NL-16-2280, Vogtle Electric Generating Plant, Units 1 & 2, Updated Final Safety Analysis Report, Figures 3D-137 Through 4.3-41
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
3E-1 REV 14 10/07 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:
00uu)u()c(322222=++ (1) in which u is the displacement vector (u 1 ,u 2 ,u 3)e it 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 3 j(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:
u i (1 ,2 ,0) = - 0 2 0 1 0 2 0 13j 0 2 2 0 1 siij 31jdd)0,,(),(G= (2) for 3 = 0. In equation 2, )0,,(G 022 0 1 01ijdenotes the ith displacement component at (1 ,2,0) generated by a unit harmonic load acting at )0,,(0 2 0 1. 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. B. The stress components j3 are assumed to have constant values )k(3j within each subregion S
- k. 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:
VEGP-FSAR-3
3E-2 REV 14 10/07 _~_~2_~1..
.u u u n = [][][][][][][][][]nn n nnn...............2122211211.........2 1 n A A)n(3j 2)2(3j 1)1(3j..
. (3) in which:
_~u 1 = _i3_i2_i1 u , u , u T are average displacements in subregion S
- i. A i = area if subregion S
- i. ij][= a 3x3 compliance submatrix relating the average displacements at subregion S i to the tractions at subregion S
- j. 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: ij][ = ),,(~~~PrrG ds ds o oSjSi (4) in which [G] is the 3x3 Green's function matrix relating the displacement at the observation point r = (1 , 2 , 0)T to a set of point loads at the source point r o~ ()~T 0 2 0 1P0,,* is the property vector associated with the underlying half space. By use of the average displacement matching approximation, the following symmetry exists even if A i A j: ][][ji T ij= (5) The property vector
~P for a horizontally layered stratum can be characterized as:
~P = (m, µi , i , i ,h i , i , i) (6) where: m = the number of layers.
µi = the shear modulus of the ith layer.
i = the shear wave velocity of the ith layer.
i = the mass density of the ith layer.
H i = the thickness of the ith layer.
VEGP-FSAR-3
3E-3 REV 14 10/07 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:
(7) where: r = 2 o22 2 o11~o~)()(rr+= = arg ()=o11 o22 1o~~tan r r b o = r {}µ=,,h,,,,mPiiiii~ , the normalized property vector in which:
µµ=µi i =i i =i i =i i h h The reference values of
µ, 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, f rr , fr , f rz , and f zz , which are the Green's 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:
=0on~ondk)kb(J),,k(F)b(I (8) for the concentrated point load applied at surface and displacement at the free surface observed at b o distance from the load point. The kernel F depends upon wave number k, frequency , and layer properties
~; whereas, the Bessel functions J n depend only upon kb
- o. 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:
[][]),,(1),,(~0~~~P b G rPrrG o=µ VEGP-FSAR-3
3E-4 REV 14 10/07 dk)kb(J]),0,k(F),,k(F[)0(I)b(Ion k 0~non+= (9) in which the upper limit of integration, k , is defined by the convergence of the dynamic integrands to the static integrands, and I n (0) represents the static (= 0) integrals. Since the radial dependence b o appears only in the Bessel functions J n , it is expedient to calculate the integrals; begin at b o = 0 and end at b o = max r (rmax is the maximum length of the foundation) in equally spaced intervals. This precalculated Green's 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
~i u , evaluated at the center of subregion S i are given by:
i 231 i 2 i 1 li)0,,(u= i 232 i 2 i 1i2)0,,(u+= (10) i 2 2 i 213 i 2 i 1i3)0,,(u+= 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 k)k(j A 3 can be expressed in terms of
~i u , by inverting the matrix []. By substituting for ~iu 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 (P 1 ,P 2 ,P 3) and the total harmonic moment with components (M 1 ,M 2 ,M 3) acting on the contact area can be expressed in terms of traction components by means of the following relationships:
P 1 = Aj n1j)j(3i= (i = 1,2, 3)
M 1 = )j(33 j 2 n1j= (11) M 2 = )j(33 j 1 n1j= M 3 = []Aj Aj)j(13 j 2)j(23 j 1 n1j= Substitution for contact tractions in terms of i and i into equation 11 leads to the desired force-displacement relationship for the rigid foundation:
=
][K M P (12)
VEGP-FSAR-3
3E-5 REV 14 10/07 where [K] is the complex frequency dependent impedanc e 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.
REV 14 10/07 MODEL AND COORDINATE SYSTEM FIGURE 3E-1
REV 14 10/07 POLAR COORDINATE SYSTEM FIGURE 3E-2
VEGP-FSAR-3
3F-1 REV 15 4/09 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 sy stem 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.
VEGP-FSAR-3
3F-2 REV 15 4/09 Loss of offsite power (LOSP) is not assumed, unless a trip of the turbine-generator system or
the reactor protection system is a direct cons equence 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 m oderate-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.
VEGP-FSAR-3
3F-3 REV 15 4/09 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:
- Duriron lines.
- Copper lines.
- ASTM A-120 galvanized lines.
- Air service lines.
- Process tubing 3/8 in.
- Instrument tubing.
- 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 comm unication 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
contents of the flooded area are then reviewed to ensure that no essential equipment is VEGP-FSAR-3
3F-4 REV 15 4/09 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 fr om 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:
- Structural enclosures.
- Structural barriers.
- Restraints.
- Seismically designed components.
- Hardening.
- Orientation. 3F.4 HAZARDS EVALUATIONS As stated previously, table 3F-1 provides a haz ards evaluation of level C and safety-related pump rooms on levels B and D of the auxiliary build ing. 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 document ed 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 turbine-
driven pump supplies twice the capacity of a motor-driven pump and is sufficient to remove VEGP-FSAR-3
3F-5 REV 15 4/09 remove decay heat and, additionally, to cool down the reactor at a rate of up to 100
°F/h not to exceed 100
°F 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.
Similarly, flooding caused by piping failures will not cause a loss of function of the AFW system because separation is provided between all three AFW pump rooms.
Watertight seals between these rooms prevent propagation of flooding and ensure that adequate capacity of the AFW system is maintained.
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 t he 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 st eam 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 sy stem, 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 ft 2 with steam generator tube bundle being uncovered have been considered for equipment qualification as discussed in paragraph 3.11.B.1.1.
VEGP-FSAR-3
3F-6 REV 15 4/09 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 ft 2 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 auxilia ry 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/M FIV 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 ft 2 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.
VEGP-FSAR-3
3F-7 REV 15 4/09 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 block wall will VEGP-FSAR-3
3F-8 REV 15 4/09 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 10 6 ft 3 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."
VEGP-FSAR-3F TABLE 3F-1 (SHEET 1 OF 84)
REV 15 4/09 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 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. 1. The high-energy line breaks have been evaluated and 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 2 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 3 OF 84)
REV 15 4/09 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 switch TISH-12210 Y
- 16. Train A raceway Conduit and cabletrays Y Y
VEGP-FSAR-3F TABLE 3F-1 (SHEET 4 OF 84)
REV 15 4/09 Room Number R-B18 Title Valve Gallery Remarks Flooding Analysis 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.
- 1. Flooding from sources located in this room will not impair the safe shutdown capability of the SRE. 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 5 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 6 OF 84)
REV 15 4/09 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 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.
- 1. The high-energy line breaks have been evaluated and will not preclude mitigation of the event or safe shutdown of the plant. 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 7 OF 84)
REV 15 4/09 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 valve HV-8920 Y 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 8 OF 84)
REV 15 4/09 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 switch. TISH-12211 Y
VEGP-FSAR-3F TABLE 3F-1 (SHEET 9 OF 84)
REV 15 4/09
(Recycle Evaporator Abandoned in Place)
VEGP-FSAR-3F TABLE 3F-1 (SHEET 10 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 11 OF 84)
REV 15 4/09
(Waste Evaporator Abandoned in Place)
VEGP-FSAR-3F TABLE 3F-1 (SHEET 12 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 13 OF 84)
REV 15 4/09 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 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. 1. Flooding from sources located in this room will not impair the safe shutdown capability of the SRE.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 14 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 15 OF 84)
REV 15 4/09 Room Number R-C83 Title Corridor Remarks Flooding Analysis 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
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. 1. The SRE located in this room is not required for safe shutdown.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 16 OF 84)
REV 15 4/09 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-conditioning (HVAC) vent valve HV-12596 N Y 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 17 OF 84)
REV 15 4/09 Room Number R-C88 Title Electric al Chase Train A Remarks Flooding Analysis 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. 1. The safe shutdown capability of the SRE located in this room is not impaired by this flood level.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 18 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 19 OF 84)
REV 15 4/09 Room Number R-C89 Title Storage Room Remarks Flooding Analysis 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. 1. Flooding from sources located in this room will not impair the safe shutdown capability of the SRE.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 20 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 21 OF 84)
REV 15 4/09 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 22 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 23 OF 84)
REV 15 4/09 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 24 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 25 OF 84)
REV 15 4/09 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 injection pump valve HV-8804B Y Y
VEGP-FSAR-3F TABLE 3F-1 (SHEET 26 OF 84)
REV 15 4/09 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 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 remark 1. 1. Flooding from sources located in this room will not impair the safe shutdown capability of the SRE.
Seismic Design Analysis Only SRE is in the room; therefore, there are no seismically induced failures (Categor y 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 27 OF 84)
REV 15 4/09 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 TE-19722E Y
VEGP-FSAR-3F TABLE 3F-1 (SHEET 28 OF 84)
REV 15 4/09 Room Number R-C99 Title Nonradioactive Pipe Chase Remarks Flooding Analysis
- 1. The flooding from sources located in this room will not impair the safe shutdown 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 29 OF 84)
REV 15 4/09 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. pipingY 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 30 OF 84)
REV 15 4/09 Room Number R-C99 Title Nonradioactive Remarks Flooding Analysis 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. 1. Safety-related function is not impaired by flooding or water spray.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 31 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 32 OF 84)
REV 15 4/09 Room Number R-C104 Title Electrical Chase Remarks Flooding Analysis 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.
- 1. The flooding from sources located in this room will not impair the safe shutdown capability of the SRE. 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 33 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 34 OF 84)
REV 15 4/09 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 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.
- 1. Safety-related function is not impaired by flooding or water spray. 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 35 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 36 OF 84)
REV 15 4/09 Room Number R-C107 Title Vestibule Remarks Flooding Analysis 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.
- 1. Safety-related function is not impaired by flooding or water spray. 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 37 OF 84)
REV 15 4/09 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 TE-15216C Y Y 3. Train D raceway Conduits and cabletrays Y Y
VEGP-FSAR-3F TABLE 3F-1 (SHEET 38 OF 84)
REV 15 4/09 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 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 remark 1. 1. Safety-related function is not impaired by flooding or water spray.
- 2. There is no essential equipment for safe shutdown located in this room; therefore, the postulated hazards have no effect on the safe shutdown capability.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 39 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 40 OF 84)
REV 15 4/09 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 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.
- 1. Safety related function is not impaired by flooding or water spray. 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 41 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 42 OF 84)
REV 15 4/09 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 43 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 44 OF 84)
REV 15 4/09 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 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.
- 1. Flooding from sources located in this room room will not impair the safe shutdown capability of the SRE. 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 45 OF 84)
REV 15 4/09 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 VEGP-FSAR-3F TABLE 3F-1 (SHEET 46 OF 84)
REV 15 4/09 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 47 OF 84)
REV 15 4/09 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 is olation 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 48 OF 84)
REV 15 4/09 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 49 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 50 OF 84)
REV 15 4/09 Room Number R-C116 Title Vestibule Remarks Flooding Analysis 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.
- 1. Safety-related equipment is not impaired by flooding or water spray. 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 51 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 52 OF 84)
REV 15 4/09 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 53 OF 84)
REV 15 4/09 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 and indicator T1SH-12215 Y
VEGP-FSAR-3F TABLE 3F-1 (SHEET 54 OF 84)
REV 15 4/09 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 55 OF 84)
REV 15 4/09 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 is olation valve HV-8111B Y 5. Charging pump discharge, trai n 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 56 OF 84)
REV 15 4/09 Room Number R-C120 Title Vestibule Remarks Flooding Analysis 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.F), 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. 1. The flooding from sources located in this room will not impair the safe shutdown capability of the SRE.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 57 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 58 OF 84)
REV 15 4/09 Room Number R-C131 Title Vestibule Remarks Flooding Analysis 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 remark 1. 1. Flooding from sources located in this room will not impair the safe shutdown, capability of the SRE.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 59 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 60 OF 84)
REV 15 4/09 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 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. 1. Flooding from sources located in this room will not impair the safe shutdown capability of the SRE.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 61 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 62 OF 84)
REV 15 4/09 Room Number R-C134 Title Pipe Penetration Room Train A Remarks Flooding Analysis
- 1. Flooding from sources located in this room will not impair the safe shutdown 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.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 63 OF 84)
REV 15 4/09 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 valve HV-9002A Y 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 64 OF 84)
REV 15 4/09 Room Number UC-C14 Title Pipe Chase Remarks Flooding Analysis 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 remark 1. 1. The flooding from sources located in this room will not impair the safe shutdown capability of the SRE.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 65 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 66 OF 84)
REV 15 4/09 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 67 OF 84)
REV 15 4/09 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 element TE-12206 Y Y 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 68 OF 84)
REV 15 4/09 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 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. X Other, see remarks. 1. The flooding from sources located in this room will not impair the safe shutdown capability of the SRE 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 69 OF 84)
REV 15 4/09 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 pump 1, valve HV 9017A Y 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 switch TISH-12207 Y Y 9. Containment spray train A 1-1206-6001
VEGP-FSAR-3F TABLE 3F-1 (SHEET 70 OF 84)
REV 15 4/09 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 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. 1. The only equipment in this room is associated with containment spray system, which is not essential in this case.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 71 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 72 OF 84)
REV 15 4/09 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 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. 1. The flooding from sources located in this room will not impair the safe shutdown capability of the SRE.
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 wi thin the room contain insufficient energ y 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 73 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 74 OF 84)
REV 15 4/09 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 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. 1. The flooding from sources located in this room will not impair the safe shutdown capability of the SRE.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 75 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 76 OF 84)
REV 15 4/09 Room Number R-D121 Title Passage Remarks Flooding Analysis 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. 1. Flooding from sources located in this room will not impair the safe shutdown capability of the SRE.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 77 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 78 OF 84)
REV 15 4/09 Room Number R-D122 Title Passage Remarks Flooding Analysis 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. 1. The flooding from sources located in this room will not impair the safe shutdown capability of the SRE.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 79 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 80 OF 84)
REV 15 4/09 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 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. 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 mitigation of the event or safe shutdown of the plant.
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 fa ilures 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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 81 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F TABLE 3F-1 (SHEET 82 OF 84)
REV 15 4/09 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 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. 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 mitigation of the event or safe shutdown of the plant.
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.
VEGP-FSAR-3F TABLE 3F-1 (SHEET 83 OF 84)
REV 15 4/09 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
VEGP-FSAR-3F REV 15 4/09 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 li ne 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.
VEGP-FSAR-3F REV 14 10/07 TABLE 3F-2 UNIT 1 AFW PUMP HOUSE DESIGN PARAMETERS Initial Conditions Design Conditions Blowdown Node No.(a) (Btu/lb)
Pressure (psia)
Temp.
(°F) Rel.
Hum.
(%)
Pressure (psia)
Temp.(b) (°F) Rel.
Hum.
(%) Flood Level (in.) Nodal Volume (ft 3) Nodes(a) Press.
Flow Area (ft 2) Calc.
after Break (psia) Peak Temp.
(°F) Time (s) Mass Flowrate (lb/s) Enthalpy Steam Line 1. 14.7 120 50 17.7 320 0-100 (c) 12.6x10 3 1-2 24.0 15.4 184 0 0 0 2. 14.7 120 50 17.7 320 0-100 (d) 10.9x10 3 1-3 20.0 15.5 249 0.004 237.4 1188.6 0.01210.51156.7 3. (atm) 14.7 120 50 - - - - - - - - - 0.1 143.8 1187.3 0.2126.71191.1 0.5110.81191.7 0.9104.51191.3 2.0100.91190.4 20100.71190.1 Auxiliar yFeedwate r 0 0 60282 1800142
- 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 VEGP-FSAR-3F TABLE 3F-3A (SHEET 1 OF 4)
REV 14 10/07 UNIT 1 MSIV/MFIV ROOM FLOODING (CASE 2) AND PRESSURE (CASE 1) ANALYSIS Initial Conditions Design Conditions Node(a) No. Pressure (psia)
Temp. (°F) Rel. Hum.
(%)
Pressure (psia) Temp (°F) Rel. Hum.
(%)
Flood Level (in.) Node Volume (ft 3) Calc Peak Pressure (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
VEGP-FSAR-3F TABLE 3F-3A (SHEET 2 OF 4)
REV 14 10/07 Initial Conditions Design Conditions Node (a No. Pressure
(psia)
Temp. (°F) Rel. Hum.
(%)
Pressure (psia) Temp (°F) Rel. Hum.
(%)
Flood Level (in.)
Node Volume (ft 3) Calc Peak Pressure (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
VEGP-FSAR-3F TABLE 3F-3A (SHEET 3 OF 4)
REV 14 10/07 Initial Conditions Design Conditions Node(a) No. Pressure
(psia)
Temp. (°F) Rel. Hum.
(%)
Pressure (psia)
Temp (°F) Rel. Hum.
(%)
Flood Level (in.)
Node Volume (ft 3) Calc Peak Pressure (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.
VEGP-FSAR-3A REV 14 10/07 TABLE 3F-3A (SHEET 4 OF 4)
Blowdown Data (Case 1) (Break at Node 10)
Time (s)
Mass Flowrate (lbm/s) Enthalpy (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
VEGP-FSAR-3B REV 15 4/09 TABLE 3F-3B (SHEET 1 OF 2)
UNIT 1 CONTROL BUILDING MSIV/MFIV VAULT AREA MSLB TEMPERATURE (CASE 3) ANALYSIS Initial Conditions Design Conditions Node No.(a) Description Pressure (psia)
Temp.
