Design Certification Application Orientation Detail of ... · This chapter includes the following...

US-APWR

Design Certification Application Orientation

Detail of FSAR

Tier2: Chapter 6

January 15,16, 2008Mitsubishi Heavy Industries, Ltd.

fUSI"RIES, LTD. UAP-HF-08010

Presenter

Andrew B. JohnsonPrincipal Engineer

Mitsubishi Nuclear Energy Systems, Inc.

IIAI•I U• rlO/t4h 4-S, LTD. UMr-nr-uou EU-I

Contents

1. Overview of Chapter

,/ Title of Chapter

/ Scope of Chapter

2. Design Features

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>Title of ChapterChapter 6:

ENGINEERED SAFETY FEATURES (ESFs)

;Scope of ChapterThis chapter includes the following ESFs:* 6.1 Engineered Safety Features Material* 6.2 Containment Systems0 6.3 Emergency Core Cooling Systems (ECCS)• 6.4 Habitability Systems• 6.5 Fission Product Removal and Control Systemsa 6.6 Inservice Inspection of Class 2 and 3 Components

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2. Design Features

> 6.2 Containment SystemsContainment systems consists of followings:

* Containment Structure (PCCV)* Containment Spray System* Containment Isolation System- Containment hydrogen monitoring and

control system

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2. Design Features

> Containment Systems (Cont'd)/ Containment Function

The containment is designed as an essentiallyleak-tight barrier that will safely and reliablyaccommodate calculated temperature andpressure conditions resulting from loss-of-coolant accident, or main steam line break.

Major Design Parameter

Type Prestressed Concrete Containment Vessel(PCCV) with Carbon Steel Liner

Design Pressure 68 psig

Design Temperature 300 deg. F

Design Leakage Rate 0. 1% air mass Iday

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2. Design Features

Containment Systems (Cont'd)" Containment Function (Cont'd)

Example analysis result (large break LOCA)

5 Contabrnent Presumr

PRI

59 CarfallnmeM Vapor Tmperatum

TV1

~I

,3W0d. F

I

Time (swc)

Pressure Transient

Tkne (see)Tmpr71tu2.reT*ransi7en

Temperature Transient

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2. Design Features

Containment Systems (Cont 'd)V Containment Heat Removal

INSIDE THE CONTAINMENT I OUTSIDE THE CONTAINMENTSpray Ring Header

i -9

F-

I -....

C Ii

9CIEHRH "H.1

•4 Independent trains-Automatic initiation byContainment Spray Signal

-Pumps and heatexchangers used for RHRfunctions during shutdown

-Common Spray RingHeader

,-~*1--- !

Hi.-

! o-. 61 -.. # - --------- '---0 . i --i

---

leo CSIHRH.

il

tT\

CSIRHRHx CS/AHRP±

Take suction from in-containment RWSPNote: Red portions are common part for CSS and RHRS

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2. Design Features S.

6.3 Emergency Core Cooling Systems (ECCS)

.- 4 Independent trains

-- ---- '-- Automatic initiation bySafety Injection Signal

,R P Rv A . Emergency Letdown Linefor Safe Shutdown

AdvancedAccumulator

Direct VesselInjection (DVI)

EmergencyLetdown Line

n • • Take suction from in-containment RWSP

-= Safety Injection1 o ..... Pump

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2. Design Features

Emergency Core Cooling Systems (EC(v' Advanced Accumulator

* Automatic switching of injection flow rate b* Integrates function of low head injection sy* Long accumulator injection time allows Ion

injection pump to start

CS) (Cont'd)

y flow damperstemger time for safety

Flow DamperAdvanced Accumulator4 IlArll UJ• h~flI~Aj AIi im

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2. Design Features

;6.4 Habitability SystemsThe habitability systems allow operators to remain safely insidethe control room envelope (CRE), that includes the main controlroom (MCR), and take the actions necessary to manage andcontrol the plant under abnormal plant conditions, including aLOCA.

VMCR HVAC Systema 2 x 100% MCR emergency filtration unitsa 4 x 50% MCR air handling units* Air tight isolation dampers* Two emergency modes

Pressurization mode ; during an accident withradiological releases.Isolation mode ; during a toxic gas event

* Automatic initiation by the MCR isolation signal

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2. Design Features

Habitability System (Cont'd)

trol Room Envelope

UAP-HF-08010-12

Air Tight

, . _ _ _ " l IL II .

2. Design Features

6.5 Fission Product Removal and Control Systems

,,The fission product removal systems remove fission products thatare released from the reactor core as a result of postulatedaccidents.

,/The containment controls the leakage of fission products to ensurethat the leakage rate from the containment is below limits.

,/The US-APWR fission product removal and control systems areas follows:

" Containment spray system

" Containment

" Annulus emergency exhaust system

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2. Design Features !NAPSW

Fission Product Removal and Control System (Cont'd)Fission product removal effects differ with the chemical forms of the radioactive iodine. The assumedchemical forms are noble gas, elemental iodine, organic iodine, and particulate (aerosol). The fissionproduct removal effects in the US-APWR containment under accident conditions are the following:

Mechanism Noble Gas Elemental Iodine Organic Iodine Particulate (Aerosol)

Containment Spray Not Applicable Slight effect, No credit Not Applicable Applicableapplied (Not t) (Based on SRP 6.5.2)

Natural Deposition Not Applicable Applicable (Note 2) Not Applicable Applicable(Based on SRP 6.5.2) (Powers natural

deposition model(NUREG/CR-6189):10t percentile)

Radioactive Decay Applicable Applicable Applicable Applicable

Containment Leakage Applicable Applicable Applicable Applicable(Note 3) (Based on Technical (Based on Technical (Based on Technical (Based on Technical

Specifications) Specifications) Specifications) Specifications)

Annulus Emergency Not Applicable Not Applicable Not Applicable ApplicableExhaust System III_ I_(HEPA filter)

Notes:1. The CSS with NaTB baskets is expected to achieve a pH of at least 7 in the RWSP. Thus, the CSS can remove elemental iodine slightly.

Therefore, we assume that the CSS does not remove elemental iodine.2. The CSS removal effects contain the removal effect by natural deposition. Because the removal effects for elemental iodine by the CSS is

not credited, the removal effects for elemental iodine by natural deposition can be credited in not only the sprayed region, but also the

unsprayed region.3. Containment Leakage to the penetration areas is treated by the annuls emergency exhaust system

13 8" j *6UiEN6W=WAv& •S"R I E S, LTD. UAP-HF-08010-14

US-APWR


Detail of FSAR

Tier2: Chapter 7, 8 and 18


•--.IVrM_- IUB WEVYN4INDUSIR1ES, LTD. UAP-HF-08011

Contents

> Chapter 7: Instrumentation and Controls (I&C)

> Chapter 8: Electric Power

> Chapter 18: Human Factor Engineering (HFE)

/ For each Chapter

1. Content Overview2. System Descriptions3. Analysis and Evaluations

> Summary

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Presenter

Ken ScarolaEngineering Manager


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Chapter 7- Content Overview ( /_US_APW,fý)_ 11

> Chapter 7 includes the following descriptions:,/ All safety related I&C systems,/ Non-safety I&C systems which are important in

maintaining safe normal operating conditions and whichsupport abnormal plant conditions

,/ Intra and inter system data, communications

> Descriptions focus on features related to:/ Performance/ Reliability/ Maintainability/ Failure modes

> Format based on RG 1.206

> Content based on RG 1.206 and SRP

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I&C System Overview (7.1) ýW

Common digital microprocessor based platform forsafety and non-safety I&C (no electro-mechanical relays)

Diverse Actuation System based on analog technology

Complete four train redundancy for safety i&C with eachdivision in separate fire area

Distributed architecture for non-safety I&C withredundancy

Fully multiplexed and duplicated signal transmissionnetworks from local areas to I&C equipment rooms andbetween I&C systems/components

Fully computerized Main Control Room and RemoteShutdown Room with no reliance on local controls

_I •_$JIES, LTD. UAP-HF-08011-4

i&C System Overview (7.1)

The digital I&C and HSI systems for the US-APWR areessentially the same as the I&C and HSI systems fornuclear plants in Japan

V First installation for non-safety digital I&C in 1987

/ Average 10 years operation for five operating plantsV Applied to all non-safety I&C, 50 applications per plant

/ Over 20 million hours.total operating experience

v/ No un-expected shut down caused by I&C since 1992

/ No system malfunction caused by S/W or H/W failure

/ The same digital platform is currently being applied to safetyand HSI systems of the Japanese APWR, and safety and HSIsystems currently being implemented for plant modernization.

• First safety and HSI application: Tomari #3, C/O 2009

jES, LTD. UAP-HF-08011-5JHI~MEA~L~WRJISIRIES, LTD. UAP-HF-0801 1-5

Main Control Room (7.1)

OperatorConsole

Safety VDU1

UAP-HF-08011-6Alarm VDU Operation VDU (Non-Safety)

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Reactor Protection System (7.2) VSG Water Leanl Low Pressurizer Pressure High

Division A-- -- .. .. .. .. .--- .. .. .. .. .. . .. .. .. .. .. . .. .. .. . . .......... -- ...... ......

Functional allocation

To otherRPS Trains

..... .. .......RPS Trains

..... . 2 . . . . . . .

La FGr. Ir

----------------------

Divisions B, C, D From other RPS Trains

are identical L-------------l I orou1 , n2 01, 0 D,2

sco RT5-B3 RS-2o

-- --- 0 0- 0

Two processors ineach divisionprovide twofunctionallydiverse trips foreach postulatedaccident.

Note

- 0MaUnII

.---- Hard wired

N E.101 Electrical to optical converter or

Optical to electrical converter

:illS BUS interface

SIsolation point

Configuration of Reactor Protection SystemZ.fMaeWUE KE . E3E E .. -,*-sueUI I• V--HpJ IXI I'1 rllI

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ESFAS, Safety Logic System, Safety HSI /-US(7.- 7.6)

SYM~ncL~nI yOULS and HSIsupport ittterlocks, acmcident

;sors inR..tn Roca R- 11c R--tc V-t IS

livision By A .. 6'-C 6 0

ESF~n Acuto Stotem S- cl Stote VDUt~~lenc IS I- I.'R~dunddnt

Arcted bu-y(CPUAll (PA l(OM! 2(0 M

enance Do) ------- 0 -----

------------ tg . . . . .... . ....... . Hrta

solid. stat outputss

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Reactor Control System (7.7) __

•,• CFunctional grouping ensures all failures.. •to , ,ro'c sare bounded by safety analysis

Rect arteton u fmoate vl e

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, o..w Word T _ p[ .. M Wpllc- ,. . • ,,,, ,D•. ., •r .... ] r

* -Gfe~ePreaosaer p~eturt ! Pwesurir a.!. ! Contro o neon I I Ietc-°oi:.I1- ~ i JL S~S -~m ntte li Gr. ' e, Gr3 i imi i, GrSG

-U-- - ~ -n -' - - - - - - ---"- - - - i' • r et r i r'- - - - i " '*

Redndacywithin eachLi,....... ....... o........... t :_it group ensures high -- .C-,"N PID IT Ccreliability --A0 "r

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Configuration of Reactor Control System

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Diverse Actuation System (7.8)

Permissive Switch for

Pw, DAS HS--k ..........i yi

, ..... F, 7- -Mo. -. . " Train

i ,log, ,1 ; I ,

. ..- -• v..... .. .. .. . Mo.-l .

P o w r S w t __n- -------- -- - - - -

1 o System LevelI /Manual Switch

Two DASdivision preventsspuriousactuation.

DAS is diverseand isolatedfrom PSMS.

O~.'l'l O.ona tOt

NOTE - The Power Breaker for DHP is located in

the MCR. It is separated from the DHP to prevent

fire propagation.

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Data Communication (7.9) /-US- -1kV1MW

A- - . PSMS I- - - - -, - w- - - . ... ..

I. -- - - -S.aporw ShlltTah l Deu P .oaa . Div.- System~, ....... ,-. ag. AdaW. C.•.... .i.. ...ed tHt Un. .

-" ~e, " -• -it-WO . D" Unk

130tor IM M .timeS~lNa R.a

S-- - I-" -"ne p -- o-" Pd Y .....

