ML20070D938
| ML20070D938 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 07/12/1994 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Boyce T Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9407130180 | |
| Download: ML20070D938 (44) | |
Text
, )
GE Nuclear Energy a.. .+ . , . c ,. c ,.,
tb , , C.) [t61 4 July 12,19')4 Docket No.52-001 Tom Boyce, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Subject:
Final SSAR and CDM Modifications
References:
- 1. Letter, Dennis M. Crutchfield to Joseph Quirk, Reliability Assurance Program (RAP) Requirements, July 8, 1994
- 2. Letter, Joseph F. Quirk to R.W. Borchardt, Submittal of Amendment 35, Non Proprietary Information to GE's ABWR S3AR and Certified Design Material Revision 4
- 3. Letter, J.h Fox to Distribution ABWR SSAR Amendment 35 and CDM Revision 4 Modification Package, June 23, 1994
Dear Tom:
In responce to Reference I and the July 12, 1994 meeting in Rockville, Maryland, we have completed the attached D RAP Design Description and ITAAC for inclusion in Section 3.6 of the CDM.
Also attached are CDM and SSAR markups addressing minor corrections.
The D-RAP Design Description and ITAAC and these markups will be included in a final modification package further supplementing References 2 and 3.
Sincerely, k $tf3 JMk Fox Advanced Reactor Programs !
1 cc: Alan Beard (GE) '
Norman Fletcher (DOE) ;
Joe Quirk (GE)
\'
JNF94-044 4
' h_ lY J 9407130180 940712 PDR A
ADOCK 05200001 PDR e
((f
l a 1 D-RAP DESIGN DESCRIPTION AND ITAAC 1
l l
l
ABWR certisedoesign Material 3.6 Design Reliability Assurance Program Design Description The Design Reliability Assurance Program (D-RAP) is a program that will be performed during the detailed design and equipment specification phase prior to initial fuel load.
The D-RAP evaluates and prioritizes the structures, systems and components (SSCs) in the design, based on their degree of risk significance. The D-RAP will identify the dominant failure modes for the risk-significant SSCs. The D-RAP will also identify the key assumptions and risk insights for the risk-significant SSCs.
The D-RAP scope includes risk-significant SSCs as determined by probabilistic, deterministic, or other methods used for design certification to identify and prioritiz.:
risk-significant SSCs.
The D-RAP purpose is to provide reasonable assurance that the plant design proceeds in a manner that is consistent with the original bases and design assumptions for the risk insights for the risk-significant SSCs.
The D-RAP objectives are to provide reasonable assurance that the plant is designed such that: (1) it is consistent with the assumptions and risk insights for these risk-significant SSCs, (2) the risk-significant SSCs will not degrade to an unacceptable level during their design life, (3) the frequency of transients that challenge these SSCs will be acceptably low, and (4) these SSCs will function reliably when challenged.
Inspections, Tests, Analyses and Acceptance Criteria Table 3.6 provides a definition of the inspections, tests, analyses, and associated .c acceptance criteria, which will be performed for Advanced Boiling Water Reactor (AlnVR)D-RAP.
I l
l l
i Design Reliability Assurance Program 3.6-1
^
Table 3.6 Design Reliability Assurance Program b
{ 03 Inspections, Tests, Analyses and Acceptance Criteria Acceptance Criteria "*
Design Commitment inspections, Tests, Analyses
- 1. The Design Reliability Assurance Program 1. Inspections of the design reliability 1.
(D-RAP) includes; scope, purpose, assurance program will be conducted. a.
Documentat. ion exists that describes objectives; the process used to evaluate the scope, purpose, and objectives of and prioritize the structures, systems and D-RAP used during plant design, and components (SSCs); and the list of SSCs conclydes that the detailed design of designated as risk-significant. For those nsk-significant SSCs i,s consistent SSCs designated as risk-significant, the with the D-RAP Design Desenption.
process used to determine dominant failure modes considered industry b. Documentation exists and concludes experience, analytical models, and that the process (probabilistic, deterministic, or other methods) used applicable requirements. Also, for those SSCs designated as risk-significant, the to evaluate and prioritize the SSCs in key assumptions and risk insights the design is based on the risk-considered operations, maintenance, and significance of the SSCs.
monitoring activities. c. A list of SSCs exists that is based on the risk-significance of SSCs.
- d. For those SSCs designated as risk significant:
i) Documentation exists and concludes that the process to determine dominant failure p modes considered industry g experience, analytical models, and applicable requirements.
E ii) Documentation exists and O
$ concludes that the key B.
( assumptions and risk insights from probabilistic, deterministic, k
g or other methods considered $
Q g
operations, maintenance, and monitoring activities.
{g to Q 8-s
a -
t OTHER CDM MARKUPS i
1 1
1 l
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s 25AS447 ROV. 2 certisedoesign usterias ABWR MUWP RCW RCW OTHERS OTHERS RCW 3 P
I (Reactor Building)
I J
! MR SURGE TANK Q
L l DG HX ] u (Reactor Butiding)
(Reactor Building) l ] 3 RCW M ! 1 Y
' Fuel Pool Cooling HX and Room Coolers ~
' [E)
- ' L TO z :' (Reactor Building)
I HECW 3 NNS NNS 3 I I l OTHER (SAFETY RELATED) HXs j (Reactor and Control Building) j OTHERS RCW RCW l OTHERS NNS FE R NNS , 3 CAD AND CUW PUMPS N, z_z sc as ~~ Td (m Building)
T~
_____3l N_NS_j i j [N_NSl"3
, , ______)
! __ _j Cuw Hx.HWs Hx b_.___ _ _ _y f___ g (Reador Bueiding)
W lCI. '!r?! 0.". m e g-=0; NON.7A8" W - M A %D NX $
~ ~~~~~~
(Control Buildog) 3 I s s
t ~ / 1 R l 8
,__ _ _ _ __._ _ _ q _~-,- p________ql L_hf wl_d oRvwEttEOuleuENTCOOLER NNS 2 2 2 NNSI
( " ' - J NNS 2 ( 2l NNS m-~:n RCW HX ' R FROM (Control Buildmg) 3 RCW C RSW W Q RSW TO RSW RCW PUMP RM I 0" "9)
RCW HX z au (Control Building) 3 RCW R TO RSW ' R RCW HX C (Control Building) 3 RCW RCW PUMP FROM (Control Building)
- 1. ALL ELECTRICAL POWER LOADS FROM THE CLASS 1E COMPONENTS SHOWN ON THIS FIGURE ARE POWERED FROM CLASS 1E DIVISION I EXCEPT FOR THE OUTBOARD CONTAINMENT ISOLATION VALVE. WHICH IS POWERED FROM D' VISION 11.
Figure 2.11.3a Reactor Building Cooling Water System (RCW-A) 2.11.3-5 Reactor Building Cooling Water System
- - na6447 Riv. 4 I
ABWR certisedoesign usterisi \
MUWP RCW RCW OTHERS 3 3
FE SPCU RCW
' ' u (Reactcf 'ouilding)
-J -
I ! SURGE TANK (React r Bukhng)
D H
' (Reactor Building) i r_.as -
' J 3 RCW M l HECW
_ y Fuel Pool Cooieng HX and Roorn Coolers V
e (Reactor Building) u L
' d 3 NNS NNS 3 I
i OTHER (SAFETY-RELATED) HXs
___d l (Reactor and Control Buiking) '
l FICW OTHERS OTHE9S RCW tJNS NNS FE R r-- - - - - - ~ q CUW PUMP q lN S I e (ReaMW) J NNS r --------'q ir-- - - - - -
I CUW Hx. HWH HX
,i 1
8
___]'
l l--- -- ( R",,4ctor Building) i i A =_ __ _ gr A _ _om _$0N-SAFETY-RELATEDf W p .______ -- 8 l R , (ControlSmiidsgendRea iktng) a M p._ , _ , ,__ _ ., ,_
_T'* b ** M fM U L -
w q onYwEtt EouieuEur cootERS s
( Reactor Budding) j ;
NNSl2 2lNNS _ .J NNSl2 /
" 2lNNS
.4 m RCW HX '
R u (Controt Buidng) 3RCW g h RSW RCW PUMP RM TORSW RCWHX I I rz (Control thmksng) 3 RCW RSW R
R TO RSW ,
R RCW Hx C (Contred Eksking) 3RCW SW RCW PUMP kR (Control Buiktng)
TO RSW ,
i NOTES.
- 1. THIS DIVISION IS POWERED FROM CLASS 1E DtvlSION H.
- PRIMARYCONTAINMENT
'- EXCEPT FOR THE CONTAINMENT OUTBOARD ISOLATION VALVE. WHICH IS POWERED FROM DMSION III.
