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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Millstone Power Station Unit 2 Safety Analysis Report Chapter1 MPS2 UFSAR 1-i Rev. 35 CHAPTER 1-INTRODUCTION AND

SUMMARY

Table of ContentsSection Title Page

1.1INTRODUCTION

...............................................................................................1.1-11.2

SUMMARY

DESCRIPTION..............................................................................1.2-1 1.2.1General........................................................................................................1.2-11.2.2Site .............................................................................................................1.2-1 1.2.3Arrangement ..............................................................................................1.2-21.2.4Reactor........................................................................................................1.2-21.2.5Reactor Coolant System..............................................................................1.2-31.2.6Containment System...................................................................................1.2-41.2.7Engineered Safety F eatures Systems..........................................................1.2-41.2.8Protection, Control and Moni toring Instrumentation.................................1.2-71.2.9Electrical Systems.......................................................................................1.2-7 1.2.10Auxiliary Syst ems.......................................................................................1.2-81.2.10.1Chemical and Volume Control System......................................................1.2-81.2.10.2Shutdown Cooling System..........................................................................1.2-91.2.10.3Reactor Building Closed C ooling Water System.......................................1.2-91.2.10.4Fuel Handling and Storage.......................................................................1.2-111.2.10.5Sampling System......................................................................................1.2-111.2.10.6Cooling Water Sy stems............................................................................1.2-111.2.10.7Ventilation Syst ems..................................................................................1.2-121.2.10.8Fire Protection System..............................................................................1.2-13 1.2.10.9Compressed Air Systems..........................................................................1.2-131.2.11Steam and Power Conversion System......................................................1.2-14 1.2.12Radioactive Waste Proc essing System.....................................................1.2-141.2.13Interrelation With Millstone Units 1 and 3...............................................1.2-151.2.14Summary of Codes and Standards............................................................1.2-171.3COMPARISON WITH OTHER PLANTS.........................................................1.3-1 1.4 PRINCIPAL ARCHITECTURAL AND ENGINEERING CRITERIA FOR DESIGN...............................................................................................................1.4-11.4.1Plant Design................................................................................................1.4-11.4.2Reactor........................................................................................................1.4-11.4.3Reactor Coolant and Au xiliary Syst ems.....................................................1.4-21.4.3.1Reactor Coolant System..............................................................................1.4-21.4.3.2Chemical and Volume Control System......................................................1.4-4 1.4.3.3Shutdown Cooling System..........................................................................1.4-51.4.4Containment System...................................................................................

1.4-5 MPS2 UFSAR Table of Contents (Continued)

Section Title Page 1-ii Rev. 351.4.5Engineered Safety Features Systems..........................................................1.4-61.4.6Protection, Control and Instrumentation System........................................1.4-61.4.7Electrical Systems.......................................................................................1.4-71.4.8Radioactive Waste Proc essing System.......................................................1.4-71.4.9Radiation Prot ection...................................................................................1.4-71.4.10Fuel Handling and Storage.........................................................................1.4-71.5RESEARCH AND DEVELOPM ENT REQUIREMENTS................................1.5-11.5.1General........................................................................................................1.5-11.5.2Fuel Assembly Flow Mixing Tests.............................................................1.5-11.5.3Control Element Assembly Drop Tests......................................................1.5-21.5.4Control Element Drive Assembly Performance Tests................................1.5-21.5.5Fuel Assembly Flow Tests..........................................................................1.5-31.5.6Reactor Vessel Fl ow Tests..........................................................................1.5-41.5.7In-core Instrumentation Tests.....................................................................1.5-41.5.8Materials Irradiati on Surveillance..............................................................1.5-51.5.9References...................................................................................................

1.5-51.6IDENTIFICATION OF CONTRACTORS.........................................................1.6-11.6.1References...................................................................................................

1.6-11.7GENERAL DESIGN CHANGES SINCE ISSUANCE OF PRELIMINARY SAFETY ANALYSIS REPORT.........................................................................1.7-11.7.1General........................................................................................................1.7-11.7.2Control Element Drive Mechanisms...........................................................1.7-11.7.3Radioactive Waste Proc essing System.......................................................1.7-11.7.3.1Clean Liquid Waste Processing System.....................................................1.7-11.7.3.2Gaseous Waste Pro cessing System.............................................................1.7-11.7.4Vital Component Closed C ooling Water Sy stem.......................................1.7-21.7.5Electrical.....................................................................................................1.7-21.7.5.1AC Power....................................................................................................1.7-21.7.5.2Diesel Genera tors........................................................................................1.7-21.7.5.3DC Supply...................................................................................................1.7-2 1.7.5.4Instrument Power........................................................................................1.7-31.7.6Axial Xenon Oscillation Protection............................................................1.7-31.7.7Number of Control Element Asse mblies and Drive Mechanisms..............1.7-31.7.8Burnable Poison Shims...............................................................................1.7-3 1.7.9Structures....................................................................................................

1.7-3 MPS2 UFSAR Table of Contents (Continued)

Section Title Page 1-iii Rev. 351.7.10High Pressure Safety Injection Pumps........................................................1.7-41.7.11Containment Purge Valve Isolation Actuation System..............................1.7-41.7.12Control Element Drive System...................................................................1.7-41.8ADVISORY 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.]..............................................................1.8-11.8.1General........................................................................................................1.8-11.8.1.1Ability of Fuel to Withstand Transien ts at End of Life and Experimental Verification of Maximum Linear Heat Genera tion Rate............................1.8-11.8.1.2Fuel Integrity Following a Loss-of-Coolant Accident................................1.8-1 1.8.1.3Primary System Quality Assurance and In-Service Insp ectability.............1.8-21.8.1.4Separation of Control and Pr otective Instrumentation...............................1.8-31.8.1.5Instrumentation for Detect ion of Failed Fuel.............................................1.8-31.8.1.6Effects of Blowdown Forces on Core and Primary System Components..1.8-41.8.1.7Reactor Vessel Th ermal Shock...................................................................1.8-41.8.1.8Effect of Fuel Rod Failure on the Capability of the Safety Injection System

.....1.8-51.8.1.9Preoperational Vibration Monitoring Program...........................................1.8-51.8.1.9.1Basis of Pr ogram.........................................................................................1.8-51.8.1.9.2Millstone Unit 2 Program...........................................................................1.8-61.8.2Special for Millstone Unit 2........................................................................1.8-7 1.8.2.1Release of Radioactivity in Case of Damaged Fuel Assemblies in Spent Fuel Pool.............................................................................................................1.8-71.8.2.2Hydrogen Cont rol.......................................................................................1.8-71.8.2.3Common Mode Failures and Anticipa ted Transients Without Scram........1.8-71.8.3References...................................................................................................

1.8-81.9TOPICAL REPORTS..........................................................................................1.9-1 1.10 MATERIAL INCORPORATED BY REFERENCE........................................1.10-11.AAEC GENERAL DESIGN CRITERIA FOR NUCLEAR POWER PLANTS..1.A-1 MPS2 UFSAR 1-iv Rev. 35 CHAPTER 1-INTRODUCTION AND

SUMMARY

List of Tables Number Title1.1-1Licensing History1.2-1Summary of Codes and Standards for Co mponents of Water-Cooled Nuclear Power Units (1)1.3-1Comparison with Other Plants 1.4-1Seismic Class I Systems and Components 1.8-1Comparison of Preoperational Vibration Monitoring Program Design Parameters1.9-1Topical Reports MPS2 UFSARNOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

1-v Rev. 35 CHAPTER 1 - INTRO DUCTION AND

SUMMARY

List of Figures Number Title1.2-1Site Layout1.2-2Plot Plan 1.2-3General Arrangement, Turbin e Building Plan at Operati ng Floor Elevation 54 Feet 6 Inches1.2-4General Arrangement, Turbine Building Plan at Mezzanine Floor Elevation 31 Feet 6 Inches1.2-5General Arrangement, Turbine Building Pl an at Ground Floor Elevation 14 Feet 6 Inches1.2-6General Arrangement Containment Plan at Floor Elevation 14 feet 6 inches and Elevation 36 feet 6 inches1.2-7General Arrangement Auxiliary Building Plan at Elevation 36 feet 6 inches and Elevation 38 feet 6 inches1.2-8General Arrangement Auxiliary Bu ilding Sections "G-G" and "H-H"1.2-9General Arrangement Auxiliary Building Ground Floor Elevati on 14 feet 6 inches and Cable Vault Elevation 25 feet 6 inches1.2-10General Arrangement Containment and A uxiliary Building Plan at Elevation (-)5 feet 0 inches and Elevation (-)3 feet 6 inches1.2-11General Arrangement Containment and Auxi liary Building Plan at Elevation (-)25 feet 6 inches and Elevat ion (-)22 feet 6 inches1.2-12General Arrangement Containment and Auxi liary Building Plan at Elevation (-)45 feet 6 inches1.2-13General Arrangement Containmen t and Auxiliary Building Section "A-A"1.2-14General Arrangement Containmen t and Auxiliary Building Section "B-B"1.2-15General Arrangement Turbine Bu ilding Sections "C-C" and "E-E"1.2-16General Arrangement Turbine Building Sections "D-D" and "F-F"1.2-17General Arrangement Intake Structure Auxiliary Steam Boiler Room Plan and Section MPS2 UFSAR1.1-1Rev. 35CHAPTER 1 - INTROD UCTION AND

1.1 INTRODUCTION

This Final Safety Analysis Repor t (FSAR) was initially submitted in support of the application of The Connecticut Light and Power Company (CL&

P), The Hartford Electric Light Company (HELCO), Western Massachusetts Electric Company (WMECO), and Northeast Nuclear Energy Company (NNECO), for a license to operate th e second nuclear powered generating unit at the site of the Millstone Power Station. Since the in itial licensing of the unit, unless otherwise indicated, the FSAR has be en updated a number of times to reflect current design and analysis information. On the basis of the information presented in the FSAR and referenced material at the time of application for operati ng license, the applicants concl uded that Millst one Unit 2 is designed and constructed a nd will be operated without undue risk to the heal th and safety of the public.Construction of Millstone Unit 2 was authorized by the United States Atomic Energy Commission (AEC) when it issued Provisional Construction Permit CPPR-76 on December 11, 1970. Commercial operation of Millstone Unit 2 commenced in December 1975 at a gross electrical output of 865 megawatts.Millstone Unit 2 is located Millstone Point in the Town of Wate rford, Connecticut. It is located immediately to the north of the first unit (Millstone Unit 1) and south of the thir d unit (Millstone Unit 3). Commercial operation of Millstone Unit 1 was auth orized by the AEC by issuing Provisional Operating License DPR-21 on Oct ober 7, 1970. Commercial ope ration of Millstone Unit 1 commenced in December, 1970. Commercial operation of Millstone Unit 3 was authorized by the United States Nuclear Re gulatory Commission (NRC) (forme rly the AEC) by issuing the Low Power License on November 25, 1985, and the Full Power License on January 31, 1986.

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

Millstone Unit 2 utilizes a pressurized water nuclear steam supply system (NSSS). The unit is similar, in this respect, to the former Yankee Atomic Electric Co mpany generating plant in Rowe, Massachusetts, (NRC Docket Nu mber 50-29), the former Hadda m Neck Plant operated by the Connecticut Yankee Atomic Powe r Company on the Connecticut Ri ver at Haddam, Connecticut (NRC Docket Number 50-213), and the Maine Yankee Atomic Power Company plant at Wiscasset, Maine (NRC Docket Number 50-309). The NSSS for Mi llstone Unit 2 is supplied by Combustion Engineering, Inc. (CE) which also supplied the steam supply system for the Maine Yankee plant. The Millst one Unit 2 NSSS is similar to the sy stems supplied by CE for the initial two units of the Baltimore Gas and Electric Calvert Cliffs Nuclear Power Plant (NRC Docket Numbers. 50-317 and 50-318).

Millstone Unit 2 has be en designed to operate safely unde r all normal operati ng conditions and anticipated transients. Although th e unit produces small am ounts of radioactive waste, the offsite disposal of these wastes is rigidly controlled and maintained below established limits.

MPS2 UFSAR1.1-2Rev. 35 In 2001, Millstone Units 1, 2 and 3 operating licenses were transf erred from Northeast Nuclear Energy Company to Dominion Nucl ear Connecticut, Inc. (DNC).

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

The transmission and distribution assets on the site will conti nue to be owned by Connecticut Light and Power (CL&P) and wi ll be operated under an Inte rconnection Agreement between CL&P and DNC.

The FSAR will retain references to Northeast Utilities and Northeast Nuclear Energy Company documents/activities when they are used in a historic context and are required to support the plant licensing bases.Upon license transfer, all reco rds and design documents necessary for operation, maintenance, and decommissioning were transferred to DNC. Some of thes e drawings are included (or referenced) in this FSAR. These drawings often have title blocks (or drawing numbers) which list Northeast Nuclear Energy Company (et. al) or Northeast Utilities Service Company (et. al). In general, no changes to these title blocks will be made at this time. Based on this general note, these drawings shall be read as if the title blocks list Do minion Nuclear Connecticut, Inc.

Millstone Unit 2 has been designed to operate re liably without accident.

Nevertheless, to ensure that no reasonably credible accide nt could result in dangerous rele ases of radioactive material, the unit incorporates a number of feat ures designed to minimize the ef fects of such an accident. The adequacy of these safety featur es under the conditions of various postulated accidents is discussed in Chapter 14.

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

Millstone Unit 2 is currently li censed for a steady state reacto r core power level of 2700 MWt, corresponding to a NSSS rating of 2715 MWt. All Chapter 14 analys es have been evaluated on the basis of these current values.

Since the construction perm it was issued, and duri ng the design and construc tion of the unit, there have been no major deviations fr om the information supplied in th e Preliminary Safety Analysis Report (PSAR). However, changes in various specific design feat ures have been found desirable and these are covered in the appropriate sections of this report. A summary of the more significant design changes incorporated in the plant since the issuance of the PSAR up to the time of application for an operating lice nse is provided in Section 1.7.

MPS2 UFSAR1.1-3Rev. 35TABLE 1.1-1 LICENSING HISTORY EVENTDATE Construction Permit IssuedDecember 11, 1970 Final Safety Analysis Report FiledAugust 15, 1972Full Term Operating Licens ing Issued September 26, 1975 Full Power License September 26, 1975 Initial Critical ityOctober 17, 1975 100% PowerMarch 20, 1976 Commercial OperationDecember 26, 1975"Stretch Power" June 25, 1979 Operating License Extensi on Requested December 22, 1986 Operating License Extension Issued January 12, 1988Full Term Operating License ExpiresDecember 11, 2010 Operating License ExpiresJuly 31, 2035 MPS2 UFSAR1.2-1Rev. 35 1.2

SUMMARY

DESCRIPTION 1.2.1 GENERAL A summary description of Millstone Unit 2 of the Millstone Nuclear Power Stat ion is provided in this section. The description includes the following:

a.Siteb.Arrangementc.Reactor d.Reactor coolant systeme.Containment systemf.Engineered safety features systems g.Protection, control and instrumentation systemh.Electrical systems i.Auxiliary systems j.Steam and power conversion systemk.Radioactive waste processing systeml.Interrelation with Millstone Units 1 and 3 m.Summary of Codes and StandardsWithheld under 10 CFR 2.390 (d) (1) 1.2.2 MPS2 UFSAR1.2-2Rev. 35 The containment houses the NSSS, consisting of the reactor, stea m generators, reactor coolant pumps, pressurizer, and some of the reactor auxiliaries. The c ontainment is equipped with a polar crane. The enclosure building completely envelopes the containment and provi des a filtration region between the containment and the environment.

The turbine building houses the turbine generator, condenser, feedwater he aters, condensate and feedwater pumps, turbine auxiliaries and certain of the switchgear assemblies.

1.2.4 REACTORThe reactor is a pressurized light water cooled and moderated type fueled by slightly enriched uranium dioxide. The uranium dioxide is in the fo rm of pellets and is c ontained in pressurized Zircaloy-4 tubes fitted with welded end caps. These rods are arranged into fuel assemblies each consisting of 176 fuel rods arranged on a 14 rod square matrix. Space is left in the fuel rod array to allow for the installation of five guide tubes. These guide tubes provide for the smooth motion of control element assembly fingers. The assembly is fitted with end fittings and spacer grids to maintain fuel rod alignment and to provide structural support. The end fittings are also drilled with flow holes to provide for the flow of cooling water past the fuel tubes. Withheld under 10 CFR 2.390 (d) (1) 1.2.3 Withheld under 10 CFR 2.390 (d) (1)

MPS2 UFSAR1.2-3Rev. 35 The reactor is controlled by a combination of chemical shim and solid absorber. The solid absorber is boron carbide pellets or stainless steel contained in tubular Inconel elements. Some earlier elements had used stainless steel as the absorber material. Five absorber elements are connected together by a spider yoke in a square matrix with a cen ter element. The five elements constitute a control element assembly (CEA). Th e 73 CEAs are connected, either singly or dually, through extension shafts, to 61 magnetic jack t ype control element driv e mechanisms (CEDMs) which are mounted on nozzles on th e reactor vessel head. Each CEA is aligned with and can be inserted into the guide tubes of fuel assemblies. The dual CEAs are utilized for shutdown rods.

The single CEAs are divide d into regulating groups.

The eight part length control rods of Cycle One were replaced by dummy flow plugs. Two of the flow plugs were replaced by reactor vessel level indication system detectors, then in Cycle Twelve, the last six remaining fl ow plugs were removed. The resulting increase in core bypass flow has been accounted for in the safety analysis.

The replacement head has a total of 78 nozzle penetrations. 67 of these nozzles are suitable for supporting control element drive mechanisms (61 ar e in use, while the othe r 6 nozzles are capped with nozzle adapters). Two nozzl es are used for heated juncti on thermocouples, which enable monitoring reactor vessel between th e top of the vessel dome and the area directly above the fuel bundles. Eight nozzles are used for nuclear instrumentation and one nozzle is used for the reactor vessel head vent. The location, size and the num ber of nozzles on the re placement reactor vessel closure head are maintained in the same c onfiguration as before (prior to cycle 16).

Chemical shim control is provide d by boric acid dissolved in the coolant water. The concentration of boric acid is maintained and controlled as required by the chemical a nd volume control system.

The reactor core rests on the core support plate assembly which is suppor ted by the core support barrel. The core support ba rrel is a right circular cylinder supported from a machined ledge on the inside surface of the vessel flange forging. The support plate assemb ly transmits the entire weight of the core to the core support barrel through a structure made of beams and vertical columns.

Surrounding the core is a shroud which serves to limit the coolant which bypasses the core. An upper guide structure, consisting of upper support st ructure, control elemen t assembly shrouds, a fuel alignment plate and a spacer ring, serves to support and align the upper ends of the fuel assemblies, prevents lifting of the fuel assemblies in the even t of a loss-of-coolant accident (LOCA) and maintains spacing of the CEAs. Chapter 3 contains more detailed information on the reactor.1.2.5 REACTOR COOLANT SYSTEM The reactor coolant system consists of two closed heat transfer l oops in parallel with the reactor vessel. Each loop contains one steam genera tor and two pumps to circulate coolant. An electrically heated pressurizer is connected to one loop hot leg. The coolant system is designed to operate at a thermal power level of 2715 MWt to produce steam at a nominal pressure of 880 psia.

The reactor vessel, loop piping, pr essurizer and steam generator pl enums are fabricated of low alloy steel, clad internally with austenitic stainless steel. The pressurizer surge line and coolant pumps are fabricated from stainless steel and the steam generator tubes are fa bricated from Inconel.

MPS2 UFSAR1.2-4Rev. 35 Overpressure protection is provided by power-operated relief va lves and spring-loaded safety valves connected to the pressurizer. Safety and relief valve discharge is released under water in the quench tank where the steam discharge is condensed. The two steam generators are vertical shell and U-tube steam generators each of which produces 5.9 x 10 6 lb/hr of steam. Steam is generated in the shell side of the stea m generator and flows upward through moisture separators. Steam outlet moisture content is less than 0.2 percent.

The reactor coolant is circulated by four electric motor-drive n, single-suction, centrifugal pumps.

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

1.2.6 CONTAINMENT SYSTEM A double containment system is used for Unit 2. The containment syst em consists of a prestressed concrete cylindrical structure referred to as the containment, which is completely enclosed by the enclosure building (EB). The enclosure buildi ng filtration region (EBF R) includes the region between the containment and the enclosure buildi ng, the penetration rooms and engineered safety feature equipment rooms. In the unlikely event of a LOCA the EBFR is main tained at a slightly negative pressure by the enclosure building filtra tion system (EBFS). Air in the EBFR would be processed through charcoal filters and released through the 375 foot Millstone stack during a LOCA.The containment uses a prestresse d post-tensioned concrete design.

The containment is a vertical right cylindrical structure with a dome and a flat base. The interior is lined with carbon steel plate to further ensure leak tightness.

Inside the containment, the r eactor and other NSSS components are shielded with concrete.

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

The containment, in conjunction with the engineered safety featur es, is designed to withstand the highest internal pressure and co incident temperature resulting fr om the main steam line break accident (Section 14.8.2). The structural design conditions are for an internal pressure of 54 psig and a coincident equilibrium temperature of 289

°F. The enclosure building is a limit ed leakage steel framed structure partially supported off the containment and auxiliary building with uninsulated metal siding and an insulated metal roof

deck. 1.2.7 ENGINEERED SAFETY FEATURES SYSTEMS The engineered safety features systems (ESF S) provide protection fo r the public and plant personnel against the incidental release of ra dioactive products from the reactor system, particularly as a result of postulated LOCA. These safety features localize, control, mitigate and MPS2 UFSAR1.2-5Rev. 35terminate such accidents to hold exposure levels below the applicable limits of 10 CFR Part 50.67. The engineered safety features consist of the following systems: a.Safety injection b.Containment sprayc.Containment air recirculation and coolingd.Enclosure building filtration e.Hydrogen controlf.Auxiliary feedwater automatic initiation system Each of these systems is divide d into two redundant independent subsystems which in turn are powered by the associated re dundant independent emer gency electrical subsystem (see Section 1.2.9). The first three are cooled by the associated redundant independent reactor building closed cooling water h eaders (see Section 1.2.10.3).

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

Four safety injection ta nks are provided, each connected to one of the four reactor inlet lines. The volume of each tank is 2019 cubic feet. Each tank contains about 1 100 cubic feet of borated water at refueling concentration and is pressurized with nitrogen at 200 psig. In the event of a LOCA, the borated water is forced into the reactor coolant system by th e expansion of the nitrogen. The water from three tanks adequately cools the entire core. Borated water is injected into the same nozzles by two low pressure and three high pressure injection pumps taki ng suction from the refueling water storage tank (RWST). For maximum reliability, the design capacity from the combined operation one high pressure and one lo w pressure pump provides adequate injection flow for any LOCA; in the event of a design basis accide nt (DBA), at least one high pressure and one low pressure pump will receive power from the emergency po wer sources if preferred power is lost and one of the emergency diesel generators is assumed to fail. When the refueling water storage tank supply is nearly depleted, the high pressure pump suctions automatically transfer to the containment sump and the low pressure pumps are shut down. One high pressure pump has sufficient capacity to cool the co re adequately at the start of recirculation. Duri ng recirculation, heat in the recirculating wate r is removed through the shutdown cooling heat exchangers via either the low pressure injection pumps or containment spray pumps.

MPS2 UFSAR1.2-6Rev. 35 The safety injection pumps are located outside the containment to permit access for periodic testing during normal operation. The pumps discharg e into separate header s which lead to the containment. Test lines are provi ded to permit running the pumps for test purposes during plant operation.

The safety injection system is designed in acco rdance with AEC Genera l Design Criteria 35, 36, and 37 in Appendix A to 10CFR50 and General Criteria as described in Section 6.1. An analysis of the performance of the safety injection system (emergency core cooling syst em) following a postulated LOCA is given in Section 14.6. Two independent, full capacity systems are provided to remove heat from the containment atmosphere by containment sprays and/or air recirculation a nd cooling after the postulated LOCA.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 recirc ulation 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 cont ainment 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 cool ed by one of two associated redundant reactor building cooling water and service water subsyste ms. They are powered by the as sociated emergency electrical subsystem.

The enclosure building filtration sy stem would collect and filter al l potential containment leakage and minimize environmental radioactivity levels resulting from the discharge of all sources of containment leakage into the encl osure building filtratio n region in the unlikely event of a LOCA.

The enclosure building filt ration system would also collect and filter any radioactive releases in the unlikely event of a fuel handli ng accident inside the containmen t or spent fuel pool areas (see Section 6.7).

The hydrogen control system is pr ovided to mix and monitor the concentration of hydrogen gas within the containment. This system consists of the post-accident recirc ulation system for mixing the containment environment and the hydrogen monitoring system for continuous monitoring of the post-accident containmen t atmosphere. The hydrogen purge system and hydrogen recombiners which are not credited in accident analyses are provided for reducing containment hydrogen concentrations.