(°F) Rel.
Hum.
(%) Pressure (psia) Temp.
(°F) Rel.
Hum.
(%) Volume (ft 3) Node Elev.
(ft) Calc. Peak (c) Temp. (°F) 1 West Steam Line Room (Lower) 14.7 120 50 29.7 320 0-100 18,361.4 224.0 483.2 (c) 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
VEGP-FSAR-3B REV 15 4/09 TABLE 3F-3B (SHEET 2 OF 2)
Flow Path Data Blowdown Data (1.0 ft 2 MSLB(b) at 102% Power)
Flow Path No. Upstream Node Downstream
Node Flow Area ft 2 Time (s)
Mass Flow-
rate (lbm/s)
Enthalpy (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 ft 2 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.
VEGP-FSAR-3F TABLE 3F-4 (SHEET 1 OF 3)
REV 14 10/07 EVALUATION OF RCS LOOP BRANCH LINE BREAKS
RCS Loop Nozzle No.
Branch Line Description Branch Line
Identification Pipe Break Evaluation Loop 1 cold leg Nozzle - 5 Boron injection tank (safety injection system (SIS)) 1204-243-1£ in. Pipe whip: Pipe whips but does not result in t he failure of any essential components. Jet impingement: Breaks do not result in the failure of any essential components.
Nozzle - 14 Accumulator injection 1204-124-10 in. Pipe whip: Two restraints were installed to prevent pipe whip.(b) Jet impingement: One barrier was installed to prevent the failure of any essential components.(b) Upstream of valve 083 1204-120-10 in.
1204-042-6 in.
1204-036-2 in. Pipe whip: Two restraints were installed to prevent pipe whip. Jet impingement: Breaks do not result in the failure of any essential components.
Nozzle - 11 CVCS normal charging li ne 1208-009-3 in. to valve 035 (second check valve) Pipe whip: Three restraints were installed to prevent pipe whip. Jet impingement: Breaks do not result in the failure of any essential components.
CVCS normal charging upstream of valve 035(second valve) 1208-008-3 in. Pipe whip: One restraint was installed to prevent pipe whip. Jet impingement: Breaks do not result in the failure of any
essential components.
Nozzle - 12 Pressurizer spray header 1201-029-4 in. Pipe whip: One restraint was installed to prevent pipe whip. Jet impingement: Breaks do not result in the failure of any
essential components.
Loop 1 crossover leg Nozzle - 18 Reactor coolant loop drain 1201-031-2 in. Pipe whip: Pipe whips but does not result in the failure of any essential components. Jet impingement: Breaks do not result in the failure of any essential components.
Loop 1 hot leg Nozzle - 15 RHR system recirculation/SIS 1201-036-12 in. (b) Pipe whip: Pipe whips but does not result in the failure of any essential components. Jet impingement: Breaks do not result in the failure of any essential components.
VEGP-FSAR-3F TABLE 3F-4 (SHEET 2 OF 3)
REV 14 10/07 RCS Loop Nozzle No.
Branch Line Description Branch Line
Identification Pipe Break Evaluation Loop 2 cold leg Nozzle - 5 Boron injection tank (SIS) 1204-244-1 1/2 in. Pipe whip: Pipe whips but does not result in the failure of any essential components. Jet impingement: Breaks do not result in the failure of any essential components.
Nozzle - 14 Accumulator injection 1204-125-10 in.(b) Pipe whip: Two restraints were installed to prevent pipe whip.(b) Jet impingement: Breaks do not result in the failure of any essential components.(b) Upstream of valve 084 1204-121-10 in.
1204-043-6 in.
1204-035-2 in. Pipe whip: Pipe whips but does not result in the failure of any essential components. Jet impingement: Breaks do not result in the failure of any essential components.
Loop 2 crossover leg Nozzle - 18 Reactor coolant loop drain 1201-042-2 in. Pipe whip: Pipe whips but does not result in the failure of any essential components. Jet impingement: Breaks do not result in the failure of any essential components.
Loop 2 hot leg Nozzle - 13 RHR SIS injection 1204-024-6 in. Pipe whip: One restraint was installed to prevent pipe whip. Jet impingement: Breaks do not result in the failure of any
essential components.
Loop 3 cold leg Nozzle - 5 Boron injection tank (SIS) 1204-245-1 1/2 in. Pipe whip: Pipe whips but does result in the failure of any essential components. Jet impingement: Breaks do not result in the failure of any essential components.
Nozzle - 14 Accumulator injection 1204-126-10 in.(b) Pipe whip: Two restraints were installed to prevent pipe whip.(b) Jet impingement: Breaks do not result in the failure of any essential components.(b) Upstream of valve 085 1204-122-10 in.
1204-044-6 in.
1204-034-2 in. Pipe whip: Two restraints were installed to prevent pipe whip. Jet impingement: Breaks do not result in the failure of any essential components.
Nozzle - 4 Letdown (CVCS) 1201-048-3 in. Pipe whip: Pipe whips but does not result in the failure of any essential components. Jet impingement: Breaks do not result in the failure of any essential components.
Loop 3 crossover leg Nozzle - 18 Reactor coolant loop drain 1201-046-2 in. Pipe whip: Pipe whips but does not result in the failure of any essential components. Jet impingement: Breaks do not result in the failure of any essential components.
VEGP-FSAR-3F TABLE 3F-4 (SHEET 3 OF 3)
REV 14 10/07 RCS Loop Nozzle No.
Branch Line Description Branch Line
Identification Pipe Break Evaluation Loop 3 hot leg Nozzle - 13 RHR SIS injection 1204-025-6 in. Pipe whip: One restraint was installed to prevent pipe whip. Jet impingement: Breaks do not result in the failure of any
essential components.
Loop 4 cold leg Nozzle - 5 Boron injection tank (SIS) 1204-246-1 1/2 in. Pipe whip: Pipe whips but does not result in the failure of any essential components. Jet impingement: Breaks do not result in the failure of any essential components.
Nozzle - 14 Accumulator injection 1204-127-10 in.(b) Pipe whip: Two restraints were installed to prevent pipe whip.(b) Jet impingement: One barrier was installed to prevent the failure of any essential components.(b) Upstream of valve 086 1204-123-10 in.(b) 1204-045-6 in.
1204-033-2 in. Pipe whip: Pipe whips but does not result in the failure of any essential components. Jet impingement: Breaks do not result in the failure of any essential components.
Nozzle - 12 Pressurizer spray header 1201-030-4 in.
1201-030-6 in.
1208-012-2 in. Pipe whip: One restraint was installed to prevent pipe whip.
Jet impingement: Breaks do not result in the failure of any
essential components.
Nozzle - 11 CVCS alternate charging line 1208-007-3 in. to valve 037 (second check valve) Pipe whip: Three restraints were installed to prevent pipe whip. Jet impingement: Breaks do not result in the failure of any essential components.
Upstream of valve 037 (second valve) 1208-488-3 in. Pipe whip: One restraint was installed to prevent pipe whip. Jet impingement: Breaks do not result in the failure of any
essential components.
Loop 4 crossover leg Nozzle - 18 Reactor coolant loop drain 1201-051-2 in. with excess letdown Pipe whip: Pipe whips but does not result in the fail ure of any essential components. Jet impingement: Breaks do not result in the failure of any essential components.
Loop 4 hot leg Nozzle - 15 RHR system recirculation/SIS injection 1201-049-12 in.(b) 1204-021-6 in. Pipe whip: One restraint was installed. Pipe whips but does not result in failure of any essential components. Jet impingement: One barrier was installed to prevent the failure of any essential components.(b)
- a. See drawing AX6DD309.
- b. Applies to Unit 1 only.
VEGP-FSAR-3 REV 14 10/07 TABLE 3F-5 JET IMPINGEMENT BARRIERS
System Line Number Break Location Target Number Barrier Number Building Accumulator injection loop-4
1204-127-10 in.
P-14DB-0411-L-1
P-14DB-0411-L-2 1201-029-4 in.
BC-3 Containment (Unit 1 only)
Main safety loops 1 and 4 from auxiliary
building to main steam tunnel
1301-008-38 in.
P-17CA-1059-C 1305-056-26 in.
BT-2 Tunnel IT1 Main feedwater loops 1-4 (tunnel to
turbine building)
1305-055-36 in.
P-12EA-1135-C 1301-007-36 in.
BT-1 Tunnel IT1
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 1410/07 TIME-HISTORY PLOT MAXIMUM COMPARTMENT COMPOSITE TEMPERATURE FOR MSIV AREA BREAK NODE FIGURE 3F-3
VEGP-FSAR-4
4.1-1 REV 15 4/09 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 + 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 new 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 previously in both VEGP Units 1 and 2. Recent fuel reloads contain an advanced zirconium alloy clad fuel known as ZIRLO, reference 2, and 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.
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 VEGP-FSAR-4
4.1-2 REV 15 4/09 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 fo rm 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. 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 UO 2 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 (ZrB
- 2) 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 eit her 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 silv er, 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.
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.
VEGP-FSAR-4
4.1-3 REV 15 4/09 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 fi xed 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.
VEGP-FSAR-4 REV 16 10/10 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 (10 6 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 (psia) 2220 2200 2220 6. Minimum DNBR for design transients Typical flow channel (LOPAR) 1.30 (g) 1.30 (VANTAGE + / VANTAGE 5) 1.24 Thimble (cold wall) flow channel (LOPAR) (g) (VANTAGE + / VANTAGE 5) 1.23 7. DNB correlation (LOPAR) R (W-3 with (g) R (W-3 with VANTAGE + / (VANTAGE 5) modified spacer factor) WRB-2 (d) modified spacer factor) Coolant flow (e) 8. Total vessel flowrate (10 6 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(ft
- 2) (LOPAR) (VANTAGE + /
VANTAGE 5) 51.1 (g) 54.1 51.1
- 11. Average velocity along fuel rods (ft/s) (LOPAR) 16.6 (g) 16.6 (VANTAGE + / VANTAGE 5) 15.3 2.62 12. Average mass velocity (10 6 lbm/h-ft 2) (LOPAR) 2.62 (g) (VANTAGE + / VANTAGE 5) 2.41 Coolant temperature
- 13. Nominal inlet (°F) 558.8 556.3 558.8
- 14. Average rise in vessel (°F) 59.4 64.2 59.4
- 15. Average rise in core (°F) 62.6 68.0 62.6 16. Average in core (°F) 591.8 592.3 591.8 VEGP-FSAR-4 REV 16 10/10 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 (°F) 588.5 588.4 588.5 18. Active heat transfer surface area, (ft
- 2) (LOPAR) 59,700 (g) 59,700 (VANTAGE + / VANTAGE 5) 57,505 19. Average heat flux (Btu/h-ft
- 2) (LOPAR) 189,800 (g) 189,800 (VANTAGE + / VANTAGE 5) 209,612 20. Maximum heat flux for normal operation (Btu/h-ft 2) (LOPAR) (VANTAGE + /
VANTAGE 5) 436,500 (a) (g) 524,030 (a) 440,300 (a) 21 Average linear power (kW/ft) 5.44 5.788 (i) 5.44 22. Peak Linear power for normal operation (kW/ft) (a) 12.5 14.47 (i) 12.6 23. Peak linear power resulting from overpower transients/operator errors, assuming a maximum overpower of 120% (KW/ft) 18.0 22.4(b) 18.0 24. Heat flux hot channel factor (F Q) 2.30 (c) 2.50 (c) 2.32 (c) 25. Peak fuel central temperature for prevention of centerline melt (ºF) 4700 4700 4700 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. UO 2 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 UO 2 (lb) (LOPAR) 222,739 (g) 222,739 (VANTAGE + / VANTAGE 5) 204,231 (j) 32. Zircaloy/ZIRLO Ž weight (lb) (active core) (LOPAR) (VANTAGE + / VANTAGE 5) 45,296 (g) 45,914 45,296 33. Number of grids per assembly (LOPAR) 8-Rtype (g) 8 - R type (VANTAGE + / VANTAGE 5) 2 - R type 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 VEGP-FSAR-4 REV 16 10/10 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 + / VANTAGE 5) 0.360 37. Diametral gap (in.) (LOPAR) 0.0065 (g) 0.0065 (VANTAGE + / VANTAGE 5) 0.0062 (non-IFBA) 38. Clad thickness (in.)
0.0225 0.0225 0.0225 39. Clad material (LOPAR) Zircaloy-4 (g) Zircaloy-4 (VANTAGE + / VANTAGE 5) Zircaloy-4/ZIRLOŽ Fuel pellets
- 40. Material UO 2 sintered UO 2 sintered UO 2 sintered 41. Density (% of theoretical) 95 95 95 42. Diameter (in.) (LOPAR) 0.3225 (g) 0.3225 (VANTAGE + / VANTAGE 5) 0.3088 (non-IFBA) 43. Length (in.) (LOPAR) 0.530 (g) 0.530 (VANTAGE + / VANTAGE 5) 0.370 (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 SS, cold-worked SS, cold-worked 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
VEGP-FSAR-4 REV 16 10/10 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. H 2O/U molecular ratio core, lattice, cold (LOPAR) 2.41 (g) 2.41 (VANTAGE + / VANTAGE 5) 2.73 First Cycle Fuel Enri chments (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 F Q = 2.32 for SNUPPS, F Q = 2.30 for VEGP (Original Design) and F Q = 2.50 for VEGP (Uprate Design).
- b. See paragraph 4.3.2.2.6.
- c. This is the maximum value of F Q 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.
VEGP-FSAR-4 REV 19 4/15 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, and stress analysis Static and dynamic modeling Blowdown code, FORCE, finite element, structural analysis code, and others 3.7.2.1 3.9.2 3.9.3 Fuel rod design Full performance characteristics (temperature, internal pressure, clad stress, etc.) Semiempirical thermal model of fuel rod with consideration of fuel density changes, heat transfer, fission gas release, etc. Westinghouse fuel rod design model 4.2.1.1 4.2.3.2 4.2.3.3 4.3.3.1 4.4.2.11 Nuclear design Cross-sections and group constants Microscopic data; macroscopic constants for homogenized core regions Modified ENDF/B library LEOPARD/CINDER type or PHOENIX-P or NEXUS/PARAGON 4.3.3.2 Group constants for control rods with self-shielding HAMMER-AIM or PHOENIX-P or NEXUS/PARAGON 4.3.3.2 X-Y and X-Y-Z power distributions, fuel depletion, critical boron concentrations, X-Y and X-Y-Z xenon distributions, reactivity coefficients 2-group diffusion theory TURTLE (2-D) or ANC (2-D or 3-D) 4.3.3.3 Axial power distributions, control rod worths, and axial xenon distribution 1-D, 2-group diffusion theory PANDA 4.3.3.3 Fuel rod power Integral transport theory LASER 4.3.3.1 Effective resonance temperature Monte Carlo weighting func tion 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 theory DOT 4.3.2.8
VEGP-FSAR-4 REV 19 4/15 TABLE 4.1-2 (SHEET 2 OF 2)
Analysis Technique Computer Code Section Referenced Thermal-hydraulic design Steady state Subchannel analysis of local fluid 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. VIPRE-01 4.4.4.5.2 Transient departure from nucleate boiling Subchannel analysis of local fluid 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. VIPRE-01 4.4.4.5.4
VEGP-FSAR-4 REV 14 10/07 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)
VEGP-FSAR-4 REV 14 10/07 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 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 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 INITIAL CORE TOP GRID TO STANDARD NOZZLE ATTACHMENT LOPAR FIGURE 4.2-5 (SHEET 1 OF 3)
REV 14 10/07 RELOAD CYCLE RECONSTITUTABLE TOP GRID TO NOZZLE ATTACHMENT LOPAR FIGURE 4.2-5 (SHEET 2 OF 3)
REV 14 10/07 THIMBLE/INSERT/TOP GRID SLEEVE BULGE JOINT GEOMETRY VANTAGE 5 FIGURE 4.2-5 (SHEET 3 OF 3)
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
REV 14 10/07 FULL-LENGTH RCCA OUTLINE FIGURE 4.2-9
REV 14 10/07 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
VEGP-FSAR-4
4.3-1 REV 19 4/15 4.3 NUCLEAR DESIGN 4.3.1 DESIGN BASES This section describes the design bases and functional requirements used in the nuclear design of the fuel and reactivity control system and relates these design bases to the general design
criteria (GDC) presented in 10 CFR 50, Appendix A. Where applicable, supplemental criteria, such as Final Acceptance Criteria for Emergency Core Cooling Systems, are addressed.
However, before discussing the nuclear design bases, it is appropriate to briefly review the four
major categories as ascribed to conditions of plant operation.
The full spectrum of plant conditions is divided into four categories, in accordance with the anticipated frequency of occurrence and risk to the public:
- Condition 1 - Normal operation.
- Condition 2 - Incidents of moderate frequency.
- Condition 3 - Infrequent faults.
- Condition 4 - Limiting faults.
In general, Condition 1 occurrences are accommodated with margin between any plant
parameter and the value of that parameter which would require either automatic or manual
protective action. Condition 2 incidents are accommodated with, at most, a shutdown of the
reactor with the plant capable of returning to operation after corrective action. Fuel damage (fuel damage as used here is defined as penetration of the fission product barrier; i.e., the fuel rod clad) is not expected during Condition 1 and Condition 2 events. It is not possible, however, to preclude a very small number of rod failures. These are within the capability of the chemical
and volume control system (CVCS) and are consistent with the plant design basis.
Condition 3 incidents do not cause more than a small fraction of the fuel elements in the reactor to be damaged, although sufficient fuel element damage might occur to preclude immediate
resumption of operation. The release of radioactive material due to Condition 3 incidents is not
sufficient to interrupt or restrict public use of those areas beyond the exclusion area boundary.
Furthermore, a Condition 3 incident does not by itself generate a Condition 4 fault or result in a
consequential loss of function of the reactor coolant or reactor containment barriers.
