Text

Technical Specifications Proposed Change No. 258 RPV Fracture Toughness and Material Surveillance Requirements
	ML030870156
Person / Time
Site:	Vermont Yankee
Issue date:	03/26/2003
From:	Balduzzi M; Entergy Nuclear Operations, Entergy Nuclear Vermont Yankee
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	BVY 03-29
	Download: ML030870156 (81)
	v • d • e

Entergy Nuclear Vermont Yankee, LLC

-.-E n terry

= EBatbrEntergy Nuclear Operations, Inc 185 Old Ferry Road Brattleboro, VT 05302-0500 March 26, 2003 BVY 03-29 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271)

Technical Specifications Proposed Change No. 258 RPV Fracture Toughness and Material Surveillance Requirements Pursuant to 10CFR50.90, Vermont Yankee' (VY) hereby proposes to amend its Facility Operating License, DPR-28, by incorporating the attached proposed change into the VY Technical Specifications. This proposed change adopts the Boiling Water Reactor Vessel and Internals Project Integrated Surveillance Program and updates pressure and temperature limitations for the reactor coolant system.

Attachments I and 2 to this letter contain supporting information and the safety assessment for the proposed change. Attachment 3 contains the determination of no significant hazards consideration.

Attachment 4 provides a proposed change to the Updated Final Safety Analysis Report regarding the Integrated Surveillance Program. Attachment 5 provides the marked-up version of the current Technical Specification and Bases pages, and Attachment 6 is the retyped Technical Specification and Bases pages.

VY has reviewed the proposed change in accordance with 10CFR50.92 and concludes that the proposed change does not involve a significant hazards consideration.

VY has also determined that the proposed change satisfies the criteria for a categorical exclusion in accordance with 10CFR51.22(c)(9) and does not require an environmental review. Therefore, pursuant to 10CFR51.22(b), no environmental impact statement or environmental assessment needs to be prepared for this change.

Upon acceptance of this proposed change by the NRC, VY requests that a license amendment be issued prior to the next scheduled refueling outage (Spring 2004) for implementation within 60 days of its effective date. A license amendment is required prior to the end of the next refueling outage because current Technical Specifications for pressure-temperature limitations are only valid through the end of the current operating cycle, and current requirements for the removal of reactor vessel surveillance specimens would necessitate the removal of a surveillance capsule during the next refueling outage. Accordingly, VY respectfully requests timely approval of this license amendment request.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. are the licensees of the Vermont Yankee Nuclear Power Station 4¢c0 1

BVY 03-29 / Page 2 If you have any questions on this transmittal, please contact Mr. Len Gucwa at (802) 258-4225.

Sincerely, Michael A. Balduzzi 7/

Vice President, Operations STATE OF VERMONT )

)ss WINDHAM COUNTY )

Then personally appeared before me, Michael A. Balduzzi, who, being duly sworn, did state that he is Vice President, Operations of the Vermont Yankee Nuclear Power Station, that he is duly authorized to execute and file the foregoing document, and that the statements therein are true to the best of his knowledge and belief.

Thomas B. Silko, Notary Public,-

My Commission Expires February 10_2007 Attachments cc: USNRC Region 1 Administrator USNRC Resident Inspector - VYNPS USNRC Project Manager - VYNPS Vermont Department of Public Service

Docket No. 50-271 BVY 03-29 Attachment 1 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 258 RPV Fracture Toughness and Material Surveillance Requirements Supporting Information and Safety Assessment of Proposed Change

BVY 03-29 / Attachment 1 / Page 1

1.0 INTRODUCTION

1.1 PURPOSE This Proposed Change to the licensing basis of the Vermont Yankee Nuclear Power Station (VYNPS) revises the Technical Specifications (TS) and Updated Final Safety Analysis Report (UFSAR) regarding reactor pressure vessel (RPV) fracture toughness and material surveillance requirements. The specific changes are summarized as follows:

1.1.1 RPV Material Surveillance Program Vermont Yankee (VY) is proposing to revise current, plant-specific RPV material surveillance requirements (SRs) by adopting the Boiling Water Reactor Vessel and Internals Project (BWRVIP) RPV integrated surveillance program (ISP) as the basis for demonstrating compliance with the requirements of Appendix H to 10CFR Part 50, "Reactor Vessel Material Surveillance Program Requirements." In a safety evaluation dated February 1, 2002 (Ref. 1), the NRC staff determined that the BWRVIP ISP was an acceptable alternative to existing BWR plant-specific RPV surveillance programs for the purpose of maintaining compliance with the requirements of Appendix H.

1.1.2 Pressure-Temperature Limitations VY is proposing to update current pressure and temperature (P-T) limit curves for the reactor coolant system that are required by TS 3.6.A, "Pressure and Temperature Limitations." Currently, TS Figures 3.6.1, 3.6.2 and 3.6.3 expire at the end of the current operating cycle. This proposed change updates the pressure and temperature limits for the reactor coolant system through the end of the current operating license. The updated P-T limits are based on a re-calculated RPV neutron fluence using an NRC staff-accepted neutron fluence methodology for boiling water reactors. The revised P-T limit curves are valid through the end of the current operating license or 32 effective full power years (EFPY) and generally satisfy the requirements of Appendix G to 10CFR Part 50, "Fracture Toughness Requirements."

1.2 DESCRIPTION

OF THE PROPOSED CHANGE 1.2.1 RPV Material Surveillance Program Current TS SR 4.6.A.5 (and associated Bases) regarding irradiated reactor vessel surveillance specimens are being revised. Specifically, the plant-specific SR 4.6.A.5 is being removed from TS, and details regarding the BWRVIP ISP (which is being adopted in place of the current plant specific requirements) are being added to the UFSAR. In addition, conforming changes are being made to the TS Bases for Sections 3.6 and 4.6.

Current TS SR 4.6.A.5 requires:

The reactor vessel irradiationsurveillance specimens shall be removed and examined to determine changes in materialproperties in accordancewith the following schedule:

BVY 03-29 / Attachment I / Page 2 CAPSULE REMOVAL YEAR 1 10 2 30 3 Standby The results shall be used to reassess materialproperties and update Figures3.6.1, 3.6.2 and 3.6.3, as appropriate. The removal times shall be referenced to the refueling outagefollowing the year specified, referencedto the date of commercial operation. of this Proposed Change provides a proposed revision to the UFSAR to adopt the provisions of the BWRVIP ISP in place of the existing plant-specific surveillance program.

Because the RPV material surveillance program requirements are being relocated from the TS and incorporated into the UFSAR, the proposed change to the UFSAR regarding the ISP is included in for NRC review.

As noted in proposed UFSAR Table 4.2.4, instead of withdrawing the second surveillance capsule after 30 years of operation, the capsule will be maintained in a "standby" status. Other, changes to the UFSAR which result from the updated P-T calculations are not included in this submittal, but will be made following issuance of a license amendment.

1.2.2 Pressure-Temperature Limitations Current TS Figures 3.6.1, 3.6.2 and 3.6.3 (and associated Bases), which establish P-T limitations for the reactor coolant system are being updated. The subject figures currently contain a restriction on their use, such that the figures are no longer valid after the end of the current operating cycle (Cycle 23). The updated set of P-T curves is valid through the end of the 40-year operating license and was re-defined based on a re-calculation of neutron fluence using an NRC staff-accepted neutron fluence methodology for BWRs. The updated curves are also clarified as described below. Otherwise, the set of P-T limits remains as shown in current TS Figures 3.6.1, 3.6.2 and 3.6.3. In addition, conforming changes are being made to the TS Bases for Sections 3.6 and 4.6.

Current TS Figures 3.6.1, 3.6.2 and 3.6.3 are being replaced by the figures in Attachment 6.

Specific changes entail:

"* Figures 3.6.1, 3.6.2 and 3.6.3 currently contain a statement that each is valid through the end of Cycle 23. That validity duration is being changed to 4.46 x 108 megawatt-hours thermal (MWH(t)).

"* To improve legibility of the curves, the grid line divisions have been changed, the ordinate axis has been identified by 100 psi increments, and more data were used to plot the curves to improve resolution.

" A Note is being added to TS Figure 3.6.2 to specify requirements for minimum temperature when using local test instrumentation during flange tensioning and detensioning operations. The new Note will specify:

During tensioning and detensioning operations with the vessel vented and the vessel fluid level below the flange region, the flange temperature may be monitored with test

BVY 03-29 / Attachment I / Page 3 instrumentation in lieu of process instrumentation for the downcomer region fluid temperature and permanent flange region outside surface temperature. The test instrumentation uncertainty must be less than +/- 2YF. The flange region temperatures must be maintained greater than or equal to 727F when monitored with test instrumentationduringtensioning, detensioning,and when tensioned.

The tabulation of pressure and temperature data on Figure 3.6.3 is being revised to more accurately reflect the plot of the curves (the curves are unchanged). At 1 16'F the bottom head pressure is changed to 413 psig, instead of the current 416 psig. At 120'F, there should be only two data points on Figure 3.6.3, and these are at 253 psig for the upper region and at 439 psig for the bottom head region. Therefore, the tabulation corresponding to a temperature of 120'F will only specify pressures of 439 psig and 253 psig for the bottom head region and upper region, respectively.

1.3 SCHEDULE VY plans to implement the proposed change to support the next refueling outage (i.e., Spring 2004) and subsequent restart. The proposed change involves the elimination of refueling outage work-scope and its approval is needed for post-outage plant restart. Because current TS SR 4.6.A.5 requires that VY remove a RPV material capsule during the next refueling outage, and the current set of P-T curves expires at the end of the current operating cycle (defined as the end of the next refueling outage), a license amendment is required before the end of the refueling outage.

The next refueling outage is currently scheduled to commence on April 3, 2004.

2.0 BACKGROUND

To ensure the structural integrity of RPVs, 10CFR50.60, "Acceptance criteria for fracture prevention measures for light water nuclear power reactors for normal operation," imposes the specific fracture toughness and material surveillance program requirements set forth in Appendices G and H to IOCFR Part 50.

2.1 RPV MATERIAL SURVEILLANCE PROGRAM Licensees of nuclear power plants are required by Appendix H to 10CFR Part 50 to implement RPV material surveillance programs (including the withdrawal and analysis of surveillance capsules) for monitoring changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region which result from neutron irradiation. These programs consist of surveillance capsules installed inside the RPV that include specimens from RPV plate, weld and heat-affected zone materials. These specimens are removed at periodic intervals, tested and analyzed to monitor the radiation embrittlement of the RPV. Appendix H provides two alternative methods for compliance:

The first alternative is the design and implementation of a plant-specific surveillance program that is consistent with ASTM E-185 (Ref. 2). In accordance with this alternative, licensees must comply with either the edition of ASTM E-185 that was current on the issue date of the American Society of Mechanical Engineers (ASME) Code to which the reactor vessel was purchased, or later editions through the 1982 edition as the basis for establishing surveillance capsule withdrawal schedules.

BVY 03-29 / Attachment 1 / Page 4 The second alternative is addressed in paragraph III.C of Appendix H to 10CFR50, "Requirements for an Integrated Surveillance Program," and involves the implementation of an integrated surveillance program in lieu of individual plant-specific RPV surveillance programs. Certain technical and regulatory criteria are set forth in paragraph III.C.

Until recently, each BWR has had its own RPV material surveillance program, and the specimen selection, testing, analysis and monitoring were conducted on a plant-specific basis. Over the past several years, the BWRVIP developed an ISP that meets the criteria defined in Appendix H for an ISP. The NRC staff approved the BWRVIP ISP in a safety evaluation (SE), which was provided to the BWRVIP by letter dated February 1, 2002 (Ref. 1).

The NRC SE concluded that the proposed ISP, if implemented in accordance with the conditions of the SE, is an acceptable alternative to all existing BWR plant-specific RPV surveillance programs for the purpose of maintaining compliance with the requirements of Appendix H to IOCFR 50 through the end of current facility 40-year operating licenses. In NRC Regulatory Issue Summary (RIS) 2002-05 (Ref. 3), NRC endorsed the BWRVIP ISP and provided guidance for BWR licensees in implementing the ISP program.

Implementation of the ISP provides certain benefits. When the original surveillance materials were selected for plant-specific surveillance programs, the state of knowledge concerning RPV material response to irradiation and post-irradiation fracture toughness was not as robust as it is today. As a result, many facilities did not include what would be identified today as the plant's limiting RPV materials in their surveillance programs. Hence, the integrated effort to identify and evaluate materials from other BWRs, which may better represent a facility's limiting materials, should improve the overall evaluation of BWR RPV embrittlement. Also, the inclusion of additional data from the testing of BWR Owners Group Supplemental Surveillance Program capsules will improve overall quality of the data being used to evaluate BWR RPV embrittlement.

Implementation of the ISP is also expected to reduce the costs associated with removing capsules from RPVs and surveillance testing and analysis, since surveillance materials that are of little or no value (either because they lack adequate unirradiated baseline Charpy V-notch data or because they are not the best representative materials) will no longer be tested. In addition, the exposure of personnel to radiation due to the removal and handling of irradiated specimens should be reduced.

By letter dated November 12, 2002 (Ref. 4), the BWRVIP submitted Proprietary Report BWRVIP-86-A (Ref. 5) to the NRC staff for information and review. BWRVIP-86-A represents a compilation of information from several sources upon which the NRC staff based its SE (Ref. 1).

The NRC staff reviewed the information in BWRVIP-86-A and, by letter dated December 16, 2002 (Ref. 6), found that it accurately incorporates all of the relevant information submitted by the BWRVIP to support NRC staff approval of the BWRVIP ISP.

A major consideration in the NRC staff's SE (Ref. 1) deals with BWR RPV fluence calculations.

Specifically, the NRC staff required as a condition to its SE that RPV neutron fluence calculations use a fluence methodology that is acceptable to the NRC staff and is consistent with the guidance found in NRC Regulatory Guide 1.190 (Ref. 7). In addition, if differing fluence methodologies are used (i.e., the methodology used to determine the neutron fluence values for a licensee's RPV differs from the methodology used to establish the neutron fluence values of the ISP surveillance capsules which represent the RPV in the ISP), the results of these differing methodologies are compatible (i.e., within acceptable levels of uncertainty).

BVY 03-29 / Attachment 1 / Page 5 2.2 P-T LIMITATIONS 2.2.1 Technical and Regulatory Basis 10CFR50.60, "Acceptance criteria for fracture prevention measures for light water nuclear power reactors for normal operation," imposes the fracture toughness requirements for the reactor coolant pressure boundary set forth in Appendix G to Part 50. Licensees of nuclear power plants are required by Appendix G to IOCFR Part 50, "Fracture Toughness Requirements," to develop and use P-T limits in order to provide adequate margins of safety during any condition of operation, including anticipated operational occurrences and system hydrostatic tests, to which the reactor coolant pressure boundary may be subjected over its service lifetime.

Appendix G to IOCFR50 describes the conditions that require P-T limits and provides the general bases for these limits. Operating limits based on the criteria of Appendix G, as defined by applicable regulations, codes, and standards, provide reasonable assurance that non-ductile or rapidly propagating failure will not occur.

Appendix G of Section XI of the ASME Boiler and Pressure Vessel Code (the Code), (Ref. 8) forms the basis for the requirements of Appendix G to IOCFR50. The operating limits for pressure and temperature are required for three categories of operation: (1) hydrostatic pressure tests and leak tests; (2) non-nuclear heatup/cooldown and low-level physics tests; and (3) core critical operation.

Pressure-retaining components of the reactor coolant pressure boundary that are made of ferritic materials (including the pressure vessel) must meet the requirements of Appendix G of the Code, as supplemented by the additional requirements in Table 1 of Appendix G to 10CFR50 for fracture toughness during system hydrostatic tests and any condition of normal operation, including anticipated operational occurrences. In addition to beltline considerations, non-beltline discontinuities such as nozzles, penetrations, and flanges may influence the construction of P-T curves.

The P-T limits are not derived from design basis accident analyses, but are prescribed for all plant modes to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause non-ductile failure of the reactor coolant pressure boundary. The P-T limits are acceptance limits because they preclude operation in an unanalyzed condition.

P-T limits are revised when necessary in accordance with Appendix H to 10CFR50 for changes in adjusted reference temperature for nil ductility transition (ARTNT) due to neutron fluence values determined from the analysis of irradiated RPV beltline materials. Upon acceptance of this Proposed Change, the ISP discussed above will provide the dosimetry data and results of fracture toughness tests as the bases for changes in ARTNDT for the VYNPS RPV.

2.2.2 Neutron Fluence Methodology IOCFR50, Appendix G requires the prediction of the effects of neutron irradiation on vessel embrittlement by calculating the ARTNDT and the Charpy Upper Shelf Energy (USE). For reactor vessel beltline materials, including welds, plates, and forgings, the values of ARTNDT must account for the effects of neutron irradiation, as part of the surveillance program of Appendix H to

BVY 03-29 / Attachment I / Page 6 10CFRS0. To predict these effects, NRC Generic Letter 88-11 (Ref. 9) imposes the use of methods described in Regulatory Guide 1.99, Revision 2 (Ref. 10). The fluence values calculated using the methodology described in Regulatory Guide 1.190 satisfy the requirements of Appendix G to 10CFR50 and Regulatory Guide 1.99.

2.2.3 Flaw Analysis The basic parameter in Appendix G to Section XI of the ASME Code (Ref. 8) for calculating P-T limit curves is the stress intensity factor (Ka), which is a function of the stress and a postulated flaw. The Code methodology specifies that licensees determine the reference Ka factors. Code Case N-640 (Ref. 14) permits use of the lower bound static initiation fracture toughness value (Kic) in lieu of Ka.

The methodology of Appendix G to the Code requires that P-T curves satisfy a safety factor of 2.0 on stress intensities arising from primary membrane and bending stresses during normal plant operations (including heatups, cooldowns, and transient operating conditions) and a safety factor of 1.5 on stress intensities arising from primary membrane and bending stresses when leak rate or hydrostatic pressure tests are performed on the reactor coolant system. Table 1 in Appendix G to IOCFR50 provides criteria for meeting P-T limitations of Appendix G to the Code and the minimum temperature requirements for normal and pressure testing operations.

3.0 SAFETY ASSESSMENT 3.1 RPV MATERIAL SURVEILLANCE PROGRAM VY is a participant in the BWRVIP, which developed the NRC staff-accepted ISP for RPV materials and will formally implement the ISP upon NRC issuance of the requested license amendment.

BWRVIP-86-A (Ref. 5) provides the technical and regulatory basis for the BWRVIP ISP and will be incorporated by reference in the VYNPS UFSAR. As noted in the NRC staff's reply to the BWRVIP dated December 16, 2002 (Ref. 6), reference to BWRVIP-86-A is acceptable in lieu of referencing the separate source documents. Attachment 4 of this proposed change is a proposed revision to the UFSAR, which will become effective upon implementation of the requested license amendment.

The BWRVIP ISP is intended to replace the existing plant-specific RPV material surveillance programs with representative weld and base materials data from host reactors. It is not intended that VYNPS be an ISP host reactor. As indicated in the Test Matrix in BWRVIP-86-A, RPV weld and plate surveillance materials from Susquehanna-1 have been selected from among all the existing plant surveillance programs (including the Supplemental Surveillance Program) to represent the corresponding limiting plate and weld material in the VYNPS RPV. Thus, in accordance with the ISP, no further capsules will be removed and tested from the VYNPS RPV. It is anticipated that the next Susquehanna-1 surveillance capsule should be removed from the vessel in year 2012.

Based on the test results of the removed capsules, fluence calculations will be reevaluated using a methodology approved by the NRC and demonstrated to be consistent with the methods described

BVY 03-29 / Attachment 1 / Page 7 in Regulatory Guide 1.190 (Ref. 7). VY used an updated fluence methodology provided by GE Nuclear Energy (GENE) (Ref. 11) and approved by NRC to develop the revised P-T curves.

As shown in Table 4-5 of BWRVIP-86-A, "Detailed Test Plan By Plant," the VYNPS RPV wall is expected to experience the lowest, end-of-life neutron fluence of all domestic BWRs.

Under the ISP, representative capsule data will be provided to each BWR vessel owner for limiting vessel weld and base materials. These data will be evaluated, as appropriate, using the methods in Regulatory Guide 1.99 (Ref. 10) in accordance with Appendix G to IOCFR50 for the determination of ARTNDT values. The relevant data (i.e., Charpy shift results) will be used to re evaluate embrittlement projections for the corresponding vessel beltline materials represented by the materials in the capsule. This re-evaluation will be conducted by VY based on the results determined from testing of representative materials. If changes in P-T limits are required due to a reassessment of the limiting ARTNDT values, changes to the licensing basis will be requested, as appropriate.

