ML20153G403
| ML20153G403 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 03/08/1988 |
| From: | Kowalchuk J, Mazzie V, Shaffer T CONNECTICUT YANKEE ATOMIC POWER CO. |
| To: | |
| Shared Package | |
| ML20153G394 | List: |
| References | |
| 83-113, 83-113-R, 83-113-R00, NUDOCS 8809080148 | |
| Download: ML20153G403 (82) | |
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s Page Project Assignment No. Rev.#
1 of 67 83-113 0 CCNCEPItRL PRCOECT DCSCRIPTICM (leo 5.18)
PRCUECT TITLE C0tNECTICUT YANKEE MODERNIZE RPS - PHASE 2 PRCUECT PREPARED BY: h % "% b *3 [ N J vN. Kowalchuk, Project Engineer Dite' 3hD APPROVED BY: -
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V. p.. Mazzie,
. JSC Supervisor Datd
-C4Gli T. A. Shaf$dt,,I&C Manager 3hhr Date h
3R . Lef Yre, Project Manager
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Date
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'G. 1.'Bouchard, CY Unit Date ,
Superintendent i AU/WORX ORDER No. : C2/25700101 Copies to:
A. R. Roby C. J. Gladding H. H. Wong T. A. Shaffer W. J. Briggs D. P. Flynn ,
V. J. Mazzie W. H. Becker M. W. Baehr i L. J. Nadeau J. R. Terraro D. D. McCory '
T. J. Galloway M. H. Brothers F. C. Libby D. W. Mazzarella
- G. E. Cornelius G. P. VanNoordennen ,
J. Chiare11a R. E. McMullen J. A. Blaisdell D. J. Ray J. J. Roncaioli R. A. Crandall G. W. Loftus J. B. Cichocki G. H. linski J. F. Bibby W. F. Kadlec R. M. bak J. H. Ashburner W. L. Varney J. M. Black B. R. Danielson R. E. Lefebvre e
0009000140 080901 PDR ADOCK 05000213 P PNV
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Page Project Assigreent No. Rev.9 2 of 67 83-113 0 OCMCEPIUP.L PM7ECT DESCRIPTIG4 (150 5.18)
PROJECT TITLE CCH4ECTICtTF YANC PCDERNIZE RPS - PHASE 2 PROJECT I TABLE OF OCMTINTS 1.0 PRCVECT SCOPE 2.0 WCLEAR ENGINEERING AND OPERATICNS GROUP INITRFACES 2.1 Engineering 2.1.1 Project Engineer 2.1.2 NUSCO Generation I&C Engineering 2
2.1.3 NUSCO Generation Engineering Mechanics 2.1.4 NUSCO Generation Electrical Engineering 2.1.5 WSCO Generation Special studies
. 2.1.6 WSCO Generation Civil Engineering - Structures ,
2.1.7 NUSCO Generation Mechanical Engineering 2.1.8 NUSCO Generation Fire Protection Engineering 2.1.9 NUSCO Safety Analysis 2.1.10 WSCO Radiological Engineering 2.1.11 NUSCO Generation Facilities Licensing 2.1.12 WSCO Reliability Engineering 2.1.13 CYAPCO Plant Engineering 2.1.14 CYAPCO I&C Engineering 2.1.15 WSCO Quality Levices 2.2 Design 2.2.1 WSCO Generation Electrical Design 2.2.2 WSCO Generation Civil Design 2.2.3 NUSCO Generation Mechanical Design 2.3 Inglementation 2.3.1 WSCO Eetterment Construction :
2.3.2 WSCO Production Test l 2.3.3 CYAPCO Instrumentation and Controls 2.3.4 CYAPCO Operations i
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l Page Project Assignment No. Rev.) l 3 of 67 83-113 , 0 l CINCEPWhL PRC12CT DESCRIPTION (NEO 5.18)
PRCUECF TIEE CCttiECTICt71' YANKEE MODERNIZE RPS - PHASE 2 PROJECT -
2.4 Nuclear Training 2.4.1 NUSCO Operator Training 2.4.2 NUSCO Technical Training 2.4.3 NUSCO Simulator Technical Support 2.5 Project Management and Administrative Support 2.5.1 NUSCO Project Management 2.5.2 NUSCO Cost Estimating 2.5.3 NUSCO Planning and Scheduling 2.5.4 NUSCO Purchasing l 3.0 CONDITIONS AND LIMITATIONS 4.0 REGUIAICRY REQUIREMDfrS 5.0 CONCEPWAL DESI(N 5.1 Bases of Current Design 5.1.1 Reactor Coolant System Flow 5.1.2 Reactor Coolant System Pressurs 5.1.3 Primary Containment Pressure ,
5.1.4 Steam Generator Harrow Range Level l 5.1.5 Steam Generator reedwater Flow 5.1.6 Steam Generator Steam Flow 5.1.7 Reactor Trip Logic System 5.2 Method of Change )
5.2.1 Reactor Coolant Systeni Flow 5.2.2 Reactor Coolant System Pressure 5.2.3 Primary Containment Pressure l
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Page Project Assignment No. Rev.9 4 of 67 83-113 0 CWCEPITAL PROTECT DESCRIPTION (150 5.18)
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FNMECT TTILE C0tNECTICt7r YANKEE MODERNIZE RPS '.rlASE 2 FNMECT ,
t 5.2.4 Steam Generator Narrow Range Level 5.2.5 Steam Generator Feedwater Flow i 5.2.6 Steam Generator Steam Flow Reactor Trip Logic 5.2.7 1
5.2.8 Reactor Protection System Testing Capability 5.2.9 Power Dependent Insertion Limit 5.3 Design Inputs 1 6.0 CWCEPIUAL DESIM NALRDCH4 t
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Page Project Assignment No. Rev.8 5 of 67 83-113 0 cmCEPItRL PROECT DESCRIPTIN (NEO 5.18)
PacuEcr TITLs CCtNECTICJr YANKEE MODERNIZE RPS - PHASE 2 PROJECT
- 1.0 PRO 7ECT SCOPE his project continues the Reactor Protection System (RPS) Modernization effort by replacing another portion of the original RPS equipment with current state-of-the-art equivalents. h e scope of this priject includes the following:
- 1) A continuation of the Phase 1 "front-end" work which includes the replacement of the sensors, transmitters, and Main Control Board mounted equipment which make up the following indication and/or trip circuitry:
a) Reactor Coolant System Flow b) Reactor Coolant System Pressure c) Primary Containment Pressure d) Steam Generator Narrow Range Levil (transmitter replacements only) e) Steam Generator Feedwater Flow (transmitter replacements only) f) Steam Generator Steam Flcw (transmitter replacements only)
- 2) the replacement of the existing Reactor Trip relay logic system with a solid state logic system and suitable field interface. Wis system will include on line testina capability and the ability to defeat and bypass instrumentation loops.
- 3) Addition of Power Dependent Insertion Limit (PDIL) circuitry to the Rod Control System.
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Page Project Assignment No. Rev.9 6 of 67 83-113 0 C[NCEP11RL PETECT DESCRIPTION (Neo 5.18)
PM7ECT TI"LE CCtMECTICUT YANKEE MODERNIZE RPS - PHASE 2 PROJECT 2.0 NUCLEAR ENGINEERING Ate OPERATIONS GROUP INIERFACES 2.1 Engineering l
2.1.1 Proittet Engin&er a) Prepare and revise the Froject Assignment and Project Description.
b) Prepare and revise Project Equipment Specifications.
1 c) Interface with vendors to provide Supplemental design criteria.
d) Review and approve preliminary and final drawings (functional, logic, schematic, wiring, P&ID, and equipment installation).
j e) Coordinate the preparation, review, and approval of the -
Project Materials Lists.
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i f) Coordinate the preparation, review, and approval of Purchase I I
j Requisitions.
g) Coordinate the preparation, review, and approval of Project l
l Installation Specifications. l l
h) Prepare the PDCR and coordinate the review, approval, and submission to the CYAPCO Engineering Supervisor.
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r Page Project Assignment No. Rev.9 7 of 67 83-113 0 GNCEP1tRL PFa7ECT DESCRIPTION (NEO 5.18) l PRCVECT TIT 12 C0tNECTICUT YANKEE PCDERNIZE RPS - PHASE 2 PROJECT ,
i) Initiate appropriate Technical Specification ChangJ Requests.
j) Joordinate the preparation, review, and approval of Safety Evaluations.
k) Review and approve vendor test procedures and witness the racility Acceptance Test.
- 1) Provide outage support for project implementation.
m) Review and approve as-built drawings and transmit to design.
n) Prepare Turnover Documentation.
o) Prepare documentation in support of PDCR close-out (rSAR change, MEPL, Reconnended Spare Parts, and PfflS updates).
p) Perform Project Close-out. l I
2.1.2 NUSCO Generation IEC Engineering i a) Supply detailed instrumentation and control design 1 requirements, b) Ensure that the instrumentation and control design is in (
accordance with appropriate codes and standards. ]
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- Page Project Assignment No. Rev.9 8 of 67 83-113 0 CONCEP!1RL PIG 7ECT DESCRIPTIG4 (MO 5.18) .
Fin 7ECT TITIJC CCtNECTICUT YANKEE MODERNIZE RPS - PHASE 2 PIO7ECT ,
c) Perform independent review and approval of design doeurantation. ,
d) Coordinate all major ItC modifications in the 1989 outage to reduce duplication of effort and define cleat boundaries between the various Project Assignments.
2.1.3 NUSCO Generation Engineering Mechanics a) Perform equipnent seismic qualification review and approval, b) Specify seismic design inputs.
c) Specify the seismic design requirements for the installation of two three-bay instrumentation cabinets in the Main Control Room.
d) Specify the seismic design requirements for the installation of cable tray and conduit in the Main control Room ceiling, e) Specify the seismic design requirements for the installation of instrument tubing in Containment.
f) Ensure that the seismic design is in accordance with
. appropriate codes and standards.
g) Review and comment on the Project Installation Specifications.
Page Project Assignment No. Rev.9 9 of 67 83-113 0 CI24CEPIUhL PIWECT DESCRIPTIG4 (NEO 5.18)
FI W ECT TITLE CCtEECTICUT YANKEE MODERNIZE RPS - PHASE 2 FIW ECT ,
h) Review and approve Civil and Mechanical drawings.
i) Prepare PDCR input.
j) Perform independent design verification reviews.
k) Review and approve Civil and Mechanical as-built drawings.
2.1.4 NUSCO Generation Electrical' Engineering a) Specify the electrical design inputs, b) Determine power requirements and provide sources.
c) Perform loading calculations.
d) Provide technical input to Electrical Design.
l e) Review and approve cable routing and electrical drawings. j f) L:sure that the electrical design is in accordance with appropriate codes and standards, g) Review and comment on the Project Installation Specifications, h) Prepare PDCR input. l l
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Page Project Assignment No. Rev.9 10 of 67 83-113 0 CCNCEP11.N. MWECT DESCRIPTICM (NED 5.18)
FNMECT TI'tu j COMECTICt7f YANKEE MODElWIZE RPS - PHASE 2 PNMECT .
- 1) Prepare the Electrical Purchase Requisitions.
j) Prepare the Electrical Safety Evaluation.
k) Perform independent design verification reviews.
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- 1) Review and approve electrical as-built drawings.
2.1.5 NL,5CO Generation Electrical Engineering - Special Studies
, a) Specify EEQ design inputs.
b) Ensure that the EEQ design is in accordance with the l appropriate codes and standards, a
! c) Review and consnent on the Project Installation Specifications.
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- l l d) Update EEQ documentation.
e) Prepare PDCR input.
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! f) Perform independent design verification reviews, l
g) Review and approve EEQ Walkdown Checklists.
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Page Project Assignment No. Rev.9 11 of 67 83-113 0 CONCEPIURL PRCL7ECT DESCRIPTION (1e0 5.18)
Fin 7ECT TITIA CCNECTICUr YANKEE MCOERNIZE RPS - PHASE 2 FIQ7ECF 2.1.6 NUSCO Generation Civil Engineering - St'ructures a) Specify the civil design inputs. ,
b) Ensure that the civil design is in accordance with appropriate codes and standards.
c) Review and comment on the Project Installation Specifications.
d) Provide technical input to Civil Design.
e) Review and approve equipnent mounting drawings.
f) Irepare PDCR input.
g) Prepare the Civil Safety Evaluation.
l h) Perform independent design verification reviews. ]
2.1.7 NJSCO Generation Mechanical Engineering l
a) Specify the mechanical design inputs, b) Ensure that the mechanical design is in accordance with appropriate codes and standards, c) Specify r.ny welding or material requirements. l l
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Page Project Assignment No. Rev.9 12 of 67 E3-113 0 OCNCEP1tRL PROJECT DESCRIPTIG4 (NEO 5.18)
PROJECT TI'IU CCmECTICitr DNKEE MODERNIZE RPS - PHASE 2 Pna7ECT .
d) Provide technical input to Mechanical Design.
e) Performindependtntdesignverificatkonreviews.
f) Review and consnent an the Project Installation Specifications.
