ML20247G666
| ML20247G666 | |
| Person / Time | |
|---|---|
| Site: | 05200002, 05000470 |
| Issue date: | 03/30/1989 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML20247G537 | List:
|
| References | |
| NUDOCS 8904040288 | |
| Download: ML20247G666 (125) | |
Text
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CECCAD ESI N E= *J e7 P1L ER CERTIFICATION (Sheet 1 of 3) pm
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'% ) {
i EFFECTIVE PAGE LISTING CHAPTER 1 l
Tabl9.__p.f_Cpj) tents i
Page Amendment I l
i E l ii E iii E 1 i
iv E v E vi E vii E viii E l
Text Page Amendment
/7 1.1-1 E 5, 1.1-2 E 1.1-3 E 1,2-1 E 1.2-2 E 1.2-3 E 1.2-4 E 1.2-5 E 1.2-6 E l 1.2-7 E 1.2-8 E 1.2-9 E 1.2-10 E 1.2-11 E 1.2-1.? E 1.2-33 E 1.2-14 E 1.2-15 E 1.2-16 E l 1.2-17 E 1.2-18 E 3.2-19 E 1.2-20 E 1.2-21 E l
- 1. ' - 2 2 E
,m, 1 2-23 E E
l t'-} 1.2-24 1.2-25 E 8904040288 390330 Amendment E ADOCK 0000 0 . December 30, 1988
{DR
CESSAR Ennneuion (Sheet 2 of 3)
O EFFECTIVE PAGE LIRTXHG (Cont'd)
CJIAPTER 1 Text (Cont'd)
P_ age Agendmqnt i 1.2-26 E 1.2-27 E 1.2-28 E 1.2-29 E 1.3-1 E 1.4-1 A 1.4-2 1.4-3 E 1.44 A 1.5-1 E 1 I
1.5-2 E 1.5-3 A 1.6-1 E 1.6-2 E 1.6-3 E I
' i 1.6-4 E 1.6-5 E 1.7-1 E 1.8-1 E l
, .i . 9 - 1 E I
Tables Amendment 1.3-1 (Sheet 1) E j 1.3-1 (Cheet 2) A i 1.3-1 (Shoot 3) E f 1.3-1 (Sheet 4) E l 1.3-1 (Sheet 5) E l 1.3-1 (Sheet 6) E ;
1.3-1 (Sheet 7) A i 1.3-1 (Sheet 8) E ..
1.3-1 (.' sheet 9 ) A j 1.3-1 (Sheet 10) E i 1.3-1 (Sheet 11) E l 1.3-1 (Sheet 12) E I 1.3-1 (Sheet 13) A O
i i
Amendment E 1 December 30, 1988 1
CESSAR nnincoi:n (Sheet 3 of 3) p l
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_ EFFECTIVE PAGE LISTING (Cont'd)
CHAPTER 1_
Tables (Cont'd) Amendment 1.3-2 E !
1.4-1 (Sheet 1) A 1.4-1 (Sheet 2) E 1.5-1 B 1.7-1 (Shoot 1) E .
1.7-1 (Sheet 2) E l 1.7-1 (Sheet 3) E l 1.7-1 (Sheet 4) E ]
1.7-2 E j l 1.'/-3 (Sheet 1) E s 1
1.7-3 (Sheet 3) E I 1.8-1 (Sheet 1) E l 1.8-1 (Sheet 2) E ;
1.8-1 (Shoot 3) E I 1.8-1 (Sheet 4) E i O 1.8-1 (Sheet 5) 3 l 1.8-1 (Sheet 6) E l
1.8-1 (Sheet 7) E
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1,8-1 (Sheet 8) E j 1.8-1 (Sheet 9) E ;
1.8-1 (Sheet 10) E j 1.8-1 (Sheet ll? E !
1.8-1 (Shect 12) E !
- 1. 8 .1 (Sheet 13) E l 1.8-1 (Sheet 14) E l 1.8-1 (Sheet 15) E 1.8-1 (Sheet 16) E l 1.8-1 (Sheet 17) E ,
1.8-1 (Sheet 18) E i 1.8-1 (Sheet 19) E l l
F_igures Amendment 1.2-1 E I 1.2-2 E !
l 1 . 2 -- 3 E 1.2-4 E l 1.2-5 E !
1.2-6 E 1.2-7 E '
i g 1.2-8 E 1.2-9 E I
(b) 1.2-10 E l 1.7-1 1
l Amendment E December 30, 1988
)! hhk bERT FICATION
(%i
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\~J TABkE_OF CONTENTS ]
Section Subiect Page No. I l
1.0 INTRODUCTION
AND GENERAL DESCRIPTION OF 1.1-1 J1LANT l t
i
1.1 INTRODUCTION
1.1-1 1.1.1 SYSTEM 80+ " STANDARD DESIGN 1.1-1 1.1.2 POWER LEVELS 1.1-2 1.1.3 SEk~RE ACCIDENT POLICY 1.1-2 E 1.2 GENERAL PLANT DESCRIPTION 1.2-1 1.2.1 PRINrrvAL SITE CHARACTERISTICS 1.2-1 l f
1.2.1.1 Site Location 1,2-1 !
l 1.2.1.2 Plant Stirroundinan 1.2-1
- N
_, 1.2.1.2.1 Meteorology 1.2-1 1.2.1.2.2 Hydrology 1.2-1 !
1.2.1.2.3 Geology and Seismology 1.2-1
).2.1.3 Plant Independence 1.2-1 E
1.2.1.4 Site Buildinq Arrangement 1.2-1 j 1.2.2 SYSTEM 80+ STANDARD DESIGN - SCOPE AND 1,2-2 U
DESCRIPTION 1.2.3 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) 1.2-2 J.2.3.1 Reactor Core 1.2-2 1.2.3.2 Reactor Internals 1.2-3 l 1.2.3.3 Reactol Coolant System (RCS) 1.2-4 1.2.4 TNGINEERED SAFETY FEATURES 1.2-5 1.2.4.1 Containment Systen 1.2-5 1.2.4.2 Safety Iniection System 1.2-6 gj 1.2.4.3 Emeroency Feedwater System 1.2-6 l
l Amendment E i December 30, 1988
y _ -~
CESSARn!Mcuiu l
TABLE OF CONTENTS (Cont'd) l CHAPTER 1 c
Sec.tio11 Subiecj; PacL No.
1.2.4.4 SA(ety Depressurization System 1.2-7 1.2.4.5 Containment Spray System 1.2-7 1.2.5 INSTRUMENTATION AND CONTROL 1. '2 - 7 1.2.5.1 Protection. Control. and 1 2-8 Instrumentation SvEt3Jns 1.2.5.1.1 Reactor Protective System 1.2-8 1.2.5.1.2 Alternate Protection System 1,2-9 1.2.5.1.3 Engineered Safety Features 1.2-9 Actuation System 1.2.5.1.4 Reactor Control Systems 1.2-9lE ]
1.2.5.1.5 Nuclear Instrumentation 1.2-10 1.2,5.1.6 Process Monitoring Systems 1.2-11 ,
1.2.6 NUCLEAR PLANT CONTROL CENTER 1,2-11 1.2.6.1 Main Control Panels 1.2-12 1.2.6.2 Remote Shutdown _P.Arle]fi 1.2-12 1.2.6.3 Alarm andj)]liplay System 1.2-13 ,
1.2.6.4 Data Processing _ System 1.2-13 E ;
1.2.6.5 Component CoILtrol System 1.2-14 1.2.7 ELECTRICAL SYSTEM 1.2-15 1.2.8 POWER CONVERS10N SY4 TEM 1.2-16 1.2.9 HEATING, VENT 1LATING, AND AIR CONDITIONING 1.2-17 SYSTEMS 1.2.10 FUEL HANDLING AND STORAGE 1.2-17 1.2.10.1 Fuel Handling 1.2--17 1.2.10.2 Fuel Storace 1.2-18 O
Amendment E ii December 30, 1988
CESSAR Eminemot t
TABLE OF CONTENTS (Cont'd) i 1
CHAPTER 1 Section Subject Face No. .j 1.2.11 AUXILIARY SYSTEMS 1.2-18 )
1.2.11.1 Shutdovr1 Cool _ing System 1.2-18 l 1.2.11.2 Chemica) and Volume control Sys_tg3 1.2-18 1.2.11.3 Procesp_ lip 2pling Syster) 1,2-20 1.2.11.4 Condensate Cleanup Systen 1.2-20 1.2.11.5 Steam Generator Blowdown System 1.2-20 1.2.11.6 l ginJensate and Feedwater System 1,2-21 l 1.2.11.7 Compressed Air Systems 1.2-21 1.2.11.8 .E,JJuipmeDt and Floor Drainale System 1.2-22 1.2.11.9 Fire Protection System 1.2-22 E I
1.2.11.10 Communi.qation Systems 1.2-23 1.2.11.11 Lichtina Systems 1,2 1.2.11.12 D1_esel Generator Furl Oil System 1.2-25 1.2.71.13 Djtesel Generator Cool _ina Water System 1.2-25 1.2.11.14 Diesel Generator Startina Air System 1.2-25 1.2.11.15 Diesel Generator Lube Oil System 1.2-25 1.2.11.16 pi ese] Generator Air Intake and 1.2-26 Exhaust System 1.2.11.17 Diesel Generator Buildina Sump 1.2-26 -
Pump System 1.2.11.18 Compressed Gas Systems 1.2-26 1.2.11.19 Potable and Sanitary Water Systems 1.2-26 1.2.11 20 Demineralized Water Makeup System 1,2-261 l
l Amendment E iii December 30, 1988 l
I______-_--____ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._ _ _ . - _ . -
I I)l5!$!$/klI SIMSficArsow 1 I
y 1
Oil TABLE OF_QONTEjiTS (Cont'd)
CHAPTER 1 Se9. tion Su_b-l e c_t _P_ A g p_ N o .
l 1.2.12 RADIOACTIVE WASTE MANAGEMENT SYSTEMS 1.2-27 )
i 1.2.13 PHYST. CAL PLANT SECURITY AND PROTECTION i FROM SABOTAGE 1. 2-27IB ]
1.2.14 COOLING WATER SYSTEMS 1.2-27 1
1.2.14.1 Condenser Circulating Water System 1.2-27 1.2.14.2 Station Service Whter Sy. stem 1.2-27 1
1.2.14.3 , Component _Coolina_ Water Systgm 1.2-28 )
i 1.2.14.4 Tyrbine Byildina Coolina Water System 1.2-28 I l
l 1.2.14.5 .gbilled Water Systen 1.2-28 1.2.14.6 Turbine B>1ildina Service Water __Sy. stem 1.2-28 1.2.15 ULTIMATE HEAT SINK 1.2-29 1
1.3 QQ11PARISON TABLES 1.3-1 E
1 1.3.1 COMPARISON WITH SIMILAR FACILITY DESIGNS 1.3-1
)
1.3.2 COMPARISON OF FINAL AND PRELIMINARY 1.3-1 INFORMATION 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.4.1 APPLICANT'S QUALIFICATIONS AND EXPERIENCE 1.4-1 1.4.2 ARCHITECT-ENGINEER'S QUALIFICATIONS AND 1.4-1 EXPERIENCE 1.4.3 COMBUSTION ENGINEERING'S QUALIFICATIONS 1.4-1 AND EXPERIENCE O8 i
Amendment E iv December 30, 1988 i
CESSAR nahuou r~N l t
t/
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i TABLE,_0_f' ColiTENTS (Cont'd)
CHAPTER 1 Section Subiect Pace No.
1.4.3.1 Pre-Commercial Reactor P:;oorams 1.4-1 1.4.3.1.1 Naval Propulsion Program 1.4-1 l 1.4.3.1.2 Boiling Nuclear Superheat (BONUS) 1.4-1 Plant l
1.4.3.2 Development and Design of Commercial 1.4-2
- PWR Systems l
1.4.3.3 Maior Component Design and Fabrication 1. 4 -3 j 1.4.3.4 Facilities 1.4-3 1.4.3.5 Commercial Reactor Operation 1.4-4 l 1.5 REQUIREMENTS FOR FURTHER TECHNICAL 1.5-1
/~N INFORMATION l ! ')
V 1.5.1 TOPICAL PROGRAM
SUMMARY
1.5-1 )
1.5.2 SYSTEM 80 - 16 x 16 ASSEMBLY TEST PROGRAM 1.5-1 l 1.5.2.1 Components Testing 1.5-1 j
1.5.2.2 Fuel Assembly Seismic Testing 1.5-1 l 1.5.2.3 _ Reactor Flow Model Testina 1.5-2 1.5.2.4 DNB Improvement 1.5-2 !
1 1.5.2.5 Fuel Development Procrams 1.5-2 1.5.3 SYSTEM 80 STEAM GENERATOR DEVELOPMENT 1.5-2 I PROGRAMS i
1.6 FIATERIAL INCORPORATED BY REFERENCE 1.6-1 l 1.7 DRAWINGS AND DIAGRAMS 1.7-1 1.7.1 ELECTRICAL, INSTRUMENTATION, AND CONTROL 1.7-1 DRAWINGS (a) s-1.7.2 PIPING AND INSTRUMENTATION DIAGRAMS 1.7-1 E l
Amendment E v December 30, 1938 I
CESSAR Enli"icariou TABLE OF CONTENTS (Cont'd)
CHAPTER 1 Section pubiect Pacre No.
1.8 REGULATORY GUIDES 1.8-1 1.9 SYSTEM 80+ STANDARD DESIGN INTERFACES 1.9-1 E l
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I Amendment E vi December 30, 1988
C E S S A R n ainc m :u l n\
\m >
LIST OF TADLES j CHAPTER 1 T_able Subiect E ;
1.3-1 Comparison of Reactor Characteristics 1.3-2 Docket Listings For C-E Recent Reactor Designs 1.4-1 C-E Pressurized Water Reactor Plants lE 1.5-1 Summary of Development Programs to Demonstrate System 80 Design Ccaservatism I
1.7-1 Safety Related Electrical, Instrumentation and control Drawings 1.7-2 Valve List Identifiers E 1.7-3 Piping and Instrumentation Diagrams
/"'N 1.8-1 Regulatory Guides
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Amendment E vii December 30, 1987
CESSAR Sini"icums O !
LIST OF FIGURES f CHAPTER 1 F_igure SMMM !
1.2-1 Reactor Building Plan at Elevation 50+0 ;
1.2-2 Reactor Building Plan at Elevation 65+0 l
1.2-3 Reactor Building Plan View of Upper Dish i
1.2-4 Reactor Building Plan at Elevation 91+9 (Top of IRWST) 1.2-5 Reactor Building Plan at Elevation 115+6 E
1.2-6 Reactor Building Plan at Elevation 146+0 (Operating Floor) (
1.2-7 Reactor Building Section From 180 to 0*
1.2-8 Reactor Building Section From 90* to 270" 1.2-9 Reactor Building Miscellaneous Sections 3.2-10 Site Plan 1.7-1 Piping and Instrumentation Diagram Symbols ,
1 i
1 9' I Amendment E viii December 30, 1988
l CESSARanecm.
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1.0 INT _RODRpTION AND GENERAL DESCRIPTION OF PLANT
1.1 INTRODUCTION
The Combustion Engineering Standard Safety 'nalysis Report -
Design Certification (CESSAR-DC) has been prep 'ed in support of Th the industry effort to standardize submittal demonstrates the compliance of the System nuclear pla m designs. 80+g ,
)
Standard Design with all current regulations for existing plants !
as well as the guidelines for future plants outlined in the I B
Commission's Severe Accident, Safety Goal, and Standardization )
Policy Statements. )
I' The starting point for development of the System 80+ Standard i
Design was the System 80 design described in CESSAR-FSAR. ,
I CESSAR-FSAR was referenced (in total or in part) in the safety {
analysis reports submitted by Arizona Public Service and Washington Public Power Supply System. In developing the System 80 design, changes from earlier C-E designs were made to respond to utility needs and provide increased conservatism (including for Loss-of-Coolant Accident (LOCA) conditions). These improvements included a larger core size, greater number of fuel L, rods and modification in the guidance method for control element j v assemblies used for reactor control and rapid shutdown. The l larger core size and increased number of fuel rods allowed for higher reactor power levels with a decrease in peak heat rating and a resultant decrease in fuel rod temperatures under LOCA i conditions. The modified Control Element Assembly guidance !
