ML16110A392

Limerick, Units 1 and 2 - 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

Exelon Generation 10

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. 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 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).

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.

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 licensee's 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 licensee's 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.

Attachment

2 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 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

Attachment

4 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.

2-1 3/4 3-18 2-1 3/4 3-18

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 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:

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:

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 FUNCTIONTRIPSETPOINT VALUE 1.MAIN STEAM LINE ISOLATIONa.Reactor Vessel Water Level1)Low, Low - Level 2> - 38 inches* > - 45 inches 2)Low, Low, Low - Level 1> - 129 inches* > - 136 inches b.DELETEDDELETEDDELETEDc.Main Steam LinePressure - Low> 756 psig > 736 psig d.Main Steam LineFlow - High< 122.1 psid < 123 psid e.Condenser Vacuum - Low 10.5 psia>10.1 psia/ 10.9 psia f.Outboard MSIV RoomTemperature - High< 192°F < 200°F g.Turbine Enclosure - Main SteamLine Tunnel Temperature - High< 165°F < 175°F h. Manual InitiationN.A. N.A. 2.RHR SYSTEM SHUTDOWN COOLING MODE ISOLATIONa.Reactor Vessel Water LevelLow - Level 3> 12.5 inches* > 11.0 inches b.Reactor Vessel (RHR Cut-inPermissive) Pressure - High< 75 psig < 95 psig c. Manual InitiationN.A. N.A. LIMERICK - UNIT 2 3/4 3-18 Amendment No. 51, 52

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 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.

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 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 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.

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

p aqe D 3 7 1c NIA *? *$ Ta f.8) f.8) 6-aae 1y 2'! a.J/1:1 ,/L Date 0 n v: p: 11/.,1/h. l.li 2Bt<( q /3.:> /15-Kl 0 Only). Print Name Obie 10/1 /16 GC't--=f--,iogame YesO 17 10/ ,/ 1"r Date 00 .......... 0 ... ....... I t D [gj of D l&J 3 D D 5 I&! D 6 jg! D 2.-..*---...-------Rev: CA I 0 0 IZl 8 0 0 jg! 9 181 D D D D I&! D D D D 13 D D if D D D D L\ -ooo I I D D 00 Cale No LI-00032 Rev OA tJuclear Group LU CALCULATION FOR PT-001-2N076C 01 Page 002 of 029 Ori 1.0 DOCTYPE.: 000 PURPOSE This section Conclusions, caicula on. 1.1 Objective includes the Objective, Limitations, and the Applicability Statement of this 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. Th.is calcl,llation analyzes the PT-001-2N076C in*struznentation loop. This calculation was performed utilizing nqrmal environmental conditions (see Sectidn 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 ' QA OA Nuclear Group LU. CALCULATION FOR Cale No LI-00032 Rev -tHT OA PT-001-'2N076C 01 Page 004 of 029 Orig. HUMPHREYS GD Date 07 /12/94 DOCTYPE: 000 Rev. WHITE AJ Date /12/94 GEORGE 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 (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 2.2 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 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 s a 1 1 e e th loop is analyzed by this calculation. Input 7 2.

• An Analytical/Process Limit PSIG has been /. utilized for this calculation based on eh¢ &8,,._ -'Fluot:tlc y:xameter speo;£2ee 3 2 .. 2. 3 .

• 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) . This calculation was performed under normal oA OA Nuclear Group LU CALCULATION FOR Cale No LI-00032 Rev OA PT-001-,2N076C 01 Page 005 of 029 Orig. HUMPHREYS GD Date 07/12/94 DOCTYPE: 2.2.4 000 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 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. 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 19 calculation PSIG which satisfies the IISCP 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 calculation to support the current station setpoint 17 345 Of PSIG, 8. 406 PSIG is "assigned

