ML20043C877
| ML20043C877 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 06/04/1990 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20043C871 | List: |
| References | |
| NUDOCS 9006060294 | |
| Download: ML20043C877 (123) | |
Text
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(v3 Amendment 12. Page change instruction I
The following pages hase been changed, please make the specified changes in your SSAR. Pages art listed as ,
page pairs (front & back), Hold page numbers represent a page that has been changed by Amendment 12.
REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No.
CHAl"TER1 4.63,4 4.6 3,4 4.66,6.1 4.6-6,6.1 1.8 ii,iii 1.8-il,lli 4.6 7,8 4.6-7,8 1.81 1.8 1 4.6 9,10 4.6 9,10 Remove all 1.8 51 pages (1.8 51 thru 1.8 51o) 4.6 11,12 4.6-11,12 and replaec them with pages 1.8 51 thru l 8 66, 4.6 15,16 4.6 15,16 Ihis is change in page numbering only. 4.6 21,22 4.6-21,22 Add 1.8 67,68 Add 1.8 69,70 CH APTER 6 Add 1.8 71,72 Add 1.8 73,74 6 3-13,14 63 13,14 Add 1.8 75,76 Add 1.8 77.78 CHAPTER 9 Add 1.8 79,80 Add 1.8 81,N2 9.1 17,18 9.1 17,18 Add 1.8 83,84 Add 1.8 85,H6 93 liia,iv 93 lila,llib
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Add 1.8 87,NN Add 93iv Add 1.8 89,90 93 2,3 93 2,3 1.9 2,3 1.9-2,3 9 3-19 93 1923 1.9 4 1.93.1 CH Al"TER 10 Cil APTER 3 10.2 3,4 10.2-3,4 3.26,7 3.26,7 3.2 7a 3.2 7a CHAPTER 15 3.29 3.2 9 3.2 10,11 3.2 10,11 15.7 13,14 15.7-13,14 3.2-13.14 3.2 13,14 15.7-15,16 15.7 15,16 3.2 21,21a 3.2 21,2Ia 3.2 21b,22 3.2 21b,22 15A.6 53,54 15A.6 53,54 3.2 23,23.1 3.2 23,23,1 l Add 3.2 23.2 CHAirTER 20 1 l
3.2 27,28 3.2 27,2S 3.2 29,30 3.2 29,30 20.2 4p,4q 20.2 4p,4q 3.71,2 3.71,2 3.9-11,11a 3.911,11a 203 vi,vii 203-vi,vii 20 3-9,10 203 9,10 Cil APTER 4 203-48,48.1 203-48,48.1 20 3 64,65 203 64,65 4.5 1,2 4.5-1,2 203 82,82.1 203-82,82.1 4.6-1,l a 4.6-1,l a 20 3 144,145 20 3-144,145 20 3-160,161 20 3 160,161
( 4.6 2 4.6 2 9006060294 900604 ADOCK0500p,{
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lAB M 23^6 oore Standard Plant krv. c -
SECTION 1.8 - ,
$ CONTENTS Section Dilt Eagt i
1.8.1 Comfonaance With Standard Review Plan 0 1.S1 !
1.8.2 Applicability of Codes and Standards 1.81 i 1.8.3 . Applicability of Experience lefonaation 1.81 '
1.8.4 laterfaces 1.81 !
- TABLES Table Htle <
Eass ;
1.8 1 Summary of Differences From SRP Section 1 : 1.8-2 b 1.82 Summary of Differences From SRP Section 2 : 1.8 3 1.83 Summary of Differences Frem SRP Section 3 / 1.8-4
_ 1.8-4 Summary of Differences From SRP Section 4- 1.8-5 1.8-5 Summary of Differences From SRP Section 5 1.8 6 1 1.86 Summary of Differences From SRP Section 6 1.8-7 1.8-7 Summary of Differences From SRP Section 7 1.8-8 1.88 Summary of Differences From SRP Section 8 ' 1.89 1.8-9 Summary of Differences From SRP Section 9 ~ 1.8 10 1.8-10 ~ Summary of Differences From SRP Section .10 ? 1.8-11' 1.8-11 Summary of Differences From SRP Section 11- 1.8 1.8 12 Summary of Differences From SRP Section 12 1.8-13 1.8 13 Summary of Differences From SRP Section 13 - 1.8-14 1.8 14 Summary of Differences From SRP Section .14 1.8-15
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1.8-15 Summary of Differences From SRP Section 15 1.8 16 1 . 8 11 Amendment 12 E-m unuma mm --mim i imi ' ' ' ' ' ' ' ' ' ' ' " ' ' ' '
..q MM 23A6100AC O RennAmed Plant - nv c SECTION 1.8 1.m TABLES m
e 1.8 16 Summary of Differences From SRP Section 16 1.8-17
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1.8 17 Summary of Differences From SRP Section 17 1.8-18 1.8-18 Summary of Differences From SRP Section 18 1.8-19 '*
- 1.8 19 Standard Review Plans and Branch Technical Positions Applicable to ABWR 1.8-20 . ,
1.8 20 NRC Regulatory Guides Applicable to ABWR 1.8 382 ~I 1.8 Industrial Codes and Standards Applicable to ABWR 1.8 51 l
1.8 22 _ Experience Information Applicable to ABWR - 1.8-67 : l i
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T 1.8 iii i
Amendment 12
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ABWR mumc Standard Plant arv. c !
1 1.8 CONFORMANCEWITH STANDARD Standard Plant / remainder of plant interface l REVIEW PLAN AND APPLICABILITY OF impact. See Subsection 1.8.4 for interface '
CODES AND STANDARDS requirements. ;
1.8.1 Conformance With Standard 1.8.4 Interfaces '
Review Plan i The SRP sections to be addressed and the The r,ubsection provides the information applicable regulatory guides and experience required by 10 CFR 5034(g) showing conformance information for the remainder of the plant are
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with the Standard Review Plan (SRP). The summary those of Table 1.8 19,1.8 20 and 1.8 22 of differences from the SRP section is presented identified in the comment column, as by SRP section in Tables 1.81 through 1.818. *in t e r f a c e". .
See Subsection 1.8.4 for interface requirements.
1.8.2 Applicability of Codes !
and Standards Standard Review Plans, Branch Technical !
l Positions, Regulatory Guides and Industrial Codes '
and Standards which are applicable to the ABWR ,
design are provided in Tables 1.819,1.8 20 and.
1.8 21. Applicable revisions are also shown, i l See Subsection 1.8.4 for interface requirements.
1.8.3 Applicability of Experience Information '
Experience information is routinely made
. available and distibuted to design personnel in the design process. Nuclear field experience is maintained in hard copy form in functional component and library files and in the GE world vide computer retrieval system.
Generic Letters and IE Bulletins, Information Notices and Circulars covering the decade including 1980 through the current issues (early 1990) were reviewed for applicability to the ABWR design. The review was enhanced by associating related experiences and tracing referenced ["
occurences. This was accomplished starting with !
the current issues of the Generic Letters and proceeding back into the decade. The Circulars, Bulletins and Notices were reviewed in that order. Interfacing experience was incuded in the review. The selection of ABWR information was based on the significance to future design and operation guidance. Included is a list of NUREGs related to the closing of current safety issues.
[ Experience that resulted in applicable rules, codes and standards was not repeated. Table 1.8 22 list the experience information that has been included in the ADWR design or has ABWR O Amendment 12 g g.g
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MM 23A61ooAC Staridard Piant arv c TABLE 1&21 INDUSTRIAL CODES AND STANDARDS APPLICABLE'ID ABWR Code or
_ Standard Number Year 'litle ACI 211.1 1981 Practice for Selecting Proportions for Normal, Heavy Weight, and Mass Concrete.
? 212 1981 Guide for Admixtures in Concrete 214 1977 Recommended Practice for Evaluation of Strength Test Results of Concrete 301 1984 Specifications for Structural Concrete for Buildings 304 1973 Practice for Measuring, Mixing, Transporting, and Placing of Concrete
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( ; 305 1977 Recommended Practice for Hot Weather Concreting 306 1978 Recommended Practice for Cold Weather Concreting 307 1979 Specification for the Design and Construction of Reinforced Concrete Chimneys L
h 308 1981 Practice for Curing Concrete
! 309 1972 Practice for Conx -lation of Concrete E i l
g 311.1R 1981 ACI Manual of Concrete laspection 7 ;)
k : 311.4R 1981 a 1 Guide for Concrete Inspection 315 1980 Details and Detailing of Concrete Reinforcement
, 318 1983 Building Code Requirements for Reinforced Concrete Code Requirements for Nuclear Safety Related Concrete 349 1980 Structures 359 (See ASME BPVC Section III)
O Anwndment 6 iA-51 M
MN 2M6100AC Standard Plant nrv. c TABLE 1.8 21 (Continued)
INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR Code or Standard
. Number Year Title AISC N690 1984 Specifications for the Design, Fabrication, and Erection of Steel Safety Related Structures for Nuclear Facilities SG 673 1986 Specification for the Design of Cold Formed Steel Structural Members E
= -- --
Manual of Steel Construction G
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dhb ' 23A6100AC Standard Plant nry c l
TABLE 13 21(Continued)
INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR
{
Code or Standard .
Number Year Title !
ANS 2.3 1983 Standard for Estimating Tornado and Other Extreme Wind l Characteristics at Nuclear Power Sites 2.8 1981 . Determining Design Basis Flooding at Power Reactor Sites 5.1 1979 Decay Heat Power in LWRs 18.1 (N237) 1984 Radioactive Source Term for Normal Operation of LWRs 52.1 1983 - Nuclear Safety Design Criteria for the Design of Stationary Boiling Water Reactor Plants ;
55.4 1979 Gaseous Radioactive Waste Processing Systems for Light Water Reactors j 57.1 1980 Design Requirements for LWR Fuel Handling Systems 57.2(N270) 1976 Design Requirements for LWR Spent Fuel Storage Facilities at NPP.
58.2 1988 Design Basis for Protiction of Light Water NPP Against Effects of Postulated Pipe Rupture 59.51 (N195) 1976 Fuel Oil Systems for Standby Diesel Generators W
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Amendment 12 1 & 53 l
MN 2W100AC Standard Plant mye TABLE 1.8 21 (Continued)
INDUSTRIAL CODES AND STANDARDS APPLICABLETO ABWR Code or Standard Number Year utie ANSI A58.1 1982 Design Loads for Buildings and other Structures, Minimum B3.5 1960 American Standard Tolerance for Ball and Roller Bearings B30.2 1983 Overhead and Gantry Crancs (Top Running Bridge, Single or Multiple Grider, Top Running Trolley Hoist)
B30.11 1980 Monorail and Underbung Crancs B31.2 1986 Power Piping Cl 1985 Specifications of General Requirements for a Quality Program D975 1981 Diesel Fuel Oils, Spec for HEl 1970 Standards for Steam Surface Condenser,6th E., Heat Exchangers Institute MC11.1 1976 Quality Standard for lastrument Air N5.12 1972 Protective Coatings (Paint) for Nuclear Industry N13.1 1%9 Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities N14.6 1986 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 Kg) or More for Nuclear Materials N509 1980 Nuclear Power Plant Air Cleaning Units and Components N510 1980 Testing of Nuclear Air-Cleaning Systems N101.2 1972 Protective Coatings (Paints) for Light Water Nuclear Containment Facilities N101.4 1972 QA for Protective Coatings Applied to Nuclear I acilities N195 (See ANS 59.51)
N237 (See ANS 18.1 9 I Amendment 12 gg l
ABM 2M6100AC E#andmed Plant arv. c 1
i TABLE 1&21(Continued)
INDUSTRIAL CODES AND STANDARDS
- APPLICABLE 1D ABWR-
! Code or i !
st dard i
Number Yrar Title ANSI
, . N270 (See ANS 52.2) i i OM3 1987 Requirements for preoperational and Initial Startup l Vibration Test Programs for Water Cooled Power Plants
- OM7 1986 Requirements for Thermal Expansion Testing of Nuclear Plant Piping Systems [ September 1986 (Draft Revision 7)]
API 620 1986 Rules for Design and Construction of Large, Welded, l Low Pressure Storage Tanks 650 1980 Welded SteelTanks for Oil Storage 1O O .
l Amendment 12 1.g.55 l
.m...a..a ABMpi... 2 mime s ary, TABLE 1&21(Continued)
INDUSTRIAL CODES AND STANDARDS i APPLICABLE 1D ABWR - !
.1 Code or Standard R Number Year title j
'I ASHRAE l
30 1978 Methods of Tc:. ting IJquid Chilling Packages ,
33 1978 Methods of Testlog Forced Circulation Air Cooling and i Air Heating Coils !
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Amendment 12 1.8-$6 a
t ABWR zwime mandard Plant nrv. c f TABLE 1.8 21 (Continned)
INDUSTRIAL CODES AND STANDARDS I APPLICABLETO ABWR Code or Standard '
Number Year- Title ASME :
l B30.2 =1983 . Overhead and Gantry Cranes .
I B30.9 1984 Slings I'
B30.10 1982 Hooks B30.11 1980 Monorails and Underbung Cranes ,
B30.16 1981 Overhead Holsts i
B31.1 1986 Power Piping B96.1 1986 Specification for Welded Alumir % Alloy Storage Tanks N45.2 1977 QA Program Requirements'e ' Juclear Power Plants ,
N45.2.1 1980 Cleaning of Fluid Systems and Associated Components ;
During construction of Nuclear Power Plants '
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Amendment 12 1.8 57
ABM - useiooxc3 Remndard Plant a r v. c -
TABLE 1.8 21 (Continued)
INDUSTRIAL CODES AND STANDARDS APPLICABLE'IO ABWR h
" Code or
- Standard Number Year Title ASME N45.2.2 1972 Packaging, Shipping, receiving, Storage, and Handling of Items for Nuclear Power Plants (During Const.ruction Phase)
N45.2.6 1978 Qualifications of Inspection, Examination and Testing.
Personnel for the Construction Phase of Nuclear Power Plants N45.2.9 1979 Requirements.for the Collecilon, Storage, and Maintenance of OA Records for Nuclear Power Plants-N45.4 1972 Leakage Rate Testing of Containment Structures for Nuclear Reactors NOA1 1983 Quality Assurance Program Requirements for Nuclear Pacilities t NOA1A .1983 Addenda to ANSI /ASME NOA 11986 9 NOA 2 1983 Quality Assurance Requirements for Nuclear Power Plants K
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/.mendment 12 gg3g e
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23A6100AC . l grandard Plant nv c ! ::
l TABLE 1.8 21 (Continued) -I O--
INDUSTRIAL CODES AND STANDARDS a
o APPLICABLE TO ABWR
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Code or Standard Number Year 'ntle !
l ASME !
I Sec 11 1989 BPVC Section 11, Material Specifications Secill
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1989 BPVC Section Ill, Rules for Construction of Nuclear h Power Plant Components i j i Sec Vill 1989 BPVC Section Vill, Rules for Construction of Pressure { ~
Vessel !
I Sec IV 1989 BPVC Section IX, Qualification Standard for Welding and .
Brazing Procedures Welder, Brazers and Welding and = !;.
Brazing Operators l Sec XI 1989 BPVC Section XI, Rutas for Inservice Inspection of -l Nuclear Power Plant Components :
ASTM (See ASME BPVC Section III) e e .
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\ 23A6100AC Riandard Plant arv. c TABLE 1&21(Continned)
INDUSTRIAL CODES AND STANDARDS ;
APPLICABLE TO AB%R Code or i Standard 1 Number Year Title AWS A4.2 - 1986 Procedures for Calibrating Magnetic Instruments toi I Measure the Delta Ferrite content of Anstenitic Stainless Steel Weld Metal D1.1 1986, Steel StructdralWelding Code l D14.1 1985 Welding of Industrial and Mill Cranes and other - .
Material Handling Equipment i Y
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Amendment 12 1.8-60 l
M 23A6100AC Riandard Plant arv c TABLE 1.8 21(Continued)
INDUSTRIAL CODES AND STANDARDS APPLICABLE TO AB%R Code or Standard -
Number Year Title A%WA D100 1984" Welded SteelTanks for Water Storage CMAA70 1983 Specification for Electric Overhead Traveling Crancs 1
ICEA l
P-46-426/IEEE 1982 Ampacities Including Effect of Shield Losses for Single S 135 Conductor Solid Dielectric Power Cable 15kV through f
69kV i
P 54 440/ NEMA '1987 Ampacities of Cables in Open Top Cable Trays W 51 S-66-524/ NEMA 1982 Cross Linked Thermosetting Polyethylene Insulated Wire W7 and Cable for Transmission and. Distributor of Electrical Ercrgy i
O Amendment 12 1.841 l e
ABM .
' 234s=4c Blandard Plant n v c' TABLE 1.8 21(Continued)
INDUSTRIAL CODES AND STANDARDS !
APPLICABLE TO AB%R l Code or .
Standard !
Number Year Title
=l 1EEE i 1
279 '
1971 Criteria for Protection Systems for NGPS -
308 1980 Criteria for Class 1E Power Systems for NPGS 317 1983 ' Electrical Penetration Assemblics in Containment .
Structures for NPGS L323 1983 Qualifying class 1E Equipment for NPGS ,
3M 1974- Motors for NPGS, Type Tests' of Continuous Duty class IE .
1 338 1977 Criteria for the Periodic Testing of NPGS Safety-Systems 344 1987- Recommended Practices for Seismic Qualifications of Class 1E Equipment for NPOS .
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379 1977 Standard Application of the Single Failure Criterion to -
NPGS Safety Systems -
i 382 1985 Qualification of. Actuators for Power Operated Valve !
Assemblies with Safety Related Functions for NPP !
383 1974 Type Test of Class IE Cables; Field Splices and Connections for NPGS 384 1981 Criteria for Independenc~e of Class 1E Equipment and Circuits 387 1984 Criteria for Diesel Generator Units Applied as Standby Power Supplies for NPGS 450 1987 Practice for Maintenance, Testing, and Replacement of Large lead Storage Batteries for Generating Stations and Substations 484 1987 Recommended Practice for the Installation Design and Installation of Large Lead Storage Batteries for NPGS ,
9 Amendment t2 1.8-62'
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ABM usamsc-Af andard Plant '
arv. c -
TABLE 1A 21(Continued) I INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR Code or Standard Number Year litle IEEE (Con't)
=485 1983 Recommended Practice for Sizing Large Lead Storage Batteries for NPGS 944 1986 Recommended Practice for the Application and Testing of .
Uninterruptable Power Supplies for Power Generating j
- Station -
ISA S7.3 1981 Quality Standard for Instrument Air ,
NEMA 1 ICS1 1983 General Standards for Industrial Control -
MG1 1987 Motors and Generators O
Amendment 12
},g43 l
3-ABM 234sioo4c .
Reandard Plant arv.c- !
TABLE 1.8 21(Continued)
INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR 1 Code or i Standars Numbr Year 'Dile -
NFPA i i
10 1981 Portable Fire Extinguishers Installation -
l 10A 1973, Portable Fire Feia-ulahers Maintenance and Use 12 1985 Carbon Dioxide Fming": *- Systems !
i 13 1985 Installation of Sprinklers Systems -
14 1986 Installation of Standpipe and Hose Systems -
24 1984 Private Service Mains and their Appurtenances ' 3 37 1984 Stationary Combustion Engines and Gas Turbines 70 1987 Hsndbook 1987 72D 1986 Proprietary Protective Signaling Systems 78 1986 Lightning Protection Code 80 19P5 Fire Doors and Windows 90A 1985 Installation of Air Conditioning and Ventilating.
Systems 91 1983 Blower and Exhaust Systems 101 1985 Life Safety Code 251 1985 Fire Test, Building Construction and Materials 252 1984 Fire Tests, Door Assemblics '
255 1984 Building Materials, Test of Surf ace Burning Characteristics 321 1987 Classification ofllammable Liquids ,
801 1986 Facilities Handlirg Rad.ioactive Materials -
O Amendment 12 t.8 M
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3 ABM 234. loo c l: Rennetard Plant arv. e TABLE 1&21(Continued)- )
INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR ;
I Code or !
Standard Number har Title 1 NFPA (Con't) .
802 1988 Nuclear Research Reactors !
1%1 1979 Fire Hose -
1%3 1985 Screw Threads and Gaskets for Fire Hose Connections i
-i O L 4
I O
Amendment 12 1.8-65 l
ABWR m3c Standard Plant nry c TABLE 1.8 21 (Continued)
INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR Code or Standard Number Year litle SSPC PA 1 1972 Shop, Field and Maintenance Painting PA 2 1973 Measurements of Paint Film Thickness with Magnetic Gages SP 1 1982 Solvent Cleaning SP 5 1985 White Metal Blast Cleaning SP 6 1986 commercial Blast Cleaning SP-10 1985 Near White Blast Cleaning OTHERS TEMA C 1978 Standards of Tubular Exchanger Manufactures Association UL-44 1983 Rubber-Insulated Wires and Cables Crane Manuf actures Association of America, Specification No. 70 Aluminum Construction Manual by Aluminum Association O
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ABWR - 2346ioore Standard Plant arv c Table 1.8-22 j
- h. EXPERIENCE INFORMATION APPLICABLE TO ABWR i*
TYPE: GENERIC LETTERS l .
hsue j but, l2 air.' Illis Comment n 80-06 4/25/80 Clarification of NRC Requirement for Emergency Response Facilities at Each Site ,
- 80 30 12/15/80 Periodic Updating of Final Safety Analysis ' Interface Reports (FSARs) 80-31 12/22/80' Control of Heavy Loads 81 03 2/26/81 Implementation of NUREG 0313m Rev.1 81 04 2/25/81 Emergency Procedures and Training for Station
- Interface Blackout Events
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81 07' 2/3/81 Control of Heavy Loads 81 10 2/18/81 Post TMI Requirements for the Emergency Operations Facility 81 11- 2/22/81 Error in NUREG 0619 81 20 /1/81. Safety Concerns Associated with Pipe Breaks in the BWR Scram System 81-37 12/29/81 ODYN Code Reanalysis Requirements 81 38 11/10/81 Storage of Low Level Radioactive Wastes Interface ;
at Power Reactor Sites 82-09 4/20/82 . Environmental Qualification of Safety Related Electrical Equipment 82 21 10/6/82 Technical Specifications for Fire Protection Audits Interface l
82 22 10/30/82 Inconsistency between Requirements of 10 CFR 73.40(d) and Standard Technical Specifications s for Performing Audits of Safeguard Contingency Plans 82-27 '11/15/82 Transmittal of NUREG 0763,' Guidelines for Confirmatory l'
In Plant Tests ot Safety Relief Valve Discharges for BWR Plants," and NUREG 0783," Suppression Pool Temperature Limits for BWR Containments
- 82 33 12/17/82 Supplement 1 to NUREG-0737 O
O Amendment 12 1.8 67
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'ABWR mome, d Standard Plant arv. c , ;l Table 1.8 22 - -
EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued)- g TYPE: GENERIC LETTERS Issue l
& Dalt. Ill).g Comment j 82 39 Problems with the Submittals of 10 CFR 73.21 Interface 12/22/82 Safeguards Information Licensing Resiew 83 05[ 2/83 Safety Evaluation of ' Emergency Procedure Guidelines, Interface Revision 2, NEDO-24934, June 1982
- j 83 07 2/16/83 The Nuclear Waste Policy Act of 1982 _ Interface !
ci 83 13 3/2/83 Clarification of Surveillance Requirements for HEPA 4 Filters and Charcoal Absorber Units in Standard Technical Specifications on ESF Cleanup Systerns .l 83 28 7/8/83 Required Actions Based on Generic Implications of Salem NIV'S Events -,
i 83-33 10/19/83 NRC Positions on Certain Requirements of Appendix Interface I R to 10 CFR 50 ,
84 15 7/2/84 Proposed Staff Actions toimprove and Maintain Diesel Generator Reliability .
85 01 1/9/85 Fire Protection Policy Steering Committee Report 86 02 1/23/86 Technical Resolution of Generic issue B-19 Thermal Hydraulic Stability -
86 10 4/24/86 Implementation of Fire Protection Requirements SM)6 3/13/S7 Periodic Verification of Leak Tight Integrity of Interface Pressure Isolation Valves 87 09 6/4 87 Sr.ctions 3.0 and 4.0 of the Standard Technical 1 Specifications (STS) on the Applicability of Limiting Conditions for Operations and Surveillance Requirements 88 01 1/25/88 NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping 88 02 1/20/88 Integrated Safety Assessment Program 11 (ISAP II)
Amendment 12 O
1.8-6R
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. ABWR - nume Standard Plant nov c
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Table 1.8-22 4 O exetaissce isroaxiriox ieetic48te ro inwa cco.ii..eo) !
TYPE: GENERIC LE'ITERS Issue -i
& h Iltig Comment 88 14 8/8/88 Instrument Air Supply System Problems Affecting- 1 Safety Related Equipment !
PAST RELATED CORRESPONDENCE 4 IE Notice 87 28, Supp.1 NUREG-1275, Volume 2 -!
88-15 9/12/88 Electric Power Systems Inadequate Control Over H I
Design Process
' PAST RELATED CORRESPONDENCE IE Notice 88-45 j 88 16 10/4/8S Removal of Cycle Specine Parameter Limits from l Technical Specifications 88 18 10/20/88 Plant Record Storage on Optical Disks Interface PAST REL ATED CORRESPONDENCE i NUREG 0800 REG. Guide 1.28, Rev. 3 I 6S 20 11/23/88 Individual Plant Examination for Severe Accident Vulnerabilities 10CFR Para. 50.54(f) 88 20 8/29/89 Generic 88-20 Supplement No.1 ;
89 01 1/31/89 Implementation of programmatic Controls for Interface Radiological Effluent Technical Specifications -
in the Administrative Controls Section of the Technical Specifications and the Relocation of ,
Procedural Details of RETS to the Off site Dose Calculation blanual or to the Process Control Program
- 89 02 3/21/89 Actions to improve the Detection of Counterfeit and Interface Fraudulently hf arketed Products PAST RELATED CORRESPONDENCE EPRI-NP 5652
- Guideline for the Utilization of Commercial Grade items in Nuclear Safety Related Applications *, Bulletins 87 02 and Supplements 1 and 2, 88 05 and Supplements 1 and 2,8810 IE Notiocer 87 66, 8819,88 35,88-46 and Supplements 1 and 2,88-48 and Supplement 1,88 97 89 04 4/3/89 Guidance on Developing acceptable Inservice Subsection f Testing Program 19B.2.2 l 89 06 4/12/S9 Task Action Plan item 1.D.2-Safety Parameter Display System CFR 50.54(f)
Amendment l' l.8-69 '
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ABWR amme Standard Plant arv. c Table 1.8 22 EXPERIENCEINFORMATION APPLICABLETO ABWR'(Continued) h TYPE: GENERIC LE1TERS issue b'n,
. Dalf. . Illlt Comment 89 07 4/28/89 Power Reactor Safeguards Contingency Planning for Surface Vehicle Bombs 89 07 4/21/89 Power Reactor Safeguards Contingency Plann!ng for .
