ML20128B758
| ML20128B758 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 01/27/1993 |
| From: | Mccracken C Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9302030164 | |
| Download: ML20128B758 (155) | |
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+ . ,o, UNITED STATES e o NUCLEAR REGULATORY COMMISSION 5 E W ASHINGT ON. D. C. 20565 t Janpry 27, 1993
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APPLICANT: GE NUCLEAR ENERGY (GE) DOCKET #: 52-001
SUBJECT:
SUMMARY
OF MEETING WITH GE TO DISCUSS INSPECTIONS, TESTS, ANALYSES, AND ACCEPTANCE CRITERIA (ITAAC) The NRC staff met with the General Electric Company, representatives of the Department of Energy, and representatives of the Nuclear industry from January 11 through January 21, 1993, to discuss the development of ITAAC as part of the certification process. Enclosure 1 is a list of attendees. Enclosure 2 contains the draf t of an Introduction (section 1.0) and also includes a set of definitions (Section 1.1) for use in conjunction sith the Desigr. Descriptions and ITAAC. This information reflects the techns:a1 agreements reached during the meeting, but it is anticipated that additional review by oGC and others may lead to the need for some wording changes. Enclosure 3 includes a draft of each of the ten example systems discussed during the meeting. These included: Residual Heat Removal Nuclear Boiler Leak Detection and Isolation Reactor Cooling Water Reactor Service Water Reactor Water Cleanup Standby Liquid Control Turbine Main Steam Condenser Control Room Habitability Area HVAC In general, these systems were chosen because the designs (and SSAR material) were considered to be relatively complete. The note.ble exception was the Control Room Habitability Area HVAC system and ao a result significant rework is required- for this Design Description and ITAAC. It is also recognized that the completion of other activities, such as the PRA insights, may impact the Design Descriptions and ITAAC.
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- Approaches for a number of common oor generic concerns were also developed. Enclosure 4 includes summarios of the proposed approaches for the following " generic" concerns:
Welding Environmental Qualification Seismic Qualification Verification of Motor-operated Valves . EMI/RFI issues.for I&C ! Instrument Setpoints Electrical Independence (Separation) Piping In conjunction with the conclusion of the ITAAC mootings, a Senior Management Meeting was held on January 21, 1993 at the GE offices in San Jose to summarize the overall ITAAC review process and other matters related to the ABWR. (For further information, see the ; minutes of that meeting.) It was decided that the team would develop staff review guidance from the' lessons learned during_the ITAAC meetings. GE and the industry are expected to also develop their lessons learned. A followup meeting at NRC headquarters is planned for the week of February 1, 1993 to discuss the results of these efforts. In addition, the need for developing an example _of a IT?.AC for a building was identified. It was decided to discuss this further~ during the upcoming meeting. , g a isigueo by Conrad E. McCracken, Chief Plant Systems Branch _ Division of Systems' Safety > L and Analysis-
Enclosures:
As stated l _
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EEw:DE ADT:NRR SPLB DSSA DThatcher WRussell CMcCracken 1/27/93 1/M/93 1[J7/93 (G:\mtgmin.dt) L I , , , - _ _ . _ _ _ . . _ .
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- 4 January 27, 1993 11EETING
SUMMARY
DISTRLDUTION w/ enclosures Central File PDR PShea PDST R/F JPartlow, 12G18 RHasselberg TBoyce EJordan, MNBB 3701 TWambach JNWilson ACRS (10) WRussell RNease, 16G15 EDoolittle, 16G15 MPleishman, 16G15 EMcKenna, 16G15 KConnaughton, 16G15 KCyr, 16G15 SFranks, DOE A.P. lleymer, NUMARC A.J. James, GE D. Antolovich, Westinghouse C. Brinkman, ABB-CE w/o enclosures: TMurley/FMiraglia MMalsch, OGC RBorchardt GMizuno, 15B18 MZobler, 15B18 RMeyer, 16G15 BBorderick, 15B18 GGrant, 17G21 i AThadani, BE2 JLyons, 8D1 ZRosztoczy, NLS 013 RTripathi, 17G21 DCrutchfield RPierson SDembek GBagchi, 7H15 JMore, 15B18 MRubin, 8E23 JGray, 16G15 JRichardson,_7D26 TCollins, 16G15 TDiPallo, NLS 013 WTravers CPosluany JJohnson, 16G15 TCollins, 16G15 MFinkelstein, 15B18 JScarborough, 16G1 2q0 09 P
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# . Irelom 1 SE/NRC ITAAC REVIEWi i
ATTENDEES l j J. LYohs NRC S. FRANKS DOE H. WALKER NRC W. Fi. ETCHER DOE W. BURTON NRC S. FRANTz NEWMAN & HoLiz!NoER (GE) G. THOMAS NRC A. P. HEYMER NUMARC D. TERAo NRC D. WILSON NIAGARA MOHAWK /EPRI l T. SULLIVAN NRC J. rec ABB-CE R. LI NRC E. WHtTAKER TVA/EPRI D. THATCHER ,NRC W. L. ZIMMERMANN AEP J. STEWART NRC D. ANToLOVICH W M. CHIMARMAL NRC J. WALDkoM CEI/SBWR T. PoLICH .NRC J. WHELEss SOUTHERN CO. T. BoYCE NRC A. J. JAMES GE J. WILSON NRC R. LouIsoN , GE R. GRAMM NRC N. HACKroRo GE
- 5. MALun NRC J. CHAMBERS GE W. russell NRC J. F. Qu1RK GE M.FzhKELsTEIN NRC C. MCCRACKEN NRC PLus SUPPORTING GE TECHNICAL PERSONNEL As NEEDED.
AJJ-4A 1/21/93
0 N GE/NRC ITAAC REVIEWS SUPP.0RTING GE TECBNICAL PERSONNEL SYSTEM /ISSUI GE ENGINEER (S)- LEAK DETECTION AND ISOLATION SYSTEM N. G. TOTAH REACTOR WATER CLEANUP SYSTEM E. V. NAZARENO i NUCLEAR BOILER SYSTEM J. K. SAWABE F. E. COOKE STANDBY LIQUID CONTROL SYSTEM P, F. BILLIG REACTOR COOLING WATER SYSTEM G. E. MILLER REACTOR SERVICE WATER SYSTEM G. E. MILLER RESIDUAL HEAT REMOVAL SYSTEM W. E. TAFT TURBINE-MAIN STEAM SYSTEM J. C. BLACK CONDENSER J. C. BLACK CONTROL ROOM HABITABILITY AREA M. MUNSON HVAC SYSTEM CONTROL BUILDING SAFETY-RELATED M. MUNSON EQUIPMENT AREA HVAC SYSTEM WELDING L. FINNEY
- 0. SANDUSKY AJJ-4B 1/21/93
O 4 , GE/NRC ITAAC REVIEWS SUPPORTING GE TECHNICAL PERSONNEL SYSTEM /ISSME GE ENGINEER 11t EQUIPMENT QUALIFICATION (50.49) N. G. LURIA D. C. RENNELS SEISMIC / DYNAMIC OUALIFICATION N. G. LURIA D. C. RENNELS MOV (89-10 ISSUES) 8. GENETTI G. L. MOORE EMI/RFI/SWC B. H. SIMON I&C ENVIRONMENTAL QUALIFICATION B. H. $1 MON INSTRUMENT SETPOINTS A. J. JAMES SECTION 1 0F THE CERTIFIED DESIGN A. J. JAMES DOCUMENT S. FRANTZ (LEGAL) PIPING DESIGN ACCEPTANCE CRITERIA J. B. KNEPP M. HERZOG E. O. SWAIN ELECTRICAL SEPARATION C. F. CHRISTENSEN
'F AJJ 4C 1/21/93
o 4 Em1mM ' ABWR Design 0:cument l 1.0 Introduction I The purpose of this document is to present the certified design for the Advanced Boiling Water Reactor (ABWR). i l The certified design consists of: ' (1) A design description, design commitments and associated inspections, l tests, analyser, and acceptance criteria (ITAAC); (2) The site parameters upon which the design is based; and (3) Interface requirements. The certified design does not include all of the material submitted as part of design certification application. Where a conflict exists bctween the Standard Safety Analysis Report for the ABWR and the certified design, the certified design is controlling. This document is structured as follows: Section 1.1-Definitions. Section 2.0-A design description, ITAAC and interface requirements for systems, structures and components within the ABWR certified design. Section 3.0-A design description, ITAAC and interface requirements for . systems, structures and components for which the design process and selected design features are being certified. Section 4.0-Identifies the interface requirements to be met by those portions of the plant for which design certification is not being sought.% wL appucants . I
~ief'cFencing the XIiWRTertifiafdesign is required to submit Derign Descriptions L-and ITAAC for the site specific design features which meet the ABWR inter requirements. ,
' Section 5.0-Describes the site parameters applicable to the ABWR certified -
design, such as tornado strength, ficod height, and cardiquake accelerations. Appendix A-A legend which defines the symbols used in figures which are part of the design description. Appendix B-List of Abbreviations. 1.0 1 1/19/93
0 % ABWR oesign occument 1.T Definitions As used in this document, the following terms are defined: Acceptance Criteria-The performance, physical condition, or analysis result for a structure, system, or component that demonstrates the design commitment is met. Analpis-A calculation or mathematical computation and/or engineering evaluation. Engineering evaluation could include, but is not limited to, comparisons with operating experience or design of similar equipment. As-built-The physical condition of the system, structure, or component following the completion ofits installadon or construction at its finallocation at the plant site. Basic Configuration (Building)-The building arrangement of structural features (e.g., floors, ceiling, walls, columns, and door ways) which are specified in the design description and the relative location of systems or components which are also specified in the building design description,if the building design description includes sntems or components, then the basic configuration (system) definition applies to those systems or components within the building. Basic Configuration (System)-The functional arrangement of structures, divisions and components specified in the design description; and includes and is limited to pressure boundary welds for ASME Code Class 1,2 and 3 components; dynamie qualification of seismic Category 1 mechanical and elecuical eghipment; environmental qualification of Class 1E electrical equipmentf and mechanical qualification of active seismic Category I motor operated valves. Inspections for basic configuration include inspection of the system functional arrangement and inspections limited to the following: (1) Inspections, including non-destructive examination (NDE) of the as-built, pressure boundary welds for ASME Code Class 1,2 or 3 components identified in the design description to demonstrate that the requirements of the ASME Code Section 111 for assuring the quality of pressure boundary welds are met. (2) Inspections of the results of the tests and/or analyses of the Seismic Category I mechanical and electrical equipment (including associated instrumentation and co- trols) identified in the design description, l including associated anchorages to demonstrate that the as-built i condition of such equipment is qualified to withstand the design basis dynamic loads without loss of their safety functions. 1.1 1 1/21/93 i I L
o b ABWR Deston Document (3) Inspection of the result of test and/or analyses of the as built Class lE elecuical equipment (including connected instrumentation and controls) identified in the design description demonstrates that the Class 1E electrical equipment is qualified to withstand the emironmental conditions associated with design basis events without loss of safety function for the time needed to be functional. Such equipment includes the connected electrical equipment (such as cabling, wiring, and terminations) and lubricants necessary to support perfonnance of the safety funetions of the components identified in the design description g As built electrical components (including aspsztea'mstrumentation and controls) are emironmentally qualified if they can withstand the emironmental conditions associated with design basis events without loss of their safety functions for the time needed to be functional.These emironmental condidons are as follows, as applicable to the bounding design basis event (s): Expected time dependent temperature and pressure profiles, humidity, chemical effects, radiation, aging, submergence, and synergistic effects which have a significant effect on equipment performance. Electrical equipment emit onmental qualification may be demonstrated through testing of an identical item of equipment under identical or similar conditions with a supporting analysis to show that the equipment to be qualified is acceptable, testing a similar item of equipment with suppo. ting analysis to show that the equipment to be qualified is acceptable, experience with identical or similar equipment under similar conditions with a supporting analysis to show that the equipment to be qualified is acceptable, or analysis in combination with partial type test data that supports the analytical assumptions and conclusions. (4) Inspection of the results of tests of active safety related motor operated valves identified in the design description demonstrate that the MOVs are qualified to perform their safety functions under design differential pressure, system pressure, fluid temperature, ambient temperature, minimum voltage, and minimum and/or maximum stroke times. Design Commitment-That portion of the design description that is verified by ITAAC. Design Description-That portion of the ABWR design that is approved by certification. 1.1 2- 1/21/93
4 6 ABWR onion oocumnt DMslon (for electrical synems/ equipment)- The designation applied to a ! given system cr set of components that enables the establishment and maintenance of physical, electrical. and functional independence from other redundant sets of components. ; Division (for mechanical systems / equipment)--The designation applied to a e specific set of components within a system. Inspect or Inspection-Visual observations, physical examinations, or review of - records based on visual observation or physical examination that compares the i as-built structure, system, or component condition to design commitments. ; Examples include walkdowns, configuration checks, measurements of ~ dimensions, and non destructive examinations. Inspections can be performed in - parts or segments over a period of time. ITAAC~~Ihe inspections, tests, analyses, and acceptance criteria that are : described in 10CFR 52.97(b) for the structures, systems, and components that are within the scope of the design certification.The ITAAC could apply to
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multiple systems and components, and they could include a design proccu for the design of systems and components that will be completed after certification. The ITAAC will be provided in tables with the following three-column format: Deshm Commitment insocctions. Tests. Analyses Acccotance Criteria , The ITAAC tables contain inspections, tests, or analyses (ITA) for each design commitraent. The idenufication of a separate ITA entry for each design commitment shall not be construed to require that separate inspections, tests, or analyses be performed for each design commitment. Instead, the activities associated with more than one ITA entry may be combined, and a single inspection, test, or analysis may be suflicient to implement more than one ITA entry. , , Simulated Signal-The intentional generation of a signal for testing. I Test-Operation of a structure, system or cornponent, to evaluate its - performance or structural integrity. 1,1 4 1/21/93
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, 6 Enclosure 3 Examole Systems Residual Heat Removal Nuclear Boiler Leak Detection and Isolation Reactor Cooling Water Reactor Service Water Reatter Water Cleanup Standby Liquid Control Turbine Main Steam Condenser Control Room Habitability Area HVAC
l ABWR ossion occumnt l-i e-9z r wms n e vis eo I ~17- 93 VIARK (LP g a g S G 2.4 Core Cooling 2.4.1 Residual Heat Removal System
//.5dl9.3 Design Description The Residual Heat Removal (RHR) System is comprised of three dhisionally separate subsystems that perform a variety of functions utilizing the following six basic modes of operation: (1) low pressure core flooder (LPFL), (2) suppression pool cooling, (3) wetwell and drywell spray cooling, (4) shutdown cooling, (5) augmented fuel pool cooling, and (6) alternating current power source (AC) independent water addition. The system configuration of each loop is shown in Figure 2.4.la, b, c (Mg j ! " 2n 4 add. he major functions of the various modes of orcration include: (1) containment heat removal, (2) reactor decay heat removal. (3) emergency reactor vecellevel makeup and (4) augmented fuel pool cooling. In line with its given functions, portions of the system are a part of the Emergency Core Cooling System (ECCS) network and the containment cooling system. Additionally, portions of the RHR System are onsidered a part of the Reactor Coolant Pressure Boundary (RCPB).
Except for the non ASME Code components of the AC independent water addition feature (Figure 4.2.1.c), the entire RHR System is designed to safety-related standards, although it performs some non-safey functions. He safety. related modes of operation include: (1) low press, e core flooding, (2) suppression pool cooling, (3) wetwell spray cooling and (4) shutdown cooling. Non safety-related modes of operation include: (1) drywell spray cooling, (2) AC independent water addition and (3) augmented fuel pool cooling. The RHR System also provides a backup, safety related fuel pool makeup capability. Ancillary modes of operation include minimum flow bypass and full flow testing. The ECCS function of the RHR System is performed by the LPFL mode.
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Following receipt of a lossof-coolant accident (LOCA) signal"--- L! ::1 er Mghiped! p=reMMA E 9e :dkd ec RHRinid;in f;ndi, the RHR System automatically initiates and operates in the LPFL mode M.a o
] mj;ned n -M er :c;n;inir dic ECCL nrc;rh1to provide emergency makeup to the reactor vesseyn.+Hb;-tn E::p 6: re?mr =: m14The LPFL injection flow for each loop begins when the Reactor Pressure Vessel (RPV) pK dome pressure is no less than 15.8 kg/cmg above the drywell pressure. For a design basis large break LOCA (LBLOCA), for RPV dome pressure no less than 2.8 kg/cm2 greater than the drywell pressure, the LPFL injection Law for each loop is 954m3/hr minimum within 36 seconds of receipt of actuation signal,wA the&4R w... ini9 nyhhsandhmode, The LPFL mode is accomplished -
by all three loops of the RHR System by transferring water from the suppression pool to the RPV, via the RHR heat exchangers. Although the LPFL mode is automatically initiated, it may also be initiated manually. The system will also 1/18,92 1 2.4
ABWR onian nocumut
- de of operation from the test mode, the IN SEh automatically revert to the LPFLsupp,,ression pool coo A a LOCA signalpch RHRToop's RPV injection valve requires a low reactor pressure permissive signal whether being opened manually or automatically in response to a LOCA signal. hemt u, n m e,1_
The containment heat remont function in the ABWR is performed by the Containment Cooling System, which is comprised of the LPFL, suppression pool cooling, and wetwell and drywell spray cooling modes of the RHR System. Following a LOCA, the energy present within the reacsor primary system is dumped either directly to the suppression pool via the Safety Relief Valves (SRVs), or indirectly via the dr>well and connecting vents. Subsequently, fission J product decay heat continues to add energy toyh: prge contamment Cooling System is designed to limit the long-term bulk temperature of the suppression pool, and thus limit the long-term peak temperatures and pressures within the wetwell and dr>well regions of the containment to within their analyzed design limits, with only two of the three loops in operation. The cocling requirements of the containment cooling function establish the necessary heat removal capacity for each loop as no less than 88.5 kcal/sec'C, which includes the RHR heat exchanger, the Reactor Building Cooling Water (RCW) system, and the Reactor Service Water (RSW) system referenced to the ultimate heat sink. ne heat remoni capacity is based on the suppression pool cooling mode with the RHR tube side heat exchanger (Hx) flow rate 954m3/hr minimum. The LPFL mode,in addition to its primary function of cooling the core, serves to cool the containment. The suppression pool cooling mode is made anilable in each of the three loops of the RHR System by circulating suppression pool water through the respective RHR heat exchanger and then directlyback to the suppression pool. tlm J1 k 'niited L =per.x :: M;;h n. ppg.dcr. pad e _ PMm.w=oThe wetwell and drywell spray modes of RHR are each anilable in
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only two of the three subsystems (Loops B and C). These functions are performed by drawing water from the suppression pool and delivering it to a common wetwell spray header and/or a common drywell spray header, botl via the associated RHR heat exchanger (s). The wetwell spray flow rate for either loop is no less than ll4m3/hr. These containment spray modes of the RHR System can be initiated manually . However, the drywell spray inlet valves can _ only be opened if there exists high dowell pressure and if the RPV injection _ ves are fully closed.yetwen and dowell sprays serve as a augmented method of containment cooling. Wetwell spray also serves to mitigate the consequences of steam bypassing the ruppression pool. __ The normal operational mode of the RHR System is in the shutdown cooling mode of operation, which is used to remove decay heat from the reactor core. This mode provides the required safety-related capability needed to achieve and maintain a cold shutdown condition . The RHR heat exchanger heat removal 2,4.1 2 1/1&92 w _
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If a drywell . spray valve-is . open, .the RHR System l automatically reverts to i 1 allow the LPFL mode in response to._the injection valve beginning to open. 1j
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O I ABWR onion 0:cumnt capacity requirements in this mode are bounded by containment cooling requirements. Shutdown cooling is initiated manually once the RPV has been depressurized below the sptem low pressure permissive. In this mode, each loop takes suction from the RPV via its dedicated suction line, pumps the water through its respective heat exchanger tubes at 954m3/hr minimum, and returns the cooled water to the RPV. Two loops (B and C) discharge water back to the RPV via tdicated spargers, while the third loop (A) utilizes the vessel spargers of one of the two feedwater lines (FW A). He heat removed in the RHR heat exchangers is transported to the ultimate heat sink via the respective dhisioa of reactor cooling water and senice water. Each shutdown cooling suction valve is interlocked with that loop's suppression pool suction and discharge valves and wetwell and drywell spray valves to prevent draining of the reactor vessel to the suppression pool. Also, to prevent draining of the reactor vessel, each shutdown cooling suction valve is interlocked with, and automatically closes on, low reactor water level. The augmented fuel pool cooling mode of the. RHR System supplements / replaces the normal fuel pool cooling system during conditions of high heat load. This mode is accomplished manually in one of two wap. When the reactor vessel head is removed, the cavity is Gooded and the fuel pool gates are removed, the RHR System cools the fuel pool in the normal shutdown cooling mode. When the fuel poolls otherwise isolated from the reactor cavity, two loops (B and C) of the RHR System _ can directly cool the pool by taking suction from and dFscharging back to thet m! fuel pool cooling system at 350 m3/hr minimum.This connection also provides forcemergenc3 fuel poopakeup f capability by supplying a safety related makeup pa to the fuel pool from a safety-related source (i.e., the suppression pool). One loop (C) of the RHR System also functions in an AC independent water addition mode. This mode provides a means of cross connecting the reactor building fire protection (FP) system header to the RHR System just outside the containment in the absence of the normal ECCS network and independent of the normal safety related AC power distribution network. The connection is accomplished by manually at the valve opening two in series valves on the cross-connection pipingjust upstream ofits tie-in to the normal RUR piping. Fire protection system water can be directed to either the RPV or the drywell spray sparger by manual opening of the loop C RHR injection valve or the two loop C drywell spray valves nese three valves also have manual hand wheels. The fire water is supplied via the system's Reactor Building distribution header. g ( Each loop of the RHR System also has both a minimum Gow mode and a full Dow pump flow testing during plant operatic >a. The minimum g# flow test mode (Cg] hat there is pump flow sufBci mode assures t anytime the pump is running by opening as needed a minimum Dow valve that directs flow back to the suppresilon pool . Upon sensing that there is sufIicient 3 Vi&92 2.4.1
, 9 ABWR nesign occument flow in the pump main discharge line, the minimum flow valve is automatically closed. In the full flow test mode, the system is en:ntdy operated in the suppression pool cooling mode, drawing suction from and discharging back to the suppression pool. The RHR System is comprised of three separate loops or subsystems, each of which includes a pump and a heat exchanger, takes suction from either the TPV or the suppression pool, and directs water back to either the RPV or the suppression pool. Two of the three loops can divert a portion of the suppression ar g pool return flow to a common wetwell spray a common _drywell spray sparger. The divisiort 7*-4Tthe sp$r+ger RHR System or direct the e x v are separatch.4 physically, ed !cWily, x '~"McEn;; php!M,4eented -e t- dN;;;n: a c;c ofi: p!=:. Each of the three ch,~ Qs powered from the 3 respective Class 1E division as shown on Figare 2 A.1a, b, c. Cooling water to each' division of RHR equipment (heat exchanger as well is supplied by the respective division of the reactor cooling water (RCW) Sptem. 0 ne RHR System also has provisions for containment isolation and reactor 3* containment pressure boundary RCPB y;es,sure isolation. Me RHR Syst 'll maintain capability to perform its intende ety.
