ML20081K549
| ML20081K549 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 11/07/1983 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20081K547 | List: |
| References | |
| TAC-52483, NUDOCS 8311100113 | |
| Download: ML20081K549 (66) | |
Text
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EDWIN I. HATCH NUCLEAR PLANT UNIT 2 RECIRCULATION PIPING REPLACEMENT REPORT TO NRC NOVEMBER 7, 1983 4
8311100113 831107 PDR ADOCK 05000366 P PDR
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HNP-2 TABLE OF CONTENTS Section Title Page 1.0 Introduction 1-1 s
2.0 Project Description 2-1 3.0 Description of Replacement of Piping 3-1 4.0 Component Design and Fabrication 4-1 ,
5.0 Analyses 5-1 6.0 Radiation Protection 6-1 7.0 Waste Management 7-1 8.0 Piping Insulation 8-1 9.0 Codes and Standards 9-1 10.0 Inspection 10-1 11.0 Training 11-1 12.0 Temporary Facilities 12-1 13.0 Project Safety Program 13-1 14.0 Technical Specifications 14-1 15.0 Safety Considerations 15-1 i
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l HNP-2 LIST OF TABLES No. Title Page 1 Observations of Carbide Precipitation 4-7 in Types 304 and 316 Stainless Steel as a Function of Heat Treatment Time 2 Existing Low Carbon Stainless Steel 4-8 in Operating Plants 3 Loading Combinations 5-4 4 List of Mockups 11-5 LIST OF FIGURES No. Title Page j 1 Project Management Organization Chart 2-6 2 Recirculation Piping Loop "A" Cutting 3-11 Diagram 3 Recirculation Piping Loop "B" Cutting 3-12 Diagram 4 Recirculation Piping Loop "A" Welding 3-13 Diagram 5 Recirculation Piping Loop "B" Welding 3-14 Diagram i
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HNP-2
1.0 INTRODUCTION
l.1 PURPOSE The Recirculation Piping Replacement Report describes the program and organization developed at Georgia Power Company (GPC) to replace the recirculation tjstem piping and certain associated system piping at the Edwin I. Hatch Nuclear Power Plant - Unit 2 (HNP-2). This report has been developed to provide an overview of the overall piping replacement project for the Nuclear Regulatory Commission.
Certain aspects of the project are still in the planning stages and are subject to change. Changes that affect safety will be subject to the plant safety reviews procedures and will involve Nuclear Regulatory Commission (NRC) notification as appropriate.
1.2 SCOPE This report includes information on all aspects of the piping replacement effort. Accordingly, information is included on the project organization, management, structure, and quality control procedures as well as information concerning the technical aspects of the job.
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, 2.0 PROJECT DESCRIPTION l 2.1 OVERALL ORGANIZATION 4
The recirculation piping replacement project (RPRP)~ management organization is made up of personnel from Georgia Power Company (GPC), Southern Company Services (SCS), and Hydro. Nuclear Services, Inc. (HNS). The project management structure is diagrammed on Figure 1 (page 2-6).
Five principal organizations are involved in the replacement project:
A. GPC - Provides overall project management and direction.
B. SCS - Provides engineering management and support in engineering, licensing, and nondestructive examination I
(NDE).
, C. Newport News Industrial (NNI) - Provides general contractor services for the field work.
D. General Electric Company (GE) - Provides the design and analyses for the new piping, and supplies the actual pipe for installation.
l E. HNS - Provides health physics managenent, Radiological
. Engineering, and ALARA services.
. 2.2 PROJECT MANAGEMENT l
Due to the magnitude and complexity of the RPRP, a matrix organization has been developed to support the overall effort.
All personnel assigned to this organization are dedicated solely to this project and do not have any additional responsibilities.
j Heading up the organization is a project manager (PM) who has i overall responsibility for the engineering, licensing, and construction effort. The PM will coordinate all activities to ensure that the project is completed on schedule and within the approved budget.
As can be seen on the organization chart (Figure 1), SCS engineering and GPC licensing will report directly to the PM.
The deputy project manager (DPM) also will report directly to the PM. The SCS lead engineer will be responsible for coordinating the overall engineering effort as well as providing functional direction to the Field Engineering Group. The licensing coordinator will interface with SCS licensing and
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coordinate the overall licensing effort. The DPM will be responsible for all field operations including health physics, field engineering, quality control, and construction.
2.3 ENGINEERING Primary engineering responsibility on the RPRP has been assigned to SCS. SCS will support the project with both a field and home office organization dedicated to the project.
The field organization is being assembled at this time and will consist of representatives from the mechanical, electrical, and civil disciplines. The field office will be staffed to provide timely and efficient support to the construction effort.
SCS's main offices in Birmingham will be available to provide overall project support and to assist the field office personnel when necessary. Items which cannot be resolved in the field will be turned over to the HNP Support Group in Birmingham for final dispositioning.
SCS is currently working to identify and analyze interferences which will have to be dealt with in conjunction with the piping replacement. Work packages are being developed to handle the removal and restoration of the various identified interferences. SCS is also working in the areas of project schedule development, temporary ventilation, and air-conditioning, replacement bills of material, and functional test procedures.
2.4 CONTRACTOR %
NNI has been selected as the general contractor for the RPRP.
NNI has a vast amount of nuclear experience with naval ships and most recently completed the Nine Mile Point RPRP where they acted as the general contractor.
NNI personnel came on site Jugust 15, 1983 and they are staffing up for the outage. NNI is involved in preparing the necessary software for the project and is developing a definitive schedule for their scope of work. Some craft people have already been hired and are involved in training, qualification, and fabrication of special tools.
NNI will utilize a number of subcontractors on this job. The areas in which subcontractors will be utilized include inspection, nondestructive examinations (NDE), and automatic welding services.
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i HNP-2 2.5 PIPING SUPPLIES GE has been selected as the supplier of the new pipe. The recirculation system design has been modified slightly. The modified design is described in section 4.0. In addition to supplying the new piping, GE is responsible for the stress analysis of the new system. Further discussion of the analyses to be performed by GE is detailed in section 5.0.
2.6 SUPPORT SERVICES 2.6.1 QUALITY CONTROL A GPC quality control (QC) surveillance organization dedicated to the RPRP will be in place prior to the start of the outage.
The QC surveillance organization will consist of a QC supervisor with a staff of QC inspectors. The QC Surveillance Group will provide coverage on both the day and night shifts to ensure that work is performed in accordance with contract and code requirements.
An NNI QA/QC Group has been activated at this time and work is underway to develop a comprehensive set of procedures for the job. The QC Group will be respcnsible for many areas including the review of material documentation and storage, the review of contractor procedures and work practices, the review of NDE work and, of course, the review and approval of all completed work.
2.6.2 HEALTH PHYSICS The health physics (HP) support organization has been developed to minimize the impact on other plant operations while providing a distinct organization with specific functions related solely to the RPRP. A manager and assistant manager report to the DPM and interface with the plant HP organization. The Health Physics Project Support Group (HPPSG) is broken down into two main functional areas with supervisors for each. The Radiological Controls Group will focus on drywell work coverage and some balance of plant (BOP) coverage including scheduling surveys, control point management, and determining protective clothing and radiation work permit requirements. The Engineering Support Group will be responsible for implementation of the ALARA program (section 6.0) including radiological engineering, ALARA engineering, and special training. The HPPSG is described in section 6.0.
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! HNP-2 2.6.3 CONSULTANTS l A number of consultants have been retained to assist in preparations for the RPRP. Additional consultants may be utilized in the future as circumstances warrant.
i HNS has been retained to provide HP and radiological engineering assistance in support of the HNP-2 RPRP. HNS will be
! responsible for developing, implementing, and managing the HPPSG t
program to ensure radiation exposures during the job are at a level that is consistent with ALARA. The HNS personnel are on j
site at this time and the development of the HP program is in progress.
GE , the supplier of the replacement piping, will be available to i
provide recommendations and technical assistance throughout the project on an as required basis.
! Bechtel, the original design architect / engineer, will also be j available to provide assistance on an as required basis.
2.6.4 LICENSING i
l The resources of SCS Nuclear Safety and Licensing Department-l (NSLD) are available on an as required basis. SCS/NSLD j coordinated the preparation of this document and will be j responsible for the preparation of other documents as requested l by GPC. GPC will remain th,e interface organization between the project and the Nuclear Regulatory Commission.
- 2.6.5 AUTHORIZED NUCLEAR INSPECTOR
! An Authorized Nuclear Inservice Inspector (ANII) will be retained to monitor the RPRP. The ANII will probably come from
. Hartford Steam Boiler Inspection and Insurance Company or other qualified sources. Having an ANII'on site will provide '
assurance that all work is performed in accordance with American Society of Mechanical Engineers Code requirements. The
, inspector will be responsible for auditing the certification of NDE personnel, the various NDE's which are performed in support of code requirements, and data sheets for completeness and accuracy. The ANII will be responsible for signing off on NDE work, thereby certifying its accuracy.
2.6.6 CONSTRUCTION COORDINATION A Construction Coordination Group working under the DPM will be in place prior to the start of the outage. The group consists of a supervisor and a staff of coordinators adequate to cover.
.both day and night shifts.
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The coordinators will function as the primary interface between NNI construction supervisors and the RPRP organization. They will also be the day-to-day interface between NNI and other contractors or power generation personnel (plant staff) working in the drywell or other plant areas affected by the RPRP.
The coordinators will monitor and report work progress, obtain necessary support from plant staff, and ensure that good housekeeping and safe work practices are followed by -he contractor to ensure that the work goes smoothly and is completed in a timely and cost effective manner.
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ASSitST ANT TO PAOJECT adANAGER >
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SCS FIELD PLANT ENGINEER FIELO COORDINAllNG OUALITY CONTROL SUPEnveSOR OP E00GINEER SUPERVISOR SUPERVISOR p
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FIGURE 1 PROJECT MANAGEMENT ORGANIZATION CilART
HNP-2
3.0 DESCRIPTION
OF REPLACEMENT OF RECIRCULATION PIPING
3.1 DESCRIPTION
OF WORK Actions are being taken for complete replacement of the existing piping made from type 304 material with GE type 316 nuclear grade (NG) material. The piping replacement will be done during the next planned refueling outage in January 1984.
