Nuclear Power Reactor Technology

Modure 2

Reactor Core and Components

Sung-Quun Zee

Manager
Core Dseign and Analysis Technology Dept.
Korea Atomic Energy Research Institute

CONTENTS 

 TABLE 

FIGURE

1. INTRODUCTION

    The design of a commercial light water reactor is such that the reactor core is loaded with a specified number of fuel assemblies that are generally identical in design but different in the amount of fissile material content. In the initial core the fuel assemblies differ in the initial enrichment of the fuel, and in subsequent fuel cycles they differ in the amount of the burnup of the fuel as well. The reactor is fueled at intervals varying from 12 to 24 moths. The refueling of a reactor consists of removing part of the core (a certain number of irradiated fuel  assemblies, the number and identity of which are determined by a fuel management scheme) and loading  an equal number of fresh and possibly previously burned fuel assemblies called the "reload batch". In general, after refueling, the neutronic, thermal-hydraulic, safety, and operating parameters of the core would be different from the previous fuel cycle. The design analyses required to determine the mechanical design, enrichment and number of assemblies of the initial or reload batches as well as the core loading pattern, the nuclear and thermal-hydraulic characteristics of the core, and the safety analyses demonstrating the safety of operation of the reactor is called core design. This section presents an overview of reactor core and components of Korean Standard Nuclear Power Plant (KSNP). In addition to this, a brief overview of the major elements of the core design process and the design criteria are provided.

2. REACTOR

    The reactor is of the pressurized water type using two reactor coolant loops. A vertical cross section of the reactor is shown in Figure. 2.1. The reactor core is composed of 177 fuel assemblies and 73 control element assemblies (CEAs). The fuel assemblies are arranged to approximate a right circular cylinder with an equivalent diameter of 123 inches (3.12 meters) and an active length of 150 inches (3.81 meters). The fuel assembly, which provides for 236 fuel rod positions (16 × 16 array), consists of five guide tubes welded to spacer girds and is closed at the top and bottom by end fittings. The guide tubes each displace four fuel rod positions and provide channels with guide the CEAs over their entire length of travel. In-core instrumentation is installed in the central guide tube of selected fuel assemblies. The in-core instrumentation is routed into the bottom of the fuel assemblies, fuel rods, and CEA guide tubes.

    The fuel is low-enrichment UO₂ in the form of ceramic pellets and is encapsulated in pre-pressurized Zircaloy tubes which from a hermetic enclosure.

    The isometric view of the reactor coolant system is shown in Figure 2.3. The reactor coolant enters the inlet nozzles of the reactor vessel, flow downward between the reactor vessel wall and the core barrel, and passes through the flow skirt section, where the flow distribution is equalized, and into the lower plenum. The coolant then flows upward through the core removing heat from the fuel rods. The heated coolant enters the core outlet region where the coolant flows around the outside of control element assembly shroud tubes to the reactor vessel outlet nozzles. The schematic reactor flow paths is shown in Figure 2.4. The control element assembly shroud tubes protect the individual neutron absorber elements of the CEAs from the effect of coolant cross-flow above the core. The reactor internals support and orient the fuel assemblies, control element assemblies, and in-core instrumentation, and guide the reactor coolant through the reactor vessel. They also absorb static and dynamic loads and transmit the loads to the reactor vessel flange. They will safety perform their functions during normal operating, upset, and faulted conditions. The internals are designed to safety withstand forces due to dead weight, handling, temperature and pressure differentials, flow impingement, vibration, and seismic acceleration. All reactor components are considered Category I for seismic design. The design of the reactor internals limits deflection where required by function. The stress values of all structural members under normal operating and expected transient conditions are not greater than those established by Section Ⅲ of the ASME Code. The effect of neutron irradiation on the materials concerned is included in the design evaluation. The effect of accident loadings in the internals is included in the design analysis.

    Reactivity control is provided by two independent system: the control element drive system and the chemical and volume control system. The control element drive system controls short-term reactivity changes and is used for long-term reactivity changes and can make the reactor subcritical without the benefit of the control element drive system. The design of the core and the reactor protection system prevents fuel damage limits from being exceeded for any single malfunction in either of the reactivity control systems. Details of the core components are discussed in the following section.

3. CORE COMPONENTS

    The core components which affect the core design most are the fuel assemblies, the fuel pins, the Gd-bearing fuel rods (shims), the control rods, the materials used for the grids, guide tubes, and fuel rod cladding, and the in-core instrumentation.