Figure 2.11.3b Reactor Building Cooling Water System (RCW.B) 2.11.3-6 Reactor Building Cooling Water System
', 25AS447 R1v. 2 certisedoesign nesterial ABWR
, MUWP RCW ra 3 RCW OTHERS OTHERS RCW SPCU RCW P
- 3 3
l FE T 1
~1
, RHR HX l L (Reactor Building) SURGE TANK L_ _
(Reactor Building) l -
~
' (Reactor Building)
' "" l V
- TO HECW l ,
, OTHER (SAFETY-RELATED) HXs a
' _(Reactor and Control Building) ,
RCW OTHERS 01HERS RCW NNS NNS FE - - - _ . ,
CRD PUMP ,
r_m a d (Reactor Building)
~
NNS l"3 3l NNS I ,,,,,,, _ _
E _ ___ - _ . ,
p_______4__.j, e**a #_W-SAW-RELME Xs p__,,j i
(Radwaste & rtune Building) i g _
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^ =C====== s ___I L______________j (Turbine Building) ,
4 o u s a es T7 7.Guw 0 R X3 V e -a i
RCW HX FROM (C< ntrol Building) 3RCW _ C RSW W RSW TO R RCW PUMP M
I U"D"9)
RCW HX (Cenot Buieng) 3 RCW FROM RSW W Q SW TO RSW ,
RCW HX C '
NOTES:
- 1. ALL ELECTRICAL POWER LOADS FOR 1HE CLASS 171 COMPONENTS SHOWN ON THIS FIGURE ARE POWERED FROM CLASS 1E DfVISION lli.
Figure 2.11.3c Reactor Building Cooling Water System (RCW-C) 2.11.3 7 Res uor Building Cooling Water System
2SAS447 Rev 3 ABWR connedonsiv mtyrist A*j 0 0 $ S4000 mm
- h. 0 ?) 0 1300 mm- * .-10500 mm~ ~ 8000 mm . 1300 mm-+
- 1300 mm . 8500 mm *
- 8000 mm - +-10500 mm~
9500 mm ,
h l w
f \
/ =
w 10500 mm e '
/ \
57000 mm A#A / '\
/ \ A/A 0 _ _ &
4 8000 mm 3 B
bg 27o' 3J ,
i 4
- 5 yd 8000 mm
_ +2 0 O r/9 y .,
j 4 ;
10500 mm \ /
b \,/
8 V 19 5
/
/
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\n %
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/ \ %/ surry-i
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,t s, --
7 I
1300 mm 16 0 '
AM FULL POWER / SHUTDOWN OPERATION (pSv/h)
As6 8 < 10 C < 50 D < 250 E < 1000 F 21000 Figure 3.2m Reactor Building Radiation Zone Map for Full Power and Shutdown Operations, Floor 4F-Elevation 31700mm gy,pg Radiation Protection
9 SSAR MARKUPS 1
, 23A6100 Rev.4 ABWR Standant Salety Analysis Report
(
Table 1.6-1 Referenced Reports ABWR SSAR Report No. Title Section No.
22A0270 UIOT Inhoid Spu b ben.GFi.2,"900 *00.0 -
l 22A7007 GESSAR II,238 Nuclear Island, BWRM Standard Plant, General 3.7 Electric Company, March 98 & Amendments 1-21. 19.2 19.3 I9So 19D.3 19D.7 19E.2 19E.3 APED-5750 Design and Performance of General Electric Boiling Water 5.4 Reactor Main Steam Line Isolation Valves. General Electric Company, Atomic Power Equipment Department, March 1969.
NEDO-10029 An Analytic Study on Brittle Fracture of GE-BWR Vessel Subject 5.3 to the Design Basis Accident, Jee i 9 GS.
NEDO-10299A H.T. Kim, Core Flow Distribution in a Modern BWR as Measured 4.4 in Monticello, October 1976.
C.J. Paone and J.A. Woolley, Rod Drop Accident Analysis for 15.4 NEDO-10527 Large Boiling Water Reactors Ucensing Topical Report, March 1972.
F.G. Brutchscy, et al., Behavior oflodine in Reactor Water 15.2 NEDO-10585 During Plant Shutdown and Startup, August 1972.
H.T. Kim, Core Flow Distribution in a Large Boiling Water 4.4 NEDO-10722 Reactok. .asDecember
- u. 3 n 1972.J m Quad CAs A~t 1 4.4 NEDO-1072jA2 H.T. Kim, Core Flow Distribution in a Large Boiling Water I
7 Reactor as Mea. cured in Quad Cities Unit 1, August 1976.
NEDO-10802 - A R.B.1Jnford, AnalyticalMethods of P ant Transients Evaluations 4.4 ,
l for the GE BWR Korif 197h 0R % 0r t964- l R.B. Linford, AnalyticalMethods of Plant Trpnsients Evaluations 4.4 l NEDO-10802-01 A for the GE BWRievision TJ A~oba ~t i Ow%# 19 BC- s R.B. Lintord, AnalyticalMethods of Plant Transients Evaluations 4.4 l NEDO 10802-02 A for the GE BWR,y"^ - .4~4mk 2., O t.cd* e t9 8G 11,1 NEDO-10871 J.M. Skarpelos and R.S. Gilbert, Technical Derivation of BWR 1 1971 Design Basis Radioactive Material Source Terms, March 1973.
~H.T. Kim GeneralElectric Thermal Analysis Basis (GETAB): 4.4 NEDO-10958-A Data, Correlation and Design Applications (LTR), January 1977. 48 NEDO-11209-04-A GE Nuclear Energy Quality Assurance Program Description, the 17.1 latest NRC-accepted version.
3B NEDE13426P l Test Mark SeriesIll5805.p Conformatory 975. Test Program -1/3 Scale impact 1.6-2 ? MaterialIncorporated by Reference - Amendment 34 T. R . to e T h r e. j e.f a I j
- 23A6100 Rsv. 4 sundanlsaletyAnsiysisa: port ABWR I 1
l 1
Table 1.6-1 Referenced Reports (Continued)
ABWR SSAR Report No. Title Section No.
NEDO 20206 D.R. Rogers, BWR Turbine Equipment N-16 Radiation Shielding 12.2 Studies, December 1973.
NEDO-20340 J. Carew, Process Computer Performance Evaluation Accuracy, 4.3 6e W.J.Bilanin, The GE Mark Ill Pressure Suppression 6.2 NEDO-20533 Containment Analytical Model, June 1974.
W.J.Bilanin, The GE Mark III Pressure Suppression 6.2 NEDO-20533-1 Containment Analytical Model, Supplement 1. September 1975 i
NEDE-205 General Electric Company Analytical Model for Loss-of-Coolant 6.3 Analysis in Accordance with 10CFR50, Appendix K September 1986.
NEDM-20609-01 P.P. Stancavage and D.G. Abbott, Liquid Discharge Doses L/DSR 12.2 Code, August 1976.
NEDO-20953A J.A. Woolley, Three-Dimensional BWR Core Simulator, January 4A.4 1977.
NEDO 21052 F.J. Moocy, Maximum Discharge Rate of Liquid-Vapor Mixtures 6.2 from Vessels, General Electric Company, September 1975.
H. Carews y, V. Nguyen, and P. Stancavege, Radiological 15.2 NEDO-21143-1 AccsJent The CONACO3 CODE, Decsmber 1981. 15.6 Airbome Releases from BWRs for Environmentalimpact 11.1 NEDO-21159 Evaluations, March 1976.-A!ce, NCDO-2115^ 2, .* .;9" 14. 12.2 NEDO-21159 hrM e.a W.s.r b ewtt.s4Ga BWRS Fuel Assembly Evaluatio w*
^ Liiusau care $hutdown 3.9 .
NEDE-21175-P Earthquake (SSE) and Loss -Coolant Accident (LOCA) L Loadings, November 1976. br**i- EvaLo.h ea r . Am 4 %T E Todme.
C 4.4 Brunswick Steam Electric Plant Unit 1 Safety Analysis Report NED)6-21215 for Plant Modifications to Eliminate SignilicJnt In-Core Vibrations, March 1976.
C 4.4 -
NED$-21251 J. Charnley, KKM Safety Analysis Report, April 1976.
l BWR Fuel Channel Mechanical Design and Deflection, 3.9 NEDE-21354-P September 1976.
L.. t.a skb eb b 3B NEDE-21471-D\ 6-kin, Analytical Model for Estimating Drag Force; on Rigid Submerged Structures Caused by '.CC' .;.2 L..'m M
- n. -s au m _. ,.... . A <n,,
NEDO-21471 F. . == , Analytical M el for Estimating Drag Forces on 3B Rigid Submerged Structu s Caused by a LOCA, D= -b- '"
1977. S'P
\ q ameke Aw MaterialIncorporated by Reference - Amendment 34 L% Dne' Ys a k Va<% osko h< < \ 9'19 - 1.6-3
\
. ,, o
- 23A6100 Rcv.4 ABWR sandsnisshrtyAnalysis neport 1
Table 1.6-1 Referenced Reports (Continued)
ABWR SSAR Report No. Title Section No.
NEDO-21506 Stability and Dynamic Forces of the GE Boiling Water Reactor 4.1 (LTR), yQr 1976.
NEDE-21514 -t t 2. BWR Scram System Reliability Analysis, December 1976, 190.6 General Electric Company.
Jamm NEDE-21526 J. Dougherty, SCAM - Subcompartment Analysis Method,6tme 6.2
- 1977. A.J.sen,t,d dj 1 e. E : J. Mark II Pressure Suppression Containment Systems:3B .
