ML110050298

Issuance of Amendment Regarding Revised Pressure and Temperature (P-T) Limit Curves and Relocation of P-T Curves to the PTLR

ML110050298
Person / Time
Site:	Pilgrim
Issue date:	01/26/2011
From:	Richard Guzman Plant Licensing Branch 1
To:	Entergy Nuclear Operations
Guzman R, NRR/DORL, 415-1030
References
TAC ME3253
Download: ML110050298 (24)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 26, 2011 Site Vice President Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360-5508

SUBJECT:

PILGRIM NUCLEAR POWER STATION - ISSUANCE OF AMENDMENT REGARDING REVISED PRESSURE AND TEMPERATURE (P-T) LIMIT CURVES AND RELOCATION OF P-T CURVES TO THE PRESSURE AND TEMPERATURE LIMITS REPORT (TAC NO. ME3253)

Dear Sir or Madam:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 234 to Facility Operating License No. DPR-35 for the Pilgrim Nuclear Power Station (Pilgrim), in response to your application dated January 24, 2010 (Agencywide Documents Access Management System (ADAMS) Accession No. ML100270054), as supplemented by letters dated September 7 (ADAMS Accession No. ML102580240) and November 4,2010 (ADAMS Accession No. ML103200208).

This amendment modifies the Pilgrim Technical Specification (TS) Section 1.0, Definitions, TS Section 3.6, Primary System Boundary Specifications 3.6.A, and TS Programs and Manuals Section 5.5, to include reference to the Pressure and Temperature Limits Report (PTLR). The PTLR includes revised 34 effective full-power years P-T curves, neutron fluence, and adjusted reference temperature values.

A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely, Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation Docket No. 50-293

Enclosures:

1. Amendment No. 234 to DPR-35

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR GENERATION COMPANY ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-293 PILGRIM NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 234 License No. 35

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment filed by Entergy Nuclear Operations, Inc. (the licensee) dated January 24, 2010, as supplemented by letters dated September 7 and November 4, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance 0) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

- 2

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-35 is hereby amended to read as follows:

8.

Technical Specifications The Technical Specifications contained in Appendix A, as revised throllgh Amendment No. 234, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Q'dJ~ ;1~2A~for ncy L. Salgado, Chief

\\ lant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: January 26, 2011

ATTACHMENT TO LICENSE AMENDMENT NO. 234 TO FACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-293 Replace the following page of the Facility Operating License with the attached revised page.

The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 3

3 Replace the following pages of Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages 1-4 1-4 1-5 1-5 1-6 1-6 3/4.6-1 3/4.6-1 3/4.6-2 3/4.6-2 3/4.6-9 3/4.6-9 3/4.6-10 3/4.6-10 3/4.6-11 3/4.6-11 5.0-11 5.0-11

- 3 B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 234, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C.

Records ENO shall keep facility operating records in accordance with the requirements of the Technical Specifications.

D.

Equalizer Valve Restriction - DELETED E.

Recirculation Loop Inoperable - DELETED F.

Fire Protection ENO shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the SER dated December 21, 1978 as supplemented subject to the following provision:

ENO may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

G.

Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Pilgrim Nuclear Power Station Physical Security, Training and Qualification, and Safeguards Contingency Plan, Revision 0" submitted by letter dated October 13, 2004, as supplemented by letter dated May 15, 2006.

Amendment 225,226,227,228,229,230,231,232,233,234

1.0 DEFINITIONS (Cont)

OPERABLE OPERABILITY OPERATING OPERATING CYCLE PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

PRIMARY CONTAINMENT INTEGRITY PROTECTIVE ACTION PROTECTIVE FUNCTION PNPS 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).

OPERATING means that a system or component is performing its intended functions in its required manner.

Interval between the end of one refueling outage and the end of the next subsequent refueling outage.

The PTLR is the Pilgrim-Specific document that provides the reactor vessel Pressure-Temperature (P-T) Curves, including heat up and cool down rates and fluence and Adjusted Reference Temperature limits for Specification 3.6.A. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.5.9.

PRIMARY CONTAINMENT INTEGRITY means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:

1.

All manual containment isolation valves on lines connected to the reactor coolant system or containment which are not required to be open during accident conditions are closed.

2.

At least one door in each airlock is closed and sealed

3.

All blind flanges and manways are closed.

4.

All automatic primary containment isolation valves and all instrument line check valves are operable or at least one containment isolation valve in each line having an inoperable valve shall be deactivated in the isolated condition.

5.

All containment isolation check valves are operable or at least one containment valve in each line having an inoperable valve is secured in the isolated position.