The replacement of the recirculation piping will call for:
A. Replacement of original type 304 material with GE type 316 NG material not exceeding 0.02-weight percent carbon.
(Recirculation piping and the stainless steel portions of the residual heat removal (RHR) and reactor water cleanup (RWCU) piping to the first isolation valve will be replaced. The recirculation piping flow element will be replaced with a flow element which has the same configuration as the existing one and is made from NG CF3 material.)
B. Modification of the piping layout to reduce the number of welds.
C. Utilization of residual stress improvement techniques.
3.2 SEQUENCE OF WORK The following subsections summarize the physical work that will be performed on the recirculation piping.
3.2.1 SYSTEM REMOVAL The following methodology describes steps involved in removing the "A" loop piping. This methodology is based on an optimized flowpath and will be varied to accommodate balance of plant work, ALARA considerations, etc., as required. The "B" loop piping removal will be similar.
Notes on system removal:
A. Removal of the severed pipe components from the drywell will be coordinated on an as needed basis due to the
-space restrictions within the drywell.
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- UNP-2 B. Critical cuts (i.e., cuts on components which will be reused) will be controlled and accurately located. The weld prep surfaces will be ex mined by the liquid penetrant method. .
C. Critical cuts will be performed with a cutting machine utilizing a combination cut / prep tool to provide the preliminary weld prep configuration. Final preps will be completed at a later time if they are not established by initial severence cuts.
d D. Suction and discharge piping will be removed concurrently, with both loops "A" and "B" worked in parallel.
E. Refer to Figures 2 and 3 (pages 3-11 and 3-12) for piping loop layout and cut locations.
3.2.1.1 Suction Piping Removal
- 1. Perform cut to sever the spoolpiece from the suction nozzle safe-end. Machine the safe-end such that the metal in the heat affected zone is removed and the final weld prep configuration is obtained. (Cut 1A.)
, 2. Perform two cuts to sever the RWCU piping from the RHR suction line. Machine the RWCU valve to its final weld prep configuration. (Cuts 2A and 4A. )
(NOTE: Optional cuts may be made to remove pipe from area.)
- 3. Perform two cuts to sever the RHR suction elbow from the RHR riser and the suction riser tee. Machine the existing RHR suction riser to its final weld prep I
configuration. (Cuts 3A and SA.)
- 4. Perform cut to sever the elbow / spool assembly (A-PS-3/A-PS-4) from the suction riser. Once the elbow / spool assembly is removed, install nozzle shielding and cover the riser and nozzle openings.
(Cut 6A.)
- 5. Perform 3 or 4 cuts to sever and ' remove the remainder of the suction riser piping from the elbow (A-PS-9) upward. (Cuts 7A, 8A, 9A, and possibly 10A.)
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- 6. Perform 1 cut to sever the elbow (A-PS-9) from the suction isolation valve. Machine the suction isolation
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. HNP-2 valve (A-PS-10) inlet to its final weld prep configuration. (Cut 11A.)
- 7. Perform 1 cut to sever the suction isolation valve from the spool / elbow (A-PS-11). Machine the uuction isolation valve outlet to its final weld prep configuration. (Cut 12A.)
- 8. Perform 1 cut to sever the spool / elbow (A-PS-11) from the pump suction. Machine the pump suction to its final weld prep configuration. (Cut 13A.)
- 9. Perform 8 cuts to remove various instrument and drain lines as necessary to facilitate the removal of th3 suction piping. (Cuts 14A, 15A, 16A, 17A, 18A, 48A, 49A, and 50A.)
3.2.1.2 Discharge Piping Removal ,
- 1. Perform 2 cuts to remove discharge riser A-PD-1.
Machine final weld prep configuration on the nozzle safe-end. (Cuts 19A and 20A.)
- 2. Perform 2 cuts to remove discharge riser A-PD-2.
Machine final weld prep configuration on the nozzle safe-end. (Cuts 21A and 22A.)
- 3. , Perform 2 cuts to remove discharge riser A-PD-3. (Cuts 23A and 24A.)
(NOTE: The weld joint between the spoolpiece and the nozzle N2J safe-end has been repaired using the weld overlay process. In order to locate the centerline of the original spool to safe-end weld, a plunge cut will be made on the spool side of the overlay. The weld centerline will then be located on the inside of the spool / elbow and then transferred to the outside surface to establish the prep cutline.)
! 4. Machine final weld prep configuration on the nozzle N2J safe-end.
- 5. Perform 2 cuts to remove discharge riser A-PD-4.
Machine the final weld prep configuration on the' nozzle safe-end. (Cuts 25A and 26A.) .
- 6. Perform 2' cuts to remove discharge riser A-PD-5.
Machine the final weld prep configuration on the nozzle safe-end. (Cuts 27A and 28A.)
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- 7. Perform 2 cuts to sever the RHR discharge piping from the discharge riser. Machine the final weld prep on the remaining RHR spool. (Cuts 29A and 30A.)
- 8. Perform 4 cuts to segment the discharge ring header.
(Cuts 33A, 34A, 35A, and 36A.)
- 9. Perform 2 cuts to remove the remainder of the discharge riser above elbow A-PD-13. (Cuts 31A and 32A.)
- 10. Perform 1 cut to remove elbow A-PD-13 from the discharge isolation valve outlet. Machine the valve outlet to its final weld prep configuration. (Cut 37A.)
- 11. Perform 1 cut to sever the discharge isolation valve from spool A-PD-15. Machine the valve inlet to its final weld prep configuration. (Cut 38A.)
- 12. Perform 1 cut to sever pipe spool A-PD-15 from the pump discharge. Machine the final weld prep configuration on the pump discharge. (Cut 39A.)
- 13. Perform 9 cuts to sever various instrument and drain lines as needed to facilitate the removal of the discharge piping. (Cuts 40A, 41A, 42A, 43A, 44A, 45A, 46A, 47A, and 51A.)
3.2.2 SYSTEM REINSTALLATION The following methodology describes steps involved in replacing "A" loop piping utilizing the heat sink welding (HSW) method of stress improvement. Induction heating stress improvement (IHSI) is also being analyzed for use in lieu of HSW in places where
- critical path time savings can be achieved. If the IHSI method proves to be more advantageous, it will be substituted for HSW.
l Some specific sequences will be varied to accommodate ALARA
- requirements, machine repair and/or replacement, etc. The "B" loop piping replacement will be similar.
Notes on system reinstallation:
A. HSW will be performed with the inside diameter (ID) of
- the joint covered with standing water. A HSW plug will l
be used to contain the water within the pipe for local HSW.
B. .All joints (root) will be radiographically inspected (RT) prior to performing any HSW. Final radiographic inspection (RT) will be performed on completed welds.
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HNP-2 C. Suction and discharge piping will be replaced concurrently, with both loops "A" and "B" worked in parallel.
D. Refer to Figures 4 and 5 (pages 3-13 and 3-14) for piping loop layout and joint locations. -
3.2.2.1 Suction Pioing Reinstallation Sequence
- 1. Template and machine upper elbow PC-28AS-1.
- 2. Make root passes for joint 28AS-2.
- 3. In parallel with 1 and 2 above, weld (28AS-4) the 28 in. x 20 in. RHR tee PC-28AS-3 to riser spool PC-28AS-2 in the shop, using HSW procedures.
- 4. Make root passes for joint 28AS-3.
- 5. Complete joint 28AS-3 using HSW procedures.
- 6. Machine riser spool PC-28A-4.
- 7. Make root passes for joint 28AS-5.
- 8. Complete joint 28AS-5 using HSW procedures.
- 9. Machine riser spool PC-28A-5.
- 10. Make root passes for joint 28AS-6.
- 11. Complete joint 28AS-6 using HSW procedures.
(NOTE: Steps 12 through 15 may be worked in parallel with steps 1 through 11 above.)
- 12. Machine super long tangent elbow PC-28AS-7.
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- 13. Make root passes for joint PC-28AS-10.
- 14. Fit-up suction isolation valve to PC-28AS-7 and make root passes for joint 28AS-9.
(NOTE: HSW suction and discharge at-same time.)
- 16. Template and machine the closure elbow PC-28AS-6.
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- 18. Complete joint 28AS-8 using HSW procedures.
- 20. In parallel with 19 above, template and machine the RHR closure elbow / spool assembly. Joint 20-2 made in shop.
- 21. Make root passes for joints 20-1 and 20-3.
- 22. Template and machine RWCU closure piping.
- 23. Root weld the RWCU closure joints out to 2/3 complete using conventional welding procedures.
- 24. Verify completion of small bore piping welds.
- 26. Complete the RHR closure joint 20-1 and 20-3 using HSW procedures.
- 27. Complete the RWCU closure joints using HSW procedures.
Fill complete system simultaneously with discharge "B" loop.
- 28. Complete the upper 28-in. elbow joint 28AS-2.
- 29. Drain the system as required to RT the joints completed in steps 27 through 29.
3.2.2.2 Discharge Piping Reinstallation Sequence
- 1. Rig ring header (HDR) pieces PC-22AD-1, PC-22AD-2, PC-28AD-4 into position. (Weld RHR elbow onto PC-28AD-4 in shop.)
- 2. Make root passes for joint 22A-1.
- 3. Make root passes for joint 22A-2.
- 5. Template and machine 12-in. riser PD-12AD-H and 12-in.
end pieces PC-12AD-K1 and PC-12AD-F1.
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- 8. In parallel with 6 and 7 above, fit-up 12-in. riser PC-12AD-H and tack weld joints 12AH-1 and 12AH-2.
- 9. In parallel with 6 and 7 above, template and machine 12-in. risers PC-12AD-J and PC-12AD-G.
- 11. In parallel with 10 above, template and machine 12-in.
risers PC-AD-K2 and PC-AD-F2.
- 12. Fit-up 12-in. risers PC-12AD-K2 and PC-12AD-F2 and tack joints AK-3, AK-2, AF-3, and AF-2.
- 13. Make root passes on safe-end joints.
- 16. Weld safe-end joints out to 2/3 complete using conventional welding procedures.
- 17. Weld all HDR joints to 2/3 T.
- 18. Fill and complete HDR joints using HSW procedures.
- 19. Template and machine PC-28AD-3.
- 20. Make root passes for joint 28AD-5.
