ML19274D650

Forwards Addendum 1 to HTGR Preliminary Safety Evaluation & Environ Info Document.Describes Proposed Commercial Gas Turbine Plant Using Helium Turbine Power Conversion Sys, Safety,Licensing & Environ Considerations

ML19274D650
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/14/1979
From:	Turner R EMVGA
To:	Lipinski W, Speis T ARGONNE NATIONAL LABORATORY, Office of Nuclear Reactor Regulation
References
NUDOCS 7902210221
Download: ML19274D650 (30)

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.,a N tr iFortNtA 92138 mo 4:53000 February 14, 1979 Dr. Walter C. Lipinski, Manager NASAP Safety / Licensing Assessment Reactor Analysis 6 Safety Division Argonne National Laboratory 9700 S. Cass Avenue Argonne, Illinois 60439 Mr. Thomas Speis Office of Nuc1 car Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Centlenen:

Attached is an appendix to the llTGR Preliminary Safety Evaluation and Environmental Information Document. This infor-cation is in response to the request forwarded by Walt Lipinski on February 9, 1979.

Sincerely, k

R. F. Turner, Manager Fuel Cycles G Systems Dept.

RIT/dn Attachment cc: J. Cleveland - OPNL T. E. Collins - DOE OY

\ '\\ (9 0 k kou 7 9 0 2 210 AA l-

f (2/14/79)

PRELD!INARY SAFLTY AND ENVIRORIENTAL INFORMATION DOCUMENT ADDENDUM 1 GAS TURBINE HIGR

1. INTRODUCTION

2. PLANT DESCRIPTION 2.1 PLANT LAYOLT 2.2 POWER CONVERSION LOOP COMPONENTS 2.2.1 HELIGi TURB0 MACHINE 2.2.2 HEAT EXCHANGERS

3. SAFETY AND LICENSING CONSIDERATIONS 3.1 HIGR-SC ISSUES APPLICABLE TO UTCR-GT 3.2 SAFETY AND LICENSING ASPECTS OF THE HTGR-GT

4. ENVIRONMENTAL CONSIDERATIONS

5. CORE AND FUEL FEATURES FOR THE HTGR-GT 5.1 CORE DESCRIPTION

5.2 DESCRIPTION

OF THE PIACTIVITY CONTROL SYSTDi

PRELIMINARY SAFETY AND ENVIRONMENTAL INFORMATION DOCUMENT ADDENDUM 1 GAS TURBINE HTGR

1. Introduction A program to design and develop a commercial gas turbine HTGR (HTGR-GT) power plant has been underway at General Atomic Company (GA) for several years with support from DOE, utilities, and manufacturing companics. The approach is based upon proceeding from the current HTCR technology base through a comprehensive power conversion system development program to a dry-cooled nuclear demonstration plant which would be the basis for Commercial Plant design.

The development and utilization of a helium turbine power conversion system operating in a direct cycle,on the hot helium delivered by the HTGR core has been shown to be technically feasibic and to have substantial incentives by international preliminary design studies and by development work, which has been underway since 1970. The work in the U.S. has been accomplished by: General Atomic, Power Systems and Prutt & Whitney Divisions of United Technologics Corporation, and General Electric Gas Turbine Projects Division, and has been supported by DOE, utilities and manufacturers.

In Europe, the program has been established as the High-temperature Helium Turbine (HHT) project led by GA affiliates, Hochtemperatur-Reaktorbau (HRB) and Kernforschungsanlage (KFA), with major industrial participation and supported by the Federal Republic of Germany (FRG) and the Swiss government.

Cooperation between the CA and HHT projects was initiated in 1973, and is currently conducted under an exchange agreement for activitics The gas turbine industry has already established as state-of-the-art for heavy-duty gas turbines the level of temperatures, unit frame size, and much of the basic technology needed for high-temperature helium turbines suitable for use with advanced HIGR-GT's.

