ML20049H276

Chapter 4 to Gessar, Reactor.

ML20049H276
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Site:	05000447
Issue date:	02/12/1982
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t GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CHAPTER 4 REACTOR i

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I SECTION 4.1 CONTENTS Section *itle Page 4.1

SUMMARY

DESCRIPTION 4.1-1 4.1.1 Reactor Vessel 4.1 4.1.2 Reactor Internal Components 4.1-1 4.1.2.1 Reactor Core 4.1-2 4.1.2.1.1 General .4.1-2 4.1.2.1.2 Core Configuration 4.1-6 4.1.2.1.3 Fuel Assembly Description 4.1-6 4.1.2.1.3.1 Fuel Rod 4.1-7 4.1.2.1.3.2 Fuel Bundle 4.1-7 4.1.2.1.4 Assembly Support and Control Rod Location 4.1-7 4.1.2.2 Shroud 4.1-8

~' 4.1.2.3 Shroud Head and Steam Separators 4.1-8 4.1.2.4 Steam Dryer Assembly 4.1-8 4.1.3 Reactivity Control Systems 4.1-8 4.1.3.1- Operation 4.1-8 4.1.3.2 Description of Control Rods 4.1-9 4.1.3.3 Supplementary Reactivity Control 4.1-11 4.1.4 Analysis Techniques 4.1-11 4.1.4.1 Reactor Internal Components 4.1-11 4.1.4.1.1 MASS (Mechanical Analysis of Space Structure) 4.1-12 4.1.4.1.1.2 Program Version and Computer 4.1-12 i

j 4.1.4.1.1.3 History of Use 4.1-12 i

j 4.1.4.1.1.4 Extent of Application 4.1-12 j 4.1.4.1.2 SNAP (MULTISHELL) 4.1-13 4.1.4.1.2.1 Program Description 4.1-13 4.1.4.1.2.2 Program Version and Computer 4.1-13 1

i. 4.1.4.1.2.3 History of Use 4.1-13

! 4.1.4.1.2.4 Extent of Application 4.1-14 4.1-i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CONTENTS (Continued)

Section Title Page 4.1.4.1.3 GASR 4.1-14 4.1.4.1.3.1 Program Description 4.1-14 4.1.4.1.3.2 Program Version and Computer 4.1-14 4.1.4.1.3.3 History of Use 4.1-14 4.1.4.1.3.4 Extent of Application 4.1-14 4.1.4.1.4 NOHEAT 4.1-15 4.1.4.1.4.1 Program Description 4.1-15 4.1.4.1.4.2 Program Version and Computer 4.1-15 4.1.4.1.4.3 History of Use 4.1-15 4.1.4.1.4.4 Extent of Application 4.1-16 4.1.4.1.5 FINITE 4.1-16 4.1.4.1.5.1 Program Description 4.1-16 4.1.4.1.5.2 Program Version and Computer 4.1-16 4.1.4.1.5.3 History of Use 4.1-16 4.1.4.1.5.4 Extent of Usage 4.1-16  ;

4.1.4.1.6 DYSEA 4.1-17 4.1.4.1.6.1 Program Description 4.1-17 4.1.4.1.6.2 Program Version and Computer 4.1-17 4.1.4.1.6.3 History of Use 4.1-17 4.1.4.1.6.4 Extent of Application 4.1-18 4.1.4.1.7 SHELL 5 4.1-18 4.1.4.1.7.1 Program Descripcion 4.1-18 4.1.4.1.7.2 Program Version and Computer 4.1-18 4.1.4.1.7.3 History of Use 4.1-19 4.1.4.1.7.4 Extent of Application 4.1-19 4.1.4.1.8 HEATER 4.1-19 4.1.4.1.8.1 Program Description 4.1-19 4.1.4.1.8.2 Program Version and Computer 4.1-19 4.1.4.1.8.3 History of Use 4.1-19 4.1.4.1.8.4 Extent of Application 4.1-20 4.1.4.1.9 FAP-71 (Fatigue Analysis Program) 4 . ~ :2 0 4.1.4.1.9.1 Program Description 4.1-20 4.1-ii

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 O

CONTENTS (Continued)

Section- Title Page 4.1.4.1.9.2 Program Version and Computer 4.1-20 j 4.1.4.1.9.3 History of Use 4.1-20 4.1.4.1.9.4 Extent of Use 4.1-21 4.1.4.1.10 CREEP / PLAST 4.1-21 4.1.4.1.10.1 Program Description 4.1-21 4.1.4.1.10.2 Program Version and Computer 4.1-21 4.1.4.1.10.3 History of Use 4.1-21 4.1.4.1.10.4 Extent of Application 4.1-22 4.1.4.1.11- ANSYS 4.1-22 4.1.4.1.11.1 Program Description 4.1-22 4.1.4.1.11.2 Program Version and Computer 4.1-23 4.1.4.1.11.3 History of Use 4.1-23 4.1.4.1.11.4 Extent of Application 4.1-23

() 4.'1.4.1.12 4.1.4.1.12.1 CLAPS-02 Program Description 4.1-23 4.1-23 4.1.4.1.12.2 Program Version and Computer 4.1-24 4.1.4.1.12.3 History of Use 4.1-24 4.1.4.1.12.4 Extent of Application 4.1-24 4.1.4.1.13 ASIST 4.1-24 4.1.4.1.13.1 Program Description 4.1-24 4.1.4.1.13.2 Program Version and Computer 4.1-25

} 4.1.4.1.13.3 History of Use 4.1-25 4.1.4.1.13.4 Extent of Application 4.1-25 4.1.4.2 Fuel Rod Thermal Analysis 4.1-25 f

4.1.4.3 Reactor Systems Dynamics 4.1-26 4.1.4.4 Nuclear Engineering Analysis 4.1-26 i

4.1.4.5 Neutron Fluence Calculations 4.1-27 4.1.4.6 Thermal Hydraulic Calculations 4.1-27 4.1.5 References 4.1-28 i

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

4. REACTOR

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SUMMARY

DESCRIPTION The reactor assembly consists of the reactor vessel, its internal components of the cor^ (shroud, steam separator and dryer assem-blies) and jet pumps. Also included in the reactor assembly are the control rods, control rod drive (CRD) housings and the control rod drives. Figure 3.9-7 (Reactor Vessel Cutaway) shows the arrangement of reactor assembly components. A summary of the important design and performance characteristics is given in Subsection 1.3.1.1, " Nuclear Steam Supply System Design Charac-teristics". Loading conditions for reactor assembly components are specified in Subsection 3.9.5.2.

4.1.1 Reactor Vessel (Q:

U The reactor vessel design and description are covered in Section 5.3.

4.1.2 Reactor Internal Components The major reactor internal components are the core (fuel, channels, control blades and instrumentation), the core support structure (including the shroud, top guide and core plate), the shroud head and steam separator assembly, the steam dryer assembly, the feed-water spargers, the core spray spargers and the jet pumps. Except for the Zircaloy in the reactor core, these reactor internals are stainless steel or other corrosion-resistant alloys of the preced-ing components; the fuel assemblies (including fuel rods and channel), control blades, in-core instrumentation, shroud head and steam separator assembly, and steam dryers are removable when the reactor vessel is opened for refueling or maintenance.

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4.1-1 i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.1.2.1 Reactor Core 4.1.2.1.1 General The design of the boiling water reactor (BWR) core, including fuel, is based on the proper combination of many design variables and operating experience. These factors contribute to the~ achievement of high reliability.

The nuclear core design described herein is based on the equilib-rium reload cycle rather than the initial cycle. The equilibrium cycle is chosen for the basis of the licensing product for two important reasons:

(1) The equilibrium cycle is more typical of the expected operating state over the life of the reactor (2) The use of the equilibrium cycle generally results in a more conservative licensing basis than the initial cycle.

The equilibrium cycle is defined as that reload cycle in which all characteristics are identical to the previous cycles. That is, the reload bundles, the reload batch size, the reload pattern, the cycle energy, etc., and the core behavior remains the same i from cycle to cycle.

Since the equilibrium cycle does, in fact, b o u r.r' the initial core from a licensing point of view, only a submittal of the descrip-tion of the initial core will be needed under the rules of 10CFR50.59.

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GESSAR II 22A7007

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/) 4.1.2.1.1 General (Continued)

NJ A number of important features of the BWR core design are summarized in the following paragraphs:

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(1) ,

The BWR core mechanical design is based on conservative application of stress limits, operating experience and experimentalftest resblts. The moderate pressure level characteristics of a direct cycle reactor (approxi-mately 1000 psia) result in moderate cladding tempera-tures and stress levels.

(2) The low coolant saturation temperature, high heat transfer coefficients and neutral water chemistry of the BWR are significant, advantageous factors in mini-mizing Zircaloy temperature and associated temperature-dependent corrosion and P buildup.

(A)

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.iThe relatively uniform fuel cla'dding temperatures

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throughout the core minimize migration of the hydrides to cold cladding zones and reduce thermal stresses.

T (3) The basic thermal and mechanical criteria applied in the design have been proven by irradiation of statis-tically significant quantitles of fuel. The design heat transfer rates and linear he'at generation rates f are similar to values proven in fuel' assembly irradiation. -

(4) The design-power distribution used in sizing the core represents an expected state of operation.

(5) The General Electric thermal analysis basis (GETAB) is applied to assure that more than 99.9% of the fuel

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G' rods in the core are expected to avoid boiling 4.1-3

GESSAR II 22A7007

'/ 238 NUCLEAR ISLAND Rev. O 4.1.2.1.l' General (Continued) , g transition for the most severe moderate frequency transient described in Chapter 15. The possibility of boiling transition occurring.during normal reactor operation is insignificant.

(6) Becaose of the large negative moderator density coeffi-cient of reactivity, the BWR has a number of inherent advantages: (a) uses of coolant flow for load following; (b) inherent self-flattening of the radial power distri-bution; (c) ease of control; (d) spatial xenon stability; and (e) ability to override xenon, in order to follow load.

Boiling water reactors do not have instability problems due to xenon. This has been demonstrated by calculations and by special tests which have been conducted on operating BWRs in an attempt to force the reactor into xenon instability. No xenon insta-bilities have ever been observed in the test results. All of these indicators have proven that xenon transients are highly damped in a BWR due to the large negative power coefficient of reactivity (Reference 1).

Important features of the reactor core arrangement are as follows:

(1) The bottom-entry cruciform control rods consist of B 4C in stainless steel tubes surrounded by a stainless steel sheath.

Rods of this design were first introduced in the Dresden-1 reactor in April 1961 and have accumulated thousands of hours of service.

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.1.2.1.1 General (Continued)

U-x (2) The fi>cd in-core fission chambers provide continuous power range neutron flux monitoring. A guide tube in each in-core assembly provides for a traversing ion chamber for calibration and axial detail. Source and intermediate range detectors are located in-core and are axially retractable. The in-core location of the startup and source range instruments provides coverage

'of the large reactor core and provides an acceptable signal-to-noise ratio and neutron-to-gamma ratio. All in-core instrument leads enter from the bottom and the instruments are in service during refueling. In-core instrumentation is discussed in Subsection 7.6.2.1.

(3) As shown by experience obtained at Dresden-1 and other plants, the operator, utilizing the in-core flux moni-f'~s tor system, can maintain the desired power distribution

\s s] within a large core by proper control rod scheduling.

(4) The Zircaloy-4 reusable channels provide a fixed flow path for the boiling coolant, serve as a guiding surface for the control rods and protect the fuel during handling operations.

(5) The mechanical reactivity control permits criticality checks during refueling and provides maximum plant safety. The core is designed to be subcritical at any time in its operating history with any one control rod fully withdrawn.

(6) The selected control rod pitch represents a practical value of individual control rod reactivity worth, and allows adequate clearance below the pressure vessel

(s between CRD mechanisms for case of maintenance and

' ' removal.

4.1-5

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.1.2.1.2 Core Configuration The reactor core is arranged as an upright circular cylinder containing a large number of fuel cells and is located within the reactor vessel. The coolant flows upward through the core. The core arrangement (plan view) and the lattice configuration are shown in Figures 4.3-1 and 4.3-2, respectively.

4.1.2.1.3 Fuel Assembly Description As can be seen from the referenced figures, the BWR core is com-posed of essentially two components--fuel assemblies and control rods. The fuel assembly and control rod mechanical configurations (Figures 4.2-2 and 4.2-4a, respectively) are basically the same as used in Dresden-1 and in all subsequent General Electric boil-ing water reactors.

4.1.2.1.3.1 Fuel Rod A fuel rod consists of UO2 Pellets and a Zircaloy-2 cladding tube.

A fuel rod is made by stacking pellets into a Zircaloy-2 cladding tube which is evacuated and backfilled with helium at 3 atmospheres pressure, and sealed by welding Zircaloy end plugs in each end of the tube. The ASME Boilcr and Pressure Vessel Code,Section III, is used as a guide in the mechanical design and stress analysis of the fuel rod. The rod is designed to withstand applied loads, both external and internal. The fuel pellet is sized to provide suffi-cient clearance wi;hin the fuel tube to accommodate axial and radial differential expansion between fuel and clad. Overall fuel rod design is conservative in its accommodation of the mechanisms affecting fuel in a BWR environment. Fuel rod design bases are dis-cussed in more detail in Subsection 4.2.1.

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.1.2.1.3.2 Fuel Bundle

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Each fuel bundle contains 62 fuel rods and two water rods which are spaced and supported in a square (8x8) array by seven spacers and a lower and upper tieplate. The fuel bundle has two important design features:

(1) The bundle design places minimum external forces on a j

j fuel rod; each fuel rod is free to expand in the axial direction.

(2) The unique structural design permits the removal and replacement, if required, of individual fuel rods.

The fuel assemblies, of which the core is comprised, are designed to meet all the criteria for core performance and to provide ease of handling. Selected fuel rods in each assembly differ from the

() others in uranium enrichment. This arrangement produces more uni-form power production across the fuel assembly, and thus allows a significant reduction in the amount of heat transfer surface required to satisfy the design thermal limitations.

4.1.2.1.4 Assembly Support and Control Rod Location A few peripheral fuel assemblies are supported by the core plate.

Otherwise, individual fuel assemblics in the core rest on fuel sup-port pieces mounted on top of the control rod guide tubes. Each

! guide tube, with its fuel support piece, bears the weight of four assemblies and is supported by a control rod drive penetration nozzle in the bottom head of the reactor vessel. The core plate provides lateral support and guidance at the top of each control

rod guide tube.

The top guide, mounted on top of the shroud, provides lateral sup-() port and guidance for the top of each fuel assembly. The reactivity 1

4 4.1-7

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.1.2.1.4 Assembly Support and Control Rod Location (Continued) of the core is controlled by cruciform control rods, containing boron carbide, and their associated mechanical hydraulic drive sys-tem. The control rods occupy alternate spaces between fuel assem-blies. Each independent drive enters the core from the bottom, and can accurately position its associated control rod during normal operation and yet exert upproximately ten times the force of gravity to insert the control rod during the scram mode of operation. Bot-tom entry allows optimum power shaping in the core, case of refuel-ing and convenient drive maintenance.

4.1.2.2 Shroud The information on the shroud is contained in Subsection 3.9.5.1.1.1.

4.1.2.3 Shroud Ilead and Steam Separators The information on the shroud head and steam separators is contained in Subsection 3.9.5.1.1.3.

4.1.2.4 Steam Dryer Assembly The information on the steam dryer assembly is contained in Sub-section 3.9.5.1.2.2.

4.1.3 Reactivity Control Systems 4.1.3.1 Operation The control rods perform dual functions of power distribution shaping and reactivity control. Power distribution in the core is controlled during operation of the reactor by manipulation of selected patterns of rods. The rods, which enter from the bottom of the near-cylindrical reactor core, are positioned to 4.1-8

1 GESSAR II 22A7007  !

238 NUCLEAR ISLAND Rev. O I i

'N 4.1.3.1 Operation (Continued) counterbalance steam voids in the top of the core and effect significant power flattening.

These groups of control elements, used for power flattening, experience a somewhat higher duty cycle and neutron exposure than the other rods in the control system.

The reactivity control function requires that all rods be available for either reactor " scram" (prompt shutdown) or reactivity regula-tion. Because of this, the control element; are mechanically designed to withstand the dynamic forces resulting from a scram.

They are connected to bottom-mounted, hydraulically actuated drive mechanisms which allow either axial positioning for reactivity regu-lation or rapid scram insertion. The design of the rod-to-drive connection permits each blade to be attached or detached from its drive without disturbing the remainder of the control system. The

(N bottom-mounted drives permit the entire control system to be left intact and operable for tests with the reactor vessel open.

4.1.3.2 Description of Control Rods The cruciform-shaped control rods contain 72 stainless steel tubes (18 tubes in each wing of the cruciform) filled with vibration compacted boron-carbido powder. The tubes are seal welded with end plugs on either end. Stainless steel balls are used to separate the tubes into individual compartments. The stainless steel balls are held in position by a slight crimp in the tube. The individual tubes provide containment of the helium gas released by the boron-neutron capture reaction.

The tubes are held in a cruciform array by a stainless steel sheath extending the full length of the tubes. A top handle (Figure 4.2-4a)

(~') aligns the tubes and provides structural rigidity at the top of the

\' / control rod. Rollers, housed in the handle, provide guidance for 4.1-9

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.1.3.2 Description of Control Rods (Continued) control rod insertion and withdrawal. A bottom casting is also used to provide structural rigidity and contains positioning rollers and a parachute-shaped velocity limiter. The handle and lower casting are welded into a single structure by means of a small cruciform post located in the center of the control rod. The con-tr'ol rods can be positioned at 6-in. steps and have a nominal with-drawal and insertion speed of 3 in./sec.

The velocity limiter is a device which is an integral part of the control rod and protects against the low probability of a rod drop accident. It is designed to limit the free-fall velocity and reactivity insertion rate of a control rod so that minimum fuel damage would occur. It is a one-way device, in that control rod scram time is not significantly affected.

Control rods are cooled by the core leakage (bypass) flow. The core leakage flow is made up of recirculation flow that leaks through the several leakage flow paths, the most important of which are:

(1) the area between the fuel channel and the tuel assembly lower tieplate; (2) holes in the lower tieplate; j (3) the area between the fuel assembly lower tieplate and the fuel support piece; (4) the area between the fuel support piece and the control rod guide tube; (5) the area between the control rod guide tube and the core support plate; and (6) the area between the core support plate and the shroud.

4.1-10

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

(}

'V 4.1.3.3 Supplementary Reactivity Control The core control requirements are met by use of the combined effects of the movable control rods, supplementary burnabic poison, and variation of reactor coolant flow. The supplementary burnable poison is gadolinia (Gd 0 I "I* " UO in selected fuel rods 23 2 in each fuel bundle.

4.1.4 Analysis Techniques 4.1.4.1 Peactor Internal Components Computer codes used for the analysis of the internal components are listed as follows:

(1) MASS (2) SNAP (MULTISilELL)

(3) GASP O (4) NCilEAT (5) FINITE (6) DYSEA (7) SilELL 5 (8) IIEATER (9) FAP-71 (10) CREEP-PLAST (11) ANSYS (12) C LAPS-02 (13) ASIST Detail description of these programs are given in the following sections.

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GESSAR II 22A7007 238 NUCLEAR ISLAND RSv. 0 4.1.4.1.1 MASS (Mechanical Analysis of Space Structure) 4.1.4.1.1.1 Program Description The program, proprietary of the General Electric Company, is an outgrowth of the PAPA (Plate and Panel Analysis) program originally developed by L. Beitch in the early 1960s. The program is based on the principle of the finite element method. Governing matrix equations are formed in terms of joint displacements using a

" stiffness-influence-coefficient" concept originally proposed by L. Beitch (Reference 2). The program offers curved beam, plate and shell elements. It can handle mechanical and thermal loads in a static analysis and predict natural frequencies and mode shapes in a dynamic analysis.

4.1.4.1.1.2 Program Version and Computer The Nuclear Energy Business Group is using a past revision of MASS.

This revision is identified as revision "0" in the computer pro-duction library. The program operates on the Honeywell 6000 computer. ,

4.1.4.1.1.3 History of Use Since its development in the early 60s, the program has been successfully applied to a wide variety of jet-engine structural problems, many of which involve extremely complex geometries. The use of the program in the Nuclear Energy Business Group also started shortly after its development.

4.1.4.1.1.4 Extent of Application Besides the Jet Engine Division and the Nuclear Energy Business Group: the Missile and Space Division, the Appliance Division, and the Turbine Division of General Electric have also applied the 4.1-12

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

/~'T 4.1.4.1.1.4 Extent of Application (Continued)

V program to a wide range of engineering problems. The Nuclear Energy Business Group (NEBG) uses it mainly for piping and reactor internals analyses.

4.1.4.1.2 SNAP (MULTISHELL) 4.1.4.1.2.1 Program Description The SNAP Program, which is also called MULTISHELL, is the General Electric Code which determines the loads, deformations and stres-ses of axisymmetric shells of revolution (cylinders, cones, discs, toroids and rings) for axisymmetric thermal boundary and surface load conditions. Thin shell theory is inherent in the solution of E. Poissner's differential equations for each shell's influence coefficients. Surface loadin g capability includes pressure, average temperature and linear throughwall gradients; the latter

(} two may be linearly varied over the shell meridian. The theoretical limitations of this program are the same as those of classical theory.

4.1.4.1.2.2 Program Version and Computer The current version maintained by the General Electric Jet Engine Division at Evandale, Ohio is being used on the Honeywell 6000 computer in GE/NEBG.

4.1.4.1.2.3 History of Use The initial version of the Shell Analysis Program was completed by the Jet Engine Division in 1961. Since then, a considerable amount of modification and addition has been made to accommodate its broadening area of application. Its application in the NEBG has

() a history longer than 10 years.

4.1-13 i

T GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.1.4.1.2.4 Extent of Application The program has been used to analyze jet engine, space vehicle and nuclear reactor components. Because of its efficiency and economy, in addition to reliability, it has been one of the main shell analysis programs in General Electric's NEBG.

4.1.4.1.3 GASP 4.1.4.1.3.1 Program Description GASP is a finite element program for the stress analysis of axisym-metric or plane two-dimensional geometries. The element representa-tions can be either quadrilateral or triangular. Axisymmetric or plane structural loads can be input at nodal points. Displacements, temperatures, pressure loads and axial inertia can be accommodated.

Effective plastic stress and strain distributions can be calculated using a bilinear stress-strain relationship by means of an inter-active convergence procedure.

4.1.4.1.3.2 Program Version and Computer The GE version, originally obtained from the developer, Pro-fessor E. L. Wilson, operates on the lioneywell 6000 computer.

4.1.4.1.3.3 IIistory of Use The program was developed by E. L. Wilson in 1965 (Reference 3).

The present version in GE/NEBG has been in operation since 1967.

4.1.4.1.3.4 Extent of Application The application of GASP in GE/NEBG is mainly for clastic analysis of axisymmetric and plane structures under thermal and pressure loads. The GE version has been extensively tested and used by engineers in General Electric Company.

4.1-14

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

/ 4.1.4.1.4 NOHEAT O

4.1.4.1.4.1 Program Description The NOHEAT program is a two-dimensional and axisymmetric, transient, nonlinear temperature analysis program. An unconditionally stable numerical integration scheme is combined with an iteration pro-cedure to compute temperature distribution within the body sub-jected to arbitrary time- and temperature-dependent boundary conditions.

This program utilizes the finite element method. Included in the analysis are the three basic forms of heat transfer, conduction, radiation, and convection, as well as internal heat generation.

In addition, cooling pipe boundary conditions are also treated.

The output includes temperature of all the nodal points for the time instants specified by the user. The program can handle 7m

() multitransient temperature input.

4.1.4.1.4.2 Program Version and Computer The current version of the program is an improvement of the pro-gram originally developed by I. Farhoomand and Professor B. L.

Wilson of University of California at Berkeley (Reference 4).

The program operates on the Honeywell 6000 computer.

4.1.4.1.4.3 History of Use The program was developed in 1971 and installed in General Electric Honeywell computer by one of its original developers, I. Farhoomand, in 1972. A number of heat transfer problems related to the reactor pedestal have been satisfactorily solved using the program.

/'N 4.1-15

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.1.4.1.4.4 Extent of Application The program using finite element formulation is compatible with the finite element, stress-analysis computer program GASP. Such compatibility simplified the connection of the two analyses and minimizes human error.

4.1.4.1.5 FINITE 4.1.4.1.5.1 Program Description FINITE is a general-purpose, finite element computer program for clastic stress analysis of two-dimensional structural problems, including: (1) plane stress; (2) plane strain; and (3) axisym-metric structures. It has provision for thermal, mechanical and body force loads. The materials of the structure may be homogen-cous or nonhomogeneous and isotropic or orthotropic. The devel-opment of the FINITE program is based on the GASP program (Subsection 4.1.4.1.3).

4.1.4.1.5.2 Program Version and Computer The present version of the program at GE/NEBG was obtained from the developer J. E. McConnelee of GE/ Gas Turbine Department in 1969 (Reference 5). The NEBG version is used on the Honeywell 6000 computer.

4.1.4.1.5.3. History of Use Since its completion in 1969, the program has been widely used in the Gas Turbine and the Jet Engine Departments of the General Electric Company for the analysis of turbine components.

4.1.4.1.5.4. Extent of Usage The program is used at GE/NEBG in the analysis of axisymmetric or nearly-axisymmetric BWR internals.

4.1-16

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O f~w 4.1.4.1.6 DYSEA 4.1.4.1.6.1 Program Description The DYSEA (Dynamic and Seismic Analysis) program is a GE proprie-tary program developed specifically for seismic and dynamic analy-sis of RPV and internals / building system. It calculates the dynamic response of linear structural systems by either temporal model superposition or response spectrum method. Fluid-structure inter-action effect in the RPV is taken into account by way of hydro-dynamic mass.

Program DYSEA was based on program SAPIV with added capability to handle the hydrodynamic mass effect. Structural stiffness and mass matrices are formulated similar to SAPIV. Solution is obtained in time domain by calculating the dynamic response mode by mode. Time integration is performed by using Newmark's 8-method.

j Response spectrum solution is also available as an option.

O]

4.1.4.1.6.2 Program Version and Computer The DYSEA version now operating on the Honeywell 6000 computer of GE, Nuclear Energy Systems Division, was developed at GE by modi-fying the SAPIV program. Capability was added to handle the hydrodynamic mass effect due to fluid-structure interaction in the reactor. It can handle three-dimension <l dynamic problems with beam, trusses, and springs. Both acceleration time histories and response spectra may be used as input.

4.1.4.1.6.3 IIistory of Use The DYSEA program was developed in the Summer of 1976. It has been adopted as a standard production program since 1977 and has been used extensively in all dynamic and seismic analysis of the RPV and internals / building system.

4.1-17

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.1.4.1.6.4 Extent of Application The current version of DYSEA has been used in all dynamic and seismic analysis since its development. Results from test prob-lems were found to be in close agreement with those obtained from either verified programs or analytic solutions.

4,1.4.1.7 SHELL 5 4.1.4.1.7.1 Program Description SHELL 5 is a finite shell element program used to analyze smoothly curved thin shell structures with any distribution of clastic material properties, boundary conceraints, and mechanical thermal and displacement loading conditions. The basic element is tri-angular whose membrane displacement fields are linear polynomial functions, and whose bending displacement field is a cubic poly-nomial function (Reference 6). Five degrees of freedom (three displacements and two bending rotations) are obtained at each nodal point.

Output displacements and stresses are in a local (tangent) surface coordinate s; stem.

Due to the approximation of element membr ane displacements by linear functions, the in-plane rotation about the surface normal is neglected. Therefore, the only rotations considered are due to bending of the shell cross-section, and application of the method is not recommended for shell intersection (or discontinu-ous surface) problems where in-plane rotation can be significant.

4.1.4.1.7.2 Program Version and Computer A copy of the source deck of SHELL 5 is maintained in GE/NEBG.

SHELL 5 operates on the internal computers.

4.1-18

GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 0

(~3 4.1.4.1.7.3 History of Use U

SHELL 5 is a program developed by Gulf General Atomic Incorporated (Reference 7) in 1969. The program has been in production status at Gulf General Atomic, General Electric, and at other major computer operating systems since 1970.

4.1.4.1.7.4 Extent of Application SHELL 5 has been used at General Electric to analyze reactor shroud support and torus. Satisfactory results were obtained.

4.1.4.1.8 HEATER 4.1.4.1.8.1 Program Description HEATER is a computer program used in the hydraulic design of feed-

) water spargers and their associated delivery header and piping.

(~'l

'ss The program utilizes test data obtained by GE using full-scale mockups of feedwater spargers combined with a series of models which represent the complex mixing processes obtained in the upper plenum, downcomer, and lower plenum. Mass and energy balances throughout the nuclear steam supply system (NSSS) are modeled in detail (Reference 8).

4.1.4.1.8.2 Program Version and Computer This program was developed at GE/NEBG in FORTRAN IV for the lioneywell 6000 computer.

4.1.4.1.8.3 History of Use The program was developed by various individuals in GE/NEBG beginning in 1970. The present version of the program has been in operation since January 1972.

('"3

(/

4.1-19

GESSAR II 22A7007 238 NUCLEAR ISLAND R2v. 0 4.1.4.1.8.4 Extent of Application The program is used in the hydraulic design of the feedwater spargers for each BWR plant, in the evaluation of design modifi-cations, and the evaluation of unusual operational conditions.

4.1.4.1.9 FAP-71 (Fatigue Analysis Program) 4.1.4.1.9.1 Program Description The FAP-71 computer code, or Fatigue Analysis Program, is a stress analysis tool used to aid in performing ASME-III Nuclear Vessel Code structural design calculations. Specifically, FAP-71 is used in determining the primary plus secondary stress range and number of allowable fatigue cycles at points of interest.

For structural locations at which the 3S m (P+Q) ASME Code limit is exceeded, the program can perform either (or both) of two elastic-plastic fatigue life evaluations: (1) the method reported in ASME Paper 68-PVP-3, or (2) the present method documented in Paragraph NB-3228.3 of the 1971 Edition of the ASME Section III Nuclear Vessel Code. The program can accommodate up to 25 transient stress states of as many as 20 structural locations.

4.1.4.1.9.2 Program Version and Computer The present version of FAP-71 was completed by L. Young of GE/NEBG in 1971 (Reference 9). The program currently is on the NEBG Honeywell 6000 computer.

4.1.4.1.9.3 History of Use Since its completion in 1971, the program has been applied to several design analyses of GE BWR vessels.

O 4.1-20

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.1.4.1.9.4 Extent of Use The program is used in conjunction with several shell analysis programs in determining the fatigue life of BWR mechanical com-ponents subject to thermal transients.

4.1.4.1.10 CREEP / PLAST 4.1.4.1.10.1 Program Description A finite element program is used for the analysis of two-dimensional (plane and axisymmetric) problems under conditions of creep and plasticity. The creep formulation is based on the memory theory of creep in which the constitutive relations are cast in the form of hereditary integrals. The material creep properties are built into the program and they represent annealed 304 stainless steel. Any other creep properties can be included if required.

The plasticity treatment is based on kinematic hardening and von Mises yield criterion. The hardening modulus can be constant or a function of strain.

4.1.4.1.10.2 Program Version and Computer The program can be used for elastic-plastic analysis with or without the presence of creep. It can also be used for creep analysis without the presence of instantaneous plasticity. A detailed description of theory is given in Reference 11. The program is operative on internal computers.

4.1.4.1.10.3 History of Use This program was developed by Y. R. Rashid (Reference ll) in 1971. It underwent extensive program testing before it was put

)

on production status.

4.1-21

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.1.4.1.10.4 Extent of Application O

The program is used at GE/NEBG in the channel cross-section mechanical analysis.

4.1.4.1.11 ANSYS 4.1.4.1.11.1 Program Description ANSYS is a general-purpose finite element computer program designed to solve a variety of problems in engineering analysis.

The ANSYS program features the following capabilities:

(1) Structural analysis, including static clastic, plastic and creep, dynamic, seismic and dynamic plastic, and large deflection and stability analysis.

(2) One-dimensional fluid flow analysis.

(3) Transient heat transfer analysis including conduction, convection, and radiation with direct input to thermal-stress analyses.

(4) An extensive finite element library, including gaps, friction interfaces, springs, cables (tension only),

direct interfaces (compression only), curved elbows, etc. Many of the elements contain complete plastic, creep, and swelling capabilities.

(5) Plotting - Geometry plotting is available for all ele-monts in the ANSYS library, including isometric and perspective views of three-dimensional structures.

O 4.1-22

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

,.m (v 4.1.4.1.11.1 Program Description (Continued)

(6) Restart Capability - The ANSYS program has restart capability for several analyses types. An option is also available for saving the stiffness matrix once it is calculated for the structure, and using it for other loading conditions.

4.1.4.1.11.2 Program Version and Computer The program is maintained current by Swanson Analysis Systems, Inc.

of Pittsburgh, Pennsylvania and is supplied to General Electric for use on the Honeywell 6000.

4.1.4.1.11.3 History of Use The ANSYS program has been used for productive analysis since Users now include the nuclear, pressure vessels and (a) early 1970.

piping, mining, structures, bridge, chemical, and automotive industries, as well as many consulting firms.

4.1.4.1.11.4 Extent of Application ANSYS is used extensively in GE/NEBG for elastic and elastic-plastic analysis of the reactor pressure vessel, core support structures, reactor internals and fuel.

4.1.4.1.12 CLAPS-02 4.1.4.1.12.1 Program Description CLAPS-02 is a general-purpose, two-dimensional finite element prog 7.am used to perform linear and nonlinear structural mechanics analysis. The program solves plane stress, plane strain and axi-() symmetri'c problems. It may be used to analyze for instantaneous 4.1-23

GESSAR Il 22A7007 238 NUCLEAh ISLAND Rev. 0 4.1.4.1.12.1 Program Description (Continued) pressure, temperature and flux changes, rapid transients and steady-state, as well as conventional elastic and inelastic buckling analyses of structural components subjected to mechanical loading.

4.1.4.1.12.2 Program Version and Computer CLAPS-02 is operational on the Honeywell-6000 computer and has a capacity of 500 nodal points, 400 elements, and 100 time steps.

4.1.4.1.12.3 History of Use CLAPS-02 is an improved version of CLAPS-01, which was developed primarily for the creep collapse analysis of BWR fuel rods.

CLAPS-02 uses a more sophisticated element, has more flexibility with regard to mechanical loading conditions and generally results in less running time than CLAPS-01.

4.1.4.1.12.4 Extent of Application CLAPS-02 has been widely used for stress analysis of fuel assem-bly components.

4.1.4.1.13 ASIST 4.1.4.1.13.1 Program Description The ASIST program is a General Electric code which can be used to obtain load distribution, deflections, critical frequencies and mode shapes in the "in-plane" or " normal-to-plane" modes for planar structures of any orientation that: (1) are statically indeterminate; (2) can be represented by straight or curved beams; and (3) are under basically any loading, thermal gradient, or 4.1-24

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O rS ) 4.1.4.1.13.1 Program Description (Continued)

(

\_/

sinusoidal excitation. Deformations and resulting load distributions are computed considering all strain energies (i.e.,

bending, torsion, shear and direct) . ASIST also considers the effects of the deflected shape on loads and provides deflections calculated for the structure. In addition to this beam column (large deflection) capability, the buckling instability of planar structures can also be calculated.

4.1.4.1.13.2 Program Version and Computer The current program version (ASIST-02) is used on the lioneywell-6000 computer in the General Electric Nuclear Energy Business Group.

4.1.4.1.13.3 History of Use

\-- The initial version of the ASIST program was developed by the General Electric Jet Propulsion Division. The program and its predecessors have been in use in the General Electric Aircraft Engine group for more than 10 years. Its applicati n in GE/NEBG has a history longer than 6 years.

4.1.4.1.13.4 Extent of Application The ASIST program has been used to determine spring constants, stresses, deflections, critical frequencies and associated modes shapes for frames, shafts, rotors, and other jet engine components.

It has been used extensively as a design and analysis tool for various components of nuclear fuel assemblies.

4.1.4.2 Fuel Rod Thermal Analysis Fuel rod thermal design analyses are performed utilizing the

('N)

\/ classical relationships for heat transfer in cylindrical 4.1-25

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.1.4.2 Fuel Rod Thermal Analysis (Continued) coordinate geometry with internal heat generation. Conditions of 100% and 116% of rated power are analyzed corresponding to ateady-state and short-term transient operation. Abnormal opera-tion transivats are also evaluated to assure that the damage limit of 1.0% cladding plastic strain is not violated. The strength theory, terminology, and strain-stress categories presented in the ASME Boiler and Pressure Vessel Code Section III are used as a guide in the mechanical design and stress analysis of the fuel rods.

4.1.4.3 Reactor Systems Dynamics The analysis techniques and computer codes used in reactor systems dynamics are described in Section 4 of Reference 10. Subsec-tion 4.4.4.6 also provides a complete stability analysis for the reactor coolant system.

4.1.4.4 Nuclear Analysis The analysis techniques are described and referenced in Subsection 4.3.3. The codes used in the analysis are:

Computer Code Function Lattice Physics Model Calculates average few-group cross sections, bundle reactivi-ties, and relative fuel rod powers within the fuel bundle.

BWR Reactor Simulator Calculates three-dimensional nodal power disuributions, exposures and thermal hydraulic charac-teristics as burnup progresses.

4.1-26

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3 4.1.4.5 Neutron Fluence Calculations l

Neutron vessel fluence calculations were carried out using a one-dimensional, discrete ordinates, Sn transport code with general anisotropic scattering.

This code is a modification of a widely used discrete ordinates code which will solve a wide variety of radiation transport problems. The program will solve both fixed source and multi-plication problems. Slab, cylinder, and spherical geometry are allowed with various boundary conditions. The fluence calcula-tions incorporate, as an initial starting point, neutron fission distributions prepared from core physics data as a distributed source. Anisotropic scattering was considered for all regions.

The cross sections were prepared with 1/E flux weighted, P sub (L) matrices for anistropic scattering but did not include reson-ance self-shielding factors. Fast neutron fluxes at locations other than the core mid-plane were calculated using a two-(~x]

-- dimensional, discrete ordinate code. The two-dimensional code is an extension of the one-dimensional code.

4.1.4.6 Thermal-Hydraulic Calculations The digital computer program uses a parallel flow path model to perform the steady-state BWR reactor core thermal-hydraulic analysis. Program input includes the core geometry, operating power, pressure, coolant flow rate and inlet enthalpy, and power distribution within the core. Output from the program includes core pressure drop, coolant flow distribution, critical power ratio, and axial variations of quality, density, and enthalpy for each channel type.

O V

4.1-27

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.1.5 References

1. R. L. Crowther, " Xenon Considerations in Design of Boiling Water Reactors," June 1968 (APED-5640).

2. L. Beitch, "Shell Structures Solved Numerically by Using a Network of Partial Panels," AIAA Journal, Volume 5, No. 3, March 1967.

3. E. L. Wilson, "A Digital Computer Program For 6.he Finite Element Analysis of Solids with Non-Linear Material Proper-ties," Aerojet General Technical Memo No. 23, Aerojet General, July 1965.

4. 1. Farhoomand and E. L. Wilson, "Non-Linear Heat Transfer Analysis of Axisymmetric Solids," SESM Report SESM71-6, University of California at Berkeley, Be-keley, California, 1971.

5. J. E. McConnelee, " Finite-Users Manual", General Electric TIS Report DF 69SL206, March 1969.

6. R. W. Clough and C. P. Johnson, "A Finite Element Approxi-mation for the Analysis of Thin Shells," International Journal Solid Structures, Vol. 4, 1968.

7. "A Computer Program for the Structural Analysis of Arbitrary O Three-Dimensional Thin Shells," Report No. GA-9952, Gulf General Atomic.

8. A. B. Burgess, " User Guide and Engineering Description of HEATER Computer Program," March 1974.

9. Young, L. J., "FAP-71 (Fatigue Analysis Program) Computer Code," GE/NED Design Analysis Unit R. A. Report No. 49, January 1972.

10. L. A. Carmichael and G. J. Scatena, " Stability and Dynamic Performance of the General Electric Boiling Water Reactor,"

January 1977 (NEDO-21506).

11. Y. R. Rashid, " Theory Report for Creep-Plast Computer Program," GEAP-10546, AEC Research and Development Report, January, 1972.

O 4.1-28

GESSAR II 22A7907

238 NUCLEAR ISLAND Rev. 0

() SECTION 4.2 CONTENTS Section Title Page 4.2 FUEL SYSTEM DESIGN 4.2-1 4.2.1. General and Detailed Design Bases 4.2-1 4.2.1.1 General Design Bases 4.2-1 4.2.1.1.1 Fuel Assembly and Its Components 4.2-1 4.2.1.1.1.1 Safety Design Bases 4.2-1 4.2.1.1.1.2 Basis for Fuel Rod Safety Evaluation 4.2-4 4.2.1.1.1.3 Design Ratios 4._-4 i

4.2.1.1.1.4 Maximum Allowable Stresses, Cycling and Fatigue Limits 4.2-7 4.2.1.1.2 Control Assembly and Its Components 4.2-8 4.2.1.2 Detailed Design Bases 4.2-9 4.2.1.2.1 Fuel Assembly and Its Components 4.2-9

() 4.2.1.2.1.1 4.2.1.2.1.2 Material Selection and Properties ~

Effects of Irradiation 4.2-9 4.2-10 4.2.1.2.1.3 Flow-Induced Vibration. 4.2-13

4.2.1.2.1.4 Fuel Densification 4.2-13 4.2.1.2.1.5 Fuel Rod Damage Mechanisms 4.2-13 4.2.1.2.1.6 Dimensional Stability 4.2-14

, 4.2.1.2.1.7 Fuel Shipping and Handling 4.2-14 4.2.1.2.1.8 Capacity for Fission Gas Inventory 4.2-15 4.2.1.2.1.9 Deflection 4.2-16 4.2.1.2.1.10 Fretting Wear and Corrosion 4.2-17 4.2.1.2.1.11 Potential for Water-Logging Rupture 4.2-17 4.2.1.2.1.12 Potential for Hydriding 4.2-18 4.2.1.2.1.13 Stress-Accelerated Corrosion 4.2-19 4.2.1.2.1.14 Fuel Reliability 4.2-19 4.2.1.2.1.15 Design Basis for Fuel Assembly Surveillance 4.2-21 4.2.1.2.2 Control Assembly and Its Components 4.2-22 4.2.1.2.2.1 Design Acceptability 4.2-22 O'x 4.2.1.2.2.2 Control Clearance 4.2-22 4.2-i

GESSAR II 22A7007 238 NUCLEAR ISLRND Rev. O CONTENTS (Continued) h Section Title 4.2.1.2.2.3 Mechanical Insertion Requirements 4.2-22 4.2.1.2.2.4 Material Selection 4.2-23 4.2.1.2.2.5 Radiation Effects 4.2-23 4.2.1.2.2.6 Positioning Requirements 4.2-23 4.2.2 General Design Description 4.2-24 4.2.2.1 Core Cell 4.2-24 4.2.2.2 Fuel Assembly 4.2-24 ,

4.2.2.2.1 Fuel Assembly Orientation 4.2-25 4.2.2.3 Fuel Bundle 4.2-25 4.2.2.3.1 Fuel Rods 4.2-26 4.2.2.3.1.1 Fuel Pellets 4.2-28 4.2.2.3.2 Water Rods 4.2-28 4.2.2.3.3 Fuel Spacer 4.2-29 4.2.2.3.4 Fuel Channe.1 4.2-30 4.2.2.3.5 Tieplates 4.2-31 4.2.2.3.6 Finger Springs 4.2-31 4.2.2.4 Reactivity Control Assembly 4.2-32 4.2.2.4.1 Control Rods 4.2-32 4.2.2.4.2 Velocity Limiter 4.2-34 4.2.3 Design Evaluations 4.2-35 4.2.3.1 Results of Fuel Rod Thermal-Mechanical Evaluations 4.2-36 4.2.3.1.1 Evaluation Methods 4.2-35 4.2.3.1.2 Fuel Damage Analysis 4.2-37 4.2.3.1.3 Steady-State Thermal-Mechanical Performance 4.2-37 4.2.3.2 Results from Fuel Design Evaluations 4.2-38 4.2.3.2.1 Flow-Induced Fuel Rod Vibrations 4.2-38 4.2.3.2.2 Potential Damaging Temperature Effects During Transients 4.2-39 4.2.3.2.3 Fretting Wear and Corrosion 4.2-39 4.2.3.2.4 Fuel Rod Cycling and Fatigue Analysis 4.2-40 4.2-ii

.i n GESSAR II 22A7007

u. 1

, 238 NUCLEAR ISLAND Rev. O I d I

i

.i .i ,

, f)7 CONTENTS (Continued) i Section Title Page

?

4.2.3.2.5 Fuel Rod Bowing 4.2-40 4.2.3.2.6 Fuel Assembly Dimensional Stability 4.2-40 4.2.3.2.'7 Temperature' Transients with a Waterlogged Fuel Element 4.2-41 4.2.3.2.7.1 Energy Release for Rupture of Waterlogged Fuel Elements 4.2-42 4.2.3.2.8 Fuel Densificatioh Analyses, 4.2-42

1. - 4.2.3.2.8.1 Power Spiking Analysis 4.2-42 i .4.2.3.2.8.2 Cladding Creep Collapso 4.2-44 4.2.3.2.8.3 Increased Linear Heat Generation ut m 4.2-44 4.2.3.2.8.4 Stored Energy Determination 4.2-45

'4.2.3.2.9 Fuel Cladding Temperatures 4.2-46 4.2.3.2.10 Incipient Fuel Center Melting 4.2-46

4.2.3.2,11 Energy Release During Fuel Emement Burnout 4.2-46 4.2.3.2.12 Fuel Rod Behavior Effects from

, Coolant Flow Blockage 4.2-50

4. 2. 3.'2.d 3 Channel Evaluation 4.2-50 4.2.3.2.14'< Fuel Shipping and Handling 4.2-50 4.2.3.2.15 Fuel Assembly - SSE and LOCA Loadings 4.2-50

! 4.2.3.3 Reactivity Control Assembly ,

Evaluation (Control Rods) 4.2-50 4.2.3.3.1 Materials Adequacy Throughout Design Liferime 4.2-50 4.2.3.3.2 Dimensional und Tolerance Analysis 4.2-50 4.2.3.3.3 Thermal Analysis of the Tendency to Warp '

4.2-51 4.2.3.3.4 Forces for Expuision 4.2-51 4.2.3.3.5 Effect of Fuel Rod Failure on Control Rod Channel Clearances 4.2-51 4.2.3.3.6 Effect of Blowdown Loads on Control Rod Channel Clearances 4.2-52

() 4.2.3.3.7 4.2.3.3.7.1 Mechanical Damage First Mode of Failure 4.2-52 4.2-53 4.2-iii

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CONTENTS (Continued)

Section Title Page 4.2.3.3.7.2 Second Mode of Failure 4.2-53 4.2.3.3.8 Analysis of Guide Tube Design 4.2-54 4.2.3.3.9 Evaluation of Control Rod Velocity Limiter 4.2-55 4.2.4 Testing and Inspection 4.2-55 4.2.4.1 Fuel, liardware and Assembly 4.2-55 4.2.4.2 Testing and Inspection (b'nrichment and Burnable Poison Concentrations) 4.2-57 4.2.4.2.1 Enrichment Control Program 4.2-57 4.2.4.2.2 Gadolinia Inspections 4.2-57 4.2.4.2.3 Reactor Control Rods 4.2-58 4.2.4.3 Surveillance Inspection and Testing of Irradiated Fuel Rods 4.2-59 4.2.5 Operating and Devslopmental Experience 4.2-61 4.2.5.1 Fuel Operating Experience 4.2-61 4.2.5.2 Fuel Development Experience 4.2-63 4.2.5.3 Fuel Rod Perforation Experience 4.2-65 4.2.5.4 Channel Operating Experience 4.2-66 4.2.6 References 4.2-67 0

4.2-iv

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

() SECTION 4.2 TABLES Tabl,e ~ Title Page 4.2-1 Fuel Cladding Conditions of Design Resulting From In-Reactor Process Conditions Combined with Earthquake Loading 4.2-71 4.2-2 Fuel Cladding Stress Intensity Limits 4.2-72 4.2-3 Fuel Cladding Estimated Number of Cycles per Each Cyclic Condition Used for Fatigue Analysis 4.2-73 4.2-4 Fuel Data 4.2-74 4.2-5 Material' Properties 4.2-75 4.2-6 Post-Shipment Fuel Inspection Plan 4.2-76 4.2-7 Inspection Equipment 4.2-77 4.2-8 Summary of Experience in Production Zircaloy-Clad UO2 Fuel 4.2-7G 4.2-9 Summary of General Electric' Operating Experience with Production Gadolinia-Bearing

() 4.2-10 Fuel General Electric Developmental Irradiating 4.2-81 Zircaloy-Clad 95% TD UO Pellet Fuel Rods 4.2-83 2

4.2-11 General Electric Developmental Irradiations Zircaloy-Clad 95% TD 00 Pellet Capsules GE 2

Test Reactor 4.2-84 4.2-12 Halden Irradiation Program Status 4.2-85 4.2-13 Fuel Rod Vibration Information 4.2-86 4.2-14 Linear Heat Generation Rate of Calculated 1%

Plastic Diametral Strain for BWR/6 Fuel 4.2-87 ILLUSTRATIONS Figure Title Page 4.2-1 Schematic of Four-Bundle Cell Arrangement 4.2-89 4.2-2 Fuel Assembly 4.2-90 4.2-3 Fuel Assembly Cross Section 4.2-91 4.2-4a Control Rod Assembly 4.2-92 4.2-4b Control Rod Information Diagram 4.2-93

(

4.2-v

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O ILLUSTRATIONS (Continued)

Figure Title Page 4.2-5 Control Rod Velocity Limiter 4.2-94 4.2-6 Fuel Cladding Average Temperature at a Fuel Column Axial Gap 4.2-95 4.2-7 Cladding Temperature versus Heat Flux, Beginning-of-Life 4.2-96 4.2-8 Cladding Temperature versus Heat Flux, End-of-Life 4.2-96 4.2-9 Fuel Energy Release as a Function of Time 4.2-97 O

O 4.2-vi

v_. , _ _ . _. _ --_._ __ _ . _

GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 0 Og 4.2 FUEL SYSTEM DESIGN 4.2.1 General and Detailed Design Bases-4.2.1.1 General Design Bases ,

\

The following paragraphs' define the general mechanical design bases that are considered in defining the design of the fuel assembly and its components and the control rod assembly and its components. In addition, the fuel design meets the Technical Specification limits on Linear IIcat Generation Rate (LIIGR) .

4.2.1.1.1 Fuel Assembly and Its Components 4.2.1.1.1.1 Safety Design Bases The fuel assembly is designed to ensure, in conjunction with the

()

\-

core nuclear characteristics (Section 4.3), the core thermal and hydraulic characteristics (Section 4.4), the plant equipment char-acteristics and the instrumentation and protection system, that fuel damage will not result in the release of radioactive materials in excess of the guideline values of 10CFR20, 50, and 100.

The mechanical design process emphasizes that:

(1) the fuel assembly provides substantial fission product retention capability during all potential operational modes, and (2) the fuel assembly provides sufficient structural integrity to prevent operational impairment of ar.y reactor safety equipment.

~

4.2-1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.1.1.1.1 Safety Design Bases (Continued)

Assurance of the design bases considerations is provided by the O

following fuel assembly capabilities:

o Pressure and Temperature Capabilities The fuel assembly and its components are capable of withstanding the predicted thermal, pressure, and mcchanical interaction loadings occurring during startup testing, normal operation, and abnormal opera-tion without impairment of operational capability.

e Handling Capability The fuel assembly and all fuel assembly components are capable of withstanding loading predicted to occur dur-ing normal handling without impairment of operational capability.

e Earthquake Loading Capability (2/3 SSE)

The fuel assembly and all fuel assembly components are capable of sustaining in-core loading predicted to occur from an operating basis earthquake (OBE), when occurring during normal operating conditions without impairment of operational capability.

e Earthquake Loading Capability (SSE)

The fuel assembly and all fuel assembly components are capable of sustaining in-core loading predicted to occur from a Safe Shutdown Earthquake (SSE) when occur-ring during normal operation without:

(1) exceeding deflection limits which could hinder control rod insertion, or 4.2-2

GESSAR II 22A7007 238 NUCLEAR ISLAtiD Rsv. 0 4.2.1.1.1.1 Safety Design Bases (Continued)

(2) fragmentation or severance of any bundle component.

e Accident Capability The capability of the fuel assembly to withstand the control rod drop accident, pipe breaks inside contain-ment accidents, fuel-handling accident, recirculation pump seizure accident, and pipe breaks outside the containment accidents is determined by analysis of.the specific event.

The ability of the fuel assembly and its components to provide the preceding capabilities is evaluated by one or more of the following:

(1) Analyses developed and design. ratios formulated to measure results against acceptance criteria (Subsec-tion 4.2.1.1.1.3).

(2) Analytical procedures based upon classical methods which are patterned after the ASME Boiler and 4

Pressure Vessel Code Section III. This procedure l allows analytical comparisons of new and old designs and maintains consistency of design characteristics (Subsection 4.2.1.1.1.4).

