ML18024A114

a Reaction to Cracking of Austenitic Stainless Steel Piping in Boiling Water Reactors (Includes Susquehanna Ses Design Modifications)

ML18024A114
Person / Time
Site:	Susquehanna
Issue date:	01/24/2018
From:	Mead E M Pennsylvania Power & Light Co
To:	Office of Nuclear Reactor Regulation
References
Download: ML18024A114 (26)

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PENNSYLVANIA POWER 8c LIGHT CCMPANY A REACTION K)CRACKING OF AUSTENITIC STAINLESS S9ZEL PIPING IN BOILING WATER REACTORS (INCLUDES SUSQUEHANNA SES DESIGN MODIFICATIONS)

Earle M.Mead./Progect Engineering Manager Susquehanna SES Table of Contents l.Introduction 2.Problem Statement Safety Significance Primary Considerations 4.1 Environmental (Coolant Chemistry) 4.2 Stress 4.3 Material 5 Susquehanna SES Preventive Measures 5.1 Environment 5.1.1 Chemical Control 5.1.2 Mechanical Control 5.1.2.1 Control Rod Drive (CRD)Pump Suction Relocation 5.1.2.2 Mechanical Vacuum Deaeration 5.1.3 Operating Procedures 5.2 Stress 5.2.1'Design Stresses 5.2.2 Fabrication Stresses 5.2.2.1 Fit-up 5.2.2.2 Initial Fabrication (Shop)5'2.2.3 Welding-Induced.

Stress 5.2.2.3.1 Heat Input 5.2.2.3.2 Joint Design 5.2.2.3.3 Filler Metal 5.2.2.3.4 Cleanline ss Table of Contents Page 5.2.3'ethods of Stress Reduction 10 5.2.3.1 Solution Heat Treatment 5.2.3.2 Heat Sink Melding 5.2.3.3 Induction Heating Stress Improvement 10 5.3 Material 5~3-1 503~2 Core Spray System Reactor Recirculation System Discharge Gate Valve Bypass Line 13 13 5~3~3 5 3~.6.References Control Rod Drive Return Line Recirculation Riser pipes 13 15 Table and Illustration Table 1 Figure 1

~~'1.Introduction This report originally resulted from an informal telecon between representatives of PPM.and the NRC which occurred in February 1976.At that time proposed Susquehanna Steam Electric Station (SSES)design changes were briefly discussed and it was agreed that this additional information should.be submitted., Subsequent to the original issue of this report there have been numerous incidents attributed to Intergranular Stress Corrosion.

Cracking (IGSCC).Based.on these incidents, PPM, has modified our position on IGSCC, which included additional design changes at SSES..For the above reason and to provide the NRC with PAL's position on NUHEG-0313, this report has been revised and updated.

2.Problem Statement From September, 1974 to October, 1976 cracking had.been discovered.

in recirculation bypass loops, core spray lines and.control rod.drive return lines of 10 GE boiling water reactors.The affected units were: a)Core spray lines-Dresden 2, Fukushima 1, and.Tsuruga b)Recirculation bypass lines-Dresden 2, Quad Cities 1, Quad Cities 2, Peach Bottom 3, Millstone Point 1, Monticello 1, Fukushima 1, Hamaoka 1, Brunswick 2, Hatch 1, and.Pilgrim l.c)Contxol rod drive return lines-Tarapur PP8cL believed that sufficient Justification existed.to recognize the cause as being Intergranular Stress Corrosion Cracking.This report was originally issued on October 12, 1976 and.summarized PP8cL's concerns as well as modifications that were implemented for SSES.From October 1976 to the present there have been'numerous reports of cracking at other plants which has been attributed to IGSCC.Not only have the above lines been affected, but there are reports of cracking in head vent lines, recirculation riser lines, instrument lines, and also, main recirculation and feedwater lines.This report has been revised to reflect PPM's position on IGSCC and.the latest modifications which have been made at SSES.