2.1.8 NUSCO Generation Fire Protection Engineering a) Specify design inputs. -
b) Perform Fire Protection reviews.
c) Ensure that the design is in accordance with appropriate codes and standards. ,
d) Review and coasnent on the Project Installation Specification.
e) Prepare PDCR input.
f) Perform independent design verification reviews.
2.1.9 NUSCO Safety Analysis a) Prepare PDCR input.
b) Prepare the Integrated Safety Evaluation.
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Page Project Assignment No. Rev.9 13 of 67 83-113 0 00NCEP11RL PROJECT DESCRIPTICH (MO 5.18)
PRDtC"! TITt2 CCMECTICtfr YANKEE MODERNIZE RPS - PHASE 2 PIQ7ECT
- 2.1.10 NUSCO Radiolegical Engineering a) Perform AIARA Design Review. ,
b) Perform c man-rem savings analysis tu justify the relocation i of PT-403 and Pr-404 in the outer annulus of containment.
c) Suhait AIARA recomendat'.ons.
2.1.11 NUSCO Generation racilities Licensing a) Identify regulatory requirements. ,
b) Review and coment on the Project Description and Design Inputs.
i j c) Provide liaison with all Regulatory Agencies as required.
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d) Coordinate agency review and approval. [
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e) Coordinate the approval of any Technical Specification Change Requests.
2.1.12 NUSCO Reliability Engineering a) Perform a Reliability Analysis c4 the plant modification in ,
accordance with IEEE Standard 577-1976, Section 4.
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Page Project Assignment No. Rev.9 14 of 67 83-113 0 ,
. CCNCEPmhL PIGECT DESCRIPTIbi (ISO 5.18)
PROJECT TI'IM CCm ECTICitr YANKEE MCCERNIZE RPS - PHASE 2 PIG ECT ,
2.1.13 CYAPCO Plant Engineering a) Coordinate the plant's review of design documents and obtain a ,
consensus of plant opinions for the detailed design.
b) Specify, review, and approve the project's Design Inputs.
c) Coordinate the preparation and performance of test procedures I to complete pr'eoperational testing.
d) Review and approve the PDCR and coordinate the PORC approval.
e) Review and coament on vendor drawings, f) Review material and installation specifications, g) Review and comment on the vendor test procedures and witness the racility Acceptance Test, h) Prepare Purchase Requisitions for the purchase of consumable items (wire lugs, spil:es, heat shrink tubing, etc...).
- 1) Provide outage support for project inglementation (coordinate work orders, testing, significant DCN processing, NCR's, etc...).
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i Page Project Assignment No. Rev.9 15 of 67 83-113 0 00NCEPIUM. PIGECT DESCRIPTIN (NEO 5.18)
PRDET'YfTIE CCtNEcrICttr YANKEE MODERNIZE RPS - PHASE 2 PI W ECT ,
1 j) Review the project construction schedule and coordinate plant activities. ,
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k) Ensure that, the modification is couplete, that all preopera-i tional requirements are couplete, and that all open items are l
identified and tracked.
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- 1) Notify the Shift Supervisor of the Engineering Release for j l
operation. .
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j m) Coordinate the collection of all as-built drawings. :
i 2.1.14 CYAPCO I&C Engineering a) Specify, review, and approve the project's Design Inputs. j b) Coordinate the preparatio.) and performance of test proceoures to complete preoperational testing.
l c) Review and comment on vendor drawings.
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. d) Review material and installation specifications and the PDCR.
1 1 e) Review and coeuwnts on the vendor test procedures and witness
- the racility Acceptance Test.
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s Page Project Assignment No. Rev.4 16 of 67 83-113 0 CDOCEP1kRL FROJECT DESCRIFFICH (Is0 5.18)
Fin 7ECT TI112 CCtNECTICtTT YANKEE MODERNIZE RPS - PHASE 2 FIE7ECT .
f) Coordinate the Software configuration of Spec 200 micro cards.
g) Provide outage support for project implementation (coordinate testing and troubleshooting).
h) Review the project construction schedule and coordinate plant ,
activities.
2,1,1$ NUSCO Quality Services a) Review and approve material and installation specifications.
b) Review and approve procurement documentation.
c) Perform supplier audits / inspections. ;
d) Perform material receipt inspections, e) Perform field surveillances. ;
f) Coordinate NCR issuance, tracking, and closure. :
I g) Perform AWO reviews and inspections.
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Page Project Assignment No. Rev.9 17 of 67 83-113 0 00NCEP1tRL PIGECT DESCRIPTIG6 (NED 5.18)
PfGECF TITIE.
CCtNECTICt7r YANKEE MODERNIZE RPS - PHASE 2 PIGECT .
2.2 Design 2.2.1 NUSCO Generation Electrical Desian a) Provide the design and drafting for the preparation and revision of schematic, wiring, panel layout, and equipment arrangement drawings, b) Provide the design and drafting for t'he preparatien and revision of loop and logic diagrams.
c) Prepare and issue cable and conduit schedules, duct and trench plans, and raceway plans.
d) Prepare and issue the Electrical Materials List.
e) Review and connent on vendor drawings.
f) Incorporate vendor drawings into the NUSCO system.
g) Prepare and issue the electrical as-built drawings.
h) Provide outage support to facilitate the installation of cable and electrical equipment.
2.2.2 NUSCO Generation Civil Design
, a) Prepare Civil design calculations.
Page Project Assignment No. Rev.9 18 of 67 83-113 0 00NCEP1UhL PIGHCf DESCRIPTIG4 (950 5.18) l PICHCr TI'In Cott1ECTICUT YANKEE MODERNIZE RPS - PHASE 2 FIO7ECT ,
b) Provide the design and drafting for the preparation and revision of equipment, conduit, cable tray, and tubing support drawings. ,
c) Prepare and issue the Civil Structural Material List or provide material information on the Civil Drawings.
l d) Prepare and issue the Civil as-built drawings.
e) Provide outage support to facilitate *the installation of equipnent and instrument tubing.
2.2.3 NUSCO Generation Mechanical Design a) Prepare Mechanical design calculations.
b) Provide the design and drafting for the preparation and revision of Mechanical P&ID, and isometric drawings.
c) Prepare and issue the Mechanical Material List.
d) Prepare and issue the Mechanical as-built drawings.
e) Provide outage support to facilitate the installation of instmment tubing.
2.3 Inglementation e
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e Page Project Assignment No. Rev.9 19 of 67 83-113 0 cmczPNM. PlWECT DESOLIPTIM (ISO 5.18)
FFDJECT TI'lu CCWECTICtfr YANKEE MODERNIZE RPS - PHASE 2 PIG ECT .
2.3.1 NUSCO Betterment Construction a) Review and coacnent on the project installation specifications.
b) Transport and install the two new Foxboro cabinets in the Main Control Room.
c) Install cable trays and conduit in the control Room between Foxboro Cabinets and Main Control Board.
d) Install new transmitter racks in containment.
e) Mount new transmitters.
f) Install conduit and supports and. pull new cables in the containment outer annulus.
g) Install instrument tubing and supports.
h) Prepare Material List for field materials.
- 1) Purchase field materials.
j) Prepare as-built drawings and transmit them to the Flent Engineer, k) Provide input for the' preparation and updating of the project schedule.
Page Project Assignment No. Rev.9 20 of 67 83-113 0 CONCEPRRL Ppn7ECT DESCRIPTIG( ( W 5.18)
Pin 7scr tim CotNECTICur YANKEE MODERNIZE RPS - PHASE 2 Pin 7ECT .
2.3.2 NUSCO Production Test a) Review and connent a the project installation specifications.
b) Pull and terminate all field side cabling in the Main Control Room.
c) Perform modifications to the Main Control Board (conponent installations, internal board wiring).
d) Prepare as-built drawings and transmit them to the Plant Engineer, e) verify the continuity and arrangement of all "field side" wiring.
f) Provide input for the preparation and updating of the project schedule.
2.3.3 CYAPCO Instrumentation and Controls a) Review and ceanent on the project installation specifications.
b) Perform all internal wiring modifications in the Foxboro cabinets.
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Project Assignment No. Rev.0 Page 21 of 67 83-113 0 M PRCOBCT DE8':RIPTICBI (NED 5.18)
PR[hTCCT TI'ILE CcmEcrICUT YANKEE Pt0DERNIZE RPS - PHASE 2 FACh7BCT .
i c) Terminate cabling at all transmitters and containment penetrations. ,
d) Verify the continuity and arrangement of all "cabinet side" wiring.
l e) Review and connent on preoperational testing procedur2s.
'f) Perform all I&C preoperational testing procedures. l t
j g) Support the integrated operability testing, i h) Prepare as-built drawings and transmit them to the Plant .
! Engineer, ,
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i i) Provide input for the preparation and updating of the project I schedule.
l r 2.3.4 CYAPCO Operations a) Review and comment on the detailed design. l
- b) Revise and review the affected operations procedures.
! c) Gupport preoperational'and integrated testing. ;
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1 Page Project Assignment No. Rev.9 22 of 67 83-113 0
'usCurnAL PIG 7ECT LESCRIPTION (Nuo 5.18)
Pin 7ECT TI'!U CCtMECTICtTF YANKEE MODERNIZE RPS - PHASE 2 FIG 7BCT .
2.4 Nuclear Training 2.4.1 NUSCO Operator Training a) Modify initial and requalification training programs to incorporate project design changes.
! b) Provide classroca and/or hands-on training for the plant operators.
2.4.2 NUSCO Technical Training a) Prepare training programs for the operation and testing of added plant equipment.
b) Provide classroom and/or hands-on training for the plant I engineers and technicians.
2.4.3 NUSCO Sism11ator Technical Support a) Update the simulator to incorporate applicable portions of this design change.
2.5 Project Managewnt and Administrative Support 2.5.1 NLE O Project Management
. a) Prepare and issue PERC packages.
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Page Project Assignment No. Rev.9 23 of 67 83-113 0 030CEFREL FIWsCT DESCRIFFI(34 (teo 5.18)
PIWucT num CotNECTICttr YANKEE MODERNIZE RPS - PHASE 2 j FI W BCT .
b) Review equipment ard construction bids.
l c) Resolve conflicts between various projects and organisations. !
i d) Provide a project plan to ensure coordination of resources.
2.5.2 NUSCO Cost Engineering a) Prepare and revise project material and' labor cost estimates.
b) Monitor, evaluate and report on cost variances. ;
2.5.3 NUSCO Planning and Scheduling
. I a) Prepare and revise project engineering, design', and i I
installation schedules. ;
b) Provide input to and coordination of the outage schedule, i I
c) Monitor, evaluate, and report on scheduling variances and critical path items. f I
l 2.5.4 NUSCO Purchasing l
i a) Prepare and issue Purchase Orders for materials and labor. !
i b) Provide expediting of materials as required.
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Page Project Assignment No. Rev.9 l 24 of 67 83-113 0 CINCEP!UM. FIG 7ECT DESCRIFFIN (l#10 5.18)
Pauser TI m CCNECTICUT YANKEE MODERNIZE RPS - PHASE 2 c) Report the expected delivery dates of equipment if other than the "Date Required" on the Purchase Requisitions.
3.0 CreeITIONS Ne LIMIMTIWS 21s project shall be installed, tested, and declared operational prior to the coupletion of the 1989 CY Refueling Outage. his project shall re-I quest the performance of pre-outage work, to the extent allowable by
- CYAPCO Engineering and Operations, to ensure that the plant modification is coupleted within the planned 50 day duration of the outage. Precau-tions shall be taken during the equipment replacements to ensure that mininum channel operability requirements are adhered to. 2e Plant Opera- ,
tions sign-off on the AHo's will control these requirements.
. Replacement of the primary containment pressure switches is required by station management to resolve maintenance problems with these switches. I W is project assignment scope includes this work to resolve station !
Controlled Routing 87-1310 & 87-1311 and Plant Incident Report 87-120. l Connecticut Yankee Management requires t.his work to be completed by start-up from the 1989 refueling outage.
4.0 cmnacRY REQUInneNIS I
W is plant design change shall be performed in accordance with 10 CFR 50.59. mis modi'ication is not expected to involve an unreviewed safety l question, however, preliminary commission notification may be prudent.