The reporting of test results to NRC, including the data required by ASTM E-185 (Ref. 2), and the results of all fracture toughness (i.e., Charpy) tests conducted on the surveillance materials will be made by the BWRVIP program administrator.

Although there are no plans to remove additional material surveillance specimens from VYNPS, the remaining two surveillance capsules will continue to reside in the RPV in accordance with the BWRVIP ISP, in case they are needed in the future as a contingency.

Consistent with the guidance provided in RIS 2002-05 (Ref. 3), and because current TS require withdrawal of RPV specimens, VY is submitting this proposed change as a license amendment request. Current TS SR 4.6.A.5 requires that the second VYNPS surveillance capsule be removed during the refueling outage following the year in which 30 years of commercial operation is reached (i.e., the Spring 2004 refueling).

NRC has previously determined, as documented in Generic Letter 91-01 (Ref. 12) that details of RPV material surveillance programs do not need to be included in the TS, because there would be duplication of controls that have been established by regulations (i.e., Appendix H to 10CFR50).

Therefore, instead of replacing the plant-specific surveillance program requirements in TS 4.6.A.5 with details regarding the ISP, VY will incorporate the ISP into the UFSAR. Because duplication of controls is unnecessary, and adequate controls already exist, it is acceptable to relocate details of the RPV surveillance program to the UFSAR.

VY is requesting a change to the VYNPS RPV material surveillance program required by 10CFR50, Appendix H, and currently implemented through TS SR 4.6.A.5, to incorporate the BWRVIP ISP into the VYNPS licensing basis. The proposed change to VY's RPV material surveillance program meets the regulatory criteria in Paragraph III. C of Appendix H to 10CFR50.

Based on the foregoing considerations, including the prior acceptance of the BWRVIP ISP by the NRC staff, this proposed change is acceptable because it provides an overall improvement in the quality of data that will be obtained, analyzed and reported to NRC for the purpose of monitoring changes in the fracture toughness properties of RPV beltline materials.

BVY 03-29 / Attachment 1 / Page 8 3.2 P-T LIMITATIONS 3.2.1 Current Licensing Basis for P-T Curves VYNPS License Amendment No. 203 (Ref. 13) revised the TS by changing the RPV P-T limit curves specified in TS Limiting Condition for Operation 3.6.A, "Reactor Coolant System Pressure and Temperature Limitations," as graphically represented in Figure 3.6.1, "Hydrostatic Pressure and Leak Tests, Core Not Critical," Figure 3.6.2, "Normal Operation, Core Not Critical,"

and Figure 3.6.3, "Normal Operation, Core Critical." However, because VY's neutron fluence estimate used at that time to support generation of the P-T curves was not based on a methodology acceptable to the NRC staff for current licensing applications, a restriction was placed on the application of the P-T curves. That restriction disallows use of the P-T curves beyond the end of the current operating cycle (i.e., Cycle 23).

3.2.2 Updated P-T Curves The updated P-T curves were established based on the requirements of Appendix G to 10CFR50 to assure that brittle fracture of the RPV is prevented. Attachment 2 to this Proposed Change provides the methodology of calculation used by VY in generating the revised P-T curves (i.e., TS Figures 3.6.1, 3.6.2 and 3.6.3). The revised P-T curves retain the same basic P-T limits as the current curves.

Composite P-T curves were generated for each of the pressure test, core not critical and core critical conditions at 32 EFPY. Attachment 6 includes proposed TS Figures 3.6.1, 3.6.2 and 3.6.3, which also incorporate a tabulation of P-T limits for both the bottom head and upper head regions.

The revised P-T curves (and current curves) differentiate between the bottom head region and upper vessel regions. The methodology used to generate the P-T curves in this submittal is similar to the methodology used to generate the curves approved in license amendment no. 203 (Ref. 13).

In this update, however, the estimate of the RPV neutron fluence was based on a new fluence methodology that follows the guidance of Regulatory Guide 1.190 (Ref. 7). Part of the analysis conducted in developing the P-T curves was to account for radiation embrittlement effects in the core region, or beltline, and ARTNDT values were determined using criteria of Regulatory Guide 1.99 (Ref. 10). However, although VY conducted an analysis in accordance with Regulatory Guide 1.99, the more conservative ARTNDT values used in the prior evaluation were retained.

For the hydrostatic pressure and leak test curve (TS Figure 3.6.1), a coolant heatup and cooldown temperature rate of 40"F/hr or less must be maintained at all times. Similarly, for the normal operation, core not critical (TS Figure 3.6.2) and the normal operation, core critical curve (TS Figure 3.6.3), the P-T curves specify a coolant heatup and cooldown temperature rate of 100"F/hr or less for which the curves are applicable.

The change to TS Figures 3.6.1, 3.6.2 and 3.6.3 to extend their applicability to 4.46 x 10' MWH(t) corresponds to an integrated plant operation of 32 EFPY. This limitation is acceptable because it is based on the re-calculated, expected neutron fluence over 40 years of operation at the current licensed power level, accounting for periods of downtime.

The enhancements made to TS Figures 3.6.1, 3.6.2 and 3.6.3 by slightly revising grid divisions, adding additional 100 psi increments to the ordinate axis, and improving curve resolution are

BVY 03-29 / Attachment I / Page 9 administrative changes of preference. They are acceptable because they do not change any technical requirement and are made to enhance user acuity.

The addition of a Note to TS Figure 3.6.2 to permit use of test instrumentation during tensioning, detensioning, and when tensioned is acceptable because test instrumentation can provide a better method of monitoring bolt-up temperatures during this phase of operations. The use of such instrumentation is limited to the condition when the vessel is vented and vessel fluid level is below the flange region. The establishment of this condition ensures that the vessel cannot be pressurized while relying on test instrumentation. Because test instrumentation is more accurate (conservatively within +/- 21F) than permanent temperature instrumentation (+/- 10'F), a limit of

> 721F may be established when using test instrumentation. A 72°F limit for test instrumentation corresponds to an 80'F limit for permanent temperature instrumentation when the respective instrumentation uncertainties are included. These values are acceptable because the analytical limit for head bolt-up is 70'F (without instrument uncertainty) as stated in current TS 3.6.A.

The changes to the tabulation in Figure 3.6.3 represent a correction of actual values used to generate the current curves. The current tabulation indicates that four different pressure limits were established corresponding to a temperature of 120'F. As can be seen from the curves, there 0

are only two such points for 120'F. Similarly, the change in bottom head pressure at 116 F to 413 psig reflects a past administrative error in transcribing the actual value from the current curve.

These changes to correct the tabulation are acceptable because they do not change actual limits (the curves are unchanged) and reflect the outputs from previous analyses.

3.2.3 Application of ASME Code Case N-640 The updated P-T limits were developed using Section XI, Appendix G of the 1995 Edition with the 1996 Addenda of the ASME Code (Ref. 8). This code edition and addenda incorporated revised stress intensity factors into the Appendix G methodology, which is used to develop the actual P-T limit curves. The revised stress intensity factors are based upon the re-orientation of the postulated defect normal to the direction of maximum stress. NRC has approved this code edition with addenda, as documented in IOCFR50.55a(b)(2).

In addition, the updated P-T limit curves are based, in part, on the application of ASME Code Case N-640 (Ref. 14). Pursuant to 10CFR50.12 and by letter dated April 16, 2001 (Ref. 15), the NRC granted an exemption to allow VY to deviate from the requirements of Appendix G to 10CFR50 in the use of this alternative method.

Code Case N-640 permits application of the lower bound static initiation fracture toughness value equation (K,, equation) as the basis for establishing the P-T curves in lieu of using the lower bound crack arrest fracture toughness value equation (i.e., the Kia equation), which is based on conditions needed to arrest a dynamically propagating crack-the method invoked by Appendix G to Section XI of the ASME Code. Use of the K,, equation in determining the lower bound fracture toughness in the development of the P-T operating limits curve is more technically correct than the use of the Kia equation because the rate of loading during a heatup or cooldown is slow and is more representative of a static condition than a dynamic condition. The Kic equation appropriately implements the use of the static initiation fracture toughness behavior to evaluate the controlled heatup and cooldown process of a reactor vessel.

BVY 03-29 / Attachment 1 / Page 10 3.2.4 Neutron Fluence Calculations In developing the updated P-T limit curves, the VYNPS neutron fluence calculations were also updated. These calculation updates were performed using the NRC-approved General Electric Nuclear Energy (GENE) methodology as documented in GENE's Licensing Topical Report NEDC-32983P-A (Ref. 11). The NRC-accepted (Ref. 16), proprietary methodology is fully described in NEDC-32983P-A and is not repeated herein. In general, GENE's methodology adheres to the guidance in Regulatory Guide 1.190 (Ref. 7) for neutron flux calculations and is based on a two-dimensional discrete ordinates code.

VY's estimate of neutron fluence is based in part on a dosimetry analysis of the first (and only) surveillance capsule removed from VYNPS on March 4, 1983, after 7.54 EFPY of irradiation.

The updated RPV fluence values demonstrate that the vessel fast fluence assumptions in the current P-T curve calculation remain conservative. The updated fluence analysis supports replacing the Cycle 23 expiration date with a 32 EFPY (4.46 x 108 MW-hour) expiration limit.

The revised calculations consist of two parts: First, the GENE methodology was applied to recalculate the surveillance coupon fluence rates. This task served to benchmark the new methodology. The second task involved updating the model to include a modem core design.

VYNPS operating Cycle 21 was selected as representative of recent, modem core designs.

Sensitivity studies of contemplated core loadings, including the current Cycle 23, indicated that peak vessel fluxes are bounded by Cycle 21. The updated fluence calculation is documented in a proprietary report prepared by GENE for VY. A summary of the VY RPV fluence analysis is presented below.

Table 1 Summary of Flux Results Location Flux (n/cm2-s)

RPV Inside Surface - max location 2.96 x 10' Surveillance Capsule (30*) 1.89 x 10' Using the core design for Cycle 21, the revised, calculated peak fast flux (E >1 MeV) at end of life is summarized in Table 1.

The fast neutron fluences at the end of plant life (32 EFPY) were conservatively calculated to be 2.99 x 1017 n/cm 2 and 1.91 x 1017 n/cm for the peak RPV location and the surveillance capsule, respectively. Through the end of calendar year 2002, VYNPS had accumulated approximately 23.8 EFPY of operation.

3.2.5 Regulatory Guide 1.99 and Adjusted Reference Temperature The current and updated P-T curves are based on bounding ARTNDT values of 89"F at 1/4T and 73"F at 3/4T. To ensure compliance with Regulatory Guide 1.99, the new fast neutron fluence at the end of plant life, 2.99 x 1017 n/cm2, was used to assess the adjusted RTNDT of beltline

BVY 03-29 / Attachment 1 / Page 11 components. The shift evaluation followed Position C.1 (surveillance data not available) and the C.1(3) attenuation formula. This evaluation is documented in Attachment 2 and demonstrates that the limiting beltline component (RPV plate 1-14) remained the same, and the ARTNDT values calculated in accordance with Regulatory Guide 1.99 remain bounded by values used to develop the current P-T curves. As demonstrated in Attachment 2, the equivalent fluence, when compared to the updated fast fluence of 2.99 x 1017 n/cm 2, remains very conservative.

Because the capsule and end-of-life (EOL) fluence values have changed, the USE equivalent margin analysis plant applicability assessment (Ref. 17) has been incorporated into Attachment 2 to demonstrate continued compliance with ASME Code Case N-512 (Ref. 18). The prediction of change in Charpy USE was calculated in accordance with Regulatory Guide 1.99. As summarized in Attachment 2, there remains ample margin between the projected decrease in weld and plate USE and the allowable value specified in NEDO-32205 (Ref. 19). Therefore, VYNPS remains in compliance with USE requirements of 10CFR50 Appendix G by demonstrating that the projected decrease in USE per the guidance of Regulatory Guide 1.99 meets bounding limits established in the topical report.

3.2.6 Non-Beltline Regions Non-beltline regions are defined as the vessel locations that are remote from the active fuel and where the EOL neutron fluence is not sufficient (i.e., < 1017 ricm2) to cause any significant embrittlement. Non-beltline components include nozzles, closure flanges, some shell plates, the top and bottom head plates, and the control rod drive penetrations.

Detailed stress analyses of the applicable non-beltline components were performed for the purpose of fracture toughness analysis. The analyses took into account the mechanical loading and anticipated thermal transients. The thermal stresses in the vessel wall are caused by a radial thermal gradient that is created by changes in the adjacent reactor coolant during transient conditions. Transients considered include 100lF/hr startup and shutdown, reactor trip, loss of feedwater heaters or flow, loss of recirculation pump flow, and transients involving emergency core cooling injections.

3.2.7 Head Closure Flange Stresses in the VYNPS RPV head closure flange (predominated by preload stress) establish limits incorporated into the updated P-T curves. For the flange evaluation, membrane and bending stresses were extracted from the original vessel stress report for pressure, preload and thermal expansion loadings. The critical location for head preload is the weld region between the upper head and the head flange. A minimum bolt-up temperature of 70"F was conservatively used and this requirement is maintained in TS 3.6.A.3. This conservatism is appropriate because bolt-up tensioning is one of the more limiting operating conditions (high stress and low temperature) for brittle fracture.

The conclusion of the revised neutron fluence analysis is that the revised TS P-T curves bound the recalculated coupon and RPV fast neutron fluences by a significant margin. The updated P-T curves are acceptable because they satisfy the requirements of 10CFR50.60(a), Appendix G to IOCFR50, and Appendix G to the ASME Code, as exempted by the methods of analyses in ASME Code Case N-640. In addition, the revised P-T curves provide an acceptable margin of safety against RPV brittle fracture.

BVY 03-29 / Attachment 1 / Page 12 3.3 Conclusion/Summary In summary, participation in the ISP will improve the quality of compliance with the regulatory requirements in Appendices G and H to IOCFR50 while reducing cost, exposure, and outage time associated with capsule removal, shipping, and testing. The methodologies used to develop the proposed P-T limit curves satisfy the requirements of the regulations (as modified by application of ASME Code Case N-640). The revised P-T curves and outputs from the ISP (which will be used as appropriate for future adjustments to P-T limits), ensure that adequate RPV safety margins against non-ductile failure will continue to be maintained during normal operations, anticipated operational occurrences, and hydrostatic testing. Together, these measures ensure that the integrity of the reactor coolant system will be maintained for the life of the plant.

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 issuance of the requested license amendment will not be inimical to the common defense and security or to the health and safety of the public.

4.0 REFERENCES

1. NRC letter from W. H. Bateman to C. Terry (BWRVIP Chairman), "Safety Evaluation Regarding EPRI Proprietary Reports 'BWR Vessel and Internals Project, BWR Integrated Surveillance Program Plan (BWRVIP-78)' and 'BWRVIP-86: BWR Vessel and Internals Project, BWR Integrated Surveillance Program Implementation Plan,"' February 1, 2002

2. American Society for Testing and Materials (ASTM) E-185, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels" July 1982

3. NRC Regulatory Issue Summary No. 2002-05, "NRC Approval of Boiling Water Reactor Pressure Vessel Integrated Surveillance Program," April 8, 2002

4. BWRVIP letter from C. Terry to NRC Document Control Desk, "Project No. 704 BWRVIP-86-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan," November 12, 2002

5. Boiling Water Reactor Vessel and Internals Project report, "BWRVIP-86-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP)

Implementation Plan," October 2002

6. NRC letter from W.H. Bateman to C. Terry (BWRVIP Chairman), "NRC Staff Review of BWRVIP-86-A, "BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan," December 16, 2002

7. NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001

BVY 03-29 / Attachment 1 / Page 13

8. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," Appendix G, 1995 Edition, including Summer 1996 Addenda

9. NRC Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Pressure Vessel Materials and its Impact on Plant Operations," July 12, 1988

10. NRC Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials,"

Revision 2, May 1988

11. NEDC-32983P-A, "Licensing Topical Report, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations," Rev. 1, December 2001

12. NRC Generic Letter 91-01, "Removal of the Schedule for the Withdrawal of Reactor Vessel Material Specimens from Technical Specifications," January 4, 1991

13. NRC letter from Robert M. Pulsifer to Michael A. Balduzzi (VYNPC), "Vermont Yankee Nuclear Power Station - Issuance of Amendment Re: P/T Curves (TAC No. MB0764,"

May 4,2001

14. American Society of Mechanical Engineers Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves,Section XI, Division 1,"

February 26, 1999

15. NRC letter from R. M. Pulsifer to M.A. Balduzzi (VYNPC), "Vermont Yankee Nuclear Power Station - Exemption from the Requirements of 10 CFR Part 50, Appendix G (TAC No. MB0763)," April 16, 2001

16. NRC letter from S. A. Richards to J. F. Klapproth (GENE), "Safety Evaluation for NEDC 32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (TAC No. MA9891)," MFN 01-050, September 14, 2001

17. NRC letter from W. R. Butler to D. A. Reid (VYNPC), "Vermont Yankee Nuclear Power Corporation, Review of Equivalent Margin Analysis (TAC No. M89225)," July 20, 1994

18. American Society of Mechanical Engineers Code Case N-512, "Assessment of Reactor Vessels With Low Upper Shelf Charpy Impact Energy Levels,Section XI, Division 1,"

February 12, 1993

19. NEDO-32205, "10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 Through BWR/6 Vessels," Revision 1, November 1993

Docket No. 50-271 BVY 03-29 Attachment 2 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 258 RPV Fracture Toughness and Material Surveillance Requirements Calculation Summary Report for Revised P-T Curves

VYC-829, Rev. 4, A7TACHMENT 1 3-18-2003 CALCULATION

SUMMARY

REPORT FOR REVISED P-t CURVES FOR VERMONT YANKEE NUCLEAR POWER STATION Preparedby: K244~ 4--51e Enrico J. 6eU PE Senior Mechanical/StructuralEngineer Reviewed by: A-1 ý- -:;,3r Michael Selling Mechan'jil/StructuralEngineer Reviewed by:./g 19 Qam s Fitzpatrick-

Senior Lead Mechanical/StructuralEngineer Approved by:

Scott Goodwin Mechanical/StructuralEngineeringSupervisor

CALCULATION

SUMMARY

REPORT FOR REVISED P-T CURVES FOR VERMONT YANKEE NUCLEAR POWER STATION 1.0 Introduction This attachment documents the revised set of pressure-temperature (P-T) curves developed for the Vermont Yankee Nuclear Power Station (VY). This work includes a full set of updated P T curves (i.e., pressure and leak test, core not critical, and core critical conditions) applicable for a gross power generation of 4.46x 108 MWHR(th) (which will bound VY power generation beyond March 12, 2012, the end of VY's current operating license (EOL)).

The curves were developed using the methodology specified in ASME Code Case N-640 [2],

the 1995 ASME Code,Section XI, Appendix G (including the Summer 1996 Addenda) [3],

and 10CFR50 Appendix G [4].

The previous revision of this report was submitted to the NRC on February 23, 2001 in support of VY's TS proposed change 244 [Attachment 2 of Reference 19]. The NRC accepted the P-T curves submitted under proposed change 244 with the condition that for operation beyond Cycle 23, VY submit an amendment request justifying the use of the curves which satisfies the guidance of RG 1.190. [21]

In response VY has revised the vessel fluence evaluation [1]. This revised assessment follows the methodology documented in the GE Licensing Topical Report (LTR) NEDC-32983P-A approved by the U.S. NRC for licensing applications in the Safety Evaluation Report [18] and in general, GE's methodology adheres to the guidance in Regulatory Guide (RG) 1.190 for neutron flux evaluation.

The new EOL fluence value remains enveloped by the conservative RTndt shift values used here and in proposed change 244. This report has been updated to incorporate the revised fluence data and demonstrates that there is no impact to the current P-T limits.

Because the capsule and EOL fluence values have changed, the upper shelf equivalent margin analysis plant applicability assessment [17] has been incorporated into this report to demonstrate continued compliance with ASME Code Case N-512. [16].

In addition to the new fluence value, the grid line divisions on the curves have been changed to make them easier to read. More data was used to plot the curves to improve resolution. In addition, specific requirements for minimum temperature using local test instrumentation have been incorporated for flange tensioning and detensioning operations.

