ML18153A835

Requests Exemption from Certain Requirements of 10CFR50.60 in Form of Change to Ts,Incorporating Revised Pressure/Temp Limits & Ltops Setpoint That Will Be Valid to end-of- License

ML18153A835
Person / Time
Site:	Surry
Issue date:	06/08/1995
From:	Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18153A836	List: ML18153A835 ML18153A837
References
95-197, NUDOCS 9506120335
Download: ML18153A835 (30)

Text

..--*-

\,

• ,r

.1-.

e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 June 8, 1995 United States Nuclear Regulatory Commission Serial No.95-197 Attention: Document Control Desk NL&P/GDM RO Washington, D. C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 Gentlemen:

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 SETPOINT Pursuant to 10 CFR 50.12 and 50.90, Virginia Electric and Power Company requests an exemption from certain requirements of 10 CFR 50.60 in the form of a change to the Technical Specifications to Facility Operating Licenses Nos. DPR-32 and DPB-37 for Surry Power Station Units 1 and 2, respectively. The proposed Technical Specifications change incorporates revised pressure/temperature limits and an associated Low Temperature Overpressure System (LTOPS) setpoint that will be valid to the end-of-license (28.8 EFPY and 29.4 EFPY for Units 1 and 2, respectively). The proposed change also incorporates analytical and operational features into the Surry design basis that provide additional pressure/temperature (PIT) operating margin and reduce the potential for an undesired power operated relief valve lift during reactor coolant pump startup. *This request also updates the unirradiated reactor vessel material toughness data presented in the Technical Specifications to reflect the data previously provided to the NRG in our response to Generic Letter 92-01, Revision 1, "Reactor Vessel Structural Integrity."

Implementation of this Technical Specification change requires an exemption from certain requirements of 10 CFR 50.60, "Acceptance criteria for fracture prevention measures for light-water nuclear power reactors for normal operation." This exemption is requested to allow the application of American Society of Mechanical Engineers (ASME) Code Case N-514, "Low Temperature Overpressure Protection," in determining the acceptable low temperature overpressure protection system (LTOPS) setpoint for Surry Power Station Units 1 and 2. The attached proposed Technical Specifications change and exemption request provide a detailed discussion* and technical justification for the application of ASME Code Case N-514 to the LTOPS setpoint for Surry Power Station.

In accordance with 10 CFR 50.55a, ASME Code Cases must be approved for use by the NRG. NRG Regulatory Guides, 1.84, 1.85 and 1.147 list the ASME Code Cases is~

p.

61

~g5g~ .09550608 0002so

• ~

PDR. .

that have been approved by the NRC. Code Case N-514 has not been added to these Regulatory Guides to date, although it has been previously approved for use at several other facilities. Code Case N-514 is discussed in the attachment to SECY-94-267, Section 4.5.2, "Status of Low-Temperature Overpressure Protection Limits Issue."

That discussion concludes with the statement "The content of Code Case N-514 has been incorporated into Appendix G of Section XI of the ASME Code and published in the 1993 Addenda to Section XI. The NRC staff is currently developing a revision to 10 CFR 50.55a that will endorse the 1993 Addenda and Appendix G of Section XI into*

the regulations." Accordingly, pursuant to 10 CFR 50.55a(a)(3),.the use of Code Case N-514 for Surry Power Station is hereby requested.

The proposed Technical Specifications change has ,been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Management Safety Review Committee. It has been determined that the proposed Technical Specifications change poses an unreviewed safety question as defined in 10 CFR 50.59 and requires NRC approval prior to implementation, since the ASME Section XI Code Case N-514 has not been generically approved for industry use by the NRC.

The proposed Technical Specification change has also been reviewed in accordance with 10 CFR 50.92, and it has been determined that the change does not pose a significant hazards consideratiO{l.

The existing Surry Units 1 and 2 pressure/temperature operating limits, and the associated LTOPS setpoint, are valid to a cumulative core burnup of 15 Effective Full Power Years (EFPY), which Unit 1 is expected to reach in February 1996.

Consequently; approval of the proposed change is requested by January 1996 to support implementation of the revised pressure/temperature operating limits and LTOPS setpoint prior to the expiration of the current limits.

If you have any questions or require additional information, please contact us.

Very truly yours, James P. O'Hanlon Senior Vice President - Nuclear Attachments cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.

Suite 2900 Atlanta, Georgia 30323 Mr. M. W. Branch NRC Senior Resident Inspector Surry Power Station

I I

COMMONWEALTH OF VIRGINIA )

)

COUNTY OF HENRICO )

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by R. F. Saunders, who is Vice President - Nuclear Operations, for J. P. O'Hanlon who is Senior Vice President - Nuclear, of Virginia Electric and Power Company. He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.

Acknowle;~~~. before me this Jl!l day of ----=g,,..,..,..t-',0~N-"'-' 19~.

My Commission Expires: \/'haey 3/ , 1931_.

~'ll!wL, Notary Public (SEAL)

ATTACHMENT 1 Discussion and Technical Justification 10 CFR 50.60 Exemption Request, ASME Code Case Approval and Proposed Technical Specifications Change Revised Pressure/Temperature Operating Limits and LTOPS Setpoint Surry Units 1 and 2 I

I

1.0 INTRODUCTION

Virginia Electric and Power Company proposes changes to the Surry Units 1 and 2 Technical Specifications to provide operating limi_ts, setpoints, and component operability requirements that ensure reactor vessel integrity during normal operation and postulated accident conditions. The existing Surry Units 1 and 2 pressure/temperature operating limits and Low Temperature Overpressure Protection System (LTOPS) setpoint are valid to a cumulative core burnup of 15 Effective Full Power Years (EFPY), which Unit 1 is expected to reach in February 1996. The proposed Units 1 and 2 limits and associated LTOPS setpoint will be valid to end-of-license (28.8 EFPY and 29.4 EFPY, respectively).

The safety evaluation demonstrates that the proposed reactor vessel protection philosophy and the associated pressure/temperature limits, LTOPS setpoint, and component operability requirements ensure reactor vessel integrity will be maintained during normal operation and design basis accident conditions. SpecificaMy, adherence to the heatup/cooldown rate-dependent pressure/temperature operating limits ensures that the assumed design basis flaw will not propagate during normal operation. Below the LTOPS enabling temperature, automatic actuation of the PORVs ensures that the assumed design basis flaw will not propagate under design basis low-temperature overpressurization accident conditions. Above the enabling temperature, two pressurizer safety valves are sufficient to relieve the overpressurization due to the inadvertent startup of two charging pumps at water solid conditions without propagation of the assumed design basis flaw.

The proposed Technical Specification change accommodates the predicted increase in the nil ductility transition reference temperature (RTNDT) due to irradiation resulting from operation through the end-of-license. The proposed changes incorporate the provisions of ASME Code Case N-514 (2), which is applicable to Surry. The Code Case permits use of 110% of the limit which satisfies the requirements of 10 CFR 50 Appendix G as the design limit for establishing the LTOPS setpoint. The Code Case also permits use of an LTOPS "enabling temperature" equal to the limiting material's RTNDT + 50°F. Although the NRC has previously approved use of Code Case N-514 in other utility submittals (3),(4), use of the Code Case has not been generically approved for industry use. Therefore, its application to Surry creates an unreviewed safety question due to a reduction in a margin of safety. Because the proposed Technical Specifications change is supported by Code Case N-514, this reduction is not significant and does not create a significant hazard consideration as defined by 10 CFR 50.92. However, an exemption permitting the use of the Code Case is required prior to implementation of the proposed operating pressure/temperature limits and LTOPS setpoint.