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Page Project Assignment No. Rev.9 25 of 67 83-113 0 GMCEF1tRL PRa7ECT DESCRIPTIM (NEO 5.18)
FROJECT TI'Iu CCtNECTICt7f YANKEE MCDERNIZE RPS - PHASE 2 Pna7ECT 2e following ISAP Topics are associated with the change:
ISAP Topic No. 1.30 - Reactor Protection System Isolation ISAP Topic No. 2.04 - Modernize Reactor Protection and Control Systems 5.0 CNCEPItRL DESIGN 5.1 sases of current Design 5.1.1 Reactor Coolant System Flow nere are two controlling permissive circuits for reactor trip due to reactor coolant less-of-flow. Wey are P7 (10% power) and P8 (74% power). No trip due to loss of flow will occur below 10%
power (P7). When in four loop operation between 10% power and 74%
power, two loops sust signal loss-of-flow to the RPS to initiate a reactor trip. During three loop operation, the idled / isolated loop's channel is in the tripped position. Above 74% power (P8),
any one loop indicating loss-of-flew will initiate a reactor trip.
Trip signals for each loop may cese from any of three sources; opening of the reactor coolant purip breaker, low voltage on the generator bus supplying the pump (4160 volts), and low pressure differential measured between the inlet and outlet plenums of the reactor coolant side of the steam generator. Iow voltage on a generator bus affects two pumps and will cause a reactor trip 4
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Page Project Assignment No. Rev.6 26 of 67 83-113 0 ,
CCNCEPfX PRC17ECT DESCRIPTION (NEO 5.18)
PROJECF TIBA CONNECTICtJT YANKEE PCDERNIZE RPS - PHASE 2 PRCA7ECT .
above 10% pow r (P7) level. Trip signals from either the pump l breaker or flow measurement indicating a one loop loss of flow ,
will cause a reactor trip above 74% power (PS). In operation [
below 74% power (P8), a coincidence of two loss of flow signals (from different loops) is necessary for a reactor trip.
Presently, there is a single Reactor Coolant system flow trans-mitter per loop. his Plant Design Change will replace the existing transmitters and add two additional transmitters per loop to provide redundancy and enhance plant reliability by preventing f inadvertent Reactor Trips.
5.1.2 Reactor Coolant System Pressure l I
RHR Interlocks I
Remotely-operated double valving is provided to isolate the !
I residual heat removal system inlet and outlet piping from the reactor coolant system. An electrical interlock between the l reactor coolant system loop 4 pressure channels and the first I (inboard) and second (outboard) set of valves prevents the valves l from being opened when reactor coolant system pressure exceeds 400 !
psig. his interlock protects the residual heat removal system piping which has a design pressure of 500 psig. Key control ,
switches provide further administrative control against misopera-tion of the second (outboard) set of valves. Rose valves are ,
also interlocked and require a pressure permissive signal, RCS pressure less than RHR. design pressure, before they can be opened.
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Page Project Assignment No. Rev.9 27 of 67 83-113 0 (23ectrit.mL PKk7sCT DESCRIrrION (NEO 5.18)
PMMECT TITU CCtNECTICtTI' YANKEE MODERNIZE RPS - PHASE 2 N .
L10PS (Low Tepperature Overpressurization System) Interlocks Remotely-operated double valving is provided to isolate the low pressure relief valves. An electrical interlock between the reactor coolant system loop 4 pressure channels ard the low pressure overpressure relief motor operated isolation valves (FR-MOV-596, -597, -598, and -599) prevents these valves from being opened when reactor coolant system pressure exceeds 345 psig. Additionally, an electrical interlock provides an alarm when either PR-MOV-596 or PR-MOV-597 are open and reactor coolant system pressure exceeds 370 psig. The low pressure relief valve setpoint is 380 psig.
Presently, there are two reactor coolant system pressure transmitters (0-3000 psig) on loop 4. @is Plant Design Change will remove the existing transmitters and install a wide range RCS pressure transmitter (0-3000 psig)' and a narrow range RCS pressure transmitter (0-600 psig) on tnth loop 1 and loop 4. his will provide further redundancy and higher setpoint accuracy for NOL and UIors valve interlocks (reference LER 86-018-00).
5.1.3 Primary Containment Pressure
%e containment isolation is initiated .:y high containment pressure (5 psig) from a two-out-of-three signal frca one of two trains of mercoids and/or low pressurizer pressure (1700 psig) 9 4
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Page Project Assignment No. Rev.9 28 of 67 43-113 0 ct3CuritmL FNNuer DESCRIFFI(30 (ISD 5.18)
P h una CCNBCTICL7f WWEEE MODERNIZE RPS - PHASE 2 PNMKT .
trip setpoints, anergency core cooling system actuation also isolates the containment. Se low pressuriser pressure signal is-a safety injection signal. Se containment isolation system actuation setpoints and two of three initiation logics were designed to ensure system reliability by preventing inadvertent system operation. Both containment isolation system trip setpoints indicate a possible leak in the Ecs and/or a pressure increase in containment.
Upon initiation of the containment isolation cignal all non-essential system penetrations are isolated from the containment.
Some of the penetrations are isolated by norns11y closed manual valves, fail-closed (on loss of instrument air pressure or elec-tric power) air-operated valws, and fail-as-is motor operated valws, while other valws are automatically closed by the containment isolation signal within 60 seconds of its initiation.
Presently, there are six mercoid containment pressure switches (three per train). mis Plant Design change will remove these pressure switches and replace them with category 1, class 1E equivalent pressure switches with higher accuracy and repeata-bility. If suitable replacement switches are not available, the existing pressure switches will be replaced with four p: essure l
transmitters. 1 5.1.4 steen Generator Narrow Ranee Lew 1 Se existing IIagan Steen Generator Harrow Range tavel transmitters e
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Page Project Assignment No. Rev.9 29 of 67 83-113 0 cascur!UhL Pga7ECF DESCRIFYIG4 (NED 5.18)
Fin 7ECT Tr M Com ECTICL7f YANKEE MODERNIZE RPS - PHASE 2 FIQ7ECT -
(LT-1301-1, -2, -3, and -4) accept an input range of 17.6 - 82.S" H3 0. he corresponding output is 1-9 VDC. Se existing accuracy is 1.0% of span. ,
I
} Wis Plant Design Change will remove these flow transmitters and !
l replace them with Category 1, class it roxboro equivalents. We i new transmitters will have a 4-20 mA output, consistent with l current industry standards. A Category 1, Class 1E unit will be
, installed to convert this output to 1-9 VDC, to be corpatible with existing Hagan rack equipment.
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l 5.1.5 steam Generator reedwater riew J
j 2e existing Hagan Ring Salance Flow transmitters (PT-1301-1Ar j -2A, -3A, and ~4A) accept an input range of 0 - 533.56" H3 0. Se ,
corresponding output is 1-9 VDC. Se existing accuracy is 1.0% of ,
span. l 4
i mis Plant Design Change will remove these flow transmitters and replace them with Category 1, Class 1E roxboro equivalents. Se I new transmitters will have a 4-20 mA output, consistent with current industry standards. A Category 1, Class 1E unit will be ;
installed to convert this output to 1-9 VDC, te be compatible with !
existing Magan rack equipment.
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1 Fage Project Assignment No. Rev.9 30 of 67 83-113 0 -
GMCErnmL PIW3CT DEerRIPPI(36 ( m 5.18) !
l t 1
N CCtMEC.TICUT YANREE j i
l I I PCDERNIZE RFS - PtmsE 2 i I
( FI m acT .
[
5.1.6 steam Generator steam Flow ;
he IM sting Negan Ring Balance Flow transmitters (PT-1201-1, -2, f
-3, and -4) accept an input range of 0 - 573.65" g o. Se correspond!ng output in 1-9 VDC. We existing accuracy is 1.0% of l span.
l 21s Plant Design change will remove these flow transmitters and f replace them with category 1, class 1E roxboro equivalents. Se !
new transmitters will have a 4-20 mA output, consistent with . I current industry standards. A Category 1, class 1E unit will be installed to convert this output to 1-9 VDC, to be compatible with l.
existing Magan rack equipent. '
5.1.7 Reactor Trip tonic system j
Se reactor trip system automatically keeps the reactor oportting .
I within a safe region by shutting down the reactor whenever the j limits of the region are approached. Se safe operating region is defined by several considerations such as mechanical / hydraulic limita:8m on equipment and heat transfer phenomena. Berefore, j the reactor trip system keeps surveillance on procoas variables which are directly related to equipmen e .:hanical limitations f j
such as pressure, pressuriser water level and cleo on variables '
which directly aff'ict the heat transfer capability of the reactor ,
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Page Project Assignment No. Rev.5 31 of 67 83-113 0 CWCEPItRL PROJECT TESCPIPTIQi (750 5.18)
P107ECT TITLE COtWECTICtTF YANKEE MODERNIZE RPS - PHASE 2 PROJECT (e.g., flow and reactor coolant temperatures). Still other par.-
meters utilized in the reactor trip system are calculated from various process variables. In any event, whenever certain direct process or calculate.< variables exceed a setpoint, the reactor will be shut down in order to protect against either gross damage to fuel cladding or loss of system integrity which could lead to release of radioactive fission products into the containment.
We reactor trip system consists of sensors with analog circuitry and redundant channels. W e sensors monitor various plant para-meters and consist of redundant logic trains. We logic trains receive inputs from the analog protection channels to complete the logic necessary to automatically open the reactor trip breakers.
Power to the control rod dri'le mechanisms is supplied by two motor generator sets operatinq from two separate 480V, three-phase buses. We output of t3e motor generator sets is 125V ac three- i phase. W e output is rectified to become 125v de which is sent to the operating coils of the control rod drive mechanisms through two reactor trip circuit breakers connected is series with the coils. Before any rod motion can take place, both breakers must be closed. Each breaker is equipped with a shunt trip coil and an undervoltage tripping device. Each shunt trip coil is so con-nected that it will be energized through separate contacts from all reactor trip devices. In addition, another set of reactor, I trip device contacts is so arranged that when reactor trip is called for they interrupt the circuit to the two undervoltage l
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Page Project Assignment No. Rev.4 32 of 67 83-113 0 cmCEPItRL PROJECT DESCRIPTIM (Neo 5.18)
PIO7ECT TITLE CC W ECTICUT YANKEE MODERNIZE RPS - PHASE 2 Pin 7ECT .
coils. hus, for a single reactor trip function there will tse two separate shunt trip signals and a single undervoltage signal to the series combination of two reactor trip breakers, cau. sing all rod mechanism coils to deenergize, and the control rods to fall, by gravity, into the core. ne rods cannot be withdrawn until the trip breakers are manually resot. he trip brec.kers are admin-istratively controlled by procedure and cannot be reset until the abnormal condition which initiated the trip is corrected.
Presently, the Reactor Trip logic is performed by a combination of parallel and series auxiliary relay contacts. his plant Design Change will remove the existing relay logic and replace it with two trains of solid-state logic. The following Reactor Trip Logic functions will be performed.
l a) High Pressurizer Level Trip l
l Logics 2/3 Compensated Press. Level Channels !
Setpoints 186%
Active: 0-100% Power runction: Prevent discharge of water through safety valves b) High Pressurizer Pressure Trip Logic: 2/3 Pressurizer Pressure Channels Setpoint: $2300 psig Active 0-100% Power Function: Protects System Integrity on '.