Condition 4 occurrences are faults that are not expected to occur but are defined as limiting faults which must be designed against. Condition 4 faults do not cause a release of radioactive
material that results in exceeding the limits of 10 CFR 100.
The core design power distribution limits related to fuel integrity are met for Condition 1 occurrences through conservative design and maintained by the action of the control system.
The requirements for Condition 2 occurrences are met by providing an adequate protection
system which monitors reactor parameters.
The control and protection systems are described
in chapter 7, and the consequences of condition 2, 3, and 4 occurrences are given in
chapter 15.
VEGP-FSAR-4
4.3-2 REV 19 4/15 4.3.1.1 Fuel Burnup 4.3.1.1.1 Basis A limitation on initial installed excess reacti vity or average discharge burnup is not required other than as is quantified in terms of other design bases, such as core negative reactivity feedback and shutdown margin discussed below. 4.3.1.1.2 Discussion Fuel burnup is a measure of fuel depletion which represents the integrated energy output of the fuel in MWd/tonne of uranium and is a convenient means for quantifying fuel exposure criteria.
The core design lifetime or design discharge burnup is achieved by installing sufficient initial excess reactivity in each fuel region and by follo wing a fuel replacement program (such as that
described in subsection 4.3.2) that meets all safety-related criteria in each cycle of operation.
Initial excess reactivity installed in the fuel, although not a design basis, must be sufficient to maintain core criticality at full-power operating conditions throughout cycle life with equilibrium
xenon, samarium, and other fission products present. The end of design cycle life is defined to
occur when the chemical shim concentration is essentially zero with control rods present to the
degree necessary for operational requirements (e.g., the controlling bank at the "bite" position).
In terms of chemical shim boron concentration, this represents approximately 10 ppm with no control rod insertion.
After the end of design cycle life is reached and reactivity is no longer sufficient to maintain criticality at full-power operating conditions, core criticality is maintained at reduced-power
conditions. 4.3.1.2 Negative Reactivity Feedbacks (Reactivity Coefficient) 4.3.1.2.1 Basis The fuel temperature coefficient will be negative, and the moderator temperature coefficient of reactivity will be nonpositive for full power operating conditions, thereby providing negative reactivity feedback characteristics at full power. Below 70-percent power, a moderator temperature coefficient of up to +7.0 pcm/
°F is allowed. From 70-percent to 100-percent power, the moderator temperature coefficient limit decreases linearly from +7.0 to 0.0 pcm/
°F. The design basis meets GDC 11. 4.3.1.2.2 Discussion When compensation for a rapid increase in reactivity is considered, there are two major effects.
These are the resonance absorption (Doppler) effects associated with changing fuel
temperature and the neutron spectrum and reactor composition change effects resulting from
changing moderator density. These basic physics characteristics are often identified by
reactivity coefficients. The use of slightly enr iched uranium ensures that the Doppler coefficient of reactivity is negative. This coefficient provi des the most rapid reactivity compensation.
VEGP-FSAR-4
4.3-3 REV 19 4/15 The core is also designed to have an overall moder ator temperature coefficient of reactivity which is nonpositive at 100-percent power and less than or equal to +7.0 pcm/
°F below 70-percent power. From 70-percent power to 100-percent power, the maximum allowed moderator temperature coefficient decreases linearly from +7.0 pcm/
°F to 0.0 pcm/
°F. At full power, void content provides another, slower compensatory e ffect. The moderator temperature coefficient is maintained at or below the above stated limit through the use of fixed burnable absorber (BA) rods, and/or integral fuel burnable absorbers (IFBA) in the form of a zirconium diboride (ZrB
- 2) coating on the enriched fuel pellets, and/or control rods by limiting the reactivity held down by soluble boron.
Burnable absorber content (quantity and distribution) is not stated as a design basis. However, for some reloads, the use of burnable absorbers may be necessary for peaking factor limit
control and for the accomplishment of meeting the moderator temperature coefficient limits
discussed above. 4.3.1.3 Control of Power Distribution 4.3.1.3.1 Basis The nuclear design basis is that, with at least a 95-percent confidence level: A. The fuel will not be operated at greater than 14.5 kW/ft under normal operating conditions, including an allowance of 2 percent for calorimetric error. B. Under abnormal conditions, including the maximum overpower condition, the fuel peak power will not cause melting, as defined in paragraph 4.4.1.2. C. The fuel will not operate with a power distribution that violates the departure from nucleate boiling (DNB) design basis (i.e., the departure from nucleate boiling ratio (DNBR) shall not be less than the design limit DNBR discussed in subsection
4.4.1) under Condition 1 and 2 events, including the maximum overpower
condition. D. Fuel management will be such as to produce values of fuel rod power and burnup consistent with the assumptions in the fuel rod mechanical integrity analysis of
section 4.2.
The above basis meets GDC 10. 4.3.1.3.2 Discussion Calculation of extreme power shapes which affe ct fuel design limits is performed with proven methods and verified frequently with measurements from operating reactors. The conditions under which limiting power shapes are assumed to occur are chosen conservatively with regard
to any permissible operating state.
Even though there is close agreement between calculated peak power and measurements, a nuclear uncertainty (paragraph 4.3.2.2.1) is applied to calculated peak local power. Such a margin is provided both for the analysis for normal operating states and for anticipated
VEGP-FSAR-4
4.3-4 REV 19 4/15 4.3.1.4 Maximum Controlled Reactivity Insertion Rate 4.3.1.4.1 Basis The maximum reactivity insertion rate due to withdrawal of rod cluster control assemblies (RCCAs) at power or by boron dilution is limited. During normal at power operation, the
maximum controlled reactivity insertion rate is limited. A maximum reactivity change rate for accidental withdrawal of two control banks is set so that peak heat generation rate and DNBR
do not exceed the maximum allowable at overpower conditions. This satisfies GDC 25.
The maximum reactivity worth of control rods and the maximum rates of reactivity insertion employing control rods are limited to preclude rupture of the coolant pressure boundary or disruption of the core internals to a degree which would impair core cooling capacity due to a
rod withdrawal or an ejection accident. (See chapter 15.)
Following any Condition 4 event (rod ejection, steam line break, etc.) the reactor can be brought to the shutdown condition, and the core will maintain acceptable heat transfer geometry. This
satisfies GDC 28. 4.3.1.4.2 Discussion Reactivity addition associated with an accidental withdrawal of a control bank (or banks) is limited by the maximum rod speed (or travel rate) and by the worth of the bank(s). For this
reactor, the maximum control rod speed is 45 in./min.
The reactivity change rates are conservatively calculated, assuming unfavorable axial power and xenon distributions. The peak xenon burnout rate is significantly lower than the maximum
reactivity addition rate for normal operation and for accidental withdrawal of two banks. 4.3.1.5 Shutdown Margins 4.3.1.5.1 Basis Minimum shutdown margin as specified in the Technical Specifications or Technical Requirements Manual is required in all operating modes, in the hot standby shutdown condition, and in the cold shutdown condition.
In all analyses involving reactor trip, the single, highest worth rod cluster control assembly is postulated to remain untripped in its full-out position (stuck rod criterion). This satisfies GDC 26.
4.3.1.5.2 Discussion Two independent reactivity control systems are pr ovided: control rods and soluble boron in the coolant. The control rod system can compensate for the reactivity effects of the fuel and water temperature changes accompanying power level changes over the range from full load to no load. In addition, the control rod system provides the minimum shutdown margin under
Condition 1 events and is capable of making the core subcritical rapidly enough to prevent VEGP-FSAR-4
4.3-5 REV 19 4/15 exceeding acceptable fuel damage limits (very sma ll number of rod failures), assuming that the highest worth control rod is stuck out upon trip.
The boron system can compensate for all xenon burnout reactivity changes and will maintain the reactor in the cold shutdown condition. Thus, backup and emergency shutdown provisions
are provided by mechanical and chemical shim control systems which satisfy GDC 26. 4.3.1.5.3 Basis When fuel assemblies are in the pressure vessel and the vessel head is not in place, k eff will be maintained at or below 0.95 with control rods and soluble boron. Further, the fuel will be maintained sufficiently subcritical that removal of all rod cluster control assemblies will not result
in criticality. 4.3.1.5.4 Discussion American National Standards Institute (ANSI) N18.2 specifies a k eff not to exceed 0.95 in spent fuel storage racks and transfer equipment flooded with pure water and a k eff not to exceed 0.98 in normally dry new fuel storage racks, assuming opt imum moderation. No criterion is given for the refueling operation. However, a 5-percent margin, which is consistent with spent fuel storage and transfer and the new fuel storage, is adequate for the controlled and continuously
monitored operations involved.
The boron concentration required to meet the refueling shutdown criteria is specified in the Core Operating Limits Report, which is referenced in the Technical Specifications. Verification that
these shutdown criteria are met, including uncertainties, is achieved using standard design
methods. The subcriticality of the core is continuously monitored as described in the Technical
Specifications. 4.3.1.6 Stability 4.3.1.6.1 Basis The core will be inherently stable to power oscillations at the fundamental mode. This satisfies GDC 12.
Spatial power oscillations within the core with a constant core power output, should they occur, can be reliably and readily detected and suppressed. 4.3.1.6.2 Discussion Oscillations of the total power output of the core , from whatever cause, are readily detected by the loop temperature sensors and by the nuclear instrumentation. The core is protected by
these systems; a reactor trip would occur if power increased unacceptably, preserving the
design margins to fuel design limits. The stability of the turbine/steam generator/core systems
and the reactor control system is such that total core power oscillations are not normally
possible. The redundancy of the protection circuits ensures an extremely low probability of
exceeding design power levels.
VEGP-FSAR-4
4.3-6 REV 19 4/15 The core is designed so that diametral and azimuthal oscillations due to spatial xenon effects
are self-damping, and no operator action or control action is required to suppress them. The
stability to diametral oscillations is so great that this excitation is highly improbable. Convergent
azimuthal oscillations can be excited by prohibited motion of individual control rods. Such oscillations are readily observable and alarmed, using the excore long ion chambers.
Indications are also continuously available from incore thermocouples and loop temperature
measurements. Movable incore detectors can be activated to provide more detailed
information. In all proposed cores, these horizontal plane oscillations are self-damping by virtue
of reactivity feedback effects designed into the core.
However, axial xenon spatial power oscillations may occur in the middle to end of core life. The control bank and excore detectors are provided for control and monitoring of axial power
distributions.
Assurance that fuel design limits are not exceeded is provided by reactor overpower T and overtemperature T trip functions, which use the measured axial power imbalance as an input.
Detection and suppression of xenon oscillations are discussed in paragraph 4.3.2.7. 4.3.1.7 Anticipated Transients Without SCRAM The effects of anticipated transients with failure to trip are not considered in the design bases of the plant. Analysis has shown that the likelihood of such a hypothetical event is negligibly
small. Furthermore, analysis of the consequences of hypothetical failure to trip following
anticipated transients has shown that no significant core damage would result, system peak
pressures would be limited to acceptable values, and no failure of the reactor coolant system
would result.
(1) 4.
3.2 DESCRIPTION
4.3.2.1 Nuclear Design Description The reactor core consists of a specified number of fuel rods held in bundles by spacer grids and top and bottom fittings. The fuel rods are constructed of Zircaloy/ZIRLOŽ cylindrical tubes
containing UO 2 fuel pellets. The bundles, known as fuel assemblies, are arranged in a pattern which approximates a right circular cylinder.
Each fuel assembly contains a 17 x 17 rod array composed of 264 fuel rods, 24 guide thimbles, and an incore instrumentation tube. Figure 4.2-1 shows a cross-sectional view of the 17 x 17
fuel assemblies. Further details of the fuel assembly are given in section 4.2.
For initial core loading, the fuel rods within a given assembly have the same uranium enrichment in both the radial and axial planes. Fuel assemblies of three different enrichments
are used in the initial core loading to establish a favorable radial power distribution. Figure 4.3-
1 shows the fuel loading pattern to be used in the first core. Two regions consisting of the two
lower enrichments are interspersed to form a checkerboard pattern in the central portion of the
core. The third region is arranged around the periphery of the core and contains the highest
enrichment. The enrichments for the VEGP initial cycle are shown in table 4.3-1.
For reload cores, VANTAGE 5 fuel assemblies may be used. Detailed descriptions of the VANTAGE 5 fuel features are given in section 4.2. Integral fuel burnable absorbers (IFBA) in
the central portion of the fuel stack may be used in reload cores to control excess reactivity.
VEGP-FSAR-4
4.3-7 REV 19 4/15 Axial blankets consisting of unenriched, mid-enriched, or fully enriched solid or annular fuel
pellets may be placed at the ends of the enriched pellet stack. The unenriched or mid-enriched
axial blanket pellets are used in reload cores to reduce 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 typical reloading pattern is also shown in figure 4.3-1 with new fuel interspersed checkerboard-style in the center and depleted fuel on the periphery. The core will normally
operate approximately 18 months between refueling, accumulating approximately 20,000
MWd/tonne of uranium per year. The exact reloading pattern, the initial and final positions of
assemblies, and the number of fresh assemblies and their placement are dependent on the
energy requirement for the next cycle and burnup and pow er histories of the previous cycles.
The core average enrichment is determined by the amount of fissionable material required to
provide the desired core lifetime and energy requi rements. The physics of the burnout process is such that operation of the reactor depletes the amount of fuel available due to the absorption
of neutrons by the U-235 atoms and their subsequent fission. In addition, the fission process
results in the formation of fission products, some of which readily absorb neutrons. These
effects, depletion and the buildup of fission products, are partially offset by the buildup of
plutonium shown in figure 4.3-2 for a typical 17 x 17 fuel assembly, which occurs due to the
nonfission absorption of neutrons in U-238. Therefore, at the beginning of any cycle a reactivity
reserve equal to the depletion of the fissionable fuel and the buildup of fission product poisons
over the specified cycle life must be built into the reactor. This excess reactivity is controlled by removable neutron-absorbing material in the form of boron dissolved in the primary coolant and
BA rods and/or ZrB 2-coated fuel pellets. The stack length of the coated fuel pellets may vary for different reload designs, with the optimum length determined on a design-specific basis.
The concentration of the soluble boron is varied to compensate for reactivity changes due to fuel burnup, fission product poisoning including xenon and samarium, BA depletion, and the
cold-to-operating moderator temperature change. Using its normal or emergency boration path, the CVCS is capable of inserting negative reactivity at a rate of approximately 30 pcm/min when
the reactor coolant boron concentration is 1000 ppm and approximately 35 pcm/min when the
reactor coolant boron concentration is 100 ppm. The peak burnout rate for xenon is 25
pcm/min. (Subsection 9.3.4 discusses the capability of the CVCS to counteract xenon decay.)
Rapid transient reactivity requirements and safe ty shutdown requirements are met with control rods. As the boron concentration is increased, the moderator temperature coefficient becomes less negative. High boron concentrations will cause the moderator temperature coefficient to be
positive at beginning of life (BOL). Therefore, burnable absorber rods and/or IFBAs are used to
reduce the soluble boron concentration sufficiently to ensure that the moderator coefficient is nonpositive at full power, is less than or equal to +7.0 pcm/
°F below 70-percent power, and below a linearly decreasing limit of +7.0 pcm/
°F to 0.0 pcm/
°F between 70-percent power and 100-percent power, respectively.
During operation, the poison content in these rods is depleted, thus adding positive reactivity to offset some of the negative reactivity from fuel depletion and fission product buildup. The
depletion rate of the BA rods and/or IFBA coated rods is not critical, since chemical shim is
always available and flexible enough to cover any possible deviations in the expected BA
depletion rate. Figure 4.3-3 is a plot of core depletion with and without BA rods, (as is typical of
the first cycle and a typical reload containing IFBAs). Note that even at end of life (EOL)
conditions some residual poison remains in the BA rods, resulting in a net decrease in the cycle
lifetime. The IFBA coating at EOL conditions leaves no residual poison.
VEGP-FSAR-4
4.3-8 REV 19 4/15 In addition to reactivity control, the BA rods and/or fuel rods containing ZrB 2-coated fuel pellets are strategically located to provide a favorable radial power distribution. Figure 4.3-4 shows the
BA distributions within a fuel assembly for the several BA patterns used in a 17 x 17 array and
also shows the standard within-assembly configurations of the fuel rods containing ZrB 2-coated fuel pellets for several IFBA patterns. Typical reload core BA loading patterns for discrete BAs (WABA or borosilicate glass) and IFBAs are shown in figure 4.3-5. These burnable absorber (IFBA) loading patterns were reassessed to achieve the most efficient absorber orientation.
The revised patterns provide improved peaking factor and reactivity holddown benefits. Use of
the revised patterns commenced with Vogtle Unit 1 Region 9 and Vogtle Unit 2 Region 7.
Tables 4.3-1 through 4.3-3 contain summaries of reactor core design parameters including reactivity coefficients, delayed neutron fraction, and neutron lifetimes. Sufficient information is
included to permit an independent calculation of the nuclear performance characteristics of the
core. 4.3.2.2 Power Distribution The accuracy of power distribution calculat ions has been confirmed through approximately 1000 flux maps during some 20 years of operation under conditions very similar to those expected.
Details of this confirmation are given in reference 2 and in paragraph 4.3.2.2.7.
4.3.2.2.1 Definitions Power distributions are quantified in terms of hot channel factors. These factors are a measure of the peak pellet power within the reactor core and the total energy produced in a coolant
channel, relative to the total reactor power output, and are expressed in terms of quantities
related to the nuclear or thermal design; namely, power density is the thermal power produced
per unit volume of the core (kW/l).
Linear power density is the thermal power produced per unit length of active fuel (kW/ft). Since fuel assembly geometry is standardized, this is the unit of power density most commonly used.
For all practical purposes, it differs from kilowatts per liter by a constant factor, which includes
geometry effects and the fraction of the total thermal power generated in the fuel rod.
Average linear power density is the total thermal power produced in the fuel rods divided by the total active fuel length of all rods in the core.