An action initiated by the protection system when a limit is reached. A PROTECTIVE ACTION can be at a channel or system level.

A system PROTECTIVE ACTION which results from the PROTECTIVE ACTION of the channels monitoring a particular plant condition.

1-4 Amendment No. +++, 234

1.0 DEFINITIONS (Cont)

REACTOR POWER OPERATION REACTOR VESSEL PRESSURE REFUELING INTERVAL REFUELING OUTAGE SAFETY LIMIT SECONDARY CONTAINMENT INTEGRITY SIMULATED AUTOMATIC ACTUATION SOURCE CHECK STAGGERED TEST BASIS REACTOR POWER OPERATION is any operation with the mode switch in the "Startup" or "Run" position with the reactor critical and above 1 % design power.

Unless otherwise indicated, REACTOR VESSEL PRESSURES listed in the Technical Specifications are those measured by the reactor vessel steam space detectors.

REFUELING INTERVAL applies only to In-service Code Testing Program surveillance tests. For the purpose of designating frequency of these code tests, a REFUELING INTERVAL shall mean at least once every 24 months.

REFUELING OUTAGE is the period of time between the shutdown of the unit prior to a refueling and the startup of the plant after that refueling. For the purpose of designating frequency of testing and surveillance, a REFUELING OUTAGE shall mean a regularly scheduled outage; however, where such outages occur within 11 months of completion of the previous REFUELING OUTAGE, the required surveillance testing need not be performed until the next regularly scheduled outage.

The SAFETY LIMITS are limits below which the reasonable maintenance of the cladding and primary systems are assured.

Exceeding such a limit is cause for unit shutdown and review by the Nuclear Regulatory Commission before resumption of unit operation. Operation beyond such a limit may not in itself result in serious consequences, but it indicates an operational deficiency subject to regulatory review.

SECONDARY CONTAINMENT INTEGRITY means that the reactor building is intact and the following conditions are met:

1. At least one door in each access opening is closed.

2. The standby gas treatment system is operable.

3. All automatic ventilation system isolation valves are operable or secured in the isolated position.

SIMULATED AUTOMATIC ACTUATION means applying a Simulated signal to the sensor to actuate the circuit in question.

A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

A STAGGERED TEST BASIS shall consist of: (a) a test schedule for.o. systems, subsystems. trains, or other designated components obtained by dividing the specified test interval into.0.

equal subintervals; (b) the testing of one system, subsystem. train or other designated components at the beginning of each subinterval.

Amendment No. +77, ~, 234 1-5

1.0 DEFINITIONS (Cont)

SURVEILLANCE FREQUENCY SURVEILLANCE INTERVAL TOTAL PEAKING FACTOR TRANSITION BOILING TRIP SYSTEM Each Surveillance Requirement shall be performed within the specified SURVEILLANCE INTERVAL with a maximum allowable extension not to exceed 25 percent of the specified SURVEILLANCE INTERVAL.

The SURVEILLANCE FREQUENCY establishes the limit for which the specified time interval for Surveillance Requirements may be extended. It permits an allowable extension of the normal surveillance interval to facilitate surveillance schedule and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities. It is not intended that this provision be used repeatedly as a convenience to extend surveillance intervals beyond that specified for surveillances that are not performed during refueling outages.

This limitation of this definition is based on engineering judgment and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.

The SURVEILLANCE INTERVAL is the calendar time between surveillance tests, checks, calibrations, and examinations to be performed upon an instrument or component when it is required to be operable. These tests may be waived when the instrument, component, or system is not required to be operable, but the instrument, component. or system shall be tested prior to being declared operable. The operating cycle interval is 24 months and the 25% tolerance of the definition of "SURVEILLANCE FREQUENCY" is applicable. The refueling interval is 24 months and the 25% tolerance specified in the definition of "SURVEILLANCE FREQUENCY" is applicable.

The ratio of the fuel rod surface heat flux to the heat flux of an average rod in an identical geometry fuel assembly operating at the core average bundle power.

TRANSITION BOILING means the boiling regime between nucleate and film bOiling. TRANSITION BOILING is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.

A TRIP SYSTEM means an arrangement of instrument channel trip Signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A TRIP SYSTEM may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action.

Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.

PNPS 1-6 Amendment No. ++7, 234

LIMITING CONDITION FOR OPERATION 3.6 PRIMARY SYSTEM BOUNDARY Applicability:

Applies to the operating status of the reactor coolant system.

Objective:

To assure the integrity and safe operation of the reactor coolant system.