. -Rod Withdrawal Incident l l
-Loss of Load Incident i
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Page ' Project Assignment No. Rev.9 33 of 67 83-113 0 GNCEPTUhL PROMICr DESCRIPTION (NEO 5.18)
PIW ECT TITLE CCt@iECTICUT YANKEE MODERNIZE RPS - PHASE 2 PIG ECT c) High Power (Flux) Trip (Overpower Trip)
Logics 2/4 Power Range Drawers Setpoint 3 Positions: 25%, 74% and 109%
Active 0-100% Power Function Protects Against Reactivity Accidents Due to:
-Rod Withdrawal Incident
-Inactive Loop Start-up
-Boron Dilution
-Excess Feed Water
' -Large Load Increase
-Rod Ejection
-Steam Line Break d) High Start-Up Rate Trip Logic: 2/4 Intermediate Drawer ,
Setpoint: $5 DPM Active Trip bypassed above 10% power l Function: Protects Against
-Uncontrolled Rod Withdrawal from subcritical
-Ejected Rod (Subcritical)
-Boron dilution l i
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Page Project Assignment No. Rev.#
34 of 67 83-113 0 CONCEPMAL PRCLTECT DESCRIPTION (NEO 5.18)
Paamcr TITLE CCNNECTILVI YANKEE MODERNIZE RPS - PHASE 2 PROHX:T .
e) High Steam Flow Trip Logic 2/4 Bistables from Elbow Tap AP's Setpoint: 110% of Full Flow (1825 Mwt)
Active 0-100% Power Functio'..: Protects Against
-Steam Line Break Also Trips the Steam Line Isolation Valves f) Steam Flow - Feed Flow Mismatch Trip Logic: 1/4 Coincident with SG, NR Low Level Setpoints Steam Flow 20% Greater tham Full Power Feedwater Flow / Steam Gen., NR Level <10%
Active: 0-100% Power Function: Protection for Loss of Teed l
g) Core Cooling Actuation Trip I Logics 2/3 PZR Pressure Setpoints 1700 psig Active 0-100% Power Function: Reactor Trip in event of a LOCA, excessive cooldown
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Page Project Assignment No. Rev.8 35 of 67 83-113 0 OCNCEmmRL pin 7ECT DESCRIPTION (Neo 5.18)
PROJECT TITLE CONNECTICtJr YANKEE MODERNIZE RPS - PHASE 2 P!nTECT h) 4160 Volt Bus Ulyler Voltage (1A & IB) Trip Loaic: 1/2 of Busses 1A & 1B (Notes 2 pumps off each bus)
Setpoint: Undervoltage Active 10% - 100% Power Function: Protection for Loss of Primary F.1.ow i) Steam Line Trip Valves Closed Trip Logic: 1/4 limit switches ,
Setpoint: Valve Closed Active 10% - 100% Power Function: Back up for steam line break and protects in the event of a isolation valve closure for any reason.
j) Variable Low Pressure Reactor Trip Logics 2/4 LPSC Signals S6tpoint: Variable with temp. & power (6T) min 1700 PSI Active 10% - 100%
Function: Protects Against DNB for
-over power - over temp incident
-Rod withdrawal incident (Hi power)
Fornula: 17.4 (Tavg + 1.17 AT) - 8850 a
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Page . Project Assignment No. Rev.9 36 of 67 83-113 0 CWCEPIURL PROJECT DESCRIPTIN (NEO 5.18)
Pin 7Ecr TITLs CCWECTICUT YANKEE MODERNIZE RPS - PHASE 2 PIG 7ECr
- k) Turbine stop valves closed Trip . ;
Logic: 2/2 Limit Switches Setpoint: Valve closed Active: 10% - 100% Power Function: Back up auto stop oil trip and protection for
-Loss of load incident 1_) Low Auto Stop Oil Pressure Logic: 2/3 Mercoid Pressure Switches Setpoint <45 PSIG Active: 10% - 100% Power Function: Protects against
-Loss of load incident m) Low Reactor Coolant Flow Trip Logic 10% (Power <74%, 2/4 SG D/P Xmitters or 1/4 SG Differential Pressure Xmitter coincident with
- 1 RCP breaker open Power >74% 1/4 SG D/P Xmitters Setpoint: 90% of rull riow Actives 10% - 100% Load Function: Protects Against
-Loss of Primary riow Incident e
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Page Project Assignment No. Rev.#
37 of 67 83-113 0 1
CONCEPItRL PfG7ECT DESCRIPTION (Neo 5.18) 1 PRC.TECT TITLE CCtNECTIQ7I YANKEE MODERNIZE RPS - PHASE 2 i
PamEcr .
n) RCP Breaker Open Trip Logic: 10%Qower<74%,2/4RCPBreakersOpenor1/4SG Differential Pressure Xmitter coincident with 1 RCP breaker open Power >74% 1/4 RCP Brekkers Open Setpoints Breaker Open Active 10% - 100% Load Function: Anticipates and Protects Against
-Loss of Primary Flos Incident , !
o) Manual Actuation i
Active: 0-100% Power Function: Provide Manual Trip Capability 5.2 Method of Change 5.2.1 Reactor Coolant System Flow i
a) Remove the existing four (4) Foxboro,10-50 ma flow transmitters FT-401, FT-402, Fr-403, and Tr-404, b) Seismically install a transmitter rack in the containment l outer annulus between columns 15 and 16 along the outer l containment wall.(El l'-6").
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Page Project Assignment No. Rev.9 38 of 67 83-113 0 ONCEPRRL PRCUECT DESCRIPTIO( (NEO 5.18)
P!nJECT TITLE CCMECTIctrr YANKEE MODERNIZE RPS - PHASE 2 PRCUECT
- c) Seismically install a transmitter rack in the containment outer annulus between colums 5 and 6 along the outer containment wall (El l'-6").
d) Seismically relocate the instrument tubing for FT-401 and FT-402 to the new transmitter rack between colums 15 and 16.
e) Seismically relocate the instrument tubing for FT-403 and ^
FT-404 to the new transmitter rack between columns 5 and 6.
f) Seismically mount six (6) roxboro, 4-20 ma flow transmitters FI-401-1, FI-401-3, Fr-401-4, FT-402-1, rr-402-3, and FI-402-4 on the new transmitter rack between columns 15 and 16.
g) Seismically mount six (6) roxborc, 4-20 ma flow transmitters l Pr-403-1, FT-403-3, FT-403-4, Fr-404-1, FT-404-3, and Pr-404-4 l en the new transmitter rack between colums 5 and 6.
h) Connect the relocated tubing from rr-401 to Pr-401-1, Pr-401-3, and rr-401-4. Install independent valving to each transmitter.
i) Connect the relocated tubing from Fr-402 to rr-402-1, Fr-402-3 and rr-402-4. Install independent valving to each trans- I mitter.
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Page Project Assignment No. Rev.9 39 of 67 83-113 0 Cl20CFFIUAL Pra7DCr DESCRIPTION (NEO 5.18)
Packtsci TItu CCttiECTICtTF YANKEE MCOERNIZE RPS - PHASE 2 PIO7ECT ,
j) Connect the relocated tubing from FT-403 to rr-403-1, Pr-403-3 and rr-403-4. Install independent valving to each trans-mitter, k) connect the relocated tubing from rr-404 to rr-404-1, Fr-404-3 and Fr-404-4. Install independent valving to each trans-mitter.
- 1) Seismically install cable and supports from the ety .
transmitters to the Main Control Room, m) Install the Foxboro I/O Modules in the Main Control Room Foxboro Cabineta.
n) Install Main Control Room cable and wiring, o) Perform a continuity test and terminate all cable and wiring, p) Energize, calibrate, and test the equipnent.
5.2.2 Reactor Coolant System Pressure l
a) Remove the erf. sting two (2) roxboro,10-50 ma wide-range pressure transmitters PT-403 and Pr-404.
.b) Seismically extend the instrument tubing from PI-407 (Reactor Coolant Pressure, Loop 1) to the new stansmitter rack betwen columns 15 and 16.
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l O 0 Page Project Assignment No. Rev.#
40 of 67 83-113 0 cmCEPIUAL PIG 7ECT DESCRIPTICH (NED 5.18)
P ETECT TITLE CCMECTICUT YANKEE MCOERNIZE RPS - PHASE 2 PETECT .
c) Cap the instrument tubing at PT-403.
d) Seismically extend the instrument tubing from Pr-404 (Reactor Coolant Pressure, Loop 4) to the net ?.ransmitter rack between colunns 5 and 6.
e) Seismically mount a wide-range Foxboro, 4-20 ma pressure transmitter PT-403 (0-3000 psig) and a narrow-range roxboro, 4-20 ma pressure transmitter PT-403N (0-600 psig) on the new transmitter rack between columns 15 and 16.
f) Seismically nount a wide-range Foxboro, 4-20 ma pressure transmitter PT-404 (0-3000 psig) and a narrow-range roxboro, 4-20 ma pressure transmitter PT-404N (0-600 psig) on the new transmitter rack between columns 5 and 6.
g) Connect the relocated tubing from Loop 1 (PI-407) to PT-403 and Pr-403N. Install independent valving to each transmitter.
h) Connect the relocated tubing from Loop 4 (Pr-404) to Pr-404 and Pr-404N. Install independent volving to each transmitter.
- 1) Seismically install cable and supports from the new tranvnitters to the Main control Room.
j) Install the roxboro I/O Modules in the Main Control Room Foxboro Cabinets.
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Page Project Assignment No. Rev.9 41 of 67 83-113 0 cmCEPmM, PIG 7ECT DESCRIPTIN (NEO 5.18)
PRCUECP TITLE CCMECTICtJP YANKEE MODERNIZE RPS - PHASE 2 PIETECP .
k) Remove the existing Taylor pressure recorders PR-403 and PR-404 on the rear of Main Control Board C.
- 1) Seismically install new Foxboro pressure recorders PR-403 and PR-404 on the rear of Main control Board C.
m) Replace the existing Digitec indicators PI-403 and PI-404 with 4-20 ma equivalents.
p) Install 4-20 ma indicators for the narrow range (0-600 psig)
RCS pressure on the Main Control Board.
o) Install Main Control Room Cable and wiring.
p) ?erform a continuity test and terminate all cable and wiring.
q) Energize, calibrate and test the equipnent.
b 5.2.3 Primary Containment Pressure a) Remove the existing six (6) Mercoid pressure switches PS-1816-1A, PS-1816-2A, PS-1816-3A, PS-1816-18, PS-1816-28, and PS-1816-35.
b) Seismically mount six (6) replacement pressure switches in the
. in the Primary Auxiliary Building (Sanpling Room).
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Page Project Assignment No. Rev.#
42 of 67 83-113 0 CONCEP1tRL P!n7ECT DESCRIPTIQ( (teo 5.18)
T 25 JECT TITLE .
CotNECTIct7r YANKEE MODERNIZE RPS - PHASE 2 PRCMECT .
c) Install Foxboro I/O dodules in the Main Control Room Foxboro Cabintets to perform the 2 of 3 coincidence logic in each
- t. rain.
d) Install Main Control Room cable and wiring.
e) Perform a continuity test and terminate all cable and wiring.
- f) Energize, calibrate and terst tha equipnent.
I Note: If suifable replacement pressure switches are not available, the existing pressure switches will be replaced with four pressure transmitters.
5.2.4 Steam Generator Narrow Pange Level I
a) Remove the existing four (4) Hagan, 1-9 VDC level transnitters '
LT-1301-1, LT-1301-2, LT-1301-3, and LT-1301-4.
b) Install four (4) equivalent Foxboro, 4-20 mA level transmitters LT-1301-1, LT-1301-2, LT-1301-3, and LT-1301-4 in place of the problematic Hagan transmitters. !
I c) Connect the existing process tubing and wiring to the '
I replacement transmitters.
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Page Project Assignment No. Rev.9 43 of 67 83-113 0 CCMCEPIUE. PRCkTECT DESCRIPTION (NED 5.18)
PROJECT TITLE CCHIECTICLT YANKEE MODERNIZE RPS - PHASE 2 PROJECT .
d) Install output conversion modules (4-20mA to 1-9 VDC),
e) Energize, calibrate and test the equipment.
5.2.5 Steam Generator reedwater riow a) Remove the existing four (4) Hagan, 1-9 VDC flow transmitters rr-1301-1A, FT-1301-2A, FT-1301-3A, and Pr-1301-4A.
b) Install four (4) equivalent Foxboro, 4-20 mA flow transmitters Pr-1301-1A, rr-1301-2A, Fr-1301-3A, and rr-1301-4A, in place of the problematic Hagan Ring Balance transmitters.
c) Connect the existing process tubing and wiring to the replacement transmitters.
d) Install outp3t conversion modules (4-2 N to 1-9 VDC).
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e) Energize, calibrate and test the equipnent.
l 5.2.6 Steam Generator Steam riow )
I a) Remove the existing four (4) Hagan, 1-9 VDC flow transmitters 1 rr-1201-1, Fr-1201-2, Fr-1201-3, ard rr-1201-4.
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dS Page Project Assignment No. Rev.9 44 of 67 83-113 0 CONCEPEAL PIGECT DESCRIPTION (NEO 5.18)
PIW ECT TITLE ,
CCtWECTICUT YANKEE MODERNIZE RPS - PHASE 2 PIGECT b) Install four (4) equivalent Foxboro, 4-20 mA flow transmicters FT-1201-1, FT-1201-2, Fr-1201-3, and FT-1201-4, in place of the problematic Hagan Ring Balance transmitters.
c) Connect the existing process tubing and wiring to the replacement transmitters, d) Install output conversion modules (4-20mA to 1-9 VDC).
e) Energize, calibrate and test the equipnent. ,
5.2.7 Reactor Trip I491c System
'Ihis plant design change will remove the existing auxiliary relays which make up the Reactor Trip coincidence logic and will replace them with two (2) redundant trains of solid-state coincidence logic. Two (2) roxboro, three-bay cabinets will be added behind ,
the Main control Board housing roxboro spec 200 and Spec 200 Micro equip,aent to accomplish this function. Individual channel reactor
' rip signals will input to these cabinets originating from the following sources:
- 1) Digital contact output signals from the existing four (4) Spec 200 racks.