Prior to approval of proposed change 244, the NRC requested that VY provide basis information to support revised initial RTndt values for beltline materials, nozzle geometry data, and stress intensity values used in the development of the P-T curves. VY provided a responce to this RAI VYC-829 R4, Attachment 1, Page 2 of 35

in reference [19]. In this revision there is no change to the initial RTndt and nozzle geometry data provided in Reference [19]. The stress intensity information previously provided [19] has been again included here to facilitate NRC review.

In summary, the revision to this report is being done to incorporate four changes:

1) Incorporate the revised fluence values provided by the GE Report [1].

2) Incorporate the revised upper shelf equivalent margin analysis (EMA) plant applicability form to demonstrate continued compliance with ASME Code Case N-512

[16].

3) Provide enhancements in curve grid division and curve resolution to facilitate operator interpretation.

4) Incorporate detailed minimum temperature requirements for flange tensioning and detensioning.

All changes, except those that are non-essential or of an administrative nature, such as correction of typographical errors, editorial changes or format preferences, are marked with margin bars.

2.0 Material Properties An assessment of the fracture toughness properties of all material used in the VY reactor vessel plate, weld and forgings is provided in Attachment 2 to VYC-829 R4. Estimation of the initial value of the nil-ductility reference temperature (RTNDT) was based on the methods described in Branch Technical Position MTEB 5-2 [5]. Charpy impact and drop weight test data from original construction Certified Materials Test Reports (CMTRs) and as-fabricated material testing [6,7], supplemented by more recent data from Battelle for one beltline plate [8], were used. The resulting initial RTNDT values are listed in Table 1.

For all material adjacent to the reactor vessel flange region, the GE vessel purchase contract required that a nil-ductility transition temperature (NDTT) of 10°F be met. Review of the CMTR data shows that the minimum Charpy energy (longitudinal specimens) was 69 ft-lb at 10°F, with 52 mils lateral expansion reported. Two "no-break" drop weight tests at 20'F were also reported. Based on MTEB 5-2, this justifies an RTNDT = 10'F.

For the limiting material adjacent to the core region, the previous submittal by VY [10] stated that the initial RTNDT of plate 1-14 was 40'F. Further evaluation justifies that the RTNDT can be conservatively taken as 30°F.

VYC-829 R4, Attachment 1, Page 3 of 35

Evaluation of the CMTR data shows that the minimum Charpy energy (from longitudinal specimens) was 42 ft-lb at a test temperature of 10°F. Lateral expansion was not reported. Two no-break drop weight tests at 40'F were reported, justifying the NDTT of

< 30'F. Based on MTEB 5-2, this justifies an initial RTNDT = 30'F.

Evaluation of the "as-fabricated" test data shows that the minimum Charpy energy (from longitudinal specimens) was 65 ft-lb at 40°F. The minimum lateral expansion was 54 mils. Two no-break drop weight tests at 20'F were reported, justifying an NDTT of

_*10°F. Based on MTEB 5-2, this justifies an initial RTNDT < 10°F.

Additional testing by Battelle exhibited relatively low Charpy energy (longitudinal specimens) [8]. At 40'F, 80'F and 120'F, the Charpy energy was 46.5 fi-lb, 57.5 ft-lb and 87.5 ft-lb, respectively with lateral expansion greater than 35 mils in all cases. From this data, it is estimated that the 50 ft-lb Charpy energy could have been achieved at <

70'F. Using the criteria from MTEB 5-2, this also justifies an RTNDT of 30'F.

Similar evaluations conducted in supporting VY calculations (Attachment 2 of VYC-829 R4) establish the initial RTNDT values for all other materials.

Table 2-1 and Table 2-2 show an evaluation of the expected irradiation shift for the beltline 2

plates. The peak end of license (EOL) fast fluence of 2.99 x 1017 n/cm (E>1.0 MeV) used in Table 2-1 is from the Reference 1 GE report. The methodology used by GE to develop this fluence value is documented in GE's Licensing Topical Report (LTR) NEDC-32983P-A [1],

which was approved by the U.S. NRC for licensing applications in the Safety Evaluation Report "Safety Evaluation for NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (TAC No. MA9891)," MFN 01-050, September 14, 2001.

For purposes of determining the P-T curves for the vessel core region materials, VY has elected to maintain the more conservatively shifted ARTNDT values previously used by VY: 89°F at the 1/4T point and 73°F at the 3/4T point. Based on guidance of Reg Guide 1.99 Rev. 2 lower values of ARTNDT could have been used. The NRC highlighted this in their Reference 11 safety evaluation.

The conservatism of employing these ARTNDT values is expressed in terms of equivalent fluence in Table 3. Based on the initial RTNDT values and chemistry factors from Table 2-2, and Regulatory Guide 1.99, Rev. 2 [12] criteria for calculating ARTNDT, the use of the conservative 2

ARTNDT values equates to a minimum end-of-life surface fluence of 1.24 x 1018 n/cm for the surface fluence, 2.99 x 1017 four core region plates. This is well beyond the peak end-of-life n/cm 2calculated for Vermont Yankee by GE [1]. This also confirms that plate 1-14, used for the VY surveillance specimens [9], is the critical plate from the standpoint of brittle failure up to fluence levels well beyond that expected at VY.

Reference 1 also provides the axial distribution of 32-EFPY fast neutron fluence at the peak azimuth of the RPV inside surface. The results of the analysis demonstrate the fast fluence 7 2 outside the active axial fuel zone at the RPV wall is less than lxI10 n/cm . The N4 feedwater nozzles are well above the top of active fuel and the N2 recirculation nozzles are below the VYC-829 R4, Attachment 1, Page 4 of 35

7 bottom of active fuel. Therefore the fluence in these locations is substantially below Ix1l 01 n/cm 2.

Based on the revised fluence projection [1], per Reg Guide 1.99 [12] requirements, we have revised the projected decrease in upper shelf energy (USE) data and reevaluated the decrease against criteria from NEDO-32205 [17], the equivalent margin topical report applicable to VY.

This topical report follows the methods provided in Code Case N-512 [18] and was accepted by the NRC [19].

As summarized in Table 15, there remains ample margin between the projected decrease in weld and plate upper shelf energy and the allowable decrease recommended in topical report NEDO 32205. Therefore VY remains in compliance with USE requirements of 10CFR50 Appendix G by demonstrating that the projected decrease in USE per the guidance of Regulatory Guide 1.99 meets bounding limits established in the topical report.

3.0 P-T Curve Methodology The P-T curve methodology is based on the requirements of References [2] through [4]. There are five regions of the reactor pressure vessel (RPV) that were evaluated in this calculation: (1) the reactor vessel beltline region, (2) the bottom head region, (3) the feedwater nozzle, (4) the recirculation inlet nozzle, and (5) the upper vessel flange region. These regions will bound all other regions in the vessel with respect to considerations for brittle fracture. For the feedwater nozzle, the limiting conditions of sudden injection of 50'F cold water into the nozzle were considered. For the remainder of the locations, 100IF/hr heatup and cooldown were considered for Service Level A/B curves and 40 'F/hr heatup and cooldown were conservatively assumed for pressure and leak test conditions. The bottom head region was independently evaluated for anticipated operational occurrences including rapid cooling following a plant scram and hot sweep transients typically associated with re-initiation of recirculation flow into a relatively colder lower head region following a reactor scram and recirculation pump trip.

3.1 General Approach for Analytical P-T Limit Curves The general approach for development of the P-T curves was as follows:

a. A temperature at the crack tip, TI/4t (i.e., 1/4t into the inside or outside vessel wall surface) is either determined using ASME Section XI, Appendix G methods or is conservatively bounded. The method for each location addressed in discussed in subsequent sections.

b. Calculate the allowable stress intensity factor, Kic, based on TI/4t using the relationship specified by Code Case N-640 [2], as follows:

Klc= 20.734 et° °2('I/'-ARTNrT) + 33.2 VYC-829 R4, Attachment 1, Page 5 of 35

where: TI/4 t = metal temperature at assumed flaw tip (°F)

ARTNDT = adjusted reference temperature for location under consideration and desired EFPY (*F)

Kic = allowable stress intensity factor (ksi 4 t inch)

c. Calculate the thermal stress intensity factor, Krr. This is calculated based on ASME Section XI, Appendix G [3] for the beltline and lower head regions, from alternate analysis for the feedwater nozzle or recirculation inlet nozzle/upper vessel regions, or using membrane and bending stresses from the reactor vessel stress report [13] for the upper flange region.

d. Calculate the allowable pressure stress intensity factor, KIp, using the following relationship:

KIp = (Kic-Krr)/SF where: Kip = allowable pressure stress intensity factor (ksi4 inch)

SF = (Code specified) safety factor

= 1.5 for pressure test conditions

= 2.0 for normal operation heatup/cooldown conditions (Level A/B)

For the upper flange region, the expression also includes an additional term that subtracts the preload stress intensity factor (multiplied by SF) from the numerator of the equation.

e. Compute the allowable pressure, P, from the allowable pressure stress intensity factor, Kip, using either ASME Appendix G [3] for the beltline or alternate analytical values for other locations.

f. Make adjustments for temperature and/or pressure uncertainties and hydrostatic head to TI/4t and P, respectively.

g. Repeat steps (a) through (f) for other temperatures to generate a series of P-T points.

3.2 Adjustments to the Curves The following additional requirements were used to define the P-T curves. These limits are established in Reference [4]:

For PressureTest Conditions (Curve A):

VYC-829 R4, Attachment 1, Page 6 of 35

If the pressure is greater than 20% of the pre-service hydrotest pressure, the temperature must be greater than RTNDT of the limiting flange material + 90'F.

If the pressure is less than or equal to 20% of the pre-service hydrotest pressure, the minimum temperature is conservatively taken as greater than or equal to the RTNDT of the limiting flange material + 60'F. This limit has been a standard GE recommendation for the BWR industry for non-ductile failure protection.

For Core Not CriticalConditions (Curve B):

If the pressure is greater than 20% of the pre-service hydrotest pressure, the temperature must be greater than RTNDT of the limiting flange material + 120'F.

If the pressure is less than or equal to 20% of the pre-service hydrotest pressure, the minimum temperature is conservatively taken as greater than or equal to the RTNDT of the limiting flange material + 60'F. This limit has been a standard GE recommendation for the BWR industry for non-ductile failure protection. This limit is applicable when the flange is tensioned or in the process of being tensioned or detensioned.

10CFR 50 Appendix G requires that temperature be maintained at or above the RTndt of the closure flange.

For Core CriticalConditions (Curve C):

The core critical P-T limits must be 40'F above any Pressure Test or Core Not Critical curve limits. Core Not Critical conditions are more limiting than Pressure Test conditions, so Core Critical conditions are equal to Core Not Critical conditions plus 40'F. In addition, when pressure is less than or equal to 20% of the pre-service hydro test pressure and water level is in the normal range for power operation, the minimum temperature must be greater than or equal to the RTNDT of the limiting flange material + 60'F.

At pressures above 20% of the pre-service hydro test pressure, the minimum Core Critical curve temperature must be at least that required for the in-service pressure test (taken as 1,100 psig), or 160'F above the highest RTNDT of the vessel flange region. As a result of these requirements, the Core Critical curve must have a step at a pressure equal to 20% of the pre-service hydro pressure to the temperature required by the Pressure Test curve at 1,100 psig, or Curve B +

401F, whichever is greater.

The resulting pressure and temperature points constitute the P-T curves. These curves relate the minimum required monitored temperature to the allowable reactor pressure. Applicable temperature and pressure adjustments (described below) are also included in Curves A, B, and C.

VYC-829 R4, Attachment 1, Page 7 of 35

The lower head area of a BWR, due to convection cooling, stratification, and cool CRD flow is subject to lower temperatures than the balance of the pressure vessel. In addition, the RTNDT of the lower head is much lower than the assumed ARTNDT being used for the beltline. The lower head is also not subject to the same high level of stress as the flange and feedwater nozzle regions: Therefore, separate curves were provided for the lower head. These curves are less restrictive than the enveloping curve used for the beltline and the balance of the vessel. This will provide Operator's with a more accurate data for assessment of PT limits for this cooler region.

3.3 Instrument Uncertainty and Hydrostatic Head A conservative evaluation of instrument uncertainty by VY derived the following bounding error due to instruments:

Temperature: +/-1OF Pressure: +/- 30 psig Thus, the derived P-T curves were shifted to the right by 10*F. When adjusted for the maximum effects of hydrostatic head (from the top head), the resulting pressure margins are shown in Table 4, where the conservatively adjusted margins are used in the P-T curves.

During vessel tensioning and detensioning the permanent flange temperature instrumentation is removed and special test instrumentation is applied to monitor flange temperature. During this procedure, the vessel is vented to atmosphere and the vessel fluid level is below the flange region. During this operation the external temperature is equal or lower than the internal temperature, therefore the external test instrumentation can be used as a more accurate and conservative assessment of flange temperature conditions. The test instrumentation is selected to have less than +/- 2'F uncertainty.

3.4 Beltline Evaluation For the beltline evaluation, the equations in ASME Section XI, Appendix G [3] are used to predict the stress intensity factors and temperature shifts for inside and outside l/4T flaws. For the cooldown, KIc was conservatively based on reactor temperature; for heatup, the ASME Section XI, Appendix G methods for estimation of temperature at the 3/4T point in the wall were used. Tables 5-8 provide detailed results for the calculations.

3.5 Flange Region For the flange evaluation, membrane and bending stresses were extracted from the original vessel stress report for pressure, preload and thermal expansion (heatup/cooldown) loadings.

The critical location was determined to be the weld region between the upper head and the head VYC-829 R4, Attachment 1, Page 8 of 35

flange [13]. Stress intensity factors were calculated based on the equations similar to ASME Section XI, Appendix G for membrane and bending stresses except that actual stresses were substituted for the pressure stresses in ASME Section XI. For this region, notes have been added to the P-T curves requiring that the minimum of the fluid or the measured vessel flange skin temperatures be used; thus this temperature may conservatively be used to compute Kic.

At temperatures in excess of the 10CFR50 Appendix G limits, the P-T limits based on the flange are much higher than those resulting from the beltline. Tables 9 and 10 provide detailed results for the critical cases (without the margins discussed in Section 3.2).

The tabulated stress intensity summary for the flange under hydrostatic pressure and leak tests has been updated in this summary report. Table 9 submitted with PC change 244 conservatively applied a 2.0 safety factor to the preload stress intensity for the Pressure Test condition. Table 9 has been updated to include the 1.5 safety factor per ASME XI. This change was done to better highlight the margin between ASME XI Appendix G temperature limits and the GE recommended minimum temperature requirement. The revised stress intensity information is included in the stress intensity summary included in Table 16-1. This change has no impact on the limiting P-T curve.

At low pressure all vessel components, except those components in the flange region, have little stress and are not at risk to brittle failure. The stress of flange region components is predominantly due to preload. With preload removed (unbolted condition) and the vessel depressurized the ASME XI Appendix G minimum temperature requirement for all vessel components are well below 0°F. In Table 17 the ASME XI P-T limits for the flange region without preload are given using the highest thermal and pressure stress intensity from the controlling flange locations. At 0°F the allowable pressure is 637 psig.

3.5 N4 Feedwater Nozzle For the feedwater nozzle, the assessment did not consider heatup and cooldown, but considered the effects of injection of 50'F feedwater into the nozzle at various reactor temperatures, this being the minimum realistic temperature for establishing flow into the feedwater nozzles. The stress intensities for pressure and for the feedwater injection were taken from the VY calculation (VYC-1005) that supported VY's NUREG-0619 feedwater nozzle inspection interval evaluation. In VYC-1005 a 1/8T flaw at the feedwater nozzle blend radius region (1.0 inches base metal, 1.1875 inches including the cladding) was evaluated.

This is considerably larger than the 0.823 maximum allowable flaw size (including cladding) that determines the blend radius inspection interval at VY and has been accepted by the NRC

[14]. Kic for the thermal shock transient was conservatively based on the mean of the injected feedwater and the reactor temperature, whereas the initial temperature is steady state at reactor temperature. The deepest point of the postulated blend radius would actually be slightly more affected by reactor temperature due to the larger exposed area for heat transfer. The results are shown in Table 11.

3.6 N2 Recirculation Nozzle VYC-829 R4, Attachment 1, Page 9 of 35

This nozzle was evaluated because of the relatively high RTNDT of one of the nozzles. An evaluation, based on the similar FW nozzle analysis discussed above, was conducted to determine a conservative stress intensity factor for a 1/4T nozzle comer crack. Cooldown was the only condition evaluated since the postulated flaw is at the inside surface in the nozzle blend radius. No credit was taken for the difference between the fluid temperature and the crack-tip temperature in computing Kic. The results are shown in Table 12 and show that significant margin exists.

3.7 Bottom Head The bottom head evaluation was conducted with methods similar to that for the beltline region.

Since the bottom head has the control rod drive penetrations, the stresses and stress intensity factors were modified. An evaluation of the effects of the penetrations showed that the membrane stresses in the bottom head would be bounded by using a factor of 2.75 times the nominal stress computed for the spherical bottom head. Then, the stress intensity factors were multiplied by a factor of 1.28 based on assuming a flaw aspect ratio (a/L) of zero instead of a 1/6 aspect ratio flaw traditionally utilized for ASME Appendix G evaluations. This approach conservatively accounted for the fact that elliptical cracks could potentially interact with the CRD penetrations in the bottom head region. For the bottom head, the P-T curves were based on the minimum of the bottom head fluid or the measured outside surface temperatures, such that Kjc is based on a minimum temperature.

Sensitivity evaluations were conducted to show that anticipated operating occurrences would not control for the bottom head region. Of significance to a BWR is a reactor scram with recirculation trip. For this transient, the lower head region can cool relatively quickly from normal reactor temperature. Then, if recirculation pumps are restarted, the relatively colder water in the bottom head can be swept out by hot water from the bottom head region.

For the cooldown transients, a transient was synthesized that bounded data taken from a reactor scram transient at VY and another BWR plant. It included cooldown from 5271F to 375°F in 10 minutes, then a 200°F/hr cooldown to 175°F, followed by a 100°F/hr cooldown. This transient showed that the limiting high pressure was 1050 psig (with margins) at the end of the initial rapid cooldown period, and that the low temperature portion of the cooldown was essentially the same as that based on the normal P-T cooldown evaluations. The resulting allowable pressure versus bottom head fluid temperature for an inside 1/4T flaw is shown in Figure 1. This evaluation is conservative since 1) there is normally a slight depressurization following a reactor scram, and 2) the initial assumed cooldown was significantly more severe than experienced at VY.

For the recirculation pump restart transient, the maximum possible pressure and temperature conditions of the water sweeping the bottom head region are at saturated conditions, coming from the upper vessel region. Analysis was conducted to evaluate a VYC-829 R4, Attachment 1, Page 10 of 35

transient temperature and stress intensity factor for an outside 1/4T flaw due to a step change transient in the bottom head. Then, using these results, a limiting step change from any initial bottom head temperature to saturated steam conditions could be iteratively determined such that the Kjc would not be exceeded at the assumed flaw.

The results are shown in Figure 2. Additional pressure margin would be available above 3501F, since the maximum possible value of the step-change temperature difference starts to decrease as a result of BWR operating pressure and temperatures conditions. Also shown on the curve is the expected pressure based on a maximum recommended top-to-bottom temperature difference of 145°F between the top and bottom head region temperatures for recirculation pump start, as recommended in GE Service Information Letter (SIL) 251 [15]. This shows that there is significant margin between the fracture limiting pressure and the pressures expected when using the SIL as a guideline for when the recirculation pumps may be restarted.

4.0 P-T Curves The resulting P-T curves, including the Appendix G to 10CFR50 margins discussed in Section 3.2 are shown in Figures 3 through 5.

During vessel tensioning and detensioning the permanent flange temperature instrumentation is removed and special test instrumentation is applied to monitor flange temperature. When monitoring external flange temperature with local test instrumentation during tensioning and detensioning the temperature should be at least:

+ 10°F (RTNDT of the of the limiting flange material)

+ 60'F (GE Margin)

+ 2*F (Maximum Test Instrument Uncertainty)

= 72'F Therefore when monitoring external flange temperature with local test instrumentation during tensioning and detensioning the flange region temperatures must be maintained greater than or equal to 72 'F. A note has been added to the P-T curve in Figure 4 to specify this requirement.

With the vessel depressurized and the flange detensioned the minimum vessel temperature per 10CFR50 Appendix G is 20°F (RTNDT of the limiting flange material, +10°F, plus instrument 0

uncertainty of permanently installed process instrumentation, 10 1F).

