ML12310A407

24-Month 10CFR50.59 and 10CFR72.48 Evaluation Summary Report for the Period July 1, 2010 Through June 30, 2012

ML12310A407
Person / Time
Site:	Limerick
Issue date:	11/05/2012
From:	Dougherty T Exelon Generation Co
To:	Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation, Document Control Desk
References
Download: ML12310A407 (28)

Text

10CFR50.59(d)(2) 10CFR72.48(d)(2)

November 5, 2012 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Limerick Generating Station, Units 1 and 2 Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352, 50-353 and 07200065

Subject:

24-Month 10CFR50.59 and 10CFR72.48 Evaluation Summary Report for the Period July 1, 2010 through June 30, 2012 Attached is the 24-Month 10CFR50.59 and 10CFR72.48 Evaluation Summary Report for Limerick Units 1 and 2 for the period of July 1, 2010 through June 30, 2012, forwarded pursuant to 10CFR50.59(d)(2) and 10CFR72.48(d)(2). The report includes brief descriptions of any changes, tests and experiments, including a summary of the evaluation of each. Seven plant changes were implemented using 10CFR50.59 Evaluations during this 24-month period. There were no plant changes implemented using 10CFR72.48 Evaluations during this 24-month period. The summaries of these changes are included in this report.

There are no commitments contained in this letter.

If you have any questions or require additional information, please do not hesitate to contact us.

Sincerely, Original signed by Thomas J. Dougherty Vice President - Limerick Generating Station Exelon Generation Company, LLC

Attachment:

Limerick Generating Station 24-Month 10CFR50.59 and 10CFR72.48 Evaluation Summary Report, July 1, 2010 through June 30, 2012 cc: Administrator Region I, US NRC USNRC Senior Resident Inspector, LGS

10 CFR 50.59 Evaluation and 10 CFR 72.48 Evaluation 24-Month Summary Report Limerick Generating Station 2012 Note: This report summarizes 10 CFR 50.59 and 10CFR72.48 Evaluations that were approved between July 1, 2010 and June 30, 2012.

Evaluation number: LG2010E003 Rev.0 50.59 Reviewer approval date: 12/10/10 PORC number: 10-026 PORC approval date: 12/16/10 Implementing document: ECR 10-00338 Rev.0 Evaluator: Mark Gift Reviewer: Ken Collier Unit 1 Unit 2 Units 1 & 2 Common Unit applicability: [ ] [ x] [ ] [ ]

Complete on: [ ] [ x] [ ] [ ]

Title:

Modify Select MOV Circuits to prevent Spurious Operations During Postulated Hot Shot Fire Scenarios Description of Activity:

The proposed activity ensures that the valves of the Residual Heat Removal (RHR),

Nuclear Boiler, High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems will remain in the desired position for safe shutdown during a postulated fire. This ECR involves modifying the circuitry of these systems specified MOVs to prevent them from re-positioning to an undesired state, initiated by hot shorts.

The activity involves: rewiring and/or the replacing of the Main Control Room (MCR) hand switches, the installation of aux. control relays within the Motor Control Centers (MCCs) and the rewiring of the MCCs.

Reason for Activity:

As part of Exelons ongoing commitment to comply with the Nuclear Regulatory Commission (NRC) requirements for post-fire safe shutdown , a select few Motor Operated Valves (MOV) have been identified by the Exelon Expert Panel that are required for the safe shutdown of Limerick Generating Station (LGS) Unit 2. A postulated fire can initiate multiple hot shorts that could result in spurious repositioning of the specific MOVs when not desired. These changes are required to address issues related to Multiple Spurious Operations (MSOs) as outlined in Nuclear Energy Institute (NEI) 00-01, Rev. 2 (Guidance for Post-Fire Safe Shutdown Circuit Analysis), update Exelons position addressing NRC Information Notice 92-18 (Potential For Loss Of Remote Shutdown Capability During A Control Room Fire) and ensure compliance with NRC Regulatory Guide 1.189, Rev 2 (Fire Protection for Nuclear Power Plants).

Effect of Activity:

The proposed activity does not change the functionality of the MOVs nor their associated components; however, the proposed activity ensures the MOVs will remain operable to safely shutdown the plant during a postulated fire. The proposed activity prevents spurious operation of the MOVs and prevents damage to the limit and torque switches during a postulated fire. The proposed activity does not impact the plant operations, design bases, or any safety analyses described in the UFSAR.

Summary of Conclusion for the Activitys 50.59 Review:

This activity modifies the circuitry of the specified MOVs to prevent them from re-positioning to an undesired state during a postulated fire scenario initiating hot shorts.

The proposed activity does not change the intended functionality of the MOVs nor their associated components. This activity will enhance compliance with the MSO requirements. All impacts of the new configuration were examined and the screening determined there may be an adverse affect on a UFSAR described design function(s). An evaluation was required due to the addition of active components in a safety related system that may increase the likelihood of occurrence of a malfunction important to safety. Since the affected systems are single active failure proof, no new failure modes are introduced, the additional components are reliable, and the change in failure rate of the affected SSCs is negligible, the 50.59 evaluation concluded there was less than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety. This activity can be implemented without obtaining prior NRC approval and a license amendment is not required.

Evaluation number: LG2011E001 Rev.0 50.59 Reviewer approval date: 3/30/11 PORC number: 11-008 PORC approval date: 4/5/11 Implementing document: ECR LG 10-00103 Rev.0 Evaluator: Robert Potter Reviewer: Jeff Holley Interface/Cross Discipline Reviewer Andy Olson Unit 1 Unit 2 Units 1 & 2 Common Unit applicability: [ ] [ ] [ x] [ ]

Complete on: [ ] [ ] [ x] [ ]

Title:

Application of TRACG04P Version 4.2.60.3 for OPRM Setpoint Determination Description of Activity:

This activity addresses the use of the General Electric Hitachi (GEH) advanced, multi-purpose NSSS thermal-hydraulic transient code TRACG04P, Version 4.2.60.3, for the purpose of determining the Oscillation Power Range Monitor (OPRM) setpoints for Limerick Generating Station. OPRM setpoints are determined for each operating cycle as part of the standard reload licensing process performed in accordance with General Electric's Standard Application for Reactor Fuel (GESTAR II) methodology. The cycle specific OPRM setpoints are presented in the Core Operating Limits Report (COLR).

Version 4.2.60.3 of TRACG04P is an upgraded version of the NRC approved TRACG02A program originally developed and licensed to determine OPRM setpoints.

