ML20212D154
| ML20212D154 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 11/14/1986 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20212D124 | List: |
| References | |
| TAC-64359, NUDOCS 8612310302 | |
| Download: ML20212D154 (63) | |
Text
f OSMUD DESIGN BASIS REPORT 2
- November 4, 1986 om e m
.n mmse nas om a ammo on as:n==
wcsa noussT e A-5415, Rev. 3 N/A 104415 001 anon si.
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PURPOSE OF DESIGN CHANGE:
See Attached E.
DESIGN CRTERIA US&
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See Attached Ill. CALCtAATIONS & DESIGN INFORMATION:
See Attached IV. FAILURE MODES:
O THIS CHANGE DOES NOT AFFECT CONTROL ROOM INSTRUMENTATION D THIS CHANGE AFFECTS CONTROL ROOM INSTRUMENTATION. SEE ANALYSIS V.
SPECIAL MAINTENANCE REQUIREMENTS:
(
See Draft Test Spec. Surveillance Requirements; Reference 11 VL SPECIAL OPERATING REQUIREMENTS:
See Attached
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V11. VER FICATION CRITERIA:
Per Individual Sub-ECN Requirements l
Vm. COMMENTS:
See Attached Nem g
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[D TABLE OF CONTEN*S, PREFACE.
. 02 s
I. PURPOSE OF DESIGN CHANGE.
02 II. DESIGN CRITERIA.
. 05 II.A.
SUMMARY
OF CHANGE.
05 II.B. DESIGN BASIS.
. 19 24 II.C. SCOPE.
II.D. EQUIPMENT CLASS & POWER REQUIREMENTS.
24 24 II.E. TESTING.
... s.......................
III. CALCULATIONS AND DESIGN INFORMATION.
26 III.A. DESIGN FEATURES.
26 III.B. FUNCTIONAL DESCRIPTION.
. 26 III.C. DESIGN CALCULATIONS.
.26 III.C.l. AFW FLOW.
. 26 i
III.C.2. CONDENSATE STORAGE TANK CAPACITY.
. 27 III.C.3. EFIC SETFOINTS
. 27 III.C.4. STEAM GENERATOR LEVEL CONTROLS.
. 28 III.C.S. MAIN FEEDWATER OVERFILL
. 28 III.C.6. LOMFW ANTICIPATORY TRIP
. 29 III.C.7. AFW PUMP RUNOUT
. 30 III.C.8. EFIC AFW POWER SOURCES.
. 30 III.C.9. UPGRADE AFW RELIABILITY-
. 31 III.C.10. HELBA & MISSILE STUDIES.
. 32 III.C.ll. EFIC SHUTDOWN BYPASS.
. 33 i
I I
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TABLE OF CONTENTS IV. FAILURE MODES.
. 33 34 IV. A AFW VALVE FAILURE.
IV.B FAILURE OF FIBEROPTIC CABLES BETWEEN CHANNELS.
35 IV.C FAILURE OF RPS INPUTS TO EFIC.
37 IV.D FAILURE OF SFAS INPUT TO EFIC.
38 IV.E FAILURE OF,EFIC TRIP INTERFACE EQUIPMENT.
39 39 IV.F EPIC POWER SOURCE TAILURES...
IV.G EPIC CONTROL FAILURE.
. 41 VI. SPECIAL OPERATING REQUIREMENTS 44 VI. A OPERATING DESCRIPTION OF EFIC CONTh0LLED DEVICES.
. 44 VII. VERIFICATION CRITERIA.
48 VIII. COMMENTS
. 48 VIII.A DESIGN VERIFICATION.
48 VIII.B DIFFERENCES BETWEEN R.S, CR-3, AND ANO EFICS.
. 49 VIII.C USE OF ORIGINAL PLANT VALVES.
. 53 FIGURE VI - EPIC CONTROLS ON HISS (E) 54 TABLE VI - EFIC CONTROLS.
55 LIST OF REFERENCES.
57 l
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ECN A-5415 MAJOR NCR Work Request 104415
} ppg Discipline I& C MOD 001 Date 10-31-86 FREFACE This DBR covers the Emergency Feedwater Initiation and Control System (EFIC) and its functional ties to associated equipment. That is, it covers the concept and equipment design of EFIC and the Trip Interface Equipment (TIE) 9s well as the functional requirements imposed by EFIC on equipment which interfaces directly with EFIC or the TIE, or is otherwise included in Mod 1.
Ocher than EFIC and TIE cabinets; this DBR does not attempt to address component specific design criteria or design information associated with new equipment (or changes to existing equipment) except when such criteria or information directly impacts the function of EFIC. For instance, this DBR explains that EFIC requires Steam Generator level taps at 6',
156", and 619', but does not cover the specific designs of the new level taps. 'That tap design and its -
attendant calculations is covered by the DBR for sub-ECN A-5415A.
Another example would be that this DBR covers required channel separation for cabling between EFIC and equipment connected to it.
However, the actual cable routing and separation studies showing that the required separation has been maintained, is to be covered by the appropriate sub-ECN.
Another example is that this DBR covers the functional necessity to have two normally open, fail open, control valve's in parallel (each with a motor operated isolation valve in series) feeding AFW to each steam generator. However the seismic design of the modified piping and valves would be covered under the sub-ECN which installs the
(
valves (AS415 J).
The scope of each sub-ECN can be found in the major ECN A-5415.
I.
PURPOSE OF DESIGN CHANGE:
Several severe transients at nuclear power plants with Babcock &
Wilcox supplied NSS's which occurred during the 1970's were caused by inappropriate, post reactor trip, steam generator feedwater and/or steam pressu're control. Reactor Coolant System overcoolings at Rancho Seco and Crystal River 3, undercoolings at Davis-Besse and TMI-2, and others were investigated by the NRC in the spring of 1980. The transients and suggested plant alterations are summarized in NUREG 0667, " Transient Response of B&W Designed Peactors."
In summer 1980 following issuance of NUREG 0667, and in the hiatus-between TMI Short-Term lessons learned (NUREG 0578) and the clarification of TMI Action Plan Requirements (NUREG 0737) SMUD agreed.in discussions with the NRC to install *EFIC" and to upgrade the AFW system to substantially comply with the NRC: ' Standard Review Plan' for auxiliary Feedwater Systems (NUREG 0800, Section 10.4.9).
5415MAJ Page 2 Rev. 2 A
ECN A-5415 MAJOR NCR Work Request 104415 Discipline I& C MOD 001 Date 10-31-86 A conceptional design for EFIC and its related plant modifications was submitted i~n draf t form to, the NRC in October of 1980 and a preliminary Safety Evaluation, Report based on a point by point comparison with SRP 10.4.9 was received f rom the NRC in January 1981. The salient features of the design at that time were to provide:
o Assured redundant availability of automatically initiated AFW for a,ll AFW design basis events.
o Redundant safety grade control of AFW to assure sufficient but not. excessive AFW flow, o
Isolation of Main Feedwater (MFW) and AFW to prevent continued feeding of a Steam Line Break (SLB) inside containment.
o Failsafe control of ADV's to prevent 'mid-range" failure on loss of control power and to prevent common mode failure which would open ADV's on both main steam lines.
Following'the District's initial commitment to install the Mod 1 changes, several licensing issues have been resolved by inclusion of the specific licensing requirement into the design base of Mod 1.
Thus, the Mod 1 (EFIC) responsibility to resolve NUREG 0737 II.F..l.1, II.E.1.2.1, and II.K.2.10 (AFW reliability, safety grade AFW initiate, and Anticipatory Reactor trip on loss of MFW), and portions of Reg Guide 1.97 (class I Steam Generator level and pressure indication).
Listed here is pertinent licensing correspondence:
1.
SMUD to NRC letter dated May 6, 198C: Responds to IE Bulletin 80-04 Steam Line Breaks in Containment.
2.
'SMUD to NRC letter dated Octcber 6, 1980: Correction of ADV failure mode on loss of NNI or ICS power.
3.
NRC to SMUD letter dated January 22, 1981; Preliminary Safety Evaluation Report of AFW upgrade.
4.
NRC to SMUD letter dated July 10, 1981; Order letter to comply with NUREG 0737.
SMUD to NRC letter dated September 8, 1981; Submitted AFW System 5.
Description (B&W Document No. 15-1120580-01) as AFW upgrade design; also submitted upgraded AFW reliability analysis.
4 5415MAJ Page 3 Rev. 2-O
bo ECN A-5415 MAJOR NCR Work Request 104415 Discipline I& C MOD 001 Date
'10-31.... _ _...
6.
SMUD to NRC letter dated April 15, 1982: Clarifies use of Flux vs. MFW flow as ultimate Anticipatory Reactor Trip (NUREG 0737 II.K.2.10).
7.
SMUD to NRC letter dated July 6,.1982; Responds to IE Bulletin 81-14: Seismic Qualification of AFW System.,
8.
NRC to SMUD letter dated September 7, 1982: Safety Evaluation Report of upgraded AFW system (per AFW System -Description B&W Document 15-1120580-01).
9.
SMUD to NRC letter dated October 22, 1982; Clarifies' intent of District response to 0737 II.E.1.2.1 and 1.2.2: Asserts SMUD ability to alter des,ign from SER'd version of AFW upgrade.
NRC to SMUD letter dated November 12, 1982; SER of Seismic design of the then " current
- AFW system.
- 11. SMUD to NRC letter dated February 18, 1983: Supplies additional AFW desigr. basis information.
- 12. NRC to SMUD letter dated April 7 1983: S.E.R.
for 0737 II..E.1.1
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but with three open items.
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- 13. SMUD to NRC letter dated April 28, 1983; Submitted B&W Document 15-1120580-03 as. latest design for AFW upgrade.,
- 14. NRC to SMUD letter dated September 26, 1983; SER for 0737
- II.E.1.1 but with one exception.
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- 16. SMUD to NRC letter dated January 17, 1986; Brief statement of Cycle 8 EFIC scope.
- 17. SMUD to t{R'C letter dated March 3, 1986; Clarification of Cycle 8 EFIC.
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ECN A-5415 MAJOR NCR Work Request _ 104415
[2 [.} -
Discipline I& C MOD 001 Date 10-31-86 II.
DESIGN CRITERIA II.A. Summary of Chance II.A.l.
This modification will place into service the Emergency Feedwater Initiation and Control System (EFIC).
EFIC is a four channel electronic logic and control system designed to meet IEEE 279 requirements for redundancy, channel independence, testability, etc.
The principal functions of EFIC are to
" initiate" AFW, to control AFW flow to assure sufficient (and not excessive). flow, to detect a leak in the Main Steam System (as indicated by low pressure in either Steam Generator), to isolate Main Feedwater (MFW) to a deprescurized or overfilled Steam Generator, to control the Atmospheric Dump Valves (ADV's), and to assure that AFW is fed only to an operable (ie, not depressuri:ed) Steam Generator.
II.A.2. "To support the EFIC and its required functions the following changes will be made-II.A.2.1.
Install level sensors on the steam generators and pressure sensors on the main steam lines. Jpecifically, this includes:
II.A.2.1.1.
Install level taps And root valves in two places on the secondary side of each steam generator at 156' above the i-top of the lower tube sheet and two places at 614' above the top of the lower tube sheet.
II.A.2.1.2 Install four wide range level transmitters on each steam generator between taps at 6' and 619' (LT-20507A, B, C, D
and L -20508a, B, C, D).
Install four low range level transmitters on each steam generator between taps at 6' and 156*.
(LT-20505 A,B,C,D and-LT-20506 A,B,C,D).
Transmitters shall be grouped as indicated on Fig. 3.1-:
sheet 2 attached to EF!C AFW System Description (Ref. 28).
5415MAJ Page 5 Fev. 2
ECN A-5415 MAJOR NCR Work Rtquant 104413
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s Discipline I& C MOD 001 Date 10-31-86 II.A.2.1.3 Install four pressure transmitters to each msin steam line..(outside... _ _. _. ___
containment).
(PT-20546 A,B,C,D and PT-20545 A,B,C,D).
Transmitters grouped ac shown on Fig.
3.1-1 Sheet 2 (Ref. 28).
II.A.2.1.4 One wide range and one low range steam generator level transmitter, and one steam generator pressure transmitter will be connected to each of the four EFIC channels. Power for the transmitters is supplied by the respective EFIC channel.
II.A.2.2 Provide non-interruptable Class 1, AC power to EFIC Cabinets H4FWA, H4FWB, H4FWC, H4 FWD and reliable AC power to the non-lE portion of H4FWC and H4 FWD.
Provide power to the Trip Interface Cabinets H4EIAl, H4EIA2, z$6 H4EIB1 and H4EIB2.
II.A.2.3 Install the electrical and fiberoptic 7
connection between the EFIC cabinets (H4FWA, B,C and D) and to the Trip Interface Cabinets (H4EIAl, H4EIB1).
II.A.2.4 IE connections between the four EFIC cabinets 3
and the plant computer (IDADS) shall be made to monitor all EFIC analog indications (24 total signals). Isolated outputs of the EPIC and TIE annunciations will also require connection to the plant computer. Indication out of the TIE that EFIC has started the AFW pumps will be annunicated in the control room and indicated on H2SF.
Indication out of t.he TIE that EFIC has isolated MFW will be annunicated in the control room.
II.A.2.5 Implementation of NI/RPS changes. B&W Field C.nange Package.3473 changes the RPS to accept MFW ficw signals; to develop a reactor powe:
to MFW flow reactor trip (See figure 4.*.1 or REF. 28); to output power /MFW flow trip, RCP running status, channel bypass, and loss of.
