the emergency power system will properly respond within the required time limit n required, the following tests are performed:

a. Manually initiated demonstration of the ability of the diesel-generators to start, synchronize and deliver power up to 2750 kW continuous, when operating in parallel with other power sources. Normal unit operation will not be affected.

b. Demonstration of the readiness of the on site generator system and the control systems of vital equipment to automatically start, or restore to operation, the vital equipment by initiating an actual loss of all normal AC station service power. This test will be conducted during each refueling interval.

Demonstration of the automatic sequencing equipment during normal unit operation. This test exercises the control and indication devices, and may be performed any time, as the sequencing equipment is redundant to normal operations. If there is a safety injection actuation signal while the test is underway, it takes precedence and immediately cancels the test. The equipment then responds to the safety injection actuation signal in the manner described in Section 8.3.

ce operation of the protective system will be infrequent, each system is periodically and inely tested to verify its operability. Each channel of the protective systems, including the sors up to the final protection element, is capable of being checked during reactor operation.

output circuit breakers are provided to permit individual testing during plant operation. See pters 7 and 8 for further details.

CRITERION 19 - CONTROL ROOM A control room is provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents (LOCA). Adequate radiation protection is 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.

Equipment at appropriate locations outside the control room is provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

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and dampers which act to shunt the intake air through the filters in the event of a high orne radioactivity level. The dampers are automatically actuated from the control room nitors. Acting on a high radiation level indication, the fresh air dampers close and recirculation pers open to provide a complete closed cycle ventilation mode with a portion of the air stream g drawn through the HEPA-charcoal filter assembly. In addition, an area radiation monitor is vided to indicate and alarm on high radiation level.

he event the operator is forced to abandon the control room, a hot shutdown panel (C21) vide the instrumentation and control necessary to maintain the plant in the hot shutdown dition (see Section 7.6.4). The potential capability for bringing the plant to a shutdown is also vided outside the control room.

Shutdown System Panels located outside the control room contain the instruments and trols necessary to achieve a hot shutdown condition should the control room become nhabitable due to fire (see Section 7.6.5). The Fire Shutdown Panel can be utilized for any rgency event which requires control room evacuation.

all indicators and controls provided on the Fire Shutdown Panel are available for all fires.

rnate methods of compliance are documented in the Millstone Unit 2 10 CFR 50 Appendix R mpliance Report.

CRITERION 20 - PROTECTION SYSTEM FUNCTIONS The protection system is designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

reactor is protected by the Reactor Protective System from reaching a condition that could lt in exceeding acceptable fuel design limits as a result of anticipated operational occurrences S-N18.2, Condition II). The Protective System is designed to monitor the reactor operating ditions and initiate a reactor trip if any of the following measured variables exceeds the rating limits:

a. High power level (variable, highest of thermal or neutron flux).

b. High pressurizer pressure.

c. Thermal margin (variable low pressure).

d. Turbine trip (equipment protection only).

e. Low reactor coolant flow.

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g. Low steam generator pressure.

h. Local power density.

i. High containment pressure.

Engineered Safeguards Actuation System detects accident conditions and initiates the Safety tures Systems which are designed to localize, control, mitigate, and terminate such accidents.

Engineered Safeguards Actuation System protects the general public from the release of oactivity by actuating components that cool the reactor core, depressurize the containment, ate the containment, and filter any containment leakage (see Section 7.3). The following meters are continuously monitored;

a. Low pressurizer pressure.

b. High/high-high containment pressure.

c. Containment gaseous and particulate radiation.

d. Low steam generator pressure.

e. High fuel handling area radiation.

f. Low refueling water storage tank level.

g. Emergency bus undervoltage.

Auxiliary Feedwater Automatic Initiation System (AFAIS) provides a dedicated source of water of sufficient capacity to remove decay heat and sensible heat following casualty ations. Automatic initiation of auxiliary feedwater occurs in response to a low Steam erator level in a two out of four (2 of 4) channel auctioneered matrix (see Section 7.3.2.2.h).

CRITERION 21 - PROTECTION SYSTEM RELIABILITY AND TESTABILITY The protection system is designed for high functional reliability and inservice testability commensurate with the safety functions to be performed. Redundancy and independence designed into the protection system is sufficient to assure that (1) no single failure results in loss of the protection function, and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. The protection system is designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.

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pendent, e.g., with respect to piping, wire routing, mounting and supply of power. This ependence permits testing and the removal from service of any component or channel without of the protection function.

h channel of the protective system, including the sensors up to the final protective element, is able of being checked during reactor operation. Measurement sensors of each channel used in ective systems are checked by observing outputs of similar channels which are presented on cators and recorders on the control board. Trip units and logic are tested by inserting a signal the measurement channel ahead of the trip units and, upon application of a trip level input, erving that a signal is passed through the trip units and the logic to the logic output relays. The c output relays are tested individually for initiation of trip action. See Chapter 7.

