ML20023D407
| ML20023D407 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 05/31/1983 |
| From: | ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT |
| To: | |
| Shared Package | |
| ML20023D404 | List: |
| References | |
| NUDOCS 8305200463 | |
| Download: ML20023D407 (166) | |
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PAGE REPLACEMENT GUIDE FOR AMENDMENT 77 CLINCH RIVER BREEDER REACTOR PLANT PRELIMINARY SAFETY ANALYSIS REPORT (DOCKET N0. 50-537)
Transmitted herein is Amendment 77 to Clinch River Breeder Reactor Plant Preliminary Safety Analysis Report, Docket 50-537. Amendment g
's) 77 consists of new and replacement pages for the PSAR text and Responses to NRC Questions.
Vertical margin lines on the right hand side of the page are used to identify changes resulting from NRC Questions and margin lines on the left hand side 'are used to identify new or changed design information.
The following attached sheets list Amendment 77 pages and instructions for their incorporation into the Preliminary Safety Analysis Report.
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\J 8305200463 830520 DR ADOCK 0500053
AMENDMENT 77 PAGE REPLACEMENT GUIDE REMOVE THESE PAGES INSERT THESE PAGES 1
Chapter 1 1.4-24, 25 1.4-24, 35 1.4-28, 29 1.4-28, 29 1.5-22, 23 1.5-22, 23 Chapter 3 3.1-63, 63a 3.1-63, 63a 3.2-8, 9 3.2-8, 9 3.2-10, 10a, 10b 3.2-10, 10a, 10b 3.2-11, lla thru 11e 3.2-11, 11a thru lie 3.2-14, 14a 3.2-14, 14a 3.8-9, 9a, 10 3.8-9, 9a, 9b, 10 3.8-18a, 19, 20, 21, 21a 3.8-18a, 19, 20, 21, 21a 3.8-39, 40 3.8-39, 40 3.8-B.14a, 15, 15a, 16 3.8-B.14a, 15, 15a, 16 kkh' kkh'
! 4.2-624, 625 4.2-624, 625 4.4-3, 4 4.4-3, 4 Chapter 5 5.6-35g, 35h 5.6-359, 35h Chapter 7 7.1-3, 3a, 4 7.1-3, 3a, 4 7.1-7, 7a 7.1-7, 7a 7.5-32 7.5-32 Chapter 9 9.1-25, 26 9.1-25, 26 9.1-65a, 65b 9.1-65a, 65b 9.13-1, 2 9.13-1, 2 A
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REMOVE THESE PAGES INSERT THESE PAGES Chapter 10 10.4-7, 8, 8a 10.4-7, 7a, 8, 8a Chapter 13 13.3-4, 5, Sa, 6 thru 10 13.3-4, 5, Sa, 6 thru 10 13.3A-3, 4 13.3A-3, 4 Chapter 15 15.1-107 thru 110 15.1-107, 107a, 108, 109, 110 15.4-33, 34 15.4-33, 34 Chapter 17 17A-1, ii 17A-1, 11 17A-1 thru 8 17A-1 thru 8 17A-15, 16 17A-15, 16 17A-25, 26 17A-25, 26 17A-57, 61 17A-57, 61 17A-64, 65, 65a. 65b 17A-64, 65, 65a, 65b 17A-68d, 68e, 68f 17A-68d, 68e, 68f 17A-75 thru 78 17A-75 thru 78
/~N 17C-1, ii, iii 17C-1, 11, iii k 17C-1 thru 4 17C-1 thru 4 17C-7 thru 20 17C-7 thru 20 17C-44, 44a 17C-44, 44a 17C-45, 46 17C-45, 46 17C-49, 50 17C-49, 50 17C-56 thru 61 17C-56 thru 61 171-41, 42 171-41, 42 APPENDIX A A-v A-v A-1, 2 A-1, la, 2 A-209, 210 A-209, 210 A-293 A-293, 294 A-295 thru 308 O 8
1 AMENDMENT 77 QUESTION / RESPONSE SUPPLEMENT This Question / Response Supplement contains an Amendment 77 tab divider to be inserted following Qi page Amndment 76, March 1983. Page Qi Amendment 77 is to be inserted following the Amendment 77 tab divider.
The following Question / Response replacement page is for the NEW NRC QUESTION SERIES RECEIVED SINCE THE FALL OF 1981 and should be inserted behind the appropriate numbered tab.
REPLACEMENT PAGE REMOVE THIS PAGE INSERT THIS PAGE QCS430.30-1 QCS430.30-1 O
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Oualltv Assurance Division l Assistant Director for Oualltv Assurance l Qualification requirements for the Assistant Director for Quality Assurance is contained in Appendix A, Section 1.4.1 of Chapter 17.
Procurement Division Assistant Director for Procurement The qualification requirements include a broad knowledge of laws and regulations applicable to Government contracting, procurement, property 1 management, and traffic managefrent functions. He must have a minimum of five years experience in contract administration and negotiations involving supply, construction, engineering, and R&D contracts and a Bachelor's degree with emphasis in business related subjects such as economics, business administration, accounting, law, and public administration.
Profect Control Division Chief of Project Control Minimum qualification requirements include ten (10) years experience in the !
Installation, operation, and maintenance of management control systems for research and development projects or programs. This experience should cover cost and schedule controls of contractors; financial controls; contracts;
/3 analysis of reports; and interfaces with ERDA Headquarters. A Bachelor's V degree in Business Administration or Engineering is required.
Financial Management Division Chief. Financial Management The qualification requirements include a knowledge of the theories underlying general accounting, industrial cost accounting, construction accounting, and government fiscal accounting sufficient to advise and assist contractors in the establishment and maintenance of accounting systems.
i Knowledge of auditing principles and practices adequate to plan and direct a l program of examinations of the financial transactions and business practices !
of contractors.
i Knowledge of the principles, theories and techniques of budget administration .
and analysis required in budget preparation and review of actions proposed or l taken in the day-to-day execution of the budget.
To meet these requirements, an individual would normally have a unive; sity degree (accounting major) and 15 years experience in government, inoustry and publIc accounting.
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1.4.4.2 Westinghouse - ARD - LRM Organization There are no specific qualification requirenents identified f or the management positions in the Westingh'ouse - ARD Organization except as defined in Appendices D and H of Chapter 17. However, the capability of their personnel is demonstrated by the experience and qualifications summarized in the folicwing paragraphs.
Over 400 ARD professionals are working directly on CRBRP. Approximately 100 of these are in management positions.
Essentially 100% of all prof essionals involved in the Project have Bachelor's degrees and approximately 40% have advanced degrees. The average professional has over seven years experience in LMFBR related work. Approximately 50% of the managers have advanced degrees and the average manager has approximately 12 years of experience in LMFBR related work.
The Bachelors and advanced degrees held by the prof essionals blanket the following fields:
Chemistry Chmical Engineering Civil Engineering E l ectr i ca l /E l ectron i cs Eng i neer i ng Industrial /Manuf acturIng Engineering Mechanical Engineering Materials Engineering Nuclear Engineering Physics 7 ARD utilizes consultants and specialists from other Westinghouse divisions whose background and experience are required for independent design reviews, ASME code expertise, manuf acturing engineering, metallurgical problems, stress / thermal / inelastic / structural analysis, and saf ety related activities.
1.4.4.3 Rockwel! Internatlonal CorooratIon There are no specific qualification requirenents identified for specific management positions at Atmics International (AI), a division of the Energy 52 Systems Group of Rockwell International Corporation except as def ined in Appendix J of Chapter 17. However, the capability of their personnel is demonstrated by the experience and qualifications summarized in the following paragraphs.
Amend. 52 Oct. 1979 1.4-25
s Oualltv Assurance Manager l
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The minimum requirements f or the Quality Assurance Manager are shown in Section 17E 1.4.1.
Protect Office Resident Manager l A minimum of 5 years of progressive responsibility for the management and supervision of technical ef forts with primary emphasis in nuclear technology.
l He must have at least a Bachelor of Science degree or equivalent experience and education and training in nuclear reactor technology.
1.4.4.6 Stone and webster Engineering Corooration Specific qualification requirements at Stone and Webster Engineering Corporation are identified for key positions identified and described in Section 1.4.2.5.6. For elI the qualification requirements, in Ileu of a degree, equivalent qualifications may be substituted based on other educational accompt ishments, experience in related f f elds and technical achievements, such as holding a license as a Professional Engineer or Certification as a Quality or RellabilITy Engineer by the American Society for Qual Ity Control .
CRBRP Senior Protect Manager A minimum of ten years of progressive responsibilitles in the supervision and management of various phases of engineering, construction, and/or quality Cj N assurance ef forts is required, with primary emphasis in the nuclear power pl ant f iel d. He must have a working knowledge of the Corporation's resources and also have a Bachelor of Science or Arts degree with additional and/or training in power plant technology.
CRBRP Deoutv Director of Construction A minimum of ten years of progressive responsibilities in the supervision and management of heavy construction projects, with emphasis on the construction of power and/or process f acilities. He must have a working knowledge of the Corporation's resources and also have a Bachelor of Science or Arts degree.
CRBRP Profect Manager A minimum of ten years of progressive responsibility in the management and supervision of technical ef forts is essential, with emphasis in the nuclear power pl ant f ield. He must have a Bachelor of Science or Arts degree, with additional education and training in management and power plant technology.
sj Amend. 63 1.4-28 Dec. 1981
CRBRP Project Oualltv Assurance Manager A minimum of ten years in quality assurance and related f101ds including manufacturing, construction, and/or installation activities. At least two years of this experience shall be associated with the nuclear field in either field or headquarters project quality assurance assignments. He must have a Bachelor of Science or Arts degree.
Senior Site Construction Reoresentative A minimum of five years in responsible assignments in field engineering and construction activities, with emphasis in the construction of power and/or process facilities. He must have a Bachelor of Science or Arts degree.
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1.4-29 Amend. 77 May 1983
1.5.1.1.4 criteria of Success The latch Component Test in Sodlum has been completed. The inconel gripper /
Inconel coupling head performed in accord with specifications for 4 times the required number of cycles. The components are considered acceptable.
The prototype system test will confirm the capability to function reliably throughout its design lIfe.
1.5.1.2 Direct Heat Removal 1.5.1.2.1 Purnose The Direct Heat Removal (DHR) service provides a supplementary means for removing long term decay heat for the remote case when none of the steam generator decay heat removal paths are available. This system must be able to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable f uel design limits and the design conditions of the reactor coolant boundary are not exceeded. This supplementary decay heat removal is performed by a cooling system incorporated into the sodium make-up/ overflow system with plant conditions as specified in Chapter 5 (Section 5.6.2). The principal uncertainty of the make-up/ overflow cooling system is short circuiting of the make-up flow with the reactor vessel. Short circuiting would occur if the Inlet fluid flows directly to the overflow line without cooling the reactor core. Tests are needed to design the system to ensure short circuiting does not compromise core cooling.
1.5.1.2.2 Procram This program is conducted by Hanford Engineering Development Laboratory at the Integral Reactor Flow Model Test Facility. A 1/21 scale outlet plenum model test was used initially to test promising OHR candidate designs for the outlet plenum. Of concern is the location of the make-up and overflow nozzles to reduce short circuiting of make-up fIow. This test wilI conceptually determine overflow nozzle locations.
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1.5-22 March 1983 L
Confirmation testing of the selected make-up and overflow concept was 53 successfully completed in the Phase I testing of the Integral Reactor Flow Model.
- 1. 5.1.2 . 3 Schedule 41 I 5
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CY 74 75 76 77 78 79 80 51 Make-up/0verflow Location Tests in 1/21 Scale Outlet 41l Plenum Model ,conpleted Confirmation Tests and Evaluation of Make-up/0verflow 4)l 53l Concept , ,completpd 411 1.5.1.2.4 Success Criteria The tests demonstrated that the distribution of the make-up flow 26 53 in the outlet plenum was adequate to assure that the DHR service will function to remove decay heat following a reactor shutdown. This system must be capable of removing heat at a rate such that specified acceptable fuel design limits and design conditions of the reactor coolant boundary are not exceeded.
531 1.5.1.2.5 Results o_f Tests 53l The 1/21 scale outlet plenum and the HEDL IRFM model tests have been completed. The results show that short circuiting of make-up flow to the over-53l flow nozzle is limited to approximately 5%. The test and results are discussed in more detail in Response 001.580.
57 l 1. 5.1.3 Blanket Failure Threshold 41 1.5.1.3.1 Purpose The CRBRP is. being designed to operate with a limited number of failed 41 fuel and blanket rods. This requires demonstration that.operat. ion _with fa.iled 57l blanket rods exposed to sodium does not r.esult in rod-to-rod failure propagation. This program investigates the potential of blanket material /
sodium reaction to cause swelling, flow blockages, and rod-to-rod failure 57l propagation in blanket assemblies.
O Amend. 57 1.5-23 Nov. 1980
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anticipated operational occurrence. Suf ficient holdup capacity shall be provided f or retention of gaseous and liquid of fluents containing radioactive materials, particularly where unf avorable site environmental conditions can be expected to impose unusual eperational limitations upon the release of such ef fl uents to the environment.
Resoonse:
The CRBRP design incorporates in its liquid processing system a division of the radioactive waste streams into two categories: an intermediate radioactivity level waste stream, and a low level radioactivity waste stream.
The normal operation of the intermediate radioactive liquid waste stream is such that the processed IIquids are not released from the plant but are recycled f or reuse. The low level radioactive waste stream will discharge to a diluent stream only a design estimate of approximately 1 pCi of activity per year excepting tritium which is expected to be 3 MCI / year.
The gaseous radioactive release from the CRBRP will be processed through the Radioactive Argon Processing System (RAPS) and the Cell Atmosphere Processing System (CAPS). These two systems are subsystems of the Inert Gas Receiving and Processing System (See Section 9.5).
The RAPS exhaust is recycled with no direct discharge to the environment. The CAPS maintains the cell atmospheres at acceptable levels. The exhaust release
,_s rate from this system is designed at 50 SCFM exhausting to the RCD HVAC N sy stem.
(V Other gaseous ef fluents f rom the CRBRP will be exhausted through the normal HVAC systems and the CRBRP design is such that activities are expected to be
<<10CFR20 limits.
The Solid Radioactive Waste System is designed to handle compactible, non-compactible and solidification of liquid wastes with cement or concrete.
Sultable weather protected f acilities are designed to prevent any release of activity to the environment during on site storage. Department of Transportation approved containers will be utilized to transport solid radioactive waste for eventual long term disposal et licensed locations.
The releases of radioactive materials f rom the CRBRP are discussed separately in Section 11.2, Liquid radioactive releases, Section 11.3, Gaseous radioactive releases, and Section 11.5 Solid radioactive releases.
Criterion 53 - Fuel Storage and Handling and Radioactivity Control The f uel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal operation, including anticipated operational occurrences, and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to saf ety, (2) with suitable shielding f or radiation protection, (3) with 7-s, appropriate contai nment, conf inement, and f iltering systems, (4) with a residual heat removal capabil ity having reliabil ity and testabil ity that 3
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Amend. 76 3.1-63 tlarch 1983
reflects the importance to safety of decay heat and other residual heat runoval, and (5) to prevent significant reduction in f uel storage coolant inventory under accident conditions. The f uel handling and its interf acing systems shall be designed to minimize the potential fcr f uel management errors that could result in f uel rod f ailure, i
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Amend. 76 3.1-63a March 1983
TABLE 3.2-1 SEISMIC CATEGORY I STRUCTURES IIII4)
- 1. Containment Building (2)(3)
- 2. Confinement Structure (2) l l
- 3. Reactor g vice Area of the Reactor Service Building I 4. Control Building (2) l 5. Steam Generator Building (2) l Sa. Electrical Equipment Building (2) l 6. Diesel-Generator Building (2)
- 7. Emergency Cooling Tower Basin
- 8. Diesel Fuel Storage Tank Foundation
- 9. Electric Manholes l 10. Class IE Electrical Duct Banks (Berled)
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(l' includes foundation support materials (Class A backfill, and concrete fill).
(2)
Includes Seismic Category 1 internal structures, safety-related radiation shielding, access hatches, airlocks, doors, and rainwater scuppers which are integral parts of the building.
' includes mechanical penetrations and the Reactor Cavity Vent System which are Integral parts of the Reactor Containment Building.
Features of the plant buildings and site necessary to provide drainage of intense local precipitation to prevent flooding of safety-related I equipment and carnponents will be subject to the appropriate provisions of design and construction Quality Assurance Pregrams.
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3.2-8 Amend. 77 May 1983
TABLE 3.2-2 PRELlMINARY LIST OF SEISMIC CATEGORY l EGANICAL SYSTEM i COMPONENTS AND ASSIGNED SAFETY Q. ASSES (3)
Safoty Quality Components Class III Group (III Location I2' Reactor Vessel & Primary Heat Transport System Reactor Vessel & Closure Heed 1 A RG Primary Sodlum Pump 1 A RG Intermediate Heat Exchanger (IHX) 1 A RG Piping 1 A RG Reactor Guard Vessel 2 B RG Pump and IHX Guard Vessels 2 B RG Upper Reactor Vessel internals 1 A RG Lower Reactor Vessel internals 1 A RG Fuel, Blanket and Control Subassembly Structures 1 A RG Primary Control Rod Drive Mechanisms Structures 1 A RG Secondary Control Rod Drive Mechanism Structures 1 A RG Auxillary Liquid Metal System Primary Sodium Overflow Vessel 1 A RG Primary Sodium Makeup Pumps 1 A RG Overflow and Primary Sodium Makeup Piping and Valves (6) 1 A RG Overflow Heat Exchanger 1 A RG Airbiast Heat Exchangers 2 B RSB EVST Sodium and NaK Forced Convection Loop Components, Piping and Valves 2 B RSB EYST Natural Convection Soditrn Loop Canponents and Piping 2 B RSB EVST Natural Convection NaK Loop Components, Valve, and Piping 3 C RSB Natural Draf t Heat Exchanger 3 C RSB Primary Loop Drain Line (6) 1 A RG Primary Cold Traps (7) 3 C RG In-Containment Pri Na Storage Vessel 3 C RG Ex-Cont. Pri Na Storage Vessel 3 C SGB EVST Na & NaK Drain Piping (8) 3 C RSB PHTS Draln L Ines (9) 2 B RCB IHTS Na Processing System 3 C SGB EVST Cold Trap 3 C RSB Accumulator Packages for Pneumatically 3 C RG, RSB Operated Vaives O
3.2-9 Amend. 74 Dec. 1982
i TABLE 3.2-2 (Continued)
PRELIMINARY LIST OF SElSMIC CATEGORY I EWANICAL SYSTEM COMPONENTS AND ASSIGNED SAFETY CLASSES I3 Saf e ty Quality Group UU IN Location (2)
Camponents Class Steam Generator System Evaporators 2 B SGB Superheaters 2 B SGB Steam Drums + Recirc Pumps 3 C SGB l
Sodium-Water Reaction Pressure Relief
- Systems (Internal to steam gen. bldg,) 3 C SGB i IHTS Na Dump Tank 3 C SGB SWRPLS Rupture Disk Assembiles (4) 2 B SGB S.G. Water and Steam Camponents, Piping and Volves 3 C SGB -
l Leak Detection Lines to isolation Valve 2 B SGB Steen Generator Auxiliary Heat Removal System Air-Cooled Condensers 3 C SGB s Auxillary Feodwater Pumps (w/o motor drives) 3 C SGB Protected Water Storage Tank (PWST) 2 B SGB Connecting Piping & Valves (Extending from PWST to and including the First Valve) 2 B SGB Turbine Drive 3 C SGB Connection Piping and Valves (except piping from PWST to and including the first volve) 3 C SGB l Exhaust Restrictors 3 C SGB Contalrunent Isolatton Valves (Within their associated fluid systems) 2 B RG, IB Containment Cleanup System Note (12) -
RSB Containment Annulus Air Cooling System Note (12) -
RSB Contalreent Annulus FIItratton System 3 C RSS Refueling System Ex-Vessei Storage Tank (EVST) 2 B RSB EVST Guard Vessel 3 C RSB EVTM Containment Pressure Boundary 3 C RSB Reactor Rotating Guide Tub 6 Assembly 1 A RW l
3.2-10 Amend. 74 Dec. 1982
TABLE 3.2-2 (Continued)
PRELIMINARY LIST OF SEISMIC CATEGORY I EWANICAL SYSTEM COMPONENTS AND ASSIGNEC SAFETY CLASSES I3)
Safety Quality IIII Component s Class III Group Location (2)
Inert Gas Receiving and Processing System '
Pricory Cover Gas Llaes (Recycle Argon) 2 B RG Equalization 1.ine Betwoon Reactor Vessel Primary Pump and Overflow Vessel 2 B RG RAPS Components (Except Ccapressors RAPS Large LN 2 Dewers, and Recycle Argon Storage Vessels) 3 C RG Recycle Argon Storage Vessels 2 B RG CAPS Ccmponen1s (Except Compressors) 3 C RG l Nuciear Isiand HVAC See Note 10 Emergency Plant Service Water System (5) 3 C SGB,DGB, Emergency Cooling Tower, l Yard (buried)
Emergency Chilled Water System (5) 3 C SGB, G , DGB
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RSB, RG Auxillary Mechanical Systems for Diesel Generators 3 C DGB Fuel Oil Storage and Transfer System including:
l Diesel Fuel Oil Storage Tanks See Note 13 YARD Fuel 01i Transfer Pumps 3 C DGB Fuel Oil Day Tanks 3 C DGB Cooling Water System including:
Water Expansion Tank 3 C DGB Jackets Cooling Heat Exchanger 3 C DGB Water Temperature Regulating Valve 3 C DGB Starting Air System including:
Air Storage Tanks 3 C DGB Lubrication System including:
LubrIcaticg 011 Heat Exchanger 3 C DGB Lubs Oil Filters and Stralners 3 C DGB O
3.2-10a Amend. 77 May 1983
TABLE 3.2-2 (Continued)
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\s / PRELIMINARY LIST OF SEISMIC CATEGORY I MECHANICAL SYSTEM COMPONENTS AND ASSIGNED SAFETY CLASSES I3)
Safety Quality III IIII
-Conponents Cl ass Group Location (2)
Recirculating Gas Cooling System 3 C RSB, RCB (Subsystems Serving: Na makeup pump cold trap pipeways, Na makeup pump and vessels, EVS pump and cold trap, EVS pumps and pipeways)
Non-Sodium Fire Protection System Piping and Valves Connecting to the 3 C OGB Emergency Plant Service Water System Notes:
(1) Saf ety Cl asses are def ined in Sections 3.2.2.1 through 3.2.2.3 (2) RCB - Reactor Containment Building IB - Intermediate Bay of the SGB
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y SGB - Steam Generator Building RSB - Reactor Service BulIdIng CB - Control Building DGB - Diesel Generator Building (3) All components will be seismically qualified by analysis unless otherwise noted; motors are included with the mechanical components they drive.
(4) The SWRPRS rupture disc assemblies will be seismically qualified by analysis based on rupture data obtained during dynamic testing.
(5) Control per.el attached to chillers will be qualif ied by test.
(6) Out to Second Isolation Valve (7) Within Dual isolation Valves (8) Downstream of Isolation Valve (9) Downstream of -Second Isolation Valve l (10) identif ication of saf ety-related ventilation and filtration equipment is provided in Tables 9.6-1, 9.6-4, 9.6-5, 9.6-6 and 9.6-8.
(11) Based on Regulatory Guide 1.26, as interpreted for an LMFBR (12) The containment annulus cooling system and containment cleanup system shall meet the safety class 3 requirements. However, these systems are provided for the mitigation of an accident beyond the design basis. Therefore, they are not classified as SC-3, but will be built to ASME Ill/3.
(13) Quality assurance is in accordance with Reg Guide 1.137 A
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3.2-10b Amend. 77 May 1983
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s s TABLE 3.2-3
- PRELIMINARY LIST OF SYSTEM COMPONENTS CLASSIFIED IE Building Electric Power System Motor Control Conters Unit Substations 125V DC Distribution Panel Battery Charges inverters Vital Reg. Transformers Batteries 4.16 KV Switchgear Diesel Generator Sets Diesel Generator XFMR and Resistors Fuel Oil Transfer Pumps Na Pump Drive Breaker Trips Connectors and Terminations Cables, Cable Trays, and Conduit Standby Lighting Panels Emergency Chil led Water System isolation Valve Operators Pressure Control Valve Operators Pressure Relief Valves Emergency Chilled Water Pump Control and instrumentation l Cables, Cable Trays, and Conduit
- The equipment contained in this list are generically identified. The specific, detailed listing of all Class IE equipment, along with the environmental qualification program to which they are subjected, is provided in Ref erence 13 of PSAR Section 1.6. Class 1E systems, raceways, battery racks, containment electrical penetrations and their supports are seismic category 1.
3.2-11 Amend. 77 May 1983
TABLE 3.2-3 (Continued)
Heating, Ventilating, and Air Conditioning System Unit Coolers Emergency Chillers Supply Fans Exhaust Fans Air Handling Units Control Room Fiiter Fans Control Room Filter Units Control Room HVAC Monitors Control Room Air Concitioning Units Aux. Building HX Air Conditioning Units Control and Instrumentation l Cables, Cable Tray, and Conduit O Reactor Containment System RG TFBDB instrumentation Panels Containment Instrumentation Panels l Cables, Cable Trays, and Conduit Recirculating Gas Cooling System Local Control Panels Solenoid Valves Moisture Switches Liquid Level Switches Cold Trap Flow Switches O
3.2-11a Amend. 77 May 1983
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TABLE 3.2-3 (Continued)
Cold Trap Temperature Switches Pump Temp. Switch Fan Motor l Cables, Cable Trays, and Conduit Nuclear Island General Purpose Maintenance Equipment System Containment isolation Valve Operators l Cables, Cable Trays, and Conduit Steam Generator Auxiliary Heat Removal System Auxiilary Feedwater Pump Motors Auxillary Feedwater Flow Meter Protected Air Cooled Condenser Condensate Flow Meter Auxillary Feedwater Valve Operators Vent Control Valve Operators I
j Water Storage Tank Fill Valve Operator AFW Turbine Steam Supply Valve Operator AFW Turbine Pressure Control Valve Operator l Cables, Cable Trays, and Conduit Steam Generator System Feedwater Valve Operators Superheater Outlet Valve Operators Steam Drum Drain Valve Operators l Cables, Cable Trays, and Conduit Sodium-Water Reaction Pressure Relief System Instrumentation and Controls f%
O 3.2-11b Amend. 77 May 1983
TABLE 3.2-3 (Continued)
Reactor Heat Transport Instrumentaiicn System Reactor inlet Pressure Transmitter Reactor Vessel Na Level Transmitter Na FIow Sensor PRI IHX Outlet Na Temp Sensor Signai Conditioning PHTS Pony Motors Primary Pump Tachometer instrument Racks Instrumentation Reaction Products Dump Line Pressure Switches IHTS PM F1cw Sensor Superheater and Steam Drum Vent Control Valve Operator Temperature / Leak Detection Instrumentation Steam Flow Meter Superheated Steam Ternperature Sensor Superheater Steam Pressure Sensor Feedwater Flow Meter Feedwater Flow Meter Feedwater Temp. Sensor Evap. Outlet Sodium Temp. Sensors Cables, Cable Trays, and Conduit O
3.2-11c Amend. 77 May 1983
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Emergency P! ant Service Water System Pump Motors l ECT Fans ,
! Makeup Pump Motors Temp. Control Valve Operators Temp. Transmitters Temp. Indicator Controllors
! Pressure Dif ferential Switches Level Switches i
j l Cables, Cable Trays, and Conduir Auxiilary Liquid Metal System Valve Operators OverfIow Thermocoupies Local Panels Pump Motors Control Cabinets Capacitor Cabinets Transformer Cabinets EVST Thermocouples Control Room Panels l Cables, Cable Trays, and Conduit EVST Cell Termperature Thermocouples O :
3.2-11 d Amend. 77 May 1983
TABLE 3.2-3 (Continued) l Inert Gas Receiving and Processing Systems Containment Isolation Valve Operators Expansion Tank Equalization Valve Operator l Cables, Cable Trays, and Conduit P1 ant Control System Main Control Panel SCRAM Breaker Cubicle l Cables, Cable Trays, and Conduit Reactor and Vessel Instrumentation System Reactor Coolant Operating Level Instrumentation l Cables, Cable Trays, and Conduit Fl ux Monitor ing System Flux Monitoring Instrumentation Cabinets Instrument Drawers l Cables, Cable Trays and Conduit Piant Protection System Containment Isolation System Cabinets Reactor Shutdown System Cabinets Plant Protection System Cabinets Cables, Cable Trays, and Conduit O
3.2-11 e Amend. 77 May 1983
TABLE 3.2-5 FRELIMINARY LIST OF ASME CONSTRUCTION CODES, CODE CLASSIFICATIONS, AND CODE CASE FCR SEISMIC CATEGORY I MECHANICAL SYSTEM COMPONENTS
- Code / Code Ciass Code Case Cnmnonent Edition / Addenda Revision Reactor Vessel & Primary Heat Transport System Reactor Vesse1 (1) ASME-1II/1 1521-1,1592-1,1593-0, 1974/ Winter '74 1594-1,1595-1,15 % -1, 1682,1690 Closure Head (1) ASME-1II/1 1521-1,1592-4,1593-1, 1974/ Winter '74 1682,1690 Primary Sodium Pump Casing (1) ASME 111/1 1521-1,1592-1,1593, 1974/ Winter '74 1594,1595-1,15 % -1, 1682 Intermediate Heat Exchangers, ASME 111/1 1521-1,1592-1,1593, lHX (Tubes and Shell) 1974/ Summer '74 1594-1,1595,15 % -1 Primary Piping ASME lil/1 1592-7,1593-1,1594-1, r3 1974/ Summer '75 1595-1,1596-1,1644-4
'] Reactor Guard Vessel (1) ASME 111/2 (2) 1521-1,1592-4,1593-1, 1974/ Summer '75 1594-1,1682-1 Pump and lHX Guard Vessels ASME 111/2 1592-4,1593-1,1594-1 1974/ Summer '74 Upper Reactor Vessel ASME lil/1 1592-12,1593-1, Internals 1977/ Summer '77 1594-1 o Class 1 Appurtenances ASME III/1 1974/ Winter '74 Lower Reactor Vessel ASfE 111/1 Internals o Horizontal BaffIe Assembiy 1974/ Winter '76 1592-11 o Bypass Flow Module 1974/ Winter '76 1592-11 o Lower inlet Module 1974/ Winter '76 1592-10 )
o Fixed Radial Shield 1977/ Winter '77 N-47-16 )
- Appl icable to Saf ety Class 1, 2 or 3 components.
Component supports f or Saf ety Class 1, 2, or 3 components meet ASME, Section lil, Subsection NF requirements as discussed in PSAR Sections 3.9.1.6 and 3.9.2.6.
m ,
X 3.2-14 Amend. 77 May 1983
TABLE 3.2-5 (Cont. )
PRELIMINARY L IST OF ASE CONSTRUCTION CODES, CODE Q. ASSIFICATIONS, AND CODE CASES FOR SEISMIC CATEGORY I EOiANICAL SYSTEM COMPONENTS Code / Code Class Code Case Comoonent Edition / Addenda Revision - _ _
o Core Former Structure 1974/ Winter '76 1592-11 o Core Support Structure 1974/Sunmer '75 1592-7,1593,1594 Primary Control Rod Drive ASE 111/1 1592-3,1593,1594, <
Mechanisms Structures 1974/ Summer '74 1595-1,1596-1 Secondary Cont ml Rod Drlye ASE 111/1 1592-7,1593-1,1594-1, Mechanism Structures 1974/ Summer '75 1595-1,1596-1 Fuel, Blanket and Control See note (3)
Subassembly Structures Auxillary Liquid and Metal System Primary Sodium Overflow Vessel ASE lil/1 1592-7,1593-1,1594-1, 1974/ Summer '76 1595-1 Primary Sodium Makeup ASE 111/1 1592-10,1593-1, Pumps 1974/ Summer '76 1594-1,1595-1, 1607-1,1685-1 EVST Sodium and NaK Pumps ASE III/2 (2) 1592-10,1593-1, 1974/ Summer '76 1594-1,1595-1, 1607-1,1685-1 IHTS Sodlun Processing Pump ASE lil/3 1592-10, 1593-1, 1974/ Summer '76 1594-1, 1595-1, 1607-1, 1685-1 OverfIow and Primary Sodium ASE III/1 1592-7,1593-1,1594-1, Makeup Piping and Fittings 1974/Sunmar '75 1595-1 EVST Sodium and NaK Forced ASE III/2 None Convection Loop Piping 1974/ Summer '76 and Fittings EVST Natural Convection Sodium ASME lil/2 None Loop Piping and Fittings 1974/ Summer '76 EVST Natural ConyectIon NaK ASE III/3 None Loop Piping and Fittings 1974/ Summer '76 0
3.2-14a Amend. 74 Dec. 1982
l Cells in which there exists a pctential for spillage of thermally hot sodium 7-~s
() will be provided with steel cell liners with insulation and cell venting features to protect the structural concrete. The liners will be designed to contain a spill of high temperature sodium once in its lifetime. For liner strain critoria under sodium spill conditions see Appendix 3.8-B.
l The reactor cavity housing the Reactor Vessel, guard vessel, and Reactor Vessel Support Ledge is located near the center of the containment system.
