Text

Response to Request for Additional Information for License Amendment Request to Revise Technical Specifications - Appendix K Measurement Uncertainty Recapture Power Uprate
	ML080990523
Person / Time
Site:	Cooper
Issue date:	04/04/2008
From:	Minahan S; Nebraska Public Power District (NPPD)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NLS2008034
	Download: ML080990523 (275)
	v • d • e

H Nebraska Public Power District "Always there when you need us" NLS2008034 April 4, 2008 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

Response to Request for Additional Information for License Amendment Request to Revise Technical Specifications.-. Appendix K Measurement Uncertainty Recapture Power Uprate Cooper Nuclear Station, Docket No. 50-298, DPR-46

References:

1. Letter from Carl F. Lyon, U.S. Nuclear Regulatory Commission, to Stewart B. Minahan, Nebraska Public Power District, dated February 22, 2008, "Cooper Nuclear Station - Request for Additional Information RE: Measurement Uncertainty Recapture Power Uprate (TAC No. MD73 85)"

2. Letter from Stewart B. Minahan, Nebraska Public Power District, to the U.S.

Nuclear Regulatory Commission, dated November 19, 2007, "License Amendment Request to Revise Technical Specifications - Appendix K Measurement Uncertainty Recapture Power Uprate"

Dear Sir or Madam:

The purpose of this letter is for the Nebraska Public Power District (NPPD) to submit responses to various Nuclear Regulatory Commission (NRC) Requests for Additional Information (RAI) related to the license amendment request (LAR) that would revise the Cooper Nuclear Station (CNS) Technical Specifications for Measurement Uncertainty Recapture power uprate. This LAR was submitted by NPPD letter dated November 19, 2007 (Reference 2).

The RAI from the Instrumentation and Controls Branch (EICB) was sent on February 22, 2008 (Reference 1). This response also contains NPPD's response to additional NRC RAI questions received from the NRC Reactor Systems Branch (SRXB), Electrical Engineering Branch (EEEB), and the Vessels and Internals Integrity Branch (CVIB), as communicated via electronic and telephone communications from the NRC CNS Project Manager. contains the response to the NRC EICB RAI questions from Reference 1.

Attachments 2 through 5 contain responses to RAI questions from the NRC SRXB, EEEB, and CVIB RAI, respectively.

COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3817 / Fax: (402) 825-5211 wvww.nppd.com

NLS2008034 Page 2 of 2 During preparation of this response to the RAIs a typographical error was discovered in the Technical Specification (TS) and Bases pages submitted by Reference 2. Attachment 6 contains a discussion of the typographical error. Attachment 7 contains a markup of the affected TS and Bases pages. Attachment 8 contains the affected TS pages in clean typed format.

Enclosures 1 through 5 contain various calculation records to support NPPD's responses to certain EICB RAI questions. These attachments and enclosures do not contain proprietary information as defined by 10 CFR 2.390. The information submitted by this letter (including attachment) does not change the conclusion of the No Significant Hazards Consideration evaluation submitted by Reference 2.

Should you have any questions regarding this submittal, please contact David Van Der Kamp, Licensing Manager, at (402) 825-2904.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on L/,4]°'- Zo6?

Sincerely, tewart B. Minahan Vice President - Nuclear and Chief Nuclear Officer

/dm Attachments (8)

Enclosures (5) cc: Regional Administrator w/ attachments and enclosures USNRC - Region IV Cooper Project Manager w/ attachments and enclosures USNRC - NRR Project Directorate IV-1 Senior Resident Inspector w/ attachments and enclosures USNRC - CNS Nebraska Health and Human Services w/ attachments and enclosures Department of Regulation and Licensure NPG Distribution w/o attachments and enclosures CNS Records w/ attachments and enclosures

ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS© C, ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS© Correspondence Number: NLS2008034 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document. Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments. Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

COMMITMENT COMMITTED DATE COMMITMENT NUMBER OR OUTAGE None N/A N/A I -t I t 4- I

4. 4

4. 4 1* I

4. I I PROCEDURE 0.42 REVISION 22 PAGE 18 OF 25

NLS2008034 Page 1 of 10 Attachment 1 Response to Instrumentation and Controls Branch Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications for Measurement Uncertainty Recapture Power Uprate Cooper Nuclear Station, Docket No. 50-298, DPR-46

Reference:

Letter from Carl F. Lyon, U.S. Nuclear Regulatory Commission, to Stewart B. Minahan, Nebraska Public Power District, dated February 22, 2008, "Cooper Nuclear Station - Request for Additional Information, RE: Measurement Uncertainty Recapture Power Uprate (TAC No. MD7385)"

The Nuclear Regulatory Commission (NRC) Request for Additional Information (RAI) questions from the Instrumentation and Controls Branch (EICB) are shown in italics. The Nebraska Public Power District (NPPD) responses are shown in block font.

NOTE: The NRC referred to 10 CFR 50.36(c) in several questions. This regulation was revised by addition of a new part (c). As a result, the applicable regulation is 50.36(d).

1. NRC Request On page 10 ofAttachment 1, the submittalproposes a 72 hourAllowed Outage time (AOT) and states that the alternate instrumentationaccuracy will gradually degrade over time associatedwith nozzle fouling and transmitterdrift, but that this degradationis likely to be imperceptiblefor a 72-hourperiod. This isa qualitative statement. Based on the transmitter drift record at the plant,please provide the calculatedeffect of the transmitterdrift on the power calorimetricduring this AOT.

NOTE: Nuclear Engineering Design Calculation (NEDC)06-035, Revision 1 (Enclosure 5) contains the Reactor Core Thermal Power Uncertainty Calculation.

This calculation had been previously submitted as Enclosure 7 of the License Amendment Request (LAR). It is being resubmitted to correct a minor error in the calculation of the lower bound of the total uncertainty. The error occurred when a negative bias (-0.0006% Power) was subtracted instead of added to the negative portion of the random uncertainty (-1.2979% Power). The negative bias results in a change of -0.0012% Power in the lower bound of the total uncertainty. This does not affect the License Amendment Request (LAR) submitted, November 19, 2007, to the NRC because the uncertainty was rounded up to the nearest 0.01% Power.

New transmitters are being installed by the Feedwater (FW) Level Control plant modification.

NLS2008034 Page 2 of 10 NPPD Response In section 3.1.2.5 of NEDC 06-035, the calculation shows that the Drift Effect (DERFFT) for the transmitter is calculated to be +0.1991% Span for an 18 month period.

DERFFT = +0.1991 % Span This effect then factors into the total loop uncertainty for FW Flow Differential Pressure (D/P) in section 3.1.3 (TLUFWF2 = +0.6756% Span). Then, this uncertainty is factored into the Total Reactor Thermal Power (i.e., Power Calorimetric) Calculation Uncertainty in Section 3.10, which results in an uncertainty that has random and bias attributes:

Total Uncertainty - +1.2979 % Power, +0.5795 % Power (Bias), -0.0006 %

Power (Bias)

= -1.2985 % Power / +1.8954 % Power In order to calculate the effect of the transmitter drift on the Power Calorimetric, the FW Flow D/P and the Power Calorimetric are calculated without the 18 month DERFFT-.

TLUFWF2 = +0.6756% Span TLUFWF2-DE18 = +0.6456% Span Total Uncertainty-DE18 = +1.2578 % Power, +0.5795 % Power (Bias), -0.0006 %

Power (Bias)

By comparing the random terms of the Total Uncertainty calculated with and without the 18 month drift effect you can determine the effect of the 18 month drift on the Power Calorimetric:

% Effect-DE18 = (1- (Total UncertaintyDEl8 / Total Uncertainty)) x 100%

= (1- (1.2578 / 1.2979)) x 100%

= +3.0893%

To determine the effect of a 72-hour period (3 days), the Drift Effect for 3 days (DERFFT3d) can be determined using the methodology of Section 3.1.2.5, and the same methodology described above can be applied for a 3 day drift.

DERFFT3d= (3/3652)". x (0.5141% Span)

= +0.0147 % Span The 3 day FW Flow D/P uncertainty is calculated with and without the Drift Effect:

NLS2008034 Page 3 of 10 TLUFWF2wDE3d = +0.6458% Span TLUFWF2-DE3d = +0.6456% Span The Total Uncertainty is calculated with and without the 3 day Drift Effect:

Total UncertaintywDE3d = +1.2580 % Power, +0.5795 % Power (Bias), -0.0006 %

Power (Bias)

Total UncertaintyDE3 = +/-1.2578 % Power, +0.5795 % Power (Bias), -0.0006 %

Power (Bias)

The Total Uncertainties are compared to determine the effect on the Power Calorimetric:

% Effect-DEd = (1- (Total UncertaintyDE3d / Total UncertaintywDE3d)) x 100%

= (1- (1.2578 / 1.2580)) x 100%

= +0.0177%

The effect of the transmitter drift on the Power Calorimetric during the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AOT is

+0.0177%. This uncertainty is considered to be imperceptible for a 72-hour period.

2. NRC Request How will reactorpower [will (sic)] be calculated when the plant computer is not operable?

How do/will plant procedures account.for when the computer is inoperable?

NPPD Response Reactor power is calculated by the backup plant computer when the primary system is not operable. In the event that both the primary and backup computers are inoperable, the reactor power can be ascertained from multiple parameters (steam flow, feed flow, turbine first stage pressure, main generator MWe). The computer malfunction abnormal procedure directs that power changes in progress be stopped when both computers malfunction and that operators then use alternate indications of reactor power (such as: APRM's, Feedwater flow, Steam flow, Turbine 1st stage pressure, etc.).

3. NRC Request RIS 2002-03, Attachment 1, Section .].F requests information to address the calibrationand maintenanceprocedures related to all instruments that affect the power calorimetric. The Nebraska Public Power Districtonly addressedthe LEFM (LeadingEdge Flow Meter) on this item in Table 4-1 ofAttachment 4. Pleaseprovide the requested informationfor the remaininginstruments.

NLS2008034 Page 4 of 10 NPPD Response For instrumentation other than the LEFM equipment, that contributes to the power calorimetric computation, calibration and maintenance is performed periodically using existing site procedures. Maintenance & Test Equipment (M&TE), setting tolerances, calibration frequencies, and instrumentation accuracy were evaluated and accounted for within the uncertainty determination of NEDC 06-035. The calibration and maintenance procedures related to the remaining instrumentation are referenced in Sections 5.2 and 5.3 of NEDC 06-035. All work will be performed in accordance with site work control procedures.

4. NRC Request Provide confirmation that the UFM (ultrasonicflow measurement) mass flow uncertainty used in the total thermalpower uncertainty determination includes uncertaintyfor the actual location of the transducers within the housing as identified in Cameron Customer Information Bulletin CIB 125, Rev. 0, dated April 23, 2007.

NPPD Response NPPD has confirmed that the UFM mass flow uncertainty used in the total thermal power uncertainty determination includes uncertainty for the actual location of the transducers within the housing. This is documented under Tab A-2 in Enclosure 4 of the LAR submittal (See line item "Transducer Installation Variability," under the input section labeled "Spool Piece Specific").

5. NRC Request The License Amendment Request (LAR) proposes the following TS changes:

a. Reference to "10% RTP (Rated Thermal Power)" has been scaled down to "9.85% RTP" in the following Technical Specifications (TS):

i. 3.1.3 CONDITIOND (page3.1.9)[sic (3.1-9)],

ii. 3.16 [sic (3.1.6] APPLICABILITY (page3.1-18),

iii. SR 3.3.2.1.2 and 3.3.2.1.3 (page3.3-17),

iv. SR 3.3.2.1.6 (page3.3-18),

v. Footnote 69 of Table 3.3.2.1-1 (page3.3-19).

b. Rejerence to "30% RTP" has been scaled down to "29.5% RTP" in the following TS:

i. 3.1.1.1 Reactor Protection (RPS)InstrumentationREQUIREDACTIONE. 1 (page 3.3-2), and related TS SR 3.3.1.1.14 (page3.3-5);

NLS2008034 Page 5 of 10 ii. Table 3.3.1.1-], FUNCTION 8 and 9, APPLICABLE MODES OR OTHER SPECIFIEDCONDITIONS (page3.3-8).

c. TS Table 3.3.1.1-1, Average Power Range Monitors ALLOWABLE VALUE of FUNCTION 2.b, Neutron Flux-High (Flow Biased), page 3.3-16, referenced by LCO 3.4.]c (RecirculationLoops Operating),is revisedfrom "< 0.66 W + 71.5% RTP(b)" to

"<0. 75 W + 62.00% RTPtb). " Footnote (b) is revisedfrom "< 0.66 W + 71.5% - 0.66" to

"<0.75 W+ 62.0%- 0.75."

d. ALLOWABLE VALUE on page 3.3-51 for TS Table 3.3.6.1-], FUNCTION 1.c., Main Steam Line Flow - High, is revisedfrom "< 144% rated steam flow" to "< 142. 7% rated steam flow. "

To supportNRC assessment of the acceptability of the LAR in regardto the setpoint change, please provide the followingfor each setpoint to be added or modified:

1. Setpoint CalculationMethodology: Provide documentation (includingsample calculations)of the methodology used for establishingthe limitingsetpoint (orNSP) and the limiting acceptable values for the As-Found and As-Left setpoints as measured in periodic surveillance testing as describedbelow. Indicate the relatedAnalytical Limits and other limiting design values (and the sources of these values)for each setpoint.

2. Safety Limit (SL)-Related Determination:Provide a statement as to whether or not the setpoint is a limiting safety system setting on which an SL has been placed as discussed in paragraph50.36(c)(1)(ii)(A) of Title 10 of the Code of FederalRegulations (10 CFR). If a setpoint is not SL-Related, explain the basisfor this determination.

3. For Setpoints That Are Determined To Be SL-Related: Please refer to the NRC letter to the Nuclear Energy Institute Setpoint Methods Task Force dated September 7, 2005 (ADAMS Accession No. ML052500004), which describes Setpoint-Related TS (SRTS) that are acceptableto the NRCfor instrumentsettings associatedwith SL-Related setpoints. Specifically: Part "A " of the Enclosure to the letterprovides Limiting Conditionfor Operations (LCO) notes to be added to the TS, and Part "B" includes a check list of the information to be provided in the TS Bases related to the proposed TS change.

a. Describe whether and how the SRTS suggested in the September 7, 2005, letter will be implemented. Ifyou do not plan to adopt the suggested SRTS, then explain how you will compliance with 10 CFR 50.36 by addressingitems 3b and 3c below.

NLS2008034 Page 6 of 10

b. As-Found Setpoint Evaluation:Describe how surveillancetest results and associated TS limits are used to establish operability of the safety system. Show that this evaluation is consistent with the assumptions and results of the setpoint calculation methodology. Discuss the plant corrective actionprocesses (including plant procedures)forrestoringchannels to operablestatus when channels are determined to be "inoperable" or "operable but degraded." If the criteriafor determining operability of the instrument being tested are located in a document other than the TS (e.g. plant test procedure) explain how the requirements of 10 CFR 50.36 are met.

c. As-Left Setpoint Control: Describe the controls employed to ensure that the instrument setpoint is, upon completion of surveillance testing, consistent with the assumptions of the associatedanalyses. If the controls are located in a document other than the TS (e.g. plant test procedure)explain how the requirements of 10 CFR 50.36 are met.

4. For Setpoints That Are Determined To Be Non-SL-related: Describe the measures to be taken to ensure that the associatedinstrument channel is capable ofperforming its specified safety functions in accordancewith applicabledesign requirements and associatedanalyses. Include in your discussion information on the controls you employ to ensure that the as left trip setting after completion ofperiodicsurveillance is consistent with your setpoint methodology. Also, discuss the plant corrective actionprocesses (includingplantprocedures)for restoringchannels to operable status when channels are determined to be "inoperable" or "operable but degraded." If the controls are located in a document other than the TS (e.g., plant testprocedure), describe how it is ensured that the controls will be implemented.

NPPD Response

1. Setpoint Calculation Methodology: Cooper Nuclear Station (CNS) is committed to using the GE Setpoint Calculation Methodology. Informational copies of the sample calculations that support the Technical Specification (TS) changes described above are provided in Enclosures 1 through 5 of this response. The calculations define the Analytical Limits, other limiting design values, As-Found and As-Left tolerances, and the reference sources for terms used to calculate the Allowable Value and Operating Setpoints.

2. Safety Limit (SL) - Related Determination: The determination of whether each change is associated with a Limiting Safety System Setting (LSSS) on which an SL has been placed, as discussed in paragraph 50.36(d)(1)(ii)(A) of Title 10 of 10CFR, is discussed as follows:

NLS2008034 Attachment I Page 7 of 10

a. Reference to "10% RTP (Rated Thermal Power)" has been scaled down to "9.85%

RTP" This change is associated with the Analytical Limit (AL) for the Rod Worth Minimizer (RWM) Low Power Permissive. The AL with respect to the accident analysis has not been changed. It is still based on 10% of the original rated thermal power (RTP) of 2381 MWth. The AL has only been scaled to the MUR power level of 2419 MWth. CNS calculation NEDC 92-050R, Revision 3C1 (Enclosure 1),

discusses the method used to scale the AL to the post-MUR implementation value of 9.84% (Sections 3.15 and 2.4 of NEDC 92-050R)(See Attachment 6 for discussion regarding the change from 9.85%). Since the AL is still based on 10% of the original RTP, the setpoints and Allowable Values derived from the AL have not changed.

Based on this, the setpoints do not change and a determination of LSSS is not applicable. In response to parts 3 and 4 of question 5, CNS will continue to address "As-Found" and "As-Left" setpoint control as prescribed in CNS operating procedures.

b. Reference to "30% RTP" has been scaled down to "29.5% RTP" This change is associated with the AL for the Turbine First Stage Pressure Permissive. The AL has not been changed with respect to the accident analysis. It is still based on 30% of the original rated turbine first stage pressure at RTP of 2381 MWth. The AL has only been scaled to correspond to the MUR power level of 2419 MWth. CNS calculation NEDC 92-050AJ, Revision 2C1 (Enclosure 2), discusses the method used to scale the AL to the post-MUR implementation value of 29.5%

(Section 3.15 of NEDC 92-050AJ). Since the AL is still based on 30% of the original rated turbine first stage pressure, the setpoints and Allowable Values derived from the AL have not changed. Based on this, the setpoints do not change and a determination of LSSS is not applicable. In response to parts 3 and 4 of question 5, CNS will continue to address "As-Found" and "As-Left" setpoint control as prescribed in CNS operating procedures.

c. TS Table 3.3.1.1-1, Average Power Range Monitors ALLOWABLE VALUE of FUNCTION 2.b, Neutron Flux-High (FlowBiased), page 3.3-16, referencedby LCO 3.4.1c (RecirculationLoops Operating),is revisedfrom "< 0.66 W + 71.5%

RTP(b)" to "< 0. 75 W+ 62.0% RTP(b). " Footnote (b) is revisedfrom "< 0.66 W+

71.5% - 0.66" to "<- . 75 W + 62.0% - 0.75."

This is the only change that involves a LSSS. This change is associated with the Allowable Value (AV) change for the Average Power Range Monitors (APRMs)

Flow Biased Neutron Flux-High trip. This function monitors neutron flux to

NLS2008034 Page 8 of 10 approximate the THERMAL POWER being transferred to the reactor coolant. The APRM neutron flux trip level is varied as a function of recirculation drive flow (i.e.,

at lower core flows, the setpoint is reduced proportional to the reduction in power experienced as core flow is reduced with a fixed control rod pattern). The APRM Neutron Flux-High (Flow Biased) Function is not specifically credited in the safety analyses, but is intended to provide protection against transients where THERMAL POWER increases slowly, and to provide protection for power oscillations which may result from reactor thermal hydraulic instability. This function is required to be operable in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). Therefore, this setpoint is a LSSS setpoint per 10CFR50.36(d)(1)(ii)(A).

d. ALLOWABLE VALUE on page 3.3-51for TS Table 3.3. 6.1-1, FUNCTION 1.c.,

Main Steam Line Flow - High, is revisedfrom ":5 144% rated steam flow" to ":5 142. 7% rated steam flow."

This change is associated with the AV for the Main Steam Line (MSL) Flow-High function. The AL from which this AV is derived, has not been changed. It is still based on 150% of the originally rated main steam flow at 2381 MWth. The AL has only been scaled to the MUR power level of 2419 MWth, resulting in an AV of 142.7% rated steam flow. CNS calculation NEDC 92-050AM, Revision ICI (Enclosure 3), discusses the method used to scale the AL to the post-MUR implementation value of 147.64% (Section 3.13 of NEDC 92-050AM). Since the AL is still based on 150% of the originally rated main steam line flow, the setpoints and AV's derived from the AL have not changed. Based on this, the setpoints do not change and a determination of LSSS is not applicable. In response to parts 3 and 4 of question 5, CNS will continue to address "As-Found" and "As-Left" setpoint control as prescribed in CNS operating procedures.

In summary, the changes described in sections a, b, and d above, do not affect the AV's or setpoints. They have only been scaled to address the MUR and they do not affect the LSSS.

3. For Setpoints That Are Determined To Be SL-Related:

Based on the response to part 2 above, only the APRM Neutron Flux-High (Flow Biased) needs to be addressed with respect to setpoints that are determined to be SL-Related.

a. The industry has generically addressed the issue of SRTS discussed in the NRC letter to the NEI dated 9/7/2005. Specifically, the industry, through Excel Services and the Technical Specification Task Force (TSTF), has conducted extensive coordination with

NLS2008034 Page 9 of 10 the NRC on this issue. As a result of that coordination the industry proposed an approach that is reflected in TSTF traveler 493, Revision 3. TSTF-493, Revision 3, was submitted to the NRC by letter to the NRC, Attn: Document Control Desk, dated January 18, 2008, and is currently under review by the NRC Staff. The NRC Staff is planning to process that traveler within their Consolidated Line Item Improvement Process (CLIIP). The current schedule is that the NRC Staff will issue the draft CLIIP documentation (i.e., a model safety evaluation and model application) for industry review by June 13, 2008, and will issue the CLIIP Notice of Availability by September 12, 2008.

Until the NRC Staff indicates their full acceptance of the approach proposed in TSTF-493, Revision 3, it would be premature for NPPD to indicate that it will follow that approach. Rather, NPPD will consider adopting that approach through the CLIIP when the Notice of Availability has been issued.

b. Cooper Nuclear Station develops Operating Limits (or Limiting Trip Settings) using the Analytical Limits defined by the inputs to the design basis analyses and the Allowable Values from CNS Technical Specifications, appropriately considering total loop uncertainty and accuracy. CNS uses the GE Setpoint Calculation Methodology for establishing instrument setpoints in the engineering calculation process.

Operability of instrument channels is established by meeting the Surveillance Requirements (SRs) and Allowable Values specified in the CNS Technical Specifications. Calibration and surveillance testing is performed according to the frequency specified in the Technical Specifications SRs. Surveillance testing is conducted to demonstrate the safety functions specified in the Bases and USAR, as applicable, can be performed.

If an as-found setpoint is discovered outside the pre-defined calibration tolerance determined in the setpoint calculations (and included in the calibration and surveillance procedures as acceptance criteria), but is conservative relative to the Allowable Value, the following action is taken:

o The condition is entered into the corrective action program (CAP) by initiation of a condition report.

o An operability determination is prepared to determine if the instrument safety function can still be met.

o If the instrument channel is functioning properly, and the instrument can be reset to within the calibration tolerance (which is conservative with respect to the Allowable Value) the instrument is considered operable and is reset appropriately.

NLS2008034 Page 10 of 10 If an as-found setpoint is discovered to be non-conservative relative to the Allowable Value, or an out-of-tolerance setpoint that is conservative to the AV cannot be reset to within the calibration tolerance, the instrument is considered inoperable. Upon discovery of an inoperable instrument channel, the following actions would be taken:

o The condition is entered into the CAP by initiation of a condition report.

o The applicable Technical Specification action statements would be followed.

o Appropriate action is taken to correct the cause of the inoperable instrument.

c. The Operating Limits (Limiting Trip Settings) and calibration tolerance are defined by calculation using the GE Setpoint Calculation Methodology in formal engineering calculations. The calculations ensure that the operating limits are set conservative relative to the Allowable Values of Technical Specifications and the Analytical Limits of design basis analyses. These Operating Limits and calibration tolerances are included as acceptance criteria in the calibration procedures. As-left calibration values are required to meet the acceptance criteria specified in the calibration procedures, which are derived in the setpoint calculations. These administrative controls are governed by station procedures.

4. For Setpoints That Are Determined To Be Non-SL-related:

No setpoint changes were identified in this category.

NLS2008034 Page 1 of 9 Attachment 2 Response to Reactor Systems Branch Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications for Measurement Uncertainty Recapture Power Uprate Cooper Nuclear Station, Docket No. 50-298, DPR-46

Reference:

Communication received from the NRC Project Manager for CNS, dated March 12, 2008.

The Nuclear Regulatory Commission (NRC) Request for Additional Information (RAI)

Questions from the Reactor Systems Branch (SRXB) are shown in italics. Nebraska Public Power District's (NPPD) response to each question is shown in block font.

1. NRC Request Reference Enclosure 7 to NLS200 7069:

Page 5: Are we correctin assuming this is providedas a form that should be usedfor guidance and that the "NO" checks do not apply? Conversely, if the "NO" checks do apply, then pleaseprovide the basesfor each check.

NPPD Response The form on page 5 is a calculation process form used to assist the calculation preparer in identifying the documents affected by the calculation. As stated on the form:

"If all answers are NO, then additional review or assistance is not required.

If any answers are YES or UNCERTAIN, then the Preparer shall obtain assistance from the System Engineer and other departments, as appropriate, to determine impacts to procedures and plant operations. Affected documents shall be listed on Attachment 2."

Attachment 2 (from the engineering procedure on calculations) is the "Design Calculation Cross-Reference Index" contained on pages 2, 3 and 4 of the calculation. Specifically, the Cross-Reference Index is used for the identification of documentation that could be affected by the implementation of the calculation.

NLS2008034 Page 2 of 9 The "NO" checks on the form are correct. Based on this the preparer did not require additional assistance in the determination of affected documents. This checklist is a Human Performance Tool. The bases for the responses to the questions on the list are not required to be documented. It should be noted that any individual preparing, revising, or reviewing a calculation must be trained on the calculation process. Therefore, they are trained on the meaning and intent of the questions contained on this form.

2. NRC Request Pages 19

23[sic (19 through 23)]: Loopfeedwater temperature uncertainty using RTDs is provided on Page 22 as +/-0.6704oF [sic (OF)] and the averagefeedwater temperature uncertainty is +/-0.4741oF[sic (OF)]. No uncertainty is identifiedfor a non-uniform feedwater temperature. In Enclosure 4, feedwater temperature uncertaintyprovided by the CheckPlus is listed on Page 6 ofAppendix A. 2 and Page 1 of Appendix A. 6 as 0. 58oF [sic (OF)] and this appears to include the CheckPlus capability of determiningan approximation of the mean feedwater temperature. Please discuss which temperature method you are using and, ifyou are using RTDs, then addressfeedwater temperature uniformity.

NPPD Response The Feedwater (FW) temperature uncertainty discussed on pages 19 through 23 is associated with the FW temperature measurements that currently exist in the system. The temperature elements are RTDs, and it is assumed that the FW temperature is uniform across the pipe.

The assumption basis is that the FW is uniformly mixed at the pumps and then again at the mixing throat just upstream of the temperature elements. The elements are located approximately 40 feet downstream of the FW mixing throat, with three 90' elbows between the mixing throat and the temperature elements, which also serve to mix the fluid to maintain uniform temperature distribution across the pipe. As stated in Section 3.2.1.1 of NEDC 06-035, the velocity of the FW is approximately 20.48 ft/sec, thus it only takes 2 seconds to travel from the mixing throat to the temperature element, further assuring that the temperature distribution is uniform.

With respect to the FW temperature uncertainty provided in Enclosure 4, it is calculated for the Caldon RTD associated with the LEFM CheckPlus spool piece.

3. NRC Request Reference Enclosure 5 and Related:

Figures 1 and 2 on Pages 4 and 5 show an "ALDEN PUMP." Is this a schematic illustrationto describe the source of water or is this the actualpump location with a

NLS2008034 Page 3 of 9 horizontalpump? If the former, then provide the missing description to describe the path from the sump to the connection with the 24" X 18" REDUCER or, alternatively, describe the missing descriptionwith respect to a reference such as Figures 1 and 3 in Trip Report dated February9, 2006 (ADAMS Accession No. ML060400418, attached).

NPPD Response Figures 1 and 2 in ER614R1 represent the actual model that was built for the test. The ALDEN pump represents the flow inlet to the model and the weigh tank represents the return to the sump. Figure 1 represents the flow input to the model by referencing the test pump (reference right side of figure 1). The return line on the left side of figure 1 represents the weigh tank that leads to the sump.

4. NRC Request Pleaseprovide C or D size schematic drawings of the plantfeedwater system that cover the flow pathfrom the feedwater pumps to a significant distance downstream of the planned CheckPlus locations and downstream of the existingfeedwater measurement instrumentation. Use a marker to illustrate the feedwaterpaths and identify all major components such as valves, existingfeedwater instrumentation,flow straighteners,and the planned CheckPlus locations.

NPPD Response Copies of the following drawings (with highlights) are submitted for information only:

Bums & Roe 2004, Sheet 3, "Flow Diagram - Condensate & Feedwater Systems,"

Revision N46.

Bums & Roe Iso Key 2004, Sheet 3, "Isometric Key - Condensate & Feedwater Systems," Revision N12.

" Jelco 2849-4, "18 Inch RF-1 from Reactor Feed Pumps IA & lB to Reactor Turbine Building," Revision N10.

The following is a brief description of the components in the Reactor FW lines from the pumps to where the lines enter the Steam Tunnel.

NLS2008034 Page 4 of 9 Reactor Feed Pump Room

" Reactor Feed Pumps (A & B) - Each Reactor Feed pump is equipped with a discharge check valve (RF-1 OCV and RF-1 ICV), and a motor operated discharge valve (RF-29MV and RF-30MV).

" Startup Flow Control Valves (FCV-1 1AA and FCV- 11 BB) - The startup flow control valves and associated motor operated isolation valves (RF-31MV, RF-32MV, RF-33MV, AND RF-34MV), bypass the RFP discharge valve during low flow conditions.

" FW Flow Elements (FE- I1A and FE- I1B) - Each Flow Element has a Flow Transmitter (FT-50A and FT-50B). The flow elements have built in flow straightening sections.

FW Mixing Throat - The 18" discharge lines from both Reactor Feed Pumps come together at the entrance of a 24" mixing throat. The FW flow exits the mixing throat as two 18" FW lines.

Caldon CheckPlus Spool Piece (Planned) - The Caldon spool pieces will be installed just downstream of the exit of the 24" mixing throat.

Heater Bay FW Flow Temperature Elements (TE-104A, TE-140B, TE-140C, and TE-140D) -

Approximately 25 feet downstream of the new Caldon CheckPlus Spool Pieces are the FW Flow Temperature Elements. These elements provide input to the Power Calorimetric calculation when the Caldon CheckPlus system is not in use. The flow elements are located in the Heater Bay, before the FW lines enter the Steam Tunnel.

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NLS2008034 Page 1 of 1 Attachment 3 Response to Reactor Systems Branch Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications for Measurement Uncertainty Recapture Power Uprate Cooper Nuclear Station, Docket No. 50-298, DPR-46

Reference:

Communication received from the NRC Project Manager for CNS, dated March 21, 2008.

The Nuclear Regulatory Commission (NRC) Request for Additional Information (RAI)

Questions from the Reactor Systems Branch (SRXB) are shown in italics. Nebraska Public Power District's (NPPD) response to each question is shown in block font.

NRC Request If CheckPlus is not operablefor greaterthan 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or there has been greaterthan a 10%

power change during an inoperablecondition lasting less than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , then the plant must be operated as though the CheckPlus was never installed. This means the venturi's are not recalibrated,licensedpower level is today's value, and the plant is operatedat a power level that is 2% less than the licensed power level to accountfor venturi uncertainty. I do not see this establishedin the licensee's submittal which, unless I've missed it, is silent on venturi calibration and does not identify a power level other than the licensedpower level.

NPPD Response The venturis will not be physically recalibrated when the CheckPlus is not operable for greater than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or there has been a change in power of greater than 10% during and inoperable condition lasting less than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . However, a software Calibration-Factor will beapplied to the venturi mass flow signal during the first 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that the CheckPlus is inoperable or out of service, provided that there are no Core Thermal Power (CTP) changes of greater than 10%.

After 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , TRM TLCO T3.3.5, "Feedwater Flow Instrumentation," requires that CTP be reduced to the originally licensed CTP of 2381 MWth (no Calibration-Factor applied).

This approach to CheckPlus Inoperability is described in Attachment 1 of the LAR submittal, Sections 4.2.2 and 4.2.8 (Criterion 1).

NLS2008034 Page 1 of 2 Attachment 4 Response to Electrical Engineering Branch Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications for Measurement Uncertainty Recapture Power Uprate Cooper Nuclear Station, Docket No. 50-298, DPR-46

Reference:

Telecon Meeting held March 14, 2008 and Communication Received from NRC Project Manager for CNS, March 18,2008.

The Nuclear Regulatory Commission (NRC) Request for Additional Information (RAI)

Questions from the Electrical Engineering Branch (EEEB) are shown in italics. Nebraska Public Power District's (NPPD) response to each question is shown in block font.

1. NRC Request Enclosure ] of the LAR (pg 5-1) refers to a grid stability analysis. Who performed the grid stability analysis? Was the analysis coordinatedwith your reliabilitycoordinator,Midwest Reliability Organization(reference the approval letterfrom the subcommittee)?

NPPD Response The grid stability analysis was performed by the Nebraska Public Power District Transmission Asset Planning engineering group.

The original grid stability analysis "Cooper Nuclear Station 22 MW Power Uprate Generation Accreditation Study" was submitted to the MAPP Design Review Subcommittee on July 15, 2005. This study documented the grid injection impacts for Cooper at 800 MW NET inlet to the transmission grid. This study was formally approved by the MAPP Design Review Subcommittee at their August 25, 2005 Meeting.

Based on the current and projected station performance and associated MAPP generation accreditation levels for CNS, the Appendix K uprate will not result in grid injection levels for CNS greater than 800 MW NET throughout the year. There is a slight chance that CNS could reach 801 MW NET during a Winter Peak Load timeframe. NPPD Transmission Asset Planning did not perform a new Accreditation Study for the potential 801 MW level since the results of that analysis would not be any different than those presented in the 800 MW Study and NPPD was not seeking approval of this new MAPP accreditation level since it was not available during the Summer Peak Season. Therefore, NPPD performed the stability sensitivity study referenced in Question # 2, to verify no grid stability issues at a 815 MW

NLS2008034 Page 2 of 2 NET injection level. The results of this sensitivity analysis validated the NPPD assumptions regarding the 801 MW level (i.e., the results were not different at 801 MW versus 800 MW).

2. NRC Request In the RAI response, letter dated March 6, 2008, the licensee states that a sensitivity analysis was performed to determine the impacts of different CNS generatoroutput levels on the LOCA analysis. Please describe what the sensiivity [sic (sensitivity)] analysis includes, specifically, the types of contingencies that were consideredin the study besides the loss of CNS.

NPPD Response The Cooper Nuclear Station Stability Sensitivity Study - Appendix K (815 MW NET) was performed to analyze the impacts of the proposed increase in CNS electrical output on the stability of the regional grid. This stability sensitivity analysis re-evaluated all of the worst case disturbances associated with the CNS area grid stability at a new 815 MW NET output level. A worst case stressed pre-disturbance base case was developed and then disturbance analysis was performed for system intact and prior outage conditions to evaluate the system response. Stuck breaker conditions were also evaluated to demonstrate the grid stability system response following delayed clearing multiple outage events.

The results of this sensitivity study demonstrated a stable system response to all disturbances under both system intact and prior outage conditions. There were no grid stability issues identified due to the injection of the increased electrical output (815 MW NET) of the CNS generator.

NLS2008034 Page 1 of 1 Attachment 5 Response to Vessels and Internals Integrity Branch Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications for Measurement Uncertainty Recapture Power Uprate Cooper Nuclear Station, Docket No. 50-298, DPR-46

Reference:

Communication Received from NRC Project Manager for CNS, March 12, 2008.

The Nuclear Regulatory Commission (NRC) Request for Additional Information (RAI)

Questions from the Vessels and Internals Integrity Branch (CVIB) are shown in italics.

Nebraska Public Power District's (NPPD) response to each question is shown in block font.

NRC Request Please confirm that Cooper belongs to the B WR VIP organizationand has implemented programs consistent with approvedB WR VIP reports and will implement programs consistent with B WR VIP reports currently under NRC review.

NPPD Response Cooper Nuclear Station currently belongs to and implements the BWRVIP Program as noted in BWRVIP letter from Carl Terry (BWRVIP) to Brian Sheron (NRC) dated May 30, 1997.

NLS2008034 Page 1 of 2 Attachment 6 Discussion of Typographical Error Regarding License Amendment Request to Revise Technical Specifications for Measurement Uncertainty Recapture Power Uprate Cooper Nuclear Station, Docket No. 50-298, DPR-46

Reference:

Letter from Stewart B. Minahan, Nebraska Public Power District, to the U.S.

Nuclear Regulatory Commission, dated November 19, 2007, "License Amendment Request to Revise Technical Specifications - Appendix K Measurement Uncertainty Recapture Power Uprate" During the development of Nebraska Public Power District's response to Instrumentation and Controls Branch (EICB) Question 2.a of this submittal (Attachment 1), it was discovered that the original License Amendment Request (LAR) submittal (referenced above) contains a typographical error. The typographical error involves the scaling of the "10% RTP" value to a new value as a result of the power uprate. The incorrect value is "9.85% RTP." The correct value is "9.84% RTP," which is derived from NEDC 92-050R (see Enclosure 1, Section 3.15).

Page 2 of 18 of Attachment 1 to the above reference states the following:

Reference to "10% RTP" has been scaled down to "9.85% RTP" in the following TS:

o 3.1.3 CONDITION D (page 3.1-9),

o 3.1.6 APPLICABILITY (page 3.1-18),

o SR 3.3.2.1.2 and 3.3.2.1.3 (page 3.3-17),

o SR 3.3.2.1.6 (page 3.3-18),

o Footnote (f) of Table 3.3.2.1-1 (page 3.3-19).

Page 3 of 18 of Attachment 1 to the above reference states the following:

Reference to "10% RTP" has been scaled down to "9.85% RTP" in the following TS BASES:

o B 3.1.3 D.1 and D.2 (two places) (page B 3.1-18),

0 E.1 (page B 3.1-19);

o B 3.

1.6 BACKGROUND

(page B 3.1-34),

0 APPLICABLE SAFETY ANALYSES (page B 3.1-35),

0 APPLICABILITY (two places) and ACTIONS (page B 3.1-36),

0 SR3.1.6.1 (pageB 3.1-37);

NLS2008034 Page 2 of 2 o B 3.3.

2.1 BACKGROUND

(page B 3.3-43),

0 APPLICABLE SAFETY ANALYSES (two places) (page B 3.3-46),

0 SR 3.3.2.1.2 and SR 3.3.2.1.3 (pages B3.3-50 and B 3.3-51),

0 SR 3.3.2.1.6 (page B 3.3-52);

o B 3.10.7 APPLICABILITY (two places) (page B 3.10-31).

As such, it is necessary to resubmit the above listed Technical Specification and Bases pages with the corrected value of "9.84% RTP." Attachment 7 contains the affected Technical Specifications and Bases pages in markup format. Attachment 8 contains the affected Technical Specification pages in clean typed format. The Technical Specification pages included incorporate changes up to Amendment 229.

NPPD has determined that this typographical error correction does not impact the Technical or Regulatory Safety Analysis, nor does it change the conclusion of the No Significant Hazards Consideration evaluation submitted in the above reference.

NLS2008034 Attachment 7 Technical Specification and Bases Pages - Markup Format Regarding License Amendment Request to Revise Technical Specifications for Measurement Uncertainty Recapture Power Uprate Cooper Nuclear Station, Docket No. 50-298, DPR-46 Technical Specification Pages 3.1-9 3.1-18 3.3-17 3.3-18 3.3-19 Bases Pages B 3.1-18 B 3.1-19 B 3.1-34 B 3.1-35 B 3.1-36 B 3.1-37 B 3.3-43 B 3.3-46 B3.3-50 B 3.3-51 B 3.3-52 B 3.10-31

Control Rod OPERABILITY 3.1.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.---------- NOTE------- D.I Restore compliance 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Not applicable when with BPWS.

THERMAL POWER OR D.2 Restore control rod 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Two or more inoperable to OPERABLE status.

control rods not in compliance with banked position withdrawal sequence (BPWS) and not separated by two or more OPERABLE control rods.

4 I E. Required Action and E.I Be in MODE 3.

associated Completion 112 hours0.0013 days 0.0311 hours 1.851852e-4 weeks 4.2616e-5 months Time of Condition A, C, or D not met.

OR Nine or more control rods inoperable.

__________________________ I ____________________________ I Cooper 3.1-9 Amendment No. i-

Rod Pattern Control 3.1.6 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Rod Pattern Control LCO 3.1.6 OPERABLE control rods shall comply with the requirements of the banked position withdrawal sequence (BPWS).

APPLICABILITY: MODES 1 and 2 with THERMAL POWER T 1.RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more OPERABLE A.1 --------NOTE-------

control rods not in Rod worth minimizer compliance with BPWS. (RWM) may be bypassed as allowed by LCO 3.3.2.1, "Control Rod Block Instrumentation."

Move associated 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months control rod(s) to correct position.

OR A.2 Declare associated 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months control rod(s) inoperable.

(continued)

Cooper 3.1-18 Amendment No.14-

Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.1.2 --------------------- NOTE ----------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after any control rod is withdrawn at < X/ RTP in MODE 2.

Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.2.1.3 --------------------- NOTE ---------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after THERMAL POWER is <:.t% RTP in MODE 1.

Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.2.1.4 --------------------- NOTE---------------

Neutron detectors are excluded.

Verify the RBM: 184 days

a. Low Power Range - Upscale Function is not bypassed when THERMAL POWER is

> 27.5% and < 62.5% RTP and a peripheral control rod is not selected.

b. Intermediate Power Range - Upscale Function is not bypassed when THERMAL POWER is > 62.5% and < 82.5% RTP and a peripheral control rod is not selected.

c. High Power Range - Upscale Function is not bypassed when THERMAL POWER is

> 82.5% RTP and a peripheral control rod is not selected.

(continued)

Amendment-29-331 3.3-17

Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.1.5 ------------------ NOTE----------------

Neutron detectors are excluded.

Perform CHANNEL CALIBRATION. 184 days SR 3.3.2.1.6 Verify the RWM is not bypassed when 18 months THERMAL POWER is _<.

SR 3.3.2.1.7 ------------------ NOTE----------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after reactor mode switch is in the shutdown position.

Perform CHANNEL FUNCTIONAL TEST. 18 months SR 3. at2--1. 8 Verify control rod sequences input to the Prior to RWM are in conformance with BPWS. declaring RWM OPERABLE following loading of sequence into RWM Cooper 3.3-18 Amendment No. -H--

Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE

1. Rod Block Monitor

a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 (h)

SR 3.3.2.1.4 SR 3.3.2.1.5

b. Intermediate Power Range - Upscale (b) 2 SR 3.3.2.1.1 (h)

SR 3.3.2.1.4 SR 3.3.2.1.5

c. High Power Range - Upscale (c),(d) 2 SR 3.3.2.1.1 (h)

SR 3.3.2.1.4 SR 3.3.2.1.5

d. Inop (d),(e) 2 SR 3.3.2.1.1 NA

e. Downscale (d),(e) 2 SR 3.3.2.1.1 > 92/125 SR 3.3.2.1.5 divisions of full scale

2. Rod Worth Minimizer 1 (f), 2 (f) 1 SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.6 SR 3.3.2.1.8

3. Reactor Mode Switch - Shutdown Position (g) 2 SR 3.3.2.1.7 NA (a) THERMAL POWER > 27.5% and < 62.5% RTP and MCPR < 1.70 and no peripheral control rod selected.

(b) THERMAL POWER > 62.5% and < 82.5% RTP and MCPR < 1.70 and no peripheral control rod selected.

(c) THERMAL POWER > 82.5% and < 90% RTP and MCPR < 1.70 and no. peripheral control rod selected.

(d) THERMAL POWER > 90% RTP and MCPR < 1.40 and no peripheral control rod selected.

(e) THERMAL POWER > 27.5% and < 90% RTP and MCPR < 1.70 and no peripheral control rod selected.

(f) With THERMAL POWER < I/oRTP.

(g) Reactor mode switch in the shutdown position.

(h) Less than or equal to the Allowable Value specified in the COLR.

Amendment-2e8- 3.3-19

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS C.1 and C.2 (continued) electrically disarmed by disconnecting power from all four directional control valve solenoids. Required Action C.1 is modified by a Note, which allows the RWM to be bypassed if required to allow insertion of the inoperable control rods and continued operation. LCO 3.3.2.1 provides additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis.

The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide time to insert and disarm the control rods in an orderly

-manner and without challenging plant systems.

D.1 and D.2 Out of sequence control rods may increase the potential reactivity worth of a dropped control rod during a CRDA. At

< 4.4W RTP, the generic banked position withdrawal qquence (BPWS) analysis (Ref. 6) requires inserted control rods not in compliance with BPWS to be separated by at least two OPERABLE control rods in all directions, including the

<j, LJ Therefore, if two or more inoperable control rods c0,-4 diagonal.

are not in compliance with BPWS and not separated by at least two OPERABLE control rods, action must be taken to restore wr compliance with BPWS or restore the control rods to OPERABLE status. Condition D is modified by a Note ndicating that the Condition is not applicable when

> RTP, since the BPWS is not required to be followed under these conditions, as described in the Bases for LCO 3.1.6. The allowed Completion Time.of 4.hours is acceptable, considering-the low probability of a CRDA occurring.

E.1 If any Required Action and associated Completion Time of Condition A, C, or D are not met, or there are nine or more inoperable control rods, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This ensures all insertable control rods are inserted and places the reactor in a condition that does not require the (continued)

Cooper B 3.1-18 Revision --

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS E.1 (continued) active function (i.e., scram) of the control rods. The number of control rods permitted to be inoperable when operating above .+Y,RTP (e.g., no CRDA considerations) could more an te value specified, but the occurrence of a large number of inoperable control rods could be indicative of a generic problem, and investigation and resolution of the potential problem should be undertaken. The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without. challenging plant systems.

SURVEILLANCE SR 3.1.3.1 REQUIREMENTS The position of each control rod must be determined to ensure adequate information on control rod position is available to the operator for determining control rod OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency of this SR is based on

. . . operating experience related to expected changes in control rod position and the availability of control rod position indications in the control room.

SR 3.1.3.2 and SR 3.1.3.3

'Control rod insertion capability is demonstrated by inserting each partially or fully;w'ithdrawn'control rod at least one notch. and observ-ing that the control rod moves.

The control rod may then be. returned.toits 'original position. This ensures the control rod is not stuck and is free to insert on a scram signal. These Surveillances are not required when THERMAL POWER is less than or equal to the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of the Banked Position Withdrawal Sequence (BPWS) (LCO 3.1.6) and the RWM (LCO 3.3.2.1). The 7 day Frequency of SR 3.1.3.2 is based on operating experience related to the changes in CRD performance and the ease of performing notch testing for fully withdrawn control rods. Partially withdrawn control (continued)

Cooper B 3.1-19 Revision 4--

Rod Pattern Control B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Rod Pattern Control BASES BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM)

(LCO 3.3.2.1, "Control Rod Block Instrumentation"), so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted 4-G/o RTP. The sequences limit, the potential outof reactivity addition that could occur in the event of a Control Rod Drop Accident (CRDA).

This Specification assures that thecontrol rod patterns are consistent with the assumptions of the CRDA analyses of References I and 2.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the CRDA are summarized in References I and 2. CRDA analyses assume that the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analysis. The RWM (LCO 3.3.2.1) provides backup to operator control of the withdrawal sequences to ensure that the initial conditions of the CRDA analysis are not violated.

Prevention or mitigation of positive reactivity insertion events is necessary to limit the energy deposition in the fuel, thereby preventing significant fuel damage which could result in the undue release of radioactivity. Since the failure consequences for.U0 2 have been shown-to be insignificant below fuel, energy deposlitions .of 300 cal/gm (Ref. 3), the fuel'damage limit; of 280. cal/gm provides a margin of safety from significant core damage which would result in release of radioactivity (Refs. 4'and 5).. Generic evaluations (Refs. 1 and 6) of a design basis CRDA (i.e., a CRDA resulting in a peak fuel energy deposition of 280 cal/gm) have shown that if the peak fuel enthalpy remains below 280 cal/gm, then the maximum reactor pressure will be less than the required ASME Code limits (Ref. 7) and the calculated offsite doses will be well within the required limits (Ref. 5).

(continued)

Cooper B 3.1-34 Revision -9

Rod Pattern Control B 3.1.6 BASES APPLICABLE Control rod patterns analyzed in Reference 1 follow the banked position SAFETY ANALYSES withdrawal sequence (BPWS). The BPWS is applica, from the (continued) condition of all control rods fully inserted to T-P (Ref. 2). For the BPWS, the control rods are required to be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions (e.g., between notches 08 and 12). The banked positions are established to minimize the maximum incremental control rod worth without being overly restrictive during normal plant operation.

Generic analysis of the BPWS (Ref. 1) has demonstrated that the 280 cal/gm fuel damage limit will not be violated during a CRDA while following the BPWS mode of operation. The generic BPWS analysis (Ref. 8) also evaluates the effect of fully inserted, inoperable control rods not in compliance with the sequence, to allow a limited number (i.e.,

eight) and distribution of fully inserted, inoperable control rods.

When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 10) may be used provided that all withdrawn control rods have been confirmed to be coupled. The rods may be inserted without the need to stop at intermediate positions since the possibility of a CRDA is eliminated by the confirmation that withdrawn control rods are coupled.

When using the Reference 10 control rod sequence for shutdown, the rod worth minimizer may be reprogrammed to enforce the requirements of the improved BPWS control rod insertion, or may be bypassed and the improved BPWS shutdown sequence implemented under LCO 3.3.2.1, Condition D controls.

In order to use the Reference 10 BPWS shutdown process, an extra check is required in order to consider a control rod to be "confirmed" to be coupled. This extra check ensures that no Single Operator Error can result in an incorrect coupling check. For purposes of this shutdown process, the method for confirming that control rods are coupled varies depending on the position of the control rod in the core. Details on this coupling confirmation requirement are provided in Reference 10. If the requirements for use of the BPWS control rod insertion process contained in Reference 10 are followed, the plant is considered to be in compliance with BPWS requirements, as required by LCO 3.1.6.

Rod pattern control satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 9).

LCO Compliance with the prescribed control rod sequences minimizes the potential consequences of a CRDA by limiting the initial conditions to those consistent with the BPWS. This LCO only applies to OPERABLE control rods. For inoperable control rods required to be inserted, separate requirements are specified in LCO 3.1.3, "Control Rod OPERABILITY," consistent with the allowances for inoperable control rods in the BPWS.

Cooper B 3.1-35

Rod Pattern Control B 3.1.6 BASES (continued) 6Th APPLICABILITY In MODES 1 and 2, when THERMAL POWER is <4.e/ 0 RTP, the CRDA is a Design Basis Accident and, therefore, compliance with the assumptions of the safety analysis is required. When THERMAL POWER is > 10% RTP, there is no credible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Ref. 2). In MODES 3, 4, and 5, since the reactor is shut down and only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will remain subcritical with a single control rod withdrawn.

ACTIONS A.1 and A.2 With one or more OPERABLE control rods not in compliance with the prescribed control rod sequence, actions may be taken to either correct the control rod pattern or declare the associated control rods inoperable within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Noncompliance with the prescribed sequence may be the result of "double notching," drifting from a control rod drive cooling water transient, leaking scram valves, or a power reduction to < 10% RTP before establishing the correct control rod pattern. The number of OPERABLE control rods not in compliance with the prescribed sequence is limited to eight, to prevent the operator from attempting to correct a control rod pattern that significantly deviates from the prescribed sequence.

Required Action A.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position. LCO 3.3.2.1 requires verification of control rod movement by a second licensed operator (Reactor Operator or Senior Reactor Operator) or by a qualified member of the technical staff. This ensures that the control rods will be moved to the correct position. A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by Required Action A.2. The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, considering the restrictions on the number of allowed out of sequence control rods and the low probability of a CRDA occurring during the time the control rods are out of sequence.