CRITERION 22 - PROTECTION SYSTEM INDEPENDENCE The protection system is designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing and postulated accident conditions on redundant channels do not result in loss of the protection function, or is demonstrated to be acceptable on some other defined basis. Design techniques, such as functional diversity or diversity in component design and principles of operation, is used to the extent practical to prevent loss of the protection function.

reactor protective systems conform to the provisions of the Institute of Electrical and ctronic Engineers (IEEE) Criteria for Nuclear Power Plant Protection Systems, IEEE-279,

1. Two to four independent measurement channels, complete with sensors, sensor power plies, signal conditioning units and bistable trip units, are provided for each protective meter monitored by the protective systems. The measurement channels are provided with a h degree of independence by separate connection of the channel sensors to the process ems. Power to the channels is provided by independent vital power supply buses. See tion 7.2.

mbustion Engineering Topical Report CENPD-11 (Reactor Protection System Diversity, W.

Coppersmith, C. I. Kling, A. T. Shesler, and B. M. Tashjian CENPD, February 1971) onstrates that functional diversity has been incorporated in the protective system design.

CRITERION 23 - PROTECTION SYSTEM FAILURE MODES The protection system is designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air) or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.

tective system instrumentation has been designed to fail into a safe state or into a state blished as acceptable in the event of loss of power supply or disconnection of the system, undancy, channel independence, and separation are incorporated in the protective system 1.A-17 Rev. 35

CRITERION 24 - SEPARATION OF PROTECTION AND CONTROL SYSTEMS The protection system is separated from control systems to the extent that failure of any single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system. Interconnection of the protection and control systems is limited so as to assure that safety is not significantly impaired.

reactor protective systems are separated from the control instrumentation systems so that ure or removal from service of any control instrumentation system component or channel does inhibit the function of the protective system. See Section 7.2.

CRITERION 25 - PROTECTION SYSTEM REQUIREMENTS FOR REACTIVITY CONTROL MALFUNCTIONS The protection system is designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.

ctor shutdown with CEA's is accomplished completely independent of the control functions e the trip breakers interrupt power to the full length CEA drive mechanisms regardless of ting control signals. The design is such that the system can withstand accidental withdrawal of trolling groups without exceeding acceptable fuel design limits. An analysis of these accidents iven in Section 14.4. The reactor protection system will prevent specified acceptable fuel gn limits from being exceeded for any anticipated transients.

ITERION 26 - REACTIVITY CONTROL SYSTEM REDUNDANCY AND CAPABILITY Two independent reactivity control systems of different design principles is provided. One of the systems uses control rods, preferably including a positive means for inserting the rods, and is capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system is capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems is capable of holding the reactor core subcritical under cold conditions.

o independent systems are provided for controlling reactivity changes. The Control Element ve System (CEDS) controls reactivity change required for power changes and power ribution shaping, and is also used for reactor protection. The boric acid shim control pensates for long term reactivity changes such as those associated with fuel burnup, variation 1.A-18 Rev. 35

er system acting independently is capable of making the core subcritical from a hot operating dition and holding it subcritical in the hot standby condition at 532°F.

er system is able to insert negative reactivity at a sufficiently fast rate to prevent exceeding eptable fuel design limits as the result of a power change (i.e., the positive reactivity added by nup of xenon).

boron addition system is capable of holding the reactor core subcritical under cold conditions.

CRITERION 27 - COMBINED REACTIVITY CONTROL SYSTEMS CAPABILITY The reactivity control system is designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.

combined capability of the reactor control systems in conjunction with dissolved boron ition by the safety injection system is such that under postulated accident conditions, even h the CEA of highest worth stuck out of the core, the core would be maintained in a geometry ch assures adequate short and long term cooling. See Criteria 26 and 28.

CRITERION 28 - REACTIVITY LIMITS The reactivity control systems are designed with appropriate limits on the potential amount of rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents include consideration of ejection (unless prevented by positive means) rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.

basis for selecting the number of control element assemblies in the core includes assuring that reactivity worth of any one assembly is within a preselected maximum value. The control ment assemblies have been separated into sets: a shutdown set and a regulating set further divided into groups as necessary. Administrative procedures and interlocks are used to permit y one shutdown group to be withdrawn at a time, and to permit withdrawal of the regulating ups only after the shutdown groups ar fully withdrawn. The regulating groups are programmed ove in sequence and within limits that prevent the rates of reactivity change and the worth of vidual assemblies from exceeding limiting values. See Sections 7.4.2, 14.4.1, 14.4.2, and 4.3.

1.A-19 Rev. 35

s associated with an inadvertent and sudden release of energy to the coolant such as that lting from CEA ejection, CEA drop, steam line rupture or cold water addition. See tions 14.4.8, 14.4.9, and 14.1.5.

boric acid system rate of reactivity addition is too slow to cause rupture of the reactor coolant sure boundary or disturb the reactor pressure vessel internals.

CRITERION 29 - PROTECTION AGAINST ANTICIPATED OPERATIONAL OCCURRENCES The protection and reactivity control systems are designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.

icipated operational occurrences have been considered in the design of the protection and tivity control systems. As is demonstrated in the safety analysis in Chapter 14 and the mbustion Engineering Report CENPD-11 (Reactor Protection System Diversity, W. C.

persmith, Cl. L. Kling, A. T. Shesler, and B. M. Tashjian, CENPD-11, February 1971), the gn is adequate to minimize the consequences of such occurrences and assures that the health safety of the public is protected from the consequences of such occurrences.

adherence to a detailed program for quality assurance, careful attention to design, component ction and system installation, coupled with the design features of redundancy, independence, testability will assure that a high probability exists that the protection and reactivity control ems will accomplish their functions. See Criteria 21 through 26.

