ML20076A927
| ML20076A927 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 06/26/1991 |
| From: | Recasha Mitchell GENERAL ELECTRIC CO. |
| To: | Chris Miller NRC |
| References | |
| EEN-9144, MFN-066-91, NUDOCS 9107090128 | |
| Download: ML20076A927 (27) | |
Text
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GE Nuclear Encryy June 26,1991 MFN No. 066-91 Docket No. STN 50-605 EEN 9144 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention:
Charles L Miller, Director Standardization and Non Power Reactor Project Direewrate w
Subject:
GE Responses to Discussion items from Conf <:ience Calls of June 3 1991 and June 5,1991 with Plant Systems Branch Enclosed are thirty four (34) copies of the GE iesponses to the subject discussion items.
It is intended that GE will amend the SSAR, as appropriate, with the enclosed responses in a future amendment.
Sincerely,
.C.
W-R.C. Mitebell, Acting Manager Re ulatory and Analysis Services M
3S2, (40S) 925-6948 cc: F. A. Ross (DOE)
D. C. Scaletti (NRC)
W. F. Burton (NRC)
D. R. Wilkins (GE)
J. F. Quirk (GE) b i} "
9107090128 910626 f
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PLANT SYSTEMS OPEN ITEMS FOR DISCUSSION 3.4.1 FLOOD PROTECTION OPEN ITEM (1)
Safety-related systems and components that may be affected by external floods are protected either beca"3e of their location above the design flood level or because they are enclosed in reinforced concrete Seismic Category I structures which have a required wall thickness of not less than two feet for portions of the structures below the flood level.
The ABWR Safety Analysis Report did not address the effects of standing water on the roofs of safety related buildings and the means to limit the amount of such water as required by the guidelines of RG 1.102.
The ability of these structures to withstand the effects of standing water remains an open issue.
RESPONSE (1)
Response to this open item is provided in revised Subsection 3.4.1.2.1 (page 3.4-1).
OPEN ITEM (2)
Analysis of rupture of moderate-energy piping larger than one-inch diameter are performed in accordance with ANSI /ANS 56.11
" Standard Design Criteria for Protection Against the Effects of
. Compartment Flooding in Light Water Reactor Plants", and Crane Co., Technical Paper No. 410, #1973," Flow of Fluids Through Valves, Fittings, and Pipe". Hinh energy line breaks in the Main Steam Line (MSL) tunnel are excluded from evaluation, since this area is instrumented for detection of leaks before a line break.
However, even in the event of feedwater line break, water will be contained in the Seismic Category I structure of the MSL tunnel and be sllowed to drain to the High Conductivity Water (HCW) sumps. The applicant described the operator response period and type of flooding source.
The staff believes that a leak-before-break analysis should use plant-specific data such as piping geometry, materials, fabrication procedures, and pipe support locations.
Therefore the staff will evaluate the acceptability of the above exclusion on a plant-specific basis.
RESPONSE (2)
Response to this open item is provided in revised Subsection 3.4.1.1.2 (page 3.4-2).
OPEN ITEM (3)
Analysis of the worst flooding due to pipe and tank failures and their consequences are performed on a floor-by-floor basis within the Reactor Building to demonstrate the safe shutdown of the reactor. However, the applicant did not discuss if the safety-related equipment within this structure (or any of the other structures analyzed) is capable of normal function while subjected to liquid spill, and completely or partially flooded.
This is an open issue.
RESPONSE (3)
Response to this open item is provided in revised Subsection 3.4.1.1.2 (page 3.4-2).
OPEN ITEM (4)
The ABWR SSAR did not include flood analysis for any structures outside the scope of the generic design housing systems or components performing a safety function, such as the ultimate heat sink pump house.
The identified interface requirements, essentially identifying only the normal groundwater and flood levels, are insufficient to insure ad3quate flood protection design for these structures.
Interface requirements that will insure the ability of the plant specific application to meet the flood protection requirements as described in SRP Section 3.4.1 (that is the requirements of GDC 2 and the guidance of RG 1.102) need to be provided.
This is an open issue.
RESPONSE'(4)
Response to this open item is provided in revised subsection 3.4.1.1.2 (page 3.4-2).
OPEN ITEM 9.1.1 (NEW FUEL STORAGE)
GE has not included in the interface requirements that the design of the new fuel storage racks will be such that the K gg will not C
e exceed 0.98 with fuel of the highest anticipated reactivity in place assuming optimum moderator conditions (foam, small droplets, spray, or fogging) as described in SRP Section 9.1.1.
This is considered an open issue.
RESPONSE 9.1.1 This open item is covered by response to Question 430.180.
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4 9.1.2 SPENT FUEL STORAGE OPEN ITEM (1)
The reactor building housing the facility is designed to seismic Category I criteria, as are the storage racks and other " fuel storage facilities", including the gates between the spent fuel pool and other pools.
However, information on the seismic classification of the spent fuel pool liner is not contained in H
the SSAR. This is considered an open issuo.
RESPONSE (1)
Response to this open item is provided in revised Table 3.2-1 (page 3.2-18) and Subsection 9.1.2.1.2 (page 9.1-?).
OPEN ITEM (2)
The design of the storage pool includes the provision of radiation monitoring systems described in SSAR Section 11.5, which were determined to satisfy in part the requirements of GDC 63, " Monitoring Fuel and Wasto Systems."
response to the request for additional information ( Responso 430.191) do not include sufficient information to conclude that there are acceptable monitoring systems for pool water level and excessivo pool liner leakage.
This is considered an open issue, i
RESPONSE (2)
Response to this open item is provided in revised Subsections 9.1.3.2 (pa,c 9.1-4) and 9.1.3.3 (page 9.1-5).
9.1.4 LIGHT LOAD HANDLING SYSTEM OPEN ITEM (1) e The refueling platform is designed to Seismic Category I standards and the entire system is housed within the reactor building-which is a Seismic Category I, flood-and tornado-protected structure.
However, the new fuel inspection stand is not classified as seismic Category I.
It is not entirely clear from the SSAR description that in the event of an earthquake the fall of the new fuel stand would not be detrimental to criticality or radiological safety.
This is considered an open issue.