• margin"* used to support the IISCP LAZ requirements as discussed in Section 2.2.5. The remaining 8.939 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 QA OA .1 LU CALCULATION FOR Cale No LI-00032 Rev OA Nuclear PT-001-2N076C 01 Page 006 of 029 Group Orig. HUMPHREYS GD Date 07/12/94 4.0 DOCTYPE: 000 Rev. WHITE Date Apr. GEORGE RT Date jpglqdeg Bpterpr Rerate ipfQrPJatipp\ aptj j5 2 2 l e§ c*1obatieia basis a:f this is IIS*P Pzaj ee'e P!W:e:M 4: 3.2 Assumptions Requiring Confirmation 3.2.1 None 07/1:4/94 07 /13/94 R.EFERENCES 4.1 Limerick Generating Station Up Analysis Report (UFSAR), Revisi ted Final Safety k") n -Section 7.. 3. 1. L 2. 4. 5 "PCRVICS -Pressure* -Sectiori 15 .. 3 Pressure Regulator F (Design Bas' reference). . 4.2 Limerick Gen rating Technica Unit 2, ti.men ent Table 3. 3. 2-2 02/17/94) (op rations and Surveilla 4.3 4.4 4.5 reference). Limerick Generating. Station Units 1&2 System Design Baseline. Document (DBD) L-S-16, Revis Section 3.2.9, Reactor Instrumentation Syst Basis reference). ! (Design the ...._,..._,. . 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). OA QA QA LU CALCULATION FOR Cale No LI-00032 Rev -etr OA Nuclear PT-00l-2N076C 01 Page 007 of 029 Group Orig. HUMPHREY$ 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 Cooling Load" (Location Da Building 4.10 Philadelphia Electric Company Letter from G.C. Storey to R. Hull General Electric Company, subjec;:t "Final OPL-3 for Limerick ARTS/MELLLAAnalysis". This document contains e . QPL-3 Forms c it*o Dae O 4. , Revision uclear e 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 CM NCR) A0851879 -Head Correction for PT-001-1N076A/B/C/D (Applicability reference). 4.16 Action Request (TyPe 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 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 I OA OA OA OA LU CALCULATION FOR Cale No LI'-00032 Rev OA Nuclear PT-001-2N076C 01 Page 016 of 029 Group Orig. HUMPHREYS GD Date 07/12/94 DOCTYPE: 000 Rev. WHITE AJ Date-07/12/94 Apr. Date 07/13/94 7. 2* DL DL DE + DT where: DE E 02 DT DTEDL 0.00006 7.3 CL CL + where: v = E (setting tolerance)* E 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 (AL CL DL .PMAr PEAr PCr)] Loop span All other variables as previously defined. TLUp TL Un 7.5 NTSP 21. 47 PSIG -21. 47 PSIG (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.+ where: limit limit = where: sigma NTSP DL PMAr *t PEAr PCr)] Loop span loop analytical PSIG 2 :;tH, $6. PSIG La22.66 or process limit joA OA LU CALCULATION FOR Cale No LI-00032 Rev OA Nuclear PT-001-2N076C 01 Page 017 of 029 Group Orig. HUMPHREYS GD Date 07/12/94 DOCTYPE: 000 Rev. WHITE AJ Date 07 /12/94 Apr. GEORGE Date 07/13/94 7.6 ATSP {increasing) ATSP -+ margin (decreasing) ATSP =.NTSP -margin where: margin margin ATSP additional margin suppl ed by calculation originator 7.7 A1lowab Value (Decreasing) AV limit [IR PMAp + PEAp + PCp PMAo + PEAo PCo /sigma ) v (AL CL PMAr PEAr PCrlJ *Loop span (Increasing). AV limit PMAn -.PEAn -PCn -PMAo -PEAo -PCo /sigma ) --.J (AL CL PMAr l?EAr + PCr)J Loop span AV 7.8 Accep Max Min All ther variables as previously 765. @?8..{P-S_I_G--B'.HG. g;;n .. PSIG fined joA OA OA LU CALCULATION FOR Cale No LI-00032 Rev OA Nuclear PT-001-2N076C 01 Page 019 of 029 Group Orig. HUMPHREYS GD Date 07/12/94 DOCTYPE: 000 Rev. WHITE 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 PIS-00l-'2N676C T s 0.00500 0.00564 0.00000 0.00000 0.00500 0.00008 o.ob2so o.boooo 0.00000 0.000-00 0.00250 0.60000 Device SPE Rad. M&TE Ace id Drift over Pres Seismic PT-00l-2N076C i?IS-001-2N676C s 0.00000 0.00000 0.00500 0.00000 0.00000 0.00250 0.00504 0.00000 0,00000 0.00000 0.00000 0.00000 Process Concerns: NORMAL Positive Negative Offsetting PMA 0.00000 0.00000 0.00000 PEA 0.00000 0. 0000.0 0.00000 IR Other 0.00000 0.00000 0.00000 Loop Results: NORMAL TLU .. AL NTSP .. AV* Ace Limits -21.4656 21.46566 0.00003 Increasing NIA N/A Decreasing 737 .. 6555 735.9990 Min .. : N/A -17.3450 746. 9227 ATSP.. N/A 756.0004 Additional Margin: -'1:8.345U DL: 0.00006 These values are n \:-ACCIDENT Positive Negative Offsetting 0.00000 0.00000 0.00000 0.00000 0.00000 0.00000 0.00000 0 .. 00000 0.00000 0.00000 ACCIDENT -21.4656 21.46566 0.00003 Increasing N/A N/A N/A NIA N/A CL: 0.00006 Decreasing 737.6555 735.9990 746.9227 765.078:1. 756.0004 765.0781 OA OA LU CALCULATION FOR (:ale No LI-00032 Rev -M-DA Nuclear PT-001-2N076C 01 Page 021 of 029 OA Group Orig. HUMPHREYS GD bate 07/12/94 DOCTYPE: 000 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 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: @@Units: PSIGRese.t: 0. OOUnits: Allw:, 93G....00Units PSIG Des/Sfty 0. 00 Units Calibration Frequency 731 Loop Settin. Tolerance : 0.000 Loop Leave Alone Zone : 6.708 Loop Cal Ac : 0.000 Analytical/Pree Lmt: ?29.99 UnitsPSIG \ 805.00 Originatorf'llJl<PHREYS GD WHITE AJ \ ( 06/01/9 OA