ML18025A645
| ML18025A645 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 09/25/1978 |
| From: | Curtis N, Mead E Pennsylvania Power & Light Co |
| To: | Parr O Office of Nuclear Reactor Regulation |
| References | |
| 841-2, 883, ER 100450, PLA-291 | |
| Download: ML18025A645 (28) | |
Text
~ REGULATORY INFORMATION DISTRIBUTION SYSTEM ( RIDS)
DI TRIBUTIGN FOR INCOMING MATERIAL
- 50-387 388 REC: PARR 0 D ORG: C!JRTIS N W DOCDATE: 09/25/78 hlR(:: PA PWR 8. I..IGHT DATL RCVD: 0'?/28/7c DO( TYPE: I ETTER NOTARIZED: NO COPIES RECEIVE~
SUBJECT:
LTR 1 ENCL 40 FORl!ARDIhlG REVISION fO APPLICAtlT"8 REPT SUBNITTED 10/13/78) CONSISTING GF REPT ENTITLED: "A REACTION TO CRACKING OF AUSTENITIC STAINLESS STEEL PIPING IVI BGILIhlG WATER REACTORS", AhlD INCLL!DIhlG S!JBJECT FACILITY" - "E-. DESlbhl M()9 IF I CAT I GhlS.
REV I EWER INI1'IAL: X iN PLANi'AME:SUSQ!JEliAhlh!A SUSQUErI ANNA - UNIT 2 llNIT 1 DISTRIB!JTER INITIAL: ~
DISTRIBUTION OF Tl-IIS MATERIAL IS AS FOLLOWS NOTES:
SEND ISE 3CYS FSAR cc ALL ANDTS AUSTENITIC STAIhlLE~S STEEL IN BWR (DISTRIBUTION CGD- B012>
FGR ACTION: ASST DIR VASSALLO++W/ENCL BR CIIIEF LWR43 BC++W/ENCL LIC ASST LWR83 /ENCL INTERNAL: REG FILFii NRC PDR<>W/ENCL
~ciW/2 ENCL GELD+>W/ENCL DIRECTOR DPM%+W/EhlCL DEPUTY DIR DPM44W/EhlCL F W ILLIAMS~~W/ENCL II SMITHii+W/ENCL DIRECTOR DSS+<W/EhlCL AD FOR ENG++W/LNCL MATERIAL ENO BR44W/ENCL CONRAD>+W/ENCL L CROCKER<">W/ENCL AD FOR ENGR 8c PROJ++W/ENCL EXTERNAL: LPDR S W ILKES BARRE PA v<W/ - NCL TERA~~W/ENCL NS I C-"'-:i W/ENCL ACRS CAT Bci+W/16 ENCL DISTRIBU! I()N: LTR 37 EllC'7 CONTROL NBR: 7'780 SIZE: 1 P+21P i%A
TWO NORTH NINTH STREET, ALLENTOWN, PA. 18101 PHONEc (215) 821-5151
$ tP ~ 8 )9+
Mr. Olan D. Parr, Chief Light Water Reactors Branch No. 3 Division of Project Management U.S. Nuclear Regulatory Commission Washington, D.C. 20555 SUSQUEHANNA SES NRC STAFF POSITION ON THE USE OF AUSTENITIC STAINLESS STEEL ER 100450 FILES 841-2, 883 PLA-291
Dear Mr. Parr:
Enclosed are 40 copies of a report entitled "A Reaction to Cracking of Austenitic Stainless- Steel Piping in Boiling Water Reactors." This report is a revision to a report provided to you on October 18, 1976.
Subsequent to the original issue of this report there have been numerous incidents attributed to Intergrannular Stress Corrosion Cracking (IGSCC).
Based on these incidents, PP5L has modified its position on IGSCC, which included additional design changes at SSES. For this reason and to provide PP5L's response to your letter of February 15, 1978, this report has been revised and updated.
