ML20044B838

Forwards SSAR Markups for Listed Sections Covering Evaluation of Control Bldg Flooding Events,Evaluation of Turbine Bldg Flooding Events & Protection of Uhs,To Support Accelerated Advanced BWR Review Schedule

ML20044B838
Person / Time
Site:	05200001
Issue date:	03/03/1993
From:	Fox J GENERAL ELECTRIC CO.
To:	Poslusny C Office of Nuclear Reactor Regulation
References
NUDOCS 9303100202
Download: ML20044B838 (17)

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GE Nuclear Energy,

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I March 3,1993 Docket No. STN 52-001 Chet Poslusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation

Subject:

Submittal Supporting Accelerated ABWR Review Schedule - ABWR SSAR Chapters 3,6 and 10

Dear Chet:

Enclosed are SSAR markups providing- (1) corrections / clarifications to Subsections 3.4.1.1.2.2,3.6.4.2 and 10.4.5.6; (2) responses to Open Items 3.4.1-1 and 3.5.1.2-1; (3) addressing COL Action items 3.4.2-1,3.5.1.4-1 and 3.5.1.4-2; (4) resubmittal of responses to Open Item 3.10-1 initially responded to February 26,1993 and (5) resubmittal of response to COL Action Item 63.4.2-1 initially responded to February 24,1993.

In addition, it has been determined that COL Action Items 3.5.1.3-1 and 1.8.1-1 are provided by existing Subsections 3.5.4.6 and 3.8.63, respectively.

Please provide a copy of this transmittal to Butch Burton and Dave Terao.

Sincerel,

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Y Fox Advanced Reactor Programs cc: Gary Ehlert (GE)

Norman Fletcher (DOE) l l

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co s cTso u mm Standard Plant Rev s (1) Where extensive flooding may occur in a blowdown will cause most of the steam to vent division rated compartment, propagation to out of the tunnel into the turbine building.

g other divisions is prevented by watertight Water or steam cannot enter the control doors or sealed hatches. Flooding in one building. See Section 3.6.1.3.2.3 for a division is limited to that division and description of the subcompartment pressurization flood water cannot propagate to other analysis performed for the steam tunnel.

divisions.

Moderate energy water services in the control (2) Leakage of water from large circulating building comprise 28-inch service water lines, water lines, such as reactor building 18-inch cooling water lines, 6-inch cooling cooling water lines may flood rooms and water lines to the chiller condenser,6-inch corridors, but through sump alarms and fire protection lines, and 6-inch chilled water leakage detection systems the control room heater lines. Smaller lines supply drinking is alerted and can control flooding by water, sanitary water and makeup for the chilled system isolation. Divisional areas are water system. Areas with water pipe routed protected by watertight doors, or where only through are supplied with floor drains and curbs limited water depth can occur, by raised to route leakage to the basement floor so that sills with pedestal mounted equipment within control or computer equipment is not subjected the protected rooms.

to water. In those areas where water infusion cannot be tolerated, the access sills are (3) Limited flooding that may occur from manual raised.

firefighting or from lines and tanks having limited inventory is restrained from Maximum Gooding may occur from leakage in a entering division areas by raised sills and 28-inch service water line at a maximum rate of elevation differences.

12.0 cubic meters / minute (3150 gpm). Early detection by alarm to control room personnel Therefore, within the reactor building, will limit the extent of flooding which will internal flooding events as postulated will not also be mitigated by drainage to exterior of the

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prevent the safe shutdown of the reactor, building. The expected release of a service water leak is limited to line volume plus 3.4.1.1.2.2 Evaluation of Control Building operator response time times leakage rate. The Flooding Events assumed operator response time is 30 minutes to close isolation valves and turn off the pump in The control building is a seven story the affected service water division. Water will building. It houses in separate areas, the be contained inside a division of closed cooling control roon proper, control and instrument water equipment rooms in the bottom level of the~ g. 3 cabinets with power supplies, closed cooling control building. A maximum of 2-M meters of water pumps and heat exchangers, mechanical water in a divisional room is expected. Water equipment (HVAC and chillers) necessary for tight doors will confine the water to a building occupation and environmental control for division.

computer and control equipment, and the steam tunnel.

