ML22112A037
| ML22112A037 | |
| Person / Time | |
|---|---|
| Site: | LaSalle, 05200002 |
| Issue date: | 01/31/1997 |
| From: | ABB Combustion Engineering |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20148A597 | List:
|
| References | |
| NUDOCS 9705090171 | |
| Download: ML22112A037 (1) | |
Text
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O Copyright C 1997 Combustion Engineering, Inc.,
All Rights Reserved.
Warning, Legal Notice and Disclaimer of Liability The design, engineering and other information contained in this document have been prepared by or for Co abustion Engineering, Inc. in connection with its application to the United States Nuclear Regulatory Commission (US NRC) for design certification of the System 80+ nuclear plant design pursuant to Title 10, Code of Federal Regulations Part 52. No use of any such information is authorized by Combustion Engineering, Inc.
except for use by the US NRC and its contrac' ors in connection with review and approval of such application. Combustion Engineering, Inc. hereby disclaims all responsibility and liability in connection with unauthorized use of such information.
Neither Combustion Engineering, Inc. nor any other person or entity makes any warranty or representation to any person or entity (other than the US NRC in connection with its review of Combustion Engineering's application) conceming such information or its use, .
except to the extent an express warranty is made by Combustion Engineering, Inc. to its customer in a written contract for the sale of the goods or services described in this document. Potential users are hereby wamed that any such information may be unsuitable for use except in connection with the performance of such a written contract by Combustion Engineering, Inc.
Such information or its use are subject to copyright, patent, trademark or other rights of Combustion Engineering, Inc. or of others, and no license is granted with respect to such rights, except that the US NRC is authorized to make such copies as are necessary for the use of the US NRC and its contractors in connection with the Combustion Engineering, Inc. application for design certification.
Publication, distribution or sale of this document does not constitute the performance of engineering or other professional services and does not create or establish any duty of care towards any recipient (other than the US NRC in connection with its review of Combustion Engineering's application) or towards any person affected by this document.
I For it formation address: Combustion Engineering, Inc., Nuclear Systems Licensing, 1 2000 Day Hill Road; Windsor, Connecticut 06095 ;
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System 80+ nesian contratDocument 7'N i Q
Introduction I
Certified Design Material l
1.0 Introduction j 2.0 System and Structure ITAAC j l 3.0 ' Non-System ITAAC i l 4.0 Interface Requirements l 5.0 Site Parameters ;
Approved Design Material - Design & Analysis l
1.0 General Plant Description 2.0 Site Ch,tracteristics 3.0 ' Design of Systems, Structuces & Components l 4.0 Reactor 5.0 RCS and Connected Systems 6.0 Engineered Safety Features 7.0 Instrumentation and Control 8.0 Electric Power 9.0 Auxiliary Systems 10.0 Steam and Power Conversion 11.0 Radioactive Waste Management l(
12.0 13.0 Radiation Protection Conduct of Operations l
14.0 Initial Test Program 15.0 Accident Analyses -
16.0 Technical Specifications 17.0 Quality Assurance i 18.0 Human Factors 19.0 Probabilistic Risk Assessment 20.0 Unresolved and Generic Safety Issues l
Approved Design Material - Emergency Operations Guidelines 1.0 Introduction ;
2.0 Standard Post-Trip Actions 3.0 Diagnostic Actions 4.0 Reactor Trip Recovery j 5.0 Loss of Coolant Accident Recovery 6.0 Steam Generator Tube Rupture Recovery 7.0 Excess Steam Demand Event Recovery 8.0 Loss of All Feedwater Recovery 9.0 Loss of Offsite Power Recovery 10.0 Station Blackout Recovery 11.0 Functional Recovery Guideline
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1 Svstam 80+ Desinn controlOceament E l
2.8 ' Steam and Power Conversion System l
- 2J.1 Turbine Generator i
Design Desedption ;
The Turbine Generator is a non-safety-related system that convens 'the energy of the steam produced in
- - the steam generators into mechanical shaft power and then into electrical energy. ,
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The Turbine Generator is located on the turbine pedestal in the turbine building.
i The Basic Configuration of the Turbine Generator is as shown on Figure 2.8.1-1.
- The flow of steam is directed from the steam generators to the turbine through th mah steam l
, . equalization header, stop valves, and control valves. After expamling through the turbine, which drives l the main generator, exhaust steam is transported to the main comlenser, j
' Turbine Generator operation is monitored and controlled by the turbine control and safety systems. Tre control and safety systems provide overspeed protection, speed and load control, and trip of the Turi,ine !
Generator by closing the turbine stop valves and control valves. The Turbine Generator trips in response to a reactor trip.
' Controls exist in the main contrr) w>om (MCR) to manually trip the Turbine Generator. ;
The Turbine Generator has an electronic overspeed trip device and a mechanical overspeed trip device. i
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' The Turbine Gland Scaling System (TGSS) is a non-safety subsystem that provides sealing at leakage l
, poims in the Turbine Generator system using sealing steam as a pressurizing medium. ;
I Inspections, Tests, Analyses, and Acceptance Criteria j
. Table 2.8,1-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Turbine I Generator. 1 a 1
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System 80+ oesign controt Document Table 2.8.1-1 Turbine Generator Design Commitment Inspections, Tests, Analyses AW-e Criteria
- l. The Basic Configuration 1. Inspection of the as-built 1. For the components and of the Turbine Generator Turbine Generator equipment shown on is as shown on Figure configuration will be Figure 2.8.1-1, the as-2.8.1 1. conducted. built Turbine Generator conforms with the Basic Configuration.
- 2. The Turbine Generator 2. Testing will be 2. The Turbine Generator trips in response to a performed on the Turbine control system generates reactor trip. Generator using a signal a turbine trip signal in simulating reactor trip. response to a signal simulating reactor trip.
- 3. Controls exist in the 3. Testing will be 3. Turbine Generator MCR to raanually trip performed using the controls in the MCR the Turbine Generator. Turbine Generator operate to manually trip contrets in the MCR. the Turbine Generator.
4 The Turbine Generator 4. Inspection of the Turbine 4. The Turbine Generator i has an electronic Generr. tor over-speed trip electronic and overspeed trip device and devices will be mechanical overspeed a mechanical overspeed conducted. trip devices are installed.
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n Cornhed Dee> Meenriel Page 2.8-3
SYtem 80+ Design ControlDocument 2.8.2 Main Steam Supply System Design Description The Main Stearn Supply System (MSSS) transports steam from the steam generators to the power conversion system and removes heat from the reactor coolant system (RCS).
The Basic Configuration of the MSSS is as shown on Figure 2.8.2-1. The safety-related portion of the MSSS consists of the main steam piping and valves located between the steam generator outlet nozzles in the Containment up to and including the main steam isolation valves (MSIVs) in the main steam valve L mes (MSVHs). The non-safety-related portions of the MSSS are located in the yard and the turbine building. Downstream of the isolation valves in the M. cVHs, the MSSS lines exit the MSVHs and cross above the yard into the turbine building.
The MSSS supplies steam to the emergency feedwater pump turbines and provides a flow path to the turbine bypass system.
The main steam safety valves (MSSVs) provide overpressure protection for the secondary side of the steam generators and for pressure boundary components in the MSSS.
The ASME Code Section III Class for the MSSS pressure retaining components shown on Figure 2.8.2-1 is as depicted on the Figure.
The safety-related equipment shown on Figure 2.8.2-1 is classified Seismic Category I.
Within a mechanical Division, the electrical controls for the atmospheric dump valve (ADV), MSIV, and MSIV bypass valve on each main steam line from a steam generator are powered from their respective .
Class IE Division.
Within a mechanical Division, the ADV blo;k valve on each main steam line from a steam generator is powered from a Class IE Division different from the Class IE Division of the respective mechanical Division.
Independence is provided between Class IE Divisions, and between Class IE Divisions and non-Class IE equipment, in the MSSS.
Controls exist in the main control room (MCR) to open and close those power operated valves shown on Figure 2.8.2-1.
The safety-related portions of the MSSS mechanical Divisions are physically separated.
Valves with response positions indicated on Figure 2.8.2-1 change position to that indicated on the Figure upon loss of motive power.
The MSIV and MSIV bypass valves close on receipt of a main steam iso!ation signal (MSIS).
A radiation monitor is installed for one steam line from each steam generator to monitor primary-to-secondary leakage.
Certified Design Matenef Page 2.8-4
System 80+ Deskn controlDocument ,
l Inspections, Tests, Analyses, and Acceptance Criteria Table 2.8.2-1 specifies the inspections, tests, analyses, and assu:iated acceptance criteria for the Main '
Steam Supply System. ,
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- 2. ASME CODE SECT 10N N CLAst COMPONENTS SHOWN ON THE PIGURE AIE SAPETY.RELATED.
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- Page 2.8-6
4 System 80+ oesion controlDocument b
w/ Table 2.8.2-l ' Main Steam Supply System Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 1. The Basic Configuration 1. Inspection of the as-built 1. For the components and of the MSSS is as shown MSSS configuration will equipment shown on on Figure 2.8.2-1. be conducted. Figure 2.8.2-1, the as-built MSSS conforms with the Basic Configuration.
- 2. The ASME Code Section 2. A pressure test will be 2. The results of the 111 MSSS components conducted on those pressure test of the shown on Figure 2.E.2-1 components of the MSSS ASME Code Section 111 retain their pressure required to be pressure components of the MSSS boundary integrity under tested by ASME Code conform with the ;
internal pressures tha' Section III. pressure testing will be experienced acceptance criteria in j during service. ASME Code Section 111.
- 3. Controls exist in the 3. Testing will be 3. MSSS controls in the MCR to open and close performed using the MCR open and close those power operated MSSS controls in the those power operated valves shown on Figure MCR. valves shown on Figure 2.8.2 1. 2.8.2-1.
- 4. The MSSVs provide 4 Testing and analysis in 4. An ASME code test and overpressure protection accordance with ASME analysis repon exists that for the secondary side of Code Section 111 will be concludes that the MSSV the steam generators and performed to determine set pressures are set such for pressure boundary set pressure. Type tests that the maximum components in the MSSS. and analyses of flow allowable pressure in the capacity of each SSV will steam generator does not be performed. exceed 110% of the steam generator design pressure, and that the total MSSV capacity is sufficient to pass 19 x 106 lbJhr. at 110% of the steam generator design pressure.
- 5. Within a tnechanical 5. Testing will be 5. Within the MSSS, a test Division, the electrical performed on the MSSS signal exists only at the controls for the ADV, by providing a test signal equipment powered from MSIV, and MSIV bypass in only one Class IE the Class IE Division valve on each main steam Division at a time. under test.
line from a steam generator are powered from their respective c . Class IE Division.
Cerened Deodon Meenrin! Page 2.8-7 l
1 Syntem 80+ Design ControlDocument Table 2.8.2-1 Main Steam Supply System (Continued)
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 6. Within a mechanical 6. Testing will be performed 6. Within the MSSS, a test Division, the ADV block on the'MSSS by providing signal exists only at the valve on each main steam a test signalin only one equipment powered from line from a steam Class IE Division at a the Class IE Division generator is powered from time. under test.
a Class IE Division different from the Class 1E Division of the respective Mechanical Division.
- 7. Independence is provided 7. Inspection of the as- 7. Physical separation exists between Class IE installed Class IE between Class IE Divisions, and between Divisions of the MSSS Divisions in We MSSS.
Class 1E Divisions and will be performed. Physical separation exists non-Class IE equipment, between Class IE in the MSSS. Divisions and non-Class IE equipment in tbe MSSS.
- 8. The safety-related portions 8. Inspection of the as-built 8. The safety-related portions of the MSSS mechanical mechanical Divisions will of the MSSS mechanical Divisions are physically be performed. Divisions are located in separated. two MSVIIs that are separated by the Containment.
Components of the system within the Containment are separated by spatial )
arratigement or barriers. I 1
- 9. Valves with response 9. Testing ofloss of motive 9. These valves change l positions indicated on power to these valves will position to the position I Figure 2.8.2-1 change be performed. indicated on Figure 2.8.2- )
position to that indicated I upon loss of motive )
on the Figure upon loss of power.
motive power.
10.a) The MSIVs close on 10.a) Testing will be performed 10.a) The MSIVs close within 5 receipt of a MSIS. using a signal simulating a sec'onds of receipt of a MSIS. signal simulating a MS!S.
10.b) The MSIV bypass valves 10.b) Testing will be performed 10.b) The MSIV bypass valves !
close on receipt of a using a signal simulating a close within 10 seconds of MSIS. MSIS. receipt of a signal I simulating a MSIS.
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- 11. A radiation monitor is 11. Inspection of the as-built 11. An N-16 detector is j installed for one steam line radiation monitors for the installed for one steam ;
from each steam generator steam lines will be line from each steam !
to monitor primary-tn. performed. generator. j secondary leakage. l I
Certihed Design acoteriel Page 2.8-8 j
System 80+ Design ControlDocument
( 2.8.3 Main Condenser Design Description The Main Condenser is a non-safety-related component that converts the turbine exhaust steam to condensate so it can be pumped back through the condensate and feedwater systems to the steam generators. The condenser circulating water system provides cooling water to the Main Condenser tubes to condense the exhaust steam from the main turbine. The Main Condenser also serves as a collection point for auxiliary equipment vents and drains, and for condensate and feedwater system makeup. The main condenser is located in the turbine building. ,
The Basic Configuration of the Main Condenser is as shown on Figure 2.8.3-1.
Displays of the Main Condenser instrumentation shown on Figure 2.8.3-1 exist in the main control room (MCR) or can be retrieved there.
- A turbine trip signal is generated in response to signals from the Main Condenser pressure instrumentation on loss of condenser vacuum.
Inspections, Tests, Analyses, and Acceptance Criteria t
Table 2.8.3-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Main Condenser.
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Design Commitment Inspections, Tests, Analyses Acceptance Criteria ,
- 1. The Basic Configuration 1. Inspection of the as-built 1. For the components and of the Main Condenser is Main Condenser equipment shown on i as shown on Figure configuration will be Figure 2.8.3-1, the as-2.8.3 1. conducted. built Main Condenser conforms with the Basic Configuration.
- 2. Displays of the Main 2. Inspection for the 2. Displays of the Condenser existence or instrumentation shown on instrumentation shown on retrieveability in the Figure 2.8.3-1 exist in Figure 2.8.3-1 exist in MCR of instrumentation the MCR or can be the MCR or can be displays will be retrieved there.
retrieved there. performed.
- 3. A turbine trip signal is 3. Testing will be 3. A turbine trip signal generated in response to performed to verify from the Main signals from the Main generation of a tutbine Condenser pressure Condenser pressure trip signal using signals instrumentation is instrumentation on loss of from the Main Condenser activated on receipt of a condenser vacuum. pressure instrumentation signal simulating loss of which simulate a loss of condenser vacuum.
. condenser vacuum.
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System 80+ Design ControlDocument 2.8.4 Main Condenser Evacuation System Design Description The Main Condenser Evacuation System is a non-safety-related system that removes air and other noncondensable gases from the main condenser. The Main Condenser Evacuation System is located in the turbine building, with the exhaust line going to the unit vent in the nuclear annex.
The Basic Configuration of the Main Condenser Evacuation System is as shown on Figure 2.8.4-1.
The Main Condenser Evacuation System has vacuum pumps and associated piping and instramentation.
The vacuum pump air discharge is routed to the unit vent and monitored for radiation.
Displays of the Main Condenser Evacuation System instrumentation shown on Figure 2.8.4-1 exist in the main control room (MCR) or can be retrieved there.
Inspections, Tests, Analyses, and Acceptance Criteria Table 2.8.4-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Main Condenser Evacuation System.
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fy3 tem 80+ Design ControlDocument Table 2.8.4-1 Main Condenser Evacuation System Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 1. The Basic Configuration 1. Inspection of the as-built 1. For the components and of the Main Condenser Main Condenser equipment shown on Evacuation System is as Evacuation System Figure 2.8.4-1, the as-shown on Figure 2.8.4-1. configuration will be built Main Condenser conducted. Evacuation System conforms with the Basic D Configuration.
- 2. Displays of the Main 2. Inspection for the 2. Displays of the Condenser Evacuation existence or instrumentation shown on System instrumentation retrievesbility in the Figure 2.8.4-1 exist in shown on Figure 2.8.41 MCR of instrumentation the MCR or can be exist in the MCR or can displays will be retrieved there, be retrieved there. performed.
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Certined Design Material Page 2.8-14
Sy~ tem 80+ Design Control Document
( 2.8.5 Turbine Bypass Systein Design Descdption The Turbine Bypass System is a non-safety-related system. For startup, shutdown, and during load shedding, the Turbine Bypass System provides the capability to take steam from the main steam ;
equalization header and discharge it directly to the main condenser, bypassing the turbine generator.
The Basic Configuration of the Turbine Bypass System is as shown on Figure 2.8.5-1.
The Turbine Bypass System consists of at least eight turbine bypass valves (TBVs), and associated piping and controls. The Turbine Bypass System is located in the turbine building.
Controls exist in the main control room (MCR) to open and close those power operated valves shown on Figure 2.8.5-1.
Valves with response positions indicated on Figure 2.8.5-1 change position to that indicated on the Figure upon loss of motive power.
The turbine bypass valves are controlled by the steam bypass control system (SBCS).
Inspections, Tests, Analyses, and Acceptance Criteria
,q Table 2.8.5-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Turbine Bypass System.
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System 80+ Design ControlDocument O
b Table 2.8.5-1 Turbine Bypass System Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 1. The Basic Configuration 1. Inspection of the as-built 1. For the components and of the Turbine Bypass Turbine Bypass System equipment shown on System is as shown on configuration will be Figure 2.8.5-1, the as-Figure 2.8.5-1. conducted. built Turbine Bypass System conforms with the Basic Configuration.
- 2. Controls exist in the 2 Testing will be 2. Turbine Bypass System ,
MCR to open and close performed using the controls in the MCR those power operated Turbine Bypass System open and close those valves shown on Figure controls in the MCR. power operated valves 2.8,5-1. shown on Figure 2.8.5-1.
- 3. The turbine bypass 3. Testing will be conducted 3. The turbine bypass valves open on receipt of using a signal simulating valves open on receipt of a turbine bypass signal, a turbine bypass signal. a signal simulating a turbine bypass signal.
- 4. Valves with response 4. Testing of loss of motive 4. These valves change ,
positions indicated on power to these valves position to the position Figure 2.8.5-1 change will be performed. indicated on Figure g
A position to that indicated 2.8.5-1 on loss of motive on the Figure upon loss power, of motive power.
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System 80+ Design ControlDocument 2.8.6 Condensate and Feedwater Systems Design Description The Condensate and Feedwater Systems transfer condensate from the main condenser hotwells to the steam generators.
The Basic Configuration of the Condensate and Feedwater Systems is as shown in Figure 2.8.6-1A and Figure 2.8.6-1B. The ASME Code Section III Class 2 components shown on Figure 2.8.6-1B are safety-related.
The Condensate System has three motor-driven condensate pumps, a deaerator storage tank, a gland seal steam condenser, low pressure heaters, associated piping, valves, instrumentation and controls located in the turbine building. The Condensate System also has a condensate cleanup system which is located in the condensate cleanup system area. The Feedwater System is located in the turbine building, above the yard, in the nuclear annex, main steam valve houses, and Containment The Feedwater System has three sets of feedwater booster and main feedwater pumps, high pressure feedwater heaters, and associated piping, valves, instrumentation and controls. Feedwater control valves are provided to regulate the feedwater flow to each steam generator and to maintain water level in the steam generator.
The ASME Code Section III Class for Feedwater System pressure retaining components shown en Figure 2.8.6-1B is as depicted on the Figure.
The safety related equipment shown on Figure 2.8.6-1B is classified Seismic Category I.
Controls exist in the main control room (MCR) to start and stop the feedwater pumps and condensate pmnps, and to open and close those power operated valves shown on Figure 2.8.6-1B.
Within a mechanical Division, the electrical controls of one main feedwater isolation valve (MFIV) of each set of two MFIVs on the main feedwater lines which penetrate the containment boundary are powered from their respective Class IE Division. The electrical controls of the other MFIV of each set of two MFIVs within a mechanical Division on the main feedwater lines which penetrate the containment boundary are powered from a Class IE Division different from the Class IE Division of the respective mechanical Division.
Independence is provided between Class IE Divisions, and between Class IE Divisions and non-Class IE equipment, in the Condensate and Feedwater Systems.
The two mechanical Divisions of the safety-related portions of the Feedwater System are physically separated.
Valves with response positions indicated on Figure 2.8.6-1B change position to that indicated on the Figure upon loss of motive power.
The MFIVs close on receipt of a rain steam isola *n signal (MSIS) or when remotely actuated from the control room.
Certr6ed Desigrr Matenal Page 2.818
- System 80 + Design ControlDocumust Check valves shown on Figure 2.8.6-1B will open, or will close, or will open and also close, under system pressure, fluid flow conditions, or temperature conditions.
Inspections, Tests, Analyses, and Acceptance Criteria ;
Table 2.8.6-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Condensate and Feedwater Systems.
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. _ . System 80+ Design ControlDocument l Table 2.8.6-1 Condensate and Feedwater Systems Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 1. The Basic Configuration 1. Inspection of the as-built I. For the components and of the Condensate and Condensate and equipment shown on Feedwater System is as Feedwater Systems Figure 2.8.6-1 A and shown on Figure configuration will be Figure 2.8.6-1B, the as-2.8.61 A and Figure performed. built Condensate and 2.8.6-1 B. Feedwater System conform with the Basic Configuration.
- 2. The ASME Code Section 2. A pressure test will be 2. The results of the
!!! Feedwater System conducted on those pressure test of ASME components shown on components of the Code Section III Figure 2.8.6-1B retain Feedwater System components of the their pressure boundary required to be pressure Feedwater System integrity under internal tested by ASME Code conform with the pressures that will be Section Ill. pressure testing experienced during acceptance criteria in the ,
service. ASME Code Section Ill. l 3.a) Within a mechanical 3.a) Testing will be 3.a) Within the Feedwater ;
Division, the electrical performed on the System, a test signal '
l controls of one MFIV of Feedwater System by exists only at the each set of two MFIVs providing a test signalin equipment powered from on the main feedwater only one Class IE the Class IE Division lines which penetrate the Division at a time. under test.
containment boundary are powered from their respective Class IE Division.
3.b) The electrical controls of 3.b) Testing will be 3.b) Within the Feedwater the other MFIV of each performed on the System, a test signal set of two MFIVs within Feedwater System by exists only at the ,
a mechanical Division on providing a test signal in equipment powered from j the main feedwater lines only one Class IE the Class IE Division i which pencuate the Division at a time. under test. !
i containment boundary are powered fmm a Class IE Division different from the Class IE Division of the respective mechanical Division.
O Certthed Destyn Material Page 2.8 22
System 80+ Design controlDocument p ,
Table 2.8.6-1 Condensate and Feedwater Systems (Continued) l Design Commitment inspections, Tests, Analyses Acceptance Criteria 3.c) Independence is provided 3.c) Inspection of the as- 3.c) Physical separation exists between Class IE installed Class IE between Class IE Divisions, and between Divisions of the Divisions in the Class IE Divisions and Condensate and Condensate and non-Class IE equipment, Feedwater Systems will Feedwater Systems.
in the Condensate and be performed. Physical separation exists :
Feedwater Systems. between Class IE Divisions and non-Class IE equipment in the Condensate and i Feedwater Systems.
- 4. The two mechanical 4. Inspection of the as-built 4. The mechanical '
Divisions of the safety- mechanical Divisions will Divisions of the safety-related portions of the be performed, related portions of the Feedwater System are . Feedwater System are physically separated. located in two main steam valve houses separated by the i Containment.
' Components of the system within the ,
Containment are separated by spatial arrangement or barriers.
- 5. Controls exist in the 5. Testing will be 5. Condensate and MCR to start and stop performed using the Feedwater System the feedwater and Condensate and controls in the MCR condensate pumps, and to Feedwater Systems operate to start and stop open and close those controls in the MCR. the feedwater and power operated valves condensate pumps, and shown on Figure to open and close those 2.8.6-1 B. power operated valves shown on Figure 2.8.6-1 B.
- 6. The Main Feedwater 6. Testing will be 6. The MFIVs close within isolation Valves close on performed using a signal 5 seconds after receipt of receipt of a MSIS. simulating a MSIS. a signal simulating a MSIS.
