L-MT-08-049, Submittal of Response to Requests for Additional Information for License Amendment Request for Power Range Neutron Monitoring System Upgrade

Submittal of Response to Requests for Additional Information for License Amendment Request for Power Range Neutron Monitoring System Upgrade

ML082620582
Person / Time
Site:	Monticello
Issue date:	09/16/2008
From:	O'Connor T Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML082620581	List: L-MT-08-049, Submittal of Response to Requests for Additional Information for License Amendment Request for Power Range Neutron Monitoring System Upgrade
References
L-MT-08-049, TAC MD8064
Download: ML082620582 (165)

Text

Monticello Nuclear Generatina - Plant Operated by Nuclear Management Company, LLC Committed to Nuclear Excellence WITHOLD ENCLOSURES 8 AND 9 FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 September 16,2008 L-MT-08-049 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Monticello Nuclear Generating Plant Docket 50-263 Renewed Facility Operating License No. DPR-22 Response to Requests for Additional Information for License Amendment Request for Power Range Neutron Monitoring Svstem Upgrade (TAC No. MD8064)

On February 6,2008, the Nuclear Management Company (NMC), LLC submitted a request to revise the Monticello Nuclear Generating Plant Technical Specifications (TS) in conjunction with the installation of the General Electric - Hitachi (GEH) Nuclear Measurement Analysis and Control (NUMAC) Power Range Neutron Monitoring (PRNM) System (Enclosure 1, Reference 1).

Additional information was requested by the U.S. Nuclear Regulatory Commission (NRC) on the basis for this proposed change by three e-mails, dated June 13, 2008 (Enclosure 1, Reference 2), June 25, 2008 (Enclosure 2, Reference 2), and July 1, 2008 (Enclosure 3, Reference 2). Responses to these NRC e-mail requests for additional information (RAI) are provided in Enclosures 1, 2 and 3, respectively. During a telephone discussion with the NRC Project Manager on August 21, 2008, it was indicated that additional information from GEH was required to respond to several of the RAls, and that a delay in submittal to September 12, 2008, would be acceptable.

Enclosure 4 provides a revised copy of the pertinent pages of the Monticello Nuclear Generating Plant (MNGP) Technical Specifications (TS) including the notes suggested by Regulatory Issue Summary (RIS) 2006-17 (Enclosure I , Reference 7) for those Limiting Safety System Settings (LSSS) that protect a safety limit in accordance with 10 CFR 50.36(c)(l)(ii)(A). Enclosure 5 provides revised TS Bases pages clarifying the functions which are LSSS that protect a safety limit and also presents a more in depth discussion of the staggered test basis for response time testing.

Enclosure 6 provides a sample calculation for several Average Power Range Monitor non-flow biased PRNM System setpoints illustrating the MNGP application of the General Electric Instrument Setpoint Methodology. Several of the responses to the RAls within Enclosures 1, 2 and 3, respectively, have a separate version of the response which contains proprietary information as defined by 10 CFR 2.390 that was 2807 West County Road 75 Monticello, Minnesota 55362-9637 Telephone: 763.295.5151 Fax: 763.295.1454

Document Control Desk L-MT-08-049 provided by General Electric - Hitachi (GEH). The proprietary RAI responses have been assembled into one proprietary enclosure, Enclosure 8. Enclosure 9 provides a GEH proprietary response clarifying the OPRM Upscale function licensing basis in response to RAI No. 2 of Enclosure 1.

GEH, as the owner of this proprietary information, has executed two affidavits provided in Enclosure 7, which identifies that the enclosed information has been handled and classified as proprietary, is customarily held in confidence, and has been withheld from public disclosure. The proprietary information contained in Enclosures 8 and 9 was provided to the MNGP in a GEH transmittal referenced by the affidavit. The proprietary information has been faithfully reproduced within the RAI responses such that the affidavit remains applicable. GEH requests that the enclosed proprietary information be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390 and 10 CFR 9.1 7. A non-proprietary version of the RAI responses containing proprietary information is provided within Enclosures 1, 2 and 3, respectively.

The NMC requests that the NRC safety evaluation for this license amendment clearly delineate those TS functions discussed herein determined not to be an LSSS that protects a safety limit, in addition to specifying those determined to be an LSSS that protects a safety limit pursuant to 10 CFR 50.36(c)(l)(ii)(A). This could reduce NRC Staff and NMC resource burdens during final resolution of the LSSS setpoint issue.

The NMC has reviewed the No Significant Hazards Consideration and the Environmental Consideration determinations provided in the February 6, 2008, license amendment request relative to the supplemental information being provided herein and has determined that no changes are required to either determination.

The MNGP Plant Operations Review Committee has reviewed these RAI responses and enclosed revised TS pages. In accordance with 10 CFR 50.91, a copy, without the eing provided to the designated Minnesota Official.

f perjury that the foregoing is true and correct. Executed

, Monticello Nuclear Generating Plant Y

Nucle hlanagement Company, LLC Enclosures cc: Administrator, Region Ill, USNRC (wlo Enclosures 8 and 9)

Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC (wlo Enclosures 8 and 9)

Minnesota Department of Commerce (wlo Enclosures 8 and 9)

Page 2 of 2

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS On February 6,2008, (Reference 1) the Nuclear Management Company, LLC (NMC) submitted a request to revise the Monticello Nuclear Generating Plant (MNGP)

Technical Specifications (TS) in conjunction with the installation of the General Electric

- Hitachi (GEH) Nuclear Measurement Analysis and Control (NUMAC) Power Range Neutron Monitoring (PRNM) System.

The following requests for additional information (RAI) concerning setpoints and setpoint methodology were received from the U.S. Nuclear Regulatory Commission (NRC) by e-mail, dated June 13, 2008 (Reference 2).

1. Setpoint Calculation Methodoloav: Provide documentation (including sample calculations) of the methodology used for establishing the limiting setpoint (or NSP) and the limiting acceptable values for the As-Found and As-Left setpoints as measured in periodic surveillance testing as described below. Indicate the related Analytical Limits and other limiting design values (and the sources of these values) for each setpoint.

Response

As discussed within Section 5.4 of Enclosure 1 to the PRNM System license amendment request (LAR), the MNGP has adopted and incorporated into the site Engineering Standards Manual (ESM) (Reference 3), the MNGP specific implementation of the General Electric - Hitachi (GEH) Instrument Setpoint Methodology (ISM) (References 4 and 5). The ESM provides plant-specific guidance on implementation of the GEH instrument setpoint guidelines and methodology. The GEH ISM has been reviewed and approved by the NRC for use by utilities as a basis for their instrument setpoint programs as discussed within the associated NRC safety evaluation for the methodology (Reference 6).

The MNGP specific implementation of the GEH ISM was applied in the determination of the setpoints for the various TS functions discussed herein.

Conceptually, the GEH method is based on ISA Standard 67.04, Method 2 but leads to more conservative setpoints. According to this approved methodology, the setpoints are calculated from the Analytic Limit (AL) using a top down approach, and the margin is calculated by ISM between the AL and the Allowable Value (AV), and between AV and the Nominal Trip Setpoint (NTSP).

The AL is a process parameter value used in the safety analysis. The AL represents a limiting value for the automatic initiation of protective actions. From the AL an AV is first calculated which, has margin to the AL, based on all measurement errors except drift. This ALIAV margin includes the Process Measurement Accuracy (PMA), Primary Element Accuracy (PEA), measuring instrument loop accuracy under trip conditions (AT),and the instrument calibration errors (C). The calibration uncertainty in the GEH ISM contains the Page 1 of 29

ENCLOSURE 1 RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS As-Left Tolerance (ALT), so the AV is already made more conservative to account for this allowance. All random errors are combined using Square Root of the Sum of the Squares (SRSS) method, and non-conservative bias errors are added algebraically. The AV represents the limiting value to which a setpoint can drift (as determined from surveillance testing) and still assure that the AL is protected. The approved GEH ISM provides a sufficient margin between the AL and AV to assure with at least 95 percent probability that the AL will not be exceeded if the setpoint has drifted to the AV. The AV is the value specified in the TSs.

The AVINTSP margin includes instrument loop accuracy under calibration conditions (Ac), instrument calibration errors (C) and instrument drift errors (D).

The approved GEH ISM basically calculates two nominal trip setpoints. The first is the setpoint with minimum required margin to the AL based on 95 percent probability of not exceeding the AL. This setpoint is called NTSPI and the AUNTSPI margin is based on all errors (PMA, PEA, AT, C, and Drift (D)).

Therefore NTSPI is equivalent to the Limiting Trip Setpoint (LSP) referred to in RIS 2006-17 (Reference 7). However, the GEH ISM also calculates a second nominal trip setpoint, referred to as NTSP2, with additional margin to provide high confidence that the setpoint will not drift beyond the AV potentially resulting in a Licensee Event Report (LER). According to the approved GEH ISM, the final NTSP has margin to the AV which provides 90 percent assurance that the AV value specified in the TSs will not be exceeded during surveillance tests. This is known as the LER Avoidance test. The final NTSP is chosen to satisfy both goals (protecting the AL and avoiding LERs) and is equivalent to the Nominal Setpoint (NSP) term used in RIS 2006-17.

Determination of the As-Found Tolerance (AFT) and ALT for the digital NUMAC PRNM System is discussed in Enclosure 6. Enclosure 6 provides a sample MNGP calculation(') CA-08-050, Revision 0, entitled "Instrument Setpoint Calculation - Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU," illustrating the MNGP specific implementation of the GEH ISM to determine the setpoints for two TS functions.

MNGP procedures require the instrument to be declared inoperable if the AV is exceeded and require that corrective actions be initiated any time the AFT is exceeded. This includes evaluating instrument performance before the channel is returned to service.

By the GEH ISM all setpoints are reset to the NTSP, within the ALT, after calibration. The ALT is a procedural allowance specified by the calibration procedure and its value is generally the same as the instrument accuracy. The

1. Two GEH setpoint calculations (Attachments 2 and 3) attached to CA-08-050 for convenient reference that contain GEH proprietary information have been removed.

Page 2 of 29

ENCLOSURE 1 RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS magnitude of ALT is generally less than the target maximum value specified by RIS 2006-17. MNGP procedures consider an instrument channel inoperable if it cannot be restored or calibrated within the specified ALT. Margin allowance for ALT is already incorporated in the calculated margins for the AV and the NTSP values according to the approved GEH ISM, so the ALT used in setpoints calculated by GEH ISM, meets the guidance of RIS 2006-17.

The applicable AFT and ALT depend on the surveillance test and the type and portion of the instrument loop that is being tested or calibrated. For example, the surveillance test for the digital electronics of the PRNM System is not vulnerable to drift or instrument inaccuracy, so the AFT and ALT for the PRNM setpoints is conservatively implemented as zero.

Enclosure 6 provides sample calculation CA-08-050, Revision 0, which illustrates the MNGP specific implementation of the GEH ISM to determine the setpoints for the following two TS functions:

TS Table 3.3.1.1-1 APRM Neutron Flux - High (Setdown) Scram (2.a)

APRM Neutron Flux - High Scram (2.c)

Separate from these TS functions, applying the GE ISM for cases involving Limiting Safety System Settings (LSSS) for which AFT or ALT are determined, or for cases where the NTSP or AV was determined, the methodologies would be documented within the MNGP Technical Requirements Manual (TRM).(~)The MNGP TRM is subject to 10 CFR 50.59 evaluation for any changes made to the document.

The NTSP, AV, and AL or the limiting design value, (as applicable) for each setpoint involved with the PRNM System implementation is provided in the response to RAI No. 2 in this enclosure.

2. TRM Appendix C was created in conjunction with the Improved Standard Technical Specifications conversion to document the methods used to calculate the AFT and ALT for several Emergency Core Cooling System (ECCS) setpoints considered LSSS.

Page 3 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS Safetv Limit (SL)-Related Determination: Provide a statement as to whether or not the setpoint is a limiting safety system setting (LSSS) for a variable on which a safety limit (SL) has been placed as discussed in 10 CFR 50.36(d)(l)(ii)(A). Such setpoints are described as "SL-Related" in the discussions that follow. In accordance with 10 CFR 50.36(d)(l)(ii)(A), the following guidance is provided for identifying a list of functions to be included in the subset of LSSS specified for variables on which SLs have been placed as defined in Standard Technical Specifications (STS)

Sections 2.1 .I, Reactor Core SLs and 2.1.2, Reactor Coolant System Pressure SLs. This subset includes automatic protective devices i n TS for specified variables on which SLs have been placed that: (1) initiate a reactor trip; or (2) actuate safety systems. As such these variables provide protection against violating reactor core safety limits, or reactor coolant system pressure boundary safety limits.

Examples of instrument functions that might have LSSS included i n this subset in accordance with the plant-specific licensing basis, are pressurizer pressure reactor trip (pressurized water reactors), rod block monitor withdrawal blocks (boiling water reactors), feedwater and main turbine high water level trip (boiling water reactors), and end of cycle recirculation pump trip (boiling water reactors). For each setpoint, or related group of setpoints, that you determined not to be SL-Related, explain the basis for this determination.

Response

As described in Sections 5.5 and 5.1.7 of the PRNM System LAR, the following functions from Specification 3.3.1. I , "Reactor Protection System Instrumentation," and Specification 3.3.2.1, "Control Rod Block Instrumentation,"

listed below are affected by installation of the PRNM System and require determination of whether they are Limiting Safety System Settings (LSSS) on which a safety limit has been placed in accordance with 10 CFR 50.36(c)(l)(ii)(A).

TS Table 3.3.1 .I-1 APRM Neutron Flux - High (Setdown) (2.a)

APRM Simulated Thermal Power - High (2.b)

APRM Neutron Flux - High (2.c)

OPRM Upscale (2.f)

Page 4 of 29

ENCLOSURE 1 RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS TS Table 3.3.1.2-1 Rod Block Monitor - Low Power Range - Upscale ( I .a)

Rod Block Monitor - lntermediate Power Range - Upscale (I .b)

Rod Block Monitor - High Power Range - Upscale ( I .c)

The NMC has reviewed these TS setpoint (or parameter setting) functions versus their associated safety analysis functions and determined which of the above Reactor Protection System (RPS) and Rod Block Monitor (RBM) functions discussed in the LAR are LSSS on which a safety limit has been placed in accordance with 10 CFR 50.36(c)(l)(ii)(A). The safety-limit related LSSS determination evaluations are provided below.

A. APRM Neutron Flux - High (Setdown)

Installation of the PRNM System introduces a new function to the MNGP TS, the APRM Neutron Flux - High (Setdown) scram function. The APRM Neutron Flux - High setdown function provides a redundant scram to the lntermediate Range Monitors (IRMs) for reactivity transients in the startup mode and is discussed in the PRNM System Licensing Topical Report (LTR).

Two BWR Owners Group documents (References 8 and 9) provide guidance on evaluating TS instruments that may be LSSS. They indicate that the APRM Neutron Flux - High (Setdown) scram function is a redundant scram function to that provided by the IRMs. This function is not credited in any design basis safety analysis for the MNGP and does not have an Analytical Limit.

APRM IOPRM Function Nominal Trip Allowable Analytical TS Table 3.3.1.I-1 Name Setpoint Value Limit APRM Neutron Flux - High I 18 % RTP r 20 % RTP NIA (Setdown) (Function 2.a)

Consistent with the MNGP licensing basis and the above guidance, the APRM Neutron Flux - High (Setdown) scram function is not a LSSS variable on which a safety limit has been placed in accordance with 10 CFR 50.36(c)(l)(ii)(A).

The TS Bases for the specification state that functions not specifically credited in the accident analysis are retained for overall redundancy and diversity of the RPS as required by the NRC approved licensing basis. This Page 5 of 29

ENCLOSURE 1 RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS function is being included in the TS since it is part of the PRNM System design and is being added to the MNGP licensing basis.

A sample calculation is provided in Enclosure 6 indicating how this setpoint is determined.

B. APRM Simulated Thermal Power - High The APRM Simulated Thermal Power (STP) - High scram function monitors neutron flux to approximate the thermal power being transferred to the reactor coolant. The APRM neutron flux is electronically filtered with a time constant representative of the fuel heat transfer dynamics to generate a simulated thermal power signal proportional to the thermal power in the reactor. The 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) but is clamped at an upper limit that is always lower than the APRM Neutron Flux - High function AV.

This setpoint function is different from the current flow-biased APRM Neutron Flux scram function which is based on the unfiltered neutron flux signal. The APRM STP signal responds more slowly to reactivity changes since it is based on a filtered (less than 7-second filter) neutron flux signal. The flow-biased APRM STP - High scram function mitigates slow reactivity transients initiated near the operating map boundary (such as a loss of feedwater heating) by reducing the over-power and delta Critical Power Ratio (CPR) for these events, but is not required to protect the Minimum Critical Power Ratio (MCPR) Safety Limit. As indicated in Table 3-1, "Limiting Safety System Settings for a Typical BWR14," of Reference 9, this setpoint function has the potential to be an LSSS requiring evaluation. A review of the MNGP design basis indicates that the APRM STP - High scram function does not have an AL and is not credited in the safety analysis.

APRM 1 OPRM Function TS Table Nominal Trip Allowable Analvtical 3.3.1.I-1 Name Setpoint Value Limit APRM Simulated 50.66 Wd + 59.6 % <- 0.66 Wd + 61.6 % NIA Thermal Power - RTP, and RTP, and

~ i g h ((Function

~) 2.b) 5 114 % RTP 5 116%RTP

4. The APRM STP - High NTSP is 5 0.66Wd + 54.6 % RTP and the APRM STP - High AV is 5 0.66(Wd- 5.4) + 61.6 % RTP when reset for single loop operation. Delta W is specified in the Core Operating Limits Report. There is no AL for the APRM STP - High function.

Page 6 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS The NRC in a safety evaluation for ARTS/MELLM(~) implementation at Susquehanna (Reference 10) states, the "APRM STP - High Function, being revised, is not SL-related, and it does provide defense-in-depth to the APRM Fixed Neutron Flux - High Function. This function is being retained in the TSs since it is part of the RPS design and the NRC-approved licensing basis. ... The NRC staff agrees that the RBM power-dependent setpoints are the only TS functions removed or altered by this LAR that are considered an SL-related LSSS." The MNGP licensing basis also indicates that the APRM STP - High scram function is not a LSSS on which a safety limit has been placed in accordance with 10 CFR 50.36(c)(l)(ii)(A).

As discussed in Item B, the TS include functions not specifically credited in the accident analysis that are retained for overall redundancy and diversity of the RPS. Since this function is part of the PRNM System design it is being included in the TS and is being added to the MNGP licensing basis.

C. APRM Neutron Flux - High (SL Related LSSS)

The APRM Neutron Flux - High scram function protects against all fast reactivity transients. The APRM Neutron Flux - High scram function generates a trip signal to prevent fuel damage or excessive Reactor Coolant System (RCS) pressure in the high power range. For the overpressurization protection analysis, high neutron flux is assumed to terminate the main steam isolation valve (MSIV) closure event and along with the safetylrelief valves (SIRVs) limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis takes credit for high neutron flux to terminate the CRDA. The AV is based on the AL assumed in the CRDA analysis.

The APRM Neutron Flux - High scram is based on the unfiltered neutron flux signal. For rapid neutron flux increase events, thermal power lags the neutron flux and the APRM Neutron Flux - High function will provide a scram signal before the flow-biased APRM Simulated Thermal Power -

High scram. The APRM Neutron Flux - High scram function AL is not changed with the installation of the PRNM System.

5. ARTSIMELLLA stands for -Average Power Range MonitorIRod Block Monitor~Technical SpecificationsIMaximum Extended Load Line Limit Analysis.

Page 7 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS APRM IOPRM Function Nominal Trip Allowable Analytical TS Table 3.3.1.I-1 Name Setpoint Value Limit APRM Neutron Flux - 1119.5%RTP r122%RTP r125%RTP High (Function 2.c)

The APRM Neutron Flux - High scram function is required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could result in the Safety Limits (e.g., MCPR and RCS pressure) being exceeded. As indicated in Table 3-1, "Limiting Safety System Settings for a Typical BWW4," of Reference 9, the APRM Neutron Flux -

High scram function is an LSSS on which a safety limit has been placed since it protects both the MCPR Safety Limit and the Reactor Pressure Safety Limit in accordance with 10 CFR 50.36(c)(l)(ii)(A).

The safety-limit related LSSS notes suggested by RIS 2006-17 are applicable and a digital version of the notes, that were approved for ARTSIMELLA implementation at Susquehanna Units 1 and 2 (Reference 10) will be applied (see the revised TS page for this function in Enclosure 4).

A sample calculation is provided in Enclosure 6 indicating how this setpoint is determined.

D. OPRM Upscale The BWROG Stability Long-Term Solution Option Ill is implemented utilizing the OPRM system. The period based detection algorithm (PBDA) is one of the three algorithms implemented in the OPRM Upscale function, but is the only algorithm credited in the safety analysis.

The BWR Owners Group developed a methodology (Reference 8) based on the requirements of 10CFR50.36 for identifying SL-LSSS and applied it to the BWRl4 and BWW6 Improved Technical Specification (ITS) NUREGs. In the BWROG methodology, only the LSSS associated with the analysis of anticipated operational occurrences (AOOs) that have the potential to challenge one of the four SLs are considered SL-LSSS. Accidents and plant capability evaluations (special events) are not included because these categories of events have event limits that typically allow exceeding the safety limits.

Page 8 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS The OPRM Upscale function is only credited as part of the reactor stability analysis. Only a Nominal Trip Setpoint is developed as part of the reload analysis which will be specified in the Core Operating Limits Report (COLR).

This function does not have an Allowable Value or an Analytical Limit.

APRM IOPRM Function Nominal Trip Allowable Analytical TS Table 3.3.1.I-1 Name Setpoint Value Limit OPRM Upscale In development NIA NIA (Function 2.9 for next Cycle.

GEH has provided a proprietary discussion and evaluation (see Enclosure

9) which clarifies the OPRM Upscale function licensing basis and use of the Safety Limit Minimum Critical Power Ratio (SLMCPR) as a specified acceptable fuel design limit (SAFDL) in reactor stability analyses in accordance with General Design Criteria (GDC) GDC-10 and 12.@)The following topics are discussed:

Review of the different Safety Limits Review of Applicable General Design Criteria Review of specified acceptable fuel design limits (SAFDL)

Discussion on the multiple uses of the SLMCPR - as a SAFDL for various analyses Discussion on the digital nature of the OPRM Unique stability setpoint methodology

1. Safetv Limits (SL) and Limiting Safetv Svstem Settings: 10 CFR 50.36 defines the SL and LSSS. Two SLs were contained in the initial BWR TS:

Fuel Claddinq Integrity (Minimum Critical Heat Flux Ratio) -

Considered a bounding value to prevent fuel rod burnout.

Reactor Coolant System (RCS) Pressure - 1325 psig reactor dome pressure - Based on ASME Code pressure limit for upset conditions, monitored by vessel pressure instrumentation.

6. 10 CFR 50, Appendix A, GDC-10 is "Reactor Design" and GDC-12 is "Suppression of Reactor Power Oscillations."

Page 9 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS Later, the fuel cladding integrity SL was split into three SLs:

Safety Limit Minimum Critical Power Ratio (SLMCPR) - Assure that greater than 99.9 % of fuel rods would not experience boiling transition.

Low Flow or Low Pressure Limit on Reactor Power -Assure nucleate boiling at conditions for which transition boiling data was not available.

Safety Limit Water Level - 1 foot above the top of active fuel - Chosen to prevent fuel failure due to heatup following core uncovery.

Based on 10 CFR 50.36(C)(l)(ii)(A), the initial BWR TS issued by the NRC contained sets of LSSS:

Reactor Protection System (RPS) trips.

Emergency Core Cooling System initiations.

ASME Code qualified safety valve opening setpoints.

((

PROPRIETARY INFORMATION REMOVED The above sets of LSSS were relatively consistently included in BWR TSs until the conversion to Improved Standard TS (ITS) occurred. In the ITS, the LSSS section was removed and the direct tie to the SL was diminished; however, similar requirements were included as Limiting Conditions for Operation (LCOs).

2. GDC-10, Reactor Design and GDC-12, Suppression of Reactor Power Oscillations:

1. GDC-10: The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

2. GDC-12: The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

Page 10 of 29

ENCLOSURE 1 RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS

3. Specified Acceptable Fuel Design Limits (SAFDLs): The two GDCs clearly separate the types of events to which they apply. SAFDLs associated with GDC-10 apply to eventslconditions during normal operation and anticipated operational occurrences (AOOs). GDC-12 states power oscillations which could exceed the SAFDL either are not possible or that the oscillations are able to be reliably and readily detected and suppressed. In other words power oscillations may exceed the SADFL as long as they are detected and suppressed. ((

PROPRIETARY INFORMATION REMOVED 11 By definition, GDC-10 is limited to AOOs, which are defined as conditions of normal operation expected to occur one or more times during the life of a nuclear power unit and include, but are not limited to, loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power. These events are the subject of the transient (or AOO) analyses in Chapter 14 of the MNGP Final Safety Analysis Report (FSAR). The current SAFDLs for AOOs (Sections listed below) from the fuel licensing topical report GESTAR (Reference 11) are:

MCPR Safety Limit 4.3.1 Fuel Pellet Centerline Melting 2.2.5 One Percent Fuel Rod Cladding Plastic Strain 2.2.7 Fuel Enthalpy Design Limit 2.2.3.1.4 This list of SAFDLs, i.e., fuel design limits, demonstrates that event limits for AOOs are more inclusive than just complying with the SLMCPR. The initial BWR SAFDL for stability was expressed in terms of decay ratio. Because of the technological changes in fuel designs, the current stability analytical limit is actually fuel integrity. ((

PROPRIETARY INFORMATION REMOVED 11 The NRC concluded in the OPRM Technical Evaluation Report (TER) associated with the NRC SE for NEDO-32465-A (Reference 12) that there is a "high-likelihood that fuel integrity will not be compromised by the likely instability events. We must note, however, that this statistical approach allows for a 5% probability that the CPR limit will be reached during an instability event." ((

PROPRIETARY INFORMATION REMOVED 11 Page 11 of 29

ENCLOSURE 1 RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS Technological Changes: As BWR designs developed, there have been technological advances in fuel design and analyses that have impacted the treatment of both AOOs and stability. These include:

((

PROPRIETARY INFORMATION REMOVED

4. Employing the SLMCPR for Multiple Uses: The association of the SLMCPR with multiple functions can be the source of considerable confusion. The three functions are:

The identification of the SLMCPR as a safety limit in TSs.

Use of the SLMCPR as one of the SAFDLs in the A 0 0 analyses.

Use of the SLMCPR as a SADFL figure of merit in stability analysis.

SLs are J no associated with protection of the fuel cladding and RCS barriers to the release of radioactive material.

The use of the SLMCPR as a SAFDL in A 0 0 analyses is appropriate for the identification of SL-LSSS. In this analysis, an operating limit minimum critical power ratio (OLMCPR) is identified and established so that the SLMCPR will not be exceeded in the event the limiting A 0 0 occurs.

Because the OLMCPR is used analytically to avoid exceeding the SLMCPR, instrument setpoints associated with the instruments assumed to function during AOOs can be identified as SL-LSSS.

The use of the SLMCPR as a figure-of-merit (SADFL) in reactor stability analysis is also appropriate. In other words because GDC-12 allows power oscillations which may exceed the SADFL as long as they are reliably and readily detected and suppressed [emphasis added], the SLMCPR in this context cannot be a SL. Rather, it is a figure-of-merit used in lieu of other possible parameters. Since the regulation, GDC-12, allows the SADFL (the SLMCPR) to be exceeded, was approved by the NRC as part of reactor stability licensing methodology, the OPRM Upscale function cannot be an LSSS and does not protect a SL. ((

PROPRIETARY INFORMATION REMOVED I] can be exceeded in stability analyses, the OPRM setpoints are not considered SL-LSSSs.

Page 12 of 29

ENCLOSURE 1 RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS Digital Instrumentation: The OPRM is designed to trip the reactor if power oscillations of sufficient magnitude are detected. The OPRM signal is a relative signal and is obtained by dividing the instantaneous reading, which could oscillate if the local core power is oscillating, by a reading which is strongly filtered and is relatively constant in time. Since the signal is a ratio, it is insensitive to the drift and calibration errors of the signal conditioning electronics that could be present if the equipment was analog. In fact the OPRM electronics are digital and the signal conditioning electronics and setpoints are implemented with digital electronics and software, and do not drift. The OPRM setpoints are not adjusted, and have no as-found and as-left tolerances. Thus the OPRM setpoints are not affected by the requirements in RIS 2006-17 (Reference 7), which is concerned with monitoring the performance of the instrument during calibration to ensure that it has not drifted excessively between calibrations so that the instrument error margins used in the setpoint calculation remain valid.

6. Unique Setpoint Methodologv: The OPRM setpoints are not derived from GE ISM (Reference 5), or any other setpoint methodology based on RG 1.I05 (Reference 13) and ISA-67.04 (Reference 14). The OPRM setpoint methodology is a comprehensive BWROG methodology for stability analysis approved by the NRC (Reference 12). According to this licensed methodology the stability analysis is based on nominal setpoints. There is no Analytic Limit or Allowable Value with defined instrument error margins to the Nominal Trip Setpoint for the OPRM setpoints. Instrument error was not specifically considered because of conse~atismsinherent in the analysis methodology. Thus the OPRM stability setpoints are based on a unique licensing basis methodology.

Use of nominal setpoints has been more recently addressed in a response to an NRC Request for Additional Information (RAI) (Reference 15) during the licensing of PRNM System at Browns Ferry Unit 1. The NRC approved the implementation of the PRNM System at Browns Ferry Unit 1 and the use of the nominal setpoints (Reference 16).

Utilization of the SLMCPR as a SAFDL (i.e., a figure-of-merit) in the NRC approved licensing stability methodology in accordance with GDC-12 indicates that the SLMCPR is not a SL when applied in this reactor stability analysis context. Based on the above discussion and review of the regulatory requirements and stability licensing basis the OPRM Upscale Function does not protect a SL. The stability event is treated as a special event in the GEH analysis methodology, not as an AOO. Therefore, consistent with the GEH stability licensing basis being applied to the MNGP and the above discussion and clarifications, the OPRM Upscale scram Page 13 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS function is not a LSSS variable on which a safety limit has been placed in accordance with 10 CFR 50.36(c)(l)(ii)(A).

Also, RIS 2006-17 is not applicable to the OPRM Upscale function since the OPRM electronics are digital and the setpoints are not subject to drift. The NRC-approved setpoint methodology is unique to the stability analysis and is not associated with a setpoint methodology, such as, RG 1.105 and ISA-67.04.

A GEH proprietary discussion is provided in Enclosure 9 expands on the discussion presented above.

Rod Block Monitor - Low, lntermediate and High Power Ranges -

Upscale (SL Related LSSS)

The Rod Block Monitor (RBM) - Low, Intermediate, and High Power Ranges - Upscale functions are designed to prevent violation of the MCPR Safety Limit and the cladding one percent plastic strain fuel design limit that may result from a single control rod withdrawal error (RWE) event. A statistical analysis of RWE events was performed to determine the RBM response for both channels for each event. From these responses, the fuel thermal performance as a function of the RBM AV was determined. The AVs are chosen as a function of power level.

RBM Function Nominal Trip Allowable Analvtical TS Table 3.3.1.2-1 Name Setpoint Value Limit Rod Block Monitor - Low 5 1 2 0 / 1 2 5 o f S120.41125 51241125 Power Range - Upscale full scale (FS) of FS of FS (Function I.a)

Rod Block Monitor - ~ 1 1 5 / 1 2 5 o f r115.4/125 5119/125 Intermediate Power Range FS of FS of FS

- Upscale (Function I .b)

RodBlockMonitor-High S110/125of S110.4/125 S114/125 Power Range - Upscale FS of FS of FS (Function I.c)

The RBM setpoints are based on the APRM, RBM and TS improvement (ARTS) program applied to the MNGP (Reference 17). The RBM setpoints are set in accordance with the results of the reload transient analysis verified each cycle as documented in the COLR.