Suppi Surface Vehicle Bombs 89 0S 5/2/89 Erosion /Corrision Induced Pipe Wall Thinning 89 10 6/28/89 Safety Related Motor Operated Valve Testing Interface and Surveillance 89 11 6/30/89 Resolution of Generic issue 101
- Boiling Water Subsection Reactor Water Level Redundancy 19B.2.16 89 13 7/18/89 Service Water System Problems Affecting Safety _ interface Related Equipment -
89 14 8/21/S9 Line item Improvements in Technical Specifications.
Removal of the 3.25 Limit on Extending Surveillance Intervals 89 15 8/21/S9 Emergency Response Data System Interface 89 18 9/6/69 Resolution of USl A 17, Systems Interactions Subsection 19B.2.3 89-19 9/20/89 Request for Action Related to Resolution of Subsection Unresolved Safety issue A 47
- Safety implication 19B.2.5 of Control Systems in LWR Nuclear Power Plants" Pursuant to 10 CFR 50.54(f) 89 22 10/19/89 Potential for Increased Roof Loads and Plant Area Plood Runoff Depth at Licensed Nuclear Power Plants Due To Recent Change in Probable Maximum Precipit.
ation Criteria Developed By The National Weather Service O
Amendment 12 1.8 70 -
i' LABWR' usime Standard Plant REV. C Table 1.8 22 O exetartsceistonairioxieeticisteroiswa cco ti. eo)
TYPE: IE BULLETINS Issue
& Dats,, Iglg Comment C.
79 08 4/14/79- Events Relevant to BWR Identified During TMI incident - q 80 01. 1/11/80 ADS Valve Pneumatic Supply >
-l 80 03 2/6/80 Loss of Charcoal from Absorber Cells ,
80 06 3/13/80 ESF Reset Controls f
80-08 4/7/80 Containment Lines Penetration Welds Interface !
80 10 5/6/80 Non Radioactive System Potential for Unmonitored Interface Release 80 12 5/9/S0 Decay Heat Removal System Operability Interface 80 13 .5/12/80 Cracking in Core Spray Spargers 80 15 6/18/80 Possible Loss of Emergency Notification System .
with Loss of Offsite Power 80-20 7/31/80 Westinghouse Type W 2 Switch Failures ;
80-21 11/6/80 Valve Yokes Supplied by Mole Interface 80 22 9/11/80 Automatic Industries, model 200 500-008 Scaled ~ Interface ,
Source Con.
80 24 11/21/80 Prevention of Damage due to H30 Leakage NUREG/CR 4524
.inside Containment 80-25 12/19/80 Operating Problems with Target Rock SRVs at BWRs 81 01 1/27/81 Surveillance of Mechanical Snubbers 81 02 4/9/81 Failure of Gate Type Valves to Close Interface 81 02, 8/19/81 Failure of Gate Type Valves to Close Interface Suppl Against Differential Pressure 81 03 4/10/81 Flow Blockage of Cooling Water to Safety System Interface i
O Amendmcm 12 1.8 "11
e MM .-
' :23A61ooAc :
Standard Plant arv e l' -
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_ Table L8 22 l
~
EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued) j j
" TYPE: IE BULLETINS i issue E ' E!L Dalt. 31tle Comment i n 82 04 ~12/3/S2 Deficiencies in Primary Containment Electrical Interface '
' Penetration Assemblies i :83-06 7/22/83 Non Conforming Materials Supplied - Interface . !
by Tube-Line Corp.
84 01 2/3/84 Cracks in Boiling Water Reactor Mark I Containment Vent Header 84 03 8/24/84 Refueling Cavity Water Seal '
85 03 11/15/85 Motor Operated Valve Common Mode Failures During Interface Plant Transients Due to improper Switch Settings 85 03, 4/27/88- Motor Operated Valve Common Mode Failure During ' Interface
=
Supp1 Plant Transients Due to improper Switch Settmgs :!
PAST RELATED CORRESPONDENCE ;
IE Bulletin 85-03,IE Notice 86-29, and 87 01 '
86-01 5/23/86 Minimum Flow Logic Problems That Could Disable '
RHR Pumps 1 86 03 10/8/86 Potential Failure of Multiple ECCS Pumps Due to Single Failure of Air-Operated Valve in Minimum Flow Recirculation Line 87 01 7/9/87 Thinning of Pipe Walls in Nuclear Power Plants 87 02 11/6/87 Fastener Testing to Determine Conformance with_ interface Applicable Material Specifications
- 87 02, 4/22/88 Fastener Testing to Determine Conformance with Interface i Supp1 Applicable Material Specifications PAST RELATED CORRESPONDENCE IE Notice 88-17 1 87 02, 6/10/88 Fastener Testing to Determine Conformance with Interface Supp 2 Applicable Material Specifications 88 04 5/5/88 Potential Safety Related Pump Loss PAST RELATED CORRESPONDENCE IE Notice 87 59 k
8847 6/15/88 Power Oscillations in Boiling Water Reactors (BWRs)
PAST RELATED CORRESPONDENCE m IE Notice 88-39 -
Amendment 12 1.B-72 a
M
a ABWR meme :!
Standard Plant REV C -l Table 1.8 22-J EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued)
TYPE: IE BULLETINS Issue I
- b'Q, 11811. Illis Comment 88 07, 12/30/88 Power Oscillations in Boiling Water Reactors (BWRs) ~ .
Supp1- !
.i E
I .
I i
s i
O Amendment 12 1.8-73
ABM - 334 iooxc .-
Standard Plant any e Tabls1.8 22 EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued)
TYPE: IE INFORMATION NOTICES g
Issue Es lhti.t. Illit
- Comment 80-12 3/31/80 Instrumentation Failure Causes PORV Opening 80-21 5/16/80 Anchorage & Support of Safety Related Electrical Equipment 80-22 5/28/80 Breakdowns in Contamination Control Programs . - Interface 80-40 11/7/80 . Excessive N SUPPl y Pressure 2
80 42 11/24/80 Effect of Radiation on Hydraulic Snubber Fluid 81 05 3/13/81 Degraded DC Systems at Palisades Interface 81 07 3/16/81 Potential Problem with Water Soluble Purge Dam Interface Materials used during inert Gas Welding 81 10 3/25/81 Inadvertent Containment Spray Interface-81 20 7/13/81 Test Failures of Electrical Penetrations 81 21 7/21/81 Potential Loss of Direct Access to Ultimate - Interface Heat Sink 81 31 10/8/81 Failure of Safety injection Valves Interface.
81 38 12/17/81 Potential Significant Equipment Failures Interface Resulting from Contamination of Air Operated Systems 82 03 3/22/82 Environmental Tests of Electrical Terminal Block 82 10 3/3/82 Following Up Symptomatic Repairs Interface .
82 12 4/21/82 Surveillance of Hydraulic Snubbers 82 22 7/9/82 Failures in Turbine Exhaust Lines 82 23 7/16/82 Main Steam Isolation Valve Leakage 82 25 7/20/82 Failures of Hiller Actuaton Upon Gradual Loss of Air Pressure 82 32 8/19/82 Contamination of Reactor Coolant System by Organics Interface Amendment 12 1,g.74 O
_____ _ _ . . _ . . . . . . . . . . . = . .
!ABWR mame '
j Standard Plant RIV C
( Table 1.8 22' j EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued) j
'IYPE: IE INFORMATION NOTICES Issue b q, Dalg, 1111g Comment .!
82 40 9/22/82 Deficiencies in Primary Containment Electrical !
Penetration Assemblics 32 43 11/16/82 Deficiencies in LWR Air Filtration / Vent System 82-49 12/16/82 Correction for Sample Conditions for Air Interface 1 ;
& Gas Monitor :
83-03 1/28/83 Calibration of Liquid LevelInstruments Interface =4 83 07 3/7/83 Nonconformities with hlaterials Supplied by - Interface Tube Line Corp.
83 0S 3/9/S3 Component Failures Caused by Elevated l DC Control Voltage 83 17 3/31/83 Electrical Control Logic Problem Resulting- !
in inoperable Auto Start of Emergency Diesel Generator 83 30 $/11/83 hiisapplication of Generic EOP Guidelines Interface 1 83-35 5/31/83 Fuel blovement with Control Rods Withdrawn Interface !
at BWRs
{
83-41 7/1/83 Damage to Redundant Safety Equipment from - :
i Backflow Through the Equipment -
83-46 7/11/83 Common hiode Valve Failures Degrade Surry's Interface -
Recirculation Spray Subsystem 83 50 8/1/83 Failure of Class IE Circuit Breakers to Close -
i 83 51 8/5/83 Diesel Generator Events 83 62 9/26/83 Failure of Toxic Gas Detectors Subsection 19B.2.31 ,
83 64 9/29/83 Lead Shiciding Attached to Safety Related Systems Interface 83 70 10/25/83 Vibration Induced Valve Failures 83 70 3/4/85 Vibration Induced Valve Failures Supp1 O Amendment 12 1.g.73
. ABWR m-c Standard Plant arv c L TaNe 1.8 22 EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued) g '
TYPE: IE INFORMATION NOTICES l: Ett. D. alt. Iltle CDantal ,
\
83 72 10/28/83 Environmental Qualification Testing Experience -
83 75 11/3/83 Improper Control Rod Manipulation Interface i i L 83 80 11/23/83 Use of Specialized
- Stiff' Pipe Clamps .
64 09 2/13/84 Lessons Learned in om NRC Inspections of Fire Protection Safe Shutdown Systems (10CFR50, App R) i 844N, 3/7/84 Lessons Learned from NRC Inspections of Irire
, Rev.1 Protection Safe Shutdown Systems (10CFR50, App. R) l i St 10 2/24/84 Motor Operating Valve Torque Switches Set Below Interface the Manufacturers Recommended Value l
84 17 3/5/84 Problems with Liquid Nitrogen Cooling Components Below the Nil Ductility Temperature 84 22 3/29/84 Deficiency in Comsip, Inc. Standard Bed Catalyst 84 23 4/5/84 Results of the NRC Sponsored Qualification Methodology on ASCO Solenoid Valves 84 32 4/18/84 Auxiliary Feedwater Sparger and Pipe Hanger Damage l 84 35 4/23/84 BWR Post Scram Drywell Pressurization 84 3S 5/17/84 Problems With Design, Maintenance, and Operation of Offsite Power Systems 84 47 6/15/84 Environmental Qualification Tests of Electrical Terminal Blocks 84 67 8/17/84 Recent Snubber Insenice Testing With High Interface Failure Rates 84 69 8/20/S4 Operation of Emergency Diesel Generators Interface 84 69, 2/24/S6 Operation of Emergency Diesel Generators Interface Supp.1 84 70 9/4/84 Reliance on Water LevelInstrumentation Interface with a Common Reference Leg Aniendment 12 1 & 76 9
f
i ABWR mame Standard Plant nrv. c
! rs Table 1.8 22 d EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued) n'PL IE INFORhiATION NOTICES Issue h Dalt. Ihlt Comment 84 70, 8/26/85 Reliance on Water Levelinstrumentation Interface Supp I with a Common Leg 84 76 10/19/84 Loss of All AC Power 64 87 12/3/84 Piping Thermal Deflection Induced by Stratified Flow 84 88 12/3/84 Standby Gas Treatment System Problems 84 93 12/17/84 Potential for Loss of Water from the Refueling Cavity 85 08 1/30/85 1. vustry Experience on Certain hinterials Used in Safety Related Equipment 85 13 2/21/85 Consequences of Using Soluble Dams Interface 85 17 4/1/85 Possible Sticking of ASCO Solenoid Valves 8517, 10/1/85 Possible Sticking of ASCO Solenoid Valves Supp.1 85 24 3/26/S5 Failures of Protective Coatings in Pipes Interface and Heat Exchancers 85 25 4/2/85 Consideration of Thermal Conditions in the Design ,
and Installativn of Supports for Diesel Generator Exhaust Silencers r
85 28 4/9/85 Partial Loss of AC Power and Diesel Generator l Degradation l
l 85 30 4/19/85 hiicrobiologically induced Corrosion of Containment l Service Water System 85 32 4/22/85 Recent Engine Failures of Emergency Diesel Generators 85 33 4/22/85 Undersized Nozzle to Shell Welded Joints in Tanks and Heat Exchangers Constructed Under the Rules of the AShiE Boiler and Pressi:re Vessel Code 85 34 4/30/85 Heat Tracing Contributes to Corrision Failure of Interface Stainless Steel Piping 85 35 4/30/85 Failure of Air Check Valves to Seat
! /]
l .Q) i
(
, Amendment 12 1.8 77
ABWR mme -
Standard Plant REY C Table 1.8 22 .
EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued)
TYPE: 10 INFORMATION NOTICES g
Issue St, Datt. Illle Comment 8505 5/17/88 Faibre of Air Check Vahts to Seat-Supp 1 85 47 6/18/85 Potential Effect of Line induced Vibration on Certain
- Target Rock Solenoid-Operated Valves 85 51 7/10/85 Inadvertent Loss ofImproper Actuation of Interface Safety Related Equipment 85 59 7/17/85 Valve Stem Corrision Failures s
85 66 8/7/85 Discrepancies Between As Built Construction Interface Drawings and Equipment installations 85 76 9/19/85 Recent Water Hammer Events 85 77 9/2C/65 Possible Loss of Emergency Notification Inteface System Due to Loss of AC Power 85 81 10/17/85 Problems Rmiting in Erroneously High Interface Reading With Thermoluminscent Dosimeters 85 81 10/30/85 Inadequate Inservice Testing of Main Steam isolation Valves 85 85 10/31/85 Systems Interaction Event Resulting in Reactor System Safety Relief Valve Opening Following a Fire Protection Duluge System Malfunction 85 F6 11/5/S5 Lightning Strikes at Nuclear Power Generating Stations 85 87 11/18/85 llazards of Incrting Atmospheres Interface 85-89 11/19/85 Potential Loss of Solid State Instrumentation Subsection Following Failure or Control Room Cooling 19B 2 31 85-90 11/19/85 Use of Scaling Componds in an Operating Plant Interface 85 91 11/27/85 Load Sequencers for Emergency Diesel Generators Interface Amendmcot 12 1 & 78 O
ABWR m um^c Standard Plant an c Table 1.8 22 EXPERIENCEINFORMATION APPLICABI.t TO ABWR (Continued)
TYPE: IE INFORMATION NOTICES Issue FfL Delt. Illig Comnant 85 92 12/2/85 Surveys of Wastes Before Disposal From Nuclear Interface Reactor Facilities 85 94 12/13/85 Potential 'or Loss of Minimum Flow Paths Leading i to ECCS Pump Damage During a LOCA 85 4 12/23/85 Temporary Strainers Left Installed in Pump Suction Interface Piping !
66 01 3/6/86 Failure of Main Teedwater Check Valves Causes Loss of j Feedwater System lntegrity and Water Hammer Damnge .;
86 03 1/14/86 Potential Deficiencies in Environmental Qualification of Limitorque Motor Yahe Operator Wiring j.
86 09 2/3/86 Failure of Check and Stop Valves Subjected to Low Flow Conditions 86 10 2/13/86 Safety Parameter Display System Malfunctions !
86 29 4/25/86 Effects of Changing Valve Motor Operator Switch Interface Settings PAST P. ELATED CORRESPONDENCE:
IE Bulletin 85 03 86-30 4/29/86 Design Limitations of Gascous Effluent Monitoring Systems l 86-39 5/20/86 Failures of RHR Pump Motors and Pump Internals 86 43 6/10/86 Problems with Silver Zeolite Sampling of Airborne Interface Radiolodine 86-4S 6/13/86 Inadequate Testing of Boron Solution Concentration in the Standby Liquid Control System 86 50 6/18/86 Inadequate Testing to Detect Failures of Safety. !
Related Pneumatic Components or Systems PAST RELATED CORRESPONDENCE:
IE Notices 82 25,85-35,85-84,85 94 86 51 6/18/86 Excessive Pneumatic Leakage in the Automatic Depressurization System PAST REL ATED CORRESPONDENCE: I 1E Bulletins 80 01,80 25;IE Notice 85 35; 1E Inspection Report 50-458/S418 (8/16/84) l Amendment 12 1.8 79 j z
l ABWR 23A6100AC .
k Standard Plant ._- ""
i Table 1.8 22 EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued)
TYPE: IE INFORMATION NOTICES 1sene i EO, Dalt. Ihlt ' Comment 86 53 6/26/86 Improper Installation of Heat Shrinkable Tubing Interface
% 57 7/11/86 Operating Problems With Solenoid Operated Valves at Nuclear Power Plants 86 60 7/28/86 Unanalyzed Post LOCA Release Paths PAST RELATED CORRESPONDENCE:
NUREG-0737 i 86 68 8/15/86 Stuck Control Rod
% 70 8/18/86 Potential Failure of All Emergency Diesel Generators 86 71 8/19/86 Recent identified Problems With Limitorque Motor Operators PAST RELATED CORRESPONDENCE: '
IE No' ice 86 03 86 76 8/20/86 Problems Noted in Control Room Emergency Subsection Vr.ntilation Systems 19B.2.31 -
i 13ST RELATED CORRESPONDENCE:
Item 111 D,3.4 of NUREG 0737 Generic Issue 83,IE Notice 85 89 - !
86 83 9/16/86 Underground Pathways into Protected Areas, Vital Interface j Areas, Material Access Areas, and Controlled Acetss '
Areas PAST RELATED CORRESPONDENCE:
NUREG 0908, ANSI 3.3 I 86 87 10/10/86 Loss of Offsite Power Upon An Automatic Bus Transfer 86 89 10/16/86 Uncontrolled Rod Withdrawal Because of A Single Failure l
%% 11/20/86 Heat Exchanger Fouling Can Cause inadequate interface Operability of Senice Water Systems PAST RELATED CORRESPONDENCE:
IE Bulletin 8103,IE Notice 8121 E100 12/12/86 Loss of Offsite Power to Vital Buses at Salem 2 86 104 12/16/86 Unqualified Butt Splice Connectors identified I in Qualified Penetrations j
Amendment 12 1.8-80 l
ABWR mme Standard Plant arv. e Table 1.8 22 O EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued) n1PE: IE INFORMATION NOTICES lasue b'n, Dak. Iltle ' Comraent 86 106 12/16/86 Feedwater Line Break 86-106, 2/13/87 Feedwater Line Break Supp.1 PAST RELATED CORRESPONDENCE:
IE Notice 82 22 EPRI Report NP 3944,4/8$
86 106, 3/18/87 Feedwater Line Break Supp.2 86 106, 10/10/88 Feedwater Line Break Supp.3 86 109 12/29/86 Diaphragm Failure in Scram Outlet Valve interface Causing Rod Insertion PAST RELATED CORRESPONDENCE: j IE Notice 86 08 87 06 1/30/87 Loss of Suction to Low Pressure Service Interface Water System Pumps Resulting From Loss of Si) hon 87 08 2/4/87 Degraded Motor Leads in Limitorque DC Motor Operators PAST RELATED CORRESPONDENCE:
(Unrelated problems invohing wiring i stalled in Limitorque motor actuators) !
IE Notices 83 72,86 03 and 86-71 87 0') 2/5/87 Emergency Diesel Generator Room Cooling Deficiency PAST RELATED CORBfSPONDENCE:
IE Notice 86 50,86 51 and 86 89 87 10 2/11/87 Potential for Water Hammer During Restart of Residual Heat Removal Pumps ;
PAST RELATED CORRESPONDENCE:
AEOD/E309,4/83 87 13 2/24/87 Potential For High Radiation Fields Following Loss of Water From Fuel Pool 4 I
PAST RELATED CORRESPONDENCE:
IE Notice 84 93,IE Bulletin 84 03 i
Amendment 12 1.8-81
i ABWR m-c .
Standard Plant krv c Table 1.8 22 EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued) h TYPE: IE INFORMATION NOTICES Issue
& Dalt- 21tjg Comment 87 14 3/23/87 Actuation of Fire Suppression System Causing inoperability of Safety Related Ventilation Equipment PAST RELATED CORRESPONDENCE:
IE Notice 83 41,85-85,86106 Supp. 2 87 28 6/22/87 Air Systems Problems at U.S. Light Water Reactors PAST RELATED CORRESPONDENCE:
AEOD C701 87 28, 12/28/88 Air Systems Problems at U.S. Light Water Reactors Sup.1 PAST RELATED CORRESPONDENCE:
AEOD C701 NUREG 1275 Vol.2 87 36 8/4/87 Significant Unexpected Erosion of Feedwater Lines i PAST RELATED CORRESPONDENCE: i
!E Notice 82 22,86106 plus Supp.1&2 IE Bulletin 87 01 87 43 9/8/87 Gaps in Neutron Absorbing Materialin High. j Density Spent Fuel Storage Racks PAST RELATED CORRESPONDENCE:
EPRI NP 4724 87 49 10/9/87 Deficiencies in Outside Containment Flooding !
Protection i 87 50 10/9/87 Potential LOCA at High and Low Pressure Interfaces from Fire Damage 87 59 11/17/87 Potential RHR Pump Loss 4
88 01 1/27/8S Safety injection Pipe Failure 88 04 2/5/88 Inadequate Qualification and Documentation '
of Fire Barrier Penetration' Seals l PAST RELATED CORRESPONDENCE: i 10CFR50 Appendix R, Appendix A to BTP APCSB 9.51, !
NUREG 0300, ASTM E 119, BTP CMEB 9.51, '
Generic Letter 8610 8S 04, 8/9/88 Inadequate Qualification and Documentation Supp.1 of Fire Barrier Penetration Seals 88 05 2/12/8S Fire in Annunciator Control Cabinets Amendment 12 1.8-82 l
ABWR msime Standard Plant arv c Table 1.8 22 O exetaitxctisronutrionierticiattroxawa cco ti ea>
TYPE: IE INFORMATION NOTICES i Issue h Dalt. Ilth Comment 88 12 4/12/88 Overg* casing of Electrical Motor Bearings Interface PAST RELATED CORRESPONDENCE: j LER 387/84 036 ]
i 88 13 4/18/88 Water Hammer and Possible Piping Damage .
Caused by Misapplication of Kerotest Packless Metal Diaphragm Globe Valves SS 17 4/22/8S Summary of Responses to NRC Bulletin 87 01,
' Thinning of Pipe Walls in Nuclear Power Plants
- PAST RELATED CORRESPONDENCE:
IE Bulletin 87 01 IE Notice 82 22,86106,87 36 SS 21 5/9/8S Inadvertent Criticality Events at Oskarshamn Interface and at U.S. Nuclear Power Plants 88 24 5/13/88 Failures of air-Operated Vakes Affecting Safety Related Systems PAST RELATED CORRESPONDENCE:
IE Notice 87 28 & Supp.1, NUREG 1275
) SS 27 5/18/8S Deficient Electrical Terminations IdentiGed Interface 4 in Safety Realted Components 8S 35 6/3/8S Inadequate Licensee Performed Vendor Audits Intciface PAST REALTED CORRESPONDENCE:
1- IE Bulletin 88-05 l
SS 37 6/14/88 Flow Blockage of Cooling Water to Safety Interface l System Components l PAST RELATED CORRESPONDENCE:
l IE Notice 8121,86 %
I 88 39 6/15/88 LaSalle Unit 2' Loss of Reci culation Pumps With Power Oscillation Event PAST RELATED CORRESPONDENCE:
Generic issue B 19 Generic Letter 86 02 l
l 88 43 6/23/SS Solenoid Valve Problems PAST RELATED CORRESPONDENCE:
l IE Notices 8517 & Supp.1,86-57 l r IE Circular 81 14 I
l Amendment 12 1.8-83 l
ABWR mac Standard Plant REV C Table 1.8 22 EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued) h TYPD IE INFORMATION NOTICES issue b'0, Dalt Illlt Comment 88 51 7/21/88 Failures of Main Steam Isolation Valves 88 61 8/11/88 Control Room Habitability Recent Reviews Subsection of Operating Experience 19B.2.31 88 63 8/15/88 High Radiation Hazards from Irradiated incore Interface Detectors and Cables 8S-65 8/18/8S Inadvertent Drainings of Spent Fuel Pools 88 70 8/29/8S Check Valve Inservice Testing Program Deficiencies PAST RELATED CORRESPONDENCE:
IE Notice 86-01 Generic Letter 87 06 88 72 9/2/88 Inadequacies in the Design of DC Motor.