- 2) Auxiliary relay contact output signals from the four'(4) channels of Nuclear Instrumentation (Reference P.A.85-066).
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Page Project Assignment No. Rev.8 45 of 67 83-113 0 CWCEPitRL PROJECT DESCRIPTION (Neo 5.18)
PROJECT TITLE CONNECTICUT YANKEE MODERNIZE RPS - PHASE 2 PRO 7ECT
- 3) Various other contact output signals form existing plant equipnent.
Note: Due to the possibility of incoming energy which could damage CMOS components on the SPEC 200 Micro equipnent, connection to dry switch contacts in the field will be buffered by Foxboro input modules.
We individual channel reactor trip signals will be input to the Train A and Train B Reactor Trip Logic Cabinets and a coincidence of channels will be performed reduntantly by each cabinet to determine if a Reactor Trip Signal is warranted. he reactor trip output signals will be input to the SCPAM breaker control cir-cuitry. Automatic Reactor Trip will occur upon loss of power to the control rods, i.e., the rods must be energized to remain withdrawn from the core. %ese trip breakers are actuated by:
- 1) Trip coils on the breakers, energized from any of the reactor trip signals.
- 2) Undervoltage trip coils (normally energized) on the breakers to ensure reactor trip in the event of loss of control rod DC voltage. %ese coils are deenergized by any of the reactor trip signals.
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l Page Project Assignment No. Rev.#
46 of 67 83-113 0 CONCEPTtIAL pin 7ECT DESCRIPTION (MO 5.18) pin 7ECT TITIE CONNECTICtTT YANKEE MODERNIZE RPS - PHASE 2 Pin 7ECT .
Note Each Reactor Trip Logic Cabinet will perform the reactor trip coincidence function twice to: (1) enhance plant reliab'ility by preventing inadvertent Keactor Trips on a single component failure, and (2) facilitate Reactor Trip System testing.
Train A will consist of Al and A2 coincidence logic while Train B will consist of B1 and B2 coincidence logic. %eir outputs will be ANDED so that a reactor trip will occur only when both Al AND A2 and/or both B1 AND B2 coincidence logic requests a reactor trip, he following reactor trip coincidence functions will be performed redundantly by the Train A and Train B Reactor Trip Logic Cabinets:
a) !. css of Reactor Coolant Flow Reactor Trips ne loss of flow in any two reactor coolant loop due to pung failure, power failure, or accidental loop valve closure may cause a reactor trip if the loss of flow occurs when operating above the permissive power setting P-7 (10 percent). Trip signals for each loop may come from any of three sources:
- 1) opening of the reactor coolant pump breaker, 2), low voltage on the main generator bus supplying the ;
reactor coolant pung, and
- 3) low D/P as measured between the inlet and outlet plenums of the steam generator.
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Page Project Assignment No. Rev.4 47 of 67 83-113 0 cmczPERL PIGECT DESCRIPTION (NED 5.18)
PfG ECT TITLE CONNECTICUT YANKEE MODERNIZE RPS - PHASE 2 PINECT .
Low voltage on a generator bus affects ,two pumps and will always cause a reactor trip above the P-7 setting. Trip signals from a reactor coolant pump breaker or a flow detector, indicating a single loop loss of flow, will cause a reactor trip only above P-8 (74 percent). Below P-8, but above P-7, a loss of flow signal from two different loops is needed for a reactor trip.
During power operations, all four reactor coolant pumps will be in service to remove core sensible heat. For a 345 kV line fault on one of the main generator buses, which will affect two reactor coolant pumps, a reactor trip will occur and the transfer of the two pumps to offsite power occurs in about one second. With this rapid transfer, enough flow is provided to prevent fuel clad damage by the high inertia of the two pumps coasting down and by the two pumps which have been transferred to the outside buses. !
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% e operation of reactor coolant pumps is sensed by auxiliary contacts on the reactor coolant pung motor circuit breakers which indicate when the breakers are opened. We coincidence !
logic for these trip signals is one of four above 74 percent full power, two of four between 74 percent and 10 percent and no reactor trip below 10 percent.
Voltage on the two 4160 volt busses supplying the reactor coolant pung motors is sensed by undervoltage relays. Since each of these buses supplies two reactor coolant pung motors, '
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Page- Project Assignment No. Rev.9 ;
48 of 67 83-113 0 ,
CCNCEPTtRL PIE 7ECT DESCRIPTION (teo 5.18) i I
Pin 7ECT TITLE P
CCt@iECTICUT YANKEE MODERNIZE RPS - PHASE 2 P!n1ECT loss of voltage on either bus initiates reactor trip whenever the plant load is above 10 percent of full power. Bere is a short time delay built into this undervoltage circuit to permit transfer for the busses from one source of supply to another.
b) Loss of Load Reactor Trip Above 10 percent power (P-7), a turbine trip will trip the ,
reactor. Any reactor trip will trip the turbine.
A turbine trip is sensed by low auto stop oil pressure or if all turbine stop valves are shut. mis signal causes the reactor trip. If the reactor trip fails to occur after a turbine trip, the temperature of the reactor coolant rises, causing a surge into the pressurizer increasing its pressure.
W e transient is then terminated by reactor trip from high pressurizer pressure or high pressurizer icvel.
c) Loss of Feedwater Flow Reactor Trip If the feedwater flow to the steam generators is decreased or
]
stopped, there is danger of losing the reactor coolant heat j- sink, first by uncovering the steam generator tubes, and ultimately by boiling the steam generators dry. Because of thermal stress consideration, it is not advisable to inject cold feedwater into a dry, high temperature steam generator.
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Page Project Ast,1gnment No. Rev.9 49 of 67 83 113 0 CCNCEPILRL PIE 7ECF DESCRIPTION (NEO 5.18)
PinTECT TITLE CONNECTICUT YANKEE MODERNIZE RPS - PHASE 2 PRCOECT To reduce steam flow quickly, a reactor trip is provided on coincident signals of large steam flow-feed flow mismatch and low steam generator water level for any one of the four steam generators. We early reactor trip will decrease the heat load on the steam generator and allow time for the auxiliary feedwater pumps to be started, preventing the steam generator from boiling dry.
This system uses the flow signals for main steam and feedwater sensed by the Steam Generator Water Level Control System. The steam flow signal for each loop is coapared to the feedwater flow signal for the same loop. If steam flow exceeds feed-water flow by more than 456,000 lb/hr (=20 percent of full power flow), a trip signal occurs and an alarm is sounded.
However, a reactor trip will not occur until the steam gener-ator level, as sensed by the narrow range level transmitter, falls to 10 percent. %e 20 percent mismatch requirement ensures that reactor trip will not occur until the icsd is high enough to require at least 456,000 lb/hr o' teedvater and steam flow per loop.
d) Steam Line Excessive Flow Reactor Trip Steam flow in each of the four main steam lines is seased as a pressure differential across elbow taps located in the first 90 degree elbow outside the reactor containment. Wis -signal is sent to a matar relay bistable in the control room. We bistable generates a trip signal when its set point is ex-
. . . . , . . . . ..c , .
_ ~
Page Project Assignment No. Rev.8 50 of 67 83-113 0 cmCEPItRL PIG 7ECT DESCRIPTICN (NEO 5.18)
PROJECT TIT!2 CONNECTICt7f YANKEE MODERNIZE RPS - PHASE 2 PIOJECT ceeded. h e set point is 110 percent of full load steam flow.
A coincidence of two of four high flows is required for the In addition to tripping the reactor, the signal will trip.
also close all four main steam line trip valves (MSWs).
We reactor will respond to a break in a main steam line as if it were a step load increase. Greater steam flow from the secondary system increases heat removal from the reactor coolant and reduces its temperature. We Re, actor Control System inserts positive reactivity in an attengt to restore coolant temperature. he lower coolant temperature also inserts positive reactivity due to the negative moderator temperature coefficient. For a large break downstream of the MSWs, the accident is terminated by' reactor trip on a coincidence of high steam flow as measured by two of four steam flow indicators, or upon closing of any of the four MSWs. For a large break upstream of the MSWs, the transient I will actuate either an overpower or low-pressure reactor trip.
In either case, if the break is not isolated, the continued cooldown will actuate the ECCS due to low RCS pressure.
e) Main Steam Line Trip Valve Reactor Trip Position switches on the MSW8 generate a reactor trip signal when any one of the valves closes for any reason. 21s trip is not in service for plant loads less than 10 percent (P-7) l
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Page Project Assignment No. Rev.9 51 of 67 83-113 0 CONCEP11RL PN:0ECT DESCRIPTION (NEO 5.16)
PIETECT TITIA CCtNECTICUI YANKEE MODERNIZE RPS - PHASE 2 PIG 7ECT ,
, If a MSW closes during power operation ?he other three steam generators will try to coopensate for the isolated steam generator. Wis can result in the steam generators attempti.ng to exceed design capacity.
We water entering the reactor for the isolated steam generator loop will be at hot leg temperature. We hotter water can cause a flux tilt and will decrease the margin to DNB.
f) Plant Startup (High SUR) i Administrative procedures are established for a safe approach to criticality. However, protection is provided for the rapid approach to criticality caused by rapid rod withdrawal, the I most adverse startup accident. his protection ist
- 1) multiple "all rods out stop" signals, initiated by a high ,
startup rate (2 DPM) from a source range or an inter- I mediate range nuclear flux detection channel. ,
l
- 2) a teactor trip signal initiated by a high startup rate (5 DPM) as measured by either intermediate range nuclear fitq detection channel (this t.:ip is prevented from reaching the reactor trip breakers above 10 percent full power).
- ( )
- Page Project Assignment No. Rev.#
52 of 67 83-113 0 CONCEPItRL PKMECT DESCRIPTION (IED 5.18)
PMMECT TITIE CormECTICUr YANKEE MODERNIZE RPS - PHASE 2 PNMECT .
- 3) the coincidence of two of four high flux level signals from the average of the upper and lower power range nuclear flux detector channels actuater a reactor trip (the trip set points are adjustable over a range con-
~
sistent with suberitical or low power operation, partial power operation, and normal full power operation).
g) Power Operation (High Flux) conditions of high power, high reactor coolant temperature and low reactor coolant pressure, or combinations of these, can ;
lead to DNB. We occurrence of DNB at high flux levels re-suits in less of heat transfer capability at the fuel clad l surface. his leads to high clad temperature with possible clad damage and release of fission products to the coolant.
An overpower rod stop (P-1) is provided to prevent abnormal high power conditions. his could result from excessive i control rod withdrawal initiated by malfunction of the Reactor control System, or by a violation of administrative
- procedures. ;
4 Protection is also provided by two independent reactor trip functions, the overpower reactor trip and the variable l low-pressure reactor trip (discussed in next section).
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Page Project Assignment No. Rev.9 53 of 67 83-113 0 .
CONCEPIUhL pin 7ECT DESCRIPTION (IED 5.18)
PIG 7ECT TITU CCN4ECTICtTF YANKEE MODERNIZE RPS - PHASE 2 PIQJECT he overpower, overtenperature and underpressure reactor trip limits are designed such that the DNB limit is never reached.
Sufficient margins are available so that adequate transient capability exists without spurious reactor trips.
Coincidence of two of four high signals from the power range actuates a reactor trip for overpower protection. The trip set points are adjustable over a range consistent with sub-critical, low power, and full power operations. ,
If the power level reaches a nominal 109 percent of full power on any two of the four channels, a reactor trip will be ,
initiated. A switch allows the operator to manually select j two lower trip set points to be used when the reactor is operated at lower power levels. %ese lower trip points are 25 percent and 74 percent of full power.
High power reactor trips protect against the following incidents:
- 1) Rod Withdrawal Incident
- 2) Inactive Loop Startup
- 3) Steam Line Break
- 4) Boron Dilution 5). Excess reed Water
- 6) Large Load Increase
- 7) Rod Ejection
- 8) 14ss of Flow j .
- t * .
- I j
Page Project Assignment No. Rev.9
. 54 of 67 83-113 0 1 cmCEP1UhL Pin 7ECT DESCRIPTIN (150 5.18)
P!nTECT TITIA J
COWECTICUT YANITE l t
MODERNIZE RPS - PHASE 2 r PIE 7ECT Note: PA 85-066 has been approved t'o replace the NIS System in 1989. Se above described operation is subject to change.
h) variable Low Pressurizer Pressure Reactor Trip Excessive boiling in the core hot channels is prevcnted by a l variable low-pressure reactor trip. We set point of this trip is calculated from average loop tenperature and loop delta temperature. he temperature rise in the core is related to the hot channel outlet temperature through the ;
design hot channel factors. We temperature rise is limited l to prevent departure from nucleate boiling (DNB). I Protection against DNB is given by the overpower (neutron flux) trip and the low-pressure reactor trip. Wese will prevent operation at a power, pressure, or temperature resulting in DNB.