,r ated fun ons du 'ng an ollowin design b acci nts co tions. cept independ w2,er addition -
' for he n n-ASME C moonents o an (Firure 4.2.1 A f he i RHR System is Seismic Category I and is housed in the Seismic Category I Reactor Building. ..:n= 6c n ir. Figurr a ' ' ..,b, =: yan'in ed fnr emde= condi'ons ex pt( the p tv he
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H: RHR p"~p' "r memrdb., c= ifug;' pump: ad =pd!: ef rupp! pag c' Ps S5 i ai O/hr go. c.than-cr c v al ;; 2.8 Eg/c;n2 'd7;;11:s ""}.The RHR pump will provide ta b/1/"ow when the surface of the Guid is saturated. The pumps are ASME Code class 2 components. Suflicient NPSH is provided to the pumps for the design conditions of 100*C water, the containment at atmospheric pressure, the suppression pool at its minimum level, and the strainers bic ked greater than or equal to 50E The pumps are interlocked from starung without tan open suction path. cenditionM tion, 5 eFM penp- ~: ix::cc;c" 'crR
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ABWR ossion Document r' m S S R.T~ B) y e nf s arc 3. m m Toblc ?.:....nThe low pressure portions of the shutdown cooling piping are protected from full reactor pressure by automatic pressure isolation valves that are interlocked with reactor pressure. Mditbaally,: d:t
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The RHR System control room . s indications and controls allow for monitoring and control during operational conditions. The control room has indication for system flows, temperatures and pressures, as well as valve open/ close and pump on/ oft status, and controls for those instruments and ponents mi, ' s (mf 1 shown c h d = hon .; ;hFigures e _ nly the 2.4. stalap d-le che eh +: ;xccpdcr. :f rim va ves do stre ofeac oop's injec ns ve h con ol , statgin 'catipy i e contr roo ). Valves how n Figu . A h,b,s 4 1 bcx...opc = r knw _ . ., , .-_.n.,m- .mm m Fmur t.[2.4.I h cl Pt.9tM e. 4
- S & fig W t-RHR System components with status indication and/or control 'nterfaces with the Remote Shutdown System (RSS) are shown on Figures 2,4.1a andg .4.lb.%e.e muhansi em-.cnime detimed s
o '- m::: /.SME codc r;ce!::. .: . ~a m 1
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Table 2.4.1.s
- ASME Design Conditions Component Code Class Pressure Tem l
i Main pum 2 35.2kg/cm2g 182*C l ! Jockey Pumps 2 28.8k 'g 182*C Heat exchangers primary e (reactor) 2 .2kg/cm2g 182 C Heat exchangers secondary side ooling) 14 kg/cmtg 182 C l 87.9kg/cm2E 302 C 1 Piping and components from the t l l to andincluding the outboard contai ent isolation valves Piping and components o n to the 2 3.16kg/cm2g 104'C containment atmosph e, out to and l including the out d containment isolation val components outside the 2 28.8kg/cm2g 182*C Piping con ment isolation valves and on the on side of the pump , (s - 1/1M2 2.4.1
1 I A sed -B T k e- y em9 & c<rrnpernsdr owtsid.e. h.e_ w % a m e 4 tsoIatton valves am/ om 2ka sucfin s//e o f the pump hue. a dwyn prestare. G 28,8 ly/cm'y ?nte system LOCA ( .C S LocA) conditiens. F 5./-
'ABWR ossion 0: cum:nt AS D 8 Conditions --
Co ent G e T e 2 ' Pip g and co ponents utside 2 .2kg/c g 18 2*C - cont 'nmen lation vah s an on the ]._ ylisch ide of the pump /~ Inspections, Tests, Analyses and Acceptance Criteria - This section provides a definition of the inspections, test and/or analyses together with associated acceptance criteria which will be undertaken for the - RHR System. . l l 2.4.1 6 1/18/92 i
9 F .% Table 2.4.1.b: Hasidisal Heat Removal Sysitam C" Inspections. Tests. Analyses and Accentance Critwin d Cart l Rad Daeign Commitment inamandana.Testat Analyana Accantansa Critaria ;
=
'[
1.- gbeelc configuration of the RHR System is 1. Inspections of the as-built eyetem 1.' ~ The as-built - ":-- '-- " RHR i shown in Figuree 2.4.1e,2.4.1b and 2.4.1c. - - ( SysteW: in : : --- t - -- T figures 1 (prov_ ides the. Mosu. -s wg., - [: "7 2":x will be conducted. 2.4.1a. f4.1b and 2.4.1c. L 5 l! 3
- 2. The RHR~Syst.F *:, _. 2 - " .r " Ti - - - - - -- '
p
.-.: i ": , .; b : . ,::::.-'q Conforr% wsth 'tke Isa.5ic.- ,
- m conftfue-dton [Shaien ;h i
e 2e,The RHR System operates in the LPFL .. *m onalteste shell VM conmted RHR How. satisfies the mode to provide emergency core make up P*rformed on the RHR LPFL mode" following
. to the reactor vessel as follows Analysis shall be performed t test results to the conditions he g CortlHed Design Commitment. The LPFL injection flow for each loop 4
y begine when the RPV dome pressure is - -t g greater then or equel to 15.8 kg/cm2 above
- r i
the drywell pressure. ' PFLinject for each loop ~ The LPFL injection flow for each loop is ;
. begine when RPV dome pressure le greater then or equel to 954 m3/hr, for e keka2 abovo i
greater then equel to 15.8 : domefessure greater then or equel to 2.81 I the eesure. ; m above the drywell pressure. Ie-2b. The L injection flow for oech . gre then or equelto 954 s//hr . m mum, for e dome pregoure et greater C 9 or equel to 2.8 kg$rh2above the1 _ _ . . _ drywell pressure.- ,. 2c.The injection y opens within 36 esc of ' 2c. Using almulated low pr re permissive . 2c.The ini valve opens within ~ 'of [ ! eseure perminolve : actuotton signol. R oction valve teste - of low pressure permi signet. l.
.(' receipt of lo
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f Ib* * . l , f & 'M- C '[_ ' y u h d' C P- y 8 W st W 4 Werr #5 b'J 4. L. -Mini *'- -k.-- _ - -
o y
" Table 2.4.1.b: R=-h Heat Removal System
.e
' trW_tana Te=*< Aamht=== and Accaatance Criteria j
F kispections. Testa. Analyses Wtence Cnterie GertNied Design Commitment Js.rft 1 x 2 b, . Q ;ffokMr_ ftQ. 10~ . ( / in ttwMcooling mode,the RHR tube ~ cooling mode,the RHR tube side - [ .In the S/P cooling mode, RHR System heet exchanger,(Hz) flow rete le 2# inthe heet exchanger, (Hul flow rete is greater functional teste shall be performed,to-- then or equel to 954 m3/hr. 3 then or equelto 954 m /hr, ---{ .].:.gr; S;- -t 2 :O 0; h::t - ' 2.C. ey flow EC,f RHR System functional tests will be ' Jer greater RHR loop 2c,7 RHR loops, either 8 or C, con provide 3 performed on loope B and C in the wetwell thenBorprovides wetwell equel to 114 m y/hr.' 7 greater then or equelto 114m /hr flow to the wetwell spray heeders. eprey mode. . ey flow .p RHR greaterloop thenCorprovides wetwel! y/hr. equal to 114 m ] I l 1 Igj,. 2de i. 1cI 4 in the shutdown cooling mode,the RHR g in the shutdown cooling mode, system' g The RHR heet exchangere tube side flow
'. functional tests will be performed to rate le grosser then or equel to 954 m3/hr. ,
tube olde. heet exchanger flow rete le , determine system flow rate through each . Heat exchanger removal cepecity in this grooter then or equel to 954 m3 /hr lheet . mode le bounded by suppreselon pool
~
exchanger heat removal cepedty in this heet exchanger. Inspectione and analysis
' shell be performed to verify the shutdown " cooling 'requ.remente : . ;
modo le bounded by suppreselon pool > cooling requiremente). .. cooling modo le bounded by supprecolon r
. pool cooling requiremente 2 2e .
in the augmented fuel pool cooling mode, g in the augmented fuel pool cooling mode, .1 2 E. p In the augmented fuel pool cooling mode, the RHR tube side heet exchanger flow rete - the RHR tube side heat exchanger flow rate ' system funcelonel tests will be performed ' 3 ' to determine system flow rete through le greater then or equel to 350 m3/hr. Host i
' le greater then or aquel to J50 m /hr lhoot . .l each heet exchanger. Inspections and . exchanger _ heat removai cepecity in this -
exchanger heat removal cepedty in thle .- . modo le bounded by suppreesson pool; l mode le bounded by suppreselon pool : analyele shall be performed to verify the . shutdown cooling modo le bcunded by cooling requiremente
- cooling requiremente). '
supprese?on pool cooling requiremonte. r --r*v w + o r w- e .
Table 2.4.1.b: Residual Heat Removal System (Continued) inspections. Tests. Analyses and Acceptance Criteria Cartified Dealen Commitment Inapactions. Testa. Analvans Accantance Critaria 2.f 2.f. 2.f. gWater is pumped at a flow rate of greater XRHR System is designed to permit pump AT*Usinginstalled controls, power supplies ' and other auxillaries, the pump flow test then or equal to 954 m3 /hr. flow during plant operation. wl!I be conducted for each RHR loop after d system Installation. Water will be pumped N 1 . **' i ti by- ; y in the test loop. System head will be f. i. -
'd i equivalent to e pressure differenCol of 2.8 kg/cm2between the RPV and the drywell.
1
-e -
'e Y ' a i RO asww rn
$4 3 3- -l k
/ he RHR p x based spen f
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ee-builte d .i u me w c. 1 _:,__ _ _ _1 . . have einffistemt WSII.
^
. ._ -- .y u .. u
%W.ametyeaswil! .
and e ..
- e. .. p .
JJ,A , s
- p. ouettee f.en the suppreseten poet with T .unter levet at she maale n value. 34
+ "..?e
" " ~""
A.TL w W/dd , ' g - doetga beats fload tempereente (les*c). so,.sa. ,.o N JA h L-
# . e t .
l . l
$W W' N . . - _
e e Table 2.4.1.b: Rapidual Heat Removal System (Continued)
- 0 b=dians.Te=*=: Anahr=== and Acemnce Criteria -
( inspassiana.Tants. Anahrams Ascentansa Criteria partlBad Danign Camsdtment l r 4, lu iSa S/P coolen moda., Q Fo, -tke. S/p coof t 4, For the S}p cool *ng modeg M the @ _ .d the effective heat removal capability of each Inspectione and enelyste effect've heat removal capability of each RHR Hx, including the effects of RCW,- _ shell be performed to determine the heat RHR Hx. Including the effects of RCW. RSW and ultimate heat sink le greater then exchanger's effective heet removal RSW and ultimate heetsinkis grooter then or equel to 885 kcsysoc 'C. capability. or aquel to 885 keeVeec *C.
/H6 fit?f h h L y'3 U h 77) 6't" l
/9 fHHL l
+
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d r e Table 2.4.1.b: Re.w.a W=t Removat System (Continued) - 4' ir winns T==en Anahrsas and Acceptance Criteria inspestlans. Tests. Ammiyams Ascentanta Critaria Cartised Damian Canunitsnant 6,El Isilletten Legle { peach RHR division telelease la eks' LPF1. win 6,Using~ simelmed be perfeemed lettission of the RHR signals, leislation signaltests 5,Upon (seneeni receipt ofata.she or, setomatic) simulased i seode la sospense se an eesematic RHR ' lettissies (LOCA) . :';- '_^~ -- t " *' ! logic. .lepus so the RHR logic, the leHowing occurs: or to a manuel teleissies signal.lfsome g=_'___.___m..___. ~ _u _ _ _ q~ a) "Ihe RHR peep receives a signal so start I b) ~the RPV injecties valve receives a .
. signal to open psevided a low ,
. reactor. pressese permissive ' is
* -c) esies pool M Tf(
volve receives a signal se adess CIMcM) l d) "I%e weteen spray valve receives a Z signal se close (Diviseems B&C
! eely)
(> g Isolotles i.egle bo . .. . 6 a.. t ladL The sheedewe coeHag labeesd and A Uslag aimeleted hMR . lselseisef AU Poe receipt of a seenmiesed RHR outbeerd leelestos volves necesseelcony signets, seems wile he performed of she lasiasion signet she' feHowieg eccors: close se seceips of an' RHR isoloales joheedews . cooling leefselee velveg imelesies logic.- s) The sheedewe .cealing inheard ' and signal.tisen>4ERS)- emeboard isolaeion velves asseseasicaHy close
- .m..__
m,,.. o br
,$ggeacbr-T
.'gh .reacter pr q s s u v-4.sA .The,P' '..u 66. <
L
- - e :: s.
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"-t_-2 __
3 Udeg(3einesseed,,_ _ __ _ _ _ , 4 Upon receipe 'of a 'simula == 9- -
-ouebesse-4estesten velves -aaleelty signey, Tasses will be p.'. ;--f of they laelassee ~signes . she geg Q . cc ,,:
j gg,,c ,,.,,cg ,, ,gw__ 7 - -- g_. _ _ . . _ _ . . signal,'"__ _ ' 9 ? leelesien logic, " -_ j - .--e _ mC --f _; og , weee, up py ,_r. ,. eman , antense;ce,a. -
% nje.c % } ;
g : Tatde 2.4.1.b: Pe " F-* Fmal Sv**=m (Continued) .. L L=== ehr.i Tee: 2 Aa=a v- _ - a A, Criteria Inanessimma.Tanta. Anatyana Assentanca Criteria , CartMind Damian Casmaniemant
~
- 7. Iwtertoc.k Lc>s ac '
7a- 7 a .. 1 7a. - / Using almulated LOCA actuation signale. g Upon receipt of simulated LOCA actuation _;
/ if already operatingin the test mode the tests will be performed on the RHR signal, RHR logic functions to suppression pool coollrg , _ . _, r or automatically reconfigure the system to .
wetwell sprey modes, the RHR System System. the LPFL mode of operation when. + automatically reverte to the LPfl. mode in oporsting in the test mode, the ; respanee to a LOCA signet. suppression pool cooling.O, ;;:, or ;::y ! wetwell.eprey modes. 7/b.' If a drywell spray valve is - 7Ab. Using an injection valve 7/b. With either drywell; spray opening signal, tests will be valve open and upon : receipt open, the' RilR System g automatically reverts to performed on _ the RilR of an injection valve opening NoT Fvay-
. signal, RilR logic functions. to T " M /'-
N allow the LPFL mode in System. ' close L the drywell Lspray - I response - to the . injection valve beginning. to open. valve.
. i.
7c.m peepLedeimum new ,dve . 7c. N mielmem Bew mode seesmo shes J7c.ueing W im' 'ud newl eigente, seses wist be ,, of she seceives a signal se opes whear these le pump New emmeness to keep -
'ipsed odeiseen Aow volvel 'ineeslock ~s ignals dedicative of she feitowing the peep cose septime the 'posey la condicione eanet . caecursemely l (for she remains by speales. se: seeded. e , logic. seguised time delay ' period):
~
soloiseen New volve' shot asecte now .
, i' back to ehe esppseeslee peel. Upon e) pump discharge pseasese is high -
semeles -shes shese is endScient Aew is- b) pteep new is low ' ~*. the pump male dischosse1Iles.' she solelemens now volve le h any The peep mielemom .. flew velve cleoed. seceives e signal to close when signets laMenetre ; of 'she Satieuing . cea seles .caises: ..
- e) pump flow is high
. , -n , , , . .
S i Tatde 2.4.1.b: posidual Heat Removal System (Continued) { :.= r= _*s.2 Tu:& _^ ";;== ==d A- = "-- = t'riteria Cartified Damien Carandtenant inamassans.Tasta. Analyans W Cnteria .
- u "Jd.Using desisted wactor prumu 7d.Tk Arv inkesion vdve is Macked froan opening, whesber. demanded
*/ J. Each 'RHR Divielen's' RFV isfacties dgeals, esets will be gir:' of the valve segelses a : low a pessomse sammally or via an RMIt inielesies RHR . a= jar *1== velve pasmissive logic.
permisolve s33 eal wh beleg - signal, if signals indicesive. of she l opened uneemally. er esteseedceMy le lollowing ' condicios esises: resposes to se ElWL lablesien. i
. a) Reactor pressere is " high -
72.The RHR posep is Wocked ison s 7e, The RHR posops are leasdecked fseen 7e,Udes doelmed Ivdve podsioel saasting unless signals .. Indicesive of startleg: without en spea section pash. signals, esses will be perfonned of she one of the following condisions esises: lRE6R penny saasti leesdock logic.' a) A section path from the" seppecsaios pool is.. available (The _ , , _ :::':- pool section - valve . is g fully opes) .. ,
- b) A section pash. from she RPV via tf3 she semedewe cooling suceien liac L
I is available . - (ths' simodows coolleg section , valve and inboard ' and ousheerd ' isolasses . valves are. all fully opea) P t i'
o Table 2.4.1.b: Resid mal Heat Removal System (Continued) -
.e.
a insoections. Tents. Analyses and Accentance Criteria tr=-Ene Tasta. Analvass Aee-atance Criteria ) Cartified D_'=a Cornmitrnerit
~
'injec.t.oa) scP
- A %
-wcD died )f,The suppression pool reform valv and lk. The suppresanoa poet roter'a' valv gthe h,Uslag' simmitted alve positi the w.aweil spesy valve - ' ' - - g signals, tests will be perfonned of the =? . :: st: blocked wetwell spray valve : ' i :--2 g ..;,;;"
4.., from opening sch.ss sigasts g .j2with
;!! :;theTRPV T;;:injectlaa ese laterlocked " c:: ^ pasiR .:. ] laaerlock Icgic. indicative cf the following conditiot valve.
e :: = :' * ; ':) C esists: [SafPvQSsion pool return Vdvg. Q -the wgj sve[j l s) The RPV injection valve it full L Sprely vg (To closed,@11 the dryw sp/ay vaes o y ] Q . I b Th och rics ywe sp y in-pclose J kv vt fully The suppression pool return valv[
... C l o.nd the wetwell spray valve _._ . - ;!' -- ;- n!.;e will p g,tonnatically as close if signals indicasive of the following condition l
' exists: l l l s) The RPV injection valve is not l fully closed l l 1 e.. .... .
'h 4 1 e Table 2.4.1.b: paidanal Heat Romantal Svstem (Continued) ,
.ls winens.T==*= Aaahrsas and Acceptance Criteria inmaassinna.Tanta. Analysaa Wtance Winna certified Damien Canuntament ,
m 1=**= *r-d' =r'ar 'd'*= c= 7yedas d- saadsignals.- a <= =*73 ,alves
~ '- ' ~ *= *r-a
- r !
7.h only be opened sisesteemeemely if these vdvs ' esses will bc are' blocked frene . being opes : sissekancously maless signals esists high drywell pseasure and - the ' perforasd 'of she well spray valve @ laserlock logic. indicardee of the following conditions I RPV lejecties valve and sheedewe emiss concurremely:. cooling 'secties. valve ass boek $sity
- closed. The drywell Spray VMyAS a) ' pressesc.is high:
b) N RPV injection' velve is folly Cgose_ wh.e.w m3 ec pIo n . vagve open5, closed c) The shutdowe cooling section valve is fully closed - The - . . - - E-
, , - , - .-7 . - -
drywell spray ~ valves wlH - g aneeneanically close it ' signals
. Jadicaries of the .#ellowing rnion .
Nl esiaes: a) The RPV .iejection valve. is aos - fully clees.d l l l-
. _-e am -_ _ _ _ _ _ _ _ . _ __ -_--
Tatdo 2 4.1.b: flesidual Heat Remnval System (Continued) , e - . C Irm===e*Inna T==*= _^_-f; == and Aee==*mnea Criteria knaassimas. Testa. W Assentanca Critaria partised w commitment C.AAAn m La t& B e 4 4 , t u ,Il k m % 4 B&,u e nEi DM% eLi
. RHrt s_p6 m powm4_ ,,ja, gag ty p ./m p, cg po vidry g %,t b
1,, Clun I E iM vis;ow b 4Lael. k Sadio! py oa J & TMF u ll=t t RRR %de,
=
2 . +,1 = s & Dt vu,t e -
-t J
a h ot. Mn. ;
'1. 'l. 9, .
JCThe RHR System utilizes jockey pumps (one .3CFunctional tests will be peifor med on the Jf Each lockey pump keeps its respective g in sech loop) to keep the pump discharge _ obility of the joci:ey pump (one in each pump discherge line fi!!ed.
% . tine fliled. loop)to koopits respective RHR pump r
@ discharge line full.
I Wad leaRen- ah
- n@ M. .
10q. X. Centret Reen ?: J: - _ l0 b.
.38* Emmpesttees will be perfesmed en the loc.K.
M - -
;mMn
-r w' :N
------'*-M sn eine i
preesded'for RHR syste. centret m.m '- --_: for the Ril R Centret Seem os detteed .ta sectten system. 2.4,7 , arm. defined la Seattee 2,4,3 { ,. l l e H y + us
g Table 2.4.1.b: finskhaaf Heat Ramaval System (Continued) , Irmwianis.T en_ Ar.d.;2 and Accantanca Criteria CartMind Danigi Cammitsnant Iraf eena== Tests. Anahrsas Assantanca Critaria ll h mLe% G s los* lu hb & ob"g g. - - tiens will be perfoseed as the # 7" P'e'* "'" 2-~2" 'M * ** ' b E88 System .3tr I
** * ~
(ass) Seamanes#ptseidad f*C 888 I**
- E 8F***""
the' R H R syeten are . , _ _ _ detimod im Scottom l 4' -, .-
- 11. .
g g, . gg, .