MPS2 UFSAR1.2-7Rev. 35The auxiliary feedwater automatic initiation system, (AFAIS), is pr ovided to ensure delivery of sufficient feedwater to the steam generators in event of the loss of main feedwater. This system automatically actuates two motor driven auxiliary feedwater pumps (see Section 10.4.5.3), and opens the two auxiliary feedwater flow control valves via the automatic initiation control circuitry (see Section 7.3.2.2.h). The AFAIS is actuated upon completion of a 2-out-of-4 logic matrix initiated by a low steam generator level. Upon recei pt of an actuation signal both pumps are started and the flow control valves to both steam generators are opened (see Section 7.3).

1.2.8 PROTECTION, CONTROL AND MONITORING INSTRUMENTATION Various instrumentation systems provide protection, control, a nd monitoring functions for the safe and efficient operation of Millstone Unit 2.Protection instrumentation system s function to shut down the reac tor and activate safety systems if continuously monitored key plant process parameters exceed predetermined limits. Specific protection instrumentation syst ems include the Reactor Protective System (RPS) and the Engineered Safety Features Actuation System (ESFAS). The RPS functions to shut down or trip the reactor if any two of four safety channels generate co incident trip signals. An RPS trip removes power from the r eactor control rods, allowi ng them to drop into the reactor, and shut it down. The ESFAS functions to actuate the engineered safety featur es systems described in FSAR Section 1.2.7. The exception to this is the containment purge valv e isolation where one of four containment air radiation detectors can generate a trip signal. Actu ation of the ESFS occurs if any two of four safety channels ge nerate coincident trip signals.

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

Indications are provided to monitor normal and abnormal plant operation. Indicators are located within the control room and thr oughout the plant. The indicators ar e used to monitor the status and operation of the protective and control syst ems, and the status of other support systems.

Major indication systems include the Control Element Assembly (CEA) Position Indication, Nuclear Instrumentation (NI), In-Core Instrument ation (ICI), Radioactiv ity Monitoring System (RMS), Integrated Computer Sy stem (ICS), Control Room A nnunciation, and Post Accident Monitoring Instrumentation (PAMI).

Details of the above and other protective, control, and monito ring instrumentation systems are provided in Chapter 7.

1.2.9 ELECTRICAL SYSTEMSThe Millstone Nuclear Power Sta tion consists of Millstone Unit 1 which is no longer generating power, Millstone Unit 2 with a 1011-MVA, 0.90 power factor generator, and Millstone Unit 3 with a 1354.7-MVA, 0.925 power factor generator (see Chapter 8).

MPS2 UFSAR1.2-8Rev. 35 The Millstone Unit 2 generator output is fed through a step up transformer bank to the 345 kV switchyard. The switchyard is connected to the high voltage tr ansmission system through four 345 kV transmission lines. The switchyard, in a ddition to carrying the el ectrical output of the station, also provides a means of supplying power to the units from external sources. Startup power and reserve auxili ary power for Millstone Unit 2 are taken from the 345 kV switchyard through the reserve station service transformer. Normal station se rvice power is taken from the generator main leads through the normal station service transformer. A second source of off site power for the engineered safety features is pr ovided from normal stat ion service transformer 15G-3SA or reserve station servi ce transformer 15G-23SA, both associated with Millstone Unit 3 via a 4160V crosstie connection. Tw o diesel generators provide the on site emergency power for Millstone Unit 2. The 4160V crosstie from Unit 3 can also be conf igured (by operator action) to supply power directly from the Unit 3 Alternate AC (SBO) diesel ge nerator to provide an alternate AC source for Unit 2 Appendix R and Station Blackout requirements.Auxiliary power for Millstone Unit 2 is provided at 6900, 4160, 480, and 120/208 volts. Direct current 125 volt systems are also available for emergency power, engineered safety feature control, and essential nuclear in strumentation, control and relaying. The preferred and on site emergency sources of electrical power are each adequate to permit prompt shutdown and maintain safe conditions under all credible circumstances. The on site emergency power source consists of two separate and redundant dies el generators. Each diesel is capable of carrying all required auxiliary loads following postulated LOCA with out exceeding its continuous rating.

Each of the two separate and re dundant station batterie s is capable of carry ing essential 125 volt DC and 120 volt AC inverter loads associated with a postulated LOCA.

The redundant channel wiring associated with these emergency el ectrical sources is physically separated.

1.2.10 AUXILIARY SYSTEMS 1.2.10.1 Chemical and Volume Control SystemThe chemistry of the reactor coolan t is controlled by purif ication of a regulate d letdown stream of reactor coolant. Wate r removed from the reactor coolant system is cooled in the regenerative heat exchanger. The fluid pressure is then reduced and flow is regulat ed by the letdown control valves. Temperature is reduced further in the letdown heat exchanger.

From there, the flow passes through a filter and a purificat ion ion exchanger to remove corrosion and fi ssion products. A small fraction of the flow is dive rted prior to entering the ion exchanger. This stream of coolant flows through a process radiation monitor. Upon leaving the ion exchanger, the coolant flows through a strainer and another filter and is then sprayed into the volume control tank.

Coolant is returned to the reactor coolant system by the chargi ng pumps, through the regenerative heat exchanger. Prior to entering the charging pum ps, the coolant boron conc entration is adjusted MPS2 UFSAR1.2-9Rev. 35to meet the reactor reactivity requirements. In addition, provision is made to inject chemical additives to the suction of the charging pumps for coolant chemistry control. The volume control system automatically controls the rate at which c oolant must be removed from the reactor coolant system to maintain the pressurizer level within the prescribed control band, thereby compensating for ch anges in volume due to coolant temperature changes. Using the volume control tank as a surge tank decreases th e quantity of liquid and gaseous wastes which would otherwise be generated.

Reactor coolant system makeup wa ter is taken from the primary water storage tank and the two concentrated boric acid storage tanks. The boric acid solution is maintained at a temperature which prevents crystallization. The makeup wa ter is pumped through the regenerative heat exchanger into the reactor coolant loop by the charging pumps. Boron concentration in the reactor coolant system can be reduced by dive rting the letdown flow away from the volume control tank to the radioactive waste processing system. Demineralized water is then used for makeup.

When the boron concentration in the reactor coolant system is low, the feed and bleed procedure previously described would genera te excessive volumes of waste to be processed. Therefore, the chemical and volume control system is equipped with a deborati ng ion exchanger which reduces boron concentration late in cycle life. A complete descript ion is given in Section 9.2.

1.2.10.2 Shutdown Cooling SystemThe shutdown cooling system (see Section 9.3) is used to reduce the reactor coolant temperature, at a controlled rate, from 300

°F to a refueling temperature of approximately 130

°F. It also maintains the proper reactor cool ant temperature during refueling. Once entry conditions are met, the shutdown cooling system can provide long term cooling capability in the event of a LOCA after the reactor coolant system has refilled (see Section 14.6.5.3).

The shutdown cooling system utilizes the low pressure safety in jection pumps to circulate the reactor coolant through two shutdown cooling heat exchangers. It is returned to the reactor coolant system through the low pressure safety injection header. The reactor building closed cooling water system (RBCCW) s upplies cooling water for the shutdown heat exchangers.

1.2.10.3 Reactor Building Closed Cooling Water System The RBCCW system consists of two separate i ndependent headers, each of which includes a RBCCW pump, a service water (s eawater)-cooled RBCCW heat exchanger , interconnecting piping, valves and controls. A third RBCCW pump and a third RBCCW heat exchanger are provided as installed spares. The co rrosion inhibited, demineralized wa ter in this closed system is circulated through the RBCCW heat ex changer where it is cooled to 85

°F by seawater which has a maximum design inlet temperature of 80

°F (see Section 9.4).

MPS2 UFSAR1.2-10Rev. 35The RBCCW system removes heat from the containment atmosphere , engineered safety feature components and various a uxiliary system/components handling the reactor c oolant. Items cooled by the RBCCW system include:

Containment air recirculation and cooling unit Reactor vessel support c oncrete cooling coils Containment spray pump seal coolers High and low pressure safety injection pump seal coolers

Shutdown cooling heat exchangers

Engineered safety feature r oom air recirculation coilsReactor 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 Each of the independent header s supply cooling water to compone nts in the associated redundant safety related sub-systems (see Section 1.2.7). The RBCCW heat ex changers, connected to each independent RBCCW headers, are c ooled by the associated independent service water header (see Section 1.2.10.6). Components in each independent RBCCW header, th e associated safety related subsystems, and the associated service water h eader are powered from the associated redundant independent emergency electrical power subsystem (see Section 1.2.9).

Remote manually operated valves allow the spar e RBCCW pump and/or heat exchanger to be operated with either of the two independent headers. The RBCCW surge tank absorbs the volumetric changes caused by temperature changes of the water within the RBCCW headers.

A chemical addition system is provided for the RBCCW system to maintain the corrosion inhibitor concentration as required.

During normal plant operation and normal shutdow n, both of the independent RBCCW headers are in service.

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

MPS2 UFSAR1.2-11Rev. 35 1.2.10.4 Fuel Handling and Storage The fuel handling systems provide for the safe handling of fu el assemblies and control element assemblies and for the required assembly, disassembly, and storage of the reactor vessel head and internals. These systems include a refueling m achine located inside the containment above the refueling pool, the fuel tr ansfer carriage, the upending machines , the fuel transfer tube, a fuel handling machine over the spent fuel pool, a new fuel elevator in the spent fuel pool, a spent fuel cask crane, a new fuel inspecti on machine in the fuel handling ar ea of the auxiliary building, and various devices used for handling the reactor vessel head and inte rnals (see Section 9.8).

New fuel is stored dry in vertic al racks within a storage vault near the spent fuel pool in the auxiliary building. Storage space is provide d for approximately one-third of a core.

The vault is designed to avoid cr iticality by spacing fu el assemblies at 20.5 inches, center to center. The spent fuel pool, located in the auxiliary building, is constructed of reinforced concrete lined with stainless steel. The spent fuel storage racks are separated into four regions, designated Regions 1, 2, 3, and 4. Section 9.8.2.1 contains a detail ed description of spent fuel storage design and components.

Cooling and purification equipmen t is provided for the spent fuel pool water (see Section 9.5).

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

1.2.10.5 Sampling System The sampling system consists of Sampling Stations 1 and 2, the Post Accident Sampling System (PASS), the Corrosion Monitoring Sample Station, and the Waste Ga s Sample Sink. These provide the means for determining physical, chem ical and radioactive conditions of process fluids, waste gas and containment air. The system is supplemented by inde pendent sampling of nonradioactive fluids in numerous locations within the unit, including samp ling of the chlorinated water. (See Section 9.6.)

1.2.10.6 Cooling Water SystemsThe exhaust steam from the main turbine and steam generator feedwater pump turbines is condensed in the condenser, whic h is cooled, in turn, by circul ating water flow ing through the condenser tubes, (see Section 9.7.1).

Four circulating water pumps, with 548,800 gpm total capacity, take suction from and discharge to Long Island Sound. The circulating water system is designed to main tain condenser back pressure at 2 inches Hg absolute with a 60.8

°F inlet circulating water temperature.

The service water system (see Section 9.7.2) provides cooling water to the RBCCW, TBCCW, diesel engine cooling water, chilled water system heat exchangers , vital switchgear room cooling coils and the circulating water pum p bearings. Three vertical, centrifugal, half capacity service MPS2 UFSAR1.2-12Rev. 35 water pumps have a design flow of 12,000 gpm, each with a total dynamic head of 100 feet of water. These pumps take suction from and discharge to Long Island Sound.

The service water system consis ts of two redundant, independent cross-connected supply headers with isolation valves to all heat exchangers and two discharge headers for the RBCCW heat exchangers. Two discharge headers exist for the emergency diesel generator cooling water; once underground these headers combine prior to entering the discharge canal. Service Water discharge from the TBCCW, chilled water system and vital switchgear room cooling heat exchangers combine into a common header prior to entering the discharge cana

l. Each of the supply headers is supplied by one of the se rvice water pumps. During norma l operation and shutdown and following a postulated LOCA, the two pumps conne cted to the two redunda nt supply headers are in service. However, only one service water pump a nd header is required to provide cooling of the RBCCW and diesel following a LOCA or for unit shutdown. Remo te manually operated valves allow the third service water pump to be connected to either of the redundant headers.

The intake structure consists of four independent bays. The intake structure is equipped with a chlorination system, consisting of two 1800 gall on sodium hypochlorite st orage tanks and two injection systems with one suppl ying sodium hypochlorite to the service water sy stem and the other to the circulating water intake.

1.2.10.7 Ventilation Systems Normally the containment environm ent is cooled by the containmen t air recirculation and cooling system. Following a postulated LOCA, these units reduce the temperature and pressure of the containment atmosphere to a safe level (s ee Sections 1.2.7, 6.5 and 9.9.1). The containment auxiliary circulation fans maintain uniform containment environmental temperature by mixing the air. Normally, the environment for the control element drive mechanisms is maintain by the CEDM fan-coil units. A forced outside air purge system is provided to maintain a suitable environment within the containment whenever access is desired. The exhaust of this containment air purge system is monitored to assure that radioactive effluents are maintained within acceptable limits. The auxiliary building is served by separate ventilation systems in the fuel handling area, the radioactive waste area a nd for the nonradioactive waste area. Ea ch area is provide d with a heating and ventilating supply unit and sepa rate exhaust fans. Exhausts from the potentially contaminated areas are filtered through high efficiency particulate air (HEPA) filters, monitored, and discharged through the Unit 2 stack. Exhaust from clean areas is discharged di rectly to the atmosphere (see Section 9.9.6).

Handling of irradiated fuel or moving a cask over the spent fu el pool does not require fuel handling area integrity or ventilation but it may be desirabl e to use the main exhaust or EBF systems, if available, as the exhaust discharge paths. If boundary integrity is set then these discharge paths provide a monito red radiological release pathway. If boundary integrity is not assured then suitable radiological monitoring is recommended per the Millstone Effluent Control Program.

MPS2 UFSAR1.2-13Rev. 35 The ventilation systems (m ain exhaust and EBFS) are normally av ailable to provid e for a filtered and monitored release pathway for effluents from the fuel handli ng area. If ventilation is not available, releases from the fu el handling area are m onitored per the Millstone Effluent Control Program to ensure appropriate radiological effluent controls are in place.Two full capacity and redundant air conditioning systems are provide d for the control room. In the event of an accident, a bypass thr ough either of the two full capacity and re dundant control room filtration systems, which contain charcoal filters, is provided to protect control room operating personnel from exposure to high radiation levels.

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

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

1.2.10.8 Fire Protection SystemThe fire protection systems' (see Section 9.10) function to protect pers onnel, structures, and equipment from fire and smoke.

The fire protection systems have been designed in accordance with the applicable National Fire Protection Association (NFPA) Codes and Standards, regulatory requirements, industry standard s, and approved procedures. Th e design of the various fire protection systems has been reviewed by American Nuclear Insurers (ANI).

The fire detection and protection systems are designed such that a fire will be detected, contained, and/or extinguished. This is accomplished through the use of noncombustible construction, equipment separation, fire walls, stops and seals, fire detecti on systems, and automatic and manual water suppression sy stems. As a minimum, portable exti nguishers, hose stations, and fire hydrants are available for all areas to control or extinguish a fire.

1.2.10.9 Compressed Air Systems The instrument air system cons ists of one 640 scfm and two 237 scfm (each) instrument air compressors, receivers, dryers, and after-filters to provide a reliab le supply of clean, oil free dry air for the unit pneumatic instrumentation and valves. Station air for normal unit maintenance is provided by a separate 630 scfm station air compressor. Operating pressu res for both systems range between 80 to 120 psig depending on how the compressors are aligne d and how the systems are interconnected.

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

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

MPS2 UFSAR1.2-14Rev. 351.2.11 STEAM AND POWER CONVERSION SYSTEM The turbine generator for Unit 2 is furnished by Gene ral Electric Company. It is an 1800 rpm tandem compound, four flow exhaust, indoor uni t designed for saturated steam conditions.

Under nominal steam conditi ons of 870 psia and 528

°F at the turbine stop valve inlets and with turbines exhausting against a condenser pressure of 2 inches Hg absolute , the gross electrical output is 935 MWe. Turbine output corresponds to a NSSS thermal power le vel of approximately 2715 MWt.

The condensate and feedwater system consists of three condensate pumps, one steam packing exhauster, two steam jet air ejectors, two external drain coolers, tw o trains each ha ving five stages of low pressure feedwater heaters, two turbine-driven steam generator feedwater pumps, two high pressure feedwater heaters as well as the associated pipi ng, valves and instrumentation. Normally, the steam generator feedwater pump turbines are driven by extraction steam. At low loads, main steam is used to drive th e steam generator feedwater pump turbines.

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

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

The radioactive waste processing system consists of the following parts. a.Clean Liquid Waste Processing System The clean liquid waste processing system collects and proces ses reactor coolant wastes from the chemical and volume c ontrol system, primary drain tank and the closed drains system. The system is comprised of pumps, filters, degasifier, demineralizers, receiver ta nks, monitor tanks and the necessary instrumentation, piping, controls and accessories. The processed clean liquid wastes are co llected 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 radi oactive liquid wastes exposed to the atmosphere, are collected in drain tanks and proces sed through filters, and demineralizers. The processed wastes are collected in a monitor tank, sampled, and MPS2 UFSAR1.2-15Rev. 35monitored prior to discharge to the circulating water system after ensuring that the predetermined limits for release are not exceeded.c.Gaseous Waste Processing SystemRadioactive waste gases are collected thro ugh 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 de cay tanks which are provided allow a minimum of 60 days storage capacity prior to release. d.Solid Waste Processing SystemRadioactive solid wastes are collected and pl aced in suitable containers for off site disposal. Spent demineralizer resins are he ld for radioactive decay prior to being dewatered and placed in a shielded cask for removal. C ontaminated 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 F acility prior to being shipped for disposal. Noncompactible soli d wastes may be shipped to an off site processor for volume reduction prior to disposal.

1.2.13 INTERRELATION WITH MI LLSTONE UNITS 1 AND 3 A number of the facilities of the Millstone Nucl ear Power Station are common to Millstone Units 1, 2, and 3. The safe shutdow n of any unit will not be impaired by the failure of the facilities and systems which are shared. A list of these facilities and systems follows:

a.Facilities Radiochemistry laboratoryRadioactive and clean change facilities , including showers, lockers, clothing storage, and toilets Radiation Protection offices

Instrument repair roomWarehouse machine shop Millstone stack (for Unit 2 waste gas), main condenser air ejector and enclosure building filtration system discharge General offices MPS2 UFSAR1.2-16Rev. 35First aid station Lunch room Visitors gallery

345 kV switch yard

Millstone Unit 3 normal stati on service transformer 15G-3SA Millstone Unit 3 reserve stati on service transformer 15G-23SA Millstone Unit 3 SBO di es el generator system Makeup water treatment (Mil lstone Units 2 and 3 only)Bulk storage chemical ton (M illstone 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 s upply and fire detection)

Auxiliary steam

Makeup water treatment

Building heating Sanitary sewers

Plant water

Communications Station Air (A system cross-tie between Un it 3 service air and Unit 2 station air headers is provided)

Operating and maintenance personnel are employed in all three units as described in Section 12.1. Both units have a double containment system with rectangular outer envelopes.The 40 CFR 190 off site radiation dose limits wi ll not be exceeded by s imultaneous operation of Millstone Units 1, 2, and 3. The Millstone Station Physical Security Plan has been im plemented in accordance with 10 CFR 73.55 "Requirements for Physical Protection of Licensed Activities in Nuclear Power Reactors Against Industrial Sabotage" to prohibit unauthorized access to vital areas.

This plan includes measures to de ter or prevent malicious actions th at could result in the release of radioactive materials into the environment through sabotage. Section 12.7 contains a description of the Security Plan.

MPS2 UFSAR1.2-17Rev. 35 1.2.14

SUMMARY

OF CODES AND STANDARDSTo ensure the integrity and operability of pres sure-containing components important to safety , established codes and st andards are used in th e design, fabrication and testing. Table 1.2-1 lists these codes and standards for comp onents relied upon to prevent or mitigate the consequences of incidents and malfunctions origin ating within the reactor coolant pressure boundary, to permit shutdown of the reactor, and to maintain the reactor in a sa fe shutdown condition.

MPS2 UFSAR 1.2-18 Rev. 35TABLE 1.2-1

SUMMARY

OF CODES AND STANDARDS FOR COMPONENTS OF WATER-COOLED NUCLEAR POWER UNITS (1)CODE CLASSIFICATION Component Group AGroup BGroup CGroup DPressure VesselsASME Boiler and Pressure Vessel Code,Section III, Class A, 1968 Edition, Addenda through Summer 1969ASME Boiler and Pressure Vessel Code,Section III, Class C (1968 Edition including Addenda through

Summer 1969)

ASME Boiler and Pressure Vessel Code,Section VIII, Division 1ASME Boiler and Pressure Vessel Code,Section VIII, Division 1 or EquivalentReactor Vessel (2) Safety Injection Tanks (4) Reactor Building Closed Cooling Water Heat Exchangers (3) Service Water Strainers (3) Pressurizer (2) Reactor Building Closed Cooling Water Surge TankVital Chilled Water System Condensers/

EvaporatorsSteam 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 Table NST-1, Class 2API-620 with the NDT Examination Requirements in Table NST-1, Class 3 API-620 or Equivalent Condenser Storage TankAtmospheric Storage Tanks -Applicable Storage Tank Codes such as API-650, AWWAD100 or ANSI B96.1 With the NDT Examination Requirements in Table NST-1, Class 2Applicable Storage Tank Codes Such as API-650 AWWAD100 or ANSI B 96.1 with the NDT Examination Requirements in Table NST-1, Class 3API-650, AWWAD100 or ANSI B 96.1 or Equivalent Diesel Oil Supply Tanks MPS2 UFSAR TABLE 1.2-1 CONTINUED 1.2-19 Rev. 35Pumps and Valves 1.ASME St andard Code for Pumps and Valves for Nuclear Power, Class 1, March 1970 DraftDraft ASME Code for Pumps and Valves, Class II, November 1968. See Footnote (5).Draft ASME Code for Pumps and Valves Class IIIValves - ANSI B 31.1.0 or Equivalent Pumps - Draft ASME Code for Pumps and Valves Class III or Equivalent2.ASME Section III, Paragraph N153 in Summer 1969 Addenda3.ASME Section III, Appendix IX Reactor Coolant Pumps and ValvesHigh Pressure Safety Injection Pumps and ValvesVital Chilled Water PumpVital Chilled Water ValvesLow Pressure Safety Injection Pumps and Valves ASME Section III 1971 Edition, 1971 Winter Addenda Service Water Pumps and Valves Standards of the Hydraulic Institute, ANSI G16.5 Class 1Reactor Coolant System Branch Connection Valves beyond Second Isolation Valves ASME Standard Code for Pumps and Valves, Class 2, March 1970 draftRBCCW Pumps and Valves Standards of

the Hydraulic Institute, ANSI B16.1, ANSI B31.1 All Containment Penetration Isolation Valves ASME Section III, 1971; Draft ASME Pump and Valve Code, 1980, 1983 Auxiliary Feedwater Pumps ASME Code for Pumps and Valves for Nuclear Power, Class II NEMA Standard SM20-1958

Hydraulic InstituteChemical and Volume Control System-Concentrated Boric Acid Service-Pumps Acid Service-Pumps and Valves Draft ASME Code for Pump and Valves, Class II, November 1968Containment Spray Pumps and ValvesPressurizer Safety Valves1.ASME Section III, Class A, 1968 Edition, Addenda through summer of 1970. Code Case 1344-1.Pressurizer Power Operated Relief Valves ASME Section III Class 1, 1977 Edition through winter 1979 AddendaTABLE 1.2-1

SUMMARY

OF CODES AND STANDARDS FOR COMPONENTS OF WATER-COOLED NUCLEAR POWER UNITS (1)CODE CLASSIFICATION Component Group AGroup BGroup CGroup D MPS2 UFSAR TABLE 1.2-1 CONTINUED 1.2-20 Rev. 351This table summarizes the Codes and Standards used for major pressure retaining components.

Not all components are listed. Lat er 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.2The 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.3Including ASME Code Case N-416.41971 ASME Boiler and Pr essure Code,Section III, Class 3.5All pressure-retaining cast part s shall be radiographed (or ultr asonically tested to equivalent standards). Where size or configuration does not permit effective volumetric examination, magnetic particle or liquid penetrant examination may be substituted. Ex amination procedures and acceptance standards shall be at least equivalent to those specified in the applicable class in the code.