Three primary heat transport system cells form an annulus around the reactor cavity on three sides. The reactor overflow vessel and primary sodlum storage vessel cell surrounds the fourth side.
The main work area inside the containment vessel is the operating floor above the reactor. Since the main work area is designed for continuous occupancy, the concrete slab for the operating floor will have suf ficient thickness to meet the structural and shielding requirements. Each cell is designed for the accident conditions and will also be designed to limit accident effects. When interior structures interact with the containment vessel, appropriate inter-active ef fects will be included in the containment vessel analysis. Thermal growth of a cell may be inhibited by neighboring elements. In such cases appropriate allowance will be made in the design to provide for ,the restrain-ing of thermal loads. Under seismic loads, the structure will act as an assemblage of shear walls. The structure will be checked to insure that the shear walls are adequate to sustain the lateral forces from seismic and other loads.
Removable slabs and plugs will be provided in areas where access is required
(,)
N,_ for operation and/or maintenance purposes. Suf ficient thickness will be provided for these slabs or plugs in order to meet the structural as well as radiation shielding requirements.
All interior structures within containment such as walls, slabs, steel framing and the El&C cubicles above the operating floor will be designed as seismic Category I structures. Therefore, no failure of structures within the containment will result from a SSE.
3.8.3.1.1 Reactor Cavity The reactor cavity is a hollow concrete cylinder closed at the bottom with a concrete slab. At the top of the cavity, a steel ledge, partially embedded in the cavity wall, is provided to support the reactor vessel. The support ledge will be designed such that it will withstand, in addition to the design basis loads, loads based upon Structural Margins Beyond the Design Base (See Ref erence 10a, PSAR Section 1.6) . The reactor cavity wall has a 40 foot internal diameter with a thickness of 7 feet. The interior of the cavity is lined with carbor steel plates. See RCB gas in Section 1.2 for arrangement detall s of the Reactor Cavity.
A ~
\ 1 V
3.8-9 Amend. 77 May 1983
Tne Reactor Vessel Support Ledge (RVSL) is a stael structure supported by, and embedded in, the Reactor Cavity as shown in Figure 3.8-9.
A systen of horizontal ring plates, radial brackets and stiffeners, and vertical cylindrical panels welded together, comprise the Resctor Vessel Ledge.
The two upper ring plates, at Elevotion 800'-7 1/4" and at Elevation 795'-1-1/4", receive the loads directly from the Reactor Vessel Support System.
The upper support ring plate receives the downward load by bearing. The lower support ring plate takes the upward loads through holddown bolts. The holddown bolts transfer the upward load to the lower support ring plate. Each holddown bolt is enclosed in a carbon steel sleeve. There are 69 holddown bolts on a circle having a 13'-1 1/4" radius.
The base plate (El. 780'-7 1/4") is embedded in concrete and is also anchored in the concrete by anchor bolts.
The exterior cylipaer (R=20'), together with the radial brackets and stiffeners, transfer the vertical loads from the support plates to the base plate. The two inner cyIinders (at R=12'-0 1/2" and at R=13'-11 1/2")
transfer loads and contribute to stif fen the intersecting plates by forming box sections.
The radial brackets continue outwards (beyond the exterior cylinder at R=20')
to constitute the radial stiffeners. These radial stiffeners are embedded in the cavity concrete and act as shear keys for horizontal load,.
l Thus the vertical loads, upward and downward, are first taken by the upper and l
lower plates respectively. These are then transferred to the base plate by (1) the radial plates and (2) the exterior vertical cylinder. The downward vertical loads are transferred to the concrete by the base plate by bearing.
! The upward vertical loads are transferred by tension in the base plate anchor bolts and shear in the concrete above the base plate.
The horizontal loads (seismic) are transfe. red through the upper horizontal plates, and the radial brackets to the stif fener plates embedded in the concrete acting as shear keys. The load is transferred to the concrete by bearing agalnst the stiffener plates.
A 1/2 inch radial gap is provided between the exterior cylinder and the concrete, to allow for free thermal expansion of the ledge under DBA conditions.
In the areas of the cut-outs where Primary Heat Transport System pipes penetrate the cavity, the cross section of the ledge is shown in Figure 3.8-8.
There are three such cut-out areas, one at each penetration.
O' 3.8-9a Amend. 77 I May 1983 f 1
O The top of the is enclosed by
. side with its reactor cavity is also the floor of the Head Access Area, which 6'-6" thick walls on three sides and a 4'-0" wall on the south inside dimension being a 44'-0" square. Figure 3.8-9 (sheet 2 of 2) presents the typical reinforcing steel in the reactor cavity at the Reactor Vessel Support Ledge.
l The temperature of the cavity structural concrete under DBA conditions will be limited to those required by the ASE Code Section 111, Division ll under paragraph CC 3440.
The ledge design will also be evaluated against a Structural Margin Beyond Design Base (SEDB) load. The 52 08 load is defined in PSAR Section 3.8.3.3.4.
The analysis and design of the Reactor Vessel Support Ledge is described in Section 3.8.3.4.8. A detailed discussion of Thermal Margins Beyond the Design Base (TEDB) in the CRBRP is provided in Reference 10b of Section 1.6. Post accident intercommunication between the Reactor Cavity (RC) and the Reactor Containment Building above the operating floor is provided by two separate and Isolable vent paths. The vents are designed to relieve at a dif ferential pressure of 16 psid between the reactor cavity and the RG atmosphere. A normally open motor operated valve is provided to isolate the reactor cavity in the unlikely event of failure of the rupture disc during normal plant operation or a minor accident.
A reactor cavity liner venting system is provided for relieving pressure from O behind the Iiner caused by steam and carbon dioxide (CO concrete of the reactor cavity and PHTS pipeway cells. 2)heT released f rom the Reactor Cavity and PHTS Pipeway Cell liner venting system is divided into several independent and physically separated zones which are vented to non-Inerted spaces either above the RG operating floor (elevation 816'-0") or below the operating floor.
4 O
3.8-9b Amend. 77 May 1983
3.8.3.1.2 Head Access Area (HAA)
The head access area is located below the operating floor level and above the reactor cav ity. The access area is of a square shape 44 feet long on each side and 14 feet high above the reactor head. The head access area is a reinforced concrete structure. Steel framing wilI be provided in this area to support the EVTM operations.
3.8.3.1.3 Primarv Heat Transoort System (PHTS) Cell Each PHTS cell is a step type rectangular reinforced concrete structure. At its widest section, the cell is approximately 48 feet wide by 72 feet long.
The celi is approximately 58 feet deep in the area housing the primary sodium pump and intermediate heat exchanger. The Interior surf aces of the cel is are provided with cell liners designed to contain hot sodium spills. The cells are designed to withstand accident pressure and temperature conditions as noted in Table 3.8-2.
3.8.3.1.4 Reactor Overflow Vessel and Primary Sodium Storage Vessel Cell This cell is a step type rectangular reinforced concrete structure. The cell is approximately 26 feet wide by 69 feet long with 62 feet height at its deepest section. The interior surface of the cell is lined with carbon steel plate similar to the PHTS cells. The cell is designed to withstand accident pressure and temperature conditions noted in Table 3.8-2.
3.8.3.1.5 Other Cells These cells are reinforced concrete structures with various sizes. The cells ,
required to maintain a nitrogen atmosphere during operatior.s wilI be lIned and l designed to the requirements noted in Table 3.8-2. Cell liners are described in : action 3A.8.
3.8.3.1.6 FIII SIab A structure fill slab of suitable thickness will be provided over the bottom containment Iiner plate.
3.8.3.2 AoolIcable Codes. Standards and SoectfIcations 3.8.3.2.1 Design Codes Applicable provisions both mandatory and recommended of the foilowing codes will be used in the design of the Internal structures:
3.8-10 O
Amend. 69 July 1982
(N Since the walls, ceilings and floors of each cell are considered as two-way
( slabs, the applied loads, used in the analysis, will be proportioned to the one-foot wide strips, in orthogonal directions, according to the ratio of their relative stif f nesses.
The cell design will be verified by using a three dimensional finite-element analysis with the computer progran NASTRAN. T:.e cell and adjacent structures will be represented in the mathematical model which will include the interaction with the containment shell and the exterior concrete wall. The appropriate loads and load combinations will be used in the analysis. 33 Further detailed analysis will be performed in areas of load concentration and penetrations as noted in 3.8.3.4.3. The reactor cavity is treated as a hollow cylinder for structural analysis. When in-house computer programs are used, their correctness will be verified against acceptable published programs. All vertical loads will be transferred to the foundation mat by three principal structural elanents, viz (a) walls of PHTS cells (b) perimeter wall around containment and (c) reacior cavity.
3.8.3.4.2 Analysis for Seismic Loads Equivalent static seismic loads as developed from the dynamic analysis of the structure will be transferred through the horizontal slab diaphragms and vertical shear walls to the foundation mat. The details of seismic analysis are described in Section 3.7.
O) 3.8.3.4.3 Analvsis for Ooeninas Structural analysis will be performed around openings in walls and slabs particularly where concentrated loads from thermal effects are induced. The 31 design will account for all the stresses in those areas and proper 28 reinforcement will be provided for the relief of such stress concentration.
3.8.3.4.4 Liner Analvsis The liner-anchor system will be designed and analyzed in accordance with the requirements and criteria specified in paragraph 3.0 of PSAR Appendix 3.8-B.
37 Liner analysis is discussed in PSAR Section 3A.8.3.5.
3.8-18a Amend. 61 Sept. 1981
i i
3.8.3.4.5 RadiatIqn Generated Heat Effect Adequate heat removal capacity will be provided for the reactor cavity so that the radiation generated heat does aot cause temperatures of the structural materials in excess of the ASME Code, Section ill, Division 2 requirements.
Where radioactive piping penetrates the concrete walls, adequate insulation or heat removal measures will be provided to control the temperature of structural materials wIthin Code Iimits.
Radiation generated heat wilt be produced as a function of the position of the Reactor core with respect to the structures and the temperature distributions will be calculated. In PHTS cells, significant heat will be generated from other sources such as piping. In alI such Instances, adequate cooling capacity will be provided to limit the temperature of the structural materials.
3.8.3.4.6 Reinforcement Design The reinforcing steel will be proportioned to meet the requirements of ACl-349. The bond and anchorage requirement of ACl-349 will be observed, since the interior structures pr.imarily provide a confinement function. The reinforcement for each wall or slab will principally consist of a set of orthogonal bars on each f ace with additional reinforcement provided in areas of penetrations or load concentration.
3.8.3.4.7 Structural Steel Design l The structural steel components other than the celi IIner and catch pan systems will be designed to the requirements of AISC specifications as identi f ied i n Section 3.8.2.2.1. When steel parts are stressed into the plastic range, an energy absorption check will be performed as described in Section 3.8.3.5.2 to assure the functional integrity. The cell liner and catch pan design criteria are given in Appendices 3.8-B and 3.8-C respectively.
3.8.3.4.8 Reactor Vessel Sucoort Ledge Design And Analysis The Reactor Vessel Support Ledge will be designed in accordance with 1rhe AISC Code with the load combinations specifled for steel structures (Section 3.8.3.3.10.2). The analysi3 w11l be performed by the finite element method with the computer program STARDYi4E.
O 3.8-19 Amend. 77 May 1983
3.8.3.5 Structural Accentance Criterla 3.8.3.5.1 Stress Structures designed by the stress limitation methods will be considered acceptable, when design stresses for the most severe combination of loads are within the limits prescribed by the appropriate codes and standards noted in Section 3.8.3.2.
3.8.3.5.2 strain Since the design of the reinforced concrete structures will be governed by ACI-349, a strain limit of 0.003 used as a basis of this code will be inherently provided in the design. For the steel structures, except for liners and catch pans, the design is based on stress limits as specified in Section 3.8.3.3. In specific Instances, where plastic behavior of the steel will be a design basis, an energy absorption check will insure that the functional requirements of the structure are not impaired. Ductility ratio, p, is a measure of the capacity of a structure to absorb energy in the plastic range. Energy absorption is satisfied when the calculated value of the required ductility ratio is less than the allowable ductility ratio for the material under a specific loadir,g condition. For discussion on liner strain, see Section 3.8.3.4.
3.8.3.5.3 Gross Deformation s i The load combinations and stresses noted in Section 3.8.3.3 will Insure that the deformation of a structure will be no greater than that ordinarily permitted f or structures of this type. Reinforced concrete structures subject to loads combined with SSE will be nearly stressed to their ultimate capacity.
However, several relieving features such as strength gain with age, relief of thermal stresses due to cracking, etc., will preclude any excessive deformation. A design check will be performed to insure that the deformation resulting from SSE and other loads will in no way impair proper functioning of the critical systems or components.
3.8.3.5.4 Factor of Safety The f actor of saf ety for the working stress design will be in accordance with the limits noted in Section 3.8.3.3. For the ultimate load design, load l
f actors wil l be in accordance with the combinations noted in Section 3.8.3.3.
3.8.3.5.5 Shear Resoonse The shear response will be established upon the basis of the classical relationship between the Young's modulus and the shear modulus. The assumed value of the Poisson's ratio and concrete moduli will be checked against the properties of concrete as determined through tests of design mixes to be used )
in the plant construction. l nU 3.8-20 Amend. 77 May 1983
(r~g V) 3.8.3.6 Materials. Quality control and soecial Construction Technlaues 3.8.3.6.1 Materials 3.8.3.6.1.1 Concrete All structural concrete work wil l conf orm to ACl-301-72 (Revised 1975) modified as necessary for the more exacting requirements of the internal structures. The concrete will have a minimum 28-day or 90-day compressive strength of 4000 psi as indicated on construction drawings.
- 3. 8 . 3 . 6.1. 2 Reinforcing Steel The reinforcing steel will conform to ASTM A-615-76a, Grade 60. When reinforcing steel is required f or arc welding the material shall conf orm to ASTM cr ASME standards.
3.8.3.6.1.3 Liner Cell liner materials are listed in Attachment B of Appendix 3.8-B.
3.8.3.6.1.4 Structural Steet All structural steel wil l conf orm to ASTM A-36-75. When a special type of steel will be used to meet a specific requirement, the material will conform
(}
U to the applicable ASTM or ASME standards.
3.8.3.6.2 Construction Technfaues Concrete construction techniques for this project are conventional type. No use of pre-stressed concrete is envisioned. If No.14 or No.18 reinf orcing bars are used, mechanical splices will be specified to meet the requirements of Regulatory Guide 1.136.
3.8.3.6.3 puality Control General information on quality control is given in Chapter 17. See Section 3.8.4.6.2 for concrete tests.
Construction of reinforced concrete and structural steel Internal structures, including cell liners and catch pans wil l comply with ANSI N45.2.5 - 1978,
" Supplementary Quality Assurance Requirements for installation, inspection and Testing of Structura! Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants."
3.8.3.6.4 Testing Reautrements for Reinforcing Steel 3.8.3.6.4.1 MIII Tests Mill tests on bar samples will be performed in accordance with AEC Regulatory Guide 1.15, Testing of Reinforcing Bars for Category I concrete structures and 7s g ) the ASTM standards noted below.
\_/
3.8-21 Amend 77 May 1983
These tests are outlined as follows:
(a) Chemical Reat-!rements Chemical analysis will be performed on reinforcing steel, for each heat of steel, in accordance with the requirements of ASTM A-615. The percentage of carbon, manganese phosphorous and sul f ur composition thus determined will be included in the test report.
O O
3.8-21a Amend. 45 July 1978
3.8.5.5 Structural Accentance Criteria 3.8.5.5.1 Stress See discussion under subsection 3.8.3.5.1 3.8.5.5.2 Strain See discussion under subsection 3.8.3.5.2.
3.8.5.5.3 Gross Deformation See discussion under subsection 3.8.3.5.3.
3.8.5.5.4 Differential settlement See discussion under subsection 3.8.5.3.1 3.8.5.6 Materlats. Quality Control and soecial Construction See discussion under subsections 3.8.3.6, 3.8.4.6 and Chapter 17. Portion of the containment structure subject to the requirements of ASME code, Section lil, Division 2 will meet all testing requirements of that code.
3.8.5.7 Testing and in-service surveillance Reauirements
(s There will be no in-service surveillance requirements.
l l
V 3.8-39 Amend. 77 May 1983
References to Section 3.8
- 1. Appendix 3.7-A, Seismic Design Criteria for the CRBRP.
- 2. Chapter 3, SEQUOYAH Nuclear Plant FSAR Docket No. 50327, Appendix 3.88.
- 3. Specification for Electric Overhead Traveling Crane, Specification No. 70, Crane Manufacturers Association of America, Inc. 28 O
l 0
3.8-40 Amend. 28 oct. 1976
4.1.2 Low Hydrogen Electrodes shall be used for manual shielded metal-arc q welding.
Where backing strips are used, they shall be analytically compatible with the materials being joined. Backing strips shall be tack welded in the joint groove of the welded preparation. All backing strips must fit closely to the underside of the weld preparation.
4.2 Welding Procedures and Performance Qualif ications A written welding procedure specification, containing Information detailed in Form QW482 of ASME,Section IX, and the qualification test data containing the information detailed in Form QW483 of ASME Section IX shall be submitted to the Purchaser for review and approval.
A certificate of welder or welding operator performance qualification test shall contain the information as detailed in Form QW484 of ASME Section IX, and be available for Purchaser review if desired.
All welding repairs shall be made in accordance with a written welding procedure.
4.3 Studs and Anchors The welding of the studs or anchors to the liner plates shall conform to the requirements of the ASME Code Divisten 2, Subarticle CC-4543.5 and ASME Section lil, Division 2.
O(/ 4.4 Storage, Conditioning and Handling of Welding Materials 4.4.1 Filler materials shall be stored, conditioned and handled in accordance with the appendices of ASME Code - Section ll, Part C which are mandatory parts of this specification.
o
\j 3.8-B.14a Amend. 64 Jan. 1982
5.0 NONDESTRUCTIVE EXAMlNAT10N RE00lREMENTS 5.1 Cell Liners (except fuel handling cell) 5.1.1 Examination of Piate Seam Welds. Cell liner piate seem welds will be full penetration and will be examined in accordance with Article CC-5500 of ,
ASME BPVC, Section Ill, Division 2. Acceptance standards for welds will be in (
accordance with Subarticle CC-5540 of ASME BPVC, Section 111, Division 2. '
Radiography is a code requirement when the plate welds are accessible, however if this is r.ut feasible due to the method of construction, the following methods of examination w il I be used:
l a) The entire length of all liner welds will be examined visually prior to placing the cell into service.
l b) The entire length of ali IIner seam welds wIII be examined by the vacuum-box method using either a bubble solution or gas detector technique.
c) The entire length of all IIner welds, except for welds examined by method d below, will be examined by liquid penetrant or magnetic particle method, d) A minimum of 10% of the total length of ali Iiner seam welds wiI! be examined ultrasonically in accordance with Article 5, " Ultrasonic Examination" of the ASME BPVC Section V. Welds examined will be located on the floor or on walls below the potential sodium pool level to the maximum extent possible.
l e) The entire length of all attachment welds will be examined by either the magnetic particle or IIquid penetrant method.
Where radiographic examination is required, double film (two separate films in the same cassette) radiographic examination procedures will be utilized with the film properly exposed and developed for single film viewing.
5.1.2 Ultrasonic Examination of Plate. Pre-selected areas in the cell liner floor and wall plates below the postulated pool depth in three Reactor Containment Building and three Reactor Service Building cells will be examined ultrasonically in accordance with Article 5, " Ultrasonic Examination" of ASME BPVC,Section V, to determine the reference plate thickness to be used in the monitoring of cell liner plate corrosion. Areas to be selected in the test cells will include:
a) Four (4) locations on the floor near the corners of the cell.
b) One (1) location on the wall below the sodium pool level.
5.1.3 Inspection of Stud Welds. Studs will be visually inspected.
Acceptance standards for all stud welds will be in accordance with Subarticle CC-5548 of ASME BPVC, Section Ill, Division 2. In addition to the above, If a visual inspection reveals any stud shear connector that does not show a f ull O
3.8-B.15 Amend. 77 May 1983
' 360 degree leg weld flash, any stud that has been repaired by welding, or any stud in which the reduction in length due to welding is less Than normal, such
[m V'
i stud wIlI be struck with a hammer and bent to an angle of 15 degrees from its original axis. For studs showing less than a 360 degree weld flash, the direction of bending will be opposite to the missing weld flash. Studs that crack in the weld, the base metal, or the shank under inspection or subsequent straightening will be removed and replaced.
5.1.4 Leak Testing of Seals. Removable access panels in the walls of Cell Liners will have seals around the periphery of the panels.
Steel removable covers over hatch concrete plugs located cver Cell Liners will have seals around the periphery of the covers.
The seals in both the access panels and the steel covers will be tested for leak tightness, the method shalI be the Gas and Bubble Formation Testing in accordance with ASE BPVC,Section V, Article 10 " Leak Testing", Subarticle T-1030, at a test pressure of 10 psig. The seals will be field adjusted as required to obtain a leak-tight joint.
5.1.5 Leak Testing of Comoteted Lined Coll. After all liner panels, doors and hatches have been completely installed and openings and penetrations for system piping, electrical and HVAC have been blanked off, the cell will be pressurized to 2 psig. Leak tightness of the liner system will be acceptable if the rate of leakage is 1 0.36% volume / day measured under atmospheric pressure as standard cubic feet per minute (SCFM). All leaks identified during this testing will be permanently repaired to meet the closure details (welds, seals, etc.).
l 5.2 Fuei HandlIna CeIl l The full penetration cell liner seam welds of fuel handling cell will be examined :n accordance with Article ND-5000 of the ASME Code, Section 111, Division I requirements.
(
Radiography is a code requirement when the liner welds are accessible, however, thIs is not feasibIe due to the method of construction. The entire length of ali liner seam welds shalI be examined by the vacuum-box method using either a bubble solution or a gas detection technique. All welds shall be examined by magnetic particle or liquid penetrant techniques.
l l
1 l
l O l
V 3.8-B.15a Amend. 64 Jan. 1982
1 59l TABLE 3.B-B-2 CELL LINER B0UNDARY MATERIALS Application Material Supplemental Requirements
- 1. Liner Plates ASME SA-516 55 ksi i U.T.S. 1 65ksi (floors, walls Grade 55 ceilings)
Grade 1020 U.T.S. > 70ksi (7/8" 0 studs olily)
- 3. Structural Shapes ASME SA-36 58ksi < U.T.S. < 71 ksi (floor beams, embedments) 59 37 Note (U.T.S. = Ultimate Tensile Strength)
O l
3.8-B.16 Amend. 59 Dec. 1980 0
l A certificate of welder or welding operator performance qualification test (xs shall contain the information as detailed in Form QW484 of ASME Section IX,
() and be available for Purchaser review if desired.
All welding repairs shall be made in accordance with a written welding procedure.
4.3 STORAGE. CONDITIONING AND HANDLING OF WELDING MATERIALS 4.3.1 Filler materials shall be stored, conditioned and handled in accordance with the appendices of ASME Code - Section 11, Part C which are mandatory parts of this specification.
5.0 NON-DESTRUCTIVE EXAMINATION REOUIREMENTS 5.1 Plate The catch pan plate seam welds shall be full penetration and will be examined in accordance with Article CQ5500 of the ASME BPVC, Section lil, Division 2 requirements. Acceptance standards for welds shall be in accordance with subarticle CC-5540.
The entire length of catch pan plate seam welds shall be examined visually prior to performing any other examination.
Where plate weld joints are made without the use of back up bars, and the weld is accessible, radiography shall be used. Where plate joints are made with the aid of backup bars or if it is not feasible to radiograph the welds, due to the method of construction, the following methods of exemination shall be used:
(V~')
- a. The entire length of catch pan plate seam welds shall be examined by the vacuum box method using either a bubble solution or gas detector 16?.hnique, and
- b. The entire length of catch pan plate seam welds, except for welds examined by method d below shall be examined by the magnetic particle method.
- c. The entire length of all attachment welds shall be examined by the magnetic particle method.
- d. A minimum of 10% of the total length of all seam welds will be examined ultrasonically in accordance with Article 5, " Ultrasonic Examination" of the ASME BPVC Section V.
Where radiographic examination is required, the builder shall use double film (two separate films in the same cassette) radiographic examination procedures with the film properly exposed and developed for single film viewing.
u 3.8-C.15 Amend. 77 May 1983
5.2 Ultrasonic Examination Pre-selected areas in the catch pan floor and wall plates below the postulated pool depth in two SGB and one RSB cell shall be examined ultrasonically in accordance with Article 5, " Ultrasonic Examination" of the ASME Code,Section V, to determine the reference plate thickness to be used in the monitoring of the catch pan plate corrosion. Areas to be selected in the test cells include:
- a. Four (4) locations on the floor near the corners of the cell.
- b. One (1) location on the wall below the postulated sodium pool level.
5.3 Metal Dec.h The side lap welding of the adjacent metal deck units shall be visually b inspected as per Article IX-2370 " Visual Examination" in Appendix IX "tiondestructive Examination Method" of the ASME Code, Section Ill, Division 2 req u i rements.
5.4 Insoection of stud Welds Anchor studs wIlI be visually inspected. Acceptance standard for alI stud welds wiII be in accordance with Subarticie CC-5548 of ASME BPVC, Section lII, Division 2.
O O) 3.8-C.16 Amend. 74 Dec. 1982
1 I
( i.ui 3 A. 4-3 A INERTED CELL DESIGNATION LIST REACTOR SERYlCE BUILDING STRUCTURAL UPSET j FLOOR RADIATION ZONE (1) DESIGN PRESS. DES. TEW. NORNAL EQUIPENT ll CFf f NO. Ti1LE ELEVATION DP_ERATION SHUTDOWN (PSIG) _(OF)(3) ATMDS.(2) CONTAINED 331 EVST Third Loop IHX 798'-6" IV lli 12 150 N EYST Backup Na Ce1I Cooler 331A EVST Third Loop 798'-6" IV 111 (12 150 N Piping i Pipeway I
337 EVST Cell 759'-4" V IV 12 150 N Ex-Vessel Storage Tank l 341 Fuel Handling Celi 781'-0" V 1Ii 12 150 A FHC Equipment 351C Pipeway 770'-0" IV lli 12 150 N Piping 351D Pipeway 770'-0" IV lli 12 150 N Piping 351A Nain EVS Cooling 775'-0" IV lil 12 150 N Piping Sys. Pipeway
?
e 351B Nain EVS CooIIng Sys. Pipeway 775'-0" IV lli 12 150 N Piping 357 EVS Cooling Loop B 779'-0" lY lli 12 150 N EYST Na Cooler Cell Pump 357A Cooling Loop B 798'-6" IV 111 12 150 N Piping
'Pipeway 357B Cooling Loop B 798'-6" IV lli 12 150 N Piping Pipeway j 360 EVS Cooling Loop A 779'-0" IV lil 12 150 N EVST Na Cooler &
CelI Pump 360A EVS Cooling Loop A 798'-6" IV 113 12 150 N Piping Pipeway 361 EVS Na Co;d Trap 779'-0" IV IV 12 150 N EVS Cold Trap &
CeiI Economizer yg 3E4 EVS SSP Cell 779'-0" IV lil 12 150 N Ex-Vessel Na SNPLG Pkg.
<< g UP 387 PTl Cell 765'-0" IV lli 12 150 N PTl
$w N Notes: (1) For definition of Radiation Zones see Table 12.3-1.
l (2) N = Nitrogen A = Argon (3) Liner Dnign Temperature
O 35 30 -
a/
7
/
/
25 - *
/
20 .
- v i /
/
O b 15 - ! SODIUM TEST DATA (T = 900*F) 8 /
{m g ,_ 100% Flow
/
E o- 40% Flow 7
10 -
[o 7 DYNALSS PREDICTIONS
/
100% Flow 40% Flow 5 __-
/
/
0 O 0.2 0.4 0.6 0.8 1.0 1.2 1.4 Time After Initiation of Rod Motion (Sec) l I
Figure 4.2-121 - SCRS R0D DISPLACEMENT VERSUS TIME Amend. 71 4.2-624 Sept. 1982
O FIGURE 4.2-122 HAS BEEN DELETED 4.2-625 Amend. 76 M,rch 1983
- 13. The control assemblies flow rate will be such as to assure adequate margin against flotation in case the driveline becomes accidentally h
v disconnected (see Section 4.2.3.1.3).
- 14. Assemblies orificing will be designed to be consistent with the requirement that the lower shinld in the f uel, blanket and control assemblies will have suf ficient solid volume fraction to limit radiation damage to the core support structure and to assure its prescribed lifetime.
- 15. The thermal-hydraulic design of the control assemblies will be such as to satisfy the scram insertion requirements during the reactor Iifetime (see Sectlon 4.2.3.1.3).
- 16. The sodium temperature shall be less than its bolling point during normal operation and anticipated and unlikely transient conditions.
- 17. The reactor will meet the aforementioned design bases operating over a range of power and flow rates, including power ranges and flow variations, from 0 to 100% of nominal conditions.
- 18. Adequate design margins (see Section 4.4.3.2) will be provided to account for design, fabrication, operational uncertainties and tolerances to ensure meeting the aforementioned Iimitations. The semi-statistical hot channel factors approach will be adopted in combining individual fuel, blanket and control assembly uncertainties.
- 19. As explained in Section 4.4.3.3.1, plant T&H design conditions are considergd in performance evaluations of permanent plant com-ponents( ), e.g., vessel, Internals, heat exchangers. Therefore,
, these conditions shall be considered in evaluation of items 7 through 10,16 through 18. On the other hand, plant expected operating conditions are adopted in steady state performance and design evaluations of replaceable components such as the reactor assembiles.
Therefore, plant expected operating conditions shall be considered in evaluation of items 1 through 6,11 through 15,17 and 18.
- 20. The design shalI prevent gas entrainment sufficient to cause significant heat transfer or reactivity changes in the core.
- 21. The design shall provide features to minimize the potential for flow blockage of incore assubiles.
V 4.4-3 Amend. 77 May 1983
l 4.4.2 Descriotion 4.4.2.1 Summarv Comoarlson This section presents a comparisen of general and core assemblies design paraneters f or the CRBRP and FFTF reactors,
- l. .CRBRP AND FFTF GENERAL PARAMETER COMPARISON Units CRBRP** Ef.IE*
Design Lif e Yrs. 30 20 Reactor Power (Thermal) MWt 975 400 Primary Coolant -
Sodium Sodium Primary Coolant Des!gn FIow Rate 106 lbm/hr 41.45 17.28 Coolant Temperature:
Reactor Vessel Inlet OF 730 600 Reactor Vessel Outlet OF 995 858 Reactor Vessel Temperature Rise CF 265 258 Pressure Drop:
Reactor inlet Nozzle-to-Outlet Nozzle (4) psi 123 110 Lower Inlet Module to Assembly Outlet Nozzle psi 116 101 Primary Pump Design (static) psi 160.3 182.5 Number of Primary Loops -
3 3 Suppressor Plate -
Yes Yes Cover Gas -
Argon Argon Cover Gas Pressure (nominal) psig 0.36 0.36 Alloweble Overpower percent 15 15
(+) Permanent plant components are those components which: 1) wilI be designed f or 30 year iIf e; and 2) cannot be east 1y rept aced.
- FFTF Initial Condition
- CRBRP T&H Design Val ues CRBRP value includes uncertaintles; FFTF value is nominal.
4.4-4 O
Amend. 69 July 1982
O Table 5.6-12 DHRS OPERATING AND SENSITIVITY EVALUATIONS DHRS NUMER NUM ER NUEER EVST NA NAK !