RESPONSE (1)
The new fuel inspection stand is used to inspect the new fuel bundles before they are placed into the spent fuel pool.
The i-fuel is moved by the jib crane from the now fuel pit to the now l
fuel inspection stand.
The jib crane is always supporting
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the fuel bundles.
The jib crane is nonseismic but is designed not to drop its load in an SSE; therefore, it is unlikely that a now fuel arsembly will_ fall into the spent fuel pool.
The new fuel inspection stand is both supported off the floor and attached to the wall.
It is also unlikely that the stand will fall into the spent fuel poal; therefore, there is no need to make the new fuel inspection stand Seismic Category I.
OPEN ITEM (2)
Dropped.
OPEN ITEM 9.2.5 (ULTIMATE HEAT sit'K)
The conceptual design for the UHS is a spray pond.
The requirements of RG 1.72 are applicable to the design of a spray
-pond as an UHS, but were not referenced as interface criteria.
The identification of RG 1.72 design criteria as interface requirements remains an open issue.
RESPONSE 9.2.5 l
Response to this open item is provided in new Subsection 9.2.17.2 (pago 9.2-13).
9.2.14 TURBINE BUILDING COOLING WATER SYSTEM OPEN ITEM (1) l The TCW system is a non-safety related system designed to provide heat removal capability for various turbine island auxiliary equipment.
The TCW is a closed loop system consisting of two l
100% pumps, two 100t heat exchangers, a surge tank and associated I
piping, valves and instrumentation.
The description of the TCW in this section of the ABWR SSAR contains several l
inconsistencies.
The component and system descriptions of Section 9.2.14.2.3 " System Operation" and of Figuro 9.2-6a do not agree with the descriptions of Sections 9.2.14.2.1 " General Description" and 9.2.14.2.2 " Component Description" and the l
responses to RAIs.
The discrepancies (number of heat exchangers, pump capacities) should be corrected.
Because the descriptions in Sections 9.2.14.2.1 and 2 are more recent versions of the system description, they were analyzed as the TCW system.
These discrepancies remain an open item.
RESPONSE (1)
Response to this open item is provided in revised Subsection 9.2.14.2.3 (pages 9.2-10.1 and 9.2-11) and revised Figure 9.2-ba (page 9.2-42).
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OPEN ITEM (?)
The TCW is located in and near the turbine building away from safety related systems.
In response to RAIs 430.206 and.207, GE asserted thr.t failure of any TCW components, including the atmospheric surge tank, would not fail any safety-related equipment.
From equipment layout diagrams this statement appears to be true for all equipment shown on the diagrams.
The atmospheric surge tank does not appear on those diagrams.
Verification that tailure of this component will not affect safety-related systems is not possible.
This remains an open issue.
RESPONSE (2)'
l Response to this open item is provided in revised Subsections 9.2.12.2 (page 9.2-7) and 9.2.14.2.1 (page 9.2-10).
9.2.15 REACTOR SERVICE WATER OPEN ITEM (1)
The RSW is en open cycle system which provides cooling water to the RCW heat exchangers.
No other heat loads are supported by the RSW system.
The RSW system picks up heat from the RCW heat j
l exchangers and rejects the heat to the ultimate heat sink which is to be designed by individual applicants referencing the ABWR.
The GE scope of the RSW includes all the piping valves, pumps, heat exchangers, instrumentation, and controls from the system intake from the ultimate heat sink to the discharge back to the ultimate heat sink.
Although the total heat rate, total flow rate, temperature drop and pressure drop at the RCW heat exchangers for all identified modes of operation were provided for the RCW system, similar parameters for the RSW system (including identification of sufficient net positive suction head at pump suction locations considering low water levels) were not provided. This is an open issue.
RESPONSE (1)
Response to this open item is provided in revised Subsection 9.2.15.2 (page 9.2-12) and new Subsection 9.2.17.4 (page 9.2-13).
OPEN ITEM (2)
GE stated that both the mechanical equipment and piping and electrical equipment including instrumentation and controls of the redundant divisions of the RSW system aro sufficiently separated and protected to ensure availability of the needed equipment to perform reactor shutdown in the event of any of the following occurrences:
pipe rupture or equipment failure induced
- flooding,
6 4
spraying steam roloaso; pipe whip and jet forces from a postulated nearby high energy line break; missiles from equipment failuret firo; non-Category I equipment failuret or a single active componont failuro in the system. However insufficient detail has boon provided to insure that this design critoria can be mot.
Specifically, location and design features for the RSW pump and associated equipment have not boon specified.
RESPONSE (2)
Rasponse to this open itom is provided in revised Subsection 9.2.15.3 (page 9.2-12) and now Subsection 9.2.17.4 (page 9.2-13).
OPEN ITEM (3)
The ability of the RSW system to perform its function cannot be verified because pump design characteristics (flow, pressure, NPSH requirements) were not included in the SSAR (The required heat loads to be removed by the RSW system are identified.)
While the design requirements of the RSW system moet the intent of GDC 44, the ability of the RSW system to moot the requirements of GDC 44 romain an open item.
RESPONSE (3)
Response to this open item is provided in now subsection 9.2.17.4 (page 9.2-13).
OPEN ITEM 9. 2. ). 5 (TURBINE SERVICE WATER SYSTEM)
The TSW is located in the intake structure (the power cycle heat sink pump house) and the turbine building.
The system does not appear to have any connections with safety related systems, although insufficient detail is provided in the system description and diagrams to verify that no such connections exist.
The applicant must demonstrate that all safety related compononts, systems, and structures are protected from flooding in the event of a pipeline break in the TSW system in order to moot position C.2 of RG 1.29 and thus comply with GDC 2.
(Due to the site specific nature of the location of some TSW components this requirement may need to be expressed as an interface requirement).
RESPONSE 9.2.16 Responso-to this open item is provided in revised Subsection 9.2.16.3 (page 9.2-12.2) and now Subsection 9.2.17.5 (pago 9.2-13).
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l ABWR muut J
nrv n Standard Plant TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued)
I i
I Quality
(
Group Quality Safety loca.
Classi.
Assurance Selimle f F.ctra b
ge ficationd Recuirement' Calego-i Princloal componenta Clau i
F2 RPV Seniclog Equlpment 1.