1

2.0 BACKGROUND INFORMATION The Surry Units 1 and 2 Reactor Coolant Systems (RCS) are protected from material failure by the imposition of restrictions on allowable pressure and temperature, and on heatup and cooldown rate. The Low Temperature Overpressure Protection System (LTOPS) ensures that material integrity limits are not exceeded during the design basis overpressurization accidents. Equipment operability requirements are imposed to ensure that the assumptions of the accident analyses remain valid.

Regulatory Guide 1.99, Revision 2 (1) provides guidance for determination of the RTNDT for reactor pressure vessel steel. RG 1.99, Revision 2 facilitates the calculation of RTNDT values either by correlation based on neutron fluence, copper content, and nickel content, or by utilization of surveillance specimen test results. The fracture resistance of reactor pressure vessel steel is indexed to RTN DT in Section XI, Appendix G of the ASME Code, facilitating the calculation of pressure/temperature operating limits which ensure that combined pressure and thermal stresses do not exceed the fracture resistance of the most limiting reactor vessel beltline material. The LTOPS setpoint is determined such that the maximum pressure predicted to occur during a mass or heat addition transient will not exceed the ASME Section XI Appendix G pressure/temperature limits. The LTOPS system is required to be active when RCS temperature is below the LTOPS enabling temperature, unless a bubble has been drawn in the pressurizer to provide adequate time for operators to correct an inadvertent mass or heat addition to the RCS, or unless the RCS is vented.

Detailed descriptions and technical justification for the exemption request and proposed Technical Specifications change are presented in the sections that follow.

2

i 3.0 SPECIFIC CHANGES The Technical Specifications change described herein applies to Surry Units 1 and 2.

In addition to the specific changes described below, editorial changes have been made to correct grammatical errors and format inconsistencies.

Technical Specification 3.1.A.2 - Steam Generator TS 3.1.A.2 has been modified to make the word "loop" plural, as the specification originally intended. This change is administrative in nature.

Technical Specification 3.1.B, Tables 3.1-1 and 3.1-2, and Figures 3.1-1 through 3.1 Heatup and Cooldown TS 3.1.B.1 has been changed to indicate that revised TS Figure 3.1-1 is valid for heatup rates up to 60°F/hr.

TS Tables 3.1-1 and 3.1-2 present unirradiated reactor vessel material toughness data for Surry Units 1 and 2. The Surry Units 1 and 2 reactor vessel materials design basis was reconstructed to support the Virginia Electric and Power Company response to Generic Letter 92-01, Revision 1, "Reactor Vessel Structural Integrity," dated March 6, 1992. Therefore, TS Tables 3.1-1 and 3.1-2 have been updated to reflect the data presented in the Virginia Electric and Power Company Response (5) to Generic Letter 92-01 (6).

Revised TS Figures 3.1-1 and 3.1-2 have been prepared to present the revised Unit 1, 28.8 EFPY and Unit 2, 29.4 EFPY pressure/temperature operating limit data (7). The curves have been modified to include a correction for the pressure difference between the point of measurement (i.e., hot leg or pressurizer) and the point of interest (i.e., the reactor vessel beltline), but do not include allowances for temperature and pressure measurement uncertainty. Instrumentation uncertainties are accommodated by ASME Section XI Appendix G margins. The 10 CFR 50 Appendix G criticality limit line has been excluded in favor of the more restrictive Technical Specification 3.1.E.4, Minimum Temperature for Criticality.

• Figures 3.1-3 and 3.1-4 are being deleted. No operational limitations or provisions are established by these figures, nor do the figures substantively contribute to the technical basis for TS 3.1 .B.

The Basis for TS 3.1.B has been modified to appropriately describe the development of the pressure/temperature operating limits presented in TS Figures 3.1-1 and 3.1-2.

Technical Specification 3.1.E.2 Minimum Temperature for Criticality The acronym "OTT," or "ductility transition temperature," in TS 3.1.E.2 and TS 3.1.E/Basis has been replaced with the phrase "the limiting value of RTNoT" to be consistent with terminology used in Technical Specifications and in applicable 3

regulatory guidance. The reference to "E-1" (TS 3.1.E.1) in TS 3.1.E.3 has been corrected to refer to "E-2" (TS 3.1.E.2). These changes are administrative in nature.

Technical Specification 3.1.G Reactor Coolant System Overpressure Mitigation The Power Operated Relief Valve (PORV) lift setpoint presented in TS 3.1.G.1.b.3 has been changed from ::;;385 psig to ::;;390 psig. The LTOPS enabling temperature of TS 3.1.G.1.b remains unchanged at 350°F. The proposed LTOPS setpoint and existing enabling temperature ensure bounding low temperature reactor vessel integrity protection during the postulated design basis mass and heat addition transients.

Provisions for venting the RCS or establishing a bubble in the pressurize'r (as alternatives to ensuring that two PORVs are operable) remain unchanged.

Grammatical changes have been made to TS 3.1.G.1.b to clarify existing requirements.

4

4.0 SAFETY SIGNIFICANCE This section presents a safety evaluation which supports the proposed changes to the Surry Units 1 and 2 Technical Specifications. The information presented in each section is summarized below.

Section 4.1 The reactor vessel material surveillance capsule results which support the proposed changes were derived from the Surry Units 1 and 2 plant-specific surveillance programs (8),(9) and the B&W Owners Group Master Integrated Reactor Vessel Surveillance Program (MIRVSP) (10). A summary of the plant-specific and MIRVSP capsules used in the determination of chemistry factors for RTN DT calculations is presented in Section 4.1 .

Section 4.2 Section 4.2 discusses the reactor vessel materials data used in calculations of irradiated RTNDT and RTPTS for Surry Units 1 and 2.

Section 4.3 Section 4.3 discusses the overpressurization analysis which supports the proposed Technical Specification LTOPS setpoint.

Section 4.4 The proposed Surry Units 1 and 2 pressure/temperature operating limits are discussed in Section 4.4.

Section 4.5 A description of the development and evaluation of the revised LTOPS setpoint and LTOPS enabling temperature applicable to both Surry units is presented in Section 4.5.

Section 4.6 Component operability requirements which ensure that the assumptions of the mass and heat addition events remain valid are described in Section 4.6.

5

4.1 Surry 1 and 2 and MIRVSP Surveillance Capsule Results Credible surveillance data for Surry Units 1 and 2 beltline materials are available from the plant-specific surveillance programs (8),(9) and the B&W Owners Group Master Integrated Reactor Vessel Surveillance Program (10). A summary of the surveillance capsules, from which data has been derived and utilized in Regulatory Guide 1.99, Revision 2 (1) RTNDT calculations, is presented below. With the exception of SA-1526, each material's calculated chemistry factor is identical to that presented in the.