The failure of a cooling water line in the mechanical rooms of the turbine building may l

The only high energy lines in the control result in a leak of 0.6 cubic meter / minute (160 building are the mainsteam lines and feedwater gpm). Early detection by control room personnel lines which pass through the steam tunnel will limit the extent of flooding. Total connecting the reactor building to the turbine release from the chilled water system will be building. There are no openings into the control limited to line inventory and surge tank volume, building from the steam tunnel. The tunnel is spillage of more than 6 cubic meters (1500 scaled at the reactor building end and open at gallons) is unlikely. Elevation differences and the turbine building end. It consists of separation of the mechanical functions from the reinforced concrete with 2 meter thick walls. remainder of the control building prevent Any break in a mainsteam or a feedwater line will propagation of the water to the control area.

flood the steam tunnel with steam. The rate of Amendment 20 34

I MM oi %41-1 mm.

Standard Plant l

nrv s ruptsires is unlikely and can be contained from and load combinations indicated in Subsection-spreading to the structures that house safety-3.8.4.3 and 3.8.5.3 using well established g

related equipment.

methods based on the general principles of engineering mechanics. All Seismic Category 1 3A.1.1.2.5 Evaluation of Turbine Building structures are in stable condition due to either Dooding Events moment.or uplift forces which result from the proper load combirations including the design Circulating water system and turbine building basis flood.

service water system are the only systems large enough to fill the condenser pit; therefore, only 3.4.3 COLLicenseInformation i

these two systems can flood into adjacent buildings.

3A.3.1 Rood Elevation A failure in either of these systems will The design basis flood elevation for the ABWR result in the total flooding of the turbine Standard Plant structures is one foot below ilding up to grade. Water is prevented from grade.

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cros to other buildings by two means. The t

first is a agally closed alarmed door in the 3A.3.2 Ground Water Elevation connecting passage \\between the turbine buildi g and service building. The second is that thc4 The design basis ground water elevation for i

radwaste tunncl will be scaled at4 e ends to the ABWR Standard Plant structures is two feet prevent water from either entering the tunnel or below grade.

! caving the tunnel. A large hydrostatic head is prevented by a large non. water-tight truck door 3A33 Mood Protection Requirements for Other l

l at grade to provide a release point for any Structures

-l water.

f The applicant referencing the ABWR design Because of the large size of the circulating will demonstrate, for the structures outside the water system, a leak will fill the condenser pit scope of the ABWR Standard Plant, that they meet quickly. Monitors were added in the condenser the requirements of GDC 2 and the guidance of RG pit of the turbine building to provide leak 1.102. (See Subsection 3.4.1.1.2) detection and an automatic means to shutdown the l

circulating water system in the event of Ilooding 3.4.4 References l

in the turbine building (see Suosection 1.

Crane Co., Flow of Fluids Through Valves, l

M 3 5 'O 4.6.2.3 arnd lo.9.s,Q '

Fittings, and Pipe, Technical Papet No.

l 3A.1.2 Permanent Dewatering System 410, 1973.

There is no permanent dewatering system 2.

ANSI /ANS 56.11, Standard, Design Criteria j

provided for in the flood design.

for Protection Against the Effects of 1

Compartment Flooding in Light Water Reactor 1

3.4.2 Analytical and Test Procedures Plants.

Since the design flood elevation is one foot 3.

Regulatory Guide 1.59, Rev. 2 Design Basis below the finished plant grade, there is no Floods for Nuclear Power Plants.

j dynamic force due to flood. The lateral l

hydrostatic pressure on the structures due to the i

design flood water level, as well as ground water l

and soil pressures, are calculated.

Structures, systems, and components in the ABWR Standard Nuclear Island designed and analyzed for the maximum b drostatic and 3

i hydrodynamic forces in accordance with the loads l

l 14-7 l

Amendment 24

- i

gg s.R.s-s Standard Plant Rev. n Table 3.4-1 STRUCTURES, PENETRATIONS, AND ACCESS OPENINGS DESIGNED FOR FLOOD PROTECTION Reactor Service Control Radwaste Turbine Structure Building Buildine Buildine Buildine Buildine Design Flood Level (mm) 11,700 11,700 11,700 11,700 11,700 Reference Plsnt Grade (mm) 12,000 12,000 12,000 12,000 12,000 Base Slab (mm)

-8,200 2150 &

-8,200

-1,500 5,300 3500 Actual Plant Grade (mm) 12,000 12,000 12,000 12,000 12,000 i

2\\000 Building Height (mm) 49,700 46;WO-49;700 22 2co 22.,2oo g

3 Penetrations Below Design Refe to None Refer to None None Flood level Table 6.2-9 Table 6.2-9 for RCWlines Au:m was Area Access Arm Auus Access Openings Below T.nad from s/B Main Entnace sumed from S/B Pipe Tunnel Twmd fromWg#

Design Floodlevel

@37m?mm Tv.s1. @ grade kvet

@SdI5f,mmJIX from R/B&T/B @ 7.900mm Area Access

@huomm.

