ML20101G175

Forwards Addl Stage 3 Tier 1 Design Certification Matl for GE ABWR Design to Suppl & Replace Matl Transmitted by

ML20101G175
Person / Time
Site:	05200001
Issue date:	06/17/1992
From:	Stark S GENERAL ELECTRIC CO.
To:	Pierson R Office of Nuclear Reactor Regulation
References
MFN-130-92, SLK-9278, NUDOCS 9206250321
Download: ML20101G175 (112)

Text

{{#Wiki_filter:. I GENuclear Ettergy unewrwarem uname rne. sr kse cus::< June 17,1992 MFN No.130 92 Docket No. STN 52-001 SLK 9278 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Robert C. Pierson, Director Standardization and Non-Power Reactor Projec: Directorate [

Subject:

Submittal af Addithnal Stage 3 Tier I Design Certification Material for the GE ABWR Design

Reference:

. Letter from P. W. Marriott to Robert C. Pierson," Tier I Design Certification Material for the GE ABWR Design, Stage 3 Submittal," MFN No.12192, Dated May 30,1992 Enclosed are thirty four (34) copies of additional Stage 3 GE ABWR Tier 1 Design Certification material to supplement and replace those transmitted by the referenced letter. The enclosed material includes newly prepared Section 3.6 Human' Factors Engineering, and the revised Section 2.7.5 Multiplexing to include Non Essential Multiplexing System. Other materials include those that were inadvertently omitted in the preparation and prmting of the stage 3 material, as well as correction of minor errors and figures reprinted to improve legibility (Figures in Section 2.15.5 HVAC). Tlw attachment provides a detailed listing of the enclosed maic ial. Sincerely, 5,. S S.J. Stark, Acting Manager Regulatory and Analysis Services M/C 444, (408) 925-6948 cc: F. A. Ross (DOE) N. D. Fletcher (DOE) 1 C. Posiusny, Jr. (NRC) R. C. Berglund - (GE)-- . J. F. Quirk (GE) 100012 9206250321 920617 j)f0 PDR ADOCK 05200001-A. PDR-I

l A'ITACilMENT The following is a list of items submitted, Where replacement is indicated, please discard the corresponding pages and replace with the enclosure. Where items are indicated to be new, please insert the enclosuie into the binder for the design certification material transmitted by the referenced letter. LTEM REPl ACE NEW COMMENTS 2.1.2 X Table 2.1.2a added 2.2.8 X Table 2.2.8, item 2 under Acceptance Criteria, "30%" was "10%", added Fig. 2.2.8 2.4.1 X Table 2.4.1 revised item 18 under Acceptance Criteria 2.7.4

• X Entry corrected 2.7.5 X

Added Section for Non-Essential Multiplexing 2.7.6

• X Entry corrected 2.12.16*

X Entry corrected 2.12.17 Section previously omitted Fig. 2.12.12a,b,c X Figures previously omitted Fig. 2.12.18a,b X Figures previously omitted comments 2.15.2 X Entry corrected Fig. 2.15.5 a - i X Improve Jegibility Fig. 2.15.10 a o X Minor corrections in notes 3.6 X New section Table 5.0 X Correct entries and typographical errors

• Not all copies previously distributed contain the wrong entry.

ABWR Design Document 2.1.2 Nucler,r Boiler System Design Description General System Description The primag functions of the Nuclear Boiler System (NBS) are: (1) to deliver steam from the Reactor Pressure Vessel (RPV) to the Main Steam System (MSS), (2) to deliver feedwater from the Condensate, Feedwater, and Air Fstraction System to the RPV, (3) to provide overpressure protection of the Reactor Coolant Pressure Bounday (RCPB), (4) to provide automatic depressurization of the RPV in the event of a Loss of Coolant Accident (LOCA) where the RPV does not depressurize rapidly and the high pressure makeup systems fail to adequately maintain the water levelin the RPV, and (5) with the exception of monitoring the neutron flux, to provide the instrumentation necessary to monitor the conditions in the RPV. This includes the RPV pressure, metal temperatur e, and water level instrumentation. Figures 2.1.2a and 2.1.2b show the general configuration of the Main Steam Lines (MSLs), the Safety / Relief Valves (SRVs), and the SRV discharge lines. The SRVs perform the dual function of overpressure protection and automatic depressurization of the RPV. Figure 2.1.2c shows the general configuration of h) the Feedwater (FW) lines. u The MSLs are designed to direct steam from the RPV to the MSS, to the FW lines to direct feedwater from the FW System to the RPV, and to the RPV instrumentation to monitor the conditions within the RPV over the full range of reactor power operation. The NBS contains the valves necessag for isolation of the MSLs, feedwater lines, and their drain lines at the pnmay containment boundag. The NBS also contains the RPV head vent line and non-condensable gas removal line. Main Steam Lines The NBS does not contain all of the MSLs. The NBS contains only the portion of the MSLs from their connection to the Reactor Pressure Vessel (RPV) to the boundary with the MSS, which occurs at the seismic interface located downstream of the outboard Main Steam Isolation Valves (MSIVs). The main steam lines are Quality Group A from the RPV out to and including Os the outboard MSIVs, and Quality Group B from the outboard MSIVs to the turbine stop valves. They are Seismic Category I from the reactor pressure vessel out to the seismic interface. 2.1.2 6/16/92

ABWR Design Document To support the safety analysis, the total steam volume of the steam lines, from the RPV to the main steam turbine stop valves and turbine bypass valves, shall be 3 greater than or equal to 113.2 m MSL Flow Limiter Each 51SL has a now limiter. The htSL flow limiter consists of a flow restricting venturi which is k>cated in each RPV MSL outlet nozzle. The restrictor limits the coolant blowdown rate from the RPV in the event a MSL break occurs outside the containment to a (choke) Dow rate equal to or less than 200% of' rated steam 2 now at 72.1 kg/cm g upstream pressure. The MSL Dow limiter also senes as a Dow element to monitor the MSL Dow. Instruments lines are provided to monitor the pressure at the throat of the MSL Gow limiter. The RPV steam dome pressure instrument lines are used to provide the pressure upstream of the MSL flow limiter. The MSL II,w limiters are designed to limit the loss of coolant from the RPV following a MSL rupture outside the containment to the utent that the RPV water level remains high enough to provide cooling within the time required to close the MSIW. The MSL flow limiter has no moving parts. h Main Steam isolation Valves Two isolation valves are welded in a horizontal run of each of the four main steam lines; one valve inside of the drywell, and the other is near the outside of the primag containment pressure boundanj. The MSIVs are Y-pattern globe vahes. The main disc or poppet is attached to the lower end of the stem. Normal steam flow tends to close the valve, and higher inlet pressure tends to hold the valve closed. The Y-pattern permits the inlet and outlet passages to be streamlined; this minimizes pressure drop during normal steam flow. The primary actuation mechanism utilizes a pneumatic cylinder; the speed at which the valve opens and closes can be adjusted. Helical springs around the spring guide shafts will close the valve if gas pressure in the actuating cylinder is reduced. The MSIV quick-closing speed is 2 3 and s 4.5 seconds when Ne or air pressure is admitted to the upper piston compartment. The valve can be test closed with a 45 to 60 second slow closing speed by admitting N or air to both the upper g 2 and low piston compartments. 2.1.2 6/16/92

ABWR oesign Docum:nt Feedwater Lines O The Nils does not contain all of the RV lines. The Nils contains only the portion of the RV lines from the seismic intedace located upsticam of the Motor-Operated Valves (MOVs) to their connections to the RPV. Figure 2.1.2c shows the portion of the RV lines within the Nils. The RV piping consists of two nominal 550 mm (22-inch) diameter lines from the RV supply header. Isolation of each line is accomplished by two containment isolation valves consisting of one check valve inside the dgwell and one positive closing check valve outside the containment. Also included in this portion of the line is a manual maintenance valve located between the inboard isolation valve and the reactor nozzle. The feedwater line upstream of the outboard isolation mlve contains a remote, manual, Motor-Operated (MO) gate valve, and a seismic interface restraint. The outboard isolation valve and the MO gate valve provide a quality group transitional point in the feedwater lines. The feedwater piping is Quality Group A from the RPV out to and including the [ outboard isolation valve, Quality Group 11 from the outboard isolation valve to and including the MO gate valve, and Quality Group D upstream of the MO gate valve. The feedwater piping and all connected piping of nominal 65 mm (21/2 -inch) or larger nominal size is Seismic Category I from the RPV to the seismic (q S interface. Safety / Relief Valves The nuclear pressure relief system consists of SRVs located on the MSI.s between the RPV and the first isolation valve, i.e. the inboard MSIV, within the drywell. These valves protect against overpressurization of the nuclear system. The rated capacity of the pressure-relieving devices shall be stdlicient to prevent a rise in pressure within the protected vessel of more than 110% of the design pressure (1.10 x 87.9 kg/cm g = 96.7 kg/cm'g) for design basis events which cause the RPV pressure to rise. 1 The SRV discharge line is designed to achieve critical flow conditions through the valve, thus providing flow independence to discharge pipe losses. Each SRV has its mm discharge line. The SRV discharge lines terminate at the quenchers located below the surface of the suppression pool. (d 2.1.2 -3 6/16/92

ABWR Design Documnt The SRVs provide three main [notection functions: (1) Overpressure safety operation: The valves f unction as salcty valves and open to prevent nuclear system ovtrpressurization-they are self-actuating by inlet steam presstue if not already signaled open for relief operation. The safety (steam pressure) mode of operation is initiated when direct and increasing static inlet steam pressure overcomes the restraining spring and frictional foices acting against the inlet steam pressure at the main disc or pilot disc and the main disc moves in the opening direction. The condition at which this action is initiated corresponds to the set-pressure value (Table 2.1.2a) stamped on the nameplate of the SRV. (2) Overpressure relief operation: The valves are opened using a pneumatic actuator upon receipt of an automatic or manually initiated signal to reduce pressure or to limit pressure riae. The relief (power actuated) mode of opennion is initiated when an electrical signal is received at any of the solenoid valves located on the pneumatic actuator assembly. The solenoid valve (s) will open, allowing pressurized air to enter the lower side of the pneumatic cylinder which pushes the piston and the rod upwards. This action pulls the lifting mechanism of the main or pilot disc thereby opening the valve to allow steam to discharge through the SRV until the inlet pressure is near or equal to zero. For overpressure relief valve operation (power-actuated mode), pressure sensors on the RPV genemte a RPV high pressure trip signal which is used to initiate opening the SRVs. When the set pressure is reached, the SRV power-actuated relief solenoid is energized, which admits pneumatic pressure to the SRV actuator, thereby opening the SRV. The SRV pneumatic operator is so arranged that, ifit malfunctions, it wdl not prevent the SRV from opening when steam inlet pressure reaches the spring lift setpoint. (3) Depressurization operation: The Automatic Depressurization System (ADS) valves open automatically as part of the Emergency Core Cooling System (ECCS) for events invohing small breaks in the nuclear system process barrier. Eight of the eighteen SRVs are designated as ADS valves and are capable of operating from either ADS logic or safety / relief logic signals. 2.1.2 6/16/92

ABWR D3 sign Docum:nt Automatic depressurization by the AD is provided to reduce the hq reactor pressure during a LOCA in which the High Pressure Core Flooder (HPCF) System and/or the Reactor Core Isolation Cooling (RCIC) System are unable to restore water level. This allows makeup of core cooling water by the low pressure makeup system, the Low Pressm e Flooder (LPFL) Mode of the Residual Heat Removal (RHR) System. The ADS consists of redundant trip channeh arranged in two separated logics that control two separate solenoid-operated gas pilots, ADS I and ADS 2, on each ADS SRV. Either pilot can operate the ADSicalve. These pilots control the pneumatic pressure applied by the accumulators and the High Pressure Nitrogen Gas Supply (HPIN) System. The power for instrumentation and logic is obtained from the Safi ty System Logic and Control (SSLC) Division I and II. Sensors from all four divisions and Division I control logic for low reactor water icvel and high dqwell pressure initiate ADS 1 pilots, and sensors from all four divisions and Division 11 initiate AT " pilots, either of which willinitiate the opening of the ADS SR\\ ' The reactor vessel low water level initiation setting for ADS is pre-selected to depressurize the reactor vesselin time to allow adequate cooling of the fuel by the network of ECCS follcwing a LOCA. Timely s depressurization of the reactor vessel is provided if the reactor water level drops below preset limits together with an indication that high drywell pressure has occurred, which signifies there is a loss of coolant into the containment with insufficient high pressure makeup to maintain reactor water level. For breaks outside the containment, timely depressurization of the reactor vesselis provided if the reactor water level drops below preset limits for a time period sufficient for the ADS high drywell pressure bypass timer and the ADS timer to time-out. All SRVs have indhidual non-safety related accumulators. In addition, those with ADS function have a separate safety-related larger capacity accumulators with sepamte redundant gas power actuators. The ADS accumulators are sized to operate the SRV two times with the dqvell i pressure at 70% of design gauge pressure following failure of the pneumatic supply to the accumulator. The SRVs can be operated individually in the power-actuated mode by remote manual switches located in the main control room. (b 2.1.2 6/16/92

r 1 ABWR Design Document NBS Instrumentation The purpose of the Nits RPV instnunentation is to monitor and provide control nput for operation variables during plant operation. The Nils contains the instnamentation for monitoring the reactor piessure, metal temperature, and water level. The reactor pressure and water level instnunents are used by multiple lioiling Water Reactor (llWR) systems, both safety related and non-safety related. Pr essure indicators and transmitters detect s cactor vesselinternal pressure from the same instnunent lines used for measuring reactor vessel water level. The RPV coolant tempemtures are determined by measuring saturation pressure (which gives the saturation temperature), outlet now tempemture to the Reactor Water Cleanup (CUW), and the RPV bottom head drain line temperature. The reactor vessel outside surface (metal) temper ~,ure is measured at the head flange and the bottom head locations. Temperatures needed for operation and for operating limits are obtained from these measurements: During normal operation, either reactor steam saturation tempemture and/or inlet temperatures of the reactor coolant to the CUW System and the RPV bottom head drain can be used determine the RPV coolant temperature. Figure 2.1.2e shows the water level and RPV penetrations for each water level mnge. The instmments that sense the water level are all differential pressure devices calibrated for a specific RPV pressure (and corresponding liquid temperature) conditions. The water level measurement design is the condensate reference chamber type. Instrument zero for all the RPV water level ranges is the top of the active fuel. The following is a description of each water level range shown on Figure 2.1.2e. .) Shutdown Range Water Level. This range is used to monitor the reactor water level during shutdown condition when the reactor system is Gooded for maintenance and head removal. The two RPV instrument penetadons elevations used for this water level measurement are located at the top of the RPV head and the instrument tapjust below the dryer skirt. (2) Narrow Range Water Level. This mnge is used to monitor reactor water level during normai power operation. This range uses the RPV taps at the elevations nur the top of the steam outlet nozzles and the taps at the elevation near t% bottom 2.1.2 6/16/92

ABWR D: sign Document g of the dryer skirt. The Feedwater Control (FDWC) System uses this (. range for its water level control and indication inputs, (3) Wide Range Water Level. This r;mge is used to monitor reactor water vel for evene where the water level excee<ls the range of the narrow n.nge water level instnamentation, and is used to generate the low reactor water level trip signals which indicate a potential LOCA. This range uses the RPV taps at the elevations near the top of the steam ou;let nozzles and the tap below the Top of the Active Puel (TAF). (4) Fuel Zone Range Water Level. This range is provided for the post accident monitoring, and prov. des the capability to monitor the re: ctor water level below the wide range water level instrumentation. This range uses the RPV taps at the elevations near the top of the steam < utlet nozzles and the taps 1 elow the TAF (above pump deck). The NBS con *ains the instnament lines to monitor the differential pressure across the P.PV pump deck and core support plate. The instrumentatic - "hich actually performs these functions is located within the Recirculation Flow Contml (RFC) System. The SRVs are provided with position sensors which provide positive indication of SRV disk / sten. position. Thermocouples are located in the discharge exhaust pipe of the SRVs. The temperature signal goes to a multipoint recorder with an alann and will be activated by any temperature in excess of a set temperature signaling that one of the SRV seats has started to leak. The NBS also contains the drywell pressure instmmentation used to generate the safety-related high drywell pressure trip LOCA signal, which is used by many of the safety-related systems to initiate safety actions. The Reactor Protection System (RPS) utilizes this signal as a scram initiation signal. The Leak Detection and Isolation Systen. 9 utilizes this signal to initiate containment isolation. The Emergency Core x>h. ' Systems (ECCSs) utilizes this signal as a system initiation signal. Control room indica

• ion and/or alarms are provided for the important plant parameters monitored by the NBS.