B.1 and B.2 If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence. Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from the prescribed sequence. Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of control rods has less impact on control rod Cooper B 3.1-36

Rod Pattern Control B 3.1.6 BASES ACTIONS B.1 and B.2 (continued) worth than withdrawals have. Required Action B.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position. LCO 3.3.2.1 requires verification of control rod movement by a second licensed operator (Reactor Operator or Senior Reactor Operator) or by a qualified member of the technical staff.

When nine or more OPERABLE control rods are not in compliance with BPWS, the reactor mode switch must be placed in the shutdown position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . With the mode switch in shutdown, the reactor is shut down, and as such, does not meet the applicability requirements of this LCO. The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the control rods out of sequence.

SURVEILLANCE SR 3.1.6.1 REQUIREMENTS The control rod pattern is verified to be in compliance with the BPWS at a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency to ensure the assumptions of the CRDA analyses are met. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency was developed considering that the primary check on compliance with the BPWS is performed by the RWM (LCO 3.3.2.1), which provides control rod blocks to enforce the required sequence and is required to be OPERABLE when operating at REFERENCES 1. NEDE-2401 1-P-A-US, "General Electric Standard Application for Reactor Fuel, Supplement for United States," Section 2.2.3.1 (Revision specified in the COLR).

2. "Modifications to the Requirements for Control Rod Drop Accident Mitigating System," BWR Owners Group, July 1986.

3. NUREG-0979, Section 4.2.1.3.2, April 1983.

4. NUREG-0800, Section 15.4.9, Revision 2, July 1981.

5. 10 CFR 100.

Cooper B 3.1-37 Cooer31-31444ieOr

Control Rod Block Instrumentation B 3.3.2.1 BASES BACKGROUND RBM averaging but remain in the display and LPRM alarm (continued) logic. Assignment of power range detector assemblies to be used in RBM averaging is controlled by the selection of control rods. The minimum number of LPRM inputs required to each RBM channel to prevent an instrument inoperative alarm is four when using eight LPRM assemblies, three when using six LPRM assemblies, and two when using four LPRM assemblies. The RBM is automatically bypassed and the output set to zero if a peripheral control rod is selected since the RBM function is not required for these rods. In addition, any one of the two RBM channels can be manually bypassed. If any LPRM detector assigned to a RBM is bypassed, the computed average signal is adjusted automatically to compensate for the number of LPRM input signals to average. When a control rod is selected, the signal conditioner gain is automatically adjusted so that the output level of the signal conditioner always corresponds to a constant level (relative to the initialization reference signal of 100/125 of full scale).

The gain set will be held constant during the movement of that rod, thus providing an indication of the change in the relative local power level. Whenever the reactor power level is below the lowest RBM operating range, the RBM is zeroed and RBM outputs are bypassed. If the indicated power increases above the preset limit, a rod block will occur.

In addition, to preclude rod movement with an inoperable RBM, a downscale trip and an inoperative trip are provided.

A rod block signal is generated if an RBM downscale trip or an inoperable trip occurs, since this could indicate a problem with the RBM channel. The downscale trip will occur if the RBM channel signal decreases below the downscale trip setpoint after the RBM channel signal has been normalized.

The inoperable trip will occur during the nulling (normalization) sequence, if the RBM channel >fails to null, too few LPRM inputs are-available,. a module is not plugged in, or the function switch is moved to any position other than "Operate."

The purpose of the RWM is to control rod patterns during startup and shutdown, such that only specified control .rod sequences and relative positions are allowed over the operating range from all control rods inserted to -tqRTP.(,.%L)

The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA. Prescribed control rod sequences are stored in the RWM, which will initiate control rod withdrawal and insert blocks when the (continued)

Cooper B 3.3-43 CReision 0

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The analytical methods and assumptions used in evaluating the CRDA are summarized in References 5 and 6. The BPWS requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the BPWS are specified in LCO 3.1.6, "Rod Pattern Control."

When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 7) may be used if the coupling of each withdrawn control rod has been confirmed. The rods may be inserted without the need to stop at intermediate positions. When using the Reference 7 control rod insertion sequence for shutdown, the rod worth minimizer may be reprogrammed to enforce the requirements of the improved BPWS control rod insertion, or may be bypassed and the improved BPWS shutdown sequence implemented under the controls in Condition D.

The RWM Function satisfies Criterion 3 of Reference 4.

Since the RWM is a system designed to act as a backup to operator control of the rod sequences, only one channel of the RWM is available and required to be OPERABLE (Ref. 7). Special circumstances provided for in the Required Action of LCO 3.1.3, "Control Rod OPERABILITY,"

and LCO 3.1.6 may necessitate bypassing the RWM to allow continued operation with inoperable control rods, or to allow correction of a control rod pattern not in compliance with the BPWS. The RWM may be bypassed as required by these conditions, but then it must be considered inoperable and the Required Actions of this LCO followed.

Compliance with the BPWS, and therefore OPERABILITY of the RWM, is required in MODES 1 and 2 when THERMAL POWER is < /o RTP.

When THERMAL POWER is >419/%RTP, there is no possible control rod 9*1,i configuration that results in a conTrol rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Ref. 5). In MODES 3 and 4, all control rods are required to be inserted into the core; therefore, a CRDA cannot occur. In MODE 5, since only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.

Cooper B 3.3-46 027'6ý

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued)

Required Actions taken. This Note is based on the reliability analysis (Ref. 9) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary.

SR 3.3.2.1.1 A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the channel will perform the intended function. It includes the Reactor Manual Control System input. It also includes the local alarm lights representing upscale and downscale trips, but no rod block will be produced at this time. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on reliability analyses (Ref. 10).

SR 3.3.2.1.2 and SR 3.3.2.1.3 A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the system will perform the intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay.

This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The CHANNEL FUNCTIONAL TEST for the RWM includes performing the RWM computer on line diagnostic test satisfactorily, attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs. For SR 3.3.2.1.2, the CHANNEL FUNCTIONAL TEST also includes attempting to select a control rod not in compliance with the prescribed sequence and verifying a selection error occurs. As noted in the SRs, SR 3.3.2.1,2 is not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after any control rod is withdrawn in MODE 2. As noted, SR 3.3.2.1.3 is not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after THERMAL POWER is < Y/o RTP in MODE 1. This allows Cooper B 3.3-50 . 1414*

I"

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued) entry into MODE 2 for SR 3. ..1.2, and entry into MODE 1 when THERMAL POWER is < oaRTP for SR 3.3.2.1.3, to perform the required Surveillance if the 92 day Frequency is not met per SR 3.0.2.

The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. The Frequencies are based on reliability analysis (Ref. 10).

SR 3.3.2.1.4 The RBM power range setpoints control the enforcement of the appropriate upscale trips over the proper core thermal power range of the Applicability Notes (a), (b), (c), (d), and (e) of ITS Table 3.3.2.1-1. The RBM Upscale Trip Function setpoints are automatically varied as a function of power. Three Allowable Values are specified in the COLR as denoted in Table 3.3.2.1-1, each within a specific power range. The power at which the control rod block Allowable Values automatically change are based on the reference APRM signal's input to each RBM channel. Below the minimum power setpoint of 27.5% RTP or when a peripheral control rod is selected, the RBM is automatically bypassed.

These power Allowable Values must be verified periodically by determining that the power level setpoints are less than or equal to the specified values. If any power range setpoint is nonconservative, then the affected RBM channel is considered inoperable. Alternatively, the power range channel can be placed in the conservative condition (i.e.,

enabling the proper RBM setpoint). If placed in this condition, the SR is met and the RBM channel is not considered inoperable. As noted, neutron detectors are excluded from the Surveillance because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.8. The 184 day Frequency is based on the actual trip setpoint methodology utilized for these channels.

SR 3.3.2.1.5 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.

Cooper B 3.3-51 1ý48ý95

Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued)

As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.8.

The Frequency is based upon the assumption of a 184 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.

SR 3.3.2.1.6 The RWM is automatically bypassed when power is above a specified value. The power level is determined from feedwater flow and steam flow signals. The setpoint where the automatic bypass feature is unbypassed must be verified periodically to be > .- /o RTP. If the RWM low power setpoint is nonconservative, then the RWM is considered inoperable.

Alternately, the low power setpoint channel can be placed in the conservative condition (nonbypass). If placed in the nonbypassed condition, the SR is met and the RWM is not considered inoperable. The Frequency is based on the trip setpoint methodology utilized for the low power setpoint channel.

SR 3.3.2.1.7 A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch - Shutdown Position Function to ensure that the channel will perform the intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The CHANNEL FUNCTIONAL TEST for the Reactor Mode Switch - Shutdown Position Function is performed by attempting to withdraw any control rod with the reactor mode switch in the shutdown position and verifying a control rod block occurs.

As noted in the SR, the Surveillance is not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links. This allows entry into MODES 3 and 4 if the 18 month Frequency Cooper B 3.3-52 --Itl4t05 I

Control Rod Testing-Operating B 3.10.7 BASES APPLICABLE As described in LCO 3.0.7, compliance with Special SAFETY ANALYSES Operations LCOs is optional, and therefore, no criteria of (continued) 10 CFR 50.36 (c)(2)(ii) (Ref. 3) apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.

LCO As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Control rod testing may be performed in compliance with the prescribed sequences of LCO 3.1.6, and during these tests, no exceptions to the requirements of LCO 3.1.6 are necessary. For testing performed with a sequence not in compliance with LCO 3.1.6, the requirements of LCO 3.1.6 may be suspended, provided additional administrative controls are placed on the test to ensure that the assumptions of the special safety analysis for the test sequence are satisfied. Assurances that the test sequence is followed can be provided by either programming the test sequence into the RWM, with conformance verified as specified in SR 3.3.2.1.8 and allowing the RWM to monitor control rod withdrawal and provide appropriate control rod blocks if necessary, or by verifying conformance to the approved test sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff. These controls are consistent with those normally applied to operation in the startup range as defined in the SRs and ACTIONS of LCO 3.3.2.1, "Control Rod Block Instrumentation."

APPLI CABILITY Control rod testing, while in MODES I and 2, with THERMAL POWER greater than 4 RTP, is adequately controlled by the existing LCOs on power distribution limits and control rod block instrumentation. Control rod movement during these conditions is not restricted to prescribed sequences and can be performed within the constraints of LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," and LCO 3.3.2.1.

With THERMAL POWER less than or equal to . RTP, the provisions of this Special Operations LCO re necessary to (conti nued)

Cooper B 3. 10.-31 Coe B3o i ntined

NLS2008034 Attachment 8 Technical Specification and Bases Pages - Final Typed Format Regarding License Amendment Request to Revise Technical Specifications for Measurement Uncertainty Recapture Power Uprate Cooper Nuclear Station, Docket No. 50-298, DPR-46 Technical Specification Pages 3.1-9 3.1-18 3.3-17 3.1-18 3.3-19

Control Rod OPERABILITY 3.1.3 Actions (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. --------- NOTE ------------- D.1 Restore compliance with 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Not applicable when BPWS.

THERMAL POWER

> 9.84% RTP. OR D.2 Restore control rod to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Two or more inoperable OPERABLE status.

control rods not in compliance with banked position withdrawal sequence (BPWS) and not separated by two or more OPERABLE control rods.

E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, C, or D not met.

OR Nine or more control rods inoperable.

Cooper 3.1-9 Amendment No.

Rod Pattern Control 3.1.6 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Rod Pattern Control LCO 3.1.6 OPERABLE control rods shall comply with the requirements of the banked position withdrawal sequence (BPWS).

APPLICABILITY: MODES 1 and 2 with THERMAL POWER < 9.84% RTP.

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. One or more OPERABLE A.1- --------- NOTE-------

control rods not in Rod worth minimizer compliance with BPWS. (RWM) may be bypassed as allowed by LCO 3.3.2.1, "Control Rod Block Instrumentation."

Move associated control 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months rod(s) to correct position.

OR A.2 Declare associated control 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months rod(s) inoperable.

(continued)

Cooper 3.1-18 Amendment No.

Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.1.2 --------------------- NOTE ---------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after any control rod is withdrawn at < 9.84% RTP in MODE 2.

92 days Perform CHANNEL FUNCTIONAL TEST.

SR 3.3.2.1.3 --------------------- NOTE ---------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after THERMAL POWER is < 9.84% RTP in MODE 1.

Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.2.1.4 --------------------- NOTE ---------------

Neutron detectors are excluded.

Verify the RBM: 184 days

a. Low Power Range - Upscale Function is not bypassed when THERMAL POWER is

> 27.5% and < 62.5% RTP and a peripheral control rod is not selected.

b. Intermediate Power Range - Upscale Function is not bypassed when THERMAL POWER is > 62.5% and < 82.5% RTP and a peripheral control rod is not selected.

c. High Power Range - Upscale Function is not bypassed when THERMAL POWER is

> 82.5% RTP and a peripheral control rod is not selected.

(continued)

Amendment 3.3-17

Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.1.5 - ----------------- NOTE ----------------

Neutron detectors are excluded.

Perform CHANNEL CALIBRATION. 184 days SR 3.3.2.1.6 Verify the RWM is not bypassed when THERMAL 18 months POWER is < 9.84% RTP.

SR 3.3.2.1.7 --------------------- NOTE ---------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after reactor mode switch is in the shutdown position.

Perform CHANNEL FUNCTIONAL TEST. 18 months SR 3.3.2.1.8 Verify control rod sequences input to the RWM are Prior to declaring in conformance with BPWS. RWM OPERABLE following loading of sequence into RWM Cooper 3.3-18 Amendment No.

Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE

1. Rod Block Monitor

a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 (h)

SR 3.3.2.1.4 SR 3.3.2.1.5

b. Intermediate Power Range - Upscale (b) 2 SR 3.3.2.1.1 (h)

SR 3.3.2.1.4 SR 3.3.2.1.5

c. High Power Range - Upscale (c),(d) 2 SR 3.3.2.1.1 (h)

SR 3.3.2.1.4 SR 3.3.2.1.5

d. Inop (d),(e) 2 SR 3.3.2.1.1 NA

e. Downscale (d),(e) 2 SR 3.3.2.1.1 > 92/125 SR 3.3.2.1.5 divisions of full scale

2. Rod Worth Minimizer 1 (f), 2 (f) 1 SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.6 SR 3.3.2.1.8

3. Reactor Mode Switch - Shutdown Position (g) 2 SR 3.3.2.1.7 NA (a) THERMAL POWER > 27.5% and < 62.5% RTP and MCPR < 1.70 and no peripheral control rod selected.

(b) THERMAL POWER > 62.5% and < 82.5% RTP and MCPR < 1.70 and no peripheral control rod selected.

(c) THERMAL POWER > 82.5% and < 90% RTP and MCPR < 1.70 and no peripheral control rod selected.

(d) THERMAL POWER > 90% RTP and MCPR < 1.40 and no peripheral control rod selected.

(e) THERMAL POWER > 27.5% and < 90% RTP and MCPR < 1.70 and no peripheral control rod selected.

(f) With THERMAL POWER < 9.84% RTP.

(g) Reactor mode switch in the shutdown position.

(h) Less than or equal to the Allowable Value specified in the COLR.

Amendment 3.3-19

NLS2008034 Enclosure 1 Rod Worth Minimizer Low Power Permissive Setpoint Calculation NEDC 92-50R, Revision 3C1 (38 Pages)

Cooper Nuclear Station, Docket No. 50-298, DPR-46 Information Only

COPY L--r- .

Page'l of 38

Title:

Rod Worth Minimizer Low Power Permissive Calculation Number: NEDC 92-050R Setpoint Calculation CED/EE Number: CED 6010820 System/Structure: RFC Setpoint Change/Part Eval Number: N/A Component: RFC-PLC-RVLC Discipline: Instrument and Control Classification: [x I Essential; [ ] Non-Essential SQAP Requirements Met? [ ] Yes; [X ] N/A Proprietary Information Included? [ ] Yes; [X] No

Description:

"3 , Ct i ý - ' - o -(

Revisiong'fto this calculation evaluates the impact of the CED 6010820 modification to the Rod Worth Minimizer Low Power Permissive Setpoint. CED 6010820 replaces both of the existing Rod Worth Minimizer (RWM) Feedwater Flow input transmitters (RFC-FT-50A and B) with Rosemount3051S transmitters. Additionally, CED 6010820 eliminates the obsolescent GEMAC analog modules that composed the Feedwater Flow channel and incorporates their functionality in the RVLCS TRICON logic. The Main Steam Flow channel analyzed by this calculation is also modified by CED 6010820. The circuit board on each of the four Steam Flow transmitters used to input the steam flow signal to the Main Steam Flow interlock are replaced by the CED, making this transmitter smart. Additionally, CED 6010820 eliminates the obsolescent GEMAC analog modules that composed the Main Steam Flow channel and incorporates their functionality in the RVLCS TRICON logic. This revision also changes the Title and Component associated with this calculation. Based on the extent of the changes to this calculation this revision is considered a complete re-write.

Conclusions and Recommendations:

The CED 6010820 installation of channel components with superior performance characteristics does not require a change to the Allowable Value or Setpoint information contained in Section 9.0. However, the calibration procedures and setpoint documentation requires changing based on the change of engineering units from mV to mA.

Rev 3C1: Revision to scale the % Rated Thermal Power (RTP) limits specified in the calculation to account for the 1.6% increase in Thermal Power that is realized with the Appendix K Power Uprate per EE 07-01. The values of the Analytical Limit (AL), Allowable Value (AV) and setpoints do not change, only the corresponding equivalent % RTP is changed. For purposes of clarification, the originally licensed RTP of 2381 MWth will hereafter be referred to as RTPo, and the Appendix K licensed RTP of 2419 MWth will be referred to as RTPK.

Conclusions and Recommendations:

Appendix K Power Uprate (EE 07-01) does not require a change to the Allowable Value or Setpoint information contained in Section 9.0. However, Technical Specifications will require revision to change the Rated Thermal Power values from RTPo to RTPK.

3C1 3 Ma E.Unruh

_ u-a M'Dwya

_____4 _____ 1',kz-a-07 1181 1

3 3 Alan Able " R. Krause /Ralph Krause W. Frewin 12/27/05 12/29/05 1/6/2006 1/6/2006 2 1 Alan Fanning 2/24/99 Ralph Krause Ralph Krause Wes Frewin Alan Able 2/25/99 2/25/99 2/25/99 2/25/99 1 1 E. M. Sverdrup R. D. Brussard R. D. Brussard Alan Boesch 2/1/93 2/3/93 2/3/93 2/3/93 Rev. Approved Number Status Prepared By/Date Reviewed By/Date IDVed By/Date By/Date Status Codes

1. Active 4. Superseded or Deleted 7. PRA/PSA

2. Information Only 5. OD/OE Support Only

3. Pending 6. Maintenance Activity Support Only

Page: 2 of 18_

NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATION CROSS-REFERENCE INDEX ITEM REV. PENDING CHANGES DESIGN INPUTS NO. TO DESIGN INP NO. N.T EINIPR 1 USAR Figure I-1-1 xxi2 N/A 2 Permutit Drawing 501-51952 1 3 Permutit Drawing 556-26386 N01 4 Permutit Drawing 528-51499 0 NA 5 Permutit Drawing 556-26811 4 'NI 6 GE Spec 21A1058 4 N/A 7 GE Spec 21A1058AR 2 N/A GE Spec 21A1058AT tl\ IN 9 GE FDI-71/10100 N/A 10 NEDC 00-095A -_,4 __N/A 11 CED 6010820 I< CCN#1 12 EQDP 46 N/A 13 GE-NE-0000-0063-6433-RO ( k

0 N/A Instrument Setpoints)

_y

Page: 3 of 1 NEDC: 92-050R Rev. Number: 301 Nebraska Public Power District DESIGN CALCULATION CROSS-REFERENCE INDEX ITEM AFFECTED DOCUMENTS R -

NO.

1 Procedure 14.15.2 2 Procedure 4.2 r --

3 Procedure 6.RWM.302 10 4 Procedure 2.1.4 96 5 Procedure 2.1.9 38 6 Procedure 2.1.1 126 V

Page: 4 of _38 NEDC: 92-050R Rev. Number: I 3C1 The purpose of this form is to assist the Preparer in screening new and revised design calculations to determine potential impacts to procedures and plant operations.@'

SCREENING QUESTIONS YES NO UNCERTAIN

1) Does it involve the addition, deletion, or manipulation of a [] [X]

component or components which could impact a system lineup and/or checklist for valves, power supplies (breakers), process control switches, HVAC dampers, or instruments?

2) Could it impact system operating parameters (e.g., temperatures, [

flow rates, pressures, voltage, or fluid chemistry)?

3) Does it impact equipment operation or response such as valve []

closure time?

4) Does it involve assumptions or necessitate changes to [] [X] [I]

of operational steps?

5) Does it transfer an electrical load to a diffe I t II [X] [I when electrical loads are added to or remo fr \e [t]

during an event?

6) Does it influence fuse, breaker, or relay coordi aa, [] [X] I]

7) Does it have the potential to affect th conditions of the I I] [X] I environment for any part of the Rea rIB g, Containment, or Control Room?

8) Does it affect TS/TS BasIX]U er Licensing Basis X] I I documents?

9) Does it affect DCDs? [] X] []

10) Does it have the po & fect procedures in any way not already mentione e teview checklists in Procedure EDP-06)? [] [X]

If so, identify If all a then additional review or assistance is not required.

If a w r re YES or UNCERTAIN, then the Preparer shall obtain assistance from the yS\ ý e .ý er and other departments, as appropriate, to determine impacts to procedures and n1t p ions. Affected documents shall be listed on Attachment 2.

Y

Page: 5 of _18 NEDC: 92-05OR Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET References General

1. CNS USAR, Figure I-1-i and Section VII-16.3.3.1, loep xxi2.

2. CNS Technical Specifications, Amendment 213, SR 3.3.2.1.6, and Bases 3.3.2.1, loep 4/19/05

3. W. H. Cooley, J. L. Leong, M. A. Smith and S. Wolf, General ElectricInstrumentSetpoint,1 k Methodology, NEDC-31336P-A, General Electric Company, San Jose, CA, September '9

4. W. H. Cooley, Setpoint CalculationGuidelinesfor the Cooper NuclearStation, EDE- - 9lo Rev. o, General Electric Nuclear Energy, San Jose, CA, January 25, 1991.

5. Not Used.

6. ANSI/ASME PTC-19.1-1985, ASME PerformanceTest Codes, Supplem n on Ins t an ApparatusPart1: Measurement Uncertainty,April 30, 1986.

7. ASME MFC-3M-1989, Reaffirmed 1995, Measurementof FluidFlow Pi s Orifice, Nozzle, and Venturi, January 31, 1990.

8. Crane Co., Technical Paper No. 410, Flow ofFluids through 1 s and Pipe.

9. NIST Standard Reference Database io, Version 2.0, NIST/A ~S1 cam P) perties.

Procedures lo. CNS Engineering Procedure 3.26, Rev. 20, "Inst e eor(, ontrol."

11. CNS Engineering Procedure 3.26.3, Rev. 5, "Ins t oint and Channel Error Calculation Methodology."

12. CNS Instrument and Control Procedure 14 u, .ev. 13, IAC Test Gauge Calibration.

13. CNS Instrument and Control Procedu *5_"*_ 13, "RFC System Calibration."

14. CNS General Operating Proce e, Start Procedure."

15. CNS Abnormal Procedure 2. K .o, "Extraction Steam Abnormal."

Correspondence

16. GE Letter, C97o111, to C nruh), "Generic Changes to CNS Setpoint Calculations,"

January 11, 1997.

17. GE Letter, C9701 ark E. Unruh), "CNS Setpoint Rounding Convention," January 9, 1997.,

ED

18. GELetter, I! 1)ý,ctibd2,1996.i G4fý2 S (Mark E. Unruh), "Compilation of Accuracies for CNS Calibration

19. S n hon Location.

.A 2Q Drawing 2004, Sheet 3, Rev. N44, "Condensate and Feed Water Systems Flow 1 4 Roe Drawing 2041, Rev. N76, "Reactor Building Main Steam System Flow Diagram."

22. Drawing 791E251, Sheet 1, Rev. No5, "Rod Worth Minimizer System Elementary Diagram."

23. GE Drawing 791E251, Sheet 2, Rev. No6, "Rod Worth Minimizer System Elementary Diagram."

24. GE Drawing 791E251, Sheet 3, Rev. No5, "Rod Worth Minimizer System Elementary Diagram."

25. GE Drawing 791E251, Sheet 4, Rev. No3, "Rod Worth Minimizer System Elementary Diagram."

26. GE Drawing 791E257, Sheet 1, Rev. N13, "Feedwater Control System Elementary Diagram."

Page: 6 of .18 NEDC: 92-05OR Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

27. GE Drawing 791E257, Sheet 2, Rev. N12, "Feedwater Control System Elementary Diagram."

28. GE Drawing 791E257, Sheet 3, Rev. N13, "Feedwater Control System Elementary Diagram."

29. GE Drawing 791E257, Sheet 4, Rev. N26, "Feedwater Control System Elementary Diagram."

30. GE Drawing 731E611, Sheet i, Rev. No4, "Primary Steam Piping."

31. Permutit (Simplex) Drawing 501-51952, Rev. 1, "Pressure Correction Factor - Steam Flow."

32. Permutit Drawing 556-26386, Rev. Nol, "Outline & Dimensional Data Steam Flow Restjri4

33. Permutit-(Simplex) Drawing 528-51499, Rev. o, "Flow Curve -Serial No's T-12125 -&212

34. Permutit Drawing 556-26811, Rev., "Outline &Assembly Type T. G. Flow Nozzle'.

35. GE Field Disposition Instruction 71/10100, Rev. 2, "Feedwater Flow El1 ent

36. CNS STP 95-o86, 9/1/98, "Ultrasonic Feedwater Flow Assessment."

37. Startup Test Results Report 20-1, 2/4/75, "ST1-2o Steam Production."

38. GE Specification 21A1058, Rev. 4 and 21A1058AR, Rev. 2, "Steam o Element. GE Specification 21A1058AT, Rev 1, "Flow Element Component' Vendor and Qualification Data

39. EQDP 46, Rev. 9, "Environmental Conditions."

40. Vendor Manual 1045, Rev. 5, "Model 86ooA Fl D .

41. Vendor Manual 0568, Rev. 5, "Rosemount 1151 P s utters and 1151 Smart Pressu te Transmitters."

42. TRICON System Accuracy Specifications, DN.86-534, Rev 1

43. Vendor Manual 2o16, Rosemount Modelýýo e Transmitters, Rev 1

44. NEDC 00-095A, EQ Normal Tqperat " Humidity, Pressure and Radiation, Rev 4

45. CED 6010820, Reactor Vess9 ILe,, Coo ystem -Phase 2

46. Kepco HSP Power Supply Tenaalc a"

47. GE-NE-0000-0063-6433-R ra k lic Power District Cooper Nuclear Station Thermal Powerr Optimization - Task T050*6: s t Setpoints", Revision 0, October 2007.

n~t

Page: 2_ of _18 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET R eferences ................................................................................................................................ 5

1. PURPOSE ...................................................................................................................... 8

2. REQUIREMENTS .................................................................................................... 8

3. ASSUM PTIONS ................................................................................................. 1 9

4. METHODOLOGY ........................................................................................

5. Evaluation of Main Steam Flow Device Uncertainties.................

6. Determinationof Main Steam Flow ChannelSetpoint Margin ............... ý

7. Evaluation of FeedwaterFlow Device Uncertainties....................... . . 29

8. Determinationof FeedwaterFlow ChannelSetpoin ar i....... .............. 33

9. CONCLUSION ........................................................ 8 9..CONCLUSION ................................... 38

Page: 8 of -8 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

1. PURPOSE 1.1. The purpose of this calculation is to determine the required setpoints for the Feedwater Flow Interlock and the Main Steam Flow Interlock which provide permissive input signals to the Rod Worth Minimizer (RWM).

1.2. The digital outputs to the RWM are set according to a scaled 4 to 20 mA signa proportional to mass water or steam flow. On increasing flow the RVLCS Tnro actuates the RWM Digtial Output to provide a signal to bypass the Rod BIi-Thctoi the RWM. This calculation evaluates the uncertainty in each fl eas6r~entthe establishes each setpoint to ensure compliance with the limiti i valuN i e eh**

e" a z Specifications.

2. REQUIREMENTS 2.1. This calculation is performed in accordance with ST ngime ing Procedure 3.26.3, Instrument Setpoint and Channel Error CalculationMe'hod>logy. (Ref. 11).

2.2. The Analytical Limit (AL) of lo% of the originally W e ) t hermal Power (RTPo) used in this calculation is based upon t &ni'SpDcn'6o'u rveillance requirement (Reference 2): "Verify the i'1t axed when THERMAL POWER is _ 1O% RTP." For the purposes of a s int (1culan this is interpreted as a permissive actuation, such that the bypass ncti is permitted when steam and feedwater flow greater than flow equivale l~t' RTPo. Rated steam flow is 9.558 x 106 lbs/hr or 9.558 Mlb/hr and fe wtler flow is 9.521 Mlb/hr (Reference i).

Consequently, the Rated Steam o ALMso is 0.9558 Mlb/hr mass flow and the Rated Feedwater Flo-'i6 .9521 Mlb/hr mass flow.

2.3. Based upon a reference em rt oY68 'F, the conversion factor for pressures in inches water colum 1 '" s-hWCi 1 psi = 27.7276 inA e e 6).

2.4. For clarificati I. :, ily licensed RTP of 2381 MWth will hereafter be referred to as RTPo, and T v-icensed RTP of 2419 MWth will be referred to as RTPK.

Page: g_ of .18 NEDC: 92-05OR Rev. Number: 301 Nebraska Public Power District DESIGN CALCULATIONS SHEET

3. ASSUMPTIONS 3.1. GE does not specify the accuracy of the feedwater flow elements RF-FE-i1A & B. Based upon the calibration test report (Reference 35) and the results of STP 95-086 ultrasonic flow assessment (Reference 36), a PEA of 0.5% of flow is assumed for this calculation.

3.2. The environmental temperature around transmitters RFC-FT-5oA & B during calibrations is assumed to be from 65 OF to loo OF. This range is 25 OF less than e-,. to 0OO OF normal range (References 39, 44). The equivalent value for MS-FT-u is65 OF to 104 OF. Normal temperatures are 40 to 104 OF (References 3 , 4 LI conditions are the same as normal for both channels.

3.3. The transmitters MS-FT-51A-D may be exposed to pressures 4p\týo their v ssu e limit of 2000 psi (Reference 41) without damage. In flow applications, th-es instruments are not exposed to differential pressures exceedingtI e'b e Range Limit (UR). Over-Pressure Effect (OPE) is zero for this c culati*rý 3.4. The transmitters RFC-FT-5oA/B may be exposed to sute4 ,pto their overpressure limit of 3626 psi (Reference 43) without damage. I nT 6)app]i ions, these instruments are not exposed to differential presse W Upper Range Limit (UR). Over-Pressure Effect (OPE) is ze r al u7mly<\ K.

3.5. Static Pressure Effects (SPE) for the M 11 r stpilters are specified by the vendor (Reference 41). The span effect (SPEs) i ao be corrected, however the zero effect (SPEz) is not.

3.6. Static Pressure Effects (SPE) fore o. el 3051S--CD-4A transmitters are specified by the vendor (Reference 43). T~he~s~,ec~ (SPEs) is = +/-o.1% of reading per 1OOO psi or maximum of 0.020 no **d h e\'r t (SPEz) is -0.035%x URL per lOOO psi.

3.7. The other istrume n10 ed to process pressures, therefore SPE will be zero for those devices. T in, ents are insensitive to atmospheric pressure changes, consequently ba mintri, r~ re effects are also zero.

3.8. These instrumen s nari> expected to function during or after a seismic event. Seismic Effect (,S )4i1ke> qual to zero.

3.9. These ' re not expected to function during or after an accident. Radiation E E efore equal to zero.

u dect (HE) for the Model 1151 and Model 3051S transmitters is not specified ele or, however the vendor states the humidity limit as o to ioo% relative ety Any effect is assumed to be included in the manufacturer's Vendor Accuracy.

ierefore, HE will equal zero for this calculation.

osemount specifies a Power Supply Effect (PSE) for the Model 1151DP6 (Smart) (Ref.

41) and Model 3o51S-1-CD-4A (Ref. 43) transmitters as 0.005% calibrated span per volt.

The Kepco power supply Model HSP 28-36 provides regulation of 28 VDC +/- o.1% or 28

- 0.028 VDC (Ref. 46). Therefore, PSE is negligible and will equal zero for this calculation.

3.12. The value for Vendor Accuracy (VA) is assumed to represent a 2-sigma confidence interval, unless otherwise stated. (Reference 4).

Page: lo of _18 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 3.13. The As Left Tolerance (ALT), Calibration Tool Accuracy (CTooL), Calibration Tool Readability Accuracy (CRED) and Calibration Standard Accuracy (CsTD) are controlled by loo% testing. Therefore, they are assumed to represent 3-sigma values (Reference 11).

The milliamp standard used to calibrate the Fluke 86ooA is assumed twice as accurate as the Fluke.

3.14. Drift Temperature Effect (DTE) is defined as the error due to external ten changes between calibrations (65 - 104 0 F or 39 0 F). The DTE for the trans included in the Rosemount Model 3o51S-1-CD-4A transmitter drift term.

states drift (long term stability) for the transmitters as +/- 0.20% of URL fo

+/- 5o°F (28 0C) temperature changes, o - ioo% relative humidi to a bar) line pressure. The temperature variations are within the 0 1I Therefore, DTE will not be considered separately. DTE = o.

3.15. For clarification, the originally licensed RTP of 2381 Mth wi eafter fi to as RTPo, and the Appendix K licensed RTP of 2419 MM w*,laf re1, ed to as R' Conversions from %RTPo to %RTPK will be performed a Ws.

%RTPo = 1.016x%RTPK

%RTPK = %RTPo 1.016 Conversions required to support this, below:

Page: ii of j1 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

4. METHODOLOGY 4.1. Instrument Channel Arrangement 4.1.1. Channel Diagram (Reference 20, 21, 28, 29, 45)

Main Steam Flow TRICON RWM Permissive

Page: 12 of ,18 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 4.1.2. Definition of Channels 4.1.2.1. Main Steam Flow The RWM main steam flow input has four differential pressure transmitters, each transmitter output is linearized and provided to the RVLCS Tricon where the signals are first validated and then sum ed The combined signal is supplied to the Tricon functional logic develop the RWM low Steam Flow interlock. (Reference 445

4.1.2.2. Feedwater Flow The RWM feedwater flow input has two diffeml"al p s U t transmitters, each transmitter output is lint an p o eo e RVLCS Tricon where the signals are first val-dahed an* hl summed.

The combined signal is supplied to the Trico X*n4t*°nele logic to develop the RWM low Feedwater Flow nrlo eence 45) 4.2. Instrument Definitions, Process aind Physica hit4ý s 4.2.1. Main Steam Flow Instrument Definitions Attribute VaI Reference CIC: M \ t114D 19 Manufacturer: Pel idtit 1 le 19 Model GE 2 S8AR 5e 19 Calibrated Span (SP): 81.2 pd-d 19, 13 Input Signal: ^ $.flOw - 3.0 Mlb/hr) 21,13 Output Signal: ortional to square of flow: dP pr. 21,38 Vendor Accuracy: 2o--

1.,/ rated flow 38 Attrit, btalue Reference CIC: MS-FT-51A through -51D 19 Manufactur K Rosemount 45 Model V 1151DP6 (Smart) 45 Upper R): loo psid 41 Ca ikte n ) 81.2 psid 13 Inj t Differential Press.: o - 81.2 psid 13 (0 - 3.0 Mlb/hr) tput gnal: Proportional: 4 - 20 mA 45 aet Accuracy: [0.2 + o.o5(UR/SP)]% of span 41 Temperature Effect (0.2% UR + o.18% SP) per 1oo OF 41 dor Drift o.1% UR for 6 months 41 Static Pressure Zero Effect 0.25% UR per 2000 psi 41 Static Pressure Span 0.25% input reading (Rd) per 1ooo 41 Correction Uncertainty psi Power Supply Effect 0.005% of span per volt 41 Power Supply Stability 28 V +/- .i% = 28 +/- 0.028 V 46 (Kepco)

EMI/RFI Effect o.1% of span 41

Page: 1_3 of _18 NEDC: 92-050R Rev. Number: 301 Nebraska Public Power District DESIGN CALCULATIONS SHEET Attribute Value Reference CIC: RFC-PLC-RXVL 45 Manufacturer: Triconex 45 Model TRICON 45 Input Signal: Prop. to MS AP: 4 - 20 mA (1-5V) 45 Output Signal: Contact out 45 Vendor Accuracy: 0.15% of FSR (Volts) 42 Vendor Temperature Effect Included in accuracy 4 Vendor Drift o 4.2.2. Feedwater Flow Instrument Definitions Attribute Value CIC: RF-FE-11A &*B 19 Manufacturer: Permutit (Six 19 Model GE Spec. 21. 19 Calibrated Output Span 114.9 psid (A 45 (SP): 115.6 psid (B Input Signal: FW flo: (on 45 Output Signal: Pr rhat :dP 35 Vendor Accuracy: 0 o'f sp; 3.1 Attribute lue Reference CIC: ,oA4&B 19 Manufacturer- unt 45 Model \ S-1-CD- 4A 45 Upper Range o00 psid 45 Calibrated Span ( 114.9 psid (A 45

~NQ\>~7 115.6 psid (B)

Differential Press.: 45 0 - 114.9 (A), 0.1 - 115.7 (B) psid =

(o - 1O.O Mlb/hr)

Proportional: 4 - 20 mA 45 Y: 0.04% of span 43 ature Effect [o.009% UR + 0.04% SP)% / 50 OF 43 0.20% UR for lo years 43 7Supply Effect 0.005% of span per volt 43 Pressure2 Zero Effect 0.035% UR per 2000 psi 43 Pressure Span Effect o.1% input reading (Rd) per lOOO 43 psi Power Supply Sttability 28 V +/- .1%= 28 +/- 0.028 V 46 (Kepco)

EMI/RFI Effect 0.1% of span 43 Attribute Value Reference CIC: RFC-PLC-RXVL 45 Manufacturer: Triconex 45 Model TRICON 45 Calibrated Span (SP): 16 mA 45

Page: IA of _18 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Attribute Value Reference Input Signal: Prop. to FW AP: 45 4 - 20 mA (1 - 5 VDC)

Output Signal: Contact out 45 Vendor Accuracy: 0.15% of FSR (Volts) 42 Vendor Temperature Effect Included in accuracy 4242 Vendor Drift 0 42 4.2.3. Signal Equivalents to the Analytical Limit As this calculation is only concerned with the "valuaýdý\V)v 0 Minimizer Bypass Permissive Setpoint and associated testing u'eati e Process Measurement Accuracy (PMA), Primary Elemen ecurc (PEA ad Static Pressure Effects (SPE) are only evaluated at the a1.'tic Limit. The signal unit equivalents of these values are aluaý belo-w rm Reference 7, the basic equation for flow measuremenes 's hzle or ventun is:

w=JU *C*Y*d 2 l P Where, W = mass flow Ju = units conversion constant C = discharge coefficient Y = Expansion factor d throat o le AP= =nozzle measufl-0 [iPsur p =density,,

f3= ratio o th oa "amer (d) to inlet diameter (D)

The loo45 a assuming only the differential pressure is changing and the for s.m1'pl ffes to:

ain Steam Flow Analytical Limit Equivalent Units The differential pressure (APALMS) at the transmitters assumes an equal division of flow (WALMs) in each steam line. If the Main Steam Analytical Limit, ALMs = 0.9558 Mlb/hr (paragraph 2.2):

WALMs _ ALMS =KMs A-AMS =0.23895Mlb/hr 4

3.0 Mlb/hr Mlb/hr Kms = Wmaa g CSMS V81.2 psid psidO- 5

'AP~u _r_ALMs; --- _2 = 0.9558 2*=o0.51515 psid 4 *

,ALM . 1 MS (14 xO.33292)

Where, Wma. = full scale mass flow, each transmitter KMs = steam flow constant (including density)

Page: 15 of1 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET pms = steam density for design conditions CSms = steam line flow transmitter calibrated span = 81.2 psid The current output from the transmitters (TXALms) is found from the transmitter scaling:

.o TX ALMS = 1.77559 * ( M )A+ mA ZERO

=1.77559 *(x]0.51515 )+4= 5.2744mA 4.2.3.2. Feedwater Flow Analytical Limit Equivalent UM The differential pressure (APALEW) at the traniin equal division of flow (WAL*w) in each feedxjt ýter Analytical Limit, ALFw = 0.9521 Mlb/hr (seepa Y ALwK r WALJ7W = Aw=FWK APA 2 - F 10.0 Alb/hr KFW - w W..

V\- 10 5 CSFW AALFW = 2* K = 0.26039 psid

.93291)2 Wheere, Wma = full sc~ueliaS flqw, each transmitter feewl.. Ae'turi discharge coefficient CFW desitfor design conditions S. fow constant (including density) f~dw,gter flow transmitter calibrated span K'eoutput e from the transmitters (TXALFW) is found from the sitfeer scaling (using FT-5oA span for conservatism):

W= 1.49266 * ()+mAZERo

=1.49266 *( 0.26039 )+4= 4.7617.mA

/I

2 Attribute Value Reference AC Transmitter Cal. Temp. 65 - loo OF (RFC-FT-5oA & B) Par. 3.2 A 65 - 104 OF (MS-FT-51A-D)

Calibration Interval 18 mo. +25% = 22.5 mo. 2 Transmitter Normal & Trip 40 - 100 OF (RFC-FT-5oA & B) 39,44 Temperature 40 - 104 OF (MS-FT-51A-D)

MS Process Temperature Design: 546 OF 38 (Saturated Steam)

MS Process Pressure 1000 psig 38 FW Process Temperature 250 to 367.1 OF 15, 1 FW Process Pressure 1175 psig 35

Page: 16 of __38 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 4.2.4.1. MS-Fr-51A-D Total Trip and Normal Temperature Range Total temperature range (ATrms) is the greater of the two differences between trip conditions and calibration temperatures:

. MTTmin- Tmax =40-104=-64 ATrMs M TTmax - cmin = 104 -

The normal temperature range (ATN) is assumed (Paragra ; t equal to the total trip temperature range:

Where, TT,min = minimum trip temperature, normal okac ident TT,max = maximum trip temperature, nor al or Tc,min = minimum calibration teneperattu.

Tc,max = maximum calibration terr h ATNMs = total normal temperatur*eJ a for transmitters 4.2.4.2. MS-FT-51A-D Drift Condition( Tn re ge Drift temperature r e ( DM) Nf'iference between the maximum and mini ca ition temperatures and is used to calculate the drift t ena re ect (DTETxMS):

ATDMS (Tc,max - ,mi (104 - 65) = 39 OF TTDMS )39 0 4.2.4.3. MS-FT-. 5ial Accuracy Temperature Range em ature range (ATANms) is the difference etnormal temperature range (ATNMs) and the drift ture range (ATDMS). ATANMS is used to calculate the normal auiray temperature effect (ATENMS) for the spurious trip avoidance an sis and the trip setpoint margin:

AS=ATNMs - ATDMs = (6 4 - 39) = 25 OF 2.4-4. RFC-FT-5oA & B Total Trip and Normal Temperature Range Total temperature range (ATThw) is the greater of the two differences between trip conditions and calibration temperatures:

TFT mn -- Tc,max = 40 10 = -160 a

TTTmM - Tkmin i00 -

The normal temperature range (ATNFw) is assumed (Paragraph 3.2) to be equal to the total trip temperature range:

[*TTFw = ATNFw = 60

Page: 17 of 318 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Where, TT,min = minimum trip temperature, normal or accident TT,max = maximum trip temperature, normal or accident Tc,min = minimum calibration temperature.

Tc,max = maximum calibration temperature ATNFW = total normal temperature range for feedwater ti 4.2.4.5. RFC-Fl-5oA & B Drift Conditions Temperature Range Drift temperature range (ATDFW) is the difference betwc maximum and minimum calibration temperatues anjl:

calculate the drift temperature effect (DTEmxTV .

ATDFw = (Tc,max - Tcmin) = (ioo -65) = 35 OF \

4.2.4.6. RFC-FT-5oA & B Normal Accu r'aiture Range Normal accuracy temperature sNs the difference between the normal temperatr: (4'n ýand the drift temperature range histeFa mflow and feedwater flow values are the sane, therc is used to calculate the normal accuracy te atur:' 'EN) for the spurious trip avoidance analysis an thletril margin for both loops:

ATANFW =ATNFW (-0 - 35) = 25 OF

ýN=ATms

=ATAN~w 250

5. Evaluation of Main Stea ce Uncertainties All device uncertainty cont h ýut ndom variables unless otherwise noted.

The identified accuracy i tiy nributions t are combined using the SRSS method to determine total device cu r6each device (i) under normal and trip conditions. The device accuracy is i1-(ta1izc( a 2 sigma confidence level, and is given by:

Ai 2 i 2 + 2 ) 2 h'\tem inside the square root sign are the random portions of the individual effects,

-t i~gma value associated with each individual effect.

e drift, normalized for a 2-sigma confidence level, is found using the SRSS method for rice (i) for calibration conditions.

, )

2+

2ý(

Page: 18 of 1 NEDC: 92-050R Rev. Number: 301 Nebraska Public Power District DESIGN CALCULATIONS SHEET The device calibration uncertainty is found similarly as follows:

5.1. Determination of MS-Fr-51A-D Uncertainty (Device 1)

The transmitter accuracy and drift uncertainties are calculated in terms of inp ti' (psid) and converted to output units (mA) to combine with the calibration un erta evaluating total device uncertainty.

5.1.1. Determination of MS-FF-51A-D Accuracy 5.1.1.1. Vendor Accuracy (VA,)

Value Reference 0.2 + 0.05 * %* SP VA VA1 =0.2+0.05* 10-0 j%*8iz.pid=ý d sýlý 5.1.1.2. Accuracy Temperatu 'E ct YE)

ValueSigma Reference ATEIN (0.2%UR+ 1 0

A 27 41

(NO 00f:=(0.0O25 1 . 018'*81.2)x-100

--03 9 2 0.0866 psid 5.1--3. a, .ressure Effects (SPE,)

Value Sigma Reference S26 UR per 2000 psi 0.0025*loopsid* 2000 2000 psig 2cy 41 psig

=0.1250 psid SPEicu =0.25% Rd per 1OOO psi (@ ALMs)

SPElcu = 0.0025*0.5152psid*l psig 2 41 x1000 psig

= 0.0013 psid 5.1.1.4. Power Supply Effect (PSE,)

Value Sigma Reference PSEI = o NA 3.11 5.1.1.5. EMI/RFI Effect (REE,)

Value Sigma Reference REEl = o.1% of span REE1 = 0.001

81.2 = 0.0812 psid 2CY 41

Page: 19 of j1 NEDC: 92-05OR Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 5.1.1.6. Other Accuracy Effects Value Sigma Reference SE= o NA Par. 3.8 RE = o NA Par. 3.9 HE= o NA Par. 3.10 OPE = o NA Par.-A 5.1.1.7. MS-FT-51A-D Accuracy (A,) x 2

A, (psid) +(ATEINj) +/-KSE i 2 )

+SP>IUj 2

ý11 250 2+(0.0013

=2*,t 0.2124jI 2+(0.0866 +20.1I 2 2).

= 0.2736psid A] - A1(psid) *16 - 0.2736*

CS 81.2 I =, ,I

[A1"*(psid)° L'

= 0.27.,6

v ipsid

+

(20`1I

. .jj

[iA, = 0.0539 nA (20)1 5.1.2. Determi 5.1.2.1.

V -iseualto 0.1% of UR for any 6-month period (VD,(6))

ýd for the 22.5 month surveillance interval (VD,)

the following relationship:

Value Sigma Reference

) (6

22.5 =0.00 1*100* F22.5 20 41, 11 6 6 936 psid

-1.2.2. Drift Temperature Effect (DTE,)

The drift temperature effect (DTE,) is cal eulated as shown below:

Value Sigma Reference DTE, = (0.2%UR + 0.1 8%SP)

AToMs 20 41 1UU 39 K = (0.002

100 + 0.0018

81.2) x 39 100

= 0.3462

0.39 = 0.1350 psid

Page: 20 of _18 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 5.1.2.3. MS-FF-51A-D Drift Value (D1)

D, (Psid) =2 (VDj ~DE, =2* 0~.1936 1 J2 +(0.1 3 5 0 2

= 0.2360 psid D, -D,(psid) '16= 0.2360 *16=0.0465 mA CS 81.2

.A, 5.1.3.

5.1.3.1. As-Left Tolerances The As-Left Tolerances for the following formula where n is th Sigma Reference ALT = (3/n)(VA) 3a 3,4

= (3/2)(o.212,

= 0.3186 psid

= 0.07mA a-5.1.3.2. Calibration TorcS ftriiitý 'alibrated with a o to 1OO psig Heise gauge at h 0ii rn is calibrated by a dead weight tester

\). Heise gauges are temperature compensated from

\ (Reference 18) and have the same accuracy as shown Nh-iad weight testers have no temperature effect (Reference ptput calibration tool is a Fluke 86oo and the calibration for milliamps is assumed to be at least twice as accurate.

uncertainty values in units of psi are converted to mA as follows:

mA = psi

16 CS Value Sigma Reference Heise Accuracy (CIToOL1) = +/-o.1 psig (o.1% FS) 3a 16, 18 CIlOOLI =0.l/81.2

16 = 0.0197 mA Heise Readability (CIRAD,) = +/-0.05 psig (1/2 MD) 30 16, 18 C1READ1 =0.05/81.2

16 =0.0099 mA Press. Std. Accuracy (CisTDm) = +/-0.1 psig = 0.0197 mA 3cy 16, 18

Page: 21 of a1 NEDC: 92-050R Rev. Number: 301 Nebraska Public Power District DESIGN CALCULATIONS SHEET Fluke Accuracy (C1ToOL2) = +/-(0.04% Rd + o.ol% FS)' 3y 40 CITOOL2 = 0.0004

20 + 0.0001

200 = 0.028mA Fluke Readability (CIREAD 2) = -o.oi mA (1 LSD) 3o 40 Fluke Temperature Effect(CTE,) = 30 40 .A

= +/- (O.0O3% of input + O.OO1% of range)/°C

- (0.00003 X 20 mA + 0.00001 x 200 mA) (40.0 -

18.3) (using 65°F to 104'F)

+ (0.00003 x 20 mA + 0.00001 x 200 mA) (40.0 -

18.3)

+ 0.0026 X 21.7 /

= 0-0.56 mA Milliamp Standard (CISTD2 ) = +/-o.o1 mAhA Par. 3.13 5.1.3.3. MS-FT-51A-D Calibration Uncertain Since the values of ALT, CTOOL,9ýu F are controlled by lOO% testing, they a ia values, so n = 3.

0.07 )2'(0 972 0O.OO99j 2 .0j (0.01)2 ( 0 .0 5 6 2 0.01) 2 3 3) 3) +(3)

=0.0663mA 1C, = 0.0663 mA (20) 5.2. Determination ol Uncertainty (Device 2)

The TRICON PLC ai and calibration uncertainties are calculated in terms of input units (mA),v device uncertainty. Only the 370oA module has error terms associated~l 42) 5.2.1. De- "em Výt U1 a'ý x i,1. Vendor Accuracy (VA2) o> Value Sigma Reference VA 0O.15% of FSR (Volts) 2a 42 0

= 0.0015 of5VDC = 7.5mV

= o.o3mA (input) 5.2.1.2. Accuracy Temperature Effect (ATE 2)

Value Sigma Reference ATE 2 = 0 NA 42 5.2.1.3. Static Pressure Effects (SPE 2)

Value Sigma Reference SPE2 = o NA 3.7

' Fluke 86ooA is used on the 200 mVdc range to indicate mA. Reading is instrument full scale (20 mA). Accuracy specification is from 15 to 35 0 C (59 to 95 OF).

Page: 22 of j1 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 5.2.1.4. Power Supply Effect (PSE)

Value Sigma Reference PSE = o NA 42 5.2.1.5. Other Accuracy Effects Value Sigma Refei SE = o NA 42 RE = o NA HE=o NA REE =o NA 5.2.1.6. RFC-PLC-RXVL Accuracy (A2 )

-+

A2 = 2 * ý(VAn 2 ) 2 +(ATEn 2), +ý8PE n 2),2 (o2 (o2 k(2) + (),

= 0.03 mA

= 0.03 mA(2) I 5.2.2.

