LD-93-012, Forwards Draft Sys 80+ Certified Design Descriptions & Associated ITAAC
| ML20128G262 | |
| Person / Time | |
|---|---|
| Site: | 05200002 |
| Issue date: | 01/28/1993 |
| From: | Brinkman C ABB ATOM, INC. (FORMERLY ASEA ATOM, INC.), ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LD-93-012, LD-93-12, NUDOCS 9302120222 | |
| Download: ML20128G262 (128) | |
Text
_ _ _ _ - _ _ _ _ - - _ - _ - _ _ _
ABB ALE A BROWN DQvf Rt January 28, 1993 LD-93-012 Docket 52-002 Attn Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
System 80+ Draft ITAAC Submittal Dear Sirst This letter transmits draft System 80+ Certified Design Descriptions and associated ITAAC (Inspections, Tests, Analyses and Acceptance criteria) for selected systems.
.-A larger package comprising the remainder of the initial submittal will be transmitted on or about February 1, 1993.
The packages for each system contains the following items, as
.applicables certified Design Description Text System Diagram ITAAC 'lable Design Commitment Inspection, Test or Analysis Acceptance criteria Supplementary Material Relevant Safety Analyses Assumptions Relevant PRA Assumptions CESSAR-DC Chapter 14: Test Description References Amplifying Information Including CESSAR-DC Section References The enclosed certified Design Descriptions and.ITAAC reflect guidance developed-in industry /NRC meetings.during 1992.
Future submittals.will update these-packages to: incorporate the
- approaches for addressing-programmatic issues' developed'during.
the-January 1993 industry /NRC review of.the lead'. plant ITAAC~and-to reflect-changes resulting from the industry review of System 80+ IT Q,$ 19h Will be conducted during the_ period: February 1st
-throu@
ti;;tthd ABB Combustion Engineering Nuclear: Power m_.
ComtutWA D$rersy bc
.1030 P94*c! HM had
' feicptume (20316MB 1919 ~
/
kDbh hob!O2:
D A --_
U.S. Nuclear Regulatory Commission LD-93-012 January 28, 1993 page 2 9-CE has attempted to maintain consistent wording betwoon tho
, LAC design commitments and the Cortified Design Description.
.rther, similar phraseology has been applied in various ITAAC when addressing tho samo issuo (e.g. configuration chocks, electrical power sourcos, etc.).
Slight differences in wording l
should not be interpreted as an intont to convoy different meanings.
Should you hava questions on tho enclosed matorial, please contact me or Mr. John Roc (203-285-2861) or Mr. George lions (203-285-5218).
Very truly yours, COMBUSTION ENGINEERING, INC.
4
&~
C.
B.
Brinkman Acting Director Nuclear Systems Licensing CDD/gdh cc R. Dorchardt (NRC)
T.
Boyce (NRC)
A. Iloymer (NUMARC)
J. Trottor (EPRI)
T.
Wambach (NRC)
SYhTEM 80+'
l.3.6 CORE SUPPORT AND REACTOR YESSEL INTERNAL STRUCTURES Design Description The Reactor Vessel Core Support Structures are safety related systems consisting of the Core Support Barrel assembly and Upper Guide Structure anemblics.The core support structures support the fuel nuemblies and provide a Dow path within the Reactor Pressure Vessel.
Reactor Vessel Internal Structures are all structures within the reactor preuure venci except the Core Support Structures, fuel, control element nssemblics and instrumentation.
The Core Support Barrel (CSil) assembly is suspended from the reactor vessel Gange.
He CSB anembly provides support and location positioning for the fuel nssembly lower end fittings. He CSB anembly contains internal structures that provide an instrumentation guide path from the lower vessel and hydraulic flow paths through the vessel from the inlet nor21cs to the upper end of the fuel assemblics.
He Upper Guide Structure (UGS) assembly is supported from the CSB upper Dange and extends into the CSB assembly to engage the top of the fuel assemblics. The UGS assembly provides an insertion path for the control element assemblics. The UGS assembly contains internal structures which provide a guide path and lateral support for the upper portion of the control element assemblics and extension shafts in the reactor vessel upper plenum region. nc UGS assembly also provides guide paths for heated junction thermocouple assemblics.
A general conceptual illustration of both structures is shown in Figure 1,3.6-1.
We Core Support Barrel and Upper Guide Structure assemblies are fabricated in accordance with ASME Code Class NF requirements and the Seismic Category I classification.
He Reactor Vessel Core support structures and internal structures withstand the effects of flow induced vibration.
Inspections, Tests, Analyses ant! Acceptance Criteria Table 1.3.61 specifies the inspections, tests, analyses and associated acceptance criteria for the Core Support and Reactor Vessel Internal Structures.
1.3.6
-1 01 28 93 I
e-
+
-7 t
J i.
i SYSTEM 80+
TABLE 13.6-1 5
CORE SUPPORT & REACIlOR VESSEL INTERNAL STRUCIURES AND CONTROL ELEMEPft' DRIVE MECHANISMS Inspections. Tests. Amalvses. and Aue., ance Criteria l
4 Certised Desian Ch I=-# n Tests. Analyses Accestmece Criseria l
L
'A basic configuration of the Reae-1.
Inspection ' of the as-built Reactor 1.
The as4eik c4-ation of the i
i ter Vessel Core Support Structures Vessel Core Support Structures will Reactor Vessel Core Support i
. is shewn in Figure 134 be performed.
Structures is in accordance with Figure 13fri for the components J
and equipment shown.
l-
' 2.
The Reactor Vessel Core Support 2.
Tests will be performed to subject 2.
He reactor vessel core ' support
[
Structures and internal structures the Reactor Vessel Core Support structures have no visible signs of i
j; withstand the, effects of flow Structure to flow induced damage, loose parts, or nmuve t
j
.. induced. vibration. l vibration. Visual inspection mi!I be wear.
- performed on the Reactor Vessel Core support Structure.
i 3
J
+
1-3 6
-2 01-28-93 e
s
?
t
..v
<m.#
L
,.-4
-.c,
.y r-w.-
.vn.
,m.,-#
,...,x.
,,c.1.,...m
.--,-.-,..-.,e
-r e
f i
TM naTc curoE PATn SYSTEM 80 +
g l
l r'
e F,\\
FN i
r' (
5 i
e a
g s
I 5;
e UPPER GUIDE
- i
\\ cEA GUIDE PATil
", o" STRUCTURE e
e I
ASSEMDLY l
a g
mi u
1
- u,y
' d ]
l!l l l
'/ '
i l
l l
l i
l i
l,.lff].,H..lI.N.. IlIOIE..E,l)..I.II.III).NINI.k
', FUEL ASSEMBLY UPPER SUPPORT-n CORE SUPPORT _
l
~
DARREL ASSEMDLY l
-l 9
l j
FUEL ASSEMBLY LOWER SUPPORT
]
N
}) IlYDRAULIc FLOW PATil
- s. *
{
r h=p=p=tryr erpri ( /
1
'j 'p ;=;.='<j Lj{j= '..
p
.a>
.(yy,
I INSTRUMENT GUIDE PATH FIGURE 1.3.6-1 REACTOR VESSEL-CORE SUPPORTSTRUCTURES
~
i
, -+
SYSTEM 80+"
1.3.6 CORE SUPPORT AND REACTOR VESSELS INTERNALS STRUCTURESITAAC SUPPORTIVE INFORMATION 1.
Amnlifying Information ne supporting material would include a description of the CVAP Program which demonstrates comp!!ance with Regulatory Guide 1.20 for a non. prototype Category I program.
See CESSAR DC Sections 3.9.2.4 and 3.9.3 2.
Relationshin of CS and RVIS ITAAC to the Safety Anahsis None 3.
Relationshin of CS and RVIS ITAAC to PRA None 4.
CESSAR DC Chapter 14 Tests Annilenble to CS and RVIS ITAAC Nonc l-
+
l l
1.9.1.2 1
01 28 93 l
l I
_SYS1EM 80+*"
1.4.1 NUCLEAR DESIGN Design Description
- Ihe nuclear design of the System 80+ reactor core is not within the scope of the certified design. The specific nuclear design that will be utilized in a facility which has referenced the System 80+ design will comply with nuclear design criteria as specified in 10 CITt 50.
1hc nuclear design of the reactor core for a referenced System 80+ plant meets the following criteria:
In the power operating range, the net prompt reactivity feedback (fuel temperature coefficient, moderator temperature cocilicient and moderator pressure cocflicient)is negative.
j 1hc values of the fuel temperature, moderator temperature and moderator picssure 1
cocllicients of reactivity are consistent with analyses that predict acceptable consequences for gmstulated accidents and anticipated operational occurrences.
The burnable poison loadings and reactivity worths are included in the plant core analyses.
1he reactor core and associated coolant, control and protection systems ensure that xenon induced power distribution oscillations do not cause the specilled acceptable fuel design limits (SAFDI2) to be exceeded.
1he potential amount and rate of reactivity insertion under normal operation and postulated reactivity accidents do not result in violation of the specified acceptable fuel design limits ',SAFDI2), damage to the scactor coolant pressure boundary (RCPB), or disruption of the core or other reactor internals which impairs the effectiveness of safety injection.
The core power distribution and power peaking allow full power operation for the design cycle length and do not result in violation of the specified fuel design limits (SAFDLs) for postulated accidents.
The amount of reacthity svallable from insertion of withdrawn CEAs meets the excess CEA worth requirement for power operating conditions.
1.4.1 01-28-93
i SYS'IEM 30t"
]
1.4.1 NUCLEAR DESIGN ITAAC i
SUPPORTIVE INFORMATION l.
Amplifying Information CESSAR.DC Section 4.3 2.
Rdelkuhin of NUCIEAR Dl?SION ITAAC to the Safety Analysis No ITAAC are provided because the nuclear design is not within the scope of the 1
certined design.
3.
Relatiomhin of NUCil!AILDESIGN ITAAC to PRA None 4.
CESSAR.DC Chanter 14 Tests Annlicable to NUCLEAR DESIGN ITAAC None applicable before fuel load 6
3 i
1.4.1 01-28-93
=,
- -,...._1
- .
KYEWh110i" 1.4.2 FUEL SYSTEM Design Description 1hc fuel system for System 80+ is not within the scope of the certified design. 'the specific fuel system that will be utilized in a facility which has referenced the System 80+ design will comply with fuel system design criteria as specified in 10 CFR 50.
1he fuel system includes fuct assemblies, fuel assembly components, fuel nxis, burnable poison components and control element assemblics. The fuel system for a referenced System 80+ plant meets the following criteria:
The fuel assembly and fuel assembly component (including the fuel rod and burnable poison rod) materials are compatible with the reactor environment.
Fuel system thermal-mechanical analyses are based on either worst tolerance assumptions or probabilistic analyses to determine statistically bounding results (i.e.,
upper 95% confidence)
The fuel assembly and fuel assembly component (excluding the fuel rod) stresses and cumulative fatigue damage factors do not exceed the limits for normal operation and design basis events.
The fuel rod and fuel assembly component analyses include consideraticn of metril thinning and associated temperature increases due to oxidation and the buildup of corrosion products to the extent that these intiuence the material properties and structural strength of the components.
1he fuel rod internal hydrogen content is controlled during manufacture of the fuel rod.
less of fuel rod mechanical integrity due to collapse of the fuel rod cladding is not predicted to occur during the design lifetim; of the fuel rod.
The fuel rod cladding stresses, strains end cumulative fatigue damage factors do not exceed the limits for normal operation and design basis events for which fuel damage does not occur.
Loss of fuel rod mechanical integrity due to excessive cladding pressure loading does not occur.
Ioss of fuel rod and fuel assembly component mechanical integrity due to fretting wear resulting from fuel rod and fuel assembly component vibration does not occur in an environment free of foreign material.
IA.2
-I-01 28-93
BWWM 80+"
he burnable poison rod cladding stresses and strains do not exceed the limits for normal operation and design basis events for which fuel damage does not occur.
Loss of burnable poison rod mechanical integrity due to excessive cladding pressure loadin3 oes not occur.
d
%c Control Element Assembly (CEA) materials are compatible with the reactor environment.
The Control Element Assemblics (CEAs) are capable of insertion into the core during all modes of plant operation within the limits assumed in the plant analyses.
Bowing or swelling of fuel rods does not result in obstruction of control element pathways which would restrict Control Element Assembly (CEA) movement.
The reactivity worth of the CEAs is included in the plant core analyses.
The Control Element Assembly (CEA) cladding stresses and strains do not exceed the limits for normal operation and design basis events.
1A.2
-2 01-28-93
SYSTEM 80+= -
1.4.2 FUEL SYSTEM ITAAC SUPPORTIVE INFORMATION 1.
Amnlifyine Information CESSAR DC Section 4.2 2.
Eglationship of NUCLEAR DESION ITAAC to the Safety Analysis None 3.
Relationship of NUCLEAR DESIGN ITAAC to PRA None 4.
CESSAR DC Chapter 14 Tests Annlicable to NUCLEAR DESTON ITAAC Nonc 3
3 i
1.4.2 - 01 28-93 a
c
--..,=
E.YSTEM 80+"
1.4.3 TilERMAL AND llYDRAULIC DESIGN REACTOR Design Description The thermal and hydraulic design of the Sptem 80+ reactor core is not within the scope of the certified design. The specific reactor thermal and hydraulic design that will be utilized in a facility which has referenced the System 80+ design will comply with thermal and hydraulic design criteria as specified in 10 CFR 50.
The thermal and hydraulic design of the reactor core for a referenced System 80+
plant meets the following criteria:
De minimum departure from nucleate boiling ratio (MDNI3R) does not occur on any fuel rod during normal operation and anticipated operational occurrences.
Flow instability within the fuel assembly flow channels does not occur during normal operation and anticipated operational occurrences.
The calculated peak temperature of the nuclear fuct in a fuel rod is less than the mciting temperature of the nuclear fuel during normal operation and anticipated operational occurrences.
The primary coolant flow rate with the four reactor coolant pumps in operation is greater than or equal to the design minimum and less than or equal to the design maximum reactor coolant system (RCS) flow rates.
1.4.3 01 38-93
,.m.-.
_m_..
SYSTEM 80+"
1A.3 TilERMAL AND ilYDRAULIC DESIGN REACTOR ITAAC SUPPORTIVE INFORMATION 1.
AIpplifying Information CESSAR.DC Section 4.4 2.
Relationship of NUCLEAR DESIGN ITAAC to the Safety Annivsis No ITAAC are provided because the thermal and hydraulle design is not within the scope of the certifled design.
3.
Relationshin of NUCLEAR DESIGN ITAAC to PRA i
None 4.
CESSAR.DC Chanier 14 Tests Applicable to NUCLEAR DESIGN ITAAC CESSAR.DC Section 14.2.12.1.57 IA.3 01 28 93
SYSTEM 80+"
1.5.1 REACTOR COOLANT SYSTEM DESIGN DESCRIPTION ne reactor coolant system (RCS) removes heat generated in the reactor core and transfers the heat to the steam generators, where feedwater is boiled to steam. %c RCS is located in the containment and is composed of a reactor vessel (RV), two stcam generators (SG), four reactor coolant pumps (RCP), one pressurizer (PZR),
four pressurizer safety valves, connecting piping, heaters and valves. %c pumps circulate reactor coolant water in parallel loops through the RV to the SO's and back to the RCP suct!ons. He PZR serves as a surge volume and provides overpressure to prevent boiling in the core region and RCS loops. %c reactor coolant system is a safety related system to the extent that it forms the pressure boundary between the reactor coolant and the containment atmosphere. Figure 1.5.11 shows a simplified system configuration.
nc RCS pressure boundary is constructed to Code Class 1 of ASME Code Section III. ASME Ccxic portions of the RCS retain their integrity under internal pressures that will be experienced during service. Components, piping and supports classified as ASME Ccxle Class 1 are Scismic Category I. Equipment that is designated as safety related is qualified for the external environments where kicated.
RCS instrumentation indications and alarms shown on Figure 1.5.1 1 are available in the control room. Controls are available in the control room to start and stop the RCPs, open and close the pressurizer spray control valves, and energize or de.
energize the pressurizer heaters.
He RCS is protected from overpressure by the PZR safety valves. Valve relief capacity is sufficient to limit RCS pressure to less than or equal to 110% of design pressure for all design batis events. PZR safety vahes of the type installed in the plant have been tested at full flow. Instrumentation is provided in the control room to indicate a not fully closed safety valve. He RCS is protected by relief valves in the shutdown cooling system (SCS) when the RCS is connected to the SCS.
Fracture toughness of RCS materials is controlled by the AS!wa Code. Delta ferrite content is controlled. The initial charpy upper shelf energy of RV beltline material is no less than a minimum requiied energy. Reference nil-ductility transition temperatures (RTsm) are derived from Charpy V notch tests of specimens taken from construction materials. De RTum so derived must be equal to or less than the required RTum. Controls are placed on residual chemical content in the reactor vessel beltline materials to limit the maximum predicted increase in RTum over the life of the plant.
The inner surface of the RV, in the active core region, is equipped with capsule holders for. accommodating material surveillance capsules. Specimens taken from 1.S.1 1-28-93 J
SYS'il?M 80+"
materials actually used in fabrication, as well as weldments typical of those used in the belt line region, are inserted in the holders before nuclear operation.1hc capsules contain Charpy V notch specimen of base metal, wcld metal and heat affected zonc material, and tensile specimen from base metal and wcld metal.
'Ihe reactor coolant pump motor includes a flywheel for extended coastdown pumping during loss of power. Fracture toughness is assured by controlling the nil-ductility transition temperature and Charpy absorbed energy. Each flywheel retains its integrity at a design overspeed condition of 125 percent of normal operating speed.
Inspections, Tests, Analyses and Acceptance Criteria Table 1.5.11 specines the inspections, tests, analyses, and associated acceptance criteria for the RCS.
h 1.5.1 1 28-93
SYSTEM 80+"
TABLE 1.5.1-1' REACTOR COOLANT SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Certi5et
- Desima Co--it--t I=s-#-
Tests. Amalvses Acceptance Criteria L
A basic configuration of the RCS is L
Visual inspections of the as-built L
De as-buik configuration of the shown in Figure. L5.1-L RCS configuration mill be con-RCS is in accordance with Figure ducted.
L5.1-1, for the components and equipment shown.
2.
ASME Code portions of the RCS 2.
A pressure test will be conducted 2.
%c resuks of the pressure test of retain their integrity under internal on those portions of the RCS re-ASME Code portions of the RCS pressures that will be experienced quired to be pressure tested by the conform with the i@ments in during service.
ASME Code.
the ASME Code Section IIL 3.
Ovupmure protection limits RCS 3.
Inspedion of the ASME Code-re-4 RCS Pim-6 is limited to @
pressure to less than or equal to quired O,upaure Prcsection Re-psia, for the design basis events 110% of design ; pressure for all
. port will be performed.
evaluated in the Ow p m-c design basis event.
Protection Report.
4.
Pressurizer safety valves of the 4.
' Inspection of the EPRI PWR Safety 4.
Type tnrhs of pw Lu safety.
type ~inoded ' in the~ plant. have -
- and Relief - Valve Test Program, valves installed in the plant has been tested at full flow, with the
- Report EPRI NP 2628-SR, and been accomplished and the resuhs inlet w4-. tion used in the vendor and inedsian records for acepted.
plant.
the safety : vahes, will be per-formed.
5.
Low temperature v,upsure pro-5.
See SCS ITAAC L5.2. -
5.
See SCS ITAAC L5.2.
tedion (LTOP) for the RCS is pro-
~
- vided by relief valves in the SCS.'
6..
Prodsions are made on the inner.
' 6.
Inspection. of the RV before 6.
Capsule holders are in place.
. surface of the RV for materit sur-closure L for presence of capsule veillance. sp,h
^
holders will be performed.
1.5.1 01-28-93
SYSTEM 80+"
TABLE 1.5.1-1 (Continued)
REACTOR COOLANT SYSTEM InSDections. Tests. Analyses and Acceptane Criteria Certified Desien Commitment Inspections. Tests. Analyses Acceptance Criteria 7.
RV material spedmens made from 7.
The process procedures and quality 7.
Surwdlance specimens are made the actual material from which the control records that trace fabri-from material used in RV fabri-vessel was constructed are inserted cation d surwillsace samples will cation, and are inserted in the in the capsule holders.
be reviewed, and inspection of the capsule boklers.
RV for presence of specimens will be performed.
8.
Each RCP motor flpheel retams 8.
Vendor test records will be in-8.