Page 14 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS The RBM is a digitally based system. As such, the system and its components are not subject to the setpoint drift attributable to typical analog systems. The applicable AFT and ALT depend on the surveillance test and the type and portion of the instrument loop that is being surveilled. For example the surveillance test for the digital RBM contained in the PRNM System instrument, tests the RBM trip setpoint which is stored digitally.

This value does not drift, and is never adjusted, so the AFT and ALT for the test is zero.

Table 3-1 of Reference 9, indicates that due to the generic nature of the RWE analysis applied to the BWRl2-5 that the Rod Block Monitor (RBM) -

Low, Intermediate, and High Power Range - Upscale setpoint functions are LSSS on which a safety limit has been placed since they protect the MCPR Safety Limit in accordance with 10 CFR 50.36(c)(l)(ii)(A).

This determination is consistent with results from recent ARTSIMELLLA implementations at Susquehanna Units 1 and 2 (Reference 10) and Nine Mile Point Unit 2. The NRC in the ARTSIMELLLA Susquehanna SE states "The NRC staff agrees that the RBM power-dependent setpoints are the only TS functions removed or altered by this LAR that are considered an SL-related LSSS." Note that this is consistent with the MNGP licensing basis which indicates that the MNGP RBM power-dependent setpoints are LSSS variables on which a safety limit has been placed in accordance with 10 CFR 50.36(c)(l)(ii)(A).

Conclusion The following instrument setpoints (or setting) functions have been determined by the NMC to be LSSS on which a safety limit has been placed for the MNGP in accordance with 10 CFR 50.36(c)(l)(ii)(A).

a. TS Table 3.3.1.I-1 APRM Neutron Flux - High (2.c)

b. TS Table 3.3.1.2-1 Rod Block Monitor - Low Power Range - Upscale ( I .a)

Rod Block Monitor - Intermediate Power Range - Upscale ( I .b)

Rod Block Monitor - High Power Range - Upscale ( I .c)

Page 15 of 29

ENCLOSURE 1 RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS NMC requests that in addition to the NRC SE for the PRNM System license amendment specifying that the above listed functions are LSSS on which a safety limit has been placed, that the SE also clearly delineate that the following functions have been reviewed by the NRC as part of this submittal and that they are not safety-limit LSSS in accordance with 10 CFR 50.36(c)(l)(ii)(A).

TS Table 3.3.1 .I-1 APRM Neutron Flux - High (Setdown) (2.a)

APRM Simulated Thermal Power - High (2.b)

OPRM Upscale (2.9 This action will avoid future repeat reviews for functions already determined by both the NMC and the NRC to not be safety limit related LSSS, reducing the time and effort involved in future resolution of the LSSS setpoint issue.

3. For the Setpoint that is determined to be SL-Related: The NRC letter to the Nuclear Energy Institute SMTF dated September 7,2005 (ADAMS Accession Number ML052500004), describes Setpoint-Related TS (SRTS) that are acceptable to the NRC for instrument settings associated with SL-Related setpoints. Specifically: Part " A of the Enclosure to the letter provides L C 0 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 changes.

a. Describe whether and how you plan to implement the SRTS suggested i n the September 7,2005 letter. If you do not plan to adopt the suggested SRTS, then explain how you will ensure compliance with 10 CFR 50.36 by addressing items 3b and 3c, below.

b. As-Found Setpoint Evaluation: Describe how surveillance test 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 action processes (including plant procedures) for restoring channels to operable status when channels are determined to be "inoperable" or "operable but degraded." If the criteria for 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.

Page 16 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS

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 associated analyses.

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.

Response to RAI 3.a For the setpoints associated with an LSSS that have been determined to be SL-Related listed below, the NMC does plan to implement a digital instrument related version of the LSSS setpoint notes that was approved by the NRC for ARTSIMELLA application at Susquehanna Units 1 and 2 (Reference 10). These notes are similar to those in the September 7, 2005, NRC letter to NEI (Reference 18), but reflect the digital nature of the PRNM System. The notes would be applied to the following TS functions:

TS Table 3.3.1.I-1 APRM Neutron Flux - High (2.c)

TS Table 3.3.1.2-1 Rod Block Monitor - Low Power Range - Upscale ( I .a)

Rod Block Monitor - Intermediate Power Range - Upscale ( I .b)

Rod Block Monitor - High Power Range - Upscale ( I .c)

The NMC has evaluated the suggested note descriptions within RIS 2006-17 (Reference 7), the NRC letter to the Nuclear Energy Institute (NEI) Setpoint Methodology Taskforce (SMTF) dated September 7, 2005 (Reference 19), and draft Technical Specification Task Force (TSTF)-493 - Revision 3 (Reference 20). Also, formats recently approved by the NRC for digital instrument LSSS notes at Susquehanna Units 1 and 2 (Reference 10) and Nine Mile Point Unit 2 (Reference 2 I) were reviewed.

Appendix C of the TRM was created by NMC in conjunction with the ITS implementation in 2006 to document the methodologies applied to several Emergency Core Cooling System setpoints considered LSSS during the ITS conversion. It is proposed to document the APRM Neutron Flux - High function

7. NMC in several previous approved LARS,including this LAR, committed to evaluate TSTF-493 after issuance. RIS 2006-17 states, "Methods and approaches different from those in this RIS may also be acceptable to the NRC." NMC intends to align the LSSS footnotes, TS Bases and TRM entries after final NRCIindustry resolution of this issue.

Page 17 of 29

ENCLOSURE 1 RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS NTSP and the methodology used in it's determination in Appendix C to the TRM.

The MNGP TS state that the Allowable Values for the Rod Block Monitor Low, Intermediate, and High Power Range - Upscale trip setpoints are specified in the COLR. These Rod Block Monitor setpoints will continue to be listed in the COLR (the NTSPs are also listed in the COLR). The methodology used to determine the Rod Block Monitor Low, Intermediate, and High Power Range - Upscale trip setpoints will be located in Appendix C to the TRM. Both the TRM and the COLR receive 10 CFR 50.59 reviews for any changes to their contents.

It is proposed that the following notes reflecting the digital nature of the PRNM System be applied to the four functions listed previously, from TS Tables 3.3.1.1-1 and 3.3.1.2-1 that are affected by this PRNM System installation:

For TS Table 3.3.1.I-1 --- APRM Neutron Flux - Hiqh (2.c):

If the as-found channel setpoint is not the Nominal Trip Setpoint but is conservative with respect to the Allowable Value, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

The instrument channel setpoint shall be reset to the Nominal Trip Setpoint at the completion of the surveillance; otherwise, the channel shall be declared inoperable. The NTSP and the methodology used to determine the NTSP are specified in the Technical Requirements Manual.

For TS Table 3.3.1.2-1 --- Rod Block Monitor - Upscale Functions (I .a, 1.b & 1.c))

If the as-found channel setpoint is not the Nominal Trip Setpoint but is conservative with respect to the Allowable Value, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

The instrument channel setpoint shall be reset to the Nominal Trip Setpoint at the completion of the surveillance; otherwise, the channel shall be declared inoperable. The NTSP shall be specified in the COLR. The methodology used to determine the NTSP is specified in the Technical Requirements Manual.

Note, the only difference between the two sets of notes is that the NTSP for the RBM trip setpoints will continue to be specified within the COLR, reflecting present practice.

Additionally, the TS Bases will describe the application of the notes to the particular TS instrumentation function. Draft, proposed, texts of the corresponding revised inserts to the TS Bases are provided in Enclosure 5.

Page 18 of 29

ENCLOSURE 1 RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS Response to 3.b and 3.c Sections 4.4.11 and 4.4.14 of the MNGP Instrument Control Manual provide guidance on performing instrument surveillance testing and conduct of work completion reviews and closeout.

Data found outside of specified limits during surveillance testing is required to be promptly entered into the corrective action process. When the AFT or ALT data does not meet the requirements the out of tolerance data must be reported to the Supervisor Maintenance (I&C). Attachment 1 to administrative procedure FP-PA-ARP-01, "CAP Action Request Process, requires under Category 13, "Technical Specifications," as part of the severity level determination process that any TS instrument that is outside of its AFT or ALT is considered a condition adverse to quality requiring entry of the condition into the Corrective Action Program (CAP) process.

The Supervisor Maintenance (I&C) (or designee) enters the condition into the CAP and the Shift Manager (or designee) is informed of the condition for review and determination of the impact on operability. The Supervisor Maintenance (I&C) is responsible for making an initial evaluation of any out of tolerance condition reported by the I&C Technician. The process is discussed in more detail below.

Surveillance procedures are assigned to I&C Technicians by the Supervisor Maintenance (I&C) or his designee for performance as required by the surveillance schedule. Prior to starting the surveillance test, the Control Room Supervisor (CRS) must sign the "Approval to Commence" line on the record copy. During surveillance testing there are four possible results:

1. The instrument setpoint is found within the ALT; the results are recorded in the procedure and, from the TS perspective, no further action is required.

2. The setpoint is outside the ALT but within the AFT, the instrument setpoint is reset to within the ALT. From a TS perspective no further action is required.

3. The instrument setpoint is found conservative with respect to the AV but outside the AFT. In this case the setpoint is reset to the LTSP (within the ALT), and the channel's response is evaluated by the Supervisor Maintenance (I&C).

The Supervisor Maintenance (I&C) makes an initial evaluation of any out of tolerance condition where the channel is outside the AFT. Generally this evaluation requires the I&C technician to attempt to restore the out of Page 19 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS tolerance device to within acceptable limits and show that it is capable of performing its design function as provided in the calibration surveillance.

When making the initial evaluation, the following items are addressed:

Does the out of tolerance condition exceed any TS limits?

Does the out of tolerance condition exceed any Section XI limits?

Does the out of tolerance condition adversely affect the operability of the associated equipment andlor system? Consultation with Plant Engineering personnel is required if this is unclear.

Does the out of tolerance device exhibit signs of a degraded1 degrading condition or indicate an unreliable instrument (repeat failures) based on available historical calibration information, maintenance log, System Engineering input, or other site resources?

If the channel is operating as expected, then the channel can be restored to service at the completion of the surveillance. A prompt verification of the channel's condition is performed after the surveillance. The channel's as-found condition is entered into the CAP for further evaluation. If the channel is not operating as expected, the channel is inoperable.

4. The instrument setpoint is found non-conservative with respect to the AV.

The Supervisor Maintenance (I&C) makes an initial evaluation of any out of tolerance condition, including a channel outside the AV. This evaluation generally follows the steps outlined above for item 3.

The MNGP Instrument Control Manual requires when a channel is outside the AV that this be reported to the Shift Manager (or his designee). The Supervisor Maintenance (I&C) informs the Shift Manager who based upon the available information makes an immediate operability determination.

The channel's as-found condition is entered into the CAP for evaluation.

The surveillance shall not be continued until approved by the Shift Manager (or his designee).

Evaluations and corrective action (maintenanceltesting) is performed to correct the condition allowing the setpoint to be reset to the NTSP (within the ALT) and the channel to be declared OPERABLE and returned to service.

Page 20 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS

4. For the Setpoint that is not determined to be SL-Related: Describe the measures to be taken to ensure that the associated instrument channel is capable of performing its specified safety functions in accordance with applicable design requirements and associated analyses. Include in your discussion information on the controls you employ to ensure that the As-Left trip setting after completion of periodic surveillance is consistent with your setpoint methodology. Also, discuss the plant corrective action processes (including plant procedures) for restoring channels 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 test procedure), describe how it is ensured that the controls will be implemented.

Response

The following new functions proposed to be added as part of this PRNM System installation were determined in the response to RAI No. 2 of this enclosure to be non-SL-Related.

TS Table 3.3.1.I-1 APRM Neutron Flux - High (Setdown) (2.a)

APRM Simulated Thermal Power - High (2.b)

OPRM Upscale (2.9 The NMC does not plan to implement the setpoint related TSs note changes described in the September 7, 2005 (Reference 19), letter for these functions since they do not meet the criteria for being SL-Related LSSS. Nonetheless, as discussed in the response to RAI No. 3, the exact same processes are applied for setpoints determined to be non-SL-Related as those determined to be SL-Related. Therefore, the same administrative control practices, including entry into the corrective action program are applied for any non-SL-Related channels found to be "inoperable" or "operable but degraded."

Page 21 of 29

-~ ~ ~ - ~-

ENCLOSURE 1 RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS OPRM Allowable Values and Setpoints: The LAR markup of TS page 3.3.1 .I-5A lists the Allowable Value for Function 2.f, "OPRM Upscale," as "As specified in COLR." However, Section 5.1.5 of Enclosure 1 of the LAR states, "There are no Allowable Values associated with the OPRM Upscale function." Additionally, Section 5.1.5 states, "The PBDA trip setpoints, which can be change with each fuel cycle, will be documented in the COLR." Please resolve this apparent discrepancy.

Response

As discussed in the NRC approved licensing topical report NEDO-32465-A (Reference 22) and acknowledged in several NRC safety evaluations (References 23 and 24) for licensee's implementing the OPRM system, there are no AVs associated with the OPRM Upscale function. The OPRM period-based detection algorithm (PBDA) Upscale trip setpoints are determined using the Option Ill licensing methodology described in Reference 22 except that a plantlcycle-specific DIVOM(~)curve slope is used due to the BWROG1sresolution of a past GEH Part 21 issue. Since the PBDA trip setpoints are cycle-dependent they will be documented in the COLR.

Some plants have listed the AV for the OPRM Upscale function in TS Table 3.3.1.1-1 as NIA, with a superscript reference to a footnote to the table indicating the OPRM Upscale function values are specified in the COLR. The NMC chose a more direct presentation for the OPRM Upscale function setpoints, since they will be provided in the COLR, that was the same as that applied in the standard technical specifications for the Control Rod RBM function (Specification 3.3.2.1),

which simply states that the RBM setpoint values for the Low, Intermediate and High Power Ranges - Upscale are "As specified in COLR". As stated in the draft proposed TS Bases changes submitted with the LAR (Enclosure 4, Insert 61, page B 3.3.1.1-8D, Item 2.f, OPRM Upscale), "There is no allowable value for this function."

8. DlVOM stands for Delta CPR over initial MCPR Versus Qscillation Magnitude Page 22 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS

6. OPRM Monitoring Period: The LAR proposes that TS Table 3.3.1 .I-1 Function 2.f, "OPRM Upscale," not be enabled until completion of an abbreviated 90 day OPRM Monitoring Period after the first plant startup following PRNM system installation. The OPRM portion of the PRNM system would operate in an indication only mode during this period. The proposed 90 day OPRM Monitoring Period is a departure from the NRC safety evaluation for NEDC 32410P, dated September 5, 1995, which recommends the OPRM Monitoring Period last for one full fuel cycle. The LAR justification for this abbreviated OPRM Monitoring Period is that Option Ill OPRM systems have accumulated more than 90 reactor years of operation and based on current industry and vendor experience with the NUMAC PRNM System that the possibility of problems with the algorithms, system performance, or hardware problems with Option Ill is considered unlikely. Provide a more detailed justification for the proposed reduction of the OPRM Monitoring Period to 90 days. The justification should include discussion of actual plant experience of algorithm problems, system performance, and hardware problems.

Response

The one cycle monitoring period in the PRNM System LTRs (References 25, 26, and 27) for the OPRM was specified because it was a new feature of the RPS.

As such, further testing, monitoring, and evaluation in the normal modes of operation was considered required to ensure that the OPRM performed as designed and did not create any unintended consequences. Since the original introduction of the OPRM, a great deal of operating experience has been obtained and the one cycle trial period is no longer needed and can be shortened to 90 days. GEH PRNM systems with the Option Ill OPRM have been installed at many plants within the U.S. and overseas. The Option Ill OPRM systems have accumulated more than 90 reactor years of fully armed operation, with the installations at Brunswick Units 1 and 2 and Browns Ferry Unit 1 being closest to the MNGP design. The MNGP is a BWR-3 plant with jet pumps, similar to Dresden Units 2 and 3 and Quad Cities Units 1 and 2. Based on the operational experience with these installations, GEH supports directly arming the OPRM system after an initial monitoring period of 90 days. Note that operating with the OPRM armed provides automatic stability mitigation. Thus, shortening the trial period with OPRM Option Ill is appropriate for the MNGP, particularly since experience in operating plants shows that it is acceptable to do so.

The MNGP intends to implement the Option Ill PRNM System during the next Refueling Outage (RFO). During the initial monitoring period when PRNM System is OPERABLE but the OPRM Upscale function (Function 2.0 is not trip-enabled, the MNGP will implement Backup Stability Protection (BSP)

Page 23 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS (Reference 28) as an alternate method for detection and suppression of instabilities proposed in this amendment application.

A non-proprietary summary of the operational problems encountered with OPRM by GEH, and the solutions for these issues relevant to the OPRM implementation at the MNGP, is described below:

1) Nine Mile Unit 2 Hiah Frequency Noise In July 2003, Nine Mile Point Unit 2 experienced an OPRM scram due to a thermal-hydraulic instability event. The Nine Mile Point Unit 2 event showed that while the OPRM system resulted in an effective scram, there were numerous successive confirmation count resets that were due to the corner frequency. The second-order Butterworth filter with a 3 Hz cutoff frequency setting allowed some residual high-frequency noise from the oscillation signal, and this led to numerous successive confirmation count resets. This successive confirmation count reset condition resulted in a safety communication (Reference 29) in which GEH recommended that the cutoff frequency be set to 1.0 Hz and the period tolerance be set to 100 msec or greater, with allowance to use a different value if applicable, based on additional justification.

The 1.0 Hz cutoff frequency was subsequently adopted at Perry and the OPRM system performed as designed during the Perry thermal-hydraulic instability event in 2004 with effective Safety Limit MCPR protection when the OPRM system generated a scram.

2) Experience with Coherent Noise A unique feature of the core thermal-hydraulic phenomenon is that reactor noise might become coherent occasionally with an oscillation period in the range of thermal-hydraulic instability events [

] This is a common phenomenon observed by many OPRM plant operators that the current alarm setpoint (which is solely based on counts) may be initiated during normal plant operation. However, it is noted while this noise might become coherent occasionally, and [

the amplitude usually will not grow in the absence of a true thermal-hydraulic instability event. Hence a high amplitude setpoint value [ ] will be effective in preventing a spurious reactor scram.

For the MNGP OPRM implementation, GEH is providing an improved feature [

I Page 24 of 29

ENCLOSURE 1 RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS

3) lnadvertent Half Scrams at Plants with the ABB Svstem Half-scrams have occurred at two plants that were using the Asea, Brown Boveri (ABB) based OPRM systems. [

4) Brunswick 2 lnadvertent Scram An inadvertent reactor scram occurred at Brunswick Unit 2 in 2007 while operating in the Single Loop Operation (SLO) with the remaining recirculation pump operating near maximum capacity. Control rods were pulled to a powerlflow state point very close to the MELLLA operating domain boundary. [

] The high noise level resulted in spiking that met the amplitude requirement of the growth rate algorithm.

As a result of the Brunswick Unit 2 inadvertent scram, GEH recommends that the operating reactor power for SLO operation be restricted to a power level that is at least 5 percent of RTP below the boundary of the EPUIMELLLA operating domain, and where acceptable SLO operation has been previously demonstrated. Also, the growth rate algorithm and amplitude based algorithm setting ranges have been revised to allow for a wider range consistent with the approved LTRs (References 30 and 31). These changes provide acceptable protection against a spurious scram due to power spiking during SLO.

The operational experience with the OPRM summarized above shows that the operational issues are related to the magnitude and coherence of the noise, and can be resolved by adjusting the OPRM system, once the noise is characterized. This noise characterization can be accomplished within the proposed 90 day period. During this time GEH will work with the NMC to measure and characterize the background APRM noise at the plant. It is expected that the MNGP plant noise will be similar to the other BWR-3 plants,

[ 1. If the APRM noise level is confirmed to be low

[ ] the OPRM system operation will likely not result in a spurious reactor scram. If the APRM noise level is high, then further review of the adequacy of the recommended amplitude setpoint will be performed. As long as the OPRM amplitude setpoint is [ ]the likelihood of an inadvertent reactor scram will be very low. In fact, all spurious trips to date would have been avoided with an amplitude setpoint [

Page 25 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS Based on the extensive GEH experience in the PRNM System installations, GEH fully supports arming of the OPRM system at the MNGP after an initial monitoring period of 90 days as this will provide for expeditious automatic stability protection while assuring that the chance of inadvertent scrams is acceptably low.

Page 26 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS REFERENCES

1. NMC letter to NRC, "License Amendment Request: Power Range Neutron Monitoring System Upgrade," (L-MT-08-004), dated February 6, 2008.

2. Email from P. Tam (NRC) to R. Loeffler (NMC) dated June 13,2008, "Monticello - Draft RAI re: Proposed Amendment on PRNM System (TAC MD8064) --- Enclosure 1 RAI Questions 1 through 6.

3. MNGP Engineering Standards Manual ESM-03.02-APP-I,Appendix I (GE Methodology Instrumentation and Controls), Revision 4.

4. GE-NE-901-021-0492, DRF A00-01932-1, Setpoint Calculation Guidelines for the Monticello Nuclear Generating Plant, October 1992.

5. NEDC-31336P-A, Class Ill, General Electric lnstrument Setpoint Methodology, September 1996.

6. NRC letter to the Boiling Water Reactor Owners Group, "Revision to Safety Evaluation Report on NEDC-31366, lnstrument Setpoint Methodology (NEDC-31336P)," dated November 6, 1995.

7 U.S. NRC Regulatory Issue Summary 2006-17, "NRC Staff Position on the Requirements of 10 CFR 50.36, "Technical Specifications," Regarding Limiting Safety System Settings During Periodic Testing and Calibration of lnstrument Channels," dated August 24, 2006.

8. GE-NE-0000-0057-2518-RO, BWR Owners Group, "Limiting Safety System Settings for BWRl4 and BWRJ6," September 2006.

9. GE-NE-0000-0062-5001--RO, BWR Owners Group, TSTF--493 lmplementation Guidance for BWR LSSS Setpoints Developed By GE Setpoint Methodology, January 2007.

10. NRC to PPL Susquehanna, LLC, "Susquehanna Steam Electric Station, Units 1 and 2 - Issuance of Amendment Re: Average Power Range MonitorlRod Block Monitorrrechnical SpecificationsIMaximum Extended Load Line Limit Analysis (ARTSIMELLLA) Implementation (TAC Nos. MC9040 and MC9041)," dated March 23, 2007.

11. GE Licensing Topical Report General Electric Standard Application for Reactor Fuel (GESTAR) - latest approved amendment.

Page 27 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing basis Methodology for Reload Applications," August 1996.

Regulatory Guide 1.105, "Setpoints for Safety Related Instrumentation",

Revision 3, 1999 ISA-S67.04, "Setpoints for Nuclear Safety-Related Instrumentation",

September 1994 N A to NRC, "Browns Ferry Nuclear Plant (BFN) - Unit 1, Techical Specifications (TS) Change TS Request for Additional Information (RAI)

Regarding Oscillation Power Range Monitor (OPRM) - (TAC No. MC9565),

NA-BFN-TS-443, October 2, 2006.

NRC to N A , Amendment No. 266 to Renewed Facility Operating License No.

DPR-33 for the Browns Ferry Nuclear Plant, Unit 1 - lssuance of Amendment Regarding Oscillation Power Range Monitor (TAC No. MC9565) (TS-443),

December 29,2006.

GE Licensing Topical Report, NEDC-30492-P, "Average Power Range Monitor, Rod Block Monitor, and Technical Specification Improvement (ARTS) Program for the Monticello Nuclear Generating Plant," April 1984.

MRC to NEI Setpoint Methods Task Force, "Technical Specification for Addressing Issues Related to Setpoint Allowable Values, dated September 7, 2005. (ADAMS Accession Number ML052500004)

NRC to NEI, NEI Setpoint Methods Task Force, "Technical Specification Allowable Values for Addressing Issues Related to Setpoint Allowable Values,"

dated September 7, 2005.

Technical Specification Task Force, Improved Standard Technical Specifications Change Traveler, TSTF-493, "Clarify Application of Setpoint Methodology for LSSS Functions," draft, Revision 3.

NRC to Nine Mile Point Nuclear Station, LLC, "Nine Mile Point Nuclear Power Station, Unit No. 2 - lssuance of Amendment Re: Implementation of ARTSIMELLA (TAC No. MD5233)," dated February 27,2008.

GE Licensing Topical Report, NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Application," August 1996.

Page 28 of 29

ENCLOSURE I RESPONSE TO THE JUNE 13,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS NRC to Exelon Nuclear, "Peach Bottom Atomic Power Station, Units 2 and 3 -

lssuance of Amendment Re: Activation of Oscillation Power Range Monitor Trip (TAC Nos. MC2219 and MC2220)," dated March 21,2005.

NRC to Exelon Nuclear, "Brunswick Steam Electric Plant, Units 1 and 2 -

lssuance of Amendment to Incorporate the General Electric Digital Power Range Neutron Monitoring System (TAC Nos. MB2321 and MB2322)," dated March 8, 2002.

NEDC-32410P-A Volume 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function," October 1995.

NEDC-32410P-A Volume 2 -- Appendices, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function," October 1995.

NEDC-32410P-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function," November 1997.

OG 02-0119-260, GE to BWR Owners' Group Detect and Suppress II Committee, "Backup Stability Protection (BSP) for Inoperable Option Ill Solution,"

July 2002.

GEH Safety Communication 03-20, "Stability Option Ill Period Based Detection Algorithm Allowable Settings, October 4, 2003.

NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

Page 29 of 29

ENCLOSURE 2 RESPONSE TO THE JUNE 25,2008 REQUESTS FOR ADDITIONAL INFORMATION TECHNICAL SPECIFICATION BRANCH QUESTIONS On February 6,2008, (Reference I ) the Nuclear Management Company, LLC (NMC) submitted a request to revise the Monticello Nuclear Generating Plant (MNGP)

Technical Specifications (TS) in conjunction with the installation of General Electric -

Hitachi (GEH) Nuclear Measurement Analysis and Control (NUMAC) Power Range Neutron Monitoring (PRNM) System.

The following requests for additional information (RAI) concerning the proposed TS changes were received from the U.S. Nuclear Regulatory Commission (NRC) by e-mail, dated June 25,2008 (Reference 2).

1. Page 15 of 168 describes the reason for the LC0 3.0.4 Note in Action 1.2 (page 89 of 168). It is unclear if the Note is applied correctly for the reasoning discussed. It is also unclear what removes the Note after the monitoring period.

Response

As described in the last two paragraphs under Section 5.1.4, "Addition of New Conditions I and J" in Reference 1, an exception to LC0 3.0.4 is proposed by the addition of a note to Required Action 1.2. This exception was not discussed within the NUMAC PRNM LTR (including the supplement) but was discussed and approved by the NRC in a 2005 Safety Evaluation for activation of the OPRM function at Peach Bottom Units 2 and 3 (Reference 3).

The NRC states in this Peach Bottom safety evaluation that while not included in the scope of the NUMAC PRNM LTR, the LC0 3.0.4 exception would allow plant restart "in the event of a shutdown during the 120-day completion time of Required Action 1.2, consistent with the orisinal intent of NEDC-32410P-A."

[emphasis added] The NRC goes on to state that the original intent "was to allow normal plant operations to continue during the recovery time from a hypothesized design problem with the Option Ill algorithms." As such, this proposed LC0 3.0.4 exception will be a permanent change to the MNGP TS and hence is not planned for removal at the end of the OPRM Monitoring Period.

The Required Action for Condition I when the OPRM Upscale function channels are inoperable requires the channels be restored to OPERABLE within 120 days.

Without the proposed LC0 3.0.4 exception, entry into the MODE or other specified condition in the Applicability would not be permitted for plant startup following PRNM System installation (or following shutdowns during the OPRM Monitoring Period) since the associated ACTIONS do not permit continued operation for an unlimited period of time. Therefore, this L C 0 3.0.4 exception is required for these reasons, also.

Page 1 of 4

ENCLOSURE 2 RESPONSE TO THE JUNE 25,2008 REQUESTS FOR ADDITIONAL INFORMATION TECHNICAL SPECIFICATION BRANCH QUESTIONS

2. Page 92 of 168 contains a Note 2 to SR 3.3.1.1.14 (Response Time Testing).

It is unclear how the Note is used. There is reasoning on page 68 and 69 of 168, however it is still unclear.

Response

The response time testing (RTT) proposed in Surveillance Requirement (SR) 3.3.1. I . I 4 of the MNGP TS will test both of the redundant OPRM or both of the redundant APRM trip outputs from each 2-Out-0f-4 Voter, i.e., Function 2.e, during each performance. This testing rate has been selected to simplify recordkeeping for the SR.

While the NUMAC PRNM LTRs justified reduced RTT, TS mark-ups were not provided to implement an "n" greater than 4 (the total number of Voter channels).

This note was added to SR 3.3.1.1. I 4 to define that "n=8" for Function 2.e. This testing rate results in a test of each APRM related Reactor Protection System (RPS) relay every 4 cycles, twice the rate justified in the LTRs. Without this notation, rigorous interpretation of four voter channels would result in a value of "n=4" for this SR.

The PRNM System modification includes redundant APRM trip and redundant OPRM trip outputs from each 2-Out-0f-4 Voter channel. There are 8 total RPS interface relays. NUMAC PRNM LTR Supplement 1 justified RTT at a rate that tested one RPS Interface relay every plant operating cycle, with tests using the APRM output for one cycle and the OPRM output for the next cycle. This yields a RTT rate of once per 8 operating cycles.

The RTT proposed in the MNGP TS will test both of the redundant OPRM or both of the redundant APRM trip outputs from each Voter during one application of the SR. This testing is consistent with the sequencing described in NUMAC PRNM LTR Supplement 1, but at twice the rate for the components.

Because this sequencing may be confusing, a more detailed description of the RTT sequence for the 2-Out-0f-4 Voter, Function 2.e, in accordance with SR 3.3.1. I . I 4 is proposed to be added to the TS Bases. A table showing an example of an acceptable test sequence is provided below.

Page 2 of 4

ENCLOSURE 2 RESPONSE TO THE JUNE 25,2008 REQUESTS FOR ADDITIONAL INFORMATION TECHNICAL SPECIFICATION BRANCH QUESTIONS An Acceptable Function 2.e Test Sequence for SR 3.3.1. I .14 "Staggering" Voter Voter Voter Voter Voter 24-Month Output Al A2 BI 62 RPS Trip Div.

Cycle Tested Output Output Output Output System 1st OPRMAI OPRM A 1 2nd APRM B1 APRM B 1 3rd OPRM A2 OPRM A 2 4th APRM 62 APRM B 2 5th APRMAI APRM A 1 6t h OPRM B1 OPRM B 1 7th APRM A2 APRM A 2 8th OPRM 62 OPRM B 2 After 8 cycles, the sequence repeats.

The pertinent draft TS Bases page has been revised to reflect this change and is provided in Enclosure 5. The specific tests will be defined in MNGP procedures.

The NRC approved this same proposed change for Susquehanna Units 1 and 2 to clarify and simplify the testing methodology. This approval is discussed in Section 3.4.3.5, "TS SR 3.3.1.1.17 Response Time Testing," of the NRC safety evaluation for Susquehanna Units 1 and 2 (Reference 4).

3. Page 94 of 168 has a Note (e) for the SR associated with Function 2.f, OPRM Upscale. There is some uncertainty regarding the Note to Operability of the OPRM, namely whether or not the OPRM is calibrated before it is declared Operable (SR 3.3.1.1. I 1 or similar test).

The OPRM will be fully calibrated before it is declared OPERABLE. Before the licensed operators can declare the OPRM system OPERABLE, they must determine that the OPRM system fully meets the definition of OPERABILITY in accordance with the TS, which demands that the applicable Surveillance Requirements, including all required testing be fully met.

