Text

Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent with TSTF-419-A (NL-13-1773)
	ML15092A856
Person / Time
Site:	Hatch
Issue date:	04/02/2015
From:	Pierce C; Southern Co, Southern Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-13-1773
	Download: ML15092A856 (146)
	v • d • e

Charles R. Pierce Regulatory Affairs Director Af'R 0 2 2015 Docket Nos.:

50-321 50-366 Southern Nuclear Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35201 Tel 205.992.7872 Fax 205.992.7601 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant SOUTHERN A COMPANY NL-13-1773 Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent with TSTF-419-A Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests an amendment to the Technical Specifications (TS) for the Edwin I. Hatch Nuclear Plant (HNP), Unit 1 and Unit 2.

The proposed amendment modifies Technical Specifications (TS) Section 1.0

("Definitions"), Limiting Conditions for Operation and Surveillance Requirement Applicability Section 3.4.9 ("RCS Pressure and Temperature (PIT) Limits"), and Section 5.0 ("Administrative Controls") to delete reference to the pressure and temperature curves, and to include reference to the Pressure and Temperature Limits Report (PTLR).

This change adopts the methodology of BWROG-TP-11-022-A Revision 1 (SIR-05-044, Revision 1-A), "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", dated June 2013, and of BWROG-TP-11-023-A, Revision 0 (0900876.401, Revision 0-A), "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations", dated May 2013, for preparation of the pressure and temperature curves, and incorporates the guidance of TSTF-419-A "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR". The HNP PTLRs have been developed based on the methodologies provided in BWROG-TP-11-022-A, Revision 1 and BWROG-TP-11-023-A, Revision 0, and based on the template provided in BWROG-TP-11-022-A, Revision 1.

The HNP PTLRs are provided as Enclosures 5 and 6.

Approval of the proposed amendment is requested by March 31, 2016. Once approved, the amendment shall be implemented within 90 days.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Georgia Official.

This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.

I U. S. Nuclear Regulatory Commission NL-1 3-1 773 Page 2 Mr. C. R. Pierce states he is Regulatory Affairs Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the facts set forth in this letter are true.

Respectfully submitted, C.t{S C. R. Pierce Regulatory Affairs Director CRP/RMJ

? =ubscribed Mfore me this _A day of /Jptk.Q Lc?t(ebr_

Notary Public My commission expires: /IJ J g/z O l1 r

Enclosures:

1. Description and Assessment

2. Proposed Technical Specification Changes (Mark-Up)

3. Revised Technical Specification Pages

'2015.

4. Proposed Technical Specification Bases Changes (Marked-Up)

5. Hatch Nuclear Plant, Unit 1 Pressure and Temperature Limits Report (PTLR) for 38 and 49.3 Effective Full-Power Years (EFPY)

6. Hatch Nuclear Plant, Unit 2 Pressure and Temperature Limits Report (PTLR) for 37 and 50.1 Effective Full-Power Years (EFPY) cc:

Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. D. R. Vineyard, Vice President-Hatch Mr. D. R. Madison, Vice President-Fleet Operations Mr. M. D. Meier, Vice President-Regulatory Affairs Mr. B. J. Adams, Vice President-Engineering Mr. G. L. Johnson, Regulatory Affairs Manager - Hatch RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. R. E. Martin, NRR Senior Project Manager - Hatch Mr. D. H. Hardage, Senior Resident Inspector-Hatch State of Georgia Mr. J. H. Turner, Environmental Director Protection Division

}

Edwin I. Hatch Nuclear Plant Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent with TSTF-419-A Description and Assessment to NL-1 3-1 773 Description and Assessment 1.0 Description The proposed amendment modifies Edwin I. Hatch Nuclear Plant (HNP) Unit 1 and Unit 2 Technical Specifications (TS) Section 1.0 ("Definitions"), Limiting Conditions for Operation and Surveillance Requirement Applicability Section 3.4.9 ("RCS Pressure and Temperature (PfT)

Limits"}, and Section 5.0 ("Administrative Controls") to delete reference to the pressure and temperature curves, and to include reference to the Pressure and Temperature Limits Report (PTLR). This change adopts the methodology of BWROG-TP-1 1-022-A, Revision 1 (SIR 044, Revision 1 -A), "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", dated June 201 3, and of BWROG-TP-1 1-023-A, Revision 0 (0900876.401, Revision 0-A), "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations", dated May 2013, for preparation of the pressure and temperature curves, and incorporates the guidance of TSTF-419-A, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR". The HNP PTLRs have been developed based on the methodologies provided in BWROG-TP-1 1-022-A, Revision 1 and BWROG-TP-1 1-023-A, Revision 0, and based on the template provided in BWROG-TP-1 1-022-A, Revision 1. The HNP PTLRs are provided as Enclosures 5 and 6.

2.0 Proposed Change The proposed change modifies:

1 ) TS Section 1.0 to add a definition of the "Pressure and Temperature Limits Report".

2) TS Limiting Conditions for Operation and Surveillance Requirement, Section 3.4.9 ("RCS Pressure and Temperature (PfT) Limits").

3) TS Section 5.6.7 is being added to include wording from TSTF-419-A concerning: 1 ) the individual TSs that address reactor coolant system pressure-temperature (P-T) limits; 2) the NRC approved topical reports that document PTLR methodologies; and 3) the requirements for providing a revised PTLR to the NRC.

The copies of the TS Bases pages are provided for NRC information only. The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control Program.

3.0 Background

By letter dated November 1 7, 201 1 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML113260534}, the Boiling Water Reactor Owners' Group (BWROG) submitted Licensing Topical Report (L TR) BWROG-TP-1 1-022, Revision 1, "Pressure Temperature Limits Report Methodology for Boiling Water Reactors," to the U.S. Nuclear Regulatory Commission (NRC) staff. By letter dated May 1 6, 201 3 (Reference 1 ), the NRC staff found that TR BWROG-TP-1 1-022, Revision 1, "is acceptable for referencing in licensing applications for boiling water reactors to the extent specified and under the limitations delineated in the TR and in the enclosed final SE. The final SE defines the basis for our acceptance of the TR." This Safety Evaluation Report (SER) permits Boiling Water Reactor (BWR) licensees to relocate their P-T curves from the facility TS to a Pressure and Temperature Limits Report (PTLR) utilizing the guidance in TS Task Force (TSTF) Traveler No. 419-A. The BWROG issued the final report (-A) on September 4, 201 3 (Reference 2) which contains the E1-1 to NL-1 3-1 773 Description and Assessment final SER, along with the NRC requests for additional information (RAis), and the BWROG's responses to the NRC RAis.

LTR BWROG-TP-1 1-022, Revision 1, is a revision of LTR SIR-05-044-A with the same title, and contains minor modifications of the P-T limit methodology. As was the case for L TR SIR 044-A, the BWROG provided this LTR to support the BWR licensees to relocate their P-T curves and associated numerical values (such as heatup I cooldown rates) from facility TS to a PTLR, a licensee-controlled document, using the guidelines provided in Generic Letter (GL) 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits."

The NRC SER contained one condition for future potential applicants to address in their application of this L TR to their plant-specific P-T limits or PTLR submittal:

Each applicant referencing this L TR shall confirm that, in addition to the requirements in the ASME Code,Section XI, Appendix G, the lowest service temperatures for all ferritic RCPB components that are not part of the RPV, are below the lowest operating temperature in the proposed P-T limits.

This condition is discussed in the Technical Analysis section of this LAR.

By letter dated November 1 7, 201 1 (ADAMS Accession No. ML113250288) the BWROG submitted LTR BWROG-TP-1 1-023, Revision 0, "Linear Elastic Fracture Mechanics (LEFM)

Evaluation of General Electric Boiling Water Reactor Water Level Instrument (WLI) Nozzles for Pressure-Temperature (P-T) Curve Evaluations", to the U.S. NRC staff for review and acceptance for referencing in subsequent licensing actions. By letter dated March 1 4, 201 3, (Reference 3) the NRC staff has found that TR BWROG-TP-1 1 -023, Revision 0, "is acceptable for referencing in licensing applications for boiling water reactors to the extent specified and under the limitations delineated in the TR and in the enclosed final SE. The final SE defines the basis for our acceptance of the TR." This SER permits BWR licensees to use BWROG-TP-1 1-023, Revision 0 as an acceptable methodology to obtain plant-specific stress intensity factors for an internal pressure load case and a 1 00 aF/hr thermal ramp load case for use in developing plant-specific P-T limit curves for reactor pressure vessel (RPV) water level instrument (WLI) nozzles. Since the analyses assumed that Alloy 600 material was used for the weld metal for all nozzles, the BWROG revised the L TR to limit the application of this L TR to only WLI nozzle configuration/design using Alloy 600 material for the weld metal. The NRC SER did not contain any additional conditions or limitations with regard to this L TR. The BWROG issued the final report (-A) on June 28, 201 3 (Reference 4) which contains the final SER, along with the NRC requests for additional information (RAis) and the BWROG's responses to the NRC RAis.

TS Task Force (TSTF) Traveler No. 419 (Reference 6) amended the Standard TS (NUREGs-1430, -1431, -1433, and -1434) to: ( 1 ) delete references to the TS LCO specifications for the P T limits in the TS definition for the PTLR, and (2) revise STS 5.6.6 to identify, by number and title, NRC-approved topical reports that document PTLR methodologies, or the NRC safety evaluation for a plant-specific methodology by NRC letter and date. A requirement was added to the reviewers note to specify the complete citation of PTLR methodology in the plant specific PTLR, including the report number, title, revision, date, and any supplements. Only the figures, values, and parameters associated with the P-T limits are relocated to the PTLR. The TSTF also specified that the methodology, and any subsequent changes, must be reviewed and E1-2 to NL-1 3-1 773 Description and Assessment approved by the NRC. In this case, the methodology was approved in the Reference 1 and 3 letters.

The HNP Unit 1 and Unit 2 PTLRs, based on the methodology provided in BWROG-TP-1 1-022-A, Revision 1 and BWROG-TP-1 1-023-A, Revision 0, and based on the template provided in BWROG-TP-1 1-022-A, Revision 1, are being submitted for review. The purpose of the HNP PTLRs is to present operating limits related to Reactor Coolant System (RCS) pressure versus temperature limits during heatup, cooldown and hydrostatic/class 1 leak testing.

4.0 Technical Analysis NRC GL 96-03 (Reference 5) allows plants to relocate their pressure-temperature (P-T) curves and numerical values of other P-T limits (such as heatup/cooldown rates) from the plant TS to a PTLR, which is a licensee-controlled document. As stated in GL 96-03, during the development of the improved standard technical specifications (STS), a change was proposed to relocate the P-T limits currently contained in the plant TS to a PTLR. As one of the improvements to the STS, the NRC staff agreed with the industry that the curves may be relocated outside the plant TS to a PTLR so that the licensee could maintain these limits efficiently. One of the prerequisites for having the PTLR option is that all of the methods used to develop the P-T curves and limits be NRC approved, and that the associated L TR for such approval is referenced in the plant TS. Based on this prerequisite, the purpose of the BWROG L TRs is to provide BWRs with NRC-approved L TRs that can be referenced in plant TS to establish BWR fracture mechanics methods for generating P-T curves/limits that allow BWR plants to adopt the PTLR option.

Historically, utilities that own BWRs have submitted license amendment requests to update their P-T curves. In addition, the current situation causes both the regulator and licensees to expend resources that could otherwise be devoted to other activities. The objective of BWROG L TRs is to avoid these situations by providing P-T curve methods that are generically approved by the NRC so that P-T curves can be documented in a PTLR.

In order to implement the PTLR, the analytical methods used to develop the P-T limits must be consistent with those previously reviewed and approved by the NRC, and must be referenced in the Administrative Controls section of the plant Technical Specifications. The L TRs provide the current BWROG methodology for developing RCS pressure test, core not critical, and core critical P-T curves for BWRs.

As discussed in Section 2.0 of the Reference 1, 1 0 CFR Part 50, Appendix G, requires licensees to establish limits on the pressure and temperature of the Reactor Coolant Pressure Boundary (RCPB) in order to protect the RCPB against brittle failure (i.e., against brittle "fast fracture"). These limits are defined by P-T limit curves for normal operations (including heatup and cooldown operations of the RCS, normal operation of the RCS with the reactor being in a critical condition, and transient operating conditions) and during pressure testing conditions (i.e.,

either inservice leak rate testing and/or hydrostatic testing conditions).

BWROG-TP-1 1-022, Revision 1 was prepared by Structural Integrity Associates (SIR-05-044, Revision 1 ) and has sections and appendices identical to its previous edition, LTR SIR-05-044-A. Since Revision 1 contains only limited modifications when compared with the previous edition, the NRC SE focuses on these modifications (excluding editorial changes) and does not repeat the evaluation of the Revision 1 contents which have been discussed and accepted in E1-3 to NL-1 3-1 773 Description and Assessment the NRC SE for the previous edition (L TR-SIR-04-055-A). As discussed in the NRC SE for L TR-05-044-A, Section 1.0 of SIR-05-044 (and therefore of SIR-05-044 Revision 1 ) describes the background and purpose for the L TR. Section 2.0 provides the fracture mechanics methodology and its basis for developing P-T limits. Section 3.0 provides a step-by-step procedure for calculating P-T limits. Appendix A provides guidance for evaluating surveillance data. Appendix 8 provides a template PTLR.

The SER that approves BWROG-TP-1 1-022 Revision 1 contains the following condition:

Each applicant referencing this L TR shall confirm that, in addition to the requirements in the ASME Code,Section XI, Appendix G, the lowest service temperatures for all ferritic RCPB components that are not part of the RPV, are below the lowest operating temperature in the proposed P-T limits.

HNP has evaluated this condition, and has determined it to be addressed for PTLR curves provided for both Unit 1 and Unit 2.

BWROG-TP-1 1-023, Revision 0 was also prepared by Structural Integrity Associates (0900876.401.RO). The SER that approves BWROG-TP-1 1-023 Revision 0 states "the staff determined that no conditions or limitations are necessary for future applicants to address in their application of this LTR to their plant-specific P-T limit submittals". Section 1, "Introduction,"

provides the relationship among P-T limits, PTLR, and the proposed LEFM evaluation for WLI nozzles. Section 2, "Methodology," provides the proposed LEFM methodology, starting from the stress analysis based on the finite element method (FEM) to the LEFM analysis based on the boundary integral/influence function (BIE/IF) method. Section 3, "Assumptions," provides assumptions adopted in each step of the proposed methodology, such as the heat transfer coefficients for the WLI nozzle and RPV external and internal surfaces, material properties for various components, and the stress free temperature for evaluating cladding stresses. Section 4, "Finite Element Model," provides information on development of the 38 FEM models first introduced in Section 2, considering various types of nozzle and FEM models with and without a crack. This section also addressed FEM model validation through mesh density check for models with and without a crack. Section 5, "Instrument Nozzle Load Cases," addresses the loads on the nozzle: internal pressure, thermal transient, and pipe reaction load. Section 6, "Pressure, Thermal, and Piping Load Results," presents FEM results (i.e., stressors and applied stress intensity factors (K1s)) under these three types of load. Section 7, "Observations and Discussions," offers the BWROG's observation of the behavior of the applied K1 and a discussion of the modeling choices that could affect the results. Section 8, "Generic Methodology for K1 Estimation," provides generic K1 Formulas derived from the FEM results for a variety of nozzles for licensees to use for WLI nozzles in their plant-specific P-T limit applications. Section 9, "Summary," provides summary findings and conclusions.

E1-4 to NL-1 3-1 773 Description and Assessment Regulatory Safety Analysis 4.1 No Significant Hazards Consideration SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 1 0 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change modifies Edwin I. Hatch Nuclear Plant (HNP) Unit 1 and Unit 2 Technical Specifications (TS) Section 1.0 ("Definitions"), Limiting Conditions for Operation and Surveillance Requirement Applicability Section 3.4.9 ("RCS Pressure and Temperature (PIT) Limits"), and Section 5.0 ("Administrative Controls"), to delete reference to the pressure and temperature curves, and to include reference to the Pressure and Temperature Limits Report (PTLR). This change adopts the methodology of BWROG-TP-1 1-022-A, Revision 1 (SIR-05-044, Revision 1 -A), "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", dated June 201 3, and of BWROG-TP-1 1-023-A, Revision 0 (0900876.401, Revision 0-A), "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure Temperature Curve Evaluations", dated May 2013, for preparation of the pressure and temperature curves, and incorporates the guidance of TSTF-41 9-A, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR". The HNP PTLRs have been developed based on the methodologies provided in BWROG-TP-1 1-022-A, Revision 1 and BWROG-TP-1 1 -

023-A, Revision 0, and based on the template provided in BWROG-TP-1 1-022-A, Revision 1. The HNP PTLRs meet all Conditions specified in the Safety Evaluation Reports (SERs) for BWROG-TP-1 1-022-A, Revision 1 and for BWROG-TP-1 1-023-A, Revision 0.

The NRC has established requirements in Appendix G to 1 0 CFR 50 in order to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants.

Additionally, the regulation in 1 0 CFR Part 50, Appendix H, provides the NRC staff's criteria for the design and implementation of reactor pressure vessel (RPV) material surveillance programs for operating lightwater reactors. Implementing these NRC approved methodologies does not reduce the ability to protect the reactor coolant pressure boundary as specified in Appendix G, nor will this change increase the probability of malfunction of plant equipment, or the failure of plant structures, systems, or components. Incorporation of the new methodologies for calculating P-T curves, and the relocation of the P-T curves from the TS to the PTLR, provides an equivalent level of assurance that the RCPB is capable of performing its intended safety functions. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No E1-5 to NL-1 3-1 773 Description and Assessment The proposed change does not affect the assumed accident performance of the RCPB, nor any plant structure, system, or component previously evaluated. The proposed change does not involve the installation of new equipment, and installed equipment is not being operated in a new or different manner. The change in methodology ensures that the RCPB remains capable of performing its safety functions. No set points are being changed which would alter the dynamic response of plant equipment. Accordingly, no new failure modes are introduced which could introduce the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The proposed change does not affect the function of the RCPB or its response during plant transients. There are no changes proposed which alter the setpoints at which protective actions are initiated, and there is no change to the operability requirements for equipment assumed to operate for accident mitigation. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based upon the above, SNC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 1 0 CFR 50.92(c), and, accordingly, a finding of no significant hazards is justified.

4.2 Applicable Regulatory Requirements I Criteria The NRC has established requirements in Appendix G of Part 50 to Title 1 0 of the Code of Federal Regulations, in order to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. The regulation at 1 0 CFR Part 50, Appendix G requires that the P-T limits for an operating light-water nuclear reactor be at least as conservative as those that would be generated if the methods of Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code,Section XI, Appendix G) were used to generate the P-T limits. The regulation at 1 0 CFR Part 50, Appendix G, also requires that applicable surveillance data from reactor pressure vessel (RPV) material surveillance programs be incorporated into the calculations of plant-specific P-T limits, and that the P-T limits for operating reactors be generated using a method that accounts for the effects of neutron irradiation on the material properties of the RPV beltline materials.

Table 1 to 1 0 CFR Part 50, Appendix G provides the NRC staff's criteria for meeting the P-T limit requirements of ASME Code,Section XI, Appendix G, as well as the minimum temperature requirements of the rule for bolting up the vessel during normal and pressure testing operations.

In addition, the NRC staff regulatory guidance related to P-T limit curves is found in Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", and Standard Review Plan Chapter 5.3.2, "Pressure-Temperature Limits and Pressurized Thermal Shock".

The regulation at 1 0 CFR Part 50, Appendix H, provides the NRC staff's criteria for the design and implementation of RPV material surveillance programs for operating lightwater reactors.

HNP Units 1 and 2 demonstrates its compliance with the Appendix H through participation in the BWRVIP Integrated Surveillance Program (ISP) (Reference 7).