Operated Valves 88-76 9/19/88 Recent Discovery of a Phenomenon Not Presiously Considered in the Design of Secondary Containment Pressure Control PAST R 4 \TED CORRESPONDENCE:
NUREG 0800 88 77 9/22/8S Inadvertent Reactor Vessel Overfill 88 81 10/7/88 Failure of AMP Window Indent Kynar Splices and Thomas and Betts Nylon Wire Caps During Emirontrental Oualification Testing SS 85 10/14/88 Broken Retaining Block Studs on Anchor Darling Check Valves 4
88 86 10/21/88 Operating with Multiple Grounds in Direct Current Distribution Systems and Supplement 1 88 89 '/88 Degradation of Kapton Electricallnsulation PAST RELATED CORRESPONDENCE:
IE Notices 87 0S,8716 8S 92 11/22/88 Potential for Spent Fuel Pool Draindown 88 95 12/8/8S Inadequate Procurement Requirements imposed Interface by Licensees on Vendors 89 01 1/4/89 Valve Body Erosion PAST RELATED CORRESPONDENCE:
IE Notice 8817 Amendment 12 t.8-84
ABWR 2mme i Standard Plant RI'Y C i
Table 1.8 22 ;
O exeeniexceistonuirioxiretic^=teroinwa cce ti oea) i i
TYPE: 10 INFORMATION NOTICES "
I Issue b'n, Dgg, Igle Comment 89 04 1/17/89 Potential Problems from the Use of Space Heaters Interface j 89-07 1/25/89 Failures of Small Diameter Tubing in Cgptrol Air, , 1 Fuel Oil, and Lube Oil Systems Which Render Emer. l gency Diesel Generators Inoperable l 89 0S 1/26/89 Pump Damage Caused by Low Flow Operation 89 10 1/27/89 Undetected installation Errors in Main Steam '
Line Pipe Tunnel Differential Temperature-Sensing Elements at Boiling Water Reactors 89 11 2/2/89 Failure of DC Motor Operated Valves to Develop ,
Rated Torque Because of Improper Cabling Sizing 89 14 2/16/89 Inadequrite Dedication Process for Commerciel Grade Components Which Could Lead to Common Mode Failure of a Safety System f 89 16 2/16/89 ExcessivcVoltage Drop in DC Sptems .
PAST RELATED CORRESPONDENCE: -l Generic Letter 88-15
, 89 17 2/22/89 Contamination and Degradation of Safety Related Battery Cells 89 20 2/24/89 Weld Failures in a Pump cf Byron Jackson Design 89 21 2/27/89 Changes in Performance Characteristics of Molded.
Case Circuit Breakers 89 26 3/7/89 Instrument Air Supply to Safety Realted Equipment PAST RELATED CORRESPONDENCE-Generic Letter 8814 89 30 3/15/89 High Temperature Emironments at Nuclear Power Plants 89-36 4/4/89 Excessive Temperatures in Emergency Core Cooling System Piping Located Outside Containment c
89 37 4/4/89 Proposed Amendments to 40CFR Part 61, Air Emission Standards for Radionuclides 89 39 4/5/89 List of Parties Excluded from Federal Procurment Interface p of Non procurement Programs ,
Q ,
Am ndment 12 LS-83
ABWR 2miw4c Standard Plant n ov. c Table 1,8 22 EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued) h!
'IYPE: IE lhTORMATION NOTICES issue F.D., Dalt.tjg Comment 89 52 6/8/89 Potential Fire Damper Operational Problems 89 61 8/30/89 Failure of Borg Warner Gate Valves to Close Against Differential Pressure 89 63 9/5/89 Possible Submergence of Electrical Circuits Located Above the Flood Level Because of Water Intrusion and Lack of Drainage - 89 64 9/7/89 Electrical Bus Bar Failures Interface 89 66 9/11/89 Qualifica.,on Lite of Solenoid Valves 89 68 9/55/89 Evaluation of lastrument Setpoints During Interface biodificatiorn 89 69 9/29/89 Loss of Thermal hlargin Caused by Channel Box Bow Imerface i 89 70 10/11/89 Possible Indications of Misrepresented Vendor Products - Interface - 89-71 10/19/89 Diverr. ion of the Residual Heat Removal Pump Seal Cooling Water flow During Recirculation Operation Following a Loss of Coolant Accident 89 72 10/24/89 Failure of Licensed Senior Operators to Classify Interface Emergency Events Properly 89 73 11/1/89 Potentail Overpressurir.ation of Low Pressure Systems Interface 89 76 11/21/89 Biofouling Agent: Zebra hiussel Interface 89 77 11/21/89 Debris in Containment Emergency Sumps and Incorrect Screen Configurations 89 79 12/1/89 Degraded Coatings and Corrosion of Steel Containment Vessels 89 80 12/1/89 Potential for Water Hamtr.<.., Thermal Stratification, and Steam Binding in High Pressure Coolant Injection Piping 89 81 12/6/89 Inadequate Control of Temporary hiodifications Interface to Safety Related Systems O Amendment 12 1.8-86
c ABWR m am^c Standard Plant RFY C Table 1.8 22
- EXPERIENCEINFORMATION APPLICABLETO ABWR (Continued)
O: 1YPE: IE INFORMATION NOTICES issue
& DEt. '[1gg . Comment ; 89 83 12/11/89 Suste.ined Degraded Voltage on the Offsite Electrical Interfr :.
Grid and less of Other Generating Stations as a Result of a Plant Trip 89 87 12/19/89 Disabling of Emergency Diesel Generators by Their Neutral Ground Fault Protection Circuitry i 89 88 12/16/89 Recent NRC Sponsored Testing of Motor Operated Vahu l i 4 i l. i !O i i i An.<ndment 12 1&B7 4
ABWR m si m e Standard Plant , _ nrv c Table 1.8 22 EXPERIENCEINFORMATION APPLICABLETO ABWR (Continued)' h! TYPE: IE CIRbULARS Issue ) b'n, j).g.lg, Illig Comment 1 l 80 03 3/6/80 Protection from Toxic Gas Hazards Interface ' 80-05 4/1/80 Emergency D/G Lube Oil Interface ! 80 08 4/18/80 RPS Response Time ,
]
80 07 4/23/80 Problems with Plant Internal Communications Systems Interface 80 10 4/29/80 Failure to Maintain Emironmental Qualification laterface of Equipment 80 11 5/13/80 Emergency Diesel Generator Lubs Oil Cecler Failures Interface , S0 14 6/24/80 Radioactive Contamination of Demin Water System Interface 80 18 8/22/80 10 CFR 50.59 Safety Evaluation foI Changes Interface , to Radioactive Waste Treatment Systems 81 03 3/2/81 Inoperable Seismic hionitoring Instrument Interface 81 05 3/31/81 Self Aligning Rod End Bushing fer Pipe Supports Interface 81 07 5/14/81 Contr01 of Radioactivity Contaminated hinterial Interface 81 0S 5/29/81 Foundation hinterials Interface SI 09 7/10/81 Containment Effluent Water 61 11 7/24/81 Inadequate Decay Heat Removal Interface 81 13 9/25/81 Torque Switch Electrical Bypass Circuit Interface
+
1 81 14 11/5/81 h1ain Steam Isolation \ alve Failures to Close Interface . O~ Amendment 12 1.8-88
\ i j 3
L ,. ,
ABWR 2mme Standard Plant arv c Table 1.8 22 EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued)
'IYPE: NUREG issue b% Dalt. Illit Comment !
l . 0313 6/88 Technical Report on Material Selection and Processmg Rev. 2 Guide. lines for BWR Coolant Pressure Boundary Piping j 1 (071 10/78 Task Action Plans for Generic Activities Category A
- j 0471 6/78 Generic Task Problem
Description:
Category B, : C & D Tasks 0588 12/79 Interim Staff Position On Emironmental Qualification of Safety Related Electrical Equipment 0619 4/80 BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking 0744 10/82 Resolution of the Task A 11 Reactor Vessel Materials Rev.1 Toughness Safety Issue , 0813 9/81 Draft Emironmental Statement Related to the rs Operation of Calloway Plant, Unit No.1 1 0977 3/83 NRC Fact FindingTask Force Report on the ATWS Events at the Salem Nuclear Generating Station, Unit 1, on February 22 and 25,1983 1150 6/89 Severe Accident Risks: An assessment for Five U.S. Nuclear Power Plants, Vol.1 & 2. ! 1161 5/80 Recommended Revisions to USNRC Seismic Design Subsection ! Criteria 19B.2.27 ! 1174' $/89 Evaluation of Systems Interactions in Nuclear Subsection l Power Plants 19B.2.3 . 1212 6/86 Statvs of Maintenance in the US Nuclear Power i Industry 1985 Vol.1,2 1217 4/88 Evaluation of Safety implications of Control Systems Subsection in LWR Nuclear Power Plants Tecimical Findings 19B.2.5 1 I Related to USl A.47 1218 4/88 Regulatery Analys4 for Resolution of USI A 47 Subsection 19B.2.5 l 1 l'.'i'9 8/89 Regulatory Analysis for Resolution of USl A 17 Subsectica 1 19B.2.3 V 19B2.27 ) ( I Amendment l'!
. 1 A-89 l l
l
. :_ -_ ) ..
'ABWR 2mme Standard Plant nrv c 1
Table 1.8 22 i 1 EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued) $ '
'IYPE:NUREG Issue Eit. Dalf. Ii11t Castittent I?ti 9/89 Regulatory Analysis for USI A 40 Subsection
- 19B.2.27 i
1289 11/88 Regulatory and Backfit Analysis: Unresolved Safety Subsection Issue A 45, Shutdown Decay Heat Removal Requirements 19B.2.29 1353 4/89 Regulatory Analysis for the Resolution of Generic Subsection Issue 82, *Beyond Design Basis Accidents in Spent 19B.2.14 Fuel Pools' 1370 9/89 Resolu' ion of USl A-18 Subsection ! 19B.2.6 CR 3922 1/85 Survey and Evaluation of System Interaction Events Subsection t and Sources Vol.1,2 - 19B.2.3 CR-4261 3/86 Assessment of Systems Interactions in Nuclear Power Subsection Plants 19B.2.3 CR 4262 5/85 Effects of Control System Failures on Transients, j Accidents at a GE BWR Yol.1 and 2 CR 4387 12/85 Effects of Control System Failures on Transient and i Accidents and Core Melt Frequencies at a GE BWR' CR 4470 5/86 Survey and Evaluation of VitalInstrumentation and Control Power Supply Events CR 5055 5/88 Atmospheric Diffusion for Control Room Habitability Subsection Assessment 19B.231 CR 50SS 1/89 Fire Risk Scoping Study: Investigation of Nuclear Power Plant Fire Risk, including Previously Unaddressed Issues. - CR 5112 3/89 Evolution of Boiling Water Peactor Water . Level Subsection Sensing Line Break and Single Failure 19B.2.16 CR 5230 4/89 Shutdown Decay Heat Removal Analysis: Plant Case Subsection Studies and SpecialIssues 19B.2.29 CR 5347 6/89 Recommendations for Resolution of Public Comments Subsection on USl A 40 - 19B.2.27 O Amendmeni 12 1890
l d. MM 23A6100AC Standard Plant rm c D Table 1.91
SUMMARY
OF ABWR STANDARD PIANT INTERFACES WITH REMAINDER OF PIANT ITEM IN'ITRFACE ! NO. SUBJECT TYPE SUBSECTION j 1.1 Standard review plan sections for remainder of Confirmatory 1.8.4 plant identified as ' Interface
- in Table 1.819 ,
1.2 Applicability of regulatory guides for remainder Confirmatory 1.8.4 of plant indentified as ' Interface'in Table 1.8 20 13 Applipbility of Experienec Informrtion Confirmatory / 1.8.4 for remainder of plant identified as Procedural l'
- Interface
- in Table 1.8-22 1.4 Emergency procedures and emergency procedures Procedural 1A3.1 j training program 1.5 Procedures for temoving salny.related systems Procedural 1A3.2 from senice 1.6 Inplant radiatiom monitoring Procedural 1A33 2.1 Envelope of ABWR Standard Plant Site Design Design & 2.2.1 Parameters Confirmatory 12 Standard Review Plan Site Characteristics Confirmatory 2.2.2 ;
3.1 Site Specific Design Basis Wind Confirmatory 333.1 ! 3.2 Site Specific Design Basis Tornado Confirmatory '333.2 33 Effect of remainder of plant structures, Confirmatory 3333 . systems and components not designed to tornado 3 loads 3.4 Flood Elevation Design 3.43.1 3.5 Ground Water Elevation Design 3.43.2 3.6 Protection of ultimate heat sink Confirmatory 35.4.1 3.7 Missels generated by natural phenomena from Confirmatory 3.5.4.2 remainder of plant 3.8 Site proximity missiles and aircraft hazards Confirmatory 3.5.43 3.9 Protection against secondary missiles inside Confirmatory 3.5.4.4 containment Amendment 14 1A2
l
~ 23A6100AC' Standard Plant ,
am e Table 1.91 i SUMMXT DF ABWR STANDARD PIANT INTERFACES NM AEMAINDER OF PLANT (Continued) ; ITEM INTERFACE I NO. SUBJECT TFI'E SUBSECTION l 3.10 Details of pipe break analysis results Confirmatory 3.6.4.1 and protection methods 3.11 leak before brut analysis results Outfirmatory 3.6.4.2 I 3.12 Foundation Waterproofing Confirmatory 3.8.6.1 ! 3.13 Site Specific Physical Properties and Confirmatory 3.8.6.2 Foundation Settlement l 3.14 Reactor Internals Vibration Analysis, Confirmatory 3.9.7.1 Measurement and Inspection Programs 3.15 ASME Class 2 or 3 Quality Group Confirmatory 3.9.7.2 Components with 60 Year Design Life 3.16 Equipment qualification report Confirmatory 3.10.5.1 3.17 Dynamic qualification report Confirmatory 3.10.5.2 1 3.18 Emironmental Oualification Document Confirmatory 3.11.6.1 3.19 Emiromental Qualification Records Confirmatory 3.11.6.2 4.1 CRD Inspection Program Procedural 4.53
$.1 Water Chemistry . Design 5.2.6 6.1 External Temperature Conarmatory 6.4.7.1 6.2 Meterology(X/Os) Confirmatory 6.4.7.2 63 Toxic Gases Confirmatory 6.4.73 ,
Effects of Sation Blackout on HVAC 7.1 Confirmatory 7.8.1 7.2 Electrostatic Discharge on Exposed Confirmatory 7.8.2 Equipmect Components 73 Localized High Her.t Spots in Semiconductor Confirmatory 7.83 Material for Computing Devices 8.1 Stability of offsite power system Confirmatry 8.1.4.1 8.2 Diesel Generator Reliability Procedural 8.1.4.2 Amendment 12 1.93 r
ABWR imi=4c . Standani Plant n,v c Table 1.9 2
SUMMARY
OF ABWR STANDARD PLANT INTERFACES WITH REMAINDER OF PIANT (Continued) ITEM INTERFACE - NO. SUBJECT TYPE SUBSECTION 83 Class IE Feeder Circuits Design 8.23.1 8.4 Non-class IE Feeders Design 8.23.2 8.5 Specific ABWR Standard Plant / remainder of plant D sign 8.233 power sysytem interfaces * , 8.o Interupting Capability of Electrical Confirmatory 83.4.1 Distribution Equipment 8.7 Diesel Generator Design Details Confirmatory 83.4.2 8.8 Certified Proof Tests on Cable Samples Confirmatory 83.43 l 8.9 Electrical Penetration Assemblies Confirmatory 83.4.4 8.10 Analysis Testing for Spatial Seperation Confirmatory 83.4.5 per IEEE 304 l 8.11 DC Voltage Analysis Confirmatory 83.4.6 8.12 Seismic Qualification of Eyewash Equipment Confirmatory 83.4.7 ; 8.13 Diesel Generator Load Table Changes Confirmatory 83.4.8 : 8.14 Offsite Power Supply Arrangements Procedural 83.4.9 8.15 Diesel Generator Qualification Tests Confirmatory 83.4.10 8.16 Defective Refurbished Circuit Breakers Confirmatory 83.411 . 8.17 Minimum Starting Voltages for Class Confirmatory 83.4.12 1E Motors 9.1 Ultimate heat sink capability Design 9.2.17.1 9.2 Makeup vr!ct system capability Design 9.2.17.2 93 Contamination of DG combustion air intakes Confirmatory 9.5.10.1 Amendment 12 1.93.1
+
1 ABM z w ioast Etandard Plant nyn TABLE 3.21 CLASSIFICATION
SUMMARY
The classification information is presented Systern wise *** In the following order: Table Tab!< 3.21 MPL 3.21 MPL llam No. Hushtt** 2hlt item No. Numbat** 2hlt B Nuclear Steam Supnh Systems E2 E22 High Pressure Core Flooder System
- B1 B11/J10 Reactor Pressure Vessel J11/J12 System'/ Fuel' E3 E31 Leak Detection and Isolation System' B2 B21 Nuclear Boiler System' E4 E51 RCIC System' B3 B31 Reactor Recirculation System F Reactor Servicina C Control and Instrument Systems F1 Fil Fuel Servicing Equipment C1 C11/C12 CRD System * !
F2 F13 RPV Senicing Equipm:nt C2 C31 Feedwater Control System F3 Fi4 RPV Intercal Senicing C3 C41 Standby Liquid Control Equipment l System F4 FIS Refueling Equipment C4 C51 Neutron Monitoring System' F5 F16 Fuel Storage Equipment C5 C61 Remote Shutdown System G Reactor Auxillary Systems C6 C71 Reactor Protection System' G1 G31 Reactor Water Cleanup System D Radiation Monitorine Systems G2 G41 Fuel Pool Cooling and Cleanup D1 D11 i Process Radiation Monitoring' System System G3 G51 Suppression Pool Cleanup D2 D23 Containment Atmospheric System Monitoring System' H Control Panels E Core Cooline Systems H1 H11 Main Control Room Panel' El E11 RHR System
- H2 H21 local Control Panels
- These systems e subsystems thereof, have a primaryfunction that is safety related. As shown in the balance of this Table, some of these systems contain non sofety related components and, conversely, some systems whose primaryfunctions are non safety r* lated contain components that have been designated safety related.
" Ma:ter Parts List Number desiptatedfor the system OI l "* Only those systems that are in the ABilR Standard Plant scope are included in this table, i Amendment 7 3.26 i
)
ABWR uuima l Standard Plant arv n TABLE 3.21 Tuble 3.21 MPL CLASSIFICATION
SUMMARY
(Continued) Table 3.21 MPL h liem No. Number *
- Injt item No. Number" Illjg j Nuclear Fuel P Station Auxillan Systems See item B1 P1 P11 Makeup Water System I (Purified)
K Radioactive Waste Systems P2 P13 Makeup Water System (Condensate)* K1 K11 Radioactive Drain Transfer System P3 P21 Reactor Building Cooling Water System
- P4 .P22 Turbine Building Cooling !
K2 Kl? Radwaste System Water System ; i PS P24/P25 HVAC Cooling Water Systems' P6 P41 Reactor Senice Water System N Peer Cvele %tems P7 P42 Turbine Service Water System !
'O N11/N21 Power Conversion System P8 P51/PS2 Instrument /Senice Air System a
$3hdN3h P9 P54 High Pressure Nitrogen System N31/N32 N33/N34 N35/N36 N37/N3S R Station Electrical)ntems N39/N41 N42lN43 R1 R42 DC Power Supply
- N44/N51 N61/N71 R2 R10/R11/ Auxiliary AC Power System' l N72 R22 N2 N62 Offgas System R3 R43 Emergency Diesel Generator i System' R4 RS2 Lighting and Senicing Power .
Supply These systems or stubsystems thereof. have a orimarv furnetion that is safetv-related. As shou in the balance of this Table. some of these systems contain non saferv . elated comvonents ond. converselv. some systems whose nrimarv fornetions are non safetv related contain comportertts that have been desic rated saferv~related. Ma<rer Pans Litt Mrmher desienated for the systent Amendment 12 n..;
1 ABWR meme Standard Plant arv n TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Table Table 3.21 MPL 3.21 MPL ltem %. Number ** Tjik Item W. Number ** Ihlt T Containment and Environmental control U Structures and Senicing Systems fuital UI U31 Cranes and Hoists T1 1 11 Primary Containment System
- U2 U41 Heating, Ventilating and Air T2 T10 Containment Internal Conditioning Systems
- Structures U3 U43 Fire Protection System l T3 T13 RPV Pedestal and Shield Wall' U4 U71/U72 Civil Structures
- T4 T22 Standby Gas Treatment System
- U73/U74 U75 T5 T31 Atmospheric Control System T6 T41 Dr>well Cooling System T7 T 19 Flammability Control System T8 T53 Suppression Pool Temperature Monitoring System
- Thesc systems or subsystems thereof, have a primaryfunction that is safety related. As shown in the balance of this Tabic, some of these systems contain non sofety related components and, conversely, some systems whose primaryfunctions are non safety related contain components that have been designated safety related.
" Master Parts List Nwnber desipwtedfor the system O Amendmou 12 3.2 7a
_ __d
MN 21A6100AE Etandard Plant ny a TABLE 3 21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality 1Aca. Classi- Assurance Selsmic Safet4 Requirmnent' ggggggy Principal Commonent" QRAA 112D' BGal193 Hggi B2 Nuclear Boller System (Continued) 4 Pipingincluding supports - 1 C,5C A B 1 main steamline (MSL) and feed-water (FW)line within outermost isolation valve
- 5. Piping incuding supports - 2 SC B B 1 MSL and FW from outermost g isolation valve to and including a seismic interface restraint and E
FW from outerrnost isolation valve ! l to and including shutoff valve ; e e- ! l
- 6. Pipingincluding supports MSL N SC,T B --- -
gg ! from the seismic interface restraint to the turbine stop valve O 7. Deleted l8. Piping FW beyond shutoff N SC,T D - - valve
- 9. Scismic Interface Restraint - 2 SC B B 1 MSL/FW
- 10. Pipe whip restraints MSL/FW 3 SC,C - B -- l R 11. Pipingincluding supports other ;
E within outermost isolation valves
- a. RPV head vent 1 C A B 1 (g)
- b. RPV head spray 3- C A B I (g)
- c. Main steam drains 1 C,SC A B 1 (g)
R 12. Pipingincudingsupports other
$ beyond outermost isolation valves
- a. RPV bead vent N C D -- --
- b. RPV head spray N SC. D -- --
- c. Main steam drains N SC D - -
Amendment 12 3.29
. . _ _ - - . - - _ . . . . ~ . . - . . . . . - . . - - . _ - .
1 i , asA61uGAE EttamAard Plant arv a TABLE 3.21 CIASSIFICATION
SUMMARY
(Continued)
- 4 1.
) QualHy
. Groep Quality , saf 14ea. Classi. Assursace Seisele Pdadpal Casapanen,a QR88 liegt Bendmad h' Catamarvf h
- B2 % Nr Beller System (Contimmed) ;
i 19 Piping imeloding supports . 2/N C,$C B/D B/- 1/~ (g) ! j l instrumentation beyond outermost isolation valves ) l 1 ; j 14. Safety / relief valves 1 C A B I j l I l 15. Valves MSL ud FW 1 C,SC ' A B I ! isolation valves, and other l FW valves within containment l 16. Valves FW,other beyond 2 SC B'. B 1 l outermost isolation vlaves up to and including shutoff valves { 17. Valves . within outermost isolation valves
- a. RPV head vent 1 C A B I (g)
. b. RPV bead spray 1 C,SC A B I (g)
- c. Main steam drains 1 C.SC A B 1 (g) .
l l 18. Valves,other
- a. RPV head vent 3 C C B 1
- b. RPV head spray 2 SC B B I
- c. Main Steam Drain N SC D - -
l 19. Valves instrumentation beyond 2/N SC B/D B/-- 1/- (g) outermost isolation valves
- 20. Mechanicalmodules.instrumen- 3 C,SC - B 1 tation with safety-related function
- 21. Electrical modules with safety. 3 C,SC,X - B I (i) related function
- 22. Cable with safety related 3 C,SC,X .- B 1 1
function O l
' Amendment 3 3.2 10 1
)
ABWR maar i nr.....a pi..r arv . i TABLE 3.21 CIASSIFICATION
SUMMARY
(Continued) Qual % i Groep Quality Imen- Classi. Assurance Selsele Primelmal commenenta Seg- game gggggggd Emphumant' Catspor/Matti Qhat l 33 Reactor Reeirculation System at r
- 1. Piping Primaryside, motor cooling system 3 C C B I I$
- 2. Pipe Supports 3 C C B 1
- 3. Pump motor cover 2 C B B I
- 4. Pump non pressure retain- N C -- -- --
ing parts including motor, instruments, electrical # cables and seals
- 5. Valves 3 C C B I (g) !
C1 CRD System
- 1. Valves with no safety related 2 SC B B I (g)
( function (not part of HCU) L e g 2. Piping including supports- 2 C,SC B B I (j) insert line
- 3. Piping other (pump suction, N SC D --- -- (g) pump discharge, drive header)
- 4. Hydraulic control unit 2 SC -- B I (k)
- 5. Fine motion drive motor N C - -- --
- 6. CRD Drive water ptars N SC D --- --
k 7. Control Rod Drive 1/3 C A/- - B 1
- 8. Electrical modules with 3 C,SC -- b I safety function
- 9. Cable with safety related 3 C,SC,X B I function 0-Amendment 12 32 11
ABWR mumre l Standard Plant arv.n TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality
- Group Quality 14ca. Classi. Assurance Seismic Safet4 lign* I Prhicinal Comnonent" Clan Acation Requirement' Catenorv b' gin C$ Remote Shutdown System 1 Components of this system are included under B2, E1, E4, G3, H2, and P2.
- 1. Electrical modules with 3 C SC,RZ, -- B 1 safety related function X !
- 2. Cable with safety related 3 RZ ... B I i function ,
C6 Reactor Protection System
- 1. Electrical modules with 3 SC,X,T, .- B I '
safety.rclated function RZ
- 2. Cable with safety 3 SC,X,T, ... B 1 I
related functions RZ
- 3. Electrical Modules, other N T,X - --- ..
(u) 1 Cable, ether N T,X - ... . . . . (u) DI Process Radiatloa Monitoring Spiem ; (includes gascous und liquid af11uent monitoring) l
- 1. Electrical modules with 3 SC,X,RZ - B 1 with safety related functions (includes monitors)
- 2. Cable with safty related 3 SC,X,RZ - B I functions
- 3. Electrical Modules, other N T,SC,RZ, - -- ...
(u) X,W
- 4. Cabic, other N T SC,RZ, - -. ...
(u) X,W I Amendment 12 3,2 13
. ampa e' - := i :: :=i :
ABWR nuims ) teandard Plant nev n TABLE 3.21 CIASSIFICATION
SUMMARY
(Continued) hI ' Qaality Group Quality imm. Classi. Assarnace h=le Safeg Principal Campamenta Qpaa an' Emilead WM Ettti
- D2 Costalament Atmospbele Moaltorlag System
- 1. Component with safety related 3 C,SC - B i '
l El RHR System ,
- 1. Heat exchangers-primary side 2 SC B B I .