3.1 Fuel Assemblies and Core Loading Pattern

    The loading pattern for Cycle 1 of KNSP, showing fuel type in each core location, is displayed in Figure 3.1-a. The Cycle 1 design features a 4-batch, mixed central-zone loading. This loading pattern facilitates the transition of the initial core to the equilibrium cores as well as higher utilization of the initial-core fuel compared with conventional three batch patterns. The initial-core design also features a low-leakage type core loading pattern. This adaptation significantly reduces the overall neutron leakage relative to a four batch out-in scheme.

    Many of the assemblies contain zones of lower enrichment fuel pins. This enrichment zoning technique is used to decrease the power peaking within an assembly and also to make intra-assembly power distribution insensitive to changes in the lattice pitch (via assembly or rod bowing). Due to the use of various enrichment zoning patterns and shim loadings, the fuel assemblies in the initial core are classified into nine different sub-batches as listed in Table and pictured in Figure 3.1-b, although only five different are shown in Figures 3.1-c and 3.1-d.

3.2 Control Rods

    The control rods are inserted into either two or four of the available CEA guide tube locations. The center guide tube is left vacant in order to accommodate an in-core instrument assembly.

    By combining 4 and 12 fingered CEAs, nearly all fuel assemblies in the core can be rodded with 73 Control Element Drive Mechanisms (CEDMs) as shown in Figure 3.2-a.

    The CEA programming scheme defines five regulating banks, two shutdown banks, and one part-strength bank. The regulating and shutdown bank CEA fingers contain pellets made of B₄C. Schematic drawings of 4 fingered and 12 fingered full length CEAs are shown in Figures 3.2-b and 3.2-c. The part strength rod (PSR), whose fingers contain inconel pellets, are used to assist in maneuvering and for controlling the axial power distribution. Figure 3.2-d shows the part strength control element assembly.

    For the shutdown banks and some control banks, the normal position is fully withdrawn. Over extended periods of operation with these CEAs at one position, flow included vibrations of the CEA fingers can cause wear on the inside of the guide tubes. Therefore these CEAs are periodically within a small range of withdrawn positions. This allows the fully withdrawn CEAs to be moved among seven different positions (steps) on a rotating basis, usually once a month, thus preventing excessive guide tube wear at any location.

3.3 Burnable Poisons

    Since the core is to be operated essentially with the unrodded state while at power, some provision must be made to offset or "hole down" the excess fuel reactivity present in the core for depletion requirements. If this hold-down were supplied only by soluble boron, the MTC would be positive over a significant portion of the cycle. To avoid this situation, lumped burnable poisons (shims) are used for the reactivity hold-down. In addition to MTC control, the lumped burnable poison plays an important role in controlling the radial power distribution by suppressing reactivity of the high enrichment assemblies. In the KNSP reactor, the shim is comprised in the top and bottom 5% of the shim rods. The axial power is small due to the axial leakage of neutrons.

    Selected assemblies contain up to Gd shim rods placed in strategic locations within the assembly lattice of enriched fuel rods. Gadolinia is used as a burnable poison in the initial core design to reduce soluble boron requirements and to provide power distribution control while reducing the thermal margin degradation associated with fuel displacing burnable poisons (e.g., B₄C/Al₂O₃ shims). This integral burnable poison also offers the advantage of the strong initial reactivity hold-down characteristic of gadolinium in combination with a low residual poison worth at end-of-cycle. This design uses pellet concentration of 4 w/o gadolinia admixed in natural urania. Reactivity control is provided using relatively few (e.g., 4 or 8) gadolinia-bearing fuel rods per fuel assembly in the fuel assemblies requiring burnable poison. The use of natural UO₂ provides flexibility in the range of concentration of Gd₂O₃ which determines the duration of the reactivity hold-down, while eliminating licensing concerns associated with fuel performance using gadolinia in enriched urania. Unlike the typical B₄C burnable absorber rods, the gadolinia fuel rods (shims) have the same dimensional specifications as the normal fuel reds. Figure 3.3-b illustrates the arrangements of burnable poison within each poisoned assembly. Figure 3.3-c shows changes in assembly multiplication factor for varied poison contents in an assembly.