Steamline plugs N
SC r
2.
Dryer and separator N
SC i
strongback and head strongback l
F3 RPV Internal Senicing Equf pment 1.
Control rod grapple N
SC 1
i l
i F4 Refueling Equipment -
(bb)
E
{
1.
Refueling platform N
SC
~
y j-2.
Refueling bellows N
SC i
l F5 Fuel Storage Equipment 1
L 1
(bb) i 1.
Fuel storage racks -
N SC c
new and spent E
(bb) 2.
Defective fuel storage N
SC
~
i.
. container I
3.- sp.J A) p. l hw.r w sc i
~ G1 Reactor Water Cleanup System li Vesselsincluding supports N
SC C
[
(filter /demineralizer) g i.
l
'A
- 2. - Regenerative beat exchaegers N
.SC C
including supports carrying
- reactoa water -
3.
Cleanup recirculation N
SC C
~
pump, motors 3218 Amendment 16
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Sundard Plant trv n 3.4 WATERI,EVEL(FLOOD) DESIGN 3.4.1.1.1 nood Prointion from Essernal Sources The types and methods used for protecting the j
ALWR safety.related structures, systems and Seismic Category I structures that may bc l
cornporents from external flooding shall conform - affected by design basis floods are desigred to to the guidelines defined in RG 1.102, withstand the floods postulated in Table 2.01 using the hardened protection approach with g
Criteria for the design basis for protection structural prov4 ions with ncorporated in the against external flooding shall conform to the plant design to protect safety.related requirements of RG 1.59. The design criteria for structures, systems, and components from protection against the effects of compartment postulated flooding. Seismic Category I flooding shall conform to the requirements of structures required for safe shutdown remain ANSI /ANS.56.11. The design basis flood levels accessible during all Hood conditions, are specified in Table 3.41.
Safety.related systems and components are 3.4.1 Flood Protecilon Hood protected either because of their location above the design Good level or because they are This section discusses the flood protection enclosed in reinforced concrete Seismic Category incasures that are applicable to the standard ABWR 1 structores which bave tbc following plant Seismic Category I structuies, systems, and requirementst components for both external flooding and R
postulated flooding from plant component (1) wall thicknesses be'ow flood level of not failures. These protection measures also apply less than two feet; l
to other structures that house systems and components important to safety which fall within (2) water stops provided in all construction the scope of plant specific, joints below flood level; 3.4.1.1 Mood !'rotection Measurts for Selsmic (3) watertight doors and equipment batches Category l Structures installed below design flood level; and m
The safety rclated systems and components of (4) waterproof coating of external surfaces.
j the ABWR Standard Plant are located in the g
reactor, control, and radwaste buildings which Waterproofing of foundations and walls of are Seismic Category.I structures. These seismic Category I structures below grade is g
structures together with those identified in accomplished principally by the use of water Table 3.41 are protected against external Good stops at expansion and construction joints. in damage. Flood protection of safety.related addition to water stops, waterproofing of the 3
systems _ and components is provided for all plant structures that house safety reJated systems and components is provided up to 8 Am (3 6
g;l postulated design flood levels and conditions -
g detcribed in Table 2.01. Postulated Gooding in) above the plant ground level to protect the from component failures in the building compart.
external surfaces from exposure to water, ments does not adversely affect plant safety nor does it represent any hazard to the public.
Additional specific provisions for flood protection include administrative procedures to Structures which house the safety.related assure that all watertight doors and batch covers are locked in the event of a flood -
gl equipment and offer flood protection are identified in Table 3.41. Descriptions of these warning. If local seepage occurs through the g
structures are provided in Subsection 3.8.4 and walls,it is controlled by sumps and sump pumps.
3.8.5.
Exterior or access openings and in the event of a flood, flood levels take a l
l penetrations that are below the design flood level are identified in Table 6.2 9.
relatively long time to develop. This allows Lf5) ecofs cu~a dass yad +o pr-eved fawb o$ \\o
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WcM r-m CAC c.o ccko6 n C C, 3,4 Amendment 16
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ABM 21ssiaaxe RIV D Standard Plant 3.4.1J.2 Compartment Mooding from Postulated After receiving a nood detection alarm, the i-Component Failures operstor has a ten minute grace period to act in cases when flooding can be identified and All piping, vessels, and heat exchangers with terminated by a remote action from the control room. In cases involving visualinspection to flooding potentialin the reactor building are seismically quallfled with one exception, and identify the specific flooding source in the complete failure of a non seismic tank or piping affected area (except ECCS areas) followed by a system is not applicable. The one exception is remote or local operator action, a minimum of 30 tbc radwaste building which contains no safe minutes is provided for the operator, p g, g-shutdown equipment, W.L2 a in allinstances of compartment nooding, a in accordance with Reference 2, leakage cracks single f ailure of an active component is
'3.4,(
are postulated in any point of moderate. energy considered for systems required to mitigate 52.
piping larger than nominal one ineb diameter. consequences of a particular flooding The leakage flow area is assumed to be a circular condition. The emergency core cooling system orifice with flow area equal to one half of the (ECCS) rooms are also evaluated on the basis of pipe outside diameter multiplied by one half of a loss of. coolant accident (LOCA) and a single the pipe nominal wall thickness. Resulting active failure or a LOCA combined with a single leakage flow rates are approximated using passive failur: 10 minutes or more after the Equation 3 2 from Reference 1 with a flow LOCA.
coefficient of 0.59 and a normal operating pressure in the pipe.