Version 4.2.60.3 of TRACG04P has not been generically approved by the NRC for OPRM setpoint determination. OPRM system trip functions are described in the UFSAR and the evaluation of OPRM PBDA setpoints is performed as part of the Limerick Generating Station cycle specific safety analysis process. NEDO-32465-A is cited in Technical Specification 3.3.1.1 by reference. Therefore, use of TRACG04P Version 4.2.60.3 constitutes a change in methodology requiring evaluation in accordance with 10 CFR 50.59.

The TRACG02A version of the TRACG thermal-hydraulic code was approved by the NRC and used in the preparation of NEDO-32465-A (Reference 5) during the original design and licensing of the GE OPRM system. In 2006 the TRACG code was upgraded to TRACG04 to support coupling with an improved kinetics model resulting from GE's transition to the PANAC11 version of the 3-dimensional core simulator program PANACEA (References 4 and 6). In 2009 GE implemented a PC-based version of the TRACG04 program, TRACG04P, Version 4.2.57.11 (Reference 3). These earlier software upgrades were evaluated under 10 CFR 50.59, as documented in 50.59 evaluations LG2009E001 and LG2007E002, respectively. This 50.59 Evaluation has been prepared to support upgrading TRACG04P to Version 4.2.60.3. Version 4.2.60.3

implements fixes to several programming deficiencies, as discussed in References 1 and

2. This 50.59 evaluation necessarily addresses all software changes implemented subsequent to the version of the program reviewed and approved by the NRC, TRACG02A.

Due to similarities between the Limerick Generating Station Unit 1 and Unit 2 design/licensing bases; this change is applicable to both Limerick Generating Station units.

Reason for Activity:

The TRACG04P code has recently been revised by the vendor, GE-Hitachi (GEH), to address a number of programming issues identified since its initial release. Descriptions of all changes implemented by TRACG04P version 4.2.60.3 subsequent to NRC approval of TRACG02A are summarized in References 1 through 4. The use of TRACG04P version 4.2.60.3 for Limerick Generating Station DIVOM analysis when determining OPRM stability setpoints constitutes a change in methodology. The upgraded version of the code was developed under the GEH NRC-approved Quality Assurance Program.

However, since TRACG04P Version 4.2.60.3 has not been reviewed and approved by NRC for DIVOM analysis, and GEH is not a license holder, the change needs to be evaluated under 10CFR 50.59 by Exelon.

Effect of Activity:

The method of determining OPRM setpoints is described in the Limerick Generating Station Technical Specification BASES 2.2.1.2, which states; "There are four "sets" of OPRM related setpoints or adjustment parameters: a)

OPRM trip auto-enable setpoints for APRM Simulated Thermal Power (30%) and recirculation drive flow (60%); b) period based detection algorithm (PBDA) confirmation count and amplitude setpoints; c) period based detection algorithm tuning parameters; and d) growth rate algorithm (GRA) and amplitude based algorithm (ABA) setpoints.

The first set, the OPRM auto-enable region setpoints, are treated as nominal setpoints with no additional margins added as discussed in Reference 5. The settings, 30% APRM Simulated Thermal Power and 60% recirculation drive flow, are defined (limit values) in a note to Table 2.2.1-1. The second set, the OPRM PBDA trip setpoints, are established in accordance with methodologies defined in Reference 4, and are documented in the COLR. There are no allowable values for these setpoints. The third set, the OPRM PBDA tuning parameters, are established or adjusted in accordance with and controlled by station procedures.

The fourth set, the GRA and ABA setpoints, in accordance with References 2 and 3, are established as nominal values only, and controlled by station procedures. "

The TRACG thermal-hydraulic code supports the determination of the second set of setpoints, period based detection algorithm (PBDA) trip setpoints. Reference 5 as specified above is BWROG Letter 96113, K. P. Donovan (BWROG) to L.E. Phillips (NRC), "Guidelines for Stability Option III 'Enable Region' (TAC M92882)," September 17, 1996. Reference 2, 3 and 4, as specified in the Technical Specification Bases 2.2.2, above are; NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995. and NEDO-31960-A Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995 and NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for reload Applications," August 1996 (Reference 5), respectively.

The TRACG thermal-hydraulic code is used to develop a conservative relationship between the change in fuel bundle critical power ratio (CPR) and the hot bundle oscillation magnitude. This conservative relationship is used to determine the Delta CPR Over Initial MCPR Verses Oscillation Magnitude (DIVOM) curve. The DIVOM curve, in conjunction with the initial maximum critical power ratio (IMCPR) and the hot bundle oscillation magnitude, is used by Global Nuclear Fuels (GNF) to determine the OPRM PBDA setpoints.

The algorithms used to detect thermal-hydraulic instability related neutron flux oscillations, described in Technical Specification BASES 2.2.1.2, are not impacted by this activity. TRACG04P is only used in the setpoint determination.

The slope of the DIVOM curve represents the thermal-hydraulic responsiveness of the fuel to a given oscillation magnitude. Thus, a steeper slope is more conservative than a flatter slope (NEDO-32465-A). Benchmarking of the NRC-approved TRACG02A code and the TRACG04P Version 4.2.60.3 code has determined that the DIVOM slope developed using TRACG04P generates a slightly more conservative (steeper) DIVOM slope. Therefore, TRACG04P Version 4.2.60.3 can be applied for Limerick Generating Station DIVOM analysis when determining OPRM stability setpoints without prior NRC approval.

Summary of Conclusion for the Activitys 50.59 Review:

The use of TRACG04P version 4.2.60.3 for Limerick Generating Station DIVOM analysis when determining OPRM stability setpoints constitutes a change in methodology that will be addressed by this 50.59 Review.

As discussed in the attached 50.59 Evaluation, GEH benchmarking analyses confirm that TRACG04P produces DIVOM curves that are slightly conservative (more limiting) than those produced by TRACG02A. Therefore, the use of TRACG04P Version 4.2.60.3 does not constitute a departure from a method of evaluation described in the UFSAR and TRACG04P can be used to support the determination of cycle specific OPRM setpoints without prior NRC approval.

The version of TRACG is below the level of detail discussed in the UFSAR and Technical Specifications, therefore a change to the UFSAR and Technical Specification BASES is not necessary. Due to the similarities between the Limerick Generating Station Unit 1 and Unit 2 design/licensing bases; this change is applicable to both units.

This 50.59 Review is being processed as part of the Limerick Generating Station Unit 2 Cycle 12 Reload Fuel Change Package, ECR LG 10-00103.