3.Safety Significance PPM, considers the statement in NUREG-0313, which reads in part,"Although the probability is extremely low that these stress corrosion cracks will propagate far enough to create a significant safety hazard to the public, the presence of such cracks is undesirable" to be an accurate assessment of the safety significance.

Beyond, that PPM believes the significance of the failures to relate only to plant reliability.

Based, on previous BWR experience, it can be expected.that brittle cracking will not occur in austenitic Type 304 stainless steel and.that small leaks will be detected, v5.sually and/or by leak detection instrumentation if they occur inside the primary containment.

The importance of detectability has been recognized, and the SSES drywell leak detection system will comp~with NRC Regulatory Guide 1.45 (May 1973).The Inservice Inspection Program at SSES is in accordance with the requirements of the ASME Boiler and Pressure Vessel Code,Section XI,"Rules for Inservice Inspection of Nuclear Power Plant Components", 1974 edition, including addenda through Summer 1975, as modified.by Appendix III to Minter 1975 addenda,"Ultrasonic Examination Method.for Class 1 and Class 2 Piping Systems Made from Ferritic Steels", and INA-2232 of the, Summer 1976 Addenda, using 50+p of the reference level as criteria for investigating reflectors.

The ISI Program will be updated, as required/allowed by 10CFR50.55a(g).

It will also be augmented to comply with the recommendations of NUREG-0313.

The degree of augmentation will depend on the outcome of PPM,'s detailed evaluation of IGSCC at SSES.This evaluation will be completed.

by January 1, 1990 and.will address all applicable lines which contain reactor coolant.It will delineate all materials and any fabrication processes which will provide a comprehensive listing'of susceptible lines in the as installed condition.

The study will then provide an evaluation of all IGSCC countermeasures which are available for use at SSES.The conclusion of the study will be a recommendation of what"course of action" should be taken for each susceptible line>from which countermeasures should.be used to how the ISI Program should.be augmented..

4'r1mary Considerations 4.1 Environment (Coolant Chemi,stry)

The coolant chemistry in Boiling Water Reactors is established.

primarily to ensure compat1bility with materials used throughout the Nuclear Steam Supply System.Hence, neutral, high-purity water is used and.halogens are stringently limited..Limitations are also placed on the silica and.copper concentrations to prevent their deposition in the turbine.Dissolved.

oxygen concentration is not normally controlled by chemical addition or mechanical deaeration.

Without chemical or mechanical control the steady state level of dissolved.

oxygen at SSES during normal operation will be a maximum of 7.1 ppb.However, during low load or no load.cond1tions the oxygen concentration can approach saturation which, under conditions of standard temperature and pressure, is 8.0 ppm.Under the combined influences of sensitization and high tensile stress this oxygen level is more than sufficient to enable stress corrosion cracking of austenitic stainless steel to occur.Since, for constant values of sensitization and.stress, tim-to-failure 1s d.irectly proportional to dissolved.

oxygen concentration, low flow and stagnant lines are high+suspect.It would., therefore, be very beneficial to reduce the oxygen level to as low as possible dur1ng all phases of plant operat1on.

4.2 Stress The design stress levels for Nuclear Steam Supp+System (NSSS)piping are established.

within the constraints of the AQS Code as well as any loading restrictions which might be imposed.by the NSSS vendor.Stress levels sufficient to result in IGSCC generally are not'ust the result of ordinarily-applied engineering loads or stresses.Rather, they result from the combin d'ffects of all sources of stress and.strain: i.e., residuals, thermal, surface, service, etc.I It is very beneficial to min1mize the amount of stress acting on the sens1tized.

material.Therefore, the amount of residual and.applied.stresses should be reduced wherever possible.4.3 Naterial Boiling Water Reactor piping is fabricated from steels which are: corrosion resistant, tough, stable dim nsional+, sufficiently strong for anticipated loadings, resistant to radiat1on damage, resistant to both acidic and basic chemical attack, economical, and.anticipated.

to be available in the foreseeable future.The most prominent materials used.are austenitic stainless steels (Types 304 and, 316)and, carbon steel.