OPERATING PEAK OF PHTS OF lHTS OF DHRS ODNSERVATIVE HEAT FLOW FLOW CASE (1) TEMPERATURE LOOPS LOOPS TRAINS DECAY HEAT LOAD GPM GPM Design 41120 f 3 0 2 Yes Yes 560 800 Case
! Updated 41052'F 3 0 2 Yes I2I Yes 560 800 ,
Design Case l Single I53 Nominal No 600 600 s1055*F 2 2 1 Failure Values ,
m Evaluation #1 Used l Single s1137 2 0 1 Yes I23 No 600 600 W Failure ;
$ Evaluation #2
! Delayed s1078 3 0 2 Yes (2) Yes 560 800 Start Design Case (ne45 minutes to DHRS operation)
(1) Conservatisms Included in all cases: Shutdown at end of cycle at f ull powers no consideration of Insulation, I no consideration of SGS heat capacity.
l (2) Decay heat values have been revised to reflect current conservative design values. ,
l (3) The ivo PriTS loops assumption conservatively envelopes the assumption that a PUTS pump or pony motor f alls to i operate.
h wN MS 8-w~ ,
N i u _ 7 ., , , , _ - . - . , _. -
Table 5.6-13 DHRS EnUIPMENT LIST AND MATERIAL SPECIFICATIONS ASME Design Design Section III Temperature Pressure Component Class Material (OF) (psii Overflow Vessel 1 SS 900 15 Makeup Pump 1 SS 900 100 48 l Overflow Heat Exchanger 1 SS 650 100 46 EVST Airblast Heat Exchanger 2 SS 650 100 EVST Nak Exp Tanks 2 SS 650 100 DHRS Nak Exp. Tank 2 SS 650 100 EVST Nak Pumps 2 SS 650 100 EVST Nak Diff Cold Traps 2 SS 650 100 Sodium Piping:
Overflow Line 1 SS 900 15 Makeup Pump Suction 1 SS 900 15 Makeup Pump Discharge to Reactor 1 SS 900 100 Nak Piping 2 SS 650 100 26 Amend. 48 5.6-35 h Feb. 1979
In the areas where the rupture of the steam or feedwater iInes can occur, the fleid instrumentation and control shalI be cualIfled to survive the resulting higher temperature and pressure transient.
The design of the PPS will include provisions for testing using a monitor.
The PPS Monitor is an autcmatic test device which will Test the PPS logic by monitoring the Primary PPS logic as test pulses are inserted into the Primary PPS. The Monitor will also trigger the initiation of the test pulses.
The design of the PPS Monitor will meet the following criteria:
- 1. The design of the PPS logic test system will provide two independent Monitors such that failures in one Monitor will not propagate to the other Monitor. The Monitors will be used to check the test results of each other.
- 2. Self-test features will be provided for each Monitor.
- 3. The automatic test of the PPS logic will be considered as a safety related f unction and will be performed by Class 1E devices. Thus, the PPS Monitors will be Class IE.
- 4. Each Monitor will be Independent from any other monitor. If isolation devices are required they will be considered IE ana designed to the requirements of IEEE 279 and qualified in accordance with IEEE 323.
- 5. A QA program meeting the requirements of 10CFR50 Appendix B (see A Section 17.1) will be applied to the PPS Monitor.
U Failure modes and ef fects analyses will be perforn.ed on the PPS logic test system and in particular on the PPS Monitors. The design of the PPS logic test system will prevent any common mode failures identified in the FMEAs.
7.1.2.2 Indeoendence of Redundant Safetv-Related Svstame To assure that independence of redundant safety-related equipment is preserved, the following_ specific physical separation criteria are imposed for safety-related instrumentation.
o All interrack PPS wiring shall be run in conduits (or equivalent) with wiring for redundant channels run in separate conduits. Only PPS wiring shall be included in these conduits. Primary RSS wiring shall not be run in the same conduit as secondary RSS wiring. Wiring for the CIS may be run in conduits containing either primary RSS wiring or conduits containing secondary shutdown system wiring, but never intermixed. Expanded criteria for physical separation of the CIS are given in Section 7.3.2.2.
o Wiring for other saf ety-related systems may be run in conduits containino either primary RSS wiring or conduits containing secondary RSS wiring, but never intermixed, provided that no degradation of the separation between primary and secondary RSS results.
A 7.1-3 Amend. 77 i May 1983
o Wiring for redundant channels shall oe brought through separate containment penetrations with only PPS wiring brought through these penetrations. Primary RSS wiring shal I not be brought through the same penetration as secondary RSS wiring. Wiring for the CIS and other safety-related systems will be brought through the same penetration as the RSS wiring with which it is routed.
o Instrumentation equipment associated with redundant channels shall be mounted in separate racks (or completely, metallically enclosed compartments). Only PPS channel instrumentation shalI be mounted in these racks. Primary RSS equipment shall not be located in the same rack as secondary RSS equipment, o The physical and electrical separation between DC and nC uninterruptible power supplles, conduits, equipment or racks of instrument channels of saf ety divisions 1, 2 and 3 shall meet the requirements of IEEE 384 and Reg. Guide 1.75. Redundant instrument channels in the primary RSS shall be physically separated f rom one ,
another in accordance with the requirements of IEEE 384 and Reg. Guide 1.75.
Physical separation of primary and secondary RSS of the same division (
(channel) shall meet the requirements of IEEE 384 and Reg. Guide 1.75 in non-hazard areas. In other areas of the plant primary and secondary RSS equipment and cables of the same division may be located in the same hazard zone.
Functional capability is maintained in the event of single design basis events which might impact more than one sensor by alternate protective functions as indicated in Table 7.2-2.
o The wiring from a PPS buffered output which is used for a non-PPS purpose may be included in the same rack as PPS equipment. The PPS wiring shall be physically separated f rom the non-PPS wiring. The amourt of separation shalI meet the requirements of IEEE 384-1974.
o Electrical power for redundant PPS equipment shal I be supplled f rom separate sources such that failure of a : ingle power source does not cause failure of more than one redundant channel. The power sources and associated wiring shalI be sepa'ated, r as specifled in Section 8.
O 7.1-3a Amend. 75 Feb. 1983
4 The criteria for cable tray fill, cable derating, cable routing in congested or hostile areas, fire detection and protection in cable areas, and cable markings are def ined in Section 8.
s V) Separation of redundant saf ety-related equipment within the control boards is described in Section 7.9.
7.1.2.3 Physical identification of Safetv-Eglated Eautoment The Plant Protection System equipment will be identified distinctively as being in the protection system. This identification will distinguish between redundant portions of the protection system such that qualifled personnel can distinguish whether the equipment is saf ety-related and, if so, which channel.
Color coding, cabinet and wire labeling and other techniques as appropriate will be used.
7.1.2.4 Conformance to Regulatorv Guldes 1.11 " Instrument Lines Penetrating Primary Reactor Containment" and 1.63 " Electric Penetration Assem-blies In Containment Structures for Watercooled Nuclear Power Plants" There are no instrument lines as def ined in Regulatory Guide 1.11 which penetrate primary reactor containment. All electric penetration assemblies in the containment vessel will be designed, constructed and installed in accordance with Regulatory Guide 1.63 and IEEE Standard 317-1972.
7.1.2.5 Conformance to IEFF Standard 323-1974 "lEEE Standard for Oualifving Class lE Eauloment for Nuclear Power Generating Stations u All Class IE equipment will be qualified to confirm the adequacy of the O. equipment design under normal, abnormal, and postulated accident conditions V for the performance Class lE f unctions. This will be accomplished through a disciplined program discussed in Ref erence 13 of PSAR Section 1.6, "CRBRP Requirements for Environmental Qualification of Class IE Equipment." 61 7.1.2.6 Conformance to IEEE Standard 336-1971 " Installation. Insoection and Testina Reaufrements fcr Instrumentation and Electric Eautoment Durina the Construction of Nuclear Power Generatina Stations" The Installation, !nspection and testing of the instrumentation, electrical and electrcnic equipment during construction will conform to the requirements of IEEE Standard 336-1971. The quality assurance program for the saf ety-related instrumentation and control equipment will conform to the requirements of Regulatory Guide 1.30. Ref er to Chapter 17 for a description of the qual ity assurance program.
7.1-4 Amend. 61 Sept. 1981
TABLE 7.1-1 fh a SAFETY-RELATED INSTRUMENTATION AND CONTROL SYSTEMS
- Reactor Shutdown Svstems includes all RSS sensors, signal conditioning calculation units, comparators, buffers, 2/3 logic, scram actuators, scram breakers, control rods, back contacts on scram breakers, HTS shutdown logic, coolant pump breakers, and mechanical mounting hardware (equipment racks).
Containment Isolation Svstem includes radiation monitoring sensors, signal conditioning, comparators, 2/3 logic, containment isolation valve actuators and valves.
Decav Heat Removal System Instrumentation and Control System includes initiating sensors, signal conditioning, calculation units, comparators, logic, auxiliary feedwater pump actuators and controls including f eedwater turbine pump, PACC DHX actuators and controls, steam rel lef valve actuators and valves; sensors, signal conditioning, logic and actuators related to decay heat removal functions of DHRS including control of sodium and NaK pumps and air blast heat exchangers; and sensors, signal conditioning, logic and actuators related to removal of heat from the EVST.
Sodium-Water Reaction Pressure Rellef System (SWRPES1 d The instrumentation, initiation and control logic which achieves adequate isolation and blow-down of the waterside of a superheater or evaporator in the event of a sodium / water reaction is Class 1E. The instrumentation used to initiate the isolation and blow-down valves are the rupture disc pressure detectors located downstream of the rupture discs. The other pressure and temperature instrumentation distributed throughout the sodium / water reaction pressure rellef subsystem is used for status indication and is not Class IE.
- The Cl inch River Breeder Reactor Plant (CRBRP) saf ety-related structures, systems, and components are designed to remain f unctional in the event of a Safe Shutdown Earthquake (SSE). These include, but are not limited to, those structures, systems and components which are necessary:
- a. To assure the integrity of the Reactor Coolant Boundary;
- b. To shutdown the reactor and maintain it in a safe shutdown condition;
- c. To prevent or mitigate the consequences of accidents which could result in potential off-site exposures comparable to the guideline exposures of 10CFR100.
A detailed listing of all Clars IE Instrumentation and Control equipment is provided in Reference 13 of PSAR Section 1.6.
fg NOTE: Class IE equipment loads are identifled in Chapter 8.
'd 7.1 -7 Amend. 77 May 1983
Rther Safetv-Related Instrumentation and Control includes instrumentation and Controls for portions of the following functions to assure the plant is maintained in a safe shutdown condition:
o Emergency Chilled Water System o Emergency Plant Service Water System o Instrumentation necessary to assure plant is maintained in safe shutdown status (See Table 7.5-4) o Heating, Ventilating, and Air Conditioning System o Recirculating Gas Cooling System o Process and Effluent Radiation Monitors and Samplers (See Table 11.4-1).
l Solid State Programmable Logic Svstem O
O 7.1-7a Amend. 77 May 1983
7.5.6.1.3 Bvoasses and Interlocks
() The control logic for the actuation of the Sodium-Water Reaction Pressure Relief System will be designed to insure reliability and freedom from spurious operation. A discussion of the bypasses and Interlock functions and their safety classifications will be provided in the FSAR. Bypasses and interlock f unctions which are required for operation of the initiation and control logic will be Class 1E.
7.5.6.1.4 Sodlum Onen No automatic action is associated with the removal of sodium from the af fected loop. However, sodium dump valves are provided for draining of sodium to the sodium dump tank, and can be initiated by operator action.
Drain valves are located in five piping runs between the IHTS sodium loop and the sodium dump tank. Each piping run contains a pair of drain valves arranged in series. Controls and indications f or all these valves are located on the Main Control Board.
7.5.6.1.5 Monitoring instrumantatten in addition to the instrumentation required for the initiating circuitry, the following parameters are measured to aid the plant operator in assessing the perf ormance of the Sodium-Water Pressure Relief System:
o Pressure in the gas space between each pair of rupture disks is monitored to detect leakage, or f ailure of the sodium side rupture O'. disk. Spark plu] leak detectors are also provided in the gas space to detect rupture disk failure.
o Thermocouple elements are provided for monitoring surface temperatures of the reaction products separator tank, centrifugal separator, centrifugal separator drain tank, and the hydrogen Igniter.
C's J
7.5-32 Amend. 77 May 1983
The third (backup) cooling circuit is designed to remove 1800 KW of heat while maintaining sodium temperatures within the EYST below
(}
' 'N /
7750F. The damper position on the natural-draf t heat exchanger is adjusted to control EVST sodium temperature under various heat loads.
NaK volumetric changes from temperature variations are accommodated in the NaK expansion tank connected to the high point in the NaK loop.
Oxide level in the NaK loop is minimized by the dif f usion cold trap.
The backup circuit is maintained in a preheated condition and is ready for immediate operation. A small sodium and NaK flow is induced by slIghtly opening the dampers of the natural-draf t heat exchanger and removing a minimum amount of heat during normal EVST cooling by circiits 1 or 2. Trace heaters are also provided fcr the sodium 59 piping and components. j c) 06eration Durino Reacter Decav Heat Removal - Reactor decay heat removal is accomplished by using the combined heat removal capability of both of the normal NaK loops in the EVS Processing System. In conjunction with the primary sodium overflow heat exchanger, each of the NaK loops circulate approximately 400 gpm through each airblast heat exchanger. The circuits are interconnected during this operating mode to provide a total NaK flow of 800 gpm which is routed through the shell-side of the overflow heat exchanger in the Primary Sodium Storage and Processing System (described in Section 9.3). When the ;
DHRS is initiated one-hal f hour af ter reactor shutdown, the NaK exits '
f rom the overflow heat exchanger at a maximum temperature of approximately 10000F. At this temperature, the NaK airblast heat m exchangers have a combined heat removal capacity of approximately 11-1/21W, which is suf ficient to remove decay heat from both the reactor and the EVST.
During the DHRS mode of operation, one of the normal Na/NaK loops remai ns i n use f or EV ST cool ing. The NaK in this loop is circulated (400 gpm) from the airbiast heat exchanger through the EYST sodium cooler prior to its flow to the overflow heat exchanger. The EVST, in effect, is cooled in series with the overflow heat exchanger. In this fIow pattern, the EVST Is Iocated in the cold ieg of the Ioop in order to minimize temperature rise of the EVST sodium. The sodium and NaK flow path in this mode of operation is shown schematically in Figure 9.1-11. Switchover from normal cooling (EVST) only to reactor decay heat removal (DHRS) is done remotely from the control room.
Switchover is accomplished by opening the isolation valves at the connections to each of the normal EVST cooling loops. The DHRS NaK (0 -
9.1-25 Amend. 59 Dec. 1980 Y _ _ _ _ _ - .
i 1
i expansion tank is isolated and the EVST NaK pump is increased to 400 gpm each. The cover gas space in the two EVST NaK expansion tanks is cross-connected to equalize tank NaK levels.
1 9.1.3.1.3 Safetv Evaluation )
l The EVST cooling capability can be provided by either of two identical, forced l convection cooling circuits, each of which can remove 1800 kW while maintaining a maximum EVST sodium outlet temperature of N510 F.
In the extremely unlikely event that the normal circuits are unavailable, heat will be removed through a third independent (backep) natural convection cool ing circuit. At 1800 kW this backup cooling circuit will maintain sodium temperatures within the EVST below 775 F.
The critical temperature- in a f uel assembly, f rom the standpoint of saf ety, is the peak fuel cladding temperature. The normal and emergency limits are given in Table 9.1-2.
The peak f uel cladding temperatures shown in Table 9.1-2a, are within the Iimits. Hence, no damage to the stored f uel assembl les w iI I occur.
The codes and standards to which the EVST vessel and the surrounding guard tank are designed and fabricated assure that leakage of sodium will be a very
,ow probability event. At the minimum level, adequate cooling is maintained with no temperature increases f rom those shown in Table 9.1-1.
The EVS Processing System components are designed to accepted industrial and nuclear standards to insure structural integrity and operational rel iabil ity.
The components, applicable design code and class, plus their seismic adequacy are listed in Table 9.3-1. Design temperatures and pressures are given in Table 9.3-7.
Each of the three sodium cooling loops is designed _ gainst the possibility of common-mode failure. Two pump suction lines are provided within the EVST for normal sodium circuit No. 2. The open end elevation of each is different, one high, one low. Each of the two Iines is separately valved externally to the EVST. After the initial fill of the loop, the Isolation valve in the low suction Iine is locked closed and remains closed (except for periodic testing) throughout the plant life. This low suction line is used only in the event of a major loop or vessel rupture. One pump suction line is provided within the EVST for normal cooling circuit No.1. The open end elevation of this Iine is between those for circuit No. 2. This Iine is valved externally to the EVST, and is called a "high" pump suction line. During normal system operation, one of the normal cooling loops is operated using the "high" pump suction line.
The suction iIne(s) in the standby normal loops are closed, in the event of a major f ail ure (rupture) of the operating normal sodium cooling loop, the isolation valve in the pump suction Iine is closed by operator action from the control room, signalled by concurrent alarms, Indicating low level in the EVST O
9.1-26 Amend. 77 May 1983 l
t
- 1. FHC Normal Fuel Handling
/
(m\ in a typical spent f uel handling sec,uence, a spent f uel assembly in a core component pot is lowered through t'..e f uel transfer port (see Figure 9.1-7) by the EVTM, into the spent fuel transfer station directly below the port. A lazy susan assembly, with three transfer positions supported by a stainless steel gridwork, provides the storage locations. Each position holds one fuel assembly, in a sodium-filled core component pot. Decay heat is removed by natural convection to the FHC atmosphere.
The spent fuel assembly is removed from the core component pot by the in-celi crane, using a gas-cooling grapple, and allowed to drip dry. If for some reason not identified as a part of normal procedures, it is deemed necessary to remove a sodium film from the exterior surf aces, the exterior surf aces will be wiped with alcohol wetted swabs.
Then the spent fuel assembly is lowered into the spent fuel shipping cask located in a shaft below the cell floor. The sequence is repeated f or the number cf assembiles necessary to fili the shipping cask. The above functions within the FHC are performed remotely by operators in the adjacent operating gallery, and can be observed through the viewing windows.
Normal core assembly handling operations in the FHC are conducted with assemblies having a decay heat of 6 kW or less. Inf requently, it may be necessary to examine a high-powered core assembly. This w'ill be done only when (1) It is necessary to complete refueling or to commence reactor startup, or (2) use of the FHC is necessary to recover from an EVTM grapple malf unction
, occurring while grappled to a high-powered core assembly.
During these operations, special precautions shalI be observed, including removal of all other spent core assemblies f rom the FHC prior to introduction of the high-powered core assembly. In addition, only one core assembly greater than 6 kW shalI be permitted in the FHC at any time.
- 2. Soent Fuel Examtriation Spent f uel examination in the FHC is lInited to inspecting the exterior surfaces of fuel assembiles to determine their geometrical condition before loading into the spent fuel shipping cask. Spent f uel assemblies will not be disassembled or sectioned in the FHC.
It is planned that only a few selected spent fuel assemblies will be examined, after the plant operation has reached its equilibrium. During the first few refuelings, it is expected that more spent fuel assemblies may be inspected.
The extent of the spent f uel examination covers the following operations, all of which will be performed in the fuel examination fixture:
- 1) Visual inspection of all exterior surfaces
- 2) Determination of axial and radial dilation of fuel assembly by measuring its length and distances across flats O
b 9.1-65a Amend. 77 May 1983
I
- 3) Measurement of the fuel assembiy bow
- 3. FHC Maintenance O' in general, standard maintenance techniques will be used to maintain FHC equipment as have been successf ully applied at Argonne West's Hot Fuel Examination Facility (HFEF) at the Idaho Nuclear Engineering 59 Laboratory, and at Al's SNAP reactor test f acilities.
Most equipment in the FHC is either modularized and sized such that it can be disassembled and positioned for removal from the cell exterior or repaired in the cell utilizing the remote manipulators.
59 Several access modes are available for the FHC including:
- 1) Removal of all or part of the roof closure gives access to the FHC from the RSB operating floor.
- 2) 28" dia. Floor Valve
- 3) 60" dia. Fuel Transfer Port 59
- 4) An FHC transfer drawer (gas lock) connects the FHC containment to the FHC operating gallery.
Large equipment (e.g., the crane bridge or trolley) can be removed, i.e.,
bagged out, from the FHC through the FHC roof closure port using the RSB bridge crane. Once outside the cell the equipment can be transferred to the 1 Large Component Cleaning Vessel (LCCV) for cleaning, if required prior to '
inspection or maintenance. An alternative is to erect a " green house" (plastic tent) over the FHC roof and part of the RSB operating floor. Large equipment could be removed frcm the FHC through the roof closure port and set down nearby on the RSB operating floor, within the green house. Through an air lock, suited-up personnel with appropriate respiratory protection could l enter the green house and perform the required maintenance tasks.
Suited-up personnel could also enter the FHC to repair nonremovable items or ,
to refurbish the cell. If such an occasion would rise, careful plans would be j i
made in cJvance for all work to be done after entry to minimize occupancy time i in the FHC. All fuel would be removed f rom the FHC, and the cell would be l thoroughly cleaned using the powered manipulator with brushes, dustpans, and vacuum cleaners equipped with HEPA filters before personnel could enter. The
! personnel would wear appropriate respiratory protection and clothing consistent with personnel safety requirements.
l Certain consumable material and spare parts (like motors, switches, light i bulbs, etc.) will be supplied for permanent storage in the FHC. Small parts can be easily removed through the transfer drawer and can be moved off the facility by bagging.
9.1-65b Amend. 59 Dec. 1980
9.13 PLANT FIRE PROTECTION SYSTEM O 9.13.1 Non-Sodium Fire Protection Svstem V
The Non-Sodium Fire Protection System (NSFPS) provides the plant with equipment, piping, valves, detectors, instrumentation and controls to prevent or mitigate the consequences of a non-sodium fire. Table 9.13-1 shows the areas covered by the Non-Sodium Fire Protection System. Firo Hazard Zones are areas in which a potential fire hazard could exist as a consequence of the credible accumulation of a significant quantity of fixed or transient combustible material.
The fire protection system design will meet the intent of Branch Technical Position CMEB 9.5-1 " Guidelines for Fire Protection For Nuclear Power Plants" with the exception that no fire protection water systems will be provided for those areas containing liquid ineral sodium.
9.13.1.1 Design Bases
- a. Fires that could indirectly or directly affect Seismic Category I safety-related structures, systems and components are identified in Table 9.13-2.
Potential fire hazards which provide the base for the design of the fire protection system in areas containing engineered safety-related structures, systems or components, are self contained tube oil systems, diesel generator fuel oil system, electrical cable insulation, and activated carbon filters. The intensities of the maximum fires involving the above combustible materials are listed in Table 9.13-2 and described p below.
V These maximum fires, together with lesser intensity design-basis fires, have served, via a preliminary fire hazards analysis, as the bases for the selection of fire protection measures involving all safety-related systems and equipment. The adequacy of the fire protection measures taken will be confirmed in conjunction with a more detailed fire hazards analysis which will be provided in the FSAR.
Lubricatino Oil The maximum fire involving lubricating oil would develop in the turbine lubricating oil equipment located in the Turbine Generator Building and would have an intensity of 20,000 BTU /lb. Fire from this source does not invofve safety-related areas and safc shutdown of the plant wIi1 not be jeopardized.
Diesel Generator Fuel Oil The largest potential source of fire from fuel oil is in the vicinity of the standby diesel generator fuel oil storage tanks, located below grade adjacent to the Diesel Generator Building. As these tanks are located below grade, the chance of an accident is reduced. Physical separation provided between the two tanks limits the spreading of fire from one i
/~N
\ )
v 9.13-1 Amend. 77 May 1983
tank to the other. Since either tank is capable of fulfilling the emergency f uel oil requirements, a saf e shutdown of the plant will not be Jeopardized by a fire in either tank.
The diesel generator fuel oil day tanks are each sized for a maximum of 1100 gallons of fuel oil and are located inside the Diesel Generator Building. Each day tank is located within a 3-hour fire rated concrete enclosure capable of containing the entire contents of the tank in the event of a leak. The day tank enclosures are maintained free of any ignition sources.
Charcoal Filters The maximum fire involving charcoal filters would develop in the Control Building. Filters are bounded and separated by fire barriers, and the filter units are redundant to each other, so safe shutdown of the plant will not be Jeopardized by a fire in either filter. A manually operated sprey system on the f ilter beds is prov ided f or f ire protection.
Electrical Cable Insulation The maximum fire involving electrical cable insulation would develop in the cable spreading rooms located in the Control Building. The cable spreading rooms are separated f rom the other areas of the Control Building by fire barrier wall. Adequate automatic detection and alarm devices and preaction spray systems are provided in the cable spreading rooms. Hence, safe shutdown of the plant wl.Il not be Jeopardized.
The use of halogenated compounds in cable insulation will be used only when substitute non-combustible materials are not available. Wherever practical Jacketing materials with high flame resistance and low smoke and of fgas characteristics shall be utilized,
- b. Discussion of Fire Characteristics Table 9.13-3 lists the saf ety related areas of the plant containing combustible materials. The burning characteristics of these materials such as maximum fire intensity, flame spreading, smoke generation and toxicity of combustion products are listed in Table 9.13-2.
Noncombustible flame retardant and heat resistant materials are used throughout the plant wherever practical to minimize the fire intensity in any combustion zone. The integrity of vital areas, components and systems is insured through the use of redundancy, physical separation and fire barriers.
- c. Design features of Building. Facilities and Structures for Fire Prevention The Fire Hazard Analysis Appendix B (Fire Protection Arrangement Dwgs) shows fire barriers which have been determined on the basis of preliminary fire load analysis.
O 9.13-2 Amend. 76 March 1983
l The ten-cell mechanical draf t wet cooling tower (two towers of five cells each) will dissipate the total heat load of the Circulating Water and Normal pI t Plant Service Water Systems at a design wet bulb temperature of 76 F. The cooling tower design approach is 11 F and the range is 21.34 F.
l The cooling tower basin is sized to maintain 1,272,000 gallons of water.
Makeup to the basin from the River Water Service System will maintain the basin level between predetermined Iimits.
10.4.5.3 Evaluation The cooling tower wIII be constructed of non-fIammable materials throughout to minimize the potential for fire. The normal drif t from the cooling tower will not affect the electrical switchgear and transformers.
The CWS is a non-safety-related system and is not required during a plant shutdown. If available, it would be utilized for residual heat removal; if not available, then the SGAHRS (Section 5.6) would be used. Failure of the CWS will not adversely af fect the f unction of any safety-related equipment.
The cooling tower is located such that its f ailure will not jeopardize the safety-related equipment.
The volume of water that could drain into the TGB as a result of a f ailure in the circulating water piping or condenser expansion joints is approximately equivalent to the inventory of the condenser (51,300 gal) plus the circulating water contents of the cooling tower basin (about 1,272,000 gal), i.e., a total of 1,323,000 gal. before the circulating water pumps would cavitate (assuming O)
\. no operator action).
There is no safety-related equipment located in the TGB. There are two potential pathways (below the maximum circulating water system flood level) for water to enter buildings where safety-related equipment is located, i.e.,
the personnel door in the Steam Generator Building Auxiliary Bay of the Steam Generator Building (SGB) and the personnel access corridor leading over the Electrical Equipment Building (EEB) to the Control Building (G) and other Nuclear Island buildings. Other openings in the TGB through which water could empty before entering the Nuclear Island (NI) buildings include doorways leading to the Maintenance Shop and Warehouse Building and the yard transformer area, as well as the roll-up door for the TGB railroad access bay.
The intended f unction of the safety-related equipment will not be impaired by the flow of this water into the Nuclear Island buildings since the potential pathways leading into the NI buildings f rom the TGB will be provided with water tight doors or will be located such that they are above the potential flood level .
10.4.5.4 Testing and insoection Reaulrements The valves and major components of the Circulating Water System are subject to hydrostatic and performance tests prior to plant operation. Hydrostatic leak test prior to initial operation will be made in accordance with the requirements of codes and standards to which the system is designed.
U) 10.4-7 Amend. 77 May 1983
10.4.5.5 InstrumentetIon AoolIcations 41 l Pressure and temperature alarms and pressure control are provided as required on the Circulating Water System. The pump and cooling tower parameters are continuously monitored to ensure that Circulating Water System performance is optimum.
10.4.6 rsndensate CI+anuo system 10.4.6.1 Design Bases The Condensate Cleanup System (I.e., Condensate PoiIshing Unit) is designed to maintain the condensate purity by removal of the following contaminants:
- a. Corrosion products that result from the corrosion that occurs in the main steam and turbine extraction piping, feedwater heater shells, drains, and condenser.
O O
Amend. 41 10.4-7a Oct. 1977
~m (C b. Suspended and dissolved solids which may be introduct.d by small leakages of circulating water through the condenser tubes.
- c. Solids carried in by the makeup water and miscellaneous drains.
At the design condensate flow and with circulating water inleakage within the capacity of the system the Condensate Clennup System wIlI produce of fIuent of the qual Ity required by the Steam Generation System given in Section 5.5.3.11.
The Condensate Cleanup System polishes 100 percent of the condensate and is designed f or continual perf ormance. Total head loss f rcen inlet to outlet terminal points will not exceed 50 psi. In addition, the syste:n design provides for removing impurities in the condensate caused by an Intermittent inleakage in the condenser of cooling water frcrn the Circulating Water System.
To assure that condensate /feedwater quality is maintained within the safe limits, the condensate polishing system is provided with capability to operate, without regeneration, for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, with permissible cool ing water inl eakage of 90 GPM. The basis for this operating condition is as f ol lows:
a) one vessel on iIne and at a point of exhaustion under normal operating conditions.
b) one vessel on lIne and haf f exhausted under normal operating conditions.
b V c) one vessel in standby and f ully regenerated.
The condensate polishing system is also capable of operating for 5 days, with cooling water inleakage of 5 GPM and meet the ef fluent requirements for operation above 5% Power. The basis f or this operating condition is that regeneration shalI not be required for more than one service vessel per day.
The circulating water analysis is given on Table 10.4-2. The maximum analysis shall be used for the condenser leak design conditions.
The design also provides a bypass of the entire Condensate Cleanup System. ;
The design of the polisher units and regoneration equipment is based on the :
Condensate Cleanup System operating on the hydrogen cycle.
]
i Piping is f urnished in accordance with ANSI B31.1 Power Piping. Pressure l
- vessels which f all within the jurisdiction of ASE Section Vill are f urnished In accordance with that code. The Condensate Cleanup System is not safety related, 10.4.6.2 Descriotion l
The Condensate Cl eanup System consists of three bal f-capacity lon' exchanger,
( each containing a bed of mixed resins in the proportion of one-part cation l resin to one-part anion resin by voinne. The third (spare) lon exchangers may l
! (q V
2 be placed in service if desired or in the event of a condenser tube leak. The Condensate Cleanup System is piped directly into the f eedwater cycle downstream of the condensate pumps.
Amend. 74 10.4-8 Dec. 1982 l
t
Each resin bed is periodically transf erred f ran the ion exchanger to an external backwash and regeneration system as required for rcrnovel of solids and/or chemical regeneration.
Spare charges of resins may be held in the externti backwash and regeneration system for immediate replacement of the exhausted tads so that an exchanger may be marie avail able f or prompt repl acement of a spent exchanger. An ef fluent strainer in the discharge pipin,r f rce each ion exchanger protects the j
f eedwater system against a discharge of resin in the event of an underdrain failure.
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O Amend. 74 10.4-8a Dec. 1982
The director of each center is responsible for directing his staff in carrying (V") out their respective responsibilities. He is delegated the authority during emergencies to locate, direct, and dispatch the personnel and equipment necessary to carry out his staff's responsibilities.
The purpose of the CECC and associated CECC staff is to provide the facilities and manpower for evaluating, coordinating, and directing the overall activities involved in coping with a radiological emergency.
During an emergency, the CECC Director and his staff will review the response to the emergency by WA and the appropriate State agencies to ensure that an effective and cooperative effort is being made. The CECC Director, after consultation with the Office of Health and Safety CECC representative, is responsible for providing TVA's recommended protective actions to the appropriate State officials.
The CECC staff will coordinate with all other WA emergency centers to ensure an effective WA effort ill response to an accident situation. The CECC staff will also provide an accurate description of the emergency situation for TVA management and public information. In addition, the CECC will coordinate with offsite Federal agencies, such as the Nuclear Regulatory Commission (NRC) and Department of Energy (DOE), to ensure availability of additional outside resources to TVA.
The DNPEC staff provides support services during a radiological emergency to the affected plant. Support services may be provided by utilizing any necessary manpower and equipment under the direct control of the Division of Nuclear Power. If the division is unable to provide adequate serches or support, requests will be made for additional support to other WA divisions, local agencies, or government installations as may be required.