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- 21. Complete joint 28AD-5 using HSW pror.edures.
(NOTE: Steps 22 through 25 may be worked in
. parallel with steps 1 through 21 above.)
I i 22. Make root passes for joint 28AD-1.
- 24. Make root passes for joint 28AD-2.
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- 25. Complete joints 28AD-1 and 2 using HSW procedures. ,
Fill suction and discharge together.
- 26. Template and machine closure elbow PC-28AD-2.
- 27. Make root passes for joints AD-3 and AD-4.
- 28. Complete joints AD-3 and AD-4 using HSW procedures.
(NOTE: HSW suction and discharge at same time.)
- 29. In parallel with step 28, template and machine the short radius RHR closure elbow assembly.
- 31. Verify completion of all small bore piping welds.
- 33. Fill the discharge piping up to and including the discharge nozzles simultaneous with suction "B" loop.
- 34. Complete the 12-in. risers to safe-end joints using HSW techniques.
- 35. Drain the system as required to RT the joints completed in step 34.
3.3 ADVANCE PREPARATIONS AND JOB PREPLANNING The advance preparations and planning for this projec't is based on past experience from similar successful projects.
Major elements of the project and certain other activities that are an integral part of the project preparation are summarized below: _
- 1. Technical working specifications are developed-based on the applicable codes and design specifications.
- 2. Engineering methodology and concepts are developed to accomplish the work in a safe, efficient manner under an ALARA radiation exposure reduction program, and to provide a completed product that meets project specifications and quality control requirements.
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- 3. Project planning and overall project schedules are developed. Materials and craft manloading are determined from these schedules. ,
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{ 4. Quality control procedures, work process proceduras, 4 and detailed step-by-step work instructions are .-
prepared. ,
- 5. Special tools and equipment are designed and s
- i, manufactured. .
- 6. Long lead time special tools and equipment-are procured.
- 7. Craft training and procedure verification mockups are designed and manufactured. ,
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- 8. Radiation exposure estimate and reduction program is developed. ,
- 9. Welding procedures are developed and qualified. '
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- 10. A well defined. organization with adequate staffing '
e-levels is established to support the project.. 'T
- 11. Craft personnel are trained in special operations using mockups to assure the operations are performed correctly and efficiently, thereby reducing radiation ** '
exposure. '
Preplanning and scheduling are important criteria for project
- control in all aspects of the work. Various planning methods
, are utilized to assure that specific activities are completed within specified times.
i The onsite project planning staff utilizes-management tools such as plan-of-the-day meetings and field problemtreports to monitor and control preoutage activities such as software > development,
! purchasing material, mockup and tool fabrication, vendor and
! subcontractor identification, and special problems. 7 A D preliminary activities schedule is issued to assure /all required activities are completed prior to reactor shutdown: A project s schedule is developed with the input from engineeringi craft 9' supervision, project management, planning, subcontractors, sand vendors. The combination of this expertise provides;for a
,! detailed schedule from which the optimum critical schedule paths
( are determined. , ;r i
Adaptations of the Critical Path Method (CPM), Program Evaluation and Review Technique (PERT), and bar charts:are used by the planning staff.- ,
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- 3.4 Interference Removal, Restoration Testine .
Southern Company Services is responsible forf the design efforth to make provision for the removal and reinst'allation of 1
equipment and structures that interfere with the recirculation.
piping replacement. Interferences include, but are not limited to, pipe, valves, conduit, cable trays, structural steel, and heating, ventilation', and air-conditioning (HVAC) components.
Each interference is evaluated to determine if operability is required during the redirculation piping replacement outage.
Design drawings and bill of materials are provided to ensure the safety and integrity of temporary operating systems. ,
Interference analysis'altro, determines impact to the Evaluat' ion of the Hatch Nuclear Plant'rire Protection Program and identifies whether or no't saZety systems are involved. Upon 3 completion of the analysis, interference documents are prepared for use in implementing removal, restoration, and testing of interferences. These documents'. include written instructions for deenergization of circuits, removal and replacement of pipe, conduit, ductwork and HVAC components, reconnectien of' circuits, nondestructive examination of welds, cable testing, and l
functional testing. Before any interference is removed detailed as-builts will be obtained to assure an exact /,
replacement during reinstallation. Once installed, interferences will be thoroughly tested to ensure system ,
fe' integrity and proper operability. Sufficient documentation is maintained to account for all engineering aspects of recirculation piping replacement interferences. {
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1 HNP-2 4.0 COMPONENT DESIGN AND FABRICATION 4.1 HNP-2 EXISTING DESIGN The HNP-2 is a boiling water reactor BWR-4 rated for 2436 MWt power output. The function of the recirculation system is to circulate the required coolant through the reactor core. The system consists of two loops, external to the reactor vessel, each containing a pump, flow element, and two shutoff valves.
Twenty internal jet pumps, ten for each loop, located within the reactor vessel are also part of the recirculation system.
RHR suction and return piping connects to the recirculation piping. The RHR system removes residual heat generated by the core during reactor shutdowns. Short sections of the connecting
, RHR piping are made from type 304 austenitic stainless steel; the remeining RHR piping is made from carbon steel.
RWCU piping is connected to the RHR suction piping. The purpose of the RWCU system is to maintain high reactor water quality by i
removing fission products, corrosion products, and other soluble and insoluble impurities.
The design basis of the existing system is:
Design pressure, suction 1250 psig at 575*F (includes RHR suction and RWCU piping)
Design pressure, discharge 1450 psig at 575*F (includes RHR return piping)
Original recirculation system American Society of (and also connecting portions Mechanical Engineers j of the RHR and RWCU piping) (ASME) Boiler and i design code Pressure Vessel Code
! Section III, Class 1 l
Figures 2 and 3 depict the arrangement of the existing recirculation system.
Recirculation piping materials currently installed were purchased.in accordance with General Electric's (GE)
- specification, Recirculation Loop Piping. The material used in l the original design was type 304 austenitic stainl'ess steel.
l Design, fabrication, testing, and inspection were in accordance I with ASME Code,Section III. The fabricated piping met the requirements of ASME Code,Section III as of the actual date of purchase.
4-1
- . _. - .- - _ _ - - . .-. . -_=
i l
HNP-2 The RHR and RWCU piping materials currently installed were i purchased in accordance with the balance of plant (BOP) supplier's specification.
i 4.2 HNP-2 REPLACEMENT DESIGN
! i 4.2.1 MATERIAL SELECTION Replacement piping for the recirculation systen and stainless steel portions of the RHR and RWCU systems will be fabricated from GE type 316 nuclear grade (NG) stainless steel.
The GE type 316 NG is an austenitic stainless steel with a high
, resistance to intergranular stress corrosion cracking (IGSCC). in the BWR environment due to its low carbon content and the J addition of molybdenum. This material has a carbon content not
- exceeding .02 weight percent. In addition, the nitrogen content of this alloy is controlled to counterbalance the loss in
! strength due to the relatively low carbon content. Increased pitting and sensitization resistance is provided by the addition of molybdenum.
4.2.1.1 Laboratory Qualification J
Several years of intensive studies of the environmental performance of NG materials within the BWR has resulted in a '
data base demonstrating the adequacy of NG stainless steel as replacement piping (section 4.3, reference 1) for types 304 and
' 316 stainless steel.
Accelerated pipe testing (8 ppm Oz, 288'C, 1 umho/cm
~
water) of welded reference type 304 stainless steel for comparison to types 316, 316-L, 316 NG, 304 NG,~CF3, and 347 stainless steels indicated that resistance to IGSCC of the
, alloys not only increases as carbon content decreases, but also j as nitrogen and manganese contents increase.
l Within the NG specifications, factors of improved resistance _to IGSCC over type 304 performance.can be expected to be at least
. 50-100 based on actual environmental pipe test results. .
Accordingly, the replacement of type 304 stainless steel piping with type 304 or 316 NG should virtually eliminate IGSCC failures within the plant design life.
In addition to normal and high oxygen BWR environments, pipes were also tested under adverse water chemistry conditions simulating water contamination occasionally found in field -
reactors during chemical' transients. Faulted environmental conditions evaluated included 8 ppm Oz, 66*C, 180 ppm 4-2
HNP-2 l
chloride, and 188 C water containing 30 ppm chloride. These simulated chloride intrusion tests were more rigorous than field intrusions, as the maximum limit of actual BWR operation under similar conditions is up to 2 weeks per year at 0.2 ppm chloride. In all cases, the NG pipe performance was equivalent to or 1atter than the referred type 304 stainless steel material indicating additional margin against off-chemistry transgranular cracking.
For supplemental information on NG material performance, the following tests were run: sensitization, crack propagation, fatigue crack initiation, and transgranular failure. Specimens were tested in the weld sensitized and low temperature sensitized (LTS) material condition. The results of all tests in 288 C, high purity BWR demonstrate the superiority or equivalence of NG stainless steels to the other alternate alloys and verify the pipe test results showing the high degree of environmental cracking resistance of the NG materials as compared to type 304 stainless steel.
4.2.1.2 Composition and Kinetics The maximum carbon content of types 304 and 316 stainless steels is 0.08 weight percent. The low carbon content of type 304 NG and 316 NG is manifested in greatly reduced kinetics of sensitization -- one of the key causes of IGSCC in stainless steels. A second notable composition characteristic of type 316 NG is the specification of 0.060 to 0.100 weight percent nitrogen. Its purpose is the recovery of alloy strength that was diminished by the reduction of carbon. Most importantly, nitrogen appears to be innocuous with respect to the kinetics and thermodynamics of sensitization; no evidence has emerged indicating that grain boundary depletion of chromium results from formation of a chromium nitride or a more complex compound. L-grade stainless steels differ from NG only in the specification for NG of a nitrogen limit and a lower limiting t
carbon content of 0.02 weight percent. Hence, data presented subsequently for type 316 L is indicative of the properties of type 316 NG -- the latter, however, being even more superior.
l Note further, that both types 316 and 316 NG contain molybdenum, whereas type 304 has none.