Reference (5) Generic Letter 92-01 response. The calculated chemistry factor for SA-1526 is slightly higher than the Reference (5) value. The chemistry factors presented below were used in the Reference (7) pressure/temperature limits analysis.

Material: Surry Unit 1 C4415-1 (Lower Shell Plate)

Surry Unit 1 Capsule T Surry Unit 1 Capsule V Resulting CF: 89.2 Material: Surry Unit 1 SA-1526 (Lower Shell Axial Weld L2)

TMI 1 Capsule E (WF-25)

TMI 1 Capsule C (WF-25)

B&WOG Capsule CR3-LG1 (WF-25)

Surry Unit 1 Capsule V Surry Unit 1 Capsule T Resulting CF: 221.3 Material: Surry Unit 1 SA-1585 (Intermediate to Lower Shell Circumferential Weld)

B&WOG Capsule CR3-LG1 Point Beach 1 Cap. V (SA-1263)

Point Beach 1 Cap. S (SA-1263)

Point Beach 1 Cap. R (SA-1263)

Point Beach 1 Cap. T (SA-1263)

Resulting CF: 149.2 Material: Surry Unit 2 C4339-1 (Lower Shell Plate)

Surry Unit 2 Capsule X Surry Unit 2 Capsule V Resulting CF: 68.4 Material: Surry Unit 2 R3008 (Intermediate to Lower Shell Circumferential Weld)

Surry Unit 2 Capsule X Surry Unit 2 Capsule V Resulting CF: 128.0 6

4.2 Surry Units 1 and 2 Reactor Vessel Materials Data and End-of-License RTN DT Values Adjusted RTNDT values have been calculated for each Surry Unit 1 and 2 reactor vessel beltline material in accordance with the methods prescribed by Regulatory Guide 1.99, Revision 2 (1). The limiting end-of-license RTNDT value occurs in the Unit 1 Intermediate to Lower Shell Circumferential Weld. A summary of the Surry Units 1 and 2 reactor vessel material properties (including beltline material chemistry, neutron fluence, and unirradiated RTNoT) and the calculated end-of-license RTNDT values are presented in Tables 1 and 2 of Reference (7). The fluence, chemistry, and unirradiated RTNDT data is identical to that presented in the response to Generic Letter 92-01 for Surry Units 1 and 2 (5). End-of-license fluence estimates are based on the Surry Units 1 and 2 Capsule V analyses (11 ),(12). These fluences are also documented in Reference (13).

4.3 Overpressurization Analysis Low temperature overpressurization protection is provided to ensure that the combined pressure and thermal stresses experienced during a design basis

. overpressurization accident remain well below those which could result in vessel fracture. The PORV setpoint is based on the analysis of two design basis accidents:

the inadvertent startup of a charging pump and the startup of a reactor coolant pump in an RCS loop with a 50°F difference between the steam generator secondary fluid temperature and the RCS temperature. Only one PORV is assumed to operate during the transients.

A two-loop RETRAN02/MOD03 (14) model was developed to analyze possible setpoints. (Information supporting the use of the RETRAN Code version RETRAN02/MOD03 is presented in Attachment 3 to Reference (15).) The overpressurization analysis results revealed that the mass addition transient produces the most limiting results. The following sections describe the inputs to the RETRAN (14) model and the analysis to determine the new PORV setpoint.

4.3.1 Mass Addition Transient The inadvertent startup of a single charging pump was selected as the design basis mass addition transient based on previous UFSAR work (Reference (16), Section 4.3.4). Because of the valve opening characteristic associated with the air-operated relief valves used on the pressurizer (17),(18), the inadvertent startup of a charging pump at water-solid conditions results in pressurization beyond the PORV lift setpoint.

The objective of the analysis was to determine the extent to which RCS pressure exceeded the pressurizer PORV lift setpoint following inadvertent startup of a charging pump during water-solid operation.

The effects of pressure measurement location were explicitly considered in the overpressurization analysis. Specifically, pressurizer PORV actuation was based on hot leg pressure in the RETRAN model. The "PORV lift setpoint overshoot" was 7

defined as the difference between the maximum reactor vessel beltline pressure and the PORV lift setpoint.

The mass addition analysis was performed at the initial conditions listed in the table below. The initial RCS temperature, pressure, and PORV setpoint were varied to observe the effects of changes in these parameters. A range of RCS temperatures between 100°F and 325°F was examined, as well as a range of initial pressures. The analysis revealed a gradually decreasing PORV lift setpoint overshoot with increasing initial RCS temperature and PORV setpoint. The peak RCS pressure was found to be relatively insensitive to the initial RCS pressure. The proposed PORV lift setpoint (Section 4.5.6) was validated by adding the PORV lift setpoint overshoot values to the proposed lift setpoint at each temperature, and verifying that the resulting pressures did not violate the design pressure/temperature limit curve. Selection of the design pressure/temperature limit curve is discussed in Section 4.5.

Reactor Coolant Temperature (°F) 100,150,200,250,300,325 Reactor Coolant Pressure (psig) 200,250,300,340,380,400 Maximum Charging Pump Flow Rate 705 gpm (Bounds the design basis flow vs. head curve)

Pressurizer Steam Volume 0 tt3 Reactor Coolant System Flow 10%

PORV OPEN Setpoint Variable PORV Closed Setpoint OPEN - 15 psi Initial Conditions for the Mass Addition Transient 4.3.2 Heat Addition Transient The heat addition transient assumes that a reactor coolant pump (RCP) is started with the maximum temperature difference allowed by Technical Specifications (50°F}

between the steam generators and the RCS. This scenario has been determined to be the design basis heat addition transient for LTOPS setpoint

• determination (Reference (16), Section 4.3.4).

The heat addition transient was modelled assuming the initial conditions listed in the table below. The secondary-to-primary heat transfer modelling included a very conservative evaluation of the local secondary side convection heat transfer coefficient, and an assumed constant bulk secondary side temperature (i.e., no credit was taken for decreasing temperature due to secondary-to-primary heat transfer). The pump startup flow characteristic was also modelled in a conservative fashion. The analysis revealed that the results of the heat addition transient are easily bounded by those of the mass addition transient.

8

~

~

Rea ctor Coolant Temperature 1oo*F Rea ctor Coolant Pressure (psig) 280,340 RC S/SG ~T 50°F Pre ssurizer Steam Volume 0 ft3 RC P Speeds In Affected Loop, startup 10% - 100%

In Unaffected Loop, coastdown 10% - 0%

PO RV Open Setpoint Variable PO RV Closed Setpoint OPEN - 15 psi Initial Conditions for the Heat Addition Transient 9

i 4.4 Revised Technical Specification Pressure/Temperature Operating Limits Surry Units 1 and 2 pressure/temperature operating limits valid to end-of-license (28.8 EFPY and 29.4 EFPY tor Units 1 and 2, respectively) are documented in Reference (7).

Heatup rates of 20°F/hr, 40°F/hr, and 60°F/hr, and cooldown rates of 0°F/hr (steady-state), 20°F/hr, 40°F/hr, 60°F/hr, and 1OQ°F/hr were considered.

The criticality limit required by 10 CFR 50 Appendix G is not included with the proposed pressure/temperature operating limits, since Limiting Condition tor Operation (LCO) 3.1.E.4 defines a minimum temperature tor criticality that is substantially more limiting than the criticality limit required by 10 CFR 50, Appendix G.

