ML24087A208

Proposed License Amendment Request Addition of Containment Limiting Condition for Operation and Surveillance Requirements

ML24087A208
Person / Time
Site:	Surry
Issue date:	03/22/2024
From:	James Holloway Virginia Electric & Power Co (VEPCO)
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
24-042
Download: ML24087A208 (1)

Text

VIRGI N IA ELECTRIC A N D POWER COMPA N Y

RICHMOND, VIRGI N IA 23261

March 22, 2024 10 CFR 50.90

U. S. Nuclear Regulatory Commission Serial No.: 24 - 042

_ Attention: Document Control Desk NRA/GDM: RO Washington, DC 20555-0001 Docket Nos.: 50-280/281 License Nos.: DPR -32/37

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 PROPOSED LICENSE AMENDMENT REQUEST ADDITION OF CONTAINMENT LIMITING CONDITION FOR OP ERATION AND SURVEILLANCE REQUIREMENTS

Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion Energy Virginia) requests an amendment to Facility Subsequent Renewed Operating License Numbers DPR-32 and DPR 37 in the form of a change to the Technical Specifications (TS) for Surry Power Station (SPS) Units 1 and 2. The proposed change revises SPS TS 3.8, Containment, to: 1) change the TS 3.8.D heading from Internal Pressure to Containment Pressure, 2) add a minimum containment air partial pressure limit to the Limiting Condition for Operation (LCO) in TS 3.8.D, and 3) add a new TS 3.8.E, Containment Air Temperature, to provide an LCO for containment average air temperature. The proposed change also revises TS 4.1, Operational Safety Review, to incorporate TS Surveillance Requirements (SRs) for containment air partial pressure and containment average air temperature. The proposed LCO and SRs are required to be included in the SPS TS pursuant to 10 CFR 50.36(c), and the changes are consistent with Standard TS (STS) 3.6.4B, Containment Pressure (Subatmospheric), and 3.6.5C, Containment Air Temperature (Subatmospheric), contained in NUREG-1431, Revision 5, Standard Technical Specifications - Westinghouse Plants (STS).

A discussion of the proposed license amendment is provided in Attachment 1. The marked-up and typed proposed TS pages are provided in Attachments 2 and 3, respectively. The associated TS Basis change supports the proposed license amendment and is provided for the NRC's information.

Dominion Energy Virginia has evaluated the proposed license amendment and has determined that it does not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for that determination is provided in Attachment 1. 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. Therefore, the proposed amendment is eligible for categorical exclusion as set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental Serial No.24-042 Docket Nos. 50-280/281 Page 2 of 3

assessment is needed in connection with the approval of the proposed change. The basis for this determination is also provided in Attachment 1. The LAR has been reviewed and approved by the Facility Safety Review Committee.

Dominion Energy Virginia requests NRC review and approval of the proposed LAR by March 31, 2025, with a 60-day implementation period.

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

Respectfully,

James E. Holloway Vice President - Nuclear Engineering and Fleet Support

Attachments:

1. Description and Assessment

2. Proposed Technical Specifications Pages (Marked-Up)

3. Proposed Technical Specifications Pages (Typed)

Commitments contained in this letter: None.

COMMONWEAL TH OF VIRGINIA )

)

COUNTY OF HENRICO )

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mr. James E. Holloway, who is Vice President - Nuclear Engineering and Fleet Support, of Virginia Electric and Power Company. He has affirmed before me that he 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 his knowledge and belief.

Acknowledged before me this ~ day of mo..vch, 2024.

My Commission Expires: q\\301'2.r

JULIE H HOUGH NOTARY PUBLIC 8066994 COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES 09-30-2027 Serial No.24-042 Docket Nos. 50-280/281 Page 3 of 3

cc: U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257

NRC Senior Resident Inspector Surry Power Station

Mr. L. John Klos NRC Project Manager - Surry U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, MD 20852-2738

Mr. G. Edward Miller NRC Senior Project Manager - North Anna U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, MD 20852-2738

State Health Commissioner Virginia Department of Health James Madison Building - 7th floor 109 Governor Street Suite 730 Richmond, VA 23219 Serial No.24-042 Docket Nos. 50-280/281

Attachment 1

DESCRIPTION AND ASSESSMENT

Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2 Serial No.24-042 Docket Nos. 50-280/281 Attachment 1

DESCRIPTION AND ASSESSMENT

1.0

SUMMARY

DESCRIPTION

Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion Energy Virginia) requests a change to the Technical Specifications (TS) for Surry Power Station (SPS) Units 1 and 2. The proposed change revises SPS TS 3.8, Containment, to:

1) change the TS 3.8.D heading from Internal Pressure to Containment Pressure, 2) add a minimum containment air partial pressure limit to the Limiting Condition for Operation (LCO) in TS 3.8.D, and 3) add a new TS 3.8.E, Containment Air Temperature, to provide an LCO for containment average air temperature. The proposed change also revises TS 4.1, Operational Safety Review, to incorporate TS Surveillance Requirements (SRs) for containment air partial pressure and containment average air temperature. The proposed LCO and SRs are required to be included in the SPS TS pursuant to 10 CFR 50.36(c), and the changes are consistent with Standard TS (STS) 3.6.48, Containment Pressure (Subatmospheric), and 3.6.5C, Containment Air Temperature (Subatmospheric), contained in NUREG-1431, Revision 5, Standard Technical Specifications - Westinghouse Plants (STS). Conforming changes to the TS 3.8 and 4.1 Bases are also being made and are provided for the NRC's information.