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CM Desipes a0etene! Page 2.8 23
System 80+ Design ControlDocument Table 2.8.6-1 Condensate and Feedwater Systems (Continued)
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 7. Check valves shown on 7. Testing will be conducted 7. Each check valve shown Figure 2.8.6-1B will to open, or close, or on Figure 2.8.6-1B open, or will close, or open and also close the opens, or closes, or will open and also close check valves shown on opens and also closes.
under system pressure, Figure 2.8.6-1B under fluid flow conditions, or system preoperational temperature conditions. pressure, fluid flow conditions, or temperature conditions.
- 8. Valves with response 8. Testing of loss of motive 8. These valves change positions indicated or power to these valves position to the position Figure 2.8.6-1B change will be performed. indicated on Figure position to that indicated 2.8.6-1B on loss of on the Figure upon loss motive power.
of motive power.
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O Cerened Design Materiat page 2.g.24
d System 80+ Design Control Document 2.8.7 Steam Generator Blowdown System Design Description The Steam Generator Blowdown System (SGBS) removes and processes steam generator fluid containing '
impurities, and returns the water to the feedwater and condensate system.
The Basic Configuration of the SGBS is as shown on Figure 2.8.7-1.
The ASME Code Section III Class 2 pressure retaining components of the SGBS shown on the Figure are safety-related.
The SGBS has a mechanical piping train from each steam generator. Each train includes two blowdown lines from the secondary side of each steam generator. The trains discharge blowdown fluid to a flash tank from which the blowdown fluid is processed and returned to the condensate and feedwater system.
The ASME Code Section III Class for the SGBS pressure retaining components shown in Figure 2.8.7-1 is as depicted on the Figure.
The safety-related equipment shown on Figure 2.8.7-1 is classified Seismic Category I.
Controls exist in the main control room (MCR) to open and close those power operated valves shown on Figure 2.8.7-1.
O b Each SGBS blowdown line penetrating containment contains valves which close upon receipt of a main steam isolation signal (MSIS),' an emergency feedwater actuation signal (EFAS), or an alternate feedwater [
actuation signal (AFAS).
Inspections, Tests, Analyses, and Acceptance Criteria Table 2.8.7-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Steam Generator Blowdown System.
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Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The Basic Configuration 1. Inspection of the as-built 1. For the components and of the SGBS is as shown SGBS configuration will equipment shown on on Figure 2.8.7-1. be conducted. Figure 2.8.7-1, the as-built SGBS conforms with the Basic Configurcion.
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- 2. The ASME Code Section 2. A pressure test will be 2. The results of the l Ill SGBS components conducted on those pressure test of ASME
! shown on Figure 2.8.7-1 components of the SGBS Code Section 111 retain their pretsure required to be pressure components of the SGBS boundary integrity under tested by the ASME conform with the internal pressures that Code Section III. pressure testing will be experienced acceptance criteria in during service. ASME Code Section Ill.
- 3. Controls exist in the 3. Testing will be 3. SGBS controls in the MCR to open and close performed using the MCR operate to open those power operated SGBS controls in the and close those power valves shown on Figure MCR. operated valves shown )
2.8.7 1. on Figure 2.8.71. l 1
- 4. Each SGBS blowdown 4. Testing will be 4. SGBS contamment l line penetrating performed using signals isolation valves close l containment contains simulating an MSIS, upon receipt of a signal ;
valves which close upon EFAS, and AFAS in simulating an MSIS, receipt of a MSIS, an individualtests. The EFAS, or AFAS.
EFAS. or a AFAS. SGBS containment isolation valves response to each signal will be observed.
\
Coraned Design A0eteniel Page 2.8 27
SyTiem 80+ Design ControlDocument l
2.8.8 Emergency Feedwater System Design Description The Emergency Feedwater System (EFWS) supplies feedwater to the steam generators for events resulting !
in loss of normal feedwater and requiring heat removal through the steam generators.
The EFWS is located within the nuclear island (NI) structures, including Containment, reactor building (RB), and nuclear annex (NA).
The Basic Configuration of the EFWS is as shown on Figure 2.8.8-1. The EFWS is safety-related as noted on Figure 2.8.8-1.
The EFWS consists of two mechanical Divisions, each with an emergency feedwater storage tank (EFWST) which is an integral part of the NI structures, two EFW pumps, a cavitating flow-limiting venturi, valves, piping, instrumentation and controls. The EFW pumps in each Division are powered
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by diverse drivers.
I A gravity-fed non-safety grade source of condensate makeup is provided to either EFWST. ]
l The flow recirculation line from each EFW pump discharge back to its associated EFWST provides j required EFW pump minimum flow and permits testing each EFW pump at full flow.
I Each EFW pump delivers at least the minimum flow required for removal of core decay heat using the steam generator (s), (SG) against steam generator feedwater nozzle pressures up to main steam safety valve lift pressure.
The cavitating flow-limiting venturis limit emergency feedwater flow to each SG with both EFW pumps running in the Division against steam generator pressures down to O psig.
Each EFWST has a volume above the EFW pump suction line penetrations to permit plant cooldown to I shutdown cooling entry conditions following the most limiting design basis event.
The ASME Code Section Ill Class for the EFWS pressure retaining components shown on Figure 2.8.8-1 l is as depicted on the Figure. l The safety-related equipment shown on Figure 2.8.8-1 is classified Seismic Category 1. 1 I
4 Displays of the EFWS instrumentation shown on Figure 2.8.8-1 exist in the main control room (MCR) {
or can be retrieved there.
i Controls exist in the MCR to start and stop the EFW pumps, and to open and close those power operated valves shown on Figure 2.8.8-1.
Alarms shown on Figure 2.8.8-1 are provided in the MCR.
Water is supplice to each EFW pump at a pressure greater than the pump's required net positive suction head (NPSii).
Corkfied Design Material Page 2.8-28
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.( Within a mechanical Division, the following components are powered from their respective Class IE Division- y l e- the motor-driven EFW pump, j i
- the two motor-operated valves in the motor-driven EFW pump's discharge line, and l t .
!- ' o' . process instrumentation, except alarms, in the motor-driven EFW pump's suction and discharge l lines.
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j Within a mechanical Division, the followit. components are powered from a Class IE Division different from the Class IE Division powering the nw.or-driven EFW pump:
e the turbine-driven EFW pump electrical cotsreis, e the two motor-operated valves in the turbine-driven EFW pump's discharge line, {w
- the motor-operated valve and electrical controls for the pneumatic valve and hydraulic valve in l the turbine-driven EFW pump's steam supply line, and ,
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- . . process instrumentation, except alanns, in the turbine 4 riven EFW pump's suction and discharge ,
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In each EFW pump's discharge line, each of the two motor-operated valves is powered from a different l Class IE bus in the same Class IE Division.
4 Independence is provided between Class IE Divisions, and between Class IE Divisions and non-Class ;
IE equipment, in the EFWS. l 3
The two mechanical Divisions of the EFWS are physically separated except for the cross-connect lines l between EFWSTs and between Divisional EFW pump discharge lines. !
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t The EFWS is actuated by an emergency feedwater actuation signal (EFAS) from the engineered safety j features actuation system (ESFAS) or by an alternate feedwater actuation signal (AFAS) from the ;
alternate protection system (APS). The engineered safety features component control system (ESF-CCS) j l'
1 includes logic to close the isolation valves and flow control valves when SG water level has risen above i a high level setpoint, and to re-open those valves when SG water level drops below a low level setpoint. f i
Motor operated valves (MOVs) having an active safety function will open, or will close, or will open and !
also close, under differential pressure or fluid flow conditions and under temperature conditions. l Check valvt.s shown on Figure 2.8.8-1 will open, or will close, or will open and also close, under system l
. pressure, fluid flow conditions,'or temperature conditions. [
i Valves with response positions indicated on Figure 2.8.8-1 change position to that indicated on the Figure j g ' upon loss of motive power.
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Syntem 80+ Design control Document Inspections, Tests, Analyses, and Acceptance Criteria Table 2.8.8-1 specifies the inspection, tests, analyses, and associated acceptance criteria for the Emergency Feedwater System.
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Sy3 tem 80+ Design ControlDocument Table 2.8.8-1 Emergency Feedwater System h Design Commitment Inspections, Tests, Analys4= Acceptance Criteria
- 1. The Basic Configuration 1. Inspection ot the as-built 1. For the components and of the EFWS is as shown EFWS configuration will equipment shown on on F.gure 2.8.8-1. be conducted. Figure 2.8.8-1, the as-built EFWS conforms with the Basic Configuration.
- 2. A gravity-fed non-safety 2. Testing of the non-safety 2. The water level increases grade source of grade source of in t'ie EFWST being fed condensate makeup is condensate makeup to the f.om the makeup source.
provided to either EFWSTs will be EFWST. performed by manually aligning the makeup source to each EFWST with the EFWST at a low water level. Water level will be observed.
- 3. The flow recirculation line 3. Testing of each EFW 3. Minimum recirculation from each EFW pump pump in the minimum flow meets or exceeds the discharge back to its fiow and the full flow test pump vendor's required associated EFWST modes will be conducted minimum flow. Full flow provides required EFW v ith flow directed to the from each pump (at least pump minimum flow and EFWST through the 500 gpm) is returc:d to permits testing each EFW pump's recirculation lines. the EFWST.
pump at full flow.
4.a) Each EFW pump delivers 4.a) Testing of each EFW 4.a) Each EFW pump delivers at least the minimum flow pump will be performed at least 500 gpm to the required for removal of to determine system flow steam generator (s) against core decay heat using the vs. steam generator a steam generator steam generator (s) against pressure. Analysis will be feedwater nozzle pressure a steam generator performed to convert the of 1217 psia.
feedwater nozzle pressure test results to the design up to main steam safety conditions.
valve lift pressure.
4.b) The cavitating flow- 4.b) Testing will be performed 4.b) The maximum flow to limiting venturis limit with both pumps in a cach SG is 800 gpm with maximum flow to each Division running. both pumps running SG with both pumps in Analysis will be used to against a steam generator the Division running convert the test results to pressure of 0 psig.
against a steam generator the conditions of the pressure of 0 psig. Design Commitment.
O Certined Design Material Page 2.8-32
System 80+ Design ControlDocument Table 2.8.8-1 Emergency Feedwater System (Continued)
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 5. Each emergency 5. Inspection of the EFWSTs 5. The internal volume feedwater storage tank has will be performed and the above the EFW pump a volume above the EFW internal volume of each suction line penetrations pump suction line tank available for of each EFWST is at least penetrations to permit emergency feedwater will 350,000 gallons.
plant cooldown to be determined.
shutdown cooling entry conditions following the most limiting design basis event.
- 6. The ASME Code Section 6. A pressure test will be 6. The results of the $
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Ill EFW components conducted on those pressure test of ASME Code Section 111 i shown on Figure 2.8.8-1 components of the EFW retain their pressure required to be pressure components of the EFW boundary integrity under tested by ASME Code conform with the pressure internal pressures that will Section III. testing acceptance criteria -
be experienced during in ASME Code Section service. 111.
7.a) Displays of the EFW 7.a) Inspection fer the 7.a) Displays of the instrumentation shown on existence or retrieveability instrumentation shown on O Figure 2.8.8-1 exist in the MCR or can be retrieved there, in the MCR of instrumentation displays will be performed.
Figure 2.8.8-1 exist in the MCR or can be retrieved there.
7.b) Controls exist in the MCR 7.b) Testing will be performed 7.b) EFW controls in the to start and stop the EFW using the EFW controls in MCR operate to start and pumps, and to open and the MCR. stop the EFW pumps, and close those power to open and close those operated valves showe on power operated valves i Figure 2.8.8-1. shown on Figure 2.8.8-1.
7.c) EFWS alarms shown on 7.c) Testing of the EFWS 7.c) The EFWS alarms shown Figure 2.8.8-1 are alarms shown on Figure on Figure 2.8.8-1 actuate provided in the MCR. 2.8.8-1 will be performed in the MCR.
using signals simulating alarm conditions.
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- 8. Water is supplied to each 8. Testing to measure EFW 8. The available NPSH EFW pump at a pressure pump suction pressure exceeds each EFW >
greater than the pump's r<ill be performed, pump's required NPSH.
required net positive Inspections and analyses suction head (NPSH). to determine NPSH available to each pump will be performed based on test data and cs-built data.
cerawed oneon neeeniet rage 2.s-33
SyTrem CO+ Design ControlDocument Table 2.8.8-1 Etnergency Feedwater System (Continued)
Design Commitment Inspections, Tests, Analyses Acceptance Criteria 9.a) Within a mechanical 9.a) Testing will be performed 9.a) Within the EFW, a test Division, the following on the EFW by providing signal exists only at the components are powered a test signal in only one equipment powered from from their respective Class IE Division at a the Class IE Division Class IE Division: time. under test.
the motor-driven EFW pump, the two motor-operated valves in the motor-driven EFW pump's discharge line, and process instrumentation, except alarms, in the motor-driven pump's suction and discharge lines.
9.b) Within a mechanical 9.b) Testing will be performed 9.b) Within the EFW, a test Division, the following on the EFW by providing signal exists only at the components are powered a test signal in only one equipment powered from from a Class IE Division Class IE Division at a the Class IE Division different from the Class time. under test.
IE Division powering the motor-driven EFW pump:
the turbine-driven EFW pump electrical controls, j the two motor-operated )
valves in the turbine- I driven EFW pump's discharge line, the motor-operated valve and electrical controls for the pneumatic valve and hydraulic valve in the turbine-driven EFW pump's steam supply line, cnd process instrumentation, except alarms, in the turbine-driven EFW pump's suction and discharge lines.
9.c) In each EFW pump's 9.c) Testing will be performed 9.c) A test signal exists only at discharge line, each of the on the EFW motor- the EFW motor-operated two motor-operated valves operated valves in each valve powered from the is powered from a EFW pump discharge line Class IE bus under test.
different Class IE bus in by providing a test signal the same Class IE in only one Class IE bus Division. at a time.
Cert l6ed Desiger Material Page 2.8 34
System 80+ Design controlDocument (O) Table 2.8.8-1 Emergency Feedwater System (Continued)
< Design Commitment Inspections, Tests, Analyses Acceptance Criteria ,
9.d) Independence is provided 9.d) Inspection of the as- 9.d) Physical separation exists between Class IE installed Class IE between Class IE Divisions, and between Divisions of the EFWS Divisions in the EFWS.
Class IE Divisions and will be performed. Physical separation exists non-Class IE equipment, between Class 1E in the EFWS. Divisions and non-Class IE equipment in the EFWS.
- 10. The two mechanical 10. Inspection of as-built 10. The two mechanical Divisions of the EFW are mechanical Divisions will Divisions of the EFW are physically separated, be performed. separated by a Divisional
, except for the wall or a fire barrier cross-connect lines except for the cross-between EFWSTs and ' connect lines between between Divisional EFW Divisional EFW pump pump discharge lines, discharge lines. Within containment, the EFWS Divisions are separated by spatial arrangement or barriers.
O ll.a) The EFV/S is actuated by ll.a) Testing will be performed 11.a) The motor-driven and V an emergency feedwater actuation signal (EFAS) or by genciating a signal simulating an EFAS for turbine-driven pumps start, and the steam an alternate feedwater its corresponding steam generator isolation and actuation signal (AFAS). generator. The test will flow control valves open, ,
be repea ed using a signal in the Division receiving ,
simulati ag an AFAS. the signal simulating an EFAS. The same components actuate in response to a signal simulating an AFAS.
Flow is delivered to the steam generator (s) in no more than 60 seconds following an EFAS or AFAS.
II.b) The ESF-CCS includes 11.t0 Testing of each EFWS 11.b) A signal simulating high SG logic to close the isolation Division will be water level signal closes valves and flow control performed using signals the SG isolation valves valves when SG water simulating high and low and the flow control level has risen above a SG water level. valves in its associated high level setpoint, and to Division. A signal rc+ pen those valves when simulating low SG water SG water level drops level signal opens the SG below a low level isolation valves and the N setpoint. flow control valves in its associated Division.
Cten9ed Deeen Mahwiet foge 2.8 35
Sy0 tem 80+ Design ControlDocument Table 2,8.8-1 Emergency Feedwater System (Continued) h Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 12. Motor-operated valves 12. Testing will be performed 12. Each MOV having an (MOVs) having an active to open, or close, or open active safety function safety function will open, and also close MOVs opens, or closes, or opens or will close, or will open having an active safety and also closes.
and also close under function under differential pressure or preoperational differential fluid flow conditions and pressure or fluid flow under temperature conditions and under conditions. temperature conditions.
- 13. Check valves shown on 13. Testing will be performed 13. Each check valve shown Figure 2.8.8-1 will open, to open, or close, or open on Figure 2.8.8-1 opens, or will close, or will open and also close check or closes, or opens and and also close under valves shown on Figure also closes.
system pressure, fluid 2.8.8-1 under system flow conditions, or preoperational pressure, temperature conditions. fluid flow conditions, or temperature conditions.
- 14. Valves with response 14. Testing of loss of motive 14. These valves change positions insticated on power to these valves will position to the position Figure 2.8,S-1 change be performed. indicated on Figure position to that indicated 2.S.8-1 on loss of motive on the Figure upon loss of power.
motive power.
O Certined Desism Matemt page 2.g4g
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System 80 + ' Deskn ControlDocument 2.8.9 Condenser Circulating Water System Design Description The Condenser Circulating Water System is a non-safety interface system that provides cooling water for the turbine condensers and transfers heat from the Turbine Building Service Water System to the normal heat sink.
The parts of the Condenser Circulating Water System that are in the Turbine Building are within the Certified Design. Those parts of the system that are outside the Turbine Building are not in the Certified Design. Figure 2.8 9-1 shows the system basic configuration and scope of the Condenser Circulating Water System within the Certified Design.
Inspections, Tests, Analyses, and Acceptance Criteria Table 2.8.9-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Condenser Circulating Water System.
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- 1. MULTIPLE LINES MAY BE USED Condenser Circulating Water System Figure 2.8.9-1 ,
W #### Page 2.8 38
Sy= tem 80 + Design controlDocument t
(Q) Table 2.8.9-1 Condenser Circulating Water System Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 1. A Basic Configuration 1. Inspection of the as-built 1. The as-built Condenser for the Condenser system will be Circulating Water Circulating Water System conducted. System conforms with I is as shown on Figure the Basic Configuration 2.8.9-1. shown on Figure 2.8.9-1.
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-Q 2.9 Radioactive Waste Managesnent 2.9.1 Liquid Waste IW- ; -- ^ System Design Mption ;
a The Liquid Waste Management System (LWMS) is used to collect, segregate, store, process, sample, and monitor radioactive liquid waste. The LWMS is non-safety-related with the exception of the containment '
isolation valves and piping in between covered in Section 2.4.5.
The LWMS is located in the radwaste building. .
The Basic Configuration of the LWMS is as shown on Figure 2.9.1-1. ,
- The LWMS has four subsystems which process radioactive or potentially radioactive liquid waste. These i four subsystems segregate liquid waste into high level waste, low level waste, laundry and hot shower / chemical waste, atxi the containment cooler condensate waste. .
i- The high level waste subsystem has 'ilters, demineralizers, provisions for batch sampling, and piping for recirculation of liquid waste for furtaer processing.
The low level waste subsystem has filters, demineralizers, provisions for batch sampling, and piping for recirculation of liquid waste for further processing. l, t
The laundry and hot shower / chemical waste subsystem has filters, demineralizers, provisions for_ batch -l' sampling, and piping for recirculation of liquid waste for further processing.
The containment cooler condensate subsystem has tanks to collect containment cooler condensate. The discharge from the tanks is monitored for radioactivity. Although not normally radioactive, this discharge ;
can be diverted to the low level waste subsystem. The containment cooler condensate tank levels and discharge flow are also monitored by level and flow instrumentation.
The LWMS subsystems have collection and storage capacity to process waste volumes expected during l 9
normal operation and from anticipated operational occurrences.
Displays of the LWMS instrumentation shown on Figure 2.9.1-1 exist in the main control room (MCR) :
or can be retrieved there. [
Controls exist in the MCR to open and close the power operated valves shown on Figure 2.9.1-1.
l The valves with the response position indicated on Figure 2.9.1-1 change position to that indicated on the Figure upon loss of motive power. l l
The LWMS has means to monitor radioactivity levels in the processed liquid waste prior to release. The c radioactivity monitor provides a signal to terminate LWMS discharge when a specified radioactivity level l is reached.' ;
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Sy~ tem 80+ Design control Document Inspections, Tests, Analyses, and Acceptance Criteria Table 2.9.1-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Liquid Waste Management System.
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Sy~ tem 80+ Design ControlDocument Table 2.9.1-1 Liquid Waste Management System Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 1. The Basic Configuration of 1. Inspection of the as-built 1. For the components and the Liquid Waste LWMS configuration will equipment shown on Management System be conducted. Figure 2.9.1-1, the as-(LWMS) is as shown on built LWMS conforms Figure 2.9.1-1. with the Basic Configuration.
- 2. The ASME Code Section 2. A pressure test will be 2. The results of the pressure III LWMS components conducted on those test of the ASME Code 6hown on Figure 2.9.1-1 components of the LWMS Section III components of retain their pressure required to be pressure the LWMS conform with boundary integrity under tested by ASME Code the pressure testing internal pressures that will Section !!!. acceptance criteria in be experienced during ASME Code Section III.
service.
- 3. The LWMS subsystems 3. Analysis of the as-built 3. An analysis exists which have collection and storage LWMS subsystems' concludes the LWMS capacity to process waste processing capability will subsystems have collection volumes expected during be perforced, and storage capacity to normal operation and from process waste volumes anticipated operational expected during normal occurrences. operation and from anticipated operational occurrences.
- 4. Displays of the LWMS 4. Inspection for the 4. Displays of the instrumentation shown on existence or retrievability instrumentation shown on Figure 2.9.1-1 exist in the in the MCR of Figure 2.9.1 1 exist in the MCR or can be retrieved instrumentation displays MCR or can be retrieved there. will be performed. there.
- 5. Controls exist in the MCR 5. Testing will be performed 5. LWMS controls in the to open and close the using the LWMS controls MCR operate to open and power operated valves in the MCR. close the power operated shown on Figure 2.9.1-1. valves shown on Figure 2.9.1-1. l 1
- 6. The valves with the 6. Testing ofloss of motive 6. The valves change !
response position indicated power to the valves will be position to the position l on Figure 2.9.1-1 changes performed. indicated on Figure l position to that indicated 2.9.1-1 upon loss of l on the Figure upon loss of motive power. ;
motive power. I
- 7. The radioactivity monitor 7. Testing of the as-built 7. LWMS discharge is provides a signal to LWMS discharge comrols terminated in response to terminate LWMS will be performed using a a signal simulating that the discharge when a specified signal which simulates radioactivity level in the radioactivity level is radioactivity levels, waste discharge line has reached. reached a specified limit.
l Certt16ed Design hinterial Page 2.9-4
__ . . ~ . . . _ _ _ _ . _ _ _ _ _ __. __. __ __ .. . ._
l System 80+ oestan controlDocument ;
1 1
2.9.2 Geseous Waste Management System Design Description l
The Gaseous Waste Management System (GWMS) is used to collect, store, process, sample, and monitor ]
radioactive gaseous waste.
]
The GWMS is located in the nuclear annex. With the exception of containment penetration isolation !
valves and the piping in between covered in Section 2.4.5, the GWMS is non-safety-related.
l The Basic Configuration of the AWMS is as shown on Figure 2.9.2-1. i The GWMS processing unit has a cooler condenser, a charcoal guard bed, and charcoal adsorbers. The f GWMS charcoal vessels, cooler, condenser, piping, components, and valves are capable of withstanding !
- a hydrogen explosion, including the pressure boundary of the analyzers. In the GWMS, one analyzer l
.is provided to monitor the concentration of hydrogen, and one analyzer is provided to monitor the l concentration of oxygen. The GWMS and its supports will not collapse in a safe shutdown earthquake.- !
Displays of the GWMS instrumentation shown on Figure 2.9.2-1 exist in the main control room (MCR)
- or can be retrieved there.
l Controls exist in the MCR to open and close the power operated valve shown on Figure 2.9.2-1. l l
The valve with the response position indicated on Figure 2.9.2-1 changes position to that indleard on l the Figure upon loss of motive power.