Evaluation number: LG2011E002 Rev.0 50.59 Reviewer approval date: 8/22/11 PORC number: 11-027 PORC approval date: 9/15/11 Implementing document: ECR 11-00098 Rev.0 Evaluator: Mark Gift Reviewer: Greg Curtin Unit 1 Unit 2 Units 1 & 2 Common Unit applicability: [ x] [ ] [ ] [ ]

Complete on: [ x] [ ] [ ] [ ]

Title:

Modify Select MOV circuits to Prevent Spurious Operations During Postulated Hot Short Fire Scenarios Description of Activity:

The proposed activity ensures that the valves of the Residual Heat Removal (RHR),

Nuclear Boiler, High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems required for safe shutdown will remain in the desired position during a postulated fire. This ECR involves modifying the circuitry of these systems specified MOVs to prevent them from re-positioning to an undesired state, initiated by hot shorts. The activity involves: rewiring and/or the replacing of the Main Control Room (MCR) hand switches, the installation of control relays within the Motor Control Centers (MCCs) and the rewiring of limit and torque switches.

Reason for Activity:

As part of Exelons ongoing commitment to comply with the Nuclear Regulatory Commission (NRC) requirements for post-fire safe shutdown , a select few Motor Operated Valves (MOVs) have been identified by the Exelon Expert Panel that are required for the safe shutdown of Limerick Generating Station (LGS), Unit 1. A postulated fire can initiate multiple hot shorts that could result in spurious repositioning of the specific MOVs when not desired. These changes are required to address issues related to Multiple Spurious Operations (MSOs) as outlined in Nuclear Energy Institute (NEI) 00-01, Rev. 2 (Guidance for Post-Fire Safe Shutdown Circuit Analysis), update Exelons position addressing NRC Information Notice 92-18 (Potential For Loss Of Remote Shutdown Capability During A Control Room Fire) and ensure compliance with NRC Regulatory Guide 1.189, Rev 2 (Fire Protection for Nuclear Power Plants).

Effect of Activity:

The proposed activity does not change the functionality of the MOVs nor their associated components; however, the proposed activity ensures the MOVs will remain operable to

safely shutdown the plant during a postulated fire. The proposed activity prevents spurious operation of the MOVs and prevents damage to the limit and torque switches during a postulated fire. The proposed activity does not impact the plant operations, design bases, or any safety analyses described in the UFSAR.

Summary of Conclusion for the Activitys 50.59 Review:

This activity modifies the circuitry of the specified MOVs to prevent them from re-positioning to an undesired state during a postulated fire scenario initiating hot shorts.

The proposed activity does not change the intended functionality of the MOVs nor their associated components. This activity will enhance compliance with the MSO requirements. All impacts of the new configuration were examined and the screen determined there may be an adverse affect on a UFSAR described design function. An evaluation was required due to the addition of active components in a safety related system that may increase the likelihood of occurrence of a malfunction important to safety. Since the affected systems are single active failure proof, no new failure modes are introduced, the additional components are reliable, and the change in failure rate of the affected SSCs is negligible, the 50.59 evaluation concluded there was less than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety. This activity can be implemented without obtaining prior NRC approval and a license amendment is not required.

Evaluation number: LG2011E003 Rev.0 50.59 Reviewer approval date: 7/11/11 PORC number: 11-022 PORC approval date: 7/13/11 Implementing document: ECR 10-00247 Rev.0 Unit 1 ECR 10-00287 Rev.0 Unit 2 Evaluator: Ken Collier Reviewer: Mark Gift Unit 1 Unit 2 Units 1 & 2 Common Unit applicability: [ ] [ ] [ x] [ ]

Complete on: [ x] [ ] [ ] [ ]

Title:

Reactor Recirculation M/G Set Replacement with ASD Units Description of Activity:

This activity will install Adjustable Speed Drives (ASDs) to replace the existing Reactor Recirculation motor-generator (M/G) Sets at Limerick Generating Station, Units 1 and 2.

The main components of the M/G Set being replaced are a drive motor, oil fluid drive coupling with scoop tube and positioner, generator, and oil cooling heat exchangers and pumps. The ASDs main components are a multiple winding step-down transformer, AC-DC-variable AC power converters, water cooling system with heat exchangers, and dual control and protective microprocessor systems with independent back-up protection relays. The ASD is a solid state static 3 phase inverter which converts the 60 Hz input line frequency to a variable output frequency to control the Reactor Recirculation Motor speed in the same manner as the existing M/G Set.

New Main Control Room (MCR) analog controls & indications will be installed on Recirculation Control panel 1(2)0-C602 and Reactor Control panel 1(2)0-C603 to interface with the ASD digital control system to provide and monitor command signals to control recirculation pump speed. A new Human-Machine-Interface (HMI) touch screen panel will be installed for each ASD on MCR panel 1(2)-C626 to echo the indications at the HMI touch screen mounted locally on each ASD drive panel. The locally mounted HMI screen provides indication & control of ASD parameters whereas the MCR HMI screen provides ASD parameter status indication only.

The existing 13.2kV feeder breaker for the M/G Sets will be reused as input for each ASD. ASD output voltage, frequency, current and volts/hertz ratio will match that currently specified for the Recirc pump motor. New digital protective relays suitable for the ASD transformer load will replace the existing analog M/G Set drive motor protective relays. New digital protective relays, configured in a 2-out-of-3 logic, will replace the existing analog relays for differential over-current protection of the recirc pump motor.

New over-frequency relays, also configured in a 2-out-of-3 logic to trip the 13.2kV breaker, will provide back-up over-speed protection for the Recirc pump motor.

Each ASD will have a new closed loop demineralized cooling water system with heat exchangers cooled by the Service Water system. Each unit will have three 100%

capacity heat exchangers, one serving each ASD with one swing heat exchanger in reserve. All three heat exchangers will be located in the same area as the existing M-G set lube oil coolers. The third swing heat exchanger allows replacement of a fouled heat exchanger while on-line, which is an improvement over the one existing lube oil cooler for each M/G Set. The closed loop cooling water system cools the ASD transformer, power cells, and panel internals using internal ASD pumps.

Reason for Activity:

Operation of the Reactor Recirculation System is critical to power production. The existing M/G Set equipment and controls are obsolete and have experienced numerous failures and spurious operation resulting in unexpected speed changes. This results in reduced unit output, unexpected reactivity changes or a plant trip. The existing Reactor Recirculation M/G Sets are original equipment. The vintage of this equipment necessitates heavy reliance on large rotating electrical and hydraulic energy conversion components and their associated moving parts, and requires a significant maintenance effort with each refueling cycle. The performance of these components has been degrading with age, and the cost of maintaining the system to achieve expected reliable performance is increasing. Additionally, some of the major components are in need of replacement/refurbishment but are obsolete and no longer supported by the Original Equipment Manufacturer (OEM), GE.