--W'The ASME Code components of the HHR 4R A hydrostatic test will be conducted on -M. The results of the hydrosta9 test of the System retain their pressure boundary those Code components of the RHR ~ ASME Code components of the RHR -
Integrity under internal pressures that will System that are required to be System conform with the requirements in be experienced during service. hydrostaucally tested by tho' ASME Code the ASME Code, Section 111 I
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CONTANMENT SYSTEM FEEDWATER 'N
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m - LEOEND: R. REMOTE SHUTDOWN SYSTEM Y , ALL ELECTRICAL POWER LOADS FOR TIE COtPONENTS SHOWN 2 N g 2 ON THIS FIGURE ARE POWERED FROM DMSION I EXCEPT FOR THE RHR 3 3 RHR OUTBOARD CONTAWMENT ISOLATION VALVE FOR SHUTDOWN g COOUNG.WHICH IS DM980N E. a , 5 y Figure 2.4.1a Residual Heat Removal (RHR-Al System
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- N.--g -
IVtR 'l FPC FPC RHR SYSTEM - - PRIMARY FPC CONTAINMENT )L 3 2 l, 3 FPC SYSTEM -- m3 2 u a mm > J n Au - i RPy . g ' unS
-ge(- _ _ _h - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ l r, , ,r l 2
1 2 g I i y--__ p . I~P 4
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@MAMPuer if TO FROM 2 mm 2 RCW C RCW C 3 g fg RHR 3 CW RCW ALL ELECifuCAL POWER LOADS FRO THE COMPONEN!C SHOWN ON THtS FIGURE ARE POWEFIED FROM DIVISION Ill EXCEPT FO81 THE OUTBOARD CONTAINMENT
. ISOLATION VALVE FOR SHUTDOWN COOLING. WHICH IS OfVIS40N L 5 ' Figure 2.4.1c Residual Heat Removal (RHR-C) System h ,
s 4 h lll2 h A+ty.fil msntg n' GB hl LMr/{
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- ABWR onian occumnt 2.1.2 Nuclear Boiler System Design Description General System Description >
The primary functions of the Nuclear Boiler System (NBS) are:- . (1) to deliver steam from the Reactor Pressure Vessel (RPV) to the Main Steam (MS) ystem. (2) provide containment isolation of the Main Steam Lines (MSLs), (3) to deliver feedwater from the Condensate, Feedwater, and Air ;' Extraction (CFDWA) System to the RPV, (4) , to provide overpressure protection of the Reactor Coolant Pressure - Boundary (RCPB), (5) to provide automatic depressurization of the RPV in the event of a IAss of Coolant Accident (LOCA)_where the RPV does not'depressurize rapidly and the high pressure makeup systems fail to_ adequately mamtain the water level in the RPV, and . (6) to provide instrumentation to monitor the drywell pressure and RPV ' pressure, metal temperature, and water level. , Within the NBS, the FW lines, the MSLs and the MSL drain' lines are located in ' the drywell and the steam tunnel. With the exception of the instrumentation - attached directly to the RPV or NBS piping, the NBS instrumentation _is located within the Reactor Building. See Figures 2.12a,2.1.2b, and 2.1.2c, for the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class. Main Steam Lines The MSLs direct steam from the RPV to the MS System. The NBS contains only _ the portion of the MSLs from their connection to the RPV to the boundarywith ' the MS System, which occurs at the seismic interface located downstream of the outboard Main Steam Isolation Valves (MSIVs). Figures 2.1.2a and 2.1.2b show._ the general configuration of the MSLs and the MSL drain lines. The MSLs are classified as Seismic Category I from the reactor pressure vessel out to the seismic interface shown in Figure 2.1.2b. The MSL drain lines provide a flow path for the Main Steam Isolation Valve s (MSIV) leakage during an accident. The pneumatic operated valve in the MS _ A drain line shown in Figure 2.1.2b opens should either electric power to the vahYs 2.1.2 1 1/21/93 n
ABWR ocsign occument __ actuating solenoid be lost, or pneumatic pressure to the valve be lost. The MSL drain lines from the MSLs to the main condenser are seismically analyzed to withstand the Safe Shutdown Earthquake (SSE). The total steam volume of the steam lines, from the RPV to the main steam turbine stop valves and turbine bypass valves, is greater than or equal to 113.2 m.3 Each MSL has a flow limiter. The MSL flow limiter consists of a flow restricting venturi which is located in eacn RPV MSL outlet nozzle. The restrictor limits the coolant blowdown rate from the RPV in the event a MSL break occurs outside the containment to a (choke) flow rate equal to or less than 200% of rated steam flow at 72.1 kg/cm 2g upstream pressure. The throat diameter of the MSL flow limiters is not greater than 355 mm. The MSL flow limiter also serves as a flow element to monitor the MSL flow. Instruments lines are prosided to monitor the pressure at the throat of the MSL flow limiter. The RPV steam dome pressure instrument lines are used to provide the pressure upstream of the MSL flow limiter. The MSL flow limiters limit the loss of coolant from the RPV following a MSL rupture outside the containmeng Main Steam isolation Valves Two isolation valves are located in a horizontal run of each of the four main steam lines; one valve is inside of the drywell, and the other is near the outside of the primary containment pressure boundary. The MSIVs are Y-pattern globe valves.The MSIV's primary actuation mechanism for opening and closing is pneumatic. Springs close the MSIV if pneumatic pressure to the, MSIV actuator is lost. The MSIV closing specci is equal to or greater than 3 and less than or equal to 4.5 seconds when Ng or air pressure is admitted to the MSIV actuator. When all the MSIVs are closed, the total leakage through the MSIVs for all four MSLs is less than or equal to 66.1 liter per minute at 20* C and one atmosphere pressure absolute. Feedwater Lines The Feedwater (RV) lines direct Feedwater from the CFDWA System to the RPV. The NBS contains only the portion of the RV lines from the seismic interface located upstream of the Motor-Operated Valves (MOVs) to their connections to the RPV. Eigure 2.1.2c shows the portion of the FW lines within the NBS. The FW piping consists of two nominal 550 mm diameter lines from the RV supply header. Isolation of each line is accomplished by two containment 2.1.2 2- 1/21/93
o 4 ABWR oesign Document isolation valves consisting of one check valve inside the drywell and o'ne positive closing check valve outside the containment. The FW line isolation check valves are qualified to withstand a FW line break outside containment. The feedwater line upstream of the outboard isoladon valve contains a hiotor-Operated (h!O) valve, and a seismic interface restraint. The feedwater piping is classified as Seismic Categon I from the hiO shutoff valves to the RPV. The ash!E Boiler and Pressure Vessel Code Class 2 piping from thg Control Rod Drive (CRD) System, the Reactor Water Cleanup (CUW)
" SystemeReactor Core Isolation Cooling (RCIC) System, and the Residual Heat Removal (RHR) System, shown in Figure 2.1.2c is also classified as Seismic Categog I.
Safety / Relief Valves The nuclear pressure relief system consists of Safety / Relief Valves (SRVs) located on the htSLs between the RPV and the first isolation valve, i.e. the inboard htSIV, within the drywell. These valves protect against overpressurization of the nuclear system. Figures 2.1.2a,2.1.2b and 2.1.2d show the general configuration of the SRVs, and the SRV discharge lines. The rated capacity of the pressure-relieving devices is sufficient to prevent a rise in p,ressure within the RPV of more than 110% of the design pressure (96.7 kg/ cm' gauge) for design basis events. The SRV discharge line is designed to achieve critical flow conditions through the valve, thus providing flow independence to discharge pipe losses. Each SRV has its own discharge line. The SRV discharge lines terminate at the quenchers located below the surface of the suppression pool. The SRV discharge lines are classified as Seismic Category I. The SRVs provide three main protection functions: (1) Overpressure safety operation: The valves function as safety vakes and open to prevent nuclear system overpressurization-they are self-actuating by inlet steam pressure if not already signaled open for rei;ef operation. Table 2.1.2a identifies the SRV spring set pressures and flow capacities. The opening time for the SRVs, from the time the pressure exceeds the valve set pressure to the time the valve is fully open,is less than or equal to 0.3 seconds. (2) Overpressure relief operation: The valves are opened using a pneumatic actuator upon receipt of an automatic or manuallyinitiated signal to reduce pressure or to limit pressure rise. 2.1.2 3 1/21/93
ABWR oesign occument For overpressure relief valve operation (power actuated mo^de).. pressure sensors on the RPV generate a RPV high pressure trip signal which is used to initiate opening the SRVs, The relief (power actuated) mode of operation is initiated when an electrical signal is received at - any of the CRV solenoid valves. The SRV pneumatic operator is so arranged that, ifit malfunctions,it does not prevent the SRV from opening when steam inlet pressure reaches the spring lift setpoint. (3) Automatic Depressmization System (ADS) operation: The ADS 5alves open automatically as part of the Emergency Core Cooling System (ECCS) for events invohing breaks in the nuclear system process barrier. Automatic depressurization by the ADS is provided to redu .- the reactor pressure during a LOCA in which the High Pressure Core Flooder (HPCF) System and/or the Reactor Core Isolation Cooling (RCIC) System are unable to restore water level. Eight of the eighteen SRVs are designated as ADS valves and are capable of operating from either ADS logic or safety /relieflogic signals. Table 2.1.2a identifies the ADS SRVs. The ADS consists of redundant trip channels arranged in two divisionally separated logics that control two separate solenoid-operated gas pilots on each ADS SRV. Either pilot can operate the ADS valve. These pilots control the pneumatic pressure applied by the accumulators and the High Pressure Nitrogen Gas Supply (HPIN) System. The DC power for instrumentation and logic is obtained from the Safety System Logic and Control (SSLC) Division I and II. Sensors from all four dhisions for low reactor _ water level and high drywell pressure and Division I control logic signal actuste one set of pilots, and sensors from all four divisions for low reactor water and high drywell pressure and Dhision II control logic signal actuate the second set of pilots, either of which initiates the opening of the ADS SRVs. Upon receipt of an kPV tow water signal the ADS automatically initiates. If the RPV low water level signal is present concurrently with high drywell pressure signal, both the main ADS timer (less than or equal to 29 seconds) and the high drywell pressure bypass timer (less than or equal to 8 niinutes) are initiated. Absent a concurrent high drywell pressure signal, only the ADS high drywell pressure bypass timer is initiated. Upon the time out of the ADS high drywell pressure bypass timer, concurrent with RPV low water level signal, the main ADS timer is initiated, if not already initiated. The main timer continues to completion and times out only in the continued presence of an RPV low water 2.1.2 1/21/93
ABWR oesign occument level signal. Upon time out of the main ADS timer, concurrent with positive indication of at least one RHR or one HPCF pump running, the ADS function is initiated.
'Be ADS can also be initiated manually. On a manual initiation signal, concurrent with positive indication of at least one RHR cr one HPCF pump is running, the ADS function is initiated.
SRVs have individual non-safety-related accumulators. In addition, those with ADS function have separate safety related accumulators with separate redundant gas power actuators / e c ch s y 03.@. The ADS accumulator capacity is sufTicient to open the SRV with the dqwell pressure at design gauge pressure following failure of the pneumatic supply to the accumulator. The SRVs can be operated individually in the power actuated mode by remote manual switches located in the main control room. They are prosided with position sensors which provide positive indication of SRV disk / stem position. Temperature sensors are located on the discharge pipe of the SRVs. NBS Instrumentation The purpose of the NBS RPVinstrumentation is to monitor and proside control input during plant operation. The NBS contains the instrument lines and instrumentadon for monitoring the reactor pressure and water level. Tir dryscggsc eq'id-RPV-meta,l
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- craperr.wslThe NBS contains the sensorrfFigure 2.1.2eThows the drywell pressure and RPV instrumentation in the NBS.
The safety-related NBS instrumentation is located in separate disisional areas. W6s Pressure instrumentation deseet reactor vessel internal pressure from the same instrument lines used for measuring reactor vessel water level. The RPV coolant temperatures are determined by measuring saturation pressure (which gives the saturation temperature), outlet flow temperature to the Reactor Water Cleanup (CUh) System ap the RPV bottom head drain line temperature (instrumentation in the dystem).The reactor vessel outside surface (metal) temperatures are measured at the head flange and the bottom head locations. During plant operation, either reactor steam saturation temperature and/or inlet temperatures of the reactor coolant to the CUW System and the RPV bottom head drain can be used detennine the RPV coolant temperature.
- 2. L2 1/21/93
.1 ABWR onion occumnt
= Figure 2.1.2e shows the water level instrumentation. The instruments that ser se the water level are differential pressure devices calibrated for specific RPV-pressure (and corresponding liquid temperature) conditions. The water level _ .
measurement design is the condensate reference chamber type. Instrument zero , for the RPV water level ranges is the top of the active fuel. The NBS contains the instrument lines to monitor the differential pressure across the_ RPV pump deck and core support plate. The instrumentation which~ actually performs these functions is located within the Reciculation Flow Control - System. The NBS also contains the drywell pressure instrumentation used to generate the safety-related high drywell pressure trip LOCA signal. The Reactor Protection System (RPS) utilizes this signal as a scram initiation signal.The Leak Detection and Isolation System (LDS) utilizes this signal to initiate containment isolation. The Emergency Core Cooling Systems (ECCSs) utilizes this signal as a system initiation signal. At % The NBS control room ^ indications and controls allows for monito ' and control during operational conditions. The control room has nTorA v h" and/or control of the ADS, RPV water level, RPV pressure, drywell pressure, SRVs, FW line outboard check valves, and RPV metal temperature. Remote Shutdown System (RSS) Interfaces NBS components with status indication and/or controlinterfaces with the RSS are shown on Figure 2.1.2a and 2.1.2c. Environmental Qualification
;. The safetyJrelated electrical equipment (including instrumentation and controis) shown on Figures 2.1.2b,2.1.2c,2.1.2d, and 2.1.2e, located in the containment, gteam tunnel and Reactor Building, is qualified for a harsh environment.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.1.2b provides a definition of the inspections, tests and/or analyses together with associated acceptance criteria which will be undertaken for the-NBS.
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2,1.2 -6 1/21/93
ABWR oesign occument Table 2,1.2a: Nuclear System Safety / Relief Valve Setpoints Set Pressures and Capac8tles Nameplate ASME Rated Capacity Number' of Spring Set Pressure at 103% Spring Set Used For SRVs Valves (kg/cmr gitt Pressure (kg/hr each)* ADS J. P 2 80.8 395,000 B,G,M,S 4 81.5 399,000 0, E, K. U 4 82.2 002,000 C,H,N,T 4 82.9 400,000 X A,F,L,R 4 83.6 409,000 X
- Eight o'the SIC , serve in the autornatic deprepurisation system function, ti Spring set prenure tolerances as perinisted ty the ASME Boiler and Pressuie Venel Code, Section !!!.
1 Minirnurn capacity per the AShlE Boiler and Prenure Venet, Section Ill. 4 i 2.1,2 +7- 1/21/93
{ Table 2.1.2b: Nuclear Boiler Systern Inspections. Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests. Analyses Acceptance Criteria
- 1. The basic configuration of the NBS is 1. Inspections will be conducted for the NBS 1. Tha as-built NBS conforms with the basic shown in Figures 2.1.2a, b, c, d, and e. System. configuration shown in Figu.es 2.1.2a. b. c.
- d. and e.
- 2. Each Main Steam Iine (MSL) has a flow 2. Inspection will be performed on the throat 2. The throat diameter of the MSL flow limiterlocated in the RPV MSL outlet diameter of the MSL flow limiters, which limiters is less than or equal to 355 mm.
- l nozzle. The throat diameter for the MSL are located within the RPV MSL outlet flow iimiter is less than or equal to 355 . nozzles.
mm.
- 3. The ASME Code components of the NBS 3. A hydrostatic test of the ASME Code 3. The results of the hydrostatic test of the System retain their pressure boundary components of the NBS System will be ASME Code components of the NBS integrity under intemal pressure.s that will conducted. System conform with the requirements in be experienced during service. The ASME Code. Section 111
- 4. The combmed volume of the four Main 4 Using as-built dimensions of the steam 4. The combined steam line volume is greater
. Steam Lines (MSLs) ard branch lines from lines volumetric analysis will be performed than or equal to 113.2 m3. ? the RPV to th9 main steam turbine stop to determine the combined main steam valves and steam bypass valves is greater line volurne.
than or equal to 113.2 m3. g. btc ^' Sa. Control Roordind!7.ations and/or controls Sa. Inspections wilig performed on the Sa. indications and/or controts exist or can be provided for NBS are defined in Section Control Roorynddations and/or controls retrieved in the Control Room as defined in 2.1.2. for the NBS. Section 2.1.2. Sb. Remote Shutdown System IRSS) Sb. Inspections will be performed on the RSS Sb. Indications and/or contro!s exist on the indications and/or controls provided for the indications and/or controls for the NBS. RSS as defined in Section 2.1.2. NBS are defined in Section 2.1.2.
- 6. The Main Steam Isolation Valve (MSIVs 6. Tests will be conducted to determine the 6.. Tbe MSIV closing time is equal to or closing time is equal to or Dreater than 3 closure time of the MSIVs. greater than 3 and less than or equalit. L5 and less than or equal to 4.S seconds when seconds.
N2or airis admitted into the valve pneumatic actuator. 6 5 w
- Table 2.1.2b: Nuclear Boiler System (Continued) ,
s inspections. Tests Analyses and Acceptance Criteria Certified Design Commitment inspections Tests, Analyses Acceptanca Criteria g A
- 7. When all four MSIVs are closed. the 7. Leakage tests will be performed to 7. MSIV 12akage for all four MSLs less than or combined leakage through the MSIVs for determine the leakage through the closed equal to 66.1 liters per minute at 20' C and all four MSts is less than or equal to 66.1 MSIVs. 1 atmosphers absolute pressure.
liters per minute at 20' C and 1 atmosphcre absolute pressure.
- 8. The SRV spring set pressure and capacities 8. Tests and analysis in accordance with the 8. The SRVs have the capacities and set are given in Table 2.1.2a. The opening time* ASME Boiler and Pressure Vessel Code will pressures shown on Table 2.11a.
for the SRVs from when the pressure bo performed to determine the spring set The opening time for the SRVs from the exceeds the valve set pressure to when the pressure, capacity and opening time of ; valve is fully open is less than or equal to each SRV. g . , 03 secod is less than or equal to 03 seconds. Sa. Upon receipt of an RPV low water leve! 9a Using simulated signals, tests will be Sa. signal the ADS logic automatically initiates. performed of the automatic ADS initiation i Upon receipt of a low water levei signal. I gic- concurrent with a high drywell pressure
? signal, at the input to the ADS initiation logic, the followQ occurs:
- 1) The main ADS timerinitiates and continues to time out in the cor.tinued presence of the RPV low water level signal The time delay for the main ADS timer is less than or equal to 29 seconds.
- 2) Upon time out of the main ADS timer.a concurrent signal that represents positive indication of at least one RHR or HPCF pump running. an ADS actuation signal is generated to the associated ADS valve solermids. .
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4 t Table 712b: Nuclear Boiler System (Continued) - 4 Inspections. Tests. Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests Analyses Acceptance Criteria 9a. (Continued) Sa. IContinued) Sa. II Upon receipt of a low water level signal. in the absence of a high dryweII pressure signal, at the input to the ADS initiation logic, the following occurs-
- 1) The ADS high dryweII pressure bypass
. timer initiates. TI.e time delay for the ADS high drywell pressure bypass timer is less than or equal to 8 minutes
- 2) Upon time out of the ADS high drywell pressure bypass timer, concurrent with a n RPV low water level signal. the main ADS timer initiates and continua to 3
time out in the continued presence of
- 2. the RPV low water level signal.
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- 3) Upon time out of the main ADS timer, i concurrent with a signal that represents posittve indication of at 4 least one RHR or HPCF pump runn~mg.
an ADS actuation signalis generated to the assocated ADS valve solenoids. 9b. Upon receipt of a manualinitiation signal 9b. Tests will be performed of the manual ADS Sb. Upon receipt of a manualinitiation signal the ADS logic initiates. initiation logic. at the input to the ADS initiation logic. concurrent with a signal that represents positive indication of at least one RHR or HPCF pump running, an ADS actuation signal is generated to the associated ADS valve solenoids. 8 w
i g Table 2.1.2b: Nuclear Boiler System (Continued) - 4 Inspections, Tests. Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests. Analyses Acceptance Criteria
- 10. The SRV ADS accumulators have the 10. An analysis and/or type test will be 10. Either-capacity to open the SRVs one time with performed to demonstrate the capacity of TN SRV ADS mMtm W W the drywell at the drywell design pressure. the SRV ADS accumulators. ca acity to lift the stem of the SRVs to the full open position one time with the drywell pressure at. or above the drywell design pressure. or
. b. the SRV ADS accumulators have the capacity to lift the stem of the SRVs to the full open position five time with the dr)well at atmospheric pressure, and an analysis that shows that 5 SRV 14fts at atmospheric pressure demonstrates the capability to open one time with the drywell at the drywelt design pressure.
- 11. Class 1E loads of the NBS are powered 11. Tests will be performed for the NBS by 11. The test signal exists only in the Ctass 1E 3*
from Class 1E Divisions, as described in providing a test signal in only one Ctass 1E Division under test in the NBS. Section 2.1.2. Division at a time.
- 12. The MSL drain lines from the MSLs to the 12. An inspection of the stress report 12. The existence of a stress report will be main condenser are seismically analyzed containing the dynamic analysis of the confirmed. This report documents that a to withstand the SSE. - piping will be conducted. dynamic seismic analysis has been performed.
- 13. Springs close the MSIV if pneumatic 13. Tests will be performed to demonstrate 13. The MSIV closes when pneumatic pressure pressure to the MSIV actuator is lost. that the MSIV will stroke to the fully closed is removed from the MStV.
position upon the loss of pneumatic pressure to the MSIV actuator.
