ML11264A028
| ML11264A028 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 09/13/2011 |
| From: | Morris J Duke Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 11-CN-002 | |
| Download: ML11264A028 (24) | |
Text
JAMES R. MORRIS DEnergy Vice President Duke Energy Catawba Nuclear Station 4800 Concord Road York, SC 29745 803-701-4251 803-701-3221 fax September 13, 2011 10 CFR 50.55a U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001
Subject:
Duke Energy Carolinas, LLC (Duke Energy)
Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Proposed Alternative Request Number 11-CN-002 for the Third Ten-Year Inservice Inspection Interval Pursuant to 10 CFR 50.55a(a)(3)(ii), Duke Energy hereby requests NRC approval of proposed alternative testing for the remainder of the third ten-year inservice inspection interval at the Catawba Nuclear Station. The details of the request are included in the enclosure and its attachment. Duke Energy requests NRC approval of this request within one calendar year of the submittal date.
This submittal document contains no regulatory commitments.
If there are any questions or if additional information is needed, please contact L.J. Rudy at (803) 701-3084.
Very truly yours, James R. Morris Enclosure/Attachment www. duke-energy.corn
U.S. Nuclear Regulatory Commission Page 2 September 13, 2011 xc (with enclosure/attachment):
V.M. McCree Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 G.A. Hutto, III NRC Senior Resident Inspector U.S. Nuclear Regulatory Commission Catawba Nuclear Station J.H. Thompson (addressee only)
NRC Project Manager (Catawba Nuclear Station)
U.S. Nuclear Regulatory Commission Mail Stop 0-8 G9A Washington, DC 20555-0001
Enclosure Duke Energy Corporation Catawba Nuclear Station, Units I and 2 Relief Request Serial #11-CN-002 Relief Requested in Accordance with 10 CFR 50.55a(a)(3)(ii) for Pressure Testing the Extended Boundary of the Reactor Coolant System Page 1 of 9
Duke Energy Corporation Relief Request Serial #11-CN-002 Page 2 of 9
- 1. ASME Code Component(s) Affected Systems Designation Legend:
NC - Reactor Coolant System (or RCS)
NV - Chemical and Volume and Control System ND - Residual Heat Removal System (or RHR)
NI - Safety Injection System WL - Liquid Radwaste System Duke Energy is requesting relief for the following ASME Class 1 piping and component segments connected to (or part of) the RCS that are isolated from direct RCS pressure during normal operation. These segments are isolated from the RCS by their configuration because they are upstream of a check valve, between two or more check valves, or between two normally closed valves that remain closed when the unit is in normal operation.
Segment 1 consists of 2 inch (NV) Class 1 piping and components upstream of Auxiliary Spray inboard check valve NV-38 up to and including outboard RCS isolation valves NV-37A (globe valve) and NV-861 (check valve).
Pipe Size / Schedule 2" NPS / Sch 160 Material Type / Grade Stainless Steel SA-376 / Grade 304 Design Pressure 2500 psia Design Temperature 650 degrees F Approximate Length 70 Feet (total for both units)
Piping and components for Segment 1 are shown on Unit 1 drawing CN-1 554-01.00 and Unit 2 drawing CN-2554-01.00 enclosed in Attachment "A".
Segment 2 consists of the following:
- 12 inch, 1 inch, and % inch Class 1 piping and components on the (ND)
Suction line between (and including) the RCS double isolation gate valves ND-1B and ND-2A ("A" Train ND Suction) and flow restrictors upstream of valves ND-4, ND-110, and ND-116.
- 12 inch, 1 inch, and % inch Class 1 piping and components on the (ND)
Suction line between (and including) the RCS double isolation gate valves ND-36B and ND-37A ("B" Train ND Suction) and flow restrictors upstream of valves ND-39, ND-111, and ND-117.
