BVY 03-115, Technical Specification Proposed Change No. 257 Implementation of Arts/Smella at VYNPS Supplemental Information and Proposed Allowable Value, Attachment 2, Calculation VYC-0693A
| ML033600061 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 12/11/2003 |
| From: | Thayer J Entergy Nuclear Northeast, Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BVY 03-115 VYC-0693A | |
| Download: ML033600061 (96) | |
Text
Docket No. 50-271 BVY 03-115 Vermont Yankee Nuclear Power Station Technical Specification Proposed Change No. 257 Calculation VYC-0693A - APRM Neutron Monitoring Trip Loops
VY CALCULATION TITLE PAGE VX'C-0693A 2
NIA N/A VY Calculation Number Rcvision Number VXendor Caleulation Number Revision Numiaber
Title:
APRM Neutron Monitorinq Trip Loops QA Statwts:
ER SG O NNS O OQA Operiting Cycle Nunmber*
N/A I The Operating Cycie Number should only he entered here If the results of the calcualation onlv apply during a spccitic operating cycle otherwise enter "NMA".
Cnalculation Supports A Design Change/Specification? (3 Ves El No 4 2003-029
'yt)cN v
1yn MIamspec Nte.
Implementation Required? IZ'Ves 4gC CAlculation DUne ns a Study Only? 0 Y es 0 No Sarety Evaluation Number:
N/A_
Superseded Cilculation Number, lille andt Revision:_ VYC-693A Rev. 1 APRM Neutron Monitoring Trip Loops For Revisions: List CCNs, I[is. or SAs incorporated/supersedted by this revision:
Computer Code(s):
NIA Are there open tiems in this enleulation/revlsion?
7 Yes 0 No Review and Approvel: (Print and Sig Name)
Preparer: Jerry Vns s
Date:
II/ I/2003 Interdisciplinc Reviewer(s%:t See Attafilment 7.7 Date:_
Independent Revievers(v Kirk Meljon Date:
11/19/2003 Appruvcd: -7Z VIFe7g
/7 ki
- L/./d3 Accepted (only Tor AP 0017 enceflations plrformed by s'cndors)
N/A Dnte:
Fimnl Turnover to DCC (Section 2):
- 1)
All upen items. if nny. have been closed.
- 2)
Implementation Confirmotion (Section 2.3.4)
Q Calculation accurately reflects existing plnnt cv'ntigritioor.
(confirmation method indicated below)
O Walkdown a As-Built input review X cusslo OR 3 N/A, calculation does not reflect existing plpant onfi
!fia
- 3)
Resolution of documents identified In the Design Output acumen tinn urVYAPf 0017.07 las hren initiated as required (Section 2.3.6, 23.7)
Printed Name Signature
_DAte_
Page I or_
Pages*
for ctlculaitions peralcrmtxl using AP 0017 thIs schc numbcroulprages in thl bodv nrtiliu fl Luliitiuil.
vYAPIF 0017.01 lFor vendor ciaculations. this is the number orpagesi ol Al' 0017 formr' udded.
AP t(l 17 Rev. X (Title page, rcvicw forms. dcaut shects. 50.S9. tcl.)
Past I oil I.l'C 4/2
VY CALCULATION DATABASE INPUT FORM Place this form in the calculation package immediately following the Title page or CCN form.
_____VYC-693A 2
N/A N/A VY Calculation/CCN Number Revision Number Vendor Calculation Number Revision Number Vendor Name:
PO Number Originating Department:
Critical References Impacted: Cl UFSAR Q DBD [ Reload. "Check" the appropriate box if any critical document is identified in the tables below.
EMPAC Asset/Equipment ID Number(s):
EMPAC Asset/System ID Number(s):
Keywords:
For Revision/CCN only Are deletions to General References, Design Input Documents or Design Output Documents required? El Yest Z No Design Input Documents and General References - The following documents provide design input or supporting information to this calculation. (Refer to Appendix A, sections 3.2.7 and section 4)
Significant
+'"'
Critical Reference REV Difference Affected Reference ft
- DOC #
X
- Document Title (including Date, if applicable)
Review tt Program
(/)
6.15 GE-NE-0000-0012-May 2003 Project Task Report "Entergy Nuclear Operations Incorporated Vermont N/A N/A 0531-01-01 Yankee Nuclear Power Station ARTS/MELLLA" Task T0506: NSSS TS Instrument Setpoints.
6.16 GE-NE-0000-0016-Qj August 2003 Project Task Report "Entergy Nuclear Operations Incorporated N/A N/A 5688-01 Vermont Yankee Nuclear Power Station Extended Power Uprate" Task T0506:
NSSS TS Instrument Setpoints.
6.27 VYC-0690 2
Recirculation Flow Loop Uncertainties to APRM 6.29 NEDC-33089P "Vermont Yankee Nuclear Power Station APRM/RBM/Technical Specifications Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)," March 2003 6.30 Nl DC-33090P 0
"Safety Analysis Report for Vermont Yankee Nuclear Power Station Constant Pressure Power Uprate," September 2003 626 OP-2429 14 Recirculation Flow System Baseline Data Collection and Instrument Calibration.
6.17 GE Performance I
"Flow Control Trip Reference (FCTR) Card".
Specification 25A5903 6.31 BVY 03-23 0
March 20,2003, Technical Specification Proposed Change No. 257 Implementation of ARTS/MELLLA 632___ __
0S1 3
c iPat Wx.
6.32 BVY 03.80 0
September 10, 2003, Technical Specif ication Proposed Change No. 263 Extended Power Uptate.
i:.:.:
t: ".:
eA69 z OF V' T
1tr
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 Design Output Documents - This calculation provides output to the following documents. (Refer to Appendix A, section 5)
+* ****
tttCritical Reference Affected Reference 0
- DOC #
REV #
Document Title (including Date, if applicable)
Program (V) 6.25 OP 4308 21 APRM Monitor Calibration.
=~~~
I=
I
.~~~
- Reference # -
- Doc # -
- Document Title -
- Affected Program -
tt ttt t11*
Assigned by preparer to identify the reference in the body of the calculation.
Identifying number on the document, if any (e.g., 5920-0264, G191172, VYC-1286)
List the specific documentation in this column. "See attached list" is not acceptable. Design Input/Output Documents should identify the specific design input document used in the calculation or the specific document affected by the calculation and not simply reference the document (e.g., VYDC, MM) that the calculation was written to support. If a DBD is used as a general reference, include the most current interim change number after the title.
List the affected program or the program that reference is related to or part of.
If "yes," attach a copy of "VY Calculation Data" marked-up to reflect deletion (See Section 3.1.8 for Revision and 5.2.3.18 for CCNs).
If the listed input is a calculation listed in the calculation database that is not a calculation of record (see definition), place a check mark in this space to indicate completion of the required significant difference review. (see Appendix A, section 4.1.4.4.3). Otherwise, enter
'NIA."
If the reference is UFSAR, DBD or Reload (IASD or OPL), check Critical Reference column and check UFSAR, DBD or Reload, as appropriate, on this form (above).
i Vennont Yankee Design Engineering Page 3 of 29 A\\.,,.
TABLE OF CONTENTS
- 1.
PURPOSE..................
I..........................................6 1.1 CALCULATION OBJECTIVES.............................................................
6 1.2 SYSTEM AND COMPONENTS.............................................................
7 1.3 INSTRUMENT LOOP FUNCTION.............................................................
7 2
METHODS AND ASSUMPTIONS.............................................................
8 2.1 GOVERNING PROCEDURES AND PROGRAMS.............................................................
9 2.2 CRITERIA.............................................................
9 2.3 ASSUMPTIONS............................................................
12 3
INPUTDATA............................................................
13 3.1 PROCESS, Loop DATA AND ANALYTICAL Lm s...............................................................
1 3 3.2 ENVIRONMENTAL CONDITIONS............................................................
14 3.3 PRIMARY ELEMENTS ND-2-1-104 DATA.............................................................
14 3.4 LPRM DATA.............................................................
15 3.5 APRM A, B, C, D, E, F, DATA............................................................
15 3.6 FLOW BIAS ERROR DATA............................................................
15 4
CALCULATION DETAIL............................................................
16 5
RESULTS AND CONCLUSIONS............................................................
16 5.1 ALLOWABLE VALUE............................................................
16 5.2 SETPOINTEVALUATION............................................................
17 5.3 CALIBRATION AND TESTRESULTS............................................................
18 5.4 CALCULATION REVIEW AND IMPACT CONSIDERATION............................................................
20 5.5
SUMMARY
OF REQUIREMENTS............................................................
21 5.6
SUMMARY
OFRECOMMENDATIONS........................................................................................................................21 6
REFERENCES............................................................
28 7
ATTACHMENTS......................................................................................................................................................29 7.1 LOOP DIAGRAM[ I PG]............................................................
29 7.2 CALCULATION DETAIL (B 1 THROUGH B4) [8 PGS]............................................................
29 7.3 DESIGN SPECICATION 22A1366, REV. 3, "NEUTRON MONITORING SYSTEM" [2 PGS]............................................... 29 7.4 DESIGN SPECIFICATION -DATA SHEET22AI366AF,REV. 1, "NEUTRONMONITORING SYSTEM"[1 PG)..................... 29 7.5 DESIGN AND PERFORMANCE SPECIFICATION 175A1 259, REV. 1, "APRM" [3 PGs]..................................................... 29 7.6 GE PERFORMANCE SPECIFICATION 25A5903 REV. l, "FLow CONTROL TRIP REFERENCE (FCTR) CARD [ 23 Po]...... 29 7.7 AP-00 17 FORMS AND INTERDEPARTMENTAL REVIEW FORM [6 PGS]........................................................................... 29 Vermont Yankee Design Engineering Page 4 of 29
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 Approval Rev. No.
Date Reason & Description of Change 0
Initial issue, Reflects new methodology described in Instrument Uncertainty and Setpoints Design Guide, Rev.
0 This sub-calculation (VYC-693A) only addresses the associated trips, with the analog output uncertainties remaining in the parent calculation (VYC-693).
1 Revised to allow the use of N=44 in the RBM flow bias equation (Setpoint < 0.66(W-AW) + N), and to reflect the use of M&TE to set the Flow Bias value in testing the RBM rod blocks. References to any specific operating cycle or COLR report were removed since this calculation is valid for any cycle where the value of N is between 42 and 44.
2 Incorporated VYC-693A CCNs 1, 2, 3 and 4. CCN 2 attachment M and table revisions had no impact on this revision of the calculation since the CCN was based on the flow bias equations used prior to ARTS/MELLLA or EPU.
This revision includes the Analytical Limits and calculates Allowable Values (ITS) Trip Setpoints and As-Found Tolerances (CTS) to support ARTS/MELLLA and EPU.
The APRM Rod Blocks have been removed from the Technical Specifications. The Rod Block Monitor has been maintained in Technical Specifications; However, the Rod Block Monitor will be treated as an indicated value (without instrument error) for settings.
- 4.
4
- 4.
I Vermont Yankee Design Engineering Page 5 of 29
..
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2
- 1.
PURPOSE 1.1 Calculation Objectives This calculation has been developed in support of the Vermont Yankee Setpoints program and covers the APRM/LPRM (Average Power Range Monitor/Local Power Range Monitor) neutron monitoring loops. This calculation has the following major objectives:
- 1)
Document the instrument loop functions and the basis for the setpoints and operator decision points associated with those functions.
- 2)
Establish the total loop uncertainty for each increasing' Setpoint and verify consistency with the design basis
- 3)
Calculate the limiting setpoints and operator decision points.
- 4)
Evaluate the adequacy of existing Setpoint Administrative Limits and procedural decision points.
- 5)
Provide As-Left and As-Found tolerances for use in instrument calibration and functional test procedures. Verify and document process corrections, instrument scaling, and calibration methods. The errors determined in this calculation are based on the vendor defined operating characteristics.
- 6)
This calculation does not include evaluations of the analog indicators or recorder functions. The accuracy of the trip functions determined by this calculation will be used as input for generic evaluations for alarm response, operating procedures, off normal operating procedures and EOP impact.
Low (decreasing) setpoints are not reviewed. They are indicative of a gross failure. Therefore, it is not necessary to determine uncertainty.
Vermont Yankee Design Engineering Page 6 of 29 S
~~~~~~~~~~~~~~~~
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 1.2 System and Components This calculation applies to the Power Range Monitoring Instrumentation of the Neutron Monitoring System. The specific components to be addressed are:
Table 1: System Components REF.
TAG lACK l SYS DESCRIPTION lIFG.
MODEL NO.
CWD NUMBER CABINT 6.18 ND-2-1-104 In-Core NM LPRM Detector GE N/A 674,675 (80 items) 6.18 LPRM 9-14 NM Local Power Range GE 135B9824G2 676,676A, Monitor 677,677A 6.18 APRM A, C, 9-14 NM Average Power GE 920D453GI
- 678, 679, E, B, D, F Range Monitor 920D453G2
- 680, 688, 692, 693 1.3 Instrument Loop Function Attachment A has a simplified loop diagram of the instruments and components described below.
1.3.1 Normal Operations During normal operation the APRMs provide the control room operators with indications of the average reactor power from about two percent to 125 percent via recorders on the operators console. This analog information is also provided to the plant computer. In addition, the APRMs are capable of generating trips when various conditions are exceeded. These trips are:
- Rod Block on APRM Upscale
- Rod Block on APRM Downscale
During normal operation the RBMs provide the control room operators with indications of the local average power immediately surrounding a control rod that has been selected for withdrawal. The RBM promotes controlled rod withdrawal by issuing rod withdrawal block signals if the reactor operator does not respond correctly to prompts requiring operator action. The RBM issues rod withdrawal block signals on:
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 The Rod Blocks associated with APRM and the Rod Block Monitor are not credited for plant protection. The APRM Rod Blocks have been removed from Technical Specifications (and placed in the VY Technical Requirements Manual The Rod Block Monitor is maintained in Technical Specifications; however, the setting for the RBM is treated as an indicated values (without instrument error applied to the setting). Therefore, only the APRM trips will be evaluated in this calculation.
The purpose of the APRM flow biased rod block function is to avoid a condition that would require Reactor Protection System action if allowed to proceed. The APRM flow biased rod block setting is selected to initiate a rod block before the APRM high neutron flux scram setting is reached. The APRM flow biased rod block setpoint value listed in the TRM is the maximum nominal trip setpoint allowable. Calibration tolerances have been established for this setpoint that will render the setpoint acceptable above or below the nominal value. The uncertainty associated with the difference between the APRM flow biased scram and rod block functions is limited to the uncertainty associated with the trip circuitry.
Common equipment (such as detectors, LPRMs, flow input and averaging circuits) will affect both functions equally.
VY has implemented long-term thermal hydraulic stability solution Option I-D.
Option I-D is only applicable to plants that can demonstrate that core wide mode instability is the predominant mode and regional mode instability is not expected.
Solution application includes demonstrating that the APRM High Flux (Flow Bias) scram line, considered an analytical limit, provides adequate Safety Limit MCPR protection.
1.3.2 Functions During an Accident The APRMs are assumed to provide the scram initiation signal (120 % power) for the mitigation of the Control Rod Drop Accident, which, according to Reference 6.4, Section 14.6.2, is only of concern when the reactor is operating at less than the RWM Low Power Setpoint (LPSP).
1.3.3 Post-Accident or EOP Functions The APRMs are not required for Post Accident Functions. The APRMs applicability to the EOPs will be addressed outside of this calculation.
2 METHODS AND ASSUMPTIONS This Calculation has been prepared in accordance with the Governing Procedures and Programs listed in step 2.1. Standard methods employed in this calculation are explained in the "Vermont Yankee Instrument Uncertainty and Setpoints Design Guide" [Ref.6.1]. This calculation is performed using the Class 1 graded approach since one of the functions (reactor scram) performed by the APRM loops is classified as Class 1, Nuclear Safety Related.
Vermont Yankee Design Engineering Page 8 of 29
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2-2.1 Governing Procedures and Programs 2.1.1 VermontYankecInstrumentUncertaintyandSetpointsDesignGuide.(Ref6.1) 2.1.2 VennontYank-eeEngineeringProcedureAP-0017,CalculationsandAnalyses (Ref. 6.9).
2.2 Criteria 2.2.1 Analytical Limits and Technical Specification Limits Ile Analytical Limits (AL) for the APRM Scram Flow Bias and Fixed trip are established per the Rod Drop Accident and Core Stability Analysis, and are defined in Ref. 6.15 for ARTS/MELLLA conditions and 6.16 for Extended Power Uprate conditions. This AL value will be used to establish Limiting Setpoints (LSp) and an Allowable Value (AV) for use in the Technical Specifications (TS) as discussed below.
This calculation will use ISA-S-67.04.02 Method I for combination of errors to determine a Trip Setpoint and an Allowable Value for Technical Specification values associated with APRM Flow Bias and Scram setpoints. The Limiting Trip Setpoint and the Allowable Value for a process increasing to a limit will be calculated as follows:
AV AL-effective uncertainty for all devices and errors not confirmed during surveillance testing.
LSp AV - effective uncertainty for testing conditions, including devices tested only.
LSp AV - (CT, + CT2 + CT,,) Confirmation of margin between LSp and AV.
All terms are as currently defined in the VY Setpoint Design Guide.
2.2.2 Numerical combinations for the calculations of instrument error, calibration error, loop error, effective error and other associated values have been calculated using Microsoft Excel. Representative calculations in Attachment(s) B were manually verified using a hand calculator, Microsoft Excel stores numbers with at least 15 digits of accuracy, all calculation outputs displayed within this calculation are rounded from the values stored by Excel. Rounding errors induced by Excel are assumed to be negligible within Us calculation.
2.2.3 No errors were found in the manual verification of the calculations performed with the Software in Attachment B. Physical evidence of the review is provided by check marks or other indications'next to each verified calculation. Where multiple calculations are generated by copying cells or formulas selected samples have been verified.
2.2.4 Technical Specification Table 4.1.2 requires the performance of a Heat Balance, for calibration of the APRM output signal, once every 7 days. The calibration of the APRMs to CoreThermal. Power (CTP) is performed under OP-4400. As a part of this procedure, the individual APRM gains are adjusted such that the individual APRMs read conservatively to the adjusted % rated CT? (+2, -00/o).
Vermont Yankee Design Engineering Page 9 of 29
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 The adjustments made to the APRM gains automatically compensate (normalize) the LPRM detector/amplifier output signal. Therefore, since the heat balance is perfomed every 7 days, the amount of LPRM drift that needs to be considered is only 8 days (7 days + 25% extension). The current LPRM drift value is valid for 700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br />. However, since there is insufficient information to estimate the reduction in this drift value for 8 days, the 700 hour0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> drift value will be used in the calculation.