(D %.) 2.1.2 6/16/92

ABWR oesign Document AGME Code Requirements O The major mechanical components me designed to meet American Society of hiechanical Engineers (AShf E) Code Re<1uirements as shown below: Component AShfE Desigu Conditions Code Class Pressure Temperature 2 FW lines from the hiOVs to 2 87.9 kg/cm g 302"C the outboard containment (1250 psig) (571"F) isolation check valves 2 RV lines from the outboard 1 87,9 kg/cm g 302 O containment isolation check (1250 psig) (575"l') valve to the RPV 2 Feedwater (FW) line 1 87.9 kg/cm g 302"C outboard containment (1250 psig) (575";, isolation check valve 2 Main Steam Isolation Valves 1 96.7 kg/cm g SOS"C (MSIVs) (1375 psig) (586. F) 2 Safety / Relief Valves (SRVs) 1 96.7 kg/cm g 308 C (' 375 prigi (586."F) 2 Main Steam Lines (MSLs). 1 87.9 kg/cm g 302 C from Reactor Pressure Vessel (1250 psig) (575"F) (RPV) to outboard MSIVs 2 MSLs from the outboard 2 87.9 kg/cm g 302"C MSIVs to the seismic (1250 psig) (575"F) interface restraint 2 SRV discharge line piping, 3 38.0 kg/cm g 250"C from the SRVs to the (540 psig) (482"F) diaphragm floor 2 SRV discharge line piping, 2 35.0 kg/cm g 250 C from the diaphragm floor to (540 psig) (482 F) the suppression pool sud' ace Inspections, Tests, Analyses and Acceptance Criteria Table 2.1.2b provides a de6nition of the inspections, tests and/or analyses together with associated acceptance criteria which will be undertaken for the NDS. I O i 1 2.1.2 6/16/92 l 1

ADWR Design Document Table 2.1.2a: Nuclear System Safety / Relief Valve Setpoints Set Pressures and Capacities ASME Rated Capacity 4 Spring Set at 103% Spring Set 2 Number' of Valves Pressure (kg/cm g) Pressure (kg/hr each) 1 80.8 395,000 1 80.8 395,000 4 81.5 399,000 4 82.2 402,000 4 82.9 406,000 4 83.C 409,000

• Eight of the SRVs serve in the automatic depressurization function.

4 1 2.1.2 6/16/92 r

} Taole 2.1.2b: Nuclear Boiler System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Anslyses Acceptance Criteria 1. A simplified cor* figuration of the Main 1. Visual field inspection will be conducted to

1. The system configuration is in accordance Steam Lines (MSLs), and Feedwater (FW) confirm that the installed equipment is in with Figures 2.1.2a,2.1.2b, and 2.1.2c.

lines within the Nuclear Boiler System compliance with the design configuration (NBS) scope, and the Safety / Relief Valves defined in Figures 2.1.2a,2.1.2b, and 2.1.2c. (SRVs) and the Safety / Relief Valve (SRV) discharge lines, as described in Section 2.1.2 and shown in Figures 2.1.2a,2.1.2b, and 2.1.2c. 2. The Reactor Coolant Pressure Boundary 2. Inspections will be conducted of ASME 2. The components have appropriate ASME (RCPB) portions of the NBS are classified Code required documents and the Code Code, Section !!!, Class 1 certifications and as American Society of Mechanical stamp on the actual components to verify Code Stamps Engineers (ASME) Code Class 1. They are that they have been manufactured per the designed, fabricated, examined and relevant ASME nequirements. 9 hydrotested per the rules of the ASME Code, Section 111. This includes the MSLs from the Reactor Pressure Vessel (RPV) to ar'd including the outboard Main Steam Isolation Valves (MSIVs), the FW lines from the outboard positive closing check valves to the RPV. 3. Each Main Steam Line (MSL) shall have a 3. Using the as-built dimensions, perform an 3. Analysis confirms that the MSL flow flow limiter located in the RPV MSL outlet analysis which shows that the MSL flow limiters perform their intended function. nozzle. The MSL flow limiter shall limit the limiters satisfy the requirement. coolant blowdown rate from the RPV in the event of a MSL break to a (choke) flow rate equal to. or less than 200% of rated s'eam 2 flow at 72.1 kg/cm g upstream pressure. 1 4. Each MSL flow iimiter has taps for two 4. Visual inspection will be conducted to 4. Inspection confirm that the MSL flow 3 instrument lines. Thr. se instrument lines confirm that the MSL instrument lines have instrument lines have been installed. are used for monitoring the flow through been installed in compliance v ith design each MSL commitment. O O O

_.... _ _.. - -... -. - - -. - ~.. - - _. - - - - -. -. _ - ~ O [ Table 2.1.2b: Nuclear Boiler System (Continued) b Inspections, Tests, Analyses and Acceptance Criteria Cutified Design Commitment inspections, Tests, Analyses Acceptance Criteria 5. The total steam line volume from the RPV 5. Using the as designed configuration of the 5. Calculations confirm that the steam line [ L to the main steam turbine stop valves and steam lines perform calculations to volume satisfies the design requirement. - l ? - steam bypass valves shall be greater than determine the main steam line volume. 3 or equal to 113.2 m. I

6. The MSlVs meet the requirements of 6.

Inspections will be conducted of ASME 6. The MSIVs have appropriate ASME Code, I ASME Code, Section Ill. Code required documents and the Code Section Ill, Class 1 certifications and code i Stamp on the actual components to verify stamps. that they have been manufactured perthe - f relevant ASME requirements. i 1 . 7. The Main Steam isolation Valve (MSIV) 7. Pre-operational tests will be conducted to 7. Pre-operational tests confirms that the I 4 i closing time shall be between 3 and 4.5 demonstrate proper operation of the MSIVs satisfy the closure time i seconds when N or air is admitted into the MSIVs, including verification of the closure requirement. 2 3 f valve pneumatic actuator. time. 9 I 8. The SRVs meet the requirements of ASME 8. Inspections will be conducted of ASME

8. The SRV have appropriate ASME Code, Code, Section 111.

Code required documents and the Code Section 111, Class 1 certifications and code i Stamp on the actual components to verify stamps. l that they have been manufactured per the f: relevant ASME requirements. l

9. There shall be 18 SRVs mounted on the 9.

Inspections will be conducted to confirm 9. Inspections confirm that the SRVs have the f {' MSLs s's shown in Figure 2.1.2a. The that the SRVs have the required (nominal) required capacities and set pressures required spring set pressure and capacities spring set pressure and (minimum) identified on their name plates. i are given in Table 2.1.2a. The SRVs shall capacity on the SRV ' nameplate. i meet the opening performance shown in inspections confirm that the proper ' [ Figure 2.1.2f. Visual inspections will be conducted to capacity and set pressure SRV has been confirm that all 18 SRVs have been mounted in its correct location. j jl insta!!ed in their proper locations. i i" Confirm that the selected SRV model j. Review of the qualification test data for the satisfies the performance requirements.- particular SRV model selected to confirm l 3 that the opening performance complies l 3 with the requirements. 1 i k w

t Table 2.1.2b: Nuclear Boiler System (Continued) a inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

10. The SRVs shall be provided with

10. Inspection will be performed that the SRVs

10. Inspection confirms that the SRVs have instrumentation which will prcvide positive have positive position indication positive position indication.

indication (i.e. by direct measurements of instrumentation, and that the SRV position. instrume ntation has been properly connected.

11. A simplified configuration of the Automatic

11. Visual field inspection will be conducted to

11. The configuration is in Eccordance with Depressurization System (ADS) SRVs and confirm that the installed ;quipment is in Figure 2.1.2d.

the non-ADS SRVs as described in Section compliance with Figure 2.1.2d. 2.1.2 and Figure 2.1.2.d. There are 8 ADS SRVs and 18 non-ADS SRVs.

12. Upon receipt of either a high drywell

12. Logic and instrument functional testing

12. The drywell pressure and RPV water level pressure trip signal current with a RPV low shall be performed to c'emonstrate that the instrumentation, as well as the ADS logic, water level 1 trip signal of sufficient ADS logic performs as required.

functions as required to generate the ADS Y duration for the ADS timer to time-out, or a initiation signal. RPV low water level 1 trip signal of sufficient duration for the ADS high drywell pressure bypass timer and the ADS timer to time-out,the ADS logic generates a ADS initiation signal to the SRV ADS solenoids.

13. The SRV discharge lines shall terminate at

13. Visual inspections will confirm

• hat the SRV

13. Inspection confirms that the SRV discharge the quenchers located below the surface of discharge line quenchers have been line quenchers have been installed.

the suppression pool. installed.

14. The RPV shall be provided with instrument

14. Visual inspections will be performed to

14. Inspection confirms that the lines and instrumentation necessary to confirm that the instrument lines and instrumentation has been properly monitor the RPV steam dome pressure and instrumentation for the RPV steam dome installed.

the RPV water level from the Bottom of the pressure, the RPV shutdown renge water Active Fuel (BAF) to top of the steam dome. level, the RPV narrow range water level, the RPV wide range water level, and the RPV fuel zone range water level sensors g has been properly installed. B

15. For the safety related NBS instrumentation,

15. Instrument functional testing shall be

15. The instrumentation functions as required.

the instrumentation t e capable of performed to demonstrate that the performing its necessary function. instrumentation performs as required. O O O

O O O t Table 2.1.2b: Nuclear Boiler System (Continued) b inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

16. Control room indication / alarms are

16. Inspection shall be performed which

16. Inspection confirms that the important provided for the important plant confirms that the important plant plant parameters have been indicated and/

parameters monitored by the NBS. parameters monitored by the NBS are or alarmed in the main control room. indicated and/or alarmed in the main I control room. 4 - 1. l Y 4 4 L 4 I 4 I J 0 5 1 1 1 I

ABWR Design Document O MAIN STEAM LINES + + + + l l ASME l l OUTBOARD l l CODE l l MAIN STEAM l l CLASS 21 I ISOLATION x I l A l j VALVE Y CONTAINMENT ASME WALL l C DECLASS 1l / l t l l l l / i i i i INBCARD DRYWELL MAIN STEAM ISOLATION VALVE i i i i k i i / i i 's / N / 's SRV ~~~~ SRV SRV SRV SRV SRV SRV SRV ~MSE ... L _ _ - MS REACTOR VESSEL MS,h _ J,SL SRV x 01 Figure 2.1.2a Safety / Relief Valves and Steamline 2.1.2 6/16/92 l

fN A ew f( 3 SRV !NBOARD MAIN (TYPICAL; STEAM ISOLATlON = VALVE SEISMIC NON-SEISMIC r f CATEGORYI ~ CATEGORYI ~ i 1 2 I TO MAIN \\ TURBINE + STEAM MAIN STEAM REACTOR LINE A FROM SEISMIC OUTBOARD MAIN VESSEL I FROM STEAMLINES INTERFACE STEAM ISOLATION l B,C &D l STEAMLINES VALVE l l l l B.C &D Jjl l l 1I I-M ~ L I l l 2lNC l l t i_v v v _ __L_ 4 _! s" i i i DRAIN LINE E + TO MAIN CONDENSER s 2 DRAIN LINE ASME NON-CODE CODE r I CLASS 1 l WETWELL l 2 I M SUPPRESSION POOL

• AT MINIMUM WATER LEVEL Ee Figure 2.1.2b Steamline

ABWR Design Document h FEEDWATER LINES V 1r // NN / \\ SEISMIC INTERFACE \\ / NC l l FROM RCIC, l l FROM RHR, ASME CUW, AND _ _, _ _ Q 4_ _ _ _ CUW, AND CODE CRD l I CRD l CLASS 2 l I I A 1r CONTAINMENT ASME WALL / CODE e cass, i l i i X X / \\ k f s 3 REACTOR VESSEL / 's Figure 2.1.2c Feedwater Line 2.1.2 -1 G-6/16/92 I

ABWR oesign Document O HPIN NBS NBS. HPIN =N : " /t ,~ RELIEF ACCUMULATOR ADS ACCUMULATOR. RV ADS 18 ADS 2 8 A& ' h h A w w I i l i i l l I I 'FROM FROM FROM POWER'- ADS ADS ACTUATED DIV.1 DIV.11 RELIEF LOGIC LOGIC LOGIC (( JSRV O ADS SRV HPIN NBS ,/ N, C T J s RELIEF RV ACCUMULATOR A 1 I.;. FROM J: POWER ig ACTUATED TSRV RELIEF LOGIC NON-ADS SRV ' O ~ Figure 2.1.2d Safety / Relief Valve Pneumatic Lines 2.1.2 17 6/16/92

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r + e '11 1 0 6' l C4 c-5 - l / E i i / I E 4 e n 1 .F \\ m ~ s g; LI M ,5E< O n SHUTDOWN RANGE j HRRMbRRRRRRRbWARRRMRtWRWABA: o C NARROW RANGE E WA4%G%%wA% m-o >Z ,z 3 Th N m o m c Oc E 4I WIDE RANGE Om m m +7/fffff/jf/jgy/ffgggf%g(9 zOm FUEL ZONE RANGE A XVZff/#/##/A%%%%%/'4 za's ~ i O O O E

I ABWR oesign occument oU O A 100 l l 5 SAFETY VALVE OPENING l CHARACTFAISTICS g 2 I B l 5 e I b O j l >[ l VALVE 4 STROKE TIME l 0 t 4 0.3 b TIME (sec) 13. TIME AT WrilCH PRESSURE EXCEEDS THE VALVE SET PRESSURE O Figure 2.1.2f Safety-Action Valve Lift Characteristics 2.1.2 6/16/92

ABWR oesign t;sument ( 2.2.8 Recirculation Flow Control System \\ Design Description The Recirculation Flow Control (RFC) System controls reactor power by controlling the recirculation Gow rate of the reactor core water. Reactor recirculation now and core flow is varied by modulating the Recirculation Internal Pump (RIP) speeds / flows through the voltage and frequency modulation of adjustable speed drive outputs. Refer to Figure 2.2.8. The RFC System consists of the triplicated process controller, solid state Adjustable Speed Drives (ASDs), switches, sensors, and alarm devices provided for operational manipulation of the ten RIPS and the smveillance of associated equipment. Recirculation flow control is achieved either by manual operation, or by automatic operation if the power level is above approximately 70% of 4 rated. The reactor internal pumps can be driven to operate anywhere between minimum speed and 100% of rated speed with the variable voltage, vadable frequency power source supplied N he ASDs. t This system is a power generation system and is classified as non-safety-related. The RFC System is designed to allow both automatic and manual operation. In the automatic mode called " Master Auto" mode

• Automatic Load Following (ALF) operation the master controller generates a demand signal for balancing out the load demand error to zero. This demand signalis forwarded to the flow controller which generates a Ouw demand signal. The Dow demand signal is adjusted by a Dow demand set down function to lower the recirculation Gow when the sensed reactor flux is above 105%. The speed controllers in the ASDs generate speed demand based on the Dow demand from the Dow controller. The speed demand causes adjustment of RIP motor power input which changes the opemting speed of the RIP and hence core flow and core power. This process continues until both the errors existing at the input of the flow controller and master controller are driven to zero. The flow controller can remain in automatic even though the master controller is in manual.