The GWMS provides a means to monitor radioactivity levels in the processed gaseous waste prior to !
release through the unit vent. The radioactivity monitor provides a signal to terminate GWMS discharge i when a specified radioactivity level is reached. !
i Inspections, Tests, Analyses, and Acceptance Criteria l Table 2.9.2-1 specifies the inspections, tests, and analyses, and associated acceptance criteria for the ;
Gaseous Waste Management System. i i
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C Table 2.9.2-1 Gaseous Waste Management System Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 1. The Basic Configuration 1. Inspection of the as-built 1. For the components and of the GWMS is as 'GWMS configuration equipment shown on shown on Figure 2.9.2-1. will be conducted. Figure 2.9.2-1, the as-built GWMS conforms with the Basic Configuration.
- 2. The GWMS and its 2. A seismic analysis of the 2. An analysis of the supports will not collapse GWMS and its supports GWMS exists that in a safe shutdown will be conducted, concludes the GWMS earthquake (SSE). and its supports do not collapse under an SSE.
- 3. The GWMS will 3. A hydrogen explosion 3. An analysis of the withstand a hydrogen pressure rise analysis of GWMS exists that explosion (i.e., twenty the GWMS will be concludes the GWMS times normal operating conducted, withstands a hydrogen pressure). explosion (i.e., twenty times normal operating pressure).
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- 4. Displays of the GWMS 4. Inspection for the 4. Displays of the (V) instrumentation shown on existence or retrievability instrumentation shown on Figure 2.9.2-1 exist in in the MCR of Figure 2.9.2-1 exist in the MCR or can be instrumentation displays the MCR or can be retrieved there, will be performed. retrieved there.
- 5. Controls exist in the 5. Testing will be 5. GWMS controls in the j MCR to open and close performed using the MCR operate to open '
the power operated valve GWMS controls in the and close the power ;
shown on Figure 2.9.2-1. MCR. operated valve shown on Figure 2.9.2-1.
- 6. The valve with the 6. Tes%.A of loss of motive 6. This valve changes response position power to this valve will position to that indicated indicated on Figure be performed. on Figure 2.9.2-1 upon 2.9.2-1 changes position loss of motive pcrwer.
to that ind.cated on the Figure upoa loss of motive power.
- 7. The radioactivity monitor 7. Testing of the as-built 7. GWMS discharge is l provides a signal to GWMS discherge terminated when the terminate GWMS controls will be simulated radioactivity discharge when a performed using a signai level in the discharge specified radioactivity which simulates waste line reaches a level is reached. radioactivity levels. specified limit.
i 4 V
l Certined Design Material Page 2.9-7
1 System 80+ Design ControlDocurnent 2.9.3 Solid Waste Management System Design Description The Solid Waste Managemer.t System (SWMS) is a non-safety-related system which is used to collect, segregate, decontaminate, process, sample, and store radioactive solid waste.
The SWMS is located in the radwaste building.
He Basic Configuration of the SWMS is as shown on Figure 2.9.3-1.
Solid waste is segregated into the following:
- High activity and low activity wetted waste, e.g. spent ion exchanger resin and spent f!hcr assemblies; and,
- Compactible and non-compactible dry solid waste, e.g. plastic sheeting, clothing, or metal tools.
The high activity and low activity spent resin processing units have collection and storage capacity to process waste volumes expected during normal operation and from anticipated operational occurrences.
These subsystems can process waste by dewatering in the shipping container.
The dry solid material subsystems have provisions for sorting of wastes, compaction of compactible waste and placement in shipping containers, and for either decontamination or direct placement of non-compactible waste into shipping containers.
Inspections, Tests, Analyses, and Acceptance Criteria Table 2.9.3-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Solid Waste Management System.
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Sy tem 80 + Design ControlDocument Table 2.9.3-1 Solid Waste Management System Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 1. The Basic Configuration 1. Inspection of the as-built 1. For the components and of the SWMS is as SWMS configuration will equipment shown on shown on Figure 2.9.3-1. be conducted. Figure 2.9.3-1, the as-built SWMS conforms with the Basic Configuration.
- 2. The high activity and low 2. Analysis of the as-built 2. An analysis exists which activity spent resin spent resin subsystems' concludes that the spent processing units have processing capability will resin processing units collection and storage be performed. have collection and capacity to process waste storage capacity to volumes expected during process waste volumes normal operation and expected during normal from anticipated operation and from operational occurrences. anticipated operational occurrences.
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System 80+ Design controlDocument 2.9.4 Process and Effluent Radiological Monitoring and Sampling System Design Description The Process and Effluent Radiological Monitoring and Sampling System (PERMSS) provides components to monitor liquid and gaseous effluents prior to release to unrestricted areas, and to monitor for inplant radioactivity.
Components of the PERMSS are located in the nuclear island stmetures, the radwaste building, the turbine building, and the station service water pump structure.
The PERMSS has components that provide radiological monitoring of gaseous and liquid prxessing systems and their effluents, airborne radioactivity, radiation areas, and specified plant equipment.3 The system provides radiological monitoring during plant operation and post-accident conditions. The two high range containment area radiation monitors provide indication of the radiation levels in Containment throughout the course of a design basis accident.
The PERMSS is non-safety-related with the exception of the following, each of which is safety-related, Seismic Category I, and Class IE:
i
- a. main control room (MCR) intake radiation monitor (2/ intake),
- b. high range containment area radiation monitor (2),
- c. containment atmosphere radiation monitor (particulate channel only),
- d. primary coolant loop radiation monitors (2).
Independence is provided between Class IE Divisions, and between Class IE Divisions and non-Class i IE equipment, in the PERMSS.
The MCR intake radiation monitors shall have the capability for auto selection and closure of the most -
contaminated intake.
Displays of the PERMSS safety-related instrumentation (the MCR air intake radiation monitors, the ,
reactor coolant radiation monitors, the high range containment arca monitors and the containment !
atmosphere particulate monitors) exist in the MCR or can be retrieved there.
Inspections, Tests, Analyses, and Acceptance Criteria Table 2.9.4-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Process and Effluent Radiological Monitoring and Sampling System.
/~N h 8 The radiation monitors that monitor gaseous and liquid processing systems and their effluents and the response of these systems to detection of radiation are addressed in the individual systems which they suppon, carwneero n 6e neaewaar rope 2.s.11
System 80+ Design Control Document Table 2.9.4-1 Process and Effluent Radiological Monitoring and Sampling System Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 1. The PERMSS has 1. Inspection of the 1. The PERMSS provides components that provide PERMSS componeva the components specified radiological monitoring will be performed. in Table 2.9.4-2.
of gaseous and liquid processing systems and their effluents, airborne radioactivity, radiation areas, and specified plant equipment.
- 2. Displays and alarms of 2. Inspection for the 2. Displays and alarms of the PERMSS safety- existence or retrievability the PERMSS safety-related instrumentation in the MCR of related instrumentation (the MCR air intake instrumentation displays (the MCR air intake radiation monitors, the and alarms will be radiation monitors, the reactor coolant radiation performed. reactor coolant radiation monitors, the high range monitors, the high range containment area contamment area monitors and the monitors and the containment atmosphere containment atmosphere particulate monitors) particulate monitors) exist in the MCR or can exist in the MCR or can be retrieved there. be retrieved there.
- 3. The MCR intake 3. Testing of each monitor 3. Each MCR intake radiation monitors shall will be conducted using monitor is activated upon have the capability for manual controls and receipt of test signals and auto selection and closure simulated automatic the associated control of the most contaminated initiation signals. room intake is closed intake. automatically.
- 4. Operation of each safety- 4. Testing cf each division 4. Each division is activated related PERMSS division (including each channel upon receipt of test can be manually activated of the safety-related signal, from the MCR or portion of the area automatically, radiation monitoring system) will be conducted using manual controls and simulated automatic initiation signals.
O Cert l6ed Desipro Meterial Page 2.9-12
i i
System 80+ Deslan ControlDocument
,O
'd Table 2.9.4-1 Process and Effluent Radiological Monitoring and Sampling System (Continued)
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 5. Each safety-related area 5. Testing of each channel 5. MCR and local alarms radiation monitor channel of the safety-related area are initiated when the ,
monitors the radiation radiation monitors will be simulated radiation level !
level in its assigned area, conducted using reaches a preset limit. l and indicates its simulated input signals. j respective MCR alarm and local audible and l visual alarm (if provided) when the radiation level reaches a preset level. ;
- 6. The following PERMSS 6. Inspection of the as-built 6. The as-built PERMSS j safety-related system will be conforms with the design instrumentation shall be conducted. description.
provided: l
- MCR intake radiation l !
monitor (2/ intake), l
- high range containment l ]
area radiation monitor e containment atmosphere l radiation monitor (particulate channel only),
e primary coolant loop l radiation monitors (2).
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System 80+ Design ControlDocument Table 2.9.4-1 Process and Effluent Radiological Monitoring and Sampling System (Continued)
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 7. The PERMSS safety- 7. Seismic analyses of the 7. An analysis report exists related instrumentation as-built PERMSS safety- which concludes that the (the MCR intake related instrumentation PERMSS safety-related radiation monitors, high will be performed. instrumentation (the range containment area MCR intake radiation radiation monitors, monitors, high range containment atmosphere contamment area radiation monitor radiation monitors, (particulate channel), and contamment atmosphere the primary coolant loop radiation monitor radiation monitors) are (particulate channel), and classified Seismic the primary coolant loop Category I. radiation monitors) are classified Seismic Category I.
- 8. Independence is provided 8. Inspection of the as- 8. Physical separation exists between Class IE installed Class IE between Class IE Divisions, and between Divisions of the Divisions in the Class IE Divisions and PERMSS will be PERMSS. Physical non-Class IE equipment, performed. separation exists between in the PERMSS. Class IE Divisions and non-Class IE equipment in the PERMSS.
O MM Deshp"r Mew (2,95) p,y ,2.g.g4 1
Sy' tem 80+ Design ControlDocument
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(Ad Table 2.9.4-2 Radiological Monitoring and Sample System Components Gaseous Process and Emuent Monitors Gaseous waste management system waste gas discharge (1)
Unit Vent - Normal (1)
Unit Vent - Post Accident (1)
Containment high purge exhaust (1)
Containment low purge exhaust (1)
Main Condenser Evacuation System (1)
Liquid Process and Emuent Monitors Component cooling water system (1/ division)
Liquid waste management system liquid waste discharge (11 Steam generator blowdown (1)
Reactor coolant gross activity (1)
Turbine floor building drains (1)
Station service water system (1/ Division)
Containment cooler condensate tank (1)
Condensate Cleanup System Neutralization Tank Discharge (1)
Airborne Radiation Monitors Containment atmosphere (1)
Radwaste building ventilation exhaust (1)
Fuel building ventilation exhaust (1)
Q Ventilation systems multisampler (2 monitors - 1/ division in Nuclear Annex) (1 in Radwaste Building)
\ j Nuclear annex building ventilation (2 monitors - 1/ division)
Main control room air intake (2/ intake)
Reactor building annulus exhaust (2 monitors - 1/ division)
Reactor building subsphere ventilation exhaust (2 monitors - 1/ division)
Portable airbome Emergency operations facility ventilation (1) v)
Cerdned Design neateriel Page 2.S-15
i lI J
Sy~ tem 80 + Design ControlDocument Table 2.9.4-2 Radiological Monitoring and Sample System Components (Continued)
Area Radiation Monitors Reactor Containment entrance Refueling bridge crane In< ore instrumentation equipment Decontamination area !
Sample room I l
l Main control room l Primary chemistry laboratory New fuel storage area Spent fuel pool bridge Fuel building area l Nuclear annex building (normal operation) i Nuclear annex building (post accident)
Reactor building subsphere (normal operation)
Reactor building subsphere (post accident)
Solid waste drum storag; and handling area Radwaste building loading bay l
llot machine shop Hot instrument shop Radwaste building areas ,
Technical support center j Special Purpose Area Radiation Monitors Main steam lines area (2 monitors - 1/ loop)
Purification filters (1/ filter)
Primary coolant loops (2 monitors - 1/ loop) liigh range containment area monitors (2) ,
Main steam lines (primary-to-secondary leakage) (2 monitors - 1/SG) j l
I, l
O Certthed Desegro Material Page 2.916
System l'0 + Design ControlDocu_ ment 2.10 Technical Support Center and Operations Support Center Design Description The Technical Support Center (TSC) performs a non-safety-related function and is located adjacent to the main control room (MCR) in the nuclear annex. The TSC provides facilities for management and technical support to plant operations during emergency conditions.
The TSC is located less than or equal to two minutes walking time from the MCR.
The TSC has floor space of at least 75 square feet per person for a minimum of 25 persons.
The TSC has radiation detection equipment for monitoring radiation levels within the TSC when the TSC is in use.
The TSC has means for voice communication to the MCR, to on-site emergency support facilities, and to off-site via dedicated or commercial telephone networks.8 Displays of the information from the discrete indication and alarm system (DIAS) and the data processing l system (DPS) exist in the TSC or can be retrieved there.2 The Operations Support Center (OSC) performs a non-safety related function and is located in the nuclear island stnictures. The OSC provides an assembly area separate from the MCR and TSC where operations 3 e support personnel can assemble in an emergency.
The OSC has equipment for voice communications with the MCR and the TSC.
l Inspections, Tests, Analyses, and Acceptance Criteria Table 2.10-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Technical Support Center.
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8 Conunnication Systems are addressed in Section 2.7.25, 2
Display aformation from the DIAS and DPS is addressed in Section 2.5.3.
cerend oneten nooserw rage 2.ro r
Sy~ tem 80+ Design ControlDocument Table 2.10-1 Technical Support Center Design Commitment inspections, Tests, Analyses Acceptance Criteria 1.a) The TSC is located less 1.a) A test of walking time 1.a) The TSC can be reached than or equal to two from the TSC to the in less than or equal to minutes walking time MCR will be performed. two minutes walking from the MCR. time from the MCR.
1.b) The TSC has floor space 1.b) Inspection of the TSC 1.b) Floor space of at least of at least 75 square feet will be performed. 1875 sq. ft. is provided per person for a in the TSC.
minimum of 25 persons.
l.c) The TSC has radiation 1.c) An inspection of the 1.c) Radiation detection detection equipment for radioactivity detection equipment to monitor monitoring radiation equipment in the TSC radiation levels within levels within the TSC will be performed. the TSC is available in when the TSC is in use. the TSC.
1.d) The TSC has means for 1.d) An inspection of the TSC 1.d) Communications voice communications to will be performed. equipment is installed, the MCR, to on-site and voice transmission emergency support and reception are facilities, and to off-site accomplished.
via dedicated or commercial telephone networks.
- 2. Displays of information 2. Inspection for the 2. Displays of information from the DIAS and the existence or retrievability from the DIAS and the DPS exist in the TSC or in the TSC of the DPS exist in the TSC or can be retrieved there. information from the can be retrieved there.
DIAS and the DPS will be performed.
- 3. The OSC is located in 3. Inspection of the location 3. The OSC is located in the nuclear island of the OSC will be the nuclear island structures. performed. structures.
- 4. The OSC has equipment 4. Testing of the equipment 4. Communications for voice communication for voice communication equipment is installed with the MCR and the will be performed, and voice transmission TSC. and reception are accomplished.
O W Des &n Materief page 2.10 2
- . . _ _. _ _ . . ~ . . . _ - _ . ..- , _ . . _ _ . - -
System 80+ oesion control cocument !
- 2.11 Initial Test Program it
^
An Initial Test Program is performed during and following construction activities and construction-related inspections and tests and extends to a declaration of commercial operation. The Initial Test Program has
. preoperational and stanup test phases. ,
i As part of the Initial Test Program, preoperational testing is conducted to demonstrate that structures, j systems, components, and design features of the as-built plant meet the performance requirements of the j
- design. Equipment functional tests, system operational tests, and system vibration and expansion ;
' measurements are conducted during preoperational testing. l t
As part of the initial Test Program, stanup testing is conducted to demonstrate that the integrated plant l with the nuclear fuel in the mactor pressure vessel meets the performance requirements of the design. l Stanup testing is performed with the plant operating at power levels ranging from zero power to full rated i 3
i power. Startup testing is conducted at five test conditions: open vessel (fuel load), heatup, low power, .
mid-power, and high power. Stanup testing includes testing of:
4 1 i
- nuclear fuel system performance, 1 1
}
- steady-state plant performance; i
e control system performance; and i
e transient performance. l The Initial Test Program is performed using detailed preoperational and stanup procedures to control the ;
conduct of testing. Detailed test procedures delineate the test methods to be used and the applicable acceptance criteria against which performance is evaluated. Test procedures are developed from {
preoperational and stanup test specifications. Approved test procedures are made available to the NRC l for review. Administrative procedures are used to control the conduct of the test program; the review, ;
evaluation and approval of test results; and test record retention. !
i i The tests specified in Section 2.0 may be a subset of the Initial Test Program. j
' This section represents a commitment that combined operating license applicants referencing the cenified .
design will implement an Initial Test Program that meets the objectives presented above. {
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-2,12 ' Human Factors ;
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12.12.1 Main Control Roosa {
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Design Description ]
i The Mtin Control Room .t (MCR) permits execution of MCR ' tasks performed by MCR operators to !
operate the plant and maintain plant safety. The MCR 'provides suitable workspace and environment for - !
continuous <-=y and use by MCR operators when the MCR is used for Plant Control. The MCR -
makes available the annunciators, displays, and controls to operate the plant and maintain plant saferf,
! including at least those annunciators, displays, and controls identified in Table 2.12.1-1. Other . .
annunciators; displays, and controls for systems operation are described in the design descriptions for the . j i respective systems. l l
- . The Basic Configuration of the MCR is as shown on Figure 2.12.1-1.
j
- The MCR comins the master control console, the auxiliary console, the safety console, the control room - !
- supervisor (CRS) Console, administrative support facilities, and the integrated process status overview j (IPSO).
l Control panels with Class 1E instrumentation are classified Seismic Category I. j i
I The MCR is located in the nuclear annex within Sre and ventilation isolation boundaries.
- t. i MCR consoles are organized functionally according to the following: !
i Master Control Console Auxiliary and Safety Consoles l i- ' Reactor Coolant System Heating. Ventilation & Air Conditioning -
Chemical & Volume Control System Cooling Water Systems ;
Plant Monitoring & Control Engineered Safety Features .!
Feedwater & Condensate Systems Safety Monitoring l Turbine Control Secondary Auxiliaries l
- Switchyard l Electrical Distribution j i The CRS console provides a workstation from which the CRS coordinates MCR ' operations. l Administrative support facilities provide office workspace. The IPSO provides safety parameter display information at a fixed location that can be viewed from the MCR consoles and administrative support l
1 facilities.
Inapection, Test, Analyses, and Acceptance Criteria ;
i Table 2.12.1-2 specifies the inspections, tests, analyses, and associated acceptance criteria for the Main l t Control Room. l
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. are addressed in Sections 2.1.1, 2.7.17, 2.7.24, 2.7.25, '2.7.26, and 3.2 respectively. !
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Table 2.12.1-1 MCR Minimum Inventory of Fixed Position Annunciators, Displays and Controls Annunciators (" - Displays Controls Offsite Bus voltage status X 120 VAC Vital load center voltage status X X 125 VDC Vitalload center voltage status X X 24 KV Main Turbine Generator output breaker position X X X 4.16 KV Class 1E bus breaker positions (supply & X X ,
crossover) 4.16 KV Class IE voltage status X X 4.16 KV Diesel Generator output breaker position X X 4.16 KV Diesel Generator stan control X X 4.16 KV Diesel Generator synchroscope X X 4.16 KV Reserve Aux Xfmr output voltage status X 4
k 480 VAC Class IE voltage status X X Annulus ventilation control setpoint X X Annulus ventilation damper position X X Annulus ventilation fan on/off X X Atmospheric dump valve position X X CEA position X CET temperature X CIAS actuation X X CIAS success monitor X X CCW HX inlet valve position X X CCW IIX outlet valve position X X CCW l{X outlet flow X CCW pumps on/off X X CCW surge tank level X Containment hydrogen level (when analyzer is in X X operation)
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Containment pressure X X CorMied Doelpre Meteriel Page 2.12 3
System 80+ Design ControlDocument l I
Tanle 2.12.1-1 MCR Minimum Inventory of Fixed Position Annunciators, Displays and Controls (Continued)
Parameter Description Annunciators m Displays Controls Containment radiation X X CSAS actuation X X Containment Spray flow X Containment Spray pump on/off X X Containment Spray pump discharge valve position X X Containment temperature X X DVI valve position X X EFAS actuation X X EFW flow control valve position X X EFW header flow X EFW motor-driven pump on/off X X EFW pump suction pressure X EFW steam-driven pump on/off X X EFW-to-SG isolation valve position X X EFW Storage Tank level X X llot Leg injection valve position X X IRWST level X Main Control Room HVAC isolation dampers X X Main Steam radiation (Area monitors & Line monitors) X SG safety valve position X MSIS actuation X X X Nuclear Annex building ventilation radiation X Primary Coolant Radiation X X Pzr Backup Heaters on/off X X Pzr Level X X Pzt Pressure X X Rapid Depressurization valve position X X RCP on/off X X Corned Design Matend Pepe 2.124
System 80+ Design controlDocument
- 1 k Table 2.12.1-1 MCR Minimum Inventory of Fixed Position Annunciators, i Displays and Controls (Continued)
Annunciators") Displays Controls RCS Cold leg temperature X l
RCS Hot Leg temperature X RCS Pressure X RCS subcooling margin X X i Reactor Building subsphere ventilation radiation X Reactor Cavity level X X Reactor Coolant gas vent valve position X X '
Reactor power (NI) X Reactor Trip (RPS) X X Reactor Vessel level X X SCS flow (while SCS is in operation) X X ,
SCS Isolation valve position (& LTOP) X X X SCS HX Bypass Valve position X X SCS HX CCW supply / isolation valve position X X SCS IlX/ Bypass Inlet & Outlet temperature (when SCS ,
is in operation) X SCS HX outlet valve position X X SCS pump on/off X X SCS/ CSS pump suction cross-connect valve position X X t
SCS/ CSS pump discharge cross-connect valve position X X SlAS actuation X X Si flow X !
SI pump on/off X X Si throttling isolation valve position X X Spent Fuel Pool level X Stanup Rate (NI) X (Oy CCW IlX Station Service Water inlet isolation valve position X X f
Cerened Coeners nieterief Page 2.12-5
System 80+ Design ControlDocument Table 2.12.1-1 MCR Minimum Inventory of Fixed Position Annunciators, Displays and Controls (Continued)
Parameter Description Annunciatorsm Displays Controls CCW IIX Station Service Water outlet isolation valve X X position CCW HX Station Service Water outlet flow X SSW pump on/off X X SG Blowdown sample radiation X SG level X X SG pressure X X Vacuum Puntp Activity X Turbine Trip X X m Annunciators are alarms and other alerting displays designed to direct opera *or attention.
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CutiM Deskre Med Page 2.12-6
Sy: tem 80+ Design controlDocument O
b' Table 2.12.1-2 Main Control Room ;
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 1. The Basic Configuration 1. Inspection of the as-built 1. For the components and of the MCR is as shown MCR configuration will equipment shown on on Figure 2.12.1 1. be conducted. Figure 2.12.1-1, the as-built MCR conforms with the Basic Configuration.
- 2. The MCR makes 2. Human Factors 2. The MCR makes available the Engineering (HFE) available the annunciators, displays, availability verification annunciators, displays, and controls identified in inspection of the as-built and controls identified in Table 2.12.1-1. MCR will be performed. Table 2.12.1-1.
- 3. The MCR provides 3. HFE suitability 3. The MCR workspace and suitable workspace and inspection against environment are i
environment for verification criteria will deternnned to be suitable continuous cccupancy be performed. for use by MCR +
and use by MCR operators.
operators when the MCR is used for plant control.
,r 4 The MCR permits 4. Testing and analysis 4. The test and analysis execution of MCR tasks against the validation results demonstrate performed by MCR criteria using a facility validation of MCR task operators to operate the that physically represents execution by MCR plant and maintain plant the MCR configuration operators to operate the ,
safety, and dynamically plant, and maintain plant represents the MCR safety.
interface characteristics and the operating characteristics and respo ses of the System 804 tgn will be perfon a;d.
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Cerened Design ate &~ rope 2.12 7
Syatem 80+ Design Control Document 2.12.2 Remote Shutdown R.oom Design Desedption The Remote Shutdown Room3 (RSR) permits execution of RSR tasks performed by RSR operators to place and maintain the plant in a safe shutdown condition. The RSR provides suitable workspace and environment separate from the main control room (MCR) for use by RSR operators in the event that the MCR becomes uninhabitable. The RSR makes available the annunciators, displays, and controls to achieve and maintain prompt shutdown of the plant and maintain safe shutdown conditions including at least those annunciators, displays, and controls identified in Table 2.12.2-1. The RSR provides capability for RSR operators to perform RSR tasks to achieve subsequent cold shutdown of the plant.