MFW Pump Anticipatory Reactor Trip information to EFIC.
Although the power /MFW flow trip modules will be installed, it will not be used by EFIC at this time.
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i ECN A-5415 MAJOR NCR W3rk R;quset 104415
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Discipline I& C MOD 001 Date 10-31-86 II.A.2.5.1 Modify the B&W NI/RPS Field Change Package FCP-3473 such that Reactor Poweruvs. MFW flow-trip logic does not trip the reactor.
II.A.2.5.2 Jumper the power /MrW flow inputs' at EFIC so that they will,not actuate EFIC.
II.A.2.5.3
' Adapt Field Change Package FCP-3473 to conform to existing NI/RPS module and terminal block layout. Modify RPS seismic and thermal analyses to conform to revised layout.
II.A.2.5.4 Install the RPS Field Change Package hardware.
II.A.2.5.5 Connect the NI/RPS outputs to their respective EFIC
- channels and to the Plant Computer (IDADS) as appropriate.
II.A.2.6 Implement B&W SFAS field change package FC-3478 Revision 5 and upgrada seismic analysis to conform to actual SFAS system configuration. The B&W Field Change Package Changes the SFAS to output to EFIC a signal to start AFW on low RC pressure and/or high containment pressure.
Indication that SFAS has signalled EFIC to start AFW and that EFIC has started AFW I
will be displayed on H2SF.
II.A.2.7 Delete SFAS control functions fer P-319, SFV-308C1, SFV-20577, and SFv-20578 from H2SF.
II.A.2.8 Modify AFW pump P-319 to receive a priority start signal from TIE cabinet H4EIA1. Manual start /3 top controls will be instal-led on ElSS in the Control Room. All other start /stop controls will be deleted except pump / motor protection and diesel generator N.S. Bus loading logic.
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Work R;guset 104415
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Discipline I& C mod 001 Date 10-31-86 II.A.2.9 The AFW pump turbine steam inlet valve HV-30801 (Formerly SFV-30801) controls shall be modif,ied to receive,,a pt,iority open, sign,al.
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from TIE cabinet H4EIBl. The SFAS auto start control for HV-30801 will be deleted.
II.A.2.10 Implement miscellaneous changes to EPIC to accommodate Rancho Seco control configuration and to add initiation, time delay module.
Specifically, modify the EFIC to implement the following items:
Make wiring changes necessary to provide for transfer of control from the Control Room to the shutdown pane} and to provide isolation of EFIC from the control room to meet Appendix R requirements for s&fe
. shutdown in the event of fire in the control room.
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Change the color of some of the LED indicators on the EPIC panel fronts to improve readability.
Make wiring changes to provide vector
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enable directly from the AFW trip initiate modules and to provide only control enable from the C/V Enable (to be i
renamed Control Enable) trip module.
This gives the capability to reset the Vector enable without resetting the control enable.
Install jumpers on any unused inputs that are fail actuated (i.e. ARTS 1, ARTS 2 and power /MFW flow trip).
Make wiring changes necessary to provide separate annunciation for the non-lE power supply and the lE power supplies.
Modify the shutdown typass permissive circuitry such that bypassing is possible when either SG pressure is below bypass setpoint.
Disconnect the overfill inputs from the vector logic. This will disable the automatic isolation of auxiliary feedwater on steam generator overfill.
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i ECN A-5415 MAJOR NCR W3rk R;qu3ct 104415
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Discipline I& C MOD 001 Date 10-31-86 Implement an overfill bistable circuit -
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for use in monitoring S.G. overfill setpoint without risk of spuriout EFIC -
-jg initiates.
Add the spare fiber optic cables between EFIC channels so they will be installed spares.
Install time delay modules in each EFIC
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cabinet such that the outputs of the following bistables can be delayed by an adjustable time of 0.0 to 9.9 seconds:
'SGA Low Level, SGB Low Level, SGA Hi Hi Level, SGB Hi Hi Level, SGA Low Pressure, SGB Low Pressure, SGA Pressure less than S,GR Pressure, and SGB Pressure less than SG-Pressure.
Note that installation of the time delay circuits is desirable but not critical to the operability of EFIC.
II.A.2.ll Modify controls for AFW Flow Test Valve FV-31855 to allow modulating control from the
/kL control room with position indication and priority close commands from TIE Cabinets H4EIAl and H4EIBl upon AFW initiate.
II.A.2.12 Disconnect ICS control to the atmcapheric dump valves and connect ADV's to EEIC control.
Y.
PV-20562 A,B and C Will be connected to EFIC channel B.
PV-20571 A,B and C will be connected to EFIC channel A.
Existing manual / auto stations on HlRI will be removed.
i New manu&l/ auto pressure control stations will- ~
be mounted on the control room panels HlRI and the shutdown panel; one set for SGA, one set fer SGB.
II.A.2.13 Reconfigure the AFW control and isolation i
valves to add isolation valves in series with l
existing AFW control valves FV-20527 and l
Add control valves in series with existing AFW valves SFV-20577 and SFV-20578.
4 This gives for each Steam Generator redundant parallel control valves each with independent isolation.
t 5415MAJ Page 9 Rev. 2
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ECN A-5415 MAJOR NCR Work R quist 104413 Discipline I& C MOD 001 Date 10-31-86 II.A.2.13.1 Controls for air operated valves FV-20527 and FV-20528 shall be f e,om,EFIC Channel.,__, _,
'A' only, via the Class 1 electric / pneumatic converters FY-20527 and FY-20528. Both valves will be normally open, fail open on loss of signal or loss qf air. Actual valve pos'ition for both valves will
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be available in the control room on HISS.
Though normally supplied with
, regular plant instrument air, a seismic Class I (non-interrupted on loss of offsite power) source of air will power the valves for two hours if necessary.
II.A.2.13.2 Controls for D.C. powered valves Fv-20531 and Fv-20532 shall be from EFIC Channel
- B' only, via transducers FY-20531 and FY-20532 i
respectively. Actual Position indication will be indicated in the control room on HISS, Both valves will be normally open, fail open on loss of control signal or power. valve power and position indication shall be powered from the same D.C.
bus which supplies power to EFIC channel "B".
II.A.2.13.3 Controls for valves EV 20577 (formerly SFV-20577) and EV-20582 will be frem the contrcl room and EFIC channel "D".
Valves are norually closed. HV-20582 and EV-20577 will be powered by a zji non-interruptable Class 1 D.C. power source.
5415MAJ Page 10 Rev. 2
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)f ECN A-5415 MAJOR NCR WJrk Rsquset 104415 Discipline I& C mod 001 Date 10-31-86 II.A.2.13.4 Control for valves EV-20578 (formerly SFV-20578) and HV-20581 will.be.from.the---- -.
control room and EFIC channel
'C'.
Valves are normally closed. EV-20581 and EV-20578 will be powered by a non-interruptable Class 1 jjg D.C. power source.
II.A.2.14 Connections from EPIC
'A' to the shutdown panel will be made to accommodate cooldown of the plant with a fire in the control
. room. Steam generator A and B pressure indication, ful'1 ran'ge level indication and hand / auto controls for valves FV-20527, FV-20528, PV-20562A, B, C and PV-20571A, B, C will be required at the shutdown panel. Isolation switches for EFIC and EPIC controlled components will be added as appropriate for Appendix
'R' isolation of components affected by a fire in the Control Room.
II.A.2.15 Modify control of the main feedwater control and block valves to accept a priority close signal from the Trip Interface Equipment (TIE). Class 1 Solenoid valves which close FV-20525, FV-20575, FV-20526, and FV-20576 should be
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commanded from TIE cabinet H4EIA1. Valves
.m EV-20529 and EV-20530 to be powered from
/$1 energency bus S2A3 will receive close commands from TIE Cabinets 84EIA1.
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II.A.2.16 Add motor operators to valves FWS-015 and FWS-016: renumber valves to FV-205'5 and Fv-20516 respectively. These two motor operated valves should be operable from the main, control room with a priority close signal from TIE cabinet H4EIBl.
Class 1 power to these valves will, be from emergency bus S2B3.
II.A.2.17 Add a motcr operated isolation valve (EV-20521) on 10" line 20529 to isolate jgg Turbine Bypass Valves PV-20561 and FV-20563. Valve power should be class 1 backed by diesel GEA with manual open/close controls from the control room.
5415MAJ Page 11 Rev. 2 9
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FCN A-5415 MAJOR NCR Work'R'qusat 104415 Discipline I& C mod 001 Date 10-31-86 II.A.2.18 Add a motor operated isolation valve HV-20522 on line 20530 to isolate Turbine Bypass Valves PV-20564 and PV-20566.
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Valve power shall be Class 1 backed by, diesel GEB with manual open/close controls from the control room.
II.A.2.19 Add a motor operator to valve MSS-017 and valve MSS-018 to provide isolation of Atmospheric Dump valv's PV-20562A and e
PV-20571A respectively. MSS-017 and MSS-018 are to be re-tagged EV-20517 and EV-20518 respectively. Valve power shall be Class 1 with manual open/close controls
/hh from the control room. EV-20517 shall have power backed by GEA; HV-20518 by GEB.
Note: ADV manual isolation valves MSS-019, MSS-021, MSS-020, MSS-022 are to remain closed during normal operation until motor operators can be added to them.
II.A.2.20 Modify MFW pump turbine steam valve HV-20565 control to receive a priority, g
close signal on AFW initiate from TIE cabinets H4EIAl and H4EIBl.
II.A.2.21 In addition to control room indication required above, the following Class 1 indication and controls will be panel mounted in the control room:
"1 II.A.2.21.1 SG
'A' low level, wide range level and pressure from EFIC channels A and B (from cabinets H4FWA, H4FWB).
II.A.2.21.2 SG *B" low level, wide range level and pressure from EFIC channels A and B (from cabinets H4FWA, 24FWB).
II.A.2.21.3 SG "A"
AND SG *B" level selection back lighted pushbuttons plus auto selected level serpoint indication lenses.
II.A.2.21.4 AFW flow rate indication (Channel A & Channel B) for both SG
'A' and SG
'B".
Rev. 2 5415MAJ Page 12
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ECN A-5415 MAJOR NCR Work R;quact 104415 Discipline I& C MOD 001 Date.
10-31-86 II.A.2.21.5 One shutdown bypass back lighted pushbutton for each EFIC channel.
II.A.2.21.6 EFIC channel A and~ channel B manual trip / reset pushbutton matrices.
II.A.2.21.7 EFIC chapnel A and channel B AFW control valve cont'rol reset switches.
II.A.2.21.8 Open/close switches for EV-31826 and HV-31827 mounted with other AFW cont,rols.
II.A.2.21.9 Condensate Storage Tank level indication (A and B channels).
(Not required for initial operation of EFIC.)
II.A.2.22 Control Room modifications required to implement the above changes include:
II.A.2.22.1 Removal from H1'RC the ICS hand / auto control stations for FV-20527 and FV-20528.
Add EPIC hand / auto controls for FV-20527, FV-20528, FV-20531 and FV-20532 to HISS.
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II.A.2.22.2 Removal of AFW flow indication from H2PS.
II.A.2.22.3 Removal of AFW-control valve hand controllers (HC-20527 and HC-20528) from H2FS.
II.A.2.22.4 Removal of steam generator level indicator LI-20503B and LI-20504B from H2PS.
II.A.2.22.5 Replacement of ICS, ADV hand / auto control station with EFIC ADV hand / auto control station on (HlRI).
II.A.2.22.6 Removal of ICS power failure ADV failure mode select switch (HS-20562C) from HlRI.
5415MAJ Page 13 Rev. 2 O
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$h/ f ECN A-5415 MAJOR NCR Work R;qu2st 104415 Discipline I&C MOD 001
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Date 10-31-86 II.A.2.22.7 Removal of Steam Line Failure Logic ' Enable / Bypass' switches from HISS.
II.A.2.22.8 Removal of ICS AFW control override switches (HS-20527 and HS-20528) from HISS.
II.A.2.22.9 Addition of open/close hand switches'for HV-20515, HV-20516 on ElRI.
II.A.2.22.10 Changing location of op'en/close hand switches for AFW cross-tie valves HV-31826 and EV-31827 f rom H2PS to BlSS.
II.A.2.22.ll Install an extension panel on the end of panel HISS to be called ElSS (E).
This panel will contain all of the AFW controls. The layout of this panel is shown in Figure VI.
II.A.2.22.12' Add motor o(.erators to valves MSS-022, MSS-020, MSS-021 and MSS-019 to provide isolation of Atmospheric Dump Valves PV-20562B, PV-20562C, PV-205171B and PV-20571C respectively. Valve power shall be Class 1 with manual open/close controls from the control room. Not required for initial operation, see II.A.2.19.
II.A.3.
Changes which are part of Mod 1, but which are to be implemented at the next refueling following implementation of the above modifications.
II.A.3.1 Modify the NI/RPS Field Change Package FCP-3473 to include the Reactor Power vs.
MFW flow Anticipatory Reactor Trip into the RPS trip string. Modify the NI/RPS Field-Change Package FCP-3311 to delete the ? Loss of Both Main Feedwater Pumps with Reactor Power gr. eater than 20t' Anticipatory Reactor Trip. Implement these changes i'n the NI/RPS.
5415MAJ yage 14 Rev. 2
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ECN A-5415 MAJOR NCR Work Request 104415 W
Discipline I&C MOD 001 Date 10-31-86 II.A.3.2 Removal of existing BlSS console and the
' new HISS (E) extension and replacement with a new one of saine cross'seFtio-'b'ut~-- "W-n 24' longer. Console structure is Class 1 seismic. Console layout consolidates AFW controls, MFW controls, Condensate Controls, Turbine controls and Generator Controls, as required, by the Control Room Design Review. Console layout is shown on Specification N25.08 drawing sheets 1 thru -
8.