The proposed change has been reviewed with respect to 10 CFR 50.92, and it has been determined that no significant hazards consideration exists. In addition, it has been determined that the change qualifies for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9); therefore, no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed change.

2.0 DETAILED DESCRIPTION

2.1 System Design and Operation

2.1.1 Containment Pressure

Containment air partial pressure is a process variable that is monitored and controlled.

The containment air partial pressure is maintained as a function of Refueling Water Storage Tank (RWST) temperature and Service Water (SW) temperature as shown on Figure 2.1 (TS Figure 3.8-1) to ensure that, following a Design Basis Accident (OBA), the containment pressure will be less than 2.0 psig within one (1) hour and less than 0.0 psig within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Controlling containment partial pressure within prescribed limits also prevents the containment pressure from exceeding the containment design negative pressure differential with respect to the outside atmosphere in the event of an inadvertent actuation of the Containment Spray (CS) system.

Page 1 of 13 Serial No.24-042 Docket Nos. 50-280/281 Attachment 1

Limitations are placed on containment internal pressure and plant operating pressure to ensure that:

1. Release of radioactive materials from the containment will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. These restrictions ensure the site boundary radiation dose is within the applicable limits of 10 CFR 50.67 or Regulatory Guide 1.183 during accident conditions.

2. For either a Loss of Coolant Accident (LOCA) or a Main Steam Line Break (MSLB):

a. the peak containment pressure will be limited to the upper containment design pressure of 45 psig,

b. following a LOCA, the containment pressure will be less than 2.0 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and less than 0.0 psig within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

c. safety-related equipment within containment will not experience temperatures greater than those to which they have previously been qualified.

3. Containment pressure will remain greater than 8 psia in the event of an inadvertent CS actuation (the bottom mat liner can withstand a pressure of 8 psia and the shell and dome liner can withstand a pressure of 3 psia).

Pursuant to TS 3.8.0, containment air partial pressure shall be maintained within the acceptable operation range as identified on Figure 3.8-1 whenever the Reactor Coolant System (RCS) temperature and pressure exceed 350°F and 450 psig, respectively. The control of containment partial pressure within these limits is achieved by the Containment Vacuum and Leakage Monitoring System.

2.1.2 Containment Temperature

Containment average air temperature is an initial condition used in the OBA analyses that establishes the containment environmental qualification operating envelope for both pressure and temperature. The limit for containment average air temperature ensures that operation is maintained within the assumptions used in the OBA analyses for containment. Specifically, two temperatures affect the maximum allowable containment partial pressure: SW temperature and containment average air temperature (see Figure 2.1 ). Both temperatures affect the allowable containment partial pressure in the same way: the higher the temperature, the lower the upper limit for allowable partial pressure.

This is because the higher any of these temperatures are at the start of a LOCA, the higher the peak accident pressure will be. However, the existing analyses for operation with a core rated power of 2587 MWth allow containment bulk air temperature to vary between 75°F and 125°F, with an upper limit on air partial pressure of 11.3 psia. The containment partial pressure LCO specifies a lower limit of not less than 10.1 psia. This lower limit is imposed to ensure the containment pressure does not decrease below the

Page 2 of 13 Serial No.24-042 Docket Nos. 50 -280/281 Attachment 1

value of 8.0 psia during inadvertent initiation of CS as noted above. Containm e nt average air temperature is computed as a weighted average of twenty -four dry bulb platinum resistance temperature detectors (RTDs).

Containment partial pressure and av e rage air temperature must be verified to be less than their upper limits and greater tha n their lower limits or the reactor must b e placed in HOT SHUTDOWN.

F igu re 2.1

Surry Technical Specification Curve For Containment Allowable Air Partial Pressure Indication VS. Service Water Temperature 11.6 11.4

- (70,11.3)

<<s 11.2 Accep table Operation "

• a5 ~ Inside the Lines '

C.

