ML050190317

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Virginia Electric and Power Company Surry Power Station Units 1 and 2 - Proposed Technical Specifications Change Request for Reactor Coolant System Pressure/Temperature Limits. Ltops Setpoint, and Ltops Enable Temperature with Exemption.
ML050190317
Person / Time
Site: Surry  Dominion icon.png
Issue date: 12/17/2004
From: Hartz L
Virginia Electric & Power Co (VEPCO)
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
04-755
Download: ML050190317 (32)
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Text

VIRGINIA ELECTRIC AND POW\II COMPANY RICIInMOND, VIRGINIA 23261 December 17, 2004 U.S. Nuclear Regulatory Commission Serial No.04-755 Attention: Document Control Desk NLOS/GDM RO Washington, D.C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGE REQUEST FOR REACTOR COOLANT SYSTEM PRESSURE/TEMPERATURE LIMITS.

LTOPS SETPOINT, AND LTOPS ENABLE TEMPERATURE WITH EXEMPTION REQUEST FOR ALTERNATE MATERIAL PROPERTIES BASIS PER 10 CFR 50.60(b)

Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion) requests amendments, in the form of changes to the Technical Specifications (TS) to Facility Operating Licenses Numbers DPR-32 and DPR-37 for Surry Power Station Units 1 and 2, respectively. The existing Reactor Coolant System (RCS) Pressure/

Temperature (P/T) operating limits, Low Temperature Overpressure Protection System (LTOPS) setpoint, and LTOPS enable temperature (Tenable) basis included in the Surry Unit 1 and Unit 2 TS are valid to cumulative core burnups of 28.8 Effective Full Power Years (EFPY) (approximately year 2012) and 29.4 EFPY (approximately year 2013) for Surry Units 1 and 2, respectively. The proposed TS change revises RCS P/T operating limits, LTOPS setpoint, and LTOPS Telable basis for cumulative core burnups up to 47.6 EFPY and 48.1 EFPY (corresponding to the period of the renewed licenses) for Surry Units 1 and 2, respectively. In addition, changes to the TS Bases reflecting the proposed changes are included for information only. An update to the NRC Reactor Vessel Integrity Database (RVID) is also provided.

A discussion of the proposed TS change is provided in Attachment 1. The marked-up and proposed TS pages reflecting the proposed change are provided in Attachments 2 and 3, respectively.

We have evaluated the proposed TS change and have determined that it does not involve a significant hazards consideration as defined in 10 CFR 50.92. We have also determined that operation with the proposed change will not result in any significant increase in the amount of effluents that may be released offsite, and no significant increase in individual or cumulative occupational radiation exposure will occur.

Therefore, the proposed amendment is eligible for categorical exclusion as set forth in

Serial No.04-755 Docket Nos. 50-280, 281 Page 2 of 3 10 CFR 51.22(c)(9), and, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed change. The bases for these two determinations are provided in .

The proposed TS change has been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Management Review Committee. After NRC approval of the proposed change, Dominion requests a six month implementation period to perform the changes necessary to implement the revised limits. Since the existing TS P/T limits, LTOPS setpoint, and LTOPS Tenable basis are valid to cumulative core burnups of 28.8 EFPY (approximately year 2012) and 29.4 EFPY (approximately year 2013) for Surry Units 1 and 2, respectively, the extended implementation time will have no impact on safe operation of Surry Unit 1 or 2.

Exemption Request Finally, a request for exemption pursuant to 10 CFR 50.12 and 50.60(b) from the requirements of 10 CFR 50.61 and 10 CFR 50 Appendix G is included in Attachment 4 to allow Dominion to revise the Surry reactor vessel material initial properties basis using BAW-2308, Revision 1. It should be noted that approval of the exemption is not required for approval of the proposed change to the Surry Units 1 and 2 Technical Specifications. Dominion requests the exemption to provide margin for possible future improvements in plant safety (e.g., reduced probability of undesired PORV lifts during reactor coolant pump startups).

If you have any further questions or require additional information, please contact Mr. Gary D. Miller at (804) 273-2771.

Very truly yours, Leslie N. Hartz Vice President - Nuclear Engineering Attachments Commitments made in this letter: None

Serial No.04-755 Docket Nos. 50-280, 281 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center Suite 23T85 61 Forsyth Street, SW Atlanta, Georgia 30303 Mr. S. R. Monarque U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8H12 Rockville, MD 20852 Mr. N. P. Garrett NRC Senior Resident Inspector Surry Power Station Commissioner Bureau of Radiological Health 1500 East Main Street Suite 240 Richmond, VA 23218

Serial No.04-755 Docket Nos. 50-280, 281

Subject:

Technical Specification Change for Revised P/T Limits, LTOPS Setpoints, LTOPS TEnable COMMONWEALTH OF VIRGINIA

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COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Leslie N. Hartz who is Vice President - Nuclear Engineering of Virginia Electric and Power Company. She has affirmed before me that she is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of her knowledge and belief.

774 / )

Acknowledged before me this /7 day of . &4YL 2004.

My Commission Expires:( Jig 3, d2oo, Notary Public

- (SEAL) *- -

Serial No.04-755 Docket Nos. 50-280, 281 ATTACHMENT 1 Discussion of Change Surry Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)

Serial No.04-755 Docket Nos. 50-280, 281 Attachment 1 Discussion of Changes 1.0 Introduction Virginia Electric and Power Company (Dominion) proposes a change to the Surry Units 1 and 2 Technical Specifications pursuant to 10CFR50.90. The proposed change is requested to provide Reactor Coolant System (RCS) pressure/temperature (P/T) operating limits, Low Temperature Overpressure Protection System (LTOPS) setpoints, and LTOPS enable temperature (Tenable) values that are valid for cumulative core burnups up to 47.6 Effective Full Power Years (EFPY) and 48.1 EFPY (corresponding to the period of the renewed licenses) for Surry Units 1 and 2, respectively. The currently licensed set of unadjusted RCS P/T limit curves (i.e., the set submitted in Reference 1, and approved in References 2 and 3) are being replaced with a new set (Reference 4, attached as Appendix D). In accordance with the ASME Code Section Xl, the higher cumulative core burnup applicability limits in this submittal are achieved through margins obtained by using K1c stress intensity factors in the development of the unadjusted RCS PIT limit curves (Reference 4, attached as Appendix D) instead of KIA stress intensity factors that represent the current licensing basis. In addition, Technical Specifications bases changes reflecting the proposed change discussed above are included for your information. The proposed TS change qualifies for categorical exclusion for an environmental assessment as set forth in IOCFR51.22(c)(9).

Therefore, no environmental impact statement or environmental assessment is needed in connection with approval of the proposed Technical Specifications change.

2.0 Background 10 CFR 50 Appendix G specifies the fracture toughness requirements for ferritic materials of pressure retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. The ASME Boiler and Pressure Vessel Code (BPVC) forms the basis for the requirements of Appendix G to 10 CFR 50. Appendix G references the requirements of ASME BPVC Section 1I1,Division 1, "Rules for Construction of Nuclear Power Plant Components," and ASME BPVC Section Xl which presents the "Rules for Inservice Inspection of Nuclear Power Plant Components."

10 CFR 50 Appendix H defines the requirements for reactor vessel materials surveillance programs. Dominion compliance with the requirements of Appendix H is documented for Surry Units 1 and 2 in References 5 and 6, respectively. Appendix H states that the purpose of the materials surveillance program is to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of light water nuclear power reactors resulting from exposure of these materials to neutron irradiation and the thermal environment. Fracture toughness data are obtained from material specimens exposed in surveillance capsules, which are withdrawn periodically from the reactor vessel. These data are used as described in Appendix G Page 1 of 15

Serial No.04-755 Docket Nos. 50-280, 281 Attachment 1 to 10 CFR 50. The current Surry Units 1 and 2 surveillance capsule withdrawal schedules are documented in Reference 7. The withdrawal schedules contained in the Surry UFSAR were approved in Reference 8.