3.4 Grids

    The fuel spacer grids maintain the fuel rod array by providing positive lateral restraint to the fuel rod but frictional restraint to axial fuel rod motion. The grids are fabricated from preformed zircaloy or inconel strips (the bottom spacer grid material is inconel) interlocked in an egg crate fashion and welded together. Each cell of the spacer grid contains two leaf springs and four arches. The leaf springs press the rod against the arches to restrict relative motion between the girds during a refueling operation. Zircaloy-4 is used for guide tubes and for girds located in the active fuel region of the fuel assembly because of its superior neutron economy properties. The ten Zircaloy-4 spacer grids are fastened to the Zircaloy-4 guide tubes by welding, and each grid is welded to each guide tube at eight location, four on the upper face of the grid and four on the lower face of the grid, where to the guide tubes due to material differences. It is supported by an Inconel 625 skirt which is welded to the spacer grid and to the perimeter of the lower end fitting. Figure 3.4 shows schematic of spaer grid.

3.5 Instrumentation

     There are 45 in-core instrument (ICI) assemblies, each containing five self-powered detectors. The 45 instruments are strategically disributed about the reactor core, and the five detectors. The 45 instruments are strategically distributed about the reactor core, and the five detectors of each instrument assembly are axially distributed along the length of the core. The individual detectors are centered at 10, 30, 50 ,70 and 90% of the core height. This permits gross three dimensional power distribution mapping of the core from 10% to 125% of full power.

    A complete in-core instrument assembly consists of five rhodium detectors, a background detector, a core exit thermocouple, and a calibration tube. The assembly is enclosed in a protective sheath, and terminated with a seal plug and an electrical connector. The individual ICI assemblies are positioned in the center guide tubes of the fuel assemblies at the 45 core locations. The core instrument pattern is optimized for the monitoring of azimuthal power tilts.

     The core exit thermocouple is a type K, grounded junction element with chromel and alumel lead wires in accordance with ASTM standards. The thermocouple extends to the end of the ICI assembly and measures the primary coolant flow outlet temperature.

    Typical in-core instrumentation assembly schematic is shown in Figure 3.5-a.

    There are also three different kinds of ex-core detectors whose ranges overlap a minimum of two decades to assure that the neutron flux is continuously monitored from source level to 200% of full power. These detectors are used in three separate electronic "channels" as follow:

Startup Channel (BF₃ proportional counters), Control Channel (dual-section uncompensated ionization chambers), and Safety Channel (three-section U235 fission chambers). Typical arrangement of In-core and Ex-core Detector system are shown Figure 3.5-b.

Summary description of the reactor core and core components are given Table 3-1 through 3-5.

4. CORE DESIGN BASES

4.1 General

    In general, the specifications of the reactor core and components are determined in the core design process, especially in the nuclear design process. It is, thus, worth understanding what are the purpose of the core design. The major objectives of core design can be summarized as follows;

1) Meet the energy production requirement (rated power x duration)

2) Meet the design criteria to ensure safety of the core and fuel

3) Maximize operational flexibility

4) Minimize the power generation cost for economics

    The bases for the core design, especially nuclear design of the fuel and reactivity control system are very important to meet the above objectives. Therefore, the core design process starts from setting up of the proper design bases and requirements. In the following subsections, some design bases closely related to the fuel and reactivity control systems designs are discussed. The General Design Criteria mostly related with core design, as an example, are listed in Table 4.1 These criteria are similar to the criteria defined in the Korean Automic Law and Regulations.

4.2 Excess Reactivity and Fuel Burnup

    The excess reactivity provided for each cycle is based on the depletion characteristics of the fuel and burnable poison and on the desired burnup for each cycle. The desired burnup is based on an economic analysis of the fuel cost and the projected operating load cycle for KNSP. The average burnup is chosen to ensure that the peak burnup is within the limits set by LOCA analysis. This design basis, along with the design basis in Subsection 4.8 satisfies General Design Criterion (GDC) 10. For KSNP, the peak pin burnup limit is recently increased and licensed from 52,000 MWD/MTU to 58,000 MWD/MTU to accommodate extended cycle operation of 18 months for subsequent cycles. The initial core average burnup amounts to 13,650 MWD/MTU.

4.3 Core Design Lifetime and Fuel Replacement Program

    For early KSNP designs, the core design lifetime and fuel replacement program are based on approximately annual refueling with approximately one-fourth of the fuel assemblies replaced at each refueling in later cycles. The first cycle design lifetime is longer than later cycles to permit a more orderly transition to equilibrium cycle conditions. This four-batch core design concept has recently been changed to three-batch such that 18 months cycle operations can be achievable from the initial core to subsequent cycles, to further enhance plant lifetime economics.

