ML22326A065

4 to the Updated Final Safety Analysis Report, Chapter 3, Design of Structures, Components, Equipment, and Systems (Part 7 of 7) and Chapter 4, Reactor (Part 1 of 2)

ML22326A065
Person / Time
Site:	Vogtle
Issue date:	10/31/2022
From:	Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML22326A145	List: ML22326A059 ML22326A060 ML22326A062 ML22326A063 ML22326A064 ML22326A065 ML22326A066 ML22326A067 ML22326A107 ML22326A108 ML22326A109 ML22326A113 ML22326A114 ML22326A116 ML22326A117 ML22326A119 ML22326A120 ML22326A121 NL-22-0810, 4 to the Updated Final Safety Analysis Report, Technical Specification Bases Changes, Technical Requirements Manual Changes, 10 CFR 50.59 Summary Report and Revised NRC Commitments Report
References
NL-22-0810
Download: ML22326A065 (1)

Text

REV 14 10/07 CONTAINMENT BUILDING EL. 323 FT.POLAR CRANE FIGURE 3D-137

REV 14 10/07 CONTAINMENT BUILDING EL. 323 FT. POLAR CRANE FIGURE 3D-138

REV 14 10/07 CONTAINMENT BUILDING EL. 323 FT. POLAR CRANE FIGURE 3D-139

REV 14 10/07 CONTROL BUILDING EL. 180 FT. BASEMAT FIGURE 3D-140

REV 14 10/07 CONTROL BUILDING EL. 180 FT. BASEMAT FIGURE 3D-141

REV 14 10/07 CONTROL BUILDING EL. 180 FT. BASEMAT FIGURE 3D-142

REV 14 10/07 CONTROL BUILDING EL. 220 FT.

FIGURE 3D-143

REV 14 10/07 CONTROL BUILDING EL. 220 FT.

FIGURE 3D-144

REV 14 10/07 CONTROL BUILDING EL. 220 FT.

FIGURE 3D-145

REV 14 10/07 CONTROL BUILDING EL. 260 FT.

FIGURE 3D-146

REV 14 10/07 CONTROL BUILDING EL. 260 FT.

FIGURE 3D-147

REV 14 10/07 CONTROL BUILDING EL. 260 FT.

FIGURE 3D-148

VEGP-FSAR-3 APPENDIX 3E IMPEDANCE FUNCTIONS FOR AN ARBITRARILY SHAPED FOUNDATION ON A LAYERED MEDIUM This appendix describes the procedure used to compute the impedance functions for an arbitrarily shaped foundation resting on a layered soil medium, for use in the soil-structure interaction analysis as specified in subsection 3.7.B.2.

The analytical techniques to obtain the impedance functions for flat rigid foundations of arbitrary shape placed on the surface of an elastic half space have been developed by Wong and Luco.(1) The equation of motion for the forced steady-state vibrations of an elastic half space excited by harmonic loads distributed over a region S of the plane surface (figure 3E-1) is:

(c 2 2 ) ( u) + 2 2u + 2u = 0 3 0 (1) it in which u is the displacement vector (u1,u2,u3)e in the cartesian system of coordinates (1,2,3) such that 3 = 0 corresponds to the surface of the half space with 3 > 0 representing the points within the half space. The symbols c and are the compressional and shear wave velocities, respectively.

Assuming that the surface tractions on the loaded region S are known, or equivalently, assuming that the stress components j3 (j = 1,2,3) on S are known, then a solution of equation (1) satisfying the mixed boundary-value problem on 3 = 0, in which displacements are prescribed along the contact between the foundation and the soil while tractions are prescribed on the soil surface not covered by the foundation, is given by:

3 ui (1,2,0) = - ³G j =1 s ij (i 10 , 2 02 ) j3 ( 10 , 02 , 0) d10 d 02 (2) for 3 = 0. In equation 2, Gij ( 10 10 , 2 02 , 0 ) denotes the ith displacement component at (1,2,0) generated by a unit harmonic load acting at ( 10 , 02 , 0 ) .

To solve this integral equation 2 for an arbitrarily shaped foundation, the following numerical procedure is used:

A. The region S is divided into n rectangular subregions Sk (k = 1,2,...,n) as indicated in figure 3E-1.

(j3k )

B. The stress components j3 are assumed to have constant values within each subregion Sk.

C. The boundary conditions are satisfied approximately by matching the average displacements within each subregion to the average value of the required compatible displacements.

Using the above approximations, integral equation 2 can be expressed in matrix equations as:

3E-1 REV 14 10/07

VEGP-FSAR-3

º

<< [ ]11 [ ]12. . . . . [ ]1 n >>

_ 1/2 << >> (1) 1/2

°u~ 1 ° << >> ° j3 A1 °

°_ ° << >> ° °

°u 2 ° << [ ] 21 [ ] 22. . . . . [ ] 2 n >> ° °

°~ ° << >> ° (j32 ) A 2°

.3/4 = <<. . . >> ° .

° 3/4 (3)

°.° <<. . . >> ° . °

°.° << >> ° . °

°_ ° <<. . . >> ° °

°u~ n ° << >> ° ( n ) n °

- ¿ << >>

<< [ ] n1 [ ] n 2. . . . . [ ] nn >>

° j3 °

- ¿

¬ 1/4 in which:

_ °_ _ _ 1/2° T u = u1i, u 2i, u 3i 3/4 are average displacements in subregion Si.

~1 ° °¿ Ai = area if subregion Si.

[]ij = a 3x3 compliance submatrix relating the average displacements at subregion Si to the tractions at subregion Sj.

To calculate the compliance submatrices, four linear integrals on the Green's functions are performed for point loads at the surface of a layered stratum. The formula for []ij is:

º

[]ij = ³ ds ³ ds o <<G ( , r~ r , P )>> (4)

Si Sj <<¬ ~o ~ >>

1/4 in which [G] is the 3x3 Greens function matrix relating the displacement at the observation point

( )T r = (1, 2, 0)T to a set of point loads at the source point r ~o 10 , 20 , 0

• P~ is the property vector associated with the underlying half space. By use of the average displacement matching approximation, the following symmetry exists even if Ai Aj:

[]ijT = [] ji (5)

The property vector P ~

for a horizontally layered stratum can be characterized as:

P = (m, i, i, i,hi, i, i) (6)

~

where:

m = the number of layers.

i = the shear modulus of the ith layer.

i = the shear wave velocity of the ith layer.

i = the mass density of the ith layer.

Hi = the thickness of the ith layer.

3E-2 REV 14 10/07

VEGP-FSAR-3 i = the Poisson's ratio of the ith layer.

i = the critical damping coefficient of the ith layer.

The matrix [G] contains six independent elements. In order to reduce the number of independent variables and to render [G] dimensionless, a new matrix [G] may be defined in the polar coordinate system {r, , z}, as shown in figure 3E-2:

[G (,r r , P )] = 1r [G (b , , P )]

~ ~o ~ 0 ~ (7) where:

r = r ro = (1 1o )2 + (2 o2) 2

~ ~

( )

= arg r~ ~ro = tan ¨ 1

§ 2 o2 *

¨ o ¸

¸

© 1 1¹ r

bo =

P~ = { m, i , i , i , hi , i , } , the normalized property vector in which:

i i =

i i =

i = i h i hi =

The reference values of , and , used to normalize and [G'], are usually taken to be

~

those of the top layer.

The dimensionless matrix [G'] is a function of four variables, frr, fr, frz, and fzz, which are the Greens functions in the polar coordinate system.

The Green's functions for three-dimensional wave propagation in layered viscoelastic media have been formulated and solved by Luco and Apsel.(2)(3) In frequency domain, solution of the Green's functions in polar coordinates, which involve the Hankel transform-type integral representations of the displacement and stress components, can be expressed in the form:

³ In (b o ) = F( k, , ~ ) Jn (kb o ) dk 0

(8) for the concentrated point load applied at surface and displacement at the free surface observed at bo distance from the load point. The kernel F depends upon wave number k, frequency , and layer properties ~ ; whereas, the Bessel functions Jn depend only upon kbo.

The F integrands are evaluated in terms of factorizations of the upgoing and downgoing wave amplitudes in each layer. The semi-infinite integral in equation 8 can be reduced to the following finite integral:

3E-3 REV 14 10/07

VEGP-FSAR-3 k"

In (b o ) = In (0) + ³ [ F ( k, , ) F ( k, 0, ) ] J 0

~ n ( kb o ) dk (9) in which the upper limit of integration, k " , is defined by the convergence of the dynamic integrands to the static integrands, and In(0) represents the static (= 0) integrals. Since the radial dependence bo appears only in the Bessel functions Jn, it is expedient to calculate the r max integrals; begin at bo = 0 and end at bo = (rmax is the maximum length of the foundation) in equally spaced intervals. This precalculated Greens function table can be repeatedly used in solving the compliance submatrices []ij in integral equation 4 using the Gaussian quadrature.

For a rigid foundation, the average displacements u~i , evaluated at the center of subregion Si are given by:

uli ( 1i , i2 , 0 ) = 1 3 i2 u2i ( 1i , i2 , 0 ) = 2 + 3 i2 (10) u3i ( 1i , i2 , 0 ) = 3 + 1 i2 2 i2 where i (i = 1, 2, 3) corresponds to the amplitudes of the translational displacements at (0,0,0),

while i (i = 1,2,3), which is assumed to be small, corresponds to the amplitudes of the rotational displacements about the i (i = 1, 2, 3) axes. From equation 3, the three corresponding traction components (j3k ) A k can be expressed in terms of u ~i

, by inverting the matrix []. By substituting for u

~i from equation 10, the surface tractions on the contact area may be expressed in terms of the translation 1 and rotation i of the rigid foundation. Finally, the total harmonic load with components (P1,P2,P3) and the total harmonic moment with components (M1,M2,M3) acting on the contact area can be expressed in terms of traction components by means of the following relationships:

n P1 =

j=1

( j) i3 Aj (i = 1,2, 3) n M1 =

j=1 j

2 (33j ) (11) n M2 =

j=1 1j (33j)

[ ]

n j

M3 = 1 (23j) Aj 2j 13

( j)

Aj j=1 Substitution for contact tractions in terms of i and i into equation 11 leads to the desired force-displacement relationship for the rigid foundation:

P 1/2 1/2 3/4 = [K ] 3/4 (12)

-M ¿ - ¿ 3E-4 REV 14 10/07

VEGP-FSAR-3 where [K] is the complex frequency dependent impedance functions for flat rigid foundations placed on the surface of an elastic half space.

3E.1 REFERENCES

1. Wong, H. L., and Luco, J. E., "Dynamic Response of Rigid Foundations of Arbitrary Shape," Earthquake Engineering and Structural Dynamics, Vol 4, pp 579-587, 1976.

2. Luco, J. E., and Apsel, R. J., "On the Green's Functions for the Layered Half Space,"

Part 1, Bulletin of the Seismological Society of America, Vol 73, pp 901-929, 1983.

3. Apsel, R. J., and Luco, J. E., "On the Green's Functions for the Layered Half Space,"

Part 2, Bulletin of the Seismological Society of America, Vol 73, pp 931-951, 1983.

3E-5 REV 14 10/07

REV 14 10/07 MODEL AND COORDINATE SYSTEM FIGURE 3E-1

REV 14 10/07 POLAR COORDINATE SYSTEM FIGURE 3E-2

VEGP-FSAR-3 APPENDIX 3F HAZARDS ANALYSIS 3F.1 INTRODUCTION The VEGP power block has been designed to provide protection for safety-related equipment from hazards and events which could reasonably be expected to occur. This protection is provided to ensure that recovery from the event is possible, to ensure the integrity of the reactor coolant pressure boundary, to minimize the release of radioactivity, and to enable the plant to be placed in a safe condition.

This appendix provides the results of integrated hazards analyses for selected areas of the plant to demonstrate the type of analyses conducted for each safety-related area of the plant to ensure that the VEGP units can withstand the postulated events. A sample of the results of the analysis of level C and safety-related pump rooms on levels B and D of the auxiliary building as shown in table 3F-1 provides this example. Analyses are also provided for the pressure/temperature effects of a pipe rupture in the Unit 1 control building main steam line and main feedwater line isolation valve compartment, the flooding effects of pipe ruptures in the Unit 1 auxiliary and control building main steam line and main feedwater line isolation valve compartments, the effects of pipe ruptures in the Unit 1 auxiliary feedwater pump rooms, and the effects of a circulating water pipe rupture.

The items considered in the evaluation of each plant area include tornadoes, floods, missiles, pipe breaks, fires, and seismic events. (Refer to sections 3.3 through 3.7 and subsection 9.5.1.)

Even though each area of the plant and each system are designed individually to properly consider the above events, an integrated analysis of rooms, systems, and events is performed to ensure that the above objectives are realized for each postulated event.

The hazards analyses are conducted on a room-by-room basis. All components within the room are reviewed for the effects of earthquake-induced failures, effects of high- and moderate-energy piping breaks (flooding, sprays, and jet impingement), and the effects of missiles.

The effects of the high-energy pipe breaks on equipment are reported in paragraph 3.6.2.5.

Fire protection and the effects of fires in the various fire areas are discussed in subsection 9.5.1.

3F.2 ANALYSIS ASSUMPTIONS In the analysis of an event or hazard, it is assumed that the plant is being operated in accordance with the requirements of the Technical Specifications. Should the event result in a turbine or reactor trip, the plant will be placed in a hot shutdown condition. If required by a limiting condition of operation or if recovery from the event will cause the plant to be shut down for an extended period of time, the plant will be taken to a cold shutdown condition. (Safe shutdown is discussed in section 7.4.)

During the hot shutdown condition, an adequate heat sink is provided to remove reactor core residual heat. Boration capability is provided to compensate for xenon decay and to maintain the required core shutdown margin. Boration is a long-term need because it is not required until approximately 25 h after shutdown.

Redundancy or diversity of systems and components is provided to enable continued operation at hot shutdown or to cool the reactor to a cold shutdown condition. If time is available, it is assumed that temporary repairs can be made to circumvent damages resulting from the hazard.

Loss of offsite power (LOSP) is not assumed, unless a trip of the turbine-generator system or 3F-1 REV 23 3/21

VEGP-FSAR-3 the reactor protection system is a direct consequence of the hazard. All available systems, including nonsafety-related systems and those systems requiring operator action, may be employed to mitigate the consequences of the hazards.

In determining the availability of the systems required to mitigate the consequences of a hazard and those required to place the reactor in a safe condition, the direct consequences of the hazard are considered. The feasibility of carrying out operator actions is based on ample time and adequate access to the controls, motor control center, switchgear, etc., associated with the component required to accomplish the proposed action.

When the postulated hazard occurs in one of two or more redundant trains of a dual-purpose moderate-energy system, single failures of components in other trains (and associated supporting trains) are not assumed.

3F.2.1 EARTHQUAKE ANALYSIS ASSUMPTIONS When evaluating the effects of any earthquake, no other major hazard or event is assumed, and no Seismic Category 1 equipment is assumed to fail as a result of the earthquake. Non-Seismic Category 1 equipment would not be available following the seismic event. Certain non-Seismic Category 1 components are designed and constructed to ensure that their failure could not reduce the functioning of a safe shutdown component to an unacceptable safety level. This criterion meets the intent of Regulatory Guide 1.29, Position C.2. Evaluation of component failure includes drop impact forces and secondary effects, such as spray and flooding from piping failures.

LOSP is assumed following a safe shutdown earthquake (SSE). An earthquake, as a single event, will affect the entire plant; hence, all the rooms dedicated to items associated with either safety-related train are considered in total.

3F.2.2 PIPE BREAK ANALYSIS ASSUMPTIONS All high- and moderate-energy lines whose failure could reduce the functioning of a safe shutdown component to an unacceptable safety level are evaluated for pipe breaks or cracks.

Thrust forces, jet impingement forces, and environmental effects are considered. Section 3.6 provides a description of the location and types of breaks and the forcing functions that are considered for analyzing pipe breaks.

Evaluation of environmental effects of moderate-energy pipe cracks has been made based on the characteristics of the flow from the postulated cracks. The locations of the cracks are discussed in paragraph 3.6.2.1. The evaluations include the effects of spraying or wetting on the safe shutdown equipment to assure that electrical safe shutdown equipment is not affected.

The evaluation also includes the effect of flooding from the worst-case pipe crack in each room or general area. Flooding volumes are based on assuming automatic isolation or operator termination of flow to the pipe failure within a reasonable period after indication of the hazard.

An interval of 30 min for operator's action after indication of flood is assumed.

3F.2.3 MISSILES ANALYSIS ASSUMPTIONS There are two general sources of postulated internally generated missiles outside the containment, rotating component failure and pressurized component failure.

Section 3.5 provides a description of the design bases for the selection of missiles. Tables 3.5.1-1, 3.5.1-2, and 3.5.1-3 provide a listing of major missiles generated within the plant.

3F-2 REV 23 3/21

VEGP-FSAR-3 Analysis of impact from missiles is done for all rotating equipment and high-energy pressurized components.

3F.2.4 FLOODING ANALYSIS ASSUMPTIONS In the event of a pipe failure, significant flooding might result and jeopardize the function of safety-related equipment required to mitigate the consequences of the pipe break or to maintain the plant in a safe shutdown condition.

Flooding rates are based on the worst-case pipe failure in each safety-related room or area.

Through-wall cracks are postulated on moderate energy lines, seismic or non-seismic piping, to substantiate the effects due to spray and flooding on components which are required to mitigate the consequences of the event and/or safely shut down the plant; i.e., essential equipment. On high energy lines, terminal and intermediate breaks are postulated based on the lines' stress analyses. Full ruptures are assumed on all postulated breaks and are analyzed in part to determine flooding effects on essential equipment for the area in which the break is postulated.

The failure mode of non-Seismic Category 1 piping will be a critical crack rather than double-ended rupture, non-Seismic Category 1 piping in areas of safety-related structures where Seismic Category 2 over 1 interactions or flooding from a double-ended rupture could result in unacceptable interactions has been supported to withstand safe shutdown earthquake loads.

The pipe support loads are determined by analyzing the piping system. For SSE loading, with the exception of the piping listed below, pipe stresses are also calculated to demonstrate they are maintained within faulted allowables:

x Duriron lines.

x Copper lines.

x ASTM A-120 galvanized lines.

x Air service lines.

x Process tubing d 3/8 in.

x Instrument tubing.

x Moderate energy, non-Seismic Category 1 piping 2 in. and smaller in Unit 2 only.

For each area the worst flooding source is identified and analyzed for spray and/or flooding effects. The flooding rate is calculated based on the flooding sources reservoir capacity, piping dimensions, and fluid parameters. The level of the flood water is based on automatic isolation or operator action after a reasonable delay time following indication of flow from the breaks or crack. The delay time of 30 min for operator's action after indication of flood is assumed. If detection is not available, the full volume of the reservoir is assumed to flood the affected area.

The maximum flood level is determined by distributing the flood volume over a surface area determined by room size and the size and geometry of large components within the room. Flow paths out of the room; i.e., open doors, grating, stairs, or large openings, are also considered.

The contents of the flooded area are then reviewed to ensure that no essential equipment is adversely affected. Additionally, flooding communication to adjacent rooms is evaluated to ensure that the event does not result in failure of essential components in adjacent rooms.

The analysis for flooding caused by failure of non-Seismic Category 1 tanks inside Category 1 buildings assumes each tank to completely rupture and empty into adjacent areas. The effects of open doors, stairs, and large openings are considered in the analysis. The tank's contents are then divided by the established boundaries' surface area to determine a flood level. The 3F-3 REV 23 3/21

VEGP-FSAR-3 contents of the flooded area are then reviewed to ensure that no essential equipment is adversely affected. The analysis considers the effects of the worst single active failure taken concurrent with the break. The effects of flooding from non-Seismic Category 1 tanks in the outside areas are discussed in paragraph 3.4.1.1.2.

Potential flooding due to probable maximum flood and probable maximum precipitation is discussed in section 3.4.

The equipment and floor drainage system is discussed in subsection 9.3.3. All water, released because of pipe breaks in the auxiliary building, drains to the common sump. Refer to paragraph 9.3.3.2.2.3 for a discussion of this design.

3F.3 PROTECTION MECHANISMS The plant layout arrangement is based on maximizing the physical separation of redundant or diverse safety-related components and systems from each other and from nonsafety- related items. Therefore, if an accident occurs within the plant, there is minimal effect on other systems or components which are required for safe shutdown of the plant or to mitigate the consequence of the hazard.

Since it is not always feasible to provide separation in every hazard situation, other protection features are employed. These protection features include the following:

x Structural enclosures.

x Structural barriers.

x Restraints.

x Seismically designed components.

x Hardening.

x Orientation.

3F.4 HAZARDS EVALUATIONS As stated previously, table 3F-1 provides a hazards evaluation of level C and safety-related pump rooms on levels B and D of the auxiliary building. Each room on those elevations is shown in drawings AX1D08A03-3, AX1D08A03-4, AX1D08A31, AX1D08A02-3, AX1D08A02-2 and AX1D08A04-4 and has been reviewed to ensure that the integrated design of the plant acceptably addresses all postulated hazards. Since the evaluations for equipment and components in all safety-related areas are documented in the project files and are available for audit, they are not provided in the FSAR.

Specific evaluations of certain areas of the plant have been of licensing concern in the past.

These evaluations are provided in the following subsections.

3F.4.1 AUXILIARY FEEDWATER (AFW) PUMP ROOMS The effects of a pipe break in the AFW pump rooms have been evaluated and the results of the Unit 1 analysis are presented in this appendix. The Unit 2 analysis is similar. The effects include room pressurization, temperature, flooding, and operability of the AFW system.

There are three separate AFW pump rooms, each housing one pump. Drawing 1X6DD300 provides plan and elevation views and nodal boundary of this area. Each of two motor-driven pumps is sized to deliver the feedwater flow required for decay heat removal. The single 3F-4 REV 23 3/21

VEGP-FSAR-3 turbine-driven pump supplies twice the capacity of a motor-driven pump and is sufficient to remove decay heat and, additionally, to cool down the reactor at a rate of up to 100qF/h not to exceed 100qF in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period. The turbine-driven pump provides system diversity to both motor-driven pumps.

The results of the pressure/temperature analysis for the Unit 1 turbine-driven AFW pump room, as shown in table 3F-2, indicate that the maximum differential pressure on the walls will be less than 1.0 psi. The walls are capable of withstanding this pressure.

Analysis of AFW piping failures shows that loss of a redundant train does not prevent decay heat removal. The capability to provide adequate feedwater flow to remove decay heat is ensured by operation of either one of two motor-driven pumps or by operation of the turbine-driven pump.

Pressure-temperature response analysis for design basis main steam and main feedwater pipe breaks in the main steam tunnel has been performed. Design basis pipe breaks and locations are selected to identify maximum tunnel pressures to be accommodated by the AFW structure adjacent to the main steam tunnel.

The maximum flooding occurs within the AFW pumphouse when considering a single failure of a check valve that results in backflow from the turbine-driven AFW pump. In this scenario, flooding will propagate to the adjacent motor-driven pump room. However, no failure of the turbine-driven pump is postulated, and the system is capable of providing the required flow. For all other cases, flooding is contained such that it will not propagate to the other pumps.

Analysis of the other hazards shows that adequate redundancy and separation are provided to ensure the operability of at least one train of the AFW system.

3F.4.2 MAIN STEAM ISOLATION VALVE (MSIV) AND MAIN FEEDWATER ISOLATION VALVE (MFIV) COMPARTMENT The MSIV/MFIV compartment is located in the wing areas of the auxiliary building and the control building. Drawings 1X6DD301, 1X6DD302, AX6DD300, AX6DD301, and AX6DD302 provide plan and elevation views of this area. The main steam and main feedwater piping in this area consists of straight piping runs extending from the containment penetrations to torsional restraints mounted in the auxiliary building and control building walls through which these lines enter the main steam tunnel. The MSIVs, main steam safety valves, atmospheric relief valves, and MFIVs are in this compartment. Also in the compartment are branch piping lines of the AFW system, chemical addition system, steam supply to the turbine-driven AFW pump, bypass loops of the MSIVs, pressure instrumentation, and drains.

3F.4.2.1 Break Size and Location Main steam and feedwater piping in this compartment is designed to the criteria stated in paragraph 3.6.2.1 for those portions of the piping passing through the primary containment and extending to the first pipe whip restraint past the first outside isolation valve. In accordance with these criteria, no specific pipe breaks are postulated in the main run of these lines in the MSIV/MFIV compartment. However, to provide an additional level of assurance of operability of safety-related equipment in this compartment, the building structure and safety-related equipment are designed for the environmental conditions (pressure, temperature, and flooding) that would result from a break, equal in area to one cross-sectional pipe area, of either a main steam line or main feedwater line without steam generator tube bundle being uncovered. Main steam line breaks up to 1.0 ft2 with steam generator tube bundle being uncovered have been considered for equipment qualification as discussed in paragraph 3.11.B.1.1.

3F-5 REV 23 3/21

VEGP-FSAR-3 Pressurization of the MSIV/MFIV compartment due to such a rupture is limited by providing adequate venting of the compartment and designing the compartment to withstand the maximum resultant pressure. Venting is accomplished by including adequate passageways between compartments or by other acceptable venting schemes. Engineered safety features required to bring the reactor to safe shutdown, which are located within these compartments, are designed to withstand the associated temperature, pressure, and humidity conditions.

The following cases are analyzed to determine the worst environmental conditions for the MSIV/MFIV compartment:

A. Case 1: Saturated steam blowdown from a main steam line break (MSLB) equivalent to the flow area of a single area rupture inside the restraint wall and a double area rupture outside the restraint wall. This case results in the maximum compartment pressure.

B. Case 2: Blowdown from a main feedwater line break equivalent to the flow area of a single area rupture. This case results in the maximum compartment flood level.

C. Case 3: Blowdown from MSLBs up to 1.0 ft2 equivalent flow area with steam generator tube bundle being uncovered producing superheated blowdown. This case results in the maximum compartment temperatures and is discussed in paragraph 3.11.B.1.1.

3F.4.2.2 Method of Analysis The case 1 analysis was performed using the Bechtel COPDA computer code, which is described in reference 1. The case 2 analysis was performed using the fluid flow equations identified in reference 2 for cold water flow. The case 3 analysis was performed using the GOTHIC computer code which is described in reference 4.

The MSIV area MSLB environmental conditions and the resulting impact on equipment qualification were evaluated using both the COPDA and GOTHIC computer codes. COPDA modeled the control and auxiliary sides of the MSIV/MFIV vault areas. Because reference 3 indicates the transient temperature for the auxiliary building MSIV/MFIV vault area is extremely similar to the transient for the same break in the control building, the detailed GOTHIC model was performed only on the control building MSIV/MFIV vault area. This area has compartments with smaller volumes and less flow area out of the break compartment than the auxiliary building MSIV/MFIV vault area (reference 3).

3F.4.2.3 Mass and Energy Release for Main Steam Line Break For cases 1 and 2, blowdown mass and energy releases were calculated for a single area rupture inside the restraint wall and a double area rupture outside the restraint wall for both the auxiliary and control building MSIV/MFIV areas. The blowdown data for the double area rupture outside the restraint wall of the auxiliary building MSIV/MFIV area (Node 10) are presented in table 3F-3A.

For case 3, a spectrum of blowdown mass and energy releases was calculated for a spectrum of break sizes and power levels. The blowdown data for a 1.0 ft2 single area rupture at 102%

power (102% of 3579 MWt) inside the restraint wall in the control building MSIV/MFIV area are presented in table 3F-3B.