VYC-829 R4, Attachment 1, Page 11 of 35

5.0 References

1. GE-NE-0000-0007-2342-RO, DRF 0000-0007-2342, Revision 0, Class 3, January 2003" Final Report Entergy Northeast Vermont Yankee Neutron Flux Evaluation".

2. ASME Boiler and Pressure Vessel Code, Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves,"Section XI, Division 1, Approved February 26, 1999.

3. ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Nonmandatory Appendix G, "Fracture Toughness Criteria for Protection Against Failure," 1995 Edition, Summer 1996 Addenda.

4. U. S. Code of Federal Regulations, Title 10, Part 50, Appendix G, "Fracture Toughness Requirements," December 1995.

5. Branch Technical Position - MTEB 5-2, "Fracture Toughness Requirements", July 1981, Rev. 1.

6. Pressure Vessel Record Exhibit E "As Fabricated Test Reports," CB&I Contract 9-6201.

7. Pressure Vessel Record Exhibit D "Certified Test Reports," CB&I Contract 9-6201.

8. Battelle Columbus Report BCL-585-84-1, "Testing of Unirradiated Pressure Vessel Surveillance Baseline Specimens for the Vermont Yankee Nuclear Generating Plant,"

3/21/84.

9. Battelle Columbus Report BCL-585-84-3, "Examination, Testing and Evaluation of Irradiated Pressure Vessel Surveillance Specimens from the Vermont Yankee Nuclear Power Station," 8/15/84.

10. Letter from Vermont Yankee Nuclear Power Corporation BVY 89-113, to U.S. NRC, "Proposed Change to Revise the Reactor Vessel Pressure-Temperature Curves in the Vermont Yankee Technical Specifications (Generic Letter 88-11)," 11/10/89.

11. Letter from Nuclear Regulatory Commission, NVY 90-077 to Vermont Yankee Nuclear Power Corporation, "Issuance of Amendment No. 120 To Facility Operating License No.

DPR Vermont Yankee Nuclear Power Station (Tac No. 75499).

12. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, May 1988.

13. Chicago Bridge & Iron Company Stress Report # 9-6201-I, Volume 3, Vermont Yankee Reactor Vessel, Revision 6, 1/06/7 1.

VYC-829 R4, Attachment 1, Page 12 of 35

14. Letter from U.S. Nuclear Regulatory Commission NVY 95-02, "Evaluation of the Request for Relief From NUREG-0619 for Vermont Yankee Nuclear Power Station (TAC No. M88803)," 2/6/95.

15. GE Service Information Letter (SIL) No. 251, "Control of RPV Bottom Head Temperature," 10/31/77.

16. ASME Boiler and Pressure Vessel Code, Code Case N-512, Assessment of Reactor Vessels With Low Upper Shelf Charpy Impact Energy Levels,Section XI, Division 1, 02 12-93.

17. NEDO-32205 Class I, November 1993, Revision 1, "10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 Through BWR/6 Vessels".

18. Letter from U.S.NRC to Chairman of BWR Owner's Group, "Acceptance for Referencing Topical report NEDO-32205, revision 1, "1 OCFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 Through BWR/6 Vessels".

19. Letter from Vermont Yankee Nuclear Power Corporation BVY 0 1-14, to U.S. NRC, "Technical Specification proposed Change No. 244, Response to Request for Additional Information," 2/23/2001.

20. Letter from Nuclear Regulatory Commission, NVY 01-046 to Vermont Yankee Nuclear Power Corporation, "Issuance of Amendment RE: P/T Limit Curves (TAC No. MB0764)

(Tac No. MB0764).

21. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U.S. NRC, March 2001.

22. VY Document TE-2002-050, "Updated RPV Fluence Calculations Evaluation in Support of P/T Curves."

VYC-829 R4, Attachment 1, Page 13 of 35

PT Limit for Recirculation Pump Trip Cooldown with Margins 1400 0

!~

I I . . . . . -,

S800 0*

4000 2000 400 0 50000 6000 00 1000 2000 3000 Bottom Head Fluid Temp, F Figure 1: Bottom Head Recirculation Pump Trip Pressure/Temperature Limit Curve PT Limit for Restart of Recirculation Pump with Margins 1200.0 1000.0

.0 U) 800.0 s PT Limit 600.0 N Psat for T + 145F 400.0 1.

I.

200.0 I I ;i A! .* *,. ! l l i i I

0.0 I-H - I,,¢7',, 7 T I"I -i-7-1-I - -r-- -I-I,!t-ý-

0.0 50.0 100.0 150.0 200.0 250.0 300 0 350 0 400 0 Bottom Head Temperature, F Figure 2: Pressure/Temperature Limit Curve for Recirculation Pump Start VYC-829 R4, Attachment 1, Page 14 of 35

Leak Test and Hydro P-T Curve 40*Flhr Heatup/Cooldown Limit Valid Through 4 46E8 MWH(t)

a a -- .. . . .... *'. .. *",

:-..--;--..-i, :.. ' . .., .." " ' . .. '*'_. ..

, /_i " l .. i.. .I'" .. i.. . . .. .. . _'",*_'. ..__'

.- "-I IlUU ¸ I _' . . . . i i i d i i i Bottom Head Curve, Uooer Regions.

... '----Useminimum of bottom 1100o

..------ fluid temperature and bottom head surface

.... i.... ---

Use minimum of downcomer region fluid temperature and flange region outside

temperature I

surface temperature, except 1000

, -.\N. . . *.

. I .f !----

when flange temperature Is greater than 110°F, use only the downcomer region fluid

.... ...r----r --- -- ........ - -- temperature

900

--i----- ......

i ----------------- :

x 800

~ ~..~ *~.. ~:

- ~ : ~ ~ "'-..

' ~ ~ ~ ~--

~i--:.....: "..

..---------- ' ---- ------- 4-"-.......

0 W

C 700

~~. ----------- - --- -- --- -

I- B to- m --

U ppe r mu o 600 - -- Temperature Head Regions

- .-------- .... - --- . ..-- (F) Hed (psig) Re in....

(psig)

- --- --- - --- - --- - - - -80

-- 0 0 80 665 253 85 712 253

, , ..--- --- . 90 764 253

0) 400 95 821 253 S---- ---- 100 885 253

105 954 253 300 110 1032 253

..,. . . . - - I- - .- - .j-- 1 10 10 32 84 2 S---- 115 1117 885 -

_..--- - ---.. . ------ ----- . '------ -120

"- 1211 932 ""

125 1316 984 ..

. . .- ----- .---- --- 1 30 1 04 2 -

135 1105 100 140 1175

..... ---------------------------.......... 145 - 1253 0 ---

60 80 100 120 140 160 180 200 TEMPERATURE (F)

Figure 3: Pressure Test P-T Curve (Curve A)

VYC-829 R4, Attachment 1, Page 15 of 35

Figure 4: Core Not Critical P-T Curve (Curve B)

Core Not Critical P-T Curve 100F/hr Heatup/Cooldown Limit Valid Through 4.46E8 MWH(t) Upoer Recions, Use minimum of downcomer region fluid temperature and flange region outside surface temperature, except when i --- -- -- - j' flange temperature is 1200 S. ---- -------- ----------

greater than 140¶F, use only the downcomer

/ '---" region fluid temperature J

1100

Bottom Head Curve, I

...- "-----Use

....}...'.:....:.... .--------- Ibottom fluid minimum of I --...I--------

r A E-otto ..erW I

1000 -7 itemperature and S..........'-.bottom head surface --

-- TerMp

(*F)

Had (9:160) eitns (064:11

- o I.... .....

..........-- : --- temperature 7/ -----

..°..

......... --- 80 0 0

--- 80 439 253 900 1--- -r Tr--

Sa._ 1---------------

i 85 474 253

.... ......... --- - . ------ --- - ------ - 90 513 253 555 253 AOO a-i, -- - +- I i. . .---

o,°. . L.l ...

95 100 603 253 mu --- ;-- - -P-i "*[ - __ ---

_°, . --- * ° -- - F - 105 655 253

. . . ....... f a 110 713 253 700 I

l i. - . .. .

115 777 253

..,...,....,.... ..,.. . . .-- ---. 120 848 253 M j....-... ,...L 125 9M 253 600 -.*. .-- - - 130 1013 253 0) zn

... =...*.....,....

...i..............

..I....I--.-.. /..:..L.L.

/-----t-----

.. L° ._..._

135 140 110B 1214 253 253

/. _' _._ _L _

140 1214 830 500

--. -- -. . . . -/--- -----------------

145 1312 889 u)

V) . . .-

...1.- --.

i- --- ---i--------- - 150 - 953

.. ... . ., . .. 155 - 1024 400

. ..

_° 160 - 1103

6-------- -- ---'..

--- 165 - 1190

- ................... . . -- -- *- -- -. 1-7-- 1258 300-~1 __

2-- 0--- 0---- --- During tensioning and detensioning operations with the vessel vented and the vessel 200 - ifluid level below the flange region, the flange temperature may be monitored with test

-------- instrumentation in lieu of process instrumentation for the downcomer region fluid l---

temperature and permanent flange region outside surface temperature The test

instrumentation uncertainty must be less than +1-2°F The flange region temperatures 100 must be maintained greater than or equal to 72 "F when monitored with test

instrumentation dunng tensioning, detensioning. and when tensioned

. ....... . .---...--n--v----,-- ------

A 0

60 80 100 120 140 160 180 200 "TEMPERATURE(F)

VYC-829 R4, Attachment 1, Page 16 of 35

Figure 5: Core Critical P-T Curve (Curve C)

Core Critical P-T Curve 10O*Fhr HeatuplCooldown Limit If Pressure < 253 pslg, Water Level must be within Normal Range for Power Operation Valid Through 4.46E8 MWH(t) 1200 "4

Bottom S. .

All

--I----I-i . . . . . .

F-4---K . . . . . .

Li! J Bottom Head 4-°{ Temp Head Regions .Recion 1100

(*F) (psial Casio) Use minimum of S

S....- ---- .--- ...

80 0 x bottom fluid temperature and

;vlI...

80 253 x 114 253 x bottom head surface

/ -.

F !.. . ;.:. .

L ---

1000- 114 402 x ..... *....*.... ....

temperature 115 407 x a, 116 413 0 900 120 439 253 CL 125 474 253 ----------------... ...--- '

--- 4 130 513 253 ----.. --- ---- --- ----- ---------...'-I CI 135 555 253 800 140 603 253 --- i~~~~~~~~~~~~~ ....... -i-i.... ---- ... ----------......---

145 655 253 .......

~ ~ ~ ~ ~ i---------- I.... .. ..i !--,----,

U, 700 150 713 253 0 155 777 253 --------... '--- - ---

. .------ *---- . .--- 1 : -- . .--

k% 160 848 253 165 926 253 . .... .. .. .. ........ ýr-r--T- ...........

r. ..... . - - -

600 170 1013 253 Z/)

fh 175 1108 253 . . . .. t.J. -J '.~~~~~~~--

-- J---

z 180 1214 253 500 180 1214 830 ... 1 ~~ ~ ~ ~ ~ . .. ....

...... ..------ . .... I..... --- J... ... .... J .

UJ 185 1312 889 190 953 (0 400 195 - 1024

'U C-1103 S..--- i-------- -.... --T--.... ....... i -.... Unoer Regions 200 -

S.... -°°.--....... -----------..... Use minimum of 205 - 1190 ----- ----

downcomer region 300 210 1258 L -I fluid temperature S....-------------- F F-and flange region

. .... 5..

outside surface temperature, except 200-I- 1

.1

.1 F when flange

,. -.- L -

;--I. ..--- ;--;- -- temperature is

... I . .- . . -.

I I greater than 180°F, Il .1',, I

,i iFln

.. I-.,.. . . __

*Il

I I- . .

use only the down.omer region 1uu C '

fluid temperature.

L . J..- .,--...

n i . ...

60 80 100 120 140 160 I18 200 TEMPERATURE (F)

VYC-829 R4, Attachment 1, Page 17 of 35

Table 1: Initial RTNDT for Materials in Vermont Yankee Reactor Vessel Initial Region Material Location RTNDT, 'F Top Head Top Head Dollar 1-1 0 Flange Region Top Head Knuckle 1-5/7 0 Top Head Knuckle 1-2/4 0 Top Head Flange 10 Vessel Shell Flange 1.0

. Upper (#4) Shell 1-10 0 Upper(#4) Shell 1-11 0 Intermediate Shell Ujper Int. (#3) Shell 1-12 10 Region Upper Int. (#3) Shell 1-13 60 Irradiated Shell Lower Int. (#2) Shell 1-14 30' Region Adjacent to Lower Int. (#2) Shell 1-1.5 ..... -10 Core Lower (#1) Shell 1-16 _ . 0 Lower (#1) Shell 1-17 0 Bottom Head Region Skirt Knuckle 17-1 40 Bottom Head Knuckle 1-18/21 30 Bottom Head Knuckle 1-22/25 0 Bottom Head Dollar 1-26 302 Bottom Head Dollar 1-27 0"

_............_Bottom Head Dollar 1-28 302 .

Nozzles Recirculation Nozzle N2B 60 Nozzles (All Others, Incl. Feedwater) 40 All Areas Welds -70

1. Limiting beltline plate used in initial surveillance capsule evaluation [9]

2. Bottom head dollar plate includes all bottom head control rod drive penetrations VYC-829 R4, Attachment 1, Page 18 of 35

Table 2-1: Calculation of Peak Fluence Values Calculation of Effective Peak Fluence Values Units EFPY years 32 Seconds per Year =3600*365*24 sec per 31536000 year Flux at Inside Surface [GE reference 1] n/cmA2/s 2.96E+08 Flux at 1/4 from inside Surface [GE reference 1] n/cmA2/s 2.05E+08 Flux at 3/4 from inside Surface [GE reference 1] n/cmA2/s 8.56E+07 Fluence at Inside Surface using GE flux = flux*EFPY*sec/yr n/cmA2 2.99E+17 Fluence at 1/4 thickness using GE flux = flux*EFPY*sec/yr n/cmA2 2.07E+17 Fluence at 3/4 thickness using GE flux = flux*EFPY*sec/yr n/cmA2 8.64E+16 Vessel Thickness inches 5.06 Fluence at 1/4 thickness by RG1.99 =GE ID Fluence *EXP(-0.24*t/4) n/cmA2** 2.20E+17 Fluence at 3/4 thickness by RG1.99 =GE ID Fluence *EXP(-0.24*3*t/4) n/cmA2** 1.20E+17

"**TheRGI.99 C.1(3) attenuation formula results in conservative Fluence Values at the 1/4t and 3/4t locations when compared to values calculated from GE flux values provided in Reference 1. Conservatively these higher values are used in the Ref Guide 1.99 Section C.A shift evaluation below.

Table 2-2: Evaluation of Shift in RTNDT for Core Region Plates Shift in accordance with 1.99 Rev. 2 Plate 1-14 1-15 1-16 1-17 Weld Initial RTNDT OF 30 -10 0 0 -70 Cu w/% 0.11 0.14 0.13 0.12 0.04 Ni w/% 0.63 0.66 0.59 0.61 1 Chemistry Factor, CF 74 102 91 83 54 delta RTNDT @ 1/4 T Based on Higher OF 13.5 18.6 16.6 15.2 9.9 RG1.99 fluence.

delta RTNDT @ 3/4 TBased on Higher OF 9.2 12.6 11.3 10.3 6.7 RG1.99 fluence.

Sig-1, Standard Deviation of Initial RTNDT 0.0 0.0 0.0 0.0 0.0 Margin@ 1/4T=2*sqrt(Sig-IA2+Sig-deltaA2) OF 13.5 18.6 16.6 15.2 9.9 Sig-delta, Standard Deviation of delta °F 6.8 9.3 8.3 7.6 4.9 RTNDT @ 1/4T Margin@ 3/4T=2*sqrt(Sig-IA2+Sig-deltaA2) OF 9.2 12.6 11.3 10.3 6.7 Sig-delta, Standard Deviation of delta °F 4.6 6.3 5.6 5.1 3.3 RTNDT @ 3/4T Adjusted RTNDT @ l/4T OF 57.0 27.3 33.2 30.3 -50.3 Adjusted RTNDT @ 3/4T OF 48 15 23 21 -57 NOTE: Sig-delta lesser value of 17'F for base metals and 28'F for welds or 1/2 delta RTNDT VYC-829 R4, Attachment 1, Page 19 of 35

Table 3: Calculation of Equivalent Peak Beltline Fluence Values Find Re2 Guide 1.99 eqiuivalent fluence Calculation of Effective Peak Beltline Fluence Units that matches ARTNDT used by VY Value Plate 1-14 1-15 1-16 Equivalent Factor on Fluence, k*2.99x10A17 k 4.13 11.15 8.85 Shift in accordance with 1.99 Rev. 2 32 EFPY 32 EFPY 32 EFPY Effective Inside Surface Fluence n/cmA2 1.24E+18 3.34E+18 2.65E+18 Value=k*2.99x OA117 Vessel Thickness inches 5.06 5.06 5.06 Fluence at 1/4 thickness n/cmA2 9.12E+17 2.46E+18 1.95E+18 Fluence at 3/4 thickness n/cmA2 4.97E+17 1.34E+18 1.06E+18 Initial RTNDT OF 30 -10 0 Chemistry Factor, CF 74 102 91 delta RTNDT @ 1/4 T OF 29.5 63.3 51.3 delta RTNDT @ 3/4 T OF 21.6 48.8 39.1 Sig-I, Standard Deviation of Initial RTNDT 0.0 0.0 0.0 Margin@ 1/4T=2*sqrt(Sig-IA2+Sig-deltaA2) OF 29.5 34.0 34.0 Sig-delta, Standard Deviation of delta RTNDT @ OF 14.7 17.0 17.0 1/4T Margin@ 3/4T=2*sqrt(Sig-IA2+Sig-deltaA2) OF 21.6 34.0 34.0 Sig-delta, Standard Deviation of delta RTNDT @ OF 10.8 17.0 17.0 3/4T Adjusted RTNDT @ 1/4T OF 89.0 87.3 85.3 0F 73 73 73 Adjusted RTNDT @ 3/4T Sig-delta lesser value of 17 0 F or 1/2 delta RTNDT NOTE:

Table 4: Pressure Margins at Locations of Interest Location Instrument Static Head Total Margin Total Margin Uncertainty, Pressure, psi Calculated, psi Used, psi psi Closure Head Flange 30 3.72 33.72 35.0 N4 FW Nozzle 30 10.54 10.54 45.0 Bottom of Core Region 30 19.87 19.87 50.0 N2 Recirculation Nozzle 30 20.65 20.65 55.0 Bottom Head 30 27.36 27.36 60.0 VYC-829 R4, Attachment 1, Page 20 of 35

Table 5: P-T Evaluation - Beltline Hydrostatic Test (Heatup)

Pressure-Temperature Curve Calculation (PressureTest wiHeatup = Curve A)

Inputs: Plant =

Component = Beltilne Vessel thickness, t = 6.0600 inches, so 4It = 2 249 ch "Iinc Vessel Radius, R = 103.1876 inches ARTNDT = 73.0 "F 40 *F/hr Heatup Rate, HU =

KIT = 1.73 ksi*inch"Iz (for cooldown rate above)

MT = 0.26 (From App G, Fig G-2214-1)

AT,14t= 6.1 °F = (KrrIMT) 0 92 using Figs G-2214-1 & G-2214-2 Safety Factor = .50, (for hydrotest)

Mm = 2.009 (for inside surface axial flaw) 1 Temperature Adjustment = 10.0 :F Pressure Adjustment = J60-0

psig (hydrostatic pressure + Uncertainty)

Fluid Calculated Adjusted Adjusted Pressure Temperature Pressure for Temperature 1/4t P for P-T Curve P-T Curve T Temperature Kic Kip (OF) (OF) (ksi*inch1 2 ) (ksi*inchI 12 ) (psig) (OF) (psig) 439 4478 28.69 700 60.0 650 500 45.99 2951 720 65.0 670 550 48 9 53.9 47.34 3040 742 70 0 692 60.0 58.9 4883 31.39 766 75 0 716 65.0 50.47 32.49 793 80 0 743 700 63.9 52.29 33.70 823 85 0 773 750 68 9 54.29 3504 855 90 0 805 80 0 73.9 78 9 5651 36.52 891 95 0 841 85.0 83.9 5896 3815 931 1000 881 90.0 61.67 3996 975 105 0 925 95.0 88.9 93.9 6467 41.96 1024 1100 974 1000 98 9 67.98 4416 1078 1150 1,028 1050 71.64 4660 1138 1200 1,088 1100 103.9 1089 7568 49.30 1203 125 0 1,153 115.0 1139 80.15 52.27 1276 1300 1,226 120.0 118.9 85.08 5557 1356 1350 1,306 125.0 VYC-829 R4, Attachment 1, Page 21 of 35