The HTGR-GT offers major improvements in plant simplification, lower capital cost, increased efficiency, and waste-heat rejection. Heat rejection is either by economical dry-cooling towers, wet / dry combination, or optionally by a low temperature secondary rankinc power cycle (HTCR-GT binary-cycle plant) that generates additional power with subsequent wet / dry or wet-cooling heat rejection. At 8500C turbine inlet temperature, the dry-cooled plant will have 40% cfficiency and the binary cycle 48%.

The HTCR-GT plant combines the existing HTGR core with closed-cycle helium gas turbine power conversion loops which operate on the reactor coolant helium. The power conversion loops (PCLs) are integrated into the prest';essed concrete reactor vessel (PCRV) on the basis of both safety and economic considerations; thus, the necessity of providing for burst protection of large external metallic pressure vessels and ducts is avoided. The PCRV is located within a secondary containment building which, together with the PCRV, incorporates safety features to limit loss of primary coolant and to limit missile damage in the event of failures in the turbonachinery, shaft scals, generator, heat exchangers, and other components.

2. Plant Description 2.1 Plant Layout The power plant plot plan concept shown on Fig. 2.1 illustrates the general layout of buildings and dry cooling towers for a twin 3000 MWt HTCR-GT plant. The react >r service building and fuc] storage facilities are shared by the two reactor units. Each unit has a separate control building and safety-related auxiliaries. A runway system is provided for turbonachinery and generator handling. Space is allocated on the plot plan for an ammonia turbine building should the binary-cycle option be selected.

Based on the utilization of an existing 3000 MWt core design, the HTGR-GT embodies three power conversion loops, each rated at 1000 MWt. The simplified isemetric diagram of the reactor and primary system (Fig. 2.2) shows the core, turbomachinery, heat exchangers, and entire helium inventory enclosed in the prestressed concrete reactor vessel (PCRV). The isometric view illustrates the integrated approach for the gas turbine plant, and changes to the major components (particularly the precooler) made since Fig. 2.2 was prepared are discussed below.

The salient cycle parameters for the non-intercooled plant are given on the simplified loop diagram (Fig. 2.3) . As shown in this diagram each loop includes a single-shaft gas turbine, a recuperative gas-to-gas heat exchanger, and a precooler (gas-to-water exchanger) for cycle heat rejection.

As shown in the plan view of the PCRV cn Fig. 2.4 the three power conversion loops are located symmetrically around and below the central core cavity.

The three turbenachines are oriented in a delta arrangement, nad the heat exchangers are installed in vertical cavitics within the PCRV sidewalls, two for each loop. This orientation of the major components results in a minimum PCRV diameter, this being economically important since the vessel is the largest cost item in the plant. The elevation views through the PCRV shown on Figs. 2.5 and 2.6 illustrate the helium gas flow path within the primary system. The components are connected by large internal ducts within the PCRV. The horizontal turbomachine cavities are located directly below their associated loop heat exchangers. The recuperator is positioned directly above the turbine exhaust, and the precooler is above the compressor inlet. A summary of the main features of the direct cycle HTGR power plant is given on Table 2.1.

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2.2 Power Conversion Loop Components i

2.2.1 Helium Turbonachine Preliminary design of the turbonachinery for the HIGR-GT plant has been donc by the Power Systems Division and Pratt and Whitney Aircraft Division of United Technologies Corporation. A simple and rugged arrangement consisting of a single-shaft, direct-drive turbonachine was chosen for the HTGR-GT. A simplified cross section of the 400 MW(c), 60-Hz machine is shown in Fig. 2.7 and the main features are outlined in Table 2.2. The design and high performance predictions for this machine reflect the influence of technology from demonstrated advanced-technology industrial gas turbinc.

The 400 MW(c) helium turbonachine has 18 compressor stages (for a pressure ratio of 2.5 with the low molecular weight gas) and 8 turbine stages. The rotor is of welded construction. Welded rotors have a long, successful history in Europe foi both gas and steam turbines. With the 60,800 kg (67-ton) rotor supported on two journal bearings (with state-of-the-art loading and peripheral speed), the overall length of the machine is 11.3 m (37 ft.). The overall dicmeter of 3.5 m (11.5 f t) is a design constraint to facilitate rail transportation of a contaminated turbonachine installed in a shielded cask. The overall machine weighs 277,000 kg (305 tons).