CRITERION 30 - QUALITY OF REACTOR COOLANT PRESSURE BOUNDARY Components which are part of the reactor coolant pressure boundary are designed, fabricated, erected and tested to the highest quality standards practical. Means are provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.

reactor coolant pressure boundary components have been designed, fabricated, erected and ed in accordance with the ASME Code Section III, 1971 through summer 1971 Addenda and SI B31.7, 1969 as specified in Criterion 14. Replacement steam generator subassemblies were icated in accordance with ASME Code Section III 1983 through summer 1984 Addenda.

replacement reactor vessel closure head including all nozzles (CEDM, HJTC, ICI and the t) is constructed in accordance with ASME Boiler and Pressure Vessel Code,Section III, section NB, 1998 Edition through 2000 Addenda.

tainment sump instrumentation is used to detect reactor coolant system leakage by providing rmation on rate of rise of sump levels and frequency of sump pump operation. Flow 1.A-20 Rev. 35

dually increasing. The containment air monitoring system (see Section 7.5.6) provides an itional means of reactor coolant system leakage detection.

CRITERION 31 - FRACTURE PREVENTION OF REACTOR COOLANT PRESSURE BOUNDARY The reactor coolant pressure boundary is designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design reflects consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.

bon and low alloy steel materials which form part of the pressure boundary meet the uirements of the ASME Code,Section III, paragraph N-330 at a temperature of +40°F.

f. Section 4.2.2). The actual nilductility transition temperature (NDTT) of the materials has n determined by drop weight tests in accordance with ASTM-E-208. For the reactor vessel e metals, Charpy tests were also performed and the results used to plot a Charpy transition ve. To address changes in regulations, the original design requirements of N-330 were plemented and the materials' initial nil-ductility reference temperatures (RTNDT) were servatively established based upon available or supplemental material toughness testing. In case of the replacement steam generators, the materials were required to satisfy NB-2331 and NDT values were established to satisfy current requirements.

bon and low alloy steel materials including weld filler metal which form part of the reactor sure boundary for replacement reactor vessel closure head satisfy ASME Section III, NB

0. Actual NDTT was established by drop weight test in accordance with ASTM-E-208 at

°F. RTNDT of the replacement head based materials was established by Charpy V-notch test at

°F. Charpy transition curves were plotted using test data for the base material of the acement reactor vessel head.

the reactor coolant pressure boundary components are constructed in accordance with the licable codes and comply with the test and inspection requirements of these codes. These test ection requirements assure that flaw sizes are limited so that the probability of failure by rapid pagation is extremely remote. Particular emphasis is placed on the quality control applied to reactor vessel, on which tests and inspections exceeding code requirements are performed.

tests and inspections performed on the reactor vessel are summarized in Section 4.6.5.

reactor vessel beltline materials receive sufficient neutron irradiation to cause embrittlement increase in RTNDT). To provide conservative margins against nonductile or rapidly pagating failure, several techniques are employed. Operating limits which account for the 1.A-21 Rev. 35

ordance with the requirements of 10 CFR 50 Appendix G (Additional details are provided in tion 4.5.1). In addition, compliance with 10 CFR 50.61 assures that the shift in the transition perature of the reactor vessel beltline materials provides adequate margins of safety against ere pressurized thermal shock events.

assure that the reactor vessel beltline materials are behaving in the predicted manner, a reactor sel material surveillance program is conducted (See Criterion 32 and Section 4.6.2).

ghness testing of unirradiated reactor vessel materials was performed to establish the baseline, the irradiated surveillance materials are periodically tested as surveillance capsules are oved during the plant's design life, in accordance with the requirements of 10 CFR 50, endix H.

activation of the safety injection systems introduces highly borated water into the reactor lant system at pressures significantly below operating pressures and will not cause adverse sure or reactivity effects.

thermal stresses induced by the injection of cold water into the vessel have been examined.

lysis shows the there is no gross yielding across the vessel wall using the minimum specified d strength in the ASME Boiler and Pressure Vessel Code,Section III. (Ref. Section 4.5.4).

erse effects that could be caused by exposure of equipment or instrumentation to containment y water is avoided by designing the equipment or instrumentation to withstand direct spray or ocating it or protecting it to avoid direct spray.

CRITERION 32 - INSPECTION OF REACTOR COOLANT PRESSURE BOUNDARY Components which are part of the reactor coolant pressure boundary are designed to permit (1) periodic inspection and testing of important areas and features to assess their structural and leak-tight integrity, and (2) an appropriate materials surveillance program for the reactor pressure vessel.

visions are made for inspection, testing, and surveillance of the Reactor Coolant System ndary as required by ASME Boiler and Pressure Vessel Code,Section XI.

Reactor vessel surveillance program was designed in accordance with ASTM E185. It plies with ASTM E185-73 and 10 CFR 50, Appendix H. Section 4.6.3 presents the details of reactor surveillance program. Sample pieces taken from the same shell plate material used in ication of the reactor vessel are installed between the core and the vessel inside wall. These ples will be removed and tested at intervals during vessel inside wall. These samples will be oved and tested at intervals during vessel life to provide an indication of the extent of the tron embrittlement of the vessel wall. Charpy tests will be performed on the samples to elop a Charpy transition curve. By comparison of this curve with the Charpy curve and drop ght tests for specimens taken at the beginning of the vessel life, the change of NDTT will be rmined and operating instructions adjusted as required.