(3) Experience and testing (Subsections 4.2.1.2, 4.2.4, and 4.2.5).

O 4.2-3

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.1.1.1.2 Basis For Fuel Rod Safety Evaluation 9

Fuel damage is defined as a perforation of the fuel rod cladding which would permit the release of fission products to the reactor coolant.

The mechanisms 'vhich could cause fuel damage in reactor operational transients and which were censidered in fuel evaluations are:

(1) rupture of the fuel rod cladding due to strain caused by relative expansion of the UO 2 Pellet, and (2) severe overheating of the fuel rod cladding caused by inadequate cooling (Sub-section 4.2.1.2.1.5).

4.2.1.1.1.3 Design Ratios Design ratios are defined by the following relationship: D.R. = A/L, where D.R. is the design ratio, L is the limi ting parameter value, and A is the actual parameter value. Design ratios of less than one are demonstrated fo'r component parameters influenced by load-ing conditions which may affect the structural or dimensional integrity of the fuel assembly or any fuel assembly component.

(S ) Limiting Parameter Values

a. Normal and Upset Design Conditions Limiting parameter values for each component are determined in the following manner as defined by Table 4.2-1.

For stress resulting from mean value or steady-state loading, the limiting value is determined by consideration of the material 0.2% offset yield strength or the equivalent strain, as established at operating temperature.

O 4.2-4

GESSAR II 22A7007 238 MUCLEAR ISLAND Rev. 0 4.2.1.1.1.3 Design Ratics (Continued) j For stress resulting from load cycling, limiting parameter values are determined from fatigue limits.

For stress resulting from loading of significant duration, the limiting. parameter is determined from considerations of stress rupture as defined by the Larson-Miller parameters. If metal temperatures are below the level of applicability of stress rupture for the material, or if the yield strength is more limiting, then the limit-ing value of stress is determined from considera-tion of the material 0.2% offset yield strength or the equivalent strain, as established at operating temperatures.

Where stress rupture and fatigue cycling are both significant, the following limiting condition is-applied:

n actual time at stress

}' allowable time at stress I=1 m

, y, actual numbercycles allouable of cycles at stress at stress

-<y I=1 Critical instability loads are derived from test data, when available, or from analytical methods when applicable test data are not available.

O 4.2-5

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.1.1.1.3 Design Ratios (Continued)

Deflection limits are those values of component O

deformation which could cause an undesirable event such as impairment of control rod movement or an excessive bypass flow rate.

b. Emergency and Faulted Design Conditions Limiting parameter values are determined in the following manner, as defined by Table 4.2-1:

(1) Stress limits are determined from considera-tion of the ultimate tensile strength or equivalent strain of the material, as established at operating temperatures.

(2) Critical instability loads are determined from test data when available or from analyt-ical methods when applicable test data are not available.

(3) Deflection limits are those values of deforma-tion that, if occurring, could lead to a more serious consequence such as prevention of control rod insertion.

(2) Actual Parameter Values l

Actual parameter values are determined from the following considerations:

a. Effective stresses are detarmined at each point of interest using the theory of constant elastic strain energy of distortion:

O 4.2-6 i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O r~S 4.2.1.1.1.3 Design Ratios (Continued)

()

2 " I ~

+ +

e X Y) ("Y ~ "Z) I"Z ^ X 2

+ 6 (T y+3 YZ + 7 ZX

b. Stress concentration is applied only to the alternating stress component.

c. Design values of instability loads are scaled up to allow for uncertainty in manner of load appli-cation, variation in modulus of elasticity, and difference between the actual case and theoretical one,

d. Calculated values of deflection for comparison with deflection limits is based on the resulting (n) permanent set after load removal; if load removal occurs before, damage may result.

4.2.1.1.1.4 Maximum Allowable Stresses, Cycling and Fatigue Limits The strength theory, terminology and stress categories presented in the ASME Boiler and Pressure Vessel Code,Section III, are used as a guide in the mechanical design and stress analysis of the reactor fuel rods. The mechanical design is based on the maximum shear stress theory for combined stresses. The equiva-lent stress intensities used are defined as the difference between the most positive and least positive principal stresses in a triaxial field. Thun, stress intensities are directly com-parable to strength values found from tensile tests. Tabic 4.2-2 presents a summary of the basic stress intensity limits that are applied for Zircaloy-2 cladding.

l d'

4.2-7

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.1.1.1.4 Maximum Allowable Stresses, Cycling and Fatigue Limits (Continued) l)

The fatigue analysis utilizes the linear cumulative damage rule (Miner's hypothesis, Reference 1) and the Zircaloy fatigue design basis of Reference 2. This correlation includes a safety factor of 2 on stress or 20 on cycles (whichever is more conservative).

The fatigue analysis is based on the cycles shown in Table 4.2-3.

The expected time duration for each of the subject cyclic load-ings is not specified and, for the startup and reduced power cycles, can vary according to the reactor status and power demand.

The cyclic condition relating to overpower transients would result from an operator error or equipment malfunction and would, therefore, be expected to be of short duration (less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />).

Additional information regarding the basis for fatigue analysis is presented in Section 6 of Reference 4.

4.2.1.1.2 Control Assembly and Its Components Safety Design Bases The reactivity control mechanism design includes control rods and gadolinia burnable poison in selected fuel rods within fuel assemblies and meets the following safety design bases:

(1) The control rods shall have sufficient mechanical strength to prevent displacement of their reactivity control material.

(2) The control rods "'all have sufficient strength and be l so designed as to prevent deformation that could inhibit their motion.

i I

I 4.2-8

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.1.1.2 Control Assembly and Its Components (Continued)

O O

(3) Each control rod shall have a device to limit its free-fall velocity sufficiently to avoid damage to the nuclear system process barrier by the rapid reactivity increase resulting from a free fall of the control rod from its fully inserted position to the position to which the control rod drive was withdrawn.

The design basis of the initial core supplementary fuel /

reactivity control rods (UGd) O2 is the same as UO2 fuel rods.

Additional information on urania-gadolinia physical and irradia-tion characteristics and material properties is provided in Reference 31.

4.2.1.2 Detailed Design Bases 4.2.1.2.1 Fuel Assembly and Its Components 7-s b

The following paragraphs present the detailed bases which are considered in defining the design of the fuel assembly and its components.

4.2.1.2.1.1 Material Selection and Properties

{

The materials will be compatible with BWR conditions and retain their design capability during reactor operation. The mechanical, chemical, thermal and radiation properties utilized as design bases are presented in Section 3 of Reference 3. The basic materials used in fuel assemblies are Zircaloy-2 and Zircaloy-4, i

Type-304 stainless steel, Alloy X-750 and ceramic uranium-dioxide and gadolinia.

i 4.2-9 I

i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.1.2.1.2 Effects of Irradiation (1) Cladding Properties, Fuel Swelling Irradiation affects both fuel and cladding material properties. The effects include increased cladding strength and reduced cladding ductility. In addition, irradiation in a thermal reactor environment results in the buildup of both gaseous and soli 1 fission products within the UO fu 3 p 11 t which tend to 2

increase the pellet diameter (i.e., fuel irradiation swelling). The irradiation swelling model is based on data reported in References 5 and 6, as well as an evaluation of all applicable high exposure data (Reference 7). Pellet internal porosity and pellet-to-cladding gap are specified such that the thermal expan-sion and irradiation swelling are accommodated for the worst-caso dimensional tolerances throughout life.

g Observations and calculations based on this refined model for relative UO 2 fuel / cladding expansion indi-cate that the as-fabricated UO p 11 t porosity is 2

adequate (without pellet dishing) to accommodate the fission product induced UO sw 11ing ut to, and beyond, 2

the peak exposures expected.

(2) Fuel Pellet-to-Cladding Gap and Gap Conductance The primary purpose of the gap between the UO f" I 2

pellet and Zircaloy cladding is to accommodate differential diametral expansion of fue: pellet and cladding and, thus, preclude the occurrence of exces-sive gross diametral cladding strain. A short time after reactor startup, the fuel cracks radially and redistributes out to the cladding. Experience has 4.2-10

.GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 0 4

4.2.1.2.1.2 Effects of Irradiation (Continued)

(U)

4 shown, however, the gap volume remains available in the form of radial cracks to accommodate' gross j diametral fuel expansion.

, The value of pellet-to-cladding thermal conductance f used in the fuel design is 1000 Btu /hr-ft 2 *F. This J

design value is empirically derived from post-irradiation data on exposed fuel with an initial pellet-to-cladding gap which-is significantly larger than that employed in the General Electric fuel design.

The use of the constant value of 1000 Btu /hr-ft - F for the pellet-cladding. thermal conductance was found to be a conservative assumption when applied in conjunction with the integral fuel design models.

(} Specifically, the design fission gas release model employed in the determination of fuel rod plenum size and cladding wall thickness has been shown to over-predict available data on fis' ion gas release when applied with a pellet-cladding thermal conductance model for relative fuel cladding expansion (pellet-to-f cladding interaction) also has been shown to be very I

conservative relative to available data when a value of 1000 Btu /hr-ft *F is used for pellet-cladding thermal conductance (Reference 7). Additional discus-sion and evaluation of the pellet-to-clad gap conduc-tance of GE BWR fuel prepressurized to 3 atmospheres is contained in References 34 through 37.

~f V

3 4.2-11

. .,,,,. -, -. ;, . ! , . . .- - , , .- - - - - - , - --,, . ,. . - - . . - . . , = - - - - . , - - - , - . . . - - . - - - ,

GESSAR Il 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.1.2.1.2 Effects of Irradiation (Continued)

(3) Axial Ratcheting Axial ratcheting of fuel cladding 1s not considered in the fuel rod design. Prototypical fuel rods have been operated in the Halden test reactor with axial elonga-tion transducers. No significant axial ratcheting was observed (Reference 8).

(4) Fuel Melting Temperature Fission product buildup tends to cause a slight reduc-tion in fuel melting temperature. The melting point of UO decreases with irradiation at the rate of 32 C/

2 10,000 mwd /Te based on data from Reference 9.

(5) Fuel Thermal Conductivity In the temperature range of interest (500 C), the fuel thermal conductivity is not significantly affected by irradiation (Reference 10).

(6) Fission Gas Release A small fraction of the gaseous fission products is

• / released from the fuel pellets to produce an increase in fuel rod internal gas pressure. In general, such irradiation effects on fuel performance have been characterized by available data and were considered in determining design features and performance. Thus, the irradiation effects on fuel performance were inherently accounted for when it was determined ahether or not the stress intensity limits and temperature limits were satisfied.

4.2-12

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O g

! i 4.2.1.2.1.3 Plow-Induced Vibration L.)

Flow-induced fuel rod vibrations depend on such parameters as flow velocity, fuel rod geometry, fuel spacer pitch, fundamental rod frequency and the excitation forces due to fluctuating pres-cures. The stresses resulting from flow-induced vibrations were considered in the mechanical design and evaluations of the fuel rods. These stresses were compared to stress intensity limits as noted in Subsection 4.2.1.1.1.4.

The flow-induced vibration affecting other fuel assembly com-ponents was based upon operational experience which, to date, has shown no significant adverse effects.

4.2.1.2.1.4 Fuel Densification The fuel densification design bases include the effects of:

O (1) power spikes due to axial gap formation; (2) increase in LHGR

(~~' ) due to pellet length shortening; (3) creep collapse of the cladding due to axial gap formation; and (4) changes in stored energy due to decreased pellet-cladding thermal conductance resultir.g from increased radial gap size. The General Electric fuel densification models are described in References 11, 12, and

13. References 11 and 12 contain a description of the most recent densification models, as modified and approved by the NRC.

Analyses of the effects of fuel densification on the design are contained in Subsection 4.2.3.2.8.

4.2.1.2.1.5 Fuel Rod Damage Mechanisms As noted in Subsection 4.2.1.1.1.2, the mechanisms which could cause fuel damage arc (1) rupture of the fuel rod cladding due to strain caused by relative expansion of the UO., pollet, and s

(2) severe overheating of the fuel rod cladding due to inadequate

[~ h cooling.

LJ 4.2-13

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.1.2.1.5 Fuel Rod Damage Mechanisms (Continued)

A value of 1% plastic strain of the Zircaloy cladding is deftned as the limit below which fuel damage due to overstraining of the fuel cladding is not expected to occur. The 1% plastic strain value is based on General Electric data on the strain capability of irradiated Zircaloy cladding segments from fuel rods operated in several BWRs (Reference 7). None of the data obtained falls below the 1% plastic strain value; however, a statistical distri-bution fit to the available data indicates the 1% plastic 7 train value to be approximately the 95% point in the total popu? tion.

This distribution implies, therefore, a small (<5%) probability that some cladding segments may have plastic clongation less than 1% at failure.

The Fuel Cladding Integrity Safety Limit (Reference 32) ensures that fuel damage due to severe overheating of the fuel rod clad-ding, caused by inadequate cooling, is avoided.

4.2.1.2.1.6 Dimensional Stability The fuel assembly and fuel components are designed to assure dimensional stability in service. The fuel cladding and channel specifications include provisions to preclude dimensional changes due to residual stresses. In addition, the fuel assembly is designed to accommodate dimensional changes that occur in service due to thermal differential expansion and irradiation effects. For example, the fuel rods are free to expand axially independent of one another.

4.2.1.2.1.7 Fuel Shipping and Handling During shipment, the fuel bundle is in a horizontal position with flexible packing separators installed between the fuel rods so that the weight of the fuel rods is supported by the shipping container rather than the spacers.

4.2-14

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

\ 4.2.1.2.1.7- Fuel Shipping and Handling (Continued)

Fuel bundle shipping procedures are qualified by a test per-formed on each design, and each individual bundle is inspected relative to important dimensional characteristics following shipment to verify that no dimensional deviations have-occurred.

The two major handling loads considered are (1) the loads due to maximum upward acceleration of the fuel assembly while grappled, and (2) the loads due to impact of the fuel assenbly into the fuel support while grappled.

4.2.1.2.1.8 Capacity for Fission Gas Inventory A plenum is provided at the top of each fuel rod to accommodate the fission gas released from the fuel during operation. The

( design basis is to provide sufficient volume to limit the fuel

) rod internal pressure so that cladding stresses do not exceed the limits given in TSble 4.2-2 during normal operation, and for short-term transients of 16% or less above the peak normal operating conditions.

(1) Fuel Rod Internal Pressure 1

( Fuel rod internal pressure is due to the helium which is backfilled during rod fabrication, the volatile con-tent of the UO 2, and the fraction of gaseous fission j products which are released from the UO The most 2

i limiting combination of dimensional tolerances is f assumed in defining the hot plenum volume used to

! compute fuel rod internal gas pressure. The available fission gas retention volume is conservatively deter-  :

i mined and the fuel rod internal pressure is calculated using the perfect gas law.

i 4.2-15 i

.,,-..e., . m- ._w,- ~ .,--...-,um._..w,,_,,_,_,m. ,--.,,mmm.,, -

v._.- ~ ,__.,.w,y,,,%,_ , . . , , -

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.1.2.1.3 Capacity for Fission Gas Inventory (Continued)

(2) Fission Gis Generation and Release A quantity of 1.35 x 10 -3 gram moles of fission gas is produced per mwd of power production. In fuel rod pressure and stress calculations, 4% of the fission Jas produced is assumed to be released from any UO 2

volume at a temperature less than 3000 F and 100%

released from any U0 above 3000 F. The above basis 2

has been demonstrated by experiment to be conservative over the complete range of design temperature and exposure conditions (Reference 7).

(3) Plenum Creepdown and Creep Collapse Creepdown and creep collapse of the plenum are not considered because significant creep in the plenum region is not expected. The fuel rod is designed to be free-standing throughout its lifetime. The temper-ature and neutron flux in the plenum region are con-siderably lower than in the fueled region; thus, the margin to creep collapse is substantially greater in the plenum. Direct measurements of irradiated fuel rods have given no indication of significant creepdown of the plenum.

4.2.1.2.1.9 Deflection The operational fuel rad deflections considered are the deflections due to:

(1) manufacturing tolerances; (2) flow-induced vibration:

4.2-16

i-GESSAR II- 22A7007 238 NUCLEAR ISLAND Rev. 0 a.

( 4.2.1.2.1.9 Deflection (Continued)

I (3) thermal effects; and 4

4 (4) axial load.

1

Two criteria limit the magnitude of these deflections

(1) the cladding stress limits must be satisfied; and (2) the fuel rod-to-j fuel rod and fuel rod-to-channel clearances must be sufficient i to allow free passage of coolant water to all heat transfer

, surfaces.

7 The fuel rod-to-fuel rod spacing limit of 0.060 in, and fuel rod-to-channel spacing limit of 0.030 in. are based upon the range of clearances that have, in-the past, been used in boiling transition testing. More recent testing to clearances below these values would indicate that a lower limit is acceptable.

t 4.2.1.2.1.10 Fretting Wear and Corrosion i

Fretting wear and corrosion have been considered in establishing

the fuel mechanical design basis. Individual rods in the fuel assembly are held in position by spacers located at intervals i

along the length of the fuel rod. Springs are provided in each j spacer cell so that the fuel rod is restrained to avoid excessive l vibration.

r 4.2.1.2.1.11 Potential for Water-Logging Rupture l

For waterlogging to occur, the fuel cladding must have a small pinhole. Pinholes are eliminated during production by 100% leak check of fuel assemblies. The Leak Detector System consists of a high vacuum system capable of attaining pressures less than

~4 1 x 10 torr, and a mass spectrometer capable of detecting leaks

() as low as 2 x 10

-11 cc/sec. Each fuel bundle is placed in the vacuum chamber, and evacuated to less than 1 x 10 torr. After 4.2-17

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.1.2.1.11 Potential for Water-Logging Rupture (Continued) lh the vacuum pressure is attained, the mass spectrometer tuned to the helium mass range is switched into the system. The output meter of the mass spectrometer indicates the presence of any helium gas in the chamber. This production procedure for the fuel is considered to preclude the potential for a waterlogging rupture.

4.2.1.2.1.12 Potential for Hydriding The fuel design bases relative to the clad hydriding mechanism assure, through a combination of engineering specifications and strict manufacturing controls, that production fuel does not contain excessive quantities of moisture or hydrogenous impur-ities. An engineering specification limit on moisture content in a loaded fuel rod is defined which is well below the threshold of fuel failure. Procedural controls are utilized in manufactur-ing to prevent introduction of hydrogenous impurities such as oils, plastics, etc., into the fuel rod. Hot vacuum outgassing (drying) of each loaded fuel rod just prior to final end plug welding is employed to assure that the level of moisture is well below the specification li'..it. As a further assurance against chemical attack from the inadvertent admission of moisture or hydrogenous impurities into a fuel rod during manufacture, a hydrogen getter material is employed in the upper plenum of all f.el rods. This getter material has been proven effective in both in-pile and out-of pile tests. The getter material is a zirconium alloy in the form of small chips. These getter chips are loosely packed in a stainless steel tube of which one end is capped, and the other end is covered by wire screening. Additional information regarding the getter is presented in Section 8 of Reference 4.

O 4.2-18

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

! 4.2.1.2.1.13 Stress-Accelerated Corrosion L)'

Stress corrosion cracking, the phenomenon whereby a ductile material, such as Zircaloy-2, experiences nonductile fracture, has been identified as a factor in pellet-cladding interaction fuel failure. The simultaneous action of certain corrosive agents and local stresses for an extended period of time has been observed to embrittle Zircaloy-2 at temperatures typical of those achieved in light water reactors. Samples of Zircaloy-2 fractured in the presence of cadmium or iodine in out-of-pile tests, for example, show very little reduction in area and the fracture surface appears non-ductile. Pellet-cladding inter-action type failures also exhibit very little reduction in area and a nonductile fracture surface appearance. Complete under-standing of the influence of stress corrosio: over certain time domains, stress levels, environments, and cladding exposure on fuel cladding damage is limited by a developing technology.

b) Consequently, stress-accelerated corrosion is not directly addressed in design analyses assessing the impact of the pellet-cladding interaction mechanism. Ilowever, qualitative observa-tions of fuel failures and their relationship to steady-state linear power resulted in the fuel design change to an 8x8 fuel rod matrix to reduce linear powerc.

4.2.1.2.1.14 Fuel Reliability The fuel component characteristics which can influence fuel reliabi'ity include: (1) fuel pellet thermal and mechanical properties, dimensions, denrity, and U-235 enrichment; (2) Zircaloy cladding thermal and mechanical properties, dimen-sions, and defects; (3) fuel rod internal void volume and impurities; (4) fuel rod-to-fuel rod and fuel rod-to-channel spacing; and (5) spring constants of the fuel rod spacer springs which maintain contact between the spacer and the fuel (n)

L.J rods. Important fuel pellet, cladding, and associated hardware 4.2-19

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.1.2.1.14 Fuel Reliability (Continued) characteristics, and dimensions are provided in Table 4.2-4 and Figure 4.2-3.

The large volume of irradiation experience to date with GE BWR fuel indicates only a few mechanisms which have actually had a direct impact on fuel reliability; namely, cladding defects, excessive deposition of system corrosion products, cladding hydriding resulting from hydrogen impurity, and pellet-cladding interaction.

The cladding defects have been virtually eliminated through implementation of improved quality inspection equipment and more stringent quality control requirements during fuel fabri-cation. Excessive deposition of corrosion products has also been virtually eliminated through improved control of corrosion product impurities in the reactor feedwater.

Cladding hydriding is the result of excessive amounts of hydrogenous impurities (moisture and/or hydrogenous material) inadvertently introduced into the rod during the fuel fabrica-tion process. The fuel fabrication process currently includes the following steps to minimize possible failures from this mechanism: (1) drying of components and pellets prior to rod loading; (2) hot vacuum outgassing of all loaded fuel rods prior to the final end-plug weld; and (3) strict control of hydrogenous materials during fabrication. In addition, as noteC in Subsection 4.2.1.2.1.12, every fuel rod contains supplementary protection in the form of a hydrogenous impurity getter which is placed in the plenum.

In early 1972, General Electric made design changes in the 7x7 l

fuel to reduce the incidence of pellet-cladding interaction in future production. The " improved 7x7" design incorporated a 4.2-20

[

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

("')

V 4.2.1.2.1.14 Fuel Reliability (Continued) reduced pellet length-to-diameter ratio, chamfered pellet ends and the climination of pellet dishing to reduce the magnitude of pellet distortions contributing to local cladding strains.

This design also employed an increased c.'. adding heat treatment temperature to reduce the statistical variability in cladding mechanical properties. Additional information regarding this cladding material is provided in Section 4 of Reference 4.

These short-term design changes have been coupled with the longer term design effort which culminated in the 8x8 design, which was introduced into operating reactors in the spring of 1974. With the 8x8 fuel, peak linear power is reduced by more than 25% relative to the 7x7 fuel design to address the strong dependence of PCI failures on bundle pcwer. The favorable fuel performance of both early General Electric BWR fuel designs with low linear heat rates, and current 8x8 reload fuel provides

/ assurance of the improved reliability of the 8x8 fuel design.

( ))

4.2.1.2.1.15 Design Basis for Fuel Assembly Surveillance General Electric maintains an active fuel assembly surveillance program specifically intended to monitor performance in operating reactors to identify and characterize unexpected phenomena which can influence fuel integrity and performance. Outage-oriented examinations are performed contingent on fuel availability as influenced by plant operation. Typically, peak duty fuel assemblies (with respect to exposure, linear heat generation rate, and the combination of both) are designated as lead assemblies for a particular design, and are selectively inspected.

Numerous other assemblics are routinely inspected employing the nondestructive techniques discussed in Subsection 4.2.4.3.

Additional information regarding fuel surveillance is contained in Subsection 4.2.4.3.

4.2-21

GCSSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.1.2.2 Control Assembly and Its Components The following paragraphs present the detailed bases which are considered in defining the design of the control assembly and its components.

4.2.1.2.2.1 Design Acceptability The acceptability of the control rod and control rod drive under scram loading condition has been demonstrated by functional testing instead of analysis or adherence to formally defined stress limits.

4.2.1.2.2.2 Control Rod Clearances The basis of the mechanical design of the control rod clearances is that there shall be no interference which will restrict the passage of the control rod.

Layout studies are performed to assure that, given the worst combination of extreme detail part tolerance ranges at assembly, no interference exists which will restrict the passage of control rods. In addition, preoperational verification will be made on each control rod system to demonstrate that the acceptable levels of operational performance are met.

4.2.1.2.2.3 Mechanical Insertion Requirements Mechanical insertion requirements during normal operation have been selected which will provide adequate operability and load following capability, and which will control the reactivity addition result-ing from burnout of peak shutdown xenon at 100% power.

Scram insertion requirements are chosen to provide sufficient negative reactivity to meet all safety criteria for plant operational transients.

4.2-22

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

/-

(\)i

\

4.2.1.2.2.4 Material Selection The selection of materials for use in the control rod design is based upon their in-reactor properties. The irradiated proper-tice of Type-304 austenitic stainless steel which comprises the major portion of the assembly, B C 4 powder, alloy X-750, and Stellite, or its equivalent, are well known and are taken into account in establishing the design of the control rod components.

The basic cruciform control rod design and materials have been operating successfully in all General Electric reactors.

4.2.1.2.2.5 Radiation Effects The corrosion rate and the physical properties (e.g., density, modulus of elasticity, dimensional aspects, etc.) of austenitic stainless steel, and Allcy X-750 are essentially unaffected by the irradiation experienced in the BWR reactor core. The effects upon D) t we the mechanical properties (i.e., yield strength, ultimate tensile strength, percent elongation and ductility) on the Type-304 stain-less steel cladding also are well known and were considered in mechanical design. The radiation effects on B C powder include 4

the release of gaseous products and swelling. The B4 C cladding is designed to sustain the resulting internal pressure buildup due to gaseous products, and the lifetime of the control rod has been established to minimize the effects of swelling.

4.2.1.2.2.6 Positioning Requirements Rod positioning increments (not lengths) are selected to provide adequate power-shaping capability. The combination of rod speed and notch length must also meet the limiting reactivity addition rate criteria.

(3

(

v) 4.2-23

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.2 General Design Description 4.2.2.1 Core Cell A core cell consists of a control rod and the four fuel assemblies which immediately surround it. Figure 4.3-2 provides core cell dimensions. Each core cell is associated with a four-lobed fuel support piece. Around the outer edge of the core, certain fuel assemblies are not immediately a'ljacent to a control rod and are supported by individual peripheral fuel support pieces.

The gridwork portion of the top guide is a solid stainless steel plate machined to accommodate 4-bundle cells. The fuel assemblies are lowered into the cell and, when seated, springs mounted at the tops of the channels force the channels into the corners of the cell such that the sides of the channel contact the grid beams (Figure 4.2-1).

4.2.2.2 Fuel Assembly A fuel assembly consists of a fuel bundle and the channel which surrounds it (Figure 4.2-2). The fuci .ssemblies are arranged in the reactor core to approximate a right circular cylinder inside the core shroud. Each fuel assembly is supported by a fuel support piece and the top guide.

The general configuration of the fuel assembly and the detailed configurations of the assembly components are the results of the evolutionary change in customer, performance, manufacturing, and serviceability requirements and the experience obtained since the initial design conception. A summary of fuel assembly mechanical data is presented in Table 4.2-4.

O 4.2-24

(

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.2.2.1 Fuel Assembly Orientation Proper orientation of fuel assemblies in the reactor core is readily verified by visual observation and is assured by veri-fication procedures during core loading. Five separate visual indications of proper fuel assembly orientation exist:

l (1) The channel fastener assemblies, including the spring and guard used to maintain clearances betwean I

channels,,are located at one corner of each fuel assembly adjacent to the center of the control rod.

l (2) The ident'ifihation boss on the fuel assembly handle

,' points toward the adjacent control rod.

1 (3) The channel spacing buttons are adjacent to the control

! rod passage area.

O (4) The assembly identification numbers, which are located on the fuel assembly handles, are all readable from the direction of the center of the cell.

t (5) There is cell-to-cell replication.

i ,

l Experience has demonstrated that these design features are clearly l visible so that any misoriented fuel assembly would be readily distinguished during core loading verification.

l 4.2.2.3 Fuel Bundle I

A fuel bundle contains 62 fuel rods and 2 water rods which are spaced and supported in a square (8x8) array by 7 spacers and l

the lower and upper tieplates. The lower tieplate has a nose-piece which supports the fuel assembly in the reactor. The upper tieplate has a handle for transferring the fuel bundle from one location to another. The identifying assembly number r 4.2-25

_ _ - _ _ . ~ _ . . _ _ _ - _ - . _ - - - _ . _ _ . _ , _ _ - . _ . . _ . _ _ _ , _ .. ,.. _ . _. .. _ . _ . _ . _ _ . _ -

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.2.3 Fuel Bundle (Continued) is engraved on the top of the handle and a boss projects from one side of the handle to aid in assuring proper fuel assembly orientation. Both upper and lower tieplates are fabricated from Type-304 stainless steel castings. Finger springs, of the same design previously used with 7x7 and 8x8 initial core and reload fuel, are also employed with the BWR/6 fuel design. The finger springs are located between the lower tieplate and the channel for the purpose of controlling the bypass flow through the flow-path (Subsection 4.2.2.3.6). Zircaloy-4 fuel rod spacers equipped with Alloy X-750 springs maintain fuel rod-to-fuel rod spacing.

4.2.2.3.1 Fuel Rods Each fuel rod consists of high density (95% TD) UO 2 fu 1 pellets stacked in a Zircaloy-2 cladding tube which is evacuated, back-filled with helium at 3 atmospheres pressure, and sealed by Zircaloy end plugs welded in each end. The 150-in. active fuel column includes a 6-in. zone of naturally enriched (0.711 wtt U-235) pellets at both the top and bottom. The fuel rod cladding thickness is adequate to be essentially free-standing under the 1000 psia BWR environment. Adequate free volume is provided within each fuel rod in the form of pellet-to-cladding gap and a plenum region at the top of the fuel rod to accommodate thermal and irradiation expansion of the UO and the internal pressures 2

resulting from the helium fill-gas, impurities and gaseous fission products liberated over the design life of the fuel. A plenum spring, or retainer, is provided in the plenum space to prevent movement of the fuel column inside the fuel rod during fuel shipping and handling (Figure 4.2-2). A hydrogen getter also has been provided in the plenum space as assurance against chemical attack from the inadvertent admission of moisture or hydrogenous impurities into a fuel rod during manufacture.

Two types of fuel rods are utilized in a fuel bundle: tie rods and standard rods (Figure 4.2-3). The eight tie rods in each 4.2-26 1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O I 'h 4.2.2.3.1 Fuel Roos (Continued)

V bundle have lower end plugs which thread into the lower tieplate casting and threaded upper end plugs which extend through the upper tie plate casting. A stainless steel hexagonal nut and locking tab are installed on the upper end plug to hold the fuel bundle together. These tie rods support the weight of the assembly only during fuel handling operations when the assembly hangs by the handle; during operation, the fuel rods are supported by the lower tieplate. Fifty-four rods in the bundle are standard fuel rods. The end plugs of the standard rods have shanks which fit into bosses in the tieplates.

An Alloy X-750 expansion spring is located over the upper end plug shank of each rod in the assembly to keep the rods seated in the lower tieplate while allowing independent axial expansion by sliding within the holes of the upper tieplate. Additional I information concerning the fuel rod expansion spring is provided C) in Section 7 of Reference 4.

The fuel bundles incorporate the use of small amounts of gadolinium as a burnable poison in selected standard fuel rods.

The irradiation products of this process are other gadolinium isotopes having low cross sections. The control augmentation effect disappears on a redetermined schedule without changes in the chemical composition of the fuel or the physical makeup of the core. Some assemblies contain more gadolinia than others to improve transverse power flattening. Also, some assemblies contain axially distributed gadolinia to improve axial power flattening, Gd 0 s uniformly distributed in the UO 2 P"1

• 23 and forms a solid solution. The gadolinia-urania fuel rods are fabricated using characteristic extended end plugs. These extended end plugs permit a positive visual check on the location of each gadolinia-bearing rod after bundle assembly, i

m 4.2-27

GESSAR II 22A7007

] 238 MUCLEAR ISLAMD Rev. O I,

4.2.2.3.1.1 Fuel Pellets l The fuel pellets consist of high density ceramic uranium-dioxide  ;

l r

j manufactured by compacting and sintering uranium-dioxide powder l

) into right cylindrical pellets with flat ends and chamfered i

edges. Some of the pellets contain small amounts of gadolinia as a burnable poison. The average pellet immersion density is

, approximately 951 of the theoretical density of UO2 Ceramic ,

uranium-dioxide is chemically inert to the cladding at operating temperatures and is resistant to attack by water. Several U-235 enrichments are used in the fuel assemblies to reduce the local i

power peaking factor. Fuel element design and manufacturing i procedures have been developed to prevent errors in enrichment  ;

i locations within a fuel assembly.

l 4.2.2.3.2 Water Rods Two rods in each fuel bundle are hollow water tubes, one of which l (the spacer-positioning water rod) positions the seven Zircaloy-4  ;

fuel rod spacers axially in the fuel bundle. The water rods are made from Zircaloy-2 tubing of slightly targer diameter and l thinner wall than the fuel rods. Several holes are punched l around the circumference of each of the water rods near each end to allow coolant water to flow through the rod. Both water rods have square lower end plugs to prevent rotation. The spacer- l positionir.g water rod is equipped with 14 tabs which are welded to its exterior. The spacer-positioning water rod and fuel l spacers are assembled by sliding the water rod through the appropriate spacer cell with the welded tabs oriented in the .

t direction of the corner of the spacer cell away from the spacer j spring. The rod is then rotated so that the tabs are positioned ,

above and below the spacer structure. The spacer-positioning i rod is prevented from rotating and unlocking the spacers by engagement of its square lower end plug with a square hole in the lower tieplate.

t 4.2-28

,-. . - .___-- - _ _- - . ~. .- . .- , . _.

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O e 4.2.2.3.2 Water Rods (Continued) k Differential thermal expansion between the fuel rods and the water rods can introduce axial loadings into the water rods through the frictional forces between the fuel rods and the spacers. The l testing which was performed to address this condition, and to verify the water rod / spacer conceptual design, is discussed in Section 2 of Reference 4.2.6-3 and in Reference 4.2.6-20.

O  !

j 4.2.2.3.3 Fuel Spacer

! The primary function of the fuel spacer is to provide lateral support and spacing of the fuel rods, with consideration of thermal-hydraulic performance, fretting wear, strength, neutron j economy, and producibility. The spacer represents an optimi-I zation of these considerations. Mechanical design of the BWR/6 spacer is similar, in concept, to that of the current 7x7 and 8x8 spacers.

The mechanical loadings on the spacer structure during normal j operation and transients result from the rod-positioning spacer spring forces, from local loadings at the water rod-spacer posi-tioning device, and a small pressure drop loading. During a seismic event, the spacer must transmit the lateral acceleration I

loadings from the fuel rods into the channel, while maintaining I

the spatial relationship between the rods.

t As noted, the spacer represents a optimization of a number of

{

l considerations. Thermal-hydraulic development effort has gone 3

into designing the particular configuration of the spacer parts.

The resultant configurations give enhanced hydraulic performance.

4 Extensive flow testing has been performed employing prototypical {

spacers to define single-phase and two phase flow characteristics. j i

I During the blowdown portion of the postulated loss-of-coolant accident (LOCA), the hydraulic (pressure differential) forces i

i

, 4.2-29 i

. , , _ , . , , . , , _ , _ , _ . , _ _ _ , . _ . _ _ . _ _ , _ _ m._m. ..-._ , , , , . . , _ , _ , , - - , , . . , - - - . - _ . . . -

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.2.3.3 Fuel Spacer (Continued) ll on the spacer are of about the same magnitude as those present during normal or transient operation of the fuel. There are no significant lateral hydraulic forces on the spacer, because the fuel channel maintains the normal flow path during the blowdown.

4.2.2.3.4 Fuel Channel The fuel channel enclosing the fuel bundle is fabricated from Zircaloy-4 and performs three functions: (1) the channel pro-vides a barrier to separate two parallel flow paths - one for flow inside the fuel bundle and the other for flow in the bypass region between channels; (2) the channel guides the control rod and provides a bearing surface for it; and (3) the channel pro-vides rigidity for the fuel bundle. The channel is open at the bottom and makes a sliding seal fit on the lower tieplate sur-face. At the top of the chann?l, two diagonally opposite corners have welded tabs, which support the weight of the channel from raised posts on the upper tieplate. One of these raised posts has a threaded hole, and the channel is attached using the threaded channel fastener assembly, which also includes the fuel assembly positioning spring. Channel-to-channel spacing is provided for by mear.s of the fuel assembly positioning spring and the spacer buttons which are located on the upper portion of channel adjacent to the control rod passage area. Axial differential expansion between the fuel bundle and its channel is accommodated at the lower tieplate.

In addition to meeting design limits, assurance is provided that the channels maintain their dimensional integrity, strength, and spatial position throughout their lifetime through specifi-cations on the channel materials and manufacturing processes and by quality measurements and process qualifications to ensure compliance with these specifications.

4.2-30

GESSAR II 22A7007 238 NUCLEAR ISLAND Ra'r . 0 4.2.2.3.4 Fuel Channel (Continued)

\

Under situations of adverse tolerance stackup, differential thermal expansion between the stainless steel tieplates and the Zircaloy. channel can result in an interference fit; however, the resultant stress and strain levels in the channel will not exceed design limits. The loads and resultant streas imposed on the fuel channel in the event of control rod interference are also within design limits.

4.2.2.3.5 Tieplates The upper and lower tieplates serve the functions of supporting the weight of the fuel and positioning the rod ends during all phases of operation and handling. The loading on the lower tie-plate during operation and transients is comprised of the fuel weight, the weight of the channel,-and the forces from the expansion springs at the top of the-fuel rods. The loading on the upper tie plate during ope:ation is due to the expansion-spring force. The expansion springs permit differential expan-sion between the fuel rods without introducing high axial forces into the rods.

Most of the loading on the lower tieplate is due to the weight of the fuel rods and the channel, which are not cyclic loadings.

During an accident, the tieplates would be subjected to the normal operational loads plus the loadings due tc the accident, such as blowdown and seismic loadings. During handling, the tieplates are subjected to acceleration and impact loadings.

4.2.2.3.6 Finger Springs a

l Finger springs are employed to control the bypass flow through

! the channel-to-lower tieplate flow path. They hav2 been used in j the initial core and reload fuel of one BWR/3 and all BWR/4 and later plants. They have also been employed on some reload 4.2-31

GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 0 1

4.2.2.3.6 Finger Springs (Continued) fuel in some additional BWR/2 and BWR/3 plants to control bypass flow through the lower tieplate to channel flow path.

Increases in channel wall permanent deflection at the lower tie-plate resulting from creep deformation at opercting conditions result in increased bypass flow through the channel-to-lower-tieplate flow path. Changes in the flow through this patil affect the total core bypass flow, which, in turn, will affect the active coolant flow, void coefficient and operational transients. Finger spring seals are employed to prov-'e control over the flow through this path over a wide range of channel wall deflections by maintaining a nearly constant flow area as the channel wall deforms. The finger springs are located between the lover tie-plate and the channel; a more detailed mechanical description is contained in Section 9 of Reference 4.

4.2.2.4 Reactivity Control Assembly 4.2.2.4.1 Control Rods The control rods perform the dual function of power shaping and reactivity control. A design drawing of the control blade is seen in Figire 4.2-4a and b. Power distribution in the core is controlled during operation of the reactor by manipulating selected patterns of control rods. Control rod displacement tends to counterbalance steam void effects at the top of the core and results in significant power flattening.

The control rod consists of a sheathed cruciform array of stainless steel tubes filled with baron-carbide powder. The control rods are nominally 9.804 in. in total span and are separated uniformly throughout the core on a 12-in. pitch. Each control rod is surrounded by four fuel assemblies.

4.2-32

. = _ , - . _ - . - - .. _.- . .

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

() 4.2.2.4.1 Control Rods (Continued)

=! The main structural member of a control rod is made of Type-304 stainless steel and consists of a top handle, a bottom casting with a velocity limiter and control rod drive coupling, a j vertical cruciform center post, and four U-shaped absorber tube 3

sheaths. The top handle, bottom casting, and center post are welded into a single skeletal structure.

1 The U-shaped sheaths are resistance welded to the center post, handle and castings to form a rigid housing to contain the i

boron-carbide-filled absorber rods. Rollers at the top and bottom of the control rod guide the control rod as it-is inserted and withdrawn from the core. The control rods are cooled by the core bypass flow. The U-shaped sheaths are perforated to allow the coolant to circulate freely about the absorber tubes. Oper-ating experience has shown that control rods constructed es I described above are not susceptible to dimensional distortions.

4

The boron-carbide (B C) powder in the absorber tubes is compacted 4

to about 70% of its theoretical density. The boron-carbide contains a minimum of 76.5% by weight natural boron. The i boron-10 (B-10) minimum content of the boron is 18% by weight.

4 Absorber tubes are made of Type-304 stainless steel. Each absorber tube is 0.220 in. in outside diameter and has a

! 0.027 in, wall thickness. Absorber tubes are sealed by a plug j welded into each end. The boron-carbide is longitudinally j separated into individual compartments by stainless steel balls at approximately 17-in. intervals. The steel balls are held in place by a slight crimp of the tube. Should boron-carbide tend to compact further in service, the steel balls will distribute the resulting voids over the length of the absorber tube.

f 4

l 4.2-33 l

i.

y y -. -q._,-y- 7,--- .,w - - - - - ,%.9 -c q,ry wyq,,w.p.y-4q-,- y,,,.eu,-,y,_ ,..g,y y-pm%,_-m-m,%m.,, _ , . _ -

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.2.4.2 Velocity Limiter The control rod velocity limiter (Figure 4.2-5) is an integral part of the bottom assembly of each control rod. This engi-neered safeguard protects against high reactivity insertion rate by limiting the control rod velocity in the event of a control rod d? o accident. It is a one-way device in that the control rod scram velocity is not significantly affected, but the control rod dropout velocity is reduced to a permissible limit.

i The velocity limiter is in the form of t**o nearly mated, conical elements that act as a large clearance piston inside the control

rod guide tube. The lower conical element is separated from j the upper conical element by four radial spacers 90 degrees apart and is at a 15-degree angle relative to the upper conical element, with the peripheral separation less than the central  !

l separation.

I The hydraulic drag forces on a control rod are proportional to approximately the square of the rod velocity and are negligible at normal rod withdrawal or rod insertion speeds. However, during the scram stroke, the rod reaches high velocity, and the drag forces munt be overcome by the drive mechanism.

To limit control rod velocity during dropout, but not during scram, the velocity limiter is provided with a streamlined j profile in the scram (upward) direction.

Thus, when the control rod is scrammed, water flows over the '

smooth surface of the upper conical element into the annulus between the guide tube and the limiter. In the dropout direction, however, water is trapped by the lower conical element and dis-  :

charged through the annulus between the two conical sections.

Because this water is jetted in a partially reversed direction 4.2-34

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O f"3 4.2.2.4.2 Velocity Limiter (Continued)

'd into water flowing upward in the annulus, a severe turbulence is created, thereby slowing the descent of the control rod assembly to less than 3.11 ft/sec.

4.2.3 Design Evaluations 4.2.3.1 Results of Fuel Rod Thermal-Mechanical Evaluations 4.2.3.1.1 Evaluation Methods Current methods for predicting fuel / cladding interaction in fuel design analyses have been discussed and compared to data in Reference 7. Important material properties used for analysis are provided in Table 4.2-5. Additional information regarding evaluation methods is provided in Section 11 of Reference 4.

(~ The mechanical evaluations reported here were performed at a

's power level equal to the license limit plus a power spike allow-ance, wherever applicable, which assures with 95% confidence that sl fuel rod in the core will exceed the maximum LHGR for which the fuel nas beer designed. Additional details regarding this method are presented in Appendix B of Reference 15.

Additional discussion of the analyses for fuel prepressurization to 3 atmospheres is contained in References 36 and 37.

During the design of BWR Zircaloy-clad UO 2 pellet fuel, continuous functional variations of mechanical properties with exposure are not employed, since the irradiation effects become saturated at very low exposure. At beginning of life, the cladding mechanical properties employed are the unirradiated values. At subsequent times in life, the cladding mechanical properties employed are the saturated irradiated values. The only exception to this is that unirradiated mechanical properties are employed above the tempera-(

\'

tures for which irradiation effects on cladding mechanical proper-ties are assumed to be annealed out. The values of clad yield 4.2-35

GESS;R II 22A7007 238 NUCLEAR ISLAMD Rev. 0 4.2.3.1.1 Evaluation Methods (Continued) strength and ultimate tensile strength employed represent the approximate lower bour.d of data on cladding fabricated by General Electric.

In the design analysis, the calculated stress and the yield strength or ultimate strength, are combined into a dimensionless quantity called the design ratio. This quantity is the ratio of calculated stress intensity to the design stress limit for a particular stress category. The design stress limit for a particular stress category is defined as a fraction of either the yield strength or ultimate strength, whichever is ?.ower. Thus, the design ratio is a measure of the fraction of the allowable stress represented by the calculated stress.

Analyses are performed to show that the stress intensity limits given in Table 4.2-2 are not exceeded during continuous operation with linear heat generation rates up to the design operating limit, or during transient operation above the design operating limit.

Stresses due to external coolant pressure, internal gas pressure, thermal effects, spacer contact, flow-induced vibration, and man-ufacturing tolerances are considered. Cladding mechanical proper-ties used in stress analyses are based on test data of fuel rod cladding for the applicable temperature.

4.2.3.1.2 Fuel Damage Analysis As noted in Subsection 4.2.1.1.2., fuel damage is defined as a perforation of the fuel rod cladding which would permit the release of fission products to the reactor coolant.

For fresh UO fuel, the calculated linear heat generation rate 2

(LHGR) corresponding to 1% diametral plastic strain of the cladding has been calculated and the results are presented in Table 4.2-14.

Due to depletion of fissionable material, the high exposure fuel has less nuclear capability and will operate at correspondingly 4.2-36

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O rT 4.2.3.1.2 Fuel Damage Analysis (Continued)

(w) lower powers, so that a significant margin is maintained throughout life between the operating LHGR and the LHGR calculated to cause ,

1% cladding diametral strain.

The addition of small amounts of gadolinia to UO2 results in a reduction in the fuel thermal conductivity and melting temperature.

The result is a reduction in the LHGRs calculated to cause 1%

plastic diametral strain for gadolinia-urania fuel rods (Reference 35). However, to compensate for this the gadolinia-urania fuel rods are designed to provide margins similar to standard UO2 r ds.

4.2.3.1.3 Steady-State Thermal-Mechanical Performance The fuel has been designed to operate at core rated power with sufficient design margin to accommodate reactor operations and

(~') satisfy the mechanical design bases discussed in detail in Sub-section 4.2.1. In order to accomplish this objective, the fuel was designed to operate at a maximum steady-state LHGR of 13.4 kW/ft, plus an allowance, wherever applicable, for densifica-tion power spiking.

Thermal and mechanical analyses have been performed which demon-strate that the mechanical design bases are met for the maximum operating power and exposure combination throughout fuel life. Design analyses have been performed for the fuel which show that the stress intensity limits given in Table 4.2-2 are not exceeded during continuous operation with LHGRs up to the operat-ing limit of 13.4 kW/f t, nor for short-term transient operation up to 16% above the peak operating limit of 13.4 kW/ft (i.e.,

15.6 kW/ft), plus an allowance for densification power spiking.

Stresscs due to external coolant pressure, internal gas pressure, thermal gradients, spacer contact, flow-induced vibration and man-ufacturing tolerances were considered. The maximum internal pres-

/~)N t

'- sure is applied coincident with the minimum applicable coolant 4.2-37

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.3.1.3 Steady-State Thermal-Mechanical Performance (Continued) pressure. Additional information regarding this type of analysis is provided in References 4 (Section 11) , 34, and 36.

The calculated maximum fission gas release fraction in the highest design power density UO r d is less than 25%. This calculation 2

is conservative because it assumes the most limiting peaking fac-tors applied to this rod. The percentage of total fuel rod radio-activity released to the rod plenum is much less tha ? 25% because of radioactivity decay during diffusion from the UO2-4.2.3.2 Results from Fuel Design Evaluations 4.2.3.2.1 Flow-Induced Fuel Rod Vibrations Experimental data on multiple rod vibrations under two-phase flow conditions have been used to develop a correlation between maximum rod displacement and fuel rod natural frequency (Reference 15).

These data indicate a decreasing maximum displacement amplitude with increasing calculated fuel rod natural frequency.

The calculated maximum vibrational amplitude for the BWR fuel is O.0007 inch. The stress levels resulting from vibration are negligibly low and well below the endurance limit of all affected components. The deflection resulting from vibration is combined with deflections from other loads and is used to demonstrate adherence to the deflection criteria of Subsection 4.2.1.2.1.9.

Flow-induced fuel rod vibration is not considered to be a viable life-limiting or failure mechanism in GE BWR fuel designs based on extensive fuel operating experience. Fuel inspections, both visual inspections during normal refueling outages, and more detailed nondestructive examinations as a part of GE surveillance programs, have not indicated any anomalous performance associated 4.2-38

GESSAR-II 22A7007 238 NUCLEAR ISLAND Rev. O r

4.2.3.2.1 Flow-Induced Fuel Rod Vibrations (Continued) f with fuel rod vibration. Table 4.2-13 presents the lower extreme of GE operating experience compared to recent fuel designs in terms of tne calculated fuel rod natural frequency. The calculated rod natural frequency for the BWR/4, 5 and 6 fuel designs is clearly well within the experience base of previously operated GE BWR fuel.

4.2.3.2.2 Potential Damaging Temperature Effects During Transients There are no predicted significant temperature effects during a power transient resulting from a single operator error or single equipment malfunction. For purposes of maintaining _ adequate thermal margin during normal steady-state operation, the minimum c_-itical power ratio (MCPR) must not be less than the required MCPR operating limit, and the maximum linear heat generation rate (MLHGR) is maintained below the design LHGR for the plant. The core and fuel design bases for steady-state operation (i.e., MCPR and LHGR

^

limits) have been defined to provide margin between the steady-state operating condition and any fuel damage condition to accom-modate uncertainties and to assure that no fuel damage results even j during the worst anticipated transient condition at any time in l life. Specifically, the MCPR operating limit is specified such that at least 99.9% of the fuel rods in the core are expected not l

to experience boiling transition during the most severe abnormal l

operational transient. The calculated fuel rod cladding strain for this class of transients is significantly below the calculated damage limit.

2 4.2.3.2.3 Fretting Wear and Corrosion Tests of typical designs, representative of the BWR/6 fuel design, have been cc.nducted both out-of-reactor as well as in-reactor prior to application in a cc.'olete reactor core basis. All tests and post-irradiation examinations have indicated that fretting corro-

, sion does not occur. Post-irradiation examination of many fuel 4.2-39

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.3.2.3 Fretting Wear and Corrosion (Continued) rods has indicated only minor fretting wear. Excessive wear at spacer contact points has never been observed with the current spacer configuration. Additional information regarding these tests and inspection of operating fuel is presented in Section 10 of Reference 4 and Sections 4.7 and 4.8 of Reference 14, 4.2.3.2.4 Fuel Rod Cycling and Fatigue Analysis During fuel life, less than 5% of the allowable fatigue life is consumed. Additional information regarding this type of analysis ic provided in Section 12 of Reference 4.

4.2.3.2.5 Fuel Rod Bowing Fuel inspections, both visual inspections during normal refueling outages and more detailed nondestructive examinations as a part of General Electric's active fuel surveillance program, have pro-vided no indication of rod bowing as a viable failure or life-limiting mechanism. This successful operating experience has been supported by fuel mechanical analyses which predict an insignifi-cant amount of fuel rod bowing (< approximately 20 mils). These analyses consider the influence of initial bow, tubing eccentri-city, fast neutron flux and thermal gradients an the potential for in-reactor creep bowing. In addition, full scale thermal-hydraulic tests have been conducted by General Electric to assess the effects of gross fuel rod bow ng. Based on results of these tests, it has been concluded that, even for severe rod bowing in the most limit-ing rods in the assembly, there is a negligible effect on critical power performance.

4.2.3.2.6 Fuel Assembly Dimensional Stability Mechanical analyses have been performed to assess the effects of the differential thermal expansion between the tieplates and spacer 4.2-40

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.3.2.6 Fuel Assembly Dimensional Stability (Continued)

(~')

'% J grids. The differential thermal expansion introduces a bending stress of less than 400 psi at the end of the fuel tube. Addi-tional information regarding the model employed in this calculation is presented in Section 4.4 of Reference 3.