"l However, it has been determined that the standard grades of 304 and.316 stainless steel with high carbon contents of approximately

.06 to.08 percent are the grades which produce the most severe sensitization and.thus, the greatest susceptibility to IGSCC.The low carbon ((.03 percent carbon)grades of 304 and 316 stainless steel are demonstrating a much greater resistance to IGSCC.These and, other alternate materials are undergoing extensive development and, evaluation for reducing the probability of IGSCC.

I I i~

5.SSES Preventive Measures PP&L efforts have centered on reducing the probability of occurrence of IGSCC for SSES.Recommended modifications to the existing SSES design were evaluated, prior to implementation against the fallowing criteria: 1.Their potential for substantially reducing the probability that XGSCC will occur;2.Their potential for substantially reducing the time required.to detect cracks or leaks resulting from XGSCC;3.Their potential for creating other problems which would either be as bad.as, or worse than, the current problem.5.1 Environment Current'nalyses indicate that among the environment related.contributors to the current ZGSCC problem in BWR's, the dissolved oxygen concentration of the reactor coolant is probably the most significant.

Oxygen levels can be controlled, either chemically, mechanically, or operationally.

5.1.1 Chemical Control Since there is no universal inhibitor'for improving BMR water chemistry and.since there are limited.data concerning the use of oxygen scavengers and neutralizers in BNR's, their use in the primary coolant is not considered.

at this time.Future developments in this area, however, will be fallowed..

5.1.2 Mechanical Control Primary oxygen removal is accomplished in the condenser during normal operation.

During this phase the oxygen level is maintained.

at approximately

'$.0 ppb.During any combination of partial load.(startup, shutdown, hot standby and some abnormal events)the oxygen level tend.s to increase toward saturation (approximately 8.0 ppm).There have been two major design changes which PPRL has made to improve the water chemistry at SSES.5.1.2.1 Control Rod.Drive (CRD)Pump Suction Relocation PPM previously relacated the CRD pump suction from the condensate storage tank to the condensate makeup/reject line.The purpose of this line is to control primary.cycle water inventory by making up from or rejecting to the condensate storage tank.Under steady load conditions this line receives a constant discharge from the condenser due to primary cycle influent water sources such as the CRD system itself.Hence', locating the CRD pump suction on the makeup/regect line results in utilization of water with the lowest oxygen concentrations available (essentially water with feedwater quality)most of the time.At the time the above change was made, PAL believed that this change, in addition to other changes made, was adequate to sufficiently reduce the probability of IGSCC.Since that time, however, ppM, has found.that controlling the 02 concentration as an effective means of controlling IGSCC requires 02 control during all phases of plant operation, not gust normal operation.

For this reason, PHkL added, a mechanical deaeration system (see section 5.1.2.2), 5.1.2.2 Mechanical Vacuum Deaeration Based on more recent data, pp&L has taken the steps to add.a mechanical vacuum deaeration system at SSES.This 02 control system will operate during all phases of plant operation except normal operation.

It will maintain the.02 concentration in the primary coolant to less than 250 ppb.This recent change was made in an effort to reduce the corrosive nature of the primary fluid as low as possible.It is PP8cL's preference to have the 02 control system installed and.operating prior to plant startup.$4 143 Operating Procedures Lines connected to the reactor vessel which are normally stagnant or experience low flow conditions may accumulate dissolved oxygen concentrations which are high relative to general reactor water.Startup, Shutdown, Hot Standby and.abnormal events are of particular significance.

To the extent practical, procedures will be developed which will minimize dissolved oxygen concentrations in stagnant or low flow lines and/or reduce the total time of stagnant conditions.