1 system design provided increased conservatism and flexibility in the reactivity control and shutdown capability.
While the System 80+ design contains most of the features of System 80_, a variety of engineering and operational improvements are included.* The changes to System 80 nre designed to provide additional safety margin and address Severe Accident, Safety !
Goal, and Standardization Policy Statements. ,
l 1.1.1 SYSTEM 80+ STANDARD DESIGN l l
A summary of the System 80+ Standard Design is presented in Section 1.2 and detailed information on specific systems is j provided in the appropriate sections of CESSAR-DC. 1
~
- Specifically, for the 8ystem 80+ Standard Design, the Electric Power Research Institute's Advanced Light Water heactor Requirements Document has been used as a guide for utility requirements regarding plant design.
Amendment E '
1.1-1 Dacember 30, 1988
i CESSARinMemou O
1.1.2 POWER LEVELS The System 80+ Standard Design, described herein, includes a reactor core designed to operate at a maximum core power level of 3800 MWt. While the System 80+ design is independent of power 3 level, the maximum core power level licensable in the United States was selected for the analysis described herein to provide limiting design and safety analysis parameters. For this core power level, the total thermal output is 3817 MWt.
E 1.1.3 SEVERE ACCIDENT POLICY The requirements to be met by future plants are:
A. Demonstration of compliance with the procedural requirements and criteria of the current Commission regulations, including the Three Mile Island requirements for new plants as reflected in the CP Rule (10 CFR 50. 34 ( f) ) .
l B. Demonstration of technical resolution of all applicable B l 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.
C. 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.
D. 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 Commicsion also recognizes the importance of such E
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."
- CESSAR-DC Will address the Severe Accident Policy and the 3 ;
unresolved generic issues. The resolution of these issues will l O
Amendment E 1.1-2 December 30, 1988 i
CESSAR !anncuiou 1 I U
be summari,Ned in CESSAR-DC Appendix A and they will take into full consideration the acceptance criteria from EPRI AIMR and DOE II ARSAP Topic Papers. A Level III PRA will be performed. This PRA, j will be described in Appendix B. Degraded core analyses will be !
included in the PRA. Results of the Sabotage Protection Program will be presented in CESSAR-DC, Appendix 13A. L l
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Amendment E 1.1-3 December 30, 1988 E____________________
CESSARnnLmu __
1.2 GE@RAL _ PLANT DESCRIPTION 1.2.1 PRINCIPAL SITE CHARACTERISTICS 1.2.1.1 Site Location The System 80+ Standard Design is designed for use at multiple sites as described in Chapter 2. T'le site-specific SAR willa '
identify the specific site for that unit. gA 1.2.1.2 Pl_ ant Surroundings _
The System 80+ Standard Design is designed for use at multiple sites. The site-specific SAR will identify the specific surroundings for that unit.
l 1.2.1.2.1 Meteorology Section 2.3 of CESSAR-DC lists, for plant radiological evaluation purposes, the short-term (accident) and long-term (rouhine) diffusion estimates (X/Q). Other meteorological design bases are listed in Table 2.0-1. Section 2.4 of the site-specific SAR will include data to show compliance with the design bases.
l \ Hydrology 1.2.1.2.2 Hydrological design bases are listed in Table 2.0-1. Section 2.4 of the site-specific SAR will include data to show compliance with the design bases.
1.:.1.2.3 Geology and Seismology l
The design of safety-related structures, 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 seicaic envelope, i
1.2.1.3 .F_lan_t Independence j 1
The System 80+ Standard Design can be used at either single-plant j or multiple-plant sites. At multiple-plant sites, the independence 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 l The layout of the System 80+ Standard Design buildings is shown in Figure 1.2-10.
Amendment E 1.2-1 December 30, 1988
CESSAR Em%mou O
1.2.2 7YSTEM 80+ STANDARD DESIGN - SCOPE AND DESCRIPTION ,
l The design scope of the System 80+ Standard Design includes all !
buildings, structures, systems, and components which can E significantly affect safe operation. The primary design characteristics are summarized in the subsections below. The seismic category, safety classification, and quality classification of mechanical components are listed in Table 3.2-1.
l 1.2.3 NUCLEAR STEAM SUPPLY SYSTEM (NSSS)
The NSSS generates approximately 3817 Mwt, producing saturated steam.
The NSSS contains two independent primary coolant loops, each of which has two reactor coolant pumps, a steam generator, a 42-inch ID outlet (hot) pipe and two 30-inch ID inlet (cold) pipes. In addition, the safety injection lines are connected directly to A -
the Reactor Vessel. An electrically heated pressurizer is connected to one of the loops of the NSSS. The pressurizer has an iricrea sed volume to enhance transient response, pressurized water is circulated by means of electric-motor-driven, l E i i
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, i
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 j grids limit lateral motion of the tubes. The control element I asscablies (CEAs) consist of NiCrFe alloy clad boron carbide absorber rods, or hafnium full strength absorber rods and solid 3 NiCrFe alloy reduced strength absorber rods, which are guided by A j tubes located within the fuel assembly. The core consists of 241 l fuel assemblics which will be initially loaded with three difforent U-235 enrichments. The NSSS full thermal output is 3817 MWt with a core thermal output of 3800 MWt.
Design criteria are established to ensure the following:
A. The minimum departure from nucleate boiling ratio during normal operation and anticipated operational occurrences will provide at least a 95% probability with 95% confidence p' that departure from nucleate boiling does not occur.
Amendment E 1.2-2 December 30, 1988
i CESSAR En#ICATION n
V' 14 The maximum 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 UO is not reached during normal operation and anticipated I operational occurrences.
C. Fuel rod clad in designed to maintain cladding integrity throughout fuel life. l l
D. Each reactor system is designed so that any xenon transients !
will be adequately damped. j l
E. The Reactor Coolant System is designed and constructed to {
maintain its integrity throughout the expected plant life.
F. Power excursions that could result from any credible ;
reactivity addition incident do not cause damage either by l deformation or rupture of the pressure vessel, or impair l operation of the engineered safety features.
G. The combined response of tne fuel temperature coefficient, the moderator temperature coefficient, the moderator void (q
j coefficient, and the moderator pressure coefficient to an increase .i n 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 scructures include the core support barrel, the lower support tructure & ICI 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 structure 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.
l The principal design bases for the reactor internals are to provide the vertical supports and horizontal restraints during i f.s all normal operating, upset, and faulted conditions.
I (V) 1.2-3
CESSAR naricarios O,
The core is supported and restrained during normal operation and i postulated accidents to ensure that coolant can be supplied to the coolant channels for heat removal.
Reactor internals are further discussed in Sections 3.9 and 4.5. !
l 1.2.3.3 Reactor Coolant System _{RCS1 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) pipe, one steam generator, two 30-inch ID inlet (cold) )
pipes, and two pumps. An electrically heated pressurizer is '
l' connected to one of the loops of the NSSS which, for System 80+,
has an increased volume to enhance transient response. l The RCS operates at a nominal pressure of 2250 psia. The reactor j l coolant enters near the top of the reactor vessel, then flows l downward between the roc , tor vessel shell and the core barrel, up through the core, leaves the reactor vessel, and flows through ,
the tube side of the two vertical U-tube (with an integral l l economizer) steam generators where heat is transferred to the {
l secondary system. Reactor coolant pumps return the reactor j coolant to the reactor vessel. j j
Two steam generators, using heat generated by the reactor core -
and carried by the primary coolant to 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 l
the tube side and secondary coolant on the shell sidci. Each unit I is designed to transfer heat from the Reactor Coolant System to !
l the secondary system to produce saturated steam when provided I with the proper input feedwater. Moisture separators and steam driers on the shell side of the steam generator limit the moisture content of the steam during normal operation at full power. An integral flow rectrictor has been designed into each steam generator steam nozzle to restrict flow in the event of a steam line break l EE The System 804 steam generator incorporates several designr enhancements including better steam dryers, increased overall heat transter area and slightly reduced full power steam a pressure. A 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 imoroving operational flexibility. Finally, the System 80+ steam generator design has a greater tube plugging allowance therefore 2 permitting the NSSS to maintain rated output with a significant number of tubes plugged.
Amendment E l 1.2-4 December 30, 1988
CESSAR 8lnbia
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The RCS is further discussed in Chapter 5.
1.2.4 ENGINEERED SAFETY FEATURES Engineered safety features function in the highly unlikely event of an accidental release of radioactive fission products from the reactor system, particularly as the result of loss-of-coolant-accidents. These safeguards function to localize, control, mitigate, or terminate such accidents to hold exposure levels below the limits of 10 CFR 100.
1.2.4.1 Containmen_t. Structur,e General arrangements for the containment and shield building are shown in Figure 1.2-1 through 1.2-9. The containment vessel is a g 200-foot diameter spherical stcol sheel with a wall thickness of I approximately one and three-quarter inches. This containment shell is supported by, but not anchored to, a spherical depression in an intermediate floor of the shield building. The shield building is a reinforced concrete cylindrical building with a hemi-spherical dome which totally encloses the containment. The outer periphery of the containment support O floor is at plant elevation 91.75 feet, which is at the same level as plant grade. Space below the containment and inside the shield building will be occupied by Engineered Safety Features equipment, e.g., emergency core cooling system equipment, containment spray system equipment, shutdown . cooling system equipment, and 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.
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 maximum containment leakage rate is defined in conjunction with performance requirements placed on the Engineered Safety Features (ESF) systems.
The capability of the containment structure to maintain design leaktight integrity and to provide a predictable environment for operation of ESF syster.s is ensured by a comprehensive design, analysis, and testing program that includes consjderatics. of:
A. The peak containment pressure and temperature aseociated with the most severe postulated accident coincident with the Operating Basis Earthquake or Safe Shutdown Earthquake.
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B. The maximum external pressure loading condition to which the containment may be subjected as a result of inadvertent p'.
containment systems operations that potentially reduce containment internal pressure below outside atmospheric pressure.
1.2.4.2 BAf ety__ Injection System In the highly unlikely event of a loss-of-coolant-accident, the Safety Injection System (SIS) for the System 80+ Standard Design, injects borated water into the Reactor Coolant System. Thee System 804 SIS incorporates a four-train safety injection 'g configuration and an In-Containment Refueling Water Storage Tank (IRWST).
The System 80+ SIS utilizes four safety injection pumps to inject borated water into the Reactor Vessel. In addition, four safety injection tanks are provided. The SI pumps are normally aligned to the IRWST and a realignment for recirculation following a LOCA is not required. This system provides cooling to limit core damage and firsion product release and ensures adequate shutdown margin.
The SIS also provides continuour. long-term, post-accident cooling of the core by recirculation of borated water from the IRWST.
The SIS is discussed further in Section 6.3.
1.2.4.3 Km.orgency_Feedwater System l'
The Emergency Feedwater System (EFWS) for the System 804 Standard Design is a dedicated safety system that is designed to perform the following functions:
A. Supply feedwater to the steam generators for the removal of 4 heat from the RCS in the event the main feedwater system is unavailable following a transient or accident.
B. Supply feedwater to the steam generators for the removal of I heat from the RCS in the event of a total loss of AC power g (station blackout).
The EFWS is comprised 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. A The EFWS is discussed further in Section 10.4.9.
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1.2.4.4 Safgty_Depressurization System The Safety Depressurization System (SDS) is a dedicated safety lg system designed to perform the following functions: 2 A. Provide a safety grade means to depressurize the RCS in the i event that pressurizer 3 pray is unavailable during plant cooldown to cold shutdown. E B. Provide a capability to rapidly diepressurize the RCS to initiate the feed and bleed method of plant cooldown subsequent to a total loss of feadwater.
The system includes the valves and piping which establishes a flow path f rom the pressurizer steam space to the In-Containment 3 Refueling Water Storage Tank (IEWST). It is manually actuated and controlled.
The SDS is discussed further in Section 6.7. l 1.2.4.5 Containment Spray J y_s_ tem The containment Spray System (CSS) for System 80+ is an independent safety system. It is designed to maintain containment pressure and temperature within 2csign linits in the unlikely event of design ba s,e s mass-energy re? eases to the containment atmosphere. A 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 over the cross sectional area of the containment. The In-Containment Refueling Water Storage Tan'. (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 System (SCS) operation. Likewise, the SCS pumps can be manually aligned toi perform the containment spray function.
Containment Spray System also provides containment air cleanup '
function. The Containment Spray System is discussed further in Section 6.2.
1.2.5 INSTRUMENTATION AND CONTROL The System 80+ Standard Design instrumentation and control systems are summarized below. These g stems are integrated with all other systems in the Nuplex 80+ Advanced s,ontrol Complex O
1 (ACC), described in Section 1.2.6.
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O Autoiaatic protection systems, cor' trol systems, and interlocks are provided, along with the administrative controls of the specific site, to assure safe operation of the plant. Sufficient instrumentation and controls are supplic: to provide manual operat.lon as a normal backup control mode on all automatic systems.
1.2.5.1 Prot _eg_tioL _ Controls and Instrumentation Jyptems A Plant Prottetion System (PPS) initiates a reactor trip if the reactor approaches prc;cribed safety limits, or provides an actuation signal to the Engineered Safety Feature syctems when a fluid system or containment para:reter apprcaches a prescribed limit.
Surficient redundancy is installed to permit periodic testlng 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 is 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 not 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 cuncontration, and the plant operating procedures.
Four independent channels of the 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 or a given parameter reach the preset limit. Should this occur, the power supplied to the Control Element Urive Mechanisms (CEDM) is interrupted, releasing tne Control Element Assemblies (CEAs) j which drop into the core to shutdowa the reactor. The l two out-of-four Mgic can be converted to two-out-of-three logic l to allow one channel to be bypassed for testing, maintenance or operation. The protection system is maintained independent of ,
and r-eparate from the manual and automatic control systems by the g J use of optical isolation and signal validation logic deceribed in j Chapter 7 Amendment E 1.2-8 December 30, 1988
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1.2.5.1.2 Alternate Protection System j l
The Alternate Protection System (APS) augments plant protection by generating an Alternate Reactor Trip Signal (ARTS) and hlternate Feedwater Actuation Signal (AFAS) that are separate and diverse from the Plant Protection System. This system is provided to address 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.
The ARTS will initiate a reactor trip when Pressurizer Pressure exceeds a predetermined value. Its sensors and circuitry are diverse from that of the RPS. The ARTS design uses a two-out-of-two logic to open the "EDM motor generator output l contactors.
The AFAS will initiate emergency feedwater when the levels in either Steam Generator decrease below a prodotermined value. Its sensors and circuitry are diverse from that of the PPS Emergency Feedwater Actuation System and Reactor Protective System.
1.2.5.1.3 Engineered 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) systems. Again, it has a l l selective two-out-of- four actuation logic that can be converted l to a selective two-out-of-three logic. The ESFAS is completely independent of the control systems.
1.2.5.1.4 Reactor Control Systems l The reactor control systens are used for startup and shutdown of the reactor, and for adjustment of the reactor power in response to turbine load demand. The NSSS control systems are capable of following ramp load changes between 15% and 100% of full power at a rate of 5% per minute and a step change of 10%, except as limited by Xenon. This control is normally accomplished by autematic movement of CEAs in response to a change in reactor coolant temperature, with manual control capable of overriding the autonatic signal at any time. If the reactor coolant l temperature is different from a programmed value, the CEAs are I 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, f
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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 boron concentration A 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. System 80+ provides additional reduced strength CEAs 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 is a Nuplex 80+ system thati automatically controls the response 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 contral system is used to dump steam in case of a large mismatch between the power being produced by the reactor and the power being usei by the turbine.