LCO 3.1.E.4 restricts the lowest operating loop average temperature to ~522°F at criticality.

The proposed pressure/temperature operating limits include a correction tor the effects of pressure measurement location. Specifically, the allowable pressures have been reduced to compensate tor the difference between the point of measurement (i.e., the hot leg or pressurizer) and the point of interest (i.e., the reactor vess~I beltline). The pressure/temperature limits do not include instrumentation uncertainties, since these uncertainties are insignificant when compared to the margin terms included in the ASME Section XI Appendix G methods (i.e., 2.0 multiplier on pressure stress). This approach was taken in the most recent LTOPS Technical Specification amendment tor North Anna Units 1 and 2 (19),(20), as well as in other utility submittals.

The LTOPS setpoint defines the operating space at all temperatures (below 350°F) and all cooldown rates with the exception of the 100°F/hr cooldown curve below approximately 190°F, and the 60°F/hr cooldown curve below approximately 130°F.

Because an administrative upper limit on heatup and cooldown rate of 50°F/hr is prescribed tor all temperatures, and because a 60°F/hr (much less, a 100°F/hr) cooldown cannot be sustained at low temperatures, the 100°F/hr and 60°F/hr cooldown curves may be eliminated from consideration below 190°F. The operating space is thereby defined by the LTOPS setpoint tor both Surry units. Therefore, it will only be necessary to verify that the 50°F/hr administrative limit on heatup and cooldown rate is maintained at all temperatures under the revised analysis.

10

- - J

'I' 4.5 LTOPS Design 4.5.1 Industry Experience The NRC's Value/Impact and Regulatory Analyses of Generic Issue 94, "Additional Low Temperature Overpressure Protection for Light. Water Reactors" (21 ),(22}

demonstrate that LTOPS events have historically occurred at essentially isothermal metal conditions. This service experience provides one component of a technical basis for using 110% of the isothermal ASME Section XI limit curve to establish the LTOPS setpoint.

An evaluation of the overpressurization data before and after 1980 revealed that the frequency of overpressurization events (per reactor year) and the severity of overpressurization events were significantly reduced by the* 1979 requirement that each plant install an overpressure mitigation system (OMS) to intercept overpressurization events and prevent them from exceeding the Technical Specifications pressure-temperature limit curves. Operating experience after 1986 indicated a further reduction in challenges to LTOP systems. Between January 1980 and December 1986, 63 PWRs logged 356 reactor years of commercial operation.

During this period there were 30 OMS challenges, driven by either the addition of mass or heat energy, or a combination of the two. Evaluation of the 30 OMS challenge events revealed that none were initiated during or immediately following significant heatup or cooldown. These results follow expectations, since pressure and temperature are not changed rapidly during low temperature operation at conditions where the OMS is enabled.

The fraction of operating time during which significant thermal stresses (e.g., those associated with a >20°F/hr heatup or cooldownf are present is small. For example, a nuclear unit may be expected to heat up and cool down 4 times per year. Assuming a 20°F/hr heatup or cooldown rate, a 400°F temperature change requires 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

Therefore, a plant may be conservatively estimated to spend 160 hour0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br />s/year with the thermal stresses associated with a 20°F/hr ramp rate. This duration represents only 1.8% of plant operating time (e.g., 160 / (365*24)). Consideration of routine low-temperature transients of short duration, such as those resulting from RCP startup or trip, would not significantly affect this estimate.

Industry experience and engineering evaluation support the conclusion that reliable ov~rpressurization protection is provided by an OMS designed to prevent pressure at the reactor vessel beltline from exceeding the isothermal (0°F/hr) pressure/temperature limit curve. This conclusion was affirmed in the NRC's Safety Evaluation Report of Wisconsin Electric and Power Company's license amendment request for modification of Technical Specifications related to the Point Beach Units 1 and 2 Overpressure Mitigating System (OMS) (23): "The zero degree heatup curve is allowed since most pressure transients occur during isothermal metal conditions."

Likewise, the isothermal limit curve was utilized in the most recent LTOPS Technical Specification amendment for North Anna Units 1 and 2 (19),(20}.

• 11

4.5.2 ASME Section XI Recommendations for LTOPS The ASME Section XI Working Group on Operating Plant Criteria (WGOPC), which has responsibility for Appendix G to Section XI, considered the burden and safety impact imposed by regulatory requirements for LTOP, and developed Code guidelines for determining the LTOP setpoint pressure and the required LTOPS enabling temperature (2):

LTOP systems shall be effective at coolant temperatures less than 200°F or at coolant temperatures corresponding to a reactor vessel metal temperature less than RTN D T + 50°F, whichever is greater.1,2 LTOP systems shall limit the maximum pressure in the vessel to 110% of the pressure determined to satisfy Appendix G of Section XI, Article G-2215.

1 The coolant temperature is the reactor coolant inlet temperature 2 The vessel metal temperature is the temperature at a distance one fourth of the vessel section thickness from the inside wetted surface in the vessel beltline region. RTNDT is the highest adjusted reference temperature (for weld or base metal in the beltline region) at a distance one fourth of the vessel section thickness from the vessel wetted inner surface as determined by Regulatory Guide 1.99, Revision 2.

These guidelines relieve some operational restrictions yet provide adequate margins against failure for the reactor pressure vessel. Further, by relieving the operational restrictions, these guidelines result in a reduced potential for activation of pressure relieving devices, thereby improving plant safety.

The philosophy adopted by the WGOPC in considering guidelines for LTOP limits was that administrative controls should be imposed to ensure that the Technical Specification pressure/temperature limits were not exceeded, and that the physical protection system must provide adequate protection against failure of the reactor pressure vessel below the enabling temperature where experience indicates the events occur. The Westinghouse standard methodology for developing LTOPS setpoints (24) endorses use of 110% of the isothermal curve as the LTOPS design limit.

12

4.5.3 Evaluation of ASME Section XI LTOPS Setpoint e

Recommendations Virginia Electric and Power Company has performed independent calculations to estimate the impact of implementing the ASME Section XI L TOPS setpoint recommendations. A 10% increase in the rate-dependent pressure/temperature limits which satisfy the requirements of ASME Section XI Appendix G can be equated to an effective reduction in the nominal factor of safety on applied pressure stresses of 2.0 ensured by ASME Section XI Appendix G methods. Utilization of 110% of the isothermal curve as the design limit for establishing the LTOPS setpoint ensures an equivalent factor of safety of 1.8 on applied pressure stresses for LTOPS events initiated at isothermal conditions.

The operating experience evaluation described in Section 4.5.1 demonstrates that events which challenge the LTOPS setpoint have historically occurred, and may be expected to continue to occur, at essentially isothermal conditions. Therefore, the presence of significant thermal stresses during an LTOPS design basis event is very unlikely. However, even if an LTOPS-challenge event occurs concurrently with a heatup or cooldown ramp rate as high as 50°F/hr, Virginia Electric and Power Company calculations demonstrate that the proposed LTOPS design ensures an equivalent factor of safety in excess of that required by ASME Section XI Appendix G for hydrostatic test conditions (i.e., 1.5).

