ML20205D751
| ML20205D751 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 10/10/1985 |
| From: | Sorensen G WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | Adensam E Office of Nuclear Reactor Regulation |
| References | |
| GO1-85-0191, GO1-85-191, NUDOCS 8510170082 | |
| Download: ML20205D751 (21) | |
Text
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Washington Public Power Supply System P.O. Box 968 3000 GeorgeWashingtonWay Richland, Washington 99352 (509)372 5000 Docket No. 50-460 REGISTERED MAIL G01-85-0191 October 10, 1985 Director of Nuclear Reactor Regulation Attention: Elinor G. Adensam, Chief Licensing Branen No. 4 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Subject:
NUCLEAR PROJECT NO. 1 ELIMINATION OF ARBITRARY INTERMEDIATE PIPE BREAKS Washington Public Power Supply System has been closely following the recent activitics of the Nuclear Regulatory Commisiton (NRC) staff and the nuclear industry related to the modification of certain aspects of current design basis pipe break in high energy piping systems.
Modification of the current NRC requirements for postulating break locations by eliminat-ing from design consideration those intermediate break locations, generally referred to as arbitrary intermediate pipe breaks (AIBS), offers considerable benefits since pipe whip restraints, jet shields and other provisions currently incorporat~ed in plant designs to mitigate the effects of these postulated breaks would no longer be required.
NRC currently requires that pipe breaks be postulated at terminal ends and at intermediate locations where stresses or cumulative usage factors exceed specified limits.
If two intermediate locations cannot be determined based on the above criteria, i.e.,
stresses and cumulative usage factors are below specified limits, additional breaks must be arbitrarily postulated at the highest intermediate stress locations so that a minimum of at least two separate intermediate pipe break locations are postulated.
The Supply System believes that current knowledge and experience supports a conclusion that designing for the arbitrary intermediate pipe breaks is not technically justified and that this requirement should be deleted.
Arbitrary intermediate pipe breaks are often postulated at locations where stresses are well below the ASME Code allowables and within a few percent of the stress levels at other points in the same system.
This results in cumbersome, corkplicated and expensive protective features being provided for specific break locations in the piping system that may not enhance overall plant safety.
Rn*
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Docket No. 50-460 ELIMINATION OF ARBITRARY INTERMEDIATE PIPE BREAKS Page 2 Application of the alternate pipe break criteria described below would not in any way compromise overall plant safety or structural design as the remaining postulated breaks and restraints would still provide an adequate level of protection in all areas containing high energy lines.
By way of this letter, we request NRC approval of the application of the following alternative pipe break criteria (excluding the RCS primary loop) for WNP-1.
e Piping systems will be designed to accommodate pipe breaks at terminal ends and at locations where (a) the stress criterion of Branch Technical Position (BTP) MEB 3-1 for Class 1, 2, and 3 piping is exceeded, or (b) the usage factor criterion of BTP MEB 3-1 for Class 1 piping is exceeded.
No arbitrary intermediate pipe breaks will be postulated where the above stress or usage factor criteria are not exceeded, e The dynamic effects (pipe whip, jet impingement and compartment pressurization loads) associated with previously postulated arbitrary intermediate pipe breaks will be excluded from the plant design basis.
e Pipe whip restraints and jet shields a:sociated with previously postulated arbitrary intermediate pipe breaks will be eliminated.
e Application of the above criteria will not affect environmental qualification of equipme nt.
Appropriate non-mechanistic breaks will continue to be postulated for equipment qualification.
In support of this request the Supply System is providing the following additional information:
e Attachment A describes benefits which will be derived from elimination of the need to postulate the arbitrary intermediate pipe breaks listed in Attachment F.
Attachment B provides the technical justification for the revised e
criteria, supporting the elimination of the need to consider arbitrary intermediate pipe breaks.
e Attachment C describes provisions in the WNP-1 design which minimize stress corrosion cracking in high energy lines.
e Attachment D describes provisions in the WNP-1 design which minimize steam / water hammer effects.
Attachment E describes provisions for minimizing the effects of e
thermal and vibration induced piping fatigue.
Docket No. 50-460 ELIMINATION OF ARBITRARY INTERMEDIATE PIPE BREAKS Page 3 o Attachment F provides a summary of arbitrary intermediate pipe breaks that would be eliminated by application of the proposed alternative break criteria.