The reactor power change resulting from the change in recirculation flow causes the pressure regulator to reposition the turbine control valves. If the original demand signal was a load / speed error signal, the turbine responds to the change in reactor power level by adjusting the control valves, and hence its power output, until the load / speed error signal is reduced to zero. In the semi-automatic mode, the operator sets the total core flow demand and the RFC System responds to maintain constant core How. Core flow control is achieved by comparing the core flow feedback. which is calculated from the core plate dif ferential pressure signals, with the operator supplied core flow set point. 2.2.8 6/1/92

l ABWR 0: sign Document In total manui control, the operator can directly manipulate the RIP speeds. Pump speeds can be controlled individually or collectively. When individually controlled, pump speed demand is obtained through the operator console and transmitted directly to the individual ASD for pump frequency control. In colleuive manual op-mtion, a common speed set point is used for controlling each IUP which has been placed in the GANG speed control mode. The recirculation flow control system is also used to control the start up of the reactor internal pumps. To minimize thennal shock to the reactor vessel, the RFC System will prevent start up of an idle RIP if the temperature difference of l the vessel bottom coolant to the saturated water tempervure corresponding to the steam dome pressure is above a predetennined v-lue. In the event of either (a) turbine trip or generator load rejection above a predetermined reactor power level, (b) reactor pressure exceeds the high dome pressure trip set point, or (c) reactor water level drops below the 1.evel 3 set point, logic will automatically be initiated to trip off a group of four RIPS. ASDs are used to provide electrical power and speed control to the pump motors in the IUP3 The ASD receives electrical power from a power plant bus at a constant AC voltzge and frequency. The ASD converts this to a variable frequency and voltage in accordance with the speed demand requested by the RFC System controller. The ASD is capable of supporting three modes of operation: start up, normal, and shutdewn. When the start up mode is selected, the invener output quickly steps up from zero to the required motor power corresponding to the minimum pump speed and holds at that output frequency. When the nonnal operation mode is selected, continuous output power frequency between minimum speed and 100% is allowed. The operation of the shutdown mode is exactly reverse that of the nonnal and start up mode; ASD output is automatically ramped to minimum speed frequency, then stepped down to zero. l The RFC System control functional logic is performed by a triply redundant, microprocessor based fault tolerant digital controller (FFDC). The FTDC consists of three identical processing channels working in parallel to provide fault tolerant operation. The RFC System design consists of two main controlloops, (1) the core flow loop, which modulates pump speed demand to provide the desired core Dow mte, and (2) the automatic load fo"owing (ALF) which modulates the core now demand in response to the load demand error. In addition, pump speed in each l RIP can be manually controlled individually or collectively. In *he tore How control mode, sensed core Dow calculated by the core plate differential pressure method is compared with the core Dow demand supplied 2.2.8 6/1/92

ABWR D: sign Docum:nt by the operator or obtained from the master controller, depending on the RFC 'v system operating mode. This flow error is input to the core flow controller to drhc pump speed demand. In ALD mode, the master controller receives a load demand signal from the steam bypass and pressure control (SP&PC) system in response to any combination oflocal opemtor load set point inputs, automatic genemtion control inputs, or grid load changes indicated by grid frequency variation. When in local control, the operator's control panel provides the operator the capability to select the operating mode of the system and to initiate certain manual actions. Indications and alarms are provided to keep the operator infonned of the system opemtional modes and equipment status, thereby allowing him to quickly determine the origin of any abnonnal conditions. Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.8 provides a definition of the inspections, tests, and/or analyses, together with the associated acceptance criteria, which will be undenaken for the RFC. O o 'd 2.2.8 W1/92

cc Table 2.2.8a: Recirculation Flow Control System Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Test, Analyses Acceptance Criteria 1. The configuration of the RFC system is 1. Inspections of the as-built RFC system shall 1. Actual RFC system configuration, for those shown in Figure 2.2.8-1. be performed components shown, conforms with Figure 2.2.8-1. 2. Reactor internal pumps (RIPS) will operate 2. Operation of the pumps at any speed 2. Pumps shall operator witt.in design at any speed between minimum spc7d and between minimum speed and 100% of specification limits at any speed between 100% of rated speed. rated shall be performed. 30% and 100% of rated. 3. The RF0 System may be operatect :n 3. The RFC System shall be operatec' at 3. The RFC System shall operate in automatic automatic mode above approximately power leve!3 greater than approximately mode within design specification limits at 70% of rated power. 70% rated. any power level above approximately 70% rated. 4. The RFC System shall be used t. .ir"I 4. The RFC System shall bo operated in the, 4. The RFC System shall operate within the the start up or shut down of the t, tart up and shutdown modes. design specification limits in the start up and shutdown modes. The pump shall be ramped from 0% to 30% and held and then shall be stepped down to 0% 5. The RFC System shall be interlocked so as 5. The RFC System shall be operated so as to 5. Th3 RFC System shall prevent start up of to prevent start up of a idle RIP if the vewel start up an idle RIP wh en the vessel bottom an idle RIP under conditions specified in bottom temperature is not within 144'F of temperature is not within 144'F of the the design specification, the saturated dome pressure equivalent saturated dome pressure temperature temperature, equivalent. 6. A select group of RIPS shall trip off in the 6. The RFC System shall be operated and the 6. The RFC System shall operate within event of either (a) turbine trip or generator following events shall be simulated; design specification limits under all load rejectior', (b reactor pressure exceeds simulated fault conditions. gp high dome pressure trep set point, or (b) generatt-load rejection (c) reactor level drops below Level 3. M re m e pressure (d) low vessel level g w O O

ABWR Design Docum:nt n'w) MASTER AUTO (ALF) FLOW M 'A GANG SDEE[ M/A ( ( l ) ) I PUMP PUMP PUMP PUMP SPEED A SPEEDO SPEEDC SPEED K M/A M/A M/A M/A G A&D A&D A&D A&D x& V PUMP SPEED CONTROL PUMP MANUAL > GANG > AUTO-FLOW CONTROL > MASTER AUTO NOTES:

1. ONLY FOUR OF TEN PUMPS ARE USED FOR ILLUSTRATION

2. > IS A RELATION MEANING *OVER-RIDE

• y~

k i Figure 2.2.8 Recirculation Flow Control System 2.2.8 5 6/1/92

ABWR ossign Document 2,4 Core Cooling -O 2.4.1 Residual Heat Removal System Design Description The Residual Heat Removal (RHR) System is comprised of three divisionally-separate subsystems that perform a variety offtmctions utilizing the following six basic modes of opemtion: (1) shutdown cooling. (2) suppression pool cooling, (3) wetwell and diywell spmy cooling, (4) low pressure core flooder (LPFL), (5) fuel pool cooling, and (6) ACindependent water addition. The configuration of - each loop is shown on its P&lD in Figure 2.4.1 (aligned in the standby _ mode). The major functions of the various modes ofoperation include: (1) containment heat removal, (2) reactor decay heat removal, (3) emergency reactor vessel level makeup and (4) augmented fuel pool cooling. In line with its given functions, portions of the system are a part of the ECCS network and the containment cooling system. Additionally, portions of the RHR System are considered a pan of the Reactor Coolant Pressure Boundary (RCPB). The entire RHR System is designed to safety-related standards, although it - performs some non-safety functions (i.e., those that are not taken credit for when evaluating design basis accidents). The safety-related modes of operation include: (1) low pressure flooding, (2) suppression pool cooling, (3) wetwell spray cooling and (4) shutdown cooling. Non safety-related modes of operation include: (1)drywell spmy cooling, (2) AC independent water addition and (3) augmented fuel pool cooling. The RHR System also provides a backup, safety-related fuel pool makeup capability, Ancillary modes of operation include minimum flow bypass and full flow testing. The ECCS function of the RHR System is performed by the LPFL mode. Following receipt of a LOCA signal ( low reactor water level or high drywell pressure ), the RHR System automatically initiates and operates in the LPFL mode (in conjunction with the remainder of the ECCS network) to provideL emergency makeup to the reactor vessel in order to keep the reactor core cooled such that the criteria of 10 CFR 50.46 are met. The LPFL mode is accomplished by all three loops of the RHR System by transferring water from the suppression pool to the RPV, via the RHR heat exchangers. Although the LPFL mode is - automatically initiated, it may also be initiated manually. The system will also automatically revert to the LPFL mode of operation from any other test or - operating mode upon receipt of a LOCA signal. Each RHR loop's RPV injection valve requires a low reactor pressure permissive signal whether being opened manually or automatically in response to a LOCA signal. The containment heat removal ftmction in the ABWR is performed by the Containment Cooling System, which is comprised of the low pressure core flooder (LPFL), suppression pool cooling, and wetwell and drywell spmy cooling - 2.4 1-6/16/92. 1

ABWR Design Document modes of the RHR System. Following ; LOCA, the energy present vcithin the teactor prirr an system is dumped citt r directly to the suppression pool via the SRVs, or indirecdy via the dnwell am onnecting vents. Subsequently, nssion product decay heat continues to add nergy to the pool. The Containment Cooling System is designed to limit - e long-term bulk tempemture of the suppression pool, and thus limit the ong-term peak temperatures and pressures within the wetwed and drywell regions of the containment to within their analyzed design limits, with only two of the three loops in operation (i.e., worse case sing!c failure). The cooling requirements of the containment cooling function establish the necessary RHR heat exchanger heat removal capacity. The LPFL mode, in addition to its priman function of cooling the core, senes to cool the containment, as the heat exchanger is designed to always be in the loop. The dedicated suppression pool cooling mode is made available in each of the three loops of the RHR System by circulating suppression pool water through the respective RHR heat exchanger and then directly back to the suppression pool. This mode of RHR is usually initiated manually but will also initiate automatically in response to high suppression pool temperature. The wetwell and dowell spray modes of RHR are each available in only two of the three subsystems (loops B and C). These functions are performed by drawing water from the suppression pool and delivering it to a common wetwell spray header and/or a common dqwell spray header, both via the associated RHR g heat achanger(s). These containment spray modes of the RHR System are typically initiated manually, with the exception of automatic initiation of wetwell spray coincident with automatic suppression pool cooling. However, the dqwell spray inlet valves can only be opened if there exists high drywell pressure and if the RPV injection valves are fully closed. Wetwell and drywell sprays sene as an augmented method of containment cooling. Wetwell spray also serves to mitigate the consequences of steam bypassing the suppression pool. The normal operational mode of the RHR System is in the shutdown cooling mode of operation, which is used to remove decay heat from the reactor core. This mode prxides the required safety-related capability needed to achieve and maintain a cold shutdown condition, including consideration of the worst case system single failure. The RHR heat exchanger heat removal capacity requirements in this mode are botmded by containment cooling requirements. Shutdown cooling is initiated manually once the RPV has been depressurized below the system low pressure permissive. In this mode och loop takes suction from the RPV via its dedicated suction line, pumps the water through its respective heat exchange i mturas the cooled water to tlm RPV. Two loops (B and C) discharge wate' < + +7 the RPV via dedicated spargers, while the third loop (A) utilizes the ael spargers of one of the two feedwater lines (FW-A). The heat removed in the RHR heat exchangers is transported to the ultimate l heat sink via the respective division of reactor cooling water and service water. i Each shutdown coohng suction valve is interlocked with that loop's suppression j 2.4.1 42-6/16/92

? ABWR o, sign Docum:nt pool suction and discharge valves and wetwell spray valve to prevent draining of O the reactor vessel to the suppression pool. Also, cac h shutdown cooling suction valve is interlocked with, and automatically closes on. low reactor water level. 'Ihe augmented fuel pool cooling mode of the RilR System supplements / replaces the normal fuel pool cooling system during infrequent conditions of high heat load. This mode is accomplished manually in one of two ways. When the reactor vessel head is removed, the cavity 00mled and the fuel pool gates are amoved, the RilR System cools the foci poolin the normal sht tdown cooling roode. When the iuel pool is otherwise isolated from the s cactor cavity, two loops (15 and C) af the RHR System can directly cool the pool by taking suction from and discharging back to the nonnal fuel pool cooling system. This connection also provides for emergency fuel pool makeup capability by supplying a safety-related makeup path to the f uel pool from a safety-related source (i.e., the suppression pool). One loop (C) of the RilR System also functions in m AC independent water addition mode. This mode provides a means of cross connecting the scactor building Ore protection system header to the RHR Systemjust outside the containment in the absence of the normal ECCS network and independent of the normal essential AC power distribution network. The connection is accompM.hed by manually opening two in-series valves on the cross-connection D pipingjust upstream ofits tie-in to the normal RHR piping. Fire nrotection system water can be dir cted to either the RPV or the drywell spmy sparger by manual opening of the loop C RHR injection valve or the two loop C drywell spray valves. These three valves also have manual hand wheels. The fire water is supplied via the system's reactor building distribution header by either the dir ect diesel-driven fire pump or from an external source utilizing a dedicated connectionjust outside the reactor building. Each loop of the RHR System also has both a minimum Dow mode and a full Dow test mode. The minimum flow mode assmes that there ' pump 00w sufficient to keep the pump cool by opening a minimum fitw valve tnat directs now back to the suppression pool anytime the pump is runn% and the main discharge valve is closed. Upon sensing that there is adequate now in the pump main discharge line, the minimum now valve is autt.matically closed, in the full now test mode, the system is essentially operated in the suppression pool cooling t a ode, drawing suction from and discharging back to the iuppression pool. a The RHR System is comprised of three separav :oeps or subsystems, each of which includes c nump and a heat exchang, r, takes suction frora cither the RPV or the suppresso a pool, and directs water back to either the RPV or the separation supr usion pool. Two of the three loops can divert a portion of the suppm Oan pool return How to a common wetwell spmy sparger or direct the 3 entire t' low to a common drywell spray sparger. The divisional subsystems of the 2.4.1 3-6/16/92 l i

ABWR 0: sign oocum:nt RilR System are separated both mechanically and electrically, as well as being physically located in diffeient ancas of the plant to addiess icquirenu nts peitaining to fire piotection and other s.epamtion criteria. Each of the thice subsystems is poweird from a sepamte divisional power distilhution bus that can be supplied 60m either an on-site or off site source. Cooling water to each division of RilR equipment (heat exchanger as well as pump and motor cooleis) is supplied by the respective division of the scactor cooling water (RCW) System. The RilR System also includes provisions for containment isolation and RCPli piessur e isolation. The RilR System will maintain the capability to perfonn its intended safety-related functions either folknving a Safe Shutdown Earthquake (SSE) or during the emironmental conditions imposed by a 1.OCA, and in each case assurning the worst case single failure. The system will also acconunodate calculated movement and thennal stresses. The system is designed so that the pumps will have necessary head / flow ci,aracteristics and available NPSil greater than required NPSil for opemting modes. The system can be powered from either nonnal off-site sources or by the emergency diesel generators. The RiiR System is Seismic Category I and is housed in the Seismic Category I scactor buihling to provide protection against tornadoes, flocxis, and other natural phenomena. The RiiR pumps are motor-driven centrifugal pumps each capable of supplying g at least 4200 gpm at 40 psid (drywell to RPV). The pumps are AShiE Code Class 2 components with a design pressure of 500 psig and a design tempemture of 360 F. The pumps are interlocked from starting without an open suction path. The RiiR pumps are protected from possible pump runcut conditions during operation. The RHR heat exchangers are horizontal U-tubc/shell type each sired to provide a minimum effective heat removal capacity (R<oeflicient) of 195 litu/sec F. The primary and secondary sides of the heat exchangers are ash 1E Ccxte Class 2 and 3, respectively. The primary side design tempemture and pressure are 500 psig and 360*F, r espectively. The secon6:7 side design tempemture and pressure are consistent with that of the RCW System. Each h>op of the RHR System has its ownjockey pump to act as a 1 e[> fill system for that k>op's pump discharge piping. Thejockey pumps are AShiE Code Class 2. The RHR System piping und valves are AShiE Code Class 1 or 2 as shown on the PMD (Figures 2.4.la, b, c). The design pressure and temperature of piping and valves varies across the system. For that piping attached to the RPV, from the RPV j out to and including the outboard containment isolation valves, the design pressure and temperature are 1250 psig and 575"F, respectively. For other piping open to the containment atmosphere, out to and including the outboard containment isolation valves, the design pressure and temperature are 45 psig and 219*F, respectively. For piping and valves outside the containment isolation valvc., the design pressure and temperature depends on whether it is located on the suction or discharge side of the main pump. Those portions on the suction 2.4.1 6/16/92

ABWR Dasign Documu.n sile aie rated at 300 psig and 360 F, while those poitions on the discharge side (q ne rated at MK) psig and 360*F, respectively. The low pressuie ;xntions of the ) .hutdown cooling piping are piotected from f all icactor pressure by automatic pressure isolation vahes that are inte locked with reactor pressure. Iligh seliability of this inteik>ck is assured by utilizing four separate and divisionally independent piessure sensors in a 2 out of-1 logic. Additionally, in-series inboard and outboard contaimnent/ pressure isolation valves in each loop are poweird from sepamte electrical divisions. Relief valves are also provided for protection from overpressure. The RHR System includes control room indication to allew for monitoring and control during design basis operational conditions, i.e., system flows, temperatures and piessures, as well as valve open/close and pump on/oll indication for those instnnnents and components shown on Figures 2A.la, b auul c, with the exception of simple check valves and overpressure relief v;dves (of the check valves shown only the testable check valves downstream of each loop's Rl'V injection valve has conto room status indication). Inspections, Tests, Analyses and Acceptance Criteria This section provides a definition of the inspections, tests and/or analyses together with associated acceptance criteria which will be undertaken for the -( RHR System. ( k 2.4.1 Ov16/92

{ Table 2.4.1: Residual Heat Removal System inspections, Tests, Analyses and Acceptance Criteria Certified Desig.n Commitment Inspections, Tests, Analyses Acceptance Criteria 1. The configuration of the RHR System is 1. Inspections of the as-built RHR 1. Actual RHM System configuration, for shown in Figures 2.4.la, b and c, which are configuration shall be performed. those components shown, conforms with each mechanically and electrically Figures 2.4.1a, b and c and separation separated from each other. requirements. 2. The RHR System operates in the LPFL 2. The ECCS LOCA performance analysis for 2. RHH System actuation and operation is mode as part of the overall ECCS network. assuring core cooling shall be validated by consistent with the ECCS performance RHR System funct%nal testing, including analysis as follows: demonstration thu the LPFL mode (of each RHR loop)is capable of automatically a. RHR Flow (each loop) d initiating and operating in response to a 2 4200 gpm (at 40 psid) LOCA signal.

b. Ti.ne to Rated Flow (each loop)

. s 36 sec p 3. The RHR System operates in the 3. The primary containment performance 3. RHR System automatically actuates in the suppression pool cooling mode to limit the analysis for long-term peak pressure and suppression pool cooling made as long-term temperature and pressure of the temperature shall be validated by RHR designed and RHR heat exchanger containment under post-LOCA conditions. System functional testing demonstrating performance is consistent with the the required flowrate through the heat containment coc!!ng system analysis as exchanger and by inspection of vendor test follows: data demonstrating the heat exchanger's effective heat removal capability. a. Effective heat removal capability of each Automatic initiation in the suppression RHR Heat Exchanger (K coefficient) pool cooling mode will also be includes effects of RCW, RSW and UHS: demonstrated. 2195 Btu /sec'F.

b. Tube side flow of each RHR Heat Exchanger 2 4200 gpm 4.