A method for performing analyses to guard against brittle fracture in reactor pressure vessels is presented in "Protection Against Non-ductile Failure," Appendix G to Section Xl of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNDT).

RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208) or the temperature 600F less than the 50 ft-lb (and 35 mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the material. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KA), which appears in Appendix G of the ASME Section Xl. The KlA curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the KlA curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors.

RTNDT and the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement or changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor vessel materials surveillance program. A surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due to irradiation is added to the original RTNDT to adjust the RTNDT for radiation embrittlement. This adjusted RTNDT (RTNDT initial plus ARTNDT) is used to index the material to the KlA curve and to set operating limits for the nuclear power plant, which reflect the effects of irradiation on the reactor vessel materials.

The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The NRC has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2 (Radiation Embrittlement of Reactor Vessel Materials, Reference 13). This methodology permits the use of credible surveillance data, such as that obtained from the capsule analysis, if it is available, in place of the calculational methodology based on a curve fit of an irradiated materials properties database. The current Surry Units 1 and 2 design and licensing basis reflects the regulatory requirements described above by the imposition of restrictions on allowable pressure and temperature (RCS P/T limits) and on heatup and cooldown rates. The Low Temperature Overpressure Protection System (LTOPS) ensures that material integrity limits are not exceeded during design basis accidents.

The replacement of the RCS P/T limit curves and the revision of the LTOPS setpoint and LTOPS Terwable value proposed in this license amendment are performed in accordance with ASME Section Xl (i.e., use of the Kic stress intensity formulation as Page 2 of 15

Serial No.04-755 Docket Nos. 50-280, 281 Attachment 1 allowed in Code Case N-641, Reference 10) and the regulatory requirements described above as well.

3.0 Discussion 3.1 Licensing and Design Basis The current Surry Units 1 and 2 Technical Specification RCS P/T limits (using a KlA stress intensity formulation), LTOPS setpoint and LTOPS Tenable value were provided to the NRC for approval in Reference 1. The NRC approved the Technical Specification change in References 2 and 3. The cumulative core burnup applicability limit for the current limits are 28.8 EFPY for Surry Unit 1 and 29.4 EFPY for Surry Unit 2. This corresponds to a 1/4-thickness (1/4-T) RTNDT of 228.40F, which conservatively represents the limiting materials for both Surry Units 1 and 2. The submittals supporting license renewal for Surry Units 1 and 2 (References 7 and 9) were approved by the NRC (Reference 8). These submittals documented that the limiting 1/4-T RTNDT value of 238.20F was predicted to bound the end-of-license-renewal limiting material for both Surry Units 1 and 2 (Unit 1 Longitudinal Weld L2, SA-1526). Reviews of the Surry Units 1 and 2 reactor vessel integrity data continue to confirm the conclusions from the license renewal effort.

To support the end-of-license-renewal limiting material for both Surry Units 1 and 2 (Unit 1 Longitudinal Weld L2, SA-1526), new Surry Units 1 and 2 RCS P/T limit curves were prepared using the Kjc stress intensity formulation and a limiting /4-T RTNDT of 238.20F. This RTNDT value corresponds to operation for 47.6 EFPY for Surry Unit 1 and 48.1 EFPY for Surry Unit 2 (Reference 4, attached as Appendix D). The K1c curve is a lower bound of dynamic, crack initiation, and static fracture toughness results obtained from several heats of pressure vessel steel. Use of the Kic stress intensity formulation is allowed by 10 CFR 50 Appendix G that, in turn, endorses the use of ASME Section Xl. To extend the cumulative core burnup applicability limit for the Surry Units 1 and 2 Technical Specification RCS P/T limits, LTOPS setpoint, and LTOPS Tenable value, the Surry Units 1 and 2 Technical Specifications governing these values are being revised to be consistent with a 1/4-T RTNDT value of 238.20F. These changes include:

1. The revised RCS PTT limits in Appendix A include modifications for pressure and temperature measurement uncertainty, as well as the pressure difference between the point of measurement (RCS hot leg) and point of interest (reactor vessel beltline).
2. The cumulative core burnup applicability limits will be extended to 47.6 EFPY for Surry Unit 1 and 48.1 EFPY for Surry Unit 2.
3. Technical Specification LTOPS setpoint and LTOPS Teable value have been prepared to reflect the extended cumulative core burnup applicability limits using methodologies that comply with the applicable regulations (e.g., 10 CFR 50 Appendix G), industry codes (e.g., ASME Section Xl), and previously approved methods.

Page 3 of 15

Serial No.04-755 Docket Nos. 50-280, 281 Attachment 1

4. The changes to the Technical Specification P/T limits, LTOPS setpoint, and Tenable value will be common for Surry Units 1 and 2 to provide more consistent operational requirements for the two units.

In addition to the Technical Specification changes, Dominion is increasing the administrative cooldown rate limit from 50°F/hr to 750F/hr. The justification for this increase is described in Section 3.4.4.

3.2 Design Inputs 3.2.1 Unadjusted Pressure/Temperature Limit Curves The current Surry Reactor Coolant System (RCS) pressure/temperature (P/T) limit curves (designed for 40 calendar years of operation) were developed by Westinghouse and were transmitted to the NRC in Reference 1. The design limit for the current Surry LTOPS setpoints is 110% of the isothermal RCS P/T curve based upon KlA stress intensity factors (Reference 1), as allowed by 10 CFR 50 Appendix G that, in turn, endorses the use of Section Xl of the ASME Code (specifically ASME Code Case N-514, References 1, 2, and 3). The KlA curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel.

Surry RCS P/T limit curves (designed for 60 years of operation) were also developed by Westinghouse, who documented their technical bases in Reference 4 (Appendix D).

Westinghouse developed the P/T limit curves for a 1/4/-T RTNDT of 238.20F. This RTNDT value is predicted to bound the end-of-license-renewal limiting material for both Surry Units 1 and 2 (Unit 1 Longitudinal Weld L2, SA-1526) as documented in Reference 7.

ASME Code Case N-641, Reference 10, supports the use of 100% of the isothermal P/T curve using Kic stress intensity factors as the design limit for LTOPS setpoints.

The Kic curve is a lower bound of dynamic, crack initiation, and static fracture toughness results obtained from several heats of pressure vessel steel. The use of Kic stress intensity factors provides greater margin in the development of P/T limit curves relative to the use of KlA. However, the use of the K1c stress intensity formulation requires that pressure and temperature measurement uncertainties be applied, and that 100% of the isothermal PIT limit curve be used instead of 110% of the isothermal P/T curve as allowed when using K1A. The P/T limit curves provided in Reference 4 (Appendix D) are valid up to 47.6 EFPY for Surry Unit 1 and 48.1 EFPY for Surry Unit 2.

The revised Surry Units 1 and 2 design basis RCS PIT limit curves (Reference 4) do not include margins for pressure and temperature measurement uncertainty, or for the pressure difference between the point of measurement (RCS hot leg) and the point of interest (reactor vessel beltline). Curves that have been modified to include pressure and temperature measurement uncertainty, and the pressure difference between the point of measurement (RCS hot leg) and the point of interest (reactor vessel beltline) are presented in Appendix A.

Page 4 of 15

Serial No.04-755 Docket Nos. 50-280, 281 Attachment 1 3.2.2 Reactor Vessel Fluence (E > 1 MeV), RTNDT, and Cumulative Core Burnup The NRC was provided (in Reference 7) information regarding reactor vessel fluence (E

> 1 MeV) versus burnup and RTNDT versus reactor vessel fluence in support of 60-year operation. The information in Reference 7 demonstrated that the limiting 1A-T RTNDT value of 238.2 0F conservatively represents the end-of-license-renewal limiting material for both Surry Units 1 and 2. The limiting material, Unit 1 Longitudinal Weld L2, SA- 1526, is predicted to have an inner surface fluence of (0.79 X 1019 n/cm2) that corresponds to a cumulative core burnup of 47.6 EFPY for Surry Unit 1 and 48.1 EFPY for Surry Unit 2. Currently available material surveillance data has not changed this conclusion. The methodology used for the fluence values given above has been shown to meet the requirements of Regulatory Guide 1.190 verbatim, or by equivalent demonstration, as documented in Reference 14. Although Reference 14 represents an RAI response for the North Anna Units, the methodology used for calculating the Surry fluences is the same as that used for North Anna.