Pipe Size / Schedule 12" NPS / Sch 140 Material Type / Grade Stainless Steel SA-376 / Grade 316 Design Pressure 2500 psia Design Temperature 650 degrees F Approximate Length 305 Feet (total for both units)
Duke Energy Corporation Relief Request Serial #1 1-CN-002 Page 3 of 9 Pipe Size / Schedule 1" NPS / Sch 160 Pipe and 6000# Fittings Material Type / Grade Stainless Steel SA-376 / Grade 304 (Pipe)
Stainless Steel SA-182 / Grade F304 (Fittings)
Design Pressure 2500 psia Design Temperature 650 degrees F Approximate Length 6 inches (total for both units)
Pipe Size / Schedule W"NPS / Sch 160 Pipe and 6000# Fittings Material Type / Grade Stainless Steel SA-376 / Grade 304 (Pipe)
Stainless Steel SA-182 / Grade F304 (Fittings)
Design Pressure 2500 psia Design Temperature 650 degrees F Approximate Length 14 inches (total for both units)
Piping and components for Segment 2 are shown on Unit 1 drawings CN-1561-01.00 and CN-1561-1.1. Unit 2 piping and components are shown on Unit 2 drawings CN-2561-01.00 and CN-2561-1.1. These drawings are enclosed in Attachment "A".
Segment 3 consists of 1% inch (NI) Class 1 piping and components within the 4 RCS Loops, between (and including) double isolation check valve pairs listed below:
NI-15 and NI-351 for Loop A NI-17 and NI-352 for Loop B NI-19 and NI-353 for Loop C N1-21 and N1-354 for Loop D Pipe Size / Schedule 1 /" NPS / Sch 160 Material Type / Grade Stainless Steel SA-376 / Grade 304 Design Pressure Valves NI-15, NI-17, NI-19, & NI-21 = 2500 psia Remaining components in Segment 3 = 2750 psia Design Temperature 650 degrees F Approximate Length 22 Feet (total for both units)
Piping and components for Segment #3 are shown on Unit 1 drawing CN-1 562-1.0 and Unit 2 drawing CN-2562-1.0 enclosed in Attachment "A".
Segment 4 consists of 2 inch and 1/2 inch (NC) Class 1 piping and components between (and including) double isolation valves and flow restrictors, as listed below:
- NC-4, NC-5 (Isolation Valves for Loop A) and Flow Restrictor upstream of valve NC-6
- NC-94, NC-95 (Isolation Valves for Loop B) and Flow Restrictor upstream of valve NC-113
- NC-13, NC-106, and NC-115 (Isolation Valves for Loop C)
- NC-1 9, NC-20, and NC-1 11 (Isolation Valves for Loop D)
Duke Energy Corporation Relief Request Serial #11-CN-002 Page 4 of 9 Segment 4 also contains 3/4 inch and 3 inch (NC) Class 1 piping and components between (and including) the following RCS double isolation valve pairs on the Reactor Vessel Head vent line:
NC-298 and NC-299 (3 inch)
NC-311 and NC-312 (%inch)
Pipe Size / Schedule 3" NPS / Sch 160 Material Type / Grade Stainless Steel SA-376 Grade 304 Design Pressure 2500 psia Design Temperature 650 degrees F Approximate Length 5 Feet Pipe Size / Schedule 2" NPS / Sch 160 Material Type / Grade Stainless Steel SA-376 I Grade 304 Design Pressure 2500 psia Design Temperature 650 degrees F Approximate Length 34 Feet Pipe Size / Schedule 3/4" NPS / Sch 160 Pipe and 6000# Fittings Material Type / Grade Stainless Steel SA-376 / Grade 304 (Pipe)
Stainless Steel SA-182 / Grade F304 (Fittings)
Design Pressure 2500 psia Design Temperature 650 degrees F Approximate Length 2 Feet Pipe Size / Schedule 1/2" NPS / Sch 160 Material Type / Grade Stainless Steel SA-376 / Grade 304 Design Pressure 2500 psia Design Temperature 650 degrees F Approximate Length 4 Feet Piping and components for Segment #4 are shown on Unit 1 drawing CN-1 553-1.0 and Unit 2 drawing CN-2553-1.0 enclosed in Attachment "A".
