ML16110A392

Response to Draft Request for Additional Information Regarding Proposed Revision to Technical Specifications in Response to GE Energy - Nuclear 10 CFR Part 21 Safety Communication SC05-03

ML16110A392
Person / Time
Site:	Limerick
Issue date:	04/19/2016
From:	David Helker Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF7263, CAC MF7264
Download: ML16110A392 (30)

Text

200 Exelon Way Exelon Generation Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.90 April 19, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353

Subject:

Response to Draft Request for Additional Information Regarding Proposed Revision to Technical Specifications in Response to GE Energy- Nuclear 10 CFR Part 21 Safety Communication SC05-03

References:

1. Letter from James Barstow (Exelon Generation Company, LLC) to U.S.

Nuclear Regulatory Commission, "License Amendment Request -

Proposed Revision to Technical Specifications in Response to GE Energy - Nuclear 10 CFR Part 21 Safety Communication SC05-03,"

dated January 15, 2016 (ADAMS Accession No. ML16015A316).

2. Electronic mail message from Richard Ennis, U.S. Nuclear Regulatory Commission, to Glenn Stewart, Exelon Generation Company, LLC, "Draft RAls - Limerick LAR to Reduce Steam Dome Pressure (CACs MF7263 & 64)," dated March 10, 2016 (ADAMS Accession No. ML16085A025).

By letter dated January 15, 2016 (Reference 1), Exelon Generation Company, LLC (Exelon) submitted a license amendment request (LAR) for Limerick Generating Station (LGS), Units 1 and 2. The proposed amendment would reduce the reactor vessel steam dome pressure associated with the Technical Specifications (TS) Safety Limits (SLs) specified in TS 2.1.1 and TS 2.1.2. The amendment would also revise the setpoint and allowable value for the main steam line low pressure isolation function in TS Table 3.3.2-2.

The proposed changes address a 10 CFR Part 21 issue concerning the potential to violate the Sls during a pressure regulator failure maximum demand (open) (PRFO) transient.

The NRC staff reviewed the information provided that supports the proposed amendment and identified the need for additional information in order to complete their evaluation of the amendment request. The draft request for additional information (RAI) was sent from the NRC to Exelon by electronic mail message on March 10, 2016 (Reference 2). A conference call was conducted on March 24, 2016, to provide clarification of the questions.

Subsequent to the teleconference, it was agreed that Exelon would respond to the RAI by April 25, 2016, which was acceptable to the NRC.

Attachment 1 to this letter provides a restatement of the RAI questions followed by our responses. Attachment 2 provides revised markups for TS page 2-1. The proposed changes to Table 3.3.2-2 on TS page 3/4 3-18 remain unchanged by this response but are

U.S. Nuclear Regulatory Commission Response to Draft Request for Additional Information Proposed Revision to Technical Specifications in Response to GE Energy - Nuclear 10 CFR Part 21 Safety Communication SC05-03 Docket Nos. 50-352 and 50-353 April 19, 2016 Page 2 included in Attachment 2 for completeness. Attachment 3 provides revised TS Bases markups (for information only). Attachment 4 provides a copy of Loop Uncertainty Calculation Ll-00032, "LU Calculation for PT-001-2N076C."

Exelon has reviewed the information supporting a finding of no significant hazards consideration, and the environmental consideration, that were previously provided to the NRC in Attachment 1 of the Reference 1 letter. Exelon has concluded that the information provided in this response does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92. In addition, Exelon has concluded that the information in this response does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.

There are no regulatory commitments in this response.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), Exelon is notifying the Commonwealth of Pennsylvania of this RAI response by transmitting a copy of this letter and its attachments to the designated State Official.

If you have any questions or require additional information, please contact Glenn Stewart at 610-765-5529.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 19th day of April 2016.

Respectfully, David P. Helker Manager, Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments:

1. Response to Draft Request for Additional Information Proposed Revision to Technical Specifications in Response to GE Energy - Nuclear 10 CFR Part 21 Safety Communication SC05-03

2. Revised Markup of Proposed Technical Specifications Pages

3. Revised Markup of Proposed Technical Specifications Bases Pages (Information Only)

4. Loop Uncertainty Calculation Ll-00032, "LU Calculation for PT-001-2N076C" cc: Regional Administrator - NRC Region I w/ attachment NRC Senior Resident Inspector - Limerick Generating Station II NRC Project Manager, NRR - Limerick Generating Station II Director, Bureau of Radiation Protection - Pennsylvania Department of Environmental Protection II

ATTACHMENT 1 License Amendment Request Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Response to Draft Request for Additional Information Proposed Revision to Technical Specifications in Response to GE Energy - Nuclear 10 CFR Part 21 Safety Communication SC05-03

Response to Draft Request for Additional Information Attachment 1 Proposed Revision to Technical Specifications in Response to GE Page 1 of 4 Energy - Nuclear 10 CFR Part 21 Safety Communication SC05-03 Docket Nos. 50-352 and 50-353 By letter dated January 15, 2016 (Reference 1), Exelon Generation Company, LLC (Exelon) submitted a license amendment request (LAR) for Limerick Generating Station (LGS), Units 1 and 2. The proposed amendment would reduce the reactor vessel steam dome pressure associated with the Technical Specifications (TS) Safety Limits (SLs) specified in TS 2.1.1 and TS 2.1.2. The amendment would also revise the setpoint and allowable value for the main steam line low pressure isolation function in TS Table 3.3.2-2. The proposed changes address a 10 CFR Part 21 issue concerning the potential to violate the SLs during a pressure regulator failure maximum demand (open) (PRFO) transient.

The NRC staff reviewed the information provided that supports the proposed amendment and identified the need for additional information in order to complete their evaluation of the amendment request. Below is a restatement of the questions followed by our responses.

SRXB-RAI-1 The current LGS TS 2.1.2 requires that the minimum critical power ratio (MCPR) be 1.09 for two recirculation loop operation and 1.12 for single recirculation loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10%

of rated flow.

An LAR dated November 19, 2015 (ADAMS Accession No. ML15323A257), for LGS Unit 1, was submitted to the NRC regarding TS 2.1, Safety Limits, to revise Safety Limit Minimum Critical Power Ratios (SLMCPRs) due to the cycle specific analysis performed by Global Nuclear Fuel for the upcoming Cycle 17. The proposed changes to the SLMCPR values are from 1.09 to 1.10 for two loop operation and from 1.12 to 1.14 for single loop operation. The NRC staff requests that the licensee clarify whether the proposed steam dome pressure change considered the SLMCPR change for TS 2.1.2 in the referenced LGS Unit 1 LAR.