Table 6: P-T Evaluation - Beltline Hydrostatic Test (Cooldown)

Pressure-Temperature Curve Calculation (PressureTest wi Cooldown = Curve A)

Inp uts: Plant = -Ya-nkee Component = Beltline Vessel thickness, t = 6.0600 inches, so 4= 2 249 4inch Vessel Radius, R = Ii03.1876 inches ARTNDT = 89.0 *~F Cooldown Rate, CR = 40 *F/hr KIT =

2.20 ksi*,nch"U (for cooldown rate above)

MT= "0.26 (From App G, Fig G-2214-1)

AT114t 3.7 ?F = (KilIMT) 044 using Figs G-2214-1 & G-2214-2 Safety Factor = 1.60, (for hydrotest)

Mm = 2.083 (for inside surface axial flaw)

Temperature Adjustment = 10.0 'F Pressure Adjustment = 60O psig (hydrostatic pressure + Uncertainty)

Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature Kip P for P-T Curve P-T Curve Kic (OF) (OF) (ksi*inch11 2

) (ksi*inch 12) (psig) (*F) (psig) 50 0 42.70 27.01 636 60.0 586 50 0 55.0 4370 27.67 651 65.0 601 550 60.0 4481 2841 669 70 0 619 600 65 0 65.0 4603 29.22 688 75 0 638 70.0 47.38 30.12 709 8o 0 659 70.0 75 0 48.87 31.12 733 85 0 683 75.0 800 80 0 50.52 32.22 758 90 0 708 85 0 85 0 5234 3343 787 95 0 737 90.0 90 0 54.35 3477 819 100.0 769 95.0 95 0 5658 36.25 853 105.0 803 1000 5904 37.89 892 1100 842 100.0 1050 1050 61 75 3971 935 1150 885 110.0 110.0 6476 41.71 982 120 0 932 1150 115.0 6808 4392 1034 125.0 984 120 0 120 0 71 74 4637 1092 1300 1,042 125.0 1250 75.80 49.07 1155 1350 1,105 130.0 130.0 80.28 52.05 1225 1400 1,175 1350 1350 8523 5535 1303 1450 1,253 VYC-829 R4, Attachment 1, Page 22 of 35

Table 7: P-T Evaluation - Beltline Level A/B (Heatup)

Pressure-TemperatureCurve Calculation (Core Not Critical!Heatup = Curve B)

Inputs: Plant =

Component = Beltline ,

Vessel thickness, t = 5.o0600 inches, so 4t = 2249 'Iinch Vessel Radius, R = 103.1875 inches ARTNOT = 73.0 FF Heatup Rate, HU = 1006 S*F/hr ksi*inchl/2 (for heatup rate above)

K4T= 4.34 MT = 0.26 (From App G, Fig G-2214-1)

AT1/4t= 15.3 '*F = (Kir/MT)* 0 92 using Figs. G-2214-1 & G-2214-2 Safety Factor = 2.00 , (for level NB)

Mm = 2.009 (for outside surface axial flaw)

Temperature Adjustment =

Pressure Adjustment = I psig (hydrostatic pressure + uncertainty)

Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature Kic K~p P for P-T Curve P-T Curve

(*F) (OF) (ksi*incht2) (ksi*inchlf2) (psig) (OF) (psig) 500 34.7 4283 1925 470 600 420 55 0 397 43.84 1975 482 65 0 432 60 0 447 4496 2031 496 70 0 446 650 49.7 4620 2093 511 75 0 461 70.0 54.7 47.57 2161 528 80 0 478 75.0 597 4908 22.37 546 85.0 496 80.0 64.7 5075 2320 566 90.0 516 85 0 69.7 5259 24.13 589 95 0 539 900 74.7 5463 25.15 614 100 0 564 950 79.7 5689 2627 641 1050 591 100.0 847 5938 27.52 672 1100 622 1050 89 7 6213 2890 705 1150 655 1100 947 6517 3042 743 1200 693 1150 997 6853 3210 784 125 0 734 120.0 104.7 72.25 3396 829 1300 779 1250 1097 76.36 3601 879 135 0 829 130.0 1147 80.90 3828 934 1400 884 135.0 1197 8591 4079 996 145 0 946 140.0 124.7 91.46 43.56 1063 150.0 1,013 145.0 1297 97.58 4662 1138 1550 1,088 150.0 134.7 10436 5001 1221 1600 1,171 1550 139.7 11184 5375 1312 1650 1,262 VYC-829 R4, Attachment 1, Page 23 of 35

Table 8: P-T Evaluation - Beltline Level A/B (Cooldown)

Pressure-TemperatureCurve Calculation (Core Not Critical!Cooldown = Curve B)

Inputs: Plant =

Component = Bleltlin~e Vessel thickness, t = 5.0600 *inches, so 4It = 2249 'Iinch Vessel Radius, R = 103.1876 inches ARTNDT =

Cooldown Rate, CR =

K= .6.49 ksilnchl/2 (for cooldown rate above)

MT = 0.26 1(From App G, Fig G-2214-1)

AT114t = 9.3 , F = (KrT/MT)

0 44 using Figs. G-2214-1 &G-2214-2 Safety Factor = 2.00 (for level NB)

Mm. = 2.083 (for inside surface axial flaw)

Temperature Adjustment = 10.0 '

Pressure Adjustment = 500Jpsig (hydrostatic pressure + uncertainty)

Fluid Calculated Adjusted Adjusted Temperature 1/4t Pressure Temperature Pressure for T Temperature Kic Kip P for P-T Curve P-T Curve (OF) ('F) (ksi*inchlt2) (ksi*inchlfl) (psig) (OF) (psig) 50 0 50 0 42.70 1861 438 60 0 388 550 550 4370 1911 450 65 0 400 60.0 60.0 44.81 1966 463 70 0 413 65 0 65.0 4603 2027 477 75 0 427 70 0 70.0 47.38 20.95 493 8o 0 443 75 0 75.0 4887 21.69 511 85 0 461 80.0 80.0 50.52 22.51 530 90 0 480 85 0 85 0 52.34 23.43 551 95.0 501 900 90 0 54.35 2443 575 1000 525 950 95 0 5658 25.54 601 1050 551 1000 1000 5904 2677 630 110.0 580 1050 1050 61.75 28 13 662 115.0 612 110.0 110.0 64.76 29.63 698 120.0 648 1150 1150 6808 31.29 737 125.0 687 1200 1200 71.74 3313 780 1300 730 1250 1250 7580 35.15 828 1350 778 1300 1300 8028 37.39 880 1400 830 135.0 1350 8523 3987 939 1450 889 1400 140.0 9070 4261 1003 1500 953 1450 145.0 9675 4563 1074 1550 1,024 1500 150.0 10343 48.97 1153 1600 1,103 1550 1550 11082 52.66 1240 1650 1,190 160.0 1600 11898 56.75 1336 1700 1,286 VYC-829 R4, Attachment 1, Page 24 of 35

Table 9: P-T Evaluation - Flange Hydrostatic Test (Heatup)

Pressure-TemperatureCurve Calculation (Pressure Test - Upper Flange 2- Heatup)

Inputs" Plant = Ya*--

Flinkee Uui Component = Upper Flange 2 :Upper FlangelHub Intersection Axial Flaw Vessel thickness, t = NIA 'inches Vessel Radius, R = NA inches ARTNOT = 10.0 'F == EFPYs 2

KIT+ 1.5 X KiPt. 70.62 ýýIksi inch" (Note Factor of 1 5 Is S afety Factor)

Safety Factor = 1.50 (for hydrotest) K, ksiinch1' 2 Kip for 1000 psig = 10.30 ' ksi inch"' KIpL=I 0*Preload = 45 Temperature Adjustment = 10.0 F Krr=Thermal = [ 2.072 Pressure Adjustment = 5.0+, psig (hydrostatic pressure + Uncertainty)

Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature Kic Kip P for P-T Curve P-T Curve

(*F) ('F) (ksi*inchlI 2) (ksi*inchI 2) (°F) (psig)

(psig) 0 0.0 5018 -1363 -1323 10 -1358 5 50 5196 -1244 -1208 15 -1243 10 10.0 5393 -11 13 -1080 20 -1115 15 150 5611 -967 -939 25 -974 20 200 5852 -806 -783 30 -818 25 250 61.19 -629 -611 35 -646 30 300 64.13 -4.33 -420 40 -455 35 350 67.38 -2.16 -210 45 -245 40 400 7098 024 23 50 -12 45 45 0 74.95 2 89 280 55 245 50 50 0 79.34 5 81 565 60 530 55 55 0 8420 905 879 65 844 60 60 0 8956 1263 1226 70 1191 65 65 0 9549 1658 1609 75 1574 67 66.9 97.93 18 20 1767 77 1732 70 70.0 10204 20.94 2033 80 1998 75 75 0 10928 25.77 2502 85 2467 80 80 0 11728 31.11 3020 90 2985 VYC-829 R4, Attachment 1, Page 25 of 35

Table 10: P-T Evaluation - Flange Level A/B (Heatup)

Pressure-Temperature Curve Calculation (Core Not Critical- UpperFlange 2- Heatup)

Inputs: Plant =

Component = Upper Flange 2 Upper Flange/Hub Intersection Axial Flaw Vessel thickness, t = NIA+ inches Vessel Radius, R = , NA inches ARTNOT = 10.0 s*F r==2e=-F>

Krr + 2 x KIPL 96.58 SkSliinch' (Note Factor of 2 Is Safety Factor)

(for level A/B) K, ksiinch11 Safety Factor = 2.00 Klp for 1000 psig = 10.30 ksi*,nchlu KIpL=1.0*Preload =

Temperature Adjustment = . 10.0 :°F KiT=Thermal = 5.18 Pressure Adjustment = *~ psig (hydrostatic pressure + uncertainty)

Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature Kic 2 Kip P for P-T Curve P-T Curve

(°F) (°F) (ksi*lnchl1 ) (ksl*inch )

1t 2 (psig) (°F) (pslg)

-15 -150 4578 -2540 -2466 -5 -2501

-10 -100 4710 -24.74 -2402 0 -2437

-5 -5 0 4856 -24.01 -2331 5 -2366 0 00 5018 -23.20 -2253 10 -2288 5 50 51.96 -22.31 -2166 15 -2201 10 100 5393 -21.32 -2070 20 -2105 15 150 5611 -2023 -1964 25 -1999 20 200 5852 -1903 -1847 30 -1882 25 250 61.19 -17.70 -1718 35 -1753 30 300 64.13 -1622 -1575 40 -1610 35 350 67.38 -1460 -1417 45 -1452 40 400 70.98 -1280 -1243 50 -1278 45 45 0 74.95 -1081 -1050 55 -1085 50 50 0 7934 -862 -837 60 -872 55 55.0 8420 -619 -601 65 -636 60 60.0 8956 -3.51 -341 70 -376 65 65 0 9549 -055 -53 75 -88 66 66 0 9675 0.08 8 76 -27 67 67.0 9803 0.73 70 77 35 68 68 0 9934 1.38 134 78 99 69 69 0 10068 2.05 199 79 164 70 700 10204 2 73 265 80 230 71 71.0 10343 342 333 81 298 72 72 0 10485 4 13 401 82 366 73 730 10630 4 86 472 83 437 74 740 107.77 560 543 84 508 75 75 0 109.28 635 616 85 581 76 76 0 110.82 7.12 691 86 656 77 77.0 112.38 7.90 767 87 732 78 78.0 11398 870 845 88 810 79 79.0 115 62 9 52 924 89 889 80 80 0 117.28 1035 1005 90 970 VYC-829 R4, Attachment 1, Page 26 of 35

Table 11: P-T Evaluation - Feedwater Nozzle Level A/B Pressure-TemperatureCurve Calculation (Core Not Critical- FW Injection - Comer Nozzle Crack)

Inputs: Plant = '-Y~ankee -

Component = FW Nozzie Blend Vessel thickness, t = - NIA inches Vessel Radius, R = NIA inches ARTNDT = 410.0 F A11IEFPYs 1 KIT for 552F - 5OF Step = 106.56 ksi*inch"' Temp. Change 502 *F Step Safety Factor = , 2.00 (for level A/B)

K1pfor 1025 psig = 33.80 ksi'inch"'

Temperature Adjustment = 10.0 (F Pressure Adjustment = ~psig (hydrostatic pressure + uncertainty)

Fluid Calculated Adjusted Adjusted Pressure Temperature Pressure for Temperature 1/8t T Temperature Kic K11 Kip P for P-T Curve P-T Curve

(*F) ('F) (ksi*inchin) (ksl*inch12) (ksi*inch'12) (psig) (°F) (psig) 50 500 5852 0 00 2926 887 60 842 55 52 5 5982 1.06 2938 891 65 846 60 550 61.19 212 2953 896 70 851 65 57.5 6262 318 2972 901 75 856 70 600 6413 4 25 2994 908 80 863 62 5 6572 531 3021 916 85 871 75 65 0 67.38 6 37 3051 925 90 880 80 67.5 69.14 7.43 3085 936 95 891 85 70 0 70.98 849 31 24 948 100 903 90 72.5 72.92 9 55 31 68 961 105 916 95 75 0 7495 10.61 3217 976 110 931 100 105 77.5 77.09 11.67 32.71 992 115 947 80 0 79.34 1274 33.30 1010 120 965 110 82.5 81 71 1380 3396 1030 125 985 115 850 84.20 1486 3467 1051 130 1006 120 87.5 8681 1592 3545 1075 135 1030 125 130 900 89.56 1698 3629 1100 140 1055 92 5 9245 1804 3720 1128 145 1083 135 95 0 9549 19.10 3819 1158 150 1113 140 145 97.5 9868 2017 3926 1191 155 1146 150 100.0 102.04 21.23 4041 1225 160 1180 102.5 105.57 22.29 41.64 1263 165 1218 155 105 0 10928 23.35 4296 1303 170 1258 160 VYC-829 R4, Attachment 1, Page 27 of 35

Tablel2: P-T Evaluation - Recirculation Nozzle Level A/B Pressure-TemperatureCurve Calculation (Core Not Critical- N2 Recirc Nozz - Cooldown)

Inputs: Plant = -Yankee Component = N2 Reclrc Noz Vessel thickness, t = NIA Vessel Radius, R =

ARTNDT = 60.0 -- -> [All1EPPYs7 KIT 25.07 ksilinchlu Safety Factor = 2.00 (for level AM)

K1p for 1025 psig = 44.25 'ksi*inch"'

Temperature Adjustment = 10.0 pF Pressure Adjustment = 5.0 __psig (hydrostatic pressure + uncertainty)

Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature Kic Kip P for P-T Curve P-T Curve

("F) (°F) (ksl*inchll 2 ) (ksi*inchlr) (psig) (°F) (psig) 7.19 166 10 111 0 0.0 3944 5 50 40.10 7.52 174 15 119 10 100 4083 7.88 183 20 128 15 15.0 41.63 8 28 192 25 137 20 20 0 42.52 8 72 202 30 147 25 250 4350 921 213 35 158 30 30 0 4458 9 75 226 40 171 35 35.0 4578 1035 240 45 185 40 40.0 47.10 11.01 255 50 200 45 45 0 48.56 11 75 272 55 217 50 50 0 50.18 12.55 291 60 236 55 550 51.96 1345 311 65 256 60 600 53.93 1443 334 70 279 65 65.0 5611 1552 360 75 305 66 664 5678 1586 367 76 312 70 70 0 5852 1673 387 80 332 70 70 3 5870 1681 389 80 334 75 75.0 61.19 1806 418 85 363 80 80.0 64.13 19.53 452 90 397 85 85.0 67.38 21.16 490 95 435 90 90.0 7098 22.95 532 100 477 95 95 0 7495 24.94 578 105 523 100 100.0 79.34 27.14 629 110 574 105 105 0 8420 2956 685 115 630 110 110.0 89.56 3225 747 120 692 115 1150 9549 3521 816 125 761 120 1200 102.04 3848 891 130 836 125 1250 109.28 42.10 975 135 920 130 1300 117.28 4611 1068 140 1013 VYC-829 R4, Attachment 1, Page 28 of 35

Table 13: P-T Evaluation - Bottom Head Hydrostatic Test (Cooldown)

Pressure-Temperature Curve Calculation (PressureTest wi Cooldown = Curve A)

Inputs: Plant = Yankee Component = BoL. head Vessel thickness, t = 56.9375 inches, so 4t = 2437 4inch Vessel Radius, R = 103.1875 inches ARTNDT = 30.0 Cooldown Rate, CR = 40 *F/hr Krr= 4.19 ksiinch"' (for cooldown rate above)

MT= NIA (From App G, Fig G-2214-1)

ATIM/ = NIA F = (KIT/MT)

0 44 using Figs. G-2214-1 &G-2214-2 Safety Factor = 1.50 (for hydrotest)

Factor M, concentration factor 1.2808 Mm = 2.256 (for inside surface axial flaw)

Temperature Adjustment p10.0(F Pressure Adjustment = 600~psig (hydrostatic pressure + Uncertainty)

Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature KIC KIp P for P-T Curve P-T Curve (OF) (OF) (ksi*inchl 2 ) (ksilinch1t 2 ) (psig) (OF) (psig) 50 0 50 0 64.13 3996 579 60 0 519 55.0 550 67.38 4213 610 65 0 550 60 0 600 7098 4452 645 70 0 585 650 65 0 7495 47.17 683 750 623 700 70 0 7934 5010 725 80 0 665 75 0 75.0 8420 5334 772 85 0 712 80.0 80.0 8956 5691 824 90 0 764 850 85 0 9549 60.86 881 95 0 821 90 0 900 10204 6523 945 100.0 885 95.0 950 10928 7006 1014 105.0 954 100.0 1000 117.28 7539 1092 1100 1,032 1050 1050 126.12 81.29 1177 1150 1,117 1100 110.0 135.90 87.80 1271 1200 1,211 1150 1150 146.70 9500 1376 1250 1,316 VYC-829 R4, Attachment 1, Page 29 of 35

Table 14: P-T Evaluation - Bottom Head Level A/B (Cooldown)

Pressure-Temperature Curve Calculation (Core Not CriticaifCooldown = Curve B)

Inputs: Plant = Ya-nke-e Component = Bot Head Vessel thickness, t = 5.9375 inches, so 41t = 2.437 4inch Vessel Radius, R = 103 .1875 inches ARTNOT = 30.0 o Cooldown Rate, CR = 100 *F/hr KIT 10.49 iksi*inchl/2 (for cooldown rate above)

MT = N/A '(From App G. Fig G-2214-1)

ATIMI = NIAA F= (KIT/MT)

0.44 using Figs G-2214-1 & G-2214-2 Safety Factor = 2.-00 (for level A/1)

Factor = 1.80'M, concentration factor Mm = 2.256 d (for inside surface axial flaw)

Temperature Adjustment = 10.0 F Height of Water for a Full Vessel N/A inches Pressure Adjustment = - 6 .0 'jpsig (hydrostatic pressure + uncertainty)

Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature Kic Kip P for P-T Curve P-T Curve (OF) (OF) (ksl*inchlr 2) (ksi*inchfl2) (psig) (OF) (psig) 50.0 50.0 6413 2682 388 60 0 328 55 0 55.0 67.38 2845 412 65 0 352 60 0 60.0 70.98 3025 438 700 378 65 0 65.0 7495 3223 467 750 407 700 70 0 79.34 3443 499 800 439 75.0 75.0 84.20 3686 534 85 0 474 80 0 800 8956 3954 573 90 0 513 85.0 850 9549 42.50 615 95 0 555 90 0 900 10204 45.78 663 100.0 603 95.0 950 10928 4940 715 105.0 655 1000 100.0 117.28 53.40 773 110.0 713 105 0 1050 126.12 57.82 837 1150 777 1100 1100 135.90 62.71 908 1200 848 1150 115.0 146.70 6811 986 125.0 926 1200 120.0 158.63 7407 1073 1300 1,013 1250 125.0 171.83 8067 1168 135.0 1,108 1300 130.0 18640 87.96 1274 1400 1,214 1350 135.0 200 00 9476 1372 145 0 1,312 VYC-829 R4, Attachment 1, Page 30 of 35