Specification:

A. Thermal and Pressurization Limitations

1. The average rate of reactor coolant temperature change during normal heatup or cool down shall not exceed the limit in the PTLR.

2. The reactor vessel shall not be pressurized for hydrostatic and/or leakage tests, and subcritical or critical core operation shall not be conducted unless the reactor vessel temperatures are above those defined by the appropriate curves in the PTLR.

SURVEILLANCE REQUIREMENT 4.6 PRIMARY SYSTEM BOUNDARY Applicability:

Applies to the periodic examination and testing requirements for the reactor cooling system.

Objective:

To determine the condition of the reactor coolant system and the operation of the safety devices related to it.

Specification A. Thermal and Pressurization Limitations

1. During heatups and cooldowns, with the reactor vessel temperature less than or equal to 450°F, verify the ReS heatup and cooldown rates are within limits every 15 minutes.

2.

Reactor vessel shell temperatures, including reactor vessel bottom head, and reactor coolant pressure shall be permanently logged at least every 15 minutes whenever the shell temperature is below 220 0 F and the reactor vessel is not vented.

Amendment No. 28, 82, 140, 163, ++6-, 234 3/4.6-1

LIMITING CONDITION FOR OPERATION 3.6 PRIMARY SYSTEM BOUNDARY (Cont)

A. Thermal and Pressurization Limitations (Cont)

In the event this requirement is not met, achieve stable reactor conditions with reactor vessel temperature above that defined by the appropriate curve and obtain an engineering evaluation to determine the appropriate course of action to take.

3. The reactor vessel head bolting studs shall not be under tension unless the temperature of the vessel head flange I

and the head is greater than the PTLR limit.

4. The pump in an idle recirculation loop shall not be started unless the temperatures of the coolant within the idle and operating recirculation loops are I within the PTLR limits.

5. The reactor recirculation pumps shall not be started unless the coolant temperatures between the dome and the bottom head drain are within the PTLR limits.

SURVEILLANCE REQUIREMENTS 4.6 PRIMARY SYSTEM BOUNDARY (Cont)

A. Thermal and Pressurization Limitations (Cont)

3. When the reactor vessel head bolting studs are tensioned and the reactor is in a Cold Condition, the reactor vessel shell temperature immediately below the head flange shall be permanently recorded.

4. Prior to and during startup of an idle recirculation loop, the temperature of the reactor coolant in the operating and idle loops shall be permanently logged.

5. Prior to starting a recirculation pump, the reactor coolant temperatures in the dome and in the bottom head drain shall be compared and permanently logged.

Amendment No. 27, 28, 42, 82, 140, 209, 2-1-Q 234 3/4.6-2

Pilgrim Reactor Vessel Pressure-Temperature Limits Hydrostatic and Leak Rates is Relocated to PTLR and TS 5.5.9 Amendment No. 28, 82, 94,140, 197,~, 234 3/4.6-9

Pilgrim Reactor Vessel Pressure-Temperature Limits Subcritical Heat up and Cool down is Relocated to PTLR and TS 5.5.9 Amendment No. 28, 82, 94,140, 197,~, 234 3/4.6-10

Pilgrim Reactor Vessel Pressure-Temperature Limits Critical Core Operation is Relocated to PTLR and TS 5.5.9 Amendment No. 28, 82, 94, 140, 197, 22+, 234 3/4.6-11

Programs and Manuals 5.5 5.5 Programs and Manuals "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision O.

d.

Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one subsystem of the MCREC System, operating at the flow rate required by the VFTP. at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.

e.

The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph

c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f.

Each Surveillance Requirement shall be performed within the specified SURVEILLANCE INTERVAL with a maximum allowable extension not to exceed 25 percent of the specified SURVEILLANCE INTERVAL. The SURVEILLANCE INTERVAL requirement is applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage. and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

5.5.9 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

a.

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

i)

Limiting conditions for Operation Section 3.6.A.2 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 document:

i)

SIR-05-044-A "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors", April 2007

c.

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

(continued)

Amendment No. 2:3-1-. 234 5.0-11

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555"()001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 234 TO FACILITY OPERATING LICENSE NO. DPR-35 ENTERGY NUCLEAR GENERATION COMPANY ENTERGY NUCLEAR OPERATIONS, INC.

PILGRIM NUCLEAR POWER STATION DOCKET NO. 50-293

1.0 INTRODUCTION

By letter dated January 24, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML100270054), as supplemented by letters dated September 7 (ADAMS Accession No. ML102580240) and November 4,2010 (ADAMS Accession No. ML103200208), Entergy Nuclear Operations, Inc. (the licensee) submitted a request for changes to the Pilgrim Nuclear Power Station (Pilgrim) Technical Specifications (TSs). The supplements dated September 7 and November 4,2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination noticed in the Federal Register on April 6, 2010 (75 FR 17443).