There are no interface requirements made upon tbc remainder of the plant from possible The dynamic effects of postulated high energy flooding in the ABWR Standard Plant buildings.
line breaks in the MSL tunnel area including Other lines, such as storm drains and normal flood analysis are excluded in the evaluation, waste lines, interface with plant yard piping.
assuming credit for detection of leaks before a However, provisions are n ade in these lines line breaks with a good accuracy and reliability that, should the yard piping become plugged, to permit shutdown and repair. The MSL tunnel crushed, or otherwise inoperable, they will vent area is instrumented with radiation and air onto the ground relieving any flooded condition.
temperature monitors that are used to automatically isolate the MSL isolation valves Conside.ing the above criteria and assump.
upon detection of high abnormal limits.
tions, analyses of piping failures and their consequences are performed to demonstrate the y,g g
a However,in the event of worst case flooding adequacy of the ABWR design. These analyses
,3 involving a feedwater line break, the maximum are provided separately for the reactor and flow rate from this high energy line break will control buildings.
not exceed 3.6 cubic meters per minute (950 gpm) over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period. Refer to Table 15.616 for 3.4.1.1.2.1 Evaluation of Reactor Building K (NWFJr feedwater line leakage parameters. Water Mood Events 3 A._1.1.2. C discharged from a postulated feedwater lire break will be contained in the Seismi: Category 2 Analysis of potential flooding within the "5 A.i structure of the MSL tunnel area and will not reactor building is considered on a floor.by.
q flood any safety related equipment in the reactor Door basis, building. The flooded area will be allowed to drain through the floor drains in the tunnel area 3.4.1.1.2.1.1 Evaluation of Moor 100 (B3D which are routed to the HCW sumps in the reactor building for collection and discharge.
Worst case flooding on this floor level would result from leakage of the RHR 18' suction line No credit is taken for operation of the drain between the containment wall and the system iso.
sump pumps although they are expected to operate lation valve (this applies also to the HPCF.
during some of the postulated floodmg events.
RCIC, and SPCU suction lines, altbough in 362 Amendment 16
.=
ABWR meu pry n Standard Plant dynamic for< c due to flood. The lateral hydrostatic pressure on the structures due to the design flood water level, as well as ground water and soil pressures, are calculated.
l Structures, systems, and components in the l
ABWR Standard Nuclear Island designed and l
analyzed for the maximum hydrostatic and hydrodynamic forces in accordance with the loads i
and load combinations indicated in Subsection 3.8.4.3 and 3.8.5.3 using well established methods based on the general principles of engineering tr.echanics. All Seismic Category I structurer are in stable condition due to either moment or uplift forces which result from the proper load combinations including tbc design basis flood.
3.4.3 In!Cliaces l
l 3.43.1 flood Eletallon l
l The design basis flood elevation for the ABWR l
Standard Plant structures is one foot below l
l
- grade, 3.43.2 Ground Water Elevation l
The design basis ground wa.er elevation for l
the ABWR Standard Plant structures is two feet I
below grade.
l 3.4.4 References
\\
l 1.
Crane Co., Flow of Fluids Through l'alves, Fittings, and Pipe. Technical Paper No.
410, 1973.
2.
nNSI/ANS 56.11, Standard, Design Criteria for Protection Against the Effects of Compartment Flooding in Light Water Reactor Plants.
l 3.
Regulatory Guide 1.59, Rev. 2 Design Basis Floods for Nuclear Poser Plants.
IN SG9 T 3,4, g 3.4.3.3 2
t N S G P.T
( '5, 4. 3. 4 3 ' 4' I 54 1
Amendment 6 i
4 INSERTS 3.4.1.1.2 INSERT 3.4.1.1.2a 1 4.l' l
wt All leak-before-break analysis will use plant-specific data such l
as piping geometry, mator.als, fabricatdon procedures, and pipe support locations.
See Subeection 3.4.3 for interface requirements.
INSERT 3.4.1.1.2b
- \\
Analysis of the worst flooding due to pipe and tank failures and F3 their consequences are performed in this subsection for the reactor building, control building, radvaste building and the service building. No credit is taken for safety-related equipment within these structures if the equipment becomes partially i
flooded.
However, in accordance with Section 3.11, all safety-related equipment is qualified to high relative humidity.
INSERT 3.4.1.1.2c For those structures outside the scope of the ABWR Standard Plant 14*I (e.g.,
the ultimate heat sink pump nouse), the applicant D4 referencing the ABWR design will demonstrate the structures j
outside the scope will meet the requirements of GDC 2 and the l
guidance of RG 1.102.
See subsection 3.3.3 for interface l
requirements.
INSERT 3.4.3 3 4.\\
3.4.3 Leak-Before-Break Analysis Leak-before break analysis will be submitted to the HRC using plant-specific data such as
,wd ping geometry, materials, fabrication procedures, and b rport locations.
Any piping qualifying for the leak-before-break approach will meet the requirements of Subsection 3.6.3.
(See Suceection 3.4.1.1.2)
INSERT 3.4.4 Flood Protection Requirements for Other Structures The applicant referencing the ABWR design will demonstrate, for
\\
4 the structures outside the scope of the ABWR Standarc Plant, that they meet the requirements of GDC 2 and the guidance of RG 1.102.
l (See Subsection 3.4.1.1.2) l l
l
ABWR imimu Standard Plant wn 9.1.2 Spent Fuel Storage The spent fuel poolis a reinforced concrete structure with a stainless steellicer, The bottoms of 9.1.2.1 Design !!ases all pool gates are sufficiently high to maintain the water level over the spent fuel storage racks form 91.2.1.1 Nuc! car Design adequate shielding and cooling. All pool fill and drain lines enter the pool above the safe shielding water (1) A full array in the loaded spect fuel rack is level. Redundant anti siphon vacuum breakers are designed to be suberitiest, by at least 59, Vk, located at the high point of the pool circulation lines Neutron absorbing material, as an integral part to preclude a pipe break from siphoning the water of the design,is employed to assure that the from the pool and jeopardizing the tafe wa'er level.
calculated k including biases and uncertainties, wilbn,ot exceed 0.95% under all The racks include individual solid tube storage normal and abnormal conditions.
compartments, which provide lateral restraint, over the entire length of the fuel assembly or bundle. The (a) Monte Carlo techniques are employed in weight of the fuel assembly or bundle is supported the calculations performed to assure that axially by the rack fuel support. Lead in guides at the k
does not exceed 0.95 under all normal top of the storage spaces provide guidance of the fuel adabnormal conditions.
during insertion.
(b) The assumption is made that the storage The racks are fabricated from materials used for array is infinite in all directions. Since no construction are specified in accordance with the credit is taken for neutron leakage, the latest issue of applicable ASTM specifications. The values reported as effective neutron racks are constructed in accordance with a quality multiplication factors are, in reality, assurance prograrn that ensures the design, infinite neutron multiplication factors.
construction and testing requirements are met.