II.A.3.3 Using operating data gathered during fuel Cycle 8, develop a procedure for setting and maintaining the EFIC MFW overfill level bistable and delay setpoints.
II.A.4.
Changes of power sources to move' the AFW components to the new TDI diesel generators (GE-A2 and GE-B2) is to be complete prior to initial EFIC operation. This was decided upon after initial issue of this DBR.
Changes to reflect the move are based upon the following:
II.A.4.1 Assumptions:
II.A.4.1.1 Motor driver for P-319 and j
P-318 must be moved to new diesels.
II.A.4.1.2 P-318 to remain a "B" train component; P-319 an ' A' train component.
II.A.4.1.3 AFW flow, AFW pump pressures, and CST levels would require no changes (i.e., signal conversion cabinets B4SCA and H4SCB are powered from SlA2 and SlB2; the same 120 VAC assured sources which power EFIC "A" and
'B' respectively.)
II.A.4.1.4 AFW should be ' associated" with the new diesels (GE-A2'&
GE-B2) while HPI/LPI should l
be ' associated" with 'old diesels (GEA & GEB).
Rev. 2 5415MAJ Page 15
t EON A-5415 MAJOR' NCR Work Request 104419 Discipline I& C MOD 001 Date 10-31-8e
- II. A. 4.1. 5 Heat tracing, ~ DADS, SPDS and Y'ard Lighting ar'e not
~ ~ ' ',
considered essential to AFW- ----
operability.
II.A.4.1.6 Interaction between GEA and GE-A2.is not considered although R.G. 1.75 separation is not observed; similar for GEB and GE-B2.
II.A.4.2 Important Constraints:
II.A.4.2.1 Battery *A2' is required to load GE-A2; Battery "52' is required to load GE-B2; Battery
'A' is required to load GEA: Battery "B" is required to load GEB.
II.A.4.2.2 AFW must be available for loss of all AC power cases,
' Station Blackout" (min. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). A single turbine driven AFW pump is acceptable for this case.
II.A.4.2.3 Although EFIC channels A, B, C and D are separated electrically, tnere are two important points of
- N commonality. First, for operation of the channels longer than two hours, and concurrent loss of offsite
~
power, battery 3A2 which powers Channel A and battery BC2 which powers, Channel C are both re-charged from diesel GE-A2 while catteries BB2 and BD2 which power Channel B and D respectively are re-charged by diesel GE-B2.
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ECN A-5415 MAJOR NCR Work Request 104415 Discipline I& C MOD 001 Date 10-31-86 I
Secondly, for Appendix
'R'
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scenarios with worst case-inte ractioi 'a~isiisptiidEs f 5f- ~
~~~~ - ~~"
fires in the tank farm or Auxiliary Building, EFIC Channels A and C or B and D are assumed to be impacted.
Therefore, for certain AppendixR' scenarios the AFW control valve arrangement effectively reverts to a two channel arrangement. With such an arrangement it is possible with assumed
' single" failure to assure feed or assure isolation but not both.
II.A.4.3 For the AFW system the following power sources were agreed to.
II.A.4.3.1 Move controls and power sources for HV-20577 RV-20578 to the NSEB. These valve motor actuators should be changed to D.C. motors.
Valve EV-20577 will retain its control signal from EFIC Channel D,but power wil,1 be from battery backed panel SOD 2.
valve EV-20578 will retain its control signal from EFIC Channel C but will be powered from battery backed panel SOC 2.
II.A.4.3.2 valves HV-31826 and EV-31827 will have their controls moved to the NSEB and will be powered from GE-B2 (S2B3) and GE-A2 (S2A3) respectively.
II.A.4.3.3 Valve HV-30801 will have its controls moved to the NSEB and will be powered by battery backed panel SOB 2.
'5415MAJ Page' 17 Rev. 2 e
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1 ECN A-5415 MAJOR NCR Work Request 104415 Discipline
!&C MOD-001 Date 10-31-86 II.A.4.3.4 The motor driver for APW pump
- -~
P-318 will have its controla__...
moved to the NSE3 and will have controls powered from switchgear S4B2 and be powered from GE-B2.
II.A.4.3.5 The motor driver for AFW pump P-319 will have its controls moved to.the NSEB and will have controls powered by
.switchgear S4A2'and be powered from GE-A2.
II.A.4.3.6 The relative physical placement of AFW isolation valves EV-20578 and EV-20582 will be re-configured from
. previous concepts such that FV-20528 and Fv-20578 are in series and FV-20532 and
- EV-20582 are in series.
From II.A.4.2.3 this yields a priority " feed' mode and requires for Appendix
'R' scenarios that either the exact routing of control and power cables will show that there is not common mode fire d
damage to channels B and D or A and C, or that other operator action to stop and/or control flow are available (e.g. cycling of the AFW pump). A detailed review of cable routing will be required to determine the appropriate failure modes and/or required operator actions.
II.A.4.4 Other components associated with either
~
MOD 001 or EFIC were assigned power sources as follows:
II.A.4.4.1 Isolation valves for ADv's and TBV's will be powered from the Bruce - GM diesels.
Power ' Train" assignment matches the steam line.
Therefore, HV-20517 and RV-10521 will be powered from S2Al; HV-20518 and EV-20522 from S2Bl. This same 5415MAJ Page 18 Rev. 2
f ECN A-5415 MAJOR NCR Work Request 104415 Discipline I& C MOD 001 Date 10-31-86
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reasoning applies to HV-20569
, __. __ and Eh2D596.the.steast.
isolation valves to the AFW pump Turbine (K-30S) which will remain on their current sources.
II.A.4.4.2 MFW isolation valves HV-20515 and EV-20516 will continue to
=
be powered by GE-B2 (S2B3).
And MFW block valves HV-20529
^,
and HV-20530-will have their contro!s moved to the NSEB and be powered from GE-A2 (S2A3).
1 II.A.4.4.3 Main Steam cross-tie HV-20565 is both a steam isolation valve and a valve necessary to assure steam availability to K-308.
Since the valve is normally closed and closure is the ' safe
- position it was decided to leave the controls
- as is' with power from GEB (S2B1).
II.B. Design Basis II.B.l.
Design Basis for the AFW system (including MFW
+
I isolation following Steam Line break) is fundamentally based upon NUREG-0800 10.4.9, Standard Review Plan Auxiliary Feedwater System (PWR).
Specific exceptions to NUREG 0800 are:
II.B.l.1 The AFW system at Rancho Seco does not normally operate (provide flov to the steam generator) during start-up, hot standby and shutdown. The AFW system is
~
an emergency back-up to the Main Feedwater System, or may be used at tne control room operator's discretion. The AFW system will initiate and function in the fully automatic mode when required (see section 2.1.4.2 Actuation Requirement; System Description; Reference 28) if the pressure in both Steam Generators had been operating at 750 psig or greater. It shall be possible to initiate the AFW System manually (in the fully manual or fully automatic mode) at any time that the Decay Heat Removal (DHR) system is not the
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principal c, ore cooling system.
5415MAJ Page 19
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ECN A-5415 MAJOR NCR Work Request 104415 Discipline I& C MOD 001 Date 10-31-86 II.B.I.2 Plant cooldown from hot standby to the DHR cut-in temperature using only safety grade
' '*~~-
- equipment controlled from the control room ----
(NUREG 0800 10.4.9 I.18) assuming worst case single active failure is not a design basis for Rancho Seco. This is however a design objective (as opposed to requirement) for the EPIC system and its controls in the contrbl room.
II.B.l.3 AFW system unreliability as analyzed using methods and data presented in NUREG-0611 and NUREG-0635 may not meet the absolute requirements of NUREG 0800 II.5.C.
This is discussed in SMUD letter to NRC dated September 8, 1981, " Auxiliary Feedwater System upgrade Reliability Analysis dated April 1981 (See I.8) and NRC's SER of EFIC dated Sept. 26, 1986 (Ref. 53).
Using l gg methods appropriate to the specific design of the ' upgraded
- AFW system at Rancho Seco, B&W conclutled that unavailability of full auto initiation was 9 x 10-5 per demand. The unavailability per demand as calculated for the NRC by Brookhaven National Labs was 7.6 x 10-4.
The
~
differences are discussed in section
/hg II.B.6 (page 33) of the SER (Ref. 53).
II.B.2.
Additional Desian Bases for EFIC and EFIC Related Changes p'er MOD 1
'=
Section 2.0 System Requirements of the EFIC AFW System Description (Ref. 28) lists design bases for the AFW system.
The EFIC AFW System Description is the most recent evolution of the documentation which forms the licensing base for the EFIC concept. The EFIC AFW System Description is a direct descendent of B&W document 15-1120580. Revision 00 of that document was the basis for the NRC's positive preliminary Safety Evaluation Report of the AFW Upgrade issued to the District by letter dated January 22, 1981.
As the design of EPIC and related system upgrades were fleshed out, revision 01 of the AFW System Description (15-1120580-01) was sent to the NRC September 8, 1981 (Ref. 25).
The NRC subsequently issued on September 7, 1982, on S.E.R. against that revision of the System Description. Following issuance of that S.E.R. there was some confusion within the NRC as to whether they had acted upon sufficient informatior.,
5415MAJ page 20 Rev. 2
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ECN_ A-5415 MAJOR NCR Work Request 104415 Discipline I& C MOD 001 Date 10-31-86 So, in a letter dated December 8, 1982 (Ref. 7) the NRC requested additional EFIC and AFW information.
' ' '" ~ ~
~
Subsequent letters' f rom SMUD to the'NRC 'Tre f erences 11 ~ ~~~~~ ~ "~ -
and 2) supplied additional information and on April 7, 1983 the NRC issued an SER covering this AFW upgrade, but noted three exceptions.
On Ap'ril 28, 1983 (Ref. 6) the D,is,trict sent a revised AFW System Description to the NRC (15-1120850-03).
- That document is the last revision of the B&W System Description sent to the NRC.
On September 26, 1983 the NRC issued an SER which covered the EFIC and,related changes to the AFW
/}g system. That SER stands as the last official approval of EFIC (Ref. 53).
B&W revised document 15-1120850 one more time (Rev.4), however, the only substantial change to the document was the addition of Rancho Seco specific EFIC setpoints for S.G.
low level initiate, etc.
The EFIC AFW System Description, which is attached to and forms a part of this D.B.R., is a SMUD Nuclear Engineering revision of B&W document 15-1120580-04.
SMUD letters to the NRC dated January 17, 1986 and March 3, 1986 (References 26 and 27) briefly describe the Mod 1 (EFIC) scope particularly for fuel cycle 8.
The latter reference contains, for the NRC, a brief description of the substantive differences between the last official system description sent to them and the design as stated in the most recent EFIC AFW System Description.
Significant permanent changes to AFW upgrade design as compared to the status of design described in System Description Doc. 15 1120580-03:
A)
AFW pump P-319 will be actuated by EFIC s'anr#1 A.
Likewise AFW pump P-318 will be actuated by EFIC channel B using it's turbine driver.
~
B)
The use of time delay circuits in the initiate logic will be utilized to minimize spurious actuations. See Figure 2.2-4.
C)
The shutdown bypass permissive signal has been altered to allow bypass-when either steam generator secondary pressure is below 725 psig.
(See Figure 2.2-4).
This simplifies operator action during a tube rupture scenario.
D)
Position' indication of manual valves.in the AFW flow path is not to be provided since each is locked in its safety position during plant' operation.
5415MAJ Page 21 Rev. 2 m--.
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jig h ECN A-5415 MAJOR NCR Work Rtquest 104415 Discipline I& C MOD _,. 001 Date 10-31-86 II.B.3 Acoendix R - Components used in various appendix
'R' proceduren are impacted by this modification, this includes At_mosp,heric Dump Valves (ADV's), AFW control valves and AFW pumps.~~Also hand / auto stations to control ADV'S and APW control valves are being added to the Appendix
'R' shutdown panel to f acilitate plant control in the event of fire in the control room.
II.B.3.1 It shall be possible to electrical'ly
~
isolate any EFIC control circuit which passes through the control room such that after isolation has occurred no adverse EFIC control action can, occur due to fire in the control room. No damage to the EFIC circuitry shall occur due to fire in the control room.
9 II.B.3.2 Automatic initiation of AFW and/or automatic isolation of MFW is acceptable but not necessary to mitigate the effects of a fire in the plant. Manual initiation of the AFW via EFIC is sufficient.
II.B.3.3 Controlled AFW flow and pressure control via'the ADV's shall be possible for at least one steam generator for all Appendix
'R' scenarios.
II.B.3.4.
EFIC powered control parameters (S.G.
level, S.G. Pressu're) shall be indicated at the place where the controlling EFIC hand / auto station is located. Indication shall be from the same EFIC channel as the controls.
II.B.3.5.
Those portions of E"FIC, its controls and indication which are used to mitigate Appendix "R' scenarios need not (for Appendix
- R*
reasons) be Class I, but must operate operate with or without the availability of offsite power. Assured power sources including both battery backed electrical power and compressed air bottle back-up !?strument air must be available for at least 2 (two) hours following loss of offsite power. After two hours, local manual control is permitted.
5415MAJ Page 22 Rev. 2 w "
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ECN A-5415 MAJOR NCR Work Request 104415 Discipline I&C MOD 001 rate 10-31-86 II.B.4 EQ Requirements - Some components in Mod 001 are sub]ect to the EQ requirements of 10CFR50.,49.
