ML18101A273
| ML18101A273 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/30/1994 |
| From: | Mcnall C, Pyle L Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML18101A272 | List: |
| References | |
| SC-CN001-01, SC-CN1-1, NUDOCS 9410110088 | |
| Download: ML18101A273 (224) | |
Text
FORM NC.DE-AP.ZZ-0002-1 CALCULATION COVER SHEET CALC. NO.: SC-CNOOl,-01 REVISION: lIRl
- CALC. TITLE: SALEM. UNIT 1,2 STEAM GENERATOR LEVEL TRIP,ALARM, IND, & REC
- SHTS. (CALC):123 ATTACHMENTS: #/TOTAL SHTS.: 10/78 TOTAL SHTS.: 201 D
INTERIM (Proposed Plant Change)
FINAL (Supports Installed condition)
DESCRIPTION OF CALCULATION REVISION (IF APPL.):
D VOID See Revision lIRl History Sheet.
REASON FOR CALCULATION REVISION (IF APPL. ) : Interim Revision lIRl was intiated to provide engineering justification for lowering the Low Steam Generator Level setpoint.
HOPE CREEK D2sh OF DR
- N/A D2 D2s Q-LIST (SALEM)? YES D NO
. D IMPORTANT TO SAFETY ? YES NO FUTURE CONFIRMATION REQUIRED ? D YES NO OTHER DOCUMENTS AFFECTED?
8/16/94 DATE 8/16/94 DATE 8//t/,Yy DATE N/A DATE B/3o I Jlli-'
' DATE If the calculation is either Q-List, Q, Qs, Qsh, F, R, or Important to Safety "YES", completion of the certification for Design Verificatio.n (Form NC.DE-AP.ZZ-0010) is required *
- Nuclear Common 9410110088
~DR
- 94-l004--*--\
ADOCK 05000272 Page 1 of 2 Revision 2 PDR
FORM NC.DE-AP.ZZ-0002-1 CALCULATION COVER SHEET CALC. NO.: SC-CNOOl-01 REVISION: 1 r~o CALC. TITLE: SALEM UNIT 1,2 STEAM GENERATOR LEVEL TRIP.ALARM, IND, & REC f SHTS. (CALC):102 ATTACHMENTS: #/TOTAL SHTS.: 11/101 TOTAL SHTS.: 203 D
INTERIM (Proposed Plant Change)
FINAL (Supports Installed Condition)
DESCRIPTION OF CALCULATION REVISION (IF APPL.):
D VOID This revision combines all the loop* for the narrow range Steam Generator Level Instrumentation: high level override and alara, low-low level trips, indication and recorder loops. This calculation super*edes SC-CHOOl-02 thru SC-CHOOl-05.
REASON FOR CALCULATION REVISION (IF APPL. ) : This calculation revision is required to include the evaluation of the replacement signal isolators being installed within DCP 2EC-H5-&. ~
~ 3118 t'Kj 2.
HOPE CREEK D2 D2s DQsh OF DR ON/A Q-LIST (SALEM)?
IMPORTANT TO SAFETY ? .
-- YES YES D
D NO NO FUTURE CONFIRMATION REQUIRED ? D YES - NO OTHER DOCUMENTS AFFECTED? ~CB~~. FSAR, etc.): 2EC-~ .3/7'8 Pr.:U z...
ORIGINATOR/COMPANY NAME: C
-r4a.__ 1n HIA M. MCNAL~&G m nate c I~
12/13/93
. . (!_.___ p ~ /:". (f..e.kt.41<-) DATE PEER REVIEWER/COMPANY NAME: AND EW F. SHAUL PSE&G 12/28/93 DATE VERIFIER/COMPANY NAME: .~TY /PSE&G 1L..l.U94 DATE REVIEWED: *. / l. l\J/A Contractor Supervisor DATE
.. (~licable)
APPROVED: (J ---:- kv'-1rL PSE&G supervisor 2-/2.
'DATE 3/J1 (Req'd)
If the calculation is' either Q-List, Q, Qs, Qsh, F, R, or Important to Safety "YES", completion of the certification for Design Verification (Form NC.DE-AP.ZZ-0001) is required.
Nuclear common Page 1 of 2 Revision 2 A COMPUTER DISK EXISTS FOR THIS CALCULATION.
REVISIONS TO THIS CALCULATION REQUIRE DISK UPDATE !!!
Title SG NR LEVEL Cover PSE&G 1--~~~~~~...-~~~~~~~~~~~~~~~~--1sheet lear Department ID Number REFERENCE 1 ulation Cover Sheet SC-CNOOl-01 of 33 Calculation Revision o IR :1 0 us:c -'3.,IJ 'l I. - 50l3 CP Number SC-CNOOl-01 2.eC- '3~s/ :r- ?DI Revision History TJ.11-:. 'PAC!il6 NO LDN&-e-R..
(Interim or Final) '41il..1Lel> 1 APP#lOUAL.~
Interim = Proposed Plant Change INTERIM Peft- ~c .t>6 - A.'P. a:~ .() ()0 z..
Final = Supports Instal- (F"oieM i) led condition Future Confirmation Required:
Originator (Initial & Date)
Reviewer (Initial & Date)
Public Service Supervisor Approval (Inital & Date)
Cover Sheet ber Pages) 1 ulations (Number Pages)
(Excluding Attachments) 32 Attachments (Number/Total pages) 5 /25 s ~5 Total Pages 58 SB Important to Safety yes x no If yes, design verification required per DE-AP-ZZ-OOlO(Q) (Design Verification, Hef. 8.3)
DE-AP.ZZ-0002(Q) Rev. 0 Exhibit 1 page 1 of 2
CALCULATION CONTINUATION/ SHEET: i OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 REVISION HISTORY (lIRl)
This calculation is in Interim Revision lIRl. Revision lIRO was initiated to support Design Change 2EC-3258, for modification to RG 1.97 isolation, and the finalization of the Emergency Response Guidelines Revision lB. The results of lIRO indicated that the Low-Low Setpoint could change but made no recommendation for the Low Steam Generator Level setpoint. This Interim revision (lIRl) was initiated to provide engineering justification for the lowering of the Low Steam Generator trip coincident with the Steam Flow/Feed Flow Mismatch function from 25% to 10% based on the desire to change it at the same time the Low-Low Setpoint is changed. The proposed changes would increase operating margin relative to steam generator level. This would help preclude unnecessary reactor trips and AFW system actuations during plant evolutions involving steam generator water level changes (e.g., plant startup), while continuing to ensure the analytical limits in the safety analysis. Both setpoints will be changed within Design Change lEC-3345 Pkg 2 and 2EC-3306 Pkg 2.
This calculation revision (lIRl) also addresses the following additions or changes:
The addition of a recommendation to change the Low Setpoint Allowable Value from 24% to 9%.
The addition of 1% additional PMA for the High-High trip function provided in Westinghouse Letter PSE-94-555,
Subject:
JPO (Justification for Past Operation) for Overpower Operation.
The addition of a recommendation to change BOP Indicated values affected by the addition of the 1% additional PMA.
Revisions to scaling corrected the "Recommended" Low-Low setpoint voltage table which erroneously showed the old setpoint (but did indicate the correct new voltage). An additional table was added for the new Low Setpoint. An error was also corrected in the High-High setpoint discussion where the setpoint was referenced as 65% instead of 67% as stated in the design inputs section.
Changes to the scaling tables based on a request from the plant to show "off scale" values in an alternate manner.
A change to Assumption 6.2 based on the replacement of the EQRR radiation maps with the new EDC PSBP 317079-01. This document also caused the Environmental table in Section 5.1.1 to be revised.
~i I
CALCULATION CONTINUATION/ SHEET: i i OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC.- No~: - SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12113193 1 CMM 8116194 I1IRi I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 A change to Assumption 6.1.4 based on interpretation changes for the Setpoint Technical Standard direction for adding Seismic uncertainties.
The change to include the EOP evaluation as part of the calculation and to finalize the data within that section based on the new revision to the EOPs and ERGs. The Attachments were renumbered based on the elimination of the EOP Attachment.
Various grooming improvements were made which did not affect the content of this document (i.e.
spelling, font sizes, grammar). All significant changes are marked with a Revision Bar to denote the changes included in this interim revision.
CALCULATION CONTINUATION/ SHEET: i i i OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 LIST OF EFFECTIVE PAGES (MAIN BODY)
Page Rev Page Rev Page Rev Page Rev Page Rev Cover lIRl 22 lIRl 49 lIRO 76 lIRO 103 lIRl 1 lIRl 23 lIRl 50 lIRO 77 lIRO 104 lIRl 11 lIRl 24 lIRl 51 lIRO 78 lIRl 105 lIRl 111 lIRl 25 lIRl 52 lIRO 79 lIRO 106 lIRl lV lIRl 26 lIRl 53 lIRO 80 lIRO 107 lIRl v lIRl 27 lIRl. 54 lIRO 81 lIRO 108 lIRl 1 lIRl 28 lIRl 55 lIRO 82 lIRO 109 lIRl 2 lIRO 29 lIRl 56 lIRl 83 lIRO 110 lIRO 3 lIRO 30 lIRl 57 lIRO 84 lIRO 111 lIRl 31 58 85 112 lIRl
-~
lIRO lIRl lIRO lIRO lIRO 32 lIRl 59 lIRO 86 lIRO 113 lIRl 6 lIRO 33 lIRl 60 lIRO 87 lIRO 114 lIRO 7 lIRO 34 lIRl 61 lIRO 88 lIRO 115 lIRO 8 lIRO 35 lIRl 62 lIRO 89 lIRO 116 lIRO 9 lIRO 36 lIRl 63 lIRO 90 lIRO 117 lIRl 10 lIRO 37 lIRl 64 lIRO 91 lIRO 118 lIRl 11 lIRO 38 lIRl 65 lIRO 92 lIRO 119 lIRl 12 lIRO 39 lIRl 66 lIRl 93 lIRO 120 lIRl 13 lIRO 40 lIRl 67 lIRl 94 lIRO 121 lIRl 14 lIRO 41 lIRl 68 lIRl 95 lIRO 122 lIRl 15 lIRl 42 lIRl 69 lIRl 96 lIRO 123 lIRl 16 lIRO 43 lIRO 70 lIRO 97 lIRO 17 lIRl 44 lIRO 71 lIRl 98 lIRO 18 lIRl 45 lIRl 72 lIRO 99 lIRl 19 lIRl 46 lIRl 73 lIRO 100 lIRO 20 lIRl 47 lIRl 74 lIRl 101 lIRl 21 lIRl 48 lIRJ 75 lIRO 102 lIRl
CALCULATION CONTINUATION/ SHEET: iv OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I 1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 LIST OF EFFECTIVE PAGES (ATTACHMENTS)
Attachment A Page i 1 2 3 4 5 6 7 8 Rev lIRl 0 0 0 0 0 0 0 lIRO Attachment B Page i 1 2 3 4 5 6 7 8 Rev lIRl 0 0 0 0 0 0 0 lIRO Attachment C Page i 1 2 Rev lIRl 0 0 Attachment D
.Page i 1 2 Rev lIRl 0 0 Attachment 10.1 Page i 1 2 3 4 5 6 7 8 9 10 11 12 13 Rev lIRl lIRO lIRO lIRO lIRO lIRl lIRl lIRl lIRl lIRl lIRO lIRl lIRl lIRl Attachment 10.2 Page i 1 2 3 4 Rev lIRl lIRO lIRO lIRO lIRO Attachment 10.3 Page i 1 2 3 4 5 Rev lIRl lIRO lIRO lIRO lIRO lIRO Attachment 10.4 Page i 1 2 3 4 5 6 7 8 9 10 Rev lIRl lIRO lIRO lIRO lIRO lIRO lIRO lIRO lIRO lIRO lIRO Attachment 10.5 Page i 1 2 3 4 5 6 7 8 Rev lIRl lIRO lIRO lIRO lIRO lIRO lIRO lIRO lIRO Attachment 10.6 Page i 1 2 3 4 5 6 7
.Rev lIRl lIRl lIRl lIRl lIRl lIRl lIRl lIRl
CALCULATION CONTINUATION/ SHEET: v OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 TABLE OF CONTENTS 1.0 PURPOSE/SCOPE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.1 Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.2 Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.0 FUNCTIONAL DESCRIPTION/DESIGN BASIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 2.1 Functional Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . 14 2.2 Design Basis Inputs ........... ~-. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
3.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... ........ .. ... .. ... 42 3.1 Setpoint Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . ... ........ .. ... .. ... 42 3.2 Updated Final Safety Analysis Report (UFSAR) . . . . . ... ........ .. ... .. ... 42 3.3 Technical Specification /BOP Design Basis .......* . . ... ........ .. ... .. ... 42
- 4.0 3.4 3.5 3.6 Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Calculations and Support Documents . . . . . . . . . . . . . .
Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
LOOP DIAGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 43 45 46 5.0 DESIGN INPUTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 5.1 General Design Inputs .............................................. 53 5.2 Process Design Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 5.3 Transmitter Design Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 5.4 Rack Design Inputs ................................................ 59 5.5 Control Room Indicator Design Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 5.6 Hot Shutdown Panel Indicator Design Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63 5.7 Recorder Design Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63 5.8 M&TE Design Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 64 6.0 ASSUMPTIONS ....................................................... 66 6.1 General Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66 6.2 Process Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 6.3 Transmitter Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69 6.4 Rack Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69 6.5 Indicator Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69 6.6 Recorder Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 70
CALCULATION CONTINUATION/ SHEET:. vi OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 l CMM 8/16/94 I lIRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.0 CALCULATION OF UNCERTAINTIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 7.1 Process Measurement Uncertainties (PM) ............................... 71 7.2 Insulation Resistance Uncertainty (IR) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 74 7.3 Process Element Accuracy (PE) ....................................... 75 7.4 Calculation of Transmitter Uncertainties (XMTR) ......................... 75 7.5 Calculation of Rack Uncertainties (RACK) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 81 7.6 Calculation Of Control Room Indicator Uncertainties (INDciJ . . . . . . . . . . . . . . . 87 7.7 Calculation Of Hot Shutdown Indicator Uncertainties (INDHs) . . . . . . . . . . . . . . . . 91 7.8 Calculation Of Recorder Uncertainties (REC) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 95 7.9 Channel Error Analysis ............................................. 99 7.10 Propagation of Error . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 100 7.11 Summary of Channel Uncertainties . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 109 8.0 CALCULATION OF SETPOINTS ........................................ 110 8.1 Calculated Setpoints . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 110 8.2 Allowable Value /Acceptable Value Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . 112 8.3 Setpoint Relationships . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 117 9.0 DISCUSSION OF RESULTS .. . .... .. ... .. ... .... . .. .. . .. . . . . .. . ... . .... 118 9.1 Low-Low Setpoint . . . . .. . .... .. ... .. ... .... . .. .. . .. . . . . .. . ... . .... 118 9.2 Low Setpoint . . . . . . . . .. . .... .. ... .. ... .... . .. .. . .. . . . . .. . ... . .... 118 9.3 High-High Setpoint . . . .. . .... .. ... .. ... .... . .. .. . .. . . . . .. . ... . .... 119 9.4 Indicator and Recorder .. . .... .. ... .. ... .... . .. .. . .. . . . . .. . ... . .... 119 9.5 EOP Evaluation . . . . . .. . .... .. ... .. ... .... . .. .. . .. . . . . .. . ... . .... 120 10.0 ATTACHMENTS A Scaling (8 pages + cover)
B Scaling (8 pages + cover)
C Scaling (2 pages + cover)
D Scaling (2 pages + cover) 10.1 Scaling (13 pages + cover) 10.2 NUS Isolator Evaluation (5 pages + cover) 10.3 Moore Isolator Evaluation (5 pages + cover) 10.4 Westinghouse Letter; S/G Water Level PMA Term Inaccuracies (10 pages + cover) 10.5 Westinghouse Letter; Safety Analysis Limits (8 pages + cover) 10.6 Westinghouse Letter; JPO for Overpower Operation (7 pages + cover)
CALCULATION CONTINUATION/ SHEET: 1 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IR'.I I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 1.0 PURPOSE/SCOPE 1.1 Purpose The purpose of this calculation is to establish the Total Channel Uncertainties for the Steam Generator Narrow Range Level Instrument loops for Low-Low Level Reactor Trip, the Low Level Steam Generator Water Level Trip, the High Level Override and Alarm setpoints, RG 1.97 Control Room Indication, Hot Shutdown Panel Indication, and Recorder. The uncertainties established in this calculation support Technical Specification Setpoints, and Surveillance Requirements.
This calculation establishes the Instrument Scaling, Calibration Tolerances and Acceptable/Allowable values to be used in Calibration or Surveillance Procedures.
Additionally, EOP Indicated Values using this instrumentation are evaluated with respect to their design basis, to determine that they are set conservatively away from applicable limits including the Total Channel Uncertainties (normal or adverse as applicable) established for the RG 1.97 Indication loops within this calculation. ~
1.2 Scope I This calculation contains the following Instrument Loop Configurations: This calculation scope does not include an evaluation of the uncertainties for the computer points and therefore, they are not included below and are not shown on the Loop Diagram. See Section 4.0 for Loop Diagrams of the configurations shown below.
CALCULATION CONTINUATION/ SHEET: 2 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 1.2.1 Steam Generator 11 (21) 1.2.1.1 Configuration A: Reactor Trip System Instrumentation : Trip Setooint Steam Generator Level Low-Low Trip - Steam Generator Water Level Low-Low Channel IV 1(2) LT-517 Level Transmitter 1(2) LC-517A-B/R I/V 1(2) LC-517A-B Signal Comparator (To BS-517B, SSPS Reactor Trip) Low-Low Level Channel III 1(2) LT-518 Level Transmitter 1(2) LC-518A-B/R I/V 1(2) LC-518A-B Signal Comparator (To BS-518B, SSPS Reactor Trip) Low-Low Level Channel II 1(2) LT-519 Level Transmitter 1(2) LC-519A-B/R I/V 1(2) LC-519A-B Signal Comparator (To BS-519B, SSPS Reactor Trip) Low-Low Level 1.2.1.2 Confi!ruration B: Engineered Safety Feature Actuation System Instrumentation:
Turbine Trip and Feedwater Isolation: Steam Generator Water Level High High.
Channel IV 1(2) LT-517 Level Transmitter 1(2) LC-517A-B/R I/V 1(2) LC-517A-B Signal Comparator (To BS-517A, SSPS Turbine Trip) High Level Override and Alarm Channel III 1(2) LT-518 Level Transmitter 1(2) LC-518A-B/R I/V 1(2) LC-518A-B Signal Comparator (To BS-518A, SSPS Turbine Trip) High Level Override and Alarm Channel II 1(2) LT-519 Level Transmitter 1(2) LC-519A-B/R I/V 1(2) LC-519A-B Signal Comparator (To BS-519A, SSPS Turbine Trip) High Level Override and Alarm
CALCULATION CONTINUATION/ SHEET: 3 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRJ I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 1.2.1.3 Configuration C: Reactor Trip System Instrumentation : Trip Setpoint Steam Generator Level Low Channel IV 1(2) LT-517 Level Transmitter 1(2) LC-517A-C/R I/V 1(2) LC-517C Signal Comparator (To BS-517C, SSPS Reactor Trip) Low Level Mismatch Trip Interlock Channel III 1(2) LT-518 Level Transmitter 1(2) LC-518A-C/R I/V i(2) LC-518C Signal Comparator (To BS-518C, SSPS Reactor Trip) Low Level Mismatch Trip Interlock 1.2.1.4 Confiwration D: Accident Monitoring Instrumentation: Steam Generator Water Level (Narrow Range) Control Room Indicator Channel IV 1(2) LT-517 Level Transmitter 1(2) LM-517A/R I/V 1(2) LM-517A Signal Isolator 1(2) LI-517 /R I/V 1(2) LI-517 Indicator Channel III 1(2) LT-518 Level Transmitter 1(2) LM-518/R I/V 1(2) LM-518 Signal Isolator 1(2) LI-518/R 1/V 1(2) LI-5.18 Indicator Channel II 1(2) LT-519 Level Transmitter 1(2) LM-519A/R 1/V 1(2) LM-519A Signal Isolator 1(2) LI-519/R I/V 1(2) LI-519 Indicator
CALCULATION CONTINUATION/ SHEET: 4 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 1.2.1.5 Configuration E: Remote Shutdown Monitoring Instrumentation - Steam Generator Level Hot Shutdown Panel Indicator Channel IV 1(2) LT-517 Level Transmitter 1(2) LM-517A/R I/V 1(2) LM-517A Signal Isolator 1(2) Ll-517A Hot Shutdown Panel Indicator 1.2.1.6 Configuration F: Accident Monitoring Instrumentation: Steam Generator Water Level (Narrow Range) Recorder Channel II 1(2) LT-519 Level Transmitter 1(2) LC-519A-B/R 1/V 1(2) LM-519M Signal Isolator 1(2) LM-500W/R I/V
- 1(2) LM-519B Signal Isolator 1(2) LA-5048 Recorder 1.2.2 Steam Generator 12 *(22) 1.2.2.1 Configuration A: Reactor Trip System Instrumentation : Trip Setpoint Steam Generator Level Low-Low Trip - Steam Generator Water Level Low-Low Channel IV 1(2) LT-527 Level Transmitter 1(2) LC-527A-B/R I/V 1(2) LC-527A-B Signal Comparator (To BS-527B, SSPS Reactor Trip) Low-Low Level Channel III 1(2) LT-528 Level Transmitter 1(2) LC-528A-B/R 1/V 1(2) LC-528A-B Signal Comparator (To BS-528B, SSPS Reactor Trip) Low-Low Level
CALCULATION CONTINUATION/ SHEET: 5 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I 1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 Channel II 1(2) LT-529 Level Transmitter*
1(2) LC-529A-B/R I/V 1(2) LC-529A-B Signal Comparator (To BS-529B, SSPS Reactor Trip) Low-Low Level 1.2.2.2 Confi!rnration B: Engineered Safety Feature Actuation System Instrumentation:
Turbine Trip and Feedwater Isolation: Steam Generator Water Level High High.
Channel IV 1(2) LT-527 Level Transmitter 1(2) LC-527A-B/R I/V 1(2) LC-527A-B Signal Comparator (To 13S-527A, SSPS Turbine Trip) High Level Override and Alarm Channel III 1(2) LT-528 Level Transmitter 1(2) LC-528A-B/R I/V 1(2) LC-528A-B Signal Comparator (To BS-528A, SSPS Turbine Trip) High Level Override and Alarm Channel II 1(2) LT-529 Level Transmitter 1(2) LC-529A-B/R I/V 1(2) LC-529A-B Signal Comparator (To BS-529A, SSPS Turbine Trip) High Level Override and Alarm 1.2.2.3 Configuration C: Reactor Trip System Instrumentation: Trip Setpoint Steam Generator Level Low Channel IV 1(2) LT-527 Level Transmitter 1(2) LC-527A-C/R I/V 1(2) LC-527C Signal Comparator (To BS-527C, SSPS Reactor Trip) Low Level Mismatch Trip Interlock Channel III 1(2) LT-528 Level Transmitter 1(2) LC-528A-C/R I/V 1(2) LC-528C Signal Comparator (To BS-518C, SSPS Reactor Trip) Low Level Mismatch Trip Interlock
CALCULATION CONTINUATION/ SHEET: 6 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IR::I I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 1.2.2.4 Configuration D: Accident Monitoring Instrumentation: Steam Generator Water Level (Narrow Range) Control Room Indicator Channel IV 1(2) LT-527 Level Transmitter 1(2) LM-527A/R I/V 1(2) LM-527A Signal Isolator 1(2) LI-527/R I/V 1(2) LI-527 Indicator Channel III 1(2) LT-528 Level Transmitter 1(2) LM-528/R I/V 1(2) LM-528 Signal Isolator 1(2) LI-528/R I/V 1(2) LI-528 Indicator Channel II 1(2)°LT-529 Level Transmitter 1(2) LM-529A/R I/V 1(2) LM-529A Signal Isolator 1(2) LI-529/R I/V 1(2) LI-529 Indicator 1.2.2.5 Configuration E: Remote Shutdown Monitoring Instrumentation - Steam Generator Level Hot Shutdown Panel Indicator Channel IV 1(2) LT-527 Level Transmitter 1(2) LM-527A/R I/V 1(2) LM-527A Signal Isolator 1(2) LI-527A Hot Shutdown Panel Indicator
CALCULATION CONTINUATION/ SHEET: 7 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IR'I I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 1.2.2.6 Configuration F: Accident Monitoring Instrumentation: Steam Generator Water Level (Narrow Range) Recorder Channel II 1(2) LT-529 Level Transmitter 1(2) LM-529A-I/R I/V 1(2) LM-529M Signal Isolator 1(2) LM-500X/R I/V 1(2) LM-529B Signal Isolator .
1(2) LA-5049 Recorder 1.2.3 Steam Generator 13 (23) 1.2.3.1 Configuration A: Reactor Trip System Instrumentation : Trip Setpoint Steam Generator Level Low-Low Trip Channel IV 1(2) LT-537 Level Transmitter 1(2) LC-537A-B/R I/V 1(2) LC-537A-B Signal Comparator (To BS-537B, SSPS Reactor Trip) Low-Low Level Channel III 1(2) LT-538 Level Transmitter 1(2) LC-538A-B/R I/V 1(2) LC-538A-B Signal Comparator (To BS-538B, SSPS Reactor Trip) Low-Low Level Channel II 1(2) LT-539 Level Transmitter 1(2) LC-539A-B/R I/V 1(2) LC-539A-B Signal Comparator (To BS-539B, SSPS Reactor Trip) Low-Low Level
CALCULATION CONTINUATION/ SHEET: 8 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 1.2.3.2 Configuration B: Engineered Safety Feature Actuation System Instrumentation:
Turbine Trip and Feedwater Isolation: Steam Generator Water Level High High.
Channel IV 1(2) LT-537 Level Transmitter 1(2) LC-537A-B/R I/V 1(2) LC-537A-B Signal Comparator (To BS-537A, SSPS Turbine Trip) High Level Override and Alarm Channel ill 1(2) LT-538 Level Transmitter 1(2) LC-538A-B/R I/V 1(2) LC-538A-B Signal Comparator (To BS-538A, SSPS Turbine Trip) High Level Override and Alarm Channel II 1(2) LT-539 Level Transmitter 1(2) LC-539A-B/R I/V 1(2) LC-539A-B Signal Comparator (To BS-539A, SSPS Turbine Trip) High Level Override and Alarm 1.2.3.3 Configuration C: Reactor Trip System Instrumentation: Trip Setpoint Steam Generator Level Low Channel IV 1(2) LT-537 Level Transmitter 1(2) LC-537A-C/R I/V 1(2) LC-537C Signal Comparator (To BS-537C, SSPS Reactor Trip) Low Level Mismatch Trip Interlock Channel III 1(2) LT-538 Level Transmitter 1(2) LC-538A-C/R I/V 1(2) LC-538C Signal Comparator (To BS-538C, SSPS Reactor Trip) Low Level Mismatch Trip Interlock
CALCULATION CONTINUATION/ SHEET: 9 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IR~ I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 1.2.3.4 Configuration D; Accident Monitoring Instrumentation; Steam Generator Water Level (Narrow Range) Control Room Indicator Channel IV 1(2) LT-537 Level Transmitter 1(2) LM-537A/R I/V 1(2) LM-537A Signal Isolator 1(2) LI-537 /R I/V 1(2) LI-537 Indicator Channel III 1(2) LT-538 Level Transmitter 1(2) LM-538/R I/V 1(2) LM-538 Signal Isolator 1(2) LI-538/R I/V 1(2) LI-538 Indicator Channel II 1(2) LT-539 Level Transmitter 1(2) LM-539A/R I/V 1(2) LM-539A Signal Isolator 1(2) LI-539/R I/V 1(2) LI-539 Indicator 1.2.3.5 Configuration E: Remote Shutdown Monitoring Instrumentation - Steam Generator Level Hot Shutdown Panel Indicator Channel IV 1(2) LT-537 Level Transmitter 1(2) LM-537A/R I/V*
1(2) LM-537A Signal Isolator 1(2) LI-537A Hot Shutdown Panel Indicator
CALCULATION CONTINUATION/ SHEET: 10 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CN001-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 1.2.3.6 Configuration Fj Accident Monitoring Instrumentationj Steam Generator Water Level (Narrow Range) Recorder Channel II 1(2) LT-539 Level Transmitter 1(2) LM-539A/R I/V 1(2) LM-539A Signal Isolator 1(2) LM-500Y /R I/V 1(2) LA-5050 Recorder 1.2.4 Steam Generator 14 (24) 1.2.4.1 Configuration A: Reactor Trip System Instrumentation : Trip Setpoint Steam Generator Level Low-Low Trip
- Channel IV 1(2) LT-547 Level Transmitter 1(2) LC-547A-B/R I/V 1(2) LC-547A-B Signal Comparator (To BS-547B, SSPS Reactor Trip) Low-Low Level Channel III 1(2) LT-548 Level Transmitter 1(2) LC-548A-B/R I/V 1(2) LC-548A-B Signal Comparator (To BS-548B, SSPS Reactor Trip) Low-Low Level Channel II 1(2) LT-549 Level Transmitter 1(2) LC-549A-B/R I/V 1(2) LC-549A-B Signal Comparator (To BS-549B, SSPS Reactor Trip) Low-Low Level
CALCULATION CONTINUATION/ SHEET: 11 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 1.2.4.2 Configuration B: Engineered Safety Feature Actuation System Instrumentation:
Turbine Trip and Feedwater Isolation: Steam Generator Water Level High-High.
Channel IV 1(2) LT-547 Level Transmitter 1(2) LC-547A-B/R I/V 1(2) LC-547A-B Signal Comparator (To BS-547~ SSPS Turbine Trip) High Level Override and Alarm Channel ill 1(2) LT-548 Level Transmitter 1(2) LC-548A-B/R I/V 1(2) LC-548A-B Signal Comparator (To BS-548~ SSPS Turbine Trip) High Level Override and Alarm Channel II 1(2) LT-549 Level Transmitter 1(2) LC-549A-B/R I/V 1(2) LC-549A-B Signal Comparator (To BS-549~ SSPS Turbine Trip) High. Level Override and Alarm 1.2.4.3 Confilruration C: Reactor Trip Svstem Instrumentation : Trip Setpoint Steam Generator Level Low Channel IV 1(2) LT-547 Level Transmitter 1(2) LC-547A-C/R I/V 1(2) LC-547C Signal Comparator (To BS-547C, SSPS Reactor Trip) Low Level Mismatch Trip Interlock Channel III 1(2) LT-548 Level Transmitter 1(2) LC-548A-C/R I/V 1(2) LC-548C Signal Comparator (To BS-548C, SSPS Reactor Trip) Low Level Mismatch Trip Interlock
CALCULATION CONTINUATION/ SHEET: 12 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
- CALC. No.: SC-CNOOl-01 ORIGINATOR. DATE REV: CMM 12/13/93
REFERENCE:
1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 1.2.4.4 Configuration D: Accident Monitoring Instrumentation: Steam Generator Water Level (Narrow Range) Control Room Indicator Channel IV 1(2) LT-547 Level Transmitter 1(2) LM-547A/R I/V 1(2) LM-547A Signal Isolator 1(2) LI-547 /R 1/V 1(2) LI-547 Indicator Channel III 1(2) LT-548 Level Transmitter 1(2) LM-548/R I/V 1(2) LM-548 Signal Isolator 1(2) LI-548/R I/V 1(2) LI-548 Indicator Channel II 1(2) LT-549 Level Transmitter 1(2) LM-549A/R I/V 1(2) LM-549A Signal Isolator 1(2) LI-549/R I/V 1(2) LI-549 Indicator 1.2.4.5 Configuration E: Accident Monitoring Instrumentation: Steam Generator Water Level Hot Shutdown Panel Indicator Channel IV 1(2) LT-547 Level Transmitter 1(2) LM-547A/R I/V 1(2) LM-547A Signal Isolator 1(2) LI-547A Hot Shutdown Panel Indicator
CALCULATION CONTINUATION/ SHEET: 13 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IR1 I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 1.2.4.6 Configuration F: Accident Monitoring Instrumentation: Steam Generator Water Level (Narrow Range) Recorder Channel II 1(2) LT-549 Level Transmitter 1(2) LM-549A/R I/V 1(2) LM-549A Signal Isolator 1(2) LI-549Z/R I/V 1(2) LI-5051 Recorder
CALCULATION CONTINUATION/ SHEET: 14 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 2.0 FUNCTIONAL DESCRIPTION/DESIGN BASIS 2.1 Functional Description (Ref. 3.1.3)
Reactor Protection- The Steam Generator Protection system prevents loss of secondary side heat transfer capability; i.e., loss of feedwater to the Steam Generators. Reactor Trip and Auxiliary Feedwater actuation occur on two out of three low-low level channels in any Steam Generator. The Low-Low Steam Generator Trip must be operable in Modes 1 and 2 when the reactor requires a heat sink. It is a primary trip function for Turbine Trip (EOL), Loss of AC Power (Station Blackout), Loss of Normal Feedwater and Feedwater System Pipe Break.
It is a backup trip for Turbine Trip (BOL). Uncertainties for accident environment are included for the Low-Low trip function with the exception of reference leg heat up uncertainties. These effects are assumed to be minimized at the time of the trip since the reference legs are insulated.
The Steam Generator Water Level Low trip (used in coincidence with a Steam/Feedwater level trip) is not used in the transient analyses but is included in the Technical Specifications Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Protection System. This trip is redundant to the Steam Generator Water Level Low-Low trip. The low trip value is set sufficiently away from normal operating values to preclude spurious trips but will initiate a reactor trip before the steam generators are dry. Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor coolant system and steam generators is minimized. Since this trip is redundant to the Low-Low Water Level trip function, the same accident uncertainties based on environmental and operating parameters will be considered in the calculation.
Engineered Safety Features Actuation - The Steam Generator Water Level High-High function is used to terminate Feedwater addition (via isolation) in order to protect the turbine from damage from steam with too high a moisture content. The capability of the upper regions of the steam generator to dry the steam is compromised when the water level in the steam generator gets too high. In addition, the steam piping supports are not designed to withstand the loading of piping plus water. Finally, the accuracy of the Steam Flow and Steamline Pressure Transmitters downstream in the steam piping would be decreased due to the addition of significant moisture in the steam. Since this function is not assumed to operate in adverse environmental conditions (no break in a pipe), instrument uncertainties are calculated for normal conditions only.
CALCULATION CONTINUATION/ SHEET: 15 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
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ORYGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1I'D1 I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 Main Control Room Indication is provided with all channels indicated and the channels used for control, recorded (FSAR Table 7.5-1). Both normal and accident uncertainties are calculated for use during normal and accident monitoring. A post accident uncertainty is calculated to demonstrate potential loop uncertainties for 120 days post accident conditions as specified in the transmitter environmental qualification.
Additionally, Remote Shutdown monitoring is. provided at Hot Shutdown Panel 213. This calculation includes normal uncertainties only for this function since a Control Room Fire and Design Basis Event are not postulated to occur concurrently.
EOP Indicated Values are evaluated based on their function and requirements for inclusion of Instrument Uncertainties as provided in the footnotes of the Emergency Response Guidelines.
