ML20086N214
| ML20086N214 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 02/15/1984 |
| From: | Tucker H DUKE POWER CO. |
| To: | Adensam E, Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20086N217 | List: |
| References | |
| NUDOCS 8402170292 | |
| Download: ML20086N214 (16) | |
Text
o DUKE POWER GOMPANY P.O. DOX 03180 CIIAHLOTTE, N.C. 28242 IIAL B. TUCKER Tzteruoxz N*"
.J".",". '""
February 15, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4 Re: Catawba Nuclear Station Docket Nos. 50-413 and 50-414
Dear Mr. Denton:
The purpose of this submittal is to provide supplemental information supporting our submittal of November 18, 1984 concerning elimination of arbitrary inter-mediate pipe breaks on Catawba Nuclear Station Unit 2.
Enclosed is the informa-tion as requested by the NRC Staff during our February 7,1984 meeting.
Additional technical justification is provided in the form of a revised and expanded Attachment B to our November 18, 1984 submittal. An additional attachment, Attachment E, is provided to indicate the location of deleted and remaining breaks, including protective devices. Areas of NRC concern, as discussed on February 7,1984, are addressed in the attached information as follows:
Item of Concern Additional Information 1.
Carbon content of Duke piping and Attachment B-1 comparison to BWR stress corrosion 2.
Thermal gradients / mixing and Attachment B-2 preoperational vibrational testing 3.
Environmental analysis methods Attachment B-4 4.
Locations of deleted and remaining Attachment E breaks, including protection devices In order to implement the revised criteria, we hereby take exception to those portions of the Standard Review Plan Sections 3.6.1 (ASB 3-1) and 3.6.2 (MEB 3-1) which deal with thh type of pipe break. An exact description of the exception is given in Table 3.6.1-3 of the Catawba FSAR, as revised per Attachment D of our November 18, 1983 submittal. The proposed criteria change will result in deletion of approximately 20 percent of the pipe rupture devices currently planned for Catawba Unit 2.
However, the remaining 80 percent of-the protective devices provide, in our opinion, an acceptable level of protection for the plant.
In addition, no support or snubber changes will be made due to the deletion of these pipe rupture devices. Technical justification for the change is presented in Attachment B to this letter. Benefits which will be realized are summarized in Attachment A to our November 18, 1983 submittal.
?O f
)Y
Mr. Harold R. Denton, Director
- February 15, 1984 Page 2 As emphasized at the Duke /NRC meeting on revised arbitrary inten11ediate break criteria, it is of critical importance to-receive a decision concerninC the break deletions in the very near future. We request a decision by March 1, 1984 in order to obtain previously stated benefits on Catawba Nuclear Station Unit 2.
After that date, we must begin analysis, design and construction of the protective devices for these arbitrary intermediate breaks in order to meet the project schedule for Unit 2.
Therefore, after March 1,1984, the benefits of a positive decision f.egin to diminish.
If I can be of further assistance, please contact me.
Very truly yours, I
Hal B. Tucker I
ROS/php Attachment cc:
(w/o attach. ment)
Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30303 NRC Resident Inspector Catawba Nuclear Station Mr. Robert Guild, Esq.
Attorney-at-Law P. O.~ Box 12097 Charleston, South Carolina 29412 Palmetto Alliance 21351 Devine Street Columbia, South Carolina 29205 Mr. Jesse L. Riley Carolina Environniental Study Group 854 Henley Place Charlott.e,- North Carolina 28207 i
f
g ATTACHMENT B Technical Justification 'for Elimination of Arbitrary Intermediate Pipe Break Postulation
.The following reasons provide generic technical justification for eliminating the arbitrary intermediate pipe break postulation required by Standard Review Plan 3.6:
1.
The pipe rupture " threshold" for all nuclear class piping is 80% of the 1974 ASME Code stress allowables. All arbitrary intermediate breaks involve stresses below this level. Hence a -large conservatism exists.
2.
Pipe rupture is recognized in Branch Technical Position MEB 3-1 as being a " rare event which may only occur under unanticipated conditions."
3.
There is no technical or other justification for postulating arbitrary intermediate breaks, other than providing additional conservatism.
4.