ESF A~tuaflblwFk. 'asseII

.l•II .. .: a...... "

.. ..t i S. t ed Safet MHSo

, .•._._.,:•. AM.,_ 802... , ."r..•DAS: iveseAcuatonSytem PSS Potc•on ndSaetyMoitoin Sste HIS: Hma Sste Iterac Sste PMS: Pan Cotrl nd ontoingSyte

II

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Chapter 7 - Analysis & Evaluations

> Chapter 7 descriptions include following design details:,/ Redundancy, separation and isolationV Data communication independence and performanceV Maintenance and operating bypassesV Test coverage (self-test and manual)V Access controls and cyber security/ Failure modes and effectsV Coping with common cause failures/ Hardware and software quality (Software life cycle),/ Hardware qualification and reliability

> Chapter 7 describes the following design processes:v/ Setpoint determination$ Software life cycle (basic and application)

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MELTAC Platform

Mitsubishi Electric Total Advanced Controller

Simple DesignV Modular and Structured Architecture

/ Single Task execution

V Cyclical Processing with No Interrupts

> Quality Assurance and ControlV Designed specifically for Nuclear Applicationsv/ Under'control of Nuclear QA/QCV Fully owned and life cycle managed by Mitsubishi

.MI.URI.HI=HE!DPU!$SJTIES, LTD. UAP-HF-08011-13

Chapter 8 - Content Overview qVS5-,._,&P0

Chapter 8 includes the following descriptions:V On-site safety related AC and DC power systemsV On-site non-safety AC and DC power systemsv/ Interface to off-site power distribution system

) Descriptions focus on features related to:V Reliability/ Maintainability/ Failure modes

Format based on RG 1.206

Content based on RG 1.206 and SRP

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Power System Overview (8.1)

Transmission Systeml

UAT1,2: Unit Auxiliary Transformer 65MVAUAT3,4 : Unit Auxiliary Transformer 53MVA

RATI 2 Reserve Auxiliary Transformer 65MVARAT3,4 : Reserve Auxiliary Transformer 53MVA

13.ikVNI t3.ikVN2Non- Safety Non -Safety

Bus Bus

Class I E (Safety Related) Buses

Non Class I E (Non Safety ATgkVBrB.BkVP1 qIVP2 B1..kVC 6 BV JISafety Saet SonnaetRlted) Buses Bus Bus Say Bs Bs Bus

A-EPS B-EPS A-AAC B-AAC C-EPS D-EPS

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Offsite Power System (8.2)

> Design Features,The two (2) sources of offsite power provided.

a) Main Transformer through Unit AuxiliaryTransformers (UAT)

b) Reserve Auxiliary Transformer (RAT)

-/The two (2) offsite power supply circuits areindependent and physically separated.

,/Both offsite power supply circuits have enoughcapacity to achieve their safety related functionduring a Design Basis Event (DBE) and meet therequirement of the applicable GDC's.

!rT BI• WPWi•gM I)UVJ_.R|ES, LTD. UAP-HF-08011-16

Onsite AC Power System (8.3)

Design Features

V/Class 1 E AC electrical power system consists of four(4) separate trains. Each train includes one Class 1 EEmergency Power Source (EPS)

,/On-Line Maintenance of any EPS is allowed withSingle-Failure Criterion remaining satisfied

V"Permanent" buses supplied from Alternate ACPower Source (AAC) are provided

V/Non-safety related loads are not supplied from class1E buses. Required non-safety related loads aresupplied from AAC in LOOP condition

,/AACs provide power to all electrical loads that arerequired to bring and maintain the unit in safe-shutdown mode upon the SBO

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Gas Turbine Generator (8.3)

Gas turbineI I

Power seotlon Gear boxCoupling

MHI selected

Gas Turbine Generators

for EPS and ACC

-bine package with exhaust silencer

Lias iuminepi-aosi erclý

OUTPUTf SHAT TURBINE

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Why Gas Turbine Generators

,/GT/G has been selected based on reliability and maintainabilityimprovements when compared to DG

Gas Turbine Generator Diesel Generator

Space Compact Large

Cooling Water Not Required Required

Routine Maintainability 1/3 the parts of a DG Complex

Once or twice during plant Periodic OverhaulLarge Scale OverhaullieRqrd life Required

Reliability (failureldemand) 104 based on Japanese 10-3experience

Starting Time 40 sec 10 sec

V'L

KuTmm!onger start time of GT/G is accommodated by the AdvancedAccumulator design of US-APWR which allows 100 sec

Au UGAVkY-NDUTRIES, LTD. UAP -HF-08011-19

Station Blackout (8.4)

Basic Concept for Coping with SBO/The AACs are available in the event of SBO, when all offsite

power sources and EPSs are not available to bring the unitto a safe shutdown condition and maintain that status

Design Basis

,/AACs of a different type (Starting System, Capacity etc.)and are provided to minimize the potential for commonmode failure with either the offsite power or the EPS system

v/ The AAC is a non-class 1 E gas turbine-generator packageconnected to a 6.9kV AC "Permanent" bus

, The AAC supplies power to loads on any class 1 E busthrough tie line circuits during SBO

V The AAC supplies power to loads for 8 hours during SBO

•JMEU"_SJItSHRDuIST.?IES, LTD. UAP-HF-08011-20

Chapter 8 - Analysis & Evaluations kAPW

Chapter 8 descriptions include the followingdesign details:

,/ Redundancy, separation and isolationv/ Failure modes and effectsV Hardware quality, Hardware qualification and reliability

>Chapter 8 describes the following designprocesses:

Class I E qualification and tests of Gas TurbineGenerator

IIAP-HF-NRNI t-•t~UT~l~I IAI~II~ CI~ I UE AP-HF-080f1 1-21

Chapter 18 - Content Overview

Chapter 18 includes the following descriptions:,/Human Factors Engineering processv/ Human Systems Interface design features

Descriptions focus on features and processesintended to:

/ Enhance human performance/ Reduce potential for errors in critical human actions

> Format based on RG 1.206> Content based on RG 1.206 and SRP

MI MriU I jI- ,V- D_).S-T.RIES, LTD. UAP-HF-08011-22

HFE Design Process Overview (18.1)

The US-APWR HSI design is based on the HSI forJapanese plants, which has been developed in phasesover the past twenty years

The Japanese HSI was developed following theNUREG 0711 HFE process

V This included dynamic validation by more than 46 Japaneseoperating crews (138 operators)

/ V&V included operability by one RO I one SRO

The US-APWR HFE program reassesses each HFEprogram element, with emphasis on changes fromprior experience

/ DCD describes applicability of prior HFE and new activitiesspecific to US-APWR, including additional dynamicvalidation by US operators using a full scope simulator.

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Operating Experience Review (18.2)

LERs and SERs Corrective action

from operating Fsystems, Maintenance

fromn Japan HFE/HS design of Logs and OperatingPWRs i standard Japanese PWR Logs from operating[2-1oop/3-1oop conventional PWRs in Japan

PWR, 4-loop APWR]

IUS-APWR HSI Design

OER process wasused for developmentof Japanese HSI.OER will be expandedfor US-APWR.

US LERs(from NUREG/CR-6400

US LERs and SERs(post NUREG/CR-6400)Corrective action

systems, MaintenanceLogs and OperatingLogs from US PWRs

.li

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Function, Task, Reliability Analysis(18.3, 4, 6)

Operating Experience ReviewI

Functional RequireFunction

-Task definition-Function allocation

ments Analysis andAllocation

j(Human - Computer)

PRA

HRA and PRA areintegrated to ensurehuman actions areaccurately modeled in thePRA, and to ensure risksignificant human actionsare given increasedattention during the HFEdesign process.

& Qualificationorganization

Human System ProcedureInterface Design Deveill pment

:Information allocation

Display & control-Protatyping

I I

i

Human FactorsVerification and Validation

* Validation test- Static test using mockup- Dynamic test using full-scope simulator

Design Implementation

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Operator Staffing (18.5)

SS*-- .. - ---T A--•,SS*& STA

*SRO

, Not Locatein MCR

RO

K

> Necessary number of Reactor Operators (RO) is reducedfrom 2 to 1 by reduction of Workload for Operation

> Minimum staff complies with 10 CFR 50.54(m)L i_ i _iis u -- Uf.i'!"' rRIES, LTD. UAP-HF-08011-26

HSI Design (18.7) QU.S5-'-1k_0

Large Display Panel HSI desian includes

i a MininInventcPositioialarms

PJi F t.yP

num•ry of Fixedn indications,and controls.

OperationalDisplay

OperatingProcedureDisplay

UAP-HF-08011-274 Alarm Display. . ..E~kflELI~E...DUSTItIES, LTD.

M1ý n JG-j, I EHý] 1

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HSI Design (18.7)

> Safety and Non-safety components can be operated from the same screen

Designconsistencybetween non-safety and safetyVDUs facilitatesoperatortransitionbetween HSIfeatures.

Mz__1_u_! ,- Y-MDUSTRIES, LTD. UAP-HF-08011-28

Procedures and Training (18.8, 9) _ Ps-

Training and procedures encompass the full range ofpersonnel, functions and systems which may affectplant safety.

; Procedures and training material are developed basedon documented Writer's Guides

> Computer based procedures include hot links whichdisplay plant information and controls on adjacentscreens

> Design consistency between computer and paperprocedures facilitates operator transition for degradedHSI conditions.

> Procedures are validated through dynamic simulation

S_ _ u '._A•• TIISTZES' LTD. UAP-HF-08011-29

HSI Validation (18.10)HSI Simulation Facility - Pittsburgh (April 2008)

-Full Scale MCR

-Interactive, full functionalityVDUs

'*High fidelity dynamic plantmodel

17ft (5m) 14ft (4m)

Used initially to validate Japanese HSI design by US operators (12/2008)Used later to validate HSI feature changes for US-APWR (6/2009)Used to validate final US-APWR HSI, including all displays, alarms,controls and procedures (ITAAC Closure)_IM_; N__•;__ ._A_ -- IIES, LTD. UAP-HF-0801 1-30

Implementation and Performance Monitoring ,-i.(18.11, 12)

Recurring implementation and changes to the HSIafter validation are in accordance with the DesignImplementation process

,, The design change process is based on a risk assessmentincluding the risk significance of effected human actions

Human performance is monitored on an ongoingbasis to ensure:

> The HSI does not create human performance problems> Actual human performance is consistent with plant analysis

assumptions regarding credited manual actions

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Summary Q f

> MHI I&C, HSI and Electrical Systems provide significantadvancements toV improve plant safety and availability/ reduce operations and maintenance costs

> The systems employ proven designs with many years ofdemonstrated reliability

> MHI suggests frequent technical meetings to minimizemisunderstandings and thereby facilitate an efficientregulatory review process

> MHI invites the NRC staff to visit the following facilities

/ MELCO digital I&C factory (Kobe)/Gas-turbine generator qualification test facility (North

Carolina)/HSI simulation facility (Pittsburgh)

LMM k MUMMOM I W INK EE AV-Y 4 N D U S T R I E S, L T D. UAP-H F-08011-32UAP-HF-08011-32~uMIXSUBISHLHEAV.Y-INDUSTRIES. LTD. UAP-HF-ARAI 1-~2

US-APWR


Detail of FSAR

Tier2: Chapter 9

January 15,16, 2008

Mitsubishi Heavy Industries, Ltd.