3.2.3 RCS Pressure Measurement Uncertainty for LTOPS The Channel Statistical Accuracy (CSA) for the RCS pressure measurement uncertainty to be used for the establishment of the LTOPS setpoint has been calculated to be

+2.05% of a 0 psig to 1000 psig instrument span, for a total CSA of 20.5 psi. For conservatism, a value of 25 psi is used in the development of the LTOPS setpoint. This uncertainty reflects the Narrow Range RCS pressure uncertainty for the actuation of the PORV bistables. Narrow Range RCS pressure is measured in the RCS hot leg for the PORV with the limiting setpoint.

3.2.4 Wide Range RCS Pressure Measurement Uncertainty for P/T Limits The CSA for the Wide Range RCS pressure measurement uncertainty has been calculated to be 2.213% of a 0 psig to 3000 psig instrument span (indication uncertainty included), for a total CSA of 67 psi. Wide Range RCS pressure is measured in the RCS hot leg. The Wide Range RCS pressure measurement channel is used for confirming RCS pressure during normal operation heatup and cooldown.

3.2.5 Wide Range RCS Temperature Measurement Uncertainty The CSA for the Wide Range RCS temperature measurement uncertainty has been calculated to be 2.0% of a 0F to 7000F instrument span, for a total CSA of 140F. For conservatism, a value of 200F is used in the development of the LTOPS enabling temperature. Wide Range RCS temperature is measured in the RCS cold leg. The Wide Range RCS temperature measurement channel is used for confirming RCS temperature during normal operation heatup and cooldown, and as input for the LTOPS enabling temperature.

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Serial No.04-755 Docket Nos. 50-280, 281 Attachment 1 3.2.6 Pressure Difference between Hot Leg and Reactor Vessel Beltline The pressure difference between the point of measurement (Narrow Range or Wide Range RCS pressure measured in the RCS hot leg) and the point of interest (reactor vessel beltline) has been determined to be 57 psi. This value was developed in consideration of one reactor coolant pump (RCP), two RCP, and three RCP operation.

This difference is applied as a bias to measured RCS pressure, to simulate pressure measurement at the reactor vessel beltline.

3.2.7 LTOPS PORV Lift Setpoint "Overshoot" Values from Mass Addition Accident Analysis The mass addition and heat addition accident analyses that support the proposed Surry Units 1 and 2 Technical Specification LTOPS setpoint are unchanged from those used for the current licensing analysis (References 1 through 3). The PORV lift setpoint "overshoot" values determined in the accident analysis are presented in Appendix B.

(See column labeled "PORV Setpoint Overshoot".) The maximum PORV lift setpoint overshoot is a function of the PORV lift setpoint and RCS temperature. Note that a pressure measurement location bias of approximately 9 psi, originally applied to the values contained in the "PORV Setpoint Overshoot" column to account for the static head difference between the RCS hot leg and the reactor vessel beltline, has been removed because it is redundant. The treatment of static head pressure measurement bias is now contained in the 57 psi value described in Item 3.2.6.

3.2.8 Margin Term for ASME Code Case N-641 In the plant specific determination of LTOPS Tenable, ASME Code Section Xl (specifically Code Case N-641, Reference 10) requires that a margin term value be determined and applied to ensure that LTOPS provides adequate protection against brittle failure at low temperatures. This value represents a margin term that accommodates the specific geometry and design pressure of the reactor vessel considered. The following is the plant specific determination of the ASME Code Section Xl margin term:

ASME Section XI Margin Term = 50 In[((1.1

  • Mm (p RI /t)) - 33.2)/20.734] (from ASME Code Section Xl)

Mm= 0.926 t/ 2 for IS axial flaw, 2 < t 2< 3.464 p= vessel design pressure = 2.5 ksia RI= 78.95 in.

t= 8.08 in.

The Surry specific values for p, R., and t are from Reference 4 (Appendix D), Section 6.

Section Xl Margin Term = 50 In[(1.1 * -0.926e8.0812 * (2.5.78.95/8.08) - 33.2)/20.734]

Section Xl Margin Term = 29.71F (Note: The ASME Code Section Xl formulation for the membrane stress correction factor, M, is valid since t" = (8.08)1'2 = 2.84, which satisfies the inequality 2 s tl2 < 3.464.)

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Serial No.04-755 Docket Nos. 50-280, 281 Attachment 1 3.3 Method of Analysis To develop the proposed Surry Units 1 and 2 Technical Specification P/T limit curves, LTOPS setpoint, and LTOPS Tenable value, the unadjusted P/T limit curves (Reference 4, attached as Appendix D)) were modified to account for pressure and temperature measurement uncertainty, and for the pressure difference between the point of measurement (RCS hot leg) and the point of interest (the reactor vessel beltline). The resulting proposed Surry Units 1 and 2 Technical Specification P/T limit curves are presented in Appendix A.

To determine the allowable TS LTOPS setpoint value, the temperature-dependent pressurizer PORV lift setpoint pressure "overshoot" values determined in the design basis mass addition and heat addition accident analysis were subtracted from the revised LTOPS design basis (i.e., isothermal) P/T limit curve. The margin between the proposed TS PORV lift setpoint allowable value pressure and the temperature-dependent LTOPS setpoint pressure allowable value is verified to be positive at each temperature. The results of the LTOPS margin assessment performed using this methodology are presented in Appendix B. Note in Appendix B that the variable MULT represents a multiplier on the design basis isothermal curve. MULT has been set to a value of 1.0 since the K1c stress intensity formulation requires that 100% of the isothermal P/T limit curve is used instead of 110% of the isothermal P/T curve as allowed when using KIA.

LTOPS Tenable is the temperature below which LTOPS must be enabled. The LTOPS Tenable value was determined using ASME Section Xl (specifically, the features of ASME Code Case N-641, Reference 10, included in ASME Section Xl) and is calculated as the sum of the following:

  • 238.20F, the 1/4-T RTNDT,
  • 29.7 0F, the margin required by ASME Code Section Xl for plant specific applications,
  • 150F, the margin for the temperature lag between the quarter-thickness vessel location and the coolant temperature during a 600F/hr heatup (i.e.,

600 F/hr heatup data from Reference 11, attached as Appendix E) provided a value of 14.20F at 3200 F; conservatively rounded to 150 F to account for a higher LTOPS Tenable),

  • 200F, the temperature measurement instrument uncertainty, and
  • 250F, the estimated temperature difference between the cold leg and the belt-line materials during a cooldown with natural circulation.

Page 7 of 15

Serial No.04-755 Docket Nos. 50-280, 281 Attachment 1 3.4 Results 3.4.1 Revised P/T Umit Curves As described in Section 3.2.1, the unadjusted P/T limit curves, including the LTOPS design basis P/T limit curve (i.e., the "steady state", "isothermal", or "0F/hr cooldown" curve), were modified to account for pressure and temperature measurement uncertainty, and for the pressure difference between the point of measurement (RCS hot leg) and the point of interest (the reactor vessel beltline). The resulting proposed revised Surry Units 1 and 2 Technical Specification P/T limit curves are presented in Appendix A.

3.4.2 Revised LTOPS Setpoint Allowable Values The Narrow Range RCS pressure measurement channel feeds the logic for opening and closing the pressurizer PORV at conditions during which the LTOPS system is enabled (i.e., temperatures that are below the LTOPS Tenable).