3F-6 REV 23 3/21

VEGP-FSAR-3 3F.4.2.4 Compartment Volumes and Vent Areas For the Unit 1 analyses of the pressure temperature-transient following a MSLB (cases 1 and 3),

the flow model of control volumes and intercompartment flow paths are illustrated in figures 3F-1 and 3F-2 and listed in tables 3F-3A and 3F-3B. The calculated Unit 1 flooding results from the case 2 analysis are shown in table 3F-3A. The calculated Unit 1 compartment pressure (case

1) and temperature (case 3) responses are listed in table 3F-3A and table 3F-3B. The Unit 2 analysis is similar.

3F.4.2.5 Initial Conditions Tables 3F-3A and 3F-3B provide the Unit 1 initial conditions for case 1, case 2, and case 3 analyses.

3F.4.2.6 Design Provisions The Plot of the time-history of the Unit 1 maximum break node compartment composite temperature for case 3 is given in figure 3F-3. The calculated Unit 1 flooding results from the case 2 analysis are shown in table 3F-3A.

Table 3F-3A provides the peak transient values of the Unit 1 compartment pressure analyses.

Table 3F-3B provides the peak transient values of the Unit 1 compartment temperature analyses. The MSIV/MFIV compartment is designed to withstand these conditions. Essential safety-related equipment is qualified to these environmental conditions as discussed in section 3.11.

3F.4.3 EVALUATION OF REACTOR COOLANT SYSTEM (RCS) LOOP BRANCH LINE BREAKS The evaluation of effects on safety-related equipment resulting from branch line breaks in the RCS is presented in table 3F-4. The evaluation shows that breaks in the RCS will not compromise the capability to safely shut down the plant.

3F.4.4 TURBINE BUILDING FLOODING EVALUATION The flooding effects of a pipe break in the turbine building have been evaluated. Although the turbine building does not contain any essential safety-related equipment (however, see paragraph 7.2.1.1.2.F for trip instrumentation in the turbine building), it is connected to other safety-related structures by piping and electrical tunnels. However, it has been shown that the propagation of flood waters into the safety-related structures is precluded by design.

The design basis pipe failure in the turbine building is postulated to be a full circumferential break in the 96-in. circulating water piping to the condensers. A complete rupture of the condenser riser expansion joints has not been postulated as such joints have not been implemented in the VEGP design. The circulating water system design pressure is 80 psig, with a normal anticipated operating pressure of 45 psig. The maximum transient analysis pressure (unscheduled shutdown) is 54 psig. The flowrate through the break is conservatively estimated to be 612,000 gal/min, which is the combined runout flow of two circulating water pumps. This flowrate will cause the water level in the turbine building to rise at a rate of 0.72 ft/min.

If it is assumed that the turbine building flood detectors and the circulating water basin level detectors fail, the flooding could continue until the basin is empty. As the water level rises in one unit, the concrete block wall between units will fail due to high static water pressure, allowing the water to spread into both halves of the building. Calculations have shown that the 3F-7 REV 23 3/21

VEGP-FSAR-3 block wall will not withstand more than 8 ft of water and will fail before the water level reaches the operating deck at el 220 ft 0 in. If the entire circulating water system volume of 1.2 x 106 ft3 is pumped into the turbine building, the resulting flood level in both units would be 211 ft.

Should flood water enter the main steam tunnels at the east or west end of the turbine building, it is precluded from entering the control and auxiliary buildings by sealed penetrations in the walls of the safety-related structures. The piping and electrical tunnels along the south wall at the centerline of the turbine building extend above grade level. Penetrations into the tunnels below grade are sealed.

A failure of the condensate feedwater system piping would result in significantly less flooding than a circulating water system failure. Even if the entire condensate/feedwater inventory flooded into the turbine building, the final flood level would be less than 1 ft above the level A floor.

3F.5 REFERENCES

1. Braddy, R.W. and Thiesing, J.W., Subcompartment Pressure and Temperature Transient Analysis, BN-TOP-4, Revision 0, Bechtel Corporation, San Francisco, California, July 1976.

2. Design for Pipe Break Effects, BN-TOP-2, Revision 2, Bechtel Corporation, San Francisco, California, June 1974.

3. Spryshak, J. J. and Iyengar, J., "Vogtle Electric Generating Plant Units 1 and 2. Main Steam Isolation Valve Area Equipment Thermal Response to Superheated Steam Releases," WCAP-13169, December 1991.

4. NAI 8907-02, Revision 8, GOTHIC Containment Analysis Package User Manual.

3F-8 REV 23 3/21

VEGP-FSAR-3F TABLE 3F-1 (SHEET 1 OF 84)

SAMPLE OF HAZARDS ANALYSIS RESULTS FOR AUXILIARY BUILDING (LEVELS B, C, AND D)

Room Number R-B15 Title Safety Injection Pump Room Train A Remarks Flooding Analysis X Flooding from sources within the room, coincident with the most limiting single 1. The high-energy line breaks have been evaluated and will not preclude mitigation of active failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G) the event or safe shutdown of the plant.

will not preclude mitigation of the event or safe shutdown of the plant.

X Flooding from sources external to the room is not credible even with a single active failure.

Other, see remarks.

Seismic Design Analysis Only safety-related equipment (SRE) is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The nonsafety-related equipment (NSRE) in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

X Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 2 OF 84)

Listing of safety-related items in room R-B15 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Safety injection pump suction header isolation valve HV-8807 A Y

2. Safety injection pump train A miniflow valve HV-8814 Y

3. SIP A to RCS Cold Leg Isolation valve HV-8821 A Y

4. Safety injection pump P6003 inlet valve HV-8923 A Y

5. Safety injection pump room leak detection switch LSH-9812 Y

6. Safety injection pump room leak detection switch LSH-9816 Y

7. Charging pump header valve HV-8924 Y Y

8. Cable tray Train A Y Y

9. Safety injection pump 1-1204-P6-003 Y 10 Air-handling units 1-1555-A7-015-000 Y

11. Piping 1-1204-003-1 1/2 in. Y Y 1-1204-008-6 in. Y Y 1-1202-117-3 in. Y Y 1-1208-411-6 in. Y Y 1-1204-011-6 in. Y Y 1-1202-118-2 in. Y Y 1-1202-116-2 in. Y Y 1-1204-028-4 in.

1-1204-014-4 in.

1-1202-115-3 in.

1-1202-113-2 in.

1-1202-115-1 in. Y Y 1-1202-117-1 in. Y Y 1-1202-281-3/4 in. Y Y 1-1202-297-1 in. Y Y 1-1202-298-3/4 in. Y Y 1-1202-372-2 in. Y Y 1-1202-373-2 in. Y Y 1-1202-475-2 in. Y Y 1-1202-476-2 in. Y Y 1-1204-008-3/4 in. Y Y 1-1204-015-4 in. Y Y 1-1204-028-3/4 in. Y Y 1-1204-052-2 in. Y Y 1-1204-128-3/4 in. Y Y 1-1204-129-3/4 in. Y Y 1-1204-130-3/4 in. Y Y 1-1204-131-3/4 in. Y Y 1-1204-280-1/2 in. Y Y 1-1555-066-1 1/2 in. Y Y 1-1609-031-3/4 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 3 OF 84)

Listing of safety-related items in room R-B15 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

12. Safety injection pump discharge, flow transmitter FT-0918 Y

13. Safety injection pump outlet, relief valve PSV-8853A Y

14. Safety injection pump inlet, relief valve PSV-8858 Y

15. Safety injection pump room cooler, high temperature TISH-12210 Y switch

16. Train A raceway Conduit and cabletrays Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 4 OF 84)

Room Number R-B18 Title Valve Gallery Remarks Flooding Analysis Flooding from sources within the room coincident with the most limiting single active 1. Flooding from sources located in this room will not impair the safe shutdown capability failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not of the SRE.

preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis There are no high-energy lines in the room.

X The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 5 OF 84)

Listing of safety-related items in room R-B18 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Piping 1-1202-162-3 in. Y Y 1-1204-010-6 in. Y Y 1-1204-004-3/4 in. Y Y 1-1204-008-8 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 6 OF 84)

Room Number R-B19 Title Safety Injection Pump Room Train B Remarks Flooding Analysis X Flooding from sources within the room, coincident with the most limiting single active 1. The high-energy line breaks have been evaluated and will not preclude mitigation of failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not the event or safe shutdown of the plant.

preclude mitigation of the event or safe shutdown of the plant.

X Flooding from sources external to the room is not credible even with a single active failure.

Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 7 OF 84)

Listing of safety-related items in room R-B19 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Safety injection pump suction header valve HV-8807B Y

2. Safety injection pump miniflow isolation train B valve HV-8813 Y

3. SIP B to RCS Cold Leg Isolation valve HV-8821B Y

4. Safety injection pump inlet and miniflow isolation HV-8920 Y valve

5. Safety injection pump P6004 inlet valve HV-8923B Y

6. Safety injection pump room leak detection switch LSH-9813 Y

7. Safety injection pump room leak detection switch LSH-9817 Y

8. Cooler motor control switch TISH 12216 Y

9. Train B raceway Conduits and cabletrays Y Y

10. Safety injection pump 1-1204-P6-004 Y

11. Air-handling unit 1-1555-A7-016-000 Y

12. Piping 1-1204-004-3 in. Y Y 1-1202-381-2 in. Y Y 1-1202-160-3 in. Y Y 1-1202-162-3 in. Y Y 1-1202-163-2 in. Y Y 1-1204-004-1 1/2 in. Y Y 1-1204-010-6 in. Y Y 1-1204-012-8 in. Y Y 1-1204-015-4 in. Y Y 1-1202-161-2 in. Y Y 1-1592-059-1 1/2 in. Y Y 1-1592-050-1 1/2 in. Y Y 1-1202-160-1 in. Y Y 1-1202-289-3/4 in. Y Y 1-1202-313-1 in. Y Y 1-1202-380-2 in. Y Y 1-1202-481-2 in. Y Y 1-1204-002-3 in. Y Y 1-1204-004-3/4 in. Y Y 1-1204-004-1 1/2 in. Y Y 1-1204-011-6 in. Y Y 1-1204-015-3/4 in. Y Y 1-1204-037-4 in. Y Y 1-1204-280-1/2 in. Y Y 1-1204-281-1/2 in. Y Y 1-1204-314-3/4 in. Y Y 1-1555-074-1 1/2 in. Y Y 1-1609-030-3/4 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 8 OF 84)

Listing of safety-related items in room R-B19 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

13. Safety injection pump discharge, flow transmitter FT-0922 Y

14. Safety injection pump inlet, relief valve PSV-8853B Y

15. Safety injection pump room cooler, high temperature TISH-12211 Y switch.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 9 OF 84)

(Recycle Evaporator Abandoned in Place)

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 10 OF 84)

Listing of safety-related items in room R-C64, C65 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Piping 1-1228-174-in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 11 OF 84)

(Waste Evaporator Abandoned in Place)

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 12 OF 84)

Listing of safety-related items in room R-C66, C67 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Piping 1-1228-168-2 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 13 OF 84)

Room Number R-C77 Title Boron Recycle Holdup Tank Room Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. Flooding from sources located in this room will not impair the safe shutdown capability failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not of the SRE.

preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 14 OF 84)

Listing of safety-related items in room R-C77 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Boron recycle holdup tank A-1210-T4-001 Y

2. Piping 1-1208-151-3/4 in. Y Y 1-1208-437-3/4 in. Y Y 1-1208-493-3/4 in. Y Y A-1561-102-10 in. Y Y 1-1561-165-10 in. Y Y Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 15 OF 84)

Room Number R-C83 Title Corridor Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. The SRE located in this room is not required for safe shutdown.

failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remark 1.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 16 OF 84)

Listing of safety-related items in room R-C83 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Recycle holdup tank/heating, ventilation, and air- HV-12596 N Y conditioning (HVAC) vent valve

2. Recycle holdup tank/HVAC vent valve HV-12597 N Y

3. Train A raceway Conduits and cabletrays N Y

4. Train B raceway Conduits and cabletrays N Y

5. Piping 1-1208-017-2 in. Y Y 2-1208-017-2 in. Y Y 1-1208-020-2 in. Y Y 1-1208-090-3 in. Y Y 1-1208-093-3 in. Y Y 1-1208-095-3 in. Y Y 1-1208-340-3/8 in. Y Y 1-1208-410-3 in. Y Y 1-1208-437-3/8 in. Y Y 1-1208-493-3/4 in. Y Y 1-1213-026-3 in. Y Y 1-1224-005-1 1/2 in. Y Y 1-1224-006-4 in. Y Y 1-1224-008-3 in. Y Y 1-1224-011-2 in. Y Y 1-1224-013-2 in. Y Y 1-1224-014-2 in. Y Y 1-1224-015-2 in. Y Y 1-1224-016-2 in. Y Y 1-1224-026-3 in. Y Y 1-1224-027-3 in. Y Y 1-1224-028-3 in. Y Y 1-1224-066-3 in. Y Y 1-1407-021-2 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 17 OF 84)

Room Number R-C88 Title Electrical Chase Train A Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. The safe shutdown capability of the SRE located in this room is not impaired by this failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not flood level.

preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 18 OF 84)

Listing of safety-related items in room R-C88 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train C Raceway Conduits and cabletrays Y Y

2. Piping 1-1202-104-8 in. Y Y 1-1202-134-8 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 19 OF 84)

Room Number R-C89 Title Storage Room Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. Flooding from sources located in this room will not impair the safe shutdown capability failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not of the SRE.

preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 20 OF 84)

Listing of safety-related items in room R-C89 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train D raceway Conduits and cabletrays Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 21 OF 84)

Room Number R-C90 Title Residual Heat Removal (RHR) Exchanger Room Train A Remarks Flooding Analysis X Flooding from sources within the room, coincident with the most limiting single active failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.1.1.6.G), will not preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 22 OF 84)

Listing of safety-related items in room R-C90 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train A and B raceway Conduits and cabletrays Y Y

2. RHR heat exchanger room leak detection switch LSH 9874 Y Y

3. RHR miniflow valve FV 0610 Y Y

4. RHR heat exchanger to pump valve HV 8804A Y Y

5. RHR exchanger 1-1205-E6-001 Y Y

6. Piping 1-1205-007-8 in. Y Y 1-1205-007-2 in. Y Y 1-1205-013-3/4 in. Y Y 1-1205-013-3 in. Y Y 1-1208-411-8 in. Y Y 1-1205-46-3/4 in. Y Y 1-1205-005-8 in. Y Y 1-1203-021-18 in. Y Y 1-1203-021-14 in. Y Y 1-1205-033-3/8 in. Y Y 1-1203-122-1 in. Y Y 1-1203-088-1 in. Y Y 1-1203-086-1 in. Y Y 1-1205-005-8 in. Y Y 1-1205-008-8 in. Y Y 1-1205-012-2 in. Y Y 1-1205-014-3/4 in. Y Y 1-1205-014-3 in. Y Y 1-1205-022-3/4 in. Y Y 1-1205-029-3/4 in. Y Y 1-1205-030-2 in. Y Y 1-1205-070-1/2 in. Y Y

7. HVAC ducts Train A Y Y

8. Residual heat exchanger outlet valve HV-0606 Y Y

9. Residual heat removal LP return bypass valve FV-0618 Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 23 OF 84)

Room Number R-C91 Title RHR Heat Exchanger Room Train B Remarks Flooding Analysis X Flooding from sources within the room, coincident with the most limiting single active failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 24 OF 84)

Listing of safety-related items in room R-C91 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train A and B raceways Conduits and cabletrays Y Y

2. RHR pump to miniflow valve FV 0611 Y Y

3. RHR heat exchanger room leak detection switch LSH 9855 Y Y

4. RHR exchanger 1-1205-E6-002 Y Y

5. Conduit - Y Y

6. Piping 1-1205-014-3/4 in. Y Y 1-1205-014-3 in. Y Y 1-1203-042-18 in. Y Y 1-1203-042-14 in. Y Y 1-1205-006-8 in. Y Y 1-1204-012-8 in. Y Y 1-1203-114-1 in. Y Y 1-1203-116-1 in. Y Y 1-1205-034-3/8 in. Y Y 1-1205-045-3/4 in. Y Y 1-1205-008-2 in. Y Y 1-1205-008-8 in. Y Y 1-1203-071-3/4 in. Y Y 1-1203-075-3/4 in. Y Y 1-1203-084-3/4 in. Y Y 1-1203-111-3/4 in. Y Y 1-1205-005-8 in. Y Y 1-1205-007-8 in. Y Y 1-1205-012-2 in. Y Y 1-1205-018-1 in. Y Y 1-1205-021-1 in. Y Y 1-1205-029-3/4 in. Y Y 1-1205-031-1 in. Y Y 1-1205-032-1 in. Y Y 1-1205-035-1 in. Y Y 1-1205-036-1 in. Y Y 1-1205-037-1 in. Y Y 1-1205-038-1 in. Y Y 1-1205-043-1 in. Y Y 1-1205-044-1 in. Y Y 1-1205-051-1 in. Y Y 1-1205-052-1 in. Y Y 1-1205-053-1 in. Y Y 1-1205-054-1 in. Y Y 1-1205-056-1 in. Y Y 1-1205-057-1 in. Y Y 1-1205-058-1 in. Y Y 1-1205-059-1 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 25 OF 84)

Listing of safety-related items in room R-C91 Item Equipment Essential Seismic Number Description Designation Equipment Category 1 1-1205-072-1/2 in. Y Y 1-1205-073-1/2 in. Y Y 1-1205-074-1/2 in. Y Y 1-1205-075-1/2 in. Y Y

7. HVAC ducts Train B Y Y

8. Residual heat exchanger outlet valve HV-0607 Y Y

9. Residual heat removal LP return bypass valve FV-0619 Y Y

10. Residual heat removal exchanger train B to H HV-8804B Y Y injection pump valve REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 26 OF 84)

Room Number R-C95 Title Pipe Chase Train B Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. Flooding from sources located in this room will not impair the safe shutdown capability failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not of the SRE.

preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remark 1.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis There are no high-energy lines in the room.

X The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 27 OF 84)

Listing of safety-related items in room R-C95 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train C raceway Conduits and cableways Y Y

2. Piping 1-1208-411-8 in. Y Y 1-1204-012-8 in. Y Y 1-1205-008-8 in. Y Y 1-1206-006-8 in. Y Y 1-1204-002-3 in. Y Y 1-1204-037-4 in. Y Y 1-1204-221-2 in. Y Y 1-1208-003-3 in. Y Y 1-1208-022-2 in. Y Y 1-1208-095-3 in. Y Y 1-1208-097-3 in. Y Y 1-1208-103-3 in. Y Y 1-1208-123-4 in. Y Y 1-1208-131-2 in. Y Y 1-1208-146-3 in. Y Y 1-1208-410-3 in. Y Y 1-1208-485-1 in. Y Y 1-1208-493-3/4 in. Y Y 1-1213-014-2 in. Y Y 1-1213-017-2 in. Y Y 1-1213-043-3 in. Y Y 1-1407-007-4 in. Y Y 1-1407-034-4 in. Y Y 1-1561-103-10 in. Y Y

3. HVAC ducts Train A Y

4. HVAC ducts Train B Y

5. Temperature elements TE-19723E Y TE-19722E REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 28 OF 84)

Room Number R-C99 Title Nonradioactive Pipe Chase Remarks

1. The flooding from sources located in this room will not impair the safe shutdown Flooding Analysis capability of the SRE.

Flooding from sources within the room, coincident with the most limiting single active failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not preclude mitigation of the event or safe shutdown of the plant.

X Flooding from sources external to the room is not credible even with a single active failure.

Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 29 OF 84)

Listing of safety-related items in room R-C99 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train B raceway Conduits and cabletrays Y Y

2. Essential chilled water 1-1592-020-2 1/2 in. piping Y Y 1-1592-020-3 in. Y Y 1-1592-031-2 1/2 in. Y Y 1-1592-031-3 in. Y Y 1-1592-052-1 1/2 in. Y Y 1-1592-057-1 1/2 in. Y Y 1-1592-044-2 1/2 in. Y Y 1-1592-054-2 1/2 in. Y Y 1-1592-026-2 in. Y Y 1-1592-038-2 in. Y Y A-1228-166-2 in. Y Y 1-1228-166-2 in. Y Y 1-1228-170-1 in. Y Y 1-1228-174-2 in. Y Y 1-1228-174-3 in. Y Y A-1561-103-10 in. Y Y 1-1592-044-2 in. Y Y 1-1592-054-2 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 30 OF 84)

Room Number R-C99 Title Nonradioactive Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. Safety-related function is not impaired by flooding or water spray.

failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 31 OF 84)

Listing of safety-related items in room R-C103 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Piping 1-1205-003-14 in. Y Y 1-1205-010-12 in. Y Y 1-1205-007-8 in. Y Y 1-1205-005-8 in. Y Y

2. Train A raceway Conduits and cabletrays Y Y

3. HVAC ducts Train A Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 32 OF 84)

Room Number R-C104 Title Electrical Chase Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. The flooding from sources located in this room will not impair the safe shutdown failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not capability of the SRE.

preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis X Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 33 OF 84)

Listing of safety-related items in room R-C104 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train A and C raceway Conduits and cabletrays Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 34 OF 84)

Room Number R-C105 Title Pipe Penetration Room Train A Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. Safety-related function is not impaired by flooding or water spray.

failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 35 OF 84)

Listing of safety-related items in room R-C105 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. HVAC ducts Train A Y

2. Containment sump isolation valve HV-8811A Y

3. Containment spray pump No. 1 suction HV-9002A Y

4. Containment spray pump No. 2 suction HV-9003A Y

5. Train D raceway Conduits and cabletrays Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 36 OF 84)

Room Number R-C107 Title Vestibule Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. Safety-related function is not impaired by flooding or water spray.

failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 37 OF 84)

Listing of safety-related items in room R-C107 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. HVAC ducts Train A Y

2. Temperature elements TE-15212C Y TE-15216C Y

3. Train D raceway Conduits and cabletrays Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 38 OF 84)

Room Number R-C108 Title Steam Generator Blowdown Heat Exchanger Room Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. Safety-related function is not impaired by flooding or water spray.

failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not 2. There is no essential equipment for safe shutdown located in this room; preclude mitigation of the event or safe shutdown of the plant. therefore, the postulated hazards have no effect on the safe shutdown Flooding from sources external to the room is not credible even with a single active capability.

failure.

X Other, see remark 1.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

X Other, see remark 2.

Missile Analysis No credible missile sources exist in the room.

X Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis There are no high-energy lines in the room.

X The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 39 OF 84)

Listing of safety-related items in room R-C108 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train C raceway Conduits and cabletrays Y Y

2. Piping 1-1206-001-12 in. Y 1-1305-067-2 1/2 in. Y Y 1-1305-068-2 1/2 in. Y Y 1-1305-080-3 in. Y Y 1-1305-080-4 in. Y Y 1-1305-080-6 in. Y Y 1-1305-098-3 in. Y Y 1-1305-098-4 in. Y Y 1-1305-099-3 in. Y Y 1-1305-107-1 1/2 in. Y Y 1-1305-107-3 in. Y Y 1-1305-108-2 1/2 in. Y Y 1-1305-109-2 1/2 in. Y Y 1-1407-001-3 in. Y Y 1-1407-002-3 in. Y Y 1-1407-003-1 in. Y Y 1-1407-003-3 in. Y Y 1-1407-004-3 in. Y Y 1-1407-005-1 in. Y Y 1-1407-006-3 in. Y Y 1-1407-015-3 in. Y Y 1-1407-029-1 in. Y Y 1-1407-035-4 in. Y Y 1-1407-038-4 in. Y Y 1-1407-039-4 in. Y Y 1-1407-062-3 in. Y Y 1-1407-066-3 in. Y Y 1-1407-066-4 in. Y Y 1-1407-067-3 in. Y Y 1-1407-098-1 1/2 in. Y Y 1-1407-124-3/4 in. Y Y 1-1407-125-3/4 in. Y Y 1-1407-126-3/4 in. Y Y 1-1407-127-3/4 in. Y Y 1-1407-128-3/4 in. Y Y 1-1407-129-3/4 in. Y Y 1-1407-130-3/4 in. Y Y 1-1407-131-3/4 in. Y Y 1-1407-132-3/4 in. Y Y 1-1592-026-1 in. Y Y 1-1592-038-1 in. Y Y

3. Temperature elements TE-15212D Y TE-15216D Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 40 OF 84)

Room Number R-C109 Title Motor Control Center (MCC) Room Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. Safety related function is not impaired by flooding or water spray.

failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis X Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 41 OF 84)

Listing of safety-related items in room R-C109 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train A raceway Conduits and cabletrays Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 42 OF 84)

Room Number R-C111 Title Chemical and Volume Control System (CVCS)

Normal Charging Pump Room Remarks Flooding Analysis X Flooding from sources within the room, coincident with the most limiting single active failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G) will not preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis No credible missile sources exist in the room.

X Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis There are no high-energy lines in the room.

X The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 43 OF 84)

Listing of safety-related items in room R-C111 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. CVCS normal charging pump 1-1208-P4-001 N Y

2. Piping 1-1204-037-4 in. Y Y 1-1204-168-3/4 in. Y Y 1-1208-002-3 in. Y Y 1-1208-003-2 in. Y Y 1-1208-016-3 in. Y Y 1-1208-123-4 in. Y Y 1-1208-140-3/4 in. Y Y 1-1208-142-3/4 in. Y Y 1-1208-187-3/4 in. Y Y 1-1208-189-3/4 in. Y Y 1-1208-190-3/4 in. Y Y 1-1208-280-1/2 in. Y Y 1-1208-411-6 in. Y Y 1-1208-440-2 in. Y Y 1-1208-482-3/4 in. Y Y 1-1208-483-3/4 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 44 OF 84)

Room Number R-C113 Title Boric Acid Batching Tank Room Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. Flooding from sources located in this room room will not impair the safe shutdown failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not capability of the SRE.

preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 45 OF 84)

Listing of safety-related items in room R-C113 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train A raceway Conduits and cabletrays Y Y

2. Piping 1-1202-122-1 1/2 in. Y Y 1-1202-122-3 in. Y Y 1-1202-125-2 in. Y Y A-1208-232-3 in. Y Y 1-1202-125-1 1/2 in Y Y 1-1202-124-3 in. Y Y 1/2-1208-241-3 in. Y Y A-1208-034-3 in. Y Y A-1208-230-1 in. Y Y A-1208-230-3/4 in. Y Y 1-1202-123-2 in. Y Y 1-1202-124-1 1/2 in. Y Y 1-1202-300-3/4 in. Y Y 1-1202-473-2 in. Y Y 1-1202-474-2 in. Y Y 1-1204-177-8 in. Y Y 1-1208-137-8 in. Y Y 1-1208-241-3 in. Y Y A-1208-492-3 in. Y Y A-1228-166-2 in. Y Y A-1228-166-1/2 in. Y Y A-1228-228-3 in. Y Y A-1609-034-3/4 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 46 OF 84)

Room Number R-C114 Title Valve Gallery Remarks Flooding Analysis X Flooding from sources within the room, coincident with the most limiting single active failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis There are no high-energy lines in the room.