Segment 5 consists of 1 inch (NC) Class 1 piping and components between (and including) the following RCS double isolation valves on the Reactor Vessel Head vent line:
NC-250A, NC-251 B, NC-252B, and NC-253A
Duke Energy Corporation Relief Request Serial #1 1-CN-002 Page 5 of 9 Pipe Size / Schedule 1" NPS / Sch 160 Material Type / Grade Stainless Steel SA-376 / Grade 304 Design Pressure 2500 psia Design Temperature 650 degrees F Approximate Length 12 Feet Piping and components are shown on Unit 1 drawing CN-1553-1.1 and Unit 2 drawing CN-2553-1.1 enclosed in Attachment "A".
2. Applicable Code Edition and Addenda
ASME Boiler and Pressure Vessel Code, Section Xl, 1998 Edition through the 2000 Addenda.
3. Applicable Code Requirement
The ASME Code, Section Xl, IWB-2500, Table IWB-2500-1, Examination Category B-P, requires a system leakage test in accordance with IWB-5220 and visual, VT-2 examination of Items B15.50 and B15.70 during each refueling outage.
IWB-5221(a) requires that the "system leakage test shall be conducted at a pressure not less than the pressure corresponding to 100% rated reactor power".
IWB-5222(b) requires that "the pressure retaining boundary during the system leakage test conducted at or near the end of each inspection interval shall extend to all Class 1 pressure retaining components within the system boundary".
4. Reason for Request
The Class 1 piping segments identified in this request are equipped with valves that provide double isolation of the reactor coolant pressure boundary. Under normal operating conditions, these isolation valves are closed and these piping segments are subject to RCS pressure and temperature only if leakage through the inboard valves occurs. To perform the Code required pressure tests, it would be necessary to place the plant in an abnormal configuration by opening the inboard valves or installing temporary jumper hoses around check valves to pressurize the piping segments. Duke Energy believes that performing these tests under these conditions causes a hardship or unusual difficulty without a compensating increase in the level of quality and safety for reasons documented herein.
- 5. Proposed Alternative and Basis for Use Pursuant to 10 CFR 50.55a(a)(3)(ii), Duke Energy proposes the following alternatives in lieu of performing system leakage tests in accordance with the requirements of IWB-5221(a) or IWB-5222(b) [as applicable] for the Class 1 components listed in Section 1 of this Relief Request:
Duke Energy Corporation Relief Request Serial #11-CN-002 Page 6 of 9 Segments 1, 2, 3, and 5 Proposed Alternative:
In lieu of the pressure requirement of IWB-5221 (a), the system leakage test shall be performed at a pressure not less than 300 psig (for Segments 1, 2, and 5) and not less than 42 psig (for Segment 3). This alternative is proposed only for the system leakage test conducted at or near the end of the inspection interval in accordance with IWB-5222(b). The VT-2 visual examination shall extend to and include the second closed valve at the boundary extremity.
Bases for the Proposed Alternative:
Pressurizing Segment 1 to NC system operating pressure (2235 psig) during unit startup with the NC system at normal operating pressure and temperature would cause a hardship because of the high risk of an inadvertent Pressurizer Auxiliary Spray Initiation to the Pressurizer. This design transient is undesirable because it would force static piping contents (cold water) into the Pressurizer spray line, resulting in an additional thermal design cycle. The plant design only allows 10 of these cycles over the plant design life. For this reason, performing the system leakage test at the pressure required by IWB-5221 (a) is a hardship for Segment 1.
Segment 2 cannot be pressurized to NC System normal operation pressure due to thermal relief check valves ND-1 16 and ND-1 17 that are routed to the NC System. TS SR 3.4.14.2 requires verification of interlock to prevent opening the loop suction valves (ND-1B, ND-2A, ND-36B, & ND-37A) with NC pressure greater than or equal to 425 psig. Per the bases for TS 3.4.14, the purpose of the interlock is to prevent an intersystem LOCA due to inadvertent opening of a loop suction isolation valve. Opening a loop suction valve to pressurize Segment 2 piping from the NC System would defeat the required interlock. Additionally, opening valves ND-1 B or ND-36B to pressurize this segment would violate the 10 CFR 50.55a(c)(2)(ii) required double isolation valve barrier of the RCS boundary from the ND system. This would create an inability to mitigate a Loss of Coolant Accident (LOCA) if a break were to occur in the 12 inch diameter piping, reducing the plant's margin of safety. Valve ND-1 B or ND-36B could not be relied upon to close against the postulated flow from the RCS through a 12 inch line break. It would also subject ND system components to potential risk of damage with only a single valve isolating ND system from RCS pressure. For these reasons, performing the system leakage test at the pressure required by IWB-5221(a) is a hardship for Segment 2.