The Surry TS LTOPS setpoint consists of the following variables:

- LTOPS PORV Setpoint

- LTOPS Tenable (Derived in Section 3.4.3)

LTOPS Tenable is the temperature below which LTOPS must be enabled. For temperatures above Tenable, adequate overpressure protection is provided by the pressurizer safety valves (PSVs). In addition to the setpoint described above, the LTOPS PORV bistables are set in a staggered fashion for each of the two pressurizer PORVs. The staggered bistable control setpoints for each PORV avoids simultaneous PORV lift. In addition, for Surry Units 1 and 2, the stagger is also required as the RCS pressure measurement for one PORV is received from the RCS hot leg and the other PORV receives RCS pressure information from the pressurizer. The pressure difference between the RCS hot leg and the pressurizer is approximately 10 psig. Note that the PORV bistable that takes pressure information from the pressurizer is set 20 psig lower than the PORV bistable that obtains pressure information from the RCS hot leg. The value of 20 psig provides sufficient margin for the pressure measurement difference between the hot leg and the pressurizer and sufficient margin to provide the necessary staggering to prevent simultaneous opening of both PORVs. For ease of comparison with Reference 1, only the higher and hence lower margin PORV setpoint (i.e., that controlled by the RCS hot leg pressure measurement) is described below.

The results of calculations performed for the LTOPS margin assessment outlined in Section 3.3 are presented in Appendix B. As Appendix B demonstrates, the revised Surry Units 1 and 2 TS LTOPS setpoint provides bounding protection for 100% of the proposed revised design basis isothermal curve under postulated mass addition and heat addition accident conditions. The analysis includes consideration of pressure and temperature uncertainties, as well as the pressure difference between the point of Page 8 of 15

Serial No.04-755 Docket Nos. 50-280, 281 Attachment I measurement (RCS hot leg) and the point of interest (reactor vessel beltline). The design basis P/T limit curves are based on a 1/4-T RTNDT of 238.20F, which conservatively bounds the most limiting /-T RTNDT values at cumulative core burnups of 47.6 EFPY and 48.1 EFPY for Surry Units 1 and 2, respectively. Therefore, the revised Surry Units 1 and 2 LTOPS setpoint is concluded to be conservative for cumulative core burnups up to 47.6 EFPY and 48.1 EFPY for Surry Units 1 and 2 respectively.

The proposed revised TS LTOPS setpoint (valid for both Surry Units 1 and 2) is shown below. The development of the bistable control setpoints will be performed in a manner to provide for additional margin to this value to accommodate postulated setpoint drift between periodic calibrations.

Surry Unit 1 and 2 Technical Specification *5395 psig @Cl I LTOPS Setpoint I *350 0F 3.4.3 Revised LTOPS Tenable As described in Section 3.3, the temperature below which LTOPS must be enabled is calculated as the summation of various components. Below is the equation in algebraic form:

LTOPS Tenabe(0 F) = RTNDT(1/4-T) + 29.7 0F + 150F[AT(1/4-T)] + 20'F(Temperature Measurement Uncertainty) + 250F (natural circulation bias)

Using a limiting 1/4-T RTNDT of 238.2 0F, which corresponds to operation for 47.6 EFPY for Surry Unit 1 and 48.1 EFPY for Surry Unit 2, a minimum value for LTOPS Tenable Of 327.90F is required. To provide additional conservatism, and to avoid unnecessary setpoint and procedure changes, the proposed value for LTOPS Tenable will remain at 3500F for Surry Units 1 and 2.

3.4.4 Administrative RCS Cooldown Rate Limit for Surry Units 1 and 2 While the maximum allowable RCS cooldown rate assumed in the development of the P/T limit curves is 100°F/hr, a 50°F/hr administrative RCS cooldown rate is currently in effect as described in Reference 1. The administrative limit was established to ensure the adequacy of the P/T limits for non-linear cooldown ramp rates (i.e., short duration temperature changes of limited magnitude that may occur during normal operation, but which may result in calculated cooldown rates in excess of the limits prescribed in the Technical Specifications). This section addresses use of a 750F/hr administrative cooldown rate limit.

The concern presented by an increase in the allowable administrative RCS cooldown rate is related to the operator's ability to control cooldown rate. It can be reasonably estimated that cooldown rate can be controlled to within 250F/hr. While it is possible to have short duration changes of limited magnitude that can exceed this rate, such Page 9 of 15

Serial No.04-755 Docket Nos. 50-280, 281 Attachment 1 changes only produce small changes in overall metal temperature, and have no significant effect on the applied stress intensity at the assumed crack tip. Studies of the effects of "step changes" in cooldown rate suggest that more restrictive limits may be appropriate if cooldown rates are not held constant at rates less than analyzed.

However, on the basis of engineering judgement, small step changes (e.g., < 250 F) do not present a significant concern since the reactor vessel material would experience insignificant temperature change at the assumed 1/4-T flaw location (i.e., limited contribution to reactor vessel stress). Increasing the administrative cooldown rate from 50°F/hr to 750F/hr continues to provide adequate margin to the analyzed rate to accommodate small unanticipated changes in cooldown rate due to short duration temperature changes of limited magnitude.

3.4.5 RTPTS Screening Reference 7 stated that the limiting material with respect to PTS screening was the Surry Unit 1 Longitudinal Weld L2, SA-1526. The information in Reference 7 demonstrated that the limiting RTPTS value of 268.50F represents the end-of-license-renewal limiting material for both Surry Units 1 and 2 (Unit 1 Longitudinal Weld L2, SA-1526) corresponding to a cumulative core burnup of 47.6 EFPY for Surry Unit 1 and 48.1 EFPY for Surry Unit 2. Currently available material surveillance data has not changed this conclusion.

4.0 Changes to Surry Units 1 and 2 Technical Specifications The following specific changes to the Surry Units 1 and 2 Technical Specifications are proposed:

  • Technical Specification 3.1.8: Figures 3.1-1 and 3.1-2 and the cumulative core burnup limits are being replaced by revised Figures 3.1-1 and 3.1-2, Reactor Coolant System Heatup and Cooldown Limitations. The proposed curves (provided in Appendix A) will be valid for both Surry Units 1 and 2. Note also that the axis labels have been clarified to identify the source of the parameter indication used (i.e., indicated wide range instrumentation). Basis changes reflect both the K1c stress intensity formulation and the extended cumulative core burnup applicability limits.

5.0 Significant Hazards Consideration Determination Virginia Electric and Power Company (Dominion) has reviewed the requirements of 10 CFR 50.92, relative to the proposed change to the Surry Units 1 and 2 Technical Specifications, and determined that a Significant Hazards Consideration is not involved.

The proposed change to the Surry Units 1 and 2 Technical Specifications modifies the Page 10 of 15

Serial No.04-755 Docket Nos. 50-280, 281 Attachment 1 Reactor Coolant System (RCS) pressure/temperature (PIT) limit curves, LTOPS setpoint, and LTOPS Tenable value, and extends the cumulative core burnup applicability limits for these parameters. The proposed P/T limit curves, LTOPS setpoint, and LTOPS Tenable value are valid to cumulative core burnups of 47.6 EFPY and 48.1 EFPY for Surry Units 1 and 2, respectively.

The following is provided to support this conclusion that the proposed change does not create a significant hazards consideration.

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

The proposed change modifies the Surry Units 1 and 2 RCS P/T limit curves, LTOPS setpoint, and LTOPS Tenable value and extends the cumulative core burnup applicability limits for these parameters. The allowable operating pressures and temperatures under the proposed RCS P/T limit curves are not significantly different from those allowed under the existing Technical Specification P/T limits. The revisions in the values for the LTOPS setpoint and LTOPS Tenable do not significantly change the plant operating space. No changes to plant systems, structures or components are proposed, and no new operating modes are established. The P/T limits, LTOPS setpoint, and Tenable value do not contribute to the probability of occurrence or consequences of accidents previously analyzed. The revised licensing basis analyses utilize acceptable analytical methods, and continue to demonstrate that established accident analysis acceptance criteria are met. Therefore, there is no increase in the probability or consequences of any accident previously evaluated.