Response

LGS Unit 1 transitioned to a full core of GNF2 fuel during the 1R16 refueling outage which was completed on April 17, 2016. The lower bound limit of 700 psia for the GEXL17 critical power correlation is justified for GNF2 fuel as indicated in the LAR for the proposed reactor vessel steam dome pressure change (Reference 1). The same correlation is used for the LGS Unit 1 TS SLMCPR change consistent with its range of applicability, which includes the lower bound limit of 700 psia. The LAR for the proposed reactor vessel steam dome pressure change, to extend the low pressure applicability, does not affect the LAR for the LGS Unit 1 TS SLMCPR proposed change. The noted LARs remain independent when the GEXL17 correlation is used within its application range.

The LGS Unit 1 SLMCPR Amendment No. 221 was issued by letter dated March 15, 2016 (Reference 2) and has been implemented by LGS. Therefore, the revised markup for TS page 2-1 for LGS Unit 1 included in Attachment 2 is based on the current LGS Unit 1 TS which incorporates the changes to the SLMCPR that were approved in Amendment No. 221.

Response to Draft Request for Additional Information Attachment 1 Proposed Revision to Technical Specifications in Response to GE Page 2 of 4 Energy - Nuclear 10 CFR Part 21 Safety Communication SC05-03 Docket Nos. 50-352 and 50-353 SRXB-RAI-2 The LAR states that main steam isolation valve (MSIV) low pressure isolation setpoint (LPIS) setting, calculated at 840 pounds per square inch gauge (psig), is based on the new analytical limit of 805 psig. The NRC staff requests that the licensee (1) provide a description of how the new analytical limit of 805 psig was arrived at, and (2) how the proposed MSIV LPIS setting of 840 psig is based on this new analytical limit.

Response

1. The current low LPIS setting (720 psig analytical limit) is not sufficient to preclude steam dome pressure from falling below 685 psig (700 psia) when above 25% power for current operation during a PRFO Anticipated Operational Occurrence (AOO) event. The approach discussed in Section 5 of the Boiling Water Reactor Owners Group (BWROG) report, NEDC-33743, Rev. 0 (Reference 3), was followed for application of the BWROG method to LGS. The results from Section 4 of the BWROG report, which are most applicable to the LGS configuration, were used. A change of analytical limit by scaling up the results in the BWROG report Table 5 for an increased LPIS analytical limit from 720 psig to 805 psig is required to meet the acceptance criterion. Accordingly, the approach considers the most limiting plant configuration and operating conditions for evaluating the effect of the SC05-03 issue.

2. The increased LPIS analytical limit of 805 psig was used as input to revise the loop uncertainty calculation for LGS. Based on this new analytical limit, the associated changes to allowable value and actual trip setpoint were established as part of the loop uncertainty calculation update (see Attachment 4).

SRXB-RAI-3 The NRC staff requests that the licensee discuss the impact of this Main Steam Line Pressure -

Low allowable value change, primarily focusing on the PRFO transient.

Response

PRFO Anticipated Transient Without Scram (ATWS) - An increased LPIS analytical limit requires an increase in the allowable value from 736 psig to 821 psig. The allowable value of 736 psig is used in the current LGS analysis of ATWS (Reference 4). The LGS ATWS analysis considers failure of the pressure regulator to maximum demand (PRFO) as a limiting event.

The event causes a drop in reactor vessel pressure and water level which continues until MSIV isolation is initiated on steam line low pressure isolation setpoint (LPIS). As the increased LPIS analytical limit will result in an increased allowable value that results in earlier steam line isolation and recirculation pump trip action, the LGS ATWS analysis remains applicable with respect to this change.

PRFO AOO - The analysis uses the LPIS analytical limit to initiate steam line isolation. The revised analytical limit of 805 psig, which is higher than the current value of 720 psig, will result in earlier steam line isolation to terminate depressurization. The change in LPIS analytical limit

Response to Draft Request for Additional Information Attachment 1 Proposed Revision to Technical Specifications in Response to GE Page 3 of 4 Energy - Nuclear 10 CFR Part 21 Safety Communication SC05-03 Docket Nos. 50-352 and 50-353 will not affect the significance of the PRFO event as a non-limiting event with respect to fuel thermal limit.

SRXB-RAI-4 The licensee proposes to reduce the reactor steam dome pressure specified in TS 2.1.1 and TS 2.1.2 from 785 psig to 685 psig based on the lower-bound pressure for the critical power correlation for the fuel currently used in the LGS, Units 1 and 2 cores. The licensees application references Global Nuclear Fuel (GNF) reports NEDC-33270P, NEDC-33292P and NEDC-32851P-A as the basis supporting the proposed change. The LGS Unit 1 core currently consists of GE14 and GNF2 fuel types and LGS Unit 2 uses GNF2 fuel.

Section 3.8.3 of GNF report NEDC-33270P discusses the critical power correlation for GNF2 fuel (i.e., GEXL17 correlation). This section includes the pressure range over which the GEXL17 correlation is valid for GNF2 fuel consistent with the information provided in Table 5-4 of GNF2 report NEDC-33292P. As discussed in Section 3.0 of Attachment 1 of the licensees application, the lower bound pressure limit for the GEXL17 correlation is 700 pounds per square inch atmospheric (psia).

GNF report NEDC-32851P-A discusses the critical power correlation for GE14 fuel (i.e.,

GEXL14 correlation). Similar to the GEXL17 correlation, Section 5.2 of the report states that the lower bound pressure limit for the GEXL14 correlation is 700 psia.

Converting 700 psia to psig, the lower bound pressure for the GEXL17 and GEXL14 correlations is approximately 685.3 psig. As such, the 685 psig value specified in the proposed TS change is slightly outside the pressure range in which the GEXL17 and GEXL14 correlations are valid for GNF2 and GE14 fuel. Please provide further justification for the proposed 685 psig value or propose a revised pressure value for this TS change that is supported by the GEXL17 and GEXL14 correlations (e.g., 700 psia).