- 2. Heat exchangers including 3 SC C B 1 supports secondary side ;
- 3. Piping including supports
- 1/2 C,SC A/B B 1 (g) within outermost isolation valves
- 4. Containment spray piping 2 C B B I including supports and
@ spargers, within and A including the outer-most isolation valves
{' 4a. Piping including supports 2/3 SC B/C B (g) l beyond outermost isolation 1 valves
- 5. Main Pumps including supports 2 SC B B 1
- 6. Main Pump motors 3 SC - B 1
- 7. Valves isolation,(LPFL 1 C,SC A B 1 (g) line) including shutdown suction line isolation valves -
- 8. Valves isolation,other 2 C,SC B B 1 (g)
(pool suction valves and pool test return valves)
- 9. Valves beyond isolation 2/3 SC B/C B I (g) valves
* 'Ihe RHR/ECCS lowpressureflooder spargers are part of the reactcvpressure vesselsystem, see item Bl.5 Amendment 11 3.2 14
ABWR a = =st Remadard Plant an n TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safety IAca. Classi- Assurance Seismic Principal Camponenta Clana' llant Scanand Ranahmaant' M h HI Main Control Room Panel
- 1. Panels 3/N X --
B/ - 1/~ (aa) i -
- 2. Electrical Modules with 3 X - B 1 safety related function
- 3. Cable with safety related 3 X -
B 1 function i I
- 4. Other mechanical and N X - - --
electrical modules H2 14 cal Control Panels
- 1. Panels or Racks 3/N C,5C,X -
B/ - I/ -- (aa)
- 2. Electrical modules with 3 C,5C,X -
B I i safety related function
- 3. Cable with safety related 3 C,SC,X, -
B 1 I function
- 4. Other nechanicaland N C,SC,X -- - --- '
electrical modules K1 Radioactive Drain Transfer System
- 1. Drain piping including supports N ALL D -- ~ i g
and valves radioactive (except) RZ,X) g 2. Drain piping including supports N ALL D -- -- g and valves nonradioactive i n
- 3. Piping and valves contain- 2 C,SC B B 1 I ment isolation '
I
- 4. Other mechanical and N ALL O
electrical modules Amendment 7 32 21
i ABWR mams
, Standard Plant nrv s TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) g! Quality Group Quality loca. Classl. Assurance Seismic t Safet4 Princinal Comnonent" Qan Man
- Gsanan Rea"lrrment' Catasen g,,,,
K2 Radwaste System
- 1. Piping including supports N C,SC B B 1 and valves forming part of containment boundary
- 2. Pressure vessels including N W - - - (p)
N C,SC,H, (p) f
- 3. Atmospheric tanks including - - -
supports T,W 4 015 PSIG Tanks and supports N W -- - - (p) , $. Heat exchangers and supports N C,5C,W - - - (p) l
- 6. Piping including supports N C,SC,H - - - (p) and valves T,W -
- 7. Pumps including supports N C,SC,H - - - (p)
T,W , NI Pmser Cornersion System
- 1. hiain steamlir. (h1SL) from N SC,T B - -
(r) [ second isolation valve to turbine stop valve
- 2. Branch line of htSL including N SC,T B B -
(r) supports between the second isolation vahe and the turbine stop valve from branch point at h1SL to and including the first valve in the branch line l
- 3. hiain feedwater line (hiFL)' N SC B B 1 including supports from second isolation valve branch lines and components beyond up to outboard shutoff valves O
Amendmem 12 3221a I
'1
'ABWR 2.was !
Standard Plant REV.B l TABLE 3.21 -l CLASSIFICATION
SUMMARY
(Continued) - l Quality , Group Quality Loca- Classi. Assurance Seismic Safet4 Beguirtment' Catenory Egigg Princinal Comnonent" Qan ll2D DEtting
- 4. Turbine bypass piping N T D - -
including supports
- 5. Turbine stop vake, turbine N T D - -
(1)(n)(o) '
'I bypass valves, and the main steam leads from the turbine control vake to the turbine ,
casing
- 6. Feedwater system components N. T D - -
beyond outboard shutoff valve
- 7. Turbine generator N T - - -
- 8. Condenser N T - - -
- 9. Air ejector equipment N T -- - -
- 10. Turbine gland sealing N T D - -
t rystem components k N2 OITgas S3 stem
- 1. Pressure vessels including - N T - -- -- .(p)(q) ;
supports
- 2. Atmospheric ta.1ks including N T - - - (p)(q) wpports
- 3. 0-15 psig tanks including N T -- -- -- (p)(q) supports
- 4. Heat exchangers including N .T .. - .- (p)(q) s supports d
- s. Pipingincluding supports N T - - -. (p)(q) and valves
- 6. Pumps including supports N T - - -- (p)(q)
( Amendment 12 3.2 21b - j
. . _ . . . _ _ ._ sh =
MM 2146100AE Standard Plant m i i' TABLE 3.21 . CLASSIFICATION
SUMMARY
(Continued) Quality. . i
. Group Quality Safet4 loca- . Classi - Assurance Seismic ,
Princinal Camponent' Qaal llen* Riallan Reanirement' Catenory Eggs I I P1 MakeupWaterSystem(Purified)
- 1. Pipingincluding supports and 2 C B B 1.
valves forming part of the con- l tainment boundary j :
- 2. Demineralizer wates storage N' O D - --
[ tank includir.g supports ;
- 3. Demineralizer water header - .2 .SC B B J piping including supports 3' and valves
- 4. Piping including supports and N O D -- --
valves
- 5. Other components N O. D . . - -- '
P2 Makeup Water System (Condensate) O
- 1. Condensate storage tank N O D - -
.(w)-
including supports .; q
- 2. Condensate header piping 2 SC B B I .
g including supports and valves
- 1
- 3. Piping including supports and N O D. - --
valves
- 4. Other components N O D - - - -
P3 - Reactor Building Coollag Water System l
- 1. Piping and valves forming part 2 SC,C B B 1 (g) of primary containment boundary P.
$ 2. Other safety related piping, 3 ~SC,C C B 1 -
j. including supports pumps and 1 valves O Amendment 12 3.2 22 J
ABWR -- Standard Plant arv n-TABLE 3 21 CLASSIFICATION
SUMMARY
(Continued) Quality , Group Quality ' Loca. Classi. Assurance Safet4 Selsi Princinal Coraggstal" Qala Uga" Rcallon Requiremment' Catag e[ 'Egiga !
- 3. Electrical moduits with 3 SC,C,X - B 1 ,
with safety 4clated function
- 4. Cabic with safety related .3 SC,C,X -- B 1 i function
- 5. Other mechanical and N SC,C,X,M - -- --
electrical modules l N Turbine Hullding Cooling N T D --- - Water System ' P5 HVAC Cooling Water Systems l
- 1. Chillers, pumps, valves, and 3 SC,X C B 1 ; I piping including supports-Safety related HVAC support i
- 2. Chillers, pumps, valves, and N C,SC,RZ, - - --
$ piping including supports -
non safety related HVAC support
- 3. Piping inclurling supports and 2 C,SC B B 1- !
valves forming part of contain-ment boundary
- 4. Electrical modules and cable 3 SC,X - B 1 i with safety related function
- 5. Other mechanical and N C,SC,RZ, - . -
eleetrical enodules T,X 14 Reactor Servlee Water System l t
- 1. Safety related piping 3 U,0,X C- B 1 including supports, piping and valves
- 2. Electrical modules and cables 3 U,0,X -
B 1 with safety related function O Amndmt ni 12 3.2 23 li ,- -
Mb ' 23A6100AE -
~
Standard Plant arv n TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group . Quality Loca. Classi- Assurance Seismic Safet4 - Princinal Comnonent" ' Qa18 112R BCallQn Requirement' Calgggty Entga
- 3. Other non safety related N U,0,X -- - ---
mechanical and electrical modules l P7 Turbine Service Water System
- 1. Non safety related piping N P,0,T - - -
including supports, piping and valves
- 2. Eletrical modules and cables N P,0,T - -- -
with non safety related function l P8 Instrument / Service Air Sptems g 1. Containment isolation includ- 2 C B B I g ing supports valves and piping -
- 2. Other non safety related N SC,RZ, - -. --
mechanical and electrical X,T,H, components W,C l P9 High Pressure Nitrogen Systems
- 1. Containment isolation includ- 2 C B B 1 ing supports valves and piping
- 2. Piping including supports with 3 SC,C C B 1 safety related function
- 3. Electric modules with 3 RZ,X - B I safety related functions
- 4. Cable with safety related 3 SC,RZ, --
B 1 function X
- 5. Other non safety related N SC,RZ, -- -- -
mechanical and electrical X components O. Amendment 12 3,2,23,3
ABWR mame Standard Plant an n TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) [
.1
. Quality .
Group Quality i Saftt- . I. ace . - Classl. Assurance Seismic Prin'cInal Comnonent" h lja* - fication Eggairement' Catenory b'nin R1 DC Power Supply Nuclear Island
-i
- 1. 125 volt batteries, battery 3 SC,C,X, -
B I l a racks, battery chargers, and RZ l" distribution equipment l O I O AmunJmou 12 3.2 23,2
-ABWR 2mme Standard Plant nry n ' ;
TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) i I. Quality ! Group - Quality l Loca- Classi- Assurance Seismic Safet4
- j 4
Princinal Comnonent" fdan llAn Dsanan Renuirement* Category Neigg i T5 Atmospheric Control System 1
- 1. Nitrogen Storage Tanks N O - - -. ,
1
- 2. Vaporizers end controls N, O - -- -
l
- 3. Piping including supports 3 C.SC -
B 1 and valves forming part of containment boundary
- 4. Electrical modules with 3 C,SC -
B I?
. safety related function 8
- 5. Cables with safety related 3 C,SC,X -- B I function l
- 6. Other piping and valves N SC,RZ,0 - - -
- 7. Oxygen monitoring component N, CSC - - .
(normal operation)- l T6 Dr3well Cooling System
- 1. Motors N C - -- -
- 2. Fans. N C - -- -
5 3. Coils, cooling N C - - -
.t
- 4. Other mechanical and N C,X - -- .
cicctrical modules T7 Flammability Control System 2 SC- B B 1 r T8 Suppression PoolTemperature Monitoring S 3stem ,
- 1. Electrical modules with 3 C,X,SC, -- B I safety related function RZ
- 2. Cable with safety related 3 _ C,X,SC, --
B 1 function RZ Amendment 12 3.2 27
)
u
MM 23A6100AE __ Remnaard Plant mn TABLE 3,21 CLASSIFICATION
SUMMARY
(Continued) h. Quality Group - Quality Loca. Classi- Assurance Seismic Principal Componenta~ fdaR1 Magt Acadead Reantremente c. ryf ggggs U1 Cranes and Holsts
- 1. Reactor Building crane N SC - - - - -
- 2. Refueling Bridge crane N SC -- --- - ,((x) x)
- 3. Fuel handlingjib crane .N SC -- - - .(x);
- 4. Upper Drywell Servicing N , C. -
-- I
- 5. Lower Drywell Senicing N C -- --- I-
- 6. Main Steam Tunnel Senicing N M - -
- 7. Special Sc,vice Rooms N SC,RZ, - --- -
T,W,X U2 Heating, Ventilating,and Air Conditionlag Systems *
- 1. Safety related equipment"
- a. Fan. coil cooling units 3 SC,X - - -
B I
- b. Heating units steam or water 3 SC,X --- B I 9
- c. Blowers Air supply or 3 SC,RZ,X - B -I:
exhaust
- d. Ductwork 3 SC,RZ,X --- B I
- c. Filters Equipment areas 3 SC,RZ ---
B I
- f. HEPA Filters, Charcoal 3 SC,RZ,X -- B I Absorbers ControlRooms and Primary Containment
- g. Valves and Dampers- 2 SC,RZ -
B' I primary containment isolation Includes Reactor Building, Control Build lng. and Service Building thermal and radiologi:al
^
environmental conuo? functions ^ . ! *~ v.tr Island scop:. Controls environment in Ma: ar: . ce . -t :uans. diesel generator rooms, battery rooms, ECCS , RCIC, pump rooms within the & ..rIslasi Amendment 9 e 3.2 28 -
ABM 2asioort 4 mandard Plant arv s-TABLE 3.21 i h CLASSIFICATION
SUMMARY
(Continued) I Quality , , Group - Quality , ' Safety IAen. Classi- Assurance Seismic ' Princinal Caennenesta gggb -ggge Acanned ht' Ca'mor / h
-i U2 Heating,Vestilatlas,and Air Condilloslag Systems * (Coatlaued)
~
- h. Other safety related 3 SC,RZ, -- B I {
valves and dampers H,X - j
- 1. Electrical modules with 3 SC,RZ, ~ - B .1.
safety related function H,X l j
- j. Cable with safety-related 3 SC,RZ, -
B 1 I function H,X ,
- 2. Non safety related equipmeni"
- a. HVAC mechanical or N SC,RZ,H, - -- -
electrical componenis X,SC, O with non safety related functions W,T - i l U3 Fire Protection System
- 1. Piping including supports and 2 C B B 1 valves forming part of the primary containment boundary
- 2. Other piping including supports N SC,C,X, D -- -
(t) (u) and valves RZ,H
- 3. Pumps N F D - -
(t) (u)
- 4. Pump motors N F - -- --
(t) (u)
- 5. Electrii modules N C SC,X, - -- -
(t) (u) RZ,H, T,W Includes thermaland radiological environmentalcontrolfunctions within the ABHR Standard Plant scope.
" Controls environment in rooms or areas containing ncn safety related equipment within the ABK'R Standard Plant.
Amendment 7 3.2 29
23A6100AE S2andard Plan' _ arv. n i TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) - Quality Group Quality Sr.fety - Loca. Classi. Assurance Seismic PrincianLCantnonent* Qaaab h' ficationd Renuirement' t'= n=arvf Estas , i U3 Fire Protection System (Continued) I I
- 6. -CO2actuation modules -N RZ,T -- -- --
- (t) (u)- !
i
- 7. Cables N SC,C,X.
-(u) ;
RZ,T,W
- 8. Sprinklers N SC,X - D -- --
(u) U4 CivilStructures I
- 1. Reactor Building (Secondary 3 SC,RZ - - -
B I ; Containment and Clean Zone)
- 2. Control Building 3 X --
B I
' 3. Senice Building ~N H - - - ... --
- 4. Radwaste Building substructure 3 W --
B I
- 5. Turbine Building N T -- -- --
s 9 Amendment 12 3.2-30
ABWR 23 m e Standard Plant nry n 3.7 SEISMIC DESIGN that earthquake which produce vibratory ground motion for which those features of the nuclear-All structures, systems, and equipment of the power plant necessary for continued operation ! facility are defined as either Seismic Category I without undue risk to the health and safety of or non.Scismic Category 1. The requirements for the public are designed to remain functional. i Seismic Category 1 identification are given in During the OBE loading condition, the safety- . Section 3.2 along with a list of systems, compo- related systems are designed to be capable of nents, and equipment which are so identified. ' continued safe operation. - Therefore, for this . loading condition, safety related structures, .! All structures, systems, components, and equip- and equipment are required to operate within ! ment that are safety related, as defined in Sec- design limits, tion 3.2, are designed to withstand earthquakes. 3 as defined herein and other dynamic loads includ- ' The seismic design fer the SSE is intended to ing those due to reactor building vibration (RBV) provide a margia in design that assures caused by suppression pool dynamics. Although capability to shut down and maintain the nuclear l this section addresses seismic aspects of design facility in a safe condition, in this case, it cnd analysis in accordance with Regulatory Guide is only necessary to ensure that the required i 1.70, the methods of this section are also systems and components do not lose their i applicable to other dynamic loading aspects, capability to perform-their safety related l cxcept for thc' range of frequencies considered, f a nction. This is referred te as the ; The cutoff frequency for dynamic analysis is 33 no loss of function criterion and the loading ' Hz for seismic lords and 80 ZHz for suppression condition as the SSE loading condition. pool dynamic loads. The definition of rigid system used in this section is applicable to Not all safety related components have the scismic design only. same functional requirements. For example, the 2 reactor containment must retain capability to The safe shutdown carthquake (SSE) is that restrict leakage to an acceptable level, earthquake which is based upon an evaluation of Therefore, based on present practice, elastic
~
the maximum cartnquake potential considering the behavior of this structure under the SSE leading regional and local geology, seismology, and condition is ensured. On the other hand, there specific characteristics of local subsurface are certain structures, components, and systems material. It is that earthquake which produces that can suffer permanent deformation without
'the maximum vibratory ground motion for which loss of function. Piping and vessels are Scismic Category I systems and components are examples of the latter where the principal designed to remain functional. These systems and requirement-is that they retain contents and components are those necessary to ensure: allow fluid flow.
(1) the integrity of the reac'or coolant pressure Table 3.21 identifies the equip.nent in boundary; various systems as Seismic Category I or non-Seismic Category 1. (2) 'hc apability to shut down the reactor and maintain it in a safe shutdown condition; and 3,7.1 SeismicInput (3) the capability to prevent or mitigate the 3.7.1.1 Design Response Spectra consequences of accidents that could result in potential offsite exposurcs comparable to The design earthquake loading is specified in the guideline exposures of 10CFR100. terms of a set of idealized, smooth curves called the design response spectra in accordance The operating basis carthquake (OBE) is that with Regulatory Guide 1.60. carthquake which, considering the regional and local geology, seismology, and specific charac- Figtere 3.71 shows the standard ABWR design teristics of local subsurface material, could values of the horizontal SSE spectra upplied at 9 during the operating life of the plant. the ground surface in the free fic',d for dampi g reasonably Itis ratios be 7.0 of 2.0, 5.0, expected to affect and 10.0% the :plant site of critical AmenJmem 12 3%t l
_ .- . - - - ~ . . -. - - 23A6100AE mandara Plant RK A values of the vertical SSE spectra applied at the- The magnitude of the SSE design time history ground surface in the free field for dampina is equal to twice the magnitude of the design ratios of 2.0, 5.0, 7.0, and 10.0% of critical OBE time history. The OBE time histories and r damping where the maximum vertical ground response spectra are used for dynamic analysis acceleration is 0.30 g-at 33Hz, same as the and evaluation of the structural Seismic System; maximum horizontal ground acceleration, the OBE results are doubled for evaluating the - structural adequacy for SSE. For development of The design values of the OBE response spectra floor response spectra for Seismic Subsystem are one. half
- of ti,e spectra shown in Figures analysis and eva'uation, see Subsection 3.7.2.5.
3.71 and 3.7 2. These spectra are shown in Figures 3.7 3 through 3.7 20. The response spectra produced from the OBE , design time histories are shown in Figures 3.7G The design spectra are constructed in through 3.7 20 along with the design OBE accordance with Regulatory Guide 1.60. The response spectra. The closeness of the two normalization factors for the maximam values in spectra in all cases indicates that the two horizontal directions are 1.0 and 1.0 as synthetic time histories are acceptable, applied to Figure 3.71. For vertical direction, the normalization factor is 1.0 as applied to The response spectra from the synthetic time Figure 3.7 2. histories for the damping values of 1,2,3 and' 4 percent conform to the requirement for an y 3.7.1.2 Design Time History enveloping procedure provided in item II.1.b of , Section 3.7.1 of NUREG-0800 (Standard Review The design time histories are synthetic FM SRP). However, the response spectra for , acceleration time histories generated to match thr, 'nigher damping values of 7 and 10 percent the design response spectra defined in Subsection show that there are some deviations from the SRP 3.7.1.1.- requirement. This deviation is considered inconsequential, because (1) generating an The design time histories considered in GESSAR artificial time history whose response spectra (Reference 1) are used. They are developed based would envelop design spectra for five different on the method proposed by Vanmarcke and Cornell damping values would result in very conservative (Reference 2) because of its intrinsic capability- time histories for use as design basis input, of imposing statistical independence among the and (2) the response spectra from the synthetic synthesized acceleration time history time histories do envelop the design spectra for components. The earthquake acceleration time the lower damping values. This is very : history components are identified as H1, H2, and important because the loads due to SSE on V. The H1 and H2 are the two horizontal structures should use 7 percent damping for components mutually perpendicular to each other. concrete components, but are obtained by Both H1 and H2 are based on the design horizontal ratioing up the response from the OBE analysis ground spectra shown in Figure 3.71. The V is involving the lower damping. The OBE analysis the vertical component and it is based on the uses only the lower damping values (up to 4%), design vertical ground spectra shown in Figure which are consi: tent with the SRP requ*rements 3.72. (See Subsection 3.7.1.3). The OBE given in Chapter 2 is one third of l the SSE, i.e., 0.10 g, for the ABWR Standard
. Nuclear Island design. However, as discussed in Chapter 2, a more conservative value of -
one half of the SSE, i.e., 0.15 g, was i employed to evaluate the structural'and component response. O Amendment 1 3.72 (: +
ABM s4sioore
- hadard Piant arv s drive during a dynamic event. A simulated test, 33.2.2.2.4 Feel Assembly (locluding imposing dynamic deflection in the fuel channels Chamael) up to values greater than the expected seismic
~
response, is performed with the CRD demonstrated GE BWR fuel channel design bases, analyti. functioning satisfactorily, cal methods, and seismic considerations are similar to those contained'in References 1 and The test was conducted in two phases due to 2. The resulting combined acceleration pro. I facility limitations. The seismic test facility files, including fuel lift for all normal / cannot be pressurized while shaking therefore the upset and faulted events are to be shown~1ess charging pressure of the hydraulic control unit than the respective design basis acceleration is reduced to simulate the back pressure that is profiles. r applied in the reactor. The-appropriate adjustment was determined by first running scram 33.2.2.2.5 Rameter latemal Pump and Motor tests with the full reactor pressure and with Assembly ~ peak transient pressure. Then with the test vessel at atmospheric pressure, the scram tests The reactor internal pump (RIP) and motor [ g were repeated with reduced charging pressures assembly, including its appurtenances and E e until the scram performance matched that of the support, is classified as Seismic Category 1, pressurized tests. This was repeated for the but not active, and is designed to withstand peak pressure also. The seismic tests were then the seismic forces, including other RBV performed with the appropriate pressure loads. The qualification of the assembly is adjustments for the conditions being tested. The done analytically, and with a dynamic test. T tests were run for various vibration levels with fuel channel deflections being the independent 33.2.2.2.6 ECCS Pump nad Motor Assembly * ~ variable. The test facility was driven to vibration levels that produced various channel A prototypc ECCS (RHR and HPCF) pump motor y deflections up to 1.6 inchus and the scram curves assembly is qualified for seismic and other recorded. The 1.6 inch channel deflection is RBV loads via a combination of dynamic several times the channel deflection calculated analysis and dynamic testing. The complete for the actual seismic condition. The motor assembly is qualified via dynamic correlation of the test with analysis is via the testing in accord. ance with IEEE 344. Tbe channel deflection not the housing structural qualification test program includes r analysis since scramability is controlled by demonstration of startup capability as well e channel deflection not housing deflection, operability during dynamic loadin. conditions. This is discussed in more detail e_ 33.2.2.2.2 Core Support (Fuel Support and in Subsection 3.9.3.2.1.4. CR Guide Tube) The pump and motor assemblics, as units A detailed analysis imposing dynamic effects operating under seismic and other RBV load due to seismic and other RBV events is performed conditions, are qualified by dynamic analy. to show that the maximum stresses developed sis and results of the analysis indicate that I during these events are much lower than the the pump and motor are capable of sustaining maximum allowed for the component material. the above loadings without exceeding the allowable stresses. This is discussed in more ' x 33.2.2.2.3 Hydraulic Control Unit (HCU) detail in Subsectioas 3.9.3.2.1.1 a n d a 5 The HCU is analyzed for the seismic and other RBV loads faulted condition and the maximum 33.2.2.2.7 RCIC Pump and Turbine Assembly stress on the HCU irame is calculated to be below the maximum allowable for th: faulted conditiota The RCIC pump construction is a horizontal, w As discussed in Subsection 3.9.1.4.1.2, t he multistage type and is supported on a . faulted condition loads are calculated to be pedestal. The RCIC pump assembly is qualified below the HCU maximum capability. Amendment 12 3.9 11 _ -..,.....,..,.r , . , , , , , . , ,
IBWR , mawn
. Standard Plant Rev. n 1 --
1 analytically by static analysis for seismic and other RBV loadings as well as the design s s operating loads of pressure, temperature, and - external piping ;onds. The results of this analysis confirm that the stresses are less than ' the allowable. This is also discussed in Subse ction 3.9.3.2.2. 1 i. The RCIC turbine is qualified for seismic and other RBV loads via a combination of static ~ analysis and dynamic testing, This is also discussed in Subsection 3.k3.2.1.5. The turbine assembly consists of rigid masses (wherein static , analysis is utilized) interconnected with control
?cvers and elec tronic control systems, necessi-
- ating final qualification via dynamic testing.
Static loading analyses are employed to verify-the structural li%cgrity of the turbine assembly and tha adequacy of bolting under operat_ing, scismicand other RBV loading conditions. The complete turbine assemt,1y is qualified via dy. x namic testing in accordance with IEEE 344. The qualification test program includes demonstration af startup capabi.lity as well as operability during dynamic loading conditions. Operability under normal load conditions is assured by com-parison to the operability of similar 'urbines in operating plants. s s
.\mendment .1 3.911a 5
\
=
y, ABM ' Awsioorn
\ Stand \ggQtant arv. c ; 4.5 RFACTORMATERIALS '
Ouide RoDer Ste1Vte No3
\ - . ,
, . 4.5.1 Control Rod Drive System Structural ; Guide Rouer Pin ' Haynes Aky No.25 !