Piping1.ANSI B 31.7 Class I, 1969 EditionANSI B 31.7, Class II 1969 EditionANSI B 31.7, Class III 1969 EditionANSI B 31.1.0 or Equivalent2.ASME Section III, Paragraph N153 in Summer 1969 Addenda3.Code Case 70 to B31.7 Primary Coolant Piping and Surge LineHigh Pressure Safety Injection Piping Low Pressure Safety Injection Piping4.Other Reactor Coolant Pressure Pressure Boundary Class I Reactor Coolant System Branch Piping beyond Second Isolation ValvesService Water Piping RBCCW Piping Piping-ASME Section III Code - 1971

Edition, Class I.Chemical and Volume Control System Concentrated

Boric Acid Service Pipi ng ANSI B31.1.0 modified (inside Containment) Containment Spray Piping All Containment Piping Penetrations1.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 EditionTABLE 1.2-1

SUMMARY

OF CODES AND STANDARDS FOR COMPONENTS OF WATER-COOLED NUCLEAR POWER UNITS (1)CODE CLASSIFICATION Component Group AGroup BGroup CGroup D Withheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-1 SITE LAYOUT MPS2 UFSAR1.2-22Rev. 35Withheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-2 PLOT PLAN

MPS2 UFSAR1.2-23Rev. 35Withheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-3 GENERAL ARRANGEMENT, TURBINE BUILDING PLAN AT OPERATING FLOOR ELEVAT ION 54 FEET 6 INCHES

MPS2 UFSAR1.2-24Rev. 35Withheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-4 GENERAL ARRANGEMENT, TURBINE BUILDING PLAN AT MEZZANINE FLOOR ELEVAT ION 31 FEET 6 INCHES

MPS2 UFSAR1.2-25Rev. 35Withheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-5 GENERAL ARRANGEMENT, TURBINE BUILDING PLAN AT GROUND FLOOR ELEVATIO N 14 FEET 6 INCHES

MPS2 UFSAR1.2-26Rev. 35Withheld 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

MPS2 UFSAR1.2-27Rev. 35Withheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-7 GENERAL ARRANGEMEN T AUXILIARY BUILDING PLAN AT ELEVATION 36 FEET 6 INCHES AND ELEVATION 38 FEET 6 INCHES

MPS2 UFSAR1.2-28Rev. 35Withheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-8 GENERAL ARRANGEMENT AUXI LIARY BUILDING SECTIONS "G-G" AND "H-H"

MPS2 UFSAR1.2-29Rev. 35Withheld 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

MPS2 UFSAR1.2-30Rev. 35Withheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-10 GENERAL ARRANGEMEN T CONTAINMENT AND AUXILIARY BUILDING PLAN AT ELEVATION (-)5 FEET 0 INCHES AND ELEVATION (-)3 FEET 6 INCHES

MPS2 UFSAR1.2-31Rev. 35Withheld under 10 CFR 2.390 (d) (1)FIGURE 1.2-11 GENERAL ARRANGEM ENT CONTAINMENT AND AUXILIARY BUILDING PLAN AT ELEVATION (-)25 FE ET 6 INCHES AND ELEVATION (-)22 FEET 6 INCHES

MPS2 UFSAR1.2-32Rev. 35Withheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-12 GENERAL ARRANGEMEN T CONTAINMENT AND AUXILIARY BUILDING PLAN AT ELEVATION (-)45 FEET 6 INCHES

MPS2 UFSAR1.2-33Rev. 35Withheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-13 GENERAL ARRANGEMEN T CONTAINMENT AND AUXILIARY BUILDING SECTION "A-A"

MPS2 UFSAR1.2-34Rev. 35Withheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-14 GENERAL ARRANGEMEN T CONTAINMENT AND AUXILIARY BUILDING SECTION "B-B"

MPS2 UFSAR1.2-35Rev. 35Withheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-15 GENERAL ARRANGEMENT TURBINE BUILDING SECTIONS "C-C" AND "E-E"

MPS2 UFSAR1.2-36Rev. 35Withheld under 10 CFR 2.390 (d) (1)FIGURE 1.2-16 GENERAL ARRANGEMENT TURBINE BUILDING SECTIONS "D-D" AND "F-F"

MPS2 UFSAR1.2-37Rev. 35Withheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-17 GENERAL ARRANGEMENT INTAKE STRUCTURE AUXILIARY STEAM BOILER ROOM PLAN AND SECTION

MPS2 UFSAR1.3-1Rev. 35 1.3 COMPARISON WITH OTHER PLANTSTable 1.3-1 presents a summ ary of the characteristics of the Mi llstone Unit 2 Nuclear Power Plant at the time of applicat ion for operating license. The table incl udes similar data for Calvert Cliffs Units 1 and 2, Maine Yankee Unit Number 1, Turkey Point Units Numbers. 3 and 4 and Palisades Unit Number 1. Bechtel Corporation and Combusti on Engineering (CE), Inc. are identified as contractors in Section 1.6. The Pali sades plant is included in the ta ble because its coolant system is similar to that of Millstone Unit 2, because both Bechtel Corporation and CE, Inc. are Palisades contractors and because it is an example of a CE, Inc. nuclear steam supply system which is operating. Calvert Cliffs and Maine Yankee were select ed because their cores are similar to that of Millstone Unit 2 and the most c ontemporaneous plants for which operating licenses have been issued with which CE is associated. Turkey Poin t is included because it is another comparable plant with which Bechtel Co rporation is associated.

MPS2 UFSAR 1.3-2 Rev. 35TABLE 1.3-1 COMPARISON WITH OTHER PLANTSHYDRAULIC and THERMAL DESIGN PARAMETERS

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)Total Core Heat Output, MWt 3.5 2,560 2,200 2,200 2,560 2,440Total Core Heat Output, Btu/hr 3.5 8,737 x 10 6 7,479 x 10 6 7,509 x 10 6 8,740 x 10 6 8,328 x 10 6 Heat Generated in Fuel, %

3.5 97.5 97.4 97.5 97.5 97.5Maximum Overpower, %

3.5 12 12 12 12 12 System Pressure, Nominal, psia 3.5 2,250 2,250 2,100 2,250 2,250System Pressure, Minimum Steady State, psia3.5 2,200 2,200 2,050 2,200 2,200 Hot Channel Factors, Overall Heat Flux, F q 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 2.45Coolant Flow: Total Flow Rate, lb/hr 3.5 134 x 10 6 101.5 x 10 6 125 x 10 6 122 x 10 6 122 x 10 6Coolant Flow: Effective Flow Rate for Head Transfer, lb/hr 3.5 130 x 10 6 97.0 x 10 6 121.25 x 10 6117.5 x 10 6117.5 x 10 6Coolant Flow: Effective Flow Area for Heat Transfer, ft 2 3.5 53.5 41.8 58.7 53.5 53.5Coolant Flow: Average Velocity along Fuel Rods, ft/sec 3.5 16 14.3 12.7 13.6 13.9Coolant Flow: Average Mass Velocity, lb/hr-ft 2 3.5 2.4 x 10 6 2.32 x 10 6 2.07 x 10 6 2.20 x 10 6 2.29 x 10 6Coolant Temperatures, °F: Nominal Inlet 3.5 542 546.2 545 543.4 538.9Coolant Temperatures, °F: Maximum Inlet due to Instrumentation Error and Deadband, °F 3.5 544 550.2 548 548 546Coolant Temperatures, °F: Average Rise in Vessel, °F 3.5 45 55.9 46 52 51.1Coolant Temperatures, °F: Average Rise in Core, °F 3.5 46 58.3 47 54 53.1Coolant Temperatures, °F: Average in Core, °F3.5 565 575.4 568.5 570.4 565.4Coolant Temperatures, °F: Average in Vessel3.5 564 574.2 568 569.5 564.4 MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONT'D) 1.3-3 Rev. 35Coolant Temperatures, °F: Nominal Outlet of Hot Channel 3.5640642642.8643636Average Film Coefficient, Btu/hr-ft 2-F3.552705400486052405300Average Film Temperature Difference, °F3.534.531.83033.533Heat Transfer at 100% Power: Active Heat Transfer Surface Area, ft 2 3.5 48,400 42,460 51,400 48,416 47,000Heat Transfer at 100% Power: Average Heat Flux, Btu/hr-ft 2 3.5 176,600 171,600 142,400 176,000 170,200Heat Transfer at 100% Power: Maximum Heat Flux, Btu/hr-ft 2 3.5 527,800 554,200 541,200 527,900 502,300Heat Transfer at 100% Power: Average Thermal Output, kw/ft 3.5 5.94 5.5 4.63 5.94 5.74Heat Transfer at 100% Power: Maximum Thermal Output, kw/ft 3.5 16.6 17.6 (2)17.6 (2) 17.816.9Maximum Clad Surface Te mperature at Nominal Pressure, °F3.5657657648657657Fuel Center Temperature, °F: Maximum at 100% Power3.53,7804,0304,0403,7803,640Fuel Center Temperature, °F: Maximum at Over Power3.54,0704,3004,3504,0703,940 Thermal Output, kw/ft at Maximum Over Power3.519.620.0 19.7 (2) 20.019.0CORE MECHANICAL DESIGN PARAMETERS

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)Fuel Assemblies: Design 3.3 CEA RCC CruciformCEA CEA Fuel Assemblies: Rod Pitch, inches 3.3 0.58 0.563 0.550 0.58 0.580 Fuel Assemblies: Cross-Section Dimensions, inches 3.3 7.98 x 7.98 8.426 x 8.4268.1135 x 8.1135 7.98 x 7.98 7.98 x 7.98HYDRAULIC and THERMAL DESIGN PARAMETERS

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONT'D) 1.3-4 Rev. 35Fuel Assemblies: Fuel Weight (as UO 2), pounds3.3 207,035 176,200 210,524 207,269 203,934Fuel Assemblies: Total We ight, pounds 3.3 282,500 226,200 295,800 282,570 279,235 Fuel Assemblies: Number of Grids per Assembly 3.3 8 7 8 8 8 Fuel Rods: Number 3.3 36,896 32,028 43,168 36,896 36,352Fuel 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 Zircal oy Zircaloy Zircaloy Zircaloy Zircaloy Fuel Pellets: Material 3.3 UO 2 Sintered UO 2 Sintered UO 2 Sintered UO 2 Sintered UO 2 SinteredFuel Pellets: Diameter, inches3.30.37950.3670.3590.37950.3795Fuel Pellets: Length, inches3.30.6500.6000.6000.6500.650Control Assemblies: Neutron Absorber3.3B 4C / S.S. Cd-In-AgCd-In-Ag (5-15-80%)Cd-In-Ag (5-15-80%) CruciformB 4C / S.S. / Cd-In-AgB 4C / S.S. / Cd-In-Ag Control Assemblies: Cladding Material3.3NiCrFe Alloy (Inconel 625)304 SS-Cold Worked Welded to 13.5 inch span304 SS Tubes, E.B.

NiCrFe AlloyNiCrFe Alloy Control Assemblies: Clad Thickness 3.3 0.040 0.109 0.016 0.040 0.040Control Assemblies: Number of Assembly, full /

part length 3.3 73 53 41 / 4 Cruciform Rods 77 / 8 77 / 8 Control Assemblies: Number of Rods per Assembly 3.3 5 20117 Tubes per Rod 5 5Core Structure: Core Barrel ID / OD, inches3.3.2.2148 / 151.5133.875 / 137.875149.75 / 152.5 148 / 149.75 148 / 149.75Core Structure: Thermal Shield ID / OD, inches3.3.2.5156.75 / 162.75142.625

/ 148.0 None None 156 / 162CORE MECHANICAL DESIGN PARAMETERS

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONT'D) 1.3-5 Rev. 35NUCLEAR DESIGN DATA

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)Structural Characteristics: Core Diameter, inches (Equivalent)3.3.1136119.5136.71136.0136.0Structural Characteristic s: Core Height, inches (Active Fuel)3.3.1136.7144132136.7136.7 H 2 O/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 UO 2 Rods per Assembly, Unshimmed / Shimmed-204212 / 208--

UO 2 Rods per Assembly, Unshimmed / Shimmed: Batch A3.3176--176176 UO 2 Rods per Assembly, Unshimmed / Shimmed: Batch B3.3164--164160 UO 2 Rods per Assembly, Unshimmed / Shimmed: Batch C3.3(176 / 164 / 164)--(176 / 164 / 164)(176 / 164 / 160)

Performance Characteristics Loading Technique3.4.13 Batch Mixed Central Zone3 Regions Non-uniform3 Batch Mixed Central Zone 3 Batch Mixed Central Zone 3 Batch Mixed Central

Zone MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONT'D) 1.3-6 Rev. 35Fuel Discharge Burnup, Mwd/MTU: Average First Cycle3.4.112,77013,00010,18013,77513,795Fuel Discharge Burnup, Mwd/MTU: First Core Average3.2.122,00014,50017,60022,55030,000 Feed Enrichment (weight percent): Region 13.4.11.931.851.652.052.01 Feed Enrichment (weight percent): Region 23.4.12.332.552.08 / 2.542.452.40 Feed Enrichment (weight percent): Region 33.4.12.823.102.54 / 3.202.992.95Feed Enrichment (weight percent): Equilibrium--2.54 / 3.20--Control Characteristics Effective Multiplication (beginning of life): Cold, No Power, Clean3.4.11.1701.1801.2121.1941.170Control Characteristics Effective Multiplication (beginning of life): Hot, No Power, Clean3.4.11.1291.381.1751.1521.129Control Characteristics Effective Multiplication (beginning of life): Hot, Full Power, Xe

Equilibrium3.4.11.0781.0771.1111.0941.075Control Assemblies: Material3.3B 4C / S.S. Cd-In-AgCd-In-Ag (5-15-80%)Cd-In-Ag (5-15-80%)

B 4C / S.S.-Cd-In-AgB 4C / S.S.-Cd-In-AgControl Assemblies: Number of Control Assemblies3.4.1735345 Cruciform8585 Number of Absorber Rods per RCC (or CEA) Assembly3.3520117 Tubes Welded to Form 13.5 inches span 55Total Rod Worth (Hot), % 3.4.111.078.6 9.69.9Boron Concentrations - To shut reactor down

with no rods inserted, clean, ppm: Cold / Hot, ppm3.4.1945 / 9351,250 / 1,2101,180 / 1,2101,120 / 1,095945 / 935Boron Concentrations - To shut reactor down

with no rods inserted, clean, ppm: To control at power with no rods inserted, clean / equilibrium

xenon, ppm3.4.1820 / 5901,000 / 6701,070 / 830960 / 725820 / 590 Kinetic Characteristics, Range Over Life: Moderator Temperature Coefficient (3) /°F 3.4.1-0.4 x 10-4 to -2.1 x 10

-4+.3 x 10-4 to -1.96 x 10

-4 -3.5-0.08 x 10

-4 to -2.25 x 10 .20 x 10-4 to -1.96 x 10 0.40 x 10

-4 to -2.20 x 10

-4NUCLEAR DESIGN DATA

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONT'D) 1.3-7 Rev. 35 Kinetic Characteristics, Range Over Life: Moderator Pressure Coefficient (3) /psi 3.4.1-0.65 x 10

-6 to +2.39 x 10 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 +2.39 x 10

-6+0.65 x 10

-6 to +2.39 x 10

-6 Kinetic Characteristics, Range Over Life: Moderator Void Coefficient (3) /% Void 3.4.1-0.41 x 10

-3 to -1.43 x 10

-3 +0.5 x 10-3 to -2.5 x 10-3-0.06 x 10

-3 to -1.0 x 10 0.41 x 10

-3 to -1.43 x 10 0.41 x 10

-3 to -1.43 x 10

-3 Kinetic Characteristics, Range Over Life:

Doppler Coefficient (4) /°F 3.4.1-1.45 x 10

-5 to -1.07 x 10 1.0 x 10-5 to -1.6 x 10 1.56 x 10

-5 to -1.08 x 10 1.46 x 10

-5 to -1.06 x 10 1.45 x 10

-5 to -1.07 x 10

-5REACTOR COOLANT SYSTEM - CODE REQUIREMENTS

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)Reactor Vessel 4.2.2ASME III Class AASME III Class AASME III Class AASME III Class AA SME III Class ASteam Generator: Tube Side 4.2.2ASME III Class AASME III Class AASME III Clas s AASME III Class AASME III Class ASteam Generator: Shell Side 4.2.2ASME III Class AASME III Class CASME III Clas s AASME III Class AASME III Class A Pressurizer 4.2.2ASME III Class AASME III Class AASME III Class AASME III Class AA SME III Class APressurizer Relief (or Quench) Tank 4.2.2ASME III Class CASME III Class CASME III Class CASME III Class CASME III Class CPressurizer Safety Valves 4.2.2ASME IIIASME IIIASME IIIASME IIIASME III Reactor Coolant Piping 4.2.2 ANSI B 31.7 US AS B 31.1 USAS B 31.1USAS B 31.7USAS 31.1PRINCIPAL DESIGN PARAMETERS OF THE COOLING SYSTEM

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)Operating Pressure, psig 4.2.1 2235 2235 2085 2235 2235Reactor Inlet Temperature, °F 4.2.1 539.7 546.2 545 544.5 540Reactor 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 2,485NUCLEAR DESIGN DATA

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONT'D) 1.3-8 Rev. 35Design Temperature, °F 4.3.4 650 650 650 650 650Hydrostatic Test Pressure (col d), psig 4.2.13,1103,1103,1103,1103,110Total Coolant Volume, cubi c feet 4.2.111,101 9,088 10,80911,10111,026PRINCIPAL DESIGN PARAMETERS OF THE REACTOR VESSEL

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)Material4.3.1, 4.5.6SA-533, Grade B Class I, low alloy steel plates and SA-508-64, Class 2 forgings, internally clad with Type 304 (5) austenitic SS SA-302, Grade B, low alloy steel, internally clad with Type 304 austenitic SS SA-302, Grade B, low alloy steel, internally clad with Type 304 austenitic SS SA-533, Grade B, Class I, steel, internally clad Type 304 austenitic SS SA-533, Grade B, forgings-A-508-64 Class 2, cladding weld deposited 304 SS

equivalentDesign Pressure, psig4.3.12,4852,4852,4852,4852,485Design 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 Nozzle s, inches 4.3.1 253 236 254 253 266-5/8Overall Height of Vessel and Enclosure Head, feet-inches to top of CRD Nozzle 4.3.141 feet 11.75 inches41 feet 6 inches 40 feet 1-13/16 inches41 feet 11.75 inches42 feet 1-3/8 inches Minimum Clad Thickness, inches 4.3.1 1/8 5/32 3/16 1/8 1/8PRINCIPAL DESIGN PARAMETERS OF THE COOLING SYSTEM

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONT'D) 1.3-9 Rev. 35PRINCIPAL DESIGN PARAMETERS OF THE STEAM GENERATORS

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)Number of Units 4.3.2 2 3 2 2 3Type4.3.2Vertical U-Tube with integral moisture separatorVertical U-Tube with integral moisture separatorVertical U-Tube with integral moisture

separatorVertical U-Tube with integral moisture separatorVertical U-tube with integral moisture separatorTube Material 4.3.2Ni-Cr-Fe Alloy Ni-Cr-Fe-Alloy Ni-Cr-Fe Alloy Ni-Cr-Fe Alloy Ni-Cr-Fe Alloy Shell Material 4.3.2SA-533 Gr. B Class 1 and SA-516 gr 70 SA-302Carbon SteelSA-533 Gr. B Class 1 and SA-516 gr 70SA-533 Gr. B Class 1

and SA-516 gr 70Tube Side Design Pressure, psig 4.3.2 2,485 2,485 2,485 2,485 2,485Tube Side Design Temperature, °F 4.3.2 650 650 650 650 650Tube Side Design Flow, lb/hr 4.3.2 61 x 10 6 33.93 x 10 6 62.5 x 10 6 61 x 10 6 40.67 x 10 6 Shell Side Design Pressure, psig 4.3.2 1,000 1,085 985 985 985Shell Side Design Temperature, °F 4.3.2 550 556 550 550 550Operating Pressure, Tube Side, Nominal, psig4.3.2 2,235 2,235 2,085 2,235 2,235Operating Pressure, Shell Side, Maximum, psig4.3.2 885 1,020 885 885 885Maximum Moisture at Outlet at Full Load, %4.3.2 0.2 0.25 0.2 0.2 0.2Hydrostatic Test Pressure, Tube Side (cold), psig4.3.23,110 3,1073,1103,1103,110Steam Pressure, psig, at full power 4.3.2 800 730 755 835 800Steam Temperature, °F, at full power 4.3.2 520.3 510 513.8 525.2 520.3 MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONT'D) 1.3-10 Rev. 35PRINCIPAL DESIGN PARAMETERS OF THE REACTOR COOLANT PUMP

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)Number of Units 4.3.3 4 3 4 4 3Type4.3.3Vertical, single stage centrifugal with bottom suction and horizontal dischargeVertical, single stage radial flow with bottom suction and horizontal dischargeVertical, single stage radial flow with bottom suction and horizontal dischargeVertical, single stage

centrifugal with bottom suction and horizontal dischargeVertical, single stage centrifugal with bottom suction and horizontal discharge Design Pressure, psig 4.3.3 2,485 2,485 2,485 2,485 2,485Design 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 2,235Suction Temperature, °F 4.3.3 540 546.5 545 543.4 538.9Design 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 290Hydrostatic Test Pressure, (col d), psig 4.3.33,110 3,1073,1103,1103,110Motor Type 4.3.3AC Induction AC Induction AC Induction AC Induction AC Induction 4.3.3Single SpeedSingle SpeedSingl e SpeedSingle Speed Single Speed Motor Rating, hp 4.3.3 6,500 6,000 6,250 7,200 9,000PRINCIPAL DESIGN PARAMETERS OF THE REACTOR COOLANT PIPING

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)Material4.3.4SA516 - GR 70 with minimum 1/8 304L SS clad Austenitic SS SA212B clad with SS SA516 - gr 70 with nominal 7/32 SS clad SA516 - gr 70 with SS

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 33.5Between Pump and Steam Generator - ID, inches4.3.4 30 31 30 30 33.5 MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONT'D) 1.3-11 Rev. 35CONTAINMENT SYSTEM PARAMETERS

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)Type5.2.1Double containment with steel lined prestressed post tensioned concrete cylinder, curved dome roof completely enclosed

by Enclosure BuildingSteel lined prestressed post tensioned concrete cylinder, shallow dome roofSteel lined prestressed post tensioned concrete cylinder, curved dome roofSteel lined prestressed

post tensioned concrete cylinder, curved dome roofSteel lined reinforced concrete flat bottom

and hemispherical dome Containment Parameters: Inside Diameter, feet5.2.1 130116116 130 135 Containment Parameters: Hei ght, feet.

5.2.1 175 169 190.5 181-2/3 169.5Containment Parameters: Free Volume, ft 3 5.2.1 1,920,000 (5)1,550,000 1,640,000 2,000,000 1,855,000 Containment Parameters: Reference Incident Pressure, psig 5.2.15459555055 Containment Parameters: Concrete Thickness, feetContainment 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 Systems6.7.2.1Completely enclosed containment has leaktight penetrations and continuous steel liner.