Each RCP !!pheel was tested and its integrity at 125% of normal spected.
passed a 125% overspeed test.
operating speed.
9.
RCP motor flywheel material has 9.
RCP material test reports will be 9.
For the RCP motor flywheels-sufficient ductility to prew nt inspected.
brittle fradure.
nil-d u ctilit y transition temperature 510*F Charpy absorbed energy 2 50 ft-Ib 10.
RCS. instrumentation indications 10.
Inspection of the control room for 10.
'the instrumentation 6(~b=
and alarms shown on Figure L5.1-the availability of instrumentation and alarms shown on Figure 1.5.1-1 are available in the Control indications and alarms identified in 1 are available in the Control Room. Controls are available to the Certified Design Commitment Room.
RCS controls operate as start and stop the RCPs, open' and will be performed.
Tests will be speedied in the Certified Design close the pressurizer spray vahrs,'
performed using the RCS controls Commitment.
and energize or de-energize the in the Control Room.
pressurizer heaters.
l IL Delta ferrite is controlled to within IL Material Test Reports and quality IL
. Weld rod and filler materiah specific limits.
control records will be inspected.
SFN to 15FN Clad-SFN to ISFN L
. rae6cc SFN to 30FN 1.5.1 01-28-93
SYSTEM 80+=
TABLE 1.5.1-1 (Continued)
REACTOR COOIANT SYSTEM Inspections. Tests. Analvscs. at.d Acceptance Criterix Certified Desien Commitment Inspections. Tests. Analyses Acceptance Criteria 1
12.
RCS materials are limited to a 12.
Material Test Reports and quality 12.
For RV belt line material, RTm is:
l maximum reference nil-ductility control records will be inspected.
forgings-s+1&F transition temperature RTm.
urld metah s -12 *F j
For other pressure-retainin g I
materuls. RTm is:
5 + 10*F 13.
The initial RV beltline Charpy 13.
Materials Test Reports and quality 13.
The initial RV beltline Charpy upper-shelf energy is no less than control records will be renewed.
upper shelf energy is no less than 75 ft-Ib.
a mammum required energy.
j 14 Instrumentation is provided in the 14.
Inspection of the control room for 14.
The temperaturr-and acoustic control room to alert operators of a presence of temperature and accus-alarms shown on Figure L5.I-I are not fully closed pressurizer safety tic alarms for each pressurizer available in the cc, trol room.
vahr.
safety vahr exit rme will be per-formed.
15.
The increase of RTm of the RV 15.
Redew the RV beltrme Material 15.
Residual chemical elements in RV beltline material over the life of Test Reports for residual chemical beltline materials are n> p eater the plant is limited by the control elements.
than-of residual chenucal elements.
Copper (in urids) 0.03 Copper (in forgings) 0.06 Nickel (in forginp) 120 Nickel (in welds) 0.10 Phosphorous 0212 l
1.5.1 01-28-93 t
4
[asuE cept c'_ ass I bd c AUXI.1ARY SPRAY (CVCS)
SAFETY VALVES PeCTE 1)
.kg owST =
4 RDS (SDS) 9 L J
i -
PeOTE 2) C hhhY CONTROL VALVE 1
u RCovSe m, =
Pss M'1'5 T
Camo:NotmE r
p (cVCS)
SEAL.
INJECTION LOOP 2A LOOP 1R wuECTION (CvCS) e,
cort._
e, BLEEDOFF (CVCS)
'I
~~'# 4 B dEDOFFK.VCS)'
'I' RCP 2A ELECTRICAL RCP 1B 5
- - -res wEATERS j. J m----_S MAIN I
MAN STEAM i
j (j ( STEAM
.l -
J ki L uuss E
tmES i
% {GEMNCL1}
[oEN NO.)2 l$ -PPS
-PPS
$D im Lg__m Lg_.m sm a
em
_e..m
+
+
\\
/
u outs -
t ESj y REACTOR PD LOOP 2 VESSEL LOOP 1 a _L h
T*
SEAL U
]f m
,,uECuoN
--m_
__m
.,Cnou SIS &
& SCS N
IN 3 7 I I LET OWN UNE (CVCS)
,, +,m_E,0,, -
_ m-i RCP 28 E3
-gRCP 1A
-- - _ g)
S _ _ _ - _m Orm j
- 1. CNE OF FOUR SAFETY VALVES SHOWN
- 2. ONE OFTWO ROS LMES SHOWN L
FIGURE 1.5.1-1
+
SYSTEM 80 REACTOR COOLANT SYSTEM t
_.3
-o
_ SYSTEM 80+"
1.5.1 REACTOR COOLANT SYSTEM ITAAC SUPPORTIVE INFORMATION 1.
Amnlifyine Information RCS Descriptiont CESSAR.DC Section 5 2.
Esigjenshin of RCS ITAAC to the Safety Analysis None 3.
Echtlonship of RCS ITAAC to PRA 1)
Each spray loop is connected to separate cold legs of the Reactor Coolant-System.
2)
Both loops are connected to a common header prior to entering the-pressurizer.
- 3)
The operation of the spray control valves can be accomplished manually.
4.
CESSAR.DC Chapter 14 Tests Annlicable to RCS 1TAAC Pre-operationalTestst CESSAR.DCSection 14.2.12.1.1,.2,.3,.37,.55 57, 58,.59--
1 1.9.1.2 1
01 28 93 t
.d
NYs1FM 80+
1.5.2 SilUTDOWN-COOLING SYSTEM Design Description 1he Shutdown Cooling System (SCS) is a safety-related system which removes heat from the reactor coolant and transfers the heat to the component c >oling water system. The SCS has two separate and redundant divisions. Each SCS division has the heat removal capacity to cool the reactor coolant from SCS entry conditions to cold shutdown conditions.
Each SCS dhision has an SCS pump, an SCS beat exchanger, valves, and connecting piping. Figure 1.5.21 shon a simplified system conDguration.
1hc SCS is built to the AShtE Code Section 111 class requirements shown on Figure 1.5.21. AShtE Code portions of the SCS retain their integrity under internal pressures that will be experienced during service. Components, piping, and supports classified as AShtE Code Class 1 or 2 are Scismic Category I. Equipment that is designated as safety.related is qualified for the environments where located.
SCS instrumentation indications and alarms shown on Figure 1.5.2-1 are available in the Control room. The SCS is controlled manually from the control room; remotely operated control valves in the SCS heat exchanger discharge line and bypass line can be positioned to control system Dow.
The SCS discharge valves are capable of opening against a differential pressure at least equal to the maximum SCS pump discharge pressure. The SCS discharge valves to the reactor coolant system are not interlocked on reactor coolant pressure. Flow-limiting devices are installed downstream from the SCS pump discharges to limit runout Dow.
The SCS pump and the Containment Spray System pump in a division are connected by piping and valves such that one pump can perform the other's function. The piping and valves in the SCS/ CSS pump suction cross-connect line permit Ocw in either direction (i.e., no check valves).
The valves in the piping from the reactor coolant system (RCS) to the suction side of the SCS pumps are interlocked so the valves cannot be opened if reactor coolant pressure exceeds SCS design pressure. There is no auto-closure interlock which shuts the SCS suction isolation valves on increasing RCS pressure during SCS operation.
The piping from the RCS to the SCS pump suction is oriented downward or horizontal, except for a short upward section connecting to the pump suction Gange.
Water is supplied to cach SCS pump at a pressure greater than the net positive suction head (NPSii) required.
i 1.S.2 1 28-93 l
r B15Eb1Mk Low temperature overpressure protection (1110P) for the llCS is provided by a relief valve kicated on each SCS suction line, as shown in l'igure 1.5.2-1, Each SCS division receives electrical power from its assigned Class 1.E bus.
A flow recirculation line around each SCS pump provides a minimum flow recirculation path. A piping line from downstream of the heat exchangers to the lilWST allows flow testing of the pumps during plant operation.
Outside containment, the two mechanical divisions of the SCS are separated by the divisional barrier well.
Inspections, Tests, Analyses and Acceptance Criterin Table 1.5.21 specifies the inspections, tests, analyses and associated acceptance criteria for the SCS.
1 9
0 1.5.2
-2 1 28-93
m 1
l l
SYSTEM 80+
Tn ELE 1. J SHUTDOWN COOLING SYSTEM Inspections. Tests. Analvsr.s. and Acceptance Criteria Certified Desien Commitment inspections Tests Analyses Acceptance Criteria 1.
A basic configuration for the SCS 1.
Inspections of the as-built SCS 1.
The as-built configuration of the is shown in Figure L5.2.L configuration will be performed.
SCS is in accordance with Figure L5.2-1 for the components and equipment shown.
2.
Water is supplied t.-each SCS pump 2.
Tests to measure SCS pump NPSH 2.
The calculated available NPSH at a pressure greater than the net will be performed. An analys to exceeds SCS pump NTSH required i
posithe suction head (NPSH) determine NPSH available to each by the wndor for the pump.
required.
pump will be prepared hsed on test data and as-built data.
3.
Relief vahrs are presided for low 3.
Inspect as-built configuration for 3.
A pressure relief vahr is installed temperature oyerpressare presence of relief vahrs.
in each SCS train suction line.
I 5.
Safety.related SCS components 5.
A test of the power availability to 5.a) The SCS pump motor in each j
described in the Design Description the SCS safety related components dhision is pourred from one of for each dhision of the SCS are will be conducted with pourt the two Class IE buses for tht powered from their respective supplied from the permanently division.
Each SCS pump derhrs Class IE busses.
mstalled electrical pourt buses.
its control power from the same Class IE bus that presides mothr power to the pump motor.
b) The SCS pump motor in each dhision is not powered from the same Class 1E bus as the CS pump motor in that dhision.
1.5.2 1-23-93
SYS1EM 80+
TABLE 1.5.2 (Continued)
SHUTDOWN COOLING SYSTEM Inspections. Tests. Anakses. and Acceptance Criteria Certified Desies Counmitment Inspections. Tests. Analyses Accendance Criteria 5.
(Contined) 5.
(Continued) 5.c) The motor for the SCS pump in each division is cooled by the CCWS for that disision.
6.
SCS instrumentation indications 6.
Inspection ' of the Control Room for 6.
The instrument = tion inhv=<
and alarms shown on Figure L5.2-the availability of instrumentation and alarms shown on I%ure L5.2-
-1 are available.. in the Control indic=tions and alarms will be 1 exist or can be nG. " in the Room. Controls are available in performed.
Tests will be Control Room.
SCS controls the control room to start and stop performed using the SG controls operate as specified in - the the SCS pumps, and open and close in the Control Room.
CertiGed Design Commitment.
the SCS remotely-operated vahrs shown in Figure L5.2.L 7.a)
Each shutdown.' cooling system has 7.a) Tests of as. built SCSu=4-.iion 7.a) Flow through the SCS heat the heat removal capacity to cool to measure the shutdown cooling s
- =-cr and HX bypass line can the reactor coolant from 'SCS entry
_ flow at the combined discharge of be a4usted while== int =ining a-n=vt;rian<
to cold ' shutdown the SG beat wh*"Ln r and HX nomin=1 flow of 5000 gpm per conditions.
bypass line mitt be performed.
division. Each SG pump devekys at least 4)0 feet of head at a flow rate no less than 5000 gpm.
b) Flow-limiting devices are installed b). With 5000 gpm Dow through the de-a - from the SG pump SCS heat <.
- _ gcz, comM SCS discharges to limit runout flow.
Ikm through the heat ~h==--r and bypass line does not exceed
!!aterlsom-1.5.2 ' 1-28-93
_~
4 4
SYsuM se+
TABLE L5.2 (Continued)
SHUTDOWN COOLING SYSTEM i-Inspections. Tests. Analyses. and Acceptance Criteria Certi5ed Desima C_
I==_ f n Tests. Amalvses Acceptance Criteria 1
8.a) Remotely operated SG suction 8.a) Tests using simulated RCS pressure 8.a) The SCS suction isolation vahn do line. isola. tion valves have greater than the SCS design not open.
indq.c acat interlocks to prevent pressure will be performed by opening if RCS pressure exceeds attempting to open the vahn from SCS design pmmo.
the control room. Each vahe will be tested indm dcatly.
i.
' b) There is no auto-closure interlock b) Tests of the component control b) The SG suction line isolation
' which shuts the SCS suction line circuits for the SCS suction line valves do not close automatically isolation vahn ~ on Le -,"q RCS isolation vahu will be performed when simulated RCS r== u pressure during SCS operation.
ming simulated RCS pressure increases.
signak.
i :
9.a) The SCS dischaise valves are 9.a)
Fn=ceba=1 tests will be performed 9.a) The SG discharge vahu open.
espable of opening agaimr a with. the SCS pmnps running at
. differential e== o at least equal
mimum flow recirm1=r6
by
~
' to maximum SCS pump discharge opening the SCS discharge vahu pressure.
from the' control room.
b) The SCS discharge vahu are not b) Tests of the SCS discharge vahes b)..The SCS disch-s,c vahes do ad
' interlocked on' RCS pressure.
will be performed ming simulated close automatie=Hy as simulated RCS pressure dansk RCS p m mu is I-aM
- 10.. ; ASME ' Code portions of the SCS 10.
A pressure test will be coeducted 10.
The resuks of the pressure test of
}L retain their integrity under internal on those portions of the SCS ASME Code portions of the SCS pressures 1that will be @caccd in ud to be em c tested by.
conform eith the 144 a in a
{
during scrdce.
the ASME Code.
the AShE Code Section IIL
' 1.5.2. '1-28-93
q l
SYs1EM se+
TABLE 1.5.2 (Continued)
SHUTDOWN COOLING SYSTEM Insocctions. Tests. Analyses. and Acceptance Criteria i
Certified Desies Cu~~- >
F _ A~- --- Tests. Anakses Acceptance Crkda r
t The SCS pumps can be flow tested IL Tests of the as-ine=IIed SCS will 11.
The SCS pumps pump up to 5000 IL-p during plant operation.
be performed by manually =Tiening gpm each through the test loop.
suction and divharge vahes to the
[
IRW5T and starting the SCS pumps manually.
12.
The SCS pump and the CSS pump 12.
An inspection of the as-built 12.
The SCS and CSS pumps suctions l
in a division 'are connected by piping will be performed.
and discharges are cross-connect-i:
piping. and valves such that one-Functional testing usine the ed by lines The valve (s) in the pump can ' perform the other's SCS/ CSS suction cross-connect line SCS/ CSS pump suction cross-function.
and the discharge cross connect connect Enes are not check valves.
l l
line will be performed.
[
~
13.
The piping from the RCS to the 11 An inspection of the as-built 11 The Certified Dc:Jgn Comm:tment SCS pump suction is oriented piping wi!I be performed.
is met du....id or bor':zontal, except for
~
{
- a short upward.'section L canarr(=g '
l to the pump suction flange.
14.
A flow recirculation line around-14 The as-built system wm"g-dius 14 Vmimum flow recirentwien rate I
- each. SCS pump ~ provides a mini-will be inspected and minimum meets or exceeds the pump l
mum flow recirculation path..
flow ' recirculation rate verified by vendor's Kvi.u-ents.
a mmunum flow measurement test.
J 15.
Outside containment, the - two 15.-
Visual inspections of SCS divisional 15.
Outside of containment, a
1 raech=aie=1. drvisions of the SCS are mechanie=1 w
4;uns will be
.SCS =ech=aie=1 disisicas.
I divisional wall separates the two e
l
. divisional barrier un pby sieally separatedby the-performed.
L:
I i
1.5.2 6-1-28-93
[
P
%.e w,+
a e...
.r-, -.,
..-..o
.m
-~r,c-
-e e--ec1-.
,-..%.+----
O=C" L
se v
T L OAv E
W D
h w
M S
Ao s uT T T L
z o
A hL N
J s
L o os
//
/ j/// //,
c i /.
M A-s ac E
hJ T
EO S
s m
}
g-Y gI
=
S t
G A
y N
y I
fg L
A~ T O
8 1 O L
- 2. C A
- 5. N L
g 1
l W
l l
X M
M E
l o
R O w
l e
U D t
c r
m T
G N
g U
l S
I S
FH A
C S
+
E 0
RU 8
RU ES
%s-PS ME E
ER KTP T
N NSS ACC S
O TRR S
DCEWH Y
N O
E G OIG 2
?
.o TA LH S
r 1,u-e OR M
P s
s R O TT o-J s
S EE T
E r
L P T NN o,o D
S UE u
c,.
L W
EV E RR C
I RL A
R R RRR I
P A T
OV U E UU I
./y/M,;//;N//h' ST T
S A CC N
L NN RMEWOO E
r EER GCCM D
e
- f/
O WYEUTT TE 1
W 2
T I ST D
SVEWWYS 1l T
C G YOU((SY Wtsg Oge 1,
I E 1
F 1
E-N AE EDD TS F
O NR URRREE N N o
J E
c
=
C P l.
NS A O N
N OT O
AL ENCTA E P L S
P S ATNR OCO C 0V p f, A S HNEEM C E I
U o
5 ECEMPNYY I
J N
Dg 8
A e S
XNNMALLOLLRN E
W Ln N
V T
O OA TEOIA ETUUTY HT CTP TTNFFCT O H E AMNWO7TA E N S a-I EOOO C 0OE A R
F DHCCLH9 MR S 0
T P$
aH~
i WSOrOCS
.T cO YXCSTWFFC S I
DHCCLNNNRS l
~
n
.e SYSTEM 80+"
1.9.1.2 SilUTDOWN COOLING SYSTEM ITAAC.
= SUPPORTIVE INFORMATION 1.
6.tnplifyine Information System
Description:
Relationshin of SCS ITAAC to the Safety Anahsis Basis: The head developed by the SCS pump is at lenst 400 feet at a flow rate no -
less than 5000 gpm.
ITAAC: ITAAC #7 tests the SCS with the pump at its design point (5000 gpm).-
3.
Relationshin of SFSR ITAAC to PRA 1)
The SCS has two separate and redundant divisions each with the heat removal capacity to cool the RCS to cold shutdown conditions.
2)
'dach SCS division has one SCS pump and one SCS heat exchanger.
3)
The SCS pumps can be aligned to the IRWST via a valve.
4)
The SCS discharge valves to the RCS are not interlocked on RCS pr_ essure and can be opened when the RCS pressure is less than or equal to the SCS pump shutoff head.
5)
The SCS discharge valves are capable of opening with a delta P equal to the SCS pump maximum discha_rge pressure.
9 6)
The SCS r imp in each division can perform the functions of the Containment Spray (CS) pump in that division for contcinment spray _ operation.
7)
The valve isolating the SCS pump suction from the IRWST is capable of passing flow in either direction.
L 8)
The SCS pump in each division can perform the function of the CS pump in.
that division to provide IRWST inventory cooling.
9)
Installed instrumentation provides the capability to monitor cooldown rate and shutdown cooling flow.
10)
Each SCS division is electrically powered from its assigned Class.1-E bus.'
E 1.9.1.2 01 28 SYSTEM 80+"
11) he SCS interfaces with the CCWS to remove to RCS decay heat loads.
12)
%c SCS can be aligned for shutdown cooling operation from the control room.
13)
The SCS pump motor in each division is powered from one of the two vital Class 1.E 4.16 Kv buses for that division. Each SCS pump derives its 125 VDC control power from the Class 1 E 125 VDC bus associated with the class 1 E 4.16 Kv bits that provides its motive power.
14)
The SCS pump motor in a cach division is not powered from the same Class 1 E 4.16 Ky bus as the CS pump motor in that division.
15)
The motor for the SCS pump in each division is cooled by the CCWS for that division.
16)
Installed instrumentation provides the capability to motiitor the performance l
of the system and the major components from the control room.
4.
CESSAR-DC Chanter 14 Tests Annlicable to SCS ITAAC Test
Description:
CESSAR-DC Section 14.2.12.1.21 I
t 1.9.1.2 01-28 93 i
k
l EYS'IF.M 80 +" -
1.9.1.1 NEW FUEL STORAGE RACKS Design Description The new fuel storage racks are safety related items that provide support and surround new fuel assemblics and maintain a geometric configuration to preclude nuclear criticality. The new fuel storage racks maintain the effective neutron multiplication factor below the required criticality limits during operation and for design buls and accident conditions.
The new fuel stusage racks are fabricated in accordance with ASME Code Section III, Class 3 requirements.
The new fuel storage racks are Seismic Category I classification.