(
An investigation of the time-temperature-sensitization curves for stainless steels reveals that the 316 L material is far less susceptible to sensitization in the heat affected' zone of a weld. About 150 minutes are available for cooling past 700*C t
within the heat-affected zone of a type 316 L weld. In general,
( this cooling time is longer than a normal air cooling time and, I therefore, the threat of weld sensitization of type 316 L is 4-3
HNP-2 essentially moot. The fundamental studies on sensitization kinetics (references 2-5) show that the low carbon content
(<.03 weight percent) and the presence of molybdenum are the cause of sensitization suppression. Additional evidence of molybdenum causing greatly reduced rates of sensitization is shown in table 1 (page 4-7). No grain boundary carbides were observed in type 316 after a 3-hour heat treatment at 650 C, .
whereas in molybdenum-free type 304 carbides were observed after 1/2 hour.
The lower carbon content of the NG material (0.02 vs 0.03 percent maximum carbon) provides further margin against sensitization.
4.2.1.3 Material Testing An intensive pipe testing program in nominal and off-chemistry environmental BWR conditions demonstrated that NG stainless
- steels show a significant improvement of IGSCC resi~ stance to types 304 and 316 stainless steel.
The results of all small specimen tests in BWR high purity water showed complete support for the use of NG stainless steels in the BWR environment.
4.2.1.4 Field Experience No IGSCC cracking incidents have been reported for types 316 NG, 304 NG, 316 L, and 304 L stainless steel piping used for several years in operating BWRs. These materials have accumulated over 85 years of operation with piping and safe-end components. (See table 2, page 4-8.)
4.2.1.5 Conclusion Nuclear grade stainless steel materials are highly resistant to IGSCC in a BWR environment. The materials provide additional margin against IGSCC when compared to the successfully applied type 304 L and 316 L piping materials. Nuclear grade materials also meet all code and regulatory requirements for use as piping in the BWR.
4.2.2 DESIGN CONDITIONS The following table lists the piping design pressures and temperatures:
4-4
= . - ..
i i
i I
HNP-2 i
Design Design Pressure Temperature (psig) ( F)
Recirculation pump suction .
1250 575 Recirculation pump discharge piping 1450 575 RHR suction piping 1250 575 RHR return piping 1450 575 i RWCU piping 1250 575 1
4.2.3 PIPING DESIGN IMPROVEMENTS Piping design improvements consist of:
, A. Use of induction bent pipe for the 12-in. risers to reduce the number of welds. This eliminates 20 welds.
B. Use of long tangent and super long tangent 28-in.
elbows to reduce'the number of welds and to increase the use of automatic inservice inspection (ISI) and i welding equipment. This eliminates 4 welds.
C. Deletion of end caps by using 12-in. induction bent pipe and a reducer on ends of the 22-in. headers to reduce the number of welds and to eliminate the crud
, traps formed by the end caps.
l D. Deletion of header sweepolets by using a header with extruded outlets. This eliminates 8 welds.
Preliminary calculations of the-stress indices for the extruded outlets show that the indices are lower than those for the original sweepolets.
E. Use of' a single forged piping fitting to replace the RER return tee, header cross and reducer. This eliminates 4 welds.
l F. Counterbore length of two times the pipe wall thickness l at all pipe weld preps to facilitate ISI. _
l G. A 2-in. minimum extension at the reducer ends and at l
the ends of elbows to facilitate ISI.
i H. To minimize problems associated with welding new pipe
, to existing equipment, such as pumps, valves, and I safe-ends, a wall thickness greater than the code required minimum wall will be provided.
4-5 ,
- , +rM .
9 p.
l HNP-2 I. Deletion of unnecessary branch connections, such as the riser pressure taps formerly used for low pressure coolant injection (LPCI) loop selection logic.
J. Mechanical polishing of the replacement pipe inside surfaces to minimize radiation buildup.
Figures 4 and 5 show the proposed recirculation replacement piping configuration. Also shown in Figures 4 and 5 are the locations where the RHR piping connects to the recirculation piping, and where the RWCU connects to the RHR piping. The stainless steel portions of the RHR and RWCU piping up to the first isolation valve will be replaced to the same configuration as the existing piping.
In summary, the recirculation piping was redesigned to eliminate as many fittings as possible and, in addition, induction bent pipe was utilized in order to reduce the number of welds. The replacement piping has 36 less circumferential welds than the existing pipe. The pipe material specified, GE type 316 NG stainless steel, has a chemical makeup specifically developed to minimize the possibility of IGSCC. To further minimize IGSCC, all of the piping material (field welds are not included) will be solution heat treated.
The physical location of all field welds has been selected to allow easier access for automated welding equipment and for subsequent ISI. This is expected to significantly reduce future inspection time and radiation exposures for inspections.
4.3 REFERENCES
- 2. K. Narita, Trans., Japan Institute of Metals, 4, 15 (1963).
- 3. J. D. Elen and A. Glas, J. Nucl. Mat., 34 182 (1970).
- 4. J. S. Armijo and H. S. Rosenbaum, J. Appl. Phys., 38, 2064 (1967).
- 5. G. J. Barnes, A. W. Aldag, and R. C. Jerner, J. Electrochem.
Soc., 119, 684 (1972).
4-6
l l
HNP-2 TABLE 1 OBSERVATIONS OF CARBIDE PRECIPITATION IN TYPES 304 AND 316 STAINLESS STEEL AS A FUNCTION OF HEAT TREATMENT TIME Heat Treatment Type 304 Type 316 Temp ( C) Time (hrs) 650 0.5 Carbides barely detectable None observed 650 1.0 Carbides clearly observed None observed 650 3.0 Carbides clearly observed None observed 650 6.0 Carbides clearly observed Carbides detectable 650 24.0 Large carbides clearly Small carbides observed clearly observed l
i 4-7 l
m - ,
~~ --- . _ _ _ _
HNP-2 l TABLE 2 EXISTING LOW CARBON STAINLESS STEEL IN OPERATING PLANTS Pipe Plant Piping Piping System Welds Diameter (in.) Installation Brunswick 1 Recire bypass 28 4 5-1976 Brunswick 2 Recire bypass 28 4 5-1976
'Hamaoka Recire bypass 40 9-12 12-1978 Dresden 1 Recirc bypass 30 4-8 5-1977 Dresden 1 Other 50 4-6 5-1977 Dresden 2 Recirc bypass 34 4 9-1976 Dresden 3 Recire bypass 34 4 9-1976 Fukushima 1 Recirc bypass 49 9-12 9-1978 Fukushima 2 Recire risers 60 9-12 12-1978 Fukushima 3 Recirc header 21 22 5-1982 Total of 56.9 reactor years exposure with no cracking to date.
Reactor Number of Years Plant Safe Ends Type of Safe End Material Safe Ends Exposure Hatch Unit 2 Recire inlet 304, 0.023%C 10 3.9 Peach Bottom Recire inlet 316 NG, 0.020%C 10 9.1 Unit 2 Peach Bottom Recire inlet 316 NG, 0.025%C 10 8.7 Unit 3 Fukushima Recire outlet 304 L 2 5.2 Unit 4 I Fukushima Recirc outlet 304 NG 2 3.8 ,
Unit 6 Total of 30.7 reactor years exposure with no cracking.
4-8
. . - - - - - - - .- .= . . - - -- . . - - . .
l l
i HNP-2 5.0 ANALYSES 5.1 SCOPE The following section summarizes the various analyses that will be performed to assure the safety of the reinstalled piping.
l 5.2 NUCLEAR CLASS 1 COMPONENT EVALUATION BASED ON AMERICAN SOCIETY OF MECHANICAL ENGINEERS CODE, SECTION III
- The piping and piping components will be analyzed in accordance with the rules of NB-3600 of American Society of Mechanical Engineers (ASME) Code, Boiler and Pressure Vessel Code,Section III, 1980 Edition through Winter 1980 Addenda.
The piping load combinations and acceptance criteria for service levels A, B, C, and D are given in table 3 (page 5-4). Those load combinations and acceptance criteria meet the limits of
- ENP-2 Final Safety Analysis Report (FSAR) Table 3.9-6.
l The primary loads considered are pressure, weight, and inertia effects of earthquake. All the loads classified as level A and B, including thermal expansion range, thermal gradients (lATt ,lATal), thermal gross discontinuity (laaTal - labTbl),
earthquake, and anchor movements, are considered in calculating primary plus secondary stress intensity range and peak stress intensity range. Stress indices in Table NB-3681(a)-1 shall be i used in qualification of most of the piping products and joints in this system. For piping products not covered by Table NE-3681(a)-1 stress indices will be established in accordance with NB-3681.
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.
i B. Primary stress intensity limits of. equation 9 in l NB-3652, 3654, 3655, and 3656 will be met for service levels A, B, C, and D.
C. Primary plus secondary stress intensity. range limits will be met for the service levels A and B pressure, temperature, and earthquake duty cycles defined in the design specification.
5-1
. _ .. ... ~ - , , - -
l l
! HNP-2 f
Primary plus secondary stress intensity range 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, the simplified elastic-plastic analysis equations 12 and 13 will be satisfied.
D. The cumuistive damage limits will be met for the service levels A and B pressure, temperature, and earthquake duty cycles defined in the design l specification. The cumulative damage will be evaluated l in accordance with NB-3653.6.
1 r
5.3 OTHER ANALYSES 5.3.1 PIPING STRESS ANALYSIS Due to the change in the piping configuration, the recirculation piping and connecting portions of the RHR piping will be reanalyzed. The recirculation piping and the RHR suction and return piping, between the tee connection and the drywell wall penetration, will be included in the analysis.
5.3.2 THERMAL EXPANSION ANALYSIS Three thermal conditions have been selected for analysis. These three thermal conditions will be used to simulate the service-levels A, B, C, and D thermal conditions.
5.3.3 SEISMIC ANALYSIS 4
The recirculation piping and connecting RHR piping 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 the piping contents plus insulation will be added to the weight of the pipe in the form of a uniformly distributed load (lbs/ft). The weight of pumps and valves will be included in the mass model. For the pump motors and valve operators, the extended mass will be modeled as ad additional weight at the respective center of gravity.
The stiffness of each support will be included in the piping mathematical model.
The earthquake analysis will be performed using the response spectra method described in HNP-2 FSAR section 3.7.
5-2
J HNP-2 In the operational basis earthquake analysis 0.5 percent of critical damping will be used and in the safe shutdown earthquake analysis 1 percent of critical damping will be used.