E1-6 to NL-1 3-1 773 Description and Assessment In March 2001, the NRC staff issued RG 1.1 90, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence". Fluence calculations are acceptable if they are done with approved methodologies or with methods which are shown to conform to the guidance in RG 1.1 90.

Section 1 82a of the Atomic Energy Act of 1 954 requires applicants for nuclear power plant operating licenses to include TS as part of the license. The Commission's regulatory requirements related to the content of TS are set forth in 1 0 CFR 50.36. That regulation requires that the TS include items in five specific categories: (1) safety limits, limiting safety system settings and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls.

The regulation at 1 0 CFR 50.36(c)(2)(ii)(B) requires that LCOs be established for the P-T limits, because the parameters fall within the scope of the Criterion 2 identified in the rule:

A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The P-T limits for BWR-designed light-water reactors fall within the scope of Criterion 2 of 1 0 CFR 50.36(c)(2)(ii)(B) and are therefore ordinarily required to be included within the TS LCOs for a plant-specific facility operating license. On January 31, 1 996, the NRC staff issued GL 96-03 to inform licensees that they may request a license amendment to relocate the P-T limit curves from the TS LCOs into a PTLR or other licensee-controlled document that would be controlled through the Administrative Controls Section of the TS. In GL 96-03, the NRC staff informed licensees that in order to implement a PTLR, the P-T limits for light-water reactors would need to be generated in accordance with an NRC-approved methodology and that the methodology to generate the P-T limits would need to comply with the requirements of 1 0 CFR Part 50, Appendices G and H; be documented in an NRC-approved topical report or plant specific submittal; and be incorporated by reference in the Administrative Controls Section of the TS.

This change implements the methodology provided in the Structural Integrity Associates reports (References 2 and 4), which will continue to ensure compliance with Appendices G and H of the Code of Federal Regulations in conjunction with plant commitments to the BWRVIP ISP program, and the associated regulatory guidance, including TSTF-41 9-A, which provides TS changes.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the Commission's regulations; and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

5.0 Environmental Assessment A review has determined that the proposed changes would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 1 0 CFR 20, or would change an inspection or surveillance requirement. However, the proposed changes do not involve: (i) a significant hazards consideration; (ii) a significant change in the E1-7 to NL-1 3-1 773 Description and Assessment types or significant increase in the amounts of any effluents that may be released offsite; or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 1 0 CFR 51.22(c)(9). Therefore, pursuant to 1 0 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed changes.

6.0 Precedent This change is generally consistent with the changes to the Improved TS described in TSTF-419-A, (Reference 6). Plants which have received approval for similar changes, in whole or in part, are listed below:

James A. Fitzpatrick Nuclear Power Plant (ADAMS Accession No. ML082630385)

Pilgrim Nuclear Power Station (ADAMS Accession No. ML 1 1 0050298)

Nine Mile Point Nuclear Station (ADAMS Accession No. ML093370002)

Oyster Creek Nuclear Generating Station (ADAMS Accession No. ML082390685) 7.0 References 1. Letter from S. Bahadur (NRC) to F. Schiffley (BWROG Chairman), "Final Safety Evaluation for Boiling Water Reactor Owners' Group Topical Report BWROG-TP-1 1-022, Revision 1, November 201 1, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors" (TAC NO. ME7649)", dated May 16, 2013

2. Letter from F. Schiffley (BWROG Chairman) to J. Golla (NRC), "Submittal of Boiling Water Reactor Owners' Group Topical Report BWROG-TP-1 1-022-A, Revision 1,

"Pressure-Temperature Limits Report Methodology for Boiling Water Reactors" (TAC NO. ME7649), dated September 4, 201 3

3. Letter from S. Bahadur (NRC) to F. Schiffley (BWROG Chairman), "Final Safety Evaluation for Boiling Water Reactors Owners' Group Topical Report BWROG-TP-1 1-023, Revision 0, November 201 1, "Linear Elastic Fracture Mechanics Evaluation Of General Electric Boiling Water Reactor Water Level Instrument Nozzles For Pressure Temperature Curve Evaluations" (TAC NO. ME7650), dated March 14, 201 3

4. Letter from F. Schiffley (BWROG Chairman) to J. Golla (NRC), "Submittal of Boiling Water Reactor Owners' Group Licensing Topical Report BWROG-TP-1 1-023-A, Revision 0, "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations" (TAC NO. ME7650)", dated June 28, 201 3

5. Generic Letter (GL) 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits", January 31, 1 996.

6. TSTF-41 9-A, "Revised PTLR Definition and References in ISTS 5.6.6, RCS PTLR,"

dated August 4, 2003.

7. Letter from L. N. Olshan (NRC) to H. L. Sumner (SNC), "Edwin I. Hatch Nuclear Plant, Units 1 and 2 Re: Issuance of Amendments (TAC NOS. M B6106 and MB61 07)", dated March 1 0, 2003.

E1-8

Edwin I. Hatch Nuclear Plant Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent with TSTF-419-A Proposed Technical Specification Changes (Mark-Up)

Definitions 1. 1 1.1 Definitions (continued)

MINIMUM CRITICAL POWER RATIO (MCPR)

MODE OPERABLE OPERABILITY PHYSICS TESTS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

RATED THERMAL POWER (RTP)

REACTOR PROTECTION SYSTEM (RPS)

RESPONSE TIME HATCH UNIT 1 The MCPR shall be the smallest critical power ratio (CPR) that exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1. 1-1 with fuel in the reactor vessel.

A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a.

Described in Section 1 3.6, Startup and Power Test Program, of the FSAR;

b.

Authorized under the provisions of 10 CFR 50.59; or

c.

Otherwise approved by the Nuclear Regulatory Commission.

The PTLR is the unit specific document that orovides the reactor vessel pressure and temperature limits, including heatup and cooldown rates. for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2804 MWt.

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

1. 1-4 (continued)

Amendment No.

3.4 REACTOR COOLANT SYSTEM (RCS)

3. 4.9 RCS Pressure and Temperature (PfT) Limits RCS PfT Limits 3.4.9 LCO 3.4.9 RCS pressure, RCS temperature, andHCS heatup and cooldown rates shall be maintained within the limits specified in the PTLR.,.aAEI-tlhe recirculation pump starting temperature requirements shall be maintained within limits.

APPLICABILITY:

At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

NOTE--------------

A.1 Restore parameter(s) to 30 minutes Required Action A.2 shall within limits.

be completed if this Condition is entered.

AND Requirements of the LCO A.2 Determine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months not met in MODES 1, 2, acceptable for and 3.

continued operation.

B.

Required Action and B.1 Be in MODE 3.

1 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C.

NOTE---------------

C.1 Initiate action to restore Immediately Required Action C.2 shall parameter(s) to within be completed if this limits.

Condition is entered.

AND Requirements of the LCO C.2 Determine RCS is Prior to entering not met in other than acceptable for MODE 2 or 3 MODES 1, 2, and 3.

operation.

HATCH UNIT 1 3.4-1 8 Amendment No. 2ee I

SURVEILLANCE REQUIREMENTS SR 3.4.9.1 SR 3.4.9.2 SR 3.4.9.3 HATCH UNIT 1 SURVEILLANCE Verify:

a.

RCS pressure and RCS temperature are within the limits specified in Figures a.4.9 1 and a.4.9 2the PLTR during RCS inservice leak and hydrostatic testing, and during RCS non-nuclear heatup and cooldown operations; and

b.

RCS heatup and cooldown rates are

< 100°F in any 1 heur periedwithio the limits speci.fm.d in the PTLR during RCS heatup and cooldown operations, and RCS inservice leak and hydrostatic testing.

N 0 T E ------------------------------

Only required to be met when the reactor is critical and immediately prior to control rod withdrawal for the purpose of achieving criticality.

Verify RCS pressure and RCS temperature are within the criticality limits specified in Figure a.4.9 athe PTLB.

NOTE------------------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation pump.

Verify the difference between the bottom head coolant temperature and the reactor pressure vessel (RPV) coolant temperature is s 145°F.

3.4-1 9 RCS PIT Limits 3.4.9 FREQUENCY In accordance with the Surveillance Frequency Control Program Once within 1 5 minutes prior to initial control rod withdrawal for the purpose of achieving criticality Once within 1 5 minutes prior to starting an idle recirculation pump (continued)

Amendment No. I

SR 3.4.9.4 SR 3.4.9.5 HATCH UNIT 1

N 0 TE -----------------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation pump.

Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature is::;; 50°F.

N 0 TE-----------------------------

Only required to be met when tensioning/

detensioning the reactor vessel head bolting studs.

Verify reactor vessel flange and head flange temperatures are > 76°Fw.itb.io the limits specifi d

n e PTL.

3.4-20 RCS PIT Limits 3.4.9 FREQUENCY Once within 1 5 minutes prior to starting an idle recirculation pump Once within 30 minutes prior to tensioning/

detensioning the reactor vessel head bolting studs and in accordance with the Surveillance Frequency Control Program Amendment No. 299 I

SURVEILLANCE REQUIREMENTS (continued SR 3.4.9.6 HATCH UNIT 1 SURVEILLANCE

N 0 TE -----------------------------

Only required to be met when the reactor vessel head is tensioned.

Verify reactor vessel flange and head flange temperatures are > 75°F'within the limits specified in the PTLR.

3.4-21 RCS Pff Limits 3.4.9 FREQUENCY Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature is s 1 06°F in MODE 4, and in accordance with the Surveillance Frequency Control Program Once within 30 minutes after RCS temperature is s 86°F in MODE 4, and in accordance with the Surveillance Frequency Control Program Amendment No. I

1400 1300 1200 1100 I

I I

I

... #JjJ.I.I.-- 20 1

32 01

.; 1000 J

,,...,__44

-48 K 900

I

,....--4-54

n.

I 11111 i

500

J I

w 400 en en w

a::

0..

300 I _ll----,..-,

r--

312 PSIG I

I 200 úûüý*4---}--4ÿ-L---

BELTLINE l l--:+--=-:-:

A=SYSTEM 100 0

AND 1

FLANGE HYDROTEST LIMITS B678M R9:!;N 1+----+-1 WITH FUEL IN THE 1---

sa*F IL-.,-------J VESSEL I

I I

0 50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE {"F)

RCS Prr Limits 3.4.9 INITIAL RTndt VALUES ARE

-20"F FOR BELTLINE, 40"F FOR UPPER VESSEL, AND 1 O"F FOR BOTTOM HEAD HEATUP /COOLDOWN RATE 20"F /HR BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT ("F) 20 130.7 BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT ("F) 32 154.6 CURVES IJ AS SHOWN:

I EF'PY SHIFT ("F) 161.0 lot" I Tl 11\\Jt" f'IIO\\/t"C

--*S SHOWN:

EFPY SHIFT ("r) 40 167.5

-s£I.l UNt. I,;UKVt.S ADJUSTED AS SHOWN:

EFPY SHIFT ("F) 44 173.7 BELTLINE CURVES ADJUSTED AS SHOWN:

EF'PY SHIFT {"r) 48 179.4 BELTLINE CURVES ADJUSTED AS SHOWN:

HPY SHIFT ("r) 54 187.2

- BELTLINE LIMITS AND UPPER VESSEL LIMITS

- - - BOTTOM HEAD LIMITS

[ ACAD I F34911 Figure 3.4.9 1 (page 1 of 1)

PressurefTemperature limits for lnservise Hydrostatic and lnservise Leakage Tests HATCH UNIT 1 3.4 22 Amendment No. 222 I

1400 1300 1200 1100 01 Vi

c. 1000 c

w

r::

Q..

900 0

1-L...ao 1-7nn z

E 500

....1 w

0::

400 en en w

0::

Q..

300 200 100 0

I I

I 20 32 I

I I

54 I

I I

I I I

I r----.

I I

II ll I

t-I

_/

I 1--

I I

I tl I

I I--

312 PSIG I BEL TUNE I

B=HEATUP/

AND FLANGE BOTIO M REGION COOLDOWN LIMITS r--

HEAD 76'F CORE NOT CRITICAL 68'F I

I 0

50 1 00 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

RCS PIT Limits 3.4.9 INITIAL RTndt VALUES ARE

-20'F FOR BELTLINE, 40'F FOR UPPER VESSEL, AND 1 O'F FOR BOTTOM HEAD HEATUP /COOLDOWN RATE 100'F/HR BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT ('r) 20 130.7 ELTLINE CURVES 11c;,Trn A SHOWN*

r r

..,. IFT ('F) 32 154.6 BELTUNE CURVES ADJUSTED AS SHOWN*

EFPY SHIFT ("F) 54 187.2

- BELTLINE LIMITS AND UPPER VESSEL LIMITS

- - - BOTTOM HEAD LIMITS

( ACMJ I F34921 Figblre J.4.Q 2 (page 1 of 1)

PressblrefTernperature Limits for Non Nblolear Heatblp, Lm'l Power Physics Tests, ami Cooldown Follm*Jing a Shbltdown HATCH UNIT 1 a.4 23s---------..P,ArRill

e0niEldiRFR*emnlt-t NieO.,.-,. 2i6292

1400 1300 1200 1100 Cl iii c..

'-" 1000 0

w

z::

a.

900 0

1-

-I

[

uuu I

5 700 f--' 1--

I

!T=¥--

J 500 w

J 400 VI VI w

a.

300 MINIMUM 200 CRITICALITY TEMPERATURE 76'F 100

(

0 20 32 54 I

II I

I I

1 I

3"12 P! IG I J

I r-C=HEATUP/

COOLDOVIIN LIMITS CORE CRITICAL RCS PIT Limits 3.4.9 INITIAL RTndt VALUES ARE

-20"F FOR BELTLINE, 40"F FOR UPPER VESSEL, AND 1 O"F FOR BOTTOM HEAD HEATUP/COOLDOWN RATE 1 OO"F /HR BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIH ("F) 20 130.7 fNF" r.LJRVF" ADJ TED AS SHOWN:

("F) 154.6 BELTLINE CURVES ADJUSTED AS SH,?N:

OT 0

187.2 54

- BELTLINE LIMITS AND UPPER VESSEL W.4ITS 0

50 100 150 200 250 300 350 400 (ACAO! F34931 I MINIMUM REACTOR VESSEL METAL TEMPERATURE {"F)

FiJI:IFe 3.4.Q 3 (paJe 8 ef :)

Press:re/TempemtuFe Limits fer Criticality HATCH UNIT 8 3.4 24 Amendment Ne. 222 l

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in NEDE-2401 1 -P-A, "General Electric Standard Application for Reactor Fuel," (applicable amendment specified in the COLR).

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by LCO 3.3.3.1, "Post Accident Monitoring (PAM)

Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Reactor Coolant System <RCS) PRESSURE AND TEMPERATURE LIMITS REPORT lPTLR) a, RCS pressure and temperature limits for heat up, cooldown. low temperature operation. criticality, and hydrostatic testing. as well as heatup and cooldown rates, sJ:tall.b.e established and documented in the PTLR for the following:

i.

Limiting Conditions for Operating Section 3.4.9 "RCS Pressure and Temperature <PITl Limits" ii.

Surveillance Requirements Section 3.4.9. "RCS Pressure and Temperature <Pff) LimitS:

b.

The analytical methods used to determine_the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC. specifically those described in the following documents:

HATCH UNIT 1

i.

BWROG-TP-11-022-A. Revision 1 <SIR-05-044. Revision 1 -A),

"Pressure-Temperature Limits Report Methodology for Boiling Water Reactors". dated August 2013 ii.

BWROG-TP-11-023-A, Revision 0 <0900876.401 Revision 0-A),

"Linear Elastic Fracture Mechanics Evaluation of General Electric Boili.Dg Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations". dated May 2013 5.0-21 Amendment No. 8

Reporting Requirements 5.6

c.

The PTLR shall be provided to the NRC upon issuance for each reactor vess.eJ ftuence period and for any revision or supplemenUhereJQ...

HATCH UNIT 1 5.0-22 Amendment No.

Definitions 1. 1 1.1 Definitions (continued)

PHYSICS TESTS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLRl RATED THERMAL POWER (RIP)

REACTOR PROTECTION SYSTEM (RPS)

RESPONSE TIME SHUTDOWN MARGIN (SDM)

PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a.

Described in Chapter 1 4, Initial Tests and Operation, of the FSAR;

b.

Authorized under the provisions of 1 0 CFR 50.59; or

c.

Otherwise approved by the Nuclear Regulatory Commission.

The PTLR is the unit specific document that orovides the reactor vessel pressure and temperature limits. including heatup and cooldown rates for the current reactor vessel fluence period. These wessure and temperature limits shall be determined for each fluence period in Sl.C..C9Lda.nce with Specification 5.6, 7.

RIP shall be a total reactor core heat transfer rate to the reactor coolant of 2804 MWt.

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a.

The reactor is xenon free;

b.

The moderator temperature is ;;:: 68°F, corresponding to the most reactive state; and

c.

All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

HATCH UNIT 2 1. 1 -5 Amendment No.--24+ I

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 RCS Pressure and Temperature (Pff) Limits RCS Pff Limits 3.4.9 LCO 3.4.9 RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within the limits specified in the PTLR...,-aoo-t_Ihe recirculation pump starting temperature requirements shall be maintained within limits.

APPLICABILITY:

At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

NOTE--------------

A.1 Restore parameter(s) to 30 minutes Required Action A.2 shall within limits.

be completed if this Condition is entered.

AND A.2 Determine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Requirements of the LCO acceptable for not met in MODES 1, 2, continued operation.

and 3.

B.

Required Action and B.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C.

NOTE---------------

C.1 Initiate action to restore Immediately Required Action C.2 shall parameter(s) to within be completed if this limits.

Condition is entered.

AND Requirements of the LCO C.2 Determine RCS is Prior to entering not met in other than acceptable for MODE 2 or 3 MODES 1, 2, and 3.

operation.

HATCH UNIT 2 3.4--1 8 Amendment No. 2-tG I

SURVEILLANCE REQUIREMENTS SR 3.4.9.1 SR 3.4.9.2 SR 3.4.9.3 HATCH UNIT 2 SURVEILLANCE Verify:

a.

RCS pressure and RCS temperature are within the limits specified in F'igblres ;.4.9 1 ana ;.4.Q 2the PTLR during RCS inservice leak and hydrostatic testing, and during RCS non-nuclear heatup and cooldown operations; and

b.

RCS heatup and cooldown rates are

< 1 00°F in any 1 hObiF periolwithjn the limits specified in the PTLR during RCS heatup and cooldown operations, and RCS inservice leak and hydrostatic testing.

N 0 T E ------------------------------

Only required to be met when the reactor is critical and immediately prior to control rod withdrawal for the purpose of achieving criticality.

Verify RCS pressure and RCS temperature are within the criticality limits specified in F'igblre ;.4.Q `the PTLR.

N 0 T E ------------------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation pump.

Verify the difference between the bottom head coolant temperature and the reactor pressure vessel (RPV) coolant temperature is s 145°F.

3.4-1 9 RCS PfT Limits 3.4.9 FREQUENCY In accordance with the Surveillance Frequency Control Program Once within 15 minutes prior to initial control rod withdrawal for the purpose of achieving criticality Once within 1 5 minutes prior to starting an idle recirculation pump (continued)

Amendment No. F I

SURVEILLANCE REQUIREMENTS continued)

SR 3.4.9.4 SR 3.4.9.5 HATCH UNIT 2 SURVEILLANCE

N 0 T E ------------------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation pump.

Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature is s 50°F.

N 0 T E ------------------------------

Only required to be met when tensioning/

detensioning the reactor vessel head bolting studs.

Verify reactor vessel flange and head flange temperatures are within the limits specified in the PTLR> 90°F.

3.4-20 RCS PfT Limits 3.4.9 FREQUENCY Once within 1 5 minutes prior to starting an idle recirculation pump Once within 30 minutes prior to tensioning/

detensioning the reactor vessel head bolting studs and in accordance with the Surveillance Frequency Control Program (continued)

Amendment No. F I

SURVEILLANCE REQUIREMENTS (continued SR 3.4.9.6 HATCH UNIT 2 SURVEILLANCE

N 0 T E Only required to be met when the reactor vessel head is tensioned.