Page 3 of 4

~~~~ ~- - - ~~

-~ ~ ~~ -

ENCLOSURE 2 RESPONSE TO THE JUNE 25,2008 REQUESTS FOR ADDITIONAL INFORMATION TECHNICAL SPECIFICATION BRANCH QUESTIONS REFERENCES

1. NMC letter to NRC, "License Amendment Request: Power Range Neutron Monitoring System Upgrade," (L-MT-08-004), dated February 6, 2008.

2. Email from P. Tam (NRC) to R. Loeffler (NMC) dated June 25, 2008, "Monticello -Additional Draft RAI re: Proposed Amendment on PRNM System (TAC MD8064) --- Enclosure 2 RAI Questions 1 through 3.

3. NRC to Exelon Nuclear, "Peach Bottom Atomic Power Station, Units 2 and 3 -

lssuance of Amendment Re: Activation of Oscillation Power Range Monitor Trip (TAC Nos. MC2219 and MC2220)," dated March 21,2005. (ADAMS Ascension No. ML05270020)

4. NRC to PPL Susquehanna, LLC, "Susquehanna Steam Electric Station, Units 1 and 2 - lssuance Of Amendment Re: Power Range Neutron Monitor System Digital Upgrade (TAC Nos. MC7486 and MC7487)," dated March 3, 2006.

Page 4 of 4

ENCLOSURE 3 RESPONSE TO THE JULY 1,2008 REQUESTS FOR ADDITIONAL INFORMATION REACTOR SYSTEMS BRANCH QUESTIONS On February 6,2008, (Reference I ) the Nuclear Management Company, LLC (NMC) submitted a request to revise the Monticello Nuclear Generating Plant (MNGP)

Technical Specifications (TS) in conjunction with the installation of General Electric -

Hitachi (GEH) Nuclear Measurement Analysis and Control (NUMAC) Power Range Neutron Monitoring (PRNM) System.

The following requests for additional information (RAI) concerning the proposed TS changes were received from the U.S. Nuclear Regulatory Commission (NRC) by email, dated July 1, 2008 (Reference 2).

1. Please provide:

( 1 description of the alternate method for detection and suppression of instabilities proposed in this amendment application; (2) identification of the differences between the proposed alternate method and the lnterim Corrective Actions (ICAs) specified in NRC Bulletin 88-07; and (3) clarification of the similarity between the alternate method and DSS-CD backup stability protection since the DSS-CD features of NUMAC PRNMS will be implemented for Monticello.

Response to Part 1 The MNGP is planning to implement the Option Ill PRNM System during the 2009 Refueling Outage (RFO). During the initial monitoring period when PRNM System is operable but the Oscillation Power Range Monitor (OPRM) trip is not enabled, the MNGP will implement Backup Stability Protection (BSP)

(Reference 3) as an alternate method for detection and suppression of instabilities. In addition, BSP will be used as a backup stability protection method (i.e., alternative method) for the duration allowed in the proposed revised MNGP Technical Specifications implementing the PRNM System if the OPRM becomes inoperable in the future.

Response to Part 2 The BSP methodology is an enhancement to the original lnterim Corrective Action (ICA) methodology. The lCAs define certain regions in the powerlflow map that are excluded from planned entry and prescribe specific actions upon unplanned entry (Reference 4). The ICA regions are based upon empirical evaluations and experience, are defined in terms of relative core flow and control rod line, and are uniformly applicable to all GE BWRs. These regions are not defined based on specific stability criteria. The ICA regions were established in Page 1 of 5

ENCLOSURE 3 RESPONSE TO THE JULY 1,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS 1994 based on original licensed thermal power, generally shorter fuel cycles, and more stable core designs. New energy-intensive core design changes have generally reduced stability margins. As a result, GEH proposed the BSP methodology as an enhancement to the ICA methodology.

A comparison of some of the characteristics between the Option Ill manual BSP and the lCAs are as follows:

The size of the base BSP regions is equivalent to the current ICA regions.

The BSP regions cannot be smaller than the ICA regions.

The BSP regions are reduced from three ICA regions (Scram, Exit, Controlled Entry) to two regions (Scram and Controlled Entry).

Decay ratio criteria are established for plantlcycle specific confirmation and, as necessary, the base BSP regions are adjusted.

Operator actions in the two BSP regions are similar to the operator actions defined for the ICA Scram and Controlled Entry regions.

Operator awareness as discussed in Reference 3, is required when operating within 10 percent of rated core flow or power from the BSP Controlled Entry region.

Response to Part 3 As stated in the response to Part 1 of this RAI, the MNGP PRNM System license amendment request (LAR) proposed the use of the Option Ill PRNM System and not the DSS-CD(') feature. The Option Ill BSP to be implemented will be similar to the manual BSP solution as in other Option Ill plants when the OPRM systems are not operable.

2. Please describe the plan to implement extended power uprate in conjunction with the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) for Monticello.

Response

As discussed on page 5 of the MNGP LAR, the NMC is not applying for an amendment to operate in the MELLLA+ operating domain by this LAR. That

1. DSS-CD stands for Detect and Suppress Solution - Confirmation Density. It includes the three BWROG Option Ill algorithms and the DSS-CD algorithm developed by GEH.

Page 2 of 5

ENCLOSURE 3 RESPONSE TO THE JULY 1,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS request will occur at a later date (to be determined) as part of a phased extended power uprate (EPU) implementation.

The DSS-CD stability solution (see References 5 and 6), an extension of the Option Ill stability solution methodology, will be necessary, however, to provide stability protection for operation within the MELLLA+ operating domain. The NMC requested in the LAR to install the DSS-CD stability solution operating in an Option Ill configuration as part of the NUMAC PRNM System retrofit.

One of the requirements for future DSS-CD implementation is the accumulation of operating data. The confirmation density algorithm will run (but will not provide a trip since it is not connected to the RPS trip output relays) to allow operational data to be gathered on its performance.

3. Provide the schedule to implement Option Ill stability solution for MNGP.

Please correct the typo for Reference I 1 which is not an approved LTR at the time submitted.

Response

Installation of the PRNM System is scheduled to occur during the spring 2009 RFO. On March 16,2009, the unit is scheduled to shutdown for the RFO.

Startup is projected, based on the current outage schedule, for April 18, 2009.

As indicated in the cover letter to the LAR, following installation of the NUMAC PRNM System, the OPRM Upscale function (TS Table 3.3.1. I Function 2.9, i.e.,

the Option Ill stability solution, will operate in an "indicate only" mode for an initial monitoring period, projected to be for 90 days of steady-state operation after startup from the 2009 RFO.(*)

Plant operation during the OPRM Monitoring Period will rely on operator action to avoid regions where instability may occur, to exit such regions when necessary, and to detect an actual instability and take mitigating action by manual means.

Following NMC review and evaluation of the operating data from the monitoring period, the OPRM Upscale function will be enabled and connected to the Reactor Protection System. The following commitment was made:

The Oscillation Power Range Monitor (OPRM) Monitoring Period is projected to be from startup following the spring 2009 Refuel Outage to until 90 days of steady-state operation have been achieved after reaching full-power. NMC

2. The OPRM Monitoring Period is conservatively projected to end 90 days after start-up and achievement of steady-state operation, projected for on or about July 18, 2009.

Page 3 of 5

ENCLOSURE 3 RESPONSE TO THE JULY 1,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS will inform the NRC of any change to the duration of the OPRM Monitoring Period.

In reference to the second part of the RAI, the February 6, 2008, LAR, Reference IIstated:

GE Nuclear Energy, Licensing Topical Report (LTR) NEDC-33075-P-A, "General Electric Boiling Water Reactor Detect and Suppress Solution -

Confirmation Density (DSS-CD)," dated July 24, 2002.

Reference 12 in the LAR refers to the NRC safety evaluation which approved the latest revision of the LTR NEDC-33075-P-A, Revision 5, for DSS-CD on November 27,2006.

NRC letter to GE Nuclear Energy, "Final Safety Evaluation for General Electric Nuclear Energy (GENE) Licensing Topical Report (LTR) NEDC-33075-P-A, Revision 5, "General Electric Boiling Water Reactor Detect and Suppress Solution - Confirmation Density," (TAC No. MC1737) dated November 27,2006.

The PRNM System LAR was submitted on February 6,2008. The GE LTR NEDC-33075-P was approved on November 27,2006.

Page 4 of 5

ENCLOSURE 3 RESPONSE TO THE JULY 1,2008 REQUESTS FOR ADDITIONAL INFORMATION CONCERNING SETPOINT QUESTIONS REFERENCES

1. NMC letter to NRC, "License Amendment Request: Power Range Neutron Monitoring System Upgrade," (L-MT-08-004), dated February 6, 2008.

2. Email from P. Tam (NRC) to R. Loeffler (NMC) dated July 1, 2008, "Monticello -

Draft Reactor Systems RAI re. Proposed Amendment on PRNM (TAC MD8064)

--- Enclosure 3 RAI Questions 1 through 3.

3. OG 02-0119-260, GE to BWR Owners' Group Detect and Suppress II Committee, "Backup Stability Protection (BSP) for Inoperable Option Ill Solution,"

dated July 2002.

4. BWROG-94078, "BWR Owner's Group Guidelines for Stability Interim Corrective Action," dated June 1994.

5. GE Nuclear Energy, Licensing Topical Report (LTR) NEDC-33075-P-A, "General Electric Boiling Water Reactor Detect and Suppress Solution - Confirmation Density (DSS-CD)," dated July 24, 2002.

6. NRC letter to GE Nuclear Energy, "Final Safety Evaluation for General Electric Nuclear Energy (GENE) Licensing Topical Report (LTR) NEDC-33075-P-A, Revision 5, "General Electric Boiling Water Reactor Detect and Suppress Solution - Confirmation Density," (TAC No. MC1737) dated November 27, 2006.

Page 5 of 5

ENCLOSURE 4 MONTICELLO NUCLEAR GENERATING PLANT RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION FOR LICENSE AMENDMENT REQUEST FOR POWER RANGE NEUTRON MONITORING SYSTEM UPGRADE (TAC NO. MD8064)

MNGP TECHNICAL SPECIFICATION REVISION REPLACEMENT PAGES FOR TS INSERTS 2 Pages Follow

REPLACEMENT INSERT 5: Adds LSSS Notes APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

2. Average Power Range Monitors

a. Neutron Flux - High, 2 3'" G SR 3.3.1.1.1 < 20% RTP (Setdown) SR 3.3.1.1.4 SR 3.3.1.1.6 SR 3.3.1.1.11 SR 3.3.1.1.15

b. Simulated Thermal Power 1 3(c) F SR 3.3.1.1.1 5 0.66 W + 61.6%

- High SR 3.3.1.1.2 RTP(~)

SR 3.3.1.1.4 and SR 3.3.1.1.6 1 116% RTP SR 3.3.1.1.11 SR 3.3.1.1.15

c. Neutron Flux - High F SR 3.3.1.1.1 1122% RTP SR 3.3.1.1.2 SR 3.3.1.1.4 SR 3.3.1 .I . 6 F SR 3.3.1.1.&

SR 3.3.1.1.15

e. 2-Out-Of-4 Voter 1,2 2 G SR 3.3.1.1.1 NA SR 3.3.1.1.4 SR 3.3.1.1.12 SR 3.3.1.1.14 SR 3.3.1.1.15

f. OPRM Upscale r 20% RTP 3(c) I SR 3.3.1.1.1(~) As specified SR 3.3.1.1.4(~) in COLR SR 3.3.1 .I .6(e)

SR 3.3.1.1.11(~)

SR 3.3.1 .I .15'~)

SR 3.3.1.1.16'~)

(b) 0.66 (W - Delta W) + 61.6% RTP when reset for single loop operation per LC0 3.4.1, "Recirculation Loops Operating." The cycle-specific value for Delta W is specified in the COLR.

(c) Each APRM IOPRM channel provides inputs to both trip systems.

(e) During the OPRM Monitoring Period the OPRM Upscale function is inoperable. Upon successful completion of the OPRM Monitoring Period (which includes time for review and acceptance of the OPRM online data by NMC) the OPRM Upscale function will initially be declared OPERABLE on-line.

Initial declaration of OPERABILITY is based upon factory acceptance testing, post-modification testing (including full or partial-surveillance performance during the RFO or during operation, as applicable), and industry experience with the PRNM System.

First performance of these new surveillance requirements is due at the end of the first surveillance interval, Upscale function was initially declared OPERABLE following the 2009 RFO.

--L-If the as-found channel setpoint is not the Nominal Trip Allowable Value, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

t (g) The instrument channel setpoint shall be reset to the Nominal Trip Setpoint at the completion of the surveillance; otherwise, the channel shall be declared inoperable. The NTSP and the methodology used to determine the NTSP are specified in the Technical Requirements Manual.

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

Control Rod Block lnstrumentation 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 As specified in 1 a COLR SR 3 . 3 . 2 .(h)(i)

SR 3.3.2.1.5 I

b. Intermediate Power Range -

Upscale (b) 2 SR 3.3.2.1

.o As specified in SR 3.3.2.1 (h)(l) COLR SR 3.3.2.1.5 I

c. High Power Range - Upscale (C), (dl 2 SR 3.3.2.1 As specified in 1 0 COLR SR 3 . 3 . 2 .(h)(i SR 3.3.2.1.5 I

d. lnop (dl, (el 2 SR 3.3.2.1. I NA

2. Rod Worth Minimizer

3. Reactor Mode Switch - Shutdown (9 2 SR 3.3.2.1.7 NA Position (a) THERMAL POWER 2 30% and < 65% RTP and MCPR is below the limit specified in COLR.

(b) THERMAL POWER 2 65% and < 85% RTP and MCPR is below the limit specified in COLR.

(c) THERMAL POWER 2 85% and < 90% RTP and MCPR is below the limit specified in COLR.

(d) THERMAL POWER 2 90% R I P and MCPR is below the limit specified in COLR.

(e) THERMAL POWER 2 30% and < 90% RTP and MCPR is below the limit specified in COLR.

(f) With THERMAL POWER 5 10% RTP Value, then the channel shall be evaluated to verify that it is functioning as required before returning the channel (i) The instrument channel setpoint shall be reset to the Nominal Trip Setpoint at the completion of the surveillance; otherwise, the channel shall be declared inoperable. The NTSP shall be specified in the COLR. The methodology

~ h ~ Z ob d w ~n c ~

Monticello . B.bbie~,~,,,, Amendment No. 446, -

ENCLOSURE 5 MONTICELLO NUCLEAR GENERATING PLANT RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION FOR LICENSE AMENDMENT REQUEST FOR POWER RANGE NEUTRON MONITORING SYSTEM UPGRADE (TAC NO. MD8064)

MNGP TECHNICAL SPECIFICATION BASES REPLACEMENT OR ADDITIONAL INSERTS 16 Pages Follow

2.c. Averaae Power Ranae Monitor Neutron Flux - Hiah The Average Power Range Monitor Neutron Flux-High Function is capable of generating a trip signal to prevent fuel damage or excessive RCS pressure. For the overpressurization protection analysis of Reference 9, high neutron flux is assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the safetylrelief valves (SlRVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 10) takes credit for high neutron flux to terminate the CRDA. The Allowable Value is based on the Analytical Limit assumed in the CRDA analyses.

The Average Power Range Monitor Neutron Flux-High Function is required to be OPERABLE in MODE Iwhere the potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being exceeded. Although the Average Power Range Monitor Neutron Flux-High Function is applicable in MODE 2, the Average Power Range Monitor Neutron Flux-High (Setdown) Function conservatively bounds the assumed trip and, together 2.d. Averaae Power Ranae Monitor l n o ~

This Function (Inop) provides assurance that the minimum numbers of APRM channels are OPERABLE.

For any APRM channel, any time its mode switch is in any position other than "Operate," an APRM module is unplugged, or the automatic self-test system detects a critical fault with the APRM channel, an lnop trip is sent to all four voter channels. lnop trips from two or more unbypassed APRM channels result in a trip output from all four voter channels to their associated trip system.

This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

There is no Allowable Value for this Function.

This Function is required to be OPERABLE in the MODES where the APRM Functions are required.

2.e. 2-Out-0f-4 Voter The 2-Out-0f-4 Voter Function provides the interface between the APRM Functions, including the OPRM Upscale Function, and the final RPS trip system logic. As such, it is required to be OPERABLE in the MODES where the APRM Functions are required and is necessary to support the safety analysis applicable to each of those Functions. Therefore, the 2-Out-0f-4 Voter Function needs to be OPERABLE in MODES 1 and 2.

All four voter channels are required to be OPERABLE. Each voter channel includes self-diagnostic functions. If any voter channel detects a critical fault in its own processing, a trip is issued from that voter channel to the associated trip system.

The 2-0ubOf-4 Voter Function votes APRM Functions 2.a, 2.b, and 2.c independently af Function 2.f. The voter also includes separate outputs to RPS for the two independently voted sets of Functions, each of which is redundant (four total outputs). The voter Function 2.e must be declared inoperable if any of its functionality is inoperable. However, due to the independent voting of APRM trips, and the redundancy of outputs, there may be conditions where the voter Function 2.e is inoperable, but trip capability for one or more of the other APRM Functions

NEW INSERT - RPS A In accordance with the guidance of Regulatory Issue Summary 2006-17 (Reference 23) Reactor Protection System function APRM Neutron Flux - High (Function 2c) is a Limiting Safety System Setting (LSSS).

RPS Instrumentation Note 3 is adde 3.3.1. I .I 1 to clarify that the recirculation B 3.3.1 . I eed the APRMs are included in the Channel BASES Calibration.

SURVEILLANCE REQUIREMENTS (continued)

\

passive devices, with minimal d simulating a meaningful signal.

sensitivity are compensated for calibration (SR 3.3.1. I .2) and th Note 2 to SR 3.3.1 .I . I 1 requires th 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of entering MODE 2 from a reasonable time in which to complete the SR.

The Frequency of SR 3.3.1. I .9 is based upon the assumption of a 92 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. The Frequency of SR 3.3.1 . I . I 1 is based The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods (LC0 3.1.3, "Control Rod OPERABILITY"), and SDV vent and drain valves (LC0 3.1.8, "Scram Discharge Volume Vent and Drain Valves"), overlaps this Surveillance to provide complete testing of the assumed safety function.

The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month ~ r e ~ u e n c ~ .

Add new paragraph: "The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e simulates APRM and OPRM trip conditions at the 2-out-of-4 voter channel inputs to check all combinations of I two tripped inputs to the 2-out-of-4 logic in the voter channels and APRM related redundant RPS relavs." I Monticello B 3.3.1.1-25 Revision No. 0

i SR 3.3.1.1.1 1 for the following RPS function(s) is modified by two Notes as identified in Table 3.3.1 .I-1. The function(s) listed below are LSSS for the protection of the reactor core Safety Limits.

Function No. RPS Function i

/ 2.c APRM Neutron Flux - High The first Note requires evaluation of channel performance for the condition where the as-found setting for the channel setpoint is not the NTSP but is conservative with respect to the Allowable Value. Evaluation of instrument performance will verify that the instrument will continue to perform in accordance with design basis assumptions.

The purpose of the assessment is to ensure confidence in the instrument performance prior to returning the instrument to service. This nonconformance will be entered into the Corrective Action Program. Entry into the Corrective Action Program will ensure required review and documentation of the condition for continued OPERABILITY.

The second Note requires that the as-left setting for the instrument be returned to the NTSP. If the as-left instrument setting cannot be returned to the NTSP, then the instrument channel shall be declared inoperable. The NTSP and the methodology used to determine the NTSP for the APRM Neutron Flux - High Function, (Function 2.c) in Table 3.3.1.I-1 are specified in Appendix C to the Technical Requirements Manual, a document controlled under 10 CFR 50.59.

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)

This SR ensures that scrams initiated from the Turbine Stop Valve -

Closure and Turbine Control Valve Fast Closure, Acceleration Relay Oil Pressure Low Functions will not be inadvertently bypassed when THERMAL POWER is > 45% RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the main turbine bypass valves must remain closed during in-service calibration at THERMAL POWER > 45% RTP, if peiforming the calibration using actual turbine first stage pressure, to ensure that the calibration is valid. The pressure switches are normally adjusted lower (30% RTP) to account for the turbine bypass valves being opened, such that 14% of the THERMAL POWER is being passed directly to the condenser.

If any bypass channel's setpoint is nonconse~ative(i.e., the Functions are bypassed at > 45% RTP, either due to open main turbine bypass valve(s) or other reasons), then the affected Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Acceleration Relay Oil Pressure

- Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition, this SR is met and the channel is considered OPERABLE.

The Frequency of 24 months is based on engineering judgment and reliability of the components.

This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. RPS RESPONSE TlME may be verified by actual response time measurements in any series of sequential, overlapping, or total channel measurements.

The RPS RESPONSE TlME acceptance criterion is 50 milliseconds.

RPS RESPONSE TlME tests are conducted on a 24 month STAGGERED TEST BASIS. Note 1 requires STAGGERED TEST BASIS Frequency to be determined based on 4 channels per trip system, in lieu of the 8 channels specified in Table 3.3.1.1-1for the MSlV -

his Frequency is based on the logic interrelationships B 3.3.1 .I-26 Revision No. 0

REPLACEMENT INSERT B5A:

APRM and OPRM RESPONSE TIME tests are conducted on a 24 month STAGGERED TEST BASIS. Note 1 requires the STAGGERED TEST BASIS to be determined based on 4 channels of APRM outputs and 4 channels of OPRM outputs, (total n = 8) being tested on an alternating basis.

This allows the STAGGERED TEST BASIS Frequency for Function 2.e to be determined based on 8 channels rather than the 4 actual 2-Out-Of-4 Voter channels.

The redundant outputs from the 2-Out-Of-4 Voter channel (2 for APRM trips and 2 for OPRM trips) are considered part of the same channel, but the OPRM and APRM outputs are considered to be separate channels for application of SR 3.3.1.1.14, so N = 8. The note further requires that testing of OPRM and APRM outputs from a 2-Out-Of-4 Voter be alternated. In addition to these commitments, References 17 and 21 require that the testing of inputs to each RPS Trip System alternate.

Combining these frequency requirements, an acceptable test sequence is one that:

1 a.

b.

Tests each RPS Trip System interface every other cycle, Alternates the testing of APRM and OPRM outputs from any specific 2-Out-Of-4 Voter channel, and

c. Alternates between divisions at least every other test cycle.

\

After 8 cycles, the sequence repeats.

Each test of an OPRM or APRM output tests each of the redundant outputs from the I

2-Out-Of-4 Voter channel for that Function and each of the corresponding relays in the RPS. Consequently, each of the RPS relays is tested every fourth cycle. The RPS relay testing frequency is twice the frequency justified by References 17 and 21.

RPS Instrumentation B 3.3.1. I BASES

- - ~ ~ p SURVEILLANCE REQUIREMENTS (continued) of the various channels required to produce an RPS scram signal. The 24 month Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience, which shows that random Add Inserts B6 and B7 failures of instrumentation components causing serious response time egradation, but not channel failure, are infrequent occurrences.

REFERENCES 1. Regulatory Guide 1. I 05, Revision 3, "Setpoints for Safety-Related lnstrumentation."

3. USAR, Section 7.6.1.2.5.

4. USAR, Chapter 14.

5. USAR, Chapter 14A.

6. USAR, Section 7.8.2.1.

7. USAR, Section 7.3.4.3.

8. ~~44& Not used.

9. USAR, Section 14.5.1.

10. USAR, Section 14.7.1.

11. USAR, Section 14.7.2.

12. USAR, Section 14.7.3.

13. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1, 1980.

14. USAR, Section 14.4.5.

15. USAR, Section 14.4.1.

16. NEDC-30851-P-A , "Technical Specification Improvement Analyses Revision No. 0

INSERT 87 This SR ensures that scrams initiated from OPRM Upscale Function (Function 2.9 will not be Inadvertently bypassed when THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is 2 25% RTP and core flow, as indicated by recirculation drive flow, is r 60% rated core flow. This normally involves confirming the bypass setpoints. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. The actual surveillance ensures that the OPRM Upscale Function is enabled (not bypassed) for the correct values of APRM Simulated Thermal Power and recirculation drive flow. SR 3.3.1.1.1 1 and the MNGP core flow measurement system calibration procedure ensure that the APRM Simulated Thermal Power and recirculationflow properly correlate with THERMAL POWER and core flow, respectively.

If any bypass setpoint is non-conservative (i.e., the OPRM Upscale Function is bypassed when APRM Simulated Thermal Power 2 25% and recirculation drive flows 60% rated), then the affected channel is considered inoperable for the OPRM Upscale Function. Alternatively, the bypass setpoint may be adjusted to place the channel in a conservative condition (non-bypass). If placed in the non-bypass condition, this SR is met and the channel is considered OPERABLE.

The Frequency of 24 months is based on engineering judgment and reliability of the components.

INSERT B8

17. NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function", October 1995.

18. NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

19. NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

20. NEDO-32465-A, "BWR Owners' Group Long-Term Stability Detect and Suppress Solutions Licensing Basis MethodologyAnd Reload Applications," August 1996.

21. NEDC-32410P-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)

Retrofit Plus Option Ill Stability Trip Function", November 1997.

Letter, LA England (BWROG) to MJ Virgilio, "BWR Owners' Group for Stability Interim Corrective Action", June 6, 1994.

i

23. U.S. NRC Regulatory Issue Summary 2006-17, "NRC Staff Position on the Requirements of 10 CFR 50.36, "Technical Specifications," Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels," dated August 24, 2006.

Control Rod Block Instrumentation B 3.3.2.1 BASES BACKGROUND (continued)

The purpose of the RWM is to control rod patterns during startup, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to 10% RTP. 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 actual sequence deviates beyond allowances from the stored sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses steam flow signals to determine when the reactor power is above the preset power level at which the RWM is automatically bypassed (Ref. 2). The RWM is a single channel system that provides input into both RMCS rod block circuits.

With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This Function prevents inadvertent criticality as the result of a control rod withdrawal during MODE 3 or 4, or during MODE 5 when the reactor mode switch is required to be in the shutdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a control rod block to all control rods.

I Rod Block Monitor  ?

Two channels of the RBM are required to be OPERABLE, with their setpoints within the appropriate Allowable Value for the associated power range, to ensure that no single instrument failure can preclude a rod block from this Function. The actual setpoints are calibrated consistent with applicable setpoint methodology.

Monticello B 3.3.2.1-2 Revision No. 0

NEW INSERT - RBM A Allowable Values are specified for each applicable Rod Block Function listed in Table 3.3.2.1-1. The NTSPs (actual trip setpoints) are selected to ensure that the setpoints are conservative with respect to the Allowable Value. A channel is inoperable if its actual trip setpoint is non-conservative with respect to its required Allowable Value.

NTSPs are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor power), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The Analytical Limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the Analytical Limits, corrected for calibration, process, and some of the instrument errors. The NTSPs are then determined, accounting for the remaining channel uncertainties. The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, and drift are accounted for. The Limiting Trip Setpoint is the value of the setpoint within its specified as-found tolerance which most closely approaches the Allowed Value. For the Rod Block Monitor, which is a digital system with a zero as-found tolerance, the Limiting Trip Setpoint is the NTSP.

The Rod Block Monitor Low, Intermediate and High Power Range - Upscale functions (Functions l a , Ib and Ic, respectively) are Limiting Safety System Setting (LSSS).

The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.

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

Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not excee the Allowable Values between successive CHANNEL CALIBRATIONS.

Operation with a trip setpoint less conservative than the nominal t r i ~

setpoint, but within its Allowable Value, is acceptable.@p setpoints are sequences of an RWE event power level, the consequences PR SL and, therefore, the RBM is not required to be OPERABLE (Ref. 3). When operating < 90% RTP, analyses have shown that with an initial MCPR 2 1.75, no RWE event will result in exceeding the MCPR SL. Also, the analyses demonstrate that when operating at 2 90% RTP with MCPR 2 1.44, no RWE event will result in exceeding the MCPR SL. Therefore, under these conditions, the RBM is also not required to be OPERABLE.