Table 4.3 summarizes the core performance characteristics of KNSP.

4.4 Negative Reactivity Feedback

    In the power operating range, the net effect of the prompt inherent nuclear feedback characteristics (fuel temperature coefficient, moderator temperature coefficient, and moderator pressure coefficient) tends to compensate for a rapid increase in reactivity. The negative reactivity feedback provided by the design as shown in Figure 4.4 satisfies General Design Criterion 11.

4.5 Reactivity Coefficients

    The values of each coefficient of reactivity are consistent with the design basis for net reactivity feeback (Subsection 4.4), and analyses that predict acceptable consequences of postulated accidents and anticipated operational occurrences, where such analyses include the response of the reactor protection system (RPS). Table 4.5 and figure 4.5 show the reactivity coefficients of KNSP.

4.6 Burnable Poison Requirements

    The burnable poison reactivity worth provided in the design is sufficient to ensure that the moderator coefficients of reactivity are consistent with the design bases in Subsection 4.5. For KSNP, burnable poison, namely Gd₂O₃ with UO₂ is utilized. Gd₂O₃ with UO₂ is so called integral burnable poison. Gd₂O₃ has been extensively used for reload cores of Korean PWRs as burnable poison material. It's properties are well known and performance in the reactor core is well proven. Typical burnable poison loading in a fuel assembly is shown in Figure 3.1-b.

4.7 Stability Criteria

    The reactor and the instrumentation and control systems are designed to detect and suppress xenon-induced power distributions that could, if not suppressed, result in conditions that exceed the specified acceptable fuel design limits (SAFDL). The design of the reactor and associated systems precludes the possibility of power level oscillations. This basis satisfies General Design Criterion 12. During the power ascension test period, the xenon oscillation control test is performed. Through this test, it is demonstrated that the reactor is properly controlled under the oscillatory xenon conditions. A typical xenon control by PSCEA for KSNP is shown in Figure 4.7

4.8 Maximum Controlled Reactivity Insertion Rate

    The core, control element assemblies (CEAs), reactor regulating system, and boron charging portion of the chemical ans volume control system are designed so thea the potential amount and rate of reactivity insertion due to normal operation and postulated reactivity accidents do not result in the following:

a. Violation of the specified acceptable fuel design limits

b. Damage to the reactor coolant pressure boundary

c. Disruption of the core or other reactor internals sufficient to impair the effectiveness of emergency core cooling system

    Typical reactivity insertion rate curve for reactors similar to KSNP is shown in Figure 4.8. This design basis, along with Subsection 4.12, satisfies General Design Criteria 25 and 28.

4.9 Power Distribution Control

    The core power distribution is controlled such that, in conjunction with other core operating parameters, the power distribution does not result in violation of the limiting conditions for operation. Limiting conditions for operation and limiting safety system settings are based on the accident analyses such that specified acceptable fuel design limits and other criteria are not exceeded for accidents. This basis, along with Subsection 4.2, satisfies General Design Criterion 10. A typical power shape due to reactor daily load swing is shown in Figure 4.9.

4.10 Excess CEA Worth with Stuck Rod Criteria

    The amount of reactivity available from insertion of withdrawn CEAs under all power operating conditions, even when the highest worth CEA fails to insert, will provide for at least 1.4% excess CEA worth after cool-down to hot zero power, plus any additional shutdown reactivity requirements assumed in the safety analyses. This basis, along with Subsection 4.11, satisfies General Design Criteria 26 and 27. Table 4.10 compares the available CEA worths with allowances.

4.11 Chemical Shim Control

    The chemical and volume control system (CVCS) is used to adjust the dissolved boron concentration in the moderator. After a reactor shutdown, this system is able to compensate for the reactivity changes associated with xenon decay and reactor coolant temperature decreases to ambient temperature, and it provides adequate shutdown margin during the refueling. This system also has the capability of controlling, independently of the CEAs, long-term reactivity changes due to fuel burnup and reactivity changes during xenon transients resulting from changes in reactor load. This design basis, along with Subsection 4.10, satisfies General Design Criteria 26 and 27.

4.12 Maximum  CEA Speeds

    Maximum CEA speeds are consistent with the maximum controlled reactivity insertion rate design basis in Subsection 4.8 The maximum CEA speed for safety analysis is 30 in/min.