A portion of the RHR System return flow (in 4 RHR System functional tests shell be 4. RHR loops B and C each separately are loops B & C) can be diverted to the wetwell performed to demonstrate wetwell spray capable of providing wetwell spray flow spray header. flow capability. consistent with the suppression pool 3 bypass analysis as follows: en in a. Wetwell spray flow.(each loop individually) 2 500 gpm. O O O

Rem val System (C:ntinued) O o Tcbla 2.4.1: Residunt H: g Inspections, Tests, Analyses and Acceptance Criteria Certified D~:4gn Commitment inspections, Tests, Analyses Acceptance Criteria 5. The RHR Sys'em operates in the shutdown S. RHR System functional tests shall be S. RHR System (each loop) is capable of cooling mode to remove reactor core performed to demonstrate operation in the taking suction from and discharging back decay heat and bring the reactor to cold shutdown cooling mode of operation. to the reactor pressure vessel. lHeat shutdown conditions. exchanger heat removal capability in this mode is bounded by containment cooling requirements-ITAAC # 31 6. The RHR System (loops B and C) operates 6. RHR System functional tests shall be 6. RHR Svstem (loops B & C) is capable of in the augmented fuel pool cooling mode performed to demonstrate operation in the taking suction from and discharging back to supply supplemental or replacement augmented fuel pool cooling mode of to the normal fuel pool cooling system. cooting to the spent fuel storage pool operation. IRequired cooling capability in this mode under abnormal conditions. bounded by containment cooling requirements - ITAAC #3] g 7. The RHR System (loop C) provides an AC 7. RHR System functional testing shall be 7. Flow capability exists for directing water independent water addition function. performed to demonstrate operation in the from the fire protection system to the RPV AC independent

• vater addition mode of and drywell spray sparger, via the RHR operation.

System (loop C), without pows being available from the essential AC distribution s/ stem.The valves are capable of being cpened by manual hand wheels. 8. The RHR System operates when powered 8. RHR System functional tests shall be 8. RHR System is capable of operating when from both normal off-site and emergency performed to demonstrate operation when supplied by either power source. on-site sources. supplied by either normal off-site power or l the emergency diesel generator (s). 9. If already operating in any other mode, the 9. Using simulated inputs. logic and 9. RHR logic functions to automatically RHR System automatically reverts to the functional testing shall be performed to reconfigure the system to the LPFL mode LPFL mode in response to a LOCA signal. demonstrate the RHR System's ability to of operation in response to a LOCA signal automatically revert to the LPFL mode from any other mode.

10. Pressure isciation valves are provided to

10. Using simulated inputs. logic and

10. Automatic isolation and interlock features protect low pressure RHR piping from functional testing shall be performed to function upon receipt of input signals.

being subjected to excessively high reactor demonstrate operation of automatic isolation and interlock functions of pressure. pressure isolation valves.

g Tcbla 2.4.1: Residu;l H:ct R:mov:.I Syst:m (C:ntinued) Inspections. Tests, Analyses and Acceptance Criteria Certified Design Commlu.4:n Inspections Tests, Analyses ActWye Wic

11. Each RHR loop operates automatically in a

11. Logic and functional testing shall be

11. RHR System op 2 % wro<epts..ly minimum flow mode to protect the pump performed to demonstrate operation of the to assure a pu ~

4. w tscw path from overheating.

.ninimum flow mode for each loop exists and no tw " h offects a (including extended minimum flow observed during exts.4 f opere on ;n the operational conditions). minimum flow mode.

12. The RHR System automatically isolates

12. Using simulated inputs, logic and valve

12. The shutdown cooline W.

' iation shutdown cooling suction valves to functional testing shall be conducted to valves automatically isou. vn a low prevent draining of the reactor vessel. demonstrate operation of the shutdown reactor water level signai. l cooling mode isolation function.

13. RHR System valve interlocks prevent

13. Using simulated inputs, logic and

13. RHR System valve interlock logic functions establishment of a drainage path from the functional testing shall be conducted to upon tsceipt of input signal, reactor vessel to the suppression pool.

demonstrate operation of interlocking between RPV suction valves and other RHR g 7 valves providing potential flow paths to the suppession pool.

14. The drywell spray inlet valves can only be

14. Using simulated inputs. logic and

14. RHR drywell spray permissive logic i

opened if there exists high drywell functional testing shall be conducted to functions to prevent drywell spray inlet pressure and the RPV injection valves are demonstrate operation of drywell spray valves from opening in the absence of fully closed. permissive logic. either a high drywell pressure signal or a signalindicating RHR RPV injection valve (s) not fully closed.

15. The RHR pumps are interlocked from

15. Logic tests shall be conducted to

15. An RHA pump start signal is not generated starting without an open suction path.

demonstrate that the RHR pumps will not in the absence of indication of an open start without an open suction path being suction path. available.

16. The RHR System utilizes jockey pumps (1

16. Functional tests will be performed to

16. Each jockey pump performs its keep-fill in each loop) to keep the pump discharge demonstrate the ability of the jockey pump function.

lines filled. (in each loop) to keep its respective RHR pump discharge line full while in the standby mode. n 9 O O

O M Table 2.4.1: Residual Heat Removal System (Continued) 9e inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

17. The RHR System full flow test mode allows

17. Functional tests will be performed to

17. Each RHR subsystem demonstrates full periodic demonstration of RHR capability demonstrate operation in the full flow test flow functional capability while during normal power operation.

mode. coproximating actual vessel injection conditions during operation in the futt flow + test mode.

18. The RHR pumps have sufficient NPSH

18. Pump vendor records will be inspected and

18. Minimum pump NPSH available, as I

during postulated operating conditions. as-procured pump NPSH compared with determined based on as-built conddions design basis analysis assumptions. Actual and the results of vendor tests and/or i system installation will be inspected, and analyses, exceeds as.orocured pump l i appropriate measurements taken,to requirements and is consistent with design l j determine available pump NPSH. basis analyses requirements that includes saturated water conditions. g

19. The RHR pumps have adequate headTiow

19. Pump vendor test records and calculations

19. RHR pumps,in as-installed system characteristics.

will be inspected, and as-installed system configuration, demonstrate head / flow flow testing conducted, to establish pump characteristics consistent with design basis head / flow characteristics. analyses assumptions.

20. Control room indications are provided for

20. Inspections wi!I be performed to verify

20. The instrumentation is present in the RHR System parameters defined in Section presence of control room indication for the control room as defined in Section 2.4.1.

2.4.1. RHR System (Section 2.4.1). 4 4 E 7 y, I 3 i S ( 4

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l ABWR D: sign D: cum:nt 2.7.4 Instrument Racks O No entry. Covered under item 2.7.3. 4 OV O e og 2.7.4 1 G/1/92 i

ABWR oesign occument 2.7.5 Multiplexing Design Description lisential Multiplexing System The Essential hiultiplexing System (131:0 piorides distributed data acquisition and control networks to support the monitoring and cono ol of the plant standby safety systems. EhtS compiiscs electrical devices and circuitry, such as Remote hiultiplexing Units (IG1Us), tmnsmission lines, and Control Room hiultiplexing Units (ChiUs), that acquire data f rom remote piocess sensors and discrete monitors located within the plant and multiplex the signals to Safety System Logic and Control (SSLC) equipment in the main control room area. SSLC piocesses the input signals and multiplexes output control signals to the final actuators of driven equipment associated with safety systems. EhtS is divided into four divisions of equipment, each with independent control of data acquisition, multiplexing, and control output functions. System timing is asynchronous among the four divisions. No conunon clock signalis transmitted among the divisions of multiplexing and no timing signals are exchanged. lbth analog and discrete sensors are connected to Rh1Us in h> cal areas, which perform signal conditioning, analog to<ligital conversion for continuous process inputs, change-of-state detection for discrete inputs, and message formatting prior to signal transmission. The IG1Us are limited to acquisition of sensor data and the output of control signals. Trip decisions and other control logic functions are performed in SSI.C processors in the main control room area. The IGIUs tmnsmit serial, time-multiplexed data streams representing the status of the plant variables via fiber optic cables to the control room ChiUs. Data transmission is made over dual redundant channels. EhtS design features automatic self-test and automatic reconfigumtion after failure of one channel (either a cable break or device failure) The system returns to nonnal operation after reconfiguration within one full scan period if an 181U or ChiU has failed, that unit will be removed from service. Faults and their location are annunciated to the operator in the mai._ control room. The ChfUs demultiplex the data and prepare the signals for use in interfacing controllers of SSLC or monitoring systems such as the process computer or display controllers. After the input data is processed in SSLC, the resulting trip logic decisions are transmitted (for Engineered Safety Features functions only) as a serial, time-multiplexed data stream via EhtS to IB1Us in the local areas, where the digital data is converted to contact closures or other signals for actuation of motor cotitrol centers or other device controllers. The data streams are dual redundant t prevent inadvertent ECCS equipment actuation after a hardware or software fault in one channel. The data reaching the IG1Us is compared in 2 out-of-2 voting logic to confirm final output to the actuators. 2.7.5 6/16/92

ABWR 0: sign Document Data can be transferied to non safety systems for control or display through isolating fiber-optic data links and bullering devices (gateways os biidges, if iequired). Data tnmsfer is made such that failures on the non-saf ety side cannot inhibit ope tion of safety-related logic functions. Data cannot be transmitted from the no a-safety side to EhtS. EhlS is capable of data trans!cr at rates sullicient to satisfy the system time response requirements of safety system functions. Data throughput capability shall be at kast 10 megabits per second. EhtS starts and nms automaticall) upon application of system power, cgardless of the sequence in which power is applied to individual controllers. EhtS and SSLC automatically establish conununications by detection of conect message paning. Logic is prmided to prevent equipment activation outputs from occurring until stable plant sensor data and interk>ck permissive data are being Icceived. Loss of power causes a controlled transition to a safe state without transients occurring that could cause inadscrtent initiation or shutdown of driven equipment. EhtS equipment is classified as safety-related, Chtss lE, and is seismically qualified. Testability EhtS includes test facilities in the control room that will monitor data tmnsmission to ensure that data tmnsport, routing, and timing specifications are accumte. Ilit error rate of each EhiS network shall be better than 1 error in 10* Out of-tolerance pammet.rs detected on-line for a particular input signal will result in an inoperative condition for that input into the trip logic ;>rocessors of SSLC. Non-Essential Multiplexing System The Non-Essential hiultiplexing System (NEhtS) provides data communications for non-safety-related plant functions. NEMS acquires non-safety-ielated data from process sensors and discrete devices located throughout the plant and transmits these signals to the non-essential control systems for control function processing. Equipment status data is transmitted to operator control panels for monitoring alarm annunciation and to the plant computer systems fer data recording and displays. NEhtS also transmits processed, non-safety-related, control signals to actuator circuits to activate valves, motor drives, alarms, monitors and indicators of the interfacing systems. NEhtS is active in all modes of plant operation. 2.7.5 6/16/92

ABWR 0: sign Docum:nt n NEhtS shall have electrical devices and circuitry independent and diverse from V that of the Essential hiultiplexing System (EhtS). Diversity shall be such as to pieclude the possibility of common-mode failure intemction between the equipment comprising the two systems. The diversity may take the form of different hardwme and sof tware for similar functions or equipment piovided by different manufacturers. The electrical desices of NEhtS, which include Remote hiultiplexing Units, tnmsmission lines, and Control Room hiultiplexing Units (Ch1Us), comprise nuious dedicated or interconnected data networks Depending on the application and required level of reliability, the networks may be interconnected as single channel, dual redundant, or triply redundant. Redundant netwm ks incoq> orate fault-tolemnt recon 0guration capability; any single f ailure of the network will cause automatic failover to the remaining good channel or channels. NEhtS will be capable of self-starting following a power interruption. Redundant networks will restart automatically after any single failure, includmg a processor failure. During system initialization or shutdown, no control output or alarm will be inadvertently activated. Control outputs will change to predetermined safe-state outputs. The various data networks comprising NEhiS will include a means to continually monitor the circuit integrity of each system. Status indicators and alanns relating to a system's status shall be displayed to the opemtor, NEh1S interfaces with non-safety systems within both the reactor building and turbine building. Rh1Us are located throughout the various buildings and may be h>cated inside the secondary containment with proper environmental qualification. However, Rh!Us shall not be kxated within the priman containment. Intercommunication is piovided between NEh1S and Eh1S to transmit safety-related signals needed by non safety systems. Such tnmsfer of data shall be made via isolated, fiber optic interconnections and shall not impair the safety-related functions of EhtS. Hardware and software failures on the non-safety side shall not affect normal operation of the safety side. Data cannot be transferred from NEh1S to EhiS. Ov 2.7.5 3-C/16/92

C ABWR oesign Document NEMS signal processing functions are as f ollows: (1) Acquire data froin process sensors and discrete desices, including opetutor's control switches, located throughout the plant. Inputs include process transmitters, voltage inputs, pulse inputs, thermocouples, RTDs, iclay contacts, and limit switches. (2) l'erform processing tasks, including signal conditioning and execution of control algoritluns as required by the system design specifications of interfacing systems. (3) Fonnat and transmit data signals via fiber optic cables to controllers, displays, plant computers, and the operator's console. (4) l'erform enor detection oficceived signals using, as a minimum,16 bit CRC checking. (5) Receive, demultiplex and prepare the signals for use in interfacing equipment. Signals will drive the following analog and digital cutput loads: (a) analog controllers, inputs and outputs (b) analog meters and recorders (c) digital indicators, such as flat panel displays, digital panel meters, light emitting diodes or CRTs (d) M/A stations (c) lights and annunciators (f) other networks (via appropriate buffering devices to match data communications protocols and transmission speeds) (6) Format and transmit processed control signals via fiber optic cables to actuators within the plant. e (7) Receive and convert the liber optic control signals to electrical signals suitable for the actuator circuits. NEMS will be capable of operating external solenoids, relays;.nd other actuators. Transmission response time and processing response time will suppon the requirements of the interfacing systems. The signal scanning rate of NEMS will be at least 10 milliseconds; signals requiring faster transmission response will use g a dedicated fiber optic data link. Bit error rate of any network or data link shall be less than 10-9, 2.7.5 G/16/92

ABWR ocsign occument q Testability , C/ NEMS is designed to pennit periodic in-service tening and inspection of instnnnents and ecluipment with the plant in nonnal operation. Supenisory instnunentation and alanns are provided Ior coinplete surveillance of systein operation. NEMS includes a self-testing function that opemtes automatically at all thnes. Self-test runs continuously and indicates detected f aulta to the board replacement l< ecl. Faults are annunciated to the operator in the main control room. Inspections, Tests, Analyses and Acceptance Criteria Tables 2.7.5a and 2.7.5b provide a definition of the visual inspections, tests and analyses, together with associated acceptance criteria, which will be used by the Essential \\1ultiplexing System and the Non-Essential Multiplexing System respectively. O d OV 2.7.5 6/16/92

{ Table 2.7.5a: Essential Multiplexing System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria 1. Four divisions of independent and 1. Visualinspection of the installed 1. EMS configuration is in accordance with redundant EMS mstrumentation acquire equipment will confirm the identity and equipment arrangement shown in sectinn and transmit the safety.related sensor Iccation of EMS instrumentation. 3.4, 'iSLC ITAAC. The figures indicate the inputs and control functions of the plant equipment panets, and their required relationship of EMS to other i standby safety systems and auxiliary interconnections: safety system processing equipment. supporting systems. 2. EMS panels and processing equipment are 2. Visual inspection of installed equipment. 2. Installed configuration of EMS conforms to Class 1E, safety-related, and seismica?ly test records, and analyses based on certified commitment. qualified. equipment location will confirm the qualification status of EMS. 3. The four divisions of redundant 3. Inspections of fabrication and instatfation 3. The installed EMS equipment conforms to instrumentation are physically and records and construction drawings or certified commitment. efectrically separated from each other. visual field inspections of the insta!!ed EMS There are no interconnections among equipment will be used to confirm e., divisions of EMS. Data communications to electrical and physical separation. the process computer or display controllers shall use an isoitting transmission medium such as fiber optic cables. 4. The RMUs and CMUs in each 4. System tests will be conducted after 4. The installed instrument channels are instrumentation division are powered installation to confirm that the electrical operational with the power sources independently from the divisional plant DC power supply configurations are in specified in the certified commitment. sources (Class 1E 125 VDC). compliance with <1esign commitments. 5. EMS meets Electromagnetic Compatibility 5. Factory tests for EMC will be conducted in 5. EMC performance of EMS is considered (EMC) requirements. Protection is provided a controlled environment on individual acceptable if tests confirm that against the effects of: EMS equipment and on the integrated electromagnetic fields static discharges, a. Electromagnetic Interference (EMI) system configuration. and electrical surges do not affect system b. Radio Frequency Interference (RFI) capability to acquire and condition data, c. Electrostatic Discharge IESC) EMC tests will also be conducted on the transmit formatted data, receive contro!

d. Electrical surge ISurge Withstand installed EMS configuration in the normal signals and send control outputs to final Capability (SWC)]

plant operating environment. actuators. 3 2 9 9 9

y Table 2.7.Sa: Essential Multiplexing Systern (Continued) 3

r Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria 6.