Local heat flux is the heat flux at the surface of the cladding (Btu/ft 2/h). For nominal rod parameters, this differs from linear power density by a constant factor.
Rod power or rod integral power is the length integrated linear power density in one rod (kW).
Average rod power is the total thermal power produced in the fuel rods divided by the number of fuel rods (assuming all rods have equal length).
The hot channel factors used in the discussion of power distributions in this section are defined as follows:
F Q heat flux hot channel factor, is defined as the maximum local heat flux on the surface of a fuel rod divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.
VEGP-FSAR-4
4.3-9 REV 19 4/15 NQ F, nuclear heat flux hot channel factor, is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and rod parameters. E Q F, engineering heat flux hot channel factor, is the allowance on heat flux required for manufacturing tolerances. The engineering factor allows for local variation in enrichment, pellet
density and diameter, surface area of the fuel rod, and eccentricity of the gap between pellet
and clad. Combined statistically, the net effect is a factor of 1.03 to be applied to fuel rod
surface heat flux. N H F, nuclear enthalpy rise hot channel factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.
Manufacturing tolerances, hot channel power distribution, and surrounding channel power distributions are treated explicitly in the calculation of the DNBR described in section 4.4.
It is convenient for the purposes of discussion to define subfactors of F Q. However, design limits are set in terms of the total peaking factor.
F Q = total peaking factor or heat flux hot channel factor.
ft kW average ft kW maximum= Without densification effects:
E Q N U N Z N XY E Q NQQFFFFFFFxxx=x= where N QF and E QF are defined above and :
N UF = factor for measurement conserv atism, assumed to be 1.05, when using 44 detector thimbles, and is 1.05 + [2.0 {3-T / (14.5)}] / 100 , where T equals the number of thimbles, when using 29 and < 44 detector thimbles.
N XYF = ratio of peak power density to average power density in the horizontal plane of peak local power.
N ZF = ratio of the power per unit core height in the horizontal plane of peak local power to the average value of power per unit core height. If the plane of peak local power coincides with the plane of maximum power per unit core height, the N ZF is the core average axial peaking factor.
To include the allowances made for densification effects, which are height dependent, the following quantities are defined: S(Z) = the allowances made for densification effects at height Z in the core (paragraph 4.3.2.2.5.).
VEGP-FSAR-4
4.3-10 REV 19 4/15 P(Z) = ratio of the power per unit core height in the horizontal plane at height Z to the average value of power per unit core height.
N XYF(Z) = ratio of peak power density to average power density in the horizontal plane of height Z.
then: F Q = total peaking factor. = ft kW average ft kW maximum Including densification allowance:
F Q = max (N XYF(Z) x P(Z) x S(Z)) x N UF x E Q F 4.3.2.2.2 Radial Power Distributions The power shape in horizontal sections of the core at full power is a function of the fuel assembly and BA loading patterns, the control rod pattern, and the fuel burnup distribution.
Thus, at any time in the cycle, a horizontal section of the core can be characterized as
unrodded or with group D control rods. These two situations combined with burnup effects
determine the radial power shapes which can exist in the core at full power. Typical first cycle values of N XYF are the radial factors (BOL to EOL) given in table 4.3-2. The effect on radial power shapes of power level, xenon, samarium, and moderator density effects are also
considered, but these are quite small. The effect of nonuniform flow distribution is negligible.
While radial power distributions in various planes of the core are often illustrated, since the
moderator density is directly proportional to enthalpy, the core radial enthalpy rise distribution, as determined by the integral of power up each channel, is of greater interest. Figures 4.3-6
through 4.3-11 show typical reload radial power distributions for one-fourth of the core for
representative operating conditions. These conditions are as follows: A. Hot full power (HFP) at BOL, unrodded, no xenon. B. HFP at BOL, unrodded, equilibrium xenon.
C. HFP near BOL, bank D in, equilibrium xenon.
D. HFP near middle of life (MOL), unrodded equilibrium xenon.
E. HFP near EOL, unrodded, equilibrium xenon.
F. HFP at EOL, bank D in, equilibrium xenon. Since the position of the hot channel varies from time to time, a single-reference radial design power distribution is selected for DNB calculations. This reference power distribution is chosen
conservatively to concentrate power in one area of the core, minimizing the benefits of flow
redistribution. Assembly powers are normalized to core average power. The radial power
distribution within a fuel rod and its variation with burnup as utilized in thermal calculations and
fuel rod design are discussed in section 4.4.
VEGP-FSAR-4
4.3-11 REV 19 4/15 4.3.2.2.3 Assembly Power Distributions For the purpose of illustration, typical rodwis e power distributions from the BOL and EOL conditions corresponding to figures 4.3-7 and 4.3-10, respectively, are given for the same
assembly in figures 4.3-12 and 4.3-13, respectively.
Since the detailed power distribution surrounding the hot channel varies from time to time, a conservatively flat radial assembly power di stribution is assumed in the DNB analysis, described
in section 4.4, with the rod of maximum integrated power artificially raised to the design value of N H F. Care is taken in the nuclear design of all fuel cycles and all operating conditions to ensure that a flatter assembly power distribution does not occur with limiting values of N H F. 4.3.2.2.4 Axial Power Distributions The shape of the power profile in the axial or vertical direction is largely under the control of the
operator through either the manual operation of the control rods or automatic motion of rods
responding to manual operation of the CVCS. Nuclear effects which cause variations in the axial
power shape include moderator density, Doppler effect on resonance absorption, spatial
distribution of xenon, burnup, and axial distribution of fuel enrichment and burnable absorbers.
Automatically controlled variations in total power output and full-length rod motion are also important in determining the axial power shape at any time. Signals are available to the
operator from the excore ion chambers, which are long ion chambers outside the reactor vessel running parallel to the axis of the core. Separate signals are taken from the top and bottom
halves of the chambers. The difference between top and bottom signals from each of four pairs of detectors is displayed on the control panel and called the flux difference, I. Calculations of core average peaking factor for many plants and measurements from operating plants under
many operating situations are associated with either I or axial offset in such a way that an upper bound can be placed on the peaking factor. For these correlations, axial offset is defined as: axial offset = btbt+ and t and b are the top and bottom detector readings.
Representative reload axial power shapes for BOL, MOL, and EOL conditions are shown in figures 4.3-14 through 4.3-16. These figures cover a wide range of axial offset, including values
achieved by skewing xenon distributions. 4.3.2.2.5 Local Power Peaking Past creep collapse methods have assumed that pellet hangup occurs and an axial gap exists
in all fuel rods within a reactor core. The size of the axial gap is estimated from conservative
early-in-life fuel densification as determined from out-of-reactor sintering tests described in
Regulatory Guide 1.126. The size of the axial gap determines the magnitude of the power spike
factor applied to fuel designs. Axial gaps greater than 0.5 in. can lead to clad collapse and
significant flux and power spiking. Axial gaps less than 0.5 in. have been shown, reference 26, to result in no clad collapse and are associated with a relatively small power spike of 1 percent
or less. Reference 26, presents data that demonstrates that no large axial gaps, i.e., > 0.3 in.,
exist in current fuel designs during in-reactor performance. Therefore, the power spike factor is
no longer considered in the nuclear design.
VEGP-FSAR-4
4.3-12 REV 19 4/15 4.3.2.2.6 Limiting Power Distributions According to the ANSI classification of plant conditions (chapter 15), Condition 1 occurrences
are those expected frequently or regularly in the course of power operation, maintenance, or
maneuvering of the plant. As such, Condition 1 occurrences are accommodated with margin
between any plant parameter and the value of that parameter which would require either
automatic or manual protective action. Inasmuch as Condition 1 events occur frequently or
regularly, they must be considered from the point of view of affecting the consequences of fault
conditions (Conditions 2, 3, and 4). In this regar d, analysis of each fault condition described is generally based on a conservative set of initial conditions corresponding to the most adverse
set of conditions which can occur during Condition 1 operation.
The list of steady-state and shutdown conditions, permissible deviations, and operational transients is given in chapter 15. Implicit in the definition of normal operation is proper and
timely action by the reactor operator; that is, the operator follows recommended operating
procedures for maintaining appropriate power distributions and takes any necessary remedial
actions when alerted to do so by the plant instrumentation. Thus, as stated above, the worst or
limiting power distribution which can occur during normal operation is to be considered as the
starting point for analysis of Conditions 2, 3, and 4 events.
Improper procedural actions or errors by the operator are assumed in the design as occurrences of moderate frequency (Condition 2). Some of the consequences which might
result are discussed in chapter 15. Therefore, the limiting power shapes which result from such
Condition 2 events are those power shapes which deviate from the normal operating condition
at the recommended axial offset band; e.g., due to lack of proper action by the operator during a
xenon transient following a change in power level brought about by control rod motion. Power
shapes which fall in this category are used for determination of the reactor protection system
setpoints to maintain margin to overpower or DNB limits.
The means for maintaining power distributions within the required hot channel factor limits are described in the Core Operating Limits Report and the Technical Specifications. A complete
discussion of power distribution control in Westinghouse pressurized water reactors (PWRs) is
included in reference 6. Detailed background information on the design constraints on local
power density in a Westinghouse PWR, on the defined operating procedures, and on the
measures taken to preclude exceeding design limits is presented in the Westinghouse topical
report on power distribution control and load following procedures.
(7) The following paragraphs summarize these reports and describe the calculations used to establish the upper bound on
peaking factors.
The calculations used to establish the upper bound on peaking factors, F Q and N H F, include all of the nuclear effects which influence the radial and/or axial power distributions throughout core life for various modes of operation, including load follow, reduced power operation, and axial
Radial power distributions are calculated for the full-power condition. Fuel and moderator temperature feedback effects are included for the average enthalpy plane of the reactor. The
steady-state nuclear design calculations are done for normal flow with the same mass flow in
each channel and flow redistribution effects neglected. The effect of flow redistribution is
calculated explicitly where it is important in the DNB analysis of accidents. The effect of xenon on radial power distribution is small (compare figures 4.3-6 and 4.3-7) but is included as part of
the normal design process.
The core average axial profile can experience significant changes, which can occur rapidly as a result of rod motion and load changes and more slowly due to xenon distribution. For the study VEGP-FSAR-4
4.3-13 REV 19 4/15 of points of closest approach to axial power distribution limits, several thousand cases are
examined. Since the properties of the nuclear design dictate what axial shapes can occur, boundaries on the limits of interest can be set in terms of the parameters which are readily
observed on the plant. Specifically, the nuclear design parameters significant to the axial power distribution analysis are as follows:
- Core power level.
- Core height.
- Coolant temperature and flow.
- Coolant temperature program as a function of reactor power.
- Fuel cycle lifetimes.
- Rod bank worths.
- Rod bank overlaps.
Normal operation of the plant assumes compliance with the following conditions: A. Control rods in a single bank move together with no individual rod insertion differing by more than 12 steps (indicated) from the bank demand position. B. Control banks are sequenced with overlapping banks. C. The control full-length bank insertion limits are not violated.
D. Axial power distribution control procedures, which are given in terms of flux difference control and control bank position, are observed.
The axial power distribution procedures referred to above are part of the required operating procedures followed in normal operation.
Limits placed on the axial flux difference are designed to assure that the heat flux hot channel factor F Q is maintained within acceptable limits. The constant axial offset control (CAOC) operating procedures described in reference 7 require control of the axial flux difference at all
power levels within a permissible operating band about a target value corresponding to the equilibrium full power value. The relaxed axial offset control (RAOC) procedures to be
implemented in Unit 1 Cycle 4 and Unit 2 Cycle 3 and beyond, described in reference 39, were
developed to provide wider control band widths and, consequently, more operating flexibility.
These wider operating limits, particularly at lower power levels, can increase plant availability by
allowing quicker plant startups and increased maneuvering flexibility without trip or reportable
occurrences.
Further operating flexibility is achieved by combining RAOC operation with an F Q surveillance technical specification. F xy (z) surveillance requires periodic plant surveillance on the height-dependent radial peaking factor, F xy (z), for partial verification that operation will not cause the F Q (z) limit to be exceeded. In the F Q surveillance technical specification to be implemented in Unit 1 Cycle 4 and Unit 2 Cycle 3, F xy (z) surveillance will be replaced by F Q (z) surveillance.
Monitoring F Q (z) and increasing the measured value for expected plant maneuvers provides a more convenient form of assuring plant operation below the F Q (z) limit while retaining the intent of using a measured parameter to verify Technical Specification compliance.
In standard CAOC analysis described in reference 7, the generation of the normal operation power distributions is constrained by the rod insertion limits (RIL) and the I band limits. The purpose of RAOC is to find the widest permissible I-Power operating space by analyzing a VEGP-FSAR-4
4.3-14 REV 19 4/15 wide range of I. Therefore, the generation of normal operation power distributions is constrained only by the RIL for RAOC.
For a CAOC analysis, load-follow simulations are performed covering the allowed CAOC operating space to generate a typical range of allowed axial xenon distributions, which in turn
are used to calculate axial power distributions in both normal operation and Condition II
accident conditions. For a RAOC analysis, however, a reconstruction model described in
reference 39 is used as a more practical method to create axial xenon distributions covering the wider I-Power operating space allowed with RAOC operation. Each resulting power shape is analyzed to determine if LOCA constraints are met or exceeded. The total peaking factor, T Q F , is determined using standard synthesis methods as described in reference 7.
The F Q (z)'s are synthesized from axial calculations combined with radial factors appropriate for rodded and unrodded planes in the first cycle. In these calculations, the effects on the unrodded radial peak of xenon redistribution that occurs following the withdrawal of a control bank (or banks) from a rodded region is obtained from two-dimensional X-Y calculations. A 1.03
factor to be applied on the unrodded radial peak was obtained from calculations in which xenon
distribution was preconditioned by the presence of control rods and then allowed to redistribute
for several hours. A detailed discussion of this effect may be found in reference 7. The calculated values have been increased by a factor of 1.05, when using 44 detector thimbles, and is 1.05 + [2.0 {3-T / (14.5)}] / 100 , where T equals the number of thimbles, when using 29 and < 44 detector thimbles for conservatism and a factor of 1.03 for the engineering factor E QF. The envelope drawn over the calculated maximum (F Q (Z) x power) points in figure 4.3-21 represents an upper bound envelope on local power density versus elevation in the core. It should be emphasized that this envelope is a conservative representation of the bounding
values of local power density.
Finally, as previously discussed, this upper bound envelope is based on procedures of load follow which require operation within specified axial flux difference limits. Operation within this
upperbound envelope is ensured by limits contained in the Technical Specifications which rely
only upon excore surveillance supplemented by the normal monthly full core map requirement and by computer-based alarms on deviation from the allowed flux difference band.
Allowing for fuel densification effects, the average linear power at 3626 MW is 5.79 kW/ft. From
figure 4.3-21, the conservative upper bound value of normalized local power density, including
power uncertainty allowance, is 2.50, corresponding to a peak linear power of 14.5 kW/ft.
To determine reactor protection system setpoints with respect to power distributions, three categories of events are considered: rod control equipment malfunctions, operator errors of commission, and operator errors of omission. In evaluating these three categories of events, the core is assumed to be operating within the four constraints described above.
The first category comprises uncontrolled rod withdrawal (with rods moving in the normal bank sequence) for full-length banks. Also included are motions of the full-length banks below their
insertion limits, which could be caused, for exam ple, by uncontrolled dilution or primary coolant cooldown. Power distributions were calculated throughout these occurrences, assuming short-
term corrective action; that is, no transient xenon effects were considered to result from the
malfunction. The event was assumed to occur from typical normal operating situations, which
include normal xenon transients. It was further assumed in determining the power distributions
that total core power level would be limited by r eactor trip to below 120 percent. Since the study
is to determine protection limits with respect to power and axial offset, no credit was taken for
trip setpoint reduction due to flux difference. Representative results are given in figure 4.3-22 in
units of kW/ft. The peak power density which can occur in such events, assuming reactor trip at VEGP-FSAR-4
4.3-15 REV 19 4/15 or below 120 percent, is less than that required for centerline melt, including uncertainties and
densification effects.
The second category, also appearing in figure 4.3-22, assumes that the operator mispositions the full-length rod bank in violation of the insertion limits and creates short-term conditions not
included in normal operating conditions.
The third category assumes that the operator fails to take action to correct a flux difference violation (such as boration/dilution transient). Representative results for peak linear power
density include an allowance for calorimetric error. The peak linear power does not exceed
22.4 kW/ft, including the above factors. This value applies to both Zircaloy and/or ZIRLO TM clad fuel. The appropriate hot channel factors QF and N H F for peak local power density and for DNB analysis at full power based on analyses of possible operating power shapes are addressed in the Core Operating Limits Report and the Technical Specifications.
The maximum allowable QF can be increased with decreasing power, as shown in the Core Operating Limits Report and the Technical Specifications. Increasing N H F with decreasing power is permitted by the DNB protection setpoints and allows radial power shape changes with rod insertion to the insertion limits, as described in paragraph 4.4.4.3. The allowance for increased N H F permitted is addressed in the Technical Specifications. This becomes a design basis criterion which is used for establishing acceptable control rod patterns and control bank sequencing. Likewise, fuel loading patterns for each cycle are selected with consideration of this design criterion. The worst values of N H F for possible rod configurations occurring in normal operation are used in verifying that this criterion is met.
Typical radial power distributions are shown in figures 4.3-6 through 4.3-11. The worst values
generally occur when the rods are assumed to be at their insertion limits. Operation with rod positions above the allowed rod insertion limits provides increasing margin to the N H F criterion.
As discussed in section 3.2 of reference 8, it has been determined that the Technical
Specifications limits are met, provided the above conditions A and B are observed. These limits
are taken as input to the thermal-hydraulic design basis, as described in paragraph 4.4.4.3.1.
When a situation is possible in normal operation which could result in local power densities in excess of those assumed as the precondition for a subsequent hypothetical accident, but which
would not itself cause fuel failure, administrative controls and alarms are provided for returning
the core to a safe condition. These alarms are described in detail in chapter 7.