2. Rod Worth Minimizer The RWM enforces the banked position withdrawal sequence (BPWS) to ensure that the initial conditions of the CRDA analysis are not violated.

~ ~~~~p Monticello B 3.3.2.1-3 Revision No. 0

/ NEW INSERT RBM B

, the calculated RBM flux (RBM channel signal). When the normalized RBM flux value exceeds the applicable trip setpoint, the RBM provides a trip output. 7 For the digital RBM, there is a normalization process initiated upon rod selection, so that only RBM input signal drift over the interval from the rod selection to rod movement needs to be considered in determining the nominal trip setpoints. The RBM has no drift characteristic with no as-left or as-found tolerances since it only performs digital calculations on the digitized input signals provided by the APRMs.

The NTSP (or Limiting Trip Setpoint) is the Limiting Safety System Setting since the RBM has no drift characteristic. The RBM Allowable Value demonstrates that the analytic limit would not be exceeded, thereby protecting the safety limit. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and environment errors are accounted for and appropriately applied for the RBM. There are no margins applied to the RBM nominal trip setpoint calculations which could mask RBM degradation.

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

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 entire 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 Specificationstests at least once per refueling interval with applicable extensions. The CHANNEL FUNCTIONAL TEST for the RWM is performed by: a) attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs; b) verifying proper annunciation of the selection error of at least one out-of-sequence control rod in each fully inserted group; and c) performing a RWM computer on-line diagnostic test. As noted in the SRs, SR 3.3.2.1.2 is not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after any control rod is withdrawn at I10% RTP in MODE 2, and SR 3.3.2.1.3 is not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after THERMAL POWER is I 10% RTP in MODE 1. This allows entry into MODE 2 for SR 3.3.2.1.2, and entry into MODE 1 when THERMAL POWER is S 10% RTP 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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. 8).

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.

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. I .2 and SR 3.3.1. I .6.

The Frequency is based upon the assumption of a interval in the determination of the magnitude of Monticello B 3.3.2.1-8 Revision No. 0

( NEW INSERT - RBM D SR 3.3.2.1.4 for the following RBM functions is modified by two Notes as identified in Table 3.3.2.1-1. These RBM functions are LSSS for the protection of the reactor core Safety Limits.

Function No. RBM Function Low Power Range - Upscale Intermediate Power Range - Upscale High Power Range - Upscale The first Note requires evaluation of channel performance for the condition where the as-found setting for the channel setpoint is not the NTSP but is conservative with respect to the Allowable Value. For digital channel components, no as-found tolerance or as-left tolerance can be specified. Evaluation of instrument performance will verify that the instrument will continue to behave in accordance with design basis assumptions. The purpose of the assessment is to ensure confidence in the instrument performance prior to returning the instrument to service. This nonconformance will be entered into the Corrective Action Program. Entry into the Corrective Action Program will ensure required review and documentation of the condition for continued OPERABILITY.

The second Note requires that the as-left setting for the instrument be returned to the NTSP. If the as-left instrument setting cannot be returned to the NTSP, then the instrument channel shall be declared inoperable. The NTSPs and Allowable Values for Rod Block Monitor Functions l a , Ib and I c are specified in the COLR. The methodology used to determine the NTSPs are specified in Appendix C to the Technical Requirements Manual, a document controlled under 10 CFR 50.59.

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

The ~ e t ar-ayl s va*as a function of power.

Thre llowable Values required in Table 3.3.2.1-1, each within a specific power range, are specified in the COLR. The power at which the control rod block Allowable Values automatically change are based on the APRM signal's input to each RBM channel. Below thi minimum power setpoint, the RBM is automatically bypassed. ~ h e s h ~ ~ setpointsa s s must be verified periodically to be 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 .I .2 and SR 3.3.1. I .6. The Frequency is based on the actual trip setpoint methodology utiliz The RWM is automatically bypassed when power is above a specified value. The power level is determined from steam flow signals. The automatic bypass setpoint must be verified periodically to be > 10% 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 24 month Frequency is based on engineering judgment considering the reliability of the components, and that indication of whether or not the RWM is bypassed is provided in the control room.

A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch - Shutdown Position Function to ensure that the entire 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 Monticello Revision No. 0

ENCLOSURE 6 MONTICELLO NUCLEAR GENERATING PLANT RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION FOR LICENSE AMENDMENT REQUEST FOR POWER RANGE NEUTRON MONITORING SYSTEM UPGRADE (TAC NO. MD8064)

CA-08-050 REVISION 0 INSTRUMENT SETPOINT CALCULATION - AVERAGE POWER RANGE MONITOR (APRM) NON-FLOW BIASED PRNM SETPOINTS FOR CLTP AND EPU 143 Pages Follow

Document Information Rev. 2 Calculation Signature Sheet e Pa e l o f 8

) Special Codes: Safeguards Proprietary I Type: Calc Sub-Type:

I I NOTE: Print and sign name in signature blocks, as required.

I) Vendor Name or Code: ( Vendor Doc No: I Description of Revision: Original Issue Method Used ( F O ~D V O ~ I ~ ) : Review Alternate Calc Test 1 Minor Revisions I Descriptionof Change:

Pages Affected:

Prepared by: I Date:

Reviewed by: I Date:

Type of Review: Design Verification Tech Review Vendor Acceptance Method Used (For DV Only): Review Alternate Calc Test Approved by: ( Date: -

(continued on next page)

Record Retention: Retain this form with the associated calculation for the life of the plant.

QF-0549 (FP-E-CAL-01), Rev. 2 Page 1 of 8 Calculation Signature Sheet N

M?

Document Information I

Title:

Instrument Setpoint Calculation - Average Power Range Monitor (APRM)

Non-Flow

- Biased PRNM Setpoints for CLTP and EPU Facility: (XI MT PB PI PL C] HUIFT I unit: [Xi 1 02 I

Safety Class: SR Aug Q 0Non SR I Special Codes: Safeguards Proprietary 1 Type: Calc Sub-Type:

I

( NOTE: I Print and sign .name in signature blocks, as required.

I I

Vendor Name or Code: 1 Vendor Doc No:

IDescription of Revision: Original Issue 1

-Prepared by: Joseph Balitski Date: 08/05/08 Reviewed by: Date:

IType of Review: [XI Design Verification Tech Review Vendor Acceptance (

Method Used (For DVOIII~): ~ e v i e Alternate a Calc Test Approved by: I Date:

Minor Revisions L ~ i n o Rev.

r- No: - - -

I Description of Change:

Paaes Affected:

(continued on next page)

Record Retention: Retain this form with the associated calculation for the life of the plant.

QF-0549 (FP-E-CAL-OI), Rev. 2 Page 2 of 8 Calculation Signature Sheet Record Retention: Retain this form with the associated calculation for the life of the plant.

reference table is used for data entry into the Passport Controlled Documents Module, (C012 Panel). It may also be used as the reference section of the calculation. The input documents, output documents and other references should all be listed here. Add additional lines as needed.

Reference Documents (Passport C012 Panel from C020)

Contro"ed* Doc Ref Type**

Document Name Document Number

1. Doc?+Type (if known) 1 PROC Engineering Standards Manual, Appendix I (GE mlnput OOutput ESM-03.02-APP-I I

Methodology Instrumentation & Controls) 2 GDOC Average Power Range Monitor Selected PRNM mlnput OOutput 0000-0077-9068 Licensing Setpoint - CLTP Operation (NUMAC) 3 GDOC Average Power Range Monitor Selected PRNM (Xllnput 00utput 0000-0081-6958 0 0 - -

7 - MNGP PRNM Licensing Setpoints - CLTP GEH-NE-0000- [XJlnput UOutput I

Operation 0076-2388 5 0GDOC Nuclear Measurement Analysis and Control B i n p u t OOutput Power Range Nuclear Monitor (NUMAC PRNM) NEDC-32410-A, Oct Retrofit Plus Option Ill Stability Trip Function, Volume 1 1995 Licensing Topical Report 6 GDOC Nuclear Measurement Analysis and Control [Xllnput 00utput Power Range Nuclear Monitor (NUMAC PRNM) NEDC-32410-A, Oct Retrofit Plus Option Ill Stability Trip Function, Volume ll 1995 Licensing Topical Report

' 7 GDOC Nuclear Measurement Analysis and Control B l n p u t UOutput Power Range Nuclear Monitor (NUMAC PRNM) NEDC-32410-A, Nov Retrofit Plus Option Ill Stability Trip Function, Supplement I 1997 Licensing Topical Report 8 CALC Instrument Setpoint Calculation - Average Power O h p u t UOutput Range Monitor (APRM) Flow-Biased Upscale CA-96-224 1 I Scram and Rod Block 1 Record Retention: Retain this form with the associated caIculation for the life of the plant.

QF-0549 (FP-E-CAL-0I), Rev. 2 Page 4 of 8 Calculation Signature Sheet N M?

g CALC Instrument Setpoint Calculation - APRM n l n p u t OOutput Downscale CR Block CA-05-153 0 10 GDOC General Electric Instrumentation Setpoint NEDC-31336P-A Oct Mlnput Ooutput Methodology Class Ill 1986 11 [XJ PROC APRM Heat Balance Calibration 0017 25 n l n p u t UOutput j2 GDoC GE-NE-901-021- Nlnput Ooutput Setpoint Calculation Guidelines far the Monticello Oct 0492, DRF AOo-Nuclear Generating Plant 1992 01932-1 13 NLIc Monticello Plant Technical Specifications Tech-Specs 155 nlnput OOutput 14 LIC Monticello Technical Requirements Manual TRM 2 Olnput UOutput 15 U L l S Instrument Setpoints for Safety-Related Systems RG 1.105 3 ulnput OOutput 16 N L I s Project Task Report, MNGP Extended Power Uprate, Technical Specifications Setpoints Task Report T0506 , mlnput OOutput 17 Lls Clarify Application of Setpoint Methodology for Dlnput UOutput LSSS Functions. Not approved but included as a TSTF-493 3

-- ---- setpoint reference document I8 LIs NRC Staff Position on the Requirements of t 0 n l n p u t UOutput CFR 50.36, "Technical Specifications," Regarding Aug RIS 2006-,,

Limiting Safety System Settings During Periodic 2006 Testing and Calibration of Instrument Channels 19 GDOC EPU - Mod 4 - Neutron Monitoring System mlnput [?Output (PRNM) EC-10856 0 20 a GDOC PRNMS Setpoint Calculations 050 (Non-Flow Biased Setpoints)

EC-I 2899 0 Olnput MOutput 21 a PROC Rod Withdraw Block C.6-005-A-03 I ulnput MOutput 22 kid PROC APRM Downscale C.6-005-A-06 3 ulnput Boutput 23 [XJ PROC APRM Hi Hi INOP Ch 1 , 2 , 3 C.6-005-A-22 3 Olnput NOutput 24 bd PROC APRM Hi Hi INOP Ch 4, 5, 6 C .6-005-A-30 3 Olnput NOutput Record Retention: Retain this form with the associated calculation for the life of the plant.

QF-0549 (FP-E-CAL-OI), Rev. 2 Page 5 of 8 Calculation Signature Sheet

' 25 PROC -

Operations Manual Section Plant Protection 8,05.06-02 18 Olnput m ~ u t p u t System 26 PROC Operations Manual Section - Power Range Olnput [XJOutput B.05.01.02-02 6 Neutron Monitoring 27 PROC Operations Manual Section - Power Range Olnput [Xloutput B.05.01.02-05 16 Neutron Monitoring 28 PROC Design Bases Document for Neutron Monitoring Olnput N o u t p u t DBD B5. C System 29 PROC APRM Calibration Readjustment for Single Loop 821 1 2 Olnput NOutput 30 PROC APRM Calibration Readjustment for Single Loop 8212 2 ulnput NOutput 31 a PROC APRMlFlow Reference Scram Functional Check 0012 41 Olnput Houtput

'32 GDOC Specification for existing Neutron Monitoring [XJlnput UOutput 257HA594 I System, 12/03/85 33 GDOC Specification for PRNM MUMAC Power Range @Input OOutput 24A5221 14 Neutron Monitor System 34 GDoC DIR TO500 alnput UOutput Design Input Request T0500, Neutron Monitoring DRF 0000-0040- 3 System 9168 35 U GDOC Mathematics of Physics and Modern Engineering, na

, 1966 [Xllnput OOutput I.S. Sokolnikoff and R. M. Redheffer 37 jX/ LIc Monticello Plant Technical Specifications Bases Bases 8 Olnput OOutput

• Controlled Doc checkmark means the reference can be entered on the C012 panel in black. Unchecked lines will be yellow. If checked, also list the Doc Type, e.g., CALC, DRAW, VTM, PROC, etc.)

• • Corresponds to these Passport "Ref Type" codes: InputstBoth = ICALC, Outputs = OCALC, Otherfunknown = blank)

Record Retention: Retain this form with the associated calculation for the life of the plant.

QF-0549 (FP-E-CAL-OI), Rev. 2 Page 6 of 8 Calculation Signature Sheet Other Passport Data Associated System (Passport C011, first three columns) OR Equipment References (Passport C025,all five columns):

Facility Unit System Equipment Type Equipment Number MT I NIP INDREC APRM-I MT 11 1 NIP 1 INDREC / APRM-2 1

/ I I 1 1 MT 1 NIP INDREC APRM-3 MT I I I NIP 1 INDREC I APRM-4 1 MT 11 I NIP / INDREC I LPRM-04-29 I MT II I NIP 1 INDREC I LPRM-12-13 MT II I NIP 1 INDREC I LPRM-12-21 I MT 11 / NIP 1 INDREC / LPRM-12-29 I MT 1 NIP 1 1

INDREC LPRM-I 2-37 MT I NIP INDREC LPRM-20-13 MT II I NIP 1 INDREC I LPRM-20-21 I MT 11 I NIP 1 INDREC I LPRM-20-29 -

1 MT II I NIP 1 INDREC 1 LPRM-20-37 I MT 1 NIP 1 INDREC LPRM-20-45 MT 1 NIP 1 INDREC LPRM-28-05 MT I NIP 1 INDREC LPRM-28-13 MT I I / NIP 1 INDREC I LPRM-28-21 I Record Retention: Retain this form with the associated calculation for the life of the plant.

QF-0549 (FP-E-CAL-0I ) , Rev. 2 Page 7 of 8 Calculation Signature Sheet NM">

MT 1 Nf P INDREC I LPRM-28-29 MT I NIP 1 INDREC LPRM-28-37 MT 1 NIP JNDREC LPRM-28-45 MT 1 NIP INDREC LPRM-36-13 MT I NIP 1 INDREC LPRM-36-2 1 MT 1 NIP 1 INDREC LPRM-36-29 MT I NIP I INDREC LPRM-36-37 MT I NIP 1 INDREC LPRM-36-45 MT I NIP 1 INDREC LPRM-44-2 I MT I NIP INDREC LPRM-44-29 -

MT I NIP INDREC LPRM-44-37 Superseded Calculations (Passport COI 9):

1 Facility I Calc Document Number / Title 1 MT CA-05-153 -

Instrument Setpoint Calculation Average Power Range Monitor (APRM) Downscale CR Block. This calculation is not affected until the installation of the PRNMS retrofit, Instrument Setpoint Calculation Average Power Range Monitor (APRM) Flow-Biased Upscale Scram and Rod Block. This calculation is not affected until the installation of the PRNMS retrofit, EC-I 0856.

Record Retention: Retain this form with the associated calculation for the life of the plant.

QF-0549 (FP-E-GAL-01), Rev. 2 Page 8 of 8 Calculation Signature Sheet NM-Description Codes Optional (Passport C018):

Code Description (optional) Code Description (optional)

Notes (Nts) = Optional (Passport X293 from C020):

I Topic Notes 1 Text I I

1 Calc Introduction Copy directly from the calculation lntro Paragraph or See write-up below Record Retention: Retain this form with the associated calculation for the life of the plant.

QF-0527 (FP-E-MOD-07) Rev. 0 Document Number1Title:

CA-08-050 IInstrument Setpoint CalculationAverage Power Range Monitor (APRM)

Non Flow-Biased PRNM Setpoints for CLTP and EPU Verifier's Name1Discipline: Charles E. Nelson, Engineering Projects Support DESIGN REVIEW CONSIDERATIONS: M m N / A

1. Were the inputs correctly selected and incorporated into design? I X I O O

2. Are assumptions necessary to perform the design activity adequately described and [XI q reasonable? Where necessary, are the assumptions identified for subsequent re-verifications when the detailed design activities are completed?

3. Are the appropriate quality and quality assurance requirements specified? N U 0

4. Are the applicable codes, standards, and regulatory requirements includin issue 9 IXI 17 and addends properly identified and are thew requirements for design met.

5. Have applicable construction and operating experience been considered? q IXI

6. Have the design interface requirements been satisfied? o n [ X I

7. Was an appropriate design method used?

8. Is the output reasonable compared to inputs?

9. Are the specified parts, equipment and processes suitable for the required [XI application?

Are the specified materials compatible with each other and the design environmental conditions to which the material will be exposed?

Have adequate maintenance features and requirements been specified? n o [ X I Are accessibility and other design provisions adequate for performance of needed q IXi maintenance and repair?

Has adequate accessibility been providy to perform the in-service inspection expected to be required during the plant Ilfe?

u o m Has the design properly considered radiation exposure to the public and plant O O I X I personnel?

Are the acceptance criteria incorporated in the design documents sufficient to allow tXI C]

verification that design requirements have been sat~sfactorilyaccompl~shed?

Have adequate pre-operational, subsequent periodic test, and inspection IXI 17 requirements been appropriately specified, including acceptance criteria?

Are adequate handling, storage, cleaning, and shipping requirements specified? o o a Are adequate identificationrequirements specified? 0 [ 7 [ X I Are requirements for record preparation, review, approval, and retention adequately q IX) specified?

COMMENTS: C] None [XI Attached (Use Form QF-0528)

Page 1 of 1

QF-0528 (FP-E-MOD-07) Rev. 0 Design Review Comment Form NMC>

Committed to Nuclear Excellence Fleet Modification Process Sheet 1 of 9 DOCUMENT NUMBER1TITLE: CA-08-050 Instrument Setpoint Calculation- Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU REVISION: Q DATE: 7/9/2008 Note: Following classification of comments is used:

(E) Editorial (P)

. . Preference/Recommendation (T) Technical ITEM I REVIEWER'S COMMENTS PREPARER'S REVIEWER'S RESOLUTION DISPOSITION Process (T) Training matrix now Preparer not yet in Qualification shows qualification Matrix, therefore, not yet qualified to prepare calculation in Engineering Qualification Matrix, but paperwork has been submitted to Engineering Director for Approval.

Process (T) 19 CFR 50.59 screening needs to be completed under the Modification EC.

General (E)

Text margins and Page size are not matched in the .pdf version.

General (P) Added list of Add list of Acronyms and Acronyms and Abbreviations Abbreviations.

General (P) Added more detail to Comments and Recommendations the calculation based sections of Inputs 4.2 and 4.3 contain on GEH PRNM discussions and detailed explanations documents that may be useful to copy into the calculation.

General (E) Standardized with %

In various places units are listed as RTP through-out

%, % Power or % RTP. Define terms if different or standardize usage.

General (P) Used Tech Spec Standardize the terms used for the terminology through-four setpoints. Recommend using the out Page 1 of 9

QF-0528 (FP-E-MOD-07)Rev. 0 Design Review Comment Form 1 II 1I Fleet Modikation Process Sheet 2 of 9 Tech Spec names or names used in NRC approved LTRs.

8. QF-0549 p l (T) Corrected to 10856 EC number incorrect;

9. QF-0549 p l (T) Review checked Check "ReviewJ'under Method Used.

10. QF-0549p4(T) This remains an open RIS-2006-17 and TSTF-493 are item because decision discussed in methodology and are has not been made on referenced in the calculation, but are RIS-2006-17 not used as inputs as indicated here. applicability. RIS-They probably should not be listed on 2006-17 and TSTF-the QF-0549. Since TSTF is not 493 are not indicated approved it probably shouldn't be as inputs.

mentioned. NRC issued the RIS and it documents a "method acceptable to the staff' but is not yet a commitment by MNGP.

11. QF-0549 pp3,4 (T) Added Passport DOC Document Types in first Column need TYPES to all to be standard DOC TYPES used in references listed Passport. All entries should have a TYPE. RG, RIS, and TSTF are not defined and should probably be TYPE "LIS." Default TYPE is "GDOC

12. QF-0549 p3 (P) Added entire Reference documents 2 through 7 documents and parts appear to meet the definition of of some documents "Obscure Reference" and should be as Attachments. See uploaded to Sharepoint under the EC TOC or attached to the calculation.

13. QF-0549 p3 Reference Documents All Section 5 (TI references have been List does not include all of the added references listed in section 5 of calculation.

14. QF-0549 p4 (T) GE documents are Reference Documents 9 and 12 not are controlled should probably be controlled documents.

15. QF-0549 p5 (T) ID'Sof all LPRMs Add LPRMs. added

16. Form 3494 (T) Added to Form 3494

QF-0528 (FP-E-MOD-07) Rev. 0 Design Review Comment Form NM" -)

CommHted to Nuclear Excellence Fleet Modiftcation Process Add information in "Other Comments:" per FP-E-CAL-01 Attachment 4 3.b.7).

17. 1 Purpose p2 0). Added statement to Include statement that the calculation Purpose section provides design basis setpoint calculation for compliance with setpoint control program and NRC commitment.

18. 1 Purpose p2 (T). Added calculations Per FP-E-CAL-01 Attachment 1 I.c. being deleted to identify calculations being deleted or Purpose superseded.

19. 2 Methodology p3 (T) Open item. Need staff Need MNGP staff agreement to agreement of using position stated on AFTIALT being set GEH or other basis for to zero in surveillance procedures. AFTIALT.

20. 4.2 Inputs p4 (E) Changed Spelling should be "Setpoints"

21. 4.3 Inputs p4 (E) Changed Spelling should be "Setpoints"

22. 4.8 Inputs p5 (E) Changed Spelling should be "manufacturer"

23. 4.5,4.6,and4.7lnputspp4,5(E) Changed Document Number should be "NEDC-3241OP-An

24. 4.9 Inputs (T) NIA FTR TO506 Rev 2 is pending. The revision does not affect the inputs used in this calculation.

25. 5.1 References p5 (E) Changed Spelling should be "Setpoints"

26. 5.2 References p5 (E) Changed Spelling should be "Setpoints"

27. 5.7 References (E) Changed Title of Calculation should be "Instrument Setpoint calculation, Average Power Range Monitor (APRM ) Flow-Biased Upscale Scram and Rod Block"

28. 5.7 References (E) Deleted issue date Issue date should be 4113/07 due to revision

29. 5.8 References (E) Deleted issue date

QF-0528 (FP-E-MOD-07) Rev. 0 4 I dl Design Review Comment Form Fleet Modification Process Sheet 4 of 9 Issue date should be 12113/07 due to revision

30. 5.12 References p6 (E) Changed Title should be "Instrument Setpoints for Safety-Related Instrumentation" 31 . 5.13 References p6 (T) NIA FTR TO506 Rev 2 is pending. The revision does not affect the inputs used in this calculation.

32. 5 . I 4 References p6 (E) Changed Document number should be NEDO-31336P-A and date should be September 1996

33. 5.15 References p6 (E) Changed Add space between % and RTP and delete extraneous comma

34. 5.16 References p6 (T) 5.15 is also Input 4.12 Extract appropriate information from and Attachment 8 this reference and attach to calculation as an obscure reference.

35. 5.18 References p6 (T) Added date and Add date of issue (18 Jan 08) and stated not approved include a note that TSTF is not yet approved.

36. 6 Assumptions p6 (P) Added reference to Assumptions may be implied Input 4.2 14.3 -

concerning seismic effect, RFIIEMI, Comment 4 in the environmental temperature uncertainty tables.

conditions, and radiation Comment 4 discusses environment. GE Reports specify seismic effect, some of effects as negligible, but the RFIIEMI, radiation, calculation needs to provide power supply effect references that these inputs are and humidity. Table applicable to MNGP or state these as updated temperatures assumptions.

37. 7.1 .Ip7 Channel Diagram (P) Channel Diagram Indicate that there are multiple updated to indicate LPRMs feeding inputs to the multiple LPRM inputs electronics.

38. 7.1.2 p7 Channel Function (T) Added 214 voter logic Per ESM-03.02-APP-I 5.1.2 Channel and other PRNM description should indicate that there APRM logic are four channels and 2-out-of-4 voter differences logic determines actual Scram.

QF-0528(FP-E-MOD-07) Rev. 0 I il NMCE CommUed to Nuclear ExceHence Design Review Comment Form Fleet Modification Process Sheet 5 of 9 Similar discussion for Rod Blocks.

39. 7.2.1 LPRM Detectors (T)

Per DIR TO500 R2 installed detectors are GE NA300 detectors. Device 1 information appears to be for Local Power Range Monitor Electronics not detectors. There are 24 strings of detectors with 4 detectors per string (96 detectors with one amplifier per detector). I believe that all equipment other than the detectors and drywell cabling will be replaced as part of PRNMS. Passport equipment database does not appear to contain information or IDSfor detectors.

Update IDSand model number with input from modification team. Correct reference since information is not in Passport as indicated. (See DIR TO500 R2.)

40. 7.2.1 LPRM Detector p8 (T)

Correct Reference for Process Element should be Section Iof lnput 4.'2 and 4.3

41. 7.2.1 p8 Device 1 (T) LPRM Detector Add data items required per ESM-03.02-APP-I 5.2.1. List as NIA if the data item is not applicable. Definition of lnput (neutron flux) and Output signals is required.

42. 7.2.1 p8 Device 1 (T) LPRM Detector ESM-03.02-APP-I 5.2.2 requires listing of Process and Physical Interfaces. LPRM detectors are exposed to in-core conditions.

Calibration is via TIP monitoring.

Provide data or explain why data isn't required for the above requirement.

43. 7.2.1 p8 Device 1 (T) LPRM Detector All listed error terms in ESM-03.02-AP-I 5.2.3 are to be explicitly addressed in each calculation.

Recommend that the specific

Design Review Comment Form Fleet Modification Process Sheet 6 of 9 abbreviations listed in 5.2.3 be used to demonstrate compliance.

44. 7.2.2 p8 Power Electronics (T) Added input and Refer to NEDC-32410P-A V1 5.3.17.2 output descriptions and 5.3.17.3 for description of input and output signals.

45. 7.2.2 p9 Power Electronics (P) Added Comment 11 Calibration Error discussion in Inputs to Section 7.2.2 4.2 and 4.3 are discussed in Section 4 Comment 11. Incorporate this discussion in the calculation.

46. 7.3.1.Ippl0,l ILoop Accuracy (T) Sections 7.2.1 and Result is correct, but does not clearly 7.2.2 have all the show that all uncertainty terms in parameters listed from ESM-03.02-APPI 5.2.3 equation are Section 5.2.3.

considered.

47. 7.3.1.2 p l 1 Drift DTE evaluation has DTE not addressed in the calculation been added per ESM-03.02-APP-I.

48. 7.3.1.3 through 7.3.1.5 pp11-12 (T) Open item. Decision lncorporate changes to ALT and AFT needs to be made on to address NRC position on basis for AFTIALT.

calculating these values for a digital Revised Section 2 to instrument system. Summary of state GEH method should also be included in methodology used for section 2 Methodology. Add Future AFTIALT.

Needs to update guidelines and procedures to document the method if different from current guidance.

49. 7.3.1.6, and 7.3.1.7 PEA (T) Attachments added See Comment #34. for several references specified in body of calculation. See TOC

50. 7.3.1.9 and 7.3.2 pp13, 14 Added description of Tabulations (P) the abbreviations to Abbreviations do not match the rest of the uncertainty tables 7.3.1 and are confusing

51. 7.4.1.1 p15 Required Margin Changed Computation includes DPEA ( . O M % )

which is incorrect. Should be APEA (0.267%).

52. 7.4.1 .Ip15 Minimum Required Changed to 1.334 Margin

QF-0528 (FP-E-MOD-07) Rev. 0 Ir II Design Review Comment Form NMC>

Commmed to ~ u d e aExcellence i

Fleet Modifcation Process Sheet 7 of 9 Value for CL in 7.3.1 was 1.334, but this computation used 1.333.

53. 7.4.1.2 p16 NTSP (E) Changed Spelling NPSP should be NTSP in first equation.

54. 7.4.1.4 p18 STA (T) Changed Incorrect value of CL.Should be 1.334%.

55. 7.4.1.4 p18 STA (T) Added Input 4.13, Per GE guidelines 4.5.4.b. Bias term Section 4.5.4.b and for APEA should be used in STA 4.5.9.b as basis for evaluation. Drift bias should not be bias term added included.

56. 7.4.1.4. p18 STA Added units Add units for CJ~TA

57. 7.4.1.4p18STA(T) Gave reference for STA test uses an a value of Operational Limit Operational Limit of 100%. This is not discussed or justified as an entry in the calculation.

58. 7.4.2 p19 Setdown Scram and Rod Deleted Block (E)

Delete extraneous paragraph number in the title

59. 7.4.2.1 p19 (AL to AV) (E) Revised Awkward wording. Revise similar to 7.4.2.2 or simply state that the "Minimum required margin between the AL and AV can be defined by:"

followed by the term equation.

60. 7.4.2.4 p21 LER Avoidance (E) Added units Add units for

61. 7.4.2.4 p21 LER Avoidance (E) Added units Add Units for NTSP*.

62. 7.4.2 STA Test Added STA An Operational Limit of 11% is calculation for this identified for the APRM Setdown section Scram. Calculation does not perform a STA test for this setpoint.

63. 7.4.3 p22 Downscale Rod Block (T) Revised discussion of lnput 4.3 establishes recommended the Downscale Rod AV and NTSP for EPU. Statement Block for both CLTP that this input establishes CLTP value and EPU sections.

QF-0528 (FP-E-MOD-07) Rev. 0 tl I 11 Design Review Comment Form N M Commftted io Nuclear Excellence Fleet Modifmtion Process Sheet 8 of 9 is incorrect. Input 4.2 did not address Revised discussion in Downscale Rod Block. Since EPU Purpose has no impact on this setpoint or its uncertainties, it is acceptable to use this as the basis for establishing the CLTP values. Discuss this in the calculation. This also needs to be included in the calculation Purpose.

64. 7.4.3.4 p24 LER Avoidance (E) Added units Add units for ~ L E R .

65. 7.5.1.2 p27 AM table (E) Re-formatted section Prevent orphaning the header and body of the table.

66. 7.5.1.3 p28 LER Avoidance (E) Added units Add units for (JLER.

67. 7.5.1.4 p29 STA (T) Added GE guidelines Per GE guidelines 4.5.4.b. Bias term 4.5.4.b and 4.5.9.b as for APEA should be used in STA basis for drift bias evaluation. Drift bias should not be included.

68. 7.5.1.4 p29 STA (T) Added reference for STA test uses an a value of Operational Limit of Operational Limit of 100%. This is not I00 %

discussed or justified as an entry in the calculation.

