ML20030A352

Chapter 10 to Final Hazards Summary Rept for Big Rock Point, R&D Program

ML20030A352
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	11/14/1961
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
References
NUDOCS 8101090371
Download: ML20030A352 (15)

Text

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.O SECTION 10 RESEARCH AND DEVELOPhiENT PROGRAhi 10.1 GENERAL CONSIDERATIONS 10.1.1 Development Program Objectives One of the purposes of the Big Rock Point plant operation is to accommodate a High Power Density Development Program described in detail in Contract AT(04-3)-361 between the AEC and General Electric Company. The specific objectives of the work to be performed in the Big Rock Point plant are as follows:

10.1.1.1 Irradiation and examination of high power density, long lifetime fuel fabricated by methods with a potential for low fabrication cost.

10.1.1. 2 hicasurennents and tests to describe core and nuclear steam supply system performance. The results will establish reactor operational limits, provide means to verify analytical procedures, and support the conceptual design of a large high power density reactor (300 MWe).

10.1. 1. 3 Installation and o peration of a data logging computer system to periodically record plant data, compute fuel exposure and plant operational data, and to demonstrate the possible economic advantages of an on-line computer 1

system for a nuclear plant.

10.1, 2 Development Program Scope e

In order to meet the above objectives, the program pro-l vides for developmental testing and operation described i

by the following sub-paragraphs. The areas to be treated therein are (1) fuel irradiation and examination, (2) reactor core performance testing, (3) stability and transient per-formance testing, (4) power distribution and physics test-ing, and (5) installation and operation of the on-line data logging computer system. None of this developmental testing will present unique safety hazards. The testing will, however, present a broader than normal series of performance tests while bringing the Big Rock Point reactor to its ultimate design rating. All operating safe-guards criteria will be met during the testing period, and the detailed knowledge of the reactor characteristics pro-vided by this Program will actually result in lessening any hazard of normal reactor operation.

10.1.2.1 Fuel bundles fabricated by processes having a potential for low fuel cost will be irradiated and examined.

This fuel will be irradiated to 10,000 to 15,000 hiWD/T of uranium average exposure and to a core average power density of 45 to 60 KW/ L.

About one-half of the approximately 1;o!o9 obT

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Section 10 Page 2 40 developmental fuel bundles to be irradiated will be operated to a core average power density of 60 KW/L.

Non-destructive examination of the development fuel will be done periodically on site at the Big Rock Point plant. Selected fuel rods will be analyzed after irradiation. This irradiation testing and examin-ation will continue for 41/2 years of developmental operation and will include the insertion, handling, and removal operations required for the test fuel bundles.

10.1. 2. 2 Performance tests of the Big Rock Point reactor core will be conducted. Steady state operating conditions prevailing with a given mode of operation will be meas-ured and evaluated to determine such performance characteristics as burnout safety margin, core exit l

quality, core and channel flow rates, and core steam void content. These tests will be conducted simultane-ously with other performance tests, including stability, and will cover a range of pressures, recirculation flow rates, total reactor power, average power density, and core inlet subcooling. The tests will be conducted in a step-wise approach to the higher power conditions. The tests will consist of measurements of reactor flows and flow distribution, inlet temperatures, neutron flux distri-bution s and core pressure drops. Four instrumented assemblies will be used to produce detailed fuel channel data. These instrumented assemblies will measure fuel channel coolant flow rates, channel and fuel bundle pres-sure drops, and coolant inlet and outlet temperatures.

Neutron flux distribution within the instrumented assemblies will be determined both by wire irradiation and with the in-core ion chambers.

10.1.2.3 Stability and transient performance tests will be con-ducted. Transient recordings of primary system variables under controlled conditions of reactor system excitation will be analyzed and interpreted to define stability margins and load regulation. The excitations to be imposed will include the oscillation of a control rod through pre-deter-mined fixed amplitudes within a range of frequencies from about 1/60 to about 2 cycles per second. Reactor system response will be measured for an excitation of turbine load demand (up to about 30% load chan ge) and for a dis-turbance in reactor pressure (up to 50 psi) by changing the pressure regulator setting. These transient tests will be conducted within a series of operating conditions embracing a defined range of pressures, recirculation flow rates, total reactor power, average power density, and core inlet subecMing. All tests will be conducted step-wise, evaluating the trend Irom previous operating conditions before proceeding to the next test.

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S2ction 10 Page 3 10.1.2.4 Power distribution and physics tests will be conducted.

The axial power profiles of selected fuel bundles will be measured and the radial power distribution across the core will be determined. These tests will be performed at prescribed points in the irradiation history of the core and will consist of wire irradiation data and gamma scans of the fuel. Selected individual bundles will be disassembled and individual fuel rods will be scanned to establish the radial power profiles within a given element.

Physics tests, performed in the cold, xenon free condition,

will be performed to establish that the reactor im sub-critical with one control rod removed, and that the void coefficient of reactivity is negative for voids produced inside of a new f uel bundle..

These physics tests will be conducted only when changes in core configurations impose substantial changes in core reactivity or control rod worth 10.1. 2. 5 The scheduling computer system, which has no reactor control or safety function, will be placed in operation.