The ASD uses solid state power conversion and highly reliable state-of-the-art digital control systems to produce the frequency and voltage applied to the pump motor. This precise control results in a finer and more stable control of the Recirc pump motor speed, and thus recirculation flow. Reactor Recirc pump will be more precise, controllable, stable, and reliable due to the fully redundant microprocessor based electronic control systems versus the existing obsolete and unreliable single failure vulnerable electro-mechanical control system. Replacing the M/G Set and fluid coupling eliminates the variability, spurious fluctuations, and instability experienced with the present antiquated analog control and fluid coupling technology. Additionally, the elimination of the fluid coupling and the associated oil support system also removes a significant maintenance effort to maintain and replace the fluid coupling and lubrication oil required for the M/G Set operation.

The ASDs are much more efficient than the existing M/G sets, and they use less power than the M/G sets. The increase in efficiency by use of the ASDs means that station loads during operation will decrease, thus providing more Megawatts to the grid. Increasing the reliability of the Recirc speed control system results in more stable, reliable plant output which ultimately results in increased grid stability.

Effect of Activity:

The Reactor Recirculation system is a non-safety-related system that is necessary for power production. The function of the Reactor Recirculation system to control reactor power level during normal power production operations is not affected by this modification.

Startup of the ASD is slightly different than startup of an M/G Set due to the changes in method of control and different types of startup permissives present in the control circuits. The ability to produce manual Reactor Recirc Pump runbacks has been added to aid Operations in system control when other plant conditions require it. The licensing basis is not otherwise affected by the changes to the Reactor Recirc system due to these control changes.

Protective actions such as reactor scram are provided by other systems external to the Reactor Recirculation system, are safety-related, and completely independent of the Reactor Recirculation Control System. There is no change to the Reactor Protection System and Engineered Safety Feature Actuation System. Thus, for any Reactor Recirculation System transient or accident, the Reactor Protection System and Engineered Safety Features Actuation System will provide exactly the same response as before the proposed activity.

Safety-Related Requirements All equipment and circuits affected by this activity are non-safety-related, and not subject to environmental qualification (EQ).

Licensing Impacts The UFSAR must be revised to document the changes made by this project. The affected UFSAR sections are:

Chapter 1, Table 1.3-8, Significant Design Changes from PSAR to FSAR Chapter 1, Table 1.10-1, Acronyms Used in UFSAR Chapter 3, Section 3.8.4.1.8, Turbine Enclosure Chapter 5, Table 5.2-3, Reactor Coolant Pressure Boundary Materials Chapter 5, Section 5.4.1, Reactor Recirculation Pumps Chapter 5, Table 5.4-1, Reactor Recirculation System Design Characteristics Chapter 7, Section 7.1.1, Identification of Safety-Related Systems Chapter 7, Section 7.7, Control Systems not Required for Safety Chapter 8, Section 8.1, Electrical Power - Regulatory Guides & IEEE Standards Chapter 8, Section 8.6, Electrical Power - Onsite Power Systems Appendix 9A, Fire Protection Evaluation Report Chapter 9, Section 9.4, HVAC Systems Chapter 9, Section 9.5, Other Auxiliary Systems Chapter 15, Section 15.3, Decrease in Reactor Coolant System Flow Rate

Chapter 15, Section 15.9, Plant Nuclear Safety Operational Analysis The revised Fire Protection Review is being reviewed separately per LS-AA-128.

The Technical Specifications, LCOs and Bases, were reviewed and no impacts were identified.

The Technical Requirements Manual Section was reviewed and no impacts were identified.

Summary of Conclusion for the Activitys 50.59 Review:

A 50.59 Screening reviewed the installation of the ASD as the replacement power drive for the M/G Set to the recirculation pump motor. The following changes associated with the proposed activity were determined to be potentially adverse since each was judged to fundamentally alter the existing means of performing or controlling design functions as described in the UFSAR:

• Change from analog to digital control system with failure modes different than the existing analog system

• Changes in acceleration capability and pump coast down time due to the ASD using direct frequency control and without any inherent inertia.

• Change in the Human-System-Interface (HSI) for recirc pump manual speed control from a continuously variable analog system to an incremental discrete step digital system.

A 50.59 Evaluation is required because of the need to consider the changes in the method of controlling the Reactor Recirculation pump with the use of digital interfaces with the digital ASD control and ultimately the control and monitoring of core flow. The use of an ASD using electronic frequency conversion changes the pump response time on ASD feeder breaker trip, and introduces the potential for higher frequency and faster acceleration than presently possible with the M/G Set. However, for the transients and accidents analyzed in the Evaluation, it is the ATWS/RPT breaker that trips the recirc pump motor and not the input breaker trip. Therefore, the faster coast down time with input breaker trip is not a factor in this analysis. Also, the internal ASD limits are backed-up by independent over-frequency trip relays to protect the motor from inadvertent overspeeding. Thus, these conditions have been analyzed and determined to be not adverse.

An FMEA/PRA performed on the ASD digital hardware concluded that failure with the ASD has less probability than with the present analog system. The ASD has multiple redundancy built into the power and control systems resulting in a very reliable system.

A System Integrity Review (SIR) was prepared to document the acceptability of the ASD for recirc control. The SIR concluded that the ASD is reliable and of sufficient quality such that the design minimizes challenges to safety systems and is acceptable for this application. Software used by the ASD was found to be of high quality and common cause software failures were determined to be bounded by existing transient analysis.

The change in HSI was determined not to result in an increased probability of operator error which could lead to an unexpected speed excursion. The ASD HSI preserves the functionality of the present recirc control system while providing additional safeguards not present in the existing HSI to prevent inadvertent operator actions. New functionality was introduced, manually initiated runbacks, to increase operator response to transients while removing operator burden. The manual runback pushbuttons have guards in place to prevent inadvertent operator initiation. The redundancy and self check software built into the HSI Input/Output modules provides further assurance against spurious system manipulations. Thus, the ASD HSI was found to not result in an increase in the frequency of occurrence of an accident nor does it increase the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

Probability Risk Assessment (PRA) model applicability review was completed. IR

1. 965797, assignment 51 has been assigned to ensure that Limerick Design / System Engineer review is performed prior to next PRA update, to ensure that the PRA modeling of the Reactor Recirculation System remains consistent with the as-built plant.

The affected sections of the UFSAR are outlined in the effects section above. UFSAR Change Request Number 2011-001 is processing these changes for Unit 1.