(c) The biases between the calculated results The racks are designed to withstand, while and experimental results, as well as the maintaining the nuclear safety design basis, the uncertainty involved in the calculations, are impact force generated by the vertical free fall drop taken into account as part of the of a fuel assembly from a I.cight of 6 feet. The rack is calculational procedure to assure that the designed to withstand a pullup force of 4000 pounds specific k limit is met.
and a horiz ntal force of 1000 pounds. There are no g
readily definable horizontal forces in excess of 1000 9.1.2.1.2 Storage Design pounds, and in the event a fuel assembly should jam, the maximum lifting force of the fuelhandling The fuel storag racks provided in the spent fuel platform grapple (a.sumes limit switches fail) is 3000 storage pnot provide storage for 2707o of one full
- pounds, core fue! load.
The fuel storage racks are designed to handle 9.1.2.1.3 Mechanical and Structural Design irradiated fuel assemblies. The expected radiation levels are well below the design levels.
The spent fuel storage racks in the reactor building contain storage space for fuel assemblies (with In accordance with Regulatory Gtide 1.29, the fuel channels) or bundles (without channels). They are storage racks are designated Safety class 2 and designed to withstand all credible static and seismic Seismic Category 1. The structuralintegrity of the loadings. The racks are designed to protect the fuct rack has been demonstrated for the load assemblies and bundies from excessive physical combinations described below using linear clastic damage which may cause the release of radioactive design methods, materials in excess of 10CFR20 and 10CFR100 requirements, under normal and abnormal The applied loaJs to the rack are:
conditions caused by impacting from either fuel assemblies, bundles or other equipment.
(1) dead loads, w.iich are weiglit of rack and fuel assemblies, and hydrostatic loads; b " '~
The U shy Poo5 m
Amendmcni 11 C\\ o s\\T s cOow tS 9, \\, 2 S eU vm c.
Pv em cA s-,,1 m rom e L 2 - 1,
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ABWR mum Standard Plant prv n (2) surface dirt dislodged from equipment control room and a local panel. Pump low suc-immersed in the pool; tion pre.asure automatically turns off the pumps. A pump low discharge pressure alarm is (3) crud and fission products emanating from the indicated iu the control room and on the local reactor or fuel bundles during refueling; panel. The circulating pump motors can be powered from the diesel generators if normal (4) debris from inspection or disp' sal opera.
power is not available. Circulating pump motor tions* and loads are considered nonessentialloads and will be operated as requiret, under accident (5) residual cleaning chemicals or flush water, eonditions.
A post strainer in the effluent stream of the The water levelin the spent fuel storage filter demineralizer limits the migration of poolis maintained at a height which is suffi-filter material. The filter holding element can cient to provide shielding for normal building withstand a differential pressure greater than occupancy. Radioactive particulates removed the developed pump head for the system.
from the fuel pool are collected in filter de-mineralizer units which are locried in shleided The filter demineralizer units are located cells. For these reasons, the exposure of plant separately in shielded cells with enough cicar-personnel to radiation from the FPC system is ance to perrait removing filter elements from the minimal. Further details of radiological vessels.
considerations for this system are described in Chapter 12.
Each cell contains only the filter deminera-lizer and piping. All valves (inlet, outlet, The circulation patterns within the reactor recycle, vent, drain, etc.) are located on the well and spent fuel storage pool are established outside of one shielding wall of the roorn, by placing the diffusers and skimmers so that together with necessary piping and headers, particles dislodged during refueling operations instrument elements and controls. Penetrations are swept away from the work area and out of the through shielding walls are located so as not to pools.
cornpromise radiation shielding requirements.
Check valves prevent the pool from siphoning The filter demineralizers are controlled from in the event of a pipe rupture, a local panel. A differential pressure and conductivity instruments pravided for each Heat from pool evaporation is handled by the filier detaineralizer unit indicate when backwash building ventilation system. Makeup water is is required. Suitable alarms, differential provided through a remote operated valve.
pressure indicators and flow indicators monitor the condition of the filter demineralizers.
9.1.3.3 Safety haluation System instrumentation is provided for both The maximum possible beat load is the decay automatic and remote. manual operations. A low-heat of the full core load of fuel at the cod of low level switch stops the circulating pumps when the fuel cycle plus the remaining decay heat of the fuel pool drain tank reserve capacity is the spent fuel discharged at previous refuel-reduced to the volume that can be pumped in ings; the maximum capacity of the spent fuel l approx.imately one minute with one pump at rated storage poolis 270% of a core. The temperature of the fuel pool water may be permitted to rise l 3
capacity (250 m /hr), A level switch is to approximately 140"F under these condi-provided in the fuel pool to alarm on high and low level. A temperature element is provided to tions. During cold shutdown conditions, if if display pool temperature in the main control appears that the fuel pool temperature will creeed 125'F, the operator can connect the l
- room, FPC system to the RHR system. Combining the ca-The cir ulating pumps are controlled from the pacities enables the two systems to keep the S. t ?-
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ABM zwimui Standard Plant REV D water temperature below 125 F. The RHR system will be used only to supplenient the fuel pool A makeup water system and pool water level cooling when the reactor is shut down. The instrumentation are provided to replace reactor will not be started up whenever portions evaporative and leakage losses. Makeup water of the RHR systems are needed to cool the fuel during normal operation will be supplied from pool. The connectinir piping from the fuel condensate. The suppression pool cleanup system storage pool to the RHR system is designed can be used as a source of makeup water in case Seismic Category I and can be isolated, assuming of failure of the normal makeup water system.
a single actise failure, from the remainder of the fuel pool system.
Connections from the RHR system to the FPC system provide a Scismic Category 1, These connections may also be utilized during safety.related makeup capability to the spent emergency conditions to assure cooling of the fuel pool. The FFC system from the RHR spent fuel regardless of the availability of the connections to the spent fuel pool are Seismic fuel pool cooling system. The volume of water in Category 1, safety related.
the storage pool is such that there is enough heat absorption capability to allow sufficient From the foregoing analysis, it is concluded time for switching over to the RHR system for that the FPC system meets its design bases, emergency cooling.