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

The proposed change modifies the Surry Units 1 and 2 RCS P/T limit curves, LTOPS setpoint, and LTOPS Tenable value and extends the cumulative core burnup applicability limits for these parameters. The allowable operating pressures and temperatures under the proposed RCS P/T limit curves are not significantly different from those allowed under the existing Technical Specification P/T limits. No changes to plant systems, structures or components are proposed, and no new operating modes are established. Therefore, the proposed changes do not create the possibility of any accident or malfunction of a different type previously evaluated.

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

The proposed revised RCS PIT limit curves, LTOPS setpoint, and LTOPS Tenable value analysis bases do not involve a significant reduction in the margin of safety for these parameters. The proposed revised RCS P/T limit curves are valid to cumulative core burnups of 47.6 EFPY and 48.1 EFPY for Surry Units 1 and 2, Page 11 of 15

Serial No.04-755 Docket Nos. 50-280, 281 Attachment 1 respectively. The proposed revised LTOPS setpoint and Tenable analyses support these same cumulative core burnup limits. The proposed revised RCS P/T limit curves utilize ASME Code Section Xl, which supports use of a conservative but less restrictive stress intensity formulation (Kic). The proposed extension of the cumulative core burnup applicability limits along with a small increase in the LTOPS PORV setpoint is accommodated by the margin provided by ASME Code Section Xl. The analyses demonstrate that established analysis acceptance criteria continue to be met. Specifically, the proposed P/T limit curves, LTOPS setpoint and LTOPS Tenable value provide acceptable margin to vessel fracture under both normal operation and LTOPS design basis (mass addition and heat addition) accident conditions. Therefore, the proposed change does not result in a significant reduction in margin of safety.

6.0 Environmental Assessment The proposed Technical Specification (TS) change to the Reactor Coolant System (RCS) pressure/temperature (P/T) limit curves, LTOPS setpoint, and LTOPS enable temperature (Tenable) value and the extended cumulative core burnup applicability limits for these parameters meet the eligibility criteria for categorical exclusion from an environmental assessment set forth in 10 CFR 51.22(c)(9), as discussed below:

(i) The license condition involves no Significant Hazards Consideration.

As discussed in the evaluation of the Significant Hazards Consideration above, the proposed change to the RCS P/T limit curves, LTOPS setpoint, and LTOPS Tenable value for Surry will not involve a significant increase in the probability or consequences of an accident previously evaluated. The possibility of a new or different kind of accident from any accident previously evaluated is also not created, and the proposed change does not involve a significant reduction in a margin of safety. Therefore, the proposed change to the RCS P/T limit curves, LTOPS setpoint, and LTOPS Tenable value meet the requirements of 10 CFR 50.92(c) and do not involve a significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The allowable operating pressures and temperatures under the proposed RCS P/T limit curves are not significantly different from those allowed under the existing Technical Specification P/T limits. No changes to plant systems, structures or components are proposed, and no new operating modes are established. Therefore, the proposed change to the RCS P/T limit curves, LTOPS setpoint, and LTOPS Tenable will not significantly change the types, or significantly increase the amounts, of effluents that may be released offsite.

Page 12 of 15

Serial No.04-755 Docket Nos. 50-280, 281 Attachment 1 (iii)There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed change modifies the Surry Units 1 and 2 RCS P/T limit curves, LTOPS setpoint, and LTOPS Tenable value, and extends the cumulative core burnup applicability limits for these parameters. The allowable operating pressures and temperatures under the proposed RCS P/T limit curves are not significantly different from those allowed under the existing Technical Specification P/T limits. No changes to plant systems, structures or components are proposed, and no new operating modes are established. In addition, the supporting analyses for the proposed changes continue to provide acceptable margin to vessel fracture under both normal operation and LTOPS design basis (mass addition and heat addition) accident conditions. Therefore, the proposed changes will not increase radiation levels compared to the existing Technical Specification PIT limits, LTOPS setpoint, and LTOPS Tenable value, so individual and cumulative occupational exposures are unchanged.

Based on the above, the proposed changes do not have a significant effect on the environment, and meet the criteria of 10 CFR 51.22(c)(9). Therefore, the proposed Technical Specification change qualifies for a categorical exclusion from a specific environmental review by the Commission, as described in 10 CFR 51.22.

7.0 Updates to the Reactor Vessel Integrity Database (RVID)

Table 1 of Appendix C of this submittal contains the RVID update based on current material properties basis (i.e., supporting operation for 47.6 EFPY and 48.1 EFPY for Surry Units 1 and 2, respectively).

In addition, Table 2 of Appendix C contains proposed alternate material properties basis that is developed from BAW-2308, Revision 1 (Reference 12). Reference 12 establishes revised (i.e., reduced) initial RTNDT values for the Unde 80 weld heat materials. Revised initial RTNDT values could be used in the future by Dominion to make various plant safety improvements (i.e., reduced probability of undesired PORV lifts during reactor coolant pump startups). An exemption request is provided in Attachment

4. Following NRC approval of this exemption request, Table 2 of Appendix C would serve as the RVID update. Note that NRC approval of the exemption is not required for approval of the proposed TS change as the TS change is supported by the current material properties basis (Table 1 of Appendix C).

8.0 Conclusions A change to the Surry Units 1 and 2 Technical Specifications is proposed to extend the cumulative core burnup applicability limit for the Surry Units 1 and 2 Technical Specification RCS P/T limits, LTOPS setpoint, and LTOPS Tenable value. This change has been developed using methodologies that comply with the applicable regulations (e.g., 10 CFR 50 Appendix G), industry codes (e.g., ASME Section Xl), and previously Page 13 of 15

Serial No.04-755 Docket Nos. 50-280, 281 Attachment 1 approved methods. The proposed change to the Surry Units 1 and 2 Technical Specifications will continue to provide acceptable margin with respect to the prevention of reactor vessel brittle fracture. Therefore, the proposed change will not adversely impact safe operation of Surry Units 1 and 2.

9.0 References

1. Letter from J. P. O'Hanlon to USNRC, 'Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Request for Exemption - ASME Code Case N-514 Proposed Technical Specifications Change, Revised Pressure/Temperature Limits and LTOPS Setpoints," dated June 8,1995.
2. Letter from USNRC to J. P. O'Hanlon, "Exemption from Requirements of 10CFR50.60, Acceptance Criteria for Fracture Prevention for Light-Water Nuclear Power Reactors for Normal Operation, Surry Power Station, Units 1 and 2, (TAC NOS. M92537 and M92538)," dated October 31, 1995.
3. Letter from USNRC to J. P. O'Hanlon, "Surry Units 1 and 2 - Issuance of Amendments Re: Surry, Units 1 and, 2 Reactor Vessel Heatup and Cooldown Curves (TAC NOS. M92537 and M92538)," dated December 28,1995.
4. WCAP-15130, "Surry Units 1 and 2 WOG Reactor Vessel 60-Year Evaluation Minigroup Heatup and Cooldown Limit Curves for Normal Operation," Revision 1, April 2001.
5. WCAP-7723, "Surry Unit 1 Reactor Vessel Radiation Surveillance Program," dated July 1971.
6. WCAP-8085, "Surry Unit 2 Reactor Vessel Radiation Surveillance Program," dated June 1973.
7. Letter from L. N. Hartz to USNRC, "Virginia Electric and Power Company (Dominion), Surry and North Anna Power Stations Units 1 and 2, Response to Request for Supplemental Information License Renewal Applications," Serial No.02-601, dated October 15,2002.
8. Letter from USNRC to D. A Christian, "License Renewal Safety Evaluation Report for North Anna, Units 1 and 2, and Surry, Units 1 And 2", Serial No.02-709, November 5,2002.
9. Letter from D. A. Christian to USNRC, 'Virginia Electric and Power Company, Surry and North Anna Power Stations Units 1 and 2, License Renewal Applications -

Submittal," Serial No. 01 -282, dated May 29,2001.