- iETIIELI U.IEAf L fIJmDUSTRIDEES, LTD. UAP-HF-08012

Presenter (AIP

Andrew B. JohnsonPrincipal Engineer


m_ 3!iltEIJ0 U L_•u, A u T RIES, LTD. UAP-HF-08012-1

Contents


/ Title of Chapter

v/ Scope of Chapter

2. Contents of Subsections

3. Design features (for example)

F -HI HEAV* 'NDUSTRIES, LTD. UAP-HF-08012-2


)Title of Chapter

Chapter 9: AUXILIARY SYSTEMS

ýScope of ChapterThis Chapter includes the following Sections andAttachment:

- 9.1 : Fuel Storage and Handling Systems

- 9.2 : Water Systems

- 9.3 : Process Auxiliaries

- 9.4 : Air Conditioning, Heating, Cooling, and

Ventilation Systems

- 9.5 : Other Auxiliary Systems- Attachment 9A : Fire Hazard Analysis

I IHADI_-IRni.2ma~u~. ~ LB~ A~U~ .afu~u ~ U 16 0a LinU. ~.- .- u


>Section 9.1 : Fuel Handling and Storage Systems

Regulatory Guide 1.206 US-APWR DCD9.1.1 Criticality Safety of Fresh and 9.1.1 Criticality Safety of New and

Spent Fuel Storage Spent Fuel Storage9.1.2 New and Spent Fuel Storage 9.1.2 New and Spent Fuel Storage

9.1.3 Spent Fuel Pool Purification and 9.1.3 Spent Fuel Pit Purification andCooling System Cooling System

9.1.4 Light Load Handling System 9.1.4 Light Load Handling System(Related to Refueling) (Related to Refueling)

9.1.6 Overhead Heavy Load Handling 9.1.5 Overhead Heavy Load HandlingSystem System

9.1.6 COL Information

9.1.7 References

S- " TIES, LTD. UAP-HF-08012-4

2. Contents of Subsectionsevink

>Section 9.2: Water SystemsRegulatory Guide 1.206 US-APWR DCD

9.2.1 Station Service Water System 9.2.1 Essential Service Water System9.2.2 Cooling System for Reactor 9.2.2 Component Cooling Water System

Auxiliary9.2.3 [ Reserved ] 9.2.3 [ Reserved 19.2.4 Potable & Sanitary Water Systems 9.2.4 Potable & Sanitary Water Systems

9.2.5 Ultimate Heat Sink 9.2.5 Ultimate Heat Sink

9.2.6 Condensate Storage Facilities 9.2.6 Condensate Storage Facilities9.2.7 Chilled Water Systems9.2.8 Turbine Component Cooling Water

SYstem9.2.9 Non-Essential Service Water

9.2.10 COL informaton

9.2.11 References

16 I I i• I IP AAAJA P_m_

-mE'- .w.U* Z UwmEur ErMUN3ES, LTD. U~r-Mlr-5UIA1-0


>Section 9.3: Process AuxiliariesRegulatory Guide 1.206 US-APWR DCD

9.3.1 Compressed Air Systems 9.3.1 Compressed Air and Ga Systems9.3.2 Process and Postaccident 9.3.2 Process and Postaccident

Sampling Systems Sampling Systems9.3.3 Equipment and Floor Drainage 9.3.3 Equipment and Floor Drainage

System Systems9.3.4 Chemical and Volume Control 9.3A Chemical and Volume Control

System (PWR Only) System9.3.5 Standby Liquid Control System 9.3.5 Not ADplicable for US-APWR

(BWR Only)- 9.3.6 COL Information

-__ _ _ 9.3.7 References

m Im

. -- MU~XKfw LTD. U~r-Mr-U3ui-IZ-


>Section 9.4: Air Conditioning, Heating, Cooling, andVentilation Systems

Regulatory Guide 1.206 US-APWR DCD9.4.1 Control Room Area Ventilation 9.4.1 Main Control Room Heating.

System Ventilation & Conditionina System9.4.2 Spent Fuel Pool Area Ventilation 9.4.2 Spent Fuel Pool Area Ventilation

System System9.4.3 Auxiliary & Radwaste Area 9.4.3 Auxiliary Building Ventilation

Ventilation System System9.4.4 Turbine Building Area Ventilation 9.4.4 Turbine Building Area Ventilation

System System9.4.5 Engineered Safety Feature 9.4.5 Engineered Safety Feature

Ventilation System Ventilation System9.4.6 Containment Ventilation System

-9.4.7 COL information

9.4.8 References

I IilUi I III AAAAA.Ih

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>Section 9.5: Other Auxiliary Systems

Regulatory Guide 1.206 US-APWR DCD9.5.1 Fire Protection Program 9.6.1 Fire Protection Program9.5.2 Communication System 9.5.2 Communication System9.5.3 Lighting System 9.5.3 Lighting System9.5.4 Diesel Generator (DG) Fuel Oil 9.5.4 Gas Turbine Generator Fuel Oil

Storage & transfer System Storage & transfer System9.5.5 DG Cooling Water System 9.5.5 Not Applicable for US-APWR9.5.6 DG Starting Air System 9.5.6 Gas Turbine Generator Starting Air

System9.5.7 DG Lubrication System 9.5.7 Gas Turbine Lubrication System9.5.8 DG Combustion Air Intake & 9.5.8 GTG Combustion Air Intake and

Exhaust System Exhaust System9.5.9 COL information

9.4.10 References

L -,.,.,u, u.-v "UmTIES, LTD. UAP-HF-08012-8

3. Design Features (for example)

> Support Systems for Safe Shutdown

V' Component Cooling Water Systems (CCWS)

V Essential Service Water Systems (ESWS)

/ HVAC systems, etc.

> Design Features of CCWS & ESWS

/ CCWS and ESWS constitute a safety cooling chain

V 4 Train configuration

/ Allows On Line Maintenance assuming single failure

U-,,,,•,u-W ,-MOHSj5 ES, LTD. UAP-HF-08012-9

3. Design Features (for example)

);oComponent Cooling Water Systemv' 4 safety train configuration

(Each train includes 1 CCWP and 1 CCW HX)v/ Separated into 2 independent sections

(Each section has 1 CCW surge tank),/The other safety components (e.g.;SFP HX) are supplied

with cooling water from 2 of 4 safety trainsCCW SURGE TANK CCW SURGE TANK

CC UPCON PUMP CCW PUMP CCW PUMP

CC XCCVW HX CCW HX CCW liX

'4traiin Saen 4trin Safety' ' 4 train Sft rt Safety'j

:Componet, .c _.nent: :Components: ....

NSafety S Non-S afety

:Components: = a. - nnSaft Componentst

MES, LTD. .UAP-HF-08012-10

3. Design Features (for example) QA ýPSW-

•Essential Service Water Systemv/ Completely independent 4 train configuration

( Each train includes 1 ESWP )/ Raw water cooling for the CCW HX and Essential

Chiller UnitCCW Hx

ESWP LItII VH

CCWHX

ESWP "1 r urn, THS

Essential

Chiller UnitI ~ ~ CIllelr~~l UniUt

UHSESWP L

IChiller UnJtI

lIAn UID tl0fl4' .44Uli•i n•o•iiR i BP illtl

-aEI ý Apirmi--*-----, LTD. uAr-rur--uou 1A- IEI

US-APWR


Detail of FSAR

Tier2: Chapter 10


"MLTSUKI•ISHIAV--IlnDU TRIES. LTD. UAP-HF-08013

Presenter/Section Leader

Yoshihiro Minami

Engineering Manager

Nuclear Turbine Plant Engineering Section

Water Reactor Engineering Department

Nuclear Energy Systems Headquarters

Mitsubishi Heavy Industries, LTD.

- EMU-BMSA• •U1SXR-IES, LTD. H APi-WPAnfl4*A-

Contents


Title of Chapter

Scope of Chapter

Overall System Flow Diagram

2. Design Features

Significant Design Features

System Design Features

ITJI EL-YQ U E T R ES, LTD. UAP-HF-08013-2


>Title of Chapter 10/ STEAM AND POWER CONVERSION SYSTEM

> Scope of Chapter

, This chapter includes the design description of thesystems and the components for power conversion

V" This chapter consists of 4 sections:

• Section 10.1 • Summary Description

* Section 10.2 Turbine-Generator

* Section 10.3: Main Steam Supply System

* Section 10.4 : Other Features of Steam and PowerConversion System

, This chapter deal with 13 systems in total

I I A • Ul• •0•4 •_* aE4QUSX1S LTD. LMr-r-lr-UoU I)

1. Overview of Chapter QUA -n! kJ#

10.1 Overall System Flow DiagramCV:CN'<-r->RB RB<--Ir->rB CV:

RB*iT rRB:10.3 Main Steam Supply System: TB:

(MSS) 1 * AB:

Containment VesselReactor BuildingTurbine BuildingAuxiliary Building

410 2Turbine-Generator (T/G)

10.4.8 Steam Generator

Blowdown System(SGBDS)

RB*3 AB AB ~RB3 RB<-Lý)

~W , boWE

10.4.2 Main Condenser

D 10.4.1 Main Condensers10.4 Gland Evacuation SystemTurbine (CS

Bypa toss

(TBS)

D I 0~~10.4.5'"culat-ing Water Sse CS

10.4,3 Gland Seal System (GSS)

10.4.6 Condensate polishingSystem (CPS)

10.4.7 Condensate and Feedwater System (CFS)

I IAEI II l I'•Oll'•Aw AIgR

2. Design Features /-US-V.-f W-

10.1 Significant Design Features

Rated NSSS power (MWt) 4,466

Steam Generator Outlet Press. (psig) 957

Quantity of Steam Generator (SG) 4

Total steam flow rate from SG (Ib/hr) 20,200,000

Steam Turbine Rating

Type of Steam Turbine (-) TC6F

Rotating Speed (rpm) 1,800

Generator Output (MWe) 1,700

Exhaust Pressure (inHga) 1.5

Generator Rating

Capacity (MVA) 1,900

Power Factor (-) 0.9

14 IIAI• UI• P•Ot•4")M14411HII LILP. Up%1EflI, -ow I .2-

2. Design Features

10.2 Turbine-Generator (T/G)

Low Pressure Turbines (LPTs)

High Pressure Turbine (HPT)

Generator

Moisture Separator/Reheater (MS/R)

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2. Design Features S.

10.2 Turbine-Generator (TIG)

,/ The T/G is non safety-related system

V The T/G could be a potential source of a high-energyturbine missile, which could cause damage to safety-related equipment or systems

/ Turbine and control/protection system are to be designedso that probability of turbine missile generation probabilitysatisfies the requirement of SRP (less than 1 x 10-5 peryear assuming proper inspection and test frequency)

V The orientation of the T/G is such that a high-energymissile to be directed at an approximately 90 degreeangle away from the safety-related structures

4 Iimll I•IRIP• I g•M. -- M---w;wwwww"ww IwEh. LTD. u/Ar-nr-uou Ia-1

2. Design Features

> 10.3 Main Steam SupplySystem (MSS)

V The MSS is to transport steamfrom the SGs to the HPT and tothe MS/R

V The MSS is provided with safety-related main steam isolationvalves (MSIVs) and main steambypass isolation valves (MSBIVs)in each main steam line for thepurpose of:

* Isolating the secondary side ofthe SGs to prevent theuncontrolled blowdown ofmore than one SG

* Isolating non safety-relatedportions of the system

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2. Design Features /_VS_ -"_jt!61-4 ýW

10.4.1 Main Condenser

V/ The main condenser is non. safety-related system

V The main condenser functions to condensate anddeaerate the exhaust steam from the main turbine andprovide a heat sink for the turbine bypass system

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STAR AURTAIS~PRAT P~PS

FRF..S.________ l)MT-,IH T•Lt ANIMCNDNE!Lee SHEET

M M MITSUBISHI TITANIUM COURSER

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2. Design Features

10.4.2 Main Condenser Evacuation System(MCES)

The iCES is non safety-related system.

/ The iCES removes noncondensable gases from themain condenser during plant startup and normaloperation

V The iCES establishes and maintains a vacuum in the.main condenser

L.wwm_1, =-Y=IW .D_._j WE S, LTD. UAP-HF-08013-10

2. Design Features

10.4.3 Gland Seal system (GSS)

V The GSS is non safety-related system.

/The GSS prevents air leakage into and steam leakageout of the casing of the steam turbine

/ Sealing steam is supplied to the turbine shaft fromeither the Auxiliary Steam Supply System (ASSS) orthe MSS

/The system returns the steam-air mixture from theturbine glands to the gland steam condenser andexhausts non-condensable gases into the atmosphere

"IXtSUBISHEIHE&V-YX4~LSISES, LTD. APH-831IUAP-HF-08013-11UAP-HF-08013-11

2. Design Features

10.4.4 Turbine Bypass system (TBS)

$ The TBS is non safety-related system.

v/ The TBS is part of the MSS and provides capability tosend the main steam flow from the SGs to the maincondenser bypassing the main turbine

V The TBS is designed to sustain a 100% load rejectionwithout reactor trip, and not requiring actuation of themain steam relief valves, main steam safety valves andpressurizer safety valves

------ D T ES, LTD. UAP-HF-08013-12

2. Design Features

10.4.5 Circulating Water System (CWS)

/ The CWS is non safety-related system.