I Multiple channels are provided to compute a low-pressute reactor trip setting based on maximum allowable boiling in the l core for prevention of DNB. We conditions of the outlet of l the hot channel are continuously calculated from the average core temperature conditions, using hot channel factors for computing the temperature rise across the hot channel. he allowable minimum pressure is continuously calculated as a function of plant temperature and power level. Should the pressurizer pressure fall below this value, a low-pressure l
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Page Project Assignment No. Rev.9 55 of 67 83-113 .0 ENCEPitRL FinTECT DESQtIPTIGf (ISO 5.18)
PIO78lCT TITUC CCttiECTICt7f YANKEE MODERNIZE RPS - PHASE 2 PIQ7ECT .
reactor trip is actuated. Should pressure fall rapidly for any power level, a preset low-pressure minimum is established at or below which a reactor trip would occur.
- he fixed lower limit on the calculation is set at 1700 psig.
For three of the four reactor coolant loops, the Tm and 6T signals are transmitted to a calculator which establishes a variable low-pressure trip set point. Each of the indepen-dently calculated set points is continuously compared with one of the three pressurizer pressure measurements. If two of the four pressurizer pressure channels are below their respective calculated minimum pressure, a reactor trip will be initiated when power is above ten porcent (P-7).
l i) High Pressurizer Pressure Reactor Trip A high-pressure reactor trip is provided to trip the reactor should pressurizer pressure rise above 2300 psig. his trip I will help maintain RCS integrity. l l
We trip signal is fed to a matrix of relay contacts which initiates a reactor trip if any two of the three trip signals exist simultaneously.
j) High Pressurizer Level Reactor Trip If pressurizer liquid level exceeds 86 percent of full level, 1
4 l
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Page Project Assignment No. Rev.9 ,
56 of 67 83-113 0 CONCEPItRL Pfa7ECT DESCRIPTIGI (NEO 5.18)
PIO7ECT TITLE CO NECTICUT YANKEE MODERNIZE RPS - PHASE 2 PRCL7ECT a trip signal is fed to a two of three coincidence matrix.
We sinultaneous existence of two high level trip signals will initiate a reactor trip at any power level above ten percent (P-7).
his trip prevents discharging liquid through the pressurizer power-operated relief valves and safety valves.
k) ECCS Actuation Reactor Trip A loss of coolant accident (!4CA) is the rupture of the RCS boundary. If the treak is small, the reactor coolant loss can i be replaced using the charging pungs, thereby permitting an orderly shutdown. Should a larger break occur, loss of system ;
pressure, boiling in the core, and eventual clad failure could !
follow. We Emergency Core Cooling System (ECCS) provides l
borated water in sufficient quantity to prevent clad failure for all credible fcCAs. In addition, the ECCS is required to !
add borated water following a SLB. his preserves the shut- ]
down margin.
Coincidence of two of three low pressurizer pressure signals
- (1700 psig), or two of three high containment pressure signals (5 psig), will actuate the ECCS. Provision is made to man-ually block safety injection below 1800 psig. Automatic unblocking occurs as system pressure is raised above 1875 psig. A signal to actuate safety injection will also cause a reactor trip. l 4
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Page Project Assignment No. Rev.9 57 of 67 83-113 0 CGICEP1tIhL PIQ7ECT DESCRIPTIGt (NEO 5.18)
PIQ7ECT TITLE 4
CCNNECTIctJr YANKEE MODERNIZE RPS - PHASE 2 PIQ7ECT *
- 1) Manual Reactor Trip Two manual reactor trip buttons are provided and positioned to ensure immediate access to the operator.
5.2.8 Reactor Protection System Testing Capability his plant design change will provide testing capability so that the Reactor Protection System will be testable during operation of the nuclear power generating station as well as those intervals when the station is shutdown. mis testability shall permit the independent testing of redundant channels and load groups while (1) maintaining the capability of those systems to respond to bona fide signals during operation, (2) tripping the output of the channel being tested, or (3) bypassing the equipnent consistent j with availability requirements.
A test panel will be installed in Nest 5 of the four (4) existing "protection-side" roxboro Racks in the Main control Room to allow calibration verification testing. h is testing will prove that with a known precision input, the instrument or channel gives the
, required output. We test panel will provide a toggle switch, banana jack and indicating laap for each protection process input.
- When the switch is in the "test position" the input will be transferred from the field transducer to the banana jack to allow signal injection and the output will be opened which will provide 4
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Page Project Assignment No. Rev.9 58 of 67 83-113 0 cmCEPME. P!n7ECT DESCRIPTIG( (NEO 5.18)
PinTECT TIT 12 CONNECTICUT YANKEE MODERNIZE RPS - PHAST 2 PIO7ECT 4
a trip input to the Reactor Trip Logic ~ Cabinets. When the in-jected precision input reaches the known reactor trip set point value, the indicating lamp will illuminate to verify that the required output is achieved.
Additionally, each Reactor Trip Logic Cabinet will be equipped with a test panel to allow coincidence logic configuration verification. he cost panel will provide a toggle switch and an indicating lany for each digital contact, coincidence input as well as an indicating lang for each coincidence output. We coincidence input langs will illuminate when the input channel is '
in a trip condition and/or when the toggle switch is in the "test position". We coincidence output langs will illuminate when the coincidence combinations are satisfied. We toggle switches will allow the testing of all possible input combinations to verify that the required output is achieved, rurthermore, bypass l capability will be provided for each digital contact, coincidence i input to allow bypass of the equipment consistent with plant !
availability requirements.
I 5.2.9 Power Dependent Insertion Limit (PDIL) rour isolated analog power range output signals from the thelear Instrumentation system will be input to the existing "con-trol-side" roxboro Rack DR. Dese analog power signals will be high auctioneered and a "maxinum insertion limit" will be j calculated based on the highest power signal. We "maximum i
l l
. l 1 l
Page Project Assignment '.4o. Rev.9
_ 59 of 67 83-113 0 CONCEP11AL EF DESCRIPTION (Neo 5.H)
PRa mCT TITLE CCm ECTICur YANKEE MODERNIZE RPS - PHASE 2 PRa7ECT insertion limit will be compared to the actual position of the control rods (Groups A and B). An alarm and an inward automatic red stop will be generated when the maximum ir,sertion limit is exceeded.
5.3 Design Inputs 5.3.1 American National Standards Institute a) ANSI N45.2, 1977 Quality Assurance Program Requirements for Nuclear Power Plants b) AN3I N45.J 2, 1972 Packa31ng, Shipping, Receiving, Storage, and Handling of Items for Nuclear Power plants c) ANSI C39.1, 1981 Requirements for Electrical Analog Indicating Instruments d) ANSI C39.2, 1964 Direct Acting Electrical Recording Instruments.
e) ANSI C39.5, 1974 Safety Requirements for Electrical and Electronic Measuring and Controlling Instrumentation.
I
e .
I I Page Project Assignment No. Rev.9 i
, 60 of 67 83-113 0 CGGGPTUhL PIGECT DESCRIPTICN (NED 5.18)
PIGECT TI'112 CCetECTICITF YANKE~J
> MODERNIZE RPS - PHASE 2 l 1
4
.l f) ANSI /IEEE-l ANS-7.4.3.2, 1982 Application Criteriin for Prograssaable i Digital computer Systems and Safety ,
! Systems of Nuclear Power Generating Stations.
5.3.2 American Society for Testing and Materials l I
i a) As m A262, 1986 Recommended Practice for Detecting j Susceptibility to Inter-granular Attack i
- in Stainless Steels.
i .
i b) AS m A380, 1978 Reccamended Practice for De-scaling and 1 Cleaning Stainless Steel Surfaces.
1 5.3.3 Institute of Electrical and Electronic Engineers a) IEEE 279, 1971 Criteria for Protection Systems for 1 clear Powar Generating Stationc. ;
i b) IEEE 308, 1980 Criteria for Cla u lE Power Systems for i Nuclear Power Generating Stations, c) IEEE 317, 1983 Electric Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations. i i
l f
, , ,,...,.e =
6
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Page Project Assignment No. Rev.0
, 61 of 67 83-113 0 CCNCEPIUhL PRCDECT DESCRIPTICH (MD 5.18) i PRCMECT TI'!!A CCMECTICUT YANKEE MODERNIZE RPS - PHASE 2 I
PRCUECT =
d) IER 323, 1974 Standard for Qualifying Class it Equipment for Nuclear Power Generating Stations. '
e) IEEE 338, 1977 Criteria for Periodic h sting of Nuclear ,
Power Generating Staticn Safety Systems. ,
f) IEEE 344, 1975 Trial Use Guide fer seismic Qualifica- t tion of Class lE Electrical Equipment
- for Nuclear Power Generating Stations. [
g) IEEE 336, 1985 Standard Installation, Inspection and l
Testing Requirements for Power, Instrumentation, and Control Equipment j et Nuclear Pacilities.
h) IEEE 379, 1977 Application of the Single Pailure Criterion to Nuclear Power Generating Station Class 1E Systems.
l i) IEEE 381, 1977 Criteria for Type Tests of Class it '
l Modules Used in Nuclear Power Generating l Stations, t
r I
l
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Page Project Assignment No. Rev.9 62 of 67 83-113 0 CG4CEPWhL PIWECT DESCRIPTIG4 (Nuo 5.18)
PRCX7ECT TITut ,
CONECTICUT YANKEE HODERNIZT RPS - PHASE 2 PRCA7ECT .
j) IEEE 383, 1974 Standard for Type Test of Class 1E Electric cablest Field Splices and Connections for Nuclear Power Generating Stations.
1 k) IEEE 384, 1981 Criteria for Independence of class 1E Equipnent and Circuits.
i 4 1) IEEE 420, 1982 Design and Qualification of Class 1E i
. Control Boards, Panels and Racks used in Nuclear Power Generating Stations.
m) IEEE 467, 1980 Quality Assurance Program Requirements
- for the Design and Manufacture of Class j lE Instrumentation and Electrical I Equipnent for Nuclear Power Generating Stations.
. l n) IEEE 472, 1974 Guide for Surge Withstand Capability (SWC) Tests. I o) IEEE 577, 1976 Requirements for Reliability Analysis in the Design and Operation of Safety Systems for Nuclear Power Generating Stations.
j t
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Page Project Assignment No. Rev.9 !
63 of 67 83-113 0 CONCEPRRL 75GECT DESCRIPTIG6 (M30 5.18) muscT Trna C043ECTICUT YANKEE MODERNIZE RPS - PHAst 2 i
p) IEEE 603, 1980 Criteria for safety Systemt for Nuclear i Power Generating Stations, q) IEEE 690, 1984 Design and Installation of Cable Systems f for Class 12 Circuits in Nuclear Power Generating Stations.
f l
5.3.4 U.S. Nuclear Requiatory Commission (NRC) ,
I a) 10CrR50, Appendix a Quality Assurance, CriterLt for Nuclear (
Power Plants. l I
b) 10CTR, Part 21 Reporting of Defects and Non- W liance.
c) 10Crh50, Appendix R Fire Protection Program for Nuclear Power Facilities Operating Prior to ;
January 1, 1979. i d) 10CrR50.49 Environmental Qualification of Electric Equipment Important to safety for Nuclear Power Plants.
e) 10CTR50.59 Changes, tests and experiments f) R.G. 1.22 Periodic Testing of Protection Systent Actuation Functions.
' 1 4
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4 Page Project Assignment No. Rev.9 i 64 of 67 83-113 0 CX30CEPRAL PROJECT DESCRIrrIGi (MD 5.18) {
i I
M ENT TA M COWECTICtTF YANKEE MODERNIZE RPS - PHASE 2 I
PROJECT ,
g) R.G. 1.29, Rev. 3 Seismic Design classification.
Sept. 1978 - ,
h) R.G. 1.38, Rev. 2 Quality Assurance Requirements for l May 1977 Packaging, IAlpping, Receiving, Storage, l and Handling of Items for Water-Cooled I Nuclear Power Plants. [
r
. i) R.G. 1.75, Rev. 2 Physical Independence of Electrical l l Sept. 1978 Systems. i I
j) R.G. 1.89, Rev. 1 Environmental Qualification of Certain i June 1984 Electrical Equipment Isportant to Safety l
1 for Nuclear Power Plants. .
I 1
i k) R.G. 1.97, Rev. 3 U. S. Regulatory Comunission Der:.1980 Instrumentation for Light-Wster-cooled !