The additional pipe rupture devices resulting from this additional
" layer" of conservatism may actually reduce rather than improve plant safety. This has been demonstrated ir. " Effects of Postulated Event Devices on Normal Operation of Piping Systems in Nuclear Power Plants," NUREG/CR-2136, Teledyne Services,1981. Included among other improvements from arbitrary break elimination is improvement in performing ISI and a reduction in unanticipated restraint of piping due to thermal growth and seismic movement.
5.
Due to system design and operating procedures at Catawba Nuclear Station, the probability of stress corrosion, thermal and vibrational fatigue, or water hanner in the arbitrary intermediate break lines is not significant.
Technical justification for this positicii is given in Attachments B-1, B-2, and B-3.
6.
Environmental qualification due to pipe rupture will not be affected by elimination of arbitrary intermediate breaks, as explained in Attachment B-4.
It is concluded that the elimination of arbitrary intermediate break postulation is technically justifiable for the foregoing reasons.
i l
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ATTACHMENT B-1 Protection of Arbitrary Break Lines From Stress Corrosion In order for stress corrosion cracking to occur in piping, the following three conditions must exist simultaneously: high tensile stresses, a susceptible material, and a corrosive environment (NUREG-0691). Since some residual stresses and some degree of material susceptibility exist in any stainless or carbon steel piping, Duke Power minimizes the potential for stress corrosion by preventing the occurrence of a corrosive environment.
Strict pipe cleaning standards prior to operation and careful control of water chemistry during plant operation are used to prevent the occurrence of this environment.
All piping involved in the elimination of arbitrary breaks at Catawba is either austenitic stainless steel or-carbon steel, as shown in Table 1. The Stainless steel is Type 304 and Type 316, and as such the carbon content is limited to a maximum of 0.08 weight percent. None of the higher carben content types (304H, 316H) have been useo.
The environments known to increase the susceptibility of austenitic stainless steel to stress corrosion are (NUREG-0691): oxygen, fluorides, chlorides, hydroxides, hydrogen peroxide, and reduced forms of sulfur (e.g., sulfides,
sulfites, and thionates). In carbon steel, these same environments plus a few additional substances such as caustics and nitrates are thought to increase susceptibility.
Prior to being put into service, piping at Catawba Nuclear Station is cleaned internally and externally, and water chemistry during flushes and pre-operational testing is controlied to maintain this cleanness according to written specifi-cations. External cleaning for Duke Class A stainless steel piping include pc'ch tests to monitor and control chloride and fluoride levels. For preopera-tional flushes influent water chemistry is controlled, with requirements on chlorides, fluorides, conductivity, and pH being included in the acceptance criteria for piping of the material type and class included in Table 1.
During plant operation, primary and secondary side water chemistry is monitored in the carbon steel and stainless steel piping. Contaminant concentrations are kept below the thresholds known to be conducive to stress corrosion cracking. Table 1 shows the major water chemistry control standards for the lines in which arbitrary breaks were previously postulated.
Oxygen content can be more strictly controlled in the Catawba PWR environment than in a BWR environment. Oxygen concentration in the fluid in the Catawba stainless steel piping is expected to be less than 0.005 ppm during nonnal power operation, whereas the steady state oxygen content in a BWR systeu is approximately 0.2 ppm. Thus this condition which facilitates cracking in BWR's is not present at Catawba.
Note that a number of the lines involve operating temperatures less than 2000F. Any stress corrosion at these temperature 3 would be extremely slow; it is an industry-wide assumption that stress corrosion is not a problem at temperatures this low. Also note that steam generator water chemistry is the major factor controlling steam generator blowdown and main steam
.i chemical composition.
2/13/84 JKR/sr
. _. _ -~. _
TABLE 1 WATER CHEMISTRY REQUIREMENTS DURING PLANT OPERATION Page 1'of 2 FOR LINES WITH PREVIOUSLY POSTULATED ARBITRARY BREAKS MAX.
DUKE -
CHLORIDES &
MAX. CATION N0. ARBITRARY PIPING PIPING OPERAT MAX. 0 HYDR 0 GEN FLOURIDES CONDUCTIVITY
- BREAKS ELIM. -
TEMP (fNG (ppm) 2 PIPING SYSTEM, MATERIAL CLASS F)
CONCENTRATION (ppm)
(umho/cm) pH*
CATAWGA 2 R: actor Coolant' SS A
557 0.10**
25-50cc/kg(H 0) 0.15 12 2
R sidual-Heat Removal SS A
557 0.10**
25-50cc/kg(H 0) 0.15 8
2 Auxiliary-i Ferdwater CS'
.B 134-445
'0.003 0.2 8.8-9.3 8
Safety Injection (l)
SS A
557 0.10**
25-50cc/kg(H 0) 0.15 11 2
Safety.
y
' Injection (2)
.A Ambient 0.15 Low 5
^
Safety 3
Injection (3) '
.SS B
Ambient
-0.15 Low 11/
Steam Generator Blowdown (l)
SS B
557 0.02 0.8 8.5-9.3 6
Steam Generator Blowdown (2).