(1) 11.0 (I) -

Cll '"

Q. -~ 10.8 "'

m. Containment Temperature Between 7~ F and 125°F "~

Q. ~ 10.6

< 10.4 " "'

. ---- (10 0,10.3) 10.2 - - (2s, 10.a: (70,10. 1) " I I (10 -I 0, 10.1) 10.0 I I 0 I I I I I I I I I I I I 0 0 I I I I I I I I I I I

• If I I I I I I I I I I I I

• I II I I I I I I I I I I 11 11 I I I I I I I I r I I ' f

~ ~ ~ ~ ~ ~ w ~ oo ~ ro ~ oo oo oo ~ 1001~

Service Water Temperature,°F

2.2 Current Technical Specifications Requirements

The SPS TS requirements for containment operability are contained in TS 3.8, Containment. TS 3.8.D, Internal Pressure, provides the TS requirement for containment a ir partial pressure as follows:

Containment air partial pressure shall be maintained within the acceptable operation range as identified in Figure 3. 8-1 whenever the Reactor Coolant System temperature and pressure exceed 350°F and 450 psig, respectively.

Page 3 of 13 Serial No.24-042 Docket Nos. 50-280/281 Attachment 1

With the containment air partial pressure outside the acceptable operation range, restore the air partial pressure to within acceptab_le limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The associated SPS TS 3.8 Basis states:

The allowable value for the containment air partial pressure is presented in TS Figure 3.8-1 for service water temperatures from 25 to 100°F. The RWST water shall have a maximum temperature of 45°F.

The horizontal upper limit line in TS F igure 3. 8-1 is based on MSLB peak calculated pressure criteria, and the sloped line from 70°F to 100°F service water temperatures is based on LOCA depressurization criteria.

If the containment air partial pressure rises to a point above the allowable value, the reactor shall be brought to the HOT SHUTDOWN condition. If a LOCA occurs at the time the containment air partial pressure is at the maximum allowable value, the maximum containment pressure will be less than design pressure (45 psig), the containment will depressurize to 2. 0 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and less than 0. 0 psig within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The radiological consequences analysis demonstrates acceptable results provided the containment pressure does not exceed 2. 0 psig for the interval from 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following the Design Basis Accident.

If the containment air partial pressure cannot be maintained greater than or equal to the minimum pressure in Figure 3.8-1, the reactor shall be brought to the HOT SHUTDOWN condition. The she/I. and dome plate liner of the containment can withstand an internal pressure as low as 3 psia, and the bottom mat liner is capable of withstanding an internal pressure as low as 8 psia.

2.3 Reason for the Proposed Change

SPS TS Figure 3.8-1, Surry Technical Specification Curve for Containment Allowable Air Partial Pressure Indication vs. Service Water Temperature, (see Figure 2.1 above),

provides the limiting operating conditions for the reactor containment including limits on containment air partial pressure and containment average air temperature. TS 3.8.D provides the LCO for containment air partial pressure; however, the SPS TS do not currently include a SR for containment air partial pressure, nor do they include a TS LCO or SR for containment average air temperature for verifying containment partial air pressure and containment average air temperature remain within their limits as indicated on TS Figure 3.8-1.

10 CFR 50.36(c)(2)(ii), Criterion 2, requires an LCO be provided for "(a) process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the

Page 4 of 13 Serial No.24-042 Docket Nos. 50-280/281 Attachment 1

integrity of a fission product barrier." Since the containment average air temperature is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a LOCA, it is required to have a TS LCO.

Likewise, 10 CFR 50.36(c)(3) requires SRs relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Consequently, SRs are necessary for containment air partial pressure and containment average air temperature to verify they are within their respective LCO limits to ensure that containment operation remains within the limits assumed for the containment analyses.

2.4 Description of the Proposed Change

As noted above, containment-related TS are included in TS 3.8, and containment partial pressure is specifically addressed by TS 3.8.D, Internal Pressure.

Pursuant to 10 CFR 50.36(c) requirements and for consistency with STS, the following specific changes will be incorporated into the SPS Units 1 and 2 TS:

• TS 3.8. D will be retitled from Internal Pressure to Containment Pressure for consistency with STS 3.6.4 and the minimum containment pressure value of ~10.1 psi a will added to the TS LCO.

• New TS 3.8.E, Containment Temperature, will be included in TS 3.8 to provide an LCO for containment average temperature as follows:

E. Containment Temperature

1. Containment average air temperature shall be ~ 75°F and ~ 125°F whenever the Reactor Coolant System temperature and pressure exceed 350°F and 450 psig, respectively.

a. If containment average temperature is not within the limits, restore the containment average temperature to within the limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

As noted in STS, the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> provided to restore the containment average temperature to within the LCO limits is acceptable considering the sensitivity of the analysis to variations in this parameter and provides sufficient time to correct minor problems.

• Two new TS SRs, 4.1, Containment Pressure, and 4.J, Containment Temperature, are being added to TS 4.1, Operational Safety Review, as follows:

Page 5 of 13 Serial No.24-042 Docket Nos. 50-280/281 Attachment 1

I. Containment Pressure - Verify containment air parlial pressure is within limits at the frequency specified in the Swveillance Frequency Control Program.

J. Containment Air Temperature - Verify containment average air temperature is within limits at the frequency specified in the Surveillance Frequency Control Program.