Rotor burst protection is incorporated in the machine design in the form of burst shicids around the compressor and turbine rotor-bladed sections (Fig. 2.7). Man-access cavities are provided in the PCRV for inspection and limited maintenance work on the journal bearings, which are of the multiple, tilting-pad, oil-lubricated type. The spaces in which the bearings are located are is ted from the main cycle working fluid by shielding (purged gas froa the purification system is used to give an acceptable radiological environment for man access). The drive to the generator is from the compressor end of the turbomachine, and the thrust bearing is located external to the PCRV to facilitate inspcetion and maintenance.

For a single-shaft helium turbonachine with a net power output of 400 MW(e), the rotating section is ccmpact and is substantially smaller than an equivalent air-breathing machine because of the high degree of pressurization (particularly at the turbine exit) and because the enthalpy drop in the helium turbine is many times greater (i.e., increased specific Table 2.1 MAIN FEATURES OF NUCLEAR CLOSED-CYCLE CAS TURBINE PLANT o Power Plant Life, 40 Years e Plant Availability. 80 Percent ,

e 3000 FM(t) Core Thermal Rating e 40 Percent Efficiency with Dry Cooling e 48 Percent Efficiency with A=:onia Bottoming Cycle e Reference Design Based on:

- Optimized Parameters for Minimun Overall Cost of Power Integrated Configuration 850*C (1562*F) Reactor Outlet Temperature Multiple Gas Turbine Loops Non-Intercooled Cycle High Degree of Heat Recuperation

- State-of-the-Art Technology, Materials and Fabrication Methods t ajor Components Within Rail Transportation Limits e Design Philosophy:

- tbke !bximun Use of Existing HTCR and Gas Turbine Technology Simplify Systen Even at the Expense of Slight Efficiency Penalty Use Conservative Design Parameters Stress Levels in Major Power Conversion Loop Components Commen-curate with Full Plant Operating Life of 280,000 Hours Major Attention Given to Safety, Reliability, and Maintenance in the Conceptual Design Phase O

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TABLE 2.2 DETAILS OF 400 MW(e) (60-itz) SINGLE-SIIAFT llELlUM GAS TURBINE Compressor Turbine No. of Stages 18 8 .

Ilub Diam, in. (mm)

First Stage 62.0 (1575) 66.6 (1691)

Last Stage 62.0 (1575) 62.6 (1590)

Tip Diam, in. (mm)

. First Stage 71.9 (1826) 76.5 (1943)

Last Stage 68.3 (1735) 86.0 (2184) llub/Tip Ratio, First/Last Stage 0.86/0.91 0.87/0.73 Blade lit, in. (mm) First/Last 4.95 (126)/3.15 (80) 4.95 (126)/11.7 (297)

Blading Adiabatic Efficiency, % 89.8 91.8 Overall Machine Length, ft (m) 37 (11.3)

Machine outer Diam, ft (m) 11.5 (3.5)

Rotor Wt, tons (kg) 67 (60,800)

Stator and Case Wt, tons (kg) 238 (216,000)

Total Machine Wt, tons (kg) 305 (276,800) 1 Type of Rotor Construction Welded Turbine Blade Material Nickel-Base Alloy (IN 100)

Rotor Burst Shield Integral part of nachine structure 1

Jottrnal Bearing Man-Access For inspcction and limited maintenance Bearing Details Number of journal bearings 2 Type of journal bearings 5 pad, tilting pad oil lubricated Thrust bearing type 8 pad, tilting pad, double acting Thrust bearing location External to PCRV I -

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power). The external dimensions of the 400 IN(c) helium gas turbine are similar to those of an air-breathing, advanced, open-cycle industrial gas turbine in the 100 MW(c) range. The fact that the helium turbine (particularly the rotor assembly and casings) is comparable in size with existing machines substantiates the claim that conventional fabrication methods and facilitates can be used.