Enclosure Building Filtration region at small negative pressure during

LCI. Automatic isolation where required. The exhaust from filtration

region passed through charcoal filters to 375 feet Millstone stack following

incident.Leak tight penetration and continuous steel liner, automatic isolation where required Leak tight penetration and continuous steel liner, automatic isolation where required Leak tight penetration and continuous steel liner, automatic isolation where required. The exhaust

from penetration rooms to vent.Leak tight penetration and continuous steel liner, automatic isolation where requiredGaseous Effluent Purge11.1.2.1.3Discharge through Unit 2 stack Through particulate filter &

monitors part of main exhaust systemDischarge through stackDischarge through ventDischarge through stack MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONT'D) 1.3-12 Rev. 35ENGINEERED SAFEGUARDS

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)Safety Injection System: Number of High Head Pumps6.3.2.134 (shared)333 (charging)

Safety Injection System: Number of Low Head Pumps6.3.2.122222Safety Injection System: Safety Injection Tank, number6.3.2.143443Containment Fan Coolers: Number of Units6.5.1.243446Containment Fan Coolers: Air Flow capacity, each at emergency condition, cfm6.5.2.234,80065,00030,00055,000N/A Post-Incident Filters Inside Containment:

Number of UnitsNoneNoneNoneNoneNone Post-Incident Filters Inside Containment: TypeNoneNoneNoneNoneNoneContainment Spray Number of Pumps6.4.2.122-23Emergency Power Diesel Generator Units8.3.1.122 total for both units23 total for both units2 Enclosure Building Filtra tion System Number of Units6.7.2.12---0RADIOACTIVE WASTE PROCESSING SYSTEMS

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)Design Failed Fuel, %11.1.1.11 1 1 1 1Gaseous Waste Processing System11.1.2.1Annual Volume of Gases Discharge, ft 3 11.1.2.114,344 (6)4,539 66,240 (6)Annual Activity Discharge, Curies11.1.2.1556 14,758 (6)6)(6)Decay Storage Time for Gases, Days11.1.2.160 (Minimum) 45 30 (Minimum) 60 (6)Compressors: Number 2 2 (7)2 2 (7)2 MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONT'D) 1.3-13 Rev. 35Compressors: Capacity, each11.1.2.225 SCFM40 CFM2.35 SCFM4 to 7 SCFM (6)Decay Tanks: Number 6 6 (7)3 3 (7)3Decay Tanks: Capacity, (each), ft 3 582 525 100 610 200LIQUID WASTE PROCESSING SYSTEMS

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)Clean Liquid Waste (Reactor Coolant Wastes)11.1.2.1Design Volume Wastes per Year11.1.2.1 14 Reactor Coolant System Volumes (840,000 Gallons)

(6)14 Reactor Coolant System (6)Expected Volume of Waste Discharge Per Year, Gallons11.1.2.1404,234(Design Incorporates Recycle of Waste to R.C. System Clean Liquid Waste

System Not Compared)724,300805,542 (6)Annual Expected Activity Discharged, curies11.1.2.1 286 (includes H 3)(6)(6)(6)Percentage of 10 CFR Part 2011.1.4.10.6%

(6)(6)(6)Degasifier: Number11.1.2.21 1 2 2 Degasifier: Type11.1.2.2Packed Column Utilizing Internal Generated Stripping SteamVacuumPacked TowerFlashing Degasifier: Design Flow Rate, gpm11.1.2.2132 160 120 100 Degasifier: Decontamination Factors11.1.2.21,000 (Kr & Xe)40 10 (6)Storage Tanks: Number11.1.2.24 4 4 2Storage Tanks: Total Capacity11.1.2.23 Reactor Coolant System (180,000 Gallons)200,000 Gallons 6 Reactor Coolant System Volumes (7)250,000 GallonsStorage Tanks: Vent Discharge11.1.2.2To Gaseous Waste System for storage and decayTo Exhaust PlenumPlant VentTo ventilation System and stackDemineralizers: Number11.1.2.23342RADIOACTIVE WASTE PROCESSING SYSTEMS

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)

MPS2 UFSAR TABLE 1.3-1 Comparison with Other Plants (CONT'D) 1.3-14 Rev. 35Demineralizers: Type11.1.2.2Mixed Bed Non RegenerativeMixed bedMixed Bed Non Regenerative Cesium RemovalDemineralizers: Decontamination11.1.2.21,00010100 (6)Demineralizers: Factors11.1.2.2(0 for Y, Mo, H 3)Evaporator (Boron Recovery): Number11.1.2.21N/A21 Evaporator (Boron Recovery): Type11.1.2.2Vacuum, Submerged U-TubeHorizontal Spray FilmForced Calculating, Single EffectEvaporator (Boron Recovery): Capacity, GPM Distillate11.1.2.2252030Evaporator (Boron Recovery): Decontamination11.1.2.2 10 5 (Nonvolatiles)

(6)Evaporator (Boron Recovery): Factors11.1.2.21,000 (Nonvolatiles), 50 (Halogens), 100 (Dissolved Gases) 10 4 (Gases)Aerated Liquid Waste Processing System (Miscellaneous Wastes)

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)Design Volume of Waste per year11.1.2.13,639,400 (Gallons)

(6)(6)(6)(6)Expected Volume of Waste Discharged per year, Gallons11.1.2.1313,000508,620 (6)1,330,320 (6)Annual Expected Activity Discharged, Curies11.1.2.11.11 (includes H 3)0.077 (6)(6)(6)Percentage of 10 CFR Part 2011.1.4.1Less than 0.1%Storage Tanks: Number11.1.2.21 2 1 2 2Storage Tanks: Total Capacity11.1.2.25,000 Ga llons 2,000 Gallons 5,500 Gallons 8,000 Gallons 24,800 Gallons Demineralizers: Number11.1.2.21 (6)N/A 1 N/ADemineralizers: Type11.1.2.2Mixed Bed Non Regenerative (6)Mixed Bed Non Regenerative Demineralizers: Decontamination Factors11.1.2.2500 (6)100LIQUID WASTE PROCESSING SYSTEMS

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)

MPS2 UFSAR 1.3-15 Rev. 351The values listed for these plants were taken from public documentation.2Based on total heat output of the core rather than heat generated in the fuel alone.3Values shown are for beginning of life full power / end of cycle full power

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

.5Measured value from pre-operational volume ve rification test and used for integrated leak rate testing. Includes volume of ven ted pressurizer , safety inj ection tanks, and other tanks.6Not Specifically Available in Public Documents.7Shared by Two (2) Units.Evaporator: 11.1.2.2N/AN/A N/A Evaporator: Number11.1.2.2 1 (7)1Evaporator: Type11.1.2.2 (6)(6)Evaporator: Capacity, Distillate GPM11.1.2.2 (6)(6)Evaporator: Decontamination Factors11.1.2.2 10 6(6)(6)Solid Waste Processing System

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)Evaporator Concentrates11.1.2.1Solidified in Concrete in 55 Gallon drums Solidified in C oncrete in 55 Gallon drums N/ASolidified in Concrete in 55 Gallon drums 55 Gallon drumsSpent Resins Shipping & Volumes11.1.2.1Shipping cask after dewatering, 225 ft 3 Dewatered 55 Gallon Drums (6)Solidified in Concrete in

55 Gallon Drums Shipping caskContaminated Filter Cartridges & Volumes11.1.2.155 Gallon drums55 Gallon drums 55 Gallon drumsSolidified in Concrete in 55 Gallon Drums Cask or 55 Gallon

drums Annual Activity Shipped, curies11.1.2.14,250 (6)(6)(6)(6)Aerated Liquid Waste Processing System (Miscellaneous Wastes)

<Parameter>REFERENCE SECTIONCYCLE 1 MILLSTONE UNIT 2 TURKEY POINT (1) UNITS 3 AND 4PALISADES (1) UNIT 1CALVERT CLIFFS (1) UNITS 1 AND 2MAINE YANKEE (1)

MPS2 UFSAR1.4-1Rev. 35 1.4 PRINCIPAL ARCHITECTURAL AND EN GINEERING CRITERIA FOR DESIGN The principal architectural and engi neering features used in the de sign of Unit 2 of the M illstone Nuclear Power Station are summarized in the following material.

1.4.1 PLANT DESIGN Principal structures and equipmen t which may serve either to prev ent accidents or to mitigate their consequences have been designed, fabricated and erected in accordance with applicable codes so as to withstand the most severe earthquakes, flooding condi tions, windstorms, ice conditions, temperature and other deleterious natural phenomena which could be reasonably assumed to occur at the site during the lifetime of this plant. Systems and components designed for Seismic Category I requirements are listed in Table 1.4-1. It should be noted that the terms

'Category' and 'Class' are us ed interchangeably throughout th e MP2 FSAR in defining seismic design classifications of Struct ures, Systems and Components. Un it 2 was designed so that the safety of one unit will not be impaired in the unlikely event of an accident in the other unit.

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

Redundancy is provided in the reacto r and safety systems so that th e single failure of any active component of either system ca nnot prevent the action necessary to avoid an unsafe condition. The unit is designed to facilitate in spection and testing of systems a nd components whos e reliabilities are important to the protection of the public and plant personnel.

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

Systems and components which are significant from th e standpoint of nuclear safety are designed, fabricated and erected to quality standards commensurate with the safety function to be performed. Appendix 1.A of th is FSAR addresses the implementation of Atomic Energy Commission (AEC) General Design Criteria for Nuclear Power Plants, 10 CFR Part 50, Appendix A. Section 12.8 describes the Quality Assurance Program.

1.4.2 REACTORThe following criteria (see Chap ter 3) apply to the reactor:a.The reactor is of the pres surized 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.

MPS2 UFSAR1.4-2Rev. 35c.Minimum departure from nucleate boi ling ratio during normal operation and anticipated transients will not be below that value which could lead to fuel rod failure or damage. The maximum fuel cen terline temperature evaluated at the design overpower conditi on will be below that value which could lead to fuel rod failure. The melting point of the UO 2 will not be reached during routine operation and anticipated transients.d.Fuel rod clad is designed to maintain cladding integrity th roughout fuel life. Fission gas release within the rods and other factors af fecting design life will be considered for the maximum expected exposures.e.The reactor and control systems are desi gned 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 co nstructed 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 deformati on or rupture, to the pressure vessel or impair operation of the engine ered safety features (ESF).i.Control element assemblies (CEA) are capa ble of holding the co re subcritical at hot zero power conditions following a tri p, and providing a safety mar gin even with the most reactive CEA stuc k in the full, withdrawn position.j.The chemical and volume cont rol 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 i ndependent of the CEA system.k.The combined response of the fuel te mperature coef ficient, 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 re mains bounded and damped in response to any expected changes in any operating variable.

MPS2 UFSAR1.4-3Rev. 35 1.4.3 REACTOR COOLANT AND AUXILIARY SYSTEMS 1.4.3.1 Reactor Coolant System The design bases in this section ar e th ose used for the integrated design of the RCS or those which apply to all of the system com ponents. The design bases unique to each component are discussed in Section 4.3.

The RCS is designed for th e normal operation of tr ansferring 2715 MWt (9.26 x 10 Btu/hr) from the reactor core (2700 MWt) and re actor coolant pumps (15 MWt) to the steam generators. In the steam generator, this heat is transferred to the seconda ry system forming 5.9 x 10 6 lb/hr of 880 psia saturated steam per generator with a 0.2 percent maximum moisture content.

The RCS is designed to acco mmodate the normal design transients listed. These transients include conservative estimates of the operational requirements of the systems and are used to make 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 fo r 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 st eam generators are limited to 200 heatup and cooldown cycles. c.15,000 power change cycles in the range between 15 and 100 per cent of full load with a ramp load change of 5 percent of full load per minute increasing or decreasing. This will occu r 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 decreas ing 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 3,1 10 psig and a temperature at least 60

°F above the nil ductility transition temperatur e (NDTT) of the component having the

highest NDTT

.g.200 cycles of leak test ing 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 st eam generators are limited to 80 cycles of leak testing at 2,485 psig.

MPS2 UFSAR1.4-4Rev. 35i.10 6 cycles of operating pres sure variations of

+/-100 psi from the normal 2,235 psig 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.

In addition to these normal de sign transients, the following abnormal transients are also considered to arrive at a satisfact ory usage factor as defined in Section III, Nuclear Vessels, of the ASME 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.Components of the RCS are desi gned and will be operated so th at no deleterious pressure or thermal stress will be imposed on the structural materials. The necessary consideration has been given to the ductile characteristics of the materials at low temperature.

1.4.3.2 Chemical and Volume Control System The 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 c oncentration in the reactor coolant.d.Provide a controlled path to the waste processing system.The system is designed to accept the dischar ge when the reactor co olant is heated at the design rate of 100

°F/hr and to provide the required makeup when the reactor coolant is cooled at the design rate of 100

°F/hr. Discharge is au tomatically diverted to the waste processing system when the volume control tank is at its highest permissible level. The sy stem will also supply makeup or accept discharge due to power decreases or increases. The design transients are

+/-10 percent of full power step changes and ramp changes of

+/-5 percent of full power per minute between 15 to 100 percent power. On power increases, the letdown flow is automatically diverted to the waste processing system when the volume control tank re aches the highest permi ssible level. On power decreases, sufficient coolant is in the volume control tank to allow a full to zero power decrease without additional makeup, in the event of a makeup system failure or override.

For an assumed 1 percent failed fuel condition, the activity in the reactor coolant does not exceed 411 µCi/cc at 77

°F. The system is also designed to mainta in the reactor coolant chemistry within the limits specified in Section 4.4.3.

MPS2 UFSAR1.4-5Rev. 35The rate of boron addition is sufficient to count eract the maximum reactivity increase due to cooldown and xenon decay. Any one of the three charging pumps is capable of injecting the required boron (as boric acid).

The maximum rate at whic h the reactor coolant boron concentration can be reduced must be substantia lly less than the equivalent maximum rate of reactivity insertion by the CEA.

Prior to refueling, the system is capable of increasing the r eactor coolant boron concentration from zero to 1720 ppm by feed and bleed when the reactor cool ant is at hot standby operating temperature.

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

1.4.3.3 Shutdown Cooling System The shutdown cooling system (see Section 9.3) is designed to c ool the RCS from approximately 300° 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 c ooling water inlet temp erature is at its maximum design value of 95

°F. The design RCS cooldown rate is 100

°F/hr. A temperature of 130°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, assu ming an infinitely exposed core. The maximum allowable pr essure for the RCS during shut down cooling is approximately 285 psig.

1.4.4 CONTAINMENT SYSTEM The containment (see Sections 5.2 and 14.8), including the asso ciated access openings and penetrations, is designed to c ontain pressures and temperatures resulting from a postulated main steamline 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 as sumed since it would tend to reduce the ener gy released into containment.d.The containment air reci rculation cooling system and the containment spray system are credited to mitigate the containment pressure and temperature consequences.Containment response to a loss-of-coolant (LOCA) accident was also analyzed. It was found that the peak containment pressure and temper ature of the MSLB accident bound the LOCA.

The containment is designed to assure integrity against postulated miss iles from equipment failures and against postulated missiles from ex ternal sources.

MPS2 UFSAR1.4-6Rev. 35 Means are provided for pressure and leak rate testing of the c ontainment system. This includes provisions for leak rate testing of individual piping and electri cal penetrations that rely on gestated seals, sealing compounds, expansion bellows, and the interior of the containment.The enclosure building (see Section 5.3) is designe d to withstand a wind loading of 115 mph, with gusts of 140 mph, snow load of 60 psf and seismic loads. The En closure Building is designed so that is structural framing will withstand tornado loads, but the siding will be blown away (see Section 5.3.3).

1.4.5 ENGINEERED SAFETY FEATURES SYSTEMS The design incorporates redundant independent full capacity engin eered safety features systems (ESFS). These, in conjunction with the containment, ensure that the release of fission products, following any postulated occurr ence, at least the minimum ESF required to terminate that occurrence are operable. The following ar e required as minimum safety features:

One high pressure safety injection (HPSI) trainOne 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 contai nment air recircul ation and cooling subsystems, or equivalent (Section 6.4)One hydrogen control subsystem One enclosure building filtration trainOne auxiliary feedwater trains Each of these subsystems is independent of it s redundant counterpart with the exception of the safety injection subsystems. Th e HPSI and LPSI subsystems (S ection 6.3) are independent up to the common pipe connections to the four reactor coolant cold legs. Remote manually operated valves provide appropriate cro ss-connections between redundant subsystems for backup and to allow maintenance. Redundant comp onents are physically separated.

The ESFS are designed to perform their functions for all break sizes in the RCS piping up to and including the double-ended rupture of the largest reactor coolant pipe. The safety injection system limits fuel and cladding damage to an amount which will not interfere with adequate emergency core cooling and holds metal-wa ter reactions to minimal amounts. Two full capacity systems, based on different principles remove heat from the containment to maintain containment integrity, the containment spray system (Section 6.4) and the containment air recirculation and cooling system (Section 6.5). The enclosur e building filtration system (EBF S) (Section 6.7) maintains the enclosure building filtration region (EBFR) at a sli ghtly negative pressure and filters the exhaust from this space. The containm ent postaccident hydrogen control system (Section 6.6) mixes and MPS2 UFSAR1.4-7Rev. 35monitors the accumulation of hydrogen gases within the containment. Purge and recombiners are not credited for any mitigating function.

1.4.6 PROTECTION, CONTROL AND INSTRUMENTATION SYSTEM A reactor protective system (RPS) (see Section 7.2) is provided which initiate s reactor trip if the reactor approaches an unsafe condition.

Interlocks and automatic protective systems ar e provided along with admi nistrative controls to ensure safe operation of the plant.

Sufficient redundancy is installed to permit periodic testing of the RPS so that failure or removal from service of any one protective system component or portion of the system w ill not preclude reactor trip or other safety action when required.

The protective system is isolated from the control instrumentati on systems so that failure or removal from service of any c ontrol instrumentation system component or channel does not inhibit the function of the protective system.

1.4.7 ELECTRICAL SYSTEMSNormal, reserve and emergency sour ces of auxiliary elec trical power are provi ded to assu re safe and orderly shutdown of the plant and to mainta in a safe shutdown condition under all credible circumstances. Onsite electrical power sources and systems are designed to provide dependability, independence, redundancy and testab ility in accordance with the requirements of 10 CFR Part 50, Appendix A. The load-carrying capability and other electrical and mechanical characteristics of emergency power systems are in accordance with the requirement s of Safety Guide Number 9. Two redundant, independent, full capacity emergency power sources and distribution subsystems are provided. Each of these subsystems powers all equipment in the associated safety related subsystems as described in Section 1.4.5.

1.4.8 RADIOACTIVE WASTE PROCESSING SYSTEM The radioactive waste pro cessing system (see Section 11.1) is designed so that discharges of radioactivity to the environment are minimized and are in accord ance with the requirements of Sections 1301 and 1302 and Appendix B of 10 CF R Part 20 and Appendix I of 10 CFR Part 50.

1.4.9 RADIATION PROTECTION Millstone Unit 2 is provi ded with a centra lized control room which has adequate shielding (see Section 11.2.2.3) and ventilation system features (see Section 9.9.10) to permit occupancy during all postulated accidents invol ving radiation releases.

The radiation shielding in Millstone Unit 2 and the radiation control pr ocedures ensure that operating personnel do not receive exposures duri ng normal operation and ma intenance in excess of the applicable limits of 10 CFR Part 20.

MPS2 UFSAR1.4-8Rev. 35 1.4.10 FUEL HANDLING AND STORAGE Fuel handling and storage facili ties (see Section 9.8) are provi ded for the safe handling and storage of fuel. The design precludes accidental criticality.

MPS2 UFSAR1.4-9Rev. 35TABLE 1.4-1 SEISMIC CLASS I SYSTEMS AND COMPONENTS System Components Safety Injection System HPSI pumps and motors LPSI pumps and motorsSafety Injection Tanks Refueling Water Storage Tank Piping and supportsValves and valve operators Containment Spray System Containment spray pumps and motors Shutdown cooling heat exchangers

Refueling water storage tank

Piping and supportsValves and valve operatorsContainment sump screenContainment Air Recirculation and Cooling SystemFans and motors Cooling CoilsHousing Enclosure Building Fi ltration System and Emergency Spent Fuel Pool CleanupFans and motors Filters and housing Electric heaters

Piping, ductwork and supports Dampers and damper operators Hydrogen Control System Hydrogen recombiners PIR fans and motors

Piping and supportsHydrogen purge valves and valve operators

Hydrogen monitoring system MPS2 UFSAR1.4-10Rev. 35 Control Room Air Conditioning System (including the control r oom filtration system)Fans and motors Direct expansion a nd condenser coils Housings Compressor

CRFS Filters Ductwork and supports Dampers and damper operators

Refrigeration piping and supports Refrigerant valves and valve operatorsTemperature control system Control Panels Engineered Safety Feature Room Air Recirculation SystemFans and motorsCooling coils

Ductwork and supports Dampers and damper operatorsDiesel Generator Ventilation SystemFans and motors Ductwork and supports DampersVital Switchgear Ventilation SystemFans and Motors Cooling Coils

Chillers and control panelsPumps and motors

Piping; valves and supports Ductwork and supports Dampers and Damper OperatorsContainment Isolation SystemPiping and sleevesValves and valve operatorsTABLE 1.4-1 SEISMIC CLASS I SYSTEMS AND COMPONENTS System Components MPS2 UFSAR1.4-11Rev. 35Electrical Power Supply SystemStation batteries, racks and chargers 125 VDC Switchgear DC/AC InvertersVital AC and DC distribution panels 4160 Volt Emergency Switchgear480 Volt Emergency Load Centers480 Volt Emergency Motor Control CentersElectrical Distribution SystemVi tal tray system and supportsVital underground duct banks Penetration assembliesReactor Coolant SystemReactor vessel and internalsControl element assemblies and drives Pressurizer

Reactor coolant pumps and motors Reactor coolant pipingPressurizer surge line and supports

Pressurizer safety and relief valvesSteam generatorsVent, 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 supportsTABLE 1.4-1 SEISMIC CLASS I SYSTEMS AND COMPONENTS System Components MPS2 UFSAR1.4-12Rev. 35Chemical and Volume Control SystemBoric acid storage tanks Boric acid pumps and drivers Boric acid piping supports and valvesCharging pumps and drivers Charging line piping, supports, valves and pulsation dampeners

Letdown line piping, su pports 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 *Spent Fuel Pool Cooling SystemPipi ng, 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) *Gaseous Waste Processing Syst emWaste gas decay tanks *Waste gas compressors *Waste gas filter

• High pressure (150 psig) service piping, supports, and valves *TABLE 1.4-1 SEISMIC CLASS I SYSTEMS AND COMPONENTS System Components MPS2 UFSAR1.4-13Rev. 35 Fuel and Reactor Component Handling EquipmentContainment polar crane 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 racksNew fuel elevator *Spent fuel inspection machine *RBCCW SystemRBCCW Pumps and MotorsRBCCW Heat Exchangers RBCCW Surge Tank

Piping and Supports Expansion JointsValves and Valve OperatorsService Water SystemPumps and Drivers Piping and Supports Valves and Valve Operators Service Water StrainersEmergency Diesel Generators Diesel Oil SystemAir Intake and Exhaust PipingControl Panels Diesel Oil Supply TanksPiping, Valves and SupportsLube Oil SystemPumps and motorsCoolers Piping and supportsHeaters Piping and supportsTABLE 1.4-1 SEISMIC CLASS I SYSTEMS AND COMPONENTS System Components MPS2 UFSAR1.4-14Rev. 35Jacket Water Cooling SystemPumps and motorsCoolers Piping and supportsHeaters Jacket water expansion tankValves and valve operators

• Designated seismic Class II components but designed for Class I earthquake basis.Air Cooling SystemPumpsCoolers Piping and supports Valve and valve operatorsStarting Air SystemAC and DC Motor Driven CompressorsStarting Air tanks

Piping and supports upstream of check valvesAuxiliary Feedwater SystemAuxiliary. feedwater pumps and drivers Condensate storage tank Piping and supportsValves and valve operatorsMain Steam System (Upstream of isolation valvesMain steam safety relief valves Atmospheric dump valvesMain Steam isolation valves

Piping and supportsValves and valve operatorsEngineered Safety Actuation System, Status Panel Reactor Protection SystemSeismic Measurement Instrumentation Main Control BoardsMain Steam Isolation PanelTABLE 1.4-1 SEISMIC CLASS I SYSTEMS AND COMPONENTS System Components MPS2 UFSAR1.4-15Rev. 35

• Designated seismic Class II components but designed for Class I earthquake basis.

Hot Shutdown Control BoardsBoric Acid Heat Tracing Panels Radiation Monitoring SystemTABLE 1.4-1 SEISMIC CLASS I SYSTEMS AND COMPONENTS System Components MPS2 UFSAR1.5-1Rev. 35 1.5 RESEARCH AND DEVELO PMENT REQUIREMENTS 1.5.1 GENERAL The design of Millstone Unit 2 is based upon concepts which have been successfully applied in the design of other pressurized water reactor power plants. However, certain programs of theoretical analysis or experimentation (constituting "research and development" as defined in the Atomic Energy Act, as amended, and in Nuclear Regulatory Commissi on (NRC) Regulations) have been undertaken to aid in plant design and to verify the pe rformance characteristics of the plant components and systems. This section describes the results and status of these analytical and test programs, including experime ntal production and testing of models, devices, equipment and materials at time of applic ation for an operating license.

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

All CE research and development programs required to ju stify the design to Mi llstone Unit 2 were completed and all test results were factored into design of the plant.

1.5.2 FUEL ASSEMBLY FLOW MIXING TESTS In 1966, a series of single-phase tests on coolant turbulent mixi ng was run on a prototype fuel assembly which was geometrically similar to the Palisades assembly. The model enabled determination of flow resist ance and vertical subchannel flow rates using pressure instrumentation and the average level of eddy flow using dye-i njection and sampling equipment.

The tests yielded the value of the inverse Peclet number charac teristic of eddy flow (0.00366).

During the course of the tests the value was shown to be insensitive to coolant temperature and to vertical coolant mass velocity. The design value of the inverse Peclet Number was established as 0.0035 on the basis of the experimental results.

As part of a CE sponsored res earch and development program, a new series of single-phase dye injection mixing tests were conduc ted in 1968. The tests were pe rformed on a model of a portion of control element assembly (CEA) type fuel assembly which was sufficiently instrumented to enable measurement, via a data reduction computer program, of the individual lateral flows across the boundaries of 12 subchannels of the model. Al though these tests were not intended for that purpose, some of the test results could be used to determine the average level of turbulent mixing in the reference design assembly. The inverse Peclet Number calculated from the average of 56 individual turbulent missing flows (two for each subchanne l boundary) obtained from the applicable data was 0.0034. With re spect to general turbulent mixing, therefore, the more recent study on the CEA verifies the cons tancy of the inverse Peclet number for moderately different fuel assembly geometries and confirms the design value of that characteristic.