The initial surveillance frequencies to be included in the plant SFCP for containment pressure and containment temperature will be 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, respectively. The 12-hour frequency for verification of containment air partial pressure is considered acceptable based on operating experience related to trending of containment pressure variations and pressure instrument drift during applicable REACTOR OPERATION conditions. Furthermore, the 12-hour frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal containment pressure condition. The 24-hour frequency for verification of containment average air temperature is considered acceptable based on observed slow rates of temperature increase within containment as a result of environmental heat sources (due to the large volume of containment). Furthermore, the 24-hour frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal containment temperature condition.

If subsequent changes to the surveillance frequencies are deemed appropriate, they would be evaluated in accordance with NEI 04-10, "Risk-informed Method for Control of Surveillance Frequencies," Revision 1, as required by TS 6.4.S, Surveillance Frequency Control Program (SFCP).

• Conforming changes to the TS 3.8 and 4.1 Bases are also being made as indicated in Attachments 2 and 3:

3.0 Technical Evaluation

Containment Internal Pressure - Containment air partial pressure is an initial condition used in the containment OBA analyses to establish the maximum peak containment internal pressure. Maintaining containment pressure within the TS limits of TS 3.8.D.1 ensures that in the event of a OBA, the resultant peak containment accident pressure will be maintained below the containment design pressure. These limits also prevent the containment pressure from exceeding the containment design negative pressure differential with respect to the outside atmosphere in the event of inadvertent actuation of the CS System. The LCO limits also ensure the containment pressure does not exceed 2.0 psig for the interval from 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a OBA, and less than 0.0 psig beyond

Page 6 of 13 Serial No.24-042 Docket Nos. 50-280/281 Attachment 1

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to achieve acceptable radiological results pursuant to the SPS updated alternative source term analyses.

The containment internal air partial pressure limits are derived from the input conditions used in the containment OBA analyses. Limiting the containment internal air partial pressure and temperature in turn limits the pressure that could be expected following a OBA, thus ensuring containment OPERABILITY. Ensur ing containment OPERABILITY limits leakage of fission product radioactivity from containment to the environment. The maximum design internal pressure for the containment is 45.0 psig. The initial conditions used in the containment design basis ana lyses for a MSLB are an air partial pressure of 11.3 psia and an air temperature of 125°F for a peak pressure of 44.8 psig. The initial conditions used in the containment design basis analyses for a LOCA are an air partial pressure of 11.55 psia and an air temperature of 125°F for a peak pressure 44.0 psig.

Both values are less than the maximum design internal pressure for the containment.

The containment was also designed for an external pressure load of 6.7 psid (i.e., a design minimum pressure of 8.0 psia). The inadvertent actuation of the CS System was analyzed to determine the reduction in containment pressure. The initial conditions used in the analysis were 9.0 psia and 125°F. This resulted in a minimum pressure inside containment of 8.15 psia, which is above the design minimum of 8.0 psia.

Containment Temperature - The containment structure serves to contain radioactive material that may be released from the reactor core following a Design Basis Accident (OBA). The containment average air temperature is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a LOCA. The containment average air temperature limit is derived from the input conditions used in the containment functional analyses and the containment structure external pressure analyses. The proposed LCO ensures that initial conditions assumed in the analysis of containment response to a OBA are not violated during unit operations. The total amount of energy to be removed from containment by the CS and containment cooling systems during post-accident conditions is dependent upon the energy released to the containment due to the event, as well as the initial containment temperature and pressure.

The higher the initial temperature, the more energy which must be removed, resulting in a higher peak containment pressure and temperature. Exceeding containment design pressure may result in leakage greater than that assumed in the accident analysis.

Operation with containment temperature in excess of the proposed LCO limit would violate an initial condition assumed in the accident analysis.

During a OBA, with an initial containment average temperature less than or equal to the LCO temperature limits, the resultant peak accident temperature is maintained below the containment design temperature. As a result, the ability of containment to perform its design function is ensured.

Page 7 of 13 Serial No.24-042 Docket Nos. 50-280/281 Atta chment 1

4.0 REG U LATORY EV AL U A TI ON

4.1 Ap plicab le Re gul a to ry Re qui rem en ts a n d Crite ria

• Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include Technical Specifications (TS) as part of the operating license. The TS ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public. The U.S.

Nuclear Regulatory Commission's (NRC) requirements related to the content of the TS are contained in Section 50.36 of Title 10 of the Code of Federal Regulations (10 CFR 50.36), which requires that the TS include items in the following specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements per 10 CFR 50.36(c)(3); (4) design features; and (5) administrative controls.

• Technical Specifications - Operability requirements for the reactor containment buildings are prescribed by SPS Units 1 and 2 TS Section 3.8, "Containment," and define the limiting operating conditions of the reactor containment. Section 4.1, Operational Safety Review, outlines the surveillance requirements applied to unit equipment and conditions.