2.2 Design Basis Inputs 2.2.1 Analytical Limits
,~
Low-Low Trip Safety Analysis Limit = 0% (Ref. 3.5.20)
High-High Trip Safety Analysis Limit = 75% (Ref. Attachment 10.5) 2.2.2 Current Technical Specification Setpoints, Allowable Values (Ref. 3.3.1.9), Table 2.2-1 Reactor Trip System Instrumentation Trip Setpoints Functional Unit Trip Setpoint Allowable Value
- 13. Steam Generator Water ~ 16% of NR Instr Span* ~ 14.8% of NR Instr Span Level Low-Low each Steam Generator each Steam Generator
- 14. Low Steam Generator ~ 25 % of NR Inst Span ~ 24% of NR Instr Span Water level each Steam Generator each Steam Generator 2.2.3 Engineered Safety Feature Actuation System Instrumentation Trip Setpoints (Ref. 3.3.1.5), Table 3.3-4 Functional Unit Trip Setpoint Allowable Value
- 5. Turbine Trip and ~ 67% of NR span ~ 68% of NR span Feedwater Isolation each Steam Generator each Steam Generator
CALCULATION CONTINUATION/ SHEET:. 16 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IR.'I I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 2.2.4 UFSAR Desi1m Basis Requirements Section 75 Safety Related Display Instrumentation Main Control Room Indicators And/Or Recorders Available to the Operator Table 7.5-1 Parameter Channels Range Ind/Rec Operational Occurrences
- 7. SG water lvl (NR) 3/SG +7 to -5 +/-4% span* All channels feet from (hot) Indicated, control full load wl channels recorded Accident Conditions
- 3. SG water lvl (NR) 3/SG +7 to -5 +/-10% span . All channels feet from (hot) Indicated, control full load wl channels recorded Main Control Room Indicators And/Or Recorders Available to the Operator to Monitor Significant Plant Parameters During Normal Operation Table 7.5-2 Parameter Channels Range Ind/Rec*
Feedwater and Steam Systems
- 2. Steam Generator level (NR) 3/SG +7 to -5 Ft +/-4% All channels Indicated, control channels recorded
CALCULATION CONTINUATION/ SHEET: 17 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 2.2.5 EOP Indicated Values (shown as Setpoint below)
(Ref. 3.3.3, 3.6.3)
EOP NUMBER AND TITLE: TRIP-1, REACTOR TRIP OR SAFETY INJECTION ERG NUMBER AND TITLE: E-0, REACTOR TRIP OR SAFETY INJECTION EOP STEP ERG FOOTNOTE SETPOINT STEP 19 16 8 8% SG LEVEL 19 16 9 12% SGLEVEL 19 16 N/A 16-33% SG LEVEL 23 19 8 8% SG LEVEL 23 19 9 12% SGLEVEL 23 19 N/A 16-33% SG LEVEL 36 25 8 8% SG LEVEL 37 28 8 8% SG LEVEL 37 N/A N/A 16-33% SG LEVEL 42 N/A N/A 16% SG LEVEL CAS F.0.3 8 8% SG LEVEL CAS F.0.3 9 12% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 17 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
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ORIGINATOR DATE REV: 93 1 CMM 8 16 94 lIR REVIEWER VERIFIER DATE 94 LFP 2.2.5 EOP Indicated Values (shown as Setpoint below)
(Ref. 3.3.3, 3.6.3)
EOP NUMBER AND TITLE: TRIP-1, REACTOR TRIP OR SAFETY INJECTION ERG NUMBER AND TITLE: E-0, REACTOR TRIP OR SAFETY INJECTION EOP STEP ERG FOOTNOTE SETPOINT STEP 19 16 8 8% SG LEVEL 19 16 9 12% SG LEVEL 19 16 N/A 16-3 !% SG LEVEL 23 19 8 8% SG LEVEL 23 19 9 12% SG LEVEL 23 19 N/A 16-33% SG LEVEL 36 25 8 8% SG LEVEL 37 28 8 8% SG LEVEL 37 N/A N/A 16-33% SG LEVEL 42 N/A N/A 16% SGLEVEL CAS F.0.3 8 8% SG LEVEL CAS F.0.3 9 12% SGLEVEL
/
CALCULATION CONTINUATION/ SHEET: 18 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I 1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 EOP NUMBER AND TITLE: TRIP-2, REACTOR TRIP
RESPONSE
ERG NUMBER AND TITLE: ES-0.1, REACTOR TRIP
RESPONSE
BOP STEP ERG FOOTNOTE SETPOINT STEP 4 1 3 8% SG LEVEL 4 5 N/A 12% SG LEVEL 13 6 3 8% SG LEVEL 13 6 N/A 16-33% SG LEVEL 16 N/A 3 16% SG LEVEL 24 12 N/A 16-33% SG LEVEL 30 12 N/A 16-33% SG LEVEL BOP NUMBER AND TITLE: TRIP-3, REACTOR TRIP
RESPONSE
ERG NUMBER AND TITLE: ES-1.1, SI TERMINATION EOP STEP ERG FOOTNOTE SETPOINT STEP 25 20 11 8% SG LEVEL 25 20 12 12% SG LEVEL 25 20 N/A 16-33% SG LEVEL 29 20 N/A 16-33% SG LEVEL 33 27 N/A 16-33% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 19 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I lIRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 EOP NUMBER AND TITLE: TRIP-4, NATURAL CIRCULATION COOLDOWN ERG NUMBER AND TITLE: ES-0.2, NATURAL CIRCULATION COOLDOWN EOP STEP ERG FOOTNOTE SETPOINT STEP 7 N/A N/A 8% SG LEVEL 7 N/A N/A 12% SG LEVEL 7 N/A N/A 16-33% SG LEVEL 12 6 N/A 16-33% SG LEVEL EOP NUMBER AND TITLE: SGTR-1 STEAM GENERATOR TUBE RUPTURE ERG NUMBER AND TITLE: E-3, STEAM GENERATOR TUBE RUPTURE EOP STEP ERG FOOTNOTE SETPOINT STEP 2 N/A N/A 16% SG LEVEL 4 3 N/A 16% SG LEVEL 6 4 4 8% SG LEVEL 6 4 5 12% SG LEVEL 10 N/A N/A 16% SG LEVEL 11 7 4 8% SG LEVEL 11 7 5 12% SG LEVEL 11 7 N/A 16-33% SG LEVEL 29 20 4 8% SG LEVEL 29 20 5 12% SG LEVEL 47 N/A N/A 16-33% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 20 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 EOP NUMBER AND TITLE: SGTR-2, POST SGTR COOLDOWN USING BACKFILL ERG NUMBER AND TITLE: ES-3.1, POST SGTR COOLDOWN USING BACKFILL (ALSO ES-3.2)
EOP STEP ERG FOOTNOTE SETPOINT STEP 4 3 N/A 16% SG LEVEL 8 4 6 8% SG LEVEL 8 4 7 12% SG LEVEL 8 4 N/A 16-33% SG LEVEL 13 6 22 16% SG LEVEL 13 6 9 62% SG LEVEL 13 6 10 53% SG LEVEL 41 4 6 8% SG LEVEL 41 4 7 12% SG LEVEL 41 4 N/A 16-33% SG LEVEL 47 8 23 16% SG LEVEL 47 8 24 62% SG LEVEL 47 8 25 53% SG LEVEL 55 N/A N/A 8% SG LEVEL 55 N/A N/A 12% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 21 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IR~ I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 EOP NUMBER AND TITLE: SGTR-3, SGTR WI1H LOCA-SUBCOOLED RECOVERY ERG NUMBER AND TITLE: ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT SUB-COOLED RECOVERY EOP STEP ERG FOOTNOTE SETPOINT STEP 7 7 6 8% SG LEVEL 7 7 7 12% SG LEVEL 14 10 6 8% SG LEVEL 14 10 7 12% SG LEVEL 14 10 7 16-33% SG LEVEL 46 35 44 16% SG LEVEL 46 35 45 62% SG LEVEL 46 35 46 53% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 22 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 EOP NUMBER AND TITLE: SGTR-4, SGTR WITH LOCA SATURATED RECOVERY ERG NUMBER AND TITLE; ECA-3.2, SGTR WITH LOCA SATURATED RECOVERY DESIRED EOP STEP ERG FOOTNOTE SETPOINT STEP 4 3 5 8% SG LEVEL 4 3 6 12% SG LEVEL 8 N/A N/A 8% SG LEVEL 8 N/A N/A 12% SG LEVEL 11 5 5 8% SG LEVEL 11 5 6 12% SG LEVEL 11 N/A N/A 16-33% SG LEVEL 32 N/A N/A 16-33% SG LEVEL 49 29 41 16% SG LEVEL 49 29 42 62% SG LEVEL 49 29 43 53% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 23 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I 1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 EOP NUMBER AND TITLE: SGTR-5, SGTR WITHOUT PZR PRESSURE CONTROL ERG NUMBER AND TITLE: ECA-3.3, SGTR WITHOUT PZR PRESSURE CONTROL EOP STEP ERG FOOTNOTE SETPOINT STEP 1 1 1 59% SG LEVEL 1 1 2 58% SG LEVEL 5 5 5 8% SGLEVEL 5 5 6 ...
12% SG LEVEL 5 N/A N/A 16-33% SG LEVEL 7 7 5 8% SG LEVEL 7 7 6 12% SG LEVEL 28 N/A N/A 16% SG LEVEL 35 29 N/A 16% SG LEVEL 35 29 1 62% SG LEVEL 35 29 2 53% SG LEVEL EOP NUMBER AND TITLE: LOSC-2 MULTIPLE SG DEPRESSURIZATION ERG NUMBER AND TITLE: ECA-2.1 UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS EOP STEP ERG FOOTNOTE SETPOINT STEP 6 CAUT2 3 8% SG LEVEL 6 CAUT2 4 12% SG LEVEL 7 2 33% SG LEVEL 25 24 33% SG LEVEL 35 16-33% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 24 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 EOP NUMBER AND TITLE: LOPA-1, LOSS OF ALt AC POWER ERG NUMBER AND TITLE: ECA-0.0, LOSS OF ALL AC POWER BOP STEP ERG FOOTNOTE SETPOINT STEP 25 N/A N/A 16-33% SG LEVEL 25 13 5 8% SGLEVEL 25 13 4 12% SG LEVEL 28 N/A N/A 16-33% SG LEVEL
- 32 50 50 50 N/A 13 13 N/A N/A 4
5 N/A 16-33% SG LEVEL 8% SG LEVEL 12% SG LEVEL 16-33% SG LEVEL 54 N/A N/A 8% SG LEVEL 54 N/A N/A 12% SG LEVEL 55 N/A N/A 8% SG LEVEL 55 N/A N/A 12% SG LEVEL 57 16 4 8% SG LEVEL 57 16 5 12% SG LEVEL 65 N/A N/A 16-33% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 25 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 EOP NUMBER AND TITLE: LOPA-2, LOSS OF ALL AC POWER/SI NOT REQUIRED ERG NUMBER AND TITLE: ECA-0.1, LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED EOP STEP ERG FOOTNOTE SETPOINT STEP 10 7 9 8% SG LEVEL 10 7 10 12% SG LEVEL 11 N/A N/A 16-33% SG LEVEL 25 N/A N/A 16-33% SG LEVEL EOP NUMBER AND TITLE: LOPA-3, LOSS OF ALL AC POWER/SI REQUIRED ERG NUMBER AND TITLE: ECA-0.2, LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED EOP STEP ERG FOOTNOTE SETPOINT STEP 13 6 3 8% SG LEVEL 13 6 4 12% SG LEVEL 14 N/A N/A 16-33% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 26 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 IlIRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 EOP NUMBER AND TITLE: LOCA-1, LOSS OF REACTOR COOLANT ERG NUMBER AND TITLE: E-1, LOSS OF REACTOR OR SECONDARY COOLANT EOP STEP ERG FOOTNOTE SETPOINT STEP 4 3 4 8% SG LEVEL 4 3 5 12% SG LEVEL 5 N/A N/A 16-33% SG LEVEL 11 N/A N/A 16% SG LEVEL
- 12 12 6
6 EOP NUMBER AND TITLE:
4 5
8% SG LEVEL 12% SG LEVEL LOCA-2, LOSS OF REACTOR COOLANT ERG NUMBER AND TITLE: ES-1.2, LOSS OF REACTOR OR SECONDARY COOLANT EOP STEP ERG FOOTNOTE SETPOINT STEP 8 6 5 8% SG LEVEL 8 6 6 12% SG LEVEL 9 6 6 16-33% SG LEVEL 26 N/A 16-33% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 27 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I 1IR1 I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 EOP NUMBER AND TITLE: LOCA-5, LOSS OF EMERGENCY RECIRCULATION ERG NUMBER AND TITLE: ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION EOP STEP ERG FOOTNOTE SETPOINT STEP 5 5 33 8% SG LEVEL 5 5 34 12% SG LEVEL 5 5 33 16-33% SG LEVEL EOP NUMBER AND TITLE: FRSM-1, RESPONSE TO NUCLEAR POWER
- ERG NUMBER AND TITLE:
BOP STEP ERG STEP GENERATION FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION-ATWS FOOTNOTE SETPOINT 10 6 5 8% SG LEVEL 10 6 6 12% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 28 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 EOP NUMBER AND TITLE: FRCC-1, RESPONSE TO INADEQUATE CORE COOLING ERG NUMBER AND TITLE: FR-C.1, RESPONSE TO INADEQUATE CORE COOLING EOP STEP ERG FOOTNOTE SETPOINT STEP 18 9 5 8% SG LEVEL 18 9 6 12% SG LEVEL 18 9 N/A 16-33% SG LEVEL IE~,
19 9 5 8% SG LEVEL 19 9 6 12% SG LEVEL 19 N/A N/A 16-33% SG LEVEL 22 NOT 11 N/A 8% SG LEVEL 22 NOT 11 N/A 12% SG LEVEL EOP NUMBER AND TITLE: FRCC-2, RESPONSE TO DEGRADED CORE COOLING ERG NUMBER AND TITLE: FR-C.2, RESPONSE TO DEGRADED CORE COOLING EOP STEP ERG FOOTNOTE SETPOINT STEP 20 9 11 8% SG LEVEL 20 9 12 12% SG LEVEL 21 9 N/A 16-33% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 29 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 l CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 EOP NUMBER AND TITLE: FRTS-1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK ERG NUMBER AND TITLE: FR-P.1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK EOP STEP ERG FOOTNOTE SETPOINT STEP 4 1 2 8% SG LEVEL 4 1 3 12% SG LEVEL EOP NUMBER AND TITLE: FRTS-2, RESPONSE TO ANTICIPATED PRESSURIZED THERMAL SHOCK ERG NUMBER AND TITLE: FR-P.2, RESPONSE TO ANTICIPATED PRESSURIZED THERMAL SHOCK EOP STEP ERG FOOTNOTE SETPOINT STEP 4 1 2 8% SG LEVEL 4 1 3 12% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 30 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 IlIRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 BOP NUMBER AND TITLE: FRHS-1, RESPONSE TO LOSS OF SECONDARY HEAT SINK ERG NUMBER AND TITLE: FR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK BOP STEP ERG FOOTNOTE SETPOINT STEP 18 8 6 8% SG LEVEL 18 8 7 12% SG LEVEL 19 8 N/A 8% SG LEVEL 19 8 N/A 12% SG LEVEL 34 20 6 8% SG LEVEL 34 20 7 12% SG LEVEL BOP NUMBER AND TITLE: FRHS-2, RESPONSE TO SG OVERPRESSURE ERG NUMBER AND TITLE: FR-H.2, RESPONSE TO SG OVERPRESSURE BOP STEP ERG FOOTNOTE SETPOINT STEP 4 3 2 92% SG LEVEL 4 3 3 91% SG LEVEL 5 CAUT4 2 92% SG LEVEL 5 CAUT4 3 91% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 31 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 lIRl I
,REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 EOP NUMBER AND TITLE: FRHS-3, RESPONSE TO SG HIGH LEVEL ERG NUMBER AND TITLE: FR-H.3, RESPONSE TO SG HIGH LEVEL EOP STEP ERG FOOTNOTE SETPOINT STEP 1 CAUTl 1 91% SG LEVEL 1 CAUTl 2 92% SG LEVEL 2 1 3 59% SG LEVEL 3 1 3 59% SG LEVEL 5 4 1 91% SG LEVEL 5 4 2 92% SG LEVEL 5 4 5 16-33% SG LEVEL 10 9 N/A 16% SG LEVEL 10 9 N/A 33% SG LEVEL 10 9 N/A 16-33% SG LEVEL EOP NUMBER AND TITLE: FRHS-4, RESPONSE TO LOSS OF SG ATMOSPHERIC AND CONDENSER DUMP VALVES ERG NUMBER AND TITLE: FR-H.4, RESPONSE TO LOSS OF NORMAL STEAM RELEASE CAPABILITIES EOP STEP ERG FOOTNOTE SETPOINT STEP 2 1 2 92% SG LEVEL 2 1 3 91% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 32 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 I 1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 EOP NUMBER AND TITLE: FRHS-5, RESPONSE TO SG LOW LEVEL ERG NUMBER AND TITLE: FR-H.5, RESPONSE TO SG LOW LEVEL EOP STEP ERG FOOTNOTE SETPOINT STEP 1 NOTE 1 1 8% SG LEVEL 1 NOTE 1 2 12% SG LEVEL 3 1 1 8% SG LEVEL 3 1 2 12% SG LEVEL 9 4 1 8% SG LEVEL 9 4 2 12% SG LEVEL 12 N/A N/A 8% SG LEVEL 12 N/A N/A 12% SG LEVEL 13 N/A N/A 8% SG LEVEL 13 N/A N/A 12% SG LEVEL EOP NUMBER AND TITLE: APPX-4, POST SI RESTORATION ERG NUMBER AND TITLE: N/A EOP STEP ERG FOOTNOTE SETPOINT STEP 4 N/A N/A 16% SG LEVEL
CALCULATION CONTINUATION/ SHEET: 33 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13.193 1 CMM 8.116.194 11-IR~ I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 2.2.6 ERG Footnotes (Ref. 3.3.3 )
E-0, REACTOR TRIP OR SAFETY INJECTION (EOP TRIP-1)
(8) Enter plant specific value showing SG level just in the narrow range, including allowances for normal channel accuracy.
(9) Enter plant specific value showing SG level just in narrow range, including allow.ances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%.
ES-0.1, REACTOR TRIP RESPONSE (EOP TRIP-2)
(3) Enter plant specific value showing SG level just in the narrow range, including
- allowances for normal channel accuracy.
ES-0.2, NATURAL CIRCULATION COOLDOWN (EOP TRIP-4)
(3) Enter plant specific value corresponding to no-load SG level.
ECA-0.0, LOSS OF ALL AC POWER (EOP LOPA-1)
(4) Enter. plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy.
(5) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%.
ECA-0.1, LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED (EOP LOPA-2)
(9) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy.
(10) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg proces errors, not to exceed 50%.
CALCULATION CONTINUATION/ SHEET: 34 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 L~
ECA-0.2, LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED (BOP LOPA-3)
(3) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy.
(4) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, _and reference leg process errors, not to exceed 50%.
E-1, LOSS OF REACTOR OR SECONDARY COOLANT (BOP LOCA-1)
(4) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy.
(5) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%.
ES-1.1, SI TERMINATION (BOP TRIP-3)
(11) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy.
(12) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, *not to exceed 50%.
ES-1.2, POST LOCA COOLDOWN AND DEPRESSURIZATION (EOP LOCA-2)
(5) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy.
(6) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%.
ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION (BOP LOCA-5)
(33) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy.
(34) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%.
CALCULATION CONTINUATION/ SHEET: 35 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CN001-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 1Il2.I ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS (EOP WSC-1)
(3) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy.
(4) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%.
E-3, STEAM GENERATOR TUBE RUPTURE (EOP SGTR-1)
(4) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy.
(5) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%. An upper limit of 50% is imposed to ensure some margin to SG overfill for control of feed flow.
ES-3.1, POST-SGTR COOLDOWN USING BACKJ:'ILL (EOP SGTR-2)
(6) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy.
(7) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%. An upper limit of 50% is imposed to ensure some margin to SG overfill for control of feed flow.
(9) Plant specific value corresponding to high-high SG level setpoint. This value was selected to provide margin for filling the ruptured SG with cold feed flow while ensuring SG overfill does not occur.
(10) Plant specific value corresponding to high-high SG level setpoint including allowances for post accident transmitter errors and reference leg process errors not less than 50%.
This value wa.S selected to provide margin for filling the ruptured SG with cold feed flow to cool the metal in the upper regions while ensuring SG overfill does not occur.
A lower limit of 50% is imposed to provide margin between uncovering the U tubes and terminating feed flow on high level.
(22) Enter plant specific value showing SG level greater than the AFW actuation setpoint (23) Enter either the plant specific value showing SG level just in range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50% or the AFW actuation setpoint, whichever is greater.
CALCULATION CONTINUATION/ SHEET: 36 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94.
(24) Enter plant specific value corresponding to high-high SG level setpoint, minus 5% for operating margin.
(25) Enter plant specific value corresponding to high-high SG level setpoint, minus 5% for operating margin, including allowances for post accident transmitter errors and reference leg process *errors, not less than 50%.
ES-3.2, POST-SGTR COOLDOWN USING BLOWDOWN (EOP SGTR-4)
(6) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy.
(7) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed. 50%. An upper limit of 50% is imposed to ensure some margin to SG overfill for control of feed flow.
(22) Enter plant specific value showing SG level greater than the AFW actuation setpoint (23) Enter either the plant specific value showing SG level just in range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50% or the AFW actuation setpoint, whichever is greater.
(24) Enter plant specific value corresponding to high-high SG level setpoint, minus 5% for operating margin.
(25) Enter plant specific value corresponding to high-high SG level setpoint, minus 5% for operating margin, including allowances for post accident transmitter errors and reference leg process errors, not less than 50%.
ES-3.3, POST-SGTR COOLDOWN USING STEAM DUMP (EOP SGTR-2)
(6) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy.
(7) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%. An upper limit of 50% is imposed to ensure some margin to SG overfill for control of feed flow.
(27) Enter plant specific value. showing SG level greater than the AFW actuation setpoint.
(28) Enter either the plant specific value showing SG level just in range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50% or the AFW actuation setpoint, whichever is
- (29) greater.
Enter plant specific value corresponding to high-high SG level setpoint, minus 5% for operating margin.
CALCULATION CONTINUATION/ SHEET: 37 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 (30) Enter plant specific value corresponding to high-high SG level setpoint, minus 5% forts....
operating margin, including allowances for post accident transmitter errors and reference leg. process errors, not less than 50%.
ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT SUBCOOLED RECOVERY DESIRED (EOP SGTR-3)
(6) Enter plant specific value showing SG ~evel just in narrow range, including allowances for normal channel accuracy.
(7) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg proces errors, not to exceed 50%.. An upper limit of 50% is imposed to ensure some margin to SG overfill for control of feed flow.
(43) Enter plant specific value showing SG level greater than the AFW actuation setpoint (44) Enter either the plant specific value showing SG level just in range, including allowances for normal channel accuracy, -post accident transmitter errors, and reference leg process errors, not to exceed 50% or the AFW actuation setpoint, whichever is greater.
(45) Enter plant specific value corresponding to high-high SG level setpoint, minus 5% for operating margin.
(46) Enter plant specific value corresponding to high-high SG level setpoint, minus 5% for operating margin, including allowances for post accident transmitter errors and reference leg process errors, not less than 50%.
ECA-3.2, SGTR WITH LOSS OF REACTOR COOLANT-SATURAIBD RECOVERY DESIRED (EOP SGTR-4)
(5) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy.
(6) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg proces errors, not to exceed 50%. An upper limit of 50% is imposed to ensure some margin to SG overfill for control of feed flow.
(40) Enter the plant specific value showing SG level greater than the AFW actuation setpoint.
(41) Enter either the plant specific value showing SG level just in range, including allowances for normal channel accuracy, post accident transmitter errors, and reference
- (42) leg process errors, not to exceed 50% or the AFW actuation setpoint, whichever is greater.
Enter plant specific value corresponding to high-high SG level setpoint, minus 5% for operating margin.
CALCULATION CONTINUATION/ SHEET: 38 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 (43) Enter plant specific value corresponding to high-high SG level setpoint, minus 5% for operating margin, including allowances for post accident transmitter errors and reference leg process errors, not less than 50%.
ECA-3.3, SGTR WITHOUT PRESSURIZER PRESSURE CONTROL (EOP SGTR-5)
(1) Enter plant specific value corresponding to high-high SG level setpoint, minus 5% for operating margin.
(2) Enter plant specific value corresponding to high-high SG level setpoint, minus 5% for operating margin, including allowances for post accident transmitter errors and reference leg process errors, not less than 50%.
(5) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy.
(6) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg proces~
errors, not to exceed 50%. An upper limit of 50% is imposed to ensure some margin to SG overfill for control of feed flow.
(33) Enter plant specific value showing SG level greater than the AFW actuation setpoint.
(34) Enter either the plant specific value showing SG level just in range, including allowances for normal channel accuracy, post accident transmitter errors~ and reference leg process errors, not to exceed 50% or the AFW actuation setpoint, whichever is greater.
(35) Enter plant specific value corresponding to high-high SG level setpoint, minus 5% for operating margin.
(36) Enter plant specific value corresponding to high-high SG level setpoint, minus 5% for operating margin, including allowances for post accident transmitter errors and reference leg process errors, not less than 50%.
FR-C.l, RESPONSE TO INADEQUATE CORE COOLING (EOP FRCC-1)
(5) Enter plant specific value showing SG level just in the narrow range, including allowances for normal channel acc~racy.
(6) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50% .
CALCULATION CONTINUATION/ SHEET: 39 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 FR-C.2, RESPONSE TO DEGRADED CORE COOLING (EOP FRCC-2)
(11) Enter plant specific value showing SG level just in the narrow range, including allowances for normal channel accuracy.
(12) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%.
FR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK (EOP FRHS-1)
(6) Enter plant specific value showing SG level just in the narrow range, including allowances for normal channel accuracy.
(7) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%.
FR-H.2, RESPONSE TO SG OVERPRESSURE (EOP FRHS-2)
(2) Enter plant specific value corresponding to SG level at the upper tap, including allowances for normal channel accuracy.
(3) Enter plant specific value corresponding to SG level at the upper tap, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors.
FR-H.3, RESPONSE TO SG HIGH LEVEL (EOP FRHS-30 (1) Enter plant specific value corresponding to SG level at the upper tap, including allowances for normal channel accuracy.
(2) Enter plant specific value corresponding to SG level at the upper tap, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors.
(3) Enter plant specific value corresponding to SG high-high level feedwateI i.mlation setpoint.
(4) Enter plant specific value showing SG level just in the narrow range, including allowances for normal channel accuracy.
(5) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process
- errors, not to exceed 50%.
CALCULATION CONTINUATION/ SHEET: 40 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IR.1 I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 FR-H.4, RESPONSE TO LOSS OF NORMAL STEAM RELEASE CAPABILITIES (EOP FRHS-4)
(2) Enter plant specific value corresponding to SG level at the upper tap, including allowances for normal channel accuracy.
(3) Enter plant specific value corresponding to SG level at the upper tap, including
- allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors.
FR-H.5, RESPONSE TO SG LOW LEVEL (EOP FRHS-5)
(1) Enter plant specific value showing SG level just in the narrow range, including allowances for normal channel accuracy.
(2) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%.
FR-P.l, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK (EOP FRTS-1)
(2) Enter plant specific value showing SG level just in the narrow range, including allowances for normal channel accuracy.
(3) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%. -
FR-P.2, RESPONSE TO ANTICIPATED PRESSURIZED THERMAL SHOCK (EOP FRTS-2)
(2) Enter plant specific value showing SG level just in the narrow range, including allowances for normal channel accuracy.
(3) Enter plant specific value showing SG level just in narrow range, including allowances for normal channd accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%.
CALCULATION CONTINUATION/ SHEET: 41 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION-ATWS (EOP FRSM-1)
(5) Enter plant specific value showing SG level just in the narrow range, including allowances for normal channel accuracy.
(6) Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%.
EOP Accident Terminology Clarification When Emergency Response Guideline footnotes specify inclusion of normal, "post accident transmitter errors, and reference leg heat up " effects, this calculation applies the total channel uncertainties for "accident" conditions. This channel uncertainty includes all channel device uncertainties including environmental effects as well as non instrument related process uncertainties / bias terms. Bias terms are only used in this evaluation when they are applicable to the direction of interest being protected. Therefore, when an increasing value is established based on a high limit, the random Channel uncertainties and negative bias terms are applicable. When a decreasing value is established based on a low limit, the random Channel uncertainties and positive bias terms are applicable. Each ERG footnote is evaluated to determine whether normal or accident uncertainties are applicable.
Post Accident Indication uncertainties were calculated in the setpoint calculation. These values are based on the total channel uncertainty including worst case reference accuracy shifts applicable for "up to a year following a design basis event". The transmitter Post DBE effect used to develop this channel uncertainty assumes that accident temperature and radiation effects are no longer applicable. The post accident uncertainties were provided in the uncertainty calculation for information only and are not considered appropriate to the time duration that is applicable for the Emergency Operating Procedures .
CALCULATION CONTINUATION/ SHEET: 42 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: sc-c:trno1-01
REFERENCE:
ORIGINATOR- DATE REV: CMM 12/13/93 1 CMV 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94
3.0 REFERENCES
3.1 Setpoint Procedures 3.1.1 SC.DE-TS.ZZ-1904(0) Rev 0, Salem Unit 1 and 2 Technical Standard for Setpoints 3.1.2 DE-TS.ZZ-1002(0) Rev 2, Technical Standard Instrument Calibration Data Cards 3.1.3 S-C-RCP-CDC-0440 Rev 2, Westinghouse Setpoint Methodology for Protection Systems 3.1.4 VTD-304209 Rev 19, Salem Nuclear GS Units 1 & 2, Precautions Limitations and Setpoints 3.2 Updated Final Safety Analysis Report (UFSAR) 3.2.1 Section 7.2, Reactor Trip System, Table 7.2-7 SG Level Control and Protection System 3.2.2 Section 7.3, Engineered Safety Features Instrumentation, Table 7.3-1 Process Instrumentation for RPS and ESF Actuation
.3.2.3 Section 7.4, Systems Required for Safe Shutdown, Section 7.4.2 Cold Shutdown Outside the Control Room 3.2.4 Section 7.5, Tables 7.5.1 "Main Control Room Indicators Available to the Operator."
3.2.5 Section 7.5, Table 7.5.2 "Main Control Room Indicators Available to the Operator to Monitor Significant Plant Parameters During Normal Operation."
3.2.6 Section 7.5, Table 7.5-3 Index Type "A" Variables 3.2.7 Section 7.5, Table 7.5-4 Summary of Instrumentation Complianc~ with RG 1.97 3.2.8 Section 15.1, Table 15.1.3 Trip Points and Time Delays to Trip Assumed in the Accident Analysis 3.3 Technical Specification /EOP Design Basis 3.3.1 Unit 1 and 2 Salem Technical Specifications 3.3.1.1 Section 3/4.3.2 Engineered Safety Feature Actuation System Instrumentation 3.3.1.2 Section 3/4.3.3 Subsection 3.3.3.7 Accident Monitoring Instrumentation, Limiting Condition for Operation 3.3.1.3 Subsection 4.3.3.7 Surveillance Requirements 3.3.1.4 Table 3.3-3 Engineered Safety Feature Actuation System Instrumentation 3.3.1.5 Table 3.3.4 Engineered Safety Feature Actuation Instrumentation Trip Setpoints 3.3.1.6 Table 3.3-11 Accident Monitoring Instrumentation 3.3.1.7 Table 4.3-11 Surveillance Requirements for Accident Monitoring Instrumentation 3.3.1.8 Table 4.3-2 Engineered Safety Feature Actuation System Instrumentation Surveillance Requirements 3.3.1.9 Section 2.2, Table 2.2.-1 Reactor Trip Instrumentation
- 3.3.2 SECL-92-049 Westinghouse Safety Evaluation 3.3.3 WOG-91-018 Westinghouse Owners Group Emergency Response Guidelines (Rev lB) 3.3.3.1 (ERG) FR-C.l, Response to Inadequate Core Cooling
CALCULATION CONTINUATION/ SHEET: 43 OPS~G. REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIG I v:
REVIEWER VERIFIER DATE 94 LFP 8 16 94 3.3.3.2 (ERG) FR-C.2, Response to Degraded Core Cooling 3.3.3.3 (ERG) FR-H.1, Response to Loss of Secondary Heat Sink 3.3.3.4 (ERG) FR-H.2, Response to SG Overpressure 3.3.3.5 (ERG) FR-H.3, Response to SG High Level 3.3.3.6 (ERG) FR-H.4, Response to Loss of Normal Steam Release Capabilities 3.3.3.7 (ERG) FR-H.5, Response to SG Low Level 3.3.3.8 (ERG) FR-S.1, Response To Nuclear Power Generation -ATWS 3.3.3.9 (ERG) FR-P.1, Response to Imminent Pressurized Thermal Conditions 3.3.3.10 (ERG) FR-P.2, Response to Anticipated Pressurized Thermal Shock 3.3.3.11 (ERG) E-1, Loss of Reactor or Secondary Coolant 3.3.3.12 (ERG) ES-1.2, Post LOCA Cooldown and Depressurization 3.3.3.13 (ERG) ECA-1.1, Loss of Emergency Coolant Recirculation 3.3.3.14 (ERG) ECA-0.0, Loss of All AC Power 3.3.3.15 (ERG) ECA-0.1, Loss of All AC Power Recovery Without SI 3.3.3.16 (ERG) ECA-0.2, Loss of All AC Power Recovery With SI 3.3.3.17 (ERG) E-2, Faulted Seam Generator Isolation 3.3.3.18 (ERG) ECA-2.1, Uncontrolled Depressurization of all SGs 3.3.3.19 (ERG) E-3, Steam Generator Tube Rupture 3.3.3.20 (ERG) ES-3.1, Post SGTR Cooldown Using Backfill 3.3.3.21 (ERG) ES-3.2, Post SGTR Cooldown Using Blowdown 3.3.3.22 (ERG) ES-3.3, Post SGTR Cooldown Using Steam Dumps 3.3.3.23 (ERG) ECA-3.1, SGTR with Loss of Reactor Coolant - Subcooled Recovery 3.3.3.24 (ERG) ECA-3.2, SGTR with LOCA Saturated Recovery Desired 3.3.3.25 (ERG) ECA-3.3, SGTR without Pressurizer Pressure Control 3.3.3.26 (ERG) E-0, Reactor Trip or Safety Injection 3.3.3.27 (ERG) ES-0.1, Reactor Trip Response 3.3.3.28 (ERG) ES-1.1, SI Termination 3.3.3.29 (ERG) ES-0.2, Natural Circulation Cooldown 3.4 Drawings 3.4.1 205302 A 8762-46 Sheet 3 of 3, Steam Generator Feed and Condensate P&ID 3.4.2 220029 B 9537 Reactor Prot & Process Cont. Systems SG Interconnections, Wiring Diagram 3.4.3 220031 B 9537 Reactor Prot & Process Cont. System SG Interconnections, Wiring Diagram 3.4.4 220033 B 9537 Reactor Prot & Process Cont. System SG Interconnections, Wiring Diagram 3.4.5 220034 B 9537 -12 Reactor Prot & Process Control Systems SG Interconnections 3.4.6 220053 B 9537 -15 Reactor Prot & Process Control Systems SG Interconnections 3.4.7 220056 B 9537-14 Reactor Prot & Process Control ~ystems SG Interconnections 3.4.8 613101 No 1 Unit, Stm Gen Feed & Cond No. 11 SG Level 1LT518, Logic Diagram 3.4.9 613102 No 1 Unit, Stm Gen Feed & Cond No 12 SG Level 1LT528, Logic Diagram 3.4.10 613103 No 1 Unit, Stm Gen Feed & Cond No 13 SG Level 1LT538, Logic Diagram
CALCULATION CONTINUATION/ SHEET: 44 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 3.4.11 613104 No 1 Unit, Stm Gen Feed & Cond No 14 SG Level 1LT548, Logic Diagram 3.4.12 613105 No 1 Unit, Stm Gen Feed & Cond No. 11 SG Level 1LT517, Logic Diagram 3.4.13 613106 No 1 Unit, Stm Gen Feed & Cond No 12 SG Level 1LT527, Logic Diagram 3.4.14 613107 No 1 Unit, Stm Gen Feed & Cond No.13 SG Level 1LT537, Logic Diagram 3.4.15 613108 No 1 Unit, Stm Gen Feed & Cond No 14 SG Level 1LT547, Logic Diagram 3.4.16 613129 No 1 Unit, Stm Gen Feed & Cond No 11 SG Level 1LT519, Logic Diagram 3.4.17 613130 No 1 Unit, Stm Gen Feed & Cond No 12 SG Level 1LT529, Logic Diagram 3.4.18 613131 No 1 Unit, Stm Gen Feed & Cond No 13 SG Level 1LT539, Logic Diagram 3.4.19 613132 No 1 Unit, Stm Gen Feed & Cond No 14 SG Level 1LT549, Logic Diagram 3.4.20 613101 No 1 Unit, Stm Gen Feed & Cond No 11 SG Level 1LT518, Logic Diagram 3.4.21 613102 No 1 Unit, Stm Gen Feed & Cond No 12 SG Level 1LT528, Logic Diagram 3.4.22 613103 No 1 Unit, Stm Gen Feed & Cond No 13 SG Level 1LT538, Logic Diagram 3.4.23 613104 No 1 Unit, Stm Gen Feed & Cond No 14 SG Level 1LT548, Logic Diagram 3.4.24 623101 No 2 Unit, Stm Gen Feed & Cond No 21 SG Level 2LT518, Logic Diagram 3.4.25 623102 No 2 Unit, Stm Gen Feed & Cond No 22 SG Level 2LTS28, Logic Diagram 3.4.26 623103 No 2 Unit, Stm Gen Feed & Cond No 23 SG Level 2LT538, Logic Diagram 3.4.27 623104 No 2 Unit, Stm Gen Feed & Cond No 24 SG Level 2LT548, Logic Diagram 3.4.28 623105 No 2 Unit, Stm Gen Feed & Cond No 21 SG Level 2LT517, Logic Diagram 3.4.29 623106 No 2 Unit, Stm Gen Feed & Cond No 22 SG Level 2LT527, Logic Diagram 3.4.30 623107 No 2 Unit, Stm Gen Feed & Cond No 23 SG Level 2LT537, Logic Diagram 3.4.31 623108 No 2 Unit, Stm Gen Feed & Cond No 24 SG Level 2LT547, Logic Diagram 3.4.32 623129 No 2 Unit, Stm Gen Feed & Cond No 21 SG Level 2LT519, Logic Diagram 3.4.33 623130 No 2 Unit, Stm Gen Feed & Cond No 22 SG Level 2LT529, Logic Diagram 3.4.34 623131 No 2 Unit, Stm Gen Feed & Cond No 23 SG Level 2LT539, Logic Diagram 3.4.35 623132 No 2 Unit, Stm Gen Feed & Cond No 24 SG Level 2LT549, Logic Diagram 3.4.36 218162 A 9783-35 No 1 Unit Control Room Annunciator Designations 3.4.37 211301 B 9508-11 No 1 Unit RC No 11 SG Level and Steam Flow Instrument Schematic 3.4.38 211302 B 9508-12 No 1 Unit, RC No 12 SG Level and Steam Flow Instrument Schematic 3.4.39 211303 B 9508-12 No 1 Unit RC No 13 SG Level and Steam Flow Instrument Schematic 3.4.40 211304 B 9508-11 No 1 Unit RC, No 14 SG Level and Steam Flow Instrument Schematic 3.4.41 240662 B 9656-10 No 2 Unit, RC No 21 SG Level and Steam Flow Instrument Schematic 3.4.42 240663 B 9656-10 No 2 Unit RC No 22 SG Level and Steam Flow Instrument Schematic 3.4.43 240664 B 9656-9 No 2 Unit, RC No 23 SG Level and Steam Flow Instrument Schematic 3.4.44 240665 B 9656-9 No 2 Unit, RC No 24 SG Level and Steam Flow Instrument Schematic 3.4.45 229928 A 1327-10 No 1 RC N-E & S-E Quadrants Ext.Tubing El 130'-0" Arrangement
. 3.4.46 229929 A 1327-13 No 1 RC N-W & S-W Quadrants Ext. Tubing El 130'-0" Arrangement 3.4.47 233026 A 1399-9 No 2 RC N-E & S-E Quadrants Ext. Tubing El 130'-0" Arrangement 3.4.48 233026 A 1399-9 No 2 RC N-W & S-W Quadrants Ext. Tubing El 130'-0" Arrangement 3.4.49 221056 B 9545-7 No 1 & 2 Units, Reactor Protection System, SG Trip Signals Loop Diagram 3.4.50 233609 B 9611-8 No 1 &2 Units, RC El 130' 11,12, 13, 14 SG Level Arrangement 3.4.51 203425 B 9790-7 No. 1 & 2 Units- Feedwater No. 11 SG Feedwater Flow Schematic
CALCULATION CONTINUATION/ SHEET: 45 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I11R~ I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 3.4.52 203426 B 9790-6 No. 1 & 2 Units- Feedwater No. 12 SG Feedwater Flow Schematic 3.4.53 203427 B 9790-6 No. 1 & 2 Units- Feedwater No. 13 SG Feedwater Flow Schematic 3.4.54 203428 B 9790-6 No. 1 & 2 Units- Feedwater No. 14 SG Feedwater Flow Schematic 3.4.55 205171Rev15, No. 1 Unit-Control Console Bezel 3.4.56 228476 Rev 12, No. 2 Unit-Control Console Bezel 3.5 Calculations and Support Documents 3.5.1 SC-MS-EQ49-001 Rev 5, Environmental Qualification for Rosemount Transmitters 3.5.2 S-C-ZZ-EEE-0625 Rev 0, Engineering Evaluation of M&TE 3.5.3 MMIS controlled database for Instrument Component Information 3.5.4 EQRR-0001 Rev 7, SGS Environmental Qualification Review Report 3.5.5 PSBP 138646 Rev 11, Westinghouse Rack Instruments 3.5.6 PSBP 312344, Dixon Edgewise Indicators 3.5.7 S-C-VAR-CEE-0811 Engineering Evaluation on the EPRI Drift Study 3.5.8 S-C-ZZ-CEE-0815 Engineering Evaluation for Acceptance Criteria for As Found Calibration Values for Salem Unit 1 & 2.