9.1.3A Inspection and Testing Requirtments The 140*F temperature limit is set to assure l hat the fuel building environment does not No special tests are required because, t
exceed equipment environmental limits.
normally, one pump, one heat exchanger and one filter.demineralizer are operating while fuelis The spent fuel storage pool is designed so stored in the pool. The spare unit is operated that no single failure of structures or equipment periodically to handle abnormal heat loads or to will cause inability to: (1) maintain irradiated replace a unit for servicing. Routine visual fuel submerged in water; (2) re establish normal inspection of the system components,instrumen.
fuel pool water level; or (3) remove decay beat tation and tr. able alarms is adequate to verify from the pool. In order to limit the possibility system operability, of pool leakage around pool penetrations, the pool is lined with stainless steel. In addition 9.12.5 Radlological Considerations to providing a high degree of integrity, the lining is designed to withstand abuse that might The water level in the spent fuel storage occur when equipment is moved about. No inlets, pool is maintained at a height which is suffi.
outlets or drains are provided that might permit cient to provide shiciding for normal building l
the pool to be drained below a safe shielding occupancy. Radioactive particulates removed level. Lines extending below this level are from the fuel pool are collected in filter.
equipped with siphon breakers, check valves, or demineralizer units which are located in other suitable devices to prevent inadvertent shielded cells. For these nasons, the exposure l
pool drainage. Interconnected drainage paths are of plant personnel to radiation from the FPC l
provided behind the liner welds. These paths are system is minimal. Further details of l
designed to: (1) prevent pressure buildup behind radiological considerations for this and other l the liner plate; (2) prevent the uncontrolled systems are described in Chapters 11,12, and loss of contaminated pool water to other rela.
15.
tively cleaner locations within the containment or fuel bandling area; and (3) provide liner leak detection and menutement. These drainage paths are designed to permit free gravity drainage *R.
4 i? sing to the equipment drain tank;.5 ors M 3 oC a JE c.e d c a p me. h cu ) / o,-
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3 Amendment 16 9.15
ABWR m-n uv n StandarFinnt 9.2.12 HVAC Normal CoolingWater Division 1 isolation valve outside containment Syuem and Class 2 piping into the drywell. The return line penetration has divisionalisol6tbn valses 9.2.12.1 Design itases inside and outside containment. These valves are motor operated.
9.2.12.1.1 Power Generation Dolgn Bases ne ilVAC normal cooling water system (nonsafe-ty related) shall provide chilled water to the cooling coils of the drywell coolers, of each building supply unit and of local air condition.
ers to maintain design thermal ensironments dur-lag normal and upset conditions. The supply tem.
perature is 44.6 F. The return temperature is 53.6* F.
9.2.12<l.2 Safety Design Bases No diesel generator power is available to
%c IIVAC normal cooling water system does not this system during a LOPP or a LOCA.
perform any safety functions, except for the containment penetration and isolation valves.
9.2.12.3 Safety Esaluation 9.1.12.2 Sptem Descriptico Operation of the llVAC normal cooling water system is not required to assure the following
%c IIVAC normal cooling water system compo. conditions:
nests are listed in Table v.2 6 and shown in Figure 9.2 2.
(1) integrity of the reactor coolant pressure boundary; System components consist of five 25% chil-lers, each with pumps, serving a common chilled (2) capability to shut down the reactor and main-water distribution system connected to the chil.
tain it in a safe shutdown condition; and led water cooling coils in the drywell coolers, the cooling coils of each building supply unit (3) ability to prevent or mitigate the and cooling coils of local air conditioners, consequences of events which could result in Condenser cooling is from the turbine building potential offsite radiological exposures.
cooling water system Each chiller and pump set has either a three way mixing valve for automati-The llVAC normal cooling water system is not cally controlling the temperature of the chilled safety related llowever, it does incorporate water delivered or a flow control valve to main-features that assume reliable operation over the tain the desired temperature. Each chiller eva-full range of normal plant operations.
potator is designed, fabricated and certified in accordance with the AShtE Code $cetion Vill, Portions of the chilled water systern which Division 1. A chemical feed tank is provided. penetrate the containment and drywell are hiakeup water is from t e tveine-buHdingTooling provided with isolation valves and penetrations was+r4ystem surge tan hich receives water from which are Seismic Category 1, Safety class 2.
the hfUWP system, is tion valves and piping for The valves may be manually operated from the primary containment penetrations are designed to control room, except when a LOCA signal assumes seismic Catego y 1, AShtE code, Section lit, class control.
2, Quality Group B, Quality Assurance B requirements. The supply line penetration has a 9 2.M 4 M gh. b SbWP b t' dh b
81 Arcadmem 14 J
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ABWR mman uv ti Elt.ttdgrd Plant evaporator, if the temperature of the chilled beat eschanger, water drops below a specified lev;i, the controller automatically adjusts the position of 9.2.14.2 Sptem Description the compressor inlet guide vanes. Flow switches prohibit the chiller from operating unless there 9.2.14.2.1 Genera! Description is water flow through both evaporator and Tbc TCW system is illustrated on Figure condenser.
9.2 6. The system is a single loop system and 9.2,14 Turbine Bullding CoolingWater Sptem consists of one surge tank, one chemical, o tank, two pumps with a capacity of 1
9.2.14.1 Design Bases 29, m each, two beat exchangers with best R
Y usef al capacity of 130 x 106 Blu/b each. !
9.2.14.1.1 Safety Design Bases (connected in parallel), and associated coolers, piping, valves, controls, and instrumentation.
The turbine building cooling water (TCW) Heat is removed from the TCW system and system serves no safety function and has no transferred to the non safety related turbine safety design basis, service water system (Subsection 9.2.16).
R There are no connections between the TCW A TCW system sample is periodically taken
{
system and any other safety related systems.
for analysis to assure that the water quality meets the chemical specifications.
9.2.14.1J Power Generation Design Bases 9.2.14.2.2 Component Description (1) The TCW system provides corrosion. inhibited, demineralized cooling water to all turbine Codes and standards applicable to the TCW island auxiliary equipment listed in Table system are listed in Table 3.21. Tbc system is 9.2 11.
designed in accordance with quality group D specifications.
