ML20043A551
| ML20043A551 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 05/02/1990 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20043A548 | List: |
| References | |
| NUDOCS 9005220232 | |
| Download: ML20043A551 (136) | |
Text
1 lr (c-ll .
o m/ -l GE Nuclear Entgy ;
LICENSING AND CONSULTING SERVICES Gene:a! lie.mc Company l
175 Curtnet kenue. Sen kse. CA 95125 '
. -i t
DATE: May 11,1990 j MEMO: EEN 9019P TO: Distribution /
FROM: E. Earl Nichols ABWR SSAR Lice sing Program Manager (408) 925-4288
SUBJECT:
ABWR SSAR AMENDMENT 11 TRANSMI' ITAL CORRECTIONS 4 ,
Reference:
Submittal of Amendment 11, Non Proprietary Information, to GE's ABWR SSAR, MFN No. 037 90, dated May 23,1990.
Submittal of Amendment 11, Proprietary Information, to GE's ABWR SSAR, MFN No. 038 90, dated May 2,1990. -
Attached is a corrected (correcting the date of issue from May 23 to May 2) copy of the subject non propreiteary transmittalletter. Please replace your copy of this letter with the attached copy.
In addition, there were a couple of printing errors that occurred in some of the packages sent out. Please make the following changes to the pages in the Amendment 11 package so that you will be assured of having the correct pages in your copy of the ABWR SSAR.
Renlacement Pull all Chapt er 3 pages from Amendment 11 package and replace with attached Chapter 3 pages.
Pull all Chapter 19B & 190 pages from Amendment 11 package and replace with attached Chapter 19B & 190 pages.
- Pull Chapter 20 pages, from 20.3-1.2 through 20.3131, and replace with attached Chapter 20 pages, from 20.31.2 through 20.3 131. 1
- Pull Chapter 20 page 20.3 288.1 and replace it with 20.3 290.1 (Proprietary Page).
Additions Insert Chapter 11 pages: 11.5 ii,iii 11.5 1,2 11.5-7,8 11.5 9,10 11.5 11,12 Insert Chapter 20 page: 20.3 141
]
CC: R. Mitchell, J.N. Fox i
9005220232 900502 PDR ADOCK 03000603 g PDC
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~
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! .f GE NucIcet Energy s-May 2,1990 (Corrected Date)
MFN No. 037 90 l
Docket No. STN 50 605 i
! Document Control Der,k U.S. Nuclear Regulatory Commission Washington, D.C. 20555 i
Attention: Charles L Miller, Director Standardization and Non Power Reactor Project Directorate
Subject:
Submittal of Amendment 11, Non Proprietary Information, to
/\
Submittal of Amendment 11, Propriettry Information, to GE's heference:
ABWR SSAR, MFN No. 038 90, dated May 2,- 1990
Dear Mr. Miller:
Enclosed are thirty foui copies of selected sections of Chapter 1, Introduction and General Description of Plant. Chapter 3, Design of Structures, Components, Equipment, and Systems, Chapter 5, Reactor Coolant System and Connected Systems, Chapter 6, Engineered Safety Features, Chppter 7, Instnanentation and Control Systems, Chapter 9, Auxiliary Systems, Chapter 10, Steam and Power Conversion, Chapter l1, Radioactive if'aste Mananment, Chapter 13, Conduct of Operation, Chapter 14, initial Test Program, Chapter 19, Response to & vere Accident Policy' Statement, Chapter 20, Guestion and Response Guide, of the Standard Safety Analysis Report-
- (SSAR) for the Advanced Boiling Water Reactor (ABWR).
This submittal includes additions covering the following: Response to standby gas treatment system questions, includes modifications to the design; Addition of a flammability control system to the atmospheric control system; Addition of the initial test program for the turbme island and radwaste facilities; Addition of the reactor service and turbine water systems; Resolution of draft safety evaluation report open items for Chapters 4,5 and 6 (except for preservice/ inservice inspection plan); Resolution of preliminary draft safety evaluation report items for Chapters 3 (except for inservice testing plan); Expansion of Chapter 7 that includes turbine island and radwaste facility instrumentation and controls; Balance of Chapter 7 and 10 question responses; i Update of applicable USIs/GSis; and Closecut of design related emergency preparedness issue.
l l
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1 j
[. -.
. i k.
O i Document Control Desk Docket No. STN 50 605 U.S. Nuclear Regulatory Commission MFN No. 037 90 May 2,1990 Page 2 In addition there are other changes to various sections of the SSAR which are identified, along with the above changes, on the page change instruction sheet.
Please note that all or parts of the following sections contain information that is designated as General Electric Company proprietary information: 7.7,19.5,19B and 20.3. This information is being submitted under separate cover.
Sincerely, r
SU 4 0 G.L. Sozzi, Actin Manager h
Lic,ensing and Consulting Services cc: D. R. Wilkins (GE)
T. E. Murley (NRC)
D. M. Crutchfield (NRC)
D.J. McGoff (DOE)
F. J. Miraglia, Jr. (NRC)
D. C. Scaletti (NRC)-
K. E. Stahlkopf (EPRI) -
F. A. Ross (DOE)
EEN9018
- MM 21A6100AE Standard Plant nv s 3.1 COhTORMANCE WITH NRC to safety shall be designed, fabricated, '
(~ GENERAL DESIGN CRITERIA erected, and tested to quality standards l
\. '
commensurate with the importance of the safety 3.L1 SUhBiARY DESCRIPTION functions to be performed. Where generally ;
recognized codes and standards are used, they i This section contains an evaluation of the shall be identified and evaluated to determine '
principal design criteria of the ABWR Standard their applicability, adequacy, and sufficiency Plant as measured against the NRC General Design and shall be supplemented or modified as Criteria for Nuclear Power Plants,10CFR50 necessary to assure a quality product in keeping Appendix A. The general design criteria, which with the required safety function. A quality are divided into six groups with the last assurance program shall be established and criterion numbered 64, are intended to establish implemented in order to provide adequate ;
minimum requirements for the principal design assurance that these structures, systems, a'nd ,
eriteria for nuclear power plants. components will satisfactorily perform their safety functions. Appropriate records of the The NRC General Design Criteria were intended design, fabrication, erection, and testing of to guide the design of all water cooled nuclear structures, systems, and components important to power plans; separate BWR specific criteria are safety shall be maintained by or under the -
not addressed. As a result, the criteria are control of the nuclear power unit licensee i subject to a variety of interpretations. For throughout the life of the unit, this reason, there are some cases where conformance to a particular criterion is not 3.1.2.1.1.2 Evaluation Against Criterion 1 directly measurable. In these cases, the conformance of the ABWR design to the interpreta- Safety.related and non. safety related tion of the criteria is discussed. For each structures, systems, and components are criterion, a specific assessment of the plant identified on Table 3.21. The total quality
(- design is made and a complete list of references assurance program is described in Chapter 17 and
\ is included to identify where detailed design is applied to the safety related iterns. The information pertinent to that criterion is quality requirements for non safety related treated in this .afety analyris report (SAR). items are controlled by the quality assurance program described in Chapter 17 in accordance Based on the content herein, the design of the with the functionalimportance of the item. The ABWR design fully satisfies and is in compliance intent of the quality assurance program is to with the NRC General Design Criteria, assure sound engineering in all phases of design and construction through conformity to 3.1.2 EVALUATION AGAINSTCRITERIA regulatory requirements and design bases described in the license application. In 3.1.2.1 Group ! . Overail Requirements addition, the program assures adherence to specified standards of workmanship and 3.1.2.1.1 Criterion 1. Quality Standards and implementation of recognized codes and standards Records in fabrication and construct!on. It also includes the observance of proper preoperational
-3.1.2.1.1.1 Criterion 1 Statement and operational testing and maintenance procedures as well as the documentation of the Structures, systems, and components important foregoing by keeping appropriate records. The total quality assurance program is responsive to and in conformance with the intent of the quality related requirements of 10CFR50 Appendix B.
l l O Amendment 11 3.11
ABM sissioare hadard Plant arv. n P
Structures, systems, and components are (1) Appropriate consideration of the most severe identified in Section 3.2 with respect to their of the natural phenomena that have been his- ,
location, service and their relationship to the torically reported for the site and sur.
safety related or non. safety related function to rounding area, with sufficient margin for be performe.d. Recognized codes and standards are the limited accuracy, quantity, and period applied to the equipment per the safety of time in which the historical data have -
classifications to assure meeting the required been accumulated; safety related function.
(2) Appropriate combinations of the effects of Documents are maintained which demonstrate normal and accident conditions with the that all the requirements of the quality effects of the natural phenomena; and assurance program are being satisfied. This documentation shows that appropriate codes, (3) The importance of the safety functions to be &
standards, and regulatory requirements are performed.
observed, specified materials are used, correct procedures are utilized, qualified personnel are 3.1.2.1.2.2 Evaluation Agalmat Criterion 2 provided, and the finished parts and components meet the applicable specifications for safe and Since the ABWR design is designated as a i reliable operation. These records are available standard plant, the design bases for so that any desired item of information is safety related (See Subsection 3.1.2.1.1.2) retrievable for reference. These records will be structures, systems, and components, cannot maintained during the life of the operating accurately reflect the most severe of the natu.
licenses, ral phenomena that have been historically re-ported for each possible site and their sur.
The detailed quality assurance program is in rounding areas. However, the envelope of site-conformance with the requirements of Criterion 1. related parameters which blanket the majority of potential sites in the conterminous United '
For further discussion, see the following States is defined in Chapter 2. The design sections: bases for these structures, systems, and compo-nents reflect this envelope of natural phenomena ,
Chapter / including appropriate combinations of the ef.
- Section Iitic fccts of normal and accident conditions with this envelope. The design bases meet the (1) 1.2 General Plant Description requirements of Criterion 2.
(2) 3.2 Classifiedon of Structures, Detailed discussion of the various phenomena Components, and Systems considered and design criteria developed are presented in the following sections:
3.1.2.1.2 Criterion 2. Design Bases for Protection Against Natural Phenomena Chapter /
SEtinn lillt 3.1.2.1.2.1 Criterion 2 Statement (1) 2.0 Summary of Site Characteristics Structures, systems, and components innportant to safety shall be designed to withstand the (2) 3.2 Classification of Structures, effects of nat:.ral phenomena such as earthquakes, Components, and Systems tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform (3) 3.3 Wind and Tornadol.cadings their safety functions. The design bases for these structure systems and components shall (4) 3.4 Water Level (Flood) Design reflect:
O l
Amendment 7 312
t
- ABWR ==s 1 Standard Plant am n g TABLE 3.21 CLASSIFICATION
SUMMARY
'Ihe classification information is presented System. wise
- La the following order: .
Table Table
' 3.21 MPL 3.21 MPL
' ham No. Malahst" Illit hem No. Enadisd* Illit B Nuclear Steam Suppit Systems E2 E22 High Pressure Core Floader System' B1 B11/J10 Reactor Pressure Vessel J11/J12 System */ Fuel' E3 E31 14,ak Detection and Isolation System' B2 B21 Nuclear Boiler System' E4 E51 RCIC System' _
B3 '
B31 Reactor Recirculation System F Reactor Senicing ,
C Control and Instrument Svatems F1 Fil Fuel Senicing Equipment C1 C11/C12 CRD System
- C2 C31 Feedwater Control System F3 F14 RPVInternal Senicing C3 C41 Standby Liquid Control Equipment System fG F4 F15 Refueling Equipment V C4 C51 Neutron Monitoring System' F5 F16 Fuel Storage Equipment C5 C61 Remote Shutdown System G D-tor Aurillary Systems C6 C71 Reactor Protection System" J
G1 G31 Reactor Water Cleanup System D Radiation Monitorine Systems G2 G41 Fuel Pool Coolir4 and Cleanup D1 D11 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 or subsysterns thereof, have a primaryfunction that is safety related. As shown in the balance of this Table, some of these systems contain non safety related components and, conversely, some systems whoseprimaryfunctions are non safety related contain components that have been designated safety-related.
p " Master Parts List Number designatedfor the system V- "* Only those systems that are in the ABH R Standard Plant scope are included in this table.
Amendment 7 - 3.24 1
ABWR - '
mums meandard Plant mv n TABLE 3.21 l CLASSIFICATION
SUMMARY
(Continued)
Table Table i 3.21 MPL 3.21 MPL hem No. h[ggdag** 3111t item No. Number ** 3111t J Nhelaar Fuel P Station Auxillars Systems See item B1 P1 P13 Makeup Water System '
(Condensate)
K Radioactive Waste Systems i P2 P21 Reactor Building Cooling Water System' K1 K11 Radioactive Drain Transfer System P3 P22 Turbine Building Cooling Water System K2 K17 Radwaste System PS P41 Reactor Cooling Water System ,
P6 P42 Turbine Cooling Water System P7 PS1/PS2 Instrument /Senice Air N power Nie Systems
/P54 /High Pressure Nitrogen Systems N1 N11/N21 Power Conversion System N22/N25 N26/N27 R Station Electrical Systems N31/N32 N33/N34 R1 R42 DC Power Supply' N35/N36 N37/N38 R2 R10/R11/ Auxiliary AC Power System' N39/N41 R22 N42/N43 i
, N44/N51 R3 R43 Emergency Diesel Generator N61/N71 System *
. N72 R4 RS2 Lighting and Senicing Power N2 N62 Offgas System Supply f
These systems or subsystems thereof. have a primarv function that is safetv-related. As shown in the balance of this Table. some of these systems contain non safety related components ond.
ronverselv. some systems whose primarv functions are non safety related contain components that have born desirnated saferv-related. ,
- hfaster Parts List Number desirnated for the system O
Amendment ?! 3.2-7
i
. . ABWR s w i.o a StandardflanL__,,_ mm a TABIE 3.21
,e .,
t CIASSIFICATION
SUMMARY
(Continued) )
l Quality
. Group Quality l Estet,y 14es- Clessi- Assursace Selsede Prinelpal Campamenta ggg e gge gd neanimment' M Egits 31 Reactor Pressert Vessel System / l Fast Assemblies J l
- 1. Reactor vessel 1 C A B 1 l
- 2. Reactor vessel support skirt 1 C A B 1 and stabilizer :
1
- 3. Reacter vessel appurtenances 1 C A B 1 (g)
)
pressure retaining portions
(
l l
- 4. Supports for CRD housing. 1 C A B ! ;
in. core housing and recircu- ;
lation laternal pump l S. Reactor internal structures - 3 C B B 1 feedwater, RHR/ECCS high ;
pressure core flooder spargers ,
\ '
- 6. Reactor internal structures- 3 C -- B 1 safety related components ;
including core support structures (See Subsection 3.9.5)
- 7. Reactor internal structures - N C ~ --- -- -
non-safety related components I.
(See Subsec6on 3.9.5) l.
l S. Control rods 3 C -- B I 9, = Power range detector hardware 3 C - B I including startup range detector
- 10. Fuelassemblies 3 C - B 1
- 11. Reactor Internal Pump 1 C A B 1 Motor Casing B2 Nuclear Boller System
- 1. Vessels levelinstrumenta- 2 C B B 1 tion condensing chambers .
O l Amendment It 328 a
ABWR noime -
-l Etandard Plant me. m TABLE 3.21 h
CLASSIFICATION
SUMMARY
(Continued)
I Quality Group Quality Safety I. ace. Classi. Assursace Seismic Princinal Componenta Caub ge gd m ge Capf h
- 2. Vessel air accumulators 3 C C B ! ;
I
- 3. Piping including supports. 2/3 C B/C B 1 (b) q safety / relief valve discharge g ,
O I
s 4
l l
i I
l O
Amendment 11 3.2-8.1
ABWR noms meandard Plant aw a
~N TABLE 3.21 '
(d CIASSIFICATION
SUMMARY
(Continued) !
1 Quality ,
hp Quality Safetr 14en. Classi- Assersaw Seismic :
Petacipal Campamenta gggb ggge gggggggd Requimment' f.ataparrf Matti B2 Nuclear Boller System (Continued) ,
- 4. Fipingincluding supports- 1 C,5C A B 1 l'
main steamline (MSL) and feed.
water (FW)line within outermost ,
isolation valve .
- 5. Pipingincuding supports . 2 SC B B 1 9 MSL and FW from outermost I isolation valve to and incluf.ing seismic interface restraint sad FW from outermost isolation to -
and including shutoff valve
- 6. Pipingincluding supports MSL N T D - -
from the seismic interface
.s q restraint to the turbine stop valve V
- 7. Deleted
- 8. Piping FW beyond seismic N T D - -
interface restraint
- 9. Seismic Interface Rest.aint - 2 SC B B I MSL/FW
.10. Pipe whipiestraints MSL/FW 3 SC,C - B ---
8 11. Piping including supports . Other
$ within outermost isolation valves ,
- a. RPV head vent 1 C A B 1 (g)
- b. RPV bead spray 1 C A B I (g)
- c. Main steam drains 1 C,SC A B 1 (g)
% 12. Pipingincudingsupports other I 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 -- -
ON !
l Amendment 11 329 i
. .RennA.=d MM Plant 214610QAE arv.n TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued)
Quality Group Quality Set 14cs. Classi- Assursace Seismic Paindmal Campamenta Clata Magt md Regniment' M h C5 Remote Shutdown System Components of this system are included under B2, Ei, E4, G3, H2, and P2.
- 1. Electrical modules with 3 C,5C,RZ, -- B 1 safety-related function X
- 2. Cable with safety related 3 RZ - B 1 function C6 Reactor Protection System
- 1. Electrical modules with 3 SC,X,T, --- B 1 safety related function RZ
- 2. Cable with safety 3 SC,X,T, - B 1 O' related functions RZ
- 3. Electrical Modules, other N T,X - -- ---
(u)
- 4. Cable, other N T,X --- -- ---
(u)
D1 Process Radiation Monitoring System (includes gaseous and liquid amuent monitoring)
- 1. Electricalmodules with 3 SC,X,RZ -- B 1 with safety related functions (includes monitors)
- 2. Cable with safty related 3 SC,X,RZ - B 1 functinns
- 3. Electrical Modules, other N T,SC,RZ, --- -- -
(u)
X,W
- 4. Cable, other N T,SC,RZ, -- - -
(u)
X,W O
Amendment 3 3.2 13 ,
l l
l
{
ABWR nome - -
meandard Plant arv m TABLE 3.21 CLASSIFICATION
SUMMARY
(Continned)
Quality Group Quality Safg I4ce. Classi. Assurance Seismic IPriadnal Cannoses,a QAAA 19A' fkation0 WM h D2 Costalement Atmospbele Monitorlag System
- 1. Component with safety-related 3 C,SC --- B I -
El RHRSystem
- 1. Heat exchangers primary side 2 SC B B 1
- 2. Heat exchangers induding 3 SC C B I supports secondaryside l 3. Piping including supports
- 1/2 C SC A/B B I (g) within outermost isolation valves
- 4. Containment spray piping 2 C B B 1 includmg supports and l
a spargers, within and induding the outer-most isolation valves 4a. Pipinginduding supports 2/3 SC B/C B (g) l beyond outermost isolation 1
valves
- 5. Main Pumps induding supports 2 SC B B I
- 6. Main Pump motors 3 SC -- P I
- 7. Valves isolation,(LPFL 1 C SC A B 1 (g) line) including shutdown suction liae isolation valves
- 8. Valves isola %other 2 C,SC B B I (g)
(pool suction valves and pooltest return valves)
- 9. Valves beyond isolation 2/3 SC B/C B 1 (g) valves
- 1he RHR/ECCS lowpressureflooder spaneers arepa's of the reactorpressure wsselsystem, see hem Bl.$.
Amendment 11 3.2 14
ABM .
.1. l Etendard Plant ma I
i TABLE 3.21 O
V '
CLASSIFICATION
SUMMARY
(Continued)
Quality Group Quality
! 14ce. Classi. Assursace Seismic Safeg Pdncinal Commonenta Qan Mont Scatland Enquirement' Catapenf Natu El RHR System (Continued)
- 10. Mechanica) modules with 3 SC C B I safety-related functions
- 11. Electrical modules with 3 C,SC,X -- B I safety related function - .
l 12. Cable with safety related 3 C,SC,X --- B I function ,
- 13. Other mechanical and N C,SC,X - - -
electrical modules l
l N
- 14. Jockey pumps including sVpports 2 SC B B I t
- 25. Jockey pump motor N SC -- --- ---
t O
Amendment 11 3 2-148
ABM nssimis Etandard Plant arv n TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued)
O, ,
t Quality Group Quality Safety Lace- Classi. Assursace Seismic Principal Componenta Qanb ge gd Requirement' Catamorvf Hein E2 High Pressure Cort Flooder
- 1. Reactor pressure vessel 1/2 C,SC A/B B I -(g) injection line and connected g
pipingincluding supports with-
, in outermost isolation vaht*
cs
- 2. All other piping including 2/3 SC,0 B/C B 1 (g) supports ** '
- 3. Main Pump 2 SC B B 1
- 4. Main Pump motor 3 SC - B I S. Valves outer isolation 1 C,SC A B I (g) and within the reactor pressure E vesselinjection line and "
connected lines
- 6. All other valves 2/3 SC B/C B I (g) l
- 7. Electrical modules with safety- 3 C,SC,X -- B 1 related function *
- 8. Cable with safety related 3 C,5C,X --- B I function E3 leak Detection and Isolation System
- 1. Temperature sensors 3/N C,SC --
B/- - I/- (z)
- 2. , Temperature switches 3/N X ~
B/ - 1/~ (z)
- 3. Pressure transmitters 3/N C,SC --
B/- 1/~ (z)
- 4. Pressure switches 3/N X ---
B/ - 1/- - (z)
- 5. Differential pressure 3/N C,SC --
B/--- I/-- (z) transmitters (flow)
The ECCS high pressure core flooder spargers are part of the Reactor Pressure Vessel System, see t item Bl.3.
l Pool suction piping, suction pipingfrom condensate storage tank, test line to pool, pump discharge piping and retum line topool.
l Amendment 11 32 15
- .MM 21A6100AE Standard Plant arv a TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued)
Quality Geoup Quality Safe Lace. Classi. Assurance Seismic Principal Commonent* Qau USAC ficationd Requirement' Qdamor/ E01gg F2 RWServicing Equipment i
- 1. Steamline plugs N SC -- - -
- 2. Dryer and separator N SC - - -
strongback and head strongback F3 RPV Internal Servicing Equipment
- 1. Control rod grapple N SC - -- --
F4 Refueling Equipment
- 1. Refueling equipment N SC . --. 1 (bb) h platform assembly "
- 2. Refueling bellows N SC - - ---
- F5 Fuel Storage Equipment
- 1. Fuel storage racks- N SC . - 1 (bb) new and spent j
N
- 2. Defective fuel storage N SC --. - -
(bb) container G1 Reactor Water Cleanup System
- 1. Vessels including supports N SC C - -
q (filter /demineralizer)
E
" 2. N Regenerative heat exchangers SC C -- -
including supports carrying reactor water
- 3. Cleanup recirculation N SC C -- -
pump, motors Amendment 11 3.2 18
MM 21A6100AE arv s ggadsrd Plant TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued)
Quality Group Quality Safet4 Classi d Assurance , Seismic g j Prinmaal Camponent, Qag e%.
gg $ganga Reaufrunent CalgDEI }idu 1 1
G1 Reactor Water Cleanup System (Contissed) l
- 4. Pipingincluding supports and 1 C,SC A B I (g) valves within and including outermost containment isolation valves on pump suction
- 5. Pump suction and discharge N SC C -- - (g) piping induding supports and valves from containment isola-tion valves back to shut off i valves at feedwater line l connections
- 6. Pipinginduding supports and 2 SC B B 1 (g) valves from feedwater lines to '
and including shut off valves
- 7. Pipinginduding supports and N SC,T C - -- (g) valves to main condenser 2 sN
- 8. Non-regenerative heat exchanger N SC C -- - (g) tube inside and pipinginduding
- supports and valves carrying l process water
- 9. Non-regenerative heat exchanger N SC D -- ---
shell and pipingincluding supports carrying closed cooling water
- 10. Fdter/demineralizer N SC D - ---
j precoat subsystem l-
- 11. Filter demin holding pumps N SC C - -
includingsupports valves and piping induding supports O
Amendment 3 3.2 19
. . MM iss62aars mem=Aard Plant nya TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued)
Quality Group Quality ;
Safety Lace. Classi. Assurance Seismic Pdadpal Componenta h Men' Md Requimment' Calaam2I E91ta
- 4. Turbine bypass piping N T D - -
including supports
- 5. Turbine stop valve, turbine N T D -- -
(1)(n)(o) bypass valves, and the main steam leads from the turbine controh31ve to the turbine casing
- 6. Feedwater system components N T D N/A N/A beyond outboard shutoff valve
- 7. Turbine generator N T - - --
l
- 8. Condenser N T -- -- -- ,
- 9. Air ejector equipment N T - - -
\
- 10. Turbine gland sealing N T D - --
- system components N1 Power Conversion System (Later)
N2 Ofigas System
- 1. Pressure vessels induding N T - - . (p)(q) supports
( 2. Atmospheric tanks including N T - -- -- (p)(q) supports
- 3. 015 psig tanksincluding N T -- -- - (p)(q) supports
- 4. Heat exchangers including N T -- .. -- (p)(q) supports
- 5. Pipingincluding supports N T -- -- -- (p)(.q) and valves
- 6. Pumps including supports N T -- - - (p)(q) J 1 l Amendment 9 3.2-21b
i ABWR . zwtou.
Remndard Plant arv s TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued)
O Quality ,
Group Quality :
Safety Emes. Classi. Assurance Seismic Principal Camponenta d Requirement' Caggenf Cla11b llant h Egigg l P1 Makeup Water System (Condensate)
- 1. Pipinginduding supports and 2 C B B I valves forming part of the con-tainment boundary ,
- 2. Condensate storage tank N O D - --
(w) g induding supports R
- 3. Condensate header piping 2 SC B B I induding supports and valves
- 4. Pipingincluding supports and N O D - -
valves
- 5. Other components N O D -- -
P2 Reactor Building Cooling Water System O
- 1. Piping and valves forming part 2 SC,C B B I (g) of primary containment boundary l
a
- 2. Other safety related piping, induding supports pumps and 3 SC,C C B I valves
- 3. Electrical modules with 3 SC,C,X -- B 1 with safety related function ;
- 4. Cable with safety related 3 SC,C,X - B I function
- 5. Other mechanical and N SC,C,X,M - -- -
electrical modules P3 Turbine Building Cooling N T D - -- !
Water System 1
Ol l Amendment 11 3.2 22 1
l
MM 22A6 ooAs a
c Standard Plant ny n TABLE 3.21 l 1
Il
\) CLASSIFICATION
SUMMARY
(Continued) I Quality j Group Quality 1 Loca. Classl. Assurance Selsmle Safet4 gan*
- Princinal Comnonent* Class Rgadog Jtenuirement' Catenon- Mcles P4 HVAC Cooling Water Systems
- 1. Chillers, pumps, vahms, and 3 SC,X C B 1 piping including supports-2 Safety related HVAC support t I,
- 2. Chillers, pumps, valves, and N C,SC,RZ, -- - -
i piping including supports . .-
non safety related HVAC tupport
- 3. Piping including supports and 2 C,SC B B I valves forming part of contain.
ment boundary
- 4. Electrical modules and cable 3 SC,X --- B 1 with safety related function -
.. Other mechanical and' N C SC,RZ, --- - ...
electrical modules T,X ,
P P5 Reactor Cooling Water System
- 1. Safety related piping 3 U,0,X C B I including supports, piping andvahts
- 2. Electrical modules and cables 3 U,0,X - B I with safety-related function -
- 3. Other non safety related N U,0,X --- --- .
mechanical and electrical modules P6 Turbine Cooling Water System
- 1. Non-safety related piping N P,0,T - - --
including supports, piping and valves rh
()
.. Amendment 11 3.2 23 I.
7<
l -
gq t Standard Plant arv.n j TABLE 3 21 -
CLASSIFICATION
SUMMARY
(Continued) i Quality Group Quality Loca- Classi. Assurance Selsmic Prineinal Comnonent" Safet4 Qgg Efglga I (Lon Renuirement' Catenorv Egtg
- 2. Eletrical modules and cables N P,0,T - -- -
with non safety related - .
function P7 Instrument / Service AirSystems 8 1. Containment isolation includ. 2 C B B I I ing supports vahts and piping
- 2. Other non safety related N SC,RZ, -- --- --
mechanical and electrical X,T H, components W,C P8 High Pressure Nitrogen Systems
- 1. Containment isolation includ- 2 C B B I ing supports valves and piping
- 2. Piping including supports with 3 SC,C C B I safety related function
- 3. Electric modules with 3 RZ,X -- B 1 safety related functions
- 4. Cable with safety related 3 SC,RZ, -- B I l function X ,
- 5. Other non safety related N SC,RZ, --- -- --
mechanical and electrical X components RI . DC Power Supply Nuclear Island l
E 1. 125 volt batteries, battery 3 SC,C,X, -- B I ,
I racks, battery chargers, and RZ distribution equipment O
Amendment 11 3.2-23,1
1 23A6100AE Reamd.ed Plant nyn TABLE 3.21 CIASSIFICATION
SUMMARY
(Continued) ;
l Quality >
Group Quality Safe Loca. Classi- Assursace Seismic Primelpal Casanonenta Qaan ann' Sca110nd Renutrement' Categorvf Enita
- 2. Controlpower cables (includ. 3 SC,C,X, - B 1 ing under ground cable system, RZ cable splices, connectors and terminalblocks)
- 3. Conduit and cable trays and 3 SC,C,X, - B I their supports RZ I
- 4. Protective relays and control 3 SC,X,RZ - B I panels
- 5. Containment electrical pene- 3 SC,C - B I trations assemblies
- 6. Motors 3 SC,C,X, - B I RZ 9 R2 Auxillary AC Power System
- 1. 6900 volt switch gear 3 SC,X,RZ - B I
- 2. 480 volt load centers 3 SC,X,RZ - B 1
- 3. 480 volt motor control 3 SC,X,RZ -- B I centers
- 4. 120 VAC safety related dis- 3 SC,X,RZ -- B I tribution equipment includ-inginverters 1
- 5. Control and power cables 3 SC,C,X - B I (including underground RZ cable systems, cable splices, connectors and terminal blocks)
- 6. Conduit and cable trays and 3 SC,C,X --- B I their supports RZ
- 7. Containment electrical 3 SC,C,X - B I penetration assemblies RZ
- 8. Transformers 3 SC,C,X - B I RZ l
Amendment 10 3.2 24 l
21A6130AE L RennAmed Plant ' arv.s _
j
- TABLE 3.21 i
L C1.ASSIFICATION
SUMMARY
(Continued) l.
l Quality Group Quality
' Safety - IAce- Classi. Assurance Seisale Princinal Cannonenta gggg e ggge gd Requiremente Catenor/ ]i91g1
- 9. Motors 3 SC,C,X,RZ - B I i 10. Imad sequencers 3 SC,X,RZ - B I E 11. Protective relays and control 3 SC,X,RZ - . B I I panels
- 12. Valve operators 3 SC,C,X,RZ - B I R3 Emergency Diesel Generstor System
- 1. Starting air receiver tanks 3 RZ C B I (y) pipingincluding supports from and including check valve and downstream pipingincluding supports and valves
- 2. Starting air compressor and N RZ - - --
motors 3.' Combustion air intake and 3 RZ,0 C B I 1 exhaust system
- 4. Safety related pipingin- 3 RZ,0 C B I cludingsupportsvalves fuel oil system, diesel cooling water system, and lube oil system
- 5. Pumpmotors fueloil 3 RZ,0 - B I system, diesel cooling water system and lube oil system
- 6. Dieselgenerators 3 RZ - B I (y)
- 7. Mechanical and electrical 3 RZ,0,X --- B I modules with safety related functions 8 Cable with safety related 3 RZ,O,X - B I functions
- 9. Other mechanical and N RZ,0 - --- -
electrical modules Amendment 10 3.2 24.1
~. ABM 2mioorn
^
- RennA.ed Plant nev. A TABLE 3.21
. /- . .