The water quality sampling system has been upgraded in order to alert the plant operators of adverse water conditions conducive to IGSCC.It includes continuous monitoring of feedwater and reactor water for dissolved pxygen concentration, conductivity and.pH and.continuous monitoring of the CRD system water for dissolved oxygen concentration and.conductivity." The sampling system, automatically alarms when water quality conditions considered to be adverse to BWR operation are reached.

Procedures will be developed.

which will minim1ze excessive oxygen concentrations in the primary cycle and, to enable plant operators to take immediate action to protect the plant from prolonged operation with adverse water quality cond.itions.

5.2 Stress Stress alone is not particularly s1gnificant so far as the overall corrosion of metals is concerned.

When combined with a corrosive env1ronment, however, the appl1cation of sufficient tensile stress in susceptible materials can lead.to stress corrosion cracking.A reduction 1n total tensile stress in lines considered.

susceptible to IGSCC could.therefore be considered.

beneficial.

In an attempt to reduce overall tensile stress, piping located inside containment and.connected.

directly to either the Reactor Pressure Vessel (RPV)or the Reactor Recirculation System was reviewed, to determine if design and/or fabricating contributions to tensile stress levels could be reduced..5.2.1 Design Stresses The combined effects of service, internal pressure, deadweight and.thermal stresses were reviewed and.all piping systems were confirmed to have layouts which limited.their contributions to total stress to a level considered to be as low as reasonably ach1evable.

Recognizing the particular susceptibility of the Core Spray System infection lines, alternate routings were chosen in order that the thermal stress component of total stress is red.uced by 25$.522 Fabrication Stresses The difficulty in assuring compliance under all shop and.field fabricating conditions, limits the effectiveness of procedures which might be developed.

in this area.However, s1nce stress levels in non-stress relieved.austenitic stainless steel piping can equal or exceed.yield., any procedures which might significantly reduce fit-up, initial fabrication, or welding-induced.stresses cannot be overlooked..

5.2.2.1 Fit-up Code tolerance for alignment are ccmplied with to assure minimum stress from misalignm nt and minimum degradat1on of fatigue resistance.

5.2.2.2 Initial.Fabrication (shop)The pipe material is purchased.

1n the solution annealed.condition.

Normally, spool p1eces are not solution annealed.due to the difficulty of maintaining desired dimensions.

5.2.2.3 Melding-Induced Stress PAL recognizes that when austenitic sta1nless steels are welded, some level of residual stress and.sens1tization is present.A compromise between heat input control and the resulting cooling rate must be achieved in order that acceptable levels of res1dual stress and.sensit1zation can be achieved without sacrificing good penetration and fusion.Unfortunately, precise quantitative values of heat input, cooling rate, etc., which wi.ll insure consistently good quality welds resistant to IGSCC are not available.

Therefore, PPM relies on past industrial exper1ence for guidance.PP8L has adopted the following measures for field welding the applicable 11nes.These measures help reduce welding-induced stresses, w1thout creating additional problems, to a level as low as can be expected using normal welding practices.

5.2.2.3.1 Heat Input No preheat (in excess of the acceptable working range of 60 F to 150 F).b.C~dO Interpass temperature limited to 350 F.Block welding prohibited.

Electrode size 11mited to 5/32" h1ax.for SMAM and l/8" Max.for GTAM which effectively limits the heat input.5.2.2.3.2 Joint Design a.The root is made with GTAM ut111'z1ng hand fed filler wire or a consumable 1nsert to insure complete penetration and good fusion.b.The extended-land)oint design has no inherent problems with lack of penetration, lack of fusion or excessive residual stress.C~An inert gas purge 1s used prior to weld.ing and inert gas backing is used.during the weld1ng of the first passes to insure a good.root contour minimizing

'the occurrence of any crevices which might lead to corrosion problems.A smooth finish contour is specified (1.e., no excessive undercut, excessive reinforcement, coarse r1pples, etc.)to reduce the occurrence of"notches" which can detrimental+

affect fatigue strength or corrosion resistance.