(2) During power operation, the TCW system operates to preside a continuous supply of Tbc chemical addition tank is located in the cooling water, at a maximum temperature of turbine building in close proximity to the TCW 1050F, to the turbine island auxiliary system surge tank.
equipment, with a service water inlet temperature not exceeding 950F.
The TCW pumps ar: 100% capacity cach and are constant speed electric motor driven, borizontal (3) The TCW system is desigt.ed to permit the centrifugal pumps. Tbc three pumps are maintenance of any single active component connected in parallel with common suction and without interruption of the cooling discharge lines.
function.
The TCW beat exchangers are 100% capacity l (4) Makeup to the TCW system is designed to each and are designed to base the TCW water permit continuous system operation with circulated oc the shell side and the power cycle design failure leakage and to permit beat sink water circulated on tbc tube side, expeditious post. maintenance system refill.
The surface area is based on normal beat load.
4 (5) The TCW system is designed to base an The TGW~ surge tank
- an atmospheric carbon atmospheric surge tank located at the steel tank located at t e bigbest point in the y, g highest point in the system.
TCW system. The surge tank is provided with a 4
level control valve that controls makeup water (6) The TCW system is designed to have a higher addition.
g pressure than tbc power cycle beat sink water to ensure leakage is from the TCW Those parts of the TCW system in the turbine f'
system to the power cycle beat sink in tbc building are located in areas that do not y
esent a tube leak occurs in the TCW system contain any safety.ietated systems. All 9. 2.1 4 @'d [5 yk%0 h >cq JL p pcw out Tcw sp %,g
- 1
9.7 14 IusGRT 9.7.14.1.1
- s The surge tank is located above the TCW pumps and heat exchangers in the turbine building in a location away from any safety-related components.
Failure of the surge tank will not affect any safety-related systems.
ABWR mstmei MV B Standard Plant safety related systems in the turbine building are located in special areas to prevent any
=
E damage from non safety related systems during seismic events. Those parts of the TCW system outside the turbine building are located away from any safety related systems.
9.2.14.2.3 System Operation During normal power operation, one of the two)67c capacity TCW system pumps circulate LOO 9.7.t 4
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9 2'IO I Arneectnent 16 1
ABWR m si -
Standard Plant P1V D inhibited d,pmineraliced watsgibrough the shell systems are preoperationally tested in u
bM side of }wtf Ntne thie850% i$acity TCW heat accordance with the requirements of Chapter 14
- 1 exchangers in service. The heat from the TCW estem is rejected to the turbine service water The :oruponents of the TCW system and
=
system which circulsas water on the tube side of associated instrumentation are accessible during the TOW system heat eschangers, piant operation for visual examination.
Periodic inspections during normal operation are The standby TCW system purnp is automatically made to ensure operability and integrity of the started oa detection of low TCv/ system pump system. Inspections include measurements of discharge pressure. The standb~y TCW system heat cooling water flows, temperaturcs, pressure.,
exchanger is placed rn service manually.
water quality, corrosion erosion rate, control positions, and set points to verify the system The cooling watee fic x rate to Oc condition.
electro hydrauile control (El-lC) c^s es, the turbine luhe oil coolers sno Wrecoltrs, and 9.2.14.5 Instrumentation Application generr. tor exeiter air cooler i. regulated by control valves. Control valves in the cooling Pressure n.nd temperature indientors are water outlet from these units are throttled in prtaided where required for testing and response to temperature signalt from the fluid balancing the system. Flow indicator taps are being cooled.
provided at strategic points in the sysam fer initial balancing of the flows and verifying The flow rate of cooling water to all of the flows during piant operation.
other coolers is manually regulated by individual throttling valves located on the cooling water Surge tank high and low level and TCW pump outlet from each unit.
discharge pressure alarms are retransmitted to the raain control room from the TCW loul control The minimum system cooling water temperature pa icts, is msintained by adjusting the TCW system heat exchanger typass salve.
Makeup flow to the TCW syvem surge tank is initiated automatically oy low surge tank water The surge tank provides a reservoit for level and is continued until the normal levet is small amounts of leakage trom the sysien aN to reestablished.
the expansion and contractinn o' (19 v fluid with changes in the system tempe smt Provisions for taking.'CW system water is conre.cted to the pumi. suction.
samples are included.
DerLneralized makeup water to tht G 'l system is controlled automatically by a vel control 9.2.15 Reactor Scrvice Water System valvt which is actusted by sensing surge tank level. A corrosion inhib'.ter is manually added 9.2.15.1 Design Bases to the system.
9.2.15.1.1 Safety Design Bases 9.2.143 Safety Evaluation (1) The reactor service water (RSW) system The TCW system has r.o safety design baste shall be designed to remove heat from and serves no safety function.
the reactor cooling water sptem which c
is required for safe reactor shutdown, 9.2.14.4 Tests and inspections and which also cools those auxiliaries whose operatica is desired following a All major components are tested and LOCA, but not essenti.1 to safe
(
in,<pected as separate components prior to shutdown.
installation, and as an integrated system after installation to ensure design performance. The (2) The RSW system shall be designed to 92l',
Amendment tt l
l' ABWR 2= = i uv n Standard Plant Seismic Category I and ASME Code, (1) flooding, spraying or steam release due Section 111, Class 3, Quality Assurance to pipe rupture or equipment failure; B, Quality Groep C, IEEE 279 and IEEE 30F requirements.
(2) pipe whip a~d jet forces resulting from postulated ppe rupture of nearby high (3) The RSW system snan oc p&ccted from energy pipes; ilooding, spr ~ing, steam impingement, pipe whip,.. forces, missiles, fire (3 missiles which result from eqvipment and the eflect of f ailure of any failure; and non Scismic Category I equipment, ag required.
(4) fire.
(4) The RSW system shall be designed to meet Liquid radiat'on monitors are provided in the the foregoing design bases during a loss RCW system. Upon detection of radiation leakage of preferred pows in a division cf the RCW system, that system is isolated by operator action from the control 9.2.15.1.2 Power Generatfor.csign Bases room, and the cooling load is met by another division of the RCW system. Consequently, The RSW system shall be de ign-d to cool the radioactive contamination released by the RSW reactor building cooling water @CW) as required system to the envircomer t does not exceed during: (a) normal operation; (b) emergency allowable limits defined by 10CFR100.
shutdown; (c) normal shutdown; and (d) testing, t',ystem low point drains and high point vents 9.2.15.2 Systa Desu'ption are provided as required.