4 -
CLASSIFICATION
SUMMARY
(Continued)
).
Quality Group Quality Loca. Assurance Seismic Principal Component" Safet4 Classi. d Reautrement' I
Qasi llan* fication Categorv halu R4 ughtlag and Servicing Power Supply N
- 1. Emergency Lighting' SC,C,X - - -
RZ T1 Primary Containment Gystem
- 1. Primary containment vessel 2 C B B I (PCV)- reinforced concrete containment vessel (RCCV)
- 2. Vent system (vertical flow 2 C B B I channels and horizontal discharges f
( .
- 3. Suppression chamber /drywell vacuum breakers 2 C B B I
- 4. PCV penetrations 2 C B B I and drywell steel head
- 5. Upper and lower drywell airlocks 2 C,SC - -- B I
- 6. Upper and lower drywell 2 C,SC --- B I equipment hatches
- 7. Lower drywell access tunnels 2 C -- B I
- 8. Suppression chamber 2 C,SC --- B I access hatch
- 9. Safety related instrumentation 2 C,SC --- B I T2 Containment Internal Structures
- 1. Reactor vessel stabilizer 3 C -- B I truss O
G Amendment 2 3.2 25
l 23A6100AE Etandmed Plant nev.s TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued)
Quality Group Quality Safe Loca. Classi- Assurance Seismic Principal Componenta Qan linn' Balland Requirtatal' C='=orvf Entu T2 Con *=l===nt Internal Structorse (Continued) g 2. Support structures for safety- 3 C -- B I 2 related pipingincluding ,
supports and equipment T3 RPV Pedestal and Shield Wall
- 1. RPV pedestal and shield wall 3 C --- B 1
- 2. Diaphragm floor 3 C - B 1 T4 Standby Gas Treatment System
- 1. . All equipment except 3 SC,C,RZ --- B I deluge piping and valves
- 2. Deluge piping and valves N SC -- -- --
Qtt)
O Amendment 11 3.2 26
< . MM 2M6100AE Standard Plant nrv. n
/
- a. A module is an assembly of interconnected components which constitute an identifiable device or piece of equipment. For example, electrical modules include sensors, power supplies, and signal processors and mechanical modules include turbines, strainers, and orifices,
- b. 1,2,3, N = Nuclear safety related function designation defined in Subsections 3.2.3 and 3.2.5.
- c. C = Primary Containment H = Service Building M = any other location O = Outdoors onsite RZ = Reactor Building Clean Zone (balance portion of the reactor building outside the Secondary Containment Zone)
SC = Secondary Containment portion of the reactor building T = Turbine Building W = Radwaste Building ,
X = Control Building -
F = Firewater Pump House U = Ultimate Heat Sink Pump House P= Power Cycle Heat Sink Pump House
- d. A,B,C,D = Ouality groups defined in Regulatory Guide 1.26 and Subsection 3.2.2. The structures, systems and components are designed and constructed in accordance with the requirements identified in Tables 3.2-2 and 3.2 3.
-~
= Ouality Group Classification not applicable to this equipment.
- c. B = the quality assurance requirements of 10CFR50, Appendix B are applied in accordance with the quality assurance program described in Chapter 17.
-~
= Requirements of 10CFR50, Appendix B are not applicable,
- f. I = The design requirements of Seismic Category I structures and equipment are applied as described in Section 3.7, Seismic Design.
-- = The seismic design requirements for the safe shutdown earthquake (SSE) are not applicable to the equipment. However, the equipment that is not safety related but which could damage Siesmic Category I equipment if its structural integrity failed is checked analytically and designed to assure its integrity under seismic loading resulting from the SSE.
- g. 1. Lines one inch and smaller which are part of the reactor coolant pressure boundary shall be ASME Code Section 111, Class 2 and Seismic Category I.
- 2. Allinstrument lines which are connected to the reactor coolant pressure boundary and are utilized to actuate and monitor safety systems shall be Safety Class 2 from the outer isolation valve or the process shutoff valve (root valve) to the sensing instrumentation.
- 3. Allinstrument lines which are connected to the reactor coolant pressure boundary and are not utilized to actuate and monitor safety systems shall be Code Group D from the outer isolation 9 valve or the process shutoff valve (root valve) to the sensing instrumentation.
1 Amendment 11 3.2 31
l ABM _ sissioors . .
Standard Plant an a NQIES (Coutinued)
- 4. All other instrument lines: ,
i Through the root valve the lines shall be of the same classification as the system to which they are attached.
ii Beyond the root valve,if used to actuate a safety system, the lines shall be of the same classification as the system to which they are attached.
iii Beyond the root valve,if not used to actuate a safety system, the lines may be.
Code Group D.
- 5. All sample lines from the outer isolation valve or the process root valve through* the ,
remainder of the sampling system may be Code Group D.
m 6. All safety relaed instrument sensing lines shall be in conformance with the criteria of
- h. Relief valve discharge piping shall be Quality Group B and Seismic Category I.
Safety / relief valve discharge line (SRVDL) piping from the safety / relief valve to the quea-chers in the suppression pool consists of two parts: the fir <t part is attached at one end to the safety / relief valve and attached at its other end to the diaphragm floor pecetration.
This first portion of the safety / relief valve discharge piping is analyzed with the main steam piping as a complete system. The second part of the safety / relief valve discharge piping extends from the penetration to the quenchers in the suppression pool. Because of the '
penetration on this part of the line, it is physically decoupled from the main steam piping .
and the first part of the SRVDL piping and is, therefore, analyzed as a separate piping system.
- i. Electrical devices include components such as switches, controllers, solenoids, fuses, j . junction boxes, and transducers which are discrete components of a larger subassembly /
l module. Nuclear safety related devices are Seismic Category 1. Fail safe devices are non Seismic Category I.
- j. The control rod drive insert lines from the drive flange up to and including the first valve on the hydraulic control unit are Safety Class 2, and non safety related beyond the first valve.
I k. The hydraulic control unit (HCU)is a factory assembled engineered module of valves, tubing, l piping, and stored water which controls two control rod drives by the application of pressures and flows to accomplish rapid insertion for reactor scram.
Although the hydraulic control unit, as a unit,is field installed and connected to process l piping, many of its internal parts differ markedly from process piping components because of the more complex functions they must provide. Thus, altt ough the codes and standards invoked by Groups A, B, C, and D pressure integrity quality levels clearly apply at all levels to the interfaces between the HCU and the connection to conventional piping components (e.g., pipe nipples, fittings, simple hand valves, etc.), it is considered that they do not apply to the l . specialty parts (e.g., solenoid valves, pneumatic components, and instruments).
I O Amendment 6 3.2-32 l
< <ABWR. nwma Etandard Plant arv. n NOTES (Coatiased)
The design and construction specifications for the HCU do invoke such codes and standards as can be reasonably applied to individual parts in developing required quality levels, but of the remaining parts and deteils. For example: (1) all welds are LP inspected; (2) all socket welds are inspected for gap between pipe and socket bottom; (3) all welding is performed by qualified welders; and (4) all work is donc per written procedures. Quality Group D is generally applicable because the codes and standards invoked by that group contain clauses which permit the use of manufacturer standards and proven design techniques which are not explicitly defined within the codes for Quality Groups A, B, or C. This is supplemented by the QC technique described.
- 1. The turt.ine stop valve is designed to withstand the SSE and maintain its integrity. ,
- m. The RCIC turbine is not included in the scope of standard codes. The assure that the turbine is fabricated to the standards commensurate with safety and performance requirements, General Electric has established specific design requirements for this component which are as follows:
- 1. All welding shall be qualified in accordance with Section IX, ASME Boiler and Pressure Vessel Code.
- 2. All pressure.containing castings and fabrications shall be hydrotested at 1.5 times the design pressure.
- 3. All high-pressure casting. shall be radiographed according to:
ASTM E 94
.- E 141 E 142 maximum feasible volume E 71,186 or 260 Severitylevel 3
- 4. As-cast surfaces shall be magnetic particle or liquid penetrant tested according to ASME Code,Section III, Paragraphs NB 2575, NC 2576, or NB 2576, and NC 2576.
- 5. Wheel and shaft forgings shall be ultrasonically tested according to ASTM A-388.
- 6. Butt welds shall be radiographed and magnetic particle or liquid penetrant tested according to the ASME Boiler and Pressure Vessel Code. Acceptance standarus shall be in accordance with ASME Boiler and Pressure Vessel Code Section III, Paragraph NB-5340, NC 5340, NB 5350, or NC 5350, respectively.
- 7. Notification shall be made on major repairs and records maintained thereof.
- 8. Record system and traceability shall be according to ASME Section III, NCA 4000,
- 9. Control and identification shall be according to ASME Section III, NCA-4000.
- 10. Procedures shall conform to ASME Section III, NB-5100 and NC 5100,
- 11. Inspection personnel shall be qualified according to ASME Section III, NB-5500 and NC-5500.
O Amendment 10 3.2 33 l
l
MM 23A6100AE Etandard Plant nry a EQ2EE (Continued)
- n. All cast pressure retaining parts of a size and configuration for which vol ~
effective are examined by radiographic methods by qualified personnel. Ultrasonic examination to equivalent standards is used as an alternt.te to radiographic methods. Examination procedures and acceptance standards are at least equivalent to those defined in Paragraph 136.4, Nonboiler External Pipng, ANSI B31.1.
- o. The following qualifications are met with respect to the certification requirements:
- 1. The manufacturer of the turbine stop valves, turbine control valves, turbine bypass valves, and main steam leads from turbine control valve to turbine casing utilizes quality control procedures -
equivalent to those defined in GE Publication GEZ-4982A, General Electric Large Steam Turbine Generator Quality Control Program.
- 2. A certification obtained from the manufacturer of these valves and steam loads demonstrates that the quality control program as defined has been accomplished.
- p. Regulatory Guide 1.143 furnishes complete design guidance relating to seismic and quality group classification and quality assurance provisions for radioactive waste management systems, structures and components,
- q. Detailed seismic design criteria for the offgas system are provided in Section 113.
- r. The portions of the MSL from the second isolation valve to the turb ne stop valve and the first valve in branch lines shall be designed so that the SSE does not cause structuralinteraction or failure that could degrade the functioning of a Seismic Category I structure system or component to an unacceptable safety level,
- s. Not used
{ t. There is a limited quality assurance program for the Fire Protection System.
- u. Special seismic qualification and quality assurance requirements are applied,
- v. Not used.
- w. The condensate storage tank will be designed, fabricated, and tested to meet the intent of API Standard API 650. In addition, the specification for this tank will require: (1) 100% surface examination of the side wall to bottom joint and (2) 100% volumetric examination of the side wa!!
weld joints.
- x. The crancs are designed to hold up their loads under conditions of OBE and to maintain their positions over the units under conditions of SSE.
- y. All off engine components are constructed to the extent possible to the ASME Code,Section III, Class 3.
- z. Components associated with a safety-related function (e.g., isolation) are safety-related.
aa. Structures which support or house safety related mechanical or electrical components are safety related, bb. A quality assurance requirements 04!! be snplied to ensure that the design, construction and testing l requirements are met. i i
l Amendment 11 3.2 M I
l
- MM 23A6100AE -
an n
[ Standard Plant 3.7 SEISMICDESIGN 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 I. The requirements for the public are designed to remain functional.
Seismic Category I 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 .
as defined herein and other dynamic loads includ- The seismic design for the SSE is intended to ing those due to reacter building vibration (RBV) provide a margin in design that assures caused by suppression pool dynamics. Although capabili ty to shut down and maintain the this section addresses seismic aspects of design nuclear facility in a safe condition. In this case, it -is only necessary to ensure that the l and 1.70,analysis the methods in accordance of this section with Regulatory are also Guide and components do not lose required systems applicable to other dynamic loading aspects, their capability to per form their except for the range of frequencies considered. safety related function. This is referred to as The cutoff frequency for dynamic analysis is 33 the no loss of function criterion and the Hz for seismic loads and 80 ZHz for suppression loading condition as the SSE loading condition.
pool dynamic loads. The definition of rigid l system used in this section is applicable to Not all safety related components have the seismic design only, same functional requirements. For example, the
~
reactor containment must retain capability to The safe shutdown earthquake (SSE) is that restrict leakage to an acceptable level, earthquake which is based upon an evaluation of Therefore, based on present practice, clastic the maximum earthquake potential considering the behavior of this structure under the SSE loading 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 Seismic 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 reactor coolant pressure Table 3.2-1 identifies the equipment in boundary; various systems as Seismic Category I or non-Seismic Category I.
(2) the capability 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 exposures comparable to The design earthquake loading is specified in the guideline exposures of 10CFR100. terms of a set of idealized, smooth curves cr.lled the design response spectra in accordance The operating basis earthquake (OBE) is that with Regulatory Guide 1.60.
carthquake which, considering the regional and local geology, seismology, and specific charac- Figure 3.71 shows the standard ABWR design teristics of local subsurface material, could values of the horizontal SSE spectra applied at L -
reasonably be expected to affect the plant site the ground surface in the free field for damping during the operating life of the plant. Itis ratios of 2.0, 5.0, 7.0 and 10.0% of critical Amendment 11 3.7-1
ABM 234stoorn Sitadard Plant nry 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 damping 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 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 i
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 the spectra shown in Figures analysis and evaluation, 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 3.7 3 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 maximum 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 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 Plan, SRP). However, the response spectra for acceleration time histories generated to match the higher damping values of 7 and 10 percent i 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 j (Reference 2) because of its intrinsic capability tirne 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 carthquake 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 consistent with the SRP requirements 3.7 2. (See Subsection 3.7.1.3).
The OBE given in Chapter 2 is one third of 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 on e-h alf of th e SSE, i.e., 0.15 g, was employed to evaluate the structural and component response.
l l
I Amendment 1 3.7-2 l
l t
t MM 21A6t00AE j
. ' Standard Plant arv. n 3.7.3.8.2.2 Effect of Differential Building adequately accounted for in the analysis. In f' '
Movements case of buried systems sufficiently flex- I
- (
~
ible relative to the surrounding or under- ;
The relative displacement between anchors is lying soil, it is assumed that the systems determined from the dynamic analysis of the will follow essentially the displacements and structures. The results of the relative anchor- deformations that the soil would have if the ;
point displacement are used in a static analysis systems were absert. When applicable, '
to determine the additional stresses due to procedures, which take into account the relative anchor point displacements. Further phenomena of wave travel and wave reflection details are given in Subsection 3.7.3.8.1.8. in compacting soil displacements from the ground displacements, are employed. '
3.7.3.9 Multiple Supported Equipment Components With Distinct inputs (2) The effects of static resistance of the surrounding soil on piping deformations or ;
The procedure and criteria for analysis are displacements, differential movements of deseribed in Subseetioas 3.7.2.1.3 and piping anchors, bent geometry and curvature 3.7.3.3.1.3. changes, etc., are considered. When-applicable, procedures utilizing the '
i 3.7J.10 Use of Constant Vertical Static principles of the theory of structures on
! Factors clastic foundations are used.
All Seismic Category I subsystems and compo- (3) When applicable, the effects due to local' nents are subjected to a vertical dynamic soil settlements, soil arching, etc., are analysis with the vertical floor spectra or time also considered in the analysis, histories defining the input. A static atsalysis l-is performed in lieu of dynamic analysis if the 3.7.3.13 Interaction of Other Piping with peak value of the applicable floor spectra times Seismic Category I Piping i a factor of 1.5 is used in the analysis. A factor of 1.0 instead of 1.5 can be used if the In certain instances, non Seismic Category I equipment is simple enough such that it behaves piping may be connected to Seismic Category I essentially as a single degree of freedom piping at locations other than a piece of equip-
$ system. If the fundamental frequency of a compo- ment which, for purposes of analysis, could be
- 't ent in the vertical direction is greater than or represented as an anchor. The transition points equal to 33 Hz, it is treated as seismically typically occur at Seismic Category I valves
- rigid and analyzed statically using the which may or may not be physically anchored.
zero pe sponse spectrum. Since a dynamic analysis must be modeled from pipe anchor point to anchor point, two options 3.7.3.11 Torsional Effects of Eccentric Masses exist:
( Torsianal effects of eccentric masses are (1) specify and design a structural anchor at included for Seismic Category I subsystems the Seismic Category I valve and analyze the similar to that for the piping systems discussed Seismic Category I subsystem; or, if in Subsection 3.7.3.3.1.2. impractical to design an anchor, 3.73.12 Buried Seismic Category I Piping and (2) analyze the subsystem from the anchor point l Tunnels in the Seismic Category I subsystem through l the valve to either the first anchor point l1 For buried Category I buried piping systems in the non Seismic Category I subsystem; or l and tunnels the following items are considered in to sufficient distance in the non. Seismic l'
the analysis: Category I Subsystem so as not to significantly degrade the accuracy of (1) The inertial effects due to an earthquake analysis of the Seismic Category I piping. 1 O upon buried systems and tunnels will be '
l \,) 1 l
Amendment tt 3.7-23 i
1
i 1
1 ABM miasinors Remndard Plant MVA Where small, non Seismic category piping is (6) recording and playback equipment; and i
'directly attached to Seismic _ Category I piping, I its effect on the Seismic Category I piping is (7) annunciators. I accounted for by lumping a portion of its mass l with the Seismic Category I piping at the point The location of seismic instrumentation is l of attachment. outlined in Table 3.7 7.
Furthermore, non Seismic Category I piping 3.7.4.2.1 'Ilme History AA.y.phs (particularly high energy piping as defined in Section 3.6) is designed to withstand the SSE to Time history accelerographs produce a record avoid jeopardizing adjacent Seismic Category I of the time varying acceleration at the sensor piping if it is not feasible or practical to location. This data is used directly for analy-isolate these two piping systems. sis and comparison with reference information and may be, by calculational methods, converted to 3.7J.14 Seismic Analysis for Reactor response spectra form for spectra comparisons laternals with design parameters.
I 1
The modeling of RPV internals is discussed in Each triaxial acceleration sensor unit con-Subsection 3.7.2.3.2. The damping values are tains three accelerometers mounted in an ortho-given in Table 3.71. The seismic model of the gonal array (two horizontal and one vertical).
RPV and internal is shown in Figure 3.7 32. All acceleration units have their principal axes oriented identically. The mounted units are '
3.7.3.15 Analysis Procedures for Damping oriented so that their axes are aligned with the building major axes used in development of the Analysis procedures for damping are discussed mathematical models for seismic analysis, in Subsection 3.7.2.15.
One THA is located on the reactor building 3.7.4 Seismiclustrumentation (RB) foundation mat, El (-) 13.2 M, at the base of an RB clean zone for the purpose of measuring.
3.7.4.1 Comparison with NRC Regulatory Guide the input vibratory motion of the foundation 1.12 mat. A second THA is located in an RB clean zone at El (+) 26.7 M on the same azimuth as the The seismic instrumentation program is foundation mat THA. They provide date on the consistent with Regulatory Guide 1.12. frequency, amplitude, and phase relationship of the seismic response of the reactor building 3.7.4.2 Location and Description of structure. A third THA is located in the free lastnamentation - field at the finished grade approx.imately 160 M from any station structures with axes oriented in The following instrumentation and associated the same direction as the reactor building equipment are used to measure plant response to accelerometers, earthquake motion:
Two seismic triggers, connected to form redun-(1) three triaxial time history accelerographs dant triggering, are provided to start the THA (THA); recording system. They are located in the free field at the finished grade 160 M from the reac-(2) three peak-recording accelerographs (PRA); tor building. The trigger unit consists of or-thogonally mounted acceleration sensors that act-(3) two triaxial seismic triggers; unte relays whenever a threshold acceleration is exceeded for any of the three axes. The trigger (4) one seismic switch (SS); is engineered to discriminate against false starts from other operating inputs such as traf.
(5) four response spectrum recorders; fic, elevators, people, and rotating equipment.
O Amendment 1 1 7-24
' MN . 23A6100AE
. standard Flant REV A SECTION 3.9 f(]
CONTENTS Section lhlt East 3.9.1 Snecial Toples for Mechanical Components 3.91 3.9.1.1 Design Transients 3.9-1 3.9.1.2 Computer Programs Used in Analyses 3.9 1 3.9.13 Experimental Stress Analysis 3.9-1 3.9.13 1 Piping Snubbers and Restraints 3.9-1 3.9.13.2 Fine Motion Control Rod Drive (FMCRD) 3.9-1 3.9.1.4 Considerations for the Evaluation of Faulted Conditions 3.91
-3.9.1.4.1 Control Rod Drive System Components 3.9 2 3.9.1.4.1.1 Fine Motion Control Rod Drive 3.9-2 l .
3.9.1.4.1.2 Hydraulic Control Unit 3.9-2 3.9.1.4.2 Reactor Pressure Vessel Assembly 3.9-2 3.9.1.43 Core Support Structures and Other Safety Reactor Internal Components 3.9-2 3.9.1.4.4 RPV Stabilizer and FMCRD and Incore Housing Restraints (Supports) 3.92 3.9.1.4.5 Main Steam Isolation Valve, Safety / Relief Valve and Other ASME Class 1 Valves 3.9-2 3.9.1.4.6 ECCS and SLC Pumps, RRS and RHR Heat Exchangers, RCIC Turbine and RRS Motor - 3.9-2 3.9.1.4.7 Fuel Storage and Refueling Equipment 3.9-3 3.9.1.4.8 Fuel Assembly (Including Channel) 3.9 3 l
l' l
O 3 .9 11 Amendment 1 L
U ._ . _: _.
- j l
ABWR m.ims Etandard Plant ma SECTION 3.9 CONTENTS (Continued)
O j Section M East I
3.9.1.4.9 ASME Class 2 and 3 Vessels 3.9-3 l 3.9.1.4.10 ASME Class 2 and 3 Pumps 3.9-3 3.9.1.4.11 ASME Class 2 and 3 Valves 3.9-3 3.9.1.4.12 ASME Class 1,2 and 3 Piping 3.93 !
3.9.1.5 Inelastic Analysis Methods 3.9-3 3.9.2 Dynamic Testing and Analysis 3.93.1 3.9.2.1 Piping Vibration, Thermal Expansion, and I Dynamic Effects 3.9-3.1 l
3.9.2.1.1 Preoperational and Initial Startup Vibration 3.9-4 and Dynamic Effects 3.9.2.1.1.1 Measurement Techniques 3.9-4 91 3.9.2.1.1.2 Monitoring Requirement 3.9-4 3.9.2.1.13 Test Evaluation and Acceptance 3.95 h 4 Criteria for Main Steam Piping "
3.9.2.1.1.4 Reconciliation and Corrective Actions 3.95 3.9.2.1.2 Thermal Expansion Testing 3.9-6 3.9.2.1.2.1 Measurement Techniques 3.9-6 ,
3.9.2.1.2.2 Monitoring Requirements 3.9-6 3.9.2.1.23 Test Evaluation and Acceptance Criteria 3.9-6 3.9.2.1.2.4 Reconciliation and Corrective Actions 3.9-7 3.9.2.2 Seismic Qualification of Safety-Related Mechanical Equipment (Including Other RBV Induced Loads 3.9-9 3 .9 111 Amcodment 11 4
. 23A6100AE 1
, Riandard Plant REV.A SECTION3.9 c.; .
'(J CONTENTS (Continued)
Sec11on Iille East l l
l 3.9.2.2.1 Tests and Analysis Criteria and Methods 3.9-9 3.9.2.2.1.1 Random Vibration loput 3.9 10 3.9.2.2.1.2 Application ofinput Modes 3.9-10 3.9.2.2.13 Fixture Design 3.9-10 3.9.2.2.1.4 Prototype Testing 3.9-10 3.9.2.2.2 Qualification of Safety-Related Mechanical Equipment 3.9-10 3.9.2.2.2.1 CRD and CRD Housing 3.9-10 r 3.9.2.2.2.2 Core Support (Fuel Support and CR Guide Tube) 3.9 11 3.9.2.2.23 Hydraulic Control Unit (HCU) 3.9 11 3.9.2.2.2.4 Fuel Assembly (Including Channel) 3.9 11 3.9.2.2.2.5 Reactor Internal Pump and Motor Assembly 3.9-11 3.9.2.2.2.6 ECCS Pump and Motor Assembly 3.9 11 3.9.2.2.2.7 RCIC Pump and Turbine Assembly 3.9-11 3.9.2.2.2.8 Standby Liquid Control Pump and Motor Assembly 3.9-12 3.9.2.2.2.9 RMC and RHR Heat Exchangers 3.9 12 3.9.2.2.2.10 Standby Liquid ControlTank - 3.9-12 3.9.2.2.2.11 Main Stcam Isolation Valves 3.9-12 3.9.2.2.2.12 Standby Liquid Control Valve (Injection Valve) 3.9 12 3.9.2.2.2.13 Main Steam Safety / Relief Valves 3.9-12 l
l 3.9.2.2.2.14 Fuel Pool Cooling and Cleanup System Pump and Motor Assembly 3.9 12 l
k 3.9iv Amendment 1 l
o L. 2%6100AE . .
< StandanI Plant nev. m '
i SECTION 3.9 a
W CONTENTS (Continued)
Sectlon M East l
l l 3.9.2.2.2.15 Other ASMEIII Equipment 3.9-13 3.9.2.2.2.16 Supports 3.9 14 ,
3.9.23 Dynamic Response of Reactor Internals Under Operational Flow Trevients and Steady-State .
Conditions 3.9-14 i 3.9.2.4 Preoperational Flow Induced Vibration Testing of Reactor Internals 3.9-16 3.9.2.5 Dynamic System Analysis of Reactor Internals l Under Faulted Conditions 3.9 17 3.9.2.6 Correlations of Reactor Internals Vibration Tests With the Analytical Results 3.9-17.2 3.93 ASME Code Class 1. 2. and 3 Comnonents.
Comnonent Sunnorts. and Core Sunnort Structures 3.9-18 3.93.1 Loading Combinations, Design Transients, and c Stress Limits 3.9-18 3.93.1.1 Plant Conditions 3.9-18.1 3.93.1.1.1 Normal Condition 3.9-18.1 3.93.1.1.2 Upset Condition 3.9-18 3.93.1.13 Emergency Condition 3.9-19 3.93.1.1.4 Faul ed Condition 3.9-19 3.93.1.1.5 Correlation of Plant Condition with Event Probability 3.9 19 3.93.1.1.6 Safety Class Functional Criteria 3.9-19 3.93.1.2 Reactor Pressure Vessel Assembly 3.9-20 3.93.13 Main Stcam (MS) System Piping 3.9-20 0
3.9-v Amendment 11
MM saA6100AE
- Senndmed Plant arv.m-SECHON 3.9 m-
-iQ CONTENTS (Continued) .,_
Section Iltle Eagt 3.9.5.2 loading Conditions 3.9-41 3S.5.2.1 Events to be Evaluated 3.9-41 3.9.52.2 Pressure Differential During Rapid Depressurization 3.9-41 3.9.5.23 Feedwater line and Main Steam Une Break 3.9-42 3S.5.23.1 Accident Definition 3.9-42 3.9.5.23.2 Effects ofInitial Reactor Power and Core Flow 3.9-42 3.9.5.2.4 Seismic and Other Reactor Building Vibration Events 3.9 42 3.9.53 Design Bases 3.9-43 3.9.53.1 Safety Design Bases 3.9-43 4 3S.53.2 Power Generation Design Bases 3.9-43 3.9.533 Design Loading Categories 3.9 43 3S.53.4 Response of Internals Due to Steam Line Brer.k Accident 3.9-43 3.9.53.5 Stress and Fatigue Limits for Core Support Structures 3.9-43 3.9.53.6 Stress, Deformation, and Fatigue Limits for Safety Class Reactor Internals (Except Core -
Support Structures) 3.9-44 3.9.6 Inservice Testine of Pumns and Vaives 3.9-44 t
3.9.6.1 Inservice Testing of Pumps 3.9-44
~
$ 3.9.6.2 Inservice Testing of Valves 3.9-44 M
3.9.7 Interfaces 3.9-45 3.9.7.1 Reactor Internals Vibration Analysis, 3.9-45 Measurement and Inspection Programs 3.9.7.2 ASME Class 2 or 3 or Quality Group 3.9-45 Components with 60 Year Life l 3.9.8 Peterences 33 45 3.9-x Amendment il 1
A 21A6100AE Rtandard Plant uw. s SECTION 3.9 i
TABLES l 1
1 M
IAbit EASc 3.9 1 Plant Events 3.9-46 3.9-2 Imd Combinations and Acceptance Criteria for Safety Related, ASME Code Class 1,2 and 3 Components, Component Supports, and Class CS Structures 3.9-49 -
3.9-3 Pressure Differentials Across Reactor Vessel Internals 3.9-53 ;
3.94 Deformation Limit, for Safety Class Reactor -I Internal Structures Only 3.9-54' ,
3.9-5 Primary Stress Limit, for Safety Class Reactor Internal Structures Only 3.9-55 l 4
3.9-6 Buckling Stability Limit, for Safety Class ReactorInternal Structures Only 3.9-57 3.9-7 Fatigue Limit, for Safety Class Reactor l Internal Structures Only 3.9-58 ILLUSTRATIONS l
Figure Iltle Eagc 1
1 3.9-la Transient Pressure Differential Following l a Steam Line Break 3.9-58.1 ;
3.9-Ib Reactor Vessel and Internal Assembly 3.9-59 -
3.9-2 Reactor Internal Flow Paths and Minimum Floodable Volume 3.9-60 -l l
l- 3.0-3 ABWR Recirculation Flow Path 3.9 61 3.9-4 Fuel Support Pieces 3 9-62 l i- )
l- 3.9-5 Pressure Nodes for Depressurization Analysis 3.9-63 j 3.9-6 Stress-Strain Curve for Blowout Restraints 3.9 64 l l
O l l- 3.9-xi Amendment 11 1
4
. ABM 234sioo^s Standard Plant' Rev. n analyzed for the faulted loading conditions. The 33.1A.10 ASME Class 2 and 3 Pumps ECCS and SLC pumps are active ASME Class 2 compo-nents. The allowable stresses for active pumps Elastic analysis methods are used for evaluat-are provided in a footnote to Table 3.9 2. ing faulted loading conditions for Class 2 and 3 pumps. The equivalent allowable stresses for The reactor coolant pressure boundary compo- nonactive pumps using clastic techniques are ob- !
nents of the reactor recirculation system (RRS) tained from NC/ND-3400 of the ASME Code Section i pump motor assembly, and recirculation motor cocl. III. These allowables are above clastic lim-ing (RMC) subsystem heat exchanger are ASME Class its. Tbc allowables for active pumps are pro-1 and Class 3, respectively, and are analyzed for vided in a footnote to Table 3.9 2.
the faulted loading conditions. All equipment stresses are within the clastic limits. 33.1A.11 ASME Class 2 and 3 Valves 3.9.1.4.7 Fuel Storage and Refueling Equipment Elastic analysis methods and standard design rules are used for evaluating faulted loading i Storage, refueling, and servicing equipment conditions for Chss 2, and 3 valves. The which is important to safety is classified as es. equivalent allowable stresses for nonactive sential components par the requirements of valves using clastic techniques are obtained -
10CFR50 Appendix A. This equipment and other from NC/ND 3500 of ASME Code,Section III.
equipment which in case of a fa9ere would de- These allowables are above clastic limits. The grade an essential component is defined in Sec- allowables for active valves are provided in a tion 9.1 and is classified as Seismic Category footnote to Table 3.9 2.