Nuclear Power Plants to Assess Plant and f j Environmental Conditions during and f
] following an accident. ,
i l
l 1) R.G. 1.100, Rev. 1 Seismic Qua?.ification of Electric !
l l Aug. 1977 E7.diment for Nuclear Power Plants. j j a) NUREG-0588, Rev. 1 Ir.teria Staff Position on Envirvmental i
) July 1981 Qtnification cd Safety-Related '
Electrical Equipment. '
i l
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5 a
4
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i l i Page Project Assignment No. Rev.8 65 of 67 83-113 0 assCurnmL PnomCT rGCRIFFICBI (teo 5.18) i l
PnamCT TITLE l
3 CCENECTICtff YANKEE !
4 MODERNIZE RPs - PHASE 2 l l M ,
2
?
] 5.3.5 Instrument society of America l i ,
) a) ISA, 367.01, 1973 Transducer and Transmitter Installation j for Nuclear safety Applications.
b) ISA, 567.06, 1984 Response Time 7tsting of Nuclect safety- (
) -
Related Instrument Channels in Nuclear ;
r Plants.
. I i 5.3.6 Mechanical Engineering standards ;
l i l a) ASME Boiler and Presture Vebsel Code, section III and IX 1983 <
Edition up to and including Winter 1983 Addenda.
l l 1
b) AsME Boiler and Pressure Vessel Code, section XI, "Inservice '
1 1
Inspection". 1983 Edition up to and including the summer 1984 ;
' 2 Addenda. I l
- e) AIsc specification for the Design, Pabrication, and Erection l
] of structurel steel for Buildings; 8th Edition. f 4
l
- d) AWs D1.3, specification for Welding sheet steel and structures !
(1981), of latest Edition in effect during installation. I i
- e) AWS D1.1, structural Welding Code (1984), or latest Edition in j effect d'iting installation.
1 i
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Page Project Assignment No. Rev.9 66 of 67 83-113 0 CONCEPIUM. PNMECT DESCRIPTICM (NEO 5.18) i FMMECT TI'IU CCNiECTICtJT YANKEE i
MCOERNIZE RPS - PHASE 2 PNMECT f) Welder Qualifications per AWS D1.1,* AWS D1.3 and ASME Section IX,as applicable. '
5.3.7 Plant Design Documents / Inputs a) CYAPCO Haddam Neck Station Updated TEAR (May, 1987).
t b) NUSCO Nuclear Training Manuals for Connecticut Yankee Systems.
c) Connecticut Yankee System Descriptions, Revision 1.
I i
d) Reacter Control and Protection System Description CYW-300, ;
Westir.ghouse Electric Corporation.
e) LER 86-018-00 ,
Note: Latest issue of such specifications, standards, and codes means the issue (including latest published case niling, interpretation, anxi addenda) in form on the date of the .
PDCR. Adoption of any subsequent issue of case rulings. l shall be subject to the owner's approval. Nortbest Utilities Welding Manual referencing earlier additions to 1 the ecde are acceptable. C. N. Flagg Welding Manual referencing earlier additions to the code are acceptable.
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Page Project Assignment No. Rev.4 67 of 67 83-113 0 f
CONCEPMhL PR3HCT DESCRIPTIM (MO 5.18)
~
PRCA7ECT TITTA CCNECTICVF YANKEE MCOERNIZE RPS - PHASE 2 PHO N F 6.0 CXNCEFIUAL DESIM Mh!AD0De(
A praliminary conceptual design walkdown was performed by the Project Engineer, the Discipline En94neering, and a Plant I&C representative to determine approximate equi p nt locations. A follow-up conceptual design :
walkdown will be performed to verify the feasibility of installa*. ion of l these locations and to provide input for the detailed design. Invulvement I of the following project participants is required: l Generation Engineering Mechanics l
Generation Electrical Engineering - Special Studie, r Generation Electrical Desigr. l
- Generation Civil Design
[
Generation Mechanica'. A sign !
Plant Engineering f Plant IEC !
Betterment Construction l l
1 I l PD2 88 l
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ATmODENT 8.B SAFETY EVALLATION WORKSHEEP Safety Evaluation Numbert (later) Revision No. O Plant Change Number (if applicable): (later) Revisic,n No. O Plant Change Title Modernize RPS - Phase 2 Referencest a) NEO 3.12, Rev. 3, Safety Evaluations b) P.A.83-113 Project File c) Connecticut Yankee UFSAR
- 1. Description of the aspects of the change being evaluated.
h e instrumentation, control, and functional operability aspects of the Reactor Protection and Control System upgrade will be evaluated to determine if the plant change is safe to implement.
- 2. Identify parameters and systems affected by the change.
Wis pro et continues the Reactor Protection System (RPS) modernization effort by replacing portions of the original RPS equipment with current state-of-the-art equivalents. his modernization effort will effect the following parameters and systems:
I) A continuation of "front-end" work which includes the replacement of the sensors, transmitters, and Main Control Board mounted equipment which make up the following indication and/or trip circuitry:
a) Reactor Coolant System flow his Plant Design Change will remove the existing four (one channel per loop) flow transmitters and will install twelve (three channels per loop) Category 1, Class lE flow transmitters.
Additionally, the existing control board circuitry and modules will be removed and replaced with three channels of Category 1, Class lE equivalents. A low flow condition in a single loop will be determined by a 2 of 3 enannel coincidence m is change will provide redundancy and enhance plant reliability by preventing inadvertent Reactor Trips.
b) Reactor Coolant System Pressure his Plant Design Change will remove the existing two RCS wide-range (0-3000 psig) pressure transmitters which are installed in the loop area and fed tf reactor coolant loop 4. rour Category 1, Class lE pressurs tr.nsmitters will be installed in the outer annulus as replacements. A wide-range (0-3000 psig) and a l narrow-range (0-600 psig) pressure transmitter will be fed by
, NEO 3.12 Rev. 3 Date: 04/25/86 Page 1 of 14 l
l
i.
- 2. cont'd. ,
reactor coolant loop 1. A redundant wide-range (0-3000 psig) and a narrow-range (0-600 psig) pressure transmitter will be fed by reactor coolant loop 4. his change will provide redundant :
sensors which are independent and sufficiently separated at the connections to the process system. Additionally, the existing l control board circuitry and modules will be removed and replaced t f
with two channels of Category 1, Class 1E equivalents. he addition of narrow-range pressure transmitters will provide the ,
l j required accuracy for Residual Heat Removal (RHR) and Low Tenperature Over9tessurization System (L10PS) interlocks. %e uvement of the transmitters from the loop area to the outer ,
annulus will reduce exposure to personnel during calibration, consistent with the NU AIARA program, and reduce the cyclt l
integrated exposure of the transmitters which is a contributor to instrument error even during normal operation.
I c) Sensor Replacements '
i 1
mis Plant Design Change will remove the following problematic l sensors and install Category 1, Class 1E equivalents in their t i
places ;
- Primary Containment Pressure Switches (6) 4
- Steam Generator Narrow Range Level Transmitters (4) '
d - Steam Generator Feedvater Fl w Transmitters (4) !
4 - Steam Generator Steam riow Transm)cters (4) i o I i II) A replacement of the existing Reactor Trip relay logic with two trains '
' of solid-state Category 1, Class 1E Reactor Trip Leg!u. Wis replace- !
ment will provide redundancy and independence to enhance plant !
i reliability and to allow Reactor Trip System testing during operation.
i l f III) @e addition of a Power Dependent Insertion Limit (PDIL) calculator to the Rod Control System. f t
- 3. Identify the credible failure modes associated with the change.
I j %e following credible failure modes are associated with the Reactor !
Protection and Control System upgrade:
i a) cable open circuit .
- b) cable short circuit
< c) cable shorted to ground ,
I d) voltage fault applied to cables l e) couponent failures :
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i t
_______________s__ . -. . . .. . .
- - - i 0
PART It IMPACT CN M ACCIDENIS EVAIAW110 AS M DESICN BASIS
- 1. Identify the design basis accidents reviewed for potential impact by the L cha'* -
We design basis accidents reviewed for potential impact by the change are:
a) Uncontrolled Rod Withdrawal Trom a Subcritical Condition b) Uncontrolled Rod withdrawal From Power c) Isolated Loop Startup ;
d) Boron Dilution .
e) Excess reedwater !
f) Excessive 14ad Increase g) Loss of Coolant Flow .
h) Loss of Load !
i) Loss of reedwater Flow j) Dropped Rod !
k) Control Rod Ejection ,
- 1) Loss of Coolant Incident t m) Steam Line Rupture n) Steam Generator hbe Rupture r o) ruel Handling Incident p) Waste Gas Incident q) Hypothetical Accident (Radiological Evaluation) r) RCP Rotor Seizure and Shaft Break
- 2. Discuss how the parameters and systems, affected by the change, impact the ,
consequences of these accidents. Use Attachment 8.A for guidance.
We parameters and systems, af fected by the change, do not impact the consequences of these accidents. Only a limited set of parameters l associated with the plant response to an accident have been designated as consequences for determining the bounds of the accident analysis. Wese parameters relate directly to the boundary performance during the accident l and are: >
a) fuel / cladding: IUBR,MCPR fuel temperature i fuel enthalpy ,
clad strain L clad temperature !
clad oxidation t
b) RCS boundary pressure c) containment: pressure d) radiological: doses l
'1 NEO 3.12 Rev. 3 Date: 04/25/S6 Page 3 of 14
. . . ... . a
- 2. cont'd An increase in consequences is defined as an increase in one of these parameters. Since the value of these parameters are based on conservative calculations, a change in the value of the parameter does not necessarily mean an increase in consequences. An increase in the value must be signi-ficant before it is called an increase in consequences. Wis change does not cause an increase in any of the above parameters since this system upgrade intends to maintain the present functionality of the Reactor Protection and Control System. Werefore, by definition, the charae does not cause e.n increase in consequences beyond the limits for a significant change, is not a change beyond the bounds of the accident analysis, itnd is not an unreviewed safety question on the basis of consequences of ac:ident analysis. .
- 3. Identify the design basis accidents, if any, for which failure modes associated with the change can be an initiating event.
The failure modes associated with the change can be an initiating event for the following design basis accidents:
a) Excess reedvater UrSAR Analysis
Conclusion:
The results of the analysis demonstrate that the minimum DiBR remins well above 1.3. In addition, there are alarms such as the high-high steam generator level that would alert the operators of the malfunc-tion in the steam generator level control system. %ese alarms in combination with the indications of increased core power, decreased RCS inlet temperature, increases SG level, and increased feedwater flow rate will alert the operator to the excess feedwater flow transient in sufficient time to prevent SG overfill.
b) Loss of Normal reedwater Flow UrSAR Analysis
Conclusion:
he peak RCS pressure resulting from a total loss of normal feedwater does not exceed 110% of design pressure. In addition, liquid discharge from the pressurizer safety valves is not predicted to occur. %e minima M?BR remains above 1.3.
- 4. Discuss the impact of the change on the probability of occurrence of the design basis accidents identified in 3. Use Attachment 8.A for guidance.
We impact of the change on the probabilities of occurrence of the design basis accidents identified above is considered negligible since the change replacas existing equipnent with upgraded equipnent to perform the identi-cal functions and installs additional equipment to compliment these func-tions. The probability of failure of the replaced and added equipnent is considered no greater thaa the probability of failure of the existing i
NEO 3.12 Rev. 3 Date: 04/25/86 Page 4 of 14 l
- 4. cont'd.
equipnent. We installation does not increase the probabilities to an unacceptabl2 level since the change: 1) does not result in a change in the probability class, 2) is not anda major
- 3) dcesincrease not increase in thethe probability ofof probability occurrence an accident, within a category,ht previously thoug incredible, to the point where it is now considered credible. W erefore, the change is not considered an Unteviewed safety Question on the basis of the probsbility of accidents.
- 5. Identify the safety systems affected by the change.
We Reactor Protection System (RPS)
- he Engineered Safety Feature Systers (EST)
We Nuclear Instrumentation System (NIS)
- 6. Discuss the inpact of the change and/or the failure sodes associated with the change on the probability of failure of these safety systems. Use Attachment 8.A for guidance.
We change a' .1/or failure modes associated with the change are assumed to have no negative impact on the probability of failure of the safety system.
We change will upgrade existing obsolete instrumentation with state-of-the-art Category 1, Class 1E instrumentation to perform the identical functions.
We following failure analysis information pertains to the roxboro equipment utilized for the PA 83-113 modernization effoet:
SPEC 200 Analog,Equipnent Lambda - 3.57 failures / millions hours MTar - 32 years SPEC 200 Micro (2CCA-S, Single control card)
Iarida - 19.5 failures /million hours e r - 5.84 years SPEC 200 Micro (2CCA-D, Dual Control Card) i Lanbda - 23.6 failures /million hours l m r - 4.84 years ne change will install three electrically redundant Reactor Coolant System rics transmitters per loop in place of the existing single flow transmitter per loop. A two-of-three logic will be utilized per loop. His change and/or the failure modes associated with this change are assumed to have a positive impact on the probability of failure of the safety syst.em since a single failure of one flow transmitter will not render the protective function inoperable.