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A system to supply reactor coolant makeup for protection against small breaks in the reactor coolant pressure boundary is provided. The system safety function shall be to assure that specified acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure boundary and rupture of small piping or other small components which are part of the boundary. The system is designed to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system safety function can be accomplished using the piping, pumps and valves used to maintain coolant inventory during normal reactor operation.

ctor Coolant System (RCS) makeup during normal operation is provided by the Chemical and ume Control System (CVCS) which includes three positive displacement charging pumps d at 44 gpm each. Two operating CVCS pumps are capable of making up the flow loss for s in the reactor coolant boundary of up to 0.250 inches equivalent diameter. Two CVCS ps are sufficient to makeup for a 0.250 inch equivalent diameter RCS break assuming either:

minimum letdown with no RCS leakage or 2) letdown isolated with maximum Technical cification allowed leakage. This CVCS design results in a substantial RCS steady state sure that is well above the shutoff head of the high pressure safety injection pumps. The ve described CVCS capability fulfills the intent of Criterion 33. Information on CVCS is tained in Section 9.2.

CRITERION 34 - RESIDUAL HEAT REMOVAL A system to remove residual heat is provided. The system safety function is to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities are provided to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

idual heat removal capability is provided by the shutdown cooling system for reactor coolant perature less than 300°F (see Section 9.3). For temperatures greater than 300°F, this function rovided by the steam generators (see Section 10.3). Sufficient redundancy, interconnections, detection, and isolation capabilities exist with these systems to assure that the residual heat oval function can be accomplished, assuming failure of a single active component. Within ropriate design limits, either system will remove fission product decay heat at a rate such that cified acceptable fuel design limits and the design conditions of the reactor coolant pressure ndary will not be exceeded.

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A system to provide abundant emergency core cooling is provided. The system safety function is to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented, and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities is provided to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

emergency core cooling system is discussed in detail in Chapter 6. It consists of the high sure safety injection subsystem, the low pressure safety injection subsystem, and the safety ction tanks (see Section 6.3).

s system is designed to meet the criterion stated above with respect to the prevention of fuel clad damage that would interfere with the emergency core cooling function, for the full ctrum of break sizes, and to the limitation of metal-water reaction. Each of the subsystems is y redundant, and the subsystems do not share active components other than the valves trolling the suction headers of the high and low pressure safety injection pumps. Minimum ty injection is assured even though one of these valves fails to function. These valves are in no associated with the function of the safety injection tanks.

ECCS design satisfies the criteria specified in 10 CFR 50.46(b).

CRITERION 36 - INSPECTION OF EMERGENCY CORE COOLING SYSTEM The emergency core cooling system is designed to permit appropriate periodic inspection of important components, such as spray rings in the reactor pressure vessel, water injection nozzles, and piping to assure the integrity and capability of the system.

pter 6 describes the arrangement and location of the components in the emergency core ling system. All pumps, the shutdown cooling heat exchangers, and valves and piping external he containment structure are accessible for physical inspection at any time. All safety injection es and piping inside the containment structure, and the safety injection tanks, may be ected during refueling.

accessibility for inspection of the reactor vessel internals, reactor coolant piping and items h as the water injection nozzles is described in Sections 4.6.3 through 4.6.6.

CRITERION 37 - TESTING OF EMERGENCY CORE COOLING SYSTEM The emergency core cooling system is designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, 1.A-24 Rev. 35

the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.

Emergency Core Cooling System (Safety Injection System) is provided with testing facilities emonstrate system component operability. Testing can be conducted during normal plant ration with the test facilities arranged not to interfere with the performance of the systems or h the initiation of control circuits, as described in Section 6.3.4.2.

safety injection system is designed to permit periodic testing of the delivery capability up to a tion as close to the core as practical. Periodic pressure testing of the Safety Injection System ossible using the cross connection to the charging pumps in the Chemical and Volume Control tem.

low pressure safety injection pumps are used as shutdown cooling pumps during normal plant ldown. The pumps discharge into the safety injection header via the shutdown cooling heat hangers and the low pressure injection lines.

h the plant at operating pressure, operation of safety injection pumps may be verified by rculation back to the refueling water storage tank. This will permit verification of flow path tinuity in the high pressure injection lines and suction lines from the refueling water storage

.

ated water from the safety injection tanks may be bled through the recirculation test line to fy flow path continuity from each tank to its associated main safety injection header.

operational sequence that brings the Safety Injection System into action, including transfer to rnate power sources, can be tested in parts as described in Chapters 6, 7, and 8.