Table 15 Equivalent Margin Upper Shelf Energy Summary RG1.99 Ratio of NEDO-32205 App B Capsule Measured Predicted Measured to Worksheet Surveillance Cu Fluence Decrease Decrease Predicted Info % n/cmA2 % % F1, Factor (Ref. Charpy (Ref. 9) (Ref. 1, 22) curves)

Surveillance Plate USE 0.11% 4.50E+16 80% 55% 1.447 Surveillance Weld USE 0.03% 4.50E+16 4.80% 4.78% 1.005 RG1.99 Adjusted EOL 1/4*T Predicted Decrease= NEDO-32205 NEDO-32205 App B Cu Fluence Decrease Pred

F1 Limit Worksheet Beltline Info % n/cmA2 % % %

(Table 2-2) (Table 2-1)

Limiting Plate USE 0 14% 2.20E+17 9.4% 13.5% 21%

Limiting Weld USE 0 04% 2 20E+17 7.3% 7.4% 34%

VYC-829 R4, Attachment 1, Page 31 of 35

Table 16-1 Stress Intensity Value Summary Pressure Test Condition Temperature KIT RPV Component Load Condition Location (deg F) (ksi*sqrt*(inch))

Bottom Head CD 40 F/HR CD lI4T note 1 4.19 Bottom Head HU 40 F/HR HU 3/4 T note 2 3.31 FW Blend HU-CD Injection Transient 1/8 T (Tfluid + 50F)/ see Table 16-2 FWBore HU-CD Injection Transient 1/8 T (Tfluid + 50F)! see Table 16-3 N2 Recirc Nozzle CI 40 F/HR CD 1/4T note 1 10.03 Temperature Krr + 1.5 x KIPL RPV Component Load Condition Location (deg F) (ksi*sqrt*ý(inch))

Upper Flange 1 CD 40 F/HR CD plus Bolt Preload 3/4T note 1 50.25 Upper Flange I HU 40 F/HR HU plus Bolt Preload 3/4T note 2 50.91 Upper Flange 2 CD 40 F/HR CD plus Bolt Preload 3/4T note 1 51.56 Upper Flange 2 HU 40 F/HR HU plus Bolt Preload 3/4T note 2 70.62 Normal Operation Condition Temperature KIT RPV Component Load Condition Location (deg F) (ksi*sqrt*(inch))

Bottom Head CD 100 F/HR CD lI4T note 1 10.49 Bottom Head HU 100 F/HR HU 3/4 T note 2 8.28 FW Blend HU-CD Injection Transient 1/8 T (Tfluid + 50)/2 see Table 16-2 FWBore HU-CD Injection Transient 1/8 T (Tfluid + 50)/2 see Table 16-3 N2 Recirc Nozzle CI 100 F/HR CD 1/4T note 1 25.07 Temperature Krr + 2 x K pL RPV Component Load Condition Location (deg F) (ksi*sqrt*(inch))

Upper Flange 1 CD 100 F/HR CD plus Bolt Preloac 3/4T note 1 67.91 Upper Flange I HU 100 F/HR HU plus Bolt Preloa 3/4T note 2 67.88 Upper Flange 2 CD 100 F/HR CD plus Bolt Preloaý 3/4T note 1 69.51 Upper Flange 2 HU 100 F/HR HU plus Bolt Preloaý 3/4T note 2 96.58 Note 1 For cooldown transients, temperature lag of metal verses fluid conservatively ignored.

Note 2 For these components both inside fluid temperature and outside skin temperature are monitored. The minimum temperature is used for monitoring PT limits. Therefore HU lag does not need to be used.

VYC-829 R4, Attachment 1, Page 32 of 35

Table 16-2 Stress Intensity Value Feedwater Nozzle Blend TemperatureandKIT Values (FWI njection (Blend)- CornerNozzle Crack)

Inputs: Plant =

Component = FV Nozzle Blend F

ARTNW = 40.0 Anlaysis Basis 502 - F Step KIT for 552F - 5OF Step= 106.56 ksitinch" Kip for 1025 psig = 33.80 Jksi-inch/2 Fluid Temperature 1/8t T Temperature Kic Kit (OF) (OF) (ksi*inch ) (ksi*inchi) 50 50.0 58.52 000 55 52.5 59.82 1.06 60 55.0 61.19 2.12 65 57.5 62.62 3.18 70 60.0 64.13 425 75 62.5 65.72 5.31 80 65.0 67.38 6 37 85 67.5 69.14 7.43 90 70.0 70.98 8.49 95 72.5 72 92 9.55 100 750 74.95 10.61 105 77.5 77.09 11.67 110 80.0 79.34 12.74 115 82.5 81.71 1380 120 85.0 8420 1486 125 87.5 8681 1592 130 90.0 89.56 1698 135 92.5 92.45 18.04 140 95.0 95.49 19.10 145 97.5 98.68 20.17 150 100.0 102.04 21.23 155 102.5 105.57 22.29 160 1050 109.28 23.35 VYC-829 R4, Attachment 1, Page 33 of 35

Table 16-3 Stress Intensity Value Feedwater Nozzle Bore Temperature andKIT Values (FW Injection (Bore)- CornerNozzle Crack)

Inputs.. Plant = e Component = FWV Nozzle, Bore ?

ARTN- 40.0 " OF Analysis Basis 502 F Step KIT for 552F - 50F Step= 13339 , ksi*inch"I2 K1p for 1025 psig =28.36 ksi*inchh" Fluid Temperature l1/8t T Temperature Kic Kit 2

(OF) (OF) (ksi*inch" ) (ksi*inch"Z) 50 50.0 58 52 0.00 55 52.5 59.82 1.33 60 55.0 61.19 266 65 57.5 62.62 3 99 70 60.0 64 13 5.31 75 62.5 65.72 6 64 80 65 0 67.38 7.97 85 67.5 69 14 9.30 90 70.0 7098 1063 95 72.5 72.92 11.96 100 75.0 74.95 13.29 105 77.5 77.09 14.61 110 80.0 79.34 15.94 115 82.5 81.71 17.27 120 850 84.20 1860 125 87.5 8681 19.93 130 90.0 8956 21.26 135 925 9245 22.59 140 95.0 95.49 2391 145 97.5 98.68 25.24 150 100.0 102.04 2657 155 102.5 105.57 27.90 160 105.0 10928 29.23 VYC-829 R4, Attachment 1, Page 34 of 35

Table 17 Bounding Flange Case with No Preload Pressure-Temperature Curve Calculation (Core Not Cntica/ - Bounding Flange Case no Preload)

Inp~uts: Plant =

Component = Upper Flange 2 Upper Flange/Hub Intersection Axial Flaw Vessel thickness, t = N/A . inches Vessel Radius, R = NA ~ inchies ARTNDT = 10.0 "OF EZAI s All==

Krr+ 2 x Kl "5.18 ksi'inch' (Note Factor of 2isSafety Factor)

Safety Factor = 2.00 ,(for level AB) K, ksii nchl' 2 K1p for 1000 psig = 31.21 ksi*inch KpL.=O 0*Preload = 0 Temperature Adjustment = 10.0 I*F Kr=Thermal=L 5.18 Pressure Adjustment =

35."0 psig (hydrostatic pressure + uncertainty)

Fluid Calculated Adjusted Adjusted Temperature 1l4t Pressure Temperature Pressure for T Temperature KIC Kip P for P-T Curve P-T Curve 2 12 (°F) (psig)

(LF) 1*F) (ksi*lnch' ) (ksi*Inchl ) (pslg)

-15 -15.0 4578 20.30 650 -5 615

-10 -100 47 10 20.96 672 0 637

-5 -5 0 4856 21.69 695 5 660 0 00 5018 2250 721 10 686 5 50 51.96 2339 749 15 714 10 100 5393 2438 781 20 746 15 150 5611 2547 816 25 781 20 20.0 58.52 2667 855 30 820 25 250 61.19 2800 897 35 862 30 30.0 6413 2948 944 40 909 35 35 0 67.38 31.10 997 45 962 40 400 7098 32.90 1054 50 1019 45 450 74.95 3489 1118 55 1083 50 500 79.34 37.08 1188 60 1153 55 55 0 84.20 3951 1266 65 1231 60 60 0 89.56 4219 1352 70 1317 65 65 0 95.49 4515 1447 75 1412 66 66 0 96.75 4578 1467 76 1432 67 67.0 98.03 4643 1488 77 1453 68 68.0 9934 47.08 1508 78 1473 69 69 0 100.68 47.75 1530 79 1495 70 70.0 10204 4843 1552 80 1517 VYC-829 R4, Attachment 1, Page 35 of 35

Docket No. 50-271 BVY 03-29 Attachment 3 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 258 RPV Fracture Toughness and Material Surveillance Requirements Determination of No Significant Hazards Consideration

BVY 03-29 / Attachment 3 / Page 1 Description of amendment request:

The Proposed Change revises the reactor pressure vessel material surveillance program as currently specified in Technical Specifications Surveillance Requirement 4.6.A.1 and the reactor coolant system Pressure-Temperature limit curves (Technical Specifications Figures 3.6.1, 3.6.2 and 3.6.3). In addition, conforming changes are also being made to the associated Technical Specification Bases and the Updated Final Safety Analysis Report. The Proposed Change incorporates contemporary methodologies and industry programs for establishing material surveillance and fracture toughness requirements that have been previously found to be acceptable to the NRC staff. The two primary components to the Proposed Change are described in the accompanying safety assessment and meet the following regulatory bases:

First, Vermont Yankee (VY) is proposing to revise the licensing basis for the Vermont Yankee Nuclear Power Station by replacing the plant-specific reactor pressure vessel (RPV) material surveillance program with the Boiling Water Reactor Vessel Internals Project (BWRVIP) Integrated Surveillance Program (ISP), which has been approved by the NRC staff as meeting the requirements of paragraph III.C of Appendix H to 10 CFR 50 for an integrated surveillance program.

Second, VY is proposing to revise the P-T limit curves for the reactor coolant system in accordance with the requirements of Appendix G to 10CFR50 and an NRC-granted allowance to use the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit CurvesSection XI, Division 1."

There are no plant modifications associated with these changes.

Basis for No Significant Hazards Determination:

Pursuant to I OCFR50.92, Vermont Yankee has reviewed the proposed change and concludes that the change does not involve a significant hazards consideration since the proposed change satisfies the criteria in 10CFR50.92(c). These criteria require that the operation of the facility in accordance with the proposed amendment will not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The discussion below addresses each of these criteria and demonstrates that the proposed amendment does not constitute a significant hazard.

The proposed change does not involve a significant hazards consideration because the changes would not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change implements an integrated surveillance program that has been previously evaluated and accepted by the NRC staff as meeting the requirements of paragraph III.C of Appendix H to IOCFR50. In addition, the proposed change revises P-T limits in accordance with Appendix G to IOCFR50 (as modified by use of an accepted ASME Code Case). Brittle fracture of the reactor pressure vessel is not a postulated or evaluated design basis accident. No evaluations of other postulated accidents are affected by this proposed change. Because the

BVY 03-29 / Attachment 3 / Page 2 applicable regulatory requirements continue to be met, the change does not significantly increase the probability of any accident previously evaluated. The proposed change provides the same assurance of RPV integrity as previously provided.

The change will require that the reactor pressure vessel and interfacing coolant system continue to be operated within their design, operational or testing limits. Also, the change will not alter any assumptions previously made in evaluating the radiological consequences of accidents.

Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2) Create the possibility for a new or different kind of accident from any previously evaluated.

The proposed change does not involve a modification of the design of plant structures, systems, or components. The change will not impact the manner in which the plant is operated and will not degrade the reliability of structures, systems, or components important to safety as equipment protection features will not be deleted or modified, equipment redundancy or independence will not be reduced, supporting system performance will not be affected, and no severe testing of equipment will be imposed. No new failure modes or mechanisms will be introduced as a result of this proposed change.

Therefore, the changes to the material surveillance program and pressure-temperature limits that compose this proposed change do not create the possibility of a new or different kind of accident than those previously evaluated.

3) Involve a significant reduction in a margin of safety.

The proposed implementation of the BWRVIP ISP has been previously evaluated generically by the NRC staff and was found to provide an acceptable alternative to plant-specific RPV material surveillance programs. The NRC staff also found that the ISP met the requirements of Appendix H to IOCFR50 for an integrated RPV material surveillance program.

Appendix G to 10CFR50 describes the conditions that require pressure-temperature (P-T) limits and provides the general bases for these limits. Operating limits based on the criteria of Appendix G, as defined by applicable regulations, codes, and standards, provide reasonable assurance that non-ductile or rapidly propagating failure will not occur. The P-T limits are not derived from design basis accident analyses (DBA); but, are prescribed for all plant modes to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause non-ductile failure of the reactor coolant pressure boundary. Calculation of P-T limits in accordance with the criteria of Appendix G to 10CFR50 and applicable regulatory requirements ensures that adequate margins of safety are maintained and there is no significant reduction in a margin of safety.

The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. There is no change or impact on any safety analysis assumption or in any other parameter affecting the course of an accident analysis supporting the Bases of any Technical Specification. The proposed change does not involve any increase in calculated off-site dose consequences. Since the proposed change for RPV material surveillance is in accordance with the NRC staff's safety evaluation for the ISP, and P-T curves were revised in accordance with the requirements of Appendix G to IOCFR50 (as modified by

BVY 03-29 / Attachment 3 / Page 3 use of ASME Code Case N-640), adequate safety margins are maintained without any significant reduction.

Conclusion On the basis of the above, VY has determined that operation of the facility in accordance with the proposed change does not involve a significant hazards consideration as defined in 10CFR50.92(c), in that it: (1) does not involve a significant increase in the probability or consequences of an accident previously evaluated; (2) does not create the possibility of a new or different kind of accident from any accident previously evaluated; and (3) does not involve a significant reduction in a margin of safety.

Docket No. 50-271 BVY 03-29 Attachment 4 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 258 RPV Fracture Toughness and Material Surveillance Requirements Revised Updated Final Safety Analysis Report

BVY 03-29 / Attachment 4 / Page 1 PROPOSED CHANGE 258 - PROPOSED UFSAR MARK-UP

1. VYNPS UFSAR, Affected Page List Current UFSAR Section 4.2.6 (pages 4.2-14 and 4.2-21)

2. Marked-up Pages See attached mark-up of UFSAR pages 4.2.14 and 4.2-21 (Table 4.2.4).

Note: Deleted text is shown by strike-through. Added text is shown by underline.

fabrication and quality control organizations and a system capable of assuring and documenting the required quality level.

The qualifications are backed up with Rotterdam's extensive experience in core structure fabrication with such United States plants as TVA I, II, and III, Peach Bottom II and III, Monticello, and Vermont Yankee. Also, Rotterdam fabricated parts of Quad Cities II reactor pressure vessels, as well as complete vessels for foreign plants, such as AKM and Nuclenor.

The Reactor Coolant System was cleaned and flushed before fuel was loaded initially. During the preoperational test program, the reactor vessel and Reactor Coolant System were given a hydrostatic test in accordance with code requirements at 125% of design pressure. The vessel temperature is maintained at a minimum of 60OF above the NDT temperature prior to pressurizing the vessel for hydrostatic test. A system leakage test at a pressure not to exceed system operating pressure is made following each removal and replacement of the reactor vessel head. Other preoperational tests include calibrating and testing the reactor vessel flange seal-ring leakage detection instrumentation, adjusting reactor vessel stabilizers, checking all vessel thermocouples, and checking the operation of the vessel flange stud tensioner.

The reactor vessel temperatures are monitored during vessel heatup and cooldown to assure that thermal stress on the reactor vessel is not excessive during startup and shutdown.

4.2.6 Inspection and Testing The plant has been designed to prevent occurrence of a gross defect. The inservice inspection program has been designed to provide for the inspection during service of those components and systems whose structural integrity must be maintained for continued safe operation of the plant. The selection of components and inspection locations is based on the ASME Code,Section XI, and 10CFR5O.55(a) . The program is presented in Reference 2.

Vermont Yankee is a participant in the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) for the purpose of monitoring changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region due to exposure of these materials to neutron irradiation. The Nuclear Regulatory Commission staff has determined that the BWRVIP ISP is an acceptable alternative to plant-specific material surveillance programs for the purpose of maintaining compliance with the requirements of Appendix H to 10CFR50, "Reactor Vessel Material Surveillance Program Requirements." Under the ISP, dosimetry data and the results of fracture toughness tests from surveillance capsules in host BWRs are shared with comparable BWRs. As required by Appendix H to 10CFR5O, VY will evaluate changes in the properties of representative materials for the purpose of determining whether changes are necessary in pressure and temperature limits and operating procedures. The report, "BWRVIP-86-A: BWR Vessel and Internals Project Updated BWR Integrated Surveillance Program VYNPS UFSAR Revision 4-ý [xx]

4.2-14 of 21

(ISP) Implementation Plan, establishes the regulatory basis for the surveillance program.

The Vermont Yankee Nuclear Power Station is not a host ISP plant for providing surveillance capsules; however, the remaining two VYNPS material surveillance capsules will continue to reside in the reactor in case they are needed in the future as a contingency. The VYNPS surveillance capsules Surveillance Test P..ogra consist of tensile and Charpy V-Notch specimens representative of the three areas of interest: reactor vessel base metal, weld Heat-Affected Zone (HAZ) metal, and weld metal from a reactor steel joint which simulates a welded joint in the reactor vessel. The specimens were placed in three separate surveillance a.r. ontained in capsules placed at three locatiQnR in the reactor e..... Iradially located adjacent to the inner vessel wall, radially adjacent to the at core mid-plane, where the neutron flux will be is highest. The specimen types contained in the capsules are listed in Table 4.2.4. In addition to the specimens listed in Table 4.2.4, sufficient specimens are provided for obtaining unirradiated base line data and for retention as archive material.

VY's neutron fluence calculations (and future re-evaluations) that support reactor coolant system pressure-temperature limits and the ISP are based on a fluence methodology that is acceptable to the NRC staff, consistent with the guidance in NRC Regulatory Guide 1.190, "Calculational Methods for Determining Pressure Vessel Neutron Fluence."

VYNPS UFSAR Revision 4-- [xx]

4.2-[xx] of [xx]

TABLE 4.2.4 SURVEILLANCE CAPSULE REMOVAL SCHEDULE Capsule Specimen Number of Specimens Vessel Withdrawal INo.tin Type (1) Azimuth Schedule (2)

No. Location 0:0t. 1)

Base Weld HAZ 1 CjL 12 12 12 300 10 years (3)

T-* 2 2 2 3tyearn s 2 C 8 8 8 1200 Standby T 2 2 2 3 C 8 8 8 3000 Standby T 2 2 2 Notes:

(1) C = standard Charpy V-Notch impact specimen T = tensile specimen (2) Specified capsules will be withdrawn during the refueling outage following the year specified, referenced to the date of commercial operation.

(3) Capsule No. 1 was removed from the vessel for analysis in March 1983.

SSpecif ied Gapsules il beaitdrw during refueling outage following the year -pecified, refrenc.ed to the date of operation.

.om.merial

-standard Charpy V-Notch impact .pecimen

T - tensile specimen VYNPS UFSAR Revision 17 [xx]

4.2-[xx] of [xx]

Docket No. 50-271 BVY 03-29 Attachment 5 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 258 RPV Fracture Toughness and Material Surveillance Requirements Marked-up Version of the Current Technical Specifications

BVY 03-29 / Attachment 5 / Page 1 Description of Technical Specification Changqes

1. Delete TS SR 4.6.A.5 on current page 116 in its entirety.

2. Modify TS Figures 3.6.1, 3.6.2 and 3.6.3 (current pages 135-137) as follows:

"* The validity of each figure is changed from the "end of cycle 23" to "4.46 E8 MWH(t)."

"* For each figure, the grid line divisions are changed, additional 100 psi increments are added to the ordinate axis, and more data are used to plot the curves.

"* A Note is added to Figure 3.6.2 for the use of test instrumentation during tensioning and detensioning operations with the vessel vented and fluid level below the flange region.

"* Corrections are made to the tabulation of pressure and temperature values in Figure 3.6.3.

3. Replace the last sentence of the 4t paragraph on current page 138 - Bases to 3.6.A and 4.6.A - with the following:

Based upon plate and weld chemistry, initialRTNDT values, predictedpeak fast neutron fluence (2.99 x 1017 n/cm2 at the reactorvessel inside surface) for a grosspower generation of 4.46 x 108 MWH(t), these core region ARTNOT values conservatively bound the guidance of Regulatory Guide 1.99, Revision 2.