NED91549'P An Analhical Modelof the Pool Swell Phenomenon, Td.1077'ates \o< v \ 9TT
- NEDO-21778-A Transient Pressure Rises Affecting Fracture Toughness 5.3 107^:
Requirements forBoiling WaterReactors,J:nu;r"Lov Drew \9-(g NEDO-21985 Functional Capability Criteria for Essential Mark Il Piping, 3.9 September 1978, prepared by Battelle Columbus Laboratories for General Electric Company.
NEDE-22056 Failure Rate Data Manual for GE BWR Components, Rev. 2 19.3 January 17,1986, Class 111, General Electric Company. 190.3 19E.2 NEDO-22155 GE Report, Generation and Mitigation of Combustible Gas 6.2 Mixtures in Inerted BWR Mark l Containments, J une 1982.
NEDE-22277-P-l G. A. Watford, Compliance of the GE BWR Fuel Design to 20.3.7 l
Stability Licensing Criteria October 1984.
NEUE-23819 P.D. Knecht, BWM DryweIIand Containment Maintenance and 12.4 Testing Access 77me Estimates, May 1978.
NEDE-23996-1 P.D. Knecht, Maintenance Access Time Estimates, BWM 12.4 Auxiliary and Fuel Buildings, May 1979.
NEDE-23996-2 A. Chappori, Maintenances Access Time Estimates, BWM 12.4 r Radweste Building, May 1979.
l yspo 2Ao s 7 Assesse BW9/5 Eof A Pl* 8>N'*
' ' ' M*"
" \ 9"II' 3'8 BWR/iew i '
NEDO 24057-P Assessment of Reactor intemals Vibration in BWR/4 and BWM 3.9 NCOO 2'^5% Plants, November 1977. .^J= MECO 21^' 7 ", .^ .;c.d c.er: Or C::: d. 1 ^ 6, nu i4EDE 2 "20057," cc.;nd.c.;M 2, Jun; 1070.
NEDE-24131 Basis for 8x8 Retrofit Fuel Theranal Analysis Application, 40.2 September 1978.'
NEDO-24154 Qualincation of the One-Dimensional Core Transient Model for 4.4 BWRs,Vol.1 & 2 October 1978.
NEDO 24154-P Qualification of the One-Dimensional Core Transient Model for 4.4 BWRs, Vol. 3 October 1978. (Proprietary) ,
1.64 MaterialIncorporatedby Refe snce- Amendment 34
=
23A6100 Roct. 4 ABWR standardsarety Analysis aeport I
Table 1.6-1 Referenced Reports (Continued)
ABWR SSAR Report Na. Title Section No.
NEDE 24222 J. Weiss, Assessment of BWR Mitigation of ATWS,1979. 15E O '2 "
M. k. tr c.Jaiwei Prg n vg NEDE-24302-P W. M. DaiyGeneric Chugging Load Definition Report, April 3B 1981.
NEDE-24326-1-P General Electric Environmental Qualification Program, 3.9 Proprietary Document, January 1983. 3.11 NEDE 24351 D. Hale, Fatigue Crack Growth in Piping and RPV Steels in 3E Simulated BWR Water Environment Update,[1g NEDE-24ol9 Study of Advanced BWR Features, Plant Definitiorbasibility 12.4 Results, Vol.lil, Part G, October 1979.
NEDO-24708 P. W. Marriot. AdditionalInformation Required for NRC Staff 7.3 Generic Report on Boiling Water Reactors, qc c p :1979.
NEDE-25100-P -C. T. Ka;;;.tu, Caorso SRV Discharge Tes ase i Test Report, 3B 1979.
May\<.
n a ,- x cohom*A s+ pg . A sg Pro n ve, NEDE-25118 G6n. K 6ta, Caorso SRV Discharge Tes s g ' se II ATR, 3B
. . . ' - - 197 9.
g NEDO-25132A E. W.Yradley, Gamma & Beta Dose to Man from Noble Gas 12.2 Release to the Atmosphere GEMAN Code, April 1980.
NEDD-25153 er, Analytical Model for Estimatin g Forces] July 3B L.
1979,E. Lfon bcpd s W4ww C% SRae co 4 ws, hen o w d CM3 y "*p NEDE 25250 A. Javid, Generic Criteria for High Frequency Cutoff of BWR 3.9 Equipment, January 1980. (Proprietary)
NEDO-25257 E. W. Bradley and V. D. Nguyen, Radiation Exposure from 12.2 Airbome Effluents-the REFAE Code, July 1980.
NEDE-25273 F. T. Dodge, Scaling Study of the General Electric Pressure 3B Suppression Test Facility - MarkIll Long Range Program, Task 2.2. 7, SwRI, March 1980. (Proprietary)
NEDE-30090 Alto Lazio Station Reliability Analysis, December 984 19D.6 l
NEDO-30130-A Bill Zarbis, Steady-State Nuclear Methods, May 1985. 4.3 (Proprietary) 4.4 NEDC-30259 H.A. Careway, D.B. Townsend, B.W. Shaffer, A Technique for 15.6 i Evaluation of BWR MSIV Leakage Contribution to Radiological Dose Rate Calculations, September 4985r t 9 8 3, NEDE-30637 B.M. Gordon, Corrosion and Corrosion Controlin BWRs, 5.2 o e mVer \ 9 8 4 19P NEDE-30640 Evaluation of Proposed Modification to the GESSAR 11 Design, Class 111, June 1984.
NEDO 30832 J.E. Torbeck., Elimination of Limit on BWR Suppression Pool 3B l
Temperature for SRV Discharge With Quenchers,jc & l*9??:
1.6-5 MaterialIncorporated by Reference - Amendment 34
- 23A610cRsv.5 ABWR standard safetyAnalysis Report Table 1.6-1 Referenced Reports (Continued)
ABWR SSAR Report No. Title Section No.
NEDC-30851P-A W. P. Sullivan, Technical Specification improvement Analyses 19D.6 for BWR Reactor Protection System, 1988.
M.v cb NEDE-31096-A GE Licensing Topical Report ATWS Response to NRC ATWS 198.2 Rule 10CFR 50.62. February 1987.
NEDE-31152-P GE Bundle Designs, December 1988. 4.2 NEDO-31331 Gerry Burnette, BWR Owner's Group Emergency Procedure 18A Guidelines, g}87.
NEDC-31336 Julie Leong, GeneralElectric Instrument Setpoint Methodology, 7.3 t u 1986.
N:x:b ete ec NEDC-31393 J. utc ABWR Containment Horizontal Vent Confirmatory Test, 3B Part I, March 1987.
NEDO-31439 C. VonDamm, The Nucic~ Measurement Analysis & Control 20.3 Wide Range Neutron Monitoring System (NUMAC-WRNMS), Vsy 1987.
NEDC-31858P Louis Lee, BWROG Report for Increasing MSIV Leakage Rate 15.6 Limits and Elimination of Leakage Control System,1991 NEDE-31906-Pf A. Chung, Laguna Verde UnitIReactorIntemals Vibration 7.4 Measurement, F b=: ; 1991.
J awwag NEDO-31960 Gien Watford, BWR Owners' Group Long-Term Stability 4.4 Solutions Licensing Methodology, June 1991.
NEDC-32267P ABWR Project Application Engineering Organization and 17.1 Procedures Manual, December 1993.
4 l
I i
MaterialIncorporated by Reference - Amendment 35 1.6-6 l
- l 23A6100 Riv. 4 '
ABWR StandardSafety Analysis Report l
l l
l Table 1.8-21 Industrial Codes and Standards
- Applicable to ABWR (Continued) l Code or i Standard Number f Year Title j NOA-2f(a (990 Quality Assurance Requirements of Computer Software for Nuclear Facility Application OM3 1990 Requirements for preoperational and initial Startup Vibration Test Program for Water Cooled Power Plants OM7 1986 Requirements for Thermal Expansion Testing of Nuclear Plant Piping Systems [ September 1986 (Draft-Revision 7)]
American Petroleum Institute (API) 620' 1986 Rules for Design and Construction of Large, Welded, Low.