Sigma Reference NA 42 5.2.2.2.

Sigma Reference NA 42 5.2.

0 mV JDL =o.o mA (2al) 5.2.3.1. As-Left Tolerances The As-Left Tolerances for the TRICON is established by use of the following formula where n is the sigma value associated with VA.

Value Sigma Reference ALT = (3/n)(VA) 3a 3,4

= (3/2)(o.o3)

= 0.045 mA

= o.o5mA (based on Fluke Readability)

Page: g3 of q8 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 5.2.3.2. Calibration Tool Uncertainty The TRICON uses the output of the transmitter as the calibrating input signal to the 370oA module. The internal components within the TRICON are self calibrated and produce a signal that provides a contact closure. Therefore, no calibration inaccuracies are required for the TRICON PLC.

((CTooL , CREAD and Cs-jf=-`( I' 5.2.3.3. RFC- PLC-RXVL Calibration Uncertainty Since the values of ALT, CTOOL, CREAD, and Csq Y co l oo%*

testing, they are assumed to represent 3 sigllh values, 0 2 0 2

.0 5 2 0 .2 1(0

=-0.0333mA 5.3. Determination of Main Steam Flo " C.

2 =o- 033 A()

5.3.1. Individual Loop Accuracy For this instrument loop,jere are vo evices in the loop; therefore, the loop accuracy (ALN) equals:,

ALN = P~ g ALN = 0.b5A9 0jd~ = 0.0617mA JJLN = O.O617 mA (20)J

6. Determinati Flow Channel Setpoint Margin 6.i. Deterndi ýPIEA and PMA iracv (PEA)

Pjer reference 38, the flow element accuracy is rated at "2 percent of rated flow."

ýThis specification could be interpreted as a constant error equivalent to 2% of the full scale flow rate at RTPo, however as shown in reference 6, the uncertainty of a flow element is a function of the nominal rate. For this

$ calculation, the accuracy of each steam flow nozzle is 2% Rd RTPo (flow). As in paragraph 4.2.3.1, the flow rate corresponding to ALMso is:

WALMS = 0.23895 Mlb/hr The flow equivalent to 2% Rd RTPo uncertainty in lb/hr at ALMs (lo% RTP) is:

PEA = +0.02

ALMSO = +0.02

0.23895 Mlb/hr = +/-4779 lb/hr Considering only an increase in AP corresponding to a positive error, the equivalent change in output for the transmitter in mA is:

Page: 24 of _8 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET WALMS + PEA = KMS

VAP +PEA , =0.23895 + 0.004779 = 0.24373 Mlb/hr APALs + PEA = K +p PEA ) 24373 0.53597 psid KM5 9 0.33292)

PEA =

-A (APAI*MS +PEA dP ) - APes *16 0.53597 -0.51515 16,

= 0.0041 mA 6.1.2. Process Measurement Accuracy (PMIA)

For this application, three significant error contributj nces and 6). Variations in the discharge coefficient due to chan gi lp tes, changes in the steam expansion factor due to pressure va r ns and-fl 6w rate, and finally changes in density due to pressur~n 1 ture variations. From reference 7, the mass flow equation is:

W=Ju *C*y*d _#

Where, W = mass steam flow in lbm/ s Ju = units conversion constant 0. 990190 C = discharge coefficie -;. - o.oo653 * °6:*fl Y = Expansio nSt d = nozzle a etý meter = 10.914 in (Reference 32)

AP=meas i e+di\foq ial pressure ininWC (at 68 OF) 13= " 1 r t diameter (d) to inlet diameter (D) = 0.5o6 (References 32, 22738

W j1d's number = D scosity in centipoise = 0.094 (Reference 8)

-- pipe upstream diameter = 21.564 in (Reference 32).

According to reference 7 (paragraph 8.5.1), the discharge coefficient is within 2% if certain fabrication conditions are satisfied and the Reynold's number is:

10 <RDo6 106 This is the PEA evaluated in paragraph 6.1.1 and if the Reynold's number is within this range, the uncertainty of the discharge coefficient (C) is included in PEA.

At WALMS = 0.23895 Mlb/hr or 66.4 lb/sec. the Reynold's number is:

RD _ 22738

66.4 = 7.45 105 0.094

21.564

Page: 25 of 38_

NEDC: 92-05OR Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET If the density decreases to some new value pmin, then for the same flow rate, a higher differential pressure is present or alternately for a constant AP, the actual flow is lower. Therefore a decrease in density results in a positive error in differential pressure and signal input to the alarm module. In the same way a decrease in the steam expansion factor to (Ymin) is equivalent to a positive error in differential pressure. , Ak Collecting all terms, including Cwhich are assumed to be fixed as constant flow:

kMs

yMs

C pms

= k.i

YAmn

L **pmnn Ys* LI)'

PMS = y.i APmini

Pmin m

PMS

YIS

/iPmin = ALPc Pmin Y2min PMAdP = APmin _ APC = PRCPMS POmin i

Y y.

2m Where, kMs = steam flow constant I?> 7\ V j pMS = steam density for desig e ibration cop itions Pmin = minimum steam density niam pressure conditions YMS = steam expansion factor for eý, alibration conditions Ymin = steam expansion fa'xtor minimum pressure conditions APc = differential presgl'{\r d0)gn calibration conditions APmin = rdiffmntiainiimumfe*

!* Imimmum pressure conditions PMVAip = diff~ertit1i presuý error limit From refe ti-n 82, the minimum reactor pressure is c)26 psig (941 psia). Usin .....pssthe minimum pressure, the minimum density (pmin) is 2.093 l}, k~e ý9). From reference 38, the design conditi ons are for 546 OF and 1015 psia. The design density (pMs) is then 2.280 9).

Ni*ihft pressure of 1015 psia (PMs) and a lOO% power differential pressure

,lb/hr of 51.51 psid, AP/PMs = 0.05 and YMS =0.97 (Reference 8, A-21).

Sinput pressure of 926 psia and APc = APALMS = 0.51515 psid, AP/Pmin =

and Ymin =0.99 (Reference 8, A-21)

,is found from:

APs *(~P~mSi

-aMdp *inyYMs mLMS 1* - 0.51515r *2.280*

21 0"9409 2.093

0.9801 / -0.04578psid PMAA- P '16

MdP *004578'16

= 0.0090mA CS 81.2 This error is always present at startup by procedure, consequently this factor should be treated as a bias.

IPMArn = 0.009 mA (bias)]j

Page: 26 of 1 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 6.2. Determination of Other Uncertainty Terms 6.2.1. Total Loop Accuracy for the Four Steam Flow Loops For this function, there are four individual loops of steam flow. These individual errors are combined as follows:

ATOTAL = */4(ALN 2 )

AToTAL=

4(0.06172) = 0.1234mA 6.2.2. Total Calibration Accuracy for the Four For this function, there are four individ individual errors are combined as follo CTOTAL = -4(C 12 )+ 4(C 2 2 )

CTOTAL = /4(0.06632)+/-+24 ICTOTAL = 0.1484 mA (21) 6.2.3. Total Drift for the Fourqt(TH. Fow Loi For this function, the e o iadivic loops of steam flow. These individual e. a .s e s follows:

DTOT4 55) +4(02) 0.(

((DTOTAL = 0.093 mA (2a)J All sig f' - I tSteam Flow uncertaintyv ten us are accounted for in the previous The setpoint margin is defined (Reference 11) as the margin between the nominal setpoint and the analytic limit. This margin is given by:

SM = (1.645/N)(SRSS OF RANDOM TERMS) + BIAS TERMS Where N represents the number of standard deviations with which all the random terms are characterized (normally 2) and 1.645 adjusts the results to a 95% probability for a single side of interest.

SM = 1.645 VA,'oAL2 + CToTAL2 + DTOTAL 2

+ (PMAY+ (PEA)2 +/- (bias terms) n

Page: 27 of 3j NEDC: 92-05OR Rev. Number: 301 Nebraska Public Power District DESIGN CALCULATIONS SHEET SM= 1.645 0.12342 + 0.14842 +0.0932 +(0.0)2 +(0.0041)2 + (0.009) 2 ISM =o.185 mA (2cF) 6.3.2. Nominal Trip Setpoint (NTSP1) Calculation The Nominal Trip Setpoint (NTSP1) for process variables that decrease4ot'ttti or for increasing permissives is given by:

NTSP = AL +SM From paragraph 2.2 the Analytic Limit (AL) is:

AL = 0.9558 Mlb/hr,-

Which from paragraph 4.2.3.1 is equivalent to:

SUMALMS = 5.2744 mA Therefore the nominal setpoint is:

NTSPxm = 5.2744 + o.185 =5.46 mA OR.

NTSPl (NTSPlmA. -mAZRO)

lax, 2.0 1.095.Mlb/hr 16 Where, E Wmax = combined total iil sqle flow for the channel As a percentage of is 1.095/9.558 = 11.46% RTPo (11.28% RTPK).

NTSP1 repr se' ts >Icy (closest to AL) at which the setpoint can be set.

6.3.3. Allowable Value Ca lelation The fud by this calculation do not have a required limiting set "I Ith chnical Specifications, consequently calculation of an za*1e.lue (AV) is not applicable.

6.3.4. > 1 ance Evaluation XT'heipirpose of the LER Avoidance Evaluation is to assure that there is ijfficient margin provided between the Allowable Value and the Nominal Trip

ýSetpoint to reasonably avoid violations of the Technical Specifications. Since there is no AV, the LER Avoidance Evaluation is not required.

\'-"5. Selection of Operating Setpoints The setpoint could be changed to NTSP1 = 5.46 mA, however the existing setpoint of 6.56 mA (20.1% RTPo or 19.78% RTPK) is conservative and should be preserved to avoid unnecessary changes to the plant, station procedures or design documents. The recommended operating setpoint is:

Os= (OSA- mAZERO) *I W. = 6.56-4*'12.0 = 1.92 Mlb/hr 16 16

Page: 28 of j1 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Where, I Wmax = combined total full scale flow for the channel As a percentage of power, OS is 1.92/9.558 = 20.o9% RTP (19.78% RTPK)

OSP = NTSPi +/- LAT OSP = 6.56 mA (20.1% RTPo or 19.78% RTPK) +/- LAT 6.3.6. Establishing Leave Alone Zone (LAZ)

These loops are not subject to periodic functional testing betweene rV s.

For the purposes of this calculation, the LAZ or Leave Alne To anc )

for RFC-PLC-RXVL is equal to the As-Left Toleranc '1U 6.3.7. Required Limits Evaluation The Required Limits Evaluation (RLE) c hjilat su-djustment to NTSP for the case when NTSP is set at the center of th ee ve ýle zone. The adjustment assures that with the stack-up of the unc a s (inclA ing [AT) for all the devices in the loop, there is en.gh..m r e peJŽ'Action Avoidance (or LER avoidance). Since there s noLE>\a *>iat with the setpoint, the RLE is not required.

6.3.8. Spurious Trip Avoidance (SPA ,atin There is no SPA evaluati , quired for a permissive such as the RWM Rod Block Bypass. Xý 6.3.9. Determinati&nofAP EII etpoint Since IeatinCorrection (EC) for a flow measurement:

Mlb/hr = 20.1% RTP (19.78% RTPK).

Page: 2c9 of _18 NEDC: 92-05OR Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

7. Evaluation of Feedwater Flow Device Uncertainties All device uncertainty contributions are random variables unless otherwise noted.

Uncertainties are calculated as in sections 5.o and 6.o for Main Steam Flow.

7.1. Determination of RFC-Fr-5oA & B Uncertainty (Device 3)

The transmitter accuracy and drift uncertainties are calculated in terms of input 1lt (psid) and converted to output units (mA) to combine with the calibration unce a*-mn evaluating total device uncertainty.

7.1.1. Determination of RFC-FT-5oA & B Accuracy 7.1.1.1. Vendor Accuracy (VA3) g e n Value igma) eference VA3 = 0.04% of span 3 43 VA3 = 0.0004

115.6psid = 0.0462 psi\

7.1.1.2. Accuracy Temperature Effect (ATE,\

Value igma Reference ATE3N = (0.009%UR +0 04:-P) 25 3a 43

=(0.00009

300 + 10.6) x -

50

= 0.0733

0.5.',7 ysid 7.1.1.3. St tik Pre u e Eff-t: SPE 3 )

,alu Sigma Reference SPE3z = o.o)' 1r.1 0bopsi SPE3z T(-b.)

SP 3 ZOsi*1175,0Pi psig 3cy 43 2000 psig psid P = %Rd per 1OOO psi (@ ALFw)

.001*0.2604psid* 1175 psig 30 43 01000 psig

= 0.0003 psid 7.1.1.4. Power Supply Effect (PSE 3)

Value Sigma Reference PSE3 = o NA 3.11 7.1.1.5. EMI/RFI Effect (REE 3)

Value Sigma Reference REE3 = o.1% of span REE 3 = 0.001

115.6 = 0.1156psid 30 43

Page: no of a1 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 7.1.1.6. Other Accuracy Effects Value Sigma Reference SE = o NA Par. 3.8 RE = o NA Par. 3.9 HE= o NA Par. 3.10 OPE = o NA Par. .

7.1.1.7. RFC-Fr-5oA & B Accuracy (A3)

SPE3CU 2 A, (Psid) = 2 * .1<Aj + (ATE 3 N)2+ (SpE3 Z ) +. +

+(0.0617 2 0.0003 )2

=2 0~.0462l)2+0.0367 )2 )

3)

= 0.0958 psid A3 - A3 (psid)

16 _ 0.0958 *16 = 0, CS 115.6 __

Ik~f(psid)

= o.0958 p)sid (2a)II q 11A3= 0.0133 mA (2(0) 7.1.2.

Therefore, VD is equal to:

Sigma Reference VD, 30y 43 Effect (DTE 3) is:

Value Sigma Reference

==0 NAa 3.14

'.1.2.3. RFC-FF-5oA & B Drift Value (D 3)

2* (VDj

0.

=1541psid D3 = D3 (psid), 16 = 0.1541 *16 = 0.02133 mA CS 115.6 JID 3 (psid) = 0.1541 psid (2C)I ID3 = 0.02133 mA (2C)]

Page: 3a of18_

NEDC: 92-05OR Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 7.1.3. Determination of RFC-FT-5roA & B Calibration Uncertainty 7.1.3.1. As-Left Tolerances The As-Left Tolerances for the transmitter is established by use of the following formula where n is the sigma value associated with VA. For conservatism VA will be represented as 2 sigma value instead of sigma value Value Sigma r 'enc" ALT = (3/n)(VA) 3cy 14

= (3/2)(O.O462)

= o0.693 psid = o.0096 mA o.olmA (based on Fluke Readabilit)

= 0.°2mA (rounded conservatively) 7.1.3.2. Calibration Tool Uncertainty These instruments are calibrate 20 psig Heise gauge at the input, which in turn is calibratb - dweight tester (Reference 12 & 13). Heise gauges ari etmpe ure compensated from

-25 OF to 125 OF (ReferenceC*18)' nd\hnReferenc tett *rtur Ine accuracy eff as shown below. The dead weig &ýv\epe at effect (Reference 18). The output calibration too' 1 luke 86oo and the calibration standard for milliamn osi's aIssu*nd tobe at least twice as accurate.

Input uncertainty valu eQsJm of psi are converted to mA as follows:

mA = psi*

e Sigma Reference Heise AccuracN\A. , -_+/-0.2 psig (o.1% FS) 30 16, 18 C 3TOOLI

96 1 = 0.0277 mA Heis ýa Aq 1-'C 3 READ1) = +/-O.1 psig (1/2 MD) 30 16, 18 R.< 15.6*16=0.0138 mA s ccuracy (C 3STD1) = +/-0.2 psig = 0.0277 mA 30 16, 18 FlJuke' ccuracy (C 3 TOOL2) = +/-(O.O4% Rd + o.o1% FS) 30 40 0oL2 = 0.0004

20 + 0.0001

200 = 0.028 mA

'Fluke Readability (C 3 READ2) = +/-O.Ol mA (1 LSD) 3cy 40

Page: 3_g of 1 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Value Sigma Reference Fluke Temperature Effect(Cim 2 ) = 30 40

= +/- (0.003%of input + o.oo1% of range)/°C

+ (0.00003 x 20 mA + 0.00001 x 200 mA) (40.0 -

18.3) (using 65°F to 104 0F)

+ (0.00003 x 20 mA + 0.00001 x 200 mA) (40.0 -

18.3)

+ 0.0026 X 21.7

= +/-0.056 mA Milliamp Standard (C3ss2) = +/-o.ol mA 7.1.3.3. RFC-Fr-5oA & B Calibration Uncertainty Since the values of ALT, CTOOL, CAD and4 ai co 1 itolled by ioo%

testing, they are assumed to rep pt.4.se} values, so n = 3.

C3 =2 56 0=0. 8 (0m1)D]

ý 0.O5268MA ri JC3 = 0.05268 mA(2r)ý 7.2. Determination of RFC-PLC-RX-VLD Uncertainty (Device 4)

The TRICON PLC accuracy w as-Rrv%*o valuated in Section 5.2. The results of this evaluation are shown(low -\'N 7.2.1. RFC-PLC- t

= 0.03 mA(20) 7.2.2. RFC=PL - Drift Value (D JD4 =0o0omA ý(20O 7.2.3(ý ALTt 7-' he /Tolerances (ALTA l[LT4 =0.5mA (2(0)

Calibration Error (C4) 1C4 =0.0333mA(0

> Determination of Feedwater Flow Loop Accuracy 7.3.1. Individual Loop Accuracy For this instrument loop, there are two devices in the loop; therefore, the loop accuracy (ALN) equals:

ALN= A3 + A4

Page: 33 of ._8 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET ALN= X0.01332 + 0.032 = 0.0328mA

[*LN = 0.0328 mA (2C)]

8. Determination of Feedwater Flow Channel Setpoint Margin 8.i. Determination of PEA and PMA 8.1.1. Primary Element Accuracy (PEA) K For this calculation, the accuracy of each feedwater flow ventu (paragraph 3.1) to be 0.5% Rd RTPo (flow). As in pargak-p2, M rate corresponding to ALiwo is:

WAFvw = 0.47605 Mlb/hr The flow equivalent 0.5% Rd RTPo uncertainty in , RTPo or 9.84% RTPK) is:

PEA = +0.005

ALFWO = +/-0.005

0.47605 Mlb/h, 38. lb/hr Considering only an increase inAP-corfeý,o i itive error, the equivalent change in output f*0 ti s \:-t r'mA is:

WALFW + PEA = K FW

VAP + PE 0 4760 + 0.00238 = 0.47843 Mlb/hr APALFW + PEA dP = ALFW +PA= 0.26300 psid KFW 0.93291 PEA m-'APALF-AAF<+ !CAP X-k I 1\)W *16 =115.

_ 0.26300 -0.26039 *16

\\V 115.6 EAmA = 0.000361 mA (2a) 8.1.2. Process aement Accuracy (PMA) 1p p ication, two significant error contributors exist (References 7 and

_)..nc, ins in the discharge coefficient due to changing flow rates and

-c ng/s in density due to pressure and temperature variations. From reference 7 , tle mass flow equation is:

W=J*C*Y*d2 AP* p Where, W = mass steam flow in ibm/sec J, = units conversion constant = 0.09970190 C = discharge coefficient = 0.9975 - 0.00653 * -*/3 SRD Y = Expansion factor = 1 for liquid d = nozzle throat outlet diameter = 7.941 in (Reference 34)

AP = measured differential pressure in inWC (at 68 OF) p = liquid density

Page: 34 of _18 NEDC: 92-05OR Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

= ratio of throat diameter (d) to inlet diameter (D) = 0.5207 (Reference 34)

RD = Reynold's number = 22738

W

,U*D

= viscosity in centipoise = 0.132 (Reference 8, page A-2)

D= pipe upstream diameter = 15.25 in (Reference 34).

According to reference 7 (Figure 8, paragraph 9.6.1), the discharge coe ,en c is within 1%if certain fabrication conditions are satisfied and the Reyn number is:

2"10 6 h e o-2* 105 -<RD This is the PEA evaluated in paragraph 8.1.1 and if theleyold uber i"/

within this range, the uncertainty of the discharge co *in t PEA.

At WALFW = 0.47605 Mlb/hr or 132.2 lb/ s.* theReod's number is:

RD-22738 0D *132.26 1.49 lO

.132*15.25 If the density decreases to somfe n vup' hn, hen for the same flow rate, a higher differential pressure is, sen or al ately for a constant AP, the actual flow is lower. Thereforead'_ rcea e in density results in a positive error in differential pressure and signaipt.- - t the alarm module.

Collecting all terms, incitd'd and Y, which are assumed to be fixed as k~w, for a constant flow: >

kFW \

kAýc* min

min APc

Ptw rapin C ilAPC APc OMSAR,,APCIPMS P)min Pmin krw - feedwater flow constant

= liquid density for design calibration conditions pmin= minimum liquid density for minimum pressure conditions APc = differential pressure for design calibration conditions APmin = differential pressure for minimum pressure conditions PMAdp = differential pressure error limit From reference 15, Attachment 1, the feedwater temperature ranges from minimum 250 OF to a maximum of 367 OF, which is the design temperature.

The density is never less than the design value, consequently there is no non-conservative errors due to density variations. In the absence of non-conservative process errors, PMA is zero for this calculation.

FMAmA =0

Page: 35 of ,18 NEDC: 92-050R Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 8.2. Determination of Other Uncertainty Terms 8.2.1. Total Loop Accuracy for the Two Feedwater Flow Loops For this function, there are two individual loops of feedwater flow. These individual errors are combined as follows:

ATOTAL = !2(ALN 2)

ATOTAL = 2(0.03282) = 0.04639mA 8.2.2. Total Calibration Accuracy for the Two Fe(

For this function, there are two individual . These individual errors are combined as follows' CTOTAL = 2(C 2)+ 2(C) 4+/-2 )

CTTA ._200283 C = V2(0.05268) + 2 _ 3-8'14mA I[

CToTAL = 0.08814 nu (2a)11 8.2.3. Total Drift for the Two FePNater Flow Loops For this function, there, a e Vi dividual loops of feedwater flow. These individual efirs a Cur e s follows:

D1'OTAL = CD2\ 4.

DTT- , ŽE7 32)+2(02) .03017mA 1E IDTOTAL =0.03017 nu (21)1J All sij ' i1, a ter Flow uncertainty terms are accounted for in the previous n of Setpoint Margin and Operating Setpoint Setpoint Margin The setpoint margin is defined (Reference 11) as the margin between the nominal setpoint and the analytic limit. This margin is given by:

SM = (1.645/N)(SRSS OF RANDOM TERMS) + BIAS TERMS Where N represents the number of standard deviations with which all the random terms are characterized (normally 2) and 1.645 adjusts the results to a 95% probability for a single side of interest.

The uncertainty terms are calculated for trip conditions, the random term is U8 and the bias term is zero. Then the setpoint margin becomes:

Page: 36_ of 1 NEDC: 92-050R Rev. Number: 301 Nebraska Public Power District DESIGN CALCULATIONS SHEET SM= I'645V/OTALo2 +CToTA2 + DToT +(PM) 2 + (PEA)2 +/- (bias terms) n SM 1.645 V0.046392 + 0.088142 +0.030172 +(o.o) 2 +(0.000361Y 2

ISM =o.o856_h>~

8.3.2. Nominal Trip Setpoint (NTSP1) Calculation The Nominal Trip Setpoint (NTSPi) for process varialeht rip or increasing permissives is given by:

NTSPi =AL +SM From paragraph 2.2 the Analytic Limit (AL) is:

AL = 0.9521 Mlb/hr4-Which from paragraph 4.2.3.2 is equivalentt, SUMALFW = 4.7617 mA Therefore the nominal setpoi ~s NTSPmA = 4.7617 + 0.o856 = nmA OR NTSP1 =TSPTSPlmn, - m 4.8516-- 4,20.0 = 1.063 Mlb/hr l 6 16 Where, E Wm. c= ) \e t al full scale flow for the channel Asýja peg6*>"6wer, NTSPi is 1.o63/9.521 = 11.16% RTPo (lO.98% RTPK) i ,ns the lower limit (closest to AL) at which the setpoint can be 8.3. Iwta Value Calculation

\The nctions covered by this calculation do not have a required limiting

ŽK tpoint in the Technical Specifications, consequently calculation of an lowable Value (AV) is not applicable.

3.. LER Avoidance Evaluation The purpose of the LER Avoidance Evaluation is to assure that there is sufficient margin provided between the Allowable Value and the Nominal Trip Setpoint to reasonably avoid violations of the Technical Specifications. Since there is no AV, the LER Avoidance Evaluation is not required.

Page: 37 of 1 NEDC: 92-05OR Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 8.3.5. Selection of Operating Setpoints The setpoint could be changed to NTSP1 = 4.85 mA, however the existing setpoint of 5.79 mA is conservative and should be preserved to avoid unnecessary changes to the plant, station procedures or design documents.

os(OSmA- mA ZERO) *5. ~79-4 OS = -16mA ) *OS. ..

16 *20.0= 2.24 Mlb/hr

=W5.79-6 Where, EWmax = combined total full scale flow for the As a percentage of power, OS is 2.24/9.521 = 2 8.3.6. Establishing Leave Alone Zone (LAZ)

These loops are not subject to periodic fii tio Lwe*n calibrations.

For the purposes of this calculation, the ,0 Tolerance (LAT) for RFC-PLC-RVLC is equal to the As-Left T'l 8.3.7. Required Limits Evaluation The Required Limits Evaluati6 (1CRLE UVates an adjustment to NTSP for the case when NTSP is set at th en r o the][eave alone zone. The adjustment assures that with the stack-up of t tr-ucerain ties (including LAT) for all the devices in the loop, ther ýejnbugh margin for *Tech. Spec Action Avoidance (or LER avoidance). Sinc hei iE associnated with the setpoint, the RLE is not requiredFJ*

8.3.8. Spurious Ti Avo aio (SPA Evaluation There is no S aeiion required for a pern Rod Block f 8.3.9. Detkihmi La Acua Field Setpoint

no Elevation Correction (EC) for a flow measurement

mA = 2.24 Mlb/hr = 23.53% RTPo (23.16% RTPK).

Page: 38 of _18 NEDC: 92-05OR Rev. Number: 3C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

9. CONCLUSION 9.1. Main Steam Flow RWM Bypass Permissive The existing setpoint is acceptable, and if additional margin is desired it could be changed to 5.46 mA (1.095 Mlb/hr). The instrument setpoints and calibration data are as follows:

RFC-PLC-RVLC Analytical Limit 0.239 Mlb/hr I 0.515 psid 5.2744 mA

.-q6 mA (2o.%?/ R`IVý -Iý. "

Alternate Setpoint Supported by this Calculation 5.46 mA As-Left Tolerance +/-+.051 Leave Alone Tolerance +/-o.o5, 9.2. Feedwater Flow RWM Bypass Permissive The existing setpoint is acceptable, and if additional 1I changed to 4.85 mA (1.o63 Mlb/hr). The instrument- e q'tpoin as follows: _

RFC-PLC-RVLC Anal ait \\o -07 6 Mlb/hr

.10261 psid° 4.7617 mA Stl 5.79 mA (23.5% RTPo, 23.16% RTPK)

Setpoint 5.79 mA (23.5% RTPo, 23.16% RTPK)

Iýulation 4.85 mA (11.16% RTPo, 1O.98% RTPK)

'olerance +/-0.05 mA Folerance +/-0.05 mA

NLS2008034 Enclosure 2 Main Turbine First Stage Pressure Setpoint Calculation NEDC 92-50AJ, Revision 2C1 (23 Pages)

Cooper Nuclear Station, Docket No. 50-298, DPR-46 Information Only

Nebraska Public Power District C O P Y DESIGN CALCULATIONS COVER SHEET :r o,,.!o

Title:

Main Turbine First Stage Pressure Setpoint Calculation Calculation No.: NEDC 92-050AJ Task Identification No.:

System/Structure: Main Steam Design Change No.: N/A Classification: [X] Essential [ ] Non-Essential Calc.

Description:

Determination of setpoints for Main Turbine First Stage Pressure Switches MS-PS-14A/B/C/D.

Rev. 2 Incorporates new analytical limit and setpoint values based on information received from Siemens Westinghouse, letter dated September 21, 1998 and attached graph WF-533.

Rev 2C1: Revision to scale the % Rated Thermal Power (RTP) limits specified in the calculation to account for the 1.6% increase in Thermal Power that is realized with the Appendix K Power Uprate. The values of the Analytical Limit (AL), Allowable Value (AV) and setpoints do not change, only the corresponding equivalent % RTP is changed. For purposes of clarification, the originally licensed RTP of 2381 MWth will hereafter be referred to as RTPo, and the Appendix K licensed RTP of 2419 MWth will be referred to as RTPK.

2C1 3 Revised the % RTP for Appendix K Power 1,ar nruh J*Ia Vi ,.

Uprate. [, (11/6l/Z/o7 2 1 Incorporate revised Analytical Limit and Rftobert 7Alan Able ' Alan Able B. Lord calculate Allowable Value and Nominal Trip Champlin 10/27/98 10/27/98 10/28/98 Setpoint 10/26/98 R Krause 10/26/98 1 Incorporate Analytical Limit and calculate Bruce Crabbs R. Krause R. Krause Elden Plettner Allowable Value and Nominal Trip Setpoint 4/20/98 4/21/98 4/21/98 4/21/98 0 4 Original Issue Brian LeCuyer B.A. McMillan B.A. McMillan Alan Boesch 3/28/94 4/5/94 4/5/94 4/25/94 Rev. Status Revision Description Prepared Checked or Design Approved No. By/Date Reviewed By/Date Verification/Date By/Date Status Codes

1. As - Built 3. For Construction

2. Information only 4. Superseded or Deleted

Nebraska Public Power District Sheet 2 of 23 DESIGN CALCULATION CROSS REFERENCE INDEX NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07 Item Rev. PENDING CHANGES TO No. DESIGN INPUTS No. DESIGN INPUTS AFFECTED DOCUMENTS 1 USAR.Section VII - 2.3.8 and 15 Procedure 6. 1RPS.303 2.3.9.9;Section IV -4.10 2 CNS Memo DED98088 dated Procedure 62ART3 3/25/98 3 NEDC-31336P-A Proce -S 4 EDE-38-1090 0 5 VM 1025 EA51 VMCF 96-0159, VMCF 96-0236, VMCF 96-0217, VMCF 96-022 1, VMCF 96-0222 6 NUREG-1433 A 7 EQDP46 4 8 EQDP 12 3 9 VM1129 6 10 VM1185 0 M 11 SCR 89-29 0 12 NEDO-10678 ---

13 Drawing 2002 Sheet 2 N30 D CN 97 4 14 Drawing 791E256 Sheet 7 N16 15 Drawing 791E256 Sheet 9 C 16 16 Drawing 791E256 Sheet 10 N, 17 Drawing 791E256 SheetII k*2 12 -

18 Drawing 791E256 Sheet 12 \N 19 GE Letter C96091 l 20 GE Letter C9612 _----

21 GE Letter C9 ---

22 GE Lett:l4l:\ý " ---

23 \r V ---

24 00 5) ---

25 t,'1ts- 5A6254 ---

Y ED694-212 0

Nebraska Public Power District Sheet 3 of 23 DESIGN CALCULATION CROSS REFERENCE INDEX NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 P 07 Ite m Rev. PENDING CHANGES TO No. DESIGN INPUTS No. DESIGN INPUTS AFFECTED DOCUMENTS 27 CNS Letter to GE Dated 8/15/96 ---

28 CNS Letter to GE Dated 9/19/96 ---

29 Deleted ---

30 SCR 85-15 0 31 SCR 94-07 0 32 TSCR NLS970002 ---

33 ST1-20 Dated 2/4/75 --- V 34 GE-NE-0000-0063-6433-RO (Task 0 T0506: TS Instrument Setpoints) _

N"N

Nebraska Public Power District Sheet 4 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07

1. PURPOSE The purpose of this calculation is to determine the Allowable Value and Nominal Trip Setpoint for the Main Turbine First Stage Pressure Switches MS-PS-14A/B/C/D.

Four Main Turbine First Stage Pressure Switches (MS-PS-14AiB/C/D) are provided to initiate the automatic bypass of the Turbine Control Valve Fast Closure and Turbine Stop Valve Closure scra* -h nh the first stage pressure is below some preset fraction of rated pressure. The switches are arrangd-& N Nbi single failure can prevent a Turbine Stop Valve Closure scram or Turbine Control Valve Fas4lq**'* *e scram.

Below 30% of the original CNS licensed rated powerRated Thermal Power (* o (238 t) see assumption 3.15), the scram signal due to Turbine Control Valve Fast Closur an Turbin Sttp Valve Closure may be bypassed because the Neutron Monitoring System High Flux sktad gh Pressure scram are adequate to protect the reactor pressure vessel. Closurf of these6alves f-m-.such a low initial power level does not constitute a threat to the integrity of any bdri., h -ease of any radioactive material. The Cooper Nuclear Station Analytical Limit for thle Tuhr iratge Pressure permissive shall be less than 30% of RTPo rated First Stage Turbine Pressure (nVrml a\ dic tion of reactor thermal power) or 210 psia (195.3 psig) per References 6.2 and 6.35, d a t .0*3ý.ý During Turbine Bypass Valve testing, the first stage pressure is reduced dulto trug*. [iass alves while the actual reactor power is unchanged. To compensate for tlce"-otential t bf rips while the Turbine Bypass Valves are being tested, the scram bypass setpoint tc urrentyl ple mented at less than or equal to 25% of RTPo rated Turbine First Stage Pressure per Referent6,

2. REQUIREMENTS 2.1 This calculation is performe cCNS Engineering Procedure 3.26.3, Instrument Setpoint and Channel Error Calculati . (Reference 6.6)

3. ASSUMPTIONS 3.1 Theeinstrum'te p are of the same model number and have a similar installation. Therefore this calculatio s bel

%e*p cable to all instruments listed in Step 4.1.3.1.

-Eab o strument has the same performance specifications as the corresponding instrument in each manufacturer, Barksdale, does not specify a temperature effect for these instruments. Therefore, it is ssumed that the TE reported on instrument B IT-M 12SS-GE (References 6.12 and 6.26) applies to these instruments. Because of the small sample size during the test, the values for ATE and DTE are used as 1a values.

Basis:

GE Letter C970115 (Reference 6.25)

Nebraska Public Power District Sheet 5 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2Cl Date: Nov. 2 2007 Date: Nov. 2 20 07 3.3 These instruments are not subject to any Over Pressure Effects (OPE). Therefore, OPE will not be considered for this calculation and OPE will be equal to zero.

Basis:

The proof pressure as listed in the vendor manual (Reference 6.9) is 1800 psig. This pressure is well the normal operating pressure of approximately 1005 psig and the peak allowable reactor pressure vevN pressure of 1375 psig (110% of design) as listed in USAR Section IV - 4.10 (Reference 6.1). 4,,

3.4 These instruments are not subject to any Static Pressure Effects (SPE). Therefore, SPE willOf considered for this calculation and SPE will be equal to zero. __>

Basis:

Static pressure effects generally apply only to differential pressure nents are measuring pressure, not differential pressure and as a result will effects.

3.5 These instruments are expected to function during or after a seismi- t. manufacturer, Barksdale, does not specify a Seismic Effect (SE) for these instrument mod \ Pre, s assumed that the seismic effects reported on instrument model B2T-A12SS (Re fjjý"o these instruments. The pressure switches in this calculation will be exercis -ata, c _me, mum acceleration (i.e., 2g peak per Reference 6.28) than what was used during theTt r erence 6.27). Also, CNS operational experience has shown that these instruiten, exhlbti \ncased sensitivity in a vibration environment, resulting in spurious actuation of the sanii. Therefore, SE will not be considered within this calculation.

Basis:

GE Letter C970115 (Refere ce ,K 3.6 The manufacturer, Barksdale o osn y Radiation Effects (RE) for these instruments. These instruments are not desigt t gtibat oy accident condition, nor are they designed to perform their function in an accident ervn*' e,-erefore, RE will not be considered for this calculation and RE will be equal to zero.

Basis:

(Reference 6.9) hufa;chire , Barksdale, does not specify Humidity Effects (HE) for these instruments. These en constructed with a NEMA 4 rated water-tight housing and contain class M switches which AI ýe"Umidity requirements of MIL-S-6743 per the information contained within Reference 6.9. It is A PThe vendor accuracy includes any effects from humidity and a separate HE will not be considered calculation. Therefore, HE will be equal to zero.

asis:

p-Barksdale Vendor Manual (Reference 6.9)

Nebraska Public Power District Sheet 6 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07 3.8 These instruments are not subject to any Power Supply Effects (PSE). Therefore, PSE will not be considered for this calculation and PSE will be equal to zero.

Basis:

The switches contained within this instrument are bistable devices that are operated by system pressure alone which do not involve a power supply.

3.9 These instruments are not subject to any RFI/EMI Effects (REE). Therefore, REE will not for this calculation and REE will be equal to zero.

Basis:

These instruments do not contain any components that may be affected by ra(

electromagnetic emissions.

3.10 All vendor accuracies are assumed to be 2 a values in accordanK es provided in EDE 1090.

Basis:

EDE-38-1090 (Reference 6.4) 3.11 Design The 100% rated steam flow is 9.542E6 lbs/h which yields a 700 psia first stage pressure.

Basis:

References 6.2 and 6.35 3.12 Normal and pla t trip conditins'have3 r'dation parameter equal to those established for instrument racks 25-5 and 25-6 1ocated in t e ddN* .ding which equals 350 R TID/40 years.

Basis:

The area wherp s are located is a mild environment and is shielded via concrete floors and walls in the E .truments are located in the Turbine Building 903' level it is reasonable to assume th tihý Iqadi t~i level documented for the sensitive instrument racks located at the reactor building

.encompass any radiation levels MS-PS-14A, B, C, or D would encounter.

temperature will not exceed 100°F as the maximum temperature in the area is established at I

A

>1

.eference 6.1 - USAR Section X.

.4

Nebraska Public Power District Sheet 7 of 23 1 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07 3.14 The analytical limit and the administrative limit are found by multiplying the first stage pressure (700 psia) by the corresponding percentage and then converting to psig by deducting 14.7.

Basis:

Reference 6.36 supplies the basis for the conversion from psia to psig.

3.15 For clarification, the originally licensed RTP of 2381 MWth will hereafter be referred to as Appendix K licensed RTP of 2419 MWth will be referred to as RTPK. Conversions from 0/

%RTPK will be performed as follows: 1 44W

%RTPo = 1.016 x%RTPK

%RTPc = %RTPo 1.016 Conversions required to support this calculation are contained

% RTPo %RTPK 100% 98.425%

30.0% 29.53%

25.0% 24.61%

Nebraska Public Power District Sheet 8 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07

4. METHODOLOGY 4.1 Instrument Channel Arrangement High Pressure Turbine RPS Logic Contacts 4.1.1 Channel Diagram (Ref. 6.19) 4.1.2 Definition of Channels Each channel consists of a single pressure switch.

4.1.3 4.1.3.1 Reference MS-PS-14A/B/C/D 6.13 Barksdale 6.13 B2T-M12SS 6.13 1200 psig 6.9 77 to 1200 psig 6.9 1: System Pressure 6.7, 6.8 Contact Closure 6.7, 6.8

)V Specifications: See Section 4.1.3.3

.3.2 Process and Physical Interfaces Reference Calibration Temperature Range: 65 to 100 OF 6.21, Assumption 3.13 Calibration Interval: 18 months + 25% (grace period) 6.10 Location: T-903-CORR 6.13

Nebraska Public Power District Sheet 9 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07

- U Reference Normal and Trip Plant Conditions I

Temperature: 50 to 100°F 6.11, 6.1 Radiation: 350 R TID/40 years 6.11, /

AssumptiT Pressure: 0.0 to -0.25 "H 20 Humidity: 20 to 90% R.H.

Temperature Range for Trip condition Error Calculations:

r Tot. Temp range (ATT) = larger of (Max trip temi )) or (Max calib temp - min trip temp)

ATT =100°F - 65°F = 35°F or 00°F - 50°F = 50OF

=

Temp range for DTE calc ) a calib temp 100OF - 65OF = 35OF Temp range for ATE cal TTATD = 50F - 35°F = 15°F Temperature Range for Nora Conditir *-rror Calculations:

Tot. Temp (MUN-.-;a I*i rati I :temp) er of (Max trip temp - min calib temp) or 0 F 35°F or ATT=100'F-50OF=50'F e n I rDTE calc (ATD) = max calib temp - min calib temp =

AT 5=o -35°F5' em'range for ATE calc (ATAT) = ATT - ATD = 50°F - 35° = 15T nost P Accident Conditions:

> Long term Post-Accident conditions do not apply to these instruments.

Seismic Conditions:

Prior to Function: Not Applicable Assumption 3.5 During Function: Not Applicable Assumption 3.5

'V

Nebraska Public Power District Sheet10 of 23 SHEET DESIGN CALCULATIONS NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07 4.1.3.3 Determination of Individual Device Accuracies All accuracy error contributions are random variables unless otherwise noted.

4.1.3.3.1 Vendor Accuracy (VA)

Utilizing a conservative approach, the value of the Upper Range (UR) (

psig) will be used as the value on which the vendor accuracy is determiI1 Value Sigma Reference VAi = 6.00 psig (+/- 0.5% UR) 2 4.1.3.3.2 Accuracy Temperature Effect (ATE)

Per the information contained within eferenc 25 ((I F t1er C9701 15), the ATE for these instruments is ATE = [(1.6197% UR)/(100 OF)] (A )

ATE = [(0.016197)(124 0 F7\),

ATE = 2.91546 psig a)

ATE = 2.92 psig / (

Reference 0 Assumption 3.3 0 Assumption 3.4 0 Assumption 3.5 0 Assumption 3.6 0 Assumption 3.7 PSE: 0 Assumption 3.8 REE: 0 Assumption 3.9 Trip Conditions The Trip Environment Conditions are the same as the Normal Plant Conditions per Step 4.1.3.2.

Nebraska Public Power District Sheet I Iof 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07 4.1.3.3.4 Accuracy Values The identified accuracy error contributions are combined using the SRSS method to determine total device accuracy. The device accuracy is normalized to a 2a confidence level, and is given by: f LA, 2 7 :ý .E E ~ P~

Where the terms inside the square root sign are the individual effects, and the "n" is the c value assocja effect.

Normal Accuracy AiN --2k(.00J)2+2.921 2 +(o)2 +

AiN = +/- 8.37 psig For this instrument 1( in the loop. Therefore, the loop accuracy (ALN)

ALN = +/- 8.37 psig same as the Normal Accuracy values due to being the same as the Normal Plant 4.1.3.4 Drift (VD)

-The manufacturer, Barksdale, does not specify a drift value for these instruments. Therefore, a drift value was determined by utilizing the vendor accuracy as the drift for a six month period (Ref. 6.4).

VD 6 = VA

Nebraska Public Power District Sheet 12 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07

-U The six month drift value is then extrapolated to a 22.5 month frequency according to the following equation:

VD 2 2 5 = (VD 6 )(SQRT (22.5/6))

VD 22.5 = (6.00 psig )(SQRT (22.5/6)) A VD 22.5 = 11.62 psig (20)

The drift value is treated as a 2a value due to the input value in extrapolated drift being 2a values.

4.1 .3.4.2 Drift Temp~erature Effect (DTE)

Per the information contained within Reference 6. C970115), the DTE for these instruments is:

DTE =[(1.6197% UR)/(100 °F)] (ATD)

DTE=[(0.016197)/(1200)/100 0 F ',

DTE = 6.80274 psig DTE = 6.81 psig (Ia) o 4.1.3.4.3 Drift Values The-'dIvice i e r i culated by SRSS coml bination of VD 22.5 and DTE.

Therto-Ie, he i ts follows:

2 2 VD22.51 +(DTEI1 J

\\ý* ere the terms inside the square root sign are the random portions of the individual effects, and the "n" is the a value associated with each individual effect.

7 2

j= 2 ~11.62) +6.8 1J Di = 17.90 psig y For this instrument loop, there is only one device in the loop. Therefore, the loop drift (DL) equals the device drift.

DL = 17.90 psig

Nebraska Public Power District Sheet 13 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07 4.1.3.5 Establishina As-Left Tolerances The As-Left Tolerances for these instruments is established in CNS Surveillance Procedures 6.1RPS.303 and 6.2RPS.303 (References 6.7 and 6.8) as +/- 6.00 psig for MS-PS-14A/B/C/D. #

ALT = +/- 6.00 psig (3o) 4.1.3.6 Determination of Device Calibration Error .4 4.1.3.6.1 Device Calibration Tool Error These instruments are calibrated with a 0 to 1200 which is in turn calibrated by a dead weight tester Since these instruments are switches nocalib'fo ired for the output.

Input Calibration Tool:

Heise, 0 to 1200 psig, .2550 Accuracy (I 1(0.1% of full scale) (3a) (References 6.14, 6.22)

Rea 0.50 psig (1/2 minor division) (3a) (References 6-1, C6 rature Effect = 0 (Temperature Compensated) (References Insfield and Green Type "T" Deadweight Tester x2 Accuracy (CSTDinp) = +/- 1.20 psig (+/- 0.1% of reading (1200 psig)) (3;)

(References 6.15, 6.22)

Outpt None Required.

4.1.3.6.2Device Calibration Error V The Calibration Error (Ci) for Device "I"is the SRSS combination of the As Left Tolerance (ALT), and the errors due to input and output calibration tools (including tool accuracy and readability and the error of the calibration standards). The calibration accuracy is normalized to a 2a confidence level:

Nebraska Public Power District Sheet 14 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07

-U 2

Ci =22ALT )2 +l n_ CTOOLinp-)

n__nn +_ CREADinp')2 + gCSTDinp -2 + CTOOLut )2 + ( CREADOut 2 + -CgTOu n1 2 + bias where "n" is the a value associated with each individual effect.

A 4.1.3.6.3 Device Calibration Error Values Since the values of ALT, CTOOL, CREAD, and CSTD are conti testing, they are assumed to represent 3a values, so n = 3 and the Calibration Error is: -N t*

2 Cl 60 +(1.20j)

)2j~ +ý0.50 2 +(1.20)2 +(-

A4 Ci = +/- 4.17 psig (2y)

In this case, the loop ci and Loop Calibration Errors are CL = +/- 4.17 psig 4.1.4 Determination of Loo For tI is calculation, the loop cotnaýih rl) ;o device. Therefore, the device error values for Accu racy, Drift, and C'brat* >tsas those for the loop. These values have been I n.

repor ted in Section 4.1.5 Determination of P !AA:a~ d PidII ,

ary Eleme A): PEA is not applicable for this type of instrument channel since Prim the prrlmy ne *ý epressure switch itself.

Pro eU ei Accuracy (PMA): The First Stage Turbine Pressure switches are not

'i t*te any accident condition, nor are they designed to perform their function in an en nment. These switches are located in the Turbine Building and a significant change

.Perature of the environment is not expected. Little or no change in the density of the

s~-i the instrument sensing lines will occur due to Turbine Building temperature changes.

fore, PMA is negligible and will not be considered for this calculation.

Set, intinof Other Error Terms 41 ,Determination of Other Error Terms All error terms to be considered have been accounted for in the previous sections.

Y

Nebraska Public Power District Sheet 15 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07

- p 4.1.7 Calculation of Setpoint Margin and Operating Setpoint 4.1.7.1 Setpoint Margin (SM)

The Setpoint Margin is defined as the margin between the limiting setpoint (Analytic l Limit) and the operating setpoint to allow for the accuracy of the instrument, inacy of the test equipment, and the instrument drift that could occur during the interv,. \

between calibration (Reference 6.5). Setpoint Margin is thus given as:

SM 1 n4 5AV.6 +C' +D + (PEA)2 +(PMA) 2 +bia \

L L L Where "n" represents the number of standard deviation~sth whi -1 e random terms are characterized (normally 2 standard deviatios n,6 adjusts the results to a 95%

probability (one-sided normal). The error termsl e lc .t d for normal environment, and the setpoint margin becomes:

SM= 1.645 2

SM = +/- 16.62 psig 4.1.7.2 Nominal Trii -ulation The Nomit 'or process variables which increase to trip is given as:

NTSPI cilLimit is < 30% of rated powerRTPo (29.53% of RTPK) or < 195.30 psig

ýes 6.2, and 6.35, 6.37 and assumption 3.15.

7,(700 x 0.3) - 14.7

= 195.30 psig F

Therefore, the Nominal Trip Setpoint is:

NTSPI = 195.30 - 16.62 psig NTSP1 = 178.68 psig NTSP1 represents the upper limit (closest to AV) which the setpoint can be set assuming zero leave alone tolerance in the direction toward the Allowable Value (AV).

K

Nebraska Public Power District Sheet 16 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07 4.1.7.3 Allowable Value Calculation For this setpoint calculation, the process variable increases to trip, therefore, the Allowable Value (AV) is calculated using the following equation from References 6.3 and 6.4:

AV=AL- 1.645 A2L +C2 +(PEA) 2 +(PMA) 2 n L Where "n" representsdai the number of standard deviations wl %hicchalht a om s are characterized (normally 2 standard deviations) and 1.4, aJjsts th r ýuls to a 5%

probability (one-sided normal).

AV = 195.30- 1.645 8.372 +4.17 AV = 187.6086 psig AV = 187.60 psig conse I r dy ference 6.23 4.1.7.4 LER Avoidance Evaluation The purpose of the LERvQidan*e aluation is to assure that there is sufficient margin provided bet en th&Al " abe alue and the Nominal Trip Setpoint to reasonably avoid violqtionsh Žalo(bA1e Value (which, when discovered during surveillance, could lead o ' c. The method of avoiding violations of the Allowable Value is tohd otm hh rrors that may be present during surveillance testing, examine the mar t cýalculated values of NTSPI and AV, and then adjust NTSP1 to provid* d liif n necessary. The following equation is used to determine the e o "ha debe expected to contribute to a potential LER situation.

'T (SRSS of Random Terms) hv e n" represents the number of standard deviations with which the random terms are S aracterized (normally 2 standard deviations).

4.1 .7.4.1 Random Terms Included in LER Avoidance The Random Terms which should be included in the LER Avoidance evaluations include:

(1) Loop Accuracy under Normal Plant Conditions (ALN)

(2) Loop Calibration Error (CL)

(3) Loop Drift (DL)

Primary Element Accuracy and Process Measurement Errors are not included due to calibration and surveillance testing being performed using input signals which simulate the process and primary element input. Therefore, the LER Avoidance Evaluation equation can be expressed as follows:

Nebraska Public Power District Sheet 17 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07 S 2 +C2 +DL2 LE L L

-LER = 8.372 +4.172 +17.902

'LER = 10.10 psig 4.1.7.4.2LER Margin Calculation Once the value of oLER is determined, the margh NTSPI and AV are calculated using the equation JA V- NTSP1I ZLER =-

O"LER This value of ZLER is tl Srob ility of violating the Allowable Value by tr *iibution as a random Normal Distribution, and then ar-a under the curve of the Normal Distribution correspon iber of standard deviations represented by ZLER" 4.

I recommended that a nominal probability of 90% for avoiding an LER ndition be used as an acceptance criterion for the LER (or Tech Spec Action)

Avoidance Evaluation. For a single instrument channel, the value of ZLER corresponding to this 90% criterion is 1.29 or greater. Since these switches are single instrument channels, the computed value of ZLER will be compared with 1.29 (Reference 6.21).

Nebraska Public Power District Sheet 18 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 . 20 07 Date: Nov. 2 20 07

-U 4.1.7.4.4Goveming Setpoint Determination A comparison Of ZLER with the limiting value for the LER (or Tech Spec Action)

Avoidance shows that:

ZLER = 0.88 < 1.29 Therefore, the LER (or Tech Spec Action) Avoidance acceptance satisfied if the setpoint is set at NTSP1. As a result, an adjusted4 must be calculated in order to meet the acceptance crition greater.

NTSP2 = AV - 1.29(aLER)

NTSP2 = 187.60 - 1.29(10.10)

NTSP2 = 174.57 psig NTSP2 = 174.5 psig (rounded conse \istrument readability)

The more conservative setpoint is thi one leý d to e the governing setpoint.