2.2.5 Primary Element Accuracy (PEA)
APRM Channel PEA is a combination of sensor sensitivity and sensor non-linearity uncertainties.
The sensitivity of the detectors decreases with neutron fluence. From Reference 6.13, Section 4.5, the average sensitivity loss, and its 2-sigma variation, for all GE LPRM detectors has been determined to be:
Sensor Sensitivity Loss = -0.33%
(bias term)
+0.20%
(random term)
The detector non-linearity and its 2 sigma variation (in the power range) has been determined to be:
Sensor Non-linearity = -0.49%
(bias term)
+/-1.0%
(random term)
The first part of these detector errors represent bias type errors which apply to all detectors, whereas the second part are random errors that represent variability amongst sensors. Assuming a worst case scenario where the APRM has the minimum number of operational detectors, the PEA, which on a percent of power basis is simply obtained by adding the bias terms and taking the SRSS of the random terms, is calculated below. In the calculation, the random error is reduced by the square root of the minimum number of operable LPRMs to one APRM channel which is 9 per Reference 6.7, Table 3.1.1, Note 5.
PEA =(0.33+0.49)+/-in)/(0.202 +12))
PEA = -0.82+/- 0.34%RP 2.2.6 Process Measurement Accuracy (PMA)
APRM PMA is a combination of APRM tracking error and the uncertainty due to neutron noise. From Reference 6.4, Section 14.5.1.3.1, the most severe event for which the APRMs are assumed to provide the scram signal is the Closure of All Main Steam Line Isolation Valves with failure of the valve position scram. Reference 6.13, Section 4.5 states that for the MSIV closure transient event, the APRM tracking error is 1.1 1% and the uncertainty due to neutron noise is typically 2.0%.
The tracking error is the uncertainty of the maximum deviation of APRM readings with LPRM failures or bypasses during a power transient. The neutron noise is the global neutron flux noise in the reactor core with a typical dominant Vermont Yankee Design Engineering Page 10 of 29
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 frequency of approximately 0.3 to 0.5 Hertz and typical maximum peak-to-peak amplitude of approximately 5 to 10 percent.
PMAS = k(2.o 2 + 1.112 ) = 2.29%RP 2.2.7 The Flow Control Trip Reference Card (FCTR) for the APRM fixed and flow biased scrams is being replaced. The replacement card is designed to operate with a 1% maximum error signal and 0.1% timing accuracy (Ref. 6.17). The 1% error signal includes both device accuracy and drift for a 36-month interval. General Electric has provided additional information (Attachment 7.7) indicating that the reference accuracy (or setting tolerance) for the FCTR is 0.5% of calibrated span.
Control Room (CR) Indication (FI-2-159A, B and FR-2-154) is scaled for a flow rate of 041,800 GPM. Rated Recirculation Drive Flow (100%) is defined as the required Drive Flow to achieve 48 MPPH (Reference TS 2.1.A.l.a). Such things as core design, changes in core and piping resistance, jet pump fouling, etc, affect the relationship between Core Flow and Drive Flow. Therefore, since the rated recirculation drive flow value could change, the percentage of rated flow for CR flow indication will vary. OP 2429 normalizes the total recirculation drive flow (output of summers FSUM-2-l lOA&B) such that the input to the APRM system is 125% of rated flow (as determined by OP 2429). This is consistent with functional testing of the APRM Flow Bias Setpoints and flow indication obtained from the APRM system (chosen via selector switch). Because of this normalization the spans between the drive flow input (VYC-690 errors) and the core flow input are assumed to be equal.
2.2.8 Technical Specification Table 4.1.2 requires the performance of a heat balance, for calibration of the APRM output signal, once every 7 days. The calibration of the APRMs to Core Thermal Power (CTP) is performed under OP 4400. The procedure begins by the calculation of CTP by performing an energy balance on the Nuclear Boiler System. Next, % rated CTP is determined (CTP/Rated CTP MWth
- 100). Then % rated CTP is divided by the applicable scale factor to find the adjusted % rated CTP. Finally, the individual APRM gains are adjusted such that the individual APRMs read conservatively to the adjusted % rated CTP (+2, -
0%). The adjustments made to the APRM gains automatically compensate (normalize) the LPRM detector/amplifier output signal. Therefore, since the heat balance is performed every 7 days, the amount of LPRM drift that needs to be considered is only 8 days (7 days + 25% extension). The uncertainty associated with the heat balance (2% of Reactor Power) is included in the transient analysis (REF. 6.19). Therefore, this uncertainty is not included in this uncertainty calculation.
2.2.9 Calibration tolerance for reactor flow biased power settings are + 0.5% CS or (0.625% RP). The tolerances for trips are +0.4% CS (0.5% RP) per Ref. 6.25 and 6.23.
Vermont Yankee Design Engineering Page 11 of 29
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 2.3 Assumptions 2.3.1 Calibration of instruments is assumed to be at a temperature within the ranges shown in the following table.
Table 2: Plant Area(s) l Plant Area I Minimum I Maximum Control Room 60 0F 180 OF 2.3.2 The temperature variation within a cabinet is the same as the variation of the room in which it is located. The temperature difference between the room and the cabinet is therefore constant. Calibration data are collected with the equipment at the operating temperature of the cabinet.
2.3.3 The vendor does not state a separate value for Humidity Effect (HE). Therefore, it is considered to be included in the accuracy and drift terms.
2.3.4 OP4305 does not give a tolerance for setting the value of the flow signal, because the flow signal is adjusted as close as possible using M&TE. This calculation assumes that only M&TE with a total accuracy of better than 0.25% of reading will be used to adjust the flow signal. Therefore, this uncertainty is assumed to be 0.25 %. This 0.25 % will be combined with the random portion of the VYC-690 flow input. This signal error will then be converted to %RP before being combined for each of the different power to flow lines. However, since the M&TE error affects the overall accuracy and the acceptance criteria for the flow bias setpoints, M&TE with accuracy of 0.05% of span will be used for setting the flow signal value.
2.3.5 All of the Power Range Neutron Monitoring System electronic equipment affected by this calculation are located in areas considered to be mild environment (control room). The LPRM detectors are the only components exposed to high radiation and they were designed for this purpose. Therefore, there are no Radiation Effects (RE) applicable to this calculation.
2.3.6 The APRM scrams are not assumed to operate for safe shutdown or for other seismic events. For operational basis earthquakes, this calculation assumes that the APRMs, LPRMs and trip functions will be recalibrated prior to continued operation. Therefore, seismic error for Neutron Monitoring is not considered in this calculation.
2.3.7 Dead Band (DB) or Readability Uncertainty (RD) are only applicable to that portion of a calculation involving indicators and recorders. Therefore, they do not apply to this calculation.
Vermont Yankee Design Engineering Page 12 of 29
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 2.3.8 Temperature Effect (TE) is not provided as a separate term by the vendor so it is assumed to be included in one of the given accuracy or drift terms.
2.3.9 The Barometric Pressure Effect (PB) is not applicable to this calculation (Section 3.6.9 of Ref. 6.1).
2.3.10 None of the Process Static Pressure Effects (SP) are applicable to this calculation (Section 3.6.13 of Ref. 6.1).
2.3.11 Power Supply Voltage Effect (VE) is considered to be included in the accuracy and drift terms since the vendor does not give a separate value for VE.
2.3.12 VYC-1 758 (Reference 6.20) shows DNM's available to support the APRM calibration with total device error of better than +-0.05% CS (1 0 VDC Range).
Therefore, the SRSS of 2 DMM's required for the calibration per Rcf. 6.25 result in an M&TE uncertainty of +/-0.07 1 % CS.
2.3.13 The neutron monitoring system is located in the control room with the exception of the sensing devices, which are located in the reactor. Therefore, unless otherwise stated, this calculation -will use the set of specifications related to the Control Room envirom-nent in Ref 6.1.
2.3.14 Technical Specification Table 3.1.1 defines the Limit for APRM High Flux Reduced as equal to 15 % RP. This value has been assumed to be the Analytical Limit for this calculation.
3 WPUT DATA Data used to calculate loop uncertainties, process corrections, setpoints, and decision points are tabulated below with the applicable reference or basis 3.1 Process, Loop Data and Analytical Limits Presented below are the input values required to calculate the Limiting Setpoints (Analytical Limits and those parameters such as calibration frequency that are common to all loop components).
Vermont Yankee Design Engineering Page 13 of 29
1 I -
I1.
-.-~,,-::-. ? :':.:..~. Z APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 Table 3: Process Data and Analytical Limits Basis Description Data Ref. 6.3 Process Span (PS) 0 to 125 % RP Analytical Limits ARTS/MELLLA APRM High Flux Scram (Two Loop Ops)
Reference 6.15 Core Flow 0 to < 31.1 %
< 0.4Wd+64.4%
Core Flow 31.1 to < 54.0 %
< 1.28Wd+37.0%
Core Flow 54.0 to < 75 %
< 0.66Wd+70.5%
Core Flow> 75%
Maximum of 120%
APRM High Flux Scram (Single Loop Ops)
Core Flow 0 to < 39.1 %
< 0.4Wd+61.2%
Core Flow 39.1 to < 61.9 %
< 1.28Wd+26.8%
Core Flow 61.9 to < 83.0 %
< 0.66Wd+65.2%
Core Flow >83%
Maximum of 120%
Extended Power APRM High Flux Scram (Two Loop Ops)
Uprate (EPU)
Core Flow 0 to < 30.9 %
< 0.33Wd+53.7%
Reference 6.16.
Core Flow 30.9 to < 66.7 %
< 1.07Wd+30.8%
Core Flow 66.7 to < 99 %
< 0.55Wd+65.5%
Core Flow> 99%
Maximum of 120%
APRM High Flux Scram (Single Loop Ops)
Core Flow0to<39.1 %
<0.33Wd+51.1%
Core Flow 39.1 to < 61.7 %
< 1.07Wd+22.2%
Core Flow 61.7 to < 119.4 %
< 0.55Wd+54.3%
Core Flow >119.4 Maximum of 120%
Assumption 2.3.13 APRM High Flux Scram (Reduced)
<15 % Power 3.2 Environmental Conditions The following table identifies the limiting environmental conditions expected for each loop instrument.
Table 4: Environmental Input Data 3.3 Basis l Description lData Ref. 6.1, Normal Drywell Temperature (Below 270 fi) 160 0F Table 2 Normal Reactor Building Temperature (Occupied Area) 106 0F Ref. 6.5 Normal Radiation N/A (Assumption 2.2.6)
Ref 6.6 Accident Radiation N/A (Assumption 2.2.6)
Primary Elements ND-2-1-104 Data Table 5: Primary Element Input Data Basis l Description Data Ref. 6.2 Maximum Temperature 600 0F Ref. 6.3 Nominal Operating Neutron Flux 1.2x1012 to 2.8x1014 nv Maximum Operating Gamma Flux 1.2xlO9R/hr Accuracy (PLEANSeTx Vermont Yankee Design Engineering Page 14 of 29
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 3.4 LPRM Data 3.5 Table 6: LPRM Ilnput Data Basis Description Data Ref. 6.2 Accuracy
+/-0.8% CS Drift
+/-0.8% CSI700 hrs APRM A, B, C, D, E, F, Data Table 7: APRM Input Data Basis j Description Data Ref. 6.11 Averaging Circuitry Accuracy
+/-0.80% CS Drift
+/-0.5% CS/2 weeks Ref 633 Trip Circuits (Reduced Scram)
Analyzed Drift
+/-0.5% CS/ 3 months (13 weeks)
Ref. 6.11 Trip Circuits (FlowBiased)
+/-1.0% CS (0 to 100% flow) 25A5903 Rev. I Accuracy and Drift valid for 36 months. Timing error not used.
(Assumption 2.2.19)
Ref 6.7 Calibration Interval:
Table 4.1.2 & 4.1.1 APRM High Flux Scram Trips 3 months (13 weeks)
LPRM 2000MWDIT Reactor Heat Balance Weekly 3.6 Flow Bias Error Data The following information is copied or extrapolated from VYC-690 Rev. 2. The accuracy associated with the cardinal flow rates for ARTS/MELLLA and EPU are used in the spreadsheet as an additional error associated with calculation of the APRM setpoints. Errors in VYC-0690 (as listed below) have been calculated based on an assumed maximum total recirculation span of 83,600 gpm (41,800 gpm per loop) to ensure maximum accuracy for recirculation flow indication in the control room.
However, the calculation of flow bias for the APRMs is based on an ideal flow of 65,000 gpm (32,500gpm per loop). Therefore to compensate for the difference in reference values for the errors associated with the flow bias a multiplier of 1.286 is used. As discussed in Section 2.2.7, the input into the FCTR is then scaled, during the core flow to recirculation flow verification, to 125% of recirculation flow. This multiplication has been performed in the spreadsheets that perform the detailed calculations for this document. The error is also multiplied by the specific flow correction value (i.e. if the flow formula is 0.33Wd+53.7 EPU 0-39.1% flow, then the random flow error would be multiplied by 0.33) in this same step.
Vermont Yankee Design Engineering Page 15 of 29
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APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 Table 8: Flow Input Errors Two Recirculation Pumps Running
% Core Flow Random Error [% Calibrated Span]
Bias Error [% Calibrated Span]
20 0.7266 0.4501 30.9 0.5941 0.2937 31.1 0.5927 0.2919 39.1 0.5513 0.2327 54 0.5143 0.1688 66.7 0.4994 0.1367 75 0.4934 0.1216 83 0.4891 0.1099 99 0.4834 0.0922__
119.4 0.4791 0.0764 Table 9: Flow Input Errors One Recirculation Pump Running
% Core Flow Random Error [% Calibrated Span]
Bias Error [% Calibrated Span]
20 0.4673 0.1137 37.5 0.4358 0.0608 39.1*
0.434776 0.058368 40 0.4342 0.057 60 0.4277 0.038 61.7 0.427428 0.03698 61.9*
0.427396 0.03686 62.5 0.4273 0.0365
- Values linearly interpolated from VYC-0690 Ref. 6.27 4
CALCULATION DETAIL The detailed calculations of the APRM loop uncertainties, setpoints, testing tolerances, and margins have been performed using Microsoft Excel spreadsheets and are documented as Attachment B.
5 RESULTS AND CONCLUSIONS 5.1 Allowable Value The Allowable Value for each required point has been determined and the results are presented in the table below. Since this is a major revision of the method and calculation of Allowable Values and Setpoints no comparison to existing values has been performed.
Vermont Yankee Design Engineering Page 16 of 29 I
I
APRMJIBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 Table 10: Allowable Values Output Instrument
% RxP mV Curve % RxP APRM A, B, C, D, E, F Scram Trips ARTS/NIELLLA Flow Biased APRM High Flux Scram (Two Loop Ops)
Core Flow 0 to <31.1 % (point based on 25% flow) 71.1 5.688
<0.4Wd+61.10%
Core Flow 31.1 to <54.0 % (point based onO50 % flow) 97.3 7.785
< 1.28Wd+33.31%
Core Flow 54.0 to <75 % (point based on 70% flow) 113.5 9.078
< 0.66Wd+67.28%
Core Flow > 75%
117.0 9.357 N/A APRMI High Flux Scram (Single Loop Ops)
Core Flow 0 to <39.1 % (point based on 25% flow) 68.1 5.447
< 0.4Wd+58.09%
Core Flow 39.1 to <61.9 % (point based on 50% flow) 87.6 7.005
< 1.28Wd+23.56%
Core Flow 61.9 to <83.0 % (point based on 70% flow) 108.3 8.664
< 0.66Wd+62.10%
Core Flow> 83.0%
117.0 9.357 N/A EPU flow Biased APRM tHigh Flux Scram (Two Loop Ops)
Core Flow 0 to <30.9 % (point based on 25% flow) 58.7 4.696
< 0.33Wd+50.45%
Fore Flow 30.9 to <66.7 % (point based on 50% flow) 80.7 6.458
< 1.07Wd+27.23%
Core Flow 66.7 to <99.0 % (point based on 75% flow) 103.6 8.287
< 0.55Wd+62.34%
Core Flow> 99.0%
117.0 9.357 N/A
[APRMI High Flux Scram (Single Loop Ops)
C ore Flow 0 to <39.1 % (point based on 25% flow) 56.3 4.500
< 0.33Wd+48.00%
Core Flow 39.1 to <61.7 % (point based on 50% flow) 72.5 5.801
< 1.07Wd+19.01%
Core Flow 61.7 to <119.4 % (point based on 75% flow) 92.5 7.398
< 0.55NVd+5122%
CoreFlow> 119.4%
117.0 9.357 N/A Note: % CS is calculated based on multiplying the calculated Vdc terms in attachments 7.2 by 10 (100% Span/ lOVdc).
5.2 Setpoint Evaluation Results are presented below for the Limiting Setpoint (LSp).
Table 11: APRM Fixed Scram (Reduced) Trip Setpoint Results Description ITS: Analytical Limit (AL);
S 15.0 CTS: Tech. Spec. Limit (TS)
Limiting Setpoint (LSp) 11.72 ITS: Allowable Value (AV) 12.97 Margin to LSp (M1 )
+0.345 Existing Setpoint 11.375 New Setpoint N/A Margin to Normal Operations There is no stable operating point when the reactor is in the "Startu "mode.
These results are presented graphically in Figure 1.
Vermont Yankee Design Engineering Page 17 of 29
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APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision
Table 12: Limiting Setpoints and Calibration Cardinal Points Output Instrument
% RiP mV Curve % RxP kPRM A, B, C, D, E, F Scram Trips ARTSINIELLLA Flow Biased NPRNI High Flux Scram (Two Loop Ops)
- ore Flow 0 to <31.1 % (point based on 25% flow) 69.9 5.588
<0.4Wd+59.85%
Core Flow 31.1 to <54.0 % (pointbased on 50 % flow) 96.1 7.685
< 1.28Wd+32.06%
Core Flow 54.0 to <75 % (point based on 70% flow) 112.2 8.978
< 0.66Wd+66.03%
Core Flow> 75%
115.7 9.257 N/A APRMI High Flux Scram (Single Loop Ops)
Core Flow 0 to <39.1 % (point based on 25% flow) 66.8 5.347
< 0.4Wd+56.84%
Core Flow 39.1 to <61.9 % (point based on 50% flow) 86.3 6.905
< 1.28Wd+22.3 1%
Core Flow 61.9 to <83.0 % (point based on 70% flow) 107.1 8.564
< 0.66Wd+60.85%
Core Flow > 83.0%
115.7 9.257 N/A EPU Flow Biased MPRA IHigh Flux Scram (Two Loop Ops)
-ore Flow 0 to <30.9 % (point based on 25% flow) 57.5 4.596
< 0.33Wd+49.20%
Core Flow 30.9 to <66.7 % (point based on 50% flow) 79.5 6.358
< 1.07Wd+25.98%
Core Flow 66.7 to <99.0 % (point based on 75% flow) 102.3 8.187
< 0.55Wd+61.09%
Core Flow > 99.0%
115.7 9.257 N/A AIPRM High Flux Scram (Single Loop Ops)
Core Flow 0 to <39.1 % (point based on 25% flow) 55.0 4.400
< 0.33Wd+46.75%
Core Flow 39.1 to <61.7 % (point based on 50% flow) 71.3 5.701
< 1.07Wd+17.76%
Core Flow 61.7 to <1 19.4 % (point based on 75% flow) 91.2 7.298
< 0.55Wd+49.97%
Core Flow> 119.4%
115.7 9.257 N/A Note: Due to the small difference between As-Left and As-Found allowances, rounding of setpoints is not recommnended.