4.2.3.2.7 Temperature Transients with a Waterlogged Fuel Element As indicated in Subsection 4.2.1.2.1.11, the potential for water-logging is considered in the fuel design. For waterlogging to occur, the fuel cladding must have a small pinhole. Pinholes are eliminated during production by a 100% leak check of fuel assem-blies. The Leak Detector System employed is described in babsec-tion 4.2.1.2.1.11. Since waterlogging is not expected and since it has not been observed in commercial power BWR fuel, no specific analysis of the consequences is performed.

p)

\

In the unlikely event that a waterlogged fuel element does exist in a BWR core, it should not have a significant potential for clad-ding burst (due to internal pressure) during a transient power increase unless the transient started from a cold or very low power condition. Normal reactor heatup rates are sufficiently slow (<100*F/hr increase in coolant temperature) that water-vapor formed inside a waterlogged fuel rod would be expected to evacuate the rod through the same passage it entered, allowing internal and external pressures to equilibrate as the coolant temperature l and pressure rise to the rated conditions. Once the internal and external pressures are at equilibrium, at rated coolant pressure l

l and temperature, transient power increases should, in general, have the effect of only slightly reducing the internal fuel rod l

plenum volume due to differential thermal expansion between fuel l and cladding, thus effecting a small short-term increase in inter-nal fuel rod pressure. The potential short-term increase in (si

,'-)

pressure due to this effect would, in general, be small (e.g., a power increase from the cold condition to peak rated power would l

l 4.2-41

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.3.2.7 Temperature Transients with a Waterlogged Fuel Element (Continued) increase internal pressure less than 15% in the peak power fuel rod). For the range of anticipated transients, the cladding pri-mary membrane stress resulting from the temporary increase in internal pressure above the coolant pressure are not expected to exceed the cladding stress design limits given in Subsection 4.2.1.1.1.4.

4.2.3.2.7.1 Bnergy Release for Rupture of Waterlogged Fuel Elements Experiments have been performed to show that waterlogged fuel elements can fail at a lower damage threshold than nonwaterlogged fuel during rapid reactivity excursions from the cold condition (References 16 and 17) (i.e., approximately 60 cal /gm as compared to

>300 ual/gm). However, it has been shown (Reference 28) that the resultant mechanical energy release for waterlogged rods, even for significant energy dispositions (approximately 400 cal /gm), is of little consequence and is well below the energy released for non-waterlogged rods subjected to comparable energy depositions.

4.2.3.2.8 Fuel Densification Analyses The amount of densification employed in the following models was determined through the use of models defined in References 11, 12, and 13.

4.2.3.2.8.1 Power Spiking Analysis The equation employed to calculate maximum gap size is as described in Reference 12:

^b ^P

+ 0.0025 L 2 4.2-42

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 h 4.2.3.2.'8.1 Power Spiking Analysis (Continued)

C/ )

where AL = ' maximum axial gap length; L = fuel column length; Ap = the average change in density as measured by ther-mal simulation for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 1700*C; 2 = anisotropic factor applied to densification; and 0.0025 =

allowance for irradie ion induced riadding growth and axial strain caused by fuel cladding mech-anical interaction.

fws The resulting power spiking penalty at the top of the core is 2.2%.

kms The power spiking penalty as a function of elevation from the bottom of the core can be conservatively expressed by:

"AP~ "AP" X

_ IE , _P. L X L where

[a = power spiking penalty at elevation X from bottom X of core;

_OE_ = power spiking penalty at top of core;

_P_

L X = elevation from bottom of core; and

\ L = fuel column length.

4.2-43

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.3.2.8.1 Power Spiking Analysis (Continued)

The power increase described by the preceding equation as a function of axial position added to the license limit LHGR (13.4 kW/ft) has been considered in design and safety analysis, wherever appli-cable. This ensures, with better than 95% confidence, that no more than one rod will exceed the power evaluated due to random occurrence of power spikes resulting from axial fuel column gaps.

The results of the power spiking analysis for normal operation have been utilized in the analysis of transients and accidents wherever applicable. The control rod drop accident is unique in the respect that it begins at the cold condition and is not affected by normal operating power level. Further, the existence of fuel column gaps can result in power spiking in the cold condi-tion during a control rod drop which should thus be considered in the evaluation of this accident. For this purpose, a separate power spiking analysis has been performed using the same assump-tions as indicated above, but employing a power spike versus gap size calculated to occur in the cold condition with zero voids.

This analysis was performed with the maximum gap size predicted at the top of the core in order to maximize the power spiking effect.

This analysis yielded a 99% probability that any given fuel rod would have a power spike < 5%.

4.2.3.2.8.2 Cladding Creep Collapse A cladding collapse analysis has been performed employing the standard General Electric finite element model (Reference 13).

Figure 4.2-6 presents the cladding midwall temperature versus time employed in the analysis. No credit is taken for internal gas pres ure due to released fission gas or volatiles. The internal pressure due to helium backfill during fabrication is considered.

Based on the analysis results, cladding collapse is not calculated to occur.

4.2-44

GESSAR II 22A7007 238 NUCLEAR ISLAND Rnv. 0

'~N 4.2.3.2.8.3 Increased Linear Heat Generation Rate A fuel pellet expands 1.2% in going from the cold to hot condition at 13.4 kW/ft. While this increase in length from the cold to hot condition is not taken credit for in either design calculations or in the process or core performance analysis during reactor oper-ation, the expansion more than offsets the decrease in pellet length due to densification.

The following expression is employed to calculate the decrease in i fuel column length due to densification in calculation of an increase in linear heat generation rate:

where O ap = the average change in density as measured by thermal simulation for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 1700*C, and 2 = anisotropic factor applied to densification.

Using the above equation, the pellet decrease in length due to densification is less than the increase in length due to thermal expansion of the pellet in going from cold to hot condition.

Therefore, no power increase is calculated due to densification.

i 4.2.3.2.8.4 Stored Energy Determination l

The effects on stored energy due to densification are accounted for in the LOCA evaluation.

O 4.2-45 i

. . . . . _ _ , _ . - . . - . . . . ,.----_.-__e. .--._.__4 -

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.3.2.9 Fuel Cladding Temperatures Fuel cladding temperatures as a function of heat flux are shown in Figure 4.2-7 for beginning-of-life (BOL) conditions and in Figure 4.2-8 for late-in-life conditions. The temperatures employed in mechanical design evaluations are calculated using a conservative design allowance for the increase in resistance to surface heat transfer due to the accumulation of system corrosion products on the surface of the rod (crud) and cladding corrosion (zirconium oxide formation).

4.2.3.2.10 Incipient Fuel Center Melting Incipient center melting is expected to occur in fresh UO 2 f" I rods at a LHGR of approximately 20.5 kW/ft. This condition cor-responds to the integral:

kdT =

93 W/cm 32 F The value of the above integral decreases slightly with burnup as a result of the decrease in fuel melting temperature with increasing exposure.

4.2.3.2.11 Energy Release During Fuel Element Burnout Boiling transition does not necessarily correspond to a fuel damage threshold. In-reactor experiments to assess the effect of operation of Zircaloy-clad UO 2 fuel rods after the onset of tran-sition boiling have been conducted by a number of different experi-menters (References 23 through 27). Post-irradiation examinations conducted on the fuel tested verified that no cladding failure and no appreciable cladding degradation occurred for fuel that experi-enced peak cladding temperatures less than approximately 2000 F.

4.2-46

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4

d Energy Release During Fuel Element Burnout (Continued)

($ 4.2.3.2.11

, LJ

! The metal-water chemical reaction between zirconium and water is given by:

Zr + 2H 2O =

>Zr02 + 2H2 - aH where -AH = 140 cal /g-mole. The reaction rate is conservatively given by the familiar Baker-Just rate equation:

2 (-45,500)

W = 33.3 x 10 6 T exp RT where W = milligrams of zirconium reacted per cm 2 of surface area;

! t = time (sec);

4

{

R T

=

=

the gas constant (cal / mole *K) ; and is the temperature of zirconium ('K).

This rate equation has been shown to be conservatively high by a factor of 2 (Reference 18). The above equation can be differentia-ted to give the rate at which the thickness of the cladding is oxi-dized. This yields:

i A

7

~ A2 th -

X exP 3--

where th = rate at which the cladding thickness is oxidizing; i

AX = oxidized cladding thickness; 1

Ay, A2 = appropriate constants; and

\

T = reaction temperature.

I 4.2-47 i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.3.2.11 Energy Release During Fuel Element Burnout (Continued)

The reaction rate is inversely proportional to the oxide buildup; therefore, at a given cladding temperature the reaction rate is self-limiting as the oxide builds up. The total energy release from this chemical reaction over a time period is given by:

T OT *

/ " rods

(- All) CLp AX dt where N = number of rods experiencing bciling transition (at rods temperature T);

- All = heat of reaction; C = cladding circumferences; L = axial length of rod experiencing boiling transi-tion; and p = density of zirconium.

This equation can be integrated and compared to the normal bundle energy release if the following conservative assumptions are made:

(1) At an axial plane, all the rods experience boiling transition and are at the same temperature. This is highly conservative since, if boiling transition occurs, it will normally occur on the high power rod (s) .

(2) Boiling transition is assumed to occur uniformly around the circumference of a rod. This generally occurs only at one spot.

4.2-48

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

- 4.2.3.2.11 Energy Release During Fuel Element Burnout (Continued)

U' (3) The rods are assumed to reach some temperature T instantaneously and stay at this temperature for an indefinite amount of time.

This integration has been performed per axial foot of bundle and the total energy release as a function of time has been compared to the total energy release of a high power bundle (approximately 6 MW) over an equal amount of time. The results are shown in Figure 4.2-9. For example, if the temperature of all the rods along a 1-ft section of the bundle were instantly increased to 1500*F, the total amount of energy that has been released at 0.1 sec is 0.4% of the total energy that has been released by the bundle (6 MW x 0.1 sec). Note that the fractional energy release decreases rapidly with time even though a constant temperature is maintained. This is because the reaction is self-limiting as

-~g previously discussed.

b)

The amount of energy released is dependent on the temperature transient and the surface area that has experienced heatup. This, of course, is dependent on the initiating transient. For example, if boiling transition were to occur during steady-state operating conditions, the cladding surface temperature would range from 1000 to 1500 F, depending on the heat fluxes and heat transfer coefficient. Even assuming all rods experience boiling transition instantaneously, the magnitude of the energy release is seen to be insignificant. Significant boiling transition is not possible at normal operating conditions or under conditions of abnormal opera-tional transients because of the thermal margins at which the fuel is operated. It can, therefore, be concluded that the energy release and potential for a chemical reaction is not an important consideration during normal operation or abnormal transients.

O 4.2-49

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.3.2.12 Fuel Rod Behavior Effects from Coolant Flow Blockage The behavior of fuel rods in the event of coolant flow blockage is covered in Reference 19.

4.2.3.2.13 Channel Evaluation Channel analytical models and evaluation results are contained in Reference 20.

4.2.3.2.14 Fuel Shipping and Handling Analyses of the major handling loads have been performed and the resulting fuel assembly component stresses are within design limits. Additional information on fuel handling and shipping loads is presented in Reference 31.

4.2.3.2.15 Fuel Assembly - SSE and LOCA Loadings An evaluation of combined safe shutdown earthquake (SSE) and loss-of-coolant accident (LOCA) loadings is contained in Reference 30, 4.2.3.3 Reactivity Control Assembly Evaluation (Control Rods) 4.2.3.3.1 Materials Adequacy Throughout Design Lifetime The adequacy of the control rod materials througout the design life was evaluated in the design of the control rods. The primary materials (B C powder and Type-304 austenitic stainless steel) 4 have been found to perform adequately for the lifetime of the con-trol rod.

4.2.3.3.2 Dimensional and Tolerance Analysis Layout studies are done to assure that, given the worst combina- l tion of extreme detail part tolerance ranges at assembly, no 4.2-30

1 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i-

! 4.2.3.3.2 Dimensional and Tolerance Analysis (Continued) 1 interference exists which will restrict the passage of control I

rods. In addition, pre-operational verification is made on each contr.1 rod system to show that the acceptable levels of opera-tional performance are met.

4.2.3.3.3 Thermal Analysis of the Tendency to Warp The various parts of the control rod ascembly remain at approx-imately the same temperature during reactor operation, minimizing the problem of distortion or warpage. Mechanical design allows for what little differential thermal growth that can exist. A minimum i gap is maintained between absorber rod tubes and the control rod frame assembly for the purpose. In addition, use of dissimilar metals is avoided in order to prevent any potential differential thermal expansion problems.

O 4.2.3.3.4 Forces for Expulsion hn analysis has been performed which evaluates the maximum pressure farces which could tend to eject a control rod from the core. i 1

i If the collet remains open, which is unlikely, calculations indicate that the steady-state control rod withdrawal velocity

! would be 2 ft/sec for a CRD pressure-under line break (the limiting

! case for rod withdrawal).

4.2.3.3.5 Effect of Fuel Rod Failure on Control Rod Channel Clearances The control rod drive mechanical design ensures a sufficiently rapid insertion of control rods to preclude the occurrence of fuel rod failures which could hinder reactor shutdown by causing sig-f

nificant distortions in channel clearances.

4.2-51

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.3.3.6 Effect of Blowdown Loads on Control Rod Channel Clearances The fuel channel load resulting from an internally applied pres-sure is evaluated utilizing a fixed beam analytical model under a uniform load. Tests to verify the applicability of the analytical model indicate that the model is conservative. If the gup between ch a'nne ls is less than the thickness of the blade or the diameter of the roller, the roller and/or blade will deflect the channel walls as it is inserted into the core. The friction force is a small percentage of the total force available to the control rod drives for overcoming such friction, and it is concluded that the main steamline break accident does not impede the insertability of the control rod.

4.2.3.3.7 Mechanical Damage Analysis has been performed for all areas of the teactivity control system showing that system mechanical damage does not affect the capability to continuously provide reactivity control.

The following discussion summarizes the analysis performed on the control rod guide tube.

The guide tube can be subjected to any or all of the following loads:

(1) inward load due to pressure differential; (2) lateral load due to flow across the guide tube; (3) dead weight; (4) seismic (vertical and horizontal); and (5) vibration.

4.2-52

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.3.3.7 Mechanical Damage (Continued)

(}

In all cases, analyses were performed considering both a recircu-lation line break and a steamline break, events which result in the largest hydraulic loadings on a control rod guide tube.

Two primary modes of failure were considered in the guide tube analysis: (1) exceeding allowabic stress, (2) excessive clastic deformation. It was found that the allowable stress limit will not be exceeded and that the clastic deformations of the guide tubo never are great enough to cause the free movement of the control rod to be jeopardized.

4.2.3.3.7.1 First Mode of Failure The first mode of failure is evaluated by the addition of all the stresses resulting from the maximum loads for the faulted condi-() tion. This results in the maximum theoretical stress value for that condition. Making a linear supposition of all calculated stresses and comparing the calculated value to the allowable limit defined by the ASME Boiler and Pressure Vessel Code yields a factor of safety of approximately 3. For faulted conditions, the factor,of -

~

safety is approximately 4.2.

4.2.3.3.7.2 Second Mode of Failure Evaluation of the second mode of failure is based on clearance reduction between the guide tube and the c,ontrol rod. The minimum allowable clearance is about 0.1 inch. This assumes maximum -

ovality and minimum diameter of the guide tube and the maximum control rod dimension. The analysis showed that, if the approx-imate 6000 psi for the faulted condition were entirely the result of differential pressure, the clearance between the control rod' and the guide tube would reduce by approximately 0.01 inch.

I 4.2-53 ,- ,

. GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.3.3.7.2 Second Mode of Failure (Continued)

This gives a design margin of 10 between the theoretically calculated maximum displacement and the minimum allowable clearance.

4.2.3.3.8 Analysis of Guide Tube Design Two types of instability were considered in the analysis of guide tube design. The first was the classic instability associated with vertically loaded columns. The second was the diametral collapse when a circular tube experiences external to internal diffcrentia1 pressure.

The limiting axially applied load is approximately 77,500 lb resulting in a material compressive stress of 17,450 psi (code allowable stress). Comparing the actual load to the yield stress level gives a design margin greater than 20 to 1. From these values, it can be concluded that the guide tube is not an unstable column.

When a circular tube experiences external to internal differen-tial pressure, two modes of failure are possible depending on whether the tube is "long" or "short". In the analysis, the guide tube is taken to be an infinitely long tube with the max-imum allowable ovality and minimum wall thickness. The conditions resulted in the lowest critical pressure calculation for the guide tube (i.e., if the tube was "short", the critical pressure calculation would give a higher number). The critical pressure ts approximately 140 psi. Iloweve r , if the maximum allowable

, . ,stresa is reached at a pressure lower than the critical pressure, then that pressure is limiting. This is the case for a BWR guide tube. The allowable stress of 17,450 psi will be reached at approxima tely 93 psi . Comparing the maximum possible pressure differential for a steamline break to the limiting pressure of 4.2-54

GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0

/}

%.)

4.2.3.3.8 Analysis of Guide Tube Design (Continued) 93 psi gives a design margin greater than 3 to 1. Therefore, the guide tube is not unstable with respect to differential pressure.

4.2.3.3.9 Evaluation of Control Rod Velocity Limiter The control rod velocity limiter limits the free-fall velocity of the control rod to a value that cannot result in nuclear system process barrier damage.

4.2.4 Testing and Inspection 4.2.4.1 Fuel, Hardware and Assembly Rigid quality control requirements are enforced at every stage of fuel manufacturing to ensure that the design specifications are met. Written manufacturing procedures and quality control plans define the steps in the manufacturing process. Fuel cladding is subjected to 100% dimensional inspection and ultrasonic inspec-tion to reveal defects in the cladding wall. Destructive tests are performed on representative samples from each lot of tubing, including chemical analysis, tensile and burst tests. Integrity of endplug welds is assured by standardization of weld processes based on radiographic and metallographic inspection of welds.

Fuel rod inspection includes metallographic and radiographic examination of fuel rods on a sample basis. Completed fuel bundles are helium leak tested to detect the escape of helium through the tubes and endplugs or welded regions. Sample tests a re performed for qualification. Production samples are tested as a check on the process and process controls. UO2 powder char-acteristics and pellet densities, composition, and surface finish are controlled by regular sampling inspections. UO 2 weights are

()

V recorded at every stage in manufacturing.

4.2-55

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.4.1 Fuel, liardware and Assembly (Continued)

Each separate pellet group is characterized by a single stamp.

Fuel rods are individually serialized prior to fuel loading to:

(1) identify which pellet group (s) is to be loaded in each fuel rod; (2) identify which position in the fuel assembly each fuel ad is to be loaded; and (3) facilitate total fuel material accounta-bility for a given project. Each finished fuel rod is gamma scanned to detect an enrichment or rod pellet loading deviations which exceed design specification.

The fuel rod upper endplugs are designed to control placement of fuel rods within a given assembly. Primary control is provided by alphanumeric symbols on the end of the upper endplug which corre-sponds to a specific enrichment and gadolinia content of the fuel rod. Secondary control is provided by sizing of the upper endplugs such that a rod of high enrichment cannot be positioned in a significantly lower enrichment location within the fuel bundle.

Additionally, the gadolinia-bearing fuel rods have extended end-plugs with characteristic markings on the shank to permit visual identification of gadolinia rod location. Correct placement is verified by recording the fuel rod serial number on the lower endplug, the alphanumeric symbols on the upper endplug, visual inspection and placement of the upper tieplate, which has been machined to accept a specific pattern of endplug diameters.

Puel assembly inspections consist of complete dimensional checks of channels and fuel bundles prior to shipment. Puel bandles are given another dimensional inspection of significant dimensions at the reactor site prior to use. The sampling rate, method and tools of the post-shipment fuel inspection are outlined in Tables 4.2-6 and 4.2-7.

I 4.2-56

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.4.2 Testing and Inspection (Enrichment and Burnable Poison 7-~3 (v) Concentrations)

The shutdown reactivity requirement is verified during initial fuel loading and at any time that core loading is changed.

Nuclear limitations for control rod drives are verified by period-ically testing the individual system. Test capabilities are described in the appropriate subsections.

The following serves to identify the various tests and inspections employed by General Electric in verifying the nuclear character-istics of the fuel and reactivity control systems.

4.2.4.2.1 Enrichment Control Program The incoming UF 6 and the resultant UO2 powder are verified by emission spectroscopy for impurities.

() The sintered pellet is also sampled for impurities by emission spectroscopy. Chemical verification of impurities is also per-formed including wet chemistry for oxygen-uranium ratio determination.

The enrichment-blended material and the green pellet enrichment are verified by the gamma scan enrichment analyzer. Each rod is gamma scanned to screen out any possible, but unlikely, enrich-ment deviations.

All assemblies and rods of a given project are inspected to assure overall accountability of fuel quantity and placement for the project.

4.2.4.2.2 Gadolinia Inspections The same rigid quality control requirements observed for standard 00 2 fuel are employed in manufacturing gadolinia-urania fuel.

4.2-57

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.4.2.2 Gadolinia Inspections (Continued)

Gadolinia bearing UO 2 fuel pellets of a given enrichment and a O

gadolinia concentration are maintained in separate groups through-out the manufacturing process. The percent enrichment and gadolinia concentration characterizing a pellet group is identi-fied by a stamp on the pellet.

Fuel rods are individually serialized prior to loading of fuel pellets into the fuel rods to: (1) identify which pellet group it to be loaded in each fuel rod; (2) identify which position in the fuel assembly each fuel rod is to be loaded; and (3) facili-tate total material accountability for a given project. Correct orientation of gadolinia-bearing rods within the fuel assembly is further assured by the longer upper endplug shanks for these rods.

The following quality control inspections are made:

(1) gadolinia concentration in the gadolinia-urania powder blend is verified; (2) sintered pellet UO -Gd 2 203 solid solution homogeneity across a fuel pellet is verified by examination of metallographic specimens; (3) gadolinia-urania pellet identification is verified; and (4) gadolinia-urania fuel rod identification is checked.

4.2.4.2.3 Reactor Control Rods Inspections and tests are conducted at various points during the manuf acture of control rod assemblies to assure that design O

4.2-58

GESSAR II 22A7007 238 NUCLEAR ISLAND Rnv. 0

,e w) 4.2.4.2.3 Reactor Control Rods (Continued)

V requirements are being met. All boron carbide lots are analyzed and certified by the supplier. Among the items tested are:

(1) chemical composition; (2) boron weight percent; (3) boron isotopic content; and 4

(4) particle size distribution.

Following receipt of the boron carbide and rev.1vs of material certificates, additional samples from each lot are tested includ-ing those previously listed. Control is maintained on the B 4 C powder through the remaining steps prior to loading into the absorber rod tubes.

Certified test results are obtained on other control rod compo-nents. The absorber rod tubing is subjected to extensive testing by the tubing supplier, including 100% ultrasonic examination.

Meta 11ographic examinations are conducted on several tubes ran-domly selected from each lot to verify cleanliness and absence of conditions resulting from improper fabrication, cleaning, or heat trea tment . Other checks are made on the subassemblies and final control rod assembly, including weld joints inspected and B 4C loading.

4.2.4.3 Surveillance Inspection and Testing of Irradiated Fuel Rods General Electric has an active program of surveillance of both production and developmental BWR fuel. The schedule of inspec-

-s tion is, of course, contingent on the availability of the fuel as g_,) influenced by plant operation.

l 4.2-59 l ,

'e

GESSAR II 22A7007 238 NUCLEAR ISLAEa Rev. 0 4.2.4.3 Surveillance Inspection and Testing of Irradiated Fuel Rods (Continued)

The lead fuel rods (with respect to exposure, LHGR, and the com-bination of both) are selectively inspected. Inspection tech-niques used include: *

(1) leak detection tests, such as " sipping";

(2) visual inspection with various aids such as binoculars, borescope, periscope, and/or underwater TV with a photographic record of observations as appropriate; (3) nondestructive testing of selected fuel rods by ultra-sonic test techniques; and (4) dimensional measurements of selected fuel rods.

Unexpected conditions or abnormalities which may arise, such as distortions, cladding perforation, or surface disturbances are analyzed. Resolution of specific technical questions indicated by site examinations may require examination of selected fuel rods in Radioactive Material Laboratory facilities.

The fuel channels are also under surveillance in continuing pro-grams. These surveillance programs are designed not only for the evaluation of present day products, but are also providing data in the areas of alternate materials and design modeling.

The results of the program are used to evaluate the BWR fuel design methods and criteria used by General Electric and are gen-erally reviewed with the Division of Reactor Licensing and docu-mented in generic fuel experience licensing topical reports.

In addition to the fuel surveillance program, lead test assemblies essentially identical to the BWR/4 and BWR/5 assemblies have been 4.2-60

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 N 4.2.4.3 Surveillance Inspection and Testing of Irradiated Fuel

) Rods (Continued) characterized in detail prior to irradiation and have been placed in service in two operating reactors (Peach Bottom 2 and Vermont Yankee). These lead test bundles will provide approximately two years of operating experience with this fuel design prior to full core loading in a BWR/4 or 5 reactor. A surveillance program similar to that described in the preceding paragraphs has been incorporated in the lead test assembly program to provide exten-sive performance monitoring. The fuel rod design for the BWR/6 fuel assemblies is the same as that employed in the lead test assemblies; therefore, the experience obtained with these assem-blies is applicable to BWR/6 fuel. A propressurized test assembly employing the same fuel rod design was placed in operation in April 1977. For a description of this fuel assembly, see Reference 36.

) 4.2.5 Operating and Developmental Experience 4.2.5.1 Fuel Operating Experience A large volume of experience with Zircaloy-clad UO2 pellet fuel has been obtained since 1960. The largest portion of this exper-ience has been obtained in operating commercial power BWRs at LHGRs representative of, or higher than, current 8x8 fuel perform-ance requirements. The large volume or production experience, starting with the first load of fuel in Dresden-1 Nuclear Power Station in 1960, has provided feedback on the adequacy of the design for, and the effects of, operation in a commercial power reactor environment.

Table 4.2-8 presents a summary of BWR experience with General Electric production Zircaloy-clad UO2 pell t fu 1. Overall, 73 s production fuel types have been designed, manufactured and oper-s ,)

m ated in 32 BWRs. When all production fuel types are considered, 4.2-61

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.5.1 Fuel Operating Experientu (Continued) a total of more than 1,250,' Jircaloy-2 clad UO O

2 fu 1 r ds have been operated in General Electric designed BWRs. The most recent fuel experience documented in Reference 21 indicates that, although the available BWR fuel experience base has increased by over 50% in the last two years, no new fuel failure mechanisms have been observed.

Peak LHGR, from approximately 10 kW/ft to approximately 18.5 kW/ft, have been experienced with the production fuel. Individual fuel assemblies have achieved average exposures greater than 25,000 mwd /t and have accumulated more than nine years in-core residence. In comparison, the current 8x8 fuel designs have the following proposed operating characteristics:

(1) 13.4 kW/ft maximum LHGR (operating limit);

(2) 40,000 mwd /t maximum local exposure; (3) 30,000 mwd /t maximum assembly exposure; and (4) 4 to 6 years in-core residence time.

Fuel rod diameters in the range of 0.425 in. to 0.593 in. outside diameter with cladding wall thickness from 30 to 40 mils and pellet-to-cladding gaps from 3 to 12 mils have been used in pro-duction fuel. Active fuel column lengths have varied from 59.8 to 146.0 in, with fission gas plenum volume per unit of fuel volume from 0.013 to 0.11.

Production fuel rods employing gadolinia-urania fuel pellets have been in use since 1965. During this time, a substantial number of gadolinia-urania rods have been successfully irradiated to appreciable exposures. Table 4.2-9 summarizes this experience.

Of these irradiated gadolinia-urania rods only a small number have 4.2-62

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.5.1 Fuel Operating Experience (Continued)

(}

experienced failure, none of which could be attributed to the fact that they contained gadolinia bearing fuel pellets.

4.2.5.2 Fuel Development Experience The production of Zircaloy-clad UO2 pellet fuel experience des-cribed in Subsection 4.2.5.1 is supplemented by a large amount of in-reactor and out-of-reactor developmental work. The develormental work to date has been employed to test a wide range of design characteristics, to investigate various mechanisms affecting the performance of the fuel rod and to extend irradiation experience to higner local combinations of fuel rod power and exposure than covered by production fuel. The following presents a discussion of the pertinent developmental fuel experience which, in combina-tion with the production fuel experience, provided the basis for

() the current fuel design and operating limits.

Tables 4.2-10 through 4.2-12 present a summary of design details and performance conditions for Zircaloy-clad UO2 pellet fuel rods and capsules

• irradiated under u;neral Electric or USAEC-General Electric development test programs. These data compliment the BWR production fuel experience by providing additional data at higher local combinations of fuel rod power and exposure. Over-all, more than 800 fuel pins with design characteristics similar to the current BWR fuel have been irradiated under General Electric or USAEC-General Electric programs. The irradiations have been performed with BWR environment in both test reactors and in commercial power BWRs. Test reactors employed in General Electric developmental irradiations summarized in Tables 4.2-10 through 4.2-12 are the Vallecitos Boiling Water Reactor (VBWR),

A capsule, as used herein, refers to a' test fuel rod, or a group of rods combined, with all features similar to production

[\ /s) fuel rods except for having reduced active fuel length (as low as approximately 3 inches).

4.2-63

GESSAR II 22R7007 238 NUCLEAR ISLAND Rev. 0 4.2.5.2 Fuel Development Experience (Continued)

O the General Electric Test Reactor (GETR) and, more recently, the IIalden Reactor. Developmental fuel irradiations have also been performed in the consumers Power Company Big Rock Point and Dresden Unit 1 commercial power BWRs.

The range of peak performance conditions covered by the various development irradiations goes beyond the design performance con-ditions for fuel in this class of reactor. The development per-formance conditions include:

(1) 13.0 to 58.0 kW/ft maximum LHGR, and (2) 1500 to 100,000 mwd /Te maximum local exposure.

The corresponding conditions for fuel to be operated in the BWR/6 class of reactor are:

O (1) 13.4 kW/ft maximum LHGR, and (2) approximately 45,000 mwd /Tc maximum local exposure.

The range of design characteristics and dimensions covered by the various developmental irradiations also encompasses the character-istics and dimensions employed in the current BWR fuel design.

The range of design characteristics and dimensions covered by the various developmental irradiations include the following:

(1) fuel rod outside diameter 250 to 0.700 in.);

(2) cladding wall thickness (0.025 to 0.060 in.);

(3) pellet-cladding gap (0.0014 to 0.016 in.); and (4) pellcr. Length (0.3 to 0.95 in.).

4.2-64

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

-s 4.2.5.2 Fuel Development Experience (Continued) v The corresponding fuel desian characteristics for this class of reactor are:

(1) fuel rod outside diameter (0.483 in.);

(2) cladding wall thickness (0.032 in.);

(3) pellet-cladding diametral gap (0.009 in.); and (4) pellet length (0.41 in.).

It has been concluded that, for the complete range of power levels and for peak fuel burnups, the calculated fuel performance has been adequately verified by experience.

4.2.5.3 Fuel Rod Perforation Experience

('N)

Q./

The early General Electric BWR fn21 experience has been exten-sively described in previous reports. In general, the Zircaloy-2 cladding performance in the very early plants was good; however, some fuel failure mechanisms were exposed and corrected and are not significantly affecting current fuel performance. Details of this experience are provided in References 7, 21 and 22.

Hydriding and pellet-cladding interaction are the failure mech-anisms which have continued to affect fuel performance. Hydriding defects were identified by significant numbers of rods perforating at relatively low power and exposure. The pellet-cladding inter-action problem currently being experienced in operating fuel did not become appreciable until a statistically significant number of Zircaloy-clad fuel rods had experienced relatively higher burnup at relatively higher power. The current fuel design incor-

[N porates improvements in design and manufacturing which provide

\/ '

confidence that a high degree of reliability can be expected.

4.2-65

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.2.5.2 Fuel Development Experience (Continued)

Operation with failed fuel rods has shown that the fission product O

release rate from defective fuel rods can be controlled by regu '

lating power level. The rate of increase in released activity apparently associated with progressive deterioration of failed rods has been deduced from chronological plots of the offgas ac'tivity measurements in operating plants. These data indicate that the activity release level can be lowered by lowering the local power density in the vicinity of the fuel rod failure.

These measured data also indicate that sudden or catastrophic failure of the fuel assembly does not occur with continued opera-tion and that the presence of a failed rod in a fuel assembly does not result in propagation of failure to neighboring rods. Shut-down can be scheduled, as required, for repairing or replacing fuel assemblics that have large defects.

Evaluation of the fission product release rate for failed fuel rods shows a wide variation in the activity release levels.

Designers have attempted to relate the relense rates to defect type, size and specific power level. These data support the qualitative observations that fission product release rates are functions of power density and that progressive deterioration is a function of time.

4.2.5.4 Channel Operating Experience General Electric Company has more than 7000 Zircaloy channels in operating reactors, and surve illance of their performance is ongoing. The preponderance of the experience has been with chan-nels that are 5.278 in. inside width with 0.080 in. wall thick-ness. Channel sizes ranging from 4.290 to 6.543 in. inside width and with 0.060 to 0.100 in. walls are included. The BWR/6 channel is 5.215 in. inside width, with 0.120 in. walls.

O 4.2-66

- _ _ . __. _. .- . _ - - _ _ _ _ -. _ _ _ = _ _ . . _ . _

GESSAR II 22A7007

'238 NUCLEAR ISLAND Rev. O

~ 4.2.5.4 Channel Operating Experience (Continued)

I 1

The performance of the channels currently in operation has shown no tendency for gross in-service deformations, although long-term

creep deformation has been identified as a potential life-limiting phenomenon. Separate reports on this subject have been provided (Reference 20).

I i

j 4.2.6 References i

1

! 1. M. A. Miner, " Cumulative Damage in Fatigue", Applied Mech.,

i 12, Transactions of the ASME, 67 (1945).

l 2. W. F. O'Donnell and B. F. Langer, " Fatigue Design Basis for

Zircaloy Components", Nuclear Science and Engineering,

! Vol. 20, 1964, pp 1-12.

/

3. BWR/6 Fuel Design, June 1976 (NEDO-20948-P).

l

4. " General Electric Boiling Water Reactor Generic Reload Application for 8x8 Fuel", March 25,-1976 (NEDO-20360-lP),

() 5.

Revision 4.

"Effect of High Burnup on Zircaloy-Clad, Bulk UO2 Plate Fuel Element Samples", September 1962 (WAPD-TM-283).

4

6. " Irradiation Behavior of Zircaloy Clad Fuel Rods Containing Dished End UO2 Pellets", July 1967 (WAPD-TM-629).

j 7. H. E. Williamson and D. C. Ditmore, " Experience with BWR Fuel Through September 1971", May 1972 (NEDO-10505).

j

8. D. C. Ditmore and R. B. Elkins, "Densification Considera-tions in BWR Fuel Design and Performance", December 1972 l (NEDM-10735). ,

9. J. A. Christenson, " Melting Point of Itradiated Uranium Dioxide", February 1965 (WACP-6065).

10. " Thermal Conductivity of Uranium Dioxide", Technical Report Series No. 59, IACA, Vienna, 1966.

i

11. Supplement 1 to the Technical Report on Densification of

! General Electric Reactor Fuels, December 1973.

1 i

l 12. V. A. Moore, letter to I. S. Mitchell, " Modified GE Model for Fuel Densification", Docket 50-321, March 22, 1974.

l 4.2-67 l

l - - - - - - - - - - , - - - - , . - _ . - - - , _ _ _ - - - - - _ - - _ . _ . - . . _ , - _ _ , _ - . . , - . . .-. -

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 References (Continued) 4.2.6 O

13. " Creep Collapse Analysis of BWR Fuel Using Safe Collapse Model", August 1974 (NEDE-20606) (Proprietary), (NEDO- 2060 5)

(Nonproprietary) .

14. "8x8 Fuel Bundle Development Support", February 1975 (NEDO-20377).

15. " General Electric Boiling Water Reactor Generic Reload Application for 8x8 Fuel", March 1976 (NEDO-20360),

(Supplement 4).

16. L. A. Stephan, "The Response of Waterlogged UO2 Fuel Rods to Power Bursts", April 1969 (IDO-ITR-105) .

17. L. A. Stephan, "The Effects of Cladding Material and Heat Treatment on the Response of Waterlogged UO2 Fuel Rods to Power Bursts", January 1979 (IN-ITR-lll).

18. " Thermal Response and Cladding Performance of an Internally Pressurized, Zircaloy Clad, Simulated BWR Fuel Bundle Cooled by Spray Under Loss-of-Coolant Conditions", April 1971 (GEAP-13112).

19. " Consequences of a Postulated Flow Blockage Incident in a BWR", October 1977 (NEDO-10174) - Rev. 1.

20. "BWR Fuel Channel Mechanical Design and Deflection",

September 1976 (NEDE-213 54-P ) (GE Proprietary), (NEDO-21354)

(Non-Proprietary).

21. R. B. Elkins, " Experience with BWR Fuel Through December 1976", July 1977 (NEDO-21660).

22. H. E. Williamson and D. C. Ditmore, " Current State of Knowledge High Performance BWR Zircaloy Clad UO2 Fuel",

May 1970 (NEDO-1017 3) .

23. H. P. Olson, "In-Pile Burnout Protection Demonstrated at HBWR", Euro Nuclear, December 1964.

24. J. E. Boyden, S. Levy, M. F. Lyons, T. Sorlic, " Experience with Operating BWR Fuel Rods Above the Critical Heat Flux",

Nucleonics, April 1965, Volume 23, No. 4.

25. G. Kjaerheim, E. Rolstad, "BWR Burnout Experiments",

Nuclear Engineering International, December 1968.

26. T. Sorlie, " Consequences of Operating Zircaloy-2 Clad Fuel Ro.is Above the Critical Heat Flux", October 1965 (APED-4986).

4.2-68

GESSAR II 22A7007 238 NUCLEAR ISLAND Rnv. O a

I 1 4.2.6 References (Continued)

L.'

27. W. Redpath, "Winfrith SGHWR In-Reactor Dryout Tests",

Journa1 of the British Nuclear Energy Society, 1974, 13(1) p. 87-97.

28. L. B. Thompson, et. al., " Light Water Reactor Fuel Behavior Program

Description:

RIA Fuel Behavior Experiment Require-ments", USAEC Report RE-S-76-170, September 1976.

29. G. A. Potts, "Urania-Gadolinia Nuclear Fuel Physical and Irradiation Characteristics and Material Properties",

January 1977 (NEDE-20943) Proprietary, (NEDO-20943)

Nonproprietary.

30. "BWR/6 Fuel Assembly Evaluation Combined SSE and LOCA Loadings", November 1976, (NEDE-21175-P).

31. " Fuel Assembly Evaluation of Shipping and Handling Loadings", March 1977, (NEDE-23542-P, NEDO-23542).

32. General Electric Thermal Analysis Bases (GETAB): Data, Correlation, and Design Application, General Electiric Co.,

- . , November 1973, (NEDO-10958).

p t  :

\> 33. BWR/4 and BWR/5 Fuel Design Amendment General Electric Co.,

January 1977, (NEDO-20944-1, Non-Proprietary).

34. R. B. Elkins, " Fuel Rod Prepressurization Amendment 1",

May 1973 (NEDO-23786-1).

35. Letter, E. D. Fuller to O. D. Parr, "NRC Request for Addi-tional Information on Fuel Rod Prepressurization",

June 8, 1978.

36. Letter, E. D. Fuller to O. D. Parr, "NRC Request for Addi-tional Information on Fuel Rod Prepressurization",

August 14, 1978.

37. R. B. Elkins, " Fuel Rod Prepressurization", March, 1978 (NEDE-23786-1-P).

/~'s i

! ) i

's_/ l 4.2-69[4.2-70

l GESSAR II 238 NUCLEAR ISLAND 22A7007 ,

Rev. 0 1

1 Table 4.2-1 FUEL CLADDING CONDITIONS OF DESIGN RESULTING FROM IN-REACTOR PROCESS CONDITIONS COMBINED WITH EARTHQUAKE LOADING i

t t CONDITIONS OF DESIGN ,

i Reactor Initial Percent of Safe Shutdown Earthquake Imposed 4 Conditions 0% 50% 100%

Startup testing Upset -- --

Normal Normal Upset Faulted Abnormal Upset -- --

I

\

t

)

i 3

i l

l t

! 4.2-71 s

, - , - - - - - - , - . . . . . . . - - , _ - . - . . - - - . - - . . . . . . . _ _ _ , . _ . - _ . _ _ _ . . . . _ - - . _ . _ . - - . - - . . . - ~ . - . . . , , _ . -

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 4.2-2 FUEL CLADDING STRESS INTENSITY LIMITS Yield Strength Ultimate Tensile Sy Strength S u Primary Membrance Stress 2/3 1/2 Primary Membrane Plus Bending Stress Intensity 1 1/2 to 3/4 Primary Plus Secondary Stress Intensity 2 1.0 to 1.5 O

I I

e d 4.2-72 l

4 I

22A7007

~

GESSAR II i 238 NUCLEAR ISLAND Rev. O O FUEL CLADDING ESTIMATED NUMBER OF CYCLES FOR EACH CYCLIC Table 4.2-3

! CONDITION USED FOR FATIGUE ANALYSIS i

i Cyclic Condition Estimated Cycles 1 6

j Room Temperature to 100% Power approximately 4/yr Hot Standby to 100% Power approximately 12/yr i

50% Power to 100% Power approximately 60/yr 75% Power to 100% Power approximately 250/yr 100% Power to 116% Power approximately 1/2 yr b

f O

i I '

l l

i lO 1

4.2-73 l

I

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 4.2-4 FUEL DATA Core Number of Fuel Assemblies 748 Fuel Cell Spacing (Control Rod Pitch) (in.) 12.0 Total Number of Fueled Rods

• 46376 Core Power Density (Rated Power) (kW/t) 54.1 Total Core Heat Transfer Area (ft ) 73303 Fuel Assembly Data Nominal Active Fuel Length (in.)** 150 Fuel Rod Pitch (in.) 0.636 Fuel Rod Spacing (in.) 0.153 Fuel Bundle Heat Transfer Area (ft ) 98 Fuel Channel Wall Thickness (in.) 0.120 Channel Width (Inside) (in. ) 5.215 Fuel Rod Data Outside Diameter (in.) 0.483 Cladding Inside Diameter (in.) 0.419 Cladding Thickness (in.) 0.032 Fission Gas Plenum Length (in.) 9.48 Pellet Immersion Density (%TD) 95 Pellet Outside Diameter (in.) 0.410 Pellet Length (in.) 0.410 Water Rod Data Outside Diameter (in.) 0.591 1

Inside Diameter (in.) 0.531

• Does not include two water rods in each assembly l ** Includes 6 in, of Natural U at the top and bottom of the fuel column.

O 1

4.2-74

. . . . - _ . . . _ .~ .

GESSAR II 22A7007 138 NUCLEAR ISLAND Rev. O Table 4.2-5 Os MATERIAL PROPERTIES

• i Zircaloy-2 Cladding Thermal Conductivity T = (600 to 800*F) k = 9-10 (Btu /hr-f t *F)

Coefficient of Linear Thermal Expansion approximately 3x10

-6 gp-1) fo - h" Total Elongation (Irradiated) > 1%

UO 2 Pellets Thermal Conduc- - 3978.1 - = + 6.02366 x 10 -12 (T+460 ) 3 Y

'692.61+T.

(Btu /hr=ft 'F)

O Melting Temperature = 5080 - 63.5 x 10

~4 E ('F)

(where E = Exposure mwd /t)

• Additional information on material properties is presented in Section 3 of Reference 4 and Section 4 of Reference 3.

2 i O 4.2-75 l

.__.-,_.,,--..__.-_,m_____,--., ._.---.._,,,____,-----,-,m. _ _ . - - _ - - . _ . . , _ . . . - . - . _ . . . - . _ . . . . _ , . - _ _

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 4.2-6 POST-S!!IPMENT FUEL INSPECTION PLAN Characteristic Method Frequency Container Damage Visual 100%

and Leak Bundle Damage Visual 100%

Shipping Separa- Visual 100%

tors Removed Cleanliness Visual 100%

Rod Integrity Visual, Gauge 100%

when required Lock Tab Washers Visual 100%

Channel Integrity Visual 100%

Channel Cleanliness Visual 100%

Guard Integrity Visual and 100%

and Installation Torque Wrench Spacer Damage Visual 100% for first 5 bundles and every 20th thereafter, otherwise the middle 3 spacers Rod to Rod Feeler Gauge 100% of first 5 bundles and every 20th thereafter, otherwise two sections, all spacers, alternate the sections Rod-to-Simulated Simulated 100% of first 5 bundles Channel Channel and and every 20th thereafter, Feeler Gauge otherwise 2 sections, 4 sides per section, alternate sec-tions excluding end sections.

Expansion Spring Visual 100% for all bundles Length Gauge 100% for first 5 bundles and every fourth thereafter, otherwise visual inspection.

Finger Spring Visual 100% for all bundles.

Seated in Pocket Gauge 100% for first 5 bundles and every fourth thereafter, otherwise visual inspection.

Note:

Deviations required 100% inspection of the next 5 bundles for that characteristic. Two deviations for 1 characteristic within 6 con-secutive bundles require revision of the AQL (acceptable quality level) with the General Electric, Wilmington, North Carolina, U.S.A.

facility.

Where a reduced inspection was performed, all inspection steps shall be designated S OK (stamped OK) .

4.2.76

GESSAR II 22A7007 238 NUCLEAR ISLAND Rov. 0 Table 4.2-7

(} INSPECTION EQUIPMENT (1) Sling, four-legged to lift containers and remove lids from shipping boxes.

(2) Bundle hold-down bars (drawing 107C4713) for clamping on to containers before pickup to the vertical position (2 each)

(3) Special spreader bar sling arrangement consisting of:

, a. spreader bar (drawing 107C4707);

b. bridle sling, similar to Bethlehem No. 206-C (5/16 in.

diameter by 2-foot length, two legs, thimble on load ends); and

c. two single-leg Bethlehem No. 110-C safety swivel hooks (4-foot length by 3/8-in. diameter wire) and 4 (7/16-in.) screw pin anchor shackles.

I (4) Special holding bar and cable sling (QCF-0015A) to support I channel for channeling operation.

t C) w/ (5) Stop plate for tilting container at bundle unloading station.

(6) Safety straps, with safety swivel shaps on each end, and takeup buckle to secure container in the vertical position.

(7) Feeler gauges, nylon: 0.105- and 0.100-in.

(8) Shim gauge

0.015-in.

(9) Spring length gauge.

(10) Simulated channel gauge.

(11) Male thread gauge, 5/16-18 UNC, go-no-go type.

O 4.2-77

i-----

I GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

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%b 4.2-78

1 s m n I Table 4.2-8 (Continued)

SUMMARY

OF EXPERIENCE IN PRODUCTION ZIRCALOY-CLAD UO2 FUEL L (DECEMBER 31, 1976) t Number Exposure l Design Active Segmente(S)

Exposure Average Time in Peak Fuel Rod 3 Cladding 3 Pellet-to- Fuel or Rods Class of Fuel No. of Peak Pellet Assembly Core IJiCR Diameter ' Thickness Cladding Cap Length Still a Reactor Reactor Type Bundles (MWJ/t) (W d/t) (years) kW/ft (in.) (alls) (Ncaninal mils) (in.) Tot al In Core I l

CEB 32 9,800 6,900 1.6 17.5 0.563 37 12 144 1.568 1,372 [

UO 276 11,900 5.233 1.6 13.4 0.493 34 9 144 17.365 17,188 GBH 8 5,600 4,000 0.8 13.4 0.493 34 9 144 504 504 3 Dresden 3 DD 724 23,500 15,600 5.9 17.5 0.563 32 12 144 35,476 16,660 t Reload CEB 52 15,900 11.100 3.6 17.5 0.563 37 12 144 2.548 2.548 l CEN 44 13,800 9,800 2.5 13.4 0.493 34 9 144 2.772 2,772 UO 132 9,300 6,000 1.3 13.4 0.493 34 9 144 8,316 8,316 GBH 8 6,200 4,500 1.3 13.4 0.493 34 9 144 504 504 U4 148 1,100 700 0.3 13.4 0.493 34 9 144 9,324 9,324 3 Millstone 1 MS 580 25,700 14,000 6.2 17.5 0.570 35.5 12 144 28,420 4,557 y [

Reload CEA 82 22,900 14,900 3.8 17.5 0.563 32 12 144 4,018 3,283 w i CEB 30 23,300 15.300 3.8 17.5 0.563 37 12 144 1,470 1,421 CO  !

MSB 125 15,900 10,000 2.1 17.5 0.563 37 12 144 6.125 6,076 U 143/124 9,000 3,300 1.1/0.1 13.4 0.493 34 9 144 16,821 16,821 Z 3 Fukushima 1 TX 404 22,300 13,800 6.4 17.5 0.570 35.5 12 144 19,796 9,800 C O ,

Reload TXA 60 13,100 7,200 5.1 17.5 0.563 32 12 144 2,940 2,401 OM f t i TXB 111 11,900 5,700 1.0 17.5 0.563 37 12 144 5,439 5,4.39 l .tm TXC 40 3,500 1,900 1.0 17.5 0.563 37 12 144 1,960 1,960 yp

• 3 Monticello MT 484 25.100 14,200 5.0 17.5 0.563 32 12 144 23,716 0 yy l M Reload 20 17,300 3.6 17.5 0.563 144 980 980

CEB 24.100 37 12 l

1 MTB 116 19,400 10,800 2.6 13.4 0.493 34 9 144 7,308 7,308 HH fj

( CBH 80 15,600 9,500 1.8 13.4 0.493 34 9 144 5,040 5,040 03 H l '

U 268 9,800 6,500 1.1 13.4 0.493 34 9 144 16,884 16,884 M  !

3 Nucienor NU 404 26.100 17,100 6.2 17.5 0.570 35.5 12 144 19,796 2.842 Reload CEA 28 26,700 17,700 4.3 17.5 0.563 32 12 144 1.372 1,176 g NL* B 68 24,800 16,200 3.6 17.5 0.563 37 12 144 3,332 3,234 j NUC 96 21,200 12,500 2.6 13.4 0.493 34 9 144 6,048 6.048 i U 60/96 14,400 5.730 0.7/1.6 13.4 0.493 34 9 144 9.828 9,828 24,100 15,100 17.5 35,476 24,696 3 3 Quad Cities 1 CX 724 5.' 0.563 32 12 144 Reload CEB 28 17,100 10,900 17.5 0.563 37 12 144 1,372 1,372 fr CEH 31 17,000 11,500 13.4 0.493 34 9 144 2.268 2.268 1 U 156 6,500 4,000 .8 13.4 0.493 34 9 144 9,828 9,828 j 3 Quad Cities 2 CY 724 23,800 14,900 4.7 17.5 0.563 32 12 144 35,476 18,816 CX 31 19,700 13,800 1.2 17.5 0.563 32 12 A.4 1,519 1,372  !

! U 298 11,900 4,300 1.7/0 9 13.4 0.493 34 9 144 18,774 18,774

. CBH 14 1,400 1,000 0.1 13.4 0.493 34 9 144 892 882 j 3 Pilgrim 1 BE 580 19,700 12,381 4.0 17.5 0.563 32 12 144 28,420 20,972 BEA 20/40 11,700 4,000 2.4/0.6 13.4 0.493 34 9 144 1,260 1,260 j U 92 4,900 3,000 0.6 13.4 0.493 34 9 144 5,796 5.796  !

j 4 UM AM 232 29,100 15,000 5.3 18.5 0.563 32 12 144 11.368 0 y i i Reload CED 12 18,400 12,900 2.9 18.5 0.563 37 12 144 588 588 my j AMA 108 21,900 13,500 2.2 13.4 0.493 34 9 144 6,804 6,804 O>

I U 120 12,900 4,600 1.3 13.4 0.493 34 9 144 7.560 7.560 < 4 i e o See footnotes at end of table o l _ . . . . __ _ . . .__ _ . .. . . . _ . _ - - - . _. . ._- _ _ _ -_ -

J

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j GESSAR II 22A7007 j 238 NUCLEAR-ISLAND Rev. O i i i

Table 4.2-9 1

SUMMARY

OF GENERAL ELECTRIC OPERATING EXPERIENCE l

WITH PRODUCTION GADOLINIA-BEARING FUEL

(JUNE 30, 1976) 1 I

Number of Number of Exposure l Design Active Cadolinia Gado11nta Average Time in Peak Fuel Rod Cladding Pellet-to- Fuel i Class of- Fue! 8 earing Bearing Assembly Core LHCR Diameter Thickness Cladding Cap - Innsth I i Reactor Reactor Type Bundles Bods (mwd /t) (years) (kW/ft) (in.) (alls) (nominal mils) (in.)

i

}

j Dresden 1 s23,000 I

1 111F 104 1042 8.5 15.5 0.5625 35 10 108.25

• v 106 324 s18,000 6.5 15.5 0.5625 35 10 108.25 1 Sig Rock Point EC 38 1807 12,000 4.5 17.7 0.5625 40 11 70 F 85 340 8,000 2.8 17.7 0.Se25 40 [

11.5 70 i l '

i Humbo3dt 1 111 176 352 12,800 4.4 16.8 0.563 32 11 79 I

) '

', 1 Tarapur 1 TA 67 134 10,400 4.1 15.8 0.563 32 12 142.25 j T3 74 148 9,100 3.0 15.8 0.563 32 12 142.5  ;

i l 1 Tarapur 2 1A 37 74 8,900 3.5 15.8 0.563 32 12 ' 142.25

! TB 1% 28 7,500 2.0 15.8 0.563 32 12 142.25 L

?

l j 2 Oyster Creek JCA 156 624 12,600 4.5 17.5 0.563 32 12 144 k-t 2 Nine Mile Fotnt 1 NMA 56 224 14,000 4.6 17.5 0.563 32 12 144

CEA 40 120 17,500 4.0 17.5 0.563 32 12 144 j_ NMC 108 432 14,700 1.0 17.5 0.563 37 12 144-l NMD  % 384 9,900 2.0 13.4 0.493 34 9 144' t WO 108 432 3,000 0.5 13.4 0.493 34 9 144 4 . W2 92 368 3.000 0.5 13.4 0.493 34 9 144 i 2 Tsuruga JAA 48 192 14.100 5.6 17.5 0.563 32 12 1. 6 l JA8 85 340 3.6 17.5 0.563 I

10.600 32 12 144 a JAC 76 304 9,000 3.0 17.5 0.563 37 12 144 JAD 47 188 6,000 2.0 17.5 0.563 37 12 144 j JAE 72 288 4,000 0.7 1'.5 0.563 37 12 144 3

3 Dresden 2 CY 215 442 12,800 5.0 17.5 0,563 32 12 144 l l

DN 509 4018 7,700 4.0 17.5 0.563 32 12 144 ,

! CEB 32 96 4,100 1.0 17.5 0.563 37 12 14* '

IJO 108 432 2,900 1.0 13.4 0.493 34 9 144

CBH 8 32 1,000 0.3 13.4 0.493 34 9 144 l

! 3 Dresden 3 CEa 52 156 9,400 3.0 17.5 0.563 37 12 144 i DDB 44 176 8,100 2.0 13.4 0.493 34 9 144 WO 108 432 4.200 0.7 13.4 0.493 34 9 144 CBH 8 32 3,000 0.7 13.4 0.493 ' 34 9 144 W2 24 96 4,100 0.7 13.4 0.493 34 9 144 3 Fukushima 1 TXA 60 240 6,300 4.5 17.5 0.56 3 32 12 144 TIB 111 444 5.100 2.8 17.5 0.i63 37 12 144 5 TIC 40 160 1,500 0.5 17.5 0.563 37 12 144

?