Very truly yours, N. W. Curtis WEB:JLI Enclosure PENNSYLVANIA POWER 8 L I GH T COMPANY
gx<) )
.0
PENNSYLVANIA POWER 8c LIGHT CCNPANY A REACTION TO CRACKING OF AUSTENITIC STAINLESS S9ZEL PIPING IN SOILING WAAR REACTORS
( INCLUDES SUSQUEHANNA SES DESIGN MODIFICATIONS)
Earle M. Mead. r Progect Engineering Manager Susquehanna SES
Table of Contents
- l. Introduction
- 2. Problem Statement
- 3. Safety Significance 4'rimary Considerations 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 50 le 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 Cleanliness
Table of Contents 5.2.3 Methods of Stress Reduction 5.2.3.1 Solution Heat Treatment 5.2.3.2 Heat Sink Melding 5.2.3.3 Induction Heating Stress Improvement
- 5. 3 Material 5.3.1 Core Spray System 5.3.2 Reactor Recirculation System Discharge Gate Valve Bypass Line 5.3.3 -
Control Rod Drive Return Line 5.3:4 Recirculation Riser Pipes
- 6. References
Table and Illustration Table 1 F1gure 1
- 1. Introduction This report original+ 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 additional information should and it was agreed. that this be submitted.
Subsequent to the original issue of this report there have been numerous incidents attributed to Intergranular Stress Corrosion Cracking (LGSCC). Based on these incidents, PP8cL has modified our position on IGSCC, which included additional design changes at SSES.
For the above reason and. to provide the NRC with PP&L's position on NUREG-0313, this report has been revised and. updated.
- 2. Problem Statement I 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) Control rod. drive return lines - Tarapur PP&L believed that sufficient Justification existed to recognize the cause as being Xntergranular Stress Corrosion Cracking. This report was originally issued on October 12, 1976 and summarized PP&L'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 PP&L'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 PP8cL 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. visually 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 comply with NRC Regulatory Guide 1.45 (May 1973).
The Inservice Inspection Program at SSES is in accordance with the requirements of the ASS 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 Winter 1975 addenda, "Ultrasonic Examination Method.
for Class 1 and Class 2 Piping Systems Made from Ferritic Steels",
and IWA-2232 of the Summer 1976 Addenda, using 50$ 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 PP8cL'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. Primary Considerations 4.1 Environment (Coolant Chemistry)
The coolant chemistry in Boiling Water Reactors is establ1shed primarily to ensure compatibility 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.l ppb. However, during low load or no load conditions the oxygen concentrat1on can approach saturation which, under conditions of standard temperature and pressure, is 8.0 ppm. Under the combined, 1nfluences of sensitization and.
high tensile stress th1s 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, t1me-to-failure is directly proportional to dissolved oxygen concentration, low flow and. stagnant lines are highly suspect.
It would, therefore, be very beneficial to reduce the oxygen level to as low as possible during all phases of plant operation.
4.2 Stress The design stress levels for Nuclear Steam Supply System (NSSS)
Pip1ng are established within the constraints of the ASME Code as well as any load1ng restrictions which might be imposed. by the NSSS vendor. Stress levels sufficient to result in IGSCC generally are not Just the result of ordinarily-applied, engineer1ng loads or stresses. Rather, they result from the combined effects of all sources of stress and strain: i.c.,
residuals, thermal, surface, service, etc.
It is very beneficial to minimize the amount of stress acting on the sensitized. material. Therefore, the amount of residual and applied stresses should. be reduced. wherever possible.
4.3 htaterial Boiling Water Reactor piping is fabricated. from steels which are:
corrosion resistant, tough, stable dimensionalIy, sufficiently strong for anticipated. loadings, resistant to radiation damage, resistant to both acidic and basic chemical attack, economical, and anticipated to be available in the foreseeable future.
prominent materials used are austenit1c stainless steels The'ost (Types 304 and. 316) and carbon steel..
However, it has been determined. that the standard. grades of 304 and. 316 stainless steel with high carbon contents of approximate+
.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.
- 5. SSES Preventive Measures PAL efforts have centered. on reducing the probability of occurrence of IGSCC for SSES.
P Recommended. modifications to the existing SSES design were evaluated.
prior to implementation against the following criteria:
- 1. Their potential for substantially reducing the probability that IGSCC will occur;
- 2. Their potential for substantiallg reducing the time required. to detect cracks or leaks resulting from IGSCC; 3, Their potential for creating other problems which would either be as bad as, or worse than, the current problem.