5.3 Calibration and Test Results Test As-Found tolerances (FI) and As-Left tolerances (C'T) are for the Fixed Scram (Clamp and Reduced settings) are defined in Tables 13 and 14.
Table 13: Calibration Tolerances As Left (CT)
Description LimitsVdc Limits % RP APRMI Fixed Scram (Reduced) Trip 0.04 Vdc 0.5 % RP Table 14: Calibration Tolerances As Found (FI)
Description l Limits Vdc l Limits % RP APRM Fixed Scram (Reduced) Trip 0.1 Vdc 1.25%
Vermont Yankee Design Engineering Page 18 of 29
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2-Table 15: Calibration Points, As-Left and As-Found Output Instrument Setting As-Left As-Left As-Found s-Found Idc Min Max Min pfax 4PRM A, B, C, D, E, F Scram Trips ARTSIMELLLA Flow Biased l
HP.Migh Flux Scram (Two Loop Ops) l ore Flow 0 to < 31.1 % (point based on 25% flow) 5.588 5.538 5.638 5.488 5.68 ore Flow 31.1 to <54.0 % (point based on 50 % flow) 7.685 7.635 7.735 7.585 7.78 ore Flow 54.0 to <75 % (point based on 70% flow) 8.978 8.928 9.028 8.878 9.07 Core Flow> 75%
9.257 9.2071 9.307 9.157 9.35 APR1I 1ligh Flux Scram (Single Loop Ops)
Core Flow 0 to <39.1 % (point based on 25% flow) 5.347 5.297 5.397 5.247 5.44 Core Flow 39.1 to <61.9 % (point based on 50% flow) 6.90 6.855 6.955 6.805 7.00' Core Flow 61.9 to <83.0 % (point based on 70% flow) 8.564 8.514 8.614 8.464 8.664 Core Flow > 83.0%
9.257 9.207 9.307 9.157 9.357 EPU Flow Biased
%.PRM H1igh Flux Scram (Two Loop Ops)
)w 0 to <30.9 % (point based on 25% flow) 4.596 4.546 4.646 4.496 4.69 Core Flow 30.9 to <66.7 % (point based on 50% flow) 6.35 6.308 6.408 6.258 6.45 Core Flow 66.7 to <99.0 % (point based on 75% flow) 8.187 8.137 8.237 8.087 8.28 Core Flow> 99.0%
9.25 9.207 9.307 9.157 9.357 APRM High Flux Scram (Single Loop Ops) ore Flow 0 to <39.1 % (point based on 25% flow) 4.400 4.350 4.450 4.300 4.500 ore Flow 39.1 to <61.7 % (point based on 50% flow) 5.701 5.651 5.751 5.601 5.801 Core Flow 61.7 to <1 19.4 % (point based on 75% flow) 7.298 7.248 7.348 7.198 7.398 Core Flow> 119.4%
9.257 9.207 9.30; 9.1557 9.357 Note: Recirculation drive flow input to APRM can be deternined by the equation RDF (Vdc) = % flow/125
- 10.
Table 16: Total Loop Uncertainty and Non-Test Error Output Instrument ILU % RxP Son Test % RiP APRM A, B, C, D, E, F Scram Trips ARTS/MELLLA Flow Biased H.M High Flux Scram (Two Loop Ops)
Core Flow 0 to < 31.1 % (point based on 25% flow)
-4.55%
-3.303' ore Flow 31.1 to <54.0 % (point based on 50 % flow)
-4.94%
-3.69o/
ore Flow 54.0 to <75 % (point based on 70% flow)
-4.47%
-3.22o/
Core Flow > 75%
-4.29%
-30404 APR.I IHigh Flux Scram (Single Loop Ops)
Core Flow 0 to <39.1 % (point based on 25% flow)
-4.36%
-3.11%
ore Flow 39.1 to <61.9 % (point based on 50% flow)
-4.49%
-3.24%
ore Flow 61.9 to <83.0 % (point based on 70% flow)
-4.35%
-3.103' Core Flow > 83.0%
-4.29%
-3.04o Vermont Yankee Design Engineering Page 19 of 29
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 Table 16: Total Loop Uncertainty and Non-Test Error Output Instrument iTLU
% RxP Non Test % RxP EPU Flow Biased PM High Flux Scram (Two Loop Ops) ore Flow 0 to <30.9 % (point based on 25% flow)
-4.50%
-3.25°k l ore Flow 30.9 to <66.7 % (point based on 50% flow)
-4.82%
-3.57Th ore Flow 66.7 to <99.0 % (point based on 75% flow)
-4.41%
-3.16l Core Flow > 99.0%
-4.29%
-3.04ol PM High Flux Scram (Single Loop Ops) ore Flow 0 to <39.1 % (point based on 25% flow)
-4.35%
-3.10%l ore Flow 39.1 to <61.7 % (point based on 50% flow)
.4.44%
-3.190/
re Flow 61.7 to <1 19.4 % (point based on 75% flow)
-4.33%
-3.08%
ore Flow> 119.4%
-4.29%
-3.04%
Table 17: Total Loop Uncertainty and Non-Test Error Putput Instrument TLU %CS Non Test % CS lPRM A, B, C, D, E, F Scram Trips ARTSIMELLLA Flow Biased High Flux Scram (Two Loop Ops) l Core Flow 0 to < 31.1 % (point based on 25% flow)
-3.64%
-2.64%
Core Flow 31.1 to <54.0 % (point based on 50 % flow)
-3.95%
-2.95o/
Core Flow 54.0 to <75 % (point based on 70% flow)
-3.58%
-2.58o/
Core Flow > 75%
-3.43%
-2.43o/
PRM High Flux Scram (Single Loop Ops) ore Flow 0 to 539.1 % (point based on 25% flow)
-3.49%
-2.49Y ore Flow 39.1 to <61.9 % (point based on 50% flow)
-3.59%
-2.59o/
ore Flow 61.9 to <83.0 % (point based on 70% flow)
-3.48%
-2.480/
Core Flow > 83.0%
-3.43%
-2.43o/
EPU Flow Biased PM High Flux Scram (Two Loop Ops) ore Flow 0 to <30.9 % (point based on 25% flow)
-3.60%
-2.60°/
ore Flow 30.9 to <66.7 % (point based on 50% flow)
-3.85%
-2.85o/
ore Flow 66.7 to _99.0 % (point based on 75% flow)
-3.53%
-2.53Y Core Flow > 99.0%
-3.43%
-2.43/
HP High Flux Scram (Single Loop Ops)
-ore Flow 0 to <39.1 % (point based on 25% flow)
-3.48%
-2.48/
Wore Flow 39.1 to <61.7 % (point based on 50% flow)
-3.56%
-2.56o/
Wore Flow 61.7 to <119A % (point based on 75% flow)
-3.47%1
-2.47o/
sore Flow> 119.4%
-3.43%
-2.43o/
5.4 Calculation Review and Impact Consideration 5.4.1 This calculation evaluates the uncertainty of loop components for design changes including Extended Power Uprate. The uncertainty determined by this calculation Vermont Yankee Design Engineering Page 20 of 29
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 will be used as input for the Reactor Scram Flow Biased and the APRM Fixed Scram (Reduced) Trip Setpoint and Allowable Value results.
5.4.2 The Design Input Considerations of ARTS/MELLLA and the VY Extended Power Uprate as well as the change in the methodology used to develop Limiting Sctpoints and Allowable Values have been considered in this calculation. The setpoints developed in this calculation are not applicable until the VY licensing amendments BVY 03-80 for Extended Power Uprate and BVY 03-23 ARTS/MELLLA are approved (as applicable).
5.4.3 The function of the instruments covered by this calculation is an assumed input in the Reload Licensing Analysis. This analysis does not assess conformance to 10 CFR 50.46, "Acceptance Criteria for ECCS for Light Water Nuclear Power Reactors," and Appendix K, "ECCS Evaluation Models". The results of this calculation do not identify errors or require changes to the Reload Licensing Analysis. Therefore, the reporting requirements of 10 CFR 50.46 are not applicable.
5.4.4 A review of the Vermont Yankee Event Report Database was conducted to identify any Event Reports that would impact this calculation. This review identified no event reports, associated with these components.
5.4.5 Precursor calculations used for design input to this calculation are not impacted by the results of this calculation. Any applicable interactions due to changes in precursor calculations will be addressed per AP 0017 during the change process for those calculations. Calculation VYC-0690 [Ref. 6.27) provides direct input to the flow errors used for this calculation.
5.5 Summary of Requirements 5.5.1 The current calibration procedure, OP 4308, Rev. 21 [Ref 6.25] requires update to accurately reflect the LSp and As-Found and As-Left tolerances. A precaution should be added to Ref. 6.25 to ensure that M&TE with an accuracy better than 0.05% of span is used to set the flow bias input value during calibration.
5.5.2 This calculation is not an implementing document and a 50.59 screen or evaluation is not required. The output of this calculation is implemented through update to applicable plant documents (i.e., OP 4308, FSAR). The downstream process that updates applicable output documents will satisfy the 50.59 evaluation requirements.
5.6 Summary of Recommendations 5.6.1 Based on the results of this calculation the following changes should be made to the evaluated trip setpoints:
5.6.1.1 APRM Flow Biased Scram Trip Setpoint.
Vermont Yankee Design Engineering Page 21 of 29
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APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 5.6.1.1.1 The formula to develop the curve for the APRM Flow Biased Scram Trip setpoints must be changed in accordance with Table 12 for the specific application of ARTS/MELLA or Extended Power Uprate.
5.6.1.1.2 Revise the As-Left and As-Found tolerances as defined in Table 15 for the new calibration points.
5.6.1.2 APRM Fixed Scram (Reduced).
5.6.1.2.1 Trip setpoints must be changed in accordance with Table 12. The values for Allowable Value and Limiting Setpoint do not change for this function for ARTS/MELLLA or EPU.
5.6.1.2.2 Revise the As-Left and As-Found tolerances as defined in Table 15 for the new calibration points.
Vermont Yankee Design Engineering Page 22 of 29
.14 I.
... 1 ;
- -: ':
...- -: I"-
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 UXL Uumr CTS: Tech. Spec. Limit (TS)
ITS: Analytical Limit (AL) = S15% RP ITS: As-Found (AV) = 12.97% RP Limiting Setpoint (LSp) = 11.72% RP Setpoint 11.375% RP Ut l
It Margint
<process>
I I -
I Normal Range of Operation Figure 1 APRM Fixed Scram (Reduced) Setpoint Vermont Yankee Design Engineering Page 23 of 29
..e...
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 A/M APRM FB SCRAM (TLO) 120 110 100 at 90 S..
.E 80 0
10.0 Cn 70 60 50 40 0
10 20 30 40 50 Drive Flow 60 70 80 90 100 I --- Scram Analytical Limit
--aScram Allowable Value
.--- Scram Nominal Trip Setpoint I Figure 2 APRI Flow Bias Scram Analytical Limit, Allowable Value, Nominal Trip Setpoint (ARTS/IMELLLA) Two Recirculation Loops Vermont Yankee Design Engineering Page 24 of 29 t
.-. ::
,
-4 APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 A/M APRM FB SCRAM (SLO) 120 110 100 D.y 90 n-so
.E 80 0
0 Cn 70 60 50 40 I.
0 10 20 30 40 50 60 Drive Flow 70 80 90 100
-*-Scram Analytical Limit
-4.-Scram Allowable Value
-u-Scram Nominal Trip Setpoint I Figure 3 APRN1 Flow Bias Scram Analytical Limit, Allowable Value, Nominal Trip Setpoint (ARTSAIELLLA) One Recirculation Loop Vermont Yankee Design Engineering Page 25 of 29
. I..
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 EPU APRM FB SCRAM (TLO) 120 110 100 z-90 0-g
.S-80 0
50 4-0
'O 70 60 50 40 0
10 20 30 40 50 60 Drive Flow 70 80 90 100 Sccram Analytical Limit
-+-Scram Allowable Value
-a-Scram Nominal Trip Setpoint w
Figure 4 APRM Flow Bias Scram Analytical Limit, Allowable Value, Nominal Trip Setpoint (EPLJ) Two Recirculation Loops Vermont Yankee Design Engineering Page 26 of 29
.4:.
.-~~~~~~~~~~~~~~~~~~
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 EPU APRM 1B SCRAM (SLO) l I i I I I
120 110 100
- a. 90
.E 80 0
40.
U0 70 60 50 40 0
I l
I 0
10 20 30 40 50 60 70 Drive Flow I
I I
80 90 100 110 120 Scram Analytical Limit
.- Scram Allowable Value
-u-Scram Nominal Trip Setpoint Figure 5 APRM Flow Bias Scram Analytical Limit, Allowable Value, Nominal Trip Setpoint (EPU) One Recirculation Loop Vermont Yankee Design Engineering Page 27 of 29
APRM/RBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 6
REFERENCES 6.1 "Instrument Uncertainty and Setpoints Design Guide," Vermont Yankee, Rev. 1.
6.2 Design Specification 22A1366, Rev. 3, "Neutron Monitoring System" 6.3 Design Specification - Data Sheet 22A1366AF, Rev. 1, "Neutron Monitoring System" 6.4 "Vermont Yankee Updated Final Safety Analysis Report, Rev.17."
6.5 "Vermont Yankee Environmental Qualification Program Manual," Rev. 38.
6.6 "Vermont Yankee Design Basis Radiation Dose Specification," Yankee Nuclear Services Calculation VYC-193. Rev. 4 6.7 Vermont Yankee Technical Specifications through Amendment 212.
6.8 Not Used 6.9 AP-0017 Calculations and Analyses Rev. 8 LPC 2.
6.10 Not Used 6.11 Design and Performance Specification 175A1259, Rev. 1, "APRM".
6.12 Not Used.
6.13 General Electric Instrument Setpoint Methodology, NEDC-31336 (GE Proprietary), October 1986, Section 3.19 and 4.5.
6.14 NEDE-2401 1-P-A-14, June 2000, General Electric Standard Application for Reactor Fuel (GESTAR II).
6.15 GE-NE-0000-0012-0531-01-01 Revision 2 November 2003 Project Task Report "EntergyNuclear Operations Incorporated Vermont Yankee Nuclear Power Station ARTS/MELLLA" Task T0506:
NSSS TS Instrument Setpoints.
6.16 GE-NE-0000-0016-5688-01 Revision 0 August 2003 Project Task Report "Entergy Nuclear Operations Incorporated Vermont Yankee Nuclear Power Station Extended Power Uprate" Task T0506: NSSS TS Instrument Setpoints.
6.17 GE Performance Specification 25A5903 Rev. 1, "Flow Control Trip Reference (FCTR) Card".
6.18 EMPAC Database.
6.19 Core Operating Limit Report (COLR).
6.20 Measuring & Testing Equipment Uncertainty Calculation, VYC-1758 Rev. 0.
6.21 Not Used.
6.22 Not Used 6.23 OP-43 108 Rev. 6, APRM Functional/Calibration (RMSS not in run).
6.24 OP-4400, Rev 21, Calibration of APRM System to Core Thermal Power.
6.25 OP 4308 Rev. 21 APRM Monitor Calibration.
6.26 OP-2429 Rev. 14 "Recirculation Flow System Baseline Data Collection and Instrument Calibration."
6.27 VYC-690 Rev. 2 "Recirculation Flow Loop Uncertainties to APRM".
6.28 NEDE-2401 I-P-A-14, June 2000, General Electric Standard Application for Reactor Fuel (GESTAR II).
6.29 NEDC-33089P, "Vermont Yankee Nuclear Power Station APRM/RBM/Technical Specifications Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)," March 2003.
6.30 NEDC-33090P, Rev. 0, "Safety Analysis Report for Vermont Yankee Nuclear Power Station Constant Pressure Power Uprate," September 2003 6.31 BVY 03-23 March 20,2003, Technical Specification Proposed Change No. 257 Implementation of ARTS/MELLLA at VY Vermont Yankee Design Engineering Page 28 of 29
APRMJRBM Neutron Monitoring Trip Loops VYC-693A, Revision 2 6.32 BVY 03-80 September 10, 2003, Technical Specification Proposed Change No. 263 Extended Power Uprate.
6.33 VYC-2252, Rev. 0 "Drift Calculation for Average Power Range Monitors."
7 ATTACHMENTS 7.1 Loop Diagram [ 1 pg]
7.2 Calculation Detail (BI through B4)
[8 pgs]
7.3 Design Specification 22A1366, Rev. 3, "Neutron Monitoring System"
[2 pgs]
7.4 Design Specification - Data Sheet 22A1366AF, Rev. 1, "Neutron Monitoring System"[1 pg]
7.5 Design and Performance Specification 175A1259, Rev. 1, "APRM" [3 pgs]
7.6 GE Performance Specification 25A5903 Rev. 1, "Flow Control Trip Reference (FCTR) Card
[ 23 pg]
7.7 AP-0017 Forms and Interdepartmental Review Form [7 pgs]
Vermont Yankee Design Engineering Page 29 of 29 Zs
LPRM/APRM NEUTRO' MONITORING (APRM LOOPS)
REACTOR I
INDICATOR DETECTOR (TYPICAL OF 40 PER LOOP)
I ALARM CIRCUIT (UPSCALE & DOWNSCALE LEVEL)
DW -
-RB-CR ALARM CIRCUIT (UPSCALE
& DOWNSCALE LEVEL)
SCRAM CIRCUIT (UPSCALE LEVEL)
ALARM CIRCUIT (UPSCALE LEVEL &
COMPARATOR)
L.
l lRECWRCULATION FLOW ALARM CIRCUIT l
(REFER TO VYC-690)
(UPSCALE& DOWNSCALE LEVEL)
I CD qA To o6 -
C., 5 VYC-693A
APRM SCRAM TRIP POW REVEW ARTSM4EUA Two Loop WC-0693A, Resion 2 Abdsnei 7.2 82. Pgs I of 2 A
BC 0
E F
I 11 I APRM Sams Cialcul nbARTSiMEULLA Two Reendisi L-OVSImm a
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SENSIfhOTY ACCURWACY REPEATABUJTY 3EOUPMIENTLto.