3 Millstonet CEA 82 246 13,700 3.3 11.5 0.563 32 12 144 i 14,500 '

GE8 30 90 3.3 17.5 0.563 37 12 144 MS8 125 500 9,200 1.6 17.5 0.363 37 12 144 WO 120 480 4.900 0.6 13.4 0.493 34 9 144 W2 23 97 4,800 0.6 13.4 0.493 34 9 144 3 Monticello CEB 20 60 15,000 3.0 17.5 0.563 37 12 144 MTB 116 4t4 9,000 2.0 13.4 0.493 34 9 144 C5H 80 320 6,800 1.4 '3.4 0.493 34 9 144 ,

U2 267 801 3.600 0.6 13.4 0.493 34 9 144 t LJ3 1 3 3.200 0.6 13.4 0.493 34 9 144 3 Nucienor

• CEA 28 84 15,500 3.7 17.5 0.563 32 12 144 Nt B 68 272 13,500 3.0 17.5 0.563 37 12 144

- NUC 96 384 9,500 2.0 13.4 0.493 34 9 144 Wo 60 240 6,500 1.0 13.4 0.493 34 9 144 4.2-81

. . - _ _ _ . _ _ . _ _ _ _ . _ - _a

GESSAR II - 22A7007 238 NUCLEAR ISLAND , Rev. O s

Table 4.2-9 (Continued) ,

SUMMARY

OF GENERAL ELECTRIC OPERATING EXPERIENCE WITIf PRODUCTION GADOLINIA-BEARING FUEL (JUNE 20, 1976)

Nimber of Sanber of E st per.u r ed Design _

Active Cadolinla CaJolint.a Average Time in re air imA Rod 41ajding Pe l le t-t e* Fuel Clama of l'ue l bearing Bearing As*cmbly ' are 1.lLii 1. ~seter Thi( kness "laddir.g Cap 1.cna t h flea, t or kc4c t or Type Bundles Reds (MWJ / t ) tyears) (k'a'I f t ) (10.) (mile) (me.ina l mils) (l".4 I Pilrrim BlA 20 80 4,600 2.0 11.4 a.49; 34 9 144

$ (paJ Cities 1 CX 724 1760 13,300 4 17.5 ' fl. 561 12 12 144 LEB 24 84 8,600 2.0 11.5 v.* p ) , 37 12 144 CEH }6 144 9,140 2.0 11.4 0.4 , 14 9 144 LJO 12 48 1,400 0.3 13.4 0.4h 34 9 144 IJ 2 144 576 1,600 0.3 11.4 0.493 34 9 144 i Quad Cit ies 2 CY 724 1760 11,200 ./ 17.5 0.563 12 12 144 CX 11 92 12,400 0./ li.5 0.$63 12 '2 l ',4

!JO 143 %2 5,POO 1.2 11.4 ' u 14) 34 9 144

• Browns Ferry 1 TY $96 2117 5,774 3.0 IM.5 0.561 37 12 14 '.

4 Browns ferry 2 TZ 596 2717 2,156 5. t) 14.5 0.561 37 12 144 4 Brunswlik BR 440 1192 4,500 1.0 18.5 0.561 17 12 144 4 Cooper CZ 420 1912 9,600 2.5 18.5 0.561 37 12 146 2M 816 7,300 2.1 13.5 0.56) 17 '* 144 4 Duinne Asnold AR LD) 4 12 1,500 0.2 18.5 0.561 17 12 144 FitrPatrisk I.A 426 1976 4,300 1.6 18.5 0.561 17 12 144

%V 4 l' uk u sh t u 2 1T es t6 2000 7.400 1. ) g18.5 0.565 t 37, 12 144 ITA 9 36 1,10 1.1 /8. 5 0.56 1 1. lj 144 4 Hitih 1 HX0 560 2003 6.300 1.8 14.5 0.561 17 12 144 4 1XM CFD 12 lb 10,h00 2.5 13.5 '

O.563 37 12 144 AM 108 4 32 11,40G 1.7 11.4 0.493 34 9 144 IJ 2 f.0 200 6,100 0.8 13.4 0.491 14 9 144 4 Peas h buttea

• PH N6 1059 10,200 2.' It.5 0. ',61 17 12 144 IJ1 20 +100 'l 11 . 0.4M) 12 9 150 l_I t IM4 920 -100 0.' 11.4 0.491 34 9 144 4 Pe n h Bertom i PS %6 2717 7,400 26 18.5 0.561 17 '2 146 m.

4 Vermont Yanke. nED 40 120 8,100 2.6 18.5 0.563 37 12 s 144

1. f o 128 964 600 1.5 11.4 0.49 3 34 9 144

~

s I

Aus rbly average e x posu r e for the e assembliss remaintog in the tore.

Includes 93 ret and a r.! as., erb livs wi th I CJ;0 3-Alumina rud; 2 *.pei t al . gra h lies wi tt -. Gd;Ol -Alumf ria negment ed rod; and 4 st.n ia l asseatblive with i Gd,0 3 -tranta segmented rud.

iniludes 3 5 st anJar d asscetilca wit h 4 bado linia-Uuy r od s 4 q.eti.41 Mir assemblics wjth 4 G.klolinia-10; r ewis , and i spect.a1 Fr aasemb11em with 8 Gadolinia-UO2 rods. \

4 Quad Cit ies ' i ne l assembly t ype CX ire .e portion of t he CX asser blies originally l o.ided inte N ,J Citics 1.

.w

+

9 n.

k

~, \

4.2-82 s.-

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! .. Table 4 2-ld 3

,Y .

, GENERAL ELECTRIC DEVELOPMENTAL IRRADIATIONS 2,IRCAIOY-CLAD '

954 TD UO, PELLET FUEL' RODS *

b Number Fuel Rod Clad Wall Pellet-to- Peak Heat Peak Peak of Diameter Thickness Cladding Flux LHCR Exposure N au:e Reactcr Rods (in.) (in.) Gap (mils) Stu/h-ft2 (kW/ft) (mwd /Te) Status Dresden Prototype VBWR 9 0.565 n.030 3.0-16.0 463,000 19.94 12,000 Completed Completed Fuel Cycle (R&D)* VBWR 144 0.424 0.022 2.0-8.0 409,000 16.6 13.800

Dresden Prototypes VBWR 52 0.265 0.028 5.0-8.0 407.000 17.64 10,000 Completed High Performance GETR 12 0.565 0.030 4.0-6.0 630,000 27.0 1.500 Completed UO 1,126,000 49.0 PJ 2 w 58.0 completed' " High Performance GETR 2 0.565 0.030 4.0-11.0 1.135,000 14,000 2 UO,b CO

                     '                                                                                                                                                               OM
             . SA-1                                 Dresden 1                           0.424       0.022                        400.000         13.0       40,000   Completed 98                                    4.0-8.0
                                                                                                                                                                                        ]

O

• D-1,2,3 d

Consumers 363 0.424 0.030 7.0 434.000 14.2 30,000 Completed yy to

         '       D-50                               Consumers            36             0.570       0.035        12.0             507.000        22.0       15.400   g.1             HH W                                                                                                                                                                             U2 H U

D-52,53 Consumers 0.700 0.040 13.0 525,000 27.0 4.600 1 V 58 p Z O "USAEC Contract AT(04-3) - 189 Project Agreement 11 USAEC Contract AT(04.3) - 189, Project Agreement 17 USAEC Contract AT(04-3) - 189, Project Agreement 41 USAEC Contract AT(04-3) - 351

                 ' Hollow Pellet USAEC Contract AT(04-3) - 189, Project Agreement 50 8 Eight     fuel rods tested during second operating cycle due to abnormal crud and scale deposition One rod failure at 49 KW/ft I

Fuel assemblies presently out of reactor pending approval for reinsertion N MN (D >

                                                                                                                                                                                     <4
                                                                                                                                                                                     . O O

O -J

Table 4.2-11 GENERAL ELECTRIC DEVELOPMENTAL IRRADIATIONS ZIRCALOY-CLAD 95% TD UO 2 PELLET CAPSULES GENERAL ELECTRIC TEST REACTOR Number Fuel Rod Cladding Wall Pellet-to Peak Heat Peak Peak of Diameter Thickness Cladding Gap Flux LHGR Exposure Capsule Rods (in.) (in.) (mils) (Btu /h-ft) (kW/ft) (mwd /Te) Status l A 3 0.425 0.024-0.032 1.4-10.2 750,000 24.5 88,000 Complete 1 0.488 0.032 11.2 785,000 29.4 34,000 Complete B 6 0.489 0.034 7.8-11.6 504,000 18.9 65,000 Complete C 5 0.557 0.036 2.0-15.0 475,000 20.3 59,000 Complete 5o O$ D 5 0.557 0.036 2.0-14.0 540,000 23.0 36,500 Complete gg ww N E 5 0.250 0.015 6.5 735,000 14.1 100,000 Complete ss E $*

• F 3 0.443 0.030 3.0-13.0 480,000 16.3 29,000 Complete 5 8

x" 35 8 O -J O O O

O O O Table 4.2-12 HALDEN IRRADIATION PROGRAM STATUS Peak LHGR Peak Exposure Number Fuel Rod Cladding Wall Pellet-to- (kW/ft) (mwd /T) of Diameter Thickness Cladding Gap (as of (as of Parameter of ] Assembly Rods (in.) (in.) (mils) 9/27/74) (9/27/74) Interest IFA-131 6 0.563 0.032 8--14 16.8 30,300 Pellet geometry ! IFA-213 7 0.563 0.032 10 16.2 27,600 Cladding heat treatment N m IFA-214 7 0.563 0.032 - 0.060 8--10 8.6 18,500 Hollow pellets 2 mixed oxide CQ

                                                                                                                  &m IFA-237      1     0.563           0.032            --

18.0 28,000 Vipac powder yy Mw y IFA-238 7 0.563 0.032 - 0.040 8--10 18.? 21,400 Cladding ss ms thickness and

• heat treat-ment h

c IFA-236 6 0.493 0.034 8 12.2 16,100 Fuel density and hollow pellets IFA-408 11 0.563 0.037 1--10 16.8 12,900 Vipac, mixed oxide, hollow i pellets IFA-409 12 0.493 0.034 9 15.4 11,300 'Densification and swelling ,N

                                                                                                                  $ Es ou 8

GESSAR II 22A7007 233 NUCLEAR ISLAND Rev. O Table 4.2-13 FUEL ROD VIBRATION INFORMATION Rod OD/ Clad Calculated Rod Fuel Design Lattice Thickness In. Natural Frequency Rowe - Yankee

• 16 x 16 0.362/0.021 25.9 Nuclenor* 7x7 0.570/0.0355 29.8 Humboldt Reload 7x7 0.486/0.033 29.9 Big Rock Point 11 x 11 0.449/0.034 31.3 Reload BWR/2-5 8x8 0.493/0.034 32.8 - 35.0 BWR/2-6 8x8 0.483/0.032 31.6 - 33.7 e
• Pilot Bundles i

l \ l l J 4.2-86

   -.                    .-       . . -     - - _ - . - . ..                     ..               - _ - _ - _ _                    -   . . . =                                              ..           - -   -.

} GESSAR II 22A7007

238 NUCLEAR ISLAND Rev. O i

,                                                                                                Table 4.2-14 LINEAR HEAT GENERATION RATE OF CALCULATED 1% PLASTIC DIAMETRAL STRAIN FOR BWR/6 FUEL Exposure                                                                    LHGR at Calculated 1%

,. (mwd /t) Plastic Strain (kw/ft)* UO 2 Gd** i 0 >24.7 >21.9 20,000 >23.0 >20.4 40,000 >19.6 >l7.3

                                *The values reported have been reduced by an amount equal to the calculated power spiking penalty (%)
                               ** Results   for gadolinia are applicable for maximum concentration used in BWR/6.

I O I 1 8 4 i 4 i i

o 4.2-87/4.2-88 l

i

      .,--_.-,--..,.----.--.c.                               - - - . . - . - - -     , , . . . , - , - - - , , - - - . , , . - - -   . - - - - - . _ - , , . , , _ , , . - - - - - - _ . , , , - , , , -          . . - -

GESSAR II 22A700' 238 NUCLEAR ISLAND Rev. O O E  % f 3( h

            /                                         s

( b g , CONTROL ROD GAP gy '

            <                  > <                        /  /)               ,

r r , 3 UPPER TIE PLATE e-CENTERING SPRING l / I Illll11Il11lll

                                                                            % top GUIDE
                                                    #                   _d CHANNEL O Figure 4.2-1.       Schematic cf Four-Bundle Cell Arrangement i

! 4.2-89

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O B All CHANNEL HANDLE A FASTNER ASSEMBLY j UPPER TIE j

                  %                   PLATE N   ,,

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 800 750 - BOL 0 INSIDE u. [ 700 - 5 4 5 0 g AVERAGE W o E 8 600 - 5 o SURFACE 550 - I I  !  !  ! I I I I 500 O 50 100 150 200 250 300 050 400 450 500 HE AT F LUX (Bru/h-f t ) x 10-3 Figure 4.2-7. Cladding Temperature versus Heat Flux, Beginning of Life 850 INSIDE 800 - EOL { AVERAGE C i i 750 -

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 100

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 i ( SECTION 4.3 CONTENTS Section Title Page 4.3 NUCLEAR DESIGN (EQUILIBRIUM CORE) 4.3-1 4.3.1 Design Bases 4.3-1

4.3.1.1 Safety Design Bases 4.3-1 4.3.1.1.1 Reactivity Basis 4.3-1 4.3.1.1.2 Overpower Bases 4.3-2 4.3.1.2 Plant Performance Design Bases 4.3-2 4.3.2 Description 4.3-3 4.3.2.1 Nuclear Design Description 4.3-4 4.3.2.1.1 Fuel Nuclear Properties 4.3-5 4.3.2.2 Power Distribution 4.3-6 4.3.2.2.1 Local Power. Distribution 4.3-7 i 4.3.2.2.2 Radial Power Distribution 4.3-9

() 4.3.2.2.3 4.3.2.2.4 Axial Power Distribution Power Distribution Calculations 4.3-9 4.3-10 4.3.2.2.5 Power Distribution Measurements 4.3-10 4.3.2.2.6 Power Distribution Accuracy 4.3-10 4.3.2.2.7 Power Distribution Anomalies 4.3-10 4.3.2.3 Reactivity Coefficients 4.3-11 4.3.2.3.1 Void Reactivity Coefficients 4.3-12 4.3.2.3.2 Moderator Temperature coefficient 4.3-13 f. ! 4.3.2.3.3 Doppler Reactivity Coefficient 4.3-14

4.3.2.3.4 Power Coefficient 4.3-15 4.3.2.4 Control Requirements 4.3-15 4.3.2.4.1 Shutdown Reactivity 4.3-16 l

4.3.2.4.2 Reactivity Variations 4.3-17 i 4.3.2.5 Control Rod Patterns and Reactivity Worths 4.3-18 4.3.2.5.1 Rod Control and Information System 4.3-19 L 4.3.2.5.2 Rod Pattern Control System (RPCS) 4.3-20 4.3.2.5.3 Rod Worth Limiter (RWL) 4.3-20 N-- 4.3.2.5.4 Control Rod Operation 4.3-21 [ ! 4.3-i i i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CONTENTS (Continued) Section Title Page 4.3.2.5.5 Scram Reactivity 4.3-21 4.3.2.6 Criticality of Reactor During Refueling 4.3-22 4.3.2.7 Stability 4.3-22 4.3.2.7.1 Xenon Transients 4.3-22 4.3.2.7.2 Thermal Hydraulic Stability 4.3-23 4.3.2.8 Vessel Irradiations 4.3-23 4.3.3 Analytical Methods 4.3-24 4.3.4 Changes 4.3-24 4.3.4.1 Reactor Core 4.3-24 4.3.4.1.1 Active Core Volume Increase 4.3-25 4.3.4.1.2 Natural Uranium Utilized 4.3-25 4.3.4.1.3 Increase in Nonbolling Water Volume 4.3-25 4.3.4.1.4 Fuel Rod Diameter Reduction 4.3-26 4.3.4.1.5 Prepressurized Fuel Rods 4.3-27 ll 4.3.5 References 4.3-27 l l O 4.3-ii

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i

   /                               SECTION 4.3 TABLES Table                         , Title                     Page 4.3-1  Reactor Core Dimensions                            4.3-29 4.3-2  . Reactivity Data for the Cold, Xenon-Free State   4.3-30 4.3-3  Reactivity and Control Fraction for Various Reactor States                                     4.3-31 4.3-4  Summary of BWR/6 Design Revisions (GE Company Proprietary                                        4.3-32 4.3-5  Calculated Neutron Fluxes (Used to Evaluate Vessel Irradiation)                                4.3-33 4.3-6  Calculated Neutron Flux at Core Equivalent Boundary                                           A 3-34 ILLUSTRATIONS Figure                         Title                      Page (J-~)

4.3-1 Equilibrium Core Loading Map 4.3-35 4.3-2 BWR/6 Lattice Nominal Dimensions, 120-mil Channel 4.3-36 4.3-3 Rod Type Distribution (GE Company Proprietary) Refer to Figure 4.3-4 4.3-37 4.3-4 Axial Fuel Rod Enrichment and Gadolinia Distribution (GE Company Proprietary) 4.3-38 4.3-5 Uncontrolled k-Infinity as a Function of l Exposure at Various Void Fractions, Dominant j Fuel Type 4.3-39 l 4.3-6 Weight Fraction - U238 as a Function of 4 Exposure Dominant Fuel Type, 40% Voids 4.3-40 4.3-7 Weight Fraction as a Function of Exposure, i Dominant Fuel Type, 40% Voids 4.3-41 4.3-8 Fission Fraction as a Function of Exposure Dominant Fuel Type, 40% Voids 4.3-42 4.3-9 Neutron Generation Time as a Function of ( Exposure at 40% Voids 4.3-43 f 4.3-10 Delayed Neutron Fraction as a Function of ! Exposure at 40% Voids 4.3-44 O 4.3-iii

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O ILLUSTRATIONS (Continued) Figure Title Page 4.3-11 Variation of Maximum Local Power Peaking as a Function of Exposure, Dominant Fuel Type, 40% l

• 1 Voids, Uncontrolled 4.3-45 4.3-12 Uncontrolled Local Power Distribution as a Function of Exposure at 40% Voids, Dominant l Fuel Type (GE Company Proprietary) 4.3-46 l 4.3-13 Uncontrolled Local Power Distribution as a Function of Void at 0.0 Lattice Exposure, Dominant Fuel Type (GE Company Proprietary) 4.3-47 4.3-14 Controlled Local Power Distribution at 40%

Voids, 0.0 Lattice Exposure, Dominant Fuel Type (GE Company Proprietary) 4.3-48 4.3-15 Uncontrolled R-Factor Distribution at 40% Voids, 0.0 Lattice Exposure, Dominant Fuel Type (GE Company Proprietary) 4.3-49 4.3-16 Variation of the Maximum Uncontrolled Bundle-Integrated R-Factor as a Function of Bundle Average Exposure 4.3-50 4.3-17 Radial Power Factors for Beginning-of-Equilibrium-Cycle and Optimal End-of-Equilibrium Cycle Conditions 4.3-51 4.3-18 Beginning-of-Equilibrium Cycle and Optimal End-of-Equilibrium-Cycle Core Average Axial Power 4.3-52 4.3-19 Equilibrium Cycle Void Coefficient for Stability Analysis as a Function of Percent Voids 4.3-53 4.3-20 Dynamic Void Reactivity Coefficient as a Function of Percent Voids at End-of-Equilibrium-Cycle 4.3-54 4.3-21 Doppler Coefficient as a Function of Average Fuel Temperature at End-of-Equilibrium-Cycle 4.3-55 4.3-22 Cold Shutdown Reactivity as a Function of Cycle Exposure Strongest Rod Withdrawn, No Xenon 4.3-56 4.3-23 Banked-Position Withdrawal Sequence, RPCS Groups 1-4, Sequence A 4.3-57 4.3-24 Banked-Position Withdrawal Sequence, RPCS Groups 5-10, Sequence A 4.3-58 4.3-25 Banked-Position Withdrawal Sequence, RPCS Groups 1-4, Sequence B 4.3-59 4.3-26 Banked-Position Withdrawal Sequence, RPCS Groups 5-10, Sequence B 4.3-60 4.3-27 Hot Operating, End-of-Equilibrium-Cycle Scram Reactivity ($), as a Function of Control Fraction 4.3-61 4.3-iv

_ -..-. - -.. . . . . _ . ~ _ _ _ _ - - - . . - _- . . . .. -.. - -.. .. - i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ILLUSTRATIONS (Continued) Figure Title Page i i 4.3-28 Model for One-Dimensional Transport Analysis of

Vessel Fluence 4.3-62

]                                                4.3-29            Radial Power Distributions Used in the Vessel Fluence Calculation                                                                                 4.3-63 i

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ( 4.3 NUCLEAR DESIGN The nuclear core design presented herein is based on the equilib-rium cycle rather than the initial cycle. The justification _for this feature is presented in Section 4.1. The design bases and licensing requirements are independent of whether an initial or equilibrium cycle is used. Only the description and the core characteristics are different. 4.3.1 Design Bases The nuclear design bases are conveniently divided into two specific categories. The safety design bases are those that are required for the plant to operate from safety considerations. The second category is the plant performance design bases that are required in order to meet the objective of producing power in an efficient manner. O Safety Design Bases 4.3.1.1 The safety design bases are requirements which protect the nuclear fuel from a violation of the design integrity limits. In general, the safety bases fall into two categories: (1) the reactivity basis which prevents an uncontrolled positive reactivity excursion, and (2) the overpower bases, which prevent the core from operating beyond the fuel integrity limits. 4.3.1.1.1 Reactivity Basis The nuclear design shall meet the following basis: The core shall be capable of being rendered suberitical at any time or at any core conditions with the highest worth control rod fully withdrawn. O 4.3-1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.1.1.2 Overpower Bases The nuclear design shall meet the following bases: (1) The void coefficient shall be negative over the entire operating range. (2) The Technical Specification limits on Linear Heat Generation Rate (LHGR), Minimum Critical Power Ratio (MCPR), and the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) shall not be exceeded during steady-state operation. (3) The nuclear characteristics of the design shall exhibit no tendency toward divergent operation. 4.3.1.2 Plant Performance Design Bases The nuclear design shall meet the following bases: (1) The design shall provide adequate excess reactivity to attain the desired cycle length. (2) The design shall be capable of operating at rated con-ditions without exceeding Technical Specification limits. (3) The nuclear design and the reactivity control system shall allow continuous, s table regulation of reactivity. (4) The nuclear design shall have adequate re,ctivity feed-back to facilitate normal operation. (5) Separate bases on the void coefficient are defined with respect to core stability analysis and plant transient analyses. 4.3-2

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 (q

 ,     ! 4.3.1.2   Plant Performance Design Bases (Continued) m/

a. The magnitude of the maximum void coefficient at any point in the cycle shall not exceed a specified design limit used to evaluate core reactivity stability.

b. The dynamic void reactivity coefficient, which is used to measure the core transient response, shall not exceed a design limit established to ensure optimum plant performance. The dynamic void reactivity coefficient is defined as the nuclear void reactivity coefficient, expressed in cents of reactivity, multiplied by the core average void fraction, expressed in percent.

fs (6) The scram reactivity response shall not violate a design ( ,/ limit established to insure optimum plant performance. This design limit must be met during the first 60% of scram insertion, since the transient response is insensi-tive to the remainder of the scram curve. The scram reactivity is evaluated at end-of-cycle for the 105% steam flow reactivity condition. (7) The Doppler reactivity coefficient shall not violate a design limit established to ensure optimum plant per-formance. This design limit is established for the end-of-cycle condition at 105% steam flow. 4.3.2 Description The BWR core design utilizes a light-water moderated reactor, fueled with slightly enriched uranium-dioxide. The use of water as a

   ,-s   moderator produces a neutron energy spectrum in which fissions are

(_) caused principally by thermal neutrons. At normal operating con-( ditions, the moderator boils, producing a spatially variable 4.3-3 L

GESSAR II 22R7007 238 NUCLEAR ISLAND Rev. 0 4.3.2 Description (Continued) distribution of steam voids in the core. The BWR design provides a O system for which reactivity changes are inversely proportional to the steam void content in the moderator. This void feedback effect is one of the inherent safety features of the BWR system. Any sys-tem input which increases reactor power, either in a local or gross sense, produces additional steam voids which reduce reactivity and thereby reduce the power. The fuel for the BWR is uranium-dioxide enriched to approximately 3 wt% in U-235. Early in the fuel life, the fissioning of the U-235 produces the majority of the energy. The presence of U-238 in the uranium-dioxide fuel leads to the production of appreciable quantities of plutonium during core operation. This plutonium contributes to both reactivity and reactor power production (i.e., approximately 50% at end-of-life). In addition, direct fission-ing of U-238 by fast neutrons yields approximately 7 to 10% of the total power and contributes to an increase of delayed neutrons in the core. Since the U-238 has a strong negative Doppler reactivity coefficient, the peak power during an excursion is limited. The reactor core is arranged roughly as a right circular cylinder containing a large number of fuel assemblies and control rods. The dimensions of the reactor core are shown in Table 4.3-1. 4.3.2.1 Nuclear Design Description At each refueling period, approximately 25% of the fuel bundles are discharged from the core and replaced with an equivalent number of fresh fuel assemblies. The fuel bundles having the highest exposure (i.e., the lowest reactivity) are discharged starting with the highest exposure and moving toward less exposure. The bundles are then shuffled in order to minimize radial power peaking and maximize the end-of-cycle reactivity. This is accom-plished by loading the lowest reactivity fuel on the periphery, h 4.3-4

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O f'm 4.3.2.1 Nuclear Design Description (Continued) Y,g] loading the relatively high reactivity fuel in a region next to the periphery toward the core center, and loading the medium reactivity fuel in the central region of the core. Within each of these zones, the fuel bundles are arranged in a nearly homogeneous manner in order to minimize reactivity mismatch. A diagram of the equilibrium core bundle loading pattern is shown in Figure 4.3-1. Each bundle contains 62 fuel rods and 2 water rods. The water rods have a slightly larger diameter than the fuel rods. The layout and dimensions of the fuel bundle are presented in Fig-ure 4.3-2. Gadolinium, in the form of Gd 023, is mixed with the UO2 and placed in selected fuel rods to provide reactivity control. The fuel rod distribution for the reload bundle is shown in Fig-ure 4.3-3. The axial distribution of enrichment and gadolinia for the rods in the fuel bundle is shown in Figure 4.3-4. Each fuel and burnable poison rod has natural uranium at the top and bottom. 4.3.2.1.1 Fuel Nuclear Properties The bundle reactivity is a complex function of several important physical properties. The important properties consist of the average bundle enrichment, the gadolinia rod location and gadolinia concentration, the void fraction and the accumulated exposure. The variation in reactivity of the infinite lattice data (k-infinity as a function of void fraction and exposure) for the dominant segment of the bundle is presented in Figure 4.3-5. At low exposure, the reactivity effect due to void formation is readily apparent; however, at higher exposure, due to the effect of void history, the curves cross. The primary reason for this behavior is the greater rate of plutonium formation at the higher void fraction. The variation of the isotopic concentrations as a function of exposure for the dominant fuel type is presented in Figures 4.3-6 and 4.3-7 for the important heavy element isotopes. 4.3-5

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.2.1.1 Fuel Nuclear Properties (Continued) Early in the fuel bundle life, approximately 93% of the power is produced by fissions in U-235 with the remainder coming from fast fissions in U-238. At high bundle exposures, typical of discharge, the power production due to plutonium exceeds that of the U-235. The fraction of total fissions in the important isotopes is shown in Figure 4.3-8 for the dominant fuel type. Other bundle parameters such as neutron generation time and delayed neutron fraction as a function of exposure at approxi-mately core average voids are shown in Figures 4.3-9 and 4.3-10, respectively, for the dominant segment of the bundle. The variation of the core-wide nuclear characteristics is a function of the characteristics of each bundle in the core. With the various reload situations, any description of the gross core characteristics can only be expressed in terms of the overall core performance. 4.3.2.2 Power Distribution The core is designed such that the resultant operating power distributions meet the plant Technical Specifications. The three criteria for thermal limits are the Maximum Linear Heat Generation Rate (MLHGR) , the Minimum Critical Power Ratio (MCPR), and a Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) (Reference 1). Each of these thermal limits is a function of both the gross three-dimensional power distribution and the local rod-to-rod power distribution. Sufficient design calculations are performed to insure the core meets these criteria. For design convenience, separate target peaking factors are used for the local and the gross power distributions. The local peaking factor is defined as the ratio of the power density in the highest power rod in the lattice, at a cross section through the bundle, to the average 4.3-6

                                                                                 .    . = . .           -                    - - _ .

I j GESSAR II 22A7007 j 238 NUCLEAR ISLAND Rev. 0 ,

                         ) 4.3.2.2     Power Distribution (Continued) power density in the 19ttice at Lhat location.                                                   In addition, the local effects on the Critical Fowar Ratio (CPR) are characterized by a quantity designated as R-Factor (Peference 2). For.the'BWR/6 j                           nuclear fuel design, the target local peaking factor is 1.13 and the i                           target R-factor is 1.05.                             The gross power peaking is defined as the ratio of the maximum power density in any axial segment of any bundle 2

in the core to the average power density in the core. The target . gross power peaking limit for this BWR/6 fuel design is 2.32. i j Variations from these target peaking factors are considered

acceptable providing the Technical Specifications are not exceeded

anywhere in the core. That is, the real criteria are the LHGR, i

the MAPLHGR, and the MCPR; not the power peaking factors. The peaking factors are used merely as a design convenience. Appro- ,

gg priate design allowances are included at the design stage to (m l provide a high degree of assurance that the Technical Specification limits will be met during plant operation. During operation of the j plant, the power distributions are measured by the in-core instru-mentation system, and thermal margins are calculated by the process I computer. " l l 4.3.2.2.1 Local Power Distribution i The local rod-to-rod power distribution and the associated R-Factor distribution are functions of the lattice fuel rod enrichment distribution. Near the outside of the lattice, where l the thermal flux peaks due to interbundle water gaps, lower i enrichment fuel rods are utilized to minimize power peaking.

Closer to the center of the bundle, higher enrichment fuel rods are used to increase the power generation and flatten the power distribution across the bundle. In addition, two water tubis

containing unvoided water are at the center of the lattice._ The j

combination of these factors results in the relatively flat local i s. i 4.3-7 l

  -- . . . . - -. -- . -                 .-      - - , - . _ - ~ . . . , _                 . - , . .        . _ - - . . - _          ,

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.2.2.1 Local Power Distribution (Continued) power distribution. The fuel rods which contain gadolinia pro-duce relatively little power early in bundle life; however, as the gadolinia is depleted, the power in these rods increases to approximately the lattice average. The variation of the maximum local peaking f actor as a function of exposure at core average condition is shown in Figure 4.3-11. The high power rods deplete at a greater rate and the local power peaking factor decreases with exposure. The local rod-to-rod power distribu-tion for core average voids at beginning-of-cycle, at an exposure typical of end-of-equilibrium-cycle, and end-of-bundle life is shown in Figure'4.3-12. Tne variation in local power distribu-tion for various lattice void fractions is shown in Figure 4.3-13. In general, the local power distribution tends to flatten with increasing void fraction. It can be seen from these data that the target local peaking factor of 1.13 is met over the range of interest. Although the 1.13 target value is exceeded at very low exposures, the reactivity of the bundle is suf ficiently low that the bundle is never limiting. At exposures where this bundle produces significant power, the local peaking factor is well within the target. The presence of a control blade adjacent to the bundle significantly skews the local power distribution within the bundle. The local power distribution for the controlled lattice is shown in Figure 4.3-14. Although the local peaking factor is quite large in this case, the gross power in a con-trolled bundle is sufficiently low such that a controlled lattice is never limiting. Figure 4.3-15 presents the uncontrolled R-Factor for each fuel rod at a planar elevation through the dominant fuel type at beginning-of-cycle at core average conditions. The variation of the maximum bundle-integrated R-Factor with exposure is presented in Fig-ure 4.3-16. The data shown demonstrate that the 1.05 target lh limit is not exceeded for the exposure range of interest. 4.3-8

7

                                      +                             . . +               -

g ] G!)GSAR II 4 22A7007 2' 238 NGCLEAR ISLA!!D Rev. 0 , 1 s 1 4.3.2.2.2 Radial Power Distributiori' .

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The integrated bundle power, commonly referred to as the radial power, is the primary factor:for' determining MCPR. 5t: rated conditions, the MCPR is directly proportional to thC radial pywer peaking. The radial power distribution in ta comple2" function of , i the control rod pattern, the fuel bundle type, the ' loading pat- 1 torn and the void condition for that bundle. The DUR.. simulator e kL is used to calculate the three-dimansional power dintribution in the core and the power is axially integrated to determine average \ '; bundle power. The bundle radial power distributions,for typical , beginning and end-of-cycle conditions are presented in Fig-ure 4.3-17. - The radial power distribution is influenced by both the radial reactivity zones and the control rods. The control rods are , selectively inserted, or withdrawn, to flatten the radial power Ce. distribution consistent with the reactivity control needed. Near. , the end-of-cycle, the region of high reactivity adjacent to the g , periphery provides the necessary radial power flattening withouts,

                                                                                                                 +- s'   .-

recourse to control rods. *

                                                                                                 .t
                                                                                         ',     d   '

4.3.2.2.3 Axial Power Distribution For the reload situation, the axial exposure sha'pe existing in the bundles which remain from previous cycles.', in combination with the selective use of shallow control rod.s, provide the s . i necessary axial power flattening. Current' reload band,le designs , contain axially uniform gadolinia. The effpct[of vdids i.s s to skew 3 the power toward the bottom of the core; the,effect of th.e

                                                             ..                 .- bottom-o         ,

entry control rods is to reduce the power'in the bottom of the core; and the effect of the exposure distribution is to flatten the power. Since the void distribution is det, ermined primarily

    -g   from the power shape, the mechanism available for further opti-s_ /  mizing the axial power shape is the control rods. Detailed 4.3-9       ,

e

GELSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.2.2.3 Aylal Power Distribution (Continued) three-diuensional calculations are performed to determine the O axial power distribution. s The Control rod patterns help to achieve an end-of-cycle exposure distribution which approximates the IIaling shape (Reference 14) . This, in turn, provides a relatively flat axial power shape with all control rods withdrawn. A typical beginning-of-equilibrium-cycle axial power shape is shdwn in Figure 4.3-18 along with the optimal end-of-equilibrium-cycle power shape.

                   '4.3.2.2.4  Power Distribution Calculations A full range of calculated power distributions along with the resultant exposure shapes and the corresponding control rod patterns are shown in Appendix 4A for a typical BWR/6 equilibrium cycle.

4.3.2.2.5 Power Distribution Measurements

                   ?ne techniques for a measurement of the power distribution within the reactor core, together with instrumentation correlations and operation limits, is discussed in Reference 4.

4.3.2.2.6 Power Distribution Accuracy

                   @hd accuracy of the calculated local rod-to-rod power distribu-tion is discussed in Reference 5. The accuracy of the radial, axial and the gross three-dimensional power distribution calcu-
 ',                 lations is discussed in Reference 6.

4.3.2.2.7 Power Distribution Anomalies

         ,         Stringent inspection procedures are utilized to ensure the cor-rect rearrangement of the core following refueling. Although a O

4.3-10

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.2.2.7 Power Distribution Anomalies (Continued) (} misplacement of a bundle in the core would be a very improbable event, calculations have been performed in order to determine the effects of such accidents on linear heat generation rate (LHGR) and critical power ratio (CPR). These results arc presented in Chapter 15. 4 The inherent design characteristics of the BWR are well suited to limit gross power tilting. The atabilizing nature of the large moderator void coefficient effectively reduces perturbations in the power distribution. In addition, the in-core instrumentation system, together with the on-line computer, provides the operator with prompt information on power distribution so that he can readily use control rods or other means to limit the undersirable effects of power tilting. Because of these design characteristics, it is not necessary to allocate a specific margin in the peaking factor to account for power tilt. If, for some reason, the power distribution could not be maintained within normal limits using control rods, then the operating power limits would have to be i reduced as prencribed in Chapter 16 (Technical Specifications) , i 4.3.2,3 Reactivity Coefficients Reactivity coefficients, the differential changes in core reactiv-ity prot 1 by differential changes in core conditions, are use-ful in ( olating the response of the core to external disturbances. The base initial condition of the system and the postulated initia-ting event determine which of the several defined coefficients are significant in evaluating the response of the reactor. There are three primary reactivity coefficients which character-ize the dynamic behavior of BWRs over all operating states: r (1) Doppler reactivity coefficient; (2) moderator temperature i ( () reactivity coefficient; and (3) moderator void reactivity coefficient. Also associated with the BWR is a power reactivity 4.3-11 I

CESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.2.3 Reactivity Coefficients (Continued) coefficient; however, this coef ficient is merely a combination of the Doppler and void reactivity coefficients in the power operating range. In order to assure optimum plant performance determined by the predicted plant tra.sient response, design limits have been established on those nuclear parameters which have a significant. effect on the plant transients. Specifically, design limits have been established for the void coefficient, the Doppler coefficient, and the scram reactivity response. These limits were established such that the proper balance was made between the nuclear design and the plant design. The design limits presented herein repre-sent the extreme values expected to occur during normal operation of the core over the lifetime of the plant. These reactor dynamics calculations are indicative of core responses to individual phenomena. The code used to calculate transient response utilizes all of the specific inputs to predict the core response to a particular transient. 4.3.2.3.1 Void Reactivity Coefficients The most important of the reactivity coefficients is the void reactivity coefficient. The void coefficient must be large enough to prevert power oscillation due to spatial xenon changes yet small enough that pressurization transients do not unduly limit plant operation. In addition, the void coefficient in a BWR has the ability to flatten the radial power distribution and to provide case of reactor control due to the void feedback mechanism. The overall void coefficient is always negative over the complete operating range since the B'.!R design is undermoderated. The reactivity change due to the formation of voids results from the reduction in neutron slowing down due to the decrease in the water-to-fuel ratio. h 4.3-12

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O "s 4.3.2.3.1 Void Reactivity Coefficients (Continued) A detailed discussion of the methods used to calculate void reactivity coefficients, their accuracy and their application to plant transient analysis is presented in Reference 7. A comparison of a detailed model using spatial variation of the important parameters and the point model is also shown. The moderator void reactivity coefficient as a function of per-cent voids is presented in Figure 4.3-19 for the exposure at which the void coefficient reaches its maximum. The most limit-ing transient response occurs at end-of-cycle. The variation of the calculated dynamic void reactivity coefficient, as a function of core average void percent is shown in Fig-ure 4.3-20 for the end-of-cycle condition. Also shown is the comparison of the design limit and the calculated nominal value () at 105% steam flow conditions multiplied by 1.25, the safety conservatism factor. These data demonstrate that the desi_a limit value is not exceeded. 4.3.2.3.2 Moderator Temperature Coefficient The moderator temperature coefficient is the least important of the reactivity coefficients, since its effect is limited to a very small portion of the reactor operating range. Once the reactor reaches the power producing range, boiling begins and the moderator temperature remains essentially constant. As with the void coefficient, the moderator temperature coefficient is associated with a change in the moderating power of the water. The temperature coefficient is negative for most of the operating cycle; however, near the end-of-cycle the overall moderator 4.3-13

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.2.3.2 Moderator Temperature Coef ficient (Continued) temperature coefficient becomes slightly positive, due to the fact that the uncontrolled BWR lattice is slightly overmoderated in the unvoided state at high exposures. This, combined with the fact that more control rods must be withdrawn from the reactor core near the end-of-cycle to establish criticality, results in the slightly positive overall moderator temperature coefficient. The range of values of moderator temperature coefficients encountered in current BWR lattices does not include any that are significant from the safety point of view. Typically, the 5 temperature cot f ficient may range from +4 x 10 6k/k'F to 5 -14 x 10 ak/k F, depending on base temperature and core exposure. The small magnitude of this coefficient, relative to that associated with transfer of heat from the fuel to the coolant, makes the reactivity contribution of moderator tempera-ture change insignificant during rapid transients. For the reasons stated previously, current core design criteria do not impose limits on the value of the temperature coefficient, and effects of minor design changes on the coefficient in mem-bers of the same class of core usually are not calculated. A measure of design control over the temperature coefficient is exercised, however, by applying a design limit to the void coefficient. This constraint implies control over the water-to-fuel ratio of the lattice; this, in turn, controls the tempera-ture coefficient. Thus, imposing a quantitative limit on the void coef ficient effectively limits the temperature coef ficient. 4.3.2.3.3 Doppler Reactivity Coef ficient The Doppler reactivity coefficient is the change in reactivity due to a change in the temperature of the fuel. This reactivity change is due to the broadening of the resonance cross sections as the fuel temperature increases. At beginning-of-life, the 4.3-14

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 y s 4.3.2.3.3 Doppler Reactivity Coefficient (Continued) Doppler contribution is primarily due to U-238; however, the buildup of Pu-240 with exposure adds to the Doppler coefficient. A detailed discussion of the methods used to calculate the Doppler coefficient, their accuracy, and application to plant transient analyses is presented in Reference 7. The application of the Doppler coefficient to the analysis of the Rod Drop Accident is discussed in Reference 8. The variation in the core average Doppler reactivity co9f ficient as a function of average lattice fuel temperature at end-of-equilibrium cycle and with the multiplier of 1.25, the safety conservatism factor, are shown in Figure 4.3-21. Also shown is the design limit for the Doppler coefficient. These data demon-strate that the design limit on the Doppler coefficient is not violated. O 4.3.2.3.4 Power Coefficient The power coefficient is determined from the composite of all the significant individual sources of reactivity change associated with a differential change in reactor thermal power assuming xenon reactivity remains constant. At end-of-equilibrium-cycle, the power coefficient at 105% steam flow conditions is approxi-mately -0.05 Ak/k i AP/P. This value is well within the range required for adequately damping power and spatial-xenon disturbances. 4.3.2.4 Control Requirements The nuclear design in conjunction with the reactivity control system provide an inherently stable system for BWRs. O 4.3-15

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.2.4 Control Requirements (Continued) The control rod system is designed to provide adequate control of the maximum excess reactivity anticipated during the equilibrium cycle operation. Since fuel reactivity is a maximum and control is a minimum at ambient temperature, the shutdown capability is evaluated assuming a cold, xenon-free core. The safety design basis requires that the core, in its maximum reactivity condition, be suberitical with the control rod of the highest worth fully withdrawn and all others fully inserted. 4.3.2.4.1 Shutdown Reactivity To assure that the safety design basis for shutdown is satisfied, an additional design margin is adopted: k-effective is calculated to be less than or equal to 0.99 with the control rod of highest worth fully withdrawn. The cold shutdown reactivity as a function of cycle exposure is shown in Figure 4.3-22. k The shutdown reactivity curve shows the calculated values of k-effective for the condition 20 C, highest worth rod withdrawn, no xenon. The minimum shutdown margin occurs when bundles which have been in the core for one cycle are at their peak reactivity. Min-imum shutdown margin can occur at any point in the equilibrium cycle depending upon the fuel loading pattern and the reload zuel characteristics. This variation is not of importance as long as adequate shutdown margin exists. Heating the reactor to hot con-ditions will increase the shutdown margin by 0.02 ok to 0.03 /k. For this reason, shutdown margin calculations are not necessary for hot conditions. The accuracy with which shutdown reactivity is calculated is discussed in Reference 6. Basically, the accuracy is character-ized as a bias and an uncertainty. The bias is a reactivity O 4.3-16

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.2.4.1 Shutdown Reactivity (Continued) correction applied directly to the calculated results. For example: k eff (Expected) =k eff (Calculated) + Ak (Bias) This bias has been incorporated into the shutdown curve shown in Figure 4.3-22. The 1% design margin is satisfied after the bias correction is applied. Reduction of control rod effectiveness during the operating cycle is not a major concern with the BWR. Using normal control rod sequencing, the control rod worth remains essentially constant over the BWR operating cycle. 4.3.2.4.2 Reactivity Variations O k,) The excess reactivity designed into the core is controlled by the control rod system supplemented by gadolinia-urania fuel rods. The reload fuel enrichment for the cycle is chosen to provide excess reactivity in the fuel assemblies sufficient to overcome the neutron losses caused by core neutron leakage, moderator heating and boiling, fuel temperature rise, equilibrium xenon and samarium poisoning, plus an allowance for fuel depletion. Control rods are used during the cycle partly to compensate for burnup and partly to flatten the power distribution. Reactivity balances are not used in describing BWR behavior because of the strong interdependence of the individual con-stituents of reactivity. Therefore, the design process does not produce components of a reactivity balance at the conditions of interest. Instead, it gives the k representing all effects eff combined. Further, any listing of components of a reactivity

  ) balance is quite ambiguous unless the sequence of the changes is clearly defined.

4.3-17

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.2.4.2 Reactivity Variations { Continued) Consider, for example, the reactivity effect of control rods and burnable poison. The combined worth of these two absorbers would be considerably different than the sum of their individual worths. Even this combined worth would be of questionable significance unless the path and conditions of other parameters (i.e., tempera-ture, void, xenon, etc.) were completely specified. Many other illustrations could be presented showing that the reactivity bal-ance approach, which may be appropriate in some types of reactors, is completely inappropriate in a BWR. This is related to the large potential excess reactivity in a BWR combined with the dependence of interaction (shadowing) factors on reactor state. In order to understand the various reactivity effects in a BWR design, certain reactivity states can be defined which provide information about BWR behavior. Typical data are presented in Table 4.3-2 and show the predicted reactivity, k eff, for v ri us cold, xenon-free conditions. The reactivity and control fraction values for a variety of oper-ating conditions are shown in Table 4.3-3. The worth of various reactivity effects can be estimated by taking the differences between reactivity states with all but one variable constant. Estimates of the temperature defect, the power defect, the xenon defect and the excess reactivity can be inferred. 4.3.2.5 Control Rod Patterns and Reactivity Worths For BWR plants, control rod patterns are not uniquely specified in advance; rather, during normal operation, the control rod patterns are selected based on the measured core power distributions, within the constraints imposed by the systems indicated in the following sections. Typical control rod patterns are calculated during the design phase tc insure that all s ifety and performance criteria l 4.3-18

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

^

4.3.2.5 Control Rod Patterns and Reactivity Worths (Continued) V are satisfied. Control rod patterns and the associated power distributions for a typical BWR are presented in Appendix 4A. These control rod patterns are calculated with the BWR Core Simulator (Reference 3). The ability of this model to predict control rod worth can be inferred from the detailed reactivity data presented in Refer-ence 6. Actual operating reactor rod patterns will be based upon the measured power distributions in the plant. All rod patterns will be such that the limits specified in the Technical Specifica-tions (Chapter 16) will be met throughout the cycle. The compari-sons of calculated and measured reactivity for the cold condition in both an in-sequence critical, where roughly 25% of the cornrol rods are withdrawn, and the stuck rod measurement, where only one or two rods are withdrawn, show the ability of the model to predict rod worth. The data presented in Table 7 of Reference 6 show that (l \- S no significant bias exists between these two configurations; there-fore, it is concluded that the worth of the rods is accurately pre-dicted for the cold condition. Figure 44 of Reference 6 shows the calculated critical reactivity for a variety of BWR cores and over a wide range of exposures. Since this represents a large variation of the number of control rods inserted and, since no significant bias is observed, it is concluded that rod worths for the hot operating condition are accurately predicted. 4.3.2.5.1 Rod Control and Information System Control rod patterns and associated control rod reactivity worths are regulated by the Rod Control and Information System (RCIS). This system utilizes redundant inputs to provide rod pattern control over the complete range of reactor operations. The control rod worths are limited to such an extent that the rod drop accident (RDA) and the power range rod withdrawal error (RNE) {" \ become unimportant. The RCIS provides for stable control of core 4.3-19

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.2.5.1 Rod Control and Information System (Continued) reactivity in both the single rod or ganged rod mode of operation. The Rod Pattern Control System (RPCS) of the RCIS provides pro-tection from a RDA from startup to the low power setpoint (LPSP), about 20% of rated power. The Rod Withdrawal Limiter (RWL) function provides protection from the RWE for all conditions above the LPSP. Each of these functions is described in the following sections. 4.3.2.5.2 Rod Pattern Control System (RPCS) The RPCS restricts control rod patterns to prescribed withdrawal sequences from the all-rods-inserted startup condition until about 20% of rated power. The withdrawal mode, called bank position withdrawal sequence, minimizes control rod worths to the extent that they are not an important concern in the operation of a BWR. The consequences of a RDA or a RWE in this range are significantly less severe than that required to violate fuel integrity limits. This system is described in detail in Reference. 9. Above 20% of rated power, control rod worths are very small due to the formation of voids in the moderator. Therefore, restrictions on control rod patterns are not required to minimize control rod worths. 4.3.2.5.3 Rod Withdrawal Limiter (RWL) Above the LPSP, the RCIS relies on the RWL function to provide regulation of control rod withdrawals in order to prevent the occurrence of a rod withdrawal error. This function limits the withdrawal of a single control rod or a gang of control rods to a predetermined increment, depending on whether the power level is above or below the high power setpoint (HPSP), typically 70% reactor power. The system senses the location of the rod or gr.ng of rods and automatically blocks withdrawal when the preset incre-ment is reached. The preset limit is determined by generic analy- g sis such that the AMCPR and ALHGR for the RWE are less than the 4.3-20

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.2.5.3 Rod Withdrawal Limiter (RWL) (Continued) (} limiting transient. Above the HPSP, the rod will block at 12 in, withdrawal. Below the HPSP, the increment is 24 in, withdrawal. 4.3.2.5.4 Control Rod Operation i The control rods can be operated either individually or in a gang composed of up to four rods. The purpose of the ganged rods is I to reduce the time required for plant startup or recovery from a I scram. The RCIS provides regulation of control rod operation re-gardless of whether rods are being moved in single or ganged mode. i The assignment of control rods to RCIS groups is shown in Fig-ures 4.3-23 and 4.3-24 for sequence A and Figures 4.3-25 and 4.3-26 i for sequence B. Also shown in these figures is the division of the groups into gangs of 1 to 4 rods which can be moved simultaneously. O j 4.3.2.5.5 Scram Reactivity 1 The Reactor Protection System (RPS) responds to certain abnormal operational transients by initiating a scram of all control rods. The RPS and the Control Rod Drive (CRD) System act quickly enough to prevent the initiating disturbance from driving the fuel beyond transient limits. The scram reactivity curve at the end-  ; of-equilibrium-cycle is shown in Figure 4.3-27. Also shown is the design limit scram curve. f i At the hot-operating condition, the control rod, power, delayed f neutron and void distributions must all be properly accounted for as a function of time in the transient analysis. Therefore, the scram reactivity is calculated with an integrated one-dimensional reactor core computer model which is coupled to recir-culation and major system control models (Reference 10). i 4.3-21 w- -

                 ,we        -c-ptw-e----      -
                                                         --r    e-r -+wm     ----w----   - - - - - -   ,w-t   ,--r--e---   - - - - - < - -~m- - -e -

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

4. 3. 2. 6 Criticality of Reactor During Refueling The maximum allowable value of k eff is <l.000 at any time. For each reload cycle, the maximum core reactivity is calculated with the highest worth rod withdrawn to show at least 1.0% Ak margin.