This allows the reactor to remain at power ins *ead of tripping.
Each steam generator's water level is maintained by a feedwater control nystem. A reactor power cutback system As used to drop selected CEAs into the core to red"ce reactor power rapidly during a loss of a feed pump or a arge loss of load. This allows the SBCS and FWCS tu maintc..n 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 detestors. Eight channels of ex-core instrumentation monitor the power. Two channels are provided for the startup, two channels are provided for power control, and four channels are provided for the protection channels. The control channels are used to control the reactor power during power operations.
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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 in-core instrumentation consists of self-powered detectors, distributed throughout the core, which provide information on flux distribution within the core.
0 1.2.5.1.6 Process Monitoring Systems Ternpe ra tu re , pressure, flow and liquid level monitoring are l provided as required to keep the operating personnel informed as
! to plant operating conditions. Protection channels will indicate i l 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 prcvided in Chapter 11. g 1.2.6 NUCLEAR PLANT CONTROL G:NTER l
p The System 80+ Standard Design includes the Nuplex 80+ Advanced g h Control Complex (ACC) to ensure a cornpletely integrated design, including human factors engineering. The ACC is subdivided into functional units which complement but do not interfere with each E
l other. These functional units include:
A. The Main Control Room - where plant control is performed. I B. Data Processing Center - where plant logs and hard copyl !
results are generated and computer programming is g accomplished.
C. Interfaces to the on-Site Technical Support Center and the Emergency Operations Facility.
The Advanced Control Complex design consists of the following major interdependent systems:
A. Main Control Panels.
B. Remote Shutdown Panel.
C. Discrete Indication and Alarm System. E D. Data Processing System.
fA)'
E. Component control System and all the systems described in Section 1.2.5.1 which were in the previous previously System 80 design.
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1.2.6.1 Main Control Panels The main control panels are designed to permit command by a i single individual during normal power operation. However, the j main control room design accommodates two control room operators 3 l
and a supervisor for all normal modes of plant operation and up {
to the full operating crew during emergencies. l Each main control panel section integrates, in a human engineered fashion, miniaturized back lighted component control switches, i meters, alarms, indicators and Video Display Units (e.g., CRTs, Plasma Displays) such that both safety-related Class 1E and y' non-Class 1E instrumentation are routinely used by the operator.
Discrete alarms and indicators are provided to allow accident and technical specification monitoring, safe shutdown and other licensing requirements for which the Data Processing System VDUs B described in 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 i is unavailable. l 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 i requirements specified in NUREG-0700. Refer to Chapter 13, Human Factors Engineering, for further information.
A Control Room ' upe rvisor 's Monitoring Console, including a DPS E driven VDU and sufficient desk space, in provided to support the ,
plant monitorir.g and daily operational ds of the Control Room i Supervisor.
B 1.2.6.2 Remote Shutdown Panels I
The Remote Shutdown Panel (RSP) design includes two isolated redundant channels of the safety-related instrumentation and j controls necessary to achieve hot standby (mode 3 plant E conditions) if the main control room must be evacuated.
A Video Display Unit (VDU) monitor is provided at the Remotel Shutdown Panel to provide operational display pages. This VDU is .
B the same as the control room VDUs and is provided fer convenience.
Local controls, RSP controls and instrumentation are l>rovided to bring the plant to cold shutdown conditions utilizing suitable;j procedures. )
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t 1.2.6.3 Alarm and Display Systems The alarm and display systems are designed to aid the operator in handling any challenges to critical plant availability or safety i functions. !
l The design integrates the information displayed from alarm windows, meters and VDUs such that the same instruments used for i accident monitoring are used for normal plant operations to {
cnable operators to use instruments with which they are most I familiar during accident situations.
I The advanced control panels include displays and alarms which allow monitoring of the following critical safety functions:
A. Reactivity Control D. RCS Inventory Control C. RCS Pressure Control D. Core Heat Removal E. RCS Heat Removal F. Containment Integrity G. Plant Radiation Emissions The set of VDU panel displays include human engineered pictorial mimic and alarm information that provides the operator a continuous real-time high level overview of the entire power plant's steam and electric conversion 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 wans of providing the operator information necessary to keep the plant operating and for monitoring during accident cor.ditions.
The alarms 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 p_ata Processing SJstem The Data Processing System (DPS) is a redundant computer based
' system which provides plant data and status information to the E operations staff. The DPS monitors the NPM and Balance of Plant Amendment E l 1.2-13 December 30, 1988 ;
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O (BOP) steam and electrical production processes. It provides the' plant operations staff the ability to obtain detailed process data via CRT information output devices.
The major functions performed by the DPS include plant wide dat3 acquisition via dedicated data links to other plant systems, validation of sensed parameters, execution of 11SSS application programs and BOP 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 CRTs, interfaced with touch-screen devices, and high speed printers are E
utilized to present the plant information to the operator.
CRT display formats incorporate Human Factor Engineering design principles which permit rapid operator comprehension of the information necessary to allow the operator to monitor, control or diagnose plant conditions. All displays are organized in a f multi-tier hierarchical structure. Touch-screen display access l mechanisms are designed to allow for easc-of-access to any display page within the hierarchy.
I 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 with modern, high speed multi-processor main frame computers which are provided in a fully redundant conf 2guration. 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 DPG component failure on plant operations.
1.2.6.5 Co_mponent Control Syst;er B
The Component Control System (CCS) is designed to control discrete state components such as pumps, valves, heaters and fans within the IJPM and BOP plant systems.
The CCS consists of the ESF-CCS and Process-CCS aim emblies to y' provide control for the different channels of Class 1E equipment, as well as non-Class 1E equipment. Although they perform different plant control functions, the CCS Clacs 1E and non-Class B 1E Assemblics utilize the same electronic and mechanical components.
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A. Component Control Logic The Component Control Logic (CCL) are microprocessor based j controllers that nonitor various digital inputs, such as manual on-off demands from the main control panel, inte rlocks ., ESFAS, Diesel sequence signals, automatic j control signals, and, as programmed, produces digital output j sfgnals to control the component (start-stop; on-of f) . The CCL generates outputs for status indication on the main control panel and plant Data Processing System.
l B. Engineered Safety Features Logic l The CCS accepts ESP Initiation signals from the Plant Protection System, The ESF Logic is used to activate the a plant's Engineered Safety Features Systems components. The ]
DTgi neered Safety Feature Actuatica System (ESFAS) Icgic )
includes modules for ESF actuation and an ESF tast '
controller. Diesel Load Sequencing (DLS) logic is included I in the design.
C. Main Control Room and Eemote Shutdown Panel Interface ,
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 i panel controls and/or the Remote Shutdown Panel may then be-used effectively to achieve an orderly, unobstructed plant rhutdown. The Remote Shutdown Panel provides control for I,'
all required hot standby components. Local CCS front panel )
controls are provided for all components, allowing complete cold shutdown from a central 1(cation outside the control room. 3 1.2.7 ELECTRICAL SYSTEM An offsite power system and an onsite power system are provided to supply the unit auxiliaries during normal operation and the Reactor Protection System and Engineered Safety Feature Systems F'
during abnormal and accident conditions.
The typical transmission grid system may consist of !
interconnected hydro plants, fossil-fueled plants, combustion turbine units, and nuclear plants supplying energy to the service area at various voltages.
The turbine-generator unit is connected to a switchyard and thereby to the transmission system via two separate and v independent transmission lines. The generator circuit breaker, along with the unit step-up trans f ormerc , allows one of these Amendment E 1.2-15 December 30, 1988
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l linen not only to supply power to the transmission system during normal operation, M also to serve as an immediate available source of preferred power. The other separate transmission line is connected, via the switchyard and a standby auxiliary transformer, to provide an indepcmdent second immediate source of of fsi te power tr> the onsite power distribution system for safety and permanent non-safety loads. )!
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, main unit transformers, the unit auxiliary transformers, standby auxiliary transformer, the di6sel generators, an alternate AC source, the batteries, and the auxiliary power system. Under normal operating conditions, the main generator supplies power through isolated phase bus and generator circuit breaker to the unit main step-up and unit-auxiliary transformers. The unit auxiliary transformers are connected to the bus between the generator circuit breaker and the unit main transformers. 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 auxiliarj power is suppl.idd from the transmission system through the unit main step-up and unit auxiliaty power transformers.
A description of the onsite power system is provided in Section E 8.3.
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 unitc The Steam and Power Conversion System utilizes a condensing cycle l vith r generative feedwater heating. Turbine exhaust steam is j condensed in a conventional surface type condenser. The condensate from the steam is returned to the steam generators through the condensate feedwater syntem.
A Turbine Bypass System capable of relieving 55% of full load main steam flow is provided to dissipate heat from Lhe Reactor Coolant System during turbine and/or reactor trip. This system consists of eight turbine bypass valves to limit pressure rise in the steam generators following cessation of fl:ow 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 i isolation volves to provide the capability to hold the plant at hot standby or, in the event of less of power to the condenser circulating water pumps, cool the plant down to the point at which the shutdown cooling system may be utilized. These valves are not part of the Turbine. Bypass System; no credit for their use is assumed in obtaining the 55% capacity of the Turbine Bypass System.
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.
Mcdulation 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 opening 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 val.ves, a main steam isolation valve, and a power operated l atmospheric dump valve. Additionally, one of the two lifnes E I utilizes a bypass line and valve arodnd the respective main steam isolation valve. Each main steam line is provided with a turbine 1 stop valve awl a control valve just upstream of the high pressure turbine.
The Steam and Power Conversion System is described further in Chapter 10.
l
- 1. 7. 9 HEATING, VENTILATING, F.ND AIR CONDITIO,NING SYSTEMS The HVAC systems for the System 80+ Standard Design are described )
in Sections 9.4.1 to 9.4.8.
1.2.10 FUEL HANDLING AND STORAGE 1.2.10.1 Fuel Handling Fuel handling equipment provides for the safe handling of fuel assemblies and CEAs under all specified conditions and for the i
required assembly, disassembly, and storage of reactor vessel I head and internals during refueling.
The major components of the syster are the refuel.ing machine, the cEA change platform, tne 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 l
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l CESSAREnnncua l O1 shipping and receiving areas during core loeding and refueling operations. Fuel is inserted and removed from the core using the refueling nach3ne. During normal operations, irradiated fuel and CEAs are always maintained in a water environment.
The principal design criteria specify the following:
A. Fuel is inserted, removed, and transported in a safe manner.
I B. Subcriticality is maintained in all operations.
Fuel handling is further discussed in Section 9.1.4, 1.2.10.2 Fuel Stora_ge k The new fuel and spent fuel storage facilities are described in Sections 9.1.3 and 9.1.2, respectively. Also included in these sections are summaries cf the criticality and safety analysis.
E 1.2.11 RUXILIARY SYSTEMS 1.2.11.1 Shutdown Coolinct 81 stem The Shutdown Cooling System (SCS) is used to reduce the temperature of the reactor coolant at a centrolled rate from 350*F to a refueling temperature of approximately 140'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 heat exchangers.
The SCS for System 80+ has a nominal design pressure of 900 psig. ;
This higher system pressure provides for greater operational flexibility and simplifies concerns with system overpressurization. The SCS pumps do not share functions with the SIS.
The SCS is further discussed in Section 5.4.7.
1.2.11.2 Chemical und Volume Control Sys.t_e_m The Chemical and Volume Control System (CVCS) controls the purity, volume, and boric acid content of the reactor coolant.
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 cooled in the regenerative heat exchanger. From there, the coolant flows to the letdown heat exchanger and then through a filter and a Amendment E 1.2-18 December 30, 1988
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domineralizer where corrosion and fission products are removed.
It is then sprayed into the volume control tank and returned by the charging pumps to the regenerative heat exchanger where it is heated prior to return to the RCS.
The Chemical and Volume Control System automatically adjusts the anount of reactor coolant in order to maintain a programmed Jevel in the pressurizer. The level program partially compensate 9 for changes in specific volume due to coolant temperature changus and reactor coolant pump controlled seal Icakage. (See Section .
9.3.4.2 for details.)
i The CVCS controls the boric acid concentration in the coolant by
- a " feed and bleed" method where the purified letdown stream is I diverted to a boron recovery section rnd either concentrated '
boric acid or demineralized water is sent to the charging pumps.
The diverted coolant stream is processed by ion exchange and dogasification and flows to a concentrator. The concentrator bottoms are cent to the in-containment refueling water 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.
Tan chnmie.n1 ' '^ d V;1m.c Control System (CVCS) for System 80+
incorporates several significant imprwements and simplificat.4ons A inc'uding the following:
I A. Reclassification as a non-safety grade system by l transferring of previously credited accident mitigation and g i safe shutdown functions to other dedicated safety systems.
B. Downgrading of all CVCS components outside containment lE l (except containment isolation valves) to a non-nuclear l safety class.
C. Improved letdown configuration.
l D. Improved charging configuration.
l l Transferring of accident mitigation and safe shutdown functions to other dedicated safety systems has permitted an overall simplj fication of plant systems. This has permitted elimination l
of redundancy for the purpose of providing single tailure i protection as well as simplification of power and qualification requirements. In addition, all components in the system (with A the exception of the containment penetrations and containment isolation valves) can be designed, purchased, and constructed to non-nuclear safety criteria. Alth? ugh not a safety grade system, clic System 80+ CVCS provides reliable makeup and depessurization capabilities for defense in depth and ease of operation.
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l Systcm 80+ employo an improved letdown configuration of which key elements are:
A. Full pressure letdown heat exchanger.
A B. presnre reduction to CVCS operating pressures downstream of the letdown heat exchanger by use of a letdown flow control valve in series with a letdown orifice. !
In addition, the charging flow is controlled by the use of centrifugal charging pumps and a charging. pump flos control or ;
throttle valve on the discharge of tne pumps. i f
1.2.11.3 Process Sampling System The 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 sample syst.em 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 E suspended impurities which can cause corrosion damage to secondary system equipment. The CCS also re:aoves radioisotopes which might enter the system in the event of a primary to secondary steam generator tube leak. The condensate polishing domineralizers will also be used to remove impurities which could 1 enter the system due to a condenser circulating water tube leak.
A doncription of this system is provided in Section 10.4.6.
1.2.11.5 S_t_eam Generator Blowdown Syst.e_m l l
The design bases for the Steam Generator Blowdown Systen are:
A. Maintain proper steam generator shell side water chemis try -
by removing non-volatile materials due to condenner tube lonks, primary to secondary tube leaks, and corrosion that vould otherwise become more concentrated in the shell side e
of the steam generators. i i
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- 8. Process steam generator blowdown for reuse as condensate.
C. Enable blowdown concurrent with steam generator tube leak (s) ,
or radioactivity present on the secondary side without release of radioactivity to the environment.
D. Process a continuous steam generator blowdoim ' rate of eithe'r 0.2% or 1% of the full power main steam flow.
E. Continuously sample the radioactivity.of the steam generator; I I
blowdown-.
F. Isolate the blowdown lines lewing the containment upon a i Containment Isolation Signal, Main Steam Isolation Signal, or Emergency Feedsator 76ctuation' Signal.
Each steam generator is equipped with its own blowdown line with i the capability of blowing down the hot leg and/or the economit:er '
l regions of the steam generator shell side. The blowdown will be
! 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 whEre thG majDr portion of the O, suspended solids are removed. After filtration, the blowriewn fluid is processed by blowdown domineralizers and returned to the l e condenser.
A description of this section is provided in Section 10.4.8. ,
l 1.2.11.6 Condensate and Feedwater Systems l
The Cundensate and Feedwater Systems are designed to return condensato from the condenser hotwell s to the steam generatorre.