The Basic Configuration of the RSR is as shown on Figure 2.12.2-1.
The RSR contains the Remote Shutdown Panel. The Remote Shutdown Panel provides a workstation from which RSR operators perform RSR operations.
Controls exist in the RSR to stop the reac'.or coolant pumps (RCPs), trip the reactor, control the emergency feedwater (EFW) steam-driven purnp turbine speed, and to start and stop those other pumps, open and close those valves, and energize or de-e vrgize those pressurizer heaters listed in Tale 2.12.2-1.
Control panels with Class IE instrumentation are claaified Seismic Category 1.
The RSR is located in the nuclear annex within fire ar.d ventilation isolation boundaries.
Inspection, Test, Analyses, and Acceptance Criteria Table 2.12.2-2 specifies the inspections, tests, analyses, and acceptance criteria for the RSR.
Nuclear Island structures, ventilation, fire protection, comrnunications, lighting, and radiation protection O
are addressed in Sections 2.1.1, 2.7.17, 2.7.24, 2.7.25, 2.7.26, and 3.2 respectively.
cenined t> esp Areter.at Page 2.r2-s
< 1 Sy:trm 80+ oesign controlDocument ;
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l Remote Shutdown Room Figure 2.12.2-1 i Cordned Design historie! Pope 2.12-9 t
System 80+ Design ControlDocument Table 2.12.2-1 Minimum Inventory of Available RSR Annunciators, Displays and Controls Parameter Description Annunciators" Displays Controls Reactor Power (Neutron Logarithmic Power) X RCS Cold Leg Temperature X RCS Ilot 12g Temperature X PZR level X PZR Pressure X Reactor Trip (RPS) X X SG Level X SG Pressure X CVCS Charging Flow X CVCS Charging Pressure X Boric Acid Storage Tank Level X l IRWST Level X EFW Steam-Driven Pump Suction Pressure X X EFW Motor-Driven Pump Suction Pressure X X EFW Steam-Driven Pump Discharge Pressure X EFW Motor-Driven Pump Discharge Pressure X EFW Steam-Driven Pump Turbine Inlet Pressure X EFW Steam-Driven Pump Flow X EFW Motor-Driven Pump Flow X EFW Steam Driven Pump Recirculation Flow X EFW Motor-Driven Pump Recirculation Flow X EFW Storage Tank level X X EFW Steam-Driven Pump Turbine Speed X X EFW Turbine Trip and Throttle (Stop) Valve Open/Close Position (Trip / Reset) X X X Ultimate lleat Sink Status X X 4.16 KV Diesel Generator Status (Emergency) X Cadfied Design Motonid Pope 2.12-10
Sy' tem 80+
m contrat Document
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Table 2.12.2-1 Minirnum Inventory of Available RSR Annunciators, Displays and Controls (Continued)
Parameter Desesiption AnnunciatoniD Displays Controls Reactor Coolant Pump Trip X X l PZR Backup Heaters (Groups 1 & 2) On/Off X X Atmospheric Dump Valve Position X X ADV Block Valve Position X X PZR Auxiliary Spray Valve Position X X Reactor Coolant Gas Vent Valve Position X X CVCS Charging Pump On/Off X X L
- Ixtdown Isolation Valve Position X X RCP Seal Bleedoff Valve Position X X MS15 Actuation X X EFW Motor Driven Pump On/Off X X b EFW Steam Driven Pump On/Off X X EFW Flow Control Valve Position X X EFW-to-SG lsolation Valve Position X X E'M Steam Supply Bypass Valve Position X X EFW Steam Supply Isolation Valve Position X X PZR Pressure Control Serpoint X X SG Pressure Control Setpoin! X X SCS Suction Line Isolation Valve Interlock Status X SCS HX/ Bypass Irdet & Outlet Temperature (when SCS is 2
in operation) X SCS Flow X X SCS HX Bypass Valve Position X X SCS Pumps On/Off X X SIT Pressure X SIT Vent Valve Position X X
- SIT isolation Valve Position X X SCS isolation Valve Position X X j cwood onw noenmer tr ttssi rege 2.1211
System 80+ Design ControlDocument Table 2.12.2-1 Minimum Inventory of Available RSR Annunciators, Displays and Controls (Continued)
Parameter Description Annunciators
- Displays Controls SCS fiX Outlet Valve Position X X SCS Wannup Bypass Valve Position X X SI Flowm x Si Discharge fleader Pressure (2) x St Pump On/Of/2) X X St Throttling Isolation Valve Position
- X X O
m Annunciators are alarms and other alerting displays designed to direct operator attention.
- Indication for two discharge headers only 9
System 80+ oesign controlDocument O
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'v' Table 2.12.2-2 Remote Shutdown Room Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 1. The Basic Configuration of 1. Inspection of the as-built 1. For the components and the RSR is as shown on RSR configuration will be equipment shown on Figure 2.12.2-1. conducted. Figure 2.12.21, the as-built RSR conforms with the Basic Configuration.
- 2. The RSR makes available 2. Human Factors 2.a) The as-built RSR makes the annunciators, displays, Engineering (HFE) available the annunciators, and controls identified in availability verification displays, and controls Table 2.12.2-1. inspection of the as-built necessary to achieve and RSR will be performed. maintain prompt hot shutdown of the reactor.
2.b) The as-built RSR provides
- capability for RSR operators to perform RSR tasks to achieve subsequent cold shutdown of the plant.
- 3. The RSR provides suitable 3 HFE suitability inspection 3. The RSR workspace and workspace and against verification criteria environment are environment for use by will be performed. determined to be cuitable RSR operators. for use by RSR operatois.
- 4. The RSR permits 4. Testing and analysis 4. The test and analysis execution of RSR tasks against the validation results demonstrate performed by RSR criteria using a facility that validation of RSR task operators to shutdown the physically represents the execution by RSR plant and maintain safe RSR configuration and operators to achieve and ,
shutdown conditions. dynamically represents the maintain safe shutdown
- RSR interface conditions.
characteristics and the operating characteristics of the System 80+ design will be performed.
- 5. Controls exist in the RSR 5. Testing will be performed 5. Controls in the RSR to stop the RCPs, trip the using the controls in the operate to stop the RCPs, reactor, control the Elv RSR. trip the reactor, control steam. driven pump turbine the EIN steam-driven speed, and to start and pump turbine speed, and stop those other pumps, to start and stop those open and close those other pumps, open and valves, and energize or de- close those valves, and .
energire those pressurizer energize or de-energize heaters listed in Tale those pressurizer heaters 2.12.2-1. listed in Table 2.12.2-1.
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I System '80 + Desion coneet Document - l g
3.0 : Non-Systesa Based Design Descriptions and ITAAC i
~3.1 - Piping Design .
4 Design Description I
The requirements for piping design in this section apply to ASME Code Class 1,2, and 3 piping that is c classified as Seismic Category I unless otherwise noted. j i
' Piping classified as Seismic Category I is required to withstand the dfects of a safe shutdown earthquake - l
< (SSE),' maintain dimensional stability, and remain functional. Seismic Category I piping, structures, }
systems, and components assure: (1) the integrity of the reactor coolant pressure boundary, and (2) the -I capability to shut down the reactoc and maintain it in a safe shutdown condition, or (3) the capability to j prevent or mitigate the consequences of accidents which could result in potential offsite exposures. (
i Piping which does not perform a safety related function but whose structural failure or interaction could degrade the functioning of a Seismic Category I structure, system, or component to an unacceptable safety l 1 i level or could result in an incapacitating injury to an occupant of the control room is classified as Seismic j
- Category II. :
3 1
- Seismic Category I piping is designed to meet the requirements of the ASME Code,Section III. Seismic j l Category 11 piping is designed and constructed such that the SSE will not cause failure in a manner to j adversc!y affect a ufety system or result in an incapacitating injury to a control room occupant.
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- Applicable piping loads due to pressure, gravity, thermal expansion, seismic excitation, wind, tornado, fluid transients, thermal stratification, missiles, and postulated pipe breaks are considered in the piping
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. ' analyses. Analytical methods and load combinations uv.d for analysis of piping systems will be !
referenced or specified in the ASME Code certified stress report. Computer programs used for piping l
- . system dynamic analysis shall be terAurked. j i' The as-built ASME Code Section III piping will be reconciled with the piping design requirements l described herein. The as-built reconciliation will be documented in the as-built piping report. ]
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, Piping systems are designed to reduce the potential for effects of erosion / corrosion, and to reduce the .) '
6 potential for waterhammer and steam hammer. Piping system supports for Seismic Category I and II
- piping systems are designed to meet the requirements of the ASME Code Section III, Subsection NF.
Pipe loads applied to attached equipment are shown to be less than the equipment allowable loads.
I For those piping systems using ferritic materials as permitted by the design specification, the materials
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and fabrication processes will be chosen to ensure that the system is not susceptible to brittle fracture U
under the expected service conditions. For those piping systems using austenitic stainless steel materials
- as permitted by the design specification, the material and fabrication process will be selected to reduce the possibility of cracking during ee ; ice. Chemical, fabrication, handling, welding, and examination requirements that reduce the potential for cracking shall be employed.
, Piping systems classified as ASME Code Section III Class 1,2, or 3 are designed to maintain dimensional stability and fanctional integrity under design loadings expected to be experienced during a 60 year design I o -
life.
j ounad Deep annanw rone1r r
System 80 + Design C^ntrolDocument Design of piping systems provides for clearances between adjacent piping, components, and other structures when the piping moves due to design static, dynamic, and thermal loadings.
The following piping systems are designed to meet leak-before-break (LBB) criteria:
- Reactor coolant system hot leg piping, reactor coolant pump (RCP) suction piping and RCP discharge piping, e Surge line,
- Main steam lines inside containment from the steam generator to the anchor at the containment penetrations,
- Shutdown cooling lines inside containment from the reactor coolant system to the anchor at the containment penetration, and e Direct vessel injection lines inside containment from the reactor vessel to the safety injection tank and the anchor at the containment penetration.
LBil acceptance criteria are established and LBB evaluations are performed for each piping system designed to meet LBB criteria. For each piping system qualified for L.BB, the as-built piping and materials will be reconciled with the bases for the LBB acceptance criteria.
Structures, components, and systems required for safe shutdown are protected from the dynamic and environmental effects of postulated pipe breaks and cracks in Seismic Category I and non-nuclear safety-related (NNS) piping systems where consideration of these dynamic effects is not eliminated by LBB. )
Each postulated pipe crack and break shall be documented in a pipe break analysis report. Design of a features which protect these items consider, as applicable, pipe whip, water spray, jet impingement, floodmg, compartment pressurization, and environmental conditions in the area where the piping is located.
Structures, systems, and components that are required to be functional during and following an SSE are protected against the effects of spraying, flooding, pressure, and temperature due to postulated pipe breaks and cracks in Seismic Category I and NNS piping systems.
Inspections, Tests, Analyses, and Acceptance Criteria Table 3.1-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Piping Design.
O Cornned Design Material Page 3.12
Sy tem 80 + Design Cont of Document s~-
LJ Table 3.1-1 Piping design Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 1. The as-built piping is 1. A reconciliation analysis 1. An as-built piping stress reconciled with the as using the as designed and report exists and designed piping as-built information will concludes that the as-configurations. be performed. built piping has been reconciled with the documents used for design. The as-built piping is reconciled with the piping design requirements described in the piping design description. For ASME Code Class piping, the as-built stress repon includes the ASME Code Cenified Stress Repon and documentation of the results of the as-built reconciliation analysis.
p 2. Piping systems classified as ASME Code Section
- 2. Inspection for the existence of ASME
- 2. ASME design repons for piping systems classified
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III Class 1,2, or 3 are design repons will be as ASME Code Section designed to maintain performed. III Class 1,2, or 3 exist dimensional stability and and conclude that the functional integrity inder design complies with the design loadings expwxi requirements of the to be experience,' .tng ASME Code,Section III.
a 60-year desiga life.
- 3. For each piping system 3. For each piping system 3. A LBB evaluation repon qualified for LBB, the qualified for LBB, an exists which documents as-built piping and inspection of the LBB that leak-before-break materials will be evaluation repon will be acceptance criteria are reconciled with the bases performed, met by the as-built piping for the LBB acceptance and piping materials.
criteria.
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Ssetem 80+ Design ControlDocument Table 3.1-1 Piping design (Continued)
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 4. Structures, components, 4. For piping systems with 4. A pipe break analysis and systems required for postulated pipe breaks, report exists and safe shutdown are an inspection of the pipe concludes that Seismic protected from the break analysis report will Category I structures, dynamic and be performed. An systems, and components environmental effects of inspection of the as-built remain functional after postulated pipe breaks high and moderate postulated pipe breaks.
and cracks in Seismic energy pipe break The pipe break analysis Category I and non- mitigation features will repon includes the nuclear safety-related be performed. results of inspections of (NNS) piping systems high and moderate where consideration of energy pipe break these dynamic effects is mitigation features not eliminated by LBB. (including spatial Each postulated pipe separation).
crack and break shall be documented in a pipe break analysis report.
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Certified Destyn Atatenet page 3.p4
System 80+ oestan conuot Document 0qj 3.2 Radiation Protection :
i Design Description Radiation Protection features in the plant provide for limitation of radiological exposures to plant personnel and to the general public from plant operations consistent with NRC radiological exposure regulations.
The Radiation Protection design includes provisions for: ;
e shielding or temporary shielding of rooms, corridors, cubicles, labyritth access, and operating ,
i areas commensurate with their expected occupancy and use; e plant shielding to permit operators to perform actions that may reqbire operator access during and following a design basis raccident;
- ventilation features to reduce airborne radioactivity levels consistent with personnel access requirements; e airborne radioactivity monitoring in those areas of the plant where, consistent with occupancy and use, airborne contamination could restrict personnel access.
In each case described above, the Radiation Protection design is based on expected radiation environments associated with operational modes and post-design basis accident conditions. ,
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i Inspections, Tests, Analyses, and Acceptance Criteria Table 3.2-1 specifies the inspections, tests, analyses, and associated acceptance criteria for Radiation Protection. i l
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System 80+ Design ControlDocument Table 3.2-1 Radiation Protection Design C==hment Inspections, Tests, Anal.v;es Acceptance Criteria
- 1. The Radiation Protection 1. An analysis of the expected 1. Maximum radiation levels design includes shielding or radiation levels in each plant are less than or equal to the l
provisions for temporary area will be performed to radiation levels in the shielding of rooms, verify the adequacy of the radiation zones specified in corridors, cubicles, labyrinth shielding design. This Table 3.2-2. Plant layout is access, and operating areas analysis shall consider the such that access to higher following: zones (areas with higher commensurate with their expected occupancy and use. a) Confirmatory calculations dose rates) is from lower shall consider significant zoned areas. Corridors and radiation sources (greater normal traffic areas are than 5% contribution) for an Zone 3 or less. Control area. Radiation source Rooms are Zone 2 or less, strength in plant systems and Radiation zone designations components will be for components during determined based on an normal operating conditions assumed source term of are listed in Table 3.2-3.
0.25% fuel cladding defects and a core inventory commensurate with a 3914 MWt equilibrium core.
Source terms shall be adjusted for radiological decay and buildup of activated corrosion and wear products, b) Commonly accepted shielding codes, using nuclec propenies derived frorn well kxwn references slull be used w model and evaluate plant radiation environments.
- 1. For nons;omplex l
geometries, point kernel shielding codes may be used.
l ii. For complex geometries, more sophisticated two or three dimensional transport codes shall be used.
- 2. The plant design shall have 2. Using the methods identified 2. Shielding design of a zone is provisions to reduce in 1. above, radiation levels such that radiation from radiation exposure from in zones shall be evaluated adjacent sources of radiation adjacent sources of for contribution from shall contribute no more radiation. adjacent sources of than a small fraction of the radiation. dose rate in the zone.
O CerenedDealgte Afstoriel (2/95) Page 3.2 2
System 80+ Deslan controlDocument O
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?'s Table 3.2-1 Radiation Protection (Continued) l l
Design Comunisment Inspections, Tests, Analyses Acceptance Criteria I I
- 3. The Radiation Protection 3. An analysis will be 3.a) The predicted individual l design includes plant performed using design personnel occupational i shielding to permit basis accident source terms doses are less than or equal l operators to perform actions and calculational methods to 5 rem to the whole body that may require operator consistent with 1.b. above total, or its equivalent, over access during and following to determine the expected the entire time period (s) l a design basis accident. radiation levels in areas of during which operator the plant that may require access is required. ,
operator access during and following a design basis 3.b) For areas requiring accident, continuous occupancy, the predicted individual personnel occupational dose
! rates do not exceed 15 mrem /hr, averaged over 30 days.
4 The Radiation Protection 4. An anal', sis will be 4 The analysis concludes that design includes provisions performed to predict airflows are from areas of for ventilation to limit airborne radioactivity lower potential airborne airborne radioactivity to concentrations in rooms, contamination to those of levels that permit persannel corridors, cubicles, and higher potential airborne access, operating areas during contamination, prior to j removal by the filters or
/ normal operations. Total ventilation flow rates and vent system, and, equipment leakages will be considered in the analysis, l which will be based on -
source terms consistent with l item 1.a. l l 1
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System 80+ Design ControlDocument 1
Table 3.2-1 Radiation Protection (Continued)
Design Commitment Inspections, Tests, Analyses Acceptance Criteria 4 (Continued) 4. (Continued) 4. (Continued) 4.a) For normally occupied areas of the plant, (i.e.,
those areas requiring routine access to operate and maintain the plant),
equilibrium concentrations of airborne radionuclides will be a small fraction of the Derived Air Concentrations in NRC dose regulations.
4.b) For areas that require infrequent access (such as for non-routine equipment maintenance), the ventilation system shall be capable of reducing radioactive airborne concentrations to the Derived Air Concentration in NRC dose regulations during the periods that occupancy is required.
4.c) For rooms where access is not anticipated to perform scheduled maintenance, plant design shall provide features to reduce airborne contamination spread to other areas of lower contamination.
O CertrnedDesiger Aceterial Page 3.24
System 80 + oesign controlDocument w
v
) Table 3.2-1 Radiation Protection (Continued)
Design Commitment Inspections, Tests, Analyses Acceptance Criteria I
- 5. The Radiation Protection 5. An analysis will be 5. Airborne radioactivity '
design includes provisions performed to identify plant monitoring in the plant is for airborne radioactivity areas that require airborne consistent with the analysis monitoring in those areas of radiation monitoring. in that:
the plant where, consistent with expected occupancy 5.a) each monitor has the and use, airborne capability of detecting the contamination could restrict time-integrated radioactivity personnel access. concentrations of the most limiting internal dose particulate and iodine radionuclide in each monitored area equivalent to the Derived Air Concentrations in NRC dose regulations for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> (10 DAC-hours);
5.b) each monitor provides a calibrated response.
representative of the radioactivity concentrations wnhin the monitored area; I
5.c) each monitor provides local audible alarms with visual I
alarms in high noise areas.
and with variable alarm setpoints and readout capability.
- 6. The plant design shall 6. Using the methods identified 6. As a result of normal provide radiation shielding in 1. above, the radiation operations, the analyzed to protect the general public dose to the maximally radiation does from direct outside of the controlled exposed member of the and scattered radiation to area during normal general public outside of the the maximally exposed operations, consistent with controlled a:ea during member of the public Federal regulations. normal operations from outside of the controlled direct and scattered area is equal to or less than radiation shall be a small fraction of the determined. Federal regulations.
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Syntem 80+ oesign ControlDocument Table 3.2-2 Radiation Zone Designations During Normal Operating Condiilons Access Acceptance Criteria Zone Designations Dose Rate (mrem /hr)* Allowed Occupancy 1 s 0.5 Uncontrolled, unlimited access (plant personnel) 2 > 0.5 to < 2.5 Controlled, limited access, 40 hr/wk to unlimited 3 2 2.5 to < 15 Controlled, limited access, 6 to 40 hr/wk 4 215 to 6 100 Controlled, limited access, I to 6 hr/wk 5 2 100 Normt'.ly unaccessible, access only permitted by radiation protection personnel I hr/wk.
- Dose rates calculated at 30 cm from the source of the radiation, not contact dose rates.
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Sy~ tem 80+ Design ControlDocument f~N
( ,/ Table 3.2-3 Radiation Zone Designations for Components During Normal Operating Conditions Zone Designation 1
" F"' Operating Shutdown Annulus Ventilation Systern Filters 2 2 Boric Acid Batch Tank 3 3 Boric Acid Concentrator 5 5 Boric Acid Concentrator lon Exchanger 5 5 Boric Acid Filter 5 5 Boric Acid Makeup Pumps 5 5 Channel A, B, C, D Battery Room i 1 Charging Pumps 5 5 Charging Pump Miniflow Heat Exchanger 5 5 Chemical and Volume Control System 5 5 Equipment Drain Sumps O\
V Condensate Cooling Water Surge Tank 2 2 Containment Spray Heat Exchanger 3 $
Containment Spray Miniflow licat Exchanger 3 5 Containment Spray Pump 3 5 Control Complex Corridors up to the Radiation 1 1 Access Control Point Control Room 1 1 Control Room IIVAC Areas 1 1 Deborating lon Exchangers 5 5 j Division 1 Battery Room 1 1 Division 2 Battery Room i 1 Emergency Feedwater Motor Driven Pump 2 2 Emergency Feedwater Turbine Driven Pump 2 2 Equipment Drain Sumps 5 5 Equipment Drain Tank 5 5 Essential Chillers 1 I Floor Drain Sumps 5 5 Ceruned Des &n Makwief Page 2.2 7
Sy~ tem 80 + Design controlDocument Table 3.2-3 Radiation Zone Designations for Components During Normal Operating I
Conditions (Continued)
Zone Designation' !
Component Operating Shutdown ;
1 Fuel Pool Cooling Pumps 3 3 Fuel Pool Filters 5 5 Fuel Pool Heat Exchanger 3 3 Fuel Pool lon Exchangers 5 5 )
Fuel Pool Purification Pumps 5 5 Fuel Transfer Tube 5 5 Gas Stripper 5 5 Hi Purge Filters 2 2 Holdup Pumps 5 5 Letdown Heat Exchanger 5 5 Lo Purge Filters 2 2 Nuclear Annex Ventilation Filters 2 2 Post-accident Hydrogen Recombiners 2 2 Preholdup lon Exchanger 5 5 Purification Filters 5 5 Purification lon Exchangers 5 5 Radiation Access Control Point i 1 Reactor Coolant System 5 5 Reactor Vessel Hot and Cold legs Steam Generators Reactor Drain Filter 5 5 Reactor Drain Pumps 5 5 Reactor Drain Tank 5 5 Reactor Makeup Water Filter 5 5 Reactor Makeup Water Pump 4 3 Regenerative Heat Exchanger 5 5 Remote Shutdown Room i 1 Certrhed Design Materd Pope 3.2 8
Sy tem 80+ Design contror Document l t
i Table 3.2-3 Radiation Zone Designations for Components During Normal Operating Conditions (Continued) ,
i Zone Designation' Component Operating Shutdown Resin Sluice Tanks 5 5 i
Safety Injection Filters 5 5 Safety Injection Pump 3 5 Sarnpling Panels 5 5 i
Sampling Panels Pipe Chase 5 5 Seal Injection Heat Exchanger 5 5 ,
Shutdown Cooling System Heat Exchanger 3 5 Shutdown Cooling System Pump 3 5 l
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t Spent Fuel Pool (Bottom) w/ fuel 5 5 Subsphere Ventilation System Filters 2 2 ,
Vital I & C Channel A, B, D D 1 1 (
Volume Control Tank 5 5 8
Radiation Zone Designations are provided in Table 3.2 2.
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3.3 Design Reliability Assurance Program The Design Reliability Assurance Program (D-RAP) is a program that will be performed during the !
detailed design and equipment specification phase prior to initial fuel load. The D-RAP evaluates and j prioritizes the structures, systems and components (SSCs) in the design, based on their degree of risk significance. The D-RAP will identify the dominant failure modes for the risk-significant SSCs. The D-RAP will also identify the key assumptions and risk insights for the risk-significant SSCs.