/ The CWS supplies cooling water to remove the heatfrom the main condenser under various plant operatingconditions and site environmental conditions

V The CWS removes the plant heat during startup,normal operation, shutdown, transient condition, orturbine trip

LMI_.7rSMJSVJJHEAVY-MDUSTRIES, LTD. UAP-HF-08013-13IMPJ~1hKE~Pr4NDU~XR!ES, LTD. UAP-HF-0801 3-13

2. Design Features QA PSý4f

> 10.4.6 Condensate Polishing System(CPS)

/ The CPS is non safety-related system.

,/The CPS is designed to remove dissolved ionic solidsand impurities from the condensate and assists in theremoval of corrosion products

10 __ IIAn Lr" flOn4o ,4AIK-B- U 0,LTD. M ~-r r-uI ,J- a.

2. Design Features

10.4.7 Condensate andFeedwater System (CFS)

,/The CFS provides feedwater at therequired temperature, pressure,and flow rate to the SGs

/ The safety-related function of theCFS is to provide containment andfeedwater isolation following adesign basis accident

/ The system provides mainfeedwater isolation valves (MFIVs)in the main feedwater lines

/The MFIVs close to limit the massand energy release to thecontainment

•MLMISU SHI1IH.AV_-Y-INDUSIRIES. LTD.

AP

RE-L -1.LOI

UAP-HF-08013-15

2. Design Features

10.4.8 Steam Generator BlowdownSystem (SGBDS) I -=

/ The SGBDS assists in maintainingsecondary side water chemistrywithin acceptable limits during normaloperation and during anticipatedoperational occurrences due to main F ' ' ''icondenser tube leakage or primary to L .-secondary steam generator tube i n _leakage [,: :•II • "

V The SGBDS has a safety-relatedfunction to isolate the secondary sideof the SGs using two isolation valvesin series in the blowdown line fromeachSG

v' This provides a heat sink for a safeshutdown or to mitigate theconsequences of a design basis ,---

accident

ETC I E fIZELEJA•, -J~UlI ES, LTD. UAP-HF-08013-16

2. Design Features

> 10.4.9 Emergency Feedwater System(EFWS)

," The EFWS is a safety-related system

/ The EFWS is designed to supplyfeedwater to the SGs and removereactor core decay heat followingtransient conditions or postulatedaccidents such as:

" Reactor trip" Loss of offsite power (LOOP)• Loss of main feedwater" Feedwater line break (FLB)* Main steam line break (MSLB)

" The EFWS consists of two motor-drivenemergency feedwater (EFW) pumps,two turbine-driven EFW pumps,emergency feedwater pits and other.necessary equipment

1_ýWv J,

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2. Design Features

> 10.4.10 Secondary Side Chemical InjectionSystem (SCIS)

, The SCIS is non safety-related system./The SCIS is designed to maintain a noncorrosive

condition within the secondary loop,/ Noncorrosive condition is maintained by controlling pH

and dissolved oxygen content in the secondary sideby:* Maintaining alkaline pH by ammonia injection" Scavenging dissolved oxygen with hydrazine

injection

_MIE.SýUBI$H I-HE-•/-Y-INDRU STRIES, LTD. UAP-HF-08013-18

2. Design Features •A4

10.4.11 Auxiliary Steam Supply System(ASSS)

V" The ASSS is non safety-related system.

v/ The ASSS is designed to provide the steam requiredfor plant use during plant startup, shutdown, andnormal operation

V Steam is supplied from either the auxiliary boiler or thesteam converter

L*L1ArURJSjkf1= jEAV_ U IES, LTM UAP-HF-08013-19LTD. UAP-HF-0801 3-19

Summary

Chapter 10 deals with the steam and powerconversion system

The steam and power conversion system isdesigned to remove the heat energy from thereactor coolant system and to convert it toelectrical energy in a safe mannerThe turbine and control/protection systems aredesigned so that the probability of turbine missileis less than the number specified in SRP

MIR :WWLMVMM , I -Y-M U 'CRIES, LTD. UAP-HF-08013-20UAP-HF-08013-20~MUISUB ISHI-KE-AVY-INDUSTRIES. LTD. UAP-HF-ORO1 ~-2fl

US-APWR


Detail of FSAR

Tier2: Chapter 11 (Dose Evaluation)


SITSUISI= AVY-INDUSTRIES, LTD. UAP-HF-08015

Presenter APW

Hiromasa Nishino

Engineering Manager

Radiation Safety Engineering Section

Reactor Safety Engineering Department



L-MijX-$jQ1%1j5 E4 X S IES, LTD."I= J_ V_ =INQ-U--T1k UAP-HF-08015-1~.MI.1S.IA~JS!.B~HEAV.Y4NDUSTRIES LTD. UAP-HF-0801 5-1

Contents


/ Title of Chapter

v' Scope of Chapter

2. Design Features

3. Dose Evaluation Methods, Criteriaand Results

4. Summary

I SUB ISHIEA-Y-INDU RIE S, LTD. UAP-HF-08015-2


> Title of ChapterChapter 11: Radioactive Waste Management

> Scope of ChapterThis chapter includes following items;

o Source Term

* Liquid Waste Management System (LWMS)o Gaseous Waste Management System (GWMS)I Solid Waste Management System (as presented

in "Detail of FSAR Tier 2 : Chapter 11(System)")Process Effluent Radiation Monitoring andSampling System (ditto)

X_1"jkQkT1K1ES, LTD. UAP-HF-08015-3~-MI.TSUBISHU-HEAV-Y-INDUSTRIES. LTD. UAP-HF-0801 5-3

1. Overview of Chapter (Cont'd) QP

> Scope of Chapter (Cont'd)

Each item described above includes subitems asfollows;

Source Termv/ Design Basis Reactor Coolant and Secondary

Coolant Activity,,'Realistic Reactor Coolant and Secondary Coolant

Activity

Notes : Fission products and activation products are considered.

M a!SEUSBISHHEvY-IDUSTREES, LTD. UAP-HF-08015-4

1. Overview of Chapter (Cont'd) AW7


• LWMS (Radioactive Effluent Releases)

/ Radioactive Effluent and Dose Calculation inNormal Operation

v' Radioactive Release due to Liquid ContainingTank Failure

GWMS (Radioactive Effluent Releases)

/ Radioactive Release and Dose Calculation inNormal Operation

/Radioactive Release and Dose Calculation due toGWMS Leak or Failure

Lmi--T-suni! ta-LH AM-X-ýIWWVATSkIES, LTD. UAP-HF-08015-5MI~SUBISHI-HEAV-Y-INDUSTRIES, LTD. UAP-HF-0801 5-5

2. Design Features /_033_ý_h'ý_ýWlffl ,

" Chapter 11 presents information on source termsof radioactive material generated within reactorcore and released via LWMS and GWMS.

* Two source term models are utilized to calculatethe radionuclide concentration in the reactorcoolant and secondary coolant.

.- M IMCU•l H MCAI/Y lIllTI lCc I Tr.. I IAP-HI=-NANt A-RinETIE~~UEUEA~~ E33~UQTDE~ TI IUAP-HF-08015-6

2. Design Features (Cont'd) AP

Design basis source term(for shielding design)

" Fuel defect:1 %" Mass balance equations described in DCD are

used to calculate each nuclide activity.

> Realistic source term(for dose evaluation during normal operation)

* Based on ANSI/ANS-18.1-1999(*)

* PWR-GALE Code is used to calculate realisticsource term and released activity during normaloperation.(*)equivalent to approximately 0.2% of fuel defect

MAO 1% 1 ý H -I- " E 4V-Y 'KIkIES, LTD. UAP-HF-08015-7~MUI3UBISHI-HEAV-Y-INDUSTRuES, LTD. UAP-HF-O8O1 5-7

3. Dose Evaluation Methods, Criteria and f-Results _____

Evaluation Item Evaluation MethodIndividual dose during Radioactive Releases: PWR-GALE Codenormal operation Dose Evaluation (Liquid) : LADTAPII Code

Dose Evaluation (Gaseous) : GASPARII Code

Dose due to GWMS Branch Technical Position 11-5leak or failure Considered operational mode of US-APWR

Radioactive Effluent Branch Technical Position 11-6Releases due to liquid NUREG-0133 Appendix A (RATAF Code)- containing tankfailure

Each evaluation is performed using assumedconservative site characteristics.

_MI=T.SUBISHI-_HEAV.Y-!ND UiS•RI ES, LTD. UAP-HF-08015-8

3. Dose Evaluation Methods, Criteria andResults (Cont'd)

Evaluation Item Criteria ResultsIndividual dose during 10 CFR 50 Appendix Inormal operation Liquid

Total body 3 mrem/y 1.98 mrem/y*

Organ 10 mrem/y 2.54 mrem/y**

Gaseous (Noble gases)Gamma dose in air 10 mrad/y 0.210 mrad/y

Beta dose in air 20 mrad/y 1.62 mrad/y

Total body 5 mrem/y 0.134 mrem/y

Skin 15 mrem/y 1.26 mrem/y

Gaseous (Iodine, Particulates)Organ 15 mrem/y 10.2 mrem/y***

Dose due to GWMS leak Branch Technical Position 11-5or failure 100 mrem 46 mrem

Radioactive Effluent 10 CFR 20 Appendix B Table 2 Col.2 0.22Releases due to liquid - (Summation of fractions of concentrationcontaining tank failure limit is equal to or less than 1.0)

*Child **Child's Liver Child's bone **** Summation of Fractions of Concentration

L-STiRIES. LTD. UAP-HF-08015-9

4. Summary

Source terms for radiation protection designand dose evaluation were evaluated usingtwo source term models, i.e. design basissource term and realistic source term.

Using assumed conservative sitecharacteristics, dose evaluations wereperformed according to standard methodsin the U.S. and complied with dose criteria.

~MIT5UISHI-EA X--IDUtl IES, LTD. IIPH~lf1~f[IAP-HF-ORt31 R-4 ClUAP-HF-08015-10

US-APWR


Detail of FSAR

Tier2: Chapter 12


I-.TV.IIDICU2 U=AI'J *PJMIC7EPCTDE I 3I% I IAP-HF-flAfltA_= UAP-HF-08016

Presenter (!APW0

Hiromasa Nishino

Engineering Manager.

Radiation Safety Engineering Section


Nuclear Energy Systems HeadquartersMitsubishi Heavy Industries, LTD.

Z UG I.-T-WBM. "I =-HE -AVr-Y=1j-MQU-%.TR1ES, LTD. UAP-HF-08016-1~MEISUBISHI-HEAV-Y-INDUS.T~RIES. LTD. UAP-HF-08016-1

Contents


/ Title of Chapter

/ Scope of Chapter

2. Design Features

3. Summary

~MLSUBSHIIEV-YINDSTIES, LTD. U/r-lrM-USUi b-/.