INSTRUMVENTTYPE Foiorst rAPort %
%CSC 4
O.W022453.7 ND1402-i-104 LPRM DETECTOR 7
.LPRM
_0.26 7%
S APRM
.000%
101 FLOW SOCA 03.%125 0.40 72.40 2.0000%
III FLOW BIASEDOSCRAM (31.12-540%)
22 120 76.81 1.0000%
12 FL WBLASEDSCRAM(540.75.0%)
125 0.80 10614 I.0000%
1.3 FLOWBIASED SCPAM(,75.0%)
i2 rmnisV
~
120.00 I.D_______DD100%
1 4 FO(ED SCRAM (REDUCED) 125 Ws, 1i.0 1.000%
15 CS__
1 6CALIBRATION EQ.U'MENT UN4CERTAINTY (MarE) 0071%
(2.3.12 IN TEXT 17 RLOWINPUTUNCERTrAN1BIAS (FROM VC.9000REV.2) 0.4501 0.2919 0.100 10 R.OW W¶JTUNCERTASOMYRAND0OM(FROMVYC*000.
REV.2)
L___
0.7206 0.5927 0.5143 CS FOR FLOW SIONAIS 128.6%
19 FLOW UNCERTAtM 00sm EFFECT
(%RxPOWER) 0.2315%
0.4805%
- 0. 1433%--
CS FOR POWER IS 125%
20 FLOWUNCEfRTANI M IJOMEFFECT
.0es.25%$TSe~Sc23.q(%R POWER) 0.3953%
1.0500 0.4054%
21 FLOW SEITTWTOLERW4CE 0.25 (SORT( I0"2C2M~2 00 l28~l0 (SORT E18-2.$C$2lA2) 00.28W011 (SORT(FIB'2.$C$21'2 i00r1.286¶DI2 22 MM"h FLOW VALUES TO BE EVALUATED RFOW) 20 31A 22 FLOW VALUES TO BE EVALUATED I% FLOW) 20-31.1%
31.1-540%
54.0-75.0%
24,
M!EUM NMAER OFLPRW s PER APRM 0
25 FRlMARY ELEMENTACCURACY
(%RP) Ss~i250.02%
Bins ~
A 0.340000%
20 ROCE 55 MUN ACRACY%
)
i2.6.)
2 320%
28 TECH SPECLUN.r REOLMED FOR CTS 120.00 %RP 20 AO4ALYTICLUMT REOUIREDFOR RS 120.00 %RP 31 CSw4slRwswsom ci
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ESTW UNERTAWI 32 RLOW BIASEDSCRAM(20-31.1%A E3%IC$10 LOOP TESTING UNJCERTAINTY LOOPTESTINGUNJCERTAINTY 0.O4SOT 5E92P24$C$2246-20'm2401194N510) 35CLIANEFFC (CE)
CALIBRATION TOLERANCE CALIBRATIONTOLERANCE
$C$I0lSORT(A3W~)
SORTIA35?2)
OB.B'SORT $E$25124$CS2B'$2.ES2'24OS194$NSI6
-5C$2540019 34
%CS I%RP 15CS
%RP
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%RP 351 1.0000% 10.625 0.50%
t.25 1.00%
2.2440%
301
.1.00%
1.0.625
.0.50%
1.25
.1.00%
32956%
381
-A35 40 FLOW BLASED SCRAl (31.1.540%)
41 CALIBRATION4 EFFECT MCALEBRATIORTOLERANCE CALIBRATION TOLERANCE LOOP TESTING UNICERTAINTY LOOP TESTING UPICERTADITY NON TESTING ERROR 42
%CS
%RP 5CS
-%RP
%CS
%RP 43 1.0000%
0.625 0.50%
1.25 1.00 2.3909%
44 l.1.000%
40.625 4050%
-1.25
.1.00%
.3.6914%
45 46 41 FLOW BASED SCRAM (54.0 -75.0%)
CALIBRATION 48 CAUBRATI0PN EFFECT (CE TOLERANCE LOOP TESTING UINCERTA*TM LOOP TESTESUN*CERTAtN1 NOTESTING ERROR 491
%CS I
%RP
%CS
%RP CS%RP 50 1.0000 0.625 0.50%
2.2 120%32595%
-1.8000%
-0.625
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53 54 FLOW BIASE0 SCRAMJ i-75.0%)
CALIBRATION 55 CALIBRATICONFFC ME TOLERANCE LOOP TESTING LUN0ERTAFMT LOOP TESTINGLESCERTAINTY NON TESTMSERROR 56
%CS2 I
%RP
%CS
%RP
%CS
%RP 57 100%0.50 0.40%
2.2 1.00%
2.2193%
a0
-1000
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-1.25
-1.00%
-3.0303 00 FIXED SCRAM (EDUCED)
CALIBRATION 61 CALIBR.ATIN EFFCT C TOLERANCE RPLOOP TESTING UKIERTAINTY LOOP TESTS LUNCERTAINTY NON TESTING ERROR
.6
%C05
%CPS0
%RP
%C05
%RP 03 2.0000%
0.50 0.40%
1.25 1.00%
2.2193%
1641
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-1.25
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%RP 5.53 IVA 5.53 6910 5478%
I S048 533 0o so 53.70%
1 5.38 61.10 59 85 39 ol AV ard Selpert Cims_
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LSP 42 V. CS RPP 33231 12r"d33 Medhd I A ITS: (AV]
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1 570 71.24 12rWdr32 0 I
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RP WA AARGN
%RP NA
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91.57% 1
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9 30 92.17%
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%CS l
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R
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10.4 1
1.97 8 39%
0.84 1
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FLOW INPUT LUCERTAINTYAS (FROM VY-6K REV0.2)
FLOW WM UKcERTAINY RAND)OU
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FLOW UNHCERTAINTY BIAS EFFECT (TO R5, POWER)
FLOWUNCERTAJTY0 RMO20AEFFECT x~nct~do25%
STC S..2.3 436%
FLOW SETTINGOTOLERANCE UINAAMd FLOW VALLES TO BE EVALLUATED
(%
FLOM%
FLOW VALUES 70089 EVALUATED 6%
FLOM~
125
-2 UMMMUNLU3E OF LPR~s P14R APRU TECH SPEC LOOT REQUIRED FOR CTS MNAILYICLWT REOUIRED FOR ITS 3RATEFONEFFECTCF
%CS 1.0400%
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20 00 t20 00 CALIBRATION TOLERANCE
%Rp 0 625 40 625 CAL IBRAT ONTOLERANCE
%RP 0137 0 4003
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112728%
(ScTPD~r2C2I'2YI003-M"E0 20 20239.i%
12000o IsO,0 (213 t2INETEXT) 0 058360 04340770 00003%
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SORT(E 10'2.6C$21'Oyl Mrl.20r0 39.1.41690%
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(SOT['18'2C%212Y1OO)
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Yankee Nuclear Senrces.3 VYC-0693A Rev. 2 Page 1 of 2 U EllERA L Q ELECTRIC ATmIC Powra EQUIPMENT OEPARUE~NT DESIGN SPECIFICATIOR I
s.cc.ac. 22AI366 S. MCI 12 CO.r a-1.1.CCV13 4.4.2.1.
The detector sensitivity shall be such that the output of the signal eonditfoner corresponds to the required percent pover reading associated with the neutron flux Levcl In the viecinity of the detector.
4.4.2.2.
The rangp of neutron flux (operatig) shall be as specSfied on the data dect.s.
iC,4.2.3.
The rage of neutron flux at 100 percent reactor power halL bc as specified on the data sheets.
4.4.2.4.
The range of gamma dose rate at 100 percent reactor pover shall be as specified oo the date sheets.
4.4.2.S.
She detector shall be designed such-that the saturation character-Istics of the detector do not caSc an error In the signal due to power supply variztion* of greater than +1 percent or the 16b percent reactor pouer value over the operating nectron flux range specified :in the data sheets.
4.4.2.6. Ihe detector shall be so designed that the true output chall not deviate fron the Ideal output by wore than +1 percenc of the 100 percent reactor power value over the top decade of the cpexatfrg fluK range specified in the data sheets.
4.4.2.7.
~t~e detector shall have aulminun lifetine of 2.0 years in the lifetime neutron flux specified In the date sheets.
End-of-life is defined as having occurred %&ea the ratio of the output uignal resulting from nentron*. -ta the output signal retulting from gas 5s tesches 5 to 1.
4.4.2.6.
The detectortcable leahage eurreat shall not exceed 41.6 percent ef the full scale output during the 1iff of the detector.
Leakage current is de-fined as that currant presented to the inafl conditioning equipment bhen the detector is at operating conditions but with no neutron or samma flux present.
4.4.2.9.
The detector and detector aseaubly nvironmnent shall he as follows:
- a.
Neutron and Cmna, Tlux at Detector as specified on the date sheets.
IL In---
IDs...
etr wrzneo Ca=&l'%,
2.4 R/hr Ieutr:Ir:
10 ren/hr
- c.
ZRactor Pressure:
Operation:
1025 pals at 546F
- Design, 1250 pcug at 600'P Maximum Emergency fressure:
1375 paig at 383'F
.d I,.. -
I FEE 9 PIZ I
.3 VYC-0693A Rev. 2 Yankee Nudear Services Page 2 of 2 GENERALCo ElECTRIC AtC~fC POcVE EQUIP90Nt OPARTMt)T D)ESIGN SPECIAT c
- o. 22M1366
.o 3
4.4.52.
The true output of the aignaL conditioning equipnent shall w t deviate from the ideal output by aore than 40.8 percent of full icale at coatrol room desizn center environmental coadttioea over the range of flux (operating) specified In the date sheets.
4.4.5.3. At the control roow design center environmental conditions specified Itn the data sheets, the equipment shall not hava a long terf drift (700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br />) gtestAr than 4(.8 percent of ful scale.
4.4.5.4.
The signal conditioning equipment shall be deaSined cc thitc the gain can be set at a dasired value within 40.1 percent of the full scale value over Urc range of adjustment.
The gain shall be adjustable in three steps and each step shall be continuously adjustable by at least a factor of 5.
4.4.5.5.
Ihe signal condirioninz equip itt *Wall provide a 0 to 160 wtV for xero to full scale output for the coaputer.
4.4.S.6.
The signal conditioning equipment shall provide a C to L0 volt A gual' to the Average Pever Range Monitoring Wystem.
Assignment o: the various signal conditioning outputs to the APIR'a is wade on the 2nctrumenc Engineering Diagram (LED) licted 'in Paragraph 2 of the data sheets.
The signal conditioner shall be designed such that two AiRHes may be driven from each LPRH signal conditioner output.
4.4.S.7.
The signal conditioning equipment shall provide a 0 to 10 volt signal to the Rod Zlock M)outor (RIM) System specifted to Paragrarp 4.6.8. The required "A core level eand C' core level couditioned detector'signale 4haIl Ub eonnected through a avItebing matrix to the RlK associated with seactor ?rotectkon Systea Trip System A.
The required '"
core level and ID" core level conditioned detector signals shall be coanceted through the Cwitching matrix to the RUK associated %dtb Reactor Protection hysteu trip System Z.
The appropriate Inpute shall be automaticelly switched upon celection of a rod.
Selection details are found on the Instrument Engincering Diagtara (IED) listed in Paragraph 2 of tbt data sheets.
4.4.5.8.
Signal Output to LPRN Meter Croup Display - The equipment shell be designed such that a gtoup of conditLoned outputs are provided to the IPRK Group Diapley upon selection of a control rod. Particular selection and routing details are as specified on the Instrument Untineerlng Diagram (lED) listed in Paragraph 2 of the data sheets, 4.4.6.
Trip end Alarm Fuetciat.
The folloving trip and alarns e required::
- a.
Upscale Level Alarm
- t.
Devgeale Level Alarm 9
f
- t, Rp Z'1 r........
....>.~'
.4 YC-0693A Rev'. 2 Yankee Nudlear Services Page 1 of I GENERAL4DEETI AMC. -PCVwt tOMPRETtOPRS4 EL~~~CrRIC
-CC 221I6A These SOMttts ate plated MXAiUsY IU the Core AS Ahovu au T1VWC !,
and w*AJLaly ts *bAxm on the Ins tusent, anglaeerizZ digram listed Lu Paraxrah 2 of theme date sheets. T.
sources; sAml be Irradiated in a swten&.
desccihed an the Lrradasticn specificato Ua ced on the ikutroa Sour"e DwdngLU lstod In TaWstagkph 2 of these data s!hect.
b~t. peuatrstiflaf shal be desitned and Installed esich that special conrdIa cables can be used for connection of detectors to signal conditioaing *quipimcat.
4.4.2t.
Intermediate R-ange bonitcorinit System.
-sam as r~arar-4.4.3.. rover 2Ante
)4oaitoriux Syses A.
Detector sod Detector Assembly eawl be VIlaced within the core iu the wmeer shav*7La FIngure 2 of these data mlieetx.
- b.
Rsapg of Operating NCeutrom Flux hAlIT be from 1.2 x IO nvU to 2.81 x 1014 CT.
- c.
lbs !kutroa Flux at !00 percent Reascci towesr La dih yicinty of the L~ctorr shall bwe aildtn the r=uFe of 9.4 x lDnUr to Z.t x :Iuv. The peak neutron flux at 120 percent of rated power wll be 3.4 I 10O'4at.
Mme detecror signal resulting froo this fins *Isi Dot detiate from the 14*al output bj date than Claterc) percent of the 100 percant rated pwer vvlue.
- d. The cam riti at too rerceat v~aco roe Late ofi.
t.ba 4etevtor shal be within the range of 4.2 X 10 Xkhr to 1.2 z la Kxar.
- v.
Detector JLifetian Uuitron Flux shall be X.8L vla i)Uav -ftexvf at the asaxtum gam, flux apectIL&Cd 4.44. ~tz
?o~r ant Ptoitorln.
1hea the teactor paver is derbcdvt e% octant e7 try. Che.qvivealtt poetlteaf sW1l be as shownu Lu Ta~le L'of these data osheet.
4.4.5.
Lmtace rwaer Passe oneitordnt.
Use a of LWrit outpuzts eeted as L~mpc to ea AMI that vu7t be bpaff0* at a co* t~ime sball &oc excee o=
four AM Vi that have ten LWC Inputs (smm coatift L1Vk power" trow aoothar Weis) and' eev for K AMV thart UaS tveuty MMihaftd 1A.K13Laet.
tiodir coed Catss wlwre all L1IM afgale firon i4wb Aasociat*A 0
~~~~it Z COP40 "v
sleasiag.
AflDIhaIontA md...e~-
a:s_..
e.
i
v *
- .5 VYC-0693A Rev. 2 Page 1 of 3 Yankee Nuclear Services GENERALE~ EUtCTRIC 4
175A12S9 l., 04-0 2
emw 175A12sg canu a~to -2 Pa
'MLE IG PERFORMANCE SpaC.
AI APR X nn~wsrmsV voz 1t:$
a G'IAJ3 I
1.0 TEMPERAIMR AXD RflQDITY 1.2 1'oll 2.0 POW1ER 2.1 4-5 volts DC
.2.2 tJ5 volt. DC 3.0 IMMUS 3.1 LPPX 3.1.1 Polarity and Range 3.1.1 Maxdm g1!obr -to operate 3.1.3 )Uinl1
%%=ber to operate in apl-c.
3.1.4 Hinimaw 'numcbtt to opexat.e 3.2 LI'RH Sypase 3.3 LPRH Coutm 3.3.1 lliaber 3.3.2 ftlxrity and Range 3.4 Calibrator rover 3.4.1 lklwmbr 3.4.2 Value 17SA9680 parr. 1.1 175A9680 pars. 1.2 175A9683 17SA9698 o to +10 volts DC 24 10 6
1 for each LP2H I
-0.5 ma per LPRM 1 for each 1X
-14 volts DC nomiusl 2 to 10 vults DC nomnaL 0 to -1 volt 1.2 ohms 0 to -160 WV 215 ohm 0 to -10 volta DC o to -10 volts DC I
4.0 OUTPUTS 4.1 4.2 4.3 4.4 Recorder~
Output IUpqda~ce Coup-car Output 7mpedance Ketor RM( Rteoroue Sigual (Isoalted) 5.0 cOMOT~LS MOD INOICAMRS~
5.1 mlode switch
. 'to.r1.
I 5.1.2 Standby.
Same SW 5.1.1 except looper~attv~e krip is tripped.
5.1.3 Zero.
LPR)~'t Aixgone.ctoa. DC anp can be adjustad to sey 5.1.4 Paver Tact.
.Operator w7~ apply &od adjustcable ximulatod rov~er aigual uhlle stiEll co~kected to the actuzal flow sigunal.
5.1.5 Po-er and Flow Teat.
Operator say apply Adjtstable A-IMulated aignolff
- Ffr-tbh
'Pvoo,-.. A V1.vi
-- P2Af~..
lc
1F Pe ji ALLrP 175A!
t2A7- -- Iio/,t47L KI, A
MMMA llLX0.4 cwo- 00
.~~~".,....
259 a..
I 3
.5 VYC-0693A Rcv. 2 Page 2 of 3 Yankee Nudear Services FEtERAL ELECTbt I lC pMAsc nt.Ei-l
~~~TrE DEI6N b PERFORKVCE :SPEC.
e e 3
_a2 175A1 Z59 A P RMt oft *o 3
-WN&
0 utsW nor ESv Co c 51,S El Z
73i_
5.2.Hater F4al-ti Swit
-c 572.1 gea Ae oaput of aeac of SLeRHs Dty sueaed.
jtu S.2.2 rcw. Output from Foberverr fnic.n
..2.3 count t terRbe of S
bc ng Averaged._
Kleter mding ic S tiec the oer.
r.
5.2.4 A1 A6, IBl-BA6, C1-06, DI-D6.
Meter rea output of each I.MK.
Select$5 calibrator power from observed L1RK.
5~
5.3 Merxr Revrse. Expand Switch 5.3.1 Reverse 1 volt f.s.
CY 5.3.2 lb 10 VoLtC f.c.
5.3.3 X10 '2W 1 volt f.s.
5.4 ?ower Test Potentiometer.
Adjustable miulated Power Signal; Range-zero to greater than fall actle.
5.5 Flow Test Potentiometer.
Adjusteble aimuLated XE3ow signal; Ren.-lesa than ceto to greater than 1001 rov.
5.6 Calibrtian Ajust Potentiometer.
Controls voltage to precision resisters of CaLibrator Range-0 to greater than -10 volta DC.
5.7 Calibration Monltor Switeh. Alovs Calibrator voltate to be viewed ac eer.
5.8 PRX 3bypucd Light.
Indieates vben LM beng Msitored is in the.Calibrate or Bypasted condition.