The speci fic-components - and : requirements -ar e --iden ti f ied.
in the DBR's for each sub ECN.
II.B.5 ADV Controls and Isolation - The three reasons that EFIC is required to control the ADV's are:
To avoid "mid-scale failure" problems associated with loss of power to the ICS.
To provide safety grade controls and components, to assist in assuring ability to cooleown from hot shutdown to the decay heat, system cut-in, temperature.
To provide three independent turbine' bypass controls to preclude common mode failures associated with the N.I. Calibration error problem; one for *A*
steamline ADV's, one for "B" steamline ADV's, and leave the TBV's under ICS Control.
The last of these is the least obvious. It stems from a Babcock & Wilcox preliminary safety ccncern PSC 7-78 N.I. calibration error. The final report documenting specific analyses for Rancho Seco which forms the basis for having no more than 28%
bypass is reference 60.
The operating license for the Cycle 7 core limits the amount of
- turbine bypass
- to less than 28%
bypass.
That.is, the combined valves wide open (VWO) flow of all operable turbine bypass valves (TBv's) and atmospheric dump valves (ADV's) must be less than 28% of total main steam flow at 100%
power. This is currently accomplished by requiring closure of the manual isolation valves in series with at least four of the ten ADV's and TBV's. The basis for re_striction is the need to limit,inadvertant core over-power conditions due to Neutron Instrumentation errors from moderate frequency overcooling transients caused by a single control system failure. This is discussed in reference 60.
5415MAJ Page 23 Rev. 2
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ECN A-5415 MAJOR NCR Work Request 104415 Discipline I& C MOD 001.
Date 10-31-86 Unless the fuel operating license is ammended, this restriction on allowable ADV/TBV flow
~ ~ "
capabiffty wiff editfn~de "Even36dgh t'hiissdridd~" '
~ ~ '
~~
basis will no longer apply. That is, with EFIC channel A controlling three ADV's, EFIC channel B
- controlling three ADV's, and JCS controlling all four TBV's no single control circuit or shared control component can cause more than,284 bypass flow. Therefore,-although MOD 1 will implement controls to all six of the ADV's, the current restriction allowing use of only two will still apply'for the duration of cycle 7.
OL Note also, that the' companion requirement for motor operated valve's in ser'ies with the ADV's and
~
TBV's (sub-ECN AS415 AC and AD) doesn't impact this design bases. These isolation valves are for isolation of a not closed ADV (or TBV) to prevent overcooling following reactor trips.
II.C. Scope Scope of design criteria is covered in II A and II B above.
II.D. Equipment Class and Power Requirements The classification of equipment and their power sources is covered in the EFIC AFW Syctem Description section 3.
Seismic classification of AFW piping is specifically covered in a SMUD letter to the NRC dated January 14, 1983 (Ref. 2).
The design of EFIC and it's relationship to components it interface with, assumes that an emergency diesel generator and it's associated emergency bus is independent of any other. -
i.e. failure of a diesel generator and/or its bus will not cause failure of any other. Also, loss of a diesel generator cannot cause loss of a battery powered L.s for as long as the batteries are able to power the bus.
II.E. Testing II.E.1.
Surveillance Testing for safety critical systems and components is covered by the Rancho Seco Technical Specifications as amended to include the Draft Tech Specs, B&W document 05-0004 dated March 25, 1985 (Ref. II).The draft tech specs cover only the EFIC, S F'AS, and RFS.
5415MAJ Page 24 Rev. 2 9
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i ECN A-5415 MAJOR NCR Work Request 104413 Dise:pline I&C MOD 001 Date 10-31-86 II.E.2.
Start-Up Testing II.E.2.1.
Insitu testing of.the EFIC circuitry.is._,~
required to assure that the CMOS Devices have not altered since the system test was performed at the Vitro factory in February 1984. Vitro Equipment Test Procedure TP-3801-4009 functionally tests each circuit and is.available.for our use.
II.E.2.2.
Test specifications for the RPS and SFAS changes (B&W Field Changes Packages FC-3473 and PC-3478 respectively) are included in the change packages. In addition, the Reactor Power vs. MFW flow Anticipatory Reactor Trip (ART) will not be used to trip the reactor until it has been tested for a fuel cycle under normal operating conditions. (See Section III C)
II.E.2.3.
Test specifications for other specific components will be addressed by the specific Sub-ECN design package.
II.E.2.4.
A test specification for the entire upgraded AFW system including EFIC, Main Feedwater Isolation and ADV control is presented in B&W Document 62-1149372 (Ref. 10).
The Test Specification was prepared prior to the start-up of the EFIC system at Arkansas Nuclear One and i.
Crystal River 3 and therefore could not reference that operating experience. It is recommended that those portions of the test spec requiring high decay heat rates not be performed if RCS and secondary system response has been demonstrated at other B&W operating plants. Rather, the adequacy of equipment response should be verified using a test procedure adapted
~
from TP-3801-4009 (Ref. 29).
5415MAJ Page 15 Rev.'2
ECN A-5415 MAJOR NCR Work RIqu20t 104415
) )
Discipline I& C MOD 001 Date 10-31-8&
III.' Calculations and Design Information
__III.A. Design Features The design features of the EFIC AFW upgrade are covered by section 3 of the System Description (Ref. 28).
Additional specific information concerning the EFIC and T.I.E. can be obtained from the vendor drawings. A listing of the Vitro (EEIC) and Consolidated Controls Co. (T.I.E.) vendor drawings is included as reference 9.
Equipment instruction manuals are also available.
A study of the ' human factors' associated with operation of the EFIC is available; see reference 13.
III.B. Functional Description A functional description of the EFIC AFW upgrade is found in the EFIC AFW System Description (REF. 28).
III.C. Design Calculation III.C.l. AFW Flow - The minimum. required AFW flow required even under the worst case single failure assumptio.n which would limit AFW flow to the steam generators is.760 gpm total flow to the SG's within 70 seconds of loss of Main Feedwat,er (LOMF). Reference 43 summarizes the calculation input and results for the 760 gpm case with a 50 second delay. Reference 45 extends the delay to 70 seconds. Another B&W calculation (Ref. 52) shows that 560 gpm of AFW would be sufficient. However that calculation was not a " safety grade
- calculation and did not employ fully defensible calculational methods. It serves to show that the 760 gpm number acknowledged by the NRC as sufficient is not the absolute minimum required AFW flow at Rancho Seco.
References 1 and 2 give considerable detail concerning the 760 gpm analysis. The NRC SER (Ref.
- 3) page 60, 61, and 62 specifically accepts the 760 gpm at 70 seconds value.
o 5415MAJ Page 26 Rev. 2 9
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ECN A-5415 MAJOR NCR Mock Request 104415 Discipline I& C MOD 001 Date 10-31-86 A value for maximum (uncontrolled) AFW flow to a l
8L"9
- 80 *** C*1C". lated to be approximately 2130 gpm at a 3G pressure o'f TOD_psTg; see Ref. 41.
III.C.2. Condensate Storage Tank Capacity - The required APW flow calculation cited above recognizes that not only decay heat, but also neactor Coolant Pump (RCP) heat. input must be considered fer A,FW cooling requirements.
A' calculation showing the effective cooling capacity of the CST which takes this into consideration is shown in reference 44.
This shows that at a minimum initial level approximately 10.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of water are available to cooldown to the Decay Heat Removal cut-in temperature.
III.C.3. EFIC Setpoints - There are several EFIC setpoints which are of interest due to their potential impact on normal and emergency operating procedures. These include S.G. low level initiate, S.G. low pressure initiate, ADV psessure control setpoint, and APW 1evel control setpoints. The various setpoints (see Table 4.2-1 of the System Description) were calculated based on operating experience, transmitter error, special S.G f,1uid condition calculations, containment environment, etc.
The calculations which are the basis for the setpoints are collected in one B&W calculation 32-1155738-00 dated 1/22/85 (Ref. 48).
References 30 and 31 became the basis for environmental temperature range input for level transmitter and liquid filled reference legs inside containment. Reference 39 documents the accuracy of the level transmitters to be used. Reference 46 documents the basis for the "ECC Level Setpoint". Note that the level transmitter string accuracies assume that the reference legs are insulated and therefore never exceed 145"F; even during accident conditions in containment.
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Work Request __ 104415 g ff)
Discipline _ I& C MOD _
001 Da t e 10-31-86 III.C.4.
Steam Generator Level Controls - The EFIC steam generator level contgols 'have the requirement to assure auf ficient. but.notaoo-much. cooling--water flow.
This requirement must be fulfilled'with two redundant flow paths. EPIC employs for each S.G.
two independent control trains each utilizing the same but independently measured feedback parameters.
This raises several control stability questions, particularly during. natural circulation conditions in the RCS.
These questions were investigated by B&W.
Reference 40 establishes the preference for a " rate limited" level rise control and Reference 38 investigates control stability and the ability of the controls to act in the full automatic mode for at least 10 minutes for all expected decay heat scenarios without requiring operator action to keep pressurizer level within limits.
III.C.S.
Main Feedwater Overfill (MFWOF) Overfilling the Steam Generator has been a concern for several years. However since the required action to mitigate overfill produces a LOMFW event, there has been reluctance to protect against it.
Additionally, there are with the OTSG design two distinctly different " overfill' problem regimes.
The one is excessive feedwater flow at high power levels.
This regime has the potential to produce hydraulic instability in the SG's and to send wet steam to the first stage of HP turbine blades.
The other regime could occur at low power levels and would literally fill the SG's witn liquid.
a RCS overcooling and deight of water in the steam lines are concerns for this event. Reactor power l
excurcions due to moderator remperature feedback I
are of interest but of no co1cern.
B&W investigations specifically aimed at proving the EFIC MFW overfill concept are pre,sented in out reference 35, 36 and 37 Reference 36 is the confirmatory calculation showing that isolation of MFW within 15 to 30 seconds is sufficient to prevent excessive moisture from carrying over into the steam lines. Reference 35 and 37 were investigations into more elaborate MFWOF detection schemes than that used by EPIC.
Their so called
" variable overfill limits
- were rejected as ineffectual or not cost effective.
5415MAJ Page 28 Rev. 2
.A-5415 MAJOR NCR Work RhquOst 104415 th Discipline I&C MOD 001 Date 10-31-86 The original EFIC concept included MFW and AFW overfill isolation logic. However*that design utilized.A. common _ bis. table. in, each. EFIC channel en._.
' decide
Even though the particular failure would have required four independent failures (or eight bistables to be grossly mis-calibrated by a single technician) the
' commonality
- aspect led to deletion of the AFW overfill portion of the designs; See also PSS of the NRC EFIC SER, Ref. 3.
Having no automatic protection for AFW overfill is justified since it is a slow developing transient
~
which would allow time for operator intervention.
The operator will still be alerted to an overfill via the MFWOF circuits, and controls and indications are available to the operator to isolate AFW from the control room.
III.C.6.
LOMFW Anticinatory Trip - As noted in several submittals to the NRC (e.g. References 6 and 24) the District intends to install as its response to NUREG 0737 II.K.2.10 an Anticipatory Reactor Trip (ART) based on a comparison of reactor power and i
MFW flow. This is described as the Reactor Power vs. MFW Flow ART's, the Flux /MFW flow ART's, or the 0/MFW Flow ART's.
The Trip is developed in the RPS in a comparitor module which compares measured i
reactor flux to total measured MFW flow.
Since MFW flow tends to be'a fairly noisy signal,
,4 there was initial concern that this scheme might lead to spurious trips of the reactor. Therefore,
' operability" evaluations were undertaken to collect data from Rancho Seco and other operating plants to. assure that spurious trips would not be a problem. References 21 and 23 transmit B&W studies 51-1135242-01 and 51-1141558-00 respectively.
These studies show that spurious trips should not be ex'pected. Reference 18, 19, 20 and 22 transmitted Rancho Seco op'etating data to 'B&W as a partial basis for the studies.
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.ECN A-5415 MAJOR NCR Work Request 104415
, Discipline I&C MOD 001 Date 10-31-86 Despite the assurances of the existing studies, it seems prudent to collect some operating data with the new hardware prior to-its-use er reactor trip. In the interim the existing LOMFW ART based upon MFWP Turbine EHC pressure, will continue.
However, the Flux /MFW flow trip may be used to tell EFIC to start AFW.
This is acceptable since based upon the considerable existing data we expect there to be no problem, and starting-the APW will not produce flow to the SG's as long as suf'ficient inventory actually exists.
III.C.7.
AFW Pump Runout - EFIC and the upgraded APW system nave provision for automatically isolating APW to a depressurized steam generator. However, various probable scenarios can have one or both APW pumps
~
pumping for a short time to depressurized S.G.'s.
A study performed by Emergency Research &
Consultants Corp. (Report No. ERCO-693 dated February 16, 1984) shows that the Rancho Sec5 AFW pumps could be ope ~ rated in the full ' runout" condition for at least 30 minutes without destruction (Reference 12).
III.C.8.
EPIC AFW Power Sources - The EFIC system initiates AFW via two 100% trains of AFW.
That is, either train is sufficient to supply the required AFW to both steam generators. The A channel of EFIC starts P-319 and controls flow to both SG's through FV-20527 and FV-20528. The B channel starts"P-318 by opening valve EV-30801 and controls flow to both
-A SG's through FV-20531 and,FV-20532.
Either channel is sufficient to assure sufficient AFW flow, but care must be taken to assure that the failure of a common electrical source will not prevent AFW from both trains from reaching at least one SG.