A control rod system interlock prevents the withdrawal of more than one rod while in the REFUEL mode, Another refueling interlock prevents the movement of the refuel-ing platform over the core when all control rods are not fully inserted. Technical Specifications (Chapter 16) allow the bypassing of these two interlocks under procedural control: more than one control blade can be withdrawn in the refuel mode only if all four fuel assemblies in the cell surrounding every withdrawn blade are removed. These special controls result in a core k ff less than that of the fully controlled core without the procedural bypasses. The k f a fully ntrolled core is typically %0.95 eff and this large margin is much more than required for the small reactivity perturbations occurring during fuel shuffling. 4.3.2.7 Stability 4.3.2.7.1 Xenon Transients Boiling water reactors do not have instability problems due to xenon. This has been demonstrated by operating BWRs for which xenon instabilities have never been observed (such instabilities would readily be detected by the LPRM's) by special tests which have been conducted on operating BWRs in an attempt to force the reactor into xenon instability, and by calculations. All of these indicators have proven that xenon transients are highly damped in a BWR due to the large negative power coefficient. Analysis and experiments conducted in this area are reported in Reference 11. 4.3-22

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.2.7.2 Thermal Hydraulic Stability l This subject is covered in Subsection 4.4.4.6. 4.3.2.8 Vessel Irradiations The neutron fluxes at the vessel have been calculated using the one-dimensional discrete ordinates transport code described in Subsection 4.1.4.5. The discrete ordinates code was used in a distributed source mode with cylindrical geometry. The geometry described six regions from the center of the core to a point beyond the vessel. The core region was modeled as a single homogenized cylindrical region. The coolant water region between the fuel channel and the shroud was described containing saturated water at 550*F and 1050 psi. The material compositions for the stainless steel in the shroud and the carbon steel in the vessel

 'g contain the mixtures by weight as specified in the ASME material specifications for ASME SA 240, 304L, and ASME SA 533 grade B.

In the region between the shroud and the vessel, the presence of the jet pumps was ignored. A simple diagram showing the regions, dimensions, and weight fractions are shown in Figure 4.3-28. The distributed source used for this analysis was obtained from the gross radial power description. The distributed source at i any point in the core is the product of the power from the power j description and the neutron yield from fission. By using the neutron energy spectrum, the distributed source is obtained for l position and energy. The integral over position and energy is normalized to the total number of neutrons in the core region. The core region is defined as a 1 centimeter thick disc with no transverse leakage. The power in this core region is set equal i to the maximum power in the axial direction. The radial and ) optimum axial power distributions are shown in Figures 4.3-29 and j 4.3-18, respectively. i 4 4.3-23 i i L

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.2.8 Vessel Irradiations (Continued) The neutron fluence is determined from the calculated flux by assuming that the plant is operated 90% of the time at 90% power level for 40 years or equivalent to 1 x 10 full power seconds. The calculated fluxes and fluence are shown in Table 4.3-5. The calculated neutron flux leaving the cylindrical core is shown in' Table 4.3-6. 4.3.3 Analytical Methods The analytical methods and nuclear data used to determine the nuclear characteristics are similar to those currently in use for design and analysis of light water moderated reactors. The Lattice Physics Model (Reference 12) is used to calculate lattice reactivity characteristics, *ew group flux averaged cross sections and local rod-to-rod power and exposure distributions. l) These data are generated for various temperature, void, exposure and control conditions as required to represent the reactor core behavior. The BWR Core Simulator (Reference 3) is a large, three-dimensional code which provides for spatially varying voids, control rods, burnable poisons, xenon and exposure. This code is used to calcu-late three-dimensional power and exposure distributions, control rod patterns and thermal-hydraulic characteristics throughout core life. These methods have been compared extensively to experiments and plant operating data. The results are presented in References 5 and 6. O 4.3-24

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.4 Changes 4.3.4.1 Reactor Core Relative to the previous core design (documented in Reference 13) this reactor core incorporates the design changes described in the following sections. Although these features are new to the BWR/6 standard plant, they have already been incorporated into reload , fuel and core designs and have been successfully demonstrated in numerous operating plants. 4.3.4.1.1 Active Core Volume Increase j The active core height is increased by 2 inches. This is made possible by the lower fission gas release in the 8x8 design, which allows a reduction in the length of the fission gas plenum. From this increase in active core height, the core average power density (~N is reduced which results in increased margins in fuel duty, thermal b hydraulics and power peaking performance. Table 4.3-4 summarizes the-changes in core design. i 4.3.4.1.2 Natural Uranium Utilized i Natural uranium is incorporated into the top and bottom 6 inches ! of the lengthened active fuel region. This change increases cold shutdown margin and reduces the core average enrichment for a fixed j energy production. i i 4.3.4.1.3 Increase in Nonboiling Water Volume The number of water tubes was increased from one to two and their i outside diameter has enlarged from 0.493 to 0.591 in. (Table 4.3-4) plus the size of the channel surrounding the fuel has been reduced. I The larger water rods tend to reduce the maximum local power factor 7 and decrease the amount of fissile inventory required to achieve , (} a fixed energy production. The increase in the ratio of nonboiling 4.3-25

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.3.4.1.3 Increase in Nonboiling Water Volume (Continued) to boiling water due to the channel and water rod changes results in an additional reduction in the magnitude of the void coefficient. The changes to the void coefficient, due to the two water tubes, flatten the axial power distribution which more than offsets the effect of the natural uranium on the axial power peaking. Ilydraulic stability margins at natural circulation and the effects of pressurization transients are also improved by the reduction in the void coefficient. Cold shutdown margin is improved since the large water rods increase the relative burnup in the top of the core and decrease the hot-to-cold reactivity swing. Part of this reduction in hot-to-cold swing can be used to increase hot excess reactivity to provide additional margin against uncertainties in nuclear predictions and to provide more power-shaping flexibility early ) in the cycle. 4.3.4.1.4 Fuel Rod Diameter Reduction To maintain standardization of design and fabrication, the fuel rod diameter is reduced from 0.493 to 0.483 inches. The 10 mil reduction in fuel rod diameter was accomplished by reducing the pellet diameter by 6 mils and decreasing the cladding thickness by 2 mils. The revisions to the fuel rod design are shown in Table 4.3-4. The MLHGR is preserved at 13.4 kW/ft as a design basis; therefore, the maximum fuel centerline temperature at full power remains very nearly the same. Although the fuel time-constant is slightly decreased as a consequence of the reduction in fuel rod diameter, analyses of core transient response have indicated the MCPR, peak fuel temperature and system pressure will be maintained below limits. 4.3-26

GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. O 4.3.4.1.4 Fuel Rod Diameter Reduction (Continued) O A summary of fuel bundle design changes can be found in Table 4.3-4. The new fuel bundle design is also illustrated in Figure 4.3-2. 4.3.4.1.5 Prepressurized Fuel Rods Although the fuel has been prepressurized to 3 atm, no effect on the nuclear response occurs. 1

 ,        4.3.5. References

1. " General Electric Company Model for Loss-of-Coolant Accident Analysis In Accordance with 10CFR50" Appendix K, January 1976 (NEDO-20566).

2. E. C. Eckert, et al., " General Electric Thermal Analysis Basis (GETAB) : Data, Correlation, and Design Application",

January 1977 (NEDO-109 5 8A) . N < 3

3. J. A. Woolley, "3D BWR Core Simulator", May 1976 (NEDO-20953) .

4. J. F. Carew, " Process Computer Performance Evaluation Accuracy", June-1974 (NEDO-20340) .

5. C. L. Martin, " Lattice Physics Methods Verification",

August 1975 (NEDO-20939).

6. G. R. Parkos, "BWR Simulator Methods Verification", January 1977 (NEDO-20946A).

7. R. C. Stirn, " Generation of Void and Doppler Reactivity Feedback For Application to BWR Plant Transient Analysis:,

August 1975 (NEDO-20964 ) .

8. R. C. Stirn, et al., " Rod Drop Accident Analysis For Large

] Boiling Water Reactors", General Electric Co., Atomic Power Equipment Department, March 1972 (NEDO-10527). (Also Supplement 1, July 1972 and Supplement 2, January 1973.)

9. C. J. Paone, " Banked Position Withdrawal Sequence",

September 1976 (NEDO-21231).

10. K. W. Cook, " Qualification of the One-dimensional Core Transient Model for Boiling Water Reactors", October 1978 (NEDO-24154, Volume 1) .

4.3-27 i

l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i 4.3.5 References (Continued)

11. R. L. Crowther, "Xr.non Cc,nsiderations in Design of Boiling Water Reactors", June 1968 (APED-5640).

12. C. L. Martin, " Lattice Physics Methods", February 1977 (NEDO-20913A). .

i

13. General Electric Standard Safety Analysis Report (GESSAR) .

14. R. K. iia ling , Operating Strategy for Maintaining an Optimum Power Distribution Throughout Life, Paper Presented at ANS Topical Meeting, San Francisco, September 1963 (TID-7672).

l

\                                                                                                                                                                                         l l

O' I I i O l l i 4.3-28

GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. O Table 4.3-1 f REACTOR CORE DIMENSIONS 1 b Active Core lleight, ft (cm) 12.5 (381) 1 8x8 BWR/6-Lattice Control Rod Pitch, in. (cm) 12.0 (30.48) Control Rod Thickness, in. (cm) 0.328 (0.8331) i Fuel Assembly Cross Section, in.2 (cm2) 5.455x5.455 (13.86x13.86) Fuel Assembly Pitch, in. (cm) 6.0 (15.24) s Channel I.D., in. (cm) 5.215 (13.25) Channel Thickness, in. (cm) 0.120 (0.3048) Fuel Rod O.D., in. (cm) 0.483 (1.227) Fuel Rod Clad Thickness, in. (cm) 0.032 (0.0813) Fuel Pellet O.D., in. (cm) 0.410 (1.041) Number of Fuel Rods 62 Water Rod O.D., in. (cm) 0.591 (1.501) Water Rod Clad Thickness, in. (cm) 0.030 (0.0762) Number of Water Rods 2 Active Fuel Length, in. (cm) 150 (381) . l 2.540 cm = 1 in. I i 4.3-29 1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 Table 4.3-2 0 REACTIVITY DATA FOR THE COLD, XENON-FREE STATE BPWS Rod k eff Groups Condition Withdrawn  % Controlled BOC MOC EOC 100 0.927 0.917 0.903 1& 2 75 0.994 0.982 0.968 1, 2, 3& 4 50 1.032 1.019 1.000 All Rods Out 0 1.112 1.097 1.073 Highest Worth Rod Withdrawn 0.973 0.961 0.953 Highest Worth Rod Core Co-Ord. (18,51) (18,51) (22,55) BOC = Beginning of cycle MOC = Middle of Cycle EOC = End of Cycle BPWS = Bank Position Withdrawal Sequence l l l 1 0 4.3-30

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 t ) (/ Table 4.3-3 REACTIVITY AND CONTROL FRACTION FOR VARIOUS REACTOR STATES Beginning- Middle- End of-Cycle of-Cycle of-Cycle Condition eff CF eff CF eff CF Cold, no xenon, 0.994 0.75 1.004 0.61 1.000 0.50 critical *, zero power Hot, no xenon, 1.006 0.50 1.000 0.50 1.010 0.44 critical *, zero power Hot, no xenon, 3 100 0.24 0.996 0.25 0.995 0.11 critical *, rated power Hot, with xenon, 0.998 0.16 0.999 0.13 0.999 0.0 critical *, rated power 7.

 !,        i Cold, no xenon,     1.032     0.50    1.019    0.50 1.031   0.44 zero power llo t , no xenon,   1.079     0.24    1.078    0.25 1.062   0.11 zero power Hot, no xenon,      1.027     0.16    1.027    0.13 1.026   0.0 rated power Hot, with           1.036     0.0     1.032    0.0  0.999   0.0 xenon,

• Control rod patterns adjusted approximately to critical. The deviations from keff = 1.000 were allowed to minimize analysis effort. The Ak between conditions with the same control frac-tion remain valid.

[ (~J i 4.3-31 1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 4.3-4

SUMMARY

OF BWR/6 DESIGN REVISIONS (GE COMPANY PROPRIETARY) 1 1 O O 4.3-32

i GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 0 O Table 4.3-5 CALCULATED NEUTRON FLUXES (USED TO EVALUATE VESSEL IRRADIATION) Flux at the Average Flux Flux at the Inside Surface Neutron Energy In the Core Core Boundary Vessel (MeV) (n/cm 2 ec) (n/cm2 -sec) (n/cm2 -sec)

             >3.0                      1.4 x 10 13                      4.2 x 1012        1.1 x 10 9 1.0 - 3.0                       3.6 x 10 13                      9.5 x 10 12       9.5 x 10 5 0.1 - 1.0                       6.1 x 1013                       1.5 x 10 12       1.6 x 10 9 18    "

Maximum Fluence > 1.0 MeV at the vessel i.d. = 4.3 x 10 (2) cm Notes:

1. The calculated flux is a maximum in the axial direction but average over the azimuthal angle.

2. The maximum fluence is calculated using the flux and a

] capacity factor of 80's or 1 x 10' full power seconds. The i fluence includes an azimuthal peaking factor and a factor to cover analytical uncertainties. The azimuthal peaking factor is derived from the results of a two-dimensional transport calculation. The two-dimensional analysis models the reactor bundle pattern in an r, e geometry. Fluxes are calculated at the cylindrical core shroud surrounding the core. The peaking factor used was 1.4. In addition to the angular peaking factor, a safety factor of 2 was applied to ensure that the pre-dicted values are conservative. O 4.3-33 i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 4.3-6 CALCULATED NEUTRON FLUX AT CORE EQUIVALENT BOUNDARY Lower Energy Flux Grout' Bound (eV) (n/cm2-sec) 1 10.0 x 10' 3.6 x 10*' 2 6.065 x 10' 5.3 x 10** 3 3.679 x 10' 2.0 x 10** 4 2.231 x 10' 3.9 x 10** 5 12 5 1.353 x 10 4.6 x 10 s 12 6 8.208 x 10 4.1 x 10 12 7 4.979 x 10' 4.0 x 10 5 12 8 3.020 x 10 2.8 x 10 9 1.P32 x 10' 2.4 x 10** 12 10 6.738 x 10" 3.4 x 10 12 11 2.479 x 10' 2.3 x 10 12 12 9.119 x 10' 2.3 x 10 12 13 3.355 x 1G' 2.1 x 10 12 14 1.234 x 10' 2.1 x 10 2 12 15 4.540 x 10 2.0 x 10 12 16 1.670 x 10* 2.1 x 10 17 6.144 x 10' l.9 x 10** 12 18 2.260 x 10' l.9 x 10 19 1.371 x 10' 9.2 x 10** 12 20 8.315 9.2 x 10 21 5.043 8.4 x 10** 22 3.059 8.7 x 10** i1 23 1.255 , 8.6 x 10 24 1.125 8.5 x 10** 25 0.616 9.1 x 10 tt 26 0.000 3.2 x 1018 4.3-34

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 (3 b l NOTE: LOADING PATTERN IS SHOWN FOR OUARTER CORE ONLY. REFLECTIVE 3 3 3 3 3 3 1 SYMMETRY APPLIES TO REMAINDER - OF CORE. 3 3 F 1 F 1 F 1 2 3 1 F 2 F 2 F 2 1 3 3 1 F 2 F 2 F 2 F 4 2 3 1 F 2 1 2 1 2 1 2 1 5 l 6 3 1 F 2 F 2 F 3 F 2 F 3 I - 3 1 F 1 1 2 2 2 1 1 1 2 1 7 3 1 F 2 F 3 F 3 F 3 3 8 F 2 F 3 F 2 1 2 1 3 2 2 1 1 1 2 1 g h 3 F 2 F 2 F 2 F 2 F 3 F 3 F 3 10 V 3 1 F 1 3 3 1 1 3 1 3 2 l _2 1 1 11 3 F 2 F 2 F 2 F 2 F 12 2 F 3 F 3 3 1 F 2 1 2 1 2 1 3 1 3 3 1 1 13 3 F 2 F 2 F 2 F 2 F 2 F 2 F 3 3 1 2 3 3 3 3 15 1 1 1 1 1 1 3 2 F+ I I 4 I I I I I J 1 2 3 5 6 7 8 9 10 11 12 13 14 15 BUNDLE TYPE DESCRIPTION TYPE NUMBER F FRESH 188 1 ONE CYCLE 188 2 TWO CYCLE 188 3 THREE CYCLE 184 \ Figure 4.3-1. Equilibrium Core Loading Map 4.3-35

GESSAR II 22P.7007 238 NUCLEAR ISLAND Rev. 0

                                                                     - * - 0  4-f00000000hfCO_                                                                              Q.E)OOO 3 a_

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j SO@OO@OOg' 3 . _ _ _ __ CHANNEL F UEL ROD PELLET WATER ROD D i*.1 1 D . A 6 C D E F G H I DIMINCHES 0.120 5.215 0 380 0 03 0.483 0 419 0.410 0591 0 531 l CON OL HOD BUNDLE L ATTICE CELL dim ID J K L P.i N O P Q R S f DIM INLHE S 1.55 4 902 0 328 0.63G 0.153 0.140 0 099 0 2725 02725 12 00 O l l Figure 4.3-2. BWR/6 Lattice Nominal Dimensions, 120-mil Channel I 4.3-36

                                                                                    =                                       . . . .                                                                                                         _.

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(GE COMPANY PROPRIETARY) v  ! I I i

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 m O 4 n THIS FIGURE IS PROPRIETARY TO THE GENERAL ELECTRIC COMPANY AND WILL BE SUPPLIED LATER UNDER SEPARATE COVER. O l I l Figure 4.3-4. Axial Fuel Rod Enrichment and Gadolinia Distribution O (GE COMPANY PROPRIETARY) 4.3-38

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ow Figure 4.3-7. Weight Fraction as a Function of Exposure, Dominant Fuel Type, 40% Voids

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• O Figure 4.3-9. Neutron Generation Time as a Function of Exposure o at 40% Voids

o.008 UNCONTROLLED AT 40% VOIDS

                                                                                                      ===--  CONTROLLEO AT 40% VOIDS
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T l l O l 1.i s l i l 1.14 - I 1.13 - 1.12 - O ti 1 g 1.11 - w 2 g

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o4 Figure 4.3-11. Variation of Mai:imum Local Power Peaking as a Function of Exposure, Dominant Fuel Type, 40% Voids, Uncontrolled

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 O THIS FIGURE IS PROPRIETARY TO THE GENERAL ELECTRIC COMPANY AND WILL BE SUPPLIED LATER UNDER SEPARATE COVER. O Figure 4.3-12. Uncontrolled Local Power Distribution as a Function of Exposure at 40% Voids, Dominant Fuel Type (GE COMPANY PROPRIETARY) O 4.3-46

l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i 4 l i ! i l i 4 1 i i i ? I i 1 f i . THIS FIGURE IS PROPRIETARY TO THE GENERAL ELECTRIC COMPANY AND WILL SE SUPPLIED LATER UNDER SEPARATE i COVER. I i 1 i I ) i f l. I i i Figure 4.3-13. Uncontrolled Local Power Distribution as a Function ' of Voids at 0.0 Lattice Exposure, Dominant Fuel Type (GE COMPANY PROPRIETARY) , l \ - i l l 4.3-47 i l

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 0 THIS FIGURE IS PROPRIETARY TO THE GENERAL ELECTRIC COMPANY AND WILL BE SUPPLIED LATER UNOFR SEPARATE COVER. O Figure 4.3-14. Controlled Local Power Distribution at 40% Voids, 0.0 Lattice Exposure, Dominant Fuel Type (GE COMPANY PROPRIETARY) O 4.3-48

    . _ . - . . . - . . _ _ -        . . . . . - . - = _ .      . . - . - _    .

i 4 GESSAR II 22A7007 , 238 NUCLEAR ISLAND Rev. O i t ' 1 l 1 s J 1 ,

                                                                                                                                              ?

f n i i i i f f 1 i i i THIS FIGURE IS PROPRIETARY TO THE GENERAL ELECTRIC CotrANY AND WILL BE SUPPLIED LATER UNDER SEPARATE $ COVER. 4 l i t ) 1 1 1 e i e t . i i i t i Figure 4.?-15. Uncontrolled R-Factor Distribution at 40% Voids, ! 0.0 Lattice Exposure, Dominant Fuel Type (GE COMPANY PROPRIETARY) f i I f 1 i 4.3-49 , t

1.06 1.05 1.04 - w w C O Z b 2

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                                                                            ---        CALCULATED R-F ACTOR 1.01                                                                                                       N O             5          10               15           20             25                30      35 xw o>

BUNDLE AVERAGE EXPOSURE (GVW1/ST) <W

• o o

Figure 4.3-16. Variation of the Maximum Uncontrolled Bundle-Integrated 04 R-Factor as a Function of Bundle Average Exposure i O O O

       .      . _ _ . _ _ _ . _ _ __ ~                 ~ . , ._.             . . - -          -   _-..~_s__              __ _ _ _ _ _ . _ - .   -    _m . _    . _ _ . _ . ___ .

O BEGINNING-OF-EQUILIBR1UM-CYCLE Integrated Power Per Chanael I 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 J 0.2773 0.3593 0.3936 0.4279 0.4487 0.4742 1 0.2773 0.3593 0.3936 0.4279' O 4487 0.4742 2 0.3165 0.4174 0.5409 0.7619 0.7514 0.8611 0.8361 0.9602 3 0.4037 0.7108 0.7540 0.6217 0.6677 0.9184 0.9690 1.0074 1.1617 4 0.4297 0.7623 0.8519 0.9469 0.9942 1.0374 1.0573 -1.0861 1.1093 1.1507 5 0.4412 0.8108 0.8957 0.9948 1.1716 1.1550 1.2362 1.1668 1.2152 1.1403 1.2870 6 0.4307 0.6145 0.9179 1.0257 1.1036 1.1683 1.1591 1.0727 1.1382 1.1224 1.1166- 1.0984 7 0.4038 0.7846 0.9043 1.1134 1.2167 1.2202 1.3141 1.2058 1.1036 1.0365 1.2666 1.1694 1.0236 8 0.3158 0.7096 0.6502 0.9950 1.0919 1.0769 1.1637 1.1369 1.1880 0.9675' l.0299 1.1793 1.1526 0.9003 9 0.4152 0.7506 0.9419 1.1626 1.1497 .l.2893 1.1297 1.3524 1.2611 1.3332 1.2302 1.3411 1.2136 1.2520. 10 0.2766 0.5362 0.8175 0.9874 1.1331 1.1473 1.1925 1.1801 1.2326 1.2390 1.1601 1.2144 1.1388 1.1671 1.0639 bJ 11 0.3580 0.7580 0.8840 1.0339 1.2276 1.0666 1.0983 0.9683 1.3336 1.1648 1.3683 1.1461 1.2992 1.1464 0.9209 LJ CD 12 0.3933 0.7486 0.9160 1.0532 1.1570 1.1345 1.0376 1.0325 1.2386 1.2348 1.2632 1.2044 1.1102 1.1081 0.7938 13 0.4266 0.6561 0.9658 1.0837 1.2124 1.1218 1.2871 1.1848 1.3514 1.1555 1.3408 1.1220 1.2827 1.0793 1.2111  :: 14 0.4457 0.8336 1.0038 1.1073 1.1408 1.1168 1.1712 1.1566 1.2210 1.1792 1.1682 1.1260 1.1769 1.1457 1.0756 c2 c) 15 0.4736 0.9576 1.1608 1.1497 1.2870 1.0986 1.0245 0.9033 1.2611 1.0709 0.9246 0.8046 1.2195 1.0773 1.1788 C) b3 t* U1

 **                                                                                                                                                                                b1 U1
 .                                                                                                                                                                                 3s >s CND-OF-EQUILIBRIUM-CYCLE (n                                                                                                                                                                                F4 F4 Fs                                                                                                                                                                                U1 F4 Integrated Power                                                                                    D Per Channel                                                                                   !!     ,

C3 , I 1 2 3 4 5 6 7 8 9 10 11 12 - 13 14 15 J 1 0.3067 0.3821 0.4182 0.4312 0.4374 0.4330 2 0.3333 0.4560 0.6639 0.6015 0.9118 0.8657 0.9357 0.8564 3 0.4061 0.7244 0.9055 0.6544 1.0563 0.9402 1.1081 0.9497' l.0101 " 4 0.4229 0.7725 0.9803 0.9236 1.1280 1.0118 1.1947 1.0450 1.1989 1.0112

• 5 0.4316 0.7901 1.0103 0.9479 1.0613 1.0682 1.1358 1.0999 1.1723 1.0952 1.1347 6 0.4244 0.7935 1.0278 0.9659 1.1824 1.0594 1.2483 1.0299 1.2810 1.1104 1.2769 1.0337 7 0.4069 0.7756 1.0197 1.0418 1.0798 1.0881 1.1550 1.1170 1.1916 1.1393 1.2226 1.1311 1.1665 8 0.3334 0.7247 0.9807 0.9504 1.1760 0.9792 1.2351 1.0276 1.2604 1.0538 1.3059 1.1268 1.2950 1.0341 9 0.4550 0.9041 0.9215 1.0565 1.0481 1.1393 1.0229 1.1808 1.1306 1.2074 1.1419 1.2229 1.1308 1.1634 10 0.3069 0.6626 0.8527 1.1240 1.0516 1.2403 1.1082 1.2741~ 1.1093 1.2909 1.0526' l.2884 1.0564 1.2781 1.0247 11 0.3819 0.8004 1.0550 1.0105 1.1309 1.0257 1.1848 1.0512 1.2058 1.0543 1.2042 1.0482 1.2159 .l.1072 1.1419 12 0.4190 0.9113 0.9399 1.1926 1.0929 1.2779 1.1386 1.3055 1.1443 1.2965 1.1304 1.2758 1.0372 1.2434 0.9872 13 0.4312 0.8643 1.1072 1.0439 1.1695 1.1081 1.2218 1.1281 - 1.2245 1.0604 1.2047 1.0401 1.1916 1.0024 1.0970 hJ 14 0.4373 0.9351 0.9483 1.1979 1.0949 1.2762 1.1312 1.2957 1.1323 1.i816 1.1134 1.2547 1.0783 1.1969 0.9397 DO h)
• 15 0.4328 0.8559 1.0100 1.0109 1.1346 1.0337 1.1666 1.0347 1.1672 1.0286 1.1451 0.9970 1.1051 0.9410 0.9744 {}

. C3 i CD i

C) %J ' ! Figure 4.3-17. Radia1 Power Factors for Beginning-of-Equilibrium-i Cycle and Optimal End-of-Equilibrium Cycle Conditions

2.6 2.4 - B E GINNIN G-OF-E QUILIBRIUM-CYCLE 2.2 - - - OPTIMAL END-OF-EQUILtBRIUM{YCLE (HALING) 2.0 - 1.8 - 1.6 - N 5 m c 2 a 1.4 -

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o.4 0.2 - I I I I I I I I I I I l N o.o 23 25 27 xM 3 5 7 9 11 13 15 17 19 21 1 o> AXIAL NODE <$O Figure 4.3-18. Beginning-of-Equilibrium-Cycle and Optimal End-of-Equilibrium-Cycle Core Average Axial Power O O O

l O i ! am l 0.00 - I i

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DESIGN LIMIT

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PERCENT VOIDS *O O OY Figure 4.3-19. Equilibrium Cycle Void Coefficient for Stability Analysis as a Function of Percent Voids l

0.0 N N N N

                                                                                                                 \                                                       CALCULATION

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                                    -20.0 0                                                                       20 w

10 30 40 50 60 70 NN PERCENT VOIDS o>

                                                                                                                                                                                          <4

• O O

Figure 4.3-20. Dynamic Void Reactivity Coefficient as a Function " of Percent Voids at End-of-Equilibrium-Cycle O O O

O

                                -0.30 DESIGN LIMIT              -
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                                                /                                                   NOTE: ALL NUMBERS INCLUDE THE DOPPLER WEIGHTING FACTOR -

OF 1.32. 1 i CALCULATED CURVE

                                                                                                     --- CALCULATED CURVE *0.95
                              -0.80                            '                                                          I                             I O                                                                                                                                                    N 500                        1000                            1500                          2000    2500                    WM o>'J

! AVER AGE FUEL TEMPERATURE PC) < .)

• O, Figure 4.3-21. Doppler Coefficient as a Function of Average Fuel Temperature at End-of-Equilibrium-Cycle 1

1.01 1% DESIGN LIMIT _ ""- CALCULATED (WITH BI AS APPLIED) w W co

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N Osa - -- -%  % 1 I I I Os7 I I I w 0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 gM l CYCLE EXPOSURE (GWd/ST) <

                                                                                                                                                                ,  g o

ow Figure 4.3-22. Cold Shutdown Reactivity as a Function of Cycle Exposure Strongest Rod Withdrawn, No Xenon , 9 O O

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 V X -+ 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 I I I I I i i l l l

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                                                                                                                                         ~~

3, 4 4 3 4 3 19 - p - 6 3 3 3 6 7 15 - - ~ l 5 2 6 't I2 2 6 I S 11 - - 1 d- 3 5 4 2 3 2 4 2 3 5 7 - - - 25 2__ _ h-h i, 2 2 2, i, f-l-h--- 27 l-h4-4---a 3__ 3 4 3 7 l _1_ lf _Ill l l j. 6 5 6 g g g Y l l l l l l l l l l l l l l l 1 I I I I I I l l I I l l l l l~ 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 X,Y = PLANT COORDINATES I, J = MODEL COORDINATES Figure 4.3-23. Banked-Position Withdrawal Sequence, RPCS (Groups 1-4, Sequence A) U 4.3-57

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 0 X~ 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 I I I I i i l i l l l 1 1 I l 3 l l 1 1 I I I I .I I I I l l t 59 _ _ _ L _I_ _l _ I I I I J

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l I , , __ _ J _ 1I - 5 2 8 i 7 2 7 i 8 2 5 2 E T ~~ - - 7'

 ,__d_'_K                            e,                e,                 10,                 e,               e,     f -l -- y - ---- 27 I       I        i       11 5'                 5                          l - H - - 29 l        T        F l            I                                                                  I         I       I         I Y         l       l        l       l      l          l          l         l        l        l        l       l         l       l         l l       l        l       l      l          l          l         l        l        I        I       I         I       I         I la 1        3        5       7      9         11         13        15       17       19       21      23        25      27       29 X, Y = PLANT COORDINATES l, J = MODEL COORDIN ATES i Figure 4.3-24.                          Banked-Position withdrawal Sequence, RPCS (Groups 5-10, Sequence A) 4.3-58

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev, 0

 / N (v/

X- 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 l l I l l I I I I i l l 1 I l  ; I l l l l I l l l l l l l l t _H - t - t- -l- - - '

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                                                                                                                     /         ;       ;

I - l-- H - - 29 I l 11 ,* 1 I f- l 1: I I I I Y l l l l l l l l l l l l l l l l l l 1 I I I  ! I I I I I l i l~ 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 X, Y = PLANT COORDINATES , 1, J = MODEL COORDINATES Figure 4.3-25. Banked-Position Withdrawal Sequence, RPCS (Groups 1-4, Sequence B)

 'O 4.3-59

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 0 X~ 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58

             !      I      l     l       l         l         l       l      l       l       l       l         l       l         [

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Figure 4.3-26. Banked-Position Withdrawal Sequence, RPCS (Groups 5-10, Sequence B) 4.3-60

;          O                                                                     O                                              O j           6.0 l

DESIGN LIMIT SCRAM CURVE 'o.95 5.0 - -- SCRAM CURVE

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M I 1 1 1 I wM o o,io o.2o o.3o 0.40 0.50 0.60 CONTROL FR ACTION ow Figure 4.3-27. Ilot Operating, End-of-Equilibrium-Cycle Scram Reactivity (S), as a Function of Control Fraction

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O 1 REACTOR CORE 4 WATER WA ER 3 SHROUD 5 VESSEL MATERI AL HADIUS VOLUME AVER AGE NO. NAME Onches) MATERIAL DENSITY 1 RE ACTOR CORE 92.58 WATER 0.318 9/cm3 UO2 2.334 g/cm3 304L STAINLESS STEE L 0.056 g/cm3 ZlRCONIUM 0 978 g/cm3 2 WATER 99 9 WATER 0.74 g/cm3 3 SHROUD 101.9 304L STAINLESS STE E L FROM ASME SA 240 4 WATER 119.0 WATER 0.74 g/cm3 5 VE SSE L 125 0 C AR BO N STE E L FROM ASME SA 533 6 AIR AIR 1.3 x 10-3 g/cc Figure 4.3-28. flodel for One-Dimensional Transport Analysis of Vessel Fluence 4.3-62

e i , 1.4 , 1.3 - ) i 12 l 1 1.1 - I

!         1.0  -

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       .J A

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   .   <  0.7  -

OM w E V in I w m M to 2 >> W

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   <h                                                                                          >

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0.3 -

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I 0.1 - ! O I I  ! I  !  ! I ' N O 10 20 30 40 50 60 70 80 90 100 [pN PERCENT OF RADIUS ,4 y o OM Fi;ure 4.3-29. Radial Power Distributions Used in the Vessel I Fluence Calculation i e i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

      )
  \~ /                                                        SECTION 4.4 CONTENTS Section                                                 Title                        Page 4.4   THERMAL-HYDRAULIC DESIGN                                                       4.4-1
              ' 4.4.1                             Design Basis                                 4.1-1 4.4.1.1                          Safety Design Bases                          4.4-1 4.4.1.2                          Powcr Generation Design Bases                4.4-1 4.4.1.3                          Requirements for Steady-State Conditions                                   4.4-2 4.4.1.4                          Requirements for Transient Conditions                                   4.4-2 4.4.1.5                          Summary of Design Bases                      4.4-3 4.4.2                            Description of Thermal-Hydraulic
                                              ,. Design of the Reactor Core                    4.4-4 4.4.2.1                          Summary Ccmparison                           4.4-4 4.4.2.2                          Critical Power Ratio                         4.4-4 4.4.2.2.1                        Boiling Correlations                         4.4-5
      )          4.4.2.3                          Linear Heat Generation Rate (LHGR)           4.4-6 4.4.2.3.1                        Design Power Distribution                    4.4-7 4.4.2.3.2                        Design Linear Heat Generation Rates (LHGR)                                 4.4-7 4.4.2.4                          Void Fraction Distribution                   4.4-8 4.4.2.5                          Core Coolant Flow Distribution and Orificing Pattern                        4.4-8 4.4.2.6                          Core Pressure Drop and Hydraulic Loads                                        4.4-9 4.4.2.6.1                        Friction Pressure Drop                       4.4-10 4.4.2.6.2                        Local Pressure Drop                          4.4-11 4.4.2.6.3                        Elevation Pressure Drop                      4.4-12 i

4.4.2.6.4 Acceleration Pressure Drop 4.4-13 4.4.2.7 Correlation and Physical Data 4.4-14 4.4.2.7.1 Pressure Drop Correlations 4.4-14 4.4.2.7.2 Void Fraction Correlation 4.4-15 4.4.2.7.3 Heat Transfer Correlation 4.4-15 4.4.2.8 Thermal Effects of Operational I

• 4.4-15 Transients

>                4.4.2.9                          Uncertainties in Estimates                   4.4-15 4.4-i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CC'! TENTS (Contittued) Section Title Page 4.4.2.10 Flux Tilt Considerations 4.4-16 4.4.3 Description of the Thermal and Hydraulic Desion of the Reactor Coolant System 4.4-16 4.4.3.1 Plant Configuration Data - 4.4-16 4.4.3.1.1 Reactor Coolant System Configuration 4.4-16 4.4.3.1.2 Reactor Coolant System Thermal Hydraulic Data 4 . 4 <-16 4.4.3.1.3 Reactor Coolant System Geometric Data 4.4-16 4.4.3.2 Operating iluatrictions on Fumps 4.4-17 4.4.3.3 Power-Flow operating Map 4.4-17 4.4.3.3.1 Limits for Normal Operation 4.4-17 4.4.3.3.1.1 Performance Characteristics 4.4-18 4.4.3.3.2 Regions of the Power Flow Map 4.4-18 g 4.4.3.3.3 Derign Features for Power-Flow Control 4.4-19. 4.4.3.3.3.1 Flow Control 4.4-21 4.4.3.4 Temperature-Power Operating Map (PWR) 4.4-22 4.4.3.6 Load-Following Characteristics 4.4-22 4.4.3.5 Thermal and Hydraulic Characteristics Summary Table 4.4-22 4.4.4 Evaluation 4.4-23 4.4.4.1 Critical Power 4.4-23 4.4.4.2 Core Hydraulics 4.4-23 4.4.4.3 Influence of Power Distributions 4.4-23 4.4.4.4 Core thermal Response , 4.4-23 4.4.4.5 Analytical Methods 4.4-24 4.4.4.5.1 Reactoi Model 4.4-24 4.4.4.5.2 System Flow Balances 4.4-26 4.4.4.5.3 System Heat Balances 4.4-27 4.4.4.6 Thermal-Hydraulic Stability Analysis 4.4,29 4.4-11

GESSAR II 22A7007 238 NUCLEAR ISLAND R:v. 0 () CONTENTS (Continued) Section Title Page 4.4.4.6.1 Introduction 4.4-29 4.4.4.6.2 Description 4.4-30 4.4.4.6.3 Stability Criteria 4.4-31 4.4.4.6.4 Mathematical Model 4.4-31 4.4.4.6.5 Analytical Confirmation 4.4-33 4.4.4.6.6 Analysis Results 4.4-33 4.4.4.6.6.1 Impact of Prepressurized Fuel on Stability 4.4-35 4.4.5 Testing and Verification 4.4-36 4.4.6 Instrumentation Requirements 4.4-36 4.4.6.1 Loose Parts 4.4-37 4.4.7 References 4.4-37 Os V i .- 4.4-iii/4.4-iv

GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. O O SECTION 4.4 TABLES Table Title Page

   / ;-1  Thermal and Ilydraulic Design Characteristics of the Reactor Core                            4.4-41 4.4-2  Axial Power Distribution Used to Calculate MCPR Operating Limit                           4.4-44 4.4-3  Void Distribution                              4.4-45 4.4-4  Axial Power Distribution Used to Generate Void and Quality Distributions                 4.4-46 4.4-5  Plow Quality Distribution                      4.4-47 4.4-6  Core Flow Distribution                         4.4-48 4.4-7  Calculated Vs Measured Core Plat Pressure Drops                                          4,4-49 4.4-8  Typical Range of Test Data                     4.4-50 4.4-9  Description of Uncertainties                   4.4-51 4.4-10 Bypass Flow Paths                              4.4-53 4.4-11 Reactor Coolant System Geometric Data          4.4-54 4.4-12 Lengths of Safety Injection Lines              4.4-55 ILLUSTRATIONS Figure                         Title                  Page 4.4-1  Schematic of Reactor Assembly Showing the Bypass Flow Paths                              4.4-57 4.4-2  Damping Coefficient versus Decay Ratio

. (Second Order Systems) 4.4-58 l 4.4-3 Ilydrodynamic and Core Stability Model 4.4-59 4.4-4 Comparison of Test Results with Reactor Core Analysis 4.4-60 4.4-5 Power-Flow Operating Map 4.4-61 4.4-6 Total Core Stability 4.4-62 I 4.4-7a 10 Psi Pressure Regulator Setpoint Step at 51.5% Rated Power (Natural Circulation) 4.4-63 4.4-7b 10-Cent Rod Reactivity Step at 51.5% Rated Power (Natural Circulation) 4.4-64 4.4-v i

' GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O ILLUSTRATIO';S (Continued) I Figure Title page 4.4-7c 6-Inch Water Level Setpoint Step at 51.5% Rated Power (Natural Circulation) 4.4-65 4.4-8 Relative Bundle Power Ilistogram for Power Distribution Used in Statistical Analysis (Basis in Reference 1) 4.4-66 i i 1

O O

4.4-vi

GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 0 4.4 TIIERMAL-l!YDRAULIC DESIGN 4.4.1 Design Basis 4.4.1.1 Safety Design Bases Thermal-hydraulic design of the core shall establish: (1) Actuation limits for the devices of.the nuclear safety systems such that no fuel damage occurs as a result of moderate frequency transient events. Specifically, the Minimum Critical Power Ratio (MCPR) operating limit is specified such that at least 99.9% of the fuel rods in the core are not expected to experience boiling transi-tion during the most severe moderate frequency transient events. () (2) The thermal-hydraulic safety limits for use in evalua-ting the safety margin relating the consequences of fuel barrier failure to public safety. (3) That the nuclear system exhibits no inherent tendency toward divergent or limit cycle oscillations which would compromise the integrity of the fuel or nuclear system process barrier. 4.4.1.2 Power Generation Design Bases The thermal-hydraulic design of the core shall provide the following operational characteristics: (1) The ability to achieve rated core power output through-out the design life of the fuel without sustaining premature fuel failure. . 4.4-1 i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.1.2 Power Generation Design Bases (Continued) (2) Flexibility to adjust core output over the range of plant load and load maneuvering requirements in a stable, pre-dictable manner without sustaining fuel damage. 4.4.1.3 Requirements for Steady-State Conditions Steady-State Limits For purposes of maintaining adequate thermal margin during normal steady-state operation, the MCPR must not be less than the required MCPR operating limit, and the MLHGR must be maintained below the design LHGR for the plant. This does not specify the operating power nor does it specify peaking factors. These parameters are determined subject to a number of constraints including the thermal limits given previonsly. The core and fuel design basis for steady-state operation (i.e., MCPR and LHGR limits) have been defined to provide margin between the steady-state operating conditions and any fuel damage condition to accommodate uncer-tainties and to assure that no fuel damage results even during the worst anticipated transient condition at any time in life. The design steady-state MCPR operating limit and the peak LHGR is given in Table 4.4-1. 4.4.1.4 Requirements for Transient Conditions Transient Limits The transient thermal limits are established such that no fuel damage is expected to occur during the most severe moderate fre-quency transient event. Fuel damage is defined as perforation of 9 4.4-2

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.1.4 Requirements for Transient Conditions (Continued) the cladding that permits release of fission products. Mechanisms that cause fuel damage in reactor transients are: (1) severe overheating of fuel cladding caused by inadequate cooling, and (2) fracture of the fuel cladding caused by relative expansion of the uranium-dioxide pellet inside the fuel cladding. For design purposes, the transient limit requirement in met if at least 99.9% of the fuel rods in the core do not experience boiling transition during any moderate frequency transient event. No fuel damage would be expected to occur even if a fuel rod actually exrerienced a boiling transition. t () A value of 1% plastic strain of Zircaloy cladding is conservatively defined as the limit below which fuel damage from overstraining the fuel cladding is not expected to occur. The LilGR required to cause this amount of cladding strain is approximately 25 kW/ft ! in unirradiated UO2 fuel, but decreases with burnup to approxi-mately 20 kW/ft for UO 2 at a local exposure of 40,000 mwd /t. Summary of Design Bases 1 4.4.1.5 i In summary, the steady-state operating limits have been established to assure that the design basis is satisfied for the most severe i l moderate frequency transient event. There is no steady-state design j overpower basis. An overpower which occurs during an incident of a moderate frequency trantient event must meet the plant transient MCPR limit. Demonstration that the transient limits are not exceeded is sufficient to conclude that the design basis is satisfied. lO 4.4-3

GESSAR II 22A7007 23U NUCLEAR ISLAND Rev. 0 4.4.1.5 Summary of Design Bases (Continued) The MCPR and peak LHGR limits are sufficiently general so that no other limits need to be stated. For example, cladding surface temperatures will always be maintained within 10 to 15 F of the coolant temperature as long as the boiling process is in the nucleate regime. The cladding and fuel bundle integrity criterion is assured as long as MCPR and LHGR limits are met. There are no additional design criteria on coolant void fraction, core coolant flow-velocities, or flow distribution, nor are they needed. The coolant flow velocities and void fraction become constraints upon the mechanical and physics design of reactor components and are partially constrained by stability and control requirements. 4.4.2 Description of Thermal-Hydraulic Design of the Reactor Core 4.4.2.1 Summary Comparison An evaluation of plant performance from a thermal and hydraulic standpoint is provided in Subsection 4.4.3. A tabulation of thermal and hydraulic parameters of the core is given in Table 4.4-1, which cive comparison of this reactor with others of similar design. 4.4.2.2 Critical Power Ratio There are three different types of boiling heat transfer to water in a forced convection system: nucleate boiling, transition boiling, and film boiling. Nucleate boiling, at lower heat transfer rates, is an extremely ef ficient mode of heat transfer, allowing large quantities of heat to be transferred with a very small temperature rise at the heated wall. As heat transfer rate 4.4-4

_. - - _ _ . ~ .. _ . - _ . _ . -- . . _ _ - _ - . _ - _ - . _ - 1 1 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O

 !                 l      4.4.2.2                 Critical Power Ratio (Continued) 1
                 /

i j is increased, the boiling heat transfer surface alternates between film and nucleate boiling, leading to fluctuations in heated wall temperatures. The point of departure from the nucleate boiling 1 region into the transition boiling region is called the boiling transition. Transition boiling begins at the critical power and is characterized by fluctuations in cladding surface temperature.

!                         Film boiling occurs at the highest heat transfer rates; it begins as transition boiling comes to an end.                                                                          Film boiling heat transfer is characterized by stable wall temperatures which are higher than i

those experienced during nucleate boiling. . I 4.4.2.2.1 Boiling Corrclations 4 The occurrence of boiling transition is a function of the local steam quality, boiling length, mass flow rate, pressure, flow I geometry, and local peaking pattern. General Electric has con- , ducted extensive experimental investigations of these parameters. These parametric studies encompass the entire design range of these

variables. In the experimental investigations, a boiling transi-tion event was associated with a 25*F rise in rod surface

• tempera- ,

                                                                                                                                                                                                   ?

ture. The (critical) quality at which boiling transition occurs as a function of the distance from the equilibrium boiling boundary i is predicted by the GEXL [ General Electric Critical Quality,

X sub(c) -Boiling Length] Correlation. This correlation is based f on accurate test data of full-scale prototypic simulations of reactor fuel assemblies operating under conditions duplicating l

those in actual reactor designs. The GEXL correlation is a "best f fit" to the data and is used together with a statistical analysis l to assure adequate reactor thermal margins (Reference 1). The figure of merit used for reactor design and operation is the l Critical Power Ratio (CPR). This is defined as the ratio of the () bundle power which would produce equilibrium quality equal to, but 4.4-5 l

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.2.2.1 Boiling Correlations (Continued) not exceeding, the correlation value (critical quality) , to the bundle power at the reactor condition of interest (i.e., the ratio of critical bundle power to operating bundle power). In this definition, the critical power is determined at the same mass flux, inlet temperature, and pressure which exist at the speci-fied reactor condition. 4.4.2.3 Linear Heat Generation Rate (LHGR) The limiting constraints in the design of the reactor core are stated in terms of the LHGR limit and the MCPR operating limit for the plant. The design philosophy used to assure that these limits are met involves the selection of one or more distributions which are more limiting than expected operating conditions and subsequent verification that under these more stringent conditions, the design limits are met. Therefore, the " design puwer distribu-tion" is an extreme condition of power. It is a fair and stringent test of the operability of the reactor as designed to comply with the foregoing limits. Expected operating conditions are less severe than those represented by a design power distribution which gives the maximum allowable LHGR and the MCPR operating limits for the plant. However, it must be established that operation with a less severe power distribution is not a necessary condition for the safety of the reactor. Because there are an infinite number of operating reactor states which can exist (with variations in rod patterns, time in cycle, power level, distribution, flow, etc.) which are within the design constraints, it is not possible to determine them all. However, constant monitoring of op rating conditions using the availble plant measurements can ensure com-pliance with design objectives. The core average and MLHGR are given in Table 4.4-1. 4.4-6

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 /) 4.4.2.3.1 Design Power Distribution Thermal design of the reactor - including the selection of the core size and effective heat transfer area, the design steam quality, the total recirculation flow, the inlet subcooling and the speci-fication of internal flow distribution - is based on the concept and application of a design power distribution. The design power distribution is an appropriately conservative representation of the most limiting therma) operating state at rated conditions and includes design allowances for the combined effects (on the fuel rod, and the fuel assembly heat flux and temperature) of the gross and local steady-state power density distributions and adjustments of the control rods. The design power distribution is used in conjunction with flow and pressure drop distribution computations to determine the thermal conditions of the fuel and the enthalpy conditions of the coolant

~'h

[d throughout the core. The design axial power distribution used in the calculation of the MCPR operating limit is given in Table 4.4-2. This distribution is consistent with that discussed in Reference 1. The design power distribution is based on detailed calculations of tne neutron flux distribution as discussed in Appendix 4A. 4.4.2.3.2 Design Linear 2: cat Generation Rates (LIIGR) The maximum and core average linear heat generation rates are shown in Table 4.4-1. The maximum linear heat generation at any location is the product of the average linear heat generation rate at that location and the total peaking factor. O V 4.4-7

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.2.4 Void Fraction Distribution h The core average and maximum exit void fractions in the core at rated condition are given in Table 4.4-1. The axial distribution of core void fractions for the average radial channel and the maximum radial channel (end of node value) for the core are given in Table 4.4-3. The core average and maximum exit value is also provided. Similar distributions.for steam quality are provided in Table 4.4-5. The core average axial power distribution used to produce these tables is given in Table 4.4-4. 4.4.2.5 Core Coolant Flow Distribution and Orificing Pattern Correct distribution of core coolant flow among the fuel assemblies is accomplished by the use of an accurately calibrated fixed orifice at the inlet of each fuel assembly. The orifices are located in the fuel support piece. They serve to control the flow distribu-tion and, hence, the coolant conditions within prescribed bounds throughout the design range of core operation. The sizing and design of the orifices ensure stable flow in each fuel assembly during all phases of operation at normal operating conditions. The core is divided into two orificed flow zones. The outer zone is a narrow, reduced-power region around the periphery of the core. The inner. zone consists of the core center region. No other con-trol of flow and steam distribution, other than that incidentally supplied by adjusting the power distribution with the control rods, is used or needed. The orifices can be changed during refueling, if necessary. Design core flow distribution calculations are made using the design power distribution which consists of a hot and average powered assembly in each of the two orifice zones. The design bundle power and resulting relative flow distribution are given in Table 4.4-6. O 4.4-8

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O o

4.4.2.5 Core Coolant Flow Distribution and Orificing Pattern ,

, (Continued) 'l The flow distribution to the fuel assemblies is calculated on the i assumption that the pressure drop-across all fuel assemblies is , 1 the same. This assumption has been confirmed by measuring the flow distribution in a modern boiling water reactor as reported in' Reference 2. There is reasonable assurance, therefore, that the calculated flow distribution throughout the core is in close agreement with the actual flow distribution of an operating reactor. t The use of the design power distribution discussed previously ensures the orificing chosen covers the range of normal opera- , tion. The expected shifts in power production during core life are less severe and are bounded by the design power distribution. P 4.4.2.6 Core Pressure Drop and Hydraulic Loads The pressure drop across various core components under the steady-state design conditions is included in Table 4.4-1. Analyses for the most limiting conditions, the recirculation line break and the steam line break, are reported in Chapter 3. The components of bundle pressure drop considered are friction, , local elevation and acceleration. Core plate pressure drop measurements have been taken on several operating BWR/3 and 4 plants containing 7x7, 8x8, and mixtures of 7x7 and 8x8 fuel. Table 4.4-7 compares measured and calculated core plate pressure drops. The measured and calculated values are in good agreement. The data is predicted with an average error of 0.04 psi. The one sigma error is 0.86 psi. i 1

)                                                        4.4-9

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.2.6 Core Pressure Drop and Hydraulic Loads (Continued) The thermal-hydraulic loads on the fuel rods during the steady-state operation, transient and accident conditions are negligible, primarily because of the channel confinement, thereby resulting in small cross flow between rods (i.e., essentially constant pressure at any given elevation in the fuel bundle). The loads (i.e., horizontal) across the control blades are minimal or negligible primarily due to the flat interchannel velocity profile as given in Reference 3. 4.4.2.6.1 Friction Pressure Drop Friction pressure drop is calculated using the model relation:

                         'l 2     gg      ,
               ^

f ~2g o D g A2 ch where aP = friction pressure drop (psi); 7 W = mass flow rate; g = Newton constant relating force and mass; c p = water density; Dy = channel hydraulic liameter; A = channel flow area; h I = length; f = friction factor; and 4 2 pg

                  =   two-phase friction multiplier.

O 4.4-10

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O 4.4.2.6.1 Friction Pressure Drop (Continued) This basic model is similar to that used throughout the nuclear power industry. The formulation for the two-phase multiplier is based on data which compare closely to that found in the open literature (Reference 4). General Electric Company has taken significant amounts of friction pressure drop data in multirod geometries representative of modern BWR plant fuel bundles and correlated both the friction factor and two phase multipliers on a best-fit basis using the above pressure drop formulation. Checks against more recent data are being made on a continuing basis to ensure the best models are used over the full range of interest to boiling water reactors. 4.4.2.6.2 Local Pressure Drop

  ) The local pressure drop is defined as the irreversible pressure loss associated with an area change such as the orifice, lower tieplates, and spacers of a fuel assembly.