I. Overview of Chapter

STitle of Chapter

Chapter 12: Radiation Protection

> Scope of ChapterThis chapter includes following items;

/ Considerations for ALARA*v/ Radiation Sourcesv/ Radiation Protection Design Features

*ALARA: As Low As Reasonably Achievable

LMI:IjSQBtSHj-KEAV- - JWU IES, LTD.XjP _STW UAP-HF-08016-3LTD. UAP-HF-0801 6-3

1. Overview of Chapter (Cont'd)

> Scope of Chapter (Cont'd)Each item described above includes subitems as follows;

Considerations for ALARAv/ Policy Considerationsv/ Design Considerations

,/ Operational Considerations - COL items

" Radiation Sources

/Contained Sources

/ Airborne Sources

_. MSUBQISH4I EAV.-INDHUS1RIES, LTD. UAP-HF-08016-4

1. Overview of Chapter (Cont'd)


Radiation Protection Design Features

v/" Plant Design Features for ALARA

'./ Shielding

/v Ventilation

/Area Radiation and Airborne Radioactivity

Monitoring Instrumentation

/ Dose Assessment

Note: Operational Radiation Protection Program-C COL Item

ýýM ITSMBi _SHI-H kAV-X-kWD-Q5-T-W1ES, LTD. UAP-HF-08016-5UAP-HF-08016-5~~METSUBISHI~HEAVY-ENDUSTRIES. LTD. UAP-HF-OBO1 6-5

2. Design Features A

> Ensuring that Occupational Radiationexposures are ALARA

*Policy Considerations

* Design Policies: Design by nuclear engineers with ALARAphilosophy and system to cooperate plant experience

* Operation Policies: Comply with RG 1.8,8.8&8.10 - COL Item

*Design Considerations

Equipment and Facility Layout are designed to minimize thepersonnel time spent in radiation areas and to minimize theradiation levels in routinely occupied plant areas

_M/.TSUBIISI-HEAV-Y-!MDU $gTkR I ES, LTD. UAP-HF-08016-6

2. Design Features (Cont'd)

> Radiation Sources*Sources for Full-Power Operation

-Contained Sources : 1% Fuel defect considered

- Airborne Sources : Constant leakage from equipments to atmosphereconsidered

*Sources for Shutdown

* Reactor Core: Specific Power of 32.1 MW/MTU and two cycles operationconsidered

* Spent Fuel : Specific Power of 32.1 MW/MTU and Burn-up of 62GWD/MTU considered

* Incore Flux Thimbles : Activated Cobalt-60 considered

*Sources for Design-Basis Accident

* Fission Products released into the containment based on RG 1.183 -considered

Aý= M-12TSLUERSH1-HEAV;Y=tjWqU.ST4t1ES, LTD. UAP-HF-08016-7I~UBISNLHEAV~Y~INDUSXRIES. LTD. UAP-HF-0801 6-7

2. Design Features (Contd) k,4AfPWRV

>Radiation Protection Design Feature*Facility Design Features " All equipment is designed to ensurethe Occupational Radiation Exposures ALARA

*Shielding Design : designed to be in

* Compliance with 10 CFR 20 under normal operation/shutdown* Compliance with 10 CFR 50 Appendix A and NUREG-0737 underDesign-Basis Accident (Main Control Room)

*Ventilation Design & Area Radiation and Airborne RadioactivityMonitoring Instrumentation Design : considered for ALARA

*Dose Assessment* Occupational Exposure: about 70 Person-rem/year* Post-Accident Actions : Radiation Exposures in Post- Accident

Sampling are compliance with 10 CFR 50.34 (f)(2)(viii)* Radiation Exposures at Site Boundary:

- Direct Radiation : negligible- Dose due to Airborne Radioactivity : given in Chapter 11

~I~1I5WjEAAI~.NDISTRESLTD. UAP-HF-08016-8

2. Design Features (Con'd)A

)Radiation Zones for Shielding Design and Radiation ControlZone Maximum

one Dose Rate Description

I • 0.25 mrem/h Controlled area, unlimited occupancy

II 1 mrem/h Restricted area, limited occupancy

III 2.5 mrem/h Restricted area, limited occupancy

IV 15 mrem/h Restricted area, limited occupancy

V 100 mrem/h Restricted area, limited occupancy

High radiation sources. Restricted area, limited occupancy forVI 1 rem/h very short periods. Access controlled as stated in the Technical

Specifications.

VII 10 rem/h Same as Zone VI above

VIII 100 rem/h Same as Zone VI above

IX 500 rad/h Same as Zone VI above

Very high radiation sources. Restricted area, very limitedX > 500 rad/h occupancy for the shortest periods. Access controlled as stated in

the Technical Specifications.

III

A2AMrTSUB1j5j"1=H.ýAVX_1NQWSTk1ES, LTD. UAP-HF-08016-9~MI ESUBISHI-HEAV-Y- ENDUSi RUES, LTD. UAP-HF-08016-9

3. Summary

Policy Considerations, Design Considerations,Radiation Sources and Radiation Protection DesignFeatures to ensure that Occupational Exposuresare ALARA are described in chapter 12.

Radiation Protection Design complies with 10 CFR20 and 10 CFR 50 for Normal Operation/Shutdownand Post-accident Actions

Dose Assessment for Occupational Exposures andpost-accident actions meet NRC's generalrequirements and/or 10 CFR 50.34

LMISU IH E V-Y-IIJUSXTJ.IES, LTD. UAP-HF-08016-10

US-APWR


Detail of FSAR

Tier2: Chapter 13


-ES, LTD. UAP-HF-08017

Presenter

Atsushi Kumaki

Engineering Manager

APWR Promoting Department



IIAD UI" I't O /'t ,4 "7 ,4________________MIMES,______ LT.U~r-nr-uou El- a

Contents


v' Title of Chapter

v Scope of Chapter

2. Topics of Section

3. Summary

IES, LTD. UAP-HF-08017-2


STitle of Chapter

Chapter 13: CONDUCT OF OPERATION

C-

> Scope of ChapterThis chapter provide informationrelating to the preparations and plansfor the design, construction, andoperation

P U5ST"IES, LTD. UAP-HF-08017-3


> 13.1 Organizational Structure of Applicant

'( This section makes clear the COL Applicant'sresponsibility to describe;

* management and technical supportorganization,

" operating organization, and

* Qualification of Nuclear Power PlantPersonnel

NE•UjjI-FlA-Y-INRU5TR ES, LTD. UAP-HF-08017-4


o 13.2 Training

V The development of training programs is theresponsibility of the COL Applicant

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13.3 Emergency Planning

v' This section provides design featuresnecessary for emergency planning, e.g.;

* technical support center (TSC),

* emergency operations facility (EOF),

* emergency response data system (ERDS),

• data communication system,

* safety parameter display system, and

* post accident monitoring system

mITEEDJEUU II -m-iN TIEES, LTD. UAP-HF-08017-6


13.4 Operational Program Implementation

/ The development of operational programimplementation is the responsibility of the COLApplicant

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13.5 Plant Procedures

v" This section makes clear the COL Applicant'sresponsibility to develop;

* administrative procedures, and

" operation and maintenance procedures

L IMWQI UinE E JI SMIES, LTD. UAP-HF-08017-8


13.6 Security

v' This section makes clear the Applicant'sresponsibility to develop;* security assessment,

• plant overall security plan,0 implementation schedule for the security

program, and0 proposed ITAAC for physical security hardware

V A security safeguards report will identify vital areasand vital equipment and other physical protectioninformation for US-APWR standard design

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13.7 Fitness for Duty

/ The development of the fitness-for-dutyprogram is the responsibility of the COLApplicant

1E5••, LTD. UAP-HF-08017-10

3. Summary

> Chapter 13 provides information relating to thepreparations and plans for the design,construction, and operation of the US-APWRplant.

> The purpose of Chapter 13 is to provideadequate assurance that the COL Applicantestablishes and maintains a staff of adequatesize and technical competence and thatoperating plans to protect the public health andsafety.

IEA I R _ UTIES, LTD. UAP-HF-08017-11

US-APWR


Detail of FSAR

Tier2: Chapter 14

January 15, 16, 2008


IIAfl Ur" fl•0l•40FA Uff#" PUA E5,LTDU. LAr'-r-1r-uou 1O

Presenter (fAPS4

Atsushi Kumaki

Engineering ManagerAPWR Promoting DepartmentNuclear Energy Systems HeadquartersMitsubishi Heavy Industries, Ltd.

I IAr'll I ir- #•Lf'•Al• A

L T D. uAr-mr-Uou -1

Contents AP


v1 Title of Chapter

v- Scope of Chapter

2. Chapter 14 Contents

3. Summary

jaOX- iMSV-H "PUS-FRIES, LTD. UAP-HF-08018-2

1. Overview of Chapter A di54 ýTitle of Chapter

Chapter 14: VERIFICATION PROGRAMS

SScope of Chapter

This chapter consists of

1) Initial Test Program Part (14.1 & 14.2)(Administrative Control & Test Abstracts)

2) ITAAC Screening Part (14.3)

3) Supplemental Information (Appendix 14A)

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~~B5N-tEAY-MNPU-F IES, LTD. u~r-mr-uouI ~--

1. Scope of Chapters and Interfaces P

2.1 Initial Test Program(Section 14.1 &14.2) S[RG 1.206

Administrative--R 14.2

Test AbstractSubpart L7 1.687

Appedix14A2.3 Supplemental

O -he -Tir -: ,Information'Other Tier 2

2.2 ITAAC Screening(Section 14.3)

ITAAC SelectionMethodology

Cross Referenceof Key Designbetween Tier 2

and Tier I

I:TierI

I ----------- Im

Chapters

-- Ch. 14 Scope RegulatoryGuidance

-- Out of ScopeI--.I

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2. Tier 2 Chapter 14 Contents

"Verification Programs"

2.1 Initial Test Program Part (14.1 & 14.2)

2.2 ITAAC Screening Part (14.3)

2.3 Supplemental Information (Appendix 14A)

LMMY ISHI HEAVY MPUSTRIES, LTD. UAP-HF-08018-5m~grALISAWJ&HI I~EAV'Y INDUSTRIES, LTD. UAP-HF-0801 8-5

2.1 Initial Test Program Part.54 'iP

Section 14.2 consists of> Administrative Control Subpart

This part addresses general commitment of theadministrative control.Site-specific administrative control is describedin the COLA phase.

> Test Abstract SubpartMost of the test abstracts is addressed in theDCD. Some of the site-specific test (e.g.personnel monitors and radiation surveyinstruments) will be added in the COLA.

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2.1 Contents Example A!PS:W4

> Test abstracts are developed based on theMHI's initial test experience and U.S. regulatoryguidance (including the past FSAR).

14.2.12.1.5 Pressurizer Relief Tank Preoperational Test

A. Objectives1. To demonstrate that design pressurizer relief tank spray flow2. To demonstrate the filling and draining operation of

B. Prerequisites .1. Required construction testing is co I2. The containment vess I react t ...... is available to the drain .....

C Tes .... tho1. Witest __ r s re ....... the required spray flow is pumped to the pressurizer relief tank.

2. While0• .. the nitrogen pressurization system operation is observed.

D. Acceptance Criteria1. The required spray flow is obtained as designed (see Subsection 5.4.11)

2. The pressurizer system ......

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2.2 ITAAC Screening Part

Section 14.3 consists of

; ITAAC Selection MethodologyAll selection methodology (including forEmergency Planning ITAAC and PhysicalSecurity ITAAC) is provided in the DCD.

>Cross Reference Table between Tier2 Key Design Features and Tier 1Description.

The significant parameters and key designfeatures in Tier 2 are listed with the applicableTier 1 description and section numbers.

!,I.S HI• HII_=P•- US, IES, LTD. UAP-HF-08018-8

2.2 Contents Example

The cross-connection clearly shows howand where most significant key design areaddressed in Tier I and Tier 2.

Tier I Ref. Key Design Features Tier 2D ILocation

Subsection The valves close within the I fter Subsection2.7.1.2.1 receipt of an actuation si 10.3.2.3.4

The main steam is I ~SIVs) closewithin 5 c D

Thees27.12-4 Thesrd capacities of the MSSVs Subsection0l,000 Ib/hr.... 10.3.2

Table 10.3.2-2

Table 2.7.1.2-4 The flow restrictor within the SG....... Subsection15.1.5.2

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2.3 Supplemental Information 1::::::A ýPS

•Comparison Table with RG 1.68/ Each item of the RG 1.68 Appendix A is

listed with applicable test abstracts.

/ Exceptions from theguidance are clearlyjustification.

regulatoryspecified with the

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lu ES, LTD. UAr-M--UbUI -IU

2.3 Contents Example

Comparison Table Example in Appendix 14A-/This cross-connection clearly shows the conformance

with the regulatory guide.

RG 1.68Appendix A

SectionNumber

i i

1.h.A7• 14.2.12.1.57

Typical Test

Safe• m ulator Testing- D •~tr Storage System

\•,•ational Test14.2.12.1.59

I .h.(8) IWot applicableThis system does not have an ESFfunction in the US-APWR.

i

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3. Summary (A#4

)Chapter 14 contents completelyconform to RG 1.206.

.Cross-connection between RG 1.68and individual test abstracts areavailable for the reviewer'sconvenience.