• 10 CFR 50 Appendix A. General Design Criteria - The regulations in Appendix A to Title 10 of the Code of Federal Regulations (10 CFR) Part 50 establish minimum principal design criteria for water-cooled nuclear power plants. The General Design Criteria (GDC) included in Appendix A to 10 CFR Part 50 did not become effective until May 21, 1971. The Construction Permits for SPS Units 1 and 2 were issued prior to May 21, 1971; consequently, SPS Units 1 and 2 were not subject to GDC requirements (Reference SECY-92 - 223 dated September 18, 1992).

• Updated Final Safety Analysis Report (UFSAR) - Surry's Updated Final Safety Analysis Report (UFSAR), Section 1.4, Compliance with Criteria, describes the station's compliance associated with its' seventy (70) design criteria. Specifically, subsection 1.4.10, "Containment," states, [in part], Containment is provided. The containment structure is designed to sustain the initial effects of gross equipment failures, such as a large coolant boundary break, without loss of required integrity, and, together with other engineered safeguards as may be necessary, to retain for as long as the situation requires the functional capability to protect the public...

UFSAR subsection 1.4.49, Containment Design Basis, states, The containment structure, including access openings and penetrations and any necessary containment heat removal systems, is designed to accommodate, without exceeding the design leakage rate, the pressures and temperatures resulting from the largest credible energy release following a LOCA, including a considerable margin for the effects of metal-water or other chemical reactions that can occur as a consequence of the failure of safety injection system.

Page 8 of 13 Serial No.24-042 Docket Nos. 50-280/281 Attachment 1

The design of the containment structure is based on the design basis accident, discussed in Sections 5.4.1 and 5.4.2 which assumes a double-ended rupture of the largest pipe in the reactor coolant system, coupled with partial loss of the redundant engineered safeguards systems (minimum safeguards). The maximum containment pressure reached in a design basis accident is less than the 45-psig design limit.

Further, the containment analyses performed assume a 2% metal-water reaction which is well above the less than 1 % expected for all accidents considered.

The containment structure, including access openings and penetrations, is designed to withstand a pressure of 45 psig and the associated thermal effects without exceeding the design leakage rate of 0. 1 weight percent of containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The heat removal capacity of the containment spray systems for the mm,mum safeguards returns the containment pressure to a subatmospheric condition in less than 60 minutes after a design-basis accident. This original design criterion was modified in conjunction with the analyses for implementation of the alternative source term. The criteria were subsequently updated to support an increase in the containment depressurization profile for the alternative source term analyses. The updated criteria require that, following the LOCA, the containment pressure be less than 2. 0 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and less than 0. 0 psig within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The radiological consequences analysis demonstrates acceptable results provided the containment pressure does not exceed 2. 0 psig for the interval from 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following the Design Basis Accident. Beyond 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, containment pressure is assumed to be less than 0. 0 psig, terminating leakage from containment.

UFSAR Chapter 5, "Containm ent System," defines the general description and design function of each unit's containment building. Subsection 5.1, "General Description,"

states [in part]: The steel-lined, reinforced-concrete containment structure, including foundations, access openings, and penetrations are designed and constructed to maintain full containment integrity when subjected to the temperatures, pressures, potential missiles resulting from the design-basis accident, and the earthquake conditions and tornados described in (UFSAR) Chapter 2. Systems are provided to remove heat from the containment and to ensure against breaching containment integrity at the time of, or following, the design-basis accident, or any lesser accident.

• Quality Assurance - 10 CFR 50 Appendix B and the licensee quality assurance program establish quality assurance requirements for the design, manufacture, construction, and operation of structures, systems, and components. Quality assurance criteria provided in 10 CFR Part 50, Appendix B, which apply to the systems and components pertinent to the proposed change include: Criteria Ill, V, XI, XVI, and XVII. Criteria Ill and V require measures to ensure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2, "Definitions," and as specified in the license application, are correctly translated into controlled

Page 9 of 13 Serial No.24-042 Docket Nos. 50-280/281 Attachment 1

specifications, drawings, procedures, and instructions. Criterion XI requires a test program to ensure that the subject systems will perform satisfactorily in service and requires that test results shall be documented and evaluated to ensure that test requirements have been satisfied. Criterion XVI requires measures to ensure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected, and that significant conditions adverse to quality are documented and reported to management. Criterion XVII requires maintenance of records of activities affecting quality.

The Dominion Energy Quality Assurance Program is described in Topical Report DOM-QA-1, "Dominion Nuclear Facility Quality Assurance Program Description (QAPD)." This topical report provides the QAPD for SPS Units 1 and 2. The Dominion Energy QAPD conforms to applicable regulatory requirements, such as 10 CFR 50, Appendix B, and approved industry standards, including equivalent alternatives, where identified. This program applies to activities during design, construction, and operation, as well as siting. The Dominion Energy QAPD describes how 10 CFR 50, Appendix B, requirements are met.