EMS includes test f acilities that will 6. Preoperational tests will be conducted on 6. Operability of the installed EMS equipment monitor data transmission to ensure that the installed EMS equipment. These tests is considered acceptable under the data transport, routing, and timing are will confirm the basic functionality of each following conditions (for each divisionh accurate. multiplexing component.The tests will a. Monitored output signais match include simulation of typicalinput simulated input signals for accuracy of parameters and monitoring of the received signal conversion and transmission transmitted parameters. Ilandom ti ne. simulations will be used to test bit error b. Bit error rate is <10'8 rate, which shall be determined to be <10-c. Simulated data errors are detected and 9-annunciated to operator. 7. Full system test of EMS with SSLC and 7. Preoperational tests will be conducted to 7. EMS support of the interfacing safety other interf acing systems connected verify safety system logic functions of each systems is considered acceptabf e if reactor 4 j confirms Ef.iS response to safetv system interfacing safety system. These tests will trip, containment isolation, and ECCS tests specified in each interfacing system verify support of SSLC and the safety response of the installed equipment meet ITAAC. Testing is conducted on the four systems for scram capability, containment the acceptance criteria stated in each 4 divisions of EMS /SSLC simultaneously to isolation capability, and ECCS initiation interfacing system lTAAC. The response 4. verify 2-out-of-4 system operation. capability.The tests will include time of each control action and trip output demonstration of ability to meet stated is within performance limits of each delay times and maximum response times. interfacing system. l Tests will be conducted such that each j display, alarm, annunciator, or other status Performance of SSLC for these same tests } indicator for each system is shown to be also confirms EMS performence. functional. See section 3.4 SSLC ITAAC, item 7 for the scope and method of testing. 8. EMS provides safe-state response to loss 8. rests will be conducted to verify that 8. EMS response to loss of power is of pnwer source. graceful degradation of EMS system acceptable for the following conditions: outputs occurs upon momentary or long-term loss of one division of the DC power a. Loss of one division of power does not source or power to individual EMS cause false output trip or inadvertent components. Tests will also confirm that initiation of final system actuators. reinitialization of system or component. Loss of power and loss of divisional after power is restored does not impair

• rip signals are annunciated.

i normal system function. a Ei i

y Table 2.7.Sa: Essential Multiplexing System (Continued) in Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

8. (continued)

b. Loss of power to individual component produces a safe-state output condition without extraneous false outputs (normally-energized outputs de-trr.ergize, normally-de-energized outputs remain de energized).

c. Restert (initialization) of component or system upon recovery of power does not caus., inadvertent output action (outputs remain in safe-stata condition until sensed inputs are evaluated in processing circuitry). d. Transient power loss (<1 second) causes no false output trip or inadvertent initiation of final system N actuators or removal of previous l tripped state. 9. EMS is fault-tolerant in each div;sion and 9. Preoperational tests will be conducted to 9. EMS response to instrument or cable provides capability for automatically verif;that a single failure of a multiplexing failure is acceptab7e for the fcliowing reconfiguring after failure of an RMU, component does net impair total system conditions-CMU, or interconnecting cable. fonction. Faults will be simulated and the a. A singla cable break does not affect response monitored. Tests specified in network operation. item 6 wili be repeated to confirm

b. Loss of one RMt1 or CMU removes that operability of network.

unit from service; network continues normal operation. c. Fault occerrence and notice of I reconfiguration is displayed to operator. Ri 9 9 9

a O O O Table 2.7.5b: Non-Essential Multiplexing System } Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspects.ms, Test, Analyses Acceptance Criteria 1. Plant-wide, distributed, data 1. Inspections c< fabrication and instsilation f. NEMS configuration cor. forms to communications networks provide the records and construction drrwir3gs or intercon section documer,tatic 1 of the means to acquire and transmit non-visual field inspections of the instaOed interfacing systems. essential plant parameter and control data NEMS equipment will be used to confrm to non.cssential monitoring and control identity and location of NEMS data systems. networks 2. NEMS instrumentation is powered from 2. System tests will 1,e conducted after lc. The installed instrument channels are the plant non-essential 125 VDC bus. installation to confirm that the electric 61 operational with the power sources power supply configrrrations are in specified in the certified commitment. compliance with design commitments. 3. Supervisory instrumentation and alarms 3. Preoperational tests will be conducted on 3. Operability of the instaffed NEMS are provided for complete surveillance of the installed NEMS equipment.These tests equipment is considered acceptable under system operation. NEMS provides a means will confirm the basic functionality of each the following conditions: to continually monitor the circuit integrity multiplexing component.The tests wi!I 1onitoted output signals match of the data networks. Status indicators and include simulation of typicalinput simulated input signals for accuracy of 4, I ^ alarms relating to NEMS status shall be parameters and monitoring of the received signal conversien and transmission displayed to the operator transmitted parameters. Rendom ,tune simulations will be c ;ed to test bit error b. Bit error rate.is <10'g. rate, which shaII in determined to be <10. c. Simulated data errors are detected and annunciated to operator. l 4. NEMS meets Electromagnetic 4. Factory tests for EMC will be conducted in 4. EMC performance of NEMS is considered Compatibility (EMC) requirements. a controlled environment on irodividual acceptable if tests confirm that Protection is provided against the effects NEMS equipment and on the irtlegrated electromagnetic f; elds, static discharges, of: system configuration. and electrical surges do not affect system capability to acquire and condition data, a. Electromagnetic Interference (EMI) EMC tests will also be conducted on the transr,it formatted data, receive control b. Radio Frecuency Interference (RFI) installed NEMS configuratica in the normal signals and send control outputs to final plant opra6ng endonment seah d. El rc I ur (Surge hstand Capability (SWC)] R 5"> d

[ Table 2.7.5b: Non-Essential Multipledug System in Inspections, Tests, Analyses and Accep?ance Criteria Certified Design Commitment inspections. Test, Analyses Acceptance Criteria 5. Full System test of each NEMS network or 5. Preoperational tests wi!! be con &.ted to 5. NEMS support of the interfacing non-data link with interfacing systems verify fogic or data processing f:$nctions of safety systems is considered acceptable if connected confirms NEMS response to each interfacing non-safety sys'am. These system control actions or monitoring ar d non-safety system tests specified in each tests will verify support of each system by display capabilities meet the acceptance interfacing system ITAAC. NEMS for data acquisition, algc*ithm criteria stated in each interfacing system processing, and control output. me tests ITAAC. The response time of each control will include demonstration of f.bility to action and trip output is within meet stated delay times and maximum performance limits of each interf acing response times. Tests will be conducted systerr.. such that each display, ala ns, annunciator, or other status indicator for each system is shown to be functional. 6. NEMS networks that employ a dual or 6. Preoperational tests will be conducted to 6. NEMS response to instrument or cable triply redundant configuration provide the verify that a single failure of a multiplexing failure is acceptable for the following capability of automatically reconfiguring component does not impair total system conditior's: L after f ailure of an RMU, CMU, or function. Fauits will be simulated and the

a. A s.ing~e cable break does not affect interconnecting cable.

response monitored. Testa specified in network operation. 9 item 3 will be repeated to confirm

b. Loss of one RMU or CMU removes that operability of network.

unit from service; network continues normal operation. c. Fault occurrence and notice cf reconfiguretion is displayed to operator. 7. NEMS incorporates facilities to provide a 7. Tests witi be conducted to verify that 7. Response to network shutdown is graceful response to power interruption or graceful degradation of NEMS system acceptable for the following conditions: other causes of inadvertent network outputs occurs upon momentary or long-NEMS will be capable of self-starting a. sht?;iown. term loss of the DC power source or power I I.' "'"U

• P "'nntenuption.

to individual NEMS components. Tests will

b. hg system M.ahadon or aise confirm that reinitialization of system shutdown, no control output or alarm or component after power is restored does wiH be snadvertently activated. Control not impair normal system function.

outputs will ch. ge to predetermined safe-state outputs. E e G G

1 ABWH 0: sign occument 1 l k.7.6 Local Control Box No entry. Covered under Itein 2.7.3. 4 l ) J 4 I ) 4 4 O 4 i l 4 O 4 2.7.6 6/1/92

i ABWR o sign Document 2.12.16 Instrument and Control Power Supply p) ( No entry. Covered under Item 2.12.15. O O 2.12.16 1 6/1/92

ABWR 0: sign Docum:nt 2.12.17 Communication System Design Description The Nonnal Plant Communication System is comprised of a telephonic system with a power actuated paging facility for plant operation, testing, and calibration and a separate sound powered telephone system for maintenance and repair, ik>th systems are non-safety-related and, therefore, do not have seismic mounting requirements beyond those required to prevent damage to essential equipment. All other communication systems (e.g. independent emergency 9 communication, wireless communication, telephone, etc.) are independent of the normal plant communication system and will be provided on a site-specific basis. (See Figure 2.12.17.) The Telephonic Communication System provides intraplant communications and broadcasting in various buildings and outside areas which are important to plant operations, personnel safety, and where communication needs are frequent. The system is a multi-channel system with independent amplifiers and distribution panels and is connected to the telephone system so as to pennit sinnultaneous broadcasting from a security telephone unit. In addition, the paging system equipment produces an audible emergency alann when actuated. The entire system is powered from a dedicated battery with a normal and standby O. battery charger and is designed for 10 hours of operation during periods when offsite power is lost. Each battery charger is sized to provide all communication needs while fully charging the battery. Each channel is provided with its own amplifier and an automatic self-check feature for amplifier output and control room alann indication for major system component failures. Manual switching to an alternate amplifier is provided. Handsets and speakers are provided in the same relative location on all floors and at unifonn intervals in corridors and rooms and are located in areas least affected by mdiation. In noisy areas, handsets are located in sound-proef booths. Redundant cables are provided between distribution panels andjunction boxes. Cables are color coded, have static electricity shields, heat and flame resistant insulation, and redundant cables are routed in separate, control level mceways. Containment penetmtions are provided for cables entering the primary containment. The Sound Powered Communication System is comprised of a central maintenance communication patch panel and a system of cables and communicationjacks local to panels and mcks where communication is frequently required. Coramunication capability is provided between the main control room and field stations and between field stations. Cables are color coded and routed in control level raceways, when available, and hase heat and flame resistant insulation. The portable sound powered telephone / handsets are the responsibility of the COL applicant. 2.12.17 -1 6/1/92 l

ABWR 0: sign Document inspections, Tests, Analyses and Acceptance Criteria Table 2.12.17 provicles a clefinition of the Inspections, Tests, ancl/or Analysis, together with tl sociate<1 Acceptance Criteria which will be unclertaken for the Cotr.inunication Systern. / O O 2.12.17 6/1/92

O O O Table 2.12.17: Communication System Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analysas Acceptance Criteria 1. The Normal Plant Telephonic 1.a inspection will be performed to assure that 1.a The Telephonic Communication System Communics. tion System is an intraplant, the installation of the Telephonic installation does not jeopardize the non-safety-related, multi-channel, battery Communication System will not jeopardize inte uity of any safety.related system. powered communication and broadcasting the integrity of any safety-related system. system with a connection to the security telephone unit and an audible emergency alarm. The system is capable of operation 1.b Testing of the Telephonic Communication 1.b The Telephonic Communication System I for 10 hours during periods of loss of System will be performed to assure that communications, broadcasting, and self offsite power with a battery charging the communications, broadcasting, and test functions perform within design l system. Multi-channels are provided with self test functions of the system perform specifications. independent amplifiers and an automatic within design specifications. self check capability and control room alarm for major systeT component j 4, failures. 2. The Normal Plant Sand Powered

2. Testing will be performed to assure sound 2.

Sound Power +.d Communications can be Communication System is a non-safety-powered communications can be achieved between the main control room 3 related system. It is comprised of a central established between the main control and the field stations and between field maintenance communication patch panel room and field locations and between field sta: ions. and a systern of cables and phone jacks locations. l-local to the panels and racks where commt,nication is frequently required. Cemmuaication capability is provided between the main control room and field stations and between field stations. l t a u r i 4 9

N .U d .CC MCC SECURITY TELEPHONE UNIT TELEPHONE & PAGING SYSTEM u BATTERY BATTERY l CONTROL l CHARGER CHAFiGER I NORMAL STANDBY AMPLIFIER AMPLIFIER 1 I 4 DISTRIBUTION I I 4 BATTERY l MDF l l MDF l BOARD SEPARATE CABLE ROUTES SOUND POWERED SYSTEM b MAINTENANCE PATCH PANEL BRANCH BRANCH PATCHING BOX BOX CORD V V 'ERMINAL SPEAKER SPEAKER BOXES O O --> H.WDSET --p-HANDSET TURBINE REACTOR BUILDING BUILDING [~~~~ ~ ~ - - -l \\ [ [ PHONE [ SETS TYPICAL CHANNELS l l g 1_ _ _ _ _ _ _ _ _ _I 3 NOTE: MDF - MAIN DISTRIBUTION FRAME SITE SPECIFIC SCOPE Figure 2.12.17 C unication System g

O O [ Table 2.12.12: Direct Current (DC) Power Supply (Continued) M 4 inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment Inspections, Tests, Analyses Acceptance Criteria

3. The non-sciety-related 250 VDC Power 3.a inspections of the non-safety-related 3.a The battery is separately housed in a Distribution System consists of one 250 distributior system arrangement will be ventilated area apart from its distribution VDC industrial type storage battery, central performed to confirm the battery is equipn.ent.

distribution panel, motor control center, separately housed in a ventilated area and distribution panels local to the apart from its associated distribution supolied losds. The battery is separately equipment. I housed in a ventilated room apart from its w*tery chargers and distribution equg ment. The battery is selected such that its 3.b A load capacity analyses will be performed 3.b The load capacity analyses confirms that warranted capacity will provide 100 showing battery terminal voltage and the battery supplies the design loads at or ~ percent of its design loads at the end.of-worst case DC load terminal voltage at above the required minimum voltage and installed-life with a minimum allowable each step of the battery loading profile to is consistent with the manufacturer's 9 voltage of 210 VDC. Tha battery is sized to assure that the battery will provide a ampere-hour ratings for the battery at a 2 supply all required loads for a minimum of minimum 210 VDC for the duration of the hour rate. 2 hours without recharging. profile. The battery is provided with a normal 3.c inspections of the normal and standby 3.c Battery charger nameplate racngs confirm battery charger supplied from two different battery charger ratings (as identified by their capacity to supply normal steady non. essential load group Power Cen:ers (P/ their nameplates) will be performed to state loads and the normal charger's C) through an interlocked bus transfer confirm their capacity to supply normal capacity to recharge the battery at a device to prevent paralleling AC load steady loads and the normal charger's maximum voltage of 280 VDC while groups.The battery charger is a self load capacity to recharge the battery at a supplying loads. limiting battery replacement type and is maximum voltage of 280 VDC while sized to supply normal steady state loads supplying loads. while restoring the battery to a full charge 3.d Tests will be performed to confirm that the 3.d Tests confirm that AC load groups or state at a mammum charging voitage of batter) charger interlocks will prevent battery chargers cannot be paralleled. 280 VDC. A smaller standby battey paralleling AC load groups or battery charger, powered frem a control bullding chargers. MCC, is also provided and sized to supply normal stead state loads during battery i 3 rnaintenance.The two battery charger 3 outputs are interlocked to prevent paralleling chargers. (Sea Figure 2.12.12c.)