Response

Exelon has decided to reference the lower bound limit for the critical power correlation in absolute pressure (i.e., 700 psia) for the GNF2 fuel currently used in the LGS, Unit 1 and Unit 2 cores, as referenced by GNF reports, NEDC-33270P and NEDC-33292P. Exelon proposes to revise the lower bound reactor steam dome pressure for the reactor core safety limits specified in TS 2.1.1 and TS 2.1.2 to reference the absolute pressure value of 700 psia. Note: The Unit 2 core already uses all GNF2 fuel. In addition, Unit 1 transitioned to all GNF2 fuel during the Unit 1 refueling outage which was completed on April 17, 2016. provides a copy of the revised TS mark-up pages that reflect the proposed change. Attachment 3 provides the corresponding revised TS Bases mark-up pages (for information only).

Response to Draft Request for Additional Information Attachment 1 Proposed Revision to Technical Specifications in Response to GE Page 4 of 4 Energy - Nuclear 10 CFR Part 21 Safety Communication SC05-03 Docket Nos. 50-352 and 50-353 EICB-RAI-1 The proposed amendment request entails changes to TS Table 3.3.2-2 and revises the trip setpoint and the allowable value for the main steam line low pressure isolation function. In order for the NRC staff to verify compliance to the regulations and the guidance pertaining to setpoint changes, the staff requests the licensee to submit the calculation for staff review. The calculation will be used to assess the methodology, the changes in assumptions, calculation of total loop uncertainty, and other pertinent information in the calculation.

Response provides a copy of Loop Uncertainty Calculation LI-00032, "LU Calculation for PT-001-2N076C" for NRC review. As discussed in a recent LGS amendment (ADAMS Accession No. ML14324A808), the LGS setpoint methodology, which is currently contained in Exelon Procedure CC-MA-103-2001, is based on the NRC-approved GE Topical Report NEDC-31336P-A, "General Electric Instrument Setpoint Methodology," dated September 1996. The NRC staff previously found the LGS setpoint methodology acceptable as discussed in an NRC letter dated February 16, 1995, "Revised Maximum Authorized Thermal Power Limit, Limerick Generating Station, Unit No. 2 (TAC No. M88393)" (ADAMS Accession No. ML011560773).

References:

1. Letter from James Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "License Amendment Request - Proposed Revision to Technical Specifications in Response to GE Energy - Nuclear 10 CFR Part 21 Safety Communication SC05-03," dated January 15, 2016 (ADAMS Accession No. ML16015A316).

2. Letter from Richard B. Ennis (U.S. Nuclear Regulatory Commission) to Bryan C. Hanson, Exelon Nuclear, "Limerick Generating Station, Unit 1 - Issuance of Amendment, RE: Safety Limit Minimum Critical Power Ratio Change (CAC No. MF7101), dated March 15, 2016 (ADAMS Accession No. ML16041A021).

3. NEDC-33743P, Revision 0, "BWR Owners' Group Reload Analysis and Core Management Committee SC05-03 Analysis Report," dated April 2012.

4. 0000-0097-1195-R0, Exelon Nuclear Limerick Units 1 and 2 Thermal Power Optimization, Task 902: Anticipated Transients Without Scram, December 2009.

ATTACHMENT 2 License Amendment Request Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Response to Draft Request for Additional Information Proposed Revision to Technical Specifications in Response to GE Energy - Nuclear 10 CFR Part 21 Safety Communication SC05-03 Revised Markup of Proposed Technical Specifications Pages Unit 1 TS Pages 2-1 3/4 3-18 Unit 2 TS Pages 2-1 3/4 3-18

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 700 psia THERMAL POWER, Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION: 700 psia With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

THERMAL POWER, High Pressure and High Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.10 for two recirculation loop operation and shall not be less than 1.14 for single recirculation loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION: 700 psia 700 psia With MCPR less than 1.10 for two recirculation loop operation or less than 1.14 for single recirculation loop operation and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with the reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

LIMERICK - UNIT 1 2-1 Amendment No. 7,30,111,127,156, 170,183,206, 221

TABLE 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE

1. MAIN STEAM LINE ISOLATION

a. Reactor Vessel Water Level

1) Low, Low - Level 2 - 38 inches* - 45 inches

2) Low, Low, Low - Level 1 - 129 inches* - 136 inches

b. DELETED DELETED DELETED 840 821

c. Main Steam Line Pressure - Low 756 psig 736 psig

d. Main Steam Line Flow - High 122.1 psid 123 psid

e. Condenser Vacuum - Low 10.5 psia 10.1 psia/ 10.9 psia

f. Outboard MSIV Room Temperature - High 192F 200F

g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High 165F 175F

h. Manual Initiation N.A. N.A.

2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION

a. Reactor Vessel Water Level Low - Level 3 12.5 inches* 11.0 inches

b. Reactor Vessel (RHR Cut-in Permissive) Pressure - High 75 psig 95 psig

c. Manual Initiation N.A. N.A.

LIMERICK - UNIT 1 3/4 3-18 Amendment No. 28, 89, 106

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 700 psia THERMAL POWER, Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION: 700 psia With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

THERMAL POWER, High Pressure and High Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.09 for two recirculation loop operation and shall not be less than 1.12 for single recirculation loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

700 psia 700 psia With MCPR less than 1.09 for two recirculation loop operation or less than 1.12 for single recirculation loop operation and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATION CONDITIONS 1, 2, 3, and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

LIMERICK - UNIT 2 2-1 Amendment No. 14, 83, 87, 97, 114, 127, 162

TABLE 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE

1. MAIN STEAM LINE ISOLATION

a. Reactor Vessel Water Level

1) Low, Low - Level 2 > - 38 inches* > - 45 inches

2) Low, Low, Low - Level 1 > - 129 inches* > - 136 inches

b. DELETED DELETED DELETED 840 821

c. Main Steam Line Pressure - Low > 756 psig > 736 psig

d. Main Steam Line Flow - High < 122.1 psid < 123 psid

e. Condenser Vacuum - Low 10.5 psia >10.1 psia/ 10.9 psia

f. Outboard MSIV Room Temperature - High < 192°F < 200°F

g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High < 165°F < 175°F

h. Manual Initiation N.A. N.A.

2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION

a. Reactor Vessel Water Level Low - Level 3 > 12.5 inches* > 11.0 inches

b. Reactor Vessel (RHR Cut-in Permissive) Pressure - High < 75 psig < 95 psig

c. Manual Initiation N.A. N.A.