For Segment 3, no intermediate test connection exists on the piping between the check valve pairs to measure the test pressure locally. Modifications would be necessary to add test gauges upstream and downstream of the check valves to verify NC pressure. Aligning an NV pump to the Boron Injection flow path in Mode 3 (during startup) and cracking open valve NI-3 would constitute a Manual
Duke Energy Corporation Relief Request Serial #1 1-CN-002 Page 7 of 9 Safety Injection, diminishing the allowed number of Cold Leg Thermal Design Transients. This action would risk degradation of piping and welds (due to thermal fatigue) for the sake of verifying the leak tightness of the piping segment.
For these reasons, performing the system leakage test at the pressure required by IWB-5221 (a) is a hardship for Segment 3.
For Segment 5, opening either of the isolation valves in each valve pair to permit pressure testing would violate the 10 CFR 50.55a(c)(2)(ii) required double isolation valve barrier of the RCS boundary during plant operation. Opening the inner isolation valve would also expose personnel to unnecessary safety hazards because personnel would have to be stationed at or near the valves in lower containment with the RCS at normal operating pressure and temperature.
Because there are no test connections to the piping in Segment 5, testing of the piping between the Reactor Head vent piping double isolation valves by hydro pump or temporary jumpers is not possible. For these reasons, performing the system leakage test at the pressure required by IWB-5221(a) is a hardship for Segment 5.
The proposed system leakage test conducted at a pressure of at least 300 psig (Segments 1, 2 and 5) and at least 42 psig (Segment 3) is acceptable because leakage (if it were to occur) would still be detectable at this reduced pressure, although at a reduced rate. Also, if leakage occurs past the first isolation valve, leakage in the piping segment would be evident during the system leakage tests and VT-2 examinations performed during each refueling outage.
VT-2 examinations that reveal no leakage at the proposed test pressures provide reasonable assurance that no leakage would be detected during a test at 2235 psig. For this reason, the proposed alternative provides an acceptable level of assurance of the leak-tight and structural integrity of the piping segments.
Segment 4 Proposed Alternative:
In lieu of the requirement of IWB-5222(b) to extend the test boundary to all Class 1 pressure retaining components within the system boundary, the test conducted at or near the end of the interval shall be performed in accordance with IWB-5222(a), with the inboard and outboard isolation valves configured in their normal reactor startup position. The VT-2 visual examination shall extend to and include the second closed valve at the boundary extremity.
Basis for the Proposed Alternative:
During normal operation, the isolation valves in this segment are maintained in the closed position. Opening either of the isolation valves in each valve pair to permit pressure testing would violate the 10 CFR 50.55a(c)(2)(ii) required double isolation valve barrier of the RCS boundary during plant operation. Opening the
Duke Energy Corporation Relief Request Serial #11-CN-002 Page 8 of 9 inner isolation valve would also expose personnel to unnecessary safety hazards because personnel would have to be stationed at or near the valves in lower containment with the RCS at normal operating pressure and temperature.
Because there are no test connections to the piping in Segment 4, testing of the piping between the double isolation valves by hydro pump or temporary jumpers is not possible. For these reasons, extending the test boundary in accordance with IWB-5222(b) during the test conducted at or near the end of the interval is a hardship for components in Segment 4.
If any leakage occurs past the first isolation valve, leakage in the piping segment would be evident during the system leakage tests and VT-2 examinations performed during each refueling outage. In addition to leakage testing, boric acid inspections performed during refueling outages provide additional assurance that leakage from these components would be detected. For these reasons, the proposed alternative provides an acceptable level of assurance of the leak-tight and structural integrity of the piping segments.