10 5.2.2.3.3 Filler Metal To minim1ze microfissuring and, sensitization problems, 308L filler metal or 309 and.309L filler metal is specified.

w1th minimum delta ferrite contents of 8 percent and.5 percent respectively.

5.2.2.3.4 Cleanliness To prevent contamination of the joint: a.Grease, oil and other contaminants are removed.from the joint and.the f1ller metal prior to making the weld.b.Only marking crayons, chalk, 1nk and temperature indicating crayons which are certified.

to be low in halogen and, sulfur content are used.c.Only cleaning solvents wh1ch are not harmful to austenitic stainless steel are used,.d..Stainless steel wire brushes are used,.e.Grinding wheels used on other materials are not used.on stainless steel.f.Grinding wheels are not used on I.D.pipe surfaces.If cleaning is necessary, flapper wheels shall be used.The existing methods and procedures for Quality Control/Quality Assurance of the above are adequate to insure that these provisions are followed and.that the results will be consistent w1th what was specified..

Restrictions consistent with those above apply equally to shop and.field subcontract welding.5.2.3 Methods of Stress Reduction Due to continued reports of IGSCC since PPM first form d.a position on the subject, PP&L has steppecl up its monitoring of the problem and its possible countermeasures.

Those countermeasures that are being investigated.

/evaluated.

which deal with stress reduction are Solution Heat Treatment (SHT), Heat Sink Welding (HSW), and.Induct1on Heating Stress Improvement (IHSI).5.2.3.1 Solution Heat Treatment This IGSCC countermeasure stress relieves shop welcls.before the sections of pipe are shipped.to the field for installation.

This method.can only be used on shop weld.s, and.then, only when the pipes have not been installed..

For SSES most of the target lines for which no other countermeasures (i.e., material changes)have been taken, have already been installed..

However, the shop welds on the recirculation get pump r1sers underwent SHT because PPM believed these lines to be extremely susceptible to IGSCC.The f1eld.weld ends of the risers were also corrosion resistant clad..This process will be d.iscussed in a later section.5.2.3.2 Heat S1nk Welding This IGSCC countermeasure requires cooling the ID of the pipe with cooling water after the root pass of the weld has been completed.

This process causes-the resultant residual stresses on the ID of the pipe to be compressive rather than tensile as would.be found.with normal welding practices.

Resultant compressive stresses prevent XGSCC, This countermeasure has not been used at SSES.However, Bechtel is presently performing a feasibility evaluation on the use of HSW for selected.lines at SSES.5.2.3.3.Induction Heating Stress Xmprovement This IGSCC countermeasure is used.after the field welds have been completed..

The process involves heating the O.D.of the pipe with an induction coil while cooling the I.D.with cooling water.This process causes the resultant stresses on the I.D.of the pipe to be compressive rather than tensile, thus preventing IGSCC.This process has not been used at SSES.Presently GE 1s in the process of developing/qualifying this procedure for use on BWR's in the United States (Reported.ly the Japanese have used.this procedure successfully on their nuclear~power plants,).PP8cL intend,s to track the progress of GE and.use the procedure on susceptible lines if and, when the procedure is determined.

to be feasible.5.3 hfaterial Due to the restrictions of coolant chemistry and total tens1le stress levels, the use of substitute materials which are less susceptible to XGSCC were considered..

The lack of significant operating data for materials other than carbon steel or Type 304 and.Type 316 stainless steels limits the options, however.Dur1ng the design of the plant, attention was given to minimizing problems related to gross corrosion.

Stainless steel was chosen for Core Spray lines inside containment, condenser and.'feedwater heater tubes, and.ASSAM A155 Grade KC 70, Class 1 feedwater pipe.PP8cL is, therefore, unwilling to use carbon steel as an IGSCC f1x.