Verify reactor vessel flange and head flange temperatures are within the limits specified in the EI!.B..> 90°P.

3.4-21 RCS PfT Limits 3.4.9 FREQUENCY Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature is 120°F in MODE 4, and in accordance with the Surveillance Frequency Control Program Once within 30 minutes after RCS temperature is 1 00°F in MODE 4, and in accordance with the Surveillance Frequency Control Program Amendment No. F I

1400 1300 1 200 1 100 01 I

"iii

c. 1 000 I

0 I

loJ

r:

0..

900 0

1-I H

800 I

I I

34 l it 600 1-

I I

I

[

'1-UU I

I 300 31 PSIG I

I 200 IELTLINI I

AND I

'LANGE BOTTOM

20 32 54 RCS PIT Limits 3.4.9 INITIAL RTndt VALUES ARE 24"F FOR BELTUNE, 26"F FOR UPPER VESSEL, AND 50"F FOR BOTTOM HEAD HEATUP /COOLDOWN RATE 20"F /HR BELTUNE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT ("F) 20 43.5

`abucd r-**qv E 'Y

""".,G("F)

..1!

B I.ES ADJUSTED AS SHOWN:

EFPY SHIFT ("F) 54 53.7

- BEL TUNE UMITS AND UPPER VESSEL UMITS A=SYSTEM HYDROTEST LIMITS 100

_,_ REGION WITH FUEL IN THE

- - - BOTTOM HEAD UMITS HEAD 90"F 68"F I

VESSEL I

0 I

0 50 100 1 50 200 250 300 350 400 (ACADJ FJ491 I MINIMUM REACTOR VESSEL METAL TEMPERATURE ("F)

Fi!iJI:IFe J.4.9 8 (page 8 ef 8)

PFessi:IFel+empeFati:IFe bimits feF lnsmvise Hydmstatis and lnseFVise beakage Tests I=IATGH YNIT 2 J.4 22 Amendment Ne. HiJ

1400 1300 1200 1100 Cl Ui

c. 1000 c

I:

a.

900 0

1-

---1 F.uu 17oo O

500 0::

400

J VI VI 0::

a.

300 200 100 0

' t-I I

I I

BELTLINI I

AND I

BOTTOM 1-L-HEAD 68'F I

I L

0 50 I

I I

I i

I I I 312 F FLANGE REGION 90'F 54 20 32 sTGl B=HEATUP/

COOLDOWN LIMITS CORE NOT CRITICAL 100 150 200 250 300 350 1--

1--

400 MINIMUM REACTOR VESSEL METAL TEMPERATURE ("F)

igFe J.4.Q 2 (page8 of: )

RCS PfT Limits 3.4.9 INITIAL RTndt VALUES ARE 24'F FOR BELTUNE, 26'F FOR UPPER VESSEL, AND SO'F FOR BOITOM HEAD HEATUP/COOLDOWN RATE 1 OO'F /HR BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT ('F) 20 43.5

- 81 ruNE CURVES ADJ\\ "ED AS SHOWN:

E

('F) 48.0.

BELTLINE CURVES ADJUSTED AS SHOWN:

t.4 23.i

')

- BEL TUNE LIMITS AND UPPER VESSEL LIMITS

- - - BOTTOM HEAD LIMITS (ACAO I F3492 I PFessFel+empeFatFe bimits feF Non NsleaF l=leatp.

bow PoweF Physiss Tests, and Cooldmo'o'R allowing a Shtdown I=IATCI=I UNIT 2 J.4 23

,A,mendment No. 8 eJ

1 400 1 300 1 200 1 1 00 Cll w

a. 1 000 0

X:

D..

900 0

.. 700

.....,.,.. 1-z r;nn

....1....

0:::

400 Vl Vl 0:::

D..

300 200 1 00 0

0 50 MINIMUM HATGH UNIT2 54 20 32 312.PSIG r--

MINIMUM C=HEATUP/

CRITICALITY COOLDOWN LIMITS TEMPERATURE t--r-r--

90"F CORE CRITICAL I

RCS Pff Limits 3.4.9 INITIAL RTndt VALUES ARE 24*F FOR BELTUNE, 26*F FOR UPPER VESSEL, AND 50*F FOR BOTTOM HEAD HEATUP /COOLDOWN RATE 1 00.F/HR BELTUNE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (*F)

LTUNE CURVES SHOWN:

FPY SHIFT (*F) 32 48.0 BELTUNE CURVES AUJU:::.I t.U A:::. :::.H.UWN:

EFPY SHIFT (*F) 54 53.7

- BEL TUNE AND NON-BELTUNE LIMITS 1 00 1 50 200 250 300 350 400 (ACAD(

F3493 I REACTOR VESSEL METAL TEMPERATURE (*F)

Fi!iJI:lFe 3.4.Q 3 (13age 8 ef :)

PFessl:lFel+emJ:)eFatl:lFe bimits feF GFitisality 3.4 24 Amendment Ne. 63 I

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT CCOLR) (continued)

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in NEDE-2401 1 -P-A, "General Electric Standard Application for Reactor Fuel," (applicable amendment specified in the COLR).

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by LCO 3.3.3.1, "Post Accident Monitoring (PAM)

Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

7 Reactor Coolant System !RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR>

a.

RCS pressure and temperature limits for heat up. cooldown, low.

te_mperature operation. criticality, and hydrostatic testing. as well as heatup and cooldown raie_s_._sllall be established and documented in tbe PTLR for the following:

i.

Limiting Conditions for Operating Section 3.4.9 "RCS Pressure and Temperature CPm Limits" ii.

Surveillance Requirements Section 3.4.9. "RCS Pressure and Temperature (Pal 1,..

b.

Tbe analvtical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC. specifically those described in the following documents:

HATCH UNIT 2

i.

BWROG-TP-11 -022-A. Revision 1 CSIR-05-044, Revision 1 -Al "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors" dated June 2013 ii BWROG-TP-1 1 -023-A. Revision 0 (0900876.401. Revision 0-Al.

"Linear Elastic Fracture Mechanics Evaluation of General E.Je.ctri Boiling Water Reactor Water Level lllsJrument Nozzles for Pressure-Temperature Curve Evaluations". dated May 2013 5.0-21 Amendment No. M

Reporting Requirements 5.6

c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplemenUhereto HATCH UNIT 2 5.0-22 Amendment No. M

Edwin I. Hatch Nuclear Plant Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent with TSTF-419-A Revised Technical Specification Pages

Definitions 1. 1 1. 1 Definitions (continued)

MINIMUM CRITICAL POWER RATIO (MCPR)

MODE OPERABLE OPERABILITY PHYSICS TESTS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

RATED THERMAL POWER (RTP)

REACTOR PROTECTION SYSTEM (RPS)

RESPONSE TIME HATCH UNIT 1 The MCPR shall be the smallest critical power ratio (CPR) that exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a.

Described in Section 1 3.6, Startup and Power Test Program, of the FSAR;

b.

Authorized under the provisions of 10 CFR 50.59; or

c.

Otherwise approved by the Nuclear Regulatory Commission.

The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2804 MWt.

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

1. 1-4 (continued)

Amendment No.

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 RCS Pressure and Temperature (PfT) Limits RCS PfT Limits 3.4.9 LCO 3.4.9 RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within the limits specified in the PTLR. The recirculation pump starting temperature requirements shall be maintained within limits.

APPLICABILITY:

At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

NOTE--------------

A.1 Restore parameter(s) to 30 minutes Required Action A.2 shall within limits.

be completed if this Condition is entered.

AND Requirements of the LCO A.2 Determine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months not met in MODES 1, 2, acceptable for and 3.

continued operation.

B.

Required Action and B.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C.

NOTE---------------

C.1 Initiate action to restore Immediately Required Action C.2 shall parameter(s) to within be completed if this limits.

Condition is entered.

AND Requirements of the LCO C.2 Determine RCS is Prior to entering not met in other than acceptable for MODE 2 or 3 MODES 1, 2, and 3.

operation.

HATCH UNIT 1 3.4-1 8 Amendment No.

SURVEILLANCE REQUIREMENTS SR 3.4.9.1 SR 3.4.9.2 SR 3.4.9.3 HATCH UNIT 1 SURVEILLANCE Verify:

a.

RCS pressure and RCS temperature are within the limits specified in the PTLR during RCS inservice leak and hydrostatic testing, and during RCS non-nuclear heatup and cooldown operations; and

b.

RCS heatup and cooldown rates are within the limits specified in the PTLR during RCS heatup and cooldown operations, and RCS inservice leak and hydrostatic testing.

N 0 T E ------------------------------

Only required to be met when the reactor is critical and immediately prior to control rod withdrawal for the purpose of achieving criticality.

Verify RCS pressure and RCS temperature are within the criticality limits specified in the PTLR.

N 0 T E ------------------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation pump.

Verify the difference between the bottom head coolant temperature and the reactor pressure vessel (RPV} coolant temperature is s 145°F.

3.4-1 9 RCS PfT Limits 3.4.9 FREQUENCY In accordance with the Surveillance Frequency Control Program Once within 1 5 minutes prior to initial control rod withdrawal for the purpose of achieving criticality Once within 1 5 minutes prior to starting an idle recirculation pump (continued}

Amendment No.

SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.9.4 SR 3.4.9.5 HATCH UNIT 1 SURVEILLANCE

N 0 TE -----------------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation pump.

Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature is s 50°F.

N 0 TE---------------------------

Only required to be met when tensioning/

detensioning the reactor vessel head bolting studs.

Verify reactor vessel flange and head flange temperatures are within the limits specified in the PTLR.

3.4-20 RCS Pff Limits 3.4.9 FREQUENCY Once within 1 5 minutes prior to starting an idle recirculation pump Once within 30 minutes prior to tensioning/

detensioning the reactor vessel head bolting studs and in accordance with the Surveillance Frequency Control Program Amendment No.

SURVEILLANCE REQUIREMENTS continued)

SR 3.4.9.6 HATCH UNIT 1 SURVEILLANCE

N 0 TE -----------------------------

Only required to be met when the reactor vessel head is tensioned.

Verify reactor vessel flange and head flange temperatures are within the limits specified in the PTLR.

3.4-21 RCS PIT Limits 3.4.9 FREQUENCY Once within 1 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after RCS temperature is s 1 06°F in MODE 4, and in accordance with the Surveillance Frequency Control Program Once within 30 minutes after RCS temperature is s 86°F in MODE 4, and in accordance with the Surveillance Frequency Control Program Amendment No.

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 0 Reactor Steam Dome Pressure Reactor Steam Dome Pressure 3.4. 1 0 LCO 3.4. 1 0 The reactor steam dome pressure shall be s 1 058 psig.

APPLICABILITY:

MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION A.

Reactor steam dome A.1 Restore reactor steam pressure not within limit.

dome pressure to within limit.

B.

Required Action and B.1 Be in MODE 3.

associated Completion Time not met.

SURVEILLANCE REQUI REMENTS SURVEILLANCE SR 3.4. 1 0. 1 HATCH UNIT 1 Verify reactor steam dome pressure is s 1 058 psig.

3.4-22 COMPLETION TIME 15 minutes 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. I

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel," (applicable amendment specified in the COLR).

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by LCO 3.3.3. 1, "Post Accident Monitoring (PAM)

Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a.

RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, as well as heatup and cooldown rates, shall be established and documented in the PTLR for the following:

i.

Limiting Conditions for Operating Section 3.4.9 "RCS Pressure and Temperature (PIT) Limits."

ii.

Surveillance Requirements Section 3.4.9, "RCS Pressure and Temperature (PIT) Limits."

b.

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

HATCH UNIT 1

i.

BWROG-TP-1 1-022-A, Revision 1 (SIR-05-044, Revision 1-A},

"Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," dated June 201 3.

(continued) 5.0-21 Amendment No.

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued) ii.

BWROG-TP-1 1-023-A, Revision 0 (0900876.401, Revision 0-A},

"Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure Temperature Curve Evaluations," dated May 201 3.

c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

HATCH UNIT 1 5.0-22 Amendment No.

5.0 ADMINISTRATIVE CONTROLS

5. 7 High Radiation Area Reporting Requirements 5.7

5. 7.1 Pursuant to 10 CFR 20, paragraph 20.1601, in lieu of the requirements of 1 0 CFR 20.1601 a, each high radiation area, as defined in 1 0 CFR 20, in which the intensity of radiation is > 1 00 mrem/hr but < 1 000 mrem/hr, measured at 30 em from the radiation source or from any surface the radiation penetrates, shall be barricaded and conspicuously posted as a high radiation area.

Entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g.,

Health Physics Technicians) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates

< 1 000 mrem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

A radiation monitoring device that continuously indicates the radiation dose rate in the area.

b.

A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.

c.

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physics supervision in the RWP.

5. 7.2 In addition to the requirements of Specification 5. 7. 1, areas with radiation levels

1 000 mrem/hr, measured at 30 em from the radiation source or from any surface the radiation penetrates, but less than 500 Rads in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months measured at 1 meter from the radiation source or from any surface that the radiation penetrates, shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervision on duty or Health Physics supervision.

HATCH UNIT 1 5.0-23 Amendment No.

Definitions 1. 1 1. 1 Definitions (continued)

PHYSICS TESTS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

RATED THERMAL POWER (RTP)

REACTOR PROTECTION SYSTEM (RPS)

RESPONSE TIME SHUTDOWN MARGIN (SDM)

PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a.

Described in Chapter 14, Initial Tests and Operation, of the FSAR;

b.

Authorized under the provisions of 1 0 CFR 50.59; or

c.

Otherwise approved by the Nuclear Regulatory Commission.

The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2804 Mwt.

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a.

The reactor is xenon free;

b.

The moderator temperature is =:: 68°F, corresponding to the most reactive state; and

c.

All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

(continued)

HATCH UNIT 2 1. 1-5 Amendment No.

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 RCS Pressure and Temperature (PfT) Limits RCS PfT Limits 3.4.9 LCO 3.4.9 RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within the limits specified in the PTLR. The recirculation pump starting temperature requirements shall be maintained within limits.

APPLICABILITY:

At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

NOTE--------------

A. 1 Restore parameter(s) to 30 minutes Required Action A.2 shall within limits.

be completed if this Condition is entered.

AND A.2 Determine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Requirements of the LCO acceptable for not met in MODES 1, 2, continued operation.

and 3.

B.

Required Action and B. 1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C.

NOTE---------------

C.1 Initiate action to restore Immediately Required Action C.2 shall parameter(s) to within be completed if this limits.

Condition is entered.

AND Requirements of the LCO C.2 Determine RCS is Prior to entering not met in other than acceptable for MODE 2 or 3 MODES 1, 2, and 3.

operation.

HATCH UNIT 2 3.4-1 8 Amendment No.

SURVEILLANCE REQUIREMENTS SR 3.4.9.1 SR 3.4.9.2 SR 3.4.9.3 HATCH UNIT 2 SURVEILLANCE Verify:

a.

RCS pressure and RCS temperature are within the limits specified in the PTLR during RCS inservice leak and hydrostatic testing, and during RCS non-nuclear heatup and cooldown operations; and

b.

RCS heatup and cooldown rates are within the limits specified in the PTLR during RCS heatup and cooldown operations, and RCS inservice leak and hydrostatic testing.

N 0 TE ------------------------------

Only required to be met when the reactor is critical and immediately prior to control rod withdrawal for the purpose of achieving criticality.

Verify RCS pressure and RCS temperature are within the criticality limits specified in the PTLR.

NOTE------------------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation pump.

Verify the difference between the bottom head coolant temperature and the reactor pressure vessel (RPV) coolant temperature is 1 45°F.

3.4-1 9 RCS Prr Limits 3.4.9 FREQUENCY In accordance with the Surveillance Frequency Control Program Once within 1 5 minutes prior to initial control rod withdrawal for the purpose of achieving criticality Once within 1 5 minutes prior to starting an idle recirculation pump (continued)

Amendment No.

SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.9.4 SR 3.4.9.5 HATCH UNIT 2 SURVEILLANCE

N 0 TE ------------------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation pump.

Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature is s 50°F.

N 0 TE ------------------------------

Only required to be met when tensioning/

detensioning the reactor vessel head bolting studs.

Verify reactor vessel flange and head flange temperatures are within the limits specified in the PTLR.

3.4-20 RCS P!T Limits 3.4.9 FREQUENCY Once within 1 5 minutes prior to starting an idle recirculation pump Once within 30 minutes prior to tensioning/

detensioning the reactor vessel head bolting studs and in accordance with the Surveillance Frequency Control Program (continued)

Amendment No.

SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.9.6 HATCH UNIT 2 SURVEILLANCE

N 0 TE ------------------------------

Only required to be met when the reactor vessel head is tensioned.

Verify reactor vessel flange and head flange temperatures are within the limits specified in the PTLR.

3.4-21 RCS PfT Limits 3.4.9 FREQUENCY Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature is

1 20°F in MODE 4, and in accordance with the Surveillance Frequency Control Program Once within 30 minutes after RCS temperature is

1 00°F in MODE 4, and in accordance with the Surveillance Frequency Control Program Amendment No.

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 0 Reactor Steam Dome Pressure Reactor Steam Dome Pressure 3.4.10 LCO 3.4.1 0 The reactor steam dome pressure shall be s 1 058 psig.

APPLICABILITY:

MODES 1 and 2.

ACTIONS CONDITION A.

Reactor steam dome A.1 pressure not within limit.

B.

Required Action and B.1 associated Completion Time not met.

SURVEILLANCE REQUIREMENTS REQUIRED ACTION Restore reactor steam dome pressure to within limit.

Be in MODE 3.

SURVEILLANCE SR 3.4. 1 0.1 HATCH UNIT 2 Verify reactor steam dome pressure is s 1 058 psig.

3.4-22 COMPLETION TIME 1 5 minutes 1 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months FREQUENCY In accordance with the Surveillance Frequency Control Program Amendment No. I

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel," (applicable amendment specified in the COLR).

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by LCO 3.3.3.1, "Post Accident Monitoring (PAM)

Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Reactor Coolant System CRCSl PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a.

RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, as well as heatup and cooldown rates, shall be established and documented in the PTLR for the following:

i.

Limiting Conditions for Operating Section 3.4.9 "RCS Pressure and Temperature (PT) Limits" ii.

Surveillance Requirements Section 3.4.9, "RCS Pressure and Temperature (PIT) Limits"

b.

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

HATCH UNIT 2

i.

BWROG-TP-1 1-022-A, Revision 1 (SIR-05-044, Revision 1-A),

"Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," dated June 201 3.

(continued) 5.0-21 Amendment No.

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued) ii.

BWROG-TP-1 1-023-A, Revision 0 (0900876.401, Revision 0-A),

"Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations," dated May 201 3.

c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

HATCH UNIT 2 5.0-22 Amendment No.

5.0 ADMINISTRATIVE CONTROLS

5. 7 High Radiation Area Reporting Requirements 5.7

5. 7.1 Pursuant to 10 CFR 20, paragraph 20.1601, in lieu of the requirements of 1 0 CFR 20.1 601 a, each high radiation area, as defined in 1 0 CFR 20, in which the intensity of radiation is > 1 00 mrem/hr but < 1 000 mrem/hr, measured at 30 em from the radiation source or from any surface the radiation penetrates, shall be barricaded and conspicuously posted as a high radiation area.

Entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g.,

Health Physics Technicians) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates

< 1 000 mrem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

A radiation monitoring device that continuously indicates the radiation dose rate in the area.

b.

A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.

c.

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physics supervision in the RWP.

5.7.2 In addition to the requirements of Specification 5.7.1, areas with radiation levels 2: 1 000 mrem/hr, measured at 30 em from the radiation source or from any surface the radiation penetrates, but less than 500 Rads in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months measured at 1 meter from the radiation source or from any surface that the radiation penetrates, shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervision on duty or Health Physics supervision.