• On the basis of the foregoing evaluations, Virginia Electric and Power Company proposes use of 110% of the isothermal limit curve to establish LTOPS setpoints. At all operating temperatures, heatup and coolaown rates will be administratively limited to the values assumed in the pressure/temperature limits analysis (7). Further, an administrative maximum upper limit on heatup and cooldown rates of 50°F/hr will be observed to ensure that the instantaneous heatup or cooldown rate does not inadvertently exceed the range of analyzed rates as a result of equipment malfunction or operator error. Adequate margin to vessel fracture is ensured by this design since:

(a) Industry experience has shown that events which challenge the LTOPS setpoint may be expected to occur at essentially isothermal conditions. (Industry experience suggests that the expected frequency of events which challenge the LTOPS system when significant (i.e., non-zero) thermal stresses may be present is on the order of one per 660 reactor years.)

(b) Physically achievable cooldown rates decrease with decreasing temperature.

(The maximum achievable cooldown rate approaches zero at the lowest allowable RCS operating temperature.)

(c) ASME Section XI Code Case N-514 (2) allows LTOPS to limit the maximum pressure in the reactor vessel to 110% of the pressure determined to satisfy Appendix G. The NRC has affirmed use of Code Case N-514 for other utilities (3),(4) ,(28) ,(29}.

(d) The Westinghouse standard methodology for developing LTOPS setpoints (24) endorses use of 110% of the isothermal curve as the LTOPS design limit.

13

e (e) For LTOPS-challenge events beyond the design basis (e.g., events occurring when thermal stresses associated with heatup or cooldown ramp rates up to 50°F/hr are present in the reactor vessel), Virginia Electric and Power Company calculations demonstrate an equivalent factor of safety in excess of that prescribed by ASME Section XI Appendix G for hydrostatic test conditions.

Operational occurrences which violate the rate-dependent Appendix G pressure/temperature limits may be evaluated in accordance with the requirements of ASME Section XI Appendix E.

4.5.4 LTOPS Enabling Temperature Evaluation Previous Virginia Electric and Power Company LTOPS analyses have established the LTOPS enabling temperature at RTNDT + ,1.T + 90°F + temperature measurement uncertainty. (,1.T is the maximum temperature difference between the water and metal at the 1/4-T and 3/4-T locations during heatup or cooldown at the maximum allowable rate.) The ASME Section XI LTOPS recommendations (2) provide for the establishment of the LTOPS enabling temperature at RTNDT + 50°F + temperature measurement uncertainty. Fifty degrees is considered adequate margin to ensure that LTOPS is enabled* at a temperature on the "upper shelf" of the beltline materials.

Above this temperature, ASME Section XI Appendix G margins are sufficient to ensure that pressures up to the RCS design pressure will not result in propagation of the design flaw. Overpressure protection below this temperature is provided by a combination of administrative and procedural controls and automatic PORV actuation.

The Westinghouse standard methodology for calculation of LTOPS setpoints and LTOPS enabling temperatures (24) includes use of RT NOT+ 50°F for establishing the LTOPS enabling temperature.

Adequate water-solid overpressurization protection above the LTOPS enabling temperature is provided by only the passive actuation of the pressurizer safety valves.

Specifically, Virginia Electric and Power Company calculations demonstrate*that two pressurizer safety valves (PSV) can accommodate sufficient flow to compensate for the inadvertent and simultaneous startup of two charging pumps. Each PSV is capable of relieving 288,000 lbm/hr of saturated steam at 2500 psia. The water relief capacity of the PSVs was assumed to be 40% of their steam relief capacity (25). The calculations considered the pressure difference between the reactor vessel beltline and the pressurizer, and a 3% PSV lift setpoint tolerance.

On the basis of the evaluations described above, Virginia Electric and Power Company proposes establishment of the LTOPS enabling temperature at the temperature corresponding to RTNDT + 50°F + temperature measurement uncertainty.

Margin is not added to compensate for the maximum calculated temperature difference between the downcomer fluid and the 1/4-T and 3/4-T reactor vessel locations, since use of the isothermal limit curve as the LTOPS design limit implies a uniform temperature distribution.

14

e 4.5.5 Selection of the LTOPS Enabling Temperature The existing TS 3.1.G.1.b specifies that LTOPS must be enabled at temperatures less than or equal to 350°F, if overpressurization protection is not provided by an RCS vent or if adequate operator response time is not provided by the maintenance of a pressurizer bubble. As demonstrated below, the existing TS 3.1.G.1.b LTOPS enabling temperature need not be modified, since it conservatively bounds the enabling temperature calculated in the manner allowed by ASME Code Case N-514.

ASME Code Case N-514 allows an LTOPS enabling temperature as low as RTNDT +

50°F + instrument uncertainty. The RTNDT for the limiting Surry Unit 1 and 2 material at end-of-license is 228.4°F. A bounding temperature measurement and instrumentation uncertainty of 21 °F is utilized. The calculated enabling temperature of 299.4°F is conservatively rounded up to 300°F. This value is conservatively bounded by the existing TS 3.1.G.1.b LTOPS enabling temperature of 350°F.

4.5.6 Proposed Design On the basis of the foregoing analyses and evaluations, Virginia Electric and Power Company proposes the following PORV lift setpoint and enabling temperature for Surry Units 1 and 2 valid to 28.8 EFPY and 29.4 EFPY, respectively:

Current:

~385 psig for RCS Avg. Temperature T~350°F Proposed:

~390 psig for RCS Avg. Temperature T~350°F The PORV lift setpoint was validated by adding the mass addition transient "setpoint overshoot" (described in Section 4.3.1) to the PORV lift setpoint pressure, and verifying that the resulting pressure is less than the isothermal limit curve.

Pressure and temperature measurement uncertainties have been excluded from consideration in the development of the LTOPS setpoint *on the basis that these uncertainties are insignificant when compared to the margin terms included in the ASME Section XI Appendix G methods. Instrumentation uncertainties have been excluded from consideration in previous submittals made by Virginia Electric and Power Company (19),(20) and other utilities (26),(27).

As of this writing, use of Code Case N-514 has not yet been formally approved by the NRC. In the interim, approval for using Code Case N-514 may be granted by the NRC in accordance with the provisions of 10 CFR 50.12, 50.55a(3} and 50.60(b).

Therefore, Virginia Electric and Power Company requests approval of ASME Code Case N-514 (2), which provides for utilization of 110% of the limit curve which satisfies the requirements of 10 CFR 50 Appendix G, and which supports establishment of the LTOPS enabling temperature at RTNDT + 50°F. Use of Code Case N-514 has been approved by the NRC in other utility submittals (3},(4),(28),(29).

15

4.6 Component Operability Requirements 4.6.1 Charging Pump Operability Requirements To ensure that plant equipment conditions are consistent with the assumptions of the inadvertent charging pump startup accident, it is necessary to require that only one charging pump be capable of automatic initiation at RCS average temperatures below the LTOPS enabling temperature. The existing TS 3.1.G.1.b specifies a maximum of one operable charging pump at RCS average temperatures less than or equal to 350°F. The existing TS 3.1.G.1.b also specifies that LTOPS must be enabled at temperatures less than or equal to 350°F, if overpressurization protection is not provided by an RCS vent or if adequate operator response time is not provided by the maintenance of a pressurizer bubble. Because the LTOPS enabling temperature calculated in accordance with ASME Code Case N-514 .is less than 350°F, the existing TS 3.1.G.1.b charging pump operability requirements and LTOPS enabling temperature are acceptable.