The modal responses and the X, Y, and Z directional responses will be combined as described in the HNP-2 ESAR.
The forces and moments due to earthquake differentia'. anchor movements will be determined by a static analysis with the movement at supports as input.
5.3.4 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 Code,Section III, Subsection NF. Due to the change in the piping configuration the existing hangers and snubbers will be reevaluated to the code to which they were purchased. If any new piping support hardware is required it will meet the materials, design fabrication and erection requirements of AISC
" Specification for Design, Fabrication, and Erection of Structural Steel for Buildings."
5.3.5 PIPE WHIP RESTRAINT DESIGN Where the piping configuration has not been changed and existing pipe whip restraints can be reused, the criteria for break location will be per the original design basis as discussed in HNP-2 FSAR section 3.6.2. If new whip restraints are required they will also be located per the original design basis.
1 l
l 5-3
l HNP-2 TABLE 3 LOADING COMBINATIONS Loading Service Stress ASME,Section III Combination Limit Limits NB-3600 References PD+DW+0BE Design 1.5 Sm NB-3652, eq. 9 Po+DW+TH+ Levels A & B 3.0 Sm NB-3653.1, eq. 10'**
OBE+ SAM +TG TH Levels A & B 3.0 Sm NB-3653.6(a), eq. 12 Po+DW+0BE+ Levels A & B 3.0 Sm NB-3653.6(b), eq. 13 SAM +TG Po+DW+TH+ Levels A & B U<1.0 NB-3653.2, eq. 11 OBE+ SAM +TG NB-3653.6 PO+DW+0BE Level B 1.8 Sm NB-3654.2, eq. 9 1.5 Sy PP+DW Level C 2.25 Sm NB-3655.2, eq. 9 1.80 Sy PO+DW+0BE Level C 2.25 Sm NB-3655.2, eq. 9 1.80 Sy PO+DW+SSE Level D 3.0 Sm NB-3656, eq. 9 2.0 Sy Table Legend PD - Design pressure PO - Service pressure PP - Peak pressure Po - Range of service pressure DW - Piping deadweight OBE - Operational basis earthquake SSE - Safe shutdown earthquake SAM - Seismic anchor movement (OBE)
TH - Loads due to thermal expansion of pipe TG - Thermal gradients i s 3
i Either equation 10 or equations 12 and 13 must be met.
5-4
HNP-2 .
6.0 RADIATION PROTECTION 6.1 RADIATION PROTECTIO!! PROGRAM PLAN Since the replacement operations will involve potentially significant radiological hecards, GPC has developed an independent radiation protection program plan (RPPP) that meets the specific needs of the recirculation piping removal and replacement operation. The primary thrust of the RPPP, summarized herein, is to provide an independent and operationally specific program, primarily using contractor personnel at all levels of operation, dedicated equipment, supplies and facilities, and GPC supervision for audit and/or liaison functions where such functions do not degrade existing programs. The basic goal of this plan is to maintain the exposures received by all personnel involved with the piping replacement operations as low as reasonably achievable (ALARA),
with a minimum of adverse impact on the existing manpower and facilities of the current GPC Health Physics Department (HPD) which will be required to handle routine operations at.HNP-1.
The primary goal of the RPPP is the implementation of effective measures to minimize and control the external and internal j exposures of individuals to radiation sources and radioactive l material. The realization of this goal requires the cooperation !
and coordination of three primary groups:
e GPC.
e NNI.
e HPPSG.
The HPPSG will ensu'e r that NNI personnel and all other outage support personnel will receive the necessary radiological information and support in order to perform all work in the drywell and other areas in a manner consistent with GPC procedures and practices and that GPC remains informed of all work conditions and control measures implemented. .
The RPPP encompasses all aspects of radiation protection including: organization, staffing, training, qualifications, engineering, ALARA, procedures, exposure, control, facilities, protective equipment, decontamination methodology, radioactive waste disposal, radioactive material control, quality control, and management oversight. The resultant plan, therefore, meets GPC's management commitment to implement and maintain an effective health physics and ALARA program during the removal and replacement of recirculation piping at HNP-2.
6-1
HNP-2 6.2 ADMINISTRATION 6.2.1 ORGANIZATION AND RESPONSIBILITIES The HPPSG is the organization which has been charged with the primary responsibility for implementing and maintaining the RPRP and maintaining radiation exposures ALARA. The HPPSG has been established as a separate and distinct functional organization which is designed to effectively interface with the GPC and HPD without adversely impacting on the ability of the existing HP program to meet the needs of routine operations and maintain exposures ALARA. The major responsibility of the HPPSG is to provide radiation protection direction, engineering, and radioactive material control during all operations involving the recirculation piping removal and replacement including all drywell operations and operations in auxiliary areas as applicable.
The organization consists of an HP manager, an assistant HP manager, and two HP supervisors in functional areas which reflect the existing HNP organization. Responsibilities of these two functional areas are organized to provide efficient radiation protection coverage and coordination of personnel. A description of primary functional responsibilities are listed below.
6.2.1.1 Health Physics Management The HP manager and the support staff are responsible for the overall management of the program including budget control, staffing, personnel exposure control, coordination with site and construction personnel, planning, direction, engineering, and control, including overall safety of project personnel. The project manager, or a designee, will also represent the group at meetings (including ALARA committee meetings), provide regulatory liaison, and ensure that GPC policies are adhered to by personnel assigned to the project.
6.2.1.2 Engineering Support The Engineering Support Group has the responsibility for providing evaluations and engineering controls to assure that operations in the drywell and auxiliary areas are carried out in a manner that maintains radiation exposures ALARA. Personnel assigned to 'this area are involved with procedure reviews, job briefings and debriefings, engineering controls (including shielding, containments, and ventilation), setting man-rem goals, tracking man-rem expenditures, exposure data reduction, and evaluating program effectiveness.
6-2
v HNP-2 6.2.1.3 Radiological Controls The Radiological Controls Group (RCG) has the responsibility of providing job coverage for all operations in the drywell and auxiliary areas that are involved with routine and nonroutine surveillance monitoring, radiological work permit (RWP) coverage, area access control, manning control points, decontamination supervision, radioactive material control, and contamination control. In addition, RCG has the responsibility of coordinating with the plant HPD which is responsible for providing the ancillary HP services necessary to support operations in the drywell and auxiliary areas and radioactive waste disposal.
The GPC manager, Chemistry and Radiation Protection, manages the HPD during this project and is responsible for providing ancillary support services to the HPPSG in the areas of logistical and technical aspects of dosimetry, routine training, main counting room activities, instrumentation, personnel decontamination, radioactive waste shipping, radioactive waste handling, protective clothing and equipment, and respiratory protection. In addition, he will assure compliance with site policies and will assist in resolving safety issues which may arise between the project construction personnel and the HPPSG.
6.2.2 INTERFACE A clear definition of management interfaces provides a system to assure efficient communications between management personnel.
The interfaces between the GPC project manager (PM), the HPPSG HP manager, the NNI PM, and the GPC manager, Chemistry and Radiation Protection are as follows:
A. The NNI site manager and the HPPSG HP manager-both report directly to the GPC DPM.
B. The HPPSG HP manager interfaces with the GPC manager, Chemistry and Radiation Protection to coordinate scheduling problems, to resolve conflicts, to expedite and support project activities.
C. The assistant. HPPSG HP manager interfaces with the GPC HP supervisor to coordinate scheduling of shift activities, to expedite problem solving, coordinate common responsibilities, and coordinate support activities.
D. The HPPSG HP supev. visor, Engineering Support interfaces with the NNI ALARA engineers to coordinate design reviews, training, and common responsibilities.
6-3
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6.2.3 MANAGEMENT OVERSIGHT The management oversight function will be accomplished by an ALARA Management Oversight Committee, including senior representatives of GPC and of project organizations. This committee will approve the project ALARA policy and review the initial, collective man-rem goals. During the course of the project this committee will review and audit the project ALARA program to evaluate program effectiveness, and will recommend improvements when applicable.
6.2.4 STAFFING AND QUALIFICATIONS 6.2.4.1 Staffing Staffing plans have been based on the HPPSG providing coverage for all work shifts. A teduced number of personnel will also be scheduled to work during the off hours in preparation for upcoming shift activities. This round-the-clock coverage will provide support to other activities required to be done so as not to interrupt drywell work, i.e., radiography operations.
Personnel in all job classifications will be divided between shifts except the HPPSG HP manager, his staff, the radiological engineer, and the engineering supervisor.
6.2.4.2 Qualifications Personnel assigned to the HPPSG will satisfy the minimum selection criteria as specified in American National Standards Institute (ANSI) N18.1, 1971. In addition, radiation protection technicians shall be qualified in accordance with the existing site qualification program. Project management will also grant final approval for each candidate, to ensure they have the skills, knowledge, experience, and attitude appropriate for j their specific assignment.
6.2.4.2.1 HP Manager The HPPSG HP manager will meet or exceed the requirements for radiation protection manager as specified in Nuclear Regulatory l Commission Regulatory Guide 1.8.
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6.2.4.2.2 HP Supervisors l The other HPPSG supervisors will satisfy the requirement for supervisors as specified in Section 4.3.2, ANSI 18.1.
6.2.4.2.3 ALARA Coordinators The HPPSG coordinator will satisfy the requirements as specified in Section 4.5.2, ANSI N18.1 and the additional requirement of at least one added year of power reactor HP experience.
6.2.4.2.4 Senior Technician The senior technicians will satisfy the requirements as specified in Section 4.5.2, ANSI N18.1.
6.2.4.2.5 Junior Technicians ,
Technicians not satisfying the senior technician criteria will be classified as junior technicians.
6.2.4.2.6 Engineers and Specialists Personnel classified as an engineer or specialist will satisfy the requirements as specified in Section 4.6.2, ANSI N18.1.
6.3 IMPLEMENTATION .
In order to coordinate implementation of the plan, the HPPSG staff will be responsible ror eight general functional areas:
drywell, BOP, control point, radiation work permits, decontamination, analytical support, procedures, and records distribution.