MPS2 UFSAR1.5-2Rev. 35 1.5.3 CONTROL ELEMENT ASSEMBLY DROP TESTSA series of tests was complete d on both single and dual CEAs in a cold water, low pressure facility to satisfy th e following objectives:a.Determine the mechanical and functional feasibility of the CEA type control rod concept.b.Experimentally determine the relations hip between CEA drop time and CEA drop weight, annular clearance be tween CEA fingers and guide tubes, and coolant flow rate within the guide tube.c.Experimentally determine the relationshi p 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 dr op time of adding a flow re striction or of plugging the flow holes in the lower portion of a gui de tube (as might occur under accident conditions).e.Determine the effects of misalignment wi thin the CEA guide tube system on drop time.The results of these tests were used as the basis for selecti ng the final CEA and guide tube geometrics. The tests have demons trated that the five-finger CEA concept is mechanic ally and functionally feasible and that the CEA design has met the criteria establ ished for drop time under the most adverse conditions. The te sting has also verified that the analytical model used for predicting the drop time s gives uniformly c onservative results.The effects on drop time of all possible combinations of frictional restraining forces in the control element drive mechanism (CEDM), angular and radial misalignment of the CEDM, bowing of the guide tubes, and misalignments of the CEA should have been e xperimentally investigated and defined. The conditions tested simulated all the effects of to lerance buildup, dynamic loadings, and thermal effects. The tests demonstrated that misalignments a nd distortions in excess of those expected from tolerance buildup or any other anti cipated cause would still result in acceptable drop times.

1.5.4 CONTROL ELEMENT DRIVE ASSEMBLY PERFORMANCE TESTS An accelerated life test of a ma gnetic jack coupled to a CEA was co mpleted. This test consisted of continuous operation of the mech anism for a total accumulated tr avel of 32,500 feet at conditions similar to those it will encounter when instal led on the operating reactor. The mechanism was operated at a speed of 40 inches per minute. without malfunction or adjustments. In addition, 200 full height drops were comple ted with all drop times less than 2.5 seconds for 90 percent insertion. Subsequent testing at various conditions was conduc ted to determine maintenance cycles.

MPS2 UFSAR1.5-3Rev. 35Tests have shown that the magne tic jack type mechanism will operate in the anticipated containment environment after a Design Basis Accident. Among va rious other tests documented in Reference 1.5-2, a magnetic jack type CEDM, similar to that insta lled at Unit 2 was verified to be capable of withstanding a comp lete 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 normal operating temperat ure and pressure (600

°F and 2250 psi) without damage to the CEDM and holding the CEA. In addition, th e coils stacks were later subjec ted to a steam environment for 15 minutes without affecting their electrical capabilities.The design of the CEDM is such that loss of CEDM cooling will not prevent the CEDM from releasing the CEA. The ability of the CEDM to release the rods is not dependent on the cooling flow provided by the CEDM cooling system. Cooling function is only to ensure reliability of the CEDM coil stack.

1.5.5 FUEL ASSEMBLY FLOW TESTSVelocity and static pressure meas urements were made in an oversiz ed model of a fuel assembly to determine the flow distributions present. Effects of the distribut ions on thermal behavior and margin are to be evaluated, where necessary, with the use of a CE version of the COBRA thermal and hydraulic code (Reference 1.5-1). Subj ects investigated in clude 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 ach ieved within 10 percent of core height.

The normal inlet flow distri bution arising from the geom etric 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 affe cted 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 hol e 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 subcha nnel during 120 percen t 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 indi cated that the upstream influence of a subchannel blockage diminished ve ry rapidly in that direction.c.Flow distribution within the assembly as af fected by complete blockage of one to nine subchannels: The flow distributions we re 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.

MPS2 UFSAR1.5-4Rev. 35d.Flow distribution below the top header pl ate, as af fected by the header plate and alignment plate flow hole geometry a nd by the presence of the CEA shroud:

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

1.5.6 REACTOR VESSEL FLOW TESTSTests were conducted with one-fifth scale models of CE reactors to determine hydraulic performance. The first tests were performed for the Palisades plant which has a reactor coolant system (RCS) similar to that of Millstone Unit

2. The tests investigated flow distribution, pressure drop and the tracing of flow paths within the vessel for all four pum ps operating and various part-loop configurations. Air was used as the test medium. CE has al so conducted tests on a one-fourth scale model of the Fort Calhoun reactor using air as the test medium.

Similar one-fifth scale model tests have been performed fo r Maine Yankee, which has a core similar to that of Millstone Unit 2. These tests were conduc ted in a cold water loop. All components for the model were geometrically simila r to those in the reactor except for the core where 217 cylindrical core tubes were substituted for the fuel bundles. The core tubes contained orifices to provide the proper axial flow resistance.Flow characteristics for Millstone Unit 2 we re determined by taki ng into consideration similarities between Millstone Un it 2 and other CE reactors in conjunction with the experimental data from the flow model programs.

1.5.7 IN-CORE INSTRUMENTATION TESTSTests on in-core thermocouples and flux detectors were perf ormed to ensure that the instrumentation will perform as ex pected at the temperatures to be encountered and that it does not vibrate excessively and caus e excessive wear or fretting.

Cold flow testing has been completed on a similar detector ca ble; no adverse vibrations or wear effects were encountered.

Hot 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 breach of mechanical integrity was observed.

Mechanical tests of the insertion and removal equipment and instrumentation were performed on thimbles of the same approximate configuration as those used on Millstone Unit 2. The top entry in-core instrumentation design prov ides a means of eliminating th e need of handling instrument assemblies separately, thus, minimizing downt ime and personnel exposur

e. A full-scale mockup was built to accommodate three in-core instrumentation thimbl e assemblies. Major components and subassemblies of the mockup included:a.An in-core instrumentation test asse mbly , including the upper guide structure support plate, three thimble guide sleeves, fuel alignment plate, three fuel bundle guide tubes, and the core support plate.

MPS2 UFSAR1.5-5Rev. 35b.A thimble assembly consisting of th e instrument plate, three in-core instrumentation thimbles and the lifting sling.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.

Insertion and withdrawal tests we re performed to determine the fri ctional forces of a multi-tube instrument thimble assembly during insertion and wi thdrawal from a set of fuel bundles. This test simulated the operation that will be performed during the refueling of the reactor. To determine whether jamming of the thimbles would occur during this operation, bendi ng loads were applied to the thimble assembly by tilting the instrument plate in 0.5 de gree increments up to a total of five degrees from horizontal. Guide tubes were filled with water. The assembly was raised and lowered approximately five times for each tilt setting. Re sults showed no discernible difference in the friction forces for the various tilt settings. The tests demonstrat ed that the repeated insertion and withdrawal of in-core instru mentation thimble assemblies in to the fuel bundle guides can be accomplished with reasonable insertion forces.Life cycle tests were performed to determine if the frictional forces increase as a result of 40 insertions and withdrawals. An automatic timer was installed in the crane electrical circuitry to automatically cycle the thimble assembly between the fully inserted and withdrawn position. The instrument plate was set for five degrees tilt and the a ssembly was cycled 60 times. The insertion and withdrawal forces were measured during the first and last five cycles. No discernible difference was noticed.An off-center lift test was performed to determine if the thimble assemb ly could be withdrawn from the core region while lifting the assembly from an extreme off center position. For a lifting point 11 inches off center, insertion was accomp lished without incident. The flexibility of the thimble is such that jamming of the assembly due to off-center lifting does not occur.

Cable insertion tests were performed to determine the forces required to completely insert and withdraw a detector cable from the in-core instrumentation thimble assembly. The guide tube routing included typical bends e qual to, or worse than, those found in the reactor. The detector cable was passed through the guide tubing and into a thimble. In all cases, the insertion and withdrawal forces were r easonable for hand insertion.

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

1.

5.9 REFERENCES

1.5-1Rowe, D. S., "Cross-Flow Mixing Betwee n Parallel Flow Channels During Boiling." COBRA Computer Program for Coolant Boiling in Rod Arrays, Part 1, BNWL-371, March 1967.

MPS2 UFSAR1.5-6Rev. 351.5-2Combustion Engineering Inc., Test Report Number TR-DT-78, dated 8/21/72, "Magnetic Jack Type Control Element Drive Mechanism Design and Test Report."

MPS2 UFSAR1.6-1Rev. 35 1.6 IDENTIFICATION OF CONTRACTORSOriginally, The Connecticut Li ght and Power Company (CL&P), the Hartford Electric Light Company (HELCO), and Western Massachusetts Electric Company (WMECO) (the Owners), and Northeast Nuclear Energy Company (NNECO) were the applicants for the operating license for Millstone Unit 2. At that time NNECO acted as the agent for the owners and was responsible for the design, construction and operation of the plant. However, in 2001, the operating license was transferred to Dominion Nuclear Connecticut, Inc., at which time they became the sole owner and operator of Millstone Unit Number 2.

Combustion Engineering (CE), Inc. was engaged to design, manufacture a nd deliver the Nuclear Steam Supply System (NSSS) and nuc lear fuel for the first core and the first two core reload batches to the site. The NSSS includes the reactor coolant syst em, reactor auxiliary system components, nuclear and certain process instrume ntation, and the reactor control and protective system. In addition, CE furnished technical assist ance for erection, initial fu el loading, testing and initial startup of the NSSS.

Bechtel Corporation was engaged as the Engineer-Constructor for this project and as such performed engineering and design work for th e balance-of-plant equipment, systems and structures not included under the CE scope of supply. Bechtel wa s engaged to perform onsite construction of the entire plant with technical advice for installation of the reactor components provided by CE.

The reactor vessel closure head was replaced during refueling outage 16 with a new head assembly fabricated from materials that are less susceptible to Primary Water Stress Corrosion Cracking (PWSCC). The new head assembly wa s manufactured by Mitsubishi Heavy Industries. Westinghouse/CE was engaged in the design, installation and testing of the head.

The pressurizer assembly was re placed in 2006 with a new assembly fabricated from materials that are less susceptible to PWSCC. AREVA was engaged in the design, fa brication, installation and testing of the replacement pressurizer.

1.

6.1 REFERENCES

1.6-1Millstone Unit 3, Final Safe ty Analysis Report, Section 13.1 - Organizational Structure.

MPS2 UFSAR1.7-1Rev. 35 1.7 GENERAL DESIGN CHANGES SINCE ISS UANCE OF PRELIMINARY SAFETY ANALYSIS REPORT 1.7.1 GENERALSince the issuing of the Preliminary Safety An alysis Report (PSAR), a number of changes were made in the design of Millstone Unit 2. These changes improved the operating characteristics and enhance plant safety and reliability. The following reflects ch anges made up to the time of operating license application.

1.7.2 CONTROL ELEMENT DRIVE MECHANISMS Magnetic jack drive mechanisms are provided for positioning the cont rol element assemblies (CEA) instead of rack and pinion drive mechanisms. The magnetic jack control element drive mechanism (CEDM) is completely sealed by a pressure boundary, el iminating the need for seals. Motion of the control element drive shaft is accomplished by sequencing five solenoid coils located around the pressure boundary.

Combustion Engineering (CE), Inc., supplied id entical CEDMs on previous plants, including Maine Yankee (Atomic Energy Commission (AEC)

Docket Number 50-309) and Calvert Cliffs Units 1 and 2 (AEC Docket Number 50-317 and 50-318).

1.7.3 RADIOACTIVE WASTE PROCESSING SYSTEM 1.7.3.1 Clean Liquid Waste Processing System A closed drains system and a 700 gallon equipment drain sump ta nk were included in the system to collect liquids containing dissolved hydrogen and fission gases from e quipment drains, valve stem leakof fs, and relief valve discharges. The liquid wastes are collected in this tank via the closed drains system. This tank was provided to minimize the releas e of radioactive gases to the atmosphere without prior proces sing by the gaseous waste system.The flash tank was replaced by a packed column-t ype degasifier utilizi ng internally generated stripping steam. The degasi fier has a better decontaminatio n factor for xenon and krypton than would have been possible with the proposed flash tank.Plant space and the necessary piping and valves were provided for incorporating two additional demineralizers into the system, if required, based on operating experience.

1.7.3.2 Gaseous Waste Processing System Four additional waste gas decay ta nks were added to the system to allow for a minimum of 60 day decay of all hydrogenated waste ga ses, including cover gases, collected by the system prior to release to the atmosphere through the Millstone stack.

MPS2 UFSAR1.7-2Rev. 35 1.7.4 VITAL COMPONENT CLOSED COOLING WATER SYSTEM The vital components closed cool ing water system was deleted and the components cooled as follows: 1.7.5 ELECTRICAL 1.7.5.1 AC Power The station service transformers supply power at 6900V and 4160V via their respective station service busses for lar ge motor loads. Further, the 4160V suppl ies power to the 480V unit substation transformers for smaller loads.To preserve redundancy and sepa ration, each motor control cente r is fed from only one 480 volt load center rather than from two.

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

Additional conditions under which th e diesel generators will star t automatically are noted in Section 8.3.3.1.

1.7.5.3 DC Supply A third station battery was added to care for the non safety-relat ed 125 volt DC lo ads associated with the turbine generator.

Each 125 volt DC distributi on panel formerly had a feeder from each of the two st ation batteries, with diodes to prevent tying the battery buses together. To maintain the independence of redundant sources, the diodes we re removed and the DC distribut ion panels fed from redundant battery buses.

ComponentCooling System Service air compressors and instrument air compressorsTurbine building clos ed cooling water (interconnecting piping provided to reactor building closed cooling water)

Auxiliary feedwater pump turbine oil coolerWater being pumped Diesel generatorService water Control room air conditioners Air MPS2 UFSAR1.7-3Rev. 35 1.7.5.4 Instrument PowerTwo 120 volt regulated AC instrument buses were provided (instead of one) to assure redundant power sources for vital instrumentation.

1.7.6 AXIAL XENON OSCILLATION PROTECTION Automatic initiation of an a ppropriate protection system fo r axial xenon os cillation was incorporated into the reactor protective syst em. This addition provided compliance with the AEC's General Design Criterion 20 as published February 20, 1971, in the Federal Register and as interpreted for preceeding reactors of similar design (see Calvert Cliffs Units 1 & 2 Amendment 15, Question 3.14). The basis for this addi tion was to provide an automatic protective backup to the operator in the unlikely event he should fail to adjust the full length CEA as required late in core life when axial xenon oscillations may become divergent.

1.7.7 NUMBER OF CONTROL ELEMENT ASSEMBLIES AND DRIVE MECHANISMS The number of CEAs in the Millstone Unit 2 re actor is 73, compared to 85 CEAs shown in the PSAR design. The number of drive mechanisms wa s changed from 65 in the PSAR to 69 for Cycle 1. Then, removal of 8 part-length CEAs in 1978 reduced the number of drive mechanisms to 61. This resulted in a net in crease in the number of single CE As (37 to 49) a nd a net reduction in the number of dual CEAs (40 to 24), thereby providing greater flexibility for optimization of CEA programming and fuel management.

1.7.8 BURNABLE POISON SHIMS Burnable poison shims were added to the fuel as semblies in Cy cle 1, replacing some fuel. These shims permitted lowering of the initial boric aci d concentration in the coolant. This provided additional assurance that the mode rator temperature coef ficient, at power at beginning of life, would not be positive.

1.7.9 STRUCTURES The 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 insp ection were not used for liner plate weld quality control.

MPS2 UFSAR1.7-4Rev. 35 1.7.10 HIGH PRESSURE SAFETY INJECTION PUMPS High Pressure Safety Injecti on (HPSI) pump P-41B (Figure 6.1-1) (Sheet 2) was connected to each of the two suction headers but is normally isolated by valvi ng. This HPSI pump served as a spare and was aligned, process wise and electrically, for opera tion only when eith er of the other two HPSI pumps is taken out of service. Two operable HPSI pumps satisfy redundancy requirements for core cooling.1.7.11 CONTAINMENT PURGE VALVE ISOLATION ACTUATION SYSTEMContainment Purge Valve Actuati on System was changed from two-out-of-four to one-out-of-four logic. See Sections 7.3.2.3 and 7.5.6.3 for details.

1.7.12 CONTROL ELEMENT DRIVE SYSTEM The Control Element Drive Syst em (CEDS) was modified to include a CEA Motion Inhibit feature which acts to help the operator assure that limits on CEA posit ion are not exceeded. The CEDS is described in Section 7.4.2.

MPS2 UFSAR1.8-1Rev. 35 1.8 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 GENERAL This section describes the status of programs initiated to investigate the items which were identified by the Advisory Committee on Reactor Safeguards (ACRS) as being of special interest and pertaining to all large water-cooled power reactors up to th e time of application for an operating license.

In carrying out these programs, in formation derived from research and development activities of the Atomic Energy Commission (AEC) and other organizations in the nuclear power industry were considered.

1.8.1.1 Ability of Fuel to Withstand Transients at End of Life and Experimental Verification of Maximum Linear Heat Generation Rate The fuel cladding was designed to limit the transi ent stresses to two-thir ds of the unirradiated value of the yield stress even during a depressurization transient near the end of life, when the internal gas pressure is highest.

Experimental verification of the maximum linear heat generation rate employed in the Millstone Unit 2 design was discussed in the original FSAR submitted at the time of application for an Operating License. Numerous irra diation tests, which bracket th e design of these units, were performed, including those in the Westinghouse test reactor, the Sh ippingport blanket irradiations, the mixed oxide irradiations in the Saxton reactor, the zirconium clad UO 2 fuel rod evaluations in the Vallecitos boiling water reactor, the large spee d blanket reactor rod ir radiations, the center melting irradiations in Big Rock, Peach Bottom 2 irradiat ions, and NRX irradiations (AECL-Canada). In these tests, fu el rods similar to those employe d in the design of the Millstone Unit 2 core were successfully ir radiated to fuel burnups varying from very short term tests up to 60,000 MWD/MTU and at linear heat ra tes ranging from 5.6 up to 27.0 kW/ft.

1.8.1.2 Fuel Integrity Following a Loss-of-Coolant Accident The ACRS had asked that informat ion be developed to show that the "...melting and subsequent disintegration of a portion of fu el assembly

...will not lead to unacceptable conditions." They referred specifically to the "...effects in terms of fission product release, local high pressure production, and the possible initiation of failure in adjacent fuel elements...".

Inquiry was made as to whet her accident conditions that might occur which cause clad temperatures to reach such high temperatures that embrittlement occurs, and whether subsequent quenching operations will cause th e embrittled portions to disint egrate and thereby prevent a sufficient flow of emer gency core coolant to the remainder of the core.

MPS2 UFSAR1.8-2Rev. 35 Fuel damage of the magnitude indicated is prevented by the inherent nuclear and thermal characteristics of the UO 2 core and by the provision of engi neered safety features (ESF).With regard to the nonexcursion mechanisms le ading to the conditions described by ACRS, the following two conditions might be conjectured:A.Fuel bundle inlet flow blockage during full power operation and s ubsequent overheating of the coolant-starved fuel, orB.loss of reactor coolant.

Condition A, inlet flow blocka ge during full-power operation and subsequent overheating and melting of the fuel, is not c onsidered possible because open (nonshrouded) fuel bundles are used, thereby providing cross-flow to th e flow-starved channel even if some of the inlet holes were blocked. Details and conclusions of the tests pe rformed at Combustion Engineering (CE), Inc. on the influence of inlet geometry on flow in the entrance region are presented in ASME paper 68-WA/HT-34 delivered at the December 1968 Winter Annual Meeting. Further analysis of these tests showed that if a gr oup of four flow holes in the core support plate at the base of the fuel bundle were blocked, the subchannels above the blocked region w ould have an inlet velocity about 21 percent of the core average bulk inlet velocity. Be cause of crossflow from the surrounding nonblocked regions, the net effect of this flow shortage, using conservative calculations, is to increase the enthalpy rise of the blocked regi on by a maximum of 35 percent. At nominal conditions, the hot channel departure from nucleate boiling ra tio (DNBR) would drop from 2.0 to 1.4, assuming that the blockage occu rred directly below the design hot channel.

Condition B was covered comprehensively in the Statement of Affirmative Testimony and Evidence of Combustion Engineering in the Matt er of Rulemaking Heari ng for the Acceptance Criteria for Emergency Core Cooling System for Light-Water-C ooled Nuclear Power Reactors, Docket Number RM-50-1. The emer gency core cooling system (ECC S) is designed to remove the decay heat from the core for the necessary peri od of time following a lo ss-of-coolant accident (LOCA). Core power distributions and LOCA temperature-time histories indicate that for peak clad temperatures below 2300

°F, the total clad oxidation will be significantly less than 1 percent.

1.8.1.3 Primary System Quality Assuran ce and In-Service Inspectability A comprehensive quality assurance program has been established to assure that Millstone Unit 2 is designed, fabricated, and cons tructed in accordance with the requirements of applicable specifications and codes. The program started with the initia l plant design and has continued through all phases of equipment procurement, fabrication, er ection, construction, and plant operation. The program provides for review of specifications to assure that quality control requirements are included and for surveillance and audits of th e manufacturing and construction efforts to assure that the specified requirements are met.

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

This program fully meets the gui delines established in the former AEC Regulation 10 CFR Part 50, Appendix B entitled "Quality Assurance Crit eria for Nuclear Power Plants." The quality MPS2 UFSAR1.8-3Rev. 35assurance organization is described in the Quality Assurance Program Description Topical Report. That information is in corporated herein by reference.

Baseline inspection and subsequently in-service inspections are performed and are further discussed in Section 4.6.6.

1.8.1.4 Separation of Control and Protective Instrumentation In addition to any redundancy and separation provided for control or for protective instrumentation, the control and protective instrume ntation are independent of each other. Control action and protective actio n derived from the same process va riable are generated by separate instrumentation loops. Malfuncti on of a single control instrume ntation loop cannot impair the operation of the protective instrumentation loop and conversely malfuncti on of the protective instrumentation loop does not af fect operation of the control loop. The instrumentation for a single protective and a single control channel may be located adjacent to one another, and their circuits may be routed in the same cable tray , but each is capable of performing its function independently of the other. Further disc ussion is provided in Chapters 7 and 8.

1.8.1.5 Instrumentation for De tection of Failed Fuel Early detection of the gro ss failure of fuel elemen ts permits early applica tions of action necessary to limit the consequences.

Based on a study of the exp ected fission and corrosion product acti vities in the reactor coolant, it was concluded that the gross gamm a plus specific isotope monitor provides a simple and reliable means for early detection fuel failures.

The design bases of the detecti on system include the following:a.Trends in fission product activity in the r eactor 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 acti vity , emitted from a fuel element cladding failure, is indicate d 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 functi ons or detection of the sp ecific 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.The location and operation of the detector , designated as a process radiation monitor, is described in Sections 7.5.6.3 and 9.2.2.

MPS2 UFSAR1.8-4Rev. 35 Note: This section provides hist orical information provided to ACRs at the time of initial Licensing and was not intended to be updated.

1.8.1.6 Effects of Blowdown Forces on Core and Primary System ComponentsThe dynamic response of reactor internals resulting from hydrodyn amic blowdo wn forces under a postulated LOCA condition was the subject of a CE topical report which contained a complete description of the theoretical basi s for methods of analysis for th e various reactor components, as well as documentation of comput er programs and the respective an alytical structural models.Reactor vessel internal structures were analyzed to ensure the re quired structural integrity during abnormal operating conditions, including the effects of blowdow n, pressure drop and buckling forces. For the LOCA, the CEFLASH-4 computer pr ogram was used to define the flow transient and the WATERHAMMER program determines the corresponding dynamic pressure load distribution. The dynamic response of the reactor vessel internals to the space and time-dependent pressure loads were obtained th rough the use of a number of stru ctural dynamic analysis codes.

Lateral and vertical dynamic response of the intern als were considered, as well as the transient response and dynamic buckling of a core support barrel in shell modes. Both the CEFLASH-4 and WATERHAMMER models were evaluated against the LOFT program results.

The loads resulting from the LOCA condition we re added to the loads resulting from normal operation and the design basis earthquake (DBE) fo r each critical component and the component deflections and stresses analyzed to ensure compli ance with the criteria specified in Section 4.2.

1.8.1.7 Reactor Vessel Thermal ShockSufficient emergency core coolin g water is available to flood the core region in the event of a major LOCA. The Millstone Unit 2 design uses a sect ion of each of the RC S cold legs to conduct the water from the safety injecti on nozzles to the reactor vessel. This water then flow s into the downcomer annulus and into the lower plenum of the reactor vessel befo re flooding the core itself. Analytical investig ations were performed to provide assurance that the resultant cooling of the irradiated inner surface of the thick-walled reactor vessel will not induce or propagate cracks sufficient to cause the reactor vessel to fail.

An analytical evaluation of pressurized thermal shock effects in CE's NSSS was issued by CE in December 1981 (CEN-189). The limi ting case is a small break LO CA with the assumption of concurrent loss of all feedwater. For Millstone Unit 2, it was f ound that crack initiation would not occur during this limiting transient th roughout the unit's desi gn life (32 EFPY).Subsequently, the Pressurized Thermal Shock Rule (10 CFR 50.61, 1986) was used for embrittlement shift prediction. Th e results confirmed that the reactor vessel was fully able to withstand a postulated pressuri zed thermal shock imposed by the ECCS through the unit's design life.