5.9 Keter kpaa6 Light. Indcates when the meter J
lReverse. XPal Switch is is a position otbeT than XCIVIA.
S.10 rip temet Switch.
Resets a1l latching trip:s In the Aram and Associated UIm's.
6.0 PERFORHAKCE 6.1 Averaging Circitry 6.1.1 Accuracy (inciudes Lizearity and Stability)
A7 Xstrictod Condltlo
'0.e2 mUll CondItions
'2.42 6.1.2 Drift 00.5X/2 ueeks 6.1.3 Respaose Time 5s
.M 6.1.4 Cain 2.5'640%
6.2 Trip Circuits (Mo-Tloov iisaed)
Spees. apply when using Quad Trip Unit 13611322 a3d Irip Reference Unit 136R1321.
I
~.tA~e~i I
,.. I7 SA
.QSE CAL1FORMA K:
O
,.a, s
._F d_.
l
.~~~~~~~~~~~~~~~~~~~*..5 TYC-0693A Red'. 2 Yankee Nuclelear Services Page 3 of 3 Yankee Nuclear Servrces CEI4ENAL~ RECTA ICt w
1 75A1259 IF V' -.A T'
DESIGN I PERFUO14ACE SPEC.
17SA1252.
lAP R K
~
M*.~
F Am* 3 InASTWO4AE~et STAKM~3 STA S~ Pat.7..3
. 3 6.2.1 AccUrRC7 Restricted Conditions Tdl Cou4LtLons 6.2.2 MMif 6.3. Trip Circultrs (Fiw Biased) spe~ca.
- pply wiben using Quad Trip Unit L36EL322.
Trip Rerereate. Unt LL36R1321, and Trip Dia" Unit 2L6X541.
6.3.1 Accuracy
- j2f.
t 2X f.is
'O.5112 getks W.=O*a PA kto 410 0 2 I :j 1
P.
It-I F". 1 0
41
- 51
- W 6.4 6.3.1.1 Ret'ticted Canditionc 50 to 1oot flou 0 to 501 floe 6.3.1.2 Full Conditions SO to 100% flowe O to 50% flou Count trip Reierence, 6.4.1 LUee~rity lestrrcted Conditions Full Couditions lestricted CoodItions Taul Canditionx 6.4.3 Drift X21 92Z
'11
- '12
'0.522 w.eeks
.4 7.0 CSTRUMctOf 7.1 c*rds 7.2 )'adulas 7.3 MAched-4-l Xig A.0. TEST REDUIROD1{T 17SA129 9 A7
.pzV.-q-fi
'.f.7 41'/b ISAIN JOSE.
CALIFORNI "1
17F541259 I.
A-L t.
C4" ADV)~A IfP Nfi yr-.t'-.-~ I..-..
0 Co E Nlu clear Energy REVISION STATUS S1{FZT.6 VYC-0693A Rev. 2 Page 1 of 23 DDC TITI.E V IAOlV CONTROL TRIP RFFERENCE CARI1 LEGEND ORPFSQR1TON OF GROUIPS I - DENOTES CHANGE MP Yns I TyPJ+
DS(
FdF.
OF)
L ITEM.:
NA 75-1Non HON [A 3N SPEC.
JEEP CLASS )E.
'11h1S ITM IS OR CONTAINS A SATTETY RELATID ITEM EQUIP CIASS CODE A
REVISION C
0 RM.01079 JUN20. 1994 I
R.A. SEARLES FEB OR. I99 RJA CN02 137 CII K BY: R.A. SEARLES ENGR:
C.J. M1IXI.ER 2
R.A. SEAR 09 i
CKN084 CHYK BY: R.A. SEAtLES ENGR:
SI). SAWYER PRINTS TO MADE BY APPROVALS GENERAL El.FXrC COMPANY 175 CURTNkR AVENUE C.J. MII1.F.R FEIv 09. 1994 P.J. KINDER JUN 10, 1991 SANJOSE CALIFORNIA 95125 CIKED 1W ISSUED 24A5215 R.A. SEARI.FS JUN09. 1994 R.J.AHMkANN JUN20. 1994 CONTONSIIEET 2 L6-XS
. Z.. -
0 Flow Conti L
1.1 1.2 2.
2.1 2.2 3.1 3.2 3.3 4-4.1 4.2 4.3 4.3.1 4.3.2 4.3.3 4.4 4.5 4.6 4.7 4.8 4.9 4.10 S.l 5.1 5.2 5.3 5.4 GE Nuclear Energy
~~.
I 24A5215 Si NO. 2 Rev 2 OF 19 rol Trip Rcference (F7CTR) Card DESIGN SPECIFICATION TABlE OF CON TFrTS DOCUNFNIT L)ESCRIPTIO-4 DOCUMENT PURPOSE AND USE 4
DOCUMENT SCOPE 4
RELATE[P DOCUMENn A
REQUIREMENTS 5
AiPLICAULE STANDARDS 5
DrESCRIPI IN GENERAL.
6 PH JYSICAL.
7 FIRMWARE 7
QUIRATIONAI, PERFORMANCE 7
ACCURACY 7
StRVJCE LIII-I 7
I'OWER REQUIREMENTS 7
POSITIVE LOGIC CIRCUIT SUPPLY VOLTAGE 8
POSITIVE ANALOG CIRCUIT StJPPI.Y VOLTAGE 8
NEGATIVE ANALOG CIRCUIT SUPPLY VOLTAGE SAMIPIUTIME 8
ANTI-ALIASING FllTFR 8
RER'S9'ONSE TIME 9
AUT1OMATIC TRIP REFERENCE SELECTION 9
RF.CIRCULATION DRIVE CORE FLOW TRANSFER FUNCTION 9
POWER BzASED ADJUSTMENT ADDER 9
Al,lTl.RNATF. TRIP REFERENCE SELECTION 9
PROCIRAMM L OAT]
-O N
TRIP REFERENCE ARRAYS 10 TRIP REFERENCE FUNCTION SPECIFICATIONS 10 SIGNAL VALIDATION 10 AUTOMATIC SETI)OWN SETP'OINT 10.6 VYC-0693A Rev. 2 Page 2 of 23
.7 1 -
.:. 1.
a GE Nuclear Energyr TAIIIQE QF CONTENTS 24A.5215 SH NO. 3 6&
SPECIFIED 1/O CHARACTERISTICS I
6.1 INPUTS It 6.1.1 ELECTRICALINPUTS 11 6.1.2 LOCAL OPERATOR'S INPUTS 1l 6.2 OUTPUTS 12 6.2.1 ELECTRICAL OUTPUTS 12 6.2.1.1 FACTORY TEST OUTPUT 12 6.2.2 LOCAL. OPERATOR'S INI)CATION 12 I.
IFIED MECJHANICAL CHARACTERISTIC 1
7.1 METAL CASE: (RW' 3 - early BWR 4) 13 7.2 SMAL.LI-44P N: (latc BWR 4 - BWR 6) 14 L
I3LOCK DIAGRAM I5 SP'EC'IF F.D OPERATING ENVIRONMENT 16 9.1 TlMPFRATURE AND HUMIDITY LIMITATIONS 16 9.2 ELECTROMAGNETIC INTERFERENCE 17 9.2.1 SUSCEPTIIRILITY 17 9.2.2 EMISSIONS 17 9.3 AMBIENT PRESSURE LIMITATIONS 17 9.4 RADIATION EXPOSURE lIMITATIONS is 9.4.1 DOSERATlE 18 9.4.2 TOTAL DOSE 18 9.5 SEISMIC DISTURBANCE LIMITATIONS iS I (L SAELTY PRF.CAUTIONS p;
10.1 PERSONNEL SAFEITY 19 I +/-
TEST 8E:QV!RF,9EMlN 12 11.1 AUTOMATIC SEIF-TEST.
19 11.2 MANUAL SURV'EIl.LANC'rE/CALIBRATION 19.6 VYC-0693A Rev. 2 Page 3 of 23
I24A25159 SH NO. 4 GE Nuclear Energ' Rev 2 OF I 9
- 1.
DOCMtENIF aT D.SCRTf'ION 1.1 DOCIIJIMENT PIRPOSE AND USE This Design Specification describes in quantitativc terms the characteristics of the FLOW CONTROL TRW RFEERENCE CARL).
Thcsc characteristics are grouped as follows:
Inptits to the card - their information content and clectrical characteristics
- Outpuls from tli card - their inrormation content and electrical characteristics The functions of the card - its information processing, transfer functions, and the intcrrelationships between thc inputs and the outputs The physical env'ironment under which the card will function properly The card's physical parameters - size, shape, type and placement of connectors Application information - potential safety hazards 1.2 D)OCUIMENT SCOPE This document describes the performance or the FLow CONTROL TRIP RIFF.KRENCfl CARD with respect to its use as a component part..6 VYC-0693A Rev. 2 Page 4 of 23
2 24AS21; Sli NO. S GE Nuclear Energy Rcv2 OF '9
- 2.
RFt ElATFI) IOCUME1FNTh 2.1 REQUIREMENTS a)
RWROG Hardware-Related Task Authorization for Stability LTS Enhanced Option 1A: 94.159.0,.S, 95.159.0..5..7. 96.159.0 b)
NUMAC Requirements Specification: 23A5082 c)
NULMAC Soflware Configuration Management PHan 23AS 161 (meets requirements of ANSIIIER 7-4.3.2-1982) d)
NEDO-32339-A Class 1. December 1996 - Licensing Topical Report, Reactor Stability Long-Ternm Solution Enhanced Option 1-A e)
)rNEDO-32339-P Supplement 2, Enhanced Option 1-A Solution Design, April 1995,
.vith errata 91I/95, 9115J95, 1/31196, 3/27/96, and 11/27/96 I)
NEDO-32339 Supplement S. Enhanced Option l-A Solution Ciosure, September 1996 2.2 a.
b.
C.
d.
e.
APPLICADIU&
STANDARDS MIL HlMIK-21717 Reliability Prediction of Electronic Equipmcnt COl 265A1 14R Specification for Printed Circuit Boards IEE-1P.323-1974 Qualifying Class lF Rquipment for Nuclear Power IE-FF' 344-1975 Seismic Qualification of Class lEEquipment for Nuclear Power GE I100 Series Electroniagnctic, Irterfercnce and Susceptibility.6 VYC-0693A Rev. 2 Page 5 of 23 II
- 1 0
GE Nuclear Encrgt,
- 3.
y)1SCRIPTION 3.1
- GFNERAI, j14AS215 St[ NO. 6 The FLOW CONTROL TRIP REFERENCE CARD is a Class IE component. This card provides all of the functions required by the original equipment and secrion 2.2 and uill not deter from the current mechanical parameters or electrical connections in order to implement new features, unless otherwise requested. Logic common, if not present at the connector, will be connected to present unused pins. In addition, this card provides input flow signal 'validation to ensure fail safe operation.
This card may be inserted and removed under power without danger and will automatically reinitiate after competion of an internal self-test (approx. 5 sec.).
This card provides for the following interface between the input flow and the output references:
Provides flow signal output following noise & EMI filtering.
Provides a Scram Trip Reference based on a derived core flow function residing in memory for the recirculation Two Loop Operation. Two Loop Setup Operation.
Singie Loop Operation and Single Loop Setup Operation settings.
- Provides A control rod block Trip Reference based on a derived core flow function residing in memory for the recirculation Two Loop Operation (TLO), TLO Setup.
Single Loop Operation (SLO), and SLO Setup settings with slope and offset
.adjustmeont.
DIP switches provide selection of alignment constants for the Recirculation Drivc/Core lFlow transfer function.
DIll switches provide selection of an alternate set of trip reference function tables.
- DIP switches provide constants for the Power Based Adjustment transfer function.
Provides a recorder output of the control rod block Trip Reference with offset adjustnent.
I.
Provides validation of incoming flow signal (out-of-range high or low).
- Provides an INOP output contact that is actuated if the card is not operable or fails self test.
- Status IXE:1) provides visual latchcel confirmation of status (INOP).
- Status pushbutton provides reset of status (INOP).
Setup LED provides visual Iatched confirmation of setpoint Setup condition.
l Setup pushbutton alternately changes between SET UP and NORMAL setpoint arrays.
Provides for autnmatic reset of trip reference setpoints to normal based on a preset control rod bloclz trip reference value.
Provides a factory test output..6 VYC-0693A Rev. 2 Page 6 of 23 i
..-1
',. -'.
0
- GE Nudclear Energi' 3.2
- PYSICAl,
'I24A5215 SH NO. 7 Rev 2 OF 19 The FLOW CONTROL TRIP REFERENCE CARD is designed in two different size configurations. The FLOW CONTROL TRIP REFERENCE CARD is designed as a pin-compatible replacement for the existing PCTR Card in the APRM page. See Secton 7, Specified Mechanical Characteristics, for details.
3.3 FIRMINVARE This card contains firnnware for providing recirculation drive flow based selpoints associated with tile input-output interface.
- 4.
01FRATIONAl, PERFORMk1AN(CE 4,1 ACCURACY The r4LoW C(ONTROL TRIP REFERENCE CAR) will be designed to operate with an accracy of il.0 % over 36 months due to environmental, initial calibration and accuracy drift; 4.2 SERVICE LlFE The target service life for the FLOW CONTROL TRIP REFERENCE CARD is to operate continuously (within the specificd environmental limits, and allowing for rcplacernent of failcd components) for at least 40 years.
4.3 P'OWER IIFQIJIREMENTS This card will require I 15 VDC. -15 VDC. Analog Common & Digital Common for nornal operation. Digital Common will bc-required for EMI grounding purposes even though +5 VDl is not being used. The +5 VI)C will be generated on board from the + 1 5 VDC. The card connections will be compatible with the present pin-out configuration in thle APRM page connectur with the D)igital Common it connected. taken from the present logic common pin-out. If not conneced. then the logic common will be connected to present unused pins. The power requirements of this card will not adversely affect the APRM pwages power supply capability. Application and tolerances arc listed below..6 VYC-0693A Rev. 2 Page 7 of 23
Ri 7~~~~~~~~~~~~~~~24A5215 S1 NC). B GE ENucearEnjergy Rev 2 OF 19 4.3.1 POSITIYV LFOI iC CIRCUIT SUPPLV VOLTAGE (Generated on board from the -ItI5 VDC & Analog Common)
Application:
+5 Volt DC power used for internal lokc.
PARAME1ER SYMBOL MIN,
OPERATING OPERATING MAX UNITS
_______ LIMIT MAX M
_AX LIMIT Logic Supply Voc 0
4.5 5 5 5.5 VDC vollagc I_ ________
Logic Supply Voer 0
l 0.00 0.05 0.10 VRMS Ripple Voltlgc I____
II II_
4.3.2 POSITI!F, ANAIO('V CIRClJT St)PPIV VOLTAGF, Application +1 5 Volt DC power used for intcrnal ADCs, DAC's, amplifiers, etc.
PARAMETER SYM11L MIN OPERATING( OPERATING MAX UNITS LIMIT MAX MAX LIMIT Positivc Analog Vdd 0
14 16 1 6 VDC Supplv'oitagc Analog Supply Vddr 0
0.0o 0.01 j
.10 VMIS Ripple V'oltagc a
4.3.3 NEGATIVE ANALOG CIRCUIT SUPPL1 1'VOLTAGE Application:
-I1 Volt DC power uscd for internal ADC's, DAC's, amplifiers. etc.,
PARAMIrI'l;R SYMBOL MIN OPERATING OPERATING MAX UNIT7S
_LIMl1 MAX MAX LIMIT Negalive Analog Vss 0
-14
-16
-16 VDC S~upply Voltag e Analog Supply Vssr 0
0.00 0.01 0.10 VRMIS Ripplce Voltagc 4.4 SAMiPLE TIME The minimnuni time for sampling the flow input is 4 trip references is 28 nisec.
4.5 ANTI-ALIASING FII.T'ER The flow input anli-aliasing filler is scl at <201 Iz.
msec. The interval for updating the.6 VYC-0693A Rev. 2 Page 8 of 23
0 GE NruclearEnergy 4.6 RESIiONSEOT1NiF.
24AS15 SH NO. 9 Rev 2 OF 19 The maximum response time from input to output is 250 msec, 4.7 AUTOMATIC TRIP RFEURENCE SELECTION Trip reference functions are identified as setup and non-setup.
Whcn the control rod block selpoint exceeds a specified Automatic Scidown Setpoint, the trip reference functions are laken fron the non-setup tables, even if the setup selection has been made with the pushbutton sciector switch provided.
4.8 RFCrlRCULATION DRIVF. CORU FLOW TRANSFER FUNCTION Thc Recirculation l)rive/Core Flow transfer function wilt be adjustable to accommodate variations in this relationship. Two four-position DIP switches. operating in binary, arc used to select the appropriate alignment constants used in the transfer fanction. There will be a maximum of sixcecn chniccs for each of two alignment constants.
4.9 POWER4 BIASED AI)Jl1ST IENT A IMER The Power Based Adjustmcnt Adder provides an adjustment to thc output trip referencc scipoints. Thc samc adjustment adder is applied to both output trip references. Three of four switches arc used to provide eight choices for the adjustmcnt adder. Selcction of an adjustment adder ol zero results in no adjustment to the output trip references.
4.10 AL F:RNA'F T)l' REFI RENCE SELECTION Two conmplete scts of trip reference arrays are stored in memory. One switch of a four I1)V switch is used to select betwcen SETI (primary) and SET2 (altcrnatc)..6 VYC-0693A Rev. 2 Page 9 of 23
'.. k.,:., %,
--'
GR2 NAclear Energr S.
PS'IOGRA j 11NG INFOR(IATION s-1 1 TIP REFERENCE ARRAYS I24AS21S SHiNO.IO0 Rev 2 OF 19 At least 512 points (9-bit resolution) composing an array of each specified function (plant specific) uill be held in meniory. There will be up to a maximum of two sets, each consisting of eight (8) transfer functions.
Thesw functions are programmed into the memory component (inslalled on a EPROM) and can only be changed by component replacement.
5.2 TRIP REFERENCE FLINU'Y]ON SPECIFICATIONS The 512 point, (9-bit resolution) tables for each plant-specific trir reference function are specifies as:
- 1.
Single Loop Operation (SIO) NORMAL
- a. Sciamt
- b. Control rod block
- 2.
Two Loop Operation (TLO) NORMAL
- a. Scram
- b. Control rod block
- 3.
SLO Setup
- a. Sclam l
.t Control rod block
- 4.
TLO Setup
- a. Scram l
- h. Control rod block-These functions are progranimcd into the memory component (installed on a socket) and can only be changed by component replacement.
5.3 SIGNAL V'ALIA)ATJ0N The input recirculation drive flow will be vaidated to confirm that it is not out-of-range N(high >130 or low < -1S%).