X The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 47 OF 84)

Listing of safety-related items in room R-C114 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train A raceway Conduits and cabletrays Y Y

2. Refuel water tank to charging pump valve LV 0112D Y Y

3. Piping 1-1202-122-1 1/2 in. Y Y 1-1202-123-2 in. Y Y 1-1202-124-1 1/2 in. Y Y 1-1202-125-2 in. Y Y 1-1592-027-1 1/2 in. Y Y 1-1592-037-1 1/2 in. Y Y 1-1208-141-6 in. Y Y 1-1208-144-4 in. Y Y 1-1203-101-2 in. Y Y 1-1203-144-4 in. Y Y 1-1208-157-8 in. Y Y 1-1208-095-3 in. Y Y 1-1208-101-2 in. Y Y 1-1208-137-8 in. Y Y 1-1208-139-8 in. Y Y 1-1208-147-3 in. Y Y 1-1208-411-8 in. Y Y 1-1208-485-1 in. Y Y 1-1208-497-2 in. Y Y 1-1208-498-2 in. Y Y 1-1208-501-2 in. Y Y

4. Charging pump miniflow isolation valve HV-8111A Y Y

5. Charging pump A suction valve HV-8471A Y Y

6. Charging pump A discharge valve HV-8485A Y Y

7. Charging pump miniflow isolation to RWST valve HV-8508A, 8509B Y Y

8. Normal charging pump miniflow isolation valve 1-1208-U4-150 Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 48 OF 84)

Room Number R-C115 Title Centrifugal Charging Pump Room Train A Remarks Flooding Analysis X Flooding from sources within the room, coincident with the most limiting single active failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis No credible missile sources exist in the room.

X Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis There are no high-energy lines in the room.

X The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 49 OF 84)

Listing of safety-related items in room R-C115 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train A raceway Conduits and cabletrays Y Y

2. CVCS centrifugal charging pump 1-1208-P6-002 Y Y

3. Charging pump room leak detection switch LSH 9826 Y

4. Charging pump room leak detection switch LSH 9830 Y

5. Air-handling unit 1-1555A7-013 Y Y

6. Piping 1-1202-122-1 1/2 in. Y Y 1-1202-123-2 in. Y Y 1-1202-124-1 1/2 in. Y Y 1-1202-125-2 in. Y Y 1-1202-370-2 in. Y Y 1-1202-371-2 in. Y Y 1-1208-141-6 in. Y Y 1-1208-144-4 in. Y Y 1-1592-027-1 1/2 in. Y Y 1-1592-037-1 1/2 in. Y Y 1-1204-012-8 in. Y Y 1-1208-132-2 in. Y Y 1-1208-191-1 in. Y Y 1-1208-197-3/4 in. Y Y 1-1208-198-1 in. Y Y 1-1555-101-1 1/2 in. Y Y 1-1555-134-1 1/2 in. Y Y

7. HVAC ducts Train A Y

8. HVAC ducts Train B Y

9. Charging pump room cooler temperature element TE-12209 Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 50 OF 84)

Room Number R-C116 Title Vestibule Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. Safety-related equipment is not impaired by flooding or water spray.

failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis There are no high-energy lines in the room.

X The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 51 OF 84)

Listing of safety-related items in room R-C116 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Piping 1-1202-003-4 in. Y Y

2. Train A and C raceway 1-1202-004-4 in. Y Y 1-1202-004-3 in. Y Y 1-1202-003-3 in. Y Y 1-1202-122-3 in. Y Y 1-1202-124-3 in. Y Y 1-1305-080-6 in. Y Y Conduits and cabletrays Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 52 OF 84)

Room Number R-C118 Title Centrifugal Charging Pump Train B Remarks Flooding Analysis X Flooding from sources within the room, coincident with the most limiting single active failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not preclude mitigation of the event or safe shutdown of the plant.

X Flooding from sources external to the room is not credible even with a single active failure.

Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis No credible missile sources exist in the room.

X Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis There are no high-energy lines in the room.

X The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 53 OF 84)

Listing of safety-related items in room R-C118 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train B raceway Conduits and cabletrays Y Y

2. CVCS centrifugal charging pump 1-1208-P6-003 Y Y

3. Charging pump room leak detection switch LSH 9827 Y

4. Charging pump room leak detection switch LSH 9831 Y

5. Air-handling unit 1-1555-A7-014 Y Y

6. Piping 1-1202-167-1 1/2 in. Y Y 1-1202-167-1 1/2 in. Y Y 1-1202-168-2 in. Y Y 1-1202-169-1 1/2 in. Y Y 1-1202-170-2 in. Y Y 1-1202-379-2 in. Y Y 1-1202-378-2 in. Y Y 1-1208-139-6 in. Y Y 1-1208-145-4 in. Y Y 1-1592-052-1 1/2 in. Y Y 1-1592-067-1 1/2 in. Y Y 1-1208-135-1 in. Y Y 1-1208-200-3/4 in. Y Y 1-1208-200-1 in. Y Y 1-1208-201-3/4 in. Y Y 1-1208-203-1 in. Y Y 1-1208-204-3/4 in. Y Y 1-1555-091-1 1/2 in. Y Y 1-1555-135-1 1/2 in. Y Y 1-1592-057-1 1/2 in. Y Y

7. HVAC ducts Train A Y

8. Charging pump room cooler, high temperature switch T1SH-12215 Y and indicator REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 54 OF 84)

Room Number R-C119 Title Valve Gallery Remarks Flooding Analysis X Flooding from sources within the room, coincident with the most limiting single active failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Pipe Break Analysis There are no high-energy lines in the room.

X The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 55 OF 84)

Listing of safety-related items in room R-C119 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train B raceway Conduits and cabletrays Y Y

2. Charging pump train B, throttling valve HV 1090B Y Y HY-1090B ZT-1090B

3. Piping 1-1592-038-2 in. Y Y 1-1592-026-2 in. Y Y 1-1202-167-1 1/2 in. Y Y 1-1202-168-2 in. Y Y 1-1202-169-1 1/2 in. Y Y 1-1202-170-2 in. Y Y 1-1208-146-3 in. Y Y 1-1208-099-3 in. Y Y 1-1208-139-6 in. Y Y 1-1208-145-4 in. Y Y 1-1204-057-4 in. Y Y 1-1208-099-2 in. Y Y 1-1208-135-1 in. Y Y 1-1208-137-8 in. Y Y 1-1208-145-1 in. Y Y 1-1208-303-2 in. Y Y 1-1208-499-2 in. Y Y 1-1592-052-1 1/2 in. Y Y 1-1592-057-1 1/2 in. Y Y

4. Charging pump miniflow isolation valve HV-8111B Y

5. Charging pump discharge, train B valve HV-8438, HV-8485V Y Y

6. Charging pump suction, train B valve HV-8471B Y Y

7. Charging pump miniflow isolation to RWST valves HV-8508B, HV-8509B REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 56 OF 84)

Room Number R-C120 Title Vestibule Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. The flooding from sources located in this room will not impair the safe shutdown failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.F), will not capability of the SRE.

preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 57 OF 84)

Listing of safety-related items in room R-C120 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train B raceway Conduit and cabletrays Y Y

2. Piping 1-1592-038-2 in. Y Y 1-1592-026-2 in. Y Y 1-1592-052-1 1/2 in. Y Y 1-1592-057-1 1/2 in. Y Y 1-1202-167-1 1/2 in. Y Y 1-1202-168-2 in. Y Y 1-1202-169-1 1/2 in. Y Y 1-1202-170-2 in. Y Y 1-1202-151-4 in. Y Y 1-1202-169-3 in. Y Y 1-1202-151-3 in. Y Y 1-1202-006-3 in. Y Y 1-1202-006-4 in. Y Y 1-1202-151-4 in. Y Y 1-1202-167-3 in. Y Y 1-1202-290-3/4 in. Y Y 1-1202-316-3/4 in. Y Y 1-1202-479-2 in. Y Y 1-1202-480-2 in. Y Y 1-1228-166-2 in. Y Y 1-1609-030-3/4 in. Y Y 1-1609-031-3/4 in. Y Y 1-1609-032-3/4 in. Y Y 1-1609-033-3/4 in. Y Y 1-1609-034-3/4 in. Y Y

3. Refueling water tank to charging pump valve LV-0112E REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 58 OF 84)

Room Number R-C131 Title Vestibule Remarks Flooding Analysis Flooding from sources within the room coincident with the most limiting single active 1. Flooding from sources located in this room will not impair the safe shutdown, failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not capability of the SRE.

preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remark 1.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis There are no high-energy lines in the room.

X The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 59 OF 84)

Listing of safety-related items in room R-C131 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Piping 1-1205-003-12 in. Y Y 1-1407-001-3 in. Y Y 1-1407-002-3 in. Y Y 1-1407-003-3 in. Y Y 1-1407-004-3 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 60 OF 84)

Room Number R-C133 Title Pipe Penetration Room Train A Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. Flooding from sources located in this room will not impair the safe shutdown capability failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not of the SRE.

preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 61 OF 84)

Listing of safety-related items in room R-C133 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train A raceway Conduits and cabletrays Y Y

2. Piping 1-1205-003-14 in. Y Y 1-1205-003-12 in. Y Y 1-1205-010-12 in. Y Y 1-1205-007-8 in. Y Y 1-1206-005-8 in. Y Y 1-1205-028-1 in. Y Y 1-1205-066-1 1/2 in. Y Y

3. HVAC ducts Train A Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 62 OF 84)

Room Number R-C134 Title Pipe Penetration Room Train A Remarks

1. Flooding from sources located in this room will not impair the safe shutdown capability Flooding Analysis of the SRE.

Flooding from sources within the room, coincident with the most limiting single active failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 63 OF 84)

Listing of safety-related items in room R-C134 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Containment sump isolation valve HV-8811A Y

2. RHR encapsulation vessel V-1-1205 V4 001 Y Y

3. Containment spray pump emergency sump isolation HV-9002A Y valve

4. Containment spray pump suction valve HV-9003A Y

5. Piping 1-1205-028-14 in. Y Y 1-1205-061-20 in. Y Y 1-1205-041-14 in. Y Y 1-1206-001-12 in. Y Y 1-1206-061-18 in. Y Y 1-1205-003-12 in. Y Y 1-1205-062-3/4 in. Y Y 1-1205-063-3/4 in. Y Y 1-1206-001-1 in. Y Y 1-1206-076-3/4 in. Y Y 1-1206-077-3/4 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 64 OF 84)

Room Number UC-C14 Title Pipe Chase Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. The flooding from sources located in this room will not impair the safe shutdown failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not capability of the SRE.

preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remark 1.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis No credible missile sources exist in the room.

X Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis There are no high-energy lines in the room.

X The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 65 OF 84)

Listing of safety-related items in room UC-C14 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Piping 1-1204-002-3 in. Y Y 1-1208-146-3 in. Y Y 1-1208-103-3 in. Y Y 1-1208-097-3 in. Y Y 1-1208-095-3 in. Y Y 1-1208-411-8 in. Y Y 1-1208-123-4 in. Y Y 1-1208-410-3 in. Y Y 1-1208-107-1 in. Y Y 1-1208-022-2 in. Y Y 1-1208-003-3 in. Y Y 1-1204-037-4 in. Y Y 1-1204-012-8 in. Y Y 1-1208-485-1 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 66 OF 84)

Room Number R-D48 Title RHR Pump Room Train A Remarks Flooding Analysis 1. The SRE will not be adversely affected by the maximum design flood.

Flooding from sources within the room, coincident with the most limiting single active failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not preclude mitigation of the event or safe shutdown of the plant.

X Flooding from sources external to the room is not credible even with a single active failure.

Other, see remark 1.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remark 2.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 67 OF 84)

Listing of safety-related items in room R-D48 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Leak detection switch LSH 9856 Y

2. Leak detection switch LSH 9860 Y

3. RHR pump 1 inlet valve HV 8812A Y Y

4. RHR 1 cold leg isolation valve HV 8716A Y Y

5. Residual heat removal pump cooler temperature TE-12206 Y Y element

6. Train A raceway Conduits and cabletrays Y Y

7. RHR pump 1-1205-P6-001 Y Y

8. Piping 1-1205-019-2 in. Y Y 1-1205-023-2 in. Y Y 1-1205-039-12 in. Y Y 1-1205-013-3 in. Y Y 1-1202-110-2 in. Y Y 1-1202-111-2 in. Y Y 1-1203-088-1 in. Y Y 1-1203-086-1 in. Y Y 1-1205-007-8 in. Y Y 1-1205-003-14 in. Y Y 1-1205-009-8 in. Y Y 1-1205-005-8 in. Y Y 1-1205-005-2 in. Y Y 1-1204-001-12 in. Y Y 1-1205-016-2 in. Y Y 1-1205-039-2 in. Y Y 1-1205-070-1/2 in. Y Y 1-1205-071-1/2 in. Y Y 1-1205-192-8 in. Y Y 1-1202-478-2 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 68 OF 84)

Room Number R-D76 Title Containment Spray Pump Room Train A Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. The flooding from sources located in this room will not impair the safe shutdown failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not capability of the SRE preclude mitigation of the event or safe shutdown of the plant.

X Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 69 OF 84)

Listing of safety-related items in room R-D76 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Leak detection switch LSH 9872 Y

2. Refueling water storage tank to containment spray HV 9017A Y pump 1, valve

3. Leak detection switch LSH 9868A Y

4. Train A raceway Conduits and cabletrays Y Y

5. Containment spray pump 1-1206-P6-001 Y Y

6. Piping 1-1206-001-12 in. Y Y 1-1206-005-8 in. Y Y 1-1202-110-2 in. Y Y 1-1202-111-2 in. Y Y 1-1206-036-3 in. Y Y 1-1206-003-10 in. Y Y 1-1202-003-2 in. Y Y 1-1202-003-3 in. Y Y 1-1204-007-10 in. Y Y 1-1202-004-2 in. Y Y 1-1202-004-3 in. Y Y 1-1202-105-2 in. Y Y 1-1202-294-3/4 in. Y Y 1-1202-477-2 in. Y Y 1-1206-005-3 in. Y Y 1-1206-022-3/4 in. Y Y 1-1206-023-3/4 in. Y Y 1-1206-031-3/4 in. Y Y 1-1206-047-3 in. Y Y 1-1206-051-3/4 in. Y Y 1-1206-070-1 in. Y Y 1-1206-152-2 in. Y Y 1-1206-223-2 in. Y Y

7. Containment sump pump relief valve PSV-9007A Y

8. Containment spray pump room cooler temperature TISH-12207 Y Y switch

9. Containment spray train A 1-1206-6001 REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 70 OF 84)

Room Number R-D77 Title Containment Spray Pump Room Train B Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. The only equipment in this room is associated with containment spray system, which failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G) will not is not essential in this case.

preclude mitigation of the event or safe shutdown of the plant.

X Flooding from sources external to the room is not credible even with a single active failure.

Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis No credible missile sources exist in the room.

X Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 71 OF 84)

Listing of safety-related items in room R-D77 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Leak detection switch LSH 9869 Y

2. Radwaste to containment spray pump 2, valve HV 9017B Y

3. Train B raceway Conduits and cabletrays Y Y

4. Containment spray pump 1-1206-P6-002 Y

5. Air-handling unit 1-1555-A7-010 Y

6. Piping 1-1202-151-2 in. Y Y 1-1202-006-2 in. Y Y 1-1204-038-10 in. Y Y 1-1206-004-12 in. Y Y 1-1206-002-12 in. Y Y 1-1206-006-8 in. Y Y 1-1592-044-1 1/2 in. Y Y 1-1592-054-1 1/2 in. Y Y 1-1202-006-3 in. Y Y 1-1202-310-3/4 in. Y Y 1-1202-483-2 in. Y Y 1-1204-221-2 in. Y Y 1-1206-004-10 in. Y Y 1-1206-006-2 in. Y Y 1-1206-019-3 in. Y Y 1-1206-027-3/4 in. Y Y 1-1206-048-3 in. Y Y 1-1206-052-3/4 in. Y Y 1-1206-071-1 in. Y Y 1-1555-070-1/2 in. Y Y 1-1609-035-3/4 in. Y Y 1-1609-036-3/4 in. Y Y

7. Containment spray pumproom cooler temperature switch TISH-12207 Y Y

8. Containment spray pumproom, train B level switch LSH-9873 Y Y

9. Containment spray pumproom cooler temperature switch TISH-12213 Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 72 OF 84)

Room Number R-D100 Title Pipe Chase Train A Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. The flooding from sources located in this room will not impair the safe shutdown failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not capability of the SRE.

preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating component generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 73 OF 84)

Listing of safety-related items in room R-D100 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Pipe chase leak detection switch LSH 9842 Y

2. Pipe chase leak detection switch LSH 9846 Y

3. Train A raceway Conduits and cabletrays Y Y

4. Piping 1-1205-007-8 in. Y Y 1-1592-031-2 1/2 in. Y Y 1-1592-020-2 1/2 in. Y Y 1-1205-003-14 in. Y Y 1-1204-001-12 in. Y 1-1202-111-2 in. Y Y 1-1202-110-2 in. Y Y 1-1205-010-12 in. Y Y 1-1205-009-8 in. Y Y 1-1204-006-24 in. Y Y 1-1204-192-8 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 74 OF 84)

Room Number R-D101 Title Pipe Chase Train B Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. The flooding from sources located in this room will not impair the safe shutdown failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not capability of the SRE.

preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 75 OF 84)

Listing of safety-related items in room R-D101 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train B raceway Conduits and cabletrays Y Y

2. Piping 1-1205-004-14 in. Y Y 1-1205-008-8 in. Y Y 1-1206-006-8 in. Y 1-1206-002-12 in. Y 1-1202-155-2 in. Y Y 1-1202-156-2 in. Y Y 1-1205-009-8 in. Y Y 1-1204-006-12 in. Y Y 1-1204-038-14 in. Y Y 1-1204-006-24 in. Y Y 1-1208-131-2 in. Y Y 1-1208-241-3 in. Y Y 1-1208-242-2 in. Y Y 1-1204-221-2 in. Y Y 1-1205-105-2 in. Y Y 1-1208-017-2 in. Y Y 1-1208-241-3 in. Y Y 1-1208-450-2 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 76 OF 84)

Room Number R-D121 Title Passage Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. Flooding from sources located in this room will not impair the safe shutdown capability failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not of the SRE.

preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

X Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 77 OF 84)

Listing of safety-related items in room R-D121 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. RHR miniflow switch FIS-0610 Y Y

2. Train A raceway Conduits and cabletrays Y Y

3. Residual heat removal LP FT-0618 Y Y Yet bypass, train A flow transmitter REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 78 OF 84)

Room Number R-D122 Title Passage Remarks Flooding Analysis Flooding from sources within the room, coincident with the most limiting single active 1. The flooding from sources located in this room will not impair the safe shutdown failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not capability of the SRE.

preclude mitigation of the event or safe shutdown of the plant.

X Flooding from sources external to the room is not credible even with a single active failure.

Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis X No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 79 OF 84)

Listing of safety-related items in room R-D122 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train A raceway Conduits and cabletrays Y Y

2. Piping 1-1205-026-3/4 in. Y Y 1-1555-061-1 1/2 in. Y Y 1-1592-020-2 1/2 in. Y Y 1-1592-028-1 1/2 in. Y Y 1-1592-031-2 1/2 in. Y Y 1-1592-036-1 1/2 in. Y Y 1-1592-053-1 1/2 in. Y Y 1-1592-056 1 1/2 in. Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 80 OF 84)

Room Number R-D128 Title Cooler Room Train A Remarks Flooding Analysis X Flooding from sources within the room, coincident with the most limiting single active 1. Missiles from sources within the room, coincident with the most limiting single active failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G) will not preclude failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G) will not mitigation of the event or safe shutdown of the plant. preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

X Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 81 OF 84)

Listing of safety-related items in room R-D128 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train A raceway Conduits and cableways Y Y

2. Piping 1-1592-036-1 1/2 in. Y Y 1-1592-028-1 1/2 in. Y Y 1-1592-020-2 1/2 in. Y Y 1-1592-031-2 1/2 in. Y Y

3. Air-handling unit 1-1555-A7-007 Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 82 OF 84)

Room Number R-D130 Title Cooler Room Train B Remarks Flooding Analysis X Flooding from sources within the room, coincident with the most limiting single active 1. Missiles from sources within the room, coincident with the most limiting single active failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not failure (assumed in accordance with paragraphs 3.6.1.1.F and 3.6.1.1.G), will not preclude mitigation of the event or safe shutdown of the plant. preclude mitigation of the event or safe shutdown of the plant.

Flooding from sources external to the room is not credible even with a single active failure.

Other, see remarks.

Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Category 2/1 interactions).

X The NSRE in the room is seismically restrained (upgraded to Category 1 requirements); therefore, no seismically induced failures of these restraints are postulated.

The NSRE, which is postulated to fail as a result of an SSE, does not impact SRE.

Other, see remarks.

Missile Analysis No credible missile sources exist in the room.

Missiles from rotating components generated within the room contain insufficient energy to escape their equipment housing(s).

X Other, see remarks.

Pipe Break Analysis X There are no high-energy lines in the room.

The high-energy line breaks have been evaluated and do not adversely affect SRE.

Other, see remarks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 83 OF 84)

Listing of safety-related items in room R-D130 Item Equipment Essential Seismic Number Description Designation Equipment Category 1

1. Train B raceway Conduits and cabletrays Y Y

2. Air-handling unit 1-1555-A7-008 Y Y REV 15 4/09

VEGP-FSAR-3F TABLE 3F-1 (SHEET 84 0F 84)

Rooms: R-C60; R-C106; R-C110; R-C125; R-C130 Effects analysis:

Pressure/temperature - There are no high-energy line breaks located in the above rooms. As the above rooms do not have high-energy line breaks and are not break nodes, they are included as a flowpath in the nodal model used in the pressure/temperature analysis of the steam generator blowdown system.

The above rooms are affected by pressure/temperature analysis.

Equipment is designed to withstand the effects of pressure, temperature, and humidity. For the pressure/temperature/humidity data, see table 3.11.B.1-1.

Structures are designed to withstand the short-term pressure/ temperature effects of high-energy line breaks.

REV 15 4/09

VEGP-FSAR-3F TABLE 3F-2 UNIT 1 AFW PUMP HOUSE DESIGN PARAMETERS Design Initial Conditions Conditions Blowdown Press. Calc.

Node Rel. Rel. Flood Nodal Flow after Peak Mass (a) (b)

No. Pressure Temp. Hum. Pressure Temp. Hum. Level Volume Area Break Temp. Time Flowrate 3 (a) 2 (Btu/lb) (psia) (°F) (%) (psia) (°F) (%) (in.) (ft ) Nodes (ft ) (psia) (°F) (s) (lb/s) Enthalpy Steam Line 3

1. 14.7 120 50 17.7 320 0-100 (c) 12.6x10 1-2 24.0 15.4 184 0 0 0 3

2. 14.7 120 50 17.7 320 0-100 (d) 10.9x10 1-3 20.0 15.5 249 0.004 237.4 1188.6 0.01 210.5 1156.7 3.

14.7 120 50 - - - - - - - - - 0.1 143.8 1187.3 (atm) 0.2 126.7 1191.1 0.5 110.8 1191.7 0.9 104.5 1191.3 2.0 100.9 1190.4 20 100.7 1190.1 Auxiliary Feedwater 0 0 60 282 1800 142

a. Refer to drawing 1X6DD300 for nodal boundary.

b. The design of the structure assumes a 2-h duration for peak temperature.

c. Node 1 Flood level (in.):

Room 104 - 86 Room 105 - 97.2

d. Node 2 Flood level (in.):

Room 106 - 11 REV 14 10/07

VEGP-FSAR-3F TABLE 3F-3A (SHEET 1 OF 4)

UNIT 1 MSIV/MFIV ROOM FLOODING (CASE 2) AND PRESSURE (CASE 1) ANALYSIS Initial Conditions Design Conditions Flood Node Calc Peak (a)

Node Pressure Temp. Rel. Hum. Pressure Temp Rel. Hum. Level Volume Pressure 3

No. (psia) (°F) (%) (psia) (°F) (%) (in.) (ft ) (psia)

I. Auxiliary Building MSIV/MFIV Area 1 14.7 120 50 29.7 320 0-100 NA 5520.0 18.4 2 14.7 120 50 29.7 320 0-100 NA 17423.0 18.0 3 14.7 120 50 29.7 320 0-100 7 3505.5 19.1 4 14.7 120 50 29.7 320 0-100 67 6033.88 19.4 5 14.7 120 50 29.7 320 0-100 NA 4138.8 20.3 6 14.7 120 50 29.7 320 0-100 67 4237.0 19.4 7 14.7 120 50 29.7 320 0-100 7 1989.8 18.8 8 14.7 120 50 29.7 320 0-100 NA 12684.8 18.0 9 14.7 120 50 29.7 320 0-100 67 22883.9 20.8 10 14.7 120 50 29.7 320 0-100 NA 9577.0 24.4 11 14.7 120 50 29.7 320 0-100 138 5415.0 23.5 12 14.7 120 50 29.7 320 0-100 NA 4651.2 18.8 13 atm REV 14 10/07

VEGP-FSAR-3F TABLE 3F-3A (SHEET 2 OF 4)

Initial Conditions Design Conditions Flood Node Calc Peak (a

Node Pressure Temp. Rel. Hum. Pressure Temp Rel. Hum. Level Volume Pressure 3

No. (psia) (°F) (%) (psia) (°F) (%) (in.) (ft ) (psia)

II. Auxiliary Building MSIV/MFIV Area Outside Restraint Wall 1 14.7 120 50 29.7 320 0-100 NA 5520.0 17.5 2 14.7 120 50 29.7 320 0-100 NA 17423.0 17.3 3 14.7 120 50 29.7 320 0-100 7 3505.5 17.4 4 14.7 120 50 29.7 320 0-100 67 6033.88 17.5 5 14.7 120 50 29.7 320 0-100 NA 4138.8 17.2 6 14.7 120 50 29.7 320 0-100 67 4273.0 17.9 7 14.7 120 50 29.7 320 0-100 7 1989.8 17.9 8 14.7 120 50 29.7 320 0-100 NA 12684.8 17.4 9 14.7 120 50 29.7 320 0-100 67 22883.9 17.4 10 14.7 120 50 29.7 320 0-100 NA 9577.0 24.4 11 14.7 120 50 29.7 320 0-100 138 5415.0 23.5 12 14.7 120 50 29.7 320 0-100 NA 4651.2 18.8 13 atm REV 14 10/07

VEGP-FSAR-3F TABLE 3F-3A (SHEET 3 OF 4)

Initial Conditions Design Conditions Flood Node Calc Peak (a)

Node Pressure Temp. Rel. Hum. Pressure Temp Rel. Hum. Level Volume Pressure 3

No. (psia) (°F) (%) (psia) (°F) (%) (in.) (ft ) (psia)

III. Control Building MSIV/MFIV Area 1 14.7 120 50 29.7 320 0-100 NA 3079.9 21.4 2 14.7 120 50 29.7 320 0-100 NA 17617.0 20.2 3 14.7 120 50 29.7 320 0-100 22 11295.2 20.4 4 14.7 120 50 29.7 320 0-100 NA 3769.6 19.8 5 14.7 120 50 29.7 320 0-100 NA 21514.5 19.5 6 14.7 120 50 29.7 320 0-100 22 13740.5 20.3 7 14.7 120 50 29.7 320 0-100 NA 8853.6 16.8 8 atm IV. Control Building MSIV/MFIV Area Outside Restraint Wall 1 14.7 120 50 29.7 320 0-100 16 28805.2 23.5

a. Refer to figure 3F-1 (sheets 1 through 4) for nodal boundary.