HATCH UNIT 2 5.0-23 Amendment No. I

Edwin I. Hatch Nuclear Plant Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent with TSTF-419-A Proposed Technical Specification Bases Changes (Marked-Up)

RCS PIT Limits B 3.4.9 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.9 RCS Pressure and Temperature (PIT) Limits BASES BACKGROUND HATCH UNIT 1 All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

This SpesifisationThe PTLR contains PIT limit curves for non-nuclear heatup and cooldown, and inservice leakage and hydrostatic testing, and also limits the maximum rate of change of reactor coolant temperature. The criticality curve provides limits for both nuclear heatup and criticality.

Each PIT limit curve defines an acceptable region of operation for a particular operating condition. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure. Therefore, the LCO limits apply mainly to the vessel.

1 0 CFR 50, Appendix G (Ref. 1 ), requires the establishment of PIT limits for material fracture toughness requirements of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the ASME Code, Section Ill, Appendix G (Ref. 2).

The actual shift in the RT Nor of the vessel material is established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 1 85 (Ref. 3) and Appendix H of 1 0 CFR 50 (Ref. 4), and the BWR Vessel and Internals Project (VIP) Integrated Surveillance Program OSP) lRef. 14). The operating PIT limit curves are adjusted, as necessary, based on the evaluation findings and the recommendations of Reference 5.

(continued)

B 3.4-39 REVISION 1 2

BASES BACKGROUND (continued)

APPLICABLE SAFETY ANALYSES LCO HATCH UNIT 1 RCS PIT Limits B 3.4.9 The PIT limit curve for inservice leak and hydrostatic testing, and the curve for non-nuclear heatup and cooldown include separate curves for the bottom head, beltline, and upper vessel and flange regions.

These curves are derived from stress analysis of these vessel regions.

The criticality limits include the Reference 1 requirement that they be at least 40°F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic testing.

The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. ASME Code,Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

The PIT limits are not derived from Design Basis Accident (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed.

References 8 and 12 aroved the muves and limits sesified in this section. References 8 and 12 establish the methodology for Q.etermining the PIT limits. Since the PIT limits are not derived from any DBA, there are no acceptance limits related to the PIT limits.

Rather, the PIT limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.

RCS PIT limits satisfy Criterion 2 of the NRC Policy Statement (Ref. 9).

The elements of this LCO are:

a.

RCS pressure and temperature are within the limits specified in Figures 3.4.9 1 and 3.4.9 2the PTLR during RCS non nuclear heatup and cooldown operations, and RCS inservice leak and hydrostatic testing. Additionally, heatup and cooldown rates are s the limits specified in the PTLR during any RCS heatup or cooldown, and inservice leak and hydrostatic testing; (continued)

B 3.4-40 REVISION 12

BASES LCO (continued)

HATCH UNIT 1

b.

The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel (RPV) coolant is s 145°F during recirculation pump startup;

c.

The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel is s 50°F during recirculation pump startup;

d.

RCS pressure and temperature are within the criticality limits specified in F'ig1:1re 3.4.Q Jthe PTLR, prior to achieving criticality; and

e.

The reactor vessel flange and the head flange temperatures are > 7e0F'within the limits soeci.fuld in the PTLR when tensioning or detensioning the reactor vessel head bolting studs.

f.

The reactor vessel flange and head flange temperatures are A: 7e0F'within the limits specified in the PTLR when the reactor vessel head is tensioned.

g.

For the case when the vessel head is either off or on but not tensioned and fuel is in the vessel, all three sections of the vessel (upper vessel, beltline, and bottom head) may be lowered to a minimum of 68°F. When the head is being tensioned, or is already tensioned, the beltline and bottom head regions may be lowered to 68°F, as long as there is not any pressure or heatup/cooldown. The upper vessel, however, has a higher minimum temperature requirement with the head tensioned, as previously delineated.

The 68°F temperature is based on fuel shutdown margin considerations, since this is a more restrictive temperature than would be obtained from 1 0 CFR 50, Appendix G, considerations. With no fuel in the vessel, the temperature may drop to as low as 40°F, because this is the highest qualification temperature to meet toughness requirements for all reactor materials.

These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.

The rate of change of temperature limits controls the thermal gradient through the vessel wall and is used as input for calculating the heatup, cooldown, and inservice leakage and hydrostatic testing PIT limit curves. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the PIT limit curves.

(continued) 8 3.4-41 REVISION 1 2

BASES ACTIONS SURVEILLANCE REQUIREMENTS HATCH UNIT 1 C.1 and C.2 (continued)

RCS PfT Limits 8 3.4.9 be completed before approaching criticality or heating up to > 212°F.

Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components. ASME Code,Section XI, Appendix E (Ref. 6), may be used to support the evaluation; however, its use is restricted to evaluation of the beltline.

Condition C is modified by a Note requiring Required Action C.2 be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits.

Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

SR 3.4.9. 1 Verification that operation is within limits is required when RCS pressure and temperature conditions are undergoing planned changes. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.

Verification of Figures 3.4.9 1 and 3.4.9 2 that RCS pressure and RCS temperature are within the limits specified in the PTLR is required during non-nuclear heatups and cooldowns, and inservice leak and hydrostatic testing. Verification of the < 1 OOoF shange in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months periodthat the RCS heat!J.p and coold.ow ates are witblnJb._e limi s specified in the PTLR is required during any heatup or cooldown operations and during RCS inservice leak and hydrostatic testing.

SR 3.4.9.2 A separate figure is used when the reactor is critical. Consequently, the RCS pressure and temperature must be verified within the appropriate limits specified in the PTLR before withdrawing control rods that will make the reactor critical.

(continued) 8 3.4-44 REVISION 69

BASES SURVEILLANCE REQUIREMENTS REFERENCES HATCH UNIT 1 SR 3.4.9.5 and SR 3.4.9.6 (continued)

RCS PIT Limits B 3.4.9 SR 3.4.9.5 is modified by a Note that requires the Surveillance to be met only when tensioning/detensioning the reactor vessel head bolting studs. SR 3.4.9.6 is modified by a Note that requires the Surveillance to be met when the head is tensioned.

1.

1 0 CFR 50, Appendix G, January 1 996.

2.

ASME, Boiler and Pressure Vessel Code, Section Ill, Appendix G.

3.

ASTM E 1 85-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," July 1 982.

4.

1 0 CFR 50, Appendix H.

5.

Regulatory Guide 1.99, Revision 2, May 1 988.

6.

ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.

7.

FSAR, Section 14.3.6.2.

8.

BWROG-TP-1 1-02.2-A. Revision 1 (SIR-05-044, Revision 1-A>.

"Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", dated June George W. Rivenbark (NRC) letter to J. T. Beckham, Jr. (GPC), Amendment 126 to the Operating bisense, dated Jl:Jne 20, 1986.

9.

NRC No. 93-1 02, "Final Policy Statement on Technical Specification Improvements," July 23, 1 993.

1 0.

GE-NE-668-1 3-0393, "Recirculation Pump Restart Without Vessel Temperature Indication for E. I. Hatch Nuclear Plant,"

December 28, 1 993.

1 1.

DRF A00-05834/6, "Safety & 1 0 CFR 50.92 Significant Hazards Consideration Assessment for RPV Stratification Prevention Improvements at Edwin I. Hatch Nuclear Plant Units 1 and 2," Apri1 1 994.

12.

BWROG-TP-1 1 -023-A. RevisiQ.QQ.lft900876.401, Revision 0-B 3.4-47 REVISION 77

BASES REFERENCES (continued)

HATCH UNIT 1 1 3.

RCS Prr Limits B 3.4.9 A). "Lin.ear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations". dated May 201 3.Leonard N. Olshan (NRC) letter to H. L. Sumner, Jr.

(SNC), Amendment 222 to Operating License, dated August 2Q, 2000.

GE SIL 517, Supplement 1, "Analysis Basis for Idle Recirculation Loop Startup," August 26, 1 998.

14.

FSAR Appendix R B 3.4-48 REVISION 23

RCS Pff Limits B 3.4.9 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.9 RCS Pressure and Temperature (Pff) Limits BASES BACKGROUND HATCH UNIT 2 All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

This SpeoifioatienThe PTLR contains Pff limit curves for non-nuclear heatup and cooldown, and inservice leakage and hydrostatic testing, and also limits the maximum rate of change of reactor coolant temperature. The criticality curve provides limits for both nuclear heatup and criticality.

Each Pff limit curve defines an acceptable region of operation for a particular operating condition. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure. Therefore, the LCO limits apply mainly to the vessel.

1 0 CFR 50, Appendix G (Ref. 1 ), requires the establishment of Pff limits for material fracture toughness requirements of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the ASME Code, Section Ill, Appendix G (Ref. 2).

The actual shift in the RT NDT of the vessel material is established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 1 85 (Ref. 3) and Appendix H of 1 0 CFR 50 (Ref. 4). and the BWR Vessel and Internals Eroject <VIP) Integrated Surveillance Program !ISP) <Ref. 14). The operating Pff limit curves are adjusted, as necessary, based on the evaluation findings and the recommendations of Reference 5.

(continued)

B 3.4-40 REVISION 77

BASES BACKGROUND (continued)

APPLICABLE SAFETY ANALYSES LCO HATCH UNIT 2 RCS Pff Limits B 3.4.9 The Pff limit curve for inservice leak and hydrostatic testing, and the curve for non-nuclear heatup and cooldown include separate curves for the bottom head, beltline, and upper vessel and flange regions.

These curves are derived from stress analysis of these vessel regions.

The criticality limits include the Reference 1 requirement that they be at least 40°F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic testing.

The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. ASME Code,Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

The Pff limits are not derived from Design Basis Accident (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed.

References 8 and 12 approved the cblrves and limits specified in this section. References 8 a.rul.12 establish the methodology for determining the Pff limits. Since the Pff limits are not derived from any DBA, there are no acceptance limits related to the Pff limits.

Rather, the Pff limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.

RCS Pff limits satisfy Criterion 2 of the NRC Policy Statement (Ref. 8).

The elements of this LCO are:

a.

RCS pressure and temperature are within the limits specified in Figblres 3.4.9 1 and 3.4.9 2the PTLR during RCS non nuclear heatup and cooldown operations, and RCS inservice leak and hydrostatic testing. Additionally, heatup and cooldown rates are ::;; the limits specified in the PTLR during any RCS heatup or cooldown, and inservice leak and hydrostatic testing; (continued)

B 3.4-41 REVISION 77

BASES LCO (continued)

HATCH UNIT 2

b.

RCS PfT Limits B 3.4.9 The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel (RPV) coolant is s 145°F during recirculation pump startup;

c.

The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel is s 50°F during recirculation pump startup;

d.

RCS pressure and temperature are within the criticality limits specified in Fig;Jre 3.4.Q Jthe PTLR, prior to achieving criticality; and

e.

The reactor vessel flange and the head flange temperatures are > QOaFwithin the limits_specified in the PTLR when tensioning or detensioning the reactor vessel head bolting studs.

f.

The reactor vessel flange and head flange temperatures are

> QOaFwithin the limits specified in the PTLR when the reactor vessel head is tensioned.

g.

For the case when the vessel head is either off or on but not tensioned and fuel is in the vessel, all three sections of the vessel (upper vessel, beltline, and bottom head) may be lowered to a minimum of 68°F. When the head is being tensioned, or is already tensioned, the beltline and bottom head regions may be lowered to 68°F, as long as there is not any pressure or heatup/cooldown. The upper vessel, however, has a higher minimum temperature requirement with the head tensioned, as previously delineated.

The 68°F temperature is based on fuel shutdown margin considerations, since this is a more restrictive temperature than would be obtained from 1 0 CFR 50, Appendix G, considerations. With no fuel in the vessel, the temperature may drop to as low as 50°F, because this is the highest qualification temperature to meet toughness requirements for all reactor materials.

These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.

The rate of change of temperature limits controls the thermal gradient through the vessel wall and is used as input for calculating the heatup, cooldown, and inservice leakage and hydrostatic testing PfT limit curves. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of (continued)

B 3.4-42 REVISION 77

BASES ACTIONS SURVEILLANCE REQUIREMENTS HATCH UNIT 2 C. 1 and C.2 (continued)

RCS PIT Limits B 3.4.9 be completed before approaching criticality or heating up to > 212°F.

Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components. ASME Code,Section XI, Appendix E (Ref. 6), may be used to support the evaluation; however, its use is restricted to evaluation of the beltline.

Condition C is modified by a Note requiring Required Action C.2 be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits.

Restoration alone per Required Action C. 1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

SR 3.4.9.1 Verification that operation is within limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.

Verification ef F'ig1:1res 3.4.Q 1 and 3.4.Q 2 that RCS pressure and RCS temperature are within the limits specified in the PTLR is required during non-nuclear heatups and cooldowns, and inservice leak and hydrostatic testing. Verification ef the < 1 00°F' change in any 1 hel:lF peried that the RCS heatup and cooldown rates are within the limits specified in the PTLR is required during any heatup or cooldown operations and during RCS inservice leak and hydrostatic testing.

SR 3.4.9.2 A separate figure is used when the reactor is critical. Consequently, the RCS pressure and temperature must be verified within the appropriate limits specified in the PTLR before withdrawing control rods that will make the reactor critical.

(continued)

B 3.4-45 REVISION 79

BASES SURVEILLANCE REQUIREMENTS REFERENCES HATCH UNIT 2 SR 3.4.9.5 and SR 3.4.9.6 (continued)

RCS PIT Limits B 3.4.9 SR 3.4.9.5 is modified by a Note that requires the Surveillance to be met only when tensioning/detensioning the reactor vessel head bolting studs. SR 3.4.9.6 is modified by a Note that requires the Surveillance to be met when the head is tensioned.

1.

1 0 CFR 50, Appendix G, January 1 996.

2.

ASME, Boiler and Pressure Vessel Code, Section Ill, Appendix G.

3.

ASTM E 1 85-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," July 1 982.

4.

10 CFR 50, Appendix H.

5.

Regulatory Guide 1.99, Revision 2, May 1 988.

6.

ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.

7.

FSAR, Section 1 5. 1.26.

8.

BWROG-TP-11-022-A. Revision 1 (SIR-05-044. Revision 1-A>.

"Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", dated June 2013.Kahtan N. Jabbour (NRC) letter toW. G. Hairston, Ill (GPC), Arnenelrnent 118 to the Operating Lisense, elates January 10, 1QQ2.

9.

NRC No. 93-1 02, "Final Policy Statement on Technical Specification Improvements," July 23, 1 993.

1 0.

GE-NE-668-1 3-0393, "Recirculation Pump Restart Without Vessel Temperature Indication for E.l. Hatch Nuclear Plant,"

December 28, 1 993.

1 1.

DRF A00-05834/6, "Safety & 1 0 CFR 50.92 Significant Hazards Consideration Assessment for RPV Stratification Prevention Improvements at Edwin I. Hatch Nuclear Plant Units 1 and 2," April 1 994.

12.

Leonarel N. Olshan (NRC) letter to H. b. Sumner, Jr. (SNC),

B 3.4-48 REVISION 87

BASES REFERENCES (continued)

HATCH UNIT 2 RCS PIT Limits B 3.4.9 Amendment 163 to the Operating License, eatee August 29, 200(},. B\\lYRQG-TP-1 1-023-A. Revision 0 !0900876.40 1.

Revision 0-A). "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level lM.tr.ument Nozzles for Pressure-Temperature Curve Evaluations". dated May 2013.

(continued) I 1 3. GE SIL 51 7, Supplement 1, "Analysis Basis for Idle Recirculation Loop Startup," August 26, 1 998.

14. FSAR. Section 5.2.4.4 B 3.4-49 REVISION 77

Edwin I. Hatch Nuclear Plant Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent with TSTF-419-A Hatch Nuclear Plant, Unit 1 Pressure and Temperature Limits Report (PTLR) for 38 and 49.3 Effective Full-Power Years (EFPY)

Southern Nuclear Operating Co.

Hatch Nuclear Plant, Unit 1 Pressure and Temperature Limits Report (PTLR) for 3 8 and 49.3 Effective Full-Power Years (EFPY)

Revision 0

Table of Contents Section 1.0 Purpose 2.0 Applicability 3.0 Methodology 4.0 Operating Limits 5.0 Discussion 6.0 References Figure I HNP-I P-T Curve A (Hydrostatic Pressure and Leak Test) for 38 EFPY Figure 2 HNP-1 P-T Curve B (Normal Operation - Core Not Critical) for 38 EFPY Figure 3 HNP-1 P-T Curve C (Normal Operation - Core Critical) for 38 EFPY Figure 4 HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 49.3 EFPY Figure 5 HNP-I P-T Curve B (Normal Operation - Core Not Critical) for 49.3 EFPY Figure 6 H NP-1 P-T Curve C (Normal Operation - Core Critical) for 49.3 EFPY Hatch Unit I PTLR Revision 0 Page 2 of40 Page 4

4 5

6 7

I 2 1 5 I 6 I 7 I 8 I 9 20

Section Table 1 Table 2 Table 3 Table 4 Table 5 Table 6 Table 7 Table 8 Table 9 Appendix A Hatch Unit 1 PTLR Revision 0 Page 3 of 40 Page HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 38 2 1 EFPY HNP-1 P-T Curve B (Normal Operation - Core Not Critical) for 38 23 EFPY HNP-1 P-T Curve C (Normal Operation - Core Critical) for 38 EFPY 26 HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 49.3 29 EFPY HNP-1 P-T Curve B (Normal Operation - Core Not Critical) for 49.3 3 1 EFPY HNP-1 P-T Curve C (Normal Operation - Core Critical) for 49.3 EFPY 34 Hatch Unit I ART Table for 38 EFPY 37 Hatch Unit 1 ART Table for 49.3 EFPY 38 Hatch Unit 1 Summary ofNozzle Stress Intensity Factors 39 Hatch Unit I Reactor Vessel Materials Surveillance Program 40

1.0 Purpose Hatch Unit 1 PTLR Revision 0 Page 4 of 40 The purpose of the Hatch Nuclear Plant, Unit 1 (HNP-1 ) Pressure and Temperature Limits Report (PTLR) is to present operating limits relating to:

1. Reactor Coolant System (RCS) Pressure versus Temperature limits during Heat-up, Cool down and Hydrostatic/Class 1 Leak Testing;

2. RCS Heat-up and Cool-down rates;

3. RPV head flange boltup temperature limits.

This report has been prepared in accordance with the requirements of Licensing Topical Reports SIR-05-044, Revision I [ 1 ] and 0900876.401, Revision 0 [2].

2.0 Applicability This report is applicable to the HNP-1 RPV for up to 38 and 49.3 Effective Full-Power Years (EFPY) [3].

The following HNP-1 Technical Specification (TS) is affected by the information contained in this report:

Limiting Condition for Operation and Surveillance Requirement 3.4.9 ("RCS Pressure and Temperature (P/T) Limits")

3.0 Methodology The limits in this report were derived as follows:

Hatch Unit I PTLR Revision 0 Page 5 of 40 I. The methodology used is in accordance with Reference [ I ] and Reference [2], which have been approved by the NRC.

2. The neutron tluence is calculated in accordance with NRC Regulatory Guide 1. 1 90 (RG 1. 1 90) [4], using the RAMA computer code, as documented in Reference [5].

3. The adjusted reference temperature (ART) values for the limiting beltline materials are calculated in accordance with NRC Regulatory Guide 1.99, Revision 2 (RG 1.99) [6], as documented in Reference [7].

4. The pressure and temperature limits were calculated in accordance with Reference [ I ],

"Pressure - Temperature Limits Report Methodology for Boiling Water Reactors," June 201 3, as documented in Reference [8].

5. This revision of the pressure and temperature limits is to incorporate the following changes:

Initial issue of PTLR.

Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation tluence data of the RPV, or other plant design assumptions in the Updated Final Safety Analysis Report (UFSAR), can be made pursuant to I 0 CFR 50.59, provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance.