The proposed amendment would revise TS Section 1.0, Definitions, TS Section 3.6, Primary System Boundary Specifications 3.6.A, and TS Programs and Manuals Section 5.5, to include reference to the Pressure and Temperature Limits Report (PTLR). The proposed PTLR would include revised 34 effective full-power years (EFPY) P-T curves, neutron fluence, and adjusted reference temperature (ART) values.

2.0 REGULATORY EVALUATION

The NRC has established requirements in Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50) to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. The NRC staff evaluates the acceptability of a facility's proposed PTLR based on the following NRC regulations and guidance: Appendix G, "Fracture Toughness ReqUirements,"

to 10 CFR Part 50; Appendix H, "Reactor Vessel Material Surveillance Program Requirements,"

to 10 CFR Part 50; Regulatory Guide (RG) 1.99, Revision 2 (Rev. 2), "Radiation Embrittlement of Reactor Vessel Materials;" Generic Letter (GL) 92-01, Rev. 1, "Reactor Vessel Structural Integrity;" GL 92-01, Rev. 1, Supplement 1, "Reactor Vessel Structural Integrity;" Standard Review Plan (SRP) Section 5.3.2; and GL 96-03. Appendix G to 10 CFR Part 50 requires that facility P-T limits for the reactor pressure vessel (RPV) be at least as conservative as those obtained by applying the linear elastic fracture mechanics methodology of Appendix G to Section

- 2 XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). Appendix H to 10 CFR Part 50 establishes requirements related to facility RPV material surveillance programs. RG 1.99, Rev. 2 contains methodologies for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation.

GL 92-01, Rev. 1 requested that licensees submit the RPV data for their plants to the NRC staff for review, and GL 92-01, Rev. 1, Supplement 1, requested that licensees provide and assess data from other licensees that could affect their RPV integrity evaluations. SRP Section 5.3.2 provides an acceptable method for determining the P-T limits for ferritic materials in the beltline of the RPV based on the ASME Code,Section XI, Appendix G methodology.

The most recent version of Appendix G to Section XI of the ASME Code which has been endorsed in 10 CFR 50.55a, and by reference in 10 CFR Part 50, Appendix G, is the 2004 Edition of the ASME Code. This edition of Appendix G to Section XI of the ASME Code incorporates the provisions of ASME Code Case N-588, "Attenuation to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels," and ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves."

Additionally, Appendix G to 10 CFR Part 50 imposes minimum head flange temperatures when system pressure is at or above 20 percent of the pre-service hydrostatic test pressure.

GL 96-03 establishes the information which must be included in an acceptable PTLR methodology and in an acceptable PTLR. The PTLR also needs to comply with TSTF-419-A, which documents revised guidance for a plant's PTLR. Since this license amendment request (LAR) requested the initial implementation of the PTLR for the Pilgrim unit, the NRC staff's review focused on both the implementation of the Pilgrim PTLR and the appropriate application of the Structural Integrity Associates Topical Report, SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," dated April 2007, to generate the proposed P-T limits.

The NRC has promulgated regulations that ensure the structural integrity of the RPV for light water-cooled power reactors. Specific fracture toughness requirements for normal operation and for anticipated operational occurrences for power reactors are set forth in Appendix G, "Fracture Toughness Requirements," of 10 CFR Part 50. The requirements of Appendix G are imposed by 10 CFR 50.60. Additionally, in response to concerns over pressurized thermal shock events in pressurized-water reactors (PWRs), the NRC issued 10 CFR 50.61, "Fracture toughness reqUirements for protection against pressurized thermal shock events."

To satisfy the requirements of both Appendix G and 10 CFR 50.61, methods for determining fast neutron fluence are necessary to estimate the fracture toughness of the pressure vessel materials. Appendix H, "Reactor Vessel Material Surveillance Program Requirements," 10 CFR Part 50 requires the installation of surveillance capsules, including material test specimens and flux dosimeters, to provide data for material damage correlations as a function of fluence.