I. These components are subjected to an clastic dynamic finite element analysis to generatt, load- 33.1A.12 ASME Class 1,2 and 3 Piping ings. This analysis utilizes appropriate floor response spectra and combines loads at freqzen. Elastic analysis methods are used for evaluat-cies up to 33 Hz for seismic loads and up to 60 ing faulted loading conditions for Class 1,2, . 1
.. Hz for other dynamic loads in three directions, and 3 piping. The equivalent allowable stresses I
. Imposed stresses are generated and combined for using clastic techniques are obtained from Appen-normal, upset, and faulted conditions. Stresses dix F (for Class 1) and NC/ND 3600 (for Class 2 are compared, depending on the specific safety and 3 piping) of the ASME Code Section III. ;
class of the equipment, to Industrial Codes, These allowables are above clastic limits. The ASME, ANSI or Industrial Standards, AlSC, allowables for functional capability of the es-allowables, sential piping are provided in a footnote to Table 3.9 2.
3.9.1A.8 Fuel Assembly (Including Channel) 33.1.5 Inelastic Analysis Methods GE BWR fuel assembly (including channel) de-sign bases, and analytical and evaluation methods Inelastic analysis is only applied to ABWR including those applicable to the faulted condi. components to demonstrate the acceptability of tions are the same as those contained in Refer. three types of postulated events. Each event is ences 1 and 2. an extermly low probability occurence and the equipment affected by ther.e events would not be 3.9.1 A3 ASME Class 2 and 3 Vessels reused. These three events are:
Elastic analysis methods are used for enluat- (1) Postulated gross piping failure.
ing faulted loading conditions for Class 2 and 3 vessels. The equivalent allowable stresses using (2) Postulated blowout of a reactor internal clastic techniques are obtained from NC/ND 3300 recirculation (RIP) motor casing due to a and NC-3200 of the ASME Code Section III. These weld failure.
allowables are above clastic limits.
(3) Postulated blowout of a control rod drive (CRD) housing due to a weld failure.
.O Amendment it 3.9-3
ABWR numu -
RennAard Plant an a The loading combinations and design criteria 3.9.2.1 Piplag Vibratloa. Thermal Espansion, for pipe whip restraints utilized to mitigate the and Dynamic Effects effects of postulated piping failures are provided in Subsection 3.6.2.3.3. The overall test program is divided into 3
two phases; the preoperational test phase and in the case of the RIP motor casing failure the initial startup test phase. Piping vibra-event, there are specific restraints applied to tion, thermal expansion and dynamic effects test- h mitigate the effects of the failure. The ing will be performed during both of these mitigation arrangement consists of lugs on the phases as described in Chapter 14. Subsections RPV bottom head to which are attached two long 14.2.12.1.51,14.2.12.2.10 and 14.2.12.2.11 re.
rodt for each RIP. The lower end of each rod late the specific role of this testing to the ov-engages two lugs on the RIP motor / cover. Tbc use erall test program. Discussed below are the gen-of inelastic analysis methods is lipiited to the eral requirements for this testing, it middle slender body of the rod itself. The attachment lugs, bolts and clevises are shown to be adequate by clastic analysis. The selection of stainless steel for the rod is based on its high ductility assumed for energy absorption during inelastic deformation.
The mitigation for the CRD housing attachment weld failure is by somewhat different means than are those of the RIP in that the components with regular functions also function to mitigate the weld failure effect. The components are specifically:
(1) Core support plate (2) Control rod guide tube (3) Control rod drive housing (4) Control rod drive outer tube (5) Bayonet fingers Only the cyli drical bodies of the control rod guide tube, cc strol rod drive housing and control rod drive outer tube are analyzed for energy absorption by inelastic deformation.
Inelastic analysis for there latter two events together with the criteria used for evaluation are consistent with the procedures described in Subsection 3.6.2.3.3 for the different components of a pipe whip restraint.
Figure 3.9 6 shows the stress strain curve used for the blowout restraints.
3.9.2 Dynamic Testing and Analysis O
Amendment 11 3.9-3.1
- - .-- .~. -- . - .- -- _. - - - - -
. ABn
' Standard Plant
. 24sioors m v. s should be noted that because one goal of the dy- to more rigorous testing and precise lastrumenta-n namic effects testing is to verify the adequacy tion requirements and, therefore, will require Q
1 of the piping support system, such components are remote monitoring techniques. Local measurement addressed in the subsections that follow. How. techniquess such as the use of a hand held ever, the more specific requirements for the de- vibrometer, are more appropriate in cases where sign and testing of the piping support system are it is expected that the vibration will be less described in Subsection 3.9.3.4.1. complex and of lessor magnitude. - Many systems that are assessable during the preoperational 3.9.21.1 Vibration and Dynamic Effects Testing test phase and that do not show significant intersystem interactions will fall into this cat- .,,
The purpose of these tests is to confirm egory.- Visual observations are utilized where 2 that the piping, components, restraints and sup. vibration is expected to be minimal and the need A ports of specified high and moderate energy sys- for a time history record of transient behavior tems have been designed to withstand the dynamic is not anticipated. However, unexpected visual effects of steady state flow induced vibration observations or local indications may require and anticipated operational transient condi- that a more sophisticated technique be used, tions. The general requirements for vibration Also, the issue of assessability should be con-and dynamic effects testing of piping systems are sidered. Application of these measurement tech-specified in R e gulatory G uid e 1.68, niques is detailed in the appropriate testing
- Preoperational and Initial Startup Test Programs specification consistent with the guidelines con-for Water Cooled Power Reactors'. More specific tained in ANSI /ASME OM3.
vibration testing requirements are defined in ANS1/ASME OM3," Requirements for Preoperational 3.9.2.1.1.2 Monitoring Requirements and Initial Startup Vibration Testing of Nuclear Power Plant Piping Systems'. Preparation of As described in Subacetion 14.2.12.1.51, detailed test specifications will be in full 14.2.12.2.10 and 14.2.12.2.11 all safety related accordance with this standard and will address piping systems will be subjected to steady state such issues as prerequisites, test conditions, and transient vibration measurements. The :, cope g precautions, measurement techniques, monitoring of such testing shall include safety related in- 8 g requirements, test hold points and acceptance strumentation piping and attached small-bore pip- R criteria. The development and specification of ing (branch piping). Special attention should the types of measurements required, the systems be given to piping attached to pumps, compres-and locations to be monitored, the test sors, and other rotating or reciprocating equip-acceptance criteria, and the corrective actions ment. Monitoring location selection consider-that may be necessary are discussed in more ations should include the proximity of isolation detail below. valves, pressure or flow control valves, flow orifices, distribution headers, pumps and other ;;
l' 3.9.2.1.1.1 Measurement Techniques elements where shock or high turbulence may be $
of concern. Location and orientation ofinstru-There are essentially three methods avail- mentation and/or measurements will be detailed able for determining the acceptability of steady in the appropriate test specification.
state and transient vibration for the affected Monitored data should include actual deflections j systems. These three measurement techniques are and frequencies as well as related system operat-visual observation, local measurements, or re- ing conditions. Time duration of data recording motely monitored / recorded measurements. The tech- should be sufficient to indicate whether the vi.
l nique used in each case will depend on such fac- bration is continuous or transient. Steady tors as the safety significance of the particular state monitoring should be performed at critical system, the expected mode and/or magnitude of the conditions such as minimum or maximum flow, or vibration, the assessability of the system during abnormal combinations or configurations of I designated testing conditions, or the need for a system pumps or valves. Transient monitoring !
time history recording of the vibratory behav- should include anticipated system and total i lor. Typically, the systems where vibration has plant operational transients where critical I the greatest safety implication will be subject piping or components are expected to show Amendment 11 394 l
- ~ .__ _ - . - - . - -. - - - - - . - - - . _ _
10
'MM EtanAmrd Plant 21A6100AE an a r
- q. significant response. Steady state conditions During the course of the tests, the remote ;
and transient events to be monitored will be measurements will be regularly checked to verify 1
"G -
detailed in the appropriate testing specification compliance with acceptance criteria. If trends
! consistent with OM3 guidelines. indicate that criteria may be violated, the men-surements should be monitored at more frequent l 3.9.2.1.1.3 Test Evaluation and Acceptance Crite- intervals. . The test will be held or terminated i ria as soon as criteria are violated. As soon as L possible after the test hold or termination ap-The piping response to test conditions propriate investigative and corrective actions shall be considered acceptable if the review of will be taken, if practicable, a walkdown of the test results indicates that the piping re- the piping and suspension system should be made sponds in a manner consistent with predictions of in an attempt to identify potential obstructions ,
the stress report and/or that piping stresses are or improperly operating suspension components.
within ASME Code Section III (NB 36000) limits. Hangers and snubbers should be positioned such ,
Acceptable limits are determined after the comple- that they can accommodate the expected deflec- i tion of piping systems stress analysis and are tions without bottoming out or extending fully, i provided in the piping test specifications. All signs of damage to piping supports or an-chors shall be investigated. -
l To ensure test data integrity and test l safety, criteria have been established to fa- Instrumentation indicating criteria failure cilitate assessment of the test while it is in shall be checked for proper operation and progress. For steady state and transient vibra- calibration including comparison with other in-l tion the pertinent acceptance criteria are usu. strumentation located in the proximity of the ex-ally expressed in terms of maximum allowable dis- cessive vibration. The assumptions used in the placement / deflection. Visual observation should calculations that generated the applicable lim. @
only be used to confirm the absence of sig- its should be verified against actual conditions A
- (/
O nificant levels of vibration and not to determine acceptability of any potentially excessive vibra-and discrepancies noted should be accounted for in the criteria limits. This may require a
@ tion. Therefore,in some cases other measurement reanalysis at actual system conditions.
" techniques will be required with appropriate quan-titative acceptance criteria. Should the investigt. tion of instrumentation and calculations fail to reconcile the criteria There are typically two levels of acceptance violations, then physical corrective actions may
,. criteria for allowable vibration displace- be required. This might include identification i ments/ deflections. Level I criteria are bounding and reduction or elimination of offending fore-type criteria associated with safety limits while ing functions, detuning of resonant piping spans Level 2 criteria are stricter criteria associated by appropriate modifications, addition of brac-with system or component expectations. For ing, or changes in operating procedures to avoid steady state vibration the Level I criteria are troublesome conditions. Any such modifications based on the endurance limit (10,000 psi) to as- will require retest to verify vibrations .have l
sure no failure from fatigue over the life of the been sufficiently reduced.
I plant. The corresponding Level 2 criteria arr.
I based on one half the endurance limit (5,000 39.2.L2 'Ihermal Expansion Testing psi). For transient vibration the Level 1 crite-o ria are based on either the ASME III code upset A thettaal expansion preoperational and startup primary stress limit or the applicable snubber testing program performed through the use of vi-load capacity. Level 2 criteria are based on a sual observation and remote sensors has been given tolerance about the expected deflection established to verify that normal unrestrained value. thermal movement occurs in specified safety-related high and moderate energy piping sys-3.9.2.1.1.4 Reconcillation and Corrective Ac- tems. The purpose of this program is to ensure tions the following:
Amendment 3
- 9-5 l
l .- .. -- .- . - . . _ __
t 23A6100AE Standard Plant arv. s (1) the piping system durihg system heatup and measurements, using a hand held scale or ruler, cooldown is free to expand and move without against a fixed reference or by recording the unplanned obstruction or restraint in the x, position of a snubber or spring can. A more y, and z directions; precise method would be using permanent or temporary instrumentation that directly measures (2) the piping system does shakedown after a few displacement, such as a lanyard potentiometer, thermal expansion cycles; that can be monitored via a remote indicator or recording device. The technique to be used will (3)- the piping system is working in a manner con- depend on such factors as the amount of movement sistent with the assumptican of the stress predicted and the ===bility of the piping.
analysis; Measurement of piping temperature is also of (4) there is adequate agreement between calcu- importance when evaluating thermal expansion, lated values and measured values of displace. This may be accomplished either indirectly via ments; and the temperature of the process fluid or by .
direct measurement of the piping wall tem-(5) there is consistency and repeatability in perature and such measurements may be obtained thermal displacements during heatup and either locally or remotely. The choice of tech-cooldown of the systems. nique use'd shall deper.? on such considerations as the accuracy require,d and the assessability The general requirements for thermal expan- of the piping.
sion testing of piping systems are specified in Regulatory Guide 1.68, "Preoperational and Ini- 3.9.2.1.2.2. Monitoring Requimnents p; l y tial Startup Testing Programs for Water Cooled g l ;; Power Reactors." More specific requirements are As described in Subsections 14.2.12.1.51
, defined in ANSI /ASME OM7 ' Requirements for and 14.2.12.2.10 all safety-related piping shall Thermal Expansion Testing of Nuclear Power Plant be included in the thermal expansion testing pro- _
l Piping Systems." Detailed test specifications gram. Thermal expansion of specified piping sys.
l will be prepared in full accordance with this tems should be mea.;ured at both the cold and hot standard and will address such issues as extremes of their expected operating condi-prerequisites, test conditions, precautions, tions. Physical walkdowns and recording of measurement techniques, monitoring requirements, hanger and snubber positions should also be con-test hold points and acceptance criteria. The ducted where possible considering assessability development and specification of the types of and local environmental and radiological condi-measurements required, the systems and locations tiens in the hot and cold states. Displacements to be monitored, the test acceptance criteria, and appropriate piping / process temperatures and the corrective actions that may be necessary shall be recorded for those systems and condi-are discussed in more detail below, tions specified. Sufficient time shall have passed before taking such measurements to ensure l 3.9.2.1.2.1 Measurement Techniques the piping system is at a steady state condi-tion. In selecting locations for monitoring pip-Verification of acceptable thermal expan- ing response, consideration shall be given to sion of.specified ' piping systems can be accom- the maximum responses predicted by the piping plished by several methods. One method is to analysis. Specific consideration should also be physically walkdown the piping system and verify given to the first run of pipe attached to l by visual observation that free thermal movement component nozzles and pipe adjacent to is unrestrained. This might include verification structures requiring a controlled gap.
that piping supports such as snubbers and spring hangers are not fully extended or bottomed out 3.9.2.1.2.3 Test Evaluation and Acceptance Crl-and that the piping (including branch lines and teria instrument lines) and its insulation is not in hard contact with other piping or support To ensure test data integrity and test structures. Another methed would involve local safety, criteria have been established to fa-
]- Amendment tt 3.9-6 i
[
c1 , ABM '
siw u, Simadard Plant an a 3.9.2.2 Seismic QumHReatles of )
-- 1 . Safety.Related Macha= leal Equipment ;
L ._ (laciudlag Other RBVInduced Imads) l This subsection describes the criteria for dynamic qualification of safety.related snechanical equipment and associated supports, and also describes the qualification testing .
and/or analysis applicable to the snajor components on a component by component basis.
Seismic and other events that may induce reactor building vibration (RBV).(see Appendix 3B) are considered. In some cases, a module or assembly consisting of mechanical and
, electrical equipment is qualified as a unit (e.g., ECCS pumps). These modules are generally discussed in this subsection and Subsection 3.9.3.2 rather than providing discussion of the ' separate electrical parts in Section 3.10. Electrical supporting equipment such as control consoles, cabinets, and panels
{ are discussed in Section 3.10.
3.9.2.2.1 Tests and Analysis Criteria and Methods The ability of equipment to perform its safety function during and after the application of a dynamic load is demonstrated !
by tests and/or analysis. The analysis is i performed in accordance with Section 3.7.
Selection of Testing, analysis or a i combination of the two is determined by the type, size, shape, and compluity of the -
equipment being considered. When practical,'
the equipment operability is demonstrated by testing. Otherwise, operability is demonstrated by mathematical analysis.
Equipment which is large, simple, and/cr consumes large amounts of power is usually qualified by analysis or static bend test to show that the loads, stresses and deflections .
are less than the allowable maximum. Analysis and/or static bend testing is also used'to show there are no natural frequencies below 33 Hz for seismic loads and 60 Hz for other RBV
- The 60 H2 frequency cutoff for dynamic loads'. If a natural frequency lower than 33 analysis of suppression pool dynamic loads is the Hz in the case of seismic loads and 60 Hz in minimum requirement based on a generic Reference 8, using the missing strain energy method,
_ performed for representative BWR equipment under high-frequency input loadings.
Amendment 11 3.99
l MM 23A6100AE '
- Remndard Plant MVA I
the case of other RBV induced loads is ~ (3) the input has sufficient intensity and 1 discovered, dynamic tests and/or mathematical duration to excite all modes to the I analyses may be used to verify operability and required magnitude so that the ter, ting l structural integrity at the required dynamic response spectra will envelop the Input conditions. l corresponding response spectra of the :
Cadividual modes. )
When the equipment is qualified by dynamic test, the response spectrum or time history of 3.9.24.1.2 Application ofloput Modes a the attachment point is used in determining input I motion. When dynamic tests are performed, the )
. _ input motion is applied to one vertical and l Natural frequency may be determined by running one horizontal axis simultaneously. However, '
a continuous sweep frequency search using a- if the equipment response along the vertical sinusoidal steady state input of low magnitude. direction is not sensitive to the vibratory Dynamic load conditions are simulated by testing motion along the horizontal direction and vice using random vibration input or single frequency versa, then the input motion is applied to one input (within equipment capability) over the direction at a time. In the case of single ,
frequency range ofinterest. Whichever method is frequency input, the time phasing of the l iused, the input amplitude during testing inputs in the vertical and horizontal
! envelopes the actual input amplitude expected directions are such that a purely rectilinear during the dynamic loading condition. resultant input is avoided.
The equipment being dynamically tested is 3.9.2.2.1.3 Fixture Design mounted on a fixture which simulates the intended service mounting and causes no dynamic coupling The fixture design simulates the actual to the equipment, service mounting and causes no dynamic coupling to the equipment.
Equipment having an extended structure, such as a valve operator, is analyzed by applying 3.9.2.2.1 A Prototype Testing static equivalent dynamic loads at the center of l gravity of the extended structure. In cases Equipment testing is conducted on l where the equipment structural complexity makes. prototypes of the equipment to be installed in l
mathematical analysis impractical, a static bend the plant.
test is used to determine spring constant and operational capability at maximum equivalent 3.9.2.2.2 Qualification of Safety.Related dynamic load conditions. Mechanical Equipment 3.9.2.2.1.1 Random Vibration loput The following subsections discuss the testing or analytical qualification of the When random vibration input is used, the safety.related major mechanical equipment, and actual input motion envelopes the appropriate other ASME III equipment, including equipment floor input motion at the individual modes, supports.
However, single frequency input such as sine beats can be use provided one of the following 3.9.2.2.2.1 CRD and CRD Housing conditions are met:
The qualification of the CRD housing (with (1) the characteristics of the required input enclosed CRD) is done analytically, and the motion is dominated by one frequency; stress results of their analysis establish the l
structural integrity of these components.
(2) the anticipated response of the equiprnent is Preliminary dynamic tests are conducted to adequately represented by one mode; or verify the operability of the control rod O
Amendment 1 3.9-10 l
-= . MM 23A6100AE Standard Plant ny a pletion of preoperational' testing, the reaclor 3.9.2J Dynamic System Analysis of Reactor (N- vessel head and the shroud head are removed, the laternals Under Faulted Conditions
( vessel is drained, and major' components arc inspectrd on a selected basis. The inspections The faulted events that are evaluated are cover the shroud, shroud head, core support defined in Subsection 3.9.5.2.1. The loads
, structures, recirculation internal pumps, the that occur as a result of these events and the peripheral control rod drive, and incore guide analysis performed to determine the response tubes. Access is provided to the reactor lower of the reactor internals are as follows:
plenum for these inspections.
(1) Reactor Internal Pressures The reactor The analysis, design and/or equipment that are internal pressure differentials (Figure to be utilized in a facility will comply with 3.9 la) due to assumed break of main steam Regulatory Guide 1.20 as explained below, or feedwater line are determined by analysis as described in Subsection Regulatory Guide 1.20 describes a 3.9.5.2.2. In order to assure that no comprehensive vibration assessment program for significant dynamic amplification ofload reactor internals during preoperational and occurs as a result of the oscillatory initial stattup testing. The vibration nature of the blowdown forces during an assessment program meets the requirements of accident, a comparison is made of the Criterion 1, Quality Standards and Record, periods of the applied forces and the Appendix A to 10CFR50 and Section 50.34, Contents natura! periods of tbe core support of Applications; Technical Information, of structures being acted upon by the applied i
10CFR50. This Regulatory Guide is applicable to forces. These periods are determined t
the core support structures and other reactor from a comprehensive vertical dynamic ,
internals. model of the RPV and internals with 12 g degrees of freedom. Besides the real A
, Vibration testing of reactor internals is masses of the RPV and core support
(' performed on all GE-BWR plants. At the time of structures, account is made for the water original issue of Regulatory Guide 1.20, test inside the RPV.
programs for compliance were instituted for the then designed reactors. The first ABWR plant is (2) External Pressure and Forces on the considered a prototype and is instrumented and Reactor Vessel An assumed break of the subjected to preoperation and startup flow main steam line, the feedwater line or the testing to demonstrate that flow induced RHR line at the reactor vessel nozzle vibrations similar to those expected during results in jet reaction and impingement operation will not cause damage. Subsequent forces on the vessel and asymmetrical
- plants which have internals similar to those of pressurization of the annulus between the the prototypes are also tested in compliance with reactor vessel and the shield wall.
the requirements of Regulatory Guide 1.20. GE is These time varying pressures are applied committed to confirm satisfactory vibration to the dynamic model of the reactor vessel performance of internals in these plants through system. Except for the nature and preoperational. flow testing followed by locations of the forcing functions, the l inspection for evidence of excessive vibration. dynamic model and the dynamic analysis Extensive vibracion measurements in prototype method are identical to those for seismic plants together with satisfactory operating analysis as described below. The experience in all BWR plants have established the resulting loads on the reactor internals, adequacy of reactor internal designs. GE defined as LOCA loads, are considered as continues these test programs for the generic shown in Table 3.9.2.
plants to verify structural integrity and to establish the margin of safety. (3) Safety / Relief Valve leads (SRV I4 ads) The ;
i discharge of the SRVs result in reactor See Subsection 3.9.7.1 for interface building vibration (RBV) due to l
l requirements of the reactor internals vibration suppression pool dynamics as described in
{- testing program. Appendix 3B. The response of the reactor Amendment Il 3.9 17
1 MM 23A6100AE Elandard Plant an a :
internals to the RBV is also determined with the reactor and internals are performed. The dynamic model and dynamic analysis method results of these analyses are used to generate '
described below for seismic analysis, the allowable vibration levels during the vibration test. The vibration data obtained "
-(4) 14CA lmeds The Assumed 1DCA also results in during the test will be analyzed in detail.
RBV due to suppression pool dynamics as described in Appendix 3B and the response of the reactor internals are again determined with the dynamic model and dynamic analysis 1 method used for seismic analysis. .Various types of LOCA loads are identified on Table 3.9 2. .
(5) Seismic Leads The theory, methods, and computer codes used for dynamic analysis of the reactor vessel, internals, attached piping and. adjoining structures are described in Section 3.7 and Subsection 3.9.1.2. Dynamic analysis is performed by coupling the lumped mass model of the L reactor vessel and internals with the l building model to determine the system 1-m natural frequencies and mode shapes. The 2 relative displacement, acceleration, and
- load response is then determined by either the time-history method or the resonse spectrum method. The load on the reactor internals due to faulted event SSE are obtained from this analysis.
The above loads are considered in combination
. as defined in Table 3.9 2. The SRV. LOCA (SBL, IBL or LBL) and SSE loads as defined in Table 3.9 2 are all assumed to act in the same direction. The peak colinear responses of the reactor internals to each of these loads are added by the square root of the sum.of the squares (SRSS) method. The resultant stresses in the reactor internal structures are directly added with stress resulting from the static and steady state loads in the faulted load combination, including the stress due to peak reactor internal pressure differential during the LOCA. The reactor internals satisfy the stress
. deformation and fatigue limits as defined in Subsection 3.9.5.3.
3.9.2.6 Correlations of Reactor laternals Vibration Tests With the Analytical Results Prior to initiation of the instrumented vibration measurement program for the prototype plant, extensive dynamic analyses of Amendment 8 3.9 17.1
_ _ . _ . _ _ _ . - _ _ ._. __ _ . ~ ____ _ _ _ _
, , MM 2&A610QAE EleanA=ed Plant arv a l The result'. of the data analyses, vibration 3.9 2 and are co %Ined in the design n amplitudes, natural frequencies, and mode shapes specifications and, >r design reports of the I
V wre then compared to those obtained from the respective equipment. (See Subsection 3.9.7.3 Georetical analysis. for interface requirements) l Such comparisons provide the analysts with Table 3.9 2 also presents the evaluation added insight into the dynamic behavior of the models and criteria. The predicted loads or i reactor laternals. The additional knowledge stresses and the design or allowable values '
gained from previous vibration tests has been for the most critical areas of each component !
sillized in the generation of the dynamic models are compared in accordance with the applicable ,
for seismic and loss of coolant accident (LOCA) code criteria or other limiting criteria. The analyses for this plant. The models used for calculated results meet the limits.
this plant are similar to those used for the vibration analysis of earlier prototype BWR The design life for the ABWR Standard plants. Plant is 60 years. A 60 year design life is a ,
requirement for all major plant components i 3.9.3 ASME Code Class 1,2, and 3 with reasonable expection of meeting this Components, Component Supports, and design life. However, all plant operational Core Support Structures components and equipment except the reactor vessel are designed to be replaceable, design 3.9.3.1 landing Combinations, Design life not withstanding. The design life Tronalents.nnd Stress Limits requirement allows for refurbihment and repair, as appropriate, to assure the design This section delineates the criteria for life of the overall plant is achieved. In selr% and definition of design limits and effect, essentially all piping systems, ~
loading combination associated with normal components and equipment are designed for a 60
< peration, postulate.4 accidents, and specified year design life. Many of these components ,
s tismic and other reactor , building vibration are classified as ASME Class 2 or 3 or Quality k (PBV) events for the design of safety.related Group D. Applicants referencing the ABWR ;
~
AWE Code components (except containment design willidentify these ASME Class 2,3 and compnents which are discussed in Section 3.8). Quality Group D components and provide the analyses required by the ASME Code, Subsection This section discusses the ASME Class 1,2, NB. These analysl= will include the and 3 equipment and associated pressure retaining appropriate operating & ration loads and for parts and identifies the applicable loadings, the effects of mixing hot and cold fluids, calculation methods, calculated stresses, and allowable stresses. A discussion of major 3.9.3.1.1 Plant Conditions equipment is included on a component.by component :
basis to provide examples. Design transients and All events that the plant will or might dynamic loading for ASME Class 1,2, and 3 credibly experience during a reactor year are equipment are covered in Subsection 3.9.1.1. evaluated to establish design basis for plant Seismic.related loads and dynamic analyses are equipment. These events are divided into four discussed in Section 3.7. The suppression plant conditions. The plant conditions pool related RBV loads are described in Appendix described in the following paragraphs are 3B Table 3.9 2 presents the cou.binations of based on event probability (i.e., frequency of dynamic events to be consideM for the design occurrence as discussed in Subsection and analysis of all ABWR ASME Code Class 1,2, 3.9.3.1.1.5) and correlated to service levels and 3 components, component supports, core for design limits defined in the ASME Boiler support structures and equipment. Specific and Pressure Vessel Code Section III as shown loading combinatkra considered for evaluation of in Tables 3.91 and 3.9 2. I each specific equ!pment are derived from Table l
Amendment 11 3.9 18 l-1.. . - . . _ - - - -- . . . - _ -, - - , - . - . . . . _ _
ABWR nwma .
Ramadmed Plant ny n 3.93.1.1.1 Normal Condition Normal conditions are any conditions in the course of system vtartup, operation in the design 9l i j
power range, normal hot standby (with condenser j available), and system shutdown other than upset, ;
emergency, faulted, or testing. l 3.93.1.1.2 Upact Condition An upset condition is any deviation from normal conditions anticipated to occur often i enough that design should include a capability to .
withstand the conditions without operational ;
impairment. The upset conditions include systr.m operational transients (SOT) which result from any single operator error or control malfunction, from a fault in a system component requiring its isolation from the system, from a loss of load or ;
power, or from an operating basis carthquake. '
Hot standby with the main condenser inclated is an upset condition.
O t
O Amendment 11 3A18.1 6
- ABWR n-s
..a.a m... m.
3.9.3.4 Component Supports correspond to those used for design of the sup-G ,
ported pipe. The component loading The design of bolts for component supports combinations are discussed in Subsection is specified in the ASME Code Section III, 3.9.3.1. The stress limits are per ASME lli, Subsection NF. Stress limits for bolts are given Subsection NF and Appendia F. Supports are in NF 3225. The rules and stress limits which generally designed either by load rating must be satisfied are those given in NF 3324.6 method per paragraph NF 3260 or by the stress multiplied by the appropriate stress limit factor limits for linear supports per paragraph for the particular service loading level and NF 3231. The critical buckling loads for the stress category specified in Table NF 3225.21. Class 1 piping supports subjected to Paulted loads that are more severe than normal, upset Moreover, on equipment which is to be, or and emergency loads, ard determined by using may be, mounted on a concrete support, sufficient the methods discussed in Appendices F and XVil holes for anchor bolts are provided to limit the of the Code. To avoid buckling in the piping anchor bolt stress to less than 10,000 psi on the supports, the allowable loads are limited to nominal bolt area in shear or tension, two thirds of the determined critical buckling loads.
Concrete anchor bolts which are used for pipe support base plates will be designed to the The design of all supports for non nuclear applicable factors of safety which are defined in piping satisfies the requirements of ANSI I&E Bulletin 79 02,' Pipe Support Base Plate B31.1, Paragraphs 120 and 121.
Designs Using Concrete Expansion Anchor Bolts,'
Revision 1 dated June 21,1979. For the major active valves identified in Subsection 3.9.3.2.4, the valve operators are 3.9.3.4.1 Piping not used as attachment points for piping supports.
Supports and their attachments for essential
. ASME Code Section !!!, Class 1,2, and 3 piping The design criteria and dynamic testing re.
late designed in accordance with Subsection NF' up quirements for the ASME 111 piping supports to the interface of the building structure. The are as follows:
h building structure component supports are de.
", signed in accordance with the AISC specification (1) Piping Supports All piping supports are for the Design, Fabrication, and Erection of designed, fabricated, and assembled so Structural Steel for buildings. The loading com- that they cannot become disengaged by the binations for the various operating conditions movement of the supported pipe of equipment after they have been installed.
All piping supports are designed in accordance with the rules of Subsection NF of the ASME Code up to the building
- structure interface as defined in the
' Augmented by the following: (1) application of project design specifications.
Code Case N 476, Supplement 89.1 which governs the design of single angle members os ASME Class (2) Spring Hangers The operating load on 1,2,3 and MC linear component supports; and (2) spring hangers is the load caused by dead when eccentric loads or other torsional loads are weight. The hangers are calibrated to en-not accommodated by designing the load to act sure that they support the operating load through the shear center or meet ' Standard for at both their hot and cold load settings.