The turbonachinery is coupled to an all water cooled generator which recently has been relocated from outside to inside the secondary ccntainment in order to climinate any shaft penetration of the secondary containment.

2.2.2 Heat Exchangars Tubular types of construction were selected for both the recuperator and precooler in the direct cycle plant. The main reasen for this selection was that it represents the only type of construction that has been proven to have the structural integrity for long-life electrical utility power service.

Initially straight tube axial counterflow configurations were selected for both the recuperator and precooler, and this is reflected in the isometric of the primary shown on Fig. 2.2. The current recuperator in the reference plant design is of straight-tube design and embodies a modular assembly having many heat transfer elements. For this gas-to-gas heat exchanger, inspection and repair is down to the nodule level.

In the case of the gas-to-water precooler concern had been expressed regarding the very large nuuber of small diameter tubes associated with the straight tube design. Recognizing the increasing importance of maintenance and inservice inspection on exchanger design, a precooler embodying a helical bundle geonctry is in the early stages of design and evolution. The large reduction in tube number associated with this flow configuration enables inspection and repair to be done down to the individual tube level, rather than the module level as in the straight tube variant. In the plant layout concept shown on Fig. 2.5 the helit 1 precooler design is shewn installed in the PCRV. Ucat exchanger dimensions and weights are given on Table 2.3.

A ground rule for the heat exchangers is that they must be designed to operate for the full life of the plant. Both units will be lowered into the PCRV cavities by a system of hydraulic jacks during construction and they are expected to remain in place during the life of the plant. In both exchanger designs provision is made for replaceability and maintenance and repair. In the case of a faulted heat transfer element (i.e., module in the case of the recuperator, and a tube in the helical precooler assembly) plugging will be performed from outside the PCRV.

Even though the single-phase working fluids (helium and water) can realize relatively high-heat transfer coefficients, large surface areas are necessary because of the high thermal conductance requirements associated with the large heat transfer rates. However, the modest metal temperatures and internal pressure differentials, compared to modern steam generators, permit the use of code-approved lower-grade alloys of reduced cost. The ferritic materials selected for both exchangers have been used extensively in industrial and nuclear plant heat exchangers. Though the exchantyr assemblics are large, state-of-the-art manufacturing methods can be used, and the modular approach in the case of the recuperator cases the fabrication, handling and assembly. The overall size and weight of both the recuperator and preccoler are similar to contemporary steam generators, transport methods, handling and installation techniques developed for these units will be equally applicable to the heat exchangers for the HTGR-GT.

TABLE 2.3 _IIEAT FXCIIANGER DETAILS FOR liTCR-GT

!!cchanical betails Recuperator Precooler I!X Construction Counterfl.ov Multipass Cross Counterflow Construction Straight tube, Modular 1:elical Fin-Tube Tubes / Exchanger 66,483 827 Tube /D, rm (in. ) 11.1 (0.4375) 28.6 (1.125)

Eundle Diameter, n (f t) 5.33 (17.5) 4.88 (16)

Cavity Dianeter, n (f t) 5.64 (18.5) 5.18 (17)

Overall IIX I!cight, n (ft) 22.4 (73.5) 22.4 (73.5)

Approx. Ucight (ton) 544 (600) 590 (650) 3.3 Cycle Parameters The plant performance is based on ISOA day conditions (15 C(59 F))

and assumes heat rejection to the atmosphere via a natural draft dry cooling tower. Figure 2.3 illustrates the cycle diagram for the 3000 IN(t) HTGP.-GT.

Table 2.4 gives cycle conditic>ns around the loop for the dry cooled cycle.

If an ammonia bottoming cycle is added plant ef ficiency increases to 47.9%.

Table 2.5 gives the cycle conditions around the loop for the binary cycle.