Above the LTOPS enabling temperature, two pressurizer safety valves are capable of relieving the flow from two charging pumps. Therefore, no additional restrictions on charging pump operability need to be implemented at temperatures above the LTOPS enabling temperature.

4.6.2 Reactor Coolant Pump Startup Criterion To ensure that plant operating conditions are consistent with the assumptions of the heat addition accident analysis, Technical Specification 3.1.A.1.f requires the steam generator secondary-to-primary temperature difference to be no greater than 50°F when a reactor coolant pump is started. This requirement is in effect when the cold leg temperature is less than or equal to the LTOPS enabling temperature. Above the LTOPS enabling temperature, overpressurization is adequately mitigated by actuation of two pressurizer safety valves.

16

I 5.0 REGULATORY BASIS FOR SPECIFIC EXEMPTION 10 CFR 50.12 Requirements 10 CFR 50.12 states that the Commission may grant an exemption from requirements contained in 10 CFR 50 provided that: 1) the exemption is authorized by law, 2) the exemption will not result in an undue risk to the public health and safety, 3) the exemption is consistent with the common defense and security, and 4) special circumstances, as defined in 10 CFR 50.12(a)(2), are present. The requested exemption to allow the use of ASME Code Case N-514 for determining the LTOPS setpoint and enabling temperature is authorized by law, will not represent an undue risk to the health and safety of the public, and is consistent with the common defense and security.

1. The requested exemption is authorized by law.

No law exists which would preclude the activities covered by this exemption request. 10 CFR 50.60(b) allows the use of alternatives to 10 CFR 50 Appendices G and H when an exemption is granted by the Commission under 10 CFR 50.12.

2. The requested exemption does not present an undue risk to the public health and safety.

A revised LTOPS setpoint and enabling temperature basis is proposed for Surry Units 1 and 2. The setpoint and enabling temperature were developed to provide bounding low temperature reactor vessel integrity protection during the design basis mass and heat addition transients. The LTOPS setpoint utilized 110% of the isothermal Appendix G curve as a design limit. The validity of this approach is supported by consideration of the conditions at which overpressurization events have been demonstrated to occur, and by an analysis which demonstrates adequate margin to reactor vessel failure for this design. The existing Technical Specification LTOPS enabling temperature was demonstrated to conservatively bound the enabling temperature calculated in accordance with the ASME Section XI guidelines (RTNoT+50°F). Above the LTOPS enabling temperature, passive actuation of two pressurizer safety valves is adequate to ensure reactor vessel integrity during the design basis LTOPS transients.

Restrictions on allowable operating conditions and equipment operability requirements have been established to ensure that operating conditions are consistent with the assumptions of the accident analysis. Specifically, RCS pressure and temperature must be maintained within the heatup and cooldown rate-dependent pressure/temperature operating limits specified in Technical Specifications. An administrative upper limit on heatup and cooldown rate of 50°F/hr will be observed. Restrictions on the number of charging pumps capable of inadvertent startup have been imposed to ensure that the assumptions of the mass addition transient analysis are not invalidated. A 17

,,I(

restriction on the allowable temperature difference between the RCS and steam generator secondary side has been imposed to ensure that the assumptions of the heat addition transient are not invalidated. Therefore, this exemption does not present an undue risk to the public health and safety.

3. The requested exemption will not endanger the' common defense and security.

The common defense and security are not endangered by this exemption request.

4. Special circumstances are present which necessitate the request for an exemption to the regulations of 10 CFR 50.60.

Pursuant to 10 CFR 50.12(a)(2), the NRC will not consider granting an exemption to the regulations unless special circumstances are present. This exemption meets the special circumstances of paragraphs (a)(2)(ii), (a)(2)(iii) and (a)(2)(v) of 10 CFR 50.12. This exemption request, as discussed below, demonstrates that the underlying purpose of the regulation will continue to be achieved [(a)(2)(ii)], would result in undue hardship or other cost that are significant if the regulation is enforced [(a)(2)(iii)] and the exemption will provide only temporary relief from the applicable regulation and the licensee has made good faith efforts to comply with the regulations [(a)(2)(v)].

10 CFR 50.12(a)(2)(ii)

Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.

The basis for the LTOPS setpoint and enabling temperature is to preclude reactor coolant system pressure from exceeding the Appendix G limits when there is a potential for non-ductile failure of the reactor vessel material.

Numerous conservatisms were included in the development of the Appendix G pressure/temperature curve calculations. These include:

• A factor of safety of 2.0 on the primary membrane (pressure) stresses.

• A margin factor applied to the shift in RTNDT of one or two standard deviations by Regulatory Guide 1.99, Revision 2.

• The use of the reference stress intensity curves (KIR) by ASME Section Ill and XI, Appendix G, bounds the dynamic crack initiation and crack arrest*

toughness. Further, the use of reference stress intensity curve bounds the crack initiation fracture toughness (KIC) properties by a factor of 1.2 to 2.5, depending on vessel temperature and RTNDT-

• Lower bound material properties are used in the analysis. Further, increased mechanical properties of the vessel which accompany 18

.r material embrittlement are not considered (increased yield strength and flow stress).

ASME Code Case N-514 recognizes the conservatism of the Appendix G curves and allows setting the LTOPS setpoint such that the ASME Section XI, Appendix G limits are not exceeded by more than 10%, and permits use of an LTOPS "enabling temperature" equal to the limiting material's RTNDT + 50°F.

This allows the implementation of a LTOPS setpoint that preserves an acceptable margin of safety while maintaining operational margins for reactor coolant pump operation at low temperatures and pressures. The LTOPS setpoint established in accordance with ASME Code Case N-514 will also minimize the unnecessary actuation of protection system pressure relieving devices (i.e., power-operated relief valves). Therefore, establishing the LTOPS setpoint in accordance with ASME Code Case N-514 criteria satisfies the underlying purpose of the ASME Code and the NRC regulations to ensure an acceptable level of safety.

10 CFR 50.12(a)(2)(iii)

Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted.

The reactor coolant system pressure/temperature operating window at low temperatures is defined by the LTOPS setpoint. Minimal operating margin is available between the LTOPS setpoint and the pressure experienced at low temperatures due to the startup of a reactor coolant pump, or as a result of normal operating pressure surges with the reactor coolant system in a water solid condition. Implementation of a LTOPS setpoint that is valid from 15 EFPY to the end-of-license without the additional margin allowed by ASME Code Case N-514 would restrict the pressure/temperature operating window and would potentially result in undesired PORV lifts. This constitutes. an unnecessary burden that can be alleviated by the application of ASME Code Case N-514. The guidelines developed by the ASME Section XI Working Group for Operating Plant Criteria (WGOPC) for LTOPS setpoints provide an acceptable margin _of safety against reactor vessel failure. Implementation of a LTOPS setpoint as allowed by ASME Code Case N-514 does not reduce the margin of safety associated with normal operational heatup and cooldown limits. Further, the LTOPS guidelines will reduce the potential for an undesired lift of power operated relief valves (PORV).

