ML20081B439
| ML20081B439 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 03/06/1984 |
| From: | Daltroff S PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| IEB-83-02, IEB-83-2, NUDOCS 8403090131 | |
| Download: ML20081B439 (21) | |
Text
. __________
PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET P.O. BOX 8699 PHILADELPHI A. PA.19101 (215: 84150o1 sHistos atinorr ELEC7m>C PnoOuCTION March 6, 1984 Docket No. 50-277 Mr. John F. Stolz, Chief Operating Reactor Branch No. 4 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Dear Mr. Stolz:
On July 4, 1983, Peach Bottom Unit 2 was shut down for augmented in-service inspection of non-conforming austenitic ntainless steel piping welds in the recirculation system and the residual heat removal system (RHR) in accordance with I.E.
During the outage, a total of 126 welds were inspected and reportable indications were detected at 26 locations: six 12-inch riser welds, eight RHR system welds, eight 28-inch recirculation welds and four 22-inch recirculation manifold welds.
Evaluations submitted by Philadelphia Electric Company to the NRC, by letters dated October 6, 1983, October 19, 1983, and November 18, 1983, reported that intergranular strrss corrosion cracking (IGSCC) had been detected at 26 weld locations.
Twenty-one of these weld locations were weld overlay
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repaired; the IGSCC at the five remaining weld locations were of F
a size that permitted Unit 2 to operate within the ASME code allowable limits for a period of 24 months without any weld overlay repair.
By Confirmatory _ Order issued November 30, 1983, from J.
F. Stolz, USNRC, to E. G. Bauer, Jr., Philadelphia Electric Company, restart of Unit 2 was authorized.
Unit 2 vas allowed _to operate for up to 4,100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> with enhancyd; surveillance measures for containment leak detection.
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PHILADELPHIA ELECTRIC COMPANY 23O1 M ARKET STREET I
P.O. BOX 8699 PHILADELPHIA. PA.19101 5HIELOS L.DALTROFF March 6, 1984 j
Docket No. 50-277 L
Mr. John F. Stolz, Chief Operating Reactor Branch No. 4 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Dear Mr. Stolz:
On July 4, 1983, Peach Bottom Unit 2 was shut down for augmented in-service inspection of non-conforming austenitic stainless steel piping welds in the recirculation system'and the residual heat removal system (RHR) in accordance with I.E.
During the outage, a total of 126 welds were inspected and reportable indications were detected at 26 locations: six 12-inch riser welds, eight RHR system welds, eight 28-inch recirculation welds and four 22-inch recirculation manifold welds, Evaluations submitted by Philadelphia Electric Company to the NRC, by letters dated October 6, 1983, October 19, 1983, and November 18, 1983, reported that intergranular stress corrosion cracking (IGSCC) had been detected at 26 weld locations.
Twenty-one of these weld locations were weld overlay repaired; the IGSCC at the five remaining weld locations were of a size that permitted Unit 2 to operate within the ASME code allowable limits for a period of 24 months without any wold overlay repair.
By Confirmatory Order issued November 30, 1983, from J.
F. Stolz, USNRC, to E. G. Bauer, Jr., Philadelphia Electric Company, restart of Unit 2 was authorized.
Unit 2 was allowed to operate for up to 4,100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> with enhanced surveillance measures for containment leak detection.
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Mr. John F. Stolz Page 2 In accordance with the instructions to submit plans for
-pipe replacement for NRC staff review one month prior to the start of such work contained in the Confirmatory Order of November 30, 1983,-the attachment to this letter documents Philadelphia Electric Company's plans for performing the pipe replacement during.the upcoming Unit 2 refuel outage.
It is the intention of Philadelphia Electric Company to proceed with work as described in the attachment and have all modification / refueling outage work completed within thirty-two weeks.
If you have any questions or required additional information in this manner, please do not hesitate to contact us.
Very truly yours,
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Attachment cc: A.