Pressure Storage Tanks 650 t 1980 Welded Steel Tanks for Oil Storage American Society of Heating, Refrigerating and Air-Conditioning Engineers,Inc. (ASHRAE) 30 1978 Methods of Testing Liquid Chilling Packages 33 1978 Methods of Testing Forced Circustion Air Cooling and Air Heating Coils American Society of Mechanical Engineers (ASME) l AG-1 t 1991 Code on Nuclear Air and Gas Treatment 830.2 t 1983 Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Grider, Top Running Trolley Hoist)
B30.9' 1984 Slings B30.10 t 1982 Hooks B30.11 t 1980 Monorails and Underhung Cranes B30.16' 1981 Overhead Hoists B31.1 t 1986 Power Piping 896.1' 1986 Specification for Welded Aluminum-Alloy Storage Tanks N45.4 1972 Leakage-Rate Testing of Containment Structures for Nuclear Reactors N509' 1989 Nuclear Power Plant Air-Cleaning Units and Components N510' 1989 Testing of Nuclear Air-Cleaning Systems N OA-1 ' 1983 Quality Assurance Program Requirements for Nuclear Facilities NOA-lh a 1983 Addenda to ANSI /ASME NOA-1-1983 I
l Conformance with Standard Review Plan and Applicability ei Codes and Standards - Amendment 34 1.8-39 l l
23A6100 R:v. 4 ABWR Standard SafetyAnalysisReport Table 1.8 21 Industrial Codes and Standards
- Applicable to ABWR (Continued)
Code or Standard Number Year Tttle OMa 1988 Operation and Maintenance of Nuclear Power Plants (Addenda to OM-1987)
Sec11 1989 BPVC Section 11, Material Specifications Sec ill 1989 BPVC Section Ill, Rules for Construction of Nuclear Power Plant Components Sec Vill 1989 BPVC Section Vlli, Rules for Construction of Pressure Vessel SecIX 1989 BPVC Section IX, Qualification Standard for Welding and Brazing Procedures Welder, Brazers and Welding and Brazing Operators Sec XI 1989 BPVC Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components C77C t9T9 siwt,4 veime Os o n 4 c. P4L Ms f.-
c934 1980 0** 7 J
- b T h,) ,A34iw .wa.a. PWe.h w American Society for Testing and Materials (ASTM)
E84 REV. A 1991 Methods of Test of Surface Burning Characteristics of Building Materials E119 1988 Standard Test Methods for Fire Tests of Building Construction and Materials E152 1981 Standard Methods of Fire Tests of Door Assemblies (See ASME BPVC Section ill for ASTM Material Specifications)
American Welding Society (AWS)
A4. 2 ' 1986 Procedures for Calibrating Magnetic Instruments to Measure the Delta Ferrite content of Anstenitic Stainless Steel Weld Metal D1.1 t 1986 Steel Structural Welding Code D14.1 ' 1985 Welding of Industrial and Mill Cranes and other Material Handling Equipment American Water Works Association ( AWWA)
D100' 1984 Welded Steel Tanks for Water Storage CMAA70 1983 Specification for Electric Overhead Traveling Cranes insulated Cable Engineer Association (ICEA)
P-46426/IEEE 1982 Ampacities including Effect of Shield Losses for Single S-135 Conductor Solid-Dielectric Power Cable 15 kV through 69 kV 1.8M Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 34
l 23A6100 Rcv. 4 sundard safety Analysis Report 1 ABWR 1
Table 1.8 21 Industrial Codes and Standards
- Applicable to ABWR (Continued)
Code or Standard Number Year Trtle P-54-440/ NEMA 1987 Ampacities of Cables in Open-Top Cable Trays WC-51 S-61-402/NEM A 1973 Thermoplastic insulated Wire & Cable for the Transmission and W C-5 Distribution of Electrical Energy S-66 524/NEM A 1982 Cross Linked Thermosetting Polyethylene Insulated Wire and WC-7 Cable for Transmission and Distributor of Electrical Energy Institute of Electrical and Electronics Eng:neers (IEEE)
Standard for Software Safety P_lans) {
1992 C37.01 t
1979 #pplication Guide for Power Circuit Breakers C37.04 t 1979 AC oower Circuit Breaker Rating Structure C37.09 t 1979 Test P.ocedure For Power Circuit Breakers C37.13' 1989 Low Voltage Power Circuit Breakers t 1987 Switchgear Assemblies and Metal-Enclosed Bus C37.20 C37.90.2 1987 Trial-Use Standard, Withstand Capability of Relay Systems to Radiated Electromagnetic interference form Transceivers C37.100 t 1992 Definitions for Power Switchgear Transformers C57.12 t 1987 General Requirements for Distribution, Power, and Regulating Transformers C57.12.11' 1980 Guide for Installation of Oil-immersed Transformers (10MVA &
Larger,69-287 kV Rating)
C57.12.80' 1978 Terminology for Power and Distribution Transformers C57.12.90' 1987 Test Code for Distribution, Power, and Regulating Transformers C62.41' 1991 Guide for Serge Voltage in Low Voltage AC Power Circuits C62.45' 1987 Guide on Surge Testing for Equipment Connected to Low-Voltage AC Power Curcuits C63.12' 1987 American National Standard for Electromagnetic Compatibility ,
Limits-Recommended Practice j l
1982 Application Criteria for Digital Computers in Safety Systems for Nuclear Facilities (to be replaced by the issued version of P 7-4.3.2, " Standard Criteria for Digital Computers Used in Safety Systems of Nuclear Power Generation Stations")
80' 1986 Guide for Safety in AC Substation Grounding 81' 1983 Guide for Measuring Earth Resistivity, Ground Impedance, and Earth Surface Potentials of a Ground System Confonnance with Standard Review Plan and Applicability of Codes and Standards - Amendment 34 1.8-41
23A6100 Rtv. 4 ABWR senadsidsafetyAnalysis Report Table 1.8-21 Industrial Codes and Standards
- Applicable te ABWn (Continued)
Code or Standard Number Year Title j 519' 1981 IEEE Standard Recommended Practices and Requirements for Harmonic Control in Electrical Power Systems, 603' 1980 IEEE Standard Criteria for Safety Systems for Nuclear Power l
Generating Stations 622 t 1987 Recommended Practice for the Design and Installation of Electric Heat Tracing Systems in Nuclear Power Generating Stations 622A' 1984 Recommended Practice for the Design and installation of Electric Pipe Heating Control and Alarm Systems in Nuclear Power Generating Stations
'730 19N Standard for Software Quality Assurance Plans l 741 t 1986 Standard Criteria for the Protection of Class 1E Power Systems and Equipment in Nuclear Power Generating Stations 765 i 1983 Standard for Prefered Power Supply for Nuclear Power Generating Stations 802.2' 1985 Standards for Local Area Networks: Logic Link Control 802.5' 1985 Token Ring Access Method and Physical Layer Specifications 828 i 1983 Standard for Software Configuration Management Plans 829 t 1983 Standard for Software Test Documentation 830' 1984 Standard for Software Requirements Specifications 845 t 1988 Guide to Evaluation of Man-Machine Performance in Nuclear Power Generating Station Control Rooms and Other Peripheries 944t 1986 Recommended Practice for the Application and Testing of Uninterruptable Power Supplies for Power Generating Station 946' 1985 Recommended Practice for the Design of Safety-Related DC l
Auyiliary Power Systems for Nuclear Power Generating Stations 1012 t 1986 Standard for Software Verification and Validation 1023' 1988 I?EE Guide to the Application of Human Factors Engineering to Systems, Equipment and Facilities of Nuclear Power Generating Stations 1033' 1985 Recommended Practice of Application of IEEE-828 to Nuclear -
Power Generation Stations 1 0 4 '2 1987 Guide to Software Configuration Management m g (orJ4) gt s4aw.La ,J.C - s ,p.su S avy N 3 Instrument Society of America (ISA)
S7.3' 1981 Quality Standard for instrument Air Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 34 -1.843
I e
- 23A6100111v. 4 ABWR standardsafetyAnalysis Report Table 1,8-21 Industrial Codes and Standards
- Applicable to ABWR (Continued)
Code or Standard Number Year Title ESD-TR-86-278 1986 Guidelines for Designing User Interf ace Software l
HDBK 761 A 1990 Human Engineering Guidelines for Management information Systems HDBK 763 1991 Hurpan Engineerina Procedures Guide, Ch. 5-7 & Appendix. A&B STD. 7.167 A oes /' ~ -
" o h 4 53 h S a b -* D e p w e- +
TDP 1 2-610 1990 Test Op,erating Procedure Part 1 l
U.S. Military (MIL)
(472D 1989 Human Engineering Design Criteria for Military Systems Jquipment and Facilities, Department of Defense F-51068 Latest Filter, Particulate High-Efficiency, Fire-Resistant Edition H 46855B 1979 Human Engineering Requirements for Military Systems,
-> Equipment and Facilities HDBK-759A 1981 Human Factors Engineering Design for Army Material STD-282 1956 Filter Units, Protective Clothing Gas-Mask Components and Related Products: Performance-Test Methods STD-461C 1987 Electromagnetic Emission and Susceptibility Requirements for the Control of Electromagnete interference STD-462 1967 Measurement of Electromagnetic Interference Characteristics STD 1472D 1989 Human Engineering Design Criteria for Military Systems, Equipment and Facilities STD 1478 1991 Task Performance Analysis l HD6e -?,5i L 4^ d 8 a l* *b' IM D* 8' $ ^ ; ** * \ A f fb"' '^# .I d'h** op s pt% > TwW ee- se.b a - S a s i c hT , - " *
- k er M Others Data Li k La 3 4,v- a w J Pk3ss cal L$er ERDA 76-21 1976 Testing of Ventilation Systems, Section 9 of Industrial Ventilation i Systems IEC 880 1986 Software for Computers in the Safety Systems of Nuclear Power I Stations IEC 964 1989 Design for Control Rooms of Nuclear Power Plants, Bureau
' Iso 7 498 %984 ' N Central de la Commission Electrotechnique internationale OSHA 1910.179 1990 Overhead and Gantry Cranes TEMA C 1978 Standards of Tubular Exchanger Manufacturers Association l UL 44 1983 Rubber-Insulated Wires and Cables l UL 489 1991 Molded Case Circuit Breakers and Circuit Breaker Enclosures 1.846 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 34
23A6100 R:v. 4 Standard Safety Analysis Report ABWR 1
1 A.2.7 Post Accident Sampling [lI.B.3] [ o,o s l 0,50 NRC Position A design and operational review of the reactor coolant and con nment atmosphere '
sampling line systems shall be performed to determ ne the car ability of personnel to obtain (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample under accident onditions nithout incurring a radiation exposure to any individual in excess of 0.0' and Q.187)Sv to the whole body l
or extremities, respectively. Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products. If the resiew indicates that personnel could not promptly and safely obtain the samples, additional design features or shielding should be provided to meet the criteria.