Page 14 of 15

Serial No.04-755 Docket Nos. 50-280, 281 Attachment 1 10.ASME Code Section Xl, Code Case N-641, "Alternate Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements",

dated January 17, 2000.

11. Letter VPA-03-193 from Westinghouse, "Thermal Stress Intensity Factors and Vessel Wall Temperatures for PT Curves from WCAP-15130, Revision 1," dated October 9, 2003.

12.BAW-2308, "Initial RTNDT of Linde 80 Weld Materials," Revision 1, dated August 2003.

13.Regulatory Guide 1.99 Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," dated May 1988.

14.Letter from L. N. Hartz to USNRC, 'Virginia Electric and Power Company (Dominion), North Anna Power Station Units 1 and 2, Request for Additional Information, Proposed Technical Specification Change Request, Reactor Coolant System Pressure/Temperature Limits, LTOPS Setpoints and LTOPS Enable Temperatures," Serial No. 04-380A, dated October 28,2004.

Page 15 of 15

Serial No.04-755 Docket Nos. 50-280, 281 APPENDIX A Pressure/Temperature Limit Curves Surry Units 1 and 2

Table 1 Surry Units 1 and 2 Heatup Data with Margins of 20 Degrees F and 67 psi for Instrumentation Errors (WCAP-15130 Rev. 1) and 57 psi for Pressure Measurement Bias 1, Heatup Rate = 20 Deg. Flhr Heatup Rate = 40 Deg. Flhr Heatup Rate = 60 Deg. F/hr Indicated Indicated Indicated Indicated Indicated Indiccated Temperature Pressure Temperature Pressure Temperature Pres sure (Deg. F) (psig) (Deg. F) (psig) (Deg. F) (pssig) 1 100 497 1 100 497 1 100 4C37 2 105 497 2 105 497 2 105 4C 3 110 497 3 110 497 3 110 4C W97 4 115 497 4 115 497 4 115 4C 5 120 497 5 120 497 5 120 4C57 6 125 497 6 125 497 6 125 4C 7 130 497 7 130 497 7 130 4$17 8 135 497 8 135 497 8 135 397 4$37 9 140 497 9 140 497 9 140 4!17 10 145 497 10 145 497 10 145 4! 37 11 150 497 11 150 497 11 150 4$97 12 150 566 12 150 566 12 150 5S25 13 155 571 13 155 571 13 155 5'32 14 160 576 14 160 576 14 160 39 15 165 582 15 165 582 15 165 . 5 16 170 589 16 170 589 16 170 5!i58 17 175 596 17 175 596 17 175 5 ;70 18 180 604 18 180 604 18 180 5E31 19 185 613 19 185 613 19 185 5!27297 20 190 623 20 190 623 20 190 61 814 21 195 634 21 195 634 21 195 6'32 22 200 646 22 200 646 22 200 6';46 23 205 659 23 205 659 23 205 6';59 24 210 673 24 210 673 24 210 67;73 25 215 690 25 215 690 25 215 6!;90 26 220 707 26 220 707 26 220 7t127 27 225 727 27 225 727 27 225 7 627 28 230 749 28 230 749 28 230 7' 29 235 773 29 235 773 29 235 7 r73 30 240 800 30 240 800 30 240 81300 31 245 829 31 245 829 31 245 8 329 32 250 861 32 250 861 32 250 8131 33 255 897 33 255 897 33 255 8!397 34 260 937 34 260 937 34 260 9 337 35 265 981 35 265 981 35 265 9J81 36 270 1029 36 270 1029 36 270 iCr49

)29 37 275 1083 37 275 1083 37 275 IC083 38 280 1142 38 280 1142 38 280 11142 39 285 1202 39 285 1201 39 285 205 40 290 1268 40 290 1262 40 290 260 41 295 1341 41 295 1328 41 295 13321 42 300 1421 42 300 1401 42 300 I 388 43 305 1510 43 305 1482 43 305 1 462 44 310 1608 44 310 1572 44 310 1E544 45 315 1716 45 315 1671 45 315 1E634 46 320 1836 46 320 1780 46 320 17734 47 325 1968 47 325 1900 47 325 1E843 48 330 2115 48 330 2033 48 330 19965 49 335 2276 49 335 2180 49 335 2C098 50 340 2342 .

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Applicable for the first 47.6 EFPY for Unit 1 and 48.1 EFPY for Unit 2 (Including Margins for Instrumentation Errors)

Table2 Surry Units 1 and 2 Cooldown Data with M argins of 20 Degrees F and 67 psi for Instrumentation Errors (WCAP-15130 Rev. 1, Modified) and 57 psi for Pressure Measurement Bias - I Cooldown Rate = 0 Deg. F/hr Cooldown Rate = 20 Deg. F/hr Cooldown Rate = 40 Deg. F/hr Indicated Indicated Indicated Indicated Indicated Indicated Temperature Pressure Temperature Pressure Temperature Pressure (Deg. F) (psig) (Deg. F) (psig) (Deg. F) (psig) 1 100 497.00 1 100 490.72 1 100 443.45 2 105 497.00 2 105 492.47 2 105 445.16 3 110 497.00 3 110 494.41 3 110 447.08 4 115 497.00 4 115 496.59 4 115 449.27 5 120 497.00 5 120 497.00 5 120 451.71 6 125 497.00 6 125 497.00 6 125 454.47 7 130 497.00 7 130 497.00 7 130 457.56 8 135 497.00 8 135 497.00 8 135 461.03 9 140 497.00 9 140 497.00 9 140 464.89 10 145 497.00 10 145 497.00 10 145 469.23 11 150 497.00 11 150 497.00 11 150 474.05 12 150 566.22 12 150 520.51 12 150 474.05 13 155 571.08 13 155 525.62 13 155 479.45 14 160 576.45 14 160 531.27 14 160 485.44 15 165 582.38 15 165 537.56 15 165 492.13 16 170 588.94 16 170 544.51 16 170 499.56 17 175 596.19 17 175 552.25 17 175 507.83 18 180 604.20 18 180 560.79 18 180 517.01 19 185 613.06 19 185 570.29 19 185 527.23 20 190 622.84 20 190 580.79 20 190 538.55 21 195 633.66 21 195 592.44 21 195 551.14 22 200 645.61 22 200 605.31 22 200 565.08 23 205 658.82 23 205 619.60 23 205 580.57 24 210 673.42 24 210 635.39 24 210 597.72 25 215 689.55 25 215 652.89 25 215 616.76 26 220 707.38 26 220 672.23 26 220 637.83 27 225 727.09 27 225 693.67 27 225 661.21 28 230 748.87 28 230 717.36 28 230 687.08 29 235 772.94 29 235 743.60 29 235 715.76 30 240 799.54 30 240 772.61 30 240 747.50 31 245 828.94 31 245 804.73 31 245 782.67 32 250 861.43 32 250 840.23 32 250 821.59 33 255 897.34 33 255 879.53 33 255 864.69 34 260 937.02 34 260 922.97 34 260 912.39 35 265 980.88 35 265 971.04 35 265 965.20 36 270 1029.35 36 270 1024.19 36 270 1023.63 37 275 1082.91 38 280 1142.11 39 285 1207.54 40 290 1279.85 41 295 1359.76 42 300 1448.08 43 305 1545.68 44 310 1653.56 45 315 1772.77 46 320 1904.53 47 325 2050.14 48 330 2211.06 SPSCOMPCURVE_NOV2004.XLS

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Serial No.04-755 Docket Nos. 50-280, 281 APPENDIX B Surry Units 1 and 2 LTOPS Margin Assessment

Surry Units 1 and 2 LTOPS Margin Assessment Cooldown Rate = 0 Deg. F/hr Cooldown Rate = 0 Deg. F/hr Odeg. F/hr Curve minus WCAP-15130, Ri (w/o Uncs & Bias) WCAP-15130, RI (Urcs, Bias, and Mult) PORV Setpolnt Overshoot PORV Setpoint Overshoot Indicated Indicated Indicated Indicated