Another EFIC f unction is to isolate AFW to a steam generator which has experienced a major steam line failure. This is accomplished using four channel logic which acts directly on the AFW isolation and control valves. As with the AFW initiation and control function, care must be taken to assure that a common electrical power source doesn't prevent required AFW isolation. Additionally, care must be taken that s. ingle electrical power failures common to more than one piece of equipment cannot prevent flow or isolation when required.
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$0-31-86 This theme is addressed in a study (Ref. 25) which shows that with power and logic channelized per the AFW system description, single. failure will netthec..__...... _..
prevent feed nor isolation of AFW flow. In mid-September 1986, a decision was made to shif t all of the AFW electrical loads to the new jg Transamerica De Laval diesel generators (GEA2 and GEB2) prior to initial EFIC startup. The power source study was updated to reflect the new sources and found to be acceptable (Ref. 59).
III.C.9.
Opqtade APW Refiability - There are several sources of information concerning reliability analysis for the Rancho Seco AFW system. The sources which figure prominently in the licensing of the upgraded APW system are:
A generic AFW reliability analysis f'or all B&W plants: B AW 1584, Dec. 1979.
This was used in early evaluations of the AFW system, including NUREG 0667. It also fccmed the analytical basis for requiring auto loading of P-319 on the emergency bus following LOOP.
A Rancho Seco specific AFW Reliability Analysis i
performed by B&W (Ref. 33).
This analysis was sent to the NRC in fulfillment of NURG 0737 II.E.1.1. (Ref. 26).
It assumed EFIC and an AFW configuration like that which we will nave after completion of MOD 1.,
A Rancho Seco specific AFW Reliability Analysis was performed by Brookhaven National Labs under contract to the NRC.
This Analysis forms the analytical basis for the EFIC S.E.R. issued.in April 1983, and is discussed and compared in the SER (Ref. 3) to both the B&W prepare analysis (Ref. 26) and the requirements of the Standard Review Plan (Ref. 4).
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ECN A-5415 MAJOR NCR Work Request 104415 Discipline I& C MOD 001 Date 10-31-86 A reliability and fault tree analysis for the EFIC system above was produced by Vitro, the
. manufacturer (Ref. 34).
This is a very detailed and conservative analys.t.s showing Abs........__......
relationship and dependability of the electronic circuits within EFIC. However, it measures success in a narrow fashion concerned only with the electronics. Therefore, its use in measuring broader mission success (e.g. not boiling the core) requires additional analytical input.
An additional reliability analysis which may be of benefit to the District but which is not a part of the licensing base for MOD 1 is a set of analyses which investigates various power sources in combination wi.th a third AFW pump.
(See B&W analysis BAW-1722 dated March 1982.)
A good summary of the various reliability analysis and their worth is contained in a B&W letter to SMUD dated July 9, 1984 (Ref. 5).
III.C.10."Helba & Missile Studies - The NRC AFW SER dated April 7, 1983 (Ref. 3) notes that at that time three unresolved items remained: AFW internal missiles study, CST protection and an AFW HELBA' analysis. These open items were resolved by References 53, 54 and 57.
An additional. HELBA analysis which looked
~
specifically at the AFW upgrade piping changes was produced by Bechtel (Bechtel Calc M21.30-363, dated 6-14-83; Ref. 32).
Hazards reviews of each sub-ECN submitted under this major will also be performed.
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III.C.ll. EPIC Shutdown Bypass - There are four conditions which can,cause EFIC to auto initiate AFW and/or isolate MFW which need to be manually-bypassed for..
a normal plant shutdown condition. Bypassing a safety system on a system level requires a coincident permissive signal (See IEEE-279).
In the original EFIC hardware specification two of' the conditions (low SG pressure, and loss of all 4 RCP's) were bypassed on a system basis, but required individual manual action locally at the four EFIC cabinets at different times during cooldown. The other two (low level in a SG, and high level in a SG) were to be handled procedurally by pulling breakers on the AFW pumps, etc.
These shutdown bypassing methods were not acceptable to the District due to complexity and remote location of the EFIC cabinets.
A shutdown bypassing concept acceptable to B&W and the District was incorporated into the EFIC hardware.
The* concept uses a single bypass permissive (pressure in the Steam Generator less than 725 psig), and can be~ enacted from the control room by pushbuttons on the BlSS console. The correspondence which documents the design is found u
in Reference 14, 15, 16, 17 and 49.
II.C.12 Sincle Failure Analysis - An analysis was performed by Vitto (Reference 55), in accordance with IEEE-279-1971 and I,EEE-379-1977, that shows that i
the EFIC System meets the single failure criterion.
IV.
FAILURE MODES Many of the operating modes of EFIC are discussed in detail in section 6.0 of the EFIC System Description (ref. 28).
The plant casualty events discussed include the following: Loss of MFW ( LM FW ),
LMFW with loss of offsite AC power, LMFW with loss of onsite and offsite AC power, plant cooldown requiring AFW, turbine trip with and without bypass, main feedline break, main steamline break /AFW line break, small break LOCA, fire outside control room, fire in the control room. In addition to these, various ErIC system f ailures are discussed in the following paragraphs.
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APW Valve Failure Since each OTSG is supplied auxiliary feedwater from a line which is controlled by a parallel combination of series sets of valves, there is no single failure which will prevent the isolation of or the feeding of the appropriate OTSG. Each of l
the four valves is powered and controlled by a separate EFIC Channel and power source. Consequently, a single channel failure will only cause the failure of one valve of the pacallel combinations, (i.e. one valve per OTSG). Each series set of valves is comprised of an isolation valve and a control valve. The control valves fail open on a loss of power or signal. 'The isolation valves, being motor operated, fail as-is.
Since only one valve in the parallel combination of the series set of valves fails, each OTSG can be either 3
fed or isolated. See also Section II.A.4.2.3 and II.A.4.3.6. l/Z\\
For failure modes of the individual valves see the DBR of the appropriate sub-ECN.
Note that the use of normally open, fail open control valves with normally clcsed isolation valves,. requires that at least one and sonetimes two specific valves operate to close off excess flow for failures of moderate probability (e.g.
loss of a single battery set such as S082). Additionally, the isolation typically requiries operator recognition of the I
problem and subsequent action. Having the isolation valves normally closed virtually requires battery backed, DC motor
/js
~--
operators, since they must operate during the loss of "all AC' power case. If they were AC powered, a potential ecmmon mode failure is also present. i.e. failure of an
'A' or "B" battery would fail open the associated control valves and loading of the associated diesel generator bus would also be prevented. Thus, the isolation valve might not be operable off of the diesel backed AC bus.
All of the above withstanding, the proposed design is, none the less, very acceptable.
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Failure of Fiberootic Cables Between Channels The fiberoptic..costmunication_#ptwe.e.n J:F,Iq,, channels is,
designed such that a single failure (such as a loss of all fiberoptic cables going into one EFIC cabinet) shall not result in a failure of EFIC functions to actuate when needed. A single event, however, can cause an inadvertent actuation of either AFW initiation or MFW Isolation if the ev'ent affects more than one channel cf initiation logic.
Initiation of AFW is of little consequence because it will only supply water to the steam generators if the levels are low and they need water. Isolation of MFW is more serious and can lead to a plant trip, but this condition is acceptable because AFW is available to cool the plant.
To meet this criteria interchannel fiberoptic cables do not require the same kind of separation that would be required of electrical cables. The difference is that the cables entering a common EFIC cabinet do not require separation from each other even though they belong to different (A, B, C and D) safety channels. There are two reasons why this is so.
The first reason is that fiberoptic cables, unlike electrical cables, cannot propagate energy along the cable in large enough quantity to damage adjacent cables. Consider a cable running between the A and the B cabinets. A disturbance to the cable in Cabinet A can in no way damage an adjacent cable in Cabinet B.
The other reason why the fiberoptic cablee that enter a common EFIC cabinet do not require separation,is that the communication system is designed so that a failure of any cable can only cause that signal to revert to an actuated state. In all cases this is the safe state which is shown in the following list of fiberoptic cable types:
1.
Test Results These cables are used to indicate at the other three channels that a test is in progress on either 'A or B channel. The purpose of this indication is to warn technician / operators at the EPIC cabinets that testing l
is in progress. A failure of these cables will cause the test confirm (i.e., test in progress) light to turn i
on.
This is the safe condition.
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Channel Bypass Tnese-cables transmit--information-used to prevent-more...-.
than one channel of EFIC being placed in maintenance bypass and to prevent other channels of EFIC being placed in maintenance bypass when an RPS channel is in bypass. A failure of these cables will both prevent placing of~ channel in maintenance bypass and will take
- out of maintenance bypass any channel that was already in bypass. The latter of these can cause a half trip of the EFW initiation and/or the Generator-A and/or B MFW solation if work in another EPIC cabinet were in progress.
3.
Vector Enable and Control Enable These cables are used to " enable
- or start the vector and control modules when EFW initiation occurs, and allows automatic level control to begin. When the cables are damaged the associate control and/or vector modules are enabled. This.is the safe condition since these modules are dormant during normal plant operations only to decrease valve stroking.
4.
Trip Inputs
(
These cables transmit the outputs of the initiate modules from the four channels to the trip modules in Channels A and B to give the four channel initiation of trips. A failure in these cables can cause a half or a full trip of the EFW initiation and/or the Generator A 3
and/or B MFW Isolation.
IE Information Notice 86-15 addressed the failare of some fiberoptic cables due to Radio Frequency Interference. The memo f rom Daniels to Lewis -(Reference
- 56) explains why this should not adversely affect the EFIC fiberoptic cables and states that EFIC Will be tested for this problem during start-up testing.
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.Each channel of.EEIC. receives _an. actuation. signal.from.the corresponding channel of RPS for the condition when both MFW pumps tripped at greater than 20 percent reactor power. The function of this signal is to initiate AFW.
Each channel also receives four Reactor Coolant Pump (RCP) trip signals from the corresponding channel of RPS.
when all four RCP's have cripped, EFIC initiates AFW and the channels A and B control modules raises their steam generator level setpoints to a level appropriate for natural. circulation. In-both of these cases, EPIC looks at the inputs as four channel and actuates based on one out of two taken twice logic.- Because of this, any single failure of the inputs to one EFIC channel will not prevent actuation of EFIC functions nor will it, by itself, cause inadvertent actuation.
~
Each channel of EF*C also receives a channel bypass signal from the corresponding channel of RPS.
The function of this signal is to prevent other channels of EFIC from being put into bypass when a channel of RPS is in bypass. A failure of this signal to actuate could allow one channel of EFIC to be put into bypass while a different channel of RPS was in bypass. In this condition it is possible to have two channels of RPS giving the signals for an EFIC initiation with both being bypassed. This does not prevent EFIC initiation, though, because there are two other channels of RPS that can initiate EFIC.
Also, this event is extremely unlikely because the bypass of both RPS and EFIC channels will be under strict administrative control.
4 A failure of the channel bypass causing bypass actuation will simply prevent other channels of EFIC from being placed in bypass or if any channel was in bypass it will take it out of bypass. The latter of these can cause a half trip of AFW initiation or a half trip of the generator A or B MFW Isolation. Also if the operator persisted in testing the bypassed RPS channel with the EFIC half trip
/hi condition and the corresponding EFIC channel not bypassed, a full EFIC AFW initiation could occur.
~
The failure modes of the RPS inputs to EFIC are listed below:
1.
Loss of Power - A loss of power in one channel of RPS will force the MFW pumps tripped and the RCP's tripped inputs to the actuated state and will force the channel bypass to the not bypassed state. A loss of the EFIC power used to sense the contacts in RPS will cause the MFW Pump Trip to appear actuated and the channel bypass to appear -bypassed but will have no affect on the RCP's tripped signals.
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toss of Signal - a loss of signal will cause the MFW pumps tripped and the RCP's tripped signals to appear actuated. A loss of signal will cause the channel bypass to appear to be in the bypassed state.
IV.D.
Failure of SFAS Input to EPIC The SFAS inputs to EFIC are designed so that a single
^
failure will not stop EFIC from initiating AFW when SFAS actuates. There are channel A and channel B initiate signals sent to EFIC from SFAS, two signals per channel; each delivering a half trip to the AFW trip module. Either channel will initiate EFIC if both of its signals are actuated (i.e. two out of two taken once). In each channel of SFAS there are two unit modules that supply the inputs to the corresponding channel of EFIC. These inputs are open contact to trip EFIC and energize to open the contacts in SFAS.
The following paragraphs summarize the failure modes:
1.
Loss of Power -
A loss of power in a SFAS channel will prevent that channel from initiating the corresponding channel of EFIC but will have no affect on the other channel. A loss of power in an EFIC channel will prevent that channel of EFIC from initiating.
2.
Signal Failure -
b~
A loss of the closed contact signal integrity in one input to EFIC will cause a half trip of EFIC and a loss of both signals in one channel will cause a full trip. Similarly, a withdrawal of a unit module from SFAS will cause a half trip in its corresponding input to EFIC.
l Since the SFAS signal is,an energize to actuate a failure of an SFAS module to actuate will result in a f ailure of receiving the half trip input in Ef!C.
i i
l l
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__ Work Request 104415 Discipline I E 'C MOD 001 Date 10-31-86 IV.E.