The independence of the various individual uncertai nties constituting the uncertainty factor on F Q enables the uncertainty (U QF) to be calculated by statistically combining the individual uncertainties on the limiting rod. The standard deviation of the resultant distribution of U Q F is determined by taking the square root of the sum of the variances of each of the contributing distributions.
(2) The measurement uncertainty is 1.05 (reference 2) and is applicable when using 44 detector thimbles. Reference 45 evaluated using 29 and < 44 detector thimbles, and the measurement uncertainty is 1.05 + [2.0 {3-T / {14.5)}] / 100 , where T equals the number of thimbles used. The value for E Q F is 1.03. The value for the rod bow factor, B QF, is 1.013, which accounts for the initial as-built rod bow.
VEGP-FSAR-4
4.3-16 REV 19 4/15 4.3.2.2.7 Experimental Verification of Power Distribution Analysis This subject is discussed in depth in reference 2. A summary of this report is given below. It
should be noted that power-distribution-related measurements are incorporated into the
evaluation of calculated power distribution information, using an incore instrumentation
processing code described in reference 9 or reference 41. The measured-versus-calculational
comparison is normally performed periodically thr oughout the cycle lifetime of the reactor, as required by Technical Specifications.
In a measurement of the heat flux hot channel factor, F Q , with the movable detector system described in subsections 7.7.1 and 4.4.6, the following uncertainties must be considered: A. Reproducibility of the measured signal.
B. Errors in the calculated relationship between detector current and local flux.
C. Errors in the calculated relationship between detector flux and peak rod power some distance from the measurement thimble.
The appropriate allowance for category A above has been quantified by repetitive measurements made with several intercalibrated detectors by using the common thimble
features of the incore detector system. This system allows more than one detector to access
any thimble. Errors in category B above are quant ified to the extent possible, by using the detector current measured at one thimble location to predict fluxes at another location, which is
also measured. Local power distribution predictions are verified in critical experiments on
arrays of rods with simulated guide thimbles, control rods, BAs, etc. These critical experiments
provide quantification of errors of categories A and C above.
Reference 2 describes critical experiments per formed at the Westinghouse Reactor Evaluation Center and measurements taken on two Westinghouse plants with incore systems of the same type used in the VEGP. The report concludes that the uncertainty associated with F Q (heat flux) is 4.58 percent at the 95-percent confidence level with only 5 percent of the measurements greater than the inferred value. This is the equivalent of a 1.645 limit on a normal distribution and is the uncertainty to be associated with a full core flux map with movable detectors reduced with a reasonable set of input data incorporating the influence of burnup on the radial power
distribution. The uncertainty is usually rounded up to 5 percent.
In comparing measured power distributions (or detector currents) with calculations for the same operating conditions, it is not possible to isolate the detector reproducibility. Thus, a comparison
between measured and predicted power distributions has to include some measurement error.
Such a comparison is given in figure 4.3-24 for one of the maps used in reference 2. Since the
first publication of reference 2, hundreds of maps have been taken on these and other reactors.
The results confirm the adequacy of the 5-percent uncertainty allowance on the calculated F Q, when using 44 detector thimbles. When using 29 and < 44 detector thimbles, the measurement uncertainty is 1.05 + [2.0 {3-T / {14.5)}]
/ 100 , where T equals the number of thimbles (reference 45).
. A similar analysis for the uncertainty in N H F (rod integral power) measurements results in an allowance of 3.60 percent at the equivalent of a 1.645 confidence level. For historical reasons, an 8-percent uncertainty factor is allowed in the nuclear design calculational basis; that is, the predicted rod integrals at full power must not exceed the design N H F less 8 percent.
A measurement in the second cycle of a 121-asse mbly, 12-ft core is compared with a simplified one-dimensional core average axial calculation in figure 4.3-25. This calculation does not give explicit representation to the fuel grids.
VEGP-FSAR-4
4.3-17 REV 19 4/15 The accumulated data on power distributions in actual operation are basically of three types: A. Much of the data is obtained in steady-state operation at constant power in the normal operating configuration. B. Data with unusual values of axial offset are obtained as part of the excore detector calibration exercise performed monthly. C. Special tests have been performed in load follow and other transient xenon conditions which have yielded useful information on power distributions.
These data are presented in detail in reference 8. Figure 4.3-26 contains a summary of
measured values of F Q as a function of axial offset for five plants from that report. 4.3.2.2.8 Testing An extensive series of physics tests has been performed on the first core. These tests and the
criteria for satisfactory results are described in chapter 14. Since not all limiting situations can
be created at BOL, the main purpose of the tests is to provide a check on the calculational
methods used in the predictions for the conditions of the test. Tests performed at the beginning
of each reload cycle are limited to verification of the selected safety-related parameters of the
reload design. 4.3.2.2.9 Monitoring Instrumentation The adequacy of instrument numbers, spatial deployment, required correlations between
readings and peaking factors, calibration, and errors are described in references 2, 6, and 8.
The relevant conclusions are summarized in paragraph 4.3.2.2.7 and subsection 4.4.6.
Provided the limitations given in paragraph 4.3.2.2.6 on rod insertion and flux difference are observed, the excore detector system pr ovides adequate online monitoring of power distributions. Further details of specific limits on the observed rod positions and flux difference
are given in the Core Operating Limits Report and the Technical Specifications, together with a
discussion of the Bases.
Limits for alarms, reactor trip, etc., are given in the Technical Specifications. Descriptions of the systems provided are given in section 7.7. 4.3.2.3 Reactivity Coefficients The kinetic characteristics of the reactor core determine the response of the core to changing
plant conditions or to operator adjustments made during normal operation, as well as the core
response during abnormal or accidental transients. These kinetic characteristics are quantified
in reactivity coefficients. The reactivity coefficients reflect the changes in the neutron
multiplication due to varying plant conditions, such as power, moderator or fuel temperatures, or
pressure or void conditions, although the latter are relatively unimportant in the VEGP reactors.
Since reactivity coefficients change during the life of the core, ranges of coefficients are
employed in transient analysis to determine the response of the plant throughout life. The
results of such simulations and the reactivity coefficients used are presented in chapter 15. The
reactivity coefficients are calculated on a corewide basis by diffusion theory methods. The effect of radial and axial power distribution on core average reactivity coefficients is implicit in
those calculations and is not significant under normal operating conditions. For example, a VEGP-FSAR-4
4.3-18 REV 19 4/15 skewed xenon distribution which results in changing axial offset by 5 percent changes the moderator and Doppler temperature coefficients by less than 0.01 pcm/
°F, and 0.03 pcm/
°F, respectively. An artificially skewed xenon distribution which results in changing the radial N H F by 3 percent changes the moderator and Doppler temperature coefficients by less than 0.03 pcm/°F and 0.001 pcm/
°F, respectively. The spatial effects are accentuated in some transient conditions, for example, in postulated rupture of the main steam line break and rupture of an RCCA mechanism housing described in subsections 15.1.5 and 15.4.8, and are included in
these analyses.
The analytical methods and calculational models used in calculating the reactivity coefficients are given in subsection 4.3.3. These models have been confirmed through extensive testing of
more than 30 cores similar to the plant described herein; results of these tests are discussed in
subsection 4.3.3.
Quantitative information for calculated reactivity coefficients including fuel-Doppler coefficient, moderator coefficients (density, temperature, pressure, and void), and power coefficient, is
given in the following sections. 4.3.2.3.1 Fuel Temperature (Doppler) Coefficient The fuel temperature (Doppler) coefficient is defined as the change in reactivity per degree
change in effective fuel temperature and is prim arily a measure of the Doppler broadening of U-238 and Pu-240 resonance absorption peaks. Doppler broadening of other isotopes is also
considered, but their contribution to the Doppler effect is small. An increase in fuel temperature
increases the effective resonance absorption cross-sections of the fuel and produces a
corresponding reduction in reactivity.
The fuel temperature coefficient is calculated by performing two-group two-dimensional or three-dimensional calculations. Moderator temperature is held constant, and the power level is varied. Spatial variation of fuel temperature is taken into account by calculating the effective
fuel temperature as a function of power density, as discussed in paragraph 4.3.3.1.
A typical Doppler temperature coefficient is shown in figure 4.3-27 as a function of the effective fuel temperature (at BOL and EOL conditions). The effective fuel temperature is lower than the
volume-averaged fuel temperature, since the neutron flux distribution is nonuniform through the
pellet and gives preferential weight to the surface temperature. A typical Doppler-only
contribution to the power coefficient, defined later, is shown in figure 4.3-28 as a function of
relative core power. The integral of the differential curve in figure 4.3-28 is the Doppler
contribution to the power defect and is shown in figure 4.3-29 as a function of relative power.
The Doppler coefficient becomes more negative as a function of life as the Pu-240 content
increases, thus increasing the Pu-240 resonance absorption, but the overall value becomes
less negative, since the fuel temperature changes with burnup, as described in paragraph
4.3.3.1. The upper and lower limits of Doppler coefficient used in accident analyses are given in
chapter 15. 4.3.2.3.2 Moderator Coefficients The moderator coefficient is a measure of the change in reactivity due to a change in specific
coolant parameters, such as density, temperature, pressure, or void. The coefficients so
obtained are moderator density, temperature, pressure, and void coefficients.
VEGP-FSAR-4
4.3-19 REV 19 4/15 4.3.2.3.2.1 Moderator Density and Temperatur e Coefficients. The moderator temperature (density) coefficient is defined as the change in reactivity per degree change in the moderator
temperature. Generally, the effects of the changes in moderator density and the temperature are considered together.
The soluble boron used in the reactor as a means of reactivity control also has an effect on the moderator density coefficient, since the soluble boron density and the water density are
decreased when the coolant temperature rises. A decrease in the soluble boron density
introduces a positive component in the moderator coefficient. If the concentration of soluble
boron is large enough, the net value of the coefficient may be positive. However, with the BA
rods and or IFBAs present, the beginning of life core hot boron concentration is sufficiently low
that the moderator temperature coefficient is nonpositive at 100-percent power, less than or equal to +7.0 pcm/
°F below 70-percent power, and less than a linearly decreasing limit of +7.0 pcm/°F to 0.0 pcm/
°F between 70-percent and 100-percent power, respectively. The effect of control rods is to make the moderator coefficient more negative, since the thermal neutron mean free path, and hence the volume affected by the control rods, increase with an increase in
temperature. With burnup, the moderator coefficient becomes more negative, primarily as a result of boric acid dilution, but also to a significant extent from the effects of the depletion of uranium and
buildup of fission products.
The moderator coefficient is calculated for a range of plant conditions by performing two-group two- or three-dimensional calculations, in wh ich the moderator temperature is varied by about
+/-5°F about each of the mean temperatures and the density is changed consistent with the temperature. The moderator coefficient is shown as a function of core temperature and boron
concentration for unrodded configurations of typical reload cores in figures 4.3-30 through 4.3-
- 32. The temperature range covered is from cold (68
°F) to about 600
°F. The contribution due to Doppler coefficient (because of change in moderator temperature) has been subtracted from these results. Figure 4.3-33 shows the hot, full-power moderator temperature coefficient plotted
as a function of lifetime for the critical boron concentration condition based on a typical reload
boron letdown.
The moderator coefficients presented here are calculated on a corewide basis, since they are used to describe the core behavior in normal and accident situations when the moderator
temperature changes can be considered to affect the entire core. 4.3.2.3.2.2 Moderator Pressure Coefficient. The moderator pressure coefficient relates the change in moderator density, resulting from a reactor coolant pressure change, to the corresponding effect on neutron production. This coefficient is of much less significance than
the moderator temperature coefficient. A change of 50 psi in pressure has approximately the
same effect on reactivity as a 1/2-degree change in moderator temperature. This coefficient
can be determined from the moderator temperature coefficient by relating change in pressure to
the corresponding change in density. The moderator pressure coefficient is negative over a portion of the moderator temperature range at BOL (-0.004 pcm/psi, BOL) but is always positive at operating conditions and becomes more positive during life (+0.3 pcm/psi, EOL). 4.3.2.3.2.3 Moderator Void Coefficient. The moderator void coefficient relates the change in neutron multiplication to the presence of voids in the moderator. In a PWR, this coefficient is not very significant because of the low void content in the coolant. The core void content is less
than one-half of 1 percent and is due to local or statistical boiling. The void coefficient varies VEGP-FSAR-4
4.3-20 REV 19 4/15 from 50 pcm/percent void at BOL and at low temperatures to -250 pcm/percent void at EOL and
at operating temperatures. The void coefficient at operating temperature becomes more
negative with fuel burnup. 4.3.2.3.3 Power Coefficient The combined effect of moderator temperature and fuel temperature change as the core power
level changes is called the total power coefficient and is expressed in terms of reactivity change
per percent power change. A typical power coe fficient at BOL and EOL conditions is given in figure 4.3-34.
The total power coefficient becomes more negative with burnup, reflecting the combined effect of moderator and fuel temperature coefficients with burnup. The power defect (integral
reactivity effect) at BOL and EOL is given in figure 4.3-35. 4.3.2.3.4 Comparison of Calculated and Experimental Reactivity Coefficients Subsection 4.3.3 describes the comparison of ca lculated and experimental reactivity coefficients
in detail. Based on the data presented there, the accuracy of the current analytical model is:
- +/-0.2 percent for Doppler and power defect.
- +/-2.0 pcm/°F for the moderator coefficient.
Experimental evaluation of the reactivity c oefficients is performed during the reload physics startup tests. 4.3.2.3.5 Reactivity Coefficients Used in Transient Analysis Table 4.3-2 gives the limiting values as well as the best-estimate values for the reactivity
coefficients for the initial cycle. The limiting values were used as design limits in the transient
analysis. The exact values of the coefficient used in the analysis depend on whether the
transient of interest is examined at the BOL or EOL, whether the most negative or the most
positive (least negative) coefficients are appropriate, and whether spatial nonuniformity must be
considered in the analysis. Conservative values of coefficients, considering various aspects of
analysis, are used in the transient analysis. This is described in chapter 15.
The reactivity coefficients shown in figures 4.3-27 through 4.3-35 are typical best-estimate values calculated for a typical reload cycle. Limiting values are chosen to encompass the best-
estimate reactivity coefficients, including the uncertainties given in paragraph 4.3.3.3 over
appropriate operating conditions. The most positive, as well as the most negative, values are
selected to form the design basis range used in the transient analysis. A direct comparison of
the best-estimate and design limit values for the initial cycle shown in table 4.3-2 can be
misleading, since in many instances the most cons ervative combination of reactivity coefficients was used in the transient analysis even though the extreme coefficients assumed may not
simultaneously occur at the conditions assumed in the analysis. The need for a reevaluation of
any accident in a subsequent cycle is contingent upon whether the coefficients for that cycle fall
within the identified range used in the analysis presented in chapter 15 with due allowance for
the calculational uncertainties given in paragraph 4.3.3.3. Control rod requirements are given in
table 4.3-3 for the initial cycle and for a hypothetical equilibrium cycle, since these are markedly
different.
VEGP-FSAR-4
4.3-21 REV 19 4/15 4.3.2.4 Control Requirements To ensure the shutdown margin required by the Technical Specifications or Technical
Requirements Manual, as applicable, under conditions where a cooldown to ambient
temperature is required, concentrated soluble boron is added to the coolant. Boron
concentrations for several core conditions are listed in table 4.3-2 for the initial cycle. For all
core conditions including refueling, the boron concentration is well below the solubility limit. The
RCCAs are employed to bring the reactor to the shutdown condition. The minimum required
shutdown margin is given in the Core Operating Limits Report.
The ability to accomplish the shutdown for hot conditions is demonstrated in table 4.3-3 by comparing the difference between the RCCA reactivity available with an allowance for the worst
stuck rod with that required for control and protection purposes. The shutdown margin includes
an allowance of 10 percent for analytic uncertainties (paragraph 4.3.2.4.9). The 10-percent
uncertainty can be reduced to 7 percent with the use of Ag-In-Cd RCCAs. Use of a 7-percent
uncertainty allowance on RCCA worth is discussed and shown to be acceptable in references
40 and 44. The largest reactivity control requirement appears at the EOL when the moderator
temperature coefficient reaches its peak negative value as reflected in the larger power defect.
The control rods are required to provide sufficient reactivity to account for the power defect from full power to zero power and to provide the required shutdown margin. The reactivity addition
resulting from power reduction consists of c ontributions from Doppler effect, moderator temperature, flux redistribution, and reduction in void content as discussed below. 4.3.2.4.1 Doppler Effect The Doppler effect arises from the broadening of U-238 and Pu-240 resonance cross-sections
with an increase in effective pellet temperature. This effect is most noticeable over the range of
zero power to full power due to the large pellet temperature increase with power generation. 4.3.2.4.2 Variable Average Moderator Temperature When the core is shut down to the hot zero-power condition, the average moderator
temperature changes from the equilibrium full-load value determined by the steam generator and turbine characteristics (steam pressure, heat transfer, tube fouling, etc.) to the equilibrium
no-load value, which is based on the steam generator shell side design pressure. The design change in temperature is conservatively increased by 6
°F to account for the control dead band and measurement errors.