The Muscle Shoals Emergency Control Center (MSECC) supports the CECC by performing environmental radiological monitoring and dose assessments and by recommending protective actions for the public to the CECC. In performing these functions, the MSECC assists the Tennessee Department of Public Health in evaluating the population exposures resulting from radiological emergencies. Real-time meteorological data will be used in dose assessment related to actual and potential releases of radioactivity. The MSECC staff directs offsite environmental monitoring for the Tennessee Department of Public Health and continues monitoring activities until a State Field Coordination Center is established to coordinate the offsite environmental monitoring effort. The MSECC will continue to evaluate the need for monitoring assistance to the Tennessee Department of Public Health. The State may request assistance from the appropriate DOE Operations Office in accordance with Interagency Radiological Assistance Plan (IRAP) for additional support. The MSECC will monitor the radiation protection problems in the plant during emergencies to provide guidance, manpower, and equipment to the Plant Health Physicist as required to control and mitigate these problems.
The Knoxville Emergency Control Center (KECC) serves as the focal point for all essential support activities involving TVA Knoxville offices. The TVA Office of Engineering Design and Construction (OEDC), Division of Engineering 13.3-4 Amend. 77 May 1983
Design (EN DES), has been delegated overall responsibility for the KECC and for providing technical support during and following a radiological emergency.
This section addresses the EN DES responsibilities for technical support during emergency conditions.
The KECC also serves as the comunication center for other essential WA offices such as the TVA Board of Directors, the General Manager, the Nuclear Safety Review Staff, and the Information Office.
Real-time meteorological data will be available in the Control Room. Technical Support Center, CECC, MSECC, and State of Tennessee Emergency Operations Center.
The radiological emergency comunications network will consist of a combination of commercial telephone circuits, radio, and microwave circuits.
South Central Bell Telephone Company lines will be used as the primary means of comunications during radiological emergency situations between plant, CECC, DNPEC, MSECC, KECC, and appropriate Federal and State agencies. This system will be augmented by the WA Private Automatic Exchange (PAX). Hard copy data transmission will be accomplished by facsimile from the CRBRP to the l DNPEC, the CC, and the MSEC. The hard copy transmission is then followed up and verified by redundant telephone comunications. TVA will provide the necessary interfaces to the CRBRP Technical Support Center Data System for transmission of necessary data to IVA emergency centers, as appropriate.
The primary means of notification of plant and offsite personnel is the comercial telephone circuits. Additionally, pocket pagers are provided to certain key individuals in the emergency organization.
Figure 13.3-1 illustrates the relationship between the TVA emergency centers and depicts the interface among TVA, Federal, State, and local agencies.
13.3.3 Coordination With Offsite Groups TVA will have agreements with other Federal agencies to assist in the evaluation and control of any radiological emergency. These agreements will include such agencies as the Department of Energy (DOE), Oak Ridge Operations Office, and the National Aeronautics and Space Administration (NASA), Marshall Space Flight Center. The CECC staff may request assistance from these outside agencies as required. The Site Emergency Director will be responsible for notification of NRC's regional office of Inspection and Enforcement.
Agreement has been made with the State of Tennessee, Tennessee Emergency Management Agency (TEMA), to provide for planning and conduct of emergency operations for emergencies at CRBRP (Reference 1). 1EMA is responsible for coordination of the efforts of all state agencies and local governments in the development of response plans that have an impact beyond the capability of a single agency or local goverment to control. The actual agreements and arrangements involved with such state agency and local government will be specifically defined in the State of Tennessee CRBRP Radiological Emergency Response Plan which will be provided in the CftBRP FSAR. The TVA CRBRP O
13.3-5 Amend. 77 May 1983
Radiological Emergency Plan will utilize the liaisons already established in developing the Sequoyah and Watts Bar Radiological Emergency Plans with the State of Tennessee. EMA will notify the State of North Carolina, Department of Crime Control and Safety, surrounding states as necessary, and coordinate assistance from the various Tennessee agencies.
TVA will maintain liaison with the Tennessee Emergency Management Agency, particularly with respect to the availability of emergency services. The Tennessee Emer6ency Management Agency will inform these agencies of actions to be taken under their respective statutory authority and assist them .in developing emergency procedures. TVA will provide any necessary training for local fire and police departments, ambulance services, and hospitals in radiological hygiene practices and recognition of raidological hazards. The attached Table 13.3-1 lists the organizations that will be participating in the Clinch River Breeder Reactor Plant Radiological Emergency Plan.
s
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13.3-Sa Amend. 77 May 1983
l Arrangements will be made for local ambulance service and fire departments to provide back-up emergency service as required to the plant. Agreements will also be culminated between TVA and a nearby hospital to provide emergency treatment to irradiated or contaminated patients as required. TVA will assist in training ambulance attendants and hospital personnel in this type of treatment and will ensure that adequate equipment is made available as necessary. An agreement will also be made with the Radiation Emergency Assistance Center and Training Site (REAC/TS) Facility operated by the Oak Ridge Associated Universities (0RAU) for emergency treatment of severely contaminated or irradiated personnel.
13.3.4 Emergency Action Levelg
'IVA will utilize the following emergency classification:
- 1. Notification of Unusual Event
- 2. Alert
- 3. Site Emergency
- 4. General Emergency This system of classification will be consistent with the system used by State and local emergency organizations.
A Notification of Unusual Event will provide early and prompt notification of minor events which could develop into or be indicative of more serious conditions which are not yet fully realized. The purposes of Notification of Unusual Event are to (1) assure that the first steps in activating emergency organizations have been carried out, and (2) provide current information on unusual events.
An Alert class will be indicated when events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. The purposes of the Alert class are to (1) assure that emergency personnel are readily available to respond if the situation becomes more serious or to perform confirmatory radiation monitoring if required, and (2) provide offsite authorities current status information.
A Site Emergency will be declared when events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. The purposes of the Site Emergency class are to (1) assure that response centers are staffed, (2) assure that monitoring teams are dispatched, (3) assure that personnel required for evacuation of nearsite areas are at duty stations if the situation becomes more serious, and (4) provide current information for and concultation with offsite authorities and the public.
A Ceneral Emergency will be declared when events are in progress or have occurred which involve actual or imminent substantial core failure with potential for loss of containment integrity. 'Ihe purposes of the General Emergency class are to (1) initiate predetermined protective actions for the 13.3-6 Amend. 77 May 1983 l
l l
O public, (2) provide continuous assessment of information from the site and
() offsite, and (3) initiate additional measures as indicated by releases or potential releases of radioactivity.
Recognition of the emergency class which may result from many initiating conditions will be primarily a judgement matter for plant operating personnel.
The initiating conditions used for recognizing and declaring the emergency class will be based on specific measured values or observable conditions de. fined as Emergency Action Levels (EAL). These can be combinations of specific instrtment readings (including their rates of change), annunciator warnings, time periods certain conditions exist, etc. The specific instrtment readings and parameters required for determination of these EAL's will be detailed in plant operating instructions. These EAL's will be used as thresholds for determining the emergency classifications.
13.3.5 Protective Measures 13.3.5.1 Plant In the event of an unplanned release of radioactivity or sudden increase in radiation levels, it will be the responsibility of the Site Emergency Director to make the decision concerning the necessity for building and/or area evacuation. In arriving at this decision, the primary considerations will be personnel eafety. The emergency siren will be used to initiate the assembly of personnel. The public address system will be used to evacuate specific areas. Upon hearing the emergency signal, all persons in the plant will go to O their preassigned areas to wait completion of radiological surveys and further V instructions. If only a specific area is to be evacuated, personnel in that area will evacuate to a safe area. Employees will be released from their assembly points when the Site Emergency Director determines it is suitable.
Should evacuation of unnecessary site personnel be considered, routes, specific radiological conditions, traffic density, and weather conditions will be evaluated prior to directing their evaluation. Any necessary personnel decontamination identified during area evacuation and accountability will be accomplished prior to evacuation from the site.
13,3.5.2 Offsite Through the assessment actions of the MSECC, the actual or potential offsite environmertal conditions will be known. The State and local agencies will be responsible for implementing actions to protect the health and safety of the public. TVA recommends protective actions to these agencies but the State 'and l local governments are responsible for deciding if any actions are needed and what they should be. WA will assist State and local governments as necessary to implement protective actions for the public. WA will also provide a prompt notification system for State and local governments to alert the public within a 10-mile area around the plant that protective actions may be required.
13.3.6 Review and Updatina m The CRBRP-REP will be reviewed annually by the Division of Nuclear Power and the Division of Occupational Health and Safety for accuracy, completeness, 13.3-7 Amend. 70 Aug. 1982
optrational readiness, and compliance with existing regn ations. All holders of these plans will acknowledge in writing, receipt of all changes.
13 3 7 Medical sunoort l The CRBRP-REP will include a description of s dical facilities at the plant and arrangements mada with other facilities to provide additional support.
One ambulance will be maintained at the site. Medical consultation will be available from TVA doctors in Chattanooga and other areas. Pembers of the plant emergency teams will be trained in first aid.
Arrangements will be made with a local hospital and with attending physicians for the emergency treatment of contaminated, injured, and exposed individuals.
The Oak Ridge Associated Universities REl.C/TS has agreed to provide treatment to severely contaminated or exposed individuals.
Arrangements will be made with a local private ambulance service to provide emergency service as required to the plant and affected areas in the event that more than one ambulance is required.
Figure 13 3-2 depicts the p: oposed locations of the Technical Support Center, the Operations Support Center, and the layout of the medical facilities and the personnel decontamination facilities.
13 3 8 Exercises and Drills An exercise will be conducted at Clinch River Breeder Reactor Plant prior to the issuance of the full power operating license to test the CRERP-REP, State, l and local plans. Exercises will be conducted yearly thereaf ter. Drills will be conducted in the following areas:
- 1. Medical Emergencies
- 2. Radiological Monitoring 3 Radiochemistry l 4. Transportation l 5. Radiological Dose Assessment l 6. Fire 13 3 9 Trainine TVA will provide General-REP training to all plant personnel and specific training to emergency response personnel. This training provices emergency staff personnel with general knowledge of the emergency plan. Training on specific duties and emergency responsibilities will be provided personnel as ne cessary. This training will be such that each of these individuals will have a working knowledge of the emergency plan and his rasponsibilities and actions upon declaration of an emergency.
O 13 3-8 Amend. 70 August 1982
l f
Training and periodic retaining will be provided to those offsite agencies who
- f. may be involved during an emergency, and will include procedures for notifica-tion, basic radiation protection, their expected roles, and site access proce-dures, as applicable. The Division of Occupational Health and Safety will provide for training to fire, police, ambulance, and hospital personnel from l agencies with which WA has agreement letters.
13.3.10 Recovery and Reentry The CRBRP-REP will provide for the development and implementation of detailed recovery and reentry plans based on evaluation of conditions existing at the time. Recovery and reentry will be a deliberste, thoroughly planned evalua-tion and all procedures developed will be reviewed by the Plant Operations Review Comittee prior to implementation.
13.3.11 Imolementation Operating instructions promulgated in the plant operating manual will be used to control plant operations during normal operating conditions. Abnormal operating instructions and emergency operating instructions which are com tained in the Plant Operating Manual will be used to specify the manipulation of controls of the plant during conditions requiring protective measures to be taken to place the plant in a safe condition. 'Ihe abnomal and emergency instructions will contain assignments of responsibility for the performance of specific tasks not otherwise established by plant practices and instructions.
py Plant instrmentation indications requiring implementation of emergency and abnormal operating instructions will be specified in these instrrctions.
Emergency action levels, also based on plant instrumentation indication, requiring implementation of the CRBRP-REP for protection of personnel' and the environment are specified in the emergency plan.
Specific actions required of offsite WA support groups will be delineated in the CRBRP-REP.
Instructions for medical treatment and handling of contaminated and exposed individuals will be contained in the CRBRP-REP.
Equipnent requirements, includirg communications equipment, for implementation of the emergency plan will be contained in CRBRP-REP. Storage and calibration requirements will be specified. Alarm signals will be described in the plant procedures.
Instructions for restoring the emergency situation to normal. from the stand-point of the hazard to personnel, plant safety, and the environment, will be contained in the CRBRP-REP and the emergency and abnomal opcrating instruc-tions. Instructions for repair of plant equipment or structures will be pre-pared after evaluation of the damage or malfunction involved.
I Reference to Section 13.3
- 1. Letter, Tanner, E.P., Director, Tennessee Emergency Management Agency, to (J
Green, H. J., TVA, July 6,1982. (Supplied to NRC in letter HQ:S:83:242, John -R.-Longenecker to J. Nelson Grace, April 8,1983.)
13.3-9 Amend. 77 May 1983
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TABLE 13.3-1 PARTICIPANTS IN CRBRP RAID 0 LOGICAL EMERGENCY PLAN Tennessee Emergency Management Agency Tennessee Department of Public Health Tennessee Department of Agriculture Tennessee Department of Public Welfare Tennessee Department of Safety Tennessee Department of Conservation Tennessee National Guard Tennessee Game and Fish Commission Tennessee Department of Transportation City and County Officials of Roane, Anderson, Loudon, Morgan, and Knox Counties Sheriff's Department of Roane, Anderson, Loudon, Morgan, and Knox Counties Civil Defense Coordinators of Roane, Anderson, Loudon, Morgan, and Knox Counties Local Police Departments Local Ambulance Service Local Fire Department Radiological Emergency Assistance Center Training site (REAC/TS)
Oak Ridge Hospital of the United Methodist Church (ORHUMC)
Department of Energy (DOE)
National Aeronautics and Space Administration Nuclear Regulatory Commission Environmental Protection Agency l Federal Emergency Management Agency O
13.3-10 Amend. 77 May 1983
C. Estimated Reaction and Response Times
- 1. The time required for the initial accident assessment of the most O
\V serious design basis accident may require 15 minutes. This time is an estimate based on the operation of the reactor instrumentation used to follow the course of accidents. Based on TVA's experience, the time required to perform an initial dose projection and notify offsite authorities can be accomplished in 15 minutes.
For the most serious design basis accident, the projected two-hour doses at the exclusion area boundary do not reach the protective action guide level for evacuation.
- 2. The time required to warn all resident and transient persons in any evacuation sector will conform to the requirements of 10 CFR 50, Appendix E-1982.
- 3. The estimated elapsed time, after the initial warning, to evacuate the 2-mile emergency planning zone (EPZ) is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The estimated evacuation time for the 5-mile EPZ is 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, 35 minutes. Tne estimated evacuation time of the 10-mile EPZ is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 45 minutes.
Each estimate contains a 1-hour 50-minute preparation time factor.
- 4. These evacuation time estimates wcre prepared by the Traffic Management Division of the Tennessee Department of Transportation.
- a. Figures 13.3A-5 and 13.3A-6 are maps showing all roads within 10 miles of the Clinch River Project. Also indicated are the 2 ,
5 , and 10-mile EPZ.
- b. Table 13.3A-1 shows the transient and resident pcpulations in the 16 directional sectors within 10 miles of the Clinch River Project. This table uses 1980 census data.
- c. Table 13.3A-2 shows the estimated transient and resident populations in the 15 directional sectors within 10 miles of.the Clinch River Project. This table uses the projected pooulation figures for year 2020. The projected population figures come from a report prepared by the Firm of Dames and Moore dated June 16,1981.
- d. Private automobiles will be the primary means for evacuating the population. Buses are expected to be used to evacuate schools and other institutions. This procedure will be specifically addressed in the CRBRP-REP.
- 5. Table 13.3-1 gives the agencies involved in the CRBRP emergency plan.
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13.3A-3 Amend. 77 Mcy 1983
TABLE 13 3A-1 MAXIMUM RESIDENT AND TRANSIENT POPULATION DISTRIBUTION WITHIN 10 MILES OF THE DEMONSTRATION PLANT FOR CENSUS YEAR 1980 lh Radial Interval (miles)
Sector Designation D-1 M 2-1 %4 a-5. 5-10.
N 0 0 14 184 0 2,000 NNE O O O O 22 4,400 NE O O O 8 80 7 ,1 91 ENE 10 10 0 8 0 4,728 E 20 30 50 398 3,568 5,172 ESE 20 30 50 187 159 2,300 SE O 59 50 460 110 7,200 SSE O 300 79 90 320 2,000 S 0 89 50 120 160 1,120 g
SSW 10 69 50 80 90 936 SW 20 80 119 110 140 1,292 WSW 20 70 80 193 340 5,000 W 0 130 114 110 991 6,764 WNW 10 94 170 10 60 4,676 NW 30 44 0 10 40 3,972 NNW 10 514 316 850 120 1,100 Sum for Radial Interval 150 1,519 1,142 2,818 6,200 59,851 Accumulative Total up to Radius Indicated 150 1,669 2,811 5,629 11,827 71,680 0
13 3A-4 Amend. 70 August 1982
Results from FORE-2M analysis are given in Figures 15.1.4-1, 2 and 3 and Table 15.1.4-1 for a 60( step reactivity insertion occuring at the worst time during t ') the SSE (see Section 15.2.3.3.1). Comparisons of the heterogeneous core V results are made with data for a homogeneous core previously reported in this section. This previously r eported data updated earlier data for the homogeneous coro analyzed in Section 15.2.3.3. The figures show the maximum hot rod temperatures for F/A #52 and 101 in the heterogeneous core as compared to similar data for F/A #6 and #8 in the homogeneous core. In the homogeneous core, F/A #8 had the maximum cladding temperature hot rod and F/A #6 had the maximum fuel temperature het rod. In the heterogeneous core, F/A #101 has the maximum fuel temperature and F/A #52 has the maximum cladding temperature. A comparison of pertinent parameters is given by Table 15.1.4-1. Although the
'smperatures for the heterogeneous core are somewhat higher, the results are well within the limits given by Table 15.1.2-2 (no sodium boiling) for extremely uniikely faults. In addition, the maximum coolant temperature is less than 1600 F. This provides considerable margin to the coolant saturation temperature which is greater than 1800 F when this maximum is attained.
One important note with regard to the above comparison is that credit has been taken in the heterogeneous core hot rod evaluation for having a programmed ,
startup to enhance the power-to-melt (i.e., improved fuel restructuring and 1 gap conductance) by using LIFE-lil analyses. The homogeneous core studies used fuel rod conditions calculated using P-19 data which assumes a " fresh rod" with no previous operation. Section 4.4.3.3 of the PSAR addresses how programmed startup is achieved at;d the consequential improvement on the minimum power-to-melt. Since the proposeo programmed startup is still not optimized, further improvement can be achieved through optimization of the p programmed startup which is scheduled for the FSAR as discussed in Section Q 4.4.3.3.
Similar data to that described above for the hot fuel assembly rod is not available for the hottest blanket assembly rod. However, it has been found from past analyses shown in Section 15.2, that the temperatures for the blanket size rod are significantly less for the SSE step reactivity insertion type of event. This is due to the large thermal inertia of the significantly bigger blanket size rods. With the extremely quick power rise for the 60d step SSE case, there is insufficient time for the temperatures to increase as much as they do for the smaller fuel size rods. Ttus, the fuel hot rod
( temperature response represents the worst case thermal transient effect to the
- core. These results support a qualitative conclusion that the design changes incurred since tho original PSAR submittal are not expected to significantly
( change the results reported in Section 15.2.
l 15.1.4.2 Undercooling Design Events Section 15.3 covers the analysis of undercooling design events. Subsequent to the earller analyses, various design changes to the plant have taken place.
The most important design change is, of course, the core design change to the heterogeneous core configuration.
V 15.1-107 Amend. 77 May 1983
An impact assessment of the significant design changes as discussed below indicated that their offects on the conecquences of undercooling events are either positive or insignificant. Nevertheless, to positively demonstrate the adequacy of the current plant design configuration against undercooling events, a detailed re-analysis was undertaken. This re-analysis was based on a ' worst-case' event selected in a systematic manner by reviewing the earlier results of the undercooling event analyses reported in Section 15.3. Unlike the overpower transients, undercooling events typically involve very smal1 fuel temperature increases as compared to that of the cladding. The selection of the ' worst-case' event was therefore primarily based on the worst consequence in terms of maximum cladding temperatures. The ' worst-case' undercooling event so established was then re-analyzed consistent with the current plant design. Details of and the results from this re-analysis are j provided herein following an item-by-item discussion of design changes below.
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0 15.1 -107 a Amend. 66
! March 1982
The significant design changes with respect to the undercooling events and their expected offects are:
- 1. The heterogeneous reactor core arrangement is described in Chapter 4.
Although the heterogeneous core arrangement is substantially different from the homogeneous arrangement, the nuclear and thermal-hydraulic design constraints are very similar. Since no design basis undercooling event presented a significant challenge to the homogeneous core (analysis indicated a margin of approximately 200 F to the onset of sodium boiling for the homogeneous arrangement), the same is predicted to hold true for the heterogeneous arrangement.
- 2. A minimum flow coastdown requirement has been specif!ed for the primary sodium coolant pumps leading to a corresponding minimum pump-and-drive system inertia requirement which is larger than the value used in the PSAR Section 15.3 analyses. This also applies to the intermediate sodium coolant pumps which use an identical design.
The impact of the increased net primary flow during the coastdown is in the direction of less severe consequences for undercooling events through additional heat removal.
- 3. Refinements made in primary sodium piping layouts which either change flow resistance or transport time are expected to have minimal, if any, effect on the undercooling event.
- 4. The differences in IHTS piping configuration between that considered for Section 15.3 analyses and the present design are expected to have fh V
minimal ef fects on the undercooling transients of Section 15.3. The changed transport delay further increases transport time for steam-system induced transients to reach the reactor inlet.
- 5. The current primary system cold Ieg check valve design is expected to allow limited reverse-flow leakage in a loop in which the pony motor is not operating. An analysis has been performed using the DEMO Code which shows that adequate core decay heat removal can be maintained with only a single primary pony motor operating, permitting reversed-flow in the other two loops. Under these conditions, loop thermal heads effectively limited the reversed flow well within the check valve leakage specification and, in fact, maintained a small forward flow during most of the transient duration.
Consequently, adequate safety performance is expected for events in which check valve leakage is a factor.
- 6. The IHX design has been changed from a remc,vable tube bundle to a fixed tube bundle design. This change does not affect the significant IHX thermal ana hydraulic parameters used in the Section 15.3 analyses.
- 7. A length increase has been made in the steam generator modules since completion of the PSAR Section 15.3 transient analyses. Neither water-side nor sodium-side pressure drop changes are large. The moduie surface increase is in the direction to decrease severity of g undercooling transients, b
15.1-108 Amend. 66 March 1982
- 8. Piping layouts have been optimized and simplified on the water / steam sida. Two drum headers have been eliminated and replaced by direct-to-drum pipe connections, in conjunction with the use of an V annular-Inlet girth-baffle in the drum (see item 9 below). Available elevation difference to assist recirculation flow has been increased (drum raised 7.0 feet).
The larger-diameter nozzles at the drum increase the available discharge area in case of large steam pipe breaks immediately at the drum. It should be noted, however, that in general, effects of severe steam-side events reach the reactor well after the reactor has been shut down by the P! ant Protection System because of the extended transport time at pony-motor flow rates. Consequently, no safety problem results at the reactor.
- 9. As noted above, the steam drum has been redesigned using an integral annular girth baffle at the inlets from the evaporator, with elimination of the steam inlet and exit headers (the recirculation exit header remains). The possible offect on discharge area available during a steam pipe break were noted above. Increased drum elevation is expected to increase water recirculation for cases in which the recirculation pump it not operating. (See pertinent comments in item 8 Immediately above.)
- 10. A revised head curve for the steam system recirculation pump is now available. Comparizon of these revised characteristics with those used in the Section 15.3 analyses indicate that for steam system A transients where the recirculation pump head has an etfect on results, the former characteristics are more conservative. (Consequently, design transient analyses have continued to use the prior characteristics.)
- 11. The Plant Control System design has progressed significantly since the Section 15.3 analyses were performed but the control concept has not changed. For the analyses of Section 15.3, however, early trip action by the Plant Protection System precluded Control System effects on the transient results (I and C tolerances and deadband are reflected in the conservative starting conditions for the transient analyses). The Plant Protection System functions having an action during the Section 15.3 events are tabulated in Table 15.1.3-1. The functions given remain valid, with the exception that the 30% trip mismatch setting (of fset) on flux-to-flow may bi; reduced to accommodate events leading to increases in power-to-flow ratio that might level-off just before the trip level. The effect of this reduction will be to decrease time required to trip and thus will lessen the effects of the resulting transient. In addition, a requirement has been established for the secondary rod system to generate a Trip Signal within 0.8 seconds after loss of electrical power on the pump busses.
(R)
LJ 15.1-109 Amend. 66 March 1982
- 12. Two Control and Protection System developments for Balance of Plant having a potential bearing on the Section 15.3 transients are:(a) a better definition of the feedwater controls and (b) incorporation of a delayed turbine trip following reactor plant trip. The latter is used lh to reduce steam pressures immediately following a plant trip to prevent saf ety-valve operation. These changes will not affect reactor conditions immediately af ter shutdown, because of the extended sodium transport times in the PHTS and lHTS at pony motor flowrates.
- 13. In the auxiliary feed system, regenerative heating of the auxiliary feedwater has been removed, with the result that cold auxiliary feedwater is now injected. For cases in which recirculation pump power is not available, mixing of the drum water with the cold feedwater may be reduced through stratification effects, resulting in lower transient cold-leg temperatures.
The re-analysis of a worst-case event is discussed below. The worst-case event selected was the Loss of Of f site Electrical Power (Section 15.3.1.1) as it resulted in the worsv maximum cladding temperature. The re-analysis was perf ormed with DEMO-4 and FORE-2M that have incorporated al l the design changes discussed above. The same PPS design data and conservative approach as described in Section 15.1.4.1 were al so used f or this analysis.
Parametric studies were performed with the earlier version of FORE-2M (Reference 1) to substantiate a worst-case canbination for the selected event for the nuclear power variation calculation. The objective was to seek a set of conditions that leads to the highest power (or the slowest decrease in power) during the transients. Conservatively, this model used a reduced level of core detail (from the 7 radial by 7 axial mode capability of the code) where the total core wide Doppler f or the f uel and blanket regions was included in the fuel region. A base case nuclear model was establishec using the following conditions:
o Minimum C of fuel; o Longest fiow coastdown of the primary pumps; o Zero decay heat; o Maximum fuel / cladding gap conductance; o Zero sodium coolant density feedback; and o Maximum Doppler coefficient All other feedbacks which ere negative (such as fuel expansion, cladding expansion, core radial expansion and bowing) were conservatively neglected.
Results of this base case and the ef fect of each worst case parameter are given by Table 15.1.4-3. Here the variation in the neutronic power is shown for each significant parameter. This demonstrates the importance of each condition in establishing the base case, in addition to these conservative studies, the base case was repeated using the current FORE-2M capability which considers all core regions. With the same base case conditions, a maximum neutronic power variation of 0.1276* was found as compared to the value of 0.1327
- I n Tab l e 15.1.4-3. Likewise, a nominal data case was run with this model and a maximum neutronic power variation of 0.1265* resulted. This confirms the conservatism of the base case model.
L *Value quoted at 2 seconds into transient for comparative purposes. l 15.1-110 Amend. 66 March 1982
l l
- l. 15.4.1.3.2 Consecuences of Blockage of a Core Assembiv The result of ' partial blockage of the f uel rod bundle inlet is a reduction in assembiy fIow rate as Illustrated in Figure 15.4.1.3-2. Whlie existence of extensive inlet blockage is quite uniIkely, as discussed earller, its effect on coolant flow is also relatively small. The Inlet flow area has to be more than 50% blocked before a signif icant reduction In ' flow takes place. As shown in Figure 15.4.1.3-1, a blockage in excess of 80% is required to raise the coolant outlet temperature by mere than 2000F. Blockages of this magnitude are extremely improbable and no mechanism or source can be postulated by which such a blockage could occur.
An analysis was perf ormed to determine the maximum cladding temperature due to complete blockage of a fuel assembly outlet nozzle during refueling. It was found that, even using a consc vative analysis, the maximum cladding temperature for thIs accident (lf It occurs af ter 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> f rom shutdown) would be less than that for operating conditions.
15.4.1.3.3 In-Core Passive Local Blockage (Non-Heat Generating)
The increase in the wake temperature behind a blockage depends on the axial location, radial location, and the blockage size. Flow rate and heat flux are also important parameters. . Two basic types of blockages have been identif ied (Ref. 34); a three-dimensional center or of f-center blockage, and _a two-dimensional edge blockage. For otherwise equal conditions, a two-dimensional blockage is worse because the wake is larger and the residence time, , is
. larger. This f actor has been verified experimentally (Ref. 35 and 36). The O' reason a two-dimensional blockage has a larger. wake is because the drag coefficient Is Iarger.
Calculations were made to ' determine the average and maximum fluid temperature in the wake region behind a central six channel blockage. Cl addi ng temperatures were also calculated. The results are shown in Table 15.4.1.3.3-1. Methods used to calculate the results are similar to those discussed in Section 15.4.3.3. Heat transfer coef ficients were obtained from the sodiuin flow blockage data (Ref. 36, and 37).
It can be seen from the table that at either the midplane or the core exit plane, the postulated six channel center blockage would cause temperatures which are considerably lower than the temperature correspo.nding to prompt cladding failure.
O Amend. 76 15.4-33 March 1983
If the six channel blockage occurred on the edge of the fuel assembly, the hydraulics would still be quast-three dimensional. Although the dimensionless residence time tR would be larger, the colder edge fluid flowing by the small blockage should partially compensate for the higher R t . Only if the six-channel blockage occurred in a corner, would there truly be a two-dimensional effect, possibly producing higher wake and cladding temperatures than shown in Tcble 15.4.1.3.3-1. Because of uncertainties in existing edge blockage data and prediction models, two-dimensional definitive blockage, calculations for edge channels were not made. An appropriate evaluation Indicates however that even a true two-dimensional corner blockage of six subchannels does not cause cxcessive mid-wall cladding temperatures.*
15.4.1.3.4 In-Core Active Blockage (Heat-Generating)
An in-core heat-generating blockage can only arise from gross fuel failure.
Dalayed neutrons from decaying fission products which are released when the failure occurs will be detectable and will alert the operator within one minute after the event. However, assuming that failure is not detectable, a hrat generating blockage would result, which, contrary to the basic essumptions made earlier concerning non-heat-generating blockages, would be porous. A general comparison of porous and nonporous blockages will be discussed first. The discussion is applicable to both core, blanket, and control assemblies.
Porous Blockages As cited previously, a mechanism which can cause a non-heat-generating complete blockage of one or more flow channels within a core, radial blanket or control rod assembly has not been identified. Nevertheless, if a complete blockage of one or more flow channels does occur by a gradual local buildup of debris, the blockage will be porous, in fact, even a gradually formed heat-generating blockage should be porous. The question to be answered is, given a porous and a nonporous blockage of otherwise similar characteristics, which one causes higher coolant and/or cladding temperatures in the region of the blockage? In the light of existing data this question is answered in the following paragraphs.
^
Leakage ilow through a blockage Increases the static base pressure immediately downstream of the blockage (Ref. 35, 38, 39, 40). The result is a decrease in the drag coefficient across the blockage and a decrease in the volume of the near-wake region. As the leakage area ratio SL increases (Data (Ref. 36) showed no safety problems for 14 edge channels blocked in a 19-rod fuel assembly.
O 15.4-34 Amend. 76 flarch 1983
CL INCH RIVER BliEEDER REACTOR PLANT A DESCRIPTION OF THE OWNER O- QUALlTY ASSURANCE PROGRAM A
TABLE OF CONTENTS PAGE NQ.
