ML18289A749

Attachment - Vogtle Unit 4 LAR-17-043 Containment Pressure Analysis

ML18289A749
Person / Time
Site:	Vogtle
Issue date:	11/07/2018
From:	Bill Gleaves NRC/NRO/DLSE/LB4
To:	Southern Nuclear Operating Co
gleaves b/415-5848
Shared Package
ML18289A742	List: ML18289A743 ML18289A745 ML18289A746 ML18289A748 ML18289A749 ML18289A750 ML18289A752 ML18289A753
References
EPID L-2018-LLA-0005, LAR-17-043
Download: ML18289A749 (8)

Text

ATTACHMENT TO LICENSE AMENDMENT NO. 146 TO FACILITY COMBINED LICENSE NO. NPF-92 DOCKET NO.52-026 Replace the following pages of the Facility Combined License No. NPF-92 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Facility Combined License No. NPF-92 REMOVE INSERT 7 7 Appendix A to Facility Combined License Nos. NPF-91 and NPF-92 REMOVE INSERT 5.5-8 5.5-8 Appendix C to Facility Combined License No. NPF-92 REMOVE INSERT C-111 C-111 C-111a C-111a C-111b C-111b C-409 C-409 C-426 C-426

(7) Reporting Requirements (a) Within 30 days of a change to the initial test program described in UFSAR Section 14, Initial Test Program, made in accordance with 10 CFR 50.59 or in accordance with 10 CFR Part 52, Appendix D, Section VIII, Processes for Changes and Departures, SNC shall report the change to the Director of NRO, or the Directors designee, in accordance with 10 CFR 50.59(d).

(b) SNC shall report any violation of a requirement in Section 2.D.(3),

Section 2.D.(4), Section 2.D.(5), and Section 2.D.(6) of this license within 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />. Initial notification shall be made to the NRC Operations Center in accordance with 10 CFR 50.72, with written follow up in accordance with 10 CFR 50.73.

(8) Incorporation The Technical Specifications, Environmental Protection Plan, and ITAAC in Appendices A, B, and C, respectively of this license, as revised through Amendment No. 146, are hereby incorporated into this license.

(9) Technical Specifications The technical specifications in Appendix A to this license become effective upon a Commission finding that the acceptance criteria in this license (ITAAC) are met in accordance with 10 CFR 52.103(g).

(10) Operational Program Implementation SNC shall implement the programs or portions of programs identified below, on or before the date SNC achieves the following milestones:

(a) Environmental Qualification Program implemented before initial fuel load; (b) Reactor Vessel Material Surveillance Program implemented before initial criticality; (c) Preservice Testing Program implemented before initial fuel load; (d) Containment Leakage Rate Testing Program implemented before initial fuel load; (e) Fire Protection Program

1. The fire protection measures in accordance with Regulatory Guide (RG) 1.189 for designated storage building areas (including adjacent fire areas that could affect the storage area) implemented before initial receipt 7 Amendment No. 146

Technical Specifications Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Safety Function Determination Program (SFDP) (continued)

b. A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed.

For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

1. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

2. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

3. A required system redundant to the support system(s) for the supported systems b.1 and b.2 above is also inoperable.

c. The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.8 Containment Leakage Rate Testing Program

a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by approved exceptions.

b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 58.1 psig. The containment design pressure is 59 psig.

c. The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.10% of primary containment air weight per day.

VEGP Units 3 and 4 5.5 - 8 Amendment No. 147 (Unit 3)

Amendment No. 146 (Unit 4)

Table 2.2.2-3 Inspections, Tests, Analyses, and Acceptance Criteria No. ITAAC No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 138 2.2.02.07b.i 7.a) The PCS delivers water from i) Testing will be performed to i) When tested, each one of the the PCCWST to the outside, top measure the PCCWST delivery three flow paths delivers water at of the containment vessel. rate from each one of the three greater than or equal to:

parallel flow paths. - 469.1 gpm at a PCCWST water level of 27.4 ft + 0.2, - 0.0 ft above the tank floor

- 226.6 gpm when the PCCWST water level uncovers the first (i.e. tallest) standpipe

- 176.3 gpm when the PCCWST water level uncovers the second tallest standpipe

- 144.2 gpm when the PCCWST water level uncovers the third tallest standpipe

- or a report exists and concludes that the as-measured flow rates delivered by the PCCWST to the containment vessel provides sufficient heat removal capability such that the limiting containment pressure and temperature values are not affected and the PCS is able to perform its safety function to remove heat from containment to maintain plant safety.