Failure of EFIC Trio Interface Ecuipment EFIC actuates various components through the trip outputs of the train A and B Trip Interface Equipment (TIE) cabinets. These components actuate in response to an A'FW Initiation or a steam generator A or B MFW Isolation. The outputs of train A are redundant to train B; therefore, a single event will not cause the failure of a required
. actuation. A loss of power to the TIE cannot prevent availability of the 1E output in that train. However, loss of power to.the 1E to non-lE TIE will prevent output of the non-1E circuit but cannot affect the lE portion of the signal. A loss of signal between the TIE and EFIC or between the TIE and a device will prevent actuation of the device.
IV.F.
Power Sources Failures for EFIC and EFIC Related Hardware As addressed in Section III.C.8. and in Reference 25 and l[hg 59, a single failure of an electrical power source will not prevent controlled feeding of AFW to either steam generator not prevent isolation of AFW to a steam generator.
Discussed here are some specific failures and their
~
effects. All of these failures assume concurrent loss of offsite power.
IV.F.1 Failure of Diesel Generator GEA or CEB No AFW components are powered from these diesel generators. However, mainsteam system branch isolation valves are. Without GES the normally closed EV-20565 would fail in its last position.
If closed, its EPIC function would be correct.
If l
open, and a major steam leak were occurring, both main steam lines would de-pressurize., In this event, P-318 would not function using the turbine driver. EFIC would feed both SG's.
To avoid overcooling, operator action would be required either to close HV-20560 or to regulate AFW flow l
manually.
I IV.F.2 Failure of Diesel Generator GE A2 i
Without GEA2 power the AFW pump P-319 would not l
operate. P-318 is sufficient for all cooling requirements and would be available in either its turbine or motor driven form.
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Without GEA2 power MFW block valves EV-20529 and EV-20530 would not function..The EFIC MFW isolation function would still be assured by RV-20515 and HV-20516.
The above scenaries hold for a minimum of two Zbi hours.
~ IV. F. 3.
Failure of Diesel Generator GEB2 Without GEB2 the motor driver for P-318 will not be available. P-318 might still be functional on its turbine driver, and P-319 would still be functional.
Without GEB2, EV-20515 and HV-20516 would fail in their last position. However, without offsite power, the condensate pumps fail and flow through these valves will not occur.
IV.F.4.
Failure of EFIC and AFW indication in the Control Room.
Power for control circuits and for backlighting of
~
pushbuttons which control EFIC comes from the EFIC channel affected e.g., if power to EFIC Channel
- A*
is lost, the ' A' channel EFIC control circuits on HISS will go dark and controls will be non-functional.
Che Class I analog indication on HISS requires two inputs to be functional signal and power. If the signal is lost, the display will go 'off scale low'.
Chat is, the digital readout will be at it's lowest possible value, and the bargraph will flash a single LED in the lowest position. If the 120 VAC power is lost the indicator will go dark.
-Power to the channel
'A' indicators is from the same battery backed inverted power which powers EFIC Channel
'A*.
Power to t,he Channel
'B' indicators is from the same battery backed inverted power which powers EFIC Channel
'B'.
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!&C MOD 001 Date 10-31-86 Since all Class I indications except AFW pump discharge pressure have redundant indicators of a
~
'different channel, the only process indication
~
lost on loss of a single power source woula be one '
of the pump discharge pressures. Control lights to the back lighted pushbuttons and the ammeter would be back-up indication showing pump operation.
IV.G.
EFIC Control Failure The EFIC has six points of process control; two AFW flow control and one main steam bleed off control per steam generator. ~ These six points are controlled by process control circuits within the A and B EFIC channels which send signals to the modulating control valves on the AFW and Main Steam Systems. The function, logic, and control setpoints are discussed at length in the EFIC Auxiliary Feedwater System Description.
The f ailure of any process control, by its nature, will cause the process variable to move away from its desired value. Failure of EFIC controls would cause the controlled variable (s) to move away from the desired value(s)' causing process changes which ultimately shift the point of control. Preciseness, responsiveness, and stability of control are kinds of control failures. They are addressed in Sections III.C.3 and III.C.4.
What we are talking about here is gross control failure which will cause the swiftest shift in control point (i.e., what happens if a single event causes valve (s) to fail open or closed).
It should be noted that for the failures below, the rate of change of RCS and Secondary System parameters is not different than would be expected for similar control failures to the existing AFW and ADV controls.
IV.G.1 Atmospheric Dump Valve Fails Closed If not controlled by Turbine Bypass Valves main steam pressure would rise.in the worst case and be controlled by the main steam relief valves. In the event the other steam generator was available, and has pressure control, RCS cooling would proceed through it and steam pressure in the impacted generator would follow saturation pressure consistent with RCS hot leg temperature.
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001 Date 10-31-86 IV.G.2 Atmospheric Dump Valve f ails open The energy release would cause main steam pressure to decrease"vith a r'isult'ing decriase~in~S.G.~ secondary tempe r a tur e.
The RCS would in turn decrease in temperature. Operator action to isolate the open ADV(s) using the motor operated ADV isolation valves is the best response. However if S.G. pressure drops below 600 psig, EFIC will isolate,MrW and AFW to the
- affected generator.
IV.G.3 ArW valve Fails Closed If an AFW control valve fails closed, the process control point would shift rapidly to the parallel control valve.
IV.G.4 2.rW valve Fails Open The energy required to heat the cold AFW to saturation will cause a temperature and subsequent pressure decrease in the steam generator. Operator action to isolate the open AFW valve using the series aligned motor operated isolation ~ valve is the best operator response. Actual valve position indication is available to identify the errant valve s
If only one S.G. is, impacted,.the'EFIC will automatically isolate AFW to that S.G. if pressure drops below 600 psig. In the event that the excess ATW develops to an cVerfill condition,,the MFW overfill protection and annunciation would alert the operator to the need to isolate the errant valve.
Note, however, that for initial operation of EFIC the MrW overfill setpoint will be at its uppermost setpoint in ord'er to avoid spurious MFW closure.
IV.G.5 Single EFIC Control Failures The four bounding EFIC control failures are: loss of power to EFIC "A' or
'B' channel, loss of a control module within EFIC ' A' or
'B' Channel, f ailure of a pressure or level sensing circuit, wacko signal to a single device. A single failure cannot simultaneously cause failure of control signals from both Channel
'A' and Channel
'B',
and control failures for either channel would be sim,ilar. Therefore, only failures of Channel
'A' will be discussed below.
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'A' would cause the ADV(s) on one main steamline to fail closed, and one AFW co.7 trol valve to each S.G. to f ail open. During normal-plant operation no change in operation would result.
If AFW has been initiated, the ADV closure would play a minor roll initially as cooling from excess AFW flow would eventually dominate secondary pressure. Manual closure of the series AFW isolation valves is required. Following re-pressurization, the failure of the ADV(s) will become apparent and the course of action is as described in IV.G.I.
Loss of one of the two control modules within EFIC Channel
'A' will cause either a control valve to the "B" S.G. to fail open (see IV.G.4.) or a control valve and the ADV (s) of the
'A' S.G.
to f ail open and closed respectively. This latter failure becomes a subset of loss of channel power.
Failure of a pressure sensor signal, though possible in either direction, would be expected to fail low.
This would cause the ADV(s) on one S.G. to fail closed and one AFW valve on the same S.C. to. fail closed (due to F.O.G.G. logic). Manual control of both valves, through EFIC, would still be possible.
Failure of.a low range level sensor, though possible in either direction, would be expected to fail low.
If it failed low, and AFW has been initiated, one AFW control valve would fail open (see.IV, G.4).
If it failed high one AFW control valve would fail closed
'd (see IV.G.3).
Failure of a wide range level sensor, though possible in either direction, would be expected to fail low.
This would lead to like scenarios for a failed low level sensor, but only if all RCP's were not running.
A failed signal to a single controlled component could cause a valve to open or close. Those events are described in IV.G.1 to IV.G.4.
However, due to the nature of the 4-20ma control circuits used, failures which produce 4 ma or less are the expected failure modes. Therefore, the expected fail state for a single component would be closed for an ADV(s) or open for an AFW control valve.
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ECN A-5415 MAJOR NCR Work Request 104415 Discipline I& C MOD 001 Date 10-31-86 This control fai:
- e discussion assumes loss of only one sens.or or loss of power to all sensors in one channel. EFIC sensors may share process sensing lines g
with other EFIC channel sensors or other system M
sensors depending upon specific sensor installation s
details. If more than one EFIC sensing parameter can be effected by a single sensing line failure, those i
failures must be analyzed as a part of the sub-ECN,
)'
and the acceptability of installation detail judged s
therein. See sub-ECN's $415A, 54158, and 5415C. The relationship of shared sensing lines to IEEE-279 is discussed in Ref. 61.
VI. Special Operatino Requirements
(
VI.A.
Operating Description of EFIC Controlled Devices VI.A.l.
EFIC indicators - EPIC control room indicators display a comprehensive set of information on the status and condition of the steam generators and the Auxiliary s
Feedwater System. During the following discussion of' these, refer to Figure VI and Table VI for panel HISS.
VI.A.l.l. For both steam generators A and B there are redundant (channels A&B) meter indications of low range level (items 2&3 and 7&8), high range l'evel (items 4&5 and 9&l0), pressure (items 27628 and 32&33) and AFW flow to the steam generators
,s (items 29&30 and 34&35).
O VI.A.l.2. The Hand / Auto (H/A) stations for 'the four' flow s
control valves, TV-20527,28,31 and 32, each.
- contain a meter that indicates the position demand signal for its associated valve (items 85,87,89 and 91).
Also actual position
\\
indications for each valve (items 84,86,88 and
- 90) are located directly above the associated H/A station.
s VI.A.1.3. The two H/A stations for ADV control (located on HlRI not shown on Figure VI) contain a meter that indicates the position demand for the ADV's.
Actual position indications for these valves are not a part of MOD 1.
\\
VI. A.'l. 4. For the AFW pumps there are meter indications f discharge pressure for each pump (items 119 &
124). Also there is a digital readout of the AFW test line flow and a meter indication of the flow valve's actual position (items 121 and 148).
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VI.A.l.5. The ' Remote / Manual / Reset
- pushbutton matrices which are. mounted on the control room consoles y:
(see items 6 and 31 of figure VI of section VI.A.) provide channelized indication of automatic actuation of EFIC functions. They also
\\i provide capability of operator manual actuation of some EFIC. functions. To assure that manual
, i
' action at these two matrices could not cause all feedwater to be inadvertently isolated, nor that
'p an overfill situation could be inadvertantly caused, a study was performed. It looked at all possible manual actions at the matrices, without regard to reason for the action, and combined those actions with all automatic EFIC functions which might affect or be affected by that action. The results show that following EFIC 2hi actuation no action or combinations of actions at the matrices will prevent all feedwater from reaching the steam generator. Also no overfill can be similarly automatically or manually commanded.
It is theoretically possible for an operator to inhibit automatic actuatien of AFW or MFW isolation. This requires simultaneously pressing at least two buttons which are spacially separated by at least eight inches. This
' correct' combination of buttons must be depressed prior to re eipt of a valid actuation and held continuously in order to inhibit s
actuation. It is also possible for the operate r to place the AFW initiate in manual af ter AFW initiation. This would not prevent AFW initiation, but it w(uld prevent subsequent " feed only good generator
- logic from isolating AFW to a depressurized Steam Generator. This selection of the manusi mode is indicated at the matrix and
{
does not prsvent the operator from manually L
isolating AFW to the af fected generator. Any manual bypass or manual inhibit of an initiate s
-~
can be immediately reversed by operator selection
't uf the actuation mode.
VI. A.1.6. In addition to analog indications, backlighted pushbuttons indicate status of EFIC initiation / manual operatien, channel bypass, control enable and the steam generator level setpoints selected. Also backlighted pushbuttons i
indicate the status or position of the AFW pumps, isolation valves and crosstie valves.
5415MAJ Page 45_
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ECN A-6415 MAJOR NCR Work R3 quest 104415 Discipline I & 'C MOD 001 Date 10-31-86 VI.A.2.
AFW control without auto Initiation of EFIC - Withedt'
~
s auto initiation of EFIC*all of the AFW contro.ls,,can' be directly operated manually.
VI.A.2.1.
AFW pumps can be started or stopped manually (items 117, 118 and 123,).
Also the AFW test flow valve can be manually operated (item 120) to. increase'or decrease test flow to the condenser.
- VI. A.2.2. " AFW flow to the steam generators can be started manually by starting an AFW pump, opening the isolation valves 'litem 56, 57, 62 and 63), placing the H/A stations for the flow control valves (item 85, 87, 89 and 91) in manual and running the valves open to obtain the desired flow.
VI.A.2.3.*
AFW flow to the steam generators can also be obtained by manually initiating EFIC channels A and/or B (items 6 and 31).
This causes starting of the AFW pumps, opening of isolation valves and initiation of automatic level control in the steam generators. In this state an operator can place the control valve H/A's,in manual if desired to manually regulat.e the AFW flow;'
l to the ste,am generators. <
VI.A.3 AFW control with Auto Initiation"of EFIC - With an 4
automatic initiation of EFIC the AFW and related controls will be automatically positioned to supply AFW regardless of their manual position at the time VI.A.3.1.
When EFIC initiates, the follo' wing equipment is automatically positioned with override of their manual controls: AFW pumps start, AFW isolation valves open, AFW test valve closes, Main team cross tie valve, EV-20565, closes. In order for i
these to be manually controlled, except -
the AFW isolation valvts, EFIC must first I
be placed in manual or if the EFIC l
initiating signals have cleared then EFIC must be reset (items 6 and 31).