Matentals -
\
Guide Shaft Stellite No. 6. 4.8.1.1 Material Speelfketions . I Guide Shaft Bushing Stellite No.12 (1) 'MatetW Ust s Seguration Spring - InconelX 750 The following material listing applies to the control rod drive tacchanism supplied for this ' Separation Magnst Alnico No. 5 - application. The position indicator and minor . i non structural items are omitted,- (c) Buffer Manaaie= 1 The properties of the materials selected for Buffer Spring . InconelX450 j the control rod drive mechanism shall be . equivalent to those given 1o. Appendix I to Buffer Sleeve 316L (Hardsurfactd g Section III of the ASME Code or parts A and B of with Colmonoy Nc. 6) M Sectica 11 of the ASME Code or are included in Regulatory Guide 1.85 except that cold worked Guide Roller Non Colbalt Base r senitic stainless Meels t. hall ha'v e's 0,2% 1 Alloys { offset yield strength no greater than 90,000 psi. . Guide Roller Pio ! Non-Cohlt Base L (a) SpoolPiece Aambly Alloys \ Spool Piece Housing, ASME182 GradaF304L Buffer Cone 316L (Hardsurfaced 9' Seal Housing ASME 181 Grade ISO 4L , (d) - Hollow Piston with Stellite No. 6) l Drive Shaft . ASME 479 Grade XM.19 (Hardsurfaced with Piston Tube XM 19 Colmoney No. f.) . Pi' toa Head 316L(Hardsurfaced Ball Bearings 440C with Stellite No. 3) - , s Gland Packing Asbestos Latch laconel X4 % Gland Packing Spring Inconel X 750 latch Spring laconelX 750 1 Ball $ crew S.) aft ASTM A 564 TP6M ! (17 4) < (e) Guide Tube . Condition H 1100 Guide Tube 316L Ball Nut .%TM A-564 TP630 (17 *) (f) Outer Tulw Assembly Condition &1100 Outer Tube XM 19 - s Ba!!s 40C Flange ASME SA1B2 Grade B04L
-1 i
Amendsent 12 dal s ( g gf 8 k / MM l 7 ffO ~ (f W " w .., i @2 ' .
ABWR nasioors Standard Plant arv c (g) Miscellaneow Parts jects selected 300 Seri ~ itainless steel compo-neets te temperatt.as in the sensitiration Ball for Check Valve Haynes Alloy range. The drive shaft, buffer sleeve, piston head and buffer are hard surfaced with Colmonoy O-Ring Seal (Between 321SS Coated with 6 (or its equivalent). Colmonoy (or its equiva-CRD Housing and Teflon lent) bard surfaced components ha~e performed CRD) successfully for the past 15 to 20 years in drive mechanisms. It is normal practice to CRD lastallation ASME SA193 remove some CRDs at each refueling outage. At aolts Grade B7 this time, the Colmonoy (or its equivalent) bard surfaced parts are accessible for visual (2) Special Materials examination. This inspection program (see Subsection 4.5.3 for iriierface requirement) is The coupling spud, latch and latch spring, ade quate to detect any incipient defects separation spring and gland packing spring are before they could become serious enough to cause fabricated from Alloy X-750 in the annealed or operating problems. The degree of conformance equalized condition, and aged 20 hours at to Regulatory Guide 1.44 is prescuted in 13000F to produce a tensile .of 165,000 pi Subsection 4.5.2.4. minimum, yield of 105,000 psi minimum, and elongation of 20% m.nimum. The ball screw shaft (2) Control of Delta Ferrite Content and ball nut are AS TM A 564, TP 630 (17 4) (or its equivalent) in condition H 1100 (aged 4 hours Discussion of this subject and the degree of at 11000F), with a tensile of 140,000 psi conformance to Regulatory Guide 1.31 is minimum, yield of 115,000 psi mininium, and presented in Subsection 4.5.2.4. elongation of 15% minimum. 4.5.1.3 Other Materials These are widely used materials, whose proper-ties are well known. The puts are readily These are presented in Subsection 4.5.1.1(2) accessible for inspection and replaceable if necessa y. 4.5.1.4 Cleaning and Cleanliness Controf All materials for use in this system shall be All the CRD parts listed in Subsection
- l selected for their compatibility with ibe reactor 4.5.1.1 are isbricated under a proc:ss
@ coolant as described in Articles NB 2160 and NB- specification which liaits contatiinants in 3120 of the ASME Code. cutting,6tinding and tapping coolants and lubricants. it uho restricts all other All materials, except SA479 or SA249 Grade proccasing materials (marking inks, tape etc.)
XM 19, have been a ccessfully used for the past to those which are completely retuovable by the 15 to 20 years in similar drive mechanisms, applied cleaning process. All contaminants are Extensive laboratory tests have demon.strated that then required to be removed by the appropriate ASME SA479 or SA249 Gtade XM 19 are suitable c! caning process prior to any of ibe fc. owing-materials and that they are resistant to stress corrosion in a BWR environment. (1) Anv e>rocessing which increases part , m temperature above 2000F. No cold worked austenitic stainless steels ) i with a yield strength greater than 90,000 psi are (2) Assembly which results in decrease of ! l employed in the control rod drive (CRD) system. accessiblity for cleaning. [
)
4.5.1.2 Austenttic Stainless Steel Con ponents (3) Release of parta for shipment. f f (1) Processes, inspections and Tests The specification for packaging and shipping ( the Control Rod Drive prnvides the following: - { i There is a special process employed which sub- - l
} T h .: drive is rinoed in hot deionized water Amendmut 8 4,52 C: '# +( -
.. g '+ -
Y
ABWR z wioars Standard Plant an c 4.6 FUNCTIONAL DESIGN OF (5) Each positioning device shall provide a 9 REACTIVITY CONTROLSYS'IEMS means to prevent or lim;t the rate of-control rod ejection from the core due to a The reactitity control system; consists of break in the drive mechanism pressure boun. control rot s and control rod drives, dary. This is to prevent fuel damage re-mplementar: reactivity control in the form of sulting from rapid insertion of reactivity. a gandolinia.u ania fuel rods (Section 4.3), and the staedby li.iuld control system (described in 4.6.1.1.2 Power Generstion Design Basis Subsection 9.2.5). l The control rod drive system (CRDS) design Evaluations of the reactivity control systems shall meet the following power generation design agalast the applicable General Design Criteria bases: (GDC) are contained in the folowing subsections: (1) The design shall provide for controlling QDS Subsection changes in core reactivity by positioning neutron. absorbing control rods within the "I 23 3.1.23.4 core. 25 3.1.23.6 26 3.1.23.7 (2) The design shall provide for movement and po-27 3.1.23.8 sitioning of control rods in increments to 28 3.1.23.9 enable optimized power control and core 29 3.1.23.10 . power shaping. 4.6.1 Inforrnation for Control Rod Drive 4.6.1.2 Descrfption System The CRDS consists of fine motion control rod 9 4.6.1.1 Design Bases 4.6.1.1.1 Safety Design Bases drive (FMCRD) mechanisms, and the CRD bydraulic system (including pumps, filters, hydraulic control units, interconnecting piping. instru-mentation and electrical controls). The CRDS, The control rod drive CRD mechanical system in conjunction with the rod control and infor-shall meet the following safety design bases: mation system (RC&lS) and reacter protection system (RPS) performs the following functions: (1) The design shall provide for rapid control rod insertion (scram) so that no fuel damage (1) Controls changes in core reactivity by 4' rer,ults from any moderately frequent event positioning ' neutron abmibing contrcl rods ~p (see Chapter 15), within the core in respon.se to control i f signals from the RC&iS.
- l (2) The desigo taall include positioning s uevices, each of which individually supports (2) Provides movement and positioning of control L and positions a control rod. rods in increments to enable optimized power , ,
control and core power shape in response to 1 (3) Each positioning device shtll be capable of control signal.s from the RC&lSL. holding the control rod in position r.nd I preventieg it from inadvematly withdrawing (3) Provides the tchility to;ponition large f.,. outward during any non. accident, accident, groups of tod* simultaneo. sly in response to
- post.acddest and seismic condition, control sipals imm the RC&lS.'
u 3 (4) Each positioning device shall be capable of (4) Protides rupid control rod insertion (scram) !I detecting the separation of tbc control rod in response to manual or automatic . signal.s - ,. l from the drive mechneism to prevent a rod from the RPS w that no fuel damage results ( e
)g drop accident; from any plau miute:st; ,
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21A6100AB Standard P] Ant REV 0 I
; (5) Gathers rod status and rod pot.ition data for rod pattern control, performance monitoring, '
operator display and scram time testing. (6) Prevents ur.s es table rod pattern or rod motions by imposing rod motion blocks in order to protect the fuel. (7) Prevents and mitigates the consequencer of a rod drop accident by detecting rod l ! k separation and controlling rod pattern. ' k
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ABWR Ef amAmed Pfiint msw.s arv c (8) Provides alternate rod insertion (ARI), an directed to each FMCRD connected to the (Q) alterante means of actuating motor driven HCUJaside each FMCRD, high pressure water lifts rod insertion should an ar.ticipated the hollow piston off the ball nut and drives ) transient without scram (ATWS) occur, the control rod into the core. A spring washer ' buffer assembly stops the , hollow piston at the (9) Automatically drives in the drive mechanisms end of its stroke. Departure from the ball nut ; with the electric motors upon scram initle- releases spring loaded latches in the hollow l tion. This provides an additional, diverse piston that engage slots in the guide tube, means of fully inserting a control rod. . These latches support the control rod in the . inserted position. The control rod'cannot be-(10) Provides selected control rod run in (SCRRI) withdrawn until the ball nut is driven up and g for reactor stability control. (See engaged with the hollow piston. Stationary fingers on the ball not then cam the latches out i subsection 7.7.1.2.2.(2)). of the slots and hold them in the retracted-(11) Prvents rod ejection by a passive motor position. A scram action is complete when every (- brake or a scram line inlet check valve. FMCRD has reached their fully inserted position. The design bases and further discussion of The un of the FMCRD n echanisms in the CRD _ both the RC&lS and RPS, and their control inter- system provides several features which enhance . 1 faces with the CRDS, we presented in Chapter 7. both the system reliability and plant:
, operations.' Some of these features are listed 4.6.1.2.1 Fine Motion Control Rod Drive and discussed briefly as follows:
Mechahma (1) Diverse Mean:of Rod Insertion Tbt fine modon control rod drive (FMCPU) w4 for positioning the control rod in-th( reactor The FMCRDs can be inserted either I core is a mechanical /nydraulic actuated mechanism hydraulically or electrically. In response A (Figurr.s l.61, 4.6 2 and 4.6 3). An electric to a scram sigral, the FMCRD is inserted motor driven ball nut and spindle assembly is- hydraulically via the stored energy in the-capable of positioning the drive at a minimum of scram accumulators. A signal is also given 18.3mm increments. Hydraulic prn.sure is used simultaneously to insert the FMCRD for fast scrams. The FMCRD penettr.ws the bottom electrically via its motor drive. This head of the reactor pressure vessel. The FMCRD diversity provides a high degrec of does not interfere with refueling and is assurance of rod insertion on demand. operative even when the head is removed from the reactor vessel. (2) Absence.of FMCRD Piston Seals
- The fine motion capability is achieved with a The FMCRD pistons have no seals and thus, do ball nut and spindle arrangement driven by an not .oquire maintenance, electric motor. The ball nut is keyed to the guide tube (roller key) to prevent its rotation (3) FMCRD Discharge and traverses axisily as the spindle rotate.s. A hollow piston rests on the ball out and upward The water which scrams the control rod motion of the ball out drives this piston and the discharges into the reactor vessel and does control rod into the core. The weight of the mot require a scram discharge volume, thus control rod keeps the bottow pistou and ball nut
- eliminating a potential for common. mode in contact during withdrawal. failure.
A single hydraulic control unit (HCU) pov ers (4) Improved Plant Maneuverability the scram action of two FMCRDs. Upon scram valve ialtiation, high pressure nitrogen from the HCU The fine motion, capability of the FMCRD O raises the piston within the accumulator forcing water through the scram piping. This water is allows rod pattern optimization in response Amndans u u2 h
.j i
.m -
t ABM ' is46ioorn Standard Plant RM C ,j to fuel burnup or load following demands. The basic elements of the FMCRD are as Such a feature complements the ability to- follows: load follow with cos e flow rate i' adjustments. Combining this with Reactor (1) Portions of the FMCRD required for Recirculation System flow control, further hydraulic scram (including hollow piston l l improves plant maneuverability, and buffer). . I 1 (5) Adaptable to Plant Automation (2) Portions of the FMCRD required for : electrical rod inst rtion (including a ; The relatively simple logic of the FMCRD motor, brake release, asociated connector, permits plant automation. This feature is ball screw shaft, ball nut; l utilized for automatic reactor startup and ' shutdowns and for automation load following. (3) Rod position indication (position synchronizing signal gt nerators). . 3 (6) Reduced Time for Reactor Startup (4) Bayonet coupilng between Ile drive and The FMCRDs can be moved in large groups, control rod. Movements of large groups of control rods (called gangs) is utilized to reduce the (5) Reed position switches for scram time for reartor startup. surveillance. ; (7) Reduced Rod Drop Accident Consequences (6) Control rod separation detection devices (Dual Class 1E CRD separation switches). The rod separation detection feature of the ' FMCRD virtually eliminates the possibility (7) Cetinuous full in indication switch, of a Rod Drop Accident by preventing rod withdrawal when control rod separation is (8) Failed buffer detection (uses the full in 9 detected. Additionally, movement of rods in large groups during reactor startup greatly indication switch, reduces the maximum relative rod worth to (9) Brake mechanism to prevent rod ejection in levels lower than current rod pattern the event of a break in the FMCRD primary controls. Rod pattern controls are retained pressure boundary and ball check valve to in order to verify proper automatic rod prevent rod ejection in the event of a movements and to mitigate the consequences failure of tbc scram insert line. , of a rod withdrawal error 3 (10) Integral internal shoot out support (to The drives are readily accessible for prevent drive shoot out) inspection and servicing. The bottom location makes maximum utilization of the water in the The'most significant features and functions reactor as a neutron shield and gives the least of the FMCRD are described below, possible tseutron exposure to the drive components. Using water from the condensate 4.6.1.2.2.1 Components for Fine Motion Control - treatment system, and/or condensate storage tanks as the operating fluid eliminates the need for The fine motion capability is achieved with a special hydraulle fluid. ball nut and spindle arrangement driven by an electric stepper motor. The ball nut is keyed 4.6.1.2.2 FMCRD Components to the guide tube (roller key) to prevent it rotation and it trave:4es axially as the spindle Figure 4.61 provides a simplified schen atic rotates. A hollow piston rests on the ball nut of the FMCRD for illustration of the drive's and upward. motion of the ball nut drives the operating principles. Figure 4.6 2 illustrates control rod in to the core. The weight of the the drive in more detail. control rod keeps the hollow piston and ball nut in contact during withdrawal. Amcndment 12 4M
0 M 23A6100AB l- Reandard Plant m e. The drive motor, located outside the pressure position when the ball nut and hollow piston are
. boundary, is connected to the spindle by a drive re engaged. ;
shaft. The drive shaft penetrates the pressure _ boundary and is scaled by conventional packings. Re engagement of the ball nut with the hollow i A splined coupling connects the drive shaft to piston following scram is automatic. the spindle. The lower half of the splined Simultaneous with the initiation of the coupling is keyed to the drive shaft and the hydraulic scram each FMCRD motor is signaled to - ! upper half keyed to the spindle. The tapered end start in order t: cause movement of the ball nut- ) of the drive shaft fits into a conical seat on upward untilit is in contas t with the hollow-the end of the spindle to keep the two axially piston. This action completes the rod full in ! aligned. The entire weight of the control rod inscrtion and leaves the drives in a condition . ! and drive laternals is carried by a drive shaft ready for restarting the reactor. With the ! thrust bearing located outside the pressure latches in the hollow piston retracted the-boundary. permanent magnets in the stepper motor provide _i' the holding torque to maintain the control rods The axially moving parts are centered and fully inserted in the core. When the motor and guided by radial rollers. The ball nut and brake are deencergized, tbc passive holding ; bottom of the hollow piston include radial torque from the brake keeps the rods fully rollers bearing against the guide tube. Radially inserted. adjustable rollers at both ends of the labyrinth ' seal keep the hollow piston precisely centered in The automatic run in of the ball nut using thb region. the electric motor drive following the hydraulic scram provides a diverse means of rod insertion The top of the rotating spindle is supported as a backup to the accumulator scram, against the iriside of the hollow piston by a . stationary guide. A hndened bushing provides 4.6.1.2.23 FMCRD Pressure Boundary the circumferential bearing between the rotating . _ spindle and stationary guide. Rollers of the The part of the drive inserted into the guide run in axial grooves in the hollow piston reactor drive housing is contained within the: to prevent the guide from rotating with the outer tube. The outer tube is the drive spindle, hydraulic pressure boundary, eliminating the need for designing the drive housing for scram a' 4.6.1.2.2.2 Components for Scram pressure. The outer tube is welded to the middle flange at the bottom and is attached to a The scram action is initiated by the hydraulic seal piece at the top. The seal piece bears control unit. High pressure water lifts the against and seals to the CRD housing. The seal hollow piston off the ball nut and drives the piece and outer tube are attached by slip type control rod into the core. A spring washer connection that accounts for any slight buffer assembly stops the hollow piston at the variation in length between the drive and the end of its stroke. Departure from the ball nut drive housing, releases spring loaded latches in the hollow piston that <.ngage slots in the guide tube. The middle and lower housing enclose the These latches support the control rod in the lower part of the drive and are a part of the inserted position. reactor pressure boundary. The middle housing is attached to the drive housing by four The control rod cannot be withdrawn until the threaded bolts. The lower housing (spool piece) ball nut is driven up and engaged with the holi:w is in turn held to the middle housing and piston. Stationary fingers on the ball nut cam secured to the drive housing by a separate set the latches in the hollow piston out of the slots of eight main mounting bolts. This arrangement in the guide tube and hold them in the retracted permits removing the lower housing, drive shaft and seal assembly without disturbing the rest of the drive. Removing the lower housing transfers the weight of the drive line from the drive _ Amndment 12 46-4 1 I
l 1 MM 23A6100AB l mandard Plant nrv. c screw nut down in to overtravel. After the drive to provide sur;> ort. This tube which is weighing sprir;g has raised the spindle to the welded to the drive middle flange, attaches by a limit of its travel, further rotation of the bayonet lock to the guide tube base. The guide spindle in the withdraw direction will drive the tube, being supported by the housing ertension, ball nut down away from the piston (assuming the prevents downward movement of the drb. in the coupling is engaged). Piston movement, if any, event of housing failure. can then be detected by a reed switch at the zero stroke position. The CRD housing support is designed to prevent ejection of a CRD and attached control 4.6.1.2.2.8 FMCRD Brake rod in the unlikely event of failure of the: The FMCRD design incorporates an (1) Drive housing to vessel attachment weld l clectromechanical brake keyed to the motor (including a failure through the housing or shaft. The brake is normally engaged by passive along the fusion line of the housing to stub l spring force. It is disengaged when the spring tube weld), load is overcome by the energized magnetic force. The braking torque between the motor (2) Flange bolting attaching the drive to the shaft and the CRD spool piece is sufficient to housing, prevent control rod ejection in the event of failure in the pressure retaining parts of the The internal support concept is illustrated ~ drive mechanism. The brake is designed so that schematically in Figure 4.6 7. With failure its failure will not prevent the control rod from assumed at point A or B, the sum of the rapid insertion (scram). mechanicalload plus pressure load acting on the drive and housing would tend to eject the The electromechanical brake is shown in Figure drive. In any of these failures, the drive 4.6 6. It is located between the stepper motor middle flange would be prevented from moving by and the synchronizing signal generators. The- the outer tube (which, in turn, is attached to O. stationary spring loaded disk and coil assembly the upper guide which is locked to the guide is contained within the brake mounting bolted to tube base). This guide tube is supported by the the bottom of the stepper motor. The rotating stub tube welded to the RPV bottom head through disk is keyed to the stepper motor shaft and the CRD housing, in the event of total failure synchro shaft, of the housing to stub tube weld, the housing would be driven downward by the weight plus vessel pressure. After moving a short distance (0.3 inch), the flange of the guide tube contacts the core plate, stopping further movement. As stated above, the CRD is positively locked to the guide tube base and it cannot move further. In this way the housing is prevented from leaving the penetration, thereby restricting the leak path to the area of the annulus tietween the CRD housing and the hole in the reactor vessel bottom head. An orderly 4.6.1.2.2.9 IntegralInternal Slowout Support shutdown would result even if this failure would occur since leakage from this annulus is less An internal CRD housing support has been than the supply from the normal make up adopted to replace the support structure of systems. The components that provide tbe beams, hanger rods, grids and support bars used anti injection function are the guide tube, core in previous product lines. plate, housing and CRD outer tube. The materials of these components are specified to This system utilizes the outer tube of the quality regiurements consistent with that function. O Amendment 12 4.66-
- AB M >>462aoss maderd Plane - nrv. c If a failure were to occur in the spool piece (point C on Figure 4.6 7), the drive would be .
prevented from moving downward by a seat provided - for the spindle adapter as part of the anti. rotation gear. 4.6.1.2.2.10 Materiala of Comatruction The materials of construction for the FMCRD. are discussed in Subsection 4.5.1. 4.6.1.23 Centrol Rod Drive Hydraulic System The CRD bydraulic system supt4ies water beder high pressure to charge the accumulators, to supply CRD purge water, and to supply purge water to the reactor internal pumps-(R(Ps) and the reactor water cleanup (RWCU) pumps. The major components are:
'O 9
Amendment 10 4.641
,. . . >l
. _ . _ _ . . . ._. . _ ._ , _ _ - _ ___ ~ _ _ _ _ _ _ _ _ _ _
3sAs10QAB Riemalard Plant uw A { (3) CRD purge flow control system Duplicate compor.ents are included where l necessary to assure continuous system operation
- (4) Strainers and filters if an inservice component should require l maintenance.
(5) Reactor internal pump and reactor water cleanup pump purge flow 44.1.2.3.2.1 CRD Supply Pump 3 l (6) Equipment protection instrumenta9on One supply pump pressurizes the system with water from the condensate treatment system
- Each HCU provides the scram flow a id purge and/or condensate storage tanks. One spare pump flow requirements for two control rod drives; is provided for standby.. A discharge check therefore, there are only half as many HCUs as valve prevents backflowf. rough the nonoperating
, there are FMCRDs. Water discharged from the HCUs pump. A portion of the pump discharge flowis to the drives during scram is injected into the diverted through a minimum flow bypass line to j- reactor through the labyrinth seals at the upper the condensate storage tank. This flow is ! end of the drive mechanism. The HCUs are not controlled by an orifice and is sufficient to used for normal poaltioning of the control rod prevent pump damage if the pump discharge is drives; this function is provided by the stepper inadvertently closed. l- motor. Condensate water is processed by two filters 44.1.2.3.1 Hydraulle Requirements in the system. The pump suction filter is a disposable element type with a 25 micron The CRD hydraulic system design is shown in absolute rating. A 250 micron strainer in the Figures 4.6 8 and 4.6 9. The hydraulic filter bypass line protects the pump when the requirements, identified by the function they filters ,are being serviced.' The drive water i perform are: filter, downstream of the pump, is a cleanable elemeta efpc with 50 micron absolute rating. A (1) An accumulator hydraulic charging pressure differential pressure indicator and control room-of approximately 2000 psig is required, alarm mcnitor the filter element as it collects Flow to the accumulators is required only foreign materials. < l during scram reset or system startup. l .' 44.1.2JJ.2 Accumulator Charging Pressure i h (2) Purge water to the drives is required at a flow rate of approximately 0.34 gpm per Accumulator charg'ag pressure is established drive unit, by precharging the :itrogen accumulator to a precisely controlled oressure at kawn temper-(3) A total of approximately 2 to 3 gpm purge . ature. During scrarr, the s:rtun valves open and flow is provided to the reactor internal permit the stored ene:3y in the accumulators to pumps and the reactor water cleanup pumps. discharge into the drives. The resulting pres- i i This flow is provided to both systems at _ sure decrease in the charging water header - l approximately CRD pump discharge pressure. allows the CRD supply pump to 'run out" (i.e., Erch system provides its own pressure flow rate to increase substantially) into the breakdown equipmerit to satisfy its control rod drives via the charging water , individual hydraulie requirements, header. The flow element upstream of the accum. ! ! ulator charging header senses high flow and pro. l 44.123.2 Sptem ibeription vides a signal to the manual auto flow control-
- . station which in turn closes the,iystem flow
( The CRD hydraulic system proddes the required. control valve. This action mainitalna increased functions with the pumps, filters, valves, control flow through the charging water header, i instrumentation ind pipin6 nhown in Figure 4.6 8, while avoiding prolonged pump operation ali red described lu the following paragraphs.- *rua cut' conditions. ) siM
' :\
f I
ABM 2mioorn Standard Plant nry c. Pressure in the charging header is monitored pressurized water for hydraulic scram, on' in the control room with a pressure indicator and signal, to two drive units. Additionally, each low pressure alarm. HCU provides the capability to adjust purge flow to the two drives. A test port is provided on During normal operation, the flow control the HCU fer connection of a portable test valve maintains a constant system flow rate, station to allow controlled venting of the scram This flow is used to purge the drives to prevent insert line to test the FMCRD ball check valve reactor water from entering the drive mechanisms. during plant shutdown. Operation of the electrical system that supplies scram signals to 4.6.1.2.3.2.3 Purge Water Hender the HCU is described in Chapter 7.. The purge water header is located downstream The basic components in each HCU are: (1) from the flow control valve. The flow control mesual, pneumatic and electrical valves; (2) an valve adjusts automatically to maintain constant accumulator; (3) related piping; (4) electrical , flow as reactor vessel pressure changes. Because conneetion s; (5) fiit e r s; a n d '(6). flow is constant, the differential pressure instrutaentation (Figure 4.6 8). The components between the reactor vessel and CRD hydraulle and their functions are described in the system is maintained constant independent of following paragraphs. l rector vessel pressure. A flow indicator in the l control room monitors purge water flow. A (1) Scram Pilot Valve Assembly differential pressure indicator in the control 3 room indicates the difference between reactor The scram pilot valve assembly is operated vessel pressure and drive purge water pressure. from the RPS. The scram pilot valve assembly, l with two solenoids, controls.the scram inlet A high differential pressure between the valve. The scram pilot valve assembly is ! reactor vessel and drive purge header could cause solenoid operated and is normally energized. On control rods to drift inward. An alarm in the loss of electrical signal to the solenoids, such control room will alert the operator of high as the loss of external AC power, the inlet port ! differential pressure at a value below that at closes and the exhaust port opens. The pilot which actual rod drift will occur. Should high valve assembly (Figure 4.6 8)'is designed so differential pressure occur, for example by that the trip system signal must be removed from failure of the flow control valve in a full open both solenoids before air pressure can be
~
position, the existence of rod drift would be discharged from the scram valve opentors. This sensed by the control rod drive separation prevents the inadvertent scram of both drives switches and alarmed to the operator in the associated with a given HCU in the event of a- , control room. Simultaneously, the control system failure of one of the pilot valve solenoids. would impose an automatic rod withdrawal block. l The ability of the system to provide hydraulic (2) Scram inlet Valve l scram would not be affected by this condition. l l The scram inlet valve opens to supply Conversely, a failure of the flow control pressurized water to the bottom of the drivc . valve in the full closed positien would result piston. This quick opening globe valve is in the loss of purge flow to the individual operated by an internal spring and system drives. The drives are fully capable of pressure. It is closed by air pressure applied .j' functioning without purge water; however, loss of to the top of its diaphragm. operator.- A purge flow will result in increased drive position indicator switch on (Sis valve contamination by leakage of reactor water in to energizes a light in the control room 'us soon as the drives. The CRD system's scram function the, valve starts to open, would not be affected by the loss o,f purge flow. ! (3) Scram Accumulator- ! .4.6.1.2J.2.4 Hydraulle Control Units l The scram accumulator stores suffadent # Etch hydraulic control unit (HCU) furnishes energy to fully insert two con:rcil vads at any AtwMmtm 12 48 i-4 r I, : i ..Jb. .. .. -.; ~ . . . . _ . 2 - .. !; - , , , , . -
l 23A6100AB l standard Plant nry c ' accumulator charging line prevents loss of water water header is indicated in the control room, pressure in the event supply pressure is lost. An alarm is provided to indicate excessive differential pressure which will lift the During normal plant operation, the accumulator control rods if not limited, piston is seated at the bottom of its cylinder. Loss of nitrogen decreases the nitrogen pressure, The pump purge water is furnished to the RIPS which actuates a pressure switch and sounds an and RWCU pumps at the required flow rates as alarm in the control room, specified on the CRDS process flow diagram (Figure 4.6 9). To ensure that the accumulator is always able to produce a scram, it is continuously monitored To assure the continuous ability to trip, the . for water leakage. A float type level switch charging water header maintains the HCU ~ actuates an alarm if water leaks past the piston accumulators at a high pressure. The scram barrier and collects in the accumulator valves remain closed except during and after instrumentation block, trip operation, so that no flow passes through the charging water header. Pressure in the 4.6.1.2.4 Control Rod Drive System Operation charging water header is monitored. A significant degradation in the charging header The operating modes of the control rod drive pressure will result in a low pressure warning l system (CRDS) are normal operation, rod insert- alarm and rod withdrawal block, and, eventually, tion / withdrawal, and scram. These operational reactor scram if further degradation occurs. modes are described in the following sections. l 4.6.1.2.4.2 Control Rod Insertion / Withdrawal r 46.1.7.4.1 Normal Operation Normal insertion and withdrawal of the During normal operation the CRDS provides the control rods is provided by the electric stepper proper amount of flow required for drive purging motor on the FMCRDs. The motors receive their
\ and for pump purging (RIPS plus RWCU pumps). A insert / withdraw commands via signals from the multi. stage centrifugal pump pressurizes the rod control and information system (RC&lS). The system with water from the treated makeup water objective of the RC&lS is to provide the system (dc. oxygenated) and/or condensate storage operator with the means to make changes in tank. CRD pump minimum bypass flow to the nuclear reactivity so that reactor power level condensate storage tank is utilized to prevent and distribution can be controlled. The system pump overheating if the pump discharge is allows the operator to manipulate control rods, blocked. The total pump flow is the sum of this The design bases and further discussion of the bypass flow, the CRD purge flow through the flow RC&lS are covered in Chapter 7.