)These contents provide sufficientinformation for NRC's review.

I I IP •6 •__mIX-S-U- UAV-HI--M5UItS-Id_;ES, LTD.

14

US-APWR


Detail of FSAR

Tier 2: Chapter 15

January 15,16, 2008


lIr-SUIH 1 WE A--N DuIlE S, LTD. UAP-HF-08019,UAP-HF-08019.

Presenter !AP&*'

Keith PaulsonSenior Technical Manager and Licensing Manager


ý-MIXVUIUR U1-U1PAff-'V-11Wn11C.'r011PC ITn LIAP-HF-08019-1-- ~EQI II~I.LJJ~~I~IlI fu i TDEE I TlIUA P-H F-08019 -1

Contents

1. Overview of Chapter 15

2. Design Features Related to Transient andAccident Analyses

3. Selection of Design Basis Events andAcceptance Criteria

4. Event Categorization and Computer CodesUsed

5. Analysis Methods

6. Analysis Results

7. Summary

EI.SUBISHI_.H EAV-Y-•.IND•UU$T•IES, LTD. UAP-HF-08019-2

1. Overview of Chapter 15

STitle of ChapterChapter 15: Transient and Accident Analyses

> Scope of ChapterTransient and Accident analyses reported in theDesign Control Document (DCD) include eight(8) categories of events to comply with theRegulatory Guide (RG) 1.206 and StandardReview Plan (SRP) NUREG-0800

MITSU BISHI-HEVVY-!NDUSRES, LTD. UAP-HF-08019-3

2. Design Features Related to Safety Analysis -

US-APWR Plant Parameter SummaryV' Larger core thermal output with improved efficiency

V Enhanced thermal margins due to the lower average linearheat rate

Features US Current4 Loop Plant

Core thermal output (MMt) 4,451 3, 565

Number of loops SGs and RCPs 4 4

Number of fuel assemblies 257 193

Fuel rod lattice 17x 17 17x 17

Active fuel length (ft) 14 12

Average linear heat rate (kW/ft) 4.6 5 7

Reactor coolant pump type Centrifugal Centrifugal

Steam generator type U-Tube U-Tube

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2. Design Features Related to Safety Analysis (Cont'djfS

US-APWR Design FeaturesV Very similar to current PWRs in the US/ Design Features and the Effects on Safety Analyses

Featumres

Neutron Reflector

I Effects on Safety AnalysesNeutron Reflector is explicitly modeled in LOCA analysesNegligible change in neutron kinetics

Simplified core lower Core inlet mixing among loops approximately the sameplenumPressurizer Larger steam space moderates pressure transients

Steam generator Smaller U-tube diameter improves transient performance in caseof SGTR*'

ECCS and EFWS*2 4 independent trains with one pump per train

Diverse actuation Satisfies design requirements to cope with A TWS*3system

Advanced Characteristics of Advanced Accumulator is modeled in LOCA

Accumulator analysesNot expected to actuate during Non-LOCA events

*1 SGTR -Steam Generator Tube Rupture;

*3 A TWS -Anticipated Transients Without Scram

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*2 EFWS -Emergency Feedwater System

UAP-HF-08019-5

3. Selection of Design Basis Events andAcceptance Criteria

> Design Basis EventsV Basic design of the US-APWR is the same as the current PWRs

in the U.S. from the viewpoint of- primary and secondary system configurations- thermal hydraulic characteristics and main plant parameters- fuel properties- core kinetics- reactor control and protection system functional design

/ The US-APWR design does not introduce any new initiatingevents for safety evaluation.

/ All transients and accidents in NRC Standard Review Plan (SRP)Chapter 15, applicable to PWRs, are included

$ 8 categories based on the causes of transients consistent withthe SRP

> Acceptance criteriaV SRP Acceptance Criteria are applied for US-APWR analyses

-MITSMUBI AHHE V 11- UAV SIES, LTD. UAP-HF-08019-6

4. Event Categorization and Computer Codes Used

> SRP Chapter 15 Events, Classification.C mpuIter CodesSection Events Category Computer Code(s) Utilized

15.1.1 Decrease in feedwater temperature AOO MARVEL-M

15.1.2 Increase in feedwater flow AOO MARVEL-M

15.1.3 Increase in steam flow AOO MARVEL-M

15.1.4 Inadvertent opening of a steam generator relief AOO MARVEL-M, ANC, VIPRE-0IMor safety valve

15.1.5 Steam system piping failures - Minor I Major AOO I PA MARVEL-M, ANC, VIPRE-01M

15.2.1 Loss of external electrical load AOO MARVEL-M

15.2.2 Turbine trip AOO Bounded by loss of load

15.2.3 Loss of condenser vacuum and other events AOO Bounded by loss of loadresulting in turbine trip

15.2.4 Inadvertent closure of main steam isolation AOO Bounded by loss of loadvalves

15.2.5 Steam pressure regulator malfunction or failure AOO No steam pressurethat results in decreasing steam flow regulators in the.US-APWR

whose malfunction or failurecould result in a steam flowtransient.

15.2.6 Loss of non-emergency AC power to the station AOO MARVEL-Mauxiliaries

15.2.7 Loss of normal feedwater flow AOO MARVEL-M

15.2.8 Feedwater system pipe break - Minor I Major AOO I PA MARVEL-M

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4. Event Categorization and Computer Codes Used /-us(Con t'd)

.. SR Chapter 15 Events, Classification, Computer CodesSection Events Category Computer Code(s) Utilized

15.3.1.1 Partial loss of forced reactor coolant flow AOO MARVEL-M, VIPRE-01M

15.3.1.2 Complete loss of forced reactor coolant flow AOO MARVEL-M, VIPRE-01M

15.3.2 Flow controller malfunctions (not applicable to - N/A - BWR EventUS-APWR)

15.3.3 Reactor coolant pump rotor seizure PA MARVEL-M, VIPRE-01M

15.3.4 Reactor coolant pump shaft break PA Bounded by rotor seizure

15.4.1 Uncontrolled RCCA bank withdrawal from a AOO TWINKLE-M, VIPRE-01M,subcritical or low power startup condition MARVEL-M

15.4.2 Uncontrolled RCCA bank withdrawal at power AOO MARVEL-M

15.4.3 RCCA misalignment AOO I PA MARVEL-M, VIPRE-01M

15.4.4 Startup of an inactive reactor coolant pump at an AOO N-1 loop operation not.incorrect temperature allowed

15.4.5 Malfunction / Failure of flow controller in BWR - N/A - BWR Eventrecirculation loop

15.4.6 CVCS malfunction that results in a decrease in AOO Evaluation without computerboron concentration in the reactor coolant code

15.4.7 Inadvertent loading and operation with fuel PA ANCassembly in improper location

15.4.8 Spectrum of RCCA ejection accidents PA TWINKLE-M, VIPRE-01M,MARVEL-M

L_ ISuBJAHiHE AV-Y-iNDUS T IES, LTD. UAP-HF-08019-8

4. Event Categorization and Computer Codes Used r•J(Cont'd)

SRP Chapter 15 Events, Classification, Computer CodesSection Event Category Computer Code(s) Utilized

15.5.1 Inadvertent actuation of the emergency core AOO N/A - shut off head of the SIcooling system during power operation pump is below nominal

operating pressure

15.5.2 CVCS malfunction that increases reactor coolant AOO MARVEL-Minventory

15.6.1 Inadvertent opening of a pressure relief valve AOO MARVEL-M

15.6.2 Radiological Consequences of the Failure of AOO RADTRADSmall Lines Carrying Primary Coolant OutsideContainment

15.6.3 Steam generator tube rupture PA MARVEL-M

15.6.4 Radiological Consequences of Main Steam Line - N/A- BWR EventFailure Outside Containment (BWR)

15.6.5 Loss-of-Coolant-Accidents Resulting from PA WCOBRA/TRAC, HOTSPOT,Spectrum of Postulated Piping Breaks Within the M-RELAP5Reactor Coolant Pressure Boundary

15.7 Radioactive Release from a Subsystem or AOOIPA RADTRADComponent

15.8 Anticipated Transient Without Scram N/A N/A

L_ U$_Tk1ES, LTD. UAP-HF-08019-9UAP-HF-08019-9LT$UBISHI~HEAV~V~-INPiJSNRIES, LTD. UAP-HF-0801 9-9

5. Analysis Methods US

>LOCA/ Large Break LOCA

WCOBRAITRAC code with the ASTRUM methodologyis implemented

Applicability of this methodology for US-APWR issubmitted in Topical Report entitled "Large BreakLOCA Code Applicability Report for US-APWR"(MUAP-07011-P(RO), July 2007) and is under review

/ Small Break LOCAM-RELAP5 code which incorporates Appendix-K

requirements is usedThis methodology is submitted in Topical Report

entitled "Small Break LOCA Methodology for US-APWR" (MUAP-07013-P(RO), July 2007) and is underreview

Plant Sensitivity analyses are provided in TechnicalReport (MUAP-07025-P(RO), December 2007)

-i._TSUSHI-HEAV-Y-!!-I•NDTUSI ES, LTD. UAP-HF-08019-10

5. Analysis Methods (Cont'd)

Non-LOCA ':1

/MARVEL-M code and TWINKLE-M code areapplied

/TWINKLE-M 3-0 neutron kineticsmethodology is used for the Rod EjectionAccident analysis from HZP condition

,/Applicability of the methodology forUS-APWR is submitted in Topical Reportentitled "Non-LOCA Methodology", (MUAP-07010-P (RO), July 2007) and is under review

III.-T.SUBISHI-H&WV4--•ND-USKTRIES, LTD. UAP-HF-08019-11

5. Analysis Methods (Cont'd) !APS4

>Radiological Consequence Analyses

-/Alternative source term is used,/Methodologies which consider decay, removal

and transport of radioactivity based on plantdesign are applied and equivalent to currentU.S. PWR

v/ RADTRAD code, approved by NRC, is used

NJ

41MVKSUBLSHI-EkV-Y-M $ISTRIES, LTD. UAP-HF-08019-12

6. Analysis Results

1. The transient and accident response of the US-APWR issimilar to that of current PWRs in the US

2. All analysis results satisfy the SRP Acceptance Criteria

3. Large thermal margin due to the lower average linear heatrate greatly enhances the safety margins

e No DNB forAOOs

> Minimal fuel failure and radiological consequences for PAs

> No PCMI fuel failure for Rod Ejection Accident

4. Enhanced ECCS performance> Large PCT margin for LOCA

> Small increase in cladding temperature due to loop sealformation.

LMI-T-SUWISHI-HEAV-Y-_INDU.ATRIES, LTD. UAP-HF-08019-13MIThUBISHLHEAVZY-INPMSTRIES. LTD. UAP-HF-O8O1 9-13

6. Analysis Results (Cont'd) 111ýw-AQOs (Anticipated Operational Occurrences)

" Minimum DNBR remains above the 95/95 limit and no fuel failuresare predicted

V" RCS pressure and main steam system pressure remain well below110% of respective system design pressures

V All AQOs do not generate any other fault that may lead to apostulated accident

> PAs (Postulated Accidents)V Minimum DNBR remains above the 95/95 limit for most PAs.

If the minimum DNBR falls below the limit, the acceptance criteriain 10 CFR 50.46 are satisfied

V RCS pressure and the main steam system pressure remain belowacceptable design limits.

V All PAs do not cause any consequential loss of required functionsof systems needed to cope with the fault.

V Resultant doses are well within the guideline values specified in10 CFR 50.34

"IiICSAJISHIJ4iE-AV2VY-.INRtU5T-RIES, LTD. UAP-HF-0801 9-14UAP-HF-08019-14

6. Analysis Results (Cont'd) QA PX

LOCAV Statistical methodology of large break LOCA demonstrates

that acceptance criteria of 10 CFR 50.46 are satisfiedPCT(95/95) = 1763 OF < 2200 OF

V Conservative analysis of small break LOCA, which is basedon Appendix-K, demonstrates that acceptance criteria of10 CFR 50.46 are satisfied

PCT = 1297 OF < 2200 OF

V Switchover to simultaneous RV and hot leg injection mode atfour hours after a LOCA prevents boric acid precipitation inthe core, then post-LOCA long term cooling is assured

IMMUIPSURISHI-HEA.Y7M-INU-STRES, LTD. UAP-HF-08019-I 5UAP-HF-08019-15

6. Analysis Results (Cont'd)

Reactivity Initiated Accident (RIA) Specific Criteria(SRP 4.2 Appendix B and SRP 15.4.8)

/ The average fuel pellet enthalpy at the hot spot remainssignificantly below 230 cal/g

/ 3-D methodology is applied to analyze Rod Ejection Accidentfrom HZP condition. Prompt fuel enthalpy rise is well belownew threshold for cladding failure.

