ML20036A058
| ML20036A058 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 04/30/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9305070344 | |
| Download: ML20036A058 (10) | |
Text
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e-6 April 30,1993 Docket No. STN 52-001 Chet Posiusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation
Subject:
Submittal Suppo: ting Accelerated ABWR Review Schedule - DFSER Open Item 3.8.3-1 and COL Action Item 9.4.81
Dear Chet:
Enclosed are SSAR markups addressing DFSER Open Item 3.8.3-1 and COL Action Item 9.4.8-1.
Please provide copies of this transmittal to Tom Cheng and Butch Burton.
Sincerely, t,
tek Fox Advanced Reactor Programs cc: Gary Ehlert (GE)
Norman Fletcher (DOE) 060033 1 s JiW124 9305070344 930403 PDR ADDCK 05200001 A
I; O.I'3.11.3 1
. ABWR mm.
Standard Plant ney. 3 internal structures of the containment conform to (2) Construction Loads--Construction loads are the applicable codes, standards, and specifica-loads which are applied to the containment tions and regulations listed in Table 3.8-4 internal structures from start to completion except where specifically stated otherwise.
of construction. The definitions for D, L and T are applicable, but are based on Structure or Specific Reference actuaT construction methods and for comoonent Number conditions.
Diaphragm Floor 14 (3) RV2-Loads from component response or direct fluid forces, on components located in the Reactor Pedestal 1 13,15-22 suppression pool, caused by safety relief valve air cleaning leads.
Reactor Shield Wall 1 13,15-22 (4) RBV--Loads due to reactor building vibra-DEPSS 15-22 tions caused by an SRV and LOCA event.
3 MisecIlaneous platforms 15 22 (5) AP--Loads and pressures directly on the reactor shield wall and loads from component L/D Equipment Tunnel 15-22 response or direct steam flow forces on components located in the reactor vessel L/D PersonnelTunnel 15-22 shield wall annulus region, caused by a 1
rupture of a pipe within the reactor vessel l
Reactor Shield Wall 15 22 shield wall annulus region.
(Stabilizer (6) SL -Loads from component response or direct 3.83.3 Imds and Load Combinations fluid forces, on components located in the sloshing zone of a pool or component, caused 3.833.1 Imad Dennitions by the sloshing phenomenon from any dynamic event.
The loads and applicable load combinations for which the structure is designed depend on the 3.83.3.2 Imad Coenbination conditions to which the particular structure is subjected.
The load combinations and associated accep-tance criteria for concrete and steel internal' The containment internal structures are de-structures of the containment are listed in Ta-signed in accordance with the loads described in ble 3.8.3-5 and Table 3.8.3-6, respectively; for Appendix 3B. These loads and the effects of the reactor shield wall refer to appendix 3B.
these loads are considered in the design of all internal structures as applicable. The loads 3.83.4 Design and Analysis Procedures within the loading combinations are combined using the absolute sum technique. (Those loads 3.83.4.1 Diaphragna Floor which are defined as reversible in algebraic sign are combined in such a way as to produce the-The design and analysis procedures used for maximum resultant stresses in the structure. All the diaphragm floor are similar to those used other loads are combined in accordance with their for the containment structure. The diaphragm direction of application to the structure.) The slab is included in the finite element model-loads are defined in Subsection 3.8.1.3 except as described in Subsection 3.8.1.4.1.1.
l follows:
3.83.4.2 F* Pedestal (1) P --Pressure loads resulting from the n8rmal operating pressure difference between The reactor pedestal is included in 'the the drywell (upper and lower) and the finite element model described in Subsection suppression chamber of the containment.
3.8.1.4.1.1.
3.8-18 Amendment 1 t
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O't 3,9.3-1 ABWR m uoont Standard Plant nrv a The design and analysis is based on the clas-respectively.
tic method. All loads are resisted by the integ-ral action of the inner and outer steel shells.
3.83.6 Materials, Quality Control, and The concrete placed in the annulus between the Special Construction Techniques inner and outer shells acts to distribute loads between the steel shells, and provides stability 3.83.6.1 Diaphragm Floor to the compression elements of the pedestal.