R. Blough, Site Inspector
m Docket No. 50-277 e
i DESCRIPTION OF PLANS FOR RECIRCUIATION AND RESIDUAL HEAT REMOVAL SYSTEMS PIPING REPLACEMENT AT PEACH BOTTOM UNIT 2 DOCKET NO. 50-277 4
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Description of Plans for Peach Bottom Unit 2 Recirculation and' Residual Heat Removal Systems Piping Replacement l
Outline Item 1.0 Introduction 1.1 Purpose 1.2 Scope 1.3 Future Reports 2.0 Project Description 2.1 Overall Organization l
2.2 Project Management 4
i-2.3 Engineering 2.4 Contractor l
2.5 Support Services 1
2.5.1 Quality Control 2.5.2 Health Physics l
2.5.3 Consultants.
l 2.5.4 Licensing l
2.5.5 Authorized Nuclear Inspector 1
1 l
3.0 Description of Work I
3.1 Defueling/ Fueling l
3.2 Work Sequence l
3.3 System Testing for Startup 4.0 Design Description
~4.1 Recirculation System & RHR System -
Existing Design 4.2 Replacement Design 5.0 Health Physics Control 6.0 Codes & Standards 7.0 Quality Assurance / Quality Control Item (Continued) 8.0 Training 8.1 Mockup Training 8.2 Welder Qualification 8.3 Health Physics 9.0 Piping Analyses 9.1 Nuclear Class I Component Evaluation Based On ASME Section III 9.2 Piping Stress Analysis 9.3 Pipe Support-Design 9.4 Pipe Whip Restraint Design ATTACHMENTS.
Existing Recirculation System and Residual Heat Removal System Piping Configuration Modified Recirculation System and Residual Heat Removal System Piping Configuration i
1.0 Introduction 1.1 Purpose This submittal describes the activities and system design in order to support the replacement of the recirculation system piping on Peach Bottom Unit 2.
1.2 Scope.
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This submittal describes the engineering, design, E
procurement, removal and installation of the
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recirculation system piping and portions of the I
residual heat removal (RHR) system piping.
1.3 Future Reports In accordance with NRC Confirmatory Order, November S
30, 1983, a justification for continued operation i
will be submitted for NRC review and approval prior to restart following this outage.
2.0 Project Description The project task is to replGce the existing 304 stainless steel piping in the Peach Bottom Atomic Power Station Unit 2 Reactor Recirculation System and RHR System (inside containment) with material which is resistant to IGSCC.
The specific piping sections included in the plans are shown on the Attachments 1 through 3.
2.1 Overall Organization The project is organized with Philadelphia Electric Company (PECo) providing the overall direction to the engineers and contractors involved.
PECo will also review and approve the specifications and proce3ures to be used.
Each engineer or contractor will perform his work under the controls provided by his or PECo's Quality Assurance Plan and implementing procedures.
An overall project interface document has been prepared to provide control of the interfaces between the individual organizations' procedures.,
2.2 Project Management PECo is responsible for overall project management.
The services of the engineers and contractor are used to prepare major documents and schedules at PECo's request.
g 2.3 Engineering Engineering and major. material procurement will be provided in general by the organizations which did so for the original plant installation.
General Electric Company (GECo) will design and provide material for the Reactor Recirculation System.
Bechtel Power Corporation (Bechtel) will design and provide material for the portions of the RHR System which are being re~ laced.
p 2.4 Contractor Chicago Bridge and Iron Company (CBI) will be'the prime contractor for the pipe replacement work.
2.5 Support Services 2.5.1 Quality Assurance Each engineer and contractor provides the required quality control services for items within his scope of work.
In addition, Philadelphia Electric Company performs periodic audits of the project participants to verify compliance with contractor's OA/QC plan and procedures.
2.5.2 Health Physics The required health physics controls are provided by PECo.
Peach Bottom personnel will be assisted by subcontractors working directly for CBI.,
e 2.5.3 consultants Consultants will be retained as needed to provide specific expertise or supplement the activities of PECo personnel.
2.5.4 Licensing i
Licensing will be directed by PECo with specific assistance being provided for submittals by the engineers and contractor.