69. 7.5.2.1 p30 (AL to AV) (E) Revised to clarify Awkward wording. Revise similar to wording 7.5.2.2 or simply state that the "Minimum required margin between the AL and AV can be defined by:"

followed by the term equation.

70. 7.5.2 p30 STA Test (T) Added STA Test for An Operational Limit of 11% is this section identified for the APRM Setdown Scram. Calculation does not perform a STA test for this setpoint.

71. 7.5.3 p33 APRM Downscale (T) Revised calculation to Add discussion similar to the GEH justify 3.5 % EPU recommendations for changes to the Downscale Rod Block AV and NTSP for Neutron Glux High. setpoint. Added Although GE recommends a change discussion of GEH for NTSP from 3.5% to 4%, there is documents on this no reason not to leave the setpoint at subject. .

QF-0528 (FP-E-MOD-07)Rev. 0 Design Review Comment Form

" r e Committed to Nudear Exce/lence Fleet ModificationProcess 4

Sheet 9 of 9 the current value of 3.5% which is established in CA-05-153. Revise the calculation accordingly to use NTSP of 3.5%

7.5.3.4 p35 LER Avoidance (E) Added units Add units for OLER.

8.3 p38 EPU Operation (T) Inserted 3.5%

Revise NTSP for DOWNSCAL Rod Downscale Rod Block Block to 3.5% in table 9.1, 9.3 p38 APRM Downscale Rod Revised discussion Block (T) since 3.5 % is setpoint Delete discussion of change to NTSP.

9.6 (E) Corrected

QF-0528 (FP-E-MOD-07) Rev. 0 Design Review Comment Form Fleet Modlficatlon Process Sheet of DOCUMENT NUMBER1TITLE:

CA-08-050 Instrument Setpoint Calculation -

Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU REVISION: 0 DATE: 7114108-ITEM REVIEWER'S COMMENTS PREPARER'S REVIEWER'S

1. RESOLUTION DISPOSITION

1. QF-0549 p. 1, Title (E): Minor spacing Corrected OR typo

2. QF-0549 p. I,Major Revisions (T): Changed EC 10856 to Isn't the calculation being performed EC 12899. EC 12899 is d K under ECI2899? for CA-08-050 impact.

3. QF-0549 p.1, Major Revisions (P): Changed to Original 06 Description of Revision field should Issue be "Original Issue" for a Rev. 0.

4. QF-0549 pp. 3-4, Reference Item 8(96-224) Not Input; Documents (T): ltems 8 through 15 Item (31336)Changed to ltem 10. ltem 10 is now and 18 through 20 are NOT listed as inputs in the calculation so INPUT referenced in for Eqs; Item 1l(0017) NO^ input; ow field should NOT be checked. Item 12 (GE) Input now, Attachment 9; ltem 13 (TS) Not lnput ltem 14 (TRM) Not lnput ltem 15 (1.105)Not lnput ltem 18 (TSTF)Not lnput ltem 19 (RIS) Not lnput ltem 20(12899) Output

5. QF-0549 p.4, Reference Documents (E): Items 10 and 16 missing - Renumbered reference 06 consider renumbering list. list

6. QF-0549 p.4, Reference Documents (T): RG, TSTF, RIS are NOT Changed to use Passport 0 controlled documents so boxes are names Page 1 of 9

Error! Reference source not found.

Error! Reference source not found., Revision Error! Reference source not found, not checked. Therefore, these mnemonics should not be listed per

"*" note at bottom of page.

QF-0549 p.4, Reference Documents Changed to EC 10856 (T): Item 20 should be EC10856.

QF-0549 p.4, Reference Documents ISP-NIP-0588, OSP-NIP-(T): Should include as PROCs APRM 0590 are not approved Calibration, ISP-NIP-0588, and yet. Same for ISP-NIP-APRM Heat Balance, OSP-NIP-0590, 0589-01, ISP-NIP-0589-as OUTPUT documents (not sure 02 about OSP-NIP-0590 though).

QF-0549 p.5, Superseded Calculations (E): Statement that calc Incorporated statement is not affected until the installation of on PRNM installation via the PRMNS retrofit should be in both EC 10856.

calculation entries - applies to CA 224 too.

3494 p.1, Title (E): Typo Biased vs. Changed Eiased.

3494 p.1, 50.59 Screening or Evaluation No (E): This field requires Changed to only EC one valid 50.59 screening number. If 10856 for screening.

one is being prepared for EC12899, then that should be listed and EC-10856 should NOT be listed as a reference. (This idea of performing the calculation outside of EC-10856 may be problematic).

3494 p.1, Other Comments field (E): Added: "Section9, Future should be "None." Needs, for list of impacted documents IAW Calc procedure FP-E-CAL-0 I ,

Attach 4, step 3.b.7 CA-08-050 All Pages, Title (E): For consistency, should be "...Non-Flow Changed to Non-Flow Biased.. . Also revise in body of l' Biased for consistency.

calculation whenever it appears.

CA-08-050 p.1, TOC (E): QF-0549 has 6 pages and Body has 43 for total Corrected TOC of 52 (unless other comments result in different values).

Error! Reference source not found.

Error! Reference source not found., Revision Error! Reference source not found.

CA-08-050 p.1, TOC (E): Consider Made TOC a stand alone, including TOC in outline at page 1. page 1 of 1 sheet.

CA-08-050 p.1, PURPOSE (E): Not only does the retrofit change how the Added setpoint name point functions but it changes its title change along with the too - a small distinction perhaps. function change.

CA-08-050 p.1, PURPOSE Q: Site The plan is to address has determined the RBM Low Power, RBM LTSP, ITSP and Intermediate Power, and High Power HTSP in CA-08-051.

Setpoints are nominal values in Persmissive setpoints technical specifications and, LPSP, IPSP and HPSP therefore, will not have uncertainties will also be discussed in associated with them. This decision CA-08-051. Changed impacts the work in CA-08-051. CA- listing to LTSP, ITSP and 08-051 should state this position. Will HTSP.

this impact whether those 3 points are considered to be calculated in CA 051 and then listed in CA-08-050 Purpose section?

CA-08-050 p.1, Purpose (E): for consistency CA-08-052 should be ~iased vs Flow

...Flow Biased.. ." without the "-"). Referenced.

CA-08-050 p.3, Purpose (E): Added the future change procedure 1383 will be renumbered to 1383 with regard to ISP-NIP-1383 under EC10856. ISP-NIP-1383 CA-08-050 p.4, lnputs (T): ltem 4.1 Changed to lnputs 4.13 the reference numbers are confusing. and 4. IIbecause both Either "(Reference 5.)" should be are referenced in the

"(Reference 5.9)" or deleted entirely if body of the calculation Reference 5.14 is the only document being referenced.

CA-08-050 p.5, lnputs (E): ltem 4.8 Deleted original passport consider referring to the EC AEL database reference instead of the entire passport (original lnput 4.8 database. deleted).

CA-08-050 pp. 5 and 6, lnputs (E): FP-E-CAL-01,Att I Why are Items 5.1 through 5.6, and states Reference Section 5.12 through 5.13 listed as references is to include all lnput and when they are already inputs? Output documents.

Eror! Reference source not found.

Error! Reference source not found., Revision Error! Reference source not found.

CA-08-050 p.6, Inputs (E): lnput 5.15 Moved to Reference is a procedure and should NOT be section, Ref 5.15, since used as the bases for the +/-2% this is not an input for the setting tolerance. calculation.

CA-08-050 p.7, 7.1.2 Channel Changed scram to Scram Function (E): "scram should be for consistency.

Scram" - check throughout document.

CA-08-050 p.8, 7.2.11.2.2 Device I Changes made to and Device 2 (T): Make for LPRMs reference manufacturer.

ought to be GEHIReuter Stokes and for Power Electronics ought to be GE.

CA-08-050 p.8, 7.2.1 Device 1 (E): Reformattedtables in Drift/APEA/DPW M i n # of LPRM Section 7.2.1 and 7.2.2 need an extra space in the Reference(s) column.

APRM analog output is:

CA-08-050 p.8, 7.2.2 Device 2 (T): - 10 to + 10Vdcfor Should include as Output signals the various devices.

Flux Recorders, Flow Recorders, Reference is made in Flow Indicators, Computer Points via 7.2.2.1 the fiber optic data link, and the APRMIODA displays.

CA-08-050 p.8, 7.2.2 Device 2 (T): LTR is a lnput reference, Should the reference for the lnput and lnput 4.5.

Output signals be the LTR?

CA-08-050 p.9, 7.2.2 Device 2 (P): Changed to PMA.

lnput Power Process Measurement Deleted Power" Accuracy should be only Process Measurement Accuracy.

CA-08-050 p.9, 7.2.2 Device 2 (T): Ti-re basis documents for Shouldn't the root basis document for AFTIALT are GEH. Also, the As Left and As Found tolerance Sections 7.3.1.3 and values be something other than 7.3.1.4 evaluated GEH's calculation? AFTIALT for uncertainty and calibration tolerance.

CA-08-050 p.9, 7.2.2 Device 2 (T):

Power Supply Effect (LPRM Detector) Added statement to indicates to see APRM PEA but no 7.3.1.6 referencing the discussion of this being included is in Power Supply Effect for section 7.3.1.6. the LPRM detector.

Error! Reference source not found.

Error! Reference source not found., Revision Error! Reference source not found.

CA-08-050 p.9, 7.2.2 Device 2 (P):

Calibration Error of nla may be GE has not provided technically correct but we still have specs for calibration not received a minimum specification instruments.

from GE for the DVM, Oscilloscope, and Frequency Counter required to perform the calibration of the APRM chassis.

CA-08-050 p.9, 7.2.2 Device 2 (P):

Consider not splitting Plant Data table Reformatted across 2 pages.

CA-08-050 p.10, 7.3.1. I Loop Added Ref 7.2.1.1 to Accuracy of Device 1 (E): Refer to Section 7.3.1.1 to item 7.2.1 for the minimum number of reference the min # of LPRMs is 14. LPRMs CA-08-050 p. 10, 7.3.1.1 (P): Consider Included discussion with an explanation that the maximum numerical calculation that accuracy value occurs at the shows accuracy error minimum number of operable LPRMs. increases with less LPRMs averaged. Min #

LPRM used.

CA-08-050 p.12, 7.3.1.6 (E): There is Rephrased the APEA some confusion between APEAL as section to define the the Loop Primary Element Accuracy Random and Bias ACCURACY and the random component variables. Included of APEA - both use the same discussion of the designation; APE&. The Loop definition of APEA for this Primary Element Accuracy is made of calculation.

the random and bias components.

Consider changing the random component to something like APEAR.

Then carry this through to wherever it is reierenced.

CA-08-050 p.12, 7.3.1.7 (E): There is Rephrased the DPEA some confusion between D P W as section to define the the Loop Primary Element Random and Bias ACCU~~CYDR~FT and the random variables. Included component of DPEA - both use the discussion of the same designation; DPEAL. The Loop definition of DPEA for this Primary Element Accuracy is made of calculation the random and bias components.

Consider changing the random

Error! Reference source not found.

Error! Reference source not found., Revision Error! Reference source not found.

component to something like DPEAR.

Then carry this through to wherever it is referenced.

CA-08-050 p.12, 7.3.1.7 (E): Made change Underline section heading.

CA-08-050 p.13, 7.3.1.8 (E): What is A is acronym for APRM.

the first "Anfor in APMA? APRM? Added acronym sheet.

CA-08-050 p.13, 7.3.1.9 (E): PMAL Changed to APMAL for should be APMAL. consistency CA-08-050 p.13, 7.3.1.9 (P): Consider Added reference column adding a column of reference for each uncertainty sections (7.3.1.1, 7.3.1.2, 7.3.1.4, calculation 7.3.1.6, 7.3.1.7,7.3.1.8, 7.3.1.3, 7.3.1.5, 7.3.1.6, 7.3.1.7).

CA-08-050 p.14,7.3.2 (P): Table Added separate section should be given its own section for Table 7.3.2.1 just like the table in 7.3.1.9.

CA-08-050 p.13, 7.3.2 (E): PMAL Changed to APMAL should be APMAL.

CA-08-050 p.13,7.3.2 (P): Consider Added Columns for adding a column of reference reference sections sections (7.3.1.1, 7.3.1.2, 7.3.1.4, 7.3.1.6, 7.3.1.7, 7.3.1.8, 7.3.1.3, 7.3.1.5, 7.3.1.6, 7.3.1.7).

CA-08-050 p.15, 7.4.1 (E): Formatting Table re-sized error - text around the table.

CA-08-050 p.15, 7.4.1 .I(E): Equation Changed to APE&,

should refer to APEAL, not APEA. Also changed APMA and This is incorrect throughout the DPEA to APMAL and document in all AV evaluation DPEAL sections. See also 7.4.1.2, 7.4.2.1, 7.4.2.2, 7.4.3.1, 7.4.3.2, 7.5.1.1, 7.5.1.2, 7.5.2.1, 7.5.2.2, 7.5.3.1, and 7.5.3.2.

CA-08-050 p.15, 7.4.1 .I(P): Should Absolute value symbol the definition of Required Margin be not required since values an absolute value to avoid addressing are not negative negative results?

Error! Reference source not found.

Error! Reference source not found., Revision Error! Reference source not found.

CA-08-050 p.15, 7.4.1 . I (T): CLvalue Changed to 1.334 is incorrect but equation result is correct. Should be 1.334.

CA-08-050 p. 15, 7.4.1.I (T): Incorrect Corrected. lncluded A P W value listed but equation result APEAL value of 0.268 is correct. Should be 0.267. (rounded up).

CA-08-050 p.16,7.4.1.2 (E): NTSP Typo corrected vs. NPSP in equation.

CA-08-050 p.16, 7.4.1.2 (E): Equation Agree, Changed to should refer to DPEAL,not DL. This is DPEALthroughout incorrect throughout the document in all NTSP evaluation sections. See also 7.4.2.2, 7.4.3.2, 7.5.1.2, 7.5.2.2, and 7.5.3.2.

CA-08-050 p.17, 7.4.1.2 (E): Should Changed for consistency say: "(5.5% versus 3.08% power)" to with other sections be consistent with other similar sections.

CA-08-050 p.18, 7.4.1.4 (T): CLvalue Changed to 1.334 is incorrect but equation result is correct. Should be 1.334.

CA-08-050 p.18, 7.4.1.4 (T): ESM Agree, APEAb of 0.49 indicates STA calculation should has been added. A Ref include any bias terms present during to Input 4.12 was added normal operation. Therefore, APEAB as a basis for the bias of 0.49 should be added. SigmasTA= term.

1.86 and Z = 9.77 which are still acceptable.

CA-08-050 p. 19, 7.4.1.4 (T): Z Corrected equation should use 1.334 instead of 1.33.

CA-08-050 p. 19, 7.4.2 (E): Extra Corrected section number in title.

CA-08-050 p.19, 7.4.2 (E): Missing Corrected colon at end of 1" sentence.

CA-08-050 p.21, 7.4.2.4 (T): Added Z and NTSP2 Calculation of Z and NTSP2should be calculation for the

Erred Reference source not found.

Error! Reference source not found., Revision.Error! Reference source not found.

documented for the rod block at Setdown Rod Block in 15.0% and 13.0% even though the Sec 7.4.2.4 (LER result is identical. Avoidance Test)

CA-08-050 p.22, (T) Missing the STA Added STA for Setdown calculation for Setdown Scram and Scram. Per lnput 4.2, Rod Block. Comment 3, GEH states STA evaluations are not performed for Rod Blocks or permissives.

lnput 4.2, Comment 3, CA-08-050 p.25, (T) Missing the STA states STA not performed calculation for Downscale.

for Rod Block.

CA-08-050 p.26,7.5.1 .I (E):

Corrected Underline section heading.

CA-08-050 p.27,7.5.1.2 (E): Don't Corrected allow table to break across pages.

Added APEA bias term CA-08-050 p.29,7.5.1.4 (T): ESM and added lnput 4.13 as indicates STA calculation should reference for the basis to include any bias terms present during include the APEA bias normal operation. Therefore, APEAB term of 0.49 should be added. Sigmas, =

1.86 and Z = 9.77 which are still acceptable.

Corrected CA-08-050 p.31, 7.5.2.2 (E): Should say: "...the NTSP from the AL."

instead of "...the AV from the AL."

CA-08-050 p.32, 7.5.2.4 (T):

Added calculation of Z Calculation of Z and NTSP2should be and NTSP* for Section documented for the rod block at 7.5.2.4 15.0% and 13.0% even though the result is identical.

Added STA for Setdown CA-08-050 p.33, (T): Missing the STA Scram. Per lnput 4.3, calculation for Setdown Scram and Comment 3, GEH states Rod Block. STA evaluations are not performed for Rod Blocks or permissives.

CA-08-050 p.33, 7.5.3 (E): Reference Revised Inputs. 4.8 is to lnput 4.10 should be 4.9. correct reference

Error! Reference source not found.

En-or! Reference source not found., Revision Error! Reference source not found.

CA-08-050 p.36, (T): Missing the STA Per Input 4.3, Comment 3, calculation for Downscale. GEH states STA 0 1.(

evaluations are not performed for Rod Blocks or permissives.

CA-08-050 p.37, 8.1 (E): Given 47 Separated APE& and OK and 52, should table items APE& DpEAL into the Random and DPEALbe revised? and bias terms GARs: 01146760, CA-08-050 p.38, 9 (T): Need to 01146761 and PCR OY include CAP numbers for 9.1, 9.2, and 9.3. 01146750 initiated CA-08-050 p.39, 9.8 (T): Need a GAR I138038 listed for 06 tracking number for DBD but it isn't a DBD changes.

1 PCR -see PRNMS Project Engineer. I I CA-08-050 p.39, 9.9, 9.10, and 9.11 LAR 011128839 for TRM, 6Y (T): Need tracking number for GARS 01146762 and licensing documents but it isn't a PCR 01146763 for TS and

- see licensing engineer. Bases CA-08-050 p.40, 9.12 and 9.1 3 (E): All Section 9 impact 0'4 Include PCR numbers for B.05.01.02- documents have tracking

05. documents listed. B.05.

01.02-05 & PCR listed CA-08-050 p.41, Attachment 1 (T):

Shouldn't APRM High Flux Scram - 119.5 NTSP - 1.34%

CLTP and EPU Operation lower As =118.16 % RTP; OK FoundIAs Left value be 118.2? Conservative valve would be 118.2 CA-08-050 p.41, Attachment 1 (T):

Shouldn't APRM Setdown Scram for 18.0 NTSP + 1.34 =

CLTP and EPU Operation As 19.34; 19.3 is more OK FoundIAs Left value be 19.3? conservative than 19-33.

CA-08-050 p.42, Attachment 1 (T):

Shouldn't APRM Downscale Rod 3.5 % - 1.34%=2.16%;

2.2 % is conservative. 01(

Block - CLTP Operation lower As FoundIAs Left value be 2.2? Changed to 2.2 % R I P

QF-0528 (FP-E-MOD-07) Rev. 0 Sheet 1 of 2 I DOCUMENT NUMBER/ TITLE: CA-08-050, Instrument Setpoint Calculation -Average Power Range Monitor (APRM) Non Flow Biased PRNM Setpoints for CLTP and EPU.

(Peer Review Comments - Rhon Sanderson) REVISI0N:O DATE:07-28-08 ITEM REVIEWER'S COMMENTS PREPARER'S REVIEWER'S

1. RESOLUTION DISPOSITION

1. Table of Contents has discrepancies Revised TOC vs. actual document with regard to Repaginated page counts and start pages for calc sections.

2. Page 3 of 43, Section 2 Changed to Input 4.1 1 "NEDC-31336 (Reference 5.7)"

change to "....(Reference 5.14)"

which is the same as 5.14. Input4.11 is 04 referenced in Calc

3. Page 3 of 43, Section 2 Changed to Input 4.13, "GE-NE-901-021-0492 (Reference 5.6)" change to ".... (Reference 5.9)"

which is same as 5.9.

lnput 4.13 is 06 referenced in Calc.

4. Page 3 of 43, Section 2 Changed "ESM-03.02-APP-1" change to "ESM-03.02-APP-I"

5. Page 10 of 43, Section 7.3.1.I Value is 0.25202, so

"+I-0.252 % Power" is not rounded that should be 0.253.

0K conservatively Corrected

6. Page 12 of 43, Section 7.3.1.6 Value is 0.26726.., so "0.267 % Power" is not rounded that should be 0.268. O H conservatively Corrected

7. Page 13 of 43, Section 7.3.1.8

"+I-2.287 % Power" is not Value is 2.28737.., so that should be 2.289.

0 C(

conservatively rounded Corrected

8. Page 15 of 43, Section 7.4.1.I Corrected to 1.334 Value of "1.333" in RM equation should be "1.334". "(0.054)A2"in RM Page 1 of 2

Error! Reference source not found.

Error! Reference source not found., Revision Error! Reference source not found.

)

equation should be "(0.267)A2". It appears that the "2.69 %" results remains correct.

9. Page 16 of 43, Section 7.4.1.2 Agree DPEAL(0.054)

RM equation is missing "(0.054)A2" was as added to all term. It appears that the "3.08 %" applicable calculations result remains correct. This error through-out calculation shows up in similar RM equations through remainder of the calc body.

10. Page 18of43,Section7.4.1.4 Sigma was Sigma value of "1.37" is not rounded recalculated to I.87 by conservatively. Bias errors should be adding APEAb bias term in accordance OK taken out of the delta between the adjusted NTSP and the Operational with Input 4.1 3. ~~

Limit prior to evaluating this delta against sigma to determine Z value - i (for conservatism). This comment 9-T' applies to the spurious trip avoidance eval. in Section 7.51.4 as well. Note that "(0.054)A2"term is missing from the sigma equation in spurious trip avoidance evals.

11. Page 3 of 43, Section 2 Eliminate discussion of Excel Hand calculator checks have been performed 0I/(

spreadsheet with regard to calculation to verify rounding is of values. Hard equations and conservative. Deleted numbers in calc need to stand on discussion of Excel their own unless Excel computations spreadsheet.

are attached and independently evaluated, which would not be necessary in this case. Number of significant digits and conservative rounding needs to be cleaned up a little in the calc, although no changes in end results are expected.

L I

MONTICELLO NUCLEAR GENERATING PLANT 3494 TITLE: CALCULATION COVER SHEET Revision 17 Page 1 of 1 ri Title Instrument Setpoint Calculation - Average C A - 0 8 - 050 Rev. 0 Power Range Monitor (APRM) Non-Flow Eiased PRNM Setpoints for CLTP and EPU 50.59 Screening not required. Calculation is 10 CFR50.59 Screening or Evaluation No: submitted as part of LAR EC 12899 Plant Impact Associated Reference(s): from CA-08-050 I Does this calculation: I YES I NO I Calc No(s), Rev(s), Add(s) I

\ Supercede another calculation? I (XI ( ( Ref QF-0549 (Calculation Signature Page, attached). r (

- 1 1 Augment (credited by) another calculation?

Affect the Fire Protection If Yes, attach Form 3765 Program per Form 3765?

I If Yes, attach Form 3544 Affect IST Program Valve or If Yes, inform IST Coordinator and provide copy of calculation What systems are affected?

DBD Section (if any): DBD-B.05.01, Neutron Monitoring System Topic Code (See Form 3805): NIP Power Range Monitors Structure Code (See Form 3805): -

RATE Reratelpower Uprate Other Comments: Section 9, Future Needs - List of impacted documents Prepared by: Date:

8-//-0 Micah I

, - w e - -

r

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 Revision 0 Acronvms and Abbreviations Page 1 of 1 AL Loop InstrumentAccuracy AFT As-Found Tolerance AFTL Loop As-Found Tolerance AGAF APRM Gain Adjustment Factor AL Analytical Limit ALT As-Left Tolerance ALTL Loop As-Left Tolerance APEAL Loop APRM Primary Element Accuracy APEAR Loop APRM Primary Element Accuracy Random APE& APRM Primary Element Accuracy bias APMk Loop APRM Process Measurement Accuracy APRM Average Power Range Monitor AV Allowable Valve CL Loop Calibration Accuracy Error CLTP Current Limiting Thermal Power DL Loop Instrument Drift DPEAL Loop Drift Primary Element Accuracy DPEAR Loop Drift Primary Element Accuracy Random DTE Drift Temperature Effect DPEAb Drift Primary Element Accuracy bias EPU Expanded Power Uprate FS Full Span GEH GE-Hitachi Nuclear Energy IRM Intermediate Range Monitor LER Licensee Event Report LPRM Local Power Range Monitor NMS Neutron Monitoring System NTSP Nominal Trip Setpoint NUMAC Nuclear Measurement Analysis and Control OL Operational Limit PIC Plant Process Computer PEA Primary Element Accuracy PMA Process Measurement Accuracy PRNM Power Range Neutron Monitoring PRNMS Power Range Neutron Monitoring System RTP Rated Thermal Power SRM Startup Range Monitor STP Simulated Thermal Power STA Spurious Trip Avoidance

MONTICELLO NUCLEAR GENERATlNG PLANT CA-08-050 Revision 0 Table of Contents Page 1 of 1 Item Description Paqes QF-0549 Calculation Signature Sheet QF-0527 Design Review Checklist QF-0528 Design Review Comment Form 3494 Calculation Cover Sheet Acronyms List of Acronyms and Abbreviations TOC Table of Contents Calculation Body Setpoint Diagrams 2 Input 4.2, GEH 0000-0077-9068 MNGP-PRNMS-APRM Calc 2008, Revision 2, March 2008 23 Input 4.3, GEH 0000-0081-6958 MNGP-PRNMS-APRM Calc 2008, Revision 0, March 2008 24 Attachment 4 Input 4.12, Mathematics of Physics and Modern Engineering, 1966, I.S. Sokolnikoff and R.M. Redheffer 3 Total 142 Section Pane #

PURPOSE METHODOLOGY ACCEPTANCE CRITERIA INPUTS REFERENCES ASSUMPTIONS ANALYSIS APRM NON-FLOW BIASED PRNM LICENSING SETPOINTS INSTRUMENT DEFINITION AND DEVICE UNCERTAINTY TERMS LOOP INSTRUMENT UNCERTAINTY EVALUATION PRNM - CLTP OPERATION SETPOINT EVALUATION PRNM - EPU OPERATION SETPOINT EVALUATION CONCLUSIONS FUTURE NEEDS

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 1 of 58 I. PURPOSE This calculation provides design basis setpoint analysis for the Allowable Values (AV) and Nominal Trip Setpoints (NTSP) for the Power Range Neutron Monitoring (PRNM)

APRM setpoints associated with the installation of EC 10856. EC-12899 documents the plant impact and configuration changes from the calculation. The following setpoints are evaluated for PRNM CLTP and EPU operation in accordance with setpoint control program and NRC commitment M87051A:

APRM Neutron Flux - High Scram APRM Neutron Flux - High (Setdown) Scram APRM Neutron Flux - High (Setdown) Rod Block APRM Downscale Rod Block The NUMAC PRNM retrofit is a digital neutron monitoring system that replaces the analog NIP System - Power Range Monitoring System. This calculation evaluates the above setpoints and determines the available margin based on PRNM retrofit uncertainty parameters for CLTP and EPU operation. The PRNM retrofit affects the above setpoints as follows:

1. PRNM adds two new neutron monitoring setpoints for CLTP and EPU operation.

These are identified above as the APRM Neutron Flux - High (Setdown) Scram and APRM Neutron Flux - High (Setdown) Control Rod Block. The function of the setpoints is described in Section 7.1.2.

2. The PRNM retrofit changes how the current APRM Flow Referenced Neutron Flux -

High High setpoint functions and changes the setpoint name. The existing APRM Flow Referenced Neutron Flux - High High setpoint is changed to a non-flow biased setpoint identified as APRM Neutron Flux - High Scram, which is independent of core recirculation flow. The function is described in Section 7.1.2.

3. The APRM Downscale Rod Block is an existing CLTP setpoint. The setpoint does not change for PRNM CLTP and EPU operation. GEH setpoint documentation, Input 4.3, recommended a NTSP setpoint of 4.0 % RTP for EPU operation. This calculation provides the design bases to use the existing CLTP NTSP setpoint of 3.5

% RPT.

This PRNM based neutron monitoring system (NMS) calculation supersedes calculations CA-05-153 (Reference 5.8) and CA-96-224 (Reference 5.7). Section 9 (Future Needs) describes the affect on these calculations due to PRNM implementation.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 2 of 58 In addition to the PRNM non-flow bias neutron monitoring setpoints identified above, the PRNM retrofit also creates new or changes other neutron monitoring setpoints. For completeness of the PRNM affected NMS setpoints, the following neutron monitoring setpoints will be evaluated in other calculations:

A. Calculation CA-08-051, lnstrument Setpoint Calculation - Rod Block Monitor (RBM) PRNM Setpoints for CLTP and EPU Operation, which includes the following sub-calculations:

RBM Low Trip Setpoint (LTSP)

RBM Intermediate Trip Setpoint (ITSP)

RBM High Trip Setpoint (HTSP)

B. Calculation CA-08-052, lnstrument Setpoint Calculation - Average Power Range Monitor (APRM) Flow Biased PRNM Setpoints for CLTP and EPU, which includes the following sub-calculations for Two Loop Operation (TLO) and Single Loop Operation (SLO):

APRM Simulated Thermal Power - High Scram (TLO)

APRM Simulated Thermal Power - High Scram (SLO)

APRM Simulated Thermal Power - High Rod Block (TLO)

APRM Simulated Thermal Power - High Rod Block (SLO)

C. Calculation CA-08-053, Average Power Range Monitor (APRM) Recirc Flow Instrumentation Calibration for PRNM CLTP and EPU, which includes the following subsections:

Recirc Flow transmitter Gain Scaling NUMAC Recirc Flow Grain Factor Equation for Procedure 1383 (Core Flow Measurement System Calibration). Note: Procedure 1383 is to be renumbered to ISP-NIP-1383 under EC 10856.

2. METHODOLOGY This calculation is performed in accordance with ESM-03.02-APP-I (Input 4.1). ESM 02-APP-I setpoint methodology is based on the following documents: General Electric lnstrument Setpoint Methodology NEDC-31336 (Input 4.11) and Setpoint Calculation Guidelines for the Monticello Nuclear Generating Plant, GE-NE-901-021-0492 (Input 4.13). The General Electric Setpoint Methodology is a statistically based methodology. It recognizes that most of the uncertainties that affect instrument performance are subject to random behavior, and utilizes statistical (probability) estimates of the various uncertainties to achieve conservative, but reasonable, predictions of instrument channel uncertainties. The objective of the statistical approach to setpoint calculations is to achieve a workable compromise between the need to ensure instrument trips when

MONTICELLO NUCLEAR GENERATlNG PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Set~ointsfor CLTP and EPU Paae 3 of 58 appropriate, and the need to avoid spurious trips that may unnecessarily challenge safety systems or disrupt plant operation.

Drift Analysis: This calculation uses GE specified drift parameters for the applicable PRNM equipment and for the existing LPRM detectors.

The uncertainties associated with the overall PRNMS including the LPRMs, APRMs and associated hardware are appropriately considered and consistent with NRC approved GE methodology in establishing the APRM setpoints. The calculation uses GEH specified ALT and AFT tolerances to calculate loop uncertainty. These parameters are specified in Inputs 4.2 and 4.3 (GE PRNM documentation) and were converted to a 2 0 value in accordance with Engineering Standards Manual ESM-03.02-APP-I, Rev 4 (Input 4.1). In addition to ALT and AFT tolerances for loop uncertainty, Sections 7.3.1.3 and 7.3.1.4 evaluated AFTIALT for digital PRNMS surveillance calibration. The setpoints are numerical values stored in the digital hardware and not subject to drift. The ALT and AFT values for the setpoint are the same as the trip setpoint. Therefore, there is no tolerance band for the surveillance calibration test. Attachment 1, Setpoint Diagrams, states AFTIALT tolerance will not be applied to surveillance calibration of the setpoints because PRNMS setpoints are digital and stored in PRNMS database.

3. ACCEPTANCE CRITERIA The Scram Setpoint and Allowable Values should be such that the Analytical Limit (AL) will not be exceeded when all applicable instrumentationuncertainties are considered.

For the Allowable Value (AV), the minimum required margin is calculated and compared to the available margin, which is AL minus AV. For the Nominal Trip Setpoint (NTSP) evaluation, the minimum required margin is calculated and compared to the available margin, which is AL minus NTSP.