3.5.9 PSE&G VTD No 312351-03 Leeds and Northrup Recorder Specification, Speedomax 100 Series 3.5.10 DE-CB.CN-0015(0) CBD for Steam Generator Feedwater & Condensate System 3.5.11 DE-CB.RCP/SEC/SSP/SPL-0032 (Q) CBD for Reactor Protection Systems 3.5.12 DE-CB.115-0017 (Q) CBD for Electrical Systems 3.5.13 PSBP 301669 SCT Transmitter, Moore Industries, Inc.
3.5.14 S-C-VAR-CEE-0807 Rev 0, Engineering Evaluation of Salem Generating Units 1&2 Insulation Resistance Effects 3.5.15 2EC-3178 Pkg 2, Design Change Package 3.5.16 PSE-92-106 Letter from Westinghouse to Mr. J. A Nichols, S/G Water Level PMA Term Inaccuracies Dated June 18, 1992 3.5.17 Rosemount Manual 4631, April 1989, 1154 Series H Alphaline Pressure Transmitters 3.5.18 Electrical Cable Database 3.5.19 ASME Steam Tables 5th Edition 3.5.20 PSE-92-043, Westinghouse Letter ET-NSL-OPL-11-92-088 Dated February 18, 1992, Safety Analysis Limits 3.5.21 PSE&G VTD No. 301129 Issue 6, Rosemount Manual, Model 1153 Series D Alphaline Pressure Transmitter and Acceptance Test Specification 3.5.22 PSE-94-532 Safety Evaluation for an Increase in SG High-High Level Setpoint Analysis 3.5.23 VTD 317079-01, Environmental Design Criteria (EDC) 3.5.24 PSE-94-555 Westinghouse Letter Dated March 24, 1994,
Subject:
JPO for Overpower Operation (Excerpt Attachment 10.6) 3.5.25 PSBP 317093 Draft Seismic Safe Shutdown Equipment List (SSEL) Unit 1 3.5.26 PSBP 317095 Draft Seismic Safe Shutdown Equipment List (SSEL) Unit 2
CALCULATION CONTINUATION/ SHEET: 46 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 3.6 Procedures 3.6.1 Calibration Procedures 3.6.1.1 S1(2).IC-CC.RCP-0033(Q) 1(2)LT-517 Steam Generator Level Protection Channel IV 3.6.1.2 S1(2).IC-CC.RCP-0034(Q) 1(2)LT-518 Steam Generator Level Protection Channel III 3.6.1.3 S1(2).IC-CC.RCP-0035(Q) 1(2)LT-519 Steam Generator Level Protection Channel II 3.6.1.4 S1(2).IC-CC.RCP-0043(Q) 1(2)LT-527 Steam Generator Level Protection Channel N 3.6.1.5 S1(2).IC-CC.RCP-0044(Q) 1(2)LT-528 Steam Generator Level Protection Channel III 3.6.1.6 S1(2).IC-CC.RCP-0045(Q) 1(2)LT-529 Steam Generator Level Protection Channel II 3.6.1.7 S1(2).IC-CC.RCP-0053(Q) 1(2)LT-537 Steam Generator Level Protection Channel IV 3.6.1.8 S1(2).IC-CC.RCP-0054(Q) 1(2)LT-538 Steam Generator Level Protection Channel II,I 3.6.1.9 S1(2).IC-CC.RCP-0055(Q) 1(2)LT-539 Steam Generator Level Protection Channel II 3.6.1.10 S1(2).IC-CC.RCP-0063(Q) 1(2)LT-547 Steam Generator Level Protection Channel IV 3.6.1.11 S1(2).IC-CC.RCP-0064(Q) 1(2)LT-548 Steam Generator Level Protection Channel III 3.6.1.12 S1(2).IC-CC.RCP-0065(Q) 1(2)LT-549 Steam Generator Level Protection Channel II 3.6.1.13 S1(2).IC-SC-RCP-0033(Q) 1(2)LT-517 Steam Generator Level Protection Channel IV 3.6.1.14 S1(2).IC-SC-RCP-0034(Q) 1(2)LT-518 Steam Generator Level Protection Channel III 3.6.1.15 S1(2).IC-SC.RCP-0035(Q) 1(2)LT-519 Steam Generator Level Protection Channel II 3.6.1.16 S1(2).IC-SC.RCP-0043(Q) 1(2)LT-527 Steam Generator Level Protection Channel IV 3.6.1.17 S1(2).IC-SC.RCP-0044(Q) 1(2)LT-528 Steam Generator Level Protection Channel III 3.6.1.18 S1(2).IC-SC.RCP-0045(Q) 1(2)LT-529 Steam Generator Level Protection Channel II 3.6.1.19 S1(2).IC-SC.RCP-0053(Q) 1(2)LT-537 Steam Generator Level Protection Channel IV 3.6.1.20 S1(2).IC-SC.RCP-0054(Q) 1(2)LT-538 Steam Generator Level Protection Channel III 3.6.1.21 S1(2).IC-SC.RCP-0055(Q) 1(2)LT-539 Steam Generator Level Protection Channel II 3.6.1.22 S1(2).IC-SC.RCP-0063(Q) 1(2)LT-547 Steam Generator Level Protection Channel IV 3.6.1.23 S1(2).IC-SC.RCP-0064(Q) 1(2)LT-548 Steam Generator Level Protection Channel III 3.6.1.24 S1(2).IC-SC.RCP-0065(Q) 1(2)LT-549 Steam Generator Level Protection Channel II.
3.6.2 NC.DE-AP.ZZ-0007(0) Specialty Reviews 3.6.3 Emergency Operating Procedures (Rev 10) 3.6.3.1 EOP-FRCC-1 Response to Inadequate Core Cooling 3.6.3.2 EOP-FRCC-2 Response to Degraded Core Cooling 3.6.3.3 EOP-FRHS-1 Response to Loss of Secondary Heat Sink 3.6.3.4 EOP-FRHS-2 Response to SG Overpressure 3.6.3.5 EOP-FRHS-3 Response to SG High Level 3.6.3.6 EOP-FRHS-4 Response to Loss of SG Atm and Condenser Dump Valves 3.6.3.7 EOP-FRHS-5 Response to SG Low Level 3.6.3.8 EOP-FRSM-1 Response to Nuclear Power Generation 3.6.3.9 EOP-FRTS-1 Response to Imminent Pressurized Thermal Conditions
- 3.6.3.10 EOP-FRTS-2 Response to Anticipated Pressurized Thermal Shock 3.6.3.11 EOP-LOCA-1 Loss of Reactor Coolant 3.6.3.12 EOP-LOCA-2 Post LOCA Cooldown and Depressurization
CALCULATION CONTINUATION/ SHEET: 47 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
- CALC. No.: SC-CNOOl-01 ORIGINATOR. DATE REV: CMM 12/13/93
REFERENCE:
1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 3.6.3.13 EOP-LOCA-5 Loss of Emergency Recirculation 3.6.3.14 EOP-LOPA-1 Loss of All AC Power 3.6.3.15 EOP-LOPA-2 Loss of All AC Power Recovery without SI 3.6.3.16 EOP-LOPA-3 Loss of All AC Power Recovery with SI 3.6.3.17 EOP-LOSC-1 Loss of Secondary Coolant 3.6.3.18 EOP-LOSC-2 Multiple Steam Generator Depressurization 3.6.3.19 EOP-SGTR-1 Steam Generator Tube Rupture 3.6.3.20 EOP-SGTR-2 Post-SGTR Cooldown 3.6.3.21 EOP-SGTR-3 SGTR With LOCA, Subcooled Recovery 3.6.3.22 EOP-SGTR-4 SGTR with LOCA - Saturated Recovery 3.6.3.23 EOP-SGTR-5 SGTR without Pressurizer Pressure Control 3.6.3.24 EOP-TRIP-1 Reactor Trip or Safety Injection 3.6.3.25 EOP-TRIP-2 Reactor Trip Response 3.6.3.26 EOP-TRIP-3 Safety Injection Termination
- 3.6.3.27 EOP-TRIP-4 Natural Circulation Cooldown
CALCULATION CONTINUATION/ SHEET: 48 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 4.0 LOOP DIAGRAM 4.1 The Loop Diagram shown below is typical for the Comparator Setpoints, Configuration A and B. Refer to Calculation Section 1.2 for differences in Component IDs.
LT-517 TP-517-1 LC-517A-B/R SSPS Turbine SSPS Reactor Trip Trip High Level Low Low Level Override and Alarm
CALCULATION CONTINUATION/ SHEET: 49 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/i6/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 4.2 The Loop Diagram shown below is typical for the Comparator Setpoints, Configuration C.
Refer to Calculation Section 1.2 for differences in Component IDs.
LT-517 TP-517-1 LC-517A-C/R LC-517C SSPS Reactor Trip Low Level Mismatch Trip Interlock
CALCULATION CONTINUATION/ SHEET: 50 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 4.3 The Loop Diagram shown below is typical for the Control Room Indicator Loops, Configuration D. Refer to Calculation Section 1.2 for Component IDs.
LT-517 TP-517-1 LM-517A/R LM-517A TP-517-2 LI-517/R LI-517 Control Console
CALCULATION CONTINUATION/ SHEET: 51 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 4.4 The Loop Diagram shown below is typical for the Hot Shutdown Panel Indication, Configuration E. Refer to Calculation Section 1.2 for Component IDs.
LT-517 TP-517-1 LM-517A/R LM-517A
- LI-517A Hot Shutdown Panel Panel 213
CALCULATION CONTINUATION/ SHEET: 52 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
- CALC. No.: SC-CNoo*1-01 ORIGINATOR- DATE REV: CMM 12113/93
REFERENCE:
1 CMM 8116/94 I1IR'I I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94
- 4. 5 The Loop Diagram shown below is typical for the Recorder loops, Configuration F. Refer to Calculation Section 1.2 for Component IDs.
LT-519 TP-519-1 LC-519A-B/R LM-519M LM-500W/R LM-519B LA-5048 Control Console
CALCULATION CONTINUATION/ SHEET: 53 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE. REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 5.0 DESIGN INPUTS 5.1 General Design Inputs 5.1.1 Equipment Locations/ Environmental Parameters (Ref. 3.5.3, 3.5.4 )
Channel IV Device Description Location 1(2)LT-517 Transmitter Bldg 05, 15, Elev 130, Area 011, Panel 444-1(2)A 1(2)LT-527 Transmitter Bldg 05, 15, Elev 130, Area 010, Panel 444-1(2)F 1(2)LT-537 Transmitter Bldg 05, 15, Elev 130, Area 007, Panel 444-1(2)G 1(2)LT-547 Transmitter Bldg 05, 15, Elev 130, Area 009, Panel 444-1(2)M 1(2)LM-517A/R Conditioner Bldg 01, 12, Elev 122, Area 002
- 1(2)LM-527A/R 1(2)LM-537A/R 1(2)LM-547A/R 1(2)LM-517A 1(2)LM-527A 1(2)LM-537A Conditioner Conditioner Conditioner Isolator Isolator Isolator Bldg 01, 12, Bldg 01, 12, Bldg 01, 12, Bldg 01, 12, Bldg 01, 12, Bldg 01, 12, Elev Elev Elev Elev Elev Elev 122, 122, 078, 122, 122, 122, Area 002 Area 002 Area 005 Area 002 Area 002 Area 002 1(2)LM-547A Isolator Bldg 01, 12, Elev 122, Area 002 1(2)LI-517A/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LI-527A/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LI-537A/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LI-547A/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LI-517 Indicator Bldg 01, 12, Elev 122, Area 002 1(2)LI-527 Indicator Bldg 01, 12, Elev 122, Area 002 1(2)LI-537 Indicator Bldg 01, 12, Elev 122, Area 002 1(2)LI-547 Indicator Bldg 01, 12, Elev 122, Area 002 1(2)LC-517A-B /R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-527A-B/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-537A-B/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-547A-B/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-517C/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-527C/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-537C/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-547C/R Conditioner Bldg 01, 12, Elev 122, Area 002
- 1(2)LC-517A-B 1(2)LC-527A-B 1(2)LC-537A-B 1(2)LC-547A-B Comparator Comparator Comparator Comparator Bldg 01, 12, Bldg 01, 12, Bldg 01, 12, Bldg 01, 12, Elev Elev Elev Elev 122, 122, 122, 122, Area 002 Area 002 Area 002 Area 002
CALCULATION CONTINUATION/ SHEET: 54 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 1(2)LC-517C Comparator Bldg 01, 12, Elev 122, Area 002 1(2)LC-527C Comparator Bldg 01, 12, Elev 122, Area 002 1(2)LC-537C Comparator Bldg 01, 12, Elev 122, Area 002 1(2)LC-547C Comparator Bldg 01, 12, Elev 122, Area 002 1(2)LI-517A Indicator Bldg 01, 12, Elev 084, Area 015 1(2)LI-527A Indicator Bldg 01, 12, Elev 084, Area 015 1(2)LI-537A Indicator Bldg 01, 12, Elev 084, Area 015 1(2)LI-547A Indicator Bldg 01, 12, Elev 084, Area 015 Channel III Device Description Location 1(2)LT-518 Transmitter Bldg 05, 15, Elev 130, Area 011, Panel 444-1(2)B 1(2)LT-528 Transmitter Bldg 05, 15, Elev 130, Area 011, Panel 444-1(2)E
- 1(2)LT-538 1(2)LT-548 1(2)LM-518A/R 1(2)LM-528A/R 1(2)LM-538A/R 1(2)LM-548A/R Transmitter Transmitter Conditioner Conditioner Conditioner Conditioner Bldg 05, 15, Bldg 05, 15, Bldg 01, 12, Bldg 01, 12, Bldg 01, 12, Bldg 01, 12, Elev Elev Elev Elev Elev Elev 130, Area 007, 130, Area 009, 122, Area 002 122, Area 002 122, Area 002 122, Area 002 Panel Panel 444-1(2)H 444-1(2)L 1(2)LM-518 Isolator Bldg 01, 12, Elev 122, Area 002 1(2)LM-528 Isolator Bldg 01, 12, Elev 122, Area 002 1(2)LM-538 Isolator Bldg 01, 12, Elev 122, Area 002 1(2)LM-548 Isolator Bldg 01, 12, Elev 122, Area 002 1(2)LI-518A/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LI-528A/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LI-538A/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LI-548A/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LI-518 Indicator Bldg 01, 12, Elev 122, Area 002 1(2)LI-528 Indicator Bldg 01, 12, Elev 122, Area 002 1(2)LI-538 Indicator Bldg 01, 12, Elev 122, Area 002 1(2)LI-548 Indicator Bldg 01, 12, Elev 122, Area 002 1(2)LC-518A-B/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-528A-B/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-538A-B/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-548A-B/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-518C/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-528C/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-538C/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-548C/R Conditioner Bldg 01, 12, Elev 122, Area 002
CALCULATION CONTINUATION/ SHEET: 55 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 1(2)LC-518A-B . Comparator Bldg 01, 12, Elev 122, Area 002 1(2)LC-528A-B Comparator Bldg 01, 12, Elev 122, Area 002 1(2)LC-538A-B Comparator Bldg 01, 12, Elev 122, Area 002 1(2)LC-548A-B Comparator Bldg 01, 12, Elev 122, Area 002 1(2)LC-518C Comparator Bldg 01, 12, Elev 122, Area 002 1(2)LC-528C Comparator Bldg. 01, 12, Elev 122, Area 002 1(2)LC-538C Comparator Bldg 01, 12, Elev 122, Area 002 1(2)LC-548C Comparator Bldg 01, 12, Elev 122, Area 002 Channel II Device D~scri~tion Location 1(2)LT-519 Transmitter Bldg 05, 15, Elev 130, Area 011, Panel 444-1(2)C 1(2)LT-529 Transmitter Bldg 05, 15, Elev 130, Area 010, Panel 444-1(2)D 1(2)LT-539 Transmitter Bldg 05, 15, Elev 130, Area 007, Panel 444-1(2)J 1(2)LT-549 Transmitter Bldg 05, 15, Elev 130, Area 009, Panel 444-1(2)K 1(2)LM-519A/R Conditioner Bld_g 01, 12, Elev 122, Area 005 1(2)LM-529A/R Conditioner Bldg 01, 12, Elev 122, Area 005 1(2)LM-539A/R Conditioner Bldg 01, 12, Elev 122, Area 005 1(2)LM-549A/R Conditioner Bldg 01, 12, Elev 122, Area 005 1(2)LM-519A Isolator Bldg 01, 12, Elev 122, Area 002 1(2)LM-529A Isolator Bldg 01, 12, Elev 122, Area 002 1(2)LM-539A Isolator Bldg 01, 12, Elev 122, Area 002 1(2)LM-549A Isolator Bldg 01, 12, Elev 122, Area 002 1(2)LI-519/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LI-529 /R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)Ll-539/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LI-549 /R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LI-519 Indicator Bldg 01, 12, Elev 122, Area 002 1(2)LI-529 Indicator Bldg 01, 12, Elev 122, Area 002 1(2)LI-539 Indicator Bldg 01, 12, Elev 122, Area 002 1(2)LI-549 Indicator Bldg 01, 12, Elev 122, Area 002 1(2)LC-519A-B/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-529A-B/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-539A-B/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-549A-B/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LC-519A-B Comparator Bldg 01, 12, Elev 122, Area 002
- 1(2)LC-529A-B 1(2)LC-539A-B 1(2)LC-549A-B 1(2)LM-519A-C/R Comparator Comparator Comparator Conditioner Bldg 01, 12, Bldg 01, 12, Bldg 01, 12, Bldg 01, 12, Elev Elev Elev Elev 122, 122, 122, 122, Area Area Area Area 002 002 002 002
CALCULATION CONTINUATION/ SHEET: 56 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16~94 l1IR~ I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 1(2)LM-529A-C/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LM-539A-CiR Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LM-549A-C/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LM-519M Isolator Bldg 01, 12, Elev 122, Area 002 1(2)LM-529M Isolator Bldg 01, 12, Elev 122, Area 002 1(2)LM-539M Isolator Bldg 01, 12, Elev 122, Area 002 1(2)LM-549M Isolator Bldg 01, 12, Elev 122, Area 002 1(2)LM-500W/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LM-500X/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LM-500Y/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LM-500Z/R Conditioner Bldg 01, 12, Elev 122, Area 002 1(2)LM-519B Isolator Bldg 01, 12, Elev 122, Area 002 1(2)LM-529B Isolator Bldg 01, 12, Elev 122, Area 002 1(2)LM-539B Isolator Bldg 01, 12, Elev 122, Area 002
- 1(2)LM-549B 1(2)LA-5048 1(2)LA-5049 1(2)LA-5050 1(2)LA-5051 Isolator Recorder Recorder Recorder Recorder Bldg Bldg Bldg Bldg Bldg 01, 01, 01, 01, 01, 12, 12, 12, 12, 12, Elev Elev Elev Elev Elev 122, 122, 122, 122, 122, Area Area Area Area Area 002 002 002 002 002 AREA/ENVIRONMENTAL PARAMETERS IAREA TEMP ACC NORM/ NORM RADS I TEMP ACC RH RADS Containment 60-120°F 351.3 °F 20-90% 2".51E6* 2.93E7 05, 15 CAL; 70-90°F 100 % RH Steam
- ~* ..
Control 55-85°F N/A 20-90% N/A N/A Room CAL; 70°F RH 01, 12
- EDC Value for 40 year TID normal is overly conservative for purposes of this calculation, see Normal Radiation Assumption 6.3.1.
r fal
CALCULATION CONTINUATION/ SHEET: 57 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/1619* l1IR:J I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 5.1.2 Loop Power Supply (Vac: Ref. 3.5.12)(Vdc: Ref. 3.5.5, 3.4.1-3.4.7) 120 Vac Vital Instrument Bus, regulated +/-2%.
Rack Power Supplies (Ref. 3.4.1-3.4.7)
Manufacturer Westinghouse (Mfg Code H015)
Model Model 121 (4111085-001)
Rating 46 Vdc +/- 5% + 200 mV ripple 5.2 Process Design Inputs Maximum Normal Operating Pressure is 756.52 psig (Unit 2, Attachment A) and 759.2 psig (Unit 1, Attachment B)
Calibrated Span is 107.673 inWC (Unit 2, Attachment A) 107.746 inWC (Unit 1, Attachment B)
Process Measurement Accuracy from Ref. 3.5.16 (Attachment 10.4)
Normal Operating High Level 44% (Ref 3.1.4)
Normal Operating Low Level 33% (Ref 3.1.4)
High High Analytical Limit 75% (Attachment 10.5) 5.3 Transmitter Design Inputs Manufacturer and Model Numbers shown below are typical for all Transmitters listed in Section 1.2, unless otherwise noted.
5.3.1 Safety/Quality Designations (Ref. 3.5.3)
SFfY RLTD/QAR: SR CLASS/QGC: IE EQ:H, SEISMIC CAT: 1
CALCULATION CONTINUATION/ SHEET: 58 OPS~G
._i------------------r-------------t CALC. No.: SC-CNOOl-01 REVISION HISTORY SHEET
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CONT'D OH SHEET:
ORIGINATO REV: C c REVIEWER VERIFIER DATE AFS SJJ 1 11 94 LFP Performance Specification - Transmitter Range Code 4 (Ref. 3.5.17)
Manufacturer Rosemount Inc.
Model No. 1154HH4RH Output: 4-20 mA de Temperature Limits +40°- 200°F Humidity Limits 0-100% RH Range Range Code 4: 0-25 to 0-150 inH20 Over pressure Limits 3000 psig without damage Accuracy: +/-0.25% of calibrated span. Includes combined effects of linearity, hysteresis, and repeatability Deadband: none Drift: +/- 0.2% of Upper Range Limit for 30 months Temperature Effect: +/- (0.15% URL + 0.35% span) per 50 Deg F ambient temperature change.
Over Pressure Effect: Maximum zero shift after 3000 psi overpressure
+/- 1.0% of URL (Range Code 4)
Static Pressure Zero Effect: +/-0.66% of URL per 1000 psi Static Pressure Span Effect: +/-0.5% of reading/1000 psi*
Power Supply Effect: Less than 0.005% of output span/volt.
Load Effect: None Mounting Position Effect: Superseded by accuracy specifications Radiation: Accuracy within +/-(0.2% of URL + 0.2% of span) during first 30 minutes; +/-(0.5% URL + 1% span) after 55 megarads TID; +/-(0.75% Upper Range Limit + 1% span) after 110 megarads TID gamma radiation exposure.
Seismic: Accuracy within +/- 0.5% of URL during and after a seismic disturbance defined by a required response spectrum with a horizontal ZPA of 8.5 g's, and a vertical ZPA of 5.2 g's.
Steam Pressure/Temperature Accuracy within+/- (1.0% of Upper Range Limit + 1.0% of span) for Range Code 4.
Post DBE Operation Accuracy at reference conditions shall be within+/- 2.5% of URL after exposure to DBE as described above for one year following DBE.
CALCULATION CONTINUATION/ SHEET: 59 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 5.4 Rack Design Inputs This calculation includes rack components for Comparator, Indication and Recording loops (See Section 1.2). The instrument rack uncertainties used in this calculation are the standard Westinghouse rack specifications as described in the Salem Setpoint Technical Standard (Ref.
3.1.1). Per Wiring Diagrams (Ref. 3.4.1-3.4.7), the loop configurations include Westinghouse rack instruments and may include a Comparator (Setpoint loop) an NUS signal isolator (for a
RG 1.97 isolation, Design Change 2EC-3258 Ref. 3.5.15) and/or Moore signal isolator (Recorder isolation). The uncertainties for racks including non-Westinghouse instruments were evaluated in Attachment 10.2 and 10.3 and the results of those evaluations provide that standard Westinghouse rack specifications are *bounding for this calculation. Therefore, this calculation includes only two typical rack total uncertainties; one typical rack which includes a bistable setting tolerance for use with the comparator loops (Rackl) and a typical rack without the bistable setting tolerance for use with the Indicator and Recorder loops (Rack2).
5.4.1 Safety/Quality Designation (Ref. 3.5.3)
SFfY RLTD/QAR: SR CLASS/QGC: IE EQ:M SEISMIC CAT: 1 5.4.2 Current to Voltage Converters (Ref. 3.5.5)
Manufacturer and Model Numbers shown below are typical for all Signal Conditioners (l/V) listed in Section 1.2, unless otherwise noted.
Manufacturer Westinghouse Model No. 3110554-000 Accuracy +/- 0.010% (Per Ref. 3.1.1, calculation uses 0.100% span)
CALCULATION CONTINUATION/ SHEET:. 60 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
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ORIGINATOR. DATE. REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 5.4.3 Signal Comparators (Ref. 3.5.5)
Manufacturer and Model Numbers shown below are typical for all Signal Comparators listed in Section 1.2, unless otherwise noted.
Manufacturer Hagan Controls/ Westinghouse (H015)
Model No. See Above Input Signal 1-5 VDC Output Digital Contact Closure 5.4.4 Signal Isolators (Ref. 3.5.3, 3.5.5; 3.5.13, 3.5.15 ) (Attachment 10.2) 1(2) LM-517A H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2) LM-527A H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2) LM-537A H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2) LM-547A H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2) LM-518 H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2) LM-528 H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2) LM-538 H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2) LM-548 H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2) LM-519A H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2} LM-529A H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2) LM-539A H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2) LM-549A H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2) LM-519M H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2) LM-529M H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2) LM-539M H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2) LM-549M H015: 4111083-001 or 089N: FIA801-05-07-08 (*)
1(2) LM-519B M422: SCT/1-5V/1-5V/AC 1(2) LM-529B M422: SCT/1-5V /1-5V /AC 1(2) LM-539B M422: SCT/1-5V/1-5V /AC 1(2) LM-549B M422: SCT/1-5V/1-5V/AC Man,ufacturer Hagan Controls/ Westinghouse (H015) INUS Corporation (089N)Moore Industries Inc (M422)
Model No. See Above Input Signal 1-5 VDC Output 1-5 VDC or 4-20 mADC
(*) These signal isolators are either the Westinghouse or the NUS model. This calculation is bounding for both (See Assumption 6.4).
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ORIGINATOR DATE REV: CMM 93 1 CMM 8 REVIEWER VERIFIER DATE AFS 94 LFP 5.4.4.1 NUS Isolator Performance Specifications (Attachment 10.2)
Accuracy +/-0.1% FS Repeatability +/-0.05% FS.
Power Supply .05% change in output for the listed variations, cumulative Linearity * +/- 0.1% FS Temp Effect +/- 0.05% FS/ 0 C.
5.4.4.2 Moore Isolator Performance Specifications (Attachment 10.3)
Accuracy +/-0.1% FS Line Voltage Effect +/-0.005% / 1% line change Temp Effect +/-0.005% per 1 Deg F over -20 to 180 Deg F.
Load Effect +/- 0.001 % span from 0 to max load resistance 5.4.5 Rack Performance General Specifications (Ref. 3.1.1, 3.1.3, 3.5.7)
Manufacturer Westinghouse Accuracy +/- 0.5% span Temperature Effect +/- 0.5% span Drift +/- 1.0% span Bistable Setting Toi +/- 0.25% span
CALCULATION CONTINUATION/ SHEET: 62 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 5.5 Control Room Indicator Design Inputs (Ref. 3.5.3, 3.5.6)
Manufacturer and Model Numbers shown below are typical for all Control Room Indicators listed in Section 1.2, unless otherwise rroted.
5.5.1 Safety/Quality Designations (Ref. 3.5.3)
SFfY RLTD/QAR: SR CIASS/QGC: IE EQ:M SEISMIC CAT: 1
- 5.5.2 Performance Specifications - Control Room Indicator (Ref. 3.5.3, 3.5.6)
Manufacturer: Dixon (Mfg Code D327)
Model: SH101AXT Input: 1-5 VDC Output: 0-100%
Accuracy @ 25 °C +/- 0.100% FS Temperature Effect; Zero Stability +/- 0.010%/Degree C Gain Stability +/- 0.020%/Degree C Maximum Accuracy Drift over time @ 25 Degrees C 0.016% / month Resolution Bar 1.000%
Digital +/- 1 count Operating Temp. Range 0 to 50 Degrees C Relative Humidity 90% maximum AC Power Requirements 118 VAC +/- 10%
CALCULATION CONTINUATION/ SHEET: 63 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRJ I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 5.6 Hot Shutdown Panel Indicator Design Inputs Manufacturer and Model Numbers shown below are typical for all Hot Shutdown Panel Indicators listed in Section 1.2, unless otherwise noted.
5.6.1 Safety/Quality Designations (Ref. 3.5.3)
SFfY RLTD/QAR: SR CLASS/QGC: IE EQ:M SEISMIC CAT: 1 5.6.2 Performance Specifications - Indicator (Ref. 3.5.3, 3.5.5)
Manufacturer: Westinghouse (W120)
Model: 107 Input: 1-5 VDC Output: 0-100 Percent Accuracy +/- 1.5% Range Temperature Effect None Supplied Drift None Supplied Resolution None Supplied 5.7 Recorder Design Inputs (Ref. 3.5.3, 3.5.9) 5.7.1 Safety /Quality Designations (Ref. 3.5.4)
SFfY RLTD/QAR: SR CLASS/QGC: IE EQ:M SEISMIC CAT: 1
CALCULATION CONTINUATION/ SHEET: 64 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 5.7.2 Performance Specifications - Recorder (Ref. 3.5.12)
Manufacturer: Leeds and Northrup Model: Speedomax 100 Series, Model 136 Input: 1-5 VDC Output: 0-100 %
Accuracy: +/- 0.5% span Deadband: +/- 0.25% of span maximum Temperature Effect: None Specified
- Drift: None Specified Readability: None Specified 5.8 M&TE Design Inputs 5.8.1 Transmitter MTE (Ref. 3.1.1, 3.5.2, 3.6)
The instruments required for calibration of the transmitter are designated in the calibration procedures as a Digital Multimeter (Fluke 8600A or equivalent), and a Dead Weight Tester (range of 140 inWC), Mansfield & Green PK or equivalent.
FLUKE 8600A Accuracy +/- 0.050% span (Ref. 3.5.2)
Dead Weight Tester Accuracy+/- 0.100% reading (Ref. 3.5.2)
Test Point Calibration performed through an installed Test Point (resistor) 250 ohm; Accuracy 0.100% span (Ref. 3.1.1) 5.8.2 Rack MTE (Ref. 3.1.1, 3.5.2, 3.6)
The instrument required for calibration of the rack per calibration procedures is the Fluke 8600A, a current simulator and a switch box. Additionally, the signal is fed thr'y~gh an installed resistor (I/V) or a test point resistor. Per Salem Technical Standard for Setpoints, the resistor MTR uncertainty is +/- 0.100% span.
FLUKE 8600A Accuracy +/- 0.050% span (Ref. 3.5.2)
I/V Accuracy+/- 0.100% span (Ref. 3.1.1)
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 5.8.3 Indicator MTE (Ref. 3.1.1, 3.5.2, 3.6)
The instrument required for calibration of the indicator per calibration procedures is the Fluke 8600A Additionally, the calibration of the Indicator is performed with readings through the installed resistor used to condition the 4-20 mADC signal to 1-5 Vde. Per Salem Technical Standard for Setpoints, the resistor contribution is within +/- 0.100% span.
FLUKE 8600A Accuracy +/- 0.050% span (Ref. 3.5.2)
I/V Accuracy+/- 0.100% span (Ref. 3.1.1) 5.8.4 Recorder MTE (Ref. 3.1.1, 3.5.2, 3.6)
- The instrument required for calibration of the rack per calibration procedures is the Fluke 8600A. Additionally, the calibration of the Recorder is performed with readings through the installed resistor used to condition the 4-20 mADC signal to 1-5 Vde. Per Salem Technical Standard for Setpoints, the resistor contribution is within the bounding uncertainty of +/-
0.100% span.
FLUKE 8600A Accuracy+/- 0.050% span (Ref. 3.5.4)
I/V Accuracy+/- 0.100% span (Ref. 3J.1)
CALCULATION CONTINUATION/ SHEET: 66 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 6.0 ASSUMPTIONS 6.1 General Assumptions 6.1.1 Drift/Surveillance Interval I
This calculation assumes that the maximum calibration interval is 30 months for all devices based on the station desire to move to a 24 month Surveillance interval and assuming a 25%
allowance on that value.
6.1.2 Sigma Determination Per the Salem Setpoint Methodology (Ref. 3.1.1) calculations are to be performed to 2 sigma (approximately 95% confidence). Also per Reference 3.1.1, where no confirmation of sigma is supplied for the instrument specifications used (supplied in support of a Nuclear Safety related system), it is reasonable to assume the data falls within a 95% confidence interval. Since no sigma was supplied for the instrumentation used in this calculation, all data was assumed to be ,,,.
2 sigma.
6.1.3 Calibration Temperatures A calibration temperature of 70 Deg F is assumed for calibration of all devices in this calculation.
CALCULATION CONTINUATION/ SHEET: 67 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
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ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 I 1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 6.1.4 Seismic Allowances Per Ref. 3.1.3, Westinghouse does not usually include seismic allowances in their RPS trip and ESF protection function uncertainty calculations. Per the Salem Setpoint Technical Standard (Ref. 3.1.1), a seismic evaluation may be required if the device is used for a Seismic Safe Shutdown. Even though the subject transmitters are included on the Seismic Safe Shutdown Equipment List (SSEL), no adverse effect on the High-High trip function based on a seismic event are assumed to be applicable to the ESFAS High-High trip function (based on normal environmental uncertainties).
For functions within this calculation that may be credited for either a Seismic Safe Shutdown and a DBE (Low trip, Low-Low trip, Indication and Recording) only the larger of the seismic or accident uncertainties would be included since only one event needs to be considered. The
- accident uncertainties provided by the vendor are significantly larger than the specified seismic uncertainties, therefore, when applicable, accident uncertainties are used in lieu of the seismic uncertainties .
CALCULATXON CONTXNUATXON/ SHEET: 68 OPS~G REVXSXON HXSTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
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ORXGXNATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1XRl I REVXEWER/VERXFXER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 6.2 Process Assumptions Per Ref. 3.5.16 (Attachment 10.4), Westinghouse provided notification to PSE&G that previously provided Process Measurement Uncertainties information applicable to the Steam Generator trip functions may not have been conservative. The notification included typical values for the Model 51 Steam Generators assuming that the plant used the same calibration and operating conditions as used in the typical calculation. The calculated values were based on the assumption that transmitter calibration was performed at 110 Deg F and that the maximum normal operating temperature is between 100 and 130 Deg F. The Salem scaling is based on 120 Deg Fat 100% power. Interim Revision lIRO of this calculation was performed assuming those .values were conservative for all functions. lttl After this revision, Westinghouse prepared a report to PSE&G,
Subject:
JPO for Overpower Operation, (Ref. 3.5.24, Attachment 10.6) confirming that based on the PSE&G assumptions for 100% RTP, the PMA typical values were conservative for the Low and Low-Low trip functions, but were not conservative for the High-High trip function as calculated in support o the JPO by Westinghouse. Per this report, the PMA values previously used in calculation Rev.
lIRO were non conservative by approximately 1% NR span. The source of this additional uncertainty is not specifically identified in the JPO, however, the 1% uncertainty is conservatively added to the calculation channel uncertainty under Revision lIRl as an additional bias to the total PMA term.
This calculation also includes indication uncertainties for use in the Emergency Operating Procedures. The impact of Process Measurement Uncertainty was not specifically addressed for the adverse Containment conditions in the Westinghouse evaluation. This calculation conservatively assumes that additional uncertainties must be included due to the elevated containment temperature for the Indication and Recorder loops.
The assumed temperature for maximum reference leg temperature at operating conditions is 224 Deg F. This temperature was chosen since the reference leg insulation is assumed to prevent heat transfer for the first hour of the accident environment. Per the EQRR Accident Temperature vs Time Profile (Ref 3.5.4), accident temperature is postulated to be below 224 Deg F after the first hour. The insulation is assumed to prevent heat up effects for temperatures exceeding this value for this time duration. This error is a positive bias since increased temperature will result in indication readings higher than actual. Therefore, this uncertainty is applicable to the EOPs that provide indicated values based on decreasing level.
Additionally, no decreasing temperature is postulated and therefore no negative uncertainties for the reference leg error are applicable. However, for ease of calculation purposes, and since the negative reference leg error contribution was small; the same uncertainty for the normal and accident negative uncertainty (used for increasing level) was used in the calculation.
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 6.3 Transmitter Assumptions 6.3.1 Transmitter Normal Radiation Uncertainty The transmitters evaluated in this calculation are located behind the biological shield and should not be exposed to the maximum normal TID exposure as provided in the EDC (Ref.
3.5.23 ). Historical "As-Found-As-Left" data was reviewed to verify this assumption. The data reviewed was primarily for the 1153 transmitters that have since been replaced by 1154 Series HH models. The qualifications for radiation provided by Rosemount are typical for both series and the 1154 transmitters are expected to perform as well or better than the 1153 transmitters. The results of the reviewed data determined that the transmitters did not drift outside of the expected allowance. Therefore, it is assumed in this calculation that normal background radiation is not causing excessive drift and that no additional uncertainty needs to be included in this calculation to account for normal radiation exposure.
- 6.4 Rack Assumptions No Rack Assumptions are required.
6.5 Indicator Assumptions 6.5.1 Readability Operator actions utilizing Indicated values based on Control Room Indicators are assumed to be based on the digital readout and not the Bargraph and therefore, this calculation does not require an Indicator readability uncertainty.
6.5.2 Hot Shutdown Panel Readability The hot shutdown panel indicator is analog with a 0-100% scale. Resolution uncertainty is not provided by the vendor, but a readability uncertainty is assumed in this calculation equal to 1/2 the smallest demarcation. The smallest demarcation; 2% for this scale, was field verified.
6.5.3 Temperature Effects Westinghouse does not publish a temperature effect for the Model 107 Indicator used in the Hot Shutdown Panel. This calculation assumes a default value of+/- 0.5% span based on the
- Salem Setpoint Technical Standard recommendation (Ref. 3.1.1).