2.5.5 Authorized Nuclear Inspector The authorized nuclear inspector will be Hartford Steam Boiler Inspection and Insurance Company.
3.0 Description of Work 3.1 Defueling/ Fueling Prior to initiation of the piping replacement, the entire core will be off-loaded and stored in the fuel pool.
Following core off-loading, it is anticipated that, due to ALARA considerations, some peripheral control rods will be removed as well.
Fcllowing completion of the pipe replacement and other maintenance tasks associated with the refueling outage, the core will be reloaded to i
complete the refueling portion of the outage.
3.2 Work Sequence The construction plan is divided into four phases.
- Phase One - Interference Removal and Piping Decontamination
- Phase Two - Sever / Remove existing piping
- Phase Three - Machine / Install new piping l
- Phase Four - Interference Restoration / Testing f
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F (1) Phase One - INTERFERENCE REMOVAL AND PIPING DECONTAMINATION During this phase, final planning will be done, drywell systems Which will interfere with the piping changeout will be checked and, where appropriate, temporarily removed.
The drywell RHR and recirculation systems piping will be prepared for removal.
Both systems will be decontaminated to the extent practical to minimize exposure to those personnel involved in the modifications.
(2) Phase Two - SEVER / REMOVE EXISTING PIPING Following system decontamination and the removal of interferences in the drywell, equipment will be set-up to cut and remove the recirculation and RHR piping.
To accelerate this removal, the existing system will be cut into manageable pieces for removal from the drywell.
System severing will be done with the use of plasma cutting or machining.
Mockups will be made and a system for containing airborne contamination for certain cutting and welding operations _will be tested prior to use inside the drywell.
(3) Phase Three - MACHINE / INSTALL NEW PIPING This phase covers the preparation of in-place weld joints, and installation of the new recirculation and RHR piping loops.
Remote automatic welding and machining processes will be utilized as much as possible for ALARA considerations.
(4) Phase Four - INTERFERENCE RESTORATION / TESTING Inventory control implemented When systems or structures are removed will assure proper restoration of replaced, modified, and temporarily modified componente and systems within the drywell prior to restarting the unit.
3.3 System Testing for Startup Non-dsstructive examination (NDE) techniaues involving radiographics, ultrasonics, liquid penetrant testing and/or magnetic particle testing will be employed as required by the applicable codes during fabricasion and installation.
Testing will be performed in accordance with PECo approved procedures.
Upon completion of installation work, a baseline examination will be made of all new piping welds in - _ _ - _ - _ _ _ _ _ _
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J accorcance with Section XI of the ASME Code and the PUCo In-Service Inspection (ISI) program.
A pre-operational test program will be developed and implemented for replaced, modified, and systems temporarily modified for interference removal.
In addition, surveillance testing will be performed prior to startup in accordance with the technical specifications.
4.0 Design Description 4.1 Recirculation System & RHR System - Existing Design Peach Botton Unit 2 is a single cycle forced circulation, BWR-4 reactor licensed to operate at 3293 Mwt.
Forced circulation through the reactor core is provided by the rec.irculation system.
This system consists of two exter. sal loops with associated pumps and twenty internal jet pumps.
The design bases of the existing recirculation system is as follows:
External Loops i
Number of Loops 2
Pipe Sizes (nominal O.D.)
Pump Suction 28 inches Pump Discharge 28 inches Discharge Manifold 22 inchcs Recirculation Inlet Line 12 inches Cross-Tie Line 22 inches Design Pressure (psig)/ Design Temperature (degrees F)
Suction Piping 1148/562 Discharge Piping 1326/562 i
Pumps 1500/575 Operation at Rated Conditions l
Recirculation Pump Flow (approximate) 45,200 gpm Flow 17.1 x 10P6 lbs/hr Total Developed Head 710 ft.
Suction Pressure (static) 1032 psia Available NPSH (min.)
500 ft.
Water Temperature (max.)
528 degrees F Pump Hydraulic HP (min.)