6.3.1 DRYWELL OPERATIONS Personnel assigned to the drywell operations will implement the I exposure control program for all work associated with the removal of pipe from the drywell. External exposure controls include special training, access control, dosimetry, posting, surveys, work surveillance, an RWP system, and other measures developed by engineering. The internal exposure control program contains many of the same features of, and is implemented
[ cencurrent with, the external exposure control program. In t
6-5
, ENP-2 addition to the above, the internal exposure control program will include respiratory protection measures and a bioassay program. The major prerequisite of an effective internal exposure control program is an effective contamination control l program.
6.3.2 BALANCE OF PLANT HP OPERATIONS I
Personnel assigned the BOP HP operations will implement the exposure control program for all work associated with the project conducted outside the drywell. Activities include radiography, repair of contaminated equipment, inspection of i
removed pipe, and work on isolation valves.
6.3.3 ACCESS CONTROL POINT Personnel assigned to the access control points will be i responsible for the smooth transfer of personnel into and out of the reactor building and/or RWP areas. These personnei ensure that only approved individuals enter and that all RWP requirements are met prior to entry.
6.3.4 RADIATION WORK PERMITS ,
Personnel assigned to providing RWP coverage will be responsible for obtaining and disseminating information concerning radiological conditions, protective clothing needs, respiratory protection requirements, and dosimetry needs. In addition, the RWP will be used to draw the workers attention to special procedures that must be employed or engineering controls that
- must be in place prior to commencement of work.
6.3.5 DECONTAMINATION .
Personnel assigned to decontamination will be responsible for decontaminating reactor components, piping, and auxiliary areas to minimize exposures and spreadable loose contamination. An ALARA evaluation will be made on each major task to determine the value of decontamination as a method to reduce exposures.
! Evaluations will include alternate methods for each i
decontamination effort including the use of hydrolazing and chemicals.
l In addition to major task related decontamination of the reactor and piping components, routine decontamination of areas,
. personnel, and equipment will be performed to control loose contamination, and to minimize the need for respiratory 6-6
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protective equipment. The drywell area will also be evaluated based on several alternatives: high-pressure waterblast, hand decontamination, and a continuous general decontamination program during the recirculation pipe change-out.
6.3.6 ANALYTICAL SUPPORT Personnel assigned to provide analytical support will be responsible for assuring rapid turn-around on survey wipes and air samples collected in areas where work is in progress and for routine (periodic) surveys. Such samples will usually consist
! of particulate filters and/or charcoal cartridges which will be routinely analyzed for alpha, beta, and gamma radiation.
Samples for which the screening process indicates a need will be transferred to the main counting facility for gamma spectrometry analysis.
6.3.7 PROCEDURES AND RECORDS DISTRIBUTION
[ Personnel assigned to control procedures and records distribution will be responsible for assuring that all activities are performed using existing or newly developed HNP procedures and for assuring that all required paperwork, records, and procedures are distributed expeditiously.
6.3.8 GPC-HNP SUPPORT t
The GPC-HNP support personnel will be responsible for general employee training, supply of clothing and respiratory protection equipment, supply of instruments, counting room, waste disposal, bioassay including whole body counting, decontamination, pro-cessing of contractor personnel, and packaging of waste for transportation.
6.4 ALARA l ,
6.4.1 POLICY GPC is committed to operating all activities at HNP in a canner that will not jeopardize employees or the public health and safety. Included in this commitment to both occupationally exposed personnel and the general public is the obligation to maintain radiation exposure at levels which are ALARA and which are in compliance with the NRC Regulation 10CFR2O and with the commitments in the Final Safety Analysis Report. To fulfill this obligation, the HPPSG will 6-7
. - = _ _ - . ._ - - - -_- -_ . _ . . -.
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I implement a radiation protection program which includes the '
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< provisions of Regulatory Guides 8.8 and 8.10 to ensure compliance with regulatory requirements and the ALARA objective.
- 6.4.2 RESPONSIBILITY !
i l i The goal of the radiation protection program shall be to i
- maintain individual and collective radiation doses to project personnel and the general public at ALARA levels through
! improved operational practices, procedures, and equipment.
Responsibility for approving the ALARA program resides with the GPC manager, Chemistry and Radiation Protection. Responsibility for developing and implementing the ALARA program resides with the HPPSG manager with primary support of the ALARA staff.
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6.4.3 PROGRAM DEVELOPMENT An ALARA program will be developed which will achieve the goal of meeting or exceeding regulatory requirements and guidance.
In addition, the program will provide the specific guidance necessary for program implementation and periodic review.
The ALARA program will include, as a minimum, a policy i statement, formal descriptions of responsibilities and i
authorities, management oversight, procedures, records, and data management.
6.4.4 PROCEDURES As part of the preplanning process operational, administrative, and engineering design ALARA procedures will be developed to support the program. Operational procedures will include, as a minimum, preplanning, man-rem estimates and tracking, cost benefit, and job debriefing. Administrative procedures will include goals, audits, program evaluations, and management oversight. Engineering procedures will include radiological considerations in design, task evaluation, and ALARA training.
+
a 6.4.5 IMPLEMENTATION Initial involvement will include involvement of the ALARA l Engineering Group in the development of plans, procedures, and l equipment. Shielding packages, prework decontamination, remote tooling, portable ventilation units, containment enclosures, and
, special training aids all will be considered. A major effort 1
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,n.. e e . - _ , , - - - . ---.y -g -. . , g,e, - - ,_w .~ ,m.- e ,-n 9 ,
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Will be expended to minimize the significant dose rates around
- the nozzles. Consultation with General Electric (GE) engineers concerning reactor shutdown conditions and shielding evaluations will be used to reduce dose rates.
~
1 Major radiation sources are being identified and their i contribution to man-rem expenditure for project evaluated. The major contributors to the project man-rem commitment identified
, to date include the contribution of the vessel internals to the i nozzle area dose rates, the contribution of the recirculation
- piping to the general area dose rates, the contribution of the t recirculation pumps and isolation valves to the general area '
dose rates on the 114 ft elevation, and the RWCU isolation valve contribution to the general area dose rates for removal of the-RWCU piping section. Techniques currently under evaluation to
, reduce the impact of the major radiation sources include the use
- of an invessel shadow shield located at the suction nozzles,
! water bladders placed in su'etion nozzle and discharge nozzle safe-ends after piping sections are removed, lead shot or lead l wool bags around all nozzle exteriors in the drywell, waterblast
! decontamination of recirculation piping internals prior to
, piping removal, and shielding of isolation valves and pumps that
! are in the vicinity of the work areas with lead blankets or lead
! sheets.
Man-rem estimates are under development and will include an estimate for the entire project, estimates by task and significant steps in each task, and weekly dose estimates.
These estimates will'be routinely updated as information is received. GPC will provide an estimate of the anticipated man-rem exposure by major task prior to initiation of the outage.
Evaluation of the survey data from previous outages and the recent shutdown in October 1983 indicate that the personnel ~will
- be exposed to the following range of dose rates during the -
1 project:
Estimated Percent of i Dose Rate (mrem /hr) Total Man-Hours
! < 25 20 i 25-100 59 l
100-1000 20 1000-1500 1 At this time, preliminary estimates indicate that the total job exposure approaches 1750 man-rem.
The ALARA-staff will maintain a continuous involvement in the progress of the recirculation piping replacement. Man-rem expenditures for each task and subtask will be monitored on a-daily basis and reported weekly to the ALARA committee. The .
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ALARA staff will review all RWP requests and written RWPs prior to issue and conduct periodic surveillances of work in progress to ensure the incorporation .of ALARA comments and that proposed engineering controls are effectively utilized. Periodic surveillances will also permit the ALARA staff to develop additional or improve existing exposure reduction methods.
At the completion of each major portion of the project and upon final completion of the recirculation piping replacement project, the ALARA staff will conduct debriefings to critique the conduct of work and compare man-rem estimates with actual man-rem expenditures. The debriefing will include specific recommendations for exposure reduction techniques on similar tasks in the future.
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] HNP-2 6.4.6 THE ALARA COMMITTEE At least weekly during early phases of the project and monthly thereafter, the HPPSG HP manager will convene a meeting of the ALARA committee. The membership of the committee shall include
~
as a minimum, the HPPSG assistant HP manager, who.shall serve as the chairman; the NNI ALARA engineer; the NNI construction s engineer; and the HPPSG Radiological Engineering. supervisor.
This committee will set goals for the maintenance =of exposures ALARA and evaluate the methods employed and the results achieved in meeting those goals.
6.4.7 DATA MANAGEMENT In order to properly preplan tasks, implement lesson's learned from previous tasks, and track man-rem totals, a historical data base n? task-specific personnel radiation exposures will be developed. This data base will contain sufficient information to allow the ALARA committee to review work planned prior to' the start of any major task so as to make recommendations for maintaining exposures ALARA, and to evaluate work in progress.or completed in order to compare actual exposures with establishe'd to goals.
, An HP/ALARA records management system will ' hie installed which:
will contain historical work-related exposure information. A historical record file suitable for incorporation into .'an r ,
automated radiological information management system will be established.
(
This data will allow evaluation of worker exposures incurred on project tasks to be categorized by type of workers, work group,,
and job function. Evaluating entry and exit times will allow - '
total man hv rs spent on particular tasks to be tabulated.
Exposure history will be collected by component,' system, and' ~, l ,
H work function.
P 6.4.8 EVALUATION OF PROGRAM EFFECTIVENESS '
t
! The ALARA staff will make frequent audits of' work in progress to monitor program effectiveness in reducing ~ exposures to ALARA levels. Audit results will be reported to HPPSG management to assure all ALARA procedures are being followed in all phases of operations. -In addition, the ALARA committee will review
, program data to determine results achieved in meeting ALARA
! goals. 3 N ! ;;
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' 7. 0 WASTE MANAGEMENT l Personnel assigned to waste management will be responsible for, i handling the significant quantities of solid low level l radioactive waste that will be generated. The waste will be composed of consumables necessary to support the operation (e.g., protective clothing, plastic sheeting, tape, etc.) and the piping and other related atructures that will be removed from the drywell areas. The consumables and small pieces of piping and related structures will be handled routinely following HNP procedures. The large pieces of piping, however, !
require special handling.