The NRC staff considered the requirements of Appendix A to 10 CFR Part 50, "General Design Criteria for Nuclear Power Plants," in its review. Specifically, General Design Criteria (GDC) 30, and 31 are applicable. GDC 14, "Reactor Coolant Pressure Boundary," requires the design, fabrication, erection, and testing of the reactor coolant pressure boundary so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. GDC 30. "Quality of Reactor Coolant Pressure Boundary," requires, among other

- 3 things, that components comprising the reactor coolant pressure boundary be designed, fabricated, erected, and tested to the highest quality standards practical. GDC 31, "Fracture prevention of reactor coolant pressure boundary," pertains to the design of the reactor coolant pressure boundary, stating:

The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions, (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating maintenance, testing and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.

RG 1.190 describes methods and assumptions acceptable to the NRC staff for determining the pressure vessel neutron fluence with respect to the regulatory requirements discussed above.

3.0 TECHNICAL EVALUATION

3.1 Licensee's Evaluation 3.1.1 PTLR Implementation The licensee stated in the January 24, 2010, submittal that, "Attachment 1 of GL 96-03 contains seven technical criteria that the contents of proposed methodology should conform to if license amendments requesting PTLR's are to be approved by the NRC staff. The NRC staff's

[evaluation] of the contents of the BWROG [Boiling Water Reactor Owner's Group] methodology against the seven criteria in Attachment 1 of GL 96-03 [is] provided in Section 3.1 of the SER

[safety evaluation report for the SIR-05-044-A report]." The licensee further stated that, "[the Pilgrim PTLR] based on the methodology and template provided in SIR-05-044-A is being supplied for review and approvaL" 3.1.2 P-T limits The adjusted reference temperature values and P-T limits valid for 34 EFPYs of facility operation using the SIR-05-044-A methodology were documented in the proposed Pilgrim PTLR. The licensee identified the lower intermediate shell longitudinal weld #3 as the limiting material for the Pilgrim RPV. The key parameters in determining the licensee's ART value for the limiting material at the one-quarter of the RPV wall thickness (1/4T) location are shown in Table 7.8 of the PTLR for 34 EFPYs.

Corresponding parameters at the three-quarter of the RPV wall thickness (3/4T) are not provided in the PTLR because PTLR Section 5, "Discussion," implied that the P-T limit curves based on the cooldown transient (the relevant critical location is at 1/4T) are more conservative than the P-T limit curves based on the heatup transient (the relevant critical location is at 3/4T),

Detailed information regarding the Pilgrim P-T limits was contained in Tables 7.1 to 7.7 of the PTLR, which include the pressure and thermal stress intensity factors (Kim and Kit) and the plane-strain fracture toughness (K1c) values at the 1/4T crack tip location for the RPV beltline, the

-4 bottom head, and the upper vessel, supporting the proposed Pilgrim P-T limits valid for 34 EFPYs. For the RPV beltline and bottom head, the ASME Code,Section XI, Appendix G methodology was used to calculate their Kim and Kit, except that a stress concentration factor of 3 was applied to the Kim value in the bottom head calculation. For the upper vessel, the nozzle corner pressure and thermal hoop stresses were based on the finite element method (FEM) results for the Pilgrim feedwater nozzle under the limiting turbine roll event. In addition, a three-dimensional FEM evaluation was also performed for the recirculation inlet nozzle to assure that it is not limiting.

These FEM analyses were performed to determine through-wall thermal and pressure stress distributions for the Pilgrim feedwater nozzles due to a step-change thermal transient, and for the recirculation inlet nozzles (N2) due to a shutdown transient at 100 ofIhr.

The formulas in the SIR-05-044-A report were then used to calculate its Kim and Kit values. In the final step, Table 7.5 of the PTLR indicated that the applied Kit values and the Klc values at the crack tip are utilized to calculate the allowable Kim at the tip of the postulated flaw at the 1/4T location. Pressure is then obtained by comparing the Kim value with the Kip Value based on an FEM model under a unit pressure.

3.2 Staff Evaluation 3.2.1 PTLR Implementation As mentioned in Section 3.1.1 of this SER, Attachment 1 of GL 96-03 requires the licensee evaluate seven technical criteria to demonstrate the acceptability of its PTLR. The NRC staff examined the proposed PTLR and determined that it was developed from the Template PTLR of the SIR-05-044-A report and meets the seven technical criteria:

(1) The PTLR methodology describes the transport calculation methods including computer codes and formula used to calculate neutron fluencies.

The Pilgrim PTLR indicated that the neutron fluence is calculated in accordance with RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," using the RAMA computer code. Hence, the first criterion is met.

(2) The PTLR methodology describes the surveillance program.

The Pilgrim PTLR indicated that Pilgrim has participated in the approved BWR Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP), which meets the requirements of 10 CFR Part 50, Appendix H. Hence, the second criterion is met.