No increases in transients, accidents, malfunctions, consequences, or increased probabilities of likelihood of an accident or malfunction have been identified as adverse in the 50.59 review process. There are no Technical Specification or Operating License changes required in this activity. Based on the 50.59 Evaluation identified below, in conjunction with the Fire Protection Review acceptability, the activity may be implemented per plant procedures without obtaining a License Amendment.

Evaluation number: LG2011E003 Rev.1 50.59 Reviewer approval date: 11/17/11 PORC number: WA-11-008 PORC approval date: 1/4/12 Implementing document: ECR LG 10-00247 Rev.3 Unit 1 ECR LG 10-00287 Rev.0 Unit 2 Evaluator: Ken Collier Reviewer: Mark Gift Unit 1 Unit 2 Units 1 & 2 Common Unit applicability: [ ] [ ] [ x] [ ]

Complete on: [ x] [ ] [ ] [ ]

Title:

Reactor Recirculation M/G Set Replacement with ASD Units Description of Activity:

• • • • • • • • • • • • • • • • • • • • • • • • Revision 1 Changes *****************************

This 50.59 Review is being revised primarily to correct a typo in the answer to Evaluation Question #1. The following sentence contains the typo:

The ASD system was reviewed for Cyber Security compliance and determined to be a non-critical system.

In the above sentence, the term non-critical should be changed to Critical Digital Asset.

The above statement is incorrect in that the ASD system was determined to be a Critical Digital Asset system as previously determined and documented by the Cyber Security engineer in the Cyber Security program database. This statement is an error since the preparer of the 50.59 Evaluation was already aware of the cyber security classification of the ASD as a Critical Digital Asset. It represents a typo that was inadvertently entered into the 50.59 Evaluation. This statement does not appear anywhere else in the 50.59 Review. This statement has no impact on the conclusion of the 50.59 Evaluation. It was included in a section discussing the various cyber security safeguards for the ASD software. Adequate justification is provided in this paragraph as to why the ASD is properly safeguarded, commensurate for a Critical Digital Asset system. The cyber security engineer determined that the ASD system is adequately safeguarded from cyber security intrusion or attacks, as is, with no added or compensatory measures required.

The characteristics that safeguard the integrity of the ASD software are already adequately and completely described in the 50.59 Evaluation. This typo has no impact on the response to the Evaluation question #1 or any other Evaluation question.

This typo was discovered during an NRC Inspection activity # 7111117 Permanent Plant Modifications which included review of LGS 50.59 Evaluations. This issue and its implications were discussed between the cyber security engineer and the NRC inspector.

The cyber security engineer provided evidence of the correct cyber security classification of Critical Digital Asset as documented in the cyber security database. The NRC inspector has accepted this response, his concerns have been addressed, and he acknowledges that this is a simple typo. This issue is being tracked by IR# 1285226.

Subsequent to PORC approval of this 50.59 Review, new UFSAR sections have been identified as impacted in addition to those already listed in this 50.59 Review. These additional UFSAR section impacts are added to the listings but do not change the responses or conclusions of the 50.59 Cover Sheet, Screening, or Evaluation. These additional impacts involve simple deletion of references to the obsolete M/G Set system and/or replace them with references to the new ASD system. They are editorial in nature.

The UFSAR change packages for both Units 1 & 2 are currently being processed and will be approved prior to ASD implementation for each respective unit. In addition, the UFSAR Change Request number for the Unit 2 changes are added to this revision 1.

All Revision 1 changes are identified with a single vertical bar at the end of the affected line.

• • • • • • • • • • • • • • • • • • • • • • End of Revision 1 Changes **************************

This activity will install Adjustable Speed Drives (ASDs) to replace the existing Reactor Recirculation motor-generator (M/G) Sets at Limerick Generating Station, Units 1 and 2.

The main components of the M/G Set being replaced are a drive motor, oil fluid drive coupling with scoop tube and positioner, generator, and oil cooling heat exchangers and pumps. The ASDs main components are a multiple winding step-down transformer, AC-DC-variable AC power converters, water cooling system with heat exchangers, and dual control and protective microprocessor systems with independent back-up protection relays. The ASD is a solid state static 3 phase inverter which converts the 60 Hz input line frequency to a variable output frequency to control the Reactor Recirculation Motor speed in the same manner as the existing M/G Set.

New Main Control Room (MCR) analog controls & indications will be installed on Recirculation Control panel 1(2)0-C602 and Reactor Control panel 1(2)0-C603 to interface with the ASD digital control system to provide and monitor command signals to control recirculation pump speed. A new Human-Machine-Interface (HMI) touch screen panel will be installed for each ASD on MCR panel 1(2)-C626 to echo the indications at the HMI touch screen mounted locally on each ASD drive panel. The locally mounted HMI screen provides indication & control of ASD parameters whereas the MCR HMI screen provides ASD parameter status indication only.

The existing 13.2kV feeder breaker for the M/G Sets will be reused as input for each ASD. ASD output voltage, frequency, current and volts/hertz ratio will match that currently specified for the Recirc pump motor. New digital protective relays suitable for

the ASD transformer load will replace the existing analog M/G Set drive motor protective relays. New digital protective relays, configured in a 2-out-of-3 logic, will replace the existing analog relays for differential over-current protection of the recirc pump motor.

New over-frequency relays, also configured in a 2-out-of-3 logic to trip the 13.2kV breaker, will provide back-up over-speed protection for the Recirc pump motor.

Each ASD will have a new closed loop demineralized cooling water system with heat exchanges cooled by the Service Water system. Each unit will have three 100% capacity heat exchangers, one serving each ASD with one swing heat exchanger in reserve. All three heat exchangers will be located in the same area as the existing M-G set lube oil coolers. The third swing heat exchanger allows replacement of a fouled heat exchanger while on-line, which is an improvement over the one existing lube oil cooler for each M/G Set. The closed loop cooling water system cools the ASD transformer, power cells, and panel internals using internal ASD pumps.

Reason for Activity:

Operation of the Reactor Recirculation System is critical to power production. The existing M/G Set equipment and controls are obsolete and have experienced numerous failures and spurious operation resulting in unexpected speed changes. This results in reduced unit output, unexpected reactivity changes or a plant trip. The existing Reactor Recirculation M/G Sets are original equipment. The vintage of this equipment necessitates heavy reliance on large rotating electrical and hydraulic energy conversion components and their associated moving parts, and requires a significant maintenance effort with each refueling cycle. The performance of these components has been degrading with age, and the cost of maintaining the system to achieve expected reliable performance is increasing. Additionally, some of the major components are in need of replacement/refurbishment but are obsolete and no longer supported by the Original Equipment Manufacturer (OEM), GE.