The general local pressure drop model is similar to the friction pressure drop and is: W2 K 2 L 2g P c where AP g = local pressure drop (psi); K = local pressure drop loss coefficient; A = reference area for local loss coefficient; and 2 = 4 p two-phase local multiplier O 4.4-11 L

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.2.6.2 Local Pressure Drop (Continued) and w, g and o are defined the same as for friction. This basic model is similar to that used throughout the nuclear power industry. The formulation for the two-phase multiplier is similar to that reported in the open literature (Reference 5) with the addition of empirical constants to adjust the results to fit data taken by General Electric Company for the specific designs of the BWR fuel assembly. Tests are performed in single-phase water to calibrate the orifice in the lower tieplate, and in both single- and two-phase flow to arrive at best-fit design values for spacer and upper tie-plate pressure drop. The range of test variables is specified to include the range of interest to boiling water reactors. New data are taken whenever there is a significant design change, to ensure the most applicable methods are in use at all times. 4.4.2.6.3 Elevation Pressure Drop The elevation pressure drop is based on the well known relation: 9 AP 9 E

                   =

[pf(1-a) + p al 9 A L g where AP U E'U E E I E AL = incremental length; F = average water density; a = average void fraction over the length AL; pg, p = saturated water and vapor density, respectively; and g = acceleration of gravity. O 4.4-12

i GESSAR II 22A7007

238 NUCLEAR ISLAND Rev. O i

4.4.2.6.4 Acceleration Pressure Drop A reversible pressure change occurs when an area change is encountered, and an irreversible loss occurs when the fluid is accelerated through the boiling process. The basic formulation for the reversible pressure change resulting from a flow area change is given by: i (l~#) " ^

                                                     =

AP ACC 2gpA22' 4=A t where 1 AP = acceleration pressure drop; ACC A2 = final flow area; and A = initial flow area and other terms are as previously defined. The basic formulation for the acceleration pressure change due to density change is: g 2 l- y-A P 9^ch _ P

                                                                                     .M   OUT    ,M.IN where 4

I 1

                                        =

x2 , (1-x)2 , P M P g" Il-")P f p g

                                       =    momentum density; and x    =    steam quality 4

! and other terms are as previously defined. The total acceleration pressure drop in BWRs is on the order of a few percent of the total pressure drop.

c;)

4 4.4-13 y- - , . - , , , .,,-r

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.2.7 Correlation and Physical Data General Electric has obtained substantial amounts of physical data in support of the pressure drop and thermal-hydraulic loads dis-cussed in Subsection 4.4.2.6. Correlations have been developed to fit these data to the formulations discussed. 4.4.2.7.1 Pressure Drop Correlations General Electric Company has taken significant amounts of friction pressure drop data in multirod geometries representative of modern BWR plant fuel bundles and correlated both the friction factor and two-phase multipliers on a test-fit basis using the pressure drop formulations reported in Subsections 4.4.2.6.1 and 4.4.2.6.2. Checks against more recent data are being made on a continuous basis to ensure the best models are used over the full range of interest to BWRs. Tests are performed in single-phase water to calibrate the orifice O and the lower tieplate, and in both single- and two phase flow to arrive at best-fit design values for spacer and upper tieplate pressure drop. The range of test variables is specified to include the range of interest to boiling water reactors. New data are taken whenever there is a significant design change to ensure the most applicable methods are in use at all times. Applicability to the single-phase and two-phase hydraulic models discussed in Subsections 4.4.2.6.1 and 4.4.2.6.2 is anfirmed by prototype (64-rod bundle) flow tests. The typical range of the test data is summarized in Table 4.4-8. O 4.4-14

_ - - .- - -- _- _ _ . . =- . . __- - _ _ . . . - l GESSAR II 22A7007

  ;                                      238 NUCLEAR ISLAND                         Rev. 0 4.4.2.7.2     Void Fraction Correlation 4

The void fraction correlation used is a version of the Zuber-Findlay model (Reference 11) where the concentration parameter and void drift coefficient are based on comparison with a large quantity l of worldwide data (References 13-24). 4.4.2.7.3 Heat Transfer Correlation f The Jens-Lottes (Reference 6) wall superheat equation is used in ] tuel design to determine the cladding-to-coolant heat transfer coefficients for nucleate boiling. i

 !               4.4.2.8   Thermal Effects of Operational Transients i

The evaluation of the core's capability to withstand the thermal effects resulting from anticipated operational transients is covered in Chapter 15 (Accident Analyses). 4.4.2.9 Uncertainties in Estimates Uncertainties in thermal-hydraulic parameters are considered in the statistical analysis which is performed to establish the f:tal cladding integrity safety limit such that at least 99.9% of the i fuel rods in the core are expected not to experience boiling transition during any moderate frequency transient event. The . statistical model and analytical procedure are described in detail l in Reference 1. The conservative power distribution used for the ! statistical analysis is shown in Figure 4.4-8 in terms of relative bundle power histogram. The uncertainties considered and their input values for the analysis are shown in Table 4.4-9. i lO I 4.4-15 i I

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O 4.4.2.10 Flux Tilt Considerations O For flux tilt considerations, refer to Subsection 4.3.2.2.7. 4.4.3 Description of the Thermal and Hydraulic Design of the Reactor Coolant System The thermal and hydraulic design of the reactor coolant system is described in this section. 4.4.3.1 Plant Configuration Data 4.4.3.1.1 Reactor Coolant System Configuration The reactor coolant system is described in Section 5.4 and shown in isometric perspective in Figure 5.4-1. The piping sizes, fittings and valves are listed in Table 5.4-1. 4.4.3.1.2 Reactor Coolant System Thermal Hydraulic Data The steady-state distribution of temperature, pressure and flow rate for each flow path in the reactor coolant system is shown in Figure 5.1-1. 4.4.3.1.3 Reactor Coolant System Geometric Data Volumes of regions and components within the reactor vessel are shown in Figure 5.1-2. Table 4.4-11 provides the flow path length, height, liquid level, minimum elevations, and minimum flow areas for each major flow path volume within the reactor vessel and recirculation loops of the reactor coolant systems. Table 4.4-12 provides the lengths and sizes of all safety injection lines to the reactor coolant system. 4.4-16

  ._ . .___ _ _ _.               --. . _ _ - _ _ _               _ _ _ . _ _ _ _ - >               _~                                              ._  _ . _ _ _                           . . _ _ _                        _. _            . _          .

} GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i i 4.4.3.2 Operating Restrictions on Pumps ExpJcted recirculation pump performance curves are shown in , f Figure 5.4-3. These curves are valid for all conditions with a l normal operating range varying from approximately 20% to 115% of ! rated pump flow. The pump characteristics, including considerations of NPSH l requirements, are the same for the conditions-of two-pump and one pump operation as described in Subsection 5.4.1. Sub-section 4.4.3.3 gives the operating limits imposed on the recir-

                                                                                     ~

culation pumps by cavitation, pump loads, bearing design flow starvation, and pump speed. ) i' i 4.4.3.3 Power-Flow Operating Map

4.4.3.3.1 Limits for Normal Operation

~ A BWR must operate with certain restrictions because of pump net positive suction head (NPS!!) , overall plant control characteris-tics, core thermal power limits, etc. The power-flow map for the power range of operation is shown in Figure 4.4-S. The nuclear

system equipment, nuclear instrumentation, and the reactor protec-tion system, in conjunction with operating procedures, maintain  ;

operations within the area of this map for normal operating condi-i tions. The boundaries on this map are as follows: i Natural Circulation Line, A: The operating state of the reactor moves along this line for the normal control rod withdrawal sequence in the absence of recirculation pump operation. 105% Steam Flow Rod Line or Rated Power (Whichever Is Less): The 105% steam flow rod line passes through 104.2% power at ,! 100% flow. The operating state for the reactor follows this 2 rod line (or similar ones) during recirculation flow changes i 4.4-17 7< .- ,_ , . _ _ _ _ _ . _ . - _ _ - . . _ _ _ _ , . _ . . _ _ _ _ . _ _ . _ . . . _ _ . . _ . _ _ , _~,---,,,-_.g.., , _ , - . _ _ _ _ _ . .

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.3.3.1 Limits for Normal Operation (Continued) with a fixed control rod pattern; however, rated power may not be exceeded. 105% steam flow rod line is based on a constant xenon concentration at 104.2% power and rated flow. Cavitation Protection Line: This line results from the recirculation pump, flow control valve and jet pump NPSH requirements. 4.4.3.3.1.1 Performance Characteristics Other performance characteristics snown on the power-flow operating map are: Constant Rod Lines: These lines show the change in power associated with flow changes, while maintaining constant control rod position. Constant Position Lines for Flow Control Valve, B, C, D, and F: These lines show the change in flow associated with power changes while maintaining flow-control valves at a constant position. 4.4.3.3.2 Regions of the Power Flow Map ! Region I - This region defines the system operational capability with the recirculation pumps and motors being driven by the low frequency motor-generator set at 25% speed. Flow is controlled by the flow control valve and power changes, during normal startup and shutdown, will be in j this region. The normal operating procedure is to start up along curve C - FCV wide open at 25% speed. i 4.4-18 1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O (_,) 4.4.3.3.2 Regions of the Power Flow Map (Continued) Region II - This region shows the area where 25% pump speed and 100% pump speed operating regimes overlap. The switching sequence from the low frequency m-g set to 100% speed will be done in this region. Region III - This is the low power area of the operating map where cavitation can be expected in the recircula-tion pumps, jet pumps, or flow control valves. Operation within this region is precluded by system interlocks which trip the main motor from the 100% speed power source to the 25% speed power source. Region IV - This represents the normal operating zone of the map where power changes can be made, by either Ox control rod movement or by core flow changes, through use of the flow control valves. 4.4.3.3.3 Design Features for Power-Flow Control The following limits and design features are employed to maintain power-flow conditions to the required values shown in Figure 4.4-5. (1) Minimum Power Limits at Intermediate and High Core Flows: l To prevent cavitation in the recirculation pumps, jet pumps, and flow control valves, the recirculation system is provided with an interlock to trip off the 100% speed ' power source and close the 25% speed power source if the difference between steamline temperature and recircula-tion pump inlet temperature is less than a preset value (9.8*F). This differential temperature is measured using l j high accuracy RTDs with a sensing error of less than

  .             0.2 F at the two standard deviation (20) confidence level.

4.4-19

GESSAR II 22A7007 238 MUCLEAR ISLAND Rev. 0 4.4.3.3.3 Design Features for Power-Flow Control (Continued) This action is initiated electronically through e 15-sec time delay. The interlock is active while in both the automatic and manual operation modes. (2) Minimum Power Limit at Low Core Flow: During low power, low loop flow operation, the temperature differ-ential interlock may not provide sufficient cavitation protection to the flow control valves. Therefore, the system is provided with an interlock to trip off the 100% speed power source and close the 25% speed power source if the feedwater flow falls below a preset level (22% of rated) and the flow control valves are below a preset position (20% open). The feedwater flow rate and recirculation flow control valve position are measured by existing process control instruments. The speed change action is electronically initiated. This interlock is active during both automatic and manual modes of operation. (3) Pump Bearing Limit: For pumps as large as the recir-culation pumps, practical limits of pump bearing design require that minimum pump flow be limited to 20% of rated. To assure this minimum flow, the system is designed so that the minimum flow control valve position will allow this rate of flow. (4) Valve Position: To prevent structural or cavitation damage to the recirculation pump due to pump suction flow starvation, the system is provided with an inter-lock to prevent starting the pumps, or to trip the pumps if the suction or discharge block valves are at less than 90% open position. This circuit is activated by a position O 4.4-20

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () 4.4.3.3.3 Design Features for Power-Flow Control (Continued) limit switch and is active before the pump is started, during manual operation mode, and during automatic operation mode. 4.4.3.3.3.1 Plow Control The principal modes of normal operation with valve flow control-low frequency motor generator (LFMG) set are summarized as follows: the recirculation pumps are started on the 100% speed power source in order to unseat the pump bearings. Suction and discharge block valves are full open and the flow control valve is in the minimum position. When the pump is near full speed, the main power source is tripped and the pump allowed to coast down to approximately 25% speed, where the LPMG set will power the pump and motor. The flow control valve is then opened to the maximum position, at which O ( ,/ point reactor heatup and pressurization can commence. When opera-ting pressure has been established, reactor power can be increased. This power-flow increase will follow a line within Region I of the flow control map shown in Figure 4.4-5. When reactor power is greater than approximately 20-28% of rated, the low feedwater flow interlock is cleared and the main recir-culation pumps can be switched to the 100% speed power source. The flow control valve is closed to the minimum position before the speed change to prevent large increases in core power and potential flux scram. This operation occurs within .cgion II of the operating map. The system is then brought to the desired power-flow level within the normal operating area of the map (Region IV) by opening the flow control valves and by withdrawing control rods. Control rod withdrawal with constant flow control valve position will result in power / flow changes along lines of constant c sub () (v) (constant position) . Flow control valve movement with constant 4.4-21

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.3.3.3.1 Plow Control (Continued) control rod position will result in power / flow changes along, or nearly parallel to, the rated flow control line. 4.4.3.4 Temperature-Power Operating Map (PWR) Not applicable. 4.4.3.5 Load-Following Characteristics Large negative operating reactivity coefficients inherent in the BWR provide the following important advantages: (1) good load-following with well-damped behavior and little undershoot or overshoot in the heat transfer response; (2) load-following with recirculation flow control; and (3) strong damping of epatial power disturbances. Design of the BWR includes the ability to follow load demands over a reasonable range without requiring operator action. Reactor power can be controlled automatically by flow control over approximately a 251 power range at, for example, approximately 1% per second for a 10% step-load change. 4.4.3.6 Thermal and Ilydraulic Characteristics Summary Table The thermal-hydraulic characteristics are provided in Table 4.4-1 for the core and tables of Section 5.4 and other portions of the reactor coolant system. O 4.4-22

I GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.4 Evaluation The design basis employed for the thermal and hydraulic character-istics incorporated in the core design, in conjunction with the plant equipment characteristics, nuclear instrumentation, and the reactor protection system, is to require that no fuel damage occur during normal operation or during abnormal operational transients. Demonstration that the applicable thermal-hydraulic limi'ts are not

 ;      exceeded is given by analyses.

4.4.4.1 Critical Power The GEXL critical power correlation is utilized in thermal-hydraulic evaluations. This correlation is discussed in more detail in i Subsection 4.4.2.2.1. i 4.4.4.2 Core Hydraulics O

                                                                                                                           ~
,       Core hydraulic models and correlations are discussed in Sub-sections 4.4.2.6, 4.4.2.7, and 4.4.4.5.

4.4.4.3 Influence of Power Distributions j The influence of power distributions on the thermal-hydraulic design is discussed in Reference 1, Appendix V. , . 4.4.4.4 Core Thermal Response , , , _ The thermal response of the core for accidents and expected" transient ! conditions is disucssed in Chapter 15 (Accident Analyses) . ! c i, S 4.4-23 I

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.4.5 Analytical Methods The analytical methods, thermodynamic data, and hydrodynamic data used in determining the thermal and nydraulic characteristics of the core are similar to those used throughout the nuclear power industry. Core thermal-hydraulic analyses are performed with the aid of a digital computer program. This program models the reactor core through a hydraulic description of orifices, lower tieplates, fuel rods, f uel rod splicers, upper tie plates, fuel channel, and the core bypass flow paths. 4.4.4.5.1 Reac tor Model The orifice, lower tieplate, fuel rod spacers, and upper tieplate are hydraulically represented as being separate, distinct local _ losses of zero thickness. The fuel channel cross section is represented by a square section with enclosed area equal to the unrodded cross-sectional area of the actual fuel channel. The fuel' channel auccably consists of three basic axial regions. The first and most important is the active fuel region which consists of the 62 fuel rods, 2 nonfueled rods, and 7 fuel rod spacers. The second is the nonfueled region consisting of 64 nonfueled rods and the upper tieplate. The third region represents the unrodded portion of the fuel channel above the upper tieplate. The active fuel region is considered in 24 independent axial segments or l I nodes over which fuel thermal properties are assumed constant and coolant properties are assumed to vary linearly. The code can handle 12 fuel channel types and 10 types of bypass l flow paths. In normal analyses, the fuel assemblies are modeled by 4 channel types - a " hot" central orifice region channel type, an average central orifice region channel type, a " hot" peripheral O 4.4-24

   .          - .    -. =

Y \ GESSAR II \ 22A7007 238 NUCLEAR ISLAND Rev. 0

                                                      >t                     ',

4.4.4.5.1 Reactor Model (Continued) ( i orifice region type and an average peripheral or,1fice region type. Usually, there is one fuel assembly representing each of- ,

;        the " hot" types.        The average types then make up the balance of                                             ,<

the core. l The computer program iterates on flow through 'eIach flow path (fuel assemblies and bypass paths) until the total differential pressure (plenum to plenum) across each path is equal, and the sum of the flows through each path equals the total core. flow. 4 Orificing is selected to optimize the core flow distributions 4 between orifice regions as discussed in Subsection 4.4.2.5. The core design pressure is determined from the required turbine throttle pressure, the steamline pressure drop, steam dryer } pressure drop, and the steam separator pressure drop. The core i () inlet enthalpy is determined from the reactor and turbine heat balances. The required core flow is then determined by applying the procedures of this section and specifications such that the thermal limits of Reference 1 are satisfied an'd the nominal expec-ted bypass flow fraction is approximately 10%. The results of applying these methods and specifications are: i { (1) flow for each bundle type; a (2) flow for each bypass path; (3) core pressure drop; I, (4) fluid property axial distribt3_ou , each bundle type; and l l l ! (5) CPR calculations for each bundle type. lO l < 4.4-25 < r

                                                        --.m-y_--,m.s.           --g.myy   --mrp--,--  -----e- - - . = , . - - . .       --

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.4.5.2 System Flow Balances The basic assumption used by the code in performing the hydraulic analysis is that the flow entering the core will divide itself between the fuel bundles and the bypass flow paths such that each assembly and bypass flow path oxperience the same pressure drop. The bypass flow paths considered are described in Table 4.4-10 and showwn in Figure 4.4-1. Due to the large flow area, the pressure drop in the bypass region above the core plate is essentially all elevation head. 'thu s , the sum of the core plate differential pressure and the bypass region elevation head is equal to the core differential pressure. The total core flow, less the control rod cooling flow, enters the lower plenum through the jet pumps. A fraction of this passes through the various bypass paths. The remainder passes through the orifice in the fuel support (experiencing a pressure loss) where more flow is lost through the fit-up between the fuel sup-port and the lower tieplate and also through the lower tieplate holes into the 'ypass region. The majority of the flow continues through the lower tieplate (experiencing a pressure loss) where some flow is lost through the flow path defined by the fuel chan-nel and lower tieplate, and restricted by the finger springs, into the bypass region. The flow through the bypass flow paths is expressed by the form: W = Ci AP +C 2 AP C" +C 3 Ap2, Full-scale tests have been performed to establish the flow coefficients for the major flow paths (Reference 12). These tests simulate actual plant configurations which have several parallel flow paths and, therefore, the flow coefficients for the individual O 4.4-26

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.4.5.2 System Flow Balances (Continued) l paths could not be separated. However, analytical models of the < l individual flow paths were developed as an independent check of the tests. The models were derived for actual BWR design dimen-i sions and considered the effects of dimensional variations. These models predicted the test results when the as-built dimensions were applied. When using these models for hydraulic design cal-culations, nominal drawing dimensions are used. This is done to yield the most accurate prediction of the expected bypass flow. With the large number of components in a typical BWR core, devia-f tions from the nominal dimensions will tend to statistically cancel, resulting in a total bypass flow best represented by that calculated using nominal dimensions. I The balance of the flow enters the fuel bundle from the lower tie-plate and passes through the fuel rod channel spaces. A small portion of the in-channel flow enters the nonfueled rods through

three orifice holes in each rod just above the lower tieplate.

] This flow, normally referred to as the water rod flow, remixes 7 2 with the active coolant channel flow below the upper tieplate. The uncertainties in calculations and the resultant uncertainty in reactor coolant system flow rate are provided in Table 4.4-9. 4.4.4.5.3 System Heat Balances Within the fuel assembly, heat balances on the active coolant are performed nodally. Fluid properties, expressed as the bundle average at the particular node of interest, are based on Reference 7. j In evaluating fluid properties, a constant pressure model is used. i The core power is divided into two parts: an active coolant power and a bypass flow power. The bypass flow is heated by neutron-i slowing down and gamma heating in the water, and by heat transfer t a 4.4-27

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.4.5.3 System Heat Balances (Continued) through the channe ' alls. Heat is also transferred to the bypass flow from structures and control elements which are, themselves, heated by gamma absorption and by n, a reaction in the control material. The fraction of total reactor power deposited in the bypass region is very nearly 2%. A similar phenomenon occurs, with the fuel bundle, to the active coolant and the water rod flows. The not effect is that 96% of the core power is conducted through the fuel cladding and appears as heat flux. In design analyses, the power is allocated to the individual fuel bundles using a relative power factor. The power distribution along the length of the fuel bundle is specified with axial power factors which distribute the bundle's power among the 24 axial nodes. A nodal local peaking factor is used to establish the peak heat flux at each nodal location. g The relative (radial) and axial power distributions when used with the bundle flow determine the axial coolant property distribution resulting in sufficient information to calculate the pressure drop components within each fuel assembly type. Once the equal pressure drop criterion has been satisfied, the critical bundle power (the power which would result in cirtical quality existing at some point in the bundle using the correlation expressed in Reference 9) is determined by an iterative process for each fuel type. In applying the above methods to core design, the number of bundles (for a specified core thermal power) and bundle geometry (8x8, rod diameter, etc.) are selected based on power density and LilGR limits. O 4.4-28

w. _ _ _ _. . __

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O 4.4.4.6 Thermal-Hydraulic Stability Analysis (O) T 4.4.4.6.1 Introduction There are many definition of stability, but for feedback processes a and control systems it can be defined as follows: a system is stable if, following a disturbance, the transient settles to a i steady, noneyclic state. A system may also be acceptably safe even if oscillatory, provided that any limit cycle of the oscillations is less than a prescribed magnitude. Instability, then, is either a continual departure from a final steady-state value or a greater-than prescribed limit cycle about the final steady-state value. The mechanism for instability can be explained in terms of fre-quency response. Consider a sinusoidal input to a feedback control system which, for the moment, has the feedback disconnected. If there were no time lags or delays between input and output, the output would be in phase with the input. Connecting the output so as to subtract from the input (negative feedback or 180 degrees 1 out-of-phase connection) would result in stable closed loop opera-tion. However, natural laws can cause phase shift between output and input and, should the phase shift reach 180 degrees, the feed-i back signal would be reinforcing the input signal rather than:sub-tracting from it. If the feedback signal were equal to or larger than the input signal (loop gain equal to one or greater), the input signal could be disconnected and the system would continue i to oscillate. If the feedback signal were less than the input I signal (loop gains less than one) , the oscillations would die out. l f It is possible for an unstable process to be stabilized by adding a control system. In general, however, it is preferable that a l process with inherent feedback be designed to be stable by itself () before it is combined with other processes and control systems. 4.4-29

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.4.6.1 Introduction (Continued) The design of the BWR is based on this premise, that individual system components are stable. 4.4.4.6.2 Description Three types of stability considered in the design of boiling water reactor are: (1) reactor core (reactivity) stability; (2) channel hydrodynamic stability; and (3) total system stability. Reactivity feedback instability of the reactor core could drive the reactor into power oscillations. Ilydrodynamic channel instability could impede heat transfer to the moderator and drive the reactor into power oscillations. The total system stability considers control system dynamics combined with basic process dynamics. A s table system is analytically demonstrated if no inherent limit cycle or divergent oscillation develops within the system as a result of calculated step disturbances of any critical variable, such as steam flow, pressure, neutron flux, and recirculation flow. The criteria to be considered are stated in terms of two com-patible parameters. First is the decay ratio x2/x o, designated as the ratio of the magnitude of the second overshoot to the first overshoot resulting from a step perturbation. This characteristic provides a graphic representation of the physical responsiveness of the system, which is readily evaluated in a time-domain analysis. Second is the da: aping coef ficient C , the definition of which corresponds to the pole pair closest to the jw axis in the s-plane for the system closed loop transfer function. This parameter also applies to the frequency-domain interpretation. The damping coefficient is related to the decay ratio as shown in Figure 4.4-2. O 4.4-30

_ . . . . . . . . . - ~ -- - -- - - - - - . GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i () 4.4.4.6.3 Stability Criteria The assurance that the total plant-is stable and, therefore, has significant safety margin shall be demonstrated analytically j when the decay ratio, x2/xo, is less than 1.0 or, equivalently, when the damping coefficient, (n, is greater than zero for each type of stability discussed. Special attention is given to differentiate between inherent system limit cycles and small, acceptable limit cycles that are always present, even in the most l i stable reactors. The latter are caused by physical nonlinearities ] (deadband , striction,-etc.) in real control system and are not representative of inherent hydrodynamic or reactivity instabilities in the reactor. The ultimate performance limit criteria for the

three types of dynamic performarce are summarized below in terms of decay ratio and damping coefficient:

Channel hydrodynamic stability x2/Xo <l, (n

J l

• Reactor core (reactivity) stability x2/xo <l, En' i

I, l Total system stability x2/xo <1, E n 0 These criteria shall be satisfied for all attainable conditions of the reactor that may be encountered in the course of plant opera- .I tion. For stability purposes, the most severe condition to which these criteria will be applied correspond to the highest attainable rod line intersection with natural circulation flow. 5 l 4.4.4.6.4 Mathematical Model The mathematical model representing the core examines the linear-ized reactivity response of a reactor system with density-dependent reactivity feedback caused by boiling. In addition, the hydro-i dynamics of various hydraulically coupled reactor channels, or 1, 1 j 4.4-31 4

     --re,, ,,     m,.e-       , ~ -      e--~n     ----w, e--              ----,,--~,,,-..<~,~~----n----e                -----r,-.-----,~n     -         ,---e w~

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.4.4.6.4 Mathematical Model (Continued) regions, are examined separately on an axially multinoded basis by grouping various channels that are thermodynamically and hydraulically similar. This interchannel hydrodynamic interaction, or coupling, exists through pressure variations in the inlet plenum, such as can be caused by disturbances in the flow distribution between regions or channels. This approach provides a reasonably accurate, three-dimensional representation of the reactor's hydrodynamics. The core model (References 25-30), shown in block diagram form in Figure 4.4-3, solves the dynamic equations that represent the reac-tor core in the frequency domain. From the solution of these dynamic equations, the reactivity and individual channel hydro-dynamic stability of the boiling water reactor is determined for a given reactor flow rate, power distribution, and total power. This gives the most basic understanding of the inherent core behavior (and hence the system behavior) and is the principal consideration in evaluating the stable performance of the reactor. As new experi-mental or reactor operating data are obtained, the model is refined to improve its capability and accuracy. The plant model considers the entire reactor system, neutronics, heat transfer, hydraulics, and the basic processes, as well as associated control systems such as the flow controller, pressure regulator, feedwater controller, ctc. Although the control systems may be stable when analyzed individually, final control system settings must be made in conjunction with the operating reactor so that the entire system is stable. The plant model yields results that are essentially equivalent to those achieved with the core model and allows the addition of the controllers, which have adjustable features permitting the attainment of the desired performance. O 4.4-32

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () 4.4.4.6.4 Mathematical Model (Con tinued) The plant model solves the dynamic equations that present the BWR system in the time domain. The variables, such as steam flow and pressure, are represented as a function of time. The extensive-ness of this model is shown in Reference 10. The model is periodically refined, as new experimental or reactor operating data are obtained, to improve its capability and accuracy. 4.4.4.6.5 Analytical Confirmation Figure 4.4-4 demonstrates the competence and inherent conser-vatism of the core stability model. The relationship of the cal-culated damping coefficient from the reactor core dynamic analytical code is related to measured results from 14 rod oscil-lator tests performed at large operating BWR plants by the General Electric Company. The correlated, Most Probable Values, () based on a least squares analysis, and the line representing a 97.5% (two sigma) confidence level, below which the actual values will fall, are presented in Figure 4.4-4. l The results show the analytical methods to be an effective and 4 useful design tool, with significant conservatism in its applica-tion to boiling water reactor core evaluation. Neal and Zivi (Ref-t erence 7) further confirm the effective application of essentially the same model to channel and core analysis, as does Reference 8. 4.4.4.6.6 Analysis Results The analysis is performed using a bounding value void coefficient { (Figure 4.3-19) which is expected to cover several cycles of ! operation. For all operating conditions in which the actual void f coefficient is less than the bounding value, the analysis remains

valid. The most sensitive reactor operating condition is that 4.4-33

GESSAR II 22A7007 238 NUC". EAR ISLAND Rev. 0 4.4.4.6.6 Analysis Results (Continued) corresponding to the highest attajnable rod line intersection with natural circulation flow. A generic power / flow operating condi-tion (i.e., 51.5% nuclear boiler rated power) is analyzed to bound the typical power / flow operating region as shown in Figure 4.4-5. Typical values of reactor core stability are as follows: Natural Circulation Reactor 51.5% Power Core Stability (105% Rod Pattern) Decay ratio, X2/Xo 0.98 Resonant frequency, Hz 0.43 The calculated values show the reactor to be in compliance with the ultimate performance criteria to the most responsive attain-able mode as cited for the reactor core stability evaluation. Figure 4.4-6 shows the cal ulated variation of the decay ratio over the normal power-flow range for the bounding core conditions. The channel hydrodynamic perormance is evaluated at the most limiting condition that occurs at the bounding core condition, peaked to the bottom of the core. The calculations yield decay ratios as presented below: Channel Natural Circulation Hydrodynamic Performance 51.5% Power Decay-ratio (X2/Xo) 0.98 Resonant frequency (Hz) 0.43 At this most responsive attainable mode, the most responsive channel conforms with the ultimate performance criteria of 1.0 decay ratio. h 4.4-34

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () 4.4.4.6.6 Analysis Results (Continued) The channel performance over the entire range of attainable operation is well below the threshold of instability. Conformance with the ultimate performance criterion is further tested by assuming that the reactor is initially operating at the-most sensitive condition. The nuclear system is then subjected to step disturbances from pressure regulator setpoint, control rods and level controller setpoint. These time responses are shown in Figures 4.4-7a, b, and c. It is-clear that the decay ratio is less than 1.0 and in conformance with the ultimate performance criterion. The analysis represents a typical BWR response with control systems similar to that designed for this plant. The control system settings used are typical values chosen within the range of equip-() ment capability. The final control system settings will be estab-lished and optimized during plant startup and, hence, the final system response will be somewhat different than that represented in Figures 4.4-7a, b, and c. The results in these figures demon-strate that a highly stable mode of operation is attainable and within the range of permitted control system settings permitted by the design. 1 t 4.4.4.6.6.1 Impact of Prepressurized Fuel on Stability i j The impact on stabil.ity parameters of GE BWR fuel prepressurized i up to 3 atmospheres has been evaluated generally and documented j in References 31 through 34. Based on these evaluations, it has !' been concluded that the effects of prepressurization up to l 3 atmospheres is bounded by the current thermal-hydraulic stability analysis and add itional analyses are not required. O ! 4.4-35 I

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ll 4.4.5 Testing and Verification The testing and verification techniques to be used to assure that the planned thermal and hydraulic design characteristics of the core have been provided, and will remain within required limits throughout core lifetime, are discussed in Chapter 14, (Initial Test Program). A summary is as follows: (1) Preoperational '.'esting Tests are performed during the preoperational test program to confirm that construction is complete and that all process and safety equipment is operational. Baseline data are taken to assist in the evaluation of subsequent tests. Heat balance instrumentation and jet pump flow and core temperature instrumentation are calibrated and set points verified. (2) Initial Startup Hot functional tests are conducted with the reactor between 5 and 10% power. Core performance is monitored continuously to assure that the reactor is operating within allowable limits (e.g., peaking factors, LHGR, etc.) and is evaluated periodically to verify the core expected and actual performance margins. 4.4.6 _ Instrumentation Requirements The reactor vessel instrumentation monitors the key reactor vessel operating parameters during planned operations. This 4.4-36 e

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O 5 4.4.6 Instrumentation Requirements (Continued) I ensures sufficient control of the parameters. The following reactor vessel sensors are discussed in Subsection 7.7.1.1 and 7.6.2.1: (1) Reactor Vessel Temperature i (2) Reactor Vessel Water Level (3) Reactor Vessel Coolant Flow Rates and Differential Pressures 4 l j (4) Reactor Vessel Internal Pressure O

     \~ #                                                            Neutron Monitoring System (5) 4.4.6.1                     Loose Parts To be supplied by Applicant.

i 1 4.4.7 References

1. " General Electric Thermal Analysis Basis (GETAB): Data, Correlation, and Design Application", General Electric I Company, January 1977 (NEDO-10958A).

1

2. " Core Flow Distribution in a Modern Boiling Water Reactor as Measured in Monticello", August 1976 (NEDO-10722A).

3. " Peach Bottom Atomic Power Station Units 2 and 3, Safety.

Analysis Report for Plant Modifications to Eliminate Signifi-cant In-Core Vibration," September 1975 (NEDO-20994). i tO 4.4-37 i _,.,.,_...__.~.,,.,,.~-_..,____.._,_._-_,-..-_.....,-.,_m. _ . . , . . . . . , . . .y _ ._ .,-_-- , . - . _

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ll 4.4.7 References (Continued)

4. R. C. Martinelli and D. E. Nelson, " Prediction of Pressure Drops During Forced Convection Boiling of Water", ASME Trans., 70, pp 695-702, 1948.

5. C. u. Baroczy, "A Systematic Correlation for Two-Phase Pressure Drop", Heat Transfer Conference (Los Angeles),

AICLE, Preprint No. 37, 1966.

6. W. H. Jens and P. A. Lottes, Analysis of Heat Transfer, Burnout, Pressure Drop, and Density Data for High Pressure Water, USAEC Report 4627, 1972.

7. L. G. Neal and S. M. Zivi, "The Stability of Boiling Water Reactors and Loops", Nuclear Science and Engineering, 30

p. 25, 1967.

8. " Stability and Dynamic Performance of the General Electric Boiling Water Reactor", General Electric Company, January 1977 (NEDO-21506).

9. S. Levy, et. al., " Experience with BWR Fuel Rods Operating Above Critical Flux", Nucleonics, April 1965.

10. " Analytical Methods of Plant Transient Evaluations for General Electric Boiling Water Reactor", General Electric Company, BWR Systems Department, February 1973 (NEDO-10802).

11. N. Zuber and J. A. Findlay, " Average Volumetric Concentration in Two-Phase Flow Systems", Trans. ASME, Journal of Heat Transfer, November 1965.

12. " Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibration", NEDE-21156 (Class III), January 1976.

13. H. S. Isbin, H. A. Rodriguez, H. C. Larson and B. D. Pattie,

      " Void Fractions in Two-Phase Flow", A.I. Ch.E. Journal, Volume 5, No. 4, pp. 427-432, December 1959.

14. H. S. Isbin, N. C. Sher, K. C. Eddy, " Void Fractions in Two-Phase Steam-Water Flow", A.I. Ch.E Journal, Volume 3, No. 1, pp. 136-142, March 1957.

O 4.4.38

_~ .- - . ... ._ . . . - _ _ ~ . -. I GESSAR II 22A7007

 ,                                                                                    238 NUCLEAR ISLAND                            Rev. O 4.4.7    References (Continued)

15. J. F. Marchaterre, "The Effect of Pressure on Boiling

Density in Multiple Rectangular Channels", ANL-5522, February 1956.

16. E. Janssen and J..A. Kervinen, "Two-Phase Pressure Drop-in Straight Pipes and Channels; Water-Steam Mixtures at 600 to 1400 psia", May 1964, (GEAP-4616).

17. W. H. Cook, " Boiling Density in Vertical Rectangular Multichannel Section with Natural Circulation", ANL-5621, November 1956.

18. G. W. Mauer, "A Method of Predicting Steady-State Boiling 4 Vapor Fractions in Reactor Coolant Channels", WAPD-BT-19,

! June 1960.

19. S. Z. Rouhani, " Void Measurements in the Region of Subcooled and Low Quality Boiling", Symposium on Two-Phase Flow,

(~N University of Exeter, Devon, England, June 1965.

20. A. Firstenberg and L. G. Neal, " Kinetic Studies of Heterogeneous Water Reactors", STL 372-38, April 15, 1966.

I l 21. J. K. Ferrel, "A Study of Convection Boiling Inside 2 Channels", North Carolina State University, Raleigh, N.C., , September 30, 1964. i

22. S. Z. Rouhani, " Void Measurements in the Region of Subcooled and Low Quality Boiling", l' art II, AE-RTL-788, Akticholaget, Atomenergi, Studsvik, Sweden, April 1966.

i 23. H. Christensen, " Power-to-Void Transfer Functions", ANL-6385, l July 1961. l l 24. R. A. Egen, D. A. Dingee, J. W. Chastain, " Vapor Formation i and Behavior in Boiling Heat Transfer", BMI-ll63, February j 1957.

25. KAPL-2170 Hydrodynamic Stability of a Boiling Channel, by

A.B. Jones, 2 October 1961.

l 26. KAPL-2208 Hydrodynamic Stability of a Boiling Channel Part 2, by A.B. Jones, 20 April 1962. l 1 l 27. KAPL-2290 Hydrodynamic Stability of a Boiling Channel Part 3, i by A.B. Jones and D.G. Dight, 28 June 1963.

4.4-39 I. -

          ., - . . . __-. ,.          _. . ,  . , _ . , _ _ . . _ _ _ _ , _ . . ~ . , _ . ,             .- _          _         _    _ , . -__

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O g 4.4.7 References (Continued)

28. KAPL-3070 Hydrodynamic Stability of a Boiling Channel Part 4, by A.B. Jones, 18 August 1964.

29. KAPL-3072 Reactivity Stability of a Boiling Reactor Part 1, by A.B. Jones and W.M. Yarbrough, 14 September 1964.

30. KAPL-3093 Reactivity Stability of a Boiling Reactor Part 2 by A.B. Jones, 1 March 1965.

31. R. B. Elkins, " Fuel Rod Prepressurization Amendment 1",

May 1978 (NEDO-2378 6-1) .

32. Letter, E. D. Fuller to O. D. Parr, "NRC Request for Addi-tional Information on Fuel Rod Prepressurization", June 8, 1978.

33. Letter, E. D. Fuller to O. D. Parr, "NRC Request for Addi-tional Information on Fuel Rod Prepressurization",

August 14, 1978.

34. R . B. Elkins, " Fuel Rod Prepressurization", March 1978 (NEDE-23786-1-P).

O 4.4-40

GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 (} Table 4.4-1 THERMAL AND HYDRAULIC DESIGN CHARACTERISTICS OF THE REACTOR CORE General Operating Conditions (238-748) Reference design thermal 3579 output (Mwt) i Power level for engineered 3730 safety features (Mwt) i Steam flow rate, at 420*F final 15.400 feedwater temperature (millions lb/hr) Core coolant flow rate 104.0 (millions lb/hr) Feedwater flow rate (millions 15.367 lb/hr) System pressure, nominal in 1040 steam dome (psia) system pressure, nominal core 1055 design (psia) Coolant saturation temperature 551 at core design pressure ( F) Average power density 54.1 (kW/ liter) Maximum Linear Heat Generation 13.4 Rate (kW/ft) Average Linear Heat Generation 5.9 Rate (kW/ft) I 73,303 Core total heat transfer area (ft 2) l i 4.4-41 _ , _ - _ . _ . , - _ _ _ _ _ . _ . ~ - . _ . ..

GESSAR II 22A70c-238 NUCLEAR ISLAI4D Rev. O Table 4.4-1 (Continued) TilERMAL AND llYDRAULIC DESIGN CilARACTERISTICS OF Tile REACTOR CORE General Operating Conditions (238-748) Maximum heat flux (Btu /hr-ft2) 361,600 Average heat flux (Btu /hr-ft2) 159,500 Design operating minimum 1.20 critical power ratio (MCPR) Core inlet enthalpy at 420 F 527.7 FFWT (Btu /lb) Core inlet temperature, at 533 420 F FFWT ( F) Core maximum exit voids within 79.0 assemblies (%) Core average void fraction, 0.414 active coolant Maximum fuel temperature ( F) 3435 Active coolant flow area per 15.164 assembly (in.2) Core average inlet velocity 6.98 (ft/sec) Maximum inlet velocity (ft/sec) 8.54 Total core pressure drop (psi) 26.4 Core support plate pressure 22.0 drop (psi) Average orifice pressure drop 5.71 Central region (psi) O 4.4-42 i

      . - .   . . . .       .-. . . - - . . .      - - . . - _ . - . _                    - - - - - .                                 . . -                                          . - -   ~ . . - . . .

1' GESSAR II 22A7007 i 238 NUCLEAR ISLAND Rev. 0 l f Table 4.4-1 (Continued) 1 O) THERMAL AND HYDRAULIC DESIGN CHARACTERISTICS' OF THE REACTOR CORE I General Operating Conditions (238-748) l l Average orifice pressure drop 18.68 Peripheral region (psi) 1 . Maximum channel pressure 15.40. loading (psi) Average-power assembly channel 14.1 pressure loading (bottom) (psi) I Shroud support ring and lower 25.7 shroud pressure loading Upper shroud pressure loading 3.7

 ;                       (psi)

I I i l 1 l i l i 4 I l i 4.4-43 l F l-_.-,,.._.,_.,._--...-..-..-_.,.--,.....-....-.-.. - . . , - . , . _ _ , _ . _ . . , - _ . . . - . . . . . . . - . . . , . . . . . . , . . , . , - - . . - , -

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O h Table 4.4-2 AXIAL POWER DISTRIBUTION USED TO CALCULATE MCPR OPERATING LIMIT Node Axial Power Factor Bottom of Core 1 0.47 2 0.55 3 0.64 4 0.74 5 n.85 6 0.97 7 1.10 8 1.21 9 1.29 10 1.34 11 1.38 12 1.40 13 1.39 14 1.36 15 1.30 16 1.23 17 1.15 18 1.08 19 1.01 20 0.93 21 0.84 22 0.74 23 0.60 Top of Core 24 0.43 l 4.4-44

f GESSAR II 22A7007 () l 238 NUCLEAR ISLAND Rev. O Table 4.4-3 l VOID DISTRIBUTION 1 Core Average Value - 0.414

 !                                                           Maximum Exit Value - 0.790 Active Fuel Length - 150 inches Core Average                                                  Maximum Channel Node              (Average Node Value)                                              (End of Node Value)

I Bottom of Core 1 0 0 2 0 0.008 l 3 0.008 0.084 4 0.042 0.204 5 0.104 0.314 6 0.178 0.402 7 0.253 0.475 8 0.323 0.532

9 0.381 0.578 10 0.429 0.614 0.467 O

11 0.644 12 0.498 0.668 13 0.524 0.687 14 0.545 0.703 ! 15 0.563 0.718 16 0.579 0.730 17 0.593 0.742 , 18 0.606 0.753 . 19 0.619 0.763 ' ! 20 0.631 0.773 21 0.640 0.780 22 0.648 0.785 23 0.654 0.789 ( Top of Core 24 0.656 0.790 0 4.4-45

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 4.4-4 AXIAL POWER DISTRIBUTION USED TO GENERATE VOID AND QUALITY DISTRIBUTIONS Node Axial Power Factor Bottom of Core 1 0.38 2 0.69 3 0.93 4 1.10 5 1.21 6 1.30 7 1.47 8 1.51 9 1.49 10 1.44 11 1.36 12 1.28 13 1.16 14 1.06 l 15 1.01 16 0.97 17 0.94 18 0.97 19 0.96 20 0.91 21 0.77 22 0.59 23 0.38 Top of Core 24 0.12 0 4.4-46

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 4.4-5 FLOW QUALITY DISTRIBUTION Core Average Value - 0.079 Maximum Exit Value - 0.332 Active Fuel 7.ength - 150 inches Core Average- Maximum Channel Node (Average Node Value) (End of Node Value) Bottom of Core 1 0 0 2 0 0 3 0 0.004 4 0.001- 0.014 5 0.004 0.031 6 0.011 0.050 7 0.020 0.072 8 0.031 0.096 9 0.043 0.118 10 0.055 0.140 11 0.066 0.161 0 12 13 14 0.077 0.087 0.097 0.181 0.199 0.215 15 0.106 0.231 16 0.114 0.245 17 0.122 0.260 18 0.130 0.275 19 0.138 0.289 20 0.146 0.303 21 0.153 0.315 22 0.159 0.324 23 0.163 0.330 Top of Core 24 0.165 0.332 O 4.4-47

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 4.4-6 CORE FLOW DISTRIBUTION Orifice Zone Central Central Peripheral Peripheral Description Hot Average Hot Average Relative Assembly Power 3.400 1.083 1.000 0.350 Relative Assembly Flow 0.926 1.051 0.563 0.641 O O 4.4-48

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 4.4-7 CALCULATED VS MEASURED CORE PLATE PRESSURE DROPS Test Condition Core Plate AP Plant Size Power (% rated) Flow (% rated) Meas (psid) Calc (psid) 183-368 83.9 100.2 25.10 24.82 251-764 95.3 94.9 18.14 17.91 99.3 96.9 18.69 18.47 224-580 70.3 60.8 5.04 5.05 99.3 99.3 14.74 14.77 218-548 86.7 100.6 17.17 19.30 90.3 96.0 16.13 17.85 218-560 66.4 59.9 7.47 6.73 79.2 94.4 18.24 17.38 O 251-764 46.9 51.3 46.6 68.0 103.3 48.0 6.13 18.50 3.99 7.55 18.00 3.52 64.9 70.3 9.42 8.90 75.9 101.0 18.51 18.50 57.8 46.4 3.79 3.46 70.1 71.0 9.57 9.32 96.4 98.9 19.50 19.25 l !O ! 4.4-49 ) i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 4.4-8 TYPICAL RANGE OF TEST DATA Measured Parametcr Test Condition Adiabatic Tests: 5 5 Spacer single phase loss N = 0.5 x 10 to 3.5 x 10 Re* coefficient Lower tieplate + orifice T= 100 to 500 F single phase loss coefficient Upper tieplate single-phase friction factor Spacer two-phase loss P = 800 to 1400 psia coefficient 6 6 Two-phase friction G = 0.5 x 10 to 1.5 x 10 multiplier lb/h-ft2 X = 0 to 40% Adiabatic Tests Ileated bundle pressure drop P = 800 to 1400 psia G = 0.5 x 106 to 1.5 x 106 lb/h=ft2 l 1 l l l }

• Reynolds Number O

f 1 4.4-50 I I

GESSAR II 22A7007 O 238 NUCLEAR ISLAND Rev. 0 Table 4.4-9 DESCRIPTION OF UNCERTAINTIES ~ Standard Deviation Quantity (%-of Point) Comment Feedwater Flow 1.76 This is'f$e largest component - ' of total reactor power uncertainty. Feedwater Temperature 0.76 These are the other signif[- ' Reactor Pressure 0.5 cant parameters in core power. determination. ' Core Inlet Temperature 0.2 Affect quality and-boiling Core Total Flow 2.5 length. Flow is not measured. directly, but is calculated from jet pump.6P. The listed uncertainty in flow corres-

     ;                                           ponds to 11.2% standard devi-
  ./                                             ation in each individual pump difference.
       -Channel. Flow Area            3.0        This accounts for manufactur-ing and-service induced variations in the free flow-area within the channel.

Friction Factor 10.0 Accounts for uncertainty in Multiplier the correlation representing two-phase pressure losses, i Channel Friction 5.0 Represents variation in the Factor Multiplier pressure loss characteristics

of individual channels. Flow area and pressure loss varia-

! tions affect the core flow l distribution, influencing the ! quality and boiling length in individual channels. i TIP Readings 6.3 These sets of data are the I base from which gross power ! distribution is determined. ! The assigned uncertainties I 4.4-51

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 4.4-9 (Continued) DESCRIPTION OF UNCERTAINTIES Standard Deviation Quantity (% of Point) Comment TIP Readings (continued) include all electric and geometrical components plus a contribution from the analytical extrapolation from the chamber location to adjacent fuel assembly seg-ment. Also included are uncertainties contributed by the LPRM system. LPRM read-ings are used to correct the power distribution calcula-tions for changes which have occurred since the last TIP survey. The assigned uncer-tainty affects power distri-bution in the same manner as the base TIP reading uncertainty. R-Factor 1.5 This is the last of the three power distribution related uncertainties. It is a function of the uncertainty l in local fuel rod power. Critical Power 3.6 Uncertainty in the GEXL correlation expressed in l terms of critical power. l 9 l 4.4-52 l l

t l gS GESSAR II , l 22A7007 (_,) 238 !TUCLEAR ISLAND Rev. 0 Table 4.4-10 BYPASS FLOW PATI!S Flow Path Description Driving Pressures, Number of Paths la. Between Fuel Support and Core Plate One/ Control Rod the control Rod Guide Differential Tube (Upper Path) lb. Between Puol Support Core Plate One/ Control Rod and the Control Rod Differential Guide Tube (Lower Path)

2. Between Core Plate and Core Plate One/ Control7Rod Control Rod Guide Tube Differential

3. Between Core Support and Core Plate One/ Instrument the In-Core Support Differential Instrument Guide Tube D 4. Between Core Plate and Core Plate One Shroud

5. Between Control Rod Core Plate One/ Control Rod Guide Tube and Control Differential ,

Rod Drive llousing

6. Between Fuel Support Channel Wall One/ Channel '

and Lower Tieplate Differential Plus Low Tieplate Differential

7. Control Rod Drive Independent of One/ Control Rod' Coolant Core

8. Between Fuel Channel Channel Wall One/ Channel and Lower Tieplate Differential

9. IIoles in Lower Tieplate Lower Tieplate/ Two/ Assembly Bypass Region Differential ,

4.4-53 f m - - g -

k GESSAR II 22A7007 238 t1UCLEAR ISLA?1D Rev. O I Tabic 4.4-11 REACTOR COOLAt1T SYSTEM GEOMETRIC DATA Ileight Elevation Flow and of Bottom Minimum Path Liquid of Each Flow Length Level Volume

• Areas (in.) (in.) (in.) (ft2)

A. Lower Plenum 213.5 213.5 -170.5 84.0 213.5 i

B. Core 164.5 164.5 43.0 146.5 t i 164.5 includes j bypass C. Upper Plenum and 179.0 179.0 207.5 57.5 Separators 179.0 D. Dome (Above IJormal 289.5 289.5 386.0 309.0 Water Level 0 E. L'- wncomer Area 311.5 311.5 -27.5 66.0 i 311.5 l

F. Recirculation Loops 114.0 ft 398.0 -392.0 132.5 in 2 and Jet Pumps (one loop) 398.0

• Reference Point is recirculation nozzle outlet centerline.