LIMERICK - UNIT 2 3/4 3-18 Amendment No. 51, 52

ATTACHMENT 3 License Amendment Request Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Response to Draft Request for Additional Information Proposed Revision to Technical Specifications in Response to GE Energy - Nuclear 10 CFR Part 21 Safety Communication SC05-03 Revised Markup of Proposed Technical Specifications Bases Pages (Information Only)

Unit 1 TS Bases Page B 2-1 Unit 2 TS Bases Page B 2-1

2.1 SAFETY LIMITS BASES

2.0 INTRODUCTION

The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that more than 99.9% of the fuel rods avoid transition boiling. Meeting the Safety Limit can be demonstrated by analysis that confirms less than 0.1% of fuel rods in the core are susceptible to transition boiling or by demonstrating that the MCPR is not less than the values specified in Specification 2.1.2 for two recirculation loop operation and for single recirculation loop operation. Less than 0.1% of fuel rods in transition boiling and MCPR greater than the values specified for two recirculation loop operation and for single recirculation loop operation represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER, Low Pressure or Low Flow 700 psia The use of the (GEXL) correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 103 lb/h, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 103 lb/h. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly criti-cal power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

700 psia LIMERICK - UNIT 1 B 2-1 Amendment No. 7, 30, 111, 127, 156 ECR 00-00209, ECR 01-00055, 170, 183 Associated with Amendment No. 206, ECR 11-00092

2.1 SAFETY LIMITS BASES

2.0 INTRODUCTION

The fuel cladding, reactor pressure vessel and primary system piping are the principle barriers to the release of radioactive materials to the environs.

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that more than 99.9% of the fuel rods avoid transition boiling. Meeting the Safety Limit can be demonstrated by analysis that confirms less than 0.1% of fuel rods in the core are susceptible to transition boiling or by demonstrating that the MCPR is not less than the values specified in Specification 2.1.2 for two recirculation loop operation and for single recirculation loop operation. Less than 0.1% of fuel rods in transition boiling and MCPR greater than the values specified for two recirculation loop operation and for single recirculation loop operation represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER, Low Pressure or Low Flow 700 psia The use of the (GEXL) correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 103 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 103 lb/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

700 psia LIMERICK - UNIT 2 B 2-1 Amendment No. 14, 83, 87, 97, 114, 127, 162, ECR LG 12-00035

ATTACHMENT 4 License Amendment Request Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Response to Draft Request for Additional Information Proposed Revision to Technical Specifications in Response to GE Energy - Nuclear 10 CFR Part 21 Safety Communication SC05-03 Loop Uncertainty Calculation Ll-00032, "LU Calculation for PT-001-2N076C"

CC*AA-309-1001 Rev. 8, Page 30 of 67 IECR LG 15-00017 Rev. o Attachment 7 Page 1 of 14 ATTACHMENT 1 Design Analysis Cover Sheet p aqe 1 Fo II owe db>Y 1A Design Analysis I Last Pa~e No*.

• Page 21 Min9r l8l

---

~*~***~-~******~

Analysis No.: ' Ll-00032 Revision:' OA Major D

Title:

3 LU .Calculation for PT-OQ1,.2N076C EC/ECR No.:* LG 15'-00017 Revision:~ 0 Station(s): 7 Limerick *Component{s ): .,.

Unit No.:* 1&2 PT-001-1N076A PT-001-2N076A Discipline:

• Electrical PT-001 "1 N076B PT-001-2N076B Descrip. Code/Keyword: 1c NIA PT-001-1N076C PT-001-2N076C Safety/QA Class: " Safety Related PT-001-1N0760 PT-001-2N0760 System Code: *? 001 Structure: *~

CONTROLLED DOCUMENT REFERENCES *$

Document No;: FromfTo Document No.: FromfTo UFSAR Section 7.3.1.1.2.4.5 Ta Tech Specs Table 3.3.2-2 Item 1.c To DBD L-S-16 Ta Is this Design Analysis Safeguards Information? ** YesO No f.8) If yes, see SY-AA-101-106 Does this Design Analysis contain Unverified Assumptions? " YesO No f.8) If yes, ATl/AR#:

This Design Analysis SUPERCEDES:,. NIA In its entirety.

Description of Revision (list changed pages when all pages of original analysis were not changed): **

This revision changes the Analytical Limit from 720 to 805 psig based on new references 4.17 and 4.18. The new Analytical Limit results in changes to the Allowable Value, Nominal Trip Setpoint. Actual Trip Setpoint. and Acceptable Limits as determined in section 7.7, 7.5, 7.6, and 7.8 respectively: Changed pages include; 2- Updated section 1.3 to reflect new allowable value, setpoint, and TOOi as basis.

4- Updated section 2.2. 1 to reflect new analytical limit value and TODI as basis ..

5- Updated section 2.2.5 and 2:2.6 to reflect new pressure margin values. and updated section 3.1.1 to delete the assumption that is no longer valid ..

6- Updated document revision numbers in sections 4.1 thru 4.5, and deleted references 4.6 and 4.7.

7- Updated section 4.9 with new revision number, deleted historical references from sections 4. 1O and 4.11, and added new references 4.17 and 4.18.

16 - Updated Limit and NTSP values in section 7.5.

17 - Updated margin and pressure values in sections 7.6, 7.7, and 7.8.

19 - Marked Attachment 2 (calculation results) for replacement by new llSCP calculation results.

21 - Updated Setpt, Allw, and Analytical/Proc Lmt values iri Attachment 4.

The use of a minor revision was approved by the Ray George (LEDE branch manager) on behalf of the SMDE on 9/28/15.

IECR LG 15-00017 Rev. 0 Attachment 7 Page 2of14 I CC-AA-309-1001 Rev. 8, Page 30 of 67*

p aae 1AFo II owe db1y 18 Preparer: 2'! Don Hoolahan

• Prim Name a.J/1:

~ '!I/~ ,/L 1

Sinn Name 9/28/2015 Date Method of Review:"' Detailed Review [] Alternate Calculations (attached) 0 Testing D Reviewer: n Mike Lazar.