• International Standards Organization TABLE 2.5 3000 FN(t) - Single Twin llTCR-GT Cycle Performance Perameters (Binary Plant - Bottoming Ammonia Cycle)

PRESSUPI TEMPERATURE FLOW / LOOP PSIA DEGREES F LB/IIR REACTOR INLET 1124.579 966.079 13783467 REACTOR OUTLET 1115.205 1562.101 13783467 DUCT INLET 1115.205 1562.301 4594489 DUCT DUTLET 1108.971 1562.000 " '4594489 TURBINE INLET 1108.971 1560.423 4601055 TURBINE OUTLET 480.209 1025.405 4647956 DUCT INLET 480.209 1023.033 4671315 DUCT OUTLET 47/.056 1022.942 4671315 RECUPERATOR HOT INLET 477.056 1022.730 4647958 RECUPERATOR HOT GUTLET 470.170 519.060 4647958 DUCT INLET 470.170 518.783 4671315 DUCT OUTLET 469.231 518.758 4671315 PREC00LER INLET 469.231 518.290 4659637 PREC00LER OUTLET 462.177 153.000 4659637 DUCT INLET 462.177 153.000 4671315 DUCT OUTLET 460.000 153.000 4671315 COMPRESSOR INLET 460.000 154.237 4690075 COMPRESSOR OUTLET 1150.000 456.883 4690075 DUCI INLET 1150.000 456.869 4599088 DUCT OUTLET 1139.484 456.744 4599088 RECUPERATOR COLD INLET 1139.484 456.708 4599038 RECUPERATOR COLD OUTLET 1128,133 966.128 4599088 DUCT INLET 1128.333 966.128 4599088 DUCT OUTLET 1124.579 966.079 4599088 COMBINED PLANT PRIMARY PLANT OUTPUT 1081.32 IN SECONDARY PLANT OUTPUT 378.06 FN ~

AUXILI Ar1Y POWER:

PRIMARY PLANT 11.00 N SECONDARY PLANT 11.90 IN NET OUTPUT 1436.48 trJ PLANT EFFICIENCY 47.88

3. Safety and Licensing Considerations As discussed elsewhere in the Steam Cycle section of this report, the HTGR-GT has a number of important features in common with the HTGR-SC.

Principal of these are the use of the prestressed concrete reactor vessel, the prismatic graphite core with encapsulated fuel particles, and the use of three independent auxiliary cooling loops. Safety features such as the control rod reserve shutdown system and liner cooling system are essentially identical. In addition, the three major inherent safety features of the HTGR-SC are also inherent in the HTGR-GT; namely:

1. The large mass of graphite in the fuel and reflector blocks which gives the core a very high heat capacity. This feature resists rapid core temperature cl.anges and is highly beneficial in limiting the consequences of design basis accidents.

2. The helium coolant does not cuase reactivity changes as density changes.

3. The enclosure of the entire reactor coolant system within a high integrity PCRV minimizes the possibility of a rupture in the primary coolant system boundary.

Section 3.1 comments on questions and answers of Section 2.4.2 of the PSEID, as they pertain to the HTGR-GT. Section 3.2 discusses major issues unique to the HTGR-GT.

A Preliminary Safety Information Document (PSID) war submitted to the NRC on July 1, 1975. The NRC returned the first Request for Additional Information (RAI) on Round 1 questions on December 15, 1975; the second Round 1 RAI was returned on April 26, 1976. However, by mid-1976 funding and manpower limitations resulted in the termination of significant activity on answering the RAls or further dialogue with the NRC.

3.1 HTGR-SC Issues Applicable to the HTGR-GT In September of 1978, the NRC submitted to DOE a list of questions on eight topics related to the proposed commercial HTGR-SC Lead Plant design. Although these questions and their answers were formulated specifically for the HTGR-SC, a significant portion of the information is applicable to the HTGR-GT.

The following list is comprised of the eight topics addressed in the NRC steam cycle questions, along with a brief description of tha applicability of the answer to the llTCR-GT.

1. Graphite as structural material - the SC answer is directly applicable. The effect of the higher temperatures in the llTGR-GT will have to be taken into account.