(3) The PTLR methodology describes how the low temperature overpressure protection system limits are calculated applying system/thermal hydraulics and fracture mechanics.

This is not applicable to BWRs, and the Pilgrim unit is a BWR.

(4) The PTLR methodology describes the method for calculating the ART values using RG 1.99, Rev. 2.

- 5 The Pilgrim PTLR indicated that RG 1.99, Rev. 2 provides the methods for determining the ARTs for the beltline materials, with their chemistry factors determined either by the RG tables or the surveillance data information from the BWRVIP ISP. Hence, the fourth criterion is met.

(5) The PTLR methodology describes the application of fracture mechanics in the construction of P-T limits based on ASME Code,Section XI, Appendix G, and the SRP.

Page 6 of the Pilgrim PTLR states that the P-T limits were calculated in accordance with Reference 9.1, i.e., the SIR-05-044-A report. This description is sufficient because the SIR-05-044-A report contains detailed information regarding the application of fracture mechanics in the construction of P-T limits based on ASME Code,Section XI, Appendix G, and the SRP. Further, the Pilgrim PTLR has provided information regarding the finite element analyses performed to generate the pertinent parts of the P-T limits. Hence. the fifth criterion is met.

(6) The PTLR methodology describes how the minimum temperature requirements in Appendix G to 10 CFR Part 50 are applied to P-T limits for boltup temperature and hydrotest temperature.

Again, referencing the SIR-05-044-A report is sufficient because the report contains detailed information regarding the minimum temperature requirements for boltup temperature and hydrotest temperature. Additionally, Page 4 of the Pilgrim PTLR describes the application of the minimum temperature requirement for boltup temperature. Hence, the sixth criterion is met.

(7) The PTLR methodology describes how the data from multiple surveillance capsules are used in the ART calculation.

The seventh criterion of GL 96-03 requires that PTLR contain "supplemental data and calculations of the chemistry factor... if the surveillance data are used in the ART calculation." The licensee's referencing the BWRVIP-135 report, Revision 2, "BWR Vessel and Internals Project, Integrated Surveillance Program (lSP) Data Source Book and Plant Evaluations/' for this information is sufficient because the BWRVIP-135 report contains such detailed chemistry factor calculation using surveillance data for the relevant Pilgrim RPV beltline material. Hence, the seventh criterion is met.

In summary, the implementation of the Pilgrim PTLR is acceptable.

3.2.2 P-T limits To evaluate the proposed P-T limits for the Pilgrim RPV. the NRC staff first confirmed the licensee's selection of the lower intermediate shell longitudinal weld # 1 as the limiting beltline material. The NRC's Reactor Vessel Integrity Database indicated that the initial reference temperature (RT NOT), copper (Cu), and nickel (Ni) values for the limiting material of the Pilgrim RPV are identical to those in the PTLR. Using these material data and RG 1.99, Rev. 2 for materials which do not have two credible surveillance data points, the NRC staff has verified the licensee's 34 EFPY ART value (1/4T) of 78.5 of. The licensee did not perform calculation for the ART value at the 3/4T location, which is relevant to the heatup P-T limit calculation, because

- 6 the SIR-05-044-A report demonstrated that P-T limits for the cooldown transient is bounding To check the overall validity of the RPV beltline material information, the NRC staff also examined material properties other than the limiting material. The NRC staff found that the RPV materials information in Table 7.8 differs from those in the recent Pilgrim license renewal application and issued a request for additional information (RAI) to request for clarification. The SER related to the license renewal of the Pilgrim unit was issued in November 2007.

The licensee's September 7,2010, response clarified that the revised chemistry factor of 300.2 OF that was used for determining the ART value for lower shell longitudinal welds #1, #2, and #3 was based on the ISP surveillance data of the BWRVIP-135 report, and confirmed that the chemistry factors for other beltline materials were derived in accordance with RG 1.99, Rev. 2. The NRC staff accepted this revision because using ISP surveillance data is in accordance with the established NRC staff position regarding use of other plants' surveillance data.

As explained in Section 3.1.2 of this SER, the proposed Pilgrim P-T limits consist of three curves: the RPV beltline, the bottom head, and the upper vessel curves. Since the bottom head P-T limits are the least limiting for the entire pressure temperature range, consistent with the SIR-05-044-A report, the current review focused on the RPV beltline P-T limits and the upper vessel P-T limits. The NRC staff performed independent calculations for both curves in the proposed P-T limits.