REV 14 10/07

VEGP-FSAR-3A TABLE 3F-3A (SHEET 4 OF 4)

Blowdown Data (Case 1)

(Break at Node 10)

Mass Flowrate Enthalpy Time (s) (lbm/s) (BTU/lbm) 0.0 0 1188 0.004 18,959 1138 0.008 19,905 1139 0.01 19,399 1137 0.03 17,567 1136 0.06 16,747 1139 0.10 16,030 1140 0.15 15,450 1140 0.20 15,312 1142 1.0 12,994 1153 1.2 12,378 1141 1.4 9304 1119 1.6 14,163 814 1.65 15,040 791 1.80 17,716 725 1.90 18,414 712 2.00 18,831 706 2.15 19,854 700 2.30 19,597 695 3.0 17,259 688 3.4 21,428 668 4.0 20,755 661 4.8 20,415 653 REV 14 10/07

VEGP-FSAR-3B TABLE 3F-3B (SHEET 1 OF 2)

UNIT 1 CONTROL BUILDING MSIV/MFIV VAULT AREA MSLB TEMPERATURE (CASE 3) ANALYSIS Initial Conditions Design Conditions Rel. Rel. Node (c)

Node Pressure Temp. Hum. Pressure Temp. Hum. Volume Elev. Calc. Peak (a) 3 No. Description (psia) (°F) (%) (psia) (°F) (%) (ft ) (ft) Temp. (°F)

(c) 1 West Steam Line Room (Lower) 14.7 120 50 29.7 320 0-100 18,361.4 224.0 483.2 2 West Aux. Feedline Room 14.7 120 50 29.7 320 0-100 11,300.0 200.5 399.0 3 East SL Room (Lower) 14.7 120 50 29.7 320 0-100 11,370.0 224.0 413.1 4 East AFL Room 14.7 120 50 29.7 320 0-100 13,740.0 200.5 360.9 5 Entrance Room (Lower) 14.7 120 50 29.7 320 0-100 4,234.0 221.0 390.2 6 Entrance Room (Upper) 14.7 120 50 29.7 320 0-100 4,234.0 239.0 398.7 7 East Penthouse 14.7 120 50 29.7 320 0-100 3,208.0 257.0 428.3 8 West Penthouse 14.7 120 50 29.7 320 0-100 2,338.6 257.0 477.4 9 East SL Room (Upper) 14.7 120 50 29.7 320 0-100 11,370.0 240.5 429.1 10 West Steam Line Room (Upper) 14.7 120 50 29.7 320 0-100 9,180.0 240.5 478.3 REV 15 4/09

VEGP-FSAR-3B TABLE 3F-3B (SHEET 2 OF 2)

Flow Path Data Blowdown Data 2 (b)

(1.0 ft MSLB at 102% Power)

Flow Upstream Downstream Flow Area Mass Flow- Enthalpy 2

Path No. Node Node ft Time (s) rate (lbm/s) (Btu/lbm) 1 4 3 145.0 0.0 0.0 0.0 2 2 1 139.0 1.5 1980.0 1192.0 3 1 3 218.6 6.5 1904.0 1194.0 4 2 4 245.0 14.0 2282.0 1188.0 5 3 5 95.0 15.0 2283.0 1189.0 6 9 6 144.0 15.5 2173.0 1191.0 7 9 7 139.5 23.5 1490.0 1201.0 8 1 8 120.0 33.0 1187.0 1204.0 9 1 9 205.5 55.5 924.2 1204.0 10 3 9 876.6 64.5 762.0 1236.0 11 1 atm 4.5 81.5 203.9 1288.0 12 2 atm 6.85 90.5 100.3 1298.0 13 3 atm 4.5 123.5 78.0 1304.0 14 4 atm 6.85 162.5 78.0 1305.0 15 5 atm 141.0 1800.0 78.0 1287.0 16 6 atm 141.0 1812.0 12.5 1295.0 17 7 atm 164.35 18 8 atm 135.85 19 5 6 268.0 20 1 break 1.0 21 1 10 725.0

a. Refer to figure 3F-2 for nodal diagrams.

b. Main steam line break.

c. The calculated peak temperatures correspond to a 1.0 ft2 MSLB at 102% power (102% of 3579 MWt). This break resulted in the overall highest peak temperature which occurred in break node 1. The peak temperatures and pressures for other nodes actually may be higher for other breaks.

REV 15 4/09

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VEGP-FSAR-3 TABLE 3F-5 JET IMPINGEMENT BARRIERS System Break Location Target Number Barrier Number Building Line Number Accumulator injection loop-4 P-14DB-0411-L-1 1201-029-4 in. BC-3 Containment (Unit 1 only) 1204-127-10 in. P-14DB-0411-L-2 Main safety loops 1 and 4 from auxiliary P-17CA-1059-C 1305-056-26 in. BT-2 Tunnel IT1 building to main steam tunnel 1301-008-38 in.

Main feedwater loops 1-4 (tunnel to P-12EA-1135-C 1301-007-36 in. BT-1 Tunnel IT1 turbine building) 1305-055-36 in.

REV 14 10/07

REV 14 10/07 NODAL BOUNDARY FOR AUXILIARY BUILDING - MSIV/MFIV ROOMS FIGURE 3F-1 (SHEET 1 OF 4)

REV 14 10/07 NODAL BOUNDARY FOR AUXILIARY BUILDING - MSIV/MFIV ROOMS FIGURE 3F-1 (SHEET 2 OF 4)

REV 14 10/07 NODAL BOUNDARY FOR CONTROL BUILDING - MSIV/MFIV ROOMS (INSIDE RESTRAINT WALL)

FIGURE 3F-1 (SHEET 3 OF 4)

REV 14 10/07 NODAL BOUNDARY FOR CONTROL BUILDING - MSIV/MFIV ROOMS (OUTSIDE RESTRAINT WALL)

FIGURE 3F-1 (SHEET 4 OF 4)

REV 14 10/07 NODAL DIAGRAM FOR MSIV/MFIV GOTHIC ANALYSIS FIGURE 3F-2

REV 14 10/07 TIME-HISTORY PLOT MAXIMUM COMPARTMENT COMPOSITE TEMPERATURE FOR MSIV AREA BREAK NODE FIGURE 3F-3

VEGP-FSAR-4 4.0 REACTOR 4.1

SUMMARY

DESCRIPTION This chapter describes:

A. The mechanical components of the reactor and reactor core, including the fuel rods and fuel assemblies.

B. The nuclear design.

C. The thermal-hydraulic design.

The initial reactor core is composed of an array of fuel assemblies that are identical in mechanical design but different in fuel enrichment. Within each fuel assembly all rods are of the same enrichment. Three different enrichments are employed in the first core: 2.10 (region 1),

2.60 (region 2), and 3.10 (region 3) weight percents. It was required that the initial core loading maximum enrichment not exceed 3.2 weight percent U-235. For subsequent reloads, the target maximum enrichment is up to 5.0 weight percent. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment not to exceed 5.0 weight percent U-235.

Reload cores are comprised of 17 x 17 VANTAGE+ with Debris Mitigation Features (VANTAGE+) and/or VANTAGE 5 fuel assemblies. The original referenced design described herein consisted of LOPAR fuel assemblies arranged in a checkered, low-leakage core loading pattern.

The significant mechanical design features of the VANTAGE 5 design, as described in reference 1, relative to the LOPAR fuel design include the following: integral fuel burnable absorbers (IFBA), intermediate flow mixer (IFM) grids, reconstitutable top nozzle (RTN),

extended burnup capability, and axial blankets. In addition, a debris filter bottom nozzle (DFBN) replacing the standard nozzle has been implemented. The RTN, DFBN, and extended burnup capability have been introduced in both VEGP Units 1 and 2. Fuel reloads contain an advanced zirconium alloy clad fuel known as ZIRLO, which is the key feature of the VANTAGE+ fuel design, reference 2, or Optimized ZIRLOTM cladding, reference 3, as well as improved design features including:

• Protective bottom grid,

• Long fuel rod end plugs,

• Annular blanket pellets,

• ZIRLO guide thimbles,

• Instrumentation tube,

• Mid grid, and

• IFM grid.

Beginning with the Region 27 (Cycle 25) reload for Unit 1 and the Region 25 (Cycle 23) reload for Unit 2, the Vogtle units implement the Westinghouse 17x17 OFA PRIMETM fuel assembly 4.1-1 REV 24 10/22

VEGP-FSAR-4 design which is an improved version of the 17x17 VANTAGE+ fuel design. The PRIMETM fuel assembly design incorporates advanced features to improve fuel reliability, dimensional stability and corrosion resistance of the 17x17 VANTAGE+ with Debris Mitigation Features fuel design.

The PRIMETM fuel features include Low Tin ZIRLO grid strap material to improve corrosion resistances, the PRIMETM Bottom Nozzle incorporates a low pressure drop design along with lower side skirts to help improve debris resistance and the external dashpot tube which stiffens the dashpot region of the fuel assembly to minimize the potential for incomplete rod insertion.

The PRIMETM advanced fuel design has been extensively evaluated, tested and analyzed and it was demonstrated that it meets all of the applicable mechanical design and safety criteria.

A fuel assembly is composed of 264 fuel rods in a 17 x 17 square array. The center position in the fuel assembly is reserved for incore instrumentation. The remaining 24 positions in the fuel assembly have guide thimbles which are joined to the top and bottom nozzles of the fuel assembly and serve to support the fuel grids. A fuel assembly may have limited substitution of zirconium alloy or stainless steel filler rods in place of fuel rods, in accordance with NRC approved applications of fuel rod configurations. The fuel grids consist of an egg crate arrangement of interlocked straps that maintain lateral spacing between the rods. The grid straps have spring fingers and dimples which grip and support the fuel rods. The middle grids also have coolant-mixing vanes. The flow mixer grid straps contain only support dimples and coolant mixing vanes.

The fuel rods consist of enriched uranium, in the form of cylindrical pellets of uranium dioxide, contained in Zircaloy-4 tubing. Commencing with Unit 2 Cycle 6, the fresh fuel region uses an advanced zirconium alloy tubing known as ZIRLO to provide improved fuel performance, reference 2. Additionally, Optimized ZIRLO cladding material has been approved for use in Vogtle 1 and 2 by the NRC, reference 4. The tubing is plugged and seal welded at the ends to encapsulate the fuel. An axial blanket of natural, mid-enriched, or fully enriched solid or annular UO2 fuel pellets may be placed at the ends of the enriched fuel pellet stack. The natural or mid-enriched axial blanket pellets are used to reduce the neutron leakage and to improve fuel utilization. The annular blanket pellets are used to increase the void volume for gas accommodation within the fuel rod. A second fuel rod type is utilized to varying degrees within some fuel assemblies. These rods use zirc diboride (ZrB2) coated fuel pellets in the central portion of the fuel stack. All fuel rods are pressurized internally with helium during fabrication to reduce clad creep-down during operation and thereby to increase fatigue life.

Depending on the position of the assembly in the core, the guide thimbles are used for rod cluster control assemblies (RCCAs), neutron source assemblies, burnable absorber (BA) assemblies, or stainless steel rod insert assemblies (SSRIA). If none of these is required, the guide thimbles may be fitted with plugging devices to limit bypass flow. Standard borosilicate glass rods were used in first cycles. Wet annular burnable absorbers (WABA) and/or IFBA coated fuel pellets are used in subsequent reloads.

The bottom nozzle is a boxlike structure which serves as the lower structural element of the fuel assembly and directs the coolant flow distribution to the assembly. The top nozzle assembly serves as the upper structural element of the fuel assembly and provides a partial protective housing for the RCCA or other components.

The RCCAs consist of 24 absorber rods fastened at the top end to a common hub or spider assembly. Each absorber rod consists of either hafnium or an alloy of silver-indium-cadmium clad in stainless steel. The control rod assemblies shall contain a nominal 142 in. of absorber material. The nominal absorber composition shall be 95.5 percent natural hafnium and 4.5 percent natural zirconium and/or 80 percent silver, 15 percent indium, and 5 percent cadmium.

All control rods shall be clad with stainless steel. The RCCAs are used to control relatively rapid changes in reactivity and to control the axial power distribution.

4.1-2 REV 24 10/22

VEGP-FSAR-4 The reactor core is cooled and moderated by light water at a pressure of 2250 psia. Soluble boron in the moderator/coolant serves as a neutron absorber. The concentration of boron is varied to control reactivity changes that occur relatively slowly, including the effects of fuel burnup and transient xenon. Burnable absorber rods are also employed in the first core to limit the amount of soluble boron required and, thereby, to maintain the desired negative reactivity coefficients. Additional boron in the form of WABAs and/or IFBAs may be employed to limit the moderator temperature coefficient and the local power peaking.

The nuclear design analyses establish the core locations for control rods and BAs and define design parameters, such as fuel enrichments and boron concentration in the coolant. The nuclear design analyses establish that the reactor core and the reactor control system satisfy all design criteria, even if the RCCA of highest reactivity worth is in the fully withdrawn position.

The core has inherent stability against diametral and azimuthal power oscillations. Axial power oscillations which may be induced by load changes and resultant transient xenon may be suppressed by the use of the RCCAs.

The thermal-hydraulic design analyses establish that adequate heat transfer is provided between the fuel clad and the reactor coolant. The thermal design takes into account local variations in dimensions, power generation, flow distribution, and mixing. The mixing vanes incorporated in the fuel assembly spacer grid design and the VANTAGE 5 fuel assembly IFMs induce additional flow mixing between the various flow channels within a fuel assembly as well as between adjacent assemblies.

The performance of the core is monitored by fixed neutron detectors outside the core, movable neutron detectors within the core, and thermocouples at the outlet of selected fuel assemblies.

The excore nuclear instrumentation provides input to automatic control functions.

Table 4.1-1 presents a comparison of the principal nuclear, thermal-hydraulic, and mechanical design parameters of the VEGP units with parameters of the SNUPPS units (Docket No. STN 50-482, STN 50-483, STN 50-485, and STN 50-486).

The analytical techniques employed in the core design are tabulated in table 4.1-2. The mechanical loading conditions considered for the core internals and components are tabulated in table 4.1-3. Specific or limiting loads considered for design purposes of the various components are listed as follows: fuel assemblies in paragraph 4.2.1.5; control rods, BA rods, neutron source rods, thimble plug assemblies, and the SSRIA in paragraph 4.2.1.6. The dynamic analyses, input forcing functions, and response loadings for the control rod drive system and reactor vessel internals are presented in subsections 3.9.4 and 3.9.5.

4.

1.1 REFERENCES

1. Davidson, S. L. and Kramer, W. R., eds, "Reference Core Report VANTAGE 5 Fuel Assembly," WCAP-10444-P-A (Proprietary) and WCAP-10445-NP-A (Nonproprietary),

September 1985.

2. Davidson, S. L. and Nuhfer, D. L., "VANTAGE+ Fuel Assembly Reference Core Report," WCAP-12610-A (Proprietary) and WCAP-14342-A (Nonproprietary), April 1995.

3. Shah, H., Optimized ZIRLOTM, WCAP-12610-P-A & CENPD-404-P-A Addendum 1-A, July 2006.

4. U.S. NRC, R. E. Martin letter to SNC, C. R. Pierce, Joseph M. Farley Nuclear Plant, Units 1 and 2, and Vogtle Electric Generating Plant, Units 1 and 2, Issuance of 4.1-3 REV 24 10/22

VEGP-FSAR-4 Amendments Regarding Use of Optimized ZIRLO (CAC Nos. MF7480, MF7481, MF7482, and MF7483), August 04, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16179A386).

4.1-4 REV 24 10/22

VEGP-FSAR-4 TABLE 4.1-1 (SHEET 1 OF 4)

REACTOR DESIGN COMPARISON TABLE VEGP VEGP Thermal and Hydraulic Design Parameters (ORIGINAL DESIGN) (UPRATE DESIGN)(h) SNUPPS

1. Reactor core heat output (MWt) 3411 3626 3411

2. Reactor core heat output (106 Btu/h) 11,639 12,372 11,639

3. Heat generated in fuel (%) 97.4 97.4 97.4

4. System pressure, nominal (psia) 2250 2250 2250 5 System pressure, minimum steady state 2220 2200 2220 (psia)

6. Minimum DNBR for design transients Typical flow channel (LOPAR) 1.30 (g) 1.30 (VANTAGE + / 1.24 VANTAGE 5)

Thimble (cold wall) flow channel (LOPAR) (g)

(VANTAGE + / 1.23 VANTAGE 5)

7. DNB correlation (LOPAR) R (W-3 with (g) R (W-3 with VANTAGE + / modified spacer WRB-2 (d) modified spacer (VANTAGE 5) factor) factor)

Coolant flow(e)

8. Total vessel flowrate (106 lbm/h) 142.1 139.5 142.1 (Based on thermal design flow) 143.0 (f)

(Based on minimum measured flow)

9. Effective flowrate for heat transfer (106 lbm/h) 133.9 130.6 133.9

10. Effective flow area for heat transfer(ft2) (LOPAR) 51.1 (g) 51.1 (VANTAGE + / 54.1 VANTAGE 5)

11. Average velocity along fuel rods (ft/s) (LOPAR) 16.6 (g) 16.6 (VANTAGE + / 15.3 VANTAGE 5) 2.62

12. Average mass velocity (106 lbm/h-ft2) (LOPAR) 2.62 (g)

(VANTAGE + / 2.41 VANTAGE 5)

Coolant temperature

13. Nominal inlet (qF) 558.8 556.3 558.8

14. Average rise in vessel (qF) 59.4 64.2 59.4

15. Average rise in core (qF) 62.6 68.0 62.6

16. Average in core (qF) 591.8 592.3 591.8 REV 23 3/21

VEGP-FSAR-4 TABLE 4.1-1 (SHEET 2 OF 4)

VEGP VEGP Thermal and Hydraulic Design Parameters (ORIGINAL DESIGN) (UPRATE DESIGN)(h) SNUPPS Heat transfer

17. Average in vessel (qF) 588.5 588.4 588.5

18. Active heat transfer surface area, (ft2) (LOPAR) 59,700 (g) 59,700 (VANTAGE + / 57,505 VANTAGE 5)

19. Average heat flux (Btu/h-ft2) (LOPAR) 189,800 (g) 189,800 (VANTAGE + / 209,612 VANTAGE 5)

20. Maximum heat flux for normal operation (LOPAR) 436,500(a) (g) 440,300(a)

(Btu/h-ft2) (VANTAGE + / 524,030 (a)

VANTAGE 5) 21 Average linear power (kW/ft) 5.44 5.788 (i) 5.44

22. Peak Linear power for normal operation 12.5 14.47 (i)

(kW/ft) (a) 12.6

23. Peak linear power resulting from overpower 18.0 22.4(b) 18.0 transients/operator errors, assuming a maximum overpower of 120% (KW/ft)

24. Heat flux hot channel factor (FQ) 2.30(c) 2.50(c) 2.32(c)

25. Peak fuel central temperature for prevention of 4700 4700 4700 centerline melt (ºF)

26. Design RCC canless RCC canless RCC canless 17 x 17 17 x 17 17 x 17

27. Number of fuel assemblies 193 193 193

28. UO2 rods per assembly 264 264 264

29. Rod pitch (in.) 0.496 0.496 0.496

30. Overall dimensions (in.) 8.426 x 8.426 8.426 x 8.426 8.426 x 8.426

31. Fuel weight, as UO2 (lb) (LOPAR) 222,739 (g) 222,739 (VANTAGE + / 204,231(j)

VANTAGE 5)

32. Zircaloy/ZIRLO/Optimized ZIRLO weight (lb) (LOPAR) 45,296 (g) 45,296 (active core) (VANTAGE + / 45,914 VANTAGE 5)

33. Number of grids per assembly (LOPAR) 8-Rtype (g) 8 - R type (VANTAGE + / 2 - R type VANTAGE 5) 6 - Z type 3 - IFM 1-P-Grid

34. Loading technique (first cycle) 3 region N/A 3 region nonuniform nonuniform Fuel rods

35. Number 50,952 50,952 50,952 REV 23 3/21

VEGP-FSAR-4 TABLE 4.1-1 (SHEET 3 OF 4)

VEGP VEGP Thermal and Hydraulic Design Parameters (ORIGINAL DESIGN) (UPRATE DESIGN)(h) SNUPPS

36. Outside diameter (in.) (LOPAR) 0.374 (g) 0.374 (VANTAGE + / 0.360 VANTAGE 5)

37. Diametral gap (in.) (LOPAR) 0.0065 (g) 0.0065 (VANTAGE + / 0.0062 (non-IFBA)

VANTAGE 5)

38. Clad thickness (in.) 0.0225 0.0225 0.0225

39. Clad material (LOPAR) Zircaloy-4 (g) Zircaloy-4 (VANTAGE + / Zircaloy-4/ZIRLO/

VANTAGE 5) Optimized ZIRLO Fuel pellets

40. Material UO sintered UO sintered UO sintered 2 2 2

41. Density (% of theoretical) 95 95 95

42. Diameter (in.) (LOPAR) 0.3225 (g) 0.3225 (VANTAGE + / 0.3088 (non-IFBA)

VANTAGE 5)

43. Length (in.) (LOPAR) 0.530 (g) 0.530 (VANTAGE + / 0.370 VANTAGE 5)

(Blanket Pellet) 0.462/0.500 Rod cluster control assemblies

44. Neutron absorber Ag-In-Cd or Ag-In-Cd or Ag-In-Cd or hafnium hafnium hafnium

45. Cladding material Type 304 Type 304 Type 304 SS, cold-worked(a) SS, cold-worked(a) SS, cold-worked(a)

46. Clad thickness 0.0185 0.0185 0.0185 Ad-In-Cd or hafnium

47. Number of clusters 53 53 53

48. Number of absorber rods per cluster 24 24 24 Core structure

49. Core barrel, ID/OD (in.) 148.0/152.5 148.0/152.5 148.0/152.5

50. Thermal shield Neutron panel Neutron panel Neutron panel design design design

51. Baffle thickness (in.) 0.88 0.88 0.88 Structure characteristics

52. Core diameter, equivalent (in.) 132.7 132.7 132.7

53. Core height, active fuel (in.) 143.7 143.7 143.7

a. 316L SS cladding material applicable to Framatome RCCA only.

REV 23 3/21

VEGP-FSAR-4 TABLE 4.1-1 (SHEET 4 OF 4)

VEGP VEGP (ORIGINAL DESIGN) (UPRATE DESIGN)(h) SNUPPS Reflector thickness and composition

54. Top, water plus steel (in.) 10 10 10

55. Bottom, water plus steel (in.) 10 10 10

56. Side, water plus steel (in.) 15 15 15

57. H2O/U molecular ratio core, lattice, cold (LOPAR) 2.41 (g) 2.41 (VANTAGE + / 2.73 VANTAGE 5)

First Cycle Fuel Enrichments (Weight Percent

58. Region 1 Units 1 and 2 N/A Core A Core B

59. Region 2 2.10 N/A 2.10 1.40

60. Region 3 2.60 N/A 2.60 2.60 3.10 3.10 2.90

a. This limit is associated with the values FQ = 2.32 for SNUPPS, FQ = 2.30 for VEGP (Original Design) and FQ = 2.50 for VEGP (Uprate Design).

b. See paragraph 4.3.2.2.6.

c. This is the maximum value of FQ for normal operation.

d. See paragraph 4.4.2.2.1 for the use of the W-3 correlation.

e. Flowrates are based on 10% steam generator tube plugging for VEGP (Uprate Design) and 0% plugging for VEGP (Original Design) and SNUPPS.

f. Inlet temperature = 557.0°F.

g. LOPAR fuel is not analyzed for the VEGP MUR power uprate to 3626 MWt.

h. The VEGP MUR power uprate increases the licensed reactor core power level from 3565 MWt to 3625.6 MWt.

i. Based on densified active fuel length.

j. The decrease in fuel weight due to annular axial blanket pellets is not considered.

REV 23 3/21

VEGP-FSAR-4 TABLE 4.1-2 (SHEET 1 OF 2)

ANALYTICAL TECHNIQUES IN CORE DESIGN Analysis Technique Computer Code Section Referenced Mechanical design of core internals loads, deflections, Static and dynamic modeling Blowdown code, FORCE, finite element, 3.7.2.1 and stress analysis structural analysis code, and others 3.9.2 3.9.3 Fuel rod design Full performance characteristics (temperature, Semiempirical thermal model of fuel rod Westinghouse fuel rod design model 4.2.1.1 internal pressure, clad stress, etc.) with consideration of fuel density changes, 4.2.3.2 heat transfer, fission gas release, etc. 4.2.3.3 4.3.3.1 4.4.2.11 Nuclear design Cross-sections and group constants Microscopic data; macroscopic constants Modified ENDF/B library 4.3.3.2 for homogenized core regions LEOPARD/CINDER type or PHOENIX-P or NEXUS/PARAGON Group constants for control rods with self- HAMMER-AIM or PHOENIX-P or 4.3.3.2 shielding NEXUS/PARAGON X-Y and X-Y-Z power distributions, fuel depletion, 2-group diffusion theory TURTLE (2-D) or ANC (2-D or 3-D) 4.3.3.3 critical boron concentrations, X-Y and X-Y-Z xenon distributions, reactivity coefficients Axial power distributions, control rod worths, and 1-D, 2-group diffusion theory PANDA 4.3.3.3 axial xenon distribution Fuel rod power Integral transport theory LASER 4.3.3.1 Effective resonance temperature Monte Carlo weighting function REPAD 4.3.3.1 Criticality of reactor and fuel assemblies 2-D, 2-group diffusion theory APX system of codes, KENO-IV 4.3.2.6 Vessel irradiation Multigroup spatial dependent transport DOT 4.3.2.8 theory REV 19 4/15

VEGP-FSAR-4 TABLE 4.1-2 (SHEET 2 OF 2)

Analysis Technique Computer Code Section Referenced Thermal-hydraulic design Steady state Subchannel analysis of local fluid VIPRE-01 4.4.4.5.2 conditions in rod bundles, including inertial and crossflow resistance terms; solution is based on a one-pass model which simulates the core and hot channels.

Transient departure from nucleate boiling Subchannel analysis of local fluid VIPRE-01 4.4.4.5.4 conditions in rod bundles during transients by including accumulation terms in conservation equations; solution is based on a one-pass model which simulates the core and hot channels.

REV 19 4/15

VEGP-FSAR-4 TABLE 4.1-3 DESIGN LOADING CONDITIONS FOR REACTOR CORE COMPONENTS

1. Fuel assembly weight and core component weights (BAs, sources, plugging devices)

2. Fuel assembly spring forces and core component spring forces

3. Internals weight

4. Control rod trip (equivalent static load)

5. Differential pressure

6. Spring preloads

7. Coolant flow forces (static)

8. Temperature gradients

9. Differences in thermal expansion

a. Due to temperature differences

b. Due to expansion of different materials

10. Interference between components

11. Vibration (mechanically or hydraulically induced)

12. One or more loops out of service

13. All operational transients listed in table 3.9.N.1-1

14. Pump overspeed

15. Seismic loads (operating basis earthquake and safe shutdown earthquake)

16. Blowdown forces (due to cold and hot leg break)

REV 14 10/07

VEGP-FSAR-4 4.2 FUEL SYSTEM DESIGN The plant conditions for design are divided into four categories in accordance with their anticipated frequency of occurrence and risk to the public:

• Condition 1 - Normal operation.

• Condition 2 - Incidents of moderate frequency.

• Condition 3 - Infrequent incidents.

• Condition 4 - Limiting faults.

Chapter 15 describes bases and plant operation and events involving each condition.

The reactor is designed so that its components meet the following performance and safety criteria:

A. The mechanical design of the reactor core components and their physical arrangement, together with corrective actions of the reactor control, protection, and emergency cooling systems (when applicable) ensure that:

1. Fuel damage is not expected during Condition 1 and Condition 2 events.(a) It is not possible, however, to preclude a very small amount of fuel damage. This is within the capability of the plant cleanup system and is consistent with plant design bases.

2. The reactor can be brought to a safe state following a Condition 3 event with only a small fraction of fuel rods damaged, although sufficient fuel damage might occur to preclude immediate resumption of operation.(b)

3. The reactor can be brought to a safe state and the core can be kept subcritical with acceptable heat transfer geometry following transients arising from Condition 4 events.