T= n nd { n.

Note 3 \\

SSOh from S/B @

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Notes:

1.

Water tight doors (bulkhead type) are provided at all reactor and control building access ways that are below grade.

2.

Water tight penetrations will be provided for all reactor and control building penetrations that b

are below grade.

3.

The lines that run through the radwaste building tunnel are not exposed to outside ground

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flooding.

4.

Penetrations below design flood level will be scaled against any hydrostatic head resulting from a moderate energy pipe failure in the tunnel or connecting building.

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• generated from other natural phenomena. The 3.5.1.6 Airtraft Hazards design basis tornado for the ABWR Standard Plant Aircraft hazards are not a design is the maximum tornado windspeed corresponding to for the Nuclear Island (i.e.110"pasis event a probability of 10E 7 per year (300 mph). The per year).

other characteristics of this 16rnado, summerized See Subsection 3.5.4.3 for COL license in Subsectica 3.3.2.1. The design basis tornado information requirements, missiles are per SRP 3.5.1.4, Spectrum 1.

3.5.2 Structures, Systems,and Components to be Protected from Externally Generated Missiles The sources of external missiles which could affect the safety of the plant are identified in Subsection 3.5.1. Certain items in the plant are required to safely shut down the reactor and maintain it in a safe condition assuming an Using the design basis tornado and missile additional single failure. These items, whether spectrum as defined above with the design of the they be structures, systems, or components, must Seismic Category I buildings, compliance with all therefore all be protected from externally of the positions of Regulatory Guide 1.117, generated missiles.

• Tornado Design Classification," Positions C.1 and C.2 is assured.

These items are the safety-related items listed in Table 3.2-1.

Appropriate safety l

The SGTS charcoal absorber beds are housed in classes and equipment locations are given in this the tornado resistant reactor building and tab!c. All of the safety-related systems listed therefore are protected from the design basis are located in buildings which are designed as tornado missiles. The offgas system charcoal tornado resistant. Since the tornado missiles absorber beds are located deep within the turbine are the design basis missiles, the systems, building and it is considered very unlikely that structures, and components listed are considered these beds could be ruptured as a result of a to be adequately protected. Provisions are made design basis tornado missile. These features to protect the charcoal delay tanks against assure compliance with Position CJ of Regulatory tornado missiles.

Guide 1.117.

See Subsection 3.5.4.1 for COL license An evaluation of all non safety-related information requirements,

{ structures, systems, and components (not housed j

I in a tornado structure) whose failure due to a 3.5.3 Barrier Design Procedures design basis tornado missile that could adversely impact the safety function of safety-related The procedures by which structures and systems and components will be provided to thej barriers are designed to resist the missiles NRC by the applicant referencing the ABWR/ described in Subsection 3.5.1 are presented in Qnigg/See WH:: LLd3 for COL license this section. The following procedures are in information requirements.

accordance with Section 3.5.3 of NUREG.0800 S a re.Ms 3 5 A.2. %d 3. s' A. 57 (Standard Review Plan).

3.5.1.5 Site Proximity Missiles Except Alteraft 3.53.1 14 cal Damage Prediction l

External missiles other than those generated The prediction of local damage in the impact by tornadop are not considered as a design basis area depends on the basic material of construc-(i.e.1 10 per year).

tion of the structure or barrier (i.e. concrete or steel). The corresponding procedures are presented separately. Composite barriers are not utilized in the ABWR Standard Plant for missile protection.

3.5.

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ABM naoore Standard Plant arv a viilent static load concentrated at the impact impact the safety function of a safety-related area is determined. The structural response to systems and cgponents will be provided to the this load, in conjunction with other appropriate NRC by theAapplicant.::f;..a: g d:

a. " '// R design loads, is evaluated using an analysis du y (See Subsection 3.5.1A).

procedure similar to that in Reference 6 for rigid missiles, and the procedure in Reference 7 3.5A.6 Turbine System Maintenance Program for deformable missiles, col. u cen wa L S.-d &

A turbine system maintenance program g

3.5.4 L=t=

including probability calculations of turbine missile generation meeting the minimum 3.5A.1 Protection of Ultimate Heat Sink requirement for the probability of missile generation shall be provided to the NRC (See Compliance with Regulatory Guide 1.27 as Subsection 3.5.1.1.3).

related to the ultimate heat sink and connecting conduits being capable of withstanding the 3 5 5 References e t 3.s. L 2 - t effects of externally generated missiles shall be demonstrated (See Subsection 3.5.2).