2. Core seismic response - the SC answer is directly applicable.

3. Fuel transient response - much of the SC answer is applicable to both GT and SC fuel. The temperature coefficient for the HTGR-GT will be stronger because of the higher average temperature of the graphite.

4. In-service inspection and testing - the criteria for ISI of the llTGR-GT will be based on the same censiderations used to establish the requirements for the llTCR-SC. These requirements are given in the proposed Section XI, Division 2 of the ASME Boiler and Pressure Vessel Code.

5. Low probability accidents - a comprehensive study of low probability accidents for the llTCR has not yet been performed.

The parts of the answer which pertain to control rod ejection, core drop and depressurization of the llTGR-SC should be generally applicable to the llTCR-GT. In addition, the answers concerning research programs, gas-cooled-reactor experience and non-probabilistic criteria are applicable to the HTGR-GT.

6. Containment requirements - the cirteria for containment design requirements are essentially the same for both the SC and GT concepts. Ilowever, the differences in primary coolant in-ventories and other operating characteristics must be taken into account for the HTCR-GT containment design.

7. Primary system integrity - even though many of the cam-ponents internal to the PCRV are not the same , the design con-siderations for the primary coolant systems of both concepts are essentially the sane.

8. Emergency core cooling provisions - the core auxiliary cooling systems (CACS) for the two concepts are essentially the same. The capacities may differ due to the different flow paths for the two concepts. The requirement for containment bachpressure has not yet been fully investigated for the IITCR-GT.

3.2 Safety and 1.icensing Aspects of the liTGR-GT lu addition to the llTGR generic issues discussed above, the llTGR-GT has a number of features that Icad to some new safety and licensing questions. The most significant of these are discussed below,

n. Shaft Seal Failure - The turbomachine/ generator shaft penetrates the primary coolant system boundary. Failure of the seal due to machine er shaft malfunction can potentially cause a rapid depret,surization to the containtxnt. Design features must be in-corporated which ensure that such accidents have an acceptably low probability of occurence,

b. Internal Pressure Equilibration Accidents - Failure of internal components such as the turbomachines or recuperators can cause rapid pressure equilibration within the PCRV. These pressure pulses / transients are much more severe than those associated with the most rapid postulated PCRV depressurization for the llTGR-SC.

The GT pressure equilibration accidents place stringent design requirements on PCRV internals and dictate component designs which may be dif ferent f rom those of the IITGR-SC.

Determination of the consequences of pressure equilibration accidents requires that the failure phenomena he defined, modeled, and verified. This in turn depends on experinental data rclated to failure, as well as experimental or other data that verify the nodeling tools. These modeling tools will include a computer code that describes the transient behavior of the compressible fluid flow following the accident.

Considerabic effort has been expended by the General Atomic Company to develop digital computer programs for the analysis of the transient thermal-fluid behavior of the primary coolant system. One such program, TUBE, was developed specifically to analyze the local consequences of rapid pressure transients. The TUBE program can represent a segment of the primary coolant system in considerable detail, accounting for shock ef fects as well as bends, contractions, expansions, etc. Considerable insight into the local pressure history associated with this type of accident can be accomplished using the TUBE program. Eventually, the analysis of ther,e accidents must be performed with a program that models the entire primary coolant system.

Application of the RATSAM program to the llTGR-GT is being studied.

The ability of RATSAM to model llTGR-GT accidents must be valider d using experimental data and/or comparison with computer programs developed elsewhere.

c. Turbomachine Failures - In addition to causing rapid pressure transients, turbomachine failures can create missiles from whici-protection must be provided. The steam cycle also has the pot <atial for internal missiles generated by circulator failures, but t a mag-nitude of the missile problem for the GT is larger. Analysis of failure consequences has proceeded at General Atomic and United Technologies Corporation as part of the conceptual design of a turbine rotor burst shield and bypass valve. Work has been performed to show that turbine bearing loads following gross machine failure are acceptable. Additional analysis is needed to determine the turbo-machine missile spectrum and the potential damage to the thermal barrier, connecting ducts, peripheral seals, instrumentation lines, and other equipment.