ii) Testing and or analysis will be ii) When tested and/or analyzed performed to demonstrate the with all flow paths delivering and PCCWST inventory provides an initial water level at 27.4 + 0.2, 72 hours3 days <br />0.429 weeks <br />0.0986 months <br /> of adequate water flow. - 0.00 ft, the PCCWST water inventory provides greater than or equal to 72 hours3 days <br />0.429 weeks <br />0.0986 months <br /> of flow, and the flow rate at 72 hours3 days <br />0.429 weeks <br />0.0986 months <br /> is greater than or equal to 100.7 gpm or a report exists and concludes that the as-measured flow rates delivered by the PCCWST to the containment vessel provides sufficient heat removal capability such that the limiting containment pressure and temperature values are not affected and the PCS is able to perform its safety function to remove heat from containment to maintain plant safety.

C-111 Amendment No. 146

Table 2.2.2-3 Inspections, Tests, Analyses, and Acceptance Criteria No. ITAAC No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 7.b) The PCS wets the outside i) Testing will be performed to i) A report exists and concludes surface of the containment vessel. measure the outside wetted that when the water in the The inside and the outside of the surface of the containment vessel PCCWST uncovers the standpipes containment vessel above the with one of the three parallel flow at the following levels, the water operating deck are coated with an paths delivering water to the top delivered by one of the three inorganic zinc material. of the containment vessel. parallel flow paths to the containment shell provides coverage measured at the spring line that is equal to or greater than the stated coverages.

- 24.1 +/- 0.2 ft above the tank floor; at least 90% of the perimeter is wetted.

- 20.3 +/- 0.2 ft above the tank floor; at least 72.9% of the perimeter is wetted.

- 16.8 +/- 0.2 ft above the tank floor; at least 59.6% of the perimeter is wetted.

ii) Inspection of the containment ii) A report exists and concludes vessel exterior coating will be that the containment vessel conducted. exterior surface is coated with an inorganic zinc coating above elevation 135'-3".

iii) Inspection of the containment iii) A report exists and concludes vessel interior coating will be that the containment vessel conducted. interior surface is coated with an inorganic zinc coating above the operating deck.

7.c) The PCS provides air flow Inspections of the air flow path Flow paths exist at each of the over the outside of the segments will be performed. following locations:

containment vessel by a natural - Air inlets circulation air flow path from the - Base of the outer annulus air inlets to the air discharge - Base of the inner annulus structure. - Discharge structure 7.d) The PCS drains the excess Testing will be performed to With a water level within the water from the outside of the verify the upper annulus drain upper annulus 10" +/- 1" above the containment vessel through the flow performance. annulus drain inlet, the flow rate two upper annulus drains. through each drain is greater than or equal to 525 gpm.

7.e) The PCS provides a flow path ii) Testing will be performed to ii) With a water supply connected for long-term water makeup to the measure the delivery rate from the to the PCS long-term makeup PCCWST. long-term makeup connection to connection, each PCS the PCCWST. recirculation pump delivers greater than or equal to 100 gpm when tested separately.

C-111a Amendment No. 146

Table 2.2.2-3 Inspections, Tests, Analyses, and Acceptance Criteria No. ITAAC No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

9. Safety-related displays Inspection will be performed for Safety-related displays identified identified in Table 2.2.2-1 can be retrievability of the safety-related in Table 2.2.2-1 can be retrieved retrieved in the MCR. displays in the MCR. in the MCR.

10.a) Controls exist in the MCR Stroke testing will be performed Controls in the MCR operate to to cause the remotely operated on the remotely operated valves cause remotely operated valves valves identified in Table 2.2.2-1 identified in Table 2.2.2-1 using identified in Table 2.2.2-1 to to perform active functions. the controls in the MCR. perform active functions.

10.b) The valves identified in Testing will be performed on the The remotely operated valves Table 2.2.2-1 as having PMS remotely operated valves in Table identified in Table 2.2.2-1 as control perform an active safety 2.2.2-1 using real or simulated having PMS control perform the function after receiving a signal signals into the PMS. active function identified in the from the PMS. table after receiving a signal from the PMS.