4.2 No Significant Hazards Consideration Determination

Virginia Electric and Power Company (Dominion Energy Virginia) has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed change to the Surry Power Station (SPS) Units 1 and 2 Subsequent Renewed Operating Licenses (SROLs) DPR-32 and DPR-37 and the Technical Specifications (TS).

The proposed change revises SPS TS 3.8, Containment, to: 1) change the TS 3.8.D heading from Internal Pressure to Containment Pressure, 2) add the minimum containment air partial pressure limit to the Limiting Condition for Operation (LCO) in TS 3.8.D, and 3) add a new TS 3.8.E, Containment Air Temperature, to provide an LCO for containment average air temperature. The proposed change also revises TS 4.1, Operational Safety Review, to incorporate TS Surveillance Requirements (SRs) for containment air partial pressure and containment average air temperature. These surveillances are necessary to ensure compliance with TS LCOs 3.8.D.1 and 3.8.E.1 and are required to be included in the SPS TS pursuant to 1 O CFR 50.36(c).

In accordance with the criteria set forth in 10 CFR 50.92, Dominion Energy Virginia has performed an analysis of the proposed change and concluded that it does not represent a significant hazards consideration. The following discussion is provided in support of this conclusion:

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

Response: No

Page 10 of 13 Serial No.24-042 Docket Nos. 50-280/281 Attachment 1

The proposed change to ensure containment pressure and temperature remain within required TS limits does not impact the condition or performance of any plant structure, system, or component, and actually confirms containment temperature and pressure are maintained within the operating limits and conditions specified in the safety analyses. The proposed change does not affect the initiators of any previously analyzed event or the assumed mitigation of accident or transient events. As a result, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated since neither accident probabilities nor consequences are being affected by the proposed change.

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

Response: No

The proposed change to ensure containment pressure and temperature remain within required TS limits does not involve any changes in station operation or physical modifications to the plant. In addition, no changes are being made in the methods used to respond to plant transients that have been previously analyzed. No changes are being made to plant parameters within which the plant is normally operated or in the setpoints that initiate protective or mitigative actions, and no new failure modes are being introduced. Therefore, the proposed change does not create the possibility of a new or different kind of accident or malfunction of equipment important to safety from any previously evaluated.

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

Response: No

The proposed change to ensure containment pressure and temperature remain within required TS limits does not affect the acceptance criteria for any analyzed event, nor is there a change to any safety limit. The proposed change does not affect any structures, systems or components or their capability to perform their intended functions and does not alter the way safety limits, limiting safety system settings or limiting conditions for operation are determined. Neither the safety analyses nor the safety analyses acceptance criteria are affected by the proposed change. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

4.3 Conclusion

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the

Page 11 of 13 Serial No.24-042 Docket Nos. 50-280/281 Attachment 1

Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL EVALUATION

The proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9) as follows:

(i) The proposed change involves no significant hazards consideration.

As described in Section 4.2 above, the proposed change involves no significant hazards consideration.

(ii) There are no significant changes in the types or significant increase in the amounts of any effluents that may be released off-site.

The proposed change to ensure containment pressure and temperature remain within required TS limits does not involve the installation of new or the removal of existing equipment. The proposed change does not modify any equipment that could affect the types or amounts of effluents that may be released off-site. The proposed change will have no impact on normal plant releases and will not increase the predicted radiological consequences of accidents postulated in the UFSAR. There are no significant changes in the types or significant increase in the amounts of any effluents that may be released off-site.

(iii) There is no significant increase in individual or cumulative occupation radiation exposure.

The proposed change to ensure containment pressure and temperature remain within required TS limits does not involve physical changes to the plant or introduce any new modes of plant operation. Therefore, there is no significant increase in individual or cumulative occupational radiation exposure.

Based on the above, Dominion Energy Virginia concludes that, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

Page 12 of 13 Serial No.24-042 Docket Nos. 50-280/281 Attachment 1

6.0 REFERENCES

6.1 Standard TS (STS) 3.6.4B, Containment Pressure (Subatmospheric), and 3.6.5C, Containment Air Temperature (Subatmospheric), contained in NUREG-1431, Revision 5, Standard Technical Specifications - Westinghouse Plants (STS).

6.2 NEI 04-10, "Risk-informed Method for Control of Surveillance Frequencies,"

Revision 1.

6.3 North Anna Power Station Units 1 and 2 Improved Technical Specifications, Section 3.6, "Containment Systems," Limiting Conditions for Operation 3.6.4, "Containment Pressure," and 3.6.5, "Cont a inment Air Temperature."