The materials, quality control, and construc-3.83.43 Reactor Shield Wall tion techniques used for the diaphragm floor and liner plate are the same as those used for the The design and analysis procedures used for containtnent wall and liner plate in Subsection the reactor shield wall are similar to those used 3.8.1.6.
for the reactor pedestal described in Subsection 3.83.4.2.
3.83.6.2 Reactor Pedestal 3.83.4.4 Drywell Equiprnent and Pipe Support The materials conform to all applicable Structure requirements of ANSI /AISC N690 and ACI 349 and comply with the following:
l The drywell equipment and pipe support struc-l!Im Specification ture is designed using the AISC working stress methods for steel safety-related structures for inner and outer shells ASTM A441 nuclear facilities (ANSI /AISC N690). The design of beams supporting pipe whip restraints allows Internal stiffeners ASTM A441 inelastic deformations due to postulated pipe rupture loads. All safety related items which Concrete fill f' c = 4000 psi the inelastic beam deformations may affect are evaluated to verify that no required safety 3.83.63 Reactor Shield Wall function would be compromised.
The materials conform to all applicable 3.83.4.5 Other Internal Structures requirements of ANSI /ASIC N690 and ACI 349 and comply with the following:
The design and analysis procedures used for other internal structures are similar to those lita Snecification used for the drywell equipment and pipe support structure as described in Subsection 3.83.4.4.
Inner and outer shells ASTM A441 3.83.5 Structural Acceptance Criteria Internal stiffeners ASTM A441
%.o 3.83.5.1 Diaphragm Floor Concrete fill f ' c = 4994 psi The calculated and allowable stresses for the Stainless Steel Clad SA-240 Type 304 L diaphragm floor are found in Tables 3H3-8 and 3H 3-9.
3.83.6.4 Drywell Equipment and Pipe Support Structure 3.83.5.2 Reactor Pedestal The materials conform to all applicable The calculated and allowable stresses for the requirements of ANSI /AISC N690 and comply with reactor pedestal are found in Tables 3H3-10 and the following:
3H 3-11.
3.8.53 Other internal Structures Structural steel and ASTM A36 The structural acceptance criteria for other connections internal concrete or steel structures are in accordance with ACl-349 and ANSI /AISC-N690, Amendmer.17 3 6-19
n ABM
@ 383" 23 m oaac Standard Plant V-REV B High strength structural ASTM A572 or A441 Other Seismic Category I structures which steel plates constitute the ABWR Standard Plant are the reactor building, control building and radwaste Bolts, studs, and nuts ASTM A325 or A490 building substructurc. Figure 1.2-1 shows the (dia. p_3/4 -)
spatial relationship of these buildings. The only other structure.in close proximity to these Bolts, studs, and nuts ASTM A307 structures is the turbine building. They are (dia. < 3/4 ")
structurally separated from the other ABWR Standard Plant buildings.
3.8.3.6.5 Other laternal Structures The Seismic Category I structure within the The materials conform to all applicable ABWR Standard Plant, other than the containment requirements of ANSI /A!SC N690 and comply with structures, that contains high energy pipes is i
the foUowing; the reactor building. The steam tunnel walls protect the reactor building from potential Im;1 Sriedfication impact by rupture of the high energy pipes.
This building is designed to accommodate the Miscellaneous platforms Same as Section guard pipe support forces.
3.83.6.4 The reactor building, steam tunnel, residual Lower drywell equipment ASTM A516 Grade 70 heat removal (RHR) system, reactor water cleanup tunnel SA-240 Type 304 L (RWCU) system, and reactor core isolation cool-ing (RCIC) system rooms are designed to handle Lower drywell personnel ASTM A516 Grade 70 the consequences of high energy pipe breaks.
tunnel SA 240 Type 304 L The RHR, RCIC, and RWCU rooms are designed for differential compartment pressures, with the Reactor shield wall stabilizer associated temperature rise and jet force.
Steam generated in the RHR compartment from the
-tube sections ASTM A501 postulated pipe break exits to the steam tunnel i
through blowout panels. The steam tunnelis k --plates ASTM A36 vented to the turbine building through the seismic interface restraint structure (SIRS).