10 CFR 50.12(a)(2)(v)

The exemption provides only temporary relief from the applicable regulation and Virginia Electric and Power Company has made a good faith effort to comply with the regulation. We request that the exemption be granted until such time that the NRC generically approves ASME Code Case N-514 for use by the nuclear industry. We are currently in compliance with the requirements of 10 CFR 50.60. However, to retain sufficient pressure/temperature operating 19

margin to the end-of license for Surry Units 1 and 2, we require the exemption to use Code Case N-514.

Basis for NRC Endorsement of Code Case N-514 - 10 CFR 50.55a(a)(3)

Compliance with the specified requirements of 10 CFR 50.55a would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The basis for the LTOPS setpoint is to preclude reactor coolant system pressure from exceeding the Appendix G limits when there is a potential for non-ductile failure of reactor vessel material. ASME Code Case N-514 allows setting the LTOPS actuation setpoint and enabling temperature such that the ASME Section XI Appendix G limits are not exceeded by more than ten percent. This proposed alternative is acceptable because the Code Case recognizes the conservatism of the Appendix G curves and allows establishing a LTOPS setpoint which retains an acceptable margin of safety while maintaining operational margins for reactor coolant pump operation at low temperatures and pressures. As discussed above, the Code Case provides an acceptable margin of safety against crack initiation and reactor vessel failure, and reduces the potential for an undesired PORV lift. Therefore, application of Code Case N-514 for Surry Units 1 and 2 continues to ensure an acceptable level of quality and safety.

20

_j

6.0

SUMMARY

AND CONCLUSIONS The Surry Units 1 and 2 pressure/temperature limits required by 10 CFR 50 Appendix G have been revised to be valid to end-of-license (28.8 EFPY and 29.4 EFPY, respectively) by including the effects of the incremental radiation exposure on the reactor vessel beltline region. The curves are based on analyses of Surry plant-specific and B&W Owners Group Master Integrated Reactor Vessel Surveillance Program surveillance capsule analysis results. The revised Appendix G curves were prepared in accordance with standard Westinghouse methodologies including Regulatory Guide 1.99, Revision 2.

A revised LTOPS setpoint and LTOPS enabling temperature basis are proposed for Surry Units 1 and 2. The setpoint and enabling temperature were developed to provide bounding low temperature reactor vessel integrity protection during the design basis mass and heat addition transients. The LTOPS setpoint utilized 110% of the isothermal Appendix G curve as a design limit. The validity of this approach is supported by consideration of the conditions at which overpressurization events have been demonstrated to occur, and by an analysis which demonstrates adequate margin to reactor vessel failure for this design. The existing Technical Specification LTOPS enabling temperature was demonstrated to conservatlvely bound the enabling temperature calculated in accordance with the ASME Section XI guidelines (RTNoT+50°F}. Above the LTOPS enabling temperature, passive actuation of two pressurizer safety valves is adequate to ensure reactor vessel integrity during the design basis LTOPS transients.

Restrictions on allowable operating conditions and equipment operability requirements have been established to ensure that operating conditions are consistent with the assumptions of the accident analysis. Specifically, RCS pressure and temperature must be maintained within the heatup and cooldown rate-dependent pressure/temperature operating limits specified in Technical Specifications. An administrative upper limit on heatup and cooldown rate of 50°F/hr will be observed.

Restrictions on the number of charging pumps capable of inadvertent startup have been imposed to ensure that the assumptions of the mass addition transient analysis are not invalidated. A restriction on the allowable temperature difference between the

• RCS and steam generator secondary side has been imposed to ensure that the assumptions of the heat addition transient are not invalidated.

21

t

7.0 REFERENCES

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

(2) ASME Boiler and Pressure Vessel Code Case N-514,Section XI, Division 1, "Low Temperature Overpressure Protection," Approval date: February 12, 1992.

(3) Letters from T. C. McMeekin (Duke Power) to USNRC, "McGuire Nuclear Station Units 1 and 2, Docket Nos. 50-369 and 50-370, Request for Exemption -

ASME Code Case N-514," June 28, 1994; August 18, 1994; and September 7, 1994.

(4) Letter from USNRC to T. C. McMeekin (Duke Power), "Exemption from Requirements of 10 CFR 50.60, Acceptance Criteria for Fracture Prevention for Light-Water Nuclear Power Reactors for Normal Operation - McGuire Nuclear Station Units 1 and 2," September 30, 1994.

(5) "Reactor Vessel Working Group Response to Closure Letters to NRC Generic Letter 92-01, Revision 1," BAW-2222 dated June, 1994.

(6) NRC Generic Letter 92-01, "Reactor Vessel Structural Integrity," dated March 6, 1992.

(7) M. J. Malone: Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," WCAP-14177, dated October, 1994.

(8) "Surry Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-7723.

(9) "Surry Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-8085.

(10) "Master Integrated Reactor Vessel Surveillance Program," BAW-1543, Revision 4, dated February, 1993. See also Supplement 1 to BAW-1543, Revision 4, dated February, 1993.

(11) S. E. Yanichko, et al.: "Analysis of Capsule V from the Virginia Electric and Power Company Surry Unit 1 Reactor Vessel Radiation Surveillance Program,"

WCAP-11415, dated February, 1987.

(12) S. E. Yanichko, et al.: "Analysis of Capsule V from the Virginia Electric and Power Company Surry Unit 2 Reactor Vessel Radiation Surveillance Program,"

WCAP-11499, dated June, 1987.

(13) C. C. Heinecke, et al.: "Surry Units 1 and 2 Reactor Vessel Fluence and RTPTS Evaluations," WCAP-11015, Revision 1, dated April, 1987.

(14) "Reactor System Transient Analyses Using the RETRAN Computer Code," VEP-FRD-41, March, 1981, as supplemented by letter from W. L. Stewart to USNRC, 22

"Virginia Electric and Power Company, Surry and North Anna Power Stations, Reactor System Transient Analysis," NRC Letter Serial No.85-753, dated November 19, 1985.

(15) Letter from W. L. Stewart to USNRC, "Virginia Electric and Power Company, North Anna Power Station Unit 1, Proposed Technical Specification Change -

Supplement," NRC Letter Serial No. 88-202B, dated June 19, 1989.

(16) Updated Final Safety Analysis Report, Surry Power Station Units 1 and 2, Virginia Electric and Power Company.

(17) "EPRI PWR Safety and Relief Valve Test Program, Safety and Relief Valve Test Report," EPRI, NP-2628-SR, December, 1982.

(18) "Safety and Relief Valves in Light Water Reactors," EPRI, NP-4306-SR, December, 1985.

(19) "Virginia Electric and Power Company North Anna Power Station Units 1 and 2 Proposed Technical Specification Change," NRC Letter Serial No.94-238, dated April 15, 1994 (North Anna Heatup and Cooldown Curve and LTOPS Submittal).

{20) "North Anna Units 1 and 2 - Issuance of Amendments Re:

Pressure/Temperature Operating Limits/Low Temperature Overpressure Protection System Setpoints/Limiting Conditions for Operation, Action Statements, and Surveillance Requirements for PORVs and Block Valves to Address GL 90-06 (TAC Nos. M77363, M77364, M77433, M77434, M89312, M89313)," NRC Letter Serial No.94-607, dated October 5, 1994.