B. The fuel assemblies are designed to withstand loads induced during shipping, handling, and core loading without exceeding the criteria of paragraph 4.2.1.5.

C. The fuel assemblies are designed to accept control rod insertions in order to provide the required reactivity control for power operations and reactivity shutdown conditions.

D. All fuel assemblies have provisions for the insertion of incore instrumentation necessary for plant operation.

E. The reactor vessel and internals, in conjunction with the fuel assemblies and incore control components, direct reactor coolant through the core. This achieves acceptable flow distribution and restricts bypass flow so that the heat transfer performance requirements can be met for all modes of operation.

The following subsection provides the fuel system design bases and design limits. This information, augmented by the clarifying information submitted to the Nuclear Regulatory a

Fuel damage as used here is defined as penetration of the fission product barrier; i.e., the fuel rod clad.

b In any case, the fraction of fuel rods damaged must be limited so as to meet the dose guidelines of 10 CFR 100.

4.2-1 REV 24 10/22

VEGP-FSAR-4 Commission (NRC) during their review of Westinghouse topical report WCAP-9500, Reference Core Report - 17 x 17 Optimized Fuel Assembly (ref. 1 and 2) is consistent with the acceptance criteria of the Standard Review Plan (SRP) 4.2.

4.2.1 DESIGN BASES The fuel rod and fuel assembly design bases are established to satisfy the general performance and safety criteria presented in section 4.2.

The fuel rods are designed for extended burnup as described in the Extended Burnup Evaluation Report (Ref. 24).

The detailed fuel rod design establishes such parameters as pellet size and density, clad/pellet diametral gap, gas plenum size, and helium prepressurization level. The design also considers effects such as fuel density changes, fission gas release, clad creep, and other physical properties which vary with burnup. The integrity of the fuel rods is ensured by designing to prevent excessive fuel temperatures (paragraph 4.2.1.2A); excessive internal rod gas pressures (paragraphs 4.2.1.3A and 4.2.1.3B) due to fission gas releases; and excessive cladding stresses, strains, and strain fatigue (paragraph 4.2.1.1C). This is achieved by designing the fuel rods so that the conservative design bases in the following subsections are satisfied during Condition 1 and Condition 2 events over the fuel lifetime. For each design basis, the performance of the limiting fuel rod must not exceed the limits specified by the design basis.

Integrity of the fuel assembly structure is ensured by setting limits on stresses and deformations due to various loads and by preventing the assembly structure from interfering with the functioning of other components. Three types of loads are considered:

A. Nonoperational loads such as those due to shipping and handling.

B. Normal and abnormal loads which are defined for Conditions 1 and 2.

C. Abnormal loads which are defined for Conditions 3 and 4.

The design bases for the incore control components are described in paragraph 4.2.1.6.

4.2.1.1 Cladding A. Zircaloy-4/ZIRLO/Optimized ZIRLO cladding material combines neutron economy (low absorption cross-section); high corrosion resistance to coolant, fuel, and fission products; and high strength and ductility at operating temperatures. Reference 3 documents the operating experience with Zircaloy-4/

ZIRLO and reference 28 documents the operating experience with Optimized ZIRLO as a clad material. Information on the materials chemical and mechanical properties of the cladding is given in reference 4 for Zircaloy-4, reference 25 for ZIRLO, and reference 28 for Optimized ZIRLO.

B. Stress-Strain Limits

1. Clad Stress The clad stresses under Condition 1 and 2 events are less than the Zircaloy/ZIRLO/Optimized ZIRLO clad material yield stress, with due consideration of temperature and irradiation effects. While the clad has some capability for accommodating plastic strain, the yield stress has been accepted as a conservative design basis.

4.2-2 REV 24 10/22

VEGP-FSAR-4

2. Clad Tensile Strain The total tensile creep strain is less than 1 percent from the unirradiated condition. The cluster tensile strain during a transient is less than 1 percent from the pretransient value. This limit is consistent with proven practice. (Ref. 3)

C. Vibration and Fatigue

1. Strain Fatigue The cumulative strain fatigue cycles are less than the design strain fatigue life. This basis is consistent with proven practice. (Ref. 3)

2. Vibration Potential fretting wear due to vibration is prevented, ensuring that the stress-strain limits are not exceeded during design life. Fretting of the clad surface can occur due to flow-induced vibration between the fuel rods and fuel assembly grid springs. Vibration and fretting forces vary during the fuel life due to clad diameter creep-down combined with grid spring relaxation.

D. Chemical properties of the cladding are discussed in reference 4 for Zircaloy-4, reference 25 for ZIRLO, and reference 28 for Optimized ZIRLO.

4.2.1.2 Fuel Material A. Thermal-Physical Properties The center temperature of the hottest pellet is below the melting temperature of the UO2 (melting point of 5080°F (ref. 4) unirradiated and decreasing by 58°F per 10,000 MWd/tonne of uranium). While a limited amount of center melting can be tolerated, the design conservatively precludes center melting. A calculated fuel centerline temperature of 4700°F has been selected as an overpower limit to ensure no fuel melting. This provides sufficient margin for uncertainties as described in paragraph 4.4.2.9.

The normal design density of the fuel is 95 percent of theoretical. Additional information on fuel properties is given in reference 4.

B. Fuel Densification and Fission Product Swelling The design bases and models used for fuel densification and swelling are provided in references 5, 6, and 23.

C. Chemical Properties Reference 4 for Zircaloy-4, reference 25 for ZIRLO, and reference 28 for Optimized ZIRLO provide the basis for justifying that no adverse chemical interactions occur between the fuel and its adjacent material.

4.2-3 REV 24 10/22

VEGP-FSAR-4 4.2.1.3 Fuel Rod Performance A. Fuel Rod Models The basic fuel rod models and the ability to predict operating characteristics are given in references 23, 25, and 27, and subsection 4.2.3. Beginning with Unit 2 Cycle 9 and Unit 1 Cycle 11, the NRC-approved fuel performance model, PAD 4.0 (ref. 27) is acceptable for use in fuel rod design.

B. Mechanical Design Limits Cladding collapse shall be precluded during the fuel rod design lifetime. The models described in references 7 and 26 are used for this evaluation.

The rod internal gas pressure remains below the value which causes the fuel/clad diametral gap to increase due to outward cladding creep during steady-state operation. Rod pressure is also limited such that extensive departure from nucleate boiling (DNB) propagation does not occur during normal operation and any accident event. Reference 8 shows that the DNB propagation criteria are satisfied.

4.2.1.4 Spacer Grids A. Mechanical Limits and Materials Properties The grid component strength criteria are based on experimental tests. The limit is established at 0.9 Pc, where Pc is the experimental collapse load. This limit is sufficient to ensure that, under worst-case combined seismic and blowdown loads from a Condition 4 loss-of-coolant accident (LOCA), the core will maintain a geometry amenable to cooling. As an integral part of the fuel assembly structure, the grids satisfy the applicable fuel assembly design bases and limits defined in paragraph 4.2.1.5.

The grid material and chemical properties are given in references 4, 5 and 25.

B. Vibration and Fatigue The grids provide sufficient fuel rod support to limit fuel rod vibration and maintain clad fretting wear within acceptable limits (defined in paragraph 4.2.1.1).

4.2.1.5 Fuel Assembly Structural Design As previously discussed in subsection 4.2.1, the structural integrity of the fuel assemblies is ensured by setting design limits on stresses and deformations due to various nonoperational, operational, and accident loads. These limit bases are applied to the design and evaluation of the top and bottom nozzles, guide thimbles, grids, and thimble joints.

The design bases for evaluating the structural integrity of the fuel assemblies are:

A. Nonoperational This is a loading of 4 g axial and 6 g lateral with dimensional stability.

B. Normal Operating and Upset Conditions 4.2-4 REV 24 10/22

VEGP-FSAR-4 The fuel assembly component structural design criteria are established for the two primary material categories, austenitic steels and Zircaloy/ZIRLO. The stress categories and strength theory presented in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section III, are used as a general guide. The maximum shear theory (Tresca criterion) for combined stresses is used to determine the stress intensities for the austenitic steel components. The stress intensity is defined as the numerically largest difference between the various principal stresses in a three-dimensional field. The allowable stress intensity value for austenitic steels, such as nickel-chromium-iron alloys, is given by the lowest of the following:

1. One-third of the specified minimum tensile strength or two-thirds of the specified minimum yield strength at room temperature.

2. One-third of the tensile strength or 90 percent of the yield strength at room temperature but not to exceed two-thirds of the specified minimum yield strength at room temperature.

The stress limits for the austenitic steel components are given below. All stress nomenclature is per the ASME Code,Section III.

Stress Intensity Limits Categories Limit General primary Sm membrane stress intensity Local primary membrane 1.5 Sm stress intensity Primary membrane plus 1.5 Sm bending stress intensity Total primary plus 3.0 Sm secondary stress intensity The Zircaloy/ZIRLO structural components, which consist of guide thimbles and fuel tubes, are in turn subdivided into two categories because of material differences and functional requirements. The fuel tube design criteria are covered separately in paragraph 4.2.1.1. The maximum shear theory is used to evaluate the guide thimble design. For conservative purposes, the Zircaloy/ZIRLO unirradiated properties are used to define the stress limits.

3. Abnormal loads during Conditions 3 or 4, worst cases represented by seismic and blowdown loads, are as follows:

a. Deflections or failures of components cannot interfere with the reactor shutdown or emergency cooling of the fuel rods.

b. The fuel assembly structural component stresses under faulted conditions are evaluated using primarily the methods outlined in Appendix F of the ASME Code,Section III. Since the current 4.2-5 REV 24 10/22

VEGP-FSAR-4 analytical methods utilize elastic analysis, the stress allowables are defined as the small value of 2.4 Sm or 0.70 Su for primary membrane and 3.6 Sm or 1.05 Su for primary membrane plus primary bending. For the austenitic steel fuel assembly components, the stress intensity is defined in accordance with the rules described in the previous section for normal operating conditions. For the Zircaloy components, the stress intensity limits are set at two-thirds of the material yield strength, Sy, at reactor operating temperature. This results in Zircaloy stress limits being the smaller of 1.6 Sy or 0.70 Su for primary membrane and 2.4 Sy or 1.05 Su for primary membrane plus bending. For conservative purposes, the Zircaloy/ZIRLO unirradiated properties are used to define the stress limits.

The material and chemical properties of the fuel assembly components are given in reference 4 for Zircaloy-4 and reference 25 for ZIRLO. See paragraph 4.2.3.4 for a discussion of the spacer grid crush testing.

c. Thermal-hydraulic design is discussed in section 4.4.

4.2.1.6 Incore Control Components The control components are subdivided into permanent and temporary devices.

The permanent components are the rod cluster control assemblies (RCCAs), secondary neutron source assemblies, and thimble plug assemblies. The temporary components are the burnable absorber (BA) assemblies, stainless steel rod insert assemblies (SSRIA), and the primary neutron source assemblies, which are normally used only in the initial core.

Materials are selected for:

A. Compatibility in a pressurized water reactor (PWR) environment.

B. Adequate mechanical properties at room and operating temperatures.

C. Resistance to adverse property changes in a radioactive environment.

D. Compatibility with interfacing components.

Material properties are given in reference 4.

The design bases for each of the mentioned components are given in the following paragraphs.

A. Control Rods For Conditions 1 and 2, the stress categories and strength theory presented in the ASME B&PV Code,Section III, Subsection NG-3000, are used as a general guide in the design of the control rod cladding.

Design conditions considered under the ASME Code,Section III, are as follows:

1. External pressure equal to the reactor coolant system (RCS) operating pressure with appropriate allowance for overpressure transients.

2. Wear allowance equivalent to 1000 reactor trips.

3. Bending of the rod due to a misalignment in the guide tube.

4.2-6 REV 24 10/22

VEGP-FSAR-4

4. Forces imposed on the rods during rod drop.

5. Loads imposed by the accelerations of the control rod drive mechanism.

6. Radiation exposure during maximum core life. The absorber material temperature does not exceed its melting temperature (1472°F for silver-indium-cadmium (Ag-In-Cd) (ref. 4) or 3913°F for hafnium (ref. 9)).

7. Temperature effects at operating conditions.

B. BA Rods For Conditions 1 and 2, the stress categories and strength theory presented in the ASME B&PV Code,Section III, Subsection NG-3000, are used as a general guide in the design of the BA cladding. For abnormal loads during Conditions 3 and 4, code stresses are not considered limiting. Failure of the BA rods during these conditions must not interfere with reactor shutdown or cooling of the fuel rods. Thus the structural elements are designed to prevent excessive slumping.

The standard burnable absorber material is borosilicate glass and is designed so that the absorber material is below its softening temperature 1510°F +/- 18°F for 12.5 weight-percent boron rod. The softening temperature for borosilicate glass is defined in ASTM C 338.

The wet annular burnable absorber (WABA) material is B4C contained in an alumina matrix. Thermal-physical and gas release properties of Al203-B4C are described in references 4 and 22. Wet annular burnable absorber rods are designed so that the absorber temperature does not exceed 1200°F during normal operation or an overpower transient. The 1200°F maximum temperature He gas release in a WABA rod will not exceed 30 percent (ref. 22).

C. Neutron Source Rods The neutron source rods are designed to withstand the following:

1. The external pressure equal to RCS operating pressure with appropriate allowance for overpressure transients.

2. An internal pressure equal to the pressure generated by released gases over the source rod life.

D. Thimble Plug Assembly The thimble plug assembly may be used to restrict bypass flow through those thimbles not occupied by absorber, source, or BA rods.

When used, the thimble plug assemblies satisfy the following:

1. Accommodate the differential thermal expansion between the fuel assembly and the core internals.

2. Maintain positive contact with the fuel assembly and the core internals.

3. Limit the flow through each occupied thimble to acceptable design value.

E. Stainless Steel Rod Insert Assembly The SSRIA may be used to provide power suppression and power shaping capability in a fuel assembly to restore design margin to meet the fuel rod internal pressure criterion. The SSRIA core component is comprised of stainless steel 4.2-7 REV 24 10/22

VEGP-FSAR-4 rods mounted to a conventional fuel assembly insert holddown device. The SSRIA satisfies all applicable core component design criteria.

4.2.1.7 Surveillance Program Paragraph 4.2.4.5 and sections 8 and 23 of references 3 and 10 discuss the testing and fuel surveillance operation experience program that has been and is being conducted to verify the adequacy of the fuel performance and design bases. Fuel surveillance and testing results, as they become available, are used to improve fuel rod design and manufacturing processes and to ensure that the design bases and safety criteria are satisfied. The improved corrosion resistance of ZIRLO fuel rod cladding has been demonstrated to high burnups in the BR-3 and North Anna demonstration programs. Cladding corrosion measurements showed that the reduced corrosion exhibited by the ZIRLO clad fuel rods was better than anticipated.

Reference 28 discusses the differences between ZIRLO and Optimized ZIRLO cladding material as well as provides information on the additional fuel surveillance programs demonstrating the similarities with ZIRLO cladding material.

4.

2.2 DESCRIPTION

AND DESIGN DRAWINGS The fuel assembly, fuel rod, and incore control component design data are given in table 4.3-1.

The VANTAGE 5 assembly has the same cross-sectional envelope as the LOPAR assembly.

The VANTAGE 5 assembly is, however, 0.210-in. longer than the initial core/Cycle 1 LOPAR assembly for both units. The grid centerline elevations of the VANTAGE 5 are identical to those of the Region 5/Cycle 3 LOPAR assembly for Unit 1 and Region 4/Cycle 2 LOPAR assembly for Unit 2. The VANTAGE 5 bottom grid is 0.355-in. lower than the LOPAR design of Cycle 1, and the top grid is 0.360-in. lower than the same LOPAR design. However, any integral contact between the two fuel assemblies will be grid-to-grid. By matching grid elevations, any crossflow maldistribution between the LOPAR and VANTAGE 5 fuel assemblies is minimized. To accommodate the VANTAGE 5 fuel assembly with ZIRLO clad fuel rods and the protective bottom grid with the elongated bottom end plug, Inconel grid, No. 1, was raised by 0.7 in., and the first Zircaloy-4/ZIRLO grid, No. 2, was lowered 0.3 in. to reduce the lower most span length in the ZIRLO fuel assembly. This resulted in span No. 1 being shortened by 1.0 in. and span No. 2 being lengthened by 0.3 in. as shown in figure 4.2-2.

Commencing with Region 10 in Unit 1 Cycle 8 and Region 9 in Unit 2 Cycle 7, the fuel assembly skeleton includes ZIRLO fabricated guide thimble tubes and instrumentation tubes. The ZIRLO guide thimble tube and instrumentation tube lengths have been reduced, resulting in a slightly shorter overall fuel assembly height to allow for additional growth at high burnup. The top nozzle holddown spring design for the shorter ZIRLO fuel assembly has a slightly increased height compared to the standard top nozzle holddown spring design to maintain comparable holddown forces. The top Inconel grid was lowered by 0.2 inch because of the shorter overall height of the ZIRLO skeleton.

Commencing with Region 12 in Unit 1 Cycle 10 and Region 11 in Unit 2 Cycle 9, the ZIRLO guide thimble tube and instrumentation tube lengths have been restored to the same length as the VANTAGE 5 design with the Zircaloy-4 skeleton. The top Inconel grid for the longer ZIRLO skeleton design has been returned to the same elevation as in the VANTAGE 5 design with the Zircaloy-4 skeleton. The standard top nozzle holddown spring is used on the longer ZIRLO fuel assembly to maintain comparable holddown forces.

4.2-8 REV 24 10/22

VEGP-FSAR-4 Commencing with Region 27 (Cycle 25) reload for Unit 1 and Region 25 (Cycle 23) reload for Unit 2, the Vogtle units implement the Westinghouse 17x17 PRIMETM fuel assembly design which is an improved version of the 17x17 VANTAGE+ fuel design. The PRIMETM fuel assembly design incorporates advanced features to improve fuel reliability, dimensional stability and corrosion resistance of the 17x17 VANTAGE+ with Debris Mitigation Features fuel design will be implemented. The PRIMETM fuel features include Low Tin ZIRLO grid strap material to improve corrosion resistances, the PRIMETM Bottom Nozzle incorporates a low pressure drop design along with lower side skirts to help improve debris resistance and the external dashpot tube which stiffens the dashpot region of the fuel assembly to minimize the potential for incomplete rod insertion. The PRIMETM advanced fuel design has been extensively evaluated, tested and analyzed and it was demonstrated that it meets all of the applicable mechanical design and safety criteria.

Optimized ZIRLO cladding material has been approved for use in Vogtle 1 and 2 by the NRC, reference 29.

Each fuel assembly consists of 264 fuel rods, 24 guide thimble tubes, and 1 instrumentation thimble tube arranged within a supporting structure. A fuel assembly may have limited substitution of zirconium alloy or stainless steel filler rods in place of fuel rods, in accordance with NRC approved applications of fuel rod configuration. The instrumentation thimble is located in the center position and provides a channel for insertion of an incore neutron detector, if the fuel assembly is located in an instrumented core position. The guide thimbles provide channels for insertion of either an RCCA, a neutron source assembly, a BA assembly, or a thimble plug assembly, depending on the position of the particular fuel assembly in the core.

Figure 4.2-1 shows a cross-section of the fuel assembly array, and figure 4.2-2 shows a fuel assembly full-length view. The fuel rods are loaded into the fuel assembly structure so that there is clearance between the fuel rod ends and the top and bottom nozzles.

Fuel assemblies are installed vertically in the reactor vessel and stand upright on the lower core plate, which is fitted with alignment pins to locate and orient each assembly. After all fuel assemblies are set in place, the upper support structure is installed. Alignment pins, built into the upper core plate, engage and locate the upper ends of the fuel assemblies. The upper core plate then bears down against the holddown springs on the top nozzle of each fuel assembly to hold the fuel assemblies in place.

Improper orientation of fuel assemblies within the core is prevented by the use of an indexing hole in one corner of the top nozzle top plate. The assembly is oriented with respect to the handling tool and the core by means of a pin inserted into this indexing hole. Visual confirmation of proper orientation is also provided by an engraved identification number on the opposite corner clamp.

4.2.2.1 Fuel Rods The fuel rods consist of uranium dioxide ceramic pellets contained in slightly cold-worked Zircaloy-4 tubing plugged and seal-welded at the ends to encapsulate the fuel. A schematic of the fuel rod is shown in figure 4.2-3. The fuel pellets are right circular cylinders consisting of slightly enriched uranium dioxide powder which has been compacted by cold pressing and then sintered to the required density. The ends of each pellet are dished slightly to allow greater axial expansion at the center of the pellets. For reloads, the ends of each pellet also have a small chamfer at the outer cylindrical surface which improves manufacturability.

4.2-9 REV 24 10/22

VEGP-FSAR-4 The VANTAGE 5 fuel rod has the same clad wall thickness as the LOPAR fuel rod, but the VANTAGE 5 fuel rod diameter is reduced to optimize the water-to-uranium ratio. The VANTAGE 5 fuel rod length is greater than the LOPAR fuel rod length to provide a longer plenum and bottom end plug. The bottom end plug has an internal-grip feature to facilitate rod loading on both designs and is longer to provide a longer lead-in for the removable top nozzle reconstitution feature.

Commencing with the Region 8 fuel rods in Unit 2 Cycle 6 and the Region 10 fuel rods in Unit 1 Cycle 8, the fuel rod bottom end plug was elongated for use with the protective bottom grid. In conjunction with the elongated bottom end plug, the fuel rod top end plug was elongated and fitted with an external gripper.

Commencing with the Region 12 fuel rods in Unit 1 Cycle 10 and Region 11 fuel rods in Unit 2 Cycle 9, the ZIRLO fuel rod length has been increased based on the growth model for ZIRLO (reference 25). In addition, the fuel rod top end plug is reduced in length by eliminating the external gripper feature. The reduced top end plug length allows a further increase in the fuel rod tube length, and, consequently, an increase in the fuel rod plenum.

Optimized ZIRLO cladding material has been approved for use in Vogtle 1 and 2 by the NRC, reference 29.

The axial blanket region is the nominal 6 inches of fuel pellets located at each end of the fuel rod pellet stack. The fuel in the axial blanket region may be natural, mid-enriched, or fully enriched, solid or annular pellets. The natural or mid-enriched axial blankets reduce neutron leakage and improve fuel utilization. The annular blanket pellets are used to increase the void volume for gas accommodation within the fuel rod. The axial blankets utilize chamfered pellets which are physically different (length) than the enriched pellets to help prevent accidental mixing during manufacturing.

The integral fuel burnable absorber (IFBA) coated fuel pellets are identical to the enriched uranium dioxide pellets except for the addition of a thin zirconium diboride (ZrB2) coating on the pellet cylindrical surface. Coated pellets occupy the central portion of the fuel stack. The number and pattern of IFBA rods within an assembly may vary depending on specific application. The ends of the enriched coated pellets are dished to allow for greater axial expansion at the pellet centerline and to increase the void volume for fission gas release. This coating may be applied with a linear B-10 loading (mg/in.) that is greater than the original IFBA design for added flexibility in the core design.

Void volume and clearances are provided within the rods to accommodate fission gases released from the fuel, differential thermal expansion between the clad and the fuel, and fuel density changes during irradiation. Shifting of the fuel within the clad during handling or shipping prior to core loading is prevented by a stainless steel helical spring which bears on top of the fuel. At assembly, the pellets are stacked in the clad to the required fuel height. The spring is then inserted into the top end of the fuel tube and the end plugs pressed into the ends of the tube and welded. All fuel rods are internally pressurized with helium during the welding process to minimize compressive clad stresses and prevent clad flattening under coolant operating pressures.

The fuel rods are prepressurized and designed so that:

• The internal gas pressure mechanical design limit referred to in paragraph 4.2.1.3 is not exceeded.

• The cladding stress-strain limits (paragraph 4.2.1.1) are not exceeded for Condition 1 and 2 events.

4.2-10 REV 24 10/22

VEGP-FSAR-4

• Clad flattening will not occur during the fuel core life.

4.2.2.2 Fuel Assembly Structure The fuel assembly structure consists of a bottom nozzle, top nozzle, guide thimbles, and grids, as shown in figure 4.2-2.

4.2.2.2.1 Bottom Nozzle The bottom nozzle serves as a bottom structural element of the fuel assembly and directs the coolant flow distribution to the assembly. The nozzle is fabricated from type 304 stainless steel and is shown in figure 4.2-2. The legs form a plenum for the inlet coolant flow to the fuel assembly. The plate prevents accidental downward ejection of the fuel rods from the fuel assembly. The bottom nozzle is fastened to the fuel assembly guide tubes by locked thimble screws which penetrate through the nozzle and mate with a threaded plug in each guide tube.

Coolant flows from the plenum in the bottom nozzle upward through the penetrations in the plate to the channels between the fuel rods. The penetrations in the plate are positioned between the rows of the fuel rods.

Axial loads (holddown) imposed on the fuel assembly and the weight of the fuel assembly are transmitted through the bottom nozzle to the lower core plate. Indexing and positioning of the fuel assembly is controlled by alignment holes in two diagonally opposite bearing plates which mate with locating pins in the lower core plate. Lateral loads on the fuel assembly are transmitted to the lower core plate through the locating pins.

A debris filter bottom nozzle (DFBN) was introduced as a replacement to the standard nozzle design described above to inhibit debris from entering the active fuel region of the core. This nozzle utilized the same material, geometry, functional, and welding requirements as the standard bottom nozzle as described above. The revised end plate flow hole patterns are shown on figure 4.2-2 to illustrate the increased number of smaller flow holes that reduce the passage of debris into the active region of the fuel assembly. Additional debris protection was provided by the protective bottom grid assembly and the elongated fuel rod bottom end plug described in paragraph 4.2.2.2.4.

A modification introducing a skirt on the DFBN was made to improve the structural integrity of the nozzle on the VANTAGE 5 design. In addition, five holes were placed on each face of this skirt to allow lateral flow communication. This change in bottom nozzle design was referred to as the modified DFBN and is illustrated on figure 4.2-2 (VANTAGE 5).

Commencing with the fresh fuel installed in Unit 1 Cycle 18, Unit 2 Cycle 17, and subsequent Vogtle reloads, the standardized debris filter bottom nozzle (SDFBN), which was developed for 17X17 fuel and is designed to have a loss coefficient that is the same, independent of supplier, will be implemented. The SDFBN has eliminated the side skirt communication flow holes as a means of improving the debris mitigation performance of the bottom nozzle, as shown in figure 4.2-2. This nozzle has been extensively evaluated and analyzed and it was demonstrated that it meets all of the applicable mechanical design criteria. In addition, specific testing was performed to demonstrate that there is no adverse effect on the thermal hydraulic performance of the SDFBN either with respect to the pressure drop or with respect to DNB.

Beginning with the Region 27 (Cycle 25) reload for Unit 1 and Region 25 (Cycle 23) reload for Unit 2, the Vogtle units implement the Westinghouse PRIMETM fuel assembly design. The PRIMETM fuel assembly bottom nozzle design has a two chamfer inlet flow hole design to 4.2-11 REV 24 10/22

VEGP-FSAR-4 improve the pressure drop (lower the loss coefficient). In addition, the PRIMETM bottom nozzle design lowers the side skirts to help improve the debris filtering capability of the bottom nozzle by eliminating the large lateral flow path between the bottom nozzle legs in the SDFBN design by lowering the side skirts of the PRIMETM bottom nozzle. In addition, small flow holes are added to the side skirts to ensure that there is adequate lateral flow to the baffle barrel region of the core to ensure that the reactor vessel former plants are cooled. These features are shown in Figure 4.2-2 (Sheet 7 of 7). This advanced bottom nozzle design has been extensively evaluated, tested and analyzed and it was demonstrated that it meets all of the applicable mechanical design and safety criteria.