Inspections, Tests, Analyses and Acceptance Criteria Table 1.9.1.11 specifies the inspections, tests, analyses and associated acceptance criteria for the new fuel storage racks.
i 1.9.1.1 Ol-28-93
E SYSTEM 80+ -
TABLE 1.9.1.1-1 NEW FUEL STORAGE RACKS Inspections. Tests. Analyses. and Accent =nce Criteria Certified Desima Connamitament Inspections. Tests. Analyses Accestance Criteria 1.
The new ; fuel storage. racks 1.
Analysis will be performed to L
Tlie cakulated effective neutron maintam -the effective neutron calculate the effective neutron
' mukiplication factor for the new :
- multiplicaten factor below the muhiplir='~ma factor for. the new fuel storage racks is less than 0.95 required criticality - limits during fuel storage racks.
for operation and design. ' basis operation _ and. design basis accident conditions (less than 0.98 accident conditions.
for immersion in a '.. uniform density aqueous foam 'or mist ' of optimum moderation ' density).
2.
The ' new fuel storage racks are
- 2...
Inspection of the construction 2.
The new fuel storage racks meet fabricated. 'in.accordance with records and the as-built new fuel the ASME Code specified physical.
ASME Code'Section III, Class.3 storage racks will be performed.
examiantina criteria for the-requirements.
ASME Code Section III,; Class 3 cimairu = tion.
.d n-1.9.1.1,
=- 2'-
01-28 93 m;,
SYSTEM 80+"
1.9.1.1 NEW FUEL STORAGE RACKS ITAAC SUPPORTIVE INFORMATION i
1.
Amplifyine Information The analysis specified in ITAAC 1 to calculate the effective neutron' multiplication factor wou.ld be describc1 See CESSAR DC, Section 9.1.1 for a discussion of the new fuel storage racks.-
2.
Relationship of NFSR ITAAC to the Safety Analysis None 3.
Belationship of NFSR ITAAC to PRA None 4.
CESSAR-DC Chapter 14 Tests Applicable to NFSR ITMC None 1.9.1.1 - -
01-28-93
- SYSTEM 80+"
1.9.1.2_
SPENT FUEL-STORAGE RACKS Design Description The spent fuel storage racks are safety related items that-provide support and surround spent fuel assemblies and maintain a geometric conGguration to preclude.
nuclear criticality. The spent fuel storage racks maintain the effective neutron multiplication factor below the required criticality limits for normal operation and postulated accident conditions.
The spent fuel sic age racks are fabricated in accordance with ASME Code, Section'-
III, Class 3 requirements.
The spent fuel storage racks ~ are Seismic Category I classification.'
Inspections, Tests, Analyses, and Acceptance Criteria Table 1.9.1.21 specifies the inspections, tests, analyses and associated acceptance -
<t criteria for the spent fuel storage racks.
t 4
1.9.1.2 - 01 28,
SYSTEM 80+='
TABLE 1.9.1.2-1 SPENT FUEL STORAGE RACKS Inspection. Tests. Analysis and Accephnce Criteria Certified Desima Comismitanent Insr=edh Test. Analysis Acceptance Criteria 1.
The-spent fuel. storage racks 1.
Analysis ' will be performed to L
The ' calculated effective neutron -
maintain the effective. neutron ca1euIate t h'c ef f ectiye multiplication factor for the spent multiplication faaor below the multiplication factor for the spent fuel storage racks is less than 0.95 required criticahty limit < :- for fuel storage racks for normal for normal operation and normal operation and postulated operation and postulated accident postulated accident conditions..
accident conditions.
condidons.
2.
The spent fuel storage racks. are
- 2..
Inspections of the construction 2.
The spent fuel storage racks meet fabricated in - accordance with records and the as-built spent fuel the ASME Code specified physical ASME Code Section III, Class 3 storage racks will be performed.
mmi= tion criteria for the ASME requirements.
Code Section III, Class.3 nauilication.
1.9.1.2 01-28-93
SYSTEM 80+"
- 1.9.1.2 SPENT FUEL STORAGE RACKS ITAAC SUPPORTIVE INFORMATION -
1.
Amplifyine Information The analysis specified in ITAAC 1 to calculate the effective neutron multiplication factor would be described.
See CESSAR-DC Section 9.1.2 for a discussion of the spent fuel storage racks.
2.
Relationship of SFSR ITAAC to the Safety Analysis -
~
None 3.
Relationship of SFSR ITAAC to PRA Nonc 4.
CESSAR DC Chapter 14 Tests Annlicable to SFSR ITAAC None
-1 1.9.1.2 101-28-93
SYSTEM 80+"
i 1.9.2.1 STATION SERVICE WATER SYSTEM Design Description The Station Senice Water System (SSWS) is a safetprelated system. It is an open loop system that takes suction from the Ultimate Heat Sink (UllS) and provides cooling water to remove heat from the Component Cooling Water System. The SSWS has the capacity to dissipate the heat loads of the CCWS during operation, shutdown, refueling, and design basis accident conditions.
The SSWS consists of two divisions. Each SSWS division is connected to its corresponding CCWS division through the component cooling water heat exchangers.
Each SSWS division has heat dissipation capacity to achieve and maintain cold shutdown.
Each division of the SSWS consists of two station senice water pumps, two station senice water strainers, and associated piping, valves, controls, and instrumentation.
The two mechanical divisions of the Station Senice Water System are physically separated. A basic configuration for the Station Service Water System is shown in Figure 1.9.2.1-1.
De ASME code classifications for the pressure retaining components of the Station Senice Water System are depicted in Figure 1.9.2.1-1. Components meeting ASME Code Class 3 requirements as depicted in the figure are safety related.
Components, piping and supports classified as ASME Code Class 3 are Seismic Category I. Equipment that is designated as safety related is qualified for the environments where located.
The station service water pumps are installed such that minimum available net positive suction head (NPSH) exceeds minimum required NPSH for each pump.
He instrumentation and alarms shown on Figure 1.9.2.1 1 are available in the Control Room. Controls are available in the Control Room to start and stop the station service water pumps. Controls are provided in the Control Room to manually align station senice water flow to the component cooling water heat exchangers.
Safety related components of each SSWS division are powered from their respective divisional Class lE busses.
Inspections, Tests, Analyses, and Acceptance Criteria Table 1.9.2.1-1 specifies the inspections, tests, analyses and associated acceptance criteria for the SSWS.
1.9.2.1 1-28-93 L-
SYSTEM 80+
TABLE 1.9.2.1-1 STATION' SERVICE WATER SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria.
Certified Desis Commitment Inseediens. Tests. Analyses Accentance Criteria 1.
A basic configuration for the L
Inspections of the as-buih SSWS L
He as-buih configuration of the Station Senice ' Water System is configuration will ' be conducted.
Station Service Water System is in.
shown in Figure L9.2.1-1.
accordance with Figure 1.9.2.1 for the composents and equipraent shown.
2.
The two mechanical divisions of 2.
Inspections of divisiona1 2.
A dhisional uall separates 'the two the SSWS are physically separated.
mechanical - separations will be SSWS mechanical dhisions.
performed.
3.a) The SSWS has the capacity to 3.a) Tests will be performed and 3.a) The heat dissipation capacity of the dissipate the.. heat loads' of the analysis prepared to determine heat CCWS exceeds the heat generation component cooling water system dissipation capacity on as-built capacity of the connected heat during operation, shutdown, CCWS serviced components and mbngers and coo:ers.Eduring
- refueling, and design basis ace;3ent measured flow rates.
operation, shutdown, ' refueling ' and :
conditions.
design basis accident condi6ons.
b) Each division has heat dissipation b) Tests w21 he performed and b) The heat dissipation capacity of to achieve and maintain cold analysis prepared for each dhision each CCWS dhision exceeds the.
shutdown.
for. heat dissipation capacity to heat-1oads generated 'for achieve and maintain cold achievement ' and maintenance Lof shutdown.
cold shutdown. -
4.
The' ASME code portions of the 4.
A pressure test will be conducted 4
The results of the pressure test of Station Service Water System. retain on those portions of the Station the ASME portions of the Sta6on their integrity - ~under internal Senice ' Water System required to-Service Water System conform with pressures experienced during be pressure tested by the code.
the requirements in thej ASME service.
Code Section III.
1.9.2.1 I-28-93 l:
T
~
~
~
SYSTEM 80+
TABLE 1.9.2.1-1 (Continued)
STATION SERVICE WATER SYSTEM Inspections. Tests. Analyses. and Acceptance Criterin Certified Desier. Comunitanent Inspections. Tests. Analyses Acceptance Criteria 5.
The station service water pumps 5.
An analysis to determine NPSH f.
The calculated available STSH
-are installed. such that minimum available to each pump will ' be exceeds SSWpump NPSH required -
available net positive suction head prepared based on..as-built data by the vendor for the SSWS pump.
_(NPSH) exceeds minimum required
. and sendor pump records.
NPSH for each pump.-
6.
SSWS instrumentation and alarms 6.
Inspection of Control Room 6.
The 'mstrumentation indications shown -on Figure 1.9111 are instrumentation indications - and and alarms shown in Figure
-l available in the Control Room.
alarms identified in the Certified 1.9.2.11. exist or can be retrieved Controls are available in the Design Commitment will be in the Control Room.
SSW3 -
Control Room to start 'and stop the performed. Tests will be performed controls operate as specified in the statien service-water pumps.
using the SSWS controls. in the Certified Design Commitment.
Controls are provided ' in the Control Room.
Control Room to manually align station service water flow to the component - cooling water heat.
- exchangers.
7.
Safety. related SSWS components.
7.
A test of power availability to the 7.
'The Certified design Commitment -
described in the Design Description safety-related ' SSWS components is met.
. for each division 'of the SSWS are will be ; conducted with power powered. from their.. respective -
supplied from the permanently
- Divisional Class IE busses.
installed electrie power. busses.
1.9.2.1 - 1 28.
STATION SERVICE WATER FROM 3
PUMP ULTIMATE j
i
{f
+
ssin HEAT SINK H
STRAINER b
1,.,,,,,
g 0 la==-1 i
bd g
1 TO T
R
? ULTIMATE 2
H HEAT SINK i ty i
O STATION H
FROM 3
PUMP ULTIMATE l
N' S
+
sE i
STRAINER L
DIVISION 1 DtvistON AL SEPAR AItON DIVISION 2 FROM N:
2 HEAT SINK f;
ssw
+
(
STATION j
N NER SERVICE WATER j
I'*"* " " I PUMP i
Ed g
i TO i
R
> ULTIMATE 3
H HEAT SINK i r-E "4HED@
"N t
il% ATE )r N
+
s3/
l srRotR STAriON FIGURE 1.9.2.1-1
"""7P STATION SERVICE WATER SYSTEM
' ""^'""
'jlYSTEM 80+"
l:
[
1.9.2.1 STATION SERVICE WATER SYSTEM SUPPORTIVE INFORMATION 1.
Amnlifying Information ITAAC 3 Confirmation of the SSWS heat dissipation capacity during operation, shutdown, _
refueling, and design basis accident conditions will be performed as part of the CCWS -
heat dissipation capacity analysis (See CCWS Tier 2-Amplifying Information).
The analysis will demonstrate that only one station service water pump matched with -
one component cooling water heat exchanger receiving component cooling water flow -
is required to operate during post accident conditions. The analysis will also-demonstrate that each division of the SSWS matched with one operating CCWS division has a heat dissipation capacity to achieve and maintain cold shutdown.
ITAAC 5 Confirmation of adequate pump NPSH will include an analysis with the following-conditions:
Station service water pump clevation.
Station service water intake level at a minimum value.
Maximum design basis station service water inlet temperature.
The calculated minimum available NPSH shall exceed NPSH required by the vendor for the SSWS pump. --
ITAAC 7 1
1 Testing of Class 1E power availability to SSWS components will include confirmation of the following:
4 Within a division, one station service water pump motor is powered from one l
Class 1E bus in that division and the other station r.ervice water pump motor is powered from the other Class 1E bus in that division iStation service water pump control circuits of the two station service water
. pumps in a division are powered from separate Clus 1E buses.
1.9.2.1 1-28-93
SYSTEM 80+"
The standby station service water pump in each division will automatically start if the running pump in that division trips.
On a loss of offsite power (LOOP), both station sersice water pumps in each division will be aligned to the diesel generator for that division. The load sequencer willload and start one pump. If that pump trips, the standby pump will then be autor.atically loaded and started.
2.
Relationshin of SSWS ITAAC to the Safety Analg,is The SSWS ITAAC does not include any specific inspections, tests, and analyses which confirm that the as. built system configuration and performance match the bases used in the evaluation models for licensing analysis. llowever,it is assumed that the SSWS is available to support the CCWS and its assumed characteristics as detailed in the evaluation models.
3.
Relationshin of SSWS ITAAC to PRA 1)
The SSWS has two redundant and separate safety related divisions with heat.
dissipation capacity to achieve and maintain safe shutdown.
2)
Each SSWS division has two SSW pumps per division.
3)
"Ihe SSWS interfaces with the CCWS to remove heat from CCWS connected loads.
4)
SSWS components in a division receive electrical power from the Class IE buses in their division.
5)
The SSW pump motors in a division are powered from the 4.16KV Class 1E power system in their division. In a division, one SSW pump is powered from one Class IE bus in that division and the other SSW pump motor is powered from the other Class 1E bus in that division.
6)
SSW pump control circuits in a division which close and trip the SSW pump breakers when required are powered from the 125 VDC Class IE power system in their division. The SSW pump control circuits of the two SSW pumps in a division are powered from separate 125 VDC Class 1E buses.
7)
Manual Start and stop actuation of the SSW pumps is provided from the control room to override automatic actuation.
8)
The two SSW divisions are physically separated and protected such that a fire or flood in one division will not affect the SSW pumps in the other division.
1.9.2.1 1-23-93
SYSTEM 80+"
9)
Installed instrumentation provides the capability to monitor the performance of the system and the major components from the control room.
4.
CESSAR-DC Chanter 14 Tests Annlicable to SSWS ITAAC See CESSAR DC Section 14.2.12.1.78 i
1.9.2.1 1-28-93
SYSTEM 80+"
1.9.2.2 COMPONENT COOLING WATER SYSTEM Design Description ne Component Cooling Water System (CCWS) is a safety-related closed loop cooling water system that, in conjunction with the Station Service Water System (SSWS) and the Ultimate Heat Sink (UllS), removes heat generated from the plant's safety related and non. safety related components and heat exchangers connected to the CCWS. The CCWS conshts of two divisions. Each CCWS division is connected to its corresponding SSWS division through the component cooling water heat exchangers.
He CCWS has the capacity to dissipate the heat loads of connected condensers, coolers, and heat exchangers during operation, shutdown, refueling, and design basis accident conditions. Each division has heat dissipation capacity to achieve and maintain cold shutdown. The CCWS provides a minimum flow to each containment spray heat exchanger.
Each division of the CCWS includes two component cooling water heat exchangers, a component cooling water surge tank, two component cooling water pumps, piping, valves, controls, and instrumentation. Outside containment, the two mechanical divisions of the CCWS are physically separated. A basic conceptual configuration of the CCWS is shown in Figure 1.9.2.2-1. Equipment depicted in Tables 1.9.2.2-2 and 1.9.2.2-3 receives cooling water flow during the plant modes indicated.
The ASME code classifications for the pressure retaining components of the Component Cooling Water System are depicted in Figure 1.9.2.21. Components meeting ASME Code Class 3 requirements as depicted in the figure are safety-related.
Additionally, cooling loops supplying component cooling water to the safety related components in Tables 1.9.2.2-2 and 1.9.2.2-3 are designated as safety related cooling loops and meet ASME Code Class 3 requirements.
Components, piping, and supports classified as ASME Code Class 3 are Seismic Category I.
Equipment that is designated as safety related is qualified for the environments wherc located.
Component cooling water is supplied to each component cooling water pump at a pressure greater than the net positive suction head (NPSH) required.
The interface from ASME Code Class 3 component cooling water piping totally outside. containment to cooling loops composed of non-ASME Code component cooling water piping is at two valves. ASME Code Class 3 requirements extend from the ASME Code Class 3 piping through both valves. These valves can be manually closed with controls in the Control Room and close automatically upon receipt of a 1.9.2.2 01-28-93
i I
SYSTEM 80+"
Safety Injection Actuation Signal (SIAS). Upon loss of motive power, these valves fail to closed positions.
The CCWS piping to the reactor coolant pumps and to the letdown heat exchanger has containment isolation valves. Containment isolation valves for the reactor coolant pumps can be operated to opened and closed positions with controls in the Control Room. Component cooling water to the reactor coolant pumps is not terminated on a Containment Isolation Actuation Signal (CIAS) or on a Safety Injection Actuation Signal (SI AS). Component cooling water flow to the letdown heat exchanger is automatically terminated on a CIAS or on an SIAS.
The instrumentation indications and alarms shown on Figure 1.9.2.21 are available in the Control Room. Controls are available in the Control Room to start and stop the component cooling water pumps. Controls are provided in the Control Room to manually align component cooling water flow to the component cooling water heat exchangers.
The following controls are available in the control room to manually initiate and/or terminate flow to components connected to the CCWS:
1)
Component cooling water flow to each shutdown cooling heat exchanger can be initiated and terminated.
2)
Component cooling water flow to each containment spray heat exchanger can be initiated and terminated.
3)
Component cooling water flow to each spent fuel pool cooling heat exchanger can be initiated and terminated.
Automatic initiation or termination of component cooling water flow is provided for the fol. lowing components connected to the CCWS:
1)
Component cooling water to cooling loops composed of non-ASME code piping is terminated upon receipt of a component cooling water low-low surge tank level signal, 2)
Component cooling water flow to each containment spray heat exchanger is initiated automatically upon receipt or a Containment Spray Actuation Signal (CSAS).
3)
Component cooling water flow to each spent fuel pool cooling heat exchanger is terminated by a Safety Injection Actuation Signal (SIAS).
Makeup water to the CCWS is supplied by the Demineralized Water Makeup System (DWMS). A safety related makeup line of Seismic Category I construction is 1.9.2.2 01-28-93
i 1
SYSTEM 80+"
provided to each division from the Station Service Water System via a spool piece which is normally removed.
1 Safety related components of each CCWS division are powered from their respective divisional Class IE busses with the exception of containment isolation valves and associated containment isolation valve instrumen;ation and controls.
1 Inspections, Tests, Analyses, and Acceptance Criteria Table 1.9.2.21 specifies the inspections, tests, analyses and associated acceptance:
criteria for the CCWS.
1.9.2.2 01 28-93
SYSTEM 80+
TABLE 1.9.2.2-1 l-COMPONENT COOLING WATER SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Certified Desien Commitment Inspections. Tests. Analyses Acceptance Criteria 1.
A, basic configuration for the 1.
Inspections of the as-built CCWS 1.
He as-built configuration of the Component Cooling 1 Water System configuration will be conducted.
Component Cooling Water System is in accordance with Figure is shown in Figure '1.9.2.2-1.
L9.2.2-1 for the components and equipment shown.
2.
Outside containment,.
the two 2.
Iaspections of divisiona1 2.
Outside containment, a dhisional mechanical ' divisions of the CCWS mechanical separations will be w~all separates the two CCWS are physically separated.
performed.
mechanical divisions.
3.a) The,CCWS has the capacity to 3.a) Test will be performed and analysis 3.a) The heat dissipation capacity of the dissipate the l heat loads. of prepared to determine heat CCWS cxceeds the heat generation connected condensers, coolers, and dissipation capacity based on as-capacity of the connected heat exchangers. during operation, built CCWS sersiced components condensers, coolers,.and heat
- shutdown, - refueling, ' and design.
and measured flow rates.
exchangers during operation, shut-basis accident ' conditions.
down, refueling and design ~ basis accident' conditions.
b) Each dhision has beat dissipation b) Test will be performed and analysis b) The heat dissipation capacity' of capacity to achieve and maintain prepared for each dhision for heat each CCWS dhision exceeds the cold shutdown.
dissipadon capacity to achieve and heat Ioads generated for maintain cold shutdown.
achievem;nt and maintenance of cold shutdown.
c) %e CCWS provides a minimum c) Test will be performed to confirm c) %e CCWS proSides at least 8000
' flow to each. containment spray CCWS flow rate to the containment gallons per minute to each heat exchanger, spray heat exchangers.
containment spray heat exchanger.
1.9.2.2 0I-28-93
(
SYSTEM 80+
' TABLE 1.9.2.2-1 (Continued)
COMPONENT COOLING WATER SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria _
Certi6ed Desian Counnaitenest Inspections. Tests. Analyses Acceptance Criteria 4.