control valve, and the recirculation and reactor water cleanup pump purge flows. A full capacity 4.6.1.2.43 Scram standby pump is available. Condensate water is processed by filters (drive water filters) in the Upon loss of electric power to both scram , pump suction and discharge. A redundant set of pilot valve solenoids, the associated HCU drive water filters are provided and are normally applies the drive insert forces to its on standby. Differential pressure monitoring respective drives using a precharged accumulator devices and control room alarms are used to contained within the HCU (The N2 /H2O l monitor the filter elements to detect plugging of accumulator having previously been charged with j the filters, charging water from the CRDS). The drives insert the control rod blades rapidly. The The purge water for each drive is provided water displaced from the drives is discharged through the purge water header. The purge water into the reactor vessel. Also, on receipt of a control valve automatically controls purge water flow to the drive mechanisms. Differential scram start andsignal, the balleach nut onFMCRD will each drive willautomatically bc l pressure between the reactor vessel and the purge driven upward by the screw shaft until it is tO Amendment 12 46-9
M 21A6100AB standard Plant nry c back in contact with the hollow piston. . This materials, B4C powder, austenitic stainless completes the rod full in lasertion and places. steel, and in the case of the alternate control . the CRDS in a condition ready for restarting the rod, Hafnium have been found suitable in meeting reactor. the demauds of the BWR environment. Specifically the B4C containing tubes are RAD After reactor scram, indication that the scram RESIST high purity 304 stainless steel with I has gone to completion (rods in full in position) special chemistry control on carbon and lowered is displayed to the operator, impurity limits to resist irradiation assisted stress corrosion cracking. The sheaths are type Following scram completion (each ball nut 316L stainless steel and the structure is type re.cagaged with its hollow piston), the scram 304,304L or type 316L stainless steel. . signals are reset and each accumulator is rechatged with water from CRDS. 4.6.2.3.1.2 Dimenalonal and Tolerance Analysis 4.6.1.2J Instrumentation Layout studies are done to assure that, given the worst combination of part tolerance ranges The instrumentation for the CRDS is defined on at assembly, no interference exists which will the system P&lD, Figure 4.6 8. Supervisory restrict the passage of control rods. In instrumentation and alarms such as accumulator addition, preoperational verification is made on trouble, low charging header pressure, purge cach control blade system to show that the water / reactor vessel high differential pressure acceptable levels of operational performance are 4 are adequate and permit surveillance of the CRD met. System's readiness. 4.6.23.13 Thermal Analysis of the Tendency The design bases and further discussion are to Warp covered in Chapter 7. The various parts of the control rod assembly 4.6.2 Evaluations of the CRDS remain at approximately the same temperature during reactor operation, negating the problem 4.6.2.1 Failure Mode and Effects Analysis of distortion or warpage. What little differential thermal growth could exist is This subject is covered in Appendix 15B. allowed for in the mechanical design. A minimum axial gap is maintained between absorber rod 4.6.2.2 Protection from Common Mode Failures tubes and the control rod from assembly for the purpose. In addition, to further this end, The position on this subject is covered in dissimilar metals are avoided. Appendix 15B. 4.6.2.3.1.4 Forces for Expulsion 4.6.2.3 Safety Evaluation An analysis has been performed which Safety evaluations of the control rods and evaluates the maximum pressure forces which control rod drives are described below. Further could tend to eject a control rod from the description of the control rods is contained in core. The results of this analysis are given in Section 4.2. Subsection 4.6.2.3.2.2.2. In summary, if the check valve in the drive flange were to fall to 4.6.23.1 Control Rods close, which is unlikely, calculations and tests indicate that the maximum ejection velocity 4.6.2.3.1.1 Materials AdequacyThroughout would exceed allowable rates. However, a Design Lifetime passive brake in the FMCRD prevents the rod ejection motion. The adequacy of the materials throughout the design life was evaluated in the mechanical design of the control rods. The primary
, Amendment 12 4 10
l 21A61W ma=Aard Plant RM A O-44.2.3.1J Functional Fallure of Critical Componesta operation. The control rod cannot rotate since it is always constrained between four fuel assemblies and the hollow piston has I l The consequences of a functional failure of rollers which operate in a track within the CRD system critical components have been FMCRD. Only structural failure would permit evaluated and the results are presented in or result in control rod to FMCRD uncoupling ! Subsection 4.6.2.3.2.2. which in turn could only result in rod drop if the redundant switches failed to sense 44.23.14 Procludlag Excessive Rate of separation. In that case the rate of rod , Reactivity Addition drop could exceed acceptable reactivity addition rates, but the combined failures of Excessive rates of reactivity addition are the structure and the redundant switches is precluded by the design of the control rod and incredible because the components are FMCRD. Prevention of rod ejection due to FMCRD designed and tested to prevent structural pressure boundary failure is described in failure through mechanical damage as Subsection 4.6.2.3.2.2. Rod drop is prevented described in the following Subsection. through: 44.2.3.1,7 Mechanical Damage (1) Two redundant Class 1E switches in the FMCRD sense separation of the hollow piston, which The results of evaluation of damage to the positions the control rod, from the ball control rod are contained in Subsection nut. These switches sense either separation 4.6.2.3.2. The control rod is designed to of the piston from the nut or separation of prevent mechanical damage that would impair the the control rod from the piston, and block essential function of containing and positioning further lowering of the nut thereby the neutron absorber material in the core. The preventing drop of either the control rod or following conditions and loads that could result , .V the contrel rod and hollow piston as an in mechanical damage are specifically considered i assembly (Refer to 4.6.1.2.2.6 for further in the control rod design and/or testing: details). (1) Absorber material containment is within RAD (2) Two redundant spring. loaded latches on the RESIST
- high purity 304 stainless steel hollow piston open to engage in openings in tubes. The chemical composition of these the guide tube within the FMCRD to catch the tubes, particularly the carbon content and hollow piston if separation from the nut the impurity level are controlled to resist were to occur. These latches open to irradiation assisted stress corrosion support the hollow piston (and control rod) cracking.
following every scram until the ball not is run in to provide the normal support for the (2) The controf rod design detalls of hollow piston (and control rod). 'Inc rate construction eliminate creviced conditions at which the hollow piston (and control rod) and stresses are controlled resist crevice would drop is less than the allowable rate corrosion and intergranular stress corrosion as evaluated in Chapter 15, cracking (IGSCC). (3) The control rod to hollow piston coupling is (3) The control rod is designed for the full a bayonet type coupling. Coupling is force that the CRD is capable of exerting on
- verified by pull test for the control rod- a stuck control rod during scram. This upon initial coupling at refueling and again loading is evaluated with critcria cach time an attempt is made to drive beyond applicable to upset events, the " full out' position during reactor operation. The control rod can only be (4) The control rod is designed for the uncoupled from the FMCRD by relative deceleration loads at the end of SCRAM
>O rotation which is not possible during strokes including the loads from a normally 4.6 11 \
ABM m sioorn , Standard Plant arv. c operating buffer under criteria for normal 44JJJ.2.2 Reptere of Hydroolic une to - and upset events and for en abnormally Driw Homelag Finage operating buffer under criteria for upset events. For the case of a scram insert line break, a partial or complete circumferential opening is (5) The control rod is designed for lateral postulated at or near the point where the line - displacements due to the maximum fuel enters the housing flange. This failure, if not channel deflection allowed within fuel mitigated by special design features, could channel design criteria under upset (OBE) result in rod ejection at speeds exceeding maxi. events and faulted (SSE) events, num allowable limits of 4 in/sec (assuming rod pattern control) or 6 inches maximum travel 44.23.2 Control Rod Drives _ distance before full stop. Failure of the scram l insert line would cause loss of pressure to the 44.23.2.1 Evaluation erseram 'I1me underside of the hollow piston. . The force resulting from full reactor pressure acting on The rod scram function of the CRD system the cross, sectional area of the hollow piston, provides the negative reactivity insertion plus the weights of the control rod and hollow i required by safety design basis 4.6.1.1.1(1). piston, is imposed on the ball nut.' The ball The scram time shown in the description is nut in turn translates this resultant force into adequate as shown by the transient analyses of a torque acting on the spindle. When this , Chapter 15. - torque exceeds the motor residual torque and seal friction, reverse rotation of the spindle 44.2J.2.2 Analysis of Malfumetion Relating to will occur permitting rod witbdrawal. Analyses Rod Withdrawal show that the forces generated during this post, m ulated event can result in rod ejection speeds I There are no known single malfunctions that which exceed the maximum allowable limits. cause the unplanned withdrawal of even a single-control rod. However, if n:ultiple malfunctions The FMCRD design provides two diverse means I are postulated, studies show that an unplanned of protection against the results of a rod withdrawal can occur at withdrawal speeds postulated scram insert line failure. The first . that vary with the combination of malfunctions means of protecthn is a ball check valve postulated. located in the middle flange of the drive at the scram port. Reverse flow during a line break 4.6.23.2.2.1 Drive Housing Fallure will cause the ball to move to the closed position. This will prevent loss of pressure to The bottom head of the reactor vessel has a the underside of the hollow piston, which in penetration for each CRD location. A drive turn will prevent the generation of loads on the housing is raised into position inside each drive which could cause rod ejection. penetration and fastened by welding. The drive is raised into the drive housing and bolted to a The second means of protection is the FMCRD flange at the bottom of the housing, brake described in Subsection 4.6.1.2.2.8. I n the event of the failure of the check valve, the In the unlikely event of a failure of the passive brake will prevent the ball spindle drive housing to vessel attachment weld rotation and rod ejection. (including a failure through the housing or along the fusion line of the housing to stub tube weld) or the flange bolting attaching the drive to the housing, ejection of the CRD and attached control 44.23.2.23 Total Failure of All Drive rod is prevented by the integral internal blowout Flange Eolts support. The details of the this internal blowout support structure are contained in The FMCRD design provides an anti. rotation Section 4.6.1.2.2.9. Anwndment 12 4 12
ABM nisioorn Standard Plant RN C , i uclear instrumentation; and O V The after installation, prestartup tests (Chapter 14) include normal and scram motion and (b) oben the rod is fully withdrawn the are primarily intended to verify that piping, first time, observe that the drive will valves, electrical components and instrumentation bot go to the overtravel position, are properly installed. The test specifications Observation of the separation switches include criteria and acceptable ranges for drive provides direct indication that the speed, scram valve response times, and control control rod is following the drive pressures. These are tests intended more to during withdrawal, but does not provide document system condition rather than tests of a direct check on coupling integrity. performance. Additionally , observation of a response from the nuclear instrumentation during As fuel is placed in the reactor, the startup an attempt to withdraw a* control rod test procedure (Chapter 14) is followed. The provides another indirect indication tests in this procedure are intended to that the rod and drive are coupled. The demonstrate that the initial operational overtravel position feature provides a . characteristics meet the limits of the positive check on the coupling _ specifications over the range of primary coolant integrity, for only an uncoupled drive temperatures and pressures from ambient to can reach the overtravel position, operating. The detailed specifications and procedures are similar to those in BWRs presently (4) During operation, accumulator pressure and under construction and in operation. level at the normal operating value are verified. 4.63.1.5 Surveillance Tests Experience with CRD systems of the same type The surveillance requirements (SR) for the CRD indicates that weekly verification of A system are described below. While these accumulator pressure and level is sufficient V requirements have not yet been formalized the intent is to follow the general pattern to assure operability of the accumulator portion of the CRD system, established for surveillance testing in BWRs presently under construction and in operation. (5) At the time of each major refueling outage,
, each operable control rod is subjected to (1) Sufficient control rods shall be withdrawn, scram time tests from the fully withdrawn following a refueling outage when core position, alterations are performed, to demonstrate with adequate shutdown margin that the core Experience indicates that the scram times of I
can be made suberitical at any time in the the control rods do not significantly change subsequent fuel cycle with the maximum worth over the time interval between refueling control rod pair having the same HCU or the outages. A test of the scram times at each single rod attached to the unpaired HCU,if refueling outage is sufficient to identify l of greater worth withdrawn and all other any significant lengthening of the scram i operable rods fully inserted. times, l t (2) Eac'n partially or fully withdrawn control 4.6.3.1.6 Functional Tests rod is exercised one or two steps at least once each week. The detailed requirements for functional testing have not as yet been formally (3) The coupling integrity shall be verified for established, but the intent is to follow the l cach withdrawn control rod as follows: general pattern established for such testing in j BWRs currently under construction and in (a) when the rod is first withdrawn, observe operation. As such, it is anticipated that the l q the control rod separation switch functional testing program of the CRDS will i tg response and discernible response of the consist of a five year maintenance life test and l Amendment 12 4 6-15 l l
~ -- . - . - - .--
MM 23A61ooAn -. Standard Plant aw c a 1.5 times design life test program. In addition to maintenance and design life tests, the program covers crud / contamination testing,
- seismic misalignment, channel bulge, failed
, buffer, rod drop (to test hollow piston latch functionality), and rod ejection (to test FMCRD l brake and scram inlet check valve functionality). 4.6.4 Information for Combined Performance of Reactivity Control Systems 4.6.4.1 Vulnerability to Common Mode Failures The reactivity control system is located such that it is protected from common mode failures due to missiles, failures of moderate and high energy piping, and fire. Sections 3.4, 3.5 and 3.6, and Subsection 9.5.1 discuss protection of essential systems against missiles, pipe breaks and fire. i 4.6.4.2 Accidents Taking Credit for Multiple Reactivity Systems There are no postulated accidents evaluated in Chapter 15 that take credit for two or more reactivity control systems preventing or mitigating each accident. 4.6.5 Evaluation of Combined Performance As indicated in Subsection 4.6.4.2, credit is not taken for multiple reactivity control systems for any postulated accidents in Chapter 15. 1 l l Ql l 1 Amendinent 12 4.6 16
.o w .n.,_. __
O \ l ~ l sCf< G- , ( 1 i i r, , !O . i I S % .
/ \ / / \/
l l l ,L - >- - 1 i l DISENGAGE ENGAGE 87 226 08 1 O . Figure 4,6-5 CONTROL ROD TO CONTROL ROD DRIVE COUPLING l 4.6 21 L i
i ! t ! ABWR maan . l
- Standard Plant aw c O
e ! l . GE PROPRIETARY - provided under separate cover e i Figure 4.6-6 FMCRD ELECTRO-MECHANICAL BRAKE Ame ndmciti 12 4M2
'M 23A6100AB Etandard Plant aw c l
, plant variables in the conservative direction See Chapter 14 for a thorough discussion of L simultaneously. The results of this calculation ' preoperational testing for these systems. ! k for the limiting case are given in Figure 6.3 67
- j. through 6.3 75 and Table 6.3 4, 6.3.4.2 Reliability Tests and Inspections The average reliability of a standby (nonoperating) safety system is a function of i the duration of the interval between periodic functional tests. The factors considered in i
/
determining the periodic test interval of the Since the ABWR results have large margins to the ECCS are: (1) the desired system availability l 10CFR50.46 licensing acceptance criteria, the (average reliability); (2) the number of ABWR licensing PCT can be based on the bounding - redundant functional system success paths; (3) PCT which is well below the 22000F PCT limit. the failure rates of the individual components in the system; and (4) the schedule of periodic 6.33.8 LOCA Analysis Conclualons- tests (simultaneous versus uniformly staggered , versus randomly staggered). Having shown compliance with the applicable acceptance criteria of Section 6.3.3.2, it is All of the active components of the HPCF concluded that the ECCS will perform its function System, ADS, RHR and RCIC Systems are designed in an acceptable manner and meet all of the so that they may be tested during normal plant t 10CFR50.46 acceptance criteria, given operation operation. Full flow test capability is at or below the MAPLHGRs in Table 6.3 7. provided by a test line back to the suction source. The full flow test is used to verify 6.3.4 Tests and Inspections the capacity of each ECCS pump loop while the plant remains undisturbed 'in the power p 63.4.1 ECCS Performance Tests generation mode. In addition, each individual i valve may be tested during normal plant i All systems of the ECCS are tested for their operation, operational ECCS function during the preoperational and/or startup test program. Each All of the active components of the ADS component is tested for power source, range,' System, except the safety / relief valves and direction of rotation, setpoint, limit switch their associated solenoid valves, are designed ' setting, torque switch setting, etc. Each pump so that they may be tested during normal plant is tested for flow capacity for comparison with operation. The SRVs and associated solenoid vendor data. (This test is also used to verify valves are all tested during plant initial power flow measuring capability). The flow tests ascension per Appendix A,- Paragraph D.2.c of involve the same suction and discharge source Regulatory Guide 1.68. SRVs are bench tested to (i.e., suppression pool), establish lift settings. Alllogic elements are tested individually and Testing of the initiating instrumentation and then as a system to verify complete system controls portion of the ECCS is discussed in response to emergency signals including the Subsection 7.3.1. The emergency power system, ability of valves to revert to the ECCS alignment- which supplies electrical power to the ECCS in from other positions. the event that offsite power is unavailable,'.is tested as described in Subsection 8.3.1. The Finally, the entire system is tested for frequency of testing is specified in the Chapter response time and flow capacity taking suction 16 Technical Specifications. Visualinspections from its normal source and delivering flow into of all the ECCS components located outside the the reactor vessel. This last series of tests is drywell can be made at any time during power performed with power supplied from both offsite operation. Components inside the drywell can be i power and onsite emergency power. visually inspected only during periods of access ' O Amendment 12 6.3-13 i I
-AB M 234sioorn Standard Plant REV C to the drywell. When the reactor vessel is open, 6.3A.2.3 RHR Testing the spargers and other internals can be inspected. The RHR pump and valves are tested periodically during reactor operation. With the 63A.2.1 HPCF Testing injection valves closed and the return line open to the suppressioc. pool, full flowing pump The HPCF can be tested at full flow with capability is demonstrated. The injection valve suppression pool water at any time during plant and the check valve are tested in a manner-operation except when a system initiation signal similar to that used for the HPCF valves. The is present. If an initiation signal occurs while system test conditions during reactor operation the HPCF is being tested, the system returns are shown on the RHR system process diagram automatically to the operating mode. The (Figure 6.3 3). motor operated valve in the line to the condensate storage system is interlocked closed 6.3.4.2A RCICTesting when the suction valve from the suppression pool is open. The RCIC loop can be tested during reactor operation. To test the RCIC pump at rated flow, A design flow functional test of the HPCF over the test bypass line valve to the suppression the operating pressure and flow range is pool and the pump suction valve from the performed by pur.iping water from the suppression suppression pool are opened and the pump is pool back through the full flow test return line started using the turbine controls in the to the suppression pool, control room. Correct operation is determined by observing the instruments in the control-The suction valve from the condensate storage room. tank and the discharge valve to the reactor remain closed. These two valves are tested If an initiation signal occurs during the separately to ensure their operability, test, the RCIC system returns to the operating mode. The valves in the test bypass lines are The HPCF test conditions are tabulated on the closed automatically and the RCIC pump discharge HPCF process flow diagram (Figure 6.31). valve is opened to assure flow is correctly routed to the vessel. 63A.2.2 ADS Testing 6.3.5 Instrumentation Requirements An ADS logic system functional test and simulated automatic operation of all ADS logic Design details including redundancy and logic channels are to be performed at least once per of the ECCS instrumentation are discumd in plant operating interval between reactor Section 7.3. refuelings. Instrumentation channels are demonstrated operable by the performance of a All instrumentation required for automatic channel functicnal test and a trip unit and manualinitiation of the HPCF, RCIC, RHR and calibration at least once per month and a ADS is discussed in Subsection 7.3.1 and is transmitter calibration at least once per designed to meet the requirements of IEEE 279 operating interval. and other applicable regulatory requirements. The HPCF, RCIC, RHR and ADS can be manually All SRVs, which include those used for ADS are initiated from the control room. bench tested to establish lift settings in compliance with ASME Code Section XI. The RCIC, HPCF, and RHR are automatically initiated on low reactor water level or high drywell pressure. The ADS is automatically actuated by sensed variables for reactor v'essel e Amendment 7 6.3-14
23A6100All Standard Plant a,v s Table 9.14 HEAVY LOAD OPERATIONS - (Continued) la plant Hardware Handling Handling Location Handling Tasks S.vstems* Equinnent Elevation
- RPY OPENING / CLOSING OPERATIONS:(Continued)
D/SP Cover plates RBS RB Crane RF 26700 Auxiliary Hoist Removal, storage and reinstallation. Lifmg slings RPV Service Platform RBS RB Crane RF 26700 Removal, storage Auxiliary Holst IRV 14500 and reinstallation. Lifing slings Steam Plugs RBS RB Crane RF 26700 ' Temporary tool Auxiliary Hoist IRV 15500 ;
~'
Installation and removal 1/2 ton Chain Holst Service Platform Refueling Platform l . Steam Separator / Shroud l Head RBS RB Crane RW 18700 ( Removal, storage and rein- Main Hoist IRV 9500 stallation. Include Dryer / Separator D/SP 18700 unbolting shroud head bolts Refueling Platform from Refueling Platform i Fuel Bundle Sampler Tool RBS Refueling Platform RW 18700 Positioning, sampling and or RB Crane IRV 9100
- removal, storage Auxiliary Hoists i
REFUELING OPERATIONS: New Fuel: 1 Receive at G/F, & lift to RBS RB Crane RB 7300 RF, Receivinginspection Auxiliary Hoist RF26700 l remove outer container
- See Table 9.19 for Legend.
O Amendment 12 9.1 17
ABWR 234siooxii. Standard Plant Rev B ! l Table 9.1-8 g HEAVY LOAD OPERATIONS
~ (Continued) -
la plant Hardware Handling Handling Location - Handling Tasks Systems
- Equipment Elevation
- i REFUELING OPERATIONS:(Continued). ;
Remove inner container and RBS - RB Crane RF 26700 store fuel bundle in new fuel Auxiliary Holst NFS 18700 vault rack. Move fuel to new NFl 18700 i fuelinspection stand, inspect I and return to storage. Move new fuel from vault RBS RB Crane NFS 18700 to fuel pool, storage of fuel - Auxiliary Holst FSP 14800 ! channel fixtures. Channel . Refueling Platform FCF 14800: new fuel and store. Move Auxiliary Holst RF 26700 channeled fuel and load Fuel Grapple RVC 9500 into reactor core. Jib Crane Soent Fuel: Remove spent fuel from . RBS Refueling Platform R W 18700 RPV core. Transport spent Auxiliary Hoists FSP 14880 ' fuel to storage racks and/or Fuel Grapple FCF 14800 fuel channel fixture remove Jib Crane RVC 9500 - ! channels and store spent Channel Handling fuel bundles Boom' Fuel Cask: Receive, lift to refut.'. RBS RB Crane G/F 7300 ; ing floor Lower into cask Main Holst RF26700 washdown pit, washdown & Auxiliary Holst FWP 18700 move to load pit. Move spent Refueling Platform FLP 14800 - fuel to cask load pit. Move Auxiliary Hoists loaded cask to eask washdown Fuel Grapple pit. Move cask to G/F for shipment.