6130 hp Flow Velocity at Fump Suction 27.5 fps (approximate)
-Drive Motor and Power Supply Frequency (at rated) 56 Hz i
Frequency (operating range) 11.5-57.5 Hz 1 I
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l Total Required Power to M-G Sets kW/ set 6730 kW kW total 13,460 kW Jet Pumps Number 20 Total Jet Pump Flow 102.5 x 10P6 lbs/hr l
Throat I.D.
8.18 inches Diffuser I.D.
19.0 inches 4
Nozzle I.D.
(representative) 3.14 inches Diffuser Exit Velocity 15.3 fps.
Jet Pump Head 76.1 ft.
The RHR system is designed for three modes of operation to aatisfy its safety objectives and design bases.
Each mode of operation is defined as a subaystem of the RHR and are as follows:
- Containment Cooling
- Low Pressure Coolant Injection
- Shutdown Cooling and Reactor Vessel Head Spray The design bases of the existing RHR system is as follows:
Pumps Number Installed per Unit -
4 Capacity per Pump -
33 1/3% (LPCI)
Design Temperature -
350 degrees F Design Pressure -
450 psig Design Conditions per pump at 20 psid*
Discharge Flow -
10,000 gpm Discharge Head -
542 ft.
Operating Conditions per Pump Discharge Flow -
O to 12,000 gpm Discharge Head -
800 to 372 ft Differential Pressure -
295 to O psid -depicts the arrangement of the existing recirculation syctem and RHR system.
The existing recirculation piping materials were purchased in accordance with General Electric's specification, " Recirculation Loop Piping".
The i
material used in the original design was Type 304
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Austenitic Stainless.Stee}. 'This piping was designed in accordance'with ANSI B31.1.
The existing RHR piping material,jalso type 304 Stainless Steel,"was-designed in accordance with ANSI B31.1.
4.2Property "ANSI code" (as page type) with input value "ANSI B31.1.</br></br>4.2" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. Replacement ~ Design
- 4. 2.1' Material Sel'ection Rehlacement piping fore he. recirculation and RER systems will he fabricated from Type 316 Nuclear Grade (NG) stainless steel.-
x Type.316'NG is an austenitic stainless steel with a high resistance to IGSCC in the BWR e.nvironment due to its low carbon condent,(less than 0.02 perce'nt by weight).
In addition, nitrogen is added-to this alloy to ad,d strength due to the relatively low carbonicontent.
Increased pitting and sensitization resistance is provided by the addition of molybdenum.
4.2.2 Design' Conditions; The replacement. piping will be designed to meet or' exceed the existing piping design pressures and temperatures stated below.
i' Design Design x
Pressure Temperature (psig)
(Degrees F)
Recirc Pump Suction Piping 1140~
562 Recirc Pump Discharge Piping 1326 562
,RHR Suction Piping 3148 562 RHR Return Piping 1326 562 4.2.3 Piping Design Improvements-Design improvements consist of:
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Use of bent pipe for the 12-inch risers to reduce the number of welds.
Use of long. tangent and super long
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-tangent;28-inch elbows to reduce the s e 4
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.. -. - -..-.-.. w number of welds and to aid in the use of automatic welding and NDE equipment.
Deletion of and caps by using 12-inch bent pipe and a reducer on ends of the 22-inch headers to reduce the number of welds and to eliminate the crud traps formed by the end caps.
Deletion of header sweepolets by using a header with extruded outlets.
Preliminary calculations of the stress l
indices for the extruded outlets show that the indices are lower than those for the original sueepolets.
Use of a single forged pipe fitting to replace the RHR return tee, header cross and reducer.
Counterbore length of two times the pipe wall thickness at all piping weld end preps to facititate ISI.
Two-inch minimum extension at the reduced ends and at the ends of elbows to facilitate ISI.
To minimize problems associated with welding new pipe to existing equipment, such as pumps, valves and safe ends, a wall thickness greater than the code required minimum wall will be provided.
Deletion of unnecessary branch connections, such as the ricer pressure taps formerly used for low pressure coolant injection (LPCI) loop selection logic.
Deletion of the recirculation loop tie valves.
Mechanical polishing to minimize / oxide radiation buildup.