Steel Support Design *, analyses will be performed Spring hangers provide a specified down in accordance with torsional analysis methods travel and up travel in excess of the such as:
- Torsional Analysis of Steel Members, specified thermal movement.
USS Steel Manual *, Publication T114 2/83 or 9 *etc. Design of Weld Structures
- by Omar W. Blodgett, Amendment 11 3.9 31 ;
1
MM 31A6100AE '
RennAmed Ple=# mn (3) Snubbers The operating loads on snubbers are the loads caused by dynamic events &
(e.g., seismic, RBV due to LOCA and SRV discharge, discharge through a relief W
valve line or valve closure) during various operating conditions. Snubbers restrain piping against response to the vi.
bratory excitation and to the associated differential movement of the piping system support anchor points. The criteria for locating snubbers and ensuring adequate load capacity, the structural and me.
chanical performance parameters used for snubbers and the lastallation and inspec.
tion consider. ations for the snubbers are as follows:
(a) Required Load Capacity and Snubber Lo-cation The entire piping system including valves and support system between an.
chor points is mathematically modeled for complete piping structural analysis. In the dynamic analysis, the snubbers are modeled as a spring with a given spring stiffness depending on the snubber size. The analysis determines the forces and moments acting on each piping components and the forces acting on the snubbers due to all dynamic loading and operating conditions defined in the piping design specification. The forces on snub.
bers are operating loads for various l
operating conditions. Those loads are assumed no to exceed the snubber design load capacity for various operating conditions, i.e., design, normal, upset, emergency and faulted, i
l-O Amendment 11 3.9-31.1 l
l I~
. . - M..Ma.e... _
2sA61ooAt mu Saubbers are generally used in agreement, they are brought in situations where dynamic support is agreement, and the system analysis -
required because thermal growth of the is redone to confirm the snubber O. piping prohibits the use of rigid loads. This iteration is continued supports. The snubber locations and until all snubber load capacities support directions are first decided by and spring coastants are estimation so that the stresses in the r e c o n cile d.
piping system will have acceptable values. The snubber locations and (c) Snubber Design and Testing support directions are refined by performing the dynamic analysis of the To assure that the required piping and support system as described struetura1 and mechan!ca1 above in order that the piping stresses performance characteristics and and support loads meet the Code product quality are achieved, the requirements, following requirements for design and testing are imposed by the The pipe support design specification design specification:
requires that snubbers be provided with position indicators to identify the rod (i) The snubbers are required by position. This indicator facilitates the pipe support design the checking of hot and cold settings of specification to be designed the snubber, as specified in the in accordance with all of the installation manual, during plant rules and regulations of the preoperational and startup testing. ASME Code Section 111 Subsection NF. This design (b) Inspection. Testing, Repair and/or requirement includes analysis Replacement of Snubbers for the norms), upset, emergency, and faulted The pipe support design specification loads. These calculated requires that the snubber supplier loads are then compared prepare an installation instruction against the allowable loads manual. This manual is required to to make sure that the contain complete instructions for the stresses are below the code testing, maintenance, and repair of the allowable limit, snubber. It also contains inspection points and the period of inspection. (ii) The snubbers are tested to insure that they can perform The pipe support design specification as required during the requires that hydraulic anubbers be seismic and other RBV events.
equipped with a fluid level indicator so and under anticipated that the level of fluid in the snubber operational transient loads can be ascertained easily. or other mechanical loads associated with the design The spring constant achieved by the requiren. cats for the plant, snubber supplier for a given load The f o11o win g test capacity snubber is compared against the requirements are included:
spring constant used in the piping system model. If the spring constants o Snubbers are subjected to are the same, then the snubber location force or displacement versus and support direction become confirmed. time loading at frequencies If the spring constants are not in wit hin the range of O
Amendment 1 3.9 32
. . ABWR -
Rtandard Plant aw n significant modes of the piping (i) There are no visible signs of i
system; damage or impaired operability as a result of o Displacements are measured to st or a ge, handling, or determine the performance installation, characteristics specified; (ii) The snubbet location, o Tests are condt.cted at various orientation, position temperatures to ensure operability setting, and configuration over the specified range; (attachments, extensions, etc.) are according to design o Peak test loads in both tension and drawings and specifications.
compression are required to be equal to or higher than the rated load (iii) Snubbers are not seized, requirements; and frozen or jammed.
o The snubbers are tested for various (iv) Adequate swing clearance is abnormal environmental conditions. provided to allow snubber ,
Upon completion of the abnormal movements, environmental transient test, the snubber is tested dynamically at a (v) If applicable, fluid is to be frequeacy within a specificd recommended level and not be frequency range. The snubber must leaking from the snubber operate normally during the dynamic system.
test.
(si) Structural connections such (d) Snubber Installation Requirements as pins, fasteners and other O -
An installation instruction manual is connecting hardware such as lock nuts, tabs, wire, cotter required by the pipe support design pins are installed correctly, specification. This manual is required to contain instructions for storage, if the period between the handling, erection, and adjustments (if initia1 pre serviee necessary) of snubbers. Each snubber examination and initial has an installation location drawing system pre operational tests which contains the installation location exceeds 6 months because of of the snubber on the pipe and unexpected situations, structure, the hot and cold settings, reexamination of Items 1,4, and additional information needed to and 5 will be performed, install the particular snubber. Snubbers which are installed
~
incorrectly or otherwise fail (c) Snubber Pre-service Examination to meet the a b o '. e requirements will be repaired The pre service examination plan of all or replaced and re examined snubbers covered by the Chapter 16 tech. in accordance with the above nical specifications will be prepared. criteria.
This examination will be made after snubber installation but not more than 6 (4) Struts - The design load on struts months prior to initial system pre oper- includes those loads caused by dead ational testing. The pre service weight, thermal expansion, seismic forces examination will verify the following: (i.e., OBE and SSE), other RBV loads, O
Amendment 7 3.9 33 c
I ABWR n-c - '
mandard Plant arv n system anchor displacements, and reaction (P/Perit) + (4/4 crit) + (f /f erit) forces caused by relief valve discharge or valve closure, etc. < (1/S.F.)
Struts are designed in accordance with ASME where:
Code Section III, Subsection NF 3000 to be capable of carrying the design loads for q = longitudinalload various operating conditions. As in case of P = external pressure snubbers, the forces on struts are obtained r = transverse shear stress from an analysis, which are assured not to S.F. = safety factor exceed the desig.n loads for various = 3.0 for design, testing, service operating conditions, levels A & S
= 2.0 for Service Level C .
3.93.4.2 Reactor Pressure Vessel Support Skirt = 1.5 for Service Level D.
The ABWR RPV support skirt is designed as an 3.9.3.4.3 Reactor Pasasure Vessel Stabillaer ASME Code Class I component per the requirements of ASME Code Section III, Subsection NF*. The The RPV stabilizer is designed as a Safety loading conditions and stress criteria are given Class I linear type component support in in Tables 3.91 and 3.9 2, and the calculated accordance with the requirements of ASME stresses meet the Code allowable stresses in the Boiler and Pressure Vessel Code Section III, critical support areas for various plant Subsection NF. The stabilizer provides a operating conditions. The stress level margins reaction point near the upper end of the RPV assure the adequacy of the RPV support skirt. An to resist horizoutal loads due to effects such analysis for buckling shows that the support as earthquake, pipe rupture and RBV. The skirt complies with Subparagraph F 1332.5 of ASME design loading conditions, and stress criteria Ill, Appendix F, and the loads do not exceed two are given in Tables 3.91 and 3.9 2, and the thirds of the critical buckling strength of the calculated stresses rocet the Code allowable skirt. The permissible skirt loads at any stresses in the critical support areas for elevation, when simultaneously applied, are various plant operating conditions, limited by the following interaction equation:
3.9.3.4.4 Floor. Mounted Major Equipment (Pumps, Heat Exchangers, and RCIC Turbine)
Since the major active valves are supported by piping and not tied to building structures, valve ' supports
- do not exist (See Subsection 3.9.3.4.1).
The HPCF, RHR, RCIC, SLC, FPCCU,
' Augmented by the following: (1) application of SPCU, and RWCU pumps; RMC, RHR, Code Case N.476, Supplement 89.1 which governs RWCU, and FPCCU beat exchangers;and RCIC the design of single angle members of ASME Class turbine are all analyzed to verify the 1,2,3 and MC linear component supports; and (2) adequacy of their support structure under when eccentric loads or other torsional loads are various plant operating conditions. In all not accommodated by designing the load to act cases, the load stresses in the critical through the shear center or meet " Standard for support areas are within ASME Code allowables.
Steel Support Design *, analyses will be performed in accordance with torsional analysis methods Seismic Category I active pump supports are such as: ' Torsional Analysis of Steel Members, qualified for dynamic (seismic and other RBV)
USS Steel Manual", Publication T114 2/83 or loads by testing when the pu:np supports
" Design of Weld Structures
- by Omar W. Blodgett, etc.
Amendment 11 3.9 34
. - i ABWR msme standard Plant ma together with the pump meet the following test conditions:
(1) almulate actual mounting conditions; (2) simulate all static and dynamic loading on the pump; (3) monitor pump operability during testing; (4) the normal operation of the pump during and after the test indicates that the supports are adequate (any deflection or .
deformation of the pump supports which precludes the operability of the pump is not accepted); and (5) supports are inspected for structuralin-tegrity after the test. Any cracking or permanent deformation is not accepted.
Dynamic qualification of component supports by analysis is generally accomplished as fol.
lows:
O r
O Amendment 11 3Mt.1
ABWR . ms t Blandard Plant an n (1) Stresses at all support elements and parts such as pump holddown and baasplate holddown O- bolts, pump support pads, pump pedestal, and foundation are checked to be within the al-lowable limits as specified in the ASME Code Session !!!, Subsection NF.
(2) For normal and upset conditions, the defle.ctions and deformations of the supports are assured to be within the elastic limits, and to not exceed the values permitted by the designer based on design verification tests.
This ensures the operability of the pump.
(3) For emergency and faulted plant conditions, the deformations do not exceed the values permitted by the designer to ensure the I operability of the pump. Elastic / plastic I analysis are performed if the deflections are above the clastic limits.
3.9.3.6 Other ASME Ill Component Supports The ASME Ill component supports and their at.
tachments (other than those discussed in preced.
ing subsection) are designed in accordance with Subsection NF of the ASME Code Section !!!* up to
.g the laterface with the building structure. The E building structure component supports are de.
signed in accordance with the AISC Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings. The inading combinations for the various operating conditions 3.9.4 ControlRod Drive System (CRDS) correspond to those used to design the supported component. The component loading combinations A control rod drive system CRDS)in an ABWR are discussed in Subsection 3.9.3.1. Active plant is equipped with an electro hydraulic fine component supports are discussed in Subsection motion control rod drive (FMCRD) system, which 3.9.3.2. The stress limits are per ASME !!!, includes the control rod drive (CRD) mechanism, Subsection NF and Appendix F. The supports are the hydraulic control unit (HCU), the condensate evaluated for buckling in accordance with ASME supply system, and power for FMCRD motor, and Ill. extends inside RPV to the coup!!ng laterface with the control rod blades.
' Augmented by the following: (1) application of Code Case N 476, Supplement 89.1 which governs the 3.9.4.1 Descriptiva Information on CRDS design of single angle members of ASME Class 1,2,3 and MC linear component supports; and (2) when Descriptive information on the CRDs as well eccendric loads or other torsionalloads are not as the entire control and drive system is con.
accommodated by designing the load to act through tained in Section 4.6.
the shear center or meet ' Standard for Steel Support Design *, analyses will be performed in 3.9.4.2 Applicable CRDS Design Specification accordance with torsional analysis methods such as: ' Torsional Analysis of Steel Members, USS CRDS is designed to meet the functional de-iO Steel Manual *, Publication T114 2/83 or ' Design of sign criteria cutlined in Section 4.6 and con.
Weld Structures
- by Omar W. Blodgett, etc.
Amendment 11 3.9 35
f MM 21A6100AE me..Amed Plant RIY B sists of the following: (2) factory quality controltests; (1) Ane motion control rod drive; (3) Five year maintenance life tests; (2) bydtmulic control unit; (4) 1.5X design life tests; (3) bydraulic power supply (pumps); (5) operationaltests; (4) electric power supply (for FMCRD motors) (6) acceptance tests; and (5) laterconnectingpiping; (7) surveillance tests.
(6) Dow and pressure and isolation valves; and All of the tests except (3) and (4) are dis.
cussed in Section 4.6. A discussion of tests (7) lastrumentation and electrical controls. (3) and (4) follows:
Those components of the CRDS forming part of (3) Five Year Maintenance Life Tests Four the primary pressure boundary are designed control rod drives are normally picked at according to ASME Code Section lit, Class 1 random from the production stock each year requirements, and subjected to various tests under simu.
lated reactor conditions and 1/6th of the The quality group classification of the service life cycles.
components of the. CRDS is outlined in Table 3.21 and they are designed to the codes and standards, Upon completion of the test program, per Table 3.2 2, in accordance with their control rod drives must meet or surpass individual quality groups. the minimum specified performance requirements.
Pertinent aspects of the design and qualifica.
tion of the CRDS components are discussed in the (4) 1.5X Design Life Tests When a signifi-following locations: transients in Subsection leant design change is made to the com.
3.9.1.1, faulted conditions in Subsection ponents of the drive, the drive is sub.
3.9.1.4, seismic testing in Subsection 3.9.2.2. jected to a series of tests equivalent to 1.5 times the service life cycles.
3.9A.3 Design Loads, Stress Limits, and Allowable Deformations 3.9.5 ReactorPressurtVesselInternals The ASME lli Code. components of the CRDS have This subsection identifies and discusses the been evaluated an@tically and the design load. structural and functionalintegrity of the major ing conditions, and stress criteria are as given reactor pressure vessel (RPV) internals, includ-in Tables 3.91 and 3.9 2, and the calculated ing core support structures.
stresses meet the Code allowable stresses. For the non Code components, the ASME 111 Code re. 3.9.5.1 Design Arrangements q'airements are used as guidelines and experimen.
tal testing is used to determine the CRD perfor. The core support structures and reactor mance under all possible conditions as described vessel internals (exclusive of fuel, control in Subsection 3.9.4.4. rods, and incore nuclear instrutnentation) are:
3.9A.4 CRD Performance Assurance Program (1) Core Support Structures The CRD test program consists of these tests: Shroud; (1) dm.lopment tests; Shroud support (including the internal pump deck);
Amendment 3 3.9-36
4 . MM 2M6100AE !
Senndard Plant an s i driven from naderneath by a pump shaft, with the 3.9.5.1.2.4 Feedwater Spargers O impeller being encircled by a diffuser shroud h assembled into the pump deck opening. These are Safety Class 2 components. They l are discussed here to describe coolant flow l
l The RM section of the RIP is located paths in the vessel and their safety !
underneath, and at the periphery of, the RPV function. Each of two feedwater lines is !
bottom head inside a pressure retaining housing connected to three spargers via three RPV termed the motor casing. The motor casing itself nozzles. One line is utilized by the RCIC is not part of the RM, but is instead a part of system; the other by the RHR shutdown cooling and welded into ao RPV nozzle (pump nozzle). The system. During the ECCS mode, the two groups motor casing thus comprises part of the reactor of spargers support diverse type of flooding coolant pressure boundary and is a Safety Class 1 of the vessel. The RCIC system side supports component, high pressure flooding and the RHR system side supports low pressure flooding, as required The principal element of the stretch tube during the ECCS operation, section is a thin walled tube configured as a hcIlow bolt fitting around the pump shaft and The feedwater spargers are stainless steel within the pump nozzle. It has an externallip headers located in the mixing plenum above the ,
(belt head) at its 9pper end and an external downcomer annulus. A separate sparger in.two threaded section at this lower end. The stretch halves is fitted to each feedwater nozz!c via tube function is to achieve tight clamping of the a tee and is shaped to conform to the curve of IP diffuser to the gasketed, internal mount end the vessel wall. The sparger tee inlet is of the RPV pump nozzle, at all extremes of connected to the RPV nozzle safe end by a
- thermal transients and pump operating conditions, double thermal sleeve arrangement, with all connections made by full penetration welds.
l 3.9.5.1.23 Steam Dryer Assembly Sparger end brackets are pinned to vessel brackets to support the spargers. Feedwater Tbc steam dryer assembly.is a non. safety class flow enters the center of the spargers and is component. it is discussed here to describe discharged radially inward to mix the cooler coolant flow paths in the vessel. The steam feedwater with the downcomer flow from the dryer removes moisture from the wet steam leaving steam separators and steam dryer before it the steam separators. The extracted moisture contacts the vessel wall. The feed. water flows down the dryer vanes to the collecting also serves to condense steam in the region troughs, then flows through tubes into the above the downcomer annulus and to subcool downcomer annulus. water flowing to the recirculation internal pumps.
The steam dryer assembly consists of multiple l
banks of dryer units mounted on a common 3.9.5.1.23 RHR/ECCS tow Pressurv Flooder structure which is removable from the reactor Sparsers pressure vessel as an integral unit. The assembly includes the dryer banks, dryer supply These are Safety Class 2 components. The l and discharge ducting, drain collecting trough, design festeres of these two spargers of the drain piping, and a skirt which forms a water RHR shutdown cooling system are similar to seal extending below the separator reference zero those of the six feedwater spargers, three of elevation. Upward and radial movement of the which belonging to one feedwater line support dryer assembly under the action of blowdown and additionally the same RHR (and ECCS) seismic loads are limited by reactor vessel function. During the ECCS mode, these internal stops which are arranged to permit dif. spargers support low pressure flooding of the ferential expansion growth of the dryer assembly vessel. The feedwater spargers are described i with respect to the reactor pressure vessel. The in Subsection 3.9.5.1.2.4.
l assembly is arranged for removal from the vessel as an integral unit on a routine basis. Two lines of RHR shutdown cooling system enter the reactor vessel through the two
\ dias,onally opposite nozzles and connect to the Amendment 11 3.9 39
ABM 2m6ioors - -
1- standard Plant arv. s spargers. The sparger tee inlet is connected to connection to the steamline is blocked. When the RPV nozzle safe end by a thermal sleeve draining the vessel during shutdown, air enters
, arrangement with all connections made by full the vessel through the vent.
penetration welds.
33J.1.2.8 Com and laternal Pump 333.1.2.6 ECCS High Preaaun Con Flooder Differvatial Preasun Unes Sparsers and Piping These lines comprise the core flow measure.
4 The core flooder spargers and piping are ment subsystem of the recirculation flow control I Safety Class 2. The spargers and piping are the system (RFCS) and provide two methods of measur.
means for directing high pressure ECCS flow to ing the ABWR core flow rates. The core DP lines the upper end of the core during accident (Safety Class 3) and internal pump DP lines conditions. (non safety class) enter the reactor vessel se.
parately through rence bottom head penetra.
Each of two high pressure cose flooder (HPCF) tions. Four pairs of the core DP lines enter system lines enters the reactor vessel through a the head in four quadrants through four penetra-diagonally opposite nozzle in the same manner as tions and terminate immediately above and below an RHR low pressure flooder line, except that the the core plate to sense the pressure in the re-curved sparger including the connecting tee is gion outside the bottom of the fuel assemblies routed around the inside of and is supported by and below the core plate during normal the cylindrical portion of the top guide. A operation, flexible coupling is interposed between the sparge: tee inlet and the sleeved inlet connector Similarly, four pairs of the internal pump DP
- inside the nozzle. The two spargers are lines terminate above and below the pump deck supported so as to accommodate thermal expansion, and are used to sense the pressure across the pump during normal pump operation. Each pair is 33.5.1.2.7 RPV Vent and Head Spray Assembly routed concentrically through a penetration and upward along a shroud support leg in the lower This is designed as a Safety Class 1 plenum, component. However, only the nozzle portion of the assembly is a reactor coolant pressure 33.5.1.23 in Core Guide Tubes and boundary, and the assembly function is not a Stabilliers safety related operation. The reactor water cleanup return flow to the reactor vessel, via These are Safety Class 3 components. The feedwater lines, can be diverted partly to a guide tubes protect the in core instrumentation spray nozzle in the reactor head in preparation from flow of water in the bottom head plenum and for refueling cooldown. The spray maintains provide a means of positioning fixed detectors saturated conditions in the reactor vessel head in the core as well as a path for insertion and volume by condensing stream being generated by withdrawal of the calibration monitors (ATIP, the hot reactor vessel walls and internals, The automated traversing incore probe subsystem).
head spray subsystem is designed to rapidly The in core flux monitor guide tubes extend from cooldown the reactor vessel head flange region the top of the in core flux monitor housing to for refueling and to allow installation of steam the top of the core plate. The power range de-line plugs before vessel floodup for refueling. tectors for the power range monitoring units and the dry tubes for the startup range neutron The head vent side of the assembly passes monitoring and average power range monitoring steam and noncondensable gases from the reactor (SRNM/APRM) detectors are inserted through the head to the steamlines during startup and opera. guide tubes.
tion. During shutdown and filling for hydro-testing, steam and noncondensable gases may be Two levels of stainless steel stabilizer vented to the drywell equipment sump while the latticework of clamps, tie bars, and spacers
. give lateral support and rigidity to the guide Amendment il 3.9 40
a- . k 21A6100AE Standard Plant REV.B 3.9.7 Interfaces 3.9.7.3 Audit of Design Specification and Design Reports O 3.9.7.1 Reactor Intemals Vibration Analysis, Measurement and laspection Program Applicants referencing the ABWR design will make available to the NRC staff Design The first applicant refereneing the ABWR Specifications and Design Reports required by design will provide, at the time of application, Be ASME Code for vessels, pumps. valves and the results of the vibration assessment program piping systems for the purpose of audit.
for the ABWR prototype internals. These results will include the following information specified 3.9.8 References in Regulatory Guide 1.20.
- 1. BWR Fuel Channel Mechanical Design an.d R G.1.20 $nbitu Deflection, NEDE 21354 P, September 1973.
C.2.1 Vibration Ana1ysis Program 2. BWR/6 fuel Assembly Evaluation of Combined C.2.2 Vibration Measurement Program Safe Shutdown Earthquake (SSE) and C.2.3 luspection Program Loss of Coolant Accident (LOCA) Loadings, C.2.4 Docurnemation of Results NEDE 21175 P November 1976.
NRC review and approval of the above 3. NEDE 24057 P (Class Ill) and NEDE 24057 information on the first applicants docket will (Class 1) Assessment of Reactor Internals, complete the vibration assessment program Vibration in BWR/4 and BWR/S Plants, requirements for prototype reactor internals. November 1977. Also NEDO-24057 P, Amendment 1, December 1978, and NEDE 2 P 24057 In addition to the information tabulated Amendment 2. June 1979.
above, the first applicant referencing the ABWR design will provide the information on the 4. General Electric Company, Analytical Model schedules in accordance with the applicable for Loss of Coolant Analysis in Accordance O- portions of position C 3 of Regulatory Guide 1.20 with 10CFR50, Appendix K, NEDE 20566P, for non. prototype internals. Proprietary Document, November 1975.
Subsequent applicants need only provide the 3. BWR reedwater Noule and Control Rod Drive informt. tion on the schedules in accordance with Retum Line Nonle Cracking, NUREG 0619.
the applicable portions of position C.3 of Regulatory Guide 1.20 for non prototype 6. General Electric En vironm ental internals. (See Subsection 3.9.2.4 for interface Qualification Program, NEDE 243261 P, requirements). Proprietary Document, January 1983.
3.9.7.2 ASME Class 2 0r 3 or Quality Group 7. Functional Capability Criteria for Components with 60 Year Design Life Essential Mark 11 Piping, NEDO 21985, September 1978, prepared by Battelle Applicants referencing the ABWR design will Columbus Lkboratories for General Electric identify ASME Class 2 or 3 or Quality Group D Company.
components that are subjected to loading,s which could result in thermal or dynamic fatigue and 8. Generic Criteria for High Frequency Cutoff provide the analyses required by the ASME Code, of BWR Equipment, NEDO 25250, Proprietary Subsection NB. These analyses will include the Document, January 1980, appropriate operating vibration loads and for the effects of mixing hot and cold fluids. (See Subsection 3.9.3.1 for interface requirements).
O Amendment 11 3.9-45
2mioore
.A.B Mpi...
.a..a in. .
Table 3.9 7 I FATIGUE LIMIT FOR SAFETY CLASS REACIDR INTERNAL STRUCTURES ONLY i
Summation of fatigue damage usage following Minor hypotheses (8):
\
limit for Service '
Levels A&B (Normal Cutnnlative Damaae in Fatione and Unset rnnditinnd Design fatigue cycle usage fro;a analysis .51.0 using the method of the ASME Code HQIE (1) Miner, M.A., Cumulative Damage in fatigue, Journal of Annlied Mechanies. Vol.
12, ASME, Vol. 67, pp A159-A164, September 1945.
O
+
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Amendment 1 3.9-58
. .- , _ , . . . _ . - ~ - - - ~ ~ .. - - - ..
3146100AE Riandmed Plant uv. n l
. I Table 3.94 h REACIVR COOIANT SYSTEM PRESSURE ISOLATION VALVES I I
STANDBY UQUID CON 11tOL n' STEM 1
C41.F006 A,B lajeaion Valves C41.F008
)
laboard Check Valve j g
RESIDUAL NEAT REMOVAL SYSIT.M
- I E11.F005 A,B,C lajection Valve loops A,B&C l E11.F006 A,B,C Testable Check Valve A,B&C l E11 F010 A,B,C Shutdown Coohng Inboard Suetion Isolation i Valve Loops A,B&C E11.F011 A,B,C Shutdown Coohng Outboard Suction Isolation Valve Loops A,B&C l S !
HIGH FRESSURE CORE FLDODER SYSTEM N l
E22 F003 B,C Injection Valve 1. cops B&C E1.2 F004 B,C Testable Check Valve loops B&C REACTOR CORE ISOLATION COOLING SYSTEM E51 F004 E51 F005 Injection Valve Testable Check Valve G1 i '
1 i
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Amendment 11 3.9 %
. ABWR Standard Plant mamt .
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PolNT P AT INTIR$tCTION POWER LAW CURVE ,
OF THE CURV! OPYU BY $=K!" CONNECTING POINT $ Y AND U ELA$ tic LINI OP $ AND $ FROM A$MI !!!,
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Figure 3.9-6 STRESS-STRAIN CURVE FOR BLOWOUT RESTRAINTS :
Amendment 11 3.9-68
. Standard ABM Plant 21A6100AE mm. n Table 3G 4-1 m
ig Site Envelope Horizontal OBE loads for R/B and RCCV I
X AXIS (RB 01110 DEG)
ELEY REACTOR BUIDING RCCV
{hQ SHEAR (TONS) MOMENT (T-Mi SHEAR (TONS) MOMENT (T M) i l
44.7 1.93E + 04 3.95E+ 03 33.2 7.67E + 04 4 8.29E+ 03 26.7 1.46E+ 05 5.44E + 04 l 135E+ 04 5.28E + 03 18.5 2.73E+ 05 8.70E+ 04 1.56E+ 04 1.47E+ 04 13.1 3.66E + 05 1.0SE + 05 I
2.10E + 04 1.63E + 04 73 4.84E+ 05 1.78E+ 05 2.10E + 04 1.63E+ 04
-0.2 4.84E+ 05 2.13E + 05 2.10E+ 04 1.63E+ 04
-6.7 4.84E + 05 232E+ 05 2.10E + 04 1.63E + 04 13.2 5.64E + 05 2.95E+ 05
[
ELEY REACTOR BUIDING Y AXIS (RB o0 270 DEG)
RCCV (hD SHEAR FRONS) MOMENT (T-M) SHEAR (TONS) MOMENT (T M) 44.7 9.53E + 03 3.93E + 03 33.2 9.11E + 04 7.55E + 03 26.7 1.79E+ 05 4.44E + 04 1.28E + 04 3.11E + 03 18.5 2.94E+ 05 1.12E + 05 1.72E + 04 7.80E + 03 13.1 3.79E + 05 1.82E+ 05 2.11E+ 04 9.73E + 03 73 4.82E+ 05 2.61E + 05 2.11E+ 04 9.73E + 03 0.2 5.09E + 05 3.06E+ 05 2.23E + 04 1.04E + 04
-6.7 5.62E+ 05 332E+05 239E+04 1.11E+ 04 13.2 637E+05 3.51E + 05 Notes:
- 1. Elevations are relative to the RPV bottom head.
- 2. Forces on the RB between EL 33.2M and 18.5M along the x. axis are the sum of maximum forces of the two sticks representing the walls as shown in Fig. 3G.2-1.
.O Amendment 4 30.4-2
- ABWR 234.ioors Standard Plant wn Table 3G.4 2 g Site Envelope OBE Loads for Selected Locations BEAM MAXIMUM FORCE MAXIMUM MOMENT LOCATION M F.[D E YY,CD MXrrMi MYrf Mi Shroud Sup't 28 98.0 263.8 286.0 1830.0 2281.0 RPV Skirt 69 577.0 7803 614.0 5322.1 3909.9 RSW Base 78 458.0 1044.0 853.0 5064.0 4168.0 1 Prdestal Base 86 2027.0 3343.0 2806.0 72077.0 58671.0 Notes:
- 1. P is verticalload due to vertical excitation
- 2. VX and MX are shcar and moment due to HOR X excitation
- 3. VY and MY are shear and moment due to HOR Y excitation l
9
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i I
O' Amendment 11 3G43
. ABWR msms Standard Plant Ru A APPENDIX 19B TABLE OF CONTENTS Sasunn 310: hat 19B RESOLUTION OF APPLICABLE UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES 19B.1 INTRODUCrlON 19B.11
'F -
19B.1.1 Purpose 19B.11 19B.1.2 Summary 19B.1 1 19B.1.3 References 19B.1 1 19B.2 ISSUE STATUS AND ABWR RESOLUTION 19B.21 19B.2.1 Quality and Reliability Assurance 19B.21 19B.2.2 In Situ Testing of Valves 19B.2 2 19B.2.3 Systems lateraction 19B.2 3 19B.2.4 Nuclear Fower Plant Vulnerability to Sabotage 19B.2 5 19B.2.5 Control System Failures 19B.218 19B.2.6 Hydrogen Control 19B.2 22 19B.2.7 Safety Related Operator Action 19B.2 25 19B.2.8 LWR Fuel 19B.2 26 19B.2.9 Effectiveness of Ultimate Heat Sinks 19B.2-28 19B.2.10 Service Water System Reliability 19B.2 29 19B.2.11 Equipment Protective Devices on Engineered Safety Features 19B.2 36 19B.2.12 Bolting Degradation or Failure 19B.2-38 19B.2.13 Design for ATWS Event 19B.2-39 19B.2.14 Design of the Spent Fuel Pool 19B.2-42 19B.ii Amendment 11
' ABWR zwms -
Standard Plant mm. 4 APPENDIX 19B >
3 TABLE OF CONTENTS (Continued)
Suur.a B11: tam 19B.2.15 High/Im Pressure Interface Design 19B.2 43 19B.2.16 Design of ABWR Water levellastrumentation 19B.2-45 19B.2.17 Probable Maximum Precipitation 19B.2-46 1
19B.2.18 Design of Main Transformer 19B.2-47 ;
19B.2.19 Accident Management 19B.2-48 19B.2.20 - Allowable Equipment Outage Times 19B.2 53
~19B.2.21 - Tendon Anchorage Design 19B.2 54 19B.2.22 Single Failure Criterion 19B.2 54 l
19B.2.23 Testing and Maintenance of Manual Valves e in Safety Related Systems 19B.2 55 19B.2.24 Electrical Power Reliability 19B.2 55 19B.2.25 Man Machine Interface 19B.2 56 19B.2.26 Maintenance and Surveil!ance Program 19B.2-60 19B.2.27 Seismic Design Criteria 19B.2 60 19B.2.28 Station Blackout 19B.2 60.1 19B.2.29 Shutdown Decay Heat Removal 19B.2-60.1 19B.2.30 Slabs, Shells and Steel Containments Ductility of Two way Slabs and Shc!!s and Buckling Behavoir of Steel Contain-ments 19B.2-60.2 19B.231 Control Room Habitability 19B.2 60.2 19B.3 INTERFACES 19B.21 <
19B.3.1 Ouality Assurance Progrun 19B.31 19B3.2 Prevention of Core Damage 19B.3-1 19B.111 Amendment 11 T
. , _ - _ ,_c
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I ABWR . nsamis !