4.4-54

i i i I ,

!                                                                                                                               l GESSAR II                            22A7007 238 NUCLEAR ISLAND i

Os Rev. 0 , l j Table 4.4-12 i LENGTilS OF SAFETY INJECTION LINES i l Nominal i Diameter Pipe Length , Loop Line (in) Schedule - (ft) 1 IIPCS IIPCS 3 16 100 57 i 12 100 48 IIPCS 4 12 80 154 i LPCS LPCS 2 14 40 92 i 12 40 14 l LPCS 3 12- 80 140 I i LPCI"A" RilR 7 18 40 23 RHR 12 18 40 4 , RIIR 9 18 40 116 ! 14 40 45 ' RiiR 10 12 80 66 i LPCI"B" RIIR 13 18 40 23 J RHR 18 18 40 4 ! RilR 15 18 40 104 i 14 40 112 RilR 16 12 80 61 , LPCI"C" RHR 21 18 40 96 RiiR 22 14 80 216 12 80 58 e i NOTE Lengths given are to the nearest foot, and are measured from the appropriate pump outlet nozzle to the RPV nozzle. l l l i j I i 4.4-55/4.4-56 ,

                    ._ .- -.          --            - - -       -_         -..-_                               . _ = _ _ -

GESSAR II 22A7007 f 238 NUCLEAR ISLAND Rev. O NOTE: PERIPHER AL FUE L SUPPORTS ARE WELDED INTO THE CORE O LOWER TIE PLATE SUPPORT PLATE. FOR THESE BUNDLES. PATH NUMBERS 1 2.6 AND 7 DO NOT EXIST j f CHANNEL i s_______' 8 f

                                                                      )                                                                      _

9 CORE SUPPORT PLATE FUE L SUPPORT l f f"

                                                                   ~

IN-COR E r O' GUIDE TUBE I - i CONTROL ROD GUIDE TU8E < 1 Ib h(#( ._ i [ lb \ 1. 2. CONTROL ROD GUIDE TUBE FUE L SUPPORT CONTROL HOD GUIDE TUBE CORE SUPPORT PLATE

3. CORE SUPPORT PLATE IN-CORE GUIDE TUBE 4 CORE SUPPORT PLATE SHROUD 5 g CONTROL ROD 5. CONTROL ROD GUIDE TUBE ORIVE HOUSING DRIVE HOUSING 6. FUEL SUPPORT LOWER TIE PLATE

7. CONTROL ROD DRIVE COOLING WATER

                             ~b                                                                  8.      CHANNEL LOWER TIE PLATE

9. ALTERN ATE FLOW PATH HOLES Figure 4.4-1. Schematic of Reactor Assembly Showing the Bypass Flow Paths 4.4-57

GCSSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O 1.0 - 0.50 ~;c 0.20 - [ 5 G I b 8 0.10 - O 0.05 - 0.02 - I I l  ! I I I I! 1 I f l f f I I 0.01 0.01 0.02 0.05 0.10 020 0.50 1.0 DECAY FI ATIO (X 2 I*0 Figure 4.4-2. Damping Coefficient versus Decay Ratio (Second Order Systems) 4.4-58

GESSAR II 22.\7007 238 NUCLEAR ISLAND rov. O NEUTRON FLUX

RESPONSE

J L REACTIVITY

                                               +

PE FITUR DATION FIEACTOR KINETICS

                                           ' (

J b TOT AL RE ACTOR RE ACTIVITY FEEOBACK TOTAL INDIVIDUAL CHANNE L TYPE REACTIVITY FEEDBACK

                                                 *            *                                           +                REACTIVITY TO POWER
                                             -                  m                                           m              TRANSFER FUNCTION                          m
                                             '(j'                                                  (j'                      AT CONSTANT INLET FLOW 4  k,                                        A s, FROM OTHER O

CHANNEL TYPES FIE ACTIVITY TO F LOW TRANSFEF1 FUNCTION AT CONSTANT POWE R J L l F LOW TO POW ER I TRANSFER I FUNCTION l l l ' i ' TO OTHEF1 I CH ANNE LS i O Figure 4.4-3. Ilydrodynamic and Core Stability'Model l 4.4-59

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1A O KHB DATA O SENN DATA 1.6 - 1.4 - 1.2 - z W 9 1 1.0 - 9 E k O O O

        ~

2 o CONFIDENCE LINE 00 k u 0.6 - U O MOST PHOBABLE CORRELATION LINE Y a A Bn 0.4 - O O O s.2 _ O O O 1

                    !          I             l              l           l        l 0

0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 DAMPlNG COEFFICIENT to, FROM TEST DATA Figure 4.4-4. Comparison of Test Results With Reactor Core Analysis 4.4-60

r ' A mC>% g- NN>JCoa Mw gOMt>% t A CM O %D<. o ( O 0 2 1 0 I 1 1 _ N 0 F O I 0 _ 1 B IG O E F R g N _ g Oi _ U I l Tl L AN I 0 TO 9 6 V O gPO I I 4 B I VG p g Q O N O I s AE CR M a 0 O G 0 g E g 9 R M A 5 O, L I 0 8 n Og 1 fg i g E, t 4 oS a

                        /g         9                                        aN iA r

e Y mg p o o5 0 S I 7 O W y O L w H F o 5 T E l A R F P 0 O - P I C U 6 T r e O T N R E w N A C o T R N OTNO I S E P OI I L P TI TS O D I A 0 I S D C I 5 S P A A I OO PMLO O P Y P L T 5 M MU C - U UMI C I T I 4 MIMNA I I T A . NI XAMM M 0 4 O O 4 MM E T VU T C p,~ bgwh - e - EEL A U j r VVA LLVF AA D O A F

                                                             /                                                                    u g

VV T O / O i _ DDE E I TI Aq G 0 F _ M M

• I 3

NEEP IOEEP P PS IL I L TSS M RE NR N AP P U WIOEP 3G O m L U M UUM P. O G P C PP CLEUE R L R 0 R . . R L I 2 I CCIC AGAG C LI RRI ECNCN I I I A CCR I R EE TWTW U RR DY O YO E L L L L TWWTAL AL AOOANONO 0 NLLRAF AF 1 ABCDE F.

            ~               -             -    -                 -              _   -        -         -            O

- 0 0 0 0 0 0 0 _ 0 0 5 . 3 2 1 1 9 8 7 1 O cw g wE w' A* ImH l:  ;'

GESSAR II 22A7007 ) 2 38 NUCLEAR ISLs.ND Rev. O l 1.4 1 0' 1.2 ULTIMATE STAalLITY LIMIT th NATUHAL o 03 - CIRCULATION x 105% ROD LINE O Y 4 E 4 0 0.6 - 0.4 - O.2

                                                           '                                  I 0

O 20 40 60 80 100 PERCENT POWER Figtt re 4.4-6. Total core stability 4.4-62

     -w                                                                      s                                                                      m b

l 1 I E ITM BI FllR IN Rf W EP9ff 2 FDM REl GN!YR TDP 2 STM LIE RIE (PSIB 150. lM$kg 5 VT.S'KL S1Em Rm T8I El* 125. l"h%IQI 5 CITT RvE vt!!D FMC (1: q 8 Ttffl!!E !y RtN tZ) 75. l100. . a a a b gA = u u

n. E^ '1 ' A '2L- 25.

7 ^ # __ 12 12 12 b. O. t * - 4 -25. 8 - a to

0. 10. 2D. 30. 40. O. 10. 20. 23. 40. w TIE (El TIE INCI cc u ZO

. CM s O CA I M CD m M> w >% i 1DELIINDHEF-EP-Stint I EUFN98 RIK 2N R!EMMD LEVEttilo fSI 2Sm R E RT F.tR g[ gg), 3 N R NNSED LEVEL tlNDES) 120. m u uru. IAf r Rai,F11 E4 5 DRIVE FUWltII ,x Z C f(E. ED. b h W. "" - - -- IE II 40. 17 u u u

            - s               s            s                  s                        _-

O. h it =* _ t D. a a a = laus . O. tO. 20, 30. 40. O. 25. 30. 75. t00. IIE 17C8 CM RtW t%I yU c>

                                                                                                                                                         < -4
                                                                                                                                                         . o Figure 4.4-7a.                  10 PSI Pressure Regulator Setpoint Step at                                                o3 51.5% Rated Power (Natural Circulation)

I ElfilOI plt 2 1 W3E. RIE (PSil 2 PUM Ful NTUI TDP 2 STM LIE RIK IPSil gg* 3 RW StrFr& FEAT F1tfX g3* 3 TLPel t 5 RISE trSil 4 t t t t>OTU FI N - 4 urf I4IT NE-[5Ttsqal 5 VESSrL S1E rn FLW 5CCFF mT veID FME (11 g g g 6 ItrHIT t y FL W (21 im. n. s a It s b u u u u M-k1_ t i_ 1 i_ f .L_ B-It2 14 12 1:n o,....a.... . s.. .n.... _

                                                                                                                                                              .      n W

D. 10. 20, 30. 40. O. 10. -

20. 30. 40.

II T ISEC8 IIT ISECl CO A '2: O + C t1 .L- O Cn i t~ Cn O t1 > A > lc

                                                    ! LIWLIIWHMT-TP-SMIRT                              I EUI M FitR                                                      s 2 N R T N5fD LEWL tINCtf 51                        2 SLRACE PERT F1.tR                                            ss 3 N R SOGO a rVit tlHCl gc y ggg y-gg79,             ,f S)                                                                                CO 150.                                                                               120.                                                                       p 5 UF11VE FLtHIttI                                                                                                 &

5 im. _ g m. s h f

50. 4h IE IE l

40. u u u u

                                                                                        ~
            - 5                   5              5                     5                       -

O. . ..n.... s o,....i.... . D. 10. 20. 30. tea. O. 25. 50. 75. 100. g IIT (SEC1 CSE FLOM 1%I xw o>

                                                                                                                                                                      <w
                                                                                                                                                                      . o o

Figure 4.4-7b. 10-Cent Rod Reactivity Step at 51.5% ow Rated Power (Natural Circulation) O O O

Cs) J (~ fS1

                                                                                                                                                                      %j i EUffEDrit.m 1VESEL         Rif (P583 2 PDE FWI) CENTEM TDP                                                                  2 STM LIE        mlE (PSil 3 RW SLN1& E RT Fttu                   125*                                            S TUW!E 9T RISE (PSIl IE-                                           4 e t ttemTE7     T15T -                                                               4 c3CIEET         18f11/t31 5 vEsm. s1EAM FL N                                                                     5 CDE RE YSID F1NC til 8 TOBIE !

T{@t Fl@ til ie. 75. n a I a b a --

                                                                                                                              =              u               u I            I1                    3

3. __ -

ja

                                                               - ; _-1       e    e               25.
         -                                                                                            ,15                  fM              iM             t4
                                                                                                 ,75.-     t                                                          .      w

3. ....i.. . . -

0. 10. 20. 30. 40. O. 10. 20. 30. 40. w IIE ITCt TIE (WCl co

     "                                                                                                                                                                       =0 CM A                                                                                                                                                                       O Cn I                                                                                                                                                                     r tn O                                                                                                                                                                       M>

I LEVELtllEH-4ET-EP-!B(IRT I EUffED4 FLtX  % 2MKE ERT FLIR H 2 W R WNED LEvalINcmst ** 3 N R SENSED LEVEttlEJESI 320~ IE- 4 uur. TRITTLmi il 1 $ 5 DRIVE FLCWIII1 $

                                                                                                                                                                             =

c 1E3. E* b I a a en

3. f"F ^ ~

                                                                                                  %D.

u a u u

                                                                                               ~
                   - s                     s             s                    s I                                                      _a

0. : . . . . t - - t O. -

0. 10. 20. 30. 40. D. 25. so. 75. 100.

TIE IWCl CffE FlfD8 til N MN c>

                                                                                                                                                                              <w
                                                                                                                                                                              . o o

Figure 4.4-7c. 6-In. Water Level Setpoint Step at o -a 51.5% Rated Power (Natural Circulation)

22 20 - 18 - 16 - 14 - E a N O z 12 - w

                                                                                                                                                                                                             =

a B ro e cm I g to 2 m>

• j >N "s

H H 8 - ro

                                                                                                                                                -                                                            E
                                                                                                                                                                                                             =

U l 6 - i i 4 - r _ r i I  ! l I I I I l I I I I I I l o

o.o c.1 0.2 o.3 o.4 0.5 o.6 0.7 o.8 o.9 1.0 1.1 1.2 1.3 1.4 1.5 1.6 xb 0M REL%TIVE BUNDLE POWER 4y

                                                                                                                                                                                                             =   O o

Figure 4.4-S. Relative Bundle Power Histogram for Power Distribution Used in Statistical Analysis (Basis in Reference 1)

G G G

_ _ _ _ . - .- - .. ~. GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

     /~

SECTION 4.5 CONTENTS . Section Title Page

         ~4 . 5 REACTOR MATERIALS                                          4.5-1 4.5.1                  Control Rod System Structural Materials                           4.5-1 4.5.1.1                Material Specifications             4.5-1 4.5.1.2                Austenitic Stainless Steel Components                          4.5-4 l                4.5.1.3                Other Materials                     4.5-6 4.5.1.4                Cleaning and Cleanliness Control    4.5-6 4.5.1.4.1              Protection of Materials During Fabrication, Shipping and Storage   4.5-6 4.5.2                  Reactor Internal Materials          4.5-7 4.5.2.1                Material Specifications             4.5-7

, 4.5.2.2 Controls on Welding 4.5-10 ("N 4.5.2.3 Nondestructive Examination of

     \_s                               Wrought Seamless Tubular Products   4.5-11 4.5.2.4                Fabrication and Processing of Austenitic Stainless Steel -

Regulatory Guide Conformance 4.5-11 4.5.2.5 Other Materials 4.5-13 4.5.3 Control Rod Drive Housing Supports 4.5-15 I i O 4.5-i/4.5-li I

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.5 REACTOR MATERIALS 4.5.1 Control Rcd System Structural Materials 1 4.5.1.1 Material Specifications i

a. Material List _

The following material listing applies to the control rod drive mechanism supplied for this application. The position indicator ! and minor nonstructural items are omitted. (1) Cylinder, Tube and Flange Assembly Flange ASME SA182 Grade F304 Plugs ASME SA182 Grade F304 Cylinder ASTM A269 Grade TP 304 O Outer Tube ASTM A269 Grade TP 304 ASME SA351 Grade CF-3 Tube Spacer ASME SA351 Grade CF-3 (2) Piston Tube Assembly ASME SA479 or SA249 Grade XM-19 ~ Piston Tube i Nose ASME SA479 Grade XM-19

Base ASME SA479 Grade XM-19 Ind. Tube ASME SA312 Type 316 i Cap ASME SA182 Grade F316 (3) Drive Line Assembly Coupling Spud Alloy X-750 Compression Cylinder ASME SA479 or SA249 Grade XM-19 Index Tube ASME SA479 or SA249 Grade XM-19 O Piston !! cad ARMCO 17-4 PII or its equivalent 4.5-1 1

 --   - ~ . , ~ . - - -                     , , , , _ _ _ , _ _ , , ._ ,      _ _ _ _     __ _    _

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.5.1.1 Material Specifications (Continued) Piston Coupling ASTM A312 Grade TP 304 or ASTM A269 Grade TP 304 Magnet Housing ASTM A312 Grade TP 304 or ASTM A269 Grade TP 304 or ASTM A312, A249, or A213 TP 316L (4) Collet Assembly Collet Piston ASTM A269 TP 304 or ASTM A312 TP 304 Finger Alloy X-750 Retainer ASTM A269 TP 304 Guide Cap ASTM A269 TP 304 (5) Miscellaneous Parts Stop Piston ARMCO 17-4 PH or its equivalent 0-Ring Spacer ASTM A240 Type 304 Nut ASME SA479 Grade XM-19 Collet Spring Alloy X-750 Ring Flange ASME SA182 Grade F304 Buffer Shaft ARMCO 17-4 PH or its equivalent Buffer Piston ARMCO 17-4 PH or its equivalent Buffer Spring Alloy X-750 Nut (hex) Alloy X-750 The austenitic 300 series stainless steels listed under Ad.M/ASME specification number are all in the annealed condition (with the exception of the outer tube in the cylinder, tube and flange assem-bly), and their properties are readily available. The outer tube is approximately 1/8 hard, and has a tensile of 90,000/125,000 psi, yield of 50,000/85,000 psi and minimum elongation of 25%. J 4.5-2 l

                                     ._- .               -. _ -_ _=_ _- --                                     -               ____ - . _ .

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i 4.5.1.1 Material Specifications (Continued)

O The coupling spud, collet fingers, buffer spring, nut (hex), and collet spring are fabricated from Alloy X-750 in the annealed or I

equalized condition, and aged 20 hours at 1300*F to produce a' ten-sile of 165,000 psi minimum, yield of 105,000 psi minimum, and -l elongation of 20% minimum. The piston head, stop piston, buffer shaft, and buffer piston are ARMCO 17-4 PH (or its equivalent) in condition H-1100 (aged 4 hours at 1100*F) , with a tensile of. 140,000 psi minimum, yield of 115,000 psi minimum, and clongation j of 15% minimum. These are widely used materials, whose properties are well known. j The parts are readily accessible for inspection and replaceable if necessary. All materials, except SA479 or SA249 Grade XM-19, have been suc-cessfully used for the past 10 to 15 years in similar drive mech-O anisms. Extensive laboratory tests have demonstrated that ASME SA479 or SA249 Grade XM-19 are suitable materials and that they are resistant to stress corrosion in a BWR environment.

b. Special Materials No cold-worked austenitic stainless steels with a yield strength greater than 90,000 psi are employed in the Control Rod Drive (CRD) system. ARMCO 17-4 PH (or its equivalent) (martensitic precipita-I tion hardened stainless steel) is used for the piston head, stop l piston, buffer shaft, and buffer piston. This material is aged

! to the H-1100 condition to produce resistance to stress corrosion cracking in the BWR environments. ARMCO 17-4 PH (or its equiva-lent) (H-1100) has been successfully used for the past 10 to 15 years in BWR drive mechanisms. I O 4.5-3 i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.5.1.2 Austenitic Stainless Steel Components

a. Processes, Inspections and Tests Two special processes are employed which subject selected 300 Series stainless steel components to temperatures in the sensitization range:

(1) The cylinder and spacer (cylinder, tube and flange assembly) and the retainer (collet assembly) are hard surfaced with Colmonoy 6 (or its equivalent) . (2) The collet piston and guide cap (collet assembly) are nitrided to provide a wear-resistant surface. Nitriding is accomplished using a proprietary process. Components are exposed to a temperature of about 1080 F for about 20 hours during the nitriding cycle. Colmonoy (or its equivalent) hard-surfaced components have per-formed successfully for the past 10 to 15 years in drive mechanisms. Nitrided components have been used in CRDs since 1967. It is nor-mal practice to remove some CRDs at each refueling outage. At this time, both the Colmonoy (or its equivalent) hard-surfaced parts and nitrided surfaces are accessible for visual examination. In addi-tion, dye penetrant examinations have been performed on nitrided surfaces of the longest service drives. This inspection program is adequate to detect any incipient defects before they could become serious enough to cause oper". ting problems. Regulatory Guide 1.44 Discussion of the degree of conformance to Regulatory Guide 1.44 is provided in Subsection 4.5.2.4. O 4.5-4

GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 0 (} V 4.5.1.2 Austenitic Stainless Steel Components (Continued)

b. Control of Delta Ferrite Content Discussion of this subject and the degree of conformance to Regula-tory Guide 1.31 is presented in Subsection 4.5.2.4.

4.5.1.3 Other Materials These are discussed in Subsection 4.5.1.1.b. 4.5.1.4 Cleaning and Cleanliness Control 4.5.1.4.1 Protection of Materials During Fabrication, Shipping and Storage All the CRD parts listed above (Subsection 4.5.1.1) are fabricated under a process specification which limits contaminants in cutting, O grinding and tapping coolants and lubricants. It also restricts all other processing materials (marking inks, tape etc.) to those which are completely removable by the applied cleaning process. All contaminants are then required to be removed by the appropriate cleaning process prior to any of the following: (1) Any processing which increases part temperature above 200*F. (2) Assembly which results in decrease of accessibility for cleaning. (3) Release of parts for shipment. The specification for packaging and shipping the Control Rod Drive provides the following: O 4.5-5

1 l GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. O I I 4.5.1.4.1 Protection of Materials During Fabrication, Shipping and Storage (Continued) The drive is rinsed in hot deionized water and dried in preparation for shipment. The ends of the drive are then covered with a vapor tight barrier with dessicant. Packaging is designed to protect the drive and prevent damage to the vapor barrier. Audits have indi-cated satisfactory protection. Semiannual examination of the humidity indicators of ten percent of the units is required to verify that the units are dry and in satis-factory condition. This inspection shall be performed with a GE-Engineering designated representative present. Position indicator probes are not subject to this inspection. Site or warehouse storage specifications require inside heated storage comparable to level 3 of ANSI N45.2.2. The degree of surface cleanliness obtained by these procedures meets the requirements of Regulatory Guide 1.37. Regulatory Guide 1.37 General Compliance or Alternate Approach Assessment: For Commit-ment and Revision Number, see Section 1.8. 4.5.2 Reactor Internal Materials l 4.5.2.1 Material Specifications Materials used for the Core Support Structure: Shroud Support - Nickel-Chrome-Iron-Alloy, ASME SB166 or SB168. O l 4.5-6

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

  ,_3     4.5.2.1   Material Specifications (Continued)
 /      \

b Shroud, core plate, and grid - ASME 3A240, SA182, SA479, SA312, SA249, or SA213 (all Type 304L) . Peripheral fuel supports - ASTM A312 Grade TP-304, A479 Type-316L, ASME SA311 Grade Type-304L Core plate and top guide studs and nuts, and core plate wedges - ASME SA479, SA193 Grade B8A, SA194 Grade 8A (all Type-304) Control rod drive housing - ASME SA312 TP-304, SA182 Type-304, and ASME SB167 Type Alloy 600. Control rod guide tube - ASME SA358 Grade 304, SA312 Grade TP-304; ASTM A358 Grade 304, A312 Grade TP-304, A351 Grade CF8, A249 TP-304. 7-s b Orificed fuel support - ASTM A249 TP-304, A240 TP-316L, A479 TP-316L. Materials Employed in Other Reactor Internal Structures. (1) Shroud llead and Separators Assembly and Steam Dryer Assembly All materials are TP-304, 304L or 316L stainless steel. Plate, Sheet and Strip ASTM A240, TP-304, 304L or 316L Forgings ASTM A182 Grade F304 or 304L Bars ASTM A276 TP-304 or 316L

     -s              Pipe                       ASTM A312 Grade TP-304 t
  'v' 4.5-7

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.5.2.1 Material Specifications (Continued) Tube ASTM A269 Grade TP-304 O Castings ASTM A351 Grade CF8 (2) Jet Pump Assemblies The components in the Jet Pump Assemblies are a Riser, Inlet Mixer, Diffuser, and Riser Brace. Materials used for these components are to the following specifications: Castings ASTM A351 Grade CF8 and ASTM SA351 Grade CF3 Bars ASTM A276 TP-304, ASTM A479 TP-316L ASTM A637 Grade 688 O Bolts ASTM A193 Grade B8 or B8M and ASME SA479 TP-316L Sheet and Plate ASTM A240 TP-304, and ASME SA240 TP-304L, 316L Pipe AFTM A358 TP-304, 316L and ASME SA312 Grade TP-304, 316L Forged or Rolled Parts ASME SA182, Grade F304, F316L, ASTM B166, and ASTM A637 Grade 688. O 4.5-6

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

, 4.5.2.1   Material Specifications (Continued)

Materials in the Jet Pump Assemblies which are not austenitic stainless steel are listed below:

a. The Inlet Mixer Adaptor casting, the wedge casting, bracket casting adjusting screw casting, and the Diffuser collar casting are hard surfaced with Stellite 6 (or its equivalent) for slip fit joints.

b. The Diffuser is a bimetallic component made by welding an austenitic stainless steel ring to a forged Alloy 600 ring, made to Specification ASTM B166.

c. The Inlet-Mixer contains a pin, insert, and beam made of Alloy X-750 to Specification ASTM A637 Grade 688.

All core support structures are fabricated from ASME specified O' materials, and designed in accordance with requirements of ASME Code, Section III, Subsection NG. The other reactor internals are noncoded, and they are fabricated from ASTM or ASME specification materials. Material requirements in the ASTM specifications are identical to requirements in corresponding ASME material specifications. 4.5.2.2 Controls on Welding Core support structures are fabricated in accordance with require-ments of ASME Code Section III, Subsection NG. Other internals are not required to meet ASME Code requirements. Requirements of ASME Section IX BPV Code, are followed in fabrication of core support structures. O 4.5-9

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.5.2.3 Nondestructive Examination of Wrought Seamless Tubular Products Wrought seamless tubular products for CRD housings, and peripheral fuel supports, were supplied in accordance with ASME Section III, Class CS, which requires examination of the tubular products by radiographic and/or ultrasonic methods according to para-graph NG-2550. Wrought seamless tubular products for other internals were supplied in accordance with the applicable ASTM or ASME material specifica-tions. These specifications require a hydrostatic test on each length of tubing. 4.5.2.4 Fabrication and Processing of Austenitic Stainless Steel - Regulatory Guide Conformance Regulatory Guide 1.31: Control of Stainless Steel Welding Cold-worked stainless steels are not used in the reactor internals O except for vanes in the steam dryers. The delta ferrite content for weld materials used in welding austenitic stainless steel assem-blies is verified on undiluted weld deposits for each heat or lot of filler metal and electrodes. The delta ferrite content is defined for weld materials as 5.0 FN minimum and 8.0 FN average (Ferrite Number). This ferrite content is considered adequate to prevent any micro-fissuring (Hot Cracking' in austenitic stainless steel welds. This procedure complies with the requirements of Regulatory Guide 1.31. Regulatory Guide 1.44: Control of the Use of Sensitized Stainless Steel Proper solution annealing of the 300 series austenitic stainless steel is verified by testing per ASTM-A262, " Recommended Practices for Detecting Susceptibility to Intergranular Attack in Stainless 4.5-10

GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. O O 4.5.2.4 Fabrication and Processing of Austenitic Stainless Steel - Regulatory Guide Conformance (Continued) Steels." Welding of austenitic stainless steel parts is performed in accordance with Section IX (Welding and Brazing Qualification) and Section II Part C (Welding Rod Electrode and Filler Metals) of the ASME Boiler and Pressure Vessel Code. Welded austenitic stain-less steel assemblies require solution annealing to minimize the possibility of the sensiti;ing. Ilowever, welded assemblies are dispensed from this requirement when there is' documentation that welds are not subject to significant sustained loads and assemblies have been free of service failure. Other reasons, in line with the regulatory guide, for dispensing with the solution annealing are that assemblies are exposed to reactor coolant during normal operation service which is below 200'F temperature or assemblies are of material of low carbon content (less than 0.025%). These controls are employed in order to comply with the intent of the fg Regulatory Guide 1.44. ' \_sl Regulatory Guide 1.37: Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants - Exposure to contaminant is avoided by carefully controlling all cleaning and processing materials which contact stainless steel during manufacture and construction. Any inadvertent surface contamination is removed to avoid potential detrimental effects. Special care is exercised to insure removal of surface contami-nants prior to any heating operation. Water quality for rinsing, flushing, and testing is controlled and monitored. The degree of cleanliness obtained by these procedures meets the requirements of Regulatory Guide 1.37. O 4.5-11

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.5.2.4 Fabrication and Processing of Austenitic Stainless Steel - Regulatory Guide Conformance (Continued) Regulatory Guides 1.31, 1.44, and 1.37 General Compliance or Alternate Approach Assessment: For Commit-ment and Revision Number, see Section 1.8. 4.5.2.5 Other Materials liardenable martensitic stainless steel and precipitation hardening stainless steels are not used in the reactor internals. Materials, other than Type-300 stainless steel, employed in reactor internals are: (1) SA479 Type XM-19 stainless steel; (2) SB166, 167, and 168, Nickel-Chrome-Iron (Alloy 600); and (3) SA637 Grade 688 Alloy X-750. Alloy 600 tubing plate, and sheet are used in the annealed condi-tion. Bar may be in the annealed or cold-drawn condition. Alloy X-750 components are fabricated in the annealed or equalized condition and aged when required. Stellite 6 (or its equivalent) hard surfacing is applied to some austenitic stainless steel castings using the gas tungsten arc welding or plasma are surfacing processes. All materials, except SA 479 Grade XM-19, have been successfully used for the past 10 to 15 years in BWR applications. Extensive laboratory tests have demonstrated that XM-19 is a suitable material and that it is resistant to stress corrosion in a BWR environment. 4.5-12

i GESSAR II 22A7007 4 238 NUCLEAR ISLAND Rev. O i 4.5.3 Control' Rod Drive Housing Supports All CRD housing support subassemblies are fabricated of ASTM-A-36 i structural steel, except for the following items:

 !                                                                                                                               Material 7

Grid ASTM A441 i

l Disc springs Schnorr Type BS-125-71-8 (or its equivalent) l Ilex bolts and nuts ASTM A307 f I 6 in. x 1 in. x 3/8 in. ASTM A500 Grade B  ! l tubes . 2 For further CRD housing support information, see Subsection 4.6.1.2. i f I i h l i t i l. i l i l 1 i l 4.5-13/4.5-14 l }

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i C\ SECTION 4.6

,  'Q CONTENTS Section                           Title                       Page 4.6   FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS         4.6-1 4.6.1               Information for Control Rod Drive System (CRDS)                       4.6-1 4.6.1.1             Control Rod Drive System Design    -4.6-1 i

4.6.1.1.1 Design Bases 4.6-1 4.6.1.1.1.1 General Design.Dases 4.6-1 4.6.1.1.1.1.1 Safety Design Bases 4.6-1 4.6.1.1.1.1.2 Power Generation Design Basis 4.6-2 4.6.1.1.2 Description 4.6-2 4.6.1.1.2.1 Control Rod Drive Mechanisms 4.6-3 4.6.1.1.2.2 Drive Components 4.6-4 4.6.1.1.2.2.1 Drive Piston 4.6-5 4.6.1.1.2.2.2 Index Tube 4.6-5 () 4.6.1.1.2.2.3 4.6.1.1.2.2.4 Collet Assembly Piston Tube 4.6-6 4.G-7 4.6.1.1.2.2.5 Stop Piston 4.6-7 4.6.1.1.2.2.6 Flange and Cylinder Assembly 4.6-8 4.6.1.1.2.2.7 Uncoupling Rod and Related Parts 4.6-9 4.6.1.1.2.3 Materials of Construction 4.5-9 4.6.1.1.2.3.1 Index Tube 4.6-10 4.6.1.1.2.3.2 Coupling Spud 4.6-10 1 4.6.1.1.2.'3.3 Collet Fingers 4.6-10 4.6.1.1.2.3.4 Seals and Bushings 4.6-10 , 4.6.1.1/2.3.5 Summary 4.6-11 4.6.1.1.2.4 Control Rod Drive Hydraulic System 4.6-12 ( 4.6.1.1.2.4.1 Hydraulic Requirements 4.6-12 4.6.1.1.2.4.2 System Description 4.6-13

             ,4.6.1.1.2.4.2.1    Supply Pump                         4.6-13 4.6.1.1.2.4.2.2     Accumulator Charging Pressure       4.6-14 l             4.6.1.1.2.4.2.3     Drive Water Pre; are                4.6-15

() ' 4.6.1.1.2.4.2.4 4.6.1.1.2.4.2.5 Cooling Water Header Scram Discharge Volume 4.6-15 4.6-16

   /

4.6-i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CONTENTS (Continued) Section Title Page , 4.6-17 ' 4.6.1.1.2.4.3 Hydraulic Control Units 4.6.1.1.2.4.3.1 Insert Drive Valve 4.6-18 4.6:1.1.2.4.3.2 Insert Exhaust Valve 4.6-18 4.6.1.1.2.4.3.3 Withdraw Drive Valve 4.6-18 4.6.1.1.2.4.3.4 Withdraw Exhaust Valve 4.6-18 4.6.1.1.2.4.3.5 Speed}'ontrolUnits 4.6-19 4.6.1.1.2.4.3.6 Scram Pilot Valve Assembly 4.6-19 4.6.1.1.2.4.3.7 Scram Inlet Valve 4.6-19 4.6.1.1.2.4.3.8 Scram Exhaust Valve 4.6-20 4.6.1.1.2.4.3.9 Scram Accumulator 4.6-20 4.6.1.1.2.5 Control Rod Drive System Operation 4.6-20 4.6.1.1.2.5.1 Rod Insertion 4.6-20 4.6.1.1.2.5.2 Rod Withdrawal 4.6-21 4.6.1.1.2.5.3 Scram 4.6-22 4.6.1.1.2.6 Instrumentation 4.6-23 4.6.1.2 Control Rod Drive Housing Supports 4.6-24 4.6.1.2.1 Safety Objective 4.6-24 4.6.1.2.2 Safety Design Bases 4.6-24 4.6.1.2.3 Description 4.6-24 4.6.2 Evaluations of the CROS 4.6-26 4.6.2.1 Failure Mode and Effects Analysis 4.6-26 4.6.2.2 Protection from Common Mode Failures 4.6-26 4.6.2.3 Safety Evaluation 4.6-27 1 4.6.2.3.1 Control Rods 4.6-27 4.6.2.3.1.1 Materials Adequacy Throughout Design Lifetime 4.6-27 4.6.2.3.1.2 Dimensional and Tolerance Analysis 4.6-27 4.6.2.3.1.3 Thermal Analysis of the Tendency to Warp 4.6-27 4.6.2.3.1.4 Forces for Expulsion 4.6-28 4.6.2.3.1.5 Punctional Failure of Critical Components 4.6-28 4.6-ii

I > si GESSAR II 22A7007 i 238 NUCLEAR ISLAND Rev. 0 () CONTENTS (Continued) Section Title. Page 4.6.2.3.1.6 Precluding Excessive Rates of Reactivity Addition 4.6-28 4.6.2.3.1.7 Effect of Rod Failure on Control j Rod Channel clearances 4.6-28 Mechanical Damage 4.6.2.3.1.8 4.6-28 4 G.2.3.1.9 Evaluation of Control Rod Velocity Limiter 4.6-29 4.6.2.3.2 Control Rod Drives 4.6-29

                                             .4.6.2.3.2.1      Evaluation of Scram Time             4.6-29 4.6.2.3.2.2      Analysis of Malfunction Relating to Rod Withdrawal                    4.6-29 4.6.2.3.2.2.1    Drive Housing Fails at Attachment Weld                                 4.6-30 4.6.2.3.2.2.2    Rupture of Hydraulic Line(s) to Drive Housing Plange                 4.6-31 4.6.2.3.2.2.2.1  Pressure-under (Insert) Line Break   4.6-31

[ 'l

        ! S~)

4.6.2.3.2.2.2.2 Pressure-over (Withdrawn) Line Break 4.6-32

         '                                    4.6.2.3.2.2.2.3  Simultaneous Breakage of the
                 )#                                            Pressure-Over (Withdrawn) and i              /                                                Pressure-Under (Insert) Lines        4.6-33

t. 4.6.2.3.2.2.3 All Drive Flange Bolts Fail in Tension 4.6-34 i 4.6.2.3.2.2.4 Wold Joining Flange to Housing Pails in Tension 4.6-35 i

! i 4.6.2.3.2.2.5 Housing Wall Ruptures 4.6-36 4.6.2.3.2.2.6 Flange Plug Blows Out 4.6-38 4.6.2.3.2.2.7 Ball Check Valve Plug Blows Out 4.6-39 l 4.6.2.3.2.2.8 Drive / Cooling Water Pressure Control Valve Closure (Reactor Pressure, 0 psig) 4.6-40

                                . , -         4.6.2.3.2.2.9    Ball Check Valve Fails to Close Passage to Vessel Ports              4.6-40 4.6.2.3.2.2.10   Hydraulic Control Unit Valve Failures                             4.6-41 4.6.2.3.2.2.11   Collet Fingers Pail to Latch         4.6-41 l

) , 4.6.2.3.2.2.12 Withdrawal Speed Control Valve Failure 4.6-42 f 4.6-iii 6 0

GESSAR II 22R7007 238 NUCLEAR ISLAND Rev. O CONTENTS (Ccntinued) Section Title Page 4.6.2.3.2.3 Scram Feliability 4.6-42 4.6.2.3.2.4 Control Rod Support and Operation 4.6-43 4.6.2.3.3 Control Rod flousing Supports 4.6-43 4.6.3 Testing and Verification of the CRDs 4.6-44 4.6.3.1 Control Rod Drives 4.6-44 4.6.3.1.1 Testing and Inspection 4.6-44 4.6.3.1.1.1 Development Tests 4.6-44 4.6.3.1.1.2 Factory Quality Control Tests 4.6-45 4.6.3.1.1.3 Operational Tests 4.6-46 4.6.3.1.1.4 Acceptance Tests 4.6-47 4.6.3.1.1.5 Surveillance Tests 4.6-48 4.6.3.1.1.6 Functional Tests 4.6-50 4.6.3.2 Control Rod Drive licusing Supports 4.6-51 4.6.3.2.1 Testing and Inspection 4.6-51 4.6.4 Information for Combined Performance of Reactivity Control Systems 4.6-52 4.6.4.1 Vulnerability to Common Mode Failures (AE) 4.6-52 4.6.4.2 Accidents Taking Credit for Multiple Reactivity Systems 4.6-52 4.6.5 Evaluation of Combined Performance 4.6-52 4.6.6 References 4.6-52 0 4.6-iv

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O SECTION 4.6 ILLUSTRATIONS Figure Title Page 4.6-1 Control Rod to* Control Rod Drive Coupling 4.6-53 4.6-2 Control Rod Drive Unit 4.6-54 4 4.6-3 Control Rod Drive Schematic 4.6-55 4.6-4 Control Rod Drive Unit (Cutaway) 4.6-56 4.6-Sa Control Rod Drive Ilydraulic System P&ID 4.6-57 4.6-5b Control Rod Drive liydraulic System P&ID 4.6-58 4.6-Sc Control Rod Drive Ilydraulic System P&ID 4.6-59 4.6-6 Control Rod Drive System Process Diagram and Data 4.6-61 4.6-7 Control Rod Drive liydraulic Control Unit 4.6-63 ! 4.6-8 Control Rod Drive llousing Support 4.6-64 4 I. l t l l O l 4.6-v/4.6-vi . 1

1 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

,    4.6   FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS 4

i

-    The reactivity control systems consist of control rods and control a

rod drives, supplementary reactivity control in the form of a burnable poison (Section 4.3), and the Standby Liquid Control System (described in Subsection 9.3.5). 4.6.1 Information for Control Rod Drive System (CRDS) 4.6.1.1 Control Rod Drive System Design i 4.6.1.1.1 Design Bases 4.6.1.1.1.1 General Design Bases 4.6.1.1.1.1.1 Safety Design Bases The CRD mechanical system shall meet the following safety design ! O bases: l (1) The design shall provide for a sufficiently rapid control rod insertion such that no fuel damage results from any moderately frequent event (see Chapter 15). } I (2) The design shall include positioning devices, each of which individually supports and positions a control rod. ) (3) Each positioning device shall:

a. prevent its control rod from initiating withdrawal as a result of a single malfunction; I

l O 4.6-1

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.1.1.1.1.1 Safety Design Bases (Continued)

b. be individually operated so that a failure in one positioning device does not affect the operation of any other positioning device; and

c. be individually energized when rapid control rod insertion (scram) is signaled so that failure of power sources external to the positioning device does not prevent other positioning devices' control rods from being inserted.

4.6.1.1.1.1.2 Power Generation Design Basis The control rod system drive design shall provide for positioning the control rods to control power generation in the core. 4.6.1.1.2 Description O The Control Rod Drive System (CRDS) controls gross changes in core reactivity by incrementally positioning neutroc absorbing control rods within the reactor core in response to manual control signals. It is also required to quickly shut down the reactor (scram) in emergency situations by rapidly inserting all control rods into the core in response to a manual or automatic signal from the Reactor Protection Trip System. The CRDS consists of locking pis-ton CRD mechanisms, and the CRD hydraulic system (including power supply and regulation, hydraulic control units, interconnecting piping, instrumentation and electrical controls). O 4.6-2

GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 4.6.1.1.2.1 Control Rod Drive Mechanisms The CRD mechanism used for positioning the control rod in the reactor core is a double-acting, mechanically latched, hydraulic cylinder using water as its operating fluid (Figures 4.6-1, 4.6-2, 4.6-3, and 4.6-4). The individual drives are mounted on the bottom head of the reactor pressure vessel. The drives do not interfere with refueling and are operative even when the head is removed from the reactor vessel. The drives are also readily accessible for inspection and servicing. The bottom location makes maximum utilization of the water in the reactor as a neutron shield and gives the least possible neutron exposure to the drive components. Using water from the condensate treatment system, and/or condensate storage tanks as the operating fluid eliminates the need for special hydraulic fluid. Drives are able to utilize simple piston seals whose leakage does not con-taminate the reactor water but provides cooling for the drive ( mechanisms and their seals. The drives are capable of inserting or withdrawing a control rod at a slow, controlled rate, as well as providing rapid insertion when required. A mechanism on the drive locks the control rod at 6-in. increments of stroke over the length of the core. A coupling spud at the top end of the drive index tube (piston l rod) engages and locks into a mating socket at the base of the t control rod. The weight of the control rod is sufficient to I engage and lock this coupling. Once locked, the drive and rod form an integral unit that must be manually unlocked by specific procedures before the components can be separated. O i 4.6-3 i

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.1.1.2.1 Control Rod Drive Mechanisms (Continued) The drive holds its control rod in distinct latch positions until O the hydraulic system actuates movement to a new position. With-drawal of each rod is limited by the seating of the rod in its guide tube. Withdrawal beyond this position to the overtravel limit can be accomplished only if the rod and drive are uncoupled. Withdrawal to the over-travel limit is annunciated by an alarm. The individual rod indicators, grouped in one control panel display, correspond to rol tive rod locations in the core. The Display Module is divided sto two sections. There is a Display section and a Rod Select soution, which are physically superimposed but independent of each other. For display purposes, the control rods are considered in groups of four adjacent rods centered around a common core volume. Each group is monitored by four LPRM strings (Subsection 7.6.2.1, Neutron Monitoring System). Rod groups at the periphery of the core may have less than four rods. A white light indicates which of the four rods is the one selected for movement. 4.6.1.1.2.2 Drive Components Figure 4.6-2 illustrates the operating principle of a drive. Figures 4.6-3 and 4.6-4 illustrate the drive in more detail. The main components of the drive and their functions are described below. O 4.6-4

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.1.1.2.2.1 Drive Piston O The drive piston is mounted at the lower end of the index tube. { The function of the index tube is similar to that of a piston rod 4 in a conventional hydraulic cylinder. The drive piston and index tube make up the main moving assembly.in the drive. The drive

 !                piston operates between positive end stops, with a hydraulic cushion provided at the upper end only. The piston has both inside and outside seal rings and operates in an annular space between an inner cylinder (fixed piston tube) and an outer cylinder (drive cylinder).                  Because the type of inner seal used is effective in only one direction, the lower sets of seal rings are mounted with one set sealing in each direction.

A pair of nonmetallic bushings prevents metal-to-metal contact i between the piston assembly and the inner cylinder surface. The outer piston rings are segmented, step-cut seals with expander springs holding the segments against the cylinder wall. A pair

     )            of split bushings on the outside of the piston prevents piston contact with the cylinder wall. The effective piston. area for 1                  downtravel, or withdrawal, is approximately 1.2 in.2 versus 4.1 in.             for uptravel, or insertion.                                     This difference in driving area tends to balance the control rod weight and assures a higher force for insertion than for withdrawal.

4.6.1.1.2.2.2 Index Tube The index tube is a long hollow shaft made of nitrided stainless steel. Circumferential locking grooves, spaced every 6 in. along i the outer surface, transmit the weight of the control rod to the i 4 collet assembly. The upper end of the index tube is threaded to receive a coupling spud. The coupling (Figure 4.6-1) accommodates a small amount of angular misalignment between the drive and the control rod. 4.6-5 _ _ . . . . - - _ . - . . _ - .~ .___. _ _ . _ _ _ . . _ _ _ - - _ - . _ _., -

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.1.1.2.2.2 Index Tube (Continued) Six spring fingers allow the coupling spud to enter the mating socket on the control rod. A plug then enters the spud and prevents uncoupling. 4.6.1.1.2.2.3 Collet Assembly The collet assembly serves as the index tube locking mechanism. It is located in the upper part of the drive unit. This assembly prevents the index tube from accidentally moving downward. The assembly consists of the collet fingers, a return spring, a guide cap, a collet housing (part of the cylinder, tube, and flange) and the collet piston. Locking is accomplished by fingers mounted on the collet piston at the top of the drive cylinder. In Lne locked or latched position the fingers engage a locking groove in the index tube. The collet piston is normally held in the latched position by a , force of approximately 150 lb supplied by a spring. Metal piston rings are used to seal the collet piston from reactor vessel pressure. The collet assembly will not unlatch until the collet fingers are unloaded by a short, automatically sequenced, drive-in l signal. A pressure, approximately 180 psi above reactor vessel pressure, must then be applied to the collet piston to overcome spring force, slide the collet up against the conical surface in the guide cap, and spread the fingers out so they do not engace a i locking groove. A guide cap is fixt3 in the upper end of the drive assembly. This member provides the unlocking cam surface for the collet fingers and serves as the upper bushing for the index tube. If reactor water is used during a scram to supplement accumulator l pressure, it is drawn through a filter on the guide cap. 4.6-6

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.1.1.2.2.4 Piston Tube The piston tube is an inner cylinder, or column, extending upward inside the drive piston and index tube. The piston tube is fixed to the bottom flange of the drive and remains stationary. Water is brought to the upper side of the drive piston through this tube. A buffer shaft, at the upper end of the piston tube, supports the stop piston and buffer components. 4.6.1.1.2.2.5 Stop Piston A stationary piston, called the stop piston, is mounted on the upper end of the piston tube. This piston provides the seal between reactor vessel pressure and the space above the drive piston. It also functions as a positive end stop at the upper limit of con-trol rod travel. Piston rings and bushings, similar to those on the drive piston, are mounted on the upper portion of the stop piston. The lower portion of the stop piston form a thinwalled cylinder O' containing the buffer piston, its metal seal ring, and the buffer piston return spring. As the drive piston reaches the upper end of the scram stroke, it strikes the buffer piston. A series of ori-fices in the buffer shaft provides a progressive water shutoff to cushion the buffer piston as it is driven to its limit of travel. The high pressures generated in the buffer are confined to the cylinder portion of the stop piston and are not applied to the stop piston and drive piston seals. The center tube of the drive mechanism forms a well to contain the ! position indicator probe. The probe is an aluminum extrusion j attached to a cast aluminum housing. flounted on the extrusion are f hermetically sealed, magnetically operated, reed switches. The ! entire probe assembly is protected by a thin-walled stainless steel tube. The switches are actuated by a ring magnet located at the I bottom of the drive piston. !O l 4.6-7 I

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.1.1.2.2.5 Stop Piston (Continued) The drive piston, piston tube and indicator tube are all of nonmagnetic stainless steel, allowing the individual switches to be operated by the magnet as the piston passes. Two switches are located at each position corresponding to an index tube grocve, thus allowing redundant indication at each latching point. Two additional switches are located at each midpoint between latching points to indicate the intermediate positions during drive motion. Thus, indication is provided for each 3 in. of travel. Duplicate switches are provided for the full-in and full-out positions. Redundant overtravel switches are located at a position below the normal full-out position. Because the limit of downtravel is nor-mally provided by the control rod itself as it reaches the backseat position, the drive can pass this position and actuate the over-travel switches only if it is uncoupled from its control rod. A convenient means is thus provided to verify that the drive and control rod are coupled after installation of a drive or at any time during plant operation. 4.6.1.1.2.2.6 Flange and Cylinder Assembly A flange and cylinder :ssembly is made up of a heavy flange welded to the drive cylinder. A sealing surface on the upper face of this flange forms the seal to the drive housing flange. The seals contain reactor pressure and the two hydraulic control pres-sures. Teflon (or its equivalent)-coated, stainless steel rings are used for these seals. The dri"e flange contains the integral ball, or two-way, check (ball-shuttle) valve. This valve directc cither the reactor vessel pressure or the driving pressure, which-ever is higher, to the underside of the drive piston. Reactor ves-sel pressure is admitted to this valve from the annular space between the drive and drive housing through passages in the flange. O 4.6-8

GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 0 l l () 4.6.1.1.2.2.6 Flange and Cylinder Assembly (Continued) Water used to operate the collet piston passes between the outer tube and the cylinder tube. The inside of the cylinder tube is

honed to provide the surface required for the drive piston seals.

4

Both the cylinder tube and outer tube are welded to the drive flange. The upper ends of these tubes have a sliding fit to allow for differential expansion.

4.6.1.1.2.2.7 Uncoupling Rod and Related Parts Two means of uncoupling are provided. With the reactor vessel head removed, the lock plug can be raised against the spring force

'             of approximately 50 lb by a rod extending up through the center of the control rod to an unlocking handle located above the control rod velocity limiter.          The control rod, with the lock plug raised,

() can then be lifted from the drive. If it is desired to uncouple a drive without removing the reactor pressure vessel head for access, the lock plug can also be pushed i up from below. In this case, the piston tube assembly is pushed up against the uncoupling rod, which raises the lock plug and allows the coupling spud to disengage the socket as the drive piston and index tube are driven down. 4 The control rod is heavy enough to force the spud fingers to enter the socket and push the lock plug up, allowing the spud to enter the socket completely and the plug to snap back into place. There-fore, the drive can be coupled to the control rod using only the weight of the control rod. 4.6.1.1.2.3 Materials of Construction 4 () Factors that determine the choice of construction materials are discussed in the following subsections. 4.6-9

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.1.1.2.3.1 Index Tube The index tube must withstand the locking and unlocking action of the coliet fingers. A compatible bearing combination must be pro-vided that is able to withstand moderate misalignment forces. Large tensile and column loads are applied during scram. The reactor environment limits the choice of materials suitable for corrosion resistance. To meet these varied requirements, the index tube is made from annealed, single-phase, nitrogen strengthened, austenitic stainless steel. The wear and bearing requirements are provided by nitriding the complete tube. To obtain suitable corro-sion resistance, a carefully controlled process of surface prepara-tion is employed. 4.6.1.1.2.3.2 Coupling Spud The coupling spud is made of Alloy X-750 that is aged for maximum physical strength and the required corrosion resistance. Because misalignment tends to cause chafing in the semispherical contact area, the part is protected by a thin chromium plating. This plating also prevents galling of the threads attaching the coupling spud to the index tube. 4.6.1.1.2.3.3 Collet Fingers Alloy X-750 is used for the collet fingers, which must function as leaf springs when cammed open to the unlocked position. Col-monoy 6 (or its equivalent) hard facing provides a long wearing surface, adequate for design life, to the area contacting the index tube and unlocking cam surface of the guide cap. 4.6.1.1.2.3.4 Seals and Bushings Graphitar-14 (or its equivalent) is selected for seals and bush-ings on the drive piston and stop piston. The material is inert and has a low friction coefficient when water-lubricated. Because 4.6-10

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1 () 4.6.1.1.2.3.4 Seals and Bushings (Continued) some loss of Graphitar strength is experienced at higher tem-peratures, the drive is supplied with cooling water to hold temperatures below 250 F. The Graphitar-14 (or its equivalent) is relatively soft, which is advantageous when an occasional particle of foreign matter reaches a seal. The resulting scratches in the seal reduce sealing efficiency until worn smooth, but the drive design can tolerate considerable water leakage past the seals into the reactor vessel. 4.6.1.1.2.3.5 Summary All drive components exposed to reactor vessel water are made of austenitic stainless steel except the following: (1) Seals and bushings on the drive piston and stop piston () are Graphitar-14 (or its equivalent) (2) All springs and members requiring spring action (collet fingers, coupling spud, and spring washers) are made of Alloy X-750. (3) The ball check valve is a 11aynes Stellite cobalt-base alloy (or its equivalent). 4 (4) Elastomeric O-ring seals are ethylene propylene. I (5) Metal piston rings are Ilaynes 25 alloy (or its equivalent) . (6) Certain wear surfaces are hard-faced with Colmonoy 6 (or its equivalent). O 4.6-11

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.1.L.2.3.5 Summary (Continued) (7) Nitriding (or its equivalent) and chromium plating are used in certain areas where resistance to abrasion is necessary. (8) The drive piston head, stop piston, buffer shaft and buffer piston are made of Armco 17-4 PH (or its equivalent). (9) Certain fasteners and locking devices are made of Alloy X-750 or 600. Pressure-containing portions of the drives are designed and fabricated in accordance with requirements of Section III of the ASME Boiler and Pressure Vessel Code. 4.6.1.1.2.4 Control Rod Drive Hydraulic System The CRD hydraulic system (Figures 4.6-5a, b) supplies and controls the pressure and flow to and from the drives through hydraulic con-trol units (HCU). The water discharged from the drives during a scram flows through the HCUs to the scram discharge volume. The water discharged from a drive during a normal control rod position-ing operation flows through the HCU, the exhaust header, and is returned to the reactor vessel via the HCUs of nonmoving drives. There are as many HCUs as the number of control rod drives. 4.6.1.1.2.4.1 Hydraulic Requirements The CRD hydraulic system design is shown in Figures 4.6-Sa, b, and c, and 4.6-6. The hydraulic requirements, identified by the function they perform are as follows: (1) An accumulator hydraulic charging pressure of approxi- l mately 1750 to 2000 psig is required. Flow to the 4.6-12

GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 0 4.6.1.1.2.4.1 Ilydraulic Requirements (Continued) ( 1 accumulators is required only during scram reset or system startup. (2) Drive pressure of approximately 260 psi above reactor vessel pressure is required. A flow rate of approxi-

                              -mately 4 gpm to insert each control rod and 2 gpm to withdraw each control rod is required.

(3) Cooling water to the drives is required at approximately 20 psi above reactor-vessel pressure and at a flow rate of approximately 0.34 gpm per drive unit. (4) The scram discharge volume is sized to receive, and contain, all the water discharged by the drives during a scram; a minimum volume of 3.34 gal per drive is () required (excluding the instrument volume) . 4.6.1.1.2.4.2 System Description The CRD hydraulic system provides the required functions with the pumps, filters, valves, instrumentation and piping shown in Fig-urcs 4.6-Sa, b, and c, and described in the following paragraphs. Duplicato components are included, where necessary, to assure continuous system operation if an in-service component requires [ maintenance. 4.6.1.1.2.4.2.1 Supply Pump

One supply pump pressurizes the system with water from the conden-l sate treatment system and/or condensate storage tanks. One spare pump is provided for standby. A discharge check valve prevento

() backflow through the nonoperating pump. A portion of the pump

!                                                        4.6-13

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.1.1.2.4.2.1 Supply Pump (Continued) h discharge flow is diverted through a minimum flow bypass line to the condensate storage tank. This flow is controlled by an orifice and is sufficient to prevent pump damage if the pump discharge is inadvertently closed. Condensate water is processed by two filters in the system. The pump suction filter is a disposable element type with a 25-micron absolute rating. A 250-micron strainer in the filter bypass line protects the pump when the filters are being serviced. The drive water filter, downstream of the pump, is a cleanable element type with a 50-micron absolute rating. A differential pressure indi-cator and control room alarm monitor the filter element as it collects foreign materials. 4.6.1.1.2.4.2.2 Accumulator Charging Pressure Accumulator charging pressure is established by procharging the O nitrogen accumulator to a precisely controlled pressure at known temperature. During scram, the scram inlet (and outlet) valves open and permit the stored energy in the accumulators to discharge into the drives. The resulting pressure decrease in the charging water header allows the CRD supply pump to "run out" (i.e., flow rate to increase substantially) into the control rod drives via the charging water header. The flow element upstream of the accumulator charging header senses high flow and provides a signal to the manual auto-flow control station which in turn closes the system flow control valve. This action maintains increased flow through the charging water header, while avoiding prolonged pump operation at "run-out" conditions. Pressure in the charging header is monitored in the control room with a pressure indicator and low pressure alarm. O 4.6-14

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O I

         ~
  \ 4.6.l.1.2.4.2.2   Accumulator Charging Pressure (Continued) d During normal operation, the flow control valve maintains a con-stant system flow rate. This flow is used for drive flow and drive cooling.