Piping and system design is an iterative process and postulated break locations may change as the system design and pipe stress analyses are finalized.
In view of anticipated changes in pipe break design, the Supply System proposes to eliminate the arbitrary intermediate pipe breaks currently identified in Attachment F as well as the need to postulate new arbitrary intermediate break locations, in the systems identified therein, during the ongoing design process.
In order to achieve the maximum benefit in terms of reduced occupational radiation exposure and improved operation and maintenance access, as well as to avoid as much of the design, purchase and installation costs as possible, we look forward to expeditious action on this request.
Included is the $150.00 application fee required by 10CFR170.12.
Y i c : r ; O..-
G. C. Sorensen, Manager Regulatory Programs (280)
GCS:LC0:11w cc:
N. D. Amaria, UE8C (898)
T. Kenyon, NRC (116)
N. Reynolds, BLCPR E. Revell, BPA (399)
NRC Doc. Control Desk G.C. Schieck, B&W FDCC (899)
Dochet No. 50-460 ATTACHMENT A BENEFIT
SUMMARY
}
FOR ELIMINATION OF ARBITRARY INTERMEDIATE BREAKS ( AIB)
I i
CATEGORY BENEFIT t
1.
Improved access for inspection Improvement in quality of in-maintenance and operation service inspection. Reduction in downtime during plant outages.
I Improved capability to recover from unusual plant conditions, eg.,
decontamination following radio-active spills, fires, etc.
1 I
Reduction of approximately 80 to 90 man-rem in radiation exposure over the 40-year plant life, a
2.
Engineering, design fabrication Estimated Saving i
f installation cost associated with
$4.1 million (1985 dollars) i~
elimination of lod restraints and jet shields.
(Included are 54 as I
designed and 54 anticipated addi-tional restraints and jet shields that would be required if aid relief was not provided.)
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Docket No. 50-460 3.
Reduction in piping heat loss at Special insulation trinning eliminated restraints required at the restraints can be eliminated. Assists in controll-ing the normal environmental temperatures improving system operational efficiency.
4.
Improvement in overall plant Elimination of potential for
, NUREG-CR-2136) unintentional restricted piping
(
safety movement.
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c Docket No. 50-460 ATTACHMENT 8 TECHNICAL JUSTIFICATION FOR ELIMINATION OF ARBITRARY INTERMEDIATE PIPE BREAKS (AIB)
The following items orovide technical justification for the elimination of
.aroitrary intermediate pipe creaks and the associated pipe whip restraints.
1.
Pipe rupture is recognized in Branch Technical Position EB 3-1 as being a " rare event which may only occur under unanticipated conditions."
2.
The combined operating nistory of commercial nuclear plants, both domestic and foreign, has not shown the need to provide protection from the dynamic effects of arbitrary intermediate pipe creaks.
3.
Intermediate pipe breaks are postulated to occur at locations where stresses exceed 80% of ASE Code allowables for Class 1, 2 and 3 or where the cumulative usage factor for Class 1 piping exceeds 10% of the Code allowable. If less than two intermediate locations exceed these criteria, then additional aroitrary intermediate pipe breaks (based on highest stress locations from the piping stress analysis) are postulated so that a minimum of at least two separate intermediate break locations are postulated. Therefore, the aroitrary break locations all exhioit stresses and usage factors below these conservative tnresholds.
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Docket No. 50-460 ATTACHMENT S (CONT'0) 4.
Consideration of these two arbitrary intermediate pipe breaks is particularly difficult because the location of high stress points may move several times as tne seismic design and analysis of structures and piping develops. The revised criteria of not having to relocate the intermediate pipe break contained in SRP 3.6.2 (tOREG 0800) coes not provide sufficient relief such that the relocated AIB need not be analysed.
5.
Arbitrary intermediate pipe breaks are only postulated to provide additional conservatism in the design. There is no technical justification for postulating these breaks.
6.
Operating procedures and piping and system designs minimize the possibility of stress corrosion cracking, thermal and vibration induced fatigue and steam / water hammer in lines where aroitrary intermediate pipe breaks are currently postulated. Detailed descriptions of the design provisions for these phenomena are provided in attachments C, D and E respectively.
7.
Generally, welded attachments are not located in close proximity to the arbitrary intermediate pipe breaks. Arbitrary intermediate break locations in proximity to welded piping attachments, other than shear lugs, will not be eliminated. Consequently, there are no local bending stresses resulting from these attachments which could affect the stress levels at the break locations being deleted.