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ORIGINATOR. DATE REV: CMM 12/13193 1 CMM 8/16/94 IlIRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 6.6 Recorder Assumptions 6.6.1 Recorder Drift No Drift uncertainty was supplied by the Vendor. This calculation assumes a default value of
+/- 0.5% span, applicable over the calibration interval; based on the Salem Setpoint Technical Standard recommendation (Ref 3.1.1).
6.6.2 Recorder Readability No Readability uncertainty was supplied by the vendor. This calculation assumes that the readability for a 0-100% scale is 1/2 the smallest demarcation. Recorders for these loops have a demarcation every 2% (Ref. 3.4.55, 3.4.56).
- 6.6.3 Recorder Temperature Effects No temperature effect was supplied by the vendor. This calculation assumes a standard default value of+/- 0.5% span based on the Salem Setpoint Technical Standard recommendatioff (Ref. 3.1.1) .
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.0 CALCULATION OF UNCERTAINTIES 7.1 Process Measurement Uncertainties (PM)
(Ref. 3.5.16, Assumption 6.2)
The following Process Measurement Uncertainties are based on the letter received from '
-1.1'.
Westinghouse informing Salem Generating Station that previous PMA terms may not have been conservative (Ref. 3.5.16). Additionally, Westinghouse letter PSE-94-555 (Ref. 3.5.24) was prepared for PSE&G to provide analysis supporting Justification for Past Operation. This letter confirmed that the low and low-low typical PMA values were conservative, and provided additional information that the High-High value was non conservative for Salem by 1%. (see Assumption 6.2). Therefore, the typical values will be utilized with an additional uncertainty of 1% included for the High-High trip, Indication and Recorder functions .
- 7.1.1 There are four Process Measurement accuracy terms provided for Steam Generator Water Level. The terms are Process Pressure Variation, Reference* Leg Temperature Variation, Fluid Velocity Effects and Downcomer Subcooling Effects. The four individual effects will be ti.)
combined together to account for the total process measurement bias.
Process Pressure Variation Effect r
After installation of the level measurement system on the steam generator, it is calibrated for a specific set of operating conditions. When the process pressure changes as a consequence of changing operating conditions, a level measurement error is created. An approximation of this measurement error, due to changes in process pressure was provided by Westinghouse. The typical error calculated for the Model 51 Steam Generator is + 1.1 % span for Low/Low-Low and -4.0% span for the High-High trip function.
7.1.2 Reference Leg Temperature Variation Effect In addition to assuming a process pressure when the level measurement system is calibrated, a reference leg temperature is assumed. This uncertainty addresses the changes in. normal operation ambient temperature, not the elevated containment ambient temperatures experienced in an inside containment high energy line break. Typically, a specific operational temperature is assumed for the purpose of calibration and an allowed operational band is assumed about the reference temperature. Westinghouse calculates two uncertainties for this variable, one in the high direction (bounded by 130 Deg F.) and one in the low direction
- (typically 100 Deg F.). The typical errors calculated for the Model 51 Steam Generator for Low/Low-Low and High-High trip functions are as follows: For the Low/ Low-Low trip, the error calculated is + 0.7% Span. For the High-High trip, the error calculated is -0.30% span.
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REV: CMM 3 1 CMM REVIEWER VERIFIER DATE AFS 94 LFP 94 7.1.3 Do\vncomer Subcooling Effects Another source of measurement error is the subcooling of the fluid in the downcomer region in conjunction with a saturated mixture around the steam generator U-tubes. The magnitude of the subcooling in the downcomer is dependent upon the following process conditions; main feedwater temperature, circulation ratio, anq location of the feedwater nozzle with respect to th~ low level tap. The typical error calculated for the Model 51 Steam Generator is +0.5%
span.
7.1.4 Fluid Velocity Effects The Fluid Velocity Effects is a measurement error introduced by fluid flow across the lower tap creating a differential pressure. The uncertainty is a bias in the indicated low level direction. The typical error calculated for the Model'51 Steam Generator is: 0.7% span.
7.1.5 Accident Process Measurement Uncertainties (PM,J.
Some of the Emergency Operating Procedures (EOPs) (See Footnote evaluation Section 9.4) utilizing the Control Room Indicators require the consideration of adverse containment conditions (See Assumption 6.2). This evaluation includes an accident uncertainty which includes additional uncertainty for reference leg heat up based on elevated containment ambient temperatures.
In addition to the Reference Leg Heat Up Effect, the other three PMA terms used for the normal Process Measurement Uncertainty are also applied to the.accident condition. The terms are Process Pressure Variation, Fluid Velocity effects and Downcomer Subcooling Effects. Values for those effects will remain the same as previously calculated. These effects are all considered applicable during the accident condition. Downcomer Subcooling effects*
will only be present for a short time due to the loss of feedwater, however, it is conservatively included in this calculation. An isolated Steam Generator may not be subject to Process Pressure Variation or Fluid Velocity effects, however, this calculation assumes that the unaffected Steam Generators' Indication loops may still be subject to these effe.:t~ .
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 Reference Leg Temperature Effect (Accident)
For calibration values see Attachment A and B. See Assumption 6.2, for process assumptions.
The analytical expression used for approximating the error due to temperature change in the reference leg (from Ref. 3.5.16), is as follows:
£ (% of span) = (HL / H){(PLc - PL) (100%)/(Pre .- Pge)}
Where:
HL vertical distance from lower tap to water level in the condensing pot (ft) at operating conditions (12.04 ft or 144.469 inWC)
H = vertical distance between upper and lower taps on the steam generator (ft) at operating conditions (12.04 ft or 144.469 inWC)
- PLC = water density at pressure and temperature for which the system was calibrated PL = water density in the reference leg at the time of interest Pre = saturated water density at the pressure for which the system was calibrated Pge = dry satUJ*ated steam density at the pressure for which the system was calibrated Specific Volumes (Ref. 3.5.19, Attachment A, B)
PLc = water at 120 °F and 771.22 PSIA = 1/0.01617 (61.84 lbm/ft3)
PL = water at 224 °P and 771.22 PSIA = 1/0.01676 (59.67 lbm/ft3)
Pre = water at 514.01 °P and 771.22 PSIA = 1/0.020765 (48.16 lbm/ft3)
Pge = steam at 514.01 °F and 771.22 PSIA = 1/0.591682 (1.69 lbm/ft3) e (% of span) = (144.469/144.469 ){(61.84-59.67) (100%)/(48.16 - 1.69)}
e (%of span) =+ 4.670% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.1.6 Total Process Measurement Accuracy Process Measurement Effect Low Level Effects High Level Effects Reference Leg Temperature Variation +0.700% -0.300%
(Normal)
+4.670%
(Accident)
Process Pressure Variation +1.100% -4.000%
Downcomer Subcooling +0.500% negligible Fluid Velocity negligible -0.700%
High-High Trip Additional Uncertainty N/A -1.000%
(Assumption 6.2)
Total Process Measurement Effect + 2.300% Normal -6.000%
+ 6.270% Accident 7.2 Insulation Resistance Uncertainty (IR)
(Ref. 3.5.14)
The insulation resistance for the subject loops was determined based on the Engineering Evaluation of Insulation Effects (Ref. 3.5.14), and the Cable Database (Ref. 3.5.18) take off lengths for the cable from the Transmitter racks to the Penetration.
A conservative uncertainty for all Transmitters of + 1.649% span is utilized in this calculation based on that evaluation (Configuration 06X) and the installation configurations and take off lengths from the Cable Data Base.
IR = + 1.649% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.3 Process Element Accuracy (PE)
No primary element (venturi, orifice or elbow) is part of this loop configuration.
Therefore, no primary element accuracy is appropriate to this calculation.
PE = +/- 0.000% span 7.4 Calculation of Transmitter Uncertainties (XMTR)
Uncertainties are from Design Inputs, Section 5.3 except where noted.
7.4.1 Transmitter Accuracy (RAXMfiJ Per design inputs, the vendor accuracy including the combined effects of linearity, hysteresis, and repeatability is +/- 0.25% span.
RAXMfR = +/- 0.250% span 7.4.2 Transmitter Drift (DRXMfiJ Per design inputs, the vendor specified drift over 30 months is +/- 0.2% of Upper Range Limit.
Per assumption 6.1.1, the drift interval is also assumed to be 30 months, therefore, the drift in terms of percent of calibrated span (107.673 inWC) is as follows:
DRXMrR = +/- 0.200% x (150 inWC /107.673 inWC)
DRXMfR = +/- 0.279% span.
7.4.3 Transmitter Temperature Effects - Normal (TE~
Per transmitter design inputs, the vendor specified temperature effect is +/- (0.15% URL +
0.35% span) per 50°F ambient temperature change.
Per Environmental Design Inputs (Section 5.1.1), the normal temperature variation inside the Containment is 60 to 120°F and per Assumption 6.1.3, the calibration temperature is 70°F.
Therefore, the maximum temperature span is 50°F.
TEXMfR = +/- [(0.15% (150 inWC/107.673 inWC) + 0.350%)(50°F/50°F)]
Therefore, TEXMTR = +/- 0.559% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I 1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.4.4 Transmitter Temperature Effects Accident (ATEXMTRl.
Since the transmitters are located in Containment, for calculation of accident conditions, the vendor specified temperature effect for temperatures above the normal specification of 130 Deg F, specified within the "Steam Pressure/Temperature" effect of+/- 1.0% URL + 1.0%
span is bounding.
ATEXMTR = +/- (1.0% x (150 inWC / 107.673 inWC)*+ 1.0%
Therefore, ATEXMTR = +/- 2.393% span 7.4.5 Transmitter Static Pressure Effects (SPEXMTJ.
Static Pressure Effects are applicable to devices directly connected to a process pressure and which measure a differential pressure. Normal operating pressure for Unit 1 and Unit 2 are slightly different; 759.2 psig for Unit 1 and 756.52 psig for Unit 2 (Ref. Attachment A and B).
The higher pressure of 759.2 psig will result in the most conservative error. The vendor stated effects for static pressure include both a zero effect and a span uncertainty that are assumed in this calculation to be dependent effects. The zero effect is +/- 0.66% URL per 1000 psi. The span uncertainty is +/- 0.5% of reading/1000 psi. The reading is conservatively taken at 140 inWC (Attachment A and B). The combined uncertainty for this application is as follows:
SPE = +/-[0.660% (150/107.673) inWC + 0.500%(140/107.673)inWC](759.2/1000)
SPEXMTR = +/- 1.192% span 7.4.6 Transmitter Over Pressure Effects (OPEXMTJ.
Per the Salem Setpoint Technical Standard (Ref. 3.1.1), the overpressure effect accounts for errors in the transmitter performance after exposure to pressures in excess of its normal design range. The design pressure limit for this range code is 3000 psig, and operation is not expected to be over this pressure, therefore, no overpressure effect is considered applicable.
OPEXMTR = +/- 0.000% span
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ORIGINATOR. DATE SC-CNOOl-01 REV: CMM 12/13/93
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1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.4.7 Transmitter Power Supply Effects (PSXMTR1 Per the Salem Setpoint Technical Specification (Ref 3.1.1), the voltage variations of the Hagan 121 power supply is +/- 2.5 VDC . The vendor states a Transmitter uncertainty of +/- 0.005% of output span per volt (input variation). The total uncertainty due to the power supply variation is PS = +/- 0.005% /volt x +/- 2.5 Volts, resulting in a power supply effect of+/- 0.0125% span.
Per the Salem Setpoint Technical Standard, if the PS is less than+/- 0.050%, the effect may be ignored, therefore:
PSXMTR = +/- 0.000% span 7.4.8 Transmitter Humidity Effects (HEXMTJ.
Per the transmitter design inputs, the humidity limits for this unit are 0-100% RH (NEMA 4X) with no additional uncertainty provided by the vendor. Therefore, no humidity effect is considered applicable.
HEXMTR = +/- 0.000% span 7.4.9 Transmitter RFI/EMI Effects (REEXMTJ.
No RFI or EMI effects were provided by the vendor. Per the Salem Setpoint Technical Standard (Ref. 3.1.1), when no effect is provided by the vendor for this effect, it may be considered not applicable.
REEXMTR = +/- 0.000% span 7.4.10 Transmitter Radiation Uncertainty (REXMTJ.
Normal Radiation Exposure Per Assumption 6.3.1, normal radiation effects are assumed to be negligible.
REXMTR = +/- 0.000% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I 1IR::I I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.4.11 Transmitter Accident Radiation Uncertainty (AREXMTiJ.
These transmitters are required to remain operable 120 days after the Design Basis Event ~
(LOCA/MSLB). Per Ref. 3.5.24, the Containment accident radiation condition is 2.93E7 JJRI Rads gamma. Per the Rosemount specification, the radiation effect is +/- 0.5% URL + 1%
span after 55 megrads TID. This uncertainty is conservatively included in the calculation for both the trip function and accident monitoring.
AREXMIR = +/- [0.500% (150 inWC/107.673 inWC) + 1.000%]
AREXMIR = +/- 1.697% span 7.4.12 Transmitter Seismic Effect (SEXMTiJ.
- No seismic consideration is included in this calculation per assumption 6.1.4.
SEXMTR = +/- 0.000% span 7.4.13 Transmitter Post DBE Effect (PDE)
The post accident uncertainty of+/- 2.5% URL will account for th~ expected reference accuracy shift when using the transmitter for up to a year after a design basis event. This error is assumed to replace normal accuracy specifications after conditions have been normalized and the elevated radiation and temperatures no longer exist.
PDE = +/- [2.500% (150 inWC / 107.673 inWC)]
PDE = +/- 3.483% span 7.4.14 Transmitter Accident Pressure Effect (APE)
The Rosemount Transmitters are qualified beyond the specified accident Containment Pressure. Per EQ-49-2 (Ref. 3.5.1), the specified pressure is 43 psig and the device was qualified for 85 psig. No additional uncertainty was provided by the vendor and therefore, this uncertainty is assumed to be included in the stated environmental effects.
APE = +/- 0.000% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I 1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.4.15 Transmitter Calibration Tolerance (CA1xMriJ Per the transmitter sensor calibration procedures (Ref. 3.6.1.13-3.6.1.24), the Calibration Tolerance (CAL) for the transmitter is established at +/- 0.5% span.
CALxMTR = +/- 0.500% span 7.4.16 Transmitter M&TE Tolerance (MTEXMriJ Per calibration procedures (Ref. 3.6.1.13-3.6.1.24), the transmitter input is measured using a Deadweight tester (+/- 0.1 % reading). The reading is conservatively taken at 140 inWC (Attachment A & B), therefore, the uncertainty for the deadweight tester will be 0.1 % x 140/107.673; or; +/-0.13% span. The output is measured with a Fluke Model 8600A (+/- 0.05%
span) .
- Additionally, th~ transmitter output is calibrated using an installed test point resistor assumed to contribute an uncertainty of+/- 0.1 % span (Ref. 3.1.1).
Total device M&TE uncertainty is the SRSS combination of the input and output M&TE uncertainties is:
MTEl = +/- 0.13% span (Ref. 3.5.2)
MTE2 = +/- 0.05% span (Ref: 3.5.2)
MTE3 = +/- 0.10% span (Ref. 3.1.1)
MTEXMrR = +/- [(MTE1) 2 + (MTE2)2+ (MTE3) 2 ] 112 MTEXMrR = +/- [(0.13%) 2 + (0.05%) 2 + (0.1%) 2 ] 112 MTEXMrR = +/- 0.171% span 7.4.17 Total Transmitter Uncertainty (Normal) (XMTRrJ All random, independent uncertainties associated with the Transmitter are combined below using the SRSS method of error combination. Since the Calibration Tolerance for this device is greater than the Accuracy, this calculation uses the term 'CAL' in the total SRSS equation.
- XMTRN = +/- [(0.5) 2 + (0.279) 2 + (0.559) 2 + (0.171) 2 + (1.192) 2]1/2 % span XMTRN = +/- 1.446% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/9* l1IR:I I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.4.18 Total Transmitter Uncertainty (Accident) (XMTR.J.
This device is required for reactor protection, accident monitoring, BOP actions and is credited for safe shutdown.
The seismic uncertainty is smaller than 'the combined accident uncertainties, therefore, the total uncertainty does not include the seismic uncertainty (Assumption 6.1.4).
All random, independent uncertainties associated with the Transmitter are combined below using the SRSS method of error combination. Since the Calibration Tolerance for this device is greater than the Accuracy, this calculation uses the term 'CAL in the total SRSS equation.
XMTRA = +/- [(0.5) 2 + (0.279)2+ (2.393)2 + (0.171)2 + (1.192)2+ (1.697) 2] 112 % span XMTRA = +/- 3.222% span 7.4.19 Total Transmitter Uncertainty (Post Accident) (XMTRp.J.
All random, independent uncertainties associated with the Transmitter are combined below using the SRSS method of error combination. Since the Calibration Tolerance is smaller than the Post DBE accuracy, this calculation uses the term 'PDE' in the total SRSS equation.
ARE and ATE are assumed to no longer be present for the condition calculated below; XMTRPA = +/- [(PDEvu-rn )2 + (DRXMTiJ2 + (TEXMTiJ 2 + (MTEXMTiJ2 + (SPE~ 2 ]1/2 XMTRPA = +/- [(3.483%)2 + (0.279%)2 + (0.559%)2 + (0.171%) 2 + (1.192%) 2]112 XMTRPA = +/- 3. 738% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.5 Calculation of Rack Uncertainties (RACK) 7.5.1. Calculation of Rack Uncertainties for RACKl (Rack including Bistable)
Uncertainties are based on design inputs Section 5.4, unless otherwise noted.
7.5.1.1 Rack Calibration Accuracy (C~cKJ Per Design Inputs, the Calibration tolerance for the rack is set at +/- 0.5% span.
C~CKl = +/- 0.500% Span 7.5.1.2 Rack Temperature Effect (TERAcKJ Per Design Inputs, the Rack Temperature Effect may be assumed to be within+/- 0.5%
span.
TERAcKt = +/- 0.500% span 7.5.1.3 Rack Drift Effect (DRRAcKJ Per Design Inputs, the Rack Drift may be assumed to be within +/- 1.0% span.
DRRAcK = +/- 1.000% span 7.5.1.4 Bistable Setting Tolerance (BSTRACKtJ.
Per Design Inputs, the Rack Bistable Setting Tolerance is set at +/- 0.25% span.
BSTRACK = +/- 0.250% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.5.1.5 Rack MTE Effect (MTERAanJ Per Design Inputs, the Rack MTE is the Fluke 8600A with an accuracy of+/- 0.05%
span.
Additionally, where a test resistor or Installed resistor (See Loop Diagram Section 4.0) is included in the loop calibration, the M&TE uncertainty shall include an additional +/-
0.1 % span to bound the resistor uncertainties (Ref. 3.1.1).
Therefore, the total M&TE for rack devices is the SRSS of MTEl and MTE2.
- [(MTE1) + (MTE2) J1 2 2 12 MTERACKI - +
2 2 2 MTERACKl = +/- [(0.05%) + (0.1 %) ]1 MTERAcKi = +/- 0.112% span 7.5.1.6 Rack Static Pressure Effects (SPERAcKJ (Ref. 3.1.1)
Static Pressure Effects are only applicable to devices directly connected to a process pressure and that measure a differential. Static pressure effects are not applicable to this device.
SPERAcKi = +/- 0.000% span 7.5.1.7 Rack Over Pressure Effects (OPERAcKJ (Ref. 3.1.1)
Over Pressure effects are only applicable to devices connected directly to a process pressure which may be exposed to an overrange condition. Over pressure effects are not applicable to this device.
- OPERAcKi = +/- 0.000% span
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ORIGINATOR. DATE REV: CMM 12/13/93 l CMM 8/16/94 IlIRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.5.1.8 Rack Power Supply Effects (PSRAcKJ (Ref. 3.1.1)
No power supply uncertainty was supplied by the vendor. Per Salem Setpoint Technical Standard (Ref. 3.1.1), since no effect is published, and these effects are typically small, no default assumption is required.
PSRAcKi = +/- 0.000% span 7.5.1.9 Rack Humidity Effects (HERAcKJ (Ref. 3.1.1)
No humidity effects were supplied by the vendor. Per the Salem Setpoint Technical Standard (Ref. 3.1.1), the effect may be assumed to be included within the stated
- 7.5.1.10 environmental effects.
HERAcKi = +/- 0.000% span Rack RFI /EMI Effects (REERAcKJ (Ref. 3.1.1)
No RFI or EMI effects were provided by the vendor. These effects are unlikely due to the shielding and regulation of the use of radios and other interference causing devices in the Control room. Per the Salem Setpoint Technical Standard (Ref. 3.1.1) since no uncertainty is provided by the vendor for this effect it may be considered not applicable:
REERAcKi = +/- 0.000% span 7.5.1.11 Rack Normal Radiation Effects (RERAcKJ (Ref. 3.1.1)
No radiation effects are specified by the vendor nor are they considered applicable to the mild environment of the Control Room.
RERAcKi = +/- 0.000% span
**-~--------
l>~~Ci CALCULATION CONTINUATION/ SHEET: 84 0
- 1-----------&-------------RE~V-IS_I_o_N__H_I_sT_o_R~Y~*-s_H_EE_T_________c_o_NT__'_D_o_N__s_H_E_E_T_:---I CALC. No.: SC-CNOOl-01
REFERENCE:
OR REV: CMM 1 c REVIEWER VERIFIER DATE AFS SJJ 1 11 94 LFP 8 16 94 7.5.1.12 Rack Seismic Effect (SERAcKJ.
(Ref. 3.1.1, 3.1.3)
Per Assumption 6.1.4, no consideration of seismic uncertainty is applicable for this calculation.
SERAcKi = +/- 0.000% span 7.5.1.13 Total Westinghouse Rack Uncertainty (including Bistable) (RACKl):
All random, independent uncertainties associated with the rack are combined below using the SRSS method of error combination. Additionally; SPE, OPE, PS, HE, REE, RE, and SE effects are all set to zero, therefore:
2 2 2 2 2 RACKl = +/- [CAL RAcK1 + DR RACK1 + TE RACKl + BST RAcK1 + MTE RACK1J112 RACKl = +/- [(0.5%)2 + (1.0%) 2 + (0.5%)2 + (0.250%)2 + (0.112%)2 ]112 RACKl = +/- 1.255% span 7.5.2 Calculation of Rack Uncertainties for Rack 2 (Rack without Bistable)
Uncertainties are from Section 5.4 unless noted otherwise.
7.5.2.1 Rack Calibration Accuracy (C~cd
- Per Design Inputs, the Calibration tolerance for the rack is set at +/- 0.500% span.
C~CK2 = +/- 0.500% span .
7.5.2.2 Rack Temperature Effect (TERAcial Per Design Inputs, the Rack Temperature Effect may be assumed to be within +/-
0.500% span.
TERAcK2 = +/- 0.500% span
CALCULATION CONTINUATION/ SHEET: 85 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
- ~~---~-----~---~~
CALC. No.: SC-CNOOl-01
REFERENCE:
OR DATE REV: CMM 12 93 1 CMM REVIEWER VERIFIER DATE AFS 94 LFP 8 7.5.2.3 Rack Drift Effect (DRRAcKJ1 Per Design Inputs, the Rack Drift may be assumed to be within +/- 1.0% span.
DRRAcK2 = +/- 1.000% span 7.5.2.4 Rack MTE Effect (MTERAcI01 Per Design Inputs, the Rack MTE is the Flll:ke 8600A with an accuracy of+/- 0.05%
span.
The M&TE uncertainty shall include +/- 0.1 % span for the I/V to bound the resistor uncertainties.
Therefore, the total M&TE for rack devices is the SRSS of MTEl and MTE2.
MTERACK2 = +/- [(MTE1) 2 + (MTE2)2 ] 112 MTERACK2 = +/- [(0.05%)2 + (0.1 %) 2]112 MTERAcK2 = +/- 0.112% span 7.5.2.5 Rack Static Pressure Effects (SPERAcd (Ref. 3.1.1)
Static Pressure Effects are only applicable to devices directly connected to a process pressure and that measure a differential. Static pressure effects are not applicable to this device.
SPERAcK2 = +/- 0.000% span 7.5.2.6 Rack Over Pressure Effects (OPERAcKJ1 (Ref. 3.1.1)
Over Pressure effects are only applicable to devices connected directly to a process pressure which may be exposed to an overrange condition. Over pressure effects are not applicable to this device.
OPERAcK2 = +/- 0.000% span
CALCULATION CONTINUATION/ SHEET: 86 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR DATE REV: CMM 12 13 93 1 CMM 8 94 lIR REVIEWER VERIFIER DATE AFS 94 LFP 7.5.2.7 Rack Power Supply Effects (PSRAaa}
(Ref. 3.1.1)
No power supply uncertainty was supplied by the vendor. Per Salem Setpoint Technical Standard (Ref. 3.1.1), since no effect is published, and these effects are typically small, no default assumption is require~.
PSRAcK2 = +/- 0.000% span 7.5.2.8 Rack Humidity Effects (HERAcial.
(Ref. 3.1.1)
No humidity effects were supplied by the vendor. Per the Salem Setpoint Technical Standard (Ref. 3.1.1), the effect may be assumed to be included within the stated environmental effects.
HERAcK2 = +/- 0.000% span 7.5.2.9 Rack RFI/EMI Effects (REERAcial.
(Ref. 3.1.1)
No RFI or EMI effects were provided by the vendor. These effects are unlikely due to the shielding and regulation of the use of radios and other interference causing devices in the Control room. Per the Salem Setpoint Technical Standard (Ref. 3.1.1) since no uncertainty is provided by the vendor for this effect it may be considered not applicable.
REERAcK2 = +/- 0.000% span 7.5.2.10 Rack Normal Radiation Effects (RERAcial.
(Ref. 3.1.l)
No radiation effects are specified by the vendor nor are they considered applicable to the mild environment of the Control Room.
RERACK2 ;,,, +/- 0.000% spau-
CALCULATION CONTINUATION/ SHEET: 87 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.5.2.11 Rack Seismic* Effect (SERAcKJl.
{Assumption 6.1.4)
Per Assumption 6.1.4, no consideration of seismic uncertainty is applicable for this calculation.
SERAcK2 = +/- 0.000% span 7.5.2.12 Total Rack Uncertainty (without Bistable) (RACK2)
All random, independent uncertainties associated with the rack are combined below using the SRSS method of error combination. Additionally; SPE, OPE, PS, HE, REE, RE, and SE effects are all set to zero, therefore:
RACK2 =+/- [(C~c1<2)2 + (DRRAcK2)2 + (TERACK2)2 + (MTERACK2)2 ]1/2 RACK2 = +/- [(0.5%) 2 + (1.0%)2 + (0.5%) 2 + (0.112%)2] 112 RACK2 = +/- 1.230% span 7.6 Calculation Of Control Room Indicator Uncertainties (INDCR)
Uncertainties are from Design Inputs Section 5.5, unless otherwise noted.
7.6.1 Indicator Accuracy (RA1Nol Per Design Inputs, the accuracy of the Indication is +/- 0.100% FS at 25 °C (77 °F assumed to be based on approximate calibration temperature).
Therefore: RAmn = 0.100% span 7.6.2 Indicator Drift (DR1Nol Per vendor specification, the Indicator drift is +/- 0.016% per month (FS) . This specification is considered a random effect for the drift interval. Per Assumption 6.1.1, the drift interval is 30 months. Therefore, the calculated Indicator drift is as follows:
- DR1ND = +/-[(0.016%)2
- 30)112 DRIND = +/- 0.088% span
CALCULATION CONTINUATION/ SHEET: 88 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.6.3 Indicator Temperature Effect (TE1Nn1 Indicator Temperature effects are stated as having a zero effect of +/- 0.010% per Degree C (+/-
0.006% per °F) and a gain of+/- 0.02% per Degree~(+/- 0.011 % per °F). These uncertainties are assumed to be dependent to each other, but independent from any other Indicator uncertainty. They are combined algebraically here for inclusion as a single random term in the total device uncertainty. Per Section 5.1.1, temperature will vary by approximately 15 °F from calibration.
Temperature effect; zero = +/- 0.006 x 15 °F = +/- 0.090% span Temperature effect; span = +/- 0.011 x 15 °F = +/- 0.165% span TEIND = +/- (0.090% + 0.165%)
TE1ND = +/- 0.255% span 7.6.4 Indicator Static Pressure Effects (SPE1Nn1 (Ref. 3.1.1)
Static Pressure Effects are only applicable to devices directly connected to a process pressure and that measure a differential. Static pressure effects are not applicable to this device.
SPE1ND = +/- 0.000% span 7.6.5 Indicator Over Pressure Effects (OPE1ND1 (Ref. 3.1.1)
Over Pressure effects are only applicable to devices connected directly to a process pressure which may be exposed to an overrange condition. Over pressure effects are not applicable to this device.
OPE1ND = +/- 0.000% span
CALCULATION CONTINUATION/ SHEET: 89 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.6.6 Indicator Power Supply Effects (PS1NOl.
(Ref. 3.1.1)
No power supply uncertainty was supplied by the vendor. Per Salem Setpoint Technical Standard, since no effect is published, and these effects are typically small, no default assumption is required.
PS1ND = +/- 0.000% span 7.6.7 Indicator Humidity Effects (HE1NOl.
(Ref. 3.1.1)
No humidity effects were supplied by the vendor. Per the Salem Setpoint Technical Standard, the effect may be assumed to be included within the other stated environmental effects.
- HE1ND = +/- 0.000% span 7.6.8 Indicator RFI/EMI Effects (REE1NOl.
(Ref. 3.1.1)
No RFI or EMI effects were provided by the vendor. These effects are unlikely due to the shielding and regulation of the use of radios and other interference causing devices in the Control room. Per the Salem Setpoint Technical Standard (Ref. 3.1.1) since no uncertainty is provided by the vendor for this effect it may be considered not applicable.
REE1ND = +/- 0.000% span 7.6.9 Indicator Radiation Effects (RE1NOl.
(Ref. 3.1.1)
No radiation effects are specified by the vendor nor are they considered applicable to the controlled environment of the Control Room.
RE1ND = +/- 0.000% span
CALCULATION CONTINUATION/ SHEET: 90 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
- ~~--~~~--~~
CALC. No.:
REFERENCE:
V: CMM REVIEWER VERIFIER DATE AFS SJJ 1 11 94 LFP 8 16 94 7.6.10 Indicator Seismic Effect (SE 1 ~
(Ref. 3.1.1)
Per Assumption 6.1.4, no consideration of seismic uncertainty is applicable for this calculation.
SEINo = +/- 0.000% span 7.6.11 Indicator Reading Error (RD 1 ~
Based on Assumption 6.5.1, no inclusion of Reading error is applicable to this calculation.
RD1ND = +/- 0.000% span 7.6.12 Indicator Calibration Tolerance (CAI.i~
Per Procedure (Ref. 3.6.1.1-3.6.1.12), the Calibration Tolerance for the Indicator is set at +/-
1.0% span.
CALiND = +/- 1.000% span 7.6.13 Indicator M&TE (MTE1 ~
Per Design Inputs, the Indicator MTE is the Fluke 8600A with an accuracy of +/- 0.05% span.
Additionally, where a test resistor or Installed resistor (See Loop Diagram Section 4.0) is included in the loop calibration, the M&TE uncertainty shall include an additional +/- 0.1 %
span to bound the resistor uncertainties. (Ref. 3.1.1)
Therefore, the total M&TE for rack devices is the SRSS of MTEl and MTE2.
MTEIND = +/- [(MTE1) 2 + (MTE2) 21112 MTEIND = +/- [(0.05%) 2 + (0.1 %) 2]112 MTE1ND = +/- 0.112% span
CALCULATION CONTINUATION/ SHEET: 91 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE . REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, .DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.6.14 Total Control Room Indicator Uncertainty (INDciJ Since all uncertainties associated with the Indicator performance are considered random and independent, they are combined using the SRSS combination method. Additionally, since Calibration Tolerance is greater than the reference accuracy, this calculation will utilize (CAL) in the SRSS equation. SPE, OPE, PS, HE, REE, RE, RD, and SE are all set to zero, therefore:
INDcR = +/- [(C.A4ND) 2 + (DR1ND)2 + (TE1ND )2 + (MTE1ND)2 ]112 INDcR = +/- [(1.000%)2 + (0.088%) 2 + (0.255%) 2 + (0.112%)2] 112 INDcR = +/- 1.042% span 7.7 Calculation Of Hot Shutdown Indicator Uncertainties (INDns)
Uncertainties for the Westinghouse indicators are from Design Inputs Section 5.6.
7.7.1 Hot Shutdown Indicator Accuracy (RA1Nnl Per Design Inputs, the accuracy of the indicator is +/- 1.5%. of full scale.
RA1ND = +/- 1.500% span 7.7.2 Hot Shutdown Indicator Drift (DR1No).
Westinghouse does not specify a value for Indicator drift. Per the Salem Setpoint Technical Standard recomraendation, this calculation includes a default value equal to instrument reference accuracy.
DR1ND =+/- 1.500% span 7.7.3 Hot Shutdown Indicator Temperature Effect (TE1No}
Westinghouse does not publish a temperature effect. Per the Salem Setpoint Technical Standard recommendation, this calculation includes a default value of+/- 0.5% span.
TE1ND = +/- 0.500% span
CALCULATION CONTINUATION/ SHEET: 92 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CN001-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.7.4 Hot Shutdown Indicator Static Pressure Effects (SPEmo).
(Ref. 3.1.1)
Static Pressure Effects are only applicable to devices directly connected to a process pressure and that measure a differential. Static pressure effects are not applicable to this device ..
SPE1ND = +/- 0.000% span 7.7.5 Hot Shutdown Indicator Over Pressure Effects (OPE1N01 (Ref. 3.1.1)
Over Pressure effects are only applicable to devices connected directly to a process pressure and which may be exposed to an overrange condition. Over pressure effects are not applicable to t~s device .
- 7.7.6 OPE1ND = +/- 0.000% span Hot Shutdown Indicator Power Supply Effects (PS 1 ~
(Ref. 3.1.1)
No power supply uncertainty was supplied by the vendors. Per Salem Setpoint Technical Standard, since no effect is published, and these effects are typically small, no default assumption is required.
PS1ND = +/- 0.000% span 7.7.7 Hot Shutdown Indicator Humidity Effects (HE1NOl (Ref. 3.1.1)
No humidity effects were supplied by the vendors. Per the Salem Setpoint Technical Standard, the effect may be assumed to be included within the other stated environmental effects.
- HE1ND = +/- 0.000% span
CALCULATION CONTINUATION/ SHEET: 93 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
OR'IGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I 1IRl I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.7.8 Hot Shutdown Indicator RFI/EMI Effects (REE 1 ~
(Ref. 3.1.1)
No RFI or EMI effects were provided by the vendor. These effects ate unlikely due to the shielding and regulation of the use of radios and other interference causing devices in the Control room. Per the Salem Setpoint Technical Standard (Ref. 3.1.1) since no uncertainty is provided by the vendor for this effect it may be considered not applicable.
REE1ND = +/- 0.000% span 7.7.9 Hot Shutdown Indicator Radiation Effects (RE 1~
(Ref. 3.1.1)
No radiation effects are specified by the vendors nor are they considered applicable to the controlled environment of the Control Room.
RE1ND = +/- 0.000% span 7.7.10 Hot Shutdown Indicator Seismic Effect (SE1Nn1 (Ref. 3.1.1)
Per Assumption 6.1.4, no consideration of seismic uncertainty is required for this calculation.
SE1ND = +/- 0.000% span 7.7.11 Hot Shutdown Indicator Reading Error (RD 1 ~
(Assumptions 6.5.2)
The Westinghouse indicator has only a bargraph scale so that a reading error is required for this calculation. The reading error is taken as one-half of a minor scale division (Ref. 3.1.1).
The scale is linear; 0-100% and the readability is assumed to be within 1.0%.
RD 1ND = +/- 1.000% span
CALCULATION CONTINUATION/ SHEET: 94 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.7.12 Indicator Calibr~.tion Tolerance (CAI,qNDl.
(Ref. 3.6.1)
The Calibration Tolerance for the Hot Shutdown Panel indicator is +/- 1.500% span.
CALiND = +/- 1.500% span 7.7.13 Indicator M&TE (MTE1NOl.
Per Design Inputs, the Indicator MTE is the Fluke 8600A with an accuracy of+/- 0.05% span.
Additionally, where a test resistor or Installed resistor (See Loop Diagram Section 4.0) is included in the loop calibration, the M&TE uncertainty shall include an additional +/- 0.1 %
span to bound the resistor uncertainties. (Ref. 3.1.1)
Therefore, the total M&TE for rack devices is the SRSS of MTEl and MTE2.
MTEIND = +/- [(MTE1)2 + (MTE2)21112 MTEIND = +/- [(0;05%)2 + (O.l%) 2]1l2 MTE1ND = +/- 0.112% span 7.7.14 Total Hot Shutdown Panel Indicator Uncertainty (INDHs).
Since all uncertainties associated with the Indicator performance are considered random and independent, they are combined using the SRSS combination method. Additionally, only the larger value of Calibration Tolerance or Reference Accuracy was utilized in the equation.
INDHs = +/- [(CAf-iND)2 + (DR1ND)2 + (TEIN0 ) 2 + (MTE1ND) 2 + (RD1ND) 2]'"
INDHs = +/- [(1.500%) 2 + (1.500%) 2 + (0.500%) 2 + (0.112%) 2 + (1.000%)2]Yz INDHs = +/- 2.401 % span
CALCULATION CONTINUATION/ SHEET: 95 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.8 Calculation Of Recorder Uncertainties (REC)
Uncertainties from Section 5.7 except where noted.
7.8.1 Recorder Reference Accuracy (RAREc:;)
Per design inputs, the accuracy for the recorder, based on Full Range is +/- 0.5%.
RAREc = +/- 0.500% span 7.8.2 Recorder Deadband (DBREc:;)
Per design inputs, the deadband for this device is+/- 0.25% span maximum. This deadband specification represents the sensitivity throughout the output range of the recorder to actual input signal change.
DBREc = +/- 0.250% span 7.8.3 Recorder Drift (DRREc:;)
No drift effect was supplied by the vendor. Per Assumption 6.6.1, this calculation includes a default value of +/- 0.5 % span.
DRREc = +/- 0.500% span 7.8.4 Recorder Temperature Effect (TEREc:;)
No temperature effects were specified by the vendor. Per Assumption 6.6.3, this calculation includes a default value of+/- 0.5% span.
TEREc = +/- 0.500% span 7.8.5 Recorder Static Pressure Effects (SPEREc:;)
Static Pressure Effects are only applicable to devices directly connected to a process pressure and which measure a differential. Static pressure effects are not applicable to this device.
SPEREc = +/- 0.000% span
CALCULATION CONTINUATION/ SHEET: 96 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/9.C I1IR'I I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.8.6 Recorder Over Pressure Effects (OPEREc;).