- 14. The pneumatic operated valve in the MSL 14. Tests will be performed to demonstrate 14. The MSL pneumatic drain line valve shown drain line shown in Figure 2.1.2b opens that the MSL drain line valve opens when in Figure 2.1.2b opens when either electric should either electric power to the valve pneumatic pressure to the valve is lost, or power to the valve actuating solenoid is actuating solenoid be lost, or pneumatic electric power to the valve actuating lost. or pneumatic pressure to the valve is pressure to the valve be Icst. solenoid is lost. lost.
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e e US/ABVR E31/LDS ITAAC Jan 20, 1993 2.4.3 Leak Detection and Isolation System (LDS) Design Description . . . . . LDS is classified as afsafety-related(fla_ss3; control and instrumentation system whose function is to detect and monitor leakage from the reactor coolant pressure boundary (RCPB) and initiate isolation of the leakage source. The system is designed to initiate automatic isolation of the process lines that penetrate the containment by closing the isolation valves. The functions of LDS include isolation of the main stenalines, the primary and secondary containment, and individual systen process lines; activation of the standby gas treatment systear the monitor of leakages inside and outside the primary containment; and indicating the monitored leakage parameters in the control room. The LDS design is fail safe, single failure croof and redundant. The LDS logic design uses two-out-of-four voting in initiation of each isolation function. Also, the logic is der,igned to incorporate channel bypass provisions to permit channel test and repair. In the bypass mode, the trip . logic utiizes two-out-of-three voting for initiation of the isolation functions. The LDS safety-related channel measurements are provided as inputs to th6 safety system and logic control (SSLC) system for signal processing, setpoint comparisons, and generation of the trip signals that initiate the isolation functions.\0nce isolation is initiated, the logic seals in the isolation signal and operator action is required to reset the logic to its normal state. The following primary and secondary containment isolation and automatic control fucctions are provided by LDS using four instrument channels to monitor leakage: (1) Closure-of the main steam line (MSL) isolation valves and main steam drain line valves on a signal indicating low reactor water level, high MSL flow in any steam line, high ambient temperature in MSL tunnel area or in Turbine Building, low main condenser vacuum, or low steam inlet pressure to the main turbine. (2) Isolation of the' Reactor Vater Clean-up (CUV) systen process lines on a signal indicating low reactor water level, high ambient MSL tunnel area temperature, high mass-differential flow, high ambient temperature in the CUV equipment areas, or when the Standby Liquid Control System (SLCS) is activated. I (3) Initiation of the Standby Gas Treatnent System (SGTS) on a signal l indicating high drywell pressure, Itw reactor water level, high i radiation in the secondary containnest or high radiation in the fuel handling area. (4) Isolation of Reactor Building Heating, Ventiletion and Air Conditioning (HVAC) system on a signal indicating high drywell pressure, low reactor water level, high radiation in the secondary containment or high j radiation in the fuel handling area. i l l
. _ - . _ - - _ _ - . - - - . -_ ~ --
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(5) Isolation of containment purge and vent lines on a signal indicating j' high drywell pressure, low reactor water level, high radiation in the secordsry containment or high radiation in the fuel handling area. l l (6) Isolation of the Reactor Building Cooling Water (RCV) and of the HV/,C l Normal Cooling Vater (HNCW) system lines on a signal indicating high ' drywell pressure or low reactor water level. (7) Isolation of the Reactor Heat Removal (RHR) shutdown cooling system loops on a sig al indicating high reactor pressure or low reactor water level. Also, h RHR shutdown cooling loop is individually isolated on a signal indi ating high ambient temperature in the RHR loop equipment area. (8) Isolation of the Reactor Core Isolation Cooling (RCIC) steam line to the RCIC turbine on a signal indicating high steam flow in the RCIC line, low steam pressure in the RCIC line, high RCIC turbine exhaust pressure, or high ambient temperature in the RCIC equipment area. (9) Isolation of the Suppression Pool Cleanup (SPCU) systek on a signal indicating high drywell pressure or low reactor water level. (10) Isolation of the riammability Control System (FCS) on a signal indicating high drywell pressure or low reactor water level. (11) Isolation of the drywell sump pump discharge lines on a signal indicating high drywell pressure or low reactor water level. Also, each discharge line is individually isolated on a signal indicating high radioactivity in the discharged liquid waste. ; (12) Isolation of the fission products monitor dryvell sample and return lines on a signal indicating high drywell pressure or low reactor water level. (13) LDSprovides(ig_the_peutronmontoringsyst6masignalindicatingahigh dryvellpressureorlowreactorwaterleveh.(J As shown in Tigures 2.4.3, the LDS isolation logic consists of safety related sensors, redundant instrument channels and logic trip units that initiate the automatic isolation functions. Also, separate manual controls in the control room are provided in LDS design for logic reset, MSIV operational control, MSIV closure tests, and for manual isolation. , LDSprovidesthefollowingcontrolsignalstoeachMSIVpilotsolenoidvalvef
- 1. Tour divisional control signals are provided to each MSIV solenoids 12 &
$3 to open the valve. MSIV closure is automatic on loss of any two divisional signals to both solenoids.
- 2. Two divisional control signals are provided to each MSIV test solenoid ,
11 to exercize partial valve closure. Division I or III is used to test [ close the outboard MSIVs and Division II or IV is used to test close the inboard MSIVs. L
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i e i i Also, LDS provides three separate divisienal isolation signals (Divisiens I, ! II and III) for automatic closure of the primary and secondary containment f isolation valves. Each LDs divisional isolation signal initiates closure of the isolation valves that are assigned in the same division. The LDS design includes the following manual controls for separate isolation , of the RCIC system, and closure of the MSIVs and the primary and secondary containment isolation valves:
- 1. Tour MSIV isolation switches - one per Divisions I, II, III, and IV.
Closure of all the MSIVs requires the actuation of two divisional MSIV isolation switches, either Divisions I and IV or II and III.
- 2. Three primary and secondary containment isolation switches - one per Divisions I, II and III.
Each isolation switch closes its respective divisional isolation valves in the primary and secondary centainment, except for the MSIVs and RCIC.
- 3. Two RCIC isolation switches - one per Divisions I and II.
Either isolation switch isolates the steam line to the RCIC turbine and causes turbine trip. Division I closes the inboard while Division II closes the outboard isolation valves. Manual reset logic functions are provided at the divisional level to initialize the logic and for logic reset after a isolation has been initiated. Separate reset functions are provided in the LDS logie design for the MSIVs, the RCIC, and the primary and secondary containment isolation circuitry. The LDS controls and indications are provided in the control room to allow for monitoring and control during operational conditions. The indications in the control room consist of the monitored leakage parameters as defined under the LDS functions. Each LDS divisional channel is powered from the same divisional power source. > Independence is provid'ed between the Class IE divisions, and also between the Class IE divisions and the non-Class IE equipment. g The LDS safety related components and associated hardware are qualified Seismic Category I. The safety related LDS sensors and associated wiring in the reactor and turbine buildings are qualified to operate in a harsh environment. Inspections, Tests, Analyses and Acceptance' Criteria Table 2.4.3 provides a definition of the 4 sp0'tions, tests. and/or analyses together with associated acceptance criteria tor the Leak Detection and Isolation System. ~
.o Table 2.4.3 .-
II.AK DETECTION & ISOLATION STSTEM Inspections. Tests., Analyses and Acceptance Criteria Certified Design Inspections. Tests Analyses Acceptance Criteria Commitment
- 1. The equipment 1. Inspection of the as built 1. The as built LDS system comprising the LDS system will be conducted. ccnforms with the description is defined in ., in Section 2.4.3.
Section 2.4.3.
- 2. LDS monitors and 2. Each IDS instrument channel 2. Each channel trips.
detects leakages shall be tested using simulated from the RCPB, and signal inputs to test the trip initiates. closure condition. of primary and l' secondary containa-ent isolation valves.
- 3. The LDS' isolation 3. The instrument channels of each 3. Isolation signal is initiated logic uses four LDS isolation function shall be when at least any two out of redundant .
tested using simulated signal four channels have tripped. instrument channels inputs. to monitor each RCPB leakage parameter. The
' isolation signal is initiated when any two out of four .
channels have tripped.
o Table 2.4 3 (CONT *D) - LEAK DETECTIDS & ISOLATION SYSTEM Inspections. Tests Analyses and Acceptance Criteria Certified Design Inspections. Tests Analyses Acceptance Criteria Commitment
- 4. The LDS isolation 4 In channel bypass mode, each 4 Isolation signal is initiated logic incorporates LDS logic isolation function when at least any two out of channel bypass shall be tested using simulated three channels have tripped.
provisions for on signal inputs. line testing and repair. In this mode, the isolation signal is initiated when any two out of three channels have tripped. sg
- 5. Each MSIV can be 5. Actuate each MSIV test switch 5. Each MSIV partially closes subjected to a to check partial closure of the and then reopens partial closure valve. automatically when its test test from the switch is actuated.
control room.
- 6. LDS provides. 6. a. Simultaneously actuate two 6. a. Closure of all the MSIVs separate manual of the four MSIV isolation occurs only when Divsions I &
controls in the switches (Div. I & IV or Div. IV or II & III switches are control room for II & III) to close all the actuated. g lj((V eclosure, for 4,el. Lion of the MSIVs Repeat the same test by b. Isolation of the RCIC actuating the other two MSIV system occurs when Div. I primary and , r.' E isolation switches. switch closes the inboard or
- secondary - b. Actuate each RCIC isolation Div. II switch closes the containmenth,dhand switch (Div. I or II) to outboard Isolation valves.
for isolation of isolate RCIC. c. Each divisional primary the RCIC system. c. Actuate each prLeary and and secondary containment secondary containment isolation isolation switch close only
^
switch (Div. I, II & III) to .its respective containment isolate the containment. Isolation valves.
e Table 2.4.3 (CONT *D)
- LIAK DETECTION En ISOIATION SYSTEM Inspections. Tests. Analyses and Acceptance Criteria certified Design Inspections. Tests. Analyses Acceptance Criteria Commitment
- 7. Manual reset 7. Tests will t,e performed using 7. The logic circuitry resets controis are the IDS reset functions in the for normal operation.
provided to perform3 control room. at < ** +ei re se t functions as described in i- Section 2.4.3.
- 8. Control room 8. Inspections will be performed 8. Controls exist and indications and on the control room indications indications exist or t be controls'for this and controls for this system. retrieved in the control room 4
system are defined as defined in Section 2.4.3. In Section 2.4.3.
- 9. IDS logic design is 9. Tests will be conducted to 9. The faulted channel trips.
fail-safe, such simulate electrical power that loss of failure to each divisional IDS electrical power to channel. '
. one LDS divisional I
logic channel initiates a channel trip.
._ - ~ - .- ._ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._
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10 The divisional LDS 10 Tests will be performed on the 10 The test signal exists only logic. channels and IDS system by providing a test in the Class IE division of associated sen:. ors signal in only one Class IE the LDS system utider test. are powered.fros division at a time. Class IE divisional power.. 11 Independence is 11 Inspection of the installed LDS 11 Physical separation exists in
.provided in,the Class,IE divisions will be LDS between Class IE system between performed. divisions, and between the Class IE divisions, Class IE divisions and the and between Class non-Class IE equipment.
IE divisions and non-Class IE
- . equipment.
A4
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C",,*" "Diagrar P Oric Of Four Divtsens** {LD3) Figure 2.4.3/ Leak Detection and Isolation System, Interface Diagram
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2.11,3 Reactor Building Cooling Water System
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{%b f j } l Design Description
'Ihe Reactor Building Coc ng Water (RCW) syster stributes cooling water through three physically nd electrically sepommi lhisions b :,. 'n u,n, >;nenen, iushiten mh.,e which me, nni ,1,r,+2iti. o.nor,ted). The system r removes heat from plant auxiliaries and transfers it to the Ultimate Heat Sink (UHS) via the Reactor Service Water (RSW) system. The RCW system removes heat from the Emergency Core Cooling system (ECCS) equipment includmg the emergency diesel generators (DGs) during a safe reactor shutdown cooling ;
function. RCW system configurations are shown in Figures 2.11.Sa, b, and c. All : components cooled by the RCW system are parts of other systems an,d are not l part of the RCW system. The RCW system performs safe reactor shutdown cooling function following a g ,_ho d loss-of coolant accident / loss-of offsite power (LOCA/ LOOP), assuming a single active failure in any mechanical or electrical dhision ?'"" '. ,r or RCW po p*j gg support system?in case of a failure which disables any one of the three RCW dhisions, the other two dhisions perform safe reactor shutdown cooling. tV W Yg,# (65 g h )dA LOCA p signal does the following: (a) starts any standby RCW pumps N ht* Y' &O
# M (b) opens any closed RCpeat hanger outlet exc&
valves 9 I N Coc fikg WD8P i (c) opens all RHR heat exchanger (tiet valves g,iP1 (d) clogay RCW containment isolation valves Q cit, d Fw " UN
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q (e) closes v;.lves to non safety rela ec. ,comporgAts t.Xc+ Angers ad ll A nts 4 overndes any low water level signal wlIich' stops operating RCW pumps) ((O (f) op s the RCW water temp aIur o Ivalve (lectbA M - esamstMa* of UCW e)tAa Sg d ew -
\{gogalves separate the safety related. portions p _ from of the the non safety-relateyC// m ^1 A ~7m during a LOCA. The isolation-Y Y,iKeiTopr u c non satery relateo RCW system are automatically or remote-manually operated, and their positions are indicated in the main control room, f9 Each RCW dhision includes two pumps which circulate cooling water through the equipment cooled by the RCW system and through thrce heat exchangers which transfer the RCW heat to the UHS via the RSW system.
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ABWR onion Docuant Ihe RCW system main control room (MCR) indication and controls allows ; monitoring and contral during operational conditions. The control room has control of motor operated valves (MOVs) and pneumatically operated valves g (POVs), the number of RCW pumps and heat exchangers in operation and flow M rate to the components being cooled. Main Control Room instrument ON indication is provided for each RCW division for surge tank level, cooling water radiation level and Residual Heat Removal (RHR) heat exchanger flow rate and temperature water supply temperature and pressure, water radiation level, surge tank water level. The RCW system components with status indication and/or control interfaces with the Remote Shutdown System (RSS) are shown in Figure 2.11.3a, b and c.y , The RCW ASME Code classilgtions for different portions of the system are "2. ll.34, M( indicated on FigutT6T1.3Ja . The safety related portions of the RCW divisions are classified as Seismic Category I and are located in Seismic Category I structures. Component design parameters are: Division A/B Division C H ;.; ey , ~ ,une, neg /m2,) 72 M g-- ' yu e- /*C 70 Discharge flow rate Ipm/ 2 21,700 2 18,200 C Nw ~2 u t'= pump) w : = = - Heat exchanger capacities are each: 2 11.s x to' kcal/hr 2 10.6 x 10' LCal/hr Connecuons to a radiauon monitor are provided in eCh division to detect radioactive contamination resulting from a tube leak in a heat exchanger. {
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The RCW pumps and heat exchangers are located in the lower floors of the
. Control Building. The equipment cooled by the RCW divisions are located in the
, Rgactor Building, Turbine Building, and Radwaste Building.
Control B '
'b./ b C 1building Figures .11.3ayj All safetyTetztEn electncal equi rn$
are m.m.o.,suely qualified fodharsh _ 3 M[I .#show which equipment receives RCW fiow during various plant operatiesg.g,
$8#' emergency conditions. The-tables alva indicate how many heat exchangers are ggjgj($
pf in senice under each condition. The following cooling loads are in all three RCW divisions: RHR heat exchangers, RHR motor and seal coolers, RHR toom coolers and HECW I refrigerators. All other cooling loads are in either one or two RCW divisions. For these coollong loads, changes may be made in the divisional assignment ifit can be shown that the cooling capacity margins in all divisions have not been significantly reduced by the changes. 1/11/93 2.11.3
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e 6 h The RCW h'e'atsexchanger heat removal capacities in 1 de 20 per cent margibb the design heat removal- quirements which aret 2.8[10 7 or division 2.84x1l6 kcal/h for division B and 26.4 6P 1/h for division C. [r(,+bLC) e Caco +0ll-<o n . /
i ABWR coston Documut ing and During 7 top requires co ling; tion, W :dsoR,s 0 iws'thrfugh Hows thr,onk s ety r ated tooler i ' equip cept [mpnghich iR i ca e h er an coo rs, as shown g , x: _
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A separate surge tank of at least 16m is provided for each RCW division. Each surge tank is shared with the corresponding division of the HVAC Emergency W Cooling the Stakeup _ ater Water (HECW) (Purified) (htCWP) system. system byStakeup automatic or water is p[rovided MCR signal. Low water level signals in the surge tanks 3rd -T, - do the following (in order of decreasing level): (a) (low) opens the MUWP makeup water valve (b) (low low) closes the pneumatic and motor operated valves which stop Dow to the non-safety related components [(c) (low low low) stops any operating RCW pumps The Suppression Pool Cleanup (SPCU) System provides a backup surge tank water supply, waK* W The pneu atic valves of safety significance fail as follows: The MUWP --jrp watervalv fall n amethe RCW water temperature control valves fail / openy 4## Ill/ AllU H sts,llNAnalyses V59 Mlb 0and 6 00Acceptance Criteria Inspections, Table 2.11.3a provides a definition of the inspections, tests, and/or analyses together with associated acceptance criteria which will be and undertaken for the RCW system. l l l l-1/11/93 3-2.11.3 I
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- Inspections, Tests, An:. lyses cnd Acceptenco Criteria I Certified Design C mmitment inspections. Tests. Analyses Acceptance Criteria "
- 1. i basic configurationh the RCW system spections of the as-built system 1. The as-built ':
isphown on Figures 2.11.3a, b and c. 1. Mses*g% c: ' ; - will be conducted -_ ;' n RCW as ~ . system i; _ . . . -
, =., Figqres , I 2.113a, b and c (f"M's 'd A "5
- 2. The ASME m m r b: ty code 2 A hydrostatic test will be conducted on co mbs3mes H en fineaalon -ek !
y components of the RCW system retain those code components of the RCW 2 The results of the hydrostatic test of the C j theig tegrity under intamal pressures that ASME code components of the RCW will be experienced durin system required to be hydrostatically system conform with the requirements in
' tested by the ASME Code. L 3a Control-room uc,.S::- :%g . ca h ols serdvi e. the ASME C grovided for RCW 3a. Inspections will be performed on the 3a % O .S kipoills c aJ com1wls system are defined in section 2.113. Control Room ' - -- gexist or can be retrieved in the +
the RCW ES g g gg system.
#^ Control Hoom as defmed in Section 2.113.
3b. Remote Shutdown System (RSS15 m
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- 36. Inspections will be performed on the RSS provided for the RCW system arch'efined ' n :for the RCW system 3bJ h e exist on the RSS as defined in e
' Sectronill 3.
in Section 2.11.3. C., N,6dt4%r etM Cobb 3 E *\ e.s -^'sM16 leoh3> i p .. ok
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- 4. M "
the RCW System are powered from ' _ Cass'A I4 Tests will De penoidieu i,.a u e 4. Safety-related eleictrical power loads or l 1E Divisions, as described in Section L components in Section 2.11.3 by checking the RCW System are powered from Class voltage at the electrical loads. 2.11.3' 1E Divisions, as described in Section 5. Wechica.I The safety-related portion of eachffivision ACLI - 2.11 3 - 1
- 5. L'is.aal nspections of the as-built system of the RCW system (ieeps A, B,C)is will be rformed. M 5. I A room outside the control room and physically separated. I gjpg primary containment does not contain i
safety-related mechanical components ' from more than one loop of the RCW ( system. 6 The RCW System responses to a LOCA
- 6. Using simulated LOCA signals, tests will be 6. Upon receipt of simulated LOCA signals.
signal arejpecified in Section 2.11.3. I g performed for the RCW System g the respons#of the RCW System ::: o
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4 ' in 7. Systemy , c : _'!: tests will be-et nductedjn 7.
- z. it.1-q /Sectiorf 2.1 d the ope tationar specified14Tabi s 2.11.3tp ' 'CKe pump discharge flowg . the r YTi[e RpW he 3 j xchanger/capacith
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AME entation rpew 5ha!! be [& .4 e nest removal capaci*N the RCW h t to detpfmine the he remova} I ex:han ca the h'est enchan ers / / Th ~'gers **' &M-'#: t eO ES A
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- 9. The RCW pump flow capac- 9. An analysis of the 9. The estimated heat ities and the RCW heat ex- as-built RCW System will removal capacities of the changer heat removal capaci- be performed. Tests will as-built RCW System divi-ties are as specified in be performed of the flow sions exceed the estimated Section 2.11.3. capacities of the heat removal requirements of installed RCW pumps. the components cooled by the Inspections and analysis RCW System divisions during will be performed to- LOCA conditions.
estimate the heat removal capacities of the RCW heat exchangers. Inspections and analysis will be performed to estimate the heat removal requirements of the as-built components which are cooled by the , RCW System during LOCA conditions. 11 0 . A surge tank with a ca- .10. Inspection and a 10. The capacity of each pacity'of at'least 16 cubic volume calculation using surge tank / is greater than meters is provided for each as-ouilt dimensions will or equal ta 16 cubic meters. RCW division. be performed.
- 11. -Motor-operated valves 11. Opening and/or clos- 11. Each I40V opencand/or" (M0V)' designated in Section closes.
ing tests'of installed 2 .11.3,as having an active valves will be conducted f safety-related function will under pre-op differential : open and/or close under .dif- pressure, fluid flew, and ferential pressure and fluid temperature conditions. flowconditions. ,
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ABWR onian Document Table 2.11.3b: Reactor Building Cooling Water Genowmens Cool --LoohS. Division A Hot - Operating Model wmol Ste Emwm Opersting Shutdown (lossof (LOCA) Co w n # Conditions AC Poweri RCW/RSW Heat 2 3 3 3 Exchangers in Service SAFETY RELATED Emergency Diesel - - X X Generator A RHR Heat Exchanger A - X X X , FPC Heat Exchanger A X X X X Others (zafety related) X X X X NON4AFETY-RELATED RW'CU Heat Exchanger X X X - Inside Drywell X X X - Others (non+sfety- X X X X related) Notes (x) = EquQment receives RCW in this mode. M = Equipment does not receive RCW in this mode.