1 l ilIllll i g Y R R LE E AG CT MR DT E RA vA ) OH / sB-: NC F 2l l ) 9 ) C 1i YV C VI C C D ora D L i N l C l 1M P LR ) V 1 AT LP I 1B R CS NY D Vs T I) f OI PT IC S LD D I l ( V D 08 YR C 4 a !)l A BE D r NR DG \\ AA y C l gK_ TH I / l SC p L p N u F C P C S M R R 1) I V C W / I l B LE T C D r V/ AG I I w e S V t C D MR D Y V RA IV @=) A V o R 0 OH C LR i D 8 ) NC D A LP P E 4 T ) DT ') CT YCS NY d OIPT vA E ) LD ( e = t sB = \\ a 2 V 1I le V I R D -y te Y R R LE D AG E C MR V 1 DTE RA 5 E s' OH / 2 vA sB"E NC F 1 2i ) 9 ) C 1 I YV a V C C 2 ora D C L D 1 LR C N A ) tM P TLP I 2 V/B R l) VCS NY V OI PT I 1 D I T LD ( 2 DC S C e I V D C 0 r 8 YR C l) VM u 4 'a )i A BE D C g DG D \\ @- @>-lgx_ iF NR AA K TH / C C '> SC O L L C N R P F E M !! B LE T l) C T R R YV C NI C D V/C AG I Y S lC MR D E V RA V K Y 0 @FiA OH C L R 8 ) I R D = NC D A T LP E 4 DT ') ) CT Y CS NY h OI PT vA g ) LD ( sB 111 \\ E 2 O T I O V D N Ndb h sau Illl(l' l'

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• LNP LNP LNP RTSID RTSID RTSID

SETON FCVC LACOL FCVC LACOL FCVC LACOL v

v v y y V - i 'a 'a - i 1 i .-.'.-. )1 i 1 i G G a E* ) ) D* ) ) C* ) ) B* ) B* ) B* )' ) ), ) )' ) I REGRAHC YBDNATS REGRAHC REGRAHC REGRAHC LAMRON LAM.PON LAMRON I I l l l ) A* ) A* )lA* T ) C PUORG BPUORG APUORG g g g DAOL DAOL DAOL YRETTAB CCM YRETTAB CCM YRETTAB CCM I CCM B/T CCU B/T CCM B/T CDV 521 B/T CDV 521 B/T CDV 521 B/T u t e O O 9

l ABWR 0:siga Documsnt O LOAD LOAD LOAD GROUP A GROUPA GROUPC 480V 250 VDC 480V 480V C/B MCC BATTERY P/C P/C 5 STANDBY ) ) ) NORMAL CHARGER CHARGER p _ _ __ 3 ) )__ _ _ _.________.) O 250 VDC CENTRAL DISTR BOARD ) ) V V COMPUTER OTHER CFCV LOADS O Figure 2.12.12c 250VDC Power Supply 2.12.12 G/1/92

O O O i P d i SwrTCH YARD l l 4 ~ MR UNK h h [RE l-OFFSITE = + mh% RAT' UATB UATC /v W gn /v W GEN .- _ _ _ _ _ e _) A3 6 B1 ogo o B2 B3 O C1 o A1' " o o C2 C3 0 i o) o)(PG) m(P!P)g) (pc)g}(PlP)9) (PG) jk/p) o) (PG) m(P!P) O) (PG) 9, g A2 (PG) og o o o o b) b) b) b) b) ?) b) b) b) l l y y y y y I y 9 9 cb TO TO TO TO TO TO t I NON4 LASS 1E, NON4 LASS 1E l NON-CLASS Sg NON-CLASS 1E NON-CLASS 1E NON-CLASS 1E NORMAL STANDBY NORMAL STANDBY NORMAL l STANDBY i UGHTING UGHTING . UGHTtNG UGHTING UGHTING UGHTING i ~ l ~ ~ ~ l 'i l 'l l I 1 i I i I f. .. ii ' 1 ' 9 4> ap E (D1) '? F (D2) ? G (D3) 7:h3 0

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TO DG-TO DG CLASS 1E STANDBY CLASS 1E STANDBY CLASS 1E STANDBY

. (D1) - (D2) N UGHTING - /. N UGHTING - f N LIGHTING j B LOADGROUP A ' LOAD GROUP B LOAD GROUP C ACTUAL UGHTING POWER SUPF1JES ARE FROM 480V SWITCHGEAR Figure 2.12.18a Normal and Standby Lighting i a

i N d kt l l CLASS 1E NON-CLASS 1 E NON-CLASS 1E 125VDC 125VDC 250VDC DIVISIONAL BATTERY NON-ESSENTIAL BATTERY NON-ESSENTIAL BATTERY DIST. PANEL DIST. P ANE-DIST. PANEL g 6 6) 6) 6) 5) 6) f) 6) i f f f f f f MCR DG RSS ELEC. CTG ELEC. RWB EQUIP. EQUIP. N / \\ / \\ / TYPICAL OF4 TYPICAL OF 3 TYPICAL OF 1 1 PER DIVISION 1 PER NON-ESSENTIAL LOAD GROUP (DIV I, !!, fil, IV) (LOAD GROUP A, B. C) I 5 B Figure 2.12.18b DC Emergency Lighting O O O

ABWR Design Document 2.15.2 Turbine Pedestal O No entry. Covered by item 2.15.11. O O 2.15.2 6/1/92 .. ~...,. - - - - - ~. - +.

O O O g um MO Mo VD VD VD Go TORNADO MISSitE BARRIER t t t t rat 1 LOUVER tJC TORNADO HECW - og ey e = + e 1 -N W f t 1=)}'I 4 EL EC "i t \\ gg BAG 12SV FILTER BATT ERY RM Dtv 1 ELEC E PRM TORNADO pg ~ y C ER VD GD RAIN LOUVER RM A t ) + f i GD i VD CONTROL BUILDING HVAC - A ) TM N HVAC 8 AND HVAC-C SIMILAR (LESS NON DMSK)NAL ROOMS) MtSS:LE BARRIER LEGENDS. VD - WOLUME DAMPER TE - TEMPERATURE ELEMENT RE - RADtATION ELEMENT C -GRAVITY DAMPER HECW - HV AC EMER COOLING WAT ER 5 m Figure 2.15.5a Control Building HVAC System

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g p RE E '@LT# y Me GD MO VD GD f LIGENDS 4 i so go ao 99 H / H e (RE TE - TEMPERATURE ELEMENT l* "o FE - FLOW ELEMENT MO W t RE - RADIATION ELEMENT e y GD GRAVITV DAMPER d# SD-SMONE DAMPER NC MO MOTOROPERATED R h"HvIc REACTOR BLDG EXHAUST E g FCU - FAN COL UNIT FJ Ivc - IP4.ET v AtvE CONTROL Figure 2.15.5c Reactor Building Secondary Containment HVAC System

N$ i.n W MO GD TORNADO MISSILE BARRIER Y j/j NC RA.'N LOUVER NO TORNADO HECW ^ GD VD / 9 ELEC I ) J ip EQUIP y VD ROOM t GD DV2 A CO t. PANEL RM BAG HVAC EOUIP FILTER ROOMS DV2 DG.B A A GD TORNADO DG 4 E [ GD RA;N LOUVER g q ? k- = / + REACTOR BUILDING ELECTRICAL EQUIPMENT HVAC (B) L TORNAoO (A)(LESS RIP CONTROL PANEL ROOM) AND (C) SIMILAR BA'SSILE BARR!ER LEC-END*4 VD - VOLUME DAMPER TE - TEMPERATURE ELEMENT GO -GAAVfTY DAMPER HECW - HVAC EMER COOUNG WATER G 6 u Figure 2.15.5d Reactor Buildi ectrical Equipment HVAC System

' A g '] f ) %d "F---~~--- m VD { RIP PS (H) vo { R!P PS f) VD { RIP PS (E) VD { RIP PS (B) i yo VD HNCW -*-- GD +4 VD t U k l GD V \\ (A) (C) (D) (0) l (S + n e - tracw RIP POWER SUPPLY (TYP) I RIP HVAC SYSTEM A C S!MILAR LEQREL VD-VOLUME DAMPER TE -TEMPERATURE ELEMENT OD-GRAVITY DAMPER HECW - HVAC NORMAL COOLNG WATER R PJP - REACTOR NTERNAL PUMP g PS - POWER SUPPLY w Figure 2.15.5e Heactor Building RIP HVAC System

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y a": NC GD c' ~ LEGENM ( GD - GRMtTY D AupER HWH - Hof W ATER HEATING Wo p HNCW HVAC NORM AL COOLING W ATER f 4 ( J Po MO MOTOROPERATED NC FROU FIG. 2 *5 5g N RE RCATION ELEMENT TE - TEMPERATURE ELEMENT TURB1NE BLDG EXHAUST VD VOLUME DAWPER Figure 2. 7 Sf Tur ~ Building Ht/AC Sg ; tem S

r (~ O \\ V TURSINE BLOG U1 ) 1 ) 590M FIG 215.5t ( 2 l TO FtG. 215 st RW FtLTER TURB!NE COND SUMP AAEA SAMPLE EAST AREA BACKWASH TORS:NE COND TANK pa A AREA CONTROL WEST STATION I ICUl BACKWASH DEMIN LAYDOWN PUMD PSV AREA AREA CONDENSER [TCU-] COMPARTMENT ~ LEGENDS AMERT AP RESIN FCU FAN Colt UN!T COOLER STORAGE R R TANK A N RW RADWA$fE TANK HTR DRA:N EHC ELECTRICAL HYDRAUUC CONTROL PUMP COND Pty TCW - TURBINE BUtLDING COOUNG WATER TRAhrs. COND AREA STEAM T&V PUW & VALVE RFP. REACTOR FE EDWATER PUMP EA ~ 54C - STOP & CONTROL EXCITER FR.TER GSC - GENERATOR ST ATOR COOGR CUB + CAL ND g4 ~' S&C VALVES [Tcg7 WATER SCRUB 8 CHEM COND. AlR RM HTR DANN j N-y TANKB ENC {Tg[J] TCW PUW iTR DRAIN AND VACUUM PUMP PUMP A - HpMTECT voD y usR csc LO. COND. RET, EAST COND. TANK g,REN Rm y RW SUW LLSE UL SAAftE y RESERV. HTR DAAN STATION POW B OFFGAS RTPPWR BE S To Fn e is s' say ( 3 g fTciT 1 ~ TURBINE BLDG LUBE O!L EXHAUST FTctn COND. PUMP MAJNT. ] 3 To Fic. r.ts se TURBINE BLDGEXHAUST Figure 215.5g Turbine Building HVAC System m

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hfE AUX BOfLER LEGU'" AREA GENERATOR VD-VOLUME DAMPER i GD -GRAVITY DAMPER FCU - FAN COfL UN!T RL-RAN LOUVER HNCW-HVAC NORMAL COOUNG WATER 3a u Figure 2.15.5h Turbine / rical Building HVAC System 9

M ~ E __g g ~ ) se > ""C"-] _, o,o "]g-my y ranti ; vo i m1um _p )- I) Vo ( -E2OiE 1 ) .- - HNCW a + ( ) ao \\ g g3 1m mmm mm L._ (I_ I_>_ yo! h-F }-, _ mu# { - f RMS2 } MRANTM 1 ECU m DCALOF_it t-l-HWH q- -{g-DwE L_) QQP G ,,7 m,, sun >tv MONrrORED 43 STACM L_ m > **aAo ) EXHAUST SEBvic1SLDG HV AC NQ?iBAplQACI]VILQONTRQU,.ER_A_R_EAS A0 D'**E" @V i s V' Go vo vo T >+ ag -}- M t s % ar 8 PLY EXHAUST p d oo /1 -}- EYr EIE$1 f ao _.J L_ _} h*,j -Ly/gp 2 _s g =__) b 5 s t' MW % 22 0 EERvlCER',0G HVAQEAQLQAC11VE CQNIflQLLEQAhEAS_ 4 ({QLN R 1 ~~ HNCW VD. VOLUME DAMPER RE - RADotTION ELDfENT h .p WE - taOesTUF4E ELEMENT O-*" M SLMV > TE TEMPERATURE ELEVENT 00 GRAvfTv DAMPER i reu-rAN eos. uurr Q Lr 'l

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Hvw-Hvac COOLER MCW HV AC NORMAL COOLNG WATER _ m FL RAMLOUVER u HVH_{TlYCA O_F 31 MB - REACTOR BULONG l BF BASEMENTFLOOR Figure 2.15.5i Service Building HVAC System

b e Table 2.15.10: Reactor Building (Continued) M f S. Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment Inspections, Tests, Analyses Acceptance Criteria . 8. Protection against pipe break event 8. Review the design documentation to. 8.Comformation that the as built structures are j dynamic effects is provided to assure that assure that the analysis of pipe break in compliance with the design the reactor can be shut down safely, that events is performed for design features documentation. For radiation protection, - the containment integrity is maintained, - such as separations, barriers, and pipe see Section 3.7, Radiation Protection. j - and that the radiological doses of a whip restraints provided for essential items j - postulated piping failure remain below the for plant safe shutdown. j limitsi i 1 9. Review RB construction records and 9. [ j 9. Secondary containment boundary l' copletely surrounds the PCV and encloses perform visual inspections of the as-built a. Per Figures 2.15.10.a through 2.15.10.o. Y all PCV's penetrations that may become a arrangement. Reference to Sec.2.14.4 for.

b. Sec.2.14.4, SGTS.

1 potential source of radioactive release after SGTS and Sec.2.15.5 for HVAC functions.

c.. Sec.2.15.5, HVAC.

j an accident. (> ; 4 l 4 t J. - ) i I i 4 . i i 1 s f i i t 2' . i. 4 'O ~.. b'> i t 1- -A 0 J b e f' 'A. ua

ABWR DGsign Document O + -- A I .,j. W Jl AEY PLAN 1100 - 10$00 10500 0000 auto 10600 9500 - - 13a0 T M $ L ePnD f 9 i i '\\ e ? ? T M S L 38200 j l q pgs T M S L 31M 5 P-i i s -'j T M S L F5LD l l r T M t L Dico I 4 l P. 8 9 F ( l T M S L IS@ STD W

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1. ALL DIMENSIONS ARE N rrirn LNLESS OTHERWISE NDICATED.

O Figure 2.15.10a Reactor Building Arrangement - L*/180* 2.15.10 6/1/92

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1. ALL DIMENSIONS ARE IN mm UNLESS OTHERWISE INDICATED. '

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I Figure 2.15.10b Reactor Building Arrangement-270*/90' 2.15.10 -15. 6/1/92 m, .-_._-.._.._,..._,_..~.__.-2._

ABWR Design Documsnt y G e e e e e e p asco q. g-g -- 8500 % ~ 1300 10500 , sooo 8000 - 10500 1300 1300 1 1 i l 1 1 i t ...._s....... Q:s.....>;[4 ... s. y s.. -.> y >... - .@s... .. y. Q R, ~ ) f 3. b m .~J Clm CUVf CRO I l { -)~ '} --v.--- T,, 1 CUW I l RHR F /2Dh (A) 2.0. I; 'M3 'osa t e i SPCU l ..I.... .......... ){,...... l l RCIC 8000 CW 1 I 270' 010 f I 90' r. i .{ r ...l l I I / I 6000 }l ).... [. (fr-- l l HPCF(( 3 10500 l RHR@) N { RHR(C) I ......64, .,_p I 1 i CRO g j CRD 10500 g [~b 5 a _w n ... i,,.....gg.......gg.. . g..... q g.......g.......y-I i I i I I 1300 180* NOTES:

1. ALL DOORS HAVE RAISED S!LLS.

2. * *

• DENOTES WATERTIGHT DOORS TO J

4 PREVENT WATER ENTERING ROOMS FROM CORRIDORS.

3. COLUMN DIMENSIONS ARE 1.6m X 1.6m (TYPICAL).

4. FLOOR SLAB THICKNESS IS 0.5m.

S. MAIN BEAM DIMENSIONS ARE 1.4m X 1.8m.

6. AU. DIMENSIONS ARE IN mm UNLESS OTHERWISE INDICATED.

O Figure 2.15.10c Reactor Building Arrangement - Elevation 8200 mm 2.15.10 16-6/1/92

ABWR Design Document O

8 8 10500 m t 300 --

---

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1300 i i a v 0 s u . ;= f I ~ 9500 ps, r,, ,T - -,.,.7 T-3..... 3 3 CUW PE 10500 li. /[ O 5'. {..............j........... ~ v v .rW NOTES:

1. COLUMN DIMENSIONS ARE 2.0m X_2.0m (TYPICAL).

2.- FLOOR SLAB THICKNESS IS 0.6m.

3. MAIN BEAM DIMENSIONS ARE 1.5m X 1.8m.

4. ALL DIMENSIONS ARE IN mm UNLESS OTHERWISE INDICATED.

O Figure 2.15.10d ' Reactor Building Arrangement-Elevation 5100 mm 2.15.10 6/1/92

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I PREVENT WATER ENTERING ROOMS FROM CORRIDORS.