Changes to the curves, limits, or parameters within this PTLR, based upon new surveillance capsule data of the RPV or other plant design assumption modifications in the UFSAR, cannot

Hatch Unit I PTLR Revision 0 Page 6 of 40 be made without prior NRC approval. Such analysis and revisions shall be submitted to the NRC for review prior to incorporation into the PTLR.

4.0 Operating Limits The pressure-temperature (P-T) curves included in this report represent steam dome pressure versus minimum vessel metal temperature and incorporate the appropriate non-beltline limits and irradiation embrittlement effects in the beltline region.

The operating limits for pressure and temperature are required for three categories of operation:

(a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) core not critical operation, referred to as Curve 8; and (c) core critical operation, referred to as Curve C.

Complete P-T curves were developed for 38 EFPY and 49.3 EFPY for HNP-1 as documented in Reference [8]. The HNP-1 P-T curves for 38 EFPY are provided in Figures I through 3, and a tabulation of the overall composite curves (by region) is included in Tables I through 3. The HNP-1 P-T curves for 49.3 EFPY are provided in Figures 4 through 6, and a tabulation of the overall composite curves (by region) is included in Tables 4 through 6. The adjusted reference temperature (ART) values for the HNP-1 vessel beltline materials are shown in Table 7 for 38 EFPY and Table 8 for 49.3 EFPY, taken from Reference [7]. The resulting P-T curves are based on the geometry, design and materials information for the HNP-1 vessel with the following conditions:

Heat-up/Cool-down rate limit during Hydrostatic Class I Leak Testing (Figures I and 4:

Curve A): - 25"F/hour1 [8].

Normal Operating Heat-up/Cool-down rate limit (Figures 2 and 5: Curve 8 - non nuclear heating, and Figures 3 and 6: Curve C - nuclear heating): - I OO"F /hour2 [8].

1 Interpreted as the temperature change in any 1 -hour period is less than or equal to 25°F.

2 Interpreted as the temperature change in any 1 -hour period is less than or equal to 1 00°F.

Minimum bolt-up temperature limit 2: 76.F [8].

Hatch Unit I PTLR Revision 0 Page 7 of 40 To address the NRC condition regarding lowest service temperature in Reference [ 1 ], the minimum temperature is set to 76 °F, which is equal to the RTNDT.max + 60 °F, for all curves.

This value is consistent with the previous minimum temperature limits developed in [9], and is higher than previous minimum bolt-up specified in [ 1 0].

The composite P-T curves are extended below 0 psig to - 1 4.7 psig based on the evaluation documented in Reference [I I ], which demonstrates that the P-T curves are applicable to negative gauge pressures. Since the P-T curve calculation methods used do not specifically apply to negative values of pressure, the tabulated results start at 0 psi g. However, the minimum RPV pressure is - 1 4.7 psig.

5.0 Discussion The adjusted reference temperature (ART) of the limiting beltline material is used to adjust beltline P-T curves to account for irradiation effects. RG 1.99 [6] provides the methods for determining the ART. The RG 1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section.

The vessel beltline copper (Cu) and nickel (Ni) values were obtained from the evaluation of the HNP-1 vessel plate, weld, and forging materials [7]. The Cu and Ni values were used with Table 1 of RG 1.99 [6] to determine a chemistry factor (CF) per Paragraph 1. 1 of RG 1.99 for welds.

The Cu and Ni values were used with Table 2 of RG 1.99 to determine a CF per Paragraph 1. 1 of RG 1.99 for plates and forgings. However, for materials where credible surveillance data exists, a fitted CF may be used if it bounds the RG I. 99 CF.

The peak RPV ID fluence value of2.43 x 1 018 n/cm2 at 38 EFPY was developed in Reference [7]

based on linear interpolation between reported fluence values for 28.4 EFPY and 49.3 EFPY from Reference [5], which were calculated in accordance with RG 1. 1 90 [4]. The peak RPV ID tluence value of 3.08 x 1 018 n/cm2 at 49.3 EFPY was obtained from Reference [5] and was

Hatch Unit 1 PTLR Revision 0 Page 8 of 40 calculated in accordance with RG 1. 1 90. These fluence values apply to the limiting beltline lower intermediate shell plate (Heat No. C4 1 1 4-2). The fluence values for the limiting lower intermediate shell plate are based upon an attenuation factor of 0.724 for a postulated 1 /4T flaw.

As a result, the 1/4T fluence for 38 EFPY and 49.3 EFPY for the limiting lower intermediate shell plate are 1.76 x 1 018 n/cm2 and 2.23 x 1 018 n/cm2, respectively, for HNP-1.

The water level instrument (WLI) nozzle is located in the lower intermediate shell beltline plates

[8]. The limiting fluence values are as described in the paragraph above. Based on the ART evaluation in Reference [7], the recirculation inlet and outlet nozzles do not exist in the beltline regton.

The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 1 /4T (inside surface flaw) and 3/4T (outside surface flaw) locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1 /4T and the 3/4T locations. This is because the thermal gradient tensile stress of interest is in the inner wall during cool-down and is in the outer wall during heat-up. However, as a conservative simplification, the thermal gradient stresses at the 1 /4T location are assumed to be tensile for both heat-up and cool-down. This results in the approach of applying the maximum tensile stresses at the 1/4T location. This approach is conservative because irradiation effects cause the allowable toughness at the 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, which is well above the P-T curve limits.

For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves specify a coolant heat-up and cool-down temperature rate of :S 1 oo*F /hr for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound Service Level A/B RPV thermal transients. For the hydrostatic pressure and leak test curve (Curve A), a coolant heat-up and cool-down temperature rate of:S 25.F/hr must be

Hatch Unit 1 PTLR Revision 0 Page 9 of 40 maintained. The P-T limits and corresponding limits of either Curve A or B may be applied, if necessary, while achieving or recovering from test conditions. So, although Curve A applies during pressure testing, the limits of Curve B may be conservatively used during pressure testing if the pressure test heat-up/cool-down rate limits cannot be maintained.

The initial RT NDT, the chemistry (weight-percent copper and nickel) and ART at the l /4T location for all RPV beltline materials significantly affected by fluence (i.e., fluence > 1 017 n/cm2 for E > 1 MeV) are shown in Table 7 for 38 EFPY and Table 8 for 49.3 EFPY [7]. The initial RT NDT values shown in Tables 7 and 8 have been previously approved for use by the NRC per Reference [ 1 9].

Per Reference [7] and in accordance with Appendix A of Reference [ 1 ], the HNP-1 representative weld and plate surveillance materials data were reviewed from the Boiling Water Reactor Vessel and Internals Project (BWRVlP) Integrated Surveillance Program (ISP) [ 1 2].

The fitted CF for the limiting plate (Heat No. C4 1 14-2), which is based on credible surveillance data, in the HNP-1 vessel bounds the RG 1.99 CF [ 1 2]. Therefore, the fitted CF is used for the limiting beltline plate. In addition, an archival plate heat (Heat No. C3985-2) from the HNP-1 vessel was included in the Supplemental Surveillance program (SSP) and irradiated data from SSP Capsules H and C are provided in Reference [ 1 2]. These data are also determined to be credible, and, consequently, a reduced margin term is used for this material as well. The HNP-1 representative weld material (2029 1 ) is contained in the Cooper and SSP Capsule C capsules [7, 1 2]. Reference [ 1 2] contains surveillance capsule test results for the HNP-1 representative weld material; however, since the material heats for the HNP-1 limiting weld material and representative surveillance capsule weld material do not match, the CF calculated using the RG 1.99 [6] tables is used.

The ANSYS finite element computer program was used to develop the stress distributions through the feedwater (FW) nozzle [1 3]. These stress distributions were used in the

Hatch Unit 1 PTLR Revision 0 Page 1 0 of 40 determination of the stress intensity factors for the FW nozzles [ 1 4]. At the time the analyses were performed, the ANSYS program was controlled under the vendor's 1 0 CFR 50 Appendix 8

[ 1 5] Quality Assurance Program for nuclear quality-related work.

The plant-specific HNP-1 FW nozzle analysis was performed to determine stress intensity factors due to through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients [ 1 4]. Pressure and thermal stress distributions were taken from Reference [ 1 3]. Detailed information regarding the analysis can be found in References [ 1 3, 1 4].

The following summarizes the development of the thermal and pressure stress intensity factors for the FW nozzle [ 1 4] :

With respect to thermal stresses, the thermal shock which represents the maximum thermal shock for the FW nozzle during normal and upset operating conditions was analyzed [ 1 3]. The thermal stress distribution, corresponding to the limiting time point presented in [ 1 3 ], along a I in ear path through the nozzle corner is used [ 1 4]. Leakage is considered in the heat transfer calculations [ 1 3]. The thermal down shock of 450°F produces the highest tensile stresses at the 1/4T location. The BIE/IF methodology presented in the SI P-T Curve LTR [ 1 ] is used to calculate the thermal stress intensity, Kn, due to the thermal shock by fitting a third order polynomial equation to the path stress distribution for the thermal shock load case [ 1 4]. Because operation is along the saturation curve, the resulting KIT can be linearly scaled to determine the KIT to reflect the worst-case step change due to the available temperature difference. It is recognized that at low temperatures, the available temperature difference is insignificant and could potentially result in a near zero stress distribution. Therefore, a minimum KIT is calculated based on the thermal ramp of 1 00°F /hr, which is associated with the shutdown transient [ 1 4]. The resulting combination ofthe thermal down shock and thermal ramp Kn values represent the bounding thermal stress intensity factors in the FW nozzle associated with the P-T curves for the non-beltline region.

Hatch Unit 1 PTLR Revision 0 Page 1 1 of 40 Boundary conditions and heat transfer coefficients used for the thermal stress analysis are as described Reference [ 1 3a]. Overall heat transfer coefficients representative of a triple sleeve sparger with Seal No. 1 failed were applied [ 1 3a].

With respect to pressure stresses, a unit pressure of 1 000 psig was applied to the internal surfaces of the finite element model (FEM) [ 1 3]. The pressure stress distribution was taken along the same path as the thermal stress distribution. Recognizing that the Reference [ 1 3] evaluation was performed using a 2-0 axi-symmetric finite element model (FEM) and that it is known that the stress intensification caused by the nozzle geometry is under predicted in a 2-D axi-symmetric representation of the nozzle, a correction factor was applied to the stresses obtained from the 2-0 axi-symmetric FEM as described in Reference [ 1 4]. The BIE/IF methodology presented in the SI P-T Curve L TR [ 1 ] is used to calculate the pressure stress intensity factor, K1p, by fitting a third order polynomial equation to the path stress distribution for the pressure load case. The resulting K1p can be linearly scaled to determine the KIP for various RPV internal pressures.

Material properties were taken from the HNP-1 code of construction [ 1 6]. Use of temperature dependent material properties is expected to have minimal impact on the results of the analysis.

The following summarizes the development of the thermal and pressure stress intensity factors for the COP nozzle [ 1 4]:

The KIT term is calculated using the ASME XI, Non-mandatory Appendix G, Paragraph G-22 1 4.3 [ 1 7] methodology for a heat-up/cool-down rate of 1 00 *F /hr as described in Reference [ 1 4].

The K1P is calculated [ 1 4] using the WRC 1 75 methodology [ 1 8].

6.0 References Hatch Unit 1 PTLR Revision 0 Page 1 2 of40 1. BWROG-TP-1 1 -022-A, Revision 1 (SIR-05-044, Revision 1 -A), "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", dated June 20 13.

2. BWROG-TP-1 1 -023-A, Revision 0 (0900876.40 1, Revision 0-A), "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations", dated May 201 3.

3. Design Input Requests:

a. DIR, Revision 2, "Revised P-T Curves for Plant Hatch Units I &2," SI File No.

1 001 527.201.

b. DIR, Revision 0, "Hatch Units 1 and 2 P-T Curve Revisions," SI File No.

1 400365.200.

4. U. S. Nuclear Regulatory Commission, Regulatory Guide 1. 1 90, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", March 200 1.

5. Transware Enterprises Inc. Report No. SNC-HA 1 -002-R-00 1 Revision 0, "Edwin I.

Hatch Unit 1 Fluence Evaluation at End of Cycle 25 and 49.3 EFPY.".

6. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel materials", May 1 988.

7. Structural Integrity-Associates Calculation No. 1 00 1 527.30 1, Revision 1, "Hatch Unit 1 RPV Material Summary and ART Calculation", July 20 14.

8. Structural Integrity Associates Calculation No. 1 00 1 527.304, Revision 2, "Hatch Unit 1 P-T Curve Calculation for 38 and 49.3 EFPY", August 20 1 4.

9. General Electric Document No. GE-NE-B 1 1 00827-00-0 1, "Plant Hatch Units 1 & 2 RPV Pressure Temperature Limits License Renewal Evaluation," March 1 999.

Hatch Unit 1 PTLR Revision 0 Page 1 3 of40 1 0. NRC Docket No. 50-32 1, "Edwin I. Hatch Nuclear Plant Unit No. 1, Amendment to Facility Operating License," Amendment No. 59, License No. DPR-57, August 1 978, ADAMS Accession No. ML012950436.

1 1. Sl Calculation No. 1 400365.30 1, Rev. 0, "Hatch RPV Vacuum Assessment."

1 2. BWRVIP-1 35, Revision 2: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA: 2009.

1 02023 1. EPRI PROPRIETARY INFORMATION. Sl File No. BWRVIP-Ol -335P.

1 3. Hatch Unit 2 NUREG-061 9 Evaluations:

a. Liffengren, D. J., et al., "Edwin I. Hatch Nuclear Power Station, Unit 1 Feedwater Nozzle Fracture Mechanics Analysis to Show Compliance with NUREG-061 9,"

NEDE-30238, DRF-30238, August 1 983, General Electric Company. Sl File No.

1 00 1 527.2 1 0.

b. Bothne, D., "Power Uprate Evaluation Report for Edwin I. Hatch Unit 1,

Feedwater Nozzle NUREG-06 1 9 Fracture Mechanics Analysis for Extended Power Uprate Conditions," GE-NE-B 1 3-0 1 869-065-0 1, July 1 997, General Electric Company. Sl File No. 1 00 1 527.2 1 0 1 4. Structural Integrity Associates Calculation No. 1 001 527.303, Revision 0, "Feedwater, Water Level Instrument, and Core DP Nozzle Fracture Mechanics Evaluation for Hatch Unit 1 and Unit 2 Pressure-Temperature Limit Curve Development", December 20 1 1 1 5. U. S. Code of Federal Regulations, Title 1 0, Energy, Part 50, Appendix B, "Quality Assurance for Nuclear Power Plants and Fuel Reprocessing Plants".

1 6. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section I l l, 1 965 Ed. Winter 1 966 Addenda.

1 7. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section XI, Rules for lnservice Inspection ofNuclear Power Plant Components, Non-mandatory

Hatch Unit I PTLR Revision 0 Page 1 4 of 40 Appendix G, "Fracture Toughness Criteria for Protection Against Failure," 200 I Ed.

through 2003 Addenda.

1 8. PVRC Recommendations on Toughness Requirements for Ferritic Materials. WRC Bulletin 1 75. August 1 972.

1 9. NUREG-1 803, "Safety Evaluation Report Related to the License Renewal of the Edwin I.

Hatch Nuclear Plant, Units I and 2," December 200 1.

20. General Electric Report No. GE-NE-B I I 0069 1 -0 I RI, "Plant Hatch Unit I RPV Surveillance Materials Testing and Analysis," March 1 997. SI File No. 1 00 1 527.202.

2 1. Letter from L. N. Olshan (NRC) to H. L. Sumner (SNC), "Edwin I. Hatch Nuclear Plant, Units I and 2 Re: Issuance of Amendments (TAC NOS. MB6 1 06 and MB6 1 07)", March 1 0, 2003.

22. BWRVIP-86, Revision 1 -A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan EPRI Product I 025 1 44, October 201 2.

Hatch Unit I PTLR Revision 0 Page 1 5 of 40 Figure 1 : HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 38 EFPY BOO 1200 1100 1000 900 Dii BOO ill 1 700 Cll

nl Cll 600 1¥

.5

'E

I 500 Cll

Gl A 400 300 200 100 0

Curve A - Pressure Test, Composite Curves

-- Beltline

Bottom Head Non-Beltline Overall I

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I Minimum Bolt-Up Temperature = 76°F I

Minimum RPV Pressure = -14.7 psig

'0 1)0 150 2L 2;()

3)()

Minimum Reactor Vessel Metal Temperature ("F)

Hatch Unit 1 PTLR Revision 0 Page 1 6 of 40 Figure 2: HNP-1 P-T Curve B (Normal Operation - Core Not Critical) for 38 EFPY Curve B - Core Not Critical, Composite Curves

-- Beltline ---

Bottom Head - -

Non-Beltline

-Overall 1300 ---

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1200 +-----f-----.--------+-,,.---

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300 +------1-----, r---

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01 Minimum Bolt-Up Temperature = 76°F 100 --t--- --1 --+-----l-1-----1-r"-=-====]

Minimum RPV Pressure = -14_7 psig 0 r---

T 1

150 210 Minimum Reactor Vessel Metal Temperature ('F)

Hatch Unit 1 PTLR Revision 0 Page 1 7 of 40 Figure 3: HNP-1 P-T Curve C (Normal Operation - Core Critical) for 38 EFPY Curve C - Core Critical, Composite Curves

-- Beltline

- - - - Bottom Head Non-Beltline

-overall 1300 I

r!

I I

I 1200 I

I I

I 1100 1000 I

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900 I

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400 300 200 100 j

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Minimum Bolt-Up l

t Temperature = 76"F ld I

M ini mum RPV Pressure = -14.7 psig 0

0 1fWJ 150 2)()

2ø 3)()

Minimum Reactor Vessel Metal Temperature ("F)

Hatch Unit I PTLR Revision 0 Page 1 8 of 40 Figure 4: HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 49.3 EFPY Curve A - Pressure Test, Composite Curves

-- Beltline

Bottom Head Non-Beltline

-overall 1300 I

I E

I I

1200 I

I I

I 1100 I

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I 1000 I

I 900 "iii 800 ill

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IV II 600 a:

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400 I

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/ I 300 200 Minimum Bolt-U p Temperature = 76°F 100 Minimum RPV Pressu re = -14.7 psig 0

'0 1

1i 2

2j()

3 Minimum Reactor Vessel Metal Temperature ("F)

Hatch Unit 1 PTLR Revision 0 Page 1 9 of 40 Figure 5: HNP-1 P-T Curve B (Normal Operation - Core Not Critical) for 49.3 EFPY Curve B - Core Not Critical, Composite Curves

-- Beltline

Bottom Head Non-Beltline

-overall 1300 I

rr I

1200 I

I 1100 1000 I

1:'

I I

900 I

I I

I iii 800 Vi

"§ 700 II 15 t:..