Therefore, NTSP2 is the governinl 4.1.7.5 Selection ofOneratina Setoint It is recommended that the metho( FSP as the center of the Leave Alone Zone be used. Thus the nominal stetppiint isý Operating S etpoint = NTSP 41 psig +/- LAT Where the I A isth ination of the Leave Alone Tolerances for all devices in the loop. -

4.1.7.6 Establi ,Iav ANne Zones Si aeýq q ly contains one device, the Loop Leave Alone Tolerance is the Device e-Tolerance. The current LAT from the surveillance procedures (References rwill be utilized in this calculation.

+/- 6.00 psig

/The Leave Alone Zone (LAZ) by definition is twice the absolute value of the LAT.

Therefore, the LAZ equals 12.00 psig.

14'.1.7.7Required Limits Evaluation The Required Limits Evaluation calculates an adjustment to the NTSP when the NTSP is set at the center of the Leave Alone Zone. This adjustment assures that with the stack-up IV of the errors (including Leave Alone Tolerances) for all devices in the loop, there is enough margin for LER (or Tech Spec Action) Avoidance.

Nebraska Public Power District Sheet 19 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2CI Date: Nov. 2 2007 Date: Nov. 2 20 07

-H The Required Limit (RL) for device "I"in the loop is:

RLi = NTSP + LATi Since this loop includes only on device the subscript "I"will be dropped.

RL = 174.50 psig + 6.00 psig RL = 180.50 psig This is compared against the AV of 187.60 psig from Sectic AV, for an increasing variable, no further adjustment to the Per References 6.3 and 6.4, aLER, RL is given by the follov a single device loop:

GLERRL 2 3 OLER,RL (1 = F1745 0)+ 4.17 +17.90 CFLER,PRL =

es*.3 and 6.4, ZLERRL is given by the following equation:

NTSP21 ZLERRL = A V-

'LER,RL ZLERPRL 1187.60-174.501 9.41 ZLE ,RL= 1.39 Per References 6.3 and 6.4, ZLERRL > 1.29 is required to meet the 90% LER Avoidance acceptance criterion. Since the calculated value for ZLERR is greater than 1.29 (i.e., it meets the LER Avoidance acceptance criterion), no further adjustment is necessary.

Nebraska Public Power District Sheet 20of 23 SHEET DESIGN CALCULATIONS NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07 4.1.7.8 Selection of Operating Setpoint The recommended Operating Setpoint is OSP = NTSP2 +/- LAT OSP = 174.50 psig +/- 6.00 psig 4.1.7.9 Spurious Trip Avoidance Test The Spurious Trip Avoidance Evaluation is not required as the's tpoifa enough away from the normal system operating pressure.

4.1.7.10 Elevation Correction The sensing line tap elevation for these instruments is At evatio'3 ,and the instrument elevation is at 907'per Reference 6.17. This-i fl, 4 i ence results in a+10.8 psig (0.432 x 25) elevation correction when calculate~l us "th enversion factor of 0.432 psig/feet as found in Reference 6.5. +!i 4.1.7.11 Determination of Actual Field SO a .00/ R (29.53"o of RTP,)

The Actual Field Setpoint (ASP)

ASP = Operating Setpoint + nvationiection ASP 174.50 psig+ 1 .8-si - 30psig ASP= 18 e co srvat )

4.2 Determination of FiJdl etpoointat25% of RTPo (24.61% of RTP,)

An administrativei f'o of RTPo (24.6 1 of RTPK) of the rated turbine first stage pressure is requirde asumption 3.14). The setpoint of 185.00 psig exceeds the 25% psig limit of 16.3 T

e** nt must provide assurance that the admin limit will not be exceeded. Therefore:

) dmin Setpoint = Admin Limit - LAT Admin Setpoint = 160.30 - 6.00 154.3 psig = 154.00 psig (rounded conservatively)

Actual Setpoint = Admin Setpoint + Elevation Correction Actual Setpoint = 154.00 + 10.8

= 164.8 psig

= 164.5 psig (rounded conservatively to device readability)

Nebraska Public Power District Sheet 21 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20_07 Date: Nov. 2 20 07

5. CONCLUSION The instrument Setpoints and Calibration data are as follows:

Analytical Limit Allowable Value Operating Setpoint Elevation Correction Setpoint (as calculated in this calculation for 30% of RTPO/2 9.53% of Actual Setpoint (CNS Setpoint to support 25% of RTPo/24.61% of R Leave Alone Tolerance (LAT) +/- 6.00 psig As Left Tolerance (ALT) +/- 6.00 psig

Nebraska Public Power District Sheet 22 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 20 07 Date: Nov. 2 20 07

6. REFERENCES 6.1 USAR Section VII - 2.3.8, Scram Bypasses; 2.3.9.9, Main Turbine First Stage Pressure Switches;Section IV - 4.10, Over pressure Protection Analysis;Section X - Tables X 1 and X- 10-2..

6.2 Siemaens Westinghouse letter, dated September 21, 1998, from Ron Macklin to Mark Metzger, "Press4re vs. Flow for Cooper Nuclear Station" with Graph WF-533 attached. 66368-0528 6.3 W.H. Cooley, J.L. Leong, M.A. Smith and S. Wolf, General ElectricInstrument Setpoint Met o . 1 NEDC-31336P-A, General Electric Company, San Jose CA, September 1996.

6.4 W.H. Cooley, Setpoint CalculationGuidelinesfor the Cooper Nuclear StatioED3A(R5ev11 \

General Electric Nuclear Energy, San Jose, CA, January 25, 1991.

6.5 CNS Engineering Procedure 3.26, Revision 10, InstrumentSetpoint and Meter an 4g§ ntrol.

6.6 CNS Engineering Procedure 3.26.3, Revision 3, Instrument Se i a nel Error Calculation Methodology.

6.7 CNS Surveillance Procedure 6.1 RPS.303, Revision 2 urbin Si g e s e Permissive Calibration and Functional(Div 1).

6.8 CNS Surveillance Procedure 6.2RPS.303, Revision' bage, Pressure Permissive Calibration 2).

and Functional(Div 6.9 CNS Vendor Manual Number 1025, GE Opr,Y-nand MaintenanceInstruction, GEK-9688A, Volume VIII, Part 3, Book 2, Switches Tab, Barksa

e* t1-*

r witches.

6.10 NUREG-1433, Volume 1, Rev. pr dard Technical Specifications, GE Plants,BWR/4s.

6.11 Deleted 6.12 EDQP 12, Revision 3, Pr.e . e. s, Barksdale Model Number B2T-M12SS, B2T-A 12SS.

6.13 Equipment Data e )

6.14 CNS Ins en CtD 1Procedure 14.1.2.1, Revision 9, ]AC Test Gauge Calibration.

6.15 ulS umber 1129, Revision 6, Mansfield and Green Type 'T Comparatorand ure DeterminationSystem.

S6.16 anual Number 1185, Revision 0, Heise Gauges Solid FrontSafety Gauge Model 't'.

eS;, tt Change Requests 89-29, Turbine First Stage Pressure Trip.

8 eneral Electric Licensing Topical Report, NEDO-10678, 72NED81, Class 1, November 1972, Seismic Qualification of Class I Electric Equipment.

Nebraska Public Power District Sheet 23 of 23 DESIGN CALCULATIONS SHEET NEDC 92-050AJ Prepared By: Mark E. Unruh Checked/Reviewed By: Ralph Krause Rev. No. 2C1 Date: Nov. 2 2007 Date: Nov. 2 20 07 6.19 Burns and Roe Drawing 2002, Sheet 2, Revision N30, Flow Diagram, Main and Exhaust and Auxiliary Steam Systems.

6.20 General Electric Drawing 791E256, Sheets 7, 9, 10, 11, 12, Revisions N16, N16, N10, N12, N12, Elementary Diagram, Reactor Protection System.

6.21 General Electric Letter C960911 to CNS, Telephone Conversation Confirmation,dated September 1996.

6.22 General Electric Letter C961202 to CNS, Compilation ofAccuraciesfor CNS CalibrationBq dated December 2, 1996.

6.23 General Electric Letter C970109-A to CNS, CNS Setpoint Rounding Conve id dated 1917 6.24 General Electric Letter C9701 11 to CNS, Generic Changes to CNSSetpoint ,i ns dated January 11, 1997.

6.25 General Electric Letter C970115 to CNS, Closure of Open Items ksdale, SOR, Barton, and Agastat Instruments, dated January 15, 1997. g-6.26 DI 0035, Transamerica DeLaval Barksdale Contro sio Aal Tes rocedure 9993.

6.27 Test Results #225A6254 for Barksdale B2T-A12S itcl, DRF# AOO-1084, Revision 0, dated May 20, 1970.

6.28 Calculation NEDC 94-212, Revision 0, Co n of Generic Response Spectrafor Cooper Nuclear Station.

6.29 CNS Letter to GE, Guidelines toe ,iIRevision 0, dated August 15, 1996.

6.30 CNS Letter to GE, Spurious i vo Calculation, Operating Limits, dated September 19, 1996.

6.31 Setpoint Change Reques S ) -, Turbine FirstStage Pressure Trip.

6.32 Setpoint Change ee -,5Turbine Trip Scram Bypass.

6.33 Setpoint C g e,e 4-07, Turbine First Stage Pressure Trip.

6.34 ( ch - *..n quest "Re (TCSR) NLS970002, ProposedChange to CNS Technical Specifications n,plwroved StandardTechnical Specifications NRC Docket No. 50-298, License No. DPR-46, t ,1997.

6.3 0 er uclear Station Startup Test Results Report, Report # 20-1 dated 2/4/75, "Startup Test ST1-20, t roduction".

6.'6 "Introduction to Fluid Mechanics" Third Edition by R. W. Fox and A. T. McDonald, pages 58-59.

6.37 GE-NE-0000-0063-6433-RO, "Nebraska Public Power District Cooper Nuclear Station Thermal Power Optimization - Task T0506: TS Instrument Setpoints", Revision 0, October 2007.

NLS2008034 Enclosure 3 MS-DPIS - 116, -117 - 118, - 119A/B/C/D Setpoint Calculation NEDC 92-50AM, Revision IC1 (18 Pages)

Cooper Nuclear Station, Docket No. 50-298, DPR-46 Information Only

Nebraska Public Power District COPY c4 0,-1Y DESIGN CALCULATIONS COVER SHEET Title MS - DPIS - 116, - 117 -118, - 119A/B/C/D Setpoint Calculation Calculation No. 92-050AM Task Identification No. 935401 (11.1)

System/Structure MS Design Change No. EE 07-01 Component MS - DPIS - 116, 117, 118M 11 9A/B/C/D Discipline Instrument and Control Classification: [ X I Essential [ I Non-Essential *ASME Stress Reports shall be approved by Registered P.E.

Calc.

Description:

Determination of Primary Containment Isolation System Group 1 Isolation on main steam high flow Allowable Value (AV) and setpoints for MS - DPIS - 116, 117, 118, 119A/B/C/D.

Revision 1 C1 is to address the affects of the Appendix K power uprate, which will allow a 1.6% Reactor Thermal Power increase (at constant dome pressure) due to increased accuracy in Feedwater Flow measurement once the Caldon LEFM system is installed at CNS. Implementing document for uprate is EE 07-01.

/1 4

Revised to address the Mark E. Unruh , f.---. .4, p..Q-implementation of the Appendix K 10/20/07 .

power Uprate (1.6% RTP increase at 07 constant dome pressure).

Rev. Status Prepared Checked or Design Approved No. Revision Description By/Date Reviewed By/Date Verification/Date By/Date Status Codes

1. As - Built 3. For Construction

2. Information only 4. Superseded or Deleted

Sheet 1 of 1 Nebraska Public Power District DESIGN CALCULATION CROSS REFERENCE INDEX NEDC 92-50AM Rev. No. ICI Prepared By: Mark E Unruh Checked/Reviewed By: J. Neddenriep Date: October 20, 2007. Date: Nov. 8, 2007 Item DESIGN INPUTS Rev. PENDING CHANGES TO No. DGIUNo. DESIGN INPUTS AFFECTED DOCUMENTS 1 USAR XIV-6.5 2 D1 1064 --

3 EQDP46 4 4 NEDC 92-050AM 0 5 MEL 13 .... _ _ __ __ __ _

6 DI 0244 --

7 DI 2423 --

8 EQDP 13 8 9 NEDC 92-050A0 1 **Vi.. 'K'..

10 GE-NE-0000-0063-6433-RO (Task 0 K

'K T0506: TS Instrument Setpoints) _ -___

K' ""

"K

-s,?;:

,, ,K ,.=: , -* K

~K

.\.".'5

\.\., '

, ,? '

" \ K',

Nebraska Public Power District Sheet I of 16 DESIGN CALCULATIONS SHEET NPPD Generated Calculation Review of Non-NPPD Calc No: NEDC 92-050AM, Generated Calculation Rev. ICI, MS-DPIS-116-119A/B/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

References

1. USAR XIV-6.5, Main Steam Line Break Accident.

2. J. E. Walker, P. D. Knecht, Analytical Limits for Cooper Nuclear Station, NEDC-32676P, General Electric Company, San Jose, CA, January, 1997.

3. W. H. Cooley, J. L. Leong, M. A. Smith and S. Wolf, General Electric Instrument Setpoint Methodology, NEDC-31336P-A, General Electric Company, San Jose, CA, September 1996.

4. W. H. Cooley, Setpoint Calculation Guidelines for the Cooper.Nuclear Station, EDE- v 0, General Electric Nuclear Energy, San Jose, CA, January 25, 1991.

5. Letter CNS to GE, Guidelines to Review GE Reference Document, Rev. 0, Dat

6. CNS Engineering Procedure 3.26.3, Rev. 2.1, Instrument Setpoint and Cha Error tion Methodology.

7. EQDP 46, Rev. 4, Environmental Conditions.

8. Equipment Data File (EDF).

9. CNS Improved Technical Specifications, April 1998.

10. GE Letter, C960911 to CNS, "Telephone Conversatr'r garding CNS Setpoint Analysis)",

dated Sept. 11, 1996.

11. CNS Surveillance Procedure 6.1PCIS.303, Rev. 0.1 C I ain Steam Line High Flow Calibration and Functional Test (DIV. 1).

12. CNS Surveillance Procedure 6.2PCIS.303 e, CIS Main Steam Line High Flow Calibration and Functional Test (DIV. 2).

13. CNS DI 1064, ITT-Barton Produc- II 88 1n 289A-2 Indicating Switches, Dated 1977.

14. NEDC 92-050AM, Rev. 0, MS 7, -118, -119 A/B/C/D Setpoint Calculation.

15. GE Design Specification Dat 1058AT, Flow Element Components, 2/19/86.

16. CNS Instrument and Co ure 14.1.2. 1, Rev. 9, IAC Test Gauge Calibration.

17. GE 719E415BB R , c ear Boiler P&ID.

18. CNS M_*y Rist (MEL), Rev. 13.

19. DI 24 eport R3-288A-1244 (Attached to BWROG Equipment Summary QSR-0270A-02).

20KýGE r, 09-A, to CNS (Mark E. Unruh), CNS Setpoint Rounding Convention, dated January 9, 21970111, to CNS (Mark E. Unruh), Generic Changes to CNS Setpoint Calculations, dated Janu 1, 1997.

22. GE Letter, C970115, to CNS (Mark E. Unruh), Generic Changes to CNS Setpoint Calculations, dated January 15, 1997.

23. DI 2423, Summary of Calculated Gamma Radiation Data for Equipment Qualification of MEL Rev. 8 Electrical Components.

24. GE Letter, C961202, to CNS (Mark E. Unruh), Compilation of accuracies for CNS Calibration Equipment, dated December 2, 1996.

Nebraska Public Power District Sheet 2 of 16 DESIGN CALCULATIONS SHEET NPPD Generated Calculation Review of Non-NPPD Calc No: NEDC 92-050AM, Generated Calculation Rev. iC1, MS-DPIS-116-119A/B/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

25. CNS Engineering Procedure 3.26, Rev. 10, Instrument Setpoint and Meter Banding Control.

26. CNS Vendor Manual 1025, Barton Differential Pressure Switches.

27. CNS Vendor Manual 1137, Ametek Model RK Pressure Tester.

28. EQDP 13, Rev. 8, ITT Barton Differential Pressure Switches Models 288, 288A, and 289A.

29. CNS Vendor Manual 0773, Barton Switches and Transmitters.

,4

30. CNS Vendor Manual 1185, Hiese Gauges Model C, CC, CM, and CCM Manual. /I

31. NEDC 92-050AO, Rev. 1, RCIC-DPIS-83 and RCIC-DPIS-84 Setpoint Calculation.

32. GE Drawing 731E61 1, Sh. 1, Rev. N04, Primary System Piping.

33. Permutit Drawing 556-26386, Rev. 7, Outline & Dimensional Data - Steani l1w Regt [tol. >1

34. Jelco Drawing X2507-217, Rev. N02, PI-15 Process Instrumentation, React B di

35. Jelco Drawing X2507-216, Rev. NOI, PI-15 Process Instrumentati n a or Buirmg.

36. GE Project Task Report GE-NE-0000-0063-6433-RO, "Nebraska I wer District cooper Nuclear Station Thermal Power Optimization, Task T0506: TS "nstrume

" evision 0, Oc:tober 2007.

Y

Nebraska Public Power District Sheet 3 of 16 DESIGN CALCULATIONS SHEET NPPD Generated Calculation Review of Non-NPPD Calc No: NEDC 92-050AM, Generated Calculation Rev. 1C1, MS-DPIS-116-119A/B/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

1. PURPOSE In consideration of Cooper setpoint verification program in conjunction with a 22.5 month surveillance interval (required 18 months plus 25% grace period) , determine the Nominal Trip Setpoint and Allowable Value for Main Steam Line (MSL) High Flow portion of the Primary Containment Isolation System (PCIS)

(Reference 1). The Analytical Limit for the MSL High Flow is 150% (of originally licensed flow corresponding to 2381 MWth), corresponding to 121.24 psid (Reference 15), as shown in Refeyefces 2 and

15. Assumption 3.13 defines how the Analytical Limit is treated after the Appendix K Powe4Yprj4e.

During an accident where the Main Steam Line has a complete double-ended shear, w ac-ftent, break, these devices function to provide input into the PCIS system and in-tttpe PI Q's the break by fast closure of the Main Steam Isolation Valves (MSIV). This ft calo "s limits the radiation exposure to the general public to within the limits of IOCFR16Q)\

Each of the four (4) Main Steam Lines has four (4) of these dc endent, reliable input to the PCIS system.

2. REQUIREMENTS 2.1 This calculation is performed in accordance with C lure 3.26.3, Instrument Setpoint and Channel Error Calculation Methodology. (Ref.

3. ASSUMPTIONS 3.1 The Barton differential pressure switches c ti *perature compensators. The auxiliary free-floating bellows attached to the high pressi.1 bel u cally protect the unit from zero and calibration drift when the instrument is subject dt c ii,ýnc ir ambient temperatures. Additionally, the differential pressure unit temperature comp ati envkl p,s the overall anticipated temperature ranges for both normal and trip conditions (-60' o fore for the purposes of this calculation, ATE and DTE are considered to equal zer . h* h operaites within the normal specified service conditions as identified in Reference 29x,-\

3.2 These devices are lo afeQ'in\ k'6actor Building South (R-903), which has as its Design Base Accident the HPCI Line B r T)ýT tsthr*ument is required to function for one hour under Trip and accident environmental c \i, n \7r Reference 28, the Humidity Effect (HE) = 0.

3.3ee calculated from test data (Ref. 19) for a Barton differential pressure transmitter, Modelý A stnstrument with an UR of 30 in. H20 and an initial setpoint of 25.38 in. H20 was tX7-subjecte, o*\(2 testing. The post test setpoint value of 25.79 in. H 20 was used to calculate an accuI I \ ii\ ient V UR as follows: (25.38 - 25.79) / 30 x 100 = 1.37% UR. These effects apply to trip

`"nll These errors are considered to be 1 sigma values per Reference 19.

3.4 aŽfatioi Effects (RE) were calculated from test data (Ref. 19). (Also see Assumption 3.3.) The test ins ent with an UR of 30 in. H20 and an initial setpoint of 25.12 in. H20 was subjected to radiation testing. The post test setpoint value of 24.90 in. H2 0 was used to calculate an accuracy in percent UR as follows: (25.12 - 24.90) / 30 x 100 = 0.733% UR. These effects apply to trip conditions only. These errors are considered to be 1 sigma values per Reference 19.

3.5 These differential pressure switches are bistable devices which operate on differential pressure and do not have a power supply. Power Supply Effects (PSE) will not be considered in this calculation. PSE = 0.

Nebraska Public Power District Sheet 4 of 16 1 DESIGN CALCULATIONS SHEET NPPD Generated Calculation Review of Non-NPPD Calc No: NEDC 92-050AM, Generated Calculation Rev. 1C1, MS-DPIS-116-119AJB/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

3.6 These devices do not contain any components that may be effected by radio frequency or electromagnetic emissions. Therefore, RFI/EMI Effect (REE) do not apply to these instruments.

3.7 The value for the Vendor Accuracy is 2 sigma unless otherwise stated (Reference 6).

3.8 The manufacturer does not specify Vendor Drift information for these instruments (Reference 13).

Therefore, the initial value used for Vendor Drift (VD) will be assumed equal to the random portion of Vendor Accuracy for 6 months, on a 2cy basis (References 3 and 4). This value and the value for DTE are replaced by analyzed CNS field data as appropriate (Reference 10).

A A

3.9 The values of the As Left Tolerance (ALT), Calibration Tool Accuracy (CTOOL) Tool Readability Accuracy (CREAD) and Calibration Standard Accuracy (CSTD) are o( 100%

testing. Therefore, they are assumed to represent 3 sigma values (Reference& 6 3.10 Vendor data for the Main Steam Line Flow Elements (Reference 15) listsý y:)0 0 2<X. 121.24 psid, this error is 3.95 psid, therefore PEA for this calculation is 3.95 ps for PMA is from Reference 3's example for calculating Main Steam Line High Flow dcfcurtion is stated as 2 % of rated flow, at a higher pressure than the design pressure of~the flo his is a conservative accuracy allowance and will be used to calculate PMA.

3.11 These instruments are not subject to Over Pressure Effects fore OPE = 0. Per Section 2.1.2 of CNS Improved Technical Specifications the pea oli pressure is 110% of vessel design pressure or 1337 psig at the Steam Do the safe working pressure of the instruments which is 1500 psig.

3.12 Static Pressure Effects (SPE) are considered neglitibje,ýf Barton differential pressure instruments therefore, SPE will equal zero for this calculatfoin. Therefore, the Static Pressure Effect (SPE) = 0. The Barton vendor manual states that a negligible-,atiý . ssure Effects (SPE) is an inherent feature of these instruments as identified in Referea-ce29. § ,--\\',I 3.13 Appendix K Power Uprate will* se tb'i, ensed Reactor Thermal Power (RTP) by 1.6%, to 2419 MWth. However, the AnalyticalI \N 7,main 121.24 psid (150% RTP at 2381 MWth) so that it will not be necessary to re-evaluat hlation exposure associated with the increased flow that would be experienced at 150% RTPat\-".N Vh (Ref. 36). At the Appendix K licensed power level of 2419 MWth, 121.24 psid will. P 147.64% RTP.

Appendix K A3atic'thi)dbi=i( (150% RTP) / 1.016 = 147.64% RTP = 121.24 psid 3.14 The

--- }lsation ea.*s ."o t'steam Valveflow expected (MSIV) in any one surveillance steamwhen testing line is one133%

MSIVofisrated flow.

closed and This occurs the normal

&n; st travel down the other three steam lines. With the implementation of the Appendix pý 133% of rated flow (at a licensed power of 2419 MWth) will correspond to 135.12% of licensed power of 2381 MWth).

Rated Flow (2419 MWth) = (133% Rated Flow) x 1.016 = 135.12% Rated Flow (2381 MWth)

Conservatively rounding up to 136% Rated Flow (at 2381 MWth), a differential pressure of 96.29 psid is obtained from reference 15.

Nebraska Public Power District Sheet 5 of 16 DESIGN CALCULATIONS SHEET NPPD Generated Calculation Review of Non-NPPD Calc No: NEDC 92-050AM, Generated Calculation Rev. 1C1, MS-DPIS-116-119A/B/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

4. METHODOLOGY 4.1 Instrument Channel Arrangement 4.1.1 Channel Diagram (Ref. 17)

LINE PCIS (GROUP 1) 4.1.2 Definition of Channel Each channel (loop) consists f 'j 4.1.3 Instrument Defim N\\ [IS Reference MS-DPIS-1 16A/B/C/ D 8 MS-DPIS-I 17A/B/C/ D 8 MS-DPIS- I 8A/B/C/ D 8 CIC: MS-DPIS-1 19A/B/C/ D 8 Manufacturer: ITT Barton 18 Model: 288A 18 Upper Range Limit (URL): 150 psid 8 Calibrated Span (SP): 0 - 150 psid 8 Input Signal: Process Pressure 15 Output Signal: Contact Closure 15

Nebraska Public Power District Sheet 6 of 16 1 DESIGN CALCULATIONS SHEET

..... No ED 20euacduatlon Review of Non-NPPD Caic No: NEDC 92-050AM, Generated Calculation Rev. 1C1, MS-DPIS-116-119A/B/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

4.1.3.2 Process and Physical Interfaces Reference 10 Calibration Temperature 65 - 104°F Range:

91 Calibration Interval 18 months +25% (grace period)

Normal Plant Conditions Temperature: 40OF- 104°F Radiation: 3.5x10 2 R (TID, 40 yrs)

Pressure: -0.10" -1.0" WG Humidity: 20% - 90% R.11 Trip Environment Conditions Temperature: 7 Radiation: 23 Pressure: 7 Humidity: 7 Long Term Accident Envi 'he -me as the Trip Environment Conditions since the instrt for service for one hour after the accident per Reference 18.

Seismic Conditions:

During Fun See Assumption 3.3 4.1.3.3 are random variables unless otherwise noted.

(VA) aigma Reference I 61 of Calibrated Span 2 29

).75 psid where Calibrated Span is 150 psid.

4.1.3.3.2 Accuracy Temperature Effect (ATE) = 0 Assumption 3.1

Nebraska Public Power District Sheet 7 of 16 DESIGN CALCULATIONS SHEET NPPD Generated Calculation Review of Non-NPPD Calc No: NEDC 92-050AM, Generated Calculation Rev. 1Cl, MS-DPIS-116-119AIB/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

4.1.3.3.3 Other Errors Value Sigma Reference OPE:= 0 Assumption 3.11 SPE: 0 Assumption 3.12 SE: = 2.0550 psid (1.37% x 150 psid) 1 Assumption 3.3 RE: = 1.0995 psid (0.733% x 150 psid) 1 Assumption 3.4 HE: = 0 Assumption 3.2 PSE: = 0 Assumption 3.5X REE: =0 4.1.3.3.4 Accuracy Values The identified accuracy error contributions are combihed'isinthe SRSS method to determine total device accuracy under ormin r trilconditions. The device accuracy is normalized to a 2 sigma con i dý ,lee,nld is given by:

Ai= 2 x SQRT[(VAi/n) 2 + (ATE +' PE')

E +,(SPEi/n) +

any biaermns Where the terms inside the sq aero)sni _ -- the random portions of the individual effects, and 'n' is the sigma val aQsocli.d with each individual effect.

Normal Accuracy Substituting values the under normal conditions is:

AiN = S ] psid For this there is only one device in the loop, thus the device and loop acc, equal. Thus the loop accuracy under normal conditions is:

S= +/- 0.75 psid ing values the Device Accuracy under trip conditions is:

= +/- 2 x SQRT[(0.75 / 2)2 + (2.0550 / 1)2 + (1.0995 / 1)2 ] psid

- 4.72 psid For this instrument loop, there is only one device in the loop, thus the device and loop accuracies (ALT)are equal. Thus the loop accuracy under trip conditions is:

ALT = +/- 4.72 psid 4.1.3.4 Determination of Individual Device Drift 4.1.3.4.1 Vendor Drift (VD)

Raw CNS field data was analyzed in the GE Instrument Trending Analysis System (GEITAS) shown in Appendix A. Results show:

VD 22 5 mo = +/- 7.26 psid (2 a).

Nebraska Public Power District Sheet 8 of 16 1 DESIGN CALCULATIONS SHEET NPPD Generated Calculation Review of Non-NPPD Calc No: NEDC 92-050AM, Generated Calculation Rev. 1C1, MS-DPIS-116-119A/B/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

4.1.3.4.2 Drift Temperature Effect (DTE)

This instrument has no drift temperature effect. Therefore DTE = 0 Assumption 3.1 4.1.3.4.3 Drift Values The Device Drift Error is calculated by SRSS combination of VD and DTE normalized to 2 sigma. A Di = 2 x SQRT[(VDMi / n) 2 + (DTEi / n) 2 ] + any bias terms Di = 2 x SQRT[(7.26/2) 2 ]

Di = +/- 7.26 psid For this instrument loop, there is only one device in ti vice and loop drift (DL)are equal. Thus:

I DL = +/- 7.26 psid This value is also shown in Appendix A.

4.1.3.5 Establishh The As-Left Tolerance for t iý'is established in CNS Surveillance Procedure 6.1(2)PCIS.303 as MS-DPIS-I 16-119A/B/C/D.

ALT = +/-3.0 psid a) (Ref. 6, 11, 12) 4.1.3.6 Determination of Devioce on Error 4.1.3.6.1 DeNt*e rror

ý~a~t .v-These instr e c ,"*ated with a 0 to 200 psig Heise gauge at the input, which in turn is q tit4by a dead weight tester (Ref. 11, 12, 16). Heise gauges are tempera ýe . *en* ted from -25 'F to 125 'F (Reference 30) and have the same accu21ji bU e elow. The dead weight testers have no temperature effect (Ref.

e instruments are switches, no calibration tool is needed for the "alibration Tool: Heise Gauge 0-200 psig

\;,ccuracy (CTOOLin) = +/-0.2 psig (0.1% of Full Scale) (3a) (Ref. 21)

Readability (CREADin) = +0.1 psid (1/2 Minor Division) (3cy) (Ref. 21)

X Calibration Standard: Mansfield & Green Model "RK" Dead Weight Tester Accuracy (CSTDin) = +/-0.1 psid(0.05% of Reading) (3c;) (Ref. 21)

Output: N/A y

Nebraska Public Power District Sheet 9 of 16 DESIGN CALCULATIONS SHEET NPPD Generated Calculation Review of Non-NPPD Calc No: NEDC 92-050AM, Generated Calculation Rev. iC1, MS-DPIS-116-119A/B/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

4.1.3.6.2Device Calibration Error The Calibration Error ( Ci ) for Device "i" is the SRSS combination of the As-Left Tolerance (ALT), and the errors due to input calibration tools (including tool accuracy and readability and the error of the calibration standards). Thus, on a 2 sigma basis the calibration error is:

Ci = 2 x SQRT[(ALTi / n ) 2 + (CTOOLin / n) 2 + (CREADin /n) 2 +

(CSTDin / n) 2 ] +/- any bias terms where 'n' is the sigma value associated with each individual term.

4.1.3.6.3 Device Calibration Error Values Since the values of ALT, CTOOL, CREAD and S..are controlled by 100%

testing, they are assumed to represent 3 sigma *alue ,so- 3, and the Device Calibration Error is.

Ci = 2 x SQRT[(3.00/3) 2 + (0.20/3)2 + 0. 0 + (0.10/3)2]

= 2.01 psid In this case the loop contain evice and Loop calibration errors are the same CL = 2 .01 psid (2 ) ,

4.1.4 Determination of Loop/Channel2l'ý{u For this calculation the lop Rbnas, nly one device, thus the device error values for Accuracy, Drift ankC I [aiio0Kae th same as those for the loop. These values have been reported in Section 4A X...

4.1.5 Determination o nd A PMA = 12 * ,("j l OWl)- 1) (Reference 31)

Where:

121 g lj

0% MSL Flow, (Reference 15)

E owk 00ýYMSL Flow, and 1 )52% (Flow1+ 2.0% Flow Error from 15 psia increase in Reactor Pressure). (Ref. 3)

\k I'v 121.24 x ((152/150) 2-1)

P +3.25 psid MA is a bias term in this calculation = + 3.25 psid.

And PEA = + 3.95 psid (2 a) (Reference 15) 4.1.6 Determination of Other Error Terms All error terms to be considered have been accounted for in the previous sections.

Nebraska Public Power District Sheet 10 of 16 DESIGN CALCULATIONS SHEET NPPD Generated Calculation Review of Non-NPPD Calc No: NEDC 92-050AM, Generated Calculation Rev. IC1, MS-DPIS-116-119A/B/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

4.1.7 Calculation of Setpoint Margin and Operating Setpoint 4.1.7.1 Setpoint Margin The setpoint margin is defined as the margin between the nominal setpoint and the analytic limit. Based on Reference 6, this margin is given by:

SM = (I.645/N)(SRSS OF RANDOM TERMS) + BIAS TERMS Where N represents the number of standard deviations with which all the rando terms are characterized (normally 2 standard deviations) and 1.645 adjusts the resuftf a 95%

probability (one-sided normal).

The error terms are calculated for trip conditions, and the margin becomes SM = (1.645/N) x SQRT[ALT2 + CL2 + DL2 + PMA 2 + PEA\(it'I-AS )

= (1.645/2) x SQRT[4.72 2 + 2.012 +7.262 +3.952]1++/-2 5 psid

= 11.25 psid 4.1.7.2 Nominal Trip Setpoint (NTSP1) Calculati t rpi The Nominal given by: Trip Setpoint (NTSP1) for pro\ýsss va bl(es which increase to trip is NTSP1 = AL - SM The Analytic Limit (AL) is:

AL = 121.24 psid, correspo jn o ;0% flow RTPo (147.64% flow RTPK).

(Reference 15, Assumption 3.I3)

Therefore the nominal tp t 19.99 psid NTSP1 NTSPI 1 09, NTSPIrPresents tlý upper limit (closest to AV) at which the setpoint can be set assumin av alone tolerance in the direction toward the Allowable Value 4.1.7.3 1 w b e lue Calculation stpoint calculation the process variable increases to trip, so the Allowable al)u(AV) is calculated using the following equation (Ref. 4):

AV = AL - (1.645/N)(SRSS OF RANDOM TERMS) - BIAS TERMS Where N represents the number of standard deviations with which all the random terms are characterized (normally 2 standard deviations) and 1.645 adjusts the results to a 95% probability (one-sided normal).

Nebraska Public Power District Sheet 11 of 16 1 DESIGN CALCULATIONS SHEET NPPD Generated Calculation Review of Non-NPPD Calc No: NEDC 92-050AM, Generated Calculation Rev. 1C1, MS-DPIS-116-119A/B/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

The random errors include the random portion of ALT, CL, PMA, PEA, but exclude drift. Thus, AV = AL - (1.645/N) x SQRT[ALT2 + CL2 + PMA 2+ PEA 2] - ( Y BIAS TERMS)

AV = 121.24 - (1.645/2) x SQRT[4.72 2 + 2.012+ 3.952] - 3.25 psid AV = 121.24 - 8.58 = 112.66 psid AV = 112.66 psid AV as % flow = -145.4% (Ref. 15)

AV = 145 % Main Steam Line Flow (Rounded down conserv (Ref.

20)

AV = 145% Main Steam Flow = 111.82 psid (Ref-However, to maintain the same AV as calculated r to the-A end Power Uprate:

AV(adj usted) =111.73 psid Since this AV was calculated based on aiceensReR of 2381 MWth, 111.73 psid will correspond to - 142.7% ai tthe Appendix K Power Uprate is implemented.

AV = 145%*! 1.016 = 142 i team Line Flow 4.1.7.4 LE K Avoidance Evaluati Th e purpose of the ce Evaluation is to assure that there is sufficient ma rgin provi~d b + eillowable Value and the Nominal Trip Setpoint to rea sonablyvavo*i *l"* lQqs of the Tech Spec Allowable Value (which, when dis covered MUIN*

lance, could lead to LER conditions). The method of avo idmigi t'1Sst the Allowable Value is to determine the errors that may be pre sent *e illance testing, examine the margin between the calculated values of N 4 a,1 -1 V, and then adjust NTSP1 to provide added margin if necessary.

+iu t~equation is used to determine the errors that would be expected to h\i t\cuto a potential LER situation.

K Sigma(LER) = (1/N)(SRSS Of RANDOM TERMS) iere N represents the number of standard deviations with which the random terms characterized (normally 2 standard deviations).

le

,4 Y

4.1.7.4.1 Random Terms Included In LER Avoidance The Random Terms that should be included in the LER Avoidance evaluations include:

Nebraska Public Power District Sheet 12 of 16 DESIGN CALCULATIONS SHEET NPPD Generated Calculation Review of Non-NPPD Calc No: NEDC 92-050AM, Generated Calculation Rev. 1C1, MS-DPIS-116-119A/B/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

Loop Accuracy under Normal plant Condition (ALN)

Loop Calibration Error (CL)

Loop Drift (DL)

Process and Primary Element Errors are not included because calibration and surveillance testing are performed using input signals which simulate the process and primary element input.

Sigma(LER) = (1/2) x SQRT[ALN2 + CL2 + DL2 ] psid Sigma(LER) = (1/2) x SQRT[0.752 +2.012 + 7.262] psid Sigma(LER) = 3.79 psid 4.1.7.4.2 LER Margin Calculation a Once the value of Sigma(LER) is determined, the( argin t em th alues of NTSPI and AV is calculated in terms of Sigma(LER) she quation below:

Z(LER) = IAV-NTSPlI / Sigma(L ER)

This value of Z is then used to determine e ro' lity of violating the Allowable Value by treating the error distribution as-a random ormal Distribution, and then determining the area under the cur-veo nD 'bution corresponding to the number of standard deviationse- h Z(LER) = 1111.73 -

Z(LER) = 0.46 4.1.7.4.3 GE Recommedati GE recommd'n thj-ýýaoniia',probability of 90% for avoiding an LER condition be used as th -iccc"t iterion for the LER Avoidance (or Tech Spec Action Avoidance)

t .= or a single instrument channel, the value of Z(LER) correspondin ,lt t 0% criterion is 1.29 or greater. For an instrument channel which ii -*in- Itiple channel logic system a value of Z(LER) greater than 0.81 can s t ech Spec Action Avoidance criterion.

es<se"iffential pressure indicating switches are treated as single units and the It value of Z(LER) will be compared with 1.29 (Ref. 10).

.1...4 Governing Setpoint Determination comparison of Z(LER) with the limiting value for LER(Tech Spec) avoidance sows that:

Z(LER) = 0.46 < 1.29 This means that the LER(Tech Spec) avoidance criterion is not met if the setpoint limit is set at NTSP 1, and therefore NTSP2 will be calculated.

1.29 = INTSP2 - AVI / SIGMA (LER)

Nebraska Public Power District Sheet 13 of 16 1 DESIGN CALCULATIONS SHEET iNrriD ueneratea Calculation Review of Non-NPPD Calc No: NEDC 92-050AM, Generated Calculation Rev. 1C1, MS-DPIS-116-119A/B/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

NTSP2 = AV-(1.29*Sigma (LER)) psid NTSP2 = 111.73 - (1.29*3.79) psid NTSP2 = 106.85 psid 4.1.7.5 Selection of Operating Setpoints It is recommended that the method of using NTSP as the center of the Leave Alone Zone be used. Thus, according to Ref. 4, the nominal setpoint is: .4 44 NTSP = 106.85 +/- LAT Where the LAT is the SRSS combination of the leave alone tol the devices in the loop. N 4.1.7.6 Establishing Leave Alone Zones Since this loop has only one device, the loop Leave) 3fe Tolerance is the device Leave Alone Tolerance. As shown in Appendix Ath( ' ',I-analysis resulted in:

LAT = +/- 5.88 psid The calculated LAT is rounded up con! iumber of decimal places in CREAD and to the next increme tqFEp, Therefore: K\ "

LAT = +/- 5.9 psid The Leave Alone Zone (I by 1d1\iition, is twice the absolute value of the Leave Alone Tolerance. 4 4.1.7.7 The Req

ý%2"aElhtion calculates an adjustment to NTSP for the case when NTSP is *e'hter of the leave alone zone. The adjustment assures that with the stacb 1 ors (including leave alone tolerances) for all the devices in the loop* .ht i margin for Tech Spec Action Avoidance (or LER avoidance).

,imit (RL) of device "i" with the largest LAT in the loop is:

"=NTSP + LATi = 106.85 + 5.9 112.75 psid Kýis is compared against AV = 111.73 psid from Section 4.1.7.3, and Since RLi = 112.75 > AV = 111.73 adjustment of NTSP is required at this stage of the calculation.

NTSP(ADJ) = AV - LAT NTSP(ADJ) = 111.73 - 5.9 psid Y NTSP(ADJ) = 105.83 psid To determine if further adjustment is needed, the Required Limits of all devices in the loop are calculated, and Sigma(LER, RL) given by the following equation (Ref.

4) is calculated:

2 2 2 Sigma(LER, RL) = (1/2) x SQRT[Y((2/3)(RL -NTSP(ADJ))) + CL + DL ]

Nebraska Public Power District Sheet 14 of 16 1 DESIGN CALCULATIONS SHEET NPPD Generated Calculation Review of Non-NPPD Calc No: NEDC 92-050AM, Generated Calculation Rev. 1C1, MS-DPIS-116-119A/B/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

Also compute Z(LER, RL) given by:

Z(LER, RL) = lAV- NTSP(ADJ)I / Sigma(LER,RL)

For this case of a single channel, If Z(LER, RL) > 1.29 then LER avoidance condition is met.

Since for this calculation there is only one device in this loop A Sigrna(LER, RL) = (1/2) x SQRT[((2/3)(RL -NTSP(ADJ)))2 +

= (1/2) x SQRT[((2/3)(112.75 -105.83))2 + 2.012 Sigma(LER, RL) = 4.42 psid and Z(LER, RL) = ABS(1 11.73 - 105.83) / 4.42

=5.9 / 4.42 = 1.33 Z(LER, RL) = 1.33 This value of Z(LER, RL) is greater tha on for a single channel of 1.29.

Therefore the criterion is met without further d'usgts.

4.1.7.8 Selection of Operating Setpoint The recommended Operating e t, OSP = 105.8 psid d do' conservatively) (Ref. 20) 4.1.7.9 Snurious Trin Avoidance Evalua n The Spurio Avoi* vNalluation is used to ensure that there is a reasonable probability to\ *)swill not occur using the selected NTSP, and an alternate in ,sl.elt~ihgya limit for the Leave Alone Zone. The method of avoiding sl t p 1 0,odetermine the errors that may be present during normal plant opera t-mine the margin between the worst applicable Operational Limit fo- is not required, and the lower limit (NTSP3) of selected NTSP (adjustei Leave Alone Zones). setpoint.

n eam

-IiaLine High Flow trip setpoint must have adequate margin from the S\

t iiin\u{plerational steam flow to ensure that spurious trips will not occur. In this

.st maximum operational steam flow would occur during Main Steam Isolation

,Valve (MSIV) surveillance testing when one of the four MSIV's is closed to verify V"perability. When this testing occurs, the steam from the tested steam line must then flow down the other three lines, causing the flow to raise to 133% of rated flow.

"I This corresponds to 96.29 psid (Assumption 3.14).

The following equation is used to determine the errors that would be expected to contribute to a potential spurious trip with consideration that the Operational Limit 2" (OL) is equal to the maximum operational steam flow of 133% (96.29 psid).

Sigma (STA) = (1/N)(SRSS OF RANDOM TERMS)

Sigma (STA) = (1/2) x SQRT[ALN2 + CL2 + DL2 + PEA 2]

Sigma (STA) = (1/2) x SQRT[(0.75) 2+ (2.0 1)2+ (3.95)2]

Sigma (STA) = (0.5)(4.495) = 2.25

Nebraska Public Power District Sheet 15 of 16 DESIGN CALCULATIONS SHEET NPPD Generated Calculation Review of Non-NPPD Calc No: NEDC 92-050AM, Generated Calculation Rev. 1C1, MS-DPIS-116-119A/B/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

References 3 & 4 recommend that a nominal probability of 95% or greater of avoiding a Spurious Trip be used as the acceptance criterion for the Spurious Trip Avoidance Evaluation. For a single instrument channel, the value of ZSTA corresponding to this 95% criterion is 1.65 (or greater). ZSTA is calculated as follows:

Z(STA) = INTSP(ADJ) - Operational Limitl / Sigma (STA)

Z (STA)= 1105.8 - 96.291 /2.25 Z(STA)= 4.23 In this case Z(STA) > 1.65. Therefore, the Spurious Trip Avoidai without further adjustments to the setpoint. 1 4.1.7.10 Elevation Correction The Main Steam flow elements are mounted in ve a ectiods--f the steam lines (Ref 32), and therefore their pressure taps are at clffer i(-jtL, lions (Taps are 21.5" apart, Ref. 33). However, the sensing lj /for4hinstruments are configured with condensing chambers at equal elevations iWtv"ide equalized pressure sensing legs (Ref. 34, 35). Therefore, there is n -onAorrection necessary for these instruments. skX\Yci*

4.1.7.11 Determination of Actual Fic The determination of Actual will be as follows:

Actual Field Setpoint = Correction Actual Field .0 Actual Fie]

0

Nebraska Public Power District Sheet 16 of 16 1 DESIGN CALCULATIONS SHEET NPPD Generated Calculation Review of Non-NPPD Calc No: NEDC 92-050AM, Generated Calculation Rev. 1C1, MS-DPIS-116-119A/B/C/D Prepared By: Mark E. Unruh Company's Name:

Setpoint Calculation Date: October 19, 2007 NPPD Reviewer:

Checked By: J. Neddenriep Date: 11-08-07 Date:

5. CONCLUSION The setpoints and calibration data for instrument MS-DPIS-1 16-119A/B/C/D are as follows:

Analytical Limit 121.24 psid Analytical Limit, as % Main Steam Line Flow 147.64 % Flow Allowable Value 111.7 Allowable Value, as % Main Steam Line Flow 142.7 Operating Setpoint Leave Alone Tolerance As-Left Tolerance

6. ATTACHMENTS Appendix A - Drift Data Summary Appendix B - GE Project Task Report GE-NE-0 i Public Power District Cooper Nuclear Station Thermal T0506: TS Instrument Setpoints", Revision 0, October 200'

NLS2008034 Enclosure 4 APRM - RBM Setpoint Calculation NEDC 98-024, Revision 5C1 (75 Pages)

Cooper Nuclear Station, Docket No. 50-298, DPR-46 Information Only

COPY OhA/

Pace 1of 75 Title APRM - RBM Setpoint Calculation Calculation Number: 98-024 CED/EE Number: EE 07-01 System/Structure NM Setpoint Change/Part Eval Number:

Component NM-NAM-AR 2, 3, 4, 5, 6, 7, 8, 9 Discipline: Instrument and Control Classification: [ X ] Essential [ ] Non-Essential SQAP Requirements Met? [ ] Yes; [X I N/A Proprietary Information Included? [I Yes; [XI No Calc.

Description:

Revision 5C1: Revision to support the implementation of the Appendix K Power Uprate (per EE 07-01). The implementation of Appendix Kwill result in changes to the Analytical Limits for the Flow Biased APRM SCRAM and Flow Biased Rod Block. Changes also included several editorial corrections to make the calculation information easier to understand.

Conclusions and Recommendations:

Setpoints and Technical Requirements Manual (Allowable Value) for the APRM Flow Biased SCRAM and Flow Biased Rod Block are listed in Section 5, Conclusions. These values will ensure that there is still adequate margin to the Rod Block and Flow Biased SCRAM values respectively.

5C1 3 JAL Changes dues to Appendix KPower Uprate. Mark Unruh I/a- 4 .461z 4/e.. f.- #/C .. ,

11/01/07 g, . z*.,,__¢. ,

11-e-4,J // i "7 Rev. Status Prepared Checked or Design Approved No. Revision Description By/Date Reviewed By/Date Verification/Date By/Date Status Codes

1. As - Built 4. Superseded or Deleted 7. PRA/SPA

2. Information only 5. OD/OE Support Only

3. For Construction 6. Maintenance Activity Support Only

Page: 2 of 75 NEDC 98-024 Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATION CROSS-REFERENCE INDEX ITEM REV. PENDING CHANGES NO. NO. TO DESIGN INPUTS 1 NEDC-32676P 1/97 2 NEDC-31892P 1 3 GENE- 187-27-1292 12/92 4 EQDP46 9 5 GE Spec. 23A1399 I 6 GE Spec. 22A2811 3 7 GE Spec. Data Sheet 22A281AC 0 8 GE IDS 248A9730NS 0 9 GE IDS 234A930INS 9 10 GE Spec. 21A1368 2 11 Design and Perf. Spec. 175A9679 0 12 Design and Perf. Spec. 235A1386 1 13 257HA392AD 4, 14 DC89-219 (a\

15 CED 1999-0117 0\

16 GE-NE-L12-00867-01-01 3 17 NEDCO03-013 18 DI-004 19 GE-NE-0000-0066-5426-RO (T sk 0 :

Neutron Monitoring System w 20 GE-NE-0000-0063-6433-RO s 1 0 0 Instrument Setpoints) ___ _

Page: 3 of 75 NEDC 98-024 Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATION CROSS-REFERENCE INDEX ITEM AFFECTED DOCUMENTS REV. NUMBER NO.

N/A A -0 NMV.

Page4 of 75 NEDC: 98-024 Rev. Number: 5C1 The purpose of this form is to assist the Preparer in screening new and revised design calculations to determine potential impacts to procedures and plant operations.@

SCREENING QUESTIONS YES NO UNCERTAIN

1. Does it involve the addition, deletion, or manipulation of a I] [x] I[]

component or components which could impact a system lineup and/or checklist for valves, power supplies (breakers), process control switches, HVAC dampers, or instruments?

2. Could it impact system operating parameters (e.g., temperatures, [X]

flow rates, pressures, voltage, or fluid chemistry)?

3. Does it impact equipment operation or response such as valve []

closure time? \\L\]

A ~ T

4. Does it involve assumptions or necessitate changes tose ,,q6eýi ing [ I Lx] [I of operational steps?

5. Does it transfer an electrical load to a different circuit oII I] [Ix []

when electrical loads are added to or remove r M ýYs during an event?

6. Does it influence fuse, breaker, or relay co [] [x]

[]

7. Does it have the potential to affect the analy' cqio lions []

environment for any part of the Reactofj3mling, Cntainn Control Room?

8. Does it affect TS/TS BasesrilSAl Ž& cnsing Basis [xI []

[]

documents?

[X] [] []

a nh-, ;+ ff.,~+ T)OTh),9

10. Does it have the po ýdures in any way not already mentioned s in Procedure EDP-06)? If [] [x] [1 so, identify: <

7 Aý If all \xth~en additional review or assistance is not required.

If an- \TES or UNCERTAIN, then the Preparer shall obtain assistance from the

,1'

and other departments, as appropriate, to determine impacts to procedures and Affected documents shall be listed on Attachment 2.

K

Page 5 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET REFERENCES

1. USAR Sections 111-7.6.4, Flow Control, VII-5.9, Average Power Range Monitor Subsystem; VII-5.10, Rod Block Monitor Subsystem; VII-7.5.3, Rod Block Interlocks.

2. J. E. Walker, P.D. Knecht, Analytical Limits for Cooper Nuclear Station, NEDC-32676P, General ctric Company, San Jose, CA, January 1997.

3. General Electric Report NEDC-31892P, Revision 1, May 1991, Extended Load Line Li ý'

Improvement Program Analyses for Cooper Nuclear Station Cycle 14.

4. W.H. Cooley, J.L. Leong, M.A. Smith and S. Wolf, General Electric Instr e et n NEDC-31336P-A, General Electric Company, San Jose, CA, September 199o.

5. W.H. Cooley, Setpoint Calculation Guidelines for the Cooper Nuclear Stat'on\ DE- 8-1090, Rev. 0, General Electric Nuclear Energy, San Jose, CA, January 25, 1991.

6. GE Report GENE-187-27-1292, DRF-AOO-05122, "Neutro ew Analytical Limits for Cooper Nuclear Station", December 1992.

7. CNS Engineering Procedure 3.26.3, Rev. 5, Instrument t o11t n tjiTnnel Error Calculation Methodology.

8. VM 1025, Volume 8, Part 4, (198-4532K16-30 5GDifferential Pressure Transmitter Instructions, Rev 79.

9. CNS Surveillance Procedure 6.IAPRM.305, Re- 20( /L . 305, Rev 21, APRM System (Flow Bias and Startup) Channel Calibration

10. CNS Surveillance Procedure 6.1 P "M.3Q2-,__.qv7126 M.302, Rev 10, RBM Channel Calibration.

11. CNS Surveillance Procedure ` R < .2RR.303, Rev 16, Reactor Recirculation Flow Unit Transmitter and Flow Unit Cyc ,46ann.l

12. GE Elementary Diagram Po R kýptron Monitoring System, 197R148, Sheet 2, Rev. N03; Sheet 3, Rev. N07; Sheet 4, Rev. N - ev. N02; Sheet 13, Rev. N05.