For parameters that do not have AL, such as Setdown Scram and Rod block and Downscale Rod Block, the difference between the minimum required margins (AL to AV and AL to NTSP) constitute the minimum required margin between AV and NTSP. This minimum required margin is compared to the available margin, which is AV minus NTSP.

For the Licensee Event Report (LER) Avoidance Test setpoint evaluation, sufficient margin is verified between the NTSP and AV setpoints to prevent an LER condition. A Spurious Trip Avoidance (STA) setpoint evaluation is performed where applicable to assure that there is a reasonable probability that spurious trips will not occur using the selected setpoints.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 4 of 58 INPUTS Engineering Standards Manual ESM-03.02-APP-I, Appendix I (GE Methodology Instrumentation & Controls), Revision 4. The ESM provides plant specific guidance on the implementation of the General Electric guidelines (Input 4.13) and methodology (Input 4.11).

GEH: 0000-0077-9068 MNGP-PRNMS-APRM Calc-2008, Revision 2, DRF: 0000-0076-1670, March 2008, Average Power Range Monitor Selected PRNM Licensing Setpoints - CLTP Operation (NUMAC). This is a GEH basis document for the digital PRNM equipment and includes setpoint functions and instrument uncertainties for PRNM CLTP operation. This document is Attachment 2.

GEH: 0000-0081-6958 MNGP-PRNMS-APRM Calc 2008, Revision 0, DRF: 0000-0081-4903, March 2008, Average Power Range Monitor Selected PRNM Licensing Setpoints - EPU Operation (NUMAC). This is a GEH basis document for the digital PRNM equipment and includes setpoint functions and instrument uncertainties for PRNM EPU operation. This document is Attachment 3.

GEH-NE-0000-0076-2388, DRF 0000-0076-2387, Revision I,MNGP PRNM Licensing Setpoints - CLTP Operation, December 2007. This document discusses the setpoint changes needed to license PRNM for CLTP operation.

NEDC-32410P-A,Volume 1 - Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function, Licensing Topical Report, October 1995. The LTR was used to provide descriptions of the PRNM equipment. Input and output signal data was obtained from this document.

NEDC-32410P-A, Volume 2 - Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function, Licensing Topical Report, October 1995. The LTR was used to provide description of the PRNM equipment.

NEDC-32410P-A, Supplement I- Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function, Licensing Topical Report, Supplement 1, November 1997. The LTR was used to provide description of the PRNM equipment.

Task Report T0506, Revision I , Project Task Report, NMC Monticello Nuclear Generating Plant Extended Power Uprate, Technical Specifications Setpoints, March 2008. This document provides PRNM CLTP and EPU setpoints addressed in this calculation.

1 MONTICELLO NUCLEAR GENERATING PLANT I CA-08-050 I TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow 1 ~iased PRNM setpoints for CLTP and EPU 1 page 5 of 58 1 4.9 Design Input Request (DIR) T0500, Rev 2, DRF 000-0040-9168, Neutron Monitoring System. This DIR provides design information on the LPRMs used for input to the PRNM equipment.

4.10 Specification 257HA594, Rev I , Neutron Monitoring System, 12/3/85.

Specification provides information on LPRM detectors and the existing analog neutron monitoring system. This document provides design specifications for the LPRMs.

4.1 1 NEDC-31336P-A, Class Ill, General Electric lnstrument Setpoint Methodology, September 1996. Setpoint equations are referenced from this document.

4.12 I.S. Sokolnikoff and R.M. Redheffer, Mathematics of Physics and Modern .

Engineering, 1966. The equation for statistical averaging of inputs is referenced from this book. Pages are contained in Attachment 4.

4.13 GE-NE-901-021-0492, DRF A00-01932-1, Setpoint Calculation Guidelines for the Monticello Nuclear Generating Plant, October 1992. This calculation references this document for the inclusion of bias for the Spurious Trip Avoidance (STA) calculation.

5. REFERENCES 5.1 GEH: 0000-0077-9068 MNGP-PRNMS-APRM Calc-2008, Revision 2, DRF: 0000-0076-1670, March 2008, Average Power Range Monitor Selected PRNM Licensing Setpoints - CLTP Operation (NUMAC) 5.2 GEH: 0000-0081-6958 MNGP-PRNMS-APRM Calc 2008, Revision 0, DRF: 0000-0081-4903, March 2008, Average Power Range Monitor Selected PRNM Licensing Setpoints - EPU Operation (NUMAC) 5.3 NEDC-32410-A, Volume I- Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function, Licensing Topical Report, October 1995.

5.4 NEDC-32410-A, Volume II - Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function, Licensing Topical Report, October 1995.

5.5 NEDC-32410-A, Supplement 1 - Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option 111 Stability Trip Function, Licensing Topical Report, Supplement 1, November 1997

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow I I ~ i & e dPRNM ~ e i ~ o i nfor t s CLTP ;nd EPU I ~ a a 6e of 58 1 5.6 GEH-NE-0000-0076-2388, Revision 1, MNGP PRNM Licensing Setpoints - CLTP Operation, December 2007 5.7 CA-96-224, Rev 1, lnstrument Setpoint Calculation - Average Power Range Monitor (APRM) Flow-Biased Upscale Scram and Rod Block.

5.8 CA-96-153, Revision 0, Instrumentation Setpoint Calculation - Average Power Range Monitor (APRM) Downscale CR Block..

5.9 GE-NE-901-021-0492, DRF AOO-01932-1, Setpoint Calculation Guidelines for the Monticello Nuclear Generating Plant, October 1992.

5.10 Monticello Nuclear Generating Plant Technical Specifications, as revised through Amendment 155. GAR 01146762 initiated to update Technical Specification in accordance with EC 10856 and calculation CA-08-050.

5.11 Monticello Nuclear Generating Plant Technical Requirements Manual (TRM), as revised through Revision 2. LAR 01 128839 updates TRM in accordance with EC 10856 and calculation CA-08-050.

5.12 Regulation Guide 1.105, R3 - lnstrument Setpoints for Safety-Related Instrumentation.

5.13 Task Report T0506, Revision 1, Project Task Report, NMC Monticello Nuclear Generating Plant Extended Power Uprate, Technical Specifications Setpoints, March 2008.

5.14 NEDC-31336P-A, Class Ill, General Electric lnstrument Setpoint Methodology, September 1996.

5.15 Procedure 0017, Revision 25, "APRM Heat Balance Calibration." This procedure is used to calibrate the APRM gains such that the absolute difference between the Average Power Range Monitor (APRM) channels and the calculated power is I2

% RTP while operating at 2 25 % RTP.

5.16 I.S. Sokolnikoff and R.M. Redheffer, Mathematics of Physics and Modern Engineering, 1966 5.17 RIS 2006-17, NRC Staff Position on the Requirements of 10 CFR 50.36, "Technical Specifications," Regarding Limiting Safety System Settings During Periodic Testing and Calibration of lnstrument Channels, August 24, 2006

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 7 of 58 TSTF-493, Rev 3, Clarify Application of Setpoint Methodology for LSSS Functions.

Date of issue 18 Jan 08. Rev 3 is not approved. Included as a reference document.

EC 10856, Rev 0, EPU - Mod 4 - Neutron Monitoring System (PRNM)

EC 12899, Rev 0, PRNMS Setpoint Calculations 050 (Non-Flow Biased Setpoints)

Specification 257HA594, Rev 1, Neutron Monitoring System, 12/3/85.

Specification provides information on LPRM detectors and the existing analog neutron monitoring system Engineering Standards Manual ESM-03.02-APP-I, Appendix I (GE Methodology Instrumentation & Controls), Revision 4 Procedure C.6-005-A-22, Rev 3, APRM Hi Hi INOP CH 1, 2, 3, is a flow-bias APRM Neutron Flux Hi Hi setpoint. PRNM retrofit converts this setpoint to a non-flow bias APRM Neutron Flux High setpoint. The PRNM APRM Neutron Flux High setpoint is part of this calculation. (2.6-005-A-22 will be revised under EC-10856.

PCR 01 129100.

Procedure C.6-005-A-30, Rev 3, APRM Hi Hi INOP CH 4,5,6, is a flow-bias APRM Neutron Flux Hi Hi setpoint. PRNM retrofit converts this setpoint to a non-flow bias APRM Neutron Flux High setpoint. The PRNM APRM Neutron Flux High setpoint is part of this calculation. C.6-005-A-30 will be revised under EC-10856.

PCR 01133816.

Procedure C.6-005-A-06, Rev 3, APRM Downscale, states a NTSP setpoint of 3.5

% RTP. This is correct for the present neutron monitoring system. Even though the PRNM CLTP and EPU operation NTSP setpoints are 3.5 % RTP, the procedure does not address that the PRNM retrofit NTSP setpoints remain the same for CLTP and EPU operation. PCR 01146778 initiated to revise procedure for EC 10856 and calculation CA-08-050.

Procedure C.6-005-A-03, Rev I, Annunciator procedure for window 5-A-3.

PRNMS adds a new rod withdraw block setpoint: APRM Neutron Flux - High (Setdown) Rod Block. PCR 01 146750 initiated to revise procedure for EC 10856 and calculation CA-08-050.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Set~ointsfor CLTP and EPU Paae 8 of 58 5.27 Procedure B.05.06-02, Rev 18, Operations Manual Section - Plant Protection System, specifies APRM Hi Hi and APRM Downscale and other setpoints. This calculation evaluates the APRM Downscale Rod Block setpoints and documents the PRNM EPU change in this setpoint. The APRM Hi Hi setpoint is flow biased and is PRNM changes this setpoint to non-flow bias APRM Neutron Flux High.

8.05.06-02 will be revised under EC-10856 by PCR 01133455.

5.28 DBD B5.1, Rev C, Design Bases Document for Neutron Monitoring System, discusses NMS setpoints, margin, uncertainty parameters such as drift, etc. This calculation validated certain NMS setpoints using the PRNM parameter uncertainties specified in GE documentation. Changes will be made under GAR 1138038.

5.29 Procedure 8211, Rev 2, APRM Calibration Readjustment for Single Loop, discusses APRM setpoint voltage adjustments including Downscale Rod Block, Hi-Hi Scram, etc. Changes have been made by PRNM and this calculation evaluates the non-flow biased PRNM setpoints. Procedure 821 1 will be deleted under EC-10856 by PCR 01133437 and replaced with directions in B.05.01.02-05 by PCR 01 133449.

5.30 Procedure 8212, Rev 2, APRM Calibration Readjustment for Two Loop, discusses APRM setpoint voltage adjustments including Downscale Rod Block, Hi-Hi Scram, etc. Changes have been made by PRNM and this calculation evaluates the non-flow biased PRNM setpoints. Procedure 8212 will be deleted under EC-10856 by PCR 01133445 and replaced with directions in B.05.01.02-05 by PCR 01 133449.

5.31 Procedure 0012, Rev 41 APRMIFlow Reference Scram Functional Check, performs the calibration of the APRM including the Neutron Flux High Scram, Setdown Scram, Setdown Rod Block, and Downscale Rod Block setpoints.

Setpoints are revised as a result of this calculation. 0012 will be deleted under EC-10856, PCR 01133332. Procedures ISP-NIP-0588, ISP-NIP-0588-01, ISP-NIP-0589-02 will be developed to replace Procedure 0012 by PCRs 01 129124, 01129125, and 01129126.

5.32 MNGP Technical Specifications Bases, Rev 8, Bases will be revised to discuss the PRNM APRM Neutron Flux High setpoint, which is non-flow bias, in place of the existing Flow Referenced Neutron Flux-High High setpoint. GAR 01 146762 initiated to update Technical Specification Bases in accordance with EC 10856 and calculation CA-08-050.

5.33 Specification 24A5221, Specification for PRNM MUMAC Power Range Neutron Monitoring System.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Set~ointsfor CLTP and EPU Paae 9 of 58 5.34 B.05.01.02-02, Rev 6, Operations Manual Section - Power Range Neutron Monitoring, specifies NMS trip setpoints, which are being changed due to PRNMS.

B.05.01.02-02 will be revised under EC-10856 by PCR 01137808.

5.35 B.05.01.02-05, Rev 16, Operations Manual Section - Power Range Neutron Monitoring, System Operation. B.05.01.02-05, Rev 16 refers to the six APRM channels, which applies to the existing NMS. PRNMS has four APRM channels as stated is Section 7.2.2.1 of this calculation. PCR 01146778 issued to revise B.05.01.02-05, Rev 16, upon implementation of EC 10856.

5.36 Design Input Request (DIR) T0500, Neutron Monitoring System, DRF 000-0040-9168. This DIR provides design information on the LPRMs used for input to the PRNM equipment.

5.37 Engineering Standards Manual ESM-03.02-APP-1, Appendix I (GE Methodology Instrumentation & Controls), Revision 4.

6. ASSUMPTIONS None.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Seboints for CLTP and EPU Paae I 0 of 58

7. ANALYSIS 7.1 APRM Non-Flow Biased PRNM Licensing Setpoints 7.1.1 Channel Diagram for APRM Neutron Flux Set~oints Output Channels:

(Device 1) Power Electronics 14 minimum of 24 APRM Neutron Flux - High Scram (Device 2) APRM Neutron Flux - High (Setdown) Scram APRM Neutron Flux - (Setdown) Rod Block APRM Downscale Rod Block 7.1.2 Channel Function:

The APRM system calculates an average of the incore Local Power Range Monitor (LPRM) chamber signals. The LPRMs are averaged such that the APRM signal is proportional to the core average neutron flux and can be calibrated as a means of measuring core thermal power. The number of APRM channels is reduced to four from six and the LPRM's are re-assigned to increase the number of LPRM's in the APRM average. The logic of the trip output signals from the APRM channels is modified from the original design to implement a Two-out-of-Four (214)trip logic that eliminates half Scrams resulting from a single PRNM channel failure. A neutron flux trip in any two APRM channels will cause a Scram.

The APRM Neutron Flux High Scram is capable of generating a trip signal to prevent fuel damage or excessive RCS pressure in high power range. For rapid neutron flux increase events, the thermal power lags the neutron flux and APRM Neutron Flux High Scram will provide a Scram signal before the APRM Flow Biased Simulated Thermal Power (STP) Scram. The APRM Neutron Flux High Scram is based on unfiltered neutron flux signal.

The APRM Setdown Scram is capable of generating a trip signal that prevents fuel damage resulting from abnormal operating transients in the low power range. The APRM Setdown Rod Block is a precursor to the APRM Setdown Scram. The setdown Scram is a redundant Scram, which overlaps the IRM region, for reactivity transients in the startup mode. This provides defense-in-depth for reactivity transients in the startup mode.

The APRM Downscale Rod Block initiation ensures that there is sufficient overlap of the operating regions of the APRMs and IRMs with the IRM detectors fully inserted. APRM Downscale Rod Block function provides indication of instrument failure or insensitivity.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: I Instrument Set~ointCalculation - Revision 0 Average Power Range 'Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 11 of 58 7.2 lnstrument Definition and Device Uncertainty Terms The APRM Non-Flow Biased Loop is composed of LPRM Detectors (Device 1) and NUMAC PRNMS Power Electronics (Device 2).

7.2.1 Device I: LPRM Detector Data 7.2.1.I lnstrument Definition:

Device 1 - LPRM Detectors Component IDS:

Total: 24 LPRM Detector strings containing 4 detectors each = 96 LPRM Detectors INDREC LPRM-04-29 INDREC LPRM-12-13 INDREC LPRM-12-21 INDREC LPRM-12-29 INDREC LPRM-12-37 INDREC LPRM-20-13 INDREC LPRM-20-21 INDREC LPRM-20-29 INDREC LPRM-20-37 INDREC LPRM-20-45 INDREC LPRM-28-05 INDREC LPRM-28-13 INDREC LPRM-28-21 INDREC LPRM-28-29 INDREC LPRM-28-37 INDREC LPRM-28-45 INDREC LPRM-36-13 INDREC LPRM-36-21 INDREC LPRM-36-29 INDREC LPRM-36-37 INDREC LPRM-36-45 INDREC LPRM-44-21 INDREC LPRM-44-29 INDREC LPRM-44-37 Sigma Reference(~)

Location Drywell nla Input 4.9, Item 2 Make GElReuter Stokes nla Input 4.9, Item 22 Model GE NA300 nla Input 4.9, Item 22 Local Power Range Monitor Input 4.2, Section 1-CLTP Process Element (LPRM) Neutron detector nla Input 4.3, Section 1-EPU Upper Range Limit (UR) n/a nla nla Calibrated Span (SP) nla nla nla Design maximum neutron flux of Input (neutron flux) 2.3E14 nv nla Input 4.9, Item 23 Output (LPRM electronics) 0.0 to 3 ma nla Input 4.5, Section 5.3.17.6 Minimum # of LPRMs per Input 4.2,Section 2.3-CLTP APRM 14 of 24 nla Input 4.3,Section 2.3-EPU

MONTICELLO NUCLEAR GENERATlNG PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 12 of 58 7.2.1.2 Process and Physical Interfaces:

Device 1 - LPRM Detectors Process and Physical Interfaces Sigma Reference(s)

Calibration Temperature n/a for LPRM detector due to in-core Range location nla nla Calibration/Surveillance Interval 7 Days nla Technical Specifications Normal Plant Conditions: LPRM detectors are exposed to reactor operating conditions:

Temperature Design Temperature 546 deg F Radiation Gamma 2.4 FUhr; Neutrons 10 FUhr Pressure Design Pressure 1250 psig Humidity at Assembly Condensation Dripping water is Connector present nla lnput4.10 Trip Environment Conditions Not applicable for setpoint calculation n/a nla Long Term Post Accident Conditions Not applicable for setpoint calculation nla nla Seismic Conditions Not applicable for setpoint calculation nla nla Process Conditions:

During Calibration Worst Case During Function Not applicable for setpoint calculation nla nla

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 13 of 58 7.2.1.3 Individual Device Accuracy I LPRM Detectors I Value Symbol Term (% RTP) Sigma Reference 0 %, LPRM detector Instrument Accuracy accuracy is included in Input 4.2, Section 2.3-CLTP A1 - LPRM Detector APEA nla Input 4.3, Section 2.3-EPU 0 %, LPRM detector drift in lnout 4.2. Section 2.3-CLTP DI I LPRM Drift I included in DPEA I nla input 4.3; Section 2.3-EPU I ( =iAP&+AP&

I APEA- Accuracy- per . I 5 1.00 % RTP + bias 0.49 1 lnput 4.2, Section 1-CLTP 1 2 I =

• DPEA, + DPEA, 1 I DPEA - Drift per I

• 0.2 % RTPI 7days + bias I L

lnput 4.2, Section 1-CLTP DPEA 1 LPRM detector 1 0.33 % RTP 1 2 In ut 4.3, Section 1-EPU 1 0 %, Design temperature is I normal in-&re temperature of 546 deg F. LPRM electronics temperature lnput 4.9, ltem 2 Accuracy effect is included in lnput 4.2, Section 2.3-CLTP ATE Temperature Effect accuracy nla nla - Overpressure effect is not applicable for LPRM Detector accuracy.

Detector is designed for

---

I OPE I Overpressure Effect 1 1250 psig. I nla In ut 4.9, ltem 2 1 nla - Static oressure effect I is not applicable for LPRM Detector accuracy due to 1

Static Pressure Effect in-core location. 1 nla n/a 1 nla - Seismic effect is not I applicable because APRM Scram and rod block are only required during normal Seismic Effect ) operating conditions. nla nla I nla - Radiation effect is not I applicable because LPRM detector is designed for a lifetime nv of 1.2E14 nv @-

Radiation Effect I 1E9 Rem/hr. I nla lnput 4.9, ltem 2 I Ma - Humidity is not I I a ~ ~ l i c a bbecause le LPRM I HE ( Humidity ( detector is located in-core. ( nla ( n/a I lnput 4.2, Section 1-CLTP lnput 4.3, Section 1-EPU Negligible for LPRM Comment 16 in each of the PSE Power Supply Effect detector nla above Inputs.

nla - RFI/EMI effect is not applicable because LPRM REE RFIIEMI Effect detector is located in-core. nla nla

MONTlCELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 14 of 58 7.2.2 Device 2: NUMAC PRNMS Power Electronic Data (LRPM, APRM, Trip Circuits) 7.2.2.1 lnstrument Definition Device 2 - NUMAC PRNMS Power Electronic Data Component ID'S:

INDREC APRM 1 INDREC APRM 2 INDREC APRM 3 INDREC APRM 4 Device 2 - NUMAC PRNMS Power Electronic Data Sigma Reference(s)

Power Electronics: Admin Building, El Input 4.2, Section 2.3-CLTP Location 951' nla Input 4.3, Section 2.3-EPU lnput 4.2, Section 2.3-CLTP Make GE nla Input 4.3, Section 2.3 -EPU lnput 4.2, Section 2.3-CLTP Model NUMAC nla Input 4.3, Section 2.3-EPU lnput 4.2, Section 2.3-CLTP Calibration Scale Full Scale = 125 % RTP nla Input 4.3, Section 2.3-EPU Input 4.2. Section 2.3-CLTP Upper Range Limit nla nla input 4.3; section 2.3-EPU Input signal 0 - 3 ma from each LPRM nla Input 4.5, Section 5.3.1 7.6 Analog Output signal:

Outputs to: Flux Recorders, Flow Recorders, Flow Indicatiors, Computer - 10 to + 10 VDC (maximum) nla Input 4.5, Section 5.3.17.7

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 15 of 58 7.2.2.2 Process and Physical Interfaces Device 2 NUMAC PRNMS Power Electronic Data sigma Reference(s)

Calibration Temperature Input 4.2, Section 2.3-CLTP Range 72 to 78 deg F n/a Input 4.3, Section 2.3-EPU CalibrationlSurveillance Input 4.2, Section 2.3-CLTP Interval 700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> 2 Input 4.3, Section 2.3-EPU .

Normal Plant Conditions:

Temperature 72 to 78 deg F n/a Radiation Negligible 2 Pressure nla 2 Input 4.2, Section 2.3-CLTP Humidity Included in Accuracy 2 Input 4.3, Section 2.3-EPU Trip Environment 72 to 78 deg F Input 4.2, Section 2.3-CLTP Conditions

- Same as Normal Plant Conditions nla Input 4.3, Section 2.3-EPU nla - Scram and Rod Block functions Long Term Post Accident are only required during normal Conditions conditions. nla nla lnput 4.2, Section 2.3-CLTP, Comment 4 lnput 4.3, Section 2.3-EPU, Seismic Conditions Included in Accuracy 2 Comment4 Process Conditions:

During Calibration Worst Case During Function nla for APRM calibration n/a n/a

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 16 of 58 7.2.2.3 Individual Device Accuracv Device 2 - N JMAC PRNMS Power Electronic Data r I Value

(% RTP) Sigma Reference Instrument Accuracy of LPRM flux channel electronics i 0.943 % RTP I1 2 I1 hput 4.2.

Input 4.3, Section 2.3-CLTP Section 2.3-EPU Power Electronics I lnput 4.2, Section 2.3-CLTP D2 Drift

• 0.50 RTP FSnOO Hours 1 2 1 lnput 4.3, Section 2.3-EPU APMA 1 1 a. tracking I Process Measure a.*1.11 %RTP I I Input 4.2, Section 1-CLTP

b. noise Accurac *

b. 2.00 % RTP 1. 2 ( Input 4.3, Section 1-EPU I I l n ~ u4.2.

t Section 2.3-CLTP CL Loop Calibration

~ c c u r a cError

~

ALT specified by the AGAF process =

• 2.0 % RTP I 3 I arid ~omrnent11 Input 4.3, Section 2.3-EPU and Comment 11

• 2.0% RTP based on APRM Gain Adjustment Input 4.2, Section 2.3-CLTP ALTL As-Left Tolerance Factor (AGAF) 3 Input 4.3, Section 2.3-EPU

= ALT, Input 4.2, Section 2.3-CLTP AFTL As-Found Tolerance *

= 2.0 % RTP 3 Input 4.3, Section 2.3-EPU Accuracy Input 4.2, Section 2.3-CLTP ATE Temperature Effect Included in Accuracy 2 Input 4.3, Section 2.3-EPU lnput 4.2, Section 2.3-CLTP and Comment 5; nla for APRM Power Input 4.3, Section 2.3-EPU 0 SPE Over ressure Effect electronic nla for APRM Power Static Pressure Effect electronic

( nla nla

(

I and Comment 5 lnput 4.2, Section 2.3-CLTP and Comment 5 ;

Input 4.3, Section 2.3-EPU and Comment 5 lnput 4.2, Section 2.3-CLTP and Comment 4; lnput 4.3, Section 2.3-EPU SE Seismic Effect Included in Accuracy 2 and Comment 4 lnput 4.2, Section 2.3-CLTP and Comment 4; lnput 4.3, Section 2.3-EPU RE Radiation Effect Negligible 2 and Comment 4 lnput 4.2, Section 2.3-CLTP and Comment 4; Input 4.3, Section 2.3-EPU HE Humidity Included in Accuracy 2 and comment 4 lnput 4.2, Sections 1 and 2.3-CLTP, Comment 4&16; Power Supply Effect Input 4.3, Sections 1 and 2.3, PSE (LPRM Detector) Negligible 2 Comment 4&16-EPU lnput 4.2, Section 2.3-CLTP and Comment 4; lnput 4.3, Section 2.3-EPU and REE RFllEMl Effect Negligible 2 Comment4

MONTlCELLO NUCLEAR GENERATlNG PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow I I Biased PRNM Setwoints for CLTP and EPU I Paae 17 of 58 1 7.3 Loop Instrument Uncertainty Evaluation 7.3.1 CLTP Operation Loop Instrument Uncertaintv; The loop uncertainty associated with the replacement of analog neutron monitoring system with the digital NUMAC PRNMS for CLTP operation is discussed and calculated in this section.

7.3.1.I Loop lnstrument Accuracv ( A d Loop lnstrument Accuracy (AL) is defined as the accuracy of the LPRM flux channel electronics. The GEH lnput documents, lnput 4.2 and 4.3, for the digital PRNMS specify a LPRM flux channel electronic accuracy for PRNM chassis. The accuracy of the LPRM detector is specified as APRM PEA or APEA (Accuracy) as defined in Section 7.2.1.3. Since the accuracy of the LPRM detector is included in APEA (Accuracy), it will be used in the calculation of Loop Primary Element Accuracy (APEAS. Calculation of Loop Accuracy (AL) will use the LPRM Electronics accuracy value.

Loop Instrument Accuracy depends on number of LPRMs averaged by the PRNM APRM. As indicated below, the statistical average is based on lnput 4.12, which is shown in Attachment 4. The fewer LPRMs averaged, the greater the accuracy error. Averaging the minimum number of LPRM detectors (14) versus a larger number (24) equals the following accuracy errors:

Averaging Calculation Example:

The following shows how number of LPRM detectors affects the accuracy error. For example, the APEA for each LPRM is +_l.O%RTP a bias of 0.49 %

RTP as indicated in Section 7.1.2.3.

lnput 4.12 (Attachment 4) can be interrupted as the following:

Accuracy Error = Accuracy(perLPRM) n = number of LPRM detectors used J;5 14 LPRM Detectors: Accuracy Error = f l'"%RTP = rt 0.267 % RPT J14 24 LPRM Detectors: Accuracy Error = a l-"%RTP =, 0.204 % RPT J24 For conservatism, the minimum number of LPRM detectors (14) is used in the accuracy calculations.

. MONTlCELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 18 of 58 Individual Device Accuracy Device 1: LPRM Detector (A11 As stated above, the LPRM Detector accuracy is included in APRM PEA (APEA) term. Section 7.3.1.6 calculates Loop Primary Element Accuracy (APEAS using this term. Therefore, LPRM flux channel electronics accuracy for Device 1 will be considered 0 % RPT.

A1 = LPRM flux channel electronics accuracy for LPRM Detector A1 = 0 % RPT since it is included in APEA term. Section 7.2.1.3 Device 2: Power Electronics (NUMAC PRNMS) (A21 As stated above, the PRNM chassis electronics for the flux generated analog signal is the term below for one LPRM detector.

Accuracy (LPRM Electronics) =

• 0.943 % RTP. (Section 7.2.2.3)

GEH does not breakdown the components of the LPRM flux channel electronics accuracy in their submitted PRNM documents. The calculation of the LPRM electronics accuracy is considered proprietary. The LPRM module, which receives analog input from the LPRM detector, is part of the new PRNM chassis. Another component is the digital processing of the LPRM detector signal. The LPRM electronics accuracy specified applies to one LPRM detector.

The APRM averages the LPRM signals to obtain reactor power indication.

The minimum number of LPRMs is 14 in accordance Section 7.2.1.I.

Per Input 4.12, the accuracy will be 1/14 of the square root of the sum of the squares of the 14 LPRMs as expressed by:

A2 = Overall (mean) accuracy for the LPRM flux channel Electronics A;!= + 0.253 % RTP LOODInstrument Accuracy (AIJ Al = + 0.00 % RTP A2= f 0.253 % RTP

MONTICELLONUCLEAR GENERATING PLANT- CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 19 of 58 AL = t J(0.000)~ + (0.0253)'

+

AL= 0.253 Oo/ RTP 7.3.1.2 Loop lnstrument Drift (DIJ Loop lnstrument Drift (DS is defined as the Square Root of the Sum of the Squares (SRSS) of the individual Device drifts.

For this calculation, Section 7.2.1.3 states the drift of the LPRM detectors (Device 1) is included in DPEA. Discussions with GE lnstrument Engineers confirmed that the total drift error for GE LPRM detectors is known and is accurate for the LPRM detector Random drift error and the Bias drift error components.

For Device 2, Section 7.2.2.3 states the NUMAC PRNM Power Electronics drift is 0.5 % RTP Full Span when calibration every 700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br />. The source of this drift value is Input 4.2.

Methodology, Section 2, states that this calculation uses GE specified drift parameters for the applicable PRNM equipment and for the existing LPRM detectors. Standards Manual ESM-03.02-APP-I, Section 5.2.4, shows a alternate methodology for determination of individual device drift. For this application, it is considered more accurate to use GE specified device drift values since both PRNM equipment and LPRM detectors are provided by the manufacturer.

Individual Device ~ h f t Device 1

+

Drift of Device 1 = Dl = 0.00 % RTP (FS) 1700 days because the LPRM (Device 1) drift is included in DPEAas specified in Section 7.2.1.3.

Device 2 Drift of Device 2 = D2 = f 0.50 % RTP FS 1700 Hours (Section 7.2.2.3)

Since the APRMs will be calibrated against the reactor heat balance every 7 days, 700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> drift value is conservative. Converting percent full span (FS) to the percent power yields:

D2= f 0.50 x (125 % RTPI 100 % RTP FS)

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 20 of 58

+

DP= 0.625 % RTP Loop lnstrument Drift (DL).

Dl = k 0.000 % RTP D2 = f 0.625 % RTP Loop lnstrument Drift = DL= k J(D,)~ +(D~)'

DL= f J(0.000)~ + (0.625)~ =

• 0.625 % RTP 7.3.1.3 Loop As-Left Tolerance (ALT3 The loop As-Left tolerance (ALTL) is being evaluated from two perspectives.

The first is based on GEH lnput 4.2, Section 2.3 and Comment 11. lnput 4.2 states that the As-Left Tolerance is equal to the Auto Gain Adjustment Factor (AGAF), which is +_ 2.00 % RTP at 3 0 . lnput 4.2, Comment 11, states the basis for AGAF equaling -?I 2.00 % RTP is as follows:

The APRM subsystem is calibrated every 7 days using the AGAF process, where the gain of the APRMs is adjusted to read the Rated Thermal Power (RTP), also called Core Thermal Power, determined by the Process Computer (PIC), within a specified As-Left Tolerance. This is equivalent to a standard calibration of the APRM electronics sub-loop (consisting of the LPRM and APRM signal conditioning electronics), where the PIC is the calibration tool and standard. The PIC and heat balance error is already accounted for in the transient analyses. Thus, the only calibration error to consider for the APRM electronics sub-loop is the As-Left Tolerance specified by the AGAF process.

A. Loop As-Left Tolerance (ALTL) used in uncertainty calculations is determined by the AGAF process:

As described above, the basis for this ALTL calculation is GEH lnput 4.2, Section 2.3.

ALTL= AGAF which is equal to: 5 2.00 % RTP at 30. (Section 7.2.2.3)

The tolerance is normalized to a 2 cr confidence level in accordance with ESM-03.02- APP-I, Section 4.3 (Input 4.1),

Converting to 2 o :

ALTL = +- 2.00 x (213) % RTP at 20.