The ASME code portions of the-4.
. A pressure test will be conducted 4.
The results of the pressure test of..
Component Cooling Water System on ' those portions 'of the the ASME portions -
of the-retain their integrity under internal Component Cooling Water System Component Cooling Water System.
pressures. ex perienced ' during required to be pressure : tested by conform with the requirements. in
- service.
the ASME code.
the ASME Code Section IIL 5.
Component -
cooling water is 5.
Tests' to measure CCWS pump 5.
The calculated available ' NPSH:
supplied ^ to each CCW pump at a NPSH will be performed.
An
- exceeds CCWpump NPSH required !
pressure greater than-the net
' analysis to determine NPSH by the vendor for the CCW pump..
positive suction head (NPSH)
. available to each pump will be required.
prepared based on. test. data, as-built data, and vendor pump records.
6.a) The interface ; from ASME Code 6.a) Inspections of the construction 6 a) The interface is as desenhd in the Class 3 component cooling ~ water records and the as-built i=ran=rion Certified. Design Commitment.
piping totally outside containment
' will be performed.
to cooling loops composed ' of non-ASME. Code. component coohng water, piping ist at two valves.
ASME Code Class 3 requirements extend from the ASME Code Cass -
3 piping through ' both valves.
b) Rese. valves can'.be manually b) L A test of Control Room' closure b). He valves can be manually closed closed with controls.in the Control
. capabihties will be performed., A
' from ' the. Control - Room and the' Room and close automatically upon.
test ' will; be ' performed :. using a valves close ' upon receipt of a receipt - of : a : Safety-Injection
' simulated. SIAS signal simulated SIAS.
Actuation Signal-(SIAS).
1.9.2.2= 01-28-93' a
~
SYSTEM 80+ -
TABLE 1.9.2.2-1 (Continucd)
COMPONENT COOLING WATER SYSTEM Inspections. Tests. Analyses. and Acceptance Cr6n'a Certified Desien Con:mituneet Inspections. Tests. Anakses Acceptase-Criteria i-6.c) Upon loss of 'modve power, thesc' 6.c)
'A test using a simulated or actual 6.c) The des close on loss of motim valves fail to closed positions.
. loss of motive power to the vahrs pourr.
L will be performed.
~7.a) ' Containment isolation valves for 7.a) A test of. containment : isolation 7.a) The Certdied -Design Cometiment
. the reactor ; coolant pumps. can be valve opening and closing is met.
' operated to opened and closed
- capabihties will be performed.
positions with controls '. in the Control Room.
b) Component cooling water to the b) Tests mill be performed using b) The Certified Design Commitment reactor. coolant pumps is not simulated CIAS and SIAS signals.
is met.
isolated on a Containment Isolation Actuation Signal L (CIAS) or on a Safety.' Injection-. Action : Signal
. (SIAS).
e)
Component. cooling water flow to.
c) Tests will be' performed using c) The Certified Design Commitment
' the Ictdown heat errhanger is ~
simulated CIAS and SIAS signals.
is met.
automatically isolated on a CIAS or :
on a SIAS.
1.9.2.2.
~
6-01-28-93 iE--
M w'
%~m'
SYSTEM 80+
TABLE 1.9.2.2-1 (Continued)
COMPONENT COOLING WATER SYSTEM Inspections. Tests ' Analyses. and Acceptance Criteria Certified Desien Commitment Inspect ~ums. Tests. Analyses Acceptance Criteria -
8.
CCWS instrumentation indications 8.
Inspection of the - Control Room R.
The instrumentation ' indications and alarms. shown. in - Figure instrumentation '
indications and and alarms shovm in.. Figure 1.9.2.2-1 are available in the alarms identified in the Certified 1.9.2.2-1 exist or can be retrieved
- Control Room.
Controls are Dedgn Commitment will be in the Control Room.
.CCWS available in the Control Room to performed.
Tests will be
. controls operate as specified in the start. and stop the. component
' performed using the CCWS controls Certdied Design Commitment..
cooling water pumps. Controls are in the Control Room.
provided ' in the Control Room to manually align the component cooling water heat nehangers.
9.a) Controls
'are available in. the 9.a) Tests of initiation and termination, 9.a) Controls are provided in the Control Room to manually initiate
.of component cooling water flow Control Room as specified below:-
andfor terminate flow to will' be performed. '
components connected
.to the
- 1) Component cooling water flow to -
CCWS.
cach shutdown cooling heat exchanger. can be initiated. and terminated.
- 2) Component cooling water flow to each. containment spray. heat exchanger can be terminated.
- 3) Component cooling. water ficw to each spent fuel pool heat neh=ger can be initiated ' and terminated.
R 1.9.2.2 01-28-93
SYSTEM 80+
TABLE 1.9.2.2-1 (Continued)
COMPONENT COOLING WATER SYSTEM Insocctions. Tests. Analyses, and Acceptance Criteria Certified Design Commitment Inspections. Tests. Analyses Acceptance Criteria 9.b) Automatic initiation or termination 9.b) Tests will be performed using 9.b) Automatic initiation or termination of component cooling flow is simulated SIAS and CSAS signals.
of component cooling water flow is provided for components connected A component cooling water surge as specified below; to the CCWS.
tank low-low level signal will also i
i be simulated.
- 1) Component cooling water f!cw to cocling loops composed of non-ASME code piping is terminated l
automatically upon the receipt of a component cooling water surge l
tank low-low level signal
- 2) Component cooling water flow to each containment spray heat exchanger isinitiated automatically upon receipt of a Containment Spray Actuation Signal (CSAS).
- 3) Component cooling water flow to each spent fuel pool coormg heat exchanger is terminated automatically by a Safety Injection Actuation Signal (SIAS).
l l
1.9.2.2 01-28-93 l
MMM Am TABLE 1.9.2.2-1 (Continued)
SYSTEM 80+
COMPONENT COOLING WATER SYSTEM Inspections. Tests. Analyses. and Acccotance Criteria Certified Design Commitment Inspections. Tests. Analyses Acceptance Criteria 10.
Safety related CCWS components 10.
A test of power availability to the 10.
The Certified Design Commitment described in the Design Description CCWS components desenkd in the is met.
for each division of the CCWS are Design Description will be powered from their respective conducted with power supplied divisional Class IE busses with the from the permanently instaIIed exception of containment isolation electric power busses.
l valves and associated containment isolation vaht irtstrumentation and controls.
l 1
l l
1 01-28-93 1.9.2.2
- m A
SYSTEM 80+
TABLE 1.9.2.2-2 COMPONENT COOLING WATER CONSUMERS Division 1 Plant' Mode / -
Normal Operation Shutdown. Cooling
' Shutdown Cooling Refueling Design Basis '
Comeponents Initial-Final Accident SAFETY-RELATED (Note a)
X X
heat exchanaer '
Cantainment : spray :
X beat erchanger = ~
Spent fuel pool
. X (Note b):
X (Note b)
X cooling heat exchanger Diesel Generator X
X-X X
X, Pump Motor Cool- -
X.
X X
X
'X ers,' MiniDow -. Heat -
Frehanaeg and .
- Essential ' Chilled Water Condensers l'
1.9.2.2 - '
01-28 --
.i 4-.
Dash um-a.
.a.
.a.
__w i4.__...im yun-_
SYSTEM 80+
TABLE 1.9.2.2-2 (Continued)
COMPONENT COOLING WATER CONSUMERS E
Division 1 l
Plant Mode /
Normal Operation.
. Shutdown ' Cooling.
Shutdown Cooling Refueling'.
Design Basis Components initial Fleal '
Accident NON-SAFETY REIATED Reactor coolant -
X
'X.
X X
X-pumps and pump motors -
q i
Charging pump X
X X
X X
motor coolers Charging pump X-
'.X X
X X'
miniflow heat -
j exchanger Normal Chilled X
X-X X
Water Condensers, Instrument Air.
Compressors, Let.
down Heat Fwhanger,. Sample Heat Fwhangers, Gas Stripper,' and Boric Acid Con- ~
centrator. (Note c) '
.l.9.2.2 -
01-28 93-
=
' SYS'mM 80+
-J TABLE 1.9.2.2-3 COMPONENT COOLING WATER CONSUMERS Division 2 Plant Mode /
Norn't Operation Shutdown Cooling Shutdown Coohng Refueling
. Design Basis Components initial Final Accident SAFETY RELATED Note a Shutdown. cooling X
X X
heat nehinger Containment spray X
heat exchanger Spent fuel pool X (Note b)
X (Note b)
X cooling. heat exchangers Diesel generator X
X X
X X
Pump Motor Cool- -
X X
X X
X ers, Miniflow Heat -
Exchangers, and; Essential Chilled.
Water Condensers l
r i
l-
.1.9.2.2 01-28-93
SYSTFM 80+
m TABLE 1.9.2.2-3 (Continued)
COMPONENT COOLING WATER CONSUMERS Division 2 1
i Plant Mode /
Normal Operation Shutdown Cooling Shutdown Cooling Refueling Design Easis Components Initial Final Accident NON-SAFETY RELATED Reactor coolant X
X X
X X
pumps and pump motors Charging pump X
X X
X X
motor coolers Charging pump
.X X
X X
X minirow. heat exchanger Normal Chilled X
X X
X Water Condensers, Instrument Air -
Compressors, let-
'down Heat F'A=ager, Sample '
Heat FrA=ngers,.-
Gas Stripper, and -
Boric Acid Con-centrator (Note c) m r-e 1.9.2.2 01-28-93 Q.
i SYSTEM 80+
NOTES FOR TABLES 1.9.2.2-2 AND 1.9.2.2-3 a.
(X) = Equipment receives compoacnt cooling water ikw in this mode
(-) = Equipment does not receive component cooling water flow in this mode b.
Either or both spent fuel pool cooling heat exchangers can reccht: flow during thic operating mode.
c.
Assignment of the Ccmponent Cooling Water source to the Ietdown Heat Exchanger, Sample Heat Exchangers, Gas Stripper and Boric Acid Concentrator is dependent upon the dhisionallocation of these components.
s 1.9.2.2 01 28 93
rs)--
,i
_t-.
I cetssestautrD waTra m m aseawce CC"#
w s,s-se e===
_rw
""" C SURGE Ml----->l CCW HX h y
TANK
{
L U
%J CCW PUMP i
O f
?/
l CCW RETURN
+
CCW FROM
-EE" CCW PUMP
- E=~"
HEAT LOADS SUPPLY TO I
4
@--f _,,,
. =.. = a.
etwwnnt suwuren Fy3-ic/71 DIVISION 2 NOTE A I
O ove-i.e uw
- 4. eein-e=urne.aren m, _
CCw s-
,,,,,,,,, g me==.amew
+
SURGE w
CCW W,,;,@
u TANK t
L U
CCW PUMP yy
[
H
~
i Af es i
ca mm ram CCW PUMP SUPPLY i
m.-
To 1
HEAT LOADS g
g HEAT l
-.l CCw Hx l :
toros
.4#,,
ore m.
sm m>.
A. ARE300VABLESPOOLPsECE!S LOCATE.D ON EACH STATION SEmMCE L'L*TL'"*""'
'^c" FIGURE 1.9.2.2-1 f
CO MPONENT COOLING WATER SYSTEM SYSTEM 80+*
8
1 1
.SYEIEM 80+"
1.9.2.2 COMPONENT COOLING WATER SYSTEM SUPPORTIVE INFORMATION i
1.
Amplifying information j
ITAAC 3 Confirmation of the CCWS heat dissipation capacity during operation, shutdown, refueling, and design basis accident conditions will be perfo:med. An analysis will be performed based upon the as built CCWS serviced components and meuured flow -
rates. 'The analysis will be based on the following:
CCWS flow to cooled components for cach plant mode Y
SSWS flow to cach component cooling water heat exchanger Design basis station service water inlet temperature a
4 Vendor heat exchanger data
'The analysis will demonstrate that only one component cooling water pump matched with one comimnent cooling water heat exchanger is required to operate during post-accident conditions. The analysis will also demonstrate that each division has a heat-dissipation capacity to achieve and maintain cold shutdown.
ITAAC5 Confirmation of adequate pump NPS!! will include testing and analysis with the
~
following conditions:
Component cooling water surge tank arid component cooling water pump locations and clevations.
Component cooling water surge tank water level at a minimum value with measured isolation valve closure times for cooling loops composed of non..
ASME code component cooling water piping.
Maximum design basis component cooling water temperature.
Pressure losses for pump inlet piping and components.'
Doth component cooling water pumps operating in a single d! visions -
1-01-28-93 1.9.2.2 i
1-
+
ede.r"w a-e-*-E fe++-e ygsM4s&<ww-e 37 yy-.ie
',,w, 4
rw-gy
.-rey
-g e-g seeg
-e.q,w-,
p.
y,,,..eyv..r.w.iaw,7i.
9,n4.;g
SYS1EM 80+"
The measured / calculated pump NPSil shall exceed the pump NPSil required by the vendor.
i ITAAC 10 Testing of Class 1 B power availability to CCWS components will include confirmation of the following:
Within a division, one component cooling water pump motor is powered from one Class IE bus in that division and the other component cooling water pump motor is powered from the other Class 1E bus in that division Component cooling water pump control circuits of the two component cooling pumps in a division are powered from separate Class IE buses.
The standby component cooling water pump in each division will automatically start if the running purnp in that division trips.
On a loss of offsite power (LOOP), both component cooling water pumps in each division will be aligned to the dicscl generator for that division. 'Ihe load sequencer willload and start one pump. If that pump trips, the standby pump will then be automatically loaded and started.
2.
Relationship of CCWS ITAAC to the Snfety Anahsis The CCWS ITAAC includes inspections, tests, and analyses which confirm that the as-built system configuration and performance match the bases used in the evaluation models for lleensing analysis. '1he CCWS characteristics and their treatment in the ITAAC are described below:
a)
IIASIS: Minimum component cooling water flow rate of 8000 g-!!ons per minute to each contninment spray heat exchanger.
ITAAC:ITAAC 3c acceptance criterion requires the CCWS to provide at least 8000 gallons per minute to each containment spray heat exchanger.
3.
Relationship of CCWS ITAAC to PRA 1)
The CCWS has two redundant and separate safety related divisions with heat dissipation capacity to achieve and maintain safe shutdown.
2)
Each CCWS division has two CCW pumps per division.
1.9.2.2 01 28-93
l I
i SYS'mM 80+"
3)
The supply and return lines to and from components in a division are completely separated from the supply and rcturn lines in the redundant dmston.
4)
The ESF Actuation System signals isolate the non safety related portion of the CCWS following an accident condition, except cooling for the RCPs, charging pump motor coolers, and charging pump miniflow heat exchangers.
5)
The CCWS interfaces with the SSWS to remove heat from CCWS connected loads.
6)
Manual Start and stop actuation of the CCW pumps is provided from the control room to override automatic actuation.
7)
The two divisions of CCWS are physically separated.
8)
Installed instrumentation provides the capability to monitor the performance of the system and the major components from the control room.
4.
CESSAR DC Chanter 14 Tests Annlicable to CCWS ITAAQ See CESSAR.DC Section 14.2.12.1.79' 1.9.2.2 01 28 93
.m
SYSTEM 80+"
1.9.2.3 DEMINERALIZED WATER MAKEUP SYSTEM Design Description The Dcmineralized Water Makeup System (DWMS) supplies water to the Condensate Storage System for makeup and to systems in the plant that require dcmineralized makeup water.
The Dcmineralized Water Makeup System does not perform any safety functions.
Failure of this system does not effect plant safety.
The DWMS includes transfer pumps, demineralizers, a vacuum degasifier, a storage tank, and associated piping, valves, and controls.
A basic configuration for the Dcmineralized Water Makeup System is shown in Figure 1.9.2.3. Two redundant dcmineralizer trains are provided.
The system is controlled from local control pancis.
Inspections, Tests, Analyses, and Acceptance Criteria Table 1.9.2.31 specifics the inspections, tests, analyses and associated acceptance criteria DWMS.
13.2.3 01 28-93
- - ~ -
. - ~, - _
t i..
SYSIEM se+
TABLE 1//.23-1 DEMINERAIJZED WATER MAKEUP SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria
_Gerti6ed Dreism C; I- -_ : ':1 - Tests. Anakses Acceptance Crieeria
+
1-L A basic-configuration for. the '
L 1mid-ss of the as-bui t sprem L
The as built configuration of the Demineralized ' Water Makeup configuration will be performed.
Demineralized Water Makeup p
System is shown in Figure L9.23.
Sprem is in accordance with Fu;;ure L9.23 for the componees and q Q 4 shown.
t I
1 L-l' 4
1-2.
..1.9.23- 01-23-93
3 u, -
- ED ""=""
~\\
O a
m z
=
g-g g
O la 8
O!!
O:i!
i ei i
=
8
!I s.3 I!'
8' d hi~1
- ED-
+
cQ s.
O!!O!!
~~~
=
- = -
si li 3-umn.
M ML=7,1".
"g* """ ~ ~~3 -
g _- - -
FIGURE 1.9.2.3 DEMINERALIZED WATER MAKEUP SYSTEM
9 SYSTEM 89E 1.9.2.3 DEMINERALIZED WATER MAKEUP SYSTEM ITAAC SUPPORTIVE INFORMATION 1.
Amplifyine Iqformation N/A-2.
Relationship of DWMS ITAAC to the Safety Analysis N/A 3.
Relationshin of DWMS ITAAC to PRA -
N/A 4.
CESSAR DC Chanter 14 Tests Aonlicabic to DWMS ITAAC N/A f
1.9.2.3 01 28-93
. ~
. ~
1 SYSTEM 80+
1.9.8.1 ESSENTIAL CIIILLED WATER SYSTEM Design Description The Essential Chilleu Water System (ECWS) is a safety related system it is a closed loop chilled water system that serves safety related IIVAC cooling loads.1he ECWS is a subsystem of the Chilled Water System (CWS) and provides chhled water to meet the cooling loads of the essential flVAC chilled water coils.
The Essential Chilled Water System consists of two divisions.1hc two mechanical divisions of the ECWS are physically separated. Each division consists of a chilled water refrigeration unit, a circulating chilled water pump, control valves, instrumentation and piping. A basic configuration of the ECWS is shown in Figure 1.9.8.1 1.
1hc ASME code classifications for the pressure retaining components of the Essential Chilled Water System are depicted in Figure 1.9.8.1. Components meeting ASME Code Class 3 requircmunts as depicted in the figure are safety related.
Components, piping and supports chtssified as ASME Code Class 3 are Scismic Category I.
Equipment that is designated as safety related is qualified for the environment where k>c.ited.
1hc ECWS can be actuated manually from the Control Room and is automatically actuated upon loss of the Normal Chilled Water System to furnish essential chilled water. less of water Cow through the chillers and high chilled water outlet temperature are indicated by alarms in the Control Room.
Makeup water to the ECWS is supplied by the Dcmineralized Water Makeup System (DWMS). A safety related makeup line of Scismic Category I construction is provided to each division from the Station Service Water System via a spool piccc which is normally removed.
Safety related components of each ECWS division are powered from their respective divisional Cbss 1E buses.
Inspections, Ti.sts, Analyses, and Acceptance Criteria Table 1.9.8.11 specifies the inspections, tests,' analyses and associated acceptance criteria for the ECWS.
1.9.8.1 12893
1 SYSEM 80+
TABLE 1.9.8.1-1 i
ESSENTIAL CHILLED WATER SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Certified Design Commitment Inspections. Tests. Analnes Acceptance Criteria l
1.
A basic configuration for the L
Inspections of the as-built Essential L
The as-built configuration of the Essential Chmed Water System is Chdled Water System configuration F"~ri=1 Chmed Water System is in i
shown in Figure L91L will be conducted.
accordance with Figure L9Al for i
the components and equipment shown.
1 i
j 2.
The two mechanical dhisions of 2.
Inspeetioas of divisiona1 2.
Ounide coatMammt a dhisional the ECWS are physically separated.
mechanical separations will be wall separates the two ECWS performed.
mechanical divisions.
[
i 3.
He ECWS presides chilled water 3.
Tests will be performed and 3.
He cooling capacity of the ECWS t
j to meet the cooling loads of the analysis prepared to determine exceeds the cooling requirements j
essential HVAC chilled water coils.
cooling capacity based on as. built of connected HVAC coolmg coils.
ECWS seniced components and measured flow rates.
l 4.
He ASME code portions of the 4.