13. GE Elementary Diagran real rection 0 System, 791E256, Sheet 9, Rev. N19; Sheet 10, Rev. N13.

14. EQDP 46, Rev. 9- io\-1 Conditions.

15. GE Letter, J1 'o o R. Bussard (NPPD), Subject "Cooper Low Power APRM Analytic Limits",

Date Oc4 1 1 9192.

16. (o \,i 928823, P. Ballinger (NPPD) to J. Leong (GE), "LPRM Information / APRM A LýN',Noember 13, 1992.

Sel~pi~a 9 o. er Nuclear Station Improved Technical Specifications.

20 GE Letter, C960911 to CNS (Gautam Sen), Telephone Conversation Confirmation (regarding CNS Setpoint Analysis), September 11, 1996.

21. CNS Instrument and Control Procedure 14.1.2. 1, Rev. 13, IAC Test Gauge Calibration.

22. GE Design Specification, 23A1399, Neutron Monitoring System (RBM/ARTS), Rev. 1.

Page 6 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

23. Not Used.

24. GE Neutron Monitoring System Design Specification, 22A281 1, Rev. 3.

25. GE Neutron Monitoring System Design Specification Data Sheet, 22A281 IAC, Rev. 0.

26. GE Neutron Monitoring System Instrument Data Sheet, 248A9730NS, Rev. 0.

27. GE Nuclear Boiler Instrument Data Sheet, 234A930 INS, Rev. 9.

28. GE Recirculation Flow Element Specification, 21A1368, Rev. 2.

29. VM 1177, RR Venturi Flow Elements, Rev. 0

30. VM 1025, Volume 4, Part 2, (GEK-34550C), Power Range Neutron Monitoring S Modification), Rev 79.

31. VM 1025, Volume 4, Part 1, (GEK-34551B), Power Range Neutron Monilring mp nts, Rev 79.

32. Flow Unit (GE Dwg 791E392NSG1; Design & Perf Spec 225 ,65. s VM 0067 (GEK-34642D),

Flow Unit OMI, Rev 9.

33. Local Power Range Monitor Design and Performance Specific oni1. v. 0.

34. APRM Page Design and Performance Specification65 *eA13

35. Nuclear Engineering Data Book - Nuclear Instrume tat n Co p r Stion, 257HA392AD Rev. 4.

36. Average Power Range & Flow Converter Specification, 7 8 0, Rev. 0

37. VM 1518, DVM Fluke 45, Rev. 1.

38. VM 1575, Pneumatic Calibrator rysta i , ev 2

39. VM 1137, Ametek Type RK Dea~t ei ev. 0.

40. VM 1045, Fluke 8600A Digita , uction Manual, Rev. 5

41. Letter, J.S. Chamley (GE) e D), Subject "Analytical Limits for Neutron Monitoring System",

December 12, 1996.

42. Memo, P. Ballingerz l. Burch (CNS), "Review of NEDC 92-50S, Rev. 3 and NEDC 95-109, Rev. 1", Dated Jk .ay'l90)7.

43. GE SusceptjbXe g Performance Specification, 225A4338, Rev. 0.

44. D 2A ATrELLA Implementation.

45. S d nation of Radio Frequency Interference (RFI) by Hello Direct Wireless Headsets in the 46^§..9....festing of Permanent Cellular Phones.

SýP- 9 09, Testing SAIC Model PDE-4 and PD-4 Teledosimetry and Repeater Units.

. r-004, Impell Design Input

49. NUREG-1433, Vol. 1, Rev. 1, Standard Technical Specifications, GE Plants, BWR/4, dated April 1995.

50. VM 1106, Fluke Model 8502A Digital Multimeter Instruction Manual, Rev. 1

51. GE Letter, from D. J. Bouchie (GE) to Elden Plettner (CNS), dated July 8, 1998, APRM Restricted Condition Definition.

Page 7 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

52. GE Letter NPPD-R-98062, from Richard Rossi (GE) to Elden Plettner (CNS), dated July 22, 1998, Impact of Questions on APRM!RBM Calculations.

53. GE Calculation GE-NE-A41-00065-01-02-04-05-06-07 Rev. 1, Average Power Range Monitor (APRM),

Rod Block Monitor (RBM) and Technical Specification (ARTS) and Power Range Monitoring Setpoint Calculations (NEDC 92-050S, Rev. 3)

54. Cooper Nuclear Station MIG Project, GE-NE-L12-00867-01-01, Rev 3, Reactor Power/Flow

55. CED 1999-0117, Cycle 20 Core Reload.

56. Supplemental Reload Analysis for Cooper Nuclear Station Reload 21 Cycle 000 - 8 ,

Rev 0, January 2003.

57. NEDC 03-013, Normal Environmental Radiation for EQ, Rev 3.

58. GE-NE-0000-0066-5426-RO, "Nebraska Public Power District Cooper Nu-ear it,4 ermal Power Optimization - Task T0500: Neutron Monitoring System w/RB1 ' evi4onV, September 2007.

59. GE-NE-0000-0063-6433-RO, "Nebraska Public Power District Coo k Station Thermal Power Nrce Optimization - Task T0506: TS Instrument Setpoints", Revision c b20-

60. CNS Nuclear Performance Procedure 10.1, "APRM ai

Page 8 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

1. PURPOSE In consideration of the Cooper setpoint verification program in conjunction with a 7.5 month surveillance interval (required 6 months plus 25% grace period), determine the Nominal Trip Setpoint and Allowable Value for the Reactor Protection System (RPS) scrams from the Average Power Range Monitoring igh Neutron Flux, Flow Biased, and Low Power (Setdown) High Neutron Flux trip functions. lso considerations of allowable APRM gain adjustment factors (AGAF) of 0.98 to 1.02 will be m S Technical Specifications SR 3.3.1.1.2).

In conjunction with a 7.5 month surveillance interval (required 6 months plus 250/ ace 0o the Nominal Trip Setpoint and Allowable Value for the Rod Block Monitori (en* ' f System (NM-NAM-AR5 and NM-NAM-AR6) monitors local neutron flux ar d a cnor withdrawal, and blocks control rod withdrawal when neutron flux exceeds re efine dep dent setpoints, Reference 1.

2. REQUIREMENTS 2.1 The APRM System (NM-NAM-AR2, NM-NAM-AR3, NM-NAt --i)O- -NAM-AR7, NM-NAM-AR8, NM-NAM-AR9) monitors average neutron flux thrOU ouu t ell 1 ' di ovides a rod block and scram at two separate flow-biased setpoints. The s) h: herequirement of providing rod blocks and scrams at other lower setpoints wheyI\t re to d itch is in a mode other than RUN (rod block in STARTUP, and scram in REFUEL or TATIRIT T STANDBY), Reference 1. Per References 2, 6, 15, 41, 54, 58 and 59, the Analytical rnitsk or e APRM Trip Channels are as follows; APRM Trip Function nal ical Limit Flow Biased Scram 0.75W + 65.6% RTP Flow Biased Rod Block 0.75W + 54.8% RTP High Neutron Flux Scram 123.0% RTP Rod Block Clamp 112.2% RTP Downscale Neutron Flux ck 0.0% RTP High Flux - Setdown So a 17.4% RTP High Flux - Setdow 3: Iod 14.4% RTP Where "W" is the lation flow rate in percent of rated (rated loop recirculation loop flow rate is that recircu ti o which provides 100% core flow at 100% power).

2.2 The APRM, onitor, and Technical Specifications (ARTS) / Extended Load Line Limit

.lo A a is(m ementation of DC-89-219 physically reconfigured the RBM and changed the A I- i he setpoints for both the RBM and APRM (in RUN mode), Reference 3. Per Refl n.i 2 e and 22, the Analytical Limits for the RBM ARTS Trip Channels and Nominal Trip

%t o t rTime Delay (Tdl) and Time Constants (Tcl, Tc2)* , Reference 44, are as follows; rip_ Function Analytical Limit o)Power Setpoint (LPSP) 30%

iidermediate Power Setpoint (IPSP) 65%

High Power Setpoint (HPSP) 85%

T

Page 9 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Analytical Limit MCPR Limit Low Trip Setpoint (LTSP) 117.0% 1.20 120.0% 1.25 123.0% 1.30 125.8% 1.35 Intermediate Trip Setpoint (ITSP) 111.2% I.

115.2%

118.0%

121.0%

High Trip Setpoint (HTSP) 107.4%

110.2%

11 - 1.30 1.35 Downscale Trip Setpoint (DTSP)

Time Delay 1 (Tdl)

Time Constant 1 (Tcl) sec.

Time Constant 2 (Tc2) 6 sec.

Time Delay I (Tdl): Delays nulling se, ction so RBM filtered signal nears equilibrium before calibration; no delay with t filt time delay from rod selection to allowable rod withdrawal start. t Z Time Constant 1 (Tcl): RBM si a ite constant.

Time Constant 2 (Tc2): Var i " gnal filter constant. Does not affect RWE transient response.

- The Downscale trip se t t( functions to prevent a rod withdrawal if the selected RBM channel power is too low fr m*os ent normalized calibration conditions (i.e. 100%). This assures that the calibration (i.e.,,, tia ý Iperformed at the time of rod selection remains valid before permitting withdrawal o l e o e Analytical Limit was changed from 91% to 89% of reference level per Refence 6. eTimFmit is not utilized in any licensing bases Rod Withdrawal Error (RWE) analysis.

Tef itation associated with the DTSP.

2.3 T"aý\ jl s performed in accordance with CNS Engineering Procedure 3.26.3, Instrument Setpoint 1I t rror Calculations (Reference 7).

Is used in this calculation are consistent with the requirements of Reg. Guide 1.105 that the GE Lsll eent Setpoint Methodology (Reference 4) is in compliance.

Page 10 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

3. ASSUMPTIONS 3.1 The GE APRMiRBM equipment accuracy specification includes the uncertainties due to seismic effect on the equipment located in the Neutron Monitoring System equipment panels. All equipment in these panels are qualified as a unit.

3.2 The recirculation loop flow transmitters are classified as non-essential instruments. These instruirr-tlare rigidly mounted and their ZPA (zero period acceleration) during a seismic event would be ins1 ncat.

Thus Seismic Effects (SE) will not be considered for this calculation.

3.3 The values of the As Left Tolerance, CTOOL, and CREAD are controlled by 100 testi e are assumed to represent 3 sigma values, Reference 5. Calibrating equipmen ccura e-sea e as hree (3) sigma values due to industry required periodic calibration with high ac-cy sta ace to NIST. The accuracy of the calibration standard is assumed the same as that h0 accur cy f the testing equipment, unless otherwise specified.4 3.4 The manufacturer does not specify Vendor drift for the RBM si / ndfng equipment (Reference 22).

Therefore the value used for Vendor Drift (VD) will be assume',i\be u to the random portion of Vendor Accuracy for 6 months, on a 2-sigma basis (References T g term Vendor Drift for the RBM trip unit, is assumed to be adequate for the allowed/V I od between surveillance tests (assumed 3 months), based on GE's experience se 1 t orm ce in BWR plants.

3.5 The manufacturer does not specify Vendor drift for hecultio flow transmitters (Reference 8).

Therefore the value used for Vendor Drift (VD) wi ss d to be equal to the random portion of Vendor Accuracy for 6 months, on a 2-sigma basis (Refe rc and 5).

3.6 For ARTS operation, setpoints for the RBM 4 i are considered (Reference 3). Table 10-5(b) of Reference 3 states, for these items that noi tt i s *o setpoint does not affect the RWE analysis or the range is restricted by design to es c s e 'i e RWE analysis). The time delay (Tdl) and time constant (Tc1, Tc2) settings crrenF \uwea sumed to be valid, and it is assumed that no setpoint calculations (using setpoint metlýoV a -)ired for these timing functions.

3.7 The APRM/RBM Technical.-S>eiati (ARTS) improvement to the RBM does not degrade the instrument accuracy and dri h -in 3.8 The Radiation Effect- to t qulpment in the specified environment does not exceed the normal integrated dose spe i d in 1P nvironmental Design Conditions document (Reference 48 and 57).

3.9 The variation t n chamber output current with +1 percent change of the ion chamber voltage in the saturat Iý ge s ligibly small or equal to zero (Reference 4).

311\1 uipment is electrical and is not subject to Overpressure Effects (OPE). The 3.10 I II ýw transmitter has a design pressure rating of 2,000 psig (Reference 8), well above the e i] n Adent pressures that will be seen by this instrument.

\fly installed NMS equipment is the same as that originally supplied by GE other than normal PC h(tJbyGE) electronic upgrades (References 30, 3 1, and 32).

ss otherwise specified, the vendor accuracies are considered to be 2 sigma values.

3.1! he manufacturer does not specify a Power Supply Effect (PSE) for the APRM!RBM Technical Specification (ARTS) equipment and it is assumed to be included in the equipment accuracy.

Page I Iof 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 3.14 The APRM/RBM Technical Specification (ARTS) equipment is subject only to normal ambient environment and are not subject to harsh, post-accident conditions. Trip and accident environmental conditions will be considered equal to normal ambient conditions for the purpose of this calculation.

Accuracy Temperature Effect (ATE) and Humidity Effect (HE) will not be considered.

3.15 Static Pressure Effects (SPE) are generally only applicable for differential pressure ins ents (References 3 and 4). The SPE will only apply to the recirculation loop flow transmitter for ca,11a ns which involve flow signal inputs. Per References 8 and 52, for an assumed 1,000 psig pro s the SPE is equal to 0.88% span per 1,000 psig.

3.16 The flow element inaccuracy is assumed by References 28, 29 to be 2% of fib 0 3.17 The As Left Tolerance (ALT) allowance for the APRM gain adjustment fa to (AGA q ter 1 (NPPD allowables are 0.98 to 1.02) is treated as an ALT of 2% power and is on idered aalue since it is adjusted within set limits during calibration. This ALT is not included in the PRM N utron Flux High Rod Block - Setdown or the Neutron Flux High Scram - Setdow- becautsfýAGAF'P-iu n performed at that low power. The ALT for the LPRMS is assumed to be the samfha o'lth APRMS, Reference 9.

3.18 The ALT for the recirculation loop flow unit summer output is a ed a] to the sum of the two aq recirculation flow loop square root unit output, Reference 11.

3.19 The APRM/RBMIFlow Unit equipment meets e IQun11-01ceIk the Susceptibility Design and Performance Specification, Reference 43. For no a I an with expected operational transient radio frequency or electromagnetic emissions, there re L 'I/EMI Effects (REE). Peak transient REE that may occur during plant maintenance that may, at rformance of the APRM/RBMIFlow Unit equipment is not considered in this calculatioq RMIR low Unit equipment has been subjected to various testing for determination of effects4o ( (References 45, 46 & 47) and the results of these tests show no adverse effect on the co oiteeintroduce REE. Therefore, REE will not be considered for this calculation.

3.20 It is assumed that for all AP ý B M\-e, ronics in the Control Room, the stated accuracy includes temperature effects, so the ATn* ues are assumed to be zero.

3.21 Reference 38 gives a temrn ýrl C cy of 0.01% per 'F for 307F to 130'F for a crystal engineering calibrator. Therefore, e em tu accuracy for calibration temperatures from 65°F to 104'F is:

Temperature A a - o/°F x 39°F = 0.39% F.S..

3.22 Leave Alone T cA for the APRM and RBM functions is assumed to be equal to +/- 1.25% power for consist! is calculation. The use of +/- 1.25% is conservative since the current procedures (f ýS )Aixvae a AT of +/- 1.25% or less for the identified APRM and RBM functions.

P.23RM Rod Block Clamp function is assumed to be 1.00% for the purpose of this calu tI'a. ALT is consistent with the related APRM functions.

3.' Accuracy of multiple signals with equivalent accuracies entering a device is determined by

( i"'ýt g the mean of the SRSS of the signals as shown in the equation below:

2 2 2 V -VAi +VA 2 +...+VA, n

Where n = the number of inputs to the device. The equation can be re-written as follows:

VATOT - (VA) 1 I (n(VA)

VA n n J

Page 12 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 3.25The Vendor Accuracy of the LPRM signals entering the APRM averaging card (VALPRMJAPRM) is determined using the equation in Assumption 3.24:

VALPRM / APRM VALP Where n = the number of LPRM's that are input to the averaging card (channel). At CNS averaging cards can have a maximum of 14 or 17 signals to an averaging card, but only functional LPRM signals are required to for the channel to be declared Operable. Based on computing the Vendor Accuracy, the most conservative value of "n" is 11. Higer v result in lower values for VALPRM/APRM. The equation can now be written as follr

VALPRM / APRM- VALPRM Since the accuracies of the LPRM's are stated in terms of Full Sca 5) e esults o 15 equat converted to units of power. Since full scale is 125% of Reactor"e,0.r (RTP), the equ ation above is multiplied by 125%.

VALPRM/ APRMI V15~,j )Ax25%

Page 13 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

4. METHODOLOGY 4.1 Instrument Channel Arrangement 4.1.1 Channel Diagram (References 12, 13, 30)

APRM Channel RMCS Flow Unit Channel RBM Channel RMCS

Page 14 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 4.1.2 Definition of Channels The APRM channel (loop) consists of the LPRM neutron detector inputs and electronic signal conditioning equipment for the neutron flux trip logic. In addition to above, the flow biased trip logic includes input from the recirculation flow signal. The APRM panel electronics is located in the main control room. .A The RBM channel (loop) consists of the LPRM neutron detector inputs along power trip reference input. The RBM panel electronics is located in the main The Flow Unit channel (loop) consist of the recirculation transmil which outputs to the APRM and RBM for flow biased trips (also Block trip). The recirculation loop flow transmitters are located, instrument rack 25-7, northwest 859' elevation (Reference 11).

4.1.3 4.1.3.1 Instrument Definition Reference APRM/RBM Channels CIC: 17 17 Manufacturer: 26, 30 Model: 52 Upper Range Limit (UR): 24, 25, 26 Calibrated Range: " 24, 25, 26 Calibrated Span (SP): 25% Power 24, 25, 26 Output Sig 0 Vdc 24, 26 Vendor P~er1TSj-,eCC RR-FT- 110A-D 17 GE 26, 30 Type 555 8,27 850 in WC 27 0-125% Flow 30

(-5.7 in WC to 403.2 in WC) 11 Span (SP): 125% 30 (408.9" WC)

"Input Signal: differential pressure 11 Output Signal: 10-50 mV 11 (across precision 1 ohm resisto r)

Vendor Perf. Specs: See Section 4.1.3.3

Page 15 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Flow Unit Manuf: GE 32 Input Signals (2) 10 - 50 mA 32 Output Signal 0 - 10 Volts 32 Vendor Perf. Specs: See Section 4.1.3.3 4.1.3.2 Process and Physical Interfaces APRM/RBM/Flow Unit Referen Calibration Temperature 60 - 90 OF Range:

Calibration Interval 6 months (+25% grace) 9 18 months (+25% grace) 9 Normal Plant Conditions Temperature: 60 - 90 OF --- /1 Radiation: 1.75x 10'* 48 Pressure: 0.10" to14 Humidity: 40% - 50% 14 Trip Environment Conditions - e iq Temperature: 0- 0 14 Radiation: 1 ID ,40 yrs) 48 Pressure: ' WG 14 Humidity: 40O° R.H. 14 Temperature Range fo njtion Error Calculations:

[(max trip temp - min calib temp)

To.Tem hrgerý.ý of I 90 - 60 =or30 °F I (max calib temp - min trip temp)

L 90 - 60 =30OF 30 OF aDTE calc ( ATD) = max calib temp - min calib temp

=90 - 60 =30 OF or ge for ATE calc (ATAT) = ATT- ATD

= 30 - 30 = 0 F T nperature Range for Normal condition Error Calculations:

[(max norm temp - min calib temp) = 30 OF Tot. Temp range (ATN) = larger of I or L(max calib temp - min norm temp) = 30 OF 30 OF Temp range for DTE calc ( ATD) max calib temp - min calib temp

= 90 - 60 = 30 OF Temp range for ATE calc (ATAN) = ATN - ATD

= 30 - 30 = 0 OF Seismic Conditions - (if required):

Prior to Function: N/A Assumption 3.1 During Function: N/A Assumption 3.1

Page 16 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Process Conditions - (if required):

During Calibration: N/A Worst Case: N/A During Function: N/A Flow Transmitter Calibration Temperature 65 - 104 OF 2 Range:

Calibration Interval 18 months (+25% grace)

Normal Plant Conditions Temperature: 40 - 104 OF Radiation: 8.8x10 R (TID, 40 y Pressure: -0.10" to -

Humidity: 20% -90%R,1 14 Trip Environment Conditions Temperature: 01(H1 Radiation: 1rs) 57 Pressure: G 14 Humidity: 2H 14 Temperature Range ** fo ri [-(max n Ction Error trip temp - min calib temp)

Calculations:

9*YI =104 - 65 = 39°OF Tot. Tern r- arger of Ior I (max calib temp - min trip temp)

L =104 -40 =64°OF

=64 OF a oDTE r calc (ATD) = max calib temp - min calib temp

= 104- 65 = 39 OF gefor ATE calc (ATAT) = ATT - ATD

=64 - 39 = 25 OF

<& ýTmperature Range for Normal condition Error Calculations:

V [-(max norm temp - min calib temp) = 39 °F Tot. Temp range (ATN) = larger of I or L(max calib temp - min norm temp) = 64 OF

-64 OF Temp range for DTE calc ( ATD) max calib temp - min calib temp

=104 - 65 = 39 OF Temp range for ATE calc (ATAN) ATN - ATD

= 64 - 39 = 25 OF Seismic Conditions - (if required):

Prior to Function: 0 Assumption 3.2 During Function: 0 Assumption 3.2

Page 17 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Process Conditions - (if required):

During Calibration: N/A Worst Case: N/A During Function: N/A 4.1.3.3 Determination of Individual Device Accuracies All accuracy error contributions are random variables 4.1.3.3.1 Vendor Accuracy (VA) 4.1.3.3.1.1 APRM Channel Value Reference VALPRM Card = 0.8% FS 10 33, 35 VALPRM/APRM = { VALPRM d

= {(0.8% Q l 125%

T3 Assumption 3.25 VAApvM Avg. ir\ X8F 2 34 x125%

1.0% Power V ji, e = 1- I%FS 2 34

= 1.25% Power rip Unit Flow-Biased = 1% FS 2 34

= 1%x (125%)

= 1.25% Power VAFlow Transmitter = 0.4% Span 2 8 VARR Flow Element = 2% (Rated Flow) 2 28 4.1.3.3.1.2 RBM Channel VALPRM Cad = 0.8 % FS 2 33,35 VALPI*MJBM = (0.8%)/ [SQRT (2 LPRM's)] x 125%

= 0.707 % Power 2 Assumption 3.24 VAsignal Conditioning Eq = 1.32% FS 2 22

= 1.32% x (125%)

= 1.65% Power

Page 18 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET VATripUnit =0.5%FS 2 22

= 0.5% x (125%)

= 0.63% Power 4.1.3.3.1.3 Flow Unit VAi Flow Unit = 2.0 % FS 2 VAi Flow Transmitter = 0.4% Span 2 VAi RR Flow Element 2% Rated Flow 4.1.3.3.2 Accuracy Temperature Effect (ATE)

ATE for the recirculation GEMAo 55 tranm er Reference 8, is

+/-1% span per 100 'F at 100% 1% to +/-2% of span per 100 'F from 49% to 20% span As shown in 4.1.3.1 the calibrat_i. ]0*t..* WC which corresponds to 48.1% of the 850 er a I-t. temperature coefficient for 48.1% span is m ey r polation to be:

Temp Coeff= 1 + 1 \1.06%span per 100 degF Therefore, fo ,.Nalculation where ATAN = 25' F (from 4.1.3.2)

TE A~~'~~ - 06% span x 250 F/100'F= 0.27% span 4.1.3.3 e ,Recirculation Flow Loop) aSigma Reference E: 0 Assumption 3.10 SPE: 0.88% span 2 Assumption 3.15 SE: 0 Assumption 3.2 RE: 0 Assumption 3.8 HE: 0 Assumption 3.14 PSE: 0 Assumption 3.13 REE: 0 Assumption 3.19

Page 19 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 4.1.3.3.4 Accuracy Values The identified accuracy error contributions are combined using the SRSS method to determine total device accuracy under normal conditions. The device accuracy is normalized to a 2 sigma confidence level, and is given by: I,'A

'Ai 2 +ýATEj n )2 + (opE, ) 2 + ý SPEj I ) 2 + ( sE, 2 nEj ) 2 HE,n Where the terms inside the square root sign ,ji'niions 6f the individual effects, and 'n' is the sigma value each individual effect. 4 4.1.3.3.4.1 Normal Accuracy several devices in the loop. Thus normal conditions will be calculate the APRM loop 2 )2 (VA AApRA4Unj, =2 0.30IJ2 +(.0)

= 1.044 % Power 2 sigma b) Accuracy of APRM Trip Unit ATU (Flow Biased Trip Unit) = 1.25 %Power ATU (Fixed Trip Unit) = 1.25 % Power c) Accuracy of devices in the flow loop

1. Accuracy of Flow Transmitter VAGMAC 555 =0.40% span SPEGMAC 555 0.88% span at 1,000 psig ATEN GMAC 555 = 0.27% span

Page 20 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 2

AT= 2 1VAGMAC555 ~2 , (SPEGMAC 555 )2 4 ATENGMAC555 2

AFT = 2 040 ) +(0.8 2 +(0.2

= 1.00% span = 4.08 in WC The flow error at the output of*

w error from both loop transi ers h e a cý"

Appendix B to be:

FT Error = 0.7366 % flow 2.

The Flow ithZventuris used in the flow Tt~). The flow error at the

'to this error from both loop flow ed in Appendix B to be:

%flow of Flow Unit 16w Unit error for is 2% FS (Ref. 32). The flow error output of the flow unit due to this error has been atted in Appendix B to be:

FU Error = 2.5 % flow

4. Total Flow Channel Accuracy The total flow error due to the 2 transmitters, 2 flow elements and the I flow unit is:

Af ) 0.7366)2 (1.414 "2+(_15 2 AFC (ft+ fe + fi) = 2ý( 2 + 2) +*2

= 2.965 % flow This flow error can be converted to power error by multiplying by the Flow Control Trip Reference (FCTR) slope, which refers to the slope of the power/flow line (Ref.

1).

FCTR slope = 0.75 (W coefficient) (Ref.58)

Therefore AFCFT+FE+FU= 0.75 x 2.965 = 2.224 % power

Page 21 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

2. RBM Channel Accuracy Accuracy of modules in the RBM loop

1. Accuracy of RBM Unit (including LPRM) from 4.1.3.3.1.2 is 0AML~J=2+1.65j) 1.707I) 2 f

= 1.80 % Power

2. Accuracy of RBM Trip Unit 4.1. .T1.2 VARBM Trip Unit = 0.63% Power

3. Total Channel AcculNIs .

I IIU LULdI id and lBM functions are:

n ii A AAPJMff V /A=2 TUFBTripU\t 2" ,- 2 ALN-j 2ý( f n )Ul+ A + nFF+EF ALNIVb = 2j(o4-4)2+(1.25 2+ 2.224)2 ALN.fb = 2.76 % Power

3. RBM Power Function

)2 '(ATURBMTrqjUni, .j 2

ALNRBM-pwr 2 ( VAAyPMUfli, ALN-RBM-pw,. 2ý 1l.044 )2 +(0.63 2

Page 22 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET ALN-Rm-p, 1.22% Power

4. RBM Trip Function ALNRBM-trip =

ALN-RBM-trip 12

-- 2 ( 1.80 (0.63 )2 ALN RBM-.lp 1.91% Power 4.1.3.3.4.2 Trip Accuracy Since the normal anl tri e same, per assumption . fC1 'CI c i tr the same as accuracy un,Mrno ao C1"itls.j j~

4.1.3.4 J'APRM system that has a drift temperature effect is ow transmitter. For this device the error temperature span per 100'F (from 4.1.3.3.2). Therefore:

.06% / 100'F) x ATD, where ATD = 39°F (section 4.1.3.2).

= (1.06/100) x 39

= 0.413% span

'This DTE value has been included in the flow transmitter drift shown in 4.1.3.4.2 1(b).

4.1.3.4.2 Vendor Drift (VD)

The drift for APRM Trip Units was derived from analysis of site calibration data, and for the rest of the APRM and RBM processing electronics channels the drifts were derived from vendor drift and accuracy specifications.

Page 23 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

1. APRM Channel Drift (6 month + 25% grace = 7.5 months) a) APRM Electronics Drift

1. LPRM and APRM Unit Drift The specified drift for the LPRM and APRM Units are:

LPRM = 0.8 % FS/ 8 Weeks 2 sigma APRM =0.5 % FS 700 Hours 2sig n the drift times specified in the above speT.ff ations r weekly calibration interval of the AP e Ctronic balance and process computer calculations.IThee.fo the above drift values will be ed as ithou do in the drift calculation.

As done for VA in 4.1.3.3.1.1 th i r due tc reduced by the square root e m i mber of LPRMs in the APRM channel. Ti ekjj IC s rft error is:

= 0.555 % FS 1.25 % Power

% Power Trip Unit Drift Ahe drift error for the Trip Units was determined by analyzing field data by program Y-GEITAS (and GEITAS) as described in Appendix A.

Results of this calculation show that the Trip Units drift for 7.5 month is 1.34 % Power:

VDTU-fxed ftp = 1.34 % Power VDTU-flow-biased tip = 1.34 % Power b) Flow Channel Drift

1. Flow transmitter Drift Specified vendor drift is:

VDGEMAC 555 = 0.40% span per 6 months Assumption 3.5

Page 24 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Therefore the drift for 22.5 months is:

VDGMAC555 = 0.40% x SQRT(22.5 mo / 6 mo) for 22.5 mo.

= 0.775 % span The DTE value for the flow transmitter is:

DTEGMAC555 = 0.413 % span from section 4.1.3.4.1.

Therefore the total drift for the flow To convert ansmife-r drift in % span to % flow, use the method shown in K ,B(equation 10) and substitute DFT in place of AFT transmitter drift is:

0.647 % flow 2.F\1 nit Drift Ref. 32 the flow unit drift is specified to be:

Du= 1.25 % FS / 700 Hours Since the flow units are checked every month, it is assumed that the above drift is applicable for calculation.

Therefore the error due to flow unit drift is:

DFU = 1.25 x 10/8 = 1.56 % flow

3. Flow Element Drift The flow element drift is assumed to be negligible.

DFE = 0

Page 25 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

4. Total Flow Channel Drift The total drift of the flow channel is:

4.

The ofFlow Channel Total total DESIGN Nebraska Drift flow drift the CALCULATIONS Power Public SHEET channel District is:

DFC= ( F 2 DU)2+DE2 2

2 2 +(1.562 DFC DFC = 1.689 % flow To convert this flow channel error in % fic multiply by the FCTR slope shown in 4.1.3.3.4.1 4 DFC = 0.75 x 1.689 %

= 1.27 % Power

'Rez22) do not specify drift for the RBM signal hence it is assumed that the drift for 6 months is racy. (Ref. 5)

=VA xSQRT( 7.5 mo. / 6mo.)

1.65% x SQRT (7.5 / 6)

= 1.84 % Power

-M described in the APRM drift calculation above, the drift of the LPRM

,electronics is 0.8% FS (or 0.8 x 1.25 = 1.0% Power). Also, for RBM, the minimum number of LPRMs is 2. Therefore the overall drift of the RBM signal conditioning electronics is:

1.00) 2__ . 19%w VDR~/Lp~m = 2 v +2 2 1.91 % Power

Page 26 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET b) RBM Trip Unit Drift For the RBM Trip Unit, the vendor specification (Ref. 22) states that the drift for the maximum calibration period (assumed to be equal to the maximum previous calibration of 3 months plus 25% grace, or 3.75 months) is 0.4 % FS. A Drift (3.75 month) = 0.4% x 125 % Power = 0.50 % Power Therefore the drift for 7.5 months is:

DTU-rbm trip= 0.5% x SQRT (7.5 mo. / 3.75 m

= 0.5% x SQRT (7.5 / 3(7

= 0.71% Power DTU-rbm power = 0.5% X S* (7.

( /i3.

3T

= 0.5% x SQW 1 )

= 0.71% P4W 4.1.3.4.3 Drift Values The total Device I calculated the Dorby SRSS combination of the random portio- d and the DTE errors, and normalizing to 2 sigma. Bias or* added (or subtracted) separately.

_____ + bias n J

)al drift for the various channels is obtained by SRSS addition of drifts of the devices in that channel, and is shown below:

a) APRM Flow Biased Channel Drift DFB = 2J( VDj ~2 1 IPM/LPRM jVDTUflow-biased-trip j2+(DFc )J2 DFB2-- 2 2) 0.694 2--+(1.34

--2 -) 2 +(1272 2

= 1.97 % Power

Page 27 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET b) APRM Fixed Channel Drift VDTU-fixed-tri 2 Df =2 (1<DAPRM / LPRM ) 2 +

Dfix =2 ~0.694)2+( 1.34)

1.51% Power c) RBM Power Channel Drift DRBMPwr

DRBMPwr =

d) RBM n/

=2 2

= 2.04 % Power The As-Left Tolerance for the APRM and RBM channels are established as shown below. All values are assumed to be 3-sigma unless otherwise specified.

1. APRM Channels The basic ALT data for the APRM functions are:

Vdc % Power Ref.

ALT 1 = 0.08 1.00 LPRM Assumption 3.17 ALT 2 = 0.08 1.00 APRM NF Fixed High SCRAM Assumption 3.23 ALT2A = 0.08 1.0 APRM Downscale Rod Block 9

Page 28 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET ALT 2B 0.08 1.00 APRM Rod B lock Clamp Assumption 3.23 ALT3 = 0.08 1.00 APRM NF F-IB SCRAM Assumption 3.23 ALT 4 = 0.08 1.00 APRM NF F-!B Rod Block Assumption 3.23 ALT 5 = 0.05 0.5 APRM NF Set ]own SCRAM 9 ALT 6 = 0.05 0.5 APRM NF Set lown Rod Block ALT 7 = 2.0 (AGAF)

2. APRM Flow Reference Channel The basic ALT data for the APRM flow reference mVdc/Vdc % FS ALT 8 = 0.20 0.5 Xmit ou \t 11 ALT9 °= 0.01 0.01 test ALT9 = 0.005 10 .

ALT,, = 0.01 u Assumption 3.18 Since the transmitter is multiply %FS by 1.25%flow to obtain %flow for the ALT 8 = 0.5 x 1.25% fl 0`

ALT 9 = 0.Qý 1.2< o 4425 %flow ALTo 0 o0.0625 %flow

'ALTl, I \I .

Took'\ t2 I w 0.125 %flow to %power for the above ALTs, multipy by FCTR 0.75.

8 02 x 0.75 = 0.469 %power

%10.0125 x 0.75 = 0.01%power A*'I'* = 0.0625 x 0.75 = 0.047%power ALT,1 = 0.125 x 0.75 = 0.094 %power

3. RBM Channels The basic ALT data for the RBM functions are:

Vdc % Power Ref.

ALT 12 = 0.08 1.00 LPRM Assumption 3.17 ALT 13 = 0.08 1.00 LPSP 10 ALT14 = 0.08 1.00 IPSP 10

Page 29 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET ALT 15 = 0.08 1.00 HPSP 10 ALT 16 = 0.08 1.00 DTSP 10 ALT 7= 0.08 1.00 LTSP 10 ALT18 = 0.08 1.00 ITSP 10 ALT 19 = 0.08 1.00 HTSP 10 4.1.3.6 Determination of Device Calibration Error (Refs. 9, 12)

1. APRM Channels DVM Fluke 45 or Fluke 8600A C2 = DVM Fluke 45 or Fluke 8600A CSTD 2 = C2 C 3 = DVM Fluke 45 or Fluke 8600A CSTD 3 = C 3

Page 30 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

2. APRM Flow Reference Channel (Refs. 9. 11. 12)

Ametek RK C5A, B 8502A, or Fluke 8600A 45, Fluke 8502A, or Fluke 8600A VM Fluke 45, Fluke 8502A, or Fluke 8600A FAB= C7A, B

= DVM Fluke 45, Fluke 8502A, or Fluke 8600A FDt = Cg

Page 31 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

3. RBM Channel (Refs. 10, 12)

RBM TRP ]

C9 C*9std Calibration Equipment C 9 = DVM Fluke 45 or Fluke A CSTD 9 = C9 /*

Tool Error and RBM channels are calibrated using Digital Voltmeters

.Q I hich can be the Fluke 45, Fluke 8502A, or Fluke 8600A per

-eetnces9, 10. The DVMs are sent off site for calibration against a scdard.. Therefore the calibration tool error is assumed to be equal to the

ýcalibration standard error. The least accurate DVM calibration tool is used as bounding in this calculation. (References 37, 40, 50)

The recirculation flow loop transmitter is calibrated with a pneumatic calibrator which can be an Ametek or Crystal Engineering, per References 11, 21. The least accurate pneumatic calibrator tool is used as bounding in this calculation. The pneumatic calibrator tool is in turn calibrated by an Ametek type RK deadweight tester, Reference 21.

Page 32 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 4.1.3.6.2 Device Calibration Error The Calibration Error ( Ci ) for Device "i" is the SRSS combination of the As Left Tolerance (ALT), and the errors due to input and output calibration tools (including tool accuracy and readability and the error of the i=

2J l2+(C0Lj, J2+f ALT )2

~j CPIADjnP )

... j +l ) +(CEA 2+(CSTDhnp 2+ CTOOLOU, + ( ',

calibration standards). Thus, on a 2 sigma basis the calibration erro where Wnis the sigma value associated with ea c individu e 4.1.3.6.3 Device Calibration Error Values Since the values of ALT, CTO nd CSTD are controlled by 100% testing, they are assumed to e es a 3 sigma values. Vendor Accuracy is written as "Vendor c W.

1. APRM Channel Item Cal. Ins nt ".e cri tion Error C1 DVM Fluk 5Vdor Accur. 0.025% reading + 6 0digits Vdc Display Exp 10"3

-*

7) Temp. Comp. 0.1 x VA per °C/(T-28)°C Resolution 100 microVdc N/A (digital) o 10.000 Vdc = 125% Power on APRM meter, range = 10 Vdc, aximum calibration temperature 90 deg F = 32 deg C. Therefore, CTOOL1 = SQRT {[0.025% x 10 Vdc + 6 digits x 10.3 ]2 2

+ [0.1 x (0.025% x 10 Vdc + 6 digits x 103 ) x (32-28) oC]

+ (100 microVdc) 2}

CTOOL1 = 0.0092 Vdc

= (0.0092 Vdc / 10 Vdc) x 100% = 0.092 % FS

= 0.092% FS x (1.25 % Power/100% FS)

= 0. 115 %Power CREADI = 0 Vdc = 0.000% FS = 0.000% Power And since items C 2 and C 3 are identical to C1 and use the same range:

CTOOL1 = CTOOL 2 = CTOOL 3 = 0.115 % Power CREAD1 = CREAD 2 = CREAD 3 = 0.000 % Power

Page 33 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET The calibration standard error for each TOOL in the fixed neutron flux channel is conservatively assumed to be equal to the calibration tool error.

Therefore:

CSTDI = CSTD 2 = CSTD 3 = 0.115 % Power

2. APRM Flow Reference Channel Item Cal. Instrument Description C4A,B Pneu calib Vendor Acc.

(4 digit display) Display Exp.

Temp. Comp.

Re2 J i(Ref. 38)

CSTD 4ANB Ametek RK VA C5A B DVM Fluke 45 (racŽ

!5%reading + 6 digits 10-2

_TInp. Comp. 0.1 x VA per °C/(T-28) 'C

'Rysolution 1 microVdc Read N/A (digital)

(range = 100 mVdc)

VA 0.025% reading + 6 digits Disp Exp 10-2 Temp. Comp. 0.1 x VA per °C/(T-28) °C Resolution 1 microVdc Read N/A (digital)

DVM Fluke 45 (range = 10 Vdc)

C8 DVM Fluke 45 (range = 10 Vdc)

Calibration Error for C4A,B Pneumatic calibrator range = 0 - 830 in WC; therefore:

CTOOL4 = SQRT[(0.1% FS x 830 inWC + I x 10-1)2

+ (0.39% FS x 830 inWC)2 ]

= 3.37 in WC over span of 408.9 in WC

= (3.37 in WC/ 408.9 in WC) x 100% = 0.824% FS

= 0.824% FS x (1.25 % Power/100% FS)

= 1.03% Power Also, CREAD 4 = 0 or 0.000% FS

Page 34 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Calibration Error for CSTD 4A,B Unit = Ametek Type RK deadweight tester; Range = 830 in WC CSTD 4A,B = 0.05% x 830

= 0.415 in WC (over span of 408.9 in WC)

= (0.415 in WC / 408.9 in WC) x 100%

= 0.101 % FS x (1.25 % Power/100% F

= 0.126% Power Calibration Error for C5A,B Unit: DVM Fluke 45; Range: 100.00 ml Max Calib Temp = 40 deg C Therefore:

CTOOL5 - SQRT{ )b25%A, 2

+ [0.1 x (0. 120C]

+ (I microVXý

'5 % Power/100% FS)

Also, or 0.000% FS

'r0'o,'r CSTDSA,B

) ation standard error is equal to the tool error

ýTD 5 = CTOOL 5 = 0.353% Power Ibalfon Error for C6A,B nit: DVM Fluke 45; Range: 100.00 mV dc, Reading = 50.00 mVdc Max Calib Temp = 32 deg C Therefore:

CTOOL 6 = SQRT{(0.025% x 50 mVdc + 6 x 10"2)2

+ [0.1 x (0.025% x 50 mVdc + 6 x 10-2) x 4°C]2

+ (1 microVdc) 2}

= 0.078 mVdc

= (0.078 mVdc/40.0 mAdc) x 100%

= 0.195% FS x (1.25 % Power/100% FS)

= 0.244% Power Also, CREAD 6 = 0 or 0.000% FS

Page 35 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Calibration Error for C6,BSTD Assume calibration standard error is equal to the tool error CSTD6 = CTOOL 6 = 0.244% Power Calibration Error for C7A,B Unit: DVM Fluke 45; Range = 10.000 Vdc at 100% FS CTOOL 7 = 0.0092 Vdc = 0.092 % FS

= 0.092% FS x (1.25 0/ e6r/lrI S

= 0.115% Power CREAD 7 = 0 or 0.000% FS Calibration Error for CSTD7A, Assume calibration standard error i equa to the tool error CSTD7 = CTOOL 0 ra Calibration Error Unit: DVM Fluke =ig 10.000 Vdc at 100% FS CTP--),,OOfJ dc = 0.092% FS (k092% FS x (1.25 % Power/100% FS)

(i 15%,Power

\=R0 0 or 0.000% FS CAiralow/rror

ý,ire calibration for standard CSTD8 error is equal to the tool error t CSTD8 = CTOOL8 = 0. 115% Power

\X. Overall APRM Channel Calibration Errors The overall 2 sigma calibration errors for the various APRM functions is obtained by SRSS addition of the 3 sigma loop errors due to calibration tools, calibration standards and As Left Tolerance (ALT). Since all the calibration equipment is well maintained and tested, it is assumed that the Ci and CiSTD values given above are 3 sigma values. Also, since the instruments are always kept within ALT after calibration, the ALT values listed in the calibration procedures (and shown in 4.1.3.5) represent 3 sigma values. Thus the overall 2 sigma calibration errors for the various APRM functions is obtained from:

+Z(~

CL = jAT,22+I(CE__

Page 36 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET a) For APRM Flow Biased SCRAM Cfb-SCRAM = 2 x SQRT {(ALT1 /3) 2

+ (ALT 3/3)22 + (ALT 7/3)2 +

2 2 (ALTl,/3)2 2(ALT,/3) + 2(ALT9/3) + 2(ALT1 o/3) +

2

+ (CTOOL, /3)2 + (CTOOL 2 /3)2 + (CTOOL 3/3)

+ 2(CTOOL4 /3)2 + 2(CTOOL5 /3) 2 + 2(CTOOL

+ 2(CTOOL 7/3)2 + (CTOOL 8/3) 2 + (C +

(CSTD 2/3)2 + (CSTD 3/3)2 + 2(CS 4 2

2(CSTD5 A,B/3) + 2(CSTD 6A,B/3) 2 ++ 2 (CSTD 8/3) 2}

C fh-SCRAM = 2 x SQRT {(1.00/3)2 + (1+00 2 . +

2(0.01/3)2 + 2(0.047/3)2 + 94/3) 1151 +

(0.115/3)2 + (0.115/3)2 + 2 2 2(0.353/3)2 +

2(0.244/3)2 + 2(0. 1153)2 + (0 /3 + (0.115/3)2 +

(0.115/3)2 + 5- 2 (0.126/3)2 + 2(0.353/3)2 +

2(0.244/3)2 +2(0. 1 0.153) 2}

Cf-scRAm 2.06% Power b)FrAPRM Floi, k 2 x SQRT ,( 1 + (ALT/3) 2 + (ALT 7/3)2 + 2(ALT8 /3) 2 +

2(ALT9 /3)2 A o/3) 2 + (ALT 11/3)2 + (CTOOLI/3) 2 2 2

+ T ,2/3)2 + (CTOOL 3/3) + 2(CTOOL4 A,B/3)

(0 A,B/ 3 )' + 2(CTOOL 6AB/3) 2 + 2(CTOOL 7AB/3) 2 2

2 2 83)2 + (CSTD1/3) + (CSTD 2/3) + (CSTD 3/3)

+ TD4 A3,B/3) 2 + 2(CSTDSA,B/3) 2 + 2(CSTD 6AB/3) 2

+*2(CST7A,B/3)2 + (CSTD8/3) 2}

IB x SQRT {(1.00/3)2 + (1.00/3)2 + (2.0/3)2 + 2(0.469/3)2

+ 2(0.01/3)2 + 2(0.047/3)2 + (0.094/3)2 + (0.115/3)2 +

(0.115/3)2 + (0.115/3)2 + 2(1.03/3)2 + 2(0.353/3)2 + 2(0.244/3)2

+ 2(0.115/3)2 + (0.115/3)2 + (0.115/3)2 + (0.115/3)2

+ (0.115/3)2 + 2(0.126/3)2 + 2(0.353/3)2 + 2(0.244/3)2

+ 2(0.115/3)2 + (0. 115/3)2}

SCf-RB = 2.06 % Power c) For APRM Neutron Flux Fixed High SCRAM 2 2 2 CflxSCRAM = 2 x SQRT {(ALT 1 /3) + (ALT 2/3) + (ALT 7/3)

+ (CTOOLI/3) 2 + (CTOOL 2/3) 2 + (CTOOL 3/3) 2

+ (CSTDI/3) 2 + (CSTD 2/3) 2 + (CSTD 3/3) 2}

Page 37 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Cn-SCRAM = 2 x SQRT {(1.00/3)2 + (1.00/3)2 + (2.0/3)2 + (0.115/3)2

+ (0.115/3)2 + (0.115/3)2 + (0.115/3)2 + (0.115/3)2

+ (0.115/3)2}

Cfx-SCRAM = 1.65% Power ,

d) For APRM Neutron Flux Rod Block Clamp 2 2 Clamp-R= 2 x SQRT {(ALT 1/3) + (ALT 2B/3)

+ (CTOOL 2/3) 2 + (CTOOL 3/3)"

+ (CSTD 2/3)2 + (CSTD3/3) 2}

{10/)2 C.ap-RB =2 x SQRT {(1.00/3 + (1.00/3) +

+ (0.115/3)2 + (0.115/3) 15 Cvnp-P = 0.96% Power e) Nck 2

+ (CTOOL,/3)

+ (CSTDI/3) 2

3)STD3/3) 2) (0.115/3)2 3)2 + (1.0/3)2 + (0.115/3)2 +

+ (0.115/3)2 + (0.115/3)2 + (0.115/3)2}

I Neutron Flux Fixed High SCRAM - Setdown 2 2 2

= 2 x SQRT {(ALT1/3) + (ALT 5/3) + (CTOOL,/3) 2 2 2

+ (CTOOL 2/3) + (CTOOL 3/3) + (CSTDI/3) 2

+ (CSTD 2 /3) + (CSTD3/3) 2}

Cset.SCRAM = 2 x SQRT {(1.00/3)2 + (0.5/3)2 + (0.115/3)2 + (0.115/3)2

+ (0.115/3)2 + (0.115/3)2 + (0.115/3)2 + (0.115/3)2}

Cset.SCRAM = 0.77% Power g) For APRM Neutron Flux Fixed Rod Block - Setdown 2 2 2 Csetp-R = 2 x SQRT {(ALT1 /3) + (ALT6/3) + (CTOOL1 /3) 2 2 2 2

+ (CTOOL 2/3) + (CTOOL 3/3) + (CSTD,/3) + (CSTD 2/3)

+ (CSTD 3/3) 2}

Cset-RB 2 x SQRT {(1.00/3)2 + (0.5/3)2 + (0.115/3)2 + (0.115/3)2

+ (0.115/3)2 + (0.115/3)2 + (0.115/3)2 + (0.115/3)2}

Cset-.B 0.77 % Power

Page 38 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

4. RBM Channel Item Cal. Instrument Description Error C9 DVM Fluke 45 Vendor Accur. 0.025% reading + 6 digits Range =10 Vdc Display Exp 10"3 /ý-

(Ref. 37) Temp. Comp. 0.1 x VA per Resolution 100 microVN CRead For DVM 10.000 Vdc 125% Power on RB e rane maximum calibration temperature 90 deg FA C e CTOOL 9 = SQRT {[0.' X 10-3 ]2 2

+ [0.1 x (0.( )25%x 1i x 10.3 ) x (32-28) °C]

+ (100 micr oV(

CTOOL 9 = 0.00 2

= (0.00 2 100% = 0.092 % FS

= 0.092 F Power

= 1z 1h 01 0 "

.000% FS 0.000% Power jandard error for the is conservatively assumed to be equal tool error. Therefore:

0. 115 % Power 112 sigma calibration error including As Left Tolerance (ALT) is from:

C = 2 x SQRT{(ALTi/n) 2 + (CTOOLi/n) 2 + (CREADi/n) 2 + (CSTDi/n) 2}

This overall calibration errors for the various RBM functions, using the values of CTOOLi , CREADi, and CSTDi from above and the appropriate ALTi values from 4.1.3.5 subheading 2, are shown below:

a) For RBM Low Power Setpoint (LPSP) 2 Clpsp = 2 x SQRT {(ALT12/3)2 + (ALT 13/3) 2 + (CTOOL9 /3)

+ (CSTD 9/3) 2}

Clpsp = 2 x SQRT {(1.00/3)2 + (1.00/3)2 + (0.115/3)2 + (0.115/3)2}

Clpsp = 0.95 % Power

Page 39 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET b) For RBM Intermediate Power Setpoint (IPSP) 2 2 Cipsp =2 x SQRT {(ALT12/3)2 + (ALT14 /3) + (CTOOL 9/3)

+ (CSTD 9/3) 2}

Cipsp = 2 x SQRT {(1.00/3)2 +(1.00/3)2 + (0.115/3)2 + (0.115/3)2}

Cipsp = 0.95% Power c) For RBM High Power Setpoint (HPSP)

Chpsp = 2 x SQRT {(ALT 12/3) 2 + (ALT 15/3

+ (CSTD 9/3) 2}

Chpsp = 2 x SQRT {(1.00/3)2 + (1.00/3)2 +

Chpsp = 0.95% Power 4 d) For Cdtsp2 x SQRT Cdt& + (0.115/3)2 + (0.115/3)2}

C&ts e) For (LTSP) 2 2

+ (ALT1 7/3) + (CTOOL 9/3)

{(1.00/3)2 + (1.00/3)2 + (0.115/3)2 + (0.115/3)2}

Power F'or RBM Intermediate Trip Setpoint (ITSP) 2 2 2 Citsp = 2 x SQRT {(ALT 12/3) + (ALT 18/3) + (CTOOL 9/3)

+ (CSTD9/3) 2}

Citsp = 2 x SQRT {(1.00/3)2 + (1.00/3)2 + (0.115/3)2 + (0.115/3)2}

Citsp = 0.95% Power g) For RBM High Trip Setpoint (HTSP) 2 2 C htsp =2 x SQRT {(ALT 12/3) + (ALT19 /3) + (CTOOL 9

/3)2

+ (CSTD 9/3)2}

ChtSp = 2 x SQRT {(1.00/3)2 + (1.00/3)2 + (0.115/3)2 + (0.1 15/3)2}

Chtsp = 0.95% Power

Page 40 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 4.1.4 Determination of Loop/Channel Values For this calculation the loop contains several devices, thus the device error values for Accuracy, Drift and Calibration are the same as those for the loop. These values have been reported in Section 4.1.3.