Lower drywell floor fill A material other The steam tunnel, which contains several pipe-material than limestone lines (e.g., main steam, feedwater, RHR), is al-concrete so designed for a compartment differential pres-sure with the associated temperature changes and 3.83.7 Testing and Inservice Inspection jet force.
Requirements Seismic Category I masonry walls are not used A formal program of testing and inservice in-in the design. The ABWR Standard Plant does not spection is not planned for the internal struc-contain seismic Category I pipelines buried in tures except the diaphragm floor, reactor pedes-soil, tal, and lower drywell access tunnels. The other internal structures are not directly related to 3.8.4.1 Description of the Structures the functioning of the containment system; therefore, no testing or inspection is performed.
3.8.4.1.1 Reactor Building Structure Testing and inservice inspection of the dia.
The reactor building (RB) is constructed of phragm floor, reactor pedestal and lower drywell reinforced concrete with a steel frame roof.
access tunnels are discussed in Subsection The RB has four stories above the ground level 3.8.1.7.
and three stories below. Its shape is a rectangle of 59 meters in the E-W direction,56 i
3.8.4 OTHER SEISMIC CATEGORY I meters in the N-S direction, and a height of
]
STRUCTL7tES about 57.9 meters from the top of the basemat.
i Amendment 14 3.8-20 i
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23A6100AE Standard Plant REV B Table 3.8-4 CODES, STANDARDS, SPECIFICATIONS, AND REGULATIONS USED IN THE DESIGN AND CONSTRUCTION OF SEISMIC CATEGORY I INTERNAL STRUCTURES OF THE CONTAINMENT (Continued)
SPECIFICATION SPECIFICATION REFERENCE OR STANDARD NUMBER DESIGNATION TITLE 15 ANSI /AISCN690 Specification for the Design, Fabrication, and Erection of SteeJ Safety.Related Structures for.
3.5.3 modifiea b SR P's Nuclear Facilities ( os.S.t, 3.B.3,cm a 3.S.tQ,
3 16 AWS DI.1 StructuralWeldirg Code 17 NCIG-02 Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants 18 ANSI /ASME Ouality Assurance Program Requirements for NOA-1-1986 Nucleat Facilities 19 (Deleted) 20 NRC Regulatory Quality Assurance Requirements for Installation, Guide 1.94 Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants 21 NRC Regulatory Materials for Concrete Containments Guide 1.136 (Article CC-2000 of the Code for Concrete Reactor Vessels and Containments) 2 22 NRC Regulatory Safety-Related Concrete Structures for Nuclear Power Guide 1.142 Plants (Other than Reactor Vessels and Containments)
Ernlanation of Abbreviations ACI American Concrete Institute AISC American Institute of Steel Construction AISI American Iron and SteelInstitute ANSI American National Standards Institute ASME American Society for Mechanical Engineers AWS American Welding Society NCIG Nuclear Construction Issues Group NRC Nuclear Regulatory Commission NOTES:
1.
Unless specified, the Edition of the Specification or Standard shall be the latest issued for industry use.
l 3 B-33 Amendment 17
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maan Standard Plant nev n in the supply system ductwork regulates the flow of air to maintain the service building clean areas at a positive pressure with respect to atmosphere.
9.4.7 Diesel Generator Area Ventilation System (4) In the : vent of a loss of offsite electric power, the service building system is shut down.
"'S N He diesel generator area ventilation system is a
part of the reactor building ventilation system de-9.412 System Description scribed in Subsection 9.4.5.
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,p (1) The schematic design of the service building )
9.4.8 Service Building Ventilation System ventilation system is shown on Pigure 9.4 7p This system serves all areas within the service I(4) The service building ventilation system supplies building, including locker rooms, men and women's filtered, heated or cooled air to the general change rooms, laundry, lunch room, instrument areas through a central fan system consisting of repair room, HVAC equipment rooms, and sersa.
an outside air intake, filters, a heating coil, a Ybe.. T5 c. -bePE.,; 2:r' re:-. This system operates during cooling coil, two 50% capacity supply air fans, all normal station condidons.
and supply air ductwork.
9A11 Design Basis 2(}) The two 50% capacity exhaust air fans induce the ventilation air from the clean areas through 9A11.1 Safety Design Basis the exhaust ducts and discharge the air to the common station vent.