3. COLUMN DIMENSIONS ARE 1.6m X 1.6m (TYPICAL).

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n

5. MAIN BEAM DIMENSIONS ARE 1.4m X 1.8m

5. ALL DIMENSIONS ARE IN mm UNLESS OTHERWISE INDICATED.

O v Figure 2.15.10f Reactor Building Arrangement-Elevation 1500 mm 2.15.?O 19-6/1/92 d .i ~ _

I 1 ABWR D: sign Docum3nt O 8 8 8 6 6 8 6 1300 -. r 8$00 10500

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1300 7

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1. ALL DOORS HAVE RAISED SILLS.

2. COLUMN DIMENStONS ARE 1.8m X 1.8m (TYPICAL).

3. FLOOR SLAB THICKNESS IS 0.6m

4. MAIN BEAN DIMENSIONS ARE 1.2m x 1.8m

5. ALL DIMENSIONS ARE IN mm UNLESS OTHERWISE INDICATED.

O Figure 2.15.10g Reactor Building Arrangement-Elevation 4800 mm 2.15.10 -2 0-6/1/92

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3. MAIN BEAM DIMENSIONS ARE 1.5m X 1.8m.

4. ALL DIMENSIONS ARE IN mm UNLESS OTHERWISE

- INDICATED. : Figure 2.15.10h Reactor Building Arrangement -Elevation 8500 mm '2.15.10 -21 6/1/92: =______---:-_.

ABWR Design Docummnt O 8 8 8 8 8 1 i 1300 - 8500 10500 8000 8000 10500 8500 - 1300 1300 I r /1 [ O' L_ - - - - - - - - b5 . f.4 _ - UC,d[Lirf.........- [ SGTS MAN UG@) 9500 g-STEAM TUNNEL '] .]..... ..........] ]........ rm d! E ~. CUW:F CUW ,LJ _ _...f... .. 'r ? N e3 270' i PS(A) D .......-.} 028.8m i -i..........-.-. j..-.... .. g' r, .,.. 90- { 010.6m -E _= i PS(B) 1.7m PS(C) -_{..... p........ ... p.. Q ha 2.0m N 10500 b ..........{............p.[1 --- Q;'g............{}...........' ./ N nr M DG(B) 00(C) gg g R FM(B) F;S(C) ~ ,,r 'A i l j( i 1300 180' NOTES:

1. ALL DOORS HAVE RAISED SILLS.

2 * * ' DENOTES WATERTIGHT DOORS TO PREVENT WATER ENTERING ROOMS FROM CORRIDORS.

3. COLUMN DIMENSIONS ARE 1.6m X 1.6m (TYPICAL).

4. FLOOR SLAB THICKNESS IS 0.5m.

5. MAIN BEAM DIMENSIONS ARE 1.4m X 1Sm.

6. ALL DIMENSIONS ARE IN mm UNLESS OTHERWISWE INDICATED.

O Figure 2.15.101 Reactor Building Arrangement.-Elevation 12300 mm 2.15.10 22-6/1/92 1 1

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1. ALL DOORS HAVE RAISED SILLS.

2. * *

• DENOTESWATERTIGHT DOORSTO PREVENT WATER ENTERING ROCMS FROM CORRIDORS.

3. COLUMN DIMENSIONS ARE 2.0m X 2.0m (TYPICAL).

4. FLOOR SLAB THICKNESS IS 0.5tn

5. MP' BEAN DIMENSIONS ARE 1.4m X 1.8nt

6. ALL DIMENSIONS ARE IN rnm UNLESS OTHERWISE INDICATED O.

V Figure 2.15.10j Reactor Building Arrangement-Elevation 18100 mm 2.15.10 6/1/92

N$ E 1 2 5 1300-. 8500 - 0500 f: 10500 1300 a L a . ~. 4,' 9500 r, f p / w, l ,g j e . -w c : : _,7 _ ' j [' u h f '--_-y.", a G > f. _ ja y o Q _,* a i 'jl ' ppc, ' ', ) [/l l ,',s l l '[ ls' j ] 10 % 10500 [.j'I. _. _.', ] y J, l I l T.M.S.L 20800 T.M.S.L 21200 NOTES:

1. ALL DOORS HAVE RAISED SILLS.

2. * *

• DENOTES WATERTIGHT DOORS TO PREVENT WATER ENTERING ROOMS FROM CORRIDORS.

3. COLUMN DIMENSIONS ARE 1.4m X 1.4m fiYPICAL).

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4. FLOOR SLAB ThlCKNESS IS 0.5m.

5. MAIN BEAM DIMENSIONS ARE 1.5m X 1.8n.

6. ALL DIMENSIONS ARE IN mm UNLESS OTHERW,SE INDICATED.

Figure 2.15.10k Reactor Bui; ding Ar ement-Elevation 20800 and 21200 mm 9

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1. ALL DOORS HAVE RAISED SILLS.

2. * *

• DENOTES WATERTIGHT DOORS TO PREVENT WATER ENTERING RCWS FROM CORROGris.

3. COLUMN DIMENSIONS ARE 1.4 + 2.2 X 1.4m (TYPICAL)

4. FLOOR SLAB THICKNESS IS 0.5m.

5. MAIN BEAM DIMENSIONS ARE 1.2m X 1.8m.

6. ALL DIMENSIONS ARE IN mm UNLESS OTHERWISE INDICATED.

i i /3 l r a LJ l Figure 2.15.101 Reactor Building Arrangement-Elevation 23f,. .um 2.15.10 6/1/92

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1. ALL DOORS HAVE RAISED SILLS.

2. * *

• DENOTES WATERTIGHT DOORS TO THEVENT WATER ENTERING ROOMS FROM CORRIDORS.

3. COLUMN DIMENSIONS ARE 1.4 % I.2 X 1.4m(TYPICAL)

4. FLOOR SLAB THICKNESS IS 0.5m

5. MAIN BEAM DIMENSIONS ARE 1.2m X 1.8m

6. ALL DEMENSIONS ARE IN mm UNLESS OTHERWISE INDIOATED.

s 1 ~ Figure 2.15.10m Reector Building Arrangement--Elevation 27200 mni 2.15.10 -20 N1/92

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1. ALL DOORS HAVE RAISED SILLS.

2. * *

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3. COLUMN DIMENSIONS ARE 2.2m X 1.6m(TYPICAL).

4, FLOOR SLAB THICKNESS IS 0.5m.

5. MAIN STEEL H.SECTION BEAM DIMENSIONS ARE BH-1.5m X 0.7m,

6. ALL DIMENSIONS ARE IN mm UNLESS OTHERWISE INDICATED.

v Figuia 2.15.10n Reactor Building Arrangement--Elevation 31700 mm 2.15.10 6/1/92

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1. COLUMN DIMENSIONS ARE 22m X t.4m.

1. COLUMN DIMErO10PG ARC 1 Am M 1 Am(TYPCAL).

2. FLOOR SLAB THICKNESS IS O &m

2. ROOF THICKNESS IS 0 Am.

3 ALL DIMENSIONS ARE IN mm ULESS OTHfRWISE

5. WAIN BEAM DIMENNONS ARE BHA 6 X 0 46.

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4. ALL DIMENSIONS ARE IN m VNLISS OTHERWISE INDICATED.

O Figure 2.15.100 Reactor Building Arrangement-Elevation 34500 and 38200 mm l 2.15.10 G/1/92 l l

ABWR nosion Docwnant q 3,6 Human Factors EnD neerino i V Design Description The AP, Wit certified design's prima 7 uman system interfaces (11SI) will be h dneloped, designed, and enduated based upon a stnictmed top 4 town human Iactors systems analysis and shall reficct state-of-the-art lunnan f actors principles. The HS! scope will include operations, maintenance, test, and inspection interfaces, operations set huical procedmes, and training needs of the Main Connol Room and Remote Shutdown System functions and equipment. To assure integration of Iluman Factors Engineering (HFE) into the plant system design, the iISI design elTort will be directed by a multi <lisciplinary iIFE 1)esign Team comprised of personnel with expertise in llFE and in other ~hnical aieas relenmt to the 1ISI design, evaluation and operations. The llFE D% n Team will establish the methods which will implement the 11S1 design J..ough the process as shown in Figure 3Al. hnplementation of that process will be as follows: (1) Plant System requisements will be analyicd to identify those functions which must be perfonned to satisfy the objectives of each functional mea. System function analysis shall determine the objective, pe fonnance requirements, and constraints of the design; and establish the funct ons which must be accomplished to meet the objectives and i required perfonnance. (2) To facilitate an allocation of functions to the human which capitalire upon areas of human strengths and avoids areas of human limitations, a structuied and well<locumented method 4alogy of allocating functions to personnel, system elements, and personnel-system combinations will be established and implemented. (3) Task analysis will be conducted and used to identify the behavioral requirements of the tasks the personnel subsystem is required to perfonn in order to achieve the functions allocated to them. A task will be a group of activities that have a common purpose, often occurring in tempond proximity, and which utilire the same displays and controls. The task analysis will be used to maintain human perfonnance requirements within human capabilities; be used as an input for developing personnel skill, personnel training, and system l communication requirements and as an input to the evaluation of established plant opemtions control room stafling levels; and fonn the p basis for specifying the requirements for the displays, data processing V and controls needed to cany out tasks. 1 3.6 1-6/16/92

ABWR 0: sign Docum:nt (4) llurnan engineering principles and ciiteria will be applied in the design dennition and evaluation of the :luinan-Sptein Intedace (11S1). (5) 1" ant and erneigency operating technical procedur es will be developed to i ipport and guide huinan interaction with plant systeins and to support and guide human interactions in the control of plant operati ns. liurnan engineering principles and criteria shall be applied in the piv eduies developrnent. (6) Through the lunnan f actors verification and validation activities, the liSI design will be evaluated as an integrated systern using 11FE evaluation procedures, guidelines, standards, and principles. Inspections, Tests, Analyses and Acceptance Criteria Table 3.6 provides a definition of the inspections, tests, and/or analyses (together) with associated acceptance criteria) which will be perforrned to dernonstrate colnpliance with the HFE conninitinents for the certined design. O 9l 3.6 6/16/92 l

P b h h a i Table 3.G: Human Factors Engineering inspections, Tests, Analyses and Acceptance Criteria I Design Commitment inspections, Tests, Analyses Design Acceptance Criteria l l 1. Human. system interfaces (HSI) shall be 1. To assure the integration of HFE into plant 1.a. The HFE design team shall include the developed, designed, and evaluated based system (i.e., system) development: following expertise: l j upon a structured top-down human factors 1.a. A HFE Design Team shall be j systems analysis and shall reflect state-of-I' ""*9'**" established; and i the-art human factors pnnciples. The HSt j i shall include operations, maintenance, test, (2) Systems Engineering and inspection interfaces, operations 4 technical procedures, and training needs of (3) Nuclear Engineering the main control room and remote ~ shutdown system functions and (4) Control and instrumentation i equipment. Engineering i (5) Architect Engineering (6) Human Factors f i i (7) Plant Operations l '(8) Computer Systems Engineering i r (9) Plant Procedure Development l l l

00) Personnel Training 1.b. A HFE Program Plan shall be 1.b The Human Factors Engineering (HFE) established to assure the proper Program Plan shall establish:

I development, execution, oversight, and documentation of the human D) Human-System Interface (HSI) design [ factors engineering program. and evaluation methods and criteria l + which are consistent with accepted [ g HFE practices and principles. i k_ (2) The primary objectives of the HFE 1 Program shallinclude, at the minimum, l the objective to develop an HSI which I makes possible safe, efficient, and reliable operator performance. l

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_ __ ~ l l g Table 3.6: Human Factors Engineering (Continued) f l Inspections, Tests, Analyses and Acceptance Criteria l Design Commitment inspections. Tests, Analyses Design Acceptance Criteria 1. Continued 1.b. Continued (4) HSI design and evaluation scope which l consists of the Main Control Room and l Remote Shutdown System ) l operations, maintenance, test, and [ inspection interfaces, operating technical procedures, and identification i-of personneltraining needs. (5) The HFE Design Team as being responsible for: j t (i) the development of HFE plans and l procedures: (, (ii) the oversight and review of HFE [ design, development, test, and i j evaluation activities; l ? (iii) the initiation, recommendation, andprovision of solutions through I designated channels for problems i identified in the implementation of ths HFE activities-l f (iv) verification of implementa-tion of team recommendations. I (v) assurance that all HFE activities j-comply to the HFE plans and procedures, and (vi) scheduling of activities and gc milestones. I i P E r

g Table 3.6: Human Factors Engineering (Continued) Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Design Acceptance Criteria 1. Continued 1.b. Continued (6) The HFE Design Team having the authority and organizational freedom to accomplish its responsibilities. The team shall have the authority to determins where its input isrequired and to access work areas, and design documentation. The Team shall have the authority to control fu ther processing, delivery, installation or use of HFE/HSI products until the disposition of a non-conformatice, deficiency or unsa sfactory condition r has been achieved. 4 (7) An HFE issue tracking system which monitors the identification and cicsure of human factors issues. The HFE issue tracking syrtem shall document and track huma7 factors engineering issues and concems, from identification until elimination or reduction to a level acceptable to the HFE Design Team. (8) The Design Control procedures through which the results of the iterative design development activities are documented and processed to maintain integration of design activities and assure that the design, design analyses and documentation are consistent and appropriately reflect the details of design implementation 3 decisions. N O O O

r Ix l g Table 3.6: Human Factors Engineering (Continued) I Inspections, Tests, Analyses and Acceptance Criteria ] Design Commitment inspections, Tests, Analyses Design Acceptance Criteria 2. Plant System requirements shall be 2.a. A System Functional Requirements 2.a. The 3ystem Functional Requirements analyzed to identify those fur.ctions which Analysis implementation Plan shall be Analysis implementation Plan shall i must be performed to satisfy the objectives developed to assure that tne analysis is establish: I of each functior al area. System function conducted according to accepted HFE analysis shall determine the objective, principles. (1) Methods and criteria for con-performance requirements, and ducting the System Functional constraints of the design; and (2) establish Requirements Analysis which are the functions which must be accomplished con sistent with accepted HFE to meet the objectives and re luired practices and principles. performance. (2) that system requirements shall define the system functions and those system functions shall pro-vide the basis for determining the id associated HS1 performance requirements. (3) that critical functions shall be defined (i.e., those functions required to achieve major system performance requirements; or those functions which,.f failed, could oose a safety hazard to plant i personnel or to the general public), i 4 1 OB u t

g Table 3.6: Human Factors Engineering (Continued) Inspections, Tests, Analyses and Acceptance Criterie Design Commitment inspections, Tests, Analyses Design Acceptance Criteria 4 2. Continued 2.a. Continued (4) that safety functions shall be identified along with any func-tional interrelationship those safety functions may have with non-safety systems. (5) that functions shall be defined as the most genera!, yet differentiable means whereby the system requirements are met, discharged, or satisfied. Functions shall be arrangeo in a logical sequence so that any specified operational usage of the system can be traced v in an end-to-end path. (6) that functions shall be described initially in graphic form. Function diagramming shall be done starting at a " top level", where major functions are described, and continuing to decompose major functions ta lower levels unt3 a specific entical end-item requirement emerges, e.g., a piece of equipment software, or an operator R ~ 9 9 9