ALTL= k 1.334 % RTP for use in uncertainty calculations The second analysis of ALTL is based on PRNM surveillance calibration of PRNM electronics. The LPRM detector loop is not involved. The electronics

MONTICELLO NUCLEAR GENERATlNG PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow L Biased PRNM Setpoints for CLTP and EPU Page 21 of 58 being calibration checked is PRNM digital equipment. The setpoints being checked are numerical values stored in the digital hardware and are not subject to drift.

B. ALTL used for PRNM surveillance calibration procedures:

ALTL= 0.00 % RTP based on PRNM digital hardware without LPRM detectors 7.3.1.4 Loop As-Found Tolerance (AFTd The As-Found Tolerance (AFTL) is being evaluated from two perspectives similar to ALTL, Section 7.3.1.3. The results are indicated below:

A. Loop As-Found Tolerance (AFTL),used in loop uncertainty calculations, is determined by GEH Input 4.2, Section 2.3, which states:

AFTL = ALTL (Section 7.2.2.3)

ALTL= +_ 2.00 % RTP at 3 0 as defined in Section 7.3.1.3.

The tolerance is normalized to a 2 0 confidence level In accordance with ESM-03.02-APP-I, Section 4.3 (Input 4.1).

AFTL = + 2.00 % x (213) % RTP at 2 cr

+

AFTL = 1.334 % RTP for use in uncertainty calculations The tolerance is normalized to a 2 a confidence level in accordance with ESM-03.02- APP-I, Section 4.3 (Input 4.1),

Converting to 2 a :

AFTL = 2.00 x (213) % RTP at 2 a.

+_

AFTL= 2 1.334 % RTP used for uncertainty calculations The second analysis of AFTL is based on PRNM surveillance calibration of only PRNM electronics. The LPRM detector loop is not involved. The setpoints being checked are numerical values stored in the digital hardware and are not subject to drift.

6 . AFTL used for PRNM surveillance calibration procedures:

AFTL= 0.00 % RTP based on PRNM digital hardware without LPRM detectors

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 22 of 58 ,

7.3.1.5 Loop Calibration Accuracy Error (CIJ In accordance with GEH specification in Input 4.2, Comment 11, the only calibration error to consider for the APRM electronics sub-loop is the loop As-Left Tolerance (ALTL) specified by the AGAF process. Calibration Accuracy Error (CL) is the As-Left Tolerance (AFTL) defined for uncertainty calculations.

Loop Calibration Accuracy Error = CL CL= ALTL= k 1.334 % RTP (Section 7.3.1.3) 7.3.1.6 Loop Primary Element Accuracv (APEAL)

APEA is equal to the Random Accuracy per LPRM detector plus the Bias Accuracy Error per LPRM detector. Section 7.2.1.3 indicates that the Power Supply Effect of the LPRM Detector in included in the APEA. This is in accordance with Inputs 4.2 and 4.3.

APEA = Random Accuracy ErrorILPRM detector + Bias AccuracyILPRM Detector

• +

Random Accuracy Error = APEAR = 1.OO % RTPI LPRM (Section 7.2.1.3)

Bias Accuracy Error = APEAb= 0.49 % RTP bias APEA = + 1.00 % RTPI LPRM + 0.49% RTP bias1LPRM (Section 7.2.1.3)

Loop Primary Element Accuracy (APEAL)= overall (mean) accuracy APEAL is equal to the Random Accuracy per LPRM detector divided by the square root of the minimum number of LPRM detectors plus the Bias Accuracy Error per LPRM detector.

APEAL = 'l."%RTP fi

+ 0.49%RTPbias (Input 4.12)

For conservatism, Bias terms are positive.

+

APEAL = 0.268 % RTP + 0.49 % RTP bias

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setooints for CLTP and EPU Paae 23 of 58 DPEA is equal to the Random Drift Error per LPRM detector plus the Bias Drift Error per LPRM detector. (Section 7.2.1.3)

Therefore, DPEA is defined as:

DPEA = Random Drift ErrorILPRM detector + Bias Drift ErrorILPRM Detector Random Drift Error = DPEAR= & 0.20 % RTPI LPRM (Section 7.2.1.3)

Bias Drift Error = DPEAb=0.33 % RTP bias DPEA = & 0.20 % RTPI LPRM + 0.33 % RTP bias1 LPRM (Section 7.2.1)

Loop Drift Primary Element Accuracy (DPEAL)= overall (mean) accuracy DPEALis equal to the Random Drift Error per LPRM detector divided by the square root of the minimum number of LPRM detectors plus the Bias Drift Error per LPRM detector.

(Input 4.12)

For conservatism, the Bias terms are positive.

APEAL = k 0.054 % RTP + 0.33 % RTP bias 7.3.1.8 Loop APRM Process Measurement Accuracy (APMAd APM&a&ing = k 1.11 % RTP (Section 7.2.2.3)

APM&Ois,= + 2.00 % RTP (Section 7.2.2.3)

APMAL= +- J(l .I 1%)' + (2.00%)~

APMAL= k 2.288 % RTP

MONTICELLO NUCLEAR GENERATlNG PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 24 of 58 7.3.1.9 -

Tabulation of Loop Uncertainties PRNM CLTP Operation Uncertainty Random Bias Type Term f % RTP + % RTP Section AL Loop Instrument Accuracy 0.253 7.3.1.1 DL Loop Instrument Drift 0.625 7.3.1.2 ALTL Loop As-Left Tolerance (uncertainty) for uncertainty calculations 1.334 7.3.1.3 ALTL Loop As-Left Tolerance for (calibration) electronic calibrations 0.00 7.3.1.3 AFTL Loop As-Found Tolerance (uncertainty) for uncertainty calculations 1.334 7.4.1.4 AFTL Loop As-Found Tolerance (calibration) for electronic calibrations 0.00 7.3.1.4 Loop Calibration CL Accuracy Error 1.334 7.3.1.5 Loop APRM Primary APEAL Element Accuracy 0.268 7.3.1.6 Loop Drift Primary Element DPEAL Accuracy 0.054 7.3.1.7 Loop APRM Process APMAL Measurement Accuracy 2.288 7.3.1.8 APRM Primary Element APEAb Accuracy Bias 0.49 7.3.1.6 Drift Primary Element DPEAb Accuracy Bias 0.33 7.3.1.7

I CA-08-050 1

MONTlCELLO NUCLEAR GENERATING PLANT TITLE: I I

Instrument S e t ~ o i nCalculation t

Average Power Range oni it or (APRM) Non-Flow 1-1 1 Revision 0 1 Biased PRNM Setpoints for CLTP &d EPU ( Page 25 of 58 1 7.3.2 EPU Operation Loop Uncertainty The comparison of Inputs 4.2 and 4.3 shows the individual device uncertainties associated with the EPU operation are identical to the individual device uncertainties associated with the CPTP operation. The uncertainty terms tabulated in Sections 7.2.1 and 7.2.2 list both CLTP (Input 4.2) and EPU (Input 4.3) references. Operating at EPU condition will not change the methodology used to combine the individual device uncertainties to produce the loop uncertainties. Therefore, the loop uncertainties calculated for the PRNM CLTP operation (Section 7.3.1) will be applicable to the PRNM EPU Operation (Section 7.3.2). The EPU uncertainties are tabulated in the table below:

7.3.2.1 Tabulation of Loop Uncertainties - PRNM EPU Operation Tabulation of Loop Uncertainties - PRNM EPU Operation Uncertainty Random Bias Type Term +-%RTP + % RTP Section AL Loop Instrument Accuracy 0.253 7.3.1 .I DL Loop Instrument Drift 0.625 7.3.1.2 ALTL Loop As-Left Tolerance (uncertainty) for uncertainty calculations 1.334 7.3.1.3 ALTL Loop As-Left Tolerance for (calibration) electronic calibrations 0.00 7.3.1.3 AFTL Loop As-Found Tolerance (uncertainty) for uncertainty calculations 1.334 7.3.1.4 AFTL Loop As-Found Tolerance (calibration) for electronic calibrations 0.00 7.3.1.4 Loop Calibration Accuracy

- CL Error 1.334 7.3.1.5 Loop APRM Primary APEAL Element Accuracy 0.268 7.3.1.6 Loop Drift Primary Element

- DPE& Accuracy 0.054 7.3.1.7 Loop APRM Process APM4 Measurement Accuracy 2.288 7.3.1.8 APRM Primary Element APE& Accuracy Bias 0.49 7.3.1.6 Drift Primary Element DPEAb Accuracy Bias 0.33 7.3.1.7

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 26 of 58 7.4 PRNM CLTP Operation Setpoint Evaluation 7.4.1 PRNM CLTP APRM Neutron Flux - High Scram lnput 4.2, Section 3, states the following Analytical Limit (AL), the recommended Allowable Value (AV) and the Nominal Trip Setpoint (NTSP) for CLTP operation with NUMAC - PRNM equipment installed.

PRNM CLTP Setpoint AL AV NTSP I APRM Neutron Flux - High Scram 125.0 1 122.0 ( 119.5 The following calculations will determine the minimum required margin between the specified AL of 125.0 % RTP, AV of 122.0 % RTP and NTSP of 119.5 %

RTP.

7.4.1.1 Allowable Value (AV) Evaluation lnput 4.11, Section 1.2.3.2, provides the following formulas for calculating the AV from the AL:

1.645 AV AL -(T),/(AL)2 + (CL)' + (PMA)' + (PEA)' k6im Minimum required margin between AL and AV can be defined by:

Minimum Required Marqin (AL-AV) = RM = AL - AV R M = A L - A V = 2 . 6 9 % RTP

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Set~ointsfor CLTP and EPU Paae 27 of 58 Available Margin (AL - AV) =AM = AL - AV AM = AL - AV = (125.0 - 122.0) % RTP A M = A L - A V = 3 . 0 0 % RTP Available Minimum PRNM CLTP Setpoint A1 AV Margin Required AV

(% RTP) - - (AL - AV)

- -Acceptable Margin APRM Neutron Flux - 125.0 122.0 3.0 2.69 Yes High Scram Since the available margin is greater than the minimum required margin (3.0 % RTP versus 2.69 % RTP), the recommended AV is acceptable.

7.4.1.2 Nominal Trip Setpoint (NTSP) Evaluation Input 4. I1, Section 1.2.3.3 provides the following formula for calculating the NTSP from the AL:

Minimum required margin between AL and NTSP can be defined by:

RM = AL - NTSP =

RM = AL - NTSP = 3.08 % RTP

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 28 of 58 Available Marqin (AL - NTSP) = AM = AL - NTSP AM = AL - NTSP = (125.0 - 119.5) % RTP = 5.50 % RTP Available Minimum PRNM CLTP AL NTSP Margin Required NTSP Setpoint (% RTP) (AL-NTSP) Margin Acceptable APRM Neutron Flux - 125.0 119.5 5.5 3.08 Yes High Scram Since the available margin is greater than the minimum required margin (5.5 %

RTP versus 3.08 % RTP), the recommended NTSP is acceptable.

7.4.1.3 Licensee Event Report (LER) Avoidance Test The purpose of the LER Avoidance Test is to assure that there is sufficient margin between the AV and NTSP to reasonably avoid violations of the AV.

lnput 4.11, Section 1.2.3.5, provides the following formula for determining LER avoidance criteria.

Where n is the number of standard deviations used (20) o,, = (;),/(0.253)' + (1.334)' + (0.625)'

cr, = 0.75 % RTP For multiple instrument channels, a Z value of greater than 0.81 provides sufficient margin between NTSP and AV as specified in ESM-03.02-APP-I, Section 5.6.3 (Input 4.1).

In accordance with lnput 4.1, Section 5.6.3, Z is defined as:

z=

IAV - NTSP~ -- l(122.0 - 11 9 . 5 ) % ~ ~=~3.33 1

OLER (0.75)% RTP Since Z is greater than 0.81, sufficient margin exists between the specified AV and NTSP.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 29 of 58 Minimum margin NTSP2 will be calculated using the minimum Z value of 0.81.

Minimum margin NTSP2 will be compared with the current NTSP of 119.5 %

RTP. This will indicate the amount of conservatism for the NTSP.

NTSP2Offset is defined as the minimum margin (% RTP) between AV and NTSP2with Z of 0.81 used.

NTSP2Offset = Z x o,,

NTSP2Offset = (0.81 x 0.75) % RTP = 0.61 % RTP NTSP2 is calculated to provide an NTSP based on the minimum LER avoidance criteria:

NTSP2 I AV - NTSP2Offset NTSP2 5 122.0 % RTP - 0.61 % RTP NTSP2 1 121.4 % RTP For the minimum valve of Z equal to 0.81, NTSP2, which is defined as the LER Avoidance NTSP, is to be less than 121.4 % RTP.

A conservative NTSP of 119.5 % RTP is used. Attachment 1, Setpoint Diagrams, shows margin for the APRM Neutron Flux - High Scram setpoint.

7.4.1.4 Spurious Trip Avoidance (STA) Test A Spurious Trip Avoidance test is performed to assure that there is a reasonable probability that spurious trips will not occur using the selected setpoint.

lnput 4.1 1, Section 1.2.3.4, provides the following formula for determining Spurious Trip avoidance. lnput 4.13, Sections 4.5.4.b and 4.5.9.b, state that any bias associated with PMA or PEA should also be included. Therefore, APEAb shown in CLTP Loop Uncertainty table in Section 7.3.1.9 is being included.

Where n is the number of standard deviations used ( 2 0 )

DPEALand DPEhwere defined and evaluated in Section 7.3.1.7. These terms are to be included in the above om equation as follows:

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setooints for CLTP and EPU Paae 30 of 58 Using APRM terms, om is defined as:

+(cL12 onA = ( : ) J ( A ~ ) ~ +(D~)' + ( A P M ) I +(APEA)' + ( D P E A ~+

) ~A P + D~P E ~A ~

om = 2.20 % RTP lnput 4.1, STA Section 5.6.8, states Z is equal to the following:

[(~dj~- t e d - (Operational - ~ i m i t ) l

~ NTSP) z=

(DSTA) lnput 4.1, Section 5.6.8, states that Adjusted NTSP is the selected setpoint minus ALTL (more conservative). Since Adjusted NTSP is (NTSP-ALTL),Z is equal to the following:

z= NTSP - ALT - (Operational -Limit)

(OSTA)

In lnput 4.2, Sections 1.3 and 1.7, GEH defines the Operational Limit (OL) as 100 % RTP for MNGP. Therefore:

As specified in lnput 4.1, Section 5.6.8, Z should be equal to or greater that 1.65 for the setpoint to be adequately separated from the Operational Limit to reasonably avoid Spurious trip conditions.

Since the actual value of 8.25 is greater than the required value of 1.65, an adequate separation exists between the NTSP and the Operational Limit (OL),

and the STA criterion is satisfied.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 31 of 58 7.4.2 PRNM CLTP: APRM Neutron Flux - High (Setdown) Scram and APRM Neutron Flux - Hiqh (Setdown) Rod Block lnput 4.2, Section 3, states that the following Analytical Limit (AL) and the following recommended Allowable Valve (AV) and the Nominal Trip Setpoint (NTSP):

PRNM CLTP Setpoints

(% RTP) AL AV NTSP APRM Neutron Flux - High (Setdown) Scram NIA 20.0 18.0 APRM Neutron Flux - High (Setdown)

Rod Block NIA 15.0 13.0 lnput 4.2 shows that Setdown Scram and Setdown Rod Block functions do not have AL. These functions only have AV and NTSP. This calculation will determine the minimum required margin for AL to AV, and AL to NTSP. The difference between the minimum required margins for AL to AV, and AL to NTSP becomes the minimum required margin for AV to NTSP.

7.4.2.1 Minimum Reauired Margin (AL to AW Evaluation lnput 4.1 1, Section 1.2.3.2, provides the following formula for calculation the AV from the AL.

1.645

+

A Y I AL -( - - I - - ) ~ ( A , )' + ( C , )' + ( P M ) ~+ (PEA)' bias Minimum required margin between the AL and AV and be defined by:

Minimum Reauired Margin (AL - AV) = RM = AL - AV RM = AL - AV = 2.69 % RTP

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU I Page 32 of 58 1 7.4.2.2 Minimum Required M a r ~ i n(AL to NTSP) Evaluation Input 4. I1, Section 1.2.3.3, provides the following formula for calculating the AV from the AL:

Minimum required margin between AL and NTSP can be defined by:

(?)J(A, )' + (C, )' + (DL)'+ (APMA, )' + (APEA, )2 + (DPEA, )' + APEA, + DPEA, Minimum Required Marqin (AL - NTSPZ = RM = AL - NTSP RM = AL - NTSP =

RM = AL - NTSP = 3.08 % RTP 7.4.2.3 Minimum Required Marain (AV to NTSP) Evaluation RM = AV- NTSP=

MinimumRequired Margin (ALto NTSP) -Minimum RequiredMargin (AL to AV)

RM = AV to NTSP = (3.08 - 2.69) % RTP RM = AV to NTSP = 0.39 % RTP I PRNM CLTP Setpoints

(% RTP)

I AV 1 NTSP 1I Available Margin (AV-NTSP)

Required APRM Neutron Flux - 1 (Setdown) Scram 20.0 18.0 2.0 0.39 1 Yes APRM Neutron Flux -

(Setdown) Rod Block 15.0 13.0 2.0 0.39 1 Yes Since the available margin is greater than the minimum required margin (2.0 %

RTP versus 0.39 % RTP), the recommended NTSP is acceptable.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow 1 Biased PRNM Setpoints for CLTP and EPU I Page 33 of 58 1 7.4.2.4 Licensee Event Report (LER) Avoidance Test The purpose of the LER Avoidance Test is to assure that there is sufficient margin between the AV and NTSP to reasonably avoid violation of the AV.

lnput 4.1 1, Section 1.2.3.5, provides the following formula for determining LER avoidance criteria.

where n is the number of the standard deviations used (20)

,a = (1)~(0.253)' + (1 334)' + (0.625)'

om = 0.75 % RTP For multiple instrument channels, a Z value of greater than 0.81 provides sufficient margin between the NTSP and AV as specified in ESM-03.02-APP-I, Section 5.6.3 (Input 4.1).

In accordance with lnput 4.1, Section 5.6.3, Z is defined as:

z=

IAV - NTSPJ

~ L E R For APRM Neutron Flux - High (Setdown) Scram setpoint:

For APRM Neutron Flux - High (Setdown) Rod Block setpoint:

Since Z is greater than 0.81, sufficient margin exists between the specified AV and NTSP.

Minimum margin NTSP2 will be calculated using the minimum Z value of 0.81.

Minimum margin NTSP2 will be compared with the current NTSP of 18.0 %

RTP. This will indicate the amount of conservatism for the NTSP.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow

~ i a s e dPRNM setpoints for CLTP and EPU I Page 34 of 58 I NTSP2 Offset is defined as the minimum required margin between AV and NTSP with Z of 0.81 used.

NTSP2Offset = Z x ,a ,

NTSP2Offset = (0.81 x 0.75) % RTP = 0.61 % RTP APRM Neutron Flux - Hiah (Setdown) Scram Setpoint For APRM Neutron Flux - High (Setdown) Scram setpoint: NTSP2 is calculated to provide a NTSP based on the minimum LER avoidance criteria:

NTSP2 I AV - NTSP2Offset NTSP2(setd, smm) I(20.0 - 0.61) % RTP NTSP2(stdomscram, I19.4 % RTP For the minimum valve of Z equal to 0.81, NTSP2,which is defined as the LER Avoidance NTSP, is to be less than 19.4 % RTP.

Therefore, a conservative NTSP of 18.0 % RTP is used for the APRM Neutron Flux - High (Setdown) Scram setpoint.

APRM Neutron Flux - Hiah (Setdown) Rod Block Setpoint For APRM Neutron Flux - High (Setdown) Rod Block setpoint: NTSP2is calculated below to indicate margin of the recommended NTSP:

NTSP2 I AV - NTSP2Offset NTSP2(setdown~ o BdI O ~ ) < ( I 5.0 - 0.61) % RTP NTSP2(setdomR O B~I O C ~ ) I 14.4 5% RTP For the minimum valve of Z equal to 0.81, NTSP2, which is defined as the LER Avoidance NTSP, is to be less than 14.4 % RTP.

Therefore, a conservative NTSP of 13.0 % RTP is used for the APRM Neutron Flux - High (Setdown) Rod Block setpoint.

Attachment 1, Setpoint Diagrams, shows the setpoint margin for CLTP APRM Neutron Flux - High (Setdown) Scram and APRM Neutron Flux - High (Setdown) Rod Block setpoints.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 35 of 58 7.4.2.5 Spurious Trip Avoidance (STA) Test A Spurious Trip Avoidance test is performed to assure that there is a reasonable probability that spurious trips will not occur using the selected setpoint.

lnput 4.1 1, Section 1.2.3.4, provides the following formula for determining Spurious Trip avoidance. lnput 4.13, Sections 4.5.4.b and 4.5.9.b, provides setpoint guidance to add bias term for APEA.

Where n is the number of standard deviations used ( 2 0 )

Terms DPEALand DPE& were defined and evaluated in Section 7.3.1.7.

These terms are to be included in the above omequation as follows:

Using APRM terms, om is defined as:

,0 = 2.20 % RTP lnput 4.1, STA Section 5.6.8, states Z is equal to the following:

I(Adjusted - NTSP) - (Operational - ~imit)l z=

~ S T A lnput 4.1, Section 5.6.8, states that Adjusted NTSP is the selected setpoint minus ALTL (more conservative). Since Adjusted NTSP is (NTSP-ALTL),Z is equal to the following:

MONTICELLO NUCLEAR GENERATING PLANT ( CA-08-050 TITLE: \ Instrument Set~ointCalculation - 1 Revision 0 J Average Power Range 'Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 36 of 58 -

z= [NTSP - ALT - (Operational - Limit)]

CSTA lnput 4.2, Section 1.I, defines the Operational Limit for the APRM Neutron Flux

- High (Setdown) Scram as 11.0 % RTP.

As specified in lnput 4.1, Section 5.6.8, Z should be equal to or greater that 1.65 for the setpoint to be adequately separated from the Operational Limit to reasonably avoid Spurious trip conditions.

Since the actual value of 2.57 is greater than the required value of 1.65, an adequate separation exists between the NTSP and the Operational Limit (OL),

and the STA criterion is satisfied.

7.4.3 PRNM CLTP - APRM Downscale Rod Block lnput 4.2 for PRNM CLTP setpoints does not address the Downscale Rod Block setpoint for PRNM CLTP operation. lnput 4.8, ltem 12, provides recommended AV and NTSP setpoints indicated below.

The PRNM CLTP APRM Downscale Rod Block setpoint is the same as the current CLTP setpoint.

PRNM CLTP Setpoint

% RTP AL AV NTSP APRM Downscale Rod Block NIA 2.0 3.5 lnput 4.8, ltem 12, indicates that the APRM Downscale Rod Block function does not have an AL. This calculation will determine the minimum required margin for AL to AV, and AL to NTSP. The difference between the minimum required margins for AL to AV, and AL to NTSP becomes the minimum required margin for AV to NTSP.

7.4.3.1 Minimum Rewired Marqin (AL to AV) Evaluation lnput 4.1 1, Section 1.2.3.2, provides the following formula for calculation the AV from the AL.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 37 of 58 Minimum required margin between AL and AV can be defined by:

(?)&A, ) ' + (C, )' + (APMa,)' + APE^, )' + APEA, Minimum Required Marain (AL - AV) = RM = AL - AV R M = A L - A V = 2 . 6 9 % RTP 7.4.3.2 Minimum Required Marain (AL to NTSP) Evaluation Input 4.1 1, Section 1.2.3.3, provides the following formula for calculating the AV from the AL:

NPsP 5 A L - ( F ) 4 ( A L ) ' +(C,)' +(D,)'+(PMA)' +(PEA)' f bias Minimum required margin between AL and NTSP can be defined by:

(?)&A,)' + ( c , ) ~+(DL)' + ( A P M ~ +) ~(APEA,)' + ( D P E A , ) ~+APE% + DPEA, Minimum Required Margin (AL - NTSP) = RM = AL - NTSP RM = AL - NTSP =

(?)~(0.253)~ + (1 .33412 + (0.625)~+ (2.288)2+ (0.268)2 + ( 0 . 0 5 4 ) ~+ 0.49 + 0.33 RM = AL - NTSP = 3.08 % RTP

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow 1 1 6iased PRNM ~ e k o i n t s for CLTP a;~d EPU I Paae 38 of 58 1 7.4.3.3 Minimum Required Margin (AV to NTSP) Evaluation = RM = AV to NTSP RM = AV to NTSP =

Minimum Required Margin (AL to NTSP) -Minimum Required Margin (AL to AV)

RM = AV to NTSP = 3.08 % - 2.69 % % RTP RM = AV to NTSP = 0.39 % RTP Available Minimum PRNM CLTP AV NTSP Margin Required AV Setpoint (% RTP) ( NTSP-AV) Margin Acceptable APRM Downscale Rod Block 2.0 3.5 1.5 0.39 Yes Since the available margin is greater than the minimum required margin (1.5 %

RTP versus 0.39 % RTP), the recommended NTSP is acceptable.

7.4.3.4 Licensee Event Report (LER) Avoidance Test The purpose of the LER Avoidance Test is to assure that there is sufficient margin between the AV and NTSP to reasonably avoid violation of the AV.

lnput 4.1 1, Section 1.2.3.5, provides the following formula for determining LER avoidance criteria.

where n is the number of the standard deviations used (2 o )

= (f)~(0.253)' + (1 .33412 + (0.625)~

o m = 0.75 % RTP For multiple instrument channels, a Z value of greater than 0.81 provides sufficient margin between the NTSP and AV as specified in ESM-03.02-APP-I, Section 5.6.3 (Input 4.1).

In accordance with lnput 4.1, Section 5.6.3, Z is defined as:

MONTICELLO NUCLEAR GENERATING PLANT 1 CA-08-050 I 1I I 1

Instrument Setpoint Calculation Averaae Power Range Monitor lAPRMI Non-Flow

~iised PRNM ~ e k o i n t sfor CLTP ;nd EPU 1I Revision 0 Paae 39 of 58 1 Since Z is greater than 0.81, sufficient margin exists between the specified AV and NTSP.

Minimum margin NTSP2will be calculated using the minimum Z value of 0.81.

Minimum margin NTSP2will be compared with the current NTSP of 3.5 % RTP This will indicate the amount of conservatism for the NTSP.

NTSP2 Offset is defined as the minimum margin (% RTP) between AV and NTSP2 with Z of 0.81.

NTSP2 Offset = Z x om NTSP2 Offset = (0.81 x 0.75) % RTP = 0.61 % RTP NTSP;! is calculated to provide an NTSP based on the minimum LER avoidance criteria:

NTSP2 2 AV + NTSP2Offset NTSP2 2 2.0 % RTP + 0.61 % RTP NTSP2 2 2.61 % RTP For the minimum valve of Z equal to 0.81, NTSP2,which is defined as the LER Avoidance NTSP, is to be greater than 2.61 % RTP.

Therefore, a conservative NTSP of 3.5 % RTP is used. Attachment 1, Setpoint Diagrams, shows the setpoint margin for CLTP APRM Downscale Rod Block setpoint.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Set~ointsfor CLTP and EPU Paae 40 of 58 7.5 PRNM EPU Operation Setpoint evaluation 7.5.1 PRNM EPU APRM Neutron Flux - Hiqh Scram The setpoints that will be implemented for APRM Neutron Flux - High are defined in lnput 4.8, Item 8. lnput 4.3, GE's recommended setpoint document, has slight differences in the setpoints. As shown in the table below, lnput 4.8 specifies slightly more conservative values for AV and NTSP. These EPU setpoints are the same values being used for PRNM CLTP operation.

PRNM EPU Setpoints (% RTP) AL AV NTSP APRM Neutron Flux - High Scram Input 4.8, Item 8 125.0 122.0 119.5 GE Recommended Setpoints (not used)

APRM Neutron Flux - High Scram Input 4.3 125.0 122.3 120.3 7.5.1 . I Allowable Value (AW Evaluation lnput 4.1 1, Section 1.2.3.2, provides the following formula for calculating the AV from the AL.

1.645 AY I AL -(1),/(A,)2 + (c,)' + (PMA)' + PEA)^ & bim Minimum required margin between AL and AV can be defined by:

Minimum Required Marqin (AL - AV) = RM = AL - AV

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 41 of 58 Available Marain (AL - AV) = AM = AL - AV AM = AL - AV = (125.0 - 122.0) % RTP A M = A L - A V = 3 . 0 % RTP Available Minimum EPU Setpoint AL AV Margin Required AV

(% RTP) (AL- AV) Margin Acceptable APRM Neutron Flux -

High Scram 125.0 122.0 3.0 2.69 Yes Since the available margin is greater than the minimum required margin (3.0 %

RTP versus 2.69 % RTP), the recommended AV is acceptable.

7.5.1.2 Nominal Trip Setpoint (NTSP) Evaluation Input 4.1 1, Section 1.2.3.3, provides the following formula for calculating the AV from the AL.

1.645 NPSP <_ AL - ( T ) , / ( A L ) 2 + (C,)' + (D,)' + (PMA)' + (PEA)' _+bias Minimum required margin between AL and NTSP can be defined by:

Minimum Required Marsin (AL - NTSP\= RM = AL - NTSP RM = AL - NTSP = 3.08 Oh RTP

I CA-08-050 MONTlCELLO NUCLEAR GENERATING PLANT TITLE: 1 Instrument Setpoint Calculation -

Average Power Range Monitor (APRM) Non-Flow I Revision 0 1 Biased PRNM Setpoints for CLTP and EPU Page 42 of 58 Available M a r ~ i n(AL - NTSP) = AM = AL - NTSP AM = AL - NTSP = (125.0 - 119.5) % RTP AM = AL - NTSP = 5.5 % RTP Available Minimum PRNM EPU Setpoint AL NTSP Margin Required AV

(% RTP) (AL- NTSP) Margin Acceptable APRM Neutron Flux 125.0 119.5 5.5 3.08 Yes

- High Scram Since the available margin is greater than the minimum required margin (5.5 %

RTP versus 3.08 % RTP), the recommended the NTSP is acceptable.

7.5.1.3 Licensee Event Report (LER) Avoidance Test The purpose of the LER Avoidance Test is to assure that there is sufficient margin between the AV and NTSP to reasonably avoid violations of the AV.

lnput 4.11, Section 1.2.3.5, provides the following formula for determining LER avoidance criteria.

where n is the number of the standard deviations used ( 2 0 )

0, = (t) ~(0.253)'+ (1.334)' + (0.625)'

oL, = 0.75 % RTP For multiple instrument channels, a Z value of greater than 0.81 provides sufficient margin between NTSP and AV as specified in ESM-03.02-APP-1, Section 5.6.3 (Input 4.1).

In accordance with lnput 4.1, Section 5.6.3, Z is defined as:

z=

IAV - NTSPI -- l(122.0 - 11~ . ~ ) % R T P=/ 3.33 OLER (0.75)%RTP Since Z is greater than 0.81, sufficient margin exists between the specified AV and NTSP.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 43 of 58 Minimum margin NTSP2 will be calculated using the minimum Z value of 0.81 Minimum margin NTSP2will be compared with the current NTSP of 119.5 %

RTP. This will indicate the amount of conservatism for the NTSP.

NTSP2Offset is defined as the minimum margin (% RTP) between AV and NTSPn with Z of 0.81 used.

NTSP2Offset = Z x ,o ,

NTSP2Offset = (0.81 x 0.75) % RTP = 0.61 % RTP NTSP2 is calculated to provide an NTSP based on the minimum LER avoidance criteria:

NTSP2 I AV - NTSP2 Offset NTSP2 I122.0 % RTP - 0.61 % RTP NTSP2 I 121.4 % RTP For the minimum valve of Z equal to 0.81, NTSP2, which is defined as the LER Avoidance NTSP, is to be less than 121.4 % RTP.

A conservative NTSP of 119.5 % RTP is used.

7.5.1.4 Spurious Trip Avoidance (STA) Test A spurious trip avoidance test is performed to assure that there is a reasonable probability that spurious trips will not occur using the selected setpoint.

lnput 4.1 1, Section 1.2.3.4, provides the following formula for determining Spurious Trip avoidance. lnput 4.13, Sections 4.5.4.b and 4.5.9.b, state that any bias associated with PMA or PEA should also be included. Therefore, APEAb shown in EPU Loop Uncertainty table in Section 7.3.2 is being included.

where n is the number of standard deviations used ( 2 0 )

Terms DPEALand DPEAbwere defined and evaluated in Section 7.3.1.7 for CLTP operation but also apply to EPU operation as discussed in Sections 7.3.2 and 7.3.2.1. These terms are to be included in the above om equation as follows:

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 44 of 58 cnA= (+)&A,)' + (c,,)'+ (DL)' + ( P M ) +~(PEA)' + (DPEA,)' + APEA, + DPEA.

Using APRM terms, om is defined as:

USA = ( ! ) , / ( A ~) + (c, + ( D )~ + ( A P M ) ~+ (APEA) + (DPEA, ) + APE% + DPEA~

,a = 2.20 % RTP lnput 4.1,Section 5.6.8, states Z is equal to the following:

I(~djusted- NTSP) - (Operational - ~ i m i t ) l z= 1

~ S ~ P A lnput 4.1,Section 5.6.8, states that Adjusted NTSP is the selected setpoint minus ALTL (more consewative). Since Adjusted NTSP is (NTSP-ALTL), Z is equal to the following:

z= NTsP - ALT - (Operational -Limit)

USA In lnput 4.3,Sections I .3 and 1.7, GEH defines the Operational Limit for MNGP as 100%

In accordance with lnput 4.1,a minimum value of Z is to be I.65. Since the actual value of 8.25 is greater than the required value of 1.65, adequate separation exists between the NTSP and the Operational Limit (OL), and the STA criterion is satisfied.

I MONTICELLO NUCLEAR GENERATING PLANT I CA-08-050 L

TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU I Page 45 of 58 1 7.5.2 EPU APRM Neutron Flux - High (Setdown) Scram and APRM Neutron Flux -

Hiqh (Setdown) Rod Block lnput 4.3, Section 3, specifies the following Analytical Limit (AL) and the following recommended Allowable Value (AV) and the Nominal Trip Setpoint (NTSP):

PRNM EPU Setpoints

(% RTP) AL AV NTSP APRM Neutron Flux - High (Setdown) Scram NIA 20.0 18.0 APRM Neutron Flux - High (Setdown) Rod Block N/A 15.0 13.0 lnput 4.3 indicates that Setdown Scram and Setdown Rod Block functions do not have AL. These functions only have AV and NTSP. This calculation will determine the minimum required margin for AL to AV, and AL to NTSP. The difference between the minimum required margins for AL to AV, and AL to NTSP becomes the minimum required margin for AV to NTSP.