A pressure test will be conducted 4
The results of the pressure test of Essential Chilled. Water. System on those portions of the Essential the ASME portions of the retain their integrity under internal Chilled Water System required to Comr.uat Cooling Water System pressures experienced during be pressure tested by the ASME conform with the requirements in senice.
code.
the ASME Code Section IIL 5.a) %e ECWS can be actuated 5.a) Tests will be performed to actuate 5.a) Components of the ECWS can be manually from the Control Room components of the ECW3 using actuated manually from the Centrol and is automatically actuated upon controls in the Control Room. A Room. He system is automatically loss of the Normal Chilled. Water tests aill also be performed ming a actuated upon the loss of the System to furnish essential chilled signal which simulates loss of the Normal Chilled Water System.
water.
Normal Chilled Water System.
i 1.9.8.1 1-28-93
sysnM 80+
TABLE 1.9.8.1-1 (Continned)
ESSENTIAL CHIILED WATER SYSTEM Inspections. Tests. Analyses. and Accretance Criteria Certined Desien Comannitament I=%f =_ Tests. Analyses Actestance C-iteria 5.b) Loss of water flow through the 5.b) Inspection of the Control Room ib) The alarms indicated in the chillers and - high --l chilled water instrumentation alarms hiH Certified Design C-mi ment' are t
outlet temperature are annunciated in the Certified Design prmided in the Control Room..
in the Cetrol Room.
Commitment will be performed.
Alarm (mak will be simulated.
6.
Safety - related ECWS componems 6.
A test of power availability to the 6.
The Certifed Design Commitment described in the Design Description ECWScomponents described in the is' met.
for each division of the ECWS are Design. Description will be im J from their respectise conducted with pourr supplied divisional. Class 1E buses.
from the permanently "mstaHed electric power buses.
1 -
(
l 4
f.
4 1
1.9.8.1 1-28-93
, i
- ~
..- w.
_.______,__m-
f
.4 l ASME CODE CLASS l l3 N-l
. STATION SERVICEWATER DEMINERAttZED WATER MAKEUP l
SYSTEM (SSWS)
SYSTEM (DWMS)
L MAKEUP-ECW EXPANSION TANK l'
QJ ECW 4
"x 1
g
+
l TO SAFETY g
LOADS I
ECW PUMP FROM SAFETY I
LOADS Tl
--4iEi:
+
ECW PUMP DMS10N 1 DMS10NAL SEPARATION DMSION 2 NOTE A i-l ASME CODE CLAGS I STATION -
DEMINERALIZED WATER MAKEUP U
ES#3) -
SYSTEM (DWMS)
MAKEUP ECW EXPANSION TANK i
QJ ECW 4
"x 1
llEill
+
1.3i TO SAFETY I
i LOADS FROM SAFETY k
LOADS ESS N a
Chit t rq NOTE- '
f ECW PUMP j
A.' A REMOVABLE SPOOL PIECEIS LOCATED ON EACH STATION SERVICEWATER SYSTEM MAKEUP LINE T EACH ECW EN'ANSION WE -
FIGURE 1.9.8.1 ESSENTIAL CHILLED WATER SYSTEM
SYSTEM 80+
1.9.8.1 ESSENTIAL CillLLED WATER SYSTEM ITAAC SUPPORTIVE INFORMATION 1.
Amnlifyine Information ITAAC 3 Confirmation of the ECWS capacity to meet the cooling load demands of the fullload of the essential llVAC chilled water coils will be performed based upon the as built ECWS serviced components and measured flow rates. The analysis will be based on t
the following:
4 ECWS flow to llVAC coollng coils Measured ECWS/NCWS chiller outlet temperatures Measured normal chilled water flow to the essential chilled water heat exchanger Vendor heat exchanger data 2.
Relationshin of ECWS ITAAC to the Safety Analysis l
N/A 3.
Relationshin of ECWS ITAAC to PRA ne PRA assumes that the Essential Chilled Water System is available for control room habitability and equipment qualification. The PRA also assumes that the ECWS is divisionally separate.
4.
CESSAR DC Chanter 14 Tests Anplicable to ECWS ITAAC See CESSAR DC Section 14.2.12.1.77 l
e l
.1.9.8.1 -12893 j..
L
~
4.
fiyElst 80+=
1.9.8.2 NORMAL CIIILLED WATER SYSTEM Design Description 1hc Normal Chilled Water System (NCWS) is a non-safety system. It is a closed loop chilled water system that serves non safety related ilVAC cooling loads.1hc NCWS is a subsystern of the Chilled Water System (CWS) and provides chilled water to connected air handling units.
A basic configuration for the Normal Chilled Water System is shown in Figure 1.9.8.2.-i
'the Nortnal Chilled Water System is made up of two divisions. The system consists of chilled water refrigeration units, chilled water circulation pumps, expansion tanks, control valves, instrumentation, and piping.
The two NCWS divisions are connected through manually operated valves. The cross connection allows one NCWS to provide chilled water to the other division.
Inspections, Tests, Analyses, and Acceptance Criteria Table 1.9.8.21 specifies the inspections, tests, analyses and associated acceptance criteria for the Normal Chilled Water Systern.
i T
1.9.8.2 1-01 28-93
.m.--
.mc..
,,.-,,,..,.e-,
--_y-i....- -.
m..~-..mn
SYSTEM 80+
TABLE L9.8.2-1 D
^
NORMAL CHILLED WATER SYSTEM Inspections. Tests. Analyses. and A&ch Criteria Certified Desies WR :
I=-:d':2 Tests. Anakses Acceptance Cdteria L
A basic conSguration -
of the L
Inspections of the as-buik system L
The as-buik @ of the Normal Chilled Wa:er System is configuration will be performed.
Normal ChiBed Water System is in shown in Figure L912-L accordance with Fgure L9A2-1 for the compoacnts and equipment shown.
2.
The two NCWS dhisions are 2.
Inspections of the as-buik sprem 2.
Manually operated vahes are-connected through manually configuration win be performed.
prmided in the piping that connect operated vahrs.
the two NCWS J;.h 9
k l
J
.1.9.8.2
- 2-1-28 93 i '
9 g-r
...es v
4 v.a en s
--,--s
,--n,
,,-c.-
w---a--r
DEMINERALIZED WATER MAKEUP SYSTEM (DWMS) MAKEUP l
f NCW EXPANSION TANK kJ NORMAL g
TO t
CHILLER y
Q
+.
I NON-SAFE'(
l LOADS AND O
NCW PUMP j
ESSENTIAL FROM NON-SAFETY j
CH:LLED
( WATER HX LOADS AND
^^
E HLLER ESSENTIAL CHILLED f WATER HX k
+
Q DIVISION 1 NCW PUMP DIVISIONAL SEPARATION DIVISION 2 NOTE A DEMINERALIZED WATER MAKEUP SYSTEM (DWMS) MAKEUP NCW EXPANSION TANK l
NY l
NORMAL g
g e
TO CHILLER s
,[
[
I NON-SAFETY l
O NCW PUMP SE FROM NON-SAFETY
(
f I
NCW PUMP NOTE:
A. THE TWO NCWS DIVISIONS ARE CONNECTED TO EACH OTHERTHROUGH NORMALLY CLOSED AND MANUALLY OPERATED VALVES.
FIGURE 1.9.8c2
- NORMAL CHILLED WATER SYSTEM
l SYS1EM 80+"
1.9.8.2 NORMAL CIIILLED WATER SYSTEM ITAAC SUPPORTIVE INFORMATION 1.
Amnlifyine Information l
N/A 2.
Relationship of NCWS ITAAC to the Safety Annhsis N/A 3.
Relationshin of NCWS ITAAC to PRA N/A l
4.
CESSAR.DC Chanter 14 Tests Annlleable to NCWS ITAAC See CESSAR.DC Section 14.2.12.1.77 t
l 1.9.8.2 - 01-28-93.
fiLTIEM 80+"
1.9.10 EQUIPMENT AND FLOOR DRAINAGE SYSTEM Design Description ne Equipment and Floor Drainage System (EFDS) segregates and transports liquid wastes to the Liquid Waste Management System (LWMS)
Liquid wastes, valve and pump leakoffs, tank overflows, and tank drains are collected by the EFDS. %c equipment and Door drains are separated according to waste types (equipment drains, Door drains, chemical wastes, and detergent wastes) to enable routing to the appropriate LWMS subsystem. Separate drain headers are provided for cach drain type so that different types of liquid wastes are not mixed.
He drainage and collection systems used for radioactive liquid wastes are separate and not connected to the systems used for non radioactive waste.
Two surnps are provided in the containment building to collect liquid waste. The containment floor drain sump is located in the holdup volume to collect floor drain wastes and leakage. He reactor cavity sump is located below the reactor vessel where no leakage is expected under operating conditions. %c containment floor drain sump and the reactor cavity sump are provided with instrumentatien to detect unidentified leakage inside containment.
The reactor building subsphere is divided into four quadrants each with a separate indepenLent sump and safety rclated sump pumps and instrumentation to collect leakage from the Engineered Safety Features (ESP) pumps and floor drainage within its respective quadrant. The sump pumps are ASME Code Class 3, Category 1 and are poweled from their respective Class 1E busses Floor drains in the Nuclear Annex are divisionally separated having no common drain lines between divisions. Each division's Door drains are directed to either the radioactive floor drain sump or the non. radioactive floor drain sump.
A separate CVCS area Door drain sump is provided in each division to collect and transport Door drainage to the floor drain waste tanks. Separate equipment drain sumps are provided in each division to collect CVCS equipment drainage in the Nuclear Annex.
Safety related backwater check valves are provided in drain lines from areas containing safety-related equipment to prevent backflow and flooding of those areas.
The safety related check valves are fabricated in accordance with ASME Code Class 3 and Seismic Category I requirements.
1.9.10 01 28-93
1 i
SYS'IV.M 80+"
Inspections, Tests, Analyses, and Acceptance Criteria Table 1.9.10-1 specifics the inspections, tests, analyses and associated acceptance criteria for the EFDS.
)
i 1.9.10 0I-28-93
1 4
SYSTEM 80+
TABLE 1.9.10-1 EOUIPMENT AND FLOOR DRAINAGE SYSTEM Inspections. Tests. Analyses. and Accendance Criteria Certi5ed Desien Comenkuneet 12----
- =_ Tests. Analyses Accendance Criteria L
A basic configuratk.a for the EFDS L
1A of the as-built system L
& as-built 4 - 4 son of the is shown. in Figures L9.101 and umfis-dion will be performed.
EFDS is in a ba with Figures
- L9.10 2.
L9.101 and L9.10-2 for the w-gas and equipment shown.
I 2.
ASME Code portions of the EFDS 2.
- A pressure test will be conduced 2.
The resuks of the pressure test of retain their integrity under ~mternal on those portions of the EFDS ASME Code portions of the EFDS pressures that will be + a.ced required to be pressure tested by conform with the r@ments in during service.
the ASME Code.
the ASME Code Section IIL I
3.
The EFDS segregates the plant's 3.
System testing will be conducted 3.
Each equipment - and floor drain i'
liquid waste wQ to waste -
after installation for each drain and drain header are properly l
type, activity, and quality and.
path.
roated to their designed routes the liquid waste to the destmation.
LWMS subsystem i M.L.5 that class of waste.-
' 4.
Reador Building Subsphere safety.
4.
A' test of the power availability to 4.
& Certified Design Commitment
. related sump pumps' and associated the Reactor Building S42.s-is met.
instrumentation. and controls are safety-related sump pumps and powered ' from : their ' respective -
associated instrumentation and riau 1E ' busses and can be controls will be conducted with powered from the diesel generators.
power supplied from. the
}
p-z.scatly - inemile d electrical F
power busses and diesel generators...
-l 9
L t
5 w.
-3 n2 e
~
SYSTEM 80+
TABLE 1.9.10-1 (Continued)
EOUIPMENT AND FLOOR DRAINAGE SYSTEM -
facspections. Tests. Ammivses. and A.-M*mce Criteria Certified Desien Co==itment
. Ira-3': = Tests. Analyses Accustance Qiteria 5.
Safety-related baciflow check S.
An inspectics will be performed on 5.
Backflow check valves areinstaBed valves are provided - in drain lines the backflow chrek vahts.
in drain Enes from areas =* "-
from areas-containing safety--
safety related c4cr.L related equipment.-
?-,.
6.
Floor drains are dhisionally 6.
An inspection will be conducted on 6.
A divisional wall separates floor separate hadng no common drain each division of floor drains.
drains in each dhision lines between dhisions.
i jt 7.
EFDS ' instrumentation
~.amed=s 7.
Inspedion of the Control Room for 7.
The instrumentation h
1-and alarms - shown in Figures the availabihty of instrumentation and alarms shown 'in Fgures 1.9.10-1 and 1.9.10.-2 are available indications and alarms identdied in 1.9.10-1 and 1.9.10-2 exist or can l
in the Control Room.
the Certified Design Commitment be retrieve in the Control Room.-
wiH be performed.
a i
l-
_ 1.9.19 -
- 41-28 93-
.< O
- v w :- m e
~
w
, ~ _,
II, o
_ _ _ _ _ w[_,
h 9I u
lll C
3 il
~
D D'
2 ji C-
\\ _s 3
g i $-
e
_____a y
1n' ljt C 3
1 I
"i C ygi l 3
n 1
ge,
&,3 m
OUADMNT A O
l'*"' '.u'u***I I
't" " """ " "I h-~ l g
i u
f Q
C Wunsign
'i l! "S'.O 3
llo l
gg
,y 23 wa b
y l Asatt C006 CLAS $l m.
l ASME CODE CLA$Sl y FL DMA HOR] $
Q alsy '0 o
+
ep z
g; ta 33,y 34_ gnea;en yan l5 l!g"5.,0 g{l 3
o e
sg
'B e QUADRANT 5 1
l Atma CODE CLA44l f F OCRDRAW M ]
78" u.pu j
5 REl 'O
-3 m
E' I!g -"tc 3s b.4---?aai:2= gil w.
a 1
2 3
l ASME CODE CLASS l P'
-~~
l AsME CODE CLASS l r FL DRA N HOR) e
. g "Xh 3y, ll b
i 1
~A
-2<-
a es--6, sui _ s
+M--A
.MrL--a 1-'cJr,
M k
J-----,-6M A-
-Wa.
ud-----a-a<L N -s
-u-e A1s._J
+4r A
L--
am n__x_
ee 5E I
"5 q'
glA
$8
,-'=m m
,q c 0
.I CM l!
sr 35 q'
--==%
,q CM ll l8
~
E o ilI 8
5m a5a I
m.
I'g l
-ig;'=:=c %
,q C 0
e o si I
=
3 I,
W K
gi 8--ar.s9 %
,q t 0
lg
- 8 C
Q is k
f f se s-Ee
!5
.m t
s=
qE O ill lf
-,== %
c o
gg y
elj li I
=
q-gg
,t 0 II g
--=:=e m
,q t
0 fj E"
s2 ij m
5 h,!
,,C-O =lI
'a q,C 0 illI
-- l$; t:"JWem 5
8 m!
lll!
- @qc o l g, 1
l' 2
--a 49%
c o l1
SYS'IEM 80+"
1.9.10 EQUIPMENT AND FLOOR DRAINAGE SYSTEM SUPPORTIVE INFORMATION 1.
AmaMine Information Not Applicable 2.
Relationshin of EOUIPMENT AND FLOOR DRAINAGE SYSTEM ITAAC to the.
Safetv Analysis i
Not Applicable 3.
Relationshin of EOUIPMENT AND FLOOR DRAINAGE SYSTEM ITAAC to EBA No direct downward flowpath that will alkiw drainage of radioactive liquids from-containment.
4.
CESSAR DC Chanter - 14 Tests Applicable-to EOUIPMENT AND I'LOOR '
DRAINAGE SYSTEM ITAAC Refer to CESSAR DC Section 14.2.12.1.84.
P l~
l-
- 1.9.10.01 28 93 l'
4..
.. _. _,. _. -., - _ _,..,.~...,._;.........._,.
_,..a.,
..~.....
.., _.. - - -. _.., ~. _, _
SYSTEM 80+"
1.9.11 CIIEMICAL AND VOLUME CONTROL SYSTEM Design Description The chemical and volurne c4mtrol system (CVCS) does not perform accident mitigation or safety functions. Portions of the system form part of the reactor coolant prer.sure boundary.
Figure 1.9.11 1 shows a simplined systern configuration.
The CVCS includes components, piping, instrumentation and controls to remove coolant water from the RCS, pass the coolant water through Ulters and ion exchangers, add and remove soluble boron from the coolant, provide spray water to the pressurizer, provide water to the RCP seals, collect controlled RCP seal bleedoff, provide water to the spent fuel pool and return water to the RCS.1he letdown portion of the CVCS consists of piping and comimnents from the RCS to the Volume Control Tank (VCl*). The components include the regenerative heat exchanger (RllX), where the letdown flow is cooled by chargine flow returning coolant to the RCS, the letdown heat exchanger where the letdown flow is cooled by component cooling water, parallel letdown flow control valves, parallel letdown flow orifices, filters and ion exchangers, and isolation valves. -1he volume control tank (VCT) receives the purified letdown Dow. A separate linc drains controlled bicedoff from the RCP seals to the VCT, lhe charging portion of the CVCS consists of piping and components from the VCr to the RCS. *lhe components includa two parallel pumps, two parallel control valves,-
isolation and check valves. A branch line directs some of the charging Dow to the RCPs for ser.1 injection.1tc remaining charging flow is heated by letdown flow in the RilX, then goes to the RCS. A line from the chemical addition unit connects to the charging line upstream of the RilX. The auxiliary spray line branches from the charging line downstream of the RilX and is connected to the RCS pressurizer spray line.
The reactor drain tank (RDT) collects drainage, leakage, and relief Gulds from the RCS (except the pressurizer safety valve discharge) and portions of systems within the reactor coolant pressure boundary (except the shutdown cooling system low temperature overpressure relief valves).1hc contents of the RDT can be pumped to the boron recovery and recycle portion of the CVCS. Dorated and unterated makeup water can be supplied to the VCr. Horated water can also tm supplied to the PCPS.
The CVCS is built to the ASME Code Section 111 Class requirements shown on Figure 1.9.11 1. Components, piping and supports classified as ASME Code Class 1, 2, or 3 are Scismic Category 1, 1.9.11
-1 01 28 93
SYSTIG180 +"
'the letdown line contains valves which close upon receipt of a safety injection actuation signal (SIAS) or by a containment isolation actuation signal (CIAS). 'Ihe RCP controlled bleedoff line contains valves which close upon receipt of a containment spray actuation signal (CSAS). 'the RDT drain line and RDT reactor water makeup line contain valves which close upon receipt of a CIAS.
Control room alarms are provided for high RilX exit temperature, low letdown line pressure, and low VCT level. Valves in the letdown line are closed automatically on high temperature by a signal from a temperature indicator in the letdown line outside containment.
Inspections, Tests, Analyses, and Acceptance Criteria Table 1.9.11-1 specifies the inspections, tests, analyses and associated acceptance criteria for the CVCS.
1.9.11 2-01 28 93
G SYSTEM 80+=
TABLE L9.11-1 CHEMICAL AND VOLUME CONTROL SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Centined Desien Comunitament-Im=-:f:s Tests. Analyses Acceptance Criteria L
A basic configuradon for the L
Inspections of the as-built CVCS L
He as-built CVCS configuration is' CVCS is shown in Figure L9.11-L configurath will be performed.
in accordance with Figure 19.11-1, (NOTE 1)
- for the components and eqmpmcat shown.
2.
ASME Code portions of the CVCS 2.'
A pressure test will be conducted 2.
. ~Ihe rsdts of the pressure test'of.
retain their integrity under internal on those portions of the CVCS ASME Code porties of ibe CVCS pressures that will be experienced required to be pressure' tested by-conform with the requirementsL in.
during senice.,
the ASME Code.
the ASME Code Section 'IIL He letdown i ne is isolated by a 3.a) Tests.will ? be performed using a 3.a) ' The ' two CVCS letdown isolation 3.a) li safety. ' injection actuation 4gnal simulated - SIAS.. The response of vahes inside containment close (SIAS).
the letdown isolation vahrs will be upon receipt of a SIAS.
observed.
b) He letdown line is isolated L by a b) Tests will ' be performed using' a b) : The two ' letdown line containment.
containment isolation actuation sig-simulated GAS. %c response of isolation valves ' close upon receipt nal (CIAS).
the letdown containment - isolation of a CIAS.
valves will be hd i -
c) The RDT. ; drain line 'and RDT c) Testi will be performed ' using a c) The - RDT. drain ~ and RWM ' line simulated : CIAS.,%c response of remote-opcrated. containment reactor water makeup line are
' isolated byf a CIAS.
the RDT drain : and RWM line isolation 7alves close upon receipt containment isolation valves lwill be of a CIAS.
observed.
i L
-1.9.11.- 01-28-93' t,
..+;.
w m
+.<
w e-
.w v.