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Docket No. 50-460 ATTACHMENT B (CONT'D) 8.
Tne pipe whip restraints necessary for postulated pipe breaks which are not ceing eliminated provide an adequate level of protection for those breaks which have some reasonaole probability of occurence.
9.
Elimination of pipe whip restraints associated with the arbitrary intermediate pipe breaks will faciliate in-service inspection, reduce neat losses from the restrained piping, and prevent the introduction of stress from the unintended restraint of piping due to thermal growth and seismic motion.
(See " Effects of Postulated Event Devices on Normal Operation of Piping Systems in Nuclear Power Plants," NUREG/CR-2136, Teledyne Engineering Services, 1981.)
10.
The elimination of arbitrary intermediate breaks will not downgrade the environmental qualification levels for Class 1E equipment.
The break postulation for environmental effects is performed independently of break postulation for pipe whip and jet impingement.
- 11. Pipe dreak Task uroup of NRC Piping Review Committee, based on their extensive survey and study, has recommended elimination of Aroitrary Intermediate dreaks as a design basis requirement (Recommendation 5 of NUREG-1061 Volume 3, " Evaluation of Potential Pipe Breaks", dated November 1984).
It is concluded that the elimination of arbitrary intermediate pipe breaks is technically justified, based on the reasons stated above.
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Docket No. 50-460 ATTACHMENT C PRuVISIONS FOR MINIMIZING STRESS CORROSION CRACKING IN HIGH ENERGY LINES The following review, encompassing a literature survey, service experience, and fabrication / installation and operational requirements, provides convincing proof that stress corrosion cracking of stainless steel and carbon steel in primary and secondary pressure boundary piping systems is an unlikely event for WNP-1.
6 Care 0N STEtL Caroon steel piping materials are considered immune to stress corrosion cracking basically Decause their overall corrosion rate in aqueous environments typical of PWR system service is high compared to the stainless steels and copper base alloys. A metal or alloy will be subject to the highly localized form of attack known as stress corrosion cracking only if the overall corrosion rate in the suoject environment is low.
i The overall corrosion rate of carbon steel piping materials will be controlled at WNP-1 by careful monitoring of the water chentistry, with the intent to minimize the oxygen level, which is the primary contributor to overall corrosion. During plant operation the secondary-side water chemistry will be carefully monitored to assure compliance with the specification requirements shown in Table 1.
In addition, the oxygen and pH control agents will be carefully selected to assure compatibility with the piping materials.
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Docket No. 50-460 STAINLESS STEEL The following review focuses primarily on austenitic stainless steel (type 304).
In order for stress corrosion cracking to occur, three conditions involving stress, terrperature and corrosive environment must occur simultaneously. Of these three, the corrosive environment is considered to be the key parameter since it is the most difficult to control. Stress and terrperature are relatively fixed parameters although residual stresses from welding or operation may produce undesiracle stress levels.
Thus, to prevent stress corrosion cracking of the pressure boundary at WNP-1, consideraD1e effort is expended to avoid susceptible corrosive environments. This is accomplished by (1) imposing strict material and fabrication / installation requirements to avoid the presence uf critical levels of contaminants known to cause stress corrosion cracking of stainless steel such as enlorides, fluorides, various forms of sulphur, caustics, and oxygen; and (2) planned rigid control of water chemistry.
Numerous measures are taken to prevent the introduction of contaminants such as (1) assuring that materials coming in contact with stainless steel during fabrication or operation do not contain harmful levels of impurities such as in crayons, insulation, gaskets, and lubricants, (2) cleaning prior to heat treatment and welding, (3) final cleaning and capping prior to shipment to site, (4) use of high quality water (low chloride, fluorides, and controlled pH) for pre-operational flusning and testing, and (5) final cleaning of outside surfaces followed by chloride and fluoride checks prior to startup testing.
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e Docket No. 50-460 In addition, other requirements are imposed on material suopliers and component manufacturers to assure the use of optimum practices to control caroide precipitation (sensitization) and cold work which are known to promote stress corrosion cracking. Precise heat treatment practices are required to be used to promote optimum metallurgical structures for resisting stress corrosion cracking. Procedures are reviewed to assure the use of effective but safe cleaning solutions. Cold working (bending) after solution annealing is prohibited except for small diameter pipe. Heavy sensitization is avoided by prohibiting stress relieving after welding and control of heat input during welding. During plant operation, primary and secondary water chemistry will be carefully monitored to assure compliance with the specification requirements shown in Table 1.