                                                      Table 3-1 Core Arrangement

 

Number of Fuel Assemblies in Core, total	177
Number of CEAs	73
Number of Fuel Rod Locations	41772
Number of Shim Rods	640
Spacing between Fuel Assemblies, Fuel Rod Surface to Surface, cm	0.528
Hydraulic diameter, nominal channel, cm	1.198
Total Flow Area (excluding guide tubes), m2	4.153
Core Equivalent Diameter, cm	312.4
Core Circumscribed diameter, cm	330.2
Total Fuel Loading, (all rod locations fuel rods), ton U	76.34

                                                         Table 3-2 Fuel Assemblies

 

Assembly Type	No.of Assemblies	wt% U-235	No.of Fuel Rods per Assembly	No.of Gd Rods per Assembly	Gd2O3 wt% in Nat. UO2
A	45	1.28	236	-	-
B	20	2.34	236	-	-
B1	8	2.34/1.28	176/52	8	4
B2	16	2.34	232	4	4
C	12	2.84/2.34	184/52	-	-
C1	32	2.84/2.34	176/52	8	4
D	12	3.34/2.84	184/52	-	-
D1	8	3.34/2.84	176/52	8	4
D2	24	3.34/2.84	128/100	8	4

                                                   Table 3-2 Fuel Assemblies (cont'd)

 

Fuel Rod Array Square	16 × 16
Square Grid
Type	Leaf Spring
Material	Zircaloy-4
Number per Assembly	10
Bottom Spacer Grid
Type	Leaf Spring
Material	Inconel 625
Number per Assembly	1
Weight of Fuel Assembly, kg	654

                                                              Table 3-3 Fuel Rods

 

Fuel Rod Material (sintered pellet)	UO2
Pellet Diameter, cm	0.826
Pellet Length, cm	0.991
Pellet Density, g/cm3	10.44
Clad Material	Zircaloy-4
Clad ID, cm	0.843
Active Length, cm	381
Plenum Length, cm	24.5

                                                 Table 3-4 Control Element Assembly

 

	Full Length	Part Length
Number	65	8
Absorber Elements, No. per Assembly	12 and 4	4
Type	Cylindrical	Cylindrical
Clad Thickness, cm	Inconel 625	Inconel 625
Clad O.D., cm	2.073	2.073
Control Element (CEA Fingers)
Poison Material (main/lower end)	B4C/Felt metal	Inconel 625
	and reduced dia. B4C	Slug
Poison Length, cm	344.17/31.75	375.92
B4C Pellet
wt % Boron, minimum	77.5	N/A

                                                        Table 3-5 Burnable Absorber

 

Absorber Material	Gd2O3-UO2
Pellet Diameter, cm	0.826
Pellet Density (% theoretical)	95.25
Theoretical density, UO2, g/cm3	10.96
Theoretical density, Gd2O3, g/cm3	7.41
Clad Material	Zircaloy-4
Clad ID, cm	0.843
Active Length, cm	342.9
Plenum Length, cm	24.53

                                                           Table 4.1   Design Criteria

     10CFR50 APP.A

                                      Table 4.3   CORE PERFORMANCE CHARACTERISTICS

 

Total Heat Output, MW(t)	2815
Heat Generated in Fuel, %	97.5
Reactor Coolant Temperature, °C
Hot Zero Power	296
Hot Full Power	296
Design Core Average, Hot Full Power	312
Nominal Primary System Pressure, kg/cm2	158
Average Liner Heat Rate, kW/ft (W/cm)	5.391 (176.9)
Specific Power, kW/kgU	36.88
Volumetric Power Density, kW/liter of core	96.26
1st Cycle Length, EFPD	370
Core Average Burnup, MWD/MTU	13650
Batch Average Burnup, MWD/MTU	~43500
Peak Pin Burnup, MWD/MTU	<52000

     APPENDIX

A. Advanced Core Design Code System - MASTER

A.1 Limitations in Applications and Inefficiency of the Commercial Design Code System

• Limitation in Code Application to Advanced Reactor and Core Design/Analyses

- Complicated Geometry in Fuel & Core Structure (Cartesian and Hexagonal)