In addition, the systems include a number of stages of regenere,tive feed and condensate heating and provisions for maintaining feedwatsr quality.
l The entire Condensate System is non-safety-related. The portions' .j of the Feedwater Syrtua that are required to mitigate the l
consequences of on accident and allow safe shutdown of the reactor are. safety-related.
A description of these systems is provided in Section 1-0.4,7.
1.2-.11.7 Cgnpressed Air Systems The Compressed Air System consists of the Instrument Air, Station Air, and Breathing Air Systems. The Instrument Air System supplies clean, oil free, dried air to all air operated 4;
1 Amendment E 1.2-21 December 30, 1988
l CESSAi?na b ou i 1
~ _- -
l 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 clean, oil free, low pressure air to various locations-in the plant, as required for breathing protection against ;
I airborne contamination while performing certain maintenance and '
cleaning operatior.s.
A description is presented in Section 9.3.1.
1.2.11.8 Kqujipment and Flo.or Drainage System The Equipment and Floor Drninage System provides the means by which wastes arn appropriately segregated and transported to the Liquid Radioactive Waste Procersing System (LRWPS) in order to l mininize the liquid and gaseous radioactive releases. This i system accomplishes this function in a manner that is concistent' ,
with normal plant operating procedures. j 1
The d rains and sumps in the Turtine Building are not normally processed by the LRWPS. Sumps in Service Buildings are separated from Turbine Building sumps and are never processed in this j systeni, j l
A description is provided in Section 9.3.3. ]
E I 1.2.11.9 Fiye Prole _ct_i_on t System !
The Fire Protection System minimizes the risks and consequences j of fires. The functions provided by the Fire Protection System )
include the following; j J
A. Prong ictection and alarm of fires. )
1 B. Ouick suppression of fires.
C. Provention of the spread of fires.
)
D. Assurance of the capability to achieve safe shutdown in the event of fires. j E. Minimi za t i c.a of radioactive exposure and the apread of contamination aa 'he rescit of fires.
F. Provision o? manual backup to automatic fire suppression syrtems.
A description of this system is presented in Section 9.5.1. ,
l 1
Amendment E 1.2-22 December 30, 198C l I I L .- .
i
CESSAR naincamu 1.2.11.10 Communication Systems The communication systems are designed to provide effective communications betwoon all areas of the plant and plent site inc]uding all vital areas of the plant. In addition, the communication systems are designed to provide an effective means to communicate to riant persennel and offsite utility and regulatcry officials during normal conditions and a'onormal/
accident cunditions such as fire, accident, and plant testing.
The Private Automatic Business Exchange (PABX) telephone syctem i cnd the Public Address (PA) system are designed to provide diverse means of communications to all critical areas of the plant during norma] and abnormal / accident conditions.
Additionally, cound-powcred 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 Li_gh ti ng_ Sy_slemy E The lighting systems are uesigned to provide adequate and effective illumination throughout the plant and plant site including all vital areas of the plant.
The luminaries are of a praven design with long life and low maintenance requirements, such as fluorescent, metal-halide, high pressure sodium lanps. Incandescent luminaries are generally only used in cases of infrequent operation. Fluorescent; luniinaries are normally used in the following cases:
A. In plant stairs and stair wells.
B. Around switchgear, motor control centers and instrumentation TLIC k S .
C. % supplement high intensity d'.senurgo (HID) luminaries in erder to provide partial illumination in areas where starting tiues (or restarting following a momentary loss of power) of HID luminaries is objectionable.
i The system design is based on the use of standard materials. The j 9 use of "special" or " custom" made fixtures or raterials is
]
I Amendment E 1.2-23 December 30, 1988 5% . .. - -______
C E S S A R n aincuum 9:
restricted to cases where the use of standard materials is demonstrated impractical.
Personnel discomfort from lighting, e.g., glare, is minimized by coordinating the design features of the lighting system with the characteristics of illuminated objects.
The lighting system components are selected to minimize the potential for danger to personnel or damage to equipment. In particular, the potential for and consequences of lamp breakage j i
are evaluated.
I Each lighting panel is provided with a main cf.rcuit breaker with spare switching capability of at least 40% to support the possible expansion of the panel's loads. -l The 1ighting panels are located in areas that are easily accessible for installation, maintenance, testing, and operation.
Similarly, the lighting fixtures are designed and located so that maintenance and relamping can be accomplished efficiently and j safely.
Provisions are made to allow the removal and reinstallation of lighting equipment in order to support room, space, or area modifications.
E The design of the plant lighting systems is in accordance with applicable industry standards for illumination fixtures, cables, g'. J u n d i n g , penetrations, conduit, controls. I The nu. mal station lighting system is used to provide normal illumination under all plant operation, maintenance and test I
conditions.
l The security lighting system provides the illumination required '
to monitor isolation zones and all outdoor areas within the plant protected perimeter, under normal conditions as well as upon loss of all AC power. The security lighting system complies with the intent of NUREG CR-1327 The emergency lightina 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 Amendment E 1.2-24 December 39, 1988
1 l
CESSARan h., 4 i
l o i 1.2.11.12 DJesel _ Generator EnAJte i Fuel Oil System l
The Diesel Generator Engine Fuel Oil System is designed to provide for storage of a coven-day supply of fuel oil for.cach 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, and to withstand the offects of natural phenomena without the loss of operability.
l A description of this system is presentnd in Section 9.5.4. .
1.2.11.13 Diesel Generator Engine Cooling Water System
)
l 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 conibust ion air aftercoolers, and the governor lube oil cooler.
A description of this system is presented in Section 9.5.5. )
l 1.2.11.14 Diesel Generator EncLi_ne Starting Air System E Tne Diesel Generator Engine Starting Air System is designed to provide fast 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 or this system is presented in Section 9.5.6.
1.2.11.15 Diesel Generator Engine Lube Oil System l
l The Diesel Generator Engine Lube Oil System is designed to I deliver clean lubricating oil to 'ho diesel generator engine, its bearings and crankshaft, and c' ar 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 i load-accepting capability. The system also provides a means by which vaed 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.
r~
{x Amendment E 1.2-25 December 30, 1988
~
CESSAR nMiricarios i I
e!
1.2.11.16 p_Jesel Generator Engi3e Air Intake and Exhaust S_ystem The Diesel Generator Engine Air Intake and Exhaust System is designed to supply clean air for combustion to the diesel generator engine and to dispose of the engines exhaust, The system is housed in a building designed to withstand the effects of natural phenomena and credible missiles. ,
i A description of this system is presented in Section 9.5.3. f i
1.2.11.27 Diesel __ Generator BuiJdina Sump Pump System The Diesel Generator Building Sump Pump System is designed to remove leahage and equipment drainage from the Diesel Generator Euilding and to protect the diesel generator units from flooding caused by a major pipe rupture.
A description of this system is provided in Section 9.5.9.
1.2.11.18 Compr_essed Gas Syyt ems i
The compressed gas supply sy; tems are provided to supply various gases for equipment and instrumentation cooling, purging, diluting, inerting, and welding. The major items of equipment E 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 _Sy_ stems The potable and sanitary water systems process water for general plant use. These systems serve no safety functions and any I malfunction has no adverse effect on any safety-related system. {
The requirements of General Design Criterion ') are met as I related to design provisions provided to control tha r91 ease of liquid offluents containing radioactive material from contaminating the PSWS.
':' hose systcas are described in Section 9.2.4.
1.2.11.20 Damineralind Water Makeup _S_ys_tp s The Demineralized Water Makeup System supplies filtered demineralized water to the ConJensate Storage System for makeup j and to other systemt throughout the plant that require high i Amendment E 1,2-26 December 30, 1988
CESSAR 8Hi% mow V(%
quality, non-safety-related, makeup water. This system, therefore, serves no safe shutdown or accident mitigation' function, and has no safety design bases.
A description of this system is presented in 9.2.3.
1.2.12 RADIOACTIVE WASTE MANAGEMENT SYSTEM 8 Radioactive sources and waste management systems are described in Chapter 11. Design considerations to minimize exposure to radioactivity are summarized in Chapter 12.
1.2.13 PHYSICAL PLANT SECURITY AND PROTECTION FROM SABOTAGE 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 protection requirements.
1.2.14 COOLING WATER SYSTEMS 1.2.14.1 Condenser Circulating Water System !
l The Condenser Circulating Water System provides cooling water for the turbine condensers and rejects heat to the atmosphere via the cooling towers. Tower sizing is such that full unit load can be maintained with one tower out of service except during the three hottest summer months of the year. The towers are capable of cooling the circulating water to 100*F with 82*F wat bulb (Ot E exceedance). Pump head, piping size, and cooling tower height j are optimized based on capital and pumping cost. Pumps are sized such that full load can be maintained with one pump down except i during the three hottest summer months of the year.
Soo Section 10.4.5 for a description of this system.
1.2.14.2 Station Service Water System The Station Service Water System (SSWS) is an open system that takes auction from the ultimate heat sink and provides cooling water flow to remove heat released from plant systems, structures and components. The SSWS then returns the heated water to the ultimate heat sink. The SSWS cools the Component Cooling Water System which in turn cools safety-related and non-safety-related reactor auxiliary loads.
A doccription of this system is prescnted in Section 9.2.1. ,
(
\
Amendment E 1.2-27 December 30, 1988
CESSAR Eanncuum l
l i
1.2.14.3 Copgonent Cooling _ Water System l The Component Cooling Wator System (CCNS) is a closed loop cooling water system which cools components and heat exchangers ,
located in the Auxiliary, Fuel, Radwa.ste and Containment !
Buildings. Ifeat transferred by these components to the CCWG is rejected by the SSWS via the CCWS heat exchangers.
A description of this system is provided in Section 9.L.2.
1.2.14.4 Turbine _Dulldi3g Cooling Water System .
l The Turbine Building Coolirig Water System (TBCWS) provides cooling for the non-safety-related components in the various ]
turbine plant auxiliary systems. Cooling is effected through heat exchangers with heat rejected to the Turbine Building Serlice Water System (TBSWS). This closed cooling water 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 and corrosion.
A det erir tion of this system is provided .in Section 9.2.8.
E 1.2.14.5 CJ)il_le_d Water System The Chilled Water Gystems (CWS) are dauigned to provide and distribute a sufficient quantity of chilled water, through a qroup of dedicated piping systems, to air hanaling units (AHUs) in specific plant areas. The CDWS is divided into two subsystemn, an Essential Chilled Water System (ECWS) that serves primarily safety-related HVAC croling loads, and a Normal Chilled Water System (UCUS) that serves non-safety-related HVAC cooling loads.
A description of this system is provided in Section 9.2.9. {
1.2.14.6 Turbine Building _S_ervice Water __Sysleig The Turbine Building Service Water System (TBSWS) removes heat from the TBCSW and rejects the heat to the cooling towers. ;
I The TBSWS uses pumps to circulate from the plant cooling towers l to remove heat from the TBCWS. Condcnser circulating water from the cooling towers, is pumped through the TBCWS heat exchangers l and is discharged back into the Condenser Circulating Water l System at a point between cne main condenser cooling water outlet I and the cooling tower inlet.
O Amendment E 1.2-28 December 30, 1988 l l
l
CESSAR naincuin 1
O V ;
A description of this system is provided in Section 9.2.10.
1.2.15 ULTIMATE HEAT SINK The ultimate heat sink described here consists of single passive 4 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 E
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 alternate ultimate heat sinks are 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 Ultimate Heat Sink is described in Section 9.2.5.
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!.aI 1.3 C_OMPARIGON TABLEG 1.3.1 COMPARISON WTTH SIMILAR FACILITY DESIGNS This section highlights the principal design features of the System 80+ arrangement and provides a comparison of the major {
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Table 1.3-1 summarizes the design and operating characteristics for the nuclear, electrical, power, structural and safeguards systems of various facilities. Combustion Engineering reactor designs of recent years are grouped into classes which are helpful in illustrating and justifying the progression of plants from lower to current power levels. The data, performance i characteristics, and other information presented herein represent the design of reactors in each clase according to the i
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. TABLE _ L.k 2 00CKET t.ISTINGS FOR C+E RECENL_RfACTOR DESIGNS Class Specific P1;pnt Docket Listino 2570 Mwt Baltimore Gas & Elect ric Co. Docket No.59-317 and 50-318 l Calvert Cliffs Units 1 & 2 Florida Power & Light Company Docket No. 50-335 and 50-389 l St. Lucie Units 1 & 2 !
Northear.t Utilities Occket No. 50 336 I Millstone Unit 2 !
3410 Mwt Louisiana Power & Light Co. Docket NO. 50-382 Wateriord Unit 3 i
Southern California Edison Docket No. 50-361 and 50-362 l San Onofre - Units 2 & 3 l rm 3817 Mwt Washing ~ ton Public Power Occket No. STN-50-508
!j (System 80) Supply System Unit 3 Dccket Ncs. STN-50-528, ^
Arizona Nuclear Power i I Project Palo Verde Nuclear 9 N 50-529 and STN-50-530 i Generating Station i i
Units 1, 2, & 3 I Combustion Engineering Decket No,. iTN 50-4 70F Standard PWR - Nuclear ;
E i Steam Supply System L !
l.
r3 Amendment L December 30, 1988 1
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t' k
1.4 It$ NOTIFICATION OF AGENTS AND CONTRACTORS.
1.4.1 APPLICANT'S QUALIFICATIONS AND EXPERIENCE (Presented in site--specific SAR. ) A I
1.4.2 AECHITECT-ENGINEE2'S QUALIFICATIONS AND EXPERIENCE (Presented in site-specific SAR.) A 1
l 1.4.3 COMBUSTION ENGINEERING'S QUALIFICATIONS AND l EXPERIENCE Combustion Engineering, Inc. (hereafter referred to as C-E or I combustion) nuclear power activities are of three general types:
design, development, construction and operation of reactor and auxiliary systems; design and fabrication of nuchar components; and, support of design, development and analytical projects.
A susmary of the company's efforts, accomplishments, and operating experience in the light water reactor field is provided below..
( 1.4.3.1 Pre-Commercial Reactor Programs l 1.4.3.1.1 Naval Propulsion Program During the period 1955 through 1960, Combustion was a major ;
contributor to the U.S. Naval Reactors program. The Company !
designed and bu il t , at its Windsor, Connecticut site, the prototype of a omall attack submarinn power plant. This prctotype (SIC) went into operation in 1959 and is still being l operated as a na"M training facility. A second plant of this type was alse designed an.1 built by Combustion for installation in the USS Tullibee (SSN-597) which has been operated as a part of the United St.ates nuclear submarine fleet.
In t.he design, development, construction and operation of the prototype system and the st.b:aa rine power plant, Combustion's j responsibilities included all safety aspects of the reactor sy s te:as .
l 1.4.3.1.2 Doiling Nuclear Superheat (BONUS) Plant Corbuction was responsible for the nuclear design and for the d rection of startup and initial operation of the BUNU::' plant in Puerto Rico.
The design of this reactor system presented a number of unique
(# problems, e.g., control and safety analysis of a two-region core, l donign of a naperheater fuel element, design of a steam control Amendment A 1.4-1 September 11, 1987 l
L -. __ __-.:_-_-___ _ - - _ _
CESSARn!Mc- ;
O.
system to assure adequate cooling of superheater fuel under all i credible conditions, and design of a containment building of the !
" total containment" type to house the entire power generating installation.
i The BONUC plant achieved full power operation in September 1965, l and was the first nuclear power plant under USAEC control l j
operating with an integral superheating core.
1.4.3.2 Development and Design of Commercial PWR Systems The development and design by Combustion of a pressurized water reactor for utility service dates back to 1958. .At that time, l the Company was selected by the AEC to undertake the design, j analysia and economic evaluation of a 250 MWe PWR plant, in j conjunction with an architect-engineer. This effort provided (
initial technical and economic guidelines for Combusticn's commercial development of the PWR.