Note in particular that oxygen levels will be maintained for the primary side by a comoination of hydrogen and hydrazine and for the secondary side by hydrazine additions.
At this time, no known stress corrosion piping failures have been reported (References 1,2,3, & 6) in PWR primary coolant systems.
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4 Docket No. 50-460 1
References:
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t 1.
S. H. Bush and R. L.. Dillon, Stress Corrosica in Nuclear Systems, Battelle Memorial Institute, Pacific Northwest Laboratories, March 8, 1973.
J 2.
A. D. Edmondson, et al., WNP-1 Intragranular Stress Corrosion Task Force Recort Supply System / United Engineers and Constructors, Inc., June 2, 1980.
3.
S..H. Bush, Stress Corrosion in Nuclear Systems, Battelle Memorial Institute, Pacific Northwest Laboratory, Septemoer 30, e
1975.
1 4.
NUREG-0691, Investigation and Evaluation of Cracking Incidents in i
l Ploing in Pressurized Water Reactors, US tRC, September 1980.
5.
Corrosion Engineering by Fontana and Greene, 1967.
l
_ 6.
Corrosion Data Survey by Norman Hamner, National Association of I
' Corrosion Engineers (NACE), Metals Section, Fifth Edition,1974.
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i Docket No. 50-460 l'
TAALE 1 WNP-1 NICLEAR PROICT I
4 WATER-DEMISTRY SPECIFICATIONS FOR LINES CONTAINIPC APalTRARY INTERWDIATE AREAKS l
Maulam operating Hydrogen Maximum Maxinn pH pH Control 02 Control i
SYSTEM i
ASME Pipe Tenp. Under Concentration Oxygen Chlorides /
(S 770F)
Agent Agent j
class Mat'l.
Normal Plt. (cc/kg H2O)
(ppm)
Fluorides i
Cond. (OF)
(ppm)
SS 569 15 - 40 0.1 0.1/0.1 4.6-8.5 Lith. Hydrox.
H2 &
(Lines to Letdowre Coolers)
(Note 1)
Hydrazine j
(Note 2)
I f
CORE FLOGOIr4 1&2 SS 305 0.1 0.1/0.1 6.0-8.0 Litn. Hydrox.
H2 &
Hydrazine (Note 2) 1 DECAY HEAT REMOVAL 1&2 SS 305 15 - 40 0.1 0.1/0.1 4.6-8.5 Lith. Hydrox.
H2 &
(Note 1)
Hydrazine (Note 2) t MAKEUP A?O PURIFICATIG4 1
SS 569 15 - 40 0.1 0.1/0.1 4.6-8,5 Lith. Hydrox.
H2 &
(HICH PRESSWE DOECTION)
(Note 1)
Hydrazine (TO LETOOWN HT. EXCH., &
(Note 2)
N.U. TO RC SYSTEM) i MAKEUP APO RJRIFICATION 1&2 SS 130 15 - 40' O.1 0.1/0.1 4.6-8.5 Lith. Hydrox.
H2 &
(HIOi PRESSWE IldCTION)
(Note 1)
Hydrazine j
(FROM LETOOWN HT. EXCH.)
(Note 2)
- KEtP Aro PURIFICATION 1&2 55 138 15 - 40 0.1 0.1/0.1 4.6-8.5 Lith. Hydrox.
H2 &
(HICH FHESSURE ItOECTION)
(Note 1)
Hydrazine s.
(M.U. PW DISOMRCE) i MAIN STEAM 2
CS 625 0.007 8.5-9.3 Ammonia Hydrazine MAlte FEEDwATER 2
CS 462 0.007 8.5-9.3 Amonia Hydrazine l
L AUXILIARY FEEDnATER 2
SS 625 0.1 8.5-9.3 Amonia Hydrazine (Note 4) i j
HIOt PRESS. STEAM TRAP RETlRN 2
CS 625 0.007 8.5-9.3 Amonia Hydrazine i
STEAM EtERATOR TEAT-UP CIRCULATION 2
CS 625 0.007 8.5-9.3 Ammonia Hydrazine j
(Note 3)
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-_ _, _.__ _,_. _. - - -. _. _... _,., - ~,,.. _., _. _,
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Docket NO. 50-460 M)TES: (1) Should be less than 0.005 ppm with proper H2 specification and reactor critical.