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4- 1/11/93 2.11.3
Some of these' cooling loads are serviced.by only one or two RCW-divisions, These components may be reassigned > to other RCW divisions if redundancy and divisional alignment of supported and supporting systems is main-tained and the design basis cooling capacity of the RCW divisions is assured. N 4 l l
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. ABWR ossian Document Table 2.11.3c: Reactor Building Cooling Water Genowmass (do /M Id Divit 5 n B ~/ ,
Het Emer N per Shutdown le g Power) 2 3 3 3 RCW/RSW Heat Exchangersin Service SAFETY 4tELATED i Emergency Dioeel - - X X Generator B RHR Heat Exchanger 8 - X X X FPC Heat Exchanger B X X X X Others (eefety-related)N X X -X X NON-SAFETYJIELATED RWCU Heat Exchanger X X X - Inside Drywell X X X - Others (non-sofety- X X X X reisted) Notes (x)- Equipment receives RCW in this mode. (.) . Equipment does not receive RCW in thle mode.
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-.ABWR ossign 0:cument .. --1 Table 2.11.3d: Reactor Building Cooling Waterh Cooll Ag kools:
Division C 0 - Het - NonnM 8' Einergency Operating Mode / Shutdown ***
"8 Components W (LOCA)
Condrtions AC-Power) ,
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RCW/RSW Heat 2 3 3 3 Exchangers in Service SAFETY 4 ELATED Emergency Diesel - - X 'X Generator B RHR Heat Exchanger - X X X B Others(safety related X X X X NON-SAFETY 4 ELATED Others (Non safety. X X X X related) Notes (x) . Equipment receives RCW in this mode. , H = Equipment does not receive RCW in this mode. Wb / ffkkI$!O' gcgyyb.g.,.po.wclm,wkeaoMsea) a week at em s
-8 1/11/93 2.11.3
o ABWR onsign Document wMuWP rn R ew , -
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ABWR Design 0:cument 3 MUWP RCW SYSTEM SYSTEM RCW OTHERS OTHERS RCW Sf M SY T M T MR } (SUROE RHRHu T R** dor BuddM)TANK l l 3 (Reactor Building) - _) I DO HX QM
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RCW OTHERS OTHERS RCW FE -- -- R 8l CUW PUMPA -3 ~ I uddng) NNS , ,'I"'"C ,, , , NNSP M 1 ___ __ __ _r T,_________] f OTH R N-SAFETY-RELATED) I (- __ _ _ _ , , _
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R = REMOTE SHUTDOWN SYSTEM THIS DIVISION IS POWERED FROM CLASS 1E DIVISION 11. EXCEPT FOR THE CONTAINMENT OUTBOARD ISOLATION vat.VE WHICH IS POWERED FROM DIVISION lit. i Figure 2,11.3b RCW Division - B
-10 1/11/91 f 2.11.3 1
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THIS DMSON IS POWE9ED FRC8A CLASS 1E DMSION ilt l' i l Figure 2.11.3c RCW Division . C l 1/11/93 2.11.3 L
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ABWR Design occument 2.11.9 RFACTOR SERVICE WATER SYSTEM Design Description The Reactor Senice Water (RSW) System removes heat from the Reactor Building Cooling Water (RCW) System and rejects this heat to the Ultimate Heat - Sink (UHS). The portions of the RSW System that are in the Control Building are within the Certified Design.Those portions of the RSW SystenLthat are outside the Control Building are not in the Certified Design. 'b I . +s ,,y ..k,,, L . ~x- ;j The RSW System is Seismic Category I and ASME Code Section III, Class 3 bd oY consists of three separate safety'related divisions. Each division is powered by its respective Class lE Division and is located in a separate room in the Control Building. The RSW System Control Room indications and controls allow for monitoring and control during operational conditions. The control room has controls and
'adications for motor-operated valve open/close status. On a LOCA signal, any closed valves for standby heat exchangers will be automgcally opened. The RSW System components with status indication and Apr control interfaces with he Remote Shutdown System (RSS) are identified in Figure 2.11.9.
Interface Requirements The portions of the RSW System which are not part of the Certified Design shall meet the following requirements. Design features shall be provided to limit the maximum flood height to 5.0 meters in each RCW heat exchanger room. The design shall have three divisions which are physically separated. Each division shall be powered by its iespective Class 1E Division. Each division shall be capable of removing the design heat capacity (as specified in Section 2.11.3) of the RCW heat exchangers in its division. Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.9 provides a definition of the inspections, tests, and/or analyses together with associated acceptance criteria for the portions of the RSW System within the Certified Design.
-1 1/21/93 2.11.9
Table 2.11.9: Reactor Service Water System Inspections, Tests, Analyses and Acceptance Criteria Certified Design Cornmitment inspections, Tests, Analyses Acceptance Criteria
- 1. The basic configuration of the RSW System 1. Inspections of the as-built system will be 1. The as-built RSW System conforms with is as shown on Figure 2.11.9. conducted. the basic configuration shown in Figure 2.11.9.
- 2. The ASME Code components of the RSW 2. A hydrostatic test will be conducted on 2. The results of the hydrostatic test of the retain their pressure boundary integrity those Code components of the RSW ASME Code components of the R$W under internal pressures that will be System required to be hydrostatically conform with the requirements in the experienced during service.
- tested by the ASME Code. ASME Code, Section lit.
3a. Control rc,om indications and controls 3a. Inspections will be performed on the 3a. Indications and controls exist or can be provided for RSW System are defined in control room indications and controls for retrieved in the control room as defined in Section 2.11.9. the RSW System. Section 2.11.9. 3b. Remote Shutdown System (RSS) 3b. Inspections will be performed on the RSS 3b. Indications and controls exist on the RSS indications and controls provided for the indications and controls for the RSW as defined in Section 2.11.9. RSW System are as defined in Section System. p 2.11.9.
- 4. Each mechanical division of the RSW 4. Inspections of the as-built system will be 4. Each mechanical division of the RSW System (Divisions A, B. C) is physically performed. System is physically separated from other separated. mechanical divisions of the RSW System.
- 5. Any closed standby heat exchanger inlet or 5. Using simulated LOCA signals, tests will be S. Upon receipt of simulated LOCA signals, outlet valves automatically open upon performed on standby heat exchanger inlet the standby heat exchanger inlet and outlet receipt of a LOCA signal. and outlet valves. valves open.
- 6. Class 1E loads for the RSW System are 6. Tests will be performed on the RSW 6. The test signal exists only in the Class 1E powered from Class 1E Divisions, as System by providing a test signal in only Division under test in the RSW System.
described in Figure 2.11.9. one Class 1E Division at a time.
- 7. Motor-operated valves designated in 7. stalled 7. Each MOV opens and/or closes.
Section 2.11.9 as having an active safety Opening and valves will be /or closing conducted and.(/6tests of ip[/fff function will open and/or close under preoperational differential pressure, fluid differential pressure and fluid flow flow, and temperature conditions. conditions. 53 8 w
- ABWR Design Document.
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- RSW SITE SPECIFIC SCOPE THIS FIGURE SHOWS ONE OF THREE IDENTICAL DIVISIONS. ALL ELECTRICAL POWER LOADS FOR THE COMPONENTS IN DIVISIONS A. B. AND C ARE POWERED FROM DIVISIONS 1,11, AND ill, RESPECTIVELY.
VALVES SHOWN ABOVE IN OtVISIONS A AND B HAVE CONTROLS AND OPEN/CLOSE STATUS INDICATION ON THE REMOTE SHUTDOWN PANEL. Figure 2.11.9 Reactor Service Water System 3- 1/21/93 2.11.9
m- . 3.
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'ABWR oesign occument 4.5 REACTOR SERVICE WATER SYSTEM--
Interface Requirements Interface requirements are in Section 2.11.9. b e
-4.5 4 1/21/93
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.. u ABWR Gesign occument 2,6 Reactor Auxiliary 2.6.1 Reactor Water Cleanup System Design Description The Reactor Water Cleanup (CUW) System as shown in Figure 2.6.1 removes particulate and dissolved impurities from the reactor coolant by circulating a portion of the reactor coolant through a filter.demineralizer.
The CUW System removes excess coolant from the reactor system dudng
~
startup, shutdown and hot standby The excess water is directed to the radwaste or suppression pool.The CUW System also provides processed water to the head spray noule for Reactor Pressure Vessel (RPV) cooldown. The CUW System reduces RPV temperature gradients by maintaining circulation in the bottom head of the RPV during periods when the reactor internal pumps are unavailable. The containment isolation valves (CIV) automatically close upon receipt of an isolation signal from the Leak Detection and Isolation System (LDS). The suction valves (containment isolation valves) are designed to close against a 2 maximum differential pressure of 87.9 kg/cm within 30 seconds upon receipt ofisolation signal. The inboard containment isolation valve is powered from
- Class 1E Division II AC bus, and the outboard containment isolation valves are fed from Class IE Division I AC bus.
The CUW suction line is provided with a flow restrictor which provides flow restricting and flow monitoring functions. Maximum throat diameter is 135 mm. ! The CUW System is classified as a non safety-related system with the exception of the primary containment isolation function. The major portion of the system is f ! located outside of the primary containment. The safety related electrical equipment (including instrumentation and controls shown in Figure 2.6.1) located in the containment and reactor building is qualified for a harsh environment. CUW system piping and components from the RPV out to and including the i outboard isolation valves are part of the reactor coolant pressure boundary and are designed to ASME Code Class I requirements and classified as Seismic Category I.The remainder of the piping system is designed to ASME Code Class l 3 requirements and classified as non-Seismic Category I. l 1 1/21/93 2.6
s s ABWR D:sion Document i The vessel bottom head drain line is connected to the main CUW suction piping . by a tee. The center line of the tee connection is at'an elevation of at least 460 - mm above the center line of the nriable leg nozzle of the RPV wide range water levelinstrument. The_CUW System control room indication and controls allows for monitoring: and control during operational conditions. The control room has indication for 7 and/or control of the containment isolation valves. Inspections, Tests, Analyses and Acceptance Criteria Table 2.6.1 provides a definition of the instructions, tests, and/or analyses together with associated acceptance criteria which will be undertaken for the - CUW System, 2 1/21/93 2.6.1
o Table 2.6.1: Reactor Water Cleanup Systeen { Inspections, Tests, Analyses an( Acceptance Criteria Certified Design Commitments inspections, Tests, Analyses Acceptance Criteria
- 1. A basic configuration _ for the CUW System 1. Inspection of the as-built system w31 be 1. The as-built CUW System conforms with is as shown in Figure 2.6.1. conducted. the basic configuration show in Figure 2.6,1.
- 2. Motor-operated valves (MOV) designated 2. Opening and/or closing tests of insta!!ed 2. Each MOV opens and/or closes. The in Section 2.6.1 as having an active safety- valves will be conducted under pre-op following valves open and/or close in the related function will open and/or close differential pressure, fluid flow, and following time limits:
under differential pressure and fluid flow temperature conditions. cond!tions. . Valve Time Open/Cose Suction line equal or Oose inboard CIV less than 30 sec. S Suction line equal or Oose outboard OV less than 30 sec.
- 3. The ASME Code components of the CUW 3. A hydrostatic test will be conducted on 3. The results of the hydrostatic test of the System retain their pressure boundary those Code components of me CUW ASME Code components of the CUW integrity under internal pressures that will System required to be hydrostatically System conform with the requirements in be experienced during service. tested by the ASME Code. the ASME Code, Section III.
- 4. Control room Indications and/or controls 4. Inspections will be performed on the 4. Indications and/or t.ontrols exist or can be provided for CUW System are as defined in control room indications and/or controls retrieved in control room as defined in Section 2.6.1 for the CUW System. Section 2.6.1.
- 5. Maximum throat diameter of the CUW 5. Inspection will be corsducted on the CUW 5. Maximum throat diameter of the CUW suction line flow restrictor is 135 mm. suction line flow restrictor throat diameter. suction line flow restrictor is 135 mm.
- 6. Center line of the vessel bottom head drain 6. inspection will be conducted on the 6. The center line of the vessel bottom head
. line tee connection is at least 460 mm elevation of the center line of the vessel drain line tee connection is at least 460 m m 5 above the center line of the variable leg bottom head dra!n line tee connection. above the center line of the variable leg 3 noIIIe of the RPV wide range water level nozzle of the RPV wide range water level - instrument. Instrument.
- . , ,6
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5
- THE INBOARD PRIMARY CONTAINMENT ISOLATION VALVE IS POWERED FROM DIVISION 11 AND THE OUTBOARD ISOLATION VALVES ARE POWERED FROM DIVISION t.
RPV CUW 1 RHX - REGENERATIVE
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Figure 2.6.1 Reactor Water Cleanup (CUW) System Schematic
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'ABWR onion Docuent 2.2,4 Standby Liquid Control System _
The Standby Liquid Control (SLC) System injects neutron absorbing poison using a boron solution into the reactor and thus providdback up reactor shutdown capability independent of the normal reactivity control system based-on insertion of control rods into the core. The SLC System is capable of . . operation over a range of reactor pressure conditions which bo'und the elevated pressures associated with an anticipated plant transient coupled with a failure to scram (ATWS) when SLC is required to operate. The SLC System is designed to bring the reactor, at any time in a core cycle, and - at design basis conditions, from full power to a subcritical condition, with the reactor in the most reactive xenon. free state, without' control rod movement. With the storage tank at minimum level and both pumps operating, the system willinject the minimum required boron solution. The SLC System as shown in Figure 2.2.4 is located in the Reactor Building which is classified as seismic Category I and consists of a boron solution' storage tank,- 4 two positive displacemen t pumps, two motor-operated injection valves which are provided in parallel for redundancy, and associated pi} ing and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The borated solution is discharged through the 'B' high pressure core . flooder (HPCF) subsystem sparger. Key equipment performance requirements - are:
.(!) - Pump flow (minimum) 22.7 m3 /hr with both pumps running
-(2) Maximum reactor pressure 88.9 kg/cm2 a (for injection)
(3) Pumpable volume in storage 23.1 m 3 tank (minimum) The SLC System is designed to be manually initiated from the main control room. Each of the two divisions is controlled by a separate switch. When it is - manually initiated to inject a liquid neutron absorber into the reactor, the following devices are actuated by each division switch:- (1) The specified division injection valve is opened. (2) The specified division suction valve is opened. (3) The specified division injection pump is started. (4)- The reactor water cleanup isolation valves are closed.
-2.2.4 . 1 1/19/93 4 ?- yy9- m-4 \--l,%u y w m- --- , h
s c ABWR oesign occument The SLC System is automatically initiated during an ATWS. When the SLC System is automatically initiated to inject a liquid neutron absorber into the reactor, both divisions are actuated. The SLC System provides borated water to the reactor core to compensate for the various reactivity effects. These effects arc xenon decay, elimination of steam voids, changing water density due to the reduction in water temperature, Doppler effect in uranium, changes in neutron leakage, and changes in control rod worth. To meet this objective, it is necessary to inject a quantity of boron which produces a minimum concentration of 850 parts per million (ppm) by weight of natural boron in the reactor core at 20 C. To allow for potential leakage and imperfect mixing in the reactor system, an additional approximately 25% (220 ppm) is added to the above requirement resulting in a total requirement of greater than or equal to 1070 ppm. The required average concentration is thus achieved in a mass of water equal to the sum of the mass of 3 water in the RPV at normal water level (less than or equal to 455 x 10 kg) plus the mass of water in the RPV shutdown cooling piping (less than or equal to 130 x 103kg). At least this quantity of boron solution is contained above the pump suction shutofilevel in the tank. The SLC System pumps are capable of producing discharge pressure to inject the solution into the reactor when the reactor is at pressure conditions corresponding to the system relief valve setpoint (109.7 kg/cm2 gauge), which is above peak ATWS pressure. The~ SLC System control room indications, alarms, and controls allow for monitoring and control during operational conditions. The control room has indication for pump discharge pressure, storage tank liquid level, injection and suction valve open/close, and pump on/off. The SLC System uses a dissolved solution of sodium pentaborate as the neutron - absorbing poison. This solution is held in a storage tank which has a heater to maintain solution temperature above the saturation temperature.Tne heaer has automatic actuation and automatic shutoff features.The SLC System storage tank, a test water tank, the two positive displacement pumps, and associated vahing are located in the secondary containment on the floor elevation below the operating floor. Each of the SLCS divisions is powered from the respective Class 1E division. The power supplied to one motor-operated injection valve, suction valve, and injection pump is powered from Division I. The power supply to the other motor-operated injection valve, suction valve, and injection pump is powered from Disision II. 1/19/S3 2.2.4
1 < , ABWR Design Document SLCS components required for RPV injection are classified as Seismic Category I. A test tank and associated piping and valves permit system testing. The tank is supplied with demineralized water which is pumped in a cl6 sed loop~"through . either pump or injected into the reactor. The SLC System is physically separated from and independent of the hydraulic ' portion of the Control Rod Drive System which performs the scram function. Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.4 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance enteria, which will be undertaken for the - SLC System. . L d 9 m i 1 L 3- 1/19/93 2.2.4 i t:
Table 2.2.4: Standby Liquid Control System {
- Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The perforrnance of the SLCS is based on 1. The as-built dimensions will be used in a 1. a. Storage tanks pumpable volume is the following plant parameters: volumetric analysis to calculate the greater than or equal to 23.1 m3.
v lumes listed below. b. RPV water inventory is less than or
- a. Storage Tank pumpable volume is greater than or equal to 23.1 m3. a. Minimum Storage tank pumpable equal to 455 x 103kg at 20*C.
volume c. RHR shutdown coolin0 system
- b. RPV water inventory is less than or inventory is less than or equal to 130x equal tn 455 x 103 kg at normalwater b. RPV water inventory at normal water level and 20*C. 103kg at 20*C.
level and 20*C. .
- c. RHR shutdown cooling system c. RHR shutdown cooling system water inventory is less than or equal to 130 x inventory at 20*C.
103kg at 20*C.
- 2. The basic configuration of the SLC System 2. Inspections of the as-built system win *n 2. The as-built SLC System conforms with the is as shown in Figure 2.2.4. conducted. basic configuration shown in Figure 2.2.4.
3a. The SLC System delivers at least 378 iiters/ 3a. Using installed controls, power supplies 3a. The SLC System injects greater than or
. minute of solution with both pumps and other auxiliaries, demineralized water equal to 378 liters / minute into the RPV with I operating against the elevated pressure will be injected item the storage tank into both pumps running against a discharge conditions in the reactor during events the RPV with both pumps running against pressure of greater than or equal to 88.9 involving SLC System initiation. a discharge pressure of greater than or kg/cm23, equal to 88.9 kg/cm2 ,,
3b. The SLC System delivers at least 189 liters / 3b. Using installed controls, power supplies 3b. The SLC System injects greater than or minute of solution with either pump and other auxiliaries, demineralized water equal to 189 liters / minute into the RPV with operating against the elevated pressure will be injected from the storage tank into either pump running against a discharge conditions in the reactor during events the RPV with one pump running against a pressure greater than or equal to 88.9 involving SLC System initiation. discharge pressure of greater than or equal kg.cm23, to 88.9 kg/cm23, 3 u
a y Table 2.2.4: Standby Liquid Control Systern (Continued) , i inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 4. The SLC System is designed to permit 4. Using installed controls, power supplies 4. a. Domineralized water is pumped with a pump flow during plant operation. and other auxiliaries, the following system flow rate greater than or equal to 189 tests will be conducted for each SLCS liters / minute in the closed loop.
division after system installation */ b. Demineralized water is injected from
- a. Demineralized water will be pumped the test tank into the RPV.
against a pressure greater than or equal to 88.9 kg/cm a in e closed loop on the test tank.
- b. Demineralized water will be injected from the test tank into the RPV.
- 5. Class 1E loads of the SLC System are 5. Tests will be performed on the SLC System 5. The test signal exists only in the Class 1E powered from Class 1E Divisions, as by providing a test signal in only one Class Division under test in the SLC System.
described in Section 2.2.4. 1E Division at a time.
- 6. The ASME Code components of the SLCS 6. A hydrostatic test will be conducted on 6. The rdsults of the hydrostatic test of the 4, retain their pressure boundary intogrity those Code components of the SLCS that ASME Code components of the SLCS under internal pressures that will be are required to be hydrostatically tested by conform with the requirements in the experienced during service. the ASME Code. ASME Code, Section llL
- 7. Control room alarms, indications, and 7. Inspections will be performed on the 7. Alarms, indications, and controls exist or controls provided for the SLC System are control room alarms, indications, and can be retrieved in the control room as defined in Section 2.2.4. coatrols for the SLC System. defined in Section 2.2.4.
- 8. The SLC pumps have sufficient NASH. 8. Tests will be conducted by injecting 8. The available NPSH exceeds the NPSH demineralized water using both SLCS required when the SLCS injects greater pumps from the storage tank to the RPV than or equal to 378 liters / minute.
with conditions in the storage tank of low level (down to pump trip level) and a temperature of greater than or. equal to 43*C.
- 9. The SLCS automatically initiates both 9. . Using simulated ATWS signals, test will be 9. Upon receipt of a simulated ATWS signal, divisions upon receipt of an ATWS signal. performed on the SLC System initiation . SLCS logic functions to automatically logic. initiate both divisions.