4.1.5 Determination of PEA and PMA Primary Element Accuracy (PEA):

APRM Channel The PEA is a combination of the GE-LPRM sensor sensitin i an s sensitivity n- - rae "t uncertainties. The sensitivity of the detectors decreases with n utr influ .( e avrage sensitivity loss, and its 2 sigma variation, for all GE LPRM detect ben determined to be:

Sensor Sensitivity loss = 0.33 % (bias term)

+/-0.20% (random Reference 4, section 4.5)

The detector non-linearity and its 2 si ia pver ge) has been determined to be: 4 Sensor Non-linearity = 0.49% Oi te

+/- 1% 11 )

nrundom (Reference 4, section 4.5)

The first part of these detector o ýr esent bias type errors which apply to all detectors whereas the second part arfyrano Y that represent variability amongst the sensors.

Assuming a worst se 1 h e e APRM has the minimum number of operational detectors, the PEA whi(. ýAt e.*. of power basis, is simply obtained by adding the bias terms and taking t andom terms, is calculated below. In the calculation, the random error is reed by iyosquare root of the minimum number of operable LPRMS to one APRM ch I h ume 11 per Reference 35.

1 number of LPRMS per APRM = 11 Cfore, P A4PRM =(0.33+0.49)+ V(0.20)2 +(1.00)2 PEAAPRM = 0.82 +/- 0.31% power The first part of the PEA (0.82%) is treated as a drift term (DPEA) and the second part (+

0.31%) as an accuracy term (APEA).

The PEA value for the Westinghouse LPRM sensors installed at the Cooper site is given as 0.7 +/- 1% per Reference 52. In the present calculations the GE LPRM PEA error values will be used as they are more conservative.

Page 41 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET RBM Channel PEA is similar to that for the APRMS and equals 0.33% (bias term) +/- 0.20% (random term) and 0.49% (bias term) +/- 1% (random term), respectively In the calculation, the random error is reduced by the square root of the operable LPRMS to one RBM channel which are 2 per Reference 35,,

Minimum number of LPRMS per RBM = 2

Therfore, PEARBM =(0.33+0.49)+/- V(0.20)2 +(1.00)2 PEARBM = 0.82 + 0.72% power The first part of the PEA (0.82%) is trel and the second part (+/-

0.72%) as an accuracy term (APEA).

The value PEA value for the Westin-bo se LPRM sensors installed at the Cooper site is given as 0.7 +/- 1% per Reference n',ne e~he GE value is larger than the Westinghouse LPRM uncertainty value, thc GE Vikh iil csed in the calculations.

Flow Unit The PEA for the flWb,,o a enturis is included in the flow channel uncertainty, therefore no additional ui e =0 irement Accuracy (PMA):

\TheIPMA is a combination of the APRM tracking and the uncertainty due to neutron noise.

-(,'sidering the APRM neutron flux, for the MSIV closure transient event, the APRM 4racking error is 1.11% and the uncertainty due to neutron noise is typically 2.0%, Reference

4. Flow noise is estimated to be 1.0% rated flow (0.75% power) per References 52 and 54.

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 frequency of approximately 0.3 to 0.5 Hertz and a typical maximum peak-to-peak amplitude of approximately 5 to 10 percent.

For neutron flux PMA = 2 x SQRT [(2.0/2)2 + (1. 11/2)2] = 2.29% power (fixed)

For flow biased, PMA = 2 x SQRT[(1 .11/2)2+ (2/2)2+ (0.75/2)2] =2.4 1%power (flow-biased)

Page 42 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET RBM The PMA of the RBM is a combination of the RBM tracking error and the uncertainty due to neutron noise. The uncertainty due to neutron noise is estimated to be the same as the RM or 2.0% (2-sigma). The error calculated by comparing the reading with all LPRMS e e to readings with different combinations of LPRM failure is estimated to be wi n o,. -

sigma) per Reference 52. A 3-sigma confidence level is used because the 10% u(o .ba on testing. "

PMA = 2 x SQRT [(2.0/2)2 + (1.0/3)2] = 2.11% power (RBM P er PMA = 2 x SQRT [(2.0/2)2 + (1.0/3)2] = 2.11% power (RBM p 4.1.6 Determination of Other Error Terms All error terms to be considered have been accounted fo the ious sections.

4.1.7 Calculation of Se oint Margin and 0 *atin ,Se o 4.1.7.1 Setpoint Margin The setpoint margin is defined as between the nominal setpoint and the analytic limit. Based on R ces 5, 7, this margin is given by:

SM = (1.645/N)(SRS ,,M TERMS) + BIAS TERMS Where N rQ~sene, of standard deviations with which all the random terms areýchara e ( ally 2 standard deviations) and 1.645 adjusts the results to a 95% rc1Ab\ ided normal).

The err~or-tert e c, culated for trip conditions, and the margin becomes A;LT) +(CLY) +(DJ) +(PM'4) +(PEA)2 + bias PRM CHANNEL Flow Biased SCRAM

-+cl + (PMAjb + (A + DPEA 6 +(2.06)2 +(1.97)2 +(2.4 1)2 +(0.3 + 0.82 sc S(2.76)2 V~. 1)2 SMfb-scRm = 4.65 % Power

Page 43 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

2. Flow Biased Rod Block sm. ((1.645 )j(A4LT- )2 + C, + (D, )2+ (PMA,,)2+ (APEA)2 + DPEA

= ~jbR 1.4 )V(2.76)2 + (2. 06)2 + (i.97)2 + (2.4 1)2 + (0.3 1)2 + 0.82 SMfb.RB= 4.65 % Power

3. Neutron Flux - Fixed High SCRAM Si

= . AT-f.)A +

fix-SCRAM +

+ýxSR (*xPM+ Dr P DEPA

-. 6 ()2 2 j) +0.82 SMfix.SCRX = 3.79% Po e

4. Neutron Flux Rod koick lamp SM =ý-R + (,x+ (PMA) + (APEA)' + DPEA SMclam 1. l 63)2 +(0.96)2 +(1.51)2 +(2.29)2 +(0.31)2 +0.82 3.58 %Power eutron Flux Downscale Rod Block

( 1 .645 2 + (C d . _ ) 2 +

+(ApEA) 2 2

S-SMdowv-RB 3.5(A8 2 P(opMA) + +DPEA SMdown-pB (1.645 )(1.63) 2 +(0.96)2 +(1.5 1)2 +(2.29)2 +(0.3 1)2 I+0.82 SMdm,0 R 3.5 8 %Power

Page 44 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

6. Neutron Flux Fixed High Scram - Setdown 2

SM4 = ( 2 (ALT ,) + _.)I + (PM4)

+ +(APEA)'

- + DPEA SMsetSCRAM= @ii1.6 45 )V(1.63)2 + (0.96Y +(1. .51)2 +(2.29)+(0.312 +0.82 SM,,,.scRAm = 3.54 %Power

7. Neutron Flux Fixed High Rod Block -

+ 0.82 b) RBM CHANNEL

1. Low Power Setp SM 1psp='

Intermediate Power Setpoint (IPSP) 2 2 SMipsp =(1.'1(ALT-PBM-pwr, +(Cipp Y +(DRBM-pwr)2 + (PMA) +(APEA) J+DPEA SMipsp =(145 1.222 +(0.95)2 +%(.99)2+P(2.1)2 +(0.72)2 +0.82 S~pp= 3.20 % Power

Page 45 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

3. High Power Setpoint (HPSP)

SMhpsp = (*-kV*(ALTRBMpwr )2+(Chpp)2 +(DRBMpr 2 ) +(PMA)2 +(APEA) 2 )+DPEA SMhPSP =(1.645- (1.22)2 +(0.95)2 +(0.99)2 +(2.11)2 +(0.72)2 )+0.82 SMhpsp = 3.20% Power

4. Downscale Trip Setpoint (DTSP) * ,

SMtp =(1.645' UP

2 ,

6.

SMitsp = 3.87 %Power

7. High Trip Setpoint (HTSP) 2 2 SMhlp = P.6- V(LT-RBM-Iip Y + (Chup )2 + (DBM trip)2 + (PMA) + (APEA) + DPEA SMhtsp 1 5V(1.91)' +(0.95)2 + (2.04)2 + (2.11)2 + (0.72)2 +p0.82 SMhtsp = 3.87 % Power

Page 46 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 4.1.7.2 Nominal Trip Setpoint (NTSP1) Calculation The Nominal Trip Setpoint (NTSP1) for process variables which increase to trip is given by:

NTSP1 = AL - SM NTSP1 represents the upper limit (closest to AV) at which the setp n es assuming zero H leave alone tolerance in the direction tow rd the oý a ePl *a)

(AV).

a) APRM CHANNEL

1. Flow Biased Scram Flow biased setpoints will be shown in t o etercept, since the slope (0.66 W) is a constant NTSP1I -SCRAM AL For this function ALfb-scR* 0.75% + .6o qPower (Reference 58)

Therefore:

NTSPlfb-scR - 4."ý465% = 60.95% Power 2.

- SMfo-RB

= 0.75W + 54.8% Power (Reference 58) lfb-,B = 54.8% - 4.65% = 50.15% Power

3. Neutron Flux - Fixed High SCRAM NTSP Ifix-scRAM = AL - SMfix-ScR For this function ALfx-scRAM = 123.0% Power (Reference 2, 58)

Therefore:

NTSPIfixSCRAM = 123.0% - 3.79% = 119.21% Power

Page 47 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

4. Neutron Flux Rod Block Clamp NTSPl1,1.p RB - AL - SMct.p RB For this function ALclanp RB = 112.2% Power (Reference 5, Therefore:

NTSPlcmppRB= 112.2%-3.58% = 108.62%Power

5. Neutron Flux Downscale Rod Block NTSP1 down RB = AL + SMdown RB For this function AL down-RB = 0.0% Power (Reference 2)

Therefore:

NTSPldowRB= 0.0%+ .

6. Neutron Flux Fixed Hi C NTSP1set-scR*M =AL- S B For this function ALset.SCPAM .4-, lwer (Reference 2, 58)

There*le

\ I 17.4%-3.54% 13.86 % Power

7. _ t lux Fixed High Rod Block - Setdown V ALsetR= 14.4% Power (Reference 2, 58)

Therefore:

NTSPlsetRB = 14.4% - 3.54% = 10.86% Power b) RBM CHANNEL

1. Low Power Setpoint (LPSP)

NTSP lpsp = AL- SMpsp For this function ALlpsp = 30.0% Power (Reference 2, 58)

Page 48 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Therefore:

NTSPl1 p p = 30.0% - 3.20% = 26.80 % Power

2. Intermediate Power Setpoint (IPSP)

NTSP lipsp = AL - SMipsp For this function ALipsp = 65.0% Power RA Therefore:

NTSPlipsp = 65.0% - 3.20% = 61.80 % Power

3. High Power Setpoint (HPSP)

NTSPl hpsp = AL- SMhpsp For this function ALhpsp = 85.0% Po r (Reference 2, 58)

Therefore:

NTSPlhp~p = 85.0%/' 0% 8 .80% Power

4. Down s-c'le ri"ýýVp t & (DTSP)

S -89.0% Power (Reference 2)

SPldtp = 89.0% + 3.87% = 92.87% Power Low Trip¶ Setpoint (LTSP)

NTSP1I tsp = AL - SMltsp For this function ALtsp = 117.0% Power (Reference 2, 58)

Therefore:

NTSPl1 ts = 117.0%-3.87% 113.13% Power

¶ Note: For the RBM trip setpoints, a MCPR of 1.20 is used, margins are the same for other MCPRs and are summarized in the conclusion (Section 5).

Page 49 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

6. Intermediate Trip¶ Setpoint (ITSP)

NTSP Iitsp = AL- SMitsp For this function ALitsp = 111.2% Power Therefore:

NTSPlitsp = 111.2% - 3.87% = 107.33 % Power

7. High Trip¶ Setpoint (HTSP)

NTSP IhtSp = AL - SMhlSp For this function ALhtsp= 107.4% Power (Reference 2, 58)

Therefore:

NTSPlhtsp= 107.4% -31 4.1.7.3 rocess variable increases to trip, so the Allowable following equation (Reference 4):

eproresents the number of standard deviations with which all the random

-eharacterized (normally 2 standard deviations) and 1.645 adjusts the results probability (one-sided normal).

random errors include the random portion of ALT, CL, PMA, PEA, but exclude (i.e., the DPEA term is not included as a bias term since it is a drift term). Thus, AV = AL - (1.645/N) x SQRT( ALT2 + CL2 + PMA2 + PEA2 ) - (Y, BIAS TERMS)

¶ Note: For the RBM trip setpoints, a MCPR of 1.20 is used, margins are the same for other MCPRs and are summarized in the conclusion (Section 5).

Page 50 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET a) APRM CHANNEL

1. Flow Biased Scram Allowable values for flow biased setpoints will be shown in terms of the intercept, since the slope (0.75 W) is a constant.

2 AVfb-scRAM = AL - (1.645/2) x SQRT (ALT.fh 2

+ Cfm.SCRAM 2 + PMA +

For this function the AL is ALfb-scRAM = 0.75W + 65.6%

Therefore:

A VjbtSCP = 65.6~j. f (2 .

2 AVe-scRAM = 62.133 % Power Rounded down conservatively toene AVe-sCRAm-- 62.0% Po

2. Flow Biased Rod Block Allowable valuej-! iased: setpoints will be shown in terms of the is a constant.

2 PMA 2 + PEA 2) 2 SQRT (ALT.f + Cfb.Ra +

+ 54.8%

--54.8- ((-1.645)(2.76)2 +(2.06)2 +(2.4 1)2 +(0.31)2 N

AVfo.RB = 51.333% Power Rounded down conservatively to nearest readable increment:

AVfb-RB = 51.0 % Power

3. Neutron Flux - Fixed High SCRAM AVfjý-scRAM = AL - (1.645/2) x SQRT (ALT-fix 2

+ Cfix.SCRAM2 + PMA 2 + PEA2 )

For this function the AL is:

ALfx-SCRAM = 123.0%

Page 51 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Therefore, A Vf.SCRAM = 123.0 -( 1.645)(i .63) 2 + (i.65)2 + (2.29)2 + (0.3 1)2 AVfx-SCRAM = 120.307 % Power Rounded down conservatively to nearest readable increment:

AVfix-SCRcM = 120.0 % Power

4. Neutron Flux Rod Block Clamp 2

AVclamp RB = AL - (1.645/2) x SQRT (ALT.fix + +

For this function the AL is:

ALclap RB = 112.2%

Therefore, AVcjamp_RB = 112.2 - I 6 AVdownRB= 109.743% e° Rounded down conser va~ o,: _tearest readable inLcrement:

AV,, r= . 0o r Wnscale Rod Block

+(1.645/2) x SQRT 2 (ALT-fix + Cdown.RB2 + PMA 2 + PEA2)

A d,,,_B= 0.0 + (1.645 )V~(1.63)2 +(0.96)2 + (2.29)2 + (0.3 1)2 ~

AVdown RB = 2.46 % Power Rounded up conservatively for ITS implementation consideration:

AVdo, RB = 3.0 % Power

Page 52 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

6. Neutron Flux Fixed High SCRAM - Setdown AVsetscRpM = AL - (1.645/2) x SQRT (ALT_fix 2

+ Cset_SCRAM2 + PMA 2 + PEA 2)

For this function the AL is ALst-sc~mv = 17.4 % A, Therefore, AVsejt-SCRAM = 17.4--(I .2-)V(1.63)2 +(0.77)2 +(2.29)2 + A AVset-SCRAM = 14.99 % Power Rounded down conservatively to nearest readable ii AVset-scRAM = 14.5% Power

7. Neutron Flux Fixed High Rod Block eI AVset.RB = AL -(645/2) T PEA 2)

For this function the AL is ALset =- 14.4 %

Therefore, AVsetRB + (0.77. +(2.29)2 + (0.31)2 AVseR B u wer readable increment:

71 6w Power Setpoint (LPSP)

+ Cipsp 2 + PMA2 +

2 AVjpsp = AL - (1.645/2) x SQRT (ALT-MPr PEA 2 )

For this function the AL is ALRB-,psp = 30.0 %

Therefore, AVpp = 30.0-- 1.645)V(1.22)2 +(0.95)2 +(2.11)2 +(0.72)2)

AV1 psp = 27.76 % Power Rounded down conservatively to nearest readable increment:

AVlpsp = 27.5% Power

Page 53 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

2. Intermediate Power Setpoint (IPSP)

AVipsp = AL - (1.645/2) x SQRT (ALT-MpWT 2

+ + PMA 2 + PEA 2)

For this function the AL is ALRB.ipsp = 65.0 %

Therefore, AVpp =65.0 - 1645 DV(1.22)2 + (0.95)2 + (2.11)2 +

K2 AVipsp = 62.76 % Power Rounded down conservatively to nearest readable i AVipsp = 62.5% Power

3. High Power Setpoint (HPSP)

AVhpSp = AL - (1.645/2) T- T PMA2 + PEA 2)

For this function the AL is ALRB-hpsp = 85.0 %

Therefore, AVh,,*, '1\.2r +(0.95) +(2.11)

AVhnsn.&8 7/o+% \er conservatively to nearest readable increment:

/4. Downscale Trip Setpoint (DTSP)

AVdtsp = AL - (1.645/2) x SQRT (ALT-RBM-p 2 2

+ Cdtsp + pMA 2 + PEA 2 )

For this function the AL is ALdtsp = 89.0 %

Therefore, A Vd,Sp = 89.0 + (1.645 / 2)C 1.645) V(l .91) 2 + (0.95)2 + (2.11)2 + (0.72)2 AVdtsp= 91.53%

Page 54 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Rounded up conservatively to nearest readable increment:

AVdtsp = 92.0 % Power

5. Low Trip¶ Setpoint (LTSP)

AVltsp = AL - (1.645/2) x SQRT (ALT.RBMftrip 2 + Cltsp2 + PMA2 +

For this function the AL is ALItsp=117%%

Therefore, AVsp=l 17.0+(1.645/2)I(I.2 .)5(1.91)2 AVltsp = 114.46 % Power Rounded down conservatively to nearest AVs~p = 114.0 % Power .

6. Intermediate Trip¶ Setpoin )

+ Citsp2 + PMA 2 + PEA2 )

2 5 )V(1.9 ly + (0.95Y + (2.1ly + (0.72Y2 IV108.66 %Power ided down conservatively to nearest readable increment:

AVitsp = 108.5 % Power

¶ Note: For the RBM trip setpoints, a MCPR of 1.20 is used, margins are the same for other MCPRs and are summarized in the conclusion (Section 5).

Page 55 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

7. High Trip¶ Setpoint (HTSP)

AVhtsp = AL - (1.645/2) x SQRT (ALTRBM-trip 2

+ Chtsp 2

+ PMA2 + PEA 2)

For this function the AL is ALhltp = 107.4 %

Therefore, A Vhsp = 107.4- 1.645)V(1.91)2 +(0.95)2 +(2.11)2 +(0.72)2 AVhtsp = 104.86 % Power Rounded down conservatively to nearest readable increm [W AVhtsp = 104.5 % Power 4.1.7.4 LER Avoidance Evaluation The purpose of the LER Av to assure that there is sufficient margin provided between th oand the Nominal Trip Setpoint to avoid violations of the Tech S c Jo .e Value (which, when discovered during surveillance, could lead to L iJ ). The method of avoiding violations of the Allowable Value i'- tdetermine the errors that may be present during surveillance testing, ema e argin between the calculated values of NTSPI and AV, and t adj N; St t o ovide added margin if necessary. The following equation isted t e he errors that would be expected to contribute to a potential <t .i

.iq(1 L) =(1iN)(SRSS Of RANDOM TERMS)

Where t* ie number of standard deviations with which the random terms are chllrt2t n4o(rmally 2 standard deviations).

4. dom Terms Included In LER Avoidance i he Random Terms that should be included in the LER Avoidance evaluations include:

Loop Accuracy under Normal plant Condition (ALN)

Loop Calibration Error (CL)

Loop Drift (DO)

Process and Primary Element Errors are not included because calibration and surveillance testing are performed using input signals which simulate the process and primary element input.

2 2 2 Sigma(LER) = (1/2)SQRT (ALN + CL + DL )

¶ Note: For the RBM trip setpoints, a MCPR of 1.20 is used, margins are the same for other MCPRs and are summarized in the conclusion (Section 5).

Page 56 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 4.1.7.4.2 LER Margin Calculation Once the value of Sigma(LER) is determined, the margin between the values of NTSP1 and AV is calculated in terms of Sigma(LER) using the equation below:

Z(LER) = JAV-NTSP II / Sigma(LER)

This value of Z is then used to determine the probability Allowable Value by treating the error distribution as a Distribution, and then determining the area unde_.the cyr.

Distribution corresponding to the numbe s ATc represented by Z.

4.1.7.4.3 GE Recommendation GE recommends that a nominal B I tbW 90% for avoiding an LER condition be used as the accept e ri'e ion r the LER Avoidance (or Tech Spec Action Avoidance) a oa rih ngle instrument channel, the value of Z(LER) esl 1 1-(\ -tl 0% xtenon is 1.29 or greater.

For an instrument c [nei of a multiple channel logic system a value of Z(LER) Ve ter can assure 90% Tech Spec Action Avoidance criterion.

GvrigSeoht- mination 4.1.7.4.4 a( - (1/2) x SQRT (ALN.h 2

+ CfhscR2 + Dm2)

) (1/2)xSQRT(2.76 2 +2.062 + 1.972) 1.99

=AV-NTSPl.scRAml / Sigma(LER)

= 162.0% - 60.95%1/ 1.99

=0.53 Since this value of Z does not correspond to a probability of more than 90% (one sided normal distribution) for a multiple channel (0.81), the NTSP is adjusted as follows:

NTSP2f.scRAm = AV - 0.81 x Sigma (LER)

= 62.0% - 0.81 x 1.99 NTSPfb-scRAM = NTSP2fb-scRAm = 60.3 8% Power

Page 57 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

2. Flow Biased Rod Block 2 2 Sigma(LER) = (1/2) x SQRT (ALN-fo + CfoR + D o )

2 Sigma(LER) = (1/2) x SQRT (2.762 + 2.062 + 1.972)

= 1.99 Z(LER) = IAV-NTSPlfoRI / Sigma(LER)

= 151.0% - 50.15%1/ 1.99

= 0.43 Since this value of Z does not correspond to 90% (one sided normal distribution) for a mt NTSP is adjusted as follows:

NTSP2f.pR = AV - 0.81 x Sigma (LER)

= 51.0%- 0.81 x 1.99 NTSPtb.RB = NTSP2fa.bR = 49.38

3. Neutron Flux -

Sigma(LER) =(1 /2 D -2 )

Sigma(LER) = (1/2) 1.652 + 1.512)

= 1.39

.211/ 1.39

,fZ does not correspond to a probability of more than normal distribution) for a multiple channel (0.81), the

-d as follows:

)2fix-scRAM = AV - 0.81 x Sigma (LER)

= 120.0%- 0.81 x 1.39 fix-SCRAM = NTSP2Sx-SCRAM = 118.87 % Power

4. Neutron Flux Rod Block Clamp Sigma(LER) = (1/2) x SQRT (ALNfix 2 + ClapR 2 + D'x 2)

Sigma(LER) = (1/2) x SQRT (1.632 + 0.962 + 1.512)

= 1.21 Z(LER) = IAV-NTSPClamp.RBI / Sigma(LER)

= 1109.5% - 108.621/ 1.21

= 0.73 Since this value of Z does not correspond to a probability of more than 90% (one sided normal distribution) for a multiple channel (0.81), the NTSP is adjusted as follows:

Page 58 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET NTSP2, 1.pp.B = AV - 0.81 x Sigma (LER)

= 109.5%- 0.81 x 1.21 NTSPCaImp-RB = NTSP2, 1ap-.R = 108.52 % Power

5. Neutron Flux Downscale Rod Block 2 2 Sigma(LER) = (1/2) x SQRT (ALNfix + Cdown.RB + Dnx 2

Sigma(LER)= (1/2)xSQRT(1.63 +0.962 + 1.512)

= 1.21 Z(LER) = lAV-NTSPIdoI / Sigma(LER)

= 13.0% - 3.581/ 1.21

= 0.48 Since this value of Z does not co e on to,,robabilty of more than 90% (one sided normal distributio a u le channel (0.81), the NTSP is adjusted as follows:

NTSP2do, RBAV 1 pv +, W.x a

=3.0% + 0.8 1 NTSPown pB = NTSP i' RB 3.9 /oPower

6. Neutron Fluxfxed i RAM - Setdown Sigma(LER QRT (ALN.fix 2 + Ct-CR +D Sigma(LT_ N QRT(1.63 2 +0.772+ 1.512) fLE TSPlset-SCRAMI / Sigma(LER)

- \0.54 4.5%-13.86%J/1.18 e this value of Z does not correspond to a probability of more than 90% (one sided normal distribution) for a multiple channel (0.81), the NTSP is adjusted as follows:

NTSP2set-SCRAM = AV - 0.81 x Sigma (LER)

=14.5% -0.81 x 1.18 NTSPset-scR 5 Am = NTSP2set-sCRAM = 13.54 % Power

7. Neutron Flux Fixed High Rod Block - Setdown 2 2 Sigma(LER) = (1/2) x SQRT (ALN.fix + Cset.B + Dfx )

2 Sigma(LER) = (1/2) x SQRT (1.632 + 0.772 + 1.5 12)

= 1.18 Z(LER) = jAV-NTSPlset.-Rl / Sigma(LER)

= 111.5% - 10.86%1/ 1.18

= 0.54

Page 59 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Since this value of Z does not correspond to a probability of more than 90% (one sided normal distribution) for a multiple channel (0.81), the NTSP is adjusted as follows:

NTSP2set,-R = AV - 0.81 x Sigma (LER)

= 11.5%-0.81 x 1.18 NTSPst.R = NTSP2seti = 10.54 % Power b) RBM CHANNEL r

1. Low Power Setpoint (LPSP) +,

Sigma(LER) = (1/2) x SQRT (ALN.RBM Pwr2 + +

Sigma(LER) = (1/2) x SQRT (1.222 + 0.952 0.2)Y

= 0.92 Z(LER) = JAV-NTSP1l psp 3 / Sigma(LR

= 127.5% - 26.80%° / 0.92 Since this value of Z des no core-spond to a probability of more than 90% (one-side norm i I I t f a single channel (1.29), the NTSP is adjusted as follows:

NTSP2s =AV -_1. x Sigma (LER)

= 27.5% -ý 1.29 0.92 NTSPpJ ,TS 26.31 %Power

3. ,High ower Setpoint ( IPSP)

Sia = (/2)

-ER x SQRT (ALN.RBMpwr 2

+PSP + DRBM-pw 2

)

4L -- (1/2) x SQRT (1.222 + 0.952 + 0.992)

- 0.92 AlE) = IAV-NTSP1 ipspl /Sigma(LER)

= ~162.5% - 61.80%1 / 0.92

=~0.76

Since this value of Z does not correspond to a probability of more than 90% (one-side normal distribution) for a single channel (1.29), the NTSP is adjusted as follows:

~NTSP2ipsp = AV - 1.29 x Sigma (LER)

= 62.5% - 1.29 x 0.92 NTSPipsp = NTSP2ipsp = 61.31 % Power

3. High Power Setpoint (HPSP)

,Sigma(LER) = (1/2) x SQRT (ALN.RBM_pwr,,2 + Chpsp2 "+ ORBM-pwr 2)

Sigma(LER) = (1/2) x SQRT (1.22 2 + 0.95 2 + 0.992 )

= 0.92

Page 60 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Z(LER) = JAV-NTSP lhpspl / Sigma(LER)

= 182.5% - 81.80%1 / 0.92

= 0.76 Since this value of Z does not correspond to a probability of more th 90% (one-side normal distribution) for a single channel (1.29), the T is adjusted as follows:

NTSP2hpsp = AV - 1.29 x Sigma (LER) 82.5%- 1.29x0.92 NTSPhpsp = NTSP 2 hpsp = 81.31 % Power

4. Downscale Trip Setpoint (DTSP)

Sigma(LER) = (1/2) x SQRT (ALN.R"BM-tp 2 C 2 _+ip2)

Sigma(LER) = (1/2) x SQRT (1 + 045 + -2.04)

= 1.48 Z(LER) IAV-NTSPldtSpl / SigW-

= 0.59 Since this value of Z o ot cyespond to a probability of more than 90% (one-side normal rsritbj n) for a single channel (1.29), the NTSP is adjusted as foil s NTSP2dts Sigma(LER)

\+ 'X 1.48 2

dtsp = 93.90% Power

'tti etpoint (LTSP)

,* ') = (1/2) x SQRT (ALN.RM.tip 2

+ Citsp2 2

+ DRBM.trip )

S(LER)= (1/2) xSQRT(1.91 2 +0.952 +2.042)

= 1.48 Z(LER) = IAV-NTSP Ilpl / Sigma(LER)

= 1114.0% - 113.131/ 1.48

= 0.59 Since this value of Z does not correspond to a probability of more than 90% (one-side normal distribution) for a single channel (1.29), the NTSP is adjusted as follows:

NTSP21tsp = AV - 1.29 x Sigma (LER)

= 114.0%- 1.29 x 1.48 NTSPItsp = NTSP21 112.09 1Pp % Power

Page 61 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

6. Intermediate Trip Setpoint (ITSP)

Sigma(LER) = (1/2) x SQRT (ALNBM.ip2 + Citsp DRMp 2 )

Sigma(LER) = (1/2) x SQRT (1.912 + 0.952 + 2.042)

= 1.48 Z(LER) = IAV-NTSP litspl / Sigma(LER)

= 1108.5% - 107.33%1/1.48

= 0.79 Since this value of Z does not correspond to a 90% (one-side normal distribution) for a singli adjusted as follows:

NTSP2itsp = AV - 1.29 x Sigma (LER) 108.5% - 1.29 x 1.48 NTSPitsp = NTSP2itsp = 106.59 %

7. High Trip Setpoint Sigma(LER) - (1/2 Ch + DRBM.-tip 2 )

Sigma(LER) = (1/2i.* + 2.04 2)

= 1.48 Z(LER) = JAV-NTSP I

=11 does not correspond to a probability of more than al distribution) for a single channel (1.29), the NTSP is

= AV - 1.29 x Sigma (LER) 104.5%- 1.29 x 1.48

= NTSP2htsp = 102.59 % Power tlno Pptrnlnte recommended that the method of using NTSP as the center of the Leave Alone e be used. Thus, according to Reference 4, the nominal setpoint is:

NTSP = NTSP2 +/- LAT Where the LAT is the SRSS combination of the leave alone tolerances for all the devices in the loop.

Page 62 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 4.1.7.6 Establishing Leave Alone Zones The LAT for APRM Flow Biased SCRAM, APRM Flow Biased Rod Block and APRM High Neutron Flux SCRAM is + 2.00% Power based on ALT 7 (AGAF) since it is the largest ALT associated with those functions. The LAT for the remaining APRM and RBM functions within this calculation is +/- 1.00% Power since tha the largest LAT associated with those functions.

4.1.7.7 Required Limits Evaluation The Required Limits Evaluation calculates an adjt NTSP is set at the center of the leave alone zone.

the stack-up of the errors (including leave alone tC loop, there is enough margin for Technical Specif (or LER avoidance).

a) APRM CHANNEL

1. Flow Biased Scram The Required Limit (RL) of d ic 'i" t t la' RLi = NTSPf-SCRAM + L

= 60.38% + 2.00% = 38 This is compared against jn-SCRAM - 32.0%

fron RLi > AV, therefore NaTSI i.,ce adjusted:

fb.scRAM- LAT = 60.0 %

To adjustment is needed, the Required Limits of all devices in the and Sigma(LER, RL) given by the following equation (Ref.

(ADJ)fm-scRAM + LATi 60% + 2.0% = 62.00%

)2

+ (CftioSCRAM)2 + (DFB

ý (I RL i - NTSP(ADJ) fb-SCRAM 07LER,RL 2

For this calculation the loop error values are used which is equivalent to using one device with the error of the whole loop. Thus:

+(2.06)2 +(1.97)2 (1(2(62.0-60.0)2 CLER,RL 2

GLER, RL = 1.58 %

Page 63 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Also compute Z(LER, RL) given by:

Z(LER, RL) = ABS(AV Th-SCRAM - NTSP(ADJ) fb-SCRAM) / GLER, RL For this case of multiple channel, If Z(LER, RL) > 0.81 then LER avoidance condition is met.

Z(LER, RL) = ABS(62.0 - 60.0) / 1.58

= 1.26 This value of Z(LER, RL) is greater than the Z criterin0r rl 0.81. Therefore the criterion is met without further adj s ent. .

NTSP (ADJ) th-SCRAM = 60.0 % Power (Rounded c ervati'

2. Flow Biased Rod Block The Required Limit (RI) of device "i' the loop is:

RLi = NTSPfo.R + LATi

=49.38% + 2.00% 8 '

This is compared against AVf. 1 m Section 4.1.7.3. Since R~i > AV, therefore NTSP fb.R netý b djusted:

NTSP (DJ) - LA, = 49.0 %

To determ*e h1 ulekir tment is needed, the Required Limits of all devices in the

5) isloop calculRý' I ,* td Sigma(LER, RL) given by the following equation (Ref.

S\

-[S, DJ)fbB + LATi

-0% + 2.0% = 5 1.00%

( - NTSP(ADJ) fb-RB + (CffB_ )2 + (DFB )I

'7LERRL = 2 For this calculation the loop error values are used which is equivalent to using one device with the error of the whole loop. Thus:

I(j((2)(51.0-49.0)I2 +(2.06)2 +(1.97)2 CLERRL 2

(FLER, RL = 1.58 %

Page 64 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Also compute Z(LER, RL) given by:

Z(LER, RL) = ABS(AV fb-R - NTSP(ADJ) fb-RB) / GLER, RL For this case of multiple channel, If Z(LER, RL) > 0.81 then LER avoidance condition is met.

Z(LER, RL) =ABS(51.0- 49.0) / 1.58

= 1.26 This value of Z(LER, RL) is greater than the Z 0.81. Therefore the criterion is met without further NTSP (ADJ) fb-RB = 49.0 % Power (Rounded re

3. Neutron Flux Fixed High SCRAM The Required Limit (RL) of device "i" wit ear the loop is:

RLi = NTSPfixscRAM- + L

= 118.87% + 2.00% 08 This is compared against AV 6 xs - 120%froi mSection 4.1.7.3. Since RLi > AVfix-scRAM, therefore NTSP needs ted:

NTSP DJ)fi A -scRAM - LAT= 12' 0.0 -2.00 = 118.0%

To deterhi~ne*urt a ustment is needed, the Required Limits of all devices in Sigma(LER, RL) given by the following equation (Ref.

ýI1j(1(1i(RLi - NTSP(ADJ) fiX SCRAM 2J+ (CfixSCRAM )2 +(Df. )

2 For this calculation the loop error values are used which is equivalent to using one device with the error of the whole loop. Thus:

O'LER,RL =

7(j((2)(I20.0-118.0))2 +(1.65)2 +(1.51)2 2

GLER, RL = 1.30 %

Page 65 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Also compute Z(LER, RL) given by:

Z(LER, RL) = ABS(AV 6x-sCRAM- NTSP(ADJ)fx.SCRAM) / Sigma(LER,RL)

For this case of multiple channel, If Z(LER, RL) > 0.81 then LER avoidance condition is met. A Z(LER, RL) = ABS(120.0 - 118.0) / 1.30

= 1.54 This value of Z(LER, RL) is greater than the Z crite 0.81. Therefore the criterion is met without further adjLf NTSP (ADJ)fix-scRAm = 118.0 % Power (Rounded

4. Neutron Flux Rod Block Clamp The Required Limit (RL) of device "i" with RLi = NTSPcimp-RB + LATi

= 108.52% + 1.00%

This is compared against AV, Section 4.1.7.3. Since RLi > AVclmp-RB, therefore NTSP needs to b NTSP (ADJ) clamp- - LAT= 109.5 - 1.00 = 108.5 %

To is needed, the Required Limits of all devices in (LER, RL) given by the following equation (Ref.

2 For this calculation the loop error values are used which is equivalent to using one device with the error of the whole loop. Thus:

8 5 )2 +(0.96)2 +(1.51)2 (109.5-10 .

- LERRL 2

(LER, RL = 0.96 %

Page 66 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Also compute Z(LER, RL) given by:

Z(LER, RL) = ABS(AV 1amp.RB- NTSP(ADJ) clamp-RB) / Sigma(LER,RL)

For this case of multiple channel, If Z(LER, RL) > 0.81 then LER avoidance condition is met.

Z(LER, RL) = ABS(109.5 - 108.5) / 0.96

= 1.05 This value of Z(LER, RL) is greater than the Z criterio4for I 0.81. Therefore the criterion is met without further adj dosent'r NTSP (ADJ) clmp-RB = 108.5 % Power (Rounded t st readab le** tl

5. Neutron Flux Downscale Rod Block The Required Limit (RL) of device "i" wifffn the loop is:

RLi = NTSPdown.RB - LAT

= 3.98% - 1.00% = 0 This is compared against AVdo from Section 4.1.7.3. Since R~i < AVdownRB, therefore NTSPdown.RB* (d adjusted:

NTS~ADJdw-,,C tB- LAT= 3.0+ 1.00 4.00 %

To determei tej h u tment is needed, the Required Limits of all devices in calcdl a 1edte

5) isloop the diSigma(LER, RL) given by the following equation (Ref.

R1\ ,DJ)dow- 0 I - LATi

)% - 1.00% =3.00%

2(RLi - NTSP(ADJ)dow-RB ) + (Cdown_)2 +(D x)2 For this calculation the loop error values are used which is equivalent to using one device with the error of the whole loop. Thus:

2 +(0.96)2 +(1.51)2

' ((2J(3.00-4.00o CLER,RL 2 0 96 CYLER, RL= . %

Page 67 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Also compute Z(LER, RL) given by:

Z(LER, RI) = ABS(AV- NTSP(ADJ)d0wn) / Sigma(LER,RL)

For this case of multiple channel, If Z(LER, RL) > 0.81 then LER avoidance condition is met.

Z(LER, RL) = ABS(3.0 - 4.00) / 0.96

= 1.05 This value of Z(LER, RL) is greater than the Z 0.81. Therefore the criterion is met without furthei Final NTSP (ADJ)do 0 , = 4.00 % Power (Rouneadre consemenly up to nearest a e`,I)crement)

6. Neutron Flux Fixed High SCRAM -Setd The Required Limit (RL) of device "i" wit,'I'll-, e ,ar the loop is:

R~i NTSPset-SCRAM + L4

=13.54% + 1.00% 51.40 This is compared against ANt % from Section 4.1.7.3. Since R~i > AVset.SCRAM, therefore NTSPset.sc e t0 be adjusted:

NTSP DJ),, s CRAM - LAT= 14.5 - 1.00 = 13.50 %

To determre( u tment is needed, the Required Limits of all devices in the

5) isloop *al Sigma(LER, RL) given by the following equation (Ref.

calctN tc '

ADJ),,ettscRAM+ LATi

.50% + 1.00% = 14.50%

( N- + (CsetSCRAM)2 + (Dfix)2 LER'RL(=- h 2 For this calculation the loop error values are used which is equivalent to using one device with the error of the whole loop. Thus:

-'LER,RL =

IZ ((2 (14.5-13.5)2 +(0.77)2 +(1.51)2 2

0 9 GLER,RL = . 1 %

Page 68 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Also compute Z(LER, RL) given by:

Z(LER, RL) = ABS(AVset.scpRM - NTSP(ADJ)setcRAM) / Sigma(LER,RL)

For this case of multiple channel, If Z(LER, RL) > 0.81 then LER avoidance condition is met.

Z(LER, RL) = ABS(14.5 - 13.5) / 0.91

= 1.10 This value of Z(LER, RL) is greater than the Z criter r 4 a s 0.81. Therefore the criterion is met without further adj2sment Final NTSP (ADJ)setSCRAM = 13.5 % Power (Rounýd C Ie invelymeto

7. Neutron Flux Fixed High Rod Block tdo The Required Limit (RI) of device "i" with th i es . T in the loop is:

RLi = NTSPset-p + LATi

= 10.54% + 1.00% F This is compared against AV, d- i ection 4.1.7.3. Since R~i > AVsetRB, therefore NTSPset.RB needs to be a ste NTSP (ADJ)sAet. LAT= 11.5 - 1.00 = 10.5%

To determn if ftrMffr- *uLjtstm t is needed, the Required Limits of all devices in the loop a ac d ma(LER, RL) given by the following equation (Ref.

5) is calcht RI NjýS I s5et-RB q-LATi

+1.00%: 11.5%

3 (R i - NTSP<ADJ)set-RB + (CsetRB)2 + (Dfix For this calculation the loop error values are used which is equivalent to using one device with the error of the whole loop. Thus:

ýj(~~((11.5-10.5 + (0.77)2 + (1.51)2 rYLERRL 2

0 9 1 GLER,RL = . %

Page 69 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET Also compute Z(LER, RL) given by:

Z(LER, RL) = ABS(AV set-E - NTSP(ADJ) set-R) / Sigma(LER,RL)

For this case of multiple channel, If Z(LER, RL) > 0.81 then LER avoidance condition is met.

Z(LER, RL) =ABS(1 1.5 - 10.5) 0.91

=1.10 This value of Z(LER, RL) is greater than the Z criter ev i*r! a ýs 0.81. Therefore the criterion is met without further adjts entntt Final NTSP (ADJ),st-pR = 10.5 % Power (Rouned' con e ively to n eai -,t !t*dat e increm ent) b) RBM Channel

1. Low Power Setpoint (LPSP)

The Required Limit (RL) of device "T wi ea in the loop is:

RLi = NTSPIpsp + LATi

= 26.31% + 1.00% .3 This is compared against AV1 , - om ection 4.1.7.3. Since R~i < AVlpsp, therefore NTSPepsp does n6i d to be adjusted:

NTSP (ADJ)Ipsp= , .0%% (Rounded conservatively readable increment) to nearest

2. Inter ed5P e tpoint (IPSP)

The Requ LrddL mitm(,L)

i~p*-LATiof device "i" with the largest LAT in the loop is:

I%+ 1.00% =62.31%

is red against AVipsp = 62.5% from Section 4.1.7.3. Since Kj-ýAVip,p I We NTSPipsp does not need to be adjusted:

NTSP (ADJ)1pP =NTSP1psp = 61.0 % (Rounded conservatively to nearest readable increment)

3. High Power Setpoint (HIPSP)

The Required Limit (RL) of device "i" with the largest LAT in the loop is:

RLi = NTSPhpsp + LATi

= 81.31% + 1.00% = 82.31%

This is compared against AVhpsp = 82.5% from Section 4.1.7.3. Since Rbi < AVhpsp, therefore NTSPhpsp does not need to be adjusted:

NTSP (ADJ)lpsp =NTSP1 psp = 81.0 % (Rounded conservatively to nearest readable increment)

Page 70 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

4. Downscale Trip Setpoint (DTSP)

The Required Limit (RL) of device "i" with the largest LAT in the loop for this decreasing setpoint is:

RLi = NTSPdtsp - LATi

= 93.90% - 1.00% = 92.90%

This is compared against AVdtsp = 92.0% from Section 4.1.7.3. Since RLi > AVdtsp, for this decreasing setpoint therefore NTSPdtsp does not need to be adjusted:

NTSP (ADJ)1 psp =NTSP1 psp = 94.0 % (Roundedl rv iv ares readable in/c eme~n~

5. Low Trip Setpoint (LTSP)

The Required Limit (RL) of device "i" with he larges L I op is:

RLi = NTSPItsp + LATi

= 112.09% + 1.00% = 113.09%

This is compared against AVltsp = 114.0% e1 .3. Since RLi < AVltsp, therefore NTSP 1 tsp does not n e ak* te NTSP (ADJ)lpsp =NTSPpsp \ -o

~readable 2. (Roundedincrement) conservatively to nearest

6. Intermediate Te4; loin TSP)

The Requir ýýLmi t o ice "i" with the largest LAT in the loop is:

RLi P 1AF Oh.59, -. 00% = 107.59%

This i a i nst AVitsp = 108.5% from Section 4.1.7.3. Since hPotP. does not need to be adjusted:

1 (ADJ)lpsp =NTSPlpsp = 106.5.% (Rounded conservatively to nearest readable increment)

High Trip Setpoint (HTSP)

The Required Limit (RL) of device "i" with the largest LAT in the loop is:

RLi = NTSPhtsp + LATi

= 102.59% + 1.00% = 103.59%

This is compared against AVhtsp = 104.5% from Section 4.1.7.3. Since RLi < AVhtsp, therefore NTSPhtsp does not need to be adjusted:

NTSP (ADJ)1 psp =NTSPpsp = 102.5 % (Rounded conservatively to nearest readable increment)

Page 71 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET 4.1.7.8 Selection of Operating Setpoint The recommended Operating Setpoints for the APRM and RBM are the NTSP(ADJ) values from section 4.1.7.7 OSP = NTSP(ADJ)

The lower limit of the setpoint (NTSP3) for purposes of performing the avoidance calculation is:

NTSP3 = OSP - (1.645/3) x SQRT( X LATi2) 4.1.7.9 Spurious Trip Avoidance Evaluation The Spurious Trip Avoidance Evaluation is used to I-ha er is a reasonable probability that spurious trips will not occ sig tsn ected -T'SP. The method of avoiding spurious trips is to determine tht od .y be present during normal plant operation and examine the margin be6e t qrst applicable operational transient for which trip is not required, th 1 it (NTSP3) of selected setpoint.

The following equation is that would be expected to contribute to a potential spL Sigma(STA) = (If TERMS)

Sigma(STA) = QRT (ALN 2 2

+ CL + DL +

2 PMA 2 + PEA 2 )

Once the value ofAi y T4is determined, the margin to the selected NTSP is calculated A= - Operational Limiti / Sigma(STA)

'Tom* , [,*u 7s m avoidance criterion (Ref. 4)

ýA > 1.65 e s scram criterion is not violated, no further adjustments are necessary.

tRM CHANNEL

1. Flow Biased Scram For the flow biased scram, the Operational Limit (OL) is considered to be the flow biased rod block Analytic Limit value.

OL = 0.75W + 54.8 Power Sigma(STA) = (1/2) x SQRT 2 2 (ALN~fo + Cmo.SCRAM + Din 2 + PEA 2 + PMA 2 )

= (1/2) x SQRT (2.762 +2.062 + 1.972 + 0.3 12 +2.412)

= 2.33

Page 72 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET For this function NTSP.scRm = 60.0, and therefore:

NTSP 3 fT-SCRAM = 60.0 - (1.645/3) x 2.00 = 58.90 % Power Z= ABSI NTSP3m-scRAM - OL I / Sigma (STA)

= ABSI 58.90 - 54.8 1/ 2.33

= 1.75 Since this value of Z corresponds to a probability of more than ok- id normalFendistribution), 1.65, the NTSPfb-sCRAM satisfies the

2. Flow Biased Rod Block For the flow biased rod block function, the Operatio al Lit () is not available.

Consequently the spurious trip avoidance aluat*in or this setpoint has not been computed.

3. Neutron Flux Fixed High SCRAM For the fixed neutron flux rathe Operational Limit (OL) is considered to be the rod bloc at 55o fow. The calculated value rod block setpoint at 75% flow is:

S Ro l om 75 + 49.0 = 105.25 % Power Sigma(STA) = (1/2)

x~ +x~ s + D* + PEAz PMA2

\v 1.6 3+ 1.65+ 1.512+0.31z+2.292 For this frod b ca fi NTSequenl 118 0utherefore:

118.0 - (1.645/3)x2.00 116.90% Power 1FtiSP3nfx.Scnk - OL Sigma (STA)

= ABSI 116.90 - 105.251 / 1.81

= 6.44 normal distribution),of 1.65, the NTSPfx_SCR*M satisfies the STA criteria.

4. Neutron Flux Rod Block Clamp For the rod block clamp function, the Operational Limit (OL) is not available.

i aConsequently the spurious trip avoidance evaluation for this setpoint has not been o ecomputed.

5. Neutron Flux Downscale Rod Block For the fixed neutron flux downscale rod block function, the Operational Limit (OL) is not available. Consequently the spurious trip avoidance evaluation for this setpoint has not been computed.

Page 73 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

6. Neutron Flux Fixed High SCRAM - Setdown For the Neutron High Flux Scram - Setdown function, the Operational Limit (OL) is considered to be that approximate power level whereby operations personnel would transfer the reactor mode switch to run or 9.5% power.

OL = 9.5% Power Sigma(STA) = (1/2) x SQRT (ALN-fix 2 + Cset.SCR M2 + Dfix2 + PEA2 +

(1/2) x SQRT 1.632 + 0.772 + 1.512 +0.3 12 +2.292)

- 1.65%

For this function NTSP(ADJ)set-SCRAM = 13.5%, therefore:

NTSP3set-SCRAM = 13.5 - (1.645/3) x 1.00 = 12.95 % w Z= ABSI NTSP3set-scRA - OL I / Sigma (

= ABSI 12.95 - 9.5 1/ 1.65

= 2.09 Since this value of Z corre if Wire than 95% (one-sided normal distribution) of 1.65, the STA criteria.

7. Neutron Flux Fixed I Setdown For the fixed neutron flux ,I b ck function, the Operational Limit (OL) is not available. Consequei6 spurious trip avoidance evaluation for this setpoint has not been compute.

4.1.7.10 Ae to the APRM and RBM channels which are electrical devices. The

.n loop flow transmitters are differential pressure devices and are not elevation correction.

of Actual Field Setpoint Since there is no elevation correction for the APRM and RBM channels:

Actual Field Setpoint (ASP ) = Operating Setpoint (OSP)

Page 74 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

5. CONCLUSION For As Left Tolerance (ALT) see section 4.1.3.5, and for the Leave Alone Tolerance (LAT) see section 4.1.7.6.

a) APRM Channel A The Analytic Limit (AL), Allowable Value (AV), calculated actual field Setpoints (equal to OSP ere is no elevation correction) for the APRM instruments NM-NAM-AR2,3,4,7,8, 9 are as follows:,

Trip Function Analytical Limit Allowable Value

1. Flow Biased Scram 0.75W + 65.6% 0.75W + 62.0%

2. Flow Biased Rod Block 0.75W + 54.8% 0.75W + 51.0%

3. High Neutron Flux Scram 123.0% 120.0%

4. Neutron Flux Rod Block Clamp 112.2% 109.5%

5. Downscale Neutron Flux Rod Block 0.0% 3.0%

6. High Flux - Setdown Scram 17.4% 14.5% 3.5%

ý.5%

7. High Flux - Setdown Rod Block 14.4% 11.5%

b) RBM Channel The Analytic Limit (AL), Allowable Value (AV), cz 11.ý6tpoints (equal to OSP since there is no elevation correction), for the ARTS / RBM ins nts NAf-AR5, 6 are as follows:

Trip Function AnaiI Allowable Setpoint

_ýLill J Value (OSP)

1. Low Power Setpoint (LPSP) A 27.5% 26.0%

2. Intermediate Power Setpoint )/ 62.5% 61.0%

3. High Power Setpoint (HPSP) Y 85o% 82.5% 81.0%

4. Downscale Trip Setpoint (DT 89.( % 92.0% 94.0%

1.20

5. Low Trip Setpoint (LTS 117.0% 114.0% 112.0%

1.25 120.0% 117.0% 115.0%

1.30 123.0% 120.0% 118.0%

1.35 125.8% 123.0% 121.0%

6. Inte;diat 1 *) 1.20 111.2% 108.5% 106.5%

1.25 115.2% 112.5% 110.5%

1.30 118.0% 115.0% 113.0%

1.35 121.0% 118.0% 116.0%

7. toint (HTSP) 1.20 107.4% 104.5% 102.5%

A 1.25 110.2%* 107.5% 105.5%

1.30 113.2% 110.5% 108.5%

1.35 116.0% 113.0% 111.0%

ReIeence 56 assumes a HTSP Analytical Limit of 111% for a RBM without filter. An unfiltered HTSP AL of 111% corresponds to a MCPR Limit of 1.25 in Table 10-5 (c) of Ref. 3. Therefore, selection of a filtered RBM Analytical Limit of 110.2% for HTSP, 115.2% for ITSP, and 120% for LTSP is equivalent to the analyzed value of 111% in Ref. 56.