II 600 Ill:

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Minimum Bolt-Up Temperature = 76"F 100 0

P Minimum RPV Pressure = -14.7 psig

-f 0

1l0 1i 2)0 3)o Minimum Reactor Vessel Metal Temperature ("F)

Hatch Unit I PTLR Revision 0 Page 20 of40 Figure 6: HNP-1 P-T Curve C (Normal Operation - Core Critical) for 49.3 EFPY 1300 1200 1 100 1000 900 iii 800 WI E;

700 Cll g

Cll 600 Ill!

r--

'E

I 500 Cll

Ill f

400 300 200 100 0

Curve C - Core Critical, Composite Curves

-- Beltline

Bottom Head Non-Beltline

-overall

!p I

I I

i 1

I ' ' '

I I

I I ' '

I I

l

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Minimum Bolt-U p Temperature = 76"F M inimum RPV Pressure = -14.7 psig 1

150 2L t

I I

Minimum Reactor Vessel Metal Temperature ("F) 31Q

Hatch Unit I PTLR Revision 0 Page 2 1 of 40 Table 1: HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 38 EFPY Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure "F

psi 76.0 0.0 76.0 365.2 98.8 415.1 114.4 465.0 126.3 5 14.9 135.9 564.8 144.0 614.7 150.9 664.6 157.0 714.5 162.4 764.5 167.3 814.4 171.8 864.3 175.8 914.2 179.6 964.1 183.1 1014.0 186.4 1063.9 189.5 1 1 13.8 192.4 1163.7 195.2 1213.6 197.8 1263.5 200.2 1313.5 202.6 1363.4

Hatch Unit 1 PTLR Revision 0 Page 22 of 40 Table 1: HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 38 EFPY (continued)

Bottom Head Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure OF psi 76.0 0.0 76.0 1226.1 78.7 1274.2 81.2 1322.4 83.6 1370.5 Non-Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure OF psi 76.0 0.0 76.0 312.6 106.0 312.6 106.0 934.2 109.5 982.6 112.7 1031.0 115.7 1079.3 118.6 1127.7 121.3 1176.1 123.9 1224.4 126.3 1272.8 128.6 1321.2

Hatch Unit I PTLR Revision 0 Page 23 of 40 Table 2: HNP-1 P-T Curve B (Normal Operation - Core Not Critical) for 38 EFPY Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure "F

psi 76.0 0.0 76.0 144.9 104.2 193.8 122.1 242.7 135.2 291.6 145.6 340.5 154.2 389.4 161.6 438.3 168.0 487.2 173.6 536.1 178.7 585.0 183.4 633.9 187.6 682.8 191.5 731.7 195.1 780.6 198.5 829.5 201.6 878.4 204.6 927.3 207.4 976.2 210.1 1025.1 212.6 1074.0 215.0 1122.9 2 17.3 1171.8 2 19.5 1220.7 221.6 1269.6 223.6 1318.5

Hatch Unit I PTLR Revision 0 Page 24 of 40 Table 2: HNP-1 P-T Curve B (Normal Operation - Core Not Critical) for 38 EFPY (continued)

Bottom Head Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure "F

psi 76.0 0.0 76.0 813.9 79.6 863.8 83.0 913.8 86.2 963.7 89.2 1013.6 92.0 1063.6 94.7 1113.5 97.3 1163.5 99.7 1213.4 102.0 1263.3 104.2 1313.3

Hatch Unit 1 PTLR Revision 0 Page 25 of 40 Table 2: HNP-1 P-T Curve B (Normal Operation - Core Not Critical) for 38 EFPY (continued)

Non-Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure "F

psi 76.0 0.0 76.0 198.2 84.6 236.4 91.8 274.5 97.9 3 12.6 136.0 3 12.6 136.0 724.2 139.0 773.5 141.8 822.9 144.4 872.2 146.8 921.5 149.2 970.9 151.4 1020.2 153.6 1069.6 155.6 1118.9 157.6 1168.3 159.4 1217.6 161.2 1266.9 163.0 1316.3

Hatch Unit 1 PTLR Revision 0 Page 26 of40 Table 3: HNP-1 P-T Curve C (Normal Operation - Core Critical) for 38 EFPY Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure "F

psi 76.0 0.0 76.0 109.3 125.1 157.8 149.3 206.2 165.6 254.7 177.9 303.1 187.7 351.6 195.9 400.0 203.0 448.5 209.1 497.0 214.6 545.4 219.6 593.9 224.1 642.3 228.2 690.8 232.1 739.2 235.6 787.7 238.9 836.1 242.0 884.6 245.0 933.1 247.7 981.5 250.3 1030.0 252.8 1078.4 255.2 1126.9 257.5 1175.3 259.6 1223.8 261.7 1272.3 263.7 1320.7

Hatch Unit 1 PTLR Revision 0 Page 27 of 40 Table 3: HNP-1 P-T Curve C (Normal Operation - Core Critical) for 38 EFPY (continued)

Bottom Head Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure

F psi 76.0 0.0 76.0 450.5 83.6 498.8 90.1 547.2 95.9 595.6 101.1 643.9 105.8 692.3 110. 1 740.7 114.1 789.1 117.8 837.4 121.2 885.8 124.4 934.2 127.4 982.5 130.2 1030.9 132.9 1079.3 135.4 1 127.7 137.9 1 176.0 140.2 1224.4 142.4 1272.8 144.5 1321.1

Hatch Unit l PTL R Revision 0 Page 28 of 40 Table 3: HNP-1 P-T Curve C (Normal Operation - Core Critical) for 38 EFPY (continued)

Non-Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure "F

psi 76.0 0.0 76.0 97.6 98.0 140.6 112.1 183.6 122.6 226.6 130.9 269.6 137.9 312.6 217.0 312.6 217.0 1563.0

Hatch Unit 1 PTLR Revision 0 Page 29 of 40 Table 4: HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 49.3 EFPY Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure

F psi 76.0 0.0 76.0 345.8 103.3 394.5 120.9 443.2 133.9 491.9 144.2 540.6 152.7 589.3 160.0 638.0 166.4 686.6 172.0 735.3 177.1 784.0 181.7 832.7 185.9 881.4 189.8 930.1 193.4 978.8 196.7 1027.4 199.9 1076.1 202.9 1124.8 205.7 1173.5 208.3 1222.2 210.8 1270.9 213.2 1319.6

Hatch Unit 1 PTLR Revision 0 Page 30 of 40 Table 4: HNP-1 P-T Curve A (Hydrostatic Pressure and Leak Test) for 49.3 EFPY (continued)

Bottom Head Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure "F

psi 76.0 0.0 76.0 1226.1 78.7 1274.2 81.2 1322.4 83.6 1370.5 Non-Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure "F

psi 76.0 0.0 76.0 312.6 106.0 312.6 106.0 934.2 109.5 982.6 112.7 1031.0 115.7 1079.3 118.6 1127.7 121.3 1176.1 123.9 1224.4 126.3 1272.8 128.6 1321.2

Hatch Unit l PTLR Revision 0 Page 3 1 of40 Table 5: HNP-1 P-T Curve B (Normal Operation - Core Not Critical) for 49.3 EFPY Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

F psi 76.0 0.0 76.0 130.4 110.3 179.8 130.4 229.2 144.7 278.6 155.9 328.0 164.9 377.4 172.6 426.8 179.3 476.2 185. 2 525.6 190.4 575.0 195. 2 624.4 199.5 673.8 203.5 723.2 207.2 772.6 210.7 822.0 213.9 871.4 216.9 920.8 219.8 970.2 222.5 1019.6 225.1 1069.0 227.5 1118.4 229.8 1167.8 232.0 1217.2 234.2 1266.6 236.2 1316.0

Hatch Unit I PTLR Revision 0 Page 32 of 40 Table 5: HNP-1 P-T Curve B (Normal Operation - Core Not Critical) for 49.3 EFPY (continued)

Bottom Head Region Curve 8 - Core Not Critical P-T Curve P-T Curve Temperature Pressure "F

psi 76.0 0.0 76.0 813.9 79.6 863.8 83.0 913.8 86.2 963.7 89.2 1013.6 92.0 1063.6 94.7 1113.5 97.3 1163.5 99.7 1213.4 102.0 1263.3 104.2 1313.3

Hatch Unit I PTLR Revision 0 Page 33 of 40 Table 5: HNP-1 P-T Curve B (Normal Operation - Core Not Critical) for 49.3 EFPY (continued)

Non-Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure "F

psi 76.0 0.0 76.0 198.2 84.6 236.4 91.8 274.5 97.9 312.6 136.0 312.6 136.0 724.2 139.0 773.5 141.8 822.9 144.4 872.2 146.8 921.5 149.2 970.9 151.4 1020.2 153.6 1069.6 155.6 1118.9 157.6 1168.3 159.4 1217.6 161.2 1266.9 163.0 1316.3

Hatch Unit I PTLR Revision 0 Page 34 of 40 Table 6: HNP-1 P-T Curve C (Normal Operation - Core Critical) for 49.3 EFPY Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure "F

psi 76.0 0.0 76.0 102.8 133.5 151.5 159.6 200.1 176.6 248.8 189.3 297.5 199.4 346.1 207.8 394.8 215.0 443.5 221.3 492.2 226.9 540.8 231.9 589.5 236.4 638.2 240.6 686.9 244.5 735.5 248.1 784.2 251.4 832.9 254.5 881.6 257.5 930.2 260.3 978.9 262.9 1027.6 265.4 1076.3 267.8 1124.9 270.1 1 173.6 272.3 1222.3 274.3 1271.0 276.3 1319.6

Hatch Unit 1 PTLR Revision 0 Page 35 of 40 Table 6: HNP-1 P-T Curve C (Normal Operation - Core Critical) for 49.3 EFPY (continued)

Bottom Head Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure "F

psi 76.0 0.0 76.0 450.5 83.6 498.8 90.1 547.2 95.9 595.6 101.1 643.9 105.8 692.3 110.1 740.7 114.1 789.1 117.8 837.4 121.2 885.8 124.4 934.2 127.4 982.5 130.2 1030.9 132.9 1079.3 135.4 1127.7 137.9 1176.0 140.2 1224.4 142.4 1272.8 144.5 1321.1

Hatch Unit 1 PTLR Revision 0 Page 36 of 40 Table 6: HNP-1 P-T Curve C (Normal Operation - Core Critical) for 49.3 EFPY (continued)

Non-Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure "F

psi 76.0 0.0 76.0 97.6 98.0 140.6 112.1 183.6 122.6 226.6 130.9 269.6 137.9 312.6 217.0 312.6 217.0 1563.0

Hatch Unit 1 PTLR Revision 0 Page 37 of 40 Description Lower Shell #1 ABC Lower Shell #2 Lower Shell#3

£ 1 Lower-In!. Shell #1 Lower-In! Shell #2 Lower-lnt Shell #3 Dea:rlpllon Lower Long. Weld ij Lower Int. Long Weld #2 Ulll Lower Int. Long Weld #1 3:

Lower - Lower Int. Girth Weld #1 Lower - Lower Int. Girth Weld #2 Ill Table 7: Hatch Unit 1 ART Table for 38 EFPY I Code No. I Heat No. I Flux Type f805-1 C4112-1 f805-2 C4112-2 G-4805-3 C4149-1 G-4803-7 C4337-1 G-4804-1 C3985-2 G-4804-2 C4114-2 I Code No. I Heat No. I Flux Type 1-307 13253/1092 1-308 1P2809/1092 1-308 1P281511092 1-313 90099/0091 1-313 33A277/0091 Flux Lot No.

Flux Lot No.

3791 3854 3854 3977 3977 Initial RTNoT ("F)

Chemistry Cu (wt %1 Nl (wt %1 8

0.13 0.64 10

0. 13 0.64

-10 0.14 0.57

-20 0.17 0.62

-20 0.11 0.60

-20 0.12 0.70 Initial RTNor("F)

Chemistry Cu (wt %1 Nl (wt %1

-50 0.221 0.732

-50 0.270 0.735

-50 0.316 0.724

-10 0.197 0.060

-50 0.258 0.165 Fluence Data Chemistry Factor rFI 92 92 99 128 65 221 Chemistry Factor rFI 189 206 219 91 126 ART NoT

("F) 44.8 44.8 48.0 68.7 34.9 119.3 ART NoT rFI 91.5 89.8 95.5 45.7 63.2 Adjustments for 114t Margin Terms a1 ("F) ao1 rFI 0.0 17.0 0.0 17.0 0.0 17.0 0.0 17.0 0.0 8.5 0.0 8.5 Adjustments for 114t Margin Terms a1 ("F) aA rFI 0.0 28.0 0.0 28.0 0.0 28.0 0.0 22.9 0.0 28.0 ART NoT rFI 86.8 88.8 72.0 82.7 31.9 116.3 ARTNDT rFI 97.5 95.8 101.5 81.4 69.2 Location Wall Thickness On.)

Full 1/4t Fluenee at iD (nlem2)

AtiDnuatlon, 1/4t = *.0.2Aa Fluence @ 1/4t (nlem2)

Fluence Factor, FF f10.28.0.101og I)

Lower Shell #1 6.375 1.594 2.05E+18 0.682 1.40E+18 0.487 Lower Shell #2 6.375 1.594 2.05E+18 0.682 1.40E+16 0.467 8111 Lower Shell#3 6.375 1.594 2.05E+16 0.682 1.40E+16 0.467

£ jl0Wer:inlgsiieii#1----------------*---s.37s

____________ 1.344----------2h43E+1ii

______ ----o:724---------i7sE+1ii-------------o.s39 ____ _

!II Lower-In! Shell #2 5.375 1.344 2.43E+18 0.724 1.76E+18 0.539 Lower-In! Shell #3 5.375 1.344 2.43E+18 0.724 1.76E+18 0.539 Lower Long. Weld 6.375 1.594 2.02E+16 0.682 1.38E+18 0.484 Lower Int. Long Weld #1 5.375 1.344 1.52E+16 0.724 1.10E+16 0.437 Lower Int. Long Weld #2 5.375 1.344 1.52E+18 I*

0.724 1.10E+16 0.437 Lower - Lower Int. Girth Weld #1 5.375 1.344 2.05E+16 0.724 i

1.48E+18 0.500 Lower - Lower Int. Girth Weld #2 5.375 1 344 2.05E+18 0.724 1.48E+18 0.500 1. W GE CF " 236 "F is used then this location becomes the limiting beltline location by 7.7 'F o-.er the current limiting location.

II II

]I a:

@I Table 8: Hatch Unit 1 ART Table for 49.3 EFPY Chemistry Description Code No.

Heat No. /

Flux Lot No.

Initial RTNorrFI Flux Type Cu (wt %1 Nl (wt %1 Lower Shell #1 G-4805-1 C41 12-1 8

0.13 0.64 Lower Shell #2 G-4805-2 C41 1 2-2 1 0

0. 13 0.64 Lower Shell#3 G-4805-3 C4149-1

-10 0.14 0.57 Lower-tnt. Shell #1 G-4803-7 C4337-1

-20 0.17 0.62 Lower-In! Shell #2 G-4804-1 C3985-2

-20

0. 1 1 0.60 Lower-tnt Shell #3 G-4804-2 C4114-2

-20 0.12 0.70 Chemistry Description Code No.

Heat No. /

Flux Lot No.

Initial RTNorrFI Flux Type Cu (wt %1 Nl (wt %1 Lower Long. Weld 1-307 1 3253/1092 3791

-50 0.221 0.732 Lower Int. Long Weld #1 1-308 1P2809/1 092 3854

-50 0.270 0.735 Lower Int. Long Weld #2 1-308 1 P2815/1092 3854

-50 0.316 0.724 Lower - Lower Int. Girth Weld #1 1-313 90099/0091 3977

-10 0, 197 0 060 Lower - Lower Int. Girth Weld #2 11>

1-313 33A277/0091 3977

-50 0,258

0. 165 Fluence Data Wall Thickness (ln.)

Fluence at ID Atlllnuatlon, Location Full 1/4t (nlcm 1/4t= e-G-2..,.

Lower Shell #1 6.375 1.594 2.56E+18 0.682 Lower Shell #2

{

6.375 1.594 2.56E+18 0.682 Lower Shell#3 6.375 1.594 2.56E+1 8 0.682 Lower-In!. Shell #1 5.375 1.344 3.08E+18 0.724 Lower-In! Shell #2 5.375 1.344 3.08E+18 0.724 Lower-In! Shell #3 5 375 1.344 3.08E+1 8 0.724 Lower Long. Weld l

6.375 1. 594 2.54E+1 8 0.682 Lower Int. Long Weld #1 5.375 1. 344 1.95E+18 0.724 Lower Int. Long Weld #2 5.375 1. 344 1.95E+18 0.724 Lower - Lower Int. Girth Weld #1 5.375 1. 344 2.56E+18 0.724 Lower - Lower Int. Girth Weld #2 5.375 1. 344 2.56E+1 8 0.724 1. W GE CF = 236 'F is used then this location becomes the limiting beltline location by 7.2 'F 01.er the current limiting location.

Chemistry Factor rFI 4RTNDT rFI 92 49.4 92 49.4 99 53.0 128 76.0 65 38.5 221 132.0 Chemistry Factor rFI 4RTNDT rFI 1 89 101.2 206 100 6 219 107.0 91 50.4 126 69.6 Hatch Unit 1 PTLR Revision 0 Page 38 of 40 Adjuslments for 1/4t Margin Tenns ARTNDT en rFI erA rF) rFI 0.0 17.0 91.4 0.0 17.0 93.4 0.0 17.0 77.0 0.0 17.0 90.0 I

0.0 8.5 35.5 0.0 8.5 129.0 Adjuslments for 1/4t Margin Tenns ARTNDT er1 rFI erA rFI rFI 0.0 28.0 107.2 0.0 28.0 106.6 0.0 28.0 1 1 3.0 0.0 25.2 90.8 0.0 28.0 75.6 Fluence Factor, FF Fluence @l 114t (nlcm2) fC0-28-0.101og q 1.75E+18 t

0.537 1.75E+1 8 0.537 1.75E+1 8 0.537 2.23E+1 8 0.596 2.23E+1 8 0.596 2.23E+1 8 0.596 1.73E+1 8 0.536 1.41E+18 0.489 1.41E+18 i

0.489 1.85E+1 8 I

0.551 1.85E+1 8 0.551

Table 9: Hatch Unit 1 Summary of Nozzle Stress Intensity Factors Nozzle Applied Pressure, Thermal, Krt Thermal, Krt K1p..apo (450"F shock)

(100 "F/hr Plate)

Feedwater 76.6 65.3 11.5 W LI 71.6 N/A 17.4 Core DP 32.3 N/A

1.7 Notes

1. K1 in units of ksi-in°*5 Hatch Unit I PTLR Revision 0 Page 39 of 40

Appendix A Hatch Unit I PTLR Revision 0 Page 40 of 40 HATCH UNIT 1 REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with I 0 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, two surveillance capsules have been removed from the Hatch Nuclear Plant Unit I (HNP-I ) reactor vessel. The first capsule was removed in I 984 after 5.75 EFPY and the second was removed in I 996 after 1 4.3 EFPY [20]. The surveillance capsule contained flux wires for neutron fluence measurement, Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel materials within the core beltline region [20].

Southern Nuclear Operating Company committed to use the ISP in the amendments issued by the NRC regarding the implementation of the Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Integrated Surveillance Program, dated March I O, 2003 [2 I]. The BWRVIP ISP meets the requirements of 1 0 CFR 50, Appendix H, for Integrated Surveillance Programs, and has been approved by NRC for the period of extended operation [22]. HNP-I continues to be a host plant under the ISP [ I 2]. Two more HNP-I capsules are scheduled to be removed and tested under the ISP in approximately 20 I 6 and 2029.

Edwin I. Hatch Nuclear Plant Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent with TSTF-419-A Hatch Nuclear Plant, Unit 2 Pressure and Temperature Limits Report (PTLR) for 37 and 50.1 Effective Full-Power Years (EFPY)

Southern Nuclear Operating Co.