--."u,-
4,v v
-m ys %
.+
EYSTEM 80+=
TABLE 1.9.11-1 (Continued)
CHEMICAL AND VOLUME CONTROL SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria -
Certified Desias Counmitment inspections. Tests. Analyses Acceptance Criteria 3.d) The RCP seal' controlled bleedoff 3.d) Tests will be performed using a 3.d) he RCP seal controlled bleedoff.
- t line. is isolated by a containment
. simulated CSAS. The response of line isolation - valves close upon spray actuation si: pial (CSAS).
the RCP seal controlled bleedoff receipt of a CSAS.
line isolation - vahrs ' will be ol sened.
~ dications I
' 4. -
CVCS instrumentation indications '
4.
Inspection of the Control Room for 4.
The instrumentation m
and alarms shown on = Figure the availability of instrumentation and alarms shown on. Figure 1.9.11-1 are available in the' indications and alarms identdied in 1.9.11-1 exist or can be retrieved Control Room.
Controls ' are the Certified Design Commitment in the Control Room.
CVCS available la the control room tu :
will be performed.
Tests will be controls operate as specified in the
~
start and stop the charging pumps, -
performed _ using the CVCS controls Certified Design Cn==;t=rar and.open. and close the CVCS in the Control Room.
remote-operated vahrs shown on Figure.1.9.11-1 5.
The letdown line has valves which 5.
Tests will be performed using a 5.
A letdown line isolation valve close upon, receipt of a high:
simulated - a high temperature inside containment.. closes on temperature signal.
signal.
The response of the
' receipt of a high temperature letdown line isolation vahrs will be signal.
observed.
1.9.11 01-28-93
- u...,
INSIDE
{
OUTSIDE CONTAINMENT CONTAINMENT HIGH TEMP
.mTE umXWVM UNE
\\
I ON HiGH TEMPERATURE CLOSES CIAS CLOSES l
STAS 8
CIAS II CLOSES T
- CLOSES s
g LETDOWN REGEN LETDOWN FROM 7 -
HEAT EXCH.
HEAT EXCH.
PURiFICATON
~
l
-I I~
{
ROS ON L
g h
V h
s EXC"^"GERS
{ASME COOf CLAM g
E V
h COO A ER P$8 II F2n LEVEL CONTROL
.--.l A
TO RCS
+
- ADDr N Nff EB I
PZR L@ N 1
m h
3 AUX)UARY
$ PRAY M+ +
VOLUVE' a
T t
OONTROL TO PZR TANK EE u
pu PS
^""
A C
i gg gg seat I
u E'OE "M lTF as XC, i
OTHER W RC s
CLOSES
--D=-
]
FROM FROM CONTROLLED I
NR ME W BLEEDOFF
?
MN "N
FROM ONE
+
l WATER -
(BORATED)
{]
ggh]
(DEMN)
R S
REACTOR sp3.-->- DRAIN TANK i
BORON RECOVERY g
I Eb AND RECYCLE FROM REACTOR eg MAKEUP WATER
' FIGURE 1.9.11-1
+
SYSTEM 80 CHEMICAL AND VOLUME CONTROL SYSTEM
SYSTEM 80+"
1.9.11 CHEMICAL AND VOLUME CONTROL SYSTEM ITAAC.
SUPPORTIVE INFORMATION 1.
Amplifyine Information System
Description:
Relationshin of CVCS ITAAC to the Safety Analysis Basis: The letdown line is isolated on a SIAS.
ITAAC: ITAAC 3 confirms that the letdown line is isolated on a SIAS.
Basis: RCP seal bleedoff is isolated by a CSAS.
ITAAC: ITAAC 3 contirms that the RCP seal bleedoff is isolated by a CSAS.
3.
Relationshin of CVCS ITAAC to PRA None 4.
CESSAR-DC Chanter 14 Tests Annlicable to SFSR ITAAC Preoperational Tests: CESSAR-DC Section 14.2.12.1.5 through 14.2.12.1.20, 14.2.12.1.60 1.9.1.2 01 28-93
=
SYSTEM 80+"
1.9.20 COMMUNICATIONS SYSTEMS Design Description The Communications System provides communications between areas within the plant site including all vital areas of the plant. In addition, the Communications System provides means to communicate to plant personnel and offsite utliity and regulatory officials.
The Communication System consists of the following subsystems:
1)
Intraplant Portable, Wireless Communication System 2)
Intraplant Private Automatic Business Exchange (PABX) Telephone System 3)
Intraplant Public Address (PA) Sptem 4)
Intraplant Sound-Powered Telephone Systems 5)
Offsite Communications System 6)
Radio Communications System The Intraplant Portable, Wireless Communication System provides the primary means of voice communication capability between plant personnel. In addition to portable
_j and wireless transmitter / receivers, the-system includes base stations, antennae, amplifiers and/or repeaters. _ Specialized system portable and wireless transmitter / receivers are provided for such applications as respirators and/or underwater diving work.
A Private Automatic Business Exchange (PABX) Telephone provides interplant communications. The Public Address (PA) System consists of audible speakers at locations in the plant.
'Ihe Sound-Powered Telephone System include the following circuits:
1)
Maintenance Circuit - consists of phone jacks at locations in the plant which can be patched together to establish communications between areas.
2)
Refueling Circuit - consists of phone jacks located in the areas required for l
refueling operations.
3)
Emergency Circuit - consists of phone jacks connecting areas of the plant where shutdown operations are conducted.
1.9.20 01-28-93
M
_a,r 1-4 J
4-a 4-1.4i e
4,Ac.
<h SYMM 80+"
Normal offsite communication is provided by the commercial telephone system and the utility private network. Utility private network lines are connected to specific telephones located in designated areas of the plant. De telephones connected to the utility private network are color coded to distinguish them from the PABX telephone system. The following networks to offsite entities are included as part of the utility private network phone system:
1)
Emergency Notification System (ENS)- provides a communications link to the Nuclear Regulatory Commission (NRC) 2)
Health Physics Network (IIPN) provides a communications link to NRC health physics personnel 3)
Ringdown System - provides a communications link to offsite agencies.
%c Portable, Wireless Communications System, PABX, PA, Sound-Powered Telephone System, and Offsite Communications System are separate networks.
However, the PABX is connected to commercial telephone system and to the utility private network. The Portable, Wireless Communications System, PABX, PA, and Sound-Powered Telephone System receive power from Class IE buses.
The Radio Communications System consisting of a security radio system and a crisis management radio system ir provided.
Inspections, Tests, Analyses and Acceptance Criteria Table 1.9.20-1 specifies the inspections, tests, analysis and associated acceptance criteria for the Communications Systems.
i 1.9.20 2-01-28-93
TABLE 1.9.20-1 SYSTEM 80+
COMMUNICATION SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Certified Design Commitment Inspections. Tests. Analyses Acceptance Criteria l
1.a) The Por t ab1e, Wi r eI e s s 1.a) Test of the Portable, Wireless 1.a) Voice transmission and reception Communication System provides Communication System will be between Ioeations are intraplant voice communication performed with plant background accomplished.
capability between plant personnet noise present.
b) Specialin :
system portable and b) Tests of respirator and diving b) Voice transmission and reception wireless transmitter / receivers are transmitters / receivers in their using respirator and diving as corresponding emironment will be transmitters / receivers in their provided for such applicadons respirators and/or underwater performed.
corresponding emironment are accomplished.
diving work.
2.
The Private Automatic Business 2.
Tests of the PABX Telephone 2
Voice transmission and reception Exchange (PABX) Telephone System from all terminals will be between plant terminals are System provides intraplant performed with plant background accomplished.
communications.
noise present.
3.
The Public Address (PA) System is 3.
Tests of the PA System will be 3.
Voice broadcast in areas where PA
)
capable of alerting plant personnel performed with plant background speakers are 1oeated is by means of audible speakers noise present.
accomplished.
located throughout the plant.
l 4
The Sound-Powered Telephone 4.
Inspections and tests of individual 4.
Interconneetion and System include the following circuit phone jacks will be communication using the Sourd-circuits -
performed with plant background Powered Telephone System are noise present.
accomplished in each circuit.
a)
Maintenance Circuit - consists of phone jacks ' at kx:ations in the plant which can be patched together to e s t a b li s h communications between areas.
01-28-93 1.9.20
SYSTEM 80+
TABLE L9.20-1 (Continued)
COMMUNICATION SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Certi5ed Desima Commmitament I==we% Tests. Analyses Acceptance Criteria 4.b) Refueling.. Circuit - consists of phone' jacks'. located in the areas required for refueling operations j
c)
Emergency Circuit. -' consists of phone jacks ~ connecting areas of the
- plant' where. shutdown operations l
are conducted. --
. 5.
The following _ networks to offsite 5.
Inspections and tests of the utihty 5.
Utility private network telephones.
i entities are included as part' of the private network ' phone terminals at are color coded to distmguish them utility private - network phone' installed locations will be f rom. PABX' telephones.
system:
performed with plant background Communication links with > the noise present.
entities identified in the Certified a) Emergency.. Notification System -
D e si g n Commitment-is.
provides a communications link to.
demonstrated
_us' g the : utility m
the Nueiear ReguIatory private network phone term'mak CommissionL (NRC) t
~ b) L Health Physics ; Network.. (HPN) -
provides / a communications.. link to.
NRC heakh physics ypersonnel;
~)
Ringdown - System"... : provides a c
communications link ! to.offsite 1 agencies.-
. The telephones' conneried to.the utility : private J network. 'are ' color--
' coded to distinguish them from. the -
PABX telephone.' system.1 1.9.20 '
01-28 93-
. q a,
.14
TABLE 1.9.20-1 (Continyed)
SYSmM 80+
COMMUNICATION SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Certified Desien Commitment Inspections. Tests. Annivses Acceptance Criteria 6.
The Portable, Wireless 6.
Simulate failure of each 6.
Each system in the Certified spr m.
Design Commitment operates Communications System, PABX, communication e
without reliance on operation of PA, Sound-Powered Telephone another communication system.
System, and Offsite
('nmmunications Sy stem are separate networks.
7.
The PABX is connected to 7.
Test of the PABX Telephone 7.
The capability to connect to PABX commercial telephone system and System will be performed.
telephone terminals to the commercial telephone system and to the utility private network.
to the utility private network is demonstrated.
8.
The
- Portab1e, Wi r eie ss 8.
A test of power availability to the 8.
The Certified Design Commitment Communications System, PABX, systems described in the Certified is met.
j j
PA, and Sound-Powered Telephone Design Commitment will be System receive power from Class conducted with power supplied i
from the yrmanenify installed IE buses.
electrica! power buses.
9.
The Radio Communications System 9.
An inspection of the security radio 9.
Two way communication is demonstrated between participating consisting of a security radio system and the crisis management system and a crisis nanagement radio system will be performed.
entities for each sysem.
radio system is provided.
Tests of each radio system will be performed.
01-28-93 L9.20
il SYSTEM 80+"
'1.9.20 COMMUNICATIONS SYSTEMS SUPPORTIVE INFORMATION 1.
Amplifying Information N/A 2.
Relationshin of CS ITAAC to the Safety Analysis N/A 3.
Relationshin of CS ITAAC to PRA The PRA assumes that the Communications Systems exist.
4.
CESSAR-DC Chapter 14 Tests Anplicable to CS ITAAC See CESSAR DC Section 14.2.12.1.87 a-9
~ 1.9.20
-1 01-28-93
SYSTEM 80+"
1.9.21 LIGIITING SYSTEM Design Description he Ughting System provides illumination at locations in the plant and the plant site, including the vital areas. The Ughting System is composed of three subsystems:
Normal Ughting System, Security Lighting System, and Emergency Lighting System.
He Normal Lighting System provides general illumination at locations in the plant.
The Normal Lighting System receives power from non. Class IE buses.
The Security Lighting System provides the illumination in isolation zones and the outdoor areas within the plant protected perimeter. The Security Lighting System receives power from the permanent non-safety buses. Portions of the Security Lighting System essentiel to maintaining plant protection are powered from an uninterruptible power supply. The Security Lighting System provides a minimum illumination of 0.2 foot-candles when measured horizontally at ground level.
He Emergency Lighting System provides illumination in the vital areas including the Control Room, Technical Support Center, Operations Support Center, the Remote Shutdown Panel Room, the stairway which provides access from the Control Room to the Remote Shutdown Panet Room, Sample Room, Ilydrogen Recombiner Rooms, routes for personnel passage and egress, and areas where operator access is required post-accident or hazard. Emergency lighting in the Control Room is installed so that alternating lighting fixtures are fed from separate Class 1E divisions. The system provides a minimum illumination level of 10 foat-candles in areas of the plant where emergency operations are performed. For other areas of the plant covered by the Emergency Lighting System, the system provides a minimum illumination level of 2 foot-candles.
The Emergency Lighting System is designed as Seismic Category I. Components of the Emergency Lighting System are_ powered from Class 1E buses. The Emergency Lighting System employs two illumination methods:
A)
AC fixtures powered l~ rom Class IE AC power sources, and B)
DC self-contained, battery-operated lighting units.
The self-contained, battery-operated lighting units have the following provisions to function without AC power:
A)
Battery life of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> at rated load, and B)
Battery loading of each unit not greater than 80% of the rated capacity with additional derating for temperature variations and 1.9.21.1 28-93 i
SYSTIT,M 80+"
C)
A shutoff time delay to continue operation for a pre-set time interval following restoration of AC power and D)
Capability to lock the power supply breakers which supply the units in the
" energized" position.
Inspections, Tests, Analyscs, and Acceptance Criteria s p eMis Table 1.9.21-1 provides the inspections, tests and/or analyses and associated acceptance criteria.
1.9.21 1-28-93.
SYSTEM 80+
TABLE 1.9.21-1 LIGHTING SYSTEM Inspections. Tests. Analyses, and Acceptance Criteria Certified Design Commitment Inspections. Tests. Analyses Acceptance Criteria 1.
The Normal Lighting System 1.
Inspections of the installed Normal 1.
The Norreal Lighting System has provides general illumination at Lighting System will be performed.
been installed.
locations in the plant.
2.a) The Security Lighting System 2.a)
Inspections of isolation zones and 2.a) The Security Lighting System provides illumination in isolation plant outdoor areas within the plant maintains illumination levels t 0.2 zones and outdoor areas within the protected perimeter will be foot-candles when measured plant protected perimeter.
performed.
Tests to measure horizontally rt ground level in illumination levels will also be isoktion zones and ou' door areas performed.
within the plant protected perimeter.
b) The Security Lighting System b) A test of power availability for the b) The Certified Design Commitment receives power from the permanent Security Lighting System will be is met.
non-safety buses.
conducted with power supplied from the permanently installed electrical pour busses.
c)
Portions of the Security Lighting e) A test of po.ver availability for the c)
Portions of the Security Lighting System essential. to maintaining Security Lighting. System will be System are powered from a battery plant protection are powered from conducted.
power source.
an uninterruptible power supply.
3.a) The Emergency Lighting System 3.a) Inspedions of areas as specified in 3.a) Emergency lighting is installed in illuminates vital areas as described the Design Description.
the areas specified in the Design in the D sign Description, routes Description.
for personnel. passage. and egress, and other areas where operator access is required post-accident or hazard.
1.9.21 01-28-93
SYSTEM 80+
TABLE 1.9.21-1 (Continued) -
LIGHTING SYSTEM Inspections. Tests. Analyses.'and Acceptance Criteria i
Certified Desien Comunitament -
. Ir=S : Tests.' Amalvses Acceptance Criteria jl 3.b) Components of the - Emergency
. 3.b) ' A test of power availability for the 3.b) The Certified Design. Commitment
' Lighting' System are powered from
' Emergency Lighting System will be is met.
Class 1E buses.
. conducted with power - supplied '
from. the. permanently installed
' electrical power - buses.
. c) L The ? Emergency < Lighting System
> c).. Tests to measure illumiution levels c) The Certified Design Commitment.:
provides ; a minimum. illumination.
' will be _ performed - at areas spec-is met.
level of '10: foot-candles - in those.
. ified in the Design Description.
areas of the plant where emergency operations : L are J performed.'
'For other areas of the plant. covered by
. the Emergency Lighting-System,
- the system provides a minimum' it-lumination ' level: of 2 foot-candles.
d) Emergencyi lighting "in the Control.
d). A test of power availability for the d) Alternating lighting fixtures in the
. Room ' is installed so that aher-Emergency. Lighting System will be -
control room j are powered :. from
.nating. lighting fixtures are conducted with' power supplied '
separate Class 1E divisions.
- powered; from. separate J Class '1E_
' from the permanently installed L divisions...
electrical ' power busses.
3
- 4..
Self-contained,1 : battery-operated 4.'
Inspection and tests of self-con.
4 The' Certdied. Design Commitment
'lightingD units of f the. Emergency [
.tained, ; battery-operated lighting is met..
. Lighting - System have the following.
units, i tests' - of i the1 emergency
- provisions 5 to function
- ' ithout AC.
lighting unit time delay, ~ and. in-w power:
L spections of1 the Kpower: suppiy
-breakers' lock ~mg - devices. will L be -
- performed.
't L9.21.
-4' 01-28.
x
~L : j, 4
ur-meo e n
a
+
-- n:-
e s.
c w
. w-
SYSTEM 80+
TABLE 1.9.21-1 (Coctinued)
LIGHTING SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Certified Desima Commitment Insocctions. Tests. Analyses Accestance Criteria 4
(Continued) -
a) Battery life: of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> at rated load.
b). Battery loading of. cach unit not greater-than-80%' of - the ' rated capacity ~ with. additional - derating -
' for temperature variations.
c) A shutoff time delay to continued operation for a pre-set. time interval following restoration ' of AC power.
d) Capability tolock the power supply -
breakers which ' supply the units in.
the " energized" position.:
~ 1.9.21 - 01-28 r.
+-
..-2,
,...., _ L.
' SYS'IV,M 80+"
1.9.21-LIGIITING SYSTEM ITAAC SUPPORTIVE INFORMATION 1.
Amplifying Information ITAAC 2 Security lighting will be provided in areas as detailed in the Security Plan. The Security plan will also detail which portions of the Security Lighting System are powered from an uninterruptible power sourec.
ITAAC 3 Vital areas, routes for personnel passage and egress, and other areas'where operator access is required post-accident or hazard will be detail' ' in plant emergency procedures and hazards analysis. The plant emergency piocedures and hazards analysis will identify areas where emergency lighting is needed.
2.
Relationship of IS ITAAC to the Safety Analysis N/A 3.
Relationship of LS ITAAC to PRA 4
=1 The PRA assumes that emergency lighting is available to allow operator access.
4.
CESSAR-DC Chanter 14 Tests Applicable to LS ITAAC See CESSAR DC Section 14.2.12.1.85 and 14.2.12.1.86 1.9.21.1-28 1 1
- l
y SYSTEM P0+"
1.9.22
- DIESEL GENERATOR SUPPORT SYSTEMS Design Descri 7lon -
l
'Ihe Diesel generator support systems described below are safety.related.
1)
Diesel Generator Engine Fuel Oil System - Storage tanks provide storage of no less t_han a seven day supply of fuel oil at full load plus a ten percent--
margin for performance testing and supplies the fuel oil to the.!!esel engine.
Each day tank has a capacity to maintain at least 60 minutes of operation from the level where fuel oil is automatically added to the tank.
2)
Diesel Generator Engine Cooling Water System - provides cooling water to the diesel engine.
3)
Diesel Generator Starting Air System - provides start capability for the diesel generrtor engine by using compressed airL to rotate _ the; engine until combustion begins and the diesel engine accelerates under its own power.
The starting air storage capacity for each diesel generator engine allows at--
least five successful engine starts without use of the starting air compressors. -
4)
Diesel Ocnerator Engine Lube Oil System - delivers lubricating oil to the:
diesel generator engine.
5)
Diesel Generator Engine Air Intake And Exhaust System - supplies air for -
combustion to the diesel engine and removes engine exhaust.
- 6) -
Diesel Generator Building Sump Pump System - removes leakage and equipment drainage from the diesel generator building.
Independent diesel generator support systems are provided for each diesel generator.