The ASD uses solid state power conversion and highly reliable state-of-the-art digital control systems to produce the frequency and voltage applied to the pump motor. This precise control results in a finer and more stable control of the Recirc pump motor speed, and thus recirculation flow. Reactor Recirc pump will be more precise, controllable, stable, and reliable due to the fully redundant microprocessor based electronic control systems versus the existing obsolete and unreliable single failure vulnerable electro-mechanical control system. Replacing the M/G Set and fluid coupling eliminates the variability, spurious fluctuations, and instability experienced with the present antiquated analog control and fluid coupling technology. Additionally, the elimination of the fluid coupling and the associated oil support system also removes a significant maintenance effort to maintain and replace the fluid coupling and lubrication oil required for the M/G Set operation.

The ASDs are much more efficient than the existing M/G sets, and they use less power than the M/G sets. The increase in efficiency by use of the ASDs means that station loads during operation will decrease, thus providing more Megawatts to the grid. Increasing the

reliability of the Recirc speed control system results in more stable, reliable plant output which ultimately results in increased grid stability.

Effect of Activity:

The Reactor Recirculation system is a non-safety-related system that is necessary for power production. The function of the Reactor Recirculation system to control reactor power level during normal power production operations is not affected by this modification.

Startup of the ASD is slightly different than startup of an M/G Set due to the changes in method of control and different types of startup permissives present in the control circuits. The ability to produce manual Reactor Recirc Pump runbacks has been added to aid Operations in system control when other plant conditions require it. The Licensing basis is not otherwise affected by the changes to the Reactor Recirc system due to these control changes.

Protective actions such as reactor scram are provided by other systems external to the Reactor Recirculation system, are safety-related, and completely independent of the Reactor Recirculation Control System. There is no change to the Reactor Protection System and Engineered Safety Feature Actuation System. Thus, for any Reactor Recirculation System transient or accident, the Reactor Protection System and Engineered Safety Features Actuation System will provide exactly the same response as before the proposed activity.

Safety-Related Requirements All equipment and circuits affected by this activity are non-safety-related, and not subject to environmental qualification (EQ).

Licensing Impacts The UFSAR must be revised to document the changes made by this project. The affected UFSAR sections include (but are not necessarily limited to) the following:

Chapter 1, Table 1.3-8, Significant Design Changes from PSAR to FSAR Chapter 1, Table 1.10-1, Acronyms Used in UFSAR Chapter 3, Table 3.2.1, LGS Design Criteria Summary Chapter 3, Section 3.8.4.1.8, Turbine Enclosure Chapter 4, Section 4.4.3, Description of Thermal and Hydraulic Design of the Reactor Coolant System Chapter 5, Table 5.2-3, Reactor Coolant Pressure Boundary Materials Chapter 5, Section 5.4.1, Reactor Recirculation Pumps Chapter 5, Table 5.4-1, Reactor Recirculation System Design Characteristics Chapter 7, Section 7.1.1, Identification of Safety-Related Systems Chapter 7, Section 7.7, Control Systems not Required for Safety Chapter 8, Section 8.1.3, Electrical Power - Onsite Power Systems

Chapter 8, Figure 8.1-4, Medium Voltage System Chapter 8, Section 8.1.6, Electrical Power - Regulatory Guides & IEEE Standards Appendix 9A, Fire Protection Evaluation Report Chapter 9, Section 9.4, HVAC Systems Chapter 9, Section 9.5, Other Auxiliary Systems Chapter 15, Section 15.3, Decrease in Reactor Coolant System Flow Rate Chapter 15, Section 15.9, Plant Nuclear Safety Operational Analysis The revised Fire Protection Review is being reviewed separately per LS-AA-128.

The Technical Specifications, LCOs and Bases, were reviewed and no impacts were identified.

The Technical Requirements Manual Section was reviewed and no impacts were identified.

Summary of Conclusion for the Activitys 50.59 Review:

A 50.59 Screening reviewed the installation of the ASD as the replacement power drive for the M/G Set to the recirculation pump motor. The following changes associated with the proposed activity were determined to be potentially adverse since each was judged to fundamentally alter the existing means of performing or controlling design functions as described in the UFSAR:

• Change from analog to digital control system with failure modes different than the existing analog system

• Changes in acceleration capability and pump coast down time due to the ASD using direct frequency control and without any inherent inertia.

• Change in the Human-System-Interface (HSI) for recirc pump manual speed control from a continuously variable analog system to an incremental discrete step digital system.

A 50.59 Evaluation is required because of the need to consider the changes in the method of controlling the Reactor Recirculation pump with the use of digital interfaces with the digital ASD control and ultimately the control and monitoring of core flow. The use of an ASD using electronic frequency conversion changes the pump response time on ASD feeder breaker trip, and introduces the potential for higher frequency and faster acceleration than presently possible with the M/G Set. However, for the transients and accidents analyzed in the Evaluation, it is the ATWS/RPT breaker that trips the recirc pump motor and not the input breaker trip. Therefore, the faster coast down time with input breaker trip is not a factor in this analysis. Also, the internal ASD limits are backed-up by independent over-frequency trip relays to protect the motor from inadvertent overspeeding. Thus, these conditions have been analyzed and determined to be not adverse.

An FMEA/PRA performed on the ASD digital hardware concluded that failure with the ASD has less probability than with the present analog system. The ASD has multiple

redundancy built into the power and control systems resulting in a very reliable system.

A System Integrity Review (SIR) was prepared to document the acceptability of the ASD for recirc control. The SIR concluded that the ASD is reliable and of sufficient quality such that the design minimizes challenges to safety systems and is acceptable for this application. Software used by the ASD was found to be of high quality and common cause software failures were determined to be bounded by existing transient analysis.

The change in HSI was determined not to result in an increased probability of operator error which could lead to an unexpected speed excursion. The ASD HSI preserves the functionality of the present recirc control system while providing additional safeguards not present in the existing HSI to prevent inadvertent operator actions. New functionality was introduced, manually initiated runbacks, to increase operator response to transients while removing operator burden. The manual runback pushbuttons have guards in place to prevent inadvertent operator initiation. The redundancy and self check software built into the HSI Input/Output modules provides further assurance against spurious system manipulations. Thus, the ASD HSI was found to not result in an increase in the frequency of occurrence of an accident nor does it increase the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

Probability Risk Assessment (PRA) model applicability review was completed. IR #

965797, assignment 51 has been assigned to ensure that Limerick Design / System Engineer review is performed prior to next PRA update, to ensure that the PRA modeling of the Reactor Recirculation System remains consistent with the as-built plant.