4. Environmental Considerations The environmental ef fects of the  ;-GT have not yet been evaulated in detail. Ilowever, general assessments lead to the conclusion that the radiological and chemical ef fects from normal po.eer operation will be similar to those of the llTCR-SC. The llTCR-GT will have minimal effect on water consumption due to the utilization of dry cooling towers in the basic plant design. The use of dry cooling is more attractive for the llTGR-GT than for competing reactor plants because the higher reject tempe ratures of the llTGR-GT facilitate transfer of heat to the atmosphere thus making the plant site independent of a water supply.

5. Cora and Fuel Features for Gas Turbine llTGR Ihe llTGR-CT is designed to accommodate the same basic core design as the IITGR-SC plant. The same fuel cycle alternatives are available for the two plant designs. The prirary dif ferences in core design and performance characteristics are related to the temperatures of the helium coolant entering and leaving the core.

The average core coolant exit temperature is 850 C (1560 F) for the ilTGR-GT and 692 C (1280 F) for the llTGR-SC design. The core inlet temper-atures are 500 C (930 F) and 318 C (605 F) for the GT and SC designs, respectively. These coolant tewperature differences would result in an increase in peak fuel temperature of about 140 C for a common fuel element design. A fuel element variation being evaluated for the llTCR-GT design utilizen a fuel rod array of 10 rows across the element radius, the same as the Fort St. Vrain fuel element (216 fuel rods / clement) . In contrast, the llTGR-SC large plant design has been based upon an 8-row fuel element (132 fuel rods / element). The 10-row element results in peak fuel temp-eratures in the llTCR-GT which are the same as those for the llTGR-SC with an 8-row element. The trade-of f for using the 10-row element is represented by the fabrication cost for a larger number of fuel rc<ls and a modestly higher core pressure drop.

5.1 Core Description The basic core paraneters are given in Table 5.1. The 3000 MW(t) core contains 534 standard fuel columns and 91 control fuel columns (Fig. 5.1).

Each fuel column consists of eight fuel elements for a total of 5000 fuel elements in the core. The fuel element and the control element are of hexagonal prism shape and their designs are identical to the Fort St. Vrain elements. The control element will contain a small control rod hole (SRC) in addition to the control rod pair (CRP) and the reserve shutdown system (RSS) . The core has a 120 symmetric layout.

The core is controlled, during normal operation with small con-trol rods (SRCs) located in each control column. The SRCs are operated in three banks, where a bank corresponds to a fuel age segment. This means that each bank is uniformly distributed throughout the core, which minimizes power perturbations due to insertion of the rods.

a 5.2 Description of the Reactivity Cor. trol Systen The control systen consists of three mechanisms: control rod pairs (CRP), small control rods (SRC) and reserve shutdown systems (RSS). Each of these controls is located in the central columrt of each region.

f 3000 K4t GT HTGR LAYOUT 77 Id ,

. s2 g 62 yy { s9. ' 6# j c' 3, 63 91 ,

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sabic 5.1 3000 MW(t) HTGR-GT Basic Core Parameters z, , . , ,,..e.. ..~ :, .- ,

Thermal Power - MW(t) 3000 Power Density - kw/l 7.1 No. of Axial Zones 4 No. of Fuel Elecents 5000 No. of Fuel Elements /Colum 8 No. of Fuel Columns Standard 534 Control 91 Core IIcight, m 6.3 Effective Core Diameter, m 8.5 6

Core Volume, 1 3.6x10 4

s, t.

REFERENCES:

1. Asmussen, K.E. and R. Rao " Core Design Study of a Very liigh Temperature Reactor" GA Report GA-A14586, August 1978.

2. Baxter, A. M. and P. A. Iyer, " Core Design Study for an Advanced IITGR",

CA Report CA-A14571, February 1978.

3. Ilanilton, C. J. and R. llackney " Nuclear Analysis of the 4000 MW(t) Gas Turbine llTGR" CA Report GA-A13993, September 1976.