(2)
H-> ls used bbove 400F', and hydrazine below 400F, (3) pA is 9.5-10.5 duriry wet lay-up.
(4) teormal is 7ppo maxinsa during long term usage.
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Docket No. 50-460 ATTACHWNT D PROVISIONS FOR MINIMIZING STEAM / WATER HAMMER EFFECTS WfP-1 Nuclear Project Piping Systems nave been designed in accordance with good engineering practice.
In general, steam systems are sloped and provided with adeouate drainage as required to prevent condensate buildup. Water systems were routed to minimize void formation.
In addition, the applicable portions of the affected piping systems have been thoroughly analyzed for aonormal operating conditions.
For safety class systems designed in accordance with AdME B&PV Code Section III, a systematic review is bein;; initiated to determine the thermal and hydraulic transient forces generated in response to an extensive matrix of initiating events, such as valve closure, pump trip and turoine trip.
The forces thus generated will be included in the final piping stress analysis and considered in the design of the pipe supports.
Systems that are designed and supplied.by Babcock & Wilcox (B&W), in general, are not susceptible to water hammer. The discussions below for the Reactor Coolant, Core Flooding, Make-up and Purification (High Pressure Injection),
and Decay Heat Removal Systems illustrate how their designs preclude water hammer. Preoperational testing, and operating experience in similar plants have verified that the d&W design is not susceptible to water hammer transients.
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i Docket No. 50-460 Lines which have any potential for water hammer are being designed to minimize 1
or preclude its effects. Water hammer mitigation effort for each system involved with AIB's are discussed below.
1.
There is a low potential for water hammer in the Reactor Coolant System because it is designed to preclude steam void formation. However, excessive depressurization of the system, which initiates safety i
injection could potentially result in water nammer.
If any such problems are encountered during startup testing, they will De corrected by 3
i modifying operating procedures.
4 i
j 2.
Core-Flooding System The Core Flooding System piping is basically water solid and as such, there is a low probability of water hammer in this system.
3.
Maketo and Purification (High Pressure Injection) System i
J The Makeup and Purification (High Pressure Injection) System is normally i
water solid.
In low temperature lines, water hammer is not expected due i
to absence of steam voids. The hign temperature lines, being water solid 1
by design, will minimize the probability of water hammer.
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Docket No. 50-460 t
4.
Normally, the Decay Heat Removal System is water solid.
It is also l
designed to preclude steam void formation.
Hence, the prouability of i
i water hammer is low.
l 5.
Main Steam System The mainsteam lines are sloped and provided with conservatively sized drip legs at suitable points for collection of condensate.
This precludes steam hammer during normal operation.
Steam hammer due to rapid condensation during an OTSG overfill situation has a very low proDacility of occurrence, because the moisture in the steam is at saturation and no potential for rapid condensation exists.
I 6.
Steam Generator Warm-up and High Pressure Steam Trap Return Systems The lines involving Aroitrary Intermediate Breaks consist of drains from tne drain pots and the headers that lead to the steam traps.
These lines are conservatively sized compared to the rate of flow, hence velocities i
are quite low.
Therefore, the intermittent operation of the trap is not l
expected to produce water hammer.
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Docket No. 50-460 7.
Main Feedwater System To eliminate possible water hammer in the main feedwater system, the volume of feedwater piping external to the steam generator which could pocket stean is minimized by using tne shortest possible horizontal run of inlet piping to tne steam generator.
The horizontal run of the main feedwater piping at the steam generator nozzle is about three (3) feet.
Tnis arrangement precludes the possibility of large amounts of steam being drawn into the feedwater piping during abnormal events, thereby preventing water hammer caused by rapid steam condensation.
In addition the fastest closing components of the system, i.e., the check valves, are located as close to the steam generators as possible, thereoy minimizing water hammer effects.
I 8.
Auxiliary Feeddater System Precautions have been taken in the routing of the auxiliary feedwater lines on WNP-1 to minimize the piping external to the Steam Genmerator which could be filled with steam, thus reducing the potential for water hammer.
Recently, damage to top mountad internal auxiliary feed rings on some operating B&W Steam Generators has been found. This was apparently the result of steam-water interaction within the header and not water hammer within the auxiliary feedwater supply piping. To preclude recurrence of this problem on WNP-1, the WNP-1 Steam Generator will be supplied with auxiliary feedwater through an external header similar in design to the l
main feedwater supply arrangement in use on operating B&W plants.