- 6. Duration of Proposed Alternative The proposed alternative is requested for use during the following inservice inspection intervals:
Station , Unit ISI Interval Interval Start Date Interval End Date Number (Tentative)
Catawba 1 3 June 29, 2005 July 14, 2014 Catawba 2 3 October 15, 2005 August 19, 2016
- 7. Related Industry Relief Requests Relief from the requirements of IWB-5222(b), for various ASME Code Editions/Addenda, has been granted to other licensees, as documented in the following relief requests. Please note that the basis for Duke's relief request is not identical to any of the cited requests listed below.
7.1. Arkansas Nuclear One - Unit 1, Grand Gulf Nuclear Station, River Bend Station, and Waterford Steam Electric Station - Unit 3, Request for Alternative No. CEP-PT-002, Approved February 23, 2009, TAC Nos. MD8819, MD8820, MD8821, and MD8822.
7.2. McGuire Nuclear Station, Units 1 and 2 - Relief Request #09-MN-005, Approved June 14, 2010, TAC Nos. ME1 732 and ME1733.
7.3. Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 - Relief Request #PT-3-01 Approved February 12, 2009, TAC Nos. ME0112 and ME0113.
7.4. Donald C. Cook Nuclear Plant, Units 1 and 2 -Relief Requests #1SIR-23, ISIR -24 and ISIR -25, Approved February 18, 2009, TAC Nos. MD9438 and MD9439.
Duke Energy Corporation Relief Request Serial #1 1-CN-002 Page 9 of 9 7.5. Joseph M. Farley Nuclear Plant, Units 1 and 2, Relief Request #FNP-ISI-ALT-04, 05, and 06, version 1.0, Approved March 30, 2010,TAC Nos. ME1767 and ME1768.
7.6. James A. Fitzpatrick Nuclear Power Plant - Relief Request #RR-01, Approved December 27, 2007, TAC No. MD4753.
7.7. South Texas Project, Units 1 and 2 - Relief Request #RR-ENG-2-51, Approved November 12, 2008, TAC Nos. MD8951 and MD8952.
7.8. Ginna Nuclear Power Plant - Relief Request #23, Approved May 5, 2009, TAC No.
ME0456.
7.9. Indian Point Nuclear Generating Unit No.3 -Relief Request #RR-3-44, Approved February 5, 2009, TAC No. ME001 1.
7.10.North Anna Power Station, Unit No.2, Relief Request #SPT-014, Approved October 9, 2009, TAC No. ME1104.
7.11 .Point Beach Nuclear Plant, Units 1 and 2 - Relief Request #RR-22, Approved July 21, 2010, TAC Nos. ME2146 and ME2147.
7.12.Seabrook Station, Unit No.1 -Relief Request, Approved October 15, 2010, TAC No. ME2418.
7.13.Shearon Harris Nuclear Power Plant, Unit 1 - Relief Request #13R-04, Approved January 21, 2009, TAC No. MD8744.
- 8. References 8.1. Catawba Flow Diagrams (Attachment "A"):
8.1.1. CN-1554-01.00 8.1.2. CN-2554-01.00 8.1.3. CN-1561-01.00 8.1.4. CN-2561-01.00 8.1.5. CN-1561-1.1 8.1.6. CN-2561-1.1 8.1.7. CN-1562-1.0 8.1.8. CN-2562-1.0 8.1.9. CN-1553-1.0 8.1.10. CN-2553-1.0 8.1.11. CN-1553-1.1 8.1.12. CN-2553-1.1
Catawba Units 1 and 2 10 CFR 50.55a Request Number 11 -CN-002 ATTACHMENT "A" Catawba Nuclear Station Flow Diagrams (12 Drawings Attached)
NOTE: These drawings are non-proprietaryand provided for Information Only
I. I I It I 13 1 1 05014 CNO
-1554-01.W00 I LL CCD
-- cm-
__D- -- -co cLmA C-=CEo4-3[I I-I FlmW a.
k ~ lfi LS f uas.~n PC IL u S.an s alfl.