Over Pressure effects are only applicable to devices connected directly to a process pressure which may be exposed to an overrange condition. Over pressure effects are not applicable to this device.
- OPEREc = +/- 0.000% span 7.8.7 Recorder Power Supply Effects (PSREc;).
No power supply uncertainty was supplied by the vendor. Per Salem Setpoint Technical Standard (Ref. 3.1.1 ), since no effect is published, and these effects are typically small, no default assumption is required.
PSREc = +/- 0.000% span 7.8.8 Recorder Humidity Effects (HEREc;).
No humidity effects were supplied by the vendor. Therefore, per the Salem Setpoint Technical Standard (Ref. 3.1.1 ), the effect is assumed to be included within the stated environmental effects.
HEREc = +/- 0.000% span 7.8.9 Recorder RFI/EMI Effects (REEREc;).
No RFI or EMI effects were provided by the vendor. These effects are unlikely due to the shielding and regulation of the use of radios and other interference causing devices in the Control room. Per the Salem Setpoint Technical Standard (Ref. 3.1.1) since no uncertainty is provided by the vendor for this effect it may be considered not applicable.
REEREc = +/- 0.000% span 7.8.10 Recorder Radiation Effects (REREc;).
No radiation effects are specified by the vendor nor are they considered applicable to the controlled environment of the Control Room.
REREc = +/- 0.000% span
CALCULATION CONTINUATION/ SHEET: 97 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I 1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.8.11 Recorder Seismic Effect (SEREc,).
Per Assumption 6.1.4, no consideration of seismic uncertainty is applicable for this calculation.
SEREc = +/- 0.000% span 7.8.12 Recorder Reading Error (RDREc,).
No readability effect was supplied by the vendor for this device. The recorder scale is 0 to 100%. Per Assumption 6.6.2, readability is 1/2 the smallest demarcation. Since the demarcations are every 2%, the readability uncertainty is +/- 1% span .
RDREc = +/- 1.000% span 7.8.13 Recorder Calibration Tolerance (Ct\Lirnc).
Per Procedure (Ref. 3.6.1.1-3.6.1.12), the calibration tolerance for the Recorder is set at+/-
1.000% span.
CALirnc = +/- 1.000% span 7.8.14 Recorder M&TE (MTEREcl Per Design Inputs, the Recorder MTE is the Fluke 8600A with an accuracy of+/- 0.05% span.
Additionally, where an Installed resistor (See Loop Diagram Section 4.0) is included in the loop calibration, the M&TE uncertainty shall include an additional +/- 0.1 % span to bound the resistor uncertainties (Ref. 3.1.1).
Therefore, the total M&TE for rack devi~es is the SRSS of MTEl and MTE2.
MTEREC = +/- [(MTE1) 2 + (MTE2) 2) 112 MTEREC = +/- [(0.05%) 2 + (0.1%) 2]1 2 MTEREc = +/- 0.112% span
CALCULATION CONTINUATION/ SHEET: 98 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.8.15 Total Recorder Uncertainty (REC)
Since all uncertainties associated with the Recorder performance are considered random and independent, they are combined using the SRSS combination method. Additionally, since Calibration Tolerance is greater than the reference accuracy, this calculation will utilize CAL in the SRSS equation. SPE, OPE, PS, HE, REE, RE, and SE are all set to zero.
REC = +/- [CAL2REC + DB2REC + DR2REC + TE2REC + RD2REC + MTE2REdl/2 REC = +/- [(1.000)2 + (0.250)2 + (0.500) 2 + (0.500)2 + (1.000) 2 + (0.112)2]1/2 % span REC = +/- 1.605% span
CALCULATION CONTINUATION/ SHEET: 99 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.9 Channel Error Analysis 7.9.1 Summary of Uncertainties:
Uncertainty Source Total Uncertainty % span Process Measurement +2.300% (For Low Trip Normal (Section 7.1) Function)
-6.000% (For High Trip Function)
Process Measurement + 6.270% (For Low Trip Accident (Section 7.1) Function)
-6.000% (For High Trip Function)
Insulation Resistance +1.649%
(Section 7.2)
Section 7.4 +/- 1.446%.
Transmitter (Normal) +/- 3.222%
Transmitter (Accident) +/- 3.738%
Transmitter (Post Ace)
Rackl (Section 7.5.1) +/- 1.255%
Rack2 (Section 7.5.2) +/- 1.230%
Control Room Indicator +/- 1.042%
(Ref. 7.6)
Hot* Shutdown Indicator +/- 2.401%
(Ref. 7.7)
Recorder +/-1.605%
(Ref. 7.8)
CALCULATION CONTINUATION/ SHEET: 100 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12113193 1 CMM 8116194 l 1IR1 I REVIEWER/VERIFIE~, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.10 Propagation of Error 7.10.1 Channel Uncertainty Determination for Configuration A and C (Transmitter through Rack/Bistable Low-Low and Low Level Trip)
See Section 1.2 for a complete listing of Component ID's.
The following is a representation of the error propagation throughout the Channel for.
Configuration A including process uncertainties calculated for the normal and trip conditions, accident transmitter uncertainties and IR effects.
LT-517 XMTRo RACK RACKlo + PMb+ + !Rb+
(XMTR) (RACKl)
RACKli PM= + 2.300% span ,
IR= + 1.649% span XMTRA = +/-3.222% span (Accident)
RACKl = +/-1.255% span (with Bistable)
XMTRo = +/- [(XMTRA)2]1/2 + PMb + + IRb +
XMTRo = +/- [(3.222%) 2]112 + 2.300% + 1.649%,
XMTRo = + 7.171 % span
- 3.222% span RACKlo = +/-[(XMTR) 2 + (RACK1) 2]112 + PMb+ + IRb+
RACKlo = +/-[(3.222%) 2 + (1.255%)2)112 + 2.300% + 1.649%,
RACKlo = + 7.407% span
- 3.458% span CU = + 7.407% span, - 3.458% span
CALCULATION CONTINUATION/ SHEET: 101 OPS~G CONT'D ON SHEET:
REVISION HISTORY SHEET CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.10.2 Channel Uncertainty Determination for Configuration B (Transmitter through Rack/Bistable High-High Trip)
See Section 1.2 for a complete listing of Component ID's.
The following is a representation of the error propagation throughout the Channel for Configuration B, including process uncertainties for the normal and trip conditions. No accident uncertainties are included.
LT-517 XMTRo RACK RACKlo -PMb-(XMTR) (RACKl)
RACKli PM= - 6.000% span XMTRN = +/-1.446% span (Normal)
RACKl = +/-1.255% span (with Bistable)
XMTRo = +/- [(XMTR) 2]1/2 - PMb-XMTRo = +/- [(1.446%) 2]112 - 6.000% span XMTRo = + 1.446% span, - 7.446% span RACKlo = +/-[(XMTR)2 + (RACK1)2]112 - PMb-RACKlo = +/-[(1.446%)2 + (1.255%) 2]112 - 6.000%
RACKlo = + 1.915% span, - 7.915% span CU = + 1.915% span, - 7.915% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IR.l I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.10.3 Channel Uncertainty Determination for Configuration D (Transmitter through Indicator) (INDc&!U 7.10.3.1 Control Room Indicator Normal Uncertainties The following is a representation of the error propagation throughout the Channel for Configuration D under normal conditions.
See Section 1.2 for a complete listing of Component ID's.
LT-517 XMTRo RACK RACK2o LI-517 (XMTR) (RACK2) (IND) INDo +PMb+ -PMb-RACK2i IN Di PM= + 2.300% span , -6.000% span IR = N/A XMTRN = +/-1.446% span (Normal)
RACK2 = +/-1.230% span XMTRo = +/- [XMTR2]1/2 + PMb+ - PMb-XMTRo = +/- [(1.446% )2] 112 + 2.300% - 6.000%
XMTRo = + 3.746% span, - 7.446% span RACK2o = +/- [(XMTR) 2 + (RACK2) 2 ] 112 + PMb+ -PMb-RACK2o = +/- [(1.446%) 2 + (1.230%) 2 1112 + 2.300% - 6.000%
RACK2o = + 4.198% span, - 7.898% span
+/- [( XMTR) 2 + (RACK2) 2 + (IND) 2] 112 + PMb+ - PMb-
+/- [(1.446%) 2 + (1.230%) 2 + (1.042) 2] 112 + 2.300% -6.000%
+ 4.466% span, - 8.166% span CU = + 4.466% span, - 8.166% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.10.3.2 Control Room Indicator Accident Uncertainties The following is a representation of the error propagation throughout the Channel for Configuration D including accident calculated process uncertainties, accident transmitter uncertainties and IR effects.
See Section 1.2 for a complete listing of Component ID's.
LT-517 XMTRo RACK RACK2o LI-517 (XMTR) (RACK2) (IND) IN Do PMb+ + IR+PMb-RACK2i INDi PM = + 6.270% span, - 6.000% span IR = + 1.649% span XMTRA= +/- 3.222% span (Accident)
RACK2= +/- 1.230% span XMTRo= +/- [(XMTR)2]112 + PMb+ + IRb+ - PMb-XMTRo= +/- [(3.222%)2]112 + 6.270% + 1.649% - 6.000%
XMTRo= + 11.141 % span, - 9.222% span RACK2o = +/- [(XMTR) 2 + (RACK2) 2 ]112 + PMb+ + IRb+ -PMb-RACK2o = +/- [(3.222%) 2 + (1.230%) 2 ]112 + 6.270% + 1.649% - 6.000%
RACK2o = + 11.368% span, - 9.449% span INDcRO = +/- [(XMTR)2 + (RACK2) 2 + (IND)2]112 + PMb+ + IRb+ - PMb-INDcRO = +/- [(3.222%) 2 + (1.230%) 2 + (1.042%) 2 ]112 + 6.270% + 1.649% - 6.000%
= + 11.522% span, - 9.603% span CU = + 11.522% span, - 9.603% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.10.3.3 Control Room Indicator Post Accident Uncertainties The following is a representation of the error propagation throughout the Channel for Configuration D for post accident conditions. The uncertainty is comprised of the normal process uncertainties and post DBE transmitter uncertainties. The IR effect is also included as it is assumed to be residual.
See Section 1.2 for a complete listing of Component ID's.
LT-517 XMTRo RACK RACK2o LI-517 (XMTR) 1----- (RACK2) (IND)
RACK2 i INDi PM= + 2.300% span , -6.000% span IR = + 1.649% span XMTRPA = +/-3.738% span (Post Accident)
RACK2 = +/-1.230% span
- XMTRo = +/- [(XMTR) 2]112 + PMb+ ,+ IRb+, * -PMb-XMTRo = +/- [(3.738%)2] + 2.300%, + 1.649%, - 6.000%
XMTRo = + 7.687% span, - 9.738% span RACK2o = +/- [(XMTR) 2 + (RACK2)2]1!2 + PMb+ + IRb+ -PMb-RACK2o = +/- [(3.738%) 2 + (1.230%)2]112 + 2.300% + 1.649% - 6.000%
RACK2o = + 7.884% span , - 9.935% span INDcRo = +/- [( XMTR) 2 + (RACK2) 2 + (IND)2]112 + PMb+ + IRb+
INDcRO = +/- [ (3.738%) 2 + (1.230%) 2 + (1.042) 2] 112 +2.300% + 1.649% -6.000%
+ 8.020% span, - 10.071 % span CU = + 8.020% span, - 10.071% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.10.4 Channel Uncertainty Determination for Configuration E (Transmitter through Hot Shutdown Indicator) (INDHsQ)
The following is a representation of the error propagation throughout the Channel for Configuration E. Only normal conditions are calculated.
See Section 1.2 for Complete listing of Component ID's.
LT-517 Xmtro RACK RACK2o LI-517A (XMTR) (RACK2) (IND) +PMb+ +IRb+PMb-RACK2i In di PM= + 2.300% span , -6.000% span .
- IR =
XMTRN RACK2 XMTRo =
XMTRo =
=
=
N/A
+/-1.446% span (Normal)
+/- 1.230% span
+/- [(XMTR) 2]112 + PMb+ - PMb-
+/- [(1.446%)2]112 + 2.300% - 6.000%
XMTRo = + 3.746% span, - 7.446% span RACK2o = +/- [(XMTR) 2 + (RACK2) 2 ]112 + PMb+ -PMb-RACK2o = +/- [(1.446%) 2 + (1.230%) 2 ]112 + 2.300% - 6.000%
RACK2o = + 4.198% span, - 7.898% span INDHso = +/- [(XMTR)2 +(RACK2) 2 + (IND)2]112 + PMb+ - PMb-INDHsO = +/- [( 1.446%) + ( 1.230%) + (2.401 % ) 11 + 2.300% - 6.000%
2 2 2 1 2
= + 5.361 % span, -9.061 % span CU = + 5.361% span, - 9.061% span
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ORIGINATOR DATE REV: CMM 93 1 CMM 8 94 1IR REVIEWER VERIFIER DATE AFS 94 LFP 7.10.5 Total Channel Uncertainty Configuration F (Transmitter through Recorder) 7.10.5.1 Recorder Normal Uncertainties The following is a representation of the error propagation throughout the Channel for Configuration F for normal conditions.
See Section 1.2 for complete listing of Component ID's.
LT-519 XMTRo RACK RACK2o LA-5048 (XMTR) (RACK2 (REC) RECo + PMb+-PMb-RACK2i RECi PM= + 2.300% span , -6.000% span IR = N/A XMTRN = +/-1.446% span (Normal)
RACK2 = +/- l.230% span REC= +/- 1.605 % span XMTRo = +/- [XMTR2]112 + PMb+ - PMb-XMTRo = +/- [(1.446%) 2] + 2.300% - 6.000%
XMTRo = + 3.746% span, - 7.446% span RACK2o = +/- [(XMTR) 2 + (RACK2)2]112 + PMb+ - PMb-RACK2o = +/- [(1.446%) 2 + (1.230%)2] 1! 2 + 2.300% - 6.000%
RACK2o = + 4.198% span, - 7.898% span RECo = +/- [( XMTR) 2 + (RACK2) 2 + (REC) 2]1/2 + PMb+ - PMb-RECo = +/- [ (1.446% )2 + (1.230% )2 + (1.605)2]112 + 2.300% - 6.000%
RECo = + 4.786% span, - 8.486% span cu= + 4. 786% span, - 8.486% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.10.5.2 Recorder Accident Uncertainties The following is a representation of the error propagation throughout the Channel for Configuration F for an accjdent condition. The uncertainty is comprised of accident transmitter effects, accident process uncertainties and IR effects.
See Section 1.2 for complete listing of Component ID's.
LT-519 XMTRo RACK RACK2o LA-5048 (XMTR) (RACK2 (REC) RECo +PMb++rRb+PMb-RACK2i RE Ci PM= + 6.270% span , - 6.000% span IR= + 1.649% span XMTRA = +/- 3.222% span (Accident)
RACK2 = +/- 1.230% span REC= +/- 1.605% span XMTRo = +/- [(XMTR)2] 1! 2 + PMb+ + IRb+ - PMb-XMTRo = +/- [(3.222%)2] + 6.270% + 1.649% - 6.000%
XMTRo = + 11.141% span, -9.222% span RACK2o = +/- [(XMTR)2 + (RACK2)2]1/2 + PMb+ + IRb+ - PMb-RACK2o = +/- [(3.222%)2 + (1.230%) 2]112 + 6.270% + 1.649% - 6.000%
RACK2o = + 11.368% span , -9.449% span RECo = +/- [(XMTR) 2 + (RACK2)2 + (REC)2]112 + PMb+ + IRb+ - PMb-RECo = +/- [(3.222%) 2 + (1.230%) 2 + (1.605%)2]112 + 6.270% + 1.649% -6.000%
RECo = + 11.723% span, - 9.804% span cu= + 11.723% span, - 9.804% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.10.5.3 Recorder Post Accident Uncertainties The following is a representation of the error propagation throughout the Channel for Configuration F for post accident conditions.
See Section 1.2 for complete listing of Component ID's.
LT-519 Xmtro RACK RACK2o LA-5048 (XMTR) (RACK2 (REC) +PMb+ +IRb+PMb-RACK2i RE Ci PM= + 2.300% span , - 6.000% span IR = + L649% span
- XMTRPA =
RACK2 =
REC=
XMTRo =
XMTRo =
+/- 3.738% span (Post Accident)
+/- 1.230% span
+/- 1.605% span
+/- [(XMTR)2] 112 + PMb+ + IRb+ - PMb-
+/- [(3.738%) 2] + 2.300% + 1.649% - 6.000%
XMTRo = + 7.687% span, - 9.738% span RACK2o = +/- [(XMTR) 2 + (RACK2)2]1!2 + PMb+ + IRb+ - PMb-RACK2o = +/- [(3.738%) 2 + (1.230%) 2]112 +2.300% + 1.649% -6.000%
RACK2o = + 1.884% span, - 9.935% span RECo = +/- [(XMTR) 2 + (RACK2) 2 + (REC)2]1/2 + PMb+ + IRb+ - PMb-RECo = +/- [(3.738%)2 + (1.230%) 2 + (1.605%) 2] 112 +2.300% + 1.649% -6.000%
RECo = + 8.199% span, - 10.250% span cu= + 8.199% span, - 10.250% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.11 Summary of Channel Uncertainties
SUMMARY
OF UNCERTAINTIES Configuration Normal Accident Post Accident A (Low-Low Trip) N/A + 7.407%, - 3.458% N/A C (Low Trip)
B (High-High Trip) + 1.915%, - 7.915% N/A N/A D (Control Room + 4.466%, -8.166% + 11.522%, -9.603% + 8.020%, -10.071 %
Indication)
E (Hot Shutdown + 5.361 %, - 9.061 % N/A N/A Indication )
F (Recorder) + 4. 786%, -8.486% + 11.723%, -9.804% + 8.199%, -10.250%
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 8.0 CALCULATION OF SETPOINTS This section includes the Calculated Setpoints and the Allowable Values used to support Technical Specification compliance. In addition, this calculation section includes the Acceptable Values for instruments other than the setpoints to provide administrative controls for total loop performance.
8.1 Calculated Setpoints 8.1.1 Calculated Setpoint for the Low-Low Trip Per Section 2.2.2, the existing Technical Specification setpoint and allowable value is:
Functional Unit Trip Setpoint Allowable Value Steam Generator Water ~ 16% of NR Instr span ~ 14.8% of NR Instr span Level Low-Low each Steam Generator each Steam Generator The Analytical Limit for the Low-Low Trip is 0% span, since a level in the Narrow Range in any intact Steam Generator is sufficient to ensure an adequate secondary inventory for a secondary heat sink.
The Calculated Setpoint for the Low Low Trip is established by adding the positive direction Channel Uncertainty (CU) to the Analytical limit. Margin is added for conservatism.
Where: AL = 0%
cu= 7.407%
CS= AL+ CU+ M cs = 0% + 7.407% + 1.593%
CS = 9.0% span The calculated setpoint shown above is the minimum value which the setpoint may be set in the field for a 95% confidence that the Analytical Limit is protected. The actual setpoint in the field is currently set conservatively away from this value.
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IR~ I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 The Low-Low trip setpoint may be set at 9.0%, and adequately protect the Analytical Limit.
This calculation recommends that this value be established.
Recommended Low-Low Setpoint Tech Spec Change Functional Unit Trip Setpoint Allowable Value Steam Generator Water ~ 9% of NR Instr span ~ 8% of NR Instr span Level Low-Low each Steam Generator each Steam Generator 8.1.2 Calculated Setpoint for the Low Trip Per Section 2.2.2, the existing Technical Specification setpoint and allowable value is:
Functional Unit Trip Setpoint Allowable Value Steam Generator Water ~ 25 % of NR Instr Span ~ 24% of NR Instr Span Level Low each Steam Generator each Steam Generator The Low Steam Generator Water Level signal coincident with the Steam Flow/Feed Flow Mismatch signal initiates a reactor trip. This signal is not credited in any safety analyses, but increases the overall reliability of the reactor protection system. The Analytical Limit of 0%
NR span (any level inside the Narrow Range is sufficient to ensure an adequate secondary inventory for a heat sink) used for the Low-Low trip is the same physical restriction impacting the determination of an adequate setting for the Low trip. Since any value equal to or greate than the Low-Low setpoint will satisfy the requirement, and since this calculation recommends that the Low-Low setpoint be revised to ~9%, this calculation supports a Low setpoint equal to or greater than this value. The Low setpoint may be established at ~10% Uust above the Low-Low trip providing an anticipatory function in the scenario of a Steam Flow/Feed Flow Mismatch). The Low Setpoint is therefore recommended at ~10% and the Allowable Value to be set at ~9% NR span.
Recommended Low Setpoint Tech Spec Change:
Functional Unit Trip Setpoint Allowable Value Steam Generator Water ~ 10% of NR Instr Span ~ 9% of NR Instr Span Level Low each Steam Generator each Steam Generator
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ORIGIN1l"'OR- DATE REV: CMM 12/13/93 1 CMM 8/16/94 l1IR1 I REVIEWER/VERIFIER. DATE AFS/SJJ 1/11/94 LFP 8/16/94 8.1.3 Calculated Setpoint for the High-High Trip The analytical limit for the High-High Trip is 75% based on Attachment 10.5.
Per Section 2.2.3, the existing Technical Specification setpoint and allowable limit is:
Functional Unit Trip Setpoint Allowable Value Turbine Trip and s 67% of NR span s 68% of NR span Feedwater Isolation each Steam Generator each Steam Generator The Calculated Setpoint for the High High Limit is established by subtracting the negative direction Channel Uncertainty (CU) from the Analytical limit (AL).
Where: AL= 75%
cu= 7.915%
CS= AL- CU cs = 75% - 7.915%
cs= 67.085%
This calculated setpoint is higher than the existing setpoint. The calculated setpoint demonstrates that the current Technical Specification value is acceptable.
8.2 Allowable Value /Acceptable Value Evaluation Allowable Values are listed within the Technical Specifications which provide a criteria for determining the operability of the trip channel upon periodic testing of the bistable 'as found' values. Exceeding these limits requires an operability determination. For devices in Technical Specification loops where no. Allowable Value is provided, such as the transmitters, indicators and recorders, an administrative*Iimit (Acceptable Value) was established to aid the plant in determining acceptable performance. Allowable values and Acceptable Values are based on the SRSS of the CAL, Drift, and M&TE Uncertainties applicable to the string calibration. This calculation evaluates existing Technical Specification Allowable Values and establishes new Acceptable Values for all applicable devices in this calculation.
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IR:J I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 8.2.1 Allowable Values 8.2.1.1 Low-Low Setpoint The existing Allowable Value for the Low Low Setpoint is 14.8%. To determine the acceptability of this value, the SRSS of the rack Calibration Tolerance, Drift, and M&TE effects was performed as follows. Uncertainties used in this evaluation are from Calculation section 7.5.
AV= +/- [(CAL RACK )2 + (DRRAaJ2 + (MTERAc0 2
+ (BSTRAaJ 2] 1 2
/
AV = +/- [(0.5%)2 + (1.0%) 2 + (0.112%) 2 + (0.25%)2]112 AV = +/- 1.151% span From Section 8.1.1, the calculated (recommended) Low-Low Setpoint is 9.0%.
Subtracting the calculated Allowable Value tolerance of+/- 1.151 % (conservatively 1%),
the minimum Technical Specification Allowable Value would be 8.0%. This value is significantly lower than the current Technical Specification Allowable Value of 14.8%,
and ~herefore, the current value is conservative and acceptable.
8.2.1.2 Low Setpoint The existing Allowable Value for the Low Setpoint is 24%. To determine the acceptability of this value, the SRSS of the rack Calibration Tolerance, Drift, and M&TE effects was performed as follows. Uncertainties used in this evaluation are from Calculation section 7.5.
AV = +/- [(CAL RACK )2 + (DRRAaJ2 + (MTERAc0 2
+ (BSTRAc0 2] 1 2
/
AV = +/- [(0.5%) 2 + (1.0%) 2 + (0.112%) 2 + (0.25%)2]112 AV = +/- 1.151% span Subtracting the calculated Allowable Value tolerance of+/- 1.151 % (conservatively 1%),
from the Setpoint of ~25%, the Technical Specification Allowable Value of ~24%, is acceptable. However, per the recommendation in Section 8.1.2, the recommendation for Allowable Value is ~9% span.
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 l 1IRi I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 8.2.1.3 High-High Setpoint The existing Allowable Value for the High-High Setpoint is 68%. To determine the acceptability of this value, the SRSS of the rack Calibration Tolerance, Drift, and M&TE effects was performed as follows. Uncertainties used in this evaluation are from Calculation section 7.5.
To determine the acceptability of the Allowable Value, the SRSS of the rack Calibration Tolerance, Drift and M&TE effects was performed as follows.
Uncertainties used in this evaluation are from Calculation Section 7.5.
2 Acceptable ValueRAcKt = +/- [CAL2RAcKt + DR2RAcKt + MTE2RACKI + BST2RAcK1]1l Acceptable ValueRAcKt = +/- [(0.5%) 2 + (1.0%) 2 + (0.112%) 2 + (0.25%) 2]112
- Acceptable ValueRACKI = +/- 1.151 % span (Conservatively 1%)
The Technical Specification is 67%. The Calculated Setpoint is 68%. Adding the Acceptable Value tolerance of 1% to the calculated setpoint, the Allowable Value would be 69%. This value is higher than the existing Allowable Value, and therefore the existing value is conservative and acceptable.
The tolerance which was used to develop the Allowable Value of +/- 1.0% span may be used to establish an administrative limit for equipment setpoints that are lower than the Technical Specification setpoint.
For setpoints established at 61 %, the Acceptable value is determined below. The Technical Specification Allowable Value is still the licensing limit, however, since the setpoint is set significantly below this point, an administrative value is also utilized.
Acceptable ValueRAcKi = +/- 1%
Acceptable Value = SP + Acceptable ValueRACKl Acceptable Value = 61% + 1% = :$62%
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 8.2.2 Acceptable Values 8.2.2.1 Transmitter Acceptable Value The Transmitter acceptable value is a device based uncertainty considering the Calibration, Drift and MTE for the device. No string devices are applicable.
CALxMra = +/- 0.500% span DRXMrR = +/- 0.279% span MTEXMTR = +/- 0.171% span Acceptable Value =+/- [(0.5%) 2 + (0.279%)2 + (0.171 %)2] 112 Acceptable ValueXMTR = +/- 0.598% span 8.2.2.2 Control Room Indicator Acceptable Value The Control Room Indicator calibration string is read through the rack components.
Therefore, the Indicator Acceptable value is comprised of the setting tolerance for the Indicator, the drifts for the rack and the Indicator and the MTE used to calibrate the string.
CALiND = +/- 1.000% span DR1ND = +/- 0.088% span DRRACK = +/- 1.000% span MTE1ND = +/- 0.112% span Acceptable Value1ND =+/- [(1.0%) 2 + (0.088%)2 + (1.0%) 2 + (0.112%)2] 112 Acceptable Value1ND = +/- 1.421 % span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 8.2.4 Hot Shutdown Panel Indicator Acceptable Value The Hot Shutdown Panel Indicator calibration string is read through the rack components.
Therefore, the Indicator Acceptable value is comprised of the setting tolerance for the Indicator, the drifts for the rack and the Indicator and the MTE used to calibrate the string.
CALiND = +/- 1.500% span DR1ND = +/- 1.500% span DRRACK = +/- 1.000% span MTE1ND = +/- 0.112% span Acceptable Value1ND = +/- [(1.5%)2 + (1.5%) 2 + (1.0%)2 + (0.112%) 2] 112 Acceptable Value1ND = +/- 2.348% span 8.2.5 Recorder Acceptable Value The Recorder calibration string is read through the rack components. Therefore, the Recorder Acceptable value is comprised of the setting tolerance for the Recorder, the drifts for the rack and the Recorder and the MTE used to calibrate the string.
= +/- 1.000% span
= +/- 0.500% span
= +/- 1.000% span
= +/- 0.112% span Acceptable ValueREc = +/- [(1.0%) 2 + (0.5%) 2 + (1.0%) 2 + (0.112%) 2]112 Acceptable ValueREc = +/- 1.504% span
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ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I lIRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 8.3 Setpoint Relationships SG High Analytical Limit 75%
Margin = 0.085%
- 74. 9% H-H Increasing Value cu= -7.915%
67% Tech Spec High-High Setpoint
- 65. 1% H-H Decreasing Value cu = +1. 915%
61% Lowest High-High Setpoint
- 59. 1% H-H Decreasing Value 44% Normal Operating High Level 33% Normal Operating Low Level cu = - 3.458%
25% Current TS Low Trip Setpoint
- 19. 5% L-L Increasing Value cu = - 3.458%
16% Current TS Low-Low Setpoint cu= +7.407%
10% Recommended Low Setpoint L-L Dec Value 8.593% ~ 9%
9% Recommended Low-Low Setpoint
- 1. 593%-*-- Decreasing Value (Low-Low)
Margin = 1. 593%
0%
SG Low Analytical Limit 0%
Frnm this diagrnm, the C.ment Technical Specification High-High Setpoint is demonstrated to be adequate. SP (67%) + r\
CU (7.915%) = 74.915% . Since Analytical Limit is 75%, positive margin exists. Channels that are set at 61 % in the field, 1~
are shown above to be conservative to the Analytical Limit and sufficiently away from the Normal Operating High Level.
From this diagram, the Current Technical Specification Low-Low Setpoint is demonstrated to be conservative (positive margin). The calculated setpoint is significantly lower than the Technical Specification setpoint and therefore, may be
- lowered. The recommended change is 9%, which is adequate to protect the Low Analytical Limit. The existing Low Setpoint is also adequate but may be changed to 10%, which is acceptable based on incorporating the recommended change to the Low-Low setpoint.
CALCULATION CONTINUATION/ SHEET: 118 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 9.0 DISCUSSION OF RESULTS 9.1 Low-Low Setpoint Per Calculation Section 8.1.1, the calculated setpoint for the Low-Low trip is ~9.0%.
The current Technical Specification Setpoint is ~16%. This value is significantly higher than the minimum requirement. Therefore, the current setpoint is conservative and acceptable, however, it could be lowered to gain operating margin.
9.1.1 Recommended Setpoint The results of this calculation support a recommended setpoint of ~9.0%. This value would
- 9.2 result in increased operating margin and still protect the Analytical limit.
Low Setpoint Per Calculation Section 8.1.2, the calculated setpoint for the Low trip is ~10.0%.
The current Technical Specification Setpoint is ~25%. This value is significantly higher than the minimum requirement. Therefore, the current setpoint is conservative and acceptable, however, it could be lowered to gain operating margin. 11'21 9.2.1 Recommended Setpoint The results of this calculation support a recommended setpoint of ~10.0%. This value would result in increased operating margin and still actuate prior to the process limit (Low-Low Analytical Limit).
CALCULATION CONTINUATION/ SHEET: 119 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IR:J I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 9.3 High-High Setpoint The re-analysis of the High-High Setpoint provided in this calculation was required primarily due to a Westinghouse Letter (Ref. 3.5.16) which indicated that additional Process Measurement Uncertainties should be evaluated. As a result of that evaluation, the Total Channel Uncertainties are larger than previously calculated.
The current Technical Sp~cif~cation Setpoint is '5.67°(a. Per the evaluation provi~e~ in Section ,~
8.1.3, the calculated setpomt is '5.67.085%. Comparmg the total channel uncertamt1es to the 1 (ll~
available margin between calculated setpoint and the Analytical Limit of 75%, the High-High -
setpoint value is still acceptable.
Current plant equipment settings for this function are set as low as 61 %. This value is acceptable since it is set below the minimum calculated setpoint and above the normal operating high level of 44% (Ref 3.1.4).
9.4 Indicator and Recorder 9.4.1 Control Room Indication Uncertainties Per Section 7.10.3, the Control Room Indication uncertainty during normal conditions is +
4.466% and - 8.166% span, + 11.522%, and - 9.603% span for accident conditions. Post accident uncertainties are + 8.020% span, and -10.071 % span.
The results of this calculation demonstrate that the calculated uncertainties are greater than the uncertainties specified in the UFSAR Tables (Ref Section 2.2.4 ). The results of this calculation recommends that the values provided in the UFSAR tables be changed from 4%
(normal and operational occurrences) and 10% (accident); to: 8% (normal and operational occurrences) and 12% (accident).
9.4.2 Hot Shutdown Panel Uncertainties Per Section 7.10.4, the Hot Shutdown Panel Indication uncertainty during normal conditions is + 5.361 % and - 9.061 % span. No accident or post accident uncertainties are applicable.
CALCULATION CONTINUATION/ SHEET: 120 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIG:INATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I 1:rRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 9.4.3 Recorder Uncertainties For Technical Specification monitoring and EOP evaluations, the Indicator Uncertainty will be utilized, since this is the primary device used by the Operator and due to its greater precision.
RG 1.97 requirements as specified in FSAR Table 7.5-1 does cre~it th~ channels used for control for the purpose of detecting steam generator tube rupture and to monitor steam generator water level following a steam line break.
Per Section 7.10.5, the Control Room Recorder uncertainty for normal conditions is +
4.786% and - 8.486% span, + 11.723% and - 9.804% span for accident conditions. The post accident uncertainty is +8.199%, -10.250% span.
The results of this calculation demonstrate that the Channel Uncertainties are greater than the
- uncertainties specified in the UFSAR Tables (Ref Section 2.2.4).
The results of this calculation determined that the values provided in the UFSAR tables be changed from 4% (normal and operational occurrences) and 10% (accident); to: 8% (normal and operational occurrences) and 12% (accident).
9.5 EOP Evaluation 9.5.1 EOP Uncertainties The table below is an excerpt from the Calculation main body, showing the Control Room Indication Channel Uncertainties. EOPs are assumed to utilize the Indicator instead of the recorder since the Indicator is more accurate. These uncertainties will be used as determined to be appropriate to the Emergency Guideline Footnote requirements.
SUMMARY
OF INDICATOR UNCERTAINTIES Configuration Normal Accident D (Indication) + 4.466% + 11.522%
- 8.166% - 9.603%
CALCULATION CONTINUATION/ SHEET: 121 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 9.5.2 BOP Results and Conclusions The EOP Indicated Values listed in Section 2.2.5 were reviewed against the requirements of the Emergency Response Guidelines (ERGs) as listed in 2.2.6.
The normal or adverse uncertainties from Section 9.4.1 were applied as specified based on the footnotes and compared against limits and/ or other limiting criteria listed in the footnotes.
Values equal or more conservative were determined to be acceptable. In some cases, EOP steps are not directly tied back to the ERGs with a footnote. These values were verified to be consistent with the values that were based on a footnote and are acceptable or not acceptable based on that criteria.
The following EOP footnotes are typical for the EOPs evaluated in Section 2.5. Wording of the individual footnotes may vary slightly from the typical footnote shown below but the values evaluated are th~ same .
Typical Footnote: Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy.
Evaluation: This footnote requires that the EDP value be established at 0% SG Level plus the positive normal channel uncertainties for the Indicator ( +4.466%) .
The cu"ent EDP value is established at 8%. This is acceptable.
Typical Footnote: Enter plant specific value showing SG level just in narrow range, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors, not to exceed 50%.
Evaluation: This footnote requires that the EDP value be established at 0% SG Level plus the accident uncertainties for the Indicator ( + 11.522%). The cu"ent EDP value is established at 12%. This is acceptable.
Typical Footnote: Enter plant specific value showing SG level greater than the AFW actuation setpoint Evaluation: This footnote requires that the EDP be based on a value equal to or greater than SG Low-Low setpoint cu"ently established at 16%, being recommended for change to a value of 9%.
CALCULATION CONTINUATION/ SHEET: 122 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRl I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 Typical Footnote: Enter plant specific value corresponding to high-high SG level setpoint, minus 5 % for operating margin.
Evaluation: The high-high setpoint is 67%. This value minus the 5% is 62%. The cu"ent EOP value is established at 62%. This is acceptable.
Typical Footnote: Enter plant specific value corresponding to high-high SG level setpoint, minus 5% for operating margin, including allowances for post accident transmitter errors and reference leg process errors, not less than 50%.
Evaluation: This footnote requires that the EOP value be established at the high high setpoint (67%f) minus 5% operating margin, minus the negative direction accident Indication uncertainties ( -9.603%.). Therefore a value of 52.397%!
satisfies the criteria. The cu"ent EOP is set at 53%. This value is slightly higher than the criteria and therefore this calculation recommends that this value be revised to encompass the additional uncertainty.
- Typical Footnote:
Evaluation:
Enter plant specific value corresponding to SG level at the upper tap, including allowances for normal channel accuracy.
This footnote requires that the EOP value be established at the 100% SG Na"ow Range Level minus the negative normal channel uncertainties of-8.166%. Therefore a value of 91.834% or less will satisfy the criteria. The l cu"ent values are established at 92%. This value is slightly higher than the I criteria and therefore this calculation recommends that this value be revised
- to encompass the additional uncertainty.
Typical Footnote: Enter plant specific value corresponding to SG level at the upper tap, including allowances for normal channel accuracy, post accident transmitter errors, and reference leg process errors.
Evaluation: This footnote requires that the EOP value be established based on SG level at 100% minus the negative directioned accident channel uncertainties of
- 9. 603%. Therefore, a value of 90.397% or less will satisfy the criteria. The cu"ent values are established at 91 %. This value is slightly higher than the ~
criteria and therefore this calculation recommends that this value be revised to encompass the additional uncertainty. LlE
CALCULATION CONTINUATION/ SHEET: 123 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.: SC-CNOOl-01
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1:rR~ I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 EOP Recommendations Since this calculation supports Unit 1 and Unit 2 Design Changes lBC-3345 Pkg 2 and 2BC-3306 Pkg 2 initiated to lower the Low and Low-Low Setpoints, this calculation recommends that the BOPs currently based on the Low-Low setpoint (i.e. various BOP values currently utilizing a 16% setpoint or range of 16%-33%) be revised to coincide with that change.
With the finalization of the Design Changes listed above, the BOP values that are currently established at 16% (16%-33%) may be revised to 12% or a range of 12% -33% and will continue to meet the intent of the footnote guidance.
Additionally, as noted above, the EOP values that are based on High-High trip values slightly exceed the footnote criteria with the recent addition of a 1% bias not previously calculated in Revision lIRO (See Assumption 6.2). Therefore, this calculation recommends that these values be revised to encompass the additional uncertainty.