4.2.2.2.2 Top Nozzle Standard Top Nozzle The standard top nozzle functions as the upper structural element of the fuel assembly in addition to providing a partial protective housing for the RCCA or other components. The top nozzle consists of an adapter plate, enclosure, top plate, and pads. Holddown springs are mounted on the assembly, as shown in figure 4.2-2. The springs and bolts are made of Inconel-718, and -600, respectively; other components are made of type 304 stainless steel.

The adapter plate is provided with round penetrations and semicircular-ended slots to permit the flow of coolant upward through the top nozzle. Other round holes are provided to accept sleeves which are welded to the adapter plate and mechanically attached to the thimble tubes.

The ligaments in the plate cover the tops of the fuel rods and prevent their upward ejection from the fuel assembly. The enclosure is a box-like structure which sets the distance between the adapter plate and the top plate. The top plate has a large hole in the center to permit access for the control rods and the control rod spiders. Holddown springs are mounted on the top plate and are fastened in place by bolts and clamps located at two diagonally opposite corners.

Integral pads which contain alignment holes for locating the upper end of the fuel assembly are positioned on the other two corners.

Reconstitutable Top Nozzle The RTN design differs from the above standard nozzle design in the following ways. A groove is provided in each thimble thru-hole in the nozzle plate to facilitate attachment and removal.

Round holes in the adapter plate are provided to accept nozzle inserts which are locked into the grooves in the adapter plate using lock tubes and mechanically attached to the thimble tubes at the lower ends of the inserts (figure 4.2-5). The nozzle plate thickness is reduced to provide additional axial space for fuel rod growth. Holddown springs are mounted on the top plate and are fastened in place by spring screws.

Commencing with the fresh fuel installed in Unit 1 Cycle 11 and in Unit 2 Cycle 9, Alloy 718 is used instead of Alloy 600 for the fuel assembly holddown spring screws on the RTN to improve resistance to primary water stress corrosion cracking.

To remove the RTN, a tool is first inserted through a lock tube (figure 4.2-5) and expanded radially to engage the bottom edge of the tube. An axial force is then exerted on the tool which overrides local lock tube deformations and withdraws the lock tube from the nozzle insert. After the lock tubes have been withdrawn, the RTN is removed by raising it off the upper slotted ends of the nozzle inserts, which deflect inwardly under the axial lift load.

With the RTN removed, direct access is provided for fuel rod examination or replacement.

Reconstitution is completed by the remounting of the RTN and the insertion of lock tubes.

4.2-12 REV 24 10/22

VEGP-FSAR-4 Additional details of this design feature, the design bases, and evaluation of the RTN are given in subsection 2.3.2 in reference 21.

A composite (cast) top nozzle was introduced in Unit 1 Cycle 8 and in Unit 2 Cycle 7, and replaced the integral welded assembly. This design change, for manufacturing process improvement, reduces the total number of components required to fabricate and assemble the top nozzle. The cast and welded top nozzles are interchangeable regarding form, fit, and function. Therefore, either of these RTN designs may be used.

Westinghouse Integral Nozzle Commencing with Vogtle Unit 1, Cycle 20, the Westinghouse Integral Nozzle (WIN) will replace the RTN. The WIN top nozzle functions as the upper structural element of the fuel assembly in addition to providing a partial protective housing for the RCCA or other components. The top nozzle consists of an adapter plate, enclosure, top plate, and pads. Holddown springs are mounted on the assembly, as shown in figure 4.2-17. For the WIN, the springs are made of Alloy 781 and the main nozzle body and the pins are made of Type 304 stainless steel.

The WIN design, while similar to the RTN, incorporates design and manufacturing improvements to eliminate the Alloy 718 spring screw for attachment of the holddown springs.

In the WIN nozzle, the springs are assembled into the nozzle pad and pinned in place. The WIN design provides a wedged rather than a clamped (bolted) joint for transfer of the fuel assembly holddown forces into the top nozzle structure. Integral pads which contain alignment holes for locating the upper end of the fuel assembly are positioned on the other two corners for the WIN. The flow plate, thermal characteristics, and method of attachment of the nozzle are all unchanged from the RTN top nozzle design.

Replacement Reconstitutable Top Nozzle (RRTN)

A replacement reconstitutable top nozzle (RRTN) design may be used in a reload cycle to replace the original reconstitutable top nozzle (RTN) or Westinghouse Integral Top Nozzle (WIN) on an irradiated fuel assembly. The mechanical features of the RRTN are the same as those for the RTN (see figure 4.2-5) or WIN with some minor dimensional differences in the top nozzle adapter plate thimble hole to facilitate attachment to an irradiated fuel assembly. The RRTN may be manufactured by either the composite (cast) process or the integral welded process. The cast and integral welded top nozzles are interchangeable regarding form, fit and function. Therefore, either of these RRTN designs may be used to replace an RTN or WIN regardless of whether the original RTN or WIN on the assembly was the composite (cast) design or the integral welded design. Either Alloy 718 or Alloy 600 may be used for the fuel assembly holddown spring screws on the RRTN.

4.2.2.2.3 Guide Thimbles and Instrument Tube The guide thimbles are structural members which also provide channels for the neutron absorber rods, BA rods, neutron source, or thimble plug assemblies. With the exception of a reduction in the guide thimble diameter and length above the dashpot, the VANTAGE 5 guide thimbles are identical to those in the LOPAR design. A reduction to the guide thimble outside diameter and inside diameter is required due to the thicker Zircaloy grid straps and reduced cell size. The VANTAGE 5 thimble tube is shorter than the LOPAR thimble tube due to the RTN or WIN and nozzle insert features. Each thimble is fabricated from Zircaloy-4/ ZIRLO tubing having two different diameters. The tube diameter at the top section provides the annular area necessary to permit rapid control rod insertion during a reactor trip. Holes are provided on the thimble tube above the dashpot to reduce the rod drop time. The lower portion of the guide thimble is swaged to a smaller diameter to reduce diametral clearances and produce a dashpot 4.2-13 REV 24 10/22

VEGP-FSAR-4 action near the end of the control rod travel during normal trip operation. The dashpot is closed at the bottom by means of an end plug, which is provided with a small flow port to avoid fluid stagnation in the dashpot volume during normal operation. The top end of the guide thimble is fastened to a tubular sleeve by three expansion swages on the standard top nozzle. The sleeve fits into and is welded to the standard top nozzle adapter plate. The top end of the guide thimble is fastened to a nozzle insert by three expansion swages. The nozzle insert fits into and is locked into the RTN or WIN adapter plate using a lock tube. The lower end of the guide thimble is fitted with an end plug, which is captured by the bottom grid insert and then fastened to the bottom nozzle by a crimp locked thimble screw.

Fuel rod support grids are fastened to the guide thimble assemblies to create an integrated structure. Since welding of the grids to the thimbles is not possible, the fastening technique depicted in figures 4.2-4 and 4.2-5 is used for all but the bottom grid in a fuel assembly.

An expanding tool is inserted into the inner diameter of the Zircaloy/ZIRLO thimble tube at the elevation of the grid sleeves that have been previously attached to the grid assembly. The four-lobed tool forces the thimble and sleeve outward to a predetermined diameter, thus joining the two components.

The top grid-to-thimble attachment is shown in figure 4.2-5 for the standard top nozzle design.

The stainless steel sleeves are brazed into the Inconel grid assembly. The Zircaloy/ZIRLO guide thimbles are fastened to the long sleeves by expanding the two members as shown in figures 4.2-4 and 4.2-6. Finally, the top ends of the sleeves are welded to the top nozzle adapter plate as shown in figure 4.2-5.

In assemblies with reconstitutable top nozzles or Westinghouse Integral Nozzles, the guide thimbles are fastened inside the top grid sleeves and nozzle inserts as shown in figure 4.2-5. A bulge in the nozzle insert is then captured in a corresponding groove in the hole in the top nozzle plate. The insert is fixed in place by the insertion of a lock tube into the insert, thus providing a mechanical connection between the guide thimble and the top nozzle.

The top Inconel grid sleeve, top nozzle insert, and thimble of the VANTAGE 5 design are joined together using bulge joint mechanical attachments as shown in figure 4.2-5. This bulge joint connection was mechanically tested and found to meet all applicable design criteria.

The intermediate mixing vane and intermediate flow mixer Zircaloy/ZIRLO/Low Tin ZIRLO grids employ a single bulge connection to the sleeve and thimble as compared to a two bulge connection used in the top Inconel grid (figure 4.2-6). Mechanical testing of this bulge joint connection was also found to be acceptable.

The bottom grid assembly is joined to the assembly as shown in figure 4.2-7. The stainless steel insert is spotwelded to the bottom grid and later captured between the guide thimble end plug and the bottom nozzle by means of a stainless steel thimble screw.

The described methods of grid fastening are standard and have been used successfully since the introduction of Zircaloy guide thimbles in 1969.

The central instrumentation tube of each fuel assembly is constrained by seating in a counterbore in the bottom nozzle at its lower end and is attached to the top and midgrid in the same manner as above for the guide thimbles. This tube has a constant diameter and guides the incore neutron detectors. The VANTAGE 5 instrumentation tube also has a reduction in diameter when compared to the LOPAR assembly instrumentation tube. This decrease still allows sufficient diametral clearance for the flux thimble to traverse the tube without binding.

Beginning with the Region 27 (Cycle 25) reload for Unit 1 and the Region 25 (Cycle 23) reload for Unit 2, the Vogtle units implement the Westinghouse PRIMETM fuel assembly design which incorporates an external dashpot tube to stiffen the dashpot region of the fuel assembly to 4.2-14 REV 24 10/22

VEGP-FSAR-4 minimize the potential for incomplete rod insertion. This advanced dashpot design has been extensively evaluated, tested and analyzed and it was demonstrated that it meets all of the applicable mechanical design and safety criteria.

4.2.2.2.4 Grid Assemblies The fuel rods, as shown in figure 4.2-2, are supported at intervals along their length by grid assemblies which maintain the lateral spacing between the rods. Each fuel rod is supported within each grid by the combination of support dimples and springs. The grid assembly consists of individual slotted straps assembled and interlocked into an egg crate arrangement with the straps permanently joined at their points of intersection.

Two types of grid assemblies are used in each fuel assembly. One type, with mixing vanes projecting from the edges of the straps into the coolant stream, is used in the high heat flux region of the fuel assemblies to promote mixing of the coolant. The other type, located at the ends of the assembly, does not contain mixing vanes on the internal straps. The outside straps on all grids contain mixing vanes which, in addition to their mixing function, aid in guiding the grids and fuel assemblies past projecting surfaces during handling or during loading and unloading of the core.

The top and bottom Inconel-718 (nonmixing vane) grids of the VANTAGE 5 fuel assemblies are similar in design to the Inconel grids of the Cycle 1, LOPAR fuel assemblies for VEGP Units 1 and 2. The six intermediate (mixing vane) grids on the VANTAGE 5 design are made of Zircaloy/ZIRLO/Low Tin ZIRLO material rather than the intermediate Inconel (mixing vane) grids which are currently used in the LOPAR design. Beginning with Unit 1 Cycle 8 and Unit 2 Cycle 7, the intermediate grids and intermediate flow mixer (IFM) grids are made of ZIRLO material.

The Zircaloy/ZIRLO/Low Tin ZIRLO grids have thicker straps than the Inconel. Also, the Zircaloy/ZIRLO/Low Tin ZIRLO grid height is higher compared to the Inconel midgrid. These dimensional changes were made to compensate for differences in material strength properties.

The Zircaloy/ZIRLO/Low Tin ZIRLO grid incorporates the same grid cell support configuration as the Inconel grid. The Zircaloy/ZIRLO/Low Tin ZIRLO interlocking strap joints and grid/sleeve joints are fabricated by laser welding, whereas the Inconel grid joints are brazed.

The VANTAGE 5 IFM grids shown on figure 4.2-2 are located in the three uppermost spans between the Zircaloy/ZIRLO/Low Tin ZIRLO mixing vane structural grids and incorporate a similar mixing vane array. Their prime function is midspan flow mixing in the hottest fuel assembly spans. Each IFM grid cell contains four dimples which are designed to prevent midspan channel closure in the spans containing IFMs and fuel rod contact with the mixing vanes. This simplified cell arrangement allows short grid cells so that the IFM grid can accomplish its flow mixing objective with minimal pressure drop.

The IFM grids, like the VANTAGE 5 mixing vane grids, are fabricated from Zircaloy/ZIRLO/Low Tin ZIRLO. This material was selected to take advantage of the material's inherent low neutron capture cross-section. The Zircaloy/ZIRLO/Low Tin ZIRLO grid straps are manufactured using the same basic techniques as the Zircaloy/ZIRLO/Low Tin ZIRLO grid assemblies used for the Westinghouse optimized fuel assembly (OFA) design and are joined to the guide thimbles via sleeves which are welded at the bottom of appropriate grid cells.

The IFM grids are not intended to be structural members. The outer strap configuration was to be similar to current fuel designs to preclude grid hang-up and damage during fuel handling.

Additionally, the grid envelope is smaller which further minimizes the potential for damage and reduces calculated forces during seismic/LOCA events. A coolable geometry is assured at the IFM grid elevation, as well as at the structural grid elevation.

4.2-15 REV 24 10/22

VEGP-FSAR-4 Commencing with the fresh feed fuel assemblies for Unit 2 Cycle 6 and Unit 1 Cycle 8, the fuel incorporated a bottom protective grid and a modification to the bottom fuel rod end plug. The protective grid illustrated in figure 4.2-2 is a partial height grid, similar in configuration to the intermediate flow mixing grid, fabricated of Inconel without mixing vanes, and positioned on the top plate of the bottom nozzle. In conjunction with the protective grid, the bottom fuel rod end plug is elongated. The protective grid and elongated bottom end plug together provide a zone below the active fuel in which debris can be entrapped.

Commencing with the fresh fuel installed in Unit 1 Cycle 18, Unit 2 Cycle 17, and subsequent Vogtle reload, the robust protective grid (RPG) which was developed as a result of observed failures in the field as noted in post irradiation exams (PIE) performed at several different plants, will be implemented. It was determined that observed failures were the result of two primary issues: 1) fatigue failure within the protective grid itself at the top of the end strap and 2) stress corrosion cracking (SCC) primarily within the rod support dimples. The RPG implements design changes such as increasing the maximum nominal height of the grid, increasing the ligament length and the radii of the ligament cutouts, and the use of four additional spacers for a total of eight spacers to help strengthen the grid. The nominal height of the grid was increased to allow V-notch window cutouts to be added to help minimize flow-induced vibration caused by vortex shedding at the trailing edge of the inner grid straps. These design changes incorporated into the RPG design help address the issues of fatigue failures and failures due to SCC. It was demonstrated that the above changes do not impact the thermal hydraulic performance of the RPG as there is no change to the pressure loss coefficient. In addition, the RPG retains the original protective grid function as a debris mitigation feature.

Beginning with the Region 27 (Cycle 25) reload for Unit 1 and the Region 25 (Cycle 23) reload Unit 2, the Vogtle units implement the Westinghouse PRIMETM fuel assembly design which incorporates Low Tin ZIRLO grid strap material to improve corrosion resistance as compared to grids manufactured with ZIRLO grid strap material. This Low Tin ZIRLO grid strap material has been extensively evaluated, tested and analyzed and it was demonstrated that all of the applicable mechanical design and safety criteria remain satisfied.

4.2.2.3 Incore Control Components Reactivity control is provided by neutron absorbing rods and a soluble chemical neutron absorber (boric acid). The boric acid concentration is varied to control long-term reactivity changes such as:

A. Fuel depletion and fission product buildup.

B. Cold to hot, zero power reactivity changes.

C. Reactivity change produced by intermediate-term fission products such as xenon and samarium.

D. Burnable poison depletion.

The chemical and volume control system (CVCS) is discussed in chapter 9.

The RCCAs provide reactivity control for:

A. Shutdown.

B. Reactivity changes due to coolant temperature changes in the power range.

C. Reactivity changes associated with the power coefficient of reactivity.

D. Reactivity changes due to void formation.

4.2-16 REV 24 10/22

VEGP-FSAR-4 It is necessary to maintain a negative power coefficient at hot full power conditions throughout the entire cycle to reduce possible deleterious effects caused by a positive coefficient during loss-of-coolant or loss-of-flow accidents. The first fuel cycle contains more excess reactivity than subsequent cycles due to the loading of all fresh (unburned) fuel. Since soluble boron alone is insufficient to ensure a negative moderator coefficient, BA assemblies are also used.

The RCCAs and their control rod drive mechanisms (CRDMs) are the only moving parts in the reactor. Figure 4.2-8 illustrates the rod cluster control (RCC) and CRDM assembly, in addition to the arrangement of these components in the reactor, relative to the interfacing fuel assembly and guide tubes. In the following paragraphs, each reactivity control component is described in detail. The CRDM assembly is described in subsection 3.9.4.

The neutron source assemblies provide a means of monitoring the core during periods of low neutron activity. The thimble plug assemblies may be used to limit bypass flow through those fuel assembly thimbles which do not contain control rods, BA rods, or neutron source rods.

4.2.2.3.1 Rod Cluster Control Assemblies The RCCAs are divided into two categories: control and shutdown. The control groups compensate for reactivity changes due to variations in operating conditions of the reactor (i.e.,

power and temperature variations). Two nuclear design criteria have been employed for selection of the control group. First, the total reactivity worth must be adequate to meet the nuclear requirements of the reactor. Second, in view of the fact that these rods may be partially inserted at power operation, the total power peaking factor should be low enough to ensure that the power capability is met. The control and shutdown groups provide adequate shutdown margin.

An RCCA is comprised of a group of individual neutron absorber rods fastened at the top end to a common spider assembly, as illustrated in figure 4.2-9.

The absorber materials used in the VEGP control rods are either solid Ag-In-Cd bar or hafnium (Hf) bar which are essentially "black" to thermal neutrons and have sufficient additional resonance absorption to significantly increase their worth. The absorber material is sealed in cold worked, high purity stainless steel tubes. (See figure 4.2-10.) A thin chrome electroplate is applied to the tubing outer surface of the Ag-In-Cd RCCA over a specified length which is in contact with the reactor internal guides for the Westinghouse RCCAs; for the Framatome RCCAs, the exterior of the rods are ion-nitrided to harden the surface. The cladding provides increased resistance to tube wear. Sufficient diametral and end clearances are provided to accommodate relative thermal expansion. In addition, the absorber diameter is slightly reduced at the lower extremity of the rodlets in order to accommodate absorber swelling and minimize cladding interaction.

The bottom plugs are made bullet-nosed to reduce the hydraulic drag during reactor trip and to guide smoothly into the dashpot section of the fuel assembly guide thimbles.

The material used in the absorber rod end plugs is type 308 stainless steel for the Westinghouse RCCAs; for the Framatome RCCAs, the absorber rod end plugs are 308L stainless steel. The design stresses used for the type 308P and 308L material are the same as those defined in the ASME Code,Section III, for type 304 stainless steel. At room temperature the yield and ultimate stresses per American Society of Testing Materials (ASTM) 580 are exactly the same for the two alloys. In view of the similarity of composition of the alloys, the temperature dependence of strength for the two materials is also assumed to be the same.

4.2-17 REV 24 10/22

VEGP-FSAR-4 The allowable stresses used as a function of temperature are listed in table 1-1.2 of Section III of the ASME Code. The fatigue strength for the type 308 material is based on the S-N curve for austenitic stainless steels in figure 1-9.2 of Section III.

The spider assembly is in the form of a central hub with radial vanes containing cylindrical fingers from which the absorber rods are suspended. Handling detents and detents for connection to the drive rod assembly are machined into the upper end of the hub. Coil springs inside the spider body absorb the impact energy at the end of a trip insertion. The Westinghouse RCCAs have two coil springs and the Framatome RCCAs have a single coil spring. The radial vanes are joined to the hub by tack-welding and brazing, and the fingers are joined to the vanes by brazing for the Westinghouse RCCAs; for the Framatome RCCAs, the spider is a one-piece casting. A bolt which holds the springs and retainer is threaded into the hub within the skirt and welded to prevent loosening in service. All components of the Westinghouse RCCA spider assembly are made from types 304 and 308 stainless steel except for the retainer, which is of 17-4 PH material, and the springs, which are Inconel-718 alloy. For the Framatone RCCA spider assembly, the spider is made from CF3M (316L) stainless steel, the retainer is 17-4 PH stainless steel, and the spring is Inconel-718 alloy.

The absorber rods are fastened securely to the spider. The rods are first threaded into the spider fingers for the Westinghouse RCCAs; for the Framatome RCCAs, the rods pass through the spider boss and have nuts that thread onto the top of the rods. The Westinghouse and Framatome RCCAs are both pinned to maintain joint tightness, after which the pins are welded in place. The end plug below the pin position is designed with a reduced section to permit flexing of the rods to correct for small operating or assembly misalignments.

The overall length is such that when the assembly is withdrawn through its full travel, the tips of the absorber rods remain engaged in the guide thimbles so that alignment between rods and thimbles is always maintained. Since the rods are long and slender, they are relatively free to conform to any small misalignments with the guide thimble.

4.2.2.3.2 BA Assembly Each BA assembly consists of BA rods attached to a holddown assembly. The BA assemblies are shown in figure 4.2-11. When needed for nuclear considerations, BA assemblies are inserted into selected thimbles within fuel assemblies.

The standard BA rods consist of borosilicate glass tubes contained within type 304 stainless steel tubular cladding plugged and seal-welded at the ends to encapsulate the glass. The glass is also supported along the length of its inside diameter by a thin-wall, tubular inner liner. The top end of the liner is open to permit the diffused helium to pass into the void volume, and the liner overhangs the glass. The liner has an outward flange at the bottom end to maintain the position of the liner with the glass. The cladding in the standard BA rods is slightly cold worked type 304 stainless steel. All other structural materials are type 304 or 308 stainless steel except for the springs, which are Inconel-718. The borosilicate glass tube provides sufficient boron content to meet the criteria discussed in subsection 4.3.1. A typical standard BA rod is shown in longitudinal and transverse cross-sections in figure 4.2-12.

Wet annual burnable absorber rods consist of annular pellets of alumina-boron carbide (AL203-B4C) burnable absorber material contained within two concentric Zircaloy tubes. These Zircaloy tubes which form the inner and the outer clad for the WABA rod are plugged, pressurized with helium, and seal-welded at each end to encapsulate the annular stack of absorber material.

The absorber stack length (figure 4.2-12) is positioned axially within the WABA rods by the use of Zircaloy bottom-end spacers.

4.2-18 REV 24 10/22

VEGP-FSAR-4 An annular plenum is provided within the rod to accommodate the helium gas released from the absorber material as it depletes during irradiation. The reactor coolant flows inside the inner tube and outside the outer tube of the annular rod. A typical WABA rod is shown in longitude and cross-section in figure 4.2-12. Additional design details are given in Section 3.0 of reference 22.

The BA rods in each fuel assembly are grouped and attached together at the top end of the rods to a holddown assembly by a flat, perforated retaining plate, which fits within the fuel assembly top nozzle and rests on the adapter plate.

The retaining plate and the BA rods are held down and restrained against vertical motion through a spring pack which is attached to the plate and is compressed by the upper core plate when the reactor upper internals assembly is lowered into the reactor. This arrangement ensures that the BA rods cannot be ejected from the core by flow forces. Each WABA rod is permanently attached to the base plate by a crimped attaching nut.

4.2.2.3.3 Neutron Source Assembly The purpose of the neutron source assembly is to provide a base neutron level to ensure that the detectors are operational and responding to core multiplication neutrons. A neutron source is placed in the reactor to provide a positive neutron count of at least one-half count per second on the source range detectors attributable to core neutrons. The detectors, called source range detectors, are used primarily when the core is subcritical and during special subcritical modes of operations.

The source assembly also permits detection of changes in the core multiplication factor during core loading, refueling, control rod testing, and approach to criticality. This can be done since the multiplication factor is related to an inverse function of the detector count rate. Changes in the multiplication factor can be detected during addition of fuel assemblies while loading the core, changes in control rod positions, and changes in boron concentration.

Both primary and secondary neutron source rods are used. The primary source rod, containing a radioactive material, spontaneously emits neutrons during initial core loading, reactor startup, and initial operation of the first core. After the primary source rod decays beyond the desired neutron flux level, neutrons are then supplied by the secondary source rod. The secondary source rod contains a stable material, which must be activated during reactor operation. The activation results in the subsequent release of neutrons.

Four source assemblies are installed in the reactor core: two primary source assemblies and two secondary source assemblies. Each primary source assembly contains one primary source rod and a number of BA rods. Each secondary source assembly contains a symmetrical grouping of four secondary source rods. Locations not filled with a source or BA rod contain a thimble plug. The source assemblies are shown in figures 4.2-13 and 4.2-14.

Neutron source assemblies are employed at opposite sides of the core. The assemblies are inserted into the RCC guide thimbles in fuel assemblies at selected unrodded locations.

As shown in figures 4.2-13 and 4.2-14, the source assemblies contain a holddown assembly identical to that of the BA assembly.

The primary and secondary source rods both utilize the same cladding material as the absorber rods. The secondary source rods contain Sb-Be pellets stacked to a height of approximately 88 in. The primary source rods contain capsules of californium (Pu-Be possible alternate) source material and alumina spacer to position the source material within the cladding. The rods in each assembly are permanently fastened at the top end to a holddown assembly.

4.2-19 REV 24 10/22

VEGP-FSAR-4 The other structural members, except for the springs, are constructed of type 304 stainless steel. The springs exposed to the reactor coolant are Inconel-718.

A double-encapsulated, secondary source rod assembly was introduced as a replacement to the single-encapsulated, secondary source rod assembly in Unit 1 Cycle 8 and Unit 2 Cycle 7.

The double-encapsulated, secondary source design provides additional margin against source material leakage. This assembly has the same exterior dimensions as the single-encapsulated, secondary source assembly. The antimony-beryllium pellets (stack height 88 in.) are encapsulated in a pressurized, 304 stainless-steel tube and then inserted in an outer-pressurized, stainless-steel tube as indicated on figure 4.2-14. This assembly makes up the new double-encapsulated, secondary source rod assembly design.

4.2.2.3.4 Thimble Plug Assembly Thimble plug assemblies may be used to limit bypass flow through the RCC guide thimbles in fuel assemblies which do not contain either control rods, source rods, or BA rods.

The thimble plug assemblies consist of a flat baseplate with short rods suspended from the bottom surface and a spring pack assembly, as shown in figure 4.2-15. The 24 short rods, called thimble plugs, project into the upper ends of the guide thimbles to reduce the bypass flow. Each thimble plug is permanently attached to the baseplate by a nut which is lock-welded or crimped to the threaded end of the plug. Similar short rods are also used on the source assemblies and BA assemblies to plug the ends of all vacant fuel assembly guide thimbles.

When in the core, the thimble plug assemblies interface with both the upper core plate and with the fuel assembly top nozzles by resting on the adapter plate. The spring pack is compressed by the upper core plate when the upper internals assembly is lowered into place.

All components in the thimble plug assembly, except for the springs, are constructed from type 304 stainless steel. The springs are Inconel-718.

4.2.2.3.5 Stainless-Steel Rod Insert Assembly The SSRIA is to provide power suppression and power shaping ability to restore design margin to meet the fuel rod internal pressure criterion.

The stainless-steel rods are mounted to a conventional insert holddown assembly, 24 rods per assembly. The rods are either solid design or solid for the top 3-foot section and stainless-steel tube for the lower 9-foot section. The rods are the same length and diameter as RCCA rodlets.