IS5.4 AUTOMATIC SUTIJOWN SETPO1NT The Automatic Setdown Sctpoint is provided fir each pair of TLO (NORMAlJSETUP) and SLO (NORMA14SE:TUI') trip references. Tblic setpoint is programmed into EPROM..6 VYC-0693A Rev. 2 Page 10 of 23
. I I
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.1 a
GE Nuclear Energr
- 6.
_SPCI J'ED 110/ CIARACTrFERISTIcS 6.1 INKITS 6.1.1 LFla.('5ICAL INPFlTS I24A.5215 S11 NO. II Rev 2 OF 19
- 1.
Input Flow: 0 to l0 volts DC (relates to 0 to 125 % Dow)
Input impedance, Z., is 100 1C.
6.1.2 ILOCAl. OPERATOR'S INPUTS I 1. Recirc TJ.OISLO (switch allows choice)
- 2.
'INOP Reset (pushbutton)
- 3.
Recorder offset adjustment
- 4.
SETUP (pushbutton - altcrnate action)
- 5.
DlIP switches for drive flow aligrnment constants, power-based adjustment' constant, and Alternate trip reference selection a)
Manual Reset:
- 1.
Pushbutton
- 2.
Change in switch status (TLO/SLO) b)
Automatic Reset:
1 I From SETUP to NORMAL of a specified rod block setpoint corresponding to 5m/. recirculation drive flow above A'not o
- 2.
flower failure
- 3.
E1xiting INOP I
Upon reset of SETUP, current recirculation TLO/SLO switch setting will determine the applicable arrays applied.6 VYC-0693A Rev. 2 Page 11 of 23 j;.
a GE Nuclear Enervg 6.2 OulTPuiTS; l4AS2l5 SH NO. 12 Rev2 0F 19 I
6.2.1 EIVC'l'RI CAI, OUTPUTIS
- 1.
Output Recire Drive Flow: 0 to 4 10 volts DC
- 2.
Scram Trip Rererence: 0 to -IO volts DC
- 3.
Control rod block Trip
Reference:
0 0o -10 volts DC
- 4.
Control rod block Recorder, 0 to I volt lC with an offset of* 100 mV
- 5.
INOP: normally open (N.O.) contact, non-latched. Open, non-energUize, on INOP.
A'nom - the highest flow in operating domain associated with Restricted Region of recirculation drive flow transfer function (plant specific),
A critical ScIf-test fault 2erocs the Scram and Rod Block trip references (forces APRM1 scram and rod block) 6.2.1.1 FACTO0R' TESt1' OITPllT Single digital fiber-optic communication output, transmitter only, at a data rate of 19.2 ki Iz. IlewIctI-Paackard I IFBR-1414T fiber-oplic. transmitter.
6.2.2 1..OCAI. OPERATOR'S INDI)CATION I.
A non-latched two color LF.D) to indicate trip reference setlup condition. 1l2[) color indications arc:
GREEN:
Normal boundaries Y Lt.WOW:
Setup boundaries
- YEULOW (slow blink):
Flow validation inhibit (normal boundaries)
- YELLOW (doublc blink):
Flow validation inhibit (setup boundaries)
- 2.
A latched two color L.ED) to indicate card INO1P Status. LED color indications are:
a 2
- GORFN:
Operating Normal REED (steady):
Current INOP (self-test fatllt)
- RED (slow blink):
Previous INOP, cleared, but not reset (Initial power-up or previous self-test fault)
- RED) (fast blink):
Flow validation fault
- RE!DIGRF.'.N (fast blink):
Previous Flow validation fault
- OFF:
No power lo card Previous INOP or flow validation fault is reported and reset in thc order ofoccurrenc.
Nonrlatcdcd Iran INOP occurs (RED) and then clears automalically, this indication will be latched RED-blinking) which indicates norinal opcration but with a past INOP.
his can bo tesct by thc WInQ Rcset pushbutton..6 VYC-0693A Rev. 2 Page 12 of 23
24A 21F GE Nuclear Energy e
E. ""~1W)M~AIALCI~~E1TG 7.1 MELTAL, CASE: (11WR 3 - txrly IRWR 4)
SH NO. '31 OF 19 0D920 ;00 1,
-. 2D.
4.250
'03 WIOP LEt)
INOP RESET PlISHBlrYTON NoRmAusInTr LED NORMAUSMP PUSHUUtIrON (ALTENATES IhEarEEN NORMAL & SUIP ARRAYS)
TLO/SLO TOGGLE SyTCH (LOWCNG LVER)
RECORDSER OFMET ADJVSTNMIP EMI sill)ltn.6 VYC-0693A Rev. 2 Page 13 of 23 i
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6 v
0 *.
- S 3- @
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24A5215 SH NO. 14 f6GE Nuclear Fnergp Rev 2 OF 19 7.2 SMALL-44 VJN: (1a(lt BWR 4 - BWR 6)
St} 6 0
4.484+ °° ECTR me VA S E I I
- OPMI~
11N0P L.ET)
PUSHBUrlON 14ORMAUSEIIJT IXK NORMAIJSJ3TIJ 1USHOIU1?TN (ALTI3RNAU1S DEIMWEiP NORMAL &- SETUP ARRAYS)
-TLIO!SLI TODGGLE SWITCH (LOCKING LEVrER)
P.P~CORPEFR OWFSEr ADJUSTMENT
]
O CARD) VECTOR.6 VYC-0693A Rev.2 Page 14 of 23
- -, : '- :,,,,, ",..,, ".......
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GE Nuclear Energy A.BOKDITl~
j24ASMt SliN.IS'
-- DIOCK DIAGRAM -.6 VYC-0693A Rev. 2 Page 15 of 23
i 0
GE Nuclear Energy 9,
NPECIFIF,) OPERATING FNVIRONMIENT I24A5215 S1IINO. 16 Rev 2 OF 19 The FLOW C:ONTROL TRIP REFERENCE, CARD will not adversely affect or be affected by the operation of any other components or equipment operating within the same environmenl.
9.1 TEMIPERATUIRE ANI) IfIIMIDITY LIMITATIONS Tbis card will perform all specified functions correctly when operated within the specified temperature range illustrated in Figure 9-1 and the, specified relative humidity range illustrated in Figure 9-2 (Applicable Standard 2.2 c)..
Figure 9-1, Temperature Limitations (0 C <Ta <70 -C) 4 Temperature (C)
-5 0
70 85
{..
(
<Range of correct operation l-
-l Temperatu re Temperature helow which above which card may lie c\\
card may be damaged Ranges in which card may or may not operate correctly damaged I
Figure 9-2, Humidify Iimitaintns (20% < Relunmiditfy 90%)
4-ReAtiV lumidity (%)
ICI 20 90 4--*
4-Rangc of correct opcration 95 Humidity above which card may be damaged llIumidity below w h;ch card may be damaged Ranges in which card may or may not operate correctly.6 VYC-0693A Rev. 2 Page 16 of23
0
.:
- -, I..:.
a a
GE Nuclear Energr 9.2 ElElcTi'6OmA(GNETIC INTERFERENCE I24A5215 SH NO. 17 Rev 2 OF 19 I
The FCTR card is qualified for electromagnetic compatibility (EMC) by type testing and analysis. The IMC testing performed eliminates the need for utilities to perforni in-plant electromagnetic environment surveys in accordancc wilh EPRI guidelines (Reference 7).
9.2.1 SUSCEPTI BLITY Thc FCTR card is mounted within the existing cabincts of the NMS and derives power from NMS power supplies. Therefore, power and signal conducted inoise immunity arc not significantly affected by the replacement or addition of these cards. The existing power.supply immunity to conducted noise and power surges remains unchanged. Thus, the only EM( susceptibility tests performed on the FCTR card are:
- Electrostatic Discharge (ESD) - 1WC-801-2 Simulated lightning Strike Conducted Immunity Test (using pulse measurements provided by Untergy Opcrations. Inc.)
- Radiaied Electric Field Test (demonstrate that adequatc margins exist for proper FCTR operation when installed into an existing NMS page under electromagnetic near-ficld emissions from adjacent cards) 9.2.2 EMISSIONS The radiated emissions test confirm that the FCTll card has sufficiently low near-field emissions as to not affect the existing NMS pages. Similar to 9.2.1 above, there is no accepted test level for this application. therefore, it must be established as follows:
- Radialed Tilcetric Field 1est (demonstrate that adequate margins exist for adjacent cards to operate without electromagnetic near-field interference from the FCTR card installed into an existing WMS page) 9.3 AMIBIENT I'RSSUIRE LIM NI TATIONS This card wVill perform to specification for any absolute pressure in the range of 13 psi to 16 psi..6 VYC-0693A Rev. 2 Page 17 of 23 r
~~~~~
~.:
I 0
GE Nuclear E7nergy 9.4 RiUIlATION EX'POSURE LIMITATIONS 9.4.1 DOSE RATE 24A$215 SH NO. 18 IRev 2
OF 19 I The card will perform to specification over its design life in a gamma field of 3 mRlhr or less.
9.4.2 TOTAL DOSE IThis card will perform to specification ovcr its service life for a total integrated gamma dose of ixlIO Rads.
9.5 SEISMIC D)ISTURBANCE LIMITATIONS The FCTR Card will be qualified based on applicable areas described in IEEE 344-1975 (Applicable Standard 2.2-d). The APRM page cnvironment will be established through analysis, and it will be determined that the addition of the FCTR into the APRM page will not significantly degradc cxisting system performance when subjected to seismic events.
Ioctimcntalion will show that qualified tests and levels (or analysis) cover APRM page environnmeint for this application.
- 10.
SAFEWT PRECAUTIONS 10.1 PERSONNEL SArETV sI.P..C.TR1CIAL No voltage greater than 28 volts + tolerance (for Status INOP contact) is prcsent on this card. This card may be removed under power without danger.
- ME1CiANICAL No moving parts will cause personal danger.
- 3iPMAt, None: no high temperaturcs are present on this card.
R~lQI O(;lClD~
Nonc-no radioactive materials are incorporated into this card.
None - no corrosive or toxic substances are incorporated into this card,.6 VYC-0693A Rev. 2 Page 18 of 23
- .1
.. -.. -.......
. 4 24AS215 SH NO. 19 GE Nuclear Erergy Rev 2 OF 19
- 11.
TEST REONRFEMNTS 11.1 AUTOMATIC SEILF-TEST An automatic self-test feature will be used on this card. This feature will automatically test at least once per minute. If an error or occurs an INOP will be initiated.
A known internal reference is used, with its outputs, Scram & Control Rod Block Trip Refercnces, internally checked against the expected outputs to verily correct system operation.
- The voltages; power, logic. & reference are monitored to verify they are within tolerancc.
The CPU monitors powcr failures to ensure fail-safe operation.
Watch dog supervisory circuit is used to ensure correct soflware operation, cycle timing and logic, power failure.
Voltages from the PAC are fed back through the ADC to ensure correct hardware conversion operation
- ScIf-lest diagnostic testing ensures correct DAC operation and sets its output to "O" if the signal is not updated (frozen).
11.2 IhIANIIAL SUKRVfIA.,ANCFJ(:AllBRATION Manual surveillance will be required at least every 36 months.
This wilt consist of inputting an existing exictnally known simulated recirc dr;ve flow signal to verify correct system operation.
Test points are provided for measuring the power voltages (+5 VDC, +I5 V'DC & -15 VDC) and the system clock (16 h1lb?)..6 VYC-0693A Rev. 2 Page 19 of 23
I a GE Nuclear Energy PROPRJUFTARY l
9.2.2 EMISSIONS Sheet 20 of 22 Rev 2 24A5215 Again, there is no accepted test level for this application, therefore, it must be established. The test levels for ernissions for the APRM page will be established by the following:
- Qualifry new FCTR CARlD per emission test at derived susceptibility qualification level above
- Document how qualified tests and levels cover APRM page environment for this application 9.3 AMBIENT PRESSURE LIMITATIONS This card will perform to specification for any absolute pressure in the range of 13 psi to 16 psi.
9.4 RADIATION EXPOSURE LIMITATIONS 9.4.1 DOSE HAT1F This card will perform to spccification over its service life in a gamma field of lxIOF-5 rads Isec or less.
9.4.2 TOTA1 1) OSF.
This card will perforni to specification over its service life for a total integrated gSamnia dose of lx01E4 rads.
9.6 SEISMIC DISTURBANCE LIMITATIONS The FCTR Card will be qualified based on applicable areas described in IEEE 344-1975 (Applicable Standard 2.2 d). The APRM page environment will be established through analysis, and it will be determined that the addition of the new FCTR into the ATRM page will not significanIly degrade existing system performance when subjected to seismic events. 1)ocumentation will show that qualified tests and levels (or analysis) cover APRM page environment for this application..6 VYC-0693A Rev. 2 Page 20 of 23
I at GE Nuclear Energy PROPRIfTAIRY
- 10. SAFETY PRECAUTIONS 10.1 PERSONNEL SAFETY Sheet 21 of 22 Rev 1 24A521 S I
IEhECTRICAl, No voltage greater than 28 volts + tolerance (for Status/INOP contact) is present on this card. This card may be removed under power without danger.
a MECHANICAI, No moving parts will cause personal danger.
RADIOLOGICAL None; no radioactive materials are incorporaled into this card.
- CHEM ICA]
None; no corrosive or toxic substances are incorporated into this card.
- 11. TEST REQUIREMENTS 11.1 AUTOMATIC SELF-TEST An automatic selr-test feature will be used on this card. This feature will automatically l test at lea-t once per minute. If an error occurs an INOP will be initiated.
l
- A known internal reference is used, with its outputs, Scram & Alarm Refercnocs, internally checked against the expccted outputs to verify correct system operation.
- The voltages; power, logic & rcfcrence are monitored to verify they are within tolerance.
l
- The CPU monitors power failures to ernsure fail-safe operation.
- Watch dog supervisory circuit is used to ensure correct software operation, cycle timing and logic power failure..6 VYC-0693A Rev. 2 Page 21 of 23 t.-I
I 0
GE Nuclear Energy PROPRIETARY Sheet 22 of 22 Rev I 24AS215 11.1 AUTOMATIC SELF-TEST (con't)
Voltages from the DAC are fedback through the ADC to ensure correct hardware conversion operation.
I Seir-test diagnostic testing ensures correct DAC operation and sets its output to 'O" if the signal is not updated (frozen).
11.2 MANUAL SURVEILLANCE I CALIBRATION Manual surveillance will be required at least every 36 months. This will consist of inputting an existing externally known simulated recirc drive flow signal to verify correct system operation.
Test points are provided for measuring the power voltages (+5 VDC, +15 VDC &
I -15 C) and the system clock (16 FMz).
S.6 VYC-0693A Rev. 2 Page 22 of 23
.6 VYC-0693A Re,. 2 Page 23 of 23 FCTR Accuracy & Drift The 14BC6411 Digital FCTR card (schematic 105E1374) has an analog front-end consisting of an analog mux (233A3701 based on the ADG516A), quad op amp (233A3709 based on the AD713),
and 10-bit AID (2-bit precision within the 233A3692P001 microcontroller based on the 80C517A).
The major contributor to accuracy over the calibrated range of Drive Flow (0 to I OV) is the resolution of the microcontroller's AID. This provides 844 counts or +/- 0.47% resolution with the 2-bit precision. The analog mux, input op amps, and precision feedback resistors provide +/- 0.12%
accuracy. This yields (SRSS) an accuracy of +/- 0.5% (all analog front-end components). The Digital FCTR performs all other operations digitally and therefore has no accuracy errors after the AID and DIA conversions. The analog outputs of the FCTR are calibrated during the system-level calibration which includes the Quad Trip card and other components of the flow loop. In addition, the accuracy error introduced from the output 12-bit DIA converters is negligible when compared with other FCTR components. Therefore, the Digital FCTR card provides an accuracy of + 0.5%.
The temperature drift effects are not shown above because the FCTR application provides stable temperature operation. However, the combined effects of accuracy and temperature drift (over the rated 0-70 0C range of operation) are better than i 1%.
Steve Sawyer Leonid Sheikman GE Electronics & Technology GE Electronics. Technology 1112412003 1112412003
.7 Vermont Yankee Setpoint Control Program Interdepartmental Review of Calculation VYC-0693A Revision 2 VERMONT YANKEE SETPOINT CONTROL PROGRAM INTERDEPARTMENTAL REVIEW OF CALCULATION:
VYC-693A Revision 2 has been prepared and independently reviewed. The Departments impacted by this calculation are requested to review the results of this calculation, concur with the results and/or recommendations, and document the department's acceptance prior to the calculation being approved.
- 1. Summarv This calculation evaluates the uncertainty & setpoint for the APRM/RBM Neutron Monitoring Trip Loops.
- 2. Calculation Open Items:
AP-0028 to be Assigned 2.1. None NA
- 3.
Department Review - contact the Setpoint Program Manager (Joe Garozzo) if not in agreement with the conclusions/statements.
3.1. Vermont Yankee I&C 3.1.a.
Procedure OP-4308 will require the following:
- 1. Add the following in the procedure discussion:
Allowable Values:
AJlowable Values Output Instrument
% RxP mV Curve % RxP kPRM A, B, C, D, E, F Scram Trips ARTS/M1ELLLA Flow Biased APRM High Flux Scram (Two Loop Ops)
Core Flow 0 to < 3 1.1 % (point based on 25% flow) 71.1 5.6880
<0.4W+61.10%
Core Flow31.1 to <54.0 % (point based on 50 % flow) 97.31 7.7848
< 1.28W+33.31%
Core Flow 54.0 to <75 % (point based on 70% flow) 113.48 9.0784
< 0.66W+67.28%
Core Flow > 75%
116.96 9.3568 N/A AI'RM High Flux Scram (Single Loop Ops)
Core Flow 0 to <39.1 % (point based on 25% flow) 68.09 5.4472
< 0.4W+58.09%
Core Flow 39.1 to <61.9 % (point based on 50% flow) 87.56 7.0048
< 1.28W+23.56%
Core Flow 61.9 to <83.0 % (point based on 70% flow) 108.3 8.6640
< 0.66W+62.10%
Core Flow> 83.0%
116.96 9.3568 N/A EPU Flow Biased APRM High Flux Scram (Two Loop Ops)
Core Flow 0 to <30.9 % (point based on 25% flow) 58.7 4.6960
< 0.33W+50.45%
Core Flow 30.9 to <66.7 % (point based on 50% flow) 80.73 6.4584
< 1.07W+27.23%
Core Flow 66.7 to <99.0 % (point based on 75% flow) 103.59 8.2872
< 0.55W+62.34%
Core Flow > 99.0%
116.96 9.3568 N/A APRMI H1igh Flux Scram (Single Loop Ops)
Core Flow 0 to <39.1 % (point based on 25% flow) 56.25 4.5000
< 0.33W+48.00%
Core Flow 39.1 to <61.7 % (point based on 50% flow) 72.51 5.8008
< 1.07WV+19.01%
Core Flow 61.7 to <119.4 % (point based on 75% flow) 92.47 7.3976
< 0.55W+51.22%
Core Flow> 119.4%
116.96 9.3568 N/A Page of 10
I
'.Z.7 Vermont Yankee Setpoint Control Program Interdepartmental Review of Calculation VYC-0693A Revision 2
- a.