L Bask: configurations for the diesel generator support systems are shown in Figures -
1.9.22.1 through 1.9.22.6.
The ASME code classifications for the pressure retaining portions of the diesel generator support systems are - depicted in Figures.1.9.22.1' through 1.9.22.6.
j Components meeting ASME Code Class 3 requirements as depicted in Ihe figures are
{
safety related.
Components, piping, and supports classified as ASME Code Class 3 are Seismic Category I. Equipment that is designated'as safety relatediis qualified _ for ths environments where located.
Safety related components of each division of the diesel generator support systems are
- powered from their associated divisional Class 1E buses.
1.9.22' -
01-28 SYEIV.M 80+"
Inspection, Tests, Analyses and Acceptance Criteria:
Table 1.9.22-1 specifics the inspections, tests, analyses and associated acceptance criteria for the Diesel Generator Engine Fuel Oil System.
1.9.22 01-28 93
SYSTEM 80+
TABLE 1.9.22-1
' DIESEL GENERA 1DR SUPPORT SYSTEMS Inspections. Tests. Analyses. and Acceptance Criteria Certified Design Commitment Inspections. Tests. Analyses
_ Acceptance Criteria 1.
Basic " configurations of the : diesel 1.
Inspections of the as-built diesel L
The as-built configuration ' of the generator support systems are.
generator support systems will be diesel generator support systems shown in Figures ' 1.9.22.1 through conducted.
are in accordance with F'gures L9.22.6.
L9.22.1 ~ through 1.9.22.6 for. the ~
components and equipment shown.
2.
Independent diesel generator '
2.
Inspections, of the as-installed 2.'
A divisional wall separates support systems' are provided for diesel support systems will' be mechanical divisions of the diesel
= cach - diesel ' generator.
performed.
generators ed their associated support systems.
3.
Storage tank for-.each diesel 3.
An analysis will be prepared based 3.
The ' two storage ranh for each generator provide a combined upon as-procured diesel generator diesel generator meet the capacity storage:of no less than a seven day data.
requirements specified in the supply of fuel oil at full load, plus Certified Design Commitment.
a ten percent margin for performance. testing.
4.
Each day tank has - a capacity to 4.
Tests of the diesel generator will be 4.
The day tank provides _ at least 60 '
maintain at least 60 minutes. of performed : at full load ' conditions.
minutes of rnnning time for the operation from the level where oil diesel generator without - fuel. oil -
is automatically added to the tank.
being added automatically to the tank.
5.
The starting - air storage espacity 5.
Tests of the diesel generator engine 5.
The diesel generator engine starts
'or each diesel generator engine starting capabilities will be at least five consecutive times allows at - least five. sucessful performed.
without use of.' the starting air engine starts without ; use of the compressors.
- starting air compressors.
1.9.22 -- 01-28-93
I SYSTEM 80+
TAELE 1.9.22-1 (Continued)
DIESEL GENERATO3 SUPPORT SYSTEMS Inspections, Tests. Analyses, and Acceptance Criteria Certified Desien Commitment Inspections. Tests. Analyses Acceptance Criteria i
l
)
6.
The ASME code portions of.the 6.
A pressure test will be conducted 6.
Tne results of the_ pressure test of l
diesel generator engine support on those portions of the diesel the ASME portions of the diese!
systems retain their integrity under generator support systems required generator support systems conform internal pressures experienced to be pressure tested by the ASME with the requirements in the ASME during service.
code.
Code Section III.
I 7.
Safety re1ated componeats 7.
A test of power availability to the 7.
The Certified Design Commitment described in the Design Description diesel generator support systems' is met.
for each division of the diesel components described in the generator support systems are Design Description will be powered from their respective conducted with power supplied dhisional Class IE buses, from the permanently "mstalled electric power buses.
1.9.22 01-28-93
.-__.A
1 7
I As e cooe ctxssi ULJI s
N s
-e>-
FUEL Oil -
STORAGE TANK mj-an
--Es3-NC I
1 FILTER
~
RECIRCULATION PUMP NC l
TANK 7
I7
"-G lasue cooe etAssi UL. 2J M
i N'
_ M.
+
FUELott gn NC 1r STORAGE TANK E
l b
DAYTANK RETAINING WALL
~
Y RD FP L COVAECTION E
1 p
STRAINER wwoR-ORivEN FUEL OfL PUMP
- DIESEL y
GENERATOR ENGINE
+N 8
J STRAINER
,or,hGINE MOUNTED COMPONENTS ARE A. E
~
"^
CONSTRUCTEDIN ACCORDANCEMTH FIGURE 1.9.22.1
'"""* " ^" "'
DIESEL GENERATOR ENGINE FUEL OIL SYSTEM' (DIVISIONAL CONRGURATION)
p.= coo,cu n I-
- d. ".Lj e.-======.o.-
=
.m Z,
~
. _O_m.m
- 3..
MEA
=0+
-_=
a.e=~
- Uns*
CCWS
'" ="
b
~
1r Ccws DfV N N 1 p.= coom cu.s l L.*
".L.!
@M u-_--===,--
=
,m
=
.: n
= d,
.T_--
~
C I-I-s -'s-^." =04
-e I
w'" E a"*ma
~
If A
RDAN E
AR 387 FIGURE 1.9.22.2 DIESEL GENERATOR ENGINE COOLING WATER SYSTEM
[ASME CODE Ct.ASSI 3
l TO DESEL :
F LTER
-COMPRESSOR AFTERCOOLER FILTER / DRYER Q
AIR RECENER I
y l
GENERATOR UNIT TANK
+
ENGINE
+
lASME CO7E CLASSI
[ N,,,y
. ) To -Sm a' ::T
~" =T""
N-E
=
mLeR c.,, ESSER
( ENomE-
+
DNISION 1 IASME CODE Ct ASSI i, TO -se.
^
MLTER COMPRESSOR UFRT TANK
+
[ ENGINE -
+
i IASWE CO M CLASSI E d T
3
} TO D:ESEL COMPRESSOR AFTERCOOLER FILTER / DRYER MR RECEIVER jGENERATOR FILTER l
UNIT
+
TANK
+
ENGME 6
DMSION 2 FIGURE 1.9.22.3 DIESEL GENERATOR ENGINE STARTING' AIR SYSTEM L
a DIESEL N
GENERATOR
+
ENGINE h
nLe JL M
PP.ELUBE OtL 1t 1P ENGNE-DRIVEN POUP LUBE Ott PUMP
-[
7 LUBEOtt
$ UMP TAhX r
nna T'
DMSON1 l Aswe coot cuss l l Asme cuoe cuss ]
= =,)
= g)O.L- -
TO USED LUBE 1F m
z SYSTEM D=
l DIESEL
%N 8
N GENERATOR
+
ENGINE l
i em
=
JL W
PRELUBEOIL 1t 1P ENGmE-OfWvEN PUMP LUBE Cst PUMP
,,,,,,, 5 LUBE OL f
D;h SUMP TANK r
nm DMSON 2
'I m.
I AseIE CODE Ct. ASS l lA&MG CN CMSS l CONsTm,cno - E-raOMc' = 3 "t" T
A. ENGME MOUNTE.D COMPONENTS ARE
- (1rOUm u,.isysTEM
.g4 h
--E Ot m NsrE rEEEsTANonnoser -
FIGURE 1.9.22.4 -
DIESEL GENERATOR ENGINE LUBE OIL SYSTEM l
3
+
FILTER I
M
?
O AIR COOLER
( MA R
TURBOCHARGER FROM EXHAUST.
i T
C
[MAMFOLD TO J
EXHAUST
\\
ATMOSPHERE {L SILENCER 3
.l
{l FROM EXHAUST
' JAMFOLD i
j l TO INTAKE -
--t>- FILTER INTAKE AIR O
AIR COOLER MANIFOLD SILENCER i
DIVISION 1 9
+
FILTER INTAKE AIR
/
TO INTAKE C
MR NW S!LENCER f MAMFOLD TURBOCHARGER lFROM EXHAUST f MAMFOLD TO EXHAUST
\\
ATMOSPHERE f
SIENCER I FROM EXHAUST MAMFOLD TO INTAKE
+
FILTER NC R C
TURBOCHARGER AIR COOLER MAMFOLD NOTES:
A. ENOMEMOUNTEDCOMPONENTS ARECONSTRUCTEDM DIVISION 2 ACCOPDANCE WITH IEEE STANDARD 387 a COMPONENTS AND PIPMG SHOWN ARE ASME SECDON N
. Ct. ASS 3 CODE APPROVED WITH THE EXCEPDON OF THE MTAKE PILTER, INTAKE SE.ENCER, AND F.XHAUST SILENCER. THESE COMPONENTS ARE SEisenCALLY OUAUFIED BY SHAKER TABLE TESTS OR ANALYSIS PERPORMED BY MANUFACTURER.
C. THE PMAL SYSTEM CONFIOURATION IS DEPENDENT UPON TwE A-OCURED D=SEtOENERATOa ENOME FIGURE 1.9.22.5 DIESEL GENERATOR ENGINE AIR INTAKE AND EXHAUST SYSTEM'
I ASME CODE CLASS l l ASME CODE CLASS l IN 31 IN 31 3
3 TO HONRADtOACTIVE I TO NONRADIOACTIVE g
-g h
nooR DRmN system n -
ROOR DRAIN SYSTEM a
n y
n N
W X
M Zn
.Zn Zn Za AREA FLOOR DRAINS AREA FLOOR DRAINS vvvvv vv:vvv M
M M
M i
P P
P P
P P
P P
U U
U U
n a
o n
1r 1r1rU 1r 1r 1r1r1r 1r FLOOR DRAIN SUMP
- FLOOR DRAIN SUMP DMSION 1 DMSION 2 FIGURE 1.9.22.6 DIESEL GENERATOR BUILDING SUMP PUMP SYSTEM
1 SYS"IT,M 80+"
1.9.22 DIESEL GENERATOR SUPPORT SYSTEMS SUPPORTIVE INFORMATION 1.
Amplifying information ITAAC 3 Confirmation that each pair of storage tanks are sized to provide a combined storage of no less than a seven day supply of fuel oil plus a ten percent margin will be performed. The analysis will be based on the following:
Fuel consumption rate at rated load or actual load as provided by _the diesel generator engine manufacturer.
Time intervals at different loads, if applicable.
2.
Relationshin of DGSS ITAAC to the Safety Anahsis N/A 3.
Relationshin of DOSS ITAAC to PRA The PRA for the diesel generator support systems assume the following:
A.
Each emergency diesel generator has two independent starting air systems.
B.
The starting air storage capacity for each emergency diesel generator is sufficient for starting the diesel generator for a minimum of five times.
C.
Each emergency diesel generator has an independent fuel oil storage system.
The storage system has sufficient fuel that allows the emergency diesel generator to operate at full power for a time period of no less than seven days.
D.
Fuel oil is transferred from the storage system to the day-tank of each emergency diesel generator. The day tank has sufficient capacity to allow the emergency diesel generator to operate at full _ load for approximately 60 minutes without being replenished. ~
E.
Transfer from the storage system to the day tanks is performed automatically.
4.
CESSAR-DC Chapter 14 Tests Annlicable to DGSS ITAAC See CESSAR-DC Section 14.2.12.1.95 i
1.9.22- 01 28-93 L
SYSTEM 80+"
1.9.24 ULTIMATE IIFAT SINK Design Description De Ultimate Heat Sink (UHS) provides the source of cooling water that transfers heat from the Station Service Water System (SSWS) to the environment.
The Ultimate Heat Sink is not within the scope of the certified design. The site specific UHS will meet the interface requirements defined below.
Interface Requirements The UIIS meets Seismic Category I requirements and its function is not lost during or after any of the following events:
1.
Natural phenomena including safe shutdown earthquake, tornado, flood, and drought.
2.
Site related events including transportation accidents, oil spills, and fires.
3.
Credible single failures of man.made structures.
4.
Sabotage.
nc UHS is capable of providing cooling to support operation, shutdown, refueling, and design basis accident conditions. The Ultimate Heat Sink is capable of providing an SSWS inlet temperature that does not exceed the maximum allowable temperature required for removing heat from the component cooling water heat exchanger during a design basis accident concurrent with a loss of offsite power.
For sites with severe winters, where ice formation in the Ultimate Heat Sink could occur, the function of the Ultimate Heat Sink is not impaired during winter months.
Where required, the intake structures will be provided with a means of deicing to prevent flow blockage of the station service water pump inlets.
A site water chemistry analysis for the Ultimate Heat Sink will be performed to determine if a water treatment system is required to minimize corrosion and fouling of the Station Senice Water System.
Inspections, Tests, Analyses, and Acceptance Criteria Table 1.9.24 specifies the inspections, tests, analyses and associated acceptance criteria for the Ultimate Heat Sink.
1.9.24 1-28-93
SYSTEM 80+
TABLE 1.9.24-1 ULTIMATE HEAT SINK Inspections. Tests. Analyses. and Acceptance Criteria Certified Design Commitment Inspections. Tests. Analyses Acceptance Criteria 1.
The UHS is capable of providing 1.
Analyses v.iIl be performed to 1.
The fecility specific SAR commits cooling to support operation, demonstrate UHS heat remov::1 that the Ultimate Heat Sink is shutdnwn, refueling, and design capacity based upon site capaNe of evidig cooling to basis accident condhions.
The meteorological conditions.
The support operation, shutdown, Ultimate Heat Sink is capable of accident analysis shall extend from icimimg. and design basis accident providing an SSWS inIet the start of the accident through a conditions and the accident analysis temperature that does not exceed 30 day time penod, shall be based demonstrates that the UHS the maximum a1Iow abi e on the worst case meteorokpcal prmides an SS%3 inlet temp.4ure temperature for removing heat conditions for the site based on the that does not mm-d the manmum from the componem coolmg water guidelines given in Regulatory a1IowabIe temperatare foc heat exrbnger during a design Guide 127, and shall consider no remming beat from the component basis. accident coincident with a water makeup to the Ubimate Heat cooling water heat Lp e
Sink for 30 days.
during a design basis accident coincident with a loss of offsite power.
2.
For sites with severe winters, 2.
An analysis mill be performed 2.
The facihty specific SAR shows where ice formation in the UHS based upon site metec ological that the function of the UHS is not could occur, the function of the conditions.
impaired due to subf. m Q UHS is not impaired during winter conditions or a method is prmided months.
to prevent flow blockage of the station service water pump inlet due to sutier.Q conditions.
3.
A site water chemistry analysis for 3.
An analpis of the site water 3.
The facility specific SAR commits the UHS will be performed to chcmhy mill be performed.
to a water treatment system if determine if a water treatment.
water Ry icements are system is ruluired to minimize at met.
corrosion and fouling of the S5%3.
1.9.24 1-23-93
i i
. NYSTEM 80+"'
l.9.24 ULTDIATE IIEAT SINK ITAAC SUPPORTIVE INFORMATION 1.
Amplifyine Information l
N/A 2.
Relationship of UliS ITAAC to the Safety Analysis The UllS ITAAC does not include any specifle inspections, tests, and analyses which l
confirm that the as. built system configuration and performance match the bases used In the evaluation models for licensing analysts. Ilowever,it is assumed that the UllS is available to support the CCWS and its assumed cheracteristics as detailed in the evaluation models.
3.
Relationshin of UllS ITAAC to PRA
'Ihe PRA assumes that the Ultimate !! cat Sink exists.
4.
CESSAR DC Chapter 14 Tests Applicable to UllS ITAAC See CESSAR DC Section 14.2.12.1.76 5.
Status of DSER Items Related to the UIIS ITAAC N/A i
i t
f!
1.9.24 1-28 3;
.}
- ~
+
u.,
SYS1EM 30+"
1.10.4 EMERGENCY FEEDWATER SYSTEM Design Description The emergency feedwater system (ERYS) is a safety related system which supplies feedwater to the steam generators for events resulting in loss of normal feedwater and requiring heat removal through the steam generators.
The ERYS consists of two mechanical divisions, cach with an ernergency feedwater storage tank (ERYST), two ERV pumps, a cavitating flow.llmiting venturi, valves, piplog, iamnentation and controls.1he pumps in cach division are powered by diverse drivers. Each rump has a separate dedicated suction line from the ERYST in its division. 'the discharge lines from the two pumps in each dhision are joined together in a common header upstream of the cavitating venturi. Flow is delivered to the downcomer feedwater line to each steam generator (50). Steam for the pump turbine driver in each division is taken from a main stmen line between the steam generator associated with that division and the main steam isolation valves for that SG. A cross-connect line with isolation valves is provided between the two EFWSTs.
Another cross-connect line with isolation valves is provided between EFWS pump discharges for the two divisions. A non safety grade source of condensate makeup is arranged for gravity feed to either ERYST. Figure 1.10.41 shows a simplified system conuguration.
liach EFWS pump deliven at least the minimum Dow required for core decay heat removal using the steam generators, against steam generator feedwater nonic pressures up to main steam strety valve lift pressure. Water is supplied to each EFW pump a a pressure greater than the net positive suction head (NPSII) required.
Each ERYST has a volume above the ERV pump suction line penetrations to permit plant cooldown to shutdown cooling entry conditions following any design basis event.
The cavitating flow limiting venturis limit emergency fecdwater flow to each SO with both ERVS pumps running in the dhision, against steam generator pressures down I
to 0 psig.
- lhe ERYS is built to the ASME Code Section III class requirements shown on Figure 1.10.41. Components, piping, and supports classified as ASME Code Class i'
I 2 or 3 are beismic Category 1. Equipment that b designated as safety related is qualified for the environments where k>cated, j
ERVS instrumentation indications and alarms shown on Figure 1.10.41 are available in the control room. Controls are available in the control room to start and stop the ERV pumps, and open end close the steam turbine supply valves, steam generator isolation valves, and now control vakes, i
ERYS safety related components and power-operated valves are supplied from the Class 1E electrical distribution busses. The ERVS is actuated by an emergency 1,10.4
-1~
12893
jiYh"IV.M 80+"
feedwater actuation signal (EFAS) from the enginecred safety features actuation system (ESFAS) or by an alternate feedwater actuation signal (AFAS) from the alternate protection system (APS). *lhe EFAS or AFAS for each steam generator starts the motor driven pump, opens the steam supply valve to the turbine driver, which starts the tmbine and pump, and opens the two steam generator isolation valves and the two E!W Dow control valves in tbc actuated EFWS division (see Figure 1.10.41). *lhe Engineered Safety Features Component C(mtrol System includes logic to close the isolation valves and flow control valves when SO water level has risen above a high level setpoint, and to re-open those valves when SG water level drops below a low level setpoint.
Outside the containment, the two rnechanical divisions of the EfMS are separated by the divisional barrier wall, except for the cross-connect lines between EIMSTs and between divisional HIM pump discharge lines.
A flow recirculation line_ from each EIMS pump provides discharge back to the EFWSTs for minimum flow protection and flow testing of the putsps.
Inspections, Tests, Annlyses, and Acceptance Criteria Table 1.10.41 specifies the insix.ctions, tests, analyses and associated acceptance criteria for the EIMS.
h 1.10.4 12893
SYSTEM 80+"
TABLE L10A-1 F.MERGENCY FEEDWATER SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Certified Desien Commitrnent Inspections. Tests. Analyses Acceptance Criteria 1.
A basic configuration for the 1.
Visual inspectens of the as-buih L
The as-built configuration of the EFWS is shown in Figure 1.10.4-L system configuration will be con-EFWS is in accordanc with Fis;ure ducted.
L10.41 for the components and i
equipment sho w.
2.
ASME Code portions of the EFWS 2.
A pressure test will be conducted 2.
The results of the pressure test of retain their integrity under internal on those portions of the EFWS re-ASME Code portions of the EFWS pressures that will be wwkad quired to be pressure tested by the conform uith the ic@ments in during service.
ASME Code.
the ASME Code Section IIL 3.
Water is supplied to each EFW 3.
Tests to measure EFW pump NPSH 3.
The calculated anilable NPSH pump at apressure greater than the will be performed. An analysis to exceed. pump NTSH required by i
net posithm suction head (NPSH) deterndne NPSH availabic to each the vendor for the pump.
required.
EFW pump will be prepared based on as-built data.
4.a) An emergency feedwater actuation 4.a) Testing will be performed by gen-4.a) The motor-driv en and tur-signal (EFAS) actuates the EFWS erating a simulated EFAS for its bine-drhrn pumps start, and the components.