Page 13 of 13 Serial No.24-042 Docket Nos. 50-280/281

Attachment 2

MARKED-UP TECHNICAL SPECIFICATIONS AND BASES PAGES (Bases Changes are Provided for NRC Information Only)

Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2 TS 3.8-3 11 08 04

c. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual valve or blind flange, or

Containment d. Otherwise, place the unit in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

2:10.l psia and

1. Containment air partial pressure shall be '1HiMfflffitflet1-within the acceptable operation range as identified in Figure 3.8-1 whenever the Reactor Coolant System temperature and pressure exceed 350°F and 450 psig, respectively.

a. With the containment air partial pressure outside the acceptable operation range, restore the air partial pressure to within acceptable limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in INSERT 1 ~ COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Basis

CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment will be restricted to those leakage paths and associated leak rates assumed in the accident analysis. These restrictions, in conjunction with the allowed leakage, will limit the site boundary radiation dose to the applicable limits of 10 CFR 50.67 or Regulatory Guide 1.183 during accident conditions.

The operability of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment.

The opening of manual or deactivated automatic containment isolation valves on an intermittent basis under administrative control includes the following considerations:

(1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and

Amendment Nos. Bases-TS 3.8-5 07 05 22

If the containment air partial pressure rises to a point above the allowable value the reactor shall be brought to the HOT SHUTDOWN condition. If a LOCA occurs at the time the containment air partial pressure is at the maximum allowable value, the maximum containment pressure will be less than design pressure ( 45 psig), the containment will depressurize to 2.0 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and less than 0.0 psig within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The ~

radiological consequences analysis demonstrates acceptable results provided the containment pressure does not exceed 2.0 psig for the interval from 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following k the Design Basis Accident.

If the containment air partial pressure cannot be maintained greater than or equal to the minimum pressure in Figure 3.8-1, the reactor shall be brought to the HOT SHUTDOWN condition. The shell and dome plate liner of the containment are capable of withstanding an internal pressure as low as 3 psia, and the bottom mat liner is capable of withstanding

~ an internal pressure as low as 8 psia.

References

UFSAR Section 4.2.2.4 Reactor Coolant Pump

UFSAR Section 5.2 Containment Isolation

UFSAR Section 5.2.1 Design Bases

UFSAR Section 5.2.2 Isolation Design

UFSAR Section 5.3.4 Containment Vacuum System

Amendment Nos. Base&

TS 4.1-2 12 0l 11

H. If the RWST Water Chemistry exceeds 0.15 PPM for ci-and/or F-, flushing of sensitized stainless steel piping as required by 4.1.E will be performed once the RWST Water Chemistry has been brought within specification limit of less than 0.15 PPM chlorides and/or fluorides. Samples will be taken periodically until the sample

INSERT 3 ~ dicates the Cl-and/or F-and levels are below 0.15 PPM.

BASIS

Check

Failures such as blown instrument fuses, defective indicators, and faulted amplifiers which result in Hupscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system.

Furthermore, such failures are, in many cases, revealed by alarm or annunciator action, and a periodic check supplements this type of built-in surveillance.

Calibration

Calibration shall be performed to ensure the presentation and acquisition of accurate information.

The nuclear flux (power level) channels shall be calibrated against a heat balance __,r standard to account for errors induced by changing rod patterns and core physics parameters. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. t

Amendment Nos. Bases-TS 4.1-5b %

06 12 19

Trending the results of this surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The SFCP 7 day Frequency considers the low probability of a gross fuel failure during this time.

Due to the inherent difficulty in detecting Kr-85 in a reactor coolant sample due to masking from radioisotopes with similar decay energies, such as F-18 and 1-134, it is acceptable to include the minimum detectable activity for Kr-85 in this calculation. If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-133 is not detected, it should be assumed to be present at the minimum detectable activity.

CONTAINMENT AIR PARTIAL PRESSURE

Verifying containment air partial pressure is within the LCO limits ensures containment operation remains within the limits assumed for the containment analyses. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

CONTAINMENTAVERAGEAIRTEMPERATURE

Verifying containment average air temperature remains within the LCO limits ensures containment operation remains within the limits assumed for the containment analyses. To determine the containment average air temperature, a weighted average is calculated using measurements taken at locations within containment selected to provide a representative sample of the overall containment atmosphere. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

Amendment Nos. ~ and ~

Serial No.24-042 Docket Nos. 50-280/281 Attachment 2

INSERT 1 - New TS LCO 3.8.E

E. Containment Temperature

1. Containment average air temperature shall be ~ 75°F and ~ 125°F whenever the Reactor Coolant System temperature and pressure exceed 350°F and 450 psig, respectively.

a. If containment ?lverage temperature is not within the limits, restore the containment average temperature to within the limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

INSERT 2 - TS 3.8 Basis

During a Design Basis Accident, with an initial containment average air temperature within the specified temperature limits, the resultant peak accident temperature is maintained below the containment design temperature. As a result, the ability of containment to perform its design function is ensured.

INSERT 3 - New TS 4.1.1 and J

I. Containment Pressure - Verify containment air partial pressure is within limits at the frequency specified in the Surveillance Frequency Control Program.

J. Containment Air Temperature - Verify containment average air temperature is within limits at the frequency specified in the Surveillance Frequency Control Program.