1. fC. V. Moore, The Design of Barricades for o%r Hazardous Pressure Systems, Nuclear 3.5A.2 Missiles Generated byhNatural Phenomena 7kngineering and Design, Vol. 5,1967

f. eu. " a.;. A. e! !";.:.: S; ";. u, Spk ms esa 2.

F. J. Moody, Prediction of Blowdown Thrust f

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tot

3. s. t. 4 - t _cmd CO t

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1. 4

't.

and Ict Forces, ASME Publication 69-HT 31,

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The remainder of plant structures, system August 1%9.

and components shall be analytically checked to ensure that during a site-specific tornado they 3.

A. Amirikan, Design of Protective Strue.

will not generate missiles exceeding the missiles tures, Bureau of Yards and Docks, Publica-g ggkconsidered under Subsection 3.5.1.4.

j tion No. NAVDOCKS P-51, Department of the

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Navy, WaMncton, D.C., August 1960.

A f 3.5A.3 Site P oximity Missiles and Aircraft Hazards.

4.

A. E. Stephenson, Full Scale Tornado-Mis-site Impact Tests, EPRI NP-440, July 1977, Analyses shall be provided that demoutrate prepared for Electric Power Research that the probability of site proximity missiles Institute by Sandia Laboratories.

(including aircraft) impacting the ABWR Standard Plant and causing consequences greater thyn 10CFR 5.

W. B. Cottrell and A. W. Savolainen, U. S.

Part 100 exposure guidelines is 5,10 per year Reactor Containment Technology, ORNL-NSIC 5, Vol.1, chapter 6, Oak Ridge Na.

(See Subsection 3.5.1.6).

tional Laboratory.

3.5A.4 Secondary Missiles inside Containment 6.

R. A. Williamson and R. R. Alvy, Impact Protection against the secondary missiles Effect of Fragments Striking Structural inside containment described in Subsection Elements, Holmes and Narver, Inc., Revised 3.5.1.2.3 shall be demonstrated.

November 1973.

oud cFABwn.

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7.

J. D. Riera, On the Stress Analysis of 3.5A.5 Impnct of Yallure a,Non Safety Related Structures Subjected to Aircraft Impact Structures, Systems, and Components Due to n Forces, Nuclear Engineering and Design, Design Basis Tornado ed of ABwA North Holland Publishing Co., Vol. 8,1968.

jsu rd Plad Sar*

An evaluation of allfos,s safety-related 8.

Deleted structures, systems, and components (not housed in a tornado structure) whose failure due to a design basis tornado missile that could advers K. K a e. - - Pa waM e-ab "C'] "

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344.2 lank-Before-Break Analysis Report i

As required by Reference 1, a BB analysis

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report shall be prepared for the piping systems proposed for exclusion from analysis for the dynamic effects due to failure of piping f ailur e. The report shall be prepared in acegadance with the guidelines presented is>.

Appendix 3E and Submitted by the COL applicant to

_ l the NRC for approval. (See Subsection 3.63).

3.6.5 References

1. Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Rupture, Federni Resister. Volume 52, No. 207, Rul s and Regulations, Pages 41288 to 41295, October 27, 1987

2. RELAP 3, A Computer Program for Reactor Blowdown Analysis, IN-1321, issued June 1970, Reactor Technolory TID-4500.

3. ANSI (ANS S8.2, Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture.

4. Standard Review Plan; Public Comments Solicited, Federal Renister. Volume 52, No.

167, Notices, Pages 32626 to 32633, August 28, 1987.

5. NUREG-1061 Volume 3, Evaluation of Forendal for Pipe Breaks, Report of the U.S. NRC Piping Review Committee, November 1984.

6. Mehta, H. S., Patel, N.T. and Ranganath, S.,

Application of the Leak-Before-Break Approach to BWR Piping, Report NP-4991. Electric Power

~ Research Institute, Palo Alto, CA, December

'1986.

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33.7 COL LicenseInformation Subsection 3.93.1.)

l 3.9.7.1 Reactor laternals Vibration Analysis, 3.9.73 Pump and Valve Inservice Testing Measurement and laspection Program Program The first COL applicant wjil provide, at COL applicants will provide a plan for the the time of application, the results of the detailed pump snd valve inservice testing and vibration assessment program for the ABWR inspection progrsm. 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. G.1.20 Subiect 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 Subseetioas 3.9.6, 3.9.6.1, 3.9.6.2.1 a nd C.23 Inspection Program 3.9.6.2.2)

C.2.4 Documentation of Results (2) 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; spec.ifically 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 i

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) g.