  • Indicated Indicated Indicated Temperature Pressure Temperature Pressure Temperature Pressure (psi) Temperature Pressure (Deg. F) (psig) (Deg. F) (psig) * (Deg. F) (Deg. F) (psig) 1 80 621.00 1 100 539.00 1 100 142.00 100 397.00 2 85 621.00 2 105 539.00 2 105 142.00 105 397.00 3 90 621.00 3 110 539.00 3 110 142.00 110 397.00 4 95 621.00 4 115 539.00 4. .115 142.00 115 397.00 5 100 621.00 5 120 539.00 5 120 142.00 120 397.00 6 105 621.00 6 125 539.00 6 125 142.00 125 397.00 7 110 621.00 7 130 539.00 7 130 142.00 130 397.00 8 115 621.00 8 135 539.00 8 135 142.00 135 397.00 9 120 621.00 9 140 539.00 9 140 142.00 140 397.00 10 125 621.00 10 145 539.00 10 145 142.00 145 397.00 11 130 621.00 11 150 539.00 .11 150 142.00 150 397.00 13 135 695.08 13 155 613.08 13 155 130.90 155 482.18 14 140 700.45 14 160 618.45 14 160 130.90 160 487.55 15 145 706.38 15 165 624.38 15 165 130.90 .165 493.48 16 150 712.94 16 170 630.94 16 170 130.90 170 500.04 17 155 720.19 17 175 638.19 17 175 130.90 175 507.29 18 160 728.20 18 180 646.20 18 180 130.90 180 515.30 19 165 737.06 19 185 655.06 19 185 130.90 185 524.16 20 170 746.84 20 190 664.84 20 190 130.90 190 533.94 21 175 757.66 21 195 675.66 21 195 130.90 195 544.76 22 180 769.61 22 200 687.61 22 200 130.90 200 556.71 23 185 782.82 23 205 700.82 23 205 101.80 205 599.02 24 190 797.42 24 210 715.42 24 210 101.80 210 613.62 25 195 813.55 25 215 731.55 25 215 101.80 215 629.75 26 200 831.38 26 220 749.38 26 220 101.80 220 647.58 27 205 851.09 27 225 769.09 27 225 101.80 225 667.29 28 210 872.87 28 230 790.87 28 230 101.80 230 689.07 29 215 896.94 29 235 814.94 29 235 101.80 235 713.14 30 220 923.54 30 240 841.54 30 240 101.80 240 739.74 31 225 952.94 31 245 870.94 31 245 101.80 245 769.14 32 230 985.43 32 250 903.43 32 250 101.80 250 801.63 33 235 1021.34 33 255 939.34 33 255 72.40 255 866.94 34 240 1061.02 34 260 979.02 34 260 72.40 260 906.62 35 245 1104.88 35 265 1022.88 35 265 72.40 265 950.48 36 250 1153.35 36 270 1071.35 36 270 72.40 270 998.95 37 255 1206.91 37 275 1124.91 37 275 72.40 275 1052.51 38 260 1266.11 38 280 1184.11 38 280 72.40 280 1111.71 39 265 1331.54 39 285 1249.54 39 285 72.40 285 1177.14
40. 270 1403.85 40 290 1321.85 40 290 72.40 290 1249.45 41 275 1483.76 41 295 1401.76 41 295 72.40 295 1329.36 42 280 .1572.08 42 300 1490.08 42 300 72.40 300 1417.68 43 285 1669.68 43 305 1587.68 43 305 58.30 305 1529.38 44 290 1777.56 44 310 1695.56 44 310 58.30 310 163726 45 .295 1896.77 45 315 1814.77 45 315 58.30 315 1756.47 46 300 2028.53 46 320 1946.53 46 320 58.30 320 188823 47 305 2174.14 47 325 2092.14 47 325 58.30 325 2033.84 48 310 2335.06 48 330 2253.06 48 330 58.30 330 2194.76 SPS MARGIN CALC NEW.XLS

Surry Units 1 and 2 LTOPS Margin Assessment PORV Setpoint Margin (Positive = Acceptable)

Indicated Indicated Indicated Pressure Minimum Temperature Pressure Temperature (psig) Margin (Deg. F) (psig) (Deg. F) (pi) (psig) 1 100 395 1 100 2.00 2.00 2 105 395 *2 105 2.00 3 110 395 3 110 2.00 4 115 395 4 115 2.00 5 120 395 5 120 2.00 6 125 395 6 125 2.00 7 130 395 7 130 2.00 8 135 395 8 135 2.00 9 140 395 9 140 2.00 10 145 395 10 145 2.00 11 150 395 11 150 2.00 13 155 395 13 155 87.18 14 160 395 14 160 92.55 15 165 395 15 165 98.48 16 170 395 16 170 105.04 17 175 395 17 175 112.29 18 180 395 18 180 120.30 19 185 395 19 185 129.16 20 190 395 20 190 138.94 21 195 395 21 195 149.76 22 200 395 22 200 161.71 23 205 395 23 205 204.02 24 210 395 24 210 218.62 25 215 395 25 215 234.75 26 220 395 26 220 252.58 27 225 395 27 225 272.29 28 230 395 28 230 294.07 29 235 395 29 235 318.14 -

30 240 395 30 240 344.74 31 245 395 31 245 374.14 32 250 395 32 250 406.63 33 255 395 33 255 471.94 34 260 395 34 260 511.62 35 265 395 35 265 555.48 36 270 395 36 270 603.95 37 275 395 37 275 657.51 38 280 395 38 280 716.71 39 285 395 39 285 782.14 40 290 395 40 290 854AS 41 295 395 41 295 934.36 42 300 395 42 300 1022.68 43 305 395 43 305 1134.38 44 310 395 44 310 1242.26 45 315 395 45 315 1361.47 46 320 395 46 320 1493.23 47 325 395 47 325 1638.84 48 330 395 48 330 1799.76 SPS MARGIN CALC NEW.XLS

Serial No.04-755 Docket Nos. 50-280, 281 APPENDIX C Material Properties Basis and RVID Update

Table 1 RVID Update Based on Current Material Properties Facility: Surry Unit 1 Vessel Manufacturer: B&W and Rotterdam Dockyard Best- Best- Assigned Estimate Estimate Material RPV Weld Wire Heat or Copper Nickel ID Fluence Chemistly Method of Initial Inner Surf. ART Material ID Location (wt%) (wt%) (xlE19) Factor Determining CF RT(NDI) Slgmapt) Sigma(detta) Margin or RT(PTS) 114-T ARr 122V109VAt Nozzle Shell Forging 0.110 0.740 0.496 76.1 Tables 40 0.0 17.0 34.0 135.2. 125.2 C4326-1 Intermedlate Shell 0.110 0.550 5.400 73.5 Tables 10 0.0 17.0 34.0 148.2 .140.5 C4326-2 Intermediate Shell 0.110 0.550 5.400 73.5 Tables 0 0.0. 17.0 34.0 138.2 . 130.5 4415-1 Lower Shell 0.102 0.493 5.400 85.0 Surv. Data 20 0.0 8.5 17.0 157.4 . 148.8 4415.2 Lower Shell 0.110 0.500 5.400 . 73.0 Tables 0 0.0 17.0 34.0 137.5 129.8 J726/25017 Nozzle to Int Shell Ctrc Weld 0.330 0.100 0.498 152.0 Tables 0 20.0 28.0 68.8 191.1 171.0 SA-1585/72445 Int. to Low Sh. Clrc (ID 40%) 0.220 0.540 4.700 131.4 Surv. Data .5 19.7 28.0 68.5 246.2 231.7 SA-1650/72445 Int. to Low Sh. CIre (OD 60%) 0.220 0.540 4.700 131.4 Sunr. Data .5 19.7 28.0 68.5 246.2 231.7 SA-149418T1t54 Int Shell Long. Welds L3 &L4 0.160 0.570 0.914 143.9 Tables .5 19.7 28.0 68.5 203.7 183.9 SA4149418T1554 LowerShell Long. Weld LI 0.160 0.570 0.790 143.9 Tables .5 19.7 28.0 68.5 197.9 178.1 SAt1526/299144 Lower Shell Long. Weld L2 0.340 0.680 0.790 220.8 Tables .7 20.6 28.0 69.5 268.5 238.2 1/4-T ARTvalue of 238.2 F was used In the determination of P/T flmits Note: Shaded cells indirate a changed value relative to Dominions most recent update to the NRCs ReactorVessel lntegrtty Database (RVID) (Last Update on 3/27/03j.