O i g Table 3.6: Human Factors Engineering (Continued) Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Design Acceptance Criteria 2. Continued 2.a. Continued (7) that detailed narrative descriptions d shall be developed for each of the identified functions and for the overall system configuration design itself. Each function shall be i identified and described in terms of inputs (observable parameters which wil? indicate system status) l functional processing (control l process and performance measures required to achieve the function), functional operations (including detecting signals, measuring information, comparing h one measurement with another, processing information, and acting upon decisions to produce a desired condition or result such as a system or component operation actuation or trip) outputs, feedback (how to determine correct discharge of function), and interface requirements from the top down so that subfunctions are I-recognized as part of larger functional elements. 2.b. An analysis of system functional require-2.b.The system functional requirements ments shell be conducted in accordance analyses shall be conducted in l with the System Functional Requirements accordance with the requirements of Analysis implementation Plan and the the Human Factors Engineering findings will be documented in System Program Plan and the System p j Functional Require ments Analysis Results Functional Requirements Analysis g Report. The analyses of the system implementation Plan. functional requirements shall be reviewed by the HFE Design Team and shall be documented in System Functiona! l l Requirements Analysis Evaluation Report. t j

g Table 3.6: Human Factors Engineering (Continued) Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Design Acceptance Criteria 3. To facilitate an allocation of functions to ?.a. An Allocation of Function implementation 3.a.The Allocation of Function the human which capitalize upon areas of Plan shall be developed to assure that the implementation Plan shall estabMsh: human strengths and avoids areas of allocation of function is conducted human limitations, a structured and well-according to accepted HFE principles. (1) The methods and criteria for the documented methodology of allocating execution of function a :ocation functions to personnel, system elements, which are consistent with accepted and personnel-system combinations shall HFE practices and principles. be established and implemented. (2) That all aspects of system and functions definition shsil be analyzed in terms of resulting human performance requirements based on the expected user popu?ation. (3) That the allocation of functions to personnel, system elements, and personnel system combinations shall reflect: (i) sensitivity, precision, time, and safety requirements, (ii) required reliability of system performance, and (iii) the number and the necessary skills of the personnel required to operate and maintain the system. g (4) The allocation criteria, rationale, analyses, and procedures shall be w 3 documented. O O O

l 4 L g Table 3.6: Human Factors Engineering (Continued) Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Design Acceptance Criteria 3. Continued 3.a. Continued (5) Analyses shall confirm that the personnel elements can correctly i perform tasks allocated to them I while maintaining operator l i situation awareness, acceptable i personnel workload, and facilitating personnel vigilance. I 3.b. An analysis of the allocation of function 3.b.The function allocation analyses shall be l l shall be conducted in accordance with the conducted in accordance with the l i Allocation of Function Implementation Plan requirements of the Human Factors j and the findings will be documented in an Engineering Program Plan and the j Allocation of Function Analysis Results Allocation of Functions Implementation l } Report The analyses of the allocation of Plan. 5 function shall be reviewed by the HFE 2 Design Team and the results of that review i l shall be documenad in an I l Aflocation of Function Evaluation Report. l i l l I I I r 5 i .~ I 4

g Table 3.6: Human Factors Engineering (Continued) inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Design Acceptance Criteria 4. Task analysis shall be conducted and used 4.a. A Task Analysis implementation Plan shall 4.a.The Task Analysis implementation Plan to identify the behavior 61 requirements of be developed to assure that the analysis is shall establish: the tasks the personnel subsystem is conducted according to accepted HFE required to perform in order to achieve the principles. (1) The methods and criteria for conduct of I funct.ons allocated to them. A task shall be the task analyses which are consistent a group of activities that have a common with accepted HFE practices and purpose, often occurring in temporal principles. ,roximity, and which utilize the same displays and controls. The task analysis (2) The scope of the task analysis which shall be used to maintain human shall i iclude all operations performed performance requirements within human at the aperator interface in the main capabilities; be used as an input for control room and at the remote developing personnel skill, personnel shutdown system. The ana!yses sha!' training, and system communication be directed to the full range of plant y requirements and as an input to the operating modes, including startup, evaluation of established p' ant operations normal operations, abnormal control room staffing levels; and form the operations, transient conditions, low basis for specifying the require-ments for power and shutdown conditions. The the displays, data processing and controls analyses shall also address operator needed to carry out tasks. interface operations during periods of maintenance test and inspection of plant systems and equipment and of the HS1 equipment. (3 ) That the analysis shall link the identified and described tasks in operational sequence diagrams. The task descriptions and operational sequence diagrams shall be used to identify which tasks are " critical" in terms ofimportance for function e achievement, potential for human g error, and impact of task failure. Human actions which are found to affect plant risk in PRA sensitivity analyses shall also be considered

• critical."

O O O

O l ) g Table 3.6: Human Factors Engineering (Continued) i Inspections, Tests, Analyses and Acceptance Criteria Design Commitment Inspections, Tests, Analyses Design Acceptance Criteria 4 4. Continued 4.a. Continued (4) Task analysis shall begin with the development of detailed narrative i descriptions of the personnel activities required for successful completion of the task. Task analyses shall define the [ input, process, and output required by and of personnel. l (5) The task analysis shall be in detail [ sufficient enough to identify l information and control requirements such that require-ments for alarms, displays, data processing, and controls for human task accomplishment may i be specified. (6) The task analysis results shall be made f 4 i. available as input to the personnel i training programs. l ? l 4 b. An analysis of tasks shall be conducted in 4.b.The task analyses shall be conducted in i l accordance with the Task Analysis accordance with the requirements of the implementation Plan and the findings will Human Factors Engineering Program Plan be documented in a Task Analysis Results and the Task Analysis implementation J 1 Report. The task analyses shall be Plan. I reviewed by the HFE Design Team and the j results of that review shall be documented in a Task Analysis Evaluation Report. k' l 4 c B I 1 PJ l i i [ i 4

g Table 3.6: Human Factors Engineering (Continued) Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections Tests, Analyses Design Acceptance Criteria 5. Human engineering principles and criteria 5.a. A Human-System Interface (HS!! Design 5.a. The HS! Design Implementation Plan sha!! shall be applied in the design definition Implementation Plan shall be developed to establish: and evaluation of the Human-System assure that human factors analyses of the (1) The methods and criteria for HS1 Interf ace (HSI). HSI Design are conducted according to equipment design; and evaluation of accepted HFE principles. HS! human performance, equipment design and associated workplace factors; which a e consistent with accepted HFE practices and pcinciples.

2) That the HS1 design shall imple-ment the information and control requirements developed through the task analyses, including the displays, controls and alarms necessary for the p

execution of those tasks identified in the task analyses as being critical tasks. (3) The methods which will assure thatthe HS1 human performance, equipment design and associated workplace factors are consistent with those modeled and evaluated in the completed task analysis. t (4) That the HSI design shall not incorporate any equipment (i.e., hardware or software function)which has not been specificallyevaluated in the task analysis. $e8 e O O

b e F g Table 3.6: Human Factors Engineering (Continuad) Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Design Acceptance Crateria 5. Continued 5.a. Continued (5) The HS! design criteria and guid-ance f for control room operationsduring periods of maintenance, test and inspection ofcontrol room HS1 i l equipment and of other plant equipment which has control room personnel interface. (6) The test and evaluation methods for resolving HFE/HSI design issues. These test and evaluation methods shall include the criteria to be used in i selecting HFE/HS1 design and evaluation tools which: 4 T. l (i) may incorporate the use of static mockups and models for evaluating access and workspace related HFE issues, and (ii! shall require dynamic simulations j' and HSI prototypes for conducting evaluations of the human performance associated with the activities in the critical tasks identified in the task analysis. 4 5.b. An ana!ysis of the human-system interface 5.b.The Human System Interface (HST) Design ' design shall be conducted in accordance Analyses shall be conducted in accordance with the HSI Design Implementation Plan with the requirements of the Human and the findings will be documented in an Factors Engineering Program Plan and the g a HSI Design implementation Analysis HSI Design implementation Plan. 3; Results Report. The analyses of the HSI l Design !mplementation shall be reviewed by the HFE Design Team and the results of that review shall be documented in an HSI i Design implementation Evaluation Report. I

g Table 3.6: Human Factors Engineering (Continued) Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Design Acceptance Criteria 6. Plant and Emergency Operating 6.a. A Plant and Emergency Operating 6.a. The Plant and Emergency Operating Procedures shall be developed to support Procedure Development Procedure Development implementation and guide human interaction with plant Implementation Plan shall be Plan shall establisn: systems and to support and guide human developed to assure that the interactions in the control of plant development of the Plant and (1) That operator actions identified in the operations. Human engineering principles Emergency Operating Procedures is task analysis shall be used as the basis and criteria shall be applied in the conducted according to accepted HFE for specifying the procedures for procedures development. principles. operations. (2) That the procedures to be devel-oped shall address normal, abnormal, and emergency plant operations including consideration of plant operations during periods when plant systems / equipment and primary operator interface (i.e., main control room) equipment is undergoing. test, maintenance or inspection. (3) Methods and criteria for develop-ment of the operating technicalproceduces which are consistentwith accepted HFE practices and principles. (4) That a Writer's Guide shall be developed which establishes the process for developing the techni-cal procedures for normal plantand system operation, abncrmal plant operations, emergency plant operations and for responding to plant g alarm conditions. The Writer's Guide 5 shall contain objective criteria which 8 wi!! require that the operations technical procedures developed are consistent in organization, style, content and usage of terms. O O O

O 4 - g Table 3.6: Human Factors Engineering (Continued) l Inspections, Tests, Analyses and Acceptance Criteria i 2 i Design Commitment inspections, Tests, Analyses Design Acceptance Criteria 1-i 6.a. Continued 6.b. The Piant and Emergency Operating 6 b.The development of the plant operations i j Procedures shall be developed in technical procedures shall be conducted in i accordance with the Plant and Emergency accordance with the requirements of the Operating Procedure Development Human Factors Engineering Program Plan Implementation Plan and the results will be and the Plant and Emergency Operating documented in a Plant and Emergency Procedure Development implementation Operating Procedure Development Plan. Report. The Plant and Emergency i Operating procedure development results shall be reviewed by the HFE Design Team and the results of that review shall be documented in a Plant and Emergency Operating Procedure Development Evaluation Report. O l 7. The HSI design shall be evaluated as an 7.a. A Human FactorsVerification and 7.a.The Human Factors Verification and i i integrated system using HFE evaluation Validation implementation Plan shall be Validation (V&V) Implementation Plan procedures, guidelines,. standards. and developed to assure that the evaluation of shall establish: principles. the integrated HSI Design is conducted in accordance with accepted HFE principles. (1) Human factors'J&V methods and criteria which are consistent with j accepted HFE practices and principles. l-(2) The methods and evaluat on criteria 4 which are consistent with accepted HFE practices and principles. s 5 g ea (

g Table 3.6: l'uman Factors Engineering (Continued) Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Design Acceptance Criteria 7. Continued 7.a. Continued (3) The scope of the evaluations of the integrated HSI shall include: (i) The Human-Systern Interface (including both the interface of he operator with the HSI equipment hardware and the interface of the operator with the HSI equipment's software driven functions) (ii) The plant and emergency operating technical procedures,and (iii) The overall HS1 work environment P (4) That static and/or "part-task ~ mode evaluations of the HSI equipment shall be conducted to confirm that the controls, displays, and data processing functions identified in the task analyses are provided and that those controls, displays and data processirsg functions are designed in accordance with accepted HFE practices and principles. 3 o O O O

m __ _ O i j g Table 3.6: Human Factors Engineering (Continued) I Inspections, Tests, Analyses and Acceptance Criteria l '.mQn Commitment inspections, Tests, Analyses Design Acceptance Criteria 7. Continued 7.a. Continued (5) The integration of HS1 equipment with each other, with the operating l 4 personnel and with the Operations Technical Procedures shallbeevaluated l through the conduct of dynamic task perform-ance testing. The dynamic taskperformance testing and evalua-tions shall be performed over the full scope of the integrated HSI design using dynamic HSI proto-types (i.e., prototypical HSI equipment which is dynamically driven by rea! time plant I simu-lation computer models), other evaluation tools and/or past dynamic task performance test and evaluation results. The methods for defining the scope and application of the dynamic HS1 prototype, past test results and I other evaluation tools shall be j documented in the implementation plan. The dynamic task perform-ance i tests and evaluations shall have astheir i objectives: j (i) Confirmation that the inte-grated HSI design facilitates achievement of the identi-fied safety functions and critical functions, (ii) Confirmation that the alloca-tion of i function and the structure of tasks g, assigned to personnelis consistent l 5; with accepted HFE principles. l l i-5 l

g Tab'e 3.6: Human Factors Engineering (Continued) Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Design Acceptance Criteria 7. Continued 7.a. Continued (iii) rmfirmation of established main witrol room staffing and the HSI design and configuration provided to support that staff in accomplishing their assigned tasks, (iv) Confirmation that Operations i Technical Procedures are complete and accurate. (v) Confirmation that the dynamic aspects of the HS1 are sufficient for task accomplishment, and h (vi) Confirmation that the integrated HSI design is condusive to eliminating the potential for operator errors. (6) That dynamic task performance test evaluations shall be conducted over the fu!I range of operational conditions and upsets. including: (i) Normal plant operations, such as plant startup, shutdown, full power operations, and plant maintenance activities; (ii) Plant system and equipment failures; F2 si (iii) HSI equipment failures: 8 (iv) Plant transients, and; (v) Postulated plant accident G G conditions.

i O O J g Table 3.6: Human Factors Engineering (Continued) t inspections, Tests, Analyses and Acceptance Criteria j Design Commitment inspections, Tests, Analyses Desir,a Acceptance Criteria 7. Continued 7.a. Continued (7) The HFE performance measuresto be used as the basis for evalu-atin < the dyna mic task perform-a nce test e esuiN. l These per-formance measures shat! include: (i) Operating crew primary task performance characteristics, such as task times and procedure violations, (ii) operating crew errors and/or error t

rates, h

(iii) cperating crew situation awareness, (iv) operating crew workload. (v) operating crew communications and coordination, (vi) anthropometry evaluations, i P P -M i

g Tab!e 3.6. Muman Factors Engineering (Continued) Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Design Acceptance Criteria 7. Continued 7.a. Continued (vii) physical pos tioning and interactions, and (viii) HSI equipment performance measures (8) The methods to confirm that HFE issues identified and documented in the Human Factors issue Tracking System have been resolved in the integrated HSI design, and (9) The methods and criteria to be used to confirm that critical human actions, as y defined by the task analysis, have been addressed in the integrated HSI design in a manner consistent with accepted HFE practices and principles. (10) The methods and criteria tc be used to confirm that the operating technical procedures are correct and can be executed within the realm of accepted human performance capabilities. i E E is e e O

~ l' g Table 3.6: Human Factors Engineering (Continued) Inspections, Tests, Analyses and Acceptance Criteria ): Design Commitment inspections, Tests, Anolyses Design Acceptance Criterie 1 7. Continued 7.b. A human factors engineering analysis of 7.b. The human factors verification and the integrated HSI design shall be validation (V&V) of the human system ( conducted in accordance with the Human interface (HSI) design shall be con-ducted Factors Verification and Vali-dation in accordance with the requirements of the Implementation Plan and the findings will Human Factors Engineuring Program Plan I be documented in Human Factors and the Human Factors V&V Verification and Validation Results Report. Implementation Plan. The analyses of the integrated HSI design - shall be reviewed by the HFE Design Team 1 and the results of that review shall be ]- documented in Human Factors Verifi-cation and Validation Evaluation Report. h i fL 1 I i i ^ ? i I 4 m l-l i l f }: & - - a e y .m n

i m l SYSTEM FUNCTIONAL REQUIREMENTS DEFINITION g_ _ __ _ _ _ q l y I ALLOCATION OF FUNCTIONS I I if I TASK ANALYSIS I Ll I 1r "UMA" YST ,g7 gpACE 9ESIGN - ~- A l N 1r i HUAN FACTORS VERIFIC/ ' t 'ND VAllDATION ~ ~ ~ ~ ~ - ~# I 1r IMPLEMENTED 5 DESIGN m i Figure 3.6 Human System in ace Design implementation Process

ABWR 0: sign Docum:nt Table 5.0: ABWR Site Parameters Maxirnum Ground Water Lesel 2 feet below grade Extr eme Wimh Itam Wind Speed: 110 mph'U/130 mph A Maximurn flood (or Tsunami) LevelW: Tor nadoA 1 foot below giade

• Maxismuu tornado wimi y>ced:

260 mph

• Iramlatmnal ulodtv 57 mph
• R,idim:

4531t

• Maxumun atm AP:

L4f> psid

• Miuile Spectra:

Per ANSI /ANS-2.3 Precipitat or (for Roof Design): Soil Properties:

• Maximiun rainfall rate:

19A in/h: 0)

• Minimum licating Capacity (demand):

JWI

• Maxinuun snow load:

50 lb/sq. It_

• Mininnun Shear Wave Velocity:

It KK)t ps"')

• liquilication Potential:

None at plant 3ite resulting irom OliE and SSE. Design Teruperatures: Seistnology:

• Arnbient
• OllE Peak Ground Acceleration (PGA):

N "' 1"E Excedmrg_i';thto 0.10g W Maumiun: 1(KfF dry bulb /

• SSE PGA:

0.3Og 77*F coir.cident wetbulb + SSE Respome Spectra: per applicable regulations Minnmun: -10*F

• SSETime Ilistory:

Envelope SSE Response Q3Accedance Vahtes (Ilistorical IJnlin Spectra ( Maximmn: 115'F dry bulb / \\ 82*F coincident wet bulb Minimum: -4 W F

• Ernergency Cooling Wate Inlet:

95+F

• Condemer Cooling Water Inlet:

s100 F (1) 50-year recurrence interval;value to be utilized for design of non-safety-related structures only. (2) 100-year recurrence interval; value to be utilized for design of safety-related structures only (3) Probable maximum flood level (PMF), as defined in ANSI /ANS-2.8, " Determining Design Basis Flooding at Power Reactor Sites." (4) 1,000,000-vear tornado recurrence interval, with associated parameters based on ANSI /ANS-2.3. (5) Maximum <alue for 1 hour 1 sq. mile PMP with ratio 5 miniutes to 1 hour PMP ao found in National Weather Source Publication HMR No. 52. Maximum short term rate; 6.2in/5 min. (6) This is the minimum shear wave velocity at low storms after the soil property uncertainties have been applied. (7) Free-field, s.: plant grade elevation. (B) For conservatism, a value of 0.15g is employed to evaluate structural and component responses of the certified design. (9) Free field, at plant grade elevation. d l l G/1/92 l .}}