S
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Discipline I&C MOD 001 Date 10-31-86 VI.A.3.1 1 The AFW isolation valves can be manually controlled while I
[
EFIC is initiated by first pressing the ove"cride button
^
(combined with the open and close button, items 56, 57, 62, and 63) and then pressing the open or close button to position the valve. When EFIC is reset or placed in manual 'the override automatically resets.
1 VI.A.3.2 When EFIC initiates AFW, the AFW flow control valves begin to automatically control the level in the steam i
generators. At any time H/A stations for these valves can be put in manual and the valves manually positioned. When EPIC is placed in manual or reset these valves continue to automatically control steam
/
generator level until the control' enable switches are reset (items 59 and 61).
VI.A.3.3.
Vector logic (Feed only good generator) can cause isolation of AFW to either, but not both, steam generators by closing the appropriate isolation valves. These i
valves then can be manually opened if desired as described in Section VI.A.3.1.1.
l VI.A.4.
EFIC isolation of the Main Feedwater (MFW) - The MFW i
flow can be shut off to the steam generators by EFIC, either automatically or manually.
VI.A.4.1 The MFW isolation, valves can be controlled manually to isolate oc allow MFW flow to the steam generators. These controls are overridden by the. actuation of the EFIC
.i Main Feed Water Isolation (M FWI) for i
either steam generator which causes the valves for the operator to close. MFWI l'
can also be manually initiated (item 6 and'
- 31) for either steam generator causing an 4-isolation of MFW to that generator, i
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VI.A.4.2.
An actuation of MFWI causes the MFW isolation valves t.o close, as stated above, and also overrides the ICS' control of the MFW flow control valves causing the valves to shut. These valves cannot be operated in manual until MrWI is placed in manual or reset (item 6 and 31).
When MFWI is placed in mangal or reset the isolation valves will remain closed until manually opened. The MFW flow valves will revert back to ICS control.
VI.A.4.3.
When Main Steam Line Isolation (MSLI) is -
actuated its only function is to actuate the same devices as MFWI. The same functions of manual and reset will apply to MSLI as are described above for MFWI.
VI,. A. 4.
EFIC Control of Atmospheric Dump Valves (ADV's)
Automatic control of the ADV's is always active in EFIC regardless of whether EFIC is initiated or not.
The ADV's can be placed,in manual control at any time from the H/A stations located on panel HlRI and manually positioned.
VII. VERIFICATION CRITERIA:
See VIII. A below.
VIII. COMMENTS:
.:=
VIII.A.
Design verification The functional design of the Emergency Feedwater Initiation and Control System was developed by Babcock &
Wilcox. As a part of the design process, B&W included a formal independent verification of EFIC and its l
relationship to the Rancho Seco upgraded AFW system. At B&W the independent verification is performed and documented by a Design Review Board. The members of that board are selected based upon their echnical background, their germain experience and their lack of prior involvement in the task. The positive findings of the Design Review Board are documented in references 8 and 62.
/hL 5415MAJ Page 48 Rev. 2
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VIII.B.
Dif ferences Between the Rancho Seco and SR-3, ANO-1 EFIC's Though known by several dif ferent names, all B&W Nuclear Plants have Emergency Feedwater Initiation and Control j
Systems which are independent from the ICS.
The form and features of these systems differ as required for site
,b specific reasons. Three of the utilities opetating B&W plants have system designs of their own or an A/E's design (Toledo Edison, GPU Nuclear, and Duke Power). The B&W i
late model (205 FA) plants have systeds designed into the safety grade plant protection systems delivered with,the plant (Scpply System, Bellefont, and Muelheim -Kaerlich).
The remaining three operating B&W 177 FA plants purchased the *EFIC* design from B&W (Crystal River-3, Arkansas r
Nuclear One, and Rancho Seco).
Each of the three EPIC plants uses the same basic concept. Hard. ware for the three plants was purchased under a generic specification from the Vitro Corporation.
However, aome differences in hardware and its application do exist..The specific differences between the EFIC as it will be utilized at Rancho Seco (RS) and the E?IC l
installation at one or both of the other EFIC plants are listed below:
B.l.
R.S. uses a single remote shutdown bypassing feature which will allow bypassing of EFIC i
initiation of the following: low SG 1evel, low SG pressure, high-high SG level loss of all 4 RCP's.
i-This is only possible if SG pressure is less than 725 psig.
I ANO bypasses the low S.G. pressure and loss of all 4 RCP's locally (individually) at the EFIC cabinets. They don't " bypass
- the low and
],
high-high initiates. They prevent operation by pulling the trip breakers located in EFIC cabinets
]
A and B.
CR-3 has a remote shutdown bypass for the low S.G.
pressure only. Bypassing of the loss of all 4 RCP's is done locally at the cabinets. Bypassing of low S.G.. level and high-high S.G.
level are not provided, but (like ANO) are blocked by pulling breakers in the EFIC A&B cabinets.
1 3
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B.2.
R.S. ' utilizes a shutdown bypass permissive such that shutdown bypassing is possible if either S.G.
pressure goes below and, stays below 725 psig. ANO
~
'~~
will use a similar permissive for bypassing their~
l /hs low S.G. pressure initiate.
CR-3 uses S.G.
pressure less than 725 psig in both S.G.'s.
CR-3 and ANO use a shutdown bypass permissive for loss of all four (4) RCP's based on reactor power
.being less than 104.
B.3.
R.S.'s EFIC uses a latch-in feature for the control
~ enable circuits such that placing the EFIC Trip circuits in manual will not drop out the automatic level controls. This, simplifies the operating procedures by not requiring manual control of AFW valves prior to placing EPIC in manual. Conversely it still allows the operator to take manual control of the control valves at anytime.
CR-3 and ANO do not at this time have this latch-in feature.
B.4.
The EPIC A & B channels have control signals for the AFW control valves emerging from both their level control and their Vector (F.O.G.G.) logics.
RS's EFIC prioritizes these signals internal to EFIC so that only one set of control signals (from one power source) is necessary to convey all EFIC information to the valve controller.
CR-3 and ANO require additional relay logic external to EFIC to perform the same function.
B.S.
The R.S. EFIC has provision for altering the ADV opening setpoint from the control room.
CR-3 and ANO do not have this designed into their EFIC.
This feature will not be used at R.S.
B.6.
The low range S.G.
level at R.S.
is between 6' and 156' above the lower tube sheet (TS) with the upper tap extending thru the downcomer annulus to sample pressure in the tube bundle. The high or wide range taps extend between 6' above the lower TS to-6' below the upper TS (6' to 619' above the lower TS).
No composite full range is necessary.
5415MAJ Page 50-Rev. 2
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Work Rsqusst. 104415 Discipline s! & ' C MOD 001 Da te 10-31-86 At ANO the low range S.G. level is between 6" and 156' above the lower TS with 7 of the 8 upper taps sampling downcomer pressure and the eighth extending through the downcomer into the tube' ~ ~ ' '
- ~
~
bundle. The high range S.G. level extends between 102' and 500* above the lower TS with lower tap m'easuring downcomer pressure, and upper tap measuring tube bundle pressure. A composite full range level ulitilized by the control module is generated with signal conditioning and switching logic.
CR-3 uses low range S.G. level taps at 6' and 277*
above the lower TS.
The upper tap samples downcomer pressure. The low range span is 150' between 6" and 156' above the lower TS.
The high range level taps are coincident with the operate range level taps at 102' and 384' above the lower T.S.
A composite full range signal is utilized by the control module.
ANO uses separate S.G. sensing taps for each of it's 16 EFIC level transmitters whereas ANO and R.S. share some taps.'
For the R.S. Specific tap scheme see sub-ECN 5415B.
B.7.
R.S. will be using Gould type PD 3200 level transmitters. ANO and CR-3 use Rosemount transmitters B.8.
All three EFIC's have relay switching internal to the cabinets, available to isolate and switch valve control commands to either the main control room or an alternate shutdown panel.
CR-3 and ANO do not use the switching capability. At R.S.
the switching is used to switch and isolate EFIC hadd/ auto. control from the Control Room to the shutdown panel for fires in the Control Room.
Also, the " manual / reset *, bypass, and Channel DC power circuits to EFIC from the Control Room will be isolatable for the same fire scenario.
B.9.
The SFAS signals which command EFIC to start AFW originate in two separate unit modules per SFAS actuation channel. This prevents f ailure of a single SFAS module from spuriously initiating AFW.
CR-3 and ANO designs use only one unit control module per SFAS actuation channel.
5415MAJ
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CR-3 and ANO use EPIC to command closure of Main Steam Isolation Valves. This function is available ih the R.S. EFIC but'is not'used.
B.ll.
The EFIC cabinets at R.S. are located in four (4) separate rooms in the NSES.
ANO has the four (4) cabinets adjacent to each other in the control room.
CR-3 has the cabinets located in four (4) rooms in the building which houses the control room.
B.12.
At R.S. each SG has parallel AFW feed arrangements of one air operated control valve (from original installation) and one Target Rock modulating solenoid control valve.
CR-3 and ANO utilize two target Rock manufactured modulating sole'noid control valves in parallel per S.G.
B.13.
At R.S.
loss of IE and non-IE power witnin the C or D EFIC channels is annunciated separately. Loss of non-IE power to the C or D channels at CR-3 and ANO is not annunciated.
B.14.
R.S. will utilize an administratively set time delay to filter pressure transients from S.G. low level, S.G. low pressure, S.G. high-high level, and S.G. differential pressure bistable initiate signals.
CR-3 utilizes a time delay filter for their S.G.
low level initiate only. ANO currently uses no time delays.
B.15.
R.S. and ANO have Trip Interface Equipment (T.I.E.)
which forms a part of the EFIC actuation logic.
These interposing relay arrays are not used at CR-3.
B.16.
R.S.
Utilizes automatic MFW overfill isolation, but not AFW overfill.
CR-3 utilizes Automatic AFW overfill isolation, but not MFW overfill. ANO does not use its automatic overfill isolation capability.
B.17.
CR-3 initiates AFW on low level in both steam -
generator. ANO and RS initiate AFW on low level in j{
either steam generator.
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!& C MOD 001 Date 10-31-86 VIII.C.
Use of Oricinal Plant Valves
' -- ~
Of the twenty-five~ active' fluid ~"s'ystem components which
~ ~
receive direct commands from EFIC, only four are new; the APW Control valves TV-20531 and FV-20532, and AFW isolation valves HV-2.0581 and HV-20582.
Additionally, new motor operators are being installed on the previously manually operated gate valves FWS-015_ and FWS-0,16.
Clearly, the
. newly installed items are purchased, designed, and installed to the applicable codes as required by the EFIC APW system description. Similarly the signals from EPIC and the transducers which receive those signals are designed and installed to the applicable codes.
Also, the yhlves and vaive actuators which receive EFIC commands but which were installed as original plant equipment, and which have been performing the same safety function, may require re-certification or other pedigree or qualification
/h(
u pg rade. Specifically, the MFW control and start-up valves, the AFW control valves, FV-20527 and FV-20528, and the ADV's will continue to be used. The functions of each of these valves has not changed, and since initial operation the valves have provided precise, reliable fluid control. What is changed with the upgraded design is that signals and transducers which tell the valves how to respond will now be safety grade, as will their motive power sources. Also, in each case, a safety grade, emergency' power backed isolation valve will have oeen installed in series with each valve. And except for the ADV's the safety grade function of each of these proven
, dependable valves is also performed redundantly by newly installed safety grade equipment.
A review of qualification documentation for these components is being performed. Upgrading of documentation, including testing as required,'will be completed prior to startup.
i 5415MAJ Page 53 Rev. 2 g
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'A-5415 MAJOR NCR Work REqusst 104415 Discipline I&C MOD 001 Date 10-31-86
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FIGURE VI - EFIC CONSOLE HISS (E) 5415MAJ Page 54 Rev. 2 l
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ECN A-5415 MAJOR NCR Work Rsquist 104415 g
Discipline' I&C MOD 001 Date 10-31-8E ITEM suuO HO. t -
DreCRIPTICN-
- tNSTP7 TAG PC.
-~ - - -
l. } CONOENSATE STORt.GE TANK.
- f 5 5 LEVEL INO:CATOR CK A lC-553O?
2 l STM GIN. A '.0W RANGE L* VEL CHANNEL A DJO:CATOR lLI-ZC505A 3 l STM GEN. A LOW RANGE LEV!L, CHANNEL S INOlCATOR lLI2050!S i STM. GEN.' A HiGH RANGE LEVEL, CHANNEL A LNO::ATOR l C-20307A 4
5 l ETM GEN.'A WIGia RANGE LEVEL, CHANNEL S!N":"ATCR lL!-205C75 6 l 'NITIATE/ TEST MATRIX, E:lC CHANNEL A l
7 l STM GEN. 5 LCW RANGE LEVEL, CH ANNEL A INO:CATOR l C-2050t.sA S l STM GEN. 5 LCW RANGE LEVEL, CHANNE 5 INOtCATOR l L1-!O50:s 5 3 l STM GEN. 5 WGH RANGE LEVEL, CHANNEL A INO:CATCR l LI-20508 A !
80 l SIN GEN. S M4H RANGE LEVEL, CHANNEL 5 INO:CATOR l LI-ICSCSS 27 l STM GEN. A ??.iS5URE, CHANNE A INDICATCR l ?!-ZO!a5 A 25 l STM GEN. A PRES 5URE. CHANNEL S *N0lCATOR IF!-205455 29 l AFW F.OW TO STH GEN. A, CHANNEL A IN0!CATOR
[ F!-!!SC2.