- Procedures will be developed and will include evaluation of 3 isotopic content, dose measurements on piping sections, dose to curie calculations, packing requirements, man-rem evaluations 4
for alternate methods, decontamination, determination of isotopic quantities, waste classification preparation f)r i shipment, and shipment.
i 4
After evaluation of' isotopic, content and dose rate meashrements are made, dose to curie calculations will be performed to determine waste classifications and to assure compliance.with 10CFR61, 49CFR173, and burial site license requirements.
Alternate procedures for waste disposal will be evaluated for man-rem estimates and cost / benefit to assure selection of the
~
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appropriate method. All shipments will be in approved packages
! and be coordinated with burial site personnel to assure ,
i' compliance with their license. conditions. It is expected that the isotopic and cupie determir.ations will require the piping be j transported in type A1 containers. Other waste generated from the project is to be processed as LSA. Based upon burial '
allotments and the classification of the waste, the first
- priority will be shipment-to Barnwell with Washington State as the backup. <
All waste handling procedures will be s b e'ct to routine quality -
control surveillances. Full health physics / coverage will be provided during all operations involving radioactive piping disposal activities. _
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HNP-2 8.0 PIPING INSULATION The piping insulation system to be used for the replacement recirculation piping system for HNP-2 will be Owens-Corning Fiberglass Corporation's Nukon or an approved equal with a separately applied 26 gauge (or heavier) American Society of Testing Materials A240 type 304 stainless steel jacketing.
Nukon insulation is comprised of quilted, light density, semi-rigid fiberous glass pad insulation, encapsulated in a woven glass cloth forming a composite blanket. Removable insulation sections will be provided to facilitate periodic inspection as required by American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Reactor Coolant Systems.
The Nukon insulation was chosen to replace the present reflection type insulation due to the improved thermal performance of the Nukon system. The Nuken system does not have the heat losses at joints between sections and conforms more easily to penetrations through the insulation such as pipe supports thus reducing heat loss at these points. The Nuclear Regulatory Commission has concluded, in Topical Report OCF-1 dated January 1979, that the Owens-Corning fiberglass insulation system is not expected to interfere with the operations of emergency recirculation cooling systems during a loss-of-coolant accident. Therefore, the Nukon type thermal insulation is acceptable for use in the containment.
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HNP-2 9.0 CODES AND STANDARDS The piping material, fabrication, and examination will be in accordance with American Society of Mechanical Engineers (ASME)
Code,Section III, 1980 Edition through Winter 1980 Addenda.
All components will be replaced in accordance with the requirements of ASME Code,Section XI, 1980 Edition through Winter 1980 Addenda. Specifically IWA-7000 and IWB-7000 will be applied to this replacement program.
Replacement piping will be manufactured in accordance with the requirements of ASME Code,Section III, Subsection NB. Shop fabrication will be done in accordance with ASME Code,Section III, Article NB-4000.
4 The piping will be installed in accordance with the requirements of NB-4000 except for welds to existing equipment not purchased to ASME Code,Section III which will meet the requirements of Section XI. Weld examinations will be done in accordance with the requirements of Article NB-5000 of ASME Code,Section III, except for welds to existing equipment or piping which shall be examined to the requirements of the original construction code.
The replaced piping will be hydrostatically tested in accordance with the requirements of Section XI prior to being placed in service. The replaced piping vill be analyzed in accordance t with the requirements of ASME Code,Section III, Article NB-3000.
The use of- the codes listed below is in compliance with ASME Code,Section XI, IWA-7210 which states that replacements may meet the original construction code or all or portions of later editions of the construction code. GPC has opted to use later
, editions of the code to replace the recirculation piping.
I A. ASME Boiler and-Pressure Vessel Code, 1980 Edition through Winter 1980 Addenda, l
l e Section II, Material Specifications; e Section III, Nuclear Power Plant Components; o Section IX, Welding Qualifications; l
o Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components - Division I.
B. Occupational Safety'and Health Administration 29CFR1910, General Industry Standards.
i j 9-1
1 HNP-2 C. Nuclear Regulatory Commission Code of Federal Regulations e 10CFR50, Licensing of Production and Utilization Facilities; o 10CFR21, Reporting Defect and Nonconformance to the Nuclear Regulatory Commission.
D. American National Standard Institute Standard N45.2, Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants, Reactor Plants and Their Maintenance.
E. American Welding Society A4.2-74, Standard Procedures for Calibrating Magnetic Instruments to Measure the Delta Ferrite Content of Austenitic Stainless Steel Weld Metal.
F. American Institute of Steel Construction, Manual of Steel Construction, 8th Edition.
G. ASME B31.1 Power Piping Code, Winter 1980 Addenda.
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HNP-2 10.0 INSPECTION A thorough inspection program will be undertaken to verify the integrity of the new recirculation piping. Nondestructive examination techniques including radiography, ultrasonics, magnetic particle, and liquid penetrant testing will be employed as required by the applicable codes. Testing will be performed by the installation contractor in accordance with an approved ,
quality control program. The testing will be further monitored by GPC quality control personnel. Hydrostatic testing will also be performed to verify the integrity of the piping system as required by the applicable codes.
Upon completion of all installation work, a baseline examination will be made of all new piping welds. Each weld will be prepped and a surface and volumetric examination performed. This work will be contracted through the Inservice Inspection Group of the SCS Mechanical Design Department. The baseline examination will be performed per American Society of Mechanical Engineers,Section XI, 1980 Edition through Winter 1980 Addenda, to the extent practical.
During startup testing the operating map will be reverified.
Piping thermal expansion and vibration tests will be performed in accordance with the General Electric specification, Preoperational and Startup Pipe Motion Test.
5
~10-1
l HNP-2 11.0 TRAINING 11.1 TRAINING PROGRAMS Extensive training programs will be utilized to minimize radiation exposure, reducesthe probability of accidents, and enhance the quality of the work. Any worker on site will be subject to the standard training required of all workers in his craft. In addition, special training will be conducted as detailed in the following sections.
11.2 MOCKUP TRAINING Based on past experience mockup use and personnel training is essential to a successful project. Mockups and training are used to:
A. Assure that craft personnel are qualified and experienced.
B. Avoid errors that may result in loss or damage to material.
C. Assure that specific operations are performed in a safe and efficient manner.
D. Reduce man-rem exposure by reducing the time required to perform the operation in a radiation environment.
E. Minimize generation of contaminated material.
F. Assure that the proper tools and equipment are available at the site to perform the required work.
- G. Assure that the work can be accomplished in a timely manner, on schedule.
See Table 4 (page 11-5) for a list of mockups that will be utilized and for training that will be accomplished.
11.3 WELDER AND WELDING OPERATOR TRAINING AND QUALIFICATION i Welders and welding operators will be qualified in accordance with ASME Code,Section IX as required by Sections III and XI.
Additional training and qualification (beyond the scope of Section IX) will be required for heat. sink and Inconel welding.
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i HNP-2 All welders and welding operators will be required to pass two manual welding qualifications. The first will be an open root groove weld in carbon steel pipe using the gas tungsten arc weld (GTAW) and shielded metal arc weld (SMAW) processes with carbon steel filler metal. This qualification will serve as an entrance test providdag an initial indication of the welder's abilities. It will also qualify all welders for structural steel welding. The second test will be a groove weld in pipe using the GTAW and SMAW processes with stainless steel filler metal and consumable insert. The welder will be qualified for non-heat sink stainless joints (such as socket welds), tacking consumable inserts using filler metal, and repairs to non-heat sink joints.
Training for automatic GTAW will consist of setup and familiarization with the equipment and instruction on proper operation. Several practice coupons will be welded to gain experience, followed by a qualification coupon. Coupons and filler metals for automatic GTAW training and qualification will be stainless steel to simulate actual plant situations.
Additional qualification will be required for heat sink welding. Qualification will consist of hands-on experience s followed by welding of a mockup coupon. Welding technique will simulate that to be used in production. The mockup coupon must pass a radiographic inspection (RT) before the welder or operator will be considered qualified to perform HSW (including repairs). Manual and automatic HSW will be qualified independently.
All welding operators who are to weld any Inconel encountered in the plant shall have proof of prior Inconel welding or train for this work. Training will consist of hands-on experience followed by the successful RT of a test coupon. Welders must also qualify for Inconel welding by successful RT of a test coupon.
Experienced welders and welding operators will receive minimal training and practice. All will be required to qualify in accordance with ASME Code,Section IX with additional qualifications for HSW.
11.4 HEALTH PHYSICS The HNP General Employee Training Program will be scheduled for i all project personnel. Where necessary, the Health Physics l Project Support Group (HPPSG) will design and conduct job l
specific training programs for select craft personnel. The training programs will be designed to compliment those programs currently being implemented by GPC at HNP, as well as meet all 11-2
l HNP-2 regulatory requirements and guidelines (10CFR19). The programs will not only stress an understanding and knowledge of the responsibilities, procedures, techniques, and philosophies i
needed by personnel to work in environments of potential radiological and industrial hazards, they will also stress the ALARA aspects of the work being performed. The HPPSG will also design a qualification program to ensure that all key personnel (senior technicians, foremen, etc.) are familiar with plant specific procedures and operations.
11.4.1 PROGRAM OBJECTIVES AND CONTENT The overall objective of project training will be to provide all project personnel with the knowledge of drywell systems and orientation, HP procedures currently in force at HNP, Nuclear Regulatory Commission rules and regulations, emergency plans, and ALARA principles to enable them to maintain their exposure to radiation at ALARA levels.
11.4.2 TASK SPECIFIC TRAINING All craft personnel who have been assigned to a task involving significant and/or unusual radiological hazards, will receive special job-specific training. This training will include, but not be limited to, indoctrination in the specific hazards of that work area or operation, appropriate special radiation protection procedures, ALARA considerations procedures, and the use of protective clothing and/or respirators. Classroom training in this regard will be supplemented by hands-on and on-the-job training whenever possible so that individuals may actually gain first hand experience of the pertinent concepts and/or equipment involved before being given the responsibility .
for independent actions. As job-specific program needs arise, field procedures will be developed with mockup utilization, use of photographs, and as-built drawings whenever possible.
11.4.3 PROGRAM IMPLEMENTATION All project training programs will be presented by qualified instructors who possess a level of knowledge of the principles involved far exceeding that of those individuals expected to receive that training. Appropriate lesson plans, or instructional guides, shall be developed and documented whenever necessary.