~flm7~C3-9 fiinM Il - - O nSoSlmlS 1f-f 47 4 1501 IC- 5
~~
l~~~i~~jlft 2 2SI!C U .5 155 =
Ii. III I 1l5111 C5 *:F IU~I -29 I *~
-~ *~J - CIII 51124 F C1 cam~f g ~ *SIfltI 45:45410 llt a~C * ~ f
II I I IA I I I II I II I IS I
-t - -
iWMC CN-2554-0I.00 i L L
-- C-= -A -'m
~T x
L
[~L ~
i i *2.
I TII .1 -c , -
~ ma m i ~
aAm L EI I-M I-X31J~st.
a' C30 .:at6' ",."
0 a
-'U "-
<T .,-T C.3J CEXll 9-'*
4-- @N 01 o -ifT.---
Fýx L..
CEOd Fo- AI;ACHMENT PAGE 2 OF IZ ZZ- R-C,1 7 -T--F C.-I I
I I i i-Am GA CONDITION 2 A.... . . . . . " ' ? - fl ,tlIE OA CONDITION 1
!i I - - I--
,* mmau1 .ti s a ma, MII am flint.
2.L,!er.
ma flfl EAft I i il i I o CATAWBA NZAIRI STATION FL.0W01I4AiMOF CFNICA. I, VLLE COTRO SYSITE(WN)
UNIT2 A
..1 JmCLNWeyIl.LI *1 -
.,,*,,,I,,, I-I*1..a1-1a ' . le. MM.. CN-2554-01. 00 I I..
I to I ft I I ft I ft I LW I Li I III I IJ I LN
I I I I I I A I a I I 7 1 S 1 S . I I In II ID0MI0CN- 1561-01.005 f L
-cum 0
Ii , ,
CEO-31 -C.ý'9 A~A 0 .
33M -0 CWC I L. I 9p-a l a%
A F
5N p A AIN*
@ý__7 c
J ýTot y
10 IF-1.
412,
- V JE?
A
_c3m CIZEM- CDEK3-I cum- r ---- ---- I CEED II-A
-c aNOT a 9mmaa
- a
@i E
- r m I
- L L~ ~J 7 1-P L-.-_!. OEM-I Cý Otis c
AIIACHMEKT'$
MAE 3 OF M2 I NAogaI-pRopatE.TAAýC OA CONDITION 2 Ata.ac St. - m aý_ CATOSSSNILELDISTATINIAIT I FR _Nn TII.? va l ata. RESIDUAL HEATllMDlVAL SnT70 INMt A
itt at, a-a-ac tm.I.c tat -
atstan. maa.CaIn ~ .A.L...........nr.fn....
a Islam vam mm N_ -
n J I I - 1 64 i
.. . . . I i I SI I I II I --
I -cam CRE3D-
-- &a Lý CAD CEE3D-C.,
POO 4
Lim CE- -29P T '- ------------Q CllE3D-F
-c=iK--+O Fum
=
63m vm.- 7Z, FOX)
-- Cori) (REm-I PM ATTACHMENT1+
PAIE4. IF IZ A-4)i "ll L'J.
"tIW- Fkpkte1ArAW
[I
¶ 0ý -i-7Ii~ i I
ý OA CONDITION 2
. CONDITION O
CATAWBNLA 1
ArIO .lt2 IM DEIJ.
ý WETVAE WON-o~~
A A
,M I II
ýlslm .. - - - -
ISCNOGPTF9]- M ,., IMAMILIS-56L-Mi. -
AA t I* I a* I a*
I I I I I I -
I~ I I L
K WL6 15-a~ I .a,~. -----
E*-z 2 C3 a
-oo
- 10 I
U (DF24 c
At1ACHMENT PAOE!5 IFI2-4'A6,4- P"EAXJ" I a GA CONDITION 2 7 ~ OU ~
0--l-r.-
C~ -.- 2.A~
pa LONDITIO 1 21 .om va m a& n, 1 1 1 1 1 1 1 C TAWBAWXLEAR STATION M IT I Aft FLW DIAORf OF RESION HEAT F*VOVAL I I i I I I I A
A. I ill NO 124 MV MEW-ICE Ale AN-M w =1 mw1ý1 mv 1, ON f6JSW I~~~~~~3- Im 5 I 7 I S I 9 S I I D I 13 560I 1 I I I I I 1 0 1 5 1 7 1 9 1 9 1 111 1 11 1 Iz 1 13 1 14
I1 1 12 1Imm NmJ3CN-2561 I-1.