When the moderator coefficient is negative (except early in core life, when higher boron concentrations may result in a moderator temperature coefficient as high as +7.0 pcm/
°F for powers below 70 percent and decreasing linearly to a value of 0.0 pcm/
°F at 100-percent power), there is a reactivity addition with power reduction. The moderator coefficient becomes more negative as the fuel depletes because the boron concentration is reduced. This effect is
the major contributor to the increased requirement at EOL. 4.3.2.4.3 Redistribution During full-power operation, the coolant density decreases with core height, and this, together
with partial insertion of control rods, results in less fuel depletion near the top of the core. Under VEGP-FSAR-4
4.3-22 REV 19 4/15 steady-state conditions, the relative power distribution will be slightly asymmetric toward the
bottom of the core. On the other hand, at hot zero-power conditions, the coolant density is
uniform up the core, and there is no flattening due to Doppler effect. The result will be a flux
distribution which at zero power can be skewed toward the top of the core. The reactivity
insertion due to the skewed distribution is calculated with an allowance for effects of xenon
distribution. 4.3.2.4.4 Void Content A small void content in the core is due to nucleate boiling at full power. The void collapse
coincident with power reduction makes a small positive reactivity contribution. 4.3.2.4.5 Rod Insertion Allowance At full power, the control bank is operated within a prescribed band of travel to compensate for
small changes in boron concentration, changes in temperature, and very small changes in the
xenon concentration not compensated for by a change in boron concentration. When the
control bank reaches either limit of this bank, a change in boron concentration is required to compensate for additional reactivity changes. Since the insertion limit is set by a rod travel limit, a conservatively high calculation of the inserted worth is made, which exceeds the normally
inserted reactivity. 4.3.2.4.6 Installed Excess Reactivity for Depletion Excess reactivity of approximately 10 percent (hot) is installed at the beginning of each cycle to provide sufficient reactivity to compensate for fuel depletion and fission product buildup
throughout the cycle. This reactivity is controlled by the addition of soluble boron to the coolant
and by BA. The soluble boron concentration for several core configurations, the unit boron
worth, and the BA worth are given in tables 4.3-1 and 4.3-2 for the initial cycle. Since the
excess reactivity for burnup is controlled by soluble boron and/or BA, it is not included in control
rod requirements. 4.3.2.4.7 Xenon and Samarium Poisoning Changes in xenon and samarium concentrations in the core occur at a sufficiently slow rate, even following rapid power level changes, that the resulting reactivity change can be controlled
by changing the soluble boron concentration. (Also see paragraph 4.3.2.4.16.) 4.3.2.4.8 pH Effects Changes in reactivity due to a change in coolant pH, if any, are sufficiently small in magnitude and occur slowly enough to be controlled by the boron system.
(11)
VEGP-FSAR-4
4.3-23 REV 19 4/15 4.3.2.4.9 Experimental Confirmation Following a normal shutdown, the total core reactivity change during cooldown with a stuck rod
has been measured on a 121-assembly, 10-ft-high core and a 121-assembly, 12-ft-high core. In
each case, the core was allowed to cool down until it reached criticality simulating the steam
line break accident. For the 10-ft core, the total reactivity change associated with the cooldown is overpredicted by about 0.3-percent with respect to the measured result. This represents an error of about 5 percent in the total reactivity change and is about half the uncertainty allowance for this quantity. For the 12-ft core, the difference between the measured and predicted reactivity change is an even smaller 0.2-percent . These measurements and others demonstrate the capability of the methods described in subsection 4.3.3. 4.3.2.4.10 Control Core reactivity is controlled by means of a chemical poison dissolved in the coolant, RCCAs, and burnable absorbers as described below. 4.3.2.4.11 Chemical Shim Boron in solution as boric acid is used to control relatively slow reactivity changes associated with: A. The moderator temperature defect in going from cold shutdown at ambient temperature to the hot operating temperature at zero power. B. The transient xenon and samarium poisoning, such as that following power changes or changes in rod cluster control position. C. The reactivity effects of fissile inventory depletion and buildup of long-life fission products. D. The depletion of the burnable absorbers.
The boron concentrations for various core conditions are presented in table 4.3-2 for the initial cycle. 4.3.2.4.12 Rod Cluster Control Assemblies The number of RCCAs is shown in table 4.3-1. The RCCAs are used for shutdown and control
purposes to offset fast reactivity changes associated with: A. The required shutdown margin in the hot zero power, stuck rods condition. B. The reactivity compensation as a result of an increase in power above hot zero power (power defect, including Doppler and moderator reactivity changes). C. Unprogrammed fluctuations in boron concentration, coolant temperature, or xenon concentration (with rods not exceeding the allowable rod insertion limits). D. Reactivity ramp rates resulting from load changes.
The allowed control bank reactivity insertion is limited at full power to maintain shutdown capability. As the power level is reduced, control rod reactivity requirements are also reduced, and more rod insertion is allowed. The control bank position is monitored, and the operator is VEGP-FSAR-4
4.3-24 REV 19 4/15 notified by an alarm if the limit is approached. The determination of the insertion limit uses
conservative xenon distributions and axial power shapes. In addition, the RCCA withdrawal
pattern determined from the analyses is used in determining power distribution factors and in
determining the maximum worth of an inserted RCCA ejection accident. For further discussion, refer to the Technical Specifications on rod insertion limits.
Power distribution, rod ejection, and rod misalignment analyses are based on the arrangement of the shutdown and control groups of the RCCAs shown in figure 4.3-36. All shutdown RCCAs
are withdrawn before withdrawal of the control banks is initiated.
In going from 0- to 100-percent power, control banks A, B, C, and D are withdrawn sequentially.
The limits of rod insertion and further discussion on the basis for rod insertion limits are
provided in the Technical Specifications. 4.3.2.4.13 Reactor Coolant Temperature Reactor coolant (or moderator) temperature control has added flexibility in reactivity control of
the Westinghouse pressurized water reactor. This feature takes advantage of the negative
moderator temperature coefficient inherent in a PWR to:
- Maximize return to power capabilities.
- Provide +/-5-percent power load regulation capabilities.
- Extend the time in cycle life to which daily load follow operations can be accomplished.
Reactor coolant temperature control supplements the dilution capability of the plant by lowering
the reactor coolant temperature to supply pos itive reactivity through the negative moderator
coefficient of the reactor. After the transient is over, the system returns the reactor coolant
temperature to the programmed value.
Moderator temperature control of reactivity, like soluble boron control, has the advantage of not significantly affecting the core power distribution. However, unlike boron control, temperature
control can be rapid enough to achieve reactor power change rates of 5 percent/min. 4.3.2.4.14 Burnable Absorbers The discrete BA rods or integral fuel burnable absorbers may be used to provide partial control
of the excess reactivity available during any fuel cycle. In doing so, these rods control peaking factors and prevent the moderator temperature coefficient from being positive at full power conditions, from exceeding +7.0 pcm/
°F at power levels below 70-percent power, and from exceeding a limit which decreases linearly from 7.0 pcm/
°F to 0.0 pcm/
°F between 70-percent power and 100-percent power, respectively. T hey perform this function by reducing the requirement for soluble boron in the moderator at the beginning of the cycle, as described
previously. For purposes of illustration, ty pical reload cycle burnable absorber patterns are
shown in figure 4.3-5, and the arrangements within an assembly are displayed in figure 4.3-4.
The boron in the rods is depleted with burnup but at a sufficiently slow rate so that the peaking
factor limits are not exceeded and the resulting critical concentration of soluble boron is such
that the moderator temperature coefficient remains within the limits stated above at all times for
power operating conditions.
VEGP-FSAR-4
4.3-25 REV 19 4/15 4.3.2.4.15 Peak Xenon Startup Compensation for the peak xenon buildup may be accomplished using the boron control
system. Startup from the peak xenon condition is accomplished with a combination of rod
motion and boron dilution. The boron dilution may be made at any time, including during the
shutdown period, provided the shutdown margin is maintained. 4.3.2.4.16 Load Follow Control and Xenon Control During load follow maneuvers, power changes are accomplished using control rod motion and
dilution or boration by the boron system as requir ed. Control rod motion is limited by the control rod insertion limits on full-length rods, as provided in the Technical Specifications and Core
Operating Limits Report and discussed in paragraph 4.3.2.4.12. The power distribution is
maintained within acceptable limits through location of the full-length rod bank. Reactivity
changes due to the changing xenon concentration can be controlled by rod motion and/or
changes in the soluble boron concentration.
Late in cycle life, extended load follow capability is obtained by augmenting the limited boron dilution capability at low soluble boron concentrations by temporary moderator temperature
reductions.
Rapid power increases (5 percent/min) from part power during load follow operation are accomplished with a combination of rod motion, moderator temperature reduction, and boron
dilution. Compensation for the rapid power increase is accomplished initially by a combination
of rod withdrawal and moderator temperature reduction. As the slower boron dilution takes
effect after the initial rapid power increase, the moderator temperature is returned to the
programmed value. 4.3.2.4.17 Burnup Control of the excess reactivity for burnup is accomplished using soluble boron and/or burnable
absorbers. The boron concentration must be limited during operating conditions to ensure that
the moderator temperature coefficient is nonpositive at full power, less than or equal to +7.0 pcm/°F below 70-percent power, and less than or equal to a limit which varies linearly from +7.0 pcm/°F to 0.0 pcm/
°F between 70-percent power and 100-percent power, respectively.
Sufficient burnable absorbers are installed at the beginning of a cycle to give the desired cycle lifetime, and limit the boron concentration to a ssure achieving the moderator temperature
coefficient limits discussed above. The practical minimum boron concentration is in the range of
0 to 10 ppm. 4.3.2.5 Control Rod Patterns and Reactivity Worths The RCCAs are designated by function as the control groups and the shutdown groups. The
terms "group" and "bank" are used synonymously th roughout this report to describe a particular grouping of control assemblies. The RCCA pattern, which is not expected to change during the
life of the plant, is displayed in figure 4.3-36. The control banks are labeled A, B, C, and D; and
the shutdown banks are labeled SA, SB, SC, SD, and SE. Each bank, although operated and
controlled as a unit, is composed of two subgroups. The axial position of the RCCAs may be
controlled manually or automatically. The RCCAs are all dropped into the core following
actuation of reactor trip signals.
VEGP-FSAR-4
4.3-26 REV 19 4/15 Two criteria have been employed for selection of t he control groups. First, the total reactivity worth must be adequate to meet the requirements specified in table 4.3-3. 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 requirements are met.
Analyses indicate that the first requirement can be met either by a single group or by two or more banks whose total worth equals at least the required amount. The axial power shape would be more peaked following movement of a single group of rods worth 3- to 4-percent Therefore, four banks (designated A, B, C, and D in figure 4.3-36), each worth approximately 1
percent , have been selected. Typical control bank worths for the initial cycle are shown in table 4.3-2.
The position of control banks for criticality under any reactor condition is determined by the concentration of boron in the coolant. On an approach to criticality, boron is adjusted to ensure
that criticality will be achieved with control rods above the insertion limits and other
considerations. (See the Technical Specifications and the Core Operating Limits Report.) Early in the cycle, there may also be a withdrawal limit at low power to maintain the moderator
temperature coefficient that is less than or equal to the limit value for that power level.
Ejected rod worths for several different conditions are given in subsection 15.4.8.
Allowable deviations due to misaligned control rods are discussed in the Technical Specifications.
A representative differential rod worth calculation for two banks of control rods withdrawn simultaneously (rod withdrawal accident) is given in figure 4.3-37.
Calculation of control rod reactivity worth versus time following reactor trip involves both control rod velocity and differential reactivity worth. The rod position versus time of travel after rod
release assumed is given in figure 4.3-38. For nuclear design purposes, the reactivity worth versus rod position is calculated by a series of steady-state calculations at various control
positions, assuming all rods out of the core as the initial position in order to minimize the initial
reactivity insertion rate. Also, to be conservative, the rod of highest worth is assumed stuck out
of the core, and the flux distribution (and thus r eactivity importance) is assumed to be skewed to the bottom of the core. The result of these calculations is shown in figure 4.3-39.
The shutdown groups provide additional negative reactivity to ensure an adequate shutdown margin. Shutdown margin is defined as the amount by which the core would be subcritical at
hot shutdown if all RCCAs were tripped, but assuming that the highest worth assembly
remained fully withdrawn and no changes in xenon or boron took place. The loss of control rod
worth due to the depletion of the absorber material is negligible, since only bank D may be in
the core under normal operating conditions (near full power).
The values given in table 4.3-3 show that the available reactivity in withdrawn RCCAs provides the design bases minimum shutdown margin, allowing for the highest worth cluster to be at its
fully withdrawn position. An allowance for the uncertainty in the calculated worth of N-1 rods is
made before determination of the shutdown margin. 4.3.2.6 Criticality of the Reactor During Refueling The basis for maintaining the reactor subcritical during refueling is presented in paragraph
4.3.1.5, and a discussion of how control requirements are met is given in paragraphs 4.3.2.4
and 4.3.2.5.
VEGP-FSAR-4
4.3-27 REV 19 4/15 4.3.2.6.1 Fuel Storage Criticality The following information describes the design criteria and analysis techniques for storage of
fuel in the spent fuel racks. The Unit 1 and Unit 2 spent fuel racks were analyzed in accordance
with the methodology contained in reference 42 which credits the soluble boron contained in the
spent fuel pool water. The analyses have been reviewed by the NRC and documented in their
safety evaluation for Technical Specification Amendments 139 and 118 for Unit 1 and Unit 2, respectively (reference 43). 4.3.2.6.1.1 Storage of Fuel Assemblies. - The fuel storage pools contain high-density fuel storage racks which are designed to store Westinghouse 17x17 fuel assemblies with a maximum enrichment of up to 5.0 wt% U-235.
One factor which prevents criticality of fuel assemblies in a fuel storage rack is the design of the rack which limits fuel assembly interaction by fixing the minimum separation between fuel
assemblies. Other factors include limits on the enrichment, burnup, fixed absorbers, burnable
absorbers, decay time, soluble boron, and placement of assemblies within the fuel storage
racks. The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95-percent probability at a 95-percent confidence level (95/95) that the effective
neutron multiplication factor, Keff , of the fuel rack array will be less than or equal to 0.95.
The spent fuel racks have been analyzed such that Keff remains less than 1.0 including uncertainties and tolerances on a 95/95 basis without the presence of any soluble boron in the
storage pool (No Soluble Boron 95/95 Conditions) as defined in reference 42. Soluble boron credit is used to provide safety margin by maintaining Keff 0.95 including uncertainties, tolerances, and accident conditions in the presence of spent fuel pool soluble boron. The amount of soluble boron required for the allowable storage conditions excluding accidents is
511 ppm (Unit 1), 394 ppm (Unit 2). The presence of Boral panels in the Unit 1 spent fuel racks
is credited in the Unit 1 criticality analyses. The presence of any Boraflex panels in the Unit 2
spent fuel racks is ignored. The criticality analyses include a calculational bias, mechanical
uncertainties, and consideration of 0.05 wt% enrichment variability.
The analysis methodology employs: (1) SCALE-PC, a personal computer version of the SCALE-4.3 code system, with the updated SCALE-4.3 version of the 44 group ENDF/B-V neutron cross
section library, and (2) the two-dimensional integral transport code DIT with an ENDF/B-VI
neutron cross section library.
SCALE-PC was used for calculations involving infinite arrays for the "2-out-of-4", "3-out-of-4", "All-Cell," and "3x3" fuel assembly storage configurations. In addition, it was employed in a full
pool representation of the storage racks to evaluate soluble boron worth and postulated
accidents.
SCALE-PC, used in both the benchmarking and the fuel assembly storage configurations, includes the control module CSAS25 and the fo llowing functional modules: BONAMI, NITAWL-Il, and KENO V.a.
The DIT code is used for simulation of in-reactor fuel assembly depletion. KENO V.a was used in the calculation of biases and uncertainties.
Models were made for each storage configuration as well as for the entire pools. Each configuration was modeled as an infinitely repeating pattern in the X-Y plane. A water reflector VEGP-FSAR-4
4.3-28 REV 19 4/15 was modeled above and below the spent fuel storage cells to account for axial reactivity effects.
The KENO model for the entire pool modeled each individual storage rack module.
KENO was used to calculate Keff for the various storage configurations considered as well as for the entire pool. For demonstrating that Keff remains below unity for zero soluble boron, SNC chose to apply an acceptance criterion of 0.995. For the storage configurations considered, the target value of Keff for these calculations was selected to be less than 0.995 by an amount sufficient to cover the magnitude of the analytical biases and uncertainties. KENO was also used to determine burnup and IFBA versus enrichment requirements to meet the target Keff for configurations that credit burnup or IFBA.
Calculations were performed for the entire pool with various fuel storage configurations to demonstrate that the Keff for the entire pool remains below 0.995 with zero soluble boron.
KENO was used to determine the soluble boron requirements for nonaccident and accident
conditions to ensure that Keff remains less than or equal to 0.95. The details of the modeling and analyses are described in reference 42.
For the analyses described in reference 43, three Westinghouse fuel assembly designs were considered: 17x17 STANDARD (STD), 17x17 Opti mized Fuel Assembly (OFA), and the 17x17 Robust Fuel Assembly (RFA). The most reactive design for each storage configuration was
used to bound the other fuel designs for both fresh and depleted fuel.
The spent fuel racks have been analyzed to allow storage of Westinghouse 17x17 fuel assemblies with nominal initial enrichments up to 5.00 wt% U-235 in storage cell locations using
credit for checkerboard configurations, burnup credit, and integral fuel burnable absorber (IFBA)
credit. Although the analysis covers nominal initial enrichments up to 5.0 wt% U-235, the
Technical Specifications allows only 5.0 wt% U-235 maximum. The presence of Boral panels in
the Unit 1 spent fuel racks is credited in the Unit 1 criticality analyses. This analysis does not
take any credit for the presence of the spent fuel rack Boraflex poison panels in the Unit 2 spent
fuel racks. The following storage configurations and enrichment limits resulted from this
analysis:
Unit 1 New or partially spent fuel assemblies with a combination of burnup and initial nominal enrichment in the "acceptable burnup domain" of figures 4.3-46 or satisfying a minimum IFBA
requirement as shown in figure 4.3-54 may be allowed unrestricted storage in the Unit 1 fuel
storage pool.
New or partially spent fuel assemblies with a maximum initial enrichment of 5.0 weight percent U-235 may be stored in the Unit 1 fuel storage pool in a 3-out-of-4 checkerboard storage
configuration as shown in figure 4.3-48.
Interfaces between storage configurations in the Unit 1 fuel storage pool shall be in compliance with figure 4.3-50. "A" assemblies are new or partially spent fuel assemblies with a combination of burnup and initial nominal enrichment in the "acceptable burnup domain" of figure 4.3-46, or
which satisfy a minimum IFBA requirement as shown in figure 4.3-54. "B" assemblies are
assemblies with initial enrichments up to a maximum of 5.0 weight percent U-235.