The D-RAP scope includes risk-significant SSCs as determined by probabilistic, deterministic, or other methods used for design certification to identify and prioritize risk-significant SSCs. 1 The D-RAP purpose is to provide reasonable assurance that the plant design proceeds in a manner that is consistent with the original bases and design assumptions for the risk insights for the risk-significant SSCs.
The D-RAP objectives are to provide reasonable assurance that the plant is designed such that: (1) it is consistent with the assumptions and risk insights for these risk-significant SSCs, (2) the risk-significant SSCs will not degrade to an unacceptable level during their design life, (3) the frequency of transients that challenge these SSCs will be acceptably low, and (4) these SSCs will function reliably when challenged.
i Inspections, Tests, Analyses, and Acceptance Criteria Table 3.3-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Design Reliability Assurance Program.
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System 80+ Desian controlDocument i Table 3.3-1 Design Reliability Assurance Program i Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 1. 'Ihe Design Reliability 1. Inspections of the design 1.a) Documentation exists that Assurance Program (D- reliability assurance describes the scope, '
RAP) includes: scope, program will be purpose, and objectives of purpose, objectives; the conducted. D-RAP used during plant process used to evaluate design, and concludes that and prioritize the the detailed design of risk ,
structures, systems and significant SSCs is components (SSCs); and consistent with the D-RAP the list of SSCs designated Design Description.
as risk-significant. For those SSCs designated as 1.b) Documentation exists and risk-significant, the concludes that the process
. process used to determine (probabilistic, dominant failure modes deterministic, or other) considered industry used to evaluate and experience, analytical prioritize the SSCs in the ,
models, and applicable design is based on the i requirements. Also, for risk-significance of the f those SSCs designated as SSCs.
risk-significant, the key assumptions and risk 1.c) A list of SSCs exists that
. is based on the risk-gO insights considered operations, maintenance, significance of the SSCs. .
and monitoring activities.
1.d) For those SSCs designated as risk-significant:
- i. Documentation exists and concludes that the process to determine dominant failure modes considered ;
industry experience, analytical models, and applicable requirements. i li. Documentation exists and concludes that the key [
assumptions and risk insights from probabilistic, deterministic, or other methods considered operations, maintenance, ,
and monitoring activities. I b :
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System 80+ Deslan controlDocument 4.0 ' Interface Requirements The System 80+ Certified Design encompasses interfaces to certain portions of the essentially complete System 80+ nuclear plant design that are not within the scope, and are not a pan of, the Cenified Design (the "out-of-scope" ponions). This section indicates the location in Section 2 where requirements for those interfaces are specified, including requirements pertaining to certain design attributes and performance characteristics of the out-of-scope ponions of the design. These interface requirements were established to allow completion of the final safety analysis and the design-specific probabilistic risk assessment.
Some interface requirements encompassed by the System 80+ Cenified Design are written to specify physical characteristics of out-of-scope portions of the design (e.g., separation of mechanical Divisions),
while other interface requirements are written to specify a functional requiremeia (e.g., heat removal capacity) of the out-of-scope portion of the design.
' An application for a combined license (COL) that references the System 80+ Certified Design must describe design provisions that comply with the interface requirements and must also provide ITAAC aimed at verifying that the as-built facility has those design provisions.
. ITAAC Jutification The interface requirements specified for the System 80+ Certified Design are similar in form, and in the type of content, to the Design Descriptions specified in Section 2 for individual System 80+ systems.
Thus, ITAAC can be developed for System 80+ interface requirements in the same manner as ITAAC were developed for System 80+ systems, j Method for ITAAC Verification f
The method to be used for verification of interface requirements must be specified in the COL ITAAC ,
- and will be identical to the methods specified for verification of Design Descriptions in Section 2. That ;
is, the interface requirements themselves will be specified in ITAAC tables as individual Design l Commitments. Verification that the requirements are met will be accomplished through Inspections, ,
, Tests, and Analyses also specified in the ITAAC tables. Finally, Acceptance Criteria to verify that the !
as-built facility conforms with the interface requirements identified in the Design Commitments will also l be specified in the ITAAC tables. Fuel load may not occur without each of those Acceptance Criteria ;
being met.
]
l i 4.1 Offsite Power System j Interface Requirements j i
Interface requirements for the Offsite Power System, including the switchyard, are provided in Section !
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System 80+ Deskn ControlDocument 4.2 Ultimate Heat Sink Interface Requirements Interface requirements for the Ultimate Heat Sink are provided in Section 2.7.5.
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4.3 Station Service Water Pump Structure Interface Requimnents .
t Interface requirements for the Station Service Water Pump Structure are provided in Section 2.7.5. .
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., 4.4 Station Service Water Pump Structure Ventilation System Interface Requirements [
Interface requirements for the Station Service Water Pump Structure Ventilation System are provided in Section 2.7.5.
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5.0 - Site Parameters This section provides a definition of the bounding site parameters used as the basis for the Certified Design, and to be used in evaluating the acceptability of a specific site.
. i Piping and components of the Certified Design may be designed for site-specific seismic requirements ;
which correspond to a SSE maximum ground acceleration of not less than 0.30g and site soil conditions l and properties.
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Ilorizontal Free Field Surface Spectra at Finished Plant Grade Level Figure 5.0-3 s,:erC4ied Design MaterW Page 5.0-4
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Syotem 80+ Design ControlDocument Table 5.0-1 Site Parameters hfaximum Ground Water Level 2 feet below finished plant grade level Maximum Flood (or Tsunami) Level 1 foot below finished plant grade level Precipitation (for Roor Design)
Probable Maximum Precipitation (PMP) 19.4 inches per hour with a ratio of 0.32 for 5 Estimate (Maximum Average Value Over linute to I hour PMP estimate. (6.2 inches per 5 One Square Mile Area for one hour) minutes)
Maximum Snow Load 50 pounds per square foot Design Ambient Temperatures 0% Exceedance Values (Historical Limit Excluding Peaks < 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />)
Maximum 115*F dry bulb 80*F coincident wet bulb temperature 81'F wet bulb (non-coincident) temperature Minimum -40
- F Extreme Wind Basic Wind Speed 110 miles per hour (50 year recurrence) 122 miles per hour (100 year recurrence) 1 Tornado Maximum Tornado Wind Speed 330 miles per hour Maximum Pressure Differential 2.4 pounds per sqwre inch Soil Properties Minimum Static Bearing Capacity 12,000 pounds per square foot at foundation level of Nuclear Island Structure l Best Estimate Minimum Shear Wave 700 feet per second Velocity Liquefaction The soils under safety-related structures and buried piping are stable against liquefaction at the site-specific Safe Shutdown Earthquake (SSE) level.
O Certined Des &n Atatorial Page 5.0-6
Sy: tem 80+ oesign contrat Document
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. SSE Response Spectra Rock Sites See Figures 5.0-1 and 5.0-2.
Soil Sites See Figures 5.0-3 and 5.04.
Dilution Factors Dilution Factor Time Period (sec/m3 )
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. System 80 + Design ControlDocument IG Introduction
. V; Certified Design Material 1.0 Introduction
~ 2.0 System and Structure ITAAC 3.0 ' . Non-System ITAAC 4.0 Interface Requirements 5.0 Site Parameters Approved Design Material - Design & Analysis 1.0 General Plant Description
, 2.0 Site Characteristics 3.0 Design of Systems, Structures & Components 4.0 Reactor I s 5.0 RCS and Connected Systems 6.0 Engineered Safety Features 7.0 . Instrumentation and Control 8.0 Electric Power 9.0 . Auxiliary Systems 10.0 Steam and Power Conversion 11.0 Radioactive Waste Management
[ 12.0 Radiation Protection
( 13.0 Conduct of Operations 14.0 Initial Test Program ;
15.0 Accident Analyses 16.0 Technical Specifications 17.0 Quality Assurance 18.0 Human Factors 19.0 Probabilistic Risk Assessment
, 20.0 Unresolved and Generic Safety Issues Approved Design Material - Emergency Operations Guidelines 1.0 Introduction -
2.0 Standard Post-Trip Actions 3.0 Diagnostic Actions 4.0 . Reactor Trip Recovery 5.0 Loss of Coolant Accident Recovery 6.0 - Steam Generator Tube Rupture Recovery ,
7.0 - Excess Steam Demand Event Recovery 8.0 Loss of All Feedwater Recovery 9.0 Loss of Offsite Power Recovery 10.0 Station Blackout Recovery 11.0 Functional Recovery Guideline ,
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Pages Date Pages- Date ' 6
-1,il 1/97 1.5-1 Original ;
- - lii . Original ;
.iv 2/95 1.6-1 through 1.6-5 Original l v Original ;
- 1.7-1 through 1.7-17 Original l 4
- 1.1-1 ' 2/95 ;
1.1-2 . Original 1.8-l' 2/95:
1.8-2 Original
- 1,2-1 ' Original 1.8-3 2/95 .l "1.2-2 2/95 1.8-4 through 1.8-32 Original. i 1.2-3 11/% 1.8-33 2/95 l- '1.2-4 Original 1.8-34 Original 2/95L 1.8-35, 1.8-36 11/96 ;
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- 1.2-6 11/% 1.8-37 Original
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Original i 1.2.12 11/% 1.8-43 through 1.8-45 1.2-13, 1.2-14 : Original 1.8-46, 1.8-47 2/95 1.2-15 through 1.2-17 2/95 1.8-48 Original 1.2-18 through 1.2-26 ' Original 1.8-49 through 1,3-51 2/95 t 1.2 27- 2/95 1.2-28, 1.2-29 Original 1.9-1, 1.9 2 Original ;
1.2-30 2/95
- 1.2-31 through 1.2103 Original 1.10-1 Original 1.10-2 1/97 ;
1.3-1 Original 1.10-3 2/95 i
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Chapter 1 Contents :
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( Page s .
1.0 Introduction and General Description of Plant . . . . . . . . . . . . . . . . . . . 1.1-1 !
1.1 ' Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-1 .
1.1.1 System 80 + Standard Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-1 l
1.1.2 ' Power Levels . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-1 .
1.1.3 Severe Accident Policy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-2 General Plant Desedption . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-1
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1.2 !
1.2.1 Principal Site Characteristics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-1 1.2.2 System 80+ Standard Design - Scope and Description . . . . . . . . . . . . . . . . . 1.2-3 t 1.2.3 Nuclear Steam Supply System . . . .' . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-3 ;
1.2.4 Engineered Safety Features . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-5 1.2.5 Instrumentation and Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-8
, 1.2.6 . Nuclear Plant Control Center . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1,2-10
. 1.2.7 Electrical System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' 2-14.. .
. '1.2.8 Power Conversion System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -1.2-14
, 1.2.9 Heating, Ventilating and Air Conditioning Systems . . . . . . . . . . . ...... 1.2-15 1.2.10 Fuel Handling and Storage ................................1.2-16 1.2.11 Auxiliary Systems . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . ......... 1.2-16 1.2.12 Radioactive Waste Management Systems . . . . . . . . . . . . . . . . . . . . . . . . 1.2-22 1.2.13 Physical Plant Security and Protection From Sabotage ................ 1.2-23 1.2.14 Cooling Water Systems ..............e.................... 1.2-23 (O/
1.2.15 Heat Sinks ............................. ..... ....... 1.2-24 1.2.16 Miscellaneous Structures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-24 1.3 Comparison Tables ..................................... 1.3-1 1.3.1 Comparison with Similar Facility Designs . . . . . . . . . . . . . . . . . . . ..... 1.3-1 1.3.2 Comparison of Final and Preliminary Information . . . . . . . . . . . . . . . . . . . . 1.3-1 1.4 Identification of Agents and Contractors ........................ 1,4-1 Applicant's Qualifications and Experience . . . . ..... .. ... .......
1.4.1 1.4-1 1.4.2 Architect Engineer's Qualifications and Experience . . . . . . . . . . . . . . . . . . . 1.4-1 1.4.3 Combustion Engineering's Qualifications and Experience ............... 1.4-1 1.5 Requirements for Further Technical Infonnation . . . . . . . . . ......... 1.5-1 i
1.6 Referenced Matedal Previously Submitted . . . . . . .. . . . . . . . . . . . . . . . . 1.6-1 -
1.7 Drawings and Diagrams . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.7-1 1.7.1 Electrical, Instmmentation, and Control Drawings . . . . . . . . . .......... 1.7-1 1.7.2 ' Piping and Instrumentation Diagrams . . . . . . . . . . . . . . . . . . ........ 1.7-1
' 1.8 Regulatory Compliance, Industry Codes and Standards ............. 1.8-1 ;
1.9 ' System 80+ Standard Design Interfaces . . . . . . . . . . . . . . . . . . . . . . . . . 1.9-1 1.10 System 80 + COL Information . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.10-1 4,,, r % ar.,w. - p., ,
System 80+ Design ControlDocument Chapter 1 Tables Page 1.2-1 System 80+ Improvements Based on Operating Experience .... . . . 1.2-29 1.3-1 Comparison of Reactor Characteristics . .... ....... ........... 1.3-2 1.3-2 Docket Listings for C-E Recent Reactor Designs . . . . . . . . . . . . . . . . . . . . 1.3-12 1.4-1 C-E NSSS Pressurized Water Reactor Plants . . . . . . . . . . . . . . . . . ..... 1.4-4 1.6-1 Referenced Material Previously Submitted . . . . . . . . . . . . . . . . . . . ..... 1.6-1 l 1.7-1 Safety-Related Electrical, Instrumentation and Control Drawing .. .. .... 1.7-1 1.7-2 Valve List Identifiers . . . . . . . . . . . . . . . ................. . .. 1.7-6 1.7-3 System 80+ Flow Diagram Matrix . . ............ .... ... .... 1.7-7 1.8-1 NRC Regulatory Guide Applicability Analysis to System 80+ ... ....... 1.8-2 1.8-2 Generic Letters Applicability Analysis to System 80+ ............... 1.8-12 1.8-3 NRC Bulletins Applicability Analysis to System 80+ . . . . . . .. .. ... 1.8-27 1.8-4 Deviations from the U.S. NRC Standard Review Plan ...... .. .. ... 1.8-30 1.8-5 Standard Review Plan Compliance Comments ......... .... ..... 1.8-35 1.8-6 System 80+ Industrial Codes and Standards . . . . . . . . . ........ . 1.8-38 1.8-7 ASME Section III Code Cases Applicable to System 80+ . . ...... .. 1.8-40 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues ....... ... . 1.8-41 1.8-9 Cross Reference for the TMI Rule (10 CFR 50.34f) . . . . . . . .......... 1.8 47 1.8-10 Cross-Reference for New NRC Policy Issues (SECY-93-087) . .... .... 1.8-50 1.9-1 Index of System, Structure or Component Interface Requirements for System 80+ 1.9-2 1.10-1 COL License Information . . . . .... ... ... . . ...... . 1.10-1 O
O AS4weved Design Material-Introdsction (2j95) page pv
System 80+ Design ControlDocument Chapter 1 Figures
(% Page t]
1.2-1 Site Arrangement; Single Unit ............. ................. 1.2-33 1.2-2 Nuclear Island General Arrangement; Section B-B . . . . . . . . . . . ...... 1.2-35 1.2-3 Nuclear Island General Arrangement; Section A-A . . . . . . . . . . . . . . . . . 1.2 -37 l 1.2-4 Nuclear Island General Arrangement; Plan At El. 50+0 . . . . . . . . . . . . . . . 1.2-39 l 1.2-5A Nuclear Island General Arrangement; Plan At El. 70+0 . . . . . . . . . . . . . . . 1.2-41 1.2-5B Nuclear Island General Arrangement; Plan At El. 81+0 . . . . . . . . . . . . .. 1.2-43 1.2-6 Nuclear Island General Arrangement; Plan At El. 91+9 . . . . . . . . . . . . . . . 1.2-45 1.2-7 Nuclear Island General Arrangement; Plan At El.115+6 . . . . . . . . . . . . . . 1.2-47 1.2-8 Nuclear Island General Arrangement; Plan At El.130+6 . . . .... ..... 1.2-49 !
1.2-9 Nuclear Island General Arrangement; Plan At El.146+0 . . . .......... 1.2-51 1.2-10A Nuclear Island General Arrangement; Plan At El.170+0 . . . .......... 1.2-53 1 1.2-10B Nuclear Island General Arrangement; Plan At El.191+0 . . . . . . . . . . . . . . 1.2-55 j 1.2-11 Nuclear Island General Arrangement; Miscellaneous Sections . . . ........ 1.2-57 1.2-12 Nuclear Island General Arrangement; Plan of Dish . . . . . . . . . . . ...... 1.2-59 1.2-13 Turbine Building General Arrangement Plan; Ground Floor . . . . . . . . . ... 1.2-61 1.2-14 Turbine Building General Arrangement Plan; Mezzanine Floor .......... 1.2-63 1.2-15 Turbine Building General Arrangement Plan; Operating Floor .. ........ 1.2-65 1.2-16 Turbine Building General Arrangement; Section 1-1. . . . . . . . . . . . . . . . . . 1.2-67 1.2-17 Turbine Building General Arrangement; Section 2-2 . . . . . . . . . . . . . . . .. 1.2-69 1.2-18 Turbine Building General Arrangement; Section 3-3 . . . . . . . . . ....... 1.2-71 1.2-19 Turbine Building General Arrangement; Section 4-4 . . . . . .......... . 1.2-73 q 1.2-20 General Arrangement; Station Service Building . . . . . . . . . . . . . . . ..... 1.2-75 h 1.2-21 1.2-22 General Arrangement; Auxiliary Boiler Structure ............ ......
Fire Pump House and Tanks . . . . . . . . . . . . . . . . . . . . . . . . .......
1.2-85 1.2-87 1.2-23 General Arrangement; Radwaste Building . . . . . . . . . . . . . . . . . . . . . . . . 1.2-89 1.2-24 General Arrangement; Diesel Fuel Storage Structure . . . . . . . . ... .. . 1.2-101 1.2-25 General Arrangement; CCW Heat Exchanger Structure . . . . . . .. .... 1.2-103
, 1.7-1 Piping and instrumentation Diagram Symbols . . . . .. .... ........ 1.7-11 1.7-2 Flow Diagram Symbols and Legends . . . . . . . . . . . . . . . . . . ........ 1.7-13 1.7-3 Flow Diagram Index . . . . . . . . . . . ................. ....... 1.7-17 (D
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I Sv'inorr 80+ Donlarr ControlDocument 1
1 1.0 Introduction and General Description of Plant j l 1.1 Introduction b
! This section of the Design Control Document describes the System 80+" W Standard Plant Design. '
Information is provided, to the extent applicable to the System 80+ Standard Design, for the final safety analysis report required under 10 CFR 50.34(b). Other information relevant to the System 80+ Standard Design, such as Three Mile Island requirements, technical resolution of Unresolved Safety Issues and medium .and high priority Generic Safety Issues, interface requirements and site parameters, and l '
important design features identified in risk assessments, is provided as required for design certification under 10 CFR 52.47(a). This descriptive information for the System 80+ Standard Plant Design has j been evaluated and accepted by the Nuclear Regulatory Commission (NRC) as documented in NUREG-1462, " Final Safety Evaluation Report Related to the Certification of the System 80+ Design." For purposes of the Design Control Document, this material on the description and analysis of the System
- 80+ Standard Design is termed " Approved Design Material." ;
4 The System 80+ Standard Design is an evolutionary development of the proven System 80 design ;
I 4 constructed and operated at the Palo Verde Nuclear Generating Station, and currently (1994) under construction at the Yonggwang and Ulchin sites in South Korea. System 80+ incorporates a variety of engineering and operational improvements' designed to provide additional reliability and safety margins when compared to the System 80 design. Further, design features to address the NRC's Severe Accident and Safety Goal Policy Statements are incorporated into the System 80+ Standard Design. j
{
A summary of the System 80+ Standard Design is presented in Section 1.2. Detailed information on structures, systems, and components that comprise the System 80+ standard design is provided in the l following chapters and sections of this Approved Design Material.
1.1.1 System 80+ Standard Design l
The scope of the System 80+ Standard Design covers an essentially complete nuclear power plant and ;
- includes all structures, systems, and components that can significantly affect safe operation. All major l j structures within the scope of the certified design are identified with a " cross-hatch" marking on the site j arrangement layout (Figure 1.2-1). Site-specific structures are shown on that arrangement layout with j 4
" slash" markings. Structures, systems and components not in or partially within the scope of the System - ;
80+ design are listed in Section 1.9.
1.1.2 Power Levels i
The System 80+ Standud Design, described herein, includes a reactor core designed to operate at a l maximum core power level of 3914 MWt. While the System 80+ design is independent of power level, ,
this core power level was se!ected for the analysis described herein to provide limiting design and safety analysis parameters. At this core power level, the total thermal output is 3931 MWr.
i i System 80+ is a trademark of Combastion Engineering, Inc.
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h L Specifically, for the System 80+ Standa d Design, the Electric Power Research Institute's Advanced Light Water Reactor Requirements Document has been used as a guide for utility requirements regarding plant ,
design. ,
i Annevent Dea 6n aseserier breenhecaen (2/96) Pege 1.1 1
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Srtem 80+ Design controlDocument 1.1.3 Severe Accident Policy The requirements to be met by future plants are:
- Demonstration of compliance with the procedural requirements and criteria of the current Commission regulations, including us Three Mile Island requirements for new plants as reflected in 10 CFR 50.34(f).
- Demonstration of technical resolution of all applicable Unresolved Safety Issues and the medium-and high-priority Generic Safety Issues, including a special focus on assuring the reliability of decay heat removal systems and the reliability of both AC and DC electrical supply systems.
- Completion of a Probabilistic Risk Assessment (PRA) and consideration of the severe accident vulnerabilities the PRA exposes along with the insights that may add to the assurance of no undue risk to public health and safety.
- Completion of a staff review of the design with a conclusion of safety acceptability using an approach that stresses deterministic engineering analysis and judgment complemented by PRA.
In addition, the Severe Accident Policy states:
"The Commission also recognizes the importance of such potential contributors to severe accident risk as human performance and sabotage. The issues of both insider and outsider sabotage threats will be carefully analyzed and, to the extent practicable, will be emphasized as special considerations in the design and in the operating procedures developed for new plants."
Severe Accidents and unresolved generic issues are addressed in this document. The resolution of these issues is summarized in Chapters 19 and 20. Degraded core analyses are included in the PRA. Results of the Sabotage Protection Program are presented in Appendix 13A.
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' Svstem 80+' Desinn contror Document i
i 41.2 ' General Plant Description
'I.2.1 Pdacipal Site Characteristics i
1.2.1.1 ' Site Location j i
The . System 80+ Standard Design is designed for use at multiple sites as described in Chapter 2. The ~
site-specific SAR will identify the specific site for that unit. !
- 1.2.1.2 - Plant Ac N:
4 The System 80+ Standard Design is designed for use at multiple sites.' The site-specific SAR will -
4 L identify the specific surroundings for that unit.
- 1.2.1.2.1 Meteorology - ,
Section 2.3 lists, for plant radiological evaluation pumoses, the short-term (accident) and long-term ;
- (routine) diffusion estimates (x/Q). Other meteorological design bases are listed in Table 2.0-1. Section
' 2.3 of the site-specific SAR will include data to show compliance with the design bases.
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-1.2.1.2.2 Hydrology t
Hydrological design bases are listed in Table 2.0-1. Section 2.4 of the site-specific SAR will include data L to show compliance with the design bases.
1.2.1.2.3 Geology and Seismology :
The design of safety-related stmetures, systems, and components of the System 80+ Standard Design is consistent with the seismic envelope given in Section 2.5. Section 2.5 of the site-specific SAR will include data to show compliance with the seismic envelope.
- 1.2.1.3 Plant Independence The System 80+ Standard Design can be used at either single-plant or multiple-plant sites. At multiple-
' plant sites, the mdependence of all safety-related systems and their support systems will be maintained between (or among) the individual plants.
1.2.1.4 Site Building Arrangement A typical layout of the System 80+ Standard Design buildings is shown in Figure 1.2-1. Sufficient open
~ f space is shown'so that a facility for dry storage of spent fuel casks can be added on a site-specific basis.
1.2.1.4.1- Site-Specific Structures Desedption and Interface Re7' M Some structures which house non-safety related and certain safety-related systems and components are
. rupplied by the licensee and are not included in the System 80+ design certification. To ensure that the des:gn of such structures is compatible with the System 80+ Standard Design, certain interface a regicirements must be met by the applicant (owner / operator). The following sections present the interface requirements and conceptual descriptions for the Administration Building, Personnel Access Portal and Anomear onow annow.no,awson rare 1.2.r 4
System 80+ Design Control Document l Warehouse. The word "shall" is used to identify interface requirements in the descriptive text. The remainder of the description is conceptual and it is not intended to be binding on the COL holder.