- shows the proposed recirculation replacement piping and the RHR system piping (replaced in same i..
i
,'co,nfiguration as existing RHR system
. piping.)
The recirculation piping was redesigned to eliminate as many fittings as possible; and in addition, bent pipe was utilized in order to~ reduce the number of welds.
The j
replacement piping has approximately 48 less circumferential welds than the existing pipe.
The pipe material specified, Type 316 Nuclear Grade stainless steel, has a low csrbon content to reduce the possibility of IGSCC.
To further minimize IGSCC, all of the piping material (shop welds and field welds are not included) will be solution heat treated.
The physical location of all field welds has been specified to allow easier access for automated welding equipment and for subsequent ISI.
This in expected to significantly reduce future inspection time and inspection radiation exposure.
5.0 Health Physics Controls A job specific radiation protection program has been formulated to assure that reasonable steps will be taken to reduce radiatioc. expoeure and contamination.
A description of the administration of the plan will be submitted to the NRC by March 9, 1984.
6.0 Codes and Standards The original piping systems were installed in accordance with ANSI B31.1.0 - 1967.
For replacement work, as designed in ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, 1980 Edition, the replacements may meet all or portions of the requirements of later editions of the Construction Code.
Based on compliance with Section XI and the current design criteria applicable to this piping, we are specifying compliance with upgraded sections of the construction codes.
The piping replacement work as related to the code requirements will be split into four areas (fabrication, design, installation, and examination).
The piping is being fabricated in accordance with ASME B&PV,Section III, 1980 Edition through Winter 1980 Addenda.
The piping systems will be designed in accordance with ASME B&PV Code,Section III, 1980 Edition through Winter 1981 Addenda.
The systems will be installed in accordance with ASME B&PV Code,Section XI, 1980 Edition through 1981 Addenda.
The nondestructive examination will be in accordance with ASME B&PV Code,Section XI, 1980 Edition through Winter 1981 Addenda.
This will, however, refer back to Section
{
III and Section V for the actual compliance requirements.
The baseline inservice inspection will comply with the Peach Botton ISI program Which requires inspection in accordance with ASME B&PV Code,Section XI, 1974 Edition through Summer 1975 Addenda.
7,0 Quality Assurance / Quality Control The replacement work will be controlled by Chicago Bridge and Iron's (CBI) Quality Assurance Program.
The program utilizes CBI's stLndard manuals, namely, " Nuclear Quality Asurance Manual for Repairs and Alternations", and " Nuclear Quality Assurance Manual for ASME Section III Products".
Both_of these program have been approved by an ASME survey
-team.
CBI's Quality Assurance and Quality Control activities will be subject to audit by PEco's Engineering &
Research Department Quality Assurance Organization.
In addition, their work will be inspected by an Authorized
. Nuclear Inspector.
f 8.0 Training B.1 Mockup Training 8.1.1 Plasma Arc Cutting and Machining Operations Operators for either the Plasma Arc Cutting or machining operation will be trained on mockups simulating expected drywell conditions.
Operators shall demonstrate their ability to perform the reotired operation (s) in the drywell.
8.1.2 Welding Processes Physical restraints such as space, position, and obstructions, etc. will be..,
I
mocked-up full scale.
Welders will be required to demonstrate an ability to make quality welds under thesa conditions prior
' to performing work inside the drywell.
8.2 Walder Qualification Each welder and welding operator will be required to pass applicable code welding' tests by producing test coupons in accordance with specified welding procedures.
Test coupons will 'ee examined non-destructively and/or destructively to prove each welder's capability.
J 8.3 Health Physics All personnel involved in the pipe replacement work will be required to receive training and pass the necessary tests administered by PECo Nuclear Training Section.
Tra!ning is provided in two categories - General Employee Training (GET) and General Respiratory Training (GRT).
GET provides the employee with training in radiation contamination and its reduction, nuclear plant rules, Radiation Work Perrit's dosimetry and dressing out in Anti-C clothing.