The service building HVAC system is not required to function in any but the normal station operating 3(4) Potentially contaminated air is routed through conditions and, therefore, has no safety bases.
mechanical moisture separators, prefilters, and high-efficiency particulate filters (HEPA) before 9A11.2 Power Generation Design Bases being routed to two 50% capacity exhaust air fans to discharge the air to the common station (1) ne service building HVAC system is deined vent.
to maintain a quality environment suitable for personnel health and safetyin the service build- /.i (5) The potentially contaminated areas are main-ing. It is designed to limit the maximum tem-tained at a slightly lower pressure then the sur-perature in the service building to 85"F. The rounding clean areas and, therefore, the air temperature in each area conforms to the flows from the clean areas to these potentially mntam' mteA areas.
equipment requirements in that area.
m (2) ne system provides a quantity of filtered out-5(6) Pressure control dampers are employed be-door air to purge any possible contamination.
tween clean and parenti=Hy contaminated areas This air is processed through the dust collec-and are of the back flow type and fait closed.
1 rs. Prefilters and HEPA filters (as required)
This minimim the back flow of contaminated c harcoa.!
s before it is monitored for radioactivity and air to clean areas when there,is a loss of power 3
released to the atmosphere, and subsequent fan system shutdown.
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(3) Both the supply air system and the exhaust air (7) Controls andInstrumentation system operate manually and continuously.
Isolation dampers at each supply fan, each (a) Each fan and each exhaust fiker package is exhaust fan, and each filter package close when controlled by hand switches located on local the respective equipment is not operating.
control panels. Pertinent system flow rates There is an additional isolation damper at the and temperatures are also indicated on the j
supply air inlet which closes when the supply air local control panels. Trouble on local con-system is not operating. An automatic damper Amendment 6 9.43
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- a. m trol panels is annunciated on the main con-Maintenance will be performed on a scheduled i
trol board.
basis in accordance with the equipment manufactur-cr's requirements.
(b) Controls are pneumatic and electric.
The system it in operation during normal plant op.
(c) Instrumentation is prended for monitoring eration.
system operating variables during normal station operating conditions. The loss of air flow; hi h and low system temperature; S
and high differential pressure across the n;;'j d.- Fir, in r"-- r, r '
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'"E 9AJ.3 Safety Evaluation (1) The service building ventilation system is not safety.related and is not required to assure either the integrity of the reactor coolant pressure boundary or the capability to shut down the reactor and main:ain it in a safe shutdown conditions.
(2) Pressure control dampers a:e employed be-y i
Iween clean and potentially contaminated arcas and are of the back flow type and fail closed.
This minimizes the back flow of contaminated air to clean areas when there is a loss of power and subsequent fan system shutdown.
(3) The system incorporates features to assure its reliable operation over the full range of normal i
station conditions.
VAJA Testing and laspection All equipment is factory inspected and tested in ac-t cordance with the applicable equipment specifica-tions and codes. System ductwork and erection of equipment is inspected during various construction
- stages. Preoperational tests are performed on all mechanical components and the system is balanced j
for the design air, and water flows and system oper-ating pressures. Controls, interlocks and safety de-l vices on each system are c'aceked, adjusted, and tested to ensure the proper sequence of operation.
A finalintegrated preoperational test is conducted with all equipment and controls operational to verify l
the system performance.
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i Figure 9.4-7. SERVICE BUILDING VENTILATION SYSTEM
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l (5) The clean areas served by the service building HVAC system has an emergency filter train.
It is manually operated.
In an emergency l
it supplies filtered air for the TSC and other normally clean areas.
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i (6) The clean areas of the service building is provided with an emergency filter train consisting of a heater /demister, prefilter, HEPA filter, 2" charcoal fileter bed, a second HEPA filter, and a fan.
insert c (8) The COL applicant will provide a detailed P&ID and an equipment list, for the service building HVAC system, including the TSC, for NRC review. (See Subsection 9.4.10.1 for COL License Information)
I 9.4.10 COL License Information i
The COL applicant will provide a detailed P&ID and an equipment list, i
for the service building HVAC system, including the TSC, for NRC review.
t (See Subsection 9.4.8.2)
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