4. Add amplifying clarification to the first sentence of the last paragraph on current page 139 - Bases 3.6.A and 4.6.A.

5. After the last paragraph on current page 139 - Bases 3.6.A and 4.6.A - insert the following two paragraphs:

Specification 3.6.A.3 requiresthat the temperatureof the vessel head flange and the head be greaterthan 70°F before tensioning. The 70'F is an analyticallimit and does not include instrumentationuncertainty,which must be procedurallyincluded depending upon which temperature monitoringinstrumentationis being used. The temperature values shown on Figures3.6.1, 3.6.2 and 3.6.3 include a 10°F instrumentation uncertainty.

A Note is included in Figure 3.6.2 that specifies test instrumentationuncertaintymust be

+/- 2°F and the flange region temperaturesmust be maintainedgreaterthan or equal to 72 0F when using such instrumentationin lieu of permanently installed instrumentation.

Qualifiedtest instrumentationmay only be used for the purpose of maintainingthe temperature limit when the vessel is vented and the fluid level is below the flange region.

If permanently installedinstrumentation(with a 10°F uncertainty)is used during head tensioning and detensioningoperations,the 80°F limit must be met.

BVY 03-29 / Attachment 5 / Page 2

6. Delete the first paragraph on current page 140 - Bases to 3.6.A and 4.6.A.

7. Delete the current, last paragraph of Bases 3.6.A and 4.6.A (on current page 140), and replace it with the following:

Vermont Yankee is a participantin the Boiling Water Reactor Vessel and Internals ProjectIntegratedSurveillance Program (ISP) for monitoring changes in the fracture toughness propertiesof ferritic materialsin the reactorpressure vessel (RPV) beltline region. (See UFSAR Section 4.2 for additionalISP details.) As ISP capsule test reports become available for RPV materialsrepresentative of VYNPS, the actualshift in the reference temperaturefor nil-ductility transition(RTNDT) of the vessel materialmay be re established. In accordancewith Appendix H to 10CFR50, VY is requiredto review relevant test reports and make a determination of whether or not a change in Technical Specifications is requiredas a result of the surveillance data.

VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION

5. The react vessel irradia on surveillance specim s shall be remov d and examined to det mine changes in ma erial properties n cordance with th following schedul CAPSULE RE OVAL YEAR The r sults shall be used to r assess material pr erties and update Figures 3.6.1, 3.6.2 nd

.6.3, as appropria .

/The removal times/hall

/be referenced to the

/ specified, fe**g~argenced to

\ the date o commercial operation/

B. Coolant Chemistry B. Coolant Chemistry

1. a. During reactor power 1. a. A sample of reactor operation, the coolant shall be radioiodine taken at least every concentration in the 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months and reactor coolant analyzed for shall not exceed radioactive iodines 1.1 microcuries of of 1-131 through 1-131 dose 1-135 during power equivalent per gram operation. In of water, except as addition, when steam allowed in jet air ejector Specification monitors indicate an 3.6.B.l.b. increase in radioactive gaseous effluents of 25 percent or 5000 pCi/sec, whichever is greater, during steady state reactor operation a reactor coolant sample shall be taken and analyzed for radioactive iodines.

Amendment No. @-3, I, 203 116

Cb1V:16J:

/l M~OaJAWS a4A/b /-AC 56 vYNPS P181 eC 6,+7-4 -rd(

VEX/ FIGURE 3.6.1 PIC 7T C Reactor Vessel Pressure-Temperature Li1mitlations, Hydrostatic Pressure and Leak Tests, Core Not Critical 40°F/hr Heatup/Cooldown Limit x/. 4(v M Valid Through nd yýc 3 12fl11 4 - 1111 7t 2i7 11-t VK 4'1-a Bottom Head Curve I . . . . ,I* . .

-I-fl-I.IF . .

I I

1I2TV7IT II I I

. I . . .

I I I I

,- -. Uooer Regions: I I I Use minimum of bottom I I I I j

f-L--,fluid temperature and - 1L -L Use minimum of downcomer I I I bottom head surface -i A L region fluid temperature and aI a I I I I IB I I I 'aI 1.J I I I I

-- 'P -I- - -

- temperature. flange region outside I a a I

- -I-- -I 1 401 surface temperature, except

-- r"--I-I- -I -

! I I when flange temperature is 1000 greater than 11 °F. use only I I I I ,i I I I I I I I I I the downcomer region fluid I a a I I I I -I I a I -* T-r-F

-I' temperature I I I I T r --- I I I I

- -- r--I--. 1- -- a a a I I I I -- I- -a-.- -- I---

I I I I -J4-4--1-I-

-I - -- I----.- --I-- - --

I I ai I

a I

S I I aI a

a a a I I Ia I a

I i

a a- ia-I I I

'" l I

5i I I

-I- I-ILa 800 6 . .. .+ . . . . . . l . .- * . . . I........... .... .. ... . .

I I l l l la o -1.- ---4 -a--I- I a i a I--I-4--

a' I I

= --

-- I1-LI

- -L J- J- L LI J. -L_. I II "- -i--- -- 4 - -- L I

I

_1 I I I I -- IJ- .1- i-T IIII 4'I I I I -II -I! a I

I I

-- 4-T-I I a-II

-a -

I r -

I I r-I a.

- -- I I- 1-- -

r - aII- I I I 1 i l i i l l l i f i l l a o 600 I I Ia I 1a I I I I a---a-t iIKII Temperature Bottom Head Upper Regions I I I I I I I I I I I a I (=F)f,- (psig) (psig)

-T-F 1 -- .. .

zW(00 Lu

"--a -a- -" - "t I I 80 0 0

-- '-a-" I --a- I I I I

I - - -- -* I I I I a a a I I - T- aI ai i IT 80 665 253 I I -- J--L--/--L I I I I I I I I I I 85 712 253 I I a-I I I I" I

I I

I I a a 90 764 253 S. . . . . . . . . . I . . . . . 95 821 253 S--

S I a I- -- I----I-- -J I

J. a; 100 885 253 I I a a I I I I I I 1 105 954 253 i I I

.-- I----I-i- - --I . - _ a_ a_

I I I I 110 1032 253

-T-r -r-i--

m a a i a 110 1032 842

- --- I--a - a a a a

"-r-0----4--I a a a a "I, -- 0--I-I 115 1117 885 a a a a 120 1211 932 aI aI a 125 1316 984

"-rI-I-- - --" - I r a 130 - 1042 a

I-1-4---a a a a

-- aJ.-a- -a- -a-I aIaIaI

-a-a-r-r a 135 1105 J- a I 140 - 1175 I

1-L-L..

145 1253 I -----------

I:::: I 100 120 140 160 180 200 TEMPERATURE (-F)

.r-CR.CA4drAl7-S DAI 4:01b MAi 7el L 4XI S Amendment No. -- , 2-G, 41-, .94, 4-24, 203 135

C/1/AJ*, G4/,b 414,1r C-)Q01C:b4r,4 7-0 P4or Ao0eG FIGURE VYS 3.6 5.2 Reactor Vessel Pressure-Temperrature Limitiatlons SoNormal Operation, core 4ot Critical

.1;W .0-,ot' I I 00OF/hr HeatupICee.4*

Valid Througl nd c e23

, ,I B B---'BB tB B -- h,BI- iB,,'-- - B --BIBl ' - -

. . B

I *

S SUpperRegions* B B

- ' 4 -- 4-I- 1,- -41. I--1--

BIll,

-B- + -- II l - Use minimum of -4 J ___ J I __ _L _ L J_ . -_,: -_S._ : 1 _: _- I LB_

_ It,3 _L_,_J_!_iL..

S B

_downcomerreglon B', B Ifluiddtemperature

- iBottom Head Curve: i, r r-i * - rr-r""1 - ""-"r- land flange region -i-,

J-i

-,-i- -i.. , ,iv B . ,B outsidesurface B B BIB ai I I'I -I- Use minmum of I I,_ B B .. I B I temperature, except ,

! ', bottom fuud , ,/, , ', I B BB 'IB ',B IB " flangeB 1000, B B when

- -i-i temperatureand I

' I B B ' B " B botlom head surface B B B aB B temperatureis B B I B I temperature. B B B -F- B- BI- -greaterthn140'F. I L ..... _..i... j_..._.luse only the -- J.

' B B I BI , ,B B

, BI, I B S I B B Id omer region r - -1 ... i rI 0 B I B I -I--I- .... FGIF 7i* T IT-F-I -i -1 -- T - fluid temperature. "-,-.

BB~ + -B -BB B BiBi B

B~~~ B I B I

I i IB ~I S ID B

B B

i B I

M B B I B S-- B"1 - T- - I I I I 1- T - r B- _ Bottom Upper I II I I Temperature Head Regions B o ' I B B 4 B!- LiB 1 BI J_

";r-1--T -r I

4--I-- -B-I I -(F) (psig) (psig) B

-_ L-_

S I , ,

-- I- " - L -r - I- !I Br-IB

- 80 0 0

-- "-T-f" --,- " - B'-BB! 1-T-r- -I - 80 439 253 Bi1 I BIB, I B I B B IB B 85 474 253

-J 600 w= 90 513 253 7, -1 T - i-- i-I-I-- - -%-I-fT -F I--T -F-,

(B) - 95 555 253 B I

- 100 603 253 I 1 ,B1 ~ , B, B iB Ba B iB ,

- Bi B BT-BB Il l B rL -,-B B - B I B I tl B-F j B Ba B BIBr l _ 105 655 253 B, Jl l yl .l,, , 411 , , ,I , 110 713 253 115 777 253 400 I II- --11 4,, 1 1 I I I 4I 1- - 120 848 253 B--,- -B- B -- FI-------F----B-------T--B -I-BiBi B BIB B Bi Bi i B i , - 125 926 253 BI I B BIB IBB I BI IBIBI I I B 130 1013 253 I BI BI B B B ~ iB__* _B _B _BB B B B B 135 1108 253 B I B BI I B,

, B IB IB I 140 1214 253 B I B IB B IB IB B I I B II I

- 140 1214 830 20 145 1312 889 Bi B i B BiB B BI B B I4B IB -I

-B-BI--B-a 150 - 953 B B BIBa 1024 BI B, B 155 - B "BIll T- .- i- -i *t - i -- .I-i- B-I ---

- 160 - 1103 BI B, B

-- 1>---,- -~1B-- - 165 - 1190 170 1258 D

60 80 100 120 140 160 180 200 r

-During tensioning and detensioning operations with the vessel vented and the vessel fluidd level below the flange region, the flange temperature may be- monitored with test instrumentation in lieu of process instrumentation for the downcomer region fluid temperature and permanent flange region outside surface temperature. The test instrumentation uncertainty must be less than +I- 20F. The flange region temperatures must be maintained greater than or equal to 72 0F when monitored with test instrumentation during tensioning, detensioning, and when tensioned.

Amendment No. -34, 93-, 203 136

bi'/gj'o,6j$ 44b 0565 A4 7o*6 b~-~4 rA VYN~PS FIGURE 3.6.3 Rea tor Vessel Pressure-Temperature Limitlations Normal Operation, Core Critical 100°F/hr Heatup/Cooldown Limit ""1 4-4 If Press re < 253 psig, Water Level must be withln 4 (a rmal Range for Power 0 eration alid Throug End yci 3

ý--ý U -- A --. L--J

\B=o Upp.. : 1 ::- : :

\ d*om Upper I I 1 t :

iI :

1200 Temltur Helld Regions I r I L II I I I I I I I I I I I ILI I

(psIN) (psig) "RegionI rrv 80 80 253 mr-I-I Il I I I I

"-- I Bottom 1

IUse minimum of Head I I

-I-I I

I I it

-r--I-II

---L --

II-4 V

114 *i I _-J- Lbottom fluid I I 114 402 0 i i temperature and I I I I

-iT bottom head -T 4F-115 I I I I surface 1000 116 I t II

temperature 120 439 (Z -

I I I L--1,.

I I I I I I ,-I- I.=

I l I I I

-r-I J-1

- Lr.

. i._

II II I l 125 474 253 I I I I

I I

I I I III I I 513 253 -'-I- 1- F -

-I II I I I -- I- --- 1 .. r-r-- , I I I I I 135 555 253 * -i--i 140 603 253 I I I I I I I I I I I I I I I I

"= 800.

n.

145 655 253 0 150 713 253 1 J- LJ LL j 1 -11 L L-- I -LL -: I --..

155 777 253 l UJ I II TII Ii l I I 160 848 253 iI I Ii i I I I JI HI ll i I I 165 926 253 170 1013 253 I I I I -I I II -III I -I I1.

0 175 1108 253 II I I I I t I I I I I I o600 180 1214 253 I - -II I I I- I I i I I I TI I I I I I I II I I I PII .

ID I III I I II lI I I I 1111 I I I IIL I I 180 1214 830 185 1312 889 i i i I i I iI i I I I ! I I I I i i 190 - 953 I I I irI I I I I I I I I I T I I I I 195 - 1024 200 - 1103 S.... ... _._,__:_ __

205 1190 m 400 210 - 1258 Ii I I I I I I I I I I I I I I %,l I I i I I I I I I I I I I I I LU

"--I--1-t- '-I- - A -L- -I- I I I i

-- I- t - I-1 1- - 1-t-II-Upper Regions

0. I I I I I I I I I I II Ii II Use minimum of

- I- -I 4.- J.-

I I I I I II I I I I I I I I I I II II II downcomer region I I It 1111 -L I -J_! I I I I fluid temperature I I I I I I I LIII I I!II I I I I 11,1 and flange region

-r-

"-" 1 1 r-rn--r -- T-r-I-I r--

- - T-r"*--

-r-i outside surface I I I I II I II [ I I I I I t I I I I I 200I t--.-l -ur. ex Lcep Si i I i i ;: 1 i i i I, i i,i enflange

-I- I I I- - - I - - - I- -i -I - -I- temper turei

._-_ _.L._ .L I J.. I I I - L...L 1A! [ _ L_ greaterthan180"F.

I I II Ieonlythe

, , , I ' i ' ' I I , I I i , , ,

T-r - -r i-I r - T-r -- r- - -1T -- r-i- downcomerregion SI I I I IIII III I l lI I I- I I I I~fluidtemperature.

III LIII- .. IIIIJ~~.... .L. IIII

. +/- . II L.... II L...

100 120 140 160 180 200 TEMPERATURE ('F) 6ýWb54jA 6 ' Pfz

/a Amendment No. , 4-, 203 137

VYNPS BASES:

3.6 and 4.6 REACTOR COOLANT SYSTEM A. Pressure and Temperature Limitations All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.2 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

The Pressure/Temperature (P/T) curves included as Figures 3.6.1, 3.6.2, and 3.6.3 were developed using 10CFR50 Appendix G, 1995 ASME Code,Section XI, Appendix G (including the Summer 1996 Addenda), and ASME Code Case N-640. These three curves provide P/T limit requirements for Pressure Test, Core Not Critical, and Core Critical. The P/T curves are not derived from Design Basis Accident analysis. They are prescribed to avoid encountering pressure, temperature or temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the reactor pressure boundary, a condition that is unanalyzed.

During heating events, the thermal gradients in the reactor vessel wall produce thermal stresses that vary from compressive at the inner wall to tensile at the outer wall. During cooling events the thermal stresses vary from tensile at the inner wall to compressive at the outer wall. The thermally induced tensile stresses are additive to the preSsure induced tensile stresses. In the flange region, bolt-preload has a significant affect on stress in the flange and adjacent plates.

Therefore heating/cooling events and bolt preload are used in the determination of the pressure-temperature limitations for the vessel.

The guidance of Branch Technical Position - MTEB 5-2, material drop weight, and Charpy impact test results were used to determine a reference nil-ductility temperature (RTND) for all pressure boundary components. For the plates and welds adjacent to the core, fast neutron (E > 1 Mev) irradiation will cause an increase in the RTND.

For these plates and welds an adjusted RTND (ARTNDT) of 89 0 F and 730 F (44 and - thickness locations) was conservatively used in development of these curves for core region com onents. sed upon plate a weld 2

chemistry, initialKT vaWues, predicted eak fluence (2.3 01' n/cm )

8 for a gross powe generation of 4.46xI0 MWH(t) (Battelle olumbus Laboratory Rep t BCL 585-84-3, dated ay 15, 1984) the core region ARTN* values onservatively bound t guidance of Reg atory Guide 1.99, Revi*on 2.

There were five regions of the reactor pressure vessel (RPV) that were evaluated in the development of the P/T Limit curves: (1) the reactor

,4 vessel beltline region, (2) the bottom head region, (3) the feedwater

' nozzle, (4) the recirculation inlet nozzle, and (5) the upper vessel flange region. These regions will bound all other regions in the vessel with respect to considerations for brittle fracture.

Amendment No. , 621, 48-, 3, 94, 4-20, 4-4-6, 203 138

VYNPS BASES: 3.6 and 4.6 (Cont'd)

Two lines are shown on each P/T limit figure. The dashed line is the Bottom Head Curve. This is applicable to the bottom head area only and includes the bottom head knuckle plates and dollar plates. Based on bottom head fluid temperature and bottom head surface temperature, the reactor pressure shall be maintained below the dashed line at all times.

Due to convection cooling, stratification, and cool CRD flow, the bottom head area is subject to lower temperatures than the balance of the pressure vessel. The RTNK of the lower head is lower than the ARTN* used for the beltline. The lower head area is also not subject to the same high level of stress as the flange and feedwater nozzle regions. The dashed Bottom Head Curve is less restrictive than the enveloping curve used for the upper regions of the vessel and provides Operator's with a conservative, but less restrictive P/T limit for the cooler bottom head region.

The solid line is the Upper Region Curve. This line conservatively bounds all regions of the vessel including the most limiting beltline and flange areas. At temperatures below the 10CFR50 Appendix G minimum temperature requirement (vertical line) based on the downcomer temperature and flange temperature, the reactor pressure shall be maintained below the solid line. At temperatures in excess of the 10CFR50 Appendix G minimum temperature requirement, the allowable pressure based on the flange is much higher than the beltline limit.

Therefore, when the flange temperature exceeds the 10CFR50 Appendix G minimum temperature requirement, the reactor pressure shall be maintained below the solid line based on downcomer temperature.

The Pressure Test carve (3.6.1) is applicable for heatup/cooldown rates up to 40*F/hr. The Core Not Critical-curve (3.6.2) and the Core Critical curve (3.6.3) are applicable for heatup/cooldown rates up to 100'F/hr. In addition to heatup and cooldown events, the more limiting anticipated operational occurrences (AOOs) were evaluated (Structural Integrity Report, SIR-00-155). For the feedwater nozzles, a sudden injection of 50°F cold water into the nozzle was postulated in the development of all three curves. The bottom head region was independently evaluated for AOOs in addition to 40°F/hr and 100'F/hr heatup/cooldown rates. This evaluation demonstrated that P/T requirements of the bottom head would be maintained for transients that would bound rapid cooling as well as step increases in temperature.

The rapid cooling event would bound scrams and other upset condition (level B) cold water injection events. The bottom head was also evaluated for a series of step heatup transients. This would depict hot sweep transients typically associated with reinitiation of recirculation flow with stratified conditions in the lower plenum.

This demonstrated that there was significant margin to P/T limits with GE SIL 251 recommendations for reinitiating recirculation flow in stratified conditions. V 1 YJ Adjustments for tempe atureýnd pressure ins rument uncertainty have been included in the lcurves. The minimum temperature requirements were all increased by 10*F to compensate for temperature loop uncertainty error. The maximum pressure values were all decreased by 30psi to account for pressure loop uncertainty error. In addition, the maximum pressure was reduced further to account for static elevation head assuming the level was at the top of the reactor and at 70*F.

S 6-4 7->

Amendment No. 203 139

VYNPS BASES: 3.6 and 4.6 (Cont'd)

The actual shift in RTNDTAf the critical plate a weld material in e core region will be es blished periodically d ing operation by removing and evaluat g, in accordance with M E185, reactor ve sel material irradiatio surveillance specimens nstalled near the side wall of the react vessel in the core ar . Since the neutro spectra at the irradiat n samples and vessel in ide radius are esse ially identical, th measured transition shi for a sample can b applied with confid ce to the adjacent sect'n of the reactor ve el.

Battelle C umbus Laboratory Report CL-585-84-3, dated ay 15, 1984, provides his information for the en-year surveillanc capsule. When data f m the next surveillance apsule is available, he predicted belt ine ARTNDT will be re-asse ed and the P/T curves revised as app priate.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures will be maintained within 50OF of each other prior to startup of an idle loop.