Hatch Nuclear Plant, Unit 2 Pressure and Temperature Limits Report (PTLR) for 37 and 50. 1 Effective Full-Power Years (EFPY)

Revision 0

Table of Contents Section 1.0 Purpose 2.0 Applicability 3.0 Methodology 4.0 Operating Limits 5.0 Discussion 6.0 References Figure 1 HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 37 EFPY Figure 2 HNP-2 P-T Curve B (Normal Operation - Core Not Critical) for 37 EFPY Figure 3 HNP-2 P-T Curve C (Normal Operation - Core Critical) for 37 EFPY Figure 4 HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 50. 1 EFPY Figure 5 HNP-2 P-T Curve B (Normal Operation - Core Not Critical) for 50. 1 EFPY Figure 6 HNP-2 P-T Curve C (Normal Operation - Core Critical) for 50. 1 EFPY Hatch Unit 2 PTLR Revision 0 Page 2 of 40 Page 4

4 5

6 7

1 2 1 5 1 6 1 7 1 8 1 9 20

Section Table I Table 2 Table 3 Table 4 Table 5 Table 6 Table 7 Table 8 Table 9 Appendix A Hatch Unit 2 PTLR Revision 0 Page 3 of 40 Page HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 37 2 1 EFPY HNP-2 P-T Curve B (Normal Operation - Core Not Critical) for 37 23 EFPY HNP-2 P-T Curve C (Normal Operation - Core Critical) for 3 7 EFPY 26 HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 50. 1 29 EFPY HNP-2 P-T Curve B (Normal Operation - Core Not Critical) for 50. 1 3 1 EFPY HNP-2 P-T Curve C (Normal Operation - Core Critical) for 50. 1 EFPY 34 Hatch Unit 2 ART Table for 37 EFPY 37 Hatch Unit 2 ART Table for 50. 1 EFPY 38 Hatch Unit 2 Summary ofNozzle Stress Intensity Factors 39 Hatch Unit 2 Reactor Vessel Materials Surveillance Program 40

1.0 Purpose Hatch Unit 2 PTLR Revision 0 Page 4 of 40 The purpose of the Hatch Nuclear Plant, Unit 2 (HNP-2) Pressure and Temperature Limits Report (PTLR) is to present operating limits relating to:

I. Reactor Coolant System (RCS) Pressure versus Temperature limits during Heat-up, Cool down and Hydrostatic/Class I Leak Testing;

2. RCS Heat-up and Cool-down rates;

3. RPV head flange boltup temperature limits.

This report has been prepared in accordance with the requirements of Licensing Topical Reports SIR-05-044, Revision I [ I ] and 0900876.40 I, Revision 0 [2].

2.0 Applicability This report is applicable to the HNP-2 RPV for up to 37 and 50. 1 Effective Full-Power Years (EFPY) (3].

The following HNP-2 Technical Specifications (TS) are affected by the information contained in this report:

Limiting Condition for Operation and Surveillance Requirement 3.4.9 ("RCS Pressure and Temperature (P/T) Limits")

3.0 Methodology The limits in this report were derived as follows:

Hatch Unit 2 PTLR Revision 0 Page 5 of 40 1. The methodology used is in accordance with Reference [ 1 ] and Reference [2], which have been approved by the NRC.

2. The neutron fluence is calculated in accordance with NRC Regulatory Guide 1. 1 90 (RG 1. 1 90) [4], using the RAMA computer code, as documented in Reference [5].

3. The adjusted reference temperature (ART) values for the limiting beltline materials are calculated in accordance with NRC Regulatory Guide 1.99, Revision 2 (RG 1.99) [6], as documented in Reference [7].

4. The pressure and temperature limits were calculated in accordance with Reference [ I ],

"Pressure - Temperature Limits Report Methodology for Boiling Water Reactors " June 20 1 3, as documented in Reference [8].

5. This revision of the pressure and temperature limits is to incorporate the following changes:

Initial issue of PTLR.

Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation fluence data of the RPY, or other plant design assumptions in the Updated Final Safety Analysis Report (UFSAR), can be made pursuant to 1 0 CFR 50.59, provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance.

Changes to the curves, limits, or parameters within this PTLR, based upon new surveillance capsule data of the RPY or other plant design assumption modifications in the UFSAR, cannot

Hatch Unit 2 PTLR Revision 0 Page 6 of 40 be made without prior NRC approval. Such analysis and revisions shall be submitted to the NRC for review prior to incorporation into the PTLR.

4.0 Operating Limits The pressure-temperature (P-T) curves included in this report represent steam dome pressure versus minimum vessel metal temperature and incorporate the appropriate non-beltline limits and irradiation embrittlement effects in the beltline region.

The operating limits for pressure and temperature are required for three categories of operation:

(a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) core not critical operation, referred to as Curve 8; and (c) core critical operation, referred to as Curve C.

Complete P-T curves were developed for 3 7 EFPY and 50. 1 EFPY for HNP-2 as documented in Reference [8]. The HNP-2 P-T curves for 37 EFPY are provided in Figures I through 3, and a tabulation of the overall composite curves (by region) is included in Tables I through 3. The HNP-2 P-T curves for 50. 1 EFPY are provided in Figures 4 through 6, and a tabulation of the overall composite curves (by region) is included in Tables 4 through 6. The adjusted reference temperature (ART) values for the HNP-2 vessel beltline materials are shown in Table 7 for 37 EFPY and Table 8 for 50. 1 EFPY, taken from Reference [7]. The resulting P-T curves are based on the geometry, design and materials information for the HNP-2 vessel with the following conditions:

Heat-up/Cool-down{ate limit during Hydrostatic Class I Leak Testing (Figures I and 4:

Curve A): 25.F/hour1 [8].

Normal Operating Heat-up/Cool-down rate limit (Figures 2 and 5: Curve 8 - non nuclear heating, and Figures 3 and 6: Curve C - nuclear heating): - I oo*F/hour2 [8].

1 Interpreted as the temperature change in any 1 -hour period is less than or equal to 25°F.

Interpreted as the temperature change in any 1 -hour period is less than or equal to I 00°F.

Minimum bolt-up temperature limit 90"F [8].

Hatch Unit 2 PTLR Revision 0 Page 7 of 40 To address the NRC condition regarding lowest service temperature in Reference [ 1 ], the minimum temperature is set to 90 °F, which is equal to the RT NDT.max + 60 °F, for all curves.

This value is consistent with the previous minimum temperature limits developed in [9] and the minimum bolt-up temperature specified in [ I 0].

The composite P-T curves are extended below 0 psig to - 1 4.7 psig based on the evaluation documented in Reference [ I I ], which demonstrates that the P-T curves are applicable to negative gauge pressures. Since the P-T curve calculation methods used do not specifically apply to negative values of pressure, the tabulated results start at 0 psig. However, the minimum RPV pressure is - 1 4.7 psi g.

5.0 Discussion The adjusted reference temperature (ART) ofthe limiting beltline material is used to adjust beltline P-T curves to account for irradiation effects. RG 1.99 [6] provides the methods for determining the ART. The RG 1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section.

The vessel beltline copper (Cu) and nickel (Ni) values were obtained from the evaluation of the HNP-2 vessel plate, weld, and forging materials [7]. The Cu and Ni values were used with Table l of RG 1.99 [6] to determine a chemistry factor (CF) per Paragraph 1. 1 of RG 1.99 for welds.

The Cu and N i values were used with Table 2 of RG 1.99 to determine a CF per Paragraph 1. 1 of RG 1.99 for plates and forgings. Since only one surveillance capsule containing the appropriate plate heat has been tested no fitted chemistry factor is available.

The RPV 10 fluence value, associated with the limiting ART, of 1.95 x 1 018 n/cm2 at 37 EFPY was developed in Reference [7] based on linear interpolation between reported fluence values for 26.6 EFPY and 50. 1 EFPY from Reference [5], which were calculated in accordance with RG 1. 1 90 [4]. The RPV ID fluence value, associated with the limiting ART, of 2.60 x 1 018 n/cm2 at

Hatch Unit 2 PTLR Revision 0 Page 8 of 40

50. 1 EFPY was obtained from Reference [5] and was calculated in accordance with RG 1. 1 90.

These fluence values apply to the limiting beltline lower shell plate (Heat No. C8553-l ). The fluence values for the limiting lower shell plate are based upon an attenuation factor of 0.682 for a postulated 1/4T flaw. As a result, the I/4T fluence for 37 EFPY and 50. 1 EFPY for the limiting lower shell plate are 1.33 x 1 01 8 n/cm2 and 1.77 x 1 018 n/cm2, respectively, for HNP-2.

The water level instrument (WLI) nozzle is located in the lower intermediate shell beltline plates

[8]. The RPV ID fluence value of2.45 x 1 01 8 n/cm2 at 37 EFPY was developed in Reference [7]

based on linear interpolation between reported fluence values for 26.6 EFPY and 50. 1 EFPY from Reference [5], which were calculated in accordance with RG 1. 1 90 [4]. The peak RPV 10 fluence value of and 3.28 x I 0 18 -/cm2 at 50. 1 EFPY was obtained from Reference [5] and was calculated in accordance with RG 1. 1 90 [ 4]. These fluence values apply to the limiting lower intermediate shell plate (Heat No. C8579-2). The fluence values for the WLI nozzle are based upon an attenuation factor of 0.724 for a postulated I/4T flaw. As a result, the I/4T fluence for 37 EFPY and 50. 1 EFPY for the limiting lower intermediate shell plate are 1.77 x I 018 n/cm2 and 2.38 x 1 018 n/cm2, respectively, for HNP-2. The recirculation inlet (N2) and outlet (N I ) nozzles do not exist in the beltline region. However, the outer edge of the recirculation inlet nozzle forging is within Y4 inch of the beltline [7]. Based on the ART evaluation in Reference [7], the N2 nozzle forging material is not limiting.

The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the I /4T (inside surface flaw) and 3/4T (outside surface flaw) locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T and the 3/4T locations. This is because the thermal gradient tensile stress of interest is in the inner wall during cool-down and is in the outer wall during heat-up. However, as a conservative simplification, the thermal gradient stresses at the 1/4T location are assumed to be tensile for both heat-up and cool-down. This results in the approach of applying the maximum tensile stresses at the 1 /4T location. This approach is conservative because irradiation effects cause the

Hatch Unit 2 PTLR Revision 0 Page 9 of 40 allowable toughness at the I /4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, which is well above the P-T curve limits.

For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves specify a coolant heat-up and cool-down temperature rate of :S I oo*F /hr for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound Service Level A/8 RPV thermal transients. For the hydrostatic pressure and leak test curve (Curve A), a coolant heat-up and cool-down temperature rate of :S 25.F/hr must be maintained. The P-T limits and corresponding limits of either Curve A or 8 may be applied, if necessary, while achieving or recovering from test conditions. So, although Curve A applies during pressure testing, the limits of Curve 8 may be conservatively used during pressure testing if the pressure test heat-up/cool-down rate limits cannot be maintained.

The initial RTNoT, the chemistry (weight-percent copper and nickel) and ART at the I/4T location for all RPV beltline materials significantly affected by tluence (i.e., tluence > 1 017 n/cm2 for E > I MeV) are shown in Table 7 for 3 7 EFPY and Table 8 for 50. 1 EFPY [7]. The initial RTNoT values shown in Tables 7 and 8 have been previously approved for use by the NRC per Reference [ I 7].

Per Reference [7] and in accordance with Appendix A of Reference [ I ], the HNP-2 representative weld and plate surveillance materials data were reviewed from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) [ 1 2].

The HNP-2 representative plate and weld materials C8554 and 5 1 9 1 2, respectively, are contained in the HNP-2 surveillance capsules [7]. BWRVIP-1 35 [ 1 2] contains surveillance capsule test results for the Hatch Unit 2 representative plate and weld materials. The representative plate heat does not match the target plate heat; however, it does match the heat for

Hatch Unit 2 PTLR Revision 0 Page 1 0 of 40 plate material used in other beltline plates. Since only one surveillance capsule containing this plate heat has been tested no fitted chemistry factor is available; therefore, the CF calculated using the RG 1.99 [6] tables is used. The representative weld material heat does not match any weld material heats used in the Hatch Unit 2 beltline; therefore, the CF calculated using the RG 1.99 tables is used.

The ANSYS finite element computer program was used to develop the stress distributions through the feedwater (FW) nozzle [ 1 3]. These stress distributions were used in the determination of the stress intensity factors for the FW nozzle [ 1 4]. At the time that each of the analyses above was performed, the ANSYS program was controlled under the vendor's I 0 CFR 50 Appendix B [ 1 5] Quality Assurance Program for nuclear quality-related work.

The plant-specific HNP-2 FW nozzle analysis was performed to determine stress intensity factors due to through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients [ 1 4]. Pressure and thermal stress distributions were taken from Reference [ 1 3]. Detailed information regarding the analysis can be found in References [ 1 3, 1 4].

The following summarizes the development of the thermal and pressure stress intensity factors for the FW nozzle [ 1 4] :

With respect to thermal stresses, the thermal shock which represents the maximum thermal shock for the FW nozzle during normal and upset operating conditions was analyzed [ 1 3]. The thermal stress distribution, corresponding to the limiting time point presented in [ 1 3], along a linear path through the nozzle corner is used [ 1 4]. Leakage is considered in the heat transfer calculations [ 1 3]. The thermal down shock of 450°F produces the highest tensile stresses at the I/4T location. The BIE/IF methodology presented in the Sl P-T Curve L TR [ I ] is used to calculate the thermal stress intensity, K1T, due to the thermal shock by fitting a third order polynomial equation to the path stress distribution for the thermal shock load case [ 1 4]. Because operation is along the

Hatch Unit 2 PTLR Revision 0 Page I I of 40 saturation curve, the resulting K1T can be linearly scaled to determine the K1T to reflect the worst-case step change due to the available temperature difference. It is recognized that at low temperatures, the available temperature difference is insignificant and could potentially result in a near zero stress distribution. Therefore, a minimum K1T is calculated based on the thermal ramp of I 00°F /hr, which is associated with the shutdown transient [ 1 4]. The resulting combination of the thermal down shock and thermal ramp KIT values represent the bounding thermal stress intensity factors in the FW nozzle associated with the P-T curves for the non-beltline region.

Boundary conditions and heat transfer coefficients were developed based on testing as described in Appendix A of Reference [ I 3a].

With respect to pressure stresses, a unit pressure of I 000 psig was applied to the internal surfaces ofthe finite element model (FEM) [ 1 3]. The pressure stress distribution was taken along the same path as the thermal stress distribution. Recognizing that the Reference [ I 3] evaluation was performed using a 2-D axi-symmetric finite element model FEM and that it is known that the stress intensification caused by the nozzle geometry is under predicted in a 2-D axi-symmetric representation of the nozzle, a correction factor was applied to the stresses obtained from the 2-D axi-symmetric FEM as described in Reference [ 1 4]. The BIE/IF methodology presented in the SI P-T Curve L TR [ I ] is used to calculate the pressure stress intensity factor, K1p, by fitting a third order polynomial equation to the path stress distribution for the pressure load case. The resulting K1p can be linearly scaled to determine the K1p for various RPV internal pressures.

Material properties were taken from the H NP-2 code of construction [ I 6]. Use of temperature dependent material properties is expected to have minimal impact on the results ofthe analysis.

6.0 References Hatch Unit 2 PTLR Revision 0 Page 1 2 of 40 I. BWROG-TP-1 1 -022-A, Revision 1 (SIR-05-044, Revision 1 -A), "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", dated June 20 1 3.

2. B WROG-TP-1 1 -023-A, Revision 0 (0900876.40 I, Revision 0-A), "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations", dated May 201 3.

3. Design Input Requests:

a. DIR, Revision 2, "Revised P-T Curves for Plant Hatch Units 1 &2," Sl File No.

1 001 527.20 1.

b. DIR, Revision 0, "Hatch Units I and 2 P-T Curve Revisions," Sl File No.

1 400365.200.

4. U. S. Nuclear Regulatory Commission, Regulatory Guide 1. 1 90, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", March 200 1.

5. Transware Enterprises Inc. Report No. SNC-HA2-00 1 -R-00 I Revision 0, "Edwin I.

Hatch Unit 2 Fluence Evaluation at End of Cycle 22 and 50. 1 EFPY."

6. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel materials", May 1 988.

7. Structural Integrity Associates Calculation No. I 00 1 527.302, Revision 1, "RPV Material Summary and ART Calculation", July 201 4.

8. Structural Integrity Associates Calculation No. I 00 1 527.305, Revision 2, "Hatch Unit 2 P-T Curve Calculation for 3 7 and 50. 1 EFPY", August 20 1 4.

9. General Electric Document No. GE-NE-B I I 00827-00-0 I, "Plant Hatch Units I & 2 RPV Pressure Temperature Limits License Renewal Evaluation," March 1 999.

I 0. NRC Docket No. 50-32 1, "Issuance of Amendment No. 1 77 to Facility Operating License DPR-57 and Amendment No. 1 1 8 to Facility Operating License NPF Edwin

Hatch Unit 2 PTLR Revision 0 Page 1 3 of 40 I. Hatch Nuclear Plant, Units 1 and 2," Amendment No. 1 77, License No. DPR-57, January 1 992, ADAMS Accession No. MLO 1 29901 00.

1 1. SI Calculation No. 1 400365.301, Rev. 0, "Hatch RPV Vacuum Assessment."

1 2. BWRVIP-1 35, Revision 2: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA: 2009.

1 02023 1. EPRI PROPRIETARY INFORMATION. SI File No. BWRVIP-01 -335P.

13. Hatch Unit 2 NUREG-061 9 Evaluations:

a. Liffengren, D. J., et al., "Edwin I. Hatch Nuclear Power Station, Unit 2 Feedwater Nozzle Fracture Mechanics Analysis to Show Compliance with NUREG-061 9,"

NEDC-30256, DRF-1 37-00 1 0, August 1 983, General Electric Company. SI File No. 1 001 527.2 1 0.

b. Stevens, G. L., "Updated Feedwater Nozzle Fracture Mechanics Analysis for Edwin I. Hatch Nuclear Power Station Unit 2," GE-NE-523-95-0991, Rev. 0, DRF 8 1 3-0 1 524, September 1 99 1, General Electric Company. SI File No.

1 001 527.2 1 0.

c. Bothne, D., "Power Uprate Evaluation Report for Edwin I. Hatch Unit 2, Feedwater Nozzle NUREG-06 1 9 Fracture Mechanics Analysis for Extended Power Uprate Conditions," GE-NE-8 1 3-0 1 869-065-02, July 1 997, General Electric Company. SI File No. 1 001 527.2 1 0

14. Structural Integrity Associates Calculation No. 1 001 527.303, Revision 0, "Feedwater, Water Level Instrument, and Core DP Nozzle Fracture Mechanics Evaluation for Hatch Unit 1 and Unit 2 Pressure-Temperature Limit Curve Development", December 201 1 1 5. U. S. Code of Federal Regulations, Title 1 0, Energy, Part 50, Appendix B, "Quality Assurance for Nuclear Power Plants and Fuel Reprocessing Plants".

Hatch Unit 2 PTLR Revision 0 Page 1 4 of 40 1 6. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section III, 1 968 Ed. through 1 970 Addenda.

1 7. NUREG-1 803, "Safety Evaluation Report Related to the License Renewal of the Edwin I.

Hatch Nuclear Plant, Units 1 and 2," December 200 1.

1 8. General Electric Report No. SASR 90-1 04, "E. I. Hatch Nuclear Power Station, Unit 2 Vessel Surveillance Materials Testing and Fracture Toughness Analysis," May 1 99 1. SI File No. l 00 1 527.205 1 9. Letter from L. N. Olshan (NRC) to H. L. Sumner (SNC), "Edwin I. Hatch Nuclear Plant, Units 1 and 2 Re: Issuance of Amendments (TAC NOS. MB6 1 06 and MB6 1 07)", March 1 0, 2003.

20. BWRVIP-86, Revision 1 -A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan, EPRI Product I 025 1 44, October 20 1 2.

Hatch Unit 2 PTLR Revision 0 Page 1 5 of 40 Figure 1: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 37 EFPY 1300 1200 1100 1000 900 I

BOO J

700 j 6QO i 500 I 400 300 200 100 0

Curve A

Pressure Test, Composite Curves

-aeltllne

- - - Bottom Head

- - Non-Beltllne OVenlll

/

,I

/ I I

, I

, I I

I i

I I I I

I I

/

.i If I I v

ll I

M inimum Bolt-Up I

Temperature = 90°F Minimum RPV Pressure = -14.7 psig 5

1ila 1!;o 2Xl 2ó MlnlmumR*ctorV.... I Mml Tempnture rF)

Hatch Unit 2 PTLR Revision 0 Page 1 6 of 40 Figure 2: HNP-2 P-T Curve B (Normal Operation - Core Not Critical) for 37 EFPY Curve B - Core Not Critical, Composite Curves

-- Beltline

Bottom Head Non-Beltline

-overall 1300 I

w I

I I

1200 I

I I

1100 f--

I I

I 1000 I

I I

I 900 I

I I

QD 800 ill

.e i

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Ill CLI 600 a::

=

e

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/i:.