7.5.2.1 Minimum Required Margin (AL to AW Evaluation lnput 4.1 1, Section 1.2.3.2, provides the following formula for calculating the AV from the AL.

AV 5 AL -(?) J(A,)' + (C,)' + (~A4.4)' +(PEA)' ibias Minimum required margin between AL and AV can be defined by:

(?),/(A,)' +(c,)' + (APMA,)' + (APEA,)' + A P E 4 Minimum Required Margin (AL - AV) = RM = AL - AV 7.5.2.2 Minimum Required M a r ~ i n(AL to NTSP) Evaluation lnput 4.11, Section 1.2.3.3, provides the following formula for calculating the NTSP from the AL.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM S e t ~ o i n t sfor CLTP and EPU Paae 46 of 58 1.645 NT'P B AL - (?--)&AL ) ' + (C, )' + (D, )' + ( P U ~ )+' (PEA)' ibias Minimum required margin between AL and NTSP can be expressed as:

Minimum Reauired Marain (AL - NTSP) = RM = AL - NTSP RM = AL - NTSP =

RM = AL - NTSP = 3.08 O h RTP 7.5.2.3 Minimum Required Marqin (AV to NTSP) Evaluation RM = AV to NTSP =

Minimum Required Margin (AL to NTSP) - Minimum Required Margin (AL to AV)

RM = AV to NTSP = 3.08 % RTP - 2.69 % RTP RM = AV to NTSP = 0.39 % RTP Available Minimum PRNM EPU Setpoint AV NTSP Margin Required AV

(% RTP) (AV- NTSP) Margin Acceptable APRM Neutron Flux -

High (Setdown) 20.0 18.0 2.0 0.39 Yes Scram APRM Neutron Flux

- High (Setdown) 15.0 13.0 2.0 0.39 Yes Rod Block Since the available margin is greater than the minimum required margin (2.0 %

RTP versus 0.39 % RTP), the recommended NTSP is acceptable.

MONTICELLO NUCLEAR GENERATING PLANT ( CA-08-050 TITLE: lnstrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow 1 ~iased PRNM setpoints for CLTP and EPU 1 Page 47 of 58 1 7.5.2.4 Licensee Event R e ~ o r(LER)t Avoidance Test The purpose of the LER Avoidance Test is to assure that there is sufficient margin between the AV and NTSP to reasonably avoid violation of the AV.

lnput 4.11, Section 1.2.3.5, provides the following formula for determining LER avoidance criteria.

where n is the number of the standard deviations used ( 2 0 ) .

0, = (+)~(0.253)' + (1.334)' + (0.625)'

o,, = 0.75 % RTP For multiple instrument channels, a Z value of greater than 0.81 provides sufficient margin between NTSP and AV as specified in ESM-03.02-APP-1, Section 5.6.3 (Input 4.1).

In accordance with lnput 4.1, Section 5.6.3, Z is defined as:

z = IAV - NTSP~

( T ~ ~ ~

For APRM Neutron Flux - High (Setdown) Scram setpoint:

For APRM Neutron Flux - High (Setdown) Rod Block setpoint:

Since Z is greater than 0.81, sufficient margin exists between the specified AV and NTSP.

Minimum margin NTSP2will be calculated using the minimum Z value of 0.81.

Minimum margin NTSP2will be compared with the current NTSP of 18.0 O h RTP. This will indicate the amount of conservatism for the NTSP.

NTSP2 Offset is defined as the minimum required margin between AV and NTSP with Z of 0.81 used.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow 1 ~ i & e dPRNM setpoints for CLTP and EPU I Page 48 of 58 1 NTSP2 Offset = Z x om NTSP20ffset = (0.81 x 0.75) % RTP = 0.61 % RTP APRM Neutron Flux - Hiah (Setdown) Scram Setpoint For APRM Neutron Flux - High (Setdown) Scram setpoint: NTSP2 is calculated to provide a NTSP based on the minimum LER avoidance criteria:

NTSP2 I AV - NTSP2 Offset NTSP2(sfdown,,sc, 5 (20.0 - 0.61) % RTP NTSP2(&fdwnscmm)s I9.4 % RTP For the minimum valve of Z equal to 0.81, NTSP2, which is defined as the LER Avoidance NTSP, is to be less than 19.4 % RTP.

Therefore, a conservative NTSP of 18.0 % RTP is used for the APRM Neutron Flux - High (Setdown) Scram setpoint.

APRM Neutron Flux - Hiqh (Setdown) Rod Block Setpoint For APRM Neutron Flux - High (Setdown) Rod Block setpoint:

NTSP2is calculated below to indicate margin of the recommended NTSP:

NTSP2 I AV - NTSP2 Offset NTSP~(&tdown~ o lock) d I (15.0 - 0.61) % RTP NTSPn(setdownR O B~ I O ~ ~ )I 14.4 % RTP For the minimum valve of Z equal to 0.81, NTSP2,which is defined as the LER Avoidance NTSP, is to be less than 14.4 % RTP.

Therefore, a conservative NTSP of 13.0 % RTP is used for the APRM Neutron Flux - High (Setdown) Rod Block setpoint 7.5.2.5 Spurious Trip Avoidance (STA) Test A Spurious Trip Avoidance test is performed to assure that there is a reasonable probability that spurious trips will not occur using the selected setpoint.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow I ~ i a s e dPRNM setpoints for CLTP and EPU 1 Page 49 of 58 lnput 4.1 1, Section 1.2.3.4, provides the following formula for determining Spurious Trip avoidance. lnput 4.13, Sections 4.5.4.b and 4.5.9.b, provides setpoint guidance to add bias term for APEA.

Where n is the number of standard deviations used ( 2 ~ ) .

Terms DPEALand DPEAb were defined and evaluated in Section 7.3.1.7 for CLTP operation but also applies to EPU operation as discussed in Section 7.3.2 and 7.3.2.1. These terms are to be included in the above ,a equation as follows:

Using APRM terms, ,a is defined as:

am = 2.20 % RTP lnput 4.1, STA Section 5.6.8, states Z is equal to the following:

I(Adjusted -NTSP) - (Operational -~imit)l z=

OSTA lnput 4.1, Section 5.6.8, states that Adjusted NTSP is the selected setpoint minus ALTL (more conservative). Since Adjusted NTSP is (NTSP-ALTL),Z is equal to the following:

z= NTSP - ALT - (Operational -Limit)

USLA lnput 4.3, Section 1. I , states the Operational Limit is 11.O%

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 50 of 58 As specified in lnput 4.1, Section 5.6.8, Z should be equal to or greater that 1.65 for the setpoint to be adequately separated from the Operational Limit to reasonably avoid Spurious trip conditions.

Since the actual value of 2.57 is greater than the required value of 1.65, an adequate separation exists between the NTSP and the Operational Limit (OL),

and the STA criterion is satisfied.

7.5.3 PRNM EPU APRM Downscale Rod Block This section of the calculation provides the basis for the NTSP setpoint to be 3.5 Oh RTP. The existing CLTP and PRNM CLTP NTSP setpoints (Section 7.4.3) are also 3.5 % RTP. EPU lnput 4.3, Section 1.8, indicates the recommended NTSP setpoint is 4.0 % RTP. This section shows there is sufficient margin to keep the NTSP setpoint at 3.5 % RTP. lnput 4.8, Item 12, also indicates the NTSP setpoint as 4.0 % RPT for EPU. This calculation provides the design basis to keep the setpoint at 3.5 % RTP.

lnput 4.3 and lnput 4.8 indicate the APRM Downscale Rod Block does not have an Analytical Limit (AL) setpoint. Both Inputs state the recommended value for the Allowable Valve (AV) is 2.0 % RTP. This section will evaluate the following Setpoints:

PRNM EPU Setpoint

(% RTP) AL AV NTSP APRM Downscale Rod Block NIA 2.0 3.5 This calculation will determine the minimum required margin for AL to AV, and AL to NTSP. The difference between the minimum required margins for AL to AV, and AL to NTSP becomes the minimum required margin for AV to NTSP.

7.5.3.1 Minimum Required Marqin (AL to AV) Evaluation lnput 4.1 1, Section 1.2.3.2, provides the following formula for calculation the AV from the AL.

Minimum required margin between AL and AV can be defined by:

1 MONTICELLO NUCLEAR GENERATING PLANT 1 CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Set~ointsfor CLTP and EPU Paae 51 af 58 Minimum Reauired Marqin (AL - AV) = RM = AL - AV RM = AL - AV = 2.69% RTP 7.5.3.2 Minimum Reauired Marqin (AL to NTSP) Evaluation Input 4.11, Section 1.2.3.3, provides the following formula for calculating the AV from the AL:

Minimum required margin between AL and NTSP can be defined by::

Minimum Required Marqin (AL - NTSP) = RM = AL - NTSP RM = AL - NTSP =

(?),/(0.253)' + (1.334)' + (0.625)' +(2.288)' + (0.268)' +(0.054)' + 0 4 9 + 0.33 RM = AL - NTSP = 3.08 % RTP 7.5.3.3 Minimum Required Marqin (AV to NTSP) Evaluation = RM = AV to NTSP RM = AV to NTSP =

Minimum Required Margin (AL to NTSP) -Minimum Required Margin (AL to AV)

RM = AV to NTSP = 3.08 % RTP - 2.69 % RTP RM = AV to NTSP = 0.39 % RTP Available Minimum EPU Setpoint AV NTSP Margin Required AV

(% RTP) (AV-NTSP) Margin Acceptable APRM Downscale Rod Block 2.0 3.5 1.5 0.39 Yes

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Set~ointsfor CLTP and EPU Paae 52 of 58 Since the available margin is greater than the minimum required margin (1.5 %

RTP versus 0.39 % RTP), the recommended NTSP is acceptable.

7.5.3.4 Licensee Event Report (LER) Avoidance Test The purpose of the LER Avoidance Test is to assure that there is sufficient margin between the AV and NTSP to reasonably avoid violation of the AV.

lnput 4.1 1, Section 1.2.3.5, provides the following formula for determining LER avoidance criteria.

= ( i ) h ~ ~ +) (c,)'

~ L E R ' +(DL)'

where n is the number of the standard deviations used ( 2 ~ )

D,, = ( : ) ~ ( 0 . 2 5 3 )+~(I .334)' + (0.625)2 om = 0.75 % RTP For multiple instrument channels, a Z value of greater than 0.81 provides sufficient margin between the NTSP and AV as specified in ESM-03.02-APP-I, Section 5.6.3 (Input 4.1).

In accordance with lnput 4.1, Section 5.6.3, Z is defined as:

Since Z is greater than 0.81, sufficient margin exists between the specified AV and NTSP.

Minimum margin NTSP2will be calculated using the minimum Z value of 0.81.

Minimum margin NTSP2will be compared with the current NTSP of 3.5 % RTP.

This will indicate the amount of conservatism for the NTSP.

NTSPz Offset is defined as the minimum margin (% RTP) between AV and NTSP:! with Z of 0.81.

NTSP2Offset = Z x oLER NTSP20ffset = (0.81 x 0.75) % RTP = 0.61 % RTP

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow

1. I ~iised PRNM ~e&ointsfor CLTP and EPU 1 Paae 53 of 58 1 NTSP2is calculated to provide an NTSP based on the minimum LER avoidance criteria:

NTSP2 2 AV + NTSP2 Offset NTSP2 2 2.0 % RTP + 0.61 % RTP NTSP2 2 2.61 % RTP For the minimum valve of Z equal to 0.81, NTSP*, which is defined as the LER Avoidance NTSP, is to be greater than 2.61 % RTP.

Therefore, a conservative APRM Downscale Rod Block NTSP setpoint of 3.5

% RTP is to be used.

1 MONTICELLO NUCLEAR GENERATING PLANT 1 CA-08-050 TITLE: I Instrument Setpoint Calculation - I Revision 0 Average Power Range 'Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 54 of 58

8. CONCLUSIONS The Analytical Limits (AL), Allowable Values (AV), and Nominal Trip Setpoints (NTSP), for APRM Neutron Flux High Scram, Setdown Scram, Setdown Rod Block and APRM Downscale Rod Block are listed below for EC-10856, which includes NUMAC PRNMS setpoints at CLTP and EPU conditions.

8.1 Loop Uncertainties for PRNM CLTP and EPU Operation Term Value Section .

4- Loop Instrument Accuracy + 0.253 % RTP 7.3.1 .I DL - LOOPInstrument Drifl + 0.625 % RTP 7.3.1.2 ALTL - Loop As-Left Tolerance for uncertainty calculations f 1.334 % RTP 7.3.1.3 ALTL - Loop As-Left Tolerance for PRNM electronic calibration 0.00 O h RTP 7.3.1.3 AFTL - Loop As-Found Tolerance for uncertainty calculations f 1.334 % RTP 7.3.1.4 AFTL - Loop As-Found Tolerance for PRNM electronic calibration 0.00 % RTP 7.3.1.4 CL- Loop Calibration Accuracy Error + 1.334 O h RTP 7.3.1.5 APEAL - Loop APRM Primary Element Accuracy + 0.268 % RTP 7.3.1.6 DPEAL- Loop Drift Primary Element Accuracy f0.054% RTP 7.3.1.7 APMAL - Loop APRM Process Measurement Accuracy + 2.288 % RTP 7.3.1.8 APEAb- APRM Primary Element Accuracy bias +0.49 % RTP bias 7.3.1.6 DPEAb- Drift Primary Element Accuracy bias + 0.33 % RTP bias 7.3.1.7

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow I I ~ i a s e dPRNM ~e&ointsfor CLTP and EPU I Paae 55 of 58 1 8.2 PRNM - CLTP Operation Setpoint (EC-10856) 8.3 PRNM - EPU Operation Setpoint (EC-108561

9. FUTURE NEEDS This calculation impacts the following documents, which are listed in EC-10856:

9.1 Calculation CA-05-153, Rev 0, Instrument Setpoint Calculation - APRM Downscale CR Block, calculated the APRM Downscale CR Block NTSP setpoint of 3.5 % RTP for ITS and CLTP operation. PRNM CLTP operation does not change this setpoint. For PRNM EPU operation, NTSP setpoint of 3.5 % RTP has also been evaluated in accordance with Section 7.5.3. CA-05-153 will be superseded when PRNM retrofit is installed because the PRNM uncertainties are used as the basis for the APRM Downscale Rod Block setpoint. GAR 01 146760 was initiated to track calculation CA-05-0153 be superseded due to EC 10856.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 56 of 58 9.2 Calculation CA-96-224, Rev 1, Instrument Setpoint calculation - APRM Flow-Biased Upscale Scram and Rod Block, includes the APRM Flow-Referenced Neutron Flux - High High setpoint. PRNM changes this setpoint to non-flow bias APRM Neutron Flux High. The current setpoint of 0.66W + 67.6 % RTP clamped at 122 % RTP is being changed to 122 % RTP. This applies to PRNM CLTP and EPU operation. GAR 01146761 was initiated to track calculation CA-96-224 be superseded due to EC 10856.

9.3 Procedure C.6-005-A-03, Rev 1, Rod Withdraw Block. This is the annunciator procedure for window 5-A-3. PRNMS adds a new rod withdraw block setpoint:

APRM Neutron Flux - High (Setdown) Rod Block. Sections 7.4.2 and 7.5.2 evaluate the setpoint. This annunciator procedure will be changed to add the new rod withdraw block setpoint. PCR 01146750 has been initiated to track changes to C.6-005-A-03 due to EC 10856 PRNM retrofit.

9.4 Procedure C.6-005-A-06, Rev 3, APRM Downscale, states a NTSP setpoint of 3.5

% RTP. This is correct for the present neutron monitoring system. Even though the PRNM CLTP and EPU operation NTSP setpoints are 3.5 % RTP, the procedure does not address that the PRNM retrofit NTSP setpoints remain the same for APRM Downscale Rod Block. PCR 01 146778 was initiated to revise C.6-005-A-03, Rev 1, when EC 10856 PRNM retrofit is installed.

9.5 Procedure C.6-005-A-22, Rev 3, APRM Hi Hi INOP CH 1, 2, 3, is a flow-bias APRM Neutron Flux Hi Hi setpoint. PRNM retrofit converts this setpoint to a non-flow bias APRM Neutron Flux High setpoint. The PRNM APRM Neutron Flux High setpoint is part of this calculation. C.6-005-A-22 will be revised under EC-10856 and PCR 01 129100.

9.6 Procedure C.6-005-A-30, Rev 3, APRM Hi Hi INOP CH 4, 5, 6, is a flow-bias APRM Neutron Flux Hi Hi setpoint. PRNM retrofit converts this setpoint to a non-flow bias APRM Neutron Flux High setpoint. The PRNM APRM Neutron Flux High setpoint is part of this calculation. C.6-005-A-30 will be revised under EC-10856 and PCR 01 133816.

9.7 Procedure 8.05.06-02, Rev 18, Operations Manual Section - Plant Protection System, specifies APRM Hi Hi and APRM Downscale and other setpoints. This calculation evaluates the APRM Downscale Rod Block setpoints and documents the PRNM EPU change in this setpoint. The APRM Hi Hi setpoint is flow biased and is PRNM changes this setpoint to non-flow bias APRM Neutron Flux High.

6.05.06-02 will be revised under EC-10856 and PCR 01133455.

1 MONTICELLO NUCLEAR GENERATING PLANT 1 CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU 1 Page 57 of 58 1 9.8 Procedure 6.05.01.02-02, Rev 6, Operations Manual Section - Power Range Neutron Monitoring, specifies NMS trip setpoints, which are being changed due to PRNMS. 6.05.01.02-02 will be revised under EC-10856 and PCR 01 137808.

9.9 8.05.01.02-05, Rev 16, Operations Manual Section - Power Range Neutron Monitoring, System Operation. B.05.01.02-05, Rev 16 refers to the six APRM channels, which applies to the existing NMS. PRNMS has four APRM channels as stated is Section 7.2.2.1 of this calculation. PCR 01146778 issued to revise B.05.01.02-05, Rev 16, upon implementation of EC 10856.

9.10 DBD B5.1, Rev C, Design Bases Document for Neutron Monitoring System, discusses NMS setpoints, margin, uncertainty parameters such as drift, etc. This calculation validated certain NMS setpoints using the PRNM parameter uncertainties specified in GE documentation. DBD B5.1 will be revised under EC 10856. GAR 1138038 tracks revision to DBD B5.1 for EC 10856 PRNM setpoint changes.

9.1 1 MNGP Technical Specification, Amendment 155, Table 3.3.1 .l-1, for APRM Flow Referenced Neutron Flux High High is replaced by PRNM APRM Neutron Flux High. New setpoint PRNM Neutron Flux-High (Setdown) is added. APRM Downscale Rod Block is being removed from Tech Specs when PRNM retrofit is installed. GAR 01146762 was initiated to track changes to the Technical Specifications due to EC 10856.

9.12 MNGP Technical Specifications Bases, Rev 8, Bases will be revised to discuss the PRNM APRM Neutron Flux High setpoint, which is non-flow bias, in place of the existing Flow Referenced Neutron Flux-High High setpoint. PRNM Neutron Flux-High (Setdown) setpoint to being added. GAR 01146763 has been initiated to track changes to the Technical Specification Bases due to EC 10856.

9.13 MNGP Technical Requirements Manual (TRM), Rev 2, New PRNM Setdown Rod Block setpoint is to be discussed in the TRM. The APRM Downscale Rod Block setpoint is being removed from the Tech Specs and will be added to the TRM.

APRM Downscale Rod Block setpoint is the same for CLTP and EPU. This calculation provides the design basis for the EPU APRM Downscale Rod Block NTSP setpoint because the GEH Input documents have a slightly different value.

LAR 01 128839 has been initiated to track PRNM setpoint changes to the TRM due to EC 10856.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 TITLE: Instrument Setpoint Calculation - Revision 0 Average Power Range Monitor (APRM) Non-Flow Biased PRNM Setpoints for CLTP and EPU Page 58 of 58 9.14 Procedure 8211, Rev 2, APRM Calibration Readjustment for Single Loop, discusses APRM setpoint voltage adjustments including Downscale Rod Block, Hi-Hi Scram, etc. Changes have been made by PRNM and this calculation evaluates the non-flow biased PRNM setpoints. Procedure 8211 will be deleted under EC-10856, PCR 01133437. SLO operation will be enabled under Operations procedure in B.05.01.02-05. PCR 1133449 has been initiated to track these procedure changes.

9.15 Procedure 8212, Rev 2, APRM Calibration Readjusfmenf for Two Loop, discusses APRM setpoint voltage adjustments including Downscale Rod Block, Hi-Hi Scram, etc. Changes have been made by PRNM and this calculation evaluates the non-flow biased PRNM setpoints. Procedure 8212 will be deleted under EC-10856, PCR 01133445. TLO operation will be enabled under Operations procedure in B.05.01.02-05. PCR 1133449 has been initiated to track these procedure changes.

9.16 Procedure 0012, APRM/Flow Reference Scram Functional Check, performs the calibration of the APRM including the Neutron Flux High Scram, Setdown Scram, Setdown Rod Block, and Downscale Rod Block setpoints. Setpoints are revised as a result of this calculation. 0012 will be deleted under EC-10856, PCR 01133332. Procedures ISP-NIP-0588, ISP-NIP-0589-01, ISP-NIP-0589-02will be developed to replace 0012. APRM Calibration will be created under EC-10856.

PCR 01129124 has been initiated to track these procedure changes.

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 Attachment I Revision o Setpoint Diagrams APRM Neutron Flux - High Scram - CLTP and EPU Operation

% RTP 120.8 As FoundIAs Left See Note 1 As Found/

As Left 119.5 NTSP ALT = AFT rt 1.34%

118.2 As FoundIAs Left APRM Neutron Flux (Setdown) Scram for CLTP and EPU Operation

% RTP 19.3 As FoundIAs Left See Note 1 As As Left 18.0 NTSP ALT = AFT +_ 1.34%

Tolerance 16.7 As FoundIAs Left

MONTICELLO NUCLEAR GENERATING PLANT CA-08-050 Attachment 1 Revision o Setpoint Diagrams I Page *of2 1 APRM Neutron Flux (Setdown) Rod Block - CLTP and EPU Operation

% RTP nla AL 15.0 AV 14.3 As FoundIAs Left See Note 1 As As Left 13.0 NTSP ALT=AFT _+ 1.34%

Tolerance 11.7 As FoundIAs Left APRM Downscale Rod Block - CLTP and EPU Operation

% RTP 4.8 As FoundIAs Left See Note 1 As As Left 3.5 NTSP ALT =AFT k 1.34 %

Tolerance 2.2 As FoundIAs Left 2.61 NTSP;!

LER Avoidance 2.0 AV Note 1: The As-Left and As-Found uncertainty tolerances are specified as 1.34 %

RTP because GEH Inputs 4.2 and 4.3 state tolerance of 2 % RTP, 3 0 .

Converting AFTIALT to 2cr confidence level results in 1.34 % RTP. However, AFTIALT tolerances are 0.00 % R I P when used for PRNM surveillance calibration. PRNMS is a digital system and the setpoint is a single number in the database not susceptible to drift. Sections 7.3.1.3 and 7.3.1.4 evaluate the uncertainty and calibration tolerances for AFT and ALT.

ATTACHMENTS 2 AND 3 TO MNGP CALCULATION CA-08-050 ARE NOT INCLUDED THEY ARE GEH PROPRIETARY CALCULATIONS (Included for ease of reference in the calculation.)

Attachment 4 Page Iof 3 CA-08-050, Rev 0 Mathematics of Phvsics and Modern Engineering I. S. SOKOLNIKOFF Professor of Mathematics University of California, Los Angela R. M. REDHEFFER Prof'cssor of Mathematics University of California, Los Angela Second Edition McGRAR-HILL ROOK COMPANY NEW YDRR .ST. LOUIS $AN FRANCISCO TORONTO LONDON SYDXEY

CA-08-050, Rev 0 Attachment 4 Page 2 of 3

3. Three coins are tossed. Let t be the number of heads show11by the frst coin, whereas y is the number of heads shown by all the coins. Compute the correlation coefficient. Your result s b u l d be smaller than the value (13-10). \Vhy?

4. Two continuous variables x and y are said to be indepndcnl if thcir joint density /(x,y) has the form j(x)g(y). Verify the theorem of compound probability for i n d e p n h ~ devcrtls. ill the form Pr(a<x<b,cy<<=h(a<z<b)Pr(c<y<d).

,4190 verily the equation E(zy) = E(t.)E(y).assuming convergeuce of the relevant inlegwls.

5. Referring to (12-9) and (12-lo), express the variance, covariance, and correlatio~tcoefficient in integral form. when x and y are continuous variables with joint c t e r s i ~ y~ ( x Y ) . By using the w u l t of Prob. 4. show that the covariancs is 0 if the variables are independenl.

6. (Chebycheo'u lemma) Let 1 be a random variable which does not assume negative valuw, and let E(t) = T. Rove that the prohhility of the inequality t < rh in at letlst 1 Iih, for every positive constant h. Oulline of mlulion: The probability of the contwy event, t > rh, irr 7 . (Chebydrea's inequality.) Let z be a random variable with' mean fi and variance L+. By Prob. 6 with 1 = (x p)? prove that the probability of the inequality Ir - 5 4 r is a t least 1 - 1/h, for every positive constant h.

8. Plot the probability of the inequality lz - 5 ka as a function of k when r is normally distributed with mean p and variance u. On the same axes, plot the minimum probability of thisinequality as given by Prob. 7. (Specific dietributions such as the Gauss. Poisson. m Maxwell-Boltnnann distributions depend on a few parameters. A melhod of estimation which does not m u m e a specific type of distribution function is called a nonpanudric method. Sicc non-parametric methods w u m e less about the random process. they give less information.)

14. Arithmetic Means. The Theory of Emors. In many applications one does not consider a single value alone, hut one obtains a mean of a large n u r n h r of values.

For exarnpIe, if rn denotes the measured value of the length of a rod one would rnake several measurements m,, rnt, .. . , m, and use the arithmetic mean iii If the true vahle of the length is 11, the errors in measurement are By adding and dividing hy n we get ij.=fi-v where 3 is the mean of the ti:

z = 2, + s +n ... +zn.

According to (14-I), the mean of (he errors is equal to h e error nf lhe rrt~rur. It irr likely to be s~nallerthan the error it1 a single measurement, b u s e positive and negative errors krld to cancel in Zs..

So far we have taken the view t.11at the s.'s are indeprrldent o k ~ r v a t i o t ~ ofaa sit~glc ral~dornvariable x, so that 2 is thc ohsertud mearc or the suntple trteori. flowever, we

CA-08-050, Rev 0 Attachment 4 Page 3 of 3 SEC. 14 ARITHYETI(: MEAXS. THE THEORY OP ERRORS 6.13 can also take the view that thin r i are il~dept.ndelttvariahlm' and that 3 is another randorn variahle whose value is the iiiean of Ihc values of ri. Witlr t.liis view it is pus-sible to get a precise descriptio~~ of the improvement in accuracy as the numl)er of measurenlentr i~lcrcases. Thc reailt is:

THEOREM. IJ Ihr turiablrs J Swe ir~dPprndenl,iJ l h ~ hcilr y the same e~.peclalionE ( r 0 = p ar~dthe $ante rvlrianc~a?, then Here 0 , denotes the standard deviation of the variable z, and oi denotes the standard deviation of 5. To prove the thwrern, observe that E(xl + ... + E ( x ~+) + E(m) 2,) = = np.

The variance of rk . . . + + r, is thcrefoce E(% + . . . + rr. - np)'

which can be written Expanding the term in brackets we obtain Since tlre varial~lesare independent, the covaria~lceof x, and xi is zero for i # j; that is, E ( x- ~ p ) ( % - p) = 0.

Also the definition of a, gives uz2 = E(x, - p)*.

Hence, taking the expectation in (14-3) yields E(a + a . 4 + xn - n ~= )nus2.~

Dividing by n* we have which gives the cmnclusion upon taking the square root.

The intuitive meaning of this result is approximately as follows: Suppose a single mcasurcment varies over an interval of length 1 about the true value, so that 1 measures t h scattcr or spread. Then the mean of n independ~ntmeasurernents will have a spread d the order of / , / A ahout the true value. This shows that the improve_merit of accuracy due to cancellation of positive and negative errors is of the order of d n , where n is the 1111rnberof ~neasurernents. Tile most important consideration justifying the analysis in practice is that systematic errors lnust be eliminated.

The foregoir~gc~)nclu~sions are independent of the density function j ( x ) that govcrns the statistical distribution of the errtjrs. However, in a p e a t variety of cases the errors have a Gaussian density 7'11is result. knt,wn as the I:anssinn la," of error, states that the variable 6 hr is r~orrnallydistributed. Specifically, tile prohahility of Tlrr: t ~ s eof lower-cuse lettern to deuote vttriabler, is custon~aryiu statistical litartiture. However, tire r, here ure aoalogotu to the .Yi of See. 12. not to the I . of Sew. & w d 5.

ENCLOSURE 7 MONTICELLO NUCLEAR GENERATING PLANT RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION FOR LICENSE AMENDMENT REQUEST FOR POWER RANGE NEUTRON MONITORING SYSTEM UPGRADE (TAC NO. MD8064)

GEH PROPRIETARY INFORMATION AFFIDAVITS 2 AFFIDAVITS ENCLOSED DATED JUNE 13,2008 SEPTEMBER 10,2008 6 Pages Follow

GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, Robert E. Brown, state as follows:

(1) I am Senior Vice President, Regulatory Affairs, Regulatory Affairs, GE-Hitachi Nuclear Energy Americas LLC ("GEH"), have been delegated the hnction of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in GEH letter, GE-MNGP-AEP-766, Monticello PRNM - GEH Responses to RAIs 1 and 6 - June 13, 2008, dated August 13, 2008. The proprietary information in Enclosure 1 entitled, GEH Responses to RAIs 1 and 6

- June 13, 2008, is identified by a dotted underline inside double square brackets.. ((This s~,Qt.e.~G.G.~.~.aneexxamP~.eee~.~~,)) In each case, the superscript notation 13) refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Proiect v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

Affidavit for MNGP-AEP-766 Affidavit Page 1 of 3

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. above.

(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GEH, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GEH. Access to such documents within GEH is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed results of GEH's analysis and evaluations of various stability events and the performance of the stability detection and suppression capability of the APRM-based detection algorithm for the BWR. The analysis is based on analytical models, methods and processes, including computer codes, which GE has developed, and applied to perform stability evaluations for the BWR. The development of the detection and suppression capability of the APRM-based detection algorithm for the BWR was achieved at a significant cost to GE.

The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GE asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and Affidavit for MNGP-AEP-766 Affidavit Page 2 of 3

analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 13th day of August 2008.

Robert E. Brown GE-Hitachi Nuclear Energy Americas LLC Affidavit for MNGP-AEP-766 Affidavit Page 3 of 3

GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, Tim E. Abney, state as follows:

(1) I am Vice President, Services Licensing, Regulatory Affairs, GE-Hitachi Nuclear Energy Americas LLC ("GEH"), have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in GEH letter, GE-MNGP-AEP-805, Monticello PRNM - GEH Response to RAI 2 June 13, 2008, dated September 8, 2008. The proprietary information in Enclosure 1 entitled, GEH Response to RAI 2 - June 13, 2008, is identified by a dotted underline inside double square brackets. ((Th~~..~.~n.t.~n.~.~..~.~..an nna-nn!s,~.~!)) In each case, the superscript notation {" refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Proiect v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

Affidavit for MNGP-AEP-805 Affidavit Page 1 of 3

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. above.

(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GEH, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GEH. Access to such documents within GEH is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2) above is classified as proprietary because it contains results of a GEH evaluation of instrumentation used at Boiling Water Reactors (BWR) to protect fuel integrity, which was developed by GEH for evaluations of Monticello's Power Range Neutron Monitor license application. The basis of the evaluation includes GEH's analysis of stability events at BWRs. Development of these analyses, techniques, and information and their application to the design, modification, and analyses methodologies and processes for the Monticello's Power Range Neutron Monitor license application was achieved at a significant cost to GEH.

The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GEH asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and Affidavit for MNGP-AEP-805 Affidavit Page 2 of 3

analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 8th day of September 2008.

Tim E. Abney Vice President, Services Licensing GE-Hitachi Nuclear Energy Americas LLC Affidavit for MNGP-AEP-805 Affidavit Page 3 of 3