CS F-557
.0.02 0.8 8.5-9.3 8/
' Chem'& Vol.
Control (1)-
SS A&B 130*
20/
Chem'& Vol Ccntrcl(2).
SS B-290 0.15 2/
. Main St:am CS B&F 557 0.3 8.8-9.3 14/
Main Steaa--
to Aux Equip CS B
557 0.3-8.8-9.3 4/
t Main Steam 0.3 8.8-9.3 8/
Vent.to Atmos CS B
.557
- - _ _. _ _._ _ _ _ _ _ _._._.m.
~ TABLE 1 WATER CHEMISTRY. REQUIREMENTS DURING PLANT OPERATION
.Page 2 of 2.-
FOR LINES WITH PREVIOUSLY POSTULATED ARBITRARY BREAKS
..c-.
++Stressicorrosion not considered a problem at this operating temperature
~* Based on EPRI-NP-2704-SR (Steam Generator Owner's Group Secondary Water Chemistry Guidelines.
.These standards ~are to be met.at Catawba Nuclear Station).
- Data is-:for the portions'of the system where arbitrary breaks were previously postulated.
- Tech.-Spec. limit; 0xygen'~ concentration expected to be less than 0.005 ppm during most power operation.
/ Estimated.- evaluation.not complete.
J 9
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'.1/ 20/84
x ATTACHMENT B-2 THERMAL AND VIBRATION FATIGUE IN PIPING ASSOCIATED WITH CATAWBA ARBITRARY INTERMEDIATE BREAKS For Catawba Nuclear Station non-class 1 ASME Code lines in general, the Code design allowables are intended to prevent fatigue failure. For Class 2 and 3 piping components, fatigue failure protection is provided for by the allowable stress range c. heck for thermal expansion stress. This stress is included in the break stress ratio for all non-class 1 breaks. And even after elimination of the arbitrary intermediate breaks, the cut-off for postulating mandatory breaks (" threshold") is still 80% of the Code allowables.
For Class 1 (Duke Class A) lines the conservatism allowed for fatigue failure is even more obvious. The ASME Code limit for the Cumulative Usage Factor (CUF) is 1.0 to assure that pipe failure will not occur. The pipe break postulation limit is 10% of this number, and most of the Class 1 arbitrary intermediate break locations involve CUF's far below this limit.
NUREG-0691 reported a number of instances of pipe cracking due to thermal fatigue in PWR systems. All cases involved either feedwater or auxiliary feedwater piping and occurred in the vicinity of steam generator nozzles.
The cracking was attributed to cyclic thermal stresses in horizontal runs at the nozzles.
Cyclic thermal stress is minimized in the main and auxiliary feedwater piping at Catawba Nuclear Station by limiting mixing of low velocity, low temperature feedwater with high temperature water in the steam generator nozzles. Mixing is prevented in the auxiliary feedwater supply by providing vertical or upward sloping piping in the vicinity of the steam generator and monitoring temper-ature near the nozzle to alarm high temperature backflow. Mixing of low velocity, low temperature feedwater with high temperature water is prevented in the main nozzle by isolating flow to-thg main nozzle and using the auxiliary feedwater nozzle, below 11% power or 250 F feedwater temperature. At this condition, temperature is monitored near the main nozzle to alarm should inflow of low temperature feedwater occur.
At Catawba Unit 2 only eight of the approximately 116 eliminated arbitrary intermediate breaks occur in the auxiliary feedwater system (none in feedwater system), and none of these are located in horizontal pipe runs at the nozzles.
Cyclic thermal stress is prevented in the arbitrary break lines in the remaining systems (listed below) by maintaining uniform temperatures with no mixing:
1.
Steam Generator Blowdown 2.
Main Steam Supply to Auriliary Equipment 4.
- 5.. Main Steam Vent to Atmosphere 6.
- 7. ' Safety Injection 8.