Page 75 of 75 NEDC 98-024:

Rev. Number: 5C1 Nebraska Public Power District DESIGN CALCULATIONS SHEET The settings (based on Reference 22 and current site settings per Reference 10) for the ARTS/RBM timing functions are:

Time Delay 1 (Tdl) 3.5 sec. +/- 0.8 sec.

Time Constant I (Tcl) 0.5 sec. +/- 0.05 Time Constant 2 (Tc2) 6.0 sec.

Since the field functional testing and calibration cannot functionally meet the 4.1.3.5 and 4.1.7.6 (as their divisions are smaller than half the smallest divisi and leave alone tolerances can be adjusted. The tolerance adjustment will tolerance band is comparable to the tolerance bands stated within this calculatih The limit closest to the Allowable Value is to be moved further from the value corresponding to half the smallest division of the meter.

The limit furthest from the Allowable Value is to be moved i e er to the next value corresponding to half the smallest division of the meter.

NLS2008034 Enclosure 5 Reactor Core Thermal Power Uncertainty Calculation NEDC 06-035, Revision 1 (64 Pages)

Cooper Nuclear Station, Docket No. 50-298, DPR-46 Information Only

Page 1 of 64

Title:

Reactor Core Thermal Power Uncertainty Calculation Number: NEDC 06-035 CED/EE Number: 07-01 System/Structure: None Setpoint Change/Part Eval Number: None Component: None Discipline: Instrumentation and Control Classification: [ ] Essential; [ X ] Non-Essential SQAP Requirements Met? [ ] Yes; [ X ] N/A Proprietary Information Included? [ I Yes; [ X ] No

Description:

The purpose of this calculation is to determine the total uncertainty in the thermal power calculation performed by the GARDEL heat balance. The calculation is being performed in support of the Meter Uncertainty Power Uprate (MUR)

Licensing Application Request (LAR).

Conclusions and Recommendations:

The total uncertainty associated with the HEABAL (Reference 5.5.1) when feedwater flow is measured with the ASME Flow nozzles is -1.2985%/+1.8954%.

The total uncertainty with the HEABAL (Reference 5.5.1) when feedwater flow is measured with the Caldon UFM is

-0.3124%/+0.3118%.

Revision 1: Corrected minor errors in formulas, calculations, nomenclature, and the calculation of the lower bound of the total uncertainty.

1. 1 1 1 I 4 4 .1 4 1 1 3 Mark E. Unruh 40 51V Alan Able 6,*---4 Alan Able 3/17/08 _ T__AC;- as-__,. Z&C, _o_,/

0 3 Mark T. Lyman R. Krause 10/19/07 R. Krause 10/19/07 M. Dixon 10/23/07 Rev. Approved Number Status Prepared By/Date Reviewed By/Date IDVed By/Date By/Date Status Codes

1. Active 4. Superseded or Deleted 7. PRA/PSA

2. Information Only 5. OD/OE Support Only

3. Pending 6. Maintenance Activity Support Only

Page: 2 of 64 NEDC:06-035 Rev. Number: 1 Nebraska Public Power District DESIGN CALCULATION CROSS-REFERENCE INDEX ITEM DESIGN INPUTS REV. PENDING CHANGES NO. NO. TO DESIGN INPUTS 1 Drawing 2004, Sheet 2 N46 N/A 2 Drawing 2039 N54 DCN 07-0899, 07-0900 3 Drawing 2049, Sheet4 N13 N/A 4 Drawing 2849-4 N10 DCN 07-0091 5 Drawing 556-26811 4 N/A 6 Drawing 3043, Sheet 12 N16 7 Drawing 3254, Sheet 13 N22 DCN 04-1569, DCN 07-0528 8 Drawing 117C3485, Sheet 1 19 N/A 9 Drawing 117C3485, Sheet 2 7 N/A 10 Drawing 15887077 2 N/A 11 Drawing 730E148BB, Sheet 1 1 N/A 12 Drawing 791 E252, Sheet 1 N10 DCN 04-0616, 07-0084 13 Drawing 791 E254, Sheet 1 N08 N/A 14 Drawing 791E263, Sheet 2 N17 N/A 15 Drawing 0199F0377 N12 DCN 04-1444 16 Drawing 0199F0380 N12 DCN 04-1445 17 Drawing E515, Sheet 9 N01 N/A 18 Drawing E515, Sheet 55 N02 N/A 19 Drawing E515, Sheet 60 N01 N/A 20 Drawing E515, Sheet 82 N01 N/A 21 Drawing E515, Sheet 88 N01 N/A 22 Drawing E515, Sheet 121 N01 N/A 23 Drawing E515, Sheet 144 N01 N/A 24 Drawing CP008, Sheet 1 N04 N/A 25 Procedure 3.26.3 6 N/A 26 Procedure 14.NBI.301 4 Rev. 5 27 Procedure 2.2.8 68 N/A 28 Procedure 15.RR.301 3 N/A 29 Procedure 2.2.66 87 N/A 30 Studsvik Scandpower Report: SSP- 0 N/A 104/414-C 31 Rosemount 00809-0100-4801 AA N/A 32 RTP Bulletin RTP7436 Series N/A N/A

Page: 3 of 64 NEDC:06-035 Rev. Number: 1 Nebraska Public Power District DESIGN CALCULATION CROSS-REFERENCE INDEX ITEM DESIGN INPUTS REV. PENDING CHANGES NO. NO. TO DESIGN INPUTS 33 RTP 981-0021-211A A N/A 34 KEPCO 146-1869 N/A N/A 35 Rosemount 00813-0100-2654 FA N/A 36 Rosemount 00813-0100-4021 FA N/A 37 Rosemount Cal. Report 1/5/06 N/A 38 Rosemount 00809-100-4360 AA N/A 39 Badger Cal. Data Report N/A 40 GE 198 4532K30-010 N/A N/A 41 Scientific Columbus Bulletin N/A N/A 42 Daniel Technical Bulletin N/A N/A 40 Barton Bulletin Gl-25 N/A N/A 43 Part Eval. Tech. No. 10511159 0 N/A 44 NEVC 70-263 (Roll 00111, Frame 0 N/A 0033) 45 GE 21A1379AR (Roll 09021, 4 N/A Frame 1898) 46 Alden Report (Roll 08118, Frame 7/70 N/A 1295) 47 GE FDI 71/101000 (Roll 17524, 1 N/A Frame 0604 48 CED 6010820 0 CCN 1-16, OSC 1-16 49 NEDC 00-095A 4 N/A 50 NEDC 94-018 3 3C1 THRU 3C21 51 CED 2000-0032 0 N/A 52 Cameron Report ER-592 I N/A 4 * +

4 4 4 4 t 1*

Page: 4 of 64 NEDC:06-035 Rev. Number: 1 Nebraska Public Power District DESIGN CALCULATION CROSS-REFERENCE INDEX ITEM AFFECTED DOCUMENTS REV. NUMBER NO.

____None.

t I t I t t

Page: 5 of 64 NEDC:06-035 Rev. Number: 1 The purpose of this form is to assist the Preparer in screening new and revised design calculations to determine potential impacts to procedures and plant operations.0l SCREENING QUESTIONS YES NO UNCERTAIN 1 Does it involve the addition, deletion, or manipulation of [ ] X] [ ]

a component or components which could impact a system lineup and/or checklist for valves, power supplies (breakers), process control switches, HVAC dampers, or instruments?

2 Could it impact system operating parameters (e.g., [] [X] [I temperatures, flow rates, pressures, voltage, or fluid chemistry)?

3 Does it impact equipment operation or response such as I] [X1 [ I valve closure time?

4 Does it involve assumptions or necessitate changes to the [ ] [X] [ ]

sequencing of operational steps?

5 Does it transfer an electrical load to a different circuit, or [] [X] []

impact when electrical loads are added to or removed from the system during an event?

6 Does it influence fuse, breaker, or relay coordination? [] [X] []

7 Does it have the potential to affect the analyzed [] [X] []

conditions of the environment for any part of the Reactor Building, Containment, or Control Room?

8 Does it affect TS/TS Bases, USAR, or other Licensing [ I [X] [I Basis documents?

9 Does it affect DCDs? [] X] [I 10 Does it have the potential to affect procedures in any way not already mentioned (refer to review checklists in [ ] [I ] [ I Procedure EDP-06)? If so, identify:

If all answers are NO, then additional review or assistance is not required.

If any answers are YES or UNCERTAIN, then the Preparer shall obtain assistance from the System Engineer and other departments, as appropriate, to determine impacts to procedures and plant operations. Affected documents shall be listed on Attachment 2.

Page: 6 of 64 NEDC:06-035 Rev. Number: 1 Nebraska Public Power District DESIGN CALCULATIONS SHEET

1. PURPOSE:

To determine the uncertainty of the reactor thermal power (heat balance) calculation performed by the process computer (GARDEL). Cooper Nuclear Station (CNS) will be installing highly accurate ultrasonic feedwater flow meters. This calculation will determine the uncertainty in reactor thermal power calculation when the reactor heat balance is performed using the installed ASME flow nozzles to measure feedwater flow and when the process computer is using a corrected feedwater flow measurement provided by the Caldon LEFM Check Plus Ultrasonic Flow Meters (UFMs).

On June 1, 2000 Appendix K to Part 50 of Title 10 of the Code of Federal Regulations was changed allowing licensees to use a power uncertainty of less than 2 percent in their LOCA analysis. The change allowed licenses to recapture power by using state-of-art devices to more precisely measure feedwater flow. Feedwater flow inaccuracy is a large contributor in the uncertainty determination of reactor power. This calculation is being performed in the support of a License Application Request (LAR) for a Measurement Uncertainty Recapture (MUR) power uprate.

2. ASSUMPTIONS:

2.1. The heat balance per Reference 5.5.1 assumes 0.0 for steam moisture fraction. The fraction assumed for steam moisture is a conservative value and results in a higher calculated power than actual power. Since this parameter is always conservative it is not accounted for in the heat balance thermal power uncertainty calculation.

2.2. The energy lost due to radiation and heat transferred through vessel is assumed to be 0.6 MW in the heat balance. There is no plant or fleet specific uncertainty number for the losses assumed in the plant process computer. Per Reference 5.6.10 the uncertainty in this number is high and a

+/-50% uncertainty in the number would result 0.015% uncertainty in CTP for a 3400 MWth plant. It is assumed a +/-50% bounding uncertainty can be associated with 0.6 MW radiative loses at CNS. This results in a radiative loss uncertainty of +/-0.3 MW.

2.3. The plant process computer uses mathematical operators, conversions and algorithms to process the signals from the process instrumentation. In addition, tables based on ASME and Keenan and Keyes are used to determine fluid enthalpies and per Reference 5.6.10 the differences in the different steam tables are negligible when compared to other uncertainties in the overall core power determinations. The operations performed by the process computer are carried to many significant digits and are considered very accurate. It is therefore concluded that no considerations are necessary for computation uncertainties of the process computer.

Page: 7 of 64 NEDC:06-035 Rev. Number: 1 2.4. Control Rod temperature is not measured and it is assumed in the process computer that Control Rod Drive (CRD) temperature is always 100 *F. The CRD pumps take suction on the hotwell level makeup line. The line is from Condensate Storage Tank A, suction of the Condensate Booster Pumps, and Condenser Hotwell (

References:

5.1.2, 5.1.5, and 5.1.9). The water is normally 102.83. 'F per Reference 5.1.60 and the water in Condensate Storage Tank A is maintained above 40°F per Reference 5.1.15. Although it is possible the CRD pumps may take suction directly on the Condensate Storage tanks it is highly unlikely except during a outage. It has therefore been assumed in this calculation that CRD temperature when at or near 100%

power is I02.83°F with a +/-10°F band added for conservatism. Furthermore, as given in Reference 5.6.10 the calculated core thermal power is insensitive to the temperature assumed for control rod water temperature.

2.5. CRD system provides Reactor Recirculation (RR) pump purge seal flow, Reactor Water Clean-up (RWCU) mini-purge pump seal flow, and continuous cold reference leg backfill. The flow paths are upstream of the CRD flow element and the flow is not accounted for in the measured CRD Flow. RR pump purge seal is adjusted to be between 1.65 and 1.85 gpm per pump (Reference 5.3.7), RWCU mini-purge is adjusted be between 1 and 2 gpm per pump (Reference 5.3.5), and each reference leg (2) backfill is adjusted to be between 0.0095 to 0.0105 gpm (Reference 5.3.8). The total maximum flow not accounted for in the CRD measured flow is 7.721 gpm. The heat balance accounts for this flow by adding a constant 6 gpm to the measured CRD flow (Reference 5.5.1). The error associated with the flows and the contribution the flows would have to the core thermal power uncertainty is assumed very small and negligible.

2.6. The original calibration data for the RWCU flow elements (orifices) was not retrievable.

However, in Reference 5.4.1, paragraph 11-111-7, when it is not possible to calibrate an orifice in the meter tube assembly the discharge coefficient can be calculated. The discharge coefficient in the calculation is not expected exceed 11.0%. Another +/-1.0% was added to the assumed and expected accuracy to conservatively account for lack of the original manufacturer's calibration data.

2.7. The calculation was performed at based line conditions for the current licensed power of 2381 MWth. However, the calculation is considered applicable when operating near 100% thermal power and for 1.7% increase in the thermal power increase. A 1.7% power increase will for most parameter result in no change and for some parameters (e.g. feeedwater parameters) only a slight change (< 2 *F and 5 psi).

Page: 8 of 64 NEDC:06-035 Rev. Number: 1

3. METHODOLOGY:

Reactor Thermal Power is calculated in the-heat balance program HEABAL (Reference 5.5.1). The thermal power is determined by the simple relationship of Q = Energyout - Energy,. HEABAL uses the below equation to determine Core Thermal Power (CTP).

CTP = Qs + Qcu + QFL - QFW - QCR - QP Where:

Qs = Steam Energy Qcu = RWCU Energy QFL = Radiative Power Losses QFW = Feedwater Energy QCR = CRD Energy Qp = Recirculation Pump Energy Per Reference 5.5.1 each of the terms of the heat balance equation are determined using the below listed plant parameters (instruments). The calculation also uses several constants for conversion and assumed factors (e.g. 0.0 for moisture).

PC Signal GARDEL Description Plant Instrument Name B031 TFWA1 FW Loop A Temperature RF-TE-140C B033 TFWA2 FW Loop A Temperature RF-TE-140D B030 TFWB1 FW Loop B Temperature RF-TE-140A B032 TFWB2 FW Loop B Temperature RF-TE-140B B015 WFWA2 FW Loop A Flow RFC-FT-50A B016 WFWB2 FW Loop B Flow RFC-FT-50B B025 PRI Steam Dome Pressure NBI-PT-53A or 53B B014 WCR1 Control Rod Drive Flow CRD-FT-204 B023 TCUI RWCI Inlet Temperature RWCU-TE-92 B024 TCUO RWCU Outlet Temperature RWCU-TE-109

Page: 9 of 64 NEDC:06-035 Rev. Number: 1 B017 WCU RWCU F/D A Flow RWCU-PE-77A B018 WCU RWCU F/D B Flow RWCU-PE-77B B019 PPI RR A Pump Motor Power RR-XFMR-TRlA B020 PP2 RR B Pump Motor Power RR-XFMR-TRIB The calculation will determine the individual instrument loop uncertainties used in the heat balance using the methodology given in Reference 5.4.2. Not each instrument loop used in the heat balance contributes equally to the thermal power calculation uncertainty. Therefore, baseline conditions will be established at 100 percent reactor power and then a sensitivity analysis will be performed to determine the effects or weighting factors that each loop uncertainty has on the heat balance calculation. The results of the loop uncertainties and weighting factors will be combined and the individual loops will than be totalized to determine the total uncertainty in the reactor thermal power heat balance calculation.

Vendor documentation (non-As-Built) has been used for determination of instrument accuracy. The information is contained mainly in Manufacturer's Equipment Specifications and Technical Bulletins. The vendor documents are prime and many times the only source for determining the accuracy of the instrument.

Listed below are the various instrument module uncertainties considered for each of the process instrument loops.

DE Instrument Drift Effect FE Flow Element Fouling Effect IE Flow Element Installation Effect MTE Meter and Test Equipment Effect OE Other Effects PSE Power Supply Effect RA Reference Accuracy RE Resistor Effect RES Resolution Effect ST Calibration Setting Tolerance

Page: 10 of 64 NEDC:06-035 Rev. Number: 1 TDE Temperature Effect on Density TE Temperature Effect TLU Total Loop Uncertainty 3.1. Feedwater Flow Uncertainty 3.1.1. Feedwater Flow Elements RF-FE-11A Mfr.: Permutit ASME Flow Nozzle Serial Number: T- 12125 RF-FE-11B Mfr.: Permutit ASME Flow Nozzle Serial Number: T-12126 3.1.1.1. Reference Accuracy:

Per Reference 5.6.2 the Feedwater Flow Element has a +/-0.5% accuracy. The accuracy is based on the discharge coefficient and represents an uncertainty on the actual flow. Per Reference 5.6.3 the flow elements were calibrated in a test line and had a calibration accuracy of 0.00096 and 0.00219 for T-12125 and T-12126, respectively. Therefore, as a matter of conservatism the accuracy of Reference 5.6.2 will be used as the reference accuracy for the flow elements.

RAFWFE = +/- 0.5% of Actual Flow 3.1.1.2. Installation Effect:

The installation effect of the flow elements is dependent of whether the flow nozzles were installed per industry standards. Per Reference 5.4.1 paragraph II-Il-3 if the installation meets the minimum pipe lengths upstream and downstream then the error due to piping configuration is less than +/-0.5 % otherwise an additional tolerance of +/-0.5 % should be applied to the flow measurement. In accordance with Figure 1I-I-1, a minimum of 16 Diameters upstream and 2.3 Diameters downstream of straight pipe should be installed for a flow nozzle with a 13ratio of slightly greater than 0.5. Per Reference 5.1.13 and Reference 5.6.3 nozzles T-12125 and T-12126 have 13ratios of 0.5192 and 0.5187, respectively.

The upstream and downstream number of pipe diameters per References 5.1.12 and 5.1.13 are given in the below table.

Page: 11 of 64 NEDC:06-035 Rev. Number: 1 T-12125 T-12126 Required Upstream >18D >18D 16D Downstream >4D >4D 2.3D Since, the installation of the flow elements meets the minimum requirements of Reference 5.4.1 the uncertainty due to the installation effect is:

IEFWFE = +/-0.5% of Actual Flow 3.1.1.3. Fouling Effect:

Feedwater Flow nozzles have been known to foul due to the plate-out of iron oxides which result from the corrosion of carbon steel condensate and feed water systems. In Reference 5.5.17 a flow verification test and analysis was performed on reactor feedwater using the Caldon External LEFM Feedwater Flow Measurement System. The report concluded the LEFM flow rate was higher than the plant instrumentation value by -0.007%. It can be inferred from the test results that there little or no fouling of the ASME flow nozzles at CNS. However, as a matter of conservatism the value given in Caldon Report (5.5.16) on flow uncertainty due to fouling will be used to envelope any fouling of the flow nozzles at CNS.

FEFWFE = +0.6% of Actual Flow (Bias Term) 3.1.1.4. Temperature Effect:

A change in feedwater temperature can affect both the density and nozzle expansion. Both nozzle expansion and density corrections are made on the feedwater mass flow rate by the Process computer. Per Reference 5.4.1, Figure 11-1-3, the Thermal Expansion Factor (Fa) is 1.0054 for the rated temperature of 367.1 'F. The calculated error for feedwater temperature is < 1'F. The error introduced on the Area Factor due to the temperature error is a very small factor, not readable on Figure 11-1-3 ( on the order of .OXI OE-5), when operating off the rated feedwater temperature (367.1 °F) and can be neglected. The affect on density due to the temperature error with respect to differential pressure will be conservatively evaluated at a +/- 1 'F at the rated conditions of 367.1 "F and 1175 psia (Reference 5.6.4).

Page: 1_2 of 64 NEDC:06-035 Rev. Number: 1 p-1 (1175 psia, 366.1 OF) = 55.291 ibm/ft3 p (1175 psia, 367.1°F) = 55.252 ibm/ft3 p+1 (1175 psia, 368.1 OF) = 55.214 lbm/ft3 The original calibration data for the transmitters based on the flow nozzles is given in Reference 5.6.4 as:

Flow T-12125 T-12126 FLOW Inches H20 at 68°F Inches H20 at 68°F MIb/HR Zero 0 0 0.0 Rated 733.7 739.6 4.761 Full Scale 1165.2 1174.6 6.000 CED6010820 will respan the transmitter to 10 Mlb/HR full scale. Therefore, the new calibration pressures will be established using the relationship for flow versus differential pressure.

W = K (pDP)°'5 Taking the ration of the two flow conditions:

W__= K (oDPi)0' 05 W2 = K (p 2 DP 2) .

Where: T1 = T2 & P1 = P2 therefore: P1 = P2 and the new full scale DP becomes:

2 DP 2 = DP1 (W2 / W1)

Flow T-12125 T-12126 FLOW Inches H20 at 68°F Inches H20 at 68°F MIb/HR Zero 0 0 0.0 Rated 733.7 739.6 4.761 Full Scale 3,236.7 3,262.8 10.000

Page: 13 of 64 NEDC:06-035 Rev. Number: 1 The new respanned transmitter differential pressure values for full scale indication will be used in the forthcoming evaluation.

The percent DP at the rated conditions:

Loop A: (733.7/3236.7) X 100 = 22.6681% DP Span Loop B: (73 9.6/3262.8) X 100 = 22.6676% DP Span Using the larger of the two DP's for conservatism and keeping flow constant (rated) for the two conditions.

DP 2 = DPI(pI/P 2) 22.6681%(55.252/55.291) = 22.6521%

22.6521% - 22.6681% = -0.0160%

p+1:

22.6681%(55.252/55.214) = 22.6837%

22.6837% - 22.6681% = +0.0156%

TDEFWFE = -0.0160%/+0.0156% or +/- 0.0160% of Span (Temperature Dependent) 3.1.2. Feedwater Flow (Differential Pressure) Transmitters RFC-FT-50A Manufacturer: Rosemount Model: 3051S Ultra RFC-FT-50B Manufacturer: Rosemount Model: 3051S Ultra Per Reference 5.6.5 the installed Rosemount 115 1DP transmitters will be replaced with Rosemount 3051 S Ultra Differential Pressure Transmitters (Part Number: 3051 S-1 -C-D-4A-2-A 11-A-1 A-D 1-M5-Q4-T 1-A1266).

3.1.2.1. Reference Accuracy:

Rosemount Reference Manual (Reference 5.5.2) lists a Reference Accuracy of:

+/-0.04% of span; for spans less than 10:1, accuracy =

+[0.005 + 0.0035(URL/Span]% of Span

Page: 1_4_4 of 64 NEDC:06-035 Rev. Number: 1 The specification is for 3 sigma, however the accuracy will not be converted to 2 sigma for added conservatism. The Upper Range Limit (URL) for a Range 4 is 300 psi. Converting to inches of water gives:

(300 psi) X (27.7277 "1E120/psi) = 8318.31 "H20 Using the smaller of the spans (3,236.7"H 20, second table is Section 3.1.1.4) for conservatism, the reference accuracy becomes:

+/-[0.005 + 0.0035(8318.31 "H2 0 /3,236.7"H 2 0 ]% of Span

=+/-0.0140% Span RARFiT = +/-0.0140% Span 3.1.2.2. M&TE Effect:

The test equipment calibration standards used to calibrate the transmitters are traceable to National Institute of Standards and Technology (NIST) and the test equipment is as least as accurate if not better than the equipment being calibrated.

Therefore, the M&TE uncertainty will be taken as the Reference Accuracy of the transmitters.

MTEpRFT = +/-0.0140% Span 3.1.2.3. Setting Tolerance:

The Setting Tolerance is taken as the As-Left tolerance of the instrument. Per Reference 5.2.5 and 5.2.6 the as Left- Tolerance is +/-0.1 mvDC for the transmitters. So the equivalent uncertainty of a 4 to 20 maDC loop is:

(0.1 ma / 16 ma) X 100 = 0.6250%

STRFFT = +/-0.6250% Span 3.1.2.4. Temperature Effect:

The transmitters are located in the Turbine Building Basement Control Corridor on Instrument Rack IR-11D per Reference 5.1.18. The Control Corridor has a minimum temperature of 78 'F and a maximum temperature of 1 10°F per

Page: 15 of 64 NEDC:06-035 Rev. Number: 1 Reference 5.6.1. The ambient change in temperature from the minimum calibration temperature to the maximum operating temperature is:

(110°F - 78°F) = 22-F Per Reference 5.5.2 the temperature effect is:

+(0.009%URL + 0.04%Span) per 50'F Using the larger of the spans (3,262.8"H 2 0) the temperature effect is:

+/--(0.00009 X 8,318.3 1"H20 + 0.0004 X 3,262.8"H 20)22/50

= +/-0.9037"H 20 Converting to percent span:

+/-(0.9037"H 20/3,262.8"H 2 0)XI 00 = -0.0277%

TERym = +/-0.0277% Span 3.1.2.5. Drift Effect:

Per Reference 5.5.2 the stability of the 305 1S is +/- 0.20% of the Upper Range Limit (URL) for 10 years. The transmitter is a Range 4 transmitter and has a URL of 300 psi. Converting pressure into inches of water at 68 'F gives a value of:

(300 psi) X (27.7277 "H20/psi) = 8318.3 1"H20 (8318.31 "H2 0) X (0.0020) = 16.6366"H 2 0 Using the lower of the two transmitter spans for conservatism to find drift per percent span:

+(16.6366"H 20 / 3236.7"H 20) X 100 = +/-0.5140% Span The transmitters are calibrated every 18 months versus 120 months and using the relationship given in Reference 5.3. 1:

Page: 16 of 64 NEDC:06-035 Rev. Number: 1 VDM = (M/120mo)" 2 X VDi 20 mo (18/120)1/2(0.5141% Span) = 0.1991% Span DELRFT +/-0.1991% Span 3.1.2.6. Static Pressure Effect:

Some differential pressure transmitters experience a shift in the output due to line pressure. The effect is on both the zero point and the span of the transmitter.

Model 3051 Series Ranges 4 and 5 are subject to this static pressure shift; however the zero error from the static pressure effect is calibrated. The span error per Reference 5.5.2 is +/-0.1% of reading per 1000 psi. Using the total span for conservatism the Static Pressure Effect is:

+/-(0.001 X 3,262.8"H20)(1175psi/1000psi)

= +/-3.8338"H 20 Converting to %Span:

(+/-3.8339"H20/3,262.8) X 100 = +/-0.1175%

SPERFFT = +-0.1175% Span 3.1.2.7. Power Supply Effect:

The transmitters have less than a +/-0.005% of calibrated span per volt per Reference 5.5.2. The transmitters are powered from (or will be powered from after the implementation of CED6010820) two 24 VDC auctioneered KEPCO power supplies. The KEPCO power supplies are powered from either 115 AC Inverters RFC-IVTR-INVA or RFC-IVTR-INVB. The inverters are powered from the 125 VDC Batteries (Reference 5.6.5). The power to the KEPCO power supplies is extremely reliable and stable. Per Reference 5.5.5 the KEPCO power supplies have a maximum reference variation of 0.3% combined effects from source, load, temperature, and time (drift). Using the maximum variation of the power supplies and stated power supply effect of the transmitters the following evaluation determines the power supply effect on the transmitters.

(24 VDC) X (0.003) X (0.005%/volt) = 0.0004%

PSERFFT = +/- 0.0004% Span I

Page: 17 of 64 NEDC:06-035 Rev. Number: 1 3.1.2.8. Resistor Effect:

The transmitter output provides a signal to the process computer by converting the current to a voltage by means of the voltage drop across a 15 Ohm +/- 0.1% resistor (Reference 5.6.5). Therefore, the effect of the resistor in the instrument loop is given as:

RERFFT = +/- 0.1% Span 3.1.2.9. Process Computer Effect:

The voltage drop off the precision resistor is used to provide an input to the MUX card which converts the analog signal to a digital signal. Per Reference E5 15, Sheet 144 the MUX card is Part Number RTP 021-5234-003 (Model Number:

RTP 7436/50-003) which is a 8-Channel Universal High-Speed Wide Range Gate Card. The card is part of a set and per computer applications personnel and the PMIS Database Information Editor the associated card in the set is an RPT 038-5097-000 (Model Number: RPT 7436/21-0008) which is the Analog-to-Digital (A/D) Converter Card. The card set per References 5.5.3 and 5.5.4 have a linearity of +/- 0.025% of full scale and a gain accuracy of less than +/- 0.025% of full scale.

The linearity and gain accuracy are combined using SRSS method to determine the Reference Accuracy.

RApc = +/- [(0.025)2 + (0.025)210.5 % Span RApc= +/-0.0354% Span The resolution effect (quantizing uncertainty) for the A/D card is +/- 1/2 LSB (Least Significant Bit). The card is 14 bits so the resolution becomes:

RESpc = +(LSB / 2 Number Bits) X 100%

RESpc = +/-(0.5 / 21") X 100%

RESpc = +/-0.0031% Span Setting Tolerance effects are errors introduced during the calibration process. Per Reference 5.5.3 the cards are not calibrated and no operator adjustments are made which could introduce an error due to a calibration process therefore, the Setting Tolerance and M&TE are negligible. Furthermore, the instrument drift is negligible.

Page: 18 of 64 NEDC:06-035 Rev. Number: 1 I The cards are located in MUX 9-86 which is located in the Computer Room (Reference 5.1.28). The Computer Room is an air-conditioned controlled environmental space. For this reason any temperature effect is considered negligible.

The process computer uses algorithms and conversions to process the plant parameters. The computer also uses tables such as the ASME steam tables.

These tables are highly accurate and the computations are digitally performed to many significant digits. The error associated with the computations is miniscule compared to loop inaccuracies and are therefore negligible.

3.1.3. Total Loop Uncertainty Feedwater Flow Tabulated below are the uncertainties associated with the Feedwater flow loop.

RAFWFE +/- 0.5% of Actual Flow IEFWFE +/- 0.5% of Actual Flow FEFWFE +0.6% of Actual Flow (Bias)

TDEFWFE +/-0.0160% (Dependent)

RARFFT +/-0.0140% Span MTERrvr +/-0.0140% Span STRFrr +/-0.6250% Span TERFFT +/-0.0277% Span DEarFT +/-0.1991% Span SPERFMr +/-0.1175% Span PSERFFT +/-0.0004% Span RERFF'r +/- 0.1% Span RApc +/-0.0354% Span RESpc +/-0.0031% Span 2

TLUFwFI = +/- (RAFWFE + IEFWFE2)0'5 TLUFwFI = +/-0.7071% of Actual Flow

Page: 19 of 64 NEDC:06-035 Rev. Number: 1 2 2 2 2 TLUFWF2 = -(RARFr 2 + MTERFFT + STRUT + TERFFT + DERFFT + PSEpJFT + SPERFsr" 2 2 SRERFFT2 + RApc 2 + RESpc 2)° 5 TLUFWF2 = +/-_0.6756% Span I FEFWFE = +0.6000% of Actual Flow (Bias)

TDEFWFE = +/-0.0160% Span (Dependent with Temperature) 3.2. Feedwater Temperature Uncertainty 3.2.1. Feedwater Temperature Elements:

RF-TE-140A Manufacturer: Rosemount Model: 0078L21C30N080AX8X9Q4 RF-TE-140B Manufacturer: Rosemount Model: 0078L21 C30N080AX8X9Q4 RF-TE-140C Manufacturer: Rosemount Model: 0078L21C30N080AX8X9Q4 RF-TE-140D Manufacturer: Rosemount Model: 0078L21C30N080AX8X9Q4 3.2.1.1. Reference Accuracy:

The RTD Model Numbers are the new RTD's to be installed per CED6010820.

The RTD's are 100 Ohm, 4-Wire, Platinum, Single Element RTD's. Per the Rosemount RTD calibration data sheets the accuracy of the RTD's is +/-0.22 'F (Reference 5.5.8).

RAFWTE = +/-0.22°F The RTD's are not adjustable and are not calibrated so there is no Setting Tolerance uncertainty or M&TE uncertainty.

The lead wire resistance effect is negligible for a 4-Wire RTD's.

The feedwater flow velocity at the piping location of the RTD's is approximately 20.48 ft/sec. Therefore, the self heating affect of the RTD's can be neglected.

3.2.2. Feedwater Temperature Transmitters:

RF-TT-168A Manufacturer: Rosemount Model: 3144PDIAINAB4M5TIC4Q4 RF-TT-168B Manufacturer: Rosemount Model: 3144PD1AINAB4M5T1C4Q4 RF-TT-168C Manufacturer: Rosemount Model: 3144PD1A1NAB4M5T1C4Q4 RF-TT-168D Manufacturer: Rosemount Model: 3144PD 1A1NAB4M5T1 C4Q4

Page: 20 of 64 NEDC:06-035 Rev. Number: 1 3.2.2.1. Reference Accuracy:

The RTD's are Class A Sensors per Reference 5.5.6. Per the I&C Calibration Sheets the span of the transmitters are from 0.0 to 150.0 mV for 280'F to 430°F.

Per Reference 5.5.7 the accuracy of the temperature transmitters is +/- 0.18 'F plus

+ 0.02% of span this gives a reference accuracy of:

+/-[((430 'F - 280 'F) X (0.0002)) + 0.18 °F)]= +/- 0.21'F RAR-rr= +0.21 'F 3.2.2.2. Setting Tolerance:

The Setting Tolerance is bounded by the "As-Left" Tolerance of the temperature transmitter: Per References 5.2.1 through 5.2.4 the "As-Left" for the old transmitter is given as +/- 0.3 mV for 0 to 150 mV span, this equates to +/-0.2% span which is the Setting Tolerance of the new transmitters.

+/-(0.002)(430 'F - 280 'F) = +/- 0.3°F STR*n = +/-0.37F 3.2.2.3. M&TE effect:

The test equipment calibration standards used to calibrate the temperature transmitters are traceable to NIST and the test equipment is as least as accurate if not better than the equipment being calibrated. Therefore, the M&TE uncertainty will be taken as the Reference Accuracy of the transmitters.

MTERFT = +/-0.21 'F 3.2.2.4. Drift Effect:

Per Reference 5.5.7 the stability (drift) is +/- 0.1 % of reading or 0.1 'C (0.18 'F) which ever is greater. At the maximum span temperature of 430 'F the error becomes:

+/-0.001(430'F) = +/- 0.43°F DERrT = +/-+0.43°F

Page: 21 of 64 NEDC:06-035 Rev. Number: I 3.2.2.5. Power Supply Effect:

Per Reference 5.5.7 the power supply effect is less than +/- 0.005% span per volt.

The temperature transmitters are to be power from the same source as the reactor feedwater transmitters (See Step 3.1.2.7), therefore:

+(24 VDC) X (0.003) X (0.00005/volt) X (150 'F) = +/-0.0005'F PSEp, 1r= +/-0.0005'F 3.2.2.6. Temperature Effect:

The temperature transmitters are located in panel 9-21 in the Cable Spreading Room per Reference 5.1.35. The Cable Spreading Room is an air-conditioned controlled environmental space. The possible change in temperature from the calibration temperature to the normal operating temperature is considered to be very small and therefore, the temperature effect is considered negligible.

3.2.2.7. Resistor Effect:

Per References 5.6.5 the voltage drop to the process computer is across a 15 Ohm

+ 0.1% resistor. In a 4 ma to 20 ma loop the span is 60 to 300 mV and the resistor effect at maximum span becomes:

+/-(300 mV X 0.001)((150 'F/(300 mV - 60 mV)) +/-0.1875°F RE --r = +/-0.1875'F 3.2.2.8. Process Computer Effect:

Per References 5.1.46, 5.1.47, 5.1.48, and 5.1.51 the reactor feedwater temperature loops uses the same mux cards as the reactor feedwater differential pressure loops (Refer to Step 3.1.2.9).

RApc = +/-0.0354% Span = +/- (150'F)(0.000354)= +/-0.0531'F RESpc = 0.0031% Span = +/- (150'F)(0.000031) = +/-0.0047°F

Page: 22 of 64 NEDC:06-035 Rev. Number: 1 3.2.3. Total Loop Uncertainty Feedwater temperature.

Tabulated below are the uncertainties associated with the Feedwater temperature loop.

RAFWTE +/-0.22°F RAjrrr +/-_0.21 'F STRFTT +/-0.3 F MTERrr +/-0.21°F DEyR-_ +/-0.43'F PSERMTT +/-0.0005°F RERFrr +/-0.1875°F RApc +/-0.0531'F RESpc +/-0.0047°F TLURFTEM 2

= +/-(RAFWTE + RARF2 + STRFTT +

2 MTERFTr 2 2

+ DERFFT + PSERFTT2 2 2 0 5 2

+ RARFFT + RESpc ) '

+ RERFrr TLURFTEMp = +/-0.6704°F There are four feedwater temperature loops and per Reference 5.5.1 one loop temperature from each loop is used to determine the loop enthalpy and loop enthalpies are averaged in the power calculation. The averaging of the loop enthalpies is equivalent to the individual feedwater loop temperatures being averaged and therefore, the TLU becomes:

TLUAvERAGE = - (TLURrTEMp/(2) 0 '5)

TLUAVERAGE = +/-0.4741 -F 3.2.4. Ultrasonic Flow Meter RF-CC-4 Manufacturer: Caldon Model: LEFM 4I+ 2000FC Measurement System The Caldon Ultrasonic Flow Meter (UFM) will be used to correct feedwater flow signal.

Per Reference 5.6,11 the UFM has a total bounding mass flow accuracy of +/- 0.30 %. The quoted accuracy is the total accuracy of the mass flow signal from the Caldon UFM it incorporates: installation effects, piping effects drift effects, power supply effects, calibration effects, temperature effects and density.

RAUFM = +/-0.30% Flow

Page: 23 of 64 NEDC:06-035 Rev. Number: 1 3.3. Vessel Pressure Uncertainty 3.3.1. Pressure Transmitter:

NBI-PT-53A Manufacturer: Rosemount Model: 3051 S Ultra NBI-PT-53B Manufacturer: Rosemount Model: 3051S Ultra 3.3.1.1. Reference Accuracy:

Per Reference 5.6.5 the installed Rosemount 15 1DP pressure transmitters will be replaced with Rosemount 3051S Pressure Transmitters (Part Number: 3051 S-1-C-G-5A-2-A11-A-iA-D1-M5-P1-Q4-TI-A1266-Q8). Either NBI-PT-53A or NBI-PT-53B is used to provide the pressure signal depending on the position of the level switch on panel 9-5 (Reference 5.1.38). The transmitters have a span of 14 psig to 1214 psig (14 psi for head correction) for a 4 ma to 20 ma output (Reference 5.3.2). A range 5 transmitter per Reference 5.5.2 has an URL of 2000 psi and has a Total Performance of +/-+0.125% of span (combined error of reference accuracy, ambient temperature and line pressure effect).

RApT = +/-0.125% of Span 3.3.1.2. Setting Tolerance:

The Setting Tolerance is bounded by the "As-Left" Tolerance of the transmitter:

Per References 5.3.2 the "As-Left" is given as +/- 0.08 Ma of the output, so the Setting Tolerance is:

STPT = +(0.08 Ma/16 Ma) x 100 = +/-0.5% of Span 3.3.1.3. M&TE:

The test equipment calibration standards used to calibrate the transmitters are traceable to NIST and the test equipment is as least as accurate if not better than the equipment being calibrated. Therefore, the M&TE uncertainty will be taken as the Reference Accuracy of the transmitters.

MTEpT = +/-0.125% of Span

Page: 24 of 64 NEDC:06-035 Rev. Number: 1 3.3.1.4. Drift Effect:

Per Reference 5.5.2 the stability of the 3051S is +/- 0.20% of the Upper Range Limit (URL) for 10 years and +/- 50 TF. The transmitter is a Range 5 transmitter and has a URL of 2000 psi.

+/- (2000 psi) X (0.0020) = +/-4 psi

+/-(4/1200) X 100 = +/-0.3333 % of Span for 10 years The transmitters are calibrated every 18 months versus 120 months and using the relationship given in Reference 5.3.1:

VDM = (M/120mo)" 2 X VD1 2omo (18/120)1/(0.3333% Span) = +/-0.1291% Span DEpT = +/-0.1291% Span 3.3.1.5. Power Supply Effect:

The transmitters have a +/-0.005% of calibrated span per volt per Reference 5.5.2.

The transmitters are powered from (or will be powered from after the implementation of CED6010820) two 24 VDC auctioneered KEPCO power supplies. The KEPCO power supplies are powered from either 115 AC Inverters RFC-IVTR-INVA or RFC-IVTR-INVB. The inverters are powered from the 125 VDC Batteries (Reference 5.6.5). The power to the KEPCO power supplies is extremely reliable and stable. Per Reference 5.5.5 the KEPCO power supplies have a maximum reference variation of 0.3% combined effect from source, load, temperature, and time (drift). Using the maximum variation of the power supplies and stated power supply effect of the transmitters the following evaluation determines the power supply effect on the transmitters.

(24 VDC) X (0.003) X (0.005%/volt) = +/-0.0004%

PSEpr = +/-0.0004% Span

Page: 25 of 64 NEDC:06-035 Rev. Number: 1 3.3.1.6. Temperature Effect:

The pressure transmitters are located on instrument rack 25-56 which is located in the Reactor Building, South East, Elevation 903'. Reference 5.6.6 gives the normal temperature variation as 70'F to 104'F. This is a delta T of less than 50

'F which is accounted for in the Total Performance (Reference Accuracy).

3.3.1.7. Resistor Effect:

The transmitter output provides a signal to the process computer by converting the current to a voltage by means of the voltage drop across a 15 Ohm +/- 0.1% resistor (Reference 5.6.5). Therefore, the effect of the resistor in the instrument loop is given as:

REpT--- +/-0.1% Span 3.3.1.8. Process Computer:

Per References 5.1.46 the reactor feedwater temperature loops uses the same mux cards as the reactor feedwater differential pressure loops (Refer to Step 3.1.2.9).

RApc = +/-0.0354% Span RESpc =+0.0031% Span

Page: 26 of 64 NEDC:06-035 Rev. Number: 1 3.3.2. Total Reactor Pressure Loop Uncertainty Tabulated below are the uncertainties associated with the Reactor Pressure loop.

RApT +/-0.125% of Span STPT +/-0.5% of Span MTEpT +/-0.125% of Span DEPT +/-0.1291% Span PSEPT +/-_0.0004% Span REPT +/-0.1% Span RApc - 0.0354% Span RESpc +/-0.0031% Span 2

TLUpT =

2

+/-(RAPT2 + STpT2 + MTEpT 2 + DEPT + PSEPr + REPT 2 + RApc 2 + RESpc 2) 0 ,5 TLUPT - +/- 0.5560% of Span 3.4. Control Rod Drive Flow Uncertainty 3.4.1. Control Rod Flow Element CRD-FE-203 Manuf.: Badger Meter Co. Style: PVF-B Serial No.: G38823-1 3.4.1.1. Reference Accuracy:

Per Reference 5.1.32 the accuracy of the venturi is +/-1% and the manufacturer calibration data sheets (Reference 5.5.10) state a resolution of 1.00% of maximum. The reference accuracy is therefore given in percent actual flow.

RACRDFE = +/--1.0% Actual Flow 3.4.1.2. Installation Effect:

The venturi is installed in 2-inch diameter pipe and has a beta ratio of 0.51 (Reference 5.5.10). Per Figure II-II-1 of Reference 5.4.1 the venture requires 32 inches of upstream straight pipe. By inspection of Reference 5.1.54 the flow

Page: 27 of 64 NEDC:06-035 Rev. Number: 1 element does not meet the installation requirement, so an additional +/-0.5%

installation error will be added to the +/-0.5% standard installation error as stated in paragraph II-I1-3 of Reference 5.4.1.

IEcRDFE = +/-1.0% Actual Flow 3.4.1.3. Temperature Effect on Nozzle Expansion:

Per Reference 5.5.10 the flow element calibration differential pressure is based on demineralized water at 150 "F and a flow rate equivalent to 100% power. The CRD pumps take suction on the hotwell level makeup line. The line is from Condensate Storage Tank A, suction of the Condensate Booster Pumps, and Condenser Hotwell (

References:

5.1.2, 5.1.5, and 5.1.9). The water is normally 103.1 °F per Reference 5.1.60 and the water in Condensate Storage Tank A is maintained above 40 °F per Reference 5.1.15. Per Reference 5.4.1, Figure 11-1-3, the Thermal Expansion Factor (Fa) is approximately 0.9995 and 1.001 for 40 "F and 150 *F, respectively. Using the equation 1-5-32 of Reference 5.4.1 for volumetric flow and for given differential pressure (calibration pressure value for fluid at 150 °F) the error in the flow measurement is determined as temperature deviates from the calibration temperature to 40 *F.

5 Q = a Fa K (2gh)° Since, the calibration differential pressures are for calibrating at 68 °F for 150 °F fluid.

Q68 = (a(2gh)° 5)/ Fa15 0 Q40 = Fa4O(Q68)

Q40 (Fa 4d/Fa1s0) (a(2gh)0 ")

0 Q4o = (0.9985/1.001) (a(2gh)° )

Q4o = 0.9975 (a(2gh)0 '5)

So actual volumetric flow at 40 is 0.9975 of the indicated flow. Since CRD fluid temperature is always less than 150 'F the error due to the thermal expansion of nozzle (actually contraction of the nozzle) is small and always conservative because the calibration data has been biased high and therefore can be ignored.

Page: 28 of 64 NEDC:06-035 Rev. Number: 1 3.4.1.4. Temperature Effect on Density:

The Process Computer does not correct for changes in density of the CRD fluid.

The calibration data used to scale the loop for 0 gpm to 160 gpm is based on a fluid temperature of 150 'F and 2000 psia, whereas the fluid temperature may be as low as 40 °F (highly unlikely except during an outage), pressure may be as low as 1425 psig (1425 + 14.7 = 1439.7 psia) and cooling water is set 50 gpm (Reference 5.3.3). The error due to density is conservatively accounted for by deviating CRD temperature from the calibration data temperature to 40*F.

DP at 50 gpm and 160 gpm, per Reference 5.5.10, is 77.14 and 789.1 inches of water, respectively and as a percent span:

(77.14/789.1) X 100 = 9.7757% DP Span The base conditions are:

pl(2000 psia, 150 *F) = 61.576 lbm/ft 3 3

4 4 0 psia, 40 °F) = 62.735 lbm/ft p2(1 DP 2 = (PI / P2) DP 1

= (61.576/62.735) X 9.7757%

= 9.5951%

9.5951% -9.7757 % =-0.1806%

Therefore, the maximum percent of span error due to density is:

TDEcpDFE = -0.1806% of Span (Temperature Bias) 3.4.2. Differential Flow Transmitter CRD-FT-204 Manufacturer: Rosemount Model: 1151DP 3.4.2.1. Reference Accuracy:

Per Reference 5.5.9 the accuracy of the 1151DP transmitter, Range Code 6 is -

0.25% of the calibrated span.

RACRDFT = +/-0.25% of Span

Page: 29 of 64 NEDC:06-035 Rev. Number: 1 3.4.2.2. Setting Tolerance:

The Setting Tolerance is bounded by the "As-Left" Tolerance of the transmitter:

Per References 5.2.8 and Instrumentation and Calibration Supervisor the Generic tolerance is +/-2%.

STCRDFT = +/-2.0% of Span 3.4.2.3. M&TE:

The test equipment calibration standards used to calibrate the test is traceable to NIST and the test equipment is as least as accurate if not better than the equipment being calibrated. Therefore, the M&TE uncertainty will be taken as the Reference Accuracy of the transmitters.

MTEcann- = +/-0.25% of Span 3.4.2.4. Drift Effect:

Per Reference 5.5.9 the stability of the 1151 Range 6 is +/-0.25% of the Upper Range Limit (URL) for 6 months. The transmitter is a Range 6 transmitter and has a URL of 100 psi.

+/-(100 psi) X (0.0025) = +/-_0.25 psi

+(0.25 psi)(27.7277 "HEO/psi) = +/-6.9319 "H20

+/-(6.9319/789.1) X 100 = +/-_0.8785% of Span for 6 months The transmitters are calibrated every 18 months versus 6 months and using the relationship given in Reference 5.3.1:

VDM = (MND 6 )l2VD 6mo (18/6)/2(0.8785% Span) = +1.5216% Span DEcIFT = +/- 1.5216% Span

Page: 30 of 64 NEDC:06-035 Rev. Number: 1 3.4.2.5. Static Pressure Effect:

The 115 1DP transmitters are subjected to a static pressure shift which is not calibrated out. The shift is constant over the range of the transmitter and can be in either direction. Per Reference 5.5.9 the zero error is +/- 0.25% URL for 2,000 psi and the span error is correctable to +/- 0.25% of input reading per 1000 psi. For conservatism the span error will be applied to the calibrated span of the transmitter. Per Reference 5.3.3 CRD pump discharge pressure is set between 1425 psig and 1475 psig.

Zero Error as a %Span is:

Zero Error = +/-[((0.0025)(100 psi ))(27.7277 "H 20/psi/789.1 "H 20))X 100] [1475 psi/2000psi]

Zero= +/-0.6479%

Span Error- +/-(0.0025%)(1475 psi/1000 psi)X100 = +/-0.3688%

Since the errors are additive: SPEcRDFT = +/-(0.64792 + 0.36882)0.5 %

SPECRDFT = +/-0.7455% Span 3.4.2.6. Power Supply Effect:

The transmitters have less than a +/- 0.005% of calibrated span per volt per Reference 5.5.9. The transmitters are powered from CRD-ES-309 (Reference 5.1.36) which is a GE Type 570 Power Supply. Per Reference 5.5.11 the line voltage of the power supply can fluctuate between 107 volts and 127 volts ac with output voltage. holding to 52.5 Vdc +/- 8%. Using the maximum variation of the power supplies and stated power supply effect of the transmitters the following evaluation determines the power supply effect on the transmitters.

(52.5 VDC) X (0.08) X (0.005%/volt) = +/-0.021%

PSEcRDvr = +/-0.021% Span

Page: 31 of 64 NEDC:06-035 Rev. Number: 1 3.4.2.7. Temperature Effect:

The transmitters are located on the south-east wall of the Reactor Building, Elevation 903' per Reference 5.1.54. Reference 5.6.6 gives the normal temperature variation as 70°F to 104 'F. This is a delta T of 34 'F. Reference 5.5.9 gives a total temperature effect of 1(0.5% URL + 0.5% of calibrated span) per 100"F.

+/-0.5% URL is:

+/-(0.005)(100 psi)(27.7277 "H 20/psi) = 13.8639 "H20

+/-0.5% URL as %Span:

(13.8639 "H120/789.1 "1120) X 100 = 1.757% Span Adding 0.5% of Span:

+/-(1.757% + 0.5%) 34 -F /100 -F = +/-0.7674%

TECRDFT = +/-0.7674% Span 3.4.2.8. Resistor Effect:

The transmitter output provides a signal to the process computer by converting the current to a voltage by means of the voltage drop across a 6 Ohm +/- 0.1% resistor (Reference 5.1.36). Therefore, the effect of the resistor in the instrument loop is given as:

RECRDFT _0.1% Span 3.4.2.9. Process Computer:

Per References 5.1.47 the CRD flow loop uses the same mux cards as the reactor feedwater differential pressure loops (Refer to Step 3.1.2.9).

RApc = +/- 0.0354% Span RESpc =-0.0031% Span

Page: 32 of 64 NEDC:06-035 Rev. Number: 1 3.4.3. Total Loop Uncertainty for CRD Flow Tabulated below are the uncertainties associated with the CRD flow loop.