0.0 INTRODUCTION
17A-1 0.1 SCOPE 17A-1 0.2 BASIS 17A-1 0.3 APPLICATION 17A-1 1.0 ORGANIZATION 17A-1 1.1 FUNCTION 17A-1 1.2 RESPONSIBILITY AND AUTHORITY 17A-2 1.3 ORGANIZATIONAL ARRANGEMENTS 17A-3 1.3.1 Oualltv Verification Branen 17A-4 f-~s 1.3.1.1 Surveillance 17A-4 (s,)
1.3.1.2 Insoection 17A-4 1.3.1.3 Audit 17A-4 1.3.2 Ouality Engineering Branch 17A-4 1.3.2.1 Planning 17A-4.
1.3.2.2 Reoorts . 17A-5 1.3.2.3 Records 17A-5 1.3.3 Ouality lmorovement Branch 17A-5 1.3.3.1 Nonconformance Control 17A-5 1.3.3.2 Trend Analvsis 17A-5 1.3.3.3 Training and Indoctrination 17A-5 1.4 OUALIFICATION REOUIREMENTS FOR OUALITY 17A-6 MANAGEMENT POSITIONS Assistant Director for Quality Assurance 17A-6 D
'J l 1.4.1 17A-i Amend. 77 May 1983
i TABLE OF CONTENTS (CCNTINUED)
PAGE NO.
l 1.4.2 Chlef, Qualltv Verification. Qualitv Engineering. 17 A-7 and Ouality Imorovement 1.5 COMMUNICATIONS 17A-7 l
2.0 QUAL ITY ASSURANCE PROGRAM 17A-8 2.1 OBJECTIVES 17A-9 2.2 EESPf((SJ tilL ITJ 17A-9 2.2.1 Dxtter 17A-9 2.2.2 Enrilr.JSitttts 17A-10 2.3 REQDj REMENTS 17A-10 2.3.1 Dysr_ajl 17A-10 2.3.2 Program Elements Executed bv the Owner 17A-11 2.3.3 Program Elements Delegated to Others 17A-11 2.3.4 3r_ansfsr of Recujr_erents to Others 17A-12 2.4 SURVE ILL AtiCf 17A-12 2.5 INTERFACE C00RDINATl0B 17A-13 2.6 PROGPIM MANAGEMEt1T 17A-14 l
3.0 DESIGN CONTPDL 17A-15 3.1 DhfiER_lFPLEMEt1T1tTJDN 17A-15 3.2 17A-20 REQUIREMENTS OF OTHER PARTICIPANTS 4.0 PROCUREMENT DOCUMENT CONTROL 17A-21 4.1 OWNER IMPLEMft[ TAT _ LOB 17A-21 4.2 EERDJEEMEtRS_Of OTHER PARTlCIPAtRS 17A-23 5.0 ltLS_TRUDI_ID115, PROCEDDRES AND DRAWINGS 17/-23 5.1 D;fNER IMPLEMENTATION 17A-23 5.2 REQU l REMEr(T.S_Df_0IEER PART l C 1 PANT _S 17A-26 17A-Ii 0
Amena. 70 Aug. 1982 j
CLINCH RIVER BREEDER REACTOR PLANT g-'g A DESCRIPTION OF THE OWNER
(/ QUALITY ASSURANCE PROGRAM
- 0. INTRODUCTION 0.1 SCOPE Contained herein is a description of the plans and actions by the Owner to assure the quality of structures, systems, and components of the Clinch River Breeder Reactor Plant (CRBRP). These plans and actions constitute the Owner's Quality Assurance Program.
0.2 BASIS The program described herein has been planned, structured and defined to fulfill the responsiblity for ultimate ef fectiveness and adequacy of the overall Project Quality Assurance Program. Responsiblity for establishing and executing portions of the overall Project Quality Assurance Program has been or will be delegated to others participating in the Project with ultimate responsibility for the adequacy of their performance retained by the Owner.
0.3 APPLICATION The Owner's Quality Assurance Program doscribed herein is applicable to the planning, design, procurement, mangf acturing, construction including testing, and operation of those safety-related structures, systems, and components
(}
%s -
identif ied in Section 3.2, 7.1 and 9.13 of this PSAR. For the Owner's purposes the program described herein is not limited to this application alone, however, and is or will be appropriately applied by the Owner to the CRBRP in its entirety including all structurer, systems and components where satisf actory perf ormance is required for the piant to operate reliably, safely and with minimum environmental ef fects. Specific examples of CRBRP non-safety related structures, systems, and components to which the appropriate portions of the owner's program applies include the Safety Parameter Display Console (TM! Item 1.D.2) and the Emergency Support Facilities (TMl item 111 A.1.2).
1.0 ORGANIZATION The Owner is RESPONSIBLE for the overall management of the Project to design, build, and operate the Clinch River Breeder Reactor Plant (CRBRP). The execution of this responsibility rests with the CRBRP Project Director, who is the principal operations officer of the Owner. Part of this responsibility is to assure that the plant is designed, built and operated in a way that will provide adequate confidence that it will perform satisf actorily in service.
To provide for this assurance, the Project Director has directed the establishment and conduct of an overall integrated quality assurance program which shall have the objectives, carry out the functions, and be executed as hereafter defined.
1.1 FUNCTION The f unctions that the Owner will perform in order to achieve the stated
( s,) objectives of the Quality Assurance Program and f ul fill its ultimate x/ responsibility for program adequacy are as follows:
17A-1 Amend. 77 May 1983
- 1. Development of an overall plan for conduct of the quality assurance program.
- 2. Assignment of program execution responsibility to appropriate program participants. These include the contractors and subcontractors who participate in the Project, as well as service contractors who only perform quality assurance activities.
- 3. Development of working plans and procedures to conduct program activities.
- 4. Organizing and staffing appropriately to implement program functions.
- 5. Implementation of Owner program activities.
- 6. Interf acing of major participant programs with the Owner's program.
- 7. Integration, coordination, eval uation, and approval of major participant quality assurance programs.
- 8. Development and impiementation of major participant programs where the Owner elects to retain execution responsibility in lieu of assigning it to another organtzation, e.g., Balance of P1 ant Supplter.
Those quality assurance f unctions which have been delegated to other organizations are explair.ed in Section 2.3 of this program description.
1.2 RESPONSIBILITY AND AUTHORITY l The Assistant Director for Quality Assurance, serving as head of the CRBRP Project Of fice Quality Assurance Division and reporting directly to the Project Director, is assigned responsibility for devising, recommending, establishment of, and assuring ef fective execution of the overali Project lQualityAssuranceProgram. The Assistant Director for Quality Assurance, is responsible that organizations, systems, and procedures at all levels will provide assurance that the Demonstration Plant is designed in accordance with requirements, is constructed as designed, and is operated in accordance with plans and procedures to achieve the demonstration objectives, in carrying out these responsibilities, he is authorized by the Project Director to:
- 1. Identify quality problems.
- 2. Initiate, recommend, or provide solutions through designated channeIs.
- 3. Verify implementation of solutions.
- 4. Determine the adequacy of faclittles and equipment provided to carry out approved procedures and instructions.
- 5. Authorize issuance of special instructions necessary to execute his responsibilitles.
17A-2 Amend. 77 May 1983 j
- 6. Notify responsible management of unsatisfactory work or unapproved practices and if necessary, stop unsatisf actory work or control further processing, delivery, or installation of nonconforming (U) materials.
The Assistant Director for Quality Assurance, is responsible for organizing l the overali Project Quality Assurance Program and for recommending further assignments of execution responsibility as appropriate. He shalI secure Charter statements from other major program participants describing their responsibilities and functions. He shall assure that in each major participant's organization the person responsible for quality assurance is granted sufficient authority to identify quality problems; to initiate, recommend, or provide solutions; and to verif y implementation of solutions.
l The Assistant Director for Quality Assurance, is responsible for recommending to the Director the organization and staffing plan for the Quality Assurance Division in the conduct of quality assurance practices necessary to fulfill the Owner responsibilities for establishment and adequacy of the program, in this position, he is responsible for the technical and administrative control (except for DOE personnel) of Individuals and groups within the Quality Assurance Division performing quality assurance activities or verifying adeqacy in the performance of quality assurance related activities of others.
The Project Office retains administrative control of DOE personnel.
1.3 ORGANIZATIONAL ARRANGEMENTS The Owner's organizational structure for performing quality-related activitle-
% associated with management, planning, design, procurement, construction and
,) operation of the CRBRP and the responsibility and authority of key positions within the organization are described in Section 1.4 of the PSAR. The Owner organization is shown in Figure 1.4-1.
To perf orm the assigned qual ity assurance f unctions, the Qual ity Assurance Division is organized as shown in Figure 17A-1. The Division is subdivided along functional Iines to perf crm qual Ity verif Ication, qual ity engineering and qual ity improvement. A description of the organizational elements are contained in subsequent paragraphs.
The personnel of the Division are located in Oak Ridge, Tennessee, in the CRBRP Project Office, Iocated at Jefferson Circle. Ali Owner program activI-ties will be executed by personnel wcrking out of that of fice. Verification that activities by contractors are in compliance with requirements will be performed through surveillance, inspection and audit practices by personnel working out of the Project Office.
The staf fing plan for the Division is based on the Project's work plan and schedule. It reflects the planned scope of work to be performed and the quality assuring activities to be applled to that work. The number of people and tneir required capabilities are identified through this planning, scheduling, and resource estimating process.
(Gv) 17A-3 Amend. 77 May 1983
1.3.1 Ouality Verification Branch The f unction of the Quality Verification Branch is to maintain surveillance h over quality assurance programs of major Project participants and to verify qual ity achievement in their work perf ormance. This branch is also responsible for monitoring the Owner's Quality Assurance Program to verify overal l adequacy.
The Qual ity Ver; f ication Branch perf orms three types of activities as described below:
1.3.1.1 Syrveillance Monitoring of the Project work and the quality assurance practices on that work is performed through this activity. It also serves as the focal point for interf ace coordination between the Owner's Quality Assurance Program and the qual ity assurance programs of other Project participants.
1.3.1.2 Insoection inspection of items and services or the monitoring of Inspections by others is accomplished through this activity. This activity incl udes perf ormance of selected civil, structural, electrical, mechanical and welding inspections, and nondestructive examinations.
1.3.1.3 Audit Planning and conducting Internal audits of the Owner's Quality Assurance Program and external audits of contractor Quality Assurance Programs is accompl ished through this activity. Scheduled and unscheduled audits are conducted.
1.3.2 Oualltv Engineering Branch The f unction of the Quality Engineering Branch is to plan, define, and develop the overal l Project Qual ity Assurance Program and the Owner's portion of that progr am. This f unction includes the preparation and maintenance of overall Project Quality Assurance Program requirements and Internal plans and procedures. This branch has lead responsibility for quality assurance program progress and status reporting and quality records management.
The Quality Engineering Branch perf orms three types of activities as described bel ow:
1.3.2.1 Planning Planning, program development and the documentation of plans and procedures f or conduct of both the overal l Project Qual ity Assurance Program and the Owner's Quality Assurance Program are perf ormed through this activity. A knowledge of industry and government standards and their appropriate application to the Project Quality Assurance Program is maintained.
O 17A-4 Amend 70 Aug. 1982
i.3.2.2 Reports
('s
() Establishment of quality assurance program progress and status reporting requirements and their maintenance is accomplished through this activity.
Collection of reports from branches of the Quality Assurance Division and the preparation of the Owner Quality Assurance Program Progress and Status Report is al so perf ormed.
1.3.2.3 Records Col lection, fil ing, and maintenance of quality records is perf ormed through this activity. The receiving, routing, and f il ing of working documentation within the Quality Assurance Division is performed. The Quality Records File will ultimately include records of the overall CRBRP Project.
1.3.3 Oualltv Imorovement Branch The f unction of the Quality improvement Branch is to provide needed training and-Indoctrination for Quality Assurance Division Personnel and to coordinate Project Office and Project-wide training and Indoctrination activities f or personnel perf orming qual ity-rel ated f unctions. This branch is also respon-sible for conducting activities wherein nonconformances are dispositioned and corrections to program deficiencies are nade to improve quality achievements and to prevent recurrence of nonconforming conditions.
The Quality improvement Branch performs three types of activities as described bel ow:
O
\s.) 1.3.3.1 Nonconformance Control Collecting unusual or abnormal occurrence reports, deviation requests, nonconformance reports, and deficiency citations, and processing them to satisf actory resolution is accomplished through this activity. A l og of quality problems identified Internally and by major program participants will be maintained and corrective actions recorded.
1.3.3.2 Trend Analvsis Activities, reports (audit, inspection, progress, status) and records are monitored through this activity to identify quality problems. Probl ems identified are studied and actions recommended to correct the problem, to improve quality achievements, and to improve the ef ficiency and ef fectiveness of quality assurance activities.
1.3.3.3 Training and Indoctrination Actions to acquaint Project personnel with the various elements of the quality assurance program and the practices needed to assure quality achievement are perf ormed through this activity. The certification of quality assurance personnel qualifications is perf ormed and personnel training and Indoctrina-tion within the Project Of fice and within other Project participant organizations is monitored to assure that:
\_/
17A-5 Amend. 70 Aug. 1982
- 1. Personnel perf orming activities af f ecting qual ity ace appropriately trained in the principles, techniques and requirements of the activity being performed.
- 2. Personnel performing activities af fecting quality are instructed as to purpose, scope, and implementation of governing manuals, pol icies, and procedures.
- 3. Appropriate training procedures are established.
- 4. Indoctrination and training activities are conducted in an effective s manner and achieve desired results.
- 5. For formal training and qualification programs, documentation includes the objective, content of the program, attendees, and date of attendance.
- 6. Proficiency evaluations or tests, as appropriate, are given to those personnel performing and verifying activi' lies affecting quality, and acceptance criteria are developed to determine if individuals are properly trained and qualified.
- 7. Certificate of qualifications clearly delineates (a) the specific functions personnel are qualified to perform and (b) the criteria used to qualify personnel in each function.
- 8. Proficiency of personnel performing and verifying activities af fect-ing quality is maintained through work experience or retraining with continued proficiency verified through reevaluating, reexaming, and/or recertifying in accordance with Project requiremants.
1.4 OUALIFICATION REOUIREMENTS FOR OUALITY ASSURANCE MANAGEMENT POSITIONS l1.4.1 Assistant Director for Ouality Assurance The Individual assigned to retain overall authority and responsibility for the l Owner's Quality Assurance Program is the Assistant Director for Quality Assurance, who is the functional manager for directing and managing the Quality Assurance Program. He will have the following qualifications:
Education - He shall be a graduate of a four-year dCCredited engineering or science College or university.
Exoerlence -
General - He shall have a minimum of 10 years experience in quality assurance or engineering, construction, or operation activities associated with nuclear f acilities or equivalent heavy industry. A minimum of six years experience shall be in quality assurance.
O 17A-6 Amend. 77 May 1983
Soecialty - He shalI possess a broad knowledge and understanding of industry and government codes, standards, and regulations defining quality assurance requirements and practices.
He shalI have a broad knowledge and understanding of quality assurance methods and their application.
He shall have experience in planning, defining and performing qual 11y assurance practices and the application of procedures.
Manaaerial - He shall be experienced in organizing, directing and administering an overall progran of activity or a major portion of an overall progran having broad scope and application.
He shall have experience in the supervision of personnel and the planning and management of other resourcos normally needed to conduct an extensive quality assurance program.
1.4.2' Chief. Qualltv Verification. Oualltv Enaineering. and OualItv Imorovement The individuals assigned to manage Quality Yerification, Quality Engineering and Quality improvement activities wIlI have the foilowing qualifications:
Education - He shal I be a graduate of a four-year accredited science or engineering college or university.
b d Exoerience -
General - He shalI have a minimum of fIve years experience in quality assurance or engineering, construction, or operation activities associated with nuclear f acilItles or equivalent heavy Industry.
Soecialtv - He shall have a broad understanding and knowledge of appl! cable industry and government codes, standards and 4 regulations def ining quality assurance requirements and practices. !
He shall have experience in planning, defining, and performing i quality assurance practices and the appiIcation of procedures to i the area of work in which he is responsible.
1 Managerial - He shalI be experienced in organizing, directing, and administering an overall progran of activity or a major portion of an overalI program having broad scope and application. He shalI have experience in the supervision of personnel and be capable of directing and coordinating the activities of the contractors to achieve objectives.
1.5 COMMUNICATIONS The f ree, continuous, unimpeded flow of communications, both horizontal ly and vertically within organizations as well as between the Owner and other progran O participants, is essential. The free exchange of Information between V
17A-7 Amend. 70 Aug. 1982
responsible individuals is also essential to the expedient execution of quetIty assurance activities.
To promote the flow of communications and to assure pcsitive attention to O
quality problans within the Project's Quality Assurance Program, lines of communication between the Owner and the organization of the other major program porticipants are established as foilows:
- 1. Communications by Senior Management - These communications will deal with such matters as major changes in the scope of the Quality Assurance Program. Communications will be addressed to the responsible senior management official with copies to the major program participants cognizant of the subject.
- 2. Communications by Oualltv Assurance Program Management - These communications will provide a direct formal or informal exchange of l
Inf ormation between the Owner's Assistant Director f or Qual ity Assurance Chief and other quality assurance managers of organizations which have a direct interface with the Owner. Copies of such communications will be distributed to other major program participants cognizant of the subject.
- 3. Communications by Individuals - These communications are encouraged to identify and evaluate quality problems and to initiate, recommend, or provide solutions. Communications may also be formal or informal, the choice of which shall depend on the significance of the subject and the judgement of the individuals involved.
Communication of quality assurance related activities within the Owner's organization is promoted through:
- 1. Periodic staff meetings of the Project Director.
- 2. Monthly quality assurance program progress and status reporting.
Communication of quality assurance related activities between the Owner's organization and the organizations of the major Program participants is promoted through:
- 1. Quarterly management revlew meetings attended by the Quality Assurance Managers of the major program participants.
- 2. Monthly quality assurance program progress and status reporting by major program participants.
2.0 OUALITY ASSURANCE PROGRAM The Owner has established and is conducting a quality assurance program that meets the criteria of 10 CFR 50, Appendix B. A description of the activities of this program is contained herein and is presented to demonstrate how the Owner is meeting the applicable criteria.
O 17A-8 Amend. 77 May 1983
both the action to be performed and the responsible person or group for performing the action. The Management Procedures Procedure directs implementation of alI CRBRP Project Of fIce procedures when approved by the Project Director. To provide positive identification and control of required procedures for quality assurance activities, manuals containing these procedures have been assembled and issued and are closely controlled by the l Quality Assurance Division and Administrative Services. These manuals each contain copies of the quality assurance program implementing documents Iisted in Figures 17A-12 and 17A-13, including a program description from which the program description in this Appendix to Chapter 17 of the PSAR was derived. A brief synopsis of those procedures is contained in Attachment 1.
The Quality Assuranco Manual is controlled using a document control log which shows the distribution of each copy by copy number inciuding the distribution of revisions. The Quality Engineering Branch of the Quality Assurance Division is responsible for this activity as well as the revision and incorporation of changes to the manual defined and approved by the Assistant Director for Quality Assurance. The contents of the CRBRP Quality Assurance Manual are reviewed and concurrence documented annually as a minimum and the manual is updated as required to maintain it current, in the execution of the program, should a disagreement arise from a difference of opinion between quality assurance personnel and other Project Of fice personnel (engineering, procurement, construction, etc.), the principals (D themseIves try to work it out. Shoutd they falI to resolve the differences,
(./ the heads of the respective divisions are briefed on the problem by the principals and they attempt to resolve the differences on their level. Should they f all also, the problem is presented to the Project Director by the heads of the Divisions involved, and he arbitrates the matter and renders a decision.
A summary description is provided in Chapter 14 of the PSAR of advanced planning for the control of management and technical interfaces between the l Constructor, A/E, NSSS Suppller, and Owner during the phaseout of design and construction and during preoperational testing and plant turnover.
3.0 DESIGN CONTROL 3.1 OWNER IMPLEMENTATION The Owner performs no design in the Project, but has assigned the design responsibility to other major participants in the Project. The Owner has established and specif ied the design guidelines for the CRBRP. In addition.
the Owner has established and specified the essential major plant parameters to be incorporated into the design by the cognizant design participants. The Owner monitors the development of the plant design through participation in design planning, review, and development meetings with the appropriate design organizations.
The Owner has established a Project-wide design control system that is based primarily on a family of System Dasign Descriptions (SDD) as the major vehicle h>)
N for design documentation, review ar.d approval . Through the SDD, the cognizant 17A-15 Amend.,77 May 1983
design contractors are responsible f or the development and def initizing of the Plant systems and components design for which they are assigned responsi bi l i ty . All SDDs are approved by the Owner. The Owner, therefore, maintains the technical supervision and administration of overall aspects of the CRBRP design. As such, the Owner has the responsibility, authority, and accountability for all aspects of the CRBRP design and design control within the specified design, cost and schedule constraints for the Project.
To implement the design control function, the Owner has established design review and approval requirements based upon a four-level classif ication system as l i sted bel ow, and has provided f or both external and internal design interface controls. Externally, the Owner has def ined interf aces and provides direction to the responsible design organizations as follows.
Type 1 Data - Requires Owner Approval l Type 2 Data - Raquires NSSS Supplier and/or AE Approval Type 3 Data - Requires RM or AE Approval Type 4 Data - May be Supplier Approved o NSSS Supplier - The Owner provides direction directly to the NSSS Supplier. For systems and equipment for which the NSSS Supplier has design responsiblity, he recommends approval of Type 1 Data directly to the Owner and the Owner takes approval action, The NSSS Supplier provides Type 2, 3, and 4 data to the Owner for information and the Owner reviews and takes action only as appropriate.
Reactor Manuf acturer - The Owner provides direction to the Reactor o
Manufacturers (RMs), who are major subcontractors to the NSSS Supplier, through the NSSS Supplier for Type 1 systems and equipment lh data. The RMs submit these date to the NSSS Supplier and the Owner in paral lel, and request approval from the NSSS Supplier. The NSSS Supplier in turn reviews and comments on these data, and suomits a recommendaticn to the Owner for approval. The Owner approves or disapproves af ter recei pt of the recommendation. The RM initiates implementation of the activity upon receipt of the Owner approval. In parallel, the NSSS Supplier issues a conf irmatory authorization f or the RM to proceed on the basis of the Owr.er Approval Action. The RM provides Type 2, 3, and 4 Data to the Owner f or inf ormation and the Owner reviews and takes action only as appropriate. (Normal ly, no Owner action is required).
o Architect-Engineer - The Owner provides direction directly to the Arch i tect-Engi neer ( AE) . For systems and equipment f or which the AE has design responsibility, the AE recommends approval of Type 1 data directly to the Owner and the Owner takes approval action. The AE provides Type 2, 3, and 4 data to the Owner for Information and the Owner reviews and takes action only as appropriate.
The Owner has established an overall Design Interf ace Control System for tne Project. The system provides f or control of system and equipment f unctional, parametric, and physical interf ace requirements for all portions of the CRBRP.
Each design contractor is required to assure the accuracy and canpleteness of interface data pertaining systems and equipment under his cognizance.
17A-16 Amend. 70 Aug. 1982
_ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ A
s o Determining, with concurrence of General Counsel, what portions of V procedures, if any, shall be communicated to the contractors.
Furnishing to the Procedures Coordinator the names of contractor personnel to whom such material together with any appropriate supplementary explanation or instructions should be distributed.
The Procedures Coordinator: assures that style, format, content, terms, titles and numbering sequence of all procedures conf orm to the requirements of the Management Procedures System.
The Chief, Administrative Services, is the prime control officer for procedures and as such:
o Effects the printing and distribution of the final approved procedures and subsequent revisions.
o Maintains a master file of alI current approved CRBRP Project office i procedures and a reference file of previously issued procedures and i their revision. J l
o Prepares and maintains an index of procedures. !
Organization Unit Managers are responsible for writing and impicmonting the procedures necessary for their division. General Administration procedures cover policles and procedures which apply to alI employees. The Project Director approves for issuance all CRBRP Project Office procedures. The g ,j
~
individual division procedures are approved by the responsible Division Manager and recommended to the CRBRP Project Director for final approval.
Each new procedure or revision of existing procedure is prepared using the Management Procedures System numbering code and format.
Each division established steps for the review of draf t procedures within the division. If a procedure applies to more than one division, the other divisions af fected receive the draf t procedure for review. A draft is sent to the Procedures Coordinator who reviews it for format, style, and numbering sequence.
The final procedure or revision of an existing procedure is approved by the appropriate Division Manager responsible for that particular subdivision of l procedures, is concurred in by the Assistant Director for Quality Assurance, and is approved by the CRBRP Project Director, and released for implementation.
Distribution of each procedure or revision of existing procedure is listed and f lied with the procedure in the procedure master file. The register shows which revision is current.
The Owner's practice for documenting, in written form, the requirements for and results of activities affecting quality is, itself, executed in accordance l with document control procedures identified under Section 6.0, Document Q Control.
. 17A-25 Amend. 77 May 1983
l The Owner's Quality Assurance Organization both participates in and monitors the execution of this prartice. Periodically the Quality Assurance organization audits or arranges for independent audit of this practice to assure implementation and adoquacy.
5.2 REOUIREMENTS OF OTHER PARTICIPANTS CRBRP Project participants, who are assigned responsibility for perf orming work activ ities af f ecting gest ity, are required by contract to establish and implement a practice of preicribing those activities in documented form that futf11Is the quality assurance requirements. This practice w11I include the preparation of the following types of documents:
o Policies, Proce<ures and Instructions o Quality Records o Quality Status Reports o Design Specifications to include Quantitative Acceptance Criteria such as Dimensions, Tolerances, and Operating Limits o Design, Manufactoring, Construction and Installation Drawings to include as-built drawings that accurately reflect the actual plant configuration o System Design Descriptions o Manufacturing, Construction, Installation, inspection and Testing Specifications, Procedures, and Instructions to include Qualitative Acceptance Celteria such as workmanship samples o Test Procedures o Topical Reports and input to SARs The Owner monitors major participant documentation practices and periodica!ly audits the participants practice to assure proper implementation and adequacy and to verify that important activities have been satisfactorily accomplished.
6.0 DOCUMENT CONTROL 6.1 OWNER IMPLEMENTATION The Owner has established and implemented a document control system that f ul fil ls the quality assurance program requirements and applies to those types tat documents prepared by the Owner and identif led in Sections 3, 4 and 5 of this description.
The controlled documents originated by the Owner are processed in a controlled manner to assure the following:
o Uniformity of format of initial and subsequent issuances.
17A-26 Amend. 77 May 1983 -
. O O O
- EM PROGRAM ACTIVITIES
! PROGRAM MANAGEMENT l
= QUALITY ASSURANCE
- ORGANIZATION
- DOCUMENTATION
- AUDITS AND REVIEWS
- CORRECTIVE ACTION
! PROGRAM I
- 1. PLANNING 1. RESPONSIBILITY 1. POLICIES AND 1. QUALITY AUDITS =. ENGINEERING HOLpS I 2. QUALITY ASSURANCE AND AUTHORITY PROCEDURES 2. MANAGEMENT l PROGRAM INDEX 2. TilA!NING AND 2. QUALITY RECORDS REVIEWS
- UNUSUAL INDOCTitlNATION 3. QUALITY STATUS OCCURANCE
'E 3. PEllSONNEL REPORTS REPORTS
@ ObALitlCATION q _____
, uw 1 r l'
- _r 2
DESIGN AND DEVELDI MENT i
PROCUREMENT MANUFACTUlllNG. FABRICATION.
, AND ASSEMBLY Y DESIGN PLANNING PROCUREMENT PLANNING INSPECTION AND TESTS b DESIGN DEFINITION AND CONTilOL PROCUREMENT REQUIREMENTS 1. PROC 6duRES
- 1. ENGINEERING STUDIES PROCUREMENT DOCUMENT REVIEW
- 2. SPECIFICATIONS, DRAWINGS AND INSTRUCTIONS EVA1.UATION AND SEl ECTION OF DOCUMENT' REVIEW AND CONTROL
- 1. DOCUMENT REVIEWS
"^ ' "
- ^ ^
- 2. DOCUMENT CONTilOL
^ ^
- 3. ENGINEERING DilAWING LISTS
- 4. INTEllCHANGE OF SOURCE DESIGN IlEVIEWS CAPABILITY INFORMATION P6' FIGURE 17A-11 MAJOR ELEMENTS OF THE EM PROGRAM
- 8
~F Q
DM REQUIREMENT OF IMPLEMENTING DOCUMCNT RDT T2-2 SECTIO'I D0C. NO. TITLE INSTRC f!ONS TITLE NUMBER REF. CDC.. ETC.
- 2. Management & Planning 2.2 Quality Assurance Program Quality Assurance Program Description Quality Assurance Charter 2.2.1 Planning Quality Assurance Program Description CRP-EN-09 Preparation and Maintenance of the Project Level 1 Schedule CRP-PC-02 Preparation and Maintenance of the CRBRP Project level 0 Schedule CPP-PC-05 Preparation and Maintenance of the Werk Breakdown Structure (WBS)
CRP-QA-02 Activity Planning w
2.3 Organization G 2.3.1 Responsibility and Authority Quality Assurance Program Description
- Quality Assurance Charter Pesponsibility All Procedures Sections CRP-DR-02 Organization Plan and Functional S tatements 2.3.2 Training and Indoctrination CRP-QA-24 Personnel Indoctrination 2.3.3 Personnel Qualification CRP-QA-25 Adminf stration of Personnel Certification and Records CRP-QA-26 Personnel Certification PMC 4.05 Confirmation of Position Requirement *, and Employee j 2.4 Documentation Qualifications 2.4.1 Policies and Procedures Qpality Assurance Program Description All Procedures Policy Sections CRBRP Manaaement Policies and Requirements CRP-AA-01 Management Precedures CRP-AA-03 Preparation of Correseeedeace l
i CRP-AA-11 Control of Project Office Procedures Manual at on Ma enance and Control of Project effice Quality P -20 r Assurance Manual l QUALITY ASSURANCE PROGRAM INDEX VERSUS REQUIREMENTS OF RDT F 2-2
?b' l wg Figure 17A-12 OWNER QUALITY ASSURANCE PROGRAM INDEX l G ."
l L~
9 9 9
m O O RE0J:RD.I*.T CF RPMS RDT F 2-2
- .:PLD:E .!!!:S DCC?T.T SECTION INSTRUCTICNS '
CMBER TITLE 00C. C. TryLg REF. 00C.. ETC.
a CRP-PS-06 l Distr % ..u... Lvelu tion, and Preparation of Responses to NRC's 10ffke of Inspection and Enfgrcement Bulletins. Circulars, and j Notices
~
CRP PS-03 rcesration and Approval of Responses to Request for Licensing and Safety information CRP-PS-04 Preparation and Compliance with Non-NRC Permits, Licenses, and Arprovals CRP-PS-05 Mainter.ance of Cor.sistency Between tr.e PSAR and the Project Baselir.e Docurer.tation CRP-CA-10 Quetity Assurance Review and Approval of Engineering Documents CRP-0A-11 Quality Assurance Review of Procurement Documents CRP-CA-12 Review of Con:ractor Quality Assurance Plans and Procedures g CRP-CA-20
%a Preparation. Maintenance and Control of Project Office Quality 3= Assurance Manual 4
3.5 Desigr. Review CRP-EN-03 Design Reviews 3.6 Development CRP-EN-06 Cevelopent Program Technical Management CRP-PC-05 Preparation and Paintenarce of the Work Breakdown Structure (WBS)
CPP-CA-10 Quality Assurance Feview and Approval of Engineering Docanents CRP-QA-12 Review of Contractor Quality Assurance Plans and Procedures LRP-EN-12 Test Article Readiness Reviews
, ,s.
CUA:.!TY ASSUR* :CE PROGPA*1 ICEX VERSUS RI*U RI"INTS CF ROT F2-2
?b' w <o a
w CL w -a FIGURE 17A 22 (CONT'D.) OWNER QUALnY ASSURANCE PriOGRAM INDEX N
~
2 RE7JIREMENT CF ;y;gg.g-;;.i3 D000*E :T REWRKS CT F 2-2 I'iSTR;0ii?is
,,e REF. C00., ETC.
"00. ?40. Tv f: q TITLE 4 Procuremen t [8, ,]
h Procure-ent Plan 91rg C"P-D:-09 P-eparetion and Faiate ance of the Fref ect Level 1 Schedule 4.2 07.PP Prefect Font *1y P-e;-ess reoert
) C R'-+ C-03 CFI-IC-05 Precera icn and Fairitenance of the
- crk Sreakcown Stracture (WE51 CRD-QA-02 Activity Planning I
Precureneet Fecuirteents CRP-Cfi-CI Processir9 of Construct 10e Ocita Subritted by the Ccastructcr 4.0 Processing Principal Onign Occurents CDP-Pi-02 CP-t :-01 Processina Ingineering Chances e r o-it:.05 Cen'ia,.." ra te t ml Peard a ctior.s
'.EP -0C-:;1 Control of Modifications to Principal Project Agreements
[
3= CRP-OP-02 Opeeatine* Division Seviw and Concurrence with Engineering 8
Design Data contract arc Subcontract Review Actions
$ CRP-PR-03 CRP-QA-10 Cuality Assurance Review and A3provai of Incince-ina DocWer.ts 4.4 Procurement Docurent Review CRP-CN-01 Procetirt c' Ccestructice Data Sutmitted by the Corstructcr CRP-EN-02 Processing Principal Oesign Docurents CDP-IN-94 rrecessiac Enciacari,0 Ctarces CRP-E?i-CS Configuraticn Con'.rol Peard Actions CF.P-PR-C3 Contract and Sutcontract Feview Actioas CFP-QA-11 Quality Assurance Review of Procurerent Cecuntnts 4.5 Evaluation and Selection of Procurement Sources CRP FR-03 Cor.trect and Sutcentract Pcvicw Actiens OUALITY ASS'JRA';CE PROGRici !?.0tt VE25;s RE+;I;.E E . S 0; ,9T F 2-2 Ok
.mCL s
g* FIGURE 17A-12 (CONT'D.) OWNER QUALITY ASSURANCE PROGRAM INDEX
@ IV 4 9 e .
a a-MQUIRDENT OF ROT F 2-2 IMPtDIDITING 00CtMFMT REMARKS SECTION FesTRUCTIO%
NUMER TITLE 00C. I:0. TITLE REF. 00C.. ETC.