30 l AFW P.0W 70 :TM GEN. A. CHANNEL S IN0!CATOR I FI-tis 05 3I l PJITIATE / TEST MATRIX, E.:lC CHANNEL S l
s 32 l STM GE.% S PRESEURE, CHANNEL A INO!CATOR- - - -
33 l STM GEN. 5 8tESE;;RE, CMANNEL S IN0!CATCR
{ 81-IOS&ea t PI-1051:s5 3A ! AFW FLOW TO STM GEN. S, CWANNE:. A IN010ATOR lF!-!!502 SS l AFr4 FLCW TO STN Cf.N. 5, CMANNEL S "N0iCATOR l F*.-!!901 A
!! l E:!C CHANNEL A 'SYFAES (MSC)-1 l
l 5: l E:ic CHANNEL 5 5YPASS (MSC)-l l
55 l 19C CHANNEL C SYPA* 5 ( M 5 0) - l
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54 l E90 CHANNEL D SYPA55 ( MSCl-l l
15 ' VOLTAGS REGUL ATOR Sum.Y SREAc (MSC)-Z l
is I AFW TO STM SEN. A !50L. VALVE HV-20151 (MSC-2 I HS-;05!!
57 l AFH TO ETM CEN. A ISCL. VALVE HV-20ST7 (M50)-2 lHS-10577 55 l STM GEN. 5:!; LEVEL CCNTROL CHANNE A l
59!::10 CHANNEL A CONTROL
- NIT: ATE 0 (M50}-1 l
60 l STM GEM. f.:';*.EVEL CCNTROL CFANNE. E l
(
61 l E.:!C "JiANNE_ E CON RO ?;1TIATE0 (M50)-f 62 l AFr4 TO ETH GEN. S '50L. VALVE hN-ZO53;(w50)-2 l W.5-?O5 5; c5 l AFW O SW C-EN. 5 !iOL. VALVE WV-205 3 (M501-2 l45-20075 TABLE VI - EFIC CONTROLS ON HlSS(E) PANEL 5415MAJ Page 55 Rev. 2 e, _
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.TW DESCRIPTION 300 N O.
' INSTR. TAG NO.
' $6 l AFM CONTROL VALVE PV-20527 POSITION INO!CATOR lZI-20527 25 l AFW TD STM C#_N. A CONTROL VALVE FV-20527 lH5-20527 Sh l AFW CONTROL VALVE FV-26551 POSITION IN0!CA*0R ll!-20!!!
$7 l AFW TO STM GEN. A- 00CROL. t:ALVE FV-2055:
!W5-20!St _
SS l AFW CONTROL VALVE FV-20523 POSITION INO:CATOR lII-2052S SSlAFW TO 5"N GEN. S CONTROL VALVE FV-20525 lWS-20!ZS 50 [ AFW CONTROL VALVE "V-20552 P05mCN INO CATOR lZ*-20552 Si l AFW TO STH-SEN. S CONTROL VALVE FV "l0551 l WS-20532 3-
$0A l AFri ~KO5573E VALVE HV-!!S26 (MSC)-Z l HS -!! alt,
- 05 l A,N CRO55T!!fVALVE HV-!!S27 (MSC)-Z l HS -!!SZT
- fir l AFW PUMP 8-510' AMMETER l
117 l ArW PUMD 8-515 (MSci-t l
IIS l AFW PUMP P-!!S STM*STOP VALVE }.N-50 SCI (M50-2 l H5-!!$0i
!!! l !.FW PUHP P-315 C:5 CHARGE FRE55URE INO:0ATOR I F~ -3!?Of 120 l AFW TFJET FLOW VALVE HV "S155 (M50-2 l
lZ1 l AFW TEST FLOW-.'.Not0A*0R (OiG17AL) l 122 l AcW R.JMP F-3f S AMMETE t l
i' 123 l AFW PUMP F-5!9 (M50)-2 l
12& l AFW PUMP F-319 0:50Ht.RGE ?RE55URE 'N010ATOR l ?!-3!SO:
I!!, l SECONOARY SYSTEMS ANNUNCIATOR /FiRST CUT RESET W50)- A l
us l AN U57 P' OW VALVE HV-$355 POS: TION IN0!O/ TOR l 21-!!!!5
!!7Al AFW POMO D-316 M.CTOR KEYLCCK l
d TABLE VI (CONT) - EFIC CONTROLS ON HISS (E) PANEL 5415MAJ Page 56 Rev. 2
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Discipline
- & C MOD-001 Da te 10-31-86 LIST OF REFERENCES s
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1.
Letter; SMUD to NRC; Feb. 18, 1983'; AFWS revied - NUREG 0737
~~
~ ~ "
II.E.1.1; Additional info requested; seismic, Tornado, pipe break,
control valve control air; AFW pump protection.
2.
Letter; SMUD to NRC; Jan.14, 1983; upgraded AFW Sys-Nureg 0737
- II.E.1.1. ; res ponds to request from 12-8-82 for additional info.
3.
S ER: NRC to SMUD; April 7, 1983; Rancho Seco - Status of the AFWS upgrade Review (NUREG 0737 item II.E.1.1).
4.
Standard Review Plan; NUREG 0800, 10.4.9 AFW System (PWR) 5.
Letter; B&W to SMUD; July 9, 1984; EFW Third Pump; explains reliability data dif ference between NRC & B&W.
6.
Letter; SMUD to NRC; April 28, 1983; EFIC upgrades AFWS; 0737 Items II. E.1.2 and II.K. 2.10.
Sent 15-1120850-03 Rev 3 of System.
Description.
7.
Letter; NRC to SMUD; Dec. 8, 1982; AFW upgrade Additional Information Request.
8.
Emergency Feedwater System, B&W Design Review Board (Doc.
680-1125704-00).
9.
Drawing List of EFIC, TIE, B&W Vendor Drawings.
5
- 10. EFW upgrade f or R.S. ; Test Spec. ; 62-1149372-00, Apr. '84
- 11. Draf t Tech Specs. EFIC/RPS/SFAS; B&W Doc. 05-0004; Mar. 25, 1985
- 12. Anticipa,ted Performance Behavior of the AFW Pumps under extreme conditions with A MSLB with both Nuclear Steam generators at Atmospheric Pressure, Elemer Makay, Report ER CO - 6 9 3, Feb. 16, 1984.
- 13. Control Room Implementation of EFIC; July 9, 1984 End Extension review.; Starkey; SMUD memo from J. Williams to Distribution.
- 14. Letter SMUD to B&W; Mar. 2, 1982; EFIC Shutdown Bypass; Rejects original shutdown bypass concepts.
l
- 15. Letter; SMUD to B&W; Dec. 13, 19 82 ; Shutdown Bypass. Says to go w/ single shutdown bypass (& Reset)
5415MAJ Page 57 Rev. 2
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17.
Memo; Wichert to Redeker Oct.
4, 1982; Evaluation of OA.82-003; acceptability of Shutdown bypass release..
~
18.
Letter SMUD to B&W (Whitney. to Holt); April 6, 1982; Feedwater Flow.
Versus Reactor Power Comparator Anticipatory Trip Operability Study; Shows F.W.
flow data for normal shutdown.'from Apr. 2, 1982.
19.
Letter SMUD to B&W: Aug. 30, 1982; FW flow vs. Reactor Power Comparator Anticipatory Trip Operability Study; Startup FW flow from August 19, 1982.
20.
Let.ter SMUD to B&W; Sept. 27, 1982; FW flow vs. Reactor Power
)
Comparator Anticipatory Trip operability Study; shows FW flow vs.
Power for the return to power of Sept. 17, 1982.
21.
Operability Evaluation of proposed Power to Main Feedwater Flow Trip; B&W Doc. 51-1135242-01; sent w/B&W letter SMUD-82-227 dated Oct. 20, 1982. Says it works.
22.
Letter SMUD to B&W, Nov. 24, 1982; Flow Signal for Flux /MFW Flow Trip.,
23.
Letter B&W to SMUD: Apr. 5, 1983; (SMUD-83-ll3;; Transmits B&W 51-1141558-00.
24.
Letter SMUD to NRC: July 20, 1983; NUREG 0737 II.K.2.10 ARTS; gives details of Flux /MFW flow ARTS.
25.
SMUD Memo; Beabe to Daniele; 12-18-85; AFW Power Sources.
26.
Letter SMUD to NRC; January 17, 1986; Status of EFIC Implementation:
Brief statement of* Cycle 8 EPIC System.
27.
Letter SMUD to NRC; March 3, 1986; EFIC Cycle 8 Scope; Clarification of cycle 8 EFIC.
28.
29.
EFIC Test Procedure, Vitro Doc. TP3801-4009, SMUD DOC. N28.02-309, 30.
Containnent Transmitter Enci cure, Inside Temperature Transmitted
~
During MSLB; Z-ZZr1M1795,17 :sges: 12-24-85; Bechtel.
31.
Temperature Rise in level Transmitter Reference Leg Under Transient condition.,(LT-20001 A/B; LT20002 A/B);. Calc. Z-RCS-I-0059; 3-28-84; Bechtel.
32.
HELB Analysis - AFWS changes: ECN A-: 006A, A-2912, A-3094, A-3622, A-3653; Bechtel Calc. M21.30-363, 34 4s: 6-14-83.
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33.
AFW upgrade Reliability Analysis for R.S. ; Document date.d April 1981.
34.
Rel,iabil,ity/, Avail _,a,bility _ Analysis. f or _ EFIC (Vitro r; May 1, 19,8 4 ;
Vitro Doc IO3801-4340 Rev. A; B&W Doc 932-1010496 01.
35.
Overcooling effects of a Variable Overfill Protection System for SMuc dated 11-5-82; Cale: B&W doc. 951-1138358-00.
36.
BAW-1655,,Jan. '81; Main Feedwater Overfill; Evaluates OTSG Dif ferential Pressure overfill parameter.
37.
B&W Study; 86-1134596: An Improved Concept for Controlling Main Feedwater Overfill (36 pages).
38.
BAW-1626; Aug. '81; Emergency Feedwater Level Rate Control-Control Evaluation sent W/B&W letter SMUD 81-120, Apr. 3, 1981. Evaluates Rate Limited followeg concept.
39.
B&W Doc. 77-1151127-00 Steam Generator Level Accuracy w/Gould Transmitter; sent w/ letter April 30, 1984.
40.
B AW 14.2 Rev. 1; March '80; Conceptual Design Study for AFWS feed rate control for B&W 177 FA plants.
41.
Estimates of max AFW flow rates for RS; dated 6-15-82; B&W Doc.
51-1134742-00.
42.
AFW flow rate Flow Orifice Calc.; SMUD Memo; Wichert to Stephenson; j
9-28-81.
3 43.
Summary of EFW upgrade LOFW analysis: April 29,1981; B&W No.
86-1123794-00.
~.
^
44.
Heat Removal Capability of SMUD Condensate Storage Tank; Mar 29, 1983; B&W No. 32-1141727-00.
45.
Letter; B&W to SMUD; 11-5-82 and memo Toney to Myers, 11-1-82; AFW Upgrade Implementation.
46.
ECCS Anal,ysis; EFW System Upgrade; B&W doc. 977-1125999-01 Substantiates 80% (actual equipment height) as absolute minumum level.
47.
Intentionally left blank.
48.
AFU upgrade Setpoints; B&W Calc #32-1155738-00; l-22-85.
49.
EPIC shutdown bypass - operator action: B&W 51-1138803-00; Nov. ' 8 2 ".
Shows at least 10 mins. to initiate AFW if MFW lost 9 750 psig during normal cooldown.
5415MAJ Page 59 Rev. 2.
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ECN A-5415 MAJOR NCR Work Raquest 104415
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Discipline I& C MOD 001 Date 10-31-86
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50.
MFW Flow Element Differential Pressure; cale. Z-FWS-M1600.
~
- 51. 'MFW Flow Element Pressure; cale. Z-M1615.
52.
Reanalysis of Minimum Required AFW Flow Rate, Calc. Z-FWS-IO102; B&W Doc. 86-1151208-00.
- 53. SER NRC to SMUD; September 26, 1983; ' Rancho Seco.- Status of the
(/bi Auxiliary Feedwater (AFWS) Upgrade Review (NUREG-0737 Item II.E.1.1) *.
- 54. Letter NRC to SMUD; April 1,1985;-
- Status of Auxiliary Feedwater (AFWS) Upgrade Review".
- 55. Single Failure Analysis of EFIC, B&W Doc. 32-1010482-01, prepared by Vitro, March 1983, vitro No. 3801-1330.
- 56. Memorandum from R. Daniels to V. Lewis; May 28, 1986; RED 86-169; 'IE Information Notice 86-15: loss of offsite power caused by problems in Fiberoptic Systems.'
- 57. Letter J. J. Mattimoe to J. F. Stolz of NRC dated May 3, 1984 transmitting report on "Ef,fect of Internally Generated Missiles on the Auxiliary Feedwater System Outside Containment" for Rancho Seco Nuclear Generating Station Unit 1".
- 58. Memo; Jerry Williams to Bob Daniels, D.C. Motor Dutz cycles for AFW valves, dated Oct. 28, 1984.
- 60. Report; Task 170 N.I. Calibration Error Final Report, March 1981, B&W document i 12-11250,41.
g
- 61. Memo; Bud Beebe to Rob Roehler, EFIC level sensing failure affecting multiple channels, Oct. 6, 1986.
- 62. Babcock & Wilcox to SMUD letter, AFE upgrade implementation shutdown bypass design review, SMUD 83-108, May 27, 1983.
5415MAJ Page 60 Rev. 2 I
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ENCLOSURE 4 EFIC Auxiliary Feedwater System Description
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