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- I I

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200 Minimum Bolt-Up Temperature = 90°F 100 M ini mum RPV Pressure = -14.7 psig 0 -

ro 1)()

2fo 2

Minimum Reactor Vessel Metal Temperature ("F)

Hatch Unit 2 PTLR Revision 0 Page I 7 of 40 Figure 3: HNP-2 P-T Curve C (Normal Operation - Core Critical) for 37 EFPY Curve C - Core Critical, Composite Curves

-- Beltline

- - -

Bottom Head Non-Beltline

-overall 1300 r

f

' I

' I I

1200 I

I 1100 I

l I

I I

r 1000 I

I I

900 I

I I

Qjj 800 VI 1 700 II I

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400 I

I I

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50 11cJ I

1 2jlo 2i0 Minimum Reactor Vessel Metal Temperature ('F)

Hatch Unit 2 PTLR Revision 0 Page 1 8 of 40 Figure 4: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 50.1 EFPY 1300 1200 1100 1000 900 1

8QO l

700 I 600 j 500 I 400 300 200 100 0

Curve A

Pressure Test, Composite Curves

-Beltllne

* * *Bottom Hud

- - Non-Seltllne Overall I

NI

/ I

I r-I I I I I

h[

I I

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Minimum R*ctor VU181 Mal T*mpentunt rF)

Hatch Unit 2 PTL R Revision 0 Page 1 9 of 40 Figure 5: HNP-2 P-T Curve B (Normal Operation - Core Not Critical) for 50.1 EFPY 1300 1200 1100 1000 900 Qjj 800 "i

700 II

í II 600 a::

.E E

I 500 II a

II E.

400 300 200 100 0

Curve B - Core Not Critical, Composite Curves

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- - -* Bottom Head

Non-Beltline

-overall w

I I

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/

I I

I A

I I

?

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1 1i 2Kl Minimum Reactor Vessel Metal Temperature {'F) 21o

Hatch Unit 2 PTLR Revision 0 Page 20 of 40 Figure 6: HNP-2 P-T Curve C (Normal Operation - Core Critical) for 50.1 EFPY 1300 1200 1100 1000 900

'DO 800 iii

]

700 cu l5 t;

cu 600 a:

.E

. E

I 500 cu cu A

400 300 200 100 0

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Minimum Bolt-Up

'+ =-

Temperature = 90°F Minimum RPV Pressure = -14.7 psig so 1)0 1i0 2l0 Minimum Reactor Vessel Metal Temperature ("F) 2i0

Hatch Unit 2 PTLR Revision 0 Page 2 1 of 40 Table 1: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 37 EFPY Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.0 506.7 98.5 554.7 105.7 602.7 112.0 650.7 117.6 698.8 122.7 746.8 127.3 794.8 131.5 842.8 135.3 890.8 138.9 938.8 142.3 986.8 145.4 1034.9 148.4 1082.9 151.2 1130.9 153.8 1178.9 156.3 1226.9 158.7 1274.9 161.0 1322.9

Hatch Unit 2 PTLR Revision 0 Page 22 of40 Table 1: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 37 EFPY (continued)

Bottom Head Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.0 869.3 94.5 918.9 98.7 968.4 102.5 1018.0 106.1 1067.5 109.4 1117.1 112.5 1166.6 115.4 1216.2 118.2 1265.7 120.8 1315.3 Non-Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.0 312.6 120.0 312.6 120.0 1310.2

Hatch Unit 2 PTLR Revision 0 Page 23 of 40 Table 2: HNP-2 P-T Curve B (Normal Operation - Core Not Critical) for 37 EFPY Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.0 248.1 101.1 296.8 110.2 345.5 117.9 394.2 124.6 442.9 130.5 491.6 135.7 540.3 140.5 589.0 144.9 637.7 148.9 686.4 152.6 735.1 156.0 783.8 159.2 832.5 162.3 881.2 165.1 929.9 167.8 978.6 170.4 1027.3 172.8 1076.0 175.1 1124.7 177.4 1173.4 179.5 1222.1 181.5 1270.8 183.5 1319.5

Hatch Unit 2 PTLR Revision 0 Page 24 of 40 Table 2: HNP-2 P-T Curve B (Normal Operation - Core Not Critical) for 37 EFPY (continued)

Bottom Head Region Curve 8 - Core Not Critical P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.0 546.3 95.8 594.7 101.0 643.1 105.7 691.5 110.1 739.9 114.0 788.4 117.7 836.8 121.1 885.2 124.3 933.6 127.3 982.0 130.2 1030.4 132.9 1078.8 135.4 1127.3 137.9 1175.7 140.2 1224.1 142.4 1272.5 144.5 1320.9

Hatch Unit 2 PTLR Revision 0 Page 25 of 40 Table 2: HNP-2 P-T Curve B (Normal Operation - Core Not Critical) for 37 EFPY (continued)

Non-Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.0 312.6 150.0 312.6 150.0 1313.5

Hatch Unit 2 PTLR Revision 0 Page 26 of 40 Table 3: HNP-2 P-T Curve C (Normal Operation - Core Critical) for 37 EFPY Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.0 140.6 112.2 189.6 127.5 238.7 139.2 287.7 148.7 336.8 156.7 385.8 163.6 434.9 169.6 483.9 175.0 533.0 179.9 582.0 184.3 631.1 188.4 680.1 192.1 729.2 195.6 778.2 198.9 827.3 202.0 876.3 204.9 925.4 207.6 974.4 210.2 1023.5 212.7 1072.5 215.0 1121.6 217.2 1170.6 219.4 1219.7 221.5 1268.7 223.4 1317.8

Hatch Unit 2 PTLR Revision 0 Page 27 of 40 Table 3: HNP-2 P-T Curve C (Normal Operation - Core Critical) for 37 EFPY (continued)

Bottom Head Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.0 330.2 102.3 379.5 112.2 428.8 120.5 478.1 127.5 527.5 133.7 576.8 139.3 626.1 144.2 675.4 148.7 724.7 152.9 774.0 156.7 823.3 160.3 872.6 163.6 922.0 166.7 971.3 169.6 1020.6 172.4 1069.9 175.0 1119.2 177.5 1168.5 179.9 1217.8 182.1 1267.1 184.3 1316.4

Hatch Unit 2 PTLR Revision 0 Page 28 of 40 Table 3: HNP-2 P-T Curve C (Normal Operation - Core Critical) for 37 EFPY (continued)

Non-Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.2 193.0 101.6 232.9 110.6 272.7 118.2 312.6 190.0 312.6 190.0 1313.5

Hatch Unit 2 PTLR Revision 0 Page 29 of 40 Table 4: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 50.1 EFPY Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.0 481.5 99.5 530.6 107.5 579.8 114.3 628.9 120.4 678. 1 125.8 727.3 130.7 776.4 135.1 825.6 139.2 874.7 142.9 923.9 146.4 973.1 149.7 1022.2 152.8 1071.4 155.7 1120.6 158.4 1169.7 161.0 1218.9 163.5 1268.0 165.8 1317.2

Hatch Unit 2 PTLR Revision 0 Page 30 of 40 Table 4: HNP-2 P-T Curve A (Hydrostatic Pressure and Leak Test) for 50.1 EFPY (continued)

Bottom Head Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.0 869.3 94.5 918.9 98.7 968.4 102.5 1018.0 106.1 1067.5 109.4 1117.1 1 12.5 1166.6 1 15.4 1216.2 118.2 1265.7 120.8 1315.3 Non-Beltline Region Curve A - Pressure Test P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.0 312.6 120.0 312.6 120.0 13 10.2

Hatch Unit 2 PTLR Revision 0 Page 3 1 of 40 Table 5: HNP-2 P-T Curve B (Normal Operation - Core Not Critical) for 50.1 EFPY Beltline Region Curve 8 - Core Not Critical P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.0 229.2 102.4 278.6 112.2 328.0 120.5 377.4 127.6 426.8 133.8 476.2 139.3 525.6 144.3 575.0 148.8 624.4 153.0 673.8 156.8 723.2 160.3 772.6 163.7 822.0 166.8 871.4 169.7 920.8 172.5 970.2 175.1 1019.6 177.6 1069.0 180.0 1 1 18.4 182.2 1 167.8 184.4 1217.2 186.5 1266.6 188.5 13 16.0

Hatch Unit 2 PTLR Revision 0 Page 32 of 40 Table 5: HNP-2 P-T Curve B (Normal Operation - Core Not Critical) for 50.1 EFPY (continued)

Bottom Head Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.0 546.3 95.8 594.7 101.0 643.1 105.7 691.5 1 10.1 739.9 1 14.0 788.4 1 17.7 836.8 121.1 885.2 124.3 933.6 127.3 982.0 130.2 1030.4 132.9 1078.8 135.4 1127.3 137.9 1 175.7 140.2 1224.1 142.4 1272.5 144.5 1320.9

Hatch Unit 2 PTLR Revision 0 Page 33 of 40 Table 5: HNP-2 P-T Curve B (Normal Operation - Core Not Critical) for 50.1 EFPY (continued)

Non-Beltline Region Curve B - Core Not Critical P-T Curve P-T Curve Temperature Pressure

F psi 90.0 0.0 90.0 3 12.6 150.0 312.6 150.0 1313.5

Hatch Unit 2 PTLR Revision 0 Page 34 of 40 Table 6: HNP-2 P-T Curve C (Normal Operation - Core Critical) for 50.1 EFPY Beltline Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.0 132.1 1 14.2 181.4 130.5 230.8 142.7 280. 1 152.5 329.5 160.7 378.8 167.8 428.1 173.9 477.5 179.4 526.8 184.4 576.2 188.9 625.5 193.0 674.8 196.9 724.2 200.4 773.5 203.7 822.9 206.8 872.2 209.7 921.6 212.5 970.9 215.1 1020.2 217.6 1069.6 220.0 1 118.9 222.2 1168.3 224.4 1217.6 226.5 1266.9 228.5 1316.3

Hatch Unit 2 PTLR Revision 0 Page 35 of 40 Table 6: HNP-2 P-T Curve C (Normal Operation - Core Critical) for 50.1 EFPY (continued)

Bottom Head Region Curve C - Core Critical P-T Curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.0 330.2 102.3 379.5 112.2 428.8 120.5 478. 1 127.5 527.5 133.7 576.8 139.3 626. 1 144.2 675.4 148.7 724.7 152.9 774.0 156.7 823.3 160.3 872.6 163.6 922.0 166.7 971.3 169.6 1020.6 172.4 1069.9 175.0 1 119.2 177.5 1168.5 179.9 1217.8 182.1 1267.1 184.3 13 16.4

Hatch Unit 2 PTLR Revision 0 Page 36 of 40 Table 6: HNP-2 P-T Curve C (Normal Operation - Core Critical) for 50.1 EFPY (continued)

Non-Beltline Region Curve C - Core Critical P-T curve P-T Curve Temperature Pressure "F

psi 90.0 0.0 90.2 193.0 101.6 232.9 1 10.6 272.7 1 18.2 312.6 190.0 312.6 190.0 1313.5

Table 7: Hatch Unit 2 ART Table for 37 EFPY Deecnpllon Code No.

Heat No.

Flux Type & Lot No.

Initial RTNoT !"F)

Chemllltry Cu (wt %)

Nl (wt %)

Lower Shell #1 G-6603-1 C8553-2

-20 0.08 0.58 Lower Shell #2 G-6603-2 C8553-1 24 0.08 0.58 jl Lower Shell#l G-6603-3 C8571-1 0

0.08 0.53 Lower-In!. Shell #1 G-6602-2 C8554-1

-20 0.08 0.57 Lower-In! Shell #2 G-6602-1 C8554-2

-10 0.08 0.58 Lower-In! Shell #3 G-6601-4 C8579-2

-4 0. 1 1 0.48 Deecnpllon Code No.

Heat No.

Flux Type & Lot No.

Initial RT NOT ("F)

Chemllltry Cu (wt %)

Nl (wt %)

Lower Long. Weld 101-842 10137

-50 0.216 0.043 Lower Int. Long Weld 101-834 51 874

-50

0. 147 0.037 4P6052 Lower - Lower Int. Girth Weld 301-871

-50 0.047 0.049 II Fluence Data Wall Thlcknees (ln.J Fluence at ID Atte nuatlon, Location Full 1/4t (n/cmz) 1/4t = e.0.24x Lower Shell #1 6.375 1.594 1. 95E+1 8 0.682 Lower Shell #2 6.375 1.594 1. 95E+1 8 0.682 Lower Shell#l 6.375 1.594 1. 95E+18 0.682 Lower-In!. Shell #1 5.375 1.344 2.45E+18 0.724 ZI Lower-In! Shell #2 5.375 1.344 2.45E+1 8 0.724 Lower-In! Shell #3 5.375 1.344 2.45E+1 8 0.724 Lower Long. Weld 6.375 1.594 1.81E+18 0.682 Lower Int. Long Weld 5.375 1.344 1.55E+1 8 0.724 Lower - Lower Int. Girth Weld 5.375 1.344 1.95E+1 8 0.724 II L

Chemllltry Factor ("F) 51 51 51 51 51 73 Chemistry Factor ("F) 98 68 31 ART NoT

("F) 24.3 24.3 24.3 27.6 27.6

-5 ART NoT

("F)

45. 1 30.0 15.2 Hatch Unit 2 PTLR Revision 0 Page 37 of 40 Adjustments for 1/4t Margin Tenns IJJ ("F) aA ("F) 0.0 1 2.2 0.0 12.2 0, 0 12.2 0.0 1 3. 8 0.0 13.8 0.0 17.0 Adjustments for 1/4t Margin Tenns IJJ("F) aA ("F) 0.0 22.5 0.0 1 5.0 0.0 7.6 ART NoT

("F) 28.6 72.6 48.6 35.2 45.2 69.5 ART NoT

("F) 40.2 1 0.0

-19.6 Fluence @ 1/4t (n/cm2)

Fluence Factor, FF j(0-21-0.101og IJ 1.33E+18 o.4n 1.33E+18 0.477 1.33E+18 0.477 1.77E+18 t

0.541 1.nE+18 0.541 1.nE+1 8 I

0.541 1.23E+18 0.460 1. 1 3E+18 0.441 1.41E+18 0.490

Table 8: Hatch Unit 2 ART Table for 50.1 EFPY Deacrlptlon Code No.

Heat No.

Flux Type & Lot No.

Initial RTNor i"F)

Chemistry Cu (wt %)

Nl (wt %)

Lower Shell #1 G-6603-1 C8553-2

-20 0.08 0.58 Lower Shell #2 G603-2 C8553-1 24 0.08 0.58 XI Lower Shell#3 G-6603-3 C8571-1 0

0.08 0.53 Lower-lnt. Shell #1 G-6602-2 C8554-1

-20 0.08 0.57 Lower-In! Shell #2 G-6602-1 C8554-2

-10 0.08 0.58 Lower-In! Shell #3 G-6601-4 C8579-2

-4 0.1 1 0.48 Des:rlptlon Code No.

Heat No.

Flux Type & Lot No.

Initial RTNor l"f)

Chemistry Cu (wt %)

Nl (wt %)

Lower Long. Weld 101-842 10137

-50 0.216 0.043 Lower Int. Long Weld 101-834 51874

-50 0.147 0.037 Lower - Lower Int. Girth Weld 301-871 4P6052

-50 0.047 0.049 il Des:rlptlon Code No.

Heat No.

Flux Type & Lot No.

Initial RTNor !"F)

Chemistry Cu (wt %)

Nl (wt %)

Forg.

Recirculation Inlet Nozzle G-6607 Q2Q24W 10 0.180 0.810 Fluence Data Wall Thlckne* (!n.J Fluence at 10 Attenuation, Location Full 1/4t (nlcmi 1/4t = e.0.2Ax Lower Shell #1 6.375 1.594 2.60E+18 0.682 Lower Shell #2 6.375 1.594 2.60E+18 0.682 Lower Shell#3 6.375 1.594 2.60E+18 0.682 Lower-In!. Shell #1 5 375 1.344 3.28E+18 0.724 YI a:

Lower-In! Shell #2 5.375 1.344 3.28E+18 0.724 Lower-In! Shell #3 5.375 1.344 3.2BE+18 0.724 Lower Long. Weld 6.375 1.594 2.42E+18 0.682 Lower Int. Long Weld 5.375 1.344 2.10E+18 0.724 Lower - Lower Int. Girth Weld 5.375 1.344 2.60E+18 0.724 il rculation Inlet Nozzle 6.375 1.594 1.00E+17 0.682 I

Chemistry Factor ("F) 51 51 51 51 51 73 Chemistry Factor ("F) 98 68 31 Chemistry Factor ("F) 141 ART NoT

("F) 27.6 27.6 27.6 31.2 31.2 44.6 ART NoT

("F) 51.4 34.4 17.2 ART NoT

("F) 1 1.9 Hatch Unit 2 PTLR Revision 0 Page 38 of 40 Adju!lbnents for 1/4t Margin Tenns CJi ("F) a6 ("F) 0.0 13.8 0.0 13.8 0.0 13.8 0.0 15.6 0.0 15.6 0.0 17.0 Adju!lbnents for 1/4t Margin Tenns a* ("F) a6 ("F) 0.0 25.7 0.0 17.2 0.0 8.6 Adju!lbnents for 1/4t Margin Tenns a* ("F) a6 ("F) 0.0 5.9 ART NOT

("F) 35.2 79.2 55.2 42.4 52.4 74.6 ART NoT

("F) 52.8 18.8

-15.6 ART NoT

("F) 33.7 Fluence @ 1/4t (nlcm2)

Fluence Factor, FF

,co.:za.o.101og l) 1.77E+18 0.541 1.77E+1 8 0.541 1.77E+1B 0.541 2.38E+18 0.61 1 2.38E+18 0.61 1 2.38E+18 0.61 1 1.65E+18 0.525 1.52E+1 8 0.506 1.88E+1B 0.555 6 82E+16 I

0.084

Table 9: Hatch Unit 2 Summary of Nozzle Stress Intensity Factors Nozzle Applied Pressure, Thermal, K,t Thermal, K,t KIP-aPP (4so*F shock)

(100 *F/hr Plate)

Feed water 78.9 46.8 12.9 WLI 80.0 N/A 19.9 Notes:

1. K1 in units of ksi-in°*5 Hatch Unit 2 PTLR Revision 0 Page 39 of 40

Appendix A Hatch Unit 2 PTLR Revision 0 Page 40 of 40 HATCH UNIT 2 REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with 1 0 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, one surveillance capsule was removed from the Hatch Nuclear Plant Unit 2 (HNP-2) reactor vessel in 1 989 following cycle 8 [ 1 8]. The surveillance capsule contained flux wires for neutron tluence measurement, Charpy V -Notch impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel materials within the core beltline region [ 1 8].

Southern Nuclear Operating Company committed to use the ISP in place of its existing surveillance programs in the amendments issued by the NRC regarding the implementation of the Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Integrated Surveillance Program, dated March I 0, 2003 [ 1 9]. The BWRVIP ISP meets the requirements of 1 0 CFR 50, Appendix H, for Integrated Surveillance Programs, and has been approved by NRC for the period of extended operation [20].

HNP-2 continues to be a host plant under the ISP

[ 1 2]. Two more HNP-2 capsule are scheduled to be removed and tested under the ISP in approximately 201 7 and 2027.

v t e Letter Sequence Request
CAC:MF6063, (Approved, Closed)
Initiation Request, Request Acceptance... Administration Questions Answers 12 November 2015 Results Approval Other: ML21159A178
MONTHYEAR2016-02-09 [Table View]

NL-13-1773, Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent with TSTF-419-A (NL-13-1773)

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NL-13-1773, Application for Amendment to Technical Specifications Regarding Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR) Consistent with TSTF-419-A (NL-13-1773)

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Enclosures:

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