Regulatory Guide 1.20 for non prototype

/gsp p t 3.4 7 3 a z f s-/

internals.

3.9.7.4 Audit of Design Specifkation and 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, l internals. (See Subsection 3.9.2.4).

pumps, valves and piping systems for the purpose of audit. (See Subsection 3.9.3.1) 3.9.7.2 ASME Class 2 or 3 or QuaDty Group D Components with 60 Year Design ute 33.8 References COL applicants will identify ASME Class 2 1.

BWR Fuel Channel Mechanical Design and or 3 or Osality Group D components that are Deflection. NEDE 21354-P, September 1976.

subjected to cyclic loadings, including operating vibration l'oads and thermal transients effects.

2.

BER/6 Fuet Assembly EWunnan of Combined of a magnitude and/or duration so severe the 60 Safe Shutdown Earthquake (SSEJ and year design life can not be assured by required Loss of-Coolant Accident (LOCA) Loadings.

Code calculations and, if similar designs have NEDE 21175-P, November 1976.

not already been evaluated, either provide an i

appropriate analysis to demonstrate the required 3.

NEDE.24057-P (Class III) and NEDE-24057 design life or provide designs to mitigate the (Cliss I) Assessment of Reactor Internals.

magnitude or duration of the cyclic loads. (See Vibration in BWR/4 and BWR/5 Plants.

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{F,~1 no]-i Standard Plant n e ocAE REV B 3.10 SEISMICQUALIFICATIONOF accomplished by test, analysis, or a combination SEISMIC CATEGORY I of the two methods.

INSTRUMENTATION AND ELECTRICAL EQUIPMENT (INCLUDING OTHER in general, analysis is used to supplement DYNAMIC LOADS) test data although simple components may lead themselves to dynamic analysis in lieu of full This section is supposed tiaddress only scale testing. The deciding f actors for seismic qualification of electrical components choosing between tests or analysis include:

and equipment in accordance with NRC Regulatory Guide 1.70 Revision 3. However, recognizing that (1) magnitude and frequency of seismic and other dynamic loads due to suppression pool dynamics RBV dynamic loadings; associated with a loss-of-coolant accident (LOCA) and safety / relief valve (SRV) discharge can have (2) environmental conditions (Subsection 3.11.1) a significant vibratory effect on the reactor associated with the dynamic loadings; building, and, hence, on the design of struc-tures, systems, and equipment in the reactor (3) nature of the safety-related function (s);

building, GE has elected to address equipment qualification for both seismic and other reactor (4) size and complexiry of the equipment; building vibration (RBV) dynamic loads in this section. The format utilized is consistent with (5) dynamic characteristics of expected failure R.G.1.70, Revision 3; thus, reference toJhe uodes (structural or functional); and gerating basia carthquaac (UtsE) and tbcTsafe shutdown earthquake (55E) in this section include (6) partial test data upon which to base the the combined seismic and other RBV dynamic analysis.

loads. The non seismic RBV dynamic loads are described in Table 3.9-2.

.T M F # T 3./ 0 The selection of qualification methods to be used is largely a matter of engineering judge-The mechanical components and equipment and ruent; however, tests, and/or analyses of assem- [

the electrical components that are integral to blies are preferable to tests or analyses on the mechanical equipment are dynamically separate components (e.g., a. er and a pump, qualified as described in Section 3.9.

including the coupling and otR. appurtenances should be tested or analyzed as an assembly).

Principal Seismic Category I structures, 3.10.1.2 input Motion systems and components are identified in Table 3.2-1. Most of these items are safety related as The input motion for the qualification of explained in Subsection 3.2.1. The safety-equipment and supports is defined by response related functions are defined in Section 3.2, and spectra. The required response spectra (RRS) include the functions essential to emergency are generated from the buildings dynamic analy-reactor shutdown, containment isolation, reactor sis, as described in Section 3.7. They are core cooling, reactor protection, containment and grouped by buildings and by elevations. This reactor heat removal, and emergency power supply, RRS definition incorporates the contribution and or otherwise are essential in preventing other RBV dynamic loads as specified by the load significant release of radioactive material to combination Table 3.9 2. Thjp esponse spectra the environment.

curves for the SSE4nd OBDare presente ' in Ap-pendix 3G. When one type of equipment is locat-3.10.1 Seismic Quallnestion Criteria ed at several elevation and/or in several build-(Including Other Dynamic Loads) ings, the governing response spectra are specificd.