As-Left and As-Found values:
Calibration Points, As-Left and As-Found Output Instrument etting s-Left As-Left VYDC VYDC
'dc Min Max As-Found As-Found Min Max XPRM A, B, C, D, E, F Scram Trips ARTS/IMELLLA Flow Biased
____l
%.PRM High Flux Scram (Two Loop Ops) l l
-ore Flow 0 to < 31.1 % (point based on 25% flow) 5.5880 5.538 5.638 5.488 5.688 Core Flow 31.1 to <54.0 % (point based on 50 % flow) 7.6848 7.635 7.735 7.585 7.785 Core Flow 54.0 to <75 % (point based on 70% flow) 8.9784 8.928 9.028 8.8781 9.078 Core Flow> 75%
9.2568 9.207 9.307 9.157 9.357 APRM High Flux Scram (Single Loop Ops)
Core Flow 0 to <39.1 % (point based on 25% flow) 5.3472 5.297 5.397 5.247 5.447 Core Flow 39.1 to <61.9 % (point based on 50% flow) 6.9048 6.855 6.955 6.805 7.005 Core Flow 61.9 to <83.0 % (point based on 70% flow) 8.564C 8.514 8.614 8.464 8.664 Core Flow > 83.0%
9.2568 9.207 9.307 9.157 9.357 EPU Flow Biased kPRM High Flux Scram (Two Loop Ops)
)w 0 to <30.9 % (point based on 25% flow) 4.5960 4.546 4.646 4.496 4.696 Core Flow 30.9 to <66.7 % (point based on 50% flow) 6.3584 6.308 6.408 6.258 6.458 Core Flow 66.7 to <99.0 % (point based on 75% flow) 8.1872 8.137 8.237 8.087 8.287 Core Flow> 99.0%
9.2568 9.207 9.307 9.157 9.357 APRIMI High Flux Scram (Single Loop Ops)
Core Flow 0 to <39.1 % (point based on 25% flow) 4.4000 4.350 4.450 4.3001 4.50C Core Flow 39.1 to <61.7 % (point based on 50% flow) 5.700 5.651 5.751 5.601 5.801 Core Flow 61.7 to <119.4 % (point based on 75%flow) 7.297 7.248 7.348 7.198 7.39E Core Flow> 11 9A%
9.2568 9.20 9.307 9.157 9.357 Note: Adjustment of the As Left Calibration Tolerance in the conservative direction is acceptable
- c.
Revise Head to reflect: NA
- d.
Insert the following M&TE requirements:
DMM's with total device error of better than +/-0.05% CS (10 VDC Range)
HP 3466A or the HP 34401A are acceptable devices to support this accuracy Requirements. VYC-1758...
- 2. In the body of the procedure and the data sheet revise as follows:
Page 2-of ID
- .. : '. :..:. 7'..., -,.:., :, ::.:
- .:.7 Vermont Yankee Setpoint Control Program Interdepartmental Review of Calculation VYC-0693A Revision 2
- b. Trip Setpoint:
Limiting Setpoints and Calibration Cardinal Points Output Instrument
%RxP mV Curve %RxP APRM A, B, C, D, E, F Scram Trips ARTS/MiIELLLA Flow Biased APRM High Flux Scram (Two Loop Ops)
Core Flow 0 to <31.1 % (point based on 25% flow) 69.85 5.5880 < 0.4W+59.85%
Core Flow 31.1 to <54.0 % (point based on 50 % flow) 96.0 7.6848< 1.28W+32.06%
Core Flow 54.0 to <75 % (point based on 70% flow) 112.2 8.9784 < 0.66W+66.03%
Core Flow > 75%
115.71 9.2568 N/A APRM High Flux Scram (Single Loop Ops)
Core Flow 0 to <39.1 % (point based on 25% flow) 66.84 5.3472 OAW+56.84%
Core Flow 39.1 to <61.9 % (point based on 50% flow) 86.31 6.9048< 1.28W+22.31%
Core Flow 61.9 to <83.0 % (point based on 70% flow) 107.05 8.5640 < 0.66NV+60.85%
Core Flow> 83.0%
115.71 9.2568 N/A EPU Flow Biased XPRM High Flux Scram (Two Loop Ops)
Core Flow 0 to <30.9 % (paint based on 25% flow) 57.45 4.5960 < 0.33W+49.20%
Core Flow 30.9 to <66.7 % (point based on 50% flow) 79.4 6.3584 < 1.07W+25.98%
Core Flow 66.7 to <99.0% (point based on 75% flow) 102.3 8.1872 <0.55W+61.09%
Core Flow> 99.0%
115.71 9.2568N/A.
APRMI 11igh Flux Scram (Single Loop Ops)
Core Flow 0 to <39.1 % (point based on 25% flow) 55 4.4000 < 0.33W+46.75%
Core Flow 39.1 to <61.7 % (point based on 50% flow) 71.26 5.7008 1.07W+17.76%
Core Flow 61.7 to <119.4 % (point based on 75% flow) 91.22 7.2976 < 0.55W+49.97%
Core Flow> 119.4%
115.71 9.2568 I/A
- c. Revise calibration data to reflect head correction of: NA
- d. Insert a 9-point calibration for all analog instruments: NAforNeutron Monitoring 3.1.b. The following comments/recommendations apply.
- l.
Change Setpoints as identified above, Sign & Date Concur 7
P I/,
b ?
Vermont Yankee I&C Representative
)k '6sct I9A r4 fey As e a vi4I 9Of f/2c-//
f~cr1</#^i/7 rL Page.3.. Of 10D
.7 Vermont Yankee Selpoint Control Program Interdepartmental Review of Calculation VYC-0693A. Revision 2 VERMONT YANKEE SETPOINT CONTROL PROGRAM INTERDEPARTMENTAL REVIEW OF CALCULATION:
VYC-693A Revision 2 has been prepared and independently reviewed. The Departments impacted by this calculation are requested to review the results of this calculation, concur with the results and/or recomnmendations, and document the department's acceptance prior to the calculation being approved.
- 1. Summary:
This calculation evaluates the uncertainty & setpoint for the APRMRBM Neutron Monitoring Trip Loops.
- 2.
Calculation Open Items:
AP-0028 to be Assigned 2.1. None NA
- 3.
Department Review - contact the Setpoint Program Manager (Joe Garozzo) if not in agreement with the conclusions/statements.
3.2. Vermont Yankee Reactor Engineering Concur/
0! veel'/ st-0 3.2.a.
None
- 1. Improved Technical Specifications
\\9. (e 0 -c)C Analytical Limits APRM Hig-h Flux Scram (Two Loot) Oss)
Core Flow 0 to <31.1 %
< 0.4W+64.4%
Core Flow 31.1 to < 54.0 %
< 1.28W+37.0%
Core Flow 54.0 to < 75 %
< 0.66W+70.S%
Core Flow > 75%
Maximum of 120%
APRM High Flux Scram (Single Loop Ops)
Core Flow 0 to <39.1 %
< 0.4W+61.2%
Core Flow 39.1 to c 61.9 %
< 1.28W+26.8%
Core Flow 61.9 to < 83.0 %
< 0.66W+65.2%
Core Flow >83%
Maximum of 120%
APRM High Flux Scram (Two Loop Ops)
Core Flow 0 to < 30.9 %
< 0.33W+53.7%
Core Flow 30.9 to < 66.7 %
< 1.07W+30.8%
Core Flow 66.7 to < 99 %
< 0.55W+65.5%
Core Flow > 99%
Maximum of 120%
APRM High Flux Scram (Single Loop Ops)
Core Flow 0 to < 39.1 %
< 0.33W+51.1%
Core Flow 39.1 to < 61.7 %
< 1.07W+22.2%
Core Flow 61.7 to < 119.4 %
< 0.55W+54.3%
Core Flow >119.4 Maximum of 120%
CcXNM-CZT P I hM % ~ C" to I-, %L %\\ tySs? ~1-rm 2zo'u TrO Lts-P.
OJOllc
((,'
Sign & Date Zap\\o I X 0k.
/12t-1-03 Vermont Yankee RV, Representative Page q of Io
- ...*..7 Vermont Yankee Setpoint Control Program Interdepartmental Review of Calculation VYC-0693A Revision 2 VERMONT YANKEE SETPOINT CONTROL PROGRAM INTERDEPARTMENTAL REVIEW OF CALCULATION:
VYC-693A Revision 2 has been prepared and independently reviewed. The Departments impacted by this calculation are requested to review the results of this calculation, concur with the results and/or recommendations, and document the department's acceptance prior to the calculation being approved.
- 1. Summarv.
This calculation evaluates the uncertainty & setpoint for the APRMIRBM Neutron Monitoring Trip Loops.
- 2.
Calculation Open Items:
AP-0028 to be Assianed NA 2.1. None
- 3.
Department Review - contact the Setpoint Program Manager (Joe Garozzo) if not in agreement with the conclusions/statements.
3.3. Vermont Yankee Operations 3.3.a.
Recahlbrate APRMs LPRM7 tc after seismic event Sign& Date v
Q.I 0F/0,/ja Vermont 'Yankee Operations Representative Concur Page A of 16
I
..7 Vermont Yankee Setpoint Control Program Interdepartmental Review of Calculation VYC-0693A Revision 2 VERMONT YANKEE SETPOINT CONTROL PROGRAM INTERDEPARTMENTAL REVIEW OF CALCULATION:
VYC-693A Revision 2 has been prepared and independently reviewed. The Departments impacted by this calculation are requested to review the results of this calculation, concur with the results and/or recommendations, and document the department's acceptance prior to the calculation being approved.
- 1. Summary.
This calculation evaluates the uncertainty & setpoint for the APRM/RBM Neutron Monitoring Trip Loops.
- 2.
Calculation Open Items:
AP-0028 to be Assigned 2.1. None NA
- 3.
Department Review - contact the Setpoint Program Manager (Joe Garozzo) if not in agreement with the conclusions/statements.
3.4. Vermont Yankee Systems Manager 3.4.a.
This analysis supports the design bases for the APRM!RBM Neutron Monitoring Trip System Sign&Date g
/ IZ t--Z0oe5 Ver nt Yankee System Engineering Representative Concur Comments Page 4
of IC)
.7 Vermont Yankee Setpoint Control Program Attachmnent 7.7 Vermont Yankee Setpoint Control Program Interdepartmental Review of Calculation VYC-0693A Revision 2 Comments:
None Page 7 of 1 6
Calculation Number: VYC-693A Revision Number: 2 CCN Number:N/A.,s pA68 s oF 1 DOCUMENTATION OF COMPUTER RESOURCE USE Izb/a1s3 CALCULATION NO.:
VYC-0693A REVISION NO.:
2 CCN No.:_N/A Computer Used (include manufacturer, CPU Type, and operating system version and level):
Dell Dimension 8200 Pentium 4 2.4 MHz; Windows XP Professional version 5.1 service pack I Microsoft Excel version 2002 Service Pack 2._
Computer Input Attached*? El Yes l No LocationlIdentifier.
Computer Output Attached*?
a Yes ED No Location/Identifier:
- Large volume input/output should be provided on CD. See Appendix E for format requirements.
List the computer codes used, and complete the following:
Approved per PP 7800 Appropriateness Verified Outstanding SPRs or Code Errorsi Code Name/Version and/or Script File Yes3_
No Yes No Yes2 No Calculation Detail and Charts
-t/
V (Attachments B1 through B4)
Microsoft Excel Version 2002 SR 2 I~~~
~~~~~~~~
I I:
- =
=
+/- =
I =
+/-
1 Software Problem Report (SPR), does not exist as a reporting method in PP 7800 and AP 6030. Contact the Code sponsor and review any outstanding SPRs or Code errors. [ER2000805]
2 If yes, fill out information below.
3 If yes, include the Code name on the Computer Code line of the title page, VYAPF 0017.01.
If a computer code was not verified in accordance with PP 7800 and AP 6030, or if there are outstanding SPRs, state below why it is appropriate.
Code Name/Script File Appropriateness Calculation Detail and Charts (Attachments Appropriateness was verified Bl through B4) Microsoft Excel Version through hand calculation 2002 SR 2 I.
_~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
I Calculation Number: VYC-693A Revision Number: 2 CCN Number: N/A Attachment 7./
_4-t14t/13 Page O,/ 1 of J1a VY CALCULATION REVIEW FORM Calculation Number:
VYC-693A Revision Number:
2 CCN Number:
N/A
Title:
APRM/RBM Neutron Monitoring-Trip Loops Reviewer Assigned:.
Kirk Melson Required Date:
El INTERDISCIPLINE REVIEW COMMENTS*
- 1. Excel Sheets - For TLU terms, the in E INDEPENDENT REVIEW RESOLUTION outs to the equation are
- 1. Made all innut terms in values which format to percent of span.
mixed with some being listed in tenns of a decimal value and labeled
--- I--
-- I-------
-
-I-
as percent of span, and others formatted as percent of span. Need to express input terms the same.
- 2. Excel Sheets - For TLU terms, the inputs to the equation are mixed, in that some terms are actually in percent of span. and others are in percent Reactor Power. Need input terms to be consistent.
- 3. Either remove flow noise as a PMA term or provide specific reference.
- 4. Calibration tolerances for Fixed Hi Scram and Reduced Fixed
- 2. Changed input values to percent of Reactor Power.
- 3. All uncertainties for flow are covered by VYC-690. Removed flow noise as an uncertainty parameter with PMA.
- 4. Corrected Cal Tolerances. based on OP-4308 and OP-43108.
Scram are different than for Flow Bias. Please revise, based on.
procedure.
- 5. Add input tables for Flow Values to Word Document.
- 5. Added Tables 8 and 9.
__ f e
A.
_/_
09/21/03 Reviewer Signature Date Method of Review:
0 Calculation/Analysis Review
le&: "/-..
)I.
L.k-"
/
09/21/03 Ca'lculation Preparer (Comments Resolved)
Date o Alternative Calculation
/_09/21/03_
ol Qualification Testing Reviewer Signature (Comments Resolved)
Date
- Comments shall be specific, not general. Do not list questions or suggestions unless suggesting wording to ensure the correct interpretation of issues.
Questions should be asked of the preparer directly.
VYAPF 0017.04 (Sample)
AP 0017 Rev. 8, LPC 2 Page 1 of 1
7 Calculation Number: VYC-693A Revision Number: 2 CCN Number: N/A Attachment 7./_
4Z1/63 Page _//
of _/J0 VY CALCULATION REVIEW FORM I Calculation Number:
VYC-693A Revision Number:
2 CCN Number:
N/A
Title:
APRM/RBM Neutron Monitoring Trip Loops Reviewer Assigned:
Kirk Melson Required Date:
El INTERDISCIPLINE REVIEW 0 INDEPENDENT REVIEW I COMMENTS*
- 6. Correct Flow Break Points for ARTSIMELLLA Single Loop Ops in Table 3 of Word document.
RESOLUTION
- 6. Corrected.
- 7. Corrected to 15.
- 8. Corrected cross references.
- 7. Chance AL from 13.5 to 15 in snreadsheets.
- 8. Correct cross references in Cells A25. A26. and El6 to reflect
- 9. LPRM drift term is extrapolated from 700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> to 1250 hours0.0145 days <br />0.347 hours <br />0.00207 weeks <br />4.75625e-4 months <br />.
which is no longer necessary. Please remove the extrapolation.
- 9. Extrapolation no longer performed. Drift for 700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> is used.
- 10. Moved these errors to non-tested uncertainties.
I1. Remove Scientific notations in the spreadsheets.
- 12. Remove all equations. values for other than Method 1.
I 1. Corrected formatting to decimal.
- 12. Removed.
- 13. Set equal to accuracv term.
- 13. Set Cal Effect equal to Accuracv. as it is larger than CT.
- 14. Biases are negative, and the negative error is to be used for AV.
- 14. Showed bias in negative uncert comp and changed to use neg for AV.
I 09/21/03 Reviewer Signature Date
' -
--s".
x A HO
{ /,;.
I
_09/21/03 Calculation Preparer (Comments Resolved)
.Y Date Method of Review:
El Calculation/Analysis Review o Alternative Calculation b9 7
/
09/21/03 o Qualification Testing Reviewer Signature (Comments Resolved)
Date
- Comments shall be specific, not general. Do not list questions or suggestions unless suggesting wording to ensure the correct interpretation of issues.
Questions should be asked of the preparer directly.
VYAPF 0017.04 (Sample)
AP 0017 Rev. 8 LPC 2 Page 1 of 1
Docket No. 50-271 BVY 03-115 Vermont Yankee Nuclear Power Station Technical Specification Proposed Change No. 257 Discussion of Changes for Revised Technical Specifications
BVY 03-115 I/Attachment 3 / Page I of 2 DISCUSSION OF CHANGES TO REVISED TECHNICAL SPECIFICATIONS TS 2.1.A.1.a (current page 6)
The heading of this section is changed from "APRM Flux Scram Trip Setting (Run Mode)" to "APRM Flux Scram Allowable Value (Run Mode)."
The Standard Technical Specifications nomenclature of "Allowable Value" is adopted for this trip function of the reactor protection system. This change is made to clarify that the specification is an Allowable Value that corresponds to the limiting value that the instrument may have for operability. This change is made to draw a distinction from other TS that may specify trip settings that differ from the definition of an Allowable Value. This change is acceptable because it represents the appropriate operability limitation for the parameter (Neutron Flux Trip Settings).
TS 2.1 A. l.a (current page 6)
The specification for this limiting safety system setting is changed from "When the mode switch is in the RUN position, the APRM flux scram trip setting shall be as shown on Figure 2.1.1 and shall be: S<0.66(W-AW)+54%" to "When the mode switch is in the RUN position, the APRM flux scram Allowable Value shall be:
Two loop operation:
S <
0.4 W +
61.10%
for 0%
<Wc<
31.1%
S<
1.28 W + 33.31%
for 31.1% <W<
54.0%
S <
0.66 W +
67.28%
for 54.0% < W <
75.0%
with a maximum of 117.0% power for W > 75.0%
Single loop operation:
S <
0.4 W +
58.09%
for 0%
<W<
39.1%
S<
1.28 W + 23.56%
for 39.1% <W<
61.9%
S'<
0.66 W +
62.10%
for 61.9%
< W <
83.0%
with a maximum of 117.0% power for W > 83.0%"
The change in the neutron flux trip setting algorithm is supported by the ARTS/MELLLA analysis provided as part of Proposed Change No. 257. The deletion of reference to TS Figure 2.1.1 is discussed below.