MPS2 UFSAR1.8-5Rev. 35 1.8.1.8 Effect of Fuel Rod Failure on the Capa bility of the Safety Injection System CE conducted experimental and an alytical investigations of fu el-rod failures under simulated LOCA conditions. The analytical work provided indications of the actual conditions to be expected in the core during a transient, in terms of potential clad heating rates, internal pressures and transient duration. The experime ntal work applied these parame ters in various combinations to establish the nature of fuel-rod deformation which might occu r under accident conditions. This subject was covered comprehensively in the Statement of Affirmative Te stimony and Evidence of Combustion Engineering in the Matter of Rulema king Hearing for the Acceptance Criteria for Emergency Core Cooling Systems for Light-Water-Cooled Nuclear Power Reactors, Docket Number RM-50-1.

1.8.1.9 Preoperational Vibration Monitoring Program A preoperational vibration monito ring program (PVMP) was comple ted for the Palisades reactor internals. Results of this program were submitted to the AEC by CE Report CENPD-36.

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

In keeping with the requirements for prototype vi bration test programs, predictions of hydraulic forcing functions and structur al response were made for the Maine Yankee and Fort Calhoun reactor internals and correlated to test program measurements. Vibration test data from all three reactors was used in demonstrating the adequacy of the Millstone Unit 2 reactor vessel internals to sustain flow-induced vibration effects. The vibration test data available, together with appropriate analyses, permitted the assessment of design or fabrication differences existing among the subject reactors as th ey related to the vibrational response characteristics of the Millstone Unit 2 reactor internals. A comparis on of applicable design parameters for the Palisades, Fort Calhoun, Maine Yankee and Millstone Unit 2 reactors as of the time of application for operating license is presented in Table 1.8-1.

The analytical methods which fo rmed the basis for the CE vibr ation response predictions were provided in the Maine Yankee and Fort Calhoun vibration m onitoring programs submittals. Palisades, Maine Yankee and Fort Calhoun Flow Model Test reports and a description of the methodology utilized to re late these data to in-reactor forcing functions were provided, as well as a description of the structur al response computer code.

1.8.1.9.1 Basis of Program The suitability of using PVMP data from Palisades, Omaha and Maine Yankee as a composite prototype was based on the following:a.Reactor internals structural response and LOCA hydraulic loadings could be adequately predicted with computer pr ograms 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).

MPS2 UFSAR1.8-6Rev. 35c.Additional instrumentation to measure or derive forcing functions was added to the Fort Calhoun reactor in accordance with Regulatory Guide 1.20 (formerly 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 pe rmits, providing the confi guration and flow modes were similar as specified in Regulator y Guide 1.20 (formerly Safety Guide 20).e.CE provided predictive methodology and predicted and limit ing values of response (acceptance criteria) on the Ma ine Yankee program. The program results were provided on a timely basis in a ccordance with the Regulatory Guide 1.20 (formerly Safety Guide 20).f.CE submitted a report on the LOCA dyna mic analysis methods and procedures.

1.8.1.9.2 Millstone Unit 2 ProgramThe PVMP to be conducted for Millstone Unit 2 reactor internals was c onsistent with those portions of the former Safety Guide 20 (a fter replaced by Regulat ory Guide 1.20), which addressed nonprototype reactors.

The following was the PVMP plan for Millstone Unit 2. As noted above, this program was contingent 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 conditio ns 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 th e preoperational functional tests, the reactor internals were removed from the reactor vessel and visual and nondestructive examination of the reactor intern als 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; MPS2 UFSAR1.8-7Rev. 35c.Those locking and bolting devices whos e failure could adve rsely af fect the structural integrity of the internals;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.A summary of the PVMP inspections described a bove was submitted after the completion of the inspection and tests in a report.

It should be pointed out that th e reactor thermal shield was re moved from the lower internals assembly because of the damage suffered due to excessive vi bratory movement. An evaluation was performed to assess the effects of thermal shield rem oval on the vibratory response of the rest of reactor internals. It was concluded that the effect would be minimal and that the conclusions of the PVMP were still valid.

1.8.2 SPECIAL FOR MILLSTONE UNIT 2 1.8.2.1 Release of Radioactivity in Case of Dama ged Fuel Assemblies in Spent Fuel Pool In the event of release or radioactivity resulti ng from damaged fuel in the spent fuel pool, the auxiliary exhaust system (AES) which is described in Sect ion 9.9.8, diverts the effluent through the enclosure building filtration system (EBFS) charcoal filter s prior to release through the Millstone stack. The AES maintain s the fuel handling area under a negative pressure to limit uncontrolled release of radioactivity.

1.8.2.2 Hydrogen Control The independent systems in th e hydrogen control systems mo nitor and mix hydrogen in the containment following a LOCA (s ee Section 6.6). Each is a full capacity, completely redundant, independent system. Air to operate the hydrogen monitoring system CIV's is provided by the instrument air system wi th a backup air bottle system that is designed to m eet single failure criteria. Two, full capacity hydrogen purge systems not credited in accident analyses are provided.

The hydrogen recombiner syst em has no mitigating function.

1.8.2.3 Common Mode Failures and Antici pated Transients Without Scram CE analyzed the response of pressurized water re actors which are typical of Millstone Unit 2 to demonstrate the diversity of the reactor protective system in mitigating common mode failures and the response of the plant to anticipated transients without scram (ATWS). Results of these studies were submitted to the AEC as topical reports.CE Report CENPD-11, entitled "Reactor Protection System Diversity" wa s submitted on March 4, 1971. This report evaluated systematic, nonrandom , concurrent failures , (i.e., common mode MPS2 UFSAR1.8-8Rev. 35 failures) of redundant devices not considered credible based on quality assurance in design, qualification testing, and periodic testing that co mmon mode failure could disable all instrument channels which measure a given process parameter, the report, nevert heless, addresses this type of failure. Monitoring of the conditi on by diverse means or principles enables a protection system to withstand common mode failures. The evaluations included the following accidents: control element assembly (CEA) w ithdrawal, CEA drop, loss of reactor coolant flow , excess load, loss of load and loss of feedwater. The results of the st udy demonstrated that the diversity of the reactor protective system is such that gross fuel damage or consequential failures in the RCS or in the main steam system will not occur for any of the accidents analyzed.

A draft of the CE report, entitled "Topical Report on Anticipated Transients Without Scram" (Proprietary) was submitted to the AEC on January 10, 1972. Evaluati ons were performed in this report based upon the assumption that no CEA are inse rted into the core during the course of the following transients: CEA withdrawal, CEA drop, idle loop startup, loss of flow, boron dilution, excess load, loss of load, loss of feedwater, sample line break, and pressurize r safety valve failure.

The transient resulting from loss of normal onsite and offsite power was also analyzed but with a conservative one percent negati ve reactivity insertion assume d following reactor trip signal generation, since for this case the failures which initiate the transient would also remove power from the control element drive mechanism (CEDM), allowing th e CEAs to insert. The final report, with results and their applicability to Millstone Unit 2, was submitted to the AEC.

1.

8.3 REFERENCES

1.8-1Millstone Unit 3, Final Safe ty Analysis Report, Section 13.1 - Organizational Structure.

MPS2 UFSARMPS2 UFSAR1.8-9Rev. 35TABLE 1.8-1 COMPARISON OF PREOPERATIONAL VIBRATION MONITORING PROGRAM DESIGN PARAMETERS

<Parameter>PalisadesFort CalhounMaine YankeeMillstone Unit 2 R mean , 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: R mean, 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: R mean, inches 75-3/860-11/16 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 141Core Cylinder OD, inches Integral Integral 145 145 Support Cylinder L, inches Integral Integral 42 42Structure SupportedIntegralIntegralCSB FlangeCSB Flange Core Shroud SupportBolted to CBSBolted to CBSBolted to CBSBolted to CBS Core Shroud: R mean, 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 MPS2 UFSAR 1.8-10 Rev. 35MPS2 UFSAR CSB = Core Support BarrelUGS = Upper Guide StructureVelocity = Design Minimum VelocityThermal ShieldNoYesYesYesNumber of Loops2232Design Minimum. Flow, 10 6 lbm/hr12571.7122139Inlet Design Temperature, F548547546544 Inlet ID, inches (a)35-1/8 28.75 39 35-3/16 Outlet ID, inches (a)48-5/8 37 40 48-1/8Inlet Pipe Velocity, ft/sec 37.7 33.7 39.2 41.6Downcomer Velocity, ft/sec 19.6 25.2 24.9 26.7Core Inlet Velocity, ft/sec 12.2 12.4 13.0 15.4Outlet 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.TABLE 1.8-1 COMPARISON OF PREOPERATIONAL VIBRATION MONITORING PROGRAM DESIGN PARAMETERS (CONTINUED)

<Parameter>PalisadesFort CalhounMaine YankeeMillstone Unit 2 MPS2 UFSAR1.9-1Rev. 35 1.9 TOPICAL REPORTS In support of the Final Safety Analysis Report, various "topica l reports" prepared by Combustion Engineering, Inc., and Bechtel Corporation were referenced throughout this document. A list of "topical reports" as of the time of application for operating li cense is given in Table 1.9-1.

MPS2 UFSAR1.9-2Rev. 35TABLE 1.9-1 TOPICAL REPORTS Combustion Engineering, Inc.Title Millstone Unit 2 Original FSAR SectionASME paper 68-WA/HT-34, December 1968 Winter Annual Meeting1.8.1.2Statement of Affirmative Testimony and Evidence of Combustion1.8.1.2 Engineering in the matter of Rule making Hearing for the Acceptance Criteria for Emergency Core Cooling System for Light-Water-Cooled Nuclear Power Reactors, Docket Number RM-50-1 1.8.1.8 Dynamic Analysis of Reactor Vessel Internals Under Loss of Coolant Accident CENPD-42-3 (Submittal to AEC in July 1972) 1.8.1.6 Thermal Shock Analysis of Reactor Vessels Due to Emergency Core Cooling System Operation, A-68-9-1, March 15,1968, submitted as part of Amendment 9 to the Maine Yankee license application 1.8.1.7 Experimental Determination of Limiting Heat Transfer Coefficients During Quenching of Thick Steel Plates in Water, A-68-10-2, December 13, 1968 1.8.1.7Finite Element Analysis of Structural Integrity of a Reactor Pressure Vessel During Emerge ncy Core Cooling, A-70-19-2, January 1970 1.8.1.7 Palisades Precritical Vibration Monitoring Program, CENPD-361.8.1.9Precritical Vibration Monitoring Program, CENPD-551.8.1.9 Reactor Protective System Diversity, CENPD-11, February 19711.8.2.3Topical Report on Anticipated Transients Without Scram, CENPD-411.8.2.3 INTHERMIC, A Computer Code fo r Analysis of Thermal Mixing, CENPD-8 3.5.3COSMO IV, A Thermal and Hydraulic Steady State Design Code for Water Cooled Reactors, CENPD-9 3.5.3 Seismic Qualification of Category I Electric Equipment for Nuclear Steam Supply Systems, CENPD-61 7.2.6.3 MPS2 UFSAR1.9-3Rev. 35TABLE 1.9-1 (CONTINUED) TOPICAL REPORTSBechtel CorporationTitle Millstone Unit 2 Original FSAR Section Consumer Power Company Palisades Nuclear Power Plant Containment Building Liner Plate Design Report, B-TOP-1 (submitted to AEC in October, 1969) 5.2.4.5Full-Scale Buttress Test for Prestressed Nuclear Containment Structures, BC-TOP-7 5.2.3.3.3Testing Criteria for Integrated Leak Rate Testing of Primary Containment Structures for Nuclear Power Plants, BN-TOP-1 5.2.9.1Design for Pipe Break Effects, BN-TOP-2 (REV. 1)Question 4.16 MPS2 UFSAR1.10-1Rev. 35 1.10 MATERIAL INCORPORATED BY REFERENCEThe following is a list of mate rial incorporated by reference in the Final Safety Analysis Report (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.

(1) Information incorporated by reference into the Final Safety Analysis Re port is subject to the update and reporting requirement s 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)).

MPS2 UFSAR1.A-1Rev. 35 1.A AEC GENERAL DESIGN CRITERIA FOR NUCLEAR POWER PLANTS 10 CFR PART 50 APPENDIX AOn February 20, 1971, the Atomic Energy Commis sion published in the Fe deral Register the General Design Criteria for Nuclear Power Plants. Prior to this date, proposed General Design Criteria for Nuclear Power Plants as issued on July 11, 1967, in the Federal Register were in effect. Before issuance of the construction permit for Millstone Unit 2, discussions reflecting the design intent in consideration of the 1967 proposed criteria were submitted in the PSAR. Design and construction was thus in itiated and has been comple ted based upon the 1967 proposed criteria.Since February 20, 1971, the applicants have attemp ted to comply with the intent of the newer General Design Criteria to the extent possible, recognizing pr evious design commitments. The extent to which this has been possible is refl ected in the discussions of the 1971 General Design Criteria which follow.CRITERION 1 - QUALITY STANDARDS AND RECORDSStructures, 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 rec ognized codes and standards ar e 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 assu rance program has be en 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, system s, and components important to safety are maintained by or under the co ntrol of the nuclear power unit licensee throughout the life of the unit.

Discussion of the quality standard s for those structures and compone nts which are essential to the prevention of incidents which would affect the public health and sa fety or to miti gation of their consequences are presented in appropriate sect ions of the FSAR. The quality assurance program in effect to assure that these structures, systems, and components will satisfactorily perform their safety functions is di scussed in Section 12.8.

For example, components of the safety injection and containmen t cooling systems are designed and fabricated in accordance with established codes and/

or standards as required to assure that their quality is in keep ing with the safety function of the component. It is not intended, however, to limit quality standards requirements to this list.

High Pressure Injection, Low Pressure Injection, and Containment Spray Pumpsa.Surfaces of pressure retaining material s for the high and low pressure safety injection pumps were examined by liquid penetrant techniques in accordance with MPS2 UFSAR1.A-2Rev. 35 the provisions of ANSI-B31.1, Paragra ph 136.5.3(d). Surfaces of pressure retaining materials for the containmen t spray pumps were examined by dye penetrant techniques in accordance with the provisions of Draft ASME Code for Pumps and Valves for Nuclear Power, Class II, 1968. Casings for al l three types of pumps have been hydrostatically tested to at least 1.5 times the design pressures.b.Pressure containing butt welds for th e 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 requi rements of the Hydraulics Institute.e.The pump seals have been designed to provi de a high degree of assurance of their proper operation, including compatibility of seal materials with water chemistry conditions and minimum dependence on ex ternally supplied cooling water

.f.Pump drive motors conform to NEMA Standards, MG-1.Safety Injection Tanks ASME Code,Section III, Class C.

Safety Injection and Containment Spray System Motor-Operated Valves and Control Valvesa.The design criteria for pressure containing parts is in accordance with ANSI B16.5.b.Radiographic inspection of pressure cont aining 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 hydrostatic ally tested in accordance with MSS-61.e.Isolation valves are designed, fabricate d, and tested in accordance with Draft ASME Code for Pumps and Valves for Nuclear Power, Class II, 1968. Control valves are designed, fabricated, and te sted in accordance with ASME Code Section III, Nuclear Power Plant Components, Class II, 1971.

MPS2 UFSAR1.A-3Rev. 35 Containment Coolersa.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 moto rs, were designed to standards of the Air Moving and Conditioning Association, AMCA-21 1A.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 c ontainment penetrations was designed in accordance with the provisions of AN SI B31.1.0. The penetrations piping was designed to ANSI B31.7, Class II and th e penetration isolation valves to the ASME Pump and Valve Code, Class II.f.Valves, other than the penetration isol ation valves, were designed in accordance with ANSI B31.1.0 and ANSI B16.5. Manually operated butterfly valves were in

accordance with AWWA-C504.Shutdown Heat Exchangersa.Pressure containing materials were tested and examined per ASME Code,Section III, Class C.b.Heat transfer design and physical design are in acco rdance 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 hydrostatic ally tested in accordance with ASME Code,Section III, Class C.

MPS2 UFSAR1.A-4Rev. 35 All tests and inspections are reviewed during material procurement an d fabrication of the components to assure conformance with the quality control techniques of the applicable codes and standards.

The appropriate sections in the FSAR discuss the specific c odes and standards invoked in fabricating or erecting the structures, systems, and components important to safety.

Appropriate records of the design, fabrication, erection, and te sting of structures, systems, and components important to safety sh all be maintained for the life of the plant. (See Section 12.8).CRITERION 2 - DESIGN BASES FOR PROTECTION AGAINSTNATURAL PHENOMENAStructures, 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 perf orm 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 surr ounding 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.

All structures, systems, and co mponents important to safety ha ve been designed to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena. The most severe natural phe nomena which are consider ed and discussed in other sections of this FSAR are as follows:a.Earthquakes / Seismology Section 2.6 b.Wind and Tornadoes / Meteorology Section 2.3c.Floods / Hydrology Section 2.5.4 Appropriate natural phenomena ar e considered in the designs of structures, systems, and components. Accepted standards for the forces imposed by natural phenom ena are used in the design.A general description of the seismic analysis program is found in Section 5.8. Additional information on major structure design against the effects of natura l phenomena is included in the following sections:Containment Structure Section 5.2 Enclosure Building Section 5.3 Auxiliary Building Section 5.4 Turbine Building Section 5.5 MPS2 UFSAR1.A-5Rev. 35Intake Structure Section 5.6 Reactor Vessel Internals Appendix 3.A Reactor Coolant System Appendix 4.A CRITERION 3 - FIRE PROTECTIONStructures, systems, and components importa nt 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 appr opriate capacity and ca pability 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.

Millstone Unit number 2 structur es, systems, and components impor tant to safety are designed and located to minimize the probability and effects of fires. Fi re protection systems (active and passive) have been provided to assure that all possible fires are de tected, controlled, and extinguished.

Fire protection and detection systems and com ponents are designed and in stalled in accordance with applicable requirements of the National Fire Protection Association (NFPA). In areas where combustible material may exist, fixed fire detection and suppression are generally provided (Section 9.10).

Fire detection and fire suppres sion systems of appropr iate types and capacities are designed to minimize the adverse effects of fires on structures, systems, and components important to safety.

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

Fire fighting systems are designed to assure that their rupt ure or inadverten t operation does not significantly impair the capabilities of any structure, system, or component important to safety.In areas where water may cause dama ge to safety equipmen t, such as vital elec trical panels or the emergency diesel generators, either shielding is provided or the water suppression system is designed such that its actuation does not affect the safety systems it protects (pre-action sprinkler system, manual activation, shielding, etc.).CRITERION 4 - ENVIRONMENTAL AND MISSILE DESIGN BASESStructures, systems, and com ponents important to safety ar e designed to acc ommodate the effects of and to be compatible with the e nvironmental conditions as sociated with normal operation, maintenance, testing, and postula ted accidents, including loss-of-coolant accidents. These structures, systems, and co mponents are appropriately protected against MPS2 UFSAR1.A-6Rev. 35dynamic effects, including the effects of missiles, pipe whi pping, and dischar ging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit.However, dynamic effects associated with postu lated pipe ruptures in nuclear power units may be excluded from the design basis when analyses, reviewed and approved by the commission, demonstrate that th e probability of fluid system piping rupture is extremely low under conditions cons istent with the design basis for the piping.All structures are designed in accordance with accepted and time proven building codes (as specified in Section 5.1.2) for the loading conditions stated in Sections 5.2.2, 5.3.3, 5.4.3, 5.5.3 and 5.6.3 which insures that they will operate under normal conditions in a safe manner. In addition, those structures and/or components which could affect public safety were designed to function safely during an eart hquake as discussed in Section 5.8. Wind and tornado storm protection design cr iteria are discussed in Sections 5.2.2.1.6, 5.3.3.1.4, 5.4.3.1.6, 5.5.3.3.2, 5.6.3.1.5, and 5.7.3.1.4. Protection agains t postulated missiles is discussed in Section 5.2.5.1.

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

Systems and components important to safety are designed to operate satisfactorily and to be compatible with environmental conditions as sociated with normal operation and postulated accident conditions. Those system s and components located in th e containment are designed to operate in an environment of 289

°F and 54 psig. Systems and compone nts important to safety are designated Seismic Class I and designed in acco rdance with the criteria given in Section 5.2.4.3.

Missile protection and pi pe whipping protection criteria fo r these systems and components are given in Sections 5.2.5.1 and 5.4.3.1.

Leak-before-break (LBB) analyses for the reacto r coolant system (RCS) main coolant loops, for the pressurizer surge line, and unisolable RCS portions of the safety injection and shutdown cooling piping, which demonstrated that the pr obability of fluid syst em piping rupture was extremely low, were reviewed and approved by the commission.

Subsequent to the commission review and approval, weld overlays were applied to dissimilar metal welds (DMWs) at the shutdown cooling, the safety injection and the pressurizer surge nozzles.

A revised LBB analysis was performed for these nozzles (see Reference A.30). Accordingly, pursuant to GDC 4, 1998 revision, the dynamic effects associated with pi pe ruptures in the above piping segments, including the effects of pipe whipping and discharg ing fluids have been excluded from the design basis 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 MPS2 UFSAR1.A-7Rev. 35Steam Generator Blow Down pipingSteam Generator Blow Down sampling pipingCRITERION 5 - SHARING OF STRUCTURES, SYSTEMS, AND COMPONENTSStructures, systems, and components important to safe ty 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, includi ng, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

Both the auxiliary and the turbine buildings of Mill stone Unit 2 are structur ally connected to their respective Millstone Unit 1 buildings. The combined buildings are isolated in the lateral direction as discussed in Secti on 5.4.1 (auxiliary building) and Section 5.5.1 (turbine building). All vertical loads which may interact between Millstone Unit 1 and Millstone Unit 2 portions of the buildings were investigated to ensure that they will function safely under all design conditions.

The Millstone Unit 2 Condensate Polishing Facilit y is located in Warehouse Number 5, which is situated North of the Millstone Unit 2 Turbin e Building and South of the Millstone Unit 3 Condensate Polishing Facility and Auxiliary Boiler Building.

A list of shared facilit ies appears in Section 1.2.13.

The safe shutdown of any unit wi ll not be impaired by the failure of the facilities and systems which are shared.CRITERION 10 - REACTOR DESIGNThe reactor core and associated coolant, c ontrol and protection syst ems 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.

Plant conditions have been cate gorized in accordance with th eir anticipated frequency of occurrence and risk to the public, and design requirements are given for each of the four categories. These categor ies covered by this crit erion are Condition I -

Normal Operation and Condition II - Faults of Moderate Frequency.

The design requirement for Condition I is that margin shall be provided between any plant parameter and the value of that parameter which would require either automatic or manual protective action; it is met by providing an adequate contro l system. The design requirement for Condition II is that such faults shall be accommodated with, at most, a shutdown of the reactor, with the plant capable of returning to operation after corrective action; it is met by providing an adequate protective system. The following design limits apply:a.The value of the departure from nucleat e boiling ratio (DNBR) will not be less than its design limit to ensure that fuel failure does not occur

.

MPS2 UFSAR1.A-8Rev. 35b.The peak temperature in the fuel will be less than the melting point of irradiated UO 2 (considering effects of ir radiation on melting point).c.The maximum primary stresses in the zi rcaloy fuel clad shall not exceed two-thirds of the minimum yield strength of the material at the operating temperature.d.Net unrecoverable circumferen tial strain shall not exceed 1 percent as predicted by computations considering clad creep and fuel-clad interaction effects.e.Cumulative strain cycling us age, defined as the sum of the ratios of the number of cycles at a given ef fective strain range (E) to the permitted number (N) at that range shall not exceed 1.0.f.The fuel rod will be designed to pr event gross clad deformation under the combined ef fects of external pressure and long term creep.The thermal margins during norma l operation ensure that the minimum thermal margins during anticipated operational occurrences do not excee d the design basis. The DNBR limit ensures a low probability of occurrence of DNB.

The occurrence of DNB does not necessarily si gnify cladding damage; it represents a local increase in temperature which may or may not cause thermal damage , depending upon severity and duration.

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

Limitation of fuel burnup will be determined by ma terial rather than nucle ar considerations. See references in Chapter 3. Sufficient margin is provide d in this core design to allow for the ratio of peak-to-average burnup.CRITERION 11 - REACTOR INHERENT PROTECTIONThe reactor core and associated coolant sy stems 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.

The combined response of the fuel temperature coef ficien t, the moderator temperature coefficient, the moderator void coefficient, and the moderator pressure coefficient to an increase in reactor power in the power operating range will be a decrease in reactiv ity; i.e., the inherent nuclear feedback characteristics will not be positive.The reactivity coefficients for this reactor are listed in Table 3.4-2 and are disc ussed in detail in Section 3.4.3.

MPS2 UFSAR1.A-9Rev. 35CRITERION 12 - SUPPRESSION OF REACTOR POWER OSCILLATIONSThe reactor core and associated coolant, c ontrol, and protection systems are designed to assure that power oscillat ions which can result in c onditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.The reactor core is desi gned not to have sustained power osc illations. If any power oscillations occur, the control system is suffic ient to suppress su ch oscillations.