4.6 Control of Configuration 4.6.1 Contract Change Control CRP-CN-02 Processing of Field Chaege Requests CRP-EN-04 Processing Engineering Changes CRP-Erl-05 Configuration Control Board Actions CRP-0C-01 Control of Modific"lons to Principal Project Agreements y CRP-PR-03 Contract and Subcontract Review Actions 4.6.2 As-B dit Vertfication Ce-C*-04 Construction Testing and Tornover G
CRP-QA-02 Activity Planning
[ CRP-QA-13 Performance of Project Survelliance CRP-QA-16 Inspection. Examination and Ten 4.7 Prasuring and Test Equipsvnt Calibration and Control CRP-QA-17 Measuring and Test Equipment Calibration and Control 4.8 Source Survefilance and Inspection CRP-QA-02 Activity Planning CRP-QA-13 Performance of Project $vrvalliance CRP-QA-14 Processing of Responses to NJclear Regulatory Commission Inspection Reports and Their Follow-t'p During Design and Construction CRP-QA-15 Arranging 'cr Nuclear Regulatory Conef ssfon Inspections CRP-Qk16 Inspection. Examination and Test -
QuACITY ASSURANCE PR0sRAM INDEx VERsus RtQutRtm eis OF ROT F 2-2
?E F$
- .a e
ME Figure 17A-12 (Cont'd.). Owner Quality Assurance Program Index s
REQUltEMENT OF ROT F 2-2 INPLEMENTING 00CtMENT REMARKS SECTION INSTRt'CTIONS TITLE .00C. NO. TITLE .AEF. 00C.. ETC.
_ NUMBER 4.9 Receiving !aspection 4.9.3 Planning and Inspection CRP-QA-02 Activity Planning CRP-QA-16 Inspection. ExaminatMn and Test 4.9.2 Documentation CRP-AA-04 Incoming Mail CRP-AA-14 Controlled DH.Wnents CRP-CN-01 Frocessing of Construction Data Sutwettted by the Constructor CRP-CN-02 Processing of Field Cnange Requests ea CRP-CN-04 Construction Testing and Turnover N CRP-EN-02 Processing Principal Design Doc'.snents Operations Division Revice and Concurrrr.ce with f
cn CRP-OP-U2 Engineering Cesign cats W CRP-OP-03 Oscrations Dtrision Review and Concurrence with 7 Licensing Data CRP-PR-03 Contract ard Subcontract Peview Actions CRP-QA-10 Quattty Amrance Revice and Approval of Engineering Documents C9P-QA-Il Quality Aswance Review of Procurement Docuents CRP-QA-12 Revtem of Contractor Quality Assurance Plan and Procedures CRP- D-Il Review. Approval .,w !ssuance of Construction Documente aant Requiring project Mff ce Approval 4.9.3 Disposition of Receld Delega W Items 4.10 Control of Moncor.forring Control of Nonconformances j
items (R*-4bO3 lCRPQA-28 lStopWorkOrders l
Delegated i 4.11 Control of Received !tems See Section 8 4.12 Quality Audits Manufacturing. Fabrication Deleseted 5.
and Assembly _
37, w=
M fy QUALITT ASSURANCE PROGRAM INDEI VER5US REQUIREMENTS OF RDT F 2-2 g .'
cn "U FIGURE 17A-12 (Cont'd.). 0WNER QUALITY ASSURANCE PROGRAM INDEX 6 9 9
1 a
0 O i
- i REQUIREMENT OF
- IMPLEMENTING DOCl,WENT REMRRit$
10 CFR 50 APPENDIX B ,
INSTRUCTIONS
] CRITERION TITLE D0C. NO. TITLE REF. D0C.. ETC.
I Inspection CRP-QA-02 Activity Planning l
CRP-QA-14 Processing of Responses to Nuclear Regulatory Commission Inspection Reports and their Follow-up During Design and Construction I
- CRP-QA-15 Arranging for Nuclear Regulatory Comission Inspections j1 CRP-QA-16 Inspection. Examination and Test XI Test Control CRP-EN-07 Technical Control of CRBRP Test Programs C CRP-CP-02 Operations Division Review and Concurrence with Engineering p Design Data cm g CRP-QA-02 Activity Planning ;
J CRP-QA-13 Perforinance of Project Surveillance i
CRP-QA-16 Inspection. Examination and Test CRP-CN-04 Construction Testing and Turnover XII Control of Measuring and Test CRP-QA-17 Measuring and Test Equipment Calibration and Cefitrol !
Equipment i
XIII Handling. Storage and Shipp;ng (Delegated)
\
XIV Inspection. Test and Operating
- Status (Delegated) h p QUALITY ASSURANCE PROGRAM I:iDEX VERSUS REQUIREMENTS OF 10 CFR 53. AFFE*CIX B .
i b.3 -
- , e J FIGURE 17A-13 (Cont'd.) OWNER QUALITY ASSURANCE PROGRAM INDEX 1
N U1 WN '
t I
. . , , ,, y -, -
- ~ -,-c - - - --- < ~' '
1C N X 'B I N 00 M ENT RN l INSTRUCTICNS CRITERION TITLE D0C. NO. TITLE REF. D00., ETC.
XV Nonconforminq Materials, Parts CRP-AA-04 Incoming Mail and Components CRP-CN-01 Processing of Constructit.n Data C3mitted by the Constructor CRP-CN-02 Processing of Field change sequests CRP-EN-02 Precess'ng Principal Design Docu e9ts CRP-OP-02 Operations Divisien Review and Concurrence with Engineering Design Data CRP-CN-04 Construction Testing and Turnover 5 CRP-QA-03 Control of Nonconformances c'n co g CRP-QA-05 P ocessing of Unusual Occurrence Reports CRP-QA-14 Processing of Responses to Nuclear Regulatory Comission Inspections Reports and Their Follow-up During Cesign and Construction CRP-04-15 Arranging for Nuclear Regalatory Comission Inspections CRP-0A-27 Unusual Occurrence Report Preparation and Disposition CRP-QA-28 Stop Work Orders XVI Corrective Action CRP-CM-02 Processing or riend change pequests CRP-EN-02 Processing Principal Design Documents CRP-OP-02 Operations Division Review and Concurreece with Engineering Design Data QUALITY ASSURANCE PROGRAM INDEX VERSUS REQUIREMENTS OF 10 CFR 53, W ENDIX B
& b'
- 8
-a 8- FIGURE 17A-13 (Cont'd.) OWNER QUALITY ASSURANCE PROGRAM INDEX
"'d 9 9 e
O J O a
REQUIREM NT OF 10 CFR 50 APPEN0!I B lhPLEMENTING 00CtMENT RUNES ,
IlfiTRUCTIONS CRITERius TITit DOC. NO. TITLE REF. 00C.. ETC.
CRP-QA-03 Control of Noncenformances CAP-QA-04 Ccerective Action Requests .
CRP-QA-05 Processing of Unusual Occurrence Reports W CLP-QA-06 Monconforsence. Unusual Occurrence and Corrective
- y Action Analysis W
CRP-QA-09 Qualfty Trend Analysis
- CRP-GA-27 Upotual Occurrence Report Preparation and 01sposttlan CRP-QA-28 Stop Work Orders XVII Quality Assurance Records CRP-AA-02 Filing Procedurws for Official Project Files CRP-AA-04 Incaning Mail CRP-QA-07 Quality Records CRP-QA-23 Preparation. Transfer. sed Receipt of Project Of fice Duality Records
, XVIII Audits CRP-0A-19 Adelnistretton of Quality Assurance Auditing l CRP-QA-21 Conduct of Product Audits CRP-QA-22 Conduct of Progressutic Audits I
QUALITY ASSURANCE PROGRAM INDEI VERSUS REQUIRLMENTS OF 10 CFR 50. APPENOII I DE i
- a tr> F k
Figure 17A-13 (Cont'd.). Owner Quality Assurance Program Index I
Review of Contractor Qualltv Assurance Plans and Procedures (CRP-0A-12)
This procedure defines the responsibilItles for review of @BRP participant quality assurance plans and procedures to determine their adequacy. The acceptance / approval process is also described.
Performance of Project Surveillance (CRP-0A-13)
This procedure detines the responsibilitles for the planning, conduct and reporting of surveillance activities performed by the CRBRP Quality Assurance Olvision.
Processing of Resoonses to Nuclear Reculatory th =lssion insoection Reoorts and Their Follow-uo During Design and Construction (CRP-0A-14)
This procedure defines the responsibilitles and actions for the review of NRC Inspection reports by the CRBRP Project Of fice. The procedure also details the actions for preparation and coordination of the formal CRBRP Project Office response to NRC.
Arranging for Nuclear Regulatorv Commission insoections (CRP-0A-15)
This procedure defines the responsibilities for internal handling of notification of NRC inspections that are to be conducted. The procedure also describes the communication links and coordination channels for the necessary arrangements.
O Q insoection. Examination and Test (CRP-OA-16)
This procedure defines the responsibilitles for the preparation for and performance of quality assurance inspections, examinations and tests during design development, procurement, construction, installation, start-up, operation, maintenance and modif Ication of the CRBRP.
Measuring and Test Eauloment Calibration and Control (CRP-0A-17)
This procedure defines the actions and responsibilities for the verification that measuring and test equipment used for inspections, examinations or tests are properly calibrated and controlled.
Administration of Oualltv Assurance Auditing (CRP-0A-19)
This procedure defines the responsibilitles for the planning, conduct, follow-up, and close out of quality assurance audits. This procedure also details the actions of the quality assurance audit administrator in documenting the audit activity.
O- 17A-75 Amend. 70 Aug. 1982
Preoaration. Maintenance and Control of Project Office Oualltv Assurance Manual ( CRP-0A-20 )
This procedure def ines the actions and responsibilities f or the preparation, O
distribution, maintenance and control of the CRBRP Quality Assurance Manual .
Conduct of Product Audits (CRP-0A-211 This procedure defines the actions and responsibilities for the preparation, conduct and reporting of quality assurance product audits by the CRBRP Project Office. The procedure also details the actions of the audit team in the course of the evaluation of selected products for conf ormance to quality requirements.
Conduct of Proarammatic Audits (CRP-0A-22)
This procedure def ines the responsibilities for the preparation, conduct, and reporting of quality assurance programmatic audits by the CRBRP Project Office. The procedure details the actions of the audit team in the course of the evaluation of programmatic practices for conformance to the quality assurance program requirements.
Preoaration. Transfer. and Recelot of ProJoct Office Oualltv Records
( CRP-0 A-23 )
This procedure defines the responsibilites and actions to be executed by each Project Of fice Division in the preperation and transfer of quality records to the Quality Assurance Division. The procedure also defines the responsibilities and action of the Quality Assurance Division when receiving quality records from other Project Of fice Divisions.
Personnel _ indoctrination (CRP-0A-24)
This procedure defines the responsibilities and actions to provide for the indoctrination of CRBRP Project Of fice personnel who carry out duties af fecting the quality of the CRBRP Plant structures, systems and components.
Administration of Personnel Certification and Records (CRP-0A-25)
This procedure def ines the responsibilities for the administration of certification for Quality Assurance Division personnel directly involved in qual ity verif ication, testi ng, eval uation or audli activ ities. The procedure also details the actions associated with collection and maintenance of records pertaining 1o personnel certi f ication.
Perr.onnel Certi f ication (CRP-0A-26)
This procedure defines the responsibilities and actions necessary to identify areas of quality importance for which qualifications or certification of personnel are required. The procedure also details the actions for verifying the adequacy of personnel training programs, certification practices and documentation.
O 17A-76 Amend. 77 Vay 1983
Unusual Occurrence Reoort Preoaration and Disoosition (CRP-0A-27)
This procedure defines the actions and responsibilities for documenting an unusual occurrence observed during the course of work on the CRBRP Project.
The procedure also details the action related to evaluation of the reportability of the event to NRC as well as the channels for reporting to NRC.
Ston Work Orders (CRP-0A-28)
This procedure defines the responsibilItles and actions required for issuing and processing Stop Work Orders (SW0s), which stop f urther fabrication, installation, or use of nonconforming items or processes which would result in a condition adverse to quality in the Plant. This procedure applies to saf ety-related work, work or activities important to saf ety, and nonsaf ety-related work which are determined to be in noncompliance with design or programmatic requirements end will be noted as such on the SWO form.
Confirmation of Position Reaylramants and Fmnlovee Qualifications (PMC 4.05)
This procedure defines the requirements and actions necessary to confirm the position requirements and employee qualifications for PMC professional employees or to assignees f rom member utilities assigned to the CRBRP Project.
p
%)
O V
17A-77 Amend. 77 May 1983
ATTAOiENT li PROCEDURE RELEASE SCHEDULE CRP Procedure Number Title Scheduled Release Date OlP-QA-28 Stop Work Orders June 1,1983 O
l l
O 17A-78 Amend. 77 May 1983
Q.lNOi RIVER BREEDER REACTOR PLANT A DESCRIFT10N OF THE BALANE OF PLANT SUPPLY QUALITY ASSURAN PROGRAM TABLE OF CONTENTS PAGE NO.
0.0 INTRODUCTION
17C-1 0.1 SCOPE 170-1 0.2 BASIS 170-1 0.3 APPLICATION 170-1 l l
1.0 ORGANIZATION 17C-1 1
1.1 FUNCTION 17C-2 )
i 1.2 RESPONSIBILITY AND AUTHORIIY 17C-2 1.3 ORGANIZATIONAL ARRANGEMENIS 17C-3 1.3.1 Oualltv Verification Branch 17C-4 1.3.1.1 Survalliance 17C-4 1.3.1.2 Insnection 17C-4 1.3.1.3 Audit 17C-4 1.3.2 Oualltv Engineering Branch 17C-4 1.3.2.1 Planning 17C-4 1.3.2.2 Reocrts 17C-5 1.3.2.3 Records 17C-5 1.3.3 Oualltv imorovement Branch 17C-5 1.3.3.1 Nonconformance Control 17C-5 1.3.3.2 Trend Analvsis 17C-5 1.3.3.3 Training and Indoctrination 17C-5 1.3.4 AE Project Oualltv Assurance Section 17C-6 1.3.4.1 Internal Audit and Survalliance 17C-6 1.3.4.2 Qual [tv Assurance Eng!neering 17C-6
(
17C-1 Amend. 70 Aug. 1982
TABLE OF CONTENTS (CONTINUED)
PAGE NQ, 1.3.4.3 Vendor Audit and Surveillance Groun 17C-7 1.4 OUALIFICATION REOUIREMENTS FOR OUALITY ASSURANCE 17C-7 MANAGEMENT POSITIONS 1.4.1 Oual!f! cat!cn Recuirements of the Assistant 17C-7 Director for Oualltv Assurance 1.4.2 Chief. Quality Verification. Qualltv Engineering. 17C-7 and Ouality Imorovement 1.4.3 Quad Itv Assurance Manager-AE Suncorted Services 17C-8 1.4.3.1 Oualltv Assurance Engineering Groun - Suoervisor 17C-9 1.4.3.2 Internal Audit and Surveillance Groun - Suoervisor 170-9 1.4.3.3 Vendor. Audit and Surveillance Groun - Suoervisor 17C-10 1.5 COMMUNICAll0NS 17C-10 2.0 OUALITY ASSURANCE PROGRAM 17C-11 2.1 PROGRAM REOUIREMENTS 17C-11 2.2 PROGRAM ELEMENTS 17C-11 2.3 PROGRAM IMPLEMENTATION 17C-12 3.0 DESIGN CONTROL 17C-13 4.0 PROCUREMENT DOCUMENT CONTROL 17C-14 5.0 _ INSTRUCTION
S. PROCEDURE
S AND DRAWINGS 17C-17 6.0 DOCUMENT CONTROL 17C-19 7.0 CONTROL OF PURCHASED MATERI AL. EOUIPMENT AND SERVICES 17C-21 8.0 IDENTIFICATION AND CONTROL OF MATERIALS. PARTS AND 17C-23 COMPONENTS 9.0 CONTROL OF SPECIAL PROCESSES 17C-24 10.0 INSPECTION 17C-24 11.0 TEST CONTROL 170-26 12.0 CONTROL GF MEASURI4G AND TEST EQUIPMENT 17C-27 g
, 17C-11 Amend. 77 May 1983
TABLE OF CONTENTS (CONTINUED)
PAGE NO.
13.0 HANDLING. STORAGE AND SHIPPING 17C-28 14.0 INSPECTION. TEST AND OPERATING STATUS 17C-28 15.0 NONCONFORMING MATERIALS. PARTS OR COMPONENTS 170-29 16.0 CORRECTIVE ACTION 170-31 17.0 OUALITY ASSURANCE RECORDS 170-3',
18.0 AUDITS 17C-33 FlGURES 1 170-1 BOP Supplier Quality Assurance Organization 170-36 17C-2 Major Elements of the BOP Supply Quality Assurance 170-37
, Program 17C-3 BOP Supply Quality Assurance Program Index 17C-41
, 17C-4 B0P Supply Quality Assurance Program index 170-45 ATTACHMENTS
- 1. Quality Assurance Procedure Descriptions 170-52 l ll. Procedure Release Schedule 17C-61 O
17C-Iii Amend. 77 May 1983
CLINCH RIVER BREEDER REACTOR PLANT A DESCRIPTION OF THE BALANCE OF PLANT SUPPLY QUALITY ASSURANCE PROGRAM
0.0 INTRODUCTION
0.1 SCOPE Contained herein is a description of the plans and actions by the Balance of Plant (BOP) Supplier to assure the quality of certain Balance of Plant (BOP) components of the Clinch River Breeder Reactor Plant (CRBRP). These plans and actions constitute the BOP Supply Quality Assurance Program.
0.2 BASIS The BOP, as def ined by the Project, includes all other structures, systems and components of the plant not included in the Nuclear Steam Supply System (NSSS). The CRBRP Owner has assigned to other selected major Project participants the action of supplying a large portion of the BOP equipment.
The CRBRP Owner has chosen to supply certain BOP equipment items and has elected to retain the execution responsibility for this major portion of the overall Project Quality Assurance Program. Assisting the Owner in this endeavor by perf orming supplier surveillance f unctions is Burns and Roe, Inc.
This combined orgaalzation is then designated the BOP Supplier with overall responsibility for execution of the BOP supply quality assurance program.
l The BOP Supply program requires that all technical documents for procurement v of equipment, such as drawings, specif Ications and statements of requirements are provided by other participants and are not covered by program activity.
0.3 APPLICATION The BOP Supply Quality Assurance Program described herein is applicable to the procurement, and through contract to the manuf acturing, of certain CRBRP BOP equipment, it includes application to all BOP suppil'er f urnished structures, systems, components and consumables (including those in Section 3.2, 7.1, and l 9.13) whose satisf actory perf ormance is required f or the plant to operate reliably, saf ely, and with minimum environmental ef fects.
1.0 ORGANIZATION The Owner is RESPONSIBLE for the overall management of the Project to design, build, and operate the Clinch River Breeder Reactor Plant (CRBRP). The execution of this responsibility rests with the CRBRP Project Director, who is the principal operations of ficer of the Owner. Part o' this responsibility is to assure that the BOP equipment is designed, manuf actured and supplled in a way that will provide adequate confidence that it will perform satisf actorliy in service. To provide assurance that the BOP equipment will be supplied so as to perf orm satisf actorily in service, a quality essurance program has been established which shall have the objectives, carry out the f unctions, and be executed as hereafter defined.
17C-1 Amend. 70 Aug. 1982
i 1.1 FUNCTION The f unctions that the Owner will perform in order to achieve the stated objectives of the BOP Supply Quality Assurance Program and f ulfill its ultimate responsibility for program adequacy are as follows:
- 1. The development of an overall plan for the conduct of the BOP Supply portion of the quality assurance program.
- 2. The development of working plans and procedures to conduct the BOP Supply Program activities.
- 3. Organizirig and staf fing appropriately to implement the BOP Supply Program activities.
- 4. The securing and directing of support services in the conduct of 80P Supply Program activities.
- 5. The surveillance over and management coordination of supplier programs.
- 6. The development and execution of program activities to meet necessary req u i rements.
1.2 RESPONSIBlLITY AND AUTHORITY l The Manager of Quality Assurance for BOP Supply is also the Assistant Director for Quali ty Assurance, serving as head of the CRBRP Project Of fice Quality Assurance Division and reporting directly to the Projct Director. The l Assistant Director for Quality Assurance, acting as the Manager of Quality Assurance for BOP Supply, is assigned responsibility for devising, recommending establishment of and assuring ef fective execution of the BOP Supply Quality Assurance Program and assuring that organization, systems, and procedures at all levels will provide assurance that the BOP equipment is designed in accordance with requirements, is manufactured as designed, and is supplied in accordance with plans and procedures. In carrying out these responsibilItles, he is authorized by the Project Director to:
- 1. Identify quality problems.
- 2. Initiate, recommend, or provide so!utions through designated channels.
- 3. Verify implementation of solutions.
- 4. Determine the adequacy of facilities and equipment provided to carry out the approved procedures and instructions.
- 5. Authorize issuance of special Instructions necessary to execute his responsibilitles.
O 170-2 Amend. 77 May 1983
- 6. Notify responsible management of unsatisfactory work or unapproved pi practices and if necessary, stop unsatisfactory work or control i further processing, delivery, or installation of nonconforming materials.
lTheAssistantDirectorforQualityAssurance, is responsible for organizing the BOP Supply Quality Assurance Program and for recommending assignments of execution responsibility as appropriate. He shall secure charter statements from other participants describing responsibilities and functions. He shall assure that in each participant's organization the person responsible for quality assurance is granted sufficient authority to identify quality problems; to initiate, recommend, or provide solutions; and to verify implementation of solutions.
l The Assistant Director for Quality Assurance, is responsible for recommending to the Director the organization and staf fing plan for the Quality Assurance Division and for managing the Quality Assurance Division in the conduct of those quality assurance practices necessary to f ulfiIl the BOP Suppller responsibliitles for establishment and adequacy of the Program. In this position, he is responsible for the technical and administrative control (except for DOE personnel) of Individuals and groups of the Quality Assurance Division, performing quality assurance activities, or verifying adequacy in the performance of quality assurance related activities of others. The Project Director retains administrative control of DOE personnel.
The AE Project Quality Assurance Manager receives technical direction and l f unctional assignment for the BOP Supply ProCram J frcm the Assistant Director T for Quality Assurance. Administratively, he and his quality assurance groups s are a part of the AE Internal organization where he reports directly to the Vice President, Breeder Reactor Division, for quality assurance program ef f ectiveness.
The supplier surveillance function is delegated to the AE Project Quality Assurance Section. The manuf acturing f unctions of the program are or will be delegated to the various equipment suppliers by contract. All organizations who are delegated quality assurance f unctions, report their progress and l quality achievement to the Assistant Director for Quality Assurance.
1.3 QBGANIZATIONAL ARRANGEMENTS The BOP Supplier organization for performing quality-related activities associated wIth management, planning and procurement of B0P equipment and the responsibility and authority of key positions within that organization are described in Section 1.4 of the PSAR for the Owner end the AE respectively.
The BOP Suppller organization is shown in Figures 1.4-1 and 1.4-7 for the Owcer and the AE respectively. The organization that will manage and implement the BOP Supply Quality Assurance Program is shown in Figure 17C-1 and is a combination of the Owner Quality Assurance Division and the Architect-Engineer Quality Assurance Section. The organization f unctions as a joint unit with support being drawn frcm the Owner's and the AE's quality assurance staff. The organIIation is subdivided along functional lines to perform quality verification, quality engineering, quality improvernent, and AE quality assurance support. A description of the organizational elements are contained in subsequent paragraphs.
17C-3 Amend. 77 May 1583
1.3.1 Oualltv Verification Branch The f unction of the Quality Verif ication Branch is to maintain surveillance over quality assurance programs of major contractors and suppliers and to verify quality achievement in their work perf ormance. This branch is also responsible for monitoring the BOP Supply quality assurance program to verify overalI adequacy.
The Quality Verif ication Branch perf orms three types of activities as described below:
1.3.1.1 Surveillance Monitoring of the Project work and the quality assurance practices on that work is perf ormed through this activity, it also serves as the focal point for interf ace coordination between the BOP Supply quality assurance progran and the quality assurance programs of other major contractors and suppliers.
The initiation, coordination and follow-up of inspections, reviews and audits is perf ormed through this activity including mding arrangements for support needed f rom other sections in the quality assurance organization, other staf f organizations or through service contracts.
1.3.1.2 Insoection inspection of items and services or the monitoring of Inspections of others is accomplished through this activity. This activity includes performance of selected civil, structural, electrical, mechanical and welding inspections, and nondestructive examinations.
1.3.1.3 Audit l Planning and conducting internal audits of the BOP Supplier's Quality l Assurance Program and external audits of contractor Quality Assurance Programs i is accomplished through this activity. Scheduled and unscheduled audits are conducted.
1.3.2 OualItv Engineering Branch The f unction of the Quality Engineering Branch is to plan, define, and develop the B0P Supply Quality Assurance Program. This f unction incl udes the prepara-tion and maintenance of BOP Supply Quality Assurance Progrcm requirements and internal pl ans and procedures. This branch has lead responsibility for quality assurance progran progress and status reporting and quality records management.
The Quality Engineering Branch perf orms three types of activities as described below:
1.3.2.1 Planning Planning, program devt lopment and documentation of plans and procedures for i conduct of the BOP Supp ler Quality Assurance Program are perf ormed through I this activity. A knowledge of Industry and goverraent standards and their 17C-4 Amend. 70 Aug. 1982 .
l
1.3.4.3 Vendor Audit and Surveillance Grouo r~~s ,
(, ) This group schedules, plans and implements vendor pre-award and in-process fabrication surveys and audits. The services of quallfled AE engineers are utilized whenever the scheduled audit or survey requires such capability. The group has authority to stop unsatir'actory or unapproved practices through contractual channels by virtue of tao Chief, Quality Assurance for BOP Supply charter. This group provides surveillance services for and performs f unc-tionally with the surveillance function of the Quality Verification Branch.
1.4 OUALlFlCATION REOUiREMENTS FOR OUALITY ASSURANCE MANAGEMENT POSITIONS l 1.4.1 OUALIFICATION REOUlREMENTS OF_THE ASSISTANT DIRECTOR FOR OUALITY ASSURANCE The Individual assigned to retain overall authority and responsibility for the l BOP Supplier Quality Assurance Program is the Assistant Director for Quality Assurance who is the f unctional manager for directing and managing the Quality Assurance Program. He will have the following qualifications:
Education - He shall be a graduate of a four-year accredited engineering or science college or university.
Excerlenqa -
General - He shall have a minimum of 10 years experience in quality
(~N assurance or engineering, construction, or operation activities
- s. associated with nuclear f acilities or equivalent heavy industry.
Soecialtv - He shall possess a broad knowledge and understanding of industry and government codes, standards, and regulations defining quality assurance requirements and practices.
He shall have a broad knowledge and understanding of quality assurance methods and their application.
He shall have experience in planning, defining and performing quality assurance practices and the application of procedures.
Managerf al - He shall be experienced in organizing, directing and adninistering an overall program of activity or a major portion of an overall program having broad scope and application.
He shall have experience in the supervision of personnel and the planning and management of other resources normally needed to conduct an extensive quality assurance program.
1.4,2 CHIEF. OUAL ITY VERIFICATION. OUAL ITY ENGINEERING. AND OUAL ITY
.lMPROVEMENT The Individuals assigned to manage Quality Verification, Quality Engineering
, and Quality improvement activities will have the following qualifications:
17C-7 Amend. 77 May 1983
Education - He shall be a graduate of a four-year accredited science or engineering college or university.
Experience -
General - He shall have a minimum of five years egorienca in quality assurance or engineering, construction, or operation activities associated with auclear f acilitles or equivalent heavy industry.
Soecialty - He shall have a broad understanding and knowledge of applicable industry and government codes, standards, and regulations def ining quality assurance requirements and practices. He shall have experience in planning, def Ining, and perf orming quality assurance practices and the application of procedures to the area of work in which he is responsible.
Managertal - He shall be experienced in organizing, directing, admin-Istering an overall program of activity or a major portion of an overall progran having broad scope and application. He shall have axperience in the supervision of personnel and be capable of direct-Ing and coordinating the activities of the contractors to achieve obj ectives.
l 1.4.3 OUALITY ASSURANCE MANAGER-AE SUPPORTED SERQCES The individual responsible for management of the AE quality assurance support services will have the following qualifications:
Education - A BS degree in Engineering / Science or an equivalent combination of education and experience is preferred.
Exoerlence -
General - He shall have a minimum of 10 years experience in quality assurance or engineering associated with design, construction, or operation of a nuclear plant f acility power generating station or heavy industry.
Specialtv - He shall possess a broad knowledge and understanding on industry and government codes, standards and regulations which define quality assurance progran requirements and practices. He shalI have a broad knowledge and understanding of quality assurance metnods and their appl ication. He shall have exper ience in planning, def ining l and Implanenting quality assurance practices and the application of procedures.
Managerial - He shall have a minimum of eight years experience in the supervision of personnel and the planning and management of other resources needed to develop and operationally maintain a corprehen-sive quality assurance program to satisfy contractually invoked req ui rements.
O 17C-8 Amend. 70 Aug. 1982
OUALITY ASSURANCE ENGINEER!% GROUP - SUPERVISOR l1.4.3.1 The Individual responsible for supervising the AE quality assurance engineering f unction will have the following qualifications:
Education - He shall be a graduate of a four-year accredited eagineering college or university.
l Exoerlence -
General - He shall have a minimum of 7 years experience in quality assurance or engineering associated with design, construction, or operation of a nuclear reactor plant f acilIty, power generating station or heavy industry.
Soecialtv - He shall possess knowledge and understanding of Industry and government codes, standards and regulations which def ine quality assurance requirennts and practices. He shal l be f amil iar with methods of application of programmatic and special quality assurance requirements in design and procurement documents. He shali be experienced in evaluating program plans, procedures and practices, and subsequently verifying conformance.
Suoervisorv - He shalI be experienced in the supervision of technical and adninistrative personnel engaged in quality assurance or engineering activities.
O V l 1.4.3.2 INTERNAL AUDIT AND SURVEILLANCE GROUP - SUPERVISOR The Individual responsible for supervising the AE quellty assuranca internal auditing and survelliance f unction will have the following qualifications:
Education - He shall be a graduate of a four year accredited college or university, or be a high school graduate and have 10 years experience in quality assurance / control in lieu of a degree.
l Exoerlence -
General - He shalI have a minimum of 7 years experience in quality assurance or engineering associated wIth design, construction, or operation of a nuclear reactor plant f acilIty, power generating station or heavy industry.
Specialt.y - He shall possess knowledge and understanding of Industry and government codes, standards and regulations which define quality usurance requirements and practices. He shalI be experienced com-mensurate with the scope, compicxity or special nature of the activi-ties to be audited. He shall possess good communicative skills.
Suoervisory - He shall be experienced in the supervision of technical personnel engaged in quality assurance auditing or surveillance I activities.
O 17C-9
/vnend. 70 Aug. 1982
1.4.3.3 VENDOR. AUDIT AND SURVEILL ANCE GROUP - SUPERVISOR The Individual responsible for supervising the AE vendor audit and survell-lance f unction will have the following qualifications:
Education - He shall be a graduate of a four year accredited college or university, or be a high school graduate and have 10 years experience in quality assurance / control in lieu of a degree.
Exoerlence -
General - He shalI have a minimum of 7 years experience in quality assurance tangineering, inspection, supervision or testing associated with design, construction, or operation of a nuciar reactor plant facility, power generating station or heavy industry.
Soecialtv - He shall possess knowledge and understanding of industry and government codes, standards and regulations which define quality assurance requirements. He shalI be experienced in establ1shing and implementing programs, plans and practices for product inspection, process surveillance, and auditing of vendor QA progrevs. He shall have a broad knowledge and understanding of NDE, special processes and equipment test methods, including evaluation of vendor quailfications and capabilities.