Unit 2 New or partially spent fuel assemblies with a combination of burnup and initial nominal
enrichment in the "acceptable burnup domain" of figure 4.3-47 may be allowed unrestricted
storage in the Unit 2 fuel storage pool.
VEGP-FSAR-4
4.3-29 REV 19 4/15 New or partially spent fuel assemblies with a combination of burnup and initial nominal
enrichment in the "acceptable burnup domain" of figure 4.3-55 may be stored in the Unit 2 fuel
storage pool in a 3-out-of-4 checkerboard storage configuration as shown in figure 4.3-48.
New or partially spent fuel assemblies with a maximum initial enrichment of 5.0 weight percent U-235 may be stored in the Unit 2 fuel storage pool in a 2-out-of-4 checkerboard storage
configuration as shown in figure 4.3-48.
New or partially spent fuel assemblies with a combination of burnup, decay time, and initial nominal enrichment in the "acceptable burnup domain" of figure 4.3-57 may be stored in the
Unit 2 fuel storage pool as "low enrichment" fuel assemblies in the 3x3 checkerboard storage
configuration as shown in figure 4.3-49. New or partially spent fuel assemblies with initial
nominal enrichments less than or equal to 3.20 weight percent U-235 or which satisfy a
minimum IFBA requirement as shown in figure 4.3-56 for higher initial enrichments may be
stored in the Unit 2 fuel storage pool as "high enrichment" fuel assemblies in the 3x3
checkerboard storage configuration as shown in figure 4.3-49.
Interfaces between storage configurations in the Unit 2 fuel storage pool shall be in compliance with figures 4.3-50, 4.3-51, 4.3-52, and 4.3-53. "A" assemblies are new or partially spent fuel
assemblies with a combination of burnup and initial nominal enrichment in the "acceptable
burnup domain" of figure 4.3-47. "B" assemblies are new or partially spent fuel assemblies with
a combination of burnup and initial nominal enrichment in the "acceptable burnup domain" of
figure 4.3-55. "C" assemblies are assemblies with initial enrichments up to a maximum of 5.0
weight percent U-235. "L" assemblies are new or partially spent fuel assemblies with a
combination of burnup, decay time, and initial nominal enrichment in the "acceptable burnup
domain" of figure 4.3-57. "H" assemblies are new or partially spent fuel assemblies with initial
nominal enrichments less than or equal to 3.20 weight percent U-235 or which satisfy a
minimum IFBA requirement as shown in figure 4.3-56 for higher initial enrichments.
The analytical methods conform with ANSI N18.2-1973, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," Section 5.7, Fuel Handling System; ANSI
57.2-1983, "Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear
Power Stations," Section 6.4.2; ANSI N16.9-1975, "Validation of Calculational Methods for
Nuclear Criticality Safety"; and the NRC Standard Review Plan, Section 9.1.2, "Spent Fuel
Storage." 4.3.2.6.1.2 Credit for Soluble Boron - Reference 42 describes how credit for fuel storage pool soluble boron is used under normal storage configuration conditions. The storage configuration is defined using Keff calculations to ensure that the Keff will be less than 1.0 with no soluble boron under normal storage conditions including tolerances and uncertainties. Soluble
boron credit is then used to maintain Keff less than or equal to 0.95. The analyses assumed 19.9% of the boron atoms have atomic weight 10 (B-10). However, to account for the effects of
variations in the natural abundance of B-10, the calculated boron concentrations, as well as the
concentrations for accidents, were adjusted to correspond to a B-10 fraction of 19.7%. The Unit
1 pool requires 511 ppm and the Unit 2 pool requires 394 ppm to maintain Keff less than or equal to 0.95 for all allowed combinations of storage configurations, enrichments, and burnups.
The soluble boron concentration required to maintain Keff less than or equal to 0.95 under accident conditions is determined by first surveying all possible events which increase the Keff value of the spent fuel pool. The accident event which produced the largest increase in spent fuel pool Keff value is employed to determine the required soluble boron concentration necessary to mitigate this and all less severe accident events. The list of accident cases
considered includes:
VEGP-FSAR-4
4.3-30 REV 19 4/15
- Dropped fresh fuel assembly on top of the storage racks,
- Misloaded fresh fuel assembly into an incorrect storage rack location,
- Misloaded fresh fuel assembly outside of the storage racks,
- Reduction in rack module-to-module water gap due to seismic event,
- Spent fuel pool temperature outside the normal range of 50 °F to 185 °F.
From a criticality standpoint, a dropped assembly accident occurs when a fuel assembly in its
most reactive condition is dropped onto the storage racks. The rack structure from a criticality
standpoint is not excessively deformed. Previous accident analysis with unborated water
showed that the dropped assembly which comes to rest horizontally on top of the rack has
sufficient water separating it from the active fuel height of stored assemblies to preclude
neutronic interaction. For the borated water condition, the interaction is even less since the
water contains boron, an additional thermal neutron absorber.
Several fuel mishandling events were simulated with KENO V.a to assess the possible increase in the Keff value of the spent fuel pools. The fuel mishandling events all assumed that a fresh Westinghouse OFA fuel assembly enriched to 5.0 weight percent U-235 (and no burnable
poisons) was misloaded into the described area of the spent fuel pool. These cases were
simulated with the KENO V.a model for the entire spent fuel pool.
For Unit 1, the fuel mishandling event which produced the largest increase in spent fuel pool Keff value is the misloading of a fresh fuel assembly between a 3-out-of-4 fuel assembly storage configuration and the pool wall. The additional soluble boron concentration necessary to
mitigate this and all less severe accident events is 340 ppm.
For Unit 2, the fuel mishandling event which produced the largest increase in spent fuel pool Keff value is the misloading of a fresh fuel assembly in an incorrect storage rack location for the 2-out-of-4 configuration. The additional soluble boron concentration necessary to mitigate this
and all less severe accident events is 704 ppm.
For the accident due to a seismic event, the gap between rack modules was reduced to zero.
For both Units 1 and 2, the reactivity increase is an order of magnitude less than that for the fuel
mishandling events.
An increase in the temperature of the water passing through the stored fuel assemblies causes a decrease in water density which results in an addition of negative reactivity for flux trap design racks such as the Unit 1 racks. However, since Boraflex is not considered to be present for the
Unit 2 racks and the fuel storage pool water has a high concentration of boron, a density
decrease causes a positive reactivity addition. The reactivity effects of a temperature range
from 32° F to 240° F were evaluated. This bounds the temperature range assumed in the
criticality analyses (50° F to 185° F). The increase in reactivity due to the decrease in temperature below 50° F is bounded by the mispla cement of a fuel assembly between the rack and pool walls for the Unit 1 racks. The increase in reactivity due to the increase in temperature
is bounded by the misload accident for the Unit 2 racks.
Including the effects of accidents, the maximum required boron concentration to maintain Keff 0.95 is 851 ppm for Unit 1 and 1098 ppm for Unit 2, which is well below the minimum Technical Specification limit of 2000 ppm.
The results of all analyses including accidents verify that there is a 95-percent probability at a 95-percent confidence level (including uncertainties) that Keff of the spent fuel storage racks will be less than 0.95 when flooded with water borated to a concentration of 851 ppm for Unit 1 and VEGP-FSAR-4
4.3-31 REV 19 4/15 1098 ppm for Unit 2. Normal nominal boron concentration in the fuel storage pool is about 2400
ppm which is greater than that used in the evaluation of pool dilution.
A spent fuel pool boron dilution evaluation was performed to determine the volume necessary to dilute the spent fuel pool from 2000 ppm to 600 ppm (the boron concentration required to maintain Keff 0.95 in the absence of an unrelated accident). The boron dilution evaluation determined that approximately 465,000 gal of water would be required to dilute the spent fuel pool from 2000 ppm to 600 ppm. A dilution event that would result in this large volume of water
would require the transfer of a large quantity of water from the dilution source and significant
increase in the spent fuel pool level, which would ultimately overflow the pool. This large volume
of water would be readily detected and terminated by plant personnel. A spent fuel pool dilution
event of this magnitude is not a credible event. 4.3.2.6.2 Rod Storage Canister The fuel rod storage canister (FRSC) must be treated as if it were an assembly with enrichment
and burnup of the rod in the canister with the most limiting combination of enrichment and
burnup. Storage of the FRSC is based upon meeting the requirements for each of the allowable
storage configurations. 4.3.2.7 Stability 4.3.2.7.1 Introduction The stability of the PWR cores against xenon-induced spatial oscillations and the control of
such transients are discussed extensively in references 6, 18, 19, and 20. A summary of these
reports is given in the following discussion, and the design bases are given in paragraph
4.3.1.6.
In a large reactor core, xenon-induced oscillations can take place with no corresponding change in the total power of the core. The oscillation may be caused by a power shift in the core which
occurs rapidly by comparison with the xenon-iodine time constants. Such a power shift occurs
in the axial direction when a plant load change is made by control rod motion and results in a
change in the moderator density and fuel temperature distributions. Such a power shift could
occur in the diametral plane of the core as a result of abnormal control action.
Due to the negative power coefficient of reactivity, PWR cores are inherently stable to oscillations in total power. Protection against total power instabilities is provided by the control
and protection system, as described in section 7.7. Hence, the discussion on the core stability
will be limited here to xenon-induced spatial oscillations. 4.3.2.7.2 Stability Index Power distributions, either in the axial direction or in the X-Y plane, can undergo oscillations due
to perturbations introduced in the equilibrium distributions without changing the total core power.
The overtones in the current PWRs and the stability of the core against xenon-induced
oscillations can be determined in terms of the eigenvalues of the first flux overtones. Writing the eigenvalue of the first flux harmonic as:
VEGP-FSAR-4
4.3-32 REV 19 4/15 ic b== (1) Then b is defined as the stability index and T = 2/c as the oscillation period of the first harmonic. The time dependence of the first harmonic ø in the power distribution can now be represented as:
ctcosaeeA)t(btt== (2) where A and a are constants. The stability index can also be obtained approximately by:
A A ln T l = b n1n+ where A n and A n+1 are the successive peak amplitudes of the oscillation and T is the time period between the successive peaks. 4.3.2.7.3 Prediction of the Core Stability The stability of the core described herein (i.e., with 17 x 17 fuel assemblies) against xenon-induced spatial oscillations is expected to be equal to or better than that of earlier designs for
cores of similar size. The prediction is based on a comparison of the parameters which are
significant in determining the stability of the core against the xenon-induced oscillations, namely: A. The overall core size is unchanged and spatial power distributions will be similar. B. The moderator temperature coefficient is expected to be similar or slightly more negative. C. The Doppler coefficient of reactivity is expected to be equal or slightly more negative at full power. Analysis of both the axial and X-Y xenon transient tests, discussed in paragraph 4.3.2.7.5, shows that the calculational model is adequate for the prediction of core stability. 4.3.2.7.4 Stability Measurements 4.3.2.7.4.1 Axial Measurements. Two axia l xenon transient tests conducted in a PWR with a core height of 12 ft and 121 fuel assemblies are reported in reference 21 and will be
briefly discussed here. The tests were performed at approximately 10 percent and 50 percent
of cycle life.
Both a free-running oscillation test and a controlled test were performed during the first test.
The second test at midcycle consisted of a free-running oscillation test only. In each of the free-
running oscillation tests, a perturbation was introduced to the equilibrium power distribution
through an impulse motion of the control bank D and the subsequent oscillation period was
monitored. In the controlled test conducted early in the cycle, the part-length rods were used to
follow the oscillations to maintain an axial offset within the prescribed limits. The axial offset of
power was obtained from the excore ion chamber readings (which had been calibrated against
the incore flux maps) as a function of time for both free-running tests, as shown in figure 4.3-40.
The total core power was maintained constant during these spatial xenon tests, and the stability index and the oscillation period were obtained from a least-square fit of the axial offset data in VEGP-FSAR-4
4.3-33 REV 19 4/15 the form of equation 2. The axial offset of power is the quantity that properly represents the
axial stability in the sense that it essentially eliminates any contribution from even-order harmonics, including the fundamental mode. The conclusions of the tests are: A. The core was stable against induced axial xenon transients, at the core average burnups of both 1550 MWd/tonne uranium and 7700 MWd/tonne uranium. The
measured stability indices are -0.041 h
-1 for the first test (curve 1 of figure 4.3-40) and -0.014 h
-1 for the second test (curve 2 of figure 4.3-40). The corresponding oscillation periods are 32.4 and 27.2 h, respectively. B. The reactor core becomes less stable as fuel burnup progresses, and the axial stability index is essentially zero at 12,000 MWd/tonne uranium. However, the
movable control rod systems can control axial oscillations, as described in
paragraph 4.3.2.7.6.1. 4.3.2.7.4.2 Measurements in the X-Y Plane. Two X-Y xenon oscillation tests were performed at a PWR plant with a core height of 12 ft and 157 fuel assemblies. The first test was
conducted at a core average burnup of 1540 MWd/tonne uranium and the second at a core average burnup of 12,900 MWd/tonne uranium. Both of the X-Y xenon tests show that the core was stable in the X-Y plane at both burnups. The second test shows that the core became
more stable as the fuel burnup increased, and all Westinghouse PWRs with 121 and 157
assemblies are expected to be stable throughout their burnup cycles. The results of these tests
are applicable to the 193-assembly VEGP core as discussed in paragraph 4.3.2.7.3. In each of the two X-Y tests, a perturbation was introduced to the equilibrium power distribution through an impulse motion of one rod cluster control unit located along the diagonal axis.
Following the perturbation, the uncontrolled oscillation was monitored, using the movable
detector and thermocouple system and the excore power range detectors. The quadrant tilt difference (QTD) is the quantity that properly represents the diametral oscillation in the X-Y
plane of the reactor core in that the differences of the quadrant average powers over two
symmetrically opposite quadrants essentially eliminates the contribution to the oscillation from
the azimuthal mode. The QTD data were fitted in the form of equation 2 through a least-square
method. A stability index of -0.076 h
-1 with a period of 29.6 h was obtained from the thermocouple data shown in figure 4.3-41. It was observed in the second X-Y xenon test that the PWR core with 157 fuel assemblies had become more stable due to an increased fuel depletion, and the stability index was not
determined. 4.3.2.7.5 Comparison of Calculations with Measurements The analysis of the axial xenon transient tests was performed in an axial slab geometry, using a
flux synthesis technique. The direct simulation of the axial offset data was carried out using the
PANDA code.
(22) The analysis of the X-Y xenon transient tests was performed in an X-Y geometry, using a modified TURTLE code.
(10) Both the PANDA and TURTLE codes solve the two-group, time-dependent neutron diffusion equation with time-dependent xenon and iodine
concentrations. The fuel temperature and moderator density feedback is limited to a steady-state model. All the X-Y calculations were performed in an average enthalpy plane.
The basic nuclear cross-sections used in this study were generated from a unit cell depletion program which has evolved from the codes LEOPARD (23) and CINDER.
(24) The detailed experimental data during the tests, including the reactor power level, the enthalpy rise, and the VEGP-FSAR-4
4.3-34 REV 19 4/15 impulse motion of the control rod assembly, as well as the plant follow burnup data, were
closely simulated in the study.
The results of the stability calculation for the axial tests are compared with the experimental data in table 4.3-5. The calculations show conservative results for both of the axial tests with a
margin of approximately -0.01 h
-1 in the stability index. An analytical simulation of the first X-Y xenon oscillation test shows a calculated stability index
of -0.081 h
-1 , in close agreement with the measured value of -0.076 h
-1. As indicated earlier, the second X-Y xenon test showed that the core had become more stable compared to the first test, and no evaluation of the stability index was attempted. This increase in the core stability in the X-Y plane due to increased fuel burnup is due mainly to the increased magnitude of the
negative moderator temperature coefficient.
Previous studies of the physics of xenon oscillations, including three-dimensional analysis, are reported in a series of topical reports.
(18)(19)(20)
A more detailed description of the experimental results and analysis of the axial and X-Y xenon transient tests is presented in reference 21 and
section 1 of reference 25. 4.3.2.7.6 Stability Control and Protection The excore detector system is utilized to provide indications of xenon-induced spatial oscillations. The readings from the excore detectors are available to the operator and also form
part of the protection system. 4.3.2.7.6.1 Axial Power Distribution. For maintenance of proper axial power distributions, the operator is instructed to maintain an axial offset within a prescribed operating band, based on the excore detector readings. Should the axial offset be permitted to move far enough
outside this band, the protection limit will be r eached, and the power will be manually reduced.
As fuel burnup progresses, 12-ft PWR cores become less stable to axial xenon oscillations.
However, free xenon oscillations are not allowed to occur, except for special tests. The full-
length control rod banks are sufficient to dampen and control any axial xenon oscillations
present. Should the axial offset be inadvertently permitted to move far enough outside the
allowed band due to an axial xenon oscillation or fo r any other reason, the protection limit on axial offset will be reached and the power will be manually reduced. 4.3.2.7.6.2 Radial Power Distribution. The core described herein is calculated to be stable against X-Y xenon-induced oscillations at all times in life. The X-Y stability of large PWRs has been further verified as part of the startup physics test program for PWR cores with 193 fuel assemblies. The measured X-Y stability of the cores with
157 and 193 assemblies was in close agreement with the calculated stability, as discussed in paragraphs 4.3.2.7.4 and 4.3.2.7.5. In the unlikely event that X-Y oscillations occur, backup
actions are possible and would be implemented, if necessary, to increase the natural stability of
the core. This is based on the fact that several actions could be taken to reduce the moderator temperature coefficient, which would increase the stability of the core in the X-Y plane. Provisions for protection against nonsymmetric perturbations in the X-Y power distribution that could result from equipment malfunctions are m ade in the protection system design. This includes control rod drop, rod misalignment, and asymmetric loss of coolant flow.