GRT provides the amployee with training in the tse and fitting of protective respiratory gear, s
9.0 Piping Anelyses 9.1 Nuclear Class 1 Component Evaluation Based on ASME Section 111 The piping will be analyzed in accor6ance with the rulee of NB-3600 of ASME Section III Boiler and Pressure Vessel Code 1980 Edition through winter 1981 addenda.
The following will be evaluated in the analysis of the piping systems a.
The pipe minimum wall thickness will be determined such that there is adequate wall thickness for the various design and operating pressures defined in the design specification.
Pressure design will be in accordance with the rules in NB-3640, 3654, 3655, 3656 and 3657.
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Primary stress intensity limits of Equation (9) in NB-3652, 3654, 3655 and 3656 will be met.
c.
Primary plus secondary stress intensity range limits will be met for the pressure, temperature and earthquake duty cycles defined in the design specification.
Primary plus secondary stress intensity range j
limit will be met by satisfying the requirement of Equation (10) in NB-3653-1.
If the stress range calculated by Equation (10) exceeds 3 Sm, i
the simplified Elastic-Plastic analysis Equations (12) and (13) will be satisfied, d.
The cumulative usage factor will be met for the B pressure, temperature and earthquake duty cycles defined in the design specification.
The cumulative usage will be evaluated in accordance with NB-3653-6.
9.2 Piping Stress Analysis Due to che change in the piping configure. tion, the recircu'J ation piping and connecting portions of the RHR piping will be reanalyzed.
The recirculation piping and tha RHR suction and return piping, between the tee connection and the drywell wall penetration, will be included in the analyses.
Thermal Expansion Analysis Three thermal conditions have been selected for analysis.
These three thermal conditions will be used to simulate the service levels contained in NB-3600.
Seismic Analysis
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The recirculation piping and connecting RHR pip *.ng j
will be modeled as a lumped mass system with enough details to accurately predict piping dynamic response up to a frequency of 33 Hz.
The' weight of l
the piping contents plus insulation will be added to the weight of the pipo in the form of a uniformly i
i distributed load (lbs/ft).
The weight of pumps and l
valves will be included in the mass model.
For the pump motors and valve operators, the extended mass will be modeled as an additional weight at the I
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The stiffness of each support will be included in the piping mathematical model.
The earthquake analysis will be performed ucing the response spectra method.
The dynamic analysis utilizes response spectrum curves for the site based on ground motion accelerations of 0.05g (design earthquake) and 0.12g (maximum credible earthquake).
The forces and moments due to earthquake differential anchor movements will be determined by a static analysis with the movement at supports as input.
9.3 Pipe Support Design The pipe support design will optimize the use of existing support hardware to minimize the number of the new components to be procured, to minimize drywell changes, and to simplify installation.
In addition, the supports will be designed to minimize the necessity for welded attachments to the piping.
The existing hangers were designed and purchased to the requirements of USAS B31.7 and the existing snubbers were designed and purchased to the requirements of the ASME Section III, Subsection NF.
Due to the change in the piping configuration, the existing hangers and snubbers will be re-evaluated to the code they were purchased to.
It any new piping support hardware is required, it will meet the materials, design fabrication and erection l
l requirements of ASME Section III, Subsection NF.
9.4 Pipe Whip Restraint Design Where the piping configuration has not been changed and e..isting pipe whip restraints can be reused, the criteria for break location will be in accordance with the original design. basis.
The original design basis was the General Electric Document, " Design Report Recirculation System Pipe Whip Restraint for the BWR 4, 218 and 251 Mark I and Mark II Product Line Plant".
Where the piping configuration has been changed, as with the header, and new pipe whip restraint design and/or analysis is needed, it will be done to the latest criteria: NRC Regulatory Guide 1.46, Standard Review Plan 3.6.2 and Branch Terminal Position MEB 3-1 requirements.
Consideration is being given to applying the ?.atest restraint design criteria to the sections of piping where the configuration has not been changed.
By applying the latest criteria, some of the existing I
restraints could to eliminated.
The elimination of some of the restrainto would facilitate piping removal, installation and ISI and would reduce the associated outage time and man-rem exposure, b
1 Docket No. 50-277 _.
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