5.1 Environment Current analyses indicate that among the environment related.
contributors to the current IGSCC problem in BNR'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 BUR 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 followed.
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 7.0 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 PP8cL has made to improve the water chemistry at SSES.
5.1 2 1 Control Rod Drive (CRD) Pump Suction Relocation PP8cL previously relocated 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
ip to primary cycle 1nfluent water sources such as the CRD system itself. Hence, locating the CRD pump suction on the makeup/reject line results in utilization of water with the lowest oxygen concentrat1ons available (essentially water with feedwater quality) most of the time.
At the time the above change was made, PP&L believed that this change, in addition to other changes made, was adequate to suff1ciently reduce the probability of IGSCC.
Since that time, however, PP&L has found. that controlling the 02 concentration as an effective means of controlling IGSCC requires 02 control during all phases of plant operation, not just normal operation. For th1s reason, PP&L added. a mechanical deaeration system (see section 5.l.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 PP&L's preference to have the 02 control system installed and. operating prior to plant startup.
50 lt3 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 relat'ive 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 t1me 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 monitor1ng of feedwater and reactor water for dissolved gxygen concentration, conductivity and. pH and. continuous monitoring of the CRD system water for dissolved oxygen concentration and. conductivity.'he sampling system automatically alarms when water quality conditions considered. to be adverse to BMH operation are reached..
Procedures will be developed. wh1ch will minimize excessive oxygen concentrations in the primary cycle and to enable plant operators to take immediate action to protect the plant from prolonged operation w1th adverse water quality conditions.
5.2 Stress Stress alone is not particular+ significant so far as the overall corrosion of metals 1s concerned. Mhen combined with a corrosive env1ronment, however, the appU.cation of sufficient tensile stress in susceptible materials can lead to stress corrosion cracking. A reduction in total tensile stress in lines considered susceptible to IGSCC could therefore be cons1dered. beneficial.
In an attempt to reduce overall tens1le 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,'eadweight 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 achievable.
Recognizing the particular susceptibility of the Core Spray System inJection lines, alternate rout1ngs were chosen in order that the thermal stress component of total stress is reduced. by 25$ .
5' ' Fabrication Stresses The difficulty in assuring compl1ance under all shop and. field fabricat1ng conditions, limits the effectiveness of procedures which might be developed in this area. However,. since 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-1nduced, stresses cannot be overlooked.
5.2 ' l Fit-up Code tolerance for alignment are canplied with to assure minimum stress from misalignm nt and. minimum degradat1on of fatigue res1stance.
5.2.2.2 Initial Fabrication (shop)
The pipe material is purchased. in the solution annealed condition. Normally, spool pieces are not solution annealed due to the difficulty of maintaining desired. dimensions.
I 5.2.2. 3 Welding-Induced Stress PP8cL recognizes that when austenitic stainless steels are welded, som level of residual stress and. sensitization 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. sensitizat1on can be ach1eved without sacr1ficing good. penetration and. fusion.
Unfortunate+, precise quantitative values of heat input, cooling rate, etc., which will insure consistently good quality welds resistant to IGSCC are not available.
Therefore, PPM. relies on past industrial experience for guidance.
PPM has adopted the following measures for field. welding the applicable lines. These measures help reduce welding-1nduced stresses, without 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. Interpass temperature limited, to 350 F.
C~ Block welding prohibited.
- d. Electrode size limited. to 5/32" Max. for &%W and l/8" Max. for GTAW wh1ch effectively limits the heat input.
5.2.2.3.2 Joint Design
- a. The root is made with GTAW utilizing hand. fed filler wire or a consumable insert to insure complete penetration and. good fusion.
The extended-land Joint design has no 1nherent problems with lack of penetration, lack of fusion or excessive residual stress.
C~ An in rt gas purge is used prior to weldin'g 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 (i.e., no excessive undercut, excessive reinforcement, coarse ripples, etc.) to reduce the occurrence of "notches" which can detrim ntally affect fatigue strength or corrosion resistance.