- 10. The SLCS can be manually initiated with' ' 10. Using the SLCS manualinitiation switch, 10. The SLCS initiates by the SLCS manual y the system initially in the standby mode for SLCS testing will be performed. initiation switches with the system initially g each division. in the standby mode for cach division.
w
c, y Table 2.2.4: Standby Liquid Control System (Continued)
- b inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 11. The SLCS pump relief valves open when 11. Shop or field tests will be performed to 11. The SLCS pump relief valves open when the inlet pressure to the valve equals or determine the relief valve setpoint. the inlet pressure to the valve equals or exceeds the setpoint (109.7 kg/cm2g). exceeds 109.7 kg/cm2g,
- 12. Motor-operated valves (MOV) designated 12. Opening and/or closing tests of instrued 12. Each MOV opens and/or closes.
in Section 2.2.4 as having an active safety- valves will be conducted under pre # . related function will open and/or close differential pressure, fluid flow, ant under differential pressure and fluid flow , temperature conditions. conditions N u
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ABWR oeston occument 2,10 Power Cycle 2.10.1 Turbine Main Steam System Design Description The Main Steam (MS) System as shown in Figure 2.10.1 supplies steam generated in the reactor to the turbine, steam auxiliaries and steam turbine bypass valves. The MS System ranges between, but does not include, the seismic interface restraint to the turbine stop and turbine bypass valves.The system includes the steam auxiliary valve (s). The MS System is designed. (1) to accommodate operational stresses such as internal pressure and dynamic loads without failures. (2) to provide a seismically analyzed fission product leakage path to the main condensers. (3) with suitable access to permit in senice testing and inspections. (4) to close the steam auxiliary valve (s) on MSIV isolation signal. These valves fall closed on loss of electrical power to nlve actuating solenoid or loss of pneumatic pressure. The MS System main steam piping consists of four lines from the seismic interface restraint to the main turbine stop valves. The header arrangement upstream of the turbine stop valves allows them to be tested on line and also supplies steam to the power cycle auxiliaries, as required. The MS System is analyzed, fabricated and examined to ASME Code Class 2 requirements, classified as non Seismic Category 1, and subject to pertinent QA requirements of Appendix II,10CFR Part 50. Insenice inspection shall be performed in accordance with ASME Section XI requirements for Code Class 2 piping. ASME authorized nuclear inspector and ASME Code stamping is not required. MS piping from the seismic interfr.cc restraint to the main stop, main turbine bypass and the steam auxiliary valve (s) is analyzed to demonstrate structural integrity under safe shutdown earthquake (SSE) loading conditions. 1 1/21/93 2.10
4 5 ABWR costo.,cocument l Inspections, Tests, Analyses and Acceptance Criteria - Table 2.10.1 provides a definidon of the inspections, tests, and/or anal >ws, i
- together with associated acceptance criteria which ill be undertaken for the MS '
System. s t l m i i I L 4 I r 1
-E 2 V20/93-2,10.1
~ . - - . _ . _ . . . . . . . _ . , _ _ . _
L *l Tatde 2.10.1: Main Steam System Inspections, Tests, Analyses and Acceptance Criteria r-Certified Design Commitment 1 ;: ^*_-5 2, Tests. Analyses Acceptsace Criterie
- 1. The besic configuration for the MS System 1. Construction records will be reviewed and 1. The as-built configuration of the MS l
is as shown in Figure 2.10.1. visual inspections will be conducted for the System is in accordance with the
~
configuration of the MS System description in Figure 2.10.1.
- 2. MS piping from the seismicinterface 2. Perform a seismic analysis of the as-built 2 The results of the seismic analysis show
. restraint to the main stop, realn turbine MS piping. that the MS piping can withstand a safe -
l bypese and the steem auxiliary Mive(s) is shutdown eenhquake withoutloss of onelyzed to demonstrates structural structuralintegrity. Integrity under SSE loeding conditions. r-.. 7ed5 3. The SA valve (s) closes following receipt of 3.~ Upon receipt of a Main Steam loolation 3. "' ; ' = = i _"^ "r ., a... will be Valve (MSIV) closure signet, the Steem performed using simulated MSIV closure e simulated MSIV closure synal Auxillery (SA) volve{s) cioees. signalf --9 ' ? ^, a _ _;;; e: . t
' 4. The pneumatically operated steem 4. ' Test will be performed one steem auxiliary 4. Steem auxiliary velvels) closes.
- auxiliary volve(slin the main steem system . velve(s). .
closes when either electrical power to the !' Y . valve actuating solenold is lost or . ' pneumatic pressure to the volve(s)is lost.'
- 5. The ASME Code components of the MS $$4. : A hydrostatic test will be conducted on 5. The results of the hydrostatic test'of the retain their pressure boundary integrity those Code components of the MS System ASME Code componentsof the MS System
- i. under internal pressures that will be ' required to be hydrostatically tested by the conform with the requirements in the experienced during service. : ASME Code. ASME Code, Sechon Ill.
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uA6:mre Standard Plant PIV A these are not available, accepted industry or classified as 00 B and inspected in engineering practice is followed. accordance with applicable portions of American Society of Mechanical Engineers 3.2.5.3 Main Steam Line Leakage Rate (ASME) Section XI. This portion of the steamline is classified as non seismic The ABWR main steam leakage path utilizes the Category I and analyzed using a dynamic large volume and surface area in the main steam seismic analysis method to demonstrate its piping, bypass line, and condenser to hold up and structural integrity under SSE loading plate out the release of fission products conditions, flowever, all pertinent QA following postulated core damage, in this requirements of Appendix D,10CFR Part 50 manner, the main steam piping, bypass line, and are applicable to ensure that the quality condenser are used to mitigate the consequences of the piping materialis commensurate with of an accident and are required to remain its importance to ser ty during normal functional during and after an SSE. operational, trat ,nt, and accident conditions. The main steam lines and all branch lines 21/2 inches in diameter and larger, up to and The seismic interface restraint provides a including the first valve (including lines and structural barrier between the selsraic valve supports) are designed by the use of au Category I portion of the main steamline in appropriate dynamic seismic system analysis to the reactor building and the non seismic withstand the safe shutdown carthquake (SSE) Category I portions of the main steamline design loads in combination with other in the turbine building. The seismic appropriate loads, within the limits specified. Interface restraint is located inside the The mathematical model for the dynamic seismic scistdic Category I building. The analyses of the main steam lines and branch !!ne classification of the main steamline in the piping includes the turbine stop valves and turbine building as non seismic Category I piping to the turbine casing and the turbine is consistent with the classification of bypass valves and piping to the condenser. The the turbine building, dynamic input loads for design of the main steam lines in the reactor building and the control At the interface between scismic and building are derived from a time history model non seismic Category I main steam piping analysis or an equivalent method as described in system, tha. seismic Category I dynamic Section 3.7. analyses will be extended to either the first anchor point in the non seismic Figure 3.21 depects the classification system or to a sufficient distance in the requirements for the main steam leakage path as non seismic system so as to not degrade the described below. validity of the selsinic Catt,gery I analysis. (1) Main steam piping from the reactor pressure vessel up to and including the outboard (4) To ensure the integrity of the remainder of isolation valve is classified as QG A (SC 1) main steamline leakage path, the following and seismic Category I. requirements are met: (2) Main steam piping beyond the outboard (a) the main steam piping between the isolation valve up to the scismic interface turbine stop valve and the turbine restraint and connecting branch lines up to inlet, the turblue bypass line from the the first normally closed valve is bypass valve to the ccadenser, and the classified as QG B (SC 2) and seismic main steam drain line fros.. the first Category I. valve to the condenser are not required to be classified as safety related nor as seismic Category 1, but are analyzed (3) The main steamline from the seismic interface restraint up to but not including using a dynamic seismic analysis to the turbine stop valve (including branch demonstrate their structural integrity lines to the first normally closed valve) is under SSE loading conditions, and Amendment W
. o a m man ABWR nry A Standard Plant (b) the condenser anchorage is seismically analyzed to demonstrate that it is capable of sustaining the SSE loading conditions without failure.
A plant specific walkdown of non seismically designed systems, structures, and components overhead, adjacent to. and attached to the main steam leakage path (i.e., the main steam piping, the bypass line, and the main condenser) shall be conducted to confirm by inspection that the as built main steam piping, bypass lines to the condenser, and the main condenser are not ' compromised by non seismically designed systems, structures and components. 3.2.6 Quality Assurance Structu.res, systems, and components that per. form nuclear safety.related functions conform to the quality assurance requirement of 10CFR50 Appendix B as shown in Table 3.21 under the heading. 'Ouality Assurance Requirements,' and in Table 3.2 2. Some NNS structures, systems, and components meet the same requirements as noted on Table 3.21. The Quality Assurance Program is described in Chapter 17. i 3201 Amendment
C 4 ABWR msimt 51111dard Plant prv A TABLE 3.21 CLASSIFICATION SUhth!ARY (Continued) Quality Group Quality j 1.aca. Classi. Assursace Selsmic i Safet4 fication Eggulttme nt' I Prineinal Comnonen1" QAM 112a' Cetenorv h I B2 Nuclear Boller System (Continued) 1
- 4. Piping including supports 1 C,5C A B !
main steamline (MSL) and feed-water (FW)line up to and in-duding the outermost isolation valve
- 5. Piping induding supports-(a) MSL (induding branch linea 2 SC B B i (t) to first valve) from outermost isolation valve up to and induding s:ismic interface restraint (b) FW (induding branch lines 2 SC B B I (ee) to first valve) from outer.
mast isolation valw to and induding the shutoff valve
- 6. Piping induding supports MSL N SC,T B E -
(r) (induding branch lines to first valve) from the seismic interface restraint up to but not induding the turb'.ne stop valve and turbine bypass valve
- 7. Piping bepad FW shutoff valw N SC D - -
(ee)
- 8. Deleted 9, Deleted -
! 10. Pipe whip restraint . MSL/FW 3 SC,C - B - (dd) if needed 11, Pipinginduding supports-other ! within outermost isolation valves
- a. RPV head vent 1 C A B 1 (g)
- b. Main steam drains 1 C.SC A B 1 (g) l l 12. Piping induding supports other beyond outermost isolation or shutoff valves 12-9 Amendment
i ABWR maime prv A Standard Plant TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safet4 Ims. Classi. Assurance Seismic Prindoal Comoonent' CIA 18 112A fication Reaulternent' Catenorv h B2 Nuclear Boller System (Continued)
- a. RPV head vent beyond N C .C E -
shutoff valves
- b. Mais steam drains to N SC,T B B I (t) first valve
- c. Main steam dtains N SC.T D - -
(c) beyond first valve MI Amendment
tMN 22AstooAs prv A Standard Plant TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Loca- Classi. Assurance Selsmic Saferg ibn* nestion Reautrement' CateroryI h Enneloal Comeetital' Can B2 Nuclev Boller System (Continued)
- 13. Piping including supports- 2/N. C,5C B/D B/E 1/- (g) instrumentation up to and beyond outermost isolation valves C A B !
- 14. Safety / relief valves 1 C,M A B I
- 15. Valves. MSL and FW 1 isolation valva, and other FW valves within containment
- 16. Valves . FW, other beyond 2 SC B B i (ee) outermost isolation valves up to and including shutoff valves
- 17. Valves . within outermost isolation valves C A B I (g)
- a. RPV bead vent!
1 C,SC A B _! (g)
- b. Main steam drains
- 18. Valves, other 3 C C B 1
- a. RPV head vent
- b. 1st Main steam drain valve N SC B B ! (r)
- c. other main steam drain valves N SC D - -
(r)
- 19. Deleted C,5C - B I
- 20. Mechanicalmodules.instrumen. 3 tatien with safety related function l
21 Electrical modules with safety- 3 C,5C,X - B I (i) related function 3 C,SC,X - B I
- 22. Cable with safety.related function l
l l 3.2 10 l Amendment
ABWR nasimin prv 4 Standard Plant TABLE 3.21 CIASSIFICATION
SUMMARY
(Continued) Quality Group Quality Saferg Loce. Classi. Assurance ~ Seismic ' Princloal Component' Cllu ll2n fication Eto'ilrement' Catenm Entn K1 Radwaste System L Drain piping induding supports N ALL D E - (p) and valves radioactive (except R LX) ALL D E - (p)
- 2. Drain piping induding supports N and valves . nontadioacive 2 C,SC D B 1
- 3. Piping and valves .
containment isolation N C,5C B B I
- 4. Piping induding supports and valves forming part of containment boundary
- 5. Pressure vessels induding N W - E - (p) supports N C SCJI, - E - (p)
- 6. Atmospheric tanks induding supports T,W N W - E - (p)
- 7. 015 PSIG Tanks and supports N C,SC,W - E - (p)
- 8. Heat exchangers and sup' ports N C.SCJI, - E - (p)
- 9. Piping induding supports and valves T,W N ALL -E - (p)
- 10. Othei mechanicaland eleancalmodules 3 SC C B I
- 11. ECCS equipment room sump backSow protea.
los check valn:s N1 Turbine Mala Steams Systene L Deleted (See B2.5) 12-2u Amendment I
kWW mum pry i Standard Plant TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality loca. Clasal. Assurmace Selsale Safet4 # I g Ettnc! cal Component
- CIAH 112 5 Ik11198 Renutrement' Catenon N1 Turbine Mala Steam System (Continued)
- 2. Deleted (See B16)
N2 Condensate, feedwater and Condensate Air Estraction System
- 1. Main feedwater line (MFL) N SC B B 1 induding supports from second isolation valve branch lines and components and including to outboard shutoff valves
- 2. Feedwater system components N T D E -
beyond outboard shutoff valve N3 !! cater. Drain and Vest System N T - E - N4 Condensate Purineation System N T - E - N5 Condensate Filter Facility N T - E - N T - E - N6 Condensate Deminerallmer N T - E - N7 Main Turbine N8 Turbine Control Systes
- 1. Turbine stop valve, turbine N T D - -
(i)(n)(o) bypass valves, and the main (r) steam leads from the turbine stop valve to the twbine l, casing l l l i i 1 3121.3 Amcodmean l
- uuioort ABWR prv 4 Standard Plant TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Locs. Classi. Assursace Seismic Safet4 I Printinal Component' QRu llan* fication Reaufrement' Catenorv bu N9 Turbine Gland Steam System N T D E - N10 Turbine labricating Oil System N T - E - N11 Motsture Separator Hester N T - E - N T - E - NL2 Extraction Systern N13 Turbine Bypass System
- 1. Turbine bypsis piping N T D - -
(r) including supports up to the condenser N14 Reactor Feedester Pump Driver N T - E - N15 Turbine Ausf!!ary Steam Systein N T - E ~ N T - E - N16 Generator N T - E - N17 11ydrogen Gas cooling System N18 Generator Cooling System N T - E - N19 GeneratorScaling011 System N T - E - N T - E - N20 Exciter N T - E - N21 Main Condenser N T - E - N22 Offgas System N T D E - N23 CirrulatingWaterSystesa N T - E - N24 Condenser Cleansp Facility 32 21.4 Atrendment
nu m ABWR pry 4 Standard Plant NOTES (Contlaued)
- n. All cast pressure. retaining parts of a size and configuration for which volumetric methods are effective are examined by radiographic methods by qualified persoanel. Ultrasonic examination to equivalent standards is used as an alternate to radiographic methods.
Examination procedures and acceptance standards are at least equivalent to those defined in Paragraph 136.4, Nonboiler External Piping, ANSI B31.1.
- o. The following qualdications 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.
'ne following requirements sha!! be met in addition to the Quality Group D requirements:
- 1. All longitudinal and circumferential butt weld joints shall be radiographed (or ultrasonically tested to equivalent standards). Where size or configuration does not peseilt effective volumetric examination, magnetic particle or liquid penetrate examination may be substituted. Examination procedures and acceptance standards shall be at least equivalent to those specified as supplementary types of craminations, Paragraph 136.4 la ANSI B31.1.
- 2. All fillet and socket welds shall be examined by either magnetic particle or liquid penetrate methods. All structural attachment welds to pressure retaining materials shall be examined by either magnetic particle or liquid penetrate methods. Examination procedures and acceptance standards shall be at least equivalent to those specified as supplementary types of examinations, Paragraph 136.4 in ANSI B31.1
- 3. Allinspection records shall be maintained for the life of the plant. These records shall include data pertaining to qualification of inspection personnel, examination procedures, and examination results.
- p. A quality assurance program meeting the guidance of Regulatory Guide 1.143 will be applied during design and construction.
- q. Detailed seismic design criteria for the offgas system are provided in Subsection 11.3.4A r, See Section 315.3.
nM Artwnomeat
ABM ursious ny 4 Standard Plant NOTES (Continued) J
- s. The recirculation motor cooling system (RMCS) is classified Quality Group B and Safety Class 2 which is consistent with the requirements of 10CFR50.55a. The RMCS, which is part of the reactor coolant l I
pressure boundary (RCPU) meets 10CFR50.55a (c)(2). Postulated failure of the RMCS piping canno cause a loss of reactor coolant in excess of normal makeup (CRD return or RCIC flow), and the RMCS l' 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 mtke 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 3, however, the system is designed and constructed in accordance with ASME Boiler and Pressure Vessel Code, Section I!!, Class I criteria as specified in Subsection 3.9.3.1.4 and Figure 5.4-4 t, A quality assurance program for the Fire Protection System meeting the guidance of Dranch Technic Position CMEB 9.51 (NUREG.08(0), is applied.
- u. Special schmic qualification and quality assurance requirements are applied.
- v. See Reg Guide 1.143, paragraph C.5 for the offgas vaidt Schmic requirements.
- w. The condensate storage tank will be designed, f abricated, and tested to meet the intent of API Standard API 650, in addition, the specification for this tank will requiret (1) 100% surface examination of the side wall to bottom joint and (2) 100% volumetric examination of the side wall weld joints,
- x. The crancs are designed to hold up their loads 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 lit.
Class 3.
- z. Components auociated with safery related function (e.g., isolation) are safety.related.
sa. Structures which support or house safety related mechanical or electrical components are safety.related. - bb All quality assurance requirements shall bc applied to ensure that the design, construction and testing requirements are met. cc. A quality assurance program, which meets or exceeds the guidance of Generic Letter 85 06,is to all non tafety related ATWS equipment. dd. The need for pipe whip restraints on the MSL/FW piping will be determined by a ' leak before break evaluation. ee. Figure 3.2 2 depicts the classification requirements for the Feedwater System. At the interfa between seismic and non seismic Category I feedwater piping system, the seismic Category I dynamic analyses will be extended to either the first anchor point in the non seismic system or tc sufficient distance in the non seismic system so as to not degrade the validity of the seismic Category I analysis. 32441 Anwedment
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SEISMIC CATEGORY I _ _ NON-SEISMIC CATEGORY I _ STRUCTUFE. SYSTEMS. AM COtrONENTS STRUCTURE. SYSTEMS, AND COhPONENTS LEGEND: A. QUALITYGROUP A . G. TUNISIE STOP VALVE' B. QUALITYGROUP B H. TUNIS 0ECONTROLVALVE 4 C.' OUALITY GROUP 8. MON-SEISMIC L TUNieIE BYPASS VALVE - J' CATEGORY LDMXY ANALYZED J. MAIN STEAM LEAD D. QUALITY GROUP D, NON-SEISMIC K. BRANCH LtdE CATEGORYI, DYNAMICALLY ANALYZED L DRAINLDIE ' D1. NON ww. CATEGORY L DYNAMICALLY M. STEAM LNE . ANALY2ED E. ISOLATION VALVE F. SEISMIC NTERFACE RESTRAedT u p%- Figure 3.2-1 QUAUTY GROUP AND N CATEGORY CLASSIFICATION 4 f *g-APPUCABLE TO POWER CONVERSION SYSTEM ,
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ABWR ossrun occument ! 2.10.21 Main Condenser Design Description The main condenser is classified as non safety-related and non seismic Category l
!. It is desigried to condense and deaerate the exhaust steam from the main turbine and provide a acat sink for the Turbine Bypass (TB) System.The main condenser is also a collection point for other steam cycle dralrt .
The main condenser hotwell provides a holdup volume for MSIV fission product leakage. The supports and anchors for the main condenser are designed to withstand a safe shutdown earthquake (SSE). The main condenser tubes are made from corrosion resistant material. The main condenser is located in the Turbine Building. i Since the main condenser operates at a vacuum, leakage is into the shell side of the main condenser. Tube side or circulating water inleakage is detected by measuring the conductivity of sample water extracted beneath the tube bundles. In addition, conductivityis monitored at the discharge of the condensate pumps and alarms provided ir the main control room. , A signal is provided to the Leak Detection and Isolation System (LDS) System on loss of vacuum. Condenser pressure indicators are located above the design basis Dood level. , inspections, Tests, Analyses and Acceptance Criterin Table 2.10.21 provides a definition of the inspections, tests, and/or analyses,- together with associated acceptance criteria which will be undertaken for the main condenser,
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oe a e ABWR outon Documnt 2.15.5 Heating, Ventilating and Air Conditioning Systems Design description, basic configuration figure and table for the inspection Test, Analyses and Acceptance Criteria are provided for the Control Room Habitability Area Ventilating and Air Conditioning (HVAC) System. Design Description Control Room Habitability Area HVAC System The Control Room Habitability Area (CRHA) HVAC System controls the - thermal, radiojogical and pressurel emironments. The system com,ists of two jafety-related Dhisions classified as Seismic Category I which are physically separated and elecuically independent. , The basic configuration for the Control / Room Habitability Area HVAC System . is shown in Figure 2.15.5a. [
+
The HVAC equipment, plenums, ducts and dampers cutside the habitability area have leak tight design features enabling the CRHA HVAC System to maintain at ! cast 3.2 mm water gauge positive differential pressure between the habitsbility area and adjacent Control Iluilding areas.- The temperature is controlled within a range of 10'C to 29'C, and a relative humidity (RH) within a range of10% to 60% l 1 The CRHA HVAC System's Control Room indication and controls allow for . monitoring and control dudn f operational conditions. The Control Room has ' indication for and%controlj of temperature and humidity. A control room pressure signal positions an atitomatic damper in the_ exhaust fan discharge duct ' to maintain a positive pressure in the Control Room. Manual control of each: I motorized damper, fan and emergency filtration unit is accomplished with-remote manual switches and indicating lights in the Control Room. A flow device : , in the emergency filtration unit discharge duct automatically starts the redundant Control Room Habitability Area HVAC System on 'oss of air flow. When the radiation mor,itor in the operating outdoor airintake detects airborne contamination, an isolation signal is generated to close the normal outdoor air - intake damper, open the emergency outdoor air intake damper, stop the normal . a exhaust fans, start the emergency air filtration tmit to decrease the - contamination before air is supplied to the Control Room Habitability Area and p' maintain a positive pressure relative to adjacent spaces of the Control Building.
- The redundant train is connected to another outdoor air intake separated from~
the or greater than or equal to 50 m. Each emergency air filtration unit treats both indoor recirculated air and outdoor air to maintain a positive pressure with , not more than 1300 m8per hour of filtered outdoor air. Leakage shall be less , 2.15.5 1 '1/20/93-h
~ ' ABWR Design occument than that required to meet the personnel dose limits of Section 3.7 Radiation Protection. The Safety-related electrical equipment including instrument and controls :
,,,,, gv located in the Control Building outside the Habitability Area is qualified for a harsh environment.
When the Products of Combustion (POC) monitor in the outdoor air intake
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detects smoke, a signal will initiate the recirculation mode by isolating dampers, stopping the exhaust fans and closing the exhaust dampers. 2.15 5 -2 1/20/93
Table 2.15.5a: Control Room Habitability Area HVAC System , u inspections, Tests, Analyses and Acceptance Criteria 5 Certified Design Commitment inspections Tests Analyses Acceptance Criteria
- 1. A basic configuration for the Control Room 1. Inspection of the system will be conducted. 1. The as-built Control Room Habitability
! Habitability Area HVAC System is as shown Area HVAC System conforms with the in Figure 2.155a. basic configuration shown in Figure 2.15.Sa.