- See Table 9.19 for Legend.
Amendment 6 9.).18
ABWR 2346ioo4>i l Standard Plant neo . i SECTION 9.3 fl v CONTENTS (Continued) Section Tult Eagt 93.7,4 Inspection and Testing Requirements 93 8 93.7.5 Instrumentation Application 938a 93.8 . Radioacthe Draln Transfer System 93 11 , 93.8.1 Design Bases 93 11 93.8.1.1 Safety Design Bases 9 3-11 93.8.1.2 Power Generation Design Bases 9 3-11 93.8.2 System Description 93 11 93.8.2.1 General Description 9 3-11 93.8.2.2 System Operation 93-11a 93.8.23 Compone nt Description 9311a 7 93.8.2.4 Safet, Evaluation 93-11b l 93.8.2.5 T.sts and Inspection 9311b 93,9 iluirocen Water Chemistrv System 93 12 93.9.1 Design Bases 93 12 93.9.1.2 Safety Design Basis 93 12 .; 93.9.13 Power Generation Design Basis 93 12 93.9.2 System Deseirption 9 3-12 93.93 Safety Evaluation 93 12 t 93.9.4 . Inspection and Testing Requirements 93-12 93.9.5 Instrumentation and Controls 93 13 93.10 Omen inlection System 93 13 9.2.10.1 Design Bases 9 3-13 93 lila Amendment 12
ABWR . 2mimn . Standard Plant nn SECTION 9.3 CONTENTS (Continued) Sectlon T[1lg Eggg ! 93.10.2 System Description 93 13 93.103 Safety Evaluation 9 3-13.1 93.10.4 Tests and Inspcetions 93 13.1 ,. 93.10.5 Instrumentation Application 9 3-13.1 93.11 Zine Infection System 93 13.1 93.11.1 Design Bases 93 13.1
! 9 3.11.2 Safety Evaluation 9 3-13.1 93.113 Test and Inspections 93 13.1 9 3.11.4 Instrumentation 93 13.1 O
l l l l 93 iiib Amendment 12 l l'
.. -- - . -. .- _ . . ~ .
1 ABWR 233a.4n Standard Plant n v. n SECTION 9.3 O TintES j Tables 2111e East 93 1 Standby Liquid Control System Operating Pressure / Temperature Conditions 9 3-14 93 2 Water Qualitylastrumentation 93-15a , 93-3 Service Air Consumption During Normal Plant Operation 9315e ILLUSTRATIONS Figure 31tig Egge 93 1 Standby Liquid Control System P&lD 9 3-16 93 2 Sodium Pentaborate Volume Concentration Requirements 9 3-17 l 93 3 Saturation Temperature of Sov.ium Pentaborate Solution 9 3-18 93 4 Sample Probe 9 3-19 93 5 Sample Probe 9 3-20 93 6 Instrument Air System P&lD 9 3-21 93 7 Service Air System P&lD 93 22 i i l 93-iv Amendment 6 I 1
. _L . -< _ . . . . ,_.
ABWR 2miooAn Standard Plant nry n p 9.3.3 Equipment and Floor approaching actual use requirements. ! Drainage System Demineralized water, rather than the actual V neutron absorber solution, can be injected i The system which collects and transfers all into the reactor to test the operation of l radioactive liquid wastes is discussed in all components of the redundant control Subsection 9.3.8. The drainage systems for system. non radioactive liquid wastes are not discussed ! because they are not a part of the ABWR Standard (5) The neutron absorber shall be dispersed Plant. within the reactor core in sufficient quantity to provide a reasonable margin for 9.3.4 Chemical and Volume Control leakage or imperfect mixing.
- System (PWR) ,
l (6) The system shall be reliable to a degree (Not applicable to a BWR) consistent with its role as a special safety system; the possibility of unintentional or 9.3.5 Standby Liquid Control System accidental shutdown of the reactor by this system shall be minimized. 93.5.1 Design Bases 9.3.5.2 System Description 93.5.1.1 Safety Design Bases The SLCS (Figure 9.31) is manually initiated The standby liquid control system (SLCS) has a through a single keyboard switch in the main safety related function and is designed as a control room to pump a boron neutron absorber Seismic Category I system, it shall meet the solution into the reactor if the operator following safety design bases: determines the reactor cannot be shut down or kept shut down with the control rods. Once the C (1) Backup capability for reactivity control operator decision for initiation of the SLCS is ( shall be provided, independent of normal re- made, the design intent is to simplify the activity control provisions in the nuclear manual process oy providing dual keylocked I reactor, to be able to shut down the reactor switches. This prevents inadvertent injection if normal control ever becomes inoperative. of neutron absurber by the SLCS. However, the insertion a the control rods is expected to (2) The backup system shall have the capacity for assure prompt shutdown of the reactor should it controlling the reactivity difference between be required, the steady-state rated operating condition of the reactor with voids and the cold shutdown The keylocked control room switch is provided condition, including shutdown margin, to to assure positive action from the main control assure complete shutdown from the most room should the need arise. Procedural controls reactive conditions at any time in core life. are applied to the operation of the keylocked control room switch. (3) The time required for actuation and effectiveness of the backup control shall be The SLCS is required only to shut down the consistent with the nuclear reactivity rate reactor and keep the reactor from going critical of change predicted between rated operating again as it cools, and cold shutdown conditions. A fast scram of the reactor or operational control of fast The SLCS is needed only in the improbable reactivity transients is not specified to be event that not enough control rods can be accomplished by this system. inserted in the reactor core to accomplish shutdown and cooldown in the normal manner. (4) Means shall be provided by which the functional performance capability of the The boron solution tank, the test water tank, p backup control system components can be the two positive displacement pumps, the two Q verified periodically under conditions motor operated injection valves, the two motor. Amendment 6 93 2
l MM 23A6t00All - ! l standard Plant REV B { operated pump suction valves, and associated Each positive displacement pump is sized to j local valves, panel, and controls are located in inject the solution into the reactor in 60 to _ the secondary containment outside the drywell and 150 minutes, independent of the amount of : wetwell. The liquid is piped into the reactor solution in the tank. The pump and system ! vessel throughout the high pressure core flooder design pressure between the injection valves and (HPCF) 1ine downstream of the HPCF inboard check the pump and system design pressure between ' valve. relief valves are approximately 1560 psig. To i prevent bypass flow from one pump in case of The boron absorbs thermal neutrons and thereby relief valve failure in the line from the other terminates the nuclear fission chain reaction in pump, a check valve is installed downstream of the uranium fuel. each relief valve line in the pump discharge pipe. The specified neutron absorber solution is
- sodium pentaborate (Na2 Bi oOl6,10H 2O). The SLCS is actuated by either of two l It is prepared by dissolving stoichiometric keylocared, spring return switches on the control l quantitles of borax and boric acid in deminera- room console. This assures that switching from lized water. An air sparger is provided in the the OFF position is a deliberate act. Changing tank for mixing. To prevent system plugging, the either switch status to RUN starts an injection 1 tank outlet is raised above the bottom of the pump, opens one motor operated injection valve, '
tank, opens one pump suction motor operated vahe, and closes both of the reactor cleanup system At all times when it is possible to make the outboard isolation valves to prevent loss of reactor critical, the SLCS shall be able to boron, deliver enough sodium pentaborate solution into the reactor (Figure 9.3 2) to assure reactor A light in the control room indicates that shutdown. This is accomplished by placing sodium power is available to the pump motor contactor pentaborate in the standby liquid control tank and that the contactor is deenergized (pump not - and filling it with demineralized water to at running). Another light indicates that the - least the low level alarm point. The solution contactor is energized (pump running). can be diluted with water to within 14 inches of the overflow level volume to allow for Storage tank liquid level, tank outlet valve evaporation losses or to lower the saturation position, pump discharge pressure, and injection temperature. valve position indicate that the system is functioning. If any of these items indicates ! The minimum temperature of the fluid in the that the liquid may not be flowing, the operator tank and piping shall be consistent with that rshall immediately change the other switch to the obtained from Figure 9.3 3 for the solution RUN mode, thereby activating the redundant train temperature. The saturation temperature of the of the SLCS. The local switch will not have a l recommended solution is 59 F at the low level STOP position. This prevents the isolation of alarm volume and a lower temperature at 14 inches the pump from the control room. Pump discharge below the tank overflow volume (Figures 9.3 2 and ' pressure and valve status are indicated in the 9.3 3). The equipment containing the solution is control room, installed in a room in which the air temperature is to be maintained within the range of 50 to Equipment drains and tank overflow are not l 1000F. An electrical resistance heater piped to the radwaste system but to separate system provides a backup heat source which containers (such as 55 gallon drums) that can be maintains the solution temperature at 75 F removed and disposed of independently to prevent (automatic operation) to 850F (automatic any trace of boron from inadvertently reaching shutoff) to prevent precipitation of the sodium the reactor. pentaborate from the solution during storage. High or low temperature, or high or low liquid Instrumentation consisting of solution tem-level, causes an alarm in the control room. perature indication and control, solution level Amendment 12 93 3
i ABWR m6,mu , , - Standard Plant arv. n ( i O ! i GE PROPRIETARY provided under separate cover (Ineludes pages 9.319 through 9.3 23) Eagg Amendment 9.3 19 6 V ' 9.3 20 6 9.3 21 6 9.3 22 6 9.3 23 6 s o Amendmcni 12 9..t19 23
ANR 23^sioo^2 l Standard Plant am ! Moisture Separatnr Reheaters Four horizon- system pressure. The latermediate stop vahe and ;
. tal cylindrical.shell, combined moisture separator / intercept valve are designed to close in approximately reheaters (MSRs) are installed in the steam path 0.2 second.
between the high and low pressure turbines. The MSRs serve to dry and reheat the HP turbine steam Low Preature Turbines Each LP turbine re-eahaust (crossaround steam), before it enters the low ceives steam from two CIVs. The steam is expanded pressure turbines. This improves cycle efficiency and axially across several stages of stationary and moving reduces moisture related erosion and corrosion in buckets. Turbine stages are numbered consecutively, the low pressure turbines. Crossaround steam is starting with the first HP turbine stage, piped into the bottom of the MSR Moisture is ec-moved in chenon type moisture separators, and is Extraction steam from the LP turbines supply drained to the moisture separator drain tank and the first four stages of feedwater heating. A fifth from there to the heater drain tank. The dry l extraction stage is provided to remove moisture and j crossaround steam next passes upward across the protect the last stage buckets from erosion induced by ' reheater which is supplied with main steam. Finally, water droplets. This extraction is drained directly to the crossaround steam is routed to the combined in- the condenser, termediate valves (CIVs) which are located just up-stream of the low pressure turbines inlet nozzles. Egyption Nonreturn Valves - Upon loss of load, the steam contained downstream of the turbine The reheaters drain, via drain tanks, to the extractions could flow back into the turbine, across pumped heater drain system that discharges to the the remaining turbine stages, and into the condenser. scactor feedwater pump s' action. Safety valves are Associated condensate could flash to steam under this provided on the MSR for overpressure protection. condition and contribute to the backflow of steam or could be entrained with the steam flow and damage Combined Intermedbte Valves Two com- the turbines. Extraction nonreturn valves are bined intermediate valves (CIVs) are provided for installed in the extraction lines to the first, third and, cach LP turbine, one in each steam supply line, if required, fourth stage of turbine extractions to O.. called the hot reheat line. The CIV consists of two guard against this backflow and the resulting potential valves sharing a common casing. The two vahes are damage due to water entrainment or overspeed the intercept valve and the intermediate stop valve, condition. Although they utilize a common casing, these valves have entirely separate operating mechanisms and Generator The generator is a direct driven, controls. The function of the CIVs is to protect the three phase,60Hz,1800 rpm, four poles synchronous turbine against overspeed from steam and water generator with water cooled stator and hydrogen energy stored between the main stop and control cooled rotor, valves and the CIVs. One ClV is located on each side of each LP turbine. The rotor is manufactured from a one piece forging and includes layers of field windings Steam from the MSRs caters the single inlet of embedded in milled slots. The windings are held each valve casing, passes through the permanent radially by steel slot wedges at the rotor outside basket strainer, past the intercept valve and stop diameter. The wedge material maintains its mechani-valve discs, and enters the LP turbine through a cal properties at elevated temperature. The magnetic single inlet. The CIVs are located as close to the LP field is generated by de power which is fed to the turbine as possible to limit the amount of uncon- windings through collector rings located outboard of trolled steam available for overspeeding the turbine, the main generator bearings, The rotor body and Upon loss of load, the intercept valve first closes shaft is machined from a single, solid steel forging. . then throttles sv.am to the LP turbine, as required to Detailed examinationsinclude: control speed and maintain synchronization. It is capable of opening against full system pressure. The (1) Material property checks on test specimens intermediate stop valves close only if the intercept taken from the forging valves fail to operate properly. These valves are
~
capabic of opening against a pressure differential of (2) Photomicrographs t. :xamination of micro-approximately 15 percent of the maximum expected structure l Amendment 8 10.2 3
ABM 23xsioox, ! Standard Plant am x l l (3) Magnetic particle and ultrasonic examination - individual channel in order to increa a reliability. (4) Surface finish tests of slots for indication of a The overspeed sensing devices are located in the stress riser front bearing standard, and therefore are protected from the effects of missiles or pipe break. The hy-hlk Hvdronen Svuem . The bulk hydrogen draulic lines are fail safe, that is, if one were to be and CO system is illustrated on Figure 10.2 4. The broken,less of hydraulic pressure would result in a-i hydrogen, system is designed to provide the necessary turbine trip. The electric trip signals are redundant. l flow and pressure at the main generator for purging One circuit could be disabled by damage to the wir-carbon dioxide during startup and supply makeup by- ing, but the other system is fail safe, i.e., loss of signal drogen for generator leakage during normal opera- results in a turbint. trip. These features provide tion. inherent protection against failure of the overspeed system caused by missiles or pipe whipping. l The system wasists of hydrogen supply piping with all the necessary valves, instrumentation, gas The electrical backup overspeed trip consists of purity measuring equipment, hydrogen gas dryers, independent circuits. Each circuit monitors a and bulk hydrogen storage unit. separate speed signal voltage and activates voltage comparators at various speed levels. The output of Fires and explosions during filling and/or these circuits is used in tripping and monitoring of the purging of the generator are prevented byinerting turbine, the generator with CO, so that a flammable mixture of hydrogen and oxygen cannot be produced. Un- Two air relay dump valves are provided which needed hydrogen is vented outside through a flame actuate on turbine trip. The valves control air to the arrestor. extraction nonreturn valves which limit contributions to turbine overspeed from steam and water in the ex. 10.2.2.3 Normal Operation traction lines and feedwater heaters. The closing time of the extraction nonreturn valves is less than 2 I During normal operation, the main stop valves seconds. $ and CIVs are wide open. Operation of the T G is ! under the control of the electro- hydraulic control Upon loss of generator load, the electro-hydrau-(EHC) system. The EHC system is comprised of lic control (EHC) system acts to prevent rotor speed l three basic subsystems: the speed control unit, the from exceeding design overspeed. Refer to Table load control unit, and the flow control unit, The 10.21 for the description of the sequence of events normal function of the EHC system is to generate following loss of turbine load. Failure of any single the position signals for the four main stop valves, component will not result in rotor speed exceeding four main control valves, and six CIVs. design overspeed (i.e,120 percent of rated speed). The following component redundancies are employed 10.2.2.4 Turbine Overspeed Protection System to guard against overspeed: In addition to the normal speed control (1) Main stop valves / Control valves function provided by the turbine control system, a separate turbine overspeed protection system is (2) Intermediate stop valves / Intercept valves included. The turbine overspeed system is a highly (CIVs) reliable and redundant system which is classified as non safety related. (3) Primary speed control / Backup speed control Protection against turbine overspeed is provided -(4) Fast acting solenoid valves / Emergency trip fluid by the mechanical overspeed trip and electrical system (ETS) backup overspeed trip. Redundancy is achieved by using two independent channels from the signal (5) Speed control /Overspeed trip / Backup over-source to the output device. The sensing device,line speed trip and output desice are of a different nature for each The main stop valves and control valves provide Anwndment 12 10.2-4
23A6100AB Rem =Aard Plant arv.A l Table 15.7 5
- O minwisre svsrsa riituas icciosur riai srsas ;
1 ! I Data and assumptions used to estimate source terms. A. Power level 4005 M Wt B. Fraction ofiodine released 10 % C. Duration of accident 2 hrs 11 Dispersion and Dose Data - - A. Meteorology Table 15.7-7 B. Boundary and LPZ distances Table 15.7 7 C. Method of Dose Calculation ' Reference 1 D. Dose conversion Assumptions Reference 1 and RG 1.109 E. ActivityInventory/ releases Table 15.7 6 F. Dose Evaluations Table 15.7 7
- O I t
!O 13.7 13 1
ABM. 23463oo4o Standard Plant arv e , Table 15.7 6 - ISOTOPIC RELEASE TO ENVIRONMENT IN CURIES O ISOTOPE 1 MIN 10 MIN 1 HOUR -- 2 HOUR I 131 1.3E + 0 1.2E + 1 4.2E + 1 5.1E + 1 1132 1.1E 11 9.4E1 3.0E 0 3.5E + 0 1133 7.7E1 6.9E + 0 2.4E + 1 2.9E + 1 1134 7.0E 2 5.9E 1 1.7E + 0 1.9E + 0 1135 33E1 3.0E+0 1.0E + 1 1.2E + 1 . TOTALI 2.6E+0 23E+1 8.0E + 1 9.8E + 1 i 1 O i i i
]
i O Amendment 12 g3,7,34
l 23A6100AB Remndard Plant nry c Table 15.7 7 RADWASTE SYSTEM FAILURE ACCIDENT METEOROLOGY AND DOSE RESULTS , THYROID WHOLE DISTANCE x/Q DOSE BODY (meters) (sec/m *) (mm) (rem) 300 1.18E 3a 3.5E + 1 2.0E 2 500 4.83E 4 1.4E + 1 83E 3 . 4 800 2.19E-4 6.4E + 0 3.8E 3 1000 1.77E 4 5.2E + 0 3.1E 3 1200 1.48E-4 4.4E + 0 2.6E 3 1500 1.19E 4 3.7E + 0 2.1E 3 2000 9.01E 5 2.7E + 0 1.6E 3 2500 7.22E 5 2.1E + 0 13E 3 ' 3000 6.02E 5 1.8E + 0 1.0E 3 3500 5.16E 5 1.5E + 0 8.9E-4 i a 1.18E 3 = 1.18 X 10
- O O
Amendment 12 13.t.g3 l
ABM 2 moor. Riandmed Plane nev.n Table 15.7 8 FUEL HANDLING ACCIDENT PARAMETERS h 1 Data and assumptions used to estimate source terms. I A. Power level 4005 M Wt B. Radial Peaking Factor 1.5 C. Duration of accident 2 hours D. No. Rods Damaged 124 rods E. Minimum time to accident 24 hrs' F. Peak linear power density 13.4 kW/ft G. Average buro up 32,000 MWD /t H. Maximum Fuel centerline temperature 3315 F l~
- 1. Fraction of activity released 10% of allisotopes except 30% Kr 85 -
11 Data and assumptions used to estimate activity released A. Species fraction . (1) organic iodine 0.25 % i (2) inorganic iodine 99.75 % l (3) noble gas 100 % l B. Pool Retention decontamination factor i (1) organiciodine 1 (2) inorganic iodine 133 i (3) noble gas ' 1 i C. SGTS filtration efficiency (1) organiciodine ' 99% < (2) inorganic iodine _ 99 % (3) noble gas 0% 4 D. Reactor Building Release Rate 300%/2hr
!!! Dispersion and Dose Data A. Meteorology Table 15.711 B. Boundary and LPZ distances Table 15.711 C. Method of Dose Calculation Reference 1 '
D. Dose conversion Assumptions Reference 1 and RG 1.109 E Activity Inventory / releases Table 15.7 9 4 Table 15.710 F, Dose Evaluations Table 15.711 ' O Amendment 2 15.716
- O O O l EVENT 27 E
m . , GENERATOR LOAD REJECTION WITH ONE [ o BYPASS VALVE FAILURE gl, STATE D g j e POWER <40% POWER 240% BYPASS 1 SYSTEM ! OPERATES (N-1) S F POWER NO 2 BYPASS CAPACITY ACCEPTABL.: STEADY STATE YES REACTOR REACTOR SCRAM SIGNAL: PRESSURE - STEAM PROTECTION = HIGH M OR PROTECTION = TURBINE HIGH PRESSURE RELIEF BYPASS SYSTEM SYSTEM CONTROL VALVE SYSTEM SYSTEM FAST CLOSURE S F SlF S F SlF I I CONTROL ROD TRIP OF CONTROL ROD i ' DRIVE SYSTEM FIVE RIPS DRIVE SYSTEM - CONTROL PRESSURE RODS RELIEF SlF SlF SlF REACTIVITY SCRAM CONTEOL TRAM s7-2sem f-s Figure 15A.6-27 PROTECTION SEQUENCES FOR GENERATOR LOAD .h
>g'
- g. REJECTION VVITH ONE BYPASS VALVE FAILURE
. _ . . . , . _ ~ ,.,y -
i I s a 2 EVENT 28 E y e r CONTROL ROD EJECTION 2 ACCIDENT [ mi 5 STATES A. 8, C AND D d 3 l .=.. I j SEE EVENT 32 I I I POWER RANGE HIGH STARTUP RANGE . HIGH FWX ! NEUTRON = FLUX NEUTRON = OR RT ; MONITORING SCRAM ppp l MONITORING SCRAM I 1 SlF ROD CONTROL AND RE FMCRD BRAKE ' INFORMATION =Ep TS MECHANISM = STOPS ROD' pp3 , SCRAM SYSTEM-SIGNAL WORTH S F s F SlF CRD SlF LIMIT REACTIVITY STOP SCRAM INSERT W CONTROL ROD RATE MUU 87-198-10 i i
=
Figure 15A.6-28 5-8
$ PROTECTION SEQUENCE FOR CONTROL ROD EJECTION ACCIDENT T
I 4 O O. O
="
e ~- * *' e _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
L ABWR m6-l Standard Plant REV D as related to the Ultimate Heat Sink. Provide a summary table listing the protective measures provided against the effects of postulated pipe failures in each of the above systems and the systems listed in SSAR Tables 3.6 2 and 3.6 4 (3.6.1) 410.23 Give details for the worst case flooding arising from a postulated pipe failure and include the mitigation features provided. Note that for flooding analysis purposes, the complete failure of non seismic Category I moderate. energy piping systems should be considered in lieu of cracks in - determining the worst case flooding condition. _ (3.6.1) 410.24 Identify all the high energy piping lines outside the containment (but within the ABNR scope), the adverse effects that may result from failures of applicable lines among them, and the protection provided against such effects for each of such lines (e.g., barriers and restraints). (3.6.1) 410.25 Cit.rify whether the reactor building steam tunnel is part of the break exclusion boundary. Also, l provide a s ibcompartment analysis for the steam tunnel. Discuss how the structural integrity of the tunnel and the equipment in the tunnel are protected against failures in the tunnel. (3.6.1) 410.26 State how the MSIV functional capability is protected. (3.6.1) 410.27 Provjde a summary table of the findings of an analysis of a postulated worst case DBA rupture of a high or moderate energy line for each of the following areas: 1) RCIC compartment,2) equipment and valve room,3) other applicable areas outside the containment (e.g., housing RHR piping) (3.6.1) 410.28 Clarify whether protection for safety related systems and components against the dynamic effects of pipe failures include their enclosures in suitable design structures or compartments, drainage systems and equipment environmental qualification as required. If so, give typical examples for the above type of protection. (3.6.1) 410.29 Regarding interfaces (Section 3.6.4.1), include results of analyses of moderate energy piping fail-urcs (currently, the interface requirements address only the high energy piping failures analyses). (3.6.1) 410.29a Appendix 31, " Equipment Qualification Environmental Design Criteria," is scheduled to be submitted in December 1988. Note that additionalinformation may be requested based on review of the above appendix (3.11) O Amendmem 12 20.2-sp
5 MM 23A6100AT Standard Plant n.v n 410.30 Although there are no detailed qualification requirements for safety related mechanical equipment in a harsh environment, GDC 1,'Ouality Standards and Records,' GDC 4, " Environmental Missile Design Vases,*- and Appendix B to 10 CFR 50,
- Quality Assurance Criteria for Nuclear Power Plants and Fuel Processing Plants" (Section 111, ' Design Control," and XVil, " Quality Assurance Records") contain the following requirements related to equipment qualifications:
a) Components shall be designed to be compatible with the postulated environmental conditions, including those associated with LOCAs b) Measure shall be established for the selection and review for suitability of application of materials, parts, acd equipment that are essential to safety-related functions. ' c) Design control measures shall be established for verifying the adequacy of design. ; 1 d) Equipment qualification records shall be maintained and shallinclude the results of tests and material analyses. 1 Clarify whether the design complies with all the above requirements for safety related mechanical equipment in a harsh environment within the ABWR scope. Provide justification for the non compliance items < abow: and identify any interface requiremeats needed to comply with the above. (3.11) 420.6 Identify the topical reports that will be provided to support any aspects of the design that are substantially j different relative to designs previously reviewed by the staff. Subjects addressed in these topical reports should ' include but not necessarily be limited to the following: l Methodology, basis and acceptance criteria for qualifying the system and equipment to the design basis electromagnetic interference (EMI) emironment. (App 31) 420.7 Identify the topical reports that will be provided to support any aspects of the design that are substantially different relative to designs previously reviewed by the staff. Subjects addressed in these topical reports should include but not necessarily be limited to the following: Methodology, basis, and acceptance criteria for qualifying the system and equipment to the design basis surge withstand capability (SWC). (App 31) 420.8 j Identify the topical reports that will be provided to support any aspects of the design that are substantially different relative to designs previously reviewed by the staff. Subjects addressed in these topical reports should include but not necessarily be limited to the following: Methodology, basis, and acceptance criteria for qualifying the system and equipment to the design basis thermal environment established by locahzed heat transfer within electronic equipment, including in non. accident I emironments; this should also address requirements for humidity controls to preclude damage from electrostatic l discharge. (App 31) O
' Amendment 9 20.2-4q
ABWR muur Standard Plant REV D
, SECTION 20.3 ILLUSTRATIONS (Continued)
Figure Iltle Ease 203 29 Flow Out of Vessel Following a 0.0218 ft' Vessel Bottom Head Drainline Break:1 RHR + 3 ADS Available 20 3 284 , 203 30 FlowIntg Vessel Following a 0.0218 ft Vessel Bottom Head Drainline Break:1 RHR + 3 ADS Available 203 284 203 31 Peak Clafding Temperature Following a 0.0218 ft Vessel Bottom Head Drainline Break:1 RHR + 3 ADS Avallable 20 3 285 203 32 Wate.( Levelin Fuel Channels Following a ; 03 ft* Break in the RHR Vessel Shutdown ; Suetion Line:1 RHR Available 203 285 l 203 33 Wate,; Levellnside Shroud Following a 03 ft* Break in the RHR Vessel Shutdown Suction Line:1 RHR Available 203 286 203 34 Wate3 Level Outside Shroud Following a 03 ft Break in the RHR Vessel Shutdown Suction Line:1 RHR Available 203 286 203 35 Vesse,I Pressure Following a 03 ft' Break in the RHR Vessel Shutdown Suction Line:1 RHR Available 203 287 . 203 36 Flow.put of Vessel Follewing a 03 ft' Break in the RHR V-ssel Shutdown Suction Line:1 RHR Availabh 20 3 287 203 37 Flow,lnto Vessel Following a 03 ft* Break in the RHR Vessel Shutdown Suction Line:1 RHR Available 20 3 288 203 38 Peak.pladding Temperature Following a 03 ft' Break in the RHR Vessel Shutdown Suction Line:1 RHR Available 20 3 288 203 39 Reactor Water Cleanup System (Response to Question 72532) 20 3-289 203 40 Characteristic Response for Damped and Undamped g System (Response to Question 725.53) 203 290 (j 203 vi Amendment 12
ABM Standard Pjant 2 m io.41 uv n . t ILLUSTRATIONS (Continued)
- 1. O: .