10 5.2.2.3. 3 Filler Metal To minim1ze microfissur1ng and sensitization" problems, 308L filler metal or 309 and 309L filler metal is specified with minimum delta ferrite contents of 8 percent and. 5 percent respectively.
5.2.2.3.4 Clean11ness To prevent contamination of the joint:
- a. Grease, oil and. other contaminants are removed. from the joint and. the filler metal prior to making the weld..
- b. Only marking crayons, chalk, 1nk and. temperature ind,icating crayons which are certified. to be low in halogen and sulfur content are used..
- c. Only cleaning solvents which are not harmful to austenitic sta1nless steel are used,.
d.. Stainless steel wire brushes are used..
- e. Grinding wheels used on other materials are not used. on stainless steel.
- f. Grind1ng 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 with what was specified.. Restrictions cons1stent 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 FPEcL f1rst formed a position on the subject, PP&L has stepped. up its monitoring of the problem and. its possible countermeasures. Those countermeasures that are be1ng investigated/evaluated. which deal with stress reduction are Solution Heat Treatment (SHT),
Heat Sink Welding (HSW), and. Induction Heating Stress Improvement (IHSI).
5.2.3.1 Solution Heat Treatment This IGSCC countermeasure stress relieves shop welds before the sections of pipe are shipped to the field for installation. This method. can on+ be used. on shop weld,s, and. then, only when the p1pes have not been installed,.
'I 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 risers underwent SHT because PP8cL believed these lines to be extremely 'susceptible to IGSCC.
The field, weld ends of the risers were also corrosion resistant clad. This process will be d,iscussed. 1n a later section.
5.2.3.2 Heat Sink Welding This XGSCC countermeasure requires cool1ng the ID of the pipe with cool1ng 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 IGSCC.
Th1s countermeasure has not been used at SSES. However, Be'chtel 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 weld.s 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.
Th1s 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 (Reportedly the Japanese have used. this procedure successfully on their nuclear power plants.). PAL intend.s to track the progress of GE and. use the procedure on susceptible lines if ancl when the procedure is determ1n d. to be feasible.
5-3 Material Due to the restrictions of coolant chemistry and total tensile stress levels, the use of substitute materials which are less susceptible to XGSCC were considered,. The lack of significant operating data for mater1als other than carbon steel or Type 304 and. Type 316 sta1nless steels limits the opt1ons, however.
During 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 AS24 A155 Grade KC 70, Class l feedwater pipe.
PPM is, therefore, unwilling to use carbon steel as an IGSCC fix.
The principal drawback to the continued use of Type 304 and Type 316 stainless steel for NSSS piping is their susceptibility to IGSCC. PP&L believes this susceptibility-is related to sensitization which occurs adjacent to a weld 1n 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, it was logical to limit the carbon content of susceptible lines.
Or1ginally, the history of the IGSCC problem formed the basis of the assumption that the combination 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 which were less than or equal to 12" diameter NPS. Any such substitutions would apply equally to pipe fittings. However, valves and conta1nment penetration flued h ads 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 max1mum carbon (with the exception of the Recirculat1on System Di,scharge Gate Valve Bypass L1n ). Stainless piping located w1thin 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. These two materials are virtually identical metallurgically but PP&L is unwilling to sacrif1ce the mechanical properties of 304 for certain piping systems.
Table 1 identifies piping for which a change in material was
)ustified, the material previously sp cif1'ed, and the replacement material chosen to mitigate 'the probability of IGSCC.
Since the time the above material changes were made, additional incidents of IGSCC have shown that the a ssumption that only those lines which are 12" diameter NPS or less are susceptible to IGSCC is incorrect. Therefore, all lines must be considered when attempting to eliminate or reduce IGSCC.
In add1tion to simple material replacement, the other "material" related XGSCC countermeasure wh1ch can be evaluated is Corrosion Resistant Cladding (CRC). This countermeasure combines cladding the field welded pipe ends with a highly corrosion resistant metal with solution annealing which effectively 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 this report, PPM utilized CRC on the recirculation jet 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 Riser Pipes are discussed below.
5.3.1 Core Spray System That portion of the core spray system which is located within the primary containment will, for Susquehanna SES, be made from 12" diameter NPS, Type 304 Stainless Steel Pipe with supplemental maximum carbon limitation of 0.030 percent.