Only one type of stainless-steel rod is used on an individual insert assembly. The SSRIA is shown in figure 4.2-16.

All components in the SSRIA, except for the springs, are constructed from type 304 and/or 308 stainless steel. The springs are Inconel-718.

4.2.3 DESIGN EVALUATION The fuel assemblies, fuel rods, and incore control components are designed to satisfy the performance and safety criteria of section 4.2, the mechanical design bases of subsection 4.2.1, and other interfacing nuclear and thermal and hydraulic design bases specified in sections 4.3 and 4.4.

4.2-20 REV 24 10/22

VEGP-FSAR-4 Effects of Conditions 2, 3, 4, or anticipated transients without trip on fuel integrity are presented in chapter 15 or supporting topical reports.

The initial step in fuel rod design evaluation for a region of fuel is to determine the limiting rod(s). Limiting rods are defined as those rods whose predicted performance provides the minimum margin to each of the design criteria. For a number of design criteria, the limiting rod is the lead burnup rod of a fuel region. In other instances, it may be the maximum power or the minimum burnup rod. For the most part, no single rod is limiting with respect to all design criteria.

After identifying the limiting rod(s), a worst-case performance analysis is performed, which considers the effects of rod operating history, model uncertainties, and dimensional variation.

To verify adherence to the design criteria, the evaluation considers the effects of postulated transient power changes during operation consistent with Conditions 1 and 2. These transient power increases can affect both rod average and local power levels. Parameters considered include rod internal pressure, fuel temperature, clad stress, and clad strain. In fuel rod design analyses, these performance parameters provide the basis for comparison between expected fuel rod behavior and the corresponding design criteria limits.

Fuel rod and assembly models used for the performance evaluations are documented and maintained under an appropriate control system. Materials properties used in the design evaluations are given in reference 4.

4.2.3.1 Cladding A. Vibration and Wear Fuel rod vibrations are flow induced. The effect of vibration on the fuel assembly and individual fuel rods is minimal. The cyclic stress range associated with deflections of such small magnitude is insignificant and has no effect on the structural integrity of the fuel rod.

The reaction force on the grid supports due to rod vibration motions is also small and is much less than the spring preload. Firm fuel clad spring contact is maintained. No significant wear of the clad or grip supports is expected during the life of the fuel assembly, based on out-of-pile flow tests, performance of similarly designed fuel in operating reactors, and design analyses.

Clad fretting and fuel vibration has been experimentally investigated, as shown in reference 11.

B. Fuel Rod Internal Pressure and Cladding Stresses A burnup-dependent fission gas release model(23, 27) is used in determining the internal gas pressures as a function of irradiation time. The plenum height of the fuel rod has been designed to ensure that the maximum internal pressure of the fuel rod will not exceed the value which would cause:

• The fuel/clad diametral gap to increase during steady-state operation.

• Extensive DNB propagation to occur.

The clad stresses at a constant local fuel rod power are low. Compressive stresses are created by the pressure differential between the coolant pressure and the rod internal gas pressure. Stresses due to the temperature gradient are not included in the volume average effective stress because thermal stresses 4.2-21 REV 24 10/22

VEGP-FSAR-4 are, in general, negative at the clad inside diameter and positive at the clad outside diameter, and their contribution to the clad volume average stress is small. Furthermore, the thermal stress decreases with time during steady-state operation due to stress relaxation. The stress due to pressure differential is highest in the minimum power rod at beginning of life (BOL) due to low internal gas pressure, and the thermal stress is highest in the maximum power rod due to steep temperature gradient.

The volume average effective stress at BOL is substantially below the unirradiated clad strength (approximately 55,500 psi) at a typical clad mean operating temperature of 700°F.

Tensile stresses could be created once the clad has come in contact with the pellet. These stresses would be induced by the fuel pellet swelling during irradiation.

Swelling of the fuel pellet can result in small clad stains (less than 1 percent) for expected discharge burnups, but the associated clad stresses are very low because of clad creep (thermal- and irradiation-induced creep). The 1-percent strain criterion is extremely conservative for fuel-swelling driven clad strain because the strain rate associated with solid fission products swelling is very slow. A detailed discussion of fuel rod performance is given in paragraph 4.2.3.3.

C. Materials and Chemical Evaluation Zircaloy-4/ZIRLO/Optimized ZIRLO clad have a high corrosion resistance to the coolant, fuel, and fission products. As shown in reference 3, there is considerable PWR operating experience on the capability of Zircaloy as a clad material. ZIRLO clad fuel rods were introduced in a Westinghouse core in 1991.

Optimized ZIRLO cladding material has been approved for use in Vogtle 1 and 2 by the NRC, reference 29. Controls on fuel fabrication specify maximum moisture levels to preclude clad hydriding.

Metallographic examination of irradiated commercial fuel rods has shown occurrences of fuel/clad chemical interaction. Reaction layers of less than 1 mil in thickness have been observed between fuel and clad at limited points around the circumference. Metallographic data indicates that this interface layer remains very thin even at high burnup. Thus, there is no indication of propagation of the layer and eventual clad penetration.

Stress corrosion cracking is another postulated phenomenon related to fuel/clad chemical interaction. Out-of-pile tests have shown that in the presence of high clad tensile stresses, large concentrations of iodine can chemically attack the Zircaloy/ZIRLO/Optimized ZIRLO tubing and lead to eventual clad cracking.

Extensive post-irradiation examination has produced no conclusive inpile evidence that this mechanism is operative in commercial fuel.

D. Rod Bowing Reference 12 presents the model used for evaluation of fuel rod bowing. To the present time, this model has been used for bow assessment in 14 x 14, 15 x 15, and 17 x 17 type cores.

E. Consequences of Power Coolant Mismatch This subject is discussed in chapter 15.

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VEGP-FSAR-4 F. Creep Collapse and Creep-Down This subject and the associated irradiation stability of cladding have been evaluated using the models described in references 7 and 26. It has been established that the design basis of no clad collapse during planned core life can be satisfied by limiting fuel densification and by having a sufficiently high initial internal rod pressure.

G. Fuel Assembly Design Evaluation with ZIRLO Clad Fuel Rods Evaluations were performed using NRC approved fuel rod performance codes to ensure that the fuel rod design bases and design criteria stated in section 4.2 were met for the fresh fuel assemblies containing ZIRLO clad fuel rods commencing in Unit 2, Cycle 6. The fuel rod design bases, criteria, and models which are particularly affected by the ZIRLO clad are described in reference 25.

The design and predicted nuclear characteristics for each fuel rod clad with ZIRLO are similar to those described in section 4.3. The evaluations have shown that the nuclear design bases for fuel rods clad with ZIRLO are satisfied. The change to ZIRLO will not affect the use of the standard nuclear design analytical models and methods noted in table 4.1-2 and described in section 4.3 to accurately evaluate the neutronic behavior of fuel rods clad with ZIRLO.

The thermal and hydraulic design bases for the fuel rods clad with ZIRLO are identical to those discussed in section 4.4. Since the use of ZIRLO clad fuel does not represent changes of any magnitude affecting the parameters which are major contributors in this area (i.e., DNB, core flow, and rod bow), the design bases described in section 4.4 remain valid.

The two non-LOCA accidents potentially affected by the use of ZIRLO clad material are the locked rotor/shaft break and RCCA ejection accidents. For the locked rotor/shaft break accident, it was determined that the ZIRLO cladding resulted in a very small increase in peak clad temperature, and the effect on the metal-to-water reaction rate was negligible when compared to Zircaloy-4.

However, sufficient margin exists in the VEGP plant safety analysis to accommodate the small PCT increase. For the RCCA ejection accident, the ZIRLO cladding results in a negligible effect in both the fraction of fuel melting at the hot spot and the fuel peak stored energy when compared to the results of Zircaloy-4. The conclusions in subsections 15.3.3, 15.3.4, and 15.4.8 for these two affected non-LOCA accidents remain valid.

The loss of coolant accident analyses for the VANTAGE 5 fuel in VEGP units were performed using the 1981 evaluation model with BASH (large break LOCA) and the NOTRUMP evaluation model (small break LOCA) described in paragraphs 15.6.5.3.1.1 and 15.6.5.3.1.2. Revisions to these evaluation models for use in the analyses of fuel with ZIRLO cladding have been identified and reported in reference 25. The revisions include the cladding specific heat, high-temperature creep (swelling), burst temperature, burst strain, and fuel assembly blockage. Calculations performed with the above revised evaluation models have shown that the effects of ZIRLO cladding on the large break and small break LOCA analysis results given in paragraphs 15.6.5.2 and 15.6.5.3 are minor.

H. Fuel Assembly Design Evaluation with Optimized ZIRLO Clad Fuel Rods 4.2-23 REV 24 10/22

VEGP-FSAR-4 Consistent with the preceding item G for ZIRLO clad fuel rods, NRC approved fuel rod performance codes and methods are utilized to ensure the design criteria presented in section 4.2 are met during normal operation, including anticipated operational occurrences on a reload basis, reference 28.

4.2.3.2 Fuel Materials Considerations Sintered, high-density uranium dioxide fuel reacts only slightly with the clad at core operating temperatures and pressures. In the event of clad defects, the high resistance of uranium dioxide to attack by water protects against fuel deterioration, although limited fuel erosion can occur. As has been shown by operating experience and extensive experimental work, the thermal design parameters conservatively account for changes in the thermal performance of the fuel elements due to pellet fracture which may occur during power operation. The consequences of defects in the clad are greatly reduced by the ability of uranium dioxide to retain fission products, including those which are gaseous or highly volatile. Observations from several operating Westinghouse PWRs(10) have shown that fuel pellets can densify under irradiation to a density higher than the manufactured values. Fuel densification and subsequent settling of the fuel pellets can result in local and distributed gaps in the fuel rods. Fuel densification has been minimized by improvements in the fuel manufacturing process and by specifying a nominal 95-percent initial fuel density.

The evaluation of fuel densification effects and the treatment of fuel swelling and fission gas release are described in references 23 and 27.

The effects of waterlogging on fuel behavior are discussed in paragraph 4.2.3.3.

4.2.3.3 Fuel Rod Performance In the calculation of the steady-state performance of a nuclear fuel rod, the following interacting factors must be considered:

A. Clad creep and elastic deflection.

B. Pellet density changes, thermal expansion, gas release, and thermal properties as a function of temperature and fuel burnup.

C. Internal pressure as a function of fission gas release, rod geometry, and temperature distribution.

These effects are evaluated using fuel rod design models (ref. 23 and 27) which include appropriate models for time dependent fuel densification. With these interacting factors considered, the model determines the fuel rod performance characteristics for a given rod geometry, power history, and axial power shape.

In particular, internal gas pressure, fuel and clad temperatures, and clad deflections are calculated. The fuel rod is divided into several axial sections and radially into a number of annular zones. Fuel density changes are calculated separately for each segment. The effects are integrated to obtain the internal rod pressure.

The initial rod internal pressure is selected to delay fuel/clad mechanical interaction and to avoid the potential for flattened rod formation. It is limited, however, by the design criteria for the rod internal pressure (paragraph 4.2.1.3).

4.2-24 REV 24 10/22

VEGP-FSAR-4 The gap conductance between the pellet surface and the clad inner diameter is calculated as a function of the composition, temperature and pressure of the gas mixture, and the gap size of contact pressure between clad and pellet. After computing the fuel temperature for each pellet annular zone, the fractional fission gas release is assessed using an empirical model derived from experimental data. (Ref. 23 and 27) The total amount of gas released is based on the average fractional release within each axial and radial zone and the gas generation rate, which in turn is a function of burnup. Finally, the gas released is summed over all zones, and the pressure is calculated.

The model shows close agreement in fit for a variety of published and proprietary data on fission gas release, fuel temperatures, and clad deflections (ref. 23, 25, and 27). These data include variations in power, time, fuel density, and geometry.

4.2.3.3.1 Fuel/Cladding Mechanical Interaction One factor in fuel element duty is potential mechanical interaction of fuel and clad. This fuel/clad interaction produces cyclic stresses and strains in the clad, and these in turn reduce clad life. The reduction of fuel/clad interaction is therefore a goal of design. The technology for using prepressurized fuel rods in Westinghouse PWRs has been developed to further this objective.

The gap between the fuel and clad is initially sufficient to prevent hard contact between the two.

However, during power operation a gradual compressive creep of the clad onto the fuel pellet occurs due to the external pressure exerted on the rod by the coolant. Clad compressive creep eventually results in the fuel/clad contact. Once fuel/clad contact occurs, changes in power level result in changes in clad stresses and strains. By using prepressurized fuel rods to partially offset the effect of the coolant external pressure, the rate of clad creep toward the surface of the fuel is reduced. Fuel rod prepressurization delays the time at which fuel/clad contact occurs and hence significantly reduces the extent of cyclic stresses and strains experienced by the clad both before and after fuel/clad contact. These factors result in an increase in the fatigue life margin of the clad and lead to greater clad reliability. If gaps should form in the fuel stacks, clad flattening will be prevented by the rod prepressurization so that the flattening time will be greater than the fuel core life.

A two-dimensional (r,) finite element model has been established to investigate the effects of radial pellet cracks on stress concentrations in the clad. "Stress concentration" herein is defined as the difference between the maximum clad stress in the direction and the mean clad stress.

The first case has the fuel and clad in mechanical equilibrium; and, as a result, the stress in the clad is close to zero. In subsequent cases the pellet power is increased in steps and the resultant fuel thermal expansion imposes tensile stress in the clad. In addition to uniform clad stresses, stress concentrations develop in the clad adjacent to radial cracks in the pellet. These radial cracks have a tendency to open during a power increase, but the frictional forces between fuel and clad oppose the opening of these cracks and result in localized increases in clad stress. As the power is further increased, large tensile stresses exceed the ultimate tensile strength of UO2, and additional cracks in the fuel are created, limiting the magnitude of the stress concentration in the clad.

As part of the standard fuel rod design analysis, the maximum stress concentration evaluated from finite element calculations is added to the volume-averaged effective stress in the clad as determined from the stress/strain calculations. The resultant clad stress is then compared to the temperature-dependent cladding yield stress in order to ensure that the stress/strain criteria are satisfied.

4.2-25 REV 24 10/22

VEGP-FSAR-4 The transient evaluation method is described in the following paragraphs.

Pellet thermal expansion due to power increases is considered the only mechanism by which significant stresses and strains can be imposed on the clad.

Power increases in commercial reactors can result from fuel shuffling (e.g., region 3 positioned near the center core for cycle 2 operation after operating near the periphery during cycle 1),

reactor power escalation following extended reduced power operation, and full-length control rod movement. In the mechanical design model, lead rods are depleted using best- estimate power histories as determined by core physics calculations. During burnup, the amount of diametral gap closure is evaluated based upon the pellet expansion cracking model, clad creep model, and fuel swelling model. At various times during the depletion, the power is increased locally on the rod to the burnup-dependent attainable power density as determined by core physics calculations. The radial, tangential, and axial clad stresses resulting from the power increase are combined into a volume average effective clad stress.

The Von Mises criterion is used to determine whether the clad yield stress has been exceeded.

This criterion states that an isotropic material in multiaxial stress will begin to yield plastically when the effective stress exceeds the yield stress as determined by an axial tensile test. The yield stress correlation is that for irradiated cladding, since fuel/clad interaction occurs at high burnup. In applying this criterion, the effective stress is increased by an allowance which accounts for stress concentrations in the clad adjacent to radial cracks in the pellet, prior to the comparison with the yield stress. This allowance was evaluated using a two-dimensional (r,)

finite element model.

Slow transient power increases can result in large clad strains without exceeding the clad yield stress because of clad creep and stress relaxation. Therefore, in addition to the yield stress criterion, a criterion on allowable clad strain is necessary. Based upon high strain rate burst and tensile test data on irradiated tubing, 1-percent strain was determined to be conservative lower limit on irradiated clad ductility and thus was adopted as a design criterion.

A comprehensive review of the available strain fatigue models was conducted by Westinghouse as early as 1968. This included the Langer-O'Donnell model,(13) the Yao-Munse model, and the Manson-Halford model. Upon completion of this review, using the results of the Westinghouse experimental programs discussed below, it was concluded that the approach defined by Langer-O'Donnell would be retained and the empirical factors of their correlation modified in order to conservatively bound the results of the Westinghouse testing program.

The Langer-O'Donnell empirical correlation has the following form:

E ln

( 100 )

+ Se Sa =

4 Nf 100 - RA where:

Sa = 1/2 E t = pseudostress amplitude which causes failure in N cycles (lb/in.2).

t = total strain range (in./in.).

E = Young's Modulus (lb/in.2).

Nf = number of cycles to failure.

RA = reduction in area at fracture in a uniaxial tensile.

Se = endurance limit (lb/in.2).

Both RA and Se are empirical constants which depend on the type of material, the temperature, and irradiation.

4.2-26 REV 24 10/22

VEGP-FSAR-4 The Westinghouse testing program was subdivided into the following subprograms:

A. A rotating bend fatigue experiment on unirradiated Zircaloy-4 specimens at room temperature and at 725°F. Both hydrided and nonhydrided Zircaloy-4 cladding was tested.

B. A biaxial fatigue experiment in gas autoclave on unirradiated Zircaloy-4 cladding, both hydrided and nonhydrided.

C. A fatigue test program on irradiated cladding from the Carolina-Virginia Tube Reactor and Yankee Core V conducted at Battelle Memorial Institute.

The results of these test programs provided information on different cladding conditions, including the effect of irradiation, hydrogen level, and temperature.

The design equations followed the concept for the fatigue design criterion according to the ASME Code, Section III:

A. The calculated pseudostress amplitude (Sa) has to be multiplied by a factor of 2 in order to obtain the allowable number of cycles (Nf).

B. The allowable cycles for a given Sa is 5 percent of Nf, or a safety factor of 20 on cycles.

The lesser of the two allowable numbers of cycles is selected. The cumulative fatigue life fraction is then computed as:

k nk N

1 fk 1

where:

nk = number of diurnal cycles of mode k.

Nfk = number of allowable cycles.

It is recognized that a possible limitation to the satisfactory behavior of the fuel rods in a reactor which is subjected to daily load follow is the failure of the clad by low-cycle strain fatigue.

During their normal residence time in the reactor, the fuel rods may be subjected to approximately 1000 cycles with typical changes in power level from 50 to 100 percent of their steady-state values.

The assessment of the fatigue life of the fuel rod clad is subject to a considerable uncertainty due to the difficulty of evaluating the strain range with results from the cyclic interaction of the fuel pellets and clad. This difficulty arises, for example, from such highly unpredictable phenomena as pellet cracking, fragmentation, and relocation. Nevertheless, since early 1968, this particular phenomenon has been investigated analytically and experimentally. (Ref. 13)

Strain fatigue tests on irradiated and nonirradiated hydrided Zircaloy-4 claddings were performed, which permitted a definition of a conservative fatigue life limit and recommendation on a methodology to treat the strain fatigue evaluation of the Westinghouse reference fuel rod designs.

It is believed that the final proof of the adequacy of a given fuel rod design to meet the load follow requirements can only come from incore experiments performed on actual reactors.

Experience in load follow operation dates back to early 1970 with the load follow operation of the Saxton reactor. Successful load follow operation has been performed on reactor A 4.2-27 REV 24 10/22

VEGP-FSAR-4 (approximately 400 load follow cycles) and reactor B (approximately 500 load follow cycles). In both cases, there was no significant coolant activity increase that could be associated with the load follow mode of operation.

4.2.3.3.2 Irradiation Experience Westinghouse fuel operational experience is presented in reference 3. Additional test assembly and test rod experience is given in sections 8 and 23 of reference 10. Optimized ZIRLO operating experience is documented in reference 28.

4.2.3.3.3 Fuel and Cladding Temperature The methods used for evaluation of fuel rod temperatures are presented in paragraph 4.4.2.11.

4.2.3.3.4 Waterlogging Local cladding deformations typical for waterlogging (c) bursts have never been observed in commercial Westinghouse fuel. Experience has shown that the small number of rods which have acquired clad defects, regardless of primary mechanism, remain intact and do not progressively distort or restrict coolant flow. In fact, such small defects are normally observed through reductions in coolant activity to be progressively closed upon further operation due to the buildup of zirconium oxide and other substances. Secondary failures which have been observed in defected rods are attributed to hydrogen embrittlement of the cladding. Post-irradiation examinations point to the hydriding failure mechanism rather than a waterlogging mechanism; the secondary failures occur as axial cracks in the cladding and are similar regardless of the primary failure mechanism. Such cracks do not result in flow blockage or increase the effects of any postulated transients.

More information is provided in references 14 and 15.

4.2.3.3.5 Potentially Damaging Temperature Effects During Transients The fuel rod experiences many operational transients (intentional maneuvers) during its residence in the core. A number of thermal effects must be considered when analyzing the fuel rod performance.

The clad can be in contact with the fuel pellet at some time in the fuel lifetime. Clad/pellet interaction occurs if the fuel pellet temperature is increased after the clad is in contact with the pellet. Clad/pellet interaction is discussed in paragraph 4.2.3.3.1.

The potential effects of operation with waterlogged fuel are discussed in paragraph 4.2.3.3.4, which concluded that waterlogging is not a concern during operational transients.

Clad flattening, as shown in reference 7, has been observed in some operating power reactors.

Thermal expansion (axial) of the fuel rod stack against a flattened section of clad could cause c

Waterlogging damage of a previously defected fuel rod has occasionally been postulated as a mechanism for subsequent rupture of the cladding. Such damage has been postulated as a consequence of a power increase on a rod after water has entered such a rod through a clad defect of appropriate size. Rupture is postulated upon power increase if the rod internal pressure increase is excessive due to insufficient venting of water to the reactor coolant.

4.2-28 REV 24 10/22

VEGP-FSAR-4 failure of the clad. This is no longer a concern because clad flattening is precluded during the fuel residence in the core (paragraph 4.2.3.1).

Potential differential thermal expansion between the fuel rods and the guide thimbles during a transient is considered in the design. Excessive bowing of the fuel rods is precluded because the grid assemblies allow axial movement of the fuel rods relative to the grids. Specifically, thermal expansion of the fuel rods is considered in the grid design so that axial loads imposed on the fuel rods during a thermal transient will not result in excessively bowed fuel rods.

4.2.3.3.6 Fuel Element Burnout and Potential Energy Release As discussed in paragraph 4.4.2.2, the core is protected from DNB over the full range of possible operating conditions. In the extremely unlikely event that DNB should occur, the clad temperature will rise due to the steam blanketing at the rod surface and the consequent degradation in heat transfer. During this time there is a potential for chemical reaction between the cladding and the coolant. However, because of the relatively good film boiling heat transfer following DNB, the energy release resulting from this reaction is insignificant compared to the power produced by the fuel.

4.2.3.3.7 Coolant Flow Blockage Effects on Fuel Rods This evaluation is presented in paragraph 4.4.4.7.

4.2.3.4 Spacer Grids The coolant flow channels are established and maintained by the structure composed of grids and guide thimbles. The lateral spacing between fuel rods is provided and controlled by the support dimples of adjacent grid cells. Contact of the fuel rods on the dimples is maintained through the clamping force of the grid springs. Lateral motion of the fuel rods is opposed by the spring force and the internal moments generated between the spring and the support dimples.

Grid testing is discussed in reference 16.

4.2.3.5 Fuel Assembly 4.2.3.5.1 Stresses and Deflections The fuel assembly component stress levels are limited by the design. For example, stresses in the fuel rod due to thermal expansion and Zircaloy/ZIRLO/Optimized ZIRLO irradiation growth are limited by the relative motion of the rod as it slips over the grid spring and dimple surfaces.

Clearances between the fuel rod ends and nozzles are provided so that Zircaloy/ZIRLO/Optimized ZIRLO irradiation growth does not result in rod end interferences.

Stresses in the fuel assembly caused by tripping of the RCCA have little influence on fatigue because of the small number of events during the life of an assembly. Assembly components and prototype fuel assemblies made from production parts have been subjected to structural tests to verify that the design bases requirements are met.

The fuel assembly design loads for shipping have been established at 4 g axial and 6 g lateral.

Accelerometers are permanently placed in the shipping cask to monitor and detect fuel 4.2-29 REV 24 10/22

VEGP-FSAR-4 assembly accelerations that would exceed the criteria. Past history and experience have indicated that loads which exceed the allowable limits rarely occur. Exceeding the limits requires reinspection of the fuel assembly for damage. Tests on various fuel assembly components, such as the grid assembly, sleeves, inserts, and structure joints, have been performed to ensure that the shipping design limits do not result in impairment of fuel assembly function. Seismic analysis of the fuel assembly is presented in reference 16.

To demonstrate that the fuel assemblies will maintain a geometry that is capable of being cooled under the worst-case accident Condition 4 event, Westinghouse has performed the following analyses.

The fuel assembly response resulting from safe shutdown earthquake (SSE) condition was analyzed using the history numerical techniques. The vessel motion for this type of accident primarily causes lateral loads on the reactor core. Consequently, the finite element seismic model as described in references 16 and 17 was used to assess the fuel assembly deflections and impact forces. However, the input parameters were modified to appropriately represent the Westinghouse 17 x 17 8-ft Inconel grid fuel assembly to be used in the VEGP units.

The motions of the upper and lower core plates and the barrel at the upper core plate elevation, which are simultaneously applied to the simulated reactor core model as input motion, were obtained from the time-history analysis of the reactor vessel and internals. The fuel assembly response, namely the displacements and impact forces, was obtained with the reactor core model by using the motions resulting from a reactor pressure vessel inlet nozzle break which produced the limiting structural loads for the fuel assembly.

4.2.3.5.1.1 Grid Analyses. The maximum grid impact force obtained from seismic analyses is less than the allowable grid strength. With respect to the guidelines of Appendix A of SRP Section 4.2, Westinghouse has demonstrated that a simultaneous SSE and LOCA event is highly unlikely. The fatigue cycles, crack initiation, and crack growth due to normal operating and seismic events will not realistically lead to a pipe rupture. (Ref. 18)

Based on the deterministic fracture mechanics evaluation of small flaws in piping components, Westinghouse has demonstrated that the dynamic effects of a large pipe rupture in the primary coolant piping system for the VEGP units is reduced. An exemption from a portion of the requirements of General Design Criterion 4 of Appendix A to 10 CFR Part 50 has been granted to the VEGP units. (Ref. 19)

The result of a pipe leakage accident would be to impose insignificant asymmetric loadings on the reactor core system. The fuel assembly grid load due to pipe ruptures was not applied to the analysis results.

4.2.3.5.1.2 Nongrid Analyses. The stresses induced in the various fuel assembly nongrid components were assessed based on the most limiting seismic condition. The fuel assembly axial forces resulting from the normal spring hold-down load together with its own weight distribution are the primary source of the stresses in the thimble guide tube and fuel assembly nozzles. The fuel-rod-accident-induced stresses, which are generally very small, are caused by bending due to the fuel assembly deflections during the seismic accident. A summary of the seismic-induced stresses, which are expressed in terms of a percentage of allowable stress limits for the fuel assembly major components, is given in table 4.2-1. The component stresses, which include normal operating stresses, are substantially below the established allowable limits. Consequently, the structural designs of the fuel assembly components are acceptable under the postulated accident design conditions for the VEGP.

4.2-30 REV 24 10/22

VEGP-FSAR-4 4.2.3.5.2 Dimensional Stability A prototype fuel assembly has been subjected to column loads in excess of those expected in normal service and faulted conditions. (Ref. 16)

No interference with control rod insertion into thimble tubes will occur during a postulated seismic transient due to fuel rod swelling, thermal expansion, or bowing. In the early phase of the transient following the coolant break, the high axial loads, which could be generated by the difference in thermal expansion between fuel clad and thimbles, are relieved by slippage of the fuel rods through the grids. The relatively low drag force restraint on the fuel rods will induce only minor thermal bowing, which is insufficient to close the fuel rod-to-thimble tube gap.