The principal drawback to the continued use of Type 304 and.Type 316 stainless steel for NSSS piping is their susceptibility to IGSCC.PP8cL believes this susceptib1lity is related to sensitization which occurs adjacent to a weld in the heat-affected zone.Sensitization is a temperature dependent metallurgical phenomenon which results in the formation of chromium carbide at grain boundaries located in the heat-affected.zone of a weld.Therefore, 1t was logical to 11mit the carbon content of susceptible lines.Originally, the history of the IGSCC problem formed the basis of the assumption that the comb1nation of residual stress and sensitization was insufficient to result in a high probability of failure due to IGSCC for lines in the reactor coolant pressure boundary larger than 12" diameter NPS.The focus for possible material substitution was, therefore, on those lines wh1ch were less than or equal to 12" diameter NPS.Any such substitutions would apply equally to pipe fittings.However', valves and.containment penetration flued, heads are not included as they are of sufficient mass to be substantially less susceptible to IGSCC.~Based on the above assumption, Type 304L stainless steel was used for all stainless piping within the reactor coolant pressure boundary which is 4" diameter NPS or smaller with a supplemental requirement of 0.030 percent maximum carbon (with the exception of the Recirculat1on System Discharge Gate Valve'ypass Line).Sta1nless piping located within the reactor coolant pressure boundary which is greater than 4" but less than 12" diameter NPS will be Type 304 stainless steel with a maximum carbon content of 0.030 percent.1hese two materials are virtually identical metallurgically but PPEcL is unwill1ng to sacrifice the mechanical properties of 304 for certain piping systems.Table l ident1fies piping for which a change in material was justified, the material previous+specified, and the replacement material chosen to mitigate the probability of IGSCC.Since the time the above material changes were made, additional inc1dents of IGSCC have shown that the a ssumption that only those lines which are 12" d1ameter NPS or less are susceptible to IGSCC is incorrect.

Therefore, all lines must be considered when attempting to eliminate or reduce IGSCC.In addition to simple mater1al replacement, the other"mater1al" related IGSCC countermeasure which can be evaluated is Corrosion Res1stant Cladding (CRC).This countermeasure combines cladding the field welded pipe ends with a highly corrosion resistant metal with solution annealing which effect1vely provides a corrosion resistant barrier between the heat affected base metal and, the oxygenated reactor coolant (corrosive fluid.).

13 As discussed.

in Section 5.3.4 of th1s report, PPM, utilized CRC on the recirculation get pump risers.Due to their particular significance with regard to the current problem, the Core Spray, Reactor Recirculation System Discharge Gate Valve Bypass Line, Control Rod Drive Return Line, and Recirculation R1ser Pipes are discussed below.5.3.1 Core Spray System That portion of the core spray system which is located within the primary containm nt will, for Susquehanna SES, be ma'de from 12" diameter NPS, Type 304 Stainless Steel Pipe with supplemental maximum carbon lim1tation of 0.030 percent.pipe and fittings will be handled similarly.

532 Reactor Recirculation Syst m Discharge Gate Valve Bypass Zine From an operational standpoint pp&L does not wish to delete this lin.It is considered important from the following standpoints; l.It provides a means of preheating an idle recirculation loop.2.It reduces thermal shock seen by the components of an 1dle loop+/3.It provides pressure equalization.

on both sides of the discharge gate valve to assure proper venting and closure of the valve.It eliminates cutting and wire drawing of the discharge gate valve seat.The 4" diameter NPS line will be fabricated.

from Type 304 stainless steel with a supplemental maximum carbon limitation of 0.030 percent.This material choice results from the des1re to limit the probability of IGSCC while retaining the mechanical properties of and the existing stress analysis for Type 304 stainless steel.Pipe and fittings will be handled similarly.