VEGP-FSAR-4
4.3-35 REV 19 4/15 A more detailed discussion of the power distribution control in PWR cores is presented in
references 6 and 7. 4.3.2.8 Vessel Irradiation A brief review of the methods and analyses used in the determination of neutron and gamma
ray flux attenuation between the core and the pressure vessel is given below. A more complete
discussion on the pressure vessel irradiation and surveillance program is given in section 5.3.
The materials that serve to attenuate neutrons originating in the core and gamma rays from both the core and structural components consist of the core baffle, core barrel, neutron panels, and
associated water annuli, all of which are within the region between the core and the pressure vessel.
In general, few group neutron diffusion theory codes are used to determine fission power density distributions within the active core, and the accuracy of these analyses is verified by
incore measurements on operating reactors.
Region and rodwise power-sharing information from the core calculations is then used as source information in two-dimensional S N transport calculations which compute the flux distributions throughout the reactor.
The neutron flux distribution and spectrum in the various structural components vary significantly from the core to the pressure vessel. Representative values of the neutron flux
distribution and spectrum are presented in table 4.3-6. The values listed are based on time-
averaged equilibrium cycle reactor core param eters and power distributions and thus are suitable for long-term neutron velocity time projections and for correlation with radiation damage
estimates.
As discussed in section 5.3, the irradiation surveillance program utilizes actual test samples to verify the accuracy of the calculated fluxes at the vessel. 4.3.3 ANALYTICAL METHODS Calculations required in nuclear design consist of three distinct types, which are performed in
sequence:
- Determination of effective fuel temperatures.
- Generation of microscopic few-group parameters.
- Space-dependent, few-group diffusion calculations.
These calculations are carried out by computer codes which can be executed individually.
However, at Westinghouse most of the codes required have been linked to form an automated
design sequence which minimizes design time, avoids errors in transcription of data, and
standardizes the design methods. 4.3.3.1 Fuel Temperature (Doppler) Calculations Temperatures vary radially within the fuel rod, depending on the heat generation rate in the
pellet; the conductivity of the materials in the pellet, gap, and clad; and the temperature of the
coolant.
The fuel temperatures for use in most nuclear design Doppler calculations are obtained from a simplified version of the Westinghouse fuel rod design model described in paragraph 4.2.1.3, VEGP-FSAR-4
4.3-36 REV 19 4/15 which considers the effect of radial variation of pellet conductivity, expansion coefficient and
heat generation rate, elastic deflection of the clad, and a gap conductance which depends on
the initial fill gas, the hot open gap dimension, and the fraction of the pellet over which the gap
is closed. The fraction of the gap assumed closed represents an empirical adjustment used to
produce close agreement with observed reactivity data at BOL. Further gap closure occurs with
burnup and accounts for the decrease in Doppler defect with burnup which has been observed
in operating plants.
Radial power distributions in the pellet as a function of burnup are obtained from LASER (26) calculations.
The effective U-238 temperature for resonance absorption is obtained from the radial temperature distribution by applying a radially dependent weighting function. The weighting
function was determined from REPAD (27) Monte Carlo calculations of resonance escape probabilities in several steady-state and transient temperature distributions. In each case, a flat
pellet temperature was determined which produced the same resonance escape probability as
the actual distribution. The weighting function was empirically determined from these results.
The effective Pu-240 temperature for resonance absorption is determined by a convolution of the radial distribution of Pu-240 densities from LASER burnup calculations and the radial
weighting function. The resulting temperature is burnup dependent, but the difference between
U-238 and Pu-240 temperatures, in terms of reactivity effects, is small.
The effective pellet temperature for pellet dimensional change is that value which produces the same outer pellet radius in a virgin pellet as that obtained from the temperature model. The
effective clad temperature for dimensional change is its average value.
The temperature calculational model has been validated by plant Doppler defect data, as shown in table 4.3-7, and Doppler coefficient data, as shown in figure 4.3-42. Stability index
measurements also provide a sensitive measur e of the Doppler coefficient near full power (paragraph 4.3.2.7). It can be seen that Doppler defect data is typically within 0.2 percent of prediction. 4.3.3.2 Macroscopic Group Constants There are two lattice codes which have been used for the generation of macroscopic group constants needed in the spatial, few-group diffusion codes. One is PHOENIX-P which has historically been the sources of the macroscopic group constants. The other is PARAGON, which will be used in forthcoming reload designs. Following is a detailed description of each. PHOENIX-P has been approved by the NRC as a lattice code for the generation of macroscopic and microscopic few-group cross-sections for PWR analysis (reference 37). PHOENIX-P is a two-dimensional, multigroup, transport-based lattice code capable of providing all necessary data for PWR analysis. Since it is a dimensional lattice code, PHOENIX-P does not rely on predetermined spatial/spectral interaction assumptions for the heterogeneous fuel lattice and can provide a more accurate multigroup flux solution.
The solution for the detailed spatial flux and energy distribution is divided into two major steps in PHOENIX-P (reference 37). First, a two-dimensional fine energy group nodal solution is obtained, coupling individual sub-cell regions (pellet, clad, and moderator) as well as surrounding pins, using a method based on Carivik's collision probability approach and heterogeneous response fluxes which preserve the heterogeneity of the pin cells and their surroundings. The nodal solution provides an accurate and detailed local flux distribution, which is then used to homogenize the pin cells spatially to fewer groups. Then, a standard S4 VEGP-FSAR-4
4.3-37 REV 19 4/15 discrete ordinates calculation solves for the angular distribution, based on the group-collapsed and homogenized cross-sections from the first step. These S4 fluxes normalize the detailed spatial and energy nodal fluxes, which are then used to compute reaction rates and power distributions and to deplete the fuel and burnable absorbers. A standard B1 calculation evaluates the fundamental mode critical spec trum, providing an improved fast diffusion coefficient for the core spatial codes. PHOENIX-P employs a 70 energy group library deriv ed from the ENDF/B-6 basic data. This library was designed to capture the integral properties of the multigroup data properly during group collapse and to model important resonanc e parameters properly. It contains all neutronics data necessary for modeling fuel, fission products, cladding and structural materials, coolant, and control and burnable absorber materials present in PWRs. Group constants for burnable absorber cells, control rod cells, guide thimbles and instrumentation thimbles, or other nonfuel cells, can be obtained directly from PHOENIX-P without adjustments such as those required in the cell or 1D lattice codes.
Paragon has been approved by the NRC as the new generation of Westinghouse lattice code (reference 47). PARAGON is a replacement for PHOENIX-P and its primary use will be to provide the same types of input data that PHOENIX-P generates for use in three-dimensional core simulator codes. This includes macroscopic cross-sections, microscopic cross-sections, pin factors for pin-power reconstruction calculations, discontinuity factors for a nodal method solution, and other data needed for safety analysis or other downstream applications.
PARAGON is based on collision probability-interface current cell coupling methods. PARAGON provides flexibility in modeling that was not available in PHOENIX-P including exact cell geometry representation instead of cylinderization, multiple rings and regions within the fuel pin and the moderator cell geometry, and variable cell pitch. The solution method permits flexibility in choosing the quality of the calculation through both increasing the number of regions modeled within the cell and the number of angular current directions tracked at the cell interfaces.
The calculation scheme in PARAGON is based on the conventional lattice modules: resonance calculation, flux solution, leakage correction, and depletion. The detailed theory of these modules is described in reference 47. The cross-section resonance calculation module is based on the space-dependent Dancoff method (reference 47); it is a generalization of the PHOENIX-P methodology that permits to subdivide the fuel pin into many rings and, therefore, generates space-dependent self-shielded isotopic cross-sections. The flux solution module uses the interface current collision probability method and permits a detailed representation of the fuel cells (reference 47). The other two modules (leakage and depletion) are similar to the ones used in PHOENIX-P.
The current PARAGON cross-section library is a 70-group library, based on the ENDF/B basic nuclear data, with the same group structure as the library currently used with PHOENIX-P. The PARAGON qualification library has been improved through the addition of more explicit fission products and fission product chains (reference 47). PARAGON is, however, designed to employ any number of energy groups. The new NEXUS cross-section generation system uses PARAGON as the lattice code (reference 48). 4.3.3.3 Spatial Few-Group Diffusion Calculations Spatial few-group diffusion calculations are performed using 3D ANC (reference 36). A two-group, two- and three-dimensional nodal code will be used for dimensional modeling of the VEGP-FSAR-4
4.3-38 REV 19 4/15 core. The three-dimensional nature of this code provides both radial and axial power
distributions. For some applications, the updated version of the PANDA will continue to be used
for axial calculations, and a two-dimensional collapse of 3D ANC that properly accounts for the three-dimensional features of the fuel will be used for X-Y calculations.
Nodal calculations (four radial mesh per assembly) are carried out to determine the critical boron concentrations and power distributions. The moderator coefficient is evaluated by varying the inlet temperature in the same kind of calculations as those used for power distribution and
reactivity predictions.
Validation of the reactivity calculations is associated with the validation of the group constants themselves, as discussed in paragraph 4.3.3.2. Validation of the Doppler calculations is
associated with the fuel temperature validation discussed in paragraph 4.3.3.1. Validation of
the moderator coefficient calculations is obt ained by comparison with plant measurements at HZP conditions, as shown in table 4.3-11.
Axial calculations are used to determine differential control rod worth curves (reactivity versus rod insertion) and axial power shapes during steady-state and transient xenon conditions (flyspeck curve). Group constants are obtained fr om the three-dimensional nodal model by flux-volume weighting on an axial slicewise basis. Radial bucklings are determined by
varying parameters in the buckling model while forcing the one-dimensional model to reproduce the axial characteristics (axial offset, midplane power) of the three-dimensional model.
Validation of the spatial codes for calculating power distributions involves the use of incore and excore detectors and is discussed in paragraph 4.3.2.2.7.
Based on comparison with measured data, it is estimated that the accuracy of current analytical methods is:
- +/-0.2 percent for Doppler defect.
- +/-2 x 10-5/°F for moderator coefficient.
- +/-50 ppm for critical boron concentration with depletion.
- +/-3 percent for power distributions.
- +/-0.2 percent for rod bank worth.
- +/-4 pcm/step for differential rod worth.
- +/-0.5 pcm/ppm for boron worth.
- +/-0.1 percent for moderator defect. 4.
3.4 REFERENCES
- 1. "Westinghouse Anticipated Transients Without Reactor Trip Analysis," WCAP-8330, August 1974. 2. Spier, E. M., "Evaluation of Nuclear Hot Channel Factor Uncertainties," WCAP-7308-L-P-A (Proprietary) and WCAP-7308-L-A, (Nonproprietary), June 1988. 3. Hellman, J. M., ed, "Fuel Densificati on Experimental Results and Model for Reactor Application," WCAP-8218-P-A (Proprietary) and WCAP-8219-A (Nonproprietary), March 1975. 4. Meyer, R. O., "The Analysis of Fuel Densification," Division of Systems Safety, U.S.
Nuclear Regulatory Commission, NUREG-0085, July 1976.
VEGP-FSAR-4
4.3-39 REV 19 4/15 5. Hellman, J. M., Olson, C. A., and Yang, J. W., "Effects of Fuel Densification Power Spikes on Clad Thermal Transients," WCAP-8359, July 1974. 6. "Power Distribution Control of Westinghouse Pressurized Water Reactors," WCAP-7811, December 1971. 7. Morita, T., et al., "Power Distribution Control and Load Following Procedures," WCAP-8385 (Proprietary) and WCAP-8403 (Nonproprietary), September 1974. 8. McFarlane, A. F., "Power Peaking Factors," WCAP-7912-P-A (Proprietary) and WCAP-7912-A (Nonproprietary), January 1975. 9. Meyer, C. E., and Stover, R. L., "Incore Power Distribution Determination in Westinghouse Pressurized Water Reactors," WCAP-8498, July 1975. 10. Barry, R. F., and Altomare, S., "The TURTLE 24.0 Diffusion Depletion Code," WCAP-7213-A (Proprietary) and WCAP-7758-A (Nonproprietary), February 1975. 11. Cermak, J. O., et al., "Pressurized Water Reactor pH -Reactivity Effect Final Report," WCAP-3696-8 (EURAEC-2074), October 1968. 12. Ford, W. E., III, "CSRL-V: Processed ENDF1B-V 227-Neutron-Group and Pointwise Cross-Section Libraries for Criticality Safety, Reactor and Shielding Studies,"
ORNL/CSD/TM-160, June 1982. 13. Deleted. 14. Petrie, L. M., and Cross, N. F., "KENO IV--An Improved Monte Carlo Criticality Program," ORNL-4938, November 1975. 15. Deleted.
- 16. Deleted.
- 17. Deleted.
- 18. Poncelet, C. G., and Christie, A. M., "Xenon-Induced Spatial Instabilities in Large Pressurized Water Reactors," WCAP-3680-20 (EURAEC-1974), March 1968. 19. Skogen, F. B., and McFarlane, A. F., "Control Procedures for Xenon-Induced X-Y Instabilities in Large Pressurized Water Reactors," WCAP-3680-21 (EURAEC-2111), February 1969. 20. Skogen, F. B., and McFarlane, A. F., "Xenon-Induced Spatial Instabilities in Three Dimensions," WCAP-3680-22 (EURAEC-2116), September 1969. 21. Lee, J. C., et al., "Axial Xenon Transient Tests at the Rochester Gas and Electric Reactor," WCAP-7964, June 1971. 22. Barry, R. F., and Minton, G., "The PANDA Code," WCAP-7084-P-A (Nonproprietary), February 1975. 23. Barry, R. F., "LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for the IBM-7094," WCAP-3269-26, September 1963. 24. England, T. R., "CINDER - A One-Point Depletion and Fission Product Program," WAPD-TM-334, August 1962. 25. Eggleston, F. R., "Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries - Winter 1977 - Summer 1978," WCAP-8768, Revision 2, October 1978.
VEGP-FSAR-4
4.3-40 REV 19 4/15 26. Poncelet, C. G., "LASER - A Depletion Program for Lattice Calculations Based on MUFT and THERMOS," WCAP-6073, April 1966. 27. Olhoeft, J. E., "The Doppler Effect for a Non-Uniform Temperature Distribution in Reactor Fuel Elements," WCAP-2048, July 1962. 28. Nodvik, R. J., "Supplementary Report on Evaluation of Mass Spectrometric and Radiochemical Analyses of Yankee Core I Spent Fuel, Incuding Isotopes of Elements
Thorium Through Curium," WCAP-6086, August 1969. 29. Drake, M. K., ed, "Data Formats and Procedure for the ENDF/B Neutron Cross Section Library," BNL-50274, ENDF-102, Vol. 1, 1970. 30. Suich, J. E., and Honeck, H. C., "The HAMMER System, Heterogeneous Analysis by Multigroup Methods of Exponentials and Reactors," DP-1064, January 1967. 31. Flatt, H. P., and Buller, D. C., "AIM-5, A Multigroup, One Dimensional Diffusion Equation Code," NAA-SR-4694, March 1960. 32. "Nuclear Design of Westinghouse Pressurized Water Reactors with Burnable Poison Rods," WCAP-7806, December 1971. 33. Strawbridge, L. E., and Barry, R. F., "Criticality Calculation for Uniform Water-Moderated Lattices," Nuclear Science and Engineering 23, p 58, 1965. 34. Nodvik, R. J., "Saxton Core II Fuel Performance Evaluation," WCAP-3385-56, Part II, "Evaluation of Mass Spectrometric and Radiochemical Materials Analyses of Irradiated
Saxton Plutonium Fuel," July 1970. 35. Leamer, R. D., et al., "PuO 2 -UO 2 Fueled Critical Experiments," WCAP-3726-1, July 1967. 36. Davidson, S. L., ed, et al., "ANC: Westinghouse Advanced Nodal Computer Code," WCAP-10965-P-A, September 1986. 37. Nguyen, T. Q., et al., "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," WCAP-11596-P-A, June 1988. 38. Mildrum, C. M., Mayhue, L. T., Baker, M. M., and Isaac, P. G., "Qualification of the PHOENIX/POLCA Nuclear Design and Analysis Program for Boiling Water Reactors,"
WCAP-10841 (Proprietary), and WCAP-10842 (Nonproprietary), June 1985. 39. WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control, FQ Surveillance Technical Specification," February 1994. 40. Chao, y. a., et al., "Westinghouse Dynamic Rod Worth Measurement Technique," WCAP-13360-P-A, January 1996. 41. Beard, C.L. and T. Morita, "Beacon Core Monitoring and Operations Support System," WCAP-12472-P-A Addendum 1-A, January 2000. 42. Vogtle Electric Generating Plant, Request to Revise Technical Specifications to Reflect Updated Spent Fuel Rack Criticality Analyses for Units 1 and 2, SNC Letter NL-04-0973, J. T. Gasser to NRC, August 13, 2004. 43. Issuance of Amendments that Revise the Spent Fuel Pool Rack Criticality Analyses, NRC to D. E. Grissette, September 22, 2005 (Amendments 139/118). 44. WCAP-16260-P-A, Revision 0, "The Spatially Corrected Inverse Count Rate (SCICR)
Method for Subcritical Reactivity Measurement," September 2005.
VEGP-FSAR-4
4.3-41 REV 19 4/15 45. GP-18735, "Evaluation of a Reduction in the Required Number of Movable Incore Detector Thimbles," January 31, 2011. 46. GP-18767," Southern Nuclear Operating Company, Vogtle Electric Generating Plant Units 1 and 2, Cycle 17 Movable Incore Detector Thimble Evaluation," April 4, 2011. 47. Ouisloumen, M., et al., "Qualification of the Two-dimensional Transport Code PARAGON," WCAP-16045-P-A, Westinghouse, 2004. 48. Zhang, B., et al., "Qualification of the NEXUS Nuclear Data Methodology," WCAP-16045-P-A, Addendum 1, Westinghouse, 2005. 4.3.5 BIBLIOGRAPHY Dominick, I. E., and Orr, W. L., "Experimental Verification of Wet Fuel Storage Criticality
Analyses," WCAP-8682 (Proprietary) and WCAP-8683 (Nonproprietary), December 1975.