With a subsequent decision by the Company to concentrate on the development of the PWR for large nuclear power stations, a program was initiated to guide required deaign and development work along appropriate lines. The following is representative of the types of PWR-oriented work which have been performed:
A. Evaluation of overall plant and systems to establish optimum physical arrangement and design criteria from the standpoint of economics and safety. Much of this work has been performed in conjunction with qualified architect-engineering organizations; i
B. Design and development of nuclear components such as control assemblies, centrol assembly drive mechanisms, and auxiliary systems equipment.
C-. Extensive testing of PWR nuclea; components, such as fuel ,
assemblics and reactor control components, under actual i service pressure, temperature and flow conditions.
I Combustion Engineering's Nuclear Laboratories have been engaged in the development and testing of fuels, fuel elements, control assemblies, reactor components and materials for reactor application. Particular emphasis has been given to UO 2 and Zircaloy cladding technology, involving both in-pile and out-of pile investigations. The initial efforts in the laboratories were associated with submarine reactor programs.
Since 1960, the personnel of the nuclear laboratories have actively participated in the joint U.S. AEC - Euratom research and development program for fuels development. In addition to 1.4-2
CESSAR Enhior i 1
l O
these programs, personnel in the Nuclear Laboratories have been i responsible for materials design activities for the HWOCR study- )
,i and for pressurized water, boiling water, nuclear superheat, and fast breeder reactor systems. I l
1.4.3.3 Jiador Component Desian and Fabrication During the period of 1955-1961, Combustion Engineering (C-E) was !
a major supplier of nuclear cores for naval propulsion service, l C-E has fabricated the boiling and the superheating fuel for the BONUS reactor. The boiling section of the Bonus core was made up of Zircaloy clad, rod type, UO fuel elements fundamentally similar to those being utilized id the C-E Standard fuel design. .
The superheater fuel utilizes Inconel-clad, rod type, UO fuel !
The superheater cladding is designed for an opebating elements.
temperature of 1250*F.
Combustion Engineering has performed the design engineeritig and fabrication of control rod drive mechanisms and fuel rods for all of the commercial power reactors listed in Table 1.4-1 at its l facilities in Windsor and East Windsor, Connecticut. l A Combustion Engineering has fabricated and shipped many reactor Q vessels for utility plant service and Additional vessels for plant sizes up through 1300 Mwe are now in A for naval service.
service.
l l Combustion Engineering has fabricated nuclear steam generators !
for naval service and for all its commercial PWR plants. In i addition, the company designed and fabricated the 10 steam generators in the Hanford New Production Reactor facility. '
Combustion Engineering manufactures reactor vessel internal structures at its Newington, N.H. facility. E j Combustion Engineering extended its manufacturing capability to l include fabrication of reactor coolant pumps by its entry in 1974 into joint ownership of the CE/KSB Pump Company.
1.4.3.4 Facilities The C-E laboratories at Windsor, Connecticut, and Chattanooga, Tennessee, provide complete facilities for the development, ,
design, analysia and testing of PWR components and systems. l These laboratories include equipment for: ;
Amendment E 1.4-3 December 30, 1988 E ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _
i l CESSAR UNncuen !
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9l A. Mechanical Testing.
B. X-ray and Radiography.
C. Metallography.
D. Ceramics Development. )
1 E. Analytical and Radio-Chemistry I F. Fuel Fabrication Development G. Corrosion Testiner H. 2500 psi Component Performance Tecting I. 2500 psi and 5000 psj Steam Generation J. Welding Development ,
The Windsor facilities of Combustion Eng '...ee ring , Inc. arc equipped to fabricate, and ptovide the necessary quality control for, fuel assemblies, control assemblies, control assembly drive mechanisms, and other specialized nuclear componcnts.
C-E has in -operation a PWR power plant simulator at the Nuclear Training Facility in Windsor which simulates the Calvert Cliffs reactor. Used by C-E to train utility operator personnel, the facility provjdes simulation of the full spectrum of nuclear steam supply system plant operations, including hot and cold startup and full load operation, mar.euvers throughout the power range, equipment malfunctions, emergency conditions and recovery, and plant shutdown. In addition, C-E completed a full scope simulator for BG&E's Calvert Cliffs facility.
Combustion Engineering's Chattanooga Plant includes a separate f acility to desit;n , fabricate, and provide quality control for large reactor precaure components. The facility has such special equiprent as heavy cuty cranes and large capacity machine too~s capable of performing work on large, heavy parts to close tolerances and fine surface finishes. It is also equipped with j the latest testing and quality control equipment, including a l linear e.cceleratcr for weld examination.
1.4.3.5 Commercial Re3M or Operation )
Table 1.4-1 lists all Combustion Engineering Pressurized Water Reactors designed and built to date. A O
i Amendment A j 1,4-4 September 11, 1987 I L -. --_ A
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- CESSAR!alb .
1.5 REQUJEEMENTS FOR FURTSIER TQC_HN1 CAL INFORMATION This section, identified in Regulatory Guide 1.70, is not applicable to the System 80+ Standard Design sinca there was no PSAR in which needs for fui: Ar technical needs were identified. '.
This section does, however, describe the programs whose results were used in the development of the System 80 design, which was the starting point for the System 90+ Standard Design. Table 1.5-1 lists those programs and identifics where results are documented.
1.5.1 TOPICAL PROGRAM
SUMMARY
References 1 through 8 present descriptic,ns of safety related A Research and Develop.nent programs which were carried out by, or in conjunction with, combustion Engineering, Inc. during the development of the System BC Nuclear Steam Supply System.
1.5.2 SYSTEM 80 - 16 x 16 ASSEMBLY TEST PROGRAM A devepment schedule, compatible with the project schedules Y having iystem 80 designs, was iml:lemented such that results would be avai able before each plant van completed.
1.5.2.1 Components Testing Component test programs have been conducted in support of all C-E PWP, designs. The tests subjected a fall-scale reactor cora module comprising fuel assemblies, control element assembly, control element drive mechanism, and reactor vessel internals components to the hydraulic environment of the reactor under all norm.1 operating conditions. The progran was a continuing series of tests wnercin components introduced as part of a particular design were tested in the C-E TF-2 hot loop test facility at Windsor, Connecticut. The tests were designed to proof-test and life-test the integrated fuel assembly and control element drive mechanis~.n under a variety of simulated operating conditions to evaluate component fretting and wear characteristics, scram perfo~mance, r and fuel assembly uplift and pressure drop.
Information and a description of the testing is provided in Section 4.4.4.2.
1.5.2.2 Fuel Assembly _S_e_ismic Testing The prog ram was divided into three areas -
spacer grid tests, fuel assenb]y static tests and fuel assembly dynamic tests. The results were utilized in developing the seismic models of the fucJ described in Section J.7.3.14 and Section 4.2.
Amendment E 1.5-1 December 30, 1988
CESSAR 8lnha e.
1.5.2.3 Reactor Flow M_odel Testing k
A scale flow model of the C-E System 80 reactor vessel and internals has been tested. A detailed discussion of the flow model and test facility appears in Section 4.4.4.2. The purpose of these tests was to establish or verify design hydraulic parameters. In particular, core inlet flow distributions and {
pressure lossen along the flow path segments within the reactor i vessel have been measured for operating configurations. !
Additional information is provided in Appendix 4A. B 1.5.2.4 DNS Improvement I
A substantial test program was undertaken to verify the thermal performance capability of the System 60 fuel assembly. The test program was an extension of the experimental studies conducted with rod bundles representative of the C-E 14 x 14 fuel cssembly.
Those studies were described in more detail in Section 4.4.4.5 and were used in System 80 analyses.
1.5.2.5 Fuel Deve_1.opment Programs Combustion Engineering has participated in several fuel E irradiation programs. In addition, several cooperative fuel development programs were performed with Kraftwerk Union as partg of a technical agreement. A A complete description of the programs and the results obtained is provided in Section 4.2.
1.5.3 SYSTEM 80 STEhM GENERATOR DEVELOPMENT PROGRAMS q C-E's System 80 Steam Generator is a vertical U-tube component as E described in Section 5.4.2. Development efforts were conducted by C-E to confirm the structural integrity of the steam generator during thermal, MSLB and FWLB transients. The results of these .
programs are given in Appendices 5B and SC for the MSLB and FWLB !
conditions, respectively. !
i o!
Amendment E 1.5-2 December 30, 1988
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CESSAR nui"icaricu O ;
1 1.5.2.3 Reac.tgr Flow Model Testing A scale flow model of the C-E System 80 reactor vessel and internals has been tested. A detailed discussion of the flow model and test f acility appears in Section 4.4.4.2. The purpose of these tests was to establish or verify design hydraulic j parameters. In particular, core inlet flow distributions and I pressure losses along the flow path segments within the reactor vessel have been measured for cperating configurations.
Additional information is provided in Appendix 4A. 3 1.5.2.4 DNB Improvement l A substantial test program was undertaken to verify the thermal performance capability of the System 80 fuel assembly. The test program was an extension of the experimental studies conducted with rod bundles representative of the C-E 14 x 14 fuel assembly, Those studies were described in more detail in Section 4.4.4.5 and were used in System 80 analyses.
1.S.2.5 Fuel Development Programs Combustion Engineering has participated in several fuel E irradiation programs. In addition, several cooperative fuel development programs were performed with Kraftwerk Union as part j of a technical agreement. A j A complete description of the progra.m.s and the results obtained is provided in Scotion 4.2.
1.5.3 SYSTEM 80 STEAM GENERATOR DEVELOPMENT PROGRAMS C-E's System 80 Steam Generator is a vertical U-tube component as E described in Section 5.4.2. Development ef forts were conducted i by C-E to confirm the structural integrity of the stea'.n generator !
during thermal, MSLB and FWLB transients. The results of these programs are given in Appendices SB and SC for the MS'LB and FWLB ,
conditions, respectively. l 1
1 0'
i Amendment E 1.5-2 December 30, 1988 l
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REFERENCES FOR SECTION 1.5
- 1. " Safety Related Research'and Development for C-E Pressurized Water Reactors Program Summaries: January 1973," Combustion Engineering, Inc. , GNPD-87, March 19734
- 2. " Safety Related Research and Development for C-E Pressurized Water Reactors Program Summaries: January 1973 through February '1974," Combustion Engineering, Inc., CENPD-143, 4 June 1974. l
- 3. " Safety Related Research and Development for C-E Pressurized Water Reactors: 1974 Program Summaries," Combustion 3 Engineering, Inc., CENPD-184, May 1975. i
- 4. " Safety Related Research and Development for C-E Pressurized Water Reactors: 1975 Program Summaries," Combustion Engineering,. .Inc., CENPD-229, June 1976.
- 5. " Safety Related Research and Development for.C-E Pressurized Water Reactors: 1976 Program Gummaries," Combustion Engineering, Inc., CENPD-258, October 1977.
('~g) 1
" Safety Related Research and Development for C-E Pressurized 6.
Woter Reactors: 1977/1978 Program Summeries," Combustion E7.'ineering, Inc., CENPD-262, March 1979. H A
- 7. "Safet y Related Research and Development for C-E Pressurized Water Reactors: 1979/1980 Program Summaries,"' Combustion Engineering, Inc., CENPD-265, September 1981.
- 8. " Safety Related Research and Development for C-E Pressurized Water Reactors: 1981 Program Summaries," Combustion l
Engineering, Inc., CENPD-267, July 1982.
l l
Amendment A 1.5-3 September 11, 1987
C E S S A R En nne. m .
O TABLE 1.5-1
SUMMARY
OF DEVELOPMENT PROGRAMS TO DEMONSTRATE SYSTEM 80 DESIGN CON _SERVATISM FROGRAM RESULTS DOCUMENTED IN: IA 3
- 1. Component Tests . Appendix 48
- 2. Fuel Assembly Seismic Tests CENPD-178 Rev. 1 B
- 3. Reactor Flow Model Test Appendix 4A
- 4. DN8 1mprovement and Flow Mixing Tests CENPD-162A and CENPD-207
- 5. Fuel Densification Program Section 4.2.3.2.10 and l
CENPD-139 l
- 6. LOCA Refill Program CENPD-134
- 7. Blowdown Heat Transfer Program CENPD-132, Suop. 1, 2, & 3 1
- 8. Reflood Test CENPD-213
- 10. Steam Generator Program Appendices 5B & SC
- 11. CPC Program CEN-72A and CEN-73A O
Amendment B March 31, 1988 L____________________________________..___.__..________.____._
I CESSAR 83nricuior O
1.6 liATERI AL INCORPORA'i'ED BY REFERENCE The following list is a tabulation of all material' which is incorporated by refetence as part of this application. Other material not incorporated by reference .is listed in the individual chapter and section referenccs for information i purposes.
DATE CESSAR
-l REPORT NO. TITLE ISSUED CHAPJf8 CENPD-26 Description of loss-of-Coolant August 1971 3.
Calculational Procedures 6.
E Suppl. #1 Description of Loss-of-Coolant October'1971 Calculational Procedures Suppl. #2 Steam Venting Experiments January 1972 Suppl. #3 Moisture Carryover during a PWR January 1972 post-LOCA Core Refill CENPD-67 Combustion Engineering, Inc. September 1973 10.
Suppl . #1 " Iodine Decontamination May 19/4 1).
Suppl . #2 Factors During PWR Steam June 1974 O Addendum 1 Addendum 2 Generation and Steam Venting" November 1974 August 1975
)
CENPD-80 Moisture Carryover During an January 1973 6.
NSSS Steam Line Break Accident CENPD-98-A C0AST Code Description April 1975 5. l E
15.
CENPD-107 Combustion Engineering, Inc. April 1974 15.
"CESEC" Suppl. #1 ATWS Models Modifications September 1974 i Suppl . #1 ATWS Model Modifications November 1975 j Amend. 1-P to CESEC .j Suppl. #2 ATWS Models for Reactivity September 1974 i Feedback and Effects of :
Pressure on Fuel l Suppl. #3 ATWS Model Modification . August 1975 I suppl. #4 To CESEC December 1975 i Suppl. #5 June 1975 l Enclosure 1-P CESEC Digital Simulation of December 1981 to LD-82 001 a C-E Nuclear Steam Supply System 1
4 l
Amendment E 1.6-1 December 30, 1988 l
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DATE CESSAR I REPORLM9. TJJLE ISSUED CHAPTB CENP-J13 Test Data Sheets, Maine-yankee August 197.?
Core Crud Removal i 1
CENPD-132-i Combastion Engineering, Inc. August 1974 6.
" Calculation Methods for the '
C-E large Break LOCA i Evaluation Model" e Suppl.. #1-P February 1975 6. j Ju!y 1975 Suppl. #2-P 6.
Suppl. #3-P June 1985 6, CENPD-133 Combustion Engineering, Inc. August 1974 5.
"CEFLASH-4A Fortian IV ,
0(gital Computer Program j for Reactor Blowdown Analysis" I i i
Suppl. #1 CL' FLASH-4AS, A Computer September 1974 Program for Reactor Blowdown !
Analysis of the Small Rreak !
Loss of Coolant Accident 1; l Suppl. #2 CEFLASH-4A, A FORTRAN IV Digital March 1975 c j Computer Program for Reactor i
. Blowdown Analysis (Modifications)
Suppl. #3 February 1977 CENPD-134 Combustion Enghteering, Inc. August 1974 6.
"COMPERC-II A Program for Suppl. #1 Emergency Refill - Reflooi of February 1975 the Core" CENPD-la5-P Combustien Engineering. Inc. August 1974 4. lB l Suppi. #2 "STRIKIN-Il A Cylindrical February 1975 6. j Suppl. #4 Geometry Fuel Rod Heat . August 1976 Suppl. #5 Transfer Program" April 1977 CENPD-136-P High Temperature Properties 4 1 of Zircaloy and UO2 f r Use a ;
in LOCA Evaluat;on Models CENPD-l?7 Combustion Engineering, IN. AugMst 1974 6.