The RSW systev' provides cooling water during System components and piping materials are various operatics. odes, d iring shutdown and selected to be compatible with the available rost LOCA operations. The system removes beat site cooling water in order to minimize from the RCW system and transfers it to the corrosio. Adequate corrosion safety factors ultimate heat sink. Figure 9.2 7 snows the RSW are used to assure the integrity of the sy; tem system diagiam.
during the life of the plant.
The RSW system is able to function during During all plant operatic 6 modes each abnormally high or low water levels and steps are division shall have at least one service water taba to prevent organic fouling that may degrade pump operating. Therefore,if a LOCA occurs, system performance. These steps include trash the system is already in operation. If a loss racks and provisions for biocide treatment (where of offsite power occurs during a LOCA, the pumps discharge is allowed). Where discharp of momentarily stop until transfer to standby biocide. is not allowed, non biocide treatment diesel generator power is completed. The pumps will be pro.ided. Thermal backwa.hing capability are ru. tarted autoniatically according to the will be provided at sea water sites where diesel loading sequence. No operator action is infestations of macrobial growth can occur, required, following a LOCA, to start the RSW system in its LOCA operating mode.
p 9.2.15.3 Safety Evaluation 9.2.15.4 Testhg and Inspection Requirements Th: components of the RSW system are 3
separated and protected to the extent necessary The RSW system is designed for periodic to assure that sufficient equipment remains pressure and functional testing to assure:
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operating to permit shutdown of the unit in the
/
cvent of any of the following (Separation is (1) the structural and leaktight integrity applied to electrical equipment and by visible inspection of the components; instrumentation and ccntrols at well as to 4g g,,
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sj mechanical equipment and p,iping.):
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. continuously during normal power operation conditions.
The standby pump is started automstically in the event the normally operating pump trips or the discharge header pressure drops below a preset limit.
9.2.16.3 Safety Evah. Jon The TSW system does not serve or support any Th <
safet function and has nf safety design bases.f W'dk a
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9.l:.16.4 hests and Inspectious dw y '3et iti d<m5. S44-N b 5 * #" ^ 9. 2. W. 5 0
- f., m All major components are tested and inspected as seperste components prior to installation, and as an integrated system after installetion to ensure design performance. Th'c systems are preoperationally tested in accordance with the requirements of Chapter 14.
The components *.,f the TSW system and associated instrumniation are accessible during plant operation for visual examination. Periodic inspections duilog normal operation are made to en+ure operability and integrity of the system.
laspections include measurement of the TSW system flow, temperatures, pressures, differential pressures and valve positions to verify the system condition.
9.2.16J Instrumentation Application Pressu.c and temperature indicators are provided where reg'uired for testing the system.
TSW system pump status is indicated in the main control room.
TSW systern trip is alarmed and the automatic startup of the standby pump is annunciated in the main control room.
High differential pressure across the duplex n
l-filters is alarmeo in the main control room.
l i
l 9.2122 Amendment 11
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23A6t00All prv s Standard Plant -
9.2.17 Interfaces 9.2.17.1 Ultimate Heat Sink Capability.
Interface requirements pertaining to ultimate
~#
heat sink capability are delineated in Subsection 9.2.5 as follows:
Submtion Illig 9.2.5.1 Safety Design Bases J
9.2.5.2 Power Generation Design Bases R
9.2,5.6 Evaluation of UHS Performance f'.2.5.7 Safety Evaluation 9.2.5.8 Conformance to Regulato>y Guide 1.27 9.2.5.9 Instrumentation and Alarms 9.2.5.10 Tests and Inspections 9.2.17.2 51aleup Water System Capabillty The raw water treatment and preparation of the demineralized water is sent to the makeup water system (purified) described in Subsection 9.2.10.
g,q, g
cy, p, y u
The makeup water preparation system shall be located in a building which does not contain any safety related structures, systems or components. If the system is not available.
- E demineralized water can be obtained frocs mobile equipment. The system shall be designed so that.
any. failure in the system, including any that cause flooding, shall not result in the failure of any safety related structure. system or component.
9.2.17J Potable and Sanitary Water System The potable and sani;ary water system shall l
be designed with ro interconnections with systems E
having the potential for containing radioactive materials. Protection shall be provided through the use of air ' gaps, where necessary. (Sce Subsection 9.2.4).
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- 9. 2. i"T. A l9 2 V15 I
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Amendment 16 l
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9.2,5 IwSERT 9, 2. t '7, '2.
If any spray pond piping is made from fiberglass-reinforced thornosetting resin, the applicant shall provide information to show that all applicable requirements of Regulatory Guide 1.72 are met.
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4 mSGF.T 9. '2. MA / 9 '2.G I 9.2.17.4 Reactor Service Water System described in tablo 9.2-13.
The applicant ((j The RSW punk e
shall provide a following 4dditional information which is site dependent. ( Sa
- S b s *. A ~ 9. 7. I s. t a r d
- 9. 2, g t. 3 )
(1) temperature increase and pressure drop across the heat exchangers (2) the required and available not positivo suction head for the RSW pumps at pump suction locations considering anticipated low water levels (3) the location of the RSW pump house (4) the design features to assure that the requirements in section 9.2.15.1.1(3) are met (5) an analysis of a pipeline break and a single active component failure shall show that flooding shall not affect the main control room or more than one division of the RSW system.
9.2.17.5 Turbine Service Water System 3 The applicant shall demonstrate that all safety-related com-9' 2' 4 ponents, systems, and structures are protected from flooding
- in the event of a pipeline break in the TSW system. ( S44 s # b r a c+s o w 9.'?.t6.1) i
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Table 9.2-t 3 Reactor Service Water System RSW Pumps (two per division)
Discharge Flow Rate 7,920 gpm Pump Total Head 50 psi Design Pressure 115 psi Design Temperature 122 F RSW Piping and Valves Design Pressure 115 psi 122 F Design Temperature CL O' " "
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ABWR ummui S9mndard Plant nn
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Figure 9.24a TURBINE COOLING WATER SYSTEM DIA(,..,.M 1
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