- Mixed Burnable Poison Fuel

- High Burnup Core

- MOX Fuel Loaded Core

• Limitation and Inefficiency in Code Maintenance and Upgrade

- Limitation due to Complicated Evolution History

- Inefficiency due to Complicated Design Procedure and Code System

- Incompatibility between the Commercial Code System

A.2 Basic Code Functional Requirements

• Utilization of Commercially Available Cell Code

• Microscopic Depletion over Macroscopic Depletion

• Seamless Integration of Steady State Dimensional Depletion / Transient Core Condition

  Simulation and Detailed Thermal Hydraulic Feedbacks

• Use of First Principle Models and Minimized Bias

• Higher or Comparable Code Accuracy and Efficiency over the Commercial Code Systems

• Versatile Code Applicability to Various Type of Reactor Analyses

A.1 FUNCTIONAL FEATURES

 

FUCTION	FEATURES
GENERAL
• Geometry	Rectangular 3-D, 2-D for 1/1, 1/2, 1/4 Core Analysis
• Group	2-Group Diffusion Theory
• Calc. State	Steady State Depletion and Transient Core Analyses with free switching-over
• Incore Instruments	Modeling for Fixed and Movable Incore Detectors
• Burnable Poison	Core Modeling with more than 2 Different Types of Burnable Poisons
• Fuel Type	UO2 and MOX with Extended Depletion Chain and (n,2n) Treatment
• User Friendliness	Advanced I/O Structure with I/O Identifier (Group-Block Identifier Structure)
	Summary Output for Frequently Used Design Parameters
	Restart and Shuffling Files
	Automatic Design Parameter Generator
CROSS SECTIONS
• Interface Program	PROLOG
• Source	CASMO or HELIOS
• Type	Microscopic
• From	Table and Function
NEUTRONIC CALC.
• Methods	Response Matrix From of NEM, NIM, AFEN with Multi- Level & CMR
	Non-liner From of NEM, ANM with Wielandt / Krylov Subspace Method
• Pin-powers	Analytic Function Expansion Method
• Reflectors	Effective Reflector X-sections or 2-Group Albedo
• Burnup Correction	Polynomial Function Expansion Method
DEPLETION CALC.
• Type	Microscopic Depletion
• Chains	Multiple Treatment of BP
	Extended Chain and (n,2n) Treatment
• Methods	Full or Semi- Predictor Corrector Method with Fixed Factor for Heavy Nuclides
	Full Predictor-Corrector with Fine Time Step for Poison
TRANSIENT CALC.
• Time	Implicit Euler Method with Fine Frequency Transformation
• Source	Analytic Integration
• Nodal Correction	Automatic Nodal Correction Factor with Time Advancement
T/H CALC.
• Model	Single Phase, Energy Conservation by COBRA-3CP
• Fuel Temperature	Fuel Temperature Table as a Function of Fuel Burnup and Power
OTHERS
• Adjoint Solution	Yes
• Xenon Dynamics	Yes
• X-section Adapt.	Yes
• Searches	Power, Control Rod, Boron, Eigenvalue
• Haling Depletion	Yes
• Shape Matching	Yes
• Automatic Design    Parameter Gen.	Yes

A.2 SPECIFICATIONS

 

Item	Specification
GENERAL
• Language	F90
• Platform	WS od PC
• Program Size	~85,000 Executable Lines (Neutronics : ~60,000 Lines T/H : ~25,000 Lines)
• Memory	Execution File Size : ~2.7 MB Required Memory : ~30 MB (for 3-D, 1/1C, 4Box/FA Modeling)
• Restart File	~10MB max / single burnup step (for 1/1C, 4Box/FA Modeling)
• Speed1	~10 CPU Sec. max / single burnup Step (for 1/1C, 4Box/FA Modeling)
• GUI	Quick Win based On-line Graphics

   (¹) Speed is generally based in HP775-125MHz

   Figure 8.1 Comparison of Assembly Power Distribution

 

Y/Z	H	G	F	E	D	C	B	A
	.729	1.281	1.422	1.193	.610	.953	.959	.777	--a
	-.002	-.005	-.005	-.003	.000	.000	.002	.002	--b
8	-.002	-.001	.002	-.001	-.001	-.001	-.002	-.003	--c
		1.397	1.432	1.291	1.072	1.055	.976	.757
		-.006	-.006	-.004	-.002	.000	-.001	.002
9		.001	-.001	.000	-.002	-.001	-.004	-.005
			1.368	1.311	1.181	1.089	1.000	.711
			-.005	-.004	-.002	.000	.001	.001
10			-.001	.000	.001	-.001	-.007	.002