2O0

175

150

125

u. 100

0 75

50

(0.04,150)

Cladding Failure

(0.08, 75)(0.20, 0)

0 0.04 0.

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08 0.12

d./Wall Thick.*=

0.16 0.2

UAP-HF-08019-16

6. Analysis Results (Cont'd) /-Us-

RADIOLOGICAL CONSEQUENCE ANALYSES/ The exclusion area boundary (EAB) and the outer boundary

of low population zone (LPZ) doses are shown to meet the10 CFR 50.34 dose guidelines

The dose results (LOCA) are 13rem < 25rem at EAB,13rem < 25rem at LPZ

V The dose for the MCR personnel is shown to meet the dosecriteria given in GDC 19

The dose results (LOCA) are 4.5rem < 5rem in MCR

&.-MTr#tXB1%M HEAVY IMPUSTRIES, LTD. UAP-HF-08019-17L..MLTWW4$HI HEAVY INDUSTRIES, LTD. UAP-HF-0801 9-17

7. Summary AP

1. US-APWR DCD FSAR Chapter 15 format andcontent comply with RG 1.206 and satisfy the SRPrequirements

2. All results of transient and accident analyses meetthe acceptance criteria

3. Methodologies and codes for US-APWR arediscussed in Topical Reports for NRC review

4. Supplemental information is provided in TechnicalReport to support DCD review

L. IVCI IIMBU UMAXI~ IffJnZUIC~TUPC I Vff% I IA P.I-IFP.lRfllQ•-4RI UAP-HF-OBOIQ-18

US-APWR


Detail of FSAR

Tier2: Chapter 16


; _MfI.SUBISHI-HEVJ-Y--INDUSlRIES, LTD. UAP-HF-08020

Presenter

Katsunori KawaiEngineering Manager

APWR Promoting Department



T.SýVW$""EW-Y-FIM IIES, LTD..Tjk UAP-HF-08020-1~e~MIT5UBI5HU~HEAV-Y-INDU5jR!ES, LTD. UAP-HF-08020-1

Contents A ýPM

Contents H


Title of Chapter

Scope of Chapter

2. Features of TechnicalSpecifications (TS)

L~41$!BEL!EAVY4DUSRISLTD. UAP-HF-08020-2

1. Overview of Chapterus -3_

'-ýW 1 0

Title of Chapter

Chapter 16: Technical Specifications

SScope of Chapter* This chapter includes the following categories

of information as required by 10 CFR 50.36 and10 CFR 50.36a

* Safety limits, limiting safety system settings,LCOs, surveillance requirements, designfeatures and administrative controls

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2. Features of TS/-Uso/,-.-ý_ýt,,

AA

Features of US-APWR safety system design0 Design concept is based on current

PWRs in the USA

H

* Four-train safety systems are one ofcharacteristic design features

Features of US-APWR TechnicalSpecifications

* Basically follow the Standard TS* (STS)

* Maximize the benefits of on-line maintenance (OLM)

* Apply Risk-Managed Technical Specifications

* NUREG-1431, Rev.03, "Standard Technical Specifications Westinghouse Plants"

LR!Th!AtIlkIU$EAV--IDU5JRIES, LTD. UAP-HF-08020-4

2. Features of TS (cont'd)

2.1 Utilization of STS

* US-APWR Technical Specifications are almost sameas the STS of NUREG-1431

* US-APWR Technical Specifications differ from STSonly as necessary to reflect technical differencesbetween conventionalUS- PWRs design and US-APWR design

* Justification for deviations between STS and US-APWR TS is described in technical report *

•: Justification for Deviations between NUREG-1431 and

US--APWR Technical Specifications (Dec. 2007)

'- MEESUBISHI-HE-AV-Y•4•ND-U-S-TRIES, LTD. UAP-HF-(08020-5

2. Features of TS (cont'd)

2.2 Safety Benefits of Four-train systems

* Enhanced redundancy (50% x 4)

vCapability beyond single failure criterion

* Maximize the benefits of on-line maintenance

v/Establish LCO requiring three trainsoperable

v/Establish completion time when one of thethree required trains inoperable

MIISU BISHI-HEAV-Y-IND U STRIES, LTD. UAP-HF-08020-6

2. Features of TS (cont'd) W

2.3 Main deviations between STS (NUREG-1431) and US-APWR TSCharacteristic design features

* Four train safety systems- e.g. : LCO is Three of four SIS trains shall beOPERABLE

* Gas turbine generators- e.g. Fuel oil testing program

* Digital Platform- e.g. Actuation logic test interval increased

Surveillance Interval* 24 month refueling cycle

•M•I.T.SUBiSHIHEV=ND US "IES, LTD. UAP-HF-08020-7

2. Features of TS (cont'd) - IF,

2.4 Adoption of Risk-Managed TechnicalSpecifications (RMTS)

Risk-Informed Completion Times (CTs)* A front-stop CT and Commitment to

Configuration Risk Management Program(CRMP)

* 30-day limit as a back-stop CT

* Reference to Risk-Managed TechnicalSpecifications Initiative 4b*

•: NEI 06-09 (Revision 0) "Risk-Informed Technical SpecificationsInitiative 4b Risk- Managed Technical Specifications (RMTS)Guidelines," November 2006.

.MT.SUB tISHi-HEkV.=HY-IU$SjR IES, LTD. UAP-HF-08020-8

2. Features of TS (cont'd) .......

Coming works for RMTS to be completed

Establishment of the station procedure ofthe Configuration Risk ManagementProgram (CRMP)

Training of responsible personnel

Preparation of a PRA model to meet thetechnical adequacy requirement of NEI 06-09

Preparation of an appropriate CRM tool

~MI~UBESI-HEV~I~MLSTIESLTD. UPH-82-UAP-HF-08020-9UAP-HF-08020-9

2. Features of TS (cont'd) QVPS5

Other risk-informed initiatives will be considered

Initiative 5b: Relocation of all SR frequencyrequirements out of TS

> Initiative 1: Actions end states modification$ This initiative would permit, for some system, entry into hot

shutdown rather than cold shut down to repair equipment

> Initiative 7: Non-TS support system impact onTS operability determinations

V This initiative would permit a risk-informed delay time beforeentering LCO actions for Inoperability due to loss of supportfunction provided by equipment outside of technicalspecifications

I;;ýMfl5rUBUH1_WEEAV-Y IMDqSTRIES. LTD. UAP-HF-08020-1 0UAP-HF-08020-10

.5111,(SAP2. Features of TS (cont'd)

P lao SstSubmitdat e~achstage in app~lying RMTS

Stage Tech. Spec. Associated Documents(Incl. RMTS)

DC Design-specific DCD Chapter 19

Plant-specificPlant-specific PRA results consistent with

CL(abllshCd)s FSAR Chapter 19 to support RMTSestablished)

0 Technical report describing PRA

Prior to Plant-specific technical adequacy, CRM tools, CRMP,(All CTs Organization, Training of personnel, etc*

fuel load established) *Implementation manual

*All required ITAAC*: In accordance with NEI 06-09

M--,jM9jQ$TR1ES, LTD. UAP-HF-08020-11*~MI.TSUBISIII-HEAV-Y-INDUSJRIES, LTD. UAP-HF-08020-1 I

US-APWR


Detail of FSAR

Tier2: Chapter 17


L -J_ DLISTRIES, LTD. UAP-HF-08021

Presenter

Naoki Miyakoshi

General ManagerNuclear Energy Systems Quality and SafetyManagement Department



&'=WT-SUB1SH11-HE-AM-, Y-UfflPV5jR1ES, LTD. UAP-HF-08021-1LTD. UAP-HF-08021 -1

Contents


Title of Chapter

v' Scope of Chapter

2. Chapter 17 Contents

3. Summry

IIA[• U• rtO/•FJ4 ,•D

ýMITSUISHI HIE AVY ImNDUSTRIES L TD - V- -- - -- -LAr-r-1r-uouL 1-4


STitle of Chapter

Chapter 17: QUALITY ASSURANCE AND

RELIABILITY ASSURANCE

SScope of ChapterQuality Assurance Program performed duringthe design certification phase

Design Reliability Assurance Program (Phase ID-RAP; design certification phase)

MITSUBISHI HEAVY INDUSTRIES, LTD. UAP-HF-08021-3

1. Overview of Chapter (cont'd) usAzD-

Scope of Chapter (cont'd)

17.1 Quality Assurance During Certification Phase

17.2 Quality Assurance During the Construction andOperation Phase

17.3 Quality Assurance Program Description

17.4 Design Reliability Assurance Program

17.5 Quality Assurance Program Guidance

17.6 Description of the Applicant's Program forImplementation of 10CFR 50.65, theMaintenance Rule

• MITSUBISHI HEAVY INDUSTRIES, LTD.

DC Phase

(COL)

DC Phase

DC Phase

DC Phase

(COL)

UAP-HF-08021-4

2. Chapter 17 Conte

Quality Assurance Program

nts

]CFR Part 50,,/ QAP meetsAppendix B,

requirements of 1C10CFR Part2I anc 10CFR Part52.

v/ QAP is based on the requirements of ASMENQA-1 -1994 "Quality Assurance Requirementsfor Nuclear Facilities Applications,"Parts I and II.

v/ QAP Description for DC phase has beenprepared on the basis of the NRC approvedQAP template (NEI 06-14A Rev.4)

MITSUBISHI HEAVY INDUSTRIES, LTD. UAP-HF-0)Rf021-58021-5

2. Chapter 17 Contents (cont'd)

> Quality Assurance Program_ -"--l ato ' - 14n CF 5 0vt n App•ndix _2

IA Requirements Applicable Remarks (MHI QAP on US-APWR)Aertatutt,1. Organization /2. QA Program /

3. Design Control /

4. Procurement Document Control5. Instructions, Procedures and Drawings v/

6. Document Control I/7. Control of Purchased Materials, Items and - At DC stage this aplies to services

Rervinr_-• lch _ n lv i T=t

8. Identification and Control of Items andMaterials Not Applicable (NUREG-0800 17.5)

9. Control of Special Processes

10. Inspection - At DC stage this applies toinspections for test facilities

11. Test Control , - At DC stage this applies toqualification tests

12. Control of Measuring and Test Equipment13. Handling, Storage and Shipping14. Inspection, Test and Operating status

15. Control of Nonconforming Items16. Corrective Action17. QA Records /: Comply -: N/A

M18. AuditMI~TSU BISHI HEAVY INDUSTRIES, LTD. UAP-HF-08021-6

2. Chapter 17 Contents (cont'd) WI

Reliability Assurance Programv1 The sco e of DCD chapter 17 is Phase I D-RAP.

Phase I Design Certification phase

Phase II Site-specific phase

Phase III Last phase of D-RAP(procurement, fabrication, construction

preoperational testing)

V US-APWR D-RAP identifies risk-significantSSCs and provides risk insights and reliabilityassumptions.

MITSUBISHI HEAVY INDUSTRIES, LTD. UAP-HF-(08021-708021-7

2. Chapter 17 Contents (cont'd) Q

Reliability Assurance Program (cont'd)SResponsibility (Phase I D-RAP)

v' General Manager, APWR project:- Establishment of US-APWR D-RAP program

-/General Manager, Reactor and Plant Safety:- Use of the PRA results and risk insights for the

Expert Panel- Conduct and coordination of the Expert Panel

V General Manager, QA:- Assuring proper implementation of QA program

(Organization, design control, procedure andinstruction, records, corrective actions, audit)


2. Chapter 17 Contents (cont'd)

List of risk-significant SSCs/ The risk and reliability organization is

responsible to provide the RAP related inputs inthe design process.