Pipe and fittings will be handled similarly.
532 Reactor Recirculation System Discharge Gate Valve Bypass Line From an operational standpoint PPM. does not wish to delete this line.
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 idle loop+
- 3. It provides pressure equalization on both sides of the discharge gate valve to'ssure proper venting and closure of the valve.
- 4. 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 desire to limit th 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.
533 Control Rod Drive Return Line This 3" diameter line was changed from 304SS to 304L SS which, it was believed, would. solve the problem of IGSCC. Subsequent to this change there were numerous reported incidents 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 11ne and make other changes to the system to maintain. the system design function. PAL concurred. with GE's recommendation and deleted the return lin . This action (1) eliminated the problem of nozzle cracking due to thermal gradients and (2) eliminated the possibility of pipe cracking due to IGSCC.
Recirculation Riser Pipes The 10" recirculation riser p1pes leading from the recirculation header to the jet pumps have recently experien'ced IGSCC at other operating plants. The cracks have been formed. in the heat affected. zones of the thermal sleeve to safe end. attachment welds. The pipes have been fabricated from 304 stainless steel. The safe ends are Imonel 600, and. the nozzles are carbon steel clad with stainless steel.
PPEcL 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 CHC process.
- 3. These welds will then be field welded using 308L weld1ng rod s ~
Th1s process w111 prevent 304 stainle'ss steel in the heat affected, area from coming in contact with the process fluid.
15
- 6. Re ference s Source Document Date GE NEDO-21000 July, lgf5 NUREG 75/067 October, lg75 ASST "Stress Corrosion Cracking of October, 1971 Metals - A State of the Art" NATO "The Theory of Stress October, 1971 Corrosion Cracking in Alloys" "Fundamental Aspects of September, 1967 Stress Corrosion Cracking" United States Joint Hearing Concerning February, 1975 Senate "Nuclear Regulatory Commission Action Requiring Safety Inspections Which Resulted in Shutdown of Certain Nuclear Power Plants"
16 SUSQUEHANNA STEAM ELECTRIC STATION TABLE 1 MATERIAL CHANGES Size Previous New Pipe Description ~EBB) Material Material Head. Spray 6II Type 304SS
~
Carbon-Limited.
304SS Core Spray Influent Control Rod Drive 3" CRD Return Line HyLraulic Return Has Been Deleted.
Standby Liquit 1-1/2" Control Reactor Water Type 304SS to Type 304LSS to Cleanup Effluent first valve, first valve, then then Carbon to Carbon Steel Steel II Instrument Piping P Type 304SS Type 304L SS 41 I Vent, Drain, and, Type 304SS Type 304L SS Test Connections Shown on Figure 1 4II Type 304SS Carbon-Limited, Recirculation'ystem Bypass Type 304SS Bottom Drain 4II Type 304SS Type 304L SS TFO:bah
CONTAINMENT 1 M PENETRATION TEST HEAD SPRAY CONNECTION ATMVENT VENT REACTOR VENT V CONNECTION VENT TO MSL TO INST VFSSELVFNT ORAIN D
STANOSY LIOIIIO SLC CONTROl SYSTEM MAINS7EAM hlAIN STEAM REACTOR WATER CLEANIIP SYST EM l RESIOVAL HEAT REMOVALSYSTEM FEEDYTATER F EEDWATER CONtROL ROO ORIVE SYSTEM TO VESSEL TO VESSEL I
1 2N 12 12" CORE SPRAY CORESPRAY TN REACTOR PRESSURE VESSEL CRD RETURN (DEXZTED)
LINE R ECIRC r LOOP J + RECIRC LOOP M 24" 24 24 24" 24 RHR RETURN RHR RETURN 20 RECIRC/
OOP ~ 28" 4
4 ~
9 1
RECIRC LOOP 1/2" 1 1/2" SLC RHR SUCTION 4N 2" .
4 BOTTOM INST CONN DRAIN M D
TM 4w M
V 4 1 RY/CU SVCTION LT hl
~
B INST I CONN D
G I 001XSGI