CALCULATION CONTINUATION/ SHEET: i OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment A
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/16/93 1 CMM 8/16/941 lIRJ I
REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 ATTACHMENT A N/R STEAM GENERATOR LEVEL TRANSMITTER SPAN AND COLD CALIBRATION VALUES
N/R STEAM GENERATOR LEVEL TRANSMITTER SPAN CONTROL ID. NO..f llA~NI A D COLD CALIBRATION VALUES :$(., -CN~I -IJ/
1.0 Instrument nurnber(s): PREPARE Y. A J 1.f J.hJHC.ll.JfF/
2-fll;/1 2.
1.1 2LT517, 2LT518, 2LT519 1.2 2LT527, 2LT528, 2LT529 J.C'AsN 1.3 2LT537, 2LT538, 2LT539 1.4 2LT547, 2LT548, 2LT549 PAGLi.OF.-d 2.0
References:
2.1 ASME STEAM TABLES 2.2 ASME B&PV Code (1974),Section III, Table I-5.0 2.3 UNIT 2 STEAM FI.OW COMPENSATION CALCUIATION (CALC # S-2-MS-CDC-0827 REV 0.)
2.4 DWG 240662-B-9656-8, N0.21 S/G LEVEL AND STEAM FI.OW INSTRUMENT SCHEMATIC 2.5 DWG 240663-B-9656-8, N0.22 S/G LEVEL AND STEAM FI.OW INSTRUMENT SCHEMATIC 2.6 DWG 240664-B-9656-9, N0.23 S/G LEVEL AND STEAM FI.OW INSTRUMENT SCHEMATIC 2.7 DWG 240665-B-9656-9, N0.24 S/G LEVEL AND STEAM FI.OW INSTRUMENT SCHEMATIC 2.8 DWG 233609-B-9611-8, S/G LEVEL ARRANGEMENT, PANELS 444-lA-lM 2.9 DWG 233025-A-1399-8, RX COMP EXT TUBING, NE & SE QUADRANTS EL 130 1 -0 11 2.10 DWG 233026-A-1399-8, RX COMP EXT TUBING, NW & SW QUADRANTS EL 130 1 -0 11
3.0 Assumptions
3.1 Normal temperature in containment is 120 DEGF.
3.2 The high pressure side of the transmitter senses the head from the reference leg pressure and the low pressure side of the transmitter senses the vessel head due to level in S/G.
3.3 Containment temperature during calibration is assumed to be 70 DEGF.
3.4 Condensate pots are supported directly from the S/G vessel.
Therefore, condensate pot elevation varies due to thermal growth of the S/G.
3.5 Cold distance between level taps (H) recorded at 70 DEGF.
3.6 Instrument tubing between the S/G taps and level transmitter is routed togther to the maximum extent possible. Therefore, since there will be no variations in fluid density in the two sensing lines, the location of the level transmitter relative to the lower level connection does not influence the results of the head correction calculation.
3.7 Final transmitter scaling will be based on saturated conditions at the mean average steam pressure for 100% load as taken from section 4.4 of reference 2.3.
3.8 The condensate pots are located at the same elevation as the upper S/G level connections .
CONTROL lO. NO.A-ITM./lltll~NI A
- 4.0 Information Given: (From References) 4.1 Specific Volumes; water at 70 DEGF and 14.7 PSIA v *-
.- 0.01605 w70 water at 120 DEGF and 771.22 PSIA v *-
.- 0.01617 w120 water at 514.01 DEGF and 771.22 PSIA v *-
.- 0.020765 SW steam at 514.01 DEGF and 771.22 PSIA v .- 0.591682 SS 4.2 Vessel and instrument installation dimensions; Centerline of the condensate pot to the lower level tap (cold) A .- 144 c
Upper level tap to the centerline of condensate pot B *-
.- 0 Reference level (0%) to centerline of the lower level tap c .-*- 0 Distance between upper and lower level connections (cold) H := 144 c
5.0 Calculations
5.1 Calculate input and output values for bench calibration.
5.1.1 Thermal growth of vessel between cold (70) and operating (514.01) temperature.
Mean coefficient of thermal expansion of vessel material,
-6 ex := 7.33* 10 in/in/DEGF Hot distance between taps, H := H * (1 +ex* (514.01 - 70))
c H = 144.469
Hot distance between con ower evel tap, A:= A * (1+0<* (514.01 - 70))
c A = 144.469 NOTE: Since the instrument tubing between the S/G taps and level transmitter is routed togther to the maximum extent possible there will be no variations in fluid density in the two sensing lines. Therefore, the difference in elevation between the lower level connection and level transmitter can be ignored when calculating the head sensed at the transmitter's HP and LP connections.
5.1.2 Weight of reference leg under normal conditions Height of the reference leg, h *-
.- A r
h = 144.469 r
Weight of the reference leg, v
w70 w *-
.- h r r v wl20 w = 143.397 r
5.1.3 Weight of water inside the S/G; Normal operating conditions at 100 % level, Height of water above lower level tap, h .- H wlOO
Weight of water in S/G, v
w70 w := h WlOO WlOO V SW w = 111.665 wlOO 5.1.5 Weight of steam inside the S/G; Normal operating conditions at o % level, Height of steam above lower level tap, h := H so Weight of steam in S/G, v
w70 w := h so so v SS w = 3. 919 so 5.1.6 Calculate the DP sensed by the transmitter.
At 100 % level, dp := w - w 100 r WlOO dp = 31. 732 100 At O % level, dp := w - w o r so dp = 139.478 0
P TRANSMITTER STATIC PRESSURE CORREcrION ALCULATION TE INSTRUMENT NUMBERS: 2LT517, 2LT518, 2LT519 z//Y9z_
2LT527, 2LT528, 2LT529 2LT537, 2LT538, 2LT539 PAGE 3 OFL 2LT547, 2LT548, 2LT549 Assume transmitter manufacturer/model; Rosemount 1154Hll4RA
- 1. Uncorrected Span:
Input; Min (i) Max (I) Units i := 139.478 I := 31. 732 INWC Output; Min (o) Max (0) Units 0 := 1 0 := 5 VDC
- 2. Correction Factor:
Vendor's correction factor (K) expressed in precent; K :=
- 75 Percent Span Normal static pressure (NOP);
NOP := 756.52 PSIG Calculated correction factor (CF), expressed in percent; NOP CF:= K * - -
1000 CF = 0.567 Span
- 3. Zero adjustment in terms of input units (ZP):
ZP := CF*%*i ZP = 0.791 INWC
- 4. Zero adjustment in terms of span (ZS) :
ZP ZS .-
I - i ZS = -0.007 Span
- 5. Correction in terms of output span (ZC):
ZC := ZS* (0 - o)
ZC = -0.029 VDC
- 6. Ideal Zero output + correction (oC) :
oC := 0 + zc oc = 0.971 VDC
- 7. Full Scale adjustment in terms of input units (FSP):
- 8. Full Scale adjustment in terms of span (FSS):
I - i FSS = -0.002 Span
- 9. Correction in terms of output span (FSC):
FSC := FSS* (0 - o)
FSC = -0.007 VDC
- 10. Ideal Full Span output+ correction (OC):
OC := 0 + FSC OC = 4.993 VDC
- 11. Revised Signal Span:
SC := (OC - oC)
SC = 4.023 VDC
- 12. Calibration information Number of calibration points (n);
n := 1 ** 9 Calibration points (INPUT)
INPUT :=
n 140 113 86 59 33 59 86 113 140 Calibration Inputs (PCT)
INPUT - i n
PCT .- - - - - -
- 100 n I - i Calibration Outputs (OUTPUT)
INPUT - i n
OUTPUT .- - - - - -
- S C + oC n I - i
.: *,_:,... .,...-""<. .... -,-.....,- ~ i
°$C,. -C/V&;<.J / --4 ' I
~ I ,
r-~
i I
1J. Calibration Table:
OUTPUT n n
-0.484 n 140 o,qc;1 24.574 llJ 49.6JJ 1.959 86 2.967 74.692 59 98.823 J.975
)) 4.946 74.692 59 49.6JJ ).975 86 2.967 24.574 llJ
-0.484 1. 959 140 0.951 J, t1.02.3uOC- ~ a.<=t11 \UJC... :: .3.~fdo \lOC...
O.E;,7 C-107./'i~)= -72.1~0 +-13<}.~18. :. ~7 RcuNO :i-c W\-IOL.e .z.Nwc..!
G,7-13"l."1 "l./.oz:3
- 101. 7l/6 TE~ PT ~ ~1
.-...~~~~.,..._~~~~~
~.~, t.. 4.023\JOC.. ~ ,q/l\lOC,.
6.C..I x.. l-107,-, tl~"i.'17d =. 1..3. 75'.3 1t.Jc.L.
~'-WQ WHCt.£ I ~ v..JC.*. 7~ I u 7..,-1 .'OB~ ~.OU+ ~.':Of :
/07. 74'-
1-
,-~-,A-r.>c..--3-.-4l=f,=\JQC.---pr
CALCULATION CONTINUATION/ SHEET: i OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment B
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/16/93 1 CMM 8/16/941 lIRl I
REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 ATTACHMENT B N/R STEAM GENERATOR LEVEL TRANSMITIER SPAN AND COLD CALIBRATION VALUES 11:e1
1.0 Instrument number(s):
j,,,Sllclt.14 FT 1.1 1LT517, 1LT518, 1LT519 1.2 1LT527, 1LT528, 1LT529 1.3 1LT537, 1LT538, 1LT539 1.4 1LT547, 1LT548, 1LT549 PAGE~OF~
2.0
References:
2.1 ASME STEAM TABLES 2.2 ASME B&PV Code (1974),Section III, Table I-5.0 2.3 DCP lEC-3039 PKG. 2 CALCUIATION FOR MULTIPLIER/SQUARE
-ROOT EXTRACTOR CALIBRATION (S-l-CN-CDC-0611 REV 0.)
2.4 DWG 211301-B-9508-11, N0.11 S/G LEVEL AND STEAM FLOW INSTRUMENT SCHEMATIC 2.5 DWG 211302-B-9508-12, N0.12 S/G LEVEL AND STEAM FLOW INSTRUMENT SCHEMATIC 2.6 DWG 211303-B-9508-12, N0.13 S/G LEVEL AND STEAM FLOW INSTRUMENT SCHEMATIC 2.7 DWG 211304-B-9508-11, N0.14 S/G LEVEL AND STEAM FLOW INSTRUMENT SCHEMATIC 2.8 DWG 233609-B-9611-8, S/G LEVEL ARRANGEMENT, PANELS 444-lA-lM 2.9 DWG 229928-A-1327-10, RX COMP EXT TUBING, NE & SE QUADRANTS EL 130 1 -0 11 2.10 DWG 229929-A-1327-12, RX COMP EXT TUBING, NW & SW QUADRANTS EL 130 1 -0 11
3.0 Assumptions
3.1 Normal temperature in containment is 120 DEGF.
3.2 The high pressure side of the transmitter senses the head from the reference leg pressure and the low pressure side of the transmitter senses the vessel head due to level in S/G.
3.3 Containment tempecature during calibration is assumed to be 70 DEGF.
3.4 Condensate pots are supported directly from the S/G vessel.
Therefore, condensate pot elevation varies due to thermal growth of the S/G.
3.5 Cold distance between level taps (H) recorded at 70 DEGF.
3.6 Instrument tubing between the S/G taps and level transmitter is routed togther to the maximum extent possible. Therefore, since there will be no variations in fluid density in the two sensing lines, the location of the level transmitter relative to the lower level connection does not influence the results of the head correction calculation.
3.7 Final transmitter scaling will be based on saturated conditions at the mean average steam pressure for 100% load as taken from section 4.4 of reference 2.3.
3.8 The condensate pots are located at the same elevation as the upper S/G level connections.
4.0 Information Given: (From References) 4.1 Specific Volumes; water at 70 DEGF and 14.7 PSIA v .- 0.01605 w70 water at 120 DEGF and 773.9 PSIA v *-
.- 0.01617 w120 water at 514.40 DEGF and 773.9 PSIA v .- 0.020776 SW steam at 514.40 DEGF and 773.9 PSIA v *-
.- 0.589511 SS 4.2 Vessel and instrument installation dimensions; Centerline of the condensate pot to the lower level tap (cold) A .- 144 c
Upper level tap to the centerline of condensate pot B *-
.- 0 Reference level (0%) to centerline of the lower level tap c .- 0 Distance between upper and lower level connections (cold) H *-
.- 144 c
5.0 Calculations
5.1 Calculate input and output values for bench calibration.
5.1.1 Thermal growth of vessel between cold (70) and operating (514.40) temperature.
Mean coefficient of thermal expansion of vessel material,
-6 O< := 7.33*10 in/in/DEGF Hot distance between taps, H := H * (l + 0<* (514.40 - 70))
c H = 144.469
Hot distance between tap, A :=A * (1 + O(* (514.40 - 70))
c A = 144.469 NOTE: since the instrument tubing between the S/G taps and level transmitter is routed togther to the maximum extent possible there will be no variations in fluid density in the two sensing lines. Therefore, the difference in elevation between the lower level connection and level transmitter can be ignored when calculating the head sensed at the transmitter's HP and LP connections.
5.1.2 Weight of reference leg under normal conditions Height of the reference leg, h *-
.- A r
h = 144.469 r
Weight of the reference leg, v
w70 w .- h r r v w120 w = 143.397 r
5.1.3 Weight of water inside the S/G; Normal operating conditions at 100 % level, Height of water above lower level tap, h := H wlOO
Weight of water in S/G, v
w70 w .- h wlOO wlOO v SW w = 111.606 WlOO 5.1.5 Weight of steam inside the S/G; Normal operating conditions at O % level, Height of steam above lower level tap, h .- H so Weight of steam in S/G, v
w70 w := h so so v SS w = 3.933 so 5.1.6 Calculate the DP sensed by the transmitter.
At 100 % level, dp := w - w 100 r w100 dp = 31. 791 100 At O % level, dp := w - w o r so dp = 139.464 0
P TRANSMITTER STATIC PRESSURE CORRECTION ALCUIATION INSTRUMENT NUMBERS: 1LT517, 1LT518, 1LT519 1LT527, 1LT528, 1LT529 1LT537, 1LT538, 1LT539 1LT547, 1LT548, 1LT549 Assume transmitter manufacturer/model; Rosemount 1154HH4RA
- 1. Uncorrected Span:
Input; Min (i) Max (I) Units i := 139.464 I := 31. 791 INWC Output; Min (o) Max (0) Units 0 := 1 0 := 5 VDC
- 2. Correction Factor:
Vendor's correction factor (K) expressed in precent; K := . 75 Percent Span Normal static pressure (NOP);
NOP := 759.2 PSIG Calculated correction factor (CF), expressed in percent; NOP CF:= K - - -
1000 CF = 0.569 Span
- 3. Zero adjustment in terms of input units (ZP):
ZP := CF*%*i ZP = 0.794 INWC
- 4. Zero adjustment in terms of span (ZS) :
ZP ZS *-
I - i ZS = -0.007 Span
- 5. Correction in terms of output span (ZC):
ZC := ZS* (0 - o) zc = -0.03 voe
- 6. Ideal Zero output+ correction (oC):
oC := o + ZC oC = 0.97 VDC
- 7. Full Scale adjustment in terms of input units (FSP):
FSP := CF*%*I FSP = 0.181 INWC
- 8. Full Scale adjustment in terms of span (FSS):
I - i FSS = -0.002 Span
- 9. Correction in terms of output span (FSC):
FSC := FSS* (0 - o)
FSC = -0.007 VDC
- 10. Ideal Full Span output+ correction (OC):
OC := 0 + FSC OC = 4.993 VDC
- 11. Revised Signal Span:
SC := (OC - oC)
SC = 4.023 VDC
- 12. Calibration information Number of calibration points (n);
n := 1 ** 9 Calibration points (INPUT)
INPUT :=
n 140 113 86 59 33 59 86 113 140 Calibration Inputs (PCT)
INPUT - i n
PCT *-
.- -----*100 n I - i Calibration Outputs (OUTPUT)
INPUT - i n
OUTPUT *-
.- -----*SC+ oC n I - i
i CQ1'1 ~L .o. NO . .+r~ r 6 lJ. Calibration Tal:lle:
PCT INPUT n n
-ir,40A un n
n, SI~
24.578 113 l. 959 49.654 86 2.968
- 74. 73 59 J ,977
- 98. 877 33 74.73 4.948 59 3.977 49.654 86 24.578 2.968 113 l. 959
-0.498 140 0.95 Tk!.iP 7°b Cl"'7.( "l.O'Z..3uoc.. ~. 970 .:. .1.~S\JO:..
~.<Cl l-1c7.C.7~J t 139.~~ =- G7.32.3 lwL.J Ra.a.~ 1"'C ~f{~.J:N'NC...: C7~Nwc:!..
(Q7-139.Y6lJ ;t. "l.cl.3 +.6.<=J?D
-107.'=>73
.] . ~ 77 \.lOC....
r Pr<§! 61'7o
- o.2.:ix. ~.oz
- c. ?S L-'07. G )
~ 7 i~
tC.9"70::.
t" I .464 1,q7" \lOC...*
=- /I Z..54b ;NV.:C...
I 13 (~we_
II 3 - L3' Y6'f
-101~7~
~T' Pr ~1o
~---------.:.<r--------,
.1"t lf.o~ t ~.<no=- bl'"ll.JO!
- 0. 1'7 (-107. 7 i) ~ \3q. 'iG,~ :. I NWC...
~ T?> Wl40L&. 1~; 122.. we_
1Z..2.-1.J.,.~E,~:A 4.c'2.~,0.910 =- r.c.2..3~
-101.~1.2:i TS.Sr ~ IC..'b
~ll~L~l=1*M=INC.,=~=====\.=C.Z.==3-~-0C...---,
CALCULATION CONTINUATION/ SHEET: i OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment C
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/16/93 1 CMM 8/16/941 1IR~
I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 ATTACHMENT C LINEAR INSTRUMENT SCALING CALCULATION
INEAR INSTRUMENT SCALING CALCUIATION 1.0 Instrument number(s):
1.1 1IM517A, 11M518, 1I.M519A J.~
1.2 1IM527A, 11M528, 11M529A 1.3 1I.M537A, 11M538, 1I.M539A 1.4 1IM547A, 11M548, 11M549A 1.5 2I.M517A, 21M518, 21M519A 1.6 2I.M527A, 21M528, 21M529A 1.7 2I.M537A, 21M538, 2I.M539A 1.8 2I.M547A, 21M548, 2I.M549A 2.0
References:
2.1 IISCS DATA BASES 3.0 Information Given: (From References) 3.1 Instrument Input Range; Min ( i) Max (I) Units i := 1 I := 5 VDC 3.2 instrument output Range; Min (o) Max (0) Units 0 := 1 0 := 5 VDC
4.0 Calculations
4.1 Calculate input and output values for bench calibration.
4.1.1 Instrument input span (INSPAN)
INSPAN := II - ii INSPAN = 4 VDC 4.1.2 Instrument output span (OUTSPAN)
OUTSPAN := IO - ol OUTSPAN = 4 VDC
- 4.1.3 Calibration information NUlllber of calibration points (n);
n := 1 ** 9 Calibration points (INPUT)
INPUT :=
n 1
2 3
4 5
4 3
2 1
Calibration Inputs (PCT)
INPUT - i n
PCT .-
- - -----*100 n INS PAN Calibration Outputs (OUTPUT)
INPUT - i n
OUTPUT := ~----*OUTSPAN + o n INSPAN 4.2 Calibration Table:
PCT INPUT OUTPUT n n n 0 1 1 25 2 2 50 3 3 75 4 4 100 5 5 75 4 4 50 3 3 25 2 2 0 1 1
CALCULATION CONTINUATION/ SHEET: i OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CN001-01 Attachment D
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/16/93 1 CMM 8/16/941 lIRl I
REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 ATTACHMENT D LINEAR INSTRUMENT SCALING CALCULATION
![NEAR INSTRUMENT SCALING CALCULATION 1.0 Instrument nurnber(s): J.~
1.1 1LI517, 1LI517A, 1LI518, 1LI519 1.2 1LI527, 1LI527A, 1LI528, 1LI529 1.3 1LI537, 1LI537A, 1LI538, 1LI539 1.4 1LI547, 1LI547A, 1LI548, 1LI549 1.5 2LI517, 2LI517A, 2LI518, 2LI519 1.6 2LI527, 2LI527A, 2LI528, 2LI529
- 1. 7 2LI537, 2LI537A, 2LI538, 2LI539 1.8 2LI547, 2LI547A, 2LI548, 2LI549 2.0
References:
2.1 IISCS DATA BASES 3.0 Information Given: (From References) 3.1 Instrument Input Range; Min ( i) Max (I) Units i .- 1 I := 5 VDC 3.2 instrument Output Range; Min (o) Max (0) Units 0 .- 0 0 := 100 PERCENT SPAN
4.0 Calculations
4.1 Calculate input and output values for bench calibration.
4.1.1 Instrument input span (INSPAN)
INSPAN := II - ii INSPAN = 4 VDC 4.1.2 Instrument output span (OUTSPAN)
OUTSPAN := lo - ol OUTSPAN = 100 PERCENT SPAN
4.1.3 Calibration information
--~~~~~--------------'
Number of calibration points (n) ;
n := 1 ** 9 Calibration points (INPUT)
INPUT :=
n 1
2 3
4 5
4 3
2 1
Calibration Inputs (PCT)
INPUT - i n
PCT .-
- - -----*100 n INSPAN Calibration Outputs (OUTPUT)
INPUT - i n
OUTPUT := - - - - -
- OUTSPAN + o n INS PAN 4.2 Calibration Table:
PCT INPUT OUTPUT n n n 0 1 0 25 2 25 50 3 50 75 4 75 100 5 100 75 4 75 50 3 50 25 2 25 0 1 0
CALCULATION CONTINUATION/ SHEET: i OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.1
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRJ I REVIEWER/VERIFIER, DATE AFS 1/11/94 LFP 8/16/94 ATTACHMENT 10.1 SCALING ADDENDUM FOR ALWWABLE VALUE AND SETPOINTS - STEAM GENERATOR LEVEL INSTRUMENTATION
CALCULATION CONTINUATION/ SHEET: 1 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.1
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IR. I REVIEWER/VERIFIER, DATE AFS 1/11/94 LFP 8/16/94 1.0 INSTRUMENT NUMBERS See Calculation Section 1.2 for complete list of Component ID's.
2.0 PURPOSE The purpose of this scaling calculation is to provide the five point calibration tables necessary to ensure proper transmitter, rack, indicator and recorder input to output relationships. This document is provided to supplement previous scaling performed in Attachment A
3.0 REFERENCES
See setpoint Calculation SC-CNOOl-01, Section 3.0 4.0 ASSUMPTIONS 4.1 This calculation assumes that where readings are not possible due to resolution or demarcations on the scale that it is acceptable to round in a conservative manner.
5.0 DESIGN INPUTS (See Calculation SC-CNOOl-01)
Calibrated Span: 33 TO 140 in wc Indicated Span: 0-100% Level (Narrow Range)
CALCULATION CONTINUATION/ SHEET: 2 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.1
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 IlIR: I REVIEWER/VERIFIER, DATE AFS 1/11/94 LFP 8/16/94 6.0 Linear Device Scaling Individual scaling is not provided for rack components including signal conditioners and isolators unless used as an output monitoring point in the calibration procedure. These devices may be checked individually, if desired using the linear scaling formulas below.
6.1 Linear Device scaling is based on the following:
Input Span: = I I-i I Min (i) Max (I) Units i =4 I = 20 mADC i =1 I =5 VDC Input Span = 16 mADC Input Span = 4 VDC Output Span: = I 0-o Min (o) Max (0) Units 0 = 1 0 =5 VDC Output Span = 4 VDC No non-linear devices are included in this calculation.
CALCULATION CONTINUATION/ SHEET: 3 OPS~G CONT'D ON SHEET:
REVISION HISTORY SHEET CALC. No.:SC-CNOOl-01 Attachment 10.1
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 I lIRJ I REVIEWER/VERIFIER, DATE AFS 1/11/94 LFP 8/16/94 7.0 ACCEPTABLE OR ALLOWABLE VALUES 7.1 Transmitters 1(2)LT-517, 1(2)LT-527, 1(2)LT-537, 1(2)LT-547 1(2)LT-518, 1(2)LT-528, 1(2)LT-538, 1(2)LT-548 1(2)LT-519, 1(2)LT-529, 1(2)LT-539, 1(2)LT-549 Manufacturer: Rosemount Model No. 1154HH4RH Input 33 TO 140 INWC Output 1-5 Vdc (at test point)
The Calibrated spans for the following transmitters is compensated for operating conditions.
See Attachment A and B for determination of these values.
Unit 1 Steam Generator Narrow Range Level lLT-517 (typical)
Required Input (in WC) Required Tolerance Acceptable Value (Vdc)
+/- 0.02 Vdc +/- 0.024 Vdc 140.0 0.950 ( 0.930 to 0.970) (0.926 to 0.974) 113.0 1.959 (1.939 to 1.979) (1.935 to 1.983) 86.0 2.968 (2.948 to 2.988) (2.944 to 2.992) 59.0 3.977 (3.957 to 3.997) ( 3.953 to 4.001) 33.0 4.948 (4.928 to 4.968) ( 4.924 to 4.972)
CALCULATION CONTINUATION/ SHEET: 4 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.1
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 I lIR" I REVIEWER/VERIFIER, DATE AFS 1/11/94 LFP 8/16/94 Unit 2 Steam Generator Narrow Range Level 2LT-517 (typical)
Required Input (in WC) Required Tolerance Acceptable Value (Vdc)
+/- 0.02 Vdc +/- 0.024 Vdc 140.0 0.951 ( 0.931 to 0.971) (0.927 to 0.975) 113.0 1.959 (1.939 to 1.979) (1.935 to 1.983)
- 86.0 59.0 2.967 (2.947 to 2.987) 3.975 (3.955 to 3.995)
(2.943 to 2.991)
( 3.951 to 3.999) 33.0 4.946 (4.926 to 4.966) ( 4.922 to 4.970)
CALCULATION CONTINUATION/ SHEET: 5 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.1
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 IlIRJ I REVIEWER/VERIFIER, DATE AFS 1/11/94 LFP 8/16/94 7.2 Indicators 1(2)LI-517, 1(2)LI-527, 1(2)Ll-537, 1(2)LI-547 1(2)LI-518, 1(2)LI-528, 1(2)LI-538, 1(2)Ll-548 1(2)Ll-519, 1(2)LI-529, 1(2)LI-539, 1(2)LI-549 Manufacturer: Dixon Model: SH101AXT Input : 1-5 Vdc Output : 0-100%
CONTROL ROOM INDICATOR SCALING Monitoring 1(2) LI-517 Acceptable Value (%)
Point (typical)
/\
I~
1(2)TP-517-1 Cal Tol = +/-
1% span* +/- 1.421 % span
- Required Required Input Vdc Tolerance(%)
1.000 0 (0 to 1.4)
(0 to 1.0) 2.000 25.0 (23.6 to 26.4)
(24.0 to 26.0) 3.000 50.0 (48.6 to 51.4)
(49.0 to 51.0) 4.000 75.0 (73.6 to 76.4)
(74.0 to 76.0) 5.000 100.0 (98.6 to 100.0)
(99.0 to 100.0)
This calculation supports +/- values for all points. Off scale values are acceptable if jaj\
determined to be within the specified tolerance.
CALCULATION CONTINUATION/ SHEET: 6 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.1
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 I1IRJ I REVIEWER/VERIFIER, DATE AFS 1/11/94 LFP 8/16/94 7.3 Hot Shut Down Indicators 1(2)Ll-517A, 1(2)LI-527A, 1(2)Ll-537A, 1(2)Ll-547A Manufacturer: Westinghouse Model: 107 Input : 1-5 Vdc Output : 0-100%
Monitoring 1(2) LI-517A Point 1(2)TP-517-1 Cal Tolerance Acceptable Value (%)
+/- 1.5% span* +/- 2.0% span*
Required Required Input Vdc Tolerance ( % )
1.000 0 (0 to 2.0)
(0 to 1.5) 2.000 25.0 (23.0 to 27.0)
(23.5 to 26.5) 3.000 50.0 (48.0 to 52.0)
(48.5 to 51.5) 4.000 75.0 (73.0 to 77.0)
(73.5 to 76.5) 5.000 100.0 (98.0 to 100.0)
(98.5 to 100.0)
- This calculation supports +/- values for all points. Off scale values are acceptable if determined to be within specified tolerance.
CALCULATION CONTINUATION/ SHEET: 7 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CN001-01 Attachment 10.1
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 IlIR .. I REVIEWER/VERIFIER, DATE AFS 1/11/94 LFP 8/16/94 7.4 Recorders 1(2)LA-5048, 1(2)LA-5049, 1(2)LA-5050, 1(2)LA-5051 Manufacturer: Leeds and Northrup Model No.: Speedomax 136 Input Range : 1-5 Vdc Output: 0-100%
Monitoring 1(2)LA-5048 Acceptable Value ( % ) *)r~
Point Cal Tol = +/- AV = +/-1.5% span*
TP-517-1 1.0% span*
Required (%) (%)
Input Vdc 1.000 0 (0 to 1.5)
(0 to 1.0) 2.000 25.0 (23.5 to 26.5)
(24.0 to 26.0) 3.000 50.0 (48.5 to 51.5)
(49.0 to 51.0) 4.000 75.0 (73.5 to 76.5)
(74.0 to 76.0) 5.000 100.0 (98.5 to 100.0)
(99.0 to 100.0)
This calculation supports+/- values for all points. Off scale values are acceptable if determined to be within the specified tolerance.
ri\
CALCULATION CONTINUATION/ SHEET: 8 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CN001-01 Attachment 10.1
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 IlIR' I REVIEWER/VERIFIER, DATE AFS 1/11/94 LFP 8/16/94 7.5 Comparators 1(2)LC-517A-B, 1(2)LC-527A-B, 1(2)LC-537A-B, 1LC-547A-B 1(2)LC-517C, 1(2)LC-527C, 1(2)LC-537C, 1(2)LC-547C 1(2)LC-518A-B, 1(2)LC-528A-B, 1(2)LC-538A-B, 1LC-548A-B 1(2)LC-518C, 1(2)LC-528C, 1(2)LC-538C, 1(2)LC-548C 1(2)LC-519A-B, 1(2)LC-529A-B, 1(2)LC-539A-B, 1LC-549A-B Manufacturer: Westinghouse Model No. Model 118 Input: 1-5 Vdc Output: Contact 7.5.1 High-High Setpoint 7.5.1.1 Where trip is established at $.67%:
Voltage = 0.67 x 4.000 Vde + 1.000 = 3.680 V de Where Allowable Value is established at s68%:
Voltage = 0.68 x 4.000 Vdc + 1.000 = 3.720 Vdc 1BS-517A, 1BS-518A, 1BS-519A, 1BS-527A, 1BS-528A, 1BS-529A, 1BS-537A, 1BS-538A, 1BS-539A, 1BS-547A, 1BS-548A, 1BS-549A 2BS-517A, 2BS-519A, 2BS-527A, 2BS-528A, 2BS-537A, 2BS-539A, 2BS-548A, 2BS-549A Output Monitoring Point Setpoint Allowable Value (s68%)
BS-517A $.67% (Cal Tol 0.25%)
Steam Generator Level Trip (inc) 3.680 Vdc (s 3.720 Vdc)
High High Trip (3.670 to 3.680 Vdc)
Reset (dee) 40 mV (30 to 50 mV) from trip
CALCULATION CONTINUATION/ SHEET: 9 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
- CALC. No.:SC-CNOOl-01 Attachment 10.1 ORIGINATOR, DATE REVIEWER/VERIFIER, DATE REV: CMM 12/13/93 AFS 1/11/94
REFERENCE:
1 CMM 8/16/94 LFP 8/16/94 I1IR: I 7.5.1.2 High High Setpoint (Field Adjusted Lower than Tech Spec)
\
Due to process requirements, the following Comparator Outputs are set at 61 % which G.e.\
is conservative to the existing Technical Specification Setpoint of 67% . (Ref main calculation section 8.0). While no adjustment to these setpoints is necessary, it is important to note that, while the Technical Specification Allowable Value will assure that the Analytical Limit is protected, it will not serve as an adequate means to determine that the loop is performing correctly, since it is set significantly beyond the calculated Acceptable Value ( 1% = 0.04 V de) for the Rack Components. This calculation recommends an administrative tolerance of 1% be established for technician alert of questionable loop performance.
2BS-518A, 2BS-529A, 2BS-538A, 2BS-547A Where trip is set at 61 %:
Voltage = 0.61 x 4.000 Vdc + 1.000 = 3.440 Vdc Scaling to support the Setpoint from Calibration procedures:
Output Monitoring Point Setpoint (Currently set at Acceptable Value 2BS-518A (typical above) 61 %) (Cal Toi 0.25%) 1% (0.040 Vdc) (note 1)
Steam Generator Level Trip (inc) 3.440 Vdc (s 3.480 Vdc)
High High Trip (3.430 to 3.440 Vde)
Reset (dee) 40 mV (30 to 50 mV) from trip Note 1. The Acceptable Value shown above is provided to allow the technician to determine acceptable loop performance for the equipment setpoints set at 61%. The actual Technical Specification Allowable Value which is the licensing limit, is s68% ors 3.720 Vdc.
v\
if£.!
CALCULATION CONTINUATION/ SHEET: 10 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.1
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 1lIR I REVIEWER/VERIFIER, DATE AFS 1/11/94 LFP 8/16/94 7.5.2 Current Low Low Setpoint Where trip is Trip ~ 16%
Voltage = 0.160 x 4.000 Vde + 1.000 = 1.640 Vde Where Allowable Value is~ 14.8%
Voltage = 0.148 x 4.000 Vde + 1.000 = 1.592 Vde The current low-low setpoint is based on previously calculated uncertainties. Per the current analysis, this setpoint is conservative and acceptable, but may be lowered for operational margin. The following is the setpoint calibration tolerances and Allowable Value to support the existing setpoint.
- Output Monitoring Point BS-517B Steam Generator Low Setpoint
~16%
(Cal Toi 0.25%)
Trip (dee) 1.640 Vde Allowable Value
~ 14.8%
(~1.592 Vdc)
Low Trip (1.640 - 1.650 Vdc)
Reset (inc) at 40 mV (30 to 50 mV) from trip
CALCULATION CONTINUATION/ SHEET: 11 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.1
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 I lIR: I REVIEWER/VERIFIER, DATE AFS 1/11/94 LFP 8/16/94 7.5.3 RECOMMENDED Low-Low Setpoint The calculation recommends setting the Low-Low trip setpoint at 9%. The following calibration information is provided to support that recommendation.
Where trip is Trip ~9%
Voltage = 0.09 x 4.000 Vdc + 1.000 = ~1.360 Vdc Where Allowable Value is ~8%
Voltage = 0.08 x 4.000 Vdc + 1.000 = ~1.320 Vdc
- Output Monitoring Point BS-517B (typical)
Steam Generator Low Setpoint
~9%
(Cal Tol 0.25%)
Trip (dee) 1.360 Vde Allowable Value
~8%
(~1.320 Vdc)
/\
~
Low Trip (1.360 - 1.370 Vdc)
Reset (inc) at 40 mV (30 to 50 mV) from trip
CALCULATION CONTINUATION/ SHEET: 12 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
- CALC. No.:SC-CNOOl-01 Attachment 10.1 ORIGINATOR, DATE REVIEWER/VERIFIER, DATE REV: CMM 12/13/93 AFS 1/11/94
REFERENCE:
1 CMM 8/16/94 I1IR:
LFP 8/16/94 I
7.5.4 Current Steam Generator Low Setpoint Where trip is Trip ~25%
Voltage = 0.250 x 4.000 Vdc + 1.000 = 2.000 Vdc Where Allowable Value is ~24%
Voltage = 0.240 x 4.000 Vde + 1.000 = *l.960 Vde Low Trip Setpoint Output Monitoring Point Setpoint Allowable Value BS-517C (typical) ~25% ~ 24%
(Cal Toi 0.25%)
Steam Generator Low Trip (dee) 2.000 Vdc (~1.960 Vdc)
Low Trip (2.000 - 2.010 Vdc)
Reset (inc) at 40 mV (30 to 50 m V) from trip
CALCULATION CONTINUATION/ SHEET: 13 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.1
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 IlIR" I REVIEWER/VERIFIER, DATE AFS 1/11/94 LFP 8/16/94 7.5.5 RECOMMENDED Steam Generator Low Setpoint Where trip is Trip ~ 10%
Voltage = 0.10 x 4.000 Vdc + 1.000 = 1.400 Vdc Where Allowable Value is ~9%
Voltage = 0.09 x 4.000 Vdc + 1.000 = 1.360 Vdc Output Monitoring Point Setpoint Allowable Value BS-517C (typical) ~10% ~9%
(Cal Toi 0.25%)
Steam Generator Low Trip (dee) 1.400 Vde (~1.360 Vdc)
Low Trip (1.400 to 1.410 Vdc)
Reset (inc) at 40 mV (30 to 50 m V) from trip
CALCULATION CONTINUATION/ SHEET: i OPS~G REVISION HISTORY SHEET CONT'D ON SHE:ET:
- CALC. No.:SC-CN001-01 Attachment 10.2 ORIGINATOR, DATE REV: CMM 12/13/93
REFERENCE:
1 CMM 8/16/9411IR1 I
REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 AlTACHMENT 10.2 NUS SIGNAL ISOLATOR /RACK EVALUATION
CALCULATION CONTINUATION/ SHEET: 1 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.2
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/94 lIRl REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 1.0 Purpose The purpose of this evaluation is to determine if the total rack uncertainties currently used for a standard Westinghouse Rack (with no comparator) are bounding if the Westinghouse Model 4111083-001 is replaced with the equivalent model manufactured by NUS, Model FIA801 07-08. This evaluation is possible since the rack being evaluated does not contain any other significant modules other than the signal isolators.
2.0 Scope This evaluation is applicable to the Instrument rack for Narrow range Steam Generator Level Indicator and Recorder loops. The isolators that are scheduled for replacement include the following tag numbers. Unit 2 will be replaced first, under DCP 2EC-3178, Pkg 2, the Unit 1 DCP number has not been determined at this calculation issuance.