TS Figure 2.1.1 (current page 11)
TS Figure 2.1.1 does not provide any requirement not included in the stated algorithms for this function. Because the figure is redundant, it can be eliminated from TS without any change in technical requirements.
In addition, elimination of this figure is consistent with Standard Technical Specifications.
BVY 03-115 / Attachment 3 / Page 2 of 2 TS Table 3.1.1 (current page 21)
Trip function no. 4, "APRM (APRM A-F) High Flux (flow bias)," is changed consistent with the algorithm change described above for TS 2.1.A.1.a.
The algorithm specified in the "Trip Settings" column is changed to the Allowable Value algorithms resulting from the adoption of ARTS/MELLLA.
In addition, Footnote (4) to Table 3.1.1 is changed to add a clarifying statement: "The specified APRM High Flux scram (flow bias) Trip Setting is an Allowable Value, which is the limiting value that the trip setpoint may have when tested periodically. The actual scram trip setting is conservatively set in relation to the Allowable Value." The change is made to emphasize that the specification is an Allowable Value that corresponds to the limiting value trip setpoint that the instrument have for operability.
This change is made to draw a distinction from other TS that may specify trip settings that differ from the definition of an Allowable Value.
In addition, Footnote (4) is also modified to eliminate the statement:
"AW is the difference between the two loop and single loop drive flow at the same core flow. This difference must be accounted for during single loop operation. AW = 0 for two recirculation loop operation." This statement can be eliminated because separate algorithm specifications are now provided for two loop and single recirculation loop operation, and the term "AW" has been eliminated from TS.
Bases Changes The TS Bases provide explanation and rationale for associated TS requirements, and in some cases, how they are to be implemented. Associated changes to the TS Bases are being made to conform to the changed TS and to add clarity to existing requirements. Bases do not establish actual requirements, and as such do not change technical requirements of the TS. The Bases changes are therefore acceptable, since they administratively document the reasons and provide additional understanding for the associated TS requirements.
Docket No. 50-271 BVY 03-115 Vermont Yankee Nuclear Power Station Technical Specification Proposed Change No. 257 Replacement Mark-Ups of the Current Technical Specifications
VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY Applicability:
Applies to the interrelated variable associated with fuel thermal behavior.
Objective:
To establish limits below which the integrity of the fuel cladding is preserved.
Specification:
A. Bundle Safety Limit (Reactor Pressure >800 psia and Core Flow
>10% of Rated)
When the reactor pressure is
>800 psia and the core flow is greater than 10% of rated:
2.1 FUEL CLADDING INTEGRITY Applicability:
Applies to trip setting of the instruments and devices which are provided to prevent the nuclear system safety limits from being exceeded.
Objective:
To define the level of the process variable at which automatic protective action is initiated.
Specification:
A. Trip Settings The limiting safety system trip settings shall be as specified below:
- 1. Neutron Flux Trip Settings I
I
- 1. A Minimum Critical Power Ratio (MCPR) of less than 1.10 (1.12 for Single Loop Operation) shall constitute violation of the Fuel Cladding Integrity Safety Limit (FCISL).
- a. APRM Flux ScramR (Run Mode)
When the mode switch i in the RUN position, I Ls
- he APRMf~uxscram i
sett6g hal Heas sonon Fissue 2. 1.1 I
(
-w, N.,a,,Ma+/-+/- De:
RT#1
S = setting in percent of rated thermal power (1593 MWt)
W - percent rated two loop drive flow where 100%
rated drive flow is that flow equivalent to 48 x 10l lbs/hr core flow Amendment No. 4&, 4., 44,.94,.4, 4-4, 4-64, 449, 176 6
VYNPS APRM Setpoints shall be <
on the graph.
20 80 100 For single loop operation, the APRM Scram setting is adjusted according to Technical Specification 2.1.A.1.a Amendment No.
G,
.94, l 8 27, 211 11
VYNPS BASES:
t llowable Value 2.1 FUEL CLADDING INTEGRITY A.
Trip Settings The bases for individual trip settings are discussed in the following paragraphs.
- 1. Neutron Flux Trip Settings
- a.
The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (1593 MWt).
Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux.
During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant Csetting wof the fuel.
Therefore, during abnormal operational sients, the thermal power of the fuel will be less than that
- !N~cated by the neutron flux at the scram setting.
Analyses ar formed to demonstrate that the APRM flux scram over the f settings from a maximum of 120% to the minimum flow bias provide protection from the fuel safety limit or a a ormal operational transients including those that may result in a thermal hydraulic instability.
An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached. The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. Reducing this operating margin would increase the frequency of spurious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the
- INSERT #2>
APRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit yet allows operating margin that reduces the possibility of
\\
~~unnecessary scrams.*
The scram tr esetting must be ad jsted to ensure at the LHGR transi t peak is not incr sed for any co
- nation of
<INSERT #3>
MFLPD an reactor core therma power. If the s am requires a chan due to an abnormal eating condition it will be acco pished by increasing he APRM gain by e ratio in Spe ification 2.1.A.l.a, hus assuring a rector scram at Seer than de ove s
For single recirculation loop operation, the APRH flux scram trip setting is reduced in accordance with the analysis presented in NEDO-30060, February 1983. This adjustment accounts for the difference between the single loop and two loop drive flow at the same core flow, and ensures that the margin of safety is not reduced during single loop operation.
Analyses of the limiting transients show that no scram adjustment is required to assure fuel cladding integrity when the transient is initiated from the operating limit MCPR defined in the Core Operating Limits Report.
AN, 4<INSERT #4>
Amendment No.
i@-, id
+9, 4>, Ad, 8-4,
- Ail44, 14
VYNPS TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT REQUIREMENTS Modes in Which Functions M be Operating lust Minimum Number Operating Instrument Channels Per Run Trip System (2)
X 1
Required ACTIONS When Minimum Conditions For Operation Are Not Satisfied (3)
A I
Trip Function
- 1.
Mode Switch in Shutdown (5A-Sl)
- 2.
Manual Scram (SA-S3A/B)
- 3.
IRM (7-41(A-F))
High Flux INOP
- 4.
High Flux (flow bias)
High Flux (reduced)
INOP
- 5.
High Reactor Pressure (PT-2-3-5S(A-D)(M))
- 6.
High Drywell Pressure (PT-5-12(A-D)(M))
- 7.
Reactor Low (6)
Water Level (LT-2-3-57A/B(M))
(LT-2-3-58A/B(M))
- 8.
Scram Discharge Volume High Level (LT-3-231 (A-H) (M))
Trip Settings Refuel (1)
X X
X X
1 A
<120/125 X
X X
x 2
2 A
A X
2 A or B A
A or B Startup (12)
X
<15%
X X
<1055 psig X
X 2
2 (5) c2.5 psig
>127.0 inches X
x X
X X
X 2
A X
x 2
A 2
A
<21 gallons X
X X
2 (per volume)
A Amendment No. G, 44, 4, S",
6, a14, -9, 44, 94-, 4l64,
-1G, 18 7 21
VYNPS TABLE 3.1.1 NOTES (Cont'd)
- 3.
When the requirements in the column "Minimum Number of Operating Instrument Channels Per Trip System' cannot be met for one system, that system shall be tripped.
If the requirements cannot be met for both trip systems, the appropriate ACTIONS listed below shall be taken:
_iate insertion of operable rods and complete insertion of all
< INSERT#6>
able rods within four hours.
- -uuce power level to IRM range and place mode switch in the "Startup/Hot Standby" position within eight hours.
c) Reduce turbine load and close main steam line isolation valves within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
d) Reduce reactor power to less than 30% of rated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- 4.
"WI' is percent rated two loop drive flow where 100l rated drive flow is that flow equivalent to 48 x 106 lbs/hr core flow. AW is he differenc etween e two loop d single loop d ve flow at the sa core flow, is difference ust be accounte or during single op operation AW 0 for /
two rec culation loop op ation.
- 5.
To be considered operable an APRM must have at least 2 LPRM inputs per level and at least a total of 13 LPRM inputs, except that channels A, C,
D, and F may lose all LPRM inputs from the companion APRM Cabinet plus one additional LPRM input and still be considered operable.
- 6.
The top of the enriched fuel has been designated as 0 inches and provides common reference level for all vessel water level instrumentation.
- 7.
Deleted.
- 8.
Deleted.
- 9.
Channel signals for the turbine control valve fast closure trip shall be derived from the same event or events which cause the control valve fast closure.
- 10.
Turbine stop valve closure and turbine control valve fast closure scram signals may be bypassed at <30% of reactor Rated Thermal Power.
- 11.
Not used.
- 12.
While performing refuel interlock checks which require the mode switch to be in Startup, the reduced APRM high flux scram need not be operable provided:
- a.
The following trip functions are operable:
- 1. Mode switch in shutdown,
- 2. Manual scram,
- 4.
High flux SRM scram in noncoincidence,
- 5. Scram discharge volume high water level, and;
- b. No more than two (2) control rods withdrawn.
The two (2) control rods that can be withdrawn cannot be face adjacent or diagonally adjacent.
Amendment No.
'4, Go, G, 644, 46&,
4, 94, 94,
- 4i44, 74-3-, 4-4, 4-94, 212 24
INSERT #1 Two loop operation:
S <
0.4W +
61.10%
for 0% <W<
31.1%
S<
1.28W
+
33.31%
for 31.1%
<W<
54.0%
S <
0.66 W +
67.28%
for 54.0%
< W <
75.0%
with a maximum of 117.0% power for W > 75.0%
Single loop operation:
S <
0.4 W
+
58.09%
for 0%
<W<
39.1%
S<
1.28 W
+
23.56%
for 39.1%
<W<
61.9%
S<
0.66 W
+
62.10%
for 61.9%
< W <
83.0%
with a maximum of 117.0% power for W > 83.0%
INSERT #2 The relationship between recirculation drive flow and reactor core flow is non-linear at low core flows. Due to stability concerns, separate APRM flow biased scram trip setting equations are provided for low core flows.
INSERT #3 The APRM flow biased flux scram Allowable Value is the limiting value that the trip setpoint may have when tested periodically, beyond which appropriate action shall be taken. For Vermont Yankee, the periodic testing is defined as the calibration. The actual scram trip is conservatively set in relation to the Allowable Value to ensure operability between periodic testing.
INSERT #4 The single loop operation equations are based on a bounding (maximum) difference between two loop and single loop drive flow at the same core flow of 8%.
INSERT #5 Two loop operation: (4)
S <
0.4 W +
61.10%
for 0%
<W<
31.1%
S<
1.28 W +
33.31%
for 31.1% <W<
54.0%
S <
0.66 W +
67.28%
for 54.0%
< W <
75.0%
with a maximum of 117.0% power for W > 75.0%
Single loop operation: (4)
S <
0.4 W +
58.09%
for 0%
<V W<
39.1%
S<
1.28 W + 23.56%
for 39.1%
<W<
61.9%
S<
0.66W
+
62.10%
for 61.9% <W<
83.0%
with a maximum of 117.0% power for W > 83.0%
INSERT #6 The specified APRM High Flux scram (flow bias) Trip Setting is an Allowable Value, which is the limiting value that the trip setpoint may have when tested periodically. The actual scram trip setting is conservatively set in relation to the Allowable Value.
Docket No. 50-271 BVY 03-1 1 5 Vermont Yankee Nuclear Power Station Technical Specification Proposed Change No. 257 Replacement Re-typed Technical Specifications Pages
VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY Applicability:
Applies to the interrelated variable associated with fuel thermal behavior.
Objective:
To establish limits below which the integrity of the fuel cladding is preserved.
Specification:
A. Bundle Safety Limit (Reactor Pressure >800 psia and Core Flow >10% of Rated)
When the reactor pressure is
>800 psia and the core flow is greater than 10% of rated:
- 1. A Minimum Critical Power Ratio (MCPR) of less than 1.10 (1.12 for Single Loop Operation) shall constitute violation of the Fuel Cladding Integrity Safety Limit (FCISL).
2.1 FUEL CLADDING INTEGRITY Applicability:
Applies to trip setting of the instruments and devices which are provided to prevent the nuclear system safety limits from being exceeded.
Objective:
To define the level of the process variable at which automatic protective action is initiated.
Specification:
A. Trip Settings The limiting safety system trip settings shall be as specified below:
- 1. Neutron Flux Trip Settings
When the mode switch is in the RUN position, the APRM flux scram Allowable Value shall be:
I I
Two loop operation:
SS 0.4W+ 61.10% for 0% < W SS 1.28W+ 33.31% for 31.1% < W SS 0.66W+ 67.28% for 54.0% < W With a maximum of 117.0% power 75.0%
Single loop operation:
SS 0.4W+ 58.09% for 0% < W SS 1.28W+ 23.56% for 39.1% < W SS 0.66W+ 62.10% for 61.9% < W With a maximum of 117.0% power 83.0%
- 31.1%
S 54.0%
- 75.0%
for W >
- 39.1%
- 61.9%
- 83.0%
for W >
where:
S = setting in percent of rated thermal power (1593 MWt)
Amendment No. 14-, 4-, 44,.9,.94, 4G9,
- 4S4, l-5-9, 4-76 6
VYNPS Amendment No.
14, 94, 8,
24111 11
VYNPS BASES:
2.1 FUEL CLADDING INTEGRITY A. Trip Settings The bases for individual trip settings are discussed in the following paragraphs.
- 1. Neutron Flux Trip Settings
- a.
APRM Flux Scram Allowable Value (Run Mode)
The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (1593 MWt).
Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux.
During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel.
Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting.
Analyses are performed to demonstrate that the APRM flux scram over the range of settings from a maximum of 120t to the minimum flow biased setting provide protection from the fuel safety limit for all abnormal operational transients including those that may result in a thermal hydraulic instability.
An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached.
The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation.
Reducing this operating margin would increase the frequency of spurious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses.
Thus, the APRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit yet allows operating margin that reduces the possibility of unnecessary scrams.
The relationship between recirculation drive flow and reactor core flow is non-linear at low core flows.
Due to stability concerns, separate APRM flow biased scram trip setting equations are provided for low core flows.
The APRM flow biased flux scram Allowable Value is the limiting value that the trip setpoint may have when tested periodically, beyond which appropriate action shall be taken.
For Vermont Yankee, the periodic testing is defined as the calibration.
The actual scram trip is conservatively set in relation to the Allowable Value to ensure operability between periodic testing. For single recirculation loop operation, the APRM flux scram trip setting is reduced in accordance with the analysis presented in NEDO-30060, February 1983.
This adjustment accounts for the difference between the single loop and two loop drive flow at the same core flow, and ensures that the margin of safety is not reduced during single loop operation.
The single loop Amendment No. 4A, aS, 347, 4-%
S-1, 94, a4-G, A4 16 14
VYNPS BASES:
2.1 (Cont'd) operation equations are based on a bounding (maximum) difference between two loop and single loop drive flow at the same core flow of 8%.
Analyses of the limiting transients show that no scram adjustment is required to assure fuel cladding integrity when the transient is initiated from the operating limit MCPR defined in the Core Operating Limits Report.
Amendment No.
14a
VYNPS TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT REQUIREMENTS Modes in Which Functions Must be Operatii Refuel (1)
Startup(12)
Minimum Number rig Operating Instrument Channels Per Run Trip System (2)
Required ACTIONS When Minimum Conditions For Operation Are Not Satisfied (3)
Trip Function
- 1.
Mode Switch in Shutdown (SA-Sl)
- 2.
Manual Scram (SA-S3A/B)
- 3.
IRM (7-41(A-F))
High Flux Trip Settings X
X x
X x
1 A
X 1
A
<120/125 X
X 2
A INOP
- 4.
High Flux (flow bias)
X X
2 A
Two loop operation:
(4)
S5 0.4W+
61.10% for 0% < W S SS 1.28W+ 33.31% for 31.1% < W S SS 0.66W+ 67.28% for 54.0% < W S With a maximum of 117.0% power for W > 75.0V X
2 A or B 31.1%
54.0%
75.0%
Single loop operation:
(4)
SS 0.4W+
58.09% for 0% < W S 39.1%
SS 1.28W+ 23.56% for 39.1% < W S 61.9%
SS 0.66W+ 62.10% for 61.9% < W 5 83.0%
With a maximum of 117.0% power for W > 83.0%
<15%
High Flux (reduced)
INOP X
X 2
A X
X 2(5)
A or B
- 5.
High Reactor Pressure (PT-2-3-55(A-D)
(M)
<1055 psig X
X X
2 A
Amendment No.
, 44, 44, 68, 47, 48, 44, 94, 94,
- l46,
-84, 182 21
VYNPS TABLE 3.1.1 NOTES (Cont'd)
- 3.
When the requirements in the column "Minimum Number of Operating Instrument Channels Per Trip System" cannot be met for one system, that system shall be tripped.
If the requirements cannot be met for both trip systems, the appropriate ACTIONS listed below shall be taken:
a) Initiate insertion of operable rods and complete insertion of all operable rods within four hours.
b) Reduce power level to IRM range and place mode switch in the "Startup/Hot Standby" position within eight hours.
c) Reduce turbine load and close main steam line isolation valves within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
d) Reduce reactor power to less than 30t of rated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- 4.
The specified APRM High Flux scram (flow bias) Trip Setting is an Allowable Value, which is the limiting value that the trip setpoint may have when tested periodically.
The actual scram trip setting is conservatively set in relation to the Allowable Value. "W" is percent rated two loop drive flow where 100% rated drive flow is that flow equivalent to 48 x 106 lbs/hr core flow.
- 5.
To be considered operable an APRM must have at least 2 LPRM inputs per level and at least a total of 13 LPRM inputs, except that channels A, C, D, and F may lose all LPRM inputs from the companion APRM Cabinet plus one additional LPRM input and still be considered operable.
- 6.
The top of the enriched fuel has been designated as 0 inches and provides common reference level for all vessel water level instrumentation.
- 7.
Deleted.
- 8.
Deleted.
- 9.
Channel signals for the turbine control valve fast closure trip shall be derived from the same event or events which cause the control valve fast closure.
- 10.
Turbine stop valve closure and turbine control valve fast closure scram signals may be bypassed at <30% of reactor Rated Thermal Power.
- 11. Not used.
- 12.
While performing refuel interlock checks which require the mode switch to be in Startup, the reduced APRM high flux scram need not be operable provided:
- a. The following trip functions are operable:
- 1. Mode switch in shutdown,
- 2. Manual scram,
- 5. Scram discharge volume high water level, and;
- b. No more than two (2) control rods withdrawn.
The two (2) control rods that can be withdrawn cannot be face adjacent or diagonally adjacent.
Amendment No. -14,
, Ql, 64, A6s,
-A, 9G, 894,
-4, 4-, a4
, 4o, 2122 24