RACRDFE +/-1.0% Actual Flow IECRDFE +1.0% Actual Flow TDEcPDFE - 0.1806% Span RACRDFT +/-0.25% Span STCRDFT +/-2.0% Span MTEcRDFr +/-0.25% Span DECRDFT +L1.5216% Span SPECRDFT +/-0.7455% Span PSEcDF-r +0.021% Span TECRDFT +0.7674% Span RECRDFT +0.1% Span RApc +L0.0354% Span RESpc +0.0031% Span 205 TLUCRDFI = +( RACRDFE + IEcRDFE ) .

TLUCRDFI = +1.4142% Flow TLUCRDF2 = - 0.1806% Span (Temperature Bias)

TLUCRDF3 = +( RACRDFT 2+ STcRFj 2 + MTECRDFT2 + DEcRDFr 2 + SPECRDFT + PSEcRDFT 2 2 2 2 05

+ TECRDFT2 + RECRDFT + RApc + RESpc ) "

TLUcRDF3 = +/-2.7562% Span 3.5. CRD Temperature Uncertainty The process computer does not have a CRD temperature, the heat balance program HEABAL (Reference 5.5.1) assume CRD temperature is 100 *F. However, the condensate pumps are 33% capacity pumps and continuously recirculate a portion of the discharge to maintain proper system flow (Reference 5.3.6) and the recirculated flow is through the same line as the

Page: 33 of 64 NEDC:06-035 Rev. Number: 1 suction of the CRD pumps. It has therefore been assumed in this calculation that CRD temperature when at or near 100% power is 103.1 *F (Reference 5.1.60) with a +10*F band added for conservatism.

TLUCRDT =10 'F 3.6. Reactor Recirculation Power Uncertainty 3.6.1. Reactor Recirculation Pump Power Watt Meter RR-XFMR-TR1A Manuf.: Scientific Columbus Model: XL3IK5A2-SC-ER RR-XFMR-TR1B Manuf.: Scientific Columbus Model: XL31K5A2-SC-ER 3.6.1.1. Reference Accuracy:

Reference 5.5.12 states an accuracy of +/-(0.2% Reading + 0.01% RO), where RO is the rated output for the AC Watt Transducers. In accordance with Reference 5.5.12 the rated output of the transducer is +/-1 mAdc. The 0 to 160 mV signal to the Process Computer is taken off a 160 ohm +/- 0.1% resistor for 0 to 6.0 MW indication (References 5.1.43 and 5.1.44). Using the maximum span value for conservatism the reference accuracy becomes:

+(0.002(160) + 0.0001(160))/160 X 100 = - 0.21% Span RARRwm = +0.21% Span 3.6.1.2. Setting Tolerance:

The Setting Tolerance is bounded by the "As-Left" Tolerance of the transmitter:

Per References 5.3.4 the "As-Left" is given as +/- 8 mV of the output, so the Setting Tolerance is:

+(8mV/160mV) X 100 = +/-5.0% Span STRRwM = +/-5.0% Span

Page: 34 of 64 NEDC:06-035 Rev. Number: 1 3.6.1.3. M&TE:

The test equipment calibration standards used to calibrate the watt meter are traceable to NIST and the test equipment is as least as accurate if not better than the equipment being calibrated. Therefore, the M&TE uncertainty will be taken as the Reference Accuracy of the transducer.

MTERRwM = +/-0.21% Span 3.6.1.4. Drift Effect:

Per Reference 5.5.12 the stability per year of the transducer is +/- 0.1% of the RO, and since the RO is same as the span the drift effect becomes:

+/-( 18/12)0.5(0.1% Span) = +/-0.1225% Span DERawM = +/-0.1225% Span 3.6.1.5. Power Supply Effect:

The transducers have an external power requirement of 85 to 135 Vac (120vAC nominal) and a frequency of 50 to 500 Hz per volt per Reference 5.5.12. The accuracy quoted is maintained if the external power requirements are within specification. The transducers are powered from power panels 2-184-12A and 2-184-12B which in turn are powered from Critical Instrument and Control Power Panels EE-PNL-CCP 1A and EE-PNL-CCP I B, respectively (Reference 5.6.8).

The power supplies are highly reliable and have a minimum and maximum voltage of 102.8 and 127.9 Vac per Reference 5.6.7. Therefore, the power supply affect can be neglected.

3.6.1.6. Temperature Effect:

The transducers are located in panels 2-184-12A and 2-184-12B (Reference 5.6.8) which are located in the Reactor Building, West, Elevation 976'. Per Reference 5.6.6 the normal temperature for the area is 70 'F to 104 °F and Reference 5.5.12 states a temperature effect of J-0.005%/1C.

+/-[(34 0F)(5/9 °C/-F)(+/- 0.005%//C) = +/-0.0944%

TERxwM = +/-0.0944% Span

Page: 35 of 64 NEDC:06-035 Rev. Number: 1 3.6.1.7. Resistor Effect:

The transmitter output provides a signal to the process computer by converting the current to a voltage by means of the voltage drop across a 160 Ohm +/- 0.1%

resistor (References 5.1.43 and 5.1.44); Therefore, the effect of the resistor in the instrument loop is given as:

RERjwM= +/-0.1% Span 3.6.1.8. Process Computer Per References 5.1.46 and 5.1.47 the RR power loops uses the same mux cards as the reactor feedwater differential pressure loops (Refer to Step 3.2.1.9).

RApc = +/-0.0354% Span RESpc = +0.0031% Span 3.6.2. Reactor Recirculation Power Total Loop Uncertainty Tabulated below are the uncertainties associated with the RR Power loops.

RARRwM +0.21% Span STRRwM +5.0% Span MTERRwM +/-0.21% Span DERRwM +/-0.1225% Span TERRWM +0.0944% Span RERRwM +0.1% Span RApC +0.0354% Span RESpc +/-0.0031% Span 2 2 TLURRwM = +( RARRwM + STRRWM + MTERRWM2 + DERRWM2 + TERRWM2 + RApc2 +

RESpc) 0°5 TLUaRwM = +/-5.0123% Span

Page: 36 of 64 NEDC:06-035 Rev. Number: 1 3.7. Reactor Water Clean-up (RWCU) Flow Loop Uncertainty 3.7.1. Flow Elements RWCU-FE-74A Manuf.; Daniel Industries Inc. Model: 520 Paddle RWCU-FE-74B Manuf.; Daniel Industries Inc. Model: 520 Paddle 3.7.1.1. Reference Accuracy The Daniel Model 520 Paddle orifice is a standard concentric flat plate orifice and the orifice bore tolerance is in strict accordance with A.P.I., Chapter 14, Section 3 (Reference 5.5.13). The original manufacturer's calibration data sheets for the orifice plates are no longer retrievable. However, in Reference 5.4.1, paragraph 11-111-7 (also Chapter II-V), when it is not possible to calibrate an orifice in the meter tube assembly the discharge coefficient can be calculated. The discharge coefficient in the calculation is not expected exceed +/- 1.0%. Another +/- 1.0% will be added to the assumed accuracy to account for manufacturer's tolerances and to ensure conservatism. The bore of the orifice is 1.522 inches by visual inspection of the stamping and the piping is 3"CU-3S which is schedule 40 stainless steel (3.068" ID).

RWCU Flow = (101 gal/min)(min/60sec)(233 in 3/gal)(ft3/1728 in 3)(1/(ar4(d/2) 2)

= 17.965 ft/sec Rd = pVd/g Rd = ((61.92 lbm/ft3)(17.965ftIsec)(1.522 in/12 in/ft))/(3.8E-4 lbm/sec-ft)

Rd = 371,285 f3 = d/D 1.522 in/3.068 in f3=0.4961 Per Reference 5.4.1, Tables II-111-2(a) and (b) the discharge coefficient is approximately 0.6058 and meets the requirement of paragraph 11-111-7.

RACUFE = +/-2.0% Flow

Page: 3_7 of 64 NEDC:06-035 Rev. Number: 1 3.7.1.2. Installation Effect:

The flow elements do not meet the installation requirements of Reference 5.4.1 and so 1/2% error must be added to the standard error of Y2% for the installation effect.

IEcUFE = +/-1.0% Flow 3.7.1.3. Temperature Effect on Nozzle Expansion Per Reference 5.3.5the fluid temperature to the filter/deminerlizers is maintained

< 130*F and per Reference 5.1.33 the normal temperature is 120 *F. The flow elements are just upstream the filter/demineralizers and since the system temperature operating band is small there is little change in the nozzle expansion factor, Fa. The change in Fa from Figure 114--3 of Reference 5.4.1 is on the order of 0.0002 or less for temperature range of I IOF to 130%. The temperature affect on nozzle expansion can be neglected.

3.7.1.4. Temperature Effect on Density During normal operations the temperature band for the fluid entering the filter/demineralizers is approximately 120 'F to 130TF. Although the fluid temperature is normally maintained in a small band, the affect temperature has on density will conservatively be evaluate at +/- 10 'F from the base condition of 120°F at 1143 psia at Flow Element (Reference 5.1.33).

P.10 (1143 psia, 110 OF) = 62.073 lbm/ft3 p (1143 psia, 120 OF) = 61.920 lbm/ft3 3

p+10 (1143 psia, 130'F) = 61.767 Ibm/ft The flow through each filter/demineralizer during normal operations is approximately 101 gpm (Reference 5.1.33). The loop is spanned 0 to 120 gpm per References 5.2.9 and 5.2.10.

Per Reference 5.2.9 and 5.2.10:

103.9 gpm = 158.80 "H2 0 120gpm = 211.74 "H20

Page: 38 of 64 NEDC:06-035 Rev. Number: 1 So the D/P at 101 gpm is:

D/Plol = 158.80 "H20 (101 gpm/103.9gpm)2

= 150.06"H20 As a percent of the total Span:

(150.06 "H 20/211.74 "H 20) X 100 = 70.87%

Therefore:

DP 2 = (PI / p2)(DP1)

DP-10 = (61.920 lbm/ft3/62.073 lbm/ft3)( 70.87%)

= 70.695%

70.695% -70.87% = -0.1750%

DP+i0 = (61.920 lbm/ft3/61.767 lbm/ft3)(70.87%)

= 71.0455%

71.0455% - 70.87% = +0.1755%

TDECuFE = -0.1750%/+0.1755%

TDEcUFE = +/-0.1755% Span 3.7.2. Flow Transmitter RWCU-FT-75A Manufacturer: Barton Inst. Systems Model: 273A RWCU-FT-75B Manufacturer: Barton Inst. Systems Model: 273A 3.7.2.1. Reference Accuracy:

Per Reference 5.5.14 the Model 273A Pneumatic Transmitters have a linearity of

+/- 1/2 % of full range output pressure for differential pressure span to 150 psi.

Converting the total span to pressure:

211.74 "H 20 X 0.036065 = 7.636 psi at 68 'F RAcutr = +/-0.5% Span

Page: 39 of 64 NEDC:06-035 Rev. Number: 1I 3.7.2.2. Setting Tolerance:

The Setting Tolerance is bounded by the "As-Left" Tolerance of the transmitter:

Per References 5.2.9 and 5.2.10 the "As-Left" is given as +/- 0.06 psi of the output and the output span of the instrument is 0 to 15 psi. However, the instrument is calibrated as a loop so the Setting Tolerance is applied to the output of the final instrument in the loop prior to the process computer which is the Pneumatic to Current Converter.

3.7.2.3. M&TE:

The test equipment calibration standards used to calibrate the test is traceable to NIST and the test equipment is as least as accurate if not better than the equipment being calibrated. Therefore, the M&TE uncertainty will be taken as the Reference Accuracy of the transducer.

MTECUFT = +/-0.5% Span 3.7.2.4. Drift Effect:

Due to the mechanical nature of the instrument the drift effect is negligible.

3.7.2.5. Temperature Effect:

Due to the mechanical nature of the instrument the temperature effect is negligible.

3.7.2.6. Static Pressure Effect:

Reference 5.5.14 does not state any static pressure effect for the flow transmitter.

However, conversations with Prime Measurement Products (formerly Barton) indicated there is a slight static pressure shift. For a differential pressure of 60 psi and below the shift is +/- 0.1% per 1000 psi on the span and for above 60 psi the shift is +/- 0.25% per 1000 psi. Because the shift was listed on an internal document the Vendor was unwilling to provide it as a reference. The span is 15 psi and the operating pressure is 1143 psia (Reference 5.1.33).

+/-0.1% (1143 psi/I000 psi) = +/-0.1143%

SPECUFT = +/-0.1143% Span

Page: 40 of 64 NEDC:06-035 Rev. Number: 1 3.7.3. Pneumatic to Current Converter RWCU-PE-77A Manufacturer: Moore Model: PIX/33-15PSIG/10-50MA/12-42DC RWCU-PE-77B Manufacturer: Moore Model: PDC/33-15PSIG/10-50MA/12-42DC 3.7.3.1. Reference Accuracy:

Reference 5.5.15 the accuracy of pneumatic to current converter is +0.2% of span.

Since, the error should be in both directions the reference accuracy will be taken as +/-0.2% Span RAcup = +/-0.2% Span 3.7.3.2. Setting Tolerance:

The Setting Tolerance is bounded by the "As-Left" Tolerance of the transmitter:

Per References 5.2.9 and 5.2.10 the "As-Left" is given as +/- 0.1 mA of the output for a 10 to 50 mA loop, so the Setting Tolerance is:

+/--(0.1 mA/40 mA) X 100 = +/-0.25% Span STcup = +/-0.25% Span 3.7.3.3. M&TE:

The test equipment calibration standards used to calibrate the converter are traceable to NIST and the test equipment is as least as accurate if not better than the equipment being calibrated. Therefore, the M&TE uncertainty will be taken as the Reference Accuracy of the transducer.

MTEcup = +/-0.2% Span

Page: 41 of 64 NEDC:06-035 Rev. Number: 1 3.7.3.4. Temperature Effects:

The transducers are located in the Reactor Building, West, Elevation 958', RWCU Valve Room. Per Reference 5.6.6 the normal temperature for the area is 70°F to 104°F and Reference 5.5.15 states a temperature effect of less than +/-2.0% of full scale input over the specified ambient temperature operating range of 30°F to 130*F. The input full scale is the span, so:

+/-(2%)((1040 F - 70°F )/(130'F- 30TF) = - 0.68%

TEcup = +/-0.68% Span 3.7.3.5. Other Effects:

The Moore Products (Reference 5.5.15) does not state any effects for drift and states power supply effect of less than +/-0.01% of rated span be volt of change in line voltage. Power supply effects are very small and drift effects have been less than 0.2% so the other effects will be conservatively assumed as +/-0.25%.

OEcup = +/-0.25% Span.

3.7.3.6. Resistor Effect:

The transmitter output provides a signal to the process computer by converting the current to a voltage by means of the voltage drop across a 6 Ohm +/- 0.1% resistor (Reference 5.1.39). Therefore, the effect of the resistor in the instrument loop is given as:

REcup = +/-0.1% Span 3.7.3.7. Process Computer Per References 5.1.48 and 5.1.51 the RWCU Flow loops uses the same mux cards as the reactor feedwater differential pressure loops (Refer to Step 3.1.2.9).

RApc = +/-0.0354% Span RESpc = +/-0.0031% Span

Page: 42 of 64 NEDC:06-035 Rev. Number: 1 3.7.4. Reactor Water Cleanup Flow Total Loop Uncertainty Tabulated below are the uncertainties associated with the RWCU flow loops. Since the instruments are calibrated as a loop only the final Setting Tolerance (STcUP) is considered in the total loop uncertainty.

RACUFE +/-2.0% Flow IEcUFE +/-1.0% Flow TDECUFE +/-0.1755% Span RAcuFr +/-0.5% Span MTEcuFr +/-0.5% Span SPEcuT +/-0.1143% Span RAcup +/-0.2% Span STcup +/-0.25% Span MTEcup +/-0.2% Span TEcup +/-0.68% Span OEcup +/-0.25% Span REcup +/-0.1% Span RApc +/-0.0354% Span RESpc +/-0.0031% Span 5

TLUCUF1 = _( RAcurEI + IEcUFE 2 + TDECUFE2) 0 TLUcuFI = +/- 2.2429% Flow 2 2 2 2 TLUcuF2 = +/-( RACUFT 2 + MTEcuFT + SPECUFT + RAcup + STcup 05 2 2 2

+MTEcup2+ TEcul 2 + OEcUP + REcUP + RApc +RESpc)

TLUcuF2 = +/-1.0917% Span 3.8. Reactor Water Clean-up Temperature Loop Uncertainty RWCU-TE-92 GE Provided P/N: 117C3485P017 IST CONAX Nuclear Inc.

RWCU-TE- 109 GE Provided P/N: 117C3485P017 ARI Industries Inc.

Page: 43 of 64 NEDC:06-035 Rev. Number: 1 3.8.1. Temperature Elements The temperature elements used to determine inlet and outlet RWCU temperature are Type T (Copper - Contantan) thermocouples (References 5.1.30 and 5.1.31). Per Reference 5.1.39 the temperature loop span is 0 "F to 600 *F and the Standard Limit Error (ANSI MC 96.1, Reference 5.4.4) for a type T thermocouple in the temperature range of> 32 °F to 662 'F is 1.8 *F or 0.75% which ever is greater. Reference 5.1.33 gives a normal operating temperature at the inlet thermocouple of 532 'F and at the outlet thermocouple of 435 OF:

+/-(5320F)(0.0075) = +/-3.9900°F RAcuTi = -3.9900°F

+/-(435-F)(0.0075) = +/-3.26250F RACUT2 = +/-3.26250 F There are no adjustments to be made on the thermocouples, no calibrations are performed and no resistors in the temperature loops. The thermocouples are feed directly to the process computer mux cards.

3.8.2. Process Computer Per References 5.1.49 and 5.1.50 the temperature loops uses the same mux cards as the reactor feedwater differential pressure loops (Refer to Step 3.1.2.9).

RApc= +/-0.0354% Span +/- (600 'F)(0.000354) = +/-0.2124°F RESpc = +0.0031% Span = t (600 'F)(0.000031) = +/-0.0186°F

Page: 44 of 64 NEDC:06-035 Rev. Number: 1 3.8.3. Total RWCU Temperature Loop Uncertainty Tabulated below are the uncertainties associated with the Temperature loops.

RAcuTr + 3.9900 *F RACUT2 +/- 3.2625 *F RApc +/- 0.2124*F RESpc - 0.0186 *F TLUcuTr =( RACUTI +RApc 2+ RESpc2) 0 5" TLUCUTI = +/-3.9957 'F 2 205 TLUcuT2 = +/-( RACUT 22 + RApc + RESpc )° TLUcur2 = +/-3.2695 "F

Page: 45 of 64 NEDC:06-035 Rev. Number: 1 3.9. Sensitivity Analysis The individual loop uncertainties do not contribute equally to the uncertainty in the core thermal power calculation. To determine the contribution that the error in a single parameter makes to the uncertainty in the power calculation baseline conditions are established and the sensitivity of that measured parameter is determined. The sensitivities are determined from the baseline plant conditions at 100% thermal power The baseline conditions are those values given in the Thermal Kit (Reference 5.1.60) for CNS main turbine at 100% thermal power. The operating points established in the Kit represent the most accurate representation of the present performance of the station secondary components and systems. It will provide the baseline conditions for the most important of the parameters of interest (e.g., FW flow, temperature, pressure, and temperature). CRD parameters with exception of temperature will be taken from Procedure 2.2.8 (Reference 5.3.3) and RWCU parameters at 100% power will be taken from the GE process drawing for the system (Reference 5.1.33). Using the baseline parameters weighting factor are established for each of the instrument loops and then the loop uncertainties are multiplied by the weighing factor to determine the contribution to the power uncertainties. The values derived are then combined statistically to determine the uncertainty in the thermal power calculation. In general a 5% span on the parameter baseline value will be used to determine the parameter weighting factor.

Derived values of enthalpy and density are taken from the ASME Steam Tables - 1967 (Reference 5.4.3). These values were also used in the loop uncertainty calculations.

BASELINE VALUES I Parameter Value Reference FW Total Flow 9.489 Mlbm/hr 5.1.60 FW A Flow 4.745 Mlbmihr 5.1.60 FW B Flow 4.745 Mlbm/hr 5.1.60 FW Pressure 1165 psia 5.1.60 FW Temperature 364.68 *F 5.1.60 FW Enthalpy 338.73 Btu/lbm 5.1.60 Vessel Dome Pressure 1019.99 psia 5.1.60 Saturated Steam 1191.55 Btu/lbm 5.1.60 Enthalpy

Page: 46 of 64 NEDC:06-035 Rev. Number: 1 BASELINE VALUES Parameter Value Reference CRD Flow 50 gpm 5.3.3 CRD Pressure 1450 psig (1465 5.3.3 psia)

CRD Temperature 102.83 TF 5.1.60 CRD Enthalpy 75.105 Btu/lbm 5.4.3 RWCU Flow 202 gpm 5.1.33 RWCU A Flow 101 gpm 5.1.33 RWCU B Flow 101 gpm 5.1.33 RWCU Pressure (FE) 1143 psia 5.1.33 RWCU Inlet 532 T 5.1.33 Temperature RWCU Outlet 435 T 5.1.33 Temperature RWCU Inlet Enthalpy 527.606 Btu/lbm 5.4.3 RWCU Enthalp Outlett Enthalpy 414.015 btu/lbm 5.4.3 RWCU Inlet Pressure 1193 psia 5.1.33 RWCU Outlet Pressure 1075 psia 5.1.33 RR A Pump Power 1.9 Mw 5.5.1 RR B Pump Power 1.9 Mw 5.5.1 3.9.1. FW Flow Sensitivity The uncertainties calculated for FW flow measurement was in percent actual flow and percent differential pressure.

For percent actual flow both feedwater flow element and Ultrasonic Flow Meter uncertainties are significantly less than 1% , however a 5% will be used to determine the weighing factor.

Error Flow (lbrn/hr) = 0.05(9.489 Mlbm/hr) = 474,450 lbm/hr

Page: 47. of 64 NEDC:06-035 Rev. Number: 1 The difference from feedwater enthalpy to steam enthalpy:

Ah = 1191.55 Btu/lbm - 338.73 Btu/lbm = 852.82 Btu/lbm The error in power due to a flow error:

Error (MW) = (474,450 lbm/hr)( 852.82 Btu/lbm)(1 hr/60 min)

(17.5796 Watts/BTU/min)(1MW/iE6 Watts)

=+/-118.5511 MW As percent Reactor Power:

+/-(118.5511 MW/2381 MW) X 100 =4.9790%

+/-4.9790%Pwr Error/5% Flow Error = 0.9958 So error in Thermal Power per error in Flow:

WFFWF = 0.9958 %Pwr Error/% Flow Error Using 22.6681% DP span from Section 3.1.1.4 the error for a 5% span becomes:

+5% = 27.6681% and -5% = 17.6681%

In terms of flow:

W+5 %= 100(27.6681%/100)°- = 52.6005%

W 5 %= 100(17.6681%/100)°" = 42.0334%

The 100% flow value per loop as percent span is:

(4.745 Mlbm/ 10 Mlbm) X 100 = 47.45%

52.6005% - 47.45% = +5.1505%

42.0334% - 47.45% = -5.4166%

Using the higher value for conservatism:

Error Flow (ibm/hr) = 0.054166(10.0 Mlbm/hr) = 541.660 Ibm/hr

Page: 48 of 64 NEDC:06-035 Rev. Number: 1 The individual loops are summed, so:

Total Flow Error = (541,6602 + 541,6602)0.5 = 766,022.92 ibm/hr Error (MW) = (766,022.92 lbm/hr)( 852.82 Btu!Ibm)(1 hr/60 min)

(17.5796 Watts/BTU/min)(1MW/I E6 Watts)

Error(MW) 191.4066 In Percent:

(191.4066 MW/2381 MW) X 100 = 8.0389%

So the error in Thermal Power per error in DP:

8.0389%/5% = 1.6078 %Power/%DP Span WFFWDP = 1.6078 %Pwr Error/% Span 3.9.2. FW Temperature Sensitivity Per Section 3.2.2.1 the FW Temperature has a span of 150 *F so a +/- 5% error will result in +/- 7.5

and using 100% baseline temperature of 364.68 °F and baseline pressure of 1165 psia:

Hf-7.5 (357.18 'F, 1165 psia) 330.877 Btu/lbm Hf (364.68 *F, 1165 psia) = 338.737 Btu/lbm Hf+7.5 (372.18 °F, 1165 psia) 346.621 Btu/lbm 330.877 Btu/lbm - 338.737 Btu/lbm = -7.860 Btu/lbm 346.621 Btu/lbm - 338.737 Btullbm = +7.884 Btu/lbm Pwr Error (MW) = (9.489E6 lbm/lhr)( 7.884 Btu/lbm)(1 hr/60 min)

(17.5796 Watts/BTU/min)(1MW/IE6 Watts)

Pwr Error (MW) = 21.9192 MW

% Pwr Error = (21.9192 MW/2381 MW) X 100 = 0.9206%

%Power/Temp Error = 0.9206%/7.5 *F = 0.1227 %Power/*F WFFWr = 0.1227 %Pwr Error/°F

Page: 49 of 64 NEDC:06-035 Rev. Number: 1 3.9.3. Reactor Pressure Sensitivity Reactor pressure is used to determine steam enthalpy and is used to determine the energy added to the CRD flow and Feedwater flow. So CRD and Feedwater flows are added together to determine the error in Reactor Thermal Power per error in Steam Pressure.

The instrument spans is 0 to 1200 psig so +/-5% is 60 psi.

Vessel Dome Pressure = 1019.99 psia - 1020 psia Hg5%(Saturation, 960 psia) = 1194.4 Btu/Ibm Hg (Saturation, 1020 psia) = 1192.2 Btu/lbm Hg+5% (Sturation, 1080 psia) = 1189.9 Btu/lbm 1194.4 Btu/lbm - 1192.2 Btu/lbm = +2.2 Btu/lbm 1189.9 Btu/lbm - 1192.2 Btu/lbm = -2.3 Btu/lbm 3 3 3 3 CRD Flow = (50 gal/m)(60m/hr)(62.22 lbm/ft )(233 in /gal)(ft /1728 in )

-25169 lbm/hr Pwr Error (MW) = (9.489E6 +25169 lbmihr)(2.3 Btu/lbm)(1 hr/60 min)

(17.5796 WattsiBTU/min)(1MW/iE6 Watts)

= 6.4115 MW (6.4115 MW/2381 MW) X 100 = 0.2693 %Pwr Error 0.2693 %Pwr Error/5% Press. Span Error = 0.0539 WFRp = 0.0539 %Pwr Error/% Span Error 3.9.4. CRD Flow Sensitivity CRD flow uncertainty is expressed in % Actual Flow and % DP span so two weighting factors are developed.

(50 gpm)(0.05) = 2.5 gpm CRD Flow = (2.5 gal/m)(60m/hr)(62.22 lbrn/ft 3)(233 in3/gal)(ft3/1728 in 3)

= 1258.443 lbm/hr Ah =1191.55 -75.105 = 1116.445

Page: 50 of 64 NEDC:06-035 Rev. Number: 1 Pwr Error (MW) = (1258.433 lbmihr)(1 116.455 Btu/lbm)(1 hr/60 min)

(17.5796 Watts/BTU/min)(1MW/1E6 Watts)

= 0.4117 MW (0.4117 MW/2381 MW) X 100 = 0.01729 %Pwr Error 0.01729 %Pwr Error/5% Flow Error = 0.0035 WFcRDF = 0.0035 %Pwr Error/% Flow Error Per Section 3.4.1.4, 50 gpm is at 9.7757% DP Span.

So +/-5% = 4.7757% and 14.7757%

Taking the square root(converting to %flow): 21.853% and 38.439%. The difference with nominal percent flow of 50 gpm/160 gpm X 100 = 31.25%.

21.853 - 31.25% = -9.397%

38.439% - 31.25% = +7.189%

(160 gpm)(0.09397) = 15.0352 gpm CRD Flow = (15.0352 gal/m)(60m/hr)(62.22 Ibm/ft3)(233 in3/gal)(ft 3/1728 in3)

= 7568.3751 ibm/hr Pwr Error (MW) = (7568.3751 lbm/hr)(1 116.445 Btu/lbm)(1 hr/60 min)

(17.5796 Watts/BTU/min)(1MW/iE6 Watts)

= 2.4757 MW (02.4757 MW/2381 MW) X 100 = 0.1040 %Pwr Error 0.1040 %Pwr Error/5% Press. Span Error = 0.0208 WFCRDDP = 0.0208 %Pwr Error/% Span Error

Page: 51 of 64 NEDC:06-035 Rev. Number: 1 3.9.5. CRD Temperature Sensitivity It has therefore been assumed in Section 3.5 that CRD temperature when at or near 100%

power is 102.83 'F with a +/-10 "F band added for conservatism. Therefore, +/-10 *Fband will be used in determining the CRD Temperature Weighting Factor.

Hf-1o (92.83 'F, 1465 psia) = 64.743 Btu/lbm Hf (102.83 °F, 1465 psia) = 74.664 Btu/lbm Hf+1o (112.83 *F, 1465 psia) = 84.589 Btu/lbm 64.743 Btu/lbm - 74.664 Btu/lbm = -9.921 Btu/lbm 84.5 89 Btu/lbm - 74.664 Btu/lbm = +9.925 Btu/lbm CRD Flow = (50 gal/m)(60m/hr)(62.22 lbm/ft3 )(233 in 3/gal)(ft3/1728 in 3)

= 25,168.854 Ibm/hr Pwr Error (MW) = (25,168.854 lbmrhr)( 9.925 Btu/lbm)(1 hr/60 min)

(17.5796 Watts/BTU/min)(1MW/iE6 Watts)

Pwr Error (MW) = 0.0732 MW

% Pwr Error = (0.0732 MW/2381 MW) X 100 = 0.0031%

%Power/Temp Error = 0.0031%/10 °F = 0.0003 %Power/*F WFCRDT = 0.0003 %Pwr Error/*F 3.9.6. RR Watt Meter Sensitivity The Watt Meter indication is 0 to 6 MW. Since this is one to one correspondence to megawatts, the weighting factor can conservative be taken as the fraction of 100%

Reactor Power the watt meter power can represent.

6 MW/2381MW =0.0025 WFwM = 0.0025 %Pwr Error/% Span Error

Page: 52 of 64 NEDC:06-035 Rev. Number: 1 3.9.7. RWCU Flow Sensitivity RWCU flow uncertainty is expressed in % Actual Flow and % DP span so two weighting factors are developed. Per Reference 5.1.33 the temperature out of the non-regenerative Heat exchanger is 120 *F.

(202 gpm)(0.05) = 10.1 gpm 3 3 3 3 RWCU Flow = (10.1 gal/m)(60m/hr)(61.921bm/ft )(233 in /gal)(ft /1728 in )

= 5,059.60 Ibm/hr Hin (533 *F, 1193 psia) = 527.606 Btu/lbm Hour (435 "F, 1075 psia) = 414.015 Btu/lbm Ah = 527.606 Btu/lbm - 414.015 Btu/lbm = 113.591 Pwr Error (MW) = (5,059.60 lbm/hr)(1 13.591 Btu/lbm)(1 hr/60 min)

(17.5796 Watts/BTU/min)(1MW/1E6 Watts)

= 0.1684 MW (0.1684 MW/2381 MW) X 100 = 0.0071 %Pwr Error 0.0071 %Pwr Error/5% Flow Error = 0.0014 WFcUr = 0.0014 %Pwr Error/% Flow Error Per Section 3.7.1.4 the normal flow of 101 gpm is 70.87% DP Span.

So +/-5 % = 65.87% and 75.87%

Taking the square root(converting to %flow): 81.160% and 87.103%. The loop has 120 gpm span. So the percent of span the nominal percent flow is:

101 gpm/ 120 gpm X 100 = 84.167%.

And the difference becomes:

81.160% - 84.167% = -3.007%

87.103% - 84.167% = +2.936%

Page: 53 of 64 NEDC:06-035 Rev. Number: 1 RWCU Flow = (120 gal/m)(60m/hr)(61.921bm/ft 3)(233 in 3/gal)(ft3/1728 in3)

= 60,114 lbm/hr (60,114 lbm/hr)(0.03007) = 1807.628 ibm/hr The individual loops are combined in the process computer:

((1807.628)2 + (1807.628)2)0.5 = 2556.372 Pwr Error (MW) = (2556.372 Ibm/hr)(1 13.841 Btu/lbm)(1 hr/60 min)

(17.5796 Watts/BTU/min)(1 MW/i E6 Watts)

= 0.0853 MW (0.0853 MW/2381 MW) X 100 = 0.0036 %Pwr Error 0.0036 %Pwr Error/5% Press. Span Error = 0.0007 WFCuDP = 0.0007 %Pwr Error/% Span Error 3.9.8. RWCU Temperature Sensitivity The inlet and outlet temperature (532 *F and 435 *F, respectively) are evaluated at the baseline conditions independent of each other. The temperature span of the instruments is 0 to 600 *F per Section 3.7.1. and 5% would be 30 0F which is too large considering the total inlet and outlet loop uncertainties are +/- 4.0031 °F and +/- 3.2693*F, respectively, (Section 3.7.3), so a +/-5 'F will be used instead.

H. 5% (527 'F, 1193 psia) = 520.183 Btu/lbm H (532 °F, 1193 psia) = 526.357 Btu/Ibm H5st (537 -F, 1193 psia) = 532.603 Btu/lbm 520.183 Btu/lbm - 526.357 Btu/lbm =- 6.174 Btu/Ibm 532.603 Btu/lbm - 526.357 Btu/lbm-= +6.246 Btu/lbm RWCU Flow = (202 gal/m)(60m/hr)(61.92ft3)(233 in 3/gal)(fl3 /1728 in 3)

= 101,191.9 lbmlhr

Page: 54 of 64 NEDC:06-035 Rev. Number: I1 Pwr Error (MW) = (101,191.9 lbmrhr)(6.246 Btu/lbm)(1 hr/60 min)

(17.5796 Watts/BTU/min)(1MW/1E6 Watts)

= 0.1852MW (0.1852 MW/2381 MW) X 100 = 0.0078%Pwr Error 0.0078 %Pwr Error/5"F = 0.0016 WFcutrr = 0.0016 %Pwr Error/* F H-5 % (430 TF, 1075 psia) = 408.520 Btu/lbm H (435 TF, 1075 psia) = 414.015 Btu/lbm H+o5%t(440 TF, 1075 psia) = 419.5 10 Btu/lbm 408.520 Btu/lbm - 414.015 Btu/lbm = -5.495 Btu/Ibm 419.510 Btu/lbm - 414.015 Btu/lbm = +5.495 Btu/lbm Pwr Error (MW) = (101,191.9 lbm/hr)(5.495 Btu/lbm)(1 hr/60 min)

(17.5796 Watts/BTU/min)(IMW/1E6 Watts)

= 0.1629 MW (0.1629 MW/2381 MW) X 100 = 0.0068%Pwr Error 0.0001 %Pwr Error/5*F = 0.0014 WFcUr2 = 0.0014 %Pwr Error/*F 3.9.9. Radiative Power Sensitivity The weighting factor for radiative power losses is the ratio of percent power to megawatts thermal power.

100%/2381 MW = 0.0420 WFRP = 0.0420%/MW

Page: 55 of 64 NEDC:06-035 Rev. Number: 1 3.10. Total Reactor Thermal Power Calculation Uncertainty The reactor thermal power calculation uncertainty is determined for two cases. The first case is for feedwater flow measurements using the installed ASME flow nozzles and the second case is for feedwater flow measurements using the Caldon Ultrasonic Flow Meters. In both cases the loop uncertainties are multiplied by the respective weighting factor and then the individual contributions to reactor thermal power is combined using SRSS. Bias errors are added to the total SRSS.

Case I (ASME Flow Nozzle)

Parameter Uncertainty Sensitivity Contribution FW Flow +/-0.7071 0.9958 +/-0.7041 FW D/P +/-0.6756 1.6078 +/-1.0863 FW Dependent +/-0.0160 1.6078 -+/-0.0257 FW Temperature +/-0.4741 0.1227 +/-0.0582 Reactor Pressure +/-0.5560 0.0539 +/-0.0300 CRD Flow +/-1.4142 0.0035 +/-0.0049 CRD D/P +/-2.7562 0.0208 +/-0.0573 CRD Temp. +/-10.000 0.0003 +/-0.0030 RR Pump Power +/-5.0123 0.0025 +/-0.0125 RWCU Flow +/-2.2429 0.0014 +/-0.0031 RWCU D/P +/-1.0917 0.0007 +/-0.0008 RWCU Inlet Temp. +/-3.9957 0.0016 +/-0.0064 RWCU Outlet +/-3.2695 0.0014 +/-0.0046 Temp.

Radiative Losses +/-0.3000 0.0420 +/-0.0126 Total SRSS +/-1.2979 FW Nozzle Bias +0.6000 0.9958 +0.5975 CRD Bias -0.1806 0.0035 -0.0006 Total Uncertainty -1.2985%/+1.8954%

Page: 56 of 64 NEDC:06-035 Rev. Number: 1 Case 2 (Ultrasonic Flow Meter)

Parameter Uncertainty Sensitivity Contribution FW Ultrasonic +/-0.3000 0.9958 10.2987 Flow FW Temperature 4-0.4741 0.1227 +/-0.0582 Reactor Pressure +/-0.5560 0.0539 A0.0300 CRD Flow +/-1.4142 0.0035 +/-0.0049 CRD D/P +/-2.7562 0.0208 +/-0.0573 CRD Temp. +/-10.000 0.0003 +/-0.0030 RR Pump Power +/-5.0123 0.0025 +/-0.0125 RWCU Flow +/-2.2429 0.0014 +/-0.0031 RWCU D/P ++/-1.0917 0.0007 +/-0.0008 RWCU Inlet Temp. +/-3.9957 0.0016 +/-0.0064 RWCU Outlet L3.2695 0.0014 +/-0.0046 Temp.

Radiative Losses +/-0.3000 0.0420 +/-0.0126 Total SRSS +/-0.3118 CRD Bias -0.1806 0.0035 -0.0006 Total Uncertainty 1 -0.3124%/+0.3118%

4. CONCLUSION:

The total uncertainty associated with the HEABAL (Reference 5.5.1) when feedwater flow is measured with the ASME Flow nozzles is -1.2985%/+1.8954%.

The total uncertainty with the HEABAL (Reference 5.5.1) when feedwater flow is measured with the Caldon UFM is -0.3124%/+0.3118%.

Page: 57 of 64 NEDC:06-035 Rev. Number: 1

5.

REFERENCES:

5.1. Drawings

5.1.1. Retention Number 454003582, Drawing 2004 Sheet 1, Rev. N30.

5.1.2. Retention Number 454003583, Drawing 2004 Sheet 2, Rev. N46.

5.1.3. Retention Number 454003584, Drawing 2004 Sheet 3, Rev. N46.

5.1.4. Retention Number 454003616, Drawing 2026 Sheet 1, Rev. N64.

5.1.5. Retention Number 454003632, Drawing 2039, Rev. N54.

5.1.6. Retention Number 454003635, Drawing 2042, Sheet 1, Rev. N32.

5.1.7. Retention Number 454003683, Drawing 2042, Sheet 2, Rev. N13.

5.1.8. Retention Number 454003684, Drawing 2042, Sheet 3, Rev. N20.

5.1.9. Retention Number 454003678, Drawing 2049, Sheet 4, Rev. N1 3.

5.1.10. Retention Number 454012990, Drawing 2608-9, Rev. 9.

5.1.11. Retention Number 45401277 1, Drawing 2608-10, Rev. 7.

5.1.12. Retention Number 454012666, Drawing 2849-4, Rev. N 1.

5.1.13. Retention Number 452005574, Drawing 556-26811, Rev. 4.

5.1.14. Retention Number 454003883, Drawing 3010, Sheet 1, Rev. N67.

5.1.15. Retention Number 454003926, Drawing 3043, Sheet 12, Rev. NI16.

5.1.16. Retention Number 454004197, Drawing 3254, Sheet 11, Rev. NI .

5.1.17. Retention Number 454004198, Drawing 3254, Sheet 12, Rev. N 18.

5.1.18. Retention Number 454004199, Drawing 3254, Sheet 13, Rev. N22.

5.1.19. Retention Number 45400420 1, Drawing 3254, Sheet 15, Rev. N05.

Page: 58 of 64 NEDC:06-035 Rev. Number: 1 5.1.20. Retention Number 454004222, Drawing 3255, Sheet 8, Rev. N31.

5.1.21. Retention Number 45004248, Drawing 3255, Sheet 34, Rev. N23.

5.1.22. Retention Number 454004250, Drawing 3255, Sheet 36, Rev. N18.

5.1.23. Retention Number 454004333, Drawing 3257, Sheet 13, Rev. N07.

5.1.24. Retention Number 454004349, Drawing 3257, Sheet 29, Rev. NI 6.

5.1.25. Retention Number 454004351, Drawing 3257, Sheet 31, Rev. N17.

5.1.26. Retention Number 454240980, Drawing 3257, Sheet 89B, Rev. NOO.

5.1.27. Retention Number 454240981, Drawing 3257, Sheet 89C, Rev. NOO 5.1.28. Retention Number 454240990, Drawing 3257, Sheet 90D, Rev. NOO.

5.1.29. Retention Number 454241010, Drawing 3257, Sheet 92F, Rev. NO0.

5.1.30. Retention Number 452209416, Drawing 117C3485, Sheet 1, Rev. 19.

5.1.31. Retention Number 452209417, Drawing 11 7C3485, Sheet 2, Rev. 7.

5.1.32. Retention Number 4540006910, Drawing 158B7077, Rev. 2.

5.1.33. Retention Number 454005663, Drawing 730E148BB, Sheet 1, Rev.1 5.1.34. Retention Number 454005392, Drawing 730E197BB, Sheet 5, Rev. N06.

5.1.35. Retention Number 454006837, Drawing 791E252, Sheet 1, Rev. N10.

5.1.36. Retention Number 454006497, Drawing 791E254, Sheet 1,Rev. N08.

5.1.37. Retention Number 454006804, Drawing 791E257, Sheet 2, Rev. N12.

5.1.38. Retention Number 454006806, Drawing 791 E257, Sheet 4, Rev. N26.

5.1.39. Retention Number 454006668, Drawing 791E263, Sheet 2, Rev. N17.

Page: 59 of 64 NEDC:06-035 Rev. Number: 1 5.1.40. Retention Number 452006647, Drawing 791E446 Sheet 1, Rev. N22.

5.1.41. Retention Number 452006747, Drawing 791E540, Rev. N19 5.1.42. Retention Number 454006752, Drawing 791E523, Sheet 2, Rev. N18 5.1.43. Retention Number 454006290, Drawing 0199F0377, Rev. N12.

5.1.44. Retention Number 454006962, Drawing 0199F0380, Rev. N12.

5.1.45. Retention Number 452001513, Drawing E507, Sheet 115, Rev. N07.

5.1.46. Retention Number 450213825, Drawing E515 Sheet 9, Rev. NO0.

5.1.47. Retention Number 450218224, Drawing E515 Sheet 55, Rev. N02.

5.1.48. Retention Number 450218225, Drawing E515 Sheet 60, Rev. NOI.

5.1.49. Retention Number 450213874, Drawing E515, Sheet 82, Rev. NO0.

5.1.50. Retention Number 450213879, Drawing E515, Sheet 88, Rev. NO0.

5.1.51. Retention Number 450218236, Drawing E515 Sheet 121, Rev. N01.

5.1.52. Retention Number 450213972, Drawing E515, Sheet 144, Rev. NO0.

5.1.53. Retention Number 453225711, Drawing E507, Sheet 228, Rev.N07.

5.1.54. Retention Number 454015176, Drawing CP008, Sheet 1, Rev. N04.

5.1.55. Retention Number 450245488, Drawing 10741-980500, Sheet 1.

5.1.56. Retention Number 450245489, Drawing 10741-980500, Sheet 2.

5.1.57. Retention Number 451245490, Drawing 10741-980500, Sheet 3.

5.1.58. Retention Number 451245491, Drawing 10741-980500, Sheet 4.

5.1.59. Retention Number 451245492, Drawing 10741-980500, Sheet 5.

5.1.60. Retention Number 451245493, Drawing 10741-980500, Sheet 6.

Page: 60 of 64 NEDC:06-035 Rev. Number: 1 5.1.61. Retention Number 451245494, Drawing 10741-980500, Sheet 7.

5.1.62. Retention Number 451245495, Drawing 10741-980500, Sheet 8.

5.1.63. Retention Number 451245496, Drawing 10741-980500, Sheet 9.

5.1.64. Retention Number 451245497, Drawing 10741-980500, Sheet 10.

5.2. Instrument Data Sheets:

5.2.1. 14.5.1 Instrument Calibration Data Sheet, RF-TT-168A.

5.2.2. 14.5.1 Instrument Calibration Data Sheet, RF-TT-168B.

5.2.3. 14.5.1 Instrument Calibration Data Sheet, RF-TT-168C.

5.2.4. 14.5.1 Instrument Calibration Data Sheet, RF-TT-168D.

5.2.5. 14.15.2 Instrument Calibration Data Sheet, RFC-LOOP-8.

5.2.6. 14.15.2 Instrument Calibration Data Sheet, RFC-LOOP-9 5.2.7. RWCU Generic Calibration Data, RWCU-LOOP-6 5.2.8. CRD Generic Calibration Data Sheet, CRD-LOOP-6.

5.2.9. 14.5.1 Instrument Calibration Data Sheet, RWCU-LOOP-3 5.2.10. 14.5.1 Instrument Calibration Data Sheet, RWCU-LOOP-4

5.3. Procedures

5.3.1. Procedure 3.26.3, "Instrument Setpoint and Channel Error Calculation Methodology",

Revision 6.

5.3.2. Procedurel4.NBI.301, "Reactor Pressure Channel Calibration", Revision 4.

5.3.3. Procedure 2.2.8, "Control Rod Hydraulic System", Revision 68.

Page: 61 of 64 NEDC:06-035 Rev. Number: 1 5.3.4. Procedure 15.RR.301, "RRMG Calibration of MG Output Power Transducer and Wattmeter", Revision 3.

5.3.5. Procedure 2.2.66, "Reactor Water Cleanup", Revision 87.

5.3.6. Procedure 2.2.6, "Condensate System", Revision 64.

5.3.7. Procedure 2.2.68, "Reactor Recirculation System", Revision 65.

5.3.8. Procedure 4.6.1, "Reactor Vessel Water Level Indication", Revision 28.

5.4. Codes and Standards:

5.4.1. ASME, "Fluid Meters Their Theory and Application" Sixth Edition, 1971.

5.4.2. ISA-RP67.04 - Part II - 1994, "Methodologies for the Determination of Setpoints For Nuclear Safety-Related Instrumentation", September 30, 1994.

5.4.3. ASME Steam Tables, 1967.

5.4.4. ANSI MC96.1-1975, "Temperature Measurement Thermocouples."

5.5. Vendor Documentation:

5.5.1. Studsvik Scandpower Report: SSP-04/414-C, "GARDEL-BWR for Cooper Nuclear Station Heat Balance Method Description", Revision 0.

5.5.2. Rosemount Reference Manual 00809-0100-4801, "Model 3051S Series Pressure Transmitter Family", Revision AA.

5.5.3. RTP Technical Bulletin, RTP7436 Series.

5.5.4. RTP Technical Manual 981-0021-211 A, "RTP8436 Series Universal Analog Input Card Set", Revision A.

5.5.5. KEPCO Technical Bulletin 146-1869.

5.5.6. Rosemount Temperature Sensor Product Data Sheet, 00813-0100-2654, Revision FA.

5.5.7. Rosemount 3144P Temperature Transmitter Product Data Sheet, 00813-0100-4021, Revision FA.

Page: 62 of 64 NEDC:06-035 Rev. Number: 1 5.5.8. Rosemount, Report of Calibration, Model Option Code X9Q4, 1/5/2006.

5.5.9. Rosemount Technical Manual 00809-0100-4360, "Model 1151 Alphaline Pressure Transmitters" Rev. AA.

5.5.10. Badger Meter Inc., "Differential Meter Flow vs Differential Calculations".

5.5.11. GE Instructions, 198 4532K30-010, "Type 570-06, 07 Isolated Power Supply.

5.5.12. Scientific Columbus Technical Bulletin, "Exceltronic AC Watt or Var Transducers".

5.5.13. Daniel Technical Bulletin, "Orifice Plates and Plate Sealing Units".

5.5.14. Barton Product Bulletin G1-25, "Models 273A and 274A Pneumatic Transmitters."

5.5.15. Part Evaluation Technical Evaluation Number 10511159, Revision 0, 5.5.16. Caldon Report ML 162, "Caldon Experience in Nuclear Feedwater Flow Measurement,"

Revision 2.

5.5.17. Caldon Report FR04.PM5. "Feedwater Assessment Program Report for Nebraska Public Power District Cooper Nuclear Station", Revision 0.

5.6. Miscellaneous 5.6.1. Calculation NEVC 70-263, Rev. 0 (Roll 00111, Frame 0033).

5.6.2. GE Purchase Specification 21A1379AR, "Cooper Flow Element Data Sheets", Revision 4 (Roll 09021, Frame 1898).

5.6.3. Alden Research Laboratories, "Calibration Two Flow Nozzles Serial Numbers T- 12125 and T-12126", July 1970 (Roll 08118, Frame 1295).

5.6.4. GE FDI Number 71/10100, " Feedwater Flow Element", Revision 1 (Roll 17524, Frame 0604).

5.6.5. CED6010820 5.6.6. Calculation NEDC 00-95A, Revision 4 5.6.7. Calculation NEDC 94-018, Revision 3,

Page: 63 of 64 NEDC:06-035 Rev. Number: 1 5.6.8. CED 2000-0032 5.6.9. Contract E69-4, Piping Specifications.

5.6.10. GE DRF A13-00461-02, "Impact of Steam Table Basis on Process Computer Heat Balance Calculations," Revision 1.

5.6.11. Cameron Engineering Report ER-592, "Bounding Uncertainty Analysis For Thermal Power Determination At Cooper NPPD Using LEFM\1 + System," Revision 1.

6. ATTACHMENTS:

6.1. Rosemount Reference Manual 00809-0100-4801, "Model 3051S Series Pressure Transmitter Family", Revision AA (Excerpt).

6.2. RTP Technical Bulletin, RTP7436 Series.

6.3. RTP Technical Manual 981-0021-211 A, "RTP8436 Series Universal Analog Input Card Set",

Revision A (Sections 1 & 2).

6.4. KEPCO Technical Bulletin 146-1869.

6.5. Rosemount 3,144P Temperature Transmitter Product Data Sheet, 00813-0100-4021, Revision FA.

6.6. Rosemount, Report of Calibration, Model Option Code X9Q4, 1/5/2006.

6.7. Rosemount Technical Manual 00809-0100-4360, "Model 1151 Alphaline Pressure Transmitters" Rev. AA (Excerpt).

6.8. Badger Meter Inc., "Differential Meter Flow vs Differential Calculations".

6.9. GE Instructions, 198 4532K30-010, "Type 570-06, 07 Isolated Power Supply.

6.10. Scientific Columbus Technical Bulletin, "Exceltronic AC Watt or Var Transducers".

6.11. Daniel Technical Bulletin, "Orifice Plates and Plate Sealing Units".

6.12. Barton Product Bulletin G 1-25, "Models 273A and 274A Pneumatic Transmitters."

Page: 64 of 64 NEDC:06-035 Rev. Number: 1 6.13. Fisher & Porter Instruction Bulletin 50EW1020, "Series 50EW1020 & 50EW1030 Pneumatic-To-Current Converters," Revision 3.

6.14. Caldon Report ML162, "Caldon Experience in Nuclear Feedwater Flow Measurement,"

Revision 2.

6.15. Cameron Engineering Report ER-592, "Bounding Uncertainty Analysis For Thermal Power Determination At Cooper NPPD Using LEFM4 + System," Revision 1.

v t e Letter Sequence Response to RAI
TAC:MD7385, Clarify Application of Setpoint Methodology for LSSS Functions (Approved, Closed)
Initiation Acceptance... Administration Questions 2 May 2008 Answers 6 March 2008 12 March 2008 4 April 2008 9 May 2008 Results Approval Other: ML073300571, ML081540285, ML082690556, NLS2008083, Revision to Commitment Referenced in Amendment 231 Safety Evaluation
MONTHYEAR2008-05-02 [Table View]

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1.51% Power c) RBM Power Channel Drift DRBMPwr

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5.3. Procedures

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NLS2008034, Response to Request for Additional Information for License Amendment Request to Revise Technical Specifications - Appendix K Measurement Uncertainty Recapture Power Uprate

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REFERENCES:

5.1. Drawings

5.3. Procedures

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