4.6.1.1.2.4.2.3 Drive Water Pressure Drive water pressure required in the drive header is maintained by the drive pressure control valve, which is manually adjusted from the control room. A flow rate of approximately 16 gpm (the sum of the flow rate required to insert 4 control rods) normally passes from the drive water pressure stage through eight solenoid-operated stabilizing valves (arranged in parallel) into the cooling water header. The flow through two stabilizing valves equals the drive insert flow for one drive; that of one stabilizing valve equals the drive withdrawal flow for one drive. When operating a () drive (s), the required flow is diverted to the drives by closing the appropriate stabilizing valves, at the same' time opening the drive directi.onal control and exhaust solenoid valves. Thus, flow through Lne drive pressure control valve is always constant. Flow indicators in the drive water header and in the line down-stream from the stabilizing valves allow the flow rate through the stabilizing valves to be adjusted when necessary. Differential pressure between the reactor vessel and the drive pressure stage is indicated in the control room. 4.6.1.1.2.4.2.4 Cooling Water Header The cooling water header is located downstream from the drive / cooling pressure valve. The drive / cooling pressure control valve is manually adjusted from the control room to produce the required drive / cooling water pressure balance. ( 4.6-15

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.1.1.2.4.2.4 Cooling Water Header (Continued) The flow through the flow control valve is virtually constant. Therefore, once adjusted, the drive / cooling pressure control valve will maintain the correct drive pressure and cooling water pres-sure, independent of reactor vessel pressure. Changes in setting of the pressure control valves are required only to adjust for changes in the cooling requirements of the drives, as the drive seal characteristics change with time. A flow indicator in the control room monitors cooling water flow. A differential pressure indicator in the control room indicates the difference between reactor vessel pressure and drive cooling water pressure. Although the drives can function without cooling water, seal life is shortened by long-term exposure to reactor temperatures. The tem-perature of each drive is indicated and recorded, and excessive temperatures are annunciated in the control room. 4.6.1.1.2.4.2.5 Scram Discharge Volume The scram discharge volume consists of header piping which con-nects to each HCU and drains into an instrument volume. The header piping is sized to receive and contain all the water discharged by the drives during a scram, independent of the instrument volume. During normal plant operation, the scram discharge volume is empty and vented to atmosphere through its open vent and drain valve. When a scram occurs, upon a signal from the safety circuit these vent and drain valves are closed to conserve reactor water. Lights in the control room indicate the position of these valves. During a scram, the scram discharge volume partly fills with water discharged from above the drive pistons. After scram is completed, the CRD seal leakage from the reactor continues to flow into the sciam discharge volume until the discharge volume pressure equals the reactor vessel pressure. A check valve in each HCU prevents reverse flow from the scram discharge header volume to the 4.6-16

GESSAR.II 22A7007 238 NUCLEAR ISLAND Rsv. 0 ( 4.6.1.1.2.4.2.5 Scram Discharge Volume (Continued) drive. When the initial scram signal is cleared from the reactor - protection system (RPS), the scram discharge volume si'gnal is over-ridden with a keylock override switch, and the scram discharge volume is drained and returned to atmospheric pressure. Remote manual switches in the pilot valve solenoid circuits allow the discharge volume vent and drain valves to be tested without disturbing the RPS. ~ Closing the scram discharge volume valves allows the outlet scram valve seats to be leak-tested by timing the accumulation of leakage inside the scram discharge volume. Seven liquid-level switches activated by six transmitters con-nected to the instrument volume, monitor-the volume for abnormal water level. They a re set at three different levels. At the lowest level, a switch actuates to indicate that the volume is not com-() pletely empty during post-scram draining or to indicate that the volume starts to fill through leakage accumulation at other times during reactor operation. At the second level, two switches pro-duce a rod withdrawal block to prevent further withdrawal of any control rod when leakage accumulates to half the capacity of the instrument volume. The remaining four switches are interconnected with the trip channels of the Reactor Trip System and will initiate l a reactor scram should water accumulation fill the instrument l volume. T 1 4.6.1.1.2.4.3 Ilydraulic Control Units Each hydraulic control unit (llCU) furnishes pressurized water, on signal, to a drive unit. The drive then positions its control rod as required. Operation of the electrical system that supplies scram and normal control rod positioning signals to the IICU is 1 j described in Subsection 7.7.1.2 (Rod Control and Information System). !O 4.6-17

 , , - - - , ~  ,- , ,-- - . ,.,., ,.,    -

n . - - .,-,,-- ,n ,n , _ _n e,,- ,, , , , - - - - , . . . _ , - . , , , , , - , , , . , ~ - , , . - , , , . , - - ,

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.1.1.2.4.3 Ilydraulic Control Units (Continued) The basic components in each IICU are: (1) manual, pneumatic and electrical valves; (2) an accumulator; (3) related piping; (4) electrical connections; (5) filters; and (6) instrumentation (Figures 4.6-5a, 4.6-5b, 4.6-5c, 4.6-6, and 4.6-8). The compo-nents and their functions are described in the following paragraphs. 4.6.1.1.2.4.3.1 Insert Drive Valve The insert drive valve is solenoid-opc .d and opens on an insert signal. The valve supplies drive watec co the bottom side of the main drive piston. 4.6.1.1.2.4.3.2 Insert Exhaust Valve The insert exhaust solenoid valve also opens on an insert signal. The valve discharges water from above the drive piston to the exhaust water header. 4.6.1.1.2.4.3.3 Withdraw Drive Valve The withdraw dritt valve is solenoid-operated and opens on a withdraw signal. The valve supplies drive water to the top of the drive piston. 4.6.1.1.2.4.3.4 Withdraw Exhaust Valve l The solenoid-operated withdraw exhaust valve opens on a withdraw signal and discharges water from below the main drive piston to the exhaust header. It also serves as the settle valve, which opens, following any normal drive movement (insert or withdraw), to allow the control rod and its drive to settle back into the nearest latch position. 4.6-18

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.1.1.2.4.3.5 Speed Control Units The insert drive valve and withdraw exhaust valve have a speed control unit. The speed control unit regulates the control rod insertion and withdrawal rates during normal operation. The manually-adjustable flow control unit is used t, regulate the water flow to and from the volume beneath the main drive piston. A correctly adjusted unit does not require readjustment except to compensate for changes in drive seal leakage. 4.6.1.1.2.4.3.6 Scram Pilot Valve Assembly The scram pilot valve assembly is operated frcm the RPS. The scram pilot valve assembly, with two solenoids, controls both the scram inlet valve and the scram exhaust valve. The scram pilot valve assembly is solenoid-operated and is normally energized. On loss of electrical signal to the solenoids, such as the loss of exter-nal a-c power, the inlet port closes and the exhaust port opens. ( The pilot valve assembly (Figures 4>6-Sa and b) is designed so that the trip system signal must be removed from both solenoids before air pressure can be discharged from the scram valve operators. This prevents inadvertent scram of a single drive in the event of a failure of one of the pilot valve solenoids. 4.6.1.1.2.4.3.7 Scram Inlet Valve The scram inlet valve opens to supply pressurized water to the bottom of the drive piston. This quick opening globe valve is operated by an internal spring and system pressure. It is closed by air pressure applied to the top of its diaphragm operator. A position indicator switch on this valve energizes a light in the control room as soon as the valve starts to open. O 4.6-19

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.1.1.2.4.3.8 Scram Exhaust Valve The scram exhaust valve opens slightly before the scram inlet valve, exhausting water from above the drive piston. The exhaust valve opens faster than the inlet valve because of the higher air pressure spring setting in the valve operator. 4.6.1.1.2.4.3.9 Scram Accumulator The scram accumulator stores sufficient energy to fully insert a control rod at any vessel pressure. The accumulator is a hydraulic cylinder with a free-floating piston. The piston separates the water on top from the nitrogen below. A check valve in the accumu-lator charging line prevents loss of water pressure in the event supply pressure is lost. During normal plant operation, the accumulator piston is seated at the bottom of its cylinder. Loss of nitrogen decreases the nitrogen pressure, wlich actuates a pressure switch and sounds an alarm in the control room. To ensure that the accumulator is always able to produce a scram, it is continuously monitored for water leakage. A float-type level switch actuates an alarm if water leaks past the piston barrier and collects in the accumulator instrumentation block. 4.6.1.1.2.5 Control Rod Drive System Operation The Control Rod Drive System (CRDS) performs rod insertion, rod withdrawal and scram. These operational functions are described in the following sections. 4.6.1.1.2.5.1 Rod Insertion Rod insertion is initiated by a signal from the operator to the insert valve solenoids. This signal causes both insert valves to open. The insert drive valve applies reactor pressure plus 4.6-20

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.1.1.2.5.1 Rod Insertion (Continued) approximately 90 psi to the bottom of the drive piston. The insert exhaust valve allows water from above the drive piston to discharge to the exhaust header. As shown in Figure 4.6-3, the locking mechanism is a ratchet-type device and does not interfere with rod insertion.- The speed at which the drive moves is determined by the flow through the insert speed control valve,_which is set for'approximately 4 gpm for a shim speed (nonscram operation) of 3 in./sec. During normal'inser-tion, the pressure on the downstream side of the speed control valve is 90 to 100 psi above reactor vessel pressure. However, if the drive slows for any reason, the flow through and pressure drop across the insert speed control valve will decrease; the full differential pressure (260 psi) will then be available to cause continued insertion. With 260-psi differential pressure acting on the drive piston, the piston exerts an upward force of 1040 lb. 4.6.1.1.2.5.2 Rod Withdrawal Rod withdrawal is, by design, more involved than insertion. The collet finger (latch) must be raised to reach the unlocked position (Figure 4.6-3). The notches in the index tube and the collet fingers are shaped so that the downward force on the index tube holds the collet fingers in place. The index tube must be lifted before the collet fingers can be released. This is done by opening the drive insert valves (in the manner described in the preceding paragraph) for approximately 1 sec. The withdraw valves are then opened, applying driving pressure above the drive piston and opening the area below the piston to the exhaust header. Pressure is simul-l taneously applied to the collet piston. As the piston raises, the { collet fingers are cammed outward, away from the index tube, by the i guide cap. l '} The pressure required to release the latch is set and maintained at a level high enough to overcome the force of the latch return spring 4.6-21

GESSAR II 22A7007 283 NUCLEAR ISLAND Rev. 0 4.6.1.1.2.5.2 Rod Withdrawal (Continued) plus the force of reactor pressure opposing movement of the collet piston; when this occurs, the index tube is unlatched and free to move in the withdraw direction. Water displaced by the drive piston flows out through the withdraw speed control valve, which is set to give the control rod a shim speed of 3 in./sec. The entire valving sequence is automatically controlled and is initiated by a single operation of the rod withdraw switch. 4.6.1.1.2.5.3 Scram During a scram, the scram pilot valve assembly and scram valves are operated as previously described. With the scram valves open, accumulator pressure is admitted under the drive piston, and the area over the drive piston is vented to the scram discharge volume. The large differential pressure (approximately 1750 pai, initially and always several hundred psi, depending on reactor vessel pres-sure) produces a large upward force on the index tube and control rod. This force gives the rod a high initial acceleration and provides a large margin of force to overcome friction. After the initial acceleration is achieved, the drive continues at a diminish-ing velocity. This characteristic provides a high initial rod inaertion rate. As the drive piston nears the top of its stroke, the piston reaches the buffor and the driveline is brought to a stop at the full-in position. Prior to a scram signal, the accumulator in the IICU has 1750-2000 psig on the water side and 1750 psig on the nitrogen side. As the inlet scram valve opens, the full water side pressure is available at the CRD acting on a 4.1 in.2 area. As CRD motion begins, this pressure drops to the gas side pressure less line losses between the accumulator and the CRD. When the drive reaches the full-in position, the accumulator completely discharges with a resulting gas side pressure of approximately 1200 psig. 4.6-22

GESSAR II 22A7007 ' 238 NUCLEAR ISLAND Rev. 0 4.6.1.1.2.5.3 S6 ram (Continued) 2 The CRD accumulators are necessary to scram the control rods within the required time. Each drive, however, has an internal ball-check valve;which allows reactor pressure-to be admitted under the drive piston. If the reactor is above 600 psi, this valve ensures rod insertion in the event the accumulator is not charged or the j inlet scram valve fails to open. The insertion time, however, will be slower than the scram time with a properly' functioning scram system. i ,l The CRDS, with accumulators, provides the following scram perform-ances at full power operation, in terms of average elapsed time after the opening of the RPS trip actuator (scram signal) for the drives to attain the scram strokes listed: From Full-Out (Notch Position 48) To: () Notch Position Stroke (in.) 44 12 28 60 108 12 Time (sec) 0.28 0.91 1.620 4.6.1.1.2.6 Instrumentation The instrumentation for both the control rods and control rod drives is defined by that given for the rod control and information system. The objective of the rod control and information system is to provide the operator with the means to make changes in nuc-lear reactivity so that reactor power level and power distribution i can be controlled. The system allows the operator to manipulate control rods. The design bases and further discussion are covered in Chapter 7,

               " Instrumentation and Control System."

O 4.6-23

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.1.2 Control Rod Drive llousing Supports 4.6.1.2.1 Safety Objective O The control rod drive (CRD) housing supports prevent any signifi-cant nuclear transient in the event a drive housing breaks or separates from the bottom of the reactor vessel. 4.6.1.2.2 Safety Design Bases The CRD housing supports shall meet the following safety design bases: (1) Following a postulated CRD housing failure, control rod downward motion shall be limited so that any resulting nuclear transient could not be sufficient to cause fuel damage. (2) The clearance between the CRD housings and the supports shall be sufficient to prevent vertical contact stresses caused by thermal expansion during plant operation. 4.6.1.2.3 Description The CRD housing supports are shown in Figure 4.6-8. Horizontal beams are installed immediately below the bottom head of the reactor vessel, between the rows of CRD housings. The beams are welded to brackets which are welded to the steel form liner of the drive room in the reactor support pedestal. 11 anger rods , approximately 10 ft long and 1-3/4 1. n . in diameter, are supported from the beams on stacks of disc springs. These springs compress approximately 2 in. under the design load. The support bars are bolted between the bottom ends of the hanger rods. The spring pivots at the top, and the beveled, loose fitting 4.6-24

                                                 -  a GESSAR II                         22A7007 238 NUCLEAR ISLAND                      Rev. 0 4.6.1.2.3  Description (Continued) 7-w V
                                                                      ~

ends on the support bars prevent substantial bending moment i$ the hanger rods if the support bars are overloaded. Individual grids rest on the support bars between adjacent beams. Because a single-piece grid would be difficult to handle'in the limited work space and because it is necessary that control rod drives, position indicators, and in-core instrumentation components be accessible for inspection and maintenance, each grid is designed for in-place assembly or disassembly. Each grid assembly is made from two grid plates, a clamp, and a bolt. The top part of the clamp guides the grid to its correct position directly below the respective CRD housing that it would support in the postulated accident. When the support bars and grids are installed, a gap of approx-(}

 \s '

imately 1 in. at room temperature (approximately 70*F) is provided between the grid and the bottom contact surface of the CRD flange. During system heatup, this gap is reduced by a net downward expan-sion of the housings with respect to the supports. In the hot operating condition, the gap is approximately 3/4 inch. In the postulated CRD housing failure, the CRD housing supports are loaded when the lower contact surface of the CRD flange con-tacts the grid. The resulting load is then carried by two grid plates, two support bars, four hanger rods, their disc springs, and two adjacent beams. The American Institute of Steel Construction (AISC) Manual of Steel Construction (Specification for the Design, Fabrication and Erection of Structural Steel for Buildings) was used in designing the CRD housing support system. However, to provide a structure that absorbs as much energy as practical without yielding, the

  '~  allowable tension and bending stresses used were 90% of yield v

4.6-25

GESSAR II 22A7007 238 NUCLEAR IS LAND Rev. 0 4.6.1.2.3 Description (Continued) and the shear stress used was 60% of yield. These design stresses O are 1.5 times the AISC allowable stressen (60% and 40% of yield, respectively). For purposes of mechanical design, the postulated failure resulting in the highest forces is an instantaneous circumferential separa-tion of the CRD housing from the reactor vessel, with the reactor at an operating pressure of 1086 psig (at the bottom of the vessel) acting on the area of the separated housing. The weight of the separated housing, CRD and blade, plus the pressure of 1086 psig acting on the area of the separated housing, gives a force of approximately 32,000 lb. This force is used to calculate the impact force, conservatively assuming that the housing travels through a 1-in. gap before it contactc the supports. The impact force (109,000 lb) is then treated as a static load in design. The CRD housing supports are designed as category I (seismic) equipment in accordance with Section 3.2. Loading conditions and examples of stress analysis results and limits are shown in Table 3.9-10(19). Safety evaluation is discussed in Subsection 4.6.2.3.3. 4.6.2 Evaluations of the CRDS 4.6.2.1 Failure Mode and Effects Analysis This subject is covered in Appendix 15A "NSOA". 4.6.2.2 Protection from Common Mode Failures The position on this subject is covered in Appendix 15A "NSOA". O 4.6-26

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.2.3 Safety Evaluation 1 Safety evaluation of the control rods, CRDS, and CRD housing supports is described below. Further description of control rods is contained in Section 4.2. 4.6.2.3.1 Control Rods 4.6.2.3.1.1 Materials Adequacy Throughout Design Lifetime The adequacy of the materials throughout the design life was evaluated in the mechanical design of the control rods. The pri-mary materials, B C powder and Type-304 austenitic stainless steel, 4 have been found suitable in meeting the demands of the BWR environment. 4.6.2.3.1.2 Dimensional and Tolerance Analysis () Layout studies are done to assure that, given the worst combina-tion of part tolerance ranges at assembly, no interference exists which will restrict the passage of control rods. In addition, preoperational verification is made on each control blade system to show that the acceptable levels of operational performance are met. 4.6.2.3.1.3 Thermal Analysis of the Tendency to Warp The various parts of the control rod assembly remain at approx-imately the same temperature during reactor operation, negating the problem of distortion or warpage. What little differential thermal growth could exist is allowed for in the mechanical design. A minimum axial gap is maintained between absorber rod tubes and the control rod frame assembly for the purpose. In addition, to further thid end, dissimilar metals are avoided. O 4.6-27

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.2.3.1.4 Forces for Expulsion An analysis has been performed which evaluates the maximum pressure O forces which could tend to eject a control rod from the core. The results of this analysis are given in Subsection 4.6.2.3.2.2.2, (Rupture of Hydraulic Line(s) to Drive Housing Flange). In sum-mary, if the collet were to remain open, which is unlikely, calcu-lations indicate that the steady-state control rod withdrawal velocity would be 2 ft/sec for a pressure-under line break, the limiting case for rod withdrawal. 4.6.2.3.1.5 Functional Failure of Critical Components The consequences of a functional failure of critical components have been evaluated and the results are covered in Subsection 4.6.2.3.2.2 (Analysis of Malfunction Relating to Rod Withdrawal). 4.6.2.3.1.6 Precluding Excessive Rates of Reactivity Addition In order to preclude excessive rates of reactivity addition, analysis has been performed both on the velocity limiter device and the effect of probable control rod failures (Subsection 4.6.2.3.2.2). 4.6.2.3.1.7 Effect of Fuel Rod Failure on Control Rod Channel Clearances The CRD mechanical design ensures a sufficiently rapid insertion of control rods to preclude the occurrence of fuel rod failures which could hinder reactor shutdown by causing significant distortions in channel clearances. 4.6.2.3.1.8 Mechanical Damage In addition to the analysis performed on the CRD (Subsection 4.6.2.3.2.2) and Subsection 4.6.2.3.2.3 (Scram Reliability) and l the control rod blade, analyses were performed on the control rod l l 4.6-28

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.2.3.1.8 Mechanical Damage (Continued) guide tube (see Subsections 4.2.3.3.7 through 4.2.3.3.8 for these analysca). 4.6.2.3.1.9 Evaluation of Control Rod Velocity Limiter The control rod velocity limiter limits the free-fall velocity of the control rod to a value that cannot result in nuclear system process barrier damage. This velocity is evaluated by the rod drop accident analysis in Chapter 15 (Accident Analysis). 4.6.2.3.2 Control Rod Drives 4.6.2.3.2.1 Evaluation of Scram Time The rod scram function of the CRD system provides the negative reactivity insertion required by safety design basis 4 . 6.1.1.1.1. l (1) . The scram time shown in the description is adequate as shown by the transient analyses of Chapter 15. 4.6.2.3.2.2 Analysis of Malfunction Relating to Rod Withdrawal There are no known single malfunctions that cause the unplanned withdrawal of even a single control rod. However, if multiple malfunctions are postulated, studies show that an unplanned rod withdrawal can occur at withdrawal speeds that vary with the combination of malfunctions postulated. In all cases, the subse-quent withdrawal speeds are less than that assumed in the rod drop accident analysis discussed in Chapter 15. Therefore, the physical and radiological consequences of such rod withdrawals are less than those analyzed in the rod drop accident. O 4.6-29

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.2.3.2.2.1 Drive Housing Fails at Attachment Weld The bottom head of the reactor vessel has a penetration for each O CRD location. A drive housing is raised into position inside each penetration and fastened by welding. The drive is raised into the drive housing and bolted to a flange at the bottom of the housing. The CRD housing material at the vessel penetration is seamless, Alloy 600 tubing with a minimum tensile strength of 80,000 psi, and Type-304 stainless steel pipe below the vessel with a minimum strength of 75,000 psi. The basic failure considered here is a complete circumferential crack through the housing wall at an elevation just below the J-weld. Static loads on the housing wall include the weight of the drive and the control rod, the weight of the housing below the J-weld, and the reactor pressure acting on the 6-in. diameter cross-sectional area of the housing and the drive. Dynamic loading results from the reaction force during drive operation. If the housing were to fail as described, the following sequence of events is foreseen. The housing would separate from the vessel. The CRD and housing would be blown downward against the support structure, by reactor pressure acting on the cross-sectional area of the housing and the drive. The downward motion of the drive and associated parts would be determined by the gap between the bottom of the drive and the support structure and by the deflection of the support structure under load. In the cur-rent design, maximum deflection is approximately 3 in. If the collet were to remain latched, no further control rod ejection would occur ; the housing would not drop far enough to clear the vessel penetration; reactor water would leak at a rate of approxi-mately 180 gpm through the 0.03-in. diametral clearance between the housing and the vessel penetration. O 4.6-30

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.2.3.2.2.1 Drive llousing Fails at Attachment Weld () (Continued) If the basic housing failure were to occur while the control rod is being withdrawn (this is a small fraction of the total drive operating time) and if the collet were to stay unlatched, the following sequence of events is foreseen. The housing would separate from the vessel; the drive and housing would be blown downward against the CRD housing support. Calculations indicate that the steady-state rod withdrawal velocity would be 0.3 ft/sec. During withdrawal, pressure under the collet piston would be approximately 250 psi greater than the pressure over it. There-fore, the collet would be held in the unlatched position until driving pressure was removed from the pressure-over port. 4.6.2.3.2.2.2 Rupture of !!ydraulic Line(s) to Drive llousing Flange There are three types of possible rupture of hydraulic lines to

 }

the drive housing flange: (1) pressure-under (insert) line break; (2) pressure-over (withdrawn) line break; and (3) coincident breakage of both of these lines. 4.6.2.3.2.2.2.1 Pressure-under (Insert) Line Break For the case of a pressure-under (insert) line break, a partial or complete circumferential opening is postulated at or near the point where the line enters the housing flange. Failure is more likely to occur after another basic failure wherein the drive housing or housing flange separates from the reactor vessel. Failure of the housing, however, does not necessarily lead directly to failure of the hydraulic lines. If the pressure-under (insert) line were to fail and if the collet were latched, no control rod withdrawal would occur. There would be no pressure differential across the collet piston and, (} therefore, no tendency to unlatch the collet. Consequently, the 4.6-31

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.2.3.2.2.2.1 Pressure-under (Insert) Line Break (Continued) associated control rod could not be withdrawn; but if reactor O pressure is greater than 600 psig, it will insert on a scram signal. The ball check valve is designed to seal off a broken pressure-under line by using reactor pressure to shift the check ball to its upper seat. If the ball check valve were prevented from seating, reactor water would leak to the containment. Because of the broken line, cooling water could not be supplied to the drive involved. Loss of cooling water would cause no immediate damage to the drive. However, prolonged exposure of the drive to tem-peratures at or near reactor temperature could lead to deteriora-tion of material in the seals. High temperature would be indi-cated to the operator by the thermocouple in the position indi-cator probe. A second indication would be high cooling water flow. If the basic line failure were to occur while the control rod is h being withdrawn, the hydraulic force would not be sufficient to hold the collet open, and spring force normally would cause the collet to latch and stop rod withdrawal. However, if the collet were to remain open, calculations indicate that the steady-state control rod withdrawal velocity would be 2 ft/sec. 4.6.2.3.2.2.2.2 Pressure-over (Withdrawn) Line Break The case of the pressure-over (withdrawn) line breakage considers the complete breakage of the line at or near the point where it enters the housing flange. If the line were to break, pressure over the drive piston would drop from reactor pressure to atmo-spheric pressure. Any significant reactor pressure (approximately 600 psig or greater) would act on the bottom of the drive piston and fully insert the drive. Insertion would occur regardless of the operational mode at the time of the failure. After full insertion, reactor water would leak past the stop piston seals. 4.6-32

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.2.3.2.2.2.2 Pressure-over (Withdrawn) Line Break [)' 'ud (Continued) This leakage would exhaust to the containment through the broken pressure-over line. The leakage rate at 1000 psi reactor pressure is estimated to be 1 to 3 gpm; however, with the Graphitar-14 (or its equivalent) seals of the stop piston removed, the leakage rate could be as high as 10 gpm, based on experimental measurements. lf the reactor were hot, drive temperature would increase. This situation would be indicated to the reactor operator by the drift alarm, by the fully inserted drive, by a high drive temperature annunciated in the control room and by operation of the drywell sump pump. 4.6.2.3.2.2.2.3 Simultaneous Breakage of the Pressure-over (Withdrawn) and Pressure-under (Insert) Lines For the simultaneous breakage of the pressure-over (withdrawn) and /} pressure-under (insert) lines, pressures above and below the drive

  piston would drop to zero, and the ball check valve would close the broken pressure-under line. Reactor water would flow from the annulus outside the drive, through the vessel ports, and to the space below the drive piston. As in the case of pressure-over line breakage, the drive would then insert (at reactor pressure approx-imately 600 psig or greater) at a speed dependent on reactor pressure. Full insertion would occur regardless of the operational mode at the time of failure. Reactor water would leak past the drive seals and out the broken pressure-over line to the contain-ment, as described above. Drive temperature would increase. Indi-cation in the control room would include the drift alarm, the fully inserted drive, the high drive temperature annunciated in the con-trol room, and the operation of the drywell sump pump.

O) t s_s 4.6-33

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.2.3.2.2.3 All Drive Flange Bolts Fail in Tension Each CRD is bolted to a flange at the bottom of a drive housing. O The flange is welded to the drive housing. Bolts are made of AISI-4140 steel, with a minimum tensile strength of 125,000 psi. Each bolt has an allowable load capacity of 15,200 lb. Capacity of the 8 bolts is 121,600 lb. As a result of the reactor design pressure of 1250 psig, the major load on all 8 bolts is 30,400 lb. If a progressive or simultaneous failure of all bolts were to occur, the drive would separate from the housing. The control rod and the drive would be blown downward against the support struc-ture. Impact velocity and support structure loading would be slightly less than that for drive housing failure, because reactor pressure would act on the drive cross-sectional area only and the housing would remain attached to the reactor vessel. The drive would be isolated from the cooling water supply. Reactor water would flow downward past the velocity limiter piston, through the large drive filter, and into the annular space between the thermal h sleeve and the drive. For worst-case leakage calculations, the large filter is assumed to be deformed or swept out of the way so it would offer no significant flow restriction. At a point near the top of the annulus, where pressure would have dropped to 350 psi, the water would flash to steam and cause choke-flow conditions. Steam would flow down the annulus and out the space between the housing and the drive flanges to the drywell. Steam formation would limit the leakage rate to approximately 840 gpm. If the collet were latched, control rod ejection would be lim-ited to the distance the drive can drop before coming to rest on the support structure. There would be no tendency for the collet to unlatch, because pressure below the collet piston would drop to zero. Pressure forces, in fact, exert 1435 lb to hold the collet in the latched position. O 4.6-34

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.2.3.2.2.3 All Drive Flange Bolts Fail in Tension (Continued) {~N If the bolts failed during control rod withdrawal, pressure below the collet piston would drop to zero. The collet, with 1650 lb return force, would latch and stop rod withdrawal. 4.6.2.3.2.2.4 Weld Joining Flange to llousing Fails in Tension The failure considered is a crack in or near the weld that joins the flange to the housing. This crack extends through the wall and completely around the housing. The flange material is forged, Type-304 stainless steel, with a minimum tensile strength of 75,000 psi. The housing material is seamless, Type-304 stainless steel pipe, with a minimum tensile strength of 75,000 psi. The conventional, full penetration weld of Type-308 stainless steel has a minimum tensile strength approximately the same as that for the parent metal. The design pressure and temperature are 1250 psig (/) w. and 575 F. Reactor pressure acting on the cross-sectional area of the drive, the weight of the control rod, drive, and flange, and the dynamic reaction force during drive operation result in a maximum tensile stress at the weld of approximately 5100 psi. If the basic flange-to-housing joint failure occurred, the flange and the attached drive would be blown downward against the support structure. The support structure loading would be s'.ightly less than that for drive housing failure, because reactor pressure would act only on the drive cross-sectional area. Lack of differ-ential pressure across the collet piston would cause the collet to remain lat hed and limit control rod motion to approximately 3 inches. Downward drive movement would be small; therefore, most of the drive would remain inside the housing. The pressure-under and pressure-over lines are flexible enough to withstand the small displacement and remain attached to the flange. Reactor water would follow the same leakage path described above for the flange-() bolt failure, except that exit to the drywell would be through the gap between the lower end of the housing and the top of the flange. 4.6-35

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.2.3.2.2.4 Weld Joining Flange to llousing Fails in Tension (Continued) Water would flash to steam in the annulus surrounding the drit ". The leakage rate would be approximately 840 gpm. If the basic failure were to occur during control rod withdrawal (a small fraction of the total operating time) and if the collet were held unlatched, the flange would separate from the housing. The drive and flange would be blown downward against the support structure. The calculated steady-state rod withdrawal yelocity would be 0.13 ft/sec. Because pressure-under and pressure-over lines remain intact, driving water pressure would continue to the drive, and the normal exhaust line restriction would exist. The pressure below the velocity limiter piston would drop below normal as a result of leakage from the gap between the housing and the flange. This differential pressure across the velocity limiter piston would result in a net downward force of approximately 70 lb. Leakage cut of the housing would greatly reduce the pressure in the annulus surrounding the drive. Thus, the not downward force on the drive piston would be less than normal. The overall effect of these events would be to reduce rod withdrawal to approximately one-half of normal speed. With a 560-psi differential across the collet piston, the collet would remain unlatched; however, it should relatch as soon as the drive signal is removed. 4.6.2.3.2.2.5 Ilousing Wall Ruptures This failure is a vertical split in thr drive housing wall just below the bottom head of the reactor 'essel. The flow area of the hole is considered equivalent to the annular area between the drive and the thermal sleeve. Thus, flow through this annular area, rather than flow through the hole in the housing, would govern leakage flow. The CRD housing is made of Alloy 600 seam-less tubing (at the penetration to the vessel), with a minimum tensile strength of 80,000 psi, and of Type-304 stainless steel 4.6-36

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.2.3.2.2.5 Housing Wall Ruptures (Continued) seamless pipe below the vessel with a minimum tensile strength of 75,000 psi. The maximum hoop stress or 9,000 psi results primarily from the reactor design pressure (1250 psig) acting on the inside of the housing. If such a rupture were to occur, reactor water would flash to steam and leak through the hole in the housing to the drywell at approximately 1030 gpm. Choke-flow conditions would exist, as described previously for the flange-bolt failure. However, leak-age flow would be greater because flow resistance would be less, that is, the leaking water and steam would not have to flow down the length of the housing to reach the drywell. A critical pres-sure of 350 psi causes the water to flash to steam. There would be no pressure differential acting across the collet piston to unlatch the collet; but the drive would insert as a O(_/ result of loss of pressure in the drive housing causing a pressure drop in the space above the drive piston. If this failure occurred during control rod withdrawal, drive withdrawal would stop, but the collet would remain unlatched. The drive would be stopped by a reduction of the net downward force action on the drive line. The net force reduction would occur when the leakage flow of 1030 gpm reduces the pressure in the j annulus outside the drive to approximately 540 psig, thereby i reducing the pressure acting on top of the drive piston to the same I' value. A pressure differential of approximately 710 psi would t exist across the collet piston and hold the collet unlatched as long as the operator held the withdraw signal. I i O l 1 4.6-37 i

  ~,    ,    , - ,  ,., ,. --,, ---,,,,,            . . , , ,      - - , - .          -   n,.n,--,, , , ,, ,,,-              -,        , ,n         ....-

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.2.3.2.2.6 Flange Pluu Blows Out

                         ~

To connect the vessel ports with the bottom of the ball check G valve, a hole of 3/4-in. diameter is drilled in the drive flange. The outer end of this hole is scaled with a plug of 0.812-in. diameter and 0.25-in. thickness. A full-penetration, Type-308 stainless steel weld holds the plug in place. The postulated failure is a full circumferential crack in this weld and subse-quent blowout cf the plug. If the weld were to fail, the plug were to blow out and the collet remained latched, there would be no control rod motion. There would be no pressure differential acting across the collet piston to unlatch the collet. Reactor water would leak past the velocity limiter piston, down the annulus between the drive and the thermal sleeve, through the vessel ports and drilled passage, and out the open plug hole to the drywell at approximately 320 gpm. Leakage calculations assume only liquid flows from the flange. Actually, hot reactor water would flash to steam, and choke-flow conditions h would exist. Thus, the expected leakage rate would be lower than the calculated value. Drive temperature would increase and initi-ate an alarm in the control room. If this failure were to occur during control rod withdrawal and if the collet were to stay unlatched, calculations indicate that con-trol rod withdrawal speed would be approximately 0.24 ft/sec. Leakage from the open plug hole in the flange would cause reactor water to flow downward past the velocity limiter piston. A small differential pressure across the piston would result in an insig-nificant driving force of approximately 10 lb, tending to increase withdraw velocity. A pressure differential of 295 psi across the collet piston would hold the collet unlatched as long as the driving signal was maintained. O 4.6-38

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ("N 4.6.2.3.2.2.6 Flange Plug Blowa Out (Continued) U Flow resistance of the exhaust path from the drive would be normal because the ball check valve would be seated at the lower end of its travel by pressure under the drive piston. 4.6.2.3.2.2.7 Dall Check Valve Plug Blows out An a means of access for machining the ball check valve cavity, a 1.25-in. diameter hole has been drilled in the flange forging. This hole in scaled with a plug of 1.31-in diameter and 0.38-in. thicknesa. A full-penetration weld, utilizing Type-308 stainless steel filler, holds the plug in place. The failure postulated la a circumferential crack in thia weld leading to a blowout of the plug. If the plug were to blow out while the drive was latched, therc would be no control rod motion. No pressure differential would {/) '- exist across the collet platon to unlatch the collet. As in the previous failure, reactor water would flow past the velocity limiter, down the annulus between the drive and thermal alcove, through the vennel ports and drilled passage, through the ball check valve cage and out the open plug hole to the drywell. The leakage calculations indicate the flow rate would be 350 gpm. This calculation assumes liquid flow, but flashing of the hot reactor water to steam would reduce this rate to a lower value. Drive temperature would rapidly increase and initiate an alarm in the control room. If the plug failure were to occur during control rod withdrawal (it would not be possible to unlatch the drive after auch a failure), the collet would relatch at the first locking groove. If the collet were to stick, calculations indicate the control rod withdrawal speed would be 11.8 ft/sec. There would be a large retarding force exerted by the velocity limiter due to a 35 pai prennure differential acronn the velocity limiter piston. 4.6-39

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.2.3.2.2.8 Drive / Cooling Water Pressure Control Valve Closure (Reactor Pressure, 0 psig) The pressure to move a drive is generated by the pressure drop of practically the full system flow through the drive / cooling water pressure control valve. This valve is either a motor-operated valve or a standby manual valve; either one is adjusted to a fixed opening. The normal pressure drop across this valve develops a pressure 260 psi in excess of reactor pressure. If the flow through the drive / cooling water pressure control valve were to be stopped (as by a valve closure or flow blockage) , the drive pressure would increase to the shutoff pressure of the supply pump. The occurrence of this condition during withdrawal of a drive at zero vessel pressure will result in a drive pressure increase from 260 psig to no more than 2000 psig. Calculations indicate that the drive would accelerate from 3 in./sec to approx-imately 7 in./sec. A pressure differential of 1970 psi across the collet piston would hold the collet unlatched. Flow would be upward, past the velocity limiter piston, but retarding force would be negligible. Rod movement would stop as soon as the driving signal was removed. 4.6.2.3.2.2.9 Ball Check Valve Fails to Close Passage to Vessel Ports Should the ball check valve sealing the passage to the vessel ports be dislodged and prevented from rescating following the insert portion of a drive withdrawal sequence, water below the drive piston would return to the reactor through the vessel ports and the annulus between the drive and the housing rather than through the speed control valve. Because the flow resistance of this return path would be lower than normal, the calculated withdrawal speed vould be 2 ft/sec. During withdrawal, differential pressure across the collet piston would be approximately 40 psi. Therefore, the ollet would tend to latch and would have to stick open before con-t inuous withdrawal at 2 ft/sec could occur. Water would flow 4.6-40

GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 0 4.6.2.3.2.2.9 Ball Check Valve Fails to Close Passage to

 /~h                     Vessel Ports (Continued) upward past the velocity '.imiter piston, generating a small retarding force of approximately 120 lb.

4.6.2.3.2.2.10 Ilydraulic Control Unit Valve Failures Various failures of the valves in the HCU can be postulated, but none could produce differential pressures approaching those described in the preceding paragraphs and none alone could produce a high velocity withdrawal. Leakage through either one or both of the scram valves produces a pressure that tends to insert the control rod rather than to withdraw it. If the pressure in the scram discharge volume should exceed reactor pressure following a scram, a check valve in the line to the scram discharge header prevents this pressure from operating the drive mechanisms. j 4.6.2.3.2.2.11 Collet Fingers Fail to Latch ( The failure is presumed to occur when the drive withdraw signal is removed. If the collet fails to latch, the drive continues to withdraw at a fraction of the normal speed. This assumption is made because there is no known means for the collet fingers to become unlocked without some initiating signal. Because the collet fingers will not cam open under a load, accidental application of a down signal does not unlock them. (The drive must be given a short insert signal to unload the fingers and cam them open before the collet can be driven to the unlock position.) If the drive withdrawal valve fails to close following a rod withdrawal, the collet would remain open and the drive continue to move at a reduced speed. O 4.6-42

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.2.3.2.2.12 Withdrawal Speed Control Valve Failure O Normal withdrawal speed is determined by differential pressures in the drive and is set for a nominal value of 3 in./sec. Withdrawal speed is maintained by the pressure regulating system and is inde-pendent of reactor vessel pressure. Tests have shown that acci-dental opening of the speed control valve to the full-open position produces a velocity of approximately 5 in./sec. The CRDS prevents unplanned rod withdrawal, and it has been shown above that only multiple failures in a drive unit and in its con-trol unit could cause an unplanned rod withdrawal. 4.6.2.3.2.3 Scram Reliability liigh scram reliability is the result of a number of features of the CRD system. For example: (1) An individual accumulator is provided for each CRD with sufficient stored energy to scram at any reactor pressure. L - reactor vessel itself, at pressures above 600 psi, will supply the necessary force to insert a drive if its accumulator is unavailable. (2) Each drive mechanism has its own scram valves and a dual solenoid scram pilot valve; therefore, only one drive can be affected if a scram valve fails to open. Both pilot valve solenoids must be de-energized to initiate a scram. (3) The RPS and the IICUs are designed so that the scram sig-nal and mode of operation override all others. (4) The collet assembly and index tube are designed so they will not restrain or prevent control rod insertion during scram. 4.6-42

GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 3 4.6.2.3.2.3 Scram Reliability (Continued) (v) (5) The scram discharge voluae is monitored for accumulated water and the reactor will scram before the volume is reduced to a point that could interfere with a scram. 4.6.2.3.2.4 Control Rod Support and Operation As described above, each control rod is independently supported and controlled as required by safety design bases. 4.6.2.3.3 Control Rod Drive Housing Supports Downward travel of the CRD housing and its control rod following the postulated housing failure equals the sum of these distances: (1) the compression of the disc springs under dynamic loading, and (2) the initial gap between the grid and the bottom contact

    ~. surface of the CPD flange. If the reactor were cold and pres-
 \_ ,/ surized, the downward motion of the control rod would be limited to the spring compression (approximately 2 in.) plus a gap of approximately 1 inch. If the reactor were hot and pressurized, the gap would be approximately 3/4 in, and the spring compression would be slightly less than in the cold condition. In either case, the control rod movement following a housing failure is substan-tially limited below one drive " notch" movement (6 in.). Sudden withdrawal of any control rod through a distance of one drive notch at any position in the core does not produce a transient sufficient to damage any radioactive material barrier.

The CRD housing supports are in place during power operation and when the nuclear system is pressurized. If a control rod is ejected during shutdown, the reactor remains subcritical because it is designed to remain suberitical with any one control rod fully withdraw- at any time. CN

\~-)

4.6-43

GESSAR II 22A7007 2 38 NUCLEAR ISLAND Rev. 0 4.6.2.3.3 Control Rod Drive ilousing Supports (Continued) At plant operating temperature, a gap of approximately 3/4 in. exists between the CRD housing and the supports. At lower tem-peratures the gap is greater. Because the supports do not contact any of the CRD housing except during the postuluted accident con-dition, vertical contact stresses are prevented. Inspection and testing of CRD housing supports is discussed in 4.6.3.2.1. 4.6.3 Testing and Verification of the CRDs 4.6.3.1 Control Rod Drives 4.6.3.1.1 Testing and Inspection 4.6.3.1.].1 Development Tests The initial development drive (prototype of the standard locking piston design) testing included more than 5000 scrams and approxi-mately 100,000 latching cycles. One prototype was exposed to simulated operating conditions for 5000 hours. These tests demonstrated the following: (1) The drive casily withstands the forces, pressures and temperatures imposed. (2) Wear, abrasion and corrosion of the nitrided stainless l parts are negligible. Mechat :al performance of the nitrided surface is superior to that of materials used in earlier operating reactors. (3) The basic scram speed of the drive has a satisfactory margin above minimum plant requirements at any reactor vessel pressure. O 4.6-44

     ..              .-     . -                 _          . .                   .              _   - = . .       .--           --.                 - _     _ . . -   --

GESSAR II 22A7007 i 238 NUCLEAR ISLAND Rev. 0 4.6.3.1.1.1 Development Tests (Continued) I (a) Usable seal lifetimes in excess of 1000 scram cycles can be expected. i 4.6.3.1.1.2 Factory Quality Control Tests Quality control of welding, heat treatment, dimensional tolerances, material verification and similar factors is maintained throughout the manufacturing process to assure reliable performance of the mechanical reactivity control components. Some of the quality contro'l tests performed on the control rods, CRD mechanisms, and IICUs are listed below: i

!                       (1)    CRD mechanism tests:

a. Pressure welds on the drives are hydrostatically tested in accordance with ASME codes.

b. Electrical components are checked for electrical continuity and resistance to ground.

c. Drive parts that cannot be visually inspected for dirt are flushed w,4th filtered water at high velocity. No significant foreign material is permitted in effluent water.

i

d. Seals are tested for leakage to demonstrate correct seal operation.

l e. Each drive is tested for shim motion, latching, and control rod position indication.

f. Each drive is subjected to cold scram tests at various reactor pressures to verify correct scram l

performance. 4.6-45 4

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.3.1.1,2 Factory Quality Control Tests (Continued) O (2) HCU tests:

a. Hydraulic systems are hydrostatically tested in accordance with the applicable code.

b. Electrical components and systems are tested for electrical continuity and resistance to ground.

c. Correct operation of the accumulator pressure and level switches is verified.

d. The unit's ability to perform its part of a scram

_ - is demonstrated.

c. Correct operation and adjustment of the insert and withdrawal valves is demonstrated.

O 4.6.3.1.1.3 Operational Tests After installation, all rods and drive mechanisms can be tested through their full stroke for operability. During normal operation, each time a control rod is withdrawn a notch, the operator can observe the in-core monitor indications to verify that the control rod is following the drive mechanism. All control rods that are partially withdrawn from the core can be tested for rod-following by inserting or withdrawing the rod one notch and returning it to its original position, while the oper-ator observes the in-core monitor indications. To make a positive test of control rod to CRD coupling integrity, the operator can withdraw a control rod to the end of its travel and then attempt to withdraw the drive to the overtravel position. 4.6-46

_ _ _ . - _ _ _ _m - . _ _ _ _ _ . _ . . . - . _ _ _= . . _ _ _ . _ _ .__ - - _ . . _ _ - - _ I GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ,

     ,                                4.6.3.1.1.3                                    Operational Tests (Continued)

Failure of the drive to overtravel demonstrates rod-to-drive ,

;                                     coupling integrity.

! Ilydraulic supply subsystem pressures can be observed from instrumentation in the control room. Scram accumulator pressures can be observed on the nitrogen pressure gages. 4.6.3.1.1.4 Acceptance Tests ' Criteria for acceptance of the individual CRD .nechanisms and . the associated control and protection systems will be incorporated in specifications and test procedures covering three distinct phases: (1) pre-installation; (2) after installation prior to startup; i and (3) during startup testing. 4 l . The pre-installation specification will define criteria and ' l acceptable ranges of such characteristics as seal leakage, fric-I tion and scram performance under fixed test conditions which must 4 be met before the component can be shipped. The after-installation, prestartup tests (Chapter 14) include l normal and scram motion and are primarily intended to verify that piping, valves, electrical components and instrumentation are properly installed. The test specifications will include criteria and acceptable ranges for drive speed, timer settings, scram valve response times, and control pressures. These are tests intended j more to document system condition rather than tests of performance. As fuel is placed in the reactor, the startup test procedure (Chapter 14) will be followed. The tests in this procedure are intended to demonstrate that the initial operational characteris-tics meet the limits of the specifications over the range of pri-I i mary coolant temperatures and pressures from ambient to operating. The detailed specifications and procedures have not as yet " l l 4.6-47

         . _ _ _ . - ~ . - _ .._.-._ ,.. - _,.-. _ ..._ -. _ .- - _ .,. _-.. - . -                                                _-_- _ _.__.-. ._--._-._ _ _._.                               . - - _ _ - - _ _ . -

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.3.1.1.4 Acceptance Tests (Continued) been prepared but will follow the general pattern established for O such specifications and procedures in BWRs presently under con-struction and in operation. 4.6.3.1.1.5 Surveillance Tests The nurveillance requirements (SR) for the CRD system are described below. (1) Sufficient control rods shall be withdrawn, following a refueling outage when core alterations are performed, to demonstrate with a margin of 0.25% ok that the core can be made subcritical at any time in the subsequent fuel cycle with the strongest operable control rod fully withdrawn and all other operable rods fully inserted. (2) Each partially or fully withdrawn control rod shall be exercised one notch at least once each week. In the event that operation is continuing with three or more rods valved out of service, this test shall be performed at least once each day. The weekly control rod exercise test serves as a periodic check against deterioration of the control rod system and also verifies the ability of the control rod drive to scram. If a rod can be moved with drive pressure, it may be expected to scram since higher pressure is applied during scram. The frequency of exercising the control rods under the conditions of three or more con-trol rods valved out of service provides even further assurance of the reliability of the remaining control rods. O 4.6-48

_ . _ _ . _ _ _ _ _ __ _ _ - ._ _ _ = . .. 2 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.3.1.1.5 Surveillance Tests (Continued) (3) The coupling integrity shall be verified for each 1 withdrawn control rod as follows:

a. when the rod is first withdrawn, observe discernible

!                                               response of the nuclear instrumIentation; and

b. when the rod is fully withdrawn the first time, observe that the drive will not go to the over-l travel position.

Observation of a response from the nuclear instrumenta-tion during an attempt to withdraw a control rod indi-cates indirectly that the rod and drive are coupled.

The overtravel position feature provides a positive' check on the coupling integrity, for only an uncoupled g-~g drive can reach the overtravel position. (4) During operation, accumulator pressure and level at the normal operating value shall be verified. Experience with CRD systems of the same type indicates ! that weekly verification of accumulator pressure and lovel is sufficient to assure operability of the accumu-lator portion of the CRD system. l i i (5) At the time of each major refueling outage, each operable control rod shall be subjected to scram time tests from the fully withdrawn position. l Experience indicates that the scram times of the control rods do not significantly change over the time interval l between refueling outages. A test of the scram times at each refueling outage is sufficient to identify any f i significant lengthening of the scram times. > l ' 4.6-49

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.3.1.1.6 Functional Tests O The functional testing program of the CRDs consists of the five-year maintenance life and the 1.5X design life test programs as described ir Subsection 3.9.4.4. There are a number of failures that can be postulated on the CRD but it would be very difficult to test all possible failures. A partial test program with postulated accident conditions and imposed single failures is available. The following tests with imposed single failures have been per-formed to evaluate the performance of the CRDs under these conditions: (1) Simulated Ruptured Scram Line Test (2) Stuck Ball Check Valve in CRD Flange (3) IICU Drive Down Inlet Flow Control Valve (V122) Failure (4) IICU Drive Down Outlet Flow Control Valve (V120) Failure (5) CRD Scram Performance with V120 tialfunction (6) IICU Drive Up Outlet Control Valve (V121) Failure (7) IICU Drive Up Inlet Control Valve (V123) Failure (8) Cooling Water Check Valve (V138) Leakage (9) CRD Flange Check Valve Leakage l 4.6-50

GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 0 4.6.3.1.1.6 Functional Tests (Continued) (10) CRu Stabilization Circuit Failure (11) IICU Filter hestriction (12) Air Trapped in CRD llydraulic System (13) CRD Collet Drop Test (14) CR Qualification Velocity Limiter Drop Test Additional postulated CRD failures are discussed in Sub-sections 4.6.2.3.2.2.1 through 4.6.2.3.2.2.11. 4.6.3.2 Control Rod Drive llousing Supports 4.6.3.2.1 Testing and Inspection O . CRD housing supports are removed for inspection and maintenance of the control rod drives. The supports for one control rod can be removed during reactor shutdown, even when the reactor is pres-surized, because all control rods are then inserted. When the support structure is reinstalled, it is inspected for correct assembly with particular attention to maintaining the correct gap between the CRD flange lower contact surface and the grid. . l O 4.6-51 4

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4.6.4 Information for Combined Performance of Reactivity Control Sy5tems 4.6.4.1 Vulnerability to Common Mode Failures The system is located such that it is protected from common mode failures due to missiles, failures of moderate and high energy piping, and fire. Sections 3.4, 3.5 and 3.6, and Subsection 9.5.1 discuss protection of essential systems against missiles, pipe breaks and fire. 4.6.4.2 Accidents Taking Credit for Multiple Reactivity Systems There are no postulated accidents evaluated in Chapter 15 that take credit for two or more reactivity control systems preventing or mitigating each accident. 4.6.5 Evaluation of Combined Performance As indicated in Subsection 4.6.4.2, credit is not taken for multiple O reactivity control systems for any postulated accidents in Chapter 15. 4.6.6 References

1. J.E. Benecki, " Impact Testing on Collect Assembly for Control Rod Drive Mechanism 7RD B144A", General Electric Company, Atomic Power Equipment Department, November 1967 (APED-5555).

O 4.6-52

1 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O - D 0 l D 0 -

                                  ~
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            /.

GESSAR II 22A7007 23S NUCLEAR ISLAND Rev. 0

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 (D \vl s CONTfr')L ROD [~

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

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ISOLATION V ALVE CHARGING WATER HISER SCH AM V ALVE PILOT AIR _ ISOLATION V ALVE

                                                                         '                h-                  ISOLATION V ALV E WITHDR AWA L RISER EXHAUST WATEH HISEFt                                ,                    d ISOLATION V ALVE ISOLATION V ALVE                                         '
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                                                        .,,#                                                  SCHAM PILOT VALVE ASSEMBLY U                          --

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                                                                                                 .            INDIC A TOR CARTHIOGE VALVE ACCUMU L ATOil N2 CH AHGtNG ACCUMU L A TOFT INSTHUMF NT ATION ASSEMBLY figure 4.6-7.        Control Rod Drive Ilydraulic Control Unit 4.6-63

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 O r s -. - .h ,c" m,- y#D / -

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