" Calculative Methods for the Guppl. #1 C-E Small break LOCA Anuary 1977 {
Evaluation Model" y CENPO-138 PARCH A FORTRAN IV August 1974 6.
Suppl. #1 Digital Computer Program to February 1975 ;
Guppl. #2 Evaluate Pool - Boiling Axial Januar 1977 q Rod, and Coolant Heatup l l
Amendment E l 1.6-2 December 30, 1988
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DATE CESSAR REPORT _NO. TITLE __.
ISSUED CHAPTER CENPD-139-P C-E fuel Evaluation Model July 1974 4.
Rev. 01 Topical _ Report Suppl. #1 CENPD-145 A Method of Analyzing In-Core April ~1975 4.
Detector Data in Power Reactor CENPD-153 Evaluation of Uncertainty in May 1980 4 ..
Rev. I the Nuclear Power Peaking Measured by the Self-Powered, Fixed In Core Detector System l CENPD 158 Analysis of Anticipated May 1976 15.
Transients Without Scram in Corc,bustion Engineering NSSS CENPD-161-P TORC Code: A Computer Code for July 1, 1975 4.
Determining the Thermal l Margin of a Reacter Core rh Q CENPD-162-P-A Combustion Engineering, Inc.
"CHF Correlation for C-E Fuel September 1976 4. ;
Suppl . #1- A Assemblies with Standard February 1977 Spacer Crids - Part 1; Uniform ,
j Axial Power Distribution" E CENPD-170 Combustion Engineering, Inc. August 1975 Suppl. #1 " Assessment af the Accuracy November 1975 7.
of the PWR Safety System l Actuation as Performed by the Core Protection Calculators' l
l l <
l I I
D l (V i i
Amendment E 1.6-3 December 30, 1988 l l
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DATE CESSAR REf_0fti NO. _, TIT 1f ._ ___JSSUE0 CHAPTER .
1 CENPD-178 Structural Analys1s of fuel August 1981 4. 3 Rev. 1 Assemblies for Seismic and ,
Loss of Coolant Accident i trading E CENPD-180 Radiolodine Behavior in the March 1976 15. )
Suppl. #1 Reactor Coolant System During March 1977 Transient Operation CEFD-182 Combustion Engineering, Inc. Novembu 1975 3. ,
" Seismic Qualification of C E j Control E.aipment" l l
CENPD-183 Combustion Engineering, Inc. July 1975 15.
)
"C-E Methods for loss of flow Analysis" (ENPD-187 P-A Combustion Engineering, Inc. March 1976 4. E "CEPAN, Method of Anal.yzing Creep Suppl. #1.A Collapse of Oval Cladding" June 1977 CENPD-188 A HERMITE, A Multi-Dimensiorial March 1976 4. B Space-Time Kinetics Code for PWR Transients CENPD-190-A rombustion Engineering, Inc. January 1976 15. B "C E Method for Control Element Assembly Ejection Analysis" CENPD-198-P Combustion Engineering, Inc. December 1975 4. 3 Suppl. #1 "Zirchloy Growth-in-Reactor Dimensional Changes in Zircaloy-4 fuel Assemblies J CfNPD-201-A Reactor Coolant Pump Performance April 1976 5.
CENPD-206-P Combustloh Engineering, Inc. January 1977 4. 3 "10RC Code Verification and Simplified Modeling Method" i
O Amendment E 1.6-4 December 30, 1988
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O i DATE CESSAR
_ REPORT NO. TITLE ISSUED CHAPTER CENPD-207-P Combustion Engineering, Inc. June 1976 4.
B
" Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids, Part 2, flon-Uniform Axial Power Distri-butions" j CENPD-210 Quality Assurance Program July 1977 17.
l Rev. 5 A Description of the C-E Nuclear January 1987 .E j
' Steam Supply System Quality September 1988 Assurance Program I CENPD-213 Combustion Engineering, Inc. January 1976 1.5 Suppl. #1 " Application of FLECHT Reflood March 1976 Heat Transfer Coefficients to lE Combustion Engineering 16 x 16 Fuel Bundles" l
CEN-214(A)-P CETOP-D Code Structure and July 1982 15.
q Modeling Methods for Arkansas )
y Nuclear One-Unit 2 j CENPD-221 Joint C-E/EPRI Fuel Performance December 1975 4. j Evaluation Program, Task C, B
{
Evaluation of Fuel Rod l Performance on Maine-Yankee l Core 1 CENPD-225-P Combustion Engineering, Inc. October 1976 4. !
" Fuel and Poison Rod Bowing" l l
Suppl. #1 February 1977 )
Suppl. #2 June 1978 Suppl. #3 July 1979 )
CENPD-254 " Post-LOCA Long Term Cooling June 1977 6.
l Evaluation Model" CENPD-255-A " Qualification of Combustion October 1985 3.
Engineering Class lE Instrumentation" CENPD-266-P-A The ROCS and DIT Computer Codes April 1983 4 ..
for Nuclear Design O CENPD-269-P Extended Burnup Operation July 1984 4.
B of Combustion Engineering PWR Fuel Amendment E 1.6-5 December 30, 1998 l
CESSAR nuiricuia O; ,
DATE CESSAR I REPORT NO. TITLE ISSUED CHAPTER j CENPD-275-P C-E Methodology for Core March 1987 4. g Designs Containing Gadolinia Urania Burnable Absorbers CESSAR PSAR Appendix 6B 6.
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Amendment B 1.6-6 March 31, 1988 l l
CESSAR88L mu !
l Ok V !
1.7 DRAWINGS AND DIAGRAMS g [
l 1.7.1 ELECTRICAL, INSTRUMENTATION, AND CONTROL DRAWINGS The systems of interest are the Reactor Protective System (RPS), B Alternate Protection System (APS), and the Engineered Safety Features Actuation System (ESFAS). These three systems provide reactor trips and Engineered Safety Feature (ESP) systems actuation for limiting events as determined by the safety analysis for the plant. W The functional block diagrams for the RPS, ESFAS, and APS are E shown in Figures 7.2-12, 7.3-3 and 7.7-12. The interface logic is shown in Figure 7.2-13. Other figures at the end of Sections j i
7.2 and 7.3 provide more detailed logic on various portions of I
these systems.
Operating logic for the control of the Shutdown Cooling System !
(SCS) are shown in Section 7.4. Operating logics for the SCS Suction Line Valve Interlocks and the Safety Injection Tank Isolation Valve Interlocks are shown in Section 7.6. I i l l Measurement Channel Block Diagrams. (MCBDs), Plant Protection B l
( System (PPS) design drawings and component Functional Logic ,
Diagrams are identified in the applicable CESSAn sections.
l l The MCBDs show all channels which are safety-related. The drawings apply to the RPS, APS, ESFAS, and to the Post-Accident B Monitoring requirements in CESSAR. j Component Logic Diagrams provide NSSS interface requirements to l the specific site. They show controls on individual components i for which C-E has system design responsibility. They describe how a component is to operate. That is, khat signals cause a component to start or stop, open or close, and what operator l controls are required.
1.7.2 PIPING AND INSTRUMENTATION DIAGRAMS Table 1.7-2 provides a list of valve identifiers which are used E on the pipina and instrumentation diagrams listed in Table 1.7-3.
Piping and instrumentation diagram symbols are presented in Figure 1.7-1.
O Amendment E 1.7-1 December 30, 1988
I CESSAREnacma TA3LLi.7-1 (G
(Sheet 1 of 4)
SAFETY RELATED ELECTRICAlu INS 1RUMENTATION A.N_D_ CONTROL DRAWINGS I
Drawing No. .. Title Section Figure 7.2-1 PPS Basic Block Diagram 7.2 Figure 7.2-2 PPS Functional interface and Testing Diagram 7.2 Figure 7.2-3 Typical PPS Low Reactor Coolant Flov: Trip 7.2 Setpoint Operation Figure 7.2-4 Typical PPS Measurement Channel functional 7.2 Diagram (Pressurizer Pressure Wfde Range)
Figure 7.2-5 Reed Switch Position Transmitter 7.2 Assembly Scheratic Figure 7.2-6 Reed Switch Position Transmitter Cable 7.2 Assemblies O Figure 7.2-7 DNBR/LPD Calculator System (CEA Calculators) 7,2 V
Figure 7.2-8 Ex-Core Neutron Flux Monitoring System 7.2 Figure 7.2-9 Reactor Coolant Pump Speed Sensors Typical 7.2 for Each Reactor Coolant Pump Figure 7.2-10 Core Protection Calcelator functional Block 7.2 Diagram E
Figure 7.2-11 PPS Bistable Trip Logic Functional Block 7.2 Diagram
,igure 7.2-12 PPS Reactor Trip System Functional .2 7
Logic Diagram Figure 7.2-13 Typical PPS Channel Functional Bistable 7.2 ;
Trip Channel Bypass Figure 7.2-14 Typical PPS Channel Functional RPS 7.2 Initiation Logic figure 7.2-15 Typical PPS Variable Setpoint Operation 7.2 (Manual Reset)
( Figure 7.2-16 PPS Testing Overlap 7.2 l
Amendment E December 30, 1988
C '
CESSAR n!#,cuion TABLE2 1 7 & Cont'd)
(Sheet 2 of 4) 1
{
SAFETY RELATED J ECTRICAL. INSTRUMENTATION AND CONTROL DRA_ WINGS Drawino No. Title Section figure 7.2-17 PPS Manual Bistable Trip Test functional 7.2 81 cit Diagram figure 7.2-18 Typical PPS Channel Contact L2istable 7.2 ;
Interface Diagram j l
Figure 7.2-19 Plant Protection System Interface Logic 7.2 !
Diagram j Figure 7.2-20a MCBD Symbols, Notes and Abbreviations 7.2 1
Figure 7.2-20b MCBD Symbols, Notes and Abbreviations 7.2 l Figure 7.2-21a RCS Loop 1 Temperatures (Narrow) MCBD 7.2 E
Figure 7.2-21b RCS Loop 2 Temperatures (Narrow) MCBD 7.2 Figure 7.2-22a RCS Loop 1 Temperatures (Wide) MCBD 7.2 Figure 7.2-22b RCS Loop 2 Temperatures (Wide) MCBD 7.2 Figure 7.2-23a Reactor Coolant Pump Pressure MCBD 7.2 Figure 7.2-23b Reactor Coolant Pump Speed MCBD 7.2 Figure 7.2-24 Pressurizer Pressure MCBD 7.2 Figure 7.2-25 Nuclear Instrumentation MCBD 7.2 Figure 7.2-26 Containment Pressure MCBD 7.2 Figure 7.2-27a Steam Generator-1 Level (Wide) MCBD 7.2 Figure 7.2-27b Steam Generator-2 Level (Wide) MCBD 7.2 Figure 7.2-28a Steam Generator-1 Pressure MCBD 7.2 Figure 7.2-28b Steam Generator-2 Pressure MCBD 7.2 Figure 7.2-29a Steam Generator-1 Level (Narrow) MCBD 7.2 Figure 7.2-29b Steam Generator-2 Level (Narrow) MCBD 7.2 Figure 7.2-30 Steam Generator Primary D/P MCBD 7.2 i Amendment E December 30, 1988
\
I CESSAR !!aincuio.
a.
t ) TABLF. 1.7-1 (Cont'd) v (Sheet 3 of 4)
S_AfETY RELATED FLECTRICAL. INSTRUMENTATION AND CONTROL DRAWINGS Drawing _No. Title Section Figure 7.3-8a Typical FCLD fcr a Solenoid-0perated Valve 7.3
]
Figure 7.3<8b Typical Electr! cal Interface for a 7.3 Solenoid-0perated Valve i l
Figure 7.3-9a Typical FCLD for a Modulating Valve with 7.3 )
Solenoid-Operator j Figure 7.3-9b Typical Electrical Interface for a Modulating 7.3 Valve with Solenoid Operator Figure 7.3-10a Typical MOV Functional Interface Design 7.3 Figure 7.3-10h Typical Electrical Interface for a Motor- 7.3 g Operated Valve figure 7.3 11 Typical FCLD for a Full Throw Motor-0perated 7.3 Valve figure 7.3-12 Typical FCLD for a Throttling Motor-0perated 7.3 Valve Figure 7.3-13a Typical FCLS for a Contactor-0perated 7.3 Component figure 7.3-13b Typical Electrical Interface for a Contactor- 7.3 Operated Component figure 7.3-14a Typical FCLD for a Circuit Breaker-0perated 7.3 Component figure 7.3-14b Typical Electrical Interface for a Circuit 7.3 Breaker-Operated Component Figure 7.3-15a Typical FCLD for a Modulating Component 7.3 Figure 7.3-15b Typical Electrical Interface for a 7.3 Modulating Component figure 7.3-18 In-containment Refueling Water Storage Tank 7.3 MCBD Amendment E December 30, 1988
CESSAR n.541ricari:w TABLE 1.7-1 (Cont'd)
(Sheet 4 of 4)
JAFETY RELATED ELECTRICAL. INSTRUMENTATION AND CONTROL DRAWINGS Drawina No. Title Section Figure 7.3-19 Reactor Drain Tank MCBD 7.3 Figure 7.3-20a Safety Injection Tank 1 MCBD 7.3 Figure 7.3-20b Safety injection Tank 2 MCBD 7.3 Figure 7.3-20c Safety injection Tank 3 MCBD 7.3 J
Figure 7.3-20d Safety Injection Tank 4 MCBD 7.3 Figure 7.3-21 Containment Spr,ay MCBD 7.3 .
} '
E Figure 7.3-22 Shutdown Cooling MCBD 7.3 Figure 7.3-23 Safety injection MCBD 7.3 Figure 7.3-24 Safety Depressurization MCBD 7.3 Figure 7.5-8 Pressurizer Level Control System MCBD 7.5 Figure 7.7-24 Pressurizer Pressure (APS) MCBD 7.7 )
i Figure 7.7-25 Steam Generator-1 Level (APS) MCBD 7.7 O
Amendment E December 30, 1988
C E S S A R En nne. m . 1 TABLE 1.7-2 VALVE LIST IDENTIFIERS Valve Type Symbol Angle A Ball B Swing Check C Packless D-Butterfly F Globe G Spring Loaded Check L Needle N Plug P Relief R Safoty S Gate T {
Vacuum Breaker V Three Way W Excess Flow Check X Pressure Regulator gperator Type Pneumatic Diaphragm D Hand H '
Motor M None N Motor Modulating (Valve Position Proportion) O Pneumatic Piston .P Solenoid S Pneumatic Vane V ]
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Amendment E j December 30, 1988 )
_ _ _ - - - - _ _ - _ _ _ _ _ _ _ _ - _ - - _ - - . _ - . - - _ . - . . _ _ . . _ _ . - _ _ _ . _ - _ - . _ - _ _ .__.---.-_.----..a_ - - . _ _ _ _ . . . _ - - . _ _-_._---_-__J
CESSARE h m.,
i m TABLE 1.7-3 (Sheet 1 of 2)
PIPING AND INSTRUMENTATION DIAGRAMS Diagram No. Title Section Figure 1.7-1 Piping and Instrumentation Diagram 1.7 Symbols Figure 5.1.2-1 Reactor Coolant System P&ID 5.1 Figure 5.1.2-2 Reactor Coolant Pump P&ID 5.1 l Figure 5.4.7-3 Shutdown Cooling System Flow Diagram 5.4 Figure 6.3.2-1A Safety Injection System P&ID 6.3 Figure 6.3.2-1B Safety Injection System P&ID 6.3 Figure 6.3.2-1C Safety Injection System Frow Diagram 6.3
(N (Short-Term) l
\,
Figure 6.3.2-1D Safety Injection System Flow Diagram 6.3
, (Short-Term) l l
Figure 6.3.2-1E Safety Injection System Flow Diagram 6.3 (Long-Term)
Figure 6.3.2-1F Safety Injection System Flow Diagram 6.3 (Long-Term)
Figure 6.3.2-2 Engineered Safeguards System 6.3 E Schematic Diagram Figure 6.7-1 Safety Depressurization System Flow 6.7 Diagram Figure 3.5 Iodine Removal System P&ID App 6B Figure 9.1-3 Spent Fucl Pool Cooling and Cleanup 9.1 P&ID Figure 9.3.4-1 Chemical & Volume Control System 9.3 Flow Diagram g Figure 9.3.4-2 Chemical & Volume Control System 9.3
) Flow Diagram Amendment E December 30, 1988 L_-_------------------
CESSAR nninumu O
TADLE 1.7-3 l
J (Sheet 2 of 2)
PIPING AND INSTRUMENTATION DIAGRAMS DJ3 gram No. Title Section Figure 9.3.4-3 Chemical & Volume Control System 9.3 ,
i Flow Diagram Figure 9.3.4-4 Chemical & Volume Control System 9.3 Flow Diagram Figure 10.1-3 Main Steam System P&ID 10.1 ,
i Figure 10.4.8-1 Steam Generator Blowdown System 10.4 E
Flow Diagram Figure 10A-1 Emergency Feedwater System Schematic App. 10A O
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Amendment E ,
December 30, 1988 l
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I OVERSIZE .