_Suoervisorv - He shall be experienced in the supervision of technical personnel engaged in Inspection and test verification, vendor surveillance or auditing activities.
1.5 COMMUNICATIONS The free, continuous, unimpeded flow of communications both horizontally and vertically within organizations as well as between the BOP Supplier's organization and other program participants is essential. The free exchange of information between responsible individuals is also essential to the expedient execution of quality assurance activities.
l l To promote the flow of communications and to assure positive attention to i quality problans within the 80P Supplier's Quality Assurance Program, lines of I
communication between the Owner and the organization of the major contractors and suppllers are estabiished as foilows:
- 1. Communications by Senior Management - These communications will deal wIth such matters as major changes in the scope of the Quality Assurance Program. Communications will be addressed to the responsible senior management official with copies to the major program participants cognizant of the subject.
- 2. Communications by OualIty Assurance Procram Management - These communications will provide a direct formal or Informal exchange of l Info ~mation beiveen the Owner's Assistant Director for Quality Assurance and other quality assurance managers of organizations 0
170-10 Amend. 77 May 1983
which will have a direct Interface with the Owner. Copies of such consnunications will be distributed to other major program participants cognizant of the subject.
- 3. I'- unications hv Individuals - These communications are encouraged to identify and evaluate quality problems and to initiate, recommend, or provide solutions. Communications may also be formal or informal, the choice of which shalI depend on the significance of the subject and the judgment of the Individuals Involved.
Consnunication of quality assurance related activities within the BOP Supplier's organization is promoted through:
- 1. Periodic staff meetings of the Project Director.
- 2. Monthly quality assurance program progress and status reporting.
Communication of quality assurance related activities between the BOP Supplier's organization and the organizations of the major contractors and suppliers is promoted through:
- 1. Quarterly management review meetings attended by the Quality Assurance Managers of the major program participants.
- 2. Monthly quality assurance program progress and status reporting by major contractors and suppliers.
2.0 OUALlTY ASSURANCE PROGRAM 2.1 EEQGBAE.EE2HREERTS The requirements for the BOP Supplier Quality Assurance Program are contained in RDT F 2-2, 1973, with Amendments 1, 2, and 3, Quality Assurance Program Requirements. Execution of a program which meets these requirements will comply w!th 10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear i
Power Plants and Fuel Reprocessing Plants. A description of the activities of this program is contained herein and is presented to demonstrate how the BOP Supplier is meeting the applicable criteria.
1 2.2 PROGRAM ELEMENIX l The elements of the BOP Supplier Quality Assurance Program which will be executed by the BOP Supplier organization are those identified as "Progran Management" complimented by those for design and development, procurement and manufacturing shown in Figure 17C-2. This is the BOP Supplier portion of the program, and it will be executed in accordance with plans, procedures and management practices of the BOP Suppller. Figures 170-3 and 17C-4 contain a matrix of the plans and procedures and Indicates the program elements as l defined in 10 CFR 50, Appendix B and RDT F 2-2 respectively, that will be implemented in accordance with specific documents. A listing of these procedures with a brief description of each is contained in Attachment 1.
These documents are contained in the manuals of procedures which have been developed by the BOP Supplier organizations to implement the quality assurance policies, goals and objectives described in this program description.
17C-11 Amend. 70 Aug. 1982 l
2.3 PROGRAM IMPLEMENTATION As described in Section 1.0, of this Appendix, the BOP Supplier's principal operations of ficer is the Project Director of the CRBRP Project. The Director has day-to-day management overylew involvement in the BOP Suppller Quality Assurance Progra and the execution of frie procram.
Responsibility for the execution of the BOP Supplier Program rests with the 80P Suppller Quality Assurance Organization. Therefore, the quality assurance progem management practices are perf ormed primarily by this Quality Assurance Organization. Other units of the BOP Suppller participant organizations provide assistance for quality assurance purposes, in the execution of program activities.
To determine progra status and to evaluate program adequacy, the BOP Suppller executes an overall program review practice in the form of a regular management review of the quality assurance program to assess the adequacy of i
its scope, implementation and ef f ectiveness. This review is perf ormed by the BOP Supplier Quality Assurance organization using resources and information at its disposal.
The Project Director receives the results of these management reviews in the f orm of regular monthly progress and status reports and other special reports as appropriate. These reports outline the progress and status of quality assurance activitics, probims and nonconformances, quality trends and results of audits. The Project Director reviews these reports and initiates whatever management action is required to improve conditions and f urther implement the progrm.
In addition to the Project Director's review and assessment of the quality assurance progrm, an annual review and evaluation of the Project including the quality assurance program, is perf ormed by a select committee appointed by the Breeder Reactor Corporation (BRC). The program is also subject to periodic review by the Project Steering Committee (PSC), either by itsel f or by some other organization on an ad hoc basis as they may choose.
The delegation of execution responsibility for program elments is accomplish-ed through contracts. These contracts will specify applicable requirements of 10 CFR 50, Appendix B for contractor quality assurance programs.
The BOP Suppller has established a management procedures system in which sig-nificant plans and actions are documented including those af fecting quality.
This system provides a mechanism wherein the policies and objectives of the Project and the Project Quality Assurance Program are def ined, documented and promulgated throughout the BOP Supplier organization. Within this system, detailed procedures for mandatory actions have been prepared, approved, and issued for use as described in Section 5 of this Appendix. Each procedure is issued by a directive signed by the Project Director or by the AE Vice-President, Breeder Reactor Division requiring implementation. These proce-dures def ine both the action to be perf ormed and the responsible person or group f or perf orming the action. To provide positive Identification and control of required procedures for quality assurance activities, manuals containing these procedures have been assembled and issued and are closely controlled by the BOP SuppIIer participating organizations. These manuals 17 C-12 Amend. 70 Aug. 1982
each contain copies as appropriate of the Quality Assurance Program implement-Ing documents listed in Figures 170-3 and 17C-4, including the program description contained in this Appendix to Chapter 17 of the PSAR. A brief l
synopsis of those procedures is contained in Attachment 1.
The Quality Assurance Manuals are controlled using a document control log
- which shows the distribution of each copy by copy number including the
- distribution of revisions. The Quality Engineering Branch of the Project i Office Quality Assurance Division is responsible for this activity as well as the revision and incorporation of changes to the manual defined and approved l by the Assistant Director for Quality Assurance. The contents of the Quality Assuronce i4anual, including the program description contained in Appendix C to Chapter 17 of the PSAR, Is reviewed annually as a minimum and is updated as required to maintain it current, i
in the execution of the program, should a disagreement arise from a dif ference of opinion between quality assurance personnel and other Project Of fice i personnel (engineering, procurement, construction, etc.), the principals i
themselves try to work it out. Should they falI to resolve the difforences, the heads of the respective Divisions are briefed on the problem by the principals and they attempt to resolve the differences on their level. Should they f all also, the problem is presented to the Project Director by the heads of the Divisions involved, and he arbitrates the matter and renders a decision.
Personnel of the Owner's organization, who are assigned responsibility for verifying that contractor performance is in accordance with requirements are O selected and assigned to their area of responsibiIty based upon experience,-
education, and management's assessment of their performance capabilitles.
They are observed for performance evaluation on a continuing basis by appropriate management. Orgoing training and Indoctrination programs are conducted to famillarize personnel with technical objectives of the activity being monitored, the requirements that it must meet, the practices and procedures to be executed in verifying conformance to requirements, and the documentation of results.
3.0 DESIGN CONTROL i
The responsibility for execution of design control practices relative to BOP equipment manuf acturing will be delegated to equipment suppliers by contract.
Each equipment suppller that is assigned responsiblIty for design is required by contract to exercise design control practices in accordance with specified requirements. These practices include the following, as appropriates o Design Planning o Design Dsfinition and Control
- 1. Design Criteria
- 2. Codes, Standards and Practices
- 3. Engineering Studies and Analyses
- 4. Parts, Materials and Processes
- 5. Design Descriptions
- 6. Specifications, Drawings and instructions i
17C-13 Amend. 77 May 1983
- 7. Identification
- 8. Acceptance Criteria
- 9. Interface Control
- 10. Engineering Holds
- 11. Calculations
- 12. Computer Codes o Document Review and Control
- 1. Document Reviews
- 2. Document Control
- 3. Engineering Drawing Lists
- 4. Drawing Checks o Design Reviews (Required design verification for the level of design activity accomplished is to be perf ormed prior to release for procurement, manuf acture, construction, or release to another organization for use in other design activities. In all cases, the design verification is to be completed prior to relying upon the component system or structure to perf orm its f unction.)
o DeveIopment (including provisions that prototype component or feature testing is perf ormed as early as possible prior to Installation of plant equipment or prior to the point when the Installation would become irreversible.)
o Failure Reporting and Corrective Action The required practices are to include the review of design drawings and specifications by the Quality Assurance organization to assure that the documents are prepared, reviewed, and approved in accordance with internal procedures and that the documents contain the necessary quality assurance requirements such as inspection and test require-ments, acceptance requirements, and the extent of documenting Inspection and test resul ts.
The BOP Supplier monitors major suppIIers' design control practices and periodically audits their practices to specified requirements to assure proper implementation and adequacy of the practice.
4.0 PROCUREMENT DOCUMENT CONTROL The BOP Supplier has established and implemented a practice for control of procurament documents to assure that procurament f unctions are accomplished in accordance with the applicable codes, standards, drawings, and specifications.
This practice is carried out under written procedures which provide for coordination and implementation of procurement planning and review of procure-ment documents such as preprocurement plans and purchase orders, and changes thereto by designated personnel to assure that these documents are complete and correct.
O 17C-14 Amend. 70 )
Aug. 1982 1
.__-_ ________________---___ a
Procurement documents are prepared by the AE and submitted to the Owner for ,
review and approval. This practice will include the preparation of procure-ment documents to contain the following:
o Scope of Work o Technical Requirements o Quality Assurance Program Requirements o Right of Access
, o Special Quality Assurance Requirements o Documentation Control o Nonconformance Control o Transfer of Requirements to Lower Tier Participants This practice wil l al so incl udet o Procurement Document Review and Approval o Document Control (Release, Distribution and Changes)
This practice also provides:
- 1. That procurement documents require suppliers to have and implement a documented quality asurance program for purchased materials, equip-ment, and services to an extent consistent with their importance to saf ety and utility;
- 2. That the purchaser has evaluated the supplier before ihe award of the purchase order or contract to assure that the supplier can meet the procurement requir.sments; and
- 3. That procurement documents for spara or replacement items will be subject to controls at least equivalent to those used for the original equipment.
An integral part of the document preparation process is the provision that the originating engineer is required to prepare a list of documents required by j the procurement actions including the location of the technical specification t requesting the document, together with the reason for the request. This Information arves as the input to a vendor document control program maintain-ed by the AE Computer Department. This computerized file maintains a continu-ous record of the status of ali documentation requested from the vendor.
Whenever there are no code requirements involved, the vendor is directed to send copies of all documents to the AE for collection and eventual forwarding i to the Owner. Whenever code requirements are involved, the vendor is directed to comply with the code for the accumulation of documents and the storage period required by the code. At the completion of the specifled code storage
, period, the vendor is contractually directed to contact the Owner for
- Instructions concerning the disposition of all records. The general and
( special conditions to each procurement action include contractual requirements l granting the procurement agency's right of access to vendor f acilitles and 1
( records for source inspection or audit. l O 17C-15 Amend. 70 Aug. 1982
l When acquisition of spare or replacement parts is a soparate procurement action from that involved in procurement of original equipment, the procure-mont action is subjected to the same system of internal technical and quality assurance review as the original component. All quality assurance require-monts and acceptance criteria are of the same level as the original procure-ment action.
When a plant procurement document is received by the BOP Supplier, it is routed to the Assistant Director for Administration and Contract Management (AD/AC). The AD/AC coordinates the BOP Supplier review in conjunction with the cognizant engineer. The review is conducted thoroughly, but as promptly as possible. To conduct reviews and expedite approval of procurement documents, a checklist is used by procurement document reviewers. These checklists are rather extensive and include such check points as whether or not the procurement documents:
- 1. Identify the documentation (e.g., drawings, specif ications, procedures, inspection and fabrication plans, inspection and tost reoerds, personnel and procedure qualifications, and material, chomical and physical test results) to be prepared, maintained, and submitted as applicable to the purchaser for review and approval.
- 2. Contain or reference the design basis technical requirements including the applicable regulatory requirements, components and material identification requirements, drawings, specifications, codes and Industrial standards, test and inspection requirements, and special process Instructions for such activities as welding, heat treating, nondestructive testing, and cleaning, g)
- 3. Identify the applicable 10 CFR 50, Appendix B requirements which must be complied with and described in the supplier's QA program.
- 4. Identify those records which shall be retained, controlled, maintain-ed, or delivered to the purchaser prior to use or installation of the hardware.
- 5. Contain the procuring agency's right of access to supplier's f acilities and records for source inspection and audit.
- 6. Provide for spare or replacement parts of safety-related structure, systems, and components are subject to controls at lease equivalent to those used for the original equipment-All changes and revisions to procurement documents are subject to the same review and approval requirements as the original document.
l Procurement document reviewers are specified by the AD/AC, and the cognizant engineer. Reviewers are selected on the basis of their qualifications and their ability to provide a meaningf ul input to a particular document. The l selected BOP Supplier reviewers include as a minimum, the AD/AC of his designated representative, the head of the BOP Supplier quality assurance organization or his designated representative and the BOP Supplier Cognizant Engineer. The selected reviewers also include contract and procurement staff and, as appropriate, counsel, patent and finance representatives. The 17C-16 Amend. 77 May 1983
I respective cognizant engineer has the principal responsibility for determining
, the sul1rability and adequacy of technical specifications included in procurement documents.
O The quality assurance reviewer has the principal responslos s i;y for deter-mining the adequacy of the procurement documents regarding quality assurance requirements. He is trained and qualifled in quality assurance proctices and concepts to make this determination. This review is to determine that quality and quality assurance requirements are correctly stated, inspectable, and controllable; there are adequate acceptance and rejection criterla; and the procurement document has been prepared, revleved, and approved in accordance with QA program requirements. . ,
l The AD/AC has the principal responsibility for determining the overall adequacy of administrative, financial and contractual aspects of procurement ;
documents. .- ,
All BOP Supplier comments f rom reviewers of .proetrement documen'ts areyade in writing. -
l After receiving and resolving all comments, the AD/AC prepares formal chres-pondence to the appropriate participant and reflects concents, approval, or notifies the participant of the reasons for disapproval. '
TheBOPSuppllerorganizationbothparticipatesinandmorsitorsti$eexecution of this practice. Periodically, the organization audits or arranges for-Independent audit of this practice to assure proper implementation and adequacy. '
The execution responsibility for procurement document. control practices related to BOP equipment manufacturing is delegated to equipment suppliers-by
- contract. ,
5.0 INSTRUCTIONS. PlW wnllRES AND DRAWINGS
~
l The BOP Suppller has prepared its procedures and instructions in accorIiance with procedures that prescribe the format to be followed and the identi- '
fication system to be used. These procedures cover al l activities of manage-ment, document review and control, procurement, surveill'ance activities, audits, and records management. These procedures prescribe methods for performing quality-related activities in conformance with the applicable requirements of 10 CFR 50, Appendix B.
The BOP, Supplier procedures are organized under a Management Procedures System which is administrated by a Procedurce Coordinator from within the BOP Supplier. organization. The Procedures Coordinator, which is a staff function e reporting to the Project Director, is assigned the function'of controlling the issuance of procedures to assure coordination and consistency in format, content, etc. The procedure system itself is organized along divisional lines (Engineering, Construction, Qual ity Assurance, Public Saf ety, .0perations, - 1 l Administration and Contract Management, Management Systems, and others) which give the respensible managers the responsibilities for: "
- L, u,
17C-17 Amend. 77 May.I983 x
~ . s.
o Assuring that policies of a continuing nature are incorporated in the Idanagement Procedures System (MPS).
o incorporating applicable laws, star.dards such as 10 CFR 50, Appendix B, Executivo Orders, decisions and directives of the Project Steering Committee (F M) into the procedures to the extent necessary to show the requiremanis placed upon the BOP Supplier.
o Determining the coverage and content of management directives mecessary to carry out their assigned functions, assuring the accuracy and currency of the procedures and arranging for the cancellation of thoso that.become obsolete.
I o Approvfng procedures for which they are responsible. Obtaining l, review, comment and document concurrences by other organizational r units when appropriate.
o Submitting L the Procedures Coordinator:
- a. Draft procedures for review of format.
l b. Final p ocedures for Director cpproval, Issuance and distributton.
o Determining, with concurrence of General Counsel, what portions of procedure:, if any, shall be communicated to the contractors.
Furnishing fo the Procedures Coordfr.ator the names of contractor personnel to whom such material together with any appropriate supplementary explanation or instructions should be distributed.
The Procedures Coordinator is the prime control officer for BOP Supplier Procedures and as sucht o Assures tha't stylc, format, content, terms, titles and numbering sequence conform to the m quirements of the Management Procedures Control System.
c After final epproval, oversees printing and distribution of the proceduras.
'o Assures That a master. file containing the originals of all approved CRBRP Project Office procedures and a log of issu3d procedures and 5
their revisions is maintained.
o Assures that the Index of procedures Is maintained current.
o Assures that all revisions and changes to existlig procedures, are distributed promptly, according to the distribution in the master file.
Organizational Unit W nagers are responsible for writing and implementing the procedures necessary fo- their division. General Administration procedures cover policies and procedures which apply to all employees. The Project Director approves for issuance all CRBRP Project Office procedures. The Individual divisicn procedures are approved by the responsible Division l Manager and are forwarded by the Procedures Coordinator to the CRBRP Project Director for final approval.
170-18 Amend. 70 Aug. 1982 l
Each new procedure or revision of existing procedure is preparea using the Management Procedures system numbering code and f ormat.
Each Division establishes steps for the review of draf t procedures within the Divisions. If a procedure applies to more than one Division, the other Divisions af fected receive the draf t procedure for review. A draft is sent to the Procedures Coordinator who reviews it for format, style, and numbering sequence. -
The final procedure or revision of existing procedure is approved by the appropriate Division manager responsible for that particular subdivision of l procedures, is concurred in by the Assistant Director for Quality Assurance and is approved by the CRBRP Project Director, and released for implementation.
Distribution of each procedure or revision of an existing procedure is listed and filed with the procedure copy in the procedure master file. The rejlster shows which revision is current.
The BOP SuppIIer practice for documenting, in written form, the requirements f or and results of activities af fecting quality is, itself, oxecuted in i accordance with document control procedures identified under Section 6.0, )
Document Control.
l The BOP Supplier Quality Assurance organization both participates in and '
monitors the execution of this practice. Periodically, the Quality Assurance Organization audits or arranges for Independent audit of this practice to .
's assure implementation and adequacy. l
)
Responsibility for prescribing in instructions, procedures and drawings activities af fecting quality relative to equipment manuf acturing will be delegated to suppliers by contract. These activities include the assurance that procedures are established and controlled to provide for the preparation of as-built drawings and related documentation in a timely manner to accurately reflect the actual plant configuration.
6.0 DOCUMENT CONTROL The BOP Supplier has established and implemented a document control system that fulfills the Quality Assurarce Program requirements and applies to those types of documents prepared by the 80P Supplier and identified in Sections 3, 4, and 5 of this description.
The controlled documents originated by the BOP Supplier are processed in a controlled manner to assure the following:
o Unif ormity of format of initial and subsequent issuances.
o Proper identification as to the originator and date of origin of a 1 document, and a mechanism for verification of the authenticity of Inf ormation.
o Positive review and approval by persons qualified to determine the correctness of the information presented and to judge its ultimate y usefulness. ,
170-19 Amend. 77 May 1983
o Prompt and accurate distribution, including a mechanism for receipt control, of both the original document and subsequent revisions to prevent inadvertant use of superseded material and to place documents in work areas in a timely manner.
o Efficient revision of documents when necessary to clarify, correct, augment or up-date the content of a document, while preserving the inrogrity of originally approved and released information, e
o Documents are avallable at the iocation whers the activity wIIl be performed prior to commencing the work.
o Quality Assurance re'q uirements are properly stated, are adequate and are incluaed prior to implementa1!on.
Controlled documents are standardized by proceduro as to identification, format, and numbering. These documents are reviewed for adequacy by Division Assistant Directors and/or the CRBRP Project Director, as appropriate. The Assistant Director of the Division originating the controlled document determines the extent of necessary reviews. The draft controlled document is routed to the appropriate reviewing personnel / organizations. The quality assurance organization reviews and concurs with these documents with regard to quality assurance related aspects. Comments of reviewing personnel are resolved prior to final approval of the document. A record of the review sequence which has been accomplished is documented and retained. Changes or revisions are reviewed and cpproved by the same Divisions that porformed the original review and approval. If the controlled document will be issued only to personnel of the originating Division, the respective Division Assistent Director may approve the document for issue upon completion of necessary reviews. If the controlled document is to be issued to personnel outside of the originating Division, the respectivo Division As Istant Director secures any necessary highce. level approval s. Tne Assistant Director of the Division originating a controlled document establishes an appropriate minimum periodic review schedule for the approved document. The primary purpose of these reviews is to detcrmine if changes in Project status have resulted in the need for revisions to the controlled documents.
The originating Division establishes and maintains an appropriate listing of the distribution of the document upon issue. A receipt page is attached to tho transmitted controlled document which requests the person receiving the document to sign and date the page and return it to the originating Division.
A designated person initials the respective distribution listing upon receipt
- of the signed page to reflect accomplishment of transmittal and receipt. A designatsd person also reviews the Division's controlled document distribution listing at least bimonthly to follow-up on any delinquent receipt pages. This distribution listing is a master list which is updated periodically to show current rev ision, number distributed, location, etc. Revisions to controlled documents are systematically processed with the same procedure as the original. Changes are also reviewed and approved by the same Divisions that performed the original review and approvals.
The BOP Suppller quality assurance organization both participates in and monitc. s the execution of the document control system. Periodically, the quality assurance organization audits er arranges for independent audit of the document control system to assure implementation and adequacy.
17C-20 Amend. 77 May 1983
(v %
REftt!REMENT OF 10 CFR 50 APPEISIX B IWLDIENTING DOCISENT REMARKS INSTRUCTIONS CRITERION TITLE D0C. NO. TITLE REF. 00C.. ETC.,
i BRD-QA-3.101-3 Preparation for and Performaere of Surveillances l BRD 'jA-3.101-4 Preparation and Issuance of Source Surveillances Vill Identification and Control of Materials. Parts and Compo- )
nents j l It Control of Special Processes Delegated X Inspection CRP-QA-02 Activity Planning N CRP-QA-16 Inspection. Examination and Test y RC-t;A-3.101 Source Surveillance Source Surveillance Planning e BC-QA-3.101-4 4
4 XI Test Control CRP-QA-02 Activity Planning CRP-QA-13 Perfor1 nance of Project Surveillance CRP-QA-16 Inspection. Examination and Test ,
ORD-E-2.4 Vendor / Contractor Documents l BRD-44-1.16 QA Review of Submittals XII control of Measuring and Test Equipment N + 17 Nasuring and Test Equipment Callbrition and Control XIII Handlieg. Storage and Shtpping Delegated XIV tion. Test and Operating BRD-QA-3.101-2 Administration of the Source Surveillance Program I
QUALITT ASSURANCE PPMPAM INDEK YtRSUS REQUIREMENTS OF 10 CFR 50. ##PENDIA B
(( Figure 17C-3 (Cont'd.). Balance of Plant (B0P) Supply Quality Assurance Program Index a
G; -
mu NO
I l
l REQUIRDtENT OF 10 C rR 50 APPENDIX B IMPLEPENTING DOCUMENT REMARKS INSTRUCT!DNS CRITERION. TITLE DUC NO. TITLE REF. 00C.. ETC .
IV Nonconforming Materials CRP-AA-04 Incomir.g Hall
(
i Parts and Components CRP-TN-02 Processing Principal Design Dccinents CRP-OP-02 Operations Division Review and Concurrence with Engineering Design Data CRP-PR-02 TVA Purchases of CR8RP lters CRP-QA-03 Control of Noncor.formances CRP-QA-05 Processing of unusual Occurrerece Reports CRP-QA-27 Unusual Occurrence Report Preparation and Disposition BRD-E-2.5 Vendor / Contractor Walver Requests BRD-QA-1.13 Corrective Action Request (CAR)
Z n
PRC.QA-1.25 BRD-QA-1.1000 Monconformance Review Board (N23)
Deviatton Reporting and Control e RRD-L-2.3 Reporting of Defects and Noncompliance Stop Work Orders su XVI Corrective Action CRP-QA-28 CRP-EN-07 Processing Principal Design Documents Engineer.ng Design Data CRP-QA-03 Control of Wonconformances CRP-QA-04 Corrective Action Requests CRP-QA-05 Processing of Unusual Occurrence Reports CRP-QA 06 Nonconfomave. Unusual Occurrence and Corrective Action Analysis CRP + -09 Quality Trend Analysis CRP-QA-27 Unusual Occurrence Report Preparation and Disposition BRD-QA-1.13 Corrective Action Request (CAR)
BRD-QA-1.25 Nonconformance Review Board (NRB)
BRD-0A-1.1003 Deviation Reporting and Control CRP-QA128 Stop Work Orders QUALITY ASSURANCE PROGRAM INDEI VER$US REQUIRLMENTS OF 10 CFR 50. APPENDII B M :s Figure 17C-3 (Cont'd.). Balance of Plant (B0P) Supply Quality Assurance Program Index LD
- co 4 9 e
O O F i
]
REQUIREKMT OF IMPLEMENT!WG DOCUMENT REMARKS RDT F2-2 SECTION TITLE D0C. M3. TITLE INSTRUCT!0NS NUMBER REF. 00C.. ETC.
4 2 Management & Planning 2.2 Quality Assurance Program Quality Assurance Program Cescription Quality Assurance Charter
, BRD-PC-1.4 Qiaality Assurance Plan 2.2.1 Planning Quality Assurance Program Description BRD-PC-1.4 Quality Assurance Plan CRP-QA-02 Ativity Planning y 2.3 Organization l 2.3.1 Responsibility and Authority Quality Assurance Program Description A Quality Assurance Charter
- All Procedures Responsibility BRD-PC-1.4 Quality Assurance Plan Sections CRP-DR-02 Organization Plaa *-d Functional Statements 2.3.2 Training and Indoctrination CRP-QA-24 Personnel Indoctrination BRD-PC-7.1 Indoctrination and Trahing 2.3.3 Personnel Qualification CRP-QA-25 Administration of Personnel Certification and Records CRP-QA-26 Personnel Certification BRD-0A-1.18 Training and Certification of Quality Assurance Personnel PPC 4,05 Confirmation of Position Requirements and Employee Qualifications 2.4 Documentation 2.4.1 Politics and Procedums Quality Assurance Program Description All Procedures Policy Sections CRBRP Managem nt Policies and Requirements CRP-AA-01 Manageant Procedures CRP-AA-03 Preparation of Correspondence CRP-AA-11 Control of Project Office Procedures Manual QUALITY ASSURANCE PROGRAM INDEX VERSUS REQUIRE M NTS OF RDT F 2-2 Figure 17C-4 BALANCE OF PLANT (B0P) SUPPLY QUALITY
- [
ASSURANCE PROGRAM INDEX L~
__ . . _ _ _ _ _ _ _ _ _ . _ _ - __ _ _ _ _ _ _ _ _ 4
REQUIRD*ENT OF RDT F 2 2 IMPLEMENTING 00CtMENT REPARKS TIM INSTRUCTIONS R TITLE ,REF. 000.. FTC.
TITLE C00. NO.
CPP-AA-14 Controlled Documents CRP-QA-20 Preparation. Maintenance and Control of Project Office Quality Assurance Nnual tRD-PC-1.4 Quality Assurance Plan. l BRO-QA-1.2 Preparation. Control and Distribution of Quality I Assurance Instructions i BRD-PC-1.5 Precedure Preparation j ERD-PC-3.6 Distribution BRD-QA-1,3 Preparation of Quality Assurance ProceJures BRD-QA-1.1E Procedure Writing Format w
N 2.4.2 Quellty Records CRP-AA-02 Filing Procedure far Official Froject Files
?
A CRP-QA-07 Qaality Records Preparation. Transfer, and Receipt of Project Office cxP-QA-23 O Quality Records RND-PC-3.1 Filing 2.4.3 Quality Status Reports CRP-AA-07 Seports control Progres
' CRP-PC-03 ~ CR8AP Project Monthly Progress Report CRP-QA-01 Quality Assurance Program Management Review Meeting CRP-QA-08 Quality Assurance Program Prcgrass end Status Review and Reporting CRD-PC-1.4 Quality Assurance Plan 2.5 Audits and Reviews 2.5.1 Quility Audtts See Section 8 2.5.2 Management Reviews CRP-QA-01 Quality Assurance Program Panagement Review Meettrgs 2.6 Ccrrective Action CRP-AA-06 Centralized Action Correspondence Control System CRP-QA-04 Corrective Action Requests CRP-QA-05 Processtng of Unusual Occurre x e Reports QUALITY ASSURANCE PROGRMI INDEX VERSUS REQUIREMENTS OF RDT F 2-2
=L' w
G.
$cn
- Figure 17C-4 (Cont'd.). Balance of Plant (B0P) Supply Quality Assurance Program Index 9 9 e
m O b MQUIRDENT OF RDT F 2-2 IIFLDENTING OOQNNT RDIAPR$
SECTION INSTRUCTIONS M IR TITLE 00C. NO. TITLE REF. 00C.. ETC.
4.6 Control of Configuration Partially Delegated 4.6.1 Contract Change Control 4.6.2 As-Built Verification CRP-QA-02 Activity Planning CRP-QA-13 Performance of Project Surveillance CRP-QA-16 Inspection. Examination and Test y BRD-y4-3.101 Source Surveillance.
BRD-QA-3.101-2 Administrative of the Wrce Survelliance Program e
A 4.7 Measu-ing and Test Equipment
- Calibration and Control CRP-QA-17 Measuring and Test Equipment Calibration and Control 4.8 Source Surveillance and Inspection CRP-QA-02 Activity Planning CRP-QA-13 Perforamnce of Project Surveillance CRP-QA-16 Inspection, taamination and Test BRD-QA-3.101 Source Survelliance BRD-QA-3.101-4 Preparation and Issuance of Source Surveillan*.e Reports BRD-QA-3.1000 Project surveillance BRD-QA-3.1000-1 Preparation of Project Surveillance / Acceptance Deck-lists. Sunenary and Report For1ms BR0 QA-3.101-3 Preparation for and Perfonnance of Survelliances QUALITY ASSURANCE PROGRAM INDEX VERSUS REQUIREMENTS OF RDT F 2-2 EE
- g. Figure 17C-4 (Cont'd.). Balance of Plant (B0P) Supply Quality Assurance Program Index 105 t
REQUIRDtENT OF REMARS RDT F 2-2 IMPLtWNTING D0ctMtWT INSTRUCTIONS SECTION NUMBER TITtt D0C. NO. TITLE REF. 00C.. ETC.
4.9 Receiving lespection 4.9.1 Planning an.f Inspectica CRP-QA-02 Activity Planning CRP-QA-16 Inspotion, t samination and Test 4.9.2 Documentation CRP-AA-04 Incoming N 11 CRP-AA-14 Contro114J Documents CRP-[N-02 Procesitag Principal Design Documents CRP-OP-02 Cyrations Diviston Review and Concurrence with
!naWering Design Data s CRP-PR-02 TVA Purchases of ERMP ltems N CF#-QA-10 Quality Assurance Review and Approval of Engineering
?
m CRP-QA-12 Docwents Review of Contractor Quality Assurance Plans and o Prreases U.3 E-2.4 Vendor / Contractor Dccuments 3;t3JA-1.16 Quality Assurance Review of Submittals BRDe'A 1.16-1 Review of Design / Document Submittals 4.9.3 Disposition of Received ! Lees 4.10 Control er Noncorforming Control of ftonconfonnances Delegated CRP-QA-03 Items BRD-E-2.4 Vender / Contractor Documents BRD-QA-1.25 !!oeconformance Revleur Board (ES)
BRD-QA-1.1000 Deviation Reporting and Control BRD-E-2.5 sendor/ Contract Wilver R* quests BRD-L-2.3 Reporting of Defects and honcompliances BRD-CA-1.13 Corrective Action Request (CAR)
CRP-QA-28 Stop Work Orders Delegated 4.11 Control of Received Items 4.12 Quality Audits See Section 8 QUALITY A$$URANCE PROGR#l INDEX VEE $US REQUIRDr.1413 0F RDT f 2-2 EN ta ."