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. HNP-2 11.4.4 PROGRAM EVALUATION AND DOCUMENTATION Internal quality assurance surveillance will be provided in order to maintain the adequacy of the proposed training program. GPC will document and keep records on all individuals selected for training / retraining and provide written examinations in order to demonstrate personnel attainment of given standards and/or preestablished acceptance criteria.
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- TABLE 4 i
l LIST OF MOCKUPS Description Purpose
- 1. 28-in. suction nozzle Train craft installing shield-mockup ing in and around the reactor pressure vessel nozzle.
- 2. 1-in., 2-in., 6-in., Train craft on machine setup 12-in., 22-in., 24-in., and cutting of various size i and 28-in. pipe cutting piping using specialized mockups and commercial cutting /end preparation machines.'1'
- 3. Machining mockup for Train machine shop craft in straight pipe and elbows machining pipe to match NNI-templates.'1'
! 4. Quality control Train inspection personnel
- inspection mockup in dimensional verification techniques.
- 5. Pipe templating equipment Train pipefitting craft'on cor-J rect use of special templates.
1.
- 6. HSW plug mockup, 22-in. Verify water plug design.
and 28-in.
Qualify and train craft in
! installation of plug, filling and_ draining, and purging operations.
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- 7. Fit-up and handling fix- Train craft on correct ture, 12-in., 22-in., 24-in., attachment and use of the and 28-in. special fixtures.
- 8. " Fire Plug" fixture Train craft-on correct
- attachment and use of the special fixture.
Craft personnel will be trained on the following machines:
e NNI special purpose cutting machines.
e Standard Portsmuth design cutting machines.
11-5
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HNP-2 TABLE 4 (Continued)
Description Purpose 4
- 9. Welder qualification mockups Qualify welders.
- 10. 28-in. pump suction elbow Train craft on cutting mockup machine fit-up in restricted area.
- 11. 12-in. long radius elbow Train craft on 12-in.
mockup cutting machine fit-up.
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HNP-2 12.0 TEMPORARY FACILITIES 12.1 SCOPE As many of the normal systems that serve the drywell will be disabled due to interference removal, several temporary facilities will be required. . Temporary facilities will also be provided as required for housing necessary support functions.
The following section provides a description of these facilities.
12.2 DRYWELL HEATING, VENTILATION, AND AIR-CONDITIONING The HNP-2 drywell temperature is maintained during operation by four cooling units with two standby cooling units. Of these six units, three will be removed for recirculation piping replacement. The remaining three units will be utilized to cool the drywell during.the replacement work. Temporary routing of piping, ductwork, and wiring will be required and will be provided in interference removal packages. Piping and ductwork will consist of high-pressure hoses and flexible duct, respectively.
Drywell ventilation during recirculation piping replacement will be accomplished using portable air handlers with high efficiency particulate air filters. These air handlers will process the air from the drywell and exhaust clean air into the reactor building. Charcoal beds will be utilized if necessary to maintain acceptable release rates. Temporary wiring and ductwork will be provided to support these units.
12.3 HEALTH PHYSICS FACILITIES AND EQUIPMENT 12.3.1 STORAGE
),
Designated radioactive material control areas will be used to store the large sections of pipe and other radioactive materials awaiting shipping and disposal. These areas will be large l
enough to accommodate the disposal preparations and interim storage of radioactive contaminated recirculation piping removed from the drywell and for other contaminated equipment and material generated by recirculation piping removal and replacement activities.
In addition to the interim storage areas, temporary radioactive storage areas in the reactor building will be needed to assure prompt removal of radioactive materials and piping from the 12-1.
a
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HNP-2 occu' pied area of the drywell. Residence times in these temporary areas will be kept to a minimum.
12.3.2 PERSONNEL CONTROL A contractor access point and change room will be utilized to control personnel entering the reactor facility. The existing contractor control point will be enlarged and equipped to handle up to 300 people per shift. A second large area, for the drywell control point, will be large enough and equipped to handle up to 100 people per shift. Smaller, temporary control points will be set up for balance of plant operations as work activities dictate. Each control point will be supported by easily accessible personnel staging and change areas. Equipment needed to support control point operations include friskers, HP supplies and instruments, protective clothing, dosimetry, respiratory protection equipment, waste containers, phones, and pages.
12.3.3 COUNTING ROOM Plant counting room facilities will be used whenever applicable, but due to anticipated numbers of surveys to be conducted and samples taken, an area will be established as a project counting room for smears and air samples. Equipment needed to support this counting room may include gas flow proportional counters, thin window G-M counters, and check sources.
12.3.4 HEALTH PHYSICS An HP storage space will be provided as an area to-support project radiation protection activities. Equipment needed to i
support these activities may include respiratory protection l equipment, air monitoring equipment, area radiation monitors, _
portable radiation monitoring equipment, friskers, personnel dosimetry, and decontamination equipment.
12.3.5 DECONTAMINATION AREAS Personnel decontamination showers presently available at HNP will be utilized as necessary for personnel with extensive contamination situations. Tool and equipment decontamination l activities will also~ utilize existing HNP facilities.
12-2
HNP-2 12.3.6 TRAINING Training areas will be designated to provide job specific -
training. These areas will be large enough and equipped to handle mockups. See section 11.0 for more information on training.
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13.0 PROJECT SAFETY PROGRAM 13.1 SCOPE
- The contractor's Industrial Safety Program as described below has been reviewed and approved by the GPC.
The site safety representative will perform surveillance of the contractor's program, including monitoring of the regularly scheduled safety meetings.
13.2 NEWPORT NEWS INDUSTRIAL SAFETY PROGRAM An industrial safety program will be administered and supervised i
in accordance with the NNI Safety Program to supplement and
- support GPC's existing safety policies and procedures. NNI's safety program is designed to comply with all applicable ,
occupational safety and health standards promulgated under the i
1970 Williams-Steiger Occupational Safety and Health Act (OSHA) and will meet or exceed all state and local government requirements. To stress safety, NNI will conduct biweekly i safety meetings to be attended by all construction personnel.
1 The dates and times along with the agenda of tha meeting will be made known to the HNP representative at least 2% hours in advance.
l i Examples of specific elements in the safety program include, but are not limited to, the following:
i l A. Maintaining proper accident and illness records.
B. Accident investigation and loss of time from accident l control.
C. Inspections to ensure that requirements are met for proper fire protection, ventilation, housekeeping, illumination, and sanitation.
D. Equipment specification review and inspection of power-activated equipment for proper safeguards and operation under the federal standards.
E. Inspection to' ensure that personal protective l equipment, when required, is in serviceable condition l and worn properly.
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HNP-2 F. Survey and evaluation of noise and vibration producing equipment with engineering and administrative controls ,
recommended for the elimination of control of hazardous noise.
G. Personnel monitoring, evaluation, and control of dust, fumes, gases, and vapor produced in various work environments.
H. Inspections to ensure that all hazardous material is handled and stored properly, and its management is adequately documented.
I. Regularly scheduled safety meetings held at onsite locations. Typical topics of meetings include the following:
e Initial safety orientation for new employees.
- Introduction to OSHA.
e Personnel safety equipment.
e First aid.
o Fire prevention safety.
e Ventilation, air supply safety.
e Radiological safety.
e On-the-job safety.
e Accident reporting.
e Responsibility for accident prevention.
e Combustible and toxic substance.
e Job safety analysis, o Electrical / welding safety.
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i HNP-2 i
I 14.0 TECHNICAL SPECIFICATIONS l
The performance of the piping replacement work will affect certain safety systems that are presently covered by the HNP-2, j, Technical Specifications. A request for a Technical Specifica-tion amendment was submitted to the Nuclear Regulatory Commission by letter dated October 20, 1983 for these systems.
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HNP-2 15.0 SAFETY CONSIDERATIONS 15.1 SAFETY EVALUATION Replacement of the recirculation piping and the nonisolable portions of the RHR and RWCU piping will not result in any unreviewed safety questions as defined in 10CFR50.59(a)(2). It is anticipated that the plant can be operated in accordance with the current operating license with only minor modifications to the existing Technical Specifications.
No impact on the acciden evaluations contained in chapter 15 of the Final Safety Analysis Report is anticipated as a result of the proposed replacement because the recirculation system break areas are not affected by the recirculation pipe replacement.
An analysis will be documented which evaluates the hydraulic effects of the changes to the recirculation system external risers and ring header. This document will show that the piping configuration changes produce only a negligible change in the total flow characteristics of the recirculation system. This will result in insignificant changes in the jet pump flow distribution pattern which will not noticeably change the core flow distribution.
Piping stress calculations described in section 5.0 will demonstrate that the stress levels are within the ASME Code,Section III allowable stresses.
No changes in operating parameters are anticipated as a result of these repairs. Replacement of existing type 304 material with low carbon GE 316 nuclear grade in accordance with NUREG 0313, Rev. 1 can be expected to measurably enhance plant safety. Adequate evidence exists to conclude that the possibility of intergranular stress corrosion cracking during the life of the plant will be greatly reduced as a result'of the proposed modification. -
15.2 IMPACT ASSESSMENT OF DISABLING THE HNP-2 DRYWELL It is anticipated that with the fuel in the fuel storage pool and gates in place, only a small portion of the drywell's equipment may be required to be operable during the pipe replacement operation.
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HNP-2 Components / systems that will be disabled include:
A. All emergency core cooling systems. All the wide range level and drywell pressure instruments will be disabled. This will preclude unwanted signals resulting from inadvertent or planned disablin'g of the instrumentation. Additionally, pumps and valves will be locked out (i.e., breakers will be racked out under administrative controls) to prevent unwanted operational conditions that could result from instrument calibration or other maintenance occurring in parallel with the pipe replacement.
B. The closed cooling water system seal purge connections to the recirculation pump and motor will be isolated.
C. The portion of the RWCU that connects to the stainless steel section of the RWCU piping which will be replaced will also be isolated. Steps will be taken through procedural controls to assure that reactor vessel water quality is maintained throughout the RPRP.
D. The neutron monitoring system (source range monitor, intermediate range monitor, local power range monitor and traversing incore probe) will be disabled and the instrument cables / conduit and tip tubes may be removed for access, as necessary. The fail-safe scram signal is jumpered out (i.e., neutron flux scram is disabled).
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