203 41 Pressure Time History After A Feedwater Line Break Available ECCS: 9 RHR System 20 3 288.1 203 42 Temperature Time History After A , Feedwater Line Break Available ECCS: ' 1 RHR System 20 3-288.1 203 43 Simplified Functional Control Diagram of Turbine Digital Control & Monitoring System (Typical of One of Three Channels) 20 3 310 I i l l O l f 203 vii Amendment 11
ABWR m-r Etandard Plant arv n QUESTION 252.11 Subsection 5.23.4.23 states that the ABWR design meets the intent of this Regulatory Guide (1.71) by utilizing the alternate approach given in Section 1.8. We cannot review this subsection because we have not received Section 1.8. In addition, this subsection should be rewritten because it lacks detailed discussion about welder qualification. RESPONSE 252.ll Response to this question is provided in revised Subsection 5.23.4.23. QUESTION 281.1 In Section 5.1 (page 5.12) the function of the reactor cleanup system filter demineralizer should include the removal of radioactive corrost n and fission products in addition to particulate and dissolved impurities, l RESPONSE 281.1 Response to this question is providcd in revised Section 5.1. QUESTION 2H1.2 i in Subsection 5,23.2.2 (page 5.2 7) irradiation assisted stress corrosion cracking (IASCC) of reactor internal components and its mitigation are not discussed. Present laboratory data and plant experience has shown that IASCC can be initiated even at low conductivity (< 03mS/cm) after long exposure to radiation. t RESPONSE 2RI.2 Response to this question is provided in the new Subsection 5.23.2.2.4,14CC Considcrations. l QUESTION 281J In Subsection 5.23.2.2 (pages 5.2 7 and 8) the ABWR standard plant design does not clearly incorporate hydrogen water chemistry to mitigate IGSCC. Since the plant design life is 60 years, hydrogen water chemistry may be of greater importance in reducing reactor coolant electrochemical i corrosion potential to prevent IGSCC as well as IASCC If hydrogen water chemistry is the referenced ABWR standard design, the following documents should be cited: EPRI Nr.528.%SR.A, Guidelines for Permanent Bil'R Hydrogen li'ater Chemistry installations 1987 ; Revision. EPRI NP.4947.SR.LD, Bil'R Hydrogen li'ater Chemistry Guidelines .1987 Revision (to be published). RESPONSE 2813 Response to this question is provided in revised Subsection 5.23.2.2. Amendmcni 12 203 9
MN tsA61ooAT standard Plant arv n QUESTlON 281.4 in Subsection 5.2.3.2.2 (page 5.2 9) the utilization of the General Electric rinc injection ! passivation (GEZIP) process for radiation buildup control for the ABWR is not discussed. GEZIP was identified as a requited design feature in the ABWR presentation to NRC staff. , RESPONSE 281.4 1 The General Electric rinc injection passivation process (GEZIP) is not in the ABWR scope. However, an interface requirement has been added (see new Subsection 5.7.6) that requires the , remainder of the plant to meet the water quality requirements of Table 3.2 5. l QUESTION 2AIJ i l In Subsection 5.2.3.2.2 (page 5.2 9) prefilming of stainless steel appears to be a promising method to reduce the buildup rate of activated corrosion products during subsequent plant operation. SIL No. 428 recommends preoperational testing of the recirculation system conducted at temperatures j 230*F be done with the dissolved crygen level controlled to between 200 and 400 ppb. Is control , of radiation buildup through preoperational oxygen control being considered for the BWR standard ' plant? Are mechanical polishing and electropolishing of piping internal surfaces also being ! considered for reducing radiation buildup? RESPONSE 2NI.5 Since the recirculation system piping has been climinated from the AllWR design, SIL No. 428 does not apply. Preoxidation, mechanical polishing, and electropolishing are not being considered for other ABWR components at this time. However, these methods are available as promising techniques to reduce radiation buildup on all internal stainless steel surfaces. QUESTlON 281.6 In Subsection 5.2.3.2.2.2 (page 5.2 9) cobalt 60 is identified as the principle contributor to shutdown radiation levels, especially the recirculation piping system of BWRs. Stellite contributes about 90% of the total cobalt 59 input to the reactor water (EPRI NP 2263, Bil'R Cobalt Source Identification, February 1982). Since irradiation of cobalt 59 yleids cobalt 60, reduction in the source of cobalt 59 is needed to reduce the buildup of shutdown radiation levels. Indicate Sicilite surface areas (square feet) in nuclear steam supply system and balance of plant. Provide the criteria for selecting Stellite plant materials for the designed application. Provide evaluation of noncobalt.containing materials whose properties arc adequate to replace Stellite in plant applications. RESPONSE 2NI 6 I (1) Stellite Surface Area for BWR/6: Total Nuclear Steam Supply System: 7439 Sq. Ft. Toral Balance of Plant: 138.0 Sq. Ft. For ABWR design, the above numbers are greatly reduced. Cobalt. based alloys have been climinated from fuel assemblies, and control rod blades and drives. l O> Amendment 3 20.3 10 i
21A6100AT Standard Plant arv n RESPONSE 43031 The plant protection signals that automatically isolate the secondary containment and activate the SGTS are: (1) Secondary containment high radiation signal. ' 1 (2) Refueling floor high radiation signal. (3) Drywell pressure high signal. (4) Reactor water levellow signal. (5) Secondary containment HVAC aupply/eahaust fans stop. Isolation of the secondary containment is accom)48 ted by closure of the secondary containment IIVAC supply / exhaust line ducts which pass through the secondary containment boundary. The HVAC isolation valves consist of two valves in series in each of the supply /exhausi lines. These valves are air operated, normally open, fall closed butterfly vales. Further details are provided in Subsection 6.23,9.4.5.1 and Section 6.5 QUESTION 43032 Identify and tabulate by sire, piping which is not provided with isolation features. Provide an analysis to demonstrate the capability of the Standby Gas Treatment System to maintain the dedgn ne. gative pressure following a design basis accident with all non isolated lines open and the event of ( the worst single failure of a secondary containment isolation valve to close, (6.2) RESPONSE 43032
, Response to this question will be prcvided in a future amendment.
QUESTION 43033
. Discuss the design provisions that prevent primary containment leakage from bypassing the secondary containment standby gas treatment system and escaping directly to the environment. Include a tabulation of potential bypass leakage paths, including the types of information indicated in Table 618 of Regulatory Guide 1.70, Revision 3. Provide an evaluation of potential bypass leakage paths considering equipment design limitations and test sensitivities. Specify and justify the maximum allowable fraction of primary containment leakage that miy bypass the secondary containment
, structure. The guidelines of BTP 6 3 should be addressed in considering potential bypass leakage l paths. (6.2) RESPONSE 43033 ! The secondary containment completely surrounds the primary containment except at the basemat, in i addition the lower third of the secondary containment is surrounded by soil, thereby reducing leakage paths. No measurable leakage is expected through its walls except at penetrations. The secondary containment will be maintained at subatmospheric conditions to prevent leakage from bypassing the secondary containment. Only valve leakage through process piping can bypass the secondary containment. This leakage will be monitored via the containment leakage test type C on the outboard l containment isolation valves. The secondary containment leak rate calculation is provided in the [d) response to Question 430.52c. Amendment 12 20..b18
21A61DQAT Etanda si Pl==e arv c QUESTION W34 Provide a list of the secondary containment openings and the instrumentation means by which each is assured to be closed during a postulated design basis accident. (6.2) RESPONSE G.34
. Response to this question is provided in revised Subsection 6.23.2 and new Table 6.2-9. ,
QUEST 10N m35 Provide a table of design information regarding the containment isolation provisions for fluid system lines and fluid instrument lines penetrating the containment which are within the GE scope of the ABWR design. Include as a minimum the following information: (1) General design criteria or regulatory guide recommendations that have been met or other defined ' bases for acceptability; ' (2) System name; (3) Fluid contained; 1 (4) L.ine size; 4 0 O Amendment 9 20.3 48.1
ABWR zwwwr Standard Plant nrv n QUESTION 440.4 In Figure 4.6 Ba, CRD system P&lD, sheet 1, piping quality classes AA D, FC.D FD D, FD B, etc.
'(~) are shown. Submit the document which crplains these classes and relates them to ASME code classes.
RESPONSE 440.4 This information is scheduled to be included in Section 1.7. Essentially, the first two letters of the codes specify the pipe primary pressure rating (150 lb.,900 lb., etc.) the type of service (condensate or reactor water, steam, etc.), and material (carbon or stainless steel). The symbols
*A*, *B* and 'C' represent AShiE Section III, code Classes 1,2, and 3, respectively. The symbol *D*
- represents AShtE Section 8, or ANSI B31.1 or other equivalent codes. j l
QUESTION 440.5 l l In figure 4.6 Sb, the leak receiver tank is shown. What is the function of this tank? How big is this tank? Will a high level in the tank impact the operation of the control rod drive? RESPONSE 440.5 This leakage collection tank is no longer part of the design. The intent of the leakage collec. tion system was to assist the operator in identifying which drives were potential candidates for seal replacement during plant outages, which would facilitate plant maintenance planning. Flowever, the de. sign could not provide the level of differentiation of leakage between individual drives needed for this purpose and was therefore deleted. Figure 4.6 8b will be updated at its next revision to document this change. ( QUESTION 440.6 Identify the essential portions of the CRD system which are safety related. Confirm that the safety related portions are isolable from non essential portions. (4.6) RESPONSE 440.6 The essential portions of the CRD system which are safety related are:
- (a) The hydraulic control units (llCUs),
[ (b) The scram insert piping from the HCUs to the fine motion control rod drives (FhtCRDs), and l (c) The ThtCRDs (except the motors) The non. essential portions of the CRD system interface with the essential portions at the follow. ing connections to the HCUs: (a) The accumulator charging water line (b) The Ph1CRD purge water line, and (c) The scram valve air supply from the scram air header. lf) i Amendment 12 20..tM i
MM 2146100AT Etand=4 Plant krv s i The safety related portions of the HCU and the scram function are protected against failure in the non cssential portions of the charging water and purge water lines by check valves. Also,instrumen-tation in the charging water line provides signals to the reactor protection system to cause reactor scram in the event of loss of charging water pressure. Loss of pressure in the scram air header causes the scram valves to actuate, resulting in reactor scram. This fall safe feature is the same as provided on current BWR designs using locking piston type control rod drive. The FMCRD provides the following functions:
- 1. Normal rod positioning in response to commands from the rod control and information system. ,
- 2. Rapid contr01 rod insertion (scram)
- 3. Scram follow function (post scram electric motor run in)
- 4. ATWS FMCRD run.in, and j 5. Selected control rod run in (for stability control). :
Of these, only the hydraulic scram function (Item 2) is classified as a safety related function. The other functions, all of which involve positioning the rods by means of the FMCRD ekctric motors, are designed for high reliability but are not classified as safety related. Therefore, because the FMCRD motor does not preform a safety related function,it is classified as non Class IE. ! With respect to ATWS in particular, the FMCRD run in function is considered an ATWS mitigating l system. As such, the NRC design guidance regarding system and equipment specifications for 10CTR$0.62, was applied. This guidance states that the ATWS mitigating systems are not required to be safety related; therefore, the FMCRD motors are not required to be Class 1E. However, the automatic FMCRD run in function for ATWS is designed to provide high reliability. Some features to be noted are:
- 1. FMCRD run in provides means of control rod insertion that is diverse from the hydraulic insertion of both the normal scram (RPS generated) and ATWS ARI (venting of Scram air header). Both of these independent functions must fall before FMCRD run in is needed to shut down the nactor.
- 2. The automatic FMCRD run in utilizes initiation signals (high reactor pressure or low water level 2) which are diverse from the RPS.
- 3. The FMCRD run in controls and instrumentation are powered from non. divisional, l
non interruptible DC power independent from RPS power.
- 4. The FMCRD motors are connected to the divisional power buses which are connected to the emergency diesel generators. This allows FMCRD run in during any loss of offsite power event. The divisional power assignments throughout the core for the FMCRD motor power supplies are in a ' checkerboard" pattern. This arrangement provides the capability to achieve hot shutdown even with failure of offsite power and one of the diesel generators (a degraded ATWS condition beyond the design basis). Under these circumstances the operator would have time to reestablish offsite power or startup of the failed diesel generator to achieve cold shutdown. As a last resort, manualinitiation of the SLCS would always be available to achieve cold shutdown.
- Amendment 8 20M5
MNk 23A6100AT
- RennAmed Plant MY B QUESTION 410.6 Discuss the means by which stored spent fuel is protected from damage by internally generated missiles. (3.5.1.1)
RESPONSE 410.6 There h no credible missile generating equipment on the refueling floor except for the load drop from the reactor building cranc. The reactor building crane will not be allowed to travel over the spent fuel pool when carrying heavy loads (see Section 9.1.5 for details). QUESTION 410.7 Section 3.5.1.1.1.4,'Other Missile Analysis," discusses the example of analysis of a containment high purge exhaust fan for a thrown blade. Provide the details of this analysis, such as the maximum penetration of the blade and the thickness of the fan casing. Discuss whether this analysis is conservative with respect to other rotating equipment missile sources. (3.5.1.1) RESPONSE 410.7 Because the ABWR is a standard plant, the specific details will only be available following procurement of equipment on a specific application referencing the ABWR design. Therefore, when performing potential missile assessment analysis, representative equipment is selected with judgement applied for worse case analysis. The containment high purge exhaust fan is judged to represent a worst case analysis. The containment high purge exhaust fan from the standard BWR 6 plant was analyzed for a thrown O blade at rated speed conditions as stated in Subsection 3.5.1.1.1.4. It was determined that the maximum thickness this blade could penetrate was conservatively 0.12 inches. Since the fan casing is 10 gage (0.134 inches), the blade would not escape and consequently P i si less than 10'7 per year. QUESTION 410.8 Regarding Section 3.5.1.1.2.2, ' Missile Analysis," provide the details of the rack, strap and cover assemble design for the pheumatic system air bottles, showing the thickness of the steel cover i and the distance to the concrete slab. (3.5.1.1) RESPONSE 410.8 Because the ABWR is a standard plant most of the specific details requested will only be available following procurement of equipment on a specific application referencing the ABWR design. The statement under item (3), Pressure Vessels, of Subsection 3.5.1.1.2.2 indicating that the bottles are not considered a credible source of missiles is based on a qualitative rather than a quantitative analysis, item (3) of Subsection 3.5.1.1.2.2 has been revised accordingly. QUESTION 410.9 Regarding Section 3.5.1.1.3,' Missile Barriers and Loadings, provide a list of all local shields and barriers outside intended to mitigate missile effects, giving their specific locations and design data. Provide an example of an analysis showing that the design of the shield or barrier will ' withstand the most energetic missile which could credible impact it. (3.5.1.1) Amendment 7 20.342 ____________________________o
r-ABWR maar Standard Plant nrv n RESPONSE 410.9 No local barriers or shields outside the containment are used for mitigating missile effects. Each safety related system is contained in its own room of a Seismic Category I building. The walls, floor and ceiling of this room act as the missile barrier or shield from missiles generated outside l this room. 1 O i l l O Amendment 12 20.3 82.1
ABWR awxwr Standard Plant RD' H > 20.3.6 Response to Sixth RAI Reference 6 , QUESTION 440.75 In the ABWR design, the HPCF is tested by taking suction from and returning water to the suppression pool. Normally the suppression pool water is a lower quality than that of the CSTt Therefore, draining, flushing and refilling the system is required prior to returning the system to standby after testing. Please discuss the pros and cons of using the CST for testing the HPCF system. (6.3) RESPONSE 440.75 The arguments supporting this design decision for the HPCF system are similar to those for the RCIC system. Please refer to Response to Question 440.41 for that discussion. In addition it is even more unlikely that marginally less clean suppression pool water would be injected to the vessel with HPCF since its initiation set point on low unter level has been lowered to Level 11/2 as compared to Level 2 for RCIC. QUESTION 440.76 Address the following TMI 2 action items related to ECCS. (6.3) (a) II.K.1.5 (b) II.K.1.10 (c) II.K.3.17 (d) II.K.3.18 (d) II.K.3.18 (c) II.K.3.21 (Q) (f) II.K.3.25 (g) li,K.3.30 (h) II.K.3.31 RESPONSE 440.76 These TMI 2 action items will be addressed in Appendix 1A scheduled to be submitted by January 30,1989. QUESTION 440.77 Confirm that the HPCF system meets the guidelines of Regulatory Guide 1.1 regarding pump Net Positive Suction Head (NPSH) (6.3) RESPONSE 440.77 Response to this question is in revised Subsection 6.2.2.3.1 and new Table 6.2 2C, QUESTION 440.78 SRP 6.3 identifies GDCs 35,36 and 37 in the acceptance criteria. Confirm that the HPCF system, described in Chapter 6.3 of the SSAR, meets the requirements of the above GDCs. (6.3) O N/ Amendment 12 2031M
ABWR awau Standard Plant arv n ; J RESPONSE 440.78 The HPCF system, as described in Section 6.3, meets tbc requirements of GDCs 35,36 and 37 as discussed in Subsections 3.1.2.4.6,3.1.2.4.7 and 3.1.2.4.8, respectively. QUESTION 440.79 , Normally, the HPCF pump takes suction from the Condensate Storage Tank (CST). But, the CST is not seismically qualified or safety related. Confirm that the system piping and level transmitters, which interface with CST, will be designed and installed such that the automatic switchover to the suppression pool takes place without failure. (6.3) RESPONSE 440.79 , The redundant CST level transmitters that input to the HPCF suction source swapover logli are Class IE. The HPCF suppression pool and CST suction valves are siesmically qualified and safety.related, as is the HPCF suction piping to a boundary interface upstream of the CST suction ' valve. This will ensure that the awomatic swapover takes place without failure. QUESTION 440.80 What is the minimum quantity of water required in the condensate storage tank (CST) for HPCF operation? Give the basis for the required quantity of water in the CST. (6.3) RESPONSE 440.80 There is no minimum CST water volume specifically reserved for HPCF makeup to the reactor. As - stated in the response to Question 440.54, the 150,480 gallons (Table 9.2 3) of water in the CST are stored for use during station blackout. This volume is that required for eight hours operation of > the RCIC. There is no minimum quantity of water required in the CST for HPCF operation. The RCIC ' l and HPCF share in the CST water and both can use water from the suppression pool. ( QUESTION 440.81 What is the closing time of test return valves F009.01C, F011.01C, F015B and F016B7 They should close earlier than 36 seconds to prevent any flow diversion to the suppression pool during a LOCA (6.3) RESPONSE 440.81 , The HPCF test return valves are specified as manufacturers standard with a nominal closing speed of 12 inches per minute. For an eight inch line, this will result in a closing time of about i 40 seconds which is reasonably consistant with the required system initiation time. The present design is considered adequate given the very low probability of a LOCA occurring while the system is lined up in the test mode. Additionally, standard valves of this design are known to be more i reliable than special fast closing valves. QUESTION 440.82 Since HPCF is part of the ECCS network, the HPCF pump minimum flow line should be designed to operate for a reasonable length of time. How long can HPCF run in minimum flow mode? (6.3) O Amendmcni 5 20.3 145
ABM m.m pi... sinaisasi ma No of Addit. % d Roda la lamal Failures Failed Cannamani MCER BalMaqi'hanaltian (a) 0 - 1.10 Meet traanient criteria (b) 1 1 Bypass valve fails to open 1.10 Meet transiest criteria (c) 1 1 RIP fails to trip ce RIT 1.10 Meet treasient criteria (d) 2.3 Allbypass valm failto open 1.07 Meet transient criteria (c) >3 RPT fails 1.04 0.3 (f) >6 Allbypass valves failto opea 0.97 3.0 and RPT fails Events (b) to (f) above meet all criteria for limiting faults, la fact, no fuel failure is expected since the time in the boiling transition ig very short (<5 seconds), and the peak clad temperature (PCT) would be less than 1100 F even without rewetting credit. QUESTION 440.113 ; You have classified the trip of all reactor internal pumps as a limiting fault. Thg is based on your assumption that the loss of greater than three reactor laternal pumps is 10 per year. Provide operating experience data to justify this failure rate. (15) RESPONSE 440.113 There is no U.S. operating experience data for plants with reactor internal pumps (RIPS). The following data are obtained from an European reactor vendor: (1) With 6 BWRs having 44 RIPS in operation of about 40 operating years, there is no event invohing trip of more than one RIP. All trips involve only one RIP. (2) All single RIP trips were caused by faulty trips in adjustable speed drive (ASD) due to noise in ASD protective circuits. ! From these data, it is coecluded that a trip of more than one RIP simultaneously is extremely unlikely. The ASD design has also been modified to reduce the probability of occurrence much less than once per plant life. So it is proper to classify the trip of all RIPS as an accident, and it is conservative to treat a trip of 3. RIPS as a moderate. frequency event. 4 QUESTION 440.114 The ABWR feedwater control system and the steam bypass and pressure control system use a triplicated digital system. You claim that no single failure in these systems will cause a minimum demand to all turbine control valves and bypass valves or the runout of two feedwater pumps. (15) QUESTION 440.114a , 1 What is the reliability of the system? l O Amendment 4 20.M60 i i
ABM 2246ioort Atandard Plant Rrt n ! RESPONSE 440.114a The systems are required to have a Mean Time Between Failure (MTBF) of greater than 1x105 hours. This requiremei.ts is on the failure rate of the system function, not single components. QUESTION 440.II4h What design feature of these systems prevent common mode failure to more than one channel? RESPONSE 440.114b The source of any common mode failure would be a design error in the hardware or processor - software. Design procedures for control and instrumentation design were designed to preclude such design errors. These procedures specify requirements for all stages of software design, verification and validation. Specifically, requirements for the verification and validation of hardware and software specify procedures required to exhaustively test hardware and software prior to use. Through this process, common mode failures due to hardware and software design errors shall be avoided. QUESTION 440.114e What protection is provided in these systems against a technician disabling a second channel while performing maintenance on the first. RESPONSE 440.llde There are many features of the process control systems which protect against errors propagating between channels regardless of the cause of the initial error. The following discussion focuses on interfaces between channels and the features that provide protection against the propagation of faults. The process control system channels operate independently using separate hardware components, different power supplies and separate, independent input signals. Interprocessor communication links are used to detect faults and avoid divergence of integrator states in each processor. The process of selecting the appropriate integrator state to use in all 3 processing channels is accomplished through the use of middle value voters included in each channel's software. This assures that a failure in one channel cannot adversely afftet the operation of the other 2 channels, in addition, the interprocessor communication paths are required to have adequate transient protection to prevent prorgation of faults between processors. For these reasons the interprocessor communication feature of the fault tolerant digital controllers en not be the source of common mode failures. Each of the three channels generates its own control signal. The three resulting control signals (1 from each channel) are converted to a single control actuator signal, locally, by a fault j tolerant, middle value voter. The output of 'he voter (the final control signal) is fed back to each of the three processors for validation. This sllows the controllers to detect a voter failure, if 2 of the 3 channels determine that a voter has failed, a lockup signalis sent to the control actuator, which locks the control actuator at its last p3sition. This feature prevents the possibility of the final output voter failure causing a failure of the system function. While a ' technician is performing maintenance on an out of service channel, the other 2 channels will recognize that the out of service channelis faulted. The technician, at worst could cause the out of service channel to generate faulty signals on the interprocessor communication links or output control signal voters, these faulty signals would be detceted as faulty and ignored. Amendmrnt 12 20.3 161}}