Interface requirements for structures which are related to a specific mechanical or electrical system are l covered in the appropriate chapter, e.g. the Station Service Water Pump Structure is covered in Section 9.2.1, Station Service Water System. Section 1.9 contains an index of all System 80+ interface l requirements. j 4
1.2.1.4.1.1 Administration Building a An Administration Building shall be provided by the licensee. ((This building provides office and support space for station administration and management personnel who have no need to be located within the Protected Area.
A typical Administration Building is designed as non-safety related, non-seismic structure with the following conceptual features. The building is a steel-framed structure with a steel deck roof covered by non<ombustible roofing. Walls are insulated metal siding or masonry. Roof drainage and clean floor drainage are discharged to the storm and waste water system. The building is located immediately outside the Protected Area fence at the entrance to the plant, near the Personnel Access Portal building. Air conditioning and heating is provided to meet normal office environment conditions.))
1.2.1.4.1.2 Personnel Access Portal The Personnel Access Portal (PAP) shall be provided by the licensee, and shall be designed to provide the following functions:
- Serve as access point through the Protected Area Boundary.
- Provide facilities to search, badge, and permit access to the Protected Area.
- Provide the Secondary Alarm Stations.
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- Provide the required bullet-resistant features to support security force functions.
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((A typical PAP building is a masonry building with non-combustible roofing on a metal deck. The PAP i l
building is located along the Protected Area fence at the entrance to the plant, near the Administration Building. A PAP building vendlation system is provided to maintain the building within design temperature limits.))'
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1.2.1.4.1.3 Warehouse ;
1 The licensee shall provide a warehouse to accommodate the foilowing:
- Material access to the Protected Area incorporating the Vehicle (and Cargo) Access Portal (VAP).
- Loading docks, search areas, QA inspection and QA Hold Areas, along with systems and fixtures j to provide bulk storage of QA and non QA parts and supplies.
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I Conceptual Design information; see DCD Introduction Section 3.4 Attvered Design Materia! kurodiction (2/95) Page 1.2 2
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e A material issue area to control dispensing of materials for maintenance.
((A typical warehouse is founded at grade on a reinforced concrete foundation. The warehouse is typically a metal enclosed building with a structural steel frame and non-combustible roofing. The i warehouse ventilation system is designed to meet the appropriate requirements of ANSI /ASME NQA-2 -
Part 2.2.))!
1.2.2 ' Systesa 80+ hadant Design - Scope and Description The scope of the System 80+ Standard Design (see Section 1.1.1) covers an essentially complete nuclear ;
power plant and includes all structures, systems, and components that can significantly affect safe
! operation. The primary design characteristics are summarized in the subsections below and improvements '
based. on operating experience are summarized in Table 1.2-1. The seismic category, safety classification, and quality assurance requirement of structures, systems, and components are listed in ;
-. Table 3.2-1. :
1.2.3 Nuclear Steam Supply Systesa ,
The' Nuclear Steam Supply System (NSSS) generates approximately 3931 MWt, producing saturated steam. l j
4 The NSSS contains two primary coolant loops, each of which has two reactor coolant pumps, a steam 1 l generator, a 42-inch ID hot leg pipe and two 30-inch ID cold leg pipes. In addition, the safety injection I lines are connected directly to the Reactor Vessel. An electrically heated pressurizer is connected to one l I
\ ef the loops of the NSSS. The pressurizer has an increased volume (relative to previous design) to enhance transient response. Pressurized water is circulated by means of electric-motor-driven, single-stage, centrifugal reactor coolant pumps. Reactor coolant flows downward between the reactor vessel shell and the core support barrel, upward through the reactor core, through the hot leg piping, through
! the tube side of the vertical U-tube steam generators, and back to the reactor coolant pumps. The saturated steam produced in the steam generators is passed to the turbine.
1.2.3,1 Reactor Core The reactor core is fueled with uranium dioxide pellets enclosed in zircaloy tubes with welded end caps. ,
The tubes are fabricated into assemblies in which end fittings limit axial motion and grids limit lateral '
motion of the tubes. The control element assemblies (CEAs) consist of NiCrFe alloy clad boron carbide or silver indium cadmium absorber rods and solid NiCrFe alloy reduced strength absorber rods, which are guided by tubes located within the fuel assembly. The core consists of 241 fuel assemblies which will be typically loaded with three different U-235 enrichments. The NSSS full thermal output is 3931 MWt :
with a core thermal output of 3914 MWt. l Design criteria are established to ensure the following:
i e The minimum departure from nucleate boiling ratio during normal operation and anticipated operational occurrences will provide at least a 95% probability with 95% confidence that departure from nucleate boiling does not o'utr. !
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U 1 I Conceptual Design infonnation; see DCD Introduction Section 3.4. j 4pmeer Deadye aseeerder. amedweise Ir r/ps> pape r.2-3 i i
System 80+ Design ControlDocument
- The manmum fuel centerline temperature evaluated at the design overpower condition is below that value which could lead to centerline fuel melting. The melting point of the UO2 is not reached during normal operation and anticipated operational occurrences.
- Fuel rod clad is designed to maintain cladding integrity throughout fuel life.
- The reactor system is designed so that any xenon transients will be adequately damped.
- The Reactor Coolant System is designed and constructed to maintain its integrity throughout the expected plant life.
- The reactor and Plant Protection System are designed such that power excursions that could result from any credible reactivity addition incident do not cause damage either by deformation or rupture of the pressure vessel, or impair operation of the engineered safety features.
- The reactor is designed such that the combined response of the fuel temperature coefficient, the moderator temperature coefficient, the moderator void coefficient, and the moderator pressure coefficient to an increase in reactor thermal power is a decrease in reactivity. In addition, reactor power transients remain bounded and damped in response to any expected changes in any operating variable.
The reactor core is further discussed in Chapter 4.
1.2.3.2 Reactor Internals The internal structures include the core support barrel, the lower support structure and in-core instmmentation nozzle assembly, the core shroud, and the upper guide structure assembly. The core support barrel is a right circular cylinder supported by a ring flange from a ledge on the reactor vessel.
It carries the entire weight of the core. The lower support structure transmits the weight of the core to the core support barrel by means of a beam structure. The core shroud surrounds the core and minimizes the amount of bypass flow. The upper guide rtructure provides a flow shroud for the CEAs, and limits upward motion of the fuel assemblies during pressure transients. Lateral snubbers are provided at the lower end of the core support barrel assembly.
The principal design bases for the reactor internals are to provide the vertical supports and horizontal restraints during all normal operating, upset, and faulted conditions.
a The core is supported and restrained during normal operation and postulated accidents to ensure that coolant can be supplied to the coolant channels for heat removal. l Reactor internals are further discussed in Sections 3.9 and 4.5.
1.2.3.3 Reactor Coolant System (RCS)
The RCS is arranged as two closed loops connected in parallel to the reactor vessel. Each loop consists of one 42-inch ID outlet hot leg, one steam generator, two 30-inch ID cold leg pipes, and two pumps.
An electrically heated pressurizer is connected to one of the loops of the RCS. )
The RCS operates at a nominal pressure of 2250 psia. The reactor coolant enters the reactor vessel, then flows downward between the reactor vessel shell and the core barrel, up through the core, leaves the Approved Design Material . kstrocuction Page 1.2-4 l
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^ Sv: tem 80+ Desinn controlDocument 7g reactor vessel, and ilows through the tube side of the two venical U-tube steam generators (with an f integral economizer) where heat is transferred to the secondary system. Reactor coolant pumps return :
i the reactor coolant to the reactor vessel. :
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Two steam generators, using heat generated by the reactor core and carried by the primary coolant to l 3
each steam generator, produce steam for driving the plant turbine-generator. Each steam generator is a !
vertical U tube heat exchanger with an integral economizer which operates with the reactor coolant on :
- the tube side and secondary coolant on the shell side. Each unit is designed to transfer heat from the ;
t Reactor Coolant System to the secondary system to produce saturated steam when provided with the ;
- proper input feedwater. Moisture separators and steam dryers on the shell side of the steam generator l limit the moisture content of the steam during normal operation at full power. An integral flow restrictor !
has been designed into each steam generator steam nozzle to restrict flow in the event of a steam line i
break.
E The System 80 + steam generator incorporates several design enhancements including better steam dryers, increased overall heat transfer area and slightly reduced full power steam pressure.
l' The System 80 + steam generator also has a larger secondary feedwater inventory which extends the " boil dry" time, thus enhancing the NSSS's capability to tolerate upset conditions and improving operational
'l flexibility. Finally, the System 80+ steam generator design has a greater tube plugging allowance, thus; permitting the NSSS to maintain rated output with a significant number of tubes plugged. ;
The RCS is further discussed in Chapter 5. >
l 1.2.4 N-ed Safety Features ;
o Engineered safety features function in the highly unlikely event of an accidental release of radioactive
[ fission products from the reactor coolant system, particularly as the result of loss-of-coolant-accidents.
These safeguards function to localize, control, mitigate, or terminate such accidents to hold exposure 4 levels below the limits of 10 CFR 100.
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1.2.4.1 Coatninment Structure 1
4 General arrangements for the reactor building, including the containment vessel, are shown in Figures t
- 1.2-2 through 1.2-12. The containment vessel is a 200-foot diameter spherical shell with a wall thickness i
j of approximately one and three-quarter inches. This containment shell is supported by a spherical support pedestal which is part of the reactor building. The reactor building is a reinforced concrete cylindrical :
I 4 building with a hemispherical dome which totally encloses the containment, internal struwire and subsphere. The exterior walls of the reactor building, including the dome, are referred to as the shield j building. Space below the containment and inside the shield building is referred to as the subsphere and is occupied by Engineered Safety Features equipment, e g., emeigency core cooling system equipment,
- containment spray system _ equipment, shutdown cooling system equipment, anri emergency feedwater equipment.
A more detailed physical description of the containment and the design criteria relating to the construction techniques, static loads, and seismic loads are provided or referenced in Section 3.8.
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- The containment design basis is to provide an essentially leak-tight barrier against the release of radioactive materials subsequent to postulated accidents. In order to meet this requirement, a maxunum 4prmatDenQrt aseewd heaehedwr W96) hoe 1.2-5 r________ ._.__.u.. -_. - , _ . . . . . . _ . . . _ , _ _ _- . - , , . . . . _. _ , , _ _ . _ . . ._ _
System 80+ Design ControlDocument containment leakage rate is defined in conjunction with performance requirements placed on the Engineered Safety Features (ESP) systems.
The capability of the containment structure to maintain design leaktight mtegrity and to provide a predictable environment for operation of ESF systems is ensured by a comprehensive design, analysis, and testing program that includes consideration of:
- The peak containment pressure and temperature associated with the most severe postulated accident coincident with the Safe Shutdown Earthquake.
- The maximum external pressure loading condition to which the containment may be subjected as a result ofinadvertent containment systems operations that potentially reduce containment internal l pressure below outside atmospheric pressure, coincident with the Safe Shutdown Earthquake.
1.2.4.2 Safety Irdection System The Safety Injection System (SIS) is designed to satisfy NRC regulatory requirements. These requirements are specified as the Licensing Design Basis for the System 80+ design. In addition, the EPRI ALWR Requirements Document has been used to define a Safety Margin Design Basis for the SIS design. The Safety Margin Design Basis contains requirements which go beyond the minimum required by the Code of Federal Regulations, thereby providing additional safety assurance in the SIS design.
In the highly unlikely event of a loss-of-coolant-accident, the SIS injects borated water into the Reactor Coolant System. The System 80+ SIS incorporates a four-train safety injection configuration and an In-Containment Ri. fueling Water Storage Tank (IRWST).
The System 80+ SIS utilizes four safety injection pumps to inject borated water directly into the Reactor .
Vessel. In addition, four safety injection tanks are provided. The SI pumps are normally aligned to the j IRWST and a realignment for recirculation following a LOCA is not required. This system provides !
cooling to limit core damage and fission product release and ensures adequate shutdown margin.
The SIS also provides continuous long-temi, post-accident cooling of the core by recirculation of borated water from the IRWST. Water, drawn from the IRWST by the SI pumps and the containment spray (CS) peups, is injected into the reactor vessel and the containment. The SI water then enters the containment throgh the primary pip break. This water and the CS water return through floor drains and the holdup j vr.lume irl te the IRWST. During this process, heat is removed from the IRWST water by the containment spray kat exchanger. The SIS and the IRUT are discussed further in Sections 6.3 and 6.8, respect;yely-The SIS is capable of providing an alternate means of decay heat removal for those events beyond the licensing design basis in which the steam generators are not available. Decay heat removal, via feeding and bleeding of the RCS, would be accomplished using the SIS to feed, the Safety Depressurization l System (SDS) to bleed, and the Shutdown Cooling System (SCS) for cooling of the IRWST water.
1.2.4.3 Emergency Feedwater System 1
The Emergency Feedwater System (EFWS) for the System 80+ Standard Design is a dedicated safety l system that is designed to perform the following functions:
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System 80+ Design ControlDocument
/ e Supply feedwater to the steam generators for the removal of heat from the RCS in the event the main feedwater system is unavailabic following a transient or accident.
e Supply feedwater to the steam generators for the removal of heat from the RCS in the event of ;
a total loss of AC power (station blackout).
The EFWS consists of two storage tanks, four pumps, and associated piping and valves. Two pumps are motor-driven and two are steam-driven. The EFWS is designed to be automatically or manually initiated.
The EFWS is discussed further in Section 10.4.9.
1.2.4.4 Safety Depressurization System The Safety Depressurization System (SD S) is a dedicated safety system designed to perform the following functions:
- Provide a safety grade means to depressurize the RCS in the event that pressurizer spray is ,
unavailable during plant cooldown to cold shutdown.
e Provide a capability to rapidly depressurize the RCS to initiate the feed and bleed method of plant ,
cooldown subsequent to a total loss of feedwater.
The system includes the valves and piping which establishes a flow path from the pressurizer steam space p to the in-Containment Refueling Water Storage Tank (IRWST). It is manually actuated and controlled. ]
V l The SDS is discussed further in Section 6.7.
l 1.2.4.5 Containment Spray System The Containment Spray System (CSS) for System 80+ is designed to maintain containment pressure and ]
temperature within design limits in the unlikely event of design basis mass-energy releases to the containment atmosphere. :
The CSS is a fully redundant two-train system. Two containment spray pumps supply water through two heat exchangers to the upper region of the containment. Spray headers are used to provide a relatively uniform distribution of spray ove 'he cross sectional area of the containment. The In-Containment Refueling Water Storage Tank (IRWST) is used as the water source for the system. The Containment Spray Pumps can be manually aligned and used as residual heat removal pumps during Shutdown Cooling l System (SCS) operation. Likewise, the SCS pumps can be manually aligned to perform the containment spray function.
The Containment Spray System also provides a containment air cleanup function to reduce the l ,
concentration of fission products in the containment atmosphere after an accident. No spray additives are required.
The CS pump; and CS heat exchangers can also be used as a backup to the SCS ptsnps and heat exchangers to provide cooling of the IRWST water during post-accident feed and bleed operations when l
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V the steam generators are not available to cool the RCS.
The CSS is discussed further in Sections 6.2.2 and 6.5.
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System 80+ oesign control Document 1.2.5 Instrumentation and Control The System 80+ Standard Design instrumentation and control systems are summarized below. These systems are integrated with all other systems in the Nuplex 80+= m Advanced Control Complex (ACC), described in Section 1.2.6.
Automatic protection systems, control systems, and interlocks are provided to assure safe operation of the plant. Sufficient instrumentation and controls are supplied to provide manual operation as a normal backup control mode on all automatic systems.
1.2.5.1 Protection, Control, and Instrumentation Systems The Plant Protection System (PPS) initiates a reactor trip if the reactor approaches prescribed safety limits, or provides an actuation signal to the Engineered Safety Feature systems when a fluid system or containment parameter approaches a prescribed limit.
Sufficient redundancy is installed to permit periodic testing of the PPS so that removal from service of any one protection system component or portion of the system will not preclude reactor trip or other protective action when required. Additionally, no single failure can preclude the PPS providing a reactor trip or other protective action when required.
The protection system and associated instrumentation is separated from the control systems and their associated instrumentation such that failure, or removal from service, of any control system, component, or instrument channel will act inhibit the functioning of the protection system (see Chapter 7 for details).
1.2.5.1.1 Reactor Protective System The controllable reactor parameters are normally maintained within acceptable operating limits by the inherent characteristics of the reactor, the Reactor Regulating System (RRS), soluble boron concentration, and the plant operating procedures.
Four independent channels of the Reactor Protective System (RPS) normally monitor each of the selected plant parameters. The RPS logic is designed to initiate protective action whenever the signal of any two channels of a given parameter reach the preset limit. Should this occur, the power supplied to the i Control Element Drive Mechanisms (CEDMs) is interrupted, releasing the Control Element Assemblies I (CEAs) which drop into the core to shutdown the reactor. The two-out-of-four logic can be converted to two-out-of-three logic to allow one channel to be bypassed for testing, maintenance or operation. The protection system is maintained independent of and separate from the manual and automatic control systems by the use of optical isolation and signal validation logic described in Chapter 7.
1.2.5.1.2 Alternate Protection System The Alternate Protection System (APS) augments plant protection by generating an Alternate Reactor Trip Signal (ARTS) and Alternate Feedwater Actuation Signal (AFAS) that are separate and diverse from the l Plant Protection System. This system is provided to address Anticipated Transients Without Scram (ATWS) and ATWS Mitigating Systems Actuation Circuitry (AMSAC) design requirements. The added equipment provides a simple, yet diverse mechanism to significantly decrease the possibility of an ATWS. I Oll t Nuplex 80+ is a trademark of Combustion Engineering, Inc.
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The AKT4 will initiate a reactor trip when the Pressurizer pressure exceeds a predetermined value. Its Q} senscrs and circuitry including the final actuation devices are diverse from that of the RPS. The A RTS l
v.esign uses a two-out-of-two logic to open the CEDM motor generator output contactors.
The AFAS will initiate emergency feedwater when the level in either Steam Generator decreases below a predetermined value. Its sensors and circuitry are independent and diverse from that of the PPS Emergency Feedwater Actuation System and the Reactor Protective System.
1.2.5.1.3 Fngnmed Safety Features Actuation System The Engineered Safety Features Actuation System (ESFAS) operates in a manner similar to the RPS to automatically actuate the Engineered Safety Feature (ESP) syrnns. Again, it has a selective two-out-of-four actuation logic that can be converted to a selective two-out-of-three logic. The ESFAS is independent of the control systems.
1.2.5.1.4 Reactor Control Systems The reactor control systems are used for startup and shutdown of the reactor, and for adjustment of the reactor power in response to turbine load demand. Reactor control functions are performed by the Power Control System (PCS) and the Process-Component Control System (Process-CCS) as described in Section 7.7.1.1. The PCS performs CEDMCS, MDS, RPCS and RRS functions. The Process-CCS performs SBCS, FWCS and pressurizer control functions. Reactor power control is normally accomplished by automatic movement of CEAs in response to a change in reactor coolan, temperature, with manual control
/7 capable of overriding the automatic signal at any time. If the reactor coolant temperature is different V from a programmed value, the CEAs are adjusted until the difference is within the prescribed control band. Regulation of the reactor coolant temperature, in accordance with this program, maintains the ,
secondary steam pressure within operating limits and matches reactor power to load demand.
The reactor is controlled by a combination of CEA motion and dissolved boric acid in the reactor coolant.
Boric acid is used for reactivity changes associated with large but gradual changes in water temperature, xenon concentration, and fuel burnup. Addition of boric acid also provides an increased shutdown margin during the initial fuel loading and subsequent refuelings. The boric acid solution is prepared and stored at a temperature sufficient to prevent precipitation (maximum boric acid concentration in any ,
storage tank is 2.5 weight percent).
CEA movement provides changes in reactivity for shutdown or power changes. The CEAs are moved by CEDMs mounted on the reactor vessel head. The CEDMs are designed to permit rapid insertion of the CEAs into the reactor core by gravity. CEA motion can be initiated manually or automatically. In addition to full strength CEAs, the System 80+ design provides reduced strength CEAs which can be '
used for reactivity control during maneuvers, thus minimizing the need for changes in RCS boron concentration during intended maneuvers and operational transients.
The pressure in the Reactor Coolant System is controlled by regulating the temperature of the coolant in the pressurizer where steam and water are held in thermal equilibrium. Steam is formed by the pressurizer heaters or condensed by the pressurizer spray to reduce variations caused by expansion and contraction of the reactor coolant due to system temperature changes.
The Megawatt Demand Setter (MDS) is a Nuplex 80+ system that automatically controls the response b of the station's main turbine to changes in power demand relayed from the utility's grid by the automatic
. dispatch system. A Steam Bypass Control System (SBCS) is used to dump steam in case of a large
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System 80+ Design control Document mismatch between the power being produced by the reactor and the power being used by the turbine.
This allows the reactor to remain at power instead of tripping. The water level in each steam generator is maintained by a Feedwater Control System (FWCS). A Reactor Power Cutback System (RPCS) is used to drop selected CEAs into the core to reduce reactor power rapidly during a loss of a feed pump or a large loss of load. This allows the SBCS and FWCS to maintain the NSSS in a stable condition, without a reactor trip, and without lifting any safety valves during loss of load transients.
1.2.5.1.5 Nuclear Instrumentation The nuclear instrumentation includes ex-core and in-core neutron flux detectors. Eight channels of ex-core instrumentation monitor the power. Two channels are provided for startup, two channels are provided for low power monitoring, and four channels are provided for the protection channels. The startup and low power channels are used for monitoring the low neutron flux levels during plant startup.
The protection channels are used to provide inputs to the overpower, logarithmic power, Departure from Nucleate Boiling Ratio (DNBR), and Local Power Density (LPD) trips in the RPS. The four safety channels are signal validated and used to control the reactor power during power operations.
The in-core instrumentation consists of self-powered detectors, distributed throughout the core, which provide information on flux distribution within the core. Refer to Sections 7.2.1.1 and 7.7.1.1 for a description of ex-core and in-core instrumentation.
1.2.5.1.6 Process Monitoring Systems Temperature, pressure, flow and liquid level monitoring are provided as required to keep the operating personnel informed as to plant operating conditions. Protection channels indicate the various parameters used for protective action as well as providing trip and pre-trip alarms from the RPS.
The plant liquid and gaseous effluents are monitored to assure that they are maintained within applicable radioactivity limits. Additional information is provided in Chapter 11.
1.2.6 Nuclear Plant Control Center In addition to the systems described in Section 1.2.5.1, the System 80+ Standard Design includes the Nuplex 80+ Advanced Control Complex (ACC) to ensure a completely integrated design, including human factors engineering. The ACC is subdivided into functional units which complement but do not interfere with each other. These functional units include:
- The Main Control Room - where plant control is performed.
- The Computer Room - where the Data Processing System (DPS) plant logs and hard copy results are generated and computer programming is accomplished. The DPS provides an interface to the Emergency Operations Facility.
- The Onsite Technical Support Center - which contains intnfoces to the DPS to support plant emergencies.
- The Remote Shutdown Room - provides a centralized remote control location to perform safe plant shutdown and cooldown after control has been transferred from the Main Control Room.
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- Vital Instrumentation and Equipment Rooms - four redundant separated rooms which contain their l respective channels (A, B, C or D) of safety-related equipment. l !
- . Non-Essential Electrical Equipment Rooms - two separated rooms which contain non-safety j related instrumentation, controls and electrical equipment.
I The Advanced Control Complex design includes the following major interdependent systems:
o Main Control Panels l
- Remote Shutdown Panel ,
e Discrete Indication and Alarm System i
e Data Processing System e Component Control System
- Plant Protection System i
- Power Control System - ;
1.2.6.1 Main Control Panels .
The main control panels are designed to permit commami by a single individual during normal power operation. However, the main control room design acconunodates two control room operators and a supervisor for all normal modes of plant operation and up to the full operating crew during emergencies.