Standard Plant hvA q APPENDIX 19B 7-TABLE OF CONTENTS (Continued)
Assion nie East ii 19B33 Protection from ExternalThreats 19B31 19B3.4 Ultimate Heat Sink Models 19B3-1
]
1 198 3.5 Ultimmt Heat Sink Reliability 19831 ]
1 198 3.6 Main Transformer Design 19B3-1 i i
19B3.7 Plant Siting 19B3-1 19B3.8 Interdisciplinary Design Reviews 19B3-1 5
3 0 .
1 l
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1 19B-iv l,
,O Amendment 11 l
- MM 23A6100AS Standard Plant Rn A SECTION 19B.1 GE PROPRIETARY . provided under separate cover East Ammadment 19B.1 1 11 19B.1 2 11 19B.12.1 11 19B.13 11 19B.14 11 0
Amendment 11 19B.1 1
ABWR was Standard Plant an. A Q APPENDIX 19B.2 TABLE OF CONTENTS Bstion Iltle East 195.2.1 Ouality and Reliability Assurance 19B.21 19B.2.2 In-Situ Testing of Valves 19;3.2-2 19B.2.3 Systems Interaction 19B.2 3 19B.2.4 Nuclear Power Plant Vulnerability to Sabotane 19B.2 5 19B.2.5 Control System Failures 19B.218 19B.2.6 Hydrogen Control 19B.2 22 19B.2.7 Safety Related Onerator Action 19B.2 25 19B.2.8 LWR Fuel 19B2 26 19B.2.9 Effectiveness of Ultimate Heat Sinks 19B.2 28 19B.2.10 Service Water System Rellabilltv 19B.2 29 Eaulnment Protective Devices on Ennineered O.
19B.2.11 Safety Features 19B.2 36 19B.2.12 Boltine Deeradation or Failure 19B.2 38 19B.2.13 Desinn for ATWS Event 19B.2 39 19B.2.14 Denien of the Snent Fuel Pool 19B.2 42 19B.2.15 Hich/ Low Pressure Interface Desinn 19B.2-43 19B.2.16 Deslan of ABWRWaterlsvelInstrumentation 19B.2-45 19B.2.17 Probable Maximum Precinitation 19B.2-46 19B.2.18 Desinn of Main Transformer 19B.2-47 19B.2.19 Accident Mananement 19B.2 48 19B.2.20 Allomble Eouloment Outane Times 19B.2 53 19B.2.21 Tendon Anchorane Desinn 19B.2-54 19B.2-il O Amendment 7
- ABWR 2 mms Standard Plant Rev.A SECTION 19B.2 g_
CONTENTS (Continued) secuna nic East 198.2.22 Sinale Failure Criterion 19B.2 54 19B.2.23 Testina and Maintenance of Manual Valves la Safety Related Systems 19B.2-55 19B.2.24 Electrical Power Reliability 19B.2 55 19B.2.25 Man-Machine Interface 19B.2 56 19B.2.26 Malatenance and Surveillance Praarna 19B.2-60 19B.2.27 Seismic Deslan Criteria 19B.2-60 19B.2.28 Station Blackout 19B.2 60.1 19B.2.29 Shutdown Decay Heat Removal 19B.2-60.1 19B.2.30 Slabs. Shells and Steel Containment Ductility of Two-way Stabs and Shells and Buckline Behavior of Steel Containments 19B.2-60.2 Control Room Habitability 19B.2 31 19B.2 60.2 TABLES Table Iule East 19B.2-1 Plant Interface Functions for Site Security System 19B.2-61 ILLUSTRATIONS Firure IRlt Eagt 19B.21 Ilhetration of Security Zone Terminology 19B.2-62 19B.2-lii Amendment 11 i
4
- . ABWR 23^62oo^s Standard Plant am ^
SECTION 19B.2
< GE PROPRIETARY . provided under separate cover fagg Amendment Eagt Amendment l 19B.2-1 11 19B.2-32 7
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19B.2 2 7 19B.2-33 7 19B.2 3 11 19B.2 34 7 19B.2 4 11 19B.2 35 7 19B,2-5 11 19B.2 36 7 19B.2 6 7 19B.2 37 7 19B.2 7 7 19B.2 38 7 19B.2 8 7 19B.2 39 - 7 19B.2 9 7 19B.2 40 7 19B.2-10 7 19B.2-41 7 19B.211 7 , 19B.2 42 11 19B.212 7 19B.2-43 11 19B.213 7 19B.2 44 7 19B.214 7 19B.2-45 11 19B.215 7 19B.2 46 11 19B.216 7 19B.2-47 7 19B.217 11 19B.2 48 7 19B.218 11 19B.2-49 7 a 19B.2-19 11 19B.2 50 7 19B.2 20 11 19B.2 51 7 19B.2 21 11 19B.2 52 7 19B.2-22 11 19B.2 53 7 19B.2 23 11 19B.2 54 7 19B.2-24 11 19B.2 55 7 19B.2 25 11 19B.2 56 7 19B.2 26 7 19B.2 57 7 19B.2-27 7 19B.2 58 11 l 19B.2 28 7 19B.2 59 7 19B.2 29 7 19B.2-60 11 19B.2-30 7 19B.2-60.1 11 19B.2 31 7 19B.2-60.2 11 19B.2 60.3 11 19B.2 61 7 19B.2-62 7 O Amendment 11 19D.21
- MM .
23A6100AS Standard Plant - ,,
ny a i
SECTION 19H.4 ;
O GE PROPRIETARY - pmvided under separate cover I
I' fagt Amendment 19H.4-1 8 19H.4 2 - 8 19H,4-3 8 19H.4 4 8 19H,4 5 8 l
- .l 19H.4-6 8 19H.4 7 8 19H,4 8 8 19H.4 9 11 19H.410 - 11
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Amendment 11 1911 4 1
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men =Aard Plant ggy.J 7]:: Response RAl"
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Branch Review Area Question Number SSAR Subsection . Sobeection letter 251.4 53.1.6.1 203.1 1 251.5 53.1.63 203.1 1 251.6 53.2.1 203.1 1 251.7 53.2.1.1 20 3.1 1 5.3.2.1.2 53.2.13 53.2.1.5 251.8 533 203.1 1 251.9 533.1.1.1 203.1 1 251.10 533.2 203.1 1 251.11 - $33.6 203.1 1 251.12 3.1.2.5.2.1 2033 3 251.13 ~ 3.5.1.1.13 2033 3.
p 251.14 3.5.4.1 2033 3 l'
l Materials 252.1 4.5.1.1(1) 203.1 1 Application 252.2 4.5.1.1(2) 20 3.1 1 252 3 4.5.2.2 203.1 1 252.4 4.5.23 203.1 1 252.5 4.5.2.4 203.1 1 252.6 4.5.2.5 203.1 1 .<
252.7 5.23.2.2 20 3.1 1 I
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1 252.10 5.23.4.1.1 203.1 1
, 252.11 5.23.4.23 203.1 1 l- 252.15 3.63 2033 3 i.
LOAB Ouality 260,1 17.0 203.8 8 Assurance 260.2 17.0 203.8 8 f 260 3 17.0 203.8 8 EMEB Seismic and 271.1 3.10.13 203.5 5 l Dynamic 271.2 3.10.1 3 203.5 5 l Load Oualification ECEB Chemical 281.1 5.1 20 3.1 1 Technology 281.2 5.23.2.2 20 3.1 1 l 2813 5.23.2.2 203.1 1 L 281.4 5.23.2.2 203.1 1 l; 281.5 5.23.2.2 203.1 1 281.6 5.23.2.2.2 203.1 1 281.7 5.23.2.23(4) 20 3.1 1 281.8 5.23.2.23(13) 20 3.1 1 281.9 6.4.9.2 20 3.1 1 281.10 Chap. 5 203.1 1 ,
/ 281.11 9.13 20 3.7 7 m
Amendment 11 20.1 1.2 l
T _ - _ _ _ _ . - _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ -
p.:
MM 23A6100AT Remndard Plant nyn NRC* - Revlem Questlos SSAR Response RAl**
Branch Area Number Subsection Subsectin 14tter 281.12 9.13 203.7 7 l 281.13 11.1 203.7 7
, i 281.14 11.5.2 203.7 7 l- 281.15 10.4 203.11 11 281.16 10.4.6 3 203.11 11 281.17 10.4.7 203.11 11 281.18 10.4 20 3.11 11 SPLB Auxiliary 410.1 3.5.1 2033 3 l- Systems 410.2 3.5.1 2033 3 4103 3.5.1.1 2033 3 410.4 3.5.1.1 2033 3 410.5 3.5.1.1 2033 3 410.6 3.5.1.1 2033 3 410.7 3.5.1.1 2033 3 410.8 3.5.1.1 2033 3 410.9 3.5.1.1 2033 3 410.10 3.5.1.2 2033 3 410.11 3.5.1.2 2033 3 410.12 3.5.1.2 2033 3 ',
i 410.13 3.5.1.2 2033 3 410.14 3.5.1.2 2033 3 410.15 3.5.1.2 2033 3 410.16 3.5.1.4- 2033 3 410.17 3.5.2 2033 3 .
410.18 3.5.2 2033 3 410.19 3.5.2 2033 - 3 410.19a 3.6.1 - 2033 3 410.20 3.6.1 2033 3 410.21. 3.6.1 2033 3 410.22 3.6.1' 2033 3 410.23 3.6.1 2033 3 l 410.24 3.6.1 2033 3 l 410.25 3.6.1 2033 3 1 410.26 3.6.1 2033 3 l
410.27 3.6.1 2033 3 410.28 3.6.1 2033 3 410.29 3.6.1 2033 3 410.29a APP 3I 2033 3 41030 3.11 2033 3 41031 9.13 203.7 7 41032 9.13 203.7 7 41033 9.13 203.7 7 41034 9.13 20 3.7 -7 ,
41035- 9.13 203.7 7 41036 9.13 203.7 7 41037 9.13 20 3.7 7 i 41038 9.13 203.7 7 ,
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41039 9.13 203.7 7 A 410.40- 9.1.5 203.7 7 -
410.41 9.1.5 203.7 7 ,. ;
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410.46 9.2.9 203.7 7 410.47 9.2.9 20 3.7 7 410.48 9.2.9 203.7 7 r 410.49 9.2.9 203.7 7 410.50 1.t IG 203.7 7 410.51 A2iJ 203.7 7 410.52. % 10 203.7 7 410.53 9.2.10 203.7 7 i 410.54 9.2.10 203.7 7 410.55 9.2.11 203.7 7 410.56- 9.2.11 203.7 7 410.57 9.2.11 203.7 7 410.58 9.2.11 203.7 7 410.59 9.2.11 203.7 7 f 410.60 9.2.11 203.7 7 l:
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7 410.63 9.2.12 20 3.7 7 L
410.64 9.2.13 203.7 7 SCIB I&C 420.1 ~ Chap 7 203.8 = 8. i
' 420.2 Chap 7 203.8- 8 4203 ' Chap 7 203.8 8
- 420.4 Chap 7 203.8 8 420.5- Chap 7 203.8 8 420.6 App 31 203.8- 8 .
i 420.7. App 31 203.8 8 420.8 App 31 203.8 8 t 420.9 App 31 203.8 8 l , 420.10 Chap 7 203.8 8 g 420.11 7.6.1.1 203.8 8
, 420.12 7.4.2.2.2 203.8 8
/ 420.13 Chap 7 203.8 8 420.14 7.1.23.9 203.8 8 420.15 7.4 20 3.8 8 420.16 7.4 203.8 8 7, 420.17 Chap 7 203.8 8 420.18 Chap 7 203.8 8 420.19 7.1 203.10 10 l- 203.8
.1Q 420.20 420.21 Chap 7 Chap 7 203.8 8
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O Branch Area Number Sutmection Subsection - letter 420.22 Chap 7 203.8 8 i
420.23 ' Chap 7 203.8 8 l
p 420.24 - Chap 7 - 203.8 8 i 420.25 Chap 7 203.8 8 420.26 7.1.2.1.6 203.8 8 l 420.27 Chap 7 203.8 8
- j. 420.28 App 15A -203.8 8 _!
J 420.29 7.1.1 203.8 8 42030 ' 7.1.2.2 203.8 8 42031 7.1.23.2 203.8 8 42032 7.1.23.2 203.8 8 e 42033 7.1.23.2 203.8 8 42034 7J 23.7 203.8 8 42035 7.1.2.6.5 203.8 8 42036 7.1.2.6.6 203.8 8 42037 7.1.2.6.7 203.8 8 ;
42038 7.1 203.8 8 42039- 7.1 .-203.8 8 420.40 73.1.1.1.1 203.8 8 420.41 73.1.1.1.1 203.8 .8 420.42' 73.1.1.1.1 203.8 ~8 73.1.1.2 420.43 20 3.8 8 -
420.44- 73.1.1.13 203.8 8 420.45 73.1.1.13 203.8 8 420.46- 73.1.1.1.4 203.8 8 420.47 73.1.1.1.4 203.8 8 420.48 7.1.2.1.6 203.8 8 420.49 Chap 7 203.8 8 420.50- 7.1 203.8 8 420.51 7.1 203.8 8 .
420.52 Chap 7 - 203.8 8 !
420.53 - Chap 7 203.8' 8 ,
420.54- Chap 7 20 3.8 8 !
420.55 Chap 7 203.8. 8
' -q 420.56 Chap 7 203.8 8 420.57 Chap 7 203.8 8 ]i 420.58- Chap 7 203.8 8 1 420.59 Chap 7 20 3.8 8 '2 420.60 7.1.2.2 203.8 8 420.61 7.1.2.2 203.8 8 .l 420.62 7.1.2.10.11 203.8 8 1 420.63 Chap 7 20 3.8 8 i 420.64 Chap? 203.8 8 1 420.65 Chap 7 203.8 8 ')
420.66 Chap 7 203.8 8 j 420.67 Chap 7 203.8 8 ;
420.68 Chap 7 203.8 8 1 420 69 Chap 7 203.11 11 l
1 Amendment 11 20.1 1.5 e !
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ABWR ==r Atandard Plant arv.n I i NRC* Review Question SSAR Response RAl**
E Branch Area Nunnber Sub6ection Subsection 14tter 4
420.70 7.1.2.1.6 203.8 8 420.71 7.1.2.1.6 203.8 8 420.72 7.1.2.1.6 203E 8 420.73 7.1.2.1.6 203.8 8 420.74 7.1.2.1.6 203.8 8 420,75 ' 7.1.2.2 203.8 8 420.76 7.1.23.2 203.8 8 420.77 7.1.2.1.4.1 203.8 8 420.78 7.1.2.1.4.1- 203.8 8 420.79 7.1.2.1.4.1 20 3.8- 8 420.80 - 7.1.23.1 203.8 8 420.81 7.1.23.1 203.8 8 420.82 7.1.233 20 3.8 8 420.83 7.1.23.4 203.8 8 42034 App 31 203.8 8 420.85 Chap 7 203.8 6
? 420.86 Chap 7 20 3.8 8 420.87 - Chap 7 203.8 8 420.88 Chap 7 20 3.8 8 420.89 Chap 7 203.8 8 420.90 Chap 7 203.8 8 420 S 1 '- Chap 7 203.8 8
- 420.92 Chap 7 203.8 8 420.93 Chap 7 203.8_ 8
- 420.94 Chap 7 203.8 8 420.95 Chap 7 203.8 8 42096 15A.6 203.8 8 420.97 73.1.1.4 203.8 8-420.98 Chap 7 203.8 8 420.99- Chap 7 203.8 8 420.100- Chap 7 203.8 8 420.101 Chap 7 203.8 8 420.102 Chap 7 203.8 8-420.103 Chap 7 203.8 8 420.104 Chap 7 -203E 8 420.105- Chap 7 203.8 8 420.106 Chap 7 203.8- 8 420.107 93.5.2 203.8 8 420.108 7.1.2.2 2033 8 420.109 7.1.23.1 20 3.8 8 420.110 7.1.23.1 20 3.8 8 420.111 7.1.23.7 203.8 ' 8 420.112 7.1.2.43 203.8 8 420.113 7.1.2.6.1.1 203.8 8 420.114 App 7A 203.8 8 420.115 73.1.1.13 2032 8 420.116 1.2.2.4.8.1.2 203.8 8 420.117 93.5.1.1 203.8 8 Amendment 11 20.1 1.6
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MM Riandard Plant 23A61rioAT REV.B' NRC*- Review Question SSAR- Response RAP' 4
Branch Area Number = Subsection Subsection Letter
. 420.118 15.2.4.5.1 20 3.8 8
- 203.8 420.119 7.4.1.2 8 420.120 73.2.1.2 - 203.8 8 420.121 73.1.2 20 3.8 '8 j 420.122 15.2.2.2.1.4 203.8 8 420.123 15B.4 203.11 11 L 420.124 15B.4 20 3.11- 11 420.125 - 7.4.1.4 203.11 11 420.126 7A.7 203.11 11 420.127 Chap 7 20 3.11 11 420.128 7A.7 20.2.11 11 -
420.129 Chap 7 203.11 11 l 420.131 19.23.4 203.11 11 420.132 193.13.1 203.11 11 420.133 193.13.1 203.11 11 420.134 19D3.4 203.11 11 !
420.135 19D.6 203.11 11 I 420.136 App 7A 20 3.11 11 !
SPLB Plant 430.1 4.6 20 3.2 2 Systems 430.2 5.2.5 203.2 2 4303 5.2.5 20 3.2 2 i 430.4 5.2.5.4.1 203.2 2 -
430.5 5.2.5 203.2 2 430.6 5.2.5 203.2 2 430.7 6.2 20 3.2 2 430.8 6.2 ' 203.2 2 430.9 6.2 203.2 2 430.10 6.2 203.2 2 430.11 6.2 20 3.2 2 ;
430.12 6.2 203.2 2 1 430.13 6.2.1.13 20 3.2 2 430.14 6.2 - 20 3.2 2 430.15 6.2 203.2 2- 1 430.16 6.2- 20 3.1 2- i 430.17 6.2.1.23 2M.2 2 430.18 6.2 20 3.2 2 430.19 6.2 203.2 2 430.20 6.2 203.2 2 430.21 6.2 203.2 2 430.22 6.2 20 3.2 2 .j 430.23 6.2 20 3.2 2 430.24 6.2 203.2 2 430 25 6.2 203.2 2 430.26 6.2 203.2 2 430.27 6.2 203.2 2 430.28 6.2 203.2 2 430.29 6.23 20 3.2 2 Amendment 11 20.1 1.7 1
-Standard MM Plant ~ 21A6100AT arv. s
! NRC* Review Question ' SSAR Response . RAl**
Branch Area Number Subsection . Subsection htter 1 43030 6.2 '203.2 2 43031 6.2 203.2 2 43032 6.2 203.2 2 43033 6.2 203.2 2 43034 6.2 20 3.2 2 t 43035 6.2- ' 203.2 ' 2 43036 6.2 203.2 2'
- 43037 6.2 203.2 -2 43038 6.2 20 3.2 2 43039 6.2.4 20 3.2 2 430.40 6.2 203.2 2 430.41 6.2- 203.2 2 430.42 6.2 203.2 2 430.43 6.2 203.2 2 430.44 6.2 203.2 ' 2-430.45 6.2 203.2 2 430.46 6.2 20 3.2 2 430.47 6.2.53 203.2 2 430.48 6.2.6 20 3.2 2 430.49 6.2.6 20 3.2 2 430.50 6.2.6 203.2 2
=
430.51 6.2.6 203.2 2 430.52 - 6.2.6 203.2 2 430.53 6.2.6 203.2 2 430.54 6.4 203.2 2 430.55 6.5.1 20 3.2 2-430.56 6.53 20 3.2 2 430.57 6.7 203.2 2 430.58 15.7 3 20 3.2 2 430.59 10.1 203.11 11 430.60 10.2' 203.11 11 430.61 10.2.2.2 203.11- - 11 430.62 10.2 203.11 11 430.63 10.2.2.4 203.11 11 430.64 10.2.2.4 203.11 11 430.65 10.2 203.11 11 430.66 10.2 203.11 11 430.67 103.2.1 203.11 11 430.68 1033 203.11 11 430.69 10 3 203.11 11 430.70 10 3 203.11 11 430.71 10.4.1 20 3.11 11 430.72 10.4.1 203.11 11 430.73 10.4.1 20 3.11 11 430.74 10.4.2 203.11 11 430.75 10.4.2 20 3.11 11 430.76 10.4.2 203.11 11
. 430.77 10.4.2 203.11 11 430.78 10.4.2 203.11 11 i Amendment 11 10.1 1.8
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2346ioorr Standard Plant nrv. n i
NRCi Review Question SSAR Response RAl" :
Branch Area Number Subsection Subsection letter 430.79 10.4.2 203.11- 11 430.80 10.43 203.11 11 430.81 10.4 3 203.11 11 430.82 10.4 3 203.11 11 430.83 10.4 3 203.11 11 430.84 10.4.4 203.11 11 430.85 10.4.5 20 3.11 11 430.86 10.4.7 203.11 11 430.87 Chap 10 203.11 11 .
430.88 Chap 10 203.11 11 j 430.89 10.4.7 203.11 11 430.90 10.4.7 203.11 11 SELS Power 435.1 8.1.2.1 20 3.8 8 !
Systems 435.2 - Chap 8 203.8 8 435 3 - 8.1.2.1 203.8 8 435.4 8.23 203.8 8 435.5 8.23 203.8 8 435.6 83.1.1.4.1 20 3.8 8 435.7 83.1.1.4.2.2 203.8 8 435.8 83 203.8 8 435.9 83.1.1.4.23 20 3.8 8 lj 435.10 83.1.1.4.2.4 20 3.8 8 4 435.11 83.1.1.5.1 20 3.8 - 8 :
435.12 83.1.1.5.2 203.8 8 435.13 83.1.1.6.4 203.8 8 i 435.14 83.1.1.7 203.8 8
- 435.15 83.1.1.7 20 3.8 8 435.16 83.1.1.7 203.8 8 ;
435.17 83.1.1.7 20 3.8 8 435.18 83.1.1.7 20 3.8 8 435.19 83.1.1.7 20 3.8 8 c
435.20 83.1.1.7 203.8 8 ;
435.21 83.1.1.8.2 203.8 8 435.22 83.1.1.8.5 203.8 8 435.23 83.1.2.1 203.8 8 435.24 83.1.2.1 203.8 8 435.25 83.1.1.23 203.8 8 t 435.26 83.1.2.2 203.8 8 435.27 83.1.2.2 203.8 8 435.28 83.1.2.4 203.8 8 435.29 83.13.1 20 3.8 8 43530 Chap 8 203.8 8 43531- 83.1.4.1.2 203.8 8 43532 83.1.4.71 203.8 8 43533 83.1.4.2.2.2 20 3.8 8 43534 83.1.4.2.2.4 203.8 8 43535 83.1.4.23.1 203.8 8 43536 83.1.4.23.2 20 3.8 8 43537 83.2.1 20 3.8 8 Amendment 11 20.1 1.9 W l19 ' l' l'. C""."l""9FF'IFl I II Il m I Il' . . .. . . '
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- NRC* - Review Question SSAR- Response RAl**
- . Branch Area Number Subsection Subsection Letter 43538 -83.2.1 203.8 8 43539 83.2.1 203.8 8 435.40 83.2.1 203.8 8' 435.41 83.2.1.2 '203.8 8 43542 53.2.13 2033 8 435.43 83.2.133: 203.8 8 435.44 83.2.2.1 203.8 8 435.45 833.1- 20 3.8 8 435.46 - Table 83-1,2,3 203.8 8 435.47 Figure 83-1. 203.8 8 435.48 Figure 83-2 203.8 8 435.49 Figure 83-3 203.8 8 435.50. : Figure 83-4 203.8 8 435.51 Figures 83 5, 203.8 8 6,7,8 435.52 . Figure 83-7 20 3.8 8 435.53 . Figure 83-8 ' 203.8 8 435.54 Table 3.21 203.8 8 435.55 8.2.1.1.8.9 203.8 8 435.56 Chap 8 20 3.8 8 435.57 Chap 8 203.8 8 435.58 Chap 8 203.8 8 435.59 Chap 8 203.8 8 435.60 83.1.2.1 203.8 8
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435.61 Chap 8 20 3.8 8 435.62 Chap 8 203.8 8 SRXB Reactor 440.1 4.6 203.2 2-Systems 440.2 4.6.23.2.2 203.2 2 440 3 4.6.1.2 203.2 2 440.4 4.6 20 3.2 2 440.5 4.6 - 203.2 2 440.6 - 4.6 - 203.2 2
' 440.7 4.6 20 3.2 2 440.8 4.6 20 3.2 2 440.9 4.6 203.2 2 440.10 4.6.23.1 203.2 2 440.11 4.6 203.2 2 440.12 4.6 20 3.2 2 440.13 5.2.2 203.4 4 440.14 5.2.2 203.4 4 440.15 5.2.2 203.4 4 440.16 5.2.2 20 3.4 4 440.18 5.2.2 20 3.4 4 440.19 5.2.2 203.4 4 440.20 5.2.2 203.4 4 440.21 5.2.2 203.4 4 440.22 5.1 203.4 4 440.23 5.2.2 203.4 4
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2 440.28 1.8 203.4 '4 440.29 5.2.2 20 3.4 4 Amendment 11 20.1 1.10
M 23A6100AT Remndard Plant REV.B j NRC* . Review Question SSAR Response RAl" -
Branch Area Nunnber Subsection Subsection Letter 44030 4.6 203.4 - 4
, 44031 4.6 203.4 4 440.17 5.2.2 20 3.4 4 44032 4.6 203.4 4 44033 4.6 203.4 4 44034 5.4.1 203.4 4 44035 5.4.1 203.4 4 44036 5.4.1 203.4 4 44037 5.4.6 203.4 4 44038 5.4.6 203.4 4 4 44039 5.4.6 203.4 4 'I 440.40 5.4.6 20 3.4 4 g 440.41 5.4.6 203.4 4 440.42 5.4.6 203.4 4
, 440.43 5.4.6 20 3.4 4 440.44 5.4.6 203.4 4 440.45 5.4.6 203.4 4 440.46 5.4.6 203.4 4 440.47- 5.4.6 203.4 4 440.48 5.4.6 203.4 4 440.49 5.4.6 20 3.4 4-440.50 - 5.4.6 203.4 4 . .;
440.51 5.4.6 203.4 4 --
l 440.52 5.4.6 203.4 4 440.53 5.4.6 203.4 4 440.54 5.4.6 203.4 4 440.55 5.4.6 203.4 4 440.56 5.4.6 20 3.4 4 440.57 5.4.6 20 3.4 4 440.58 5.4.6 203.4 4 440.59 5.4.7 203.4 4 /
440.60 5.4.7 203.4 4 T.
440.61 5.4.7 203.4 4 440.62 5.4.7 203.4 4 440.63 5.4.7 203.4 4 440.64 5.4.7 203.4 4 440.65 5.4.7 20 3.4 4 i 440.72 5.4.7 203.4 4 440.73 5.4.7 203.4' 4 440.74 5.4.7 203.4 4 i 440.75 63 20 3.6 6 440.76 63 203.6 6 440.77 63 203.6 6 440.78 63 203.6 6 440.79 63 203.6 6 440.80 63 203.6 6 440.81 63 20 3.6 6 440.82 63 203.6 6 : <
440.83 63 20 3.6 6 -
Amendment 11 20.1 1.11
4 1 4: .
ABM . 2 461oo^r maderd Plant REV.B
'NRC* Review Question SSAR Response RAl**
Branch Area Number Subsection' Subsection latter 440.84 63 203.6 6 440.85 63 203.6 6 440.86- 63 203.6. 6 440.87 63 203.6 6 440.88 63 203.6 6 440.89 63 203.6 6
' 440.90 63 20 3.6 6 440.91 63 203.6 6 440.92 63 20 3.6 6 440.93 63 203.6 6 440.94 63 203.6 6 440.95 63 203.6 6 440.% ~ 63 203.6 -6 440.97 63 203.6 6 440.98 63 - 20 3.6 6-440.99 63 203.6 6 440.100 63 20 3.6 6 440.101 93.5 203.6 6 440.102 93.5 203.6 6 440.103 93.5 203.6 6 440.104 93.5 203.6 6 440.105 93.5 203.6 6 440.106 93.5 20 3.6 6 9 440.107 440.108 440.109.
93.5 Chap 15 Chap 15 203.6 203.6 203.6 6
6 6
440.110 Chap 15 10 3.6 6 440.111 Chap 15 20 3.6 6 440.112 Chap 15 20 3.6 6 440.113 Chap 15 203.6 6 440.114 Chap 15 203.6 - 6 440.115 Chap 15 203.6 6 440.116 Chap 15 203.6 6 PRPB Meteorology 451.1 2.0 2033 3-451.2 2.4 2033 3 SPLB Effluent 460.1 11.1 20 3.7 7 Treatment 460.2 11.1 20 3.7 7 4603 11.1 20 3.7 7 g 460.4 11.1 20 3.7 7 460.5 11.5 203.7 7 PRPB Radiological 470.1 15.5.2 203.1 1 Report 470.2 15.6.2 20 3.1 1 470 3 15.6.4.5.1.1 203.1 1 470.4 15.6.5.5 203.1 1 470.5 15.6.5 203.1 1 470.6 15.7.5 20 3.1 1 203.1
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470.7 15.7 1 203.1 470.8 15.7 1
- Amendment 11 20.1 1.12 4
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ABM 2m6ioorr REV.B 3: Remndard Plant NRC* Review Question - SSAR Response RAl**
Branch Area Number - Subsection Subsection letter 470 9 15.7 203.1 1 470.10 15.7 203.1 1 PRPB Radiation 471.1 12.1.1.2 -203.7 7 Protection 471.2 12.1.2.2.1 20 3.7 7 4713 12.1.2.2.2 20 3.7 7 471.4 12.1.23 20 3.7 7 471.5 12.1.2 3 203,7 7 471.6 12.1.23.2 203.7 7 471.7 12.1.2 3.2 203.7 7 471.8 12.2 20 3.7 7 471.9 11.1 20 3.7 7 471.10 12.1.2.23 203.7 7 471.11 123 20 3.7 ' 7 471.12 11.1 20 3.7 7 471.13 12.2.2 20 3.7 7 471.14 12.2.2 203.7. 7 471.15 - Chap.12 20 3.7 7 471.16 Chap.12 - 20 3.7 7 471.17 12.2 203.7 7 471.18 Chap.12 203.7 7 471.19 12.2 20 3.7 7 471.20 Chap.12 20 3.7 7 471.21 123 203.7 7 471.22 Chap.12 203.7 7 471.23 123.13 20 3.7 7 471.24 12 3 20 3.7 7 471.25 123 20 3.7 7 471.26- 12 3 20 3.7 i 471.27 123 20 3.7 7 471.28 - Chap.12 203.7 7 471.29 123 20 3.7 - 7 47130- Chap.12 20 3.7 7 47131 12.2.2.1 20 3.7 7 47132- Chap.12 203.7 7 47133 123 20 3.7 7 47134 123.5 20 3.7 7
- 47135 Chap.12 203.7- 7 47136 123 203.7 7 47137 123 20 3.7 7 47138 12 3.1 203.7- 7 47139 Chap.12 203.7 7 471.40 12 3 20 3.7 7 471.41 123 20 3.7 7 RES Probabilistic 725.1 _ App.19D 20 3.9 - 9 Risk 725.2 App.19D 203.9 9 Assesment 7253 App.19D 203.9 9 725 3 App.19D 20 3.9 9 725.4 App.19D 20 3.9 9 725.5 App.19D 20 3.9 9 Amendment 11 20.1 1.13 k
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, Ad . Branch- Area Number. Subsection Subsection Letter .