3.10.1.1 Selection of QuallSention Method 3.10.1.3 Dfnamic Quallfleation Program Dynamic qualification of Seismic Category I instrumentation and electrical equipment is The dynamic qualification program is de-DisEFr 3.10 The COL applicant m

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must ensure that specific seismic and dynamic input response spectra are properly defined and enveloped in tha methodology for its specific plant and implemented in its equipment qualification program.

ABM naawaz Standard Plant nrv a by dynamic analysis using appropriate exceeded when the tubing is subjected to the response spectra, loads specified in Subsection 3.9.2 for Class 2 and 3 piping.

(b) Moor Response Spectra 3.10.4 Operating License Review (Tests and (i) Roor response spectr4 used are Analyses Results) those generated for the supporting fl o o r. In case supports are See Subsection 3.10.5.2 for COL license attached to the walls or to two information requirements.

different locations, the upper bound envelope spectra obtained by 3.10.5 COL License Information superimposing are used.

3.10.5.1 Equipment Qualification @tecords) eot (ii) In many cases, to facilitate the

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3,f p f design, several floor response The equipment qualification records spectra are combined by an upper including the reports (see Subsections bound envelope obtained by 3.10.2.1.4 and 3.10.2.2.3) shall be maintained superimposing.

in a permanent file and shall be readily available for audit.

3.10.3.2.3 Local Instrument Supports 3.10.5.2 Dynamic Qualification Report For field-mounted Seismic Category I instruments, the following is applicable:

A dynamic qualification report (DQR) shall be prepared identifying all Seismic Category I (1) The mounting structures for the instruments instrumentation and electrical parts and have a fundamental frequency above the equipment therein and their supports. The DOR excitation frequency of the RRS.

shall contain the following: (1) A table or file for each system that is identified in (2) The stress level in the mounting structure Table 3.21 to be safety related or having does not exceed the material allowable Seismic Category I equipment shall be included stress whea the mounting structure is in the DOR "wa6% the MPL item nr.mber and subjected to the maximum acceleration level name, the qualification method and the input for its location.

motion for all Seismic Category I equipment and the supporting structure in the system, 3.10.3.2.4 Instrument Tubing Support and the corresponding qualification summary table or vendor's qualification report. (2)

The following bases are used in the seismic The mode of safety related operation (i.e.,

and other RBV dynamic loads design and analysis active, manual active or passive) of the of Seismic Category I instrument tubing supports:

instrumentation and equipment along with the manufacturer identification and model numbers (1) The supports are qualified by the response shall also be tabulated in the DOR. T h'e spectrum method; operational mode identifies the instrumentation or equipment (a) that performs (2) Dynamic load reatraint measures and analpis the safety related functions automatically, for the suppr "s are based on combined (b) that is used by the operators to perform i

limiting valas for static load, span the safety related functions manually, or (c) i length, and computed dynamic response; and whose failure can prevent the satisfactory accomplishment of one or more safety.related (3) The Seismic Category I instrument tubing functions. (See Subsection 3.10.4).

systems are supported so that the allowable stress permitted by Section III of ASME Boiler and Pressure Vessel Code are not IMee r 1.10.r.)

COL applasnts wIIkwide. f% tit'ew.& reim/N ml Ow"'c l'wameio:t 4r % egu@meda uRf,calim pnfrem nu runeum n i n O cw d nes a.nD S& b ra c.hin 110,

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ABM m61omo gev c Standard Plant conservative direction simultaneously. The 6.3.4.2 Reliability Tests and Inspections results of this calculation for the limiting case are given in Figure 6.3 67 through 6.3-75 and The average reliability of a standby Table 6.3-4 Since the ABWR results have large (nonoperating) safety system is a function of margins to the 10CFR50.46 licensing acceptance the duration of the interval between periodic criteria, the ABWR licensing PCT can be based on functional tests. The factors considered in the bounding PCT which is well below the 22000F determining the periodic test interval of the i

PCT limit.

ECCS are: (1) the desired system availability (average reliability); (2) the number of 633.8 LOCA Analysis Conclusions redundant functional system success paths; (3) the failure rates of the individual components Having shown compliance with the applicable in the system; and (4) the schedule of periodic acceptance criteria of Section 6.3.3.2, it is tests (simultaneous versus uniformly staggered concluded that the ECCS will perform its function versus randomly staggered).

in an acceptable manner and meet all of the criteria in Appendix 43, given operation at or All of the active components of the HPCF below the MAPLHGRs provided by the utility for System ADS, RHR and RCIC Systems are designed l

cach fuel bundle. See Subsection 6.3.6.