The affected sections of the UFSAR are outlined in the effects section above. UFSAR Change Request Number 2011-001 is processing these changes for Unit 1. UFSAR Change Request Number 2011-021 is processing these changes for Unit 2.

No increases in transients, accidents, malfunctions, consequences, or increased probabilities of likelihood of an accident or malfunction have been identified as adverse in the 50.59 review process. There are no Technical Specification or Operating License changes required in this activity. Based on the 50.59 Evaluation identified below, in conjunction with the Fire Protection Review acceptability, the activity may be implemented per plant procedures without obtaining a License Amendment.

Evaluation number: LG2011E004 Rev.0 50.59 Reviewer approval date: 10/6/11 PORC number: 11-031 PORC approval date: 10/18/11 Implementing document: IR 1208490-07 Evaluator: Joe Mittura/Greg Curtin Reviewer: Mark Gift Unit 1 Unit 2 Units 1 & 2 Common Unit applicability: [ ] [ ] [ x] [ ]

Complete on: [ ] [ ] [ x] [ ]

Title:

Assessment of the Affect of ECR 01-00816 on Station Black out Coping Description of Activity:

The original, licensed description of the Station Blackout (SBO) scenario, describes the initial plant conditions as the loss of all off site power on both units and the loss of all Emergency Diesel Generators (EDG) on one unit. Additionally, the unit that has EDGs available (the non-blacked out (NBO) unit) must consider the additional failure of a diesel due to the consideration of a single active failure. No other failures or design basis events are postulated. If the EDG that fails on the NBO unit powers an ESW pump motor, then the associated ESW pump would not start and the ESW loop would not be available to cool the associated EDG on the NBO unit. There are two ESW pump motors powered from the EDGs of each unit, therefore, there is a 1 in 2 probability that one of the three EDGs that were postulated to be operating, would be shut down due to high jacket water temperature. Thus, soon after the initiation of an SBO event, there would be two EDGs operating on the NBO unit.

Limerick is licensed as an alternate AC (AAC) plant for coping with an SBO event. The blacked out unit (BO), with no off site power or EDGs must cope without AC for one hour. During this initial hour, the NBO unit will cross connect the 101 and 201 busses from the operating EDGs to the BO unit's safeguard busses to start necessary loads on the BO unit. The station procedures indicate that priority should be given to restoring ESW pumps. When power is restored to the NBO's ESW pump motor and the ESW pump is operating, then the EDG that was shut down due to high jacket water temperature may be restarted and utilized to power plant loads. This cross connecting activity, resupply of ESW to the EDG and the EDG restart is expected to be completed within the first hour of the SBO event. This meets the license condition of having two EDGs (stated in the licensing correspondence as more than one diesel but less than two) available on the NBO unit to cope with safe shutdown loads and to have an additional (three total) EDGs to support the BO unit within one hour.

As an alternate method to provide cooling to the third EDG that would have lost ESW cooling due to the failure of the EDG on the NBO unit, transfer switches in the main control room (MCR) were available to transfer cooling for each EDG from their primary ESW loop to their alternate ESW loop. In the original plant design, these transfer switches would be available to quickly transfer the cooling water without having to leave the MCR. This action was not specifically described in operating procedures, nor was it described in the UFSAR.

In 2000, ECR 01-00816, Revision 1 was dispositioned to resolve an NCR condition identified under Al322730. ECR 01-00816, Revision 1addressed an unanalyzed configuration that would be created by the inadvertent repositioning of the ESW cooling valves to the EDGs. ECR 01-00816 de-energizes the motor operators associated with the alternate ESW loop supply/return valves HV-01l-*3lB/D, HV-0l1-*33A/C, HV-011-32B/D, and HV-011-*34A/C for the EDGs by opening the valves associated motor control center breaker and locking them open. De-energizing the motor operators maintain the valves in their normally closed position and therefore eliminate the potential for inadvertent valve opening. The remote operating capabilities of the subject valves from the main control room are defeated when the breaker is de-energized. The ECR also defeats the input to the associated EDG Out of Service annunciator by installing a jumper in the valve logic in each breaker cubical.

Reason for Activity:

ECR 01-00816 did not discuss the impact of the removal of the remote operating capability on the SBO coping scenario. This 50.59 screening and evaluation assess the removal of the remote operating capabilities of the ESW cooling valves to the EDGs during normal and emergency, including SBO, conditions.

Effect of Activity:

ECR 01-00816 causes a loss of remote operation and position indication for the subject valves. ECR 01-00816 also removes one option for recovering the cooling to an EDG on the NBO unit when the failed EDG supports an ESW pump motor. SBO coping is still consistent with the licensing description since cooling to the third EDG on the NBO unit can be made available after cross connecting to the BO unit within the first hour.

Summary of Conclusion for the Activitys 50.59 Review:

The screening identifies two questions that are answered yes. Question 1 identifies that the change could have an adverse affect on an SSC design function that is described in the UFSAR. Question 2 identifies that the change may adversely affect how a UFSAR described SSC design function is performed or controlled. Thus, an evaluation is also performed. The resultant evaluation showed that the affects of this change were minimal under SBO conditions since the SBO and the BO unit can both be safely shutdown using existing procedures without complications. The license conditions for safe shutdown and

reactor cooling have not changed. NRC prior approval is not required for the implementation of the 01-00816 activity.

Evaluation number: LG2012E001 Rev.0 50.59 Reviewer approval date: 2/15/12 PORC number: 12-007 PORC approval date: 2/17/12 Implementing document: 1GP-6.1 Rev.25 GP-2 Rev.144 S53.3.B Rev.15 S53.3.I Rev.23 S53.0.A Rev.25 Evaluator: Mark Gift Reviewer: Greg Curtin Unit 1 Unit 2 Units 1 & 2 Common Unit applicability: [ x] [ ] [ ] [ ]

Complete on: [ x] [ ] [ ] [ ]

Title:

Provide an Alternate Means of Monitoring a Reactor Well Seal Leak for Unit 1 Description of Activity:

As a pre-outage activity for 1R14, the reactor well seal rupture drain high flow alarm (FSH-053-101) is flushed and tested to ensure that the function of the alarm is adequately tested. ST-4-015-490-1 failed because legacy foreign material cannot be retrieved and clogs the flow switch sensing line which prevents the alarm function. The step to verify the operation of the alarm (4.4.7) is a required step to pass the ST. ST-4-015-490-1 is required to start flood-up.

UFSAR Section 9.1.3.5 states that high leakage rates through the refueling bellows or reactor well seals are annunciated at the refueling floor control panel and by a common trouble alarm in the main control room. Due to blockage in the instrument line, this alarm function is no longer available. The description is not part of a method of evaluation and draws no conclusions as to its impact on design basis.