1(2) LM-517A 1(2) LM-527A 1(2) LM-537A 1(2) LM-547A 1(2) LM-518 1(2) LM-528 1(2) LM-538 1(2) LM-548 1(2) LM-519A 1(2) LM-529A 1(2) LM-53~A 1(2) LM-549A 3.0 References 3.1 NUS Signal Isolator Performance Specification Sheets {Attached) 3.2 Salem Setpoint Technical Standard SC. DE-TS.ZZ-1904 (Q) 3.3 For Additional References See Calculation Section 3.0
CALCULATION CONTINUATION/ SHEET: 2 OPS~G REVISION HISTORY SHEET CONT'D OH SHEET:
CALC. No.:SC-CN001-01 Attachment 10.2
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/9411IR1 I
REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 4.0 Design Inputs 4.1 Standard Westinghouse Rack Performance Specifications (Ref 3.2)
Accuracy +/- 0.500% span
- Temperature Effect +/- 0.500% span Drift +/- 1.000% span RACK = +/- [(RARAaJ2 + (DRRAc )2 + (TERAciJ2 ]1/2 RACK = +/- [(0.500%)2 + (1.000%~ + (0.500%)2 ]112 RACK = +/- 1.225% span 4.2 NUS Performance Specification (Ref 3.1)
Accuracy +/- 0.1 % FS Repeatability +/- 0.05% FS Temperature Effect +/- 0.05% span per Deg C Power Supply 0.05% change in output for the listed variations, cumulative Linearity +/- 0.1 % FS 5.0 Comparison of NUS to Rack Uncertainties 5.1 Calculation of NUS Uncertainties 5.1.1 NUS Reference Accuracy (RANUs}:
Per Design Inputs, the Accuracy for the NUS Isolator is +/- 0.100% span. In addition, accuracy includes the specified linearity of 0.100% and repeatability of 0.050%. Total reference accuracy is considered the SRSS of the components of accuracy such that:
RANUs = +/- [(0.100%)2 + (0.050%)2 + (0.100%)2 ]112 RANUs = +/- 0.150% span
CALCULATION CONTINUATION/ SHEET: 3 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CN001-0l Attachment 10.2
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/9411IR1 I
REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 5.1.2 NUS Temperature Effect (TENUs).:
Per Design Inputs, the Isolator Temperature Effect is less than the 0.050% of output full scale change for a 1 Deg C change in temperature over the specified range. Per the Design Inputs Section 5.1.1 of Ref 3.3, the temperature difference between the maximum room temperature vs calibration temperature and the difference between minimum room temperature and calibration temperature are both 15 deg F.
TENUs = +/- (0.050% /1.8) x 15 Deg F TENUs = +/- 0.417% span 5.1.3 NUS Drift Effect (DRNUs).
No drift was supplied by the vendor for this device. Per the Salem Technical Standard, (Ref 3.2) a default value should be established. A drift of +/-0.250% span is established for this device which is greater than the reference accuracy.
DRNUs = +/- 0.250% span 5.1.4 NUS Rack Power Supply Effects (PSNUs).
(Ref 3.1)
The NUS specified power supply effect is 0.050% change in output for the listed variations.
Per Salem Setpoint Technical Standard (Ref 3.2), a power supply effect of this magnitude or less may be ignored. Therefore,
- PSNUs = +/- 0.000% span 5.1.5 NUS Rack Humidity Effects (HENUs).
(Ref 3.2)
No humidity effects were supplied by the vendor. Per the Salem Setpoint Technical Standard (Ref 3.2), the effect may be assumed to be included within the stated effects.
HENUs = +/- 0.000% span
CALCULATION CONTINUATION/ SHEET: 4 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
- CALC. No.:SC-CN001-01 Attachment 10.2 ORIGINATOR, DATE REVIEWER/VERIFIER, DATE REV: CMM 12/13/93 AFS/SJJ 1/11/94
REFERENCE:
1 CMM 8/16/9411IR1 LFP 8/16/94 I
5.1.6 NUS Rack RFI/EMI Effects (REENUs}
(Ref 3.2)
No RFI or EMI effects were provided by the vendor. These effects are unlikely due to the shielding and regulation of the use of radios and other interference causing devices in the Control room. Per the Salem Setpoint Technical Standard (Ref 3.2) since no uncertainty is provided by the vendor for this effect it may be considered not applicable.
REENUs = +/- 0.000% span 5.1.7 NUS Rack Normal Radiation Effects (RENUs}
(Ref 3.2)
- No radiation effects are specified by the vendor nor are they considered applicable to the mild environment of the Control Room.
RENUs = +/- 0.000% span 5.1.8 Total NUS Uncertainty (NUS):
All random, independent uncertainties associated with the NUS isolator are combined below using the SRSS method of error combination. PS, HE, REE, and RE effects are negligible.
therefore:
NUS = +/- [(RANU5 ) 2 + (DRNU5) 2 + (TENU5)2 ]112 NUS = +/- [(0.150%)2 + (0.250%)2 + (0.417%) 2 ]112 NUS = +/- 0.509% span 6.0 Conclusions The standard rack uncertainties are +/- 1.225% span (Reference Section 4.1)
The total NUS Isolator uncertainties are +/- 0.509% span (analysis performed above)
Based on the above, the Standard Westinghouse Rack total uncertainties are greater than those calculated with the NUS performance specification. Therefore, for a rack including an NUS isolator the Standard Westinghouse rack uncertainties are bounding.
I* ~
CALCULATION CONTINUATION/ SHEET: i OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.3
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CHM 8/16/941 lIR:
I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 ATTACHMENT 10.3 MOORE SIGNAL ISOLATOR /RACK EVALUATION
CALCULATION CONTINUATION/ SHEET: 1 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.3
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/941 lIR:
I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 1.0 Purpose The purpose of this evaluation is to determine if the total rack uncertainties currently used for a standard Westinghouse Rack are bounding if the Moore signal isolator Model SCT is considered part of the rack. This evaluation assumes that the rack already includes either a standard Westinghouse Model 110 isolator or an NUS Model FIA801-05-07-08. (Ref Attachment 10.2 for NUS evaluation).
2.0 Scope This evaluation is applicable to the Instrument rack for Narrow range Steam Generator Level Recorder loops.
3.0 References 3.1 PSBP 301669, Moore Isolator specification sheet (Attached) 3.2 Salem Setpoint Technical Standard SC.DE-TS.ZZ-1904 (Q) 3.3 Calculation SC-CNOOl-01Attachment10.2 3.4 Calibration Procedures S1(2).IC-CC-RCP-0033 (Channel IV)
Calibration Procedures S1(2).IC-CC-RCP-0034 (Channel III)
Calibration Procedures S1(2).IC-CC-RCP-0035 (Channel II)
CALCULATION CONTINUATION/ SHEET: 2 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.3
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/941 1IR' I
REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 4.0 Analysis Assumption Westinghouse Isolator Model 131-110 is a high impedance, solid-state differential de amplifier with a single input and a single-ended de isolator circuit used in control circuits requiring de stability and floating de output. The NUS Model FCA-801 isolator is a comparable model with the exception of having 4 separate isolated outputs.
The Westinghouse specifications do not include uncertainties specific to the isolators.
Westinghouse does provide a general* specification for the overall rack performance which is comprised of a +/-0.5% span accuracy and a+/- 0.5% temperature effect. Considering that the rack general specification is applicable for racks with multiple instruments, it can be assumed that the individual uncertainty of the isolator may be significantly less than the overall rack specification. The NUS isolator is being purchased as an equivalent replacement for this model, therefore, this calculation assumes that the individual uncertainties for the Westinghouse isolator are comparable to the NUS specifications.
The NUS specifications were evaluated (see Ref 3.3), and the calculated uncertainty is
+/-0.509% span. The total uncertainty for the Westinghouse rack (Ref 3.2) including the NUS or the Westinghouse isolator for the effects of Rack Accuracy, Drift and Temperature Effects, is +/- 1.225% span.
Based on the assumption that the NUS uncertainties and the Westinghouse Isolator uncertainties are comparable, this calculation assumes that the addition of another device within this rack would not affect the analysis, as long as the uncertainty of the additional device combined SRSS with the isolator is still within+/- 1.225%.
CALCULATION CONTINUATION/ SHEET: 3 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.3
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/941 1IR:l I
REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 5.0 Design Inputs 5.1 Westinghouse Rack Accuracy +/-1.225% (Ref 3.3) 5.2 NUS Accuracy +/- 0.509% span (Ref 3.3) 5.3 Moore Isolator Performance Specification (Ref 3.1)
Accuracy +/- 0.1 % of span (linearity and repeatability)
Load Effect +/- 0.01 % span from 0 to max load resistance .
Temperature Effect+/- 0.005% span per Deg F over -20 to 180 Deg F.
Line Voltage Effect +/- 0.005%/1 % line change 6.0 Determination of Moore Isolator Uncertainty 6.1 Uncertainties for the Moore Signal Isolator are from Design Inputs (taken from Reference 3.1) unless otherwise noted.
Signal Isolator Reference Accuracy (RA18ol Per Design Inputs, the Reference accuracy is 0.1 % span.
RA180 = +/- 0.1 % span 6.2 Isolator Temperature Effect (TE18ol Per Design Inputs, the Isolator Temperature Effect may be assumed to be within+/- 0.005%
span per Deg F change within the specified operational range of -20 to 180 Deg F. The control room temperature variation from calibration is 15 Deg F. Therefore, the total temperature effect is:
TE180 = +/- (0.005% / 1 Deg F ) x 15 Deg F.
TE180 = +/- 0.075% span
CALCULATION CONTINUATION/ SHEET: 4 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.3
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/941 1IR:
I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 6.3 Isolator Power Supply Effects (PS15o).
Per design inputs, the line voltage effect is +/- 0.005%/1 % variation. Per the power supply regulation, the variation may be +/- 2%. Therefore, the total uncertainty will be 0.010% span.
PS150 = +/- 0.010% span 6.4 Isolator Drift (DR1sol No drift was supplied by the vendor for this device. Per the Salem Technical Standard, (Ref 3.2), a default value should be established. A drift of +/- 0.250% span is established for this device which is greater than twice the reference accuracy.
DR150 = +/- 0.250% span 6.5 Isolator Miscellaneous Effects (ME15o).
Per design inputs, the isolator has a load effect of +/- 0.010% of span.
ME150 = +/- 0.010% span 6.6 Total Isolator Uncertainty (ISO)
The total uncertainty is the SRSS combination of all calculated uncertainties associated with the isolator.
ISO = +/- [(RA1so) 2 + (TE1so) 2 + (PS1so) 2 + (ME1so )2 + (DR1so )2]112 ISO = +/- [(0.100%)2 + (0.075%)2 + (0.010%)2 + (0.010%)2 + (0.250%)2]112 ISO = +/- 0.280% Span
CALCULATION CONTINUATION/ SHEET: 5 OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CN001-01 Attachment 10.3
REFERENCE:
ORIGINATOR, DATE REV: CMM 12/13/93 1 CMM 8/16/941 lIR:l I
REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 7.0 Rack ~th NUS/Westinghouse Isolator and the Moore Isolator Per Assumption 4.1, the Westinghouse Isolator is assumed to have the same accuracy as the NUS isolator at 0.509% span.
Rack with Moore = +/- [(NUS/Westinghouse)2 + (Moore)2]1f2 Rack with Moore = +/- [(0.509%)2 + (0.280%)2] 112 Rack with Moore = +/- 0.581 % span 8.0 Conclusions Per the analysis of section 4.0, the addition of the Moore uncertainty is acceptable with no change in the standard rack uncertainties, providing the Moore uncertainties are within the remaining standard rack uncertainties available after the Westinghouse or NUS isolator uncertainties are accounted for.
Based on the Moore uncertainty calculated above, and assuming that the Westinghouse or NUS isolators in the loop are within 0.509% span, the addition of the Moore isolator uncertainty of +/- 0.280% span results in a combined uncertainty of 0.581 % span. This uncertainty is within the 1.225% which is used as the standard rack assumption for Accuracy, Temperature Effect and Drift. Since the rack only includes the two isolators, the Moore uncertainties do not cause the rack to exceed the uncertainties assumed for the rack.
Therefore, the standard rack assumptions are bounding for this particular case.
CALCULATION CONTINUATION/ SHEET: i OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CN001-01 Attachment 10.4
REFERENCE:
ORIGINATOR. DATE REV: CMM 12/13/93 1 CMM 8/16/941 lIR I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 ATTACHMENT 10.4 WESTINGHOUSE LEITER; S/G WATER LEVEL PMA TERM INACCURACIES
~ \ OF- 10
':.:*' ?- 6 *q7,._
Westinghouse Energy Systems Electric Corporation Mr. J. A. Nichols, Manager Reliability & Assessment Public Service Electric and Gas Company P. O. Box 236 Hancocks Bridge, New Jersey 08038 Public Service Electric and Gas Company Salem Units No. 1 and 2 SIG Water Level PMA Term Inaccuracies
Dear Mr. Nichols:
The purpose of this letter is to inform your plant that the Process Measurement Accuracy CPMA) term, based on standard Westinghouse Methodology, for Steam Generator Water Level instrumentation uncertainty calculations may.
be non-conservative. This would impact the protection functions which use this parameter, i.e. Steam Generator Water Level - Low, Low-Low, and High-High. The magnitude of the impact is plant specific. It is affected by the steam generator model* number and is sensitive to the calibration conditions used by the plant (process pressure and reference leg temperature) and tap locations.
The standard Westinghouse methodology used a random value of +/-2.0t span for this term in setpoint uncertainty calculations for all models of steam generator design. This value was based on the density variation as a function of power ~nd level, and the assumption that calibration was performed at soi power conditions. For several of the models, the fluid velocity effect was known to introduce a significant bias in the low direction and a separate allowance was incorporated for this effect for Steam Generator Water Level -
High-High.
More recently, *an improved understanding of ~P level measurement system errors based on scientific work documented in an Instrument Society of America paper CG. E~ Lang and J. p. Cunningham, 0elta-P Level Measurement Systems.
11 II 11 Instrumentation, Controls, and Automation in the Power Industry, Vol. 34, 11 Proceedings of the Thirty-Fourth Power Instrumentation Symposium, June 1991),
has led to a reinvestigation of the Steam Generator Level Process Measurement Accuracy terms. The conclusions are that two other error components should be accounted for explicitly Ci.e~, reference leg temperature changes from calibration temperature, and downcomer subcooling) and that fluid velocity
fd- 2 DF- JO June 18, 1992 Page 2 effects should be considered for alt steam generator models. In add1tion, the assumption of calibration at sot power may not be conservative with respect to actual calibration conditions used by the plants. These error components are not considered to be random in nature, and should therefore be treated as biases.
Two cases were evaluated to determine the potential magnitude of the 1mpact of the additional errors on the total channel uncertainty, and are discussed in detail i~ the attachment. The first case used plant specific data for a three loop plant with a Model 51, and is expected to be typical of the effects for that steam generator model. The second case is considered to be a bounding evaluation for a Model F steam generator. Based on these evaluations, the previous uncertainty calculations for Steam Generator Hater Level - Low and Low-Low may be nonconservative by approximately 1 to 2i span. The potential nonconservatism for Steam Generator Hater Level - High-High ranges from 1 to 161. It must be emphasized that the magnitude of the impact is plant specific as well as model specific, and is sensitive to the calibration conditions used by the plant (process pressure and reference leg temperature) and tap locations.
- aased on engineering judgement, Hestinghouse believes that, although potentially outside the existing licensing basis, the required safety functions can still be performed by either 1) existing automatic systems on a best estimate basis, or 2) operator action. Therefore, this issue would not constitute a Substantial Safety Hazard pursuant to the requirements of 10 CFR Part 21. It is recommended, however, that the impact of this issue on your plant be evaluated.
If there are any questions, please contact the undersigned.
Very truly yours, J. N. Steinmetz, Central Area Domestic Customer Projects 0269K
.se_.c.,..,Joo 1-0 I r&-J I A.ilA,~NT IO' t/
WESTINGHOUSE PROPRIETARY Cl.ASS 2 ~ 3 oi= .10 Steam Generator Water Level Process Measurement Accuracy Terms and Setpoint Uncertianties Westinghouse has determined that the Process Measurement Accuracy term, based on standard Westinghouse methodology, for Steam Generator Water Level instrumentation uncertainty calculations may be nonconservative. This would impact the protection functions which use this parameter, i.e. Steam Generator Water Level - Low, Low-Low, and High-High. Westinghouse has performed evaluations of two cases which are described below to determine the potential impact of the increased uncertainty. Eased on these evaluations, the previous uncertainty calculations for Steam Generator Water Level - Low and Low-Low may be nonconservative by approximately 1 to 2% span. The potential nonconservatism for Steam Generator Water Level - High-High ranges from 1 to 16%. The magnitude of the impact is plant specific.
It is.affected by the steam generator model and is sensitive to the calibration conditions used by the plant (process pressure and reference leg temperature) and tc.p locations. A:though poten~ially outside the existing licensing basis, the required safety functions can still be performed by either 1) existing automatic systems on a best estimate basis, or 2) operator action. It is recommended, however, that the impact of this issue on your plant be evaluated.
ISStm: DBSClUPTION Ba*ic Compon~t The basic component involved in this issue is the Steam Generator Water Level instrumentation and the associated uncertainty analysis.
This uncertainty analysis, if based on standard Westinghouse methodology, includes an uncertainty term to account for Process Measurement Accuracy. Historically, a random value of +/-2.0% span has been used for this term in setpoint uncertainty calculations for all models of steam generator design. This value was based on the density variation as a function of power and level, and the assumption that calibration was performed at 50% power conditions.
For several of the models, the fluid velocity effect was known to introduce a significant bias in the low direction and a separate allowance was incorporated for this effect for Steam Generator Water Level - High-High.
Deviation The issue concerning the Steam Generator Water Level Process Measurement Accuracy term is that the use of a random +/-2.0% value may be nonconservative. A paper presented at an Instrument Society of America (ISA) conference in June, 1991 (G. E. Lang and J. P.
Cunningham, "Delta-P Level Measurement Systems","Instrumentation, Controls, and Automation in the Power Industry", Vol. 34, Proceedings of the Thirty-Fourth Power Instrumentation Symposium) provided information leading to the conclusion that this uncertainty term should be re-evaluated. In particular, process pressure variation
effects on density, reference leg temperature changes under normal operating conditions, downcomer subcooling, and fluid velocity effects for all steam generator models should be accounted for explicitly. In addition, the assumption of calibration at 50% power may not be conservative with respect to actual calibration conditions used by a particular plant. These error components are not considered to be random in nature, and should therefore be treated as biases. The equations used in determining these error components were present~d in the ISA paper referenced and are repeated below.
Process Pressure Variations After installation of the level measurement system on the steam generator, it is calibrated for a specific set of operating conditions, i.e., a reference leg temperature and process pressure.
If the process pressure changes as a conseque~c& ~f changing operating conditions, then a level measurement error is created. An approximation of the measurement error, due to changes in process press'ure (assuming the temperature of the fluid in the vessel is at the saturation temperature corresponding to the steam pressure) is:
£, r[.:; Jr P;, - P,.) [i] [p, - PJ - t~i]
+ (100) t .[ Fl 2
pre - p H where:
L
£p
-- measurement uncertainty in percent of the*levelspan actual water level in the vessel above the lower tap (ft)
Hi:. - maximum vertical distance from the lower tap to water level H
- in the condensate pot at the upper tap (ft) vertical distance between upper and lower taps on the vessel, i.e., the level span (ft) pg - dry saturated steam density at the process pressure (lbm/ft 3 )
Pt Pqc -
- II saturated water density at the process pressure (lbm/ft 3 )
saturated steam density at the calibration pressure Ptc - (lbm/ft 3 )
saturated water density at the cali:br:it!.on pressure ::.f. the
- system is hot calibratedi or water density at the calibration pressure and temperature if the system is cold calibrated (lbm/ftl).
For a given protec~ion function, this uncertainty will be a bias, e.g., for Steam Generator Water Level - Low-Low, assuming calibration at 100% Rated Thermal Power (RTP) conditions, the process pressure variation is a negative bias for 0% level at 0% RTP and a positive bias for 100% level at 0% RTP .(the two limiting conditions for instrumentation for !eedwater line break and feedwater malfunction) .
SC.- c...JcoL-OI eaJ f WESTINGHOUSE PROPRIETARY CL.ASS 2 Reference Leg Temcerature Variations In addition to assuming a process pressure when the level measurement system is calibrated, a reference leg temperature is assumed. This uncertainty addresses the changes in normal operation ambient temperature, not the elevated containment ambient temperatures experienced in an inside containment high energy line break.
Typically, a specific operational temperature is assumed for the purpose of calibration and an allowed operational band is assumed about the reference temperature. Westinghouse calculates two uncertainties for this variable, one in the high direction (bounded by 130 °F) and one in the low direction (typically 100 °F) . These are considered reasonable operational limits for this purpose. The equation used to determine the reference leg temperature variation uncertainty is:
[-i ](P:... - Pi:} (100 >
[Pre - Pgc]
where:
ET measurement uncertainty in percent of level span Hi. maximum vertical distance from the lower tap to the water level in the condensate pot at the upper tap (ft)
H
- vertical distance between the upper and lower taps on the vessel, i.e., the level span (ft)
- Pr.c water density at the calibration ambient conditions (process pressure and reference leg temperature at which the calibration was per~ormed) (lbm/£*::= i Pi.
Pt..
-- water density in the reference leg (lbm/ft 3 )
saturated water density at the calibration pressure if the system is hot calibrated, or water density at the calibration pressure and temperature if the system is cold calibrated (lbm/ft 3 ) .
- saturated steam density at the calibration pressure (lbm/ft 3 )
- For a given protection function, this uncertainty will be a bias, e.g., for Steam Generator Water Level - Low-Low, assuming calibration at a reference condition of 110 °F and allowed temperature swings of up to 130 nF and down to 100 °F, th~ 130 °F error is a bias in the indicated high level direction and the 100 °F error is a bias in the indicated low direction. Thus to be conservative, for the Low-Low reactor trip, the indicated high error is used. For Steam Generator Water Level - High-High the indicated low direction is conservative and the low temperature error is used.
Fluid Velocity Effects When performing a calibration of the Steam Generator Water Level channels, the fluid velocity near the tap locations has been assumed to be negligible such that a differential pressure would not be induced due to fluid flow. Howeve,, this is nc~ ~~a case for ~~e
- ---* - *-~--.a
- sc-Lt--icol-oi ~v 1 ~"flA'-Hrne.i.J1 lO, Y
.fd. ~ or . -jO WESTlNUHOUSE PROPRlEfARY CLASS 2 lower tap due to shell and internals design. The upper tap is assumed to be in the steam space. An approximation of the error introduced by fluid velocity effects past the lower tap is:
_ r~r [i. o
- Kt] (100)
£ v ..
2 (H) (gc) <Prrrl [Pt.. - Pgc]
where:
measurement uncertainty in percent of the level span fluid flow rate normal to the lower tap (lbm/sec)
- flow area at the lo#er tap (ft2) friction and form loss factor vertical distance between the upper and lower taps on the vessel, i.e., the level span (ft) gravitational constant (ft/sec2)
- water density in the vicinity of the lower tap (lbm/ft 3 )
saturated water density at the calibration pressure if the system is hot calibrated, or water density at the calibration pressure and temperature i! the system is cold calibrated (lbm/ftl).
saturated steam density at the calibration pressure (lbm/ftl) .
pressure drop in the downcomer to the lower tap (lbm)
- steam flow at rated thermal power conditions circulation ratio at rated thermal power conditions.
This uncertainty is a bias in the indicated low level direction. The magnitude varies as a function of power, thus an appropriate value must be used !or each specific protection function depending on the conditions for the event, e.g., Steam Generator Water Level - Low-Low is used !or both zero power and 100% power events. The smallest magnitude negative error is at zero power, thus it is acceptable to use the zero power value for both zero power and 100% power events.
The highest magnitude negative error typically occurs between 50 and 70% power, thus for Steam Gen<;!rator Water Lev:-l
- High-High, .:t is conservative -to u~e the part power value for a Feedwater Malfunction event.
Downcomer Suhcooling Effects*
Another source o! measurement error is the subcooling of the fluid in the downcomer region in conjunction with a saturated mixture around the steam generator U-tubes. The magnitude of the subcooling in the downcomer is dependent upon the following process conditions; main
!eedwater temperature, circulation ratio, and location of the
!eedwater nozzle with respect to the low level tap. This uncertainty is determined by the following:
,b.~MENT 10.t.J-fS 7()r rV WESTINGHOUSE PROPRIETARY CLASS 2
£ s ..
(p irr - pt] (1 0 0 ) rLH"H. L]ain 71 (p .-.. - Pgc:J I where:
measurement uncertainty in percent of the level span water density in the vicinity of the lower tap (lbm/ft 3 )
saturated water density at the process pressure (lbm/ft 3 )
saturated steam density at the calibration pressure Ptc .. (lbm/ftJ) saturated water density at the calibration pressure if the system is hot calibrated, or water density at the calibration pressure and temperature if the system is cold calibrated (lbm/ftl).
- the maximum height of the water column above the lower tap that is assumed to be subcooled (ft)
H. vertical distance between upper and lower taps on the vessel, i.e., the level span (ft)
L actual water level in the vessel above the lower tap (ft)
- This uncertainty is a bias in the indicated high direction, thus it is non-conservative for the Steam Generator Water Level - Low-Low function (and should therefore be accounted for) and conservative for the High-High function (and may be ignored) .
- Two cases were evaluated to determine the potential magnitude of the impact of the additional errors on the total channel uncertainty, and are discussed in detail below. The first case used plant specific data from a three loop plant, and is expected to be typical of the effects for a Model 51 steam generator. The second case is considered to be a bounding evaluation for a Model F steam generator.
Based on these evaluations, the previous uncertainty calculations for Steam Generator Water Level - Low and Low-Low may be nonconservative by approximately 1 to 2% span. The potential nonconservatism for Steam Generator Water Level - High-High ranges from l to 16%. It must be emphasized that the magnitude of the impact is plant specific as well as model specific, and is sensitive to the calibration conditions used by the plant (process pressure and reference leg temperature) and tap locations.
For the three loop plant the reference conditions and the magnitudes of the effects are as follows:
Reference conditions ll0°F Reference Leg Temperature, 792 psia, 100% RTP
~recess pressure +l.1% span (ll0°F, 1020 psia, 0% level]
variation -4.0% span [ll0°F, 1020 psia, 100% lev~ll Reference leg +0.7% span (l30°F, 790 psia, any level]
temperature -0.3% span (l00°F, 790 psia, any level]
WESTINGHOUSE PROPRIETARY Cl.ASS 2 Fluid velocity effects -0.7% span Downcomer subcooling +0.5% span For Steam Generator Water Level - Low-Low, the summation of the applicable terms is:
+1.1 + 0.7 - 0.7 + 0.5 * +1.6% span.
For Steam Generator Water Level - High-High, the summation of the applicable terms is:
-4.0 - 0.3 - 0.7 + 0.5 * -4.5% span.
Using these values as biases in the uncertainty calculations, resulted in increases of approximately 1.0% span in the total channel uncertainty for Steam Generator Water Level - Low-Low, and approximately 3.9% span for High-High, relative to using the +/-2.0 random term. It was determined that sufficient margin existed to accommodate these values* for this specific plant.
For a generic case assuming a Model F Steam Generator, the reference conditions and the magnitudes of the effects are based on a two loop plant and are considered to be bounding:
Reference conditions 110°F, 760 psi a, 100% RTP Process pressu:ce +1.0% span [110 °F I 954 psi a, 0% level]
variation -3.4% span [110°F, 954 psi a, 100% level]
Reference l.eg +0.7% span [130°F, 760 psi a, any level]
temperature -0.3% span [100 °F, 760 psia, any level]
Fluid velocity effects -14.2% span [70% RTPJ 0.0 to -2.0% span [0% RTPJ Downcomer subcooling +1.6% span For Steam Generator Water Level. - Low-Low, the summation of the applicable terms is:
+1.0 + 0.7 - 0.0 + 1.6 * +3.3% span (conservative cal.culation) or
+1.0 + 0.7 - 2.0 + 1.6 * +l.3% span (better estimate calcul.ation)
- For Steam Generator Water Level - High-High, the summation of the applicable is:
-3.4 - 0.3 - 14.2 + 1.6 - -16.3% span.
- Using these values as biases in the uncertainty calculations, resulted in increases of approximately 1.7 to 2.7% span in the total channel uncertainty for Steam Generator Water LQ~e~. - Lov-Low, .:.nd approximately 15.7% span for High-High, relative to using the +/-2.0 random term. It should be noted that calibration at any power level less than 100% RTP will result in decreases in the process pressure variation terms for both lov and high levels. Therefore the above is a worst case situation. Calibration at 50% RTP results in process pressure variation terms of +0.7% span and -1.6% span for the low and
WESTINGHOUSE PROPRIETARY CLASS 2 high levels, respectively. This change results in correspondingly lower total channel uncertainties since the ter:n is applied as a bias.
The safety significance of this issue is a function of calibration conditions, steam generator model, and the margin present in the existing trip setpoints and safety analyses. Without specific knowledge of the calibration conditions used at the plants, i Westinghouse cannot make a definitive determination of safety significance. However, based on engineering judgement as discussed below, the increase in uncertainty on Steam Generator Water Level
..I Low-Low is small, and on a best estimate basis the existing acceptance criteria for currently analyzed events would be
=--**
maintained.
. For the Steam Generator Water Level Low-Low uncertainty calculation, there is typically a small degree of margin (0.5 to 1.0% spc~)
between the total channel aliowance (Safety Analysis Limit minus Nominal Trip Setpoint) and the total channel uncertainty. Based on a bounding increase in total channel uncertainty of 2.0%, an additional 1.0 to 1.5% must be accommodated. This can be found on an interim basis in the Environmental Allowance (EA) term. Westinghouse typically specifies an EA magnitude based on the postulated Steambreak environment, which is enveloped by a maximu.~ temperature of 420°F. Each transmitter supplied by Westinghouse is temperature compensated at a steady state 320°F based on the belief that the electronics will not see the maximum temoerature due to thermal shielding and inertia of the transmitter.casing. Typical maximum Feedwater line break aml:>ient temperatures are postulated to be approximately 350°F several minutes into the event, while a typical time of reactor trip is less than 60 seconds into the event. This would indicate that the transmitter will see an ambient temperature significantly less than 350°F and the electronics would see an even lower temperature. Assuming that the EA magnitude is linear from 6%
at 320°F to 0% at 130°F, a steady state temperature of approximately 250°F would result in an EA ter.n of 4% span. This is significantly more than postulated containment ambient temperatures at 60 seconds into a Feedwater line break. Thus it is reasonable to assume that an additional 2% span is available for interim margin considerations.
These assumptions are based on engineering judgment and may be outside the plant licensing basis, but may be considered in developing the basis for contlnued plant operation until a plant specific evaluat~on can be completed.
The Steam Generator Water Level - High-High reactor trip is provided for a Feedwater Malfunction event which results in an uncontrolled increase in level. The primary effect of this event is a.n increase in moisture carryover which can cause significant turbine blade erosion if not corrected. Thi3 is primarily a commercial concern, i.e. not an immediate safety concern, and there would be time for operator action to terminate the event if thi3 were the only concern .
. An additional concern, however, is the filling of the steam generator
WESTINGHOUSE PROPRIETARY CLASS 2 with subsequent filling of the steam lines with water. The steam line piping supports may not be designed to support the dead weight of water in the steam lines, therefore the event must be terminated prior to creation of a steam line break due to piping support failure.
A small increase in uncertainty in the High-High trip (on the order of 5% span or less) can typic3ll] be accommoG-t~d within the margin in this channel or by increases in the Safety Analysis Limit.
However for large increases as exhibited in the bounding Model F evaluation above, more detailed evaluation may be necessary.
Although typically outside plant licensing bases, control system alarms are available to initiate operator action for mitigation.
That is, when level deviates outside the control band (typically 5%
span), or when there is a significant mismatch between steam flow and feedwater flow, an alarm sounds for operator notification. This would initiate operator action to terminate the event, thus preventing the filling of the steam generator or the steam lines.
Westinghouse has concluded that this issue would not constitute a Substantial Safety Hazard pursuant to the requirements of 10 CFR Part 21 based on the availability and use of existing automatic systems on a best estimate basis or operator action. Since Westinghouse does not have the capability (i.e., Westinghouse does not have knowledge of plant specific calibration conditions) to perform a plant specific evaluation of this issue, it is being communicated so that a regulatory evaluation can be performed.
The potential increase in tota*l channel uncertainty for those channels involving Steam Generator Level should be evaluated based on the above discussion, and a determination made as to whether the current trip setpoints are acceptable.
As described in the technical evaluation section Westinghouse performed evaluations for two cases, one of which was for a specific plant with a Mode1*s1 steam generator and the* other was for a bounding configuration with a Model F steam generator. A letter will be sent to all utilities for whom Westinghouse has performed a setpoint uncertainty evaluation informing them of the potential issue. In addition, in all future setpoint uncertainty evaluations performed by Westinghouse, these Process Measurement Accuracy terms will be explicit~y included. -- __ .-
- CALCULATION CONTINUATION/ SHEET: i OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.5
REFERENCE:
ORIGINATOR, DATE REV: CMK 12/13/93 1 CMK 8/16/941 lIR:
I REVIEWER/VERIFIER, DATE AFS/SJJ 1/11/94 LFP 8/16/94 ATTACHMENT 10.5 WESTINGHOUSE LETTER; SAFETY ANALYSIS LIMITS
SC- e."1 oo /- oI IZcrl I FEB* 11 '94 15:59 FROM OPL LICENSING TO 86093391234 PAGE.002/009 Westinghouse Becrrtc Corporation Energy Systems NUCLEAR FUEL GROUP FES 1 4 '94 IV Fs- t:ff-O 'l 2
- PSE-94-532
!?Rec'd By /J:::...~
Q""Copy to Tl<&; p.ac.. ./
Gr' CoPyiP ::s. :J.c.....e.fy<':i:\c) \/"'
- Mr. E.S.Rosenfeld,Manager 0 Route to- - - - - -
Nuclear Fuel Gr' Col'Y t& II. Otlbr+!z> (t.*e.) February 11, 1994 Public Setvice Electric & Gas C.ompa.ny 0 Return to ET-NSL-OPL-Il-94--063 P.O. Box 236 MC N20 ICF Orig.Ul"Copy 0 NSR CJ Hancocb Bridge, NJ 08038 Cover only to ICF O Attachments filed Public Savice Electric & Gas Company Salem Units I and 2
Subject:
Safety Evaluation for an Increase in Steam Generator High-High Level Setpoint Analysis Value
Dear Mf. Rosenfeld:
The pmpose of this letter is to transmit to you a safety evaluation of an increase in the steam
- generator water level high-high ESF safety analysis limit setpoint from 73 % NRS to 7S %- .NRS. The current Technical Specification nominal setpoi:at (f. S. Table 3.3-4) is 67% of narrow I3Dge span (NRS). The Salem UFSAR (Section 15.1.3) indicates that the accident analysis assumes a safety analysis limit setpoint of 75% NRS, while more recent information such as the Setpoint Study, WCAP-12103, indicar.es that the accident analysis iwumes 73% NRS.
The effect of die increased steam generat.or water level high-high ESF safety analysis limit setpoint oo the Salem UFSAR Chapter 15 accident analyses bas be.en evaluated and shown to be acceptable for both current operation and after implementation of the FUIMRP analyses.
If you have any questions or comments, please contact the undersigned.
Very truly yours,
(\--rJ--,~~r.
J~
Special Sales Repr~ve
+-
Power Sysiem., Field Sales AMS/
cc: R. S. Keat MCN20 1L, IA T. K. Ross MCN20 lL, IA Attachmem: SECL-94-042, ~Inc~ in Steam Generator High-High Level Setpoint Analysis Value, 7 pages .
SC!.- ct-loo 1-01 Rctl I FEB 11 '94 15:59 FROM OPL LICENSING TO 86093391234 PAGE.003/009 SECL-94-042 Page 1 of7 W::&m:NGllOUSE SAFETY EVALUATION CHECK LISI'
- 1) NUO...EAR PLANTS: Salem Unffsl and 2
- 2) CHECK UST APPLICABLE TO: Increase in Stram Geuttator ffi&&-H'lfh Level SeJpoint Analysis Yalue -
- 3) The written safety eval~on of the revised procedure, de.sign change or modification required by 10CFR5Q.59 has been prepared to the extent re.quired and is attached. If a safety evaluation is not required or is iDcomplete for any reason, explain on Page 2. Parts A and B of this Safety Evaluation Oieclc List are to be completed only on the b~is of the safety evaluation pCrformed.
CHECK UST - PART A 3_1) Yes_ No.X.. A change.to the plant as described in the FSAR?
3_2) Yes_ No..X.. A change to procedures as described in the FSAR?
3.3) Yes_ No-X. A test or experiment not described in the FSAR?
3.4) Yes_ No.X.. A change to the plant redinical specificatious (Appendix A to the Operating License)?
- 4) CHECK LIST - PART B f]ustificad.on for Part B llll5Wers must be included on page 2.)
4.1) Yes_ NoX Will the probability of an accident previously evaluated in the FSAR be.
lncreased?
4.2) Yes_ NoJL Will the consequences of an accident previously evaluated in the :FSAR.
be increased?
4_3) Yes_ NoX Ma.y the possibility of an accideDt whic.h is different than any ah:eady evaluated in the FSAR be created?
4-4) Yes_ No.X.. Will the probability of a malfunction of equipment important to safety previously evalwired in the FSAR be increased?
4.5) Yes_ NoX Will the consequences of a malfunction of equipment important to safety previo'USly evaluated in the FSAR be increased'!
4.6) Yes_ Nol(_ May the possibility of a malfunction of equipment important to safety different 1han any already evaluated in the FSAR be created?
- 4. 7) Yes_ No_x_ Will the margin of safety as defined in the bases to any technical specification be reduced?
sc- c.Fdoo 1-0 I REI/ I 4rr;::;c!lm~ ,...;r /0.6' stl .3 or<?'
FEB 11 '94 15:59 FROM OPL LICENSING TO 86093391234 PAGE.004/009 SECL-94-00 Page 2 or7 If the answers to any of the above q~ions are unknown, indicate under S) REMARKS and explain below.
If the answec to any of the above questions in Part (3-4) or Part B cannot be answered in the negative, the change review .requires an application for license amendment in accordance with 10 CFR 50.59 (c) and submitted to the NRC pursuant to 10 CFR 50.90.
S) REMARKS; The answers given in Section 3, Part A, and Section 4, Part B, of the Safety Evaluation Checklist. are based on the attached Safety Evaluation.
Reference document(s):
FOR FSAR UPDATE Section: N/A Pages: _ _ __ Tables: _ __ Figures:._ _ __
Reason for I Description of Change:
No change was made to the FSAR value of the S/G High-High level seq>oint.