Prin1Name v: p: 11/.,1/h. ~ M1k_*~

Sign Name '

l.li 2Bt<(

q /3.:>

Date

/15-Review Notes: ,, Independent review Kl Peer review 0 Calculation has been independently. reviewed per CC-AA-309 and CC-AA-309-1001. All comments have been satisfactorily incorporated.

(F0t External Analyses Only).

c:rl~"!J~

~~~~

External Approver: ,. John Pelliccone Print Name Obie Exelon Reviewer: ~ Ai...J\'\.\.t>~~ ~~~ GC't- -=f--,iogame 10/1 /16

!U Print Name Oil le Independent 3rd Party Review Reqd? ,,. YesO Exelon Approver: 17

\<ai~~ T~"~~~ *~ 10/ ,/ 1"r Date

Exelon Confidential/Proprietary CC-AA-103-1003 IECR LG 15-00017 Rev. 0 Attachment 7 Page 3 of 14 Revision 11 Page 7of11 ATTACHMENT 2 Owner's Acceptance Review Checklist for External Design Analyses Page 1B Followed by 1G Design Analysis No.:_L ........\_-....00 .......2......._ _ _ _ _ _ _ Rev:

..........0...~ Qf:i.;

No Question Instructions and Guidance Yes I No IN/A t Do assumptions have All Assumptions should be stated in clear terms with enough D D [gj sufficient documented justification to confirm that the assumption is conservative.

rationale?

For example, 1) the exact value of a particular parameter may not be known or that parameter may be known to vary over the range of conditions covered by the Calc_ulation. It is appropriate to represent or bound the parameter with an assumed value. 2) The predicted performance of a specific piece of equipment in lieu of actual test data. It is appropriate to use the documented opinion/position ofa recognized expert on that equipmentto represent predicted equipment performance.

Consideration should also be given as to any qualification testing that may be needed to validate the Assumptions. Ask yourself, would you provide more justification if you were .

performing this analysis? If yes, the rationale is likely incomplete, Are assumptions Ensure the documentation for source and rationale for the D D l&J 2 compatible with the assumption supports the way the plant is currently or will be way the plant is operated post change and they are not in conflict with any operated and with the design parameters. If the Analysis purpose is to establish a licensing basis? new licensing basis, this question can be answered yes, if the assumption suooorts that new basis.

3 Do all unverified If there are unverified assumptions without a tracking D D !&l assumptions have a mechanism indicated, then create the tracking item either tracking and closure through an ATI or a work order attached to the implementing mechanism in place? WO. Due dates for these actions need to support verification prior to the analysis becoming operational or the resultant plant change beinq op authorized.

4 Do the design inputs The origin of the input, or the source should be identified and !XI D D have sufficient be readily retrievable within Exelon's documentation system.

rationale? If not, then the source should be attached to the analysis. Ask yourself, would you provide more justification if you were performing this analysis? If yes, the rationale is likely incomolete.

5 Are design inputs correct and reasonable The expectation is that an Exelon Engineer should be able to I&! D D clearly understand which input parameters are critical to, the with critical parameters outcome of the analysis. That is, what is the impact of a identified, if change in the parameter to the results of the analysis? If the aooropriate? impact is large, then that parameter is critical.

6 Are design inputs Ensure the documentation for source and rationale for the jg! D D compatible with the inputs supports the way the plant is currently or will be way the plant is operated post change and they are not in conflict with any operated and with the design parameters.

licensinq basis?

Exelon Confidential/Proprietary CC-AA-103-1003 Revision 11 jECR LG 15-00017 Rev. 0 Attachment 7 Page 4of14 Page 8of11 ATTACHMENT 2 Owner's Acceptance Review Checklist for External Design Analyses Page 1C Followed by 10 Design Analysis No.:_L.......,\_-_.O......,QQ

.....2.-..*---...-------Rev: CA No Question Instructions and Guidance Yes I No IN/A 7 Are Engineering See Section 2.13 in CC-AA-309 for the attributes. toat are 0 0 IZl Judgments clearly sufficient to justify Engineering Judgment. Ask yourself, documented and would you provide more justification if you were performing justified? this analysis? If yes, the rationale is likely incomplete.

8 Are Engineering Ensure the justification for the engineering judgment 0 0 jg!

Judgments compatible. supports the way the plant is currently or will be operated with the way the plant is post change and is not in conflict with any design operated and with the parameters~ If the Analysis purpose is to establish a new licensing basis? licensing basis, then this question can be answered yes, if the judoment suooorts that new basis.

9 Do the results and Why was the analysis being performed? Does the stated 181 D D conclusions satisfy the purpose match the expectation from Exelon on the proposed purpose and objective of application of the results? If yes, th~n the analysis meets the Desion Analysis? the needs of the contract..

10 Are the results*and Make sure that the results support the UFSAR defined [XI D D conclusions compatible system design and operating conditions, or they support a with the way the plant is. proposed change to those conditions. If the analysis operated and with the supports a change, are all of the. other changing documents licensino basis? included on the cover sheet as impacted documents?

11 Have any limitations on Does the analysis support a temporary condition or I&! D D the use of the results procedure change? Make sure that any other documents been identified and needing to be updated are included and clearly delineated in transmitted to the the design analysis. Make sure that the cover sheet appropriate includes the other documents where the results of this.

oroanizations? analysis provide the input 12 Have margin impacts Make sure that the impacts to margin are clearly shown ~ D D been identified and within the body of the analysis. If the: analysis results in docum.ented reduced margins ensure thatthis has been appropriately appropriately for any. dispositioned in the EC being used to issue .the analysis.

negative impacts (Reference ER-AA-2007)?

13 Does the Design Are there sufficient documents included to support the ~ D D Analysis include the sources of input, and other reference material that is not applicable design basis readily retrievable in Exelon controlled Documents?

documentation?

14 Have all affected design Determine if sufficient searches have been performed to ~ D D analyses been identify any related analyses that need to be revised along documented on the with the base analysis. It may be necessary to perform Affected Documents List some basic searches to validate this.

(AOL) for the associated Confiauration Change?

15 Do the sources of inputs Compare any referenced codes and standards to the current IX] D D and analysis design basis and ensure that.any differences are reconciled.

methodology used meet If the input sources or analysis methodology are based on committed technical and an out-of-date methodology or code, additional reconciliation regulatory may be required if the site has since committed to a n:iore reauirements? recent code

Exelon Confidential/Proprietary CC-AA-103-1003 IECR LG 15-00017 Rev. O Attachment 7 Page 5of14 Revision.1.1 Page 9of11 ATTACHMENT 2 Owner's Acceptance Review Checklist for External Design Analyses Page 1D Followed by 2 Design Analysis No.: L\ - ooo "!:>2. Rev: oA No Question Instructions and Guidance Yes I No I NIA.

1q Have vendor supporting Based on the risk assessment performed during the pre-job l&I D D technical documents brief for the analysis (per HU-AA-1212), ensure that

• and references sufficient reviews of any supporting documents not provided (including GE DRFs) with the final analysis are. performed~

been reviewed when necessary?

17 Do operational limits Ensure the Tech Specs. Operating Procedures, etc. contain 00 u u support assumptions operational limits that support the analysis assumptions and and inputs? inputs.