Chemical and Volume Control 90 e
e -
The potential for vibration fatigue in Catawba piping systems is minimized through pre-operational vibration tests. Table 14.2.12-1 of the Catawba FSAR describes the Piping System Vibration Test which is part of the Preoperational Test Program at Catawba. The purpose of the test is to verify the-following:
a.
Piping layout and support / restraints are adequate to withstand normal transients without damage to piping systems, and b.
Flow induced vibration is sufficiently small to cause no fatigue or stress failures in the piping system.
During the test, points on the piping systems with large displacements are selected for measurement of piping velocity. These measurements are evaluated with respect to the following acceptance criteria and any required modif f-cations to achieve acceptability are made:
A.
Steady State Vibration Testing Acceptance criteria are based on conservatively estimated stresses which are derived from measured velocities and conservatively assumed mode shapes.
B.
Transient Vibration Testing a.
No permanent deformation or damage in any system, structure, or component important to nuclear safety is observed.
b.
All suppressors and restraints respond within their allowable ranges, between stops or with indicators on scale.
c.
The measured piping vibration for Reactor Coolant System during reactor coolant pump starts and trips do not exceed the values specified by Duke Power Company Design Engineering Department.
All piping systems that contain rupture devices to be deleted under our proposal are included in the Vibration Test Program. Catawba FSAR Tables 3.9.2-1 and 3.9.2-1 (a) list the systems to be included in the steady state and transient test programs, respectively. These tables are included on the following two pages.
Based on the information presented in this attachment, no problems from thermal or vibrational fatigt.e would be expected at Catawba Nuclear Station.
Table 3.9.2-1 Piping Systems Included In Vibration Test Program System Reactor Coolant System Safety Injection System Residual Heat Removal System Containment Spray System Chemical and Volume Control System Boron Recycle System Baron Thermal Regeneration System Component Cooling System Liquia Radwaste System
(
Fuel Pool Cooling and Cleanup System i
Diesel Generator Fuel Oil System Diesel Generator Cooling Water System Diesel Generator Lub Oil System Nuclear Servict Water System Refueling Water System Main Steam System Feedwater System Auxiliary Feedwater System Steam Dump System Control Area Chilled Water System Steam Generator Blowdown Recycle System l
Recirculated Cooling Water System
(
1 Rev. 1
,s
-s
~,
O TABLE 3.9.2-la l
Piping Systems Included In Transient Vibration Test Program
(
\\
Systern Transient Type Type Measurement Simultaneous Test NC NC Pump Start Vibration measurement HFT at selected points NC Pump Trip Vibration measurement HFT at selected points NC PROV Cycling Post transient inspection HFT BB Initiation of S/G Post transient inspection S/G BD Test Blowdown Isolation of S/G Post transient inspection S/G BD rest Blowdown CA (Motor driven Pump Post transient inspection Aux. FDW F.T.
Start)
(Motor driven Pump Post transient inspection Aux. FDW F.T.
trip AFWPT Cold Start Post transient-inspection Aux. F0W F.T.
AFWPT Trip Post transient inspection Aux. FDW F.T.
CF Isolation Valve Post transient inspection HFT Closure NI NI Pump Start Post transient inspection ESF of pump discharge piping CCP Pump Start Post transient inspection ESF of pump discharge piping NV Letdown isolation Post transient inspection HFT SM Main Steam Isolation Post transient inspection SM isolation HFT (individually)
Main Steam PORV Post transient inspection HFT Discharge CF, SM Loss gf Electrical Post transient inspection
-Power Escalation Load 50, 100% FP Testing A.
Rev.-7 s
ATTACHMENT B-3 UATER/ STEAM HAMMER IN PIPING ASSOCIATED WITH CATAWBA ARBITRARY INTERMEDIATE BREAKS The potential for water hammer may exist in some of the systems involving lines where arbitrary intermediate breaks are being eliminated (e.g.,
Auxiliary Feedwater and Steam Generator Blowdown). However, steps have been taken to minimize the probability of significant water hammer actually occurring in these lines. Water hammer in each of the systems involved in elimination of arbitrary breaks is discussed below, i
1.
Steam Generator Blowdown Fluid flow through the Steam Generator Blowdown Lines is normally two-phase and of 0-10% quality. There is very little chance of water hammer in these lines inside Containment. A greater susceptibility to water hammer exists outside Containment, with the greatest potential occurring upon reinitiation of flow following containment isolation. Some problems have occurred in this area at the operating McGuire Nuclear Station.