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Page 5 of 14 1 1
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- 6. cont'd.
n e change will install a wide-range Reactor Coolant System Pressure transmitter (0-3000 psig) and a narrow-range Reactor Coolant System Pressure transmitter (0-600 psig) fed from Reactor Coolant Loop 4. Iden-tical, redundant Reactor Coolant System Pressure transmitters will be in-stalled and fed from Reactor Coolant Loop 1. Wis change and/or the failure modes associated with this change are assumed to have a positive impact on the probability of failure of the safety system since indepen-dence and redundancy will be provided such that no single failure of a component will interfere with the proper operation of an inde ndent or redundant counterpart. Additionally, the narrow-range tran tters will provide a more accurate pressure signal to the Residual Heat Removal (PHR)
System and the Low Temperature Overpressurization System (L10PS) inter-locks. W e new transmitters will be installed in the outer annulus to l ieduce their cyclic integrated exposure which is a contributor to instru-ment error (Presently, two wide-range Reactor Coolant System Pressure transmitters, 0-3000 psig, are installed in the loop area and are fed by Reactor Coolant Loop 4).
Two seismically mounted transmitter racks will be installed in containment to support the Reactor Coolant System Flow and Pressure transmitters. One transmitter rack will support the six flow transmitters for loops 1 and 2, and the two pressure transmitters for Loop 1. We other transmitter rack will support the six flow transmitters for Loops 3 and 4, and the two pressure transmitters for Loop 4. Cabling to redundant transmitters will be f physically separated and installed in separate conduit in containment.
Outside of containment, cabling to redundant transmitters will be phyically separated and installed in separate cable tray. he conduits and cable trays containing redundant transmitter cabling may be supported 4
by the same support. We transmitter rack, conduit supports, and cable i
tray supports will be seismically installed such that their failure will not be considered credible. Thus, the single fcilure criteria remains intact.
W e change will remove the existing six Primary Containment Pressure Switches and install four Primary Containment Pressure Transmitters. The logic will change from 2-out-of-3 (twice) to 2-out-of-4. 21s change and/or the failure modes associated with this change are assumed to have a
'l positive impact on the probability of failure of the safety system since !
the pressure transmitters will provide higher accuracy and better repeata-bility than the pressure switches.
W e change will replace the following problematic transmitters with Category 1, class 1E equivalents:
- Steam Generator Narrow Range Level Transmitters (4)
- Steam Generator reedwater riow Transmitters (4)
- Steam Generator Steam Flow Transmitters (4) m is change and/or the failure modes associated with this change are assumed to have no negative impact on the probability of failure of the ,
j safety systems since state-of-the-art equipment will replace existing problematic obsolete equipnent. l I
. NEO 3.12 Rev. 3 l Date: 04/25/86 Page 6 of 14 l 1
i 6 cont'd he change will remove the existing Reactor Trip relay logic and replace it ,
with two trains of solid-state Reactor Trip logic. Individual sensor !
channels will provide isolated bistable inputs to both Train A and Train B ,
- of the redundant Reactor Trip Logic Cabinets. Each Reactor Trip Logic l Cabinet will perform the reactor trip ceincidence function twice. Train A .
will consist of Al and A2 coincidence logic while Trcin B will consist of i
al and 52 coincidence logic. We individual coincidence logic outputs will l
be ANDED in each train such that a reactor trip will occur only when both Al AND A2 and/or both B1 AND B2 coincidence logic request a reactor trip.
- nis change and/or the failure modes associated with this change are ;
1 assumed to have a positive impact on the probability of failure of the i safcty ystem since independence and redundancy will be provided such that ,
i any sin le failure within the protection system will not prevent proper ,
protect ve action at the system level when required. Additionally, this 1
plant design change will provide testing capability so that the Reactor Prote': tion System will be testable during operation of the nuclear power genera
- station as well as those intervals when the station is shutdown.
- his tt .uility shall
' and loau groups while (permit thethe
- 1) maintaining independer capability oft testing of redundant those systems to channel t respond to bona fide signals during operation, (2) tripping the output of
. the channel being tested, or (3) bypassing the equipment consistent with availability requirements.
4 I rinally, the change will install a "control-grade" Power Dependent Insertion Limit (PDIL) calculator to the Rod Control System. Tour analog power range output signals will be high auctioneered and a "maximum l i insertion limit" will be calculated based on the highest power signal. !
1 nis change and/or the failure modes associated with the change are assumed [
] to have no negative impact on the probability of failure of the safety j system since the power range channels will be independently and electri-cally isolated from the PDIL calculator. Se safety system will not be degraded.
i ne independence of added equipment is verified by observing that no l 1
equipnent or points of vulnerasility are in comon with counterpart I
- redundant equipment channels, systems, or combination of systers. One exception to this, is the use of the existing single pair of instment tubing in each loop as a process connection to the three electrically !
redundant RCS flow transmitters per loop. However, the net effect of this l change is positive since the increased electrical redundancy will allow a (
- single electrical component failure without rendering the protective [
function inoperable. Prior to this change, the protective function was j vulnerable to a single electrical component failure since there was only !
one flow transmitter per loop. Electrical isolation of instrumentation and i control circuits will be achieved through the use of Class lE isolation ;
devices applied to interconnections of Class 1E and non-Class lE circuits, !
as well as, Class lE logic circuits of redundant divisions. Separation of ;
added redundant equipaent begins at the process sensors and is maintained ,
in the field wiring, containment penetrations, and analog protection ,
cabinets to the redundant trains in the logic cabinets.
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. NEO 3.12 Rev. 3 l
- Date: 04/25/86 l Page 7 of 14 l i
1
- 7. Discuss the impact of the change on the performance of these safety systems. Use Attachment 8.A for guidance, a) he change does not degrade the rformance of a safety system assumed to function in the accident anal sis, b) ne change does not increase the challen es of the safety systems assumed to function in the accident anal sis.
c) he change does not increase the probability of failure of systems designed to reduce challenges to safety systems assumed to function in i the safety analysis.
j d) %e change does not alter the performance *of systens designed to l reduce challenges to safety systems, such that the system no longer l reduces challenges to the safety system. i i
Additionally, the change will not increase the overall safety system ,
l response times beyond the limits given in the Updated Final safety Analysis l Report (UTSAR).
%erefore, the change is within the bounds of the accident analysis and is not an Unreviewed safety Question on the basis of the performance of safety l systems.
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NEO 3.12 Rev. 3 i
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Date: 04/25/86 Page 8 of 14 i J
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Yes No e
o Based upon 2, does the change increase the consequences of a design basis
]
i accident? [] [x)
) o Based upon 4, does the change increase J. the probability of a design basis
- accident? () (x) 1 1 .
- o Based u 6, does the change increase l the pr bility of a failure of a !
- safety system? () (x) ;
a
'f l o Based upon 7, does the change degrade ,
the performance of a safety system !
below that assumed in the design basis
'l analysis? () (x) l 1
If any of the above is answered yes, the change is an unreviewed safety question. <
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NEO 3.12 Rev. 3 !
Date: 04/25/86 I Page 9 of 14 !
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t E.E_'______
O PART II: Poll!NTIAL FOR QTATICN OF A NDf LNANALYZED EVENT
- 1. Based upon Part I, assess the impact of the change ar4/or failure edes associated with the change, to deterr.ine if the igact has modified the plant response to the point where it can be considered a new accident.
Discuss the basis for this determination.
Part I determined that the change, and the failure modes associated with the change, have not degraded the performance of a safety syr. tem below that assumed in the design basis analysls. We protective functions provided by ,
J the system will be maintainad and will be performed by Category 1, Class lE equip.nent. Based on this, it is determined that the impact of the change and/or failure modes associated with the change will not modify the plant response to the point where it can be considered a new accident.
- 2. Detemine if the failure modes associated with the change represent a new unanalyzed accident. Discuss the basis for this determination, ne failure medes asseeiated with this cMnge are similar to the failure modes of existing equipment and do not represent a new unanalyzed accident. l Mditionally, since independent and rhysically separated redundant channels are provided in the instrumentation cabinets, any single failure within the protection system cabinets will not prevent proper protective action at the
< systec level when required.
- 3. Determine if the change, or a failure mode associated with the ,
increases the probability of an accident to the point where it d be considered within the design basis.
1 The change and the failure redes associated with the change, do net increase the probability of an accident to the point where it should be considered within the design basis since the change vill be f.inctionally equivalent to existing equipnent, at the system level, and the failure modes will be similar to those of existing equipnent such that the consequences of a failure would not be worsened, i
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i i NEO 3.12 Rev. 3 Date: 04/25/86 Page 10 of 14
o: .
PuremRY Yes No Based upon 1, 2, and 3, does the change create the potential for a new unanalyzed accident? () (x)
If the answer is yes, the change represents an unreviewed safety question.
. NEO 3.12 Rev. 3 Date: 04/25/86 Page 11 of 14
~ ___ . _
PART III1 IMPACT G4 'IHE MMGIN OF SMTIT
~
- 1. Based upon the consequences identified in Part I, discuss thr. impact of the !
consequences on the protective boundaries. Use Attachment 8.4 for guidance. ;
No increase in consequences was identified in Part I as a result of the change. Werefore, the impact of the consequences on the protective boundaries is unchanged.
- 2. Identify how the protective boundaries, if any, are directly affected by l the change or a failure mode of the change. Use Attachment U.A for l guidance.
We protective boundaries re not directly affected by the change or a failure mode of the change.
- 3. Discuss the impact of the change on the safety limits for the protective boundaries identified in 2. Use Attachment 8.A for guidance. e i
N/A - see 2 above.
{
- 4. Discuss the impact of the change on the basis of hchnical dpecifications.
Use Attachment 8.A for guidance, i mis change does not negatively impact the basis of Technical Specifica- l tions, n e margin of safety is not reduced since the safety limits, as
- described in the Technical specifications, are unchanged. Additionally, ;
the assumptions made on system performance for accident analys:.s remain ,
valid since system performance is not degraded. It is ersumed that the ;
replacement of existing obsolete Protection System equipent with state-of-the-art, category 1, Class lE equipment will increase system reliability
- and decrease instrucent drif t, ne change will install this equipment in a manner to provide independence and redundancy. Physical separation and electrical isolation will be provided internal to the protection cabinets to reduce the likelihood of interactions between channels in the event of channel malfunction, n ree redundant Reactor c olant System riow trans- :
mitters will be installed on each loop in place of the existing single flow (
transmitter. ;
ne assumptions made on system performance for accident analysis will be guaranteed by administrative controls and Technical Specifications requirements foc surveillance and availability to ensure that the Single ;
railure criteria remains valid. Administrative controls wi)1 require !
testing at intervals determined from consideration of manuft uturer's i specifications and historical experience with usage of similar equipment.
W e effect of the testing interval upon performance of the protection equipment will be evaluated periodically, as appropriate, to determine if i the interval used is an effective factor in maintaining the equipment in an l oprational status.
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NEO 3.12 Rev. 3 Date: 04/25fM Page 12 of 14
4 SLMMARY Yes No
- 1. Based u pn 1, do the consequences of the design Msis accidents exceed the limits for an unacceptable change given in Attachment 8.A? [] [x]
- 2. Based upon 2, 3, and 4, does the change reduce the margin of safety provided for the protective boundaries?
[] [x]
If any of these questions are answered yes, the change is unsafe and should not be implemented.
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tax) 3.12 Rev. 3 Date: 04/25/36 Page 13 of 14
PART IV: SAFETY EVAIARTIM CWCIAJ5ICM ,
assed upon the evaluation in Parts I, II, and III, the changes i
E
[x] is safe and in not an unreviewed safety question ;
[] is safe but is an unreviewed safety question I
[] is unsafe and cannot be inplemented ;
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J. N. Kowalchuk Date f I&C Engineering !
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V. J. Mazzie, Supervisor Date l I&C Engineering ,
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r T. A. Shaffer, Manager Date !
IEC Engineering [
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A. R. Roby, System Manager Date Generation Electrical Engineering l
r G. L. Johnson, Director Date i Generation Engineering and Design ,
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. NEO 3.12 Rev. 3 i Date: 04/25/86 i Page 14 of 14 ;
.- o, Docket No. 50-213 B12999 i
Attachment II Replacement of Nuclear Instrumentation System Conceptual Project Description and
- Preliminary Safety Evaluations l
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September 1988