Create an SFMS entry as required by CG-AA-4008~ SFMS Number: .1:\-'5\ 49?

jECR LG 15-00017 Rev. 0 Attachment 7 Page 6 of14 LU CALCULATION FOR Cale No LI-00032 Rev OA QA tJuclear PT-001-2N076C 01 Page 002 of 029 Group Ori DOCTYPE.: 000 Rev.Y-...w.~~-AioJ.-~~~~~~-1.1<~'4--Y-.++~+>:J.<j...

AprA~-tr~"!'tb~-fto't---~~~~~-tt.:tt<~-i:I'-++~~~

1.0 PURPOSE

\_Update header with Orig.

Rev. Apr. and dates.

This section includes the Objective, Limitations, Conclusions, and the Applicability Statement of this caicula ti on.

1.1 Objective The. objective of this calculation is to determine the Nominal Trip Setpoint (NTSP), Actual Trip Setpoint (ATSP) and the Allowable Value (AV) for the Main Steam Line Low Pressure Isolation Actuation Instrumentation as 'described in the Limerick Unit 2 Technical Specifications Table 3.3.2-2, Item 1.c (Ref. 4.~). Th.is calcl,llation analyzes the PT-001-2N076C in*struznentation loop. This calculation was performed utilizing nqrmal environmental conditions (see Sectidn 2~2.3),

The normal NTSP, ATS!? and AV results .of this calculation are documented in Section 7.

Results of this "base calculation" ar.e also applicable to the loops listed in Section 1.4. ..

1.2 Limitations The Max and Min Acceptable Limits calculated in Section 7.8 are not authorized for use in the PECo maintenance progirun by this revision of the calculation.

This calculation is run for a normal environment and does not account for any uncertainties associated with accident scenarios {see Section 2.2.3).

The. appropriate use of this calculation to support design or

.Station activities, other than those specified in Section 1.1 of this calc* ponsibility of the user.

1.3 Conclusions SIG was calculated d includes opercit;:ional in Section 7.7 OA 840.oo___,/ is based on TODI ES1400026 (Ref. 4.17) and General Electric Safety Concern SC05-03 (Ref 4.18).

IECR LG 15-00017 Rev. 0 Attachment 7 Page 7of14 Nuclear LU. CALCULATION FOR PT-001-'2N076C 01 Cale No LI-00032 Page 004 of 029 Rev -tHT OA oA Group Orig. HUMPHREYS GD Date 07 /12/94 DOCTYPE: 000 Rev. WHITE AJ Date 07 /12/94 Apr. GEORGE RT Date 07/13/94 installations which result in a head correction of +5.6 PSIG for both units. This was documented by the issuance of IISCP Anomaly No, 114, "Head Corrections for PT-00l~i(2)N076A/B/CJD" (Ref- 4.14) which describes the iss1.1.ance. of Action Requests A0851879 (Ref, 4 .15) artd A0852289 (Ref. 4 .16) (Type CM-:-NCR) for correcting these discrepancies. The.se discrepancies have no affect on this calculation as the head correction pertains only to J:;he scaling of the transmitter. The scaling of PT-001-2N076C was done using +5.6 PSIG in accordance with the field iristallations.

2. 0 DESIGN *BASIS This section includes the Tech.,ical Background and Design Input information relevant to the ca,lculation.

2.1 Technical Background Low steam pressure at the turbine inlet While the reactor is operating could indicate a malfunction of the steam pressure controller in which the turbine control valves or turbine bypass valves become fully open and cause rapid depressurization of the reactor vessel. Instrumentation is installed to monitor the steam line pressure in order to mitigate the consequences of this type of occurrence. The signals generated.by this monitoring instrumentation input into the NSSSS isolation logic. which automatically closes the Main Steam Isolation Valves (MSIVs) whenever the Mode Sw' s a 1 1 e e th loop is analyzed by this calculation.

2.2

2.

• TOOi ES 1400026 (Ref. 4.17) and General Input aos.oo_§_!~~!ric Safetv Concern SC05-03 (Ref. 4.18).

An Analytical/Process Limit o?-~ PSIG has been /.

7 utilized for this calculation based on eh¢ &8,,._ -

"i'udsbze 'Fluot:tlc ?'11:~t (MSl:U &:ast CJe-sci. OA use~ y:xameter speo;£2ee i~she--8i~l G.9~ 3 (~~n...-

4+-l~

~::120~~~~

.@S;.~~~=~~~

~di . ~

• lil+.-

T is calcu ation inc u es any applica le System Rerate Design/Operating Conditions and Impacts as a result of the Power Rerate analyses per the guidelines. contained in Specification NE-177 (Ref.

4 .12) .

2.. 2. 3 This calculation was performed under normal

IECR LG 15-00017 Rev. 0 Attachment 7 Page 8of14 LU CALCULATION FOR Cale No LI-00032 Rev ~

OA Nuclear PT-001-,2N076C 01 Page 005 of 029 OA Group Orig. HUMPHREYS GD Date 07/12/94 DOCTYPE: 000 Rev. WHITE AJ Date 07/12/94 Apr. GEORGE RT Date 07 /13/94 environmental conditions based on the design information contained in Section 15 .. 1. 3 of the Limerick Generating Station Updated F.inal S~fety Analysis Report (UFSAR) (Ref. 4.1). UFSAR Section 15 .1. 3 indicates that the design bases event for the isolation of the ma*in steam line as a result of low steam line pressure is a failure of the main turbine pressure regulator. This failure will resuJ,t in no release of steam to the Turbine Enclosure environment. Therefore PT-001-2N076C will not be subjected to any harsh environment*

effects when accomplishing its intended safety function.

2.2.4 Process consideration has been included to provide support for additional operational flexibility. This process consideration appears within the calculation as consideration Sl. This consideration is based on engineering judgement and reflects an amount approximately twice the accuracy of the transmitter plus an: additional amount which resu1ts in a conservatively rounded 2.2.5 The delta between the Allowable Value (AV) and the Actual Trip Set Point (ATSP) within this calculation i~~ PSIG which satisfies the IISCP 19 Leave Alone Zone Requirement to provide at least one LAZ between AV and ATSP.

l:'.:--17 .345 2.2.6 Additional margin of ~ PSIG was added to this QA calculation to support the current station setpoint Of thi~~ PSIG, 8. 406 PSIG is "assigned 17 *345 margin"* used to support the IISCP LAZ requirements as discussed in Section 2.2.5. The remaining 8.939 __:.7~PSIG is "unassigned margin" which is considered additional conservatism that may be utilized in future analyses.

2.2-7 All other design inputs* to this calculation are documented on the Supporting Data Sheet Attachments.

3..0 ASSUMPTIONS OA None.

.1'

jECR LG 15-00017 Rev. 0 Attachment 7 Page 9of14j LU CALCULATION FOR Cale No LI-00032 Rev 0- OA OA Nuclear PT-001-2N076C 01 Page 006 of 029 Group Orig. HUMPHREYS GD Date 07/12/94 DOCTYPE: 000 Rev. WHITE AJ Date 07/1:4/94 Apr. GEORGE RT Date 07 /13/94 jpglqdeg Bpterpr Rerate ipfQrPJatipp\ aptj j5

~tea ii:P.&eetiea 2 2 l e§ t;;~is c:a:l c*1obatieia QA

~ basis a:f this assl:h~ia;:;; is IISP Pzaj ee'e

~s P!W:e:M p l?!i<OO;, Od.62 (l'lef 4: 231 3.2 Assumptions Requiring Confirmation 3.2.1 None Current revision is 16 dated September 2012.

4.0 R.EFERENCES 4.1 Limerick Generating Station Up ted Final Safety Analysis Report (UFSAR), Revisi n ~(dated~)-

k") QA

-Section 7.. 3. 1. L 2. 4. 5 "PCRVICS -

Pressure*

-Sectiori 15 . . 3 Pressure Regulator F (Design Bas' reference). . 52

. ~

4.2 Limerick Gen rating ~atton Technica Unit 2, ti.men ent ~ Table 3. 3. 2-2 02/17/94) (op rations and Surveilla reference).

Limerick Generating. Station Units 1&2 System 4.3 Design Baseline. Document (DBD) L-S-16, Revis n~.k:,;

Section 3.2.9, Reactor Instrumentation Syst (Design Basis reference).

4.4 the 4.5

...._,..._,..._....:'G-"-~~~~~-'(.~.~ew~ou::.;::m..-"il.ll""-d.fll.CJin:.!~~~l.l.4ilA--'

PECo procedure IC-11-50014 for PT-001-2N076C dated 06/28/88, PIS-001-2N676C dated 01/16/87. Master Loop Sheet for PT-001-2N076C dated 06/28/88 {Applicability reference).

!ECR LG 15-00017 Rev. 0 Attachment 7 Page 10of14 I LU CALCULATION FOR Cale No LI-00032 Rev -etr Nuclear PT-00l-2N076C 01 Page 007 of 029 OA I OA Group Orig. HUMPHREY$ GD Date 07/12/94 DOCTYPE: 000 Rev. WHITE AJ Date 07 /12/94 Apr. G 0 GE RT Date 07/13/94 4.9 Calculation M-75-12, Revis Building OA Cooling Load" (Location Da 4.10 Philadelphia Electric Company Letter from G.C. Storey to G *. R. Hull General Electric Company, subjec;:t "Final OPL-3 for Limerick ARTS/MELLLAAnalysis". This document contains r-v-~~K.J""'Cf--lV..C~~~-'1~~'7--25µ.-.,~:.i;si.l.l,;1 e . QPL-3 Forms that~,~-~~

c *n it*o s Dae 3 9 9 *