Figure 17C-4 (Cont'd.). Ba' lance of Plant (B0P) Supply Quality Assurance Program Index
$3 9 9 9
Pranaratlon. Malnhnence and Control of Profact Office OualItv Assurance Manua1 (CR M A-20)
This procedure defines the actions and responsibilItles for the preparation, distribution, maintenance and control of the CfBRP Quel!ty Assurance Manual.
Conduct of Product Audits (CRfM)A-21)
This proceduro defines the actions and responsibliitles for the preparation, conduct and reporting of quality assurance product audiis by the CRBRP Project Office. The procedure also details the actions of the audit team in the course of the evaluation of selected products for conformance to, quality requirements.
Conduct of Procr- - -t Ic Aud I ts ( CRf'-OA-22 )
This procedure defines the responsibilitles for the preparation, conduct, and reporting of quality assurance progranmatic audits by the CRBRP Project Office. The procedure also details the actions of the audit team in the cc,urse of the evaluation of progranmatic practices for conformance to quality program requirements.
Prenaration. Transfer. and Recelot of Prolact Office Oualltv Records (CRP-OA-23)
This procedure defines the responsibilities and actions to be executed by each Project Office Division in the preparation and transfer of quality records to the Quality Assurance Division. The procedure also defines the ;
Os responsibliities and action of the Quality Assurance Division when receiving j quality records f rom other Project Of fice Divisions.
Escapnnel Indoctrination (CRP-OA-24) l This procedure defines the responsibilities and actions to provide for the
! indoctrination of CRBRP Project Office personnel who carry out duties af fecting the quality of CRBRP Plant structures, systems and components.
Administration of Personnel Certification and Records (CRP-OA-25)
This procedure defines the responsibilities for the administration ot r certification for Quat ip/ Assurance Division personnel directly involved in quality verification, testing, evaluation or audit activities. The procedure also details the actions associated with collection and maintenance of records pertaining to personnel certification.
Personnel Certification (CRP-OA-26)
This procedure defines the responsibilities and actions necessary to identify
, areas of quality importance for which qualification or certification of personnel are required. The procedure also details the actions for verifying the adequacy of personnel training programs, certification practices and i documentation.
O 17C-56 Amend. 70 August 1982
- - .-._ - -- - - .- . - - - -.-,-.. - _ - .- .-.._ ~ .. - - _ - - _._ - --. _
l l
l Unusual Occurrence Reoort Preoaration erJl DIsoo<.ition (CRP-OA-27)
This procedure definos the actions and responsibilities for documenting en unusual occurrence observed during the course of work on the CRBRP Project.
The procedure also details the action related to evaluation of the reportability of the event to NRC as well as tne channels for reporting to NRC.
Stoo Work Orders (CRP-0A-28)
This procedure defines the responsib!Iitles and actior,s required for issuing and processing Stop Work Orders (SW0s), which stop further f abrication, installation, or use of nonconforming items or processes which would result in a condition adverse to quality in the Plant. This procedure applios to saf ety-related work, work or act ivities important to saf ety, and nonsafety-related work which are determined to be in noncompilance with design or programmatic requirements and will be noted as such on the SWO form.
Confirmation of Position Reautrements and Ernolovee Oualifications (PMC 4.05)
This procedure defines the requirements and actions nocessary to confIr9 the position requirements and employee qualifications for PMC professional employees or to assignees from member utiiItles assigned to the CRBRP Project.
AE PROCEDURES Procadure Preoaration (BRD-PC-1.5)
This procedure establishes the method for preparation, review, approval and updating of ali Project procedures except those numbered "BRD-QA-xx."
Filing (BRD-PC-3.1)
This procedure establishes the methods and categories to be used in Project files; it contains provisions for a QA history file in compliance with Criterion Vil of 10CFR50, Appendix B.
1)lstribution (BRD-PC-3.6)
This procedure establishes the method for maintaining distribution requirements of Project documents.
.lnrygetrination and Training (BRD-FO-7.1)
This procedure establishes the requirements for Indoctrination of Project personnel in the goals, policies and procedures of the Project, training in the Project work methods and provides documentation of accomplishment of the procedure activities.
O 17C-57 Amend. 77 May 19o3
Preauallfication of Bidders (BRD-E-2 21
() This procedure establishes the method of prequalification of prospective bidders for technical and financial capability. It also provides for generation of a prospective bidder's list using the above information and data accumulated via BRD-QA-1.11.
Technical Evaluation of Bids (BRD-E-2.3)
This procedure defines the methods to be used for engineering review and evaluction of bids.
Vendor / Contractor Documents (BRD-E-2.4)
This procedure provides the methods for receipt, logging, review, processing and return of vendor / contractor documents.
Vender / Contractor Walver Reauests (BRD-E-2.5)
This procedure provides the method by which vondor/ contractor waiver requests are received, evaluated, dispositioned and the vendor notified of the results.
Reoortino of Defects and Non-Comollance (BRD-L-2.3)
'his procedure establishes the method for review of defects or non-compliance as defined by 10CFR21 and significant deficiencies as defined by Paragraph 50.55(e) of 10CFR50.
() Preoaration. Control and Distribution of Ouality Assurance Instructions (BRD-0A-1.2)
This procedure establishes the guidelines for preparation, Issue, control and use of quality assurance Instructions by QA personnel.
Prenaration of OA Procedures (BRD-0A-1.3)
This procedure establishes the eethod for preparation and control of "QA" designated procedures. .
Vendor Qualltv Assurance Preauallfication Program (BRD-0A-1.11)
The procedure establishes the meihod of prequalifying a prospective bidder's quality assurance prcgram for a bidder's list.
Vendor Qualltv Assurance Oualification Survey ( BRD-0A-1.12 )
This procedure establishes the method and criteria for conducting a preaward survey and evaluation of a prospective vendor's or subcontractor's quality assurance / quality control system.
Corrective Action Reauest (CAR) ( BPD-0 A-1.13 )
This procedure establishes the methods and documentation used in requesting corrective / preventive actions via a system of graded requests.
(~'N 17C-58 Amend. 70 August 1982
Oualltv Assurance Review of Submittals (BRD-0A-1.16)
This procedure provides guidelines for a standard approach to quality assurance review of submitted documents.
Trainina and Certification of Oualltv Assurance Personnel ( BRD-0A-1.18 )
This procedure establishes the training and certification methods for quality assurance personnel who perform ncndestructive examinations and inspection of materials, parts, structures or systems.
Procedure Writing Format (BRD-0A-1.L91 This procedure provides a guide to the standardized fccmat to be used in writing procedures fw the BRD.
Bid Review for Qualltv Reautrements (BRD-0A-1.21)
This procedure provides guicallnes for a standard approach for quality assurance review of bids.
Nonconformance Review Board (NRB) ( BRD-0 A-1. 25 )
This procedure provides guidelines for the composition, operation and generation of documentation by the Nonconformance Review Board. It provides for membership of the Authorized ASE inspector when considering items under the jurisdiction of the ,^.SE Code.
Deviation Reoorting and Control (BRD-OA-1.1000)
This procedure establishes the methods and documentation required to report and disposition deviations.
Source Surveil lance (BRD-OA-3.101)
This procedure establishes the scope, guidelines, responsibilitles and control of source survelllance activities, from initial planning through release for shipment.
Pro f ect Survei l l ance (BRD-0A-3.1000)
This procedure provides a means and guidelines for examining the ef fediveness of the Project Quality Assurance Program on a less formal basis than auditing.
l Aud its (BRD-0A-4.2)
This procedure establishes the guidelines for auditing of the Qua!Ity Assurance Program.
17C-59 Amend. 70 August 1982
Evaluation of Preoual if ication Ouestionnaire (BRD-0A-1.11-1)
Tnis Instruction provides direction for evaluating a Q. A. Prequalification O' Questionnaire when considering the suitability of a vendor as an acceptable source.
Performance. Evaluation and Reoorting of Preaward Survevs (BRD-0A-1.12-11 This Instruction defines the actions required in planning, performing and reporting the results of a preaward survey.
Review of Design / Document Sybmittals (BRD-0A-1.16-1)
This instruction provides the checklist that defines the minimum QA review of design documents and vendor submitted documents.
Preoaration of Pro tect Surveil lance /Accentance Check l ists. Sn==arv and Reoort Ecrms (BRD-0A-3.1000-1)
This instruction defines the requirements for a surveillance checklists, methods of summarizing the results and reporting them.
Quality Assurance Comoletion Record (BRD-0A-3.101-2)
This instruction provides for recording the condition of a component and its documentation immediately prior to shipment.
Preoaration For and Performance of Surveil lance (BRD-0A-3.101-3)
This instruction describes the methods and considerations addressed when preparing and performing surveillance activities.
Preoaration and issuance of Source Survei l lance Reoorts (BRD-0A-3.101-4)
This instruction defines the requirements and provides the guidance for preparation and issuance of Source Surveillance Reports, i
1
(
17C-60 Amend. 70 August 1982
ATTACHE NT II PROCEDURE RELEASE SCMEDULE Procedure Nnmher Title Scheduled Release Date CRP-QA-28 Stop Work Orders June 1, 1983 l
l l
l I
l O
I O
17C-61 Amend. 77 May 1983
4 O s F J
GE ENERGY SYSTEMS &
TECHNOLOGY DIVISION i
ADVANCED RE ACTOR SYSTEMS DEPARTMENT 1
1 1 I I 7
A ARSD FINANCIAL CLINCH RIVER APPLICATIONS ENGINEERING PRODUCT ASSURANCE ARSD-LEGAL ARSD EMPLOYEE W PROJECT
+
SECTION & PLANNING & SERVICES RELATIONS
~
l
. TECHNOLOGY &
SPECIAL PROJECTS ENGINEERING 1
January 1982
- N 4m Figure 17I-2. GE-ARSD Quality Program Management Organization '
' F R.
M Ri ~
e es
NDCLEAR FUELS SPECI AL PROJECTS DIVISION
~~~~~~~~]
l
.-.- ._. l l
ADVANCED !
r-------- REACTOR SYSTEMS l DEPARTMENT l
l --. . .
1 PRODUCT ASSURANCE !
l & -_____a SERVICES g
I PRODUCT ASSURANCE PRODUCT ASSURANCE CLINCH RIVER DEVELOPMENT PROJECT PROJECTS
- MANAGEMENT -
GUALITY ENGINEERING SYSTEMS
_., & VERIFICATi0fi PROCURED EQUIPMENT DRAFTING QUALITY DOCUMENTATION ENGINEERING
& VERIFICATION ARSD PRODUCTS TECHNICAL OPERATIONS SUPPORT
~
PRODUCT ASSURANCE AUDITS FACILITIES -
- DOES NOT MAKE DIsIECT CONTRIBUTION TO CLINCH RIVER ACTIVITIES (SUPPORTS BASE PROGRAMS AND OTHER WORK)
NOTE: OPERATIONS -
BROKEN LINES REPRESENT ORGANIZA TIONAL g gggylggg ACCESS FOR FUNCTIONAL REPORTING.
Figure 1713 GE-ARSD PRODUCT ASSURANCE ORGANIZATION Amend. 77 May 1983 171-42
TABtE OF CONTENTS (Continued)
Page No.
A.101 WESDYN A-264 A.102 W-2DB (Westinghouse Proprietary) A-266
, A.103 WRAPUP D A-268 A.104 XSRES-WlDX A-270 A.105 MPHI A-271 A.106 (RESERVED) 4 '
A.107 772 A-283 A.108 7 81 A-284 A.109 907 A-205 A.110 1017 A-286 A.111 1027 A-287 A.112 1036 A-288 A.113 1374 A-289 A.114 1392 A-290 A.115 1671 A-291 A.116 1691 A-292 A.117 1823 A-293 A.118 B0S0R4 A-294 A.119 CODES A-295 A.120 EQUILIN A-296 A.121 FLUSH A-299 A.122 MODPROP A-300 A.123 RESPECTPLOT A-303 A.124 STRUDL A-304 A.125 THAVSA A-305 A.126 2DGENFRt.ME -
A-308 O
Amend. 77 A-v May 1983
. . - - - - - . - . . , . , . - _ - - . _ - - . _ . - . . . - . . - ~ . . . . . . . . . - . _ - - _ - . _ - . -
t I
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APPENDIX A COMPUTER CODES This Appendix is a compilation of brief abstract descriptions of the computer codes used in the evaluation of the CRBRP. Where the code or a particular modification or maintained version of the codes listed below is a proprietary item of the Contractor, it is so noted in the abstract description.
- 1. AlD 27. ELTEMP 53. KENO-IV 79. SCAP-BR
- 2. ANISN-W 28. ET0X 54. LIFE-Ill 80. SETS
- 3. ANSYS 29. E0984A 55. LION 81. SNAP
- 4. APPROPOS 30. E1682A 56. LSD-2 82. SOFIRE-ll 56.A. NUPIPE 82.A. SPCA
- 5. APSA 31. E1755A 57. MAP 83. SPECEQ/ SPEC
- 6. ASHSD2 32. E1739A 58. MARC 84. SPHINX
- 7. AUTOTEM-Il1 33. FATHOM-360 & 59. ENAC 85. SPRAY-l
- 8. CACECO 3605 60. MINX 86. SUPERPIPE 8A. CATFISH 34. FBRDSAP 61. MRI/STARDYNE 87. TAP-A
- 9. CHERN 35. FESAP 62. NAPALM 88. TAP-B
- 10. CINDA-3G 36. FLODISC 63. NASTRAN 89. TAP-4F
- 11. COBRA 37. FORE-2M 64. NICER 89A. TEMPEST
- 12. 00 mADEX llI 38. FRST 65. NONSAP 90. TFEATS
- 13. (X)NL IFE 39. FULMIX 66. OCTOPUS 91. TGRV 13A. CORINTH 40. FURFAN 67. ORIGEN 92. THl-3D
- 14. COTEC 41. GAR.EG-W 68. PDA 93. TNTD/THTE
%) 15. CRA8 42. GASA 69. PERT-Y 94. TRANSWRAP
- 17. CRSSA 44. GSAP4 71. PLAP 96. TRITON
- 18. DAHRS 45. HAA-3B 72. PUMA 97. TRUMP
- 19. DEAP 46. HAFMAT 73. QAD 98 VARR-Il
- 20. DEBLIN2 47. HAP 74. RIBD-Il 99. VENTURE
- 21. DEMO 48. HAP-lI (SAP) 75. S-4 100. WECAN
- 23. DRIPS 51 . HYTRAN 78. SCAP 103. WRAPUP D 24 DUNHAM's 52. KALNINS 104. XSRES-WlDX
- 25. DYNALSS 105. MPHI
- 26. DYNAPLAS 106. (RESERVED) 107. 772 108. 781 109. 907 110. 1017 111. 1027 112. 1036 113. 1374 114. 1392 115. 1671 116. 1691 117. 1823 O
Amend. 77 A-1 May 1983
\
APPENDIX A (Cont'd.)
a 118. BOS0R4 119. CODES 120. EQUILIN 121. FLUSH 122. MODPROP 123. RESPECTPLOT 124. STRUDL 125. THAVSA 126. 2DGENFRAME O
A-la Amend. 77 May 1983
O A.1 AID AID (Accident Inhalation Dose) is a computer program written primarily for the parame tric analysis of the control room in-halation dose following a major reactor accident. AID can calculate thyroid, whole-body, bone and lung doses under various reactor containment and control building ventilation system conditions. Based on input data for meteorological conditions and ventilation system setups, the program first calculates: (1) the atmospheric diffusion factors as a func-tion of distance and time after an accident; (2) the dilution factor inside the control room according to ventilation system parameters such as filtered air intake rate, recirculation rate, distance between containment building and control room, point of air-inlet and the filter efficiency. Subsequently, the external whole-body, internal whole-body, thyroid, bone and lung doses, are calculated by the program. The AID code is based upon the nuclear power plant control room ventilation and meterological models as described by Murphy and Campe ;
(Reference 1). AID can also be used to calculate on and off-site radiation exposures for any other major releases of radioactive material.
Availability d The AID code is available at Burns and Roe in Oradell, New
, Jersey on an IBM 370/168 computer.
Verification The AID code verification has been by hand calculations as recorded in Burns and Roe,17c. internal, non-proprietary, documentation.
Application AlD is used to caiculate the internal and external whole body, thyroid; bone and lung doses to control room operators following a major release of radioactivity to ensure compliance with Design Criterion 17.
References:
(1) K. G. Murphy and R. M. Campe, 13th AEC Air Cleaning Con-ference " Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19" August 1974).
(2) TY Byoun and J. N. Conway, The Proceedings of the 14th ERDA Air Cleaning Conference, " Evaluation of Control Room Radiation Exposure," CONF-760822, August 1976.
Amend. 4t, July 1978
l A.60 MINX
. MINX calculates fine-group averaged infinitely dilute cross sections, self-shielding factors, and group-to-group transfer matrices from ENDF data. MINX incorporates and improves upon the resonance capabilities and the high-legendre-order transfer matrices of existing codes. Group structure, Legendre order, weight function, temperature, dilutions, and processing tolerances are all under user control.
Availability MINX as released in September 1976, is currently available at the Los Alamos Scientific Laboratory (LASL) computer facility in Los Alamos, New Mexico.
Verification MINX will become an integral part of the final design phase for CRBRP.
It will be used in the reference design method for calculating neutron cross section data for both the plate geometry critical experiments and the pin geometry CRBRP core. This cross section data will be used in nuclear analysis computer codes as part of the reference calculational technique to analyze critical experiments and establish bias factors and uncertainties for criticality, reactivity coefficients, control rod worths, reaction rates and other neutronic parameters. Thase bias factors and uncertainties will then be applied to the analysis of the CRBRP core, again (V) using reference design methods employing cross sections generated, in part, by MINX. This proposed verification scheme is similar to the current scheme which used the ET0X, XSRES-lDX and ANISN codes to generate cross section data.
Application MINX generates pseudo-composition independent multigroup libraries in the standard Committee on Computer Code Coordination (CCCC)-III interface formats for use in the design and analysis of nuclear systems. The out-put from MINX is used as input to the SPHINX code.
Reference C. R. Weisbin, P. D. Soran, R. E. MacFarlane, D. R. Harris, R. J. LaBauve, J. S. Hendricks, H. E. White and R. B. Kidman, " MINX, A Multigroup Inter-pretation of tiuclear X-Sections from ENDF/B," LA-6486-MS, September 1976.
O Amend. 45 A-209 July 1978
O 1
l A.61 MRI/STARDYNE The STARDYNE Analysis System is a series of compatible computer programs
- for analyzing the finite element method linear elastic structural models l for a full range of static and dynamic input conditions. The static capa-t bility includes the calculation of structural defonnations and members l loads / stresses caused by an arbitrary set of thennal and/or applied loads, l Prescribed displacement vectors can be used as input to compute resulting I
internal deformations, loads and stresses. The dynamic capability includes normal mode response analyses for a wide range of loading conditions including l transient, steady state harmonic, random and shock spectra. Dynamic response I results can, in general, be presented as structural deformations (displacements, velocities, or accelerations), and/or internal member loads / stresses.
Availability STARDYNE(CDC-8400-2500) had been available on the CDC 7600 computer of Lawrence-Berkeley Laboratory.
l l STARDYNE is also available on the CDC-Cybernet System.
Verification Documentation of verification of the STARDYNE computer code per SRP Section 3.9.1.II.2.c primary and intermediate can be found in Ne reference.
j Application l
The STARDYNE code is being used to perform global dynamic analysis of the pumps including preliminary normal modes analysis and seismic and rotor imbalance response analyses.
l STARDYNE has also been used for static and seismic analysis of structures.
l Reference MRI/STARDYN Finite Element Demonstration Problems, Document No. 84002500, Control Data Corporation, Minneapolis, Minn.
O A-210 Amend. 77 May 1983
A.117 1823, (revised 10/79) Structural Analysis for Personnel Lock Bulkhead, pV Door, Hinge, and Latching Device This progran calculates stresses and some reacticas for the personnel lock barrel, bulkhead, door, hinge, and latching device due to pressure and seismic loads. Loads and standard or modified coef ficients for each stress summary sheet are inputted. Stresses or reactions are computed by multiplying the
, loads by the coef ficients.
AvalIabIIItv This program (proprietary) is available at the G&l of fice in Oak Brook, IilInois, on an IBM 3033N computer.
Verification This program was verified by methods described and transmitted by Reference (1), Section A.107.
Ano1IcatIon The program is, used to calculate stresses and reactions in the air lock of the containment vessel.
References Not Applicable (uses f undamental mathematical equations only)
O l
i O
V Amend. 76 A-293 March 1983
l O
A.118 BOSOR4 BOS0R4 is a computer program for stress, stability and vibration analysis of shells of revolution. The program was developed by D. Bushnell of Lockheed Missiles and Space Company (Reference 1).
The computer code is based upon the linear, elastic, thin shell theory. The structure should be axisymmetric. The program can handle various kinds of wall materials and loadings. Both mechanical and thermal loads are permitted in the analysis. In cases involving stress analysis of a shell for non-axisymmetric loading, the program finds the Fourier series for the loads, calculates the shell response in each harmonic to the load components with-that harmonic, and superposes the results for all harmonics.
The program has an option by which the stability analysis of a shell can be treated as a bifurcation buckling problem and mathematically it is treated as an eigenvalue problem. The program also handles shell vibration as an eigenvalue problem and finds mode shapes and frequencies.
B0SOR4 uses a finite-difference scheme as a numerical technique in the solution of shell problem.
Availability This program is available through CDC - Cybernet.
Verification B050R4 is recognized and widely used in industry with a sufficient history of successful use to justify its validity.
Application BOSOR4 has been used in the analysis of axisymmetrical structures.
Reference (1) Bushnell D-Stress Stability and Vibration of Complex Branched Shells of Revolution: Analysis and User's Manual for BOSOR4. NASA / Langley Research Center, Hampton, Virginia. Contract NASI-10929.
O A-294 Amend. 77 May 1983
G V
A. 119 CODES
. CODES is a Burns and Roe computer program that designs reinforced concrete wall and shell structures per the requirements of ACI 318. The program cal-culates the reinforcement requirements due to axial load and bending, torsional moment, longitudinal and transverse shear. This is done for each of 14 load combinations and the maximum reinforcement areas for each group of elements are tabulated. In addition, the program contains an option wherein various loadings are combined and converted to principal forces. The components of the principal forces in the meridional and hoop directions are combined with the bending moments for the design of shell reinforcement.
Availability CODES is developed by Burns and Roe.
Verification
- The CODES verification has been done by hand calculation and is available in internal Burns and Roe documents.
/}
'%ul Application g
CODES is used in the design of reinforced concrete wall and shell structures.
x A-295 Amend. 77 May 1983
O A.!20 EQUILIN The computer program EQUILIN is designed to calculate an equivalent linear temperature distribution from a given non-linear distribution, according to the technique described in Reference (1).
The computer program is provided with infonnation of a concrete section subjected to a non-linear temperature distribution. The program determines mean temperature of the section from the non-linear temperature. Also, an equivalent linear temperature distribution is found such that it produces the same uncracked moment about the centerline cf the section as does the non-linear temperature distribution.
Availability EQUILIN is developed by Burns and Roe and is available in the CYBER 176 Computer of CDC-CYBERNET System.
Verification EQUILIH has been verified against hand calculations. A verification problem is attached and the results are found satisfactory, i l
Application
]
l l EQUILIN has been used to determine equivalent linear temperature distributions !
! in thennal analysis of concrete structures. I l Reference j (1) ACI-349-76, Appendix A Connentary, Code Requirements for Nuclear Safety Related Concrete Structures i
l l
Amend. 77 A-296 May 1983
I EQUILIN "JERIFICATION Attached figure shows a temperature profile through a concrete section.
The following table provides a comparison between FQUILIN and hand calculations:
EQUILIN HAND CALCULATIONS Mean Temperature 216.4 216.5 Difference in Temperature Through Wall Thickness 286.4 286.5 Inside Wall Temperature 359.6 359.7 l
Outside W:sil Temperature 73.2 73.2 Equivalent linear temperature profile is shown dotted in Figure A.120-1.
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A-297 Amend. 77 May 1983
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Amend. 77 A-298 May 1983
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.V A.121 FLUSH This is a computer program for approximate three dimensional analysis of soil-structure interaction problems.
The program was developed by J. Lysmer, T. Udaka, C. F. Tsai, and H. B. Seed of the University of California, Berkeley. The program is a further development of the complex response finite element program LUSH. FLUSH includes additional features sucli as transmitting boundaries, beam elements, an approximate three dimensional capabilities, deconvolution within trie program, etc.
Availability This program is available through CDC-Cybernet.
Verification It is recognized and widely used in industry with a sufficient history of successful applications to justify its validity.
Application This program has been used for soil-structure interaction in seismic analysis.
Reference Lysmer, J.; Udaka, T.; Tsai, C.; See, H. B.; FLUSH, A Computer Program for Approximate 3-D Analysis of Soil-Structure Interaction Problems. Report No.
EERC-75-30, November 1975, University of California, Berkeley, California.
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A-299 Amend. 77 May 1983
O A.122 MODPROP This program calculates properties of a structure to be used in a " lumped mass" seismic analysis. The program calculates the masses and mass moments of inertia at specified elevations. It also calculates the member properties, such as area, shear area in two orthogonal axes (reference coordinates), area moment of inertia about these two axes, torsional rigidity, center of rigidity and centroid.
Availability MODPROP has been produced by Burns and Roe and is availavle in the CYBER 176 computer of CDC - CYBERNET.
Verification MODPROP has been verified against hand calculations. A verification problem is attached.
Application MODPROP has been used to determine mass and stiffness properties of structures for seismic analysis.
l l
l A-300 Amend. 77 l May 1983 llll
() MODPROP VERIFICATION Figure A.122-1 shows a plan and elevation of a structure with concrete walls and slabs.
Section properties of the walls between Elevations 0.0 ft and 20.0 ft. were calculated. Also, mass properties (in weight units) were calculated between i
Elevation 10.0 ft and -2.0 ft. Both MODPROP and hand calculations were used.
Results Units are feet and Kips. Moments of inertia are about centroidal axes.
M00 PROP HAND CALCULATIONS Shear Areas Ax = 224.43 224.4 i Ay = 200.43 200.4 I Total Area A = 415.68 415.6 Area Mass of Inertia Ixx = 0.37012E5 37012.0 Iyy = 0.37512E5 37512.0 Torsional Rigidity Kt = 0.7063E9 7.063 x 10 8
, Weight, W = 953.9 953.9 Weight Moment of Inertia Iwx = 0.91827E5 91,829.0
(~~'N Iwy = 0.96290E5 96,292.0 l' \~ l Iwz = 0.162192E6 162,195.0 Coordinates of Center of Rigidity, X = 14.53 14.53 Y= 10.36 10.36 Coordinates of Centroid: X= 12.74 12.735 Y= 9.96 9.957 f
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l A.123 RESPECTPLOT l This program computes response spectre, from earthquake accelerograms digitized at equal time intervals. The generated response spectra represent the maximum responses of a damped single degree of freedom system. The program is based upon the techniques described in Reference (1).
Availability RESPECTPLOT is a Burns and Roe modified version of SPECEQ/SPECUQ (Reference 1).
It is available in the CYBER 176 Computer of CDC - Cybernet System.
Verification RESPECTPLOT results were verified against STARDYNE Program of Reference (2) and the results are found satisfactory.
Application RESPECTPLOT has been used to develop spectra from the time-histories.
References (1) N. C. Nigam, P. C. Jennings, Digital Calculations of Response Spectra from Strong Motion Earthquake Forces, Earthquake Engineering Research Laboratory, California Institute of Technology, Pasadena, California, June 1968 (2) STARDYNE, Control Data Corporation, Publication No. 76079900 A-303 Amend. 77 p\_,/ May 1983
O l A.124 STRUDL STRUDL is a comprehensive structural static and dynamic analysis and design program. It stands for Structural , Design Language and conceived, developed and initially released by the Department oT Civil Engineering, Massachusetts Institute of Technology, Cambridge, Mass.
Availability The computer program is available through Georgia Institute of Technology, Atlanta, Georgia (GTSTRUDL) and McDonnell Douglas Automation Company (MCAUT0) of St. Louis, Missouri.
Verification It is recognized and widely used in the industry with sufficient history of successful applications to justify its validity.
Application It will be used for structural frame analysis of steel structures and supports.
References (1) McDonnell Douglas Automation Company, " ICES STRUDL User Manual" (2) Georgia Institute of Technology, "GTSTRUDL User Information Manual", Report No. SCEGIT-79-179, January 1979 A-304 Amend. 77 (
May 1983
A.125 'THAVSA This program calculates combined floor response spectra from spectra produced from independent seismic analyses for each of the earthquake directions (North-South, East-West, Veritical). It combines translational and rotational effects. The combination is based on equations 11,12 and 13 of Appendix B of the Peference.
. Availability THAVSA was developed by Burns and Roe and is available in the CYBER 176 Computer of CDC - CYBERNET.
Verification l
THAVSA was verified against hand calculations. The verification consisted of three steps: l
- 1) Interpolation in a semilog spectrum plot (Figure A.125-1).
- 2) Verification of the calculations with equations 11, 12, 13 of Appendix B of WARD-D-0037. Results: 1 THAVSA Hand Calculations Eq. 11 1.2630 1.2629 Eq. 12 1.3254 1.3254 Eq. 13 0.9084 0.9085
- 3) Plot Verification. '
Figure A.125-2 shows a comparison of a spectral plot produced by THAVSA and the results of' hand calculations.
( Application l
l THAVSA has been used to produce Design Acceleration Response Spectra for Equipment Specifications and for the design of structural components.
Reference WARD-D-0037 - CRBRP Seismic Design Criteria (PSAR Appendix 3.7-A).
A-305 Amend. 77 May 1983
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O A.12- 2DGENFRAME l 2DGENFRAME is a time-shared computer program which performs the bending analysis of two-dimensional frames. Frame members can be rigidly attached or pin connected. )
i Locding condition include combinations of forces at the joints, concentrated forces on the members. Distributed loads on the members, concentrated moments on the members. Distorsions of the members, and temperature changes in the members. Data can be entered interactively or from data files.
Availability 2DGENFRAME is commercially available to users through Control Data Corporation CDC KRONOS Time Sharing Computer Systems.
Verification Verification of this program was done by hand calculation and is available in internal Burns and Roe documer,ts.
Application 2DGENFRAME is used in the analysis of 2 dimensional frame structures.
O O
A-308 Amend. 77 May 1983
O AMENDMENT 77 LIST OF RESPONSES TO NRC QUESTIONS There are no new NRC Questions in Amendment 77.
l t
9 Qi G
l Ouestion CS430.30 (9.12) n.
() Identify the types of lighting that will be provided in the above tabulated vital areas. Show that lighting will be available in the event of a design basis accident, including the safe shutdown earthquake.
Resoonse: s The CRBRP Lighting System providos normal, standby and emergency lighting as described in Section 9.12 of the PSAR. The Normal Lighting Sys1em providos Illumination under sli normal plant operating conditions with power avallable f rom the Plant, Pref erred, or Reserve poner supply systems. The Standby Lighting System provides adequate illumination under all normal and emergency plant operating conditions with power available from the Plant, Preferred, Reserve or . Class IE Onsite AC Power System. Under an emergency conditica, resulting in loss of all offsite power sources, the standby lighting system will be powered from the Class 1E onsite AC power system (Emergoney Diesel Generators). Both Normal and Standby Lighting Systems utilize high pressure sodium and fluorescent light fixtures. The Emergency Lighting System provides adequate illumination at points of egress, in the Control Rom, at remote shutdown locations and at all locations required for access to safety-related equipment. The Emergency Lighting System utilizes self-contained individual eight (8) hour rated battery powered units with sealed beam lanps and seif-contained eight (8) hour rated battery powered exit signs.
All lighting fixtures in Nuclear Island buildings are seismically qualified to maintain structural integrity in accordance with IEEE Std. 344-1975. The Q,
v lighting fixtures and raceways are supported to meet Seismic Category 1 requirements as described in Sections 3.7.2 and 3.7.3 of the PSAR.
The Standby Lighting System is classified as 1E up to and including the lighting panel. The circuits to the Standby Lighting System light fixtures are also IE and are routed to maintain required separation from non-Class 1E or Class IE cables of other Divisions as described in Section 8.3.1.2 of the PSAR. However, the Iighting fixtures are non-Class 1E and as such these circuits from the 11 hting 0 panels to the lighting fixtures of the standby lighting system are considered associated 1E.
PSAR Section 9.12 has been revised to reflect the above.
O V
Amend. 74 QCS430.30-1 Dec. 1982