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Docket No. 50-460 ATTACHMENT E PROVISIONS FOR MINIMIZING THE EFFECTS OF f
THERMAL AND VIERATION INDUCED PIPING FATIGUE I. GENERAL FATIGUE DESIGN CONSIDERATIONS Class I piping fatigue considerations are addressed by the use of a cumulative usage factor (CUF).
In order to ensure that piping does not 1
fail due to fatigue, the ASME Code limits the CUF to a maximum of 1.0.
Aroltrary intermediate break locations are limited to a CUF below 0.1.
Because the CUF is below 0.1, fatigue will not be a concern at these locations.
For Class 2 and 3 lines, fatigue is considered in the ASME Code allowable stress Iange check for thermal expansion stresses.
This stress is included in the total stress value used to determine postulated break locations. All arbitrary break locations are limited to stresses less than 80% of the code allowables.
If the number of thermal cycles is expected to be greater than 7,000, then the allowaole stresses are further reduced by an amount dependent on the numoer.of cycles.
I An appropriately colservative condition already exists for postulating break locations which exceed tne OUF (Class 1) and stress limitations above. Postulation of additional arbitrary intermediate pipe breaks with CUF and stress values witnin these limits does not significantly add to l
the protection against fatigue failure.
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Docket No. 50-460 II. THERMAL-DESIGN CONSIDERATION In WNP-1 design, the auxiliary feedwater system piping is not connected to the main feedwater system and the system generators have separate nozzles for auxiliary feedwater and main feedwater.
Tnermal fatigue, therefore, is minimized in the main feedwater piping by preventing the introduction of cold auxiliary feedwater.
During. start-up, the heat-up of the steam generator brings up the temperature of the feedwater downstream of tne isolation valves to a U
maximum of 425 F, while tne feedwater upstream of tne isolation valves 0
is brought up by auxiliary steam to a temperature of 390 F.
Thus a close temperature match is achieved before the isolation valves are opened. Although it is proposed that the aroitrary intermediate break locations in the inain feedwater system be eliminated, the terminal end break locations at the steam generator nozzles, wnich have experienced pipe cracks in PWRS, are unaffected.
WNP-1 is designed to mitigate the consequences of a circumferential pipe break at these locations and the ability to bring the plant to safe snutdown condition is maintained.
III. VIBRATION DESIGN CONSIDERATIONS Piping is designed and supported to minimize transient and steady state vibration. Piping system vibration testing will be performed as described in FSAR Section 3.9.2 to ensure that piping system vibration is witnin allowable levels E-2
-' Dodet No. 50-460 ATTACHNENT F
SUMMARY
OF ARBITRARY INTERFEDIATE PIPE BREAKS TO BE ELIMINATED Pipe Building No. of Breaks No. of Restraints Eliminat System Size Location Eliminated Designea Anticipatec w/o AIB Reli Auxiliary Feed Water 6"
CTMT 4*
4 (FWA)
Core Flood 14" CTMT 4
2 1
Decay Heat 12" CTMT 4
1 4
Removal (DHR)
7 (FWS) 16" CTMT 7
2 10 6"
MSFIA 2
3" CTMT 6
1 1/2" CTMT 4
Hign Pressure Steam 6"
CTMT 4
5 4
Trap Return (HPR) 3" CTMT 4
3" MSFIA 8
1 2
2 1/2" CTMT 8
1 1/2" M5FIA 11 2
Main Stream 28" CTMT 8
16 8
(MSS) 6" CTMT 2
4 6"
GS8 2
3 4
3" CTMT 4
2" CTNU 4
1 1/2" MSFIA 1
Make-Up & Purifi-4" GSB 7
2 2
cation (High Pressure 3"
CTMT 1
Injection) (MUS) 3" GS8 4
2 1
2 1/2" CTMT 16 4
2 1/2" GS8 2
1 1
2" GS8 10 1 1/2" CTMT 11 4
1 1 1/2" GSB 7
8 Reactor 3"
CTMT 1
2 Coolant (RCS)
Steam Generator 2 1/2" CTMT 1
Heat-Up Circulating (SGC)
TOTAL 152 54 54
- Estimateo - Evaluation not complete.
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