~~~~~~~~~~~~~~~ (Rev. O historical reference)

OA

4. , Revision uclear e y e a e Specification for Limerick Generating Station Units 1&2 Power Rerate Operating Conditie>ns (!'ower Rerate Information reference) .

4.13 IISCP Project Letter to File M-P-PEOOl-0152 - Utilization of OPL-3 (Assumptions reference) .

4.14 IISCP Project Anomaly No. 114, Head Corrections for PT-001-1 (2)N076A/B/C/D .(Applicability reference) *.

4.15 Action Request (Type CM NCR) A0851879 - Head Correction for PT-001-1N076A/B/C/D (Applicability reference).

4.16 Action Request (TyPe CM NCR) A0852289 - Head Correction for PT-001-2N076A/B/C/D (Applicability reference).

5.0 ATTACHMENTS 5.1 See Supporting Data Sheet Attachments located within this calculation.

6.0 ANALYSIS 6.1 Loop Effects 6.1.1 Loop ID No. PT-001-2N076C Config 01 6.1.2 Loop Function MAIN STEAM LINE c LOW PRESSURE - NS4 ISOLATION 6.1.3 Configuration Description MN STN LN C PRESS INDICATION 6.1.4 Loop Instrument List Add new references shown below. ---***-****---

4.17 Transmittal of Design Information (TOOi) ES1400026, Rev. 0, "Low Pressure OA Isolation Setpoirit for the Limerick Station Loop Uncertainty Calculation" 4.18 General Electric Safety Concern SC05-03, dat~d 3/29/2005, "Potential to exceed Low Pressure Technical S ecification Safety Limit"

IECR LG 15-00017 Rev. 0 Attachment 7 Page 11of14 j Nuclear LU CALCULATION FOR PT-001-2N076C 01 Cale No LI'-00032 Page 016 of 029 Rev -&a-* OA joA Group Orig. HUMPHREYS GD Date 07/12/94 DOCTYPE: 000 Rev. WHITE AJ Date- 07/12/94 Apr. GEOR(.;E RT Date 07/13/94

7. 2* DL DL = DE + DT where:

DE E 02 DT - ~ DTE 1 DL = 0.00006 7.3 CL CL =V + M where:

v = E (setting tolerance)*

M E MTE 2 CL = 0.00006 7.4 TLU (Positive)TLUp = [IR + PMAp + PEAp + PCp + PMAo + PEAo + PCo +

V(AL +CL+ DL + PMAr + PEAr + PCr)] *Loop span (Negative)TLUn = (- PMAn - PEAn - PCn - PMAo - PEAo - PCo +

-v (AL + CL + DL + .PMAr + PEAr + PCr)]

• Loop span All other variables as previously defined.