The basic stability of a pre ssurized water reactor with UO 2 fuel is due to the fast acting negative contribution to the power coeffici ent provided by the Doppler effect.

Any trend toward xenon oscillati ons which may occur in the core are controlled and suppressed by movement of the control elemen t assemblies (CEAs) so that the thermal design bases are not exceeded. Xenon oscillations ar e characterized by long periods and slow changes in power distribution. The nuclear instrument ation will provide the informat ion necessary to detect these changes.Xenon stability analysis fo r Millstone Unit 2 is discussed in Section 3.4.5. The re actor protective system is discussed in Section 7.2.

The reactor protective system automatically tr ips the reactor if axia l xenon oscillations are permitted to approach unsafe limits (Sections 7.2.3.3.10 and 1.7.6).CRITERION 13 - INSTRUMENTATION AND CONTROL Instrumentation are provided to monitor va riables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, includi ng those variables and systems that can affect the fission process, the inte grity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls are provided to maintain these variables a nd systems within prescribed operating ranges.Instrumentation is provided, as required, to monitor and maintain significant process variables

which can af fect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Controls are provided for the purpose of maintaining these variables within th e limits prescribed for safe operation.

The principal variables and systems to be mon itored include neutron le vel (reactor power); reactor coolant temperature, flow, and pressure; pressurizer liquid level; steam generator level and pressure; and containment pressure and temperature. In addition, instrumentation is provided for continuous automatic monitoring of process radiation level and bor on concentration in the reactor coolant system.

MPS2 UFSAR1.A-10Rev. 35 The following is provided to m onitor and maintain control ove r the fission process during both transient and steady state periods over the lifetime of the core:a.Ten independent channels of nuclear inst rumentation, 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 us ed to initiate a reactor shutdown in the event of overpower; two Reactor Regulating cha nnels 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.

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

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

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

The containment pressure and temperature instrumentation is designed to monitor these parameters during normal operation and th e full range of postulated accidents.

The instrumentation and control systems are described in detail in Chapter 7.CRITERION 14 - REACTOR COOLANT PRESSURE BOUNDARY The reactor coolant pressure bounda ry is designed, fabricated, er ected and tested so as to have an extremely low probabili ty of abnormal leakage, of rapidly propagating failure and of gross rupture.

Reactor coolant system com ponents are designed in acco rdan ce with the ASME Code,Section III, Pump and Valve Code (reactor co olant system pumps), and ANSI B31.7 (see Section 4 for codes and effective dates). Quality control, inspection, and te sting as required by these standards and allowable react or pressure-temperature operations ensure the integrity of the reactor coolant system.

The reactor coolant system components ar e considered Class I for seismic design.

MPS2 UFSAR1.A-11Rev. 35CRITERION 15 - REACTOR COOLANT SYSTEM DESIGNThe reactor coolant system and associated auxiliary, contro l, and protection system is designed with sufficient margin to assure th at the design conditions of the reactor coolant pressure boundary are not exceeded during a ny condition of normal operation, including anticipated operational occurrences.The design criteria and bases fo r the reactor coolant pressure boundary are described in the response to Criterion 14.

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

The reactor protective syst em functions to minimize the deviat ion from normal ope rating limits in the event of certain anticipated operational occurrences. The results of analyses show that the design limits of the reactor cool ant pressure boundary are not exceeded in the event of such occurrences.CRITERION 16 - CONTAINMENT DESIGNReactor containment and associated systems are provided to es tablish an essentially leak-tight barrier against the uncontrolled release of radioactivit y to the environment and to assure that the containment design conditions im portant to safety are not exceeded for as long as postulated accide nt conditions require.

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

Designed as a pressure vessel, the containment structure is capable of withstanding all design postulated accident conditions including a loss-of-coolant accident (LOCA). All containment penetrations are sealed as described in Section 5.2.6. Isolation valves ar e provided for all piping systems which penetrate the containm ent, as described in Section 5.2.7.As an extra measure of safety, an enclosure building completely surrounds the containment. In the event of an accident, the enclosure building fi ltration region (EBFR), described in Section 6.7.2, is maintained at a slightly negative pressure to preclude leakage to th e environment. Potential leakage from the containment is channeled into the enclosure building filtration system as described in Section 6.7. Throughline leakage th at can bypass the EBFR is discussed in Section 5.3.4.

CRITERION 17 - ELECTRIC POWER SYSTEMS An on site electric power system and an of f site electric power system are provided to permit functioning of structures, systems, and components important to safety. The safety MPS2 UFSAR1.A-12Rev. 35function for each system (assuming the other system is not functioning) is to provide sufficient capacity and capability to assure that (1) specifie d acceptable fuel design limits 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 trans mission network to the on site el ectric distribution system is supplied by two physically independent circuits (not necessa rily on separate rights-of-way), designed and located so as to minimize to the extent practical, the likelihood of their simultaneous failure under operating a nd postulated accident and environmental con ditions. 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 ci rcuit, 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, a nd other vital safety functions are maintained.

Provisions are included to mini mize the probability of losi ng electric power from any of the remaining supplies as a result of, or coinci dent with, the loss of power generated by the nuclear power unit, the trans mission network, or from the on site electric power supplies.The off site power supplies system is described in Sections 8.1 and 8.2. The preferred source of auxiliary power for unit shutdown is from or through the rese rve station service transformers.

System interconnection is provi ded by four 345 kV circuits. These transmission lines are on a single right-of-way with each line installed on an independent set of structures. A description of the structure routing configurat ion is described in Section 8.1.2.1.

The combination breaker-and-a-half and double breaker-double bus switching arrangement in the 345 kV substation includes two full capacity main buses. Pr imary and back up relaying are provided for each circuit along wi th circuit breaker failure bac kup protection. These provisions permit the following:a.Any circuit can be switch ed under normal or fault co nditions without af fecting 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 secti on of bus are isolated without interrupting service to any element other than those connect ed to the faulty bus section.

MPS2 UFSAR1.A-13Rev. 35d.The failure of any circuit br eaker to trip within a set time initiates the automatic tripping of the adjacent breakers and thus may result in the loss of a line or 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.

Overhead lines from the switchyard to the reserv e station service tr ansformers are separated at the switchyard structure and are carri ed on separate towers. These transformers are located near each Unit, and are physically is olated from the normal st ation service transforme rs and from the main transformers.

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

The on site power supply system is described in Sections 8.3 and 8.5. Two full capacity, separate and redundant batteries are provided for all DC loads and for 120 volt AC vital instrument loads. In the event that off site power is not availabl e when needed, a "start" signal is given to both emergency diesel generators (DG).

These generators and their auxiliaries are entirely separate and redundant, and each genera tor feeds one 4,160 volt emergency bus. A generator is automatically connected to its bus only if there is no bus voltage and only if the dead bus did not result from protective relay action.

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

A single failure and any possible related failures in that channel cannot adversely affect equipment and components on the other redundant channel.

Due to the redundancy and separation of power s upplies, distribution and control required for vital functions, all components can be readily inspected and tested. Similarly, mo st subsystems can be tested in their entirety.CRITERION 18 - INSPECTION AND TESTING OF ELEC TRIC POWER SYSTEMS Electric power systems importa nt 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 c ontinuity of the system s and the condition of their components. The systems shall be designe d 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, swit ches, and buses, and (2) the ope rability 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, incl uding operation of applicable portions of the MPS2 UFSAR1.A-14Rev. 35 protection system, and the tran sfer of power among the nuclear power unit, the of f site power system, and the on site power system.

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

To verify that the emergency power system will properly respond within the required time limit when required, the following tests are performed:a.Manually initiated demonstration of the ab ility of the diesel-generators to start, synchronize and deliver power up to 2750 kW continuous, when operating in parallel with other power sources. Norm al unit operation will not be af fected.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 act ual loss of all normal AC station service power. This test will be conducted during each refueling interval.

Demonstration of the au tomatic sequencing equi pment during normal unit operation. This test exercises the contro l and indication devices, and may be performed any time, as the sequencin g equipment is redundant to normal operations. If there is a safe ty 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.Since operation of the protective system will be infrequent, ea ch system is periodically and routinely tested to verify its operability. Each channel of the protective systems, including the sensors up to the final protecti on element, is capable of bei ng checked during reactor operation.

The output circuit breaker s are provided to permit individua l testing during pl ant operation. See Chapters 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). Ad equate 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 locatio ns 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.

MPS2 UFSAR1.A-15Rev. 35 The control room is provided wi th two separate air conditioning systems and two particulate, absolute, charcoal filter unit asse mblies, an airborne radioactivity detector in the fresh air supply line and dampers which act to shunt the intake air through the filters in the event of a high airborne radioactivity level.

The dampers are automatically act uated from the control room monitors. Acting on a high radiatio n level indication, the fresh air dampers clos e and recirculation dampers open to provide a complete closed cycle ventila tion mode with a portion of the air stream being drawn through the HEPA-charcoal filter assembly. In addition, an area radiation monitor is provided to indicate and al arm on high radiation level.

In the event the operator is forced to aba ndon the control room, a ho t shutdown panel (C21) provide the instrumentation and control necessary to maintain the plant in the hot shutdown condition (see Section 7.6.4). The poten tial capability for bringing th e plant to a shutdown is also provided outside the control room.

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

Not all indicators and controls provided on the Fire Shutdown Pane l are available for all fires.

Alternate methods of compliance are documented in the Millstone Unit 2 10 CFR 50 Appendix R Compliance 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 an d to initiate the operation of systems and components important to safety.

The reactor is protecte d by the Reactor Protective System from reaching a condition that could result in exceeding acceptable fuel design limits as a result of anticipated operational occurrences (ANS-N18.2, Condition II). The Prot ective System is designed to monitor the reactor operating conditions and initiate a reactor trip if any of the following measured variables exceeds the operating limits:a.High power level (variable, highe st of thermal or neutron flux).b.High pressurizer pressure.c.Thermal margin (variable low pressure).

d.Turbine trip (equipm ent protection only).

e.Low reactor coolant flow.

MPS2 UFSAR1.A-16Rev. 35f.Low steam generator level.g.Low steam generator pressure.h.Local power density.i.High containment pressure.The Engineered Safeguards Actuation System detects accident conditions and initiates the Safety Features Systems which are designe d to localize, control, mitigate, and terminate such accidents.

The Engineered Safeguards Actuation System prot ects the general public from the release of radioactivity by actuating components that cool the reactor core, depressurize the containment, isolate the containment, and fi lter any containment leakage (see Section 7.3). The following parameters are continuously monitored;a.Low pressurizer pressure.b.High/high-high containment pressure.

c.Containment gaseous a nd particulate radiation.d.Low steam generator pressure.

e.High fuel handling area radiation.

f.Low refueling water storage tank level.g.Emergency bus undervoltage.The Auxiliary Feedwater Automati c Initiation System (AFAIS) pr ovides a dedicated source of feedwater of sufficient capacity to remove decay heat and sensible heat following casualty situations. Automatic initiation of auxiliary feedwater occurs in response to a low Steam Generator 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. Redunda ncy 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 requi red 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.

MPS2 UFSAR1.A-17Rev. 35 The protective system is designed to provide a high functional reliability and inservice testability.

No single failure will result in the loss of the protective function. The protective channels are independent, e.g., with respect to piping, wire routing, mounting and supply of power. This independence permits testing and the removal from service of an y component or channel without loss of the protection function.

Each channel of the protective system, including th e sensors up to the final protective element, is capable of being checked during reactor operation.

Measurement sensors of each channel used in protective systems are checked by observing outputs of similar ch annels which are presented on indicators and recorders on the control board. Trip units and logic are tested by inserting a signal into the measurement channel ahead of the trip units and, upon application of a trip level input, observing that a signal is passed through the trip uni ts and the logic to the logic output relays. The

logic output relays are test ed 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 prot ection function, or is demonstrated to be acceptable on some other defined basis. Design techniques, such as functional diversity or diversity in component de sign and principles of operation, is used to the extent practical to prevent loss of the protection function.

The reactor protective sy stems conform to the pr ovisions of the Institute of Elec trical and Electronic Engineers (IEEE) Cr iteria for Nuclear Power Plan t Protection Systems, IEEE-279, 1971. Two to four independent me asurement channels, complete with sensors, sensor power supplies, signal conditioning units and bistable trip units, ar e provided for each protective parameter monitored by the protective systems. The measurement channels are provided with a high degree of independe nce by separate connection of the channel sensors to the process systems. Power to the channels is provided by independent vital power supply buses. See Section 7.2.Combustion Engineering Topical Report CENPD-11 ("Reactor Protection System Diversity," W. C. Coppersmith, C. I. Kling, A. T. Shesler, and B. M. Tashjian CENPD, February 1971) demonstrates 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 in to a state demonstrated to be acceptable on some other defined basis if condi tions 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 radia tion) are experienced.Protective system instrumentation has been designed to fail into a safe state or into a state established as acceptab le in th e event of loss of power supply or disconnection of the system, Redundancy, channel independence, and separation are incorporated in th e protective system MPS2 UFSAR1.A-18Rev. 35 design to minimize the possib ility of the loss of a pr otection function under adverse environmental conditions. See Sections 7.2 and 7.3.CRITERION 24 - SEPARATION OF PR OTECTION AND CONTROL SYSTEMS The protection system is separa ted from control systems to th e extent that failure of any single control system component or channel, or failure or removal from service of any single protection system component or cha nnel which is common to the control and protection systems leav es 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.The reactor protective systems ar e separated from the control in strumentation systems so that failure or removal from service of any control instrumentatio n system component or channel does not inhibit the function of the protective system. See Section 7.2.

CRITERION 25 - PROTECTI ON SYSTEM REQUIREMENTS FOR REACTIVITY CONTROL MALFUNCTIONS The protection system is designed to assure that specified acc eptable 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.Reactor shutdown with CEA's is accomplished co mpletely independent of the control functions since the trip breakers interrupt power to the full length CEA drive mechanisms regardless of existing control signals. The design is such that the system can withstand a ccidental withdrawal of controlling groups without exceeding acceptable fuel design limits. An analys is of these accidents is given in Section 14.4. The reac tor protection system will prevent specifi ed acceptable fuel design limits from being exceeded for any anticipated transients.CRITERION 26 - REACTIVITY CONTROL SYSTEM REDUNDANCY AND CAPABILITYTwo 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 antic ipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivi ty control system is capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assu re acceptable fuel design li mits are not exceeded. One of the systems is capable of holding the reac tor core subcritical under cold conditions.Two independent systems are provided for contro lling reactivity changes. The Control Element Drive System (CEDS) controls reactivity change required for pow er changes and power distribution shaping, and is also used for reactor protec tion. The boric acid shim control compensates for long term reactivit y changes such as those associ ated with fuel burnup, variation MPS2 UFSAR1.A-19Rev. 35 in the xenon and samarium c oncentrations, and plant cooldow n and heatup. See Sections 7.4.2 and 9.2.2.1.

Either system acting independently is capable of maki ng the core subcritical from a hot operating condition and holding it s ubcritical in the hot standby condition at 532

°F.Either system is able to insert negative reactivity at a sufficiently fast rate to prevent exceeding acceptable fuel design limits as th e result of a power change (i.e

., the positive reac tivity added by burnup of xenon).

The boron addition system is capa ble 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.The combined capability of the reactor cont rol systems in conjuncti on with dissolved boron addition by the safety injection system is such that under pos tulated accident conditions, even with the CEA of highest worth stuck out of the co re, the core would be ma intained in a geometry which 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 react ivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor co olant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impa ir 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.

The basis for selecting the number of control elemen t assemblies in the core inclu des assuring that the reactivity worth of any one assembly is within a preselected maximum value. The control element assemblies have been separated into sets: a shutdown set and a regulating set further subdivided into groups as necessary. Administrative procedures and interloc ks are used to permit only one shutdown group to be wit hdrawn at a time, and to permit withdrawal of the regulating groups only after the shutdown gr oups ar fully withdrawn. The re gulating groups are programmed to move in sequence and within limits that prevent the rates of reactivity change and the worth of individual assemblies from exceeding limiting values. S ee Sections 7.4

.2, 14.4.1, 14.4.2, and 14.4.3.

MPS2 UFSAR1.A-20Rev. 35 The reactor coolant pressure boundary and reactor vessel internals are designed to be capable of accommodating without rupture, and with limited plastic defo rmation, the static and dynamic loads associated with an inadvertent and sudden release of energy to the coolant such as that resulting from CEA ejection, CEA drop, steam li ne rupture or cold water addition. See Sections 14.4.8, 14.4.9, and 14.1.5.The boric acid system rate of react ivity addition is too slow to cause rupture of the reactor coolant pressure boundary or disturb the reactor pressure vessel internals.

CRITERION 29 - PROTECTION AGAINST ANTICIPATEDOPERATIONAL OCCURRENCESThe protection and reactivity control systems are designed to assure an extremely high probability of accomplishing their safety func tions in the event of anticipated operational occurrences.Anticipated operational occurrences have been considered in the design of the protection and reactivity control s ystems. As is demonstrated in the safety analysis in Chapter 14 and the Combustion Engineering Report CENPD-11 ("Reactor Protection System Diversity", W. C. Coppersmith, Cl. L. Kling, A. T. Shesler, and B. M. Tashjian, CENPD-11, February 1971), the design is adequate to minimize the consequences of such occurrences and assures that the health and safety of the public is protected from the consequences of such occurrences.The adherence to a detailed program for quality assurance, careful attent ion to design, component selection and system installati on, coupled with the design features of redundancy, independence, and testability will assure that a high probability exists that the protection and reactivity control systems will accomplish their f unctions. See Criteria 21 through 26.CRITERION 30 - QUALITY OF REACTOR COOLANT PRESSURE BOUNDARY Components which are part of the reac tor coolant pressure boundary are designed, fabricated, erected and tested to the highe st quality standards practical. Means are provided for detecting and, to the extent practi cal, identifying the locat ion of the source of reactor coolant leakage.The reactor coolant pressure boundary components have been designed, fabricated, erected and tested in accordance with the ASME Code Section III, 1971 through summer 1971 Addenda and ANSI B31.7, 1969 as specified in Criterion 14. Repl acement steam generator subassemblies were fabricated in accordance with ASME Code Section III 1983 through summer 1984 Addenda.The replacement reactor vessel closure head including all nozzles (CEDM, HJTC, ICI and the vent) is constructed in accordance with ASME Boiler and Pressure Vessel Code,Section III, Subsection NB, 1998 Edition through 2000 Addenda.Containment sump instrumentation is used to detect reactor coolant system leakage by providing information on rate of rise of sump levels and frequency of sump pump operation. Flow MPS2 UFSAR1.A-21Rev. 35instrumentation indicates and records makeup flow rate and volumes from the primary water system. This instrumentation al lows detection of suddenly occu rring leaks or those which are gradually increasing. The contai nment air monitoring system (see Section 7.5.6) provides an additional means of reactor coolant system leakage detection.CRITERION 31 - FRACTURE PREVENTION OF REACTORCOOLANT 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 manne r and (2) the probability of rapidly propagating fracture is minimized. The de sign 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 irradiat ion on material properti es, (3) residual, steady state and transient stresses, and (4) size of flaws.

Carbon and low alloy steel materials which fo rm part of the pressure boundary meet the requirements of the ASME Code,Section III , paragraph N-330 at a temperature of +40

°F. (Ref. Section 4.2.2). The actual ni lductility transition temperatur e (NDTT) of the materials has been determined by drop weight tests in accord ance with ASTM-E-208. For the reactor vessel base metals, Charpy tests were also performed a nd the results used to plot a Charpy transition curve. To address changes in regulations, th e original design require ments of N-330 were supplemented and the materials' initial nil-ductility reference temperatures (RTNDT) were conservatively established ba sed upon available or supplemental material toughness testing. In the case of the replacement steam generators, the materials were required to satisfy NB-2331 and RTNDT values were established to satisfy current requirements.

Carbon and low alloy steel materials including weld filler meta l which form part of the reactor pressure boundary for replacement reactor vessel closure head satisfy ASME Section III, NB 2000. Actual NDTT was established by drop weight test in accordance with ASTM-E-208 at - 40°F. RTNDT of the replacement head based materials was established by Charpy V-notch test at - 40°F. Charpy transition curves were plotted usi ng test data for the ba se material of the replacement reactor vessel head.

All the reactor coolant pressure boundary compone nts are constructed in accordance with the applicable codes and comply with the test and in spection requirements of th ese codes. These test inspection requirements assure that flaw sizes are limited so that the probability of failure by rapid propagation is extremely remote. Pa rticular emphasis is placed on the quality control applied to the reactor vessel, on which tests and inspections exceeding c ode requirements are performed. The tests and inspections performed on the reactor vessel are summarized in Section 4.6.5.

The reactor vessel beltline materials receive sufficient neutron ir radiation to cause embrittlement (an increase in RTNDT). To provide conservative marg ins against nonductile or rapidly propagating failure, seve ral techniques are employed. Operat ing limits which account for the MPS2 UFSAR1.A-22Rev. 35 RTNDT of all pressure boundary mate rials, both unirradiated and ir radiated, are established in accordance with the requirements of 10 CFR 50 Appendix G (Additi onal details are provided in Section 4.5.1). In addition, compliance with 10 CFR 50.61 assures that the shift in the transition temperature of the reactor vessel beltline materials provides adequate margins of safety against severe pressurized thermal shock events.To assure that the reactor vessel beltline materials are behaving in the predicted manner, a reactor vessel material surveillan ce program is conducted (See Criterion 32 and Section 4.6.2). Toughness testing of unirradiated reac tor vessel materials was perfor med to establish the baseline, and the irradiated surveillance materials are periodically tested as surveillance capsules are removed during the plant's desi gn life, in accordance with the requirements of 10 CFR 50, Appendix H.

The activation of the safety injection systems introduces highly borated water into the reactor coolant system at pressures significantly below operating pressures and wi ll not cause adverse pressure or reactivity effects.

The thermal stresses induced by the injection of cold water in to the vessel have been examined.

Analysis shows the there is no gr oss yielding across the vessel wa ll using the minimum specified yield strength in the ASME Boiler and Pressure Vessel Code,Section II I. (Ref. Section 4.5.4).Adverse effects that could be cau sed by exposure of equi pment or instrumentation to containment spray water is avoided by designing the equipment or instrumentation to withstand direct spray or by locating it or protecting it to avoid direct spray.CRITERION 32 - INSPECTION OF REACTOR COOLANT PRESSURE BOUNDARY Components which are part of the reactor coolant pre ssure boundary are designed to permit (1) periodic inspection and testing of im portant areas and features to assess their structural and leak-tight integrity, and (2) an appropriate materials surveillance program for the reactor pressure vessel.

Provisions are made for inspection, testing, a nd surveillance of the Reactor Coolant System boundary as required by ASME Boiler and Pressure V e ssel Code,Section XI.

The Reactor vessel surveillance program was designed in acc ordance with ASTM E185. It complies with ASTM E185-73 and 10 CFR 50, Appendix H. Section 4.6.3 presents the details of the reactor surveillance program.

Sample pieces taken from the same shell plate material used in fabrication of the reactor vessel are installed between the core a nd the vessel inside wall. These samples will be removed and tested at intervals during vessel inside wall. These samples will be

removed and tested at intervals during vessel life to provide an indication of the extent of the neutron embrittlement of the ve ssel wall. Charpy tests will be performed on the samples to develop a Charpy transition curve. By comparison of this curve with the Charpy curve and drop weight tests for specimens taken at the beginning of the vessel life, the change of NDTT will be determined and operating instructions adjusted as required.

MPS2 UFSAR1.A-23Rev. 35CRITERION 33 - REACTOR COOLANT MAKEUP A system to supply reactor c oolant makeup for protection ag ainst 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 react or 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 sy stem operation (assuming onsite power is not available) the system safety function can be accomplished using the piping, pumps and valves used to ma intain coolant inventory dur ing normal reactor operation.

Reactor Coolant System (RCS) makeup during normal operation is provided by the Chemical and V olume Control System (CVCS) which includes three positive displacement charging pumps rated at 44 gpm each. Tw o operating CVCS pumps are capable of making up the flow loss for leaks in the reactor coolant boundary of up to 0.250 inches equivalent diameter. Two CVCS pumps are sufficient to makeup for a 0.250 inch equivalent diamet er RCS break assuming either:

1) minimum letdown with no RCS leakage or 2) letdown isolated with maximum Technical Specification allowed leakage. This CVCS de sign results in a substantial RCS steady state pressure that is well above the shutoff head of the high pressu re safety injection pumps. The above described CVCS capability fulfills the intent of Criterion 33. Info rmation on CVCS is contained 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.Residual heat removal capability is provided by the shutdown cooling system for reactor coolant temperature less than 300

°F (see Section 9.3). For temperatures greater than 300

°F, this function is provided by the steam generators (see Section 10.3). Sufficient redundancy, interconnections, leak detection, and isolation capabil ities exist with these systems to assure that the residual heat removal function can be accomplished, assuming failure of a single active component. Within appropriate design limits, either system will remove fission product d ecay heat at a rate such that specified acceptable fuel design li mits and the design conditions of the reactor coolant pressure boundary will not be exceeded.