- 2. The Control Room Habitability Area is 2. The Control Room Habitability Area HVAC 2.. The Control Room Habitability Area is maintained at a pos!tive pressure greater System will be tested in all modes of maintained at a positive pressure of at than or equal to 3.2 mm water gauge operation. least 3 2 mm water gauge relative to relative to the atmosphere and other areas atmosphere and other areas of the Control of the Controt Building. Building.
- 3. Control Room Indicators and controls 3. Inspections will be performed on the 3. Indicators and controts exist or can be provided for the Control Room Habitability Control Room indicators and controls for retrieved in the Control Room, as defineo Area HVAC Systems are as defined in the Control Room Habitabi!!ty Area HVAC in Section 2.15.5.
Section 2.15.5. System.
- 4. Class 1E losds for the Control Room 4. Tests will be performed on the Controi 4. The test signal exists only in the safety-y Habitability Area HVAC System are Room Habitability Area HVAC System by reisted electrical power loads for Class 1E powered from Class 1E Divisions, as oroviding a test signalin enty en ciass 1E Division under test in the Control Room described in Section 2.15.5 division at a time. Habitabil!ty Area HVAC System.
2.
- 5. Each mechan! cal Division of the Contret 5. Inspection of the as-built system will be 5. Each mechanical Division of the Control Room Hab3tability Area HVAC System is perfo med. Room Habitability Area HVAC System is physically separated. physically separated from the other mechanical Divisions of the Control Roorn Habitability Area HVAC System.
- 6. On detection of smoke at the outdoor 6. Tests with sirnulated outdoor smoke signal 6. The Control Room Habitability Area intake, the outdoor air intaka dampers shall will be performed with the Control Room SysteEHVAC is in the recircu!stico mode close, recirculation dampers shall open, Habitability Area HVAC System in the when sYoke is present outside the outdoor and the exhaust f ans and exhaust dampers recircutation mode. airintakes of the Control Room Habitability Isolate. Area HVAC System.
- 7. Two outdoor air intakes of the Control 7. Inspect the distance between the outdoor 7. The control Room Habitability Area HVAC Room Habitability Area HAVC System are airintakes of the Control Room Habitability System outdoor air intakes are 50 m apart.
at least 50 m apart. Area System. 6 w
o oL l u .. . lrwions. Tests. Analyses and Acceptance Criteria ' lg. - Certined Design Cr; --0..c.; in Sm : ': -; Tests. Artslyses Acceptance Criterie
' 3. A low flow condition during operation of 8. Test the automatic transfer from the the Emergency Filter Unit in the emergency 8. The standby Control Room Habitability Controf Room Heb6tability Aree HVAC mode will cause the redundant Control Area HVAC Division starts in the Division operating in the emergency mode emergency mode when a low flow Room Heb!tability Area HVAC System to with a simulated low flow signal to the l start in the emergency mode condition exists in the operating Control standby Control Room Habitability Area Room Hebetability Area HVAC Division in i HVAC Division in the emergency mode. the emergency mode.
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Enclosure 4 Generic Concern Summaries
- 1. Welding
- 2. Environmental Qualifications Seismic Qualification
- 3. Verification of MOV capabilities
- 4. Electromagnetic Interference (EMI/RFI)
Instrument Setpoints
- 5. Electrical Independence (Separation)
- 6. Piping 4
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- 1. Summary of Weldinn
?
The staff and GE reached agreement'on the approach for verifying welding. The verification will be performed as a part of an ITAAC basic configuration check in each system. ASME Code Class 1, pressure-boundary welds will be inspected to ensure the physical quality of the welds. The supporting SSAR material (attached in draf t form) will specify.the codes / standards and the acceptance criteria for evaluating the quality of welds for ASME Boiler and Pressure Vessel Code components and supports, non-ASME pressure retaining piping, structural and building steel, electrical cable tray and conduit supports', HVAC supports, and refueling cavity and spent fuel pool liners. . GE will decide later in which sections of the SSAR these welding design criteria and acceptance criteria will appear. l I l l I
o .. ;c ;; ~w MerHCbS Weldino,and 4+M Acceptance Criteria As-sequ4mnents-44+ts# teles Ere ce%idered by the :t:fr to be c::ential ja
. s tre!' %g acid ug-est44444e:. My change will require the C0!.- appWrnt-tv ertst-the-changeHo-the-NRC-tt-eff-for -ravic.; and apprrvai prior to ese:-
of. 5 GCTl04 IIUL AS MPL.tcA 6L&, ASME Code Weldina Fog PKGssunk LwaDMLY Aq CDi& S4ffCA.T STCDcTVLES Welding act vities shall be performed in accordance with the requirements of Section Ill of 3 th ASME Code. The required nondestructive examination and acceptan4c criteria is stated in Tabic 1. Component supports shall be in accordance with the requirements of Subsection NF of Section !!! ffabricateds of the ASME Code n e pt th S 4h: ti:L:1 a:ld ::: ptece crittric chaH-be-the-Nue+eee-fenMructica 1350e Greep (NCIG) 5tendeid NCIG-01, "Gui Wid-Acceptance Criteri; for StructSr:1 "ciding of Nuclear-Power Plants," h ision t
' 44D E4HWEb yeldino of non-ASME oressure retainino Pioina Welding activities involving non-ASME pressure retaining piping shall be accomplished in accordance with written procedures and shall meet the requirements of the ANSI B31.1, Code. The weld acceptance criteria shall be as defined for the applicable nondestructive examination method described in ANSI B31.1 Code Weldina of Structural and Buildine Steel Welding activities shall be accomplished in accordance with written procedures and shall meet the requirements of the American Institute of Steel The visual acceptance Construction (AISC) Manual of Steel Construction.
criteria shall be as defined in NCIG-01, 3 Revision 2 v /N.5ERT Weldina of Electrical Cable Trav and Conduit Succorts, Welding activities shall be accomplished in accordance with the leerk:n "elding Sc:icty fAWSyStructural Welding Code, D1.1 The weld visual acceptance NCIG-01, Revision 2. criteria shall be as defined in%3.sfecTucAL WGLbW6C0ff. bl.1 A40 Weldina of Heatina Ventilatino and Air Conditi",ino SuoDorts Welding activities shall be accomplished in accordance with the fm:rican--
"elding Sc:icty IAWS7 Structural Weldttg Code, 01.1 The weld visual acceptance criteria shall be as defined in NCIG-01, Revision 2.
t W5 A s-rat 4(TuRAL klELbW6 C@E hu MD Weldina of Refuel Cavity and Soent Fuel Pool liners wnw AccuSI5LE Welding activities shall be accomplished n accordance with the 5:rican-WeldingSccietyAAWSyStructuralWeldin Code, 01.1 The welded seams of the liner plates shall be spot radiographedt liquid penetrant and vacuum box The exemined after fabrication to ensure thlt the liner do not leak. acceptance criteria for these examination shall meet the acceptance criteria stated in subsection NE-5200 of Section III of the ASME Code. INSEAT - AMERICAN wELDiac, s&tETt'( AWS) STRUCTURAL WELbWG CO6E b I.! Mb NitCL sAR CoNSTRUlil0d 1,5SKE GRodP [ N'CIC7) SBOARD A C l C, .:i, " VISUAL WELD A CCEPTAAICE CRIT &AIA FOR .57AUCmeAL W5LDING AT NUCLEAR POWER PLANG,
1 s TABLE 1 Welding Activities and Weld Examination Requirements for ASME Code, Section III Welds Class 1 Components (1)(2)(3) Component Weld Type , NDE Reoutrements, Categacy A (longitudinali RT plus NT or PT Vessel Vessel, Pipe, Category B RT plus MT or PT Pumo. Valve (Circumferential) Pipe, Pump, Butt weld RT plus MT or PT fillet and socket welds MT or PT Valve Category C and similar RT plus MT or PT. RT must be Vessels (6) multiple exposure welds Partial penetration and MT or PT on all accessible surf aces - fillet welds Category D a) Butt welds, all RT plus MT or PT Vessels (6) RT plus MT or PT
& Branched b) Corner welded nozzles RT plus NT or PT Connections c) Corner welded branch and piping connection exceed-ing 4" nominal diameter MT or PT d) Corner welds branch and piping 4" and less e) Weld buildup deposits at UT plus a, b, e above if connected to openings nozzle or pipe f) Partial penetration MT or PT progressive and final surface g) Oblique full penetration RT or UT plus MT or PT. In addition, branch and piping UT of weld,-fusion zone, and parent connections metal beneath attachment surface.
Fillet, partial penetration, MT or PT General socket tids Structural attachment welds MT or PT General Special welds 1) Specially designed seals MT or PT
- 2) Weld metal cladding PT PT
- 3) Hard surfacing _
None a) Valves 4" or less PT
- 4) Tube-tube sheet welds VT
- 5) Brated joints 1
. .- e - - . . ~ . . .- - . . . .-. _. . - . . . . . ~ - ; a: aza ..
4 i '* Class 2 Components;_(1)(2)(4) Wald-Type- NDE'Recuirements~ Component - Vessel Category A-(Longitudinal) :
,RT-a) Either of the members exceeds -
3/16_ inch . _ _ _ MT, PT,-or-RT b) Each member 3/16 inch or less RT Pipe, Pump, Longitudinal Valve Vessel Category B-(Circumferential)- : RT1 ' a) Either of the members exceeds-
'3/16-in. .-
MT, PT, or RT
-b) Each member 3/16" or:less Pipe, Pump and Circumferential RT - ,
Valve a) Butt welds .
'MT or PT b) Fillet and partial penetration a vessel (6) and Category C similar joints a) Corner joints,:either of_the RT in other members: exceeds 3/16"lof.
components thickness-
~
MT, PT, forLRT : . b) Each member 3/16" or less . MT=or PT.
?
c). Partial penetration and-fillet welds ' and Category 0 .. . . Vessel similar we(6)lds a) Full penetration joints when RT ' in other either members exceed 3/16" of
. components thickness ;
b) Full: penetration corner MT or PT joints when either member exceeds 3/16" MT-or_PT' c)-Both members 3/16" or lessi NT or:PT-d) Partial: penetration and fillet weld ~ joints RT-Branch Con. a) Nominal size exceed 4" - MT or PT (externa 11and L and Nozzles in b) Nominalisize 4" or smaller accessible internal- surf aces) . pipe, valve, DumD N _..,f
a n s Class 2 Components (Cont'd)(1)(2)(4) Comoonent Weld __ Type NDE Re_quirements' , Cat. A RT Vessels Cat. B RT. designed to NC-3200 Cat. C, Butt weld RT Cat. C, Full penetration corner UT or RT ' Cat. C, Partial penetration corner MT or PT both' sides and fillet wtids Cat. D, Full penetration (6) RT Cat. D, Partial penetration 'HT or PT both sides Fillet, Partial Penetration, socket,. HT or PT and structural attachment welds Special Welds a) specially designed seals NT or PT b) weld metal cladding MT or PT c) hard surfacing PT d)- hard surfacing for valves with Hone inlet connection 4" nominal pipe , size or less PT e) tube-tube sheet welds ~ f) Brazed joints VT Storage Tanks a) side joints RT (Atmospheric) VT b) roof and-roof-to-sidewall vacuum box testing of at least c) bottom joints 3 psi d) bottom to sidewall vacuum box + MT or PT e) Nozzle to tank side HT or PT VT f) Nozzle to roof g) Joints in nozzles RT h) others Similar welds in vessels RT Storage Tanks a) sidewall (0-15 psi) RT b) roof RT if not possible MT or PT c) roof-to-sidewall vacuum box method + MT or PT d) bottom & bottom-to-side e) nozzle tank MT or PT f) joints to nozzles RT same as similar vessel joints g) others
- c. 'o_
Class 3 Components (1)(2)(5) Weld Type NDE Reouirements-Component
-Vessels : Category A (Longitudinal)-
- 1. a) Thickness exceeding the limits- RT of Table NO. 5211.2-1 b) Welds based on joint effi-- .
RT. ciency permitted by NO.3351.1 c) butt welds in nozzles attached RT to vessels in a or b above - Spot RT each 50 ft of weld.'
- 2. Welds not included in 1 above addittoral RT to cover each welders work.
Nonferrous vessels exceeding 3/8 RT 3. inch pipes greater than 2 in, size RT, NT, or PT Pump, Valve, Pipe - according to the product form-
- pumps & valves greater than 2 in, Vessel Category B (Circumferential)-
RT
- 1. a) Thickness exceeds Table ND.5211.2 for Ferrous metals RT b) thickness exceeds 3/8 in for-nonferrous metals RT c) joint efficiency according to NO.3352.1(a)
-RT d) attachments to vessels'and exceeds nominal pipe size 10" or thickness l-1/8 in.
RT 6 in, long sections - the 2, welds not involved in 1 above- intersections of Cat. A welcs Greater than 2" nominal. pipe ~ size RT, PT, or NT pipe, pump and valve Vessel Category C: RT
- 1. a) Thickness exceeds Table ND-5211.2 or NO-5211.3 RT b) Attachments exceed 10 inch NPS or 1 1/8 inch wall thickness Spot RT to cover each
- 2. Welds not_ involved in 1 or-2 welders work above RT PT. or MT Pipe, Pumo. Valves Greater than 2" nominal pipe size
-4
o c-Class 3 Components (Cont'd)(1)(2)(5) Vessel Category 0:
- 1. Full penetration butt' welds RT designed for joint efficiency per NO.3352.1(a)
- 2. In nozzles or communicating RT chambers attached to vessels or heads requiring full RT
- 3. Welds not covered' by I and 2 Spot-RT to cover each-above welders work Pipe, Pump and Greater than 2" nominal pipe size RT, PT, or MT -
Valve . a) weld metal cladding PT Special Welds b) hard surfacing PT (i) hard surfacing for valves none with inlet connection 4" nominal pipe size or less PT c) tube-tube sheet welds d) Brazed joints VT a) sidewall joints Same as Category A or B Storage Tanks vessel joints (Atmospheric) VT b) roof and roof-to-sidewall vacuum box testing of at-c) bottom joints least 3 psi, or PT or MT plus VT during pressure-test Same as' bottom joints d) bottom to sidewall MT or PT e) Nozzle to tank side VT f) Nozzle to roof _ g) Joints in nozzles ex. roof nozzles MT or PT h) others Sim,ilar welds in -vessels Same as Category'A--or B. Storage Tanks a) sidewall vessel joints (0-15 psi) Same as Category'A vessel b) roof joints ~ Same as above if possible. c) roof-to-sidewall or MT or PT OR Vacuum box te at least d) bottom & bottom-to-side 3 psi, or PT MT plus VT during pressu test HT or PT e) nozzle to tank MT-or PT f) joints in nozzles - g) others_ same as similar vessel joints 5 l:
- o. <. a l
Containment Vessel (1)(2)(6) Component Weld Type NDE Reouirements Containment Category A. Butt Welds (Long'l) RT Containment Category B Butt Welds (Circ.) RT Containment Category C, Butt weld RT Containment Category C, Nonbutt Welds UT or MT or PT Cont ainmer.t Category 0, Butt Welds RT
~
Containment Category 0, Nonbutt Welds UT or NT or PT Containment Structural attachment welds RT a) Butt Welds UT or HT or PT b) Nonbutt Welds Special welds Weld Metal Cladding PT Components Supports (1)(2)(7) Weld Type NDE Requirements Component Class 1 Primary Member, Full Penetration Butt Welds RT Supports All other welds MT or PT Secondary Member Welds VT Class 2 and MC Primary Member, Full Penetration Butt Welds RT Supports Partial Penetration or fillet MT or PT welds throat greater than 1" VT All other Welds Secondary Member Welds VT Class 3 Primary Member, Groove or HT or PT Supports throat greater than l' VT All other welds Secondary Member Weld VT Special Welds Transmitting Loads in the UT base metal beneath the weld Requirements, Through Thickness Direction in All Classes Members Greater than 1" i w co gs svPPG AT 57RucTVRGS OXD(8) b.ompoaenf weid Tyf e A/De Reg %enf
- ore Supp ort StrucFures Cahegory A, long;lyd;nal baTT wells '
.Provrd e dTrect sqporf Ca +ey ory 5, cfrcmferediaI battadis or cesfras'nro f % fvul, Category c,Elage to shell weus de, un der noro,,al oper - Cafegory D, so etle +- sb ell w eI &s tfingco,,di+rons.), C alegory E, bea u., eud connectrons
+o ot h r Struc h ,-eG, a Repair we Us vnder % Tech or M T or PT.
LO'To deef, 3 Refair welas avec /g inch or MTor PT plus 1070 deep. RT or V T*
- nfernal Stevetures
- l. . n bq an y oS er StYuc- Sam e as a hove same fvre winin Yk reacter v essel). tdons an d z 46ty, Temparary Attack men +s (RemoveJ 6efue oper All MT or PT etion )
+ E x a m in a+ ion s,ay he by any +ec h n rqu a of cer+arn com Lin -
afron s of tech n igu a s, -Fro m s;w,plo VT To MT or PT plus RT ce vT. pua Ii+y factor m and M;pe %% f ate depen d ent on +La tech a ig ve cs) se lec+ed, in acc orda a c e wi# Table NG -3352-1,
e c.a a NOTES: o 1)- The required confirmation'that facility welding-activities ~are-in compliance with.the certified design commitments will-include' the following_ third party verifications: .
- a. Facility welding specifications and. procedures; meet the- applicable- '
ASME Code requirements
- b. -Facility welding activities are performed in accordance with the:
applicable ASME Code requirements- '
-c. Welding. activities related records are prepared, evaluated and maintained in accordance with the ASME Code requirements
- d. Welding processes used-to weld dissimilar base metal and welding filler metal combinations- are compatible for the intended applications
- e. The facility has established procedures :for: qualifications'of welders.
and welding _ operators in accordance with the appl.1 cable ASME' Code requirements- -
- f. Approved procedures are available and used for preheating and post' '
heating of welds,-and those procedures meet'the applicable requirements of the ASME Code '
- g. Completed welds are examined in accordance with the applicable examination method required by the ASME Code
- 2) Radiographic film will, be reviewed and accepted by the COL applicant!s -
nondestructive _ examination (NDE), Level III examiner prior to ' final acceptance-
- 3) The NDE' requiremints for Class 1 components-will be as stated in subarticle N8-5300 of Section III of the ASME Code
- 4) The NDE requirements for class 2 components will be as stated in subarticle NC-5300 of Section III of the ASME Code
- 5) The NDE requirements for Class 3 components will be as stated in subarticle ND-5300 of:Section III of the ASME. Code
- 6) The NDE requirements for containment vessels will be as stated in subarticle NE-5300 of:Section III of the ASME Code
- 7) The NDE requirements for component:su'pports will be:as stated in
[v F)penetration-corner f subarticle NF-5300 81 For corner joints _UT may be used instead of RT. 4 ofzone, weld joints, if RT is used, the fusion Section For Type 2 full-metal beneath the attachment surface shall be UT examined after welding. and parent III o u LEGEND:
'RT - Radiographic Examination; UT - Ultrasonic Examination; MT - Magnetic- '
Particle Examination; LP - Liquid Penetrant Examination; -VT . Visual Examination w ~ y j; pg reqv ?? m e s h SW W" W,('Y b [ u m ?aar+atw-w nsuwn. MplG Cbdc.
y' l_ -
-c
' Aa A Y 4 .
k 0 L s ' ~8 ['g.._ s a 0 @~ Iis, f;0-0051-1= Welded joint locations typical of categories A. 8, C, and 0 l l l i I 6 i-' Uis. IU0--))52-I Ijyisi$ bwli jsisiC5
o +en
' 2. Summary- of Seismic Oualification and Environmental Oualification The staff and GE reached agreement on the approach to be used for verifying (1) the seismic qualification of mechanical and electrical equipment and-(2):
~
the environmental qualification of electrical equipment, lhe verificatica will be performed as part of an ITAAC basic configuration check in each system. The supporting SSAR material has been compiled from GE report NEDE-24326-1-P and is under review by GE for its proprietary nature.. 1 l l { L l t i
, w
- 3. Summary of MOVs The staff and GE reached an agreement on Tier 2 description for MOVs that covers design and qualification as well as pre-operational testing commitments (see attached). We also reached an agreement on Tier 1 description of.this issue. The agreement was that MOVs will be included in the basic configuration ITAAC for qualification testing. For each system there will also be a MOV ITAAC in the three column format'(see attached). The staff also indicated the need for an ITAAC for check valves. Tier 2 descriptions for check valves must also be written.
Although it was concluded that other valves and pumps would not'need further Tier 1 treatment, Tier 2 (the SSAR) does not contain sufficient information on the design and qualification and pre-operational testing of other types of valves as well as pumps. The staff will be working with GE to reach agreement on the information that will be added to the ABWR SSAR.
, , , .eu -
j 4:- O.
- Table..
r_ ;
'(System Name'Used Here (Motor Operated Valves))
Inspections, Tests,' Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. Motoroperated valves (MOV) designated in . 1. Opening and/or closing tests of installed . 1. Each >*OV opens and/or closes. The -
- - Section as having an active .
valves will be conducted under ' following valves open and/or close in the-
- safety-related function will open and/or dose . . preoperationaldifferentialpressure, fluid following time limits:
under differential pressure and fluid flow flow, and ternperature condtions.- conditions .. 4
' Velve ' Time (sec)" . .
m open' close
. close ^,
A h
. g"f m
-i
- .. E
- g-'
4 4
- g f
g r + v s e-'*.. w -e, .w-r ---'a * ~w --
C 9 ABWR m moos Standard Plant _
"YB 14A.6 METIIOD FOR MEETING torque produced by the actuator and thrust delivered to EQUIPMENT ITAAC the stem for increasing differential pressure and flow 14A.6.1 Motor Operated Valves conditions (referred to as load sensitive behavior). The licensee will compare the design torque and thrust requirements to the control switch trip torque and thrust Design and Qualification subtracting margin for load sensitive behavior, control switch repeatability, and degradadon. De licensee will For each motor-operated valve assembly (MOV) measure the total thrust and torque delivered by the with an acuve safety related funcdon, the design basis MOV under wttb and dynamic conditions (includmg and requtred operating condidons (includmg testmg) will diagnostic equipment inaccuracy and control switch be established. repeatability) to compare to the allowable structural capability limits for the individual parts of the MOV.