A design and operational review of the radiological spectrum analysis facilities shall be performed to determine the capability to quantify (in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) certain mdionuclides that are indicators of the degree of core damage. Such radionuclides are noble gases (which indicate cladding failure), iodines and cesiums (which indicate high fuel temperatures), and nonvolatile isotopes (which indicate fuel melting). The initial reactor coolant spectmm should correspond to a Regulatory Guide 1.3 or 1.4 release.
The review should also consider the effects of direct radiation from piping and components in the auxiliary building and possible contamination and direct radiation from airborne effluents. If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria.
In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions. Procedures shall be provided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term). Both analyses shall be capable of being completed promptly (i.e., the boron sample analysis within an hour and the chloride sample analysis within a shift).
Response
Discharges From Plant and Containment- During the development of an accident, samples ofliquid and gaseous discharges from both the plant and containment will be obtained. Chemical and radiochemical analyses will be performed for protection of the health and safety of the public and the plant operators. These samples will be obtained from the Process Sampling System. The Post Accident Sampling Systems will not be required to obtain these samples.
t Cor.' Damage Assessment-During this inidal period, instrumentation will provide -l sub ient informadon for the operators to perform their duties. For example, the Co.c Jnment high range radiation meters will give instant information concerning the radiation level in containment (To obtain data from the PASS several hours may be required for sampling and analyses.). Calculations can be performed to relate lA 7 Response to TMIRelated Matters- Arnendment 34
', 33A6100 RCV. 4 l stendedsareryAnstraiswat ABWR j
, 5.
main condenser tubing, the condensate polishing demineralizers and ,
the pumps and valves in the condensate /feedwater systems. These pumps are stopped and these valves closed during upset conditions.
Thus, because both factors are not present, the time to complete the analysis is increased to 4 days.)
O.05 0.50 (6) It must be possible to obtain and yze a sample without diation exposures l
to any individual exceedin 0m v for whole body and 750 m Sv for extremities. Meets the requirements of h"JREG-073~. 50. 4 (O UN U M-(7) Ability to sample and analyze for reactor coolant boron must be provided.
Meets the requirements of NUREG4737.
(8) Ifinline monitoring is used, backup sampling and analysis capability must be provided. Inline monitoring is not used. Meets the requirements of NUREG0737.
(9)
(a) Capability to identify and quantify a specified number ofisotopes over a range of nuclide concentrations from approximately 37,000 Bq/g to 370,000 M Bq/g. Capability is provided to identify and quantify the desired isotopes in samples over a range from approximately 37,000 Bq/g to 37,000 M Bq/g. Samples obtained during the accident recovery ,
phase would be within this range for most core damage accidenu. If the l
gross radioactivity levels are higher than 37,000 M Bq/g, this would confirm that severe core damage has occurred.
(b) Restrict background levels of radiation in the laboratory and provide proper ventilation. Meets the requirements of NUREG4737.
(10) Provide adequate accuracy, range and sensitivity to provide pertinent information. Meets the requirements of NUREG-0737.
(11)
(a) Provide sample lines with proper features for sampling during accident conditions. Meets the requirements of NUREG4737.
(b) PASS ventilation exhaust should be filtered with charcoal adsorbers and HEPA filters. Meets the requirements of NUREG.0737.
'1A 10 Response to TMI Related Matters - Amendment 34 4"
l
, 23A6100 R:v. 3 ABWR standard safety Analysis Report I
1 I
Strain energy weighted modal damping can also be used in the dynamic analysis. Stram energy weighting is used to obtain u.e modal damping coefficient due to the contribations of elements with different damping properties in the model. The element dampir.g values are specified in Table 3.7-1. Strain energy weighted modal damping is calculated as specified in Subsection 3.12.h5 3 7 2,/5' In direct integration analysis, damping is input in the form of cx and Q damping constants, which give the percentage of critical damping, A as a function of the circular frequency, co.
cx Sco 1=g+7 (3.7-27) 3.7.3.8.1.8 Effect of Differential Building Movements In most cases, subsystems are anchored and restrained to floors and walls of buildings that may have differential movements during a seismic event. The movements may range from insignificant differential displacements between rigid walls of a common building at low elevations to relatively large displacements between separate buildings at a high seismicity site.
Differential endpoint or restraint deflections cause forces and moments to be induced into the system. The stress thus produced is a secondary stress. It isjustifiable to place this stress, which results from restraint of free end displacement of the system, in the secondary stress category because the stresses are self-limiting and, when the stresses exceed yield strength, minor distortions or deformations within the system satisfy the condition which caused the stress to occur.
The earthquake thus produces a stresxxhibiting property much like a thermal expansion stress and a static analysis can be used to obtain actual stresses. The differential displacements are obtained from the dynamic analysis of the building. The displacemenu are applied to the anchors and restraints corresponding to the maximum differential displacements which could occur. The static analysis is made three times:
once for one of the horizontal differential displacements, once for the other horizontal differential displacement, and once for the vertical.
The inertia (primary) and displacement (secondary) loads are dynamic in nature and their peak values are not expected to occur at the same time. Hence, the combination of the peak values ofinertia load and anchor displacement load is quite consenstive. In addition, anchor movement effects are computed from static analyses in which the displacements are applied to produce the most conservative loads on the components.
Therefore, the primary and secondary loads are combined by the SRSS method.
3737 Seismic Design - Amendment 33
f !
l 23A3100 Rev. S ABWR Standard Safety Analysis Repon 3.7.6 Referencesgj 3.7-1 [ Mral Electnc Company BWR/6-238 Standard Safety Analysis Repon (CESSAR), Docket No. STN 50-447, >'~=nbs 7,1975, J h 3 ojL 9 7 3.
OMJ y) 3.7-2 h E. H. Vanmarcke and C. A. Cornell, Seismic Risk and Design Response Spectra, ASCE Specialty Conference on Safety and Reliability of Metal Stmctures, Pittsburgh, Pennsyhania, November 1972. j 3.7-3 p N .G-0800, Standard Review Plan, Section 3.7.1.
3.7-4 L K. Liu, Seismic Analysis of the Boiling WaterReactor, symposium on seismic analysis of pressure vessel and piping components, First National. Congress on .
Pressure Vessel and Piping, San Francisco, California, May 1971.
3.7-5 EPRI NP-5930, A CnterionforDetermining Exceedance of the Operating Basis Eanhquake, July 1988.
3.7-6 EPRI TR-100082, Standardization of Cumulative Absolute Velocity, December 1991.
3.7-7 EPR1 NP-6695, Guidelinesfor NucitarPlant Response to an Eanhquake, December 1989.
3.7-8 P. Koss, Seismic Testing ofElectn' cal Cable Support Systenu, Structural Engineers of l
California Conference, San Diego, September 1979.
3 l'AI Seismic Design - Amendment 35 l
, . - - ~ = _ ..
23A6100 Rsv. 4 standard safety Analysis Report A,BWR Table 3.8-2 Major Allowable Stresses in Concrete and Reinforcing Steel Concrete Reinforcing Steel Compression Tangential Shear Tension 16.54 MPa (1) Provided by concrete 206.8 MPa l Service Load Combination v=0 c
(2) Provided by orthogonal 310.3 MPa (For test 1
reinforcement pressure case) l v o = 1.2f =e 1.96 MPa 23.44 MPa
? 372.4 Mpa l Factored Load (1) Provided by&J~wunm M8
.e;.t::mr-M come v-Combination v=0 c
(2) Provided by orthogonal .
reinforcement l
v o = 2.48e = 3.92 MPa Table 3.8-3 Stressintensity Limits Primary Stresses Primary &
Bending & Secondary Local Mem. Local Mem. Stresses Gen. Mem Pt P+Pt P+P+0 t
Pm 0.75 Sy 1.15 Sy 1.15 Sy N/A-Test Condition 1.5 S m 1.5Sm N/A Design Condition 1.0 S m
- The larger of The larger of The larger of 3Sm' Post-LOCA 1.8 Smc or 1.5 Sy Flooding 1.2 Smc or 1.0 Sy 1.8 Smc or 1.5 Sy
- The allowable stress intensity Sm is the Sm listed in Table 110.0 and Sy is the yield strength listed in Table 12.0 of Appendix I of ASME Code Section Ill.
3 8*51 Seismic Category I Structures - Amendment 34 .
, 23A6100 Rov. 2 ABWR standud Safety Analysis Reput The site-envelope maximum relative displacements of the nodal points wit.h respect to the base of each respective stick model are shown in Tables 3A-27a througi 3A-27d and 3A-28. These results may be used for design ofcomponents located within the respective building.