, h umber o /reactor vessel irradiation surveillance scimens and the*

J. frequenci for removing and tes ng these specimens Ee provided to

/* assure *mpliance with the req qrements of Appendi~i to 10CFR Part 50.1 B. Coolant Chemistry A steady-state radioiodine concentration limit of 1.1 [tCi of 1-131 dose equivalent per gram of water in the Reactor-Coolant System can be reached if the gross . adioactivity in the gaseous effluents is near the limit, as set forth in the Offsite Dose Calculation Manual, or if there is a failure or prolonged shutdown of the cleanup demineralizer. In the event of a steam line rupture outside the drywell, the NRC staff ca-lctrlations show the resultant-radiological dose at the site boundary to be less than 30 Rem to the thyroid. This dose was calculated on the basis of the radioiodine concentration limit of 1.1 IiCi of 1-131 dose equivalent per gram of water, atmospheric diffusion from an equivalent elevated release of 10 meters at the nearest site boundary (190 m) for a X/Q = 3.9 x 10-3 sec/mr3 (Pasquill D and 0.33 m/sec equivalent), and a steam line isolation valve closure time of five seconds with a steam/water mass release of 30,000 pounds.

The iodine spike limit of four (4) microcuries of 1-131 dose equivalent per gram of water provides an iodine peak or spike limit for the reactor coolant concentration to assure that the radiological consequences of a postulated LOCA are within 10CFR Part 100 dose guidelines.

The reactor coolant sample will be used to assure that the limit of Specification 3.6.B.1 is not exceeded. The radioiodine concentration would not be expected to change rapidly during steady-state operation over a period of 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months . In addition, the trend of the radioactive gaseous effluents, which is continuously monitored, is a good indicator of the trend of the radioiodine concentration in the reactor coolant.

When a significant increase in radioactive gaseous effluents is indicated, as specified, an additional reactor coolant sample shall be taken and analyzed for radioactive iodine.

Amendment No. -3, , 91, 43, 4-64, 4-9-3, 203 140

Docket No. 50-271 BVY 03-29 Attachment 6 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 258 RPV Fracture Toughness and Material Surveillance Requirements Retyped Technical Specification Pages

BVY 03-29 / Attachment 6 / Page 1 Listing of Affected Technical Specifications Pages Replace the Vermont Yankee Nuclear Power Station Technical Specifications pages listed below with the revised pages included herein. The revised pages contain vertical lines in the margin indicating the areas of change.

Remove Insert 116 116 135 135 136 136 137 137 138 138 139 139 140 140

VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION B. Coolant Chemistry B. Coolant Chemistry

1. a. During reactor power 1. a. A sample of reactor operation, the coolant shall be radioiodine taken at least every concentration in the 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months and reactor coolant analyzed for shall not exceed radioactive iodines 1.1 microcuries of of 1-131 through 1-131 dose 1-135 during power equivalent per gram operation. In of water, except as addition, when steam allowed in jet air ejector Specification monitors indicate an 3.6.B.l.b. increase in radioactive gaseous effluents of 25 percent or 5000 pCi/sec, whichever is greater, during steady state reactor operation a reactor coolant sample shall be taken and analyzed for radioactive iodines.

Amendment No. 3-3, 91, 241 116

VYNPS Figure 3.6.1 Reactor Vessel Pressure-Temperature Limitations Hydrostatic Pressure and Leak Tests, Core Not Critical 40Flhr HeatuplCooldown Limit Valid Through 4.46E8 MWH(t)

S* ._..-- *....

.... ,..---.r-- ..,,...-...--.....-... .. .... ....* ._...* .

1200 I I

....... Bottom Head Curve Use minimum of bottom ---.. .F -*-, /,.. --

Use minimum of downoomer

,--- fluid temperature and .. .1...- , .

._*. . *.-I . .

. . .-4 Sregion fluid temperature and 1100 bottom head surface -.... flange region outside -..

- -- temperature . .. - ----. surface -.temperature, except ........ ;

I:. :*: ----. when flange temperature is 1000 S. . . . . ; f l greater than 11 0F, use only Is 900 I I,....... . / the downcomer region fluid ---

temperature.

S....-----:---------

SI

.- -i - -; -

,- I1. S.....- :--i.. . i.. . .

S.... ;....,

S.... =*...*

800 0

O.

I.

M

- -------- '. ..---- . . . . --- ----i... ...

. .,.. ., .. . . ÷ . ., .. ,.. .= _

-J W .----.. _.--.-. ..... *........

700 .. ...." ....' ... *... ;. - ----------.. .. . .. , .... ------- ' ._....* . *.... ,. . ...." .... .. . . .. . .. . .

0 Co - ----- .. ..---.... . . .. . --- . -- . -- .----------

- -- ------- -i- - --- Bo fo -- --------- U pe "*

I B to p e Rfnf I ... :..."..... .....

--- '-.... ..... ..... -- i-- Temperature Head Regions

.....-i..'-----'--

.. -. .. ..-*-- i- -(P ) (psig) (psig)

.. .... .... . ... .... .. . ... .. . ...,.. . . .. i -- :- -8 0 0 0 Co RInn 80 665 253

---.. :-----,------ ----- ,---" .... ...--:----'--- g 6 5

"..... - . . - - - -- -- - - -- 85 712 253 0,. 90 764 253

,aOO 95 821 253

'a 100 885 253

..- -....- .... .......... 105 954 253 300 _____ --- -110 1032 253

.... ....-----,----""....--- ---- 110 1032 842

.. .. -- -- -- 115 11178 5 200 ---~"~ ..~ --~*120 ~ ~ ~ .......

..~ ".....;.. ~~~-- ....

i------13-104 1211 932

--- ---- ... --...--- -- ----- .. 125 1316 984

..... 3 P . 130 - 1042 100

- -140 135 -

-1175 1105

-- ".. -- S.

. .-- . ".m----I..

.... . "-... ---- . ;- . 145 - 1253 0

60 80 100 120 140 160 180 200 TEMPERATURE (°F)

Amendment No. 3, 62, 8--1, --3, 4-2-, 243 135

VYNPS FIGURE 3.6.2 Reactor Vessel Pressure-Temperature Limitations Normal Operation, Core Not Critical Upper Regions 100°F/hr Heatup/Cooldown Limit Use minimum of Valid Through 4.46E8 MWH(t) downcomer region fluid temperature and flange region outside surface 1200 "_--_-- .... --- ...I- -....----- Ft +--... / - temperature, except when flange temperature is

...----------... greater than 140°F, use

....... only the downcomer region fluid temperature 1111Il 11111A 1 .------- Bottom

------- Head Curve

. *

J

============ ........ ...

1000

-+/----temperature Use minimum of

'"---'"Ibottom fluid and

[- /... "'!

Te"p Bottorn U~ppeir Rgons

._.J

-.......-------- bottom head surface

temperature r/::::: j:/

-- 4

-L...

"'i 80 9F) (psial 0

(WaQ) 0 82 439 253 85 474 253 0

I..-.. . 4 _.---. .........--..--

.- 90 513 253 95 555 253 800-J... _ _ _ _ _ _ _ _

100 6M3 253 0 . ........

...... --- ------ 105 665 253 I

7nn -_ . :::_ _ _ 110 713 253 115 777 253 120 848 253 125 Sm 253 U.

0 600 130 1013 253 135 1105 253

---4----- . .. ... .. --- r . -- -+ - ... ... - - - ---

-.. r -. . .4-- -" " . . . .

140 1214 253 (I

500 --4-----

4 --- ------ -. ' " '"-..... -- - --- ----- '. . . . 140 1214 830 0.

.. ..- -- --- --...- ... 145 1312 889

---i---' ---- .. . . . . ., - , - , . , . . - - - - ---4--------- -------- ----- ...--... 953 150 -

155 - 1024 400 160 1103

............... 165 - 1190

................ 170 - 1258 300

________ - .d SL . . . . . . . ..L . . .

Dunng tensioning and detensioning operations with the vessel vented and the vessel 200 fluid level below the flange region, the flange temperature may be monitored with test

instrumentation in lieu of process instrumentation for the downcomer region fluid

temperature and permanent flange region outside surface temperature The test nstrumentation uncertainty must be less than +I- 2°F. The flange region temperatures i--- .......

100- .. . 4------....- must be maintained greater than or equal to 72 °F when monitored with test

... instrumentation during tensioning, detensioning, and when tensioned 0

--- - ---- ----------- I ---,--- *---- - - - - - -- ----., ----- - ---- - - _

60 80 100 120 140 160 180 200 TEMPERATURE (F)

Amendment No. 3, -3, 2-0-3 136

VYNPS FIGURE 3.6.3 Reactor Vessel Pressure-Temperature Limitations Normal Operation, Core Critical 100°lhr HeatuplCooldown Limit If Pressure < 253 psig, Water Level must be within Normal Range for Power Operation Valid Through 4.46E8 MWH(t)

1'"

" -4i "4... .. ...

. ..1 . 1 "* 1*"I .. ... : ' ' " l..

',.'" . .. ' ',--"--- -- ----.. :/

Bottom A1 1200 . . . . .. * +

f Temp Head Regions ; I ; ; ;

(°F) (psig) (psig) Bottom Head

-- .... 1--l ........

...-. . Recion 80 0 S.... .... Use minimum of ----

S.... i

[ ......

1100 80 114 253 253 bottom fluid temperature and ..i-- -i-------

114 402

- -- --.. S........ bottom head surface ---

1000 115 407 ......J....

116 413 0 S.......... temperature ...S.... ,*.. . . .... ... .

120 439 253 S....,..... ...S.... , . .. . . ----- -------

ci g00 125 130 474 513 253 253 ------ ...................

135 555 253

..-............ _ .... *.. _... _.......;... *...--- -... ... ..- .. .. .......L- -........ ..

S 800 140 603 253

' --: -- - -- - ---- : ------....---- --- * --4-- *--

C 145 655 253 o

I. 150 713 253 ..-- ...

---.. . . .. .. he...

155 777 253 0 700 "...--.. ...---- :-- ........ if - ------------ - ---.. .. " ---- -- ------ *-

160 848 253 0I. 165 926 253 ---- ...I...

. -+1........ --- _ ._I .......-- .. ..... ... ...I...I... ........ . .

ix IC 170 1013 253 600 175 1108 253 ----- -- { --- "-----... * ;r :.............

C 180 1214 253 .. . .. . -.. . -- .~~~~~~1 -- ----.. .---- I---

. - . .--- ------ +--- -

180 1214 830 ------- --- ------- --- ---. ... . . .

2 500 185 1312 889 190 - 953 195 - 1024 a..

W 400 200 - 1103 ix 205 - 1190 ...

... ---. ---- ----- .... Ucoer Recions+

210 - 1258 --

S.... i i.. : ', : Use minimum of

-- ----.. ----.-.--- ,------ ------------------. r---------,+------ downcomer region 300 fluid temperature

,.---[ .

..-.... . . . . . . . . . outside surface temperature, except 200 I I I I I I I I when flange

- i I I I i ------. -....

-- i ---- ----,. .

--.J..t

...L...... ----

r---- --- *-----....

.- .- . - -----.... -- ....... i.... temperature is greater than 180°F.

100 use only the

__*.. .... S....

.... °... __-- - .---- i----'-------- downcomer region

-I---

II ------------.

~~

,j

.. .... . ....... __.....*... .... S.... ........

S....

i--.'----' --- fluid temperature 0 S, , , i r 6000 8000 10000 12000 14000 16000 18000 20000 TEMPERATURE (F)

Amendment No. 3, -3, 2-O3 137

VYNPS BASES:

3.6 and 4.6 REACTOR COOLANT SYSTEM A. Pressure and Temperature Limitations All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.2 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

The Pressure/Temperature (P/T) curves included as Figures 3.6.1, 3.6.2, and 3.6.3 were developed using 10CFR50 Appendix G, 1995 ASME Code,Section XI, Appendix G (including the Summer 1996 Addenda), and ASME Code Case N-640. These three curves provide P/T limit requirements for Pressure Test, Core Not Critical, and Core Critical. The P/T curves are not derived from Design Basis Accident analysis. They are prescribed to avoid encountering pressure, temperature or temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the reactor pressure boundary, a condition that is unanalyzed.

During heating events, the thermal gradients in the reactor vessel wall produce thermal stresses that vary from compressive at the inner wall to tensile at the outer wall. During cooling events the thermal stresses vary from tensile at the inner wall to compressive at the outer wall. The thermally induced tensile stresses are additive to the pressure induced tensile stresses. In the flange region, bolt preload has a significant affect on stress in the flange and adjacent plates.

Therefore heating/cooling events and bolt preload are used in the determination of the pressure-temperature limitations for the vessel.

The guidance of Branch Technical Position - MTEB 5-2, material drop weight, and Charpy impact test results were used to determine a reference nil-ductility temperature (RTNDT) for all pressure boundary components. For the plates and welds adjacent to the core, fast neutron (E > 1 Mev) irradiation will cause an increase in the RTN*.

For these plates and welds an adjusted RTNDT (ARTND) of 89'F and 73'F (14 and ý% thickness locations) was conservatively used in development of these curves for core region components. Based upon plate and weld chemistry, initial RTN* values, predicted peak fast neutron fluence 2

(2.99 x 1017 n/cm at the reactor vessel inside surface) for a gross power generation of 4.46 x 108 MWH(t), these core region ARTND values conservatively bound the guidance of Regulatory Guide 1.99, Revision 2.

There were five regions of the reactor pressure vessel (RPV) that were evaluated in the development of the P/T Limit curves: (1) the reactor vessel beltline region, (2) the bottom head region, (3) the feedwater nozzle, (4) the recirculation inlet nozzle, and (5) the upper vessel flange region. These regions will bound all other regions in the vessel with respect to considerations for brittle fracture.

Two lines are shown on each P/T limit figure. The dashed line is the Bottom Head Curve. This is applicable to the bottom head area only and includes the bottom head knuckle plates and dollar plates. Based on bottom head fluid temperature and bottom head surface temperature, the reactor pressure shall be maintained below the dashed line at all times.

Amendment No. .- a, -&9, 9, .93, .94, 44G, -4", 243 138

VYNPS BASES: 3.6 and 4.6 (Cont'd)

Due to convection cooling, stratification, and cool CRD flow, the bottom head area is subject to lower temperatures than the balance of the pressure vessel. The RTnD of the lower head is lower than the ARTNDT used for the beltline. The lower head area is also not subject to the same high level of stress as the flange and feedwater nozzle regions. The dashed Bottom Head Curve is less restrictive than the enveloping curve used for the upper regions of the vessel and provides Operator's with a conservative, but less restrictive P/T limit for the cooler bottom head region.

The solid line is the Upper Region Curve. This line conservatively bounds all regions of the vessel including the most limiting beltline and flange areas. At temperatures below the 10CFR50 Appendix G minimum temperature requirement (vertical line) based on the downcomer temperature and flange temperature, the reactor pressure shall be maintained below the solid line. At temperatures in excess of the 10CFR50 Appendix G minimum temperature requirement, the allowable pressure based on the flange is much higher than the beltline limit.

Therefore, when the flange temperature exceeds the 10CFR50 Appendix G minimum temperature requirement, the reactor pressure shall be maintained below the solid line based on downcomer temperature.

The Pressure Test curve (3.6.1) is applicable for heatup/cooldown rates up to 40 0 F/hr. The Core Not Critical curve (3.6.2) and the Core Critical curve (3.6.3) are applicable for heatup/cooldown rates up to 100*F/hr. In addition to heatup and cooldown events, the more limiting anticipated operational occurrences (AOOs) were evaluated (Structural Integrity Report, SIR-00-155). For the feedwater nozzles, a sudden injection of 50*F cold water into the nozzle was postulated in the development of all three curves. The bottom head region was independently evaluated for AOOs in addition to 40 0 F/hr and 100 0 F/hr heatup/cooldown rates. This evaluation demonstrated that P/T requirements of the bottom head would be maintained for transients that would bound rapid cooling as well as step increases in temperature.

The rapid cooling event would bound scrams and other upset condition (level B) cold water injection events. The bottom head was also evaluated for a series of step heatup transients. This would depict hot sweep transients typically associated with reinitiation of recirculation flow with stratified conditions in the lower plenum.

This demonstrated that there was significant margin to P/T limits with GE SIL 251 recommendations for reinitiating recirculation flow in stratified conditions.

Adjustments for temperature and pressure instrument uncertainty have been included in the P/T curves (Figures 3.6.1, 3.6.2 and 3.6.3). The minimum temperature requirements were all increased by 10°F to compensate for temperature loop uncertainty error. The maximum pressure values were all decreased by 30psi to account for pressure loop uncertainty error. In addition, the maximum pressure was reduced further to account for static elevation head assuming the level was at the top of the reactor and at 70 0 F.

Specification 3.6.A.3 requires that the temperature of the vessel head flange and the head be greater than 70*F before tensioning. The 70°F is an analytical limit and does not include instrumentation uncertainty, which must be procedurally included depending upon which temperature monitoring instrumentation is being used. The temperature values shown on Figures 3.6.1, 3.6.2 and 3.6.3 include a 10'F instrumentation uncertainty.

Amendment No. 243 139

VYNPS BASES: 3.6 and 4.6 (Cont'd)

A Note is included in Figure 3.6.2 that specifies test instrumentation uncertainty must be +/- 2'F and the flange region temperatures must be maintained greater than or equal to 72°F when using such instrumentation in lieu of permanently installed instrumentation.

Qualified test instrumentation may only be used for the purpose of maintaining the temperature limit when the vessel is vented and the fluid level is below the flange region. If permanently installed instrumentation (with a 10OF uncertainty) is used during head tensioning and detensioning operations, the 80'F limit must be met.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures will be maintained within 50OF of each other prior to startup of an idle loop.

Vermont Yankee is a participant in the Boiling Water Reactor Vessel and Internals Project Integrated Surveillance Program (ISP) for monitoring changes in the fracture toughness properties of ferritic materials in the reactor pressure vessel (RPV) beltline region. (See UFSAR Section 4.2 for additional ISP details.) As ISP capsule test reports become available for RPV materials representative of VYNPS, the actual shift in the reference temperature for nil-ductility transition (RTT) of the vessel material may be re-established. In accordance with Appendix H to 10CFR50, VY is required to review relevant test reports and make a determination of whether or not a change in Technical Specifications is required as a result of the surveillance data.

B. Coolant Chemistry A steady-state radioiodine concentration limit of 1.1 gCi of 1-131 dose equivalent per gram of water in the Reactor Coolant System can be reached if the gross radioactivity in the gaseous effluents is near the limit, as set forth in the Offsite Dose Calculation Manual, or if there is a failure or prolonged shutdown of the cleanup demineralizer. In the event of a steam line rupture outside the drywell, the NRC staff calculations show the resultant radiological dose at the site boundary to be less than 30 Rem to the thyroid. This dose was calculated on the basis of the radioiodine concentration limit of 1.1 gCi of 1-131 dose equivalent per gram of water, atmospheric diffusion from an equivalent elevated release of 10 meters at the nearest site boundary (190 m) for 3

a X/Q = 3.9 x 10-3 sec/M (Pasquill D and 0.33 m/sec equivalent), and a steam line isolation valve closure time of five seconds with a steam/water mass release of 30,000 pounds.

The iodine spike limit of four (4) microcuries of 1-131 dose equivalent per gram of water provides an iodine peak or spike limit for the reactor coolant concentration to assure that the radiological consequences of a postulated LOCA are within 10CFR Part 100 dose guidelines.

The reactor coolant sample will be used to assure that the limit of Specification 3.6.B.1 is not exceeded. The radioiodine concentration would not be expected to change rapidly during steady-state operation over a period of 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months . In addition, the trend of the radioactive gaseous effluents, which is continuously monitored, is a good indicator of the trend of the radioiodine concentration in the reactor coolant.

When a significant increase in radioactive gaseous effluents is indicated, as specified, an additional reactor coolant sample shall be taken and analyzed for radioactive iodine.

Amendment No. -3, 4-2, , 94, 44-4, 193, 203 140

ML030870156

Contents

Text

Subject:

1.0 INTRODUCTION

1.2 DESCRIPTION

2.0 BACKGROUND

4.0 REFERENCES

SUMMARY

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ML030870156

Text

Subject:

1.0 INTRODUCTION

1.2 DESCRIPTION

2.0 BACKGROUND

4.0 REFERENCES

SUMMARY

SUMMARY

============ ........ ...

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