I Each main control panel section integrates miciaturized back lighted component control switches, meters, alarms, indicators and Video Display Units (VDUs) (e.g., CRTs, Plasma Displays) such that both safety- l
- - Discrete alarms ami indicators are provided (o allow accident and technical specification monitoring, safe shutdown and other licensing requirements for which the Data Processing System VDUs described in I
Section 1.2.6.4 cannot be credited. The discrete alarms and indicators are also designed to permit continued plant operation for unlikely instances when the Data Processing System is unavailable. ,
The panel arrangements and layouts for all controls and indicators on the main control panels are designed ' verified and validated in accordance with human factors design guidelines and requirements specified in NUREG-0700. Refer to Chapter 18, Human Factors Engineering, for further information. >
- A Control Room Supervisor's Monitoring Console, including a DPS driven VDU and sufficient desk space, is provided to support the plant monitoring and daily operational needs of the Control Room i Supervisor.
- 1.2.6.2. Resnote Shutdown Panels -
I O
The Remote Shutdown Panel (RSP) design includes a minimum of two isolated redundant channels of the i safety related instrumentation and controls necessary to achieve hot standby (mocle 3 plant conditions)if !
. the main control room must be evacuated. !
Anomed Doesn annander.areenmmten (2/nst rage 1.21r ,
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..-m _- _
System 80+ Design ControlDocument A VDU monitor is provided at the Remote Shutdown Panel to provide operational display pages. This VDU is the same as the control room VDUs and is provided for convenience.
Local controls, RSP controls, and instrumentation are provided to bring the plant to cold shutdown conditions utilizing suitable procedures.
1.2.6.3 Discrete Indication and Alarm System The alarm and display systems are designed to aid the operator in handling any challenges to critical plant availability or safety functions.
The design integrates the information displayed from alarm windows, meters and VDUs such that the same instruments used for accident monitoring are used for normal plant operations to enable operators to use instruments with which they are most familiar during accident situations.
The advanced control panels include displays and alanns which allow monitoring of parameters associated with the following critical safety functions:
e Reactivity Connel e RCS Inventory Control e RCS Pressure Control e Core Heat Removal e RCS Heat Removal e Containment Integrity e Plant Radiation Emissions The set of VDU panel displays include human engineered pictorial mimic and alarm information that provides the operator with a continuous real-time high level overview of the entire power plant's steam and electrical production process.
The alarm and display systems are designed such that no single failure will result in the loss of plant information presented to the operator. The design includes diverse means of providing the operator information necessary to keep the plant operating and for monitoring during accident conditions.
The alanns are designed to identify their priority through the use of hierarchical physical location and color coding. Alarm processing techniques (plant mode adaptation and suppression) based on validated process parameter inputs are used to increase operator comprehension and reduce nuisance alarms.
1.2.6.4 Data Processing System The Data Processing System (DPS) is a fault tolerant multiprocessor computer based system which provides plant data and status information to the operations staff. The DPS monitors the steam and electrical production processes. It provides the plant operations staff the ability to obtain detailed process data via VDU information output devices.
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in, 1 The major functions performed by the DPS include plant wide data acquisition via dedicated data links
- d to other plam systems, validation of sensed parameters, execution of application programs and performance calculations, monitoring of general plant status and plant safety status, generation of logs and reports, the determination of alarm conditions, sequence of events recording and post-trip review. Multicolor VDUs, interfaced with touch-screen devices, and high speed printers are utilized to present the plant information to the operator.
VDU display formats incorporate Human Factor Engineering design principles which permit rapid operator comprehension of the information necessary to allow the operator to monitor, control and diagnose plant conditions. All displays are organized in a multi-tier hierarchical structure. Touch-screen display access mechanisms are designed to allow for ease-of-access to any display page within the hierarchy.
The DPS is designed to reliably provide the plant operations staff with complete and timely information for the safe and efficient operation of the plant. The DPS is implemented using a modern, high speed distributed computer system configured to be fault tolerant. The DPS is designed to tolerate the loss of any major system component without loss of functionality.
The design includes automatic fail over and sufficient redundant peripherals necessary to minimize the effects of a DPS component failure on plant operations.
1.2.6.S Component Control System n The Component Control System (CCS) is designed to control discrete-state components such as pumps,
( ! valves, heaters and fans within the plant systems.
The CCS consists of the ESF-CCS and Process-CCS assemblies to provide control for the different channels of Class IE equipment, as well as non-Class IE equipment. Although they perform different plant control functions, the CCS Class IE and non-Class IE Assemblies utilize diverse software and software dependent electronic components.
1 e Component Control Logic l The Component Control Logic (CCL) is a set of microprocessor based controllers that monitor various digital inputs, such as manual on-off demands from the main control panel, interlocks, ESFAS, Diesel sequence signals, automatic control signals, and, as programmed, produces digital output signals to control the component (start-stop; on-off). The CCL generates outputs for status indication on the main control panel and plant Data Processing System.
e Engineered Safety Features Logic i l
The CCS accepts ESF Initiation signals from the Plant Protection System. The ESF Logic is used to activate the plant's Engineered Safety Features Systems components. The Engineered Safety Feature Actuation System (ESFAS) logic includes modules for ESF actuation and an ESF test controller. Diesel Load Sequencing (DLS) logic is included in the design.
e Main Control Room and Remote Shutdown Panel Interface n i U All main control room interfaces are isolated to prevent fault propagation into the CCS logic in the event of control room damage. Upon main control room evacuation, CCS local panel Approved Design Materie! kutroduction Page 1.213
System 80+ Design ControlDocument controls and/or the Remote Shutdown Panel may then be used effectively to achieve an orderly, unobstructed plant shutdown. The Remote Shutdown Panel provides control for all required hot standby components. Local CCS front panel controls are provided for all components, allowing complete cold shutdown from a central location outside the control room.
1.2.7 Electrical System An offsite power system and an onsite power system are provided to supply the unit auxiliaries with power during nom! operation and the Reactor Protection System and Engineered Safety Feature Systems with power de ibnormal and accident conditions.
The turbine-genemor unit is connected to a switchyard and thereby to the transmission system via a transmission line. The generator circuit breaker, along with the unit main transformer, allows this line to supply power to the transmission system during normal operation.
A second separate and independent transmission line is connected to a separate switchyard which provides power directly to the safety buses via divisional Reserve Auxiliary Transformers.
A description of the offsite power system is provided in Section 8.2.
The onsite power system for the unit consists of the main generator, the generator circuit breaker, the unit main transformer, the unit auxiliary transformers, Reserve Auxiliary Transformers, the diesel generators, an alternate AC source, the batteries, and the auxiliary power system. Under normal operating conditions, the main generator supplies power through an isolated phase bus and the generator circuit breaker to the unit main and unit auxiliary transformers. The unit auxiliary transformers are connected to the bus between the generator circuit breaker and the unit main transformer. During normal operation, station auxiliary power is supplied from the main generator through these unit auxiliary transformers. During startup and shutdown, the generator circuit breaker is open, and station auxiliary power is supplied from the transmission system through the unit main and unit auxiliary power transformers.
1 A description of the ansite power system including the alternate AC source is provided in Section 8.3.
A description of the ccmbustion turbine generator and its fuel storage facility is provided in Section 1.2.16.2 and Section 8.3.1.
1 1.2.8 Power Conversion System The function of the Steam and Power Conversion System is to convert the heat energy generated by the nuclear reactor into electrical energy. The heat energy produces steam in two steam generators capable of driving a turbine generator unit.
The Steam and Power Conversion System utilizes a condensing cycle with regenerative feedwater heating. l Turbine exhaust steam is condensed in a conventional surface type condenser. The condensate from the i steam is returned to the steam generators through the condensate and feedwater system. l l
A Turbine Bypass System capable of relieving 55% of fullload main steam flow is provided to dissipate i heat from the Reactor Coolant System during turbine and/or reactor trip. This system consists of eight i turbine bypass valves to limit pressure rise in the steam generators following cessation of flow to the !
turbine. Once the steam flow path to the turbine has been blocked by the closing of the turbine valves, <
decay heat is removed by directing steam to the condenser.
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.) In addition to the above, atmospheric steam dump valves are connected to the main steam lines upstream of the main steam line isolation valves to provide the capability to hold the plant at hot standby or, in the l event of loss of power to the condenser circulating water pumps, cool the plant down to the point at l which the shutdown cooling system may be utilized. These valves are not part of the Turbine Bypass l System; no credit for their use is assumed in obtaining the 55 % capacity of the Turbine Bypass System.
I Overpressure protection for the shell side of the steam generators and the main steam line piping up to the inlet of the turbine stop valve is provided by spring-loaded safety valves. Modulation of the turbine bypass valves discussed earlier would normally prevent the safety valves from opening. The steam bypass system, coupled with the Reactor Power Cutback System, would prevent apening of the safety valves following a turbine and/or reactor trip.
Each steam generator has two steam discharge lines. Each line is provided with a flow measuring device, five spring-loaded safety relief valves, a main steam isolation valve, a power operated atmospheric dump valve and a bypass line and valve around each main steam isolation valve. Each main steam line is provided with a turbine stop valve and a control valve just upstream of the high pressure turbine.
The Steam and Power Conversion System is described further in Chapter 10.
General arrangements for the turbine building are shown in Figures 1.2-13 through 1.2-19.
1.2.9 Heating, Wntilating and Air Conditioning Systems G The Heating, Ventilating and Air Conditioning (HVAC) Systems for all plant buildings are designed for l (V 3 personnel comfort and/or equipment operation with the exception of Annulus Ventilation System. In addition, the following systems have been provided with protection features described below.
- Control Room HVAC Subsystem is designed for unintemipted safe occupancy of the control room during normal operation and post-accident shutdown.
- Fuel Building Ventilation System is a once through ventilation system designed to limit the radiation release following a fuel handling accident to meet the 10 CFR 100 guidelines. It maintains the building under negative pressure and directs the air flow from less-contaminated to more-contaminated areas before exiting.
- Nuclear Annex and Radwaste Building Ventilation Systems are once through ventilation systems with fr . red exhausts. They maintain negative building pressures and direct the air flow from less-contaminated to more-contaminatei areas before exiting.
- The Annulus Ventilation System (AVS) sewes the space between the primary containment and the secondary containment. The system does tot perform any normal ventilation function.
However, it does provide additional assurance against the release of radioactivity to the environment; therefore, it is designed as an engineered safety feature and should be capable of operating and perfonning its function during startup, power operation, hot standby and hot shutdown.
- Subsphere Building Ventilation System is a once through ventilation system designed to filter the m
(d'
) post-accident contaminated leakages before exiting to meet the 10 CFR 100 guidelines. It maintains building under negative pressure and directs air flow from cleaner to dirtier areas before exiting.
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- The Containment Cooling and Ventilation System is provided with post-accident containment isolation features and filtration units for air cleanup during normal and refueling operations. It limits the radiation release to meet the 10 CFR 100 guidelines in case of a fuel handling accident inside containment.
1.2.10 Fuel IIandling and Storage 1.2.10.1 Fuel IIandling Fuel handling equipment provides for the safe handling of fuel assemblies ar.d CEAs under all specified conditions and for the required assembly, disassembly, and storage of reactor vessel head and internals during refueling.
The major components of the system are the refueling machine, the CEA change platform, the fuel transfer system, the spent fuel handling machine, and the new fuel and CEA elevators. This equipment is provided to transfer new and spent fuel between the fuel storage facility, the containment building, and the fuel shipping and receiving areas during core loading and refueling operations. Fuel is inserted and removed from the core using the refueling machine. During normal operations, irradiated fuel and CEAs are always maintained in a water environment.
The principal design criteria specify the following:
- Fuel is inserted, removed, and transported in a safe manner.
- Subcriticality is maintained in all operations.
Fuel handling is further discussed in Section 9.1.4.
1.2.10.2 Fuel Storage The new fuel and spent fuel storage facilities are described in Sections 9.1.1 and 9.1.2, respectively.
Also included in those sections are sununaries of the criticality and safety analysis.
1.2.11 Auxiliary Systems 1.2.11.1 Shutdown Cooling System The Shutdown Cooling System (SCS) is used to reduce the temperature of the reactor coolant, at a controlled rate, from 350'F to a refueling temperature of 120*F and to maintain the proper reactor ,
coolant temperature during refueling. This system utilizes the shutdown cooling pumps to circulate the reactor coolant through two shutdown heat exchangers, returning it to the reactor coolant system. The component cooling water system supplies cooling water for the shutdown cooling heat exchangers.
l l The SCS for System 80+ has a design pressure of 900 psig. This system pressure provides for greater operational flexibility and simplifies concerns regarding system overpressurization. The SCS pumps do not share functions with the SIS. i I
The SCS is further discussed in Section 5.4.7.
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System 80+ oeslan controlDocument 1.2.11.2 Chem .:al and Volume Control System 8
The Chemical and Volume Control System (CVCS) controls the purity, volume, and boric acid content of the reactor coolanL The CVCS is not required for any safe shutdown or accident mitigation function.
The coolant purity level in the Reactor Coolant System (RCS) is controlled by continuous purification of a bypass stream of reactor coolant. Water removed from the RCS is ce 31ed in the regenerative heat exchanger. From there, the coolant flows to the letdown heat exchanger and then through a filter and a demineralizer where corrosion and fission products are removed. It is then sprayed into the volume control tank and retumed by the charging pumps to the regenerative heat exchanger where it is heated prior to return to the RCS loops and reactor coolant pump seal injection.
4 The CVCS automatically adjusts the amount of reactor coolant in order to maintain a progranuned level in the pressurizer. The level program partially compensates for changes in specific volume due to coolant temperature changes and reactor coolant pump controlled seal leakage. (See Section 9.3.4.2 for details.)
The CVCS controls the boric acid concentration in the coolant by a " feed and bleed" method where the purified letdown stream is diverted to a boron recovery subsystem and either concentrated boric acid or demineralized water is sent to the charging pumps. The divtrted coolant stream is processed by ion exchange and degasification and flows to a concentrator, The coxentrator bottoms are sent to the boric acid storage tank for reuse as boric acid solution and the distillate is first passed through an ion exchanger and then stored for reuse as demineralized water in the reactor makeup water tank.
A description of the CVCS system Boric Acid Storage, Holdup, and Reactor Makeup Water Tanks, and l
- pd Dike is provided in Section 1.2.16.5.
The CVCS for System 80+ incorporates several significant improvements and simplifications including the following:
- Reclassification as a non-safety related system by transferring of previously credited accident mitigation and safe shutdown functions to other dedicated safety systems.
- Improved letdown configuration.
- Improved charging configuration.
Transferring of accident mitigation and safe shutdown functions to other dedicated safety systems has permitted an overall simplification of plant systems. Although not a safety related system, the System 80+ CVCS provides reliable makeup and depressurization capabilities for defense in depth and ease of operation.
System 80+ employs an improved letdown configuration, of which key elements are the following:
- A full pressure letdown heat exchanger.
- Pressure reduction to CVCS operating pressures downstream of the letdown heat exchanger by use of a letdown orifice in series with a letdown flow control valve. i G ,
b i
l Annroved Deskn nietarief hvooduceion (2/95) Page 1.2-17 l j
Sy^ tem 80 + Design ControlDocument In addition, the charging flow is controlled by the use of centrifugal charging pumps and a charging pump flow control valve on the discharge of the pumps.
A description of the CVCS is provided in Section 9.3.4.
1.2.11.3 Process Sampling System The process sampling system is designed to collect and deliver representative samples of liquids and gases in various process systems to sample stations for chemical and radiological analysis. The system permits sampling during reactor operation, cooldown and post-accident modes without requiring access to the containment. Remote samples can be taken of fluids in high radiation areas without requiring access to these areas. The process sampling system performs no safety function.
A description of this system is presented in Section 9.3.2.
1.2.11.4 Condensate Cleanup System The Condensate Cleanup System (CCS) is an integral part of the Condensate System. The CCS is designed to remove dissolved and suspended impurities which can cause corrosion damage to secondary system equipment. The CCS also removes radioisotopes which might enter the system in the event of a primary to secondary steam generator tube leak. The condensate polishing demineralizers will also be used to remove impurities which could enter the system due to a condenser circulating water tube leak.
A description of this system is provided in Section 10.4.6.
1.2.11.5 Steam Generator Blowdown System The design bases for the Steam Generator Blowdown System are:
- Maintain proper steam generator shell-side water chemistry by removing non-volatile materials due to condenser tube leaks, primary to secondary tube leaks, and corrosion products that would otherwise become more concentrated.
- Process steam generator blowdown for reuse as condensate.
- Enable blowdown concurrent with steam generator tube leak (s) or radioactivity present on the secondary side without release of radioactivity to the environment.
- Process a continuous steam generator blowdown rate of either 0.2% or 1% of the full pawer main steam flow.
- Continuously sample the radioactivity of the steam generator blowdown.
- Isolate the blowdown lines leaving the containment upon a Containment Isolation Signal, Main Steam Isolation Signal, or Emergency Feedwater Actuation Signal.
Each steam generator is equipped with its own blowdown line with the capability of blowing down the hot leg and/or the economizer regions of the steam generator shell side. The blowdown is directed into a flash tank where the flashed steam is returned to the cycle via the low pressure feedwater heaters. The liquid portion flows to a heat exchanger where it is cooled, and then directed through a blowdown filter ApprowdDesign Ataterant Introduction Page 1.2-18
System 80+ Design ConW Document l
, /O where the major portion of the suspended solids are removed. After filtration, the blowdown fluid is
- (.) processed by blowdown demineralizers and returned to the condenser. l A description of this system is provided in Section 10.4.8.
1.2.11.6 Condensate and Feedwater Systems i
The Condensate and Feedwater Systems are designed to return condensate from the condenser hotwells to the steam generators. In addition, the systems include a number of stages of regenerative feed and l condensate heating and provisions for maintaining feedwater quality.
The en' ire Condensate System is non-safety-related. The portions of the Feedwater System that are required to mitigate the consequences of an accident and allow safe shutdown of the reactor are safety-related.
~ A description of these systems is provided in Sections 9.2.6 and 10.4.7.
1.2.11.7 Compressed Air Systems The Compressed Air Systems are non-safety related systems consisting of the instrument Air, Station Air, and Breathing Air Systems. The Instrument Air System supplies clean, oil free, dried air to all air operated instrumentation and valves. The Station Air System supplies compressed air for air operated tools, miscellaneous equipment, and various maintenance purposes. The Breathing Air System supplies p clean. oil free, low pressure air to various locations in the plant, as required for breathing protection against airborne contamination while performing certain maintenance and cleaning operations.
A description of these systems is presented in Section 9.3.1.
1.2.11.8 Equipment and Floor Drainage System The Equipment and Floor Drainage System provides the means by which wastes are appropriately segregated and transported to the Liquid Waste Management System (LWMS) in order to minimize the liquid and gaseous radioactive releases. This system accomplishes this function in a manner that is consistent with normal plant operating procedures.
The drains and sumps in the Turbine Building are not normally processed by the LWMS.
A description of the Equipment and Floor Drainage System is provided in Section 9.3.3.
1.2.11.9 Fire Protection System The Fire Protection System minimizes the risks and consequences of fires. The functions provided by the Fire Protection System include the following:
I
- Prompt detection and alarm of fires.
- - Quick suppression of fires.
V e Prevention of the spread of fires. ]
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- Assurance of the capability to achieve safe shutdown in the event of fires.
- Minimization of radioactive exposure and the spread of contamination as the result of fires.
- Provision of manual backup to automatic fire suppression systems.
A description of the Fire Pump House and Tanks is provided in Section 1.2.16.3.
A description of the Fire Protection System is presented in Section 9.5.1.
1.2.11.10 Communication Systems The communication systems are designed to provide effective communications between all areas of the plant and plant site including all vital areas of the plant. In addition, the communication systems are designed to provide an effective means to communicate to plant personnel and offsite utility and regulatory officials during normal conditions and abnormal / accident conditions such as fire, accident, and plant testing.
The Portable Wireless Com nunication System is the primary dedicated means of communication for the plant. In addition to the Portable Wireless Communication System, the Private Automatic Business Exchange (PABX) telephone system and the Public Address (PA) system are designed to provide diverse means of communications to all critical areas of the plant. Additionally, sound-powered telephone systems are provided between selected critical areas of the plant for auxiliary shutdown and other required functional purposes. Finally, multiple offsite communications lines, both direct and through the PABX are provided for effective communications during normal and abnormal / accident conditions. All of these diverse communications systems are independent of each other to assure effective communications assuming a single failure.
A description of these systems is presented in Section 9.5.2.
1.2.11.11 Lighting System The lighting system is designed to provide adequate and effective illumination throughout the plant and plant site including all vital areas of the plant.
The normal lighting system is used to provide normal illumination under all plant operation, maintenance and test conditions.
i The security lighting system provides the illumination required to monitor isolation zones and all outdoor l areas within the plant protected perimeter.
The emergency lighting system is used to provide acceptable levels of illumination throughout the station !
and particularly in areas where emergency operations are performed, such as control rooms, battery rooms, containment, etc., upon loss of the normal lighting system.
A description of these systems is presented in Section 9.5.3.
O A lvvovest Design heaterial- hstroductmn Page 1.2-20
System 80 + Desl.qn ControlDocument 1.2.11.12 Diesel Generator Engine Fuel Oil System The Diesel Generator Engine Fuel Oil System is designed to provide for storage of a seven-day supply of fuel oil for each diesel generator engine and to supply the fuel oil to the engine, as necessary, to drive the emergency generator. The system is designed to meet the single failure criterion. ,
A description of the Diesel Fuel Storage Structure is provided in Section 1.2.16.7.
A description of the Diesel Generator Engine Fuel Oil System is presented in Section 9.5.4.
1.2.11.13 Diesel Generator Engine Cooling Water System The Diesel Generator Engine Cooling Water System is designed to maintain the temperature of the diesel generator engine within an optimum operating range during standby and during full-load operation in order to assure its fast starting and load-accepting capability and to reduce thermal stresses. The system is also designed to supply cooling water to the engine lube oil cooler, the combustion air aftercoolers, and the governor lube oil cooler.
A description of this system is presented in Section 9.5.5.
4 1.2.11.14 Diesel Generator Engine Starting Air System The Diesel Generator Engine Starting Air System is designed to provide start capability for the diesel generator engine by using compressed air to rotate the engine until combustion begins and it accelerates
( under its own power.
A detailed description of this system is presented in Section 9.5.6.
1.2.11.15 Diesel Generator Engine Lube Oil System The Diesel Generator Engine Lube Oil System is designed to deliver clean lubricating oil to the diesel generator engine its bearings and crankshaft, and other moving parts. By means of heaters, the lube oil system is designed to deliver warmed oil to the engine during standby to assure its fast-starting and load- )
accepting capability. The system also provides a means by which used oil may be drained from the ]'
engine and its components, and replaced with clean oil.
A description of this system is presented in Section 9.5.7.
1.2.11.16 Diesel Generator Engine Air Intake and Exhaust System The Diesel Generator Engine Air intake and Exhaust System is designed to supply clean air for i combustion to the diesel generator engine and to dispose of the engine's exhaust. I I
A description of this system is presented in Section 9.5.8. '
1.2.11.17 Diesel Generator Building Sump Pump System O The Diesel Generator Building Sump Pump System is designed to remove leakage and equipment b drainage from the Diesel Generator Building and to protect the diesel generator units from flooding caused by a major pipe rupture.
ANweved Deemn Maeorial- kreodwesion Pope 1.2 21
System 80+ Design ControlDocument A description of this system is provided in Section 9.5.9.
1.2.11.18 Compressed Gas Systems
((The compressed gas supply systems are provided to supply various gases for equipment and instrumentation cooling, purging, diluting, inerting, and welding. The major items of equipment are the high pressure gas cylinders and pressure regulators to control the pressure and distribution of the various gases used throughout the plant. These compressed gas supply systems are non-safety-related and any failure does not jeopardize the operation of any safety-related components or systems))!.
A description of these systems is provided in Section 9.5.10.
1.2.11.19 Potable and Sanitary Water Systems Those portions of the Potable and Sanitary Water Systems (PSWS) that are within the Reactor Building, Nuclear Annex, Turbine Building, Radwaste Building, and Service Building are within the scope of the Certified Design. Those portions of the PSWS that are not within the Reactor Building, Nuclear Annex, Turbine Building, Radwaste Building, and Service Building are not within the scope of the Certified Design. Out of scope portions of the system are licensee supplied and are site specific.
These systems process water for general plant use. These systems serve no safety functions and any malfunction has no adverse effect on any safety-related system. The requirements of General Design Criterion 60 are met as related to design provisions provided to control the release of liquid effluents containing radioactive material from contaminating the PSWS.
These systems are described in Section 9.2.4.