so that they may be tested during normal plant i

operation. Full flow test capability is 6.3.4 Tests and Inspections provided by a test line back to the suction source. The full flow test is used to verify 63.4.1 ECCS Performance Tests the capacity of each ECCS pump loop while the plant remains undisturbed in the power All systems of the ECCS are tested for their generation mode. In addition, each individual l

operational ECCS function during the valve may be tested during normal plant preoperational and/or startup test program. Each operation.

component is tested for power source, range, direction of rotation, setpoint, limit switch All of the active components of the ADS setting, torque switch setting, etc. Each pump System, except the safety / relief valves and is tested for flow capacity for comparison with their associated solenoid valves, are designed j

vendor data. (This test is also used to verify so that they may be tested during normal plant i

flow measuring capability). The flow tests operation. The SRVs and associated solenoid l

involve the same suction and discharge source valves are all tested during plant initial power (i.e., suppression pool),

ascension per Appendix A, Paragraph D.2.c of Regulatory Guide 1.68. SRVs are bench tested to t

Alllogic elements are tested individually and establish lift settings.

then as a system to verify complete system response to emergency signrls including the Testing of the initiating instrumentation and ability of valves to revert to the ECCS alignment controls portion of the ECCS is discussed in from other posit'ons.

Subsection 7.3.1. The emergency power system, which supplies electrical power to the ECCS in j

Finally, the entire system is tested for the event that offsite power is unavailable, is response time and flow capacity taking suction tested as described in Subsection 8.3.1. The i

from its normal source and delivering flow into frequency of testing is specified in the Chapter the reactor vessel. This last series of tests is 16 Technical Specifications. Visual inspections performed with power supplied from both offsite of all the ECCS components located outside the power and onsite emergency power.

drywell can be made at any time during power operation. Components inside the drywe!! can be See Chapter 14 for a thorough discussion of visually inspected only during periods of access preoperational testing for these systems.

See S A rs k as s. 3. 6. L Nor COG li c.e w s a mSo%ab e r* 9M'9 O

E CCS &

yW 0 13 Arnendrnent 15 i

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23A6100AB Standard Plant uy c low water level and drywell high pressure plus indication that at least one RHR or HPCF pump is operating. He HPCF, RCIC, and RHR automatically return from system flow. test modes to the emergency core cooling mode of operation following receipt of an automatic invitation signal. The RHR LPFL mode injection into the RPV begins when reactor pressure decreases to the RHR's pump discharge shutoff pressure.

HPCF injection begins as soon as the HPCF pump is up to speed and the injection valve is open, since the HPCF is capable of injection water into the RPV over a pressure range from 1177 psid to 100 psid or pressure difference between the vessel and drywell.

l 6.3.6 COL License Information 63.6.1 ECCS Performance Results ne exposure dependent MAPLHGR, peak cladding temperature, and oxidation fraction for each fuel bundle design based on the limiting break size will be provided by the COL applicant to the co t. (,, 3, g 1- )

USNRC for information. (See Subsection 633).

2-

+

6.3.7 Reference 1.

General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFRSO, Appendir K, (NEDE-20$66-P-A),

September 1986.

r 6.3.6.2 ECCS Testing Requirements The COL applicant will perform a test every refueling of each ECCS subsystem in accordance with Technical Specification SR 3.5.1.7.

(See Subsection 6.3.4);

)

6.3.t5 Amendment 23 l

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ABWR 23A6100AJ Standard Plant mma suction side of the drain pump. This switch will tion valves are interlocked with the circulating water automatically stop the pump in the event of low pumps so that when a pump is started, its discharge water lev:1 in the standpipe to protect the pump valve will be opening while the pump is coming up to from mmive cavitadon.

speed, thus assuring there is water flow through the pump. When the pump is stopped, the discharge 10A.5.3 Evaluadon valve closes automatically to prevent or minimize backward rotation of the pump and motor.

The CWS is not a safey related system; however, a flooding analysis of the turbine building is Level switches monitor water levelin the con-performed on the CWS postulating a complete denser discharge water boxes and provide a permis-rupture of a single expansion joint. The analysis sive for starting the circulating water pumps. These assumes that the flow into the condenser pit comes level ewitches ensure that the supply piping and the from both the upstream and downstream side of the condenser are full of water prio to circulating water break and, for conservatism, it assumes that one pump startup thus preventing water pressure surges system isolation valve does not fully close.

from damaging the supply piping or the condenser.