DOCUMENT PAGE PULLED .- .
SEE APERTURE CARDS .
i NUMBER OF OVERSIZE PAGES FILMED ON APERTURE CARDS APERTURE CARD /MARD COPY AVAILABLE r10M RECORDS AND REPORTS MANAGEMENT BRANCH
d CESSAR!annem.
O ;
J 1.8 BEGULATORY GUIDES l B Table 1.8-1 lists the applicable Regulatory- Guides which arel addressed in the System 80+ Standard Design.
E
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Amendment E
'1.8~1 December 30, 1988~
i CESSAR E!'accia h)
TABLE 1.8-1 (Sheet 1 of 19)
REGULATORY GUIDES Original or Revision Reference Rocument/ Title GDG References Issue Date CESSAR Section Reg. Guide 1.1 - 11/70 6.3, 6.5 Net Positive Suction Head for Emergercy Core Cooling and Containment Heat Removal Systen Pumps Reg. Guide 1.2 - 11/70 5.2.3.3.1.1 Thermal Shock to Reactor Pressere Vessels - GDC 35 l
E Reg. Guide 1.3 - Not Applicable (BWR) ,
l Reg. Guide 1.4 - Revision 2 6.3.3.6 .
Assumptions Used for Evaluating 6/74 the Potential Radiological I I
n Consequences of a LOCA for
( Pressurized Water Reactors Reg. Guide 1.5 Not Applicable (BWR) l Reg. Guide 1.6 - 3/71 8.1.4.2 Independence Between Redundant l Standby (On-Site) Power Sources and Between Their Distribution Systems
)
l j Reg. Guide 1.7 - Revision 2 6.5 Supplement - Control of 11/78 I
l Combustible Gas Concentration I in Containment following LOCA Reg. Guide 1.8 - Not Applicable Personnel Selection and Training Reg. Guide 1.9 - Revision 2 8.1.4.2 j Selection of Diesel Generator 12/79 ]
Set Capacity for Standby Power i Supplies l
l O i Amendment E December 30, 1988
- -- - - - - - - - - --- - - - - - - - - - - - - - - _ - - - - - _ 1
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n - n-S A R 8 B lin c e TABLE 1,8-1 (Cont'd)
(Sheet 2 of 19)
REGULATORY GUIDES C'riginal or Revision Reference ;
Newellt/Ilt}# 900 iWfNntrL _ Issue Date CESSAR Section Reg. Guide 1.10 - Withdrawn Mechanical (Cadweld) 5pilces in Reinforcing Bars of icnc~ete r Containments Reg. Guide 1.11 - 3[l! 7.1.2.15 Instrument Lines Penet n ting Primary Reactor Containco t g Reg. Guide 1.12 - Rcvision . 3.7 ,
Instrumentation for Eartbaufkes 4/74 i Reg. Guide 1.13 - Revisits 2 9.1 l fuel Stcrage Facility Design 12/81 Basis 1 E
Reg. Guide 1.14 - Revision 1 5.4.1.1 Reactor Coolant Pcmp Flywincei 8/75 Integrity Reg. Guide 1.15 - Witiidrawn l3 Testing of Reinforcing Bars I for Concrete Structures Reg. Guide 1.16 - Not Applicable l Reporting of Operating l Information Reg. Guide 1.17 - Revision i 7.1.1.16, App. 13A l,'
Protection Against Industrial GT/3 Sabotage B
Reg. Guide 1 18 - !Hthdr w; Structura! Acceptance Tests for Concrete Primary Reactor Containments ]
B !
Reg. Guide 1.19 - Nithdrawn Nondestructive Examination of Primary Containment Welds f I l
A w.u M .a e n t E DecN3her 30, 1988
CESSAREanncum TABLE 1.8-1 (Cont'd)
(Sheet 3 of 19)
REGULATORY GUIDES Original or Revision Reference Document / Title GDC References Issue Qate CESSAR Section Reg. Guide 1.20 - Revision 2 3.9.2.4 Vibration Measurements on 5/76 Reactor Internals Reg. Guide 1.21 - Not Applicable Measuring and Renorting of l Effluents from Nuclear Poeer Plants Reg. Guide 1.22 - 2/72 7.1.2.17, 8.1.4.2 Periodic Testing of Protection Systems Actuation functions -
1 l Reg. Guide 1.23 - Not Applicable Onsite Meteorological Programs 1
Q(3 Rea Gcide 1.24 3/72 15.7 !
resumptions Ur.cd for Evaluating
, h e Potential Radiological l
Consequences of a Pressuri ad e t!ater Reactor Radioactive Gas S?arage icnk Failure l Rep. Guide 1.25 - 3/72 15.7 l Assumptions Used for Evaluating l
t% htential Radiological Cc xoquences 01 a f:iel Handling and Storage f acilt ty for Boiling 20 Pressurized Water Reactors Rs-; GuMe 1. 26 - Revision 3 3.2.2, 10.4 Quality Group Classifiuliou 2/76 and Stadards Reg G' tide L D - Revision 2 9.2.5 i Ultiste Heat tank 1/76 l
Reg. Guide i.2F? - rcvision 3 17 Ouality Awuranct Pre; ram 8/85 Eequi Du nt s Amendment E December 30, 1988 L_____.______________.._
8 l
CESSAR Enac-1 TABLE 1.8-1 (Cont'd)
(Sheet 4 of 19)
REGULATORY GUIDES Original or Revision Reference f
Documer_t] Title GDC Reference _s_ Issue Date _
CESSAR Section Reg. Guide 1.29 - Revision 3 3.2.1, 7.1.2.18, I Seismic Design Classification 9/78 10.4.9 )
f Reg. Guide 1.30 - 8/72 17 !
Quality Assurance Requirements l for the Installation, !nspection and Testing of instrumentation and Electrical Equipment.
Reg. Guide 1.31 - Revision 3 5.2.3.4.2.1 Control of Stainless Steel 4/78 Welding Reg. Guide 1.32 - Revision 2 8.1.4.2 Use of IEEE Std. 308-1971, 2/78
" Criteria for Class IE Electric i E
Systems for Nuclear Power Generating Stations" Reg. Guide 1.33 - Not Applicable Quality Assurance Program Requirements (Operation)
Reg. Guide 1.34 - 12/72 5.2.3.3.2 2 Control of Electroslag Weld Properties Reg. Guide 1.35 - Not Applicable l Inservice Surveillance of (Concrete containment)
Ungrouted Tendons in Prestre3 sed Concrete Containment Structures - ., ;
Reg. Guide 1.36 2/73 5.2.3.2.3+ a Nonmetallic Thermal Insulation 10.3.2.3.4 for Austenitic Stainless Steel l
l 9
Amendment E December 30, 1988
_____.____________________________J
i C E S S A R SE M?ic.1,:n N
TABLE 1.8-1 (Cont'd)
(h (Sheet 5 of 19)
REGULATORY GUli)ES Original or Revision Reference Document / Title tiDC References Issue Date CESSAR Section j
Reg. Guide 1.37 - 3/73 5.2.3.4.1.2.1, -
Quality Assurance Requirements 10.3.6.2 {
l for Lleaning of Fluid Systems ]
and Associated Components of
]
h ter-Cooled Nuclear Power Plants j a
Reg. Guide 1.38 - Revision 2 17 i Quality Assurance Requirements 5/77 i for Packag;ag, Shipping, )
Receiving, Storage, and Handling nf items for Water Cooled Nuclear Power Plants E Reg. Guide 1.39 - Not Applicable p Housekeeping Requirements for gj Water Cooled Nuclear Power Plants l
Reg. Guide 1.40 - 3/73 3.11 ;
Qualification Tests of !
Continuous - Duty Motors J l
Installed Inride of Containment of Water Cooled Nuclear Power Plants Reg. Guide 1.41 - 3/73 8.1.4.2, 14 E Preoperational Testing of '
Redundant On-site Electric Power Systems to Verify Proper Load Group Assignments Reg. Guide 1.42 - Withdrawn lB Interim Licensing Policy l on as low as Practicable for Gaseous Radiciodine Releases from Light Water-Cooled Nuclear Power Reactors O
Amendment E December 30, 1988 1 .
i CESSAREannem TABLE 1.8-1 (Cont'd)
(Sheet 6 of 19) i REGULATORY GUIDES !
Original or Revision Reference Document / Title GOC Ref(rent issue Date CESSAR Section l 1
Reg. Guide 1.43 - 5/73 5.2.3.3.2.1 Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components Reg. Guide 1.44 - 5/73 5.2.3.4.1.1.1 ,
Control of the Use of Sensitized l Stainless Steel l
Reg. Guide 1.45 - 5/73 5.2.5, 7.1.2'20 Reactor Coolant Pressure Boundary Leakage Detection E
Systems Reg. Guide 1.46 - Withdrawn Protection Against Pipe t l
Whip Inside Containment -
Reg. Guide 1.47 - 5/73 7.1.2.21, 8.1.4.2 Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems Reg. Guide i.48 - Withdrawn a Design Limits and Loading Combinations for Seismic Category 1 Fluid System Components Reg. Guide 1.49 - Revision 1 1.1.2 Power Levels of Water - 12/73 Cooled Nuclear Power Plants Reg. Guide 1.50 - 5/73 5.2.3.3.2.1 i Control of Preheat Temperature i J
for Welding of Low-Alloy Steel Reg. Guide 1.51 - Withdrawn B Inservice Inspection of ASME I Code Class 2 Nuclear Power Cot..ponen t s ei Amendment E December 30, 1988
CESSAR unincuna lABIE B 1.8-1 (Cont'd)
(Sheet 7 of 19)
REGULATORY GUIDES Original or Revision Reference Document / Title GDC References _ Issue Date CESSAR Section Reg. Guide 1.52 - Revision 2 6.5 E Design, Testing, and Maintenance 3/78 Criteria for Atmosphere Cleanup System Air Filtration and Absorption Units of Light-Water- !
Cooled Nuclear Power Plants j 1
Reg. Guide 1.53 - 6/73 7.1.2.9, 8.1.4.2 E j Application of the Single- ]
Failure Criterion to Nuclear 1 Power Plant Protection Systems 3 Reg. Guide 1.54 - 6/73 6.1.2.1 Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power O Plants Reg. Guide 1.55 - Withdrawn ln Concrete Placement in Category 1 l
Structures ;
i 1 Reg. Guide 1.56 - Not Applicable (BWR f Maintenance of Water Purity .
in Boiling Water Reactors Reg. Guide 1.57 - 3.8 )
6/73 '
Design Limits and Loading Combinations for Metal Primary l Reactor Containment System j i Components j Reg. Guide 1.58 - Revision 1 14.2.2.11 Qualification of Nuclear 9/80 Power Plant inspection, Examination and Testing j Personnel Reg. Guide 1.59 - Revision 2 2 Design Basis floods for 8/78 Nuclear Power Plants 1
Amendment E I December 30, 1988
CESSAR !!nincari:n TABLE 1.8-1 (Cont'd)
(Sheet 8 of 19)
REGLATORY GUIDES Original or Revision Reference Docume_nt/ Title GDC Refgr_qnces Issue Date CESSAR Section Reg. Guide 1.60 - Revision 1 2, 3.7 E Design Response Spectra for 12/73 Seismic Design of Nuclear Power Plants Reg. Guide 1.61 - 10/73 3.7.1.3 Damping Values for Seismic Design of Nucleir Power Plants Reg. Guide 1.62 - 10/73 7.1.2.22, 8.1.4.2 Manual Initiation of Protective Actions ,
Reg. Guide 1.63 - Revision 3 3.8.2, 8.1.4.2 Electric Penetration Assemblies 2/87 in Containment Structures for Water-Cooled Nuclear Power Plants Reg. Guide 1.64 - Revision 2 17 O
Quality Assurance Requirements 6/76 for the Design of Nuclear Power Plants Rec. .'uide 1.65 - 10/73 5.3.1.7 Mat: ,al and Inspection for Reactor Vessel Closure Studs Reg. Guide 1.66 - Withdrawn Nondestructive Examination of Tubular Products B
Reg. Guide 1.67 - Withdrawn Installatini, of overpressure Protection Devices Reg. Guide 1.68 - Revision 2 14.2.7.1, 7.1.2.24, Preoperational and Initial 8/78 8.1.4.2 E Startup Test Program for Water-Cooled Power Reactors Reg. Guide 1.68.1 - Not Applicable (BWR)
Preoperational and initial Startup Testing of Feedwater and Condensate Systems for <
Boiling Water Power Plants Amendment E December 30, 1988
CESSAR Emincune,.
i (9 TABLE 1.8-1 (Cont'd)
'u)
(Sheet 9 of 19)
RECULfTORY GUIDES J l
Original or Revision Reference .
Document / Title GDC References - Yssue Date CESSAR Section Reg. Guide 1.68.'2 - Revision 1 14.2.7.3 Initial Startup Test Program 7/78 to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear
, Power Plants Reg. Cuide 1.68.3 - 4/82 14 Preoperational Testing of Instrument and Control Air Reg. Guide 1.69 - 12/73 12 Concrete Radiation Shields for Nuclear Power Plants Reg. Guide 1.70 - Revision 3 CESSAR-DC Intro- ,
e' Standard Format and Contents 11/78 ductory Statement
- l. of Safety Analysis Reports for Nuclear Power Plants Reg. Guide 1.71 - 12/73 5.2.3.3.2.3 Welder Qualification for Areas of Limited Accessibility l Reg. Guide 1.72 - Revi sior, 2 9.2.5 E Spray Pond Plastic Piping 11/78
.i Reg. Guide 1.73 - 1/74 3.11, 7.1.2.25 Qualification Tests cf Electric Valve Ope ators I stalled Inside ;
the Containment of Nuclear Power Plants !
Reg. Guide 1,74 - 2/74 17 Quality Assurance Terms and Definitions Reg. Guide 1.75 - Revision 2 7.1.: 10, 8.1.4.2 Physical Independence of 9/78 Electric Systems
- g. Reg. Guide 1.76 - 4/74 ?
Design 8ases Tornado for Nuclear Power Plants
'7endment E December 30, 1988
CESSARHM%ua 1 i
i A
T_ABLE __1.8-1 (Cont'd)
(Sheet 10 of 19)
REGULATORY GUJ0 Q briginal or Revision Reference ,
Document / Title GDC References _ Issue Date _
CESSAR Section t
Reg. Guide 1.77 - 5/74 15.4.5 Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors .
Assumptions for Evaluating 6/74 6.4, 9.4 IE the Habitability of a Nuclear Power Plant Control Room During a Fostulated Hazardous Chemical Release Reg. Guide 1.79 - Revision 1 3.1.33, 14.2.7.2 Preoperational Testing of 9/75 Emergency Core Cooling Systems for Pressurized Water Systems ;
Reg. Guide 1.80 - Withdrawn Preoperationa'l Testing of Instrument Air Systems Reg. Guide 1.81 - Revision 1 1.2.1.3 Shared Emergency and Shutdown 1/7S Electric Systems for Multi-Unit Nuclear Power Plants RW , 6ulde 1.82 - Revision 1 1.2, 3.8 i Wa.er Sources for Long Term 11/85 i Recirculation Cooling Following a E Loss-of-Coolant Accident Reg. Guide 1,83 - Revicion 1 5.2.4.1 Inservice Inspection of 7/75 )
Pressurized Water Reactor l i
Steam Generator Tubes Reg. Guide 1.84 - Revision 25 5.2.1.2 Code Case Acceptability ASME 5/88 l Sec+. ion 111 Design and l Fabrication s O
i Amendment E ;
December 30, 1988 I
CESSAR nnincm.
= j j
h.g -TABLE'l.8-1 (Cont'd)
(Sheet 11 of 19)
REGULATORY GlqD.L]
D Original or Revision Reference Document / Title GDC References Issue Date CESSAR Section Reg. Guide 1.85 - Revision 25 5.2.1.2 E t
.j Code Case Acceptability ASME 5/88 4 i Section III Materials l Reg. Guide 1.86 - Not Applicable E Termination of Operating ,
Licenses for Nuclear Reactors l l Reg. Guide 1.87 Not Applicable (ETR)
Construction Criteria for Class 1 Components in Elevated Temperature Reactors Reg. Guide 1.88 - Revision 2 17 Collection, Storage, and 10/76 B l
( Maintenance of Nuclear Power l V] Plant Quality Assurance Records Reg. Guide 1.89 - Revision 1 3.11, 7.1. 2. 5, 8.1. 4. 2 l E I Environmental Qualification of 6/84 J !
Certain Electric Equipment l Important to Safety for Nuclear i Power Plants l 3
Reg. Guide 1,90 - Not Applicable Inservice inspection of Prestressed (Concrete containment)
Concrete Containment Structures with Grooted Tendons Reg. Guide 1.91 - Not Applicable
- Evaluations of Explosions Postulated E l to Occur en Transportation Routes Near Nuclear Power Plants Reg. Guide 1.92 - Revision 1 3.7.2.7 a Combining Modal Responses and 2/76 r Spatial Components in Seismic Response Analysis Reg. Guide 1.93 - 12/74 8.1.4.2 E
. Availability of Electric Power
( Sources l
Amendment E December 30, 1988
l CESSAREnnnema TABLE 1.8-1 (Cont'd)
-(Sheet 12 of 19) ]
B_EGULATORY GUIDES Original or Revision Reference Document / Title GDC References Issue Date CESSAR Section I
Reg. Guide 1.94 - Revision 1 3.8 E Quality A-ssurance Requirements 4/76 f for Installation, inspection, and Testing of Structural L Concrete and Structural Steel During the Construction Phase of j 1
Nuclear Power Plants Reg. Guide 1.95 - Revision 1 6.4 E
Protection of Nuclear Power Plant 1/77 Centrol Room Operators Against an Accidental Chlorine Release Reg. Guide 1.96 - Not Applicable .(BWRs 3 Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear .~
Power Plants Reg. Guide 1.97 - Revision 3 3.1, 7.1-.2.26, 7,5, E
instrumentation for Light-Water- 5/83 10.4.9 ,
Cooled Nuclear Power Plants To {
Assess Plant and Environs Conditions During and following
,an Accident. {
1 Reg. Guide 1.98 - Not Applicable (BWRs) ,
Assumptions Used for Evaluating In the Potential Radiological i Consequences of a Radioactive )
Offgas System failure in a l i
Boiling Water Reactor Reg. Guide 1.99 - Revision 2 5.3.1.6.7 E <
Effects of Residual Elements on 5/88 Predicted Radiation Damage to Reactor Vessel Materials l
} Reg. Guide 1,100 - Revision 2 3.10 !E i
Seismic Qualification of Electric 6/88 Equipment for Nuclear Power Plants !
l Amendment E :
l December 30, 1988 1
1 l
CESSAR 8Esncma TABLE 1.8-1 (Cont'd)
(Sheet 13 of 19)
REGULATORY GUIDES i Original or Revision Reference l Document / Title GDC References Issue Date _ CESSAR Section I Reg. Guide 1.101 - Withdrawn j B
Emergency Planning for !
Nuclear Power Plants i Reg. Guide 1.102 - Revision 1 2 E Flood Protection for Nuclear 9/76 Power Plants I Reg. Guide 1.103 - Withdrawn Post Tensioned Prestressing Systems for Concrete Reactor B Vessels and Containment Reg. Guide 1.104 - Withdrawn i Overbecd Crane Handling i Systems for Nuclear Power O Plants E !
Reg. Guide 1.105 - Revision 2 7.1.2.27 Instrument Setpoints for 2/86 Safety-Related Systems <
j Reg. Guide 1.106 - Revision 1 7.1.2.28 l' Thermal Overload Protection 3/77 for Electric Motors on l Motor 0perated Valves Reg. Guide 1.107 - Not Applicable 3 l l
Qualifications for Cement (Concrete containment)
Grouting for Prestressing -
Tendons in Containment Structures l l
Reg.~ Guide 1.108- Revision 1 8.1 E j Periodic Testing of Diesel 8/77 Conerator Units Used as Onsite Electric Power Systems at Nuclear Power Plants Reg. Guide 1.109 - Not Applicable E l Calculation of Annual Doses to Man ;
From Routine Releases of Reactor O
V Effluents for the Purpose of Evaluating Compliance with 10 CFR i
l Part 50, Appendix I i
Amendment E December 30, 1988
( !
s t
i CESSAR 8lahi.
TABLE _1.8-1 (Cont'd)
(Sheet 14 of 19) {
1 RE.GULATORY GUIDEji j Original or Revisior, Reference l DAcumerttITltle GDC References Issue _ DATA ___ _CflSAR Section Reg. Guide 1.110 - 3/76 12 ]
Cost-Benefit Analysis for Radwaste i Systems for Light-Water-Cooled i Nuclear PoWr Reactors Reg. Guide 1.111 - Not Applicable l Methods for f.stimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors Reg. Guide 1.112 - Not Applicable ,
Calculation of Releas es of l Radioactive Mater.als in 1 Gaseous and Liquid Effluents {
from Light-WateF-Cooled 1 Power Reactors e l i
Not Applicable ;
Estimating Aquatic Dispersion ]
of Effluents from Accidental
) and Routine Reactor Releases )
for the Purpose of Implementing Appendix 1 Reg. Guide 1.114 - Not Applicable Guidance on Being Operator at the Controls of a Nuclear Power Plant Reg. Guide 1.115 - Revision 1 3.5 Protection Against Low- 7/77 Trajectory Turbine Missiles Reg. Guide 1.116 - 5/77 3.9 Quality Assurance Requirements I for Installation, Inspectiou, and Testing of Mechanical Equipment and Systems l
Reg. Guide 1.117 - Revision 1 3.1.2 Tornado Design Classification 4/78 Amendment E December 30, 1988
CESSAR nainmin
__ _ q
( l TABLE 1.8-1 (Cont'd)
%)
(Sheet 15 of 19) I REGULATORY GUIDES 1
i Original or Revision Reference 4
} Document / Title GOC References issue Date _
CESSAR Sectior, 1
Periodic Testing of Electric Power and Protection Systems Revision 2 6/78 7.1.2.7, 8 lr )
{
i 8
Reg. Guide 1.119 - Withdrawn Surveillance Program for ls l New Fuel Assembly Designs ]
Reg. Guide 1.120 -- Revision 1 7.1.2.29 L j Fire Protection Guidelines for 11/77 Nuclear Power Plants
(
Reg. Guide 1.l'21 - 8/76 S.4 -E Bases for Plu0ging Degraded PWR Steam Genertor Tubes Reg. Guide 1.122 - Revision 1 3.7.2 p(,) Development of Floor Design Response 2/78 a 1
j Spectra for Seismic Design of Floor-Supported Equipment or Components !
i Reg. Guide 1.123 - Revision 1 Quality Assurance Requirements for 7/77 17 IE i l
Control of Procurement of items ;
and Services for Nuclear Power Plants l B !
Reg. Guide 1.124 - Revision 1 5.4.14 l Service Limits and Loading 1/78 i Combinations .ror Class 1 Linear Type Component Supports Reg. Guide 1.125 - Revision 1 3 E
Physical Models for Design and 10/78 Operation of Hydraulic Structures .
) and Systems for Nuclear Power
! Plants ,
Reg. 6uide 1.126 - Revision 1 4.2 E An Acceptable Model and Related 3/78 Statistical Methods for the Analysis l of Fuel Densification 3
(O Amendment E December 30, 1988 i
F"
?
CESSAR 88Lmu 1ABLE 1 & 1 (Cont'd)
(Sheet 16 of '19)
BEGULATORY GUIDES Original or Revision Reference ,
i Rocument/ Title GDC Reference _L _ __jssue Date _ _, CESSt.R Section _
Reg. Guide 1.127 - Not Applicable E Inspection of Water-Control 8 j i
Structures Associated with Nuclear Power Plants Reg. Guide 1.128 - Revision 1 8.1, 8.3.2 g Installation Design and Installation 10/76 ;
of large Lead Storage Batteries j for Wuclear Power Plants !
Reg. Guide 1.129 - Revision 1 8.1,4.2 2,'78 B Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Nuclear Power Plants Reg. Guide 1.130 - Revision 1 5.4.14 i
Service Limits and Loading 10/76 Combinations for Class 1 Plate-and-Shell-Type Component Supports Reg. Guide 1.131 - 8/77 8.'l.4.2 g Qualification Tests of Electric Cables, Field Splitet, and Connections for Light-Water-Cooled 3 Nuclear Power Plants Reg. Guide 1.132 - Not Applicable E Site Investigations f.or foundations ,
of Nuclear Power Plants l Reg. Guide 1.133 - Revision 1 7.1.2.30, 7.7.1.6.3 lE Loose-Part Detection Program for 5/81 i the Primary Systems of Light-Water- ]
Cooled Reactors lB j l
Reg. Guide 1.134 - Not Applicable j Medical Evaluation of Nuclear Power le Plant Personnel Requiring Operator Licenses lB Reg. Guide 1.135 - Not Applicable g i Normal Water lovel and Discharge at i Nuclear Power Plants {
Amendment E December 30, 1988
CESSARnaincmo
=_
(J TABLE 1.8-1 (Cont'd)
(Sheet 17 of 19}
REGULATORf GUIDES l Original or Revisi0n Reference Document / Title GDC References Essue Date CESSAR Section :
Reg. Guide 1.136 - Not Applicable Materials, Construction, and Testing (Concrete ccatainment of Concrete Containments (Articles CC-1000, -2000, and -4000 through
-6000 of the " Code for Concrete Reactor Vessels and Containments" l Reg. Guide 1 137 - Revision 1 9.5 Fuel-0il Systems for Standby Diesel 10/79 Generators Reg. Guide 1.138 - Not Applicable Laboratory Investigations of Soils for Engineering Analysis and Design of Nuclear Power Plants Reg. Guide 1.139 - 5/78 5.4.7 Guidance for Residual Heat Removal Reg. Guide 1.140 - Revision 1 94 Design, Testing, and Maintenance 10/79 E Criteria for Normal Ventilation ;
Exhaust System Air Filtration and l Absorption Units of Light-Water-Cooled Nuclear Power Plants Reg. Guide 1.141 - 4/78 6.2.4 Containment Isolation Provisions for Fluid Systems Reg. Guide 1.142 - Revision 1 3.8 Safety-Related Concrete Structures 10/81 i for Nuclear Pcwer Plants (Other Than Reactor Vessels and Containment)
Reg. Guide 1.143 - Revision 1 11 ,
Design Guidance for Radioactive 10/79 Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants
()
V Amendment E December 30, 1988
CESSAR nni"lCATION TABLE 1.8-1 (Cont'd)
(Sheet 18 of 19)
REGULATORY GUIDES Original or Revision Reference Issue Date CESSAR Settlpr Document / Title GDC References _
Reg. Guide 1.144 - Not Applicable Auditing of Quality Assurance Programs for Nuclear Power Plants !
Reg. Guide 1.145 - Revision 1 ?
Atmospheric Dispersion Models for 11/82 Potential Accident Consequence Assessment at Nuclear Power Plants Reg, Guide 1.146 - Not Applicable Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants Reg. Guide 1.147 - Revision 6 5.2.1.2 In-service inspection Code 5/88 Case Acceptability, ASME '
Section XI, Division 1 Reg. Guide 1.148 - 3/81 3,5,6 Functional Specification for i Active Valve Assemblias in lg Systems Important to Safety in Nuclear Power Plants
}
Reg. Guide 1.149 - Revision 1 Not Applicable j Nuclear Power Plant Simulators 4/87 i for Use in Operator Training l Reg. Guide 1.150 - Revision 1 5.1.4, 5.3.1.3 I Ultrasonic Testing of Reactor 2/83 Vessel Welds During Pre-service and in-service ,
Examinations {
l Reg. Guide 1.151 - 7/83 5.1.4, 7.1.2.31, !
Instrument Sensing Lines 7.1.3(E) j l
Reg. Guide 1.152 - 11/85 7.1.2.32 i Criteria for Programmable Digital Computer System Software in Safety Systems of Nuclear Power Generating Stations Amendinent E l December 30, 1988 i
~
m CESSAR En'JPICATl*.N TABLE 1.8-1 (Cont'd)
(Sheet 19 of 19) l REGULATORY GUIDES Original or Revision Reference Document / Title GDC References Issue Date CESSAR Section i Reg. Guide 1.153 - 12/85 5.1.4, 7.1.2.13 Criteria for Power, Instrumentation, and Control Portion of Safety Systems Reg. Guide 1.154 - Not Applicable Format and Content of Plant-Specific (Plant-Specific) ;
Pressurized Thermal Shock Safety Analysis Reports for PWRs j l
Reg. Guide 1.155 - 8/88 8.1, 10.4.8, 10.4.9 i Station Blackout 1 Reg. Guide 1.156 - 11/87 7.1.2.33 Environmental Qualification of Connection Assemblies for Nuclear Power Plants Reg. Guide 5.65 - 9/86 13.6, App. 13A I
' Vital Area Access Controls, Protection i of Physical Security Equipment, and ;
Key and Lock Controls j Reg. Guide 8.8 - Revision 3 12 E
l laformation Relevant to Ensuring 6/78 l Lhe Occupational Radiation
- Exposures at Nuclear Power l Stations will be ALARA j 1 ;
Reg. Guide 8.12 - Revision 2 7.1.2.34, 7.7.1.1.10 j i Criticality Accident Alarm 10/88 i Systems Reg. Guide 8.19 - Revision 1 12 ,
Occupational Radiation Dose 5/79 ;
Assessment in Light-Water i Reactor Plants-Design Stage Man-Rem Estimates i
I V
l l 1 l
Amendment E December 30, 1988 j E____________________________________________
s CESSAR !!nincamu .
l 1.9 BYSTEM 80+ STANDARD DESIGN INTE_RFACES The System 80+ Standard Design includes all buildings, structures, systems, and components which can significantly affect plant safety. (Any interfaces with site-specific systems or references to design implementation documents will be E identified in this section.)-
1 I
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l l l 1
)
I l 3 l
i r
l 9 \.
Amendment E l 1.9-1 December 30, .1988 L:-_-__-________ - :_______-_________________ . _ _ _ _ _ _ _ _ _ _ _ _ _