1.2.11.20 Demineralized Water Makeup System The Demineralized Water Makeup System supplies filtered, demineralized water to the Condensate Storage System for makeup and to other systems throughout the plant that require high quality, non-safety-related, makeup water. This system, therefore, serves no safe shutdown or accident mitigation function. and has no safety design bases.
The Demineralized Water Makeup System Demineralizer trains are located in the Station Services Building. A description of the Demineralized Water Makeup System is presented in 9.2.3.
1.2.12 Radioactive Waste Management Systems Radioactive sources and waste management systems are described in Chapter 11. Design considerations to minimize exposure to radioactivity are summarized in Chapter 12.
A description of the Radwaste Building which houses the Solid and Liquid Waste Management Systems is provided in Section 1.2.16.4.
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Conceptual Design information; see DCD Introduction Section 3.4 Anproved Design Matemis! . hstrocketion Page 1.2-22
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.1.2.13 Physical Plant Secudty and Protection From Sabotage l The System 80+ Standard Design features which protect against sabotage are listed in Appendix A to Chapter 13. ((The owner-specific plan will provide details on implementation of certain sabotage ;
j
- . protection requirements.])2 i
- 1.2.14 Cooling Water Systens j s 1.2.14.1- Condenser Circulating Water System !
l The Condenser Circulating Water System provides cooling water for the turbine condensers and rejects {
7 f
heat to the normal heat sink. :
l 1
See Section 10.4.5 for a description of this system. !
?
~t
- l. 1.2.14.2 Station Service Water System l
The Station Service Water System (SSWS) is an open system that takes suction from the Ultimate Heat i i
Sink (UHS) and provides cooling water to remove heat released from plant systems, structures and components. The SSWS returns the heated water to the UHS. The SSWS cools the Component Cooling J
Water System (CCWS) which in turn cools essential and non-essential reactor auxiliary loads.
, i i A description of the Station Service Water System is presented in Section 9.2.1. ]
f \
\ 1.2.14.3 Component Cooling Water System The Component Cooling Water System is a closed loop cooling water system that, in conjunction with ;
the SSWS and the UHS, removes heat generated from the plant's essential and non-essential components !
connected to the CCWS. Heat transferred by these components to the CCWS is rejected to the SSWS l via the component cooling water heat exchangers.
A description of the Component Cooling Water Heat Exchanger Structure is provided in Section 1.2.16.8.
4 A description of the Component Cooling Water System is provided in Section 9.2.2.
j 1.2.14.4 Tmbine Building Cooling Water System ,
The Turbine Building Cooling Water System (TBCWS) provides cooling for the non-safety-related i components in the various turbine plant auxiliary systems. Cooling is effected through heat exchangers I with heat rejected to the Turbine Building Service Water System (TBSWS). This closed cooling water l system is used in lieu of direct cooling by the TBSWS because the quality of the water being circulated )
- in the TBSWS could result in a greater tendency for equipment fouling ami corrosion. i l
A description of the Turbine Building Cooling Water System is provided in hection 9.2.8. l t
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Systern 80+ Design controlDocument 1.2.14.5 Chilled Water System The Chilled Water System (CWS) is designed to provide and distribute a sufficient quantity of chilled water, through a group of dedicated piping systems, to air handling units (AHUs) in specific plant areas.
The CWS is divided into two subsystems, an Essential Chilled Water System (ECWS) that serves primarily safety-related IIVAC coobng loads, and a Normal Chilled Water System (NCWS) that serves non-safety-related HVAC cooling loads.
A description of these systems is provided in Section 9.2.9.
1.2.14.6 Turbine Building Service Water System The Turbine Building Service Water System (TBSWS) removes heat from the TBCWS and rejects the heat to the cooling towers.
The TBSWS uses pumps to circulate water from the plant cooling towers to remove heat from the TBCWS. Condenser circulating water from the cooling towers, is pumped through the TBCWS heat exchangers and is discharged back into the Condenser Circulating Water System at a point between the main condenser cooling water outlet and the cooling tower inlet.
A description of the Turbine Building Service Water System is provided in Section 9.2.10.
1.2.15 lieat Sinks The ultimate heat sink described in Section 9.2.5 consists of a single passive independent cooling water pond. However, it is recognized that site-specific conditions may require the use of two ponds to meet Regulatory Guide 1.27. The design brackets alternative ultimate heat sinks which may be specified for a particular site if environmental restrictions limit the use of a cooling pond or if an alternative water supply is more reliable. Acceptable altern.:te ultimate heat sinks include an ocean, a large lake, a large river, a lake and a cooling pond, a river and a cooling pond, or a cooling tower and cooling pond.
The normal heat sink is not within the scope of the System 80+ Standard Design, but a conceptual design (cooling towers) is provided in Section 10.4.5. The normal heat sink receives the heat load from the Condenser Circulating Water System and the Turbine Building Service Water System.
1.2.16 Miscellaneous Structures 1.2.16.1 Station Services Building / Auxiliary Boiler Structure The Station Services Building (SSB) houses the following typical station equipment and systems:
- Station staff office space
- Plant Makeup Water Treatment
- Food Service and Break Facilities for Personnel Inside the Protected Area
- Turbine Plant Locker, Shower, and Toilet Facilities
- Tool Issue and Storage Areas Approved Design Materset . Irrtrodocta Page 1.2-24
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i System 80 + . Design controlDocument j
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- Cold Machine Shop Conventional Laboratory l
The Station Services Building is designed as non-safety related, with a steel-framed structure and steel deck roof covered by non-combustible roofing. Walls are insulated metal siding. Roof drainage and clean floor drainage are discharged to the storm and waste water.
' A description of the structural elements of the Station Services Building is given in Section 3.8.4.1.9.
The building general arrangement is shown on Figure 1.2-20, sheets 1 through 5. The location of the SSB is given on Figure 1.2-1.
The A"tiliary Boiler Structure (ABS) is a non-safety related, steel framed structure which houses the auxil m boiler and its associated auxiliaries. The location of the ABS is shown on Figure 1.2-1. The structut design is discussed in Section 3.8.4.1.9. The building arrangement is shown on Figure 1.2-21.
, 1.2.16.2 Combustion Turbine Generator and Fuel Storage Facility The Combustion Turbine Generator (CTG) serves as the Alternate AC Power (AAC) source. The CTG and fuel storage facility are non-safety related. The CTG is provided as a packaged unit and is housed
- in a metal enclosure. The CTG is mounted on a reinforced concrete foundation designed for the CTG l and enclosure dead weight, dynamic operating loads and any induced wind loadings. The location of the facility is shown on Figure 1.2-1. The AAC system is discussed in Section 8.3.1.
A Q The CTG is located in close proximity to the normal switchgear building outside of the turbine missile impact zone. The length of electrical output leads are kept to a minimum. Auxiliaries required for operation of the CTG are included, or in close proximity to the CTG enclosure.
The arrangement of the CTG enclosure and components therein provides for routine maintenance of the unit. This arrangement permits removal and replacement of the CTG as a fully assembled, skid mounted unit.
The CTG facility is constructed such that a fire within the enclosure or the CTG does not involve other buildings or equipment, or endanger plant safety or continued operations.
Adequate heating and ventilation maintain the enclosure within acceptable temperature limits. The enclosure and related auxiliary structures are designed for normal design conditions such as flood, wind, etc.
Since the CTG provides backup pov.cr !9 station security systems, the enclosure and related auxiliary structures will be located in a secure area of the plant for protection.
Fuel storage tanks are designed in accordance with National Fire Protection Association (NFPA)-30 and are located outside the turbine missile strike zone, or the tanks are hardened or otherwise shielded against missiles.
1.2.16.3 Fire Pump House and Tanks
- p (J The fire pump house contains the two main fire pumps (one diesel engine driven and one electric motor driven), the electric motor driven pressure maintenance pump, and their associated controls, drivers, Nwend auto,, meww. muon rare r.2 2s
System 80+ Design ControlDocument piping and fittings described in Section 9.5.1.5.1. The two water storage tanks are each 300,000 gallon capacity, ground level, steel construction tanks. The facility is shown on Figure 1.2-22.
The building is a non-safety related, non-seismic steel-framed structure with a steel deck roof covered by non-combustible roofing. Walls are insulated metal siding.
The building is subdivided, utilizing a 3-hour rated fire barrier, into two separate areas, one for the diesel driven main fire pump and its associated controller, fuel tank, piping, and fittings, and the other for the electric motor driven main fire pump and the electric motor driven pressure maintenance pump, their associated controllers, piping and fittings.
The building heating and ventilation system maintains the building within design temperature limits.
The fuel tank for the diesel driver is located inside the building within a concrete containment dike sized to hold the full contents of the tank.
Roof drainage and floor drainage discharge to the storm and waste water system. The diesel engine fire pump room sump pump discharges to an oil separator prior to discharge.
The water storage tanks are non-safety related, non-seismic vertical, cylindrical, steel wall tanks designed in accordance with NFPA-22, Standard for Water Tanks for Private Fire Protection.
1.2.16.4 Radwaste Building The Radwaste Building houses the following systems:
e Liquid Waste Management System (LWMS) e Solid Waste Management System (SWMS)
The Radwaste Building is located adjacent to the Nuclear Annex. Provisions are included to preclude an uncontrolled release of effluents to the environment due to a LWMS or SWMS failure. If site specific requirements preclude locating the Radwaste Building adjacent to the Nuclear Annex, the consequences of a LWMS or SWMS failure are evaluated to demonstrate compliance with 10 CFR 20, Appendix B limits. This analysis demonstrates that the concentration of the liquid effluent at the potable water source, released from the Radwaste Building due to a LWMS or SWMS failure, is within 10 CFR 20, Appendix B limits.
The Radwaste Building is located in close proximity to the Interim Onsite Storage Facility to facilitate transport of packaged waste for interim storage prior to shipment to a licensed burial facility.
Adequate space is provided for storage and processing of radwaste.
Ventilation ensures controlleil and monitored release of gaseous effluent from the Radwaste Building.
The Radwaste Building i:: equipped with area and airborne radiation monitors to provide indication of a spill and to ensure that personnel exposures are maintained ALARA.
A structural description of the Radwaste Building is given in Section 3.8.4.1.7. The general arrangement of the building is shown on Figures 1.2-23, sheets I through 6. The waste management systems are l
Approved Design Materhol . Introduction Page 1.2-26
l System 80+ oesign controlDocument i
, l described in Sections 11.2 and 11.4. The gaseous waste system which is located in the nuclear annex is described in Section 11.3.
1.2.16.5 Boric Acid Storage, Holdup, and Reactor Makeup Water Storage Tanks and Dike l The Boric Acid Storage, Holdup, and Reactor Makeup Water Storage tanks are in the Chemical and l ;
Volume Control System (CVCS), and are of vertical, cylindrical, single-wall stainless steel construction.
. The Boric Acid Storage Tank is a Seismic Category I, NNS tank located in the yard. The tank is designed to ASME III, Class 3 requirements and is contained within a common Seismic Category II ,
reinforced concrete dike structure which provides protection against natural phenomena such as :
earthquakes, tornadoes, hurricanes and floods. The structure also serves as a containment barrier to limit the release of water due to leaks or spills. The tank is equipped with an overflow line which is of sufficient size to handle a potential storage tank overflow. The Boric Acid Storage Tank is described in l Section 3.8.4.1.6. The location is shown on Figure 1.2-1.
The Holdup and Reactor Makeup Water Tanks are non-seismic and non-safety related. Both tanks are ;
4 designed to API-650 and are located outdoors within the common Seismic Category II reinforced concrete dike structure of sufficient height / size to retain potential leakage, up to the complete contents of all tanks in the event of tank ruptures. The dike for these tanks is described in Section 3.8.4.1.11. The dike location is shown on Figure 1.2-1.
A description of the CVCS system is presented in Section 9.3.4.
Condensate Storage Tank / Dike ;
1.2.16.6 The Condensate Storage Tank is designed to API-650 and is non-safety related. It is surrounded by a Seismic Category Il reinforced concrete dike of sufficient height / size to retain the entire contents of the tank in the event of a tank overflow or tank failure. The Condensate Sorage Tank and dike are described in Section 3.8.4.11. The location is shown on Figure 1.2-1.
The Condensate Storage System is described in Section 9.2.6.
1.2.16.7 Diesel Fuel Storage Structure There are two Diesel Fuel Storage Structures, one on each side of the Nuclear Annex. The structures are Seismic Category I and are designed to withstand fire, sabotage, internally and externally generated missiles, floods, tornados, hurricanes and the Safe Shutdown Earthquake. Each structure contains two, one-half capacity, steel storage tanks, separated by a 2-hour-rated fire barrier. An adjacent steel framed, non-nuclear safety, Seismic Category 11 equipment room houses auxiliary equipment.
A description of the Diesel Fuel Storage Structure is given in Section 3.8.4.1.4. The building arrangement is shown on Figure 1.2-24. A more detailed description of the Diesel Fuel Tank Structure is provided in Section 9.5.4. ,
1.2.16.8 Component Cooling Water Heat Exchanger Structure (s)
The Component Cooling Water (CCW) heat exchanger structure (s) houses the four CCW heat
( exchangers, associated piping, valves, auxiliaries and sump pumps described in Section 9.2.2. The Anwovenf Dene,s nieterset. krev&oc6ert (2/95) Pope 1.2-27
I System 80+ Design ControlDocument 1
structure (s) meets Seismic Categori I requirements and is capable of withstanding the effects of the following events:
- Natural phenomena, including SSE, floods, tornados and hurricanes
- Externally and internally generated missiles
- Fire and sabotage The structure (s) is of reinforced concrete construction and is located within the site security boundary and outside the turbine missile path. Complete separation of the two CCW divisions is provided by physical barriers for fire, single failure, pipe whip and seismic interaction effects.
The general arrangement of the CCW heat exchanger structure (s) is shown in Figure 1.2-25.
O O
Aproved Design Material Introduction pog, y,2,gg
l l
Sv' tem 80+ Deskn controlDocument Table 1.2-1 System 80+ hnprovements Based on Operating Experience (mv)
This table summarizes the major design improvements which have resulted from design and analysis experience as well as plant startup and operating experience. The experience input to the System 80+
design process has been accrued through the organizations panicipating in the System 80+ design team.
This includes architect engineering organizations (Stone & Webster Engineering Corporation and Duke ,
Engineering & Services, Inc.) which have extensive experience in plant design and, in the case of Duke Engineering & Services, actual plant operating experience. Architect engineering experience is reflected mainly in the plant layout, building design, control room, and the many " balance of plant" systems supporting the Nuclear Steam Supply System. This experience was brought to the System 80+ design team by the engineers responsible for the design of specific structures and systems in currently operating plants and by actual plant operators who also panicipated in the design process. The ALWR Utility Requirements -
Document was also used in the design of System 80+ and the design and operating experience of participating utilities reflected therein has been incorporated through the adoption of design requirements.
Experience related to the operation of the Nuclear Steam Supply System was brought to the System 80+
design through the predecessor System 80 and c"" Nuclear Steam Supply System designs and through the years of experience of individual designers. This . vidual experience was developed through review of industry experience reflected in documents such as NRC Bulletins and Generic Letters (See Tables 1.8-2 and 1.8-3), Unresolved and Generic Safety Issues (see Chapter 20), Institute for Nuclear Power Operations publications, and in the ABB-CE Corrective Actions Program. Their experience was also developed through participation on design teams for startup of plants with Nuclear Steam Supply Systems designed by ABB-CE.
Operating experience is reflected throughout the System 80+ design described in the Approved Design p Material, including shutdown risk improvements, which are reported in the shutdown risk section of Chapter Q 19. The major improvements based on operating experience are summarized below.
Integrated Design Process One organization, ABB-CE, is responsible for the design of structures, systems, and components of a plant which are important to safety (where design features depend on site-specific characteristics, interface i requirements are provided), thus facilitating an integrated design process. The major considerations in this ,
integrated design approach are as follows: j 1
- 1. The PRA is used to evaluate the design and to identify areas where significant improvement cts be obtained. Although the end product of the PRA is a calculation of core damage frequency and offsite consequences, the PRA can also be used to gain design insights and identify improvements j for handling more frequent transients and accidents (Chapter 19).
- 2. Maintainability of the plant is being addressed by using equipment that mintnuzes the need for maintenance, by assuring that equipment can be easily accessed, and by assuring that maintenance actions will be as simple as possible (so as to avoid unplanned reactor trips and plant downtime). j These same considerations apply to periodic testing and inspection of equipment.
- 3. In almost all cases for System 80+, safety and non-safety functions have been separated. This will make the plant much simpler to operate and maintain.
v 4provedDoeQn Noenrint onevehcainer Page 1.2 29
System E10+ Design ControlDocument Table 1.2-1 System 80+ Improvements Based on Operating Experience (Cont'd.)
- 4. Human factors (i.e., the man-machine interface) are considered throughout the plant and especially in the control room (Chapter 18).
- 5. ALARA considerations affect the selection of materials and location of piping and equipment that carry radioactive coolant. For example, specifications for the reactor coolant system materials have been tightened to minimize transport of contarmnation. Improvements in the steam generator tubing material and access openings greatly reduce radiation exposures for maintenance, testing, and inspection. De overall goal is to maintain personnel exposure to less than 100 man-rems per year for each reactor (Chapter 12).
- 6. Plant security (i.e., sabotage protection) and fire protection concerns have been directly addressed in determining layouts for plant safety systems (Section 13.6).
Increased RCS Design Margins and Improvements
- 1. Reactor: The core operating margin has been increased by reducing the normal operating hot leg temperature and revising core parameter monitoring methods. The ability to change operating power level (i.e., maneuver) using control rods only (without adjusting boron concentration in the coolant system) has been provided, simplifying reactivity control during plant load changes and reducing liquid waste processing requirements (Sections 4.3 and 4.4).
c
- 2. Reactor Pressure Vessel: De reactor vessel is ring-forged with material specifications that result in a sixty year end-of-life RTsor well below the current NRC screening criteria. This results in a significant reduction in the number of welds (with resulting reduction in inservice inspection) and eliminates concern for pressurized thermal shock (Section 5.3).
- 3. Pressurizer: The pressurizer volume is increased to enhance the transient response of the RCS and to reduce unnecessary challenges to safety systems (Section 5.4.10).
l l 4. Steam Generators: The steam generators include alloy 690 tubes, improved steam dryers, and a j seventeen percent increase in overall heat transfer area, which includes a ten percent margin for i potential tube plugging. The steam generators have a twenty-five percent larger secondary l feedwater inventory to extend the ' boil dry" time and improve response to upset conditions. Steam j generator improvements also have been added to facilitate maintenance and long term integrity.
These include larger and repositioned manways, a standby recirculation nozzle, and a redesigned flow distribution plate (Section 5.4.2).
- 5. Mechanical improvements based on System 80 startup and operating experience include strengthened reactor coolant pump impellers, redesigned reactor coolant temperature detector thermowells, strengthened reactor vessel upper guide structure, specification of antimony-free reactor coolant pump bearings, strengthened reactor coolant pump shafts, and redesigned steam generator economizer internals.
l O' l Apfvered Design nestere! hstrocketion (2/95) Page r.2.10 1 l
Sy* tem 80 + Design ControlDocument (qj Table 1.2-1 System 80+ hnprovements Based on Operating Experience (Cont'd.) ,
, o Advanced Control Room Design
- 1. The Advanced Control Complex (Nuplex 80+) for System 80+ has been designed to meet demanding human factor, reliability, and licensing requirements, and is characterized by state-of- ,
the-art advances, such as distributed digital processing, fiber optic data communications, and touch sensitive video displays (Chapter 18).
- 2. Nuplex 80+ is a total integration of plant-wide instrumentation and controls (l&C) systems. The Advanced Control Complex includes the Main Control Room, the Technical Support Center, the Remote Shutdown Room, Computer Room; the Vital Instrumentation and Equipment Rooms, Non-Essential Electrical Equipment Rooms and their respective control, protection, and monitoring ,
systems.
- 3. Redundancy and diversity in all information processing and display ensures the correctness of information presentation and allows continued operation with equipment failures. Sufficient diversity is provided to ensure that the plant could be brought to a safe condition even with the loss of all safety-related digital instrumentation and controls. The integration of information from the former Safety Parameter Display System and the Post Accident Monitoring Instrumentation (PAMI) into normal operating displays allows the same displays to be used during all plant conditions.
- 4. Alarms are based on validated signal inputs with logic and setpoints that account for plant and equipment operating modes. Four levels of alarm presentation are employed. Individual and global p, alarm acknowledgement features ensure that all alarms are recognized without operator task Q overload. Alarm acknowledgement provides direct access to supponing displays.
Highly Reliable Engineered Safeguards Systems
- 1. Chemical and Volume Control System (CVCS): The CVCS incorporates numerous significant improvements which include centrifugal charging pumps, a high pressure letdown heat exchanger, and simplified charging and auxiliary spray piping. A diverse positive-displacement charging pump has been added as a third source of cooling for the reactor coolant pump seals. Required safety functions previously performed by the CVCS are now delegated to other dedicated safety systems (Section 9.3.4).
- 2. Safety injection System (SIS): The SIS design has been improved to provide a simpler and more reliable system with increased redundancy, it has four mechanical trains for safety injection, direct-to-vessel injection connections, and an in<ontainment refueling water storage tank. The same size pumps and valves used in the original System 80 two train design are now used in all four trains.
The trains are not interconnected by common headers and include provision for full flow, on-line testing to eliminate the need to extrapolate bypass-flow test results to demonstrate compliance to Technical Specifications (Section 6.3.2).
- 3. In-Containment Refueling Water-Storage Tank (IRWST): The IRWST has been located in the containment building, in a torus-like configuration around the reactor vessel cavity. Containment
.. water collection points empty into the IRWST. This means that the safety injection pumps always take water from the tank, eliminating the need to switch from tank to containment sump following a loss of coolant accident (Section 6.8).
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1 Table 1.2-1 System 80+ Improvements Based on Operating Experience (Cont'd.) g l Highly Reliable Engineered Safeguards Systems (Cont'd.) ;
- 4. Safety Depressurization System (SDS): The SDS is a dedicated manually-operated system designed to permit depressurization of the Reactor Coolant System (RCS) when normal processes are not available. The SDS provides the capability to rapidly depressurize the RCS so that an operator can initiate primary system feed and bleed (using the safety injection pumps) to remove decay heat following a total loss of feedwater event. Manual control of motor operated valves enable discharge from the pressurizer to be directed to the IRWST, without the reliability concern that is associated with automatically operating valves (Section 6.7).
- 5. Emergency Feedwater System (EFWS): The EFWS is a dedicated safety system intended for emergency use only. (The Main Feedwater System includes a startup pump and a full range control system for normal startup and shutdown operations). 1 The EFWS has two separate trains. Each consists of one emergency feedwater storage tank, one full capacity motor-driven pump, one full capacity non-condensing steam-driven pump, and one cavitating venturi. The cavitating venturi minimizes excessive emergency feedwater flow to a steam generator with a ruptured feed or steam line. The EFWS therefore requires no provision for automatic isolation of emergency feedwater flow to a steam generator having a ruptured steam line or feed line (Section 10.4.9).
- 6. Shutdown Cooling System (SCS): The SCS design pressure has been increased to 900 psig. This higher pressure provides greater operational flexibility and eliminates concern for system over-pressurization. The SCS is interconnected with the Containment Spray System, which uses identical pumps. The reliability of both systems is therefore increased, and each set of pumps can serve as a backup for the other (Section 5.4.7).
Plant Structures and Arrangements
- 1. The containment for System 80+ is a 200-foot diameter steel sphere which maximizes space for equipment and maintenance while minimizing unusable volume in the upper part of the containment.
The operating floor offers 75% more usable area than a cylindrical containment of equal volume (Sections 3.8 and 6.2).
- 2. Features for mitigating the consequences of postulated severe accidents include a reactor vessel cavity designed to imprme the ability to resolidify molten core material on the cavity floor by cWling and retaining the molten core debris (Section 6.8).
- 3. The spherical containment provides a lower annulus tader the sphere which replaces a conventional safetygrade auxiliary building, and is an ideal locatica for safety systems. Placing of the safeguards equipment in the sub-sphere areas is an economically attractive approach to addressing numerous regulations associated with this equipmm. Separation for internal flood mitigation, fire protection, security, and sabotage concems e easily addressed wi'.hout adverse affect on accessibility (Section 3.8).
O AporonN1 Design Material- hstrodoction Page 1.2-32
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