CRITERION 38 - CONTAINMENT HEAT REMOVAL A system to remove heat from the reactor containment is provided. The system safety function is to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss-of-coolant accident and maintain them at acceptably low levels.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities are provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

1.A-25 Rev. 35

ucing the containment pressure and temperature following any loss-of-coolant accident CA) and maintaining them at acceptably low levels.

ficient heat removal capability is provided by any of the following combinations of ipment:

a. Two containment spray pumps with associated heat exchangers.

b. Three of the four containment air recirculation and cooling units.

c. One containment spray pump with associated heat exchanger in combination with two containment air recirculation and cooling units.

containment heat removal systems are provided with suitable interconnections such that each bination of two containment air recirculation and cooling units and one containment spray p, aligned with the associated shutdown cooling heat exchanger, are provided with cooling er from the same RBCCW header and powered by the same emergency bus. All associated ponents, such as valves, are likewise powered from the same emergency bus. Each bination of these components is capable of removing heat at a rate greater than required to t the postaccident containment pressure and temperature. A single failure of any active ponent does not render the redundant group inoperable.

containment spray system is provided with containment isolation capabilities in accordance h Criterion 56. The above containment penetration is provided with leak detection capabilities ccordance with Criterion 54.

CRITERION 39 - INSPECTION OF CONTAINMENT HEAT REMOVAL SYSTEM The containment heat removal system is designed to permit appropriate periodic inspection of important components, such as the torus, sumps, spray nozzles, piping to assure the integrity and capability of the system.

or components of the containment spray system are located to permit access for periodic ntenance and inspection. Components of the containment air and recirculation system are ted within the containment and are therefore accessible for maintenance and inspection ng shutdown.

containment sump is located in the lowest elevation of the containment at Elevation (-)22-6 is accessible during reactor shutdown for periodic visual inspections (see Section 6.2).

containment spray nozzles are accessible for periodic inspection during reactor shutdown.

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The containment heat removal system is designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole, and, under conditions as close to the design and practical, the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.

spray system and the air recirculation and cooling systems in the containment have visions for online testing to assure system operation, performance and structural and leaktight grity of the associated components. Testing procedures are described in Sections 6.4.4.2 and 4.2, respectively.

containment heat removal systems undergo preoperational testing prior to plant startup. The procedure is described in Chapter 13.

CRITERION 41 - CONTAINMENT ATMOSPHERE CLEAN UP Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment are provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.

Each system has suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities to assure that for offsite electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure.

containment is not provided with an atmosphere cleanup system. However, a second barrier, enclosure building, is provided around the containment to collect potential leakage from the tainment under postaccident conditions.

enclosure building filtration system (EBFS) is provided to collect and process potential age from the containment during postaccident operation. Potential containment leakage is the enclosure building filtration region (EBFR) which forms the outer barrier in the double tainment boundary. The EBFS is described in Section 6.7. Throughline leakage that can ass the EBFR is discussed in Section 5.3.4.

hydrogen control system is provided to mix and monitor the concentration of hydrogen in the tainment atmosphere following postulated accidents to assure the containment integrity is 1.A-27 Rev. 35

h of these cleanup systems consist of completely redundant, independent safety function.

se are provided with suitable interconnections and separations such that a single failure in any system does not render the redundant subsystem inoperable.

hydrogen control system is incorporated with containment isolation capabilities for each ng subsystem which penetrates the primary containment. Containment isolation is in ordance with Criterion 56. Provision for leak detection is incorporated in accordance with erion 54.

ITERION 42 - INSPECTION OF CONTAINMENT ATMOSPHERE CLEANUP SYSTEMS The containment atmosphere cleanup systems are designed to permit appropriate periodic inspection of important components, such as filter frames, fans, hydrogen recombiners, analyzers, valves, ducts, and piping to assure the integrity and capability of the systems.

enclosure building filtration system (EBFS) is located to permit access for periodic ection and maintenance. The components of the hydrogen control system located outside the tainment are accessible for periodic inspection and maintenance. The components located de containment are accessible for inspection and maintenance during shutdown.

hydrogen control system and EBFS are incorporated with provisions for online testing to onstrate system operation, performance and integrity. These tests procedures are described in tions 6.6.4.2 and 6.7.4.2, respectively.

CRITERION 43 - TESTING OF CONTAINMENT ATMOSPHERE CLEANUP SYSTEM The containment atmosphere cleanup systems are designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the systems such as fans, filters, dampers, pumps, and valves and (3) the operability of the systems as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the systems into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of associated systems.

enclosure building filtration system (EBFS) and hydrogen control system are incorporated h provisions for online testing. The test procedures are described in Sections 6.7.4.2 and 4.2, respectively.

containment atmosphere cleanup systems undergo preoperational tests prior to plant startup.

t procedures are described in Chapter 13.

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A system to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink is provided. The system safety function is to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions.

Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities are provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

RBCCW system, described in Section 9.4, and the service water system, described in tion 9.7.2, are provided to transfer heat from structures, systems, and components important to ty to an ultimate heat sink. The systems are designed to transfer the combined heat load of e structures, systems, and components under normal and accident conditions.

RBCCW supplies cooling water to components important to safety through two independent ders. One header provides adequate heat removal capability to safely shutdown the plant under dent conditions, but at a lesser rate. Service water is supplied to the RBCCW heat exchangers wo independent headers to assure heat removal capability. Two service water pumps are in tinuous operation with a spare pump provided. One pump supplies sufficient heat removal ability for the RBCCW heat exchangers to safely shut down the plant and for accident gation.

RBCCW and service water systems are provided with suitable redundancy in components suitable interconnections to assure heat removal capability. The systems are designed to ble isolation of system components or headers and to detect system maloperation.