Facility: Surry Unit 2 Vessel Manufacturer: B&W and Rotterdam Dockyard Best- Best- Assigned Estimate Estimate Material RPV Weld Wire Heal or Copper Nickel , ID Fluence Chemistry Method of InitIal , Inner Surf. ART Material ID Location (w%/.) (wt%) (x1 E19) Factor Determining CF RT(NDT) Sigma(l) Sigma(defta) Margin or RT(PTS) 1/4-T ART 123V303VA1 Nozzle Shell Forging 0.110 0.720 0.471 75.8 Tables 30 0.0 17.0 34.0. 123.9 114.0 C4331-2 Intermediate Shell 0.120 0.600 5.340 83.0 Tables -10 0.0 17.0 34.0 141.5 132.7 C4339-2 Inlermediale Shell 0.110 0.540 5.340 73.4 Tables .20 0.0 17.0 34.0 117.9 110.2.

C4208 2 Lower Shell 0.150 0.550 5.340 107.3 Tables -30 0.0 17.0 34.0 155.8 144.5 C4339-1 Lower Shell 0.107 0.530 5.340 70.8 Tables .10 0.0 17.0 34.0 124.2 118.8 L737/4275 Nozzle to Int Shell Clrc Weld 0.350 0.100 0.471 160.5 Tables 0 20.0 28.0 68.8 195.7 174.6 R300810227 Int. to Lower Shell Cir Weld 0.187 0.545 5.340 132.4 Surv. Data 0 20.0 14.0 48.8 238.2 222.3 WF.4/8T1762 Int. Shell Long. L4 (ID 50%) 0.190 0.570 1.080 152.4 - . Tables .5 19.7 28.0 68.5 219.1 198.0 SA.1585t72445 Int. Sh. L3 1100%). L4 IOD 50) 0.220 0.540 1.080 131.4 Surv. Data .5 19.7 28.0 88.5 197.7 179.5 WF-4t8T1762 LS L2 (ID63%). 1 (100) 0.190 0.570 1.080 152.4 Tables -5 19.7 28.0 68.5 219.1 198.0 WF-818T1762 LS Long. Weld L2 (OD 37%) 0.190 0.570 1.080 152.4 Tables -5 19.7 28.0 68.5 219.1 198.0 114-T ART value of 238.2 F was used In the determination of P/T limits Note: Shaded cells indicate a changed value relative to Dominion's most recent update to the NRCs Reactor Vessel Integrity Database (RVID) (Last Update on 3/27/03).

Table 2 RVID Update Based on Material Prtpertles as Modified by BAW-2308, Rev. 1 Facility: Surry Unit 1 Vessel Manufacturer: B&W and Rotterdam Dockyard Best- Best- Assigned Estimate Estimate Material RPV Weld Wife Heat or Copper Nickel ID Fluence Chemistry Method of Initial Inner Surf. ART Material ID Location (wt%) (wt%) (xtE19) Factor Determining CF RT(NDT) Sigmail) Slgma(delta) Margin or RT(PTS) 114-T ARF 122V109VAi Nozzle Shen Forging 0.110 0.740 0.496 78.1 Tables 40 0.0 17.0 34.0 135.2 125.2 C4326-t Intermediate Shell 0.110 0.550 5.400 73.5 Tables 10 0.0 17.0 34.0 148.2 140.5 C4.i26-2 Intermediate Shel 0.110 0.550 5.400 73.5 Tables . 0 0.0 17.0 34.0. 138.2. 130.5 4415.1 Lower Shen 0.102 0.493 5.400 85.0 Surv. Data 20 0.0 8.5 17.0 157.4 148.6 4415.2 LowerShell 0.110 0.500 5.400 73.0 Tables 0 0.0 17.0 34.0 137.5 129.8 J726125017 Nozzle to int Shen Circ Weld 0.330 0.100 0.496 152.0 Tables 0 20.0 28.0 68.8 191.-. 171.0 SA-1585r72445 int. to Low Sh. Circ (ID 40%) 0.220 0.540 4.700 131.4 Surv. Data -732 11.9 28.0 80.8 170.3 155.9 SA-1650172445 Int. to Low Sh. Circ (OD 60%) 0.220 0.540 4.700 131.4 Surv. Data -73.2 11.9 28.0 60.8 . 170.3 155.9 SA-149418T1554 Int Shell Long. Welds L3 A L4 0.160 0.570 0.914 143.9 Tables -67.9 20.1 28.0 68.9 141.3 121.4 SA.149418T1554 Lower Shen Long. WeldL1 0.160 0.570 0.790 . 143.9 Tables 47.9 . 20.1 28.0 68.9 135.4 115.7 SA-15261299L44 Lower Shen Long. Weld L2 0.340 0.680 0.790 220.6 Tables *78.8 12.0 28.0 60.9 188.1 157.8 114-T ART value d 238.2 F was used In the determination of P/T limits Note Shaded cells indicate a changed value relative to Domlnion's most recent update to the NRC's Reactor Vessel Integrity Database (RVID) (Last Update on 312i103).

Facility: Surry Unit 2 Vessel Manufacturer: B&W and Rotterdam Dockyard Best- Best- Assigned Estimate Estimate Material RPV Weld Wire Heat or Copper Nickel ID Fluence Chemistry Method of Initial Inner Surf. ART Materlal ID Location (wt%) (wt%) (xl E19) Factor Determining CF RT(NDT) Sigma(l) Sigma(delta) Margin or RT(PTS) 1t4-TART' 123V303VA1 Nozzle Shen Forging 0.110 0.720 0.471 75.8 Tables 30 0.0 17.0 34.0 123.9 114.0 C4331-2 Intermediate Shell 0.120 0.600 5.340 83.0 Tables -10 0.0 17.0 34.0 141.5 132.7 C4339-2 Intermedlate Shenl 0.110 0.540 5.340 73.4 Tables -20 0.0 17.0 34.0 117.9 110.2 C4208-2 Lower Shell 0.150 0.550 5.340 107.3 Tables -30 0.0 17.0 34.0 155.8 144.5 C4339-1 LowerShell 0.107 0.530 5.340 70.8 Tables .10 0.0 17.0 34.0 124.2 118.8 L73714275 Nozzle to Int Shell Circ Weld 0.350 0.100 0.471 160.5 Tables 0 20.0 28.0 68.8 1.95.7 . 174.8 R300810227 Int. to Lower Shen Circ Weld 0.187 0.545 5.340 132.4 Surv. Data 0 20.0 14.0 48.8 . 238.2 222.3 WF-418T1762 il. Shell Long. L4(PO50%) . 0.190 _0570 1.080 152.4 Tables. 67.9 20.1 . 28.0 68.9 .* 158.7 - 135.6 SA-1585172445 Int. Sh. L3 (100%). L4 (0D 50) 0.220 0.540 1.080 131.4 Sutv. Data *732 11.9 28.0 60.8 121.9 1037 WF-418T1762 LS L2 (ID 63%). L1(100) 0.190 0.570 1.080 152.4 Tables l 7.9 20.1 - 28.0 88.9 158.7 135.6 WF-818T1762 LS Long. Weld L2 (OD 37%) 0.190 0.570 1.060 152.4 Tables 4-7.9 20.1 28.0 88.9 158.7 135.8 114-T ART value of 2382 F was used In the determination of P/T limits Note: Shaded cells Indicate a changed value relative to Dominion's most recent update to the NRC's Reactor Vessel Integrity Database (RVID) (Last Update on 3127103).

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