However, design and operating procedures have been improved for McGuire and Catawba, and minimal water hammer problems are expected due to provisions to gradually repressurize the downstream piping before establishing full flow. Since these changes have been initiated at McGuire, there has been no reoccurrence of water hammer induced damage.
2.
Auxiliary Feedwater The Auxiliary Feedwater System is designed to minimize the probability of significant water hammer occurrence. The nozzle at the steam generator involves a 900 elbow connecting immediately to a vertical run of pipe to minimize steam voids. Also, tempering flow is maintained so the line will be filled with water at all times.
Hewever, it is recognized that some potential for water hammer in the auxiliary feedwater lines exists. Consequently, the temperature in these lines is monitored so that they may be filled slowly and flow initiated gradually when steam voids are suspected. The piping involved in the arbitrary intermediate breaks contains thennowells allowing such temperature increases to be detected and the proper operating procedures tc be implemented; also the associated valves j
are periodically checked for leaks. The Auxiliary Feedwater System at the McGuire Nuclear Station is similar to the one at Catawba, and McGuire has experienced no water hammer problcms to date.
3.
Reactor Coolant Overall, there is a very low potential for water hammer in the Reactor Coolant System since it is designed to preclude steam void formation.
Some problems were experienced ini;ially at McGuire, but shut-down procedures have been modified to eliminate water hammer at McGuire and Catawba.
@O g
4.
Residual Heat Removal The portions 'of the Residual Heat Removal System having arbitrary intermediate breaks are within the Reactor Coolant System boundary prior to (or at) the first valve off the Reactor Coolant Loops.
Therefore the explanation in part 3 above is applicable here.
5.
Safety Injection The Safety Injection System lines are all either water solid or gaseous lines. Steam voids would not be expected, especially since approximately 60% of the arbitrary intemediate breaks are in ambient temperature lines. In general, valves in the system are slow acting and operating procedures are designed to prevent water hammer. There would obviously be no problem in the gaseous lines. Therefore there is a low probability of water hammer problems in this system.
6.
Chemical and Volume Control All but two of the atbitrary intermediate breaks in the Chemical and Volume Control System are in low temperature (130 F operating) lines.
0 These nomally water solid lines would have a very small probability of steam void formation, and no water hammer events would be expected.
For the two breaks (outside Containment) in higher temperature lines (2900F), operating procedures for the system have been developed which minimize the probability of water hamer occurrence.
7.
Main Steam Supply to Auxiliary Equipment These lines have redundant heat tracing to prevent water accumulation.
Although the heat tracing is not nuclear safety related, it is required to be functional for system operation. Therefore the potential for water hamer problems is very low.
8.
~
Because of the design and function of the two-inch Main Steam System lines containing arbitrary intermediate breaks, no water hamer problem exists. The two-inch warm-up lines are normally filled with steam and the steam drain lines are designed for water flow with no accumulation.
9.
Main Steam Vent to Atmosphere The breaks in the Main Steam Vent to Atmosphere System are in the vicinity of the PORV's which release steam to the atmosphere. There is no chance of water hammer as long as the Main Steam line stays drained. McGuire experience has shown this to be the case.
There is only very minimal probability of significant steam hammer occurring in any of the lines involved in elimination of arbitrary l
breaks.
_ The function of the main steam line drains is to remove any liquid from the main steam lines. Significant water or steam hammer will not occur without liquid.
l l
l l
JKR/sr 2/6/84
ATTACHMENT B-4 ENVIRONMENTAL ANALYSES IN RELATION TO ELIMINATION OF ARBITRARY INTERMEDIATE BREAKS Concerning the environmental analysis due to pipe rupture for Catawba Nuclear Station, there will be no change in the results of this analysis due to elimination of arbitrary _ intermediate pipe breaks. In the Duke environmental analyses, a full circumferential break is assumed at every point on high and moderate energy piping in each compartment. The break which provides the worst environmental effect (temperature and humidity) is used to establish environmental qualification parameters. Therefore the break postulation for environmental effects is performed independently of break postulation for pipe whip and jet impingement, and will not be affected bv elimination of arbitrary intermediate breaks.