An alternate corresponding steam generator.
steam generator ise.lation and flow feedwater actuation signal (AFAS)
The test will 17, repeated using a control vahrs open, in the dhision actuates the EFWS components.
simulated AFAS.
recching the simulated EFAS. The same components actuate in res-pense to a simulated AFAS.
1.10.4 1-28-93
Sys'mM 80+=
TABLE 1.10A-1 (Continued)
EMERGENCY FEEDWATER SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Certified Desies Cosunitment Inspections. Tests. Analyses Acceptance Crkeda 4.b) SG water level. ~gnah open and 4.b) Fundional tests of each division 4.b) A simulated high SG water level close the SG ' isolation and flow will be performed by simniaring signal closes the SG isolation vahrs control valves.
high and low SG water k: vel and flow control valves in its signals.
associated dhision.
A simulated low SG water level tig,al opens the SG isolatH vahrs and flow control vahrs in its associated dhision.
5.a) Each EFWS pump delivers at Icast 5.a) EFWS fw Wal tests of each 5.a) Each EFWS pump delivers at least the minimum Dow. required for EFWS pump will be performed to 500 Epm to the steam generator (s) core decay heat. removal to the determine as-built system Dow n against asteam generator feedwater steam generator (s) against a steam steam generator pressure. Analyses no=de p - of 1217 psia.
l generator feedwater nozzle pressure wiD be performed to consert the up to main steam safety vahr lift test resuhs to the conditions of the Certified Design Commitment.
pressure.
L b). Cavitating flow-limiting venturis b) EFWS fadw==1 tests will be per-b) ' Maximmn flow to each SG is 800 limit maximum flow to each SG formed with both pumps in a divi-gpm with both pumps running with both pumps in.the dhision sion running.
Analyses will be again4 a steam generator pressure running, against a steam generator used to conert - the test resuks to of 0 psig.
the conditions of ' the Certdied pressure of 0 psig.'
Design Commitment.
6.
Each emergency. feedwater storage.
6.
Inspection of construaion records 6.
Each EFW5T internal volume is at tank has an internal ' volume rbove for the EFWSTs will be performed Icast 350,000 gaHons above the the EFW. pump sudion line and the internal - volume of each EFW pump suction line pene-penetrations :
to 1 permit plant tank available for emergency feed-trations. '
cooldown to shutdown ' cooling water will be calculated.
entry conditions ?. following 'any
. design basisLcvent.
1.10.4 1-28-93
SYSTEM 80+=
TABLE 1.10.4-1 (Continued)
EMERGENCY FEEDWATFR SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Certified Desis Commitment Jaspections. Tests. Analyses Acceptance Criteria 7.
EFWS instrumentation indications 7.
Inspection of the control room for 7.
The instrumentation indications Fig;ure the availability of instrumentanon and alarms shomu on Fq;ure and alarris shown-on L10.4-1 are available in the indications and alar.ns identified in L10.41 exist or can be retrbed Control Room.
Controls are the Certified Design Commitment in the Control Room.
EFW room to will be performed.
Tests will be controis operate as speofied in the available in the_ control start and stop the EFW pumps, and performed ndng the EFW controls Certified Design Commitment.
open and dose the EFW pump in the Control Room.
steam turbine supply vahts, steam generator isolation vahrs, and flow control vahrs.
8.
Safety-related EFWS components 8.
A test of the power availabtTity to 8.
The Certified Design C-mitment described in the Dedgn Description the safety related components for is met.
for each division of the EFWS are the EF%3 will be conducted with
- powered from their. respective power supplied from the perman-I
' oaIIed electrical power Class IE busses with the %duu ently m
of containment isolation vahts and buses.
assooated containment isolation valve instrumentation and controls.
9.
Outside om'=tament; the two 9.
Visual Inspections of EF%3 dhi-9.
Outside of containment, a dhi-mechanical divisions of the EFWS sional mechanical separations. will simal wall separces the two EFWS are physica!!y separated except for be performed.
mechanical dhisions.
the cross-connect lines between EFWSTs and betsten dhisional EFW pump discharge lines.
1.10.4-1-28-93
TABLE 1.10.4-1 (Continued) sys1EM 80+=
EMERGENCY FEEDWATER SYSTEM Insocctions. Tests. Analyses and Acceptance Criteria Acceptance Criteria Insocctions. Tests. Analyses Certified Desien Commitment 10.
The flow recirculation line from 10.
Tests of each EFW pump in the 10.
Minimum recarculation flow meets crinimum flow and full flow test or exceeds the pump vendor's each EFW pump discharge back to modes will be conducted with flow required flow. Full Dow from each directed to the EFWST through tbc pump (at least 510 Fpm) is returned its associated EFWST provides required EFW pwnp minimum to the EFWSTs.
flow '_ad permits testing each EFW pump's recirculation Iines.
pump at full flow.
1 28-93 1.10.4
r
--- Eras oR Aras o*Ess
.outsoE cowTAe.neENT wooE ccNTAmMEur g
u pas
. on..a.s.e.v.e.u.rs, w
MAm, STEAM
{
Iasus ccoe etAss I g
g M
l l
N r]
N '"'I Q 'AI Eu, FROM DEwW
ums on AFAsTAmts f
EFAS WATER,
C
.a"e1Cn-%
c -
- STEAM CENERATOR 1 H
EMET.CENCY 4O FEEDWATER DOW:OMER SD cAvrTATma tmE 4--@
6 vEuTuai sTORAct L
Tux i i
g
~
g v.
9 gn3 O
O*
1 I
c>c
- =~w.N r
-,tc.
-v=ve v=ve h
f EFASCR AFASSTARTS
.E
~
eras Ca AFAS CvCLES MAN STEAM l"*******'
SQ t gyyy M
souRC
\\
C pc N
- MaEu,
[
Erw pupp 54ISOUCTICM %
FROM DEMIN cowrnoLvALys vmLvt b
E*As OR AFAA STm WATER
- i. '
Z+r
-as
=
ca uAs evetes ETEAM CEft.AATOR 2 nsa
- "~~~*7 EMERGENCY FEEDWATER DOW4CoteER i
l CAVITATpf8 LNGE FEEDWATER' veMrwM STORAGE g ;I f
f Se;i r==,
^
@W w
.. ano.
r.
CoNTROLVAL*PE
. TALVE Mil i.
eFAscR AFAssTARTs SYSTEM +TM i
FIGURE 1.10.4-1 EMERGENCY FEEDWATER SYSTEM
=-
. - =. -
SYS'IEM 80+"
1.10A EMERGENCY FEEDWATER SYSTEM ITAAC SUPPORTIVE INFORMATION 1.
Amnlifyine Informatinq ERYS
Description:
EcIntionshin of EFWS ITAAC to the Safety Analv_sh, 1)
IIASIS: Minimum flow rate to a steam generator requiring emergency feedwater = 500 gallons per minute with steam generator pressure at 1200 psia.
ITAAC: ITAAC 5 acceptanco criterion requires the minimum flow of 500 gpm with steam generator feedwater nozzle pressure at 1217 psia which corresponds to design preaure of 1200 psia,12 psi safety valve tolerance and 5 psi pressure difference between nozzle and steam pressure. At 1200 psia steam pressure, the flow would therefore exceed the analysis value of 500 gpm.
2)
HASIS: Maximum flow rate to a steam generator requirisig emergency feedwater = 800 gpm at runout conditions.
ITAAC: ITAAC 5 acceptance criterion requires maximum ERV flow to a SG of 800 gpm with steam generator pressure at 0 psig. During a steam line break inside containment which depressurizes the steam generator (s), there would likely be some pressnre in the SG due to containment pressurization.
Some pressure in the SG would also be likely in the case of a steam line break outside containment due to line losses between the SG and the break lo:ation. 'Ite ITAAC conditions bound the expected conditions during analyzed events.
3)
IIASIS: Emergency feedwater storage tank capacity is 350,000 gallons each.
ITAAC: ITAAC 6 acceptance criterion requires a minimum volume of -
350,000 gallons per tank.
4)
IIASIS: No single failure in the ERVS will prevent the system from performing as stated in 1) through 3) above.
ITAAC: ITAAC 8 and 9, respectively, confirm the mechanical separation and electricalindependence aspects of the ERVS. ITAAC 5 confirms that each EFWS pump is capable of meeting the system flow delivery requirements.
1,10.4 12893
l l
EYSTEM 80+"
3.
Relationship of EFWS ITAAC to PRA 1)
He ERVS has two redundant divisions for supplying feedwater to the steam generaton to achieve heat removal from the reactor to the entry conditions for using the SCS.
2)
Each ERVS dkision has two EFW pumps, cach with a pump driver dhcrse from the other.
3)
In each EBVS dMslon, the two ERV pump discharge pipes are joined together inside containn.cnt to a single pipe that connects to the SO downcoinct feedwater line.
4)
The ERV pumps in one division can supply feedwater to the SG in the other division through a pipe having at least two normally closed isolation valves 4
installed.
5)
Each EFW Storage Tank (EFWST) can be supplied by gravity flow from the Condensate Water Storage Tank (CST). His source is isolated by at least two normally closed isolation valves.
6)
The EFW turbincariven pump in each division is supplied steam from the SG in its division via a pipe connection located upstream of the MSIV.
7)
The EFWS is actuated by a EFAS and a SPS actuation signal (Low SG Water Ixvel).
8)
Upon receipt of an actuation signal, the EFWS:
Start the associated motor-driven pump, a.
b.
De-energizes the solenoid to open the associated turbine steam supply valve.
c.
Opens the associated EFW isolation valves to the appropriate SO.
9)
Each EFW division provides at least 500 gpm to tbc downcomer line of either SG.
10)
Installed instrumentation provides the capability to monitor the performance of the system and the major components from the control room.
11)
Each ERV pump can deliver EFW flow to the SGs when the SG pressure is at the Main Steam Safety Valve (MSSV) setpoint.
12)
Each EFWST has a useable volume of at least 350,000 gallons.
1.10.4 1 28-93
I SYS'IEM 80+"
13)
Each EFW division receives power from its associated Class 1.E buses.
14)
Each EFW line has a cavitating venturi to limit FFW pump runout when feeding a %ad" steam generator.
4.
CESSAR DC Chapter 14 Tests Applicable to EFWS ITAAC Pre-operational Tests: CESSAR DC Section 14.2.12.136 i
e i
P 1
1.10A 3
12893
.,,.,-~.
I finiv.M 80+"
1.11.2 UASEOUS WASTE MANAGEMENT SYSTEM Design Description The Gaseous Waste Management System (GWMS) is a non safety system which collects, stores, processes, samples, and monitors radioactive gaseous waste. Piping from the Reactor Drain Tank to the GWMS penetrates containment and is provided wi'h containment isolation valves.
'Ile GWMS is a charcoal delay system. 'Ihe GWMS ptccesses radioactive gases generated by the plant sptems connected to it during plant operations. This system is not intended to process post. accident sources; therefore the inlet radioactive waste streams to the GWMS are isolated in post-accident conditions. A general conceptual illustration of the GWMS is shown in Figure 1.11.21.
The GWMS system contains conditioning equipment (including a coolcr-condenser for humidity control and a charcoal guard bed) to minimize moisture and contamination in the charcoal adsorbers and charcoal adsorbers to delay passage of noble gases through the equipment.
'The GWMS precludes the buildup of an explosive mixture of hydrogen and oxygen in the GWMS by the provision of parallel gas analyzers and nitrogen purge capability.
parallel gas analyzers monitor the concentration of hydrogen and/or oxygen in the GWMS. Nitrogen purge capability maintains the concentration of hydrogen and/or oxygen at icss than 4% of the gases in the GWMS. The parallel gas analyzers provide signals to alarm both k>cally and in the control room exceeding a limit on the concentrations of hydrogen and oxygen.
The GWMS processes radioactive gaseous waste so that the concentration of the gaseous radioactive effluents discharged to unrestricted areas are within limits specified by ICCFR20, Appendix 11, Table 11 during normal operating conditions and in the event of a GWMS leak or failure. Effluents from the GWMS are filtered through particulate and activated charcoal filters prior to release at the unit vent to the environment.
The GWMS can continuously monitor concentrations of radioactivity in piocessed 3
gaseous waste prior to release to the environment. The radiation monitor activates y
controls to automatically isolate the GWMS discharge if the limits of 10CFR20, Appendix I3, Table II will be exceeded, inspections, Tests, Analyses, and Acceptance Criteria Table 1.11.2-1 specifies the insputions, tests and/or analyses and their associated acceptance criteria for the GWMS, 1.11.2 1 23-93
SYsnEM 80+
_ TABLE 1.11.2-1 GASEOUS WASTE MANAGEMENT SYSTEM Insocction;. Tests. Analyses. and Acceptance Criteria Certified Desima C-=k-M I= E-: d*-
=_ Tests. Analyses Accestance Critaia 1-A basic r9afigcration of the L
An inspection of the as-buik L
De as-buik G%NS configuration GWMS is shown in Figure L1L2-L GWMS configuration will be is in accordance with Figure conducted.
LIL2-1 for the components and equipment shown.
2.
The G%MS procest s radioactive 2.
Analysis of the as-built GwMS 2.
The concentration of radioactive gaseous waste : so ' that the performance data will be isotopes in the gaseous efDuents do concentration of the radioactive performed. Testing of the G%NS not exceed 10CFR20, Appendix B, gaseous efDuents discharged to components will be performed.
Table II Emits.
unrestricted ~ areas ' within limits specified in 10CFR20, Apgendix B, Table IL 3.
The GWMS ged.dcs a buildup of 3.
The foBowing inspections and tests 3.
The G%MS gcandes the buildup an explosive mixture of hydrogen are performed to verify that the of an explosive mixture of and oxygen. The G%NS has:
G%NS can picd de a buildup of hysgcs and oxygere an capkee mixture.of hyaw and oxygen:
a) ParaIIel gas analya:rs for hydrogen a) Inspection of as-built GWMS con-a) Parai!el gas analyzers stich alarm
- and/or oxygen. which alarm ' both fquration will be performed. Tests both locally and in the control locally 'and in the control room.
. of wiE be performed usig; signk room are provided.
that simulate ' wdc. ice of limits
.to venfy the paraBel gas analyn alarm upon detection of high concentrations
'of Lyiqcs or cxygen in the GWIIS.
L11.2 1-28-93
SYSTEM 80+
'EABLE I.11.2-1 (Continued]
GASEOUS WASTE MANAGEMENT SYSTEM Inspections. Tests. Analyses. and Acceptance Criteria Certified Desicn Commitment Inspections. Tens. Analyses Acceptance Criteria 3.b) Nitrogen purge capability to 3.b) Tesis to measure nitrogen purge b) Nitrogen purge ~6ta'= bydrogen maintain the hydrogen and/or flow wiH be performed.
and/or oxygen concentratens at less than 4% of the gases in the ongen concentrations at less than G % 315.
4% of the gases in the GWMS.
e) The parallel gas analyzers proside c)
Tests if alarms will be performed c) local and control room alarms signals to alarm both locally and in using simalated analyzer signah to function.
l the control room upon exceeding a the alarms.
limit on the concentratiocs of hydrogen and oxygen.
l 4.
The GWMS can monitor 4.
Inspections of the as-built system 4.
Radiation monitoring capability is located duaw of the charcoal l
concentration of radioactivity in wiD be performed.
adsorbers.
processed gaseous waste prior to release to the emironment.
5.
The mstrementation indications i
Inspections of the as-built system i
Instrumentation indications and and controls shown in Figure will be performed.
Tests of the controls shown in Fy;ure LIL2-1 1.1L2-1 are available in the main inlet waste stream isolation will be exist or can be retrieved control room.
The GWMS inlet waste control room. The inlet radioactive performed.
streams can be isolated by manual gaseous waste streams to the acte n.
G%%iS can be isolated manually I
from the control room.
)
6.
Discharges of radioactive gaseous 6.
Tests of - the isolation capability 6.
Discharge is terminated upon l
signal that simulates receipt of simulated srnal.
effluents to the emironmem are using a terminated automatically if the exceedence of limits will be limits of 10CFR20 Appendix B,
performed.
Table 11 will be exceeded.
1-28-93 L11.2 PROCESS GAS HEADER Il STRIPPER DRAIN TANK (EDT) h 1r 1r VOLUME CONTROL F
1 I7 7
TANK (VCT)
?
Y O*""
CNCOAL CNEGAL
(
COOLER CHILLED N
WAM REACTOR DRAIN (NOTE A)
TANK (RDT)
L _.s n
a es u.>
n h
R Q
Ug WI (j
DRAW II u
TANK BYPASS m h 37 FROM LWMS AERATED VENTS LJ NOTES:
A.
CONTAINMENTISOLATION VALVES AND ASSOCIATED PIPING ARESAFETY CLASS 2 EL ALL COMPONENTS AND PIPING ARE SAFETY CLASS NNS UNLESS OTHERWISE NOTED.
FIGURE 1.11.2-1 GASEOUS WASTE MANAGEMENT SYSTEM FLOW DIAGRAM
liLSInt 80+"
1.11.2 GASEOUS WASTE MANAGEMENT SYSTEM SUPPORTIVE INFORMATION l
1.
Amplifyine Information ITAAC 1 CESSAR DC, Section 11.3 revised per DSER open item 11.3-5, provides a description of the methodology to verify compliance with to CFR 20, Appendix B limits. Included in the analysis are the following assumptions:
a)
Minimum carrier gas flow rate of at least I scfm.
l l
b)
Minimum mass of charcoal in adsorber of at least 18,000 lbm.
c)
Minimum charcoal adsorbtivity for Kgpton and Xenon of at least:
18.5 cc/gm for Krypton 330 cc/gm for Xenon d)
Average atmospheric dispersion factor of 7.2x10 sec/m' during 4
normal operation.
e)
Two hour accident atmospheric dispersion factor of 1.0x10*sec!m'for accident conditions.
IIAACs.3.
l The parallel gas analyzers provide a high alarm at 3% hydrogen and 1% oxygen and a high-high alarm at 4% hydrogen concentration in the GWMS. 'lhe high alarm l
provides ample time for the operator to take remedial action to initiate the nitrogen l
purge systems to reduce the concentration of hydrogen or oxygen in the GWMS. The high-high alarm automatically initiates the nitrogen purge system to preclude a j
buildup of an explosive mixture of hydrogen and oxygen in the GWMS in accordance l
with 10CFR50, Appendix A, (General Design Criteria 3).
L ITAAC 4. 5 and 6 L
- Ihis criteria ensures compliance with 10CFR50, Apoend!x A (General Design Criteria 60 and 64) which specify the requirement for a cona olled monitored release pathway.
In addition, leakage rates of processing equipment of the GWMS shoukt be maintained within the limits specified in ANSI /ANS 55.4, Table 9 to ensure releases i
from the GWMS to the emironment are controlled.
1.11.2 -
1 28-93 l
L l
'J
l SYhTEM 80+"
In adaition 10CFR50, Appendix I, which specides maintalning general put lic exposurc AIARA due to radioactive gascous ef0uents, is an important design objective which must be met to verify compliance with 40CFR190 (an acceptance criteria in the Radiation protection ITAAC). 40CFR190 specides a limit for exposure to the general public (l.c.,25 mrem who:e bcx!y,75 thyroid,25 any other organ) due to direct and scattered radiation, as well as radioactive effluents from a uranium fuel I
cycle. Although a failure to comply with 10CFR50, Appendh I wuuld not result in automatic shutdown of a facility, a detailed report describing why the limits were exceeded and action to be taken would be required. Ilowever, a pattern of noncompliance could result in a civil penahy based on failure to control radioactive releases per 10CFR50, Appendix A (GDC 60) and would rc0cet unfavorably on the design of a radioactive waste management system.
To verify compliance with 10CFR50, Appendix I, an analysis using Regulatory Guide 1.109 methcxiology would be performed. The following site specific information would be required to perform the analysis:
a) land use survey, such as location of nearest food pathways (e.g., potable water source, garden, cow, goat, etc.)
b)
Meteorological data (i.e., average annual atmospheric dispersion factor (X/0) and deposition factor (D/0)).
2.
Relationship of the Gaseous Waste Manacement System ITAAC to the Safety Anahsis 4
Section 15.7.1, 'Gascous Waste Management System Ixak or Failure", of the CESSAR DC is addressed in Section 11.3.7.2 of the CESSAR-DC.
e 3.
Eglationship of Gaseous Waste Manacement System ITAAC to PRA N/A l
4.
CESSAR DC Chanter 14 Tests Annlicable to Gaseous Waste Manacement System ITAAC l
l 14.2.12.1.116 I
+
l l
1.11.2 1 28-93
, - -. - -.. -.