725.6 App.19D 203S 9 l 725.7 . App.19D 2039 - 9- l App.19D 725.8 203.9 9 725.9 App.19D - 203.9 9 725.10 App.19D '203.9' 9 725.11 App.19D 20 3.9 9 ,
725.12 App.19D 20 3 S 9 725.13 ' App.19D 203.9 9 725.14 App.19D 203.9 9 725.15 - App.19D 203.9 9 e 725.16 App.19D 203.9 9 725.17 App.19D 20 3 S 9 725.18 App.19D - 203.9 9 . - .
725.19 App.19D 203.9 9 725.20 App.19D 20 3.9 9 725.21 App.19D 203.9 9 725.22 App.19D 203.9 9 725.23 App.19D 203.9 9 lT 725.24 App.19D . 203.9 9 l 725.25 App.19D 203.9 9 725.26 App.19D 20 3.9 9 725.27- App.19D 2039 9 725.28 App.19D 20 3.9 9 l: O .725.29 App.19D 203.9 9.
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( 72530 App.19D 203.9 - 9 72531 ' App.19D 203.9 9.
72532 App.19D 203.9 9 ,
72533 App.19D 203.9- 9 72534 App.19D 203.9 9 l 72535 App.19D 203.9 9 ,
72536 App.19D 203.9 9 72537 App.19D 203.9 9 l 72538 App.19D 203.9 9-l 72539 App.19D 203.9. 9 725.40 App.19D . 203.9 9-L l 725.41- App.19D 203S 9 ll 725.42 App.19D 203.9 9
- - 725.43 App.19D 2039 9 l
725.44 App.19D 20 3.9 9 L 725.45 App.19D 203.9 9 1^ 725.46 App.19D 2039 9 !
l 725.47 App.19D 203.9 9 725.48 App.19D 203.9 9 725.49 App.19D 203.9 9 725.50 App.19D 203.9 9 -
725.51 App.19D 203.9 9 725.52 App.19D ' 203.9 9 725.53 App.19D 203.9 9 725.54 App.19D - 20 3.9 9 725.55 App.19D 203.9 9 h_76 ! 725.56 App.19D 203.9 9 l
l, Amendment 11 20.1 1.14 w
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- Anch -- Area Number Subsection Subsection htter . -,
725.57 ' App.19D 203.9 9 j 725.58 App.19D 203.9 9 :j 725.59 App.19D - 203.9 9 725.60 App.19D 203.9 9 725.61 App.19D 203.9 9 RSGB Safeguards - 910.7 13.6.1 203.7 7 910.8 13.6 3.7 20 3.7 -7 910.9 13.6 3 203.7 7
. 910.10 - 13.6 3.7 203.7 7 910.11 13.633 203.7 7 910.12 13.6 3.4 20 3.7 7 910.13 13.6 3.6 20 3.7 7 910.14 13.6 3 203.7 7 910.16 13.6 203.7 7 O
l Amendment 11 20.1 1.15
ABWR 3- ,
2 m ioo n Standard Plant - wn 420.18 l For the proposed use of digital computers, show how the digital system is superior to analog alternatives to implementing the logic. Show how the analyses determined that the realiability of the digital computer based system was better than the reliability of the analog system. (7) 420.19 This section states that automatic self test is performed sequentially on all four divisions, to minimize common mode effects, and that a complete self test sequence through all four divisions takes no more than 30 minutes. The original response to Question 19 revised this section. What hardware and software design features are provided to allow sequencing and testing of the four divisions without violating independence / isolation criteria? The revised section appears to allow a common centralized test driver.
Illustrate with a block diagram. (7.1.2.1.6.(4))
420.20 Describe the fiber optic links in the safety systems, What signals are multiplexed on each link? Show how the independence criteria in accordance with IEEE Std 603 and IEEE Std 379 is sati.;fied with the proposed configuration of fiber optic links.(7) 420.21 Describe the safety computer system's interface to any non safety computer systems and other plant instrumentation. Describe if information transfer from 1E to N.1E computers is via broadcast or handshake.
(7) 420.22 Provide a table of conformance to IEEEE 603 and ANSI /IEEE 7-43.2. (7) 420.23 Provide a table of conformance to IEEE 384, indicating where credit is taken for isolation or separation, what devices or methods are used, and the basis of isolation device qualification. If specific types of components have not been chosen, provide specification level information including testing acceptance criteria.
(7) 420.24 l Are any artificial intelligence features provided in the proposed system, wi.ereby probabilistic judgements are made by the system, or whereby the system can ' learn" during its operational life? (7) 420.25 Is credit taken in the safety analysis for any rotating memory devices such as disk drives? (7) 420.26 What is the definition of " Safety Associated" as used in SAR Section 7.1.2.1.67 (7.1.2.1.6)
O Amendment It 20.2 13.2
.Mkk 23A6100AT * '
Standard Plant a,y. s 420.27. .
. Specify which parameters are to be triplicated. At what point does the triplication start (flow orifice, -
l
' -l~' sensor?) and end (transmitter, trip logic?). If there is triplication of se., sors is there dhcrsity between sensors?
(7)
L 420.29 For those systems where it has not already been done (example 7.1.1.3.5) clarify whether manual or !
automatic initiation will be used. (7.1.1) .
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Standard Plant - au s 420.30
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Define the word ' sufficient" used in section (j). (7.1.2.2) 42031 For section 7.1.23.2(1)(c,d,e) and (2)(a) defime sufficient". (7.1.23.2) 42032 The listed design basis should include instrumentation necessary to inform the operator that isolation has been completed and control should provide ability for operator to reset (with adequate safeguards against inadvertently breaking isolation). (7.1.23.2)
^
42033 -
Add to 7.1.23.2(2)(c)...'without causing plant shutdowns" or reducing safety margins. (7.1.23.2) 42034 For Section 7.1.23.7(1)(b) provide a listing of the nonessential parts of the cooling water system which should be isolated. List any nonessential parts for which isolation is not provided. (7.1.23.7) 42035 is the wetwell to drynll vacuum breaker control manual or automatic? (7.1.2.6.5)
.42036 y
If the CAMS system is only a monitoring system, why is it not always on instead of waiting for a LOCA to -
monitor radiation? (7.1.2.6.6) a 42037 What is the immediate safety action required by relief valve leakage and is it automatic? (7.1.2.6.7) 42038 The table indicates RG 1.151 applies only to safety related display and Non-1E control systems. Section 7.1.2.10.11 refers to other safety systems including RPS and ECCS. Clarify which systems RG 1.151 is to apply to.
(Table 7.12) 42039 The table lists few systems for which RG 1.97 is applicable. Address the RG 1.97 for all categories and variables. (Table 7.12) 420.40 -
The HPCF pump is interlocked (73.1.1.1.1(3)(c)) with the undervoltage monitor. If the breaker cannot close will it retry and what information is available to the operator if it doesn't close that would indicate an undervoltage problem? (73.1.1.1.1)
Amendment 9 20.2 13.3 i
k 4 y 1M R zusiooAT -
Standard Plant : nu s 420.60 -t w-
,k Provide examples fca section (g) which meet the design bases. (7.1.2.2) 420.61-Explain section (b) further. Does this mean one 480V bus,4160 bus the generator? Same question at 7.2.3.2(2)(b).- (7.1.2.2)
+ 420.62 Provide justification for going to a 2/3 scram instead of 1/3 when or e is bypassed. (7.1.2.10.11)
- a. '
= 420.63 -
What are the realiability/ availability goals for the reactor prote: tion and engineered safety features systems? (7) 420.64' Describe the reliability model and assumptions used to demonstrate achievement of the reliability goals; this should include a description of the system architecture. (7) :
' 420.65 What methodology is used in determining the system reliability / availability? (7) 420.66 I . A: s \ ,
'V' '
' Describe the data validation features in triplicated sensors. (7) l 420.67 What testing will be done to demonstrate reliability? What is the specific scope of these tests? (7)-
420.68 i What is the effect upon the number of spurious trips generated by the RPS if the digital design replaces -
, the previous analog design? Provide comparison (7) ,
420.69 Are there any limitations on the ABWR design concerning the use of expert systems? Any limitations on the use of technology not specifically described? The original response does not describe an approach for .r determining what hardware or software developments which may ocur between design certification and plant operation can be implemented without changes to the design certification and NRC review. (7A) 420.70.
Is there any system for in service testing of the ARI? (7.1.2.1.6) 420.71
' ("N Is the CRD scram discharge high water level used as the example of the fifth test valid given that there is yl no scram discharge volume? (7.1.2.1.6)
Amendment 11 20.2 13.6 &
-, a 23A6100AT
= Standard Plant - wn 420.72 Section (1) of 7.1.2.1.6 states that normal surveillance can identify failures. Discuss whether this system has the capability of transmitting this information to the plant computer so that an immediate alarm can be given in addition to waiting for the scheduled surveillance. (7.1.2.1.6) l 420.73 Section (4) notes that the four divisions are tested in sequence. When the thirty mbute sequence is '
l complete does the test system start over again or is this an operator initiated test? (7.1.2.1.6) .p 420.74 i Section (5) notes that only one division shall be bypassed at any one time. Describe the interlock protection or administrative controls which assure this. (7.1.2.1.6) 420.75 i
For section 7.1.2.2(j) clarify that the physical and electrical separation does not preclude the proper environmental qualification of redundant I&C equipment. (7.1.2.2) i 420.76 For section 7.1.23.2(1)(c,d,e) and (2)(a) define ' sufficient *. (7.1.23.2) i 420.77 One of the reasons stated for the utilization of microprocessors for the implementation of instrumentation -
and logic function:, is that less uncertainty exists in the margins between actual safety limits and the limiting safety trips. The margins are stated to be set from experimental data on setpoint drift (see Section 7.1.2.1.4.1) ;
and from quantitative reliability requirements for each system and its components. l s
Provide the documented bases for this procedure. (7.1.2.1.4.1) 420.78 One of the reasons stated for the utilization of microprocessors fo: the implementation ofinstrumentation and logic functions is that less uncertainty exists in the margins between actual safety limits and the limiting safety trips. The margins are stated to be set from experimental data on setpoint drift (see Section 7.1.2.1.4.1) and from quantitative reliability requirements for each system and its components.
Will this precedure be a topical report used as a design tool? (7.1.2.1.4.1) 420.79 !
One of the reasons stated for the utilization of microprocessors for the implementation of instrumentation and logic functions is that less uncertainty exists in the margins between actual safety limits and the limiting safety trips. The margins are stated to be set from experimental data on setpoint drift (see Section 7.1.2.1.4.1) and from quantitative reliability requirements for each system and its components.
What experimental data has been used to provide inputs to this design approach? (7.1.2.1.4.1)
O Amendment 9 20,2 13.7 l
AB%1t 2sA6too^T Standard Plant m.v n 420.113
^
_ Has consideration been given to providing the annunciators with backup diesel or battery power? (Ref.
7.1.2.6.1.1(2)(g)). (7.1.2.6.1.1) 420.114 The copy of Section 7 provided to the staff did not include Appendix 7A nor an indication that it was to be provided later. Provide this section or a schedule for providing it. (7A.1 1) 420.115 In the discussion about torque switches and thermal overloads, there is a reference to Section 3.8.4.2 which is the applicable codes and standards for seismic qualification of the Reactor and Control Buildings.
What is the correct reference? (73.1.1.13(4)(c))
420.119 Are there any other valves which must isolate upon initiation of the SLCS? (7.4.1.2(7))
420.120 List all exemptions to the requirement rather than providing an example. (73.2.1.2(3)(c))
420.121
. The first paragraph states that pipe break outside containment and feedwater line break are discussed below. The staff could not locate these items. (73.1.2(7))
420.122 Is the instrumentation required for the operator to verify bypass valve perfortnance and relief valve operator 1E or N 1E7 (15.2.2.2.1.4) 420.123 SSAR 15B.4 describes the essential multiplexing system (EMS) in some detail. SSAR Figure 7A.21 states that the design is not limited to this configuration. It is our understanding that the EMS design is still in a preliminary design stage. Is SSAR 15B.4 still accurate and is the design limited to that configuration?.
(15B.4) 420.124 The FMEA submitted in SSAR 15B.4 is inadequate for a safety evaluation supporting the design certification. The FMEA appears to the staff to be oversimplified with one line item each for component failures and does not address potential software complications. The staff requests clarification of how this FMEA was developed given that the system design has not been finalized. The staff also believes that software failures need to be evaluated. The failure modes investigated should include, as a minimum, stall, runaway, lockup, interruption / restoration, clock and timing faults, counter overflow, missing / corrupt date, and effects of hardware faults on software. (15B.4) e Amendment 11 20.2 13.12
_l ABWR ursioorr Standard Plant no n 420.12s This section provided additional clarification of the intended use of the remote shutdown system. The degree of independence and isolation from the Safety System Logic and Control (SSLC) and EMS are not clear. Is it intended in the SSAR to take credit for the RSS if there is a totalloss of EMS? (7.4.1.4) 420.126 Compared with GESSAR II, the ABWR has significantly reduced the number of input sensors by use of sharing sensors. Provide a bases to why this does not increase potential vulnerability to common mode failures by reducing sensor diversity. (7A-7) 420.127 In general, the applicant should provide a clear presentation of how the ABWR with common software and hardware modules for many functions (including SSLC logic self-test programs) conforms with IEEE 2791971 and is at least as single failure proof as GESSAR II. The discussion of shared sensors in 7A 7 does not address potential common mode software failures which may be capable of defeating the diverse parameters. Additionally, the applicant should address why diversity of software should not be a requirement to maintain system diversity. (7) 420.128 Will software be used to isolate data? If so, what are the design and qualification criteria that are to be applied? Are there any systems which have non Class 1E softwere such as keyboard or display control software that interface with the Class-1E systems? Are there any interface with the Class 1E systems which receive inputs from non Class 1E systems or other channels of IE systems. (7A.7) 420.129 List those systems or major components in the I&C design area for which the design is not complete to the " purchase specification" level. (7) 420.130 In response to Question 420.63, a MTBF goal of 100,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (11.4 years) is given for the essential multiplexing system. Is this goal for one channel or the complete system? If this goalis for the complete system,it appears to the staff that the ABWR can expect to loose control at the control room of many of the safety systems (RPS, RHR, ADS) five or six times over the lifetime of the plant. How does this compare with the reliability / availability of multiple ESF systems in the BWR/S & 6 design (ur GESSAR II)?
420.131 Are multiplexer and software failures included in these systems interactions and common cause failures?
(19.2.3.4) 420.132 Section 19.3.13.1 (b) states that "if core cooling is accomplished without the use of an RHR systems and the suppression pool cooling begins overheating, the suppression pool cooling mode of the RHR will be initiated by the operator." Is any manual action required prior to 30 minutes? (193.13.1 (b))(Response
, 420.47)
O Amendment It 20.2 13.13
LABWR msimr- 1
' Standard Plant um n
..2 M i- CO.133
- g
- Subsection 19.3.1.3.1(c)(i) describes the MSIV closure sequence with the most desirable outcome requiring opera tor action at 30 seconds to insert rods, if that fails the operator must inhibit ADS valves from t opening and initiate SLCS within 10 minutes. THese activities do not appear to be consistent with stated design goal of no operator action for 30 minutes following a transient. Provide a description of how the MSIV closure sequence meets the 30 minute rule (6.3.1.1.1) same question for loss of Offsite Power (LOOP).
4 0.134 Equipment maintenance or test unavailability are taken from GESSAR PRA and are based upon BWR ' '!
experience. In the past,I&C has been a large contributor to system downtime. How do these systems (RHR, RCIC) unavailability numbers take into account the new multiplexing and microprocessors? (19D.3.4) -
40.135 =
Provide the justification for Mcan Time To Repair (MTTR) of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for multiplexers and 30 minutes for ESFlogic. Inverters and battery chargers have restoration time given in (Table 19A.8) as 48-56 hours. Are the multiplexers designed with all test and maintenance equipment installed? (Table 19D.610) 4 0.136 ?
- The staff has reviewed the commitments in the SSAR and has reviewed the available documentation describing the verification and validation plans. To date, the information has been vague, general in nature and lacking in essential detail to demonstrate conformance with ANSI /IEEE 7 4.3.2. Does the applicant intend to L enclose the V&V Plan as Appendix B of SSAR Chapter 7 or will the V&V details be left as an interface <
em : requirement? The staff requires a formal, structured V&V plan to be in place and implemented early in the software design process. (7A) I t
-O O
Amendment It 20.2 13.14
8 K >
,; gg g, ,
E*andard Plant ' arv. m
, .i 20.2.10 Chapter 10 Questions
, 4
- p. -
NM 281.15 In a letter from Thomas E. Murley, NNR, to Ricardo Artigas, G.E., dated August 7,1987, the -
p staff provided the ABWR licensing review bases as well as the scope and content of the ABWR Standard
- i. Safety Analysis Report (SSAR). In Section 8.7, Water Chemistry Guidelines, of the referenced letter, K it states that GE has committed to using BWR Owners Group water chemistry guidelines.-~ These i
- guidelines are necessary to maintain proper water chemistry in BWR cooling systems to prevent
'intergrannular stress corrision cracking of austenitic stainless steel piping and components and to minimize corrision and erosion /corrision induced pipe wall thinning in single phase and two phase high energy carbon steel piping. Water chemistry is also important for the minimization of plant -
radiation levels due to activated corrision products. Section 10.4.6.3 of the ABWR indicates that the condensate cleanup system complies with Regulatory Guide 1.56. Section 10.4 should indicate that ;
the system meets the guidelines published in:
EPRI NP-4947 SR. BWR Hydrogen Water Chemistry Guidelines 1987 Revision, dated October 1988. .,
L. EPRI NP 5283-SR A.. guidelines for Permanent BWR Hydrogen Water Chemistry 1987 Revision, dated b
' September 1987. ,
The use of zine injection as a means of controlling BWR radiation field build up should be '
discussed.
' 281.16 i In Sectisa 10.4.6.3, the ABWR SSAR indicates that the condensate cleanup system removes some LO A/1 radioactive material, activated corrision products and fission products that are carried over from the reactor. More important functions involve removal of condensate leakage w assare meeting BWR Hydrogen Water Chemistry Guidelines. This should be discussed.
-.281.17 L The condensate (Figure 10,4.4) and feedwater (Figure 10.4.7) system diagrams do not indicate the location of the oxygen injection into the condensate system and hydrogen and zine oxide into the feedwater system. This information should be provided.
281.18 Section 10.4 does not discuss design improvements involving material selection, water chemistry, steam temperatures, piping design and hydrodynamic conditions that are necessary to control
' erosion /corrision. The EPRI CHECMATE or other erosion /corrision computer codes mat be useful design tools to minimize wall thinning due to erosion /corrision corrisionn. The ABWR SSAR should discuss design considerations to minimize erosion /corrision and procedures and administrative controls to assure that the structuralintegrity of single phase and two phase high energy carbon steel piping systems is maintained.
430.59 Provide information on the following figures and tables: (10.1)
(a) Figure 10.1-2. Heat Balance for Guaranteed Reactor Rating (b) Figure 10.1-3, Heat Balance for Valve Wide Open Amendment 11 20.2 16 i
s - _ _ _ _ , _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ .
W gg -
Standard Plant arv. n (c) Table 10.11, Summary of Important Design Features and Performance Characteristics of the Steam and Power Conversion System, with regard to: ;
Condensate pumps: total head (ft) and motor hp.
Low pressure heaters: Stage pressure (psia) and duty per shell (Btu /ht) for Heaters Nos.1, 2, 3, and 4.
High pressure heaters: Stage pressure (psia) and duty per shell (Blu/br) for Heaters Nos. 5 and 6.
law pressure turbine exhaust pressure to condenser
- 430.60 Specify the value for time 'T" in Figure 10.2-2. (10.2) 430.61 Provide a description of the bulk hydrogen storage facility mentioned in Section 10.2.2.2.
(10.2) 430.62-Provide a description of the speed control unit, the load control unit and the flow control unit of the electro bydraulic control (EHC) system. Your description should include how they perform their intended functions. Clarify whether the EHC system will fully cut off steam at 103 percent of .
rated turbine speed. (10.2) 430.63 For turbine overspeed protection system (described in Section 10,2,2,4), the SSAR referred to redundant electrical trip signals. Provide information on the power source associated with each of the trip circuits. (10.2) 430.64 As presented in Section 10.2.2.4 of the ABWR SSAR, the closing time of the extraction nonreturn valves is less than 0.2 seconds, while it is 2 seconds'at current BWR plants. Provide additional information on the design of these valves that supports the difference between the above closing time s values. (10.2) 430.65 Clarify whether at least one main stop valve, reheat stop valve and reheat intercept valve will
- be inspected at approximately 31/3 years by dismantling them, and whether visual and surface examinations will be conducted for the valve seats, disks and stems (note that the above is an acceptance criterion for SRP Section 10.2). (10.2) 430.66
' identify preoperational and startup tests of the turbine generator in accordance with Regulatory Guide 1.68," Initial Test Programs for Water Cooled Power Plants," as an interface requirement. -
- (10.2) .
Amendment 11 20.2-16.1 l
_ . _ _ _ _ . . . . . . =
_ _ _ _ _ . _ = _ . . . . . . . . . . . . . . ..
- 4
.Qg gg Remndard Plant REV B -
, i G 47 J .
As stated in Section 103.2.1,'the four main steam lines are connected to a header upstream of the turbine stop valves..,*. However, according to Figure 103 2a, the main steam header is located
. downstream of the turbine stop valves. Identify whether the statement or figure is in error and revise the item in error so that the SSAR is consistent. (103) 6 48' ProvMe information on the' leakage detection system for steam leakage from the MSSS in the event
- of a steam line break. Also provide information on the stated ' safety feature designed into the MSSS* that will prevent radiation exposures in excess of the limits of 10 CFR Part 100 in the event of a break of a main steam line or any branch line (SSAR Section 1033). (103)
G.69 For the following items identified in SSAR Figure 103-1: (103)
(a) Deaerating steam to condenser (b) Offgas system (c) . Steam jet air ejectors (d) Turbine gland sealing system (e). Reheater (f) . Main steam bypass Provide the followinginformation:
(a)- Maximum steam flow (lbs/hr)
(b) Type of shut off valve (s)
(c) Size, quality, design code, closure time, actuation mechanism and associnted motive power of the valve (s).
430.70 Provide information on the following items: (103)
(a) J Analysis for steam hammer and relief valve discharge loads issues.
1 (b) Power source to the solenoid valves for the inboard and outboard main steam isolation -
valves.
(c) Location of seismic interface restraint (e.g., interface of which buildings?).
(d) Route which the main steam lines, including the branch lines, pass up to the turbine stop valves.
(c) Specific design features provided to protect safety related portions of the main steam supply system, including the main steam isolation valves, against externally and internally generated missiles and adverse phenomena such'as floods, hurricanes and tornadoes.
G.71 Describe provisions for operation of the main condenser with (10.4.1) leaking condenser tubes.
O Amendment 11 20.2.E2 i------.--i... .i
ABM 23&sioo4r -
- Standard Plant REV.B 430.72 -
Describe the permissible cooling water inleakage rate and the allowed time of operation with inleakage. (10.4.1) 430.73
-l 1
Provide information on the following items: (10.4.1)
(a) Provisions incorporated into the main condenser evacuation system component or tube failure due to steam blowdown from the turbine bypass system.
(b) Worst possible flood level in the applicable buildings due to complete failure of main condenser and provisions for protecting safety related equipment located in the buildings against such flooding (note that ABWR SSAR Section 3.4 does not discuss the I; turbine building). l 430.74 Discu.;s how the components of the main steam condenser evacuation system (MCES) conform to the guidelines of Regulatory Guide 1.26,1.33 and 1.123 with respect to quality group classification and '
quality assurance programs.
l 430.75 j Provide the design pressure and normal operational absolute pressure for the MCES components that could contain potentially explosive gas mixtures. (10.4.2) 430.76 Identify the radiation monitoring provisions for the mechanical vacuum pump exhaust. Is the exhaust filtered by charcoal adsorber and HEPA filters prior to release?
430.77
. Identify the number, location and functions (i.e., recording and annunciating alarm) performed by the hydrogen analyzers. Clarify whether they can withstand a hydrogen detonation. (10.4.2) 430.78' Clarify whether the air ejectors are redundant in the sense that one of them is a standby.
(10.4.2) ;
430.79 Identify the components and portions of the MCES that are designed to withstand a detonation in the system. (10.4.2) 430.80 '
Discuss how the design of the turbine gland sealing system (TGS$) conforms to the guidelines of Regulatory Guide 1.26 as it relates to the quality group classification for the system, and Regulatory Guide 1.33 and 1.123 as they relate to the quality assurance programs. (10.4.3)
- - ,, .., s,
': - -MM 2 w too n ~
Standard Plant arv.a 430.81 Provide a description of the exhauster blower provided for the TGSS. (10.43) 430.82 -
g ABWR SSAR Subsection 10.43.1.2 states that the TGSS exhau'sts the noncondensable gases to the
. [/ turbine building equipment vent system; however, Subsection 10.4.3.3 states that the TGSS exhausts the noncondensable gas gases eventually to main vent. Clarify how the TGSS exhausts are monitored.
Also, clarify whether the main vent mentioned above is the plant vent referred to in SSAR Section 11.5. (10.4 3) 430.83 What is the source for the auxiliary steam? Justify why an advanced design will use essentially radioactivity free auxiliary steam (see SSAR Section 10,4 3.2.2) as a backup sealing source rather than as normal sealing source.= Note that the use of a process steam supply for sealing purpose can result in significant operational radioactivity releases. (10.43) 430.84 For turbine bypass system:
(a) Provide figures which delineate the system and its components.
m (b) Clarify whether the system includes pressure reducer assemblies for the bypass valves to i reduce steam pressure prior to steam discharge into the condenser. (10.4.4) 6e -
- For the circulating water system:
(a) Describe the function of the waterbox fill and drain subsystem mentioned in ABWR Subsection 10.4.5.2.1. Also, describe the " makeup water" shown in SSAR Figure 10.4 3.
5 ' (b) Provide the worst possible flood levels that can occur in the applicable plant buildings as a result of circulatory water system failure and indicate how safety related equipment located in the buildings is protected against such flooding.
430.86 -
How is the remote manual motor operated shutoff valve (gate valve F 2820) powered?
430.87 Describe the design features provided to protect the safety related portion of the condensate and feedwater system from laternally generated missiles.
430.88 Provide a summary of the analysis of a postulated high-enegy pipe break for the feedwater piping in the main steam tunnel including the design features provided (e.g., pipe whip restraints) for preventing adverse effects resulting from pipe whip, jet impingement and flooding.
O Amendment 11 20.2-16A E
y7 _ _ __ .-
in ABWR - '
w1w Standard Plant - nry. n
- 4 0.89 .
i 1
Provide information on the analysis that shows that the entire feedwater system piping can.
- j. accommodate water hammer events and the means to prevent water hammer loads due to hydraulic 1 transients.
4 0.90 Provide detailinformation on the feedwater control valve and controller design, including the .j features that ensure the design will be stable and compatible with the system and imposed operating - i conditions. l l
l l
i
'l ll i !
t- '
!~ -
e 1.
l l
l l
l I
Amendment 11 20.2-163 e- * , . ,
M
.-- lABWR mam r -
Ramadard Plant REV H SECTION 20.3 ILLUSTRATIONS (Continued)
Figure Dtle P. age 203-29 Flow Ouj of Vessel Following a 0.0218 ft Vessel Bottom Head Drainline Break:1 RHR + 3 ADS Available- 20S284 203-30 ' FlowIntg VesselFollowing a .
0.0218 ft Vessel Bottom Head Drainline Break:1 RHR + 3 ADS Available 20S284 203-31 Peak Clafding Temperature Following a 0.0218 ft Vessel Bottom Head Drainline Break:1 RHR + 3 ADS Available 203 285 203-32 Wateg 14wl in Fuel Channels Following a 03 ft Break in the RHR Vessel Shutdown Suction Line:1 RHR Available 20 3-285 203-33 Wateg level Inside Shroud Following a -
03 ft Break in the RHR Vessel Shatdown Suction Line: 1 RHR Available 203-286 203 34 Wateg Level Outside Shroud Following a 03 ft Break in the RHR Vessel Shutdown Suction Line:1 RHR Available 203 286 203 35 Vessp Pressure Following a 03 ft Break in the RHR Vessel Shutdown Suction Line:1 RHR Available 203-287 20 3-36 Flowput of Vessel Following a 03 ft Break in the RHR Vessel Shutdown Suction Line:1 RHR Available 20 3-287 20 3-37 Flowjnto Vessel Following a -
03 ft Break in the RHR Vessel Shutdown Suction Line:1 RHR Available 203-288 .
20 3-38 Peakpladding Temperature Following a 03 ft Break in the RHR VesselShutdown Suction Line: 1 RHR Available 20 3 288 20 3-39 Reactor Water Cleanup System (Response to Question 72532) 20 3-289 20 3-40 Characteristic Response for Damped and Undamped System (Response to Question 725 53 203 290 203-vi Amendment 10
' MN - 23A6100A7.
Standard Plant nry n ILLUSTRATIONS (Continued)
, - n e-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 Rl!R System 203 288.1 20 3-43 Simplified Functional Control Diagram of Turbine Digital Control & Monitoring .
System (Typical of One of Three Channels) 203-310 9
s
-j.
I 203-vil Amendment 11 m
s.
'/9
+
MM EsamAard Plant 23A610aAT am a TABLEI Comparison of nguiremer.ss la ABWR plandard safety analyses REPORT -j and ABWR presentatles to NRC stmN (October 21 and 22,1987) (contissed) j ABWR Presentation ABWR Standard Safety to NRC StaN Analysis Report .
?
1 7 Highly corrosion- Required Design Feature Not discussedin resistant condenser Sulsection 5.2.12.2.3.
tubes to minimize leakage into condensate system 8- Maintain electrochemical Required Design Feature Notlisted in corrosion potential Table 5.2 5.
< 0.23 V to suppressIGSCC j
^
9 Erosion / corrosion- Design Feature Not discussed in resistant materials Subsection SA.9. l in steam extraction and +
drain linena minimize O failures
] Design Feature May be in Subsection 10A.1 l - 10. Ease ofleak detection in and repair of the which has not been !
main condenser submitted yet.