For the RPV beltline P-T limits, the NRC staff utilized the ASME Code,Section XI, Appendix G methodology, using the K1c curve as resistance and the pressure dependent Kim formula and the cooldown rate dependent Kit formula as driving forces. Based on the information submitted by the licensee, including the pressure head for accounting the column of water in the RPV provided in the licensee's PTLR, the NRC staff produced almost identical beltline P-T limits for the two ends of Curve B in Figure 6.2 (or Table 7.4) of the PTLR. The NRC staff's exercise also revealed that the proposed P-T limits do not include temperature and pressure instrument uncertainties. This is acceptable because temperature and pressure instrument uncertainty determination is not specified in Appendix G to 10 CFR Part 50, Appendix G to Section XI of the ASME Code, and SRP Section 5.3.2.

Table 7.4 also indicated that the licensee's fluid temperature is the same as the RPV metal temperature at 1/4T. This is a conservative approach in accordance with the SIR-05-044-A report.

For the upper vessel P-T limits, the NRC staff utilized the Kit and Kim formulas in the SIR-05-044 A report (stress coefficients are available in the Pilgrim PTLR) to calculate driving forces, and the ASME Code,Section XI, Appendix G Klc curve to calculate resistance. These pressure and thermal stress coefficients at the nozzle corner were based on the FEM model results for the plant-specific feedwater nozzle (confirmed by the licensee's response dated September 7, 2010) under the limiting turbine roll event, a step-change thermal transient of 456 OF. Table 7.5 of the PTLR listed the calculated KIC, Kim, and P values at different fluid temperatures based on the Kit value of 117.41 ksi.(inch) 112 for the upper vessel region/core not critical Curve B (Figure 6.2).

The NRC staff noted that at low fluid temperatures, the Kit values appear to exceed the corresponding Klc values and issued an RAI, asking how the PTLR methodology is consistent with the SIR-05-044-A report methodology for the upper vessel region.

- 7 The licensee's September 7,2010, response replied that because the RPV can only experience a shock from the saturation temperature to 90 of, the P-T limits at low temperatures are calculated by scaling the KIt of 117.41 ksi.(inch) 1/2 by the ratio of (saturation temperature - 90

°F)/(546 of - 90 OF). This explains the apparent inconsistency among Kit, Kim, and Klc, and the NRC staff verified a couple of P-T limit pairs applying the ratio. Again, the NRC staff's calculation produced almost identical P-T values for a randomly selected point along the lower segment of the proposed P-T limits.

A recent view of BWR RPVs revealed that some RPVs have small-diameter, drill-hole type instrument nozzles (e.g., the water level nozzle) in the beltline region. The concern is that these small-diameter nozzles may be more limiting than the limiting material that is normally identified in the P-T limits. The licensee's September 7, 2010, response states that, H[t]he evaluation of the instrument nozzle is included in Reference 9.3 of the PTLR... To the extent applicable, its effects are included in the evaluation." Hence, this concern is resolved.

10 CFR Part 50, Appendix G contains additional requirements for the minimum metal temperature of the closure head flange and vessel flange regions. These considerations were reflected in the "notches" of the P-T limits. The NRC staff has verified that when P > 20% of the hydro test pressure (- 313 psig), the minimum temperature of 100 of for the pressure test curve, 130 OF for the normal operation/core not critical curve, and 170 of for the normal operation/core critical curve are derived from adding the RT NDT of 10°F for the limiting flange material temperature to 90 of, 120 of, and 160 OF that were specified in 10 CFR Part 50, Appendix G for the three operation conditions. The NRC staff has also verified that when P S 313 psig, the minimum temperature of 60 OF for the pressure test curve and the normal operation/core not critical curve is more conservative than the RT NDT for the limiting flange material temperature that was specified in 10 CFR Part 50, Appendix G.

Based on the above evaluation, the NRC staff determined that the licensee's proposed P-T limits are in accordance with the SIR-05-044-A report and satisfy the requirements of Appendix G to Section XI of the ASME Code and Appendix G to 10 CFR Part 50. Hence, the licensee's proposed P-T limit curves are acceptable for operation of the Pilgrim RPV valid for 34 EFPYs.

3.2.3 Fluence Methodology The licensee stated that fluence calculations supporting the proposed, updated P-T limits were performed using the RAMA fluence evaluation methodology. The RAMA methodology has been generically approved for use at boiling-water reactors, subject to the condition/limitation that plant geometry-specific validation must be performed.

Generic qualification data for the RAMA method currently includes BWR/2 and BWR/4 designs (Susquehanna, Oyster Creek, and Hope Creek). In light of this, the licensee provided qualification data based on dosimetry analysis from the Monticello Nuclear Generating Plant, which is a BWR/3 of comparable vessel design to Pilgrim.