3A.10.5 Summary i The site-envelope maximum seismic responses presented in Subsection 3A.10 envelop the maximum seismic SSE responses of the ABWR plant structures and components for a wide range of subsurface properties and conditions as well as the effect of concrete cracking and side soil-wall separation. These responses are used to design the ABWR plant structures and components.
3A.11 References A p p d t v. "3 Aj 3A-1 [ General Electric Company BWR/6-238 Standard Safety Analysis Report (GESSAR) Docket No. STN 50-447, S'ert::- 7,1075. b \ j 10 si 3 ~' 3 -
Appdy 3 A 3 3A-2 [ General Elecuic Company GESSAR II BWR/6 Nuclear Island Design (22A7007)' March 1980.
SA NUREG/CR-1161. Recommended Revisions to Nuclear Regulatory Commission Seismic Design Criteria, May 1980.
3A-4 Lysmer,J., Tabatabaie-Raissi, M., Tajirian, F., Vahdani, S., and Ostadan, F.,
SASSI-A System for Analysis of Soil-Structure Interaction, Report No.
UC/B/GT/81-02 Geotechnical Engineering, University of California, Berkeley, CA, April,1981; also Ostadan, F., Computer Program SASSI, CE '
(944), Theoretical, User's and Validation Manuals (1991), Bechtel Corporation, San Francisco, California SA-5 Schnabel, P.B., Lysmer,J., and Seed, H.B., SHAKE-A Computer Program for Earthquake Response Analysis of Horizontally Layered Sites, Report No. RC 72-12, Earthquake Engineering Research Center, University of California, ,
Berkeley, CA,1972.
-3A-6 Seed, H.B. and Idriss, I.M-Soil Moduli and Damping Factors for Dynamic j Response Analysis, Report No. RC 70-10, Earthquake Engineering Research l Center, University of California, Berkeley, CA,1970. )
l SA-7 Idriss. I.M-Response of Soft Soil Sites During Earthquakes. H. Bolton Seed Memorial Symposium Proceedings. Volume 2, Bi Tech Publishers,' May 1990, !
1
- 3A ;6 Seismic Soil Structure Interaction Analysis - Amenament 32 '>
~
2sAs a nov.2 andadsanyAnnorsisnePar
. ABWR 0* '
TEMPERATURE SENSORS 331 (REFERTOTABLE BELOW)
I SRV QUENCHERS -
v TE
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g g /18* X 81*
9
@ FOR ELEVATION SEE FIGURE 7.6-10 286*
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k' f - X l 27C/' - -
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DIVISION I DIVISION 11 OlVISION lli DIVISION IV 16*_ TE 001 A,E,J,N TE@10.G.LR 61* TE.an9R.F.K.p E GG2). .M.S 106* TE tnt.F.K.* 'E 005).- .M.S q
151' TE-004A.E.J.h E-004C.ta_
196* TE 005A.E.J.h E ^^5.G. - 4 241' "E.nnam,s.K.P "E4063.EM.S 286* "E-007B.'".K.P 'Y PJ73.E.M.S 331' TE-008A.E.J.N TE onat'/3 i RI s
NOTE: DIVISIONS 1,11, lil AND IV TEMPERATURE SENSORS AT EACH LOCATION SHALL BE SEPARATED BY 15 - 30 CM.'
Figure 7,6-9 Instrumentation Location Definition for the Suppression Pool Temperature Monitoring System -
All Other instrumentation Systems Required for Safety- Amendment 32 7.M7
. . . 23A6100 Rev. 3 ABWR studerds:teryAcrysis depon The pressure control function provides ABWR automatic load following by forcing the turbine control valves to remain under pressure control supervision, while enabling fast bypass opening for transient events requirirg fast reduction in turbine steam flow.
The steam bypass function controls reactor pressure by modulating three automatically operated, regulating bypass valves in response to the bypass flow demand signal. This control mode is assumed under the following conditions:
(a) During reactor vessel heat-up to rated pressure.
(b) While the turbine is brought up to speed and synchronized. I (c) During power operation when reactor steam generation exceeds the turbine steam flow requirements. l (d) During plant load rejections and turbine-generator trips.
(c) During cooldown of the nuclear boiler.
(7) 1&C Interface The external signal interfaces for the SB&PC System are as follows:
(a) Narrow range dome pressure signals from the SB&PC System to the Recirculation Flow Control System, g (b) Equivalent load or steam flow feedback gnal from the Turbine Control System (whichYr.:!+d A meftriplicated fault-tolerant digital control : S" " "C) .
(c) Signals to and from the main control room.
Bypass hydraulic power supply trouble signal from the Turbine Bypass (d)_ System to the SB&PC System.
(e) Output signals from the SB&PC System to the performance monitoring and control function of the process computer.
(f) Displayed variables and alarms from the SB&PC System to the main control room panel operator interface.
(g) Narrow and wide range pressure signals, MSIV position signals from the Nuclear Boiler System to the SB&PC System.
(h) Bypass valve position, servo current, position error and valve open and ~
closed signals from the Turbine Bypass System.
(1).- Emergency bypass valve fast opening signals and bypass valve flow demand signals from the SB&PC System to the Turbine Bypass System.
Control Systems Not Required for Safety - Amendment 33 7.7-69
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3 8 E REACTOR POWERCONTROL FUNCIIONS H ANY si Af us .
.3 -Ts viA CouPUI R DAI A APPRGALJG HE ATUP C taQOg RF ACTOR $4E ACYOR REACTOR ACoutS81*084 8 10 RANGE " POM R fiHu'f DOUW4 DE PRE S$upu2ATIOso e g y 6 gipsCl uOeeeG CMiiCatBIY DOWN MuOE A CCOLDowuN e Q EQUIPteENT l MODE uoog AUTOsi#Al ec STATUS LOAD FOLLOW MM M , g
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a AUTOMATC CORE SPtI(1 Rf C3 Mt IDE e POWE ft AF C Flow N WA880 (f e .AuYOhsAf tc 8e PRE &suf4E 9 REOut ATOR CE taANO cantE et Ow Core f e4GL
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- CONTROL was VF V As WF M
ACIUAIORS ACTUAIORS Y
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. _ . _ _ . _ . _ _ . Figure 73-11 Simplified Function; I Diagram of the Automatic Power Regulation System j R.O O C ON-T R O t #
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23A6100 R2v. 4 ABWR standard san Autysis aepois l
l l
(2) Prevent the uncontrolled loss of contaminated pool water to other relatively cleaner locations within the containment or fuel-handling area j l
Provide liner leak detection and measurement l (3)
These drainage paths are designed to permit free gravity drainage to the equipment l
drain tanks or sumps of sufficient capacity and/or pumped to the Radwaste Building.
A makeup water system and pool water level instrumentation are provided to replace evaporative and leakage losses. Makeup water during normal operation will be supplied from condensate. The Suppression Pool Cleanup (SPCU) System can also be used as a Seismic Category I source of makeup water in case of failure of the normal Makeup Water System.
Both FPC and SPCU Systems are Seismic Category I, Quality Group C design with the l
exception of the filter-demineralizer portion, which is shared by both systems.
Following an accident or seismic event, the filter-demineralizers are isolated from the FPC cooling portion and the SPCU System by two block valves in series at both the inlet and outlet of the common filter-deminemlizer portion. Seismic Category 1. Quality l
Group C bypass lines are provided on both FPC and SPCU Systems to allow continued flow of cooling and makeup water to the spent-fuel pool.
Connections from the RHR System to the FPC System provide a Seismic Category 1, safety-related makeup capability to the spent-fuel pool. The FPC System from the RHR connections to the spent-fuel pool are Seismic Category I, safety-related.
- N ', E ftT A Furthermore, fire hoses can be used as an alternate makeup source.The fire protection standpipes in the Reactor Building and their water supply (yard main, one diesel engine driven pump and water source) are seismically designed. A second fire pump, driven by a motor powered from the combustion turbine generator,is also provided. Engineering analysis indicates that, under the maximum abnormal heat load with the pool gates closed and no pool cooling taking place, the pool temperature will reach about 100'C in about 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. This provides sufficient time for the operator to hook up fire hoses for pool makeup. The COL applicant will develop detailed procedures and operator training for providing firewater makeup to the spent-fuel pool. See Subsection 9.1.6.9 for COL license information.
The FPC components, housed in the Seismic Category i Reactor Building, are Seismic Category 1, Quality Group C, including all components except the filter-demineralizer.