E 11 2% Reactor water cleanup Design Feature Not discussed m system to improve water Subsection 5.2.3.2.23 quality and occupational ic radiation exposure I
- 12. Fullflowrecirculation Design Feature Not discussed in Subsection 5.2.3.2.2.3. ;
to main condenser from '
cleanup outlet to reduce feedwater impurities ;
20.3 13 Amendment 9
y t, N' ' ABWR'-
- fiandard Plant , _ , . ,
2mmr REY.D ItESPONSE 281.10 6
Item 1 i, 4 '
Response to item 1 of this question is provided in revised Subsection 5.23.2.2.2.
V Item 2 g Response to item 2 of this question is provided in revised Subsection 5.23.2.2.
Item 3 l The system for adding zine to the feedwater is discussed in new Section 93.11.
Item 4 The system whleb luctudes a full flow deep bed condensate treatment system is discussed in )
revised Subsection 5.23.2.23 and new Subsection 10.4.6.
Item 5 New and improved water quality monitoring instrumentation is being constantly developed and introduced for use in BWR plants. Several useful instruments have been developed and introduced within the past few years. GE will evaluate the state of the art when a'BWR is undergoing detailed design and willincorporate such instruments that are necessary to assure proper water '
quality,- -
Item 6 Response to item 6 of this question is provided in revised Subsection 5.23.2.23.
Item 7 Response to item 7 of this question is provided in revised Subsection 5.23.2.23.
Item 8 Response to item 8 of this question is provided in revised Subsection 5.23.2.2.2 and Table 5.2 5.
Item 9 Response to item 9 of this question is provided in revised Subsection 5.23.2.23.
O Amendment 11 20.3-14
n
!- l ABM 2mioarr l mandmed Plant REV B (3/
/
Item 10 Response to item 10 of this question is provided in revised Subsection 5.23.2.23.
Item 11 !
(
Response to item 11 of this question is provided in revised Subsection 5.23.2.23. ;
Item 12 l The condensate treatment system which includes full flow recirculation to main condenser from condensate cleanup outlet is discussed in revised Subsection 10.4.6.2.
i i
QUESTION 470.1 Subsection 15.6.2 of the ABWR FSAR provides your analysis for the radiological consequences of a l failure of smalllines carrying primary coolt.nt outside of containment. This analysis only considers the failure of an instrument line with a 1/4. inch flow restricting orifice. Show that this failure scenario provides the most severe radioactive releases of any postulated failure of a smallline.
Your evaluatica should include lines that meet GDC 55 as well as small lines exempt from GDC 55.
RESPONSE 470.1 l- The analysis for failure of a smallline carrying primary coolant was conservatively analyzed as a .
i failure of an instrument line with full flow for a period of two hours. This analysis is deemed conservative for the reason given below.
(O~] Of all the lines carrying coolant penetrating the primary containment wall, only the instrument .
lines are exempt from GDC 55. All other lines use some form of check valve / motor operited valve
+
combination to stop the flow of primary coolant in the event of a line break. Typically, the ,
i motor operated valves close at the rate of two inches per ten seconds. Considering a two inch line l and assuming that a flow of 175 pounds per second would result in operator cetion within 60 seconds, the total mass released over the 70 second period would be approximately 12,000 pounds or about one j half of the assumed release over two hours from the instrument line. Using this logic and these simplified calculations,it is found that a two-hour instrument line break bounds releases for small l' lines.
l l QUESTION 470.2 l
l Provide a justification for your assumption that the plant continues to operate (and therefore no
-iodine peaking is experienced) during a smallline break outside containment (Subsection 15.6.2) ;
accident scenario. Also provide the basis for the assumption that the release duration is only two ,
hours.
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21A6100AT Erandmed Pimmt arv s RESPONSE 470.2 The analysis for failure of a smallline carrying primary coolant was based upon considering the plant remaining at full power for a period of two hours at which time flow was stopped. For 4
conservative purposes, the release was considered instantaneous in the actual computations. These parameters were chosen for conservatism and case of computation. The actual case of the rupture of an instrument line is described in Chapter 8 of NEDO 211431 (Reference 2 of SSAR Subsection 15.6.7) and results in full flow for approximately ten minutes following operator action and gradual depressurleation over a five hour period. The total mass of liquid released is approaimately 12,000 pounds or one half of the assumed release analysis, la addition, lodine spiking is considered on a release per fuel bundle basis. With the spiking term, which is estimated as a 15% initial release following release of the remaintne 85% proportional to the depressurization,it is found that the results are similar to those ar= iza Section 15.6 but slightly less conservative.
QUESTION 470.3 ; ,
Subsection 15.6.4.5.1.1 of the,9 .4 gives the iodine source term (concentration and isotopic mix) used to analyze the steam line/u,cak cetside of containment accident. The noble gas seurce term, however,is not addressed. Provide the noble gas source term used. Also the table in $ubsection 15.6.4.5.1.1 seems heavily weighted to tbc shorter lived activilles (i.e.,1134). Provide the bases for the isotopic mix used in your analysis (iodine and noble gas).
RESPONSE 470.3 Subsection 15.6.4.5.1.1 states that for case 1 the noble gas source term used was equivalent to an offgas release of 50,000 microCuries per second and 300,000 microOuries per second for case 2. In both cases, the source term is referenced to a 30. minute decay time. The isotopic distribution for such source terms are relatively standard throughout the industry and can be found in Table 2 2 of NUREG 0016. For the iodine isotopes the concentrations are technical specification limits of 0.2 microCuries per gram (case 1) and 4 microCurles per gram (case 2) dose equivalent to 1131.' The isotopic breakdown is based upon evaluations of BWR iodine chemistry in the early 1970's and is given in Reference 2 of SSAR Subsection 15.6.7. Tbc breakdown is as follows, and is similar to that found in Table 2 2 of NUREG 0016:
1131 0.073 1132 0.71 1133 0.5 1134 1.4 1135 0.73 QUESTION 470.4 Subsection 15.6.5.5 states that the analysis is based on assumptions provided in Regulatory Guide 1.3 except where noted. For all assumptions (e.g., releare assumed to occur one hour after accident initiation, the chemical species fractions for lodine, the temporal decrease in primary containment leakage rates, credit for condenser 1r.akage rates, and dose conversion fr.ctors) which deviate from NRC guidance such as regulatory guides and ICRP2, provide a detailed description of the justification for the deviation or a reference to another section of the SSAR where the deviations are discussed in detail. Provide a comparison of the dose estimates using these assumptions versus those which result from using the NRC guidance.
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Etandard Plant _ nrv. n QUESTION 00.44 identify the system lines whose containment isolation requirements are covered by Of>C 57 and ,
discuss conformance of the design to the GDC requirements. (6.2)
RESPONSE 4M.44 GDC $7 addresses closed loop systems which penetrate the containment but do not communicate with the containment interior. The system lines shown in Table 20.3 5 have been identified and are considered to conform to GDC57 with the valve configuration as shown. The heavy lines denote an '
extension of the containment boundry.
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1 QUESTION 430.45 For the combustible gas control systems design, identify clearly those areas that may not be part of *.he ABWR scope and provide relevant interface requirements. (6.2) ,
RESPONSE 430.48 -
The combustible gas control systems, consisting of flammability control system (FCS T49) and :
atmospheric control system (ACS T31), are completely within the scope covered by the ABWR SSAR. As such, there are no interfaces with equipment or systems outside the scope of this submittal. ?
Interfaces with systems or equipment within the scope of the SSAR are discussed, as necessary in .
Subsection 6.2.5. ,
QUESTION 430.46 ;
According to SRP 6.2.5 specific seceptance criteria related to the concentration of hydrogen or orygen in the containment atmosphere among others are the following:
(a) The analysis of hydrogen and oxygen production should be based on the parameters listed in Table 1 of Regulatory Guide 1.7 for the purpose of establishing the design basis for j combustible control systems. ,
(b) The fission product decay energy used in the calculation of hydrogen and oxygen production from radiolysis should be equal to or more conservative than the decay energy model given in Branch Technical Position ASB9 2 in SRP 9.2.5. (
Provide justification that the assumptions used in the ABWR in establishing the design basis for the combustible gas control systems are conservative with respect to the criteria a. and b. above. ;
(6.2)
RESPONSE 430.46 The analysis of hydrogen and oxygen production for ABWR combustible gas control design is besed on the parameters listed in Table 1 of Regulatory Guide 1.7. The fission product decay energy model used is that presented in SRP 9.2.5 Branch Technical Position ASB 9.2. Therefore, the ABWR design basis for combustible gas control system is conservative and appropriate for design.
QUESTION 430.47 Provide an analysis of the production and accumulation of combustible gases within the containment following a postulated loss of coolant accident including all applicable information specified in +
Section 6.2.5.3 of Regulatory Guide 1.70, Revision 3. ,
RESPONSE 430.47 Figure 6.2 41 provides an analysis of the oxygen and hydrogen concentrations in the primary containment after the design basis accident. Inputs to the analysis are provided in a revised j Subsection 6.2.5.3. i QUESTION 430.48 Regarding Containment Type A leakage testing. (6.2.6)
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QUESTION 00.48a }
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Provide the values for Pa and Pg.
i RESPONSE 430.48a -
P approximately 40 psig and 0.5 Pa< P <Pa- ;
QUEFTION 00.48b laclude the acceptance criterlon for L during preoperational leakage rate tests, l.c., Lt
- l La (L tm/ Lam), for the case when La (L im/ Lam) = .0.7. .!
RESPONSE 40.48b 1 Response to this question is provided in revised Subsection 6.2.6.1.1.5. ,
QUESTION 430.48c Your acceptance criterion for L m i t (SSAR Subsection 6.2.5.1.2.2, item 1) is at variance with the staff's current practice for acceptance of Lt m+ Also. It does not comply with the 10 CFR Part 50,
- Appendix J,Section III, item A.1.(a) requirement. Therefore, either provide sufficient supporting justification for the exemption from compliance with the above requirement or correct the criterion '
as appropriate to comply with the requirement. Also, correct the stated acceptance criterion (SSAR Subsection 6.2.6.1.2.2, item 3) as appropriate to comply with Appendix J, Section !!!, Jtem A.6.(b) requirement.
O RESPONSE 430.48c V
Response to this question is provided in revised Subsection 6.2.6.1.2.2. ,
QUESTION 430.48d and 430.48e ,
i Regarding ILRT, identify the systems that will not be vented or drained and provide reasons for ,
the same.
RESPONSE 430.48d and 430.48e Table 3.611ists essential systems that become available to shut down the reactor and mitigate the consequences of a postulated piping failure to acceptable limits. P&lDs and process flow '
diagrams for these systems are contained in their respective chapters. With regrad to integrated leak rate test (ILRT), the criteria described in Table 3.6 7 is applied to determine which systems will or will not be vented or drained. Provisions for venting / draining affected systems are shown on ;
the P&lDs listed in Table 3.6-7.
QUESTION 430.49 L
Regarding Type B test,(6.2.6) l l
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QUESTION 430.49a Clarify how air locks opened during periods when containm:nt integrity is required by plant's Tech- ,
nical Specifications will be tested to comply with Appendix J, Section Ill, item D.2.(b),(iii). ;
- i. RESPONSE 430.49a Response to this question is provided in resed Subsection 6.2.6.23.
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. QUESTION 430.49b :
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Provide the frequency for periodic tests of air locks and associated inflatable seals.
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RESPONSE 430.49b i Response to this question is provided by revit.cd Subsection 6.2.6.23.
l QUESTION 430.49c {
Provide the acceptance criteria for lock testing and the associated inflatable seal testing. j RESPONSE 430.49c l Response to this question is provided in revised Subsection 6.2.6.2.2.
QUESTION 430.49d .
List all containment penetrations subject to Type B tests.
t RESPONSE 430.49d Response to this question is provided in revised Subsection 6.2.611 and new Table 6.2 8.
QUESTlON 430.49e i List all those penetrations to be excluded from Tyoc B testing and the rationale for excluding f them.
RESPONSE 430.49e i
Response to this question is provided in revised Subsection 6.2.6.2.1 and Table 6.2 8.
QUESTION 430.50 ,
Regarding Type C tests (6.2.6) ;
QUESTION 430.50s -
Correct the statement (Subsection 6.2.6.3.1, Paragraph 1) as appropriate to ensure that the hydrau- .
lic Type C tests are performed only on those isolation valves that are qualified for such tests per !
Appendix.J. The current statement implies that these tests are not necessarily restricted to the !
valves that qualify for such tests. -
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I QUESTION 43034o Section 6.4.7.1, External Temperature. *provides design maximum external temperature of 2000F j and 100F. How are these values used in the design and assessments related to the ABWR7 What ;
factors, such as insulation, heat generation from control room personnel and equipment and heat losses, are taken into account? Do these values reprnent ' instantaneous' values or are they temporal and/or spatial averages?
RESPONSE 430440 1
These values represent the summer maximum dry bulb air temperature. They are used in sizing the j HVAC essential chilled water system chillers and the control room HVAC system.
QUESTION 43054p ,
Clarify your position on potential hazardous or toxic gas sources onsite of an ABWR. If l- applicable, indicate the special features provided in the ABWR design in this regard, to ensure control room habitability.
RESPONSE 43054p p Response to this question is provided in revised Subsection 6.4.7.3.
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s' QUESTION 43054q Identify all the interface requirements for control room habitability systems (e.g.,
instrumentation for protection against toxic gast,s in general and chlorine in particular; potential toxic gas release points in the environs). -
RESPONSE 43034q The ABWR control room habitability system has no interface requirements.
QUESTION 430$$
t Regarding ESF Atmosphere Cleanup Systems, (6.5.1) f QUESTION 43035a Provide a table listing the compliance status of the Standby Gas Treatment System (SGTS) with tach of the regulatory positions specified under C of RG 1.52. Provide justifications for each of those items that do not fully comply with the corresponding requirements,10 this context, you may
- note that the lack of redundancy of the SGTS filter train (the staff considers that filter trains are also active components . See SRP 6.4, Acceptance Criterion II.2.b) is not acceptable. Further, the l described sizing of the charcoal adsorbers based on assumed decontamination factors for various chemical forms of lodine in the suppression pool is not acceptable (RG 1.3 assumes a decontamination factor of 1 for all forms of iodine and RG 1.52 requires comp'iance with the above guide for the design of the adsorber section). Therefore, revise charcoal weight and charcoaliodine loading given
.O' in SSAR Table 6.51 as appropriate.
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The response to the first part of this question is provided in Appendix 6A. Justification of j single filter train is provided in revised Subsection 6.5.1.3.3. The iodine source term is discussed in Subsections 6.5.1.3.3 and 6.5.1.3.4. Tables 6.51 and 6.5 2 have been revised.
QUESTION 430J8b Specify the laboratory test criteria for methyllodine penetration that will be identified as an h
laterface requirement to be qualified for the adsorber efficiencies for lodine given in SSAR Table 15.6 8. Also, provide the depth of the charcoal beds for the control room emergency system.
RESPONSE 430J8b The response to the first part of this question is protided in Appendix 6A which assesses compliance against Regulatory Guide 1.52, Positions C.3.1, C.6.a (2) and C.6.a(3).
Control room HVAC charcoal bed depth is discussed in Subsection 9.4.1.1.3. J QUESTION 43035c Provide a table listing the con.pliance status of the instrumentation provided for the SGTS for read out, recording and alarm provisions in the control room with ush of the instrumentation items ,
identified in Table 6.5.1 1 of SRP 6.5.1. For partial or non compliance items, provide
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justifications.
RESPONSE 430J$c The response to this question is provided in Appendix 6B.
QUESTION 430J$d i Clarify whether primary containment purging during normal plant operation when required to limit the discharge of contaminants to the environment will always be through the SGTS (See SSAR Section 6.5.1.2.3.3). Clarify whether such a release prior to the purge system isolation has been considered in the LOCA dose analysis.
1 RESPONSE 43035d The response to this question is provided in Subsection 6.5.1.3.6. Note that Subsection 6.5.1.2.3.3 has been renumbered io 6.5.1.2.3.2.
QUESTION 43035e Provide the compliance status tables referred to in items (a) and (c) above for the control room ESF filter trains. (The staff notes that you have committed to discuss control room ESF filter 1 system under Section 9.4.1. However, since evaluation of the control room habitability system cannot i be completed until the information identified above is provided, the above information is requested -j now.) i RESPONSE 430J5e The response to this question is provided in Subsections 9.4.1.1.6.2 and 9.4.1.1.6.3. l 1
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QUESTION 430.S$f i
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i Identify the applicable interface requirements for the SGTS and the control room ESF atmosphere l cleanup system. ]
RESPONSE 40J5f !
Both the SGTS and the control room ESF atmosphere cleanup system are completely within the scope covered by the ABWR SSAR. As such, there are no inmfaces with equipment or systems outside the scope of this submittal. Interfaces with systems or equipment within the scope of the SSAR are discussed, as necessary,in Section 6.4 and Subsections 6.5.1 and 9.4.1.1. Testing requirements are :
described in the standards and Regulatory Guides referenced in these sections.
QUESTION 00.56 Regarding Fission Product Control Systems and Structures, (6.53) i QUESTION 00.56a Provide the drawdown time for achieving a negative pressure of 0.25 inch water gauge for the sec-ondary containment with respect to the environs during SGTS operation. Clarify whether the ,
unfiltered release of radioactivity to the environs during this time for postulated LOCA has been ,
considered in the LOCA dose analysis. (Note that the unfiltered release need not be considered provided the required negative pressure differentialis achieved within 60 seconds from the time of i the accident).
RESPONSE 40.56a The response to this question is provided in Subsection 6.5.13.2.
! QUESTION 430.56b l
l Provide justification (See SRP Section 6.53,11.4) for the decontamination factor assumed in SSAR Tables 6.5 2 and 15.6 8 for lodine in the suppression pool, correct the elemental, particulate and organic lodine fractions given in tables to be consistent with RG 1.3, and incorporate the correction -
in the LOCA analysis tables. Alternatively, taking no credit for iodine retention in the suppression pool, revise the LOCA analysis tables. Note that the revision of the LOCA analysis tables (this also ,
includes the control room doses) mentioned above is strictly in relation to the iodine retention l
factor in the suppression pool (also, there may be need for revision of other parameters (s) given in L the tables and these will be identified under the relevant SRP Sections questions). t RESPONSE CO.56b l Table 6.5-2 has been revised to be consistent with the assumptions of Regulatory Guide 13 L
l, With regard to Table 15.6-8, the LOCA analysis is performed in accordance with paragraph 8.9 of .
the . Licensing Review Bases document. An evaluation of suppression pool scrubbing using the MAAP3B code for LOCA conditions shows a scrubbing factor of 600 to 1000 (Subsection 19E.2.1). Therefore the use of a scrubbing factor of 100 is sufficiently conservative. Tbc variance between the current calculations and the prior evaluation methodologies is found in Table 20.31. !
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C i F- 21A6100AT Remndard Plant ma. m QUESTION WJSc Identify the applicable interface requirements 9i !
RESPONSE W.56c tlc response to this question can be found in the response to Question 430.55f.
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% QUESTION 2103 In Subsection 3.1.2.1.1.2, ' Evaluation Against Criterion l', a footnote states that
'important to safety
- and ' safety related* are considered equivalent in this SSAR. The staff does not agree with this definition. The staff's position on this issue rernmins as stated in NRC Generic i Letter 84 01, 'NRC Use of the Terms 'Important to Safety
- and
- Safety Related", dated January 5,1984. The staff used this position as guidance in its reviews of applications for operating licenses of nuclear power plants for a number of years prior to the issuance of GL 84 01. During these reviews, the staffs' evaluations of the quality assurance requirements in 10 CFR Part 50, .
Appendix B generally applied to the narrower class of ' Safety-related" equipment as defined in 10 CFR Part 50.49(b)(1).10 CFR Part 100, Appendix A and in Section 3.2 of this SSAR. This implied that normal industry practice for quality assurance was generally acceptable for most equipment not covered by the ' safety.related* definition. However, as pointed out in Generic Letter 84-01, there have been specific situations in the past where the staff has determined that quality assurance requirements beyond normalindustry practice were needed for components and equipment in the more broad *important to safety" class.
It is the staff's opinion that a strict interpretation of the ABWR position on this issue could result in an unacceptable classification of structures, systems and components for Table 3.21 in this SSAR>
I Revise the footnote in Subsection 3.1.2.1.1.2 and the discussion in Section 3.2 to be consistent I with the staff's position as stated in Generic Letter 84 01. It should be made clear that the staff's position will not result in a broadening of the staff's review. Rather, it provides the f-basis which the staff has been using and continues to use as guidance in its reviews of Ouality Group l-i Classification for certain components and equipment which are not included in the ' safety-related*
definition.
RESPONSE 2103 Subsection 3.1.2.1.1.2 has been raised and the footnote deleted. The ABWR is consistent with the definition of important to safety as defined in 10CFR50, Appendix A, General Design Criteria, as
- specified in the Nuclear Regulatory Commission's Memorandum and Order dated June 6,1984 (CLI 84 9).
The Commission identified the definition of important to safety as a generic issue to be resolved by l rule making. The Commission also stated that in the interim, past practice should be followed.
The past practice for all GE Nuclear Energy BWR design has been that all equipment has been identified as either safety related or non safety related. For certain non safety related equipment, the pertinent requirements (including quality assurance requirements) have been specified on a case by case basis commensurate with the functional importance of the equipment (e.g., fire protection and radioactive waste treatment systems). All prior GE BWR designs have been licensed on that basis. (This includes Shoreham for which this issue was specifically addressed and resulted in the Commission's June 6,1984 Memorandum and Order.) The ABWR is consistent with this prior practice as specified by the Commission's Memorandum and order.
QUESTION 210.4 In Subsection 3.2.3
- Safety Classifications", ANSI /ANS 52.11983, ' Nuclear Safety Criteria for the Design of Stationary BWR Plants'is referenced for the definitions of safety classes. The guidance in this document for components which are not within the scope of Regulatory Guide 1.26 has not been endorsed by the staff. Therefore, the staff does not completely accept ANSI /ANS 52.1 for Amendment 11 20.3 126
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the definitions of all safety classes. Questions 210.5, 210.13, 210.15, 210.17, 210.44, and 210.45 are based on this position. To assure that Table 3.21 will be consistent with similar tables in' ,
recently licensed BWR/6 plants, such as Perry and River Bend, the reference to ANSI /ANS 52.11983 -
should be either eliminated or revised.
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ABM isreioorr arv s SAndard Plant RESPONSE 210.4 The safety classification methodology used to classify the ABWR equipment is the same as that used for previous GE BWR designs. The safety class definitions for ABWR while based on AISI/ANS 52.1 1983 are consistent with this classification methodology and past des!gns. ANS adopted the GE BWR safety classification criteria and expanded them into LWR classification criteria. The ANS LWR classification criteria were reviewed by the NRC staff during development and resolutions for all substantive comments were developed at a meeting between the NRC staff and representatives of ANS on October 12,1982. AISI/ANS 52.1 1983 incorporated those resolutions. In addition, AISI/ANS $2.1 -
1983 is specified by the EPRI ALWR Requirements Program.
QUESTION 210.8 In Table 3.21, items 81.7,' Control Rods' and Bl.9,' Fuel Assemblies
- are classified as Safety Class 3, which is consistent with the criteria in the ANSI /ANS 52.1 1983 Standard. As stated in Question 210.4, the staff does not agree with all of the recommendations in that Standard. The staff position is that Control Rods and Fuel Assemblies should be Safety Class 2 and Quality Oroup B. To br consistent with this position and with staff reviews on recent BWR/6 plants, such as Perry and River Bend, revise Table 3.21 to change the classifications of the Control Rods and Fuel Assemblics from Safety Class 3 to 2 and add Quality Group B.
Questions 210.44 and 210.45 provides similar staff positions for item B1.5 Safety Related Reactor Internal Structures and Core Support Structures.
RESPONSE 2103 The safety requirements for non piping components such as cor. trol rods and fuel assemblies are the same whether they are designated Safety Class 2 or Safety Class 3. The Safety Class 3 designation for such items is based on a comprehensive, systematic rational. Designating the items Safety Class 2 instead of Safety Class 3 would not change the safety requirements applied to them but would cause inconsistencies with the rational. To maintain consistency, the Safety Class 3 designation will be retained.
QUESTION 210.6 In Table 3.2.1, item B2.5 identifies Main Steam Line (MSL) piping from the outermost isolation valve to and including the seismic interface restraint as being Safety Class 1 and Quality Group A.
Figure 5.13b,' Nuclear Boiler System P&lD, Sheet 2' identifies the same portion of the MSL as Quality Group B. Beyond the seismic interface restraint, the MSL piping is quality Group D, which is not acceptable to the staff. To be acceptable, the MSL should be classified as recommended in Standard Review Plant 3.2.2,' System Ouality Group Classification *, Appendix A,i.e., Quality Group B from the outermost isolation valve to the turbine stop valve. This staff position is based on the assumption that the ABWR MSL design differs from the BWR/6 design in that it does not contain a shutoff valve in addition to the two containment isolation valves. Revise Table 5.13b, Table 3.21, Subsection 3.9.3.1.3 and Subsection 5.4.9.3 to be consistent with the above staff position.
RESPONSE 210.6 Table 3.21, Subsection 3.9.3.1.3 and Subsection 5.4.9.3 are corrected. The MSL piping beyond the seismic interface restraint will be changed from Quality Group D to Quality Group B at the next revision of Figure 5.13b.
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handard Plant nv n QUESTION 210.7 O Item B2.5 in Table 3.21 does not appear to agree with Figure 5.13c,' Nuclear Boller System F&lD, Sheet 5*, item B2.5 states that piping in the Feedwater (FW) Systems from the outermost isolation valve to and including the seismic interface restraint is Safety Class 1 and Quality Group L Figure 5.13c shows the FW line as Quality Group A up to the first spring closing check valve outside containment (F262A). The FW piping is Quality Group B between valves F262A and F282A and Quality Group D beyond F262A. There does not appear to be a seismic restraint in Figure 5.13c.
Assuming that the ABWR FW line is similar to the BWR/6 designs,i.e., valve F282A is a shutoff valve in addition to the two containment isolation valves, the Quality Group classification of this line does not appear to be consistent with the guidelines of Standard Review Plan 3.2.2, Appendix B.
Revise Table 3.21, Figure 5.13c and Subsection 5.4.9.3 to be consistent with the staff position on ~
Quality Grr>up in SRP 3.2.2, Appendix B. The transition from Quality Group B to D should be at the seismic interface restraint rather than shutoff valve F282A, RESPONSE 210.7 Table 3.21 and Subsection 5.4.9.3 have been corrected. Figure 5.13c will be revised accordingly at its next revision as indicated in Figure 20.3 20.
QUESTION 210.8 In Table 3.21, Item B3.1, the primary side recirculating motor cooling system piping is classified as Safety Class 3 and Quality Group C. In Subsection 3.9.3.1.4, this piping is described as being designed to the ASME Code, Section !!!, Subsection NB 3600, which is comparable to Safety Class 1. In Figure 5.4 4, ' Reactor Recirculation System P&lD*, this piping is identified as Quality Group A. The staff's position is that this piping should be, as a minimum, Safety Class 1, Quality O Group A and meet the requirements of 10 CFR 50, Appendix B from the interface of the piping with the pump motor casing to and including the first pipe support. The remainder of this piping, should be the sarae Safety Class as the supported piping. Revise items B3.1 and B3.2 in Table 3.21 to be consistent with the staff position.
RESPONSF.210.8 The treirculation motor control system (RMCS) is classified Quality Group C and Safety Class 3 which is in accordance with the requirements of 10CFR50.55a. The RMCS, which is part of the reactor coolant pressure boundary (RCPB), meets 10CFR50.55a(c)(2). Postulated failure of the RMCS piping cannot cause a loss of reactor coolant in excess of normal makeup (CRD return or RCIC flow), and the RMCS is not an engineered safety feature. Thus, in the event of a postulated failure of the RMCS piping during normal operation, the reactor can be shutdown and cooled down in an orderly manner, and reactor coolant makeup can be provided by a normal make up system (e.g., CRD return or RCIC system).
Thus, per 10CFR50.55a(c)(2), the RMCS need not be classified Quality Group A or Safety Class 1.
Since the RMCS is not an engineered safety feature (e.g.,it does not provide emergency reactivity control, emergency core coolant, or primary reactor containment), the system need not be classified Quailty Group B or Safety Class 2. The RMCS is classified Quality Group C and Safety Class 3, however, the system is designed and fabricated in accordance with AMSE Boiler and Fressure Vessel Code, Section 111, Class 1 criteria as specified in Subsection 3.9.3.1.4 and Figure 5.414. A typographical error in Table 3.21, item B3.1, ' Piping. primary side, motor cooling *, has been corrected.
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Erandard Plant m,. g QUESTION 210.9 l In Table 3.21, and the classification summary for the Control Rod Drive Mechanism and the Low .!
Pressure Core Flooder System or provide a justification for not inc!uding this information. The staff position on the Safety Class of these systems is as stated in Question s 210.5 and 210.45.
RESPONSE 210.9 The classification summary for the control rod drive mechanism has been added to Table 3.21.
Portions of the control rod drive mechanism that are part of the reactor coolant pressure boundary are Safety Class 1. All other portions of the control rod drive mechanism are Safety Cless 3 (see
- response to Question 210.5). The low pressure core flooder system is a subsystem of the residual 5
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} Standard Plant RIV B heat removal system and its classification summary is included in items E1.1 through E1.15 of Table 3.21.
QUESTION 210.10 Provide the basis for all Control Rod Drive System valves (Item C1.1 in Table 3.21) to be classifit.d as Non Nuclear Safety and Non Scismie.
RESPONSE 210.10 item C1.1 of Table 3.21 has been clarified. All valves required to provide the scram function are part of the hydraulic control unit which is Safety Class 2. The hydraulic control unit is Item C1.4 in Table 3.2-1. All other valves do not perform a safety.related function and are Non. Nuclear Safety.
QUESTION 210.11 Provide the basis for portions of piping systems with the outermost isolation valves in the Residual Heat Removal System and the Reactor Core Isolation Cooling System (Items E1.3, E4.1, and E4.6 in Table 3.21) to be classified as Safety Class 2 and 3.
RESPONSE 210.11 Portions of piping of thu residual heat removal system, high pressure core flooder system, and reactor core isolation cooling system within the outermost isolation valve which are not part of the reactor coolant pressure boundary but are part of an engineered safety feature are Safety Class 2.
Examples are the suppression pool suction piping and containment spray piping. Portions of piping of
.- l the residual heat removal system, high pressere core flooder system, and the reactor core isolation cooling system which are part of the reactor coolant pressure boundary but are one (1) inch or less in diameter are Safety Class 2. Examples are instrument lines and drain lines. Items Bl.5, E1.3, E2.1, E2.2 and E2.5 of Table 3.21 have been clarified; Figures 5.410a and 6.2 38a will be corrected at their next revision (as indicated in Figures 20.3 20.1 and 20.3 20.2 and Subsections 9.5.1.2.4, 5.9.5.1.2.5, and 3.9.5.1.2.6 have been correeted.
QUESTION 210.12 Items E2.1 and E2.5 in Table 3.21 classifies some pumps and valves within the outermost isolation valves in the High Pressure Core Flooder System as Safety Class 2. Provide the basis for-this classification.
RESPONSE 210.12 See response to Question 210.11.
QUESTION 210.13 In Table 3.21, item F4.1,' Refueling Equipment Platform Assembly" is classified as Non Nuclear Safety. To be consistent with the staff position as stated in Ouestion 210.4 and with staff reviews on recent BWR/6 plants, such as Perry and River Bend, revise Table 3.21 to change this classification to Safety Class 2 and Quality Group B.