Through-wall cracks, as addressed in Section 3.6.2.1.2.3.c. of the FSAR (P.3.6-18 of Rev. 9), will continue to be omitted from postulation in moderate energy piping inside Containment. All scenarios involving moderate energy through-wall cracks are enveloped by high energy piping environmental analyses performed per the existing criteria as described above..
(
6 re-
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l ATTACHMENT E
]
Postulated Pipe Break Location Informatica This attachment coatains pipe break location and related information for Catawba Nuclear Station. Pipe break locations are shown in composite plan dews for the Reactor Building and Doghouses. For the Auxiliary Building break locations are indicated cc individual piping math models.
For each break, the following information is presented in tabular form:
break and hinge nodes, break elevation, protective device type, and pipe diameter and ASME class.
The data presented in this attachment for the Reactor Building pertains to Catawba Nuclear Station Unit 2, for which arbitrary break relief is being requested in this submittal. For the Doghouses and Auxiliary Building, Unit 1 data is presented rather than that for Unit 2 because the Unit 2 pipe break interaction evaluation is not complete in these buildings. The piping outside Containment for the two units is generally mirror image. Thus pipe breaks and associated protection in this area are similar but not exact between the two uriits.
NOMENCLATURE APPLICABLE TO CATAWBA NUCLEAR STATION PIPING SYSTEM /l.INE TITLE BB Steam Generator Blowdown System 7
CA Auxiliary Feedwater System CF Feedwater System NC Reactor Coolant System ND Residual Heat Removal System NI Safety Injection System NV Chemical and Volume Control System SA Main Steam to Auxiliary Equipment SM Main Steam System SV Mein Steam Vent to Atmosphere BB(Outside Steam Generator Blowdown System Contain)
BB-201 Steam Generator Blowdown System (Loop A)
BB-202 Steam Gen::rator Blowdown System (Loop B)
BB-203 Steam Generator Blowdoun System (Loop C) i BB-204 Steam Generator Blowdown System (Loop D)
CA(Outside
~ Auxiliary Feedwater System '
Contain)
CA-201 Auxiliary Feedwater System-(Loop A)
CA-202 Auxiliary Feedwater System (Loop B)
CA-203 Auxiliary Feedwater System (Loop C)
CA-204 Auxiliary Feedwater System (Loop D)
&.Inside Contain)
NC-201 Pressurizer Surge Line NC-202 Pressurizer Relief Valve Lines j
.NC-203 Pressurizer. Spray Line NC-205 Reactor Coolant System - RTD Loop A NC-206 Reactor Coolant System RTD Loop B-NC-207 Reactor Coolant System - RTD Loop' C NC-208 Reactor. Coolant System - RTD Loop D NC-210 Reactor Coolant System - Loop A Drain-NC-211
. Reactor Coolant System.- Loop B Drain t
NOMENCLATURE APPLICABLE TO CATAWBA NUCLEAR STATION PIPING - PAGE TWO SYSTEM /LINE TITLE NC-212 Reactor Coolant System - Loop C Drain & Letdown NC-213 Reactor Coolant System - Loop D Drain NC-225 Reactor Coolant System - Charging Line NC-226 Reactor Coolant System - Alternate Charging Line ND-201 Residual Heat Removal - Supply to RHR Pump ND-202 Residual Heat Removal - Supply to RHR Pump NI(Outside Safety Injection System Contain)
NI-201 Safety Injection - Upper Head Injection NI-202 Safety Injection - Upper Head Injection NI-203 Safety Injection - Upper Head Injection Crossover NI-204 Safety Injection - Accumulator Loop A NI-205 Safety Injection - Accumulator Loop B NI-206 Safety Injection - Accumulator Loop C NI-207 Safety Injection - Accumulator Loop D NI-212 Safety Injection - Safety Injection Pump Inlet NI-220 Safety Injection - Charging Pump Injection NI-221 Safety Injection - Charging Pump Injection NV(Outside Chemical and Volume Control System Contain)
NV-201 Chemical & Volume Control Charging Line NV-203 Chemical & Volume Control - Letdown Line NV-209 Chemical & Volume Control - RCP A seal water NV-210 Chemical & Volume Control - RCP B seal water NV-211 Chemical & Volume Control - RCP C seal water NV-212 Chemical & Volume Control - RCP D seal water SA(Outside Main Steam to Auxiliary Equ'; ment Contain)
SM(Outside &
Main Steam System Inside Contain)
SV(Outside Main Steam Vent to Atmosphere Contain)
8 1
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