4.2.3.6 Reactivity Control Assembly and Burnable Absorber Rods 4.2.3.6.1 Internal Pressure and Cladding Stresses During Normal, Transient, and Accident Conditions The designs of the BA and source rods provide a sufficient cold void volume to accommodate the internal pressure increase during operation. This is not a concern for the RCCA absorber rod because no gas is released by the absorber material (whether it be Ag-In-Cd or hafnium).

For the standard BA rod, the use of glass in tubular form provides a central void volume along the length of the rods (figure 4.2-12). For the WABA rod, there is also sufficient cold void volume to limit the internal pressure to a value, which satisfies the design criteria. For the source rods, a void volume is provided within the rod in order to limit the internal pressure increase until end of life (EOL) (figures 4.2-13 and 4.2-14).

The stress analysis of these rods assumes 100-percent gas release to the rod void volume in addition to the initial pressure within the rod.

During normal transient and accident conditions, the void volume limits the internal pressures to values which satisfy the criteria in paragraph 4.2.1.6. These limits are established not only to ensure that peak stresses do not reach unacceptable values, but also to limit the amplitude of the oscillatory stress component in consideration of the fatigue characteristics of the materials.

Rod, guide thimble, and dashpot flow analyses indicate that the flow is sufficient to prevent coolant boiling within the guide thimble. Therefore, clad temperatures at which the clad material has adequate strength to resist coolant operating pressures and rod internal pressures are maintained.

4.2.3.6.2 Thermal Stability of the Absorber Material, Including Changes and Thermal Expansion The radial and axial temperature profiles within the source and absorber rods have been determined by considering gap conductance, thermal expansion, neutron or gamma heating of the contained material as well as gamma heating of the clad.

The maximum temperatures of the Ag-In-Cd or Hf rod absorber material were calculated and found to be significantly less than the respective material melting point and to occur axially at only the highest flux region. The thermal expansion properties of the absorber materials and the phase changes are discussed in reference 4 for Ag-In-Cd and reference 9 for Hf.

4.2-31 REV 24 10/22

VEGP-FSAR-4 The maximum temperature of the borosilicate glass in the standard burnable absorber was calculated to be about 1300°F, which occurs following the initial rise to power. As the operating cycle proceeds, the glass and pellet temperature decreases for the following reasons:

A. Reduction in power generation due to boron-10 depletion.

B. Better gap conductance as the helium produced diffuses to the gap.

C. External gap reduction due to borosilicate glass creep.

The maximum temperature of the Al203-B4C burnable absorber pellet is calculated to be less than 1200°F which takes place following the initial rise to power. As the operating cycle proceeds the burnable absorber pellet temperature decreases for the following reasons: (1) a reduction in heat generation due to boron depletion, and (2) better gap conduction as the helium produced diffuses into the gap.

Sufficient diametral and end clearances have been provided in the control rods, burnable absorber, and source rods to accommodate the relative thermal expansions between the enclosed material and the surrounding clad and end plug.

4.2.3.6.3 Irradiation Stability of the Absorber Material, Taking into Consideration Gas Release and Swelling The irradiation stability of the Ag-In-Cd and Hf absorber material is discussed in references 4 and 9, respectively. Irradiation produces no deleterious effects in the Ag-In-Cd absorber material.

As mentioned in paragraph 4.2.3.6.1, gas release is not a concern for the control rod material because no gas is released by the absorber material. Sufficient diametral and end clearances are provided to accommodate swelling of the absorber material.

Based on experience with borosilicate glass and on nuclear and thermal calculations, gross swelling or cracking of the glass in the standard burnable absorber tubing is not expected during operation. Some minor creep of the glass at the hot spot could occur but would only continue until boron-10 depletion and helium gap closure by creep had lowered the glass temperature to values which cause negligible creep. The wall thickness of the inner liner is sized to provide adequate support in the event of slumping and to collapse locally before rupture of the exterior cladding if unexpected large volume changes due to swelling or cracking should occur. The ends of the inner liner are open to allow helium, which diffuses out of the glass, to occupy the central void.

The Al203-B4C WABA pellets are designed such that gross swelling or crumbling of the pellets is not expected to occur during reactor operation. Some minor cracking of the pellets may occur, but this cracking should not affect the overall absorber and stack integrity.

4.2.3.6.4 Potential for Chemical Interaction, Including Possible Waterlogging Rupture The structural materials selected have good resistance to irradiation damage and are compatible with the reactor environment.

Corrosion of the materials exposed to the coolant is quite low, and proper control of chloride and oxygen in the coolant will prevent the occurrence of stress corrosion. The potential for the interference with RCC movement due to possible corrosion phenomena is very low.

4.2-32 REV 24 10/22

VEGP-FSAR-4 Waterlogging rupture is not a failure mechanism associated with VEGP control rods. However, a breach of the cladding for any postulated reason does not result in serious consequences.

Both the Ag-In-Cd and Hf absorber materials are relatively inert and would still remain remote from high coolant velocity regions. Rapid loss of material resulting in significant loss of reactivity control material would not occur. Bettis test results (ref. 4) concluded that additions of indium and cadmium to silver, in the amounts to form the Westinghouse absorber material composition, result in small corrosion rates.

The consequences of a clad breach in the standard BA cladding would be small. It is anticipated that upon clad breach, the borosilicate glass would be leached by the coolant water and that localized power peaking of a few percent would occur; no design criteria would be expected to be violated.

For the WABA, in the unlikely event that the Zircaloy clad is breached, the B4C in the affected rod(s) could be leached-out by the coolant water. If this occurred early, incore instruments could detect large peaking factor changes, and corrective action would be taken, if warranted.

A postulated clad breach after substantial irradiation would have no significant effect on peaking factors since the boron will have been burned-out. Breaching of the Zircaloy clad by internal hydriding is not expected due to moisture controls employed during fabrication. Rods of this design have also performed very well with no failures observed.

4.2.4 TESTING AND INSPECTION PLAN 4.2.4.1 Quality Assurance Program The quality assurance program plan of the Westinghouse Nuclear Fuel Division for the VEGP is summarized in reference 20.

The program provides for control over all activities affecting product quality, commencing with design and development and continuing through procurement, materials handling, fabrication, testing and inspection, storage, and transportation. The program also provides for the indoctrination and training of personnel and for the auditing of activities affecting product quality through a formal auditing program.

Westinghouse drawings and product, process, and material specifications identify the inspections to be performed.

4.2.4.2 Quality Control Quality control philosophy is generally based on the following inspections being performed to a 95-percent confidence that at least 95 percent of the product meets specification, unless otherwise noted.

A. Fuel System Components and Parts The characteristics inspected depend upon the component parts; the quality control program includes dimensional and visual examinations, check audits of test reports, material certification, and nondestructive examination, such as X-ray and ultrasonic.

All material used in this core is accepted and released by quality control.

4.2-33 REV 24 10/22

VEGP-FSAR-4 B. Pellets Inspection is performed for dimensional characteristics such as diameter, density, length, and squareness of ends. Additional visual inspections are performed for cracks, chips, and surface conditions according to approved standards.

Density is determined in terms of weight per unit length and is plotted on zone charts used in controlling the process. Chemical analyses are taken on a specified sample basis throughout pellet production.

C. Rod Inspection Fuel rod, control rodlet, BA rod, and source rod inspection consists of the following nondestructive examination techniques and methods, as applicable:

1. Each rod is leak tested using a calibrated mass spectrometer, with helium being the detectable gas.

2. Rod welds are inspected by ultrasonic test or X-ray in accordance with a qualified technique and Westinghouse specifications meeting the requirements of ASTM-E-142.

3. All rods are dimensionally inspected prior to final release. The requirements include such items as length, camber, and visual appearance.

4. All fuel rods are inspected by gamma scanning or other approved methods, as discussed in paragraph 4.2.4.5, to ensure proper plenum dimensions.

5. All fuel rods are inspected by gamma scanning, or other approved methods, as discussed in paragraph 4.2.4.5, to ensure that no significant gaps exist between pellets.

6. All fuel rods are active gamma scanned to verify enrichment control prior to acceptance for assembly loading.

7. Traceability of rods and associated rod components is established by quality control.

D. Assemblies Each fuel rod, control rod, BA rod, and source rod assembly is inspected for compliance with drawing and/or specification requirements. Other incore control component inspection and specification requirements are given in paragraph 4.2.4.4.

E. Other Inspections The following inspections are performed as part of the routine inspection operation:

1. Tool and gauge inspection and control, including standardization to primary and/or secondary working standards. Tool inspection is performed at prescribed intervals on all serialized tools. Complete records are kept of calibration and conditions of tools.

4.2-34 REV 24 10/22

VEGP-FSAR-4

2. Audits are performed of inspection activities and records to ensure that prescribed methods are followed and that records are correct and properly maintained.

3. Surveillance inspection, where appropriate, and audits of outside contractors are performed to ensure conformance with specified requirements.

F. Process Control To prevent the possibility of mixing enrichments during fuel manufacture and assembly, strict enrichment segregation and other process controls are exercised.

The UO2 powder is kept in sealed containers. The contents are fully identified. A Westinghouse identification tag completely describing the contents is affixed to the containers before transfer to powder storage. Isotopic content is confirmed by analysis.

Powder withdrawal from storage can be made by only one authorized group, which directs the powder to the correct pellet production line. All pellet production lines are physically separated from each other, and pellets of only a single nominal enrichment are present in any one section of a given production line at any given time. Enrichment barriers prevent mixing of product on a production line.

Finished pellets are placed on trays and transferred to segregated storage racks within the confines of the pelleting area. Samples from each pellet lot are tested for isotopic content and impurity levels prior to acceptance by quality control.

Physical barriers prevent mixing of pellets of different nominal densities and enrichments in this storage area. Unused powder and substandard pellets are returned to storage in the appropriate item control area.

Pellets are loaded into fuel cladding tubes on isolated production lines. Each production line contains only rods of one fuel type at any one time. The cladding tubes are identified by serialized traceability codes. The code uniquely identifies each rod and is maintained with the rod throughout production.

After inspection, the rods are loaded into a magazine for subsequent loading into a fuel assembly. At the time the magazines are loaded, the fuel rod identification numbers are entered into a computer system which then identifies the location of each rod within a given fuel assembly. After the loading of the fuel assembly, the top nozzle, which is inscribed with a permanent identification number, is attached to the assembly.

Similar traceability is provided for BA, source rods, and control rodlets, as required.

4.2.4.3 Online Fuel Failure Monitoring The function of the CVCS letdown monitor is to monitor the CVCS letdown liquid process and to provide indication of abnormal activity levels in the RCS. This monitor can act as a means of failed fuel warning because failed fuel would be a cause of an increase in activity. However, confirmation of the cause of any abnormal activity levels will be made by laboratory analysis of primary coolant. For a discussion of the CVCS letdown monitor, refer to information provided 4.2-35 REV 24 10/22

VEGP-FSAR-4 on liquid process and effluent monitors presented in section 11.5, paragraph 11.5.2.3, and table 11.5.2-1.

4.2.4.4 Incore Control Component Testing and Inspection Tests and inspections are performed on each reactivity control component to verify the mechanical characteristics. In the case of the full-length RCCA, prototype testing has been conducted; and both manufacturing test/inspections and functional testing at the plant site are performed.

During the component manufacturing phase, the following requirements apply to the reactivity control components to ensure the proper functioning during reactor operation:

A. All materials are procured to specifications to attain the desired standard of quality.

B. All Westinghouse spider assemblies (a spider from each braze lot for Hf RCCAs) are proof tested by applying a 5000-lb load to the spider body, so that approximately 310 lb are applied to each vane. This proof load provides a bending moment at the spider body approximately equivalent to 1.4 times the load caused by the acceleration imposed by the CRDM.(a)

C. All rods are checked for integrity by the methods described in paragraph 4.2.4.2C.

D. To ensure proper fit with the fuel assembly, the rod cluster control, BA, and source assemblies are installed in the fuel assembly and checked for binding in the dry condition.

The RCCAs are functionally tested, following core loading but prior to criticality, to demonstrate reliable operation of the assemblies. Each assembly is operated (and tripped) one time at full-flow/hot conditions.

In order to demonstrate continuous free movement of the RCCAs and to ensure acceptable core power distributions during operations, partial movement checks are performed on every RCCA, as required by the Technical Specifications. In addition, periodic drop tests of the full-length RCCAs are performed at each refueling shutdown to demonstrate continued ability to meet trip time requirements.

If an RCCA cannot be moved by its mechanism, adjustments in the boron concentration of the coolant ensure that adequate shutdown margin would be achieved following a trip. Thus, inability to move one RCCA can be tolerated. More than one inoperable RCCA could be tolerated but would impose additional demands on the plant operator. Therefore, the number of inoperable RCCAs has been limited to one.

4.2.4.5 Tests and Inspections by Others If any tests and inspections are to be performed on behalf of Westinghouse, Westinghouse will review and approve the quality control procedures, inspection plans, etc., to be utilized to ensure that they are equivalent to the description provided in paragraphs 4.2.4.1 through 4.2.4.4 and are performed properly to meet all Westinghouse requirements.

a. Requirement B is not applicable to the Framatome RCCA spider assemblies.

4.2-36 REV 24 10/22

VEGP-FSAR-4 4.2.4.6 Inservice Surveillance As detailed in reference 3, significant 17 x 17, 12-ft Inconel grid fuel assembly operating experience has been obtained. VEGP will establish a surveillance program for inspection of post-irradiated fuel assemblies. This surveillance program will establish the schedule, guidelines, and inspection criteria for conducting visual inspection of post-irradiated fuel assemblies and/or insert components. As a minimum, the surveillance program will include a quantitative visual examination of some discharged fuel assemblies from each refueling. This program includes criteria for additional inspection requirements for post-irradiated fuel assemblies if unusual characteristics are noticed in the visual inspection or plant instrumentation and subsequent laboratory analysis indicates gross failed fuel. The VEGP post irradiated fuel surveillance program will address disposition of fuel assemblies and/or insert components receiving an unsatisfactory visual inspection. Those post-irradiated fuel assemblies receiving an unsatisfactory visual inspection shall not be reinserted into the core until a more detailed inspection and/or evaluation can be performed. Normally the fuel assemblies will be taken to the spent fuel inspection station.

The engineering group is responsible for initiating the necessary corrective actions.

4.2.4.7 Onsite Inspection Written procedures are used by the station staff for the post- shipment inspection of all new fuel and associated components, such as control rods, plugs, and inserts. Fuel handling procedures specify the sequence in which handling and inspection take place.

Loaded fuel containers, when received onsite, are externally inspected to ensure that labels and markings are intact and seals are unbroken. After the containers are opened, the shock indicators attached to the suspended internals are inspected to determine whether movement during transit exceeded design limitations.

Following removal of the fuel assembly from the container in accordance with detailed procedures, the fuel assembly plastic wrapper is examined for evidence of damage. The polyethylene wrapper is then removed, and a visual inspection of the entire bundle is performed.

Control rod, source, and BA assemblies usually are shipped in fuel assemblies and are inspected prior to removal of the fuel assembly from the container. The control rod assembly is withdrawn a few inches from the fuel assembly to ensure free and unrestricted movement, and the exposed section is visually inspected for mechanical integrity, replaced in the fuel assembly, and stored with the fuel assembly. Control rod, source, or BA assemblies may be stored separately or within fuel assemblies in the new fuel storage area.

Westinghouse designed an experimental program to obtain in-reactor creep and growth behavior data for various cladding and structural zirconium-based alloys to fluence levels that support long-term target values for fuel burnup. The program involves the irradiation of non-fueled test samples in symmetric core locations in VEGP Unit 2 commencing with Cycle 10.

For the duration of the program, the test samples will be inserted into as many as four selected host fuel assemblies in the reload core. The test samples that will be irradiated for more than one cycle will be moved to the new once-burned fuel assemblies after each cycle of operation.

Neutronically equivalent dummy samples will be used as needed to preserve core symmetry.

The reload evaluation for each cycle will address the cycle-specific test samples.

4.2-37 REV 24 10/22

VEGP-FSAR-4 The discharged test samples will be removed from the fuel assemblies and prepared for shipment. The test samples will be sheared and loaded into casks for shipment to a hot cell test facility. The remaining test parts and the dummy samples can be stored in the spent fuel pool.

4.2.4.8 References

1. Letter, T. M. Anderson (Westinghouse) to J. R. Miller (NRC), dated August 15, 1980.

2. Letter, T. M. Anderson (Westinghouse) to J. R. Miller (NRC), dated April 21, 1981.

3. Slagle, W. H., "Operational Experience with Westinghouse Cores," WCAP-8183 (revised annually).

4. Beaumont, M. D., et al., "Properties of Fuel and Core Component Materials," WCAP-9179, Revision 1 (Proprietary), and WCAP-9224 (Nonproprietary), July 1978.

5. Slagle, W. H., et al., 17x17 Next Generation Fuel (17x17 NGF) Reference Core Report, May 2011.

6. Deleted.

7. George, R. A., Lee, Y. C., and Eng, G. H., "Revised Clad Flattening Model," WCAP-8377 (Proprietary) and WCAP-8381 (Nonproprietary), July 1974.

8. Risher, D., et al., "Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8963 (Proprietary), November 1976, and WCAP-8964 (Nonproprietary),

August 1977.

9. "Hafnium," Appendix A to Reference 2, October 1980.

10. Eggleston, F. T., "Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries - Winter 1976 - Summer 1978,"

WCAP-8768, Revision 2, October 1977.

11. Demario, E. E., "Hydraulic Flow Test of the 17 x 17 Fuel Assembly," WCAP-8278 (Proprietary) and WCAP-8279 (Nonproprietary), February 1974.

12. Skaritka, J., ed, "Fuel Rod Bow Evaluation," WCAP-8691, Rev 1, July 1979.

13. O'Donnell, W. J. and Langer, B. F., "Fatigue Design Basis for Zircaloy Components,"

Nuclear Science and Engineering 20, pp 1-12, 1964.

14. Stephan, L. A., "The Effects of Cladding Material and Heat Treatment on the Response of Waterlogged UO2 Fuel Rods to Power Bursts," IN-ITR-111, January 1970.

15. Western New York Nuclear Research Center Correspondence With the U.S. Atomic Energy Commission on February 11 and August 27, 1971, Docket No. 50-57.

16. Gesinski, L. and Chiang, D., "Safety Analysis of the 17 x 17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident," WCAP-8236 (Proprietary) and WCAP-8288 (Nonproprietary), December 1973.

17. Beaumont, M. D., et al., "Verification, Testing, and Analysis of the 17 x 17 Optimized Fuel Assembly," WCAP-9401-P-A (Proprietary) and WCAP-9402-A (Nonproprietary),

August 1981.

18. Witt, F. J., Bamford, W. H., and Esselman, T. C., "Integrity of the Primary Piping Systems of Westinghouse Nuclear Power Plants During Postulated Seismic Events,"

WCAP-9283, March 1978.

4.2-38 REV 24 10/22

VEGP-FSAR-4

19. "Georgia Power Company, et al., (Vogtle Electric Generating Plant, Units 1 and 2),

Exemption," Federal Register, Volume 50, No. 27, February 8, 1985.

20. "Nuclear Fuel Division Quality Assurance Program Plan," WCAP-7800, Revision 4-A, March 1975.

21. Davidson, S. L. ed et al., "Reference Core Report VANTAGE 5 Fuel Assembly," WCAP-10444-P-A, September 1985.

22. Skaritka, J., et al., "Westinghouse Wet Annular Burnable Absorber Evaluation Report,"

WCAP-10021-P-A, Revision 1 (Proprietary), October 1983.

23. Weiner, R. P., et al., "Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations," WCAP-10851-P-A (Proprietary) and WCAP-11873-A (Nonproprietary), August 1988.

24. Davidson, S. L. ed et al., "Extended Burnup Evaluation of Westinghouse Fuel, "WCAP-10125-P-A and WCAP-10126-N1-P (Nonproprietary), December 1985.

25. Davidson, S. L., et al., "VANTAGE + Fuel Assembly Reference Core Report," WCAP-12610-P-A, April 1995.

26. Kersting, P. J., et al., "Assessment of Clad Flattening and Densification Power Spike Factor Elimination in Westinghouse Nuclear Fuel," WCAP-13589-A, March 1995.

27. Foster, J. P., et al., "Westinghouse Improved Performance Analysis and Design Model (PAD 4.0)," WCAP-15063-P-A, Revision 1, with Errata, July 2000.

28. Shah, H., Optimized ZIRLOTM, WCAP-12610-P-A & CENPD-404-P-A Addendum 1-A, July 2006.

29. U.S. NRC, R. E. Martin letter to SNC, C. R. Pierce, Joseph M. Farley Nuclear Plant, Units 1 and 2, and Vogtle Electric Generating Plant, Units 1 and 2, Issuance of Amendments Regarding Use of Optimized ZIRLO (CAC Nos. MF7480, MF7481, MF7482, AND MF7483), August 04, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16179A386).

4.2-39 REV 24 10/22

VEGP-FSAR-4 TABLE 4.2-1 FUEL ASSEMBLY COMPONENT STRESSES (PERCENT OF ALLOWABLE)

Uniform Stresses Combined Stresses Component (Membrane/Direct) (Membrane and Bending)

Thimble 54.9 37.8 Fuel rod(a) 23.7 15.8 Top nozzle plate (b) 6.6 Bottom nozzle plate (b) 12.7 Bottom nozzle leg 2.0 9.1

a. Includes primary operating stresses.

b. Indicates a negligible value.

REV 14 10/07

REV 14 10/07 FUEL ASSEMBLY CROSS-SECTION 17 X 17 LOPAR FIGURE 4.2-1 (SHEET 1 OF 2)

REV 14 10/07 FUEL ASSEMBLY CROSS-SECTION 17 X 17 VANTAGE 5 FIGURE 4.2-1 (SHEET 2 OF 2)

REV 24 10/22 FUEL ASSEMBLY OUTLINE, 17 X 17 LOPAR, REGIONS 5A, 5B, 5C 5D, 5L, 5M, 5N FIGURE 4.2-2 (SHEET 1 OF 7)

REV 24 10/22 FUEL ASSEMBLY OUTLINE, 17 X 17 LOPAR, REGIONS 5E, 5P FIGURE 4.2-2 (SHEET 2 OF 7)

REV 24 10/22 FUEL ASSEMBLY OUTLINE 17 X 17 VANTAGE 5 FIGURE 4.2-2 (SHEET 3 OF 7)

REV 21 4/18 FUEL ASSEMBLY OUTLINE 17 X 17 OFA P+ ZIRC-4/ZIRLO/

ZIRLO/OPTIMIZED ZIRLOTM FUEL RODS FIGURE 4.2-2 (SHEET 4 OF 7)

REV 24 10/22 FUEL ASSEMBLY OUTLINE 17 X 17 VANTAGE+ WITH LOW ROD INTERNAL PRESSURE FUEL RODS FIGURE 4.2-2 (SHEET 5 OF 7)

REV 24 10/22 FUEL ASSEMBLY OUTLINE VOGTLE 17 X 17 VANTAGE+ WITH LOW ROD ELECTRIC GENERATING PLANT INTERNAL PRESSURE FUEL RODS AND SDFBN UNIT 1 AND UNIT 2 FIGURE 4.2-2 (SHEET 6 OF 7)

REV 24 10/22 FUEL ASSEMBLY OUTLINE, 17 X 17 VOGTLE ELECTRIC GENERATING PLANT PRIME FUEL ASSEMBLY UNIT 1 AND UNIT 2 FIGURE 4.2-2 (SHEET 7 OF 7)

REV 14 10/07 FUEL ROD SCHEMATIC LOPAR, REGIONS 5A, 5B, 5C, 5D, 5L 5M, 5N, FIGURE 4.2-3 (SHEET 1 OF 4)

REV 14 10/07 FUEL ROD SCHEMATIC LOPAR, REGIONS 5E, 5P FIGURE 4.2-3 (SHEET 2 OF 4)

REV 14 10/07 FUEL ROD SCHEMATIC VANTAGE 5 FIGURE 4.2-3 (SHEET 3 OF 4)

REV 14 10/07 FUEL ROD SCHEMATIC LOW ROD INTERNAL PRESSURE DESIGN FIGURE 4.2-3 (SHEET 4 OF 4)

REV 14 10/07 PLAN VIEW LOPAR FIGURE 4.2-4 (SHEET 1 OF 2)

REV 14 10/07 PLAN VIEW VANTAGE 5 FIGURE 4.2-4 (SHEET 2 OF 2)

REV 14 10/07 ELEVATION VIEW, GRID TO THIMBLE ATTACHMENT LOPAR FIGURE 4.2-6 (SHEET 1 OF 2)

REV 14 10/07 GRID TO THIMBLE ATTACHMENT JOINTS VANTAGE 5 FIGURE 4.2-6 (SHEET 2 OF 2)

REV 14 10/07 GUIDE THIMBLE TO BOTTOM NOZZEL JOINT LOPAR FIGURE 4.2-7 (SHEET 1 OF 2)

REV 14 10/07 GUIDE THIMBLE TO BOTTOM NOZZEL JOINT VANTAGE 5 FIGURE 4.2-7 (SHEET 2 OF 2)

REV 14 10/07 RCC AND DRIVE ROD ASSEMBLY WITH INTERFACE COMPONENTS FIGURE 4.2-8

a A single coil spring applicable to Framatome RCCAs only b Overall assembly length of 161.0 inches applicable to Framatome RCCAs only c Spider boss diameter of 0.355 applicable to Framatome RCCAs only REV 23 3/21 FULL-LENGTH RCCA OUTLINE FIGURE 4.2-9

a 316 stainless steel low contaminant cladding material applicable to Framatome RCCA only b Overall absorber rod length of 153.831 applicable to Framatome RCCAs only REV 23 3/21 ABSORBER ROD FIGURE 4.2-10

REV 14 10/07 BURNABLE ABSORBER ASSEMBLY (STANDARD BOROSILICATE GLASS)

FIGURE 4.2-11 (SHEET 1 OF 2)

REV 14 10/07 WET ANNULAR BURNABLE ABSORBER (WABA) ASSEMBLY FIGURE 4.2-11 (SHEET 2 OF 2)

REV 14 10/07 BA ROD CROSS SECTION (STANDARD BOROSILICATE GLASS)

FIGURE 4.2-12 (SHEET 1 OF 2)

REV 14 10/07 BA ROD CROSS SECTION (WET ANNULAR BURNABLE ABSORBER)

FIGURE 4.2-12 (SHEET 2 OF 2)

REV 14 10/07 PRIMARY SOURCE ASSEMBLY FIGURE 4.2-13

REV 14 10/07 DOUBLE-ENCAPSULATED SECONDARY SOURCE ASSEMBLY (SIX ROD PATTERN)

FIGURE 4.2-14 (SHEET 1 OF 2)

REV 14 10/07 SINGLE-ENCAPSULATED SECONDARY SOURCE ASSEMBLY (FOUR ROD PATTERN)

FIGURE 4.2-14 (SHEET 2 OF 2)

REV 14 10/07 THIMBLE PLUG ASSEMBLY FIGURE 4.2-15 (SHEET 1 OF 2)

REV 14 10/07 STANDARDIZED THIMBLE PLUG ASSEMBLY FIGURE 4.2-15 (SHEET 2 OF 2)

REV 14 10/07 STAIN STEEL ROD INSERT ASSEMBLY FIGURE 4.2-16

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