The dosimetry comparison included four copper, four iron, and four nickel wires. The calculation-to-measurement ratios for the set averaged 1.03+/-11 % 1 G, which is well within the 20 percent specified in RG 1.190. There was no readily discernible bias in the validation results.

- 8 Because the licensee has calculated reactor vessel neutron fluence using a methodology that has been generically approved, and found to adhere to the recommendations set forth in RG 1.190, and because the licensee has demonstrated that the methodology is acceptably qualified for use at a BWR/3, the NRC staff finds the fluence calculations acceptable for use as input to the PTLR development.

3.2.4 NRC Staff Position on Technical Specification Task Force (TSTF)-419 By letter to the TSTF dated November 2, 2009, the NRC staff stated that it no longer supports the position reflected in TSTF Traveler 419 (ML092151016). Specifically, the NRC staff has taken issue with the removal of revision numbers and dates from references to NRC-approved methodologies that are referenced in facility TSs. Therefore, the NRC staff requested that the licensee revise its proposed TS to incorporate the revision number and approval date in the reference to Structural Integrity Associates Topical Report SIR-05-044-A, "Pressure Temperature Limits Report Methodology for Boiling Water Reactors," April 2007, that appears in proposed TS 5.5.9.b.

The licensee provided the requested revision in its November 4, 2010, supplemental letter. The NRC staff finds that the revised TS page is acceptable because it includes the revision number and approval date of SIR-05-044-A, thus addressing the concern identified in the November 2, 2009, letter to the TSTF.

3.2.5 Conclusion Based on the NRC staffs review of the information provided in the licensee's submittals dated January 24, September 7, and November 4, 2010, the NRC staff concludes that the proposed Pilgrim PTLR meets GL 96-03 requirements for implementation and, therefore, is approved as part of the Pilgrim licensing basis.

The Pilgrim RPV P-T limits are based on an approved methodology documented in the SIR-05-044-A report. The NRC staff performed independent evaluations and verified that the P-T limits were developed appropriately using the SIR-05-044-A methodology, and the proposed P-T limits valid for 34 EFPYs satisfy the requirements of Appendix G to Section XI of the ASME Code and Appendix G to 10 CFR Part 50. The TS revision to reflect the use of this acceptable methodology is also appropriate.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Massachusetts State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes SRs.

The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment

involves no significant hazards consideration, and there has been no public comment on such finding (April 6, 2010 (75 FR 17443)). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above that (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 issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: B. Parks S.Sheng Date: January 26, 2011

January 26, 2011 Site Vice President Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360-5508

SUBJECT:

PILGRIM NUCLEAR POWER STATION - ISSUANCE OF AMENDMENT REGARDING REVISED PRESSURE AND TEMPERATURE (P-T) LIMIT CURVES AND RELOCATION OF P-T CURVES TO THE PRESSURE AND TEMPERATURE LIMITS REPORT (TAC NO. ME3253)

Dear Sir or Madam:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 234 to Facility Operating License No. DPR-35 for the Pilgrim Nuclear Power Station (Pilgrim), in response to your application dated January 24, 2010 (Agencywide Documents Access Management System (ADAMS) Accession No. ML100270054), as supplemented by letters dated September 7 (ADAMS Accession No. ML102580240) and November 4, 2010 (ADAMS Accession No. ML103200208).

This amendment modifies the Pilgrim Technical Specification (TS) Section 1.0, Definitions, TS Section 3.6, Primary System Boundary Specifications 3.6.A, and TS Programs and Manuals Section 5.5, to include reference to the Pressure and Temperature Limits Report (PTLR). The PTLR includes revised 34 effective full-power years P-T curves, neutron fluence, and adjusted reference temperature values.

A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely, Ira!

Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-293

Enclosures:

1. Amendment No. 234 to DPR-35

2. Safety Evaluation cc w/encls: Distribution via Listserv Distribution:

PUBLIC RidsOgcRp RidsOGCMailCenter RidsN rrDssSrxb LPL1-1 R/F RidsNrrDorlLPL1-1 RidsNrrLASLiUle RidsRgn 1 MailCenter RidsAcrsAcnw _MailCenter RidsN rrDirsltsb RidsNrrDorlDpr RidsNrrPMPilgrim RidsNrrDciCvib BParks, NRR SSheng, NRR RGrover, NRR ADAMS Accession No.: ML110050298 OFFICE LPL1-1/PM LPL1-1/LA NAME RGuzman SUttle DATE' 1/6/11 1/6/11 AJones 1121/11