TLUp = 21. 47 PSIG TLUn -21. 47 PSIG 7.5 NTSP (increasing) NTSP = limit + (- PMAn - PEAn - PCn - PMAo ~ PEAo PCo + (1.645 /sigma ) * -v(AL +CL + DL +

PMAr + PEAr + PCr)J *Loop span (decreasing) NTSP = limit + [IR + PMAp + PEAp + PCp + PMAo +

PEAo + PCo + (i.645 /sigma)

• v{AL +CL.+

DL + PMAr *t PEAr + PCr)]

• Loop span where:

limit = loop analytical or process limit limit = ?il~ PSIG

~

where: '~805.00 OA sigma 2 NTSP  :;tH, $6. PSIG

/~

La22.66

IECR LG 15-00017 Rev. 0 Attachment 7 Page 12of14 I Nuclear LU CALCULATION FOR PT-001-2N076C 01 Cale No LI-00032 Page 017 of 029 Rev -&G-OA joA Group Orig. HUMPHREYS GD Date 07/12/94 DOCTYPE: 000 Rev. WHITE AJ Date 07 /12/94 Apr. GEORGE RT Date 07/13/94 7.6 ATSP

{increasing) ATSP - NTsp* + margin (decreasing) ATSP =.NTSP - margin where:

margin additional margin suppl ed by calculation originator margin ~"E------17.345

~ 840.00 ATSP 7.7 A1lowab Value (Decreasing) AV = limit + [IR + PMAp + PEAp + PCp + PMAo + PEAo +

v PCo + (1. 645 /sigma ) * (AL + CL + PMAr +

PEAr + PCrlJ *Loop span (Increasing). AV limit + [- PMAn - .PEAn - PCn - PMAo - PEAo -

PCo + (1. 645 /sigma ) * --.J (AL + CL + PMAr +

l?EAr + PCr)J

• Loop span AV OA 7.8 Accep Max Min *,.-....,,-,,.-,,-..,.---..---..r---..,--..r-"~

All ther variables as previously fined 765. @?8..{P-S_I_G--B 49 .0 7 B OA

'.HG. g;;n.. PSIG

\:::~--830.923

jECR LG 15-00017 Rev. 0 Attachment 7 Page 13of14 Nuclear Group LU CALCULATION FOR PT-001-2N076C 01 Cale No LI-00032 Page 019 of 029 Orig. HUMPHREYS GD Rev Date

-fttj-OA 07/12/94 I OA DOCTYPE: 000 Rev. WHITE A.J Date 07112/94.

Apr. GEORGE RT Date 07/13194 ATTACHMENT 2:

Device Accuracy Temperature Humidity Tol. P.wr Supp Norm Accid Accid PT-001-2N076C T 0.00500 0.00564 0.00000 0.00000 0.00500 0.00008 PIS-00l-'2N676C s o.ob2so o.boooo 0.00000 0.000-00 0.00250 0.60000 Device SPE Rad. M&TE Drift over Pres Seismic Ace id PT-00l-2N076C T 0.00000 0.00000 0.00500 0.00504 0.00000 0.00000 i?IS-001-2N676C s 0.00000 0.00000 0.00250 0,00000 0.00000 0.00000 Process Concerns: NORMAL ACCIDENT OA Positive Negative Offsetting Positive Negative Offsetting PMA 0.00000 0.00000 0.00000 0.00000 0.00000 0.00000 PEA 0.00000 0. 0000.0 0.00000 0.00000 0.00000 0.00000 IR 0.00000 Other 0.00000 0.00000 0.00000 0 .. 00000 0.00000 0.00000 Loop Results: NORMAL ACCIDENT TLU .. -21.4656 21.46566 -21.4656 21.46566 AL 0.00003 0.00003 Increasing Decreasing Increasing Decreasing NTSP .. NIA 737 ..6555 N/A 737.6555 AV* N/A 735.9990 N/A 735.9990 Ace Limits Min .. : N/A -17.3450 746. 9227 N/A 746.9227 Max*: N/A~ 765.0781 NIA 765.078:1.

ATSP.. N/A 756.0004 N/A 756.0004 Additional Margin: -'1:8.345U DL: 0.00006 CL: 0.00006

• These values are n \:-

\_See section 2.2.6.

\_-Replace with new llSCP software calculation results.

IECR LG 15-00017 Rev. O Attachment 7Page14of14 j Nuclear Group LU CALCULATION FOR PT-001-2N076C 01

(:ale No LI-00032 Page 021 of 029 Orig. HUMPHREYS GD Rev bate

-M- DA 07/12/94 I OA DOCTYPE: 000 Rev *. WHITE AJ Date 0.7 /12/94 Apr. GEORGE RT Date 07/13/94 ATTACHMENT 4: Loop Calibration Data Process Temperature Units Min 0.00 Max 0.00 Normal 0.00 Trip 0.00 Process Radiation Units Min O.OOOe+OOO Max O.OOOe+OOO Normal O.OOOe+OOO Trip O.OOOe+OOO Process Humidity Units Min 0.00 Max 0.00 Normal 0.00 Trip 0.00 0

Sigma Setpt:

Des/Sfty Lm,.

~ @@Units: PSIGRese.t:

Loop Settin. Tolerance

0. 00 Units

0.000

0. OOUnits: Allw:,

Calibration Frequency Loop Leave Alone Zone 93G....00Units PSIG 4'-\ 731

6.708 Loop Cal Ac : 0.000 Analytical/Pree Lmt: ?29.99 UnitsPSIG \

' / ' 805.00 OA Originatorf'llJl<PHREYS GD 05/09/9~iewer WHITE AJ \ 06/01/9 840.00 ( . 821.00 Replace 720.00 with 805.00 (New Analytical Limit from TOOi).