1 14 V4-K L
CREXý-
CEE3D--
awl),
2-1a - I.
16 -a=
r 016 CRUE:)-
FED CEE3>- CEE3D-- DIM-
-Cý arn-QZE3>-
C23m-U Pin.
- Ite~ar. b IA PR lap" C C j
AIIACHMENT ME 1, OFJZ it týg AMRi GA CONDITION 2 OA CONDITION 1 Aft.. coeý CATAWBAMCLEARSTATION UNIT2 FL. DIAORAM OF Am RESIDUA HEATREMOVAL SYSTEM (NO)
A A ! a.OAPI.t~ asi S.
~mIIW w iCm a
-ll as hL!3,T . ~
6-.
Xa&..W 1 3 1 4 1 -, - 6 1 7 1 a I I 0~~~~1 I I IS 12 I 13 I I
I n I
-4 -. -~
NC IOWM. CN-1562-1.0 it1 scQm~erI-r f43 i-.
6 14-AHlAtMENT 4 PAGE7 OFI GACONDITION 2 OA CONDITIONI
. i pi i I iI 1 1 - ccI.~~
1 ! ý 1 ! 1 CATAW114 WJCLE0A TATION UNITI Imudm -.Lm m -mml wa L -ý FLOWOJAO O ALM IVOIIM ME' V(I t A IMVIM - OW, NtL CN- 1502-1.
I I I I I I b I
-r I II I II I I A A
I I B i
I A
B I I I A a
I I a I
I is I
I
. I I,
i 1DM0.I. CN-2562-t. II-6EeqM~4T
- 3 QA CONOT ON 2 rnw' - OA CONDITION i mlan~~
- Caifl 9 BiBS I, a, - MSit uS 9 II IT CATWBSTATION 2
~. iiSlS=OBPiBP In PBPeN lOl Gti ~
AE maIqlm I ~.Il
- ~~~~~ s~
e~~~~~~~
hr. ~~.
5 1-lS si ai SAFETYINJECTION I~ll SYSTEM l
-jSftCNOO TFBI 0 I 5- 2 0 I 5 I IVI I II I I ES I IS l I =* I J I 4 I 0 I b t I I* I 14
-.4 1 -.- 10 1 11 1 12 1 13 1 L4 1.
~ Ia~4.N53-1.0 C=
-1(-1z L-Y---
L cy ( L
-- r
_ a-it A-------
S -~~tMI.
.668.-2 2t 8 ta Caa 9~~; 28
. 2E~
a~miJ i- , -
.4- ttIt ~
.15r10 FIE
= - "S 4 sit]
F E0,, Yc. Y I C r-E D 4i~28 E
atTt a-a -- -- - - La E 0 28 ~ 22I t c a S*:a- ----------
a-
~ ~AIaACHENT
-CK E
~~~aa-a amaaa v~ao~E-mtt~maa-maat 04 CONDITION 2 OA CONDITION I a.a-aia at.
a-t Ca . - ma-aaa.
m-_ ta.paH R It%
i~itatErlaitMatCPMai- ftH i.StiOtlPa-ittati ip 2fl na-a. , a Pa iita itaa a--pati SI
, I~
IN iia otwtl r-waatt.t
% uhi t
Tiaa.