RBCCW and service water systems are designed to operate with onsite power (assuming ite power is not available) and with offsite power (assuming onsite power is not available).

systems are designed such that a single failure in either system will not adversely affect safe ration, accident mitigation, or safe shutdown of the plant.

CRITERION 45 - INSPECTION OF COOLING WATER SYSTEM The cooling water system is designed to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to assure the integrity and capability of the system.

RBCCW system and service water system, excluding underground piping, are designed to mit periodic inspection of important components, such as pumps, heat exchangers, valves and ng to assure the integrity and heat removal capability of the system. The components of the CCW system located outside the containment are located in a low radiation area, which 1.A-29 Rev. 35

ng plant shutdown. Inspection of RBCCW system components is described in Section 9.4.4.2.

or service water system components, such as pumps and strainers, are accessible for periodic ection during normal operation. Inspection of the service water system is described in tion 9.7.2.5.

CRITERION 46- TESTING OF COOLING WATER SYSTEM The cooling water system is designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and the performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation for reactor shutdown and for loss-of-coolant accidents (LOCA), including operation of applicable portions of the protection system and the transfer between normal and emergency power sources.

ine testing provisions are incorporated in the RBCCW and service water systems to onstrate the operability, performance, structural and leaktight integrity of the systems. The CCW and service water systems are designed so that under conditions as close to design as tical, the performance shall be demonstrated of the full operational sequence that brings the em into operation, including operation of applicable portions of the protection system, and the sfer between normal and emergency power sources. Testing of the RBCCW and service water ems are described in Sections 9.4.4.2 and 9.7.2.5, respectively.

CRITERION 50 - CONTAINMENT DESIGN BASIS The reactor containment structure, including access openings, penetrations, and the containment heat removal system are designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and, with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident. This margin reflects consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and energy from metal-water and other chemical reactions that may result from degraded emergency core cooling functioning, (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters.

containment structure, including the access openings, penetrations and the containment heat oval system, is designed to withstand a pressure of 54 psig and a temperature of 289°F owing a loss-of-coolant accident (LOCA) or a main steam line break accident (see tion 14.8.2). Details of the methods used to analyze the containment structure are described in tion 5.2.2. To obtain an adequate margin of safety, a factored load was selected for a design ch allows a 25 percent increase over the calculated postulated accident load.

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e, such as penetration sleeves, personnel locks, and equipment hatch, are designed to meet the uirements of the ASME Boiler and Pressure Vessel Code,Section III (Nuclear Vessels) 1968 tion through the summer 1969 addenda Paragraph N-1211. Further description of the liner e is contained in Section 5.2.3.

a further check on the design a structural integrity test, composing a test pressure load of 115 ent of the design accident pressure load, is conducted prior to operation. In addition to this, a rate test will be conducted prior to operation and at certain intervals during operation. Details he leak rate test are provided in Section 5.2.8.1.

CRITERION 51 - FRACTURE PREVENTION OF CONTAINMENT PRESSURE BOUNDARY The reactor containment boundary is designed with sufficient margin to assure that under operating, maintenance, testing, and postulated accident conditions (1) its ferritic materials behave in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design reflects consideration of service temperatures and other conditions of the containment boundary material during operation, maintenance, testing, and postulated accident conditions, and the uncertainties in determining (1) material properties, (2) residual, steady state, and transient stresses, and (3) size of flaws.

containment consists of a prestressed reinforced concrete cylinder and dome connected to supported by a massive reinforced concrete slab. A one-quarter inch thick steel liner plate is ched to the inside surface of the concrete containment and its penetrations. Consideration has n given to both design and construction techniques to assure the containment pressure ndary behaves in a ductile manner and the probability of a rapidly propagating fracture is imized.

liner plate is designed to carry no load, and serves only as a leaktight barrier. Analytical ulations of the strains under an extreme and most improbably set of load conditions indicate strains are well within the ductile limits of the material. The analytical approach to liner gn is presented in the Bechtel Corporation Proprietary Report B-TOP-1.

ll penetrations the liner plate is thickened using the 1968 ASME Code,Section III for Class B sels as a guide to limit stress concentrations.

visions, as described in Section 5.2.5.1.1, are made to prevent a potential internally generated sile from rupturing the liner plate.

erials for the penetrations require satisfactory Charpy V-notch impact test results. All etrations are stress relieved. The construction materials selected for the liner plate and etrations are given in Section 5.2.1.

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CRITERION 52 - CAPABILITY FOR CONTAINMENT LEAKAGE RATE TESTING The reactor containment and other equipment which may be subjected to containment test conditions are designed so that periodic integrated leakage rate testing can be conducted at containment design pressure.

reactor containment and other equipment which is subjected to containment test conditions designed so that periodic integrated leakage rate testing can be conducted at containment gn pressure. The test procedure is described in Section 5.2.8.

CRITERION 53 - PROVISIONS FOR CONTAINMENT TESTING AND INSPECTION The reactor containment is designed to permit (1) appropriate periodic inspection of all important areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leak tightness of penetrations which have resilient seals and expansion bellows.

reactor containment is designed to permit appropriate periodic testing of all important areas.

ails of the containment testing and inspection are discussed in Section 5.2.8.