V List of risk-significant SSCs is initially based onthe result of PRA and Expert Panel.

$ The list and changes shall be approved byExpert Panel.

/ List of risk-significant SSCs and its keyassumptions shall be maintained by the riskand reliability organization.

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3. Summary Q

MHI established QAPD for Design Certificationand carried out design activities in accordancewith the QAPD. MHI submitted the QAPD as atopical report to contribute for the NRC review.

Topical Report

"Quality Assurance Program (QAP) DescriptionFor Design Certification of the US-APWR (PQD-HD-19005 Rev.1)"

MHI established D-RAP (phase 1)program and

prepared a list of risk-significant SSCs.


US-APWR


Detail of FSAR

Tier2: Chapter 19'


IS•UBISHi=HE~V-•Y U-s•T•RIES, LTD. UAP-HF-08022

Presenter

Katsuya KuroiwaEngineering Manager




MlURTSUBIS-HI-HEAV. -Y- INW ýD-U-$Rk1E S, L TD. UPH-82-UAP-HF-08022-1

Contents QA A


> Title of Chapter

> Scope of Chapter

2. Probabilistic Risk Assessment

3. Severe Accident Evaluation

4. Summary

K MIEMEDnCUS U=Alff% 1WflhE1EC'rn1WQ U Tf% I IAP-HI=-lNfn79-9UAP-HF-08022-2

1. Overview of Chapter/_VS7ý1;h

Title of ChapterChapter 19: PROBABILISTIC RISK ASSESSMENT

AND SEVERE ACCIDENT EVALUATION

SScope of ChapterProbabilistic Risk Assessment (PRA) resultsand insights including internal and externalevents during full-power operations and duringlow power and shutdown operationsSevere accident evaluations including anassessment of preventive and mitigationfeatures, containment performance capability,accident management and severe accidentmitigation design alternatives (SAMDA)

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2. Probabilistic Risk Assessment CA P-S;b,

> Methods and Approach.Basic design concept of US-APWR is similar to currentPWRs. Therefore, present guides and standards are applied.

,/Regulatory Guide 1.200 Rev.1

,/Standards endorsed by Regulatory Guide 1.200Rev.1

* ASME RA-S-2002 and the addenda ASMERA-Sa-2003, ASME RA-Sb-2005

* ANSI-ANS 58.21-2003,/Areas where no formal standards exists, previousstudies or guidance are used

MiI--UIjSHIHEA-V.Y-INDUPSIRIES, LTD. UAP-HF-08022-4

2. Probabilistic Risk Assessment (cont'd) A

-Special Design Features of US-APWRImproved plant safety as compared to currently

operating nuclear power plants

* Higher redundancy: four train mechanical andelectrical safety systems

• Simplicity: In-containment RWSP eliminatesrecirculation switchover

* Independent: Physical separation of four trainsafety systems

* Diversity: Alternative systems such as diverseactuation system, alternative AC power source

etc.

MIXISURISM-HkAV=Y=1 RkU-S-TRIES, LTD. UAP-HF-08022-5LTD. UAP-HF-08022-5

2. Probabilistic Risk Assessment (cont'd)

> Level 1 Internal Events PRA at Power* Core Damage Frequency (CDF): 1.2 x 10-61RY

FWLS LOAC0'4% -02 LOOC

ATWS SGT I L1.2% 06% 01%

TRANS VSLOCA

PLOCW 12% 01%13% LLOCA

MLOCA 0 04%

1 4%

1 6%

LOFF1 6%

RVR 49.3%65%

LOCCW25 6%

Core Damage Frequency Contribution

"T-SUDaISH HEAVY INDUSTRIES, LTD.

LLOCA Large Pipe Break LOCA

MLOCA Medium Pipe Break LOCA

SLOCA Small Pipe Break LOCA

VSLOCA Very Small Pipe Break LOCA

SGTR Steam Generator Tube Rupture

RVR Reactor Vessel Rupture

Steam Line Break/Leak(Downstream MSIV: Turbine side)Steam Line Break/Leak(Upstream MSIV: CV side)

FWLB Feed-water Line Break

TRANS General Transient

LOFF Loss of Feed-water Flow

LOCCW Loss of Component Cooling Water

PLOCW Partial Loss of Component Cooling Water

LOOP Loss of Offsite Power

LOAC Loss of Vital ac Bus

LODC Loss of Vital DC Bus

UAP-HF-08022-6

2. Probabilistic Risk Assessment (cont'd) Q--W

Level 2 Internal Events PRA at Power* Large Release Frequency (LRF) : 1.0 x 10-7IRY

TRANS VSLOCA FWLB AIWS LOACSLBO, 0 8% 0 6% 02% 01% 01% LLOCA Large Pipe Break LOCA09% 5L51RVR SL% MLOCA Medium Pipe Break LOCA11% LOC SLOCA Small Pipe Break LOCALOFF 001%-13% LLOCA VSLOCA Very Small Pipe Break LOCA

MLOCA 00% .GTR .Steam .Generator. Tube.Rup.. . .

2 1% SGTR Steam Generator Tube Rupture

SGTR RVR Reactor Vessel Rupture60% Steam Line Break/Leak

PLLOC SLBO (Downstream MSIV: Turbine side)346% 1 , -

SLBI -am=. ne= rea ea=(Upstream MSIV: CV side)

FWLB Feed-water Line Break

TRANS General Transient

LOFF Loss of Feed-water Flow

LOCCW Loss of Component Cooling Water

PLOCW Partial Loss of Component Cooling Water

LOOP Loss of Otfsite Power

LOAC Loss of Vital ac Bus

LODC Loss of Vital DC Bus

29,4%

Large Release Frequency Contribution

A-*w P41TSUBISHI HEAVY INPUSTRIES, LTD. UAP-HF-08022-7.~,MLT#IJmNJ 4IM~Y INDUSTRIES. LTD. UAP-H F-08022-7


Uncertainty Analysis for Internal Events PRAat Power

1.0E-05 1.0E-O6 :

1.OE-060

2-9E-06 95percentile

1.2E-06 Mean

7.8E-07 Median

3,OE-07 5percentile

1 .OE-07 -77 ----- - ------

95percenfile;3.OE-07

Mean; 11 E-07

Median; 6.5E-08

5 percentile;2.DE-08

1.0E-07

Core Damage Frequency

•AjIPIOl'NSill HEAVY INPUSTRIES, LTD.

Large Release Frequency

UAP-HF-08022-8


o PRA Results of Other Events

CDF LRF

Seismic (Seismic Margin Analysis)Plant HCLPF: .0.5g

Internal Fire 1.7xl0-6/RY 2.0xl0 7/RY

Internal Flood 1.5x1 0 6/RY 4.0xl 0 7/RY

Other External Site Specific

Low Power and 2.0xl07/Ry Assumed to beShutdown same with CDF

00HUHIFAVY INDUSTRIES, LTD. UAP-HF-08022-9lRI$NI HEAVY INDUSTRIES LTD. UAP-HF-08022-9


> Risk Significant Scenario and SSCs* Station blackout with common cause failure of

emergency gas turbine generators+ Alternative AC power is effective to reduce the risk

" Loss of component cooling water by common causefailure of component cooling water pumps+ Independent trains are effective to eliminate the risk of

loss of cooling water by leak* Fire in the switchyard area, causes loss of offsite power

+ Physical separation is effective to reduce the risk of fire inother areas

* Major flood in the divided area of the reactor building,causes partial loss of safety functions+ Physical separation is effective to reduce the risk

_M I.S U B I ,H -EAV-Y-I ND-U-S-TR IES, LTD. UAP-HF-08022-10


> PRA Insights and Design Features* CDF and LRF are less than the NRC goals

( less than I E-4/year for CDF and less than1 E-6/year for LRF)

* Design features of US-APWR as shown belowreduce the risk.

+ Four train safety systems+ Independent four train electrical system with

alternative AC power source

I in-containment RWSP+ Various severe accident prevention/mitigation

features

Z-ý,MI.X.SýURISH-1-HE-AX-Y=IMCtUS-FgklES, LTD. UAP-HF-08022-11LTD. UAP-HF-08022-1 I


Risk-informed Applications at Design Phase

* PRA has been used to optimize the plantdesign with respect to safety.

• Assumptions of important operating actionsare identified for the accident managementframework.

* Risk significant SSCs are identified for theReliability Assurance Program (RAP).

* PRA insights are utilized to develop risk-managed technical specifications (RMTS).

MflTSBISHI-HEAVY=INMWUSTRIES, LTD. UAP-HF-08022-12

3. Severe Accident Evaluation

Prevention and Mitigation

• Apply proven techniques for existing plants withimprovements

Analysis Approaches and Methods

* Apply analysis approaches accepted by NRC forformer DC applications

" Employ MAAP4.0.6 for severe accidentprogression analysis, and other specific codesfor specific phenomena

LCALTiSUBIj-IHKAV~INPJR - jWkt ES, LTD. UPH-82-1UAP-HF-08022-13UAP-HF-08022-13

1-UPS5 --!#A3. Severe Accident Evaluation (cont'd)

Severe Accident Prevention Features

Anticipated Transient - Four train reactor protection systemWithout Scram - Diverse actuation system

Mid-Loop Operation - Automatic let-down isolation- Alternative core cooling

Station Blackout - Four emergency gas turbine generators- Two alternative AC power sources

Fire Protection - Physically separated four train safetysystems

Intersystem Loss-of- - Up-rated RHRS pipingCoolant Accident

Others - Feed and bleed with redundancy- Alternative component cooling, etc.

~i.MUW H HEAVY INDU*TRIES, LTD. UAP-HF-08022-14

4-7JS5-P_43. Severe Accident Evaluation (cont'd)

Severe Accident Mitigation Features Addressedsevere accidentissues

(1) Hydrogengeneration andcontrol

(2) Core debriscoolability

(3) Steamexplosion

(4) HPME

(5) TISGTR

(6) MCCI

(7) Long-termcontainmentoverpressure

(8) Equipmentsurvivability

Am AWR# 5111 HEAVY INDUSTRIES, LTD.UAHF00-5 UAP-HF-08022-15

3. Severe Accident Evaluation (cont'd) (!APSIR

- Containment Performance (SECY-93-087)* Deterministic goal:

÷ Containment integrity be maintained forapproximately 24 hours following the onsetof core damage for the more likely severeaccident challenges

* Results:

÷ Containment integrity is maintained formore than 24 hours following the onset ofcore damage for most of the severeaccident conditions

-.MI.TUBISHIHEAV-Y-INDUS TRIES. LTD. UAP-HF-08022-16

3. Severe Accident Evaluation (cont'd) (P

I Accident Management* Develop a framework includes:

÷ Approach

÷ Operational and phenomenological conditions

÷ Basis of the actions

* Four countermeasures and operating actions:

+ To prevent core damage

+ To terminate the progress of core damage if itbegins and to retain the core within the reactor,vessel

÷ To maintain containment integrity as long aspossible

÷ To minimize offsite release

~MITSUBAl _-H k-E-- RArINWjr I E S, L TD. IIP.F.R9.1I ap.I -nn9-1•.R • t7

3. Severe Accident Evaluation (c ont'd) AP

> SAMDAMeet the requirement of 10 CFR 50.34(f)(1)(1) toconsider potential design improvements

* Approaches:- Guidance for regulatory analysis (NUREG/BR-0184 and

NUREG/BR-0058)+ Industry implementation guidance (NEI 05-01, Rev. A)

- consistent with SECY-99-169

* Results:

+ Ten candidate SAMDAs are selected from 156 potentialimprovements

+ The benefit of each SAMDA is observed to be significantlyless than the cost impact

* No additional design alternatives are shown to be cost-beneficial in severe accident mitigation design

LM:TSUBIS5H1,aEAV-Y-INDUSTRIES, LTD. UAP-HF-08022-18

4. Summary

" Describe the design-specific PRA* PRA results indicate the US-APWR design meets

the NRC safety goals.

* Describe design features for theprevention and mitigation of severeaccidents

!-MI.,SUBI HIEAVY-INDUSTRIES LTD. UAP-F

(-ALPW34-

IF-08022-19I

Design Certification Application Orientation Detail of ... · This chapter includes the following...

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