(21) B. F. Gore, et al.: "Value/Impact Analysis of Generic Issue 94, 'Additional Low Temperature Overpressure Protection for Light Water Reactors,"' NUREG/CR-5186, dated November, 1988.

(22) E. D. Throm: "Regulatory Analysis for the Resolution of Generic Issue 94,

'Additional Low-Temperature Overpressure Protection for Light Water Reactors,"' NUREG-1326, dated December, 1989.

(23) Letter from R. A. Clark (USNRC) to Sol Burstein, Amendment 45 to Operating License DPR-24, and Amendment 50 to Operating License DPR-27 (NRC Approval of Point Beach Units 1 and 2 LTOPS Submittal), dated May 20, 1980.

(24) "Methodology Used to Develop Cold Overpressurization Mitigating System Setpoints and RCS Heatup and Cool down Limit Curves," WCAP-14040, dated June, 1994.

(25) G. 0. Barrett, et al.: "Pressurizer Safety Valve Set Pressure Shift; Westinghouse Owners Group Project MUHP2351," WCAP-12910, dated March, 1991.

23

{26)

Letter from J. H. Goldberg (FP&L) to USNRC, St. Lucie Unit 1, Docket No 50-335, Proposed License Amendment, P-T Limits and LTOP Analysis, dated December 5, 1989.

{27) Letter from USNRC to J. H. Goldberg (FP&L), St. Lucie Unit 1 -Issuance of Amendment Re: Pressure/Temperature (PIT) Limits and Low Temperature Overpressure Protection (LTOP) Analysis {TAC No. 75386), Docket No. 50-335, dated June 11, 1990.

{28) Letter from J. N. Hannon (USNRC) to K. M. Haas (Consumers Power),

Palisades Plant - "Exemption from Requirements of 10 CFR 50.60, Acceptance Criteria for Fracture Prevention for Light-Water Nuclear Power Reactors for Normal Operation," March 2, 1995.

(29) Letter from J. F. Stolz (USNRC) to L. R. Eliason (Public Service Electric and Gas), "Exemption from Requirements of 10 CFR 50.60, Acceptance Criteria for Fracture Prevention for Light-Water Nuclear Power Reactors for Normal Operation, Salem Nuclear Generating Station, Units 1 and 2," February 13, 1995.

{30) SECY-94-267, "Status of Reactor Pressure Vessel Issues," dated October 28, 1994 24

Significant Hazards Consideration The Surry Units 1 and 2 Reactor Coolant Systems are protected from material failure by the imposition of restrictions on allowable pressure and temperature, and on heatup and cooldown rate. The Low Temperature Overpressure Protection System (LTOPS) ensures that material integrity limits are not exceeded during the design basis overpressurization accidents. Equipment operability requirements are imposed to ensure that the assumptions of the accident analyses remain valid. The operating restrictions, setpoints, and equipment operability requirements must be revised to extend their applicability to a higher cumulative burnup, and to improve operational flexibility.

The current pressure/temperature operating limits and LTOPS setpoint are valid to 15 EFPY for both Surry Units. According to the most recent estimates, the burnup applicability l_imits will be exceeded by Surry Unit 1 in February 1996. The proposed Surry Units 1 and 2 Technical Specifications include revised pressure/temperature operating limits valid to end-of-license. The reactor vessel integrity protection philosophy which supports the proposed Technical Specifications changes provides improved operational flexibility while maintaining an adequate margin of safety as demonstrated by the safety analysis.

Virginia Electric and Power Company has reviewed the proposed Technical Specification changes against the criteria of 10 CFR 50.92 and has concluded that the changes do not pose a significant hazards consideration. Specifically, operation of Surry Power Station in accordance with the Technical Specification changes will not:

a. involve a significant increase in the probability or consequences of an accident previously evaluated. The safety analysis demonstrates that the proposed reactor vessel protection philosophy, and the associated pressure/temperature limits, LTOPS setpoint, and component operability requirements, ensure that reactor vessel integrity will be maintained during normal operation and design basis accident conditions. Specifically, adherence to the heatup/cooldown rate-dependent pressure/temperature operating limits ensures that the assumed design basis flaw will not propagate during normal operation. Below the LTOPS enabling temperature, automatic actuation of the PORVs ensures that the assumed design basis flaw will not propagate under design basis low-temperature overpressurization

• accident conditions. Above the enabling temperature, two pressurizer safety valves are sufficient to relieve the overpressurization due to the inadvertent startup of two charging pumps at water solid conditions without propagation of the assumed design basis flaw.

b. create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed Technical Specifications modify pressure/temperature operating limits, LTOPS setpoint and enabling temperature, and component operability requirements. The revised pressure/temperature operating limits and LTOPS setpoint are only slightly different than those currently in the Technical Specifications. The LTOPS enabling temperature remains unchanged: No operating limits or setpoints are

- added or deleted by the proposed changes. Therefore, it may be concluded 25

that the operating limits and setpoint changes do not create the possibility of a new or different kind of accident. With regard to component operability requirements, restrictions on the number of charging pumps which may be operable, the number of PORVs which must be operable, and the allowable temperature difference between the steam generator primary and secondary remain unchanged. Only the setpoint temperature at which these restrictions apply have been modified. The proposed changes are entirely consistent with the reactor vessel integrity protection philosophy which ensures that the design basis reactor vessel flaw will not propagate under normal operation or postulated. accident conditions. Further, the proposed changes do not invalidate the any component design criteria or the assumptions of any UFSAR Chapter 14 accident analyses.

c. involve a significant reduction in a margin of safety. As described above, the reactor vessel integrity protection philosophy ensures that the design basis assumed flaw will not propagate under normal operation or design basis accident conditions. Adherence to the Technical Specification pressure/temperature operating limits ensures that the margin to vessel fracture provided by the ASME Section XI methodology is maintained. With regard to LTOPS protection, the safety analysis demonstrates that the proposed LTOPS design ensures margins consistent with those provided by ASME Section XI Appendix G methods as amended by ASME Code Case N-514. Utilization of ASME Code Case N-514 technically results in a reduction in the margin of safety, since a less restrictive LTOPS analysis design limit (i.e., 110% of the isothermal limit curve) is employed. However, the proposed design has been demonstrated to provide an acceptable margin of safety. Both industry experience and engineering evaluation support the conclusion that LTOPS design basis events may be expected to occur at essentially isothermal conditions. An engineering evaluation demonstrates that any reduction in allowable pressure due to thermal stresses which may be expected to exist during an LTOPS design basis event is insignificant when compared to margins provided by the ASME Section XI Appendix G methods for calculating pressure/temperature operating limits. This design maximizes the operating margin above the minimum RCS pressure for reactor coolant pump (RCP) operation, thereby minimizing the probability of undesired PORV lifts during RCP startup.

Virginia Electric and Power Company concludes that the activities associated with these proposed Technical Specification changes satisfy the no significant hazards consideration criteria of 10 CFR 50.92 and, accordingly, a no significant hazards consideration finding is justified.

26