SIGNATURES Prepared by (Li=smg): ~ A. Sicari Date: 1~I/- §1
~~~wcdc#tl R.H. Owoc* '-=:::::
.Date: J.h.lr'-f
.:S{!.-C/\100 I- 01 REt/ I FEB 11 '94 15:59 FROM OPL LICENSING TO 86093391234 SECl...-94-042 Page3 of7 Salem Units 1 and 2 Safety Evaluation for an Increase in the Steam Generator Bigh-ffi.gb Level Setpoint Analysis Valoe Imroduction and Summary of Results Recent Salem setpoint uncertainty calculations have indicated the need to justify a relaxed safety analysis assumption for the steam generator water level high-high F.SF setpoint, which actuates a turbine trip and feedwater isolation.
The pmpose of this evaluation iS to address aii increase in the steam generator water level high-high BSF safety analysis limit setpoint from 73 % NRS to 75 % NRS. The current Technical Specification nominal seipoint (T. S. Table 3.3-4) is 67% of D3ITOW range span (NRS). However, there is conflicting documemation concerning the accident aoalysis assumption for this setpOint. The Salem UFSAR (Section 15.1.3) indicates that the accident analysis as.sum.es a safety analysis limit setpoint of 75% NRS, while more recent infoxmation such as the Setpoint Study, WCAP-12103, indicates that the accident analysis assumes 73%
NRS.
The effect of the incteased steam generator water level high-high F.sF safety analysis limit setpoint on the Salem UFSAR Chapter 15 accident analyses bas been evaluated and shown to be acceptable for both current operation and after implementation of the FU/MRP analyses.
1bc calculared flow capacity does not result in an u.o.reviewed safety question as defined in 10CFRS0.59. .
Liceming B:ms 1be only Salem licensing basis event that assumes protection functions initiated by the bigh-bigh st.eam genentor water level secpoint is the non-LOCA feedwater malfunction event presented in UFSAR Section 15.2.10. The high-high steam generator water level protection function is as.uned to close all feedwater control and bypass valves, and the feedwater isolation valves, trip the main feedwatcr pumps, and trip the turbine. For convenience, the event is tenninated by a reactor trip on tuibine trip.
The feedwater maifnnction event is an ANS Condition II event. The Condition Il acceptance .
criteria are satisfied for this event by demonstrating that the DNB design basis is met Another concern for thin event, not specifically addresse.d in the UFSAR, is steam generator overfill.
- Sf/ S or=J>
FEB 11 '94 15:59 FROM OPL LI CENS ms TO 86093391234 PAGE.006/009 SECL-94-042 Page 4of7 Evaluation The feedwater malfunction analysis is perfonned at zero and full power. The analysis
~mes that a control system malfu.oction or ope:raror error causes one or more feedwarer control valves to open fully, resultiog in a step increase in fcedwate:r flow. The zero power analysis does not assume the high-high steam generator wau-.r level protection function and is not affected by a change in the 5e4>0int value. The full power analysis credits the high-high steam generator water level protection function t.o terminate the event by isolating the main feedwater and tripping the twbine. Reactor trip occurs on tnrbi.ne trip.
The high-high steam generator water level protection functions are not required to meet the DNBR limiL Although the UFSAR analysis shows that the minimum DNBR occurs after the tuibine trip but prior t.o reactor trip and feedwater isolation occur, the DNBR had reached a new equilibrium value well above the limit valiie. The Fuel Upgrade/Margin Recovery Program (FU/MRP) analysis DNBR also had reached a new equilibrium value which was well above the DNBR limit before the high-high steam generator water level protection functions were actuated. In either case, had the core thermal limits been approached, the overtempera...ture AT and/or ove.rpower AT reactor trips would prevent the reactor core from reaching a condition which could "result io a violation of the DNB design basis.
With respect to steam generator ovedill, the analyses assume a feedwarer isolation 32 seconds after the high-high level setpoint is reached. Although not explicitly credited in the analysis, prior to feedwater isolation valve closure, the feedwatet pumps are tripped on a high-high steam generator water level signal These functions will continue to prevent steam geneDltor overfill.
The evaluation of the pertinent non-LOCA events demonstrates that the as-installed capacity of the oondenser steam dump system does not change the conclusions of the UFSAR. In addition, the conclusions "presented in the UFSAR remain bounding for:
LOCA and LOCA-Related Accidents Steam Generat.or Tube R.uptmc Containment Integrity In.strnmentatioo and Control Systems Performance Radiological Consequences Equipment Qualificati.oolComponent Performance Technical Spccifications/Setpoints
- Emergency Operating Procedmes
.SC- Ct</OO /-0 I f:// I F~B 11 '94 16:00 FROM DPL LICENSING TD 86093391234 PAGE.007/009 SECL-94-042 Page5 of7 Conclmion This evaluation concludes that an increase in the safety analysis limit high-high steam geoemor water level setpoint from 73 % NRS to 75 % NRS will not signif'u::antly affect the safety analyses. For the feedwar.er malfunction event all applicable criteria continue r.o be met.
Assesmlent of Unreviewed Safety Question
- 1. Will the probability of an accident previously evaluated in the UFSAR be increased?
The revised safety analysis limit high-high steam generator water level setpoint does not involve an ~ in the probability of an accident previously evaluated. The high-high steam genetator water level setpoint is a part of the accident mitigation
- response and is not itself an initiator for any transient. The accident which relies on the high-high Jevel setpoint bas been evaluated and all applicable safety criteria continue to be met.. The.refore, the change will not result in any additional challenges to plant equipment. The consideration of a revised high-high level setpoint analysis value does not result in a situation where the design, material, and constroction standards that were applicable prior to the change are altered. The evaluation of the change indicates that it will not affect the operability of systems related to accident mitigation. Since the a<;tual plant configuration, pexfonna.oce of systems, and initiating event mechanisms axe not being changed as a result of this evaluation, the probability of any accident previously evaluated int.be UFSAR is not changed.
- 2. Will the consequences of an accident previously evaluated :in the UFSAR be increased?
The change to the safety analysis limit high-high steam generator water level setpoint does not increase the co~ences of an a.ccidmt previously evaluated. All applicable accident analysis acceptance criteria. continue to be met. The transient which is affected by the change to the high-high level setpoint safety analysis limit has been evaluated and all applicable safety criteria. continue to be met. The :tevised safety analysis limit does not degrade or prevent the response of other plant systems such that their function in the control of radiological consequences is adversely affected. The safety evaluation shows that the design limits continue to be met and therefore fission baxrier integrity is not challenged. The slight increase in the safety analysis value has been shown not to adversely affe.ct the response of the plant to postulated accident scenarios. Nor does this change affect the mitigation of t.he radiological consequen~ of any accident described in the UFSAR. Therefore, since the actual plant coofiguiati.on and pe.rformance of systems is not being changed, and since it bas been concluded that the transient results are unaffected by this paxameter
5fl 7 or= 8 FEB 11 '94 16:01 FRCJ1 OP!_ LICENSIMG TO 86093391234 SECJ.,.94--042 P&&e6 of7 modification, the consequences of an accident previously evaluated in the UFSAR will not be increased.
- 3. May the possibility of an accident.which is different than any already evaluated in the UFSAR be created?
The revised safety analysis limit high-high steam generator ware.r level setpoint does not create the possibility of a new or different kind of accideot from any accident previously evaluated. The evaluation of the change shows that all safety criteria continue to be mei. The setpoiut adjustment does not afi'ect the assumed accident initiation sequences. Therefore, this change neither causes the initiation of any different accident nor creates any new failure mechanisms. The possibility of an accident which is different than any already evaluated in the UFSAR is not created since the revised steam geoexator high-high level setpoint safety analysis limit does not result in a change to the main steam system or any other plant system design basis. No new operating configuration is being imposed by the setpoint adjustment that would create a new failure scenario. In addition1 no new failure modes are being created for any plant equipment. This change does not result in any event previously deemed incredible being made credible. Therefore, the types of accidents defmed in the UFSAR continue to represent the credible spectrum of events to be anal.yud which detennine safe plant opeiation and the pos.si'b.ility of an accident different than any already evaluated in the UFSAR is not created.
- 4. Will the probability of a malfunction of equipment important to safety previously evaluated in the UFSAR be increased?
The revised safety analysis limit high-high steam generator water level setpoint will not adversely affect sysrem perl'onnance or safety system functions assumed in the accident analyses. The original design specifications such as for seismic mJ,uirements, electrical sepuation and e.oviro~ntal qualification are unaffected. In addition1 this change does not result in equipment used in accident mitigation to be exposed to an adverse environment nor does it create an adverse condition for any other safety-related equipment. Component integrity is not challenged. Therefore1 it will not* affect the fai1me modes or failure probability of any equipment important to safety; no new failure modes are being created for any plant equipment. The revised setpOint will not adversely affect the operation of the Reactor Protection System, or any other device required for accident mitigation. Tberef~, probability of a malfunction of equipment important to safety previously evaluated in the UFSAR will not be increased.
se.- cf\/ oo 1- 01 J?Et/ 1 FEB li '94 16:0j FRO'! OPL LICENSING TO 8613933912:34 SECL-94-042
~e7of7
- 5. Will the consequences of a malfunction of equipment importmt of safety previously evaluated in the UFSAR be increased?
The previously identified most limiting single failures are still limiting, and the performance and effectiveness of equipment important to safety is unchanged despite the change to the safety analysis limit high-high steam generator water level setpo.int.
This change does not adversely affect the abili'ty of existing components and systems to mitigate the radiological dose consequences of any accident. The revised high-high level safety analysis limit does not result in response t.o accident scenarios different than that postulated in the UFSAR. The increased capacity does not introduce any new equipment other tllail that previously evaluated in the UFSAR, nor does ii create any new failure modes for e."tlsting safety-relaced equipment. Both the margin to DNB and fuel tempenture limits remain protected. Component integrity is not challenged. Because the licensing basis safety analysis criteria continue to be met, there is no increase in the CODSeQuences of a malfunction of equipment important to safety previously evaluated in the UFSAll
- 6. May the possibility of a xnalfunction of equipment important to safety different than
- any already evaluated in the UFSAR be.created? .
With the revised safety analysis limit high-high steam generator water level setpOint, all original design and perlormance criteria continue to be met, and no new failure modes have been created for any system, component, or piece of eqtripment. No new single failure mechanisms have been introduced nor will the core operate in excess of pertinent design basis operating limits. The li~ing basis safety analysis criteria continue to be met. The steam generator level secpoints and the Reactor Protection System will operate as designed. The change to the high-b:igh level safety analysis limt does not create a new scenario for a malfunction of equipment different from any previously evaluat.ed. Component integrity is not challenged. Since the revised setpoint is not expeciOO to n:sult in more adverse conditions and is not expected to result i.Jfany incicasc in the chalknges to safety systems, there is oo new cirtumstanee or condition created which could result in any malfunction of equipment important to safety different than already evaluated in the UFSAR.
- 7. Will the margin of safety as defined in the bases to any technical specification be reduced?
The accident analysis accepmnce criteria continue to be met ~ the .revised safety analysis limit high-high steam generator water level setpoint. There are no adversely impacted Teclmi.cal Specifications, and safety ma.rgi.n.s are 1lOt' reduced.
Thus, the margin of safety as defined in the bases for the technical sper...ifications will not be changed.
CALCULATION CONTINUATION/ SHEET: i OPS~G REVISION HISTORY SHEET CONT'D ON SHEET:
CALC. No.:SC-CNOOl-01 Attachment 10.6
REFERENCE:
ORIGINATOR, DATE REV: CMM 8/16/941 lIR.
I REVIEWER/VERIFIER, DATE LFP 8/16/94 ATTACHMENT 10.6 WESTINGHOUSE LEITER; JPO FOR OVERPOWER OPERATION
i'lrlK .jt'.J ' ':14 Id' C : ::>Id Ff'l.Jl"l OPL L. i CEr ..6 ING
_* ".SC-Ci.Joo l -cl 1
A~~H ME.r-.JI to* <;:.,
f'd- l oF= 7 PSE-94-555 N11c1ear recnnorcey Division W8'1inibouse Energy Systems Elechic CftPQration Bax 355
?ittSOOr~ Pennsylvania 15230*0355 A)FS'i 94 20 I
- Mt.E.S.Rosenfeld,:Manager Nuclear Fuels March 24, 1994 Public Service Electric & Gas Company ET-NSL-OPL-Il-94-143 P .0. Box 236 MC N20 Hancocks Bridge, NJ 08038 Pub.lie SeIVicc Electric & Gas Company Salem Units 1 and 2
Subject:
JPO for Overpower Operation
Dear Mr. Rosenfeld:
The purpose of this Jetter is to transmit the report providing the justification:of past: opeiation of Salem Unit 2 during Cycles 7 and 8 at power levels up to 104.5% rated thermal power.
The report examines each of the li~ing basis accident analyses and for each event, the impact of the overpower opetation is evaluated and it is coocluded*for*Salem Unit 2 that the safety of the plant was not compromised. For some of the licensing-basis events an engine.ering evaluation was adequate to confirm that no significant safety concern existed.
This was possible, either because the licensing analysis was not affecte.d by the ovetp0wer operation, or that more t.ban sufficient margin exim to offset the adveISe coosequences associated with oveq>0wer operation. For other events a more de.tailed analysis, including computer simulation, was needed, due to a lack of available margin, or the unavailability of sensitivities to assess the impact of overpower operation.
Included with the report are two appendices. Appendix 1 provides the detailed report of the evaluation of the :Reactor Protection System (RPS) and Engineered Safety Feature Actuation Systelll (PSFAS) setpoints. The acceptability of the actual setpoint with respect to the Technical Specifications is identified in this report. Appendix 2 contains pressure and temperature plots and digitized data for PSE&G's use in performing inside containment equipment qualification (EQ) evaluations.
- Sc.-: c:;,.:loo t- o 1 ~f'U ATTAC..~ME.N,... lO."
Pd-- 2 DF 7 Mr. E. S. Rosenfeld PSE-9+555
~h 24, 1994 Page 2 If you have any questions or comments, please contact the undersigned.
Very truly youn, I~~-ft-~
Special Sales Representative Power Systems Field Sales AMS/
cc: . T. K .. Ross. MC ~"21* . . JL,. IA .
R. S. Kent MC N21. lL, IA
. ~~G*~~~:r:.o 1 FROf'l op_ ~:r..E1.s1-1G "T"O 85093:'.9::.234 A. rn::k:..a-\ ME:µ T 10
- G, Pd '3 oF 7 Table or Contents 1.0 Introduction and Summary l 2.0 Evaluations in Support of Past ()ptt-ation 1 2.l Evaluation of RPS/ESFAS Setpoints 2 2.2 Evaluation of Non-LOCA Events 13 2.3 Evaluation of LOCA-R.elated Events 31 2.4 Evaluation of RCS Components and Fluid Systems 33 2.5 Evaluation of Radiological Doses 40 2.6 Evaluation of Nuclear Fuel 2.7 Evaluation of ContaimneDt Integrity Analyses 41 2.8 Evaluation of Out.side C.Ont.ainment Equipment QualificatioD 45 Appendix 1: Det.ennination of Impacts on the Reactor Protection System and the Engineered Safety Feature Aeruation System for Operation of Salem Unit 2 at 104-.5 % Rated Thermal Power Appendix 2: Inside Containn:r.nt Integrity Analysis Figures and Digitized Data JPO. wpf: ll>-0334
,MAR 30 94 09:23 FRCl1 OPL LICENSING TO 86093391234 PAGE.008/047 sc.-c.Joo 1-ol
~r~c~ IY\E µT 10* ~
~ 4 01= 7 Determination of Impacts on the Reactor Protection System and the Engineered Safety Feature Actuation System for Operation of Salem Unit 2 at 104.Si Rated Thermal -Power March 1994 C. F. Ciocca
MAR 30 '94 08:27 FROM OPL LICE~3ING TO 860933'31234 SC.- c."'100 1-0 I
~ --rn:a' !-tr/\ E"-)T JO
- G, P"J '5 o F= 7
~channel uncertainties the safety analyses appear preserved and the Allowable Value was not impacted.
- 12. Loss of Flow, Setpoint (~90% TDF/Loop)1 Allowable Value (589% TOF/ Loop)-
Westinghouse recommends that the Loss of Flow Setpoint be normalized to the precision flow calorimetric determination of 100% RCS flow and the bistable be set at 90% of the precision flow calorimetric. Implementation requires that each cycle would need to be re-scaled. Salem has utilized a previous cycle for determination of the precision flow calorimetric and continued to maintain the bistable setting based on the early scaling calculation. This in effect has preserved the Thermal Design Flow and as flo~ is reduced (due to tube plugging, etc.) the bistable setpoint is approached*.
- This in effect has put the plant closer to a trip than the safety analyses require. The conclusion is that the plant has the trip setpoint higher than 90% of the TDF and has been operating in a conservative manner.
Information received from PSE&G has identified that PSE&G calibration provides a 5.4i instrument span (120' flow) channel uncertainty :for this. function. - This value* corresponds '-to -a '6-.-S% .
ow* uncertainty-.- * -The ,current- TS setpoint- is - 90t *flow* and _the- *-. *.*.
- rrent Safety Analysis Limit is at 87% flow. PSE&G calculations, based on the benchmarck flow calorimetric1 set the bistable at 82087 gpm. This c:orr_esponds. to a setpoint of 94\
flow. This provides a 0.5\ flow mar9in, therefore the channel would have been within the Satety Analysis Limit. Further discussions with PSE&G representatives indicates that the above discussion on holding the bistable at a constant value between cycles is consistent with the plant past practice.
- 13. Steam Generator Water Level Low-Low, Setpoint (~16% NR Span), Allowable Value (~14.8% NR Span)-
Steam Generator Level Low and Low-Low Level for Loss of Normal Feed and Feedline Break setpoints are impacted as the chan9es in pressures and flows have impacts on the PMA biases. 'l'he extent of the impact is determined by calculating the biases for the operating conditions and comparing the results to the values previously calculated by PSE&G. The differences* are then compared to the available margin between the nominal trip .
setpoint and the safety analyses limits. Based on the PSE&G assumptions for 100\ RTP the PMA values chosen by PSE&G were conservative to those calculated by Westinghouse. Therefore the available margin in this function was adequate to protect the Safety Analysis Limit and was in accordance with the TS .
- Page 7
MAR 30 '94 08:27 FRa1 OPL LICENSING TO 86093391234 PAGE.0:.:..8/047 SC - C N~O l - e I /
ATT'Al:.H t"\ E:lJ'r i 0 * ""
r~ ~ ~rz:. 7
- 14. Steam Flow/Feed Flow Mismatch and Low SG Water Level, Setpoint (!:40% of Full Flow@ RTP & ~25% NR Span), J\llowable Value (~42.5% of Full Flow @ RTP & ~24i NR Span)-
Steam Flow Feed Flow Mismatch coincident with the Steam Generator Water Level Low function must be evaluated as Steam Flow Feed Flow Mismatch operability and Steam Generator Water Level Low operability.
For the Steam Flow Feed Flow Mismatch function, Salem normalizes the steam flow transmitters to the daily power calorimetric on a bi-monthly schedule. Given a power level higher than indicated the mismatch function would have normalized the steam input at a value lower than actual. The result is that the mismatch would not have been operating closer to the trip setpoint as the steam flow input would have been indicating 4.St RTP lower than actual with the Feedf low input being higher than indicated by a value corresponding to 4.5% RTP. This function is not explicitly modeled in the safety analyses, however the analyses should be evaluated for operating at higher than indicated steam and feed water flows.
Steam Generator Level. Low. is. impacted as-the changesin*pressures **
and .flows ;. ha.ve . impacts0 :on. the
- PMA: *biases ~
- The* extent *of -the::: -
- act* is determined by calculating the biases for the operating onditions and comparing the results to the values previously calculated by PSE&G. The differences are ..then- compared . to the***
available margin between the nominal trip setpoint and the safety
- analyses limits. The extent of the il!\pact is detennined by calculating the biases for the operating conditions and comparing the results to the values previously calculated by PSE&G. The differences are then compared to the available margin between the nominal trip setpoint and the safety analyses limits. Based on the PSE&G assumptions tor 100\ RTP the PMA values chosen by PSE&G were conservative to those calculated by Westinghouse. Therefore the available marg~n in this function was adequate to protect the Safety Analysis Limit and was in accordance with the TS.
- 15. Undervoltage RCP Volts/bus, Setpoint (~2900 Volts/bus),
Allowable Value (~2850)-
The Undervoltage RCP Volts/bus setpoint is not impacted by operating at higher reactor power levels. Plant electrical conditions are independent of the reactor power levels.
Therefore, there is no impact on this function.
- 16. Underfrequency RCPf Setpoint (~56.5 HZ),
Allowable Value (~56.4 HZ)-
- Page 8
.:i.,;: * ':>4 lclb: ..:,.:::. rril.JM OPL '--; CEl'<S I.'iG TO 86093391234 PRGE.029/047 SC.- ~t-.loeJ L- o J A'"fT'Ac,H MctSr aO .to
~ 7 oF7
- 5. Turbine Trip and Feedwater Isolation A. Steam Generator Water Level High High, Setpoint(~67% of NR Span), Allowable Value(s68% of NR Span)-
The Steam Generator Water Level High High function is impacted due to the changes in pressures and flows have impacts on the PMA biases. The extent of the impact is determined by calculating the biases for the operating conditions and comparing the results to the values previously calculated by PSE&G. The differences are then compared to the available margin between the nominal trip setpoint and the safety analyses limits. Based on the PSE&G assumptions for 100t RTP the PMA values chosen by PSE&G were non-conservative to those calculated by Westinghouse by approximately 1\ NR span. However the available margin in this function was adequate to protect the Safety Analysis Limit and the setpoint was in accordance with the TS.
- 6. Safeguards Equipment Control system, Setpoint (NA),
Allowable Value (NA)-
There is no nominal trip setpoint or allowable value which can be explicitly attributed to this £unction. Therefore there is no impact for this function for an overpower* condition .
- 7. Undervoltage, Vital Bus A. Loss of Voltage; Setpoint (~70\ bus-voltage)*
Allowable Value (~65% voltage}-
The Undervoltage, Vital Bus setpoint is not impacted by operating at higher reactor power levels. Plant electrical conditions are independent of the reactor power levels. Therefore, there is no impact on this function.
B. Sustained Degraded Voltage, Setpoint (~91.6% bus voltage for Sl3 seconds), Allowable Value(~91% voltage for S15 seconds)-
The Sustained Degraded Voltage setpoint is not impacted by operating at higher reactor ~ower levels. Plant electrical conditions are inde~endent of the reactor power levels.
Therefore, there is no impact on this function.
P;\ge 19
FORM NC.DE-AP.ZZ-0010-1 CERTIFICATION FOR DESIGN VERIFICATION Reference No. SC-<:AJoCJ I'- O/
SUMMARY
STATEMENT
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,?q2mef!ll;tf; The undeniped hereby certifie1 that the daip verifi *on for tbe subject doc:umea& bu beea completed, the questions from the pneric checklist have beea reviewed and u ~ and all comment& have been adequately incorporated.
Desip Verifier A.aipecl By Sipature of Desip Verifier I Date Deaip Verifier Allipld By Sipaan of De&ip Verifier I Date Deaip Verifier Amped By Sipamre of Deaip Verifier I Date Paae_/ of.i Nuclear Commoa Paae 2 of 4 111.92
FORM NC.DE-AP.ZZ--0010-1 CERTIFICATION FOR DESIGN VERIFICATION REFERENCE DOCUMENT NO. /REV. _ _ s __c....-__.IJ:;....N_c:::io
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COMMENTS RESOLUI10N
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- Nuclear CollllDCll P11e 4 of 4 1/1/92
FORM NC.DE-AP.ZZ-0010-l CERTIFICATION FOR DESIGN VERIFICATION I
REFERENCE DOCUMENT NO. /REV. _ _..-.5;_..-=-C-__.<J=,/Vt..;.;i; ....x;?~l--....CJ-......I_ _
COl\fMENTS RESOLurION L/. 7/7M-Jsm 1- /.kz. s./n--n *t2-
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FORM NC.DE-AP.ZZ-0010-2
- GENERIC VERIFICATION CHECJa.IST YES REP'ERENCB DOCUMENT NUMBER/REVISION SC'- CA/OO/ - o I NO N/A WHERB P'OUND PAGB NO
- I &'.:!.L COMMENTS (Y/N\
- 1. WEH DESIGN INPUTS CORRECTLY SELECTED AJID ./ f'* 2& A/
INCCRPCllATED INTO DESIGN?
- 2. AU AS. . .TIOlll NECESSAIY TO PHFCM THI DESIGN ACTIVITY ADEQUATELY DESClllED AJID REASCJllAIU'P WMEU
/ d.qo N NECESSMY, AU TH! ASSUFTIOlll IDENTIFIED FOi
-~- -
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SUISEQUEllT Rl*YHIFICATIOll WHEI THI! DETAILED DESIGN ~~~ '-~I ... ;)
ACTIVITIES ARI aJl'l!TED7
- ~
- -v/
- 3. ARE THI APPROPRIATE QUALITY AJID QUALITY ASSl*Ma REQUIREMEITS SPECIFIED?
- I
- 4. ARE THI APPLICAILE allEI, STANDARDS AND REGULATORY ltEQUIREMEITI llllCLLDllG ISSUES AND ADDENDA PROPERLY IDENTIFIED A11D. All THEIR HQUIREMEJITI FOi DESIGN Ml!T'P _/- p f 'rt ;/
- 5. HAYE APttt.ICAILE COlllTllUCTIOll A11D OPEUTlllG EXP£RIEMCI IEEll COllllDERED'P L~ /
/
6
- HAYI T* DHIGI llllTDFACI llEQUllEMEJITI HEii SATISFIED?
- - -/
- 7. WAI Al APPllONIATI DEllGI METHOD USED?
-r- ~
- a. II THE OUTPUT WIOllULI all'AUD TO lllPUTI? L - - p. 6-:;~
tJ
- 9. AU Tiii SPECIFIED PAITI, ECUIPMllT, Alm PIOCllSll SUITAIU FOi TllE IECUlllD APPUCATIClll
- /
- 10. All Tl* SlllCIFIED MTDIALI c:a.ATIU WITI EACI
- L OTHEI Alm T* DEllGI EIYllCIMlllTAL CGllDITIClll TO WlllCI TH! MTDIA!. ~ILL I! !XPOSID!
- - -/
- 11. HAYE ADECUTI MllllllllCI f'IATUlll Alm IECUIUMlllTI BEEi SPICIFIED'P
. c
- 12. ARI ACCHllllLITT -
REPAll'P OTm Diii* PllOVlllClll ADECUATI FQI PEI~ Cl' ..a MllTlllAICI Alm
- - - ~ /
-I
- 13. HAI AOECUTI ACCHllllLITY IHI l'IOVIDID TO PHFOlll THI ll*IDYICI lllPICTICll EXPICTID TO
- llCUlllD UIH T* PLAIT LIFl7
- Pase 1 of2 1/1/92
FORM NC.DE-AP.ZZ-0010-2 GENERIC VERIFICATION CHEClCLIST REFERENCE DOCUMBNT NtJMBER/REVISION (CONTINUED) oC'-(}11/00/ -a1 I kV I YES NO N/A WHEU l'Otnm COMMENTS PAGS NO. tY/Nl
- 14. HAS THE DESIGll PROPERLY CONSIDERED lADIATIOM EXPOSUIE TO THE PWLJ C ANO PLAllT PElSCNNEL7 HAV! AL.AJA / -
v p.12- ;/
- - --/
CONSIDERATIONS IEEM ADORESSED?
-/ -
- 15. Al! THE ACCEPTANCE CllTEllA IMCCRPOIATED IM THE DESIGN DCICl.llENTI SUFFICIENT TO ALLOW VEllFICATIOM THAT p /5#1b (I!
DESIGll RECIUllEMEMTI HAV! IEEM SATISFACTQllLT ACCCMPLJ SHED?
- - ____/'
- 16. HAI VElJFICATJOM OF THE ELECTllC LOAD CONTIOL PROGRAM CDE*TS.ZZ*290l(Q)J IEEN P£lFOIJllD7
- - -7
- 17. HAI THE EFFECT OM THI DIESEL GUUATQI LOAD SEQUENCE ,SMT IEEI WI. YZED?
-/
- 18. HAV! ADEQUATE PlE*OPEUTIOUL AND SUISEQUEMT PEllCDIC TEST lEQUllEJIEMTS IEEM APPlCPllJATELT SPECIFIED?
- 19. All ADEQUATE NAllDLJIHI, ITOIMm, CLIAllllHI AND SHIPPING lECIUllEMEMTI SPICIFIED?
- - d/ I/
- 20. AlE ADEQUATE IDEllTIFICATICI l!ClllllNEllTI SPECIFIED?
- - 7/ y
- - -/
- 21. ARE IECllllEMEMTI FQll IECOID PHPAIATICll lEVIEV ,
APPROVAL, RETEITICll, ETC. ADEQUATELY SPICIFIED?
- Nuclear Deputmmt Pa1e 2of2 1/1192
cmnnaTICJl !at Ilm(Jf VDirIClTICJf m nx. WJ.mN_.;.._ _ _ _ _1-
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O I
DATE Pa;e _l.of S-DE-AP.ZZ-0010 Exhibit 2 Rev.QA Page 1 of 3
CERTIFICATION FOR DESIGN VERIFICATION Reference No.
SUMMA.RY STATEMENT The undersigned hereby certifies that the design verification for the subject package has been completed and all comments have been adequately addressed.
J, ;JK /7<<<< s ///.. l. /c>).VO 1/s/1i-Design Verifier Assigned By ~ignature of Design Verifier/Date Design Verifier Assigned By Signature of Design Verifier/Date Design Verifier Assigned By Signature of Design Verifier/Date Design Verifier Assigned By Signature of Design Verifier/Date DE-AP.zz.,;.0010 Exhibit 2 Paqe '2- of 5 Rev.OA Page 2 of 3
CERTIFICATION FOR DESIGN VERIFICATION CONTINUATION Verified the applicable items as designated on the Generic Verification Check-list.
Verified that the choice of environmental conditions, (normal, accident, MSLB, etc.) was appropriate.
Verified that the values for rack accuracy and drift were assigned properly depending on whether the instruments were installed in a Westinghouse rack and with regard to the calibration procedures.
Verified that normal radiation affects were not duplicated for accident conditions.
Verified that the values used in the Attachments, if any, accurately reflected the values used in, or generated by, the calculation.
Verified that the objectives were adequately addressed, the assumptions were reasonable, and that the conclusions accurately reflected the results of the calculation.
?'-.
Page ii__ of 5 DE-AP.ZZ-0010 Exhibit 2 Rev.OA Page 3 of 3
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- 5. DS APPLIWLE a:tm1ttrr'ICJf IND CFDATI?< USED? .. 7 -$ See.
- 8. IS THE mIQl IXntmn' mscrUtZ aJl'Am> 'lO 1!E
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- 9. W: 'Im: SPttll IED PARTS. ~ IND PRCmm SUITABLE ~ 'Im: ~ APPLIClTICIP.
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- 10. ARE THE S'PELllIED Ml'l'DlALS OJllATIBIZ vrm Dal arm AND m I!ZIC1l ~ CXHlmcm '!O UHICB '1'HE
~TDUAL *"ILL BE mom>?
IO - ~ J#A .:---1--
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- 11. IQVE ~TE MlntmWa IU'nm AND ~
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- 12. AP.I AC~SI!n.m' AND O!m mIQl PRa/ISICJG ~TE rat !'!1...~ ':I Nmm> ~IND REPAIR?
1'1- - J#A 1;---1---
- 13. HT\S ADIQ.llTE ACCESSI!I:.m BID ~ TO PDFCmf 1
'n1E !N-mvICI INS?Et"!'!CJf DPD:Iil> '!O BE R!tl1IJUl) IDJN:i PUtm' IJJ EIIME? IJ_
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~ N PUBLIC AND PL.VII' r!lSCHE.? HAVE;;.>>.>.~
SIDDA'!'I~ Bml ADDPJSSEl>?
- - - I.----1---
DE-AP.ZZ-0010 Exhibit 3 Rev.o Paqe 1 of 2
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- 17. DS m IMPACl' Qf m DIEmi GDDl'lat Ian ~
mJDY BID NQI.Ym>?
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- 18. DVE m.trJATE PRE~QQL AND 18SJX1mn' PllICmC ml' ~ BID APPRCFIUlm.Y SPIXJJW>?
- -Ji/A
- 19. m ~TE !llNCl.JXi. m:uaz, aA'II1li R SillPPlJG m.;.unmmn'S SP!X:ll IW?
- - JJ/A
- 20. W: mJ;1.JATE Imn'll"ICATICll ~ ShLllm>?
- - ll/A
- 21. *ARI:~ FCll m:rD Pm'ARATICJf, IEVI!Jl, APPROYAL, Rrl'Umaf. rrc. ~y SPIXJJm>? -*-$
DE-AP.ZZ-0010 Exhibit l Rev.O Page 2 Of 2
FORM NC.DE-AP.ZZ-0010-1 CERTIFICATION FOR DESIGN VERIFICATION Reference No. X-CA./(!;:10 / -0 I (2.Et/. i r 12.1..
SUMMARY
STATEMENT
,ff. eviFt-
..-rh'C LtJN *:::H::-UO/P/.
The undersigned hereby certifies that the design verification for the subject document has been completed, the questions from the generic checklist have been reviewed and addressed as appropriate, and all comments have been adequately incorporated.
,i. J. lA J f:::.Ca/7 ~I Design Verifier Assigned By Design Verifier Assigned By Signature of Design Verifier I Date Design Verifier Assigned By Signature of Design Verifier I Date Design Verifier Assigned By Signature of Design Verifier I Date Page / of
- - Z..
Nuclear Common Page 2 of 4 111/92
- REFERENCE DOCUMENT NO. /REV.
FORM NC.DE-AP.ZZ-0010-1 CERTIFICATION FOR DESIGN VERIFICATION
-SC- c,,u &70/- '1l I Jr-2 !!
COl\fMENTS RESOLUTION
- P/t?/Yv Acceptance of DATE RESOLVED BY DATE Resolution Page '2..- of
- - Z-Nuclear Common Page 4 of 4 111192
FORM NC.DE-AP.ZZ-0010-2 GENERIC VERIFICATION CHECKLIST REFERENCE DOCUMENT NUMBER/REVISION
~C-CA/001-0 I I i:rd-1 YES NO N/A WHERE FOUND COMMENTS PAGE NO. (Y/N)
- 1. WERE DESIGN INPUTS CORRECTLY SELECTED AND / ~~ ~?~e/6~
INCORPORATED INTO DESIGN? AJ
- -- -- //0
- 2. ARE ASSUMPTIONS NECESSARY TO PERFORM THE DESIGN 6!!h 69 ACTIVITY ADEQUATELY DESCRIBED AND REASONABLE? WHERE NECESSARY, ARE THE ASSUMPTIONS IDENTIFIED FOR -t/ -- -- µ SUBSEQUENT RE-VERIFICATION WHEN THE DETAILED DESIGN ACTIVITIES ARE COMPLETED?
- - -/
- 3. ARE THE APPROPRIATE QUALITY AND QUALITY ASSURANCE REQUIREMENTS SPECIFIED?
/
- 4. ARE THE APPLICABLE COOES, STANDARDS AND REGULATORY REQUIREMENTS INCLUDING ISSUES AND ADDENDA PROPERLY IDENTIFIED AND ARE THEIR REQUIREMENTS FOR DESIGN MET? - -- -- v
- 5. HAVE APPLICABLE CONSTRUCTION AND OPERATING EXPERIENCE BEEN CONSIDERED?
t/
- 6. HAVE THE DESIGN INTERFACE REQUIREMENTS BEEN .
SATISFIED?
A
-- --v :
- 7. WAS AN APPROPRIATE DESIGN METHOD USED? / JI p'
- 8. IS THE OUTPUT REASONABLE COMPARED TO INPUTS? - ./ -- --
/00/.Aff-1"* 1 µ
- 9. ARE THE SPECIFIED PARTS, EQUIPMENT, AND PROCESSES SUITABLE FOR THE REQUIRED APPLICATION?
- -- --/
- 10. ARE THE SPECIFIED MATERIALS COMPATIBLE WITH EACH OTHER AND THE DESIGN ENVIRONMENTAL CONDITIONS TO WHICH THE MATERIAL WILL BE EXPOSED?
- -- /
- 11. HAVE ADEQUATE MAINTENANCE FEATURES AND REQUIREMENTS BEEN SPECIFIED?
z/
- 12. ARE ACCESSIBILITY AND OTHER DESIGN PROVISIONS ADEQUATE FOR PERFORMANCE OF NEEDED MAINTENANCE AND REPAIR?
/
- 13. HAS ADEQUATE ACCESSIBILITY BEEN PROVIDED TO PERFORM THE IN-SERVICE INSPECTION EXPECTED TO BE REQUIRED DURING THE PLANT LIFE?
/
A Nuclear Department Page 1 of2 111192
FORM NC.DE-AP.ZZ-0010-2 GENERIC VERIFICATION CHECKLIST REFERENCE DOCUMENT NUMBER/REVISION (CONTINUED) sc-c-A.Joo/ -o t.. I .:1712..;t.
YES NO N/A WHERE FOUND COMMENTS PAGE NO. {Y/Nl
- 14. HAS THE DESIGN PROPERLY CONSIDERED RADIATION EXPOSURE TO THE PUBLIC AND PLANT PERSONNEL? HAVE ALARA /
CONSIDERATIONS BEEN ADDRESSED? -- -- --
- 15. ARE THE ACCEPTANCE CRITERIA INCORPORATED IN TH.E DESIGN DOCUMENTS SUFFICIENT TO ALLOIJ VERIFICATION THAT //CJ ?
DESIGN REQUIREMENTS HAVE BEEN SATISFACTORILY ACCOMPLISHED?
-- / -- --
- 16. HAS VERIFICATION OF THE ELECTRIC LOAD CONTROL PROGRAM [OE*TS.ZZ-2908(Q)] BEEN PERFORMED?
/
- 17. HAS THE EFFECT ON THE DIESEL GENERATOR LOAD SEQUENCE STUDY BEEN ANALYZED?
v
- 18. HAVE ADEQUATE PRE-OPERATIONAL AND SUBSEQUENT PERIODIC TEST REQUIREMENTS BEEN APPROPRIATELY SPECIFIED?
/
- 19. ARE ADEQUATE HANDLING, STORAGE, CLEANING AND ,/
SHIPPING REQUIREMENTS SPECIFIED? -- -- --
- 20. ARE ADEQUATE IDENTIFICATION REQUIREMENTS SPECIFIED?
-- -- /
- 21. ARE REQUIREMENTS FOR RECORD PREPARATION REVIEW, APPROVAL, RETENTION, ETC. ADEQUATELY SPECIFIED?
/
- Nuclear Department Page
.i 2 of2 111192