Based on the above conservative assumptions, To satisfy the bearing lubricating water and shaft the CWS and related facir. ties are designed such that scaling water interlocks during startup, the circulat-the selected combination of plant physical arrange-ing water pump bearing lubricating and shaft seal ment and system protecdve features ensures that all flow switches, located in the lubricating seal water credible potential circulating water spills insid,e the supply lines, must sense a muursum Dow to provide turbine building remain confined inside the con-pump start permissive.

denser pit. Further, plant safety is ensured in case of multiple CWS failures or other negligible probability Monitoring the performance of the circulating CWS related events by the plant safety related gen-water system is accomplished by differential pressure eral flooding protection provisions that are discussed transducers across each half of the condenser with in Secdon 3.4.

' remote differential pressure indicators located in the r

main control room. Thermal element signals from

(

10AJA Tests and laspections the supply and discharge sides of the condenser are transmitted to the plant computer for recording, The CWS and related systems and facilities are display and condenser performance calculations.

tested and checked for leakage integrity prior to initial plant startup and, as may be appropriate, To prevent icing and freeze up when the ambient following major maintenance and inspection.

temperature of the ultimate heat sink falls below 32 F, warm water from the discharge side of the All active and selected passive components of mndenser is recirculated back to the screen house the circulating water system are accessible for intake. Thermal elements, located in each condenser inspection and maintenance / testing during normal supply line and monitored in the main control room, power station operation.

are utilized in throttling the warm water recirculation valve, which maintains the minimum inlet tempera-10A33 Instrumentation Applications ture of approximately 40 F.

J h

Temperature monitors are provided upstream 10A33 Dood Protection and downstreams of each condenser shell section.

See response to Question 430.73(b), protection Indication is provided in the control room to,

against a CWS pipe, water box or expansion joint l

identify open and closed positions of motor-operated ( failuref-butterfly valves in the CWS piping.

10A.6 Condensate Cleanup System All major circulating water system valves which control the flow path can be operated by local The condensate cleanup system (CCS) purifies controls or by remote manual switches located on

. and treats the condensate as required to maintain the main control board. The pump discharge isola-reacter feedwater purity, using filtration to remove IN SE9.T A 10A10 Amendment 17 1

ci mz.i cr i c emo awa Standard Plant wn tube bundle, initiate power reducion and faulty tube bundle drain down if required, and arrange for water box entry and leak repair at the earliest appropriate time.

QUESTION 430.72 Provide the permissible coolbg water inleakage rate and the allowed time of operation with inleakage.

(10.4.1)

RESPONSE 430.72 The polishing system is sized to meet the chemistry requirements for continuous operation while operating continuously with a condenser leak of 0.001 gpm and to maintain water quality during an orderly unit shutdown (not longer than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) with a leak of 0.1 gpm until repairs can be made. The design is adequate to c!can up the feed and condensate system during plant heatup and low power operation without limiting plant startup time. The number and sizing of the ion mhangers are such that the functional reqmrements are met while permitting the replacement of resin in one ion mh=ger at a time. The ABWR Standard Plant design features facilitate replacement of ion mh ange resin.

QUESTION 430.73 Provide information on the following items-(10.4.1)

(a) Provisions incorporated into the main condenser design to preclude component or tube failure due to steam blowdown from the turbine bypass system.

(b) Worst possible flood levelin the applicable buildings due to complete failure of main condenser and provisions for protecting safety related equipment locatealin the buildings against such floodmg (note that ABWR SSAR Section 3.4 does not discuss the turbine building).

RESPONSE 430.73 (a) Specific provisions inside the condenser to preclude condenser tube damage due to turbine bypass steam impingement are to be defined by the condenser vendor for each project. Typically the provision inside the condenser consists of a horizontal perforated steam distribution pipe enclosed in a perforated guard

]

pipe designed to protect the condenser internals from steam impingement. The perforated pipe and its guard pipe run the fulllength of the condenser and are supported above the condenser tube bundle.

See nw.I SAstchon to.4 5.6 4.c & ** 8 P m8*A*

'TM*8'h'"'

(b)hfA circulating water system pipe, waterbox, or capanuon jomt failure, if not detected and isolated, would

[ cause internal turbine building floodmg up to slightly over grade level, with excess flood waters potentially spilling over on site. If a failure occurred within the condensate system (condenser shell side), the resulting flood level would be less than grade level due to the relatively small hotwellwater inventory relative to the condenser pit capacity. In either event the floodmg of the turbine building would not affect safety related equipment since no such equipment is located inside the turbine building and all plant safety related facilities are protected agamst site surface water intrusion e

QUESTION 430.74 Discuss how the components of the main condenser evacuation system (MCES) conform to the guidelines of Regulatory Guide 1.26,1.33. and 1.123 with respect to quality group classification and quality assurance programs.(10.4.2)

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