MPS2 UFSAR1.A-24Rev. 35 CRITERION 35 - EMERGENCY CORE COOLINGA 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-wate r reaction is limited to negligible amounts.Suitable redundancy in components and features, and suitable interconnections, leak

detection, isolation, and containment capabiliti es is provided to assure that for onsite electrical power system operation (assuming of fsit e 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.The emergency core cooling system is discussed in detail in Ch apter 6. It consists of the high pressure safety injection subsystem, the low pressure safety inje ction subsystem, and the safety injection tanks (see Section 6.3).

This system is designed to meet the criterion stated above with re spect to the prevention of fuel and clad damage that would interfere with the emergency core cooling function, for the full spectrum of break sizes, and to the limitation of metal-water reac tion. Each of the subsystems is fully redundant, and the subsystems do not sh are active components other than the valves controlling the suction headers of the high and low pressure safety injection pumps. Minimum safety injection is assured even though one of these valves fails to function. These valves are in no way associated with the function of the safety injection tanks.

The ECCS design satisfies the crit eria specified in 10 CFR 50.46(b).CRITERION 36 - INSPECTION OF EMERGENCY CORE COOLING SYSTEMThe emergency core cooling system is designe d to permit appropria te periodic inspection of important components, such as spray ri ngs in the reactor pressure vessel, water injection nozzles, and piping to assure th e integrity and capability of the system.Chapter 6 describes the arrangement and locati on of the components in the emer gency core cooling system. All pumps, the shutdown cooling h eat exchangers, and valves and piping external to the containment structure are accessible for physical inspection at any time. All sa fety injection valves and piping inside the containment structure, and the safety in jection tanks, may be inspected during refueling.

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

CRITERION 37 - TESTING OF EMERGENCY CORE COOLING SYSTEMThe 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, MPS2 UFSAR1.A-25Rev. 35 (2) the operability and performance of the ac tive 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 sequenc e that brings the system into operation, including operation of applicable portions of the protection sy stem, the transfer between normal and emer gency power sources, and the operation of the associated cooling water system.The Emergency Core Cooling System (Safety Inject ion System) is provided with testing facilities to demonstrate system component operability. Testing can be conducted during normal plant operation with the test facilities arranged not to interfere with th e performance of the systems or with the initiation of control circ uits, as described in Section 6.3.4.2.

The safety injection system is designed to permit periodic testi ng of the delivery capability up to a location as close to the core as practical. Periodic pressu re testing of the Safe ty Injection System is possible using the cross connection to the charging pumps in the Chemical and Volume Control System.The low pressure safety injection pumps are us ed as shutdown cooling pumps during normal plant cooldown. The pumps discharge into the safety injecti on header via the shutdown cooling heat exchangers and the low pressure injection lines.With the plant at operating pres sure, operation of safe ty injection pumps may be verified by recirculation back to the refuel ing water storage tank. This will permit verification of flow path continuity in the high pressure injection lines and suction lines from the refueling water storage tank.Borated water from the safety injection tanks may be bled through the recirculation test line to verify flow path continuity from each tank to its associated main safety injection header.

The operational sequence that brings the Safety Inj ection System into acti on, including transfer to alternate power sources, can be tested in parts as desc ribed in Chapters 6, 7, and 8.CRITERION 38 - CONTAINMENT HEAT REMOVALA 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 of fsite power is not available) and for offsite electric power system ope ration (assuming onsite power is not available) the system safety function can be accomplishe d, assuming a single failure.

MPS2 UFSAR1.A-26Rev. 35The containment spray system (Section 6.4) and the containment air reci rculation and cooling system (Section 6.5) are provide d as redundant, independent syst ems, each fully capable of reducing the containment pressu re and temperature following any loss-of-coolant accident (LOCA) and maintaining them at acceptably low levels.Sufficient heat removal capab ility is provided by any of th e following combinations of equipment:a.Two containment spray pumps with associated heat exchangers.b.Three of the four c ontainmen t air recirculat ion and cooling units.c.One containment spray pump with associat ed heat ex changer in combination with two containment air recirc ulation and cooling units.

The containment heat removal syst ems are provided with suitable in terconnections such that each combination of two containment air recirculation an d cooling units and one containment spray pump, aligned with the associated shutdown cooling heat exchanger, are provided with cooling water from the same RBCCW head er and powered by the same emergency bus. All associated components, such as valves, are likewise powered from the same emergency bus. Each combination of these com ponents is capable of removing heat at a rate greater than required to limit the postaccident containment pressure and temperature. A single failure of any active component does not render the redundant group inoperable.

The containment spray system is provided with containment isolation capabilities in accordance with Criterion 56. The above contai nment penetration is provided wi th leak detection capabilities in accordance 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.

Major components of the contai nment spray system are located to permit access for periodic maintenance and inspection. Components of the containment air and reci rculation system are located within the containment and are theref ore accessible for maintenance and inspection during shutdown.

The containment sump is located in the lowest el evation of the containment at Elevation (-)22-6 and is accessible during reactor shutdown for periodic visual inspections (see Section 6.2).The containment spray nozzles are accessible for periodic inspection during reactor shutdown.

MPS2 UFSAR1.A-27Rev. 35CRITERION 40 - TESTING OF CONTAINMENT HEAT REMOVAL SYSTEMThe containment heat removal system is designed to permit appropria te periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the ac tive 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 operati onal 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.

The spray system and the air recirculation a nd cooling systems in the containment have provisions for online testing to a ssure system operation, performanc e and structural and leaktight integrity of the associated co mponents. Testing procedures ar e described in Sections 6.4.4.2 and 6.5.4.2, respectively.

The containment heat removal systems undergo preoperational testi ng prior to plant startup. The test procedure is described in Chapter 13.CRITERION 41 - CONTAINMENT ATMOSPHERE CLEAN UPSystems 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 oxyge n and other substances in the containment atmosphere following postulated accidents to assure th at 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 accomplishe d, assuming a single failure.

The containment is not provided with an atmosp here cleanup system. However, a second barrier, the enclosure building, is provide d around the containment to coll ect potential leakage from the containment under postaccident conditions.

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

The hydrogen control system is provided to mix and monitor the concentration of hydrogen in the containment atmosphere following postulated accidents to assure the containment integrity is MPS2 UFSAR1.A-28Rev. 35maintained. This is discussed in Section 6.6. Re duction of hydrogen concentr ation is not credited for design basis accidents.

Each of these cleanup systems consist of completely redundant , independent safety function. These are provided with suitable interconnections a nd separations such that a single failure in any subsystem does not render the redundant subsystem inoperable.

The hydrogen control system is in corporated with containment is olation capabilities for each piping subsystem which penetrates the primary containment.

Containment isolation is in accordance with Criterion 56. Provi sion for leak detection is inco rporated in accordance with Criterion 54.CRITERION 42 - INSPECTION OF CONTAINMENT ATMOSPHERE CLEANUP SYSTEMS The containment atmosphere cl eanup 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.The enclosure building filtration system (EBF S) is located to pe rmit access for periodic inspection and maintenance. The components of the hydrogen contro l system located outside the containment are accessible for periodic inspection and maintenance. The components located inside containment are accessible for inspection and maintenance during shutdown.The hydrogen control system and EBFS are incorporated with pr ovisions for online testing to demonstrate system operation, performance and integrity. These tests procedures are described in Sections 6.6.4.2 and 6.7.4.2, respectively.CRITERION 43 - TESTING OF CONTAINMENT ATMOSPHERE CLEANUP SYSTEM The containment atmosphere cl eanup 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.

The enclosure building filtration system (EBFS) and hydrogen cont ro l system are incorporated with provisions for online test ing. The test procedures are described in Sections 6.7.4.2 and 6.6.4.2, respectively.

The containment atmosphere cleanup systems undergo preoperational tests prior to plant startup. Test procedures are described in Chapter 13.

MPS2 UFSAR1.A-29Rev. 35CRITERION 44 - COOLING WATER 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, system s, 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 th at for onsite electric power

system operation (assuming of fsit e 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.

The RBCCW system, described in Section 9.4, and the service water system, described in Section 9.7.2, are provided to transfer heat from structures, systems, and components important to safety to an ultimate heat sink.

The systems are designed to transfer the combined heat load of these structures, systems, and componen ts under normal and accident conditions.

The RBCCW supplies cooling water to components important to safety through two independent headers. One header provides ad equate heat removal capability to safely shutdown the plant under accident conditions, but at a lesser rate. Service water is supplied to the RBCCW heat exchangers by two independent headers to assure heat removal capability. Two service water pumps are in continuous operation wi th a spare pump provided. One pump supplies sufficient heat removal capability for the RBCC W heat exchangers to safely shut down the plant and for accident mitigation.

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

The RBCCW and service water sy stems are designed to operate with onsite power (assuming offsite power is not available) and with offsit e power (assuming onsite pow er is not available).The systems are designed such that a single failure in either system will not adversely affect safe operation, 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 exchange rs and piping, to assure the integrity and capability of the system.

The RBCCW system and servi ce water system, excluding unde r ground piping, are designed to permit periodic inspection of impor tant components, such as pumps , heat exchangers, valves and piping to assure the integrity and heat removal ca pability of the system.

The components of the RBCCW system located outside the containment are located in a low radiation area, which MPS2 UFSAR1.A-30Rev. 35permits access for periodic inspection and maintenance during operation. Components of the RBCCW system located inside the containment are accessible for insp ection and maintenance during plant shutdown. Inspection of RBCCW system components is described in Section 9.4.4.2.

Major service water system compone nts, such as pumps and strain ers, are accessible for periodic inspection during normal operation.

Inspection of the service wate r system is described in Section 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 in tegrity of its components, (2) the operability and the performance of the active components of th e 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 syst em and the transfer between normal and emergency power sources.

Online testing provisions are incorporated in the RBCCW and service water systems to demonstrate the operability, performance, structural and leaktight integrity of the systems. The RBCCW and service water systems are designed so that under condi tions as close to design as practical, the performance shall be demonstrated of the full opera tional sequence that brings the system into operation, including ope ration of applicable portions of the protection system, and the transfer between normal and emergency power sources. Testing of the RBCCW and service water systems 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, pe netrations, 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 pressu re and temperature condi tions 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 c ooling functioning, (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational m odel and input parameters.

The containment structure, incl uding the access openings, penetrati ons and the containment heat removal system, is designed to withstand a pr essure of 54 psig an d a temperature of 289

°F following a loss-of-coolant acci dent (LOCA) or a main stea m line break accident (see Section 14.8.2). Details of the methods used to analyze the containment structure are described in Section 5.2.2. To obtain an adequate margin of safety, a factored load was selected for a design which allows a 25 percent increase over the calculated postulated accident load.

MPS2 UFSAR1.A-31Rev. 35 A high degree of leak tightness is provided by a one-quarter inch thick steel liner plate which completely encloses the interior surface of the containment stru cture. Components of the liner plate, such as penetra tion sleeves, personnel locks, and equipm ent hatch, are designed to meet the requirements of the ASME Boiler and Pressure Vessel Code,Section III (Nuclear Vessels) 1968 Edition through the summer 1969 addenda Paragraph N-1211. Furthe r description of the liner plate is contained in Section 5.2.3.As a further check on the design a structural integrity test, composing a test pressure load of 115 percent of the design accident pres sure load, is conducted prior to operation. In addition to this, a leak rate test will be conducted prior to opera tion and at certain interv als during operation. Details of the leak rate test ar e provided in Section 5.2.8.1.

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

The containment consists of a prestressed reinforced concrete cylinder a nd dome connected to and supported by a massive reinforced concrete sla

b. A one-quarter inch thick steel liner plate is attached to the inside surface of the concrete containment and it s penetrations. C onsideration has been given to both design and construction techniques to assure the containment pressure boundary behaves in a ductile manne r and the probability of a ra pidly propagating fracture is minimiz ed.The liner plate is designed to car ry no load, and serves only as a leaktight barrier. Analytical calculations of the strains under an extreme and most improbably se t of load conditions indicate the strains are well within the ductile limits of the material. The analytical approach to liner design is presented in the Bechtel Co rporation Proprietary Report B-TOP-1.

At all penetrations the liner pl ate is thickened usi ng the 1968 ASME Code,Section III for Class B Vessels as a guide to limi t stress concentrations.

Provisions, as described in Section 5.2.5.1.1, are made to prevent a potential internally generated missile from ruptur ing the liner plate.

Materials for the penetrations require satisfactory Charpy V-notch impact test results. All penetrations are stress relieve

d. The construction materials sele cted for the liner plate and penetrations are given in Section 5.2.1.

MPS2 UFSAR1.A-32Rev. 35 Additional details c oncerning the construction techniques and inspection provisions are outlined in Section 5.9.3.5.CRITERION 52 - CAPABILITY FOR CONTAINMENT LEAKAGE RATE TESTING The reactor containment and ot her equipment which may be s ubjected to containment test conditions are designed so that periodic integrat ed leakage rate testing can be conducted at containment design pressure.

The reactor containment and other eq uipment which is subjected to containment test conditions are designed so that periodic in tegrated leakage rate testing can be conducted at containment design pressure. The test procedure is described in Section 5.2.8.CRITERION 53 - PROVISIONS FOR CONTAINMENT TESTING AND INSPECTION The reactor containment is de signed to permit (1) appropria te 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 ti ghtness of penetrations which have resilient seals and expansion bellows.

The reactor containment is designed to permit appropriate peri odic testing of all important areas. Details of the containment testing and inspection are discussed in Section 5.2.8.CRITERION 54 - PIPING SYSTEMS PENETRATING CONTAINMENT Piping systems penetrating prim ary reactor containment are pr ovided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to sa fety 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 a nd to determine if valve leakage is within acceptable limits.

Piping systems penetrati ng containment are provided with su itable redundancy to assure the systems function adequately during postulated accidents such that fa ilure of a portion of a system will not create a hazard to sa fe unit operation. Piping systems are provided with containment isolatio n valves in accordance with the requi rements of Criterion 55, 56, and 57. Containment isolation valves have been sel ected and tested to provide adequate opera tion at maximum flow conditions. Provisions are incorporated for leak detection and performance testing of those piping systems penetrating the containment (Section 5.2.7.4.2).CRITERION 55 - REACTOR COOLANT PRESSURE BOUNDARY PENETRATING CONTAINMENTEach line that is part of the reactor coolan t pressure boundary and that penetrates primary reactor containment is provided with containment isolation valv es as follows, unless it can MPS2 UFSAR1.A-33Rev. 35 be demonstrated that the contai nment isolation provisions for a specific class of lines, such as instrument lines, are accepta ble on some other defined basis:

(1) One locked closed isolation valve inside and one locked clos ed 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 is olation 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 contai nment 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 requiremen ts to minimize the probability or consequences of an accidental rupture of these lines or of line s connected to them are provided as necessary to assure adequate safety. Determ ination 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.

For those piping systems penetrat ing the containment and connect ed directly to the reactor coolant pressure boundary, isolation provisions have been incorporated. Section 5.2.7 indicates applicable valve arrangements, a complete description of pene trations and valve position on air/power failure.

Provisions are made fo r leak testing as desc ribed in Section 5.2.7.4.2.CRITERION 56 - PRIMARY CONTAINMENT ISOLATION Each line that connects directly to the cont ainment atmosphere a nd penetrates primary reactor containment is provided with containment isolation valv es as follows, unless it can be demonstrated that the contai nment isolation provisions for a specific class of lines, such as instrument lines, are accepta ble on some other defined basis:

(1) One locked closed isolation valve inside and one locked clos ed isolation valve outside containment; or MPS2 UFSAR1.A-34Rev. 35 (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 is olation 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 contai nment are lo cated 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.

For those piping system penetrating the containment and connected directly to the containment atmosphere, isolation provisions have been incorporated. Section 5.2.7 indicates the applicable valve arrangements, a complete description of penetrations and valve position on air/power failure.CRITERION 57 - CLOSED SYSTEM ISOLATION VALVESEach line that penetrates primary reactor containment and is neither part of the reactor coolant pressure boundary not connected directly to the containmen t atmosphere has at least one containment isolation valve which shall be either automatic, or locked closed, or capable of remote manual operation. This valv e 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.

For those piping systems penetrati ng the containment which are neither part of the reactor coolant pressure boundary nor connected di rectly with the containment at mosphere, isolation provisions have been incorporated.Section 5.2.7 indicates appl icable valve arrangements, a complete description of penetrations and valve position on air/power failure.

CRITERION 60 - CONTROL OF RELEASES OF RADIOACTIVE MATERIALS TO THE ENVIRONMENT The nuclear power unit design includes mean s 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 material s, particularly where unfa vorable site environmental conditions can be expected to impose unusua l operational limitations upon the release of such effluents to the environment.

MPS2 UFSAR1.A-35Rev. 35 The radioactive waste processing system (RWS), as described in Section 11.1, is designed to provide controlled handling and disposal of liquid, gaseous, and solid wastes from Millstone Unit 2. The RWS is designed to ensure that the general public and plan t personnel are protected against exposure to radioactive material in accordance with 10 CFR Part 20, Sections 1301 and 1302, and Appendix B and 10 CFR Part 50, Appendix I.

All liquid and gaseous radioactive releases from the RWS ar e designed to be accomplished on a batch basis. All radioactive materials are sampled prior to rel ease to ensure compliance with 10 CFR Part 20, Sections 1301 and 1302, and Appendix B and 10 CFR Part 50, Appendix I and to determine release rates. Radioactive materials which do not meet release requi rements will not be discharged to the environment. The RWS is designed with sufficient holdup capacity and flexibility for reprocessing of wastes to ensure release limitations are met.The RWS is designed to preclude the ina dvertent release of ra dioactive material.All storage tanks in the clean liquid waste and gaseous waste systems are provided with valve interlocks which prevent the addition of waste to a tank which is being discharged to the environment. Each discharge path from the RWS is provided with a radiation monitor which alerts unit personnel and initiates automatic clos ure of redundant isolation valves to prevent further releases in the event of noncompliance to 10 CFR Part 20, Sections 1301 and 1302, and Appendix B.Section 11.1.5 describes th e plant design for the handling of solid wastes.CRITERION 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 a ssure 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 a ppropriate containment, confinem ent, and filtering systems, (4) with a residual heat removal capability ha ving reliability and testability that reflects the importance to safety of decay heat and ot her residual heat removal, and (5) to prevent significant reduction in fuel storage c oolant inventory under accident conditions.

Systems for fuel storag e and handling, and all systems contai ning radioactivity are designed to ensure adequate safety under no rmal and postulated accident condi tions. Design of these systems are described in the sections listed below:SystemSection Reactor Coolant System 4.0Engineering Safety Features Systems6.0 MPS2 UFSAR1.A-36Rev. 35 All components important to the sa fety of these systems are located to permit periodic inspection as required. Suitable shielding, as described in Section 11.2, is provided for these components to protect plant personnel and to allow inspection and testing.To ensure the containment and confinement of radioactivity, all com ponents are designed and tested in accordance with accepted Codes and St andards. All system components are visually inspected and adjusted, if required, to ensure correct installation and arrangement. The completely installed systems were subject to acceptance tests or preoperation tests as described in Chapter 13 to ensure the integrity of the systems.

The spent fuel pool cooling system described in Section 9.5, is desi gned to ensure adequate decay heat removal from stored fuel.

Sections 5.4.3 and 9.5 desc ribe how the spent fuel pool is designed to prevent significant reduction in fuel storage coolant inventory.

CRITERION 62 - PREVENTION OF CRITICALITY IN FUEL STORAGE AND HANDLINGCriticality in the fuel storage and handling system is prevented by physical systems or processes, preferably by use of ge ometrically safe configurations.

New fuel assemblies are stored in dry racks in parallel rows at elevation 38 feet 6 inches of the auxiliary building. The base of the new fuel rack s at elevation 38 feet 6 inches minimizes the possibility of flooding the fuel a ssemblies. Nevertheless, the new fu el racks maintain a center to center distance of 20.5 inches, large enough to prev ent criticality in the un likely event of flooding with unborated water.

Additional details of new fuel stor age are given in Sections 9.8.2.1.1and 9.8.4.1.1.

Spent fuel assemblies are stored in parallel rows at the bottom of the spent fuel pool. The racks are separated into 4 regions, de signated Regions 1, 2, 3, and 4.

Fuel assemblies used at Millstone Unit 2 may include reduced enri chment fuel rods adjacent to guide thimbles and reduced enrichment axial blanket regions. The criticality analyses are performed using a single enrichment in all fuel rods that is the highest initial planar average U-235 enrichment of the axial regions in the fuel assembly. Th is averaged enrichment is designated as the initial planar average enrichment.

Region 1 can store, in a 2 out of 4 storage pattern, any fuel assembly with a maximum initial planar average enrichment up to 4.85 weight percent U-235. The ot her two locations in the 2 out of 4 storage pattern ar e designated as Restrict ed Locations (shown in Figure 9.8-7). Fuel storage rack locations designated as Restricted Locations in Figure 9.8-7 shall remain empty. No fuel Auxiliary Systems9.0Radioactive Waste Processing System11.0SystemSection MPS2 UFSAR1.A-37Rev. 35assembly, no Non-standard Fuel Configuration, no non-fuel component, nor any hardware/material of any kind may be stored in a Restricted Location.

(1)Regions 2 and 4 use fuel burnup credit and store fuel assemblies in a 3 out of 4 storage pattern, in which the fourth location in a 2 x 2 storage array is designated as a Restricted Location per Figure 9.8-7.

Regions 1 and 2 contain Boraflex panels whic h are no longer credited as neutron absorbers.

Region 3 uses fuel burnup credit and has all st orage locations availabl

e. In addition, fuel assemblies stored in Region 3 mu st contain either three Borated Stainless Steel Poison Rodlets (installed in the assembly's cente r guide tube and in two diagonall y opposite guide tubes) or a full length, full strength Control Element Assembly (CEA).

There are also Non-standard Fuel Configurations in the spent fu el pool (SFP). A Non-standard Fuel Configuration is an object containing fuel that does not conform to the standard fuel configuration. The standard fuel co nfiguration is a 14 x 14 array of fu el rods (or fuel rods replaced by un-enriched fuel rods or stainless steel rods) with fi ve (5) guide tubes that occupy four lattice pitch locations each. Fuel in any other array is a "Non-st andard Fuel Configuration." Reconstituted fuel in whic h one or more fuel rods have been replaced by either un-enriched fuel rods or stainless steel rods is consider ed to be a standard fuel configuration.

Note that each of the Non-standard Fuel Configurations must ha ve a separate criticality analysis which may allow storage in one or multiple Re gions, and which may or may not require Borated Stainless Steel Poison Rodlets or a CEA if stored in Region 3.

GDC 62 states that the "Criticalit y in the fuel storage and handli ng system shall be prevented by physical systems or processes, pr eferably by use of geometrica lly safe configurations." As detailed above, the Region 1, 2, 3, a nd 4 storage racks, requi re more than just fuel geometry alone for reactivity control. All four regions credit soluble boron in the spent fuel pool water. Regions 1, 2, and 4 credit Restricted Locations per Figure 9.8-7. Regions 2, 3, and 4 use fuel burnup credit.

Region 3 requires that fuel assemblies contain either three Borated Stainless Steel Poison Rodlets or a full length, full strength CEA (note that the criticality analysis of a given Non-standard Fuel Configuration may qualify it for Region 3 storage without these inserts).

Administrative controls are used to ensure proper placements of Borated Stainless St eel Poison Rodlets and CEAs, use of soluble boron and fuel burnup credit, and control of Restricted Locations. Further, for accident conditions, soluble boron is credited in the spent fuel pool water. The NRC has concurred that the credit for these neutron pois ons, soluble boron, fuel burnup cred it, Restricted Locations, and associated administrative controls are acceptable in meeting the requirements of GDC 62.(1) Note that Region 1 and 2 S FP rack storage locations contain removable Boraflex panel boxes which house the Boraflex panels. The Boraflex panel boxes were manufactu red 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.

MPS2 UFSAR1.A-38Rev. 35Both the spent fuel and new fuel storage racks are designed to preclude any deformation of the racks during earthquake loads that would reduce the center to center spacing to a point where the fuel would approach criticality.

Fuel handling equipment is designed to ensure safe handling of fuel assemblies and to prevent criticality. Section 9.8.4 desc ribes the safety features of the fuel handling equipment.CRITERION 63 - MONITORING FUEL AND WASTE STORAGE Appropriate systems are provi ded in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditi ons that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.Section 9.5.2.1 describes the monitoring and alarm instrumentation provided for the spent fuel storage system to detect conditions that may result in loss of de cay heat removal capability and excessive radiation levels. Section 7.5.6 describes the monitoring provisions for radioactive waste handling 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.

Containment radiation is monitored by gaseous and particulate monitors as described in Sections 7.5.1.2 and 7.5.6.3.

Radiation in effluent discharge paths and the plant environs are moni tored as described in Sections 7.5.6.2 and 7.5.6.3.