11.a) The motor-operated valves iii) Tests of the motor-operated iii) Each motor-operated valve identified in Table 2.2.2-1 valves will be performed under changes position as indicated in perform an active safety-related preoperational flow, differential Table 2.2.2-1 under function to change position as pressure, and temperature preoperational test conditions.

indicated in the table. conditions.

11.b) After loss of motive power, Testing of the remotely operated After loss of motive power, each the remotely operated valves valves will be performed under remotely operated valve identified identified in Table 2.2.2-1 assume the conditions of loss of motive in Table 2.2.2-1 assumes the the indicated loss of motive power power. indicated loss of motive power position. position.

139 2.2.02.07b.ii Not used per Amendment No. 112 140 2.2.02.07b.iii Not used per Amendment No. 112 141 2.2.02.07c Not used per Amendment No. 112 142 2.2.02.07d Not used per Amendment No. 112 C-111b Amendment No. 146

1. The physical arrangement of the nuclear island structures, the annex building, and the turbine building is as described in the Design Description of this Section 3.3, and as shown on Figures 3.3-1 through 3.3-14. The physical arrangement of the radwaste building and the diesel generator building is as described in the Design Description of this Section 3.3.

2. a) The nuclear island structures, including the critical sections listed in Table 3.3-7, are seismic Category I and are designed and constructed to withstand design basis loads, as specified in the Design Description, without loss of structural integrity and the safety-related functions. The design bases loads are those loads associated with:

• Normal plant operation (including dead loads, live loads, lateral earth pressure loads, and equipment loads, including hydrodynamic loads, temperature and equipment vibration);

• External events (including rain, snow, flood, tornado, tornado generated missiles and earthquake); and

• Internal events (including flood, pipe rupture, equipment failure, and equipment failure generated missiles).

b) Site grade level is located relative to floor elevation 100-0 per Table 3.3-5. Floor elevation 100-0 is defined as the elevation of the floor at design plant grade.

c) The containment and its penetrations are designed and constructed to ASME Code Section III, Class MC.(1) d) The containment and its penetrations retain their pressure boundary integrity associated with the design pressure.

e) The containment and its penetrations maintain the containment leakage rate less than the maximum allowable leakage rate associated with the peak containment pressure for the design basis accident.

f) The key dimensions of the nuclear island structures are as defined on Table 3.3-5.

g) The containment vessel above the operating deck provides a heat transfer surface. A free volume exists inside the containment shell above the operating deck.

h) The containment free volume below elevation 107.68 provides containment floodup during a postulated loss-of-coolant accident.

3. Walls and floors of the nuclear island structures as defined on Table 3.3-1, except for designed openings and penetrations, provide shielding during normal operations.

4. a) Walls and floors of the annex building as defined on Table 3.3-1, except for designed openings and penetrations, provide shielding during normal operations.

b) The walls on the outside of the waste accumulation room in the radwaste building provide shielding from accumulated waste.

1. Containment isolation devices are addressed in subsection 2.2.1, Containment System.

C-409 Amendment No. 146

Table 3.3-6 Inspections, Tests, Analyses, and Acceptance Criteria No. ITAAC No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 772 3.3.00.02d Not used per Amendment No. 84 773 3.3.00.02e Not used per Amendment No. 84 774 3.3.00.02f 2.f) The key dimensions of nuclear An inspection will be performed A report exists and concludes island structures are defined on of the as-built configuration of that the key dimensions of the Table 3.3-5. the nuclear island structures. as-built nuclear island structures are consistent with the dimensions defined on Table 3.3-5.

775 3.3.00.02g 2.g) The containment vessel above the The maximum containment The containment vessel operating deck provides a heat transfer vessel inside height from the maximum inside height from surface. A free volume exists inside operating deck is measured and the operating deck is 146'-7" the containment shell above the the inner radius below the spring (with tolerance of +12", -6"),

operating deck. line is measured at two and the inside diameter is orthogonal radial directions at 130 feet nominal (with one elevation. tolerance of +12", -6").

776 3.3.00.02h 2.h) The free volume in the An inspection will be performed A report exists and concludes containment allows for floodup to of the as-built containment that the floodup volume of support long-term core cooling for structures and equipment. The this portion of the containment postulated loss-of-coolant accidents. portions of the containment is less than 71,960 ft3 to an included in this inspection are elevation of 107.68'.

the volumes that flood with a loss-of-coolant accident in passive core cooling system valve/equipment room B (11207). The in-containment refueling water storage tank volume is excluded from this inspection.

C-426 Amendment No. 146