This activity will revise plant procedures to allow for alternate means of monitoring reactor well seal rupture drain flow. The alternate means of monitoring is a camera trained on the seal drainage trough. This camera will be monitored on a one hour frequency during filling operations and every four hours thereafter when normal reactor well level has been established. The loss of inventory through the seals is limited by the seal plate and is determined to provide sufficient time for operator action to be taken given the frequency of surveillance. Also, in addition to the camera surveillance, several other methods of detecting seal problems exist. These include operator monitoring of the pool level at the skimmer surge tank weir, seal pressure, fuel pool low level alarm, and level monitoring in the main control room.

Reason for Activity:

Procedure ST-4-015-490-1 is required to be completed satisfactorily to begin reactor cavity flood up. The reactor well seal rupture drain high flow alarm has been determined to not be necessary if alternate means of detecting seal leakage are available. Plant procedures changes will allow for alternate monitoring if FSH-053-101 is not available.

Effect of Activity:

The subject procedures are only used in preparation for a refueling outage. The ST procedure verifies the ability of the reactor well seals to effectively separate the reactor well from the reactor enclosure. The reactor well seals are designed for zero leakage. A failure of both of the redundant seals is required for leakage to enter the leakage trough.

Seal leakage will drain into a trough below the seals and then drain into an 8 inch drain line. This drain line has a check valve and a 1 inch line that bypasses the check valve.

The bypass line has a flow orifice and differential pressure switch with an alarm.

If the alarm is out of service before or during flood up conditions, alternate monitoring methods may be employed to ensure that there is not excessive leakage into the reactor well seal rupture drain. This activity does not remove the requirement to monitor the leakage, but provides an alternate means of monitoring the leakage.

Summary of Conclusion for the Activitys 50.59 Review:

An alternate means of monitoring the reactor well seal rupture drain may be used before and during reactor well flooded operations. A 50.59 evaluation was performed and all questions were answered "no". The procedure changes may be applied without the prior concurrence of the NRC.

Evaluation number: LG2012E002 Rev.0 50.59 Reviewer approval date: 5/31/12 PORC number: 12-025 PORC approval date: 6/25/12 Implementing document: ECR LG 09-00343 Rev.0 Evaluator: Ken Collier Reviewer: Mark Gift Interface Reviewer: Mike May Unit 1 Unit 2 Units 1 & 2 Common Unit applicability: [ ] [ ] [ x] [ ]

Complete on: [ ] [ ] [ x] [ ]

Title:

Correct for Potential Reactor Protection System Level 3 (+12.5") Setpoint Error Description of Activity:

The scope of this Engineering Change Request (ECR) is to lower the Reactor Protection System (RPS) Reactor Water Level 3 (L3) scram Analytical Limit (AL) to account for a potential reactor water level process measurement error (PMA). The Tech Spec Allowable Value (AV), Nominal Trip Setpoint (NTSP) & Tech Spec Actual Trip Setpoint (ATSP) for the RPS scram L3 instrumentation loops are not impacted or changed by this activity. This activity revises the Improved Instrument Setpoint Control Program (IISCP) Instrument Loop Uncertainty (LU) calc for the RPS scram L3 instrument channels LT-042-1(2)N080A(B,C,D) to reflect this change in AL. There is no physical work associated with this activity. This activity involves only changes to the IISCP LU calculations to account for this hitherto unknown process measurement error.

Reason for Activity:

A steam flow induced error was identified by GE Hitachi Nuclear Energy (GEH-NE) via a 10CFR Part 21 Notification # SC04-14, dated 10/11/2004. Based on subsequent GEH report # GEH-NE-0000-0077-4603-R1 (10/08), "Evaluation of Steam Flow Induced Error (SFIE) Impact on the L3 Setpoint Analytical Limit", a SFIE exists where water level could reach the bottom of the dryer skirt and allow steam to bypass into the annulus during a Loss of Feedwater (LOFW) event. This steam bypass adversely affects the RPS L3 vessel water level measurement, which relies on a reference pressure tap in the annulus. The GEH-NE evaluation quantified this error and determined its impact on various analyses, including LOCA and LOFW which are two events potentially impacted by this error. One method the GEH-NE report recommends to account for this SFIE is to lower the Analytical Limit (AL) for the RPS scram L3 water level, with no impact to the station Tech Spec RPS scram L3 setpoint. This method was chosen by Limerick since it is feasible for Limerick and it results in no impact to the Limerick Tech Spec AV or ATSP for RPS L3 scram.

Effect of Activity:

This issue involves an unanalyzed sensing error in the reference leg of level instrumentation for the RPS L3 scram setpoint during the Loss of Feedwater Flow (LOFW) event. During this event, steam flow phenomena occurs which causes the L3

(+12.5 inches) trip to delay actuation, resulting in the minimum water above top of active fuel being lower than analyzed. This error is caused by steam bypassed into the annulus space between the vessel wall and steam dryer skirt when water level falls to a low level during a LOFW event. GEH-NE evaluated this error following a Part 21 notice and concluded that applying the minimum water level to all BWRs would still leave adequate water level above the safety limit for all plants. This evaluation concludes that Limerick can correct for this error by adjustment to the Analytical Limit for the RPS scram L3 setpoint. Although this ECR lowers the RPS L3 scram AL, this does not impact the current RPS L3 scram AV, NTSP, or Tech Spec ATSP. Also, this change has no impact on the current ADS confirmatory L3 AV, NTSP, or Tech Spec ATSP as documented in Loop Uncertainty calc LT-042-2N095A. Therefore, there is no impact on the system operational requirements. No Limiting Safety System Settings are impacted.

Summary of Conclusion for the Activitys 50.59 Review:

Lowering the RPS L3 AL to account for the SFIE affects the Loss of Feedwater (LOFW) analysis and the Loss of Coolant Accident (LOCA) analysis. The GEH-NE report evaluated the impact of SFIE values that are bounding for Limerick (LGS) on the LOFW and LOCA analysis. This analysis showed that the impact of the SFIE is minor and acceptable. This GEH-NE report evaluated the impact on other events, systems and functions affected by SFIE and the reduction in L3 analytical limit and determined the impact to be insignificant. These functions are not affected by the delay in L3 scram.

However, since the water level during the accident or transient is lower than previously analyzed, this is considered an adverse change to a UFSAR described design function. A 50.59 Evaluation was prepared which resulted in all questions answered 'No', therefore prior NRC approval is not required to implement this activity.

A UFSAR change package has been prepared to include notes in applicable UFSAR sections clarifying the affect the changes due to SFIE.