t ii aS C tt 1M-i i
tttB-tptaC ra~t--Iiaatt-ii tmcAi aPt-t al-u ittn tais~..,
L- ~ 'S - t-CM i-tT na
-- l i-a
`7 0 a-Lot DIotaa 5 -CA-CTOR COOLAHT SYSiTEMi a ON-1 553-1.0 1 1 2 1 I 1 7 1I a 1 w
a-4t g
- I A* I m 3 I--
i iJ '
L-<- W. T-_ NO CN-253-1.0 L i C'
i (I2 255 25 Ž i IJ- 'i AC.ICIAU , ALI_ - -------------- '25 A ~ /I 25........ 25----- -
* **t C H-1D * * . --- H
,,. ?,,, -i
- I-* 2*. , .* * '*.7
- L* '--'------,-----.---
0 tED ___ flflt~ Y/\.~ŽaCt F
, .25
-XQM-CP * -*A.,. AU~LNA2 5) 2 25 .......
--- -*1__ -- "----- -I.--2-5u , f ,-
~=25~
ý_. ... ,-, "----. . . f . -. 0
@÷ 25 ,:,_ . .. .
i25i
,"M I .. AtACKEKI E OF 1-4...$ . .
i (. -,1 , *,
O.A. CONDITION 2 IA.R WOWAA.A.AS44 IC ONCJ -M. c Oýh MWNOTl.
A..T=aAUA.S I A A.AIIA CA~lIALA~.~
IC CL C ~lSCI.l~a I CA.Is. A TnA -A.EIA LCA? ICC aCLWAS I j CtCT.
21AA 1 ~T..SCI CAIA IA IA LCCC.IA CC ACAC TAIC I I I I I I I ! IO ( 4 "I . .UNIT..
2.
C~~Al.55
AIMIICCI M =LL A~tCA~.L W.
- 1C UCI . ILAIA , 1w I I I I 1 I! I I SCO.
CCA.AAIA A.iAItA.. lSA~a....CACM.l r
-~ ~- -, . *~~I lIIIIIIIA.CCIICI. IAASI C2 L.U~U.L2.L~L.C I 1*
1 2 1 I *l I ll* I LL il I I3 1 1,.
1 1 4 1 1 7 1 . 'i I--- I'R Iia~ IG ~ N 1O 3 1.
Iowa 4m CN- 1553- 1. 1 71 I i . I ,
I1 L
K i
Gyam 21 F"" 1~
2L 1-
~oN~~n= a a,,
, ina maFe),
r-~-7 F
ao c I 0ý2\U1a~ I 1
'2~y - NPAGE 11 OF12.
f q 21 1"4 ~ (R 21 ~ E-1,
~ ---- 9 -
__ n~- -1. -- ~~1 OA CONDITION 2 OA CONDITION 2 1510- -, I Ir WOO.
AilMIHO STAIWON UN1l7I W~~Da &F ;fim me.
FTo V 0090 oF
-n ,trha pa-7-97 a ic V,
[O 9 LMTi SYSOeOV t'~a KMrV 1V WR ito a a-...i meon cli A
~~m ~Em Itn,~aOoP1 WN:i JMI =1 19 proip i mt W.aiP-in00a a O Iti~if "i'~SPt tW ~ ~ rP ONPrti i.Pmf Sie b.m RoVISIOS - - -. 1- 1 F~S JDO. NO CN-1553-I.]
4- 71 7 a 1 9 1 ~ i a pa 5
Q I I* I 11 I IEAVQNýCN-2553-1. I hy L L MM-~ A4 -M
- O I' ft' 04 *r Fini ci
'aft 0Th t~f o
<4 H G
c='
bI'f n~n -
I-cd F
r r F
- 91.1-1 '
,::A.-ýi ft E F
r-- ---- C=9~L
- e Cm-AVIACBHMENT I PAGE JZ OFf7z
\19 ~ -7f c t----------U
fcf 9
u~II-uf ----------
E
- - =lm-- .- .- 19 OA CONDITION 4 OH CONDITION 2 OA CONDITION I i~~f ~ ~ ~I~Ot~fttf~tf~ - 53 ej.E i-c -OWR ..-. w~f X; Ail i utettec ttiM~ii C ~ L ~FL EAC W HIAGRAM TORCOOLAMT OF SYSTEM "a REIVM itaclr -511 ct L.N-i2553-1.-1a