These components are protected from the effects of natual phenomena, such as:
earthquake, extemal flooding, wind, tornado and external missiles. The FPC System is non-safety-related with the exception of the RHRgystem connections for safetv-related makeup and supplemental cooling. The RHR gystem connections will be protected from the effects of pipe whip, internal flooding, internally generated missiles, and the 9.1 15 fuel Storage and Handling - Amendment 34
i
. . .. L I N S g g.T A fne m o.n u o. 5 v a \ve s w bsek pe<%d 4he l ft H R s b 4a 4a6 su c+ son %om +Le 1 sgowf hsml s 4eme. p oo\ %el c.oel 46 p o e l as a c cas s t b\ e. -97 laws o. w a c.c s ciw -
m s JAcow+ 4 % 4a ~ d am e nh4 c <
bAm +k R H R syk 40 p sv%4 4k-sem43 A d s4crx p geel 4 ew b os h w3 f
l 1
aw - - - , , . - - - - - , - - - -+
l Ob
? Table 9.2-4a Reactor Building Cooling Water Division A b tXI P: Emergency Normal (LOCA)
Operating Shutdown at 4 Shutdown at Hot Standby Hot Sta dby (Suppression Operating Mode / Components Conditions Hours 20 Hours (No Loss of AC) (Loss o AC) Pool at 97'C Heat Flow
- Heat Flow Heat Flow Heat Flow Heat low Heat Flow Essential Emergency Diesel Generator A - - - - - - - -
13.40 229 13.40 229 l
RHR Heat Exchanger A - - 108.02 1,199 34.75 1,199 - -
1,199 89.18 1.199 l
l Others (essential)' 3.18 205 3.60 205 3.81 205 3.39 205 4.10 205 4.19 205 Non Essential CUW Heat Exchanger 8 20.10 159 159 159 20.10 159 20.93 159 - -
j - -
FPC Heat Exchanger A f 7.12 279 7.12 279 7.12 279 7.12 279 7.12 279 9.63 279 l
Inside Drywell'* 5.86 320 5.86 320 5.86 320 5.86 320 3.39 320 - -
3 l
Others (non-essential)" 2.64 160 2.64 160 2.64 160 2.64 160 0.84 59 0.75 59 $
l 38.94 1,123 127.28 2,322 54.01 2,322 38.94 1,123 75.36 2,450 117.23 1,971 l Total Load s
A
- Heat x GJ/h; flow x m /h, 3 sums may not be equal due to roundino.
l t HECW refrigerator, CAMS coolers, room coolers (RHR, RCIC, CAMS), RHR motor and seal coolers.
- The heat transferred from the CUW heat exchanger at the start of cooldown is appreciable, but during the criticallast part of a cooidown, the heat removed is very little because the temperature difference between the reactor water and the RCW System is small. Sometimes, the operators may remove the CUW heat exchangers from service during cooldown. Thus, the heat removed varies from about that during normal operation at the start of cooldown to very little at the end of cooldown.
p 3 f includes FPC rcom cooler.
" Drywell (A & C) and RIP coolers. R 5
t t Instruments and service air coolers; CUW pump cooler, CRD pump oil, and RIP MG sets. A hot water exchanger is in this division which removes E
{ heat from the RCW System. &
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23A6100 Rw. 4 Stand & Selety Analysis R: port ABWR 6
Main purge flow and sample flow are in closed lines and are routed through closed drains to the reactor building equipment drain sump.
The Post-Accident Sampling System (PASS) consists of a sample holding rack, sampling rack, sample conditioning rack, local control panel and shielding casks. Samples from the sample conditioning rack, discussed above, are sent to the PASS sample holding rack. A portion of the sample flowis passed through an inline sample vessel. After adequate purging, the sample vessel is isolated and transported to the laboratory. All valves in this operation are operated remotely. The sampling system isolation valves are operated from the main control room and all other valves are operated from the local control panel. After the sample vessel has been isolated and removed, the piping is flushed with demineralized water. The water from purging and flushing is drained to the suppression pool.
The sample holding rack has an enclosure around the sample vessel to contain any leaks ofliquids or gases. The liquids drain to the radwaste system and the gases go to the reactor building exhaust system.
The PASS isolation valves shall be connected to a reliable source of power that will be available starting at least one hour after a LOCA or ATWS event. The isolation valves shall have Class 1E power and the panels and other equipment shall be powered with two offsite power supplies and one onsite power supply.
l Gas samples are obtained from a sample line connected to the Containment Atmospheric Monitoring System (CAMS). A vacuum pump is provided to transfer the gas sample from a sample holding rack to a sampling rack. The sample is mixed uniforTnly. In the sampling rack, the gas is passed through and collected in a gas sample holder. After isolation, the gas sample holder is removed and transported to the laboratory for analysis.
The upper limits for activity levels in liquid and ' gas samples are:
Liquid samples 3.70E+10 Bq/cm 3 l
Gas samples 3.70E+09 Bq/cm 3 l
Means to reduce radiation exposure are provided such as, shielding, remotely operated valves, and sample transporting casks. The radiation exposure to any individual shall l not be in excess o E-0 ~ d(18.75E-OpSv to the whole body or extremities, respectively. g (..5'O Acceptance Criterion II.K.5 of SRP Section 9.3.2 requires the capability of sampling 3
liquids of 37.0E+10 Bq/cm . The ABWR design has the capability of sampling liquids of 3.70E+10 Bq/cm3Sampling will be performed and area radiation measurementwill be Process Auxiliaries - Amendment 34 9.34
23A6100 Rev. 5 ABWR Standant SafetyAnalysis Report i
.. I O,50 (- l 19A.2.20 Post Acciden ampling [ item (2) (viii)] .
NRC Posit' n Provide a pability to promptly obtain and analyze samples from the reactor coolant system d containment that may contain TID 14844 source term radioactive materials with t radiation exposures to any individual exceeding 0.05 Sv to the whole-body or 0.75 Sv to the extremities. Materials to be analyzed and quantified include certain radionuclides that are indicators of the degree of core damage (e.g., noble gases, iodines and cesiums, and non-volatile isotopes), hydrogen in the containment atmosphere, dissolved gases, chloride, and boron concentrations. [II.B.3]
Response ~
This item is addressed in Subsection lA.2.7.
19A.2.21 Hydrogen Control System Preliminary Design [ Item (2) (ix)]
NRC Position Provide a system for hydrogen control that can safely accommodate hydrogen generated by the equivalent of a 100% fuel-clad metal-water reaction. Preliminary design information on the tentatively preferred system option of those being evaluated ,
in paragraph (1) (xii) of 10 CFR 50.34(f) is sufficient at the construction permit stage.
The hydrogen control system and associated systems shall provide, with reasonable ,
assurance, that: [II.B.8]
(1) Uniformly distributed hydrogen concentrations in the containment do not exceed 10% during and following an accident that releases an equivalent amount of hydrogen as would be generated from a 100% fuel clad metal-water reaction, or that the post-accident atmosphere will not support hydrogen combustion.
(2) Combustible concentntions of hydrogen will not collect in areas where unintended combustion or detonation could cause loss of containment integrity or loss of appropriate mitigating features.
(3) Equipment necessary for achieving and maintaining safe shutdown of the plant and maintaining containment integrity will perform its safety function during and after being exposed to the environmental conditions attendant with the release of hydrogen generated by the equivalent of a 100% fue1< lad metal water reaction including the environmental conditions created by activation of the hydrogen control system.
(4) If the method chosen for hydrogen controlis a post-accident inerting system, inadvertent actuation of the system can be safely accommodated during plant operation.
i 19A4 Response to CP/ML Rule 10 CFR 50.3Mf)- Amendment 35 l
23A6100 R2v. 5 ABWR standard SafetyAnalysis Report Safety issues index (Continued)
NRC SSAR fitle Priority . Subsection I.C.8 Pilot Monitoring of Selected Emergency Procedures for Resolved COL App.
Near-Term Operating License Applicants I.D.1 Control Room Design Reviews Resolved 1 A.2.2 1.D.2 Plant Safety Parameter Display Console Resolved 1 A.2.3 >
l.D.3 Safety System Status Monitoring Medium 19A.2.17 1.D.5(2) Plant Status and Post-Accident Monitoring Resolved 198.2.65 l.D.5(3) On-Lin Nctor Surveillance System Near Res. 19B.2.66 1.F.2(2) Include QA Personnel in Review and Approval of Plant Resolved 19A.2.43 Procedures I.F.2(3) Include QA Personnel in All Design, Construction,installa- Resolved 19A.2.43 tion, Testing, and Operation Activities 1.F.2(6) Increase the Size of Licensees' QA Staff Resolved 19A.2.43 1.F.2(9) Clarify Organizational Reporting Levels for the QA Organi- Resolved 19A.2.43 zation 1.G.1 Training Requirements Resolved 1 A.2.4
- 1.G.2 Scope of Test Program Resolved 19B.2.67 ,
ll.B.1 Reactor Coolant System Vents Resolved 1 A.2.5 COL App.
II.B.2 Plant Shielding to Provide Access to Vital Areas and Protect Resolved 1 A.2.6 Safety Equipment for Post-Accident Operation ll.B.3 Post-Accident Sampling Resolved 1 A.2.7 11.B.4 Training for Mitigating Core Damage Resolved COL App.
11.B.8 Rulemaking Proceeding on Degraded Core Accidents Resolved 19A.2.21 19 A . 7. t 8 9 A . 2.2.
II.D.1 Testing Requirements Resolved 1 A.2.9 g,2,4 l Resolved t 9 A. 2.1 ! ,
lI.D.3 Relief and Safety Valve Position indication 1 A.2.10 ll.E.4.1 Dedicated Penetrations Resolved 1 A.2.13 II.t:.4.2 Isolation Dependability Resolved 1 A.2.14 f it.E.4.4 Purging Resolved 19A.2.27 II.E.6.1 Test Adequacy Study Resolved 19B.2.68 COL App.