CRITERION 54 - PIPING SYSTEMS PENETRATING CONTAINMENT Piping systems penetrating primary reactor containment are provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. Such piping systems are designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits.

ng systems penetrating containment are provided with suitable redundancy to assure the ems function adequately during postulated accidents such that failure of a portion of a system not create a hazard to safe unit operation. Piping systems are provided with containment ation valves in accordance with the requirements of Criterion 55, 56, and 57. Containment ation valves have been selected and tested to provide adequate operation at maximum flow ditions. Provisions are incorporated for leak detection and performance testing of those piping ems penetrating the containment (Section 5.2.7.4.2).

CRITERION 55 - REACTOR COOLANT PRESSURE BOUNDARY PENETRATING CONTAINMENT Each line that is part of the reactor coolant pressure boundary and that penetrates primary reactor containment is provided with containment isolation valves as follows, unless it can 1.A-32 Rev. 35

(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

Isolation valves outside containment are located as close to containment as practical and upon loss of actuating power, automatic isolation valves are designed to take the position that provides greater safety.

Other appropriate requirements to minimize the probability or consequences of an accidental rupture of these lines or of lines connected to them are provided as necessary to assure adequate safety. Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation valves and containment, include consideration of the population density, use characteristics, and physical characteristics of the site environs.

those piping systems penetrating the containment and connected directly to the reactor lant pressure boundary, isolation provisions have been incorporated. Section 5.2.7 indicates licable valve arrangements, a complete description of penetrations and valve position on power failure.

visions are made for leak testing as described in Section 5.2.7.4.2.

CRITERION 56 - PRIMARY CONTAINMENT ISOLATION Each line that connects directly to the containment atmosphere and penetrates primary reactor containment is provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or 1.A-33 Rev. 35

(3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

Isolation valves outside containment are located as close to the containment as practical and upon loss of actuating power, automatic isolation valves are designed to take the position that provides greater safety.

those piping system penetrating the containment and connected directly to the containment osphere, isolation provisions have been incorporated. Section 5.2.7 indicates the applicable e arrangements, a complete description of penetrations and valve position on air/power ure.

CRITERION 57 - CLOSED SYSTEM ISOLATION VALVES Each line that penetrates primary reactor containment and is neither part of the reactor coolant pressure boundary not connected directly to the containment atmosphere has at least one containment isolation valve which shall be either automatic, or locked closed, or capable of remote manual operation. This valve is outside containment and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve.

those piping systems penetrating the containment which are neither part of the reactor coolant sure boundary nor connected directly with the containment atmosphere, isolation provisions e been incorporated.

tion 5.2.7 indicates applicable valve arrangements, a complete description of penetrations and e position on air/power failure.

RITERION 60 - CONTROL OF RELEASES OF RADIOACTIVE MATERIALS TO THE ENVIRONMENT The nuclear power unit design includes means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid waste produced during normal reactor operation, including anticipated operational occurrences.

Sufficient holdup capacity is provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such effluents to the environment.

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he RWS is designed to ensure that the general public and plant personnel are protected against osure to radioactive material in accordance with 10 CFR Part 20, Sections 1301 and 1302, and endix B and 10 CFR Part 50, Appendix I.

liquid and gaseous radioactive releases from the RWS are designed to be accomplished on a h basis. All radioactive materials are sampled prior to release to ensure compliance with CFR Part 20, Sections 1301 and 1302, and Appendix B and 10 CFR Part 50, Appendix I and to rmine release rates. Radioactive materials which do not meet release requirements will not be harged to the environment. The RWS is designed with sufficient holdup capacity and ibility for reprocessing of wastes to ensure release limitations are met.

RWS is designed to preclude the inadvertent release of radioactive material.

storage tanks in the clean liquid waste and gaseous waste systems are provided with valve rlocks which prevent the addition of waste to a tank which is being discharged to the ironment. Each discharge path from the RWS is provided with a radiation monitor which ts unit personnel and initiates automatic closure of redundant isolation valves to prevent her releases in the event of noncompliance to 10 CFR Part 20, Sections 1301 and 1302, and endix B.

tion 11.1.5 describes the plant design for the handling of solid wastes.

ITERION 61 - FUEL STORAGE AND HANDLING AND RADIOACTIVITY CONTROL The fuel storage and handling, radioactive waste and other systems which may contain radioactivity are designed to assure adequate safety under normal and postulated accident conditions. These systems are designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventory under accident conditions.

tems for fuel storage and handling, and all systems containing radioactivity are designed to ure adequate safety under normal and postulated accident conditions. Design of these systems described in the sections listed below:

stem Section actor Coolant System 4.0 gineering Safety Features Systems 6.0 1.A-35 Rev. 35

xiliary Systems 9.0 dioactive Waste Processing System 11.0 components important to the safety of these systems are located to permit periodic inspection equired. Suitable shielding, as described in Section 11.2, is provided for these components to ect plant personnel and to allow inspection and testing.

ensure the containment and confinement of radioactivity, all components are designed and ed in accordance with accepted Codes and Standards. All system components are visually ected and adjusted, if required, to ensure correct installation and arrangement. The completely alled systems were subject to acceptance tests or preoperation tests as described in Chapter 13 nsure the integrity of the systems.

spent fuel pool cooling system described in Section 9.5, is designed to ensure adequate decay t removal from stored fuel. Sections 5.4.3 and 9.5 describe how the spent fuel pool is designed revent significant reduction in fuel storage coolant inventory.