5 3 3 Control Rod Drive Return Line This 3" diameter line was chang'ed from 304SS to 304L SS which, 1t was believed., would.solve the problem of IGSCC.Subsequent to this change there were numerous reported inc1dents of cracking in the CRD Return Line nozzle.These incidents of cracking were attributed to excessive thermal gradients across the nozzle rather than IGSCC.General Electric's recommendation was to delete the return line and.make other changes to the system to mainta1n the system design function.PPEcL concurred..with GE's recommendation and.deleted.the return lin.This action (1)eliminated the problem of nozzle cracking due to thermal grad.1ents and.(2)eliminated the possibility of pip cracking due to IGSCC.5.3 4 Rec1rculation Riser Pipes The 10'ecirculation riser p1pes leading from the recirculation header to the get pumps have recently experienced IGSCC at other operating plants.The cracks have been formed.in the heat affected.zones of the thermal sleeve to safe end.attachment weld.s.The pipes have been fabricated from 304 sta1nless steel.The safe end.s are Imonel 600, and.the nozzles are carbon steel clad.with stainless steel.PP8:L has made the following changes which will minimize the possibility of these pipes cracking due to IGSCC of Susquehanna SES.1.The pipe to safe end and pipe to tee welds will have their ID's clad with 308L weld, material in the heat affected.area prior to welding.2.The riser pipes will be solution heat treated.to eliminate residual stresses from the elbow to pipe shop weld.s, and the CRC process.3.These welds will then be field welded.using 308L welding rods.This process will prevent 304 stainless steel in the heat affected area from coming in contact with the process fluid.

15 6.Re ference s Source GE NRC AS%1 Document NEDO-21000 NUHEG 75/067"Stress Corrosion Cracking of Metals-A State of the Art" Date Ju~, 1975 October, 1975 October, 1971 NATO"The Theory of Stress Corrosion Cracking in Alloys" October, 1971"Fundamental Aspects of Stress Corrosion Cracking" September, 1967 United States Senate Joint Hearing Concerning"Nuclear Regulatory Commission Action Requiring Safety Inspections Which Resulted in Shutdown of Certain Nuclear Power Plants" February, 1975 16 SUSQUEHANNA STEAM ELECTRIC STATION TABLZ 1 MA'JRRIAL CHANGES Pipe Description Head.Spray Core Spray Influent Size~NPS 12 II previous Material Type 304SS New Material Carbon-Limited.

Type 304SS Control Rod, Drive 3" Hydraulic Return CRD Return Line Has Been Deleted.Standby Liquid.Control Reactor Water Cleanup Effluent 1-1/2" 4II Type 304SS to fir st valve p then Carbon Steel Type 304LSS to first valve, then to Carbon Steel Instrument piping Vent, Drain, and.Test Connections Shown on Figure 1 1", 2" Type 304SS 4I I Type 304SS Type 304L SS Type 304L SS Recirculation'ystem Bypass Bottom Drain 4II 4I I Type 304SS Type 304SS Carbon-Limited, Type'304SS Type 304L SS TFO: bah 17 M~1 M CONTAINMENT PENETRATION HEAD SPRAY REACTOR VENT TO INST ATM VENT VENT TO lASt, VESSEI.VENT N LIAIN STEAM l FEEOIVATER TO VESSEL I 6>>MAIN STEAM FEEOWATER TO VESSEL SLC TEST CONN'ECTION VENT V CONNECTION DRAIN CONN f CT I ON STANDBY LIGVID CONTROL SYSTEM REACTOR IYATER CLEANUP SYSTEM RESIDUAL NEAT REMOVAL SYSTEM CONTROL AOD DRIVE SYSTEM 12" CORE SPRAY M 1>>12 1>>12" R EACTOR PRESSURE VESSEL 12" I>>1>>>>COR'E SPRAY , CRO RETURN (DELETED}LINE>>RHR RETURN 24" A ECIRCY LOOP J 24 RECIRC+LOOP 24" 24 RHR RETURN I>>M 1 20 RHR SUCTION RECIRCg 26" P I LOO 20>>20" 4 GOTT OM DRAIN 4>>4>>4>>RECIRC 9 LOOP INST CONN 1 1/2" SLC RWCU SUCTION 6" 6" INST CONN V 1>>4 1" LT M 0 I OOI SOBGI Figure 1