O Amendment 11 20.3 129
l 4 1
,- -MM 21A6100AT Standard Plant arv n 1
[3 V
RESPONSE 210.13 ;
Item F4.1 of Table 3.21, Subsection 9.1.4.1, and Table 9.12 have been revised to show that a ;
quality assurance program will be .ipplied to ensure that the design, construction and testing requirements for the refueling equipment platform assembly are met. This quality assurance ,
commitment is similar to the commitments made for other nonsafety related equipment such as fire i protection equipment, radwaste equipment and ATWS equipment, j The Non Nuclear Safety designation for the refueling equipment platform assembly is consistent with ANSI /ANS 57.11980 (that has been endorsed by SRP Section 9.1.1) ANSI /ANS 52.11983, the EPRI ALWR Requirements Progrant and past industry practice. Also,in accordance with past industry 1 practice, the assembly is Seismic Category I to prevent catastrophic collapse onto the reactor core .]
during a seismic event. The consequences of failure of this assembly are within acceptable limits for such an event.
j 1
QUESTION 210.14 If a Fuel'Ieansfer System or Tube is applicable to the ABWR, add the Classification Summary for this system under Item F4,
- Refueling Equ'pment' of Table 3.21.
RESPONSE 210.14 The ABWR design does not include a fuel transfer system or tube. The refueling arrangement and i process for the ABWR is the same as for BWR/3, BWR/4, and BWR/5. The spent fuel storage poolis at the refueling floor level. During refueling the reactor vessel is flooded up to the spent fuel storage pool level, a gate in the spent fuel storage pool is removed connecting it to the flooded ;
reactor vessel, and fuel is transferred underwater via the refueling platform.
QUESTION 210.15 6 In Table 3.21, items FS.1,' Fuel Storage Racks New and Spent
- and FS.2, ' Defective Fuel Storage Container" are classified as Non Nuclear Safety. Item FS.2 is also classified as l Non. Seismic. To be consistent with the staff position as statec in Question 210.4 and with staff reviews on recent BWR/6 plants, such as Perry and River Bend, revise Table 3.21 to change the classification of items F5.1 and FS.2 to Safety Class 3 and Quality Group C. In addition, change the seismic classification of item F5.2 to Seismic Category I and add 'B' in the Quality Assurance column for F5.2.
RESPONSE 210.15 -
l Items FS.1 and FS.2 of Table 3.21 and Subsection 9.1.2.1.3 have been revised to show that a quality assurance program will be applied to ensure that the design, construction and testing requirements are met. Since the equipment is not required to prevent or mitigate a design basis event,10CFR50 Appendix B quality assurance requirements are not required.
' The Non Nuclear Safety designation for fuel storage racks (new and spent) and the defective fuel storage container is consistent with AISI/ANS 57.11980 and ANS 57.2/ANSIN210-1976 (that have l been endorsed by SPP Sections 9.1.1 and 9.1.2), AISI/ANS.52.1 1983, the EPRI ALWR Requirements Program, and past industry practice. The fuel storage racks are Seismic Category I commensurate with their functionalimportance. The consequences of a credible failure of the racks and container are ,
within acceptable dose limits.
Amendment 11 20S130
I MM 2sA61ooAT Standard Plant arv. n ]
QUESTION 210.16 In Table 3.21, the following components in the Reactor . Water Cleanup System are correctly l classified as Quality Group C, but are also classified as Non Nuclear Safety: !
G1.1 Vessels G1.2 Regenerative Heat Exchanges G1.3 Cleanup Recirculation Pump G1.5 Pump suction and discharge piping beyond containment isolation valves G1.8 Non regenerative heat exchanger tube inside and piping and valves carrying process water . -
G1.11. Filter demineralizer holding pumps, valves and piping l
9 I
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i 9
L l-i l
1 l
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I 1-n 9
Amendment 11 20.31R1 I _ . -_ . ._-
ABM Standard Plant ursioorr REV.B l
I Ov To be consistent with the discussions in Subsections 3.2.2 and 3 2.3 and with the information in Tables 3.21 and 3.2 3, the staff is of the opinion that all of the above components should be j
i classified as Safety Class 3 in addition to Quality Group C. Revise Table 3.21, items Gill, G1.2, G1.3, G1.5, G1.8, and G1.11 to change the Safety Class from *N" to *3* or provide a justification for j mot doing so.
l RESPONSE 210.16 The piping portions of the reactor water cleanup system out to and including the outermost isolation valves are Safety Class 1. In accordance with past practice, the portions of the reactor water cleanup system beyond the outermost isolation valves are Quality Group C, Non. Seismic, Category 1, and Non Nuclear Safety. These latter portions are not part of the reactor t.oolant pressure ,
boundary or primary reactor containment and do not perform a safety related function. The ;
consequences of failure of those latter portions are within acceptable dose limits. The 3 classification summary for electrical equipment required for isolation has been added to Table 3.21.
QUESTION 210.17 In Table 3.21, items G2.3, ' Heat Exchangers*, G2.4,
- Pumps and Pump Motors *, G2.5,
- Piping.
t Valves *, and G2.7, *RHR Connections
- in the Fuel Pool Cooling and Cleanup System are all classified '
as Non. Nuclear Safety, which is consistent with the criteria in the ANSI /ANS 52.1 1983 Standard.
As stated in Question 210.4, the staff does not agree with all of the recommendations in that ,
Standard. The staff position is that all of the above items should be Safety Class 3, Seismic
. Category 1 and listed under Quality Assurance requirements of 10 CFR 50, Appendix P.. Regulatory
- l Positions C.2 in Regulatory Guide 1.26 and C.1 in Regulatory Guide 1.29 includes this position. To l be consistent with this position and with staff reviews on recent BWR/6 plants, such as Perry and River Bend, revise Table 3.21 to change the classification of items G2.3, G2.4, G2.5, aLd G2.7 from
' Non Nuclear Safety to Safety Class 3, add Seismic Category I and add *B' under Quality Assurance Requirement.
RESPONSE 210.17 ,
The fuel pool cooling and cleanup system is Non Nuclear Safety, Quality Group C in accordance with the SRP Section 9.1.3, AISI/ANS 52.1 1983, the EPRI ALWR Requirements Program and past l-industry practice. The spent fuel poolis Safety Class 3, has a Seismic Category I makeup water system and source (i.e., the RHR System) is housed in the Safety Class 3 secondary containment, and has a Safety Class 3 ventilation system. Item G2.7 of Table 3.21 has been corrected to show that the RHR Connections are Safety Class 3.
l QUESTION 210.18 A staff position is that piping and valves forming part of primary ec tainment boundary should be Seismic Category 1. In Table 3.21, piping and valves in the Reactor Building Cooling Water System which form part of the primary containment boundary are classified as Non Seismic. Revise Table 3.21 to add Seismic Category I to the classification of item P2.1 or provide a justification for not doing so. .
RESPONSE 210.18 Item P2.1 of Table 3.21 has been corrected.
20.3 t31 Amendment 3
i
- .- 3 Generol FActric Company r ABWR raorniorray isronurrion 234n0041 Standard Plant am rAn 7
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Figure 20.3-41 PRES 5URE TIME HISTORY AFTER A FEEDWATER LINE AREAK i AVAILABLE ECOS: 1 RHR SYSTEM f
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TIME-SECONDS 4.5 (,. 10' Figure 20.3-42 TEMPERATURE TIME HISTORY AFTER A FEEDWATER BREAK O- AVAILABLE ECCS: 1 RHR SYSTEM V
Amendment 11 20.3290.1
. . ABWR na6iooxx e Standard Plant arv.n SECTION 11.5 fj,., CONTENTS Section Title P. ass 11.5.1 Design Bases 11.5 1 11.5.1.1 Design Objectives 11.5-1 11.5.1.1.1 Systems Required for Safety 11.5 1 11.5.1.1.2 Systems Required for Plant Operation 11.5-1 11.5.1.2 Design Criteria 11.5-2 11.5.1.2.1 Systems Required for Safety 11.5 2 11.5.1.2.2 Systems Required for Plant Operation 11.5 2 11.5.2 System Descriction 11.5-3 11.5.2.1 Systems Required for Safety 11.5 3 11.5.2.1.1 Main Steamline Radiation Monitoring System 11.5-3 11.5.2.1.2 Containment liVAC Radiation Monitoring System 11.5 3 11.5.2.1.2.1 Operating Floor llandling Area Radiation Monitoring Subsystem 11.5-4 11.5.2.1 3 Fuel Area Ventilation Exhaust Radiation Monitoring 11.5-4 11.5.2.1.4 Standby Gas Treatment Radiation Monitoring System 11.5-4 11.5.2.1.5 Control Building IIVAC Radiation Monitoring System 11.5-5 11.5.2.2 Systems Required for Plant Operation 11.56 13.5.2.2.1 Offg .s Pretreatment Radiation Monitoring System 11.5 6 11.5.2.2.2 Offgas Post Treatment Radiation Monitor 11.5-6 11.5.2.2 3 Carbon Bed Vault Radiation Monitoring System 11.5-7 11.5.2.2.4 Plant Vent Discharge Radiation Monitoring System 11.5-8 11.5.2.2.5 Liquid Process and Effluent Monitoring Systems 11.5-8 I I
l
\ 11.5 li l
Amendment 9
r ABWR msmx . o Standard Plant RIN. D SECTION 11.5 CONTENTS (Continued) $<
Section 3111e East 11.5.2.2.5.1 Radwaste Effluent Radiation Monitoring System 1158 11.5.2.2.5.2 Reactor Building Cooling Water Radiation Monitoring System 11.5 9 11.5.2.2.53 Dcleted .
11.5.2.2.5.4 Deleted 11.5.2.2.6 Radwaste Building HVAC Radiation Monitoring System 1159 ILS.2.2.7 Offgas Vent Radiation Monitoring System 11.5 9 11.53 EI11oent Monitorine and Samnline 1159 11.5 3.1 Basis for Monitor Location Selection 11.5-10 11.5 3.2 Expected Radiation Levels 11.5-10 11.533 Instrumentation 11.5 10 11.5 3.4 Setpoints 11.5-10 11.5.4 Process Monitorine and Samnline 11.5 10 11.5.4.1 Implementation of General Design Criterion 60 11510 ,
11.5.4.2 Implementation of General Design Criterion 64 11.5-10 11.5,13 Basis for Monitor Location Selection 11510 11.5.4.4 Expected Radiation Levels 11.5-10 11.5.4.5 Instrumentation 11.5 10 11.5.4.6 Setpoints 11.5-11 11.5.5 Cn111> ration and Maintenance 11.5-11 l 11.5.5.1 Inspection and Tests 11.5-11 l
l 11.5.5.2 Calibration 11.5-11 11.5.5 3 Maintenance 11.5 12 l
l_ 11.5.5.4 Audits and Verifications 11512 115i!i Amendment 11
gg 2wstoaAK Standard Plant arv n n 11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND
. (Ui SAMPLING SYSTEMS 11J.1.1.2 Systems Required for Plant The process and effluent radiological Operation monitoring and sampling systems are provided to allow determination of the content of radioactive The main objective of operational radiation material in various gaseous and liquid process monitoring systems is to provide operating and effluent streams. The design objective and personnel with measurements of the content of criteria are primarily determined by the system radioactive material in all effluent and designation of either: important process streams. This allows demonstration of compliance with plant normal (1) instrumentation systems required for safety, operational technical specifications by or providing gross radiation level monitoring and collection of halogens and particulates on (2) instrumentation systems required for plant filters (gaseous effluents) as required by operation. Regulatory Guide 1.21. Additional objectives are to initiate discharge valve isolation on the 11.5.1 Design Bases offgas or liquid radwaste systems if predeter.
mined release rates are exceeded and to provide 11.5.1.1 Design Objectives for sampling at certain radiation monitor locations to allow determination of specific 11.5.1.1.1 Systems Required for Safety radionuclide content.
The main objective of radiation monitoring The radiation monitoring systems (RMS) systems required for safety is to initiate provided to meet these total plant objectives appropriate protective action to limit the are:
potential release of radioactive materials from V) 1 the reactor vessel and primary and secondary (1) for gaseous effluent streams containment if predetermined radiation levels are exceeded in major process /cffluent streams. An (a) plant vent discharge; additional objective is to provide control room personnel with an indication of the radiation (b) offgas exhaust vent; levels in the major process / effluent streams plus alarm annunciation if high radiation levels are (c) radwaste building ventilation; and detected.
(d) turbine building ventilation. .
The radiation mcnitoring systems (RMS) provided to meet these objectives are: (2) for liquid effluent streams (1) main steamline RMS; (a) radwaste effluent radioactivity.
(2) containment beating, ventilating, and air conditioning (HVAC) radiation monitoring system; (3) for gaseous process streams (3) fuel area HVAC radiation monitoring system; (a) offgas pre-treatment sampling; (4) control building HVAC radiation monitoring system; and (b) offgas post treatment sampling; and (5) standby gas treatment system vent radiation (c) carbon bed vault gross gamma ra:!iation monitoring system. levels. i l Amendment 11 11.5 1 l
ABWR 2-u nyn Standard Plant (4) for liquid process streams (10) register full scale output if radiation detection exceeds full scale; and (a) reactor building closed cooling water :
intersystem radiation leakage. (11) have sensitivities and ranges compatible with anticipated radiation levels. -
11.5.1.2 Design Criteria The applicable General Design Criteria of 10CFR50, Appendix A, are 1, 2, 3, 13, 20, 21, '
11 3.1.2.1 Systems Required for Safety 22,23,24, and 29. The systems shall meet the design requirements for Safety Class 2, Seismic The design criteria for the main steamline and Category 1, systems along with the quality containment ventilation exhaust plenum monitoring assurance requirements of 10CFR$0, Appendix B.
systems are that the systems shall:
11 3.1.2.2 Systems Required for Plant (1) withstand the effect of natural phenomena Operation (e.g., earthquakes) without loss of capability to perform their functions; The design criteria for operational radiation monitoring systems are that the systems shall:
(2) perform their intended safety functions in the environment resulting from normal and (1) provide continuous it.dication of radiation abnormal conditions (e.g., loss of HVAC and levels in the main control room; isolation events);
(2) provide warning of increasing radiation (3) m eet t he re1iabi!ity, testability, levels indicative of abnormal conditions by independence, and failure mode requirements alarm annunciation; of engineered safety features; (3) insof ar as practical, provide self-(4) provide continuous output on control room monitoring of components to the extent that panels; power failure or component malfunction .
causes annunciation and discharge valve (5) permit checking of the operational isolation channel trip; availability of each channel during reactor operation with provisions for calibration (4) monitor a sample representative of the bulk function and instrument checks; stream or volume; (6) assure an extremely high probability of (5) have provisions for calibration, function, accomplishing safety functions in the event and instrumentation checks; l
l of anticipated operational occurrences; (6) have sensitivities and ranges compatible (7) initiate prompt protective action prior to with anticipated radiation levels; and exceeding plant technical specification limits; (7) register full scale output if radiation i
detection exceeds full scale, l- (8) provide warning of increasing radiation l
levels indicative of abnormal conditions by The RMS monitoring discharges from the alarm annunciation; gaseous and liquid radwaste treatment system shall have provisions to alarm and to initiate (9) insof ar as practical, provide self- automatic closure of the waste discharge valve monitoring of components to the extent that on the affected treatment system prior to l power failure or comp nent malfunction exceeding the normal operation limits specified causes annunciation and channel trip; in technical specifications as required by Regulatory Guide 1.21.
11.5-2 O
Amendment 11 l; ._ _
. ABWR msima Standard Plant RfV r1 ]
radiation monitor and the common two pen trip circuit ici instrument trouble. The trip )
(T recorder. A' 120 Vac local bus supplies the circuits are capahle of convenient operational
'V' sample panel. verification by means of test signals or through the use of portable gamma sources. Power is l Each radiation monitor has three trip supplied from channel A of the containment HVAC l circuits: two upscale (high high.high and high), radiation monitoring system.
and one downscale (low)/ inoperative. Each trip is visually displayed on the radiation monitor.
The first three trips actuate corresponding main untrol room annunciators: offgas post treatment high.high high radiation, offgas post treatment i bigh radiation, and offgas post treatment i downscale. A trip circuit on the recorder actuates an offgas post treatment high.high radiation annunciator. High or low sample flow msasured at the sample panel actuates a main control room offgas vent pipe sample high low flow annunciator.
A trip auxiliary unit in the control room takes the high.high high (HHH) and downscale trip outputs and, if its logic is satisfied, initiates closure of the offgas system discharge and drain valves. The logic is satisfied if either two HHH, on one HHH and one downscale, or two downscale trips occur. The HHH trip setpoints are determined so that valve closure is initiated
(] prior to exceeding technical specification U limits. Any one high upscale trip initiates closure of offgas system bypass line valve and permits opening of the treatment line valve.
A vial sampler panel similar to the pre-treatment sampler panelis provided from grab sample collection to allow isotopic analysis and gross monitor calibration.
11.5.2.2.3 Carbon Bed Vault Radiation Monitoring System The carbon vault is monitored for gross gamma radiation level with a single instrument channel. The channel includes a sensor and converter, an indicator and trip unit, and a locally mounted auxiliary unit. The indicator and trip unit is located in the main control room. The channel provides for sensing and readout, both local and remote, of gross gamma radiatiog over a range of six logarithmic decades
, (1 to 10 mR/hr).
l l The indicator and trip unit has one adjustable l . upscale trip circuit for alarm and one downscale i Vb l'-
l Amendment 4 11,$.7 i
~
ABM 23461ooxx Standard Plant nrv. n 11.5.2.2.4 Plant Vent Discharge Radiation only background levels of radioactive Monitoring System materials. Increases in radiation level may be indicative of heat exchanger leakage or This system monitors the plant vent discharge equipment malfunction.
for gross radiation level and collects halogen and particulate samples. A representative sample power is supplied from 120-Vac non-divisional is continuously extracted from the ventilation buses for the radiation monitors and recorders ducting through an isokine. tic probe in accordance and from a 120 Vac local bus for the sample with ANSI N13.1 passed through the containment panels.
ventilation sample panel for monitoring and samp-ling, and returned to t'ne ventilation ducting. Each radiation monitor has two trip circuits: ,
The sample panel has a pair of filters (one for one upscale (high) and one downscale (Iow or particulate collection and one for halogen col. inoperative). Each trip is visually displayed lection ) in parallel (with respect to flow) with on the affected radiation monitor. These trips a continuous gross radiation detection assembly, actuate corresponding main control room The gross radiation detection assembly consists annunciators: one upscale (high sadiation) and of a shielded chamber, beta gamma sensitive GM the downscale for the affected liquid monitoring tubes, and a check source. A radiation monitor channel. Low sample flow measured at the sample in the min control room analyzes and visually panel actuates a control room low flow displays the measured gross radiation level. annunciator for the affected liquid channel.
The sample panel shielded chambers can be For each liquid monitoring location, a purged with room air by using two solenoid valves continuous sample is extracted from the liquid operated from the control room to check detector process pipe, passed through a liquid sample response to background radiation, thus checking panel which contains a detection assembly for operability of the grots radiation channel. gross radiation monitoring, and returned to the process pipe. The detection assembly consists Power is supplied from 120 Vac instrument Bus of a scintillation detector mounted in a 1 J2 for the radiation monitor and recorder and shielded sample chamber equipped with a check from 120 Vac instrument Bus E2 for the sample source. A radiation monitor in the control room panel. The recorder has one pen. analyzes and visually displays the measured gross radiation level.
The radiation monitor has three trip circuits:
two upscale (high high and high) and one down. The sample panel chamber and lines can be scale (low). Each trip is visually displayed on drained to allow assessment of background the radiation monitor. These three trips actuate buildup. The panel measures and indicates high high radiation, plant vent discharge high sample line flow. A solenold operated check radiation, and plant vent discharge downscale. source operated from the control room can be High or low sample pressure measured at the sam- used to check operability of the channel.
pie panel actuates a main control room plaat vent discharge sample high low flow annunciator. 11.5.2.2.5.1 Radwaste Emuent Radiation Monitoring System Table 11.5-2 presents the gaseous and airborne monitors for the effluent radiation monitoring This system monitors the radioactivity in the syst e m, radwaste effluent prior to its discharge.
11.5.2.2.5 Liquid Process and Emuent Liquid waste can be discharged from the Monitoring Systems sample tanks containing liquids that have been l processed through one or more treatment systems These systems monitor the gamma radiation such as evaporation, filtration, and ion levels of liquid process and effluent streams. exchange. Prior to the discharge from the tank, With the exception of the radwaste system the liquid in the tank is sampled and analyzed ef0uent, the streams monitored normally contain in the laboratory. Based upon this analysis Amendment 11 11 3-8
37 7 q a' e ABM - 234siooxx
' Standard Plant nry n discharge is permitted at a specified release separate from each other. Each subsystem l A. rate and dilution rate. consists of four channels (A, B, C, and D) and
- ) each channel has a local detector, a converter, The radiation monitor has four trip circuits. and a main control room radiation monitor.
This high high upscale trip on the radweste efflu. Power is supplied to one subsystem of four ent radiation monitor is used to initiate closure channels by the 120.Vac instrument Bus J1 and to of the radwaste system discharge valve and the other subsystem by the 120.Vac instrument simultaneously actuate an alarm in the radwaste Bus J2.
, control room.
Each radiation monitor provides two trip circuits: one for upscale (high) radiation or an inoperative circuit and one for downscale. An o '
The high upscale trip and the low downscale upscale / inoperative trip of the channel A or B 1 trip actuate annunciators in the main control radiation monitors initiates the closing of the room. A low flow switch in the sample line also radwaste tank and pump room isolation valves and ,
actuates an annunciator in the main control exhausts the air from the radwaste working room. Table 11.5 3 presents the liquid monitors areas. An upscale / inoperative trip of channel C for the process radiation monitoring system. initiates the closing of the Ml Ocparator room zone supply valves, the closing of the radwaste 11.5.2.2.5.2 Reactor Building Cooling Water exhaust fan inlet damper, the stopping of the Radiation Monitoring System radwaste exhaust fan, and the startup of the air clean up fan.
This system consists of three channels: one for each loop for monitioring intersystem The same trips on the other four main contiol radiation leakage into the rector building room radiation monitors of the redundant system cooling water system, initiate the actuation of the same valves, damper, and fan.
', j 11.5.2.2.53 Deleted High radiation and downscale main control room annunciators are actuated by the trip signals from the monitors.
Each main control room radiation monitor vis. .
ually displays the radiation level. In addi.
tion, each annunciator supplies an output signal to the computer.
11.5.2.2.5.4 Deleted 11.5.2.2.7 Origas Vent Radletion Monitoring System This system monitors the offgas vent dis, charge for gross radbtion level and collects halogen and particulate samples. The system is identical to the plant vent radiation monitoring 11.5.2.2.6 Radwaste 9uilding IIVAC Radiation system with corresponding annunciators.
Monitoring System 11.5.3 Emuent Monitoring and Sampling This system monitors the radwaste building ventilation discharge,. including radwaste storage All potentially radioactive effluent tank vents, for gross radiation level. The materials are monitored for radioactivity in system consists of two redundant instrument accordance with Criterion 64 of General Design subsystems which are physically and electrically Criteria,10CFR50, Appendix A, as follows:
.D k
Amendment 11 11.5-9
ABM 23462aoax Standard Plant REV B
'O (1) liquid releases are monitored for gross (1) offgas post treatment; gamma radioactivity; (2) containment HVAC; and (2) solid wastes are monitored for gross gamma radioactivity; and (3) liquid radwaste effluent.
(3) gaseous releases are monitored for gross The effluent isolation functions for each gamma radioactivity. monitor are given in Table 11.51.
1133.1 Basis for Monitor Location Selection 113.4.2 Implementation of General Design Criteria 64 Monitor locations are selected to assure that all effluent materials comply with regulatory Radiation levels in radioactive and poten-requirements as covered in Regulatory Guide 1.21. tially radioactive process streams are monitored by the following process monitors:
1133.2 Erpected Radiation levels (1) main steamline; Expected rad:ation levels are in the ranges listed in Tables 11.5 2 and 11.5 3. (2) offgas pretreatment and post treatment; 11.533 Instrumentation (3) carbon bed vault; and Radiation monitors used are listed in Table (4) reactor building cooling water.
11.5-1.
Grab samples are analyzed to identify and quantify the specific radionuclides in effluents 11.5.43 Basis for Monitor Location Selection and wastes. The results from the sample analysis are used to establish relationships between the Monitor locations are selected to assure gross gamma monitor readings and concentrations compliance with Regulatory Guide 1.21 in that or release rates of radionuclides in continuous sample points are located where there is a effluent releases, minimum of disturbance due to fittings and other physical characteristics of the equipment and 1133.4 Setpoints components. Sample nozzles are inserted into the flow or liquid volume to ensure sampling the Setpoints for actuation of automatic control bulk volume of pipes and tanks. In the case of features initiating actuation of isolation both liquid and gas flow, care is taken to valves, dampers or diversion valves are specified assure that individual samples are actually in the plant technical specifications. representative of the effluent mixture. A more detailed discussion is given in ANSI N13.1.
Setpoints are listed in Table 11.5-1.
11.5.4.4 Expected Radiation I4vels 11.5.4. Process Monitoring and Sarnpling Expected radiation levels are listed in Tables 11.5.4.1 Implementation of General Design 11.5-2 and 11.5 3.
Criterion 60 11.5.4.5 Instrumentation All potentially significant radioactive dis-charge paths are equipped with a control system Radiation monitors used are listed in Table to automatically isolate the discharge on indi- 11.5-1.
cation of a high radiation level. These include:
Amendment 11 11.5-10 9
t
- 4 ,
ABM 22^6200^x Standard Plant an n
~.
Grab samples are analyzed to identify and (4) offgas pretreatment; and fT quantify the specific radionuclides in process V streams. The results from the sample analysis (5) carbon bed vauh.
are used to establish relationships between the gross gamma monitor readings and concentration The following monitors include built in check and radionuclides in the process streams. sources and purge systems which can be operated from ths main control room: !
11.5.4.6 Setpoints i (1) offgas post treatment; ;
Set points are listed in Table 11.5-1.
(2) plant vent discharge; 11.5.5 Calibration and Maintenance (3) radwaste building; and ,
11.5.5.1 Inspection and Tests (4) offgas vent. ,
During reactor operation, daily checks of I system operability are made by observing channel behavior, At periodic intervals during reactor operation, the detector response of each monitor provided with a remotely positioned check source will be recorded together with the instrument background count rate to ensure proper function.
ing of the monitors. Any detector whose re- ;
sponse cannot be verified by observation during normal operation or by using the remotely posi-tioned check source will have its response checked with portable check source. A record 7- will be maintained showing the background l (V radiation level and the detector response. 11.5.5.2 Calibration !
I The system has electronic testing and cali. The continuous radiation monitor calibration i brating equipment which permits channel testing is according to certified National Bureau of ,
without relocating or dismounting channel compo. Standards of commercial radionuclide standards, l nents. An internal trip test circuit adjustable and is accurate to at least + or 15%. The j over the fuel range of the readout meter is used source detector geometry during primary cali- 4 for testing. Each channel is tested at least bration is identical to the sample detector geo- i semiannually prior to performing a calibration metry in actual use. Secondary standards which check. . Verification of valve operation, venti- were counted in reproducible geometry during the lation diversion,'or other trip function will be primary calibration are supplied with each con- ,
done at this time if it can be done without tinuous monitor for calibration after installa- !
jeopardizing plant safety. The test will be tion. Each continuous monitor is calibrated us- !
documented, ually during plant operation or during the refu- ;
eling outage if the detector is not readily ac. ;
The following monitors have alarm trip cessible. A calibration can also be performed !
circuits which can be tested by using test by using liquid or gaseous radionuclide stan- ;
signals or portable gamma sources: dards or by analyzing particulate iodine or gas-cous grab samples with laboratory instruments.
(1) main steamline; The offgas pretreatment monitor shall respond (2) containment HVAC; to a gross gamma signal obtained from the periodic analyses of grab samples. The readout (3) fuel area HVAC; units shall be mR/hr per mci /sec.
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Amendment 11 11.5-11
x s ABWR 2=mx E Standard Plant REV.B The following monitors shall respond to a 11.5J.4 Audits and Verincations gross gamma signal obtained from the periodic analyses of grab samples to read the rate in Out of Standard Plant scope, counts / min (1)' offgas post-treatment; (2) plant vent discharge; (3)- radwaste building vent; (4) fuel area vent; (5) offgas vent; (6) radwasteeffluent;and (7) reactor building cooling water.
The following monitors shall be calibrated to read the gross gamma dose rate in mR/hr:
(1) main steamhne; (2)- containment HVAC; (3) fuel area HVAC; (4) - carbon bed vault; and
-(5) control building HVAC.
11.5.53 Maintenance The channel detector, electronics, and recor-der are serviced and maintained on an annual ba-sis or in accordance with manufacturers recommen-dations to ensure reliable operations. Such maintenance includes cleaning, lubrication, and assurance of free movement of the recorder in addition to the replacement or adjustment of any components required after performing a test or calibration check. If any work is performed which would affect the calibration, a recali-bration is performed at the completion of the work.
11.5-12 O
Amendment 11 I
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Standard Plant arv.n A !
V QUESTION 210.46 Portions of the stress, deformation and buckling limits for safety class reactor internals which are listed in Tables 3.9 4,3.9 5 and 3.9 6 requires additional review by the staff. If either Equation b in Table 3.9 4, Equations e, f, and g in Table 3.9 5 or Equation e in Table 3.9 6 will be used in the design of safety class reactor internals for the ABWR, provide a commitment in each of these tables that supporting data will be provided to the staff for review.
RESPONSE 210.46 A commitment to provide supporting data for NRC review has been added as requested.
QUESTION 210.47 The information in Subsection 3.9.6 infers that only ASME Class 1,2 and 3 pumps and valves will be included in the inservice testing (IST) program for the ABWR. It is the staff's position as stated in Standard Review Plan, Sections 3.9.6.11.1 and 3.9.6.11.2 that all pumps and valves which are considered as safety related should be included in the IST program even if they are not categorized as ASME Class 1,2 or 3. Revise Subsection 3.9.6 to agree with this position.
f-RESPONSE 210.47 Response to this question is provided in revised Subsection 3.9.6. ,
j,g QUESTION 210.48 D The first paragraph in' Subsection 3.9.6 states that accessibility for inservice testing of
!- applicable pumps-and valves is provided in the plant design. However, the second paragraph and Subsection 3.9.6.3 infers that relief from ASME Section XI inservice testing will be submitted for some pumps and valves.
RESPONSE 210.48 GE is aware that there may be a need for exceptions in cases where access or configuration prohibits some types of NDE or where operability of certain equipment must be tested indirectly .
rather than by direct operation (based on the type of equipment). However,it is not GE's intention to take deliberate exception to Section XI requirements on the basis of inappropriate design. ~
Therefore, Subsection 3.9.6.3 has been removed.
QUESTION 210.49 In Subsection 3.9.6," Inservice Testing of Pumps and Valves," provide a commitment to perform periodic leak testing of all pressure isolation valves in accordance with the applicable sections of the technical Specifications for recently licensed BWR/6 plants. Normally, this information includes a list of all pressure isolation valves which will be leak tested. If such a list is not available for the ABWR, a commitment to provide the list of valves as a part of the ABWR Technical l Specifications will be acceptable.
RESPONSE 210.49 l-( Response to this question is provided in revised Subsection 3.9.6.
I-Amendment 9 20.3 141