II.F.1 Additic." Wident Monitoring instrumentation Recolved 1 A.2.15 -
II.F.2 Identification of and Recovery from Conditions Leading to Resolved 1 A.2.16 <
inadequate Core Cooling ,
ll.F.3 Instruments for Monitoring Accident Cor.ditions Resolved 1 A.2.17.
II.J.4.1 Revise Deficiency Reporting Requirements Resolved COL App.
1 Resolution of Applicable Unresolved SafetyIssues and Generic Safetyissues - Amendment 35 198-S .
23A6100 Ret 4 Standard Safety Analysis Repstt ABWR performance dunng an accident. The purpose of this issue is to improve the a _ curacy of measurement of airborne iodine concentrations.
Acceptance Criteria Airborne iodine concentrations must be accurately determined throughout tt e plant under accident conditions.
Resolution Item III.D.3.3(1) which concerns in-plant radiation monitoring is resolved in Subsection 12.3.4 which also references each area detector location on the plant layout drawings for each building (Figures 12.3-56 through 12.3-73) as well as the specific area radiation channels for each building, the detcctor map location, the channel sensitivity range, and the local alarm areas (Tables la.3-3 through 12.3-7).
References 198.2.72-1 NUREG-0660, NRC Action Plan Developed as a Result of the TMI-2 Accident, U.S. NRC, May {.rty.
\980 198.2.72-2 NUREG-0737, Clanfication of Thil Action Plan Requirements, U.S. NRC, November 1980.
19B.2.73 Ill.D.3.3(2): Set Criteria Requiring Licensees to Evaluate Need for Additional Survey Equipment Issue NUREG-0660 (Reference 19B.2.73-1) is a guideline to improve nuclear power plant worker radiation prcection to allow workers to take effective action to control the course and consequences of an accident, as well as to keep exposures as low as reasonably achievable (ALARA) during normal operation and accidents.
Acceptance Criteria This issue required the NRR to set criteria requiring licensees to evaluate in their plants the need for additional survey equipment and radiation monitors in vital areas and requiring, as necessary, installation of area monitors with remote readout. The NRR evaluated the need to specify the minimum types and quantities of portable monitoring instrumentation, including very high dose rate survey instruments. Operating reactors were reviewed for conformance with Standard Review Plan (SRP) Section 12.3.4, Area Radiation and Airborne Radioactivity MonitoringInstrumentation. The NRR revised the SRP Sections 12.5 and 12.3.4 to incorporate additional monitor requirement criteria.
Resolution Item III.D.3.3(2) which concerns licensees evaluate the need for additional radiation l suney equipment is resolved in Subsection 12.3.4. This item also concerned the need l l
to specify the rninimum types and quantities of portable monitoring instrumentation, l l
including very high dose rate survey instruments. As noted in Subsections 12.5.2, i
198 123 Resolutson of Applicable Unresolved Safety Issues and Generic SafetyIssues - Amendment 34
~. _. .
l a
- 23A6100 Rev. 5 standardsafery Analysis nevon ABWR 19A.2.39 and 19A.S.5, COL applicants will provide the portable instruments in operating reactors that accurately measure radio-iodine concentration in plant areas under accident cond.idons.
I -
References 19B.2.73-1 NUREG-0660, NRC A ion Plan Developed as a Result of the TbH-2 Accident, U.S. NRC, May 19B.2.73-2 NUREG-0737, Clarification of ThU Action Plan Requirements, U.S. NRC, November 1980.
198.3 COL License Information 198.3.1 COL Applicant Safety issues ,
The COL applicant shall provide resolutions for the issues identified as COL applicant in the Safety Issues Index consistant with the documentation format discussed in Subsection 19B.1.1. ;
198.3.2 Testing of isolators ,
As established in Section 7A.3, the COL applicantis required to establish a test program '
for fiber optic-type isolators used between safety-related and non-safety-related systems.
If other types ofisolators are used (those subject to electricalleakage due to maximum credible electrical faults), the COL applicant shall implement the required testing, inspection, and replacement isolators when needed (See Subsection 19B.2.53).
199 1 4 Resolution of Applicable Unreso!ved Safety Issues and Generic Safetyissues - Amendment 35
[.
. e 23A6100 Rev. 5 ABWR suaantsareryAnarrsis napart.
. ('.
Table 19E.2-1 Potential Suppression Pool Bypass Lines (Continued)
Pathway Basis For Exclusion Number Size (mm) isolation -(See Description of Lines From To (1 in. = 25.4 mm) Valves Notes)
SPCU Suction 1 SP RB 200 MO, MO 2 SPCU Return 1 SP RB 250 MO, CK 2 Cont. Atmosphere Monitor 6 DW RB 20 h -MO -
LDS Samples 2 DW RB 30 (SO, SO) -
Drywell Sump Drains 2 DW RB 100 MO, MO -
HVCW/RBCW Supply 4 DW RB 125 CK, MO 1 HVCW/DWCW Return 4 DW RB 125 MO, MO 1 DW Exhaust /SGTS 1 DW RB 550 AO,AO 7 Wetwall Vent to SGTS 1 WW RB 550 AO, AO 2 DW Purge 1 DW RB 350 AO -
Inerting Makeup 1 DW WW 50 AO,AO -
WW Inerting/ Purge 1 WW RB 550 AO, AO 2 g Instrument Air (and Service 2 DW RB 50 CK, MO 1 Air)
SRV Pneumatic Supply 3 DW RB 50 CK, MO 1 Flammability Control 2 DW RB 100 (AO, MO) 3-ADS /SRV Discharge 8 RPV WW 300 RV -
ACS Supply 2 DW WW 550 AO, AO -
WW/DW Vacuum Breaker 8 DW WW 500 CK -
Miscellaneous Leakage 1 DW RB -
NONE 6 Access Tunnels 2 DW RB -
NONE
~
l ' LUW Keturn as via1-eedwater NOTES:
Legends and Acronyms I
Pathway Source (From) Termination (To) l TIPV Reactor Pressure Vessel WW Wetwell DW Drywell RB Reactor Building 1 SP Suppression Pool WW Wetwe!!
ST. Steam Tannel 19E.2160 . Deterministic Analysis of Plant Perforrnance - Amendment 35 i
~
23A6100 R1v. 4 stuurd safety Aulysis Report ABWR Table 19E.2-21 Summary of Bypass Probabilities Figures 19E.219a Flow Split Bypass Probability Bypass Bypass to Pathway Fraction Equation Probability Fraction 19E.2-19k Lines from the RPV 6.7E-1 4*P1*(P3* P4+P5) 1.6E-6 1.1E-6 A Main Steam Main Steam Leakage 2.2E-5 4*P2*(P3*P4+PS) 1.1E-2 2.5E-7 A :
5.2E-1 2*P9'P9'P15 2.4E-8 1.3E-8 8 Feedwater 3.1E-5 30*P13*P9 6.0E-5 1.9E-9 D Reactor Inst. Lines HPCF Discharge 1.1E-1 2*P9'P10'P14 1.3E-7 1.5E-8* C HPCF Equalizing Line 1.0E-3 2*P10*P11*P13 6.7E-8 6.7E-11
- C 3.0E-3 1*P9'P13 3.6E-7 1.1 E-9 8 SLC injection ,
RCIC Steam Supply 6.9E-2 1*P8'P14 5.2E-9 3.6E-10 E LPFL Discharge 1.7E-1 2'P9'P10*P15 6.7E-4 1.1 E-8* C LPFL Equalizing Line 1.0E-3 2*P10*P11*P13 6.7E-8 6.7E-11' C 1.2E-1 1*P8'P14 5.2E-9 6.2E-10 E CUW Suction 3.1E-5 4*P13*P9 8.1E-6 2.5E-10 D CUW Inst Lines Post Ace Sampling 1.0E-3 4*P8'P11 3.6E-7 3.7E-10 J 3.1E-5 9'P13'P9 1.8E-5 5.7E-10 D LDS instruments SRV Discharge 6.9E-2 8'P14 1.3E-4 8.8E-6 K Lines from the Drywell P8 3,g ,,g 2,33,g g Cont Atmos Monitor 8.9E-4 6 P13 2E -14E-9 #
1.7E-3 2*P8'P11 1.6E-7 2.6E-10 J LDS Samples 3.0E-2 2*P8'P13 1.6E 7 4.7E-9 J Drywell Sump Drain DW Purge 5.4E-1 1*P6*P11 1.1E-6 6.2E 7 1 Inerting Makeup 1.2E-2 1*P6 7.4E-4 8.9E-6 l l
ACS Supply 7.5E-1 2*P12*P6 1.5E-6 1.1E-6 H j 2.6E-1 8'P9 6.7E-2 1.7E-2 G WW-DW Vac Bkri Grand Total excluding vacuum breaker 2.1E-5 Goal 8.4E-4
- These lines may be excluded for station blackout events.
t Addressed on Containment Event Trees.
1 19E.2-175 Deterministic Analysis of Plant Performance - Amendment 34
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FIGURE ' 7.7-7 RECIRCULATION FLOW CONTROL SYSTEM IBD (Sheet 2 of 9: j tmendment JS ABWR SSAR' 23A6100 Rev 5 21-440