ITERION 62 - PREVENTION OF CRITICALITY IN FUEL STORAGE AND HANDLING Criticality in the fuel storage and handling system is prevented by physical systems or processes, preferably by use of geometrically safe configurations.

w fuel assemblies are stored in dry racks in parallel rows at elevation 38 feet 6 inches of the iliary building. The base of the new fuel racks at elevation 38 feet 6 inches minimizes the sibility of flooding the fuel assemblies. Nevertheless, the new fuel racks maintain a center to ter distance of 20.5 inches, large enough to prevent criticality in the unlikely event of flooding h unborated water. Additional details of new fuel storage are given in Sections 9.8.2.1.1and 4.1.1.

nt fuel assemblies are stored in parallel rows at the bottom of the spent fuel pool. The racks are arated into 4 regions, designated Regions 1, 2, 3, and 4.

l assemblies used at Millstone Unit 2 may include reduced enrichment fuel rods adjacent to de thimbles and reduced enrichment axial blanket regions. The criticality analyses are ormed using a single enrichment in all fuel rods that is the highest initial planar average 35 enrichment of the axial regions in the fuel assembly. This averaged enrichment is gnated as the initial planar average enrichment.

ion 1 can store, in a 2 out of 4 storage pattern, any fuel assembly with a maximum initial ar average enrichment up to 4.85 weight percent U-235. The other two locations in the 2 out storage pattern are designated as Restricted Locations (shown in Figure 9.8-7). Fuel storage locations designated as Restricted Locations in Figure 9.8-7 shall remain empty. No fuel 1.A-36 Rev. 35

dware/material of any kind may be stored in a Restricted Location.(1) ions 2 and 4 use fuel burnup credit and store fuel assemblies in a 3 out of 4 storage pattern, in ch the fourth location in a 2 x 2 storage array is designated as a Restricted Location per ure 9.8-7.

ions 1 and 2 contain Boraflex panels which are no longer credited as neutron absorbers.

ion 3 uses fuel burnup credit and has all storage locations available. In addition, fuel mblies stored in Region 3 must contain either three Borated Stainless Steel Poison Rodlets talled in the assembly's center guide tube and in two diagonally opposite guide tubes) or a full th, full strength Control Element Assembly (CEA).

re are also Non-standard Fuel Configurations in the spent fuel pool (SFP). A Non-standard l Configuration is an object containing fuel that does not conform to the standard fuel figuration. The standard fuel configuration is a 14 x 14 array of fuel rods (or fuel rods replaced un-enriched fuel rods or stainless steel rods) with five (5) guide tubes that occupy four lattice h locations each. Fuel in any other array is a Non-standard Fuel Configuration.

onstituted fuel in which one or more fuel rods have been replaced by either un-enriched fuel s or stainless steel rods is considered to be a standard fuel configuration.

e that each of the Non-standard Fuel Configurations must have a separate criticality analysis ch may allow storage in one or multiple Regions, and which may or may not require Borated nless Steel Poison Rodlets or a CEA if stored in Region 3.

C 62 states that the Criticality in the fuel storage and handling system shall be prevented by sical systems or processes, preferably by use of geometrically safe configurations. As iled above, the Region 1, 2, 3, and 4 storage racks, require more than just fuel geometry alone reactivity control. All four regions credit soluble boron in the spent fuel pool water. Regions 1, nd 4 credit Restricted Locations per Figure 9.8-7. Regions 2, 3, and 4 use fuel burnup credit.

ion 3 requires that fuel assemblies contain either three Borated Stainless Steel Poison Rodlets full length, full strength CEA (note that the criticality analysis of a given Non-standard Fuel figuration may qualify it for Region 3 storage without these inserts). Administrative controls used to ensure proper placements of Borated Stainless Steel Poison Rodlets and CEAs, use of ble boron and fuel burnup credit, and control of Restricted Locations. Further, for accident ditions, soluble boron is credited in the spent fuel pool water. The NRC has concurred that the it for these neutron poisons, soluble boron, fuel burnup credit, Restricted Locations, and ciated administrative controls are acceptable in meeting the requirements of GDC 62.

Note that Region 1 and 2 SFP rack storage locations contain removable Boraflex panel boxes which house the Boraflex panels. The Boraflex panel boxes were manufactured as an integral part the original SFP racks and as such are NOT stored components in SFP rack storage locations. Criticality analysis has shown that the Restricted Locations are acceptable with or without the Boraflex panel boxes.

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would approach criticality.

l handling equipment is designed to ensure safe handling of fuel assemblies and to prevent cality. Section 9.8.4 describes the safety features of the fuel handling equipment.

CRITERION 63 - MONITORING FUEL AND WASTE STORAGE Appropriate systems are provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.

tion 9.5.2.1 describes the monitoring and alarm instrumentation provided for the spent fuel age system to detect conditions that may result in loss of decay heat removal capability and essive radiation levels. Section 7.5.6 describes the monitoring provisions for radioactive waste dling and storage areas.

CRITERION 64 - MONITORING RADIOACTIVITY RELEASES Means are provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents.

tainment radiation is monitored by gaseous and particulate monitors as described in tions 7.5.1.2 and 7.5.6.3.

iation in effluent discharge paths and the plant environs are monitored as described in tions 7.5.6.2 and 7.5.6.3.

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