Serial No.24-042 Docket Nos. 50-280/281

Attachment 3

PROPOSED TECHNICAL SPECIFICATIONS AND BASES CHANGES (Bases Changes are Provided for NRC Information Only)

Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2 TS 3.8-3

c. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual valve or blind flange, or

d. Otherwise, place the unit in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

D. Containment Pressure

1. Containment air partial pressure shall be 2: 10.1 psia and within the acceptable operation range as identified in Figure 3.8-1 whenever the Reactor Coolant System temperature and pressure exceed 350°F and 450 psig, respectively.

a. With the containment air partial pressure outside the acceptable operation range, restore the air partial pressure to within acceptable limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

E. Containment Temperature

1. Containment average air temperature shall be~ 75°F and~ 125°F whenever the Reactor Coolant System temperature and pressure exceed 350°F and 450 psig,

respectively.

a. If containment average temperature is not within the limits, restore the containment average temperature to within the limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Basis

CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment will be restricted to those leakage paths and associated leak rates assumed in the accident analysis. These restrictions, in conjunction with the allowed leakage, will limit the site boundary radiation dose to the applicable limits of 10 CFR 50.67 or Regulatory Guide 1.183 during accident conditions.

Amendment Nos.

TS 3.8-3a

The operability of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment.

The opening of manual or deactivated automatic containment isolation valves on an intermittent basis under administrative control includes the following considerations:

(1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation,

and

Amendment Nos.

TS 3.8-5

If the containment air partial pressure rises to a point above the allowable value the reactor shall be brought to the HOT SHUTDOWN condition. If a LOCA occurs at the time the containment air partial pressure is at the maximum allowable value, the maximum containment pressure will be less than design pressure ( 45 psig), the containment will depressurize to 2.0 psig within I hour and less than 0.0 psig within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The radiological consequences analysis demonstrates acceptable results provided the containment pressure does not exceed 2.0 psig for the interval from 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following the Design Basis Accident.

If the containment air partial pressure cannot be maintained greater than or equal to the minimum pressure in Figure 3.8-1, the reactor shall be brought to the HOT SHUTDOWN condition. The shell and dome plate liner of the containment are capable of withstanding an internal pressure as low as 3 psia, and the bottom mat liner is capable of withstanding an internal pressure as low as 8 psia.

During a Design Basis Accident, with an initial containment average air temperature within the specified temperature limits, the resultant peak accident temperature is maintained below the containment design temperature. As a result, the ability of containment to perform its design function is ensured.

References

UFSAR Section 4.2.2.4 Reactor Coolant Pump

UFSAR Section 5.2 Containment Isolation

UFSAR Section 5.2.1 Design Bases

UFSAR Section 5.2.2 Isolation Design

UFSAR Section 5.3.4 Containment Vacuum System

Amendment Nos.

TS 4.1-2

H. If the RWST Water Chemistry exceeds 0.15 PPM for Cl-and/or F-, flushing of sensitized stainless steel piping as required by 4.1.E will be performed once the RWST Water Chemistry has been brought within specification limit of less than 0.15 PPM chlorides and/or fluorides. Samples will be taken periodically until the sample indicates the Cl-and/or F-and levels are below 0.15 PPM.

I. Containment Pressure - Verify containment air partial pressure is within limits at the frequency specified in the Surveillance Frequency Control Program.

J. Containment Air Temperature - Verify containment average air temperature is within limits at the frequency specified in the Surveillance Frequency Control Program.

BASIS

Check

Failures such as blown instrument fuses, defective indicators, and faulted amplifiers which result in "upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system.

Furthermore, such failures are, in many cases, revealed by alarm or annunciator action, and a periodic check supplements this type of built-in surveillance.

Calibration

Calibration shall be performed to ensure the presentation and acquisition of accurate information.

The nuclear flux (power level) channels shall be calibrated against a heat balance standard to account for errors induced by changing rod patterns and core physics parameters. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Amendment Nos.

TS4.l-5b

Trending the results of this surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The SFCP 7 day Frequency considers the low probability of a gross fuel failure during this time.

Due to the inherent difficulty in detecting Kr-85 in a reactor coolant sample due to masking from radioisotopes with similar decay energies, such as F-18 and I-134, it is acceptable to include the minimum detectable activity for Kr-85 in this calculation. If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-133 is not detected, it should be assumed to be present at the minimum detectable activity.

CONTAINMENT AIR PARTIAL PRESSURE

Verifying containment air partial pressure is within the LCO limits ensures containment operation remains within the limits assumed for the containment analyses. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

CONTAINMENT A VERA GE AIR TEMPERATURE

Verifying containment average air temperature remains within the LCO limits ensures containment operation remains within the limits assumed for the containment analyses. To determine the containment average air temperature, a weighted average is calculated using measurements taken at locations within containment selected to provide a representative sample of the overall containment atmosphere. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

Amendment Nos.