De licensee will establish the following design ne licensee will test for proper control room position and qualificetion requirements and will provide accep- indicadon of the MOV. tance criteria for these requirements. By te:iting each size, type, and model the licensee will dete1mine the The parameters and acceptance criteria for torque and thrust (as applicable to the type of MOV) demonstrating that the above functionrd performance requirements to operate the MOV and will ensure the requirements have been met are as follows, adequacy of the torque and thrust that the motor operator can deliver under design (design basis and required (a) As required by the safety function: the valve operating) conditions. Dese design conditions include must fully open; the valve must fully close with fluid flow, differenual pressure (includmg pipe breaki, diagnosue indication of hard seat contact. system pressure, fluid temperature, ambient tempera-ture, minimum voltage, and minimum and maximum (b) The control switch settings must provide ade-stroke tane requirements. The ticensee will ensure that quate margin to achieve design requirements in-the structural capability hmits of the individual parts of ciuding consideration of diagnostic equipment the MOV will not be exceeded under design conditions, inaccuracy, control switch repeatability, load The licensee will demonstrate by testing each size, type, sensitive behavior, and margm for degradation. and model that torque and thrust requirements from in-situ testing conditions can be extrapolated to design (c) ne motor output capability at degraded voltage conditions. The licensee will ensure that the valve must equal or exceed the control switch setting specified for each application is not susceptible to including consideration of diagnostic equipment pressure locking and thermal binding. inaccuracy, control switch repeatability, load sensitive behavior and margin for degradanon, Preoperational Testing (d) ne maximum torque and thrust (as applicable ne licensee will test each MOV in the open and for the type of MOV) achieved by_ the MOV in-close directions under static and maximum achievable cluding diagnostic equipment inaccuracy and condition? using diagnostic equipment that measures control switch repeatability must not exceed the torque and thrust (as tpplicable to the type of MOV), allowable structural capability limits for the in-and motor parameters. The licensee will test the MOV dividual parts of the MOV. under various differential pressure and flow conditions l and perform a sufficient number of tests to reliably (e) The remote position indication testing must extrapolate the torque and thrust requirements to its verify that proper disk position is indicated in the design conditions. The licensee will determine the control room. torque and thrust requirements to close the valve for the position at which there is diagnostic indication of hard (f) Stroke-time measurements taken during valve ! seat contact. The licensee will extrapolate the torque and opening and closing must meet minimum and thrust require.nents (including diagnostic equipment maximum stroke-time reqturements. inaccuracy) from the test to design conditions for such parameters as diffesential pressure, fluid flow, under l voltage and temperature. For the point of control switch trip, the licensee will determine any loss in l Amendmern 14A 1
'l
- 4; Summary of EMI/RFI. Setooints. E0 for-I&C. 1 The structure of ITAAC and DAC affecting development, testing.:and.- .- _ _ .
Installation of digital, safety-related, Einstrumentation 'and control 1(I&C) ~ equipment will be as follows: , 1-. The concept of several_ generic ITAAC or DAC is' eliminated.. -
- 2. I&C development is viewed as an integrated process involving both-hardware and software aspects simultaneously, instead of the: previous-emphasis on a separate software process without-specific attention to ~
hardware performance.. Hardware aspects will specifically~ incorporate EQ - _ (including'seisuic qualification)-anu EMI/RFI issues.
- 3. - GE will provide one Instrumentation and Control DAC describing the1 process of integrated software-and hardware development.
- 4. _
The DAC process will_ describe capturing hardware performance requirements with regard to EQ and EMI/RFI in the hardware'and software. design specification.
- 5. Hardware testing of the integrated hardware / software equipment will be specifically addressed by the DAC in the description of the V&V plan and; overall plan.
- 6. The integrated hardware / software process willlinclude verification of--
accuracy of instrument' loops in the-installed safety-related: systems. When sensors are. processed through digital-I&C-logic, this involves bothL sensor accuracy and analog-to-digital conversion accuracy. ITAAC' statements will be_ developed to address accuracy based on a non- _ , proprietary version of GE's.setpoint methodology that GE willisubmit. . These statements 'will be _part of DAC because normal. operating- parameters that detern.ine' alarm'and control thresholds _may not be established 1 before the-various process systems 1are installed. To ensure proper . identification of Class-IE sensors that- are to be included in accuracy._ determination, individual system or-. building design descriptions : (figures'or text) shall indicate these sensors. U $ e f sr' " -
5.. g
- 5. Summary of Electrical Independence (Separation)
GE and the staff decided to treat the electrical independence (e.g. fluid system components powered from independent Class lE electrical divisions) in the fluid system ITAAC by a " standard" type ITAAC entry. See individual fluid system ITAAC. The overall treatment of electrical independence in:luding separation is to be treated in the Electrical System and I&C ITAAC. See attached " standard lar.guage for independence for electrical and 1&C systems."
Y c s , mn s I Standard Lancusee for-!1 pen hubndependence b'r 0 ' '* f 0 " 'N it c sI % %els, 5 Jcnuary 15, 1993 the iniG unsanY h n65 CDC
- x. Independence is provided between Class 1E Divisions, and between Class 1E Divisions and non Class 1E squipment, in the System.
ITA x.1. Tests will be performed in the System by providing a test signal in only one Class 1E Division at a time. AC x.1.The test signal exists only in the Class 1E division under test in the System. 1TA x.2. Inspection of the as-installed Class 1E Divisions in the System will b3 performed. AC x.2. Physical separation exists between Class 1E Divisions in the System. Physic.a1 separation exists between Class 1E Divisions and non Class 1E cquipment in the System. c:\o W \ltanc/btrplate wp January 15, 1993
o b s.,
- 6. Summary of Pioino Desion Acceptance Criteria (DAC)
The staff and GE reached agreement on the resolution of comments ~from the industry /NUMARC.and the NRC's Greybeard Committee on the' generic piping ITAAC (also referred to as Piping DAC). The piping design description has been substantially expanded to include the certified design commitments. The ITAAC have been reduced in number to consolidate those design commitments that are implicit in the ASME Code requirements and to eliminate some design criteria that were deemed not appropriate for the ITAAC treatment. Proposed additions and changes to the SSAR were provided by GE to support the piping DAC/ITA/C changes. See the attached information. l l
%k %
ABWR oesign Document 3.3 PIPING DESIGN 3.3.1 Description Piping associated with hydraulic and pneumatic systems is categorized as either nuclear safety related (i.e., Seismic Category I) or non-nuclear safety (NNS) related (i.e., Non Seismic Category I). The piping shall be designed for a design life of 60 years. Piping systems that must remain functional during and following a safe shutdown earthquake (SSE) are designated as Seismic Category I and are further classified as ASME Code Class 1,2 or 3. Unless otherwise specified in this description, piping systems means nuclear safety related piping systems. Piping systems and components are designed and constructed in accordance with the ASME Code requirements identified in the individual system Design Descriptions. Piping systems are designed to meet their ASME Code Class and Seismic Category requirements. The ASME Code Class 1,2 and 3 piping systems shall be designed to retain their pressure integrity and functional capability under internal design and operating pressures and design basis loads. Piping stresses due to static and dynamic loads shall be combined and calculated in accordance with the ASME Code and shall be shown to be less than the ASME Code allowables for each senice level. For ASME Code Class 1 piping systems, a fatigue analysis shall be performed in accordance with the ASME Code Class 1 piping requirements. Environmental effects shall be included in the fatigue analysis. The Class 1 piping fatigue analysis shall show that the ASME Code Class I piping fatigue requirements have been met. For ASME Code Class 2 and 3 piping rystems, piping stress ranges due to thermal expansion shall be calculated in accordance with the ASME Code Class 2 and 3 piping requirements. The piping stress analysis shall show that the ASME Code Class 2 and 3 piping thermal expansion stress range requirements have been met. For the ASME Code Class 2 and 3 piping systems and components which l_ are subjected to severe thermal transients, the effects of these transients shall be i included in the design. l The Feedwater lines shall be designed for thermal stratification loads. l 4 Psps.m vpt%.s w.H be deu,gned so minimhs de e @ cts of ecocos rros wnj.* - For those piping systems using ferritic materials as permitted by the design specification, the ferritic materials shall not be susceptible to brittle fracture under the expected senice conditions. For those piping systems using austenitic stainless steel materials as permitted by the design specification, the stainless steel piping material and fabrication process shall be selected to minimize the possibility of cracking during senice. 3.3 1- In1m l l
1 l} we ABWR assion Document Chemical, fabrication, handling, welding, and examination requirements that rninimize cracking shall be met. Piping system supports shall be designed to meet the requirements of ASME Code Subsection NF, For piping systems, the pipe applied loads on attached equipment shall be calculated and shown to be less than the equipment allowable loads. Analytical methods and load combinations used for analysis of piping systems shall be referenced or specified in the ASME Code certified stress report. Piping systems and their supports shall be mathematically modeled to provide results for piping system frequencies up to the analysis cut off frequency. Computer programs used for piping system dynamic analysis shall be benchmarked. Systems, structures and componcats that are required to be functional during and following an SSE, shall be protected against the dynamic effects associated with postulated high energy pipe breaks. The pipe break analyses report shall specify the criteria used to postulate breaks and the analytical methods used to perform the pipe break analysis. For postulated pipe breaks, the pipe break analysis report shall confirm; (1) piping stresses in the containment penetration - area are within their allowable stress limits, (2) pipe whip restraints andjet shield designs are capable of mitigating pipe break loads, and (3) loads on safety-related systems, structures and components are within their design loads limits. Piping systems that are qualified for leak-before-break design may exclude design features to mitigate the dynamic effects from postulated high energy pipe breaks. Piping systems shall be designed to provide clearance from structures, systems, and components where necessary for the accomplishment of the structure, system, or component's safety function as specified in the respective structure or system Design Description. The as-built piping shall be reconciled with the piping design required by this section (3.3.1). Inspections, Tests, Analyses and Acceptance Criteria Table 3.3 provides a definition of the inspections, tests, analyses, and associated acceptance criteria, which will be performed for ABWR nuclear safety related and NNS related piping systems as specified in each system's Design Description. Table 3.3 may be completed on an individual system basis. 1/2V93 3.3
( y P Table 3.3 GENERIC PIPING DESIGN Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment Inspections, Teets, Analyses Acceptance Criteria
- 1. The pip'ng system is designed to inspections of ASME Code required An ASME Code certiEed stress meet its ASME Code Class and documents will be conducted. report exists.
Seismic Category requirements. . The ASME Code Class 1,2, and 3 piping system shall be designed to retain its pressure integrity and functional capability under internal design and operating pressures and design basis loads. Piping and piping components shall be designed to show compliance with the requirements of ASME Code, Section 111.
- 2. Systems, structures, and Inspections of the pipe break A pipe break anahsis report or components, that are required to be analysis report, or icak-before-break leak-before-break report exists.
functional during and following an report, will be conducted. This report includes SSE, shall be protected against the documentation of the results of dynamic effects associated with An inspection of the as-built high inspections of high energy pipe postulated high energy pipe breaks. energy pipe break mitigation break mitigation features. Piping systems that are qualiGed for features will be performed. leak-before-break design may exclude design features to mitigate the dynamic effects from postulated high energy pipe breaks. Revision 4.2 1/21/93 r (Final Agreernents from NRC 8e NUMARK Meeting on 1/20/93)
Y 1 (/ An as-built stress report exists. P 3- The as-built piping shall be An inspection of the as-built piping reconciled with the piping design and supports will be performed. For ASME Code Ctass piping, required in section 3.3.1. the as-built stress report
- reconciliation analysis using the includes the ASME Code as-designed and as-built certified stress report and information will be performed. documentation of the results of the as-built reconciliation analysis.
1/21/93 E Revision 4.2 (Final Agreements from NRC Sc NUhtARK Efeedng on 1/20/93)
. - =... '[ Standard Plant mA61cnAE Rev n The results of the data analyses, vibration COL li-information requirements.
amplitudes, natural frequencies, and mode shapes i are then compared to those obtained from the Thermal stratification of fluids in a piping theoretical analysis. system is one of the specific operating conditions that is included in the loads and Such comparisons provide the analysts with load combinations that are contained in the added insight into the dynatnic behavior of the piping design specifications and design reactor internals. The additional knowledge reports. It is known stratification can occur gained from previous vibration tests has been in the feedwater piping during plant startup and utilized in the generation of the dynamic models when the plant is in hot standby conditions for seismic and loss of coolant accident (LOCA) following scram (see Subsection 3.9.2.1.3). If, analyses for this plant. The models used for during design or startup, evidence of thermal this plant are similar to those used for the stratification is detected in any other piping vibration analysis of earlier prototype BWR system, then stratification will be evaluated. plants. If it cannot be shown that the stresses in the pipe are low and that movement due to bowing is 3.9.3 ASME Code Class 1,2, and 3 acceptable, then stratification will be treated Components, Component Supports, and as a design load. In general, if temperature Core Support Structures differences between the top and bottom of the pipe are less than 50 F, it may be assumed 3.9.3.1 Loading Combinations, Design design specification and stress reports need not Tmnsients,and Stress Limits be revised to include stratification. This section delineates the criteria for The design life for the ABWR Standard Plant selection and definition of design limits and is 60 years. A 60 year design life is a loading combination associated with normal requirement for all major plant components with operation, postulated accidents, and specified reasonable expectation of meeting this design seismic and other reactor building vibration life. However, all plant operational components I (RBV) events for the design of safety.related and equipment except the reactor vessel are l ASME Code components (except containnaent designed to be replaceable, design life not components which are discussed in Section 3.8). withstanding. The design life requirement allows for refurbishment and repair, as This section discusses the ASME Class 1,2, appropriate, to assure the design life of the and 3 equipment and associated pressure termining overall plant is achieved. In effect, parts and identifies the applicable loadings, essentially all piping systems, components and calculation methods, calculated stresses, and equipment are designed for a 60 year design allowable stresses. A discussion of major life. Many of these components are classified equipment is included on a component-by,.-psm as ASMEp ? 73.or_ Quality Group D. In the basis to provide examples. Design transients and evenbady non-Clast I components are subisted dynamic loading for ASME Class 1,2, and 3 Je cyclic loadings, including operai'iiig ' l equipment are covered in Subsection 3.9.1.1. fvibration loads and thermal transient effects,
- Seismic related loads and dynamic analyses are of a magnitude and/or duration,so severe that l discussed in Section 3.7. The suppression the 60 ysar design life ca(Ve assured by pool related RBV loads are described in Appendix nited Code calenlations, COL aplicanteritt" 3B. Table 3.9 2 presents the combinations of idenufy these components ano either provide an dynamic events to be considered for the design appropriate analysis to demonstrate the required I
and analysis of all ABWR ASME Code Class 1,2, design life or provide designs to mitigate the and 3 components, component supports, core magnitude or duration of the cyclic loads, support structures and equipment. Specific Cocifonents excluded from this requirement are loading combinations considered for evaluation of (1) tecs where mixing of hot and cold fluids each specific equipment are derived from Table occurs and thermal sleeves have been provided in . 3.9-2 and are contained in the design accordance with the P&lDs, (2) components, such specifications and/or design reports of the as the quencher, for which a fatigue analysis respective equipment. See Subsection 3.9.7.4 for has already been performed, providing the com. Amucmasa Severe thermal transients that will be evaluated for possible #8 effect on plant life are temperature rate changes faster than 1500*F/ Hour when the total fluid temperature change is
q - y 23A6100AE
\
PP/ B Standard Plant ,, Subsection 3.9.3.1.) l 3.9.7 COL License Information 3.9.7.1 Reactor Internals Vibration Analysis, 3.9.7.3 Pump and Valve Inservice Testing Measurtment and inspection Program Program The first COL applicant will provide, at COL applicants will provide a plan for the the time of application, the results of the detailed pump and valve inservice testing and vibration assessment program for the ABWR inspection program. This plan will prototype internals. These results will include the following information specified in Regulatory (1) Include baseline pre service testing to Guide 1.20. support the periodic in service testing of the components required by technical R. U 1.20 SuNect specifications. Provisions are included to disassemble and inspect the pump, check C.2.1 Vibration Analysis valves, and MOVs within the Code and Program safety related classification as necessary, C.2.2 Vibration Measurement depending on test resuits. (See Program Subsections 3.9.6, 3.9.6.1, 3.9.6.2.1 and C.2.3 Inspection Program 3.9.6.2.2) C.2.4 Documentation of Results O Provide a study to determine the optimal frequency for valve stroking during NRC review and approval of the above inservice testing. (See Subsection information on the first COL applicant's docket 3.9.6.2.2) will complete the vibration assessment program requirements for prototype reactor internals. (3) Address the concerns and issues identified in Generic Letter 8910; specifically the In addition to the information tabulated method of assessment of the loads, the above, the first COL applicant will provide the method of sizing the actuators, and the information on the schedules in accordance with setting of the torque and limit switches. the applicable portions of position C.3 of (See Subsection 3.9.6.2.2) Regulatory Guide 1.20 for non prototype 3.9.7.4 Audit of Design Specification and interna 1s. Design Reports Subsequent COL applicants need only provide the information on the schedules in accordance COL applicants will make available to the with the applicable portions of position C.3 of NRC staff design specification and design Regulatory Guide 1.20 for non prototype reports required by ASME Code for vessels, I internals. (See Subsection 3.9.2.4), pumps, valves and piping systems for the purpose of audit. (SpSubsectionu 3.9.3.1) 3.9.7.2 ASME Class 2 or 3 or Quality Group D -d Components with 60 Year Design Ufe 3.9.8 References 1g -- J h./)-
- 1. BWR Fuel Channel Mechanical Design and COL applicants will identify ASME Class 2 or 3 or Quality Group D components that are Deflection, NEDE-21354-P, September 1976.
subjected to cyclic loadings, including operating
- 2. BWR/6 Fuel Assembly Evaluadon of Combined vibtation loads and thermal transients offects.
of a magnitude and/or duration so severe the 60 Safe Shutdown Earthquake (SSE) and year design life can not be assured by required Loss-of-Coolant Accident (LOCA) Loadingr. Code calculations and, if similar designs have NEDE 21175-P, November 1976. not already been evaluated. cither provide an appropriate analysis to demonstraic the required 3. NEDE-24057 P (Class Ill) and NEDE 24057 design life or provide designs to mitigate the (Class I) Assessment of Reactor Internak magnitude or duration of the cyclic loads. (See Vibration in BWR/4 and BWR/5 Plants. 3w Amendment 23
eun - 3.9.7.5 ABME Class 1,2 and 3 Piping System Clearance Requirements ASME Class 1,2 and 3 piping systems shall be designed to provide clearance from structures, systems, and components where necessary for the accomplishment of the structure, system, or component's safety function as specified in the respective structure or system design description. The COL licensee shall verify that the maximum calculated piping system deflections under service conditions do not exceed the minimum clearances between the piping system and nearby structures, systems, or components. The COL licensee shall document in the certified design stress report that the clearance requirements have been met. 3.9.7.6 As-Built Reconciliation Analysis For ABME Class 1,2 and 3 Piping Systems For ASME Class 1,2 and 3 piping systems, the COL licensee shall reconcile the as-built piping system with the as-designed piping system. The COL licensee will perform an as-built inspection of the pipe routing, location and orientation, the location, size, clearances and orientation of piping supports, and the location and weight of pipe mounted equipment. This inspection will be performed by reviewing the as-built drawings containing verification stamps, and by performing a visual inspection of the installed piping system. The piping configuration and component location, size, and orientation shall be within the tolerances specified in the certified as-built piping Stress Report. The tolerances to be used for reconciliation of the as-built piping system with the as-designed piping system are provided in the EPRI report, " Guidelines for Piping System Reconciliation (NCIG-05, Revision 1)," NP-5639 dated May 1988. A reconciliation analysis using the as-built and as-designed information shall be performed. The certified as-built Stress Report shall document the results of the as-built reconciliation analysis. M
ABWR n^sim^n REV D Standard Plant ,
#N (1) A summary of the dynamic analyses applicable to high cnergy piping
" g'g 7 fgg systems in accordance with Subsection 3.6.2.5 of Regulatory Guide 1.70. This shall include:
(a) Sketches of applicable piping systems showing the location, size and orientation of postulated pipe - breaks and the location of pipe whip restraints and jet impingement barriers. (b) A summary of the data developed to select postulated break locations including calculated stress intensities, cur.clative usage factors and . tress ranges as delineated in BTP MEB 31. (2) For failure in the moderate.cnergy piping systems listed in Table 3.6 6, l descriptions showing how safety related systems are protected from the resulting jets, flooding and other adverse environmental effects. (3) Identification of protective measures provided against the effects of postulated pipe failures for protection f of each of the systems listed in Tables 3.61 and 3.6-2. (4) The details of how the MSIV functional capability is protected against the effects of postulated pipe failures. (5) Typical examples, if any, where protection for safety related systems and components against the dynamic effects of pipe failures include their enclosure in suitably designed structures or compartments (including g any additional drainage system or equipment environmental qualification 3.6./ COL LicenseInfortnation needs). 3.6. Details of Pipe Break Analysis Results (6) The details of how the feedwater line l and Protection Methods check and feedwater isolation valves i functional capabilities are protected The following shall be provided by the COL against the effects of postulated pipe l applicant (See Subsection 3.6.2.5): jailures U) SEf 4th b Amendment D _. _ W7
s n acse 3.6.4 As-Built Inspection of High Energy Pipe Break Mitigation Features An as-built inspection of the high energy pipe break mitigation features shall be performed. The as-built inspection shall confirm that systems, structures and components, that are required to be functional during and following an SSE, are protected against the dynamic effects associated with high energy pipe breaks. An as-built inspection of pipe whip restraints, jet shields, structural barriers and physical separation distances shall be performed. For pipe whip restraints and jet shields, the location, orientation, size and clearances to allow for thermal expansion shall be inspected. The locations of structures, identified as a pipe break mitigation feature, shall be inspected. Where physical separation is considered to be a pipe break mitigation feature, the assumed separation distance shall be confirmed during the inspection. 3.6.5 COL License Information 3.6.5.1 Details of Pipe Break Analysis Results and Protection Methods (7) An inspection of the as-built high energy pipe break mitigation features shall be performed. The pipe break analysis report or leak-before-break report shall document the results of the as-built inspection of the high energy pipe break mitigation features. (See subsection 3.6.4, for a summarv of the as-built inspection requirements.)
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