The computer and its associated equipment will be used to measure and convert plant signals, calculate operating data and parameters, log measured and calculated data, and alarm failed sensors and out-6f-limit operating variables.

Computer system tests which involve specific reactor operating conditions will consi" primarily of steady plant operation to evaluate methods and accuracy of heat rate calcu!ations, xenon reactivity calculations, and neutron flux di s tribe tion. Specified variations of control rod patterns will be required to evaluate computer treatment of core power distribution and control rod scheduling. All other computer functions, among which are co'ntrol rod, fuel, and ion-chamber exposures, and core performance calculations, will be tested as a part of the normal ope ration of the plant. The computer is not required for normal operation of the plant. It will be purchased and operated under AEC Contract No. A T-( O c., )- 3 61.

10.1. 3 Phases of the Development Program 10.1.3.1 The above definitions of the development program scope in the Big Rock Point reactor (Par.10.1. 2) are directly applicable also as a definition of each of the two major parts of the program, Phase I and Phase II.

Phase I will begin im i ediately following the required engine ering start-up testing program described elsewhere, and will have a duration of possibly six to eight months. Phase II will begin immediately after Phase I is completed and will continue to a time 4-1/2 years after initial criticality of the Big Rock Point reactor. The fea-i tures which distinguish Phases I and II beyond the schedule are given in the following two paragraphs.

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a Section 10 Page 4 10.1. 3. 2 In Phase I the reactor operational limits are defined in relation to the design features of the 157 Mwt core of the Big Rock Pout reactor. The limits prescribed, and which will be observed, are 157 MW reactor thermal power and approximately 45 KW/L average core power density. The desigt. and operating criteria, as presented below in paragraphs 10. 2.1, are based on tested analytical methods and reactor experience embracing the features of Phase I tests and irradiations. Step-wise test procedures, grad-j ually advancing to new conditions of flow, pressure, and power level, will evaluate these criteria and determine warranted changes in test conditions.

4 10.1. 3. 3 In Phase II the reactor operational limits will be extended to correspond to the 240 Aiwt core and to a core average power j'

density of approximately 60 KW/L. The operating and testing criteria will be modified from that of Phase I based on results from (a) out-of-pile thermal-hydraulic tests, (b) V'BWR irradiation and examination, (c) Big Rock Point reactor engineering startup tests, and (d) Big Rock Point reactor Phase I tests. The tests reserved for Phase II involve conditions ~of operation for which analytical methods and reactor experience is required and will be provided from the Pre-operational and Phase I development programs. In the step-wise context of testing being pursued in this program this knowledge will be available before initiation of Phase II work.

10.2 DEVELOPMENT PROGRAM - PHASE I

10. 2. 1 Operating and Testing Criteria - Phase I Within the 157 Mwt and 45 KW/L operating limits, the following criteria will govern all tests.,

10. 2.1.1 Developmental fuel clad stress shall not exceed 70% of yield strength calculated for the end of fuel life,125% of rated heat flux, and 125% of rated reactor pressure. Clad stress shall not exceed 80% of ultimate strength calculated for the end of fuel life at zero reactor pressure.

Test conditions in Phase I will not exceed the maximum fuel temperature criterion given in paragraph 8. 2.1. 2.

Results of this Phase I testing, however, will be utilized to establish criteria for tests permitting center fuel melting during Phase II.

10. 2.1. 2 The burnoat ratio shall be shown by calculation to be at least

1. 5 for all reactor operation test conditions. The burnout ratio is defined as the minimum ratio of the design burnout limit heat flux to the reactor heat flux, as calculated using appropriate allowances for flux distributions, flow distributions, effects of power transients, and uncertainties.

10. 2.1. 3 Test conditions shall be such that the measured steady state in-core flux peak to peak amplitudes shall not exceed +25% of the average operating flux level for a given test. For -

transient tests the flux limit trip will be set at 125% of the i

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4 S2cticn 10 Page 5 Rev 1 (3/23/62) stead 7 state final operating value. The hydraulic stability of the reactor will be analytically demonstrated for each potentially limiting reactor test condition prior to conducting that test.

This will be based upon stability calculations, which will deter-mine such parameters as phase margin in open-loop frequency response analyses. Typical calculations a e reported in GEAP-I 3795, " Consumers Big Rock Point Nuclear Power Reactor Stability Analysis. "

10.2.1.4 The peak-to-peak flux variation will not exceed 1257, of the oper-ating power level during any stability test which may be conducted.

When rod oscillation is a part of the test, the initial amplitude of 4

disturbance will be limited to one control rod notch. When results i

of this limited controlled disturbance give evidence that multiple-rod or multiple-notch oscillation will not exceed the stated flux criteria, these larger disturbances may be applied as part of the stability testing.

10.2.1.5 The reactor will remain suberitical with any single control rod completely removed from the core. This will be determined by test whenever core loading is altered such as to increase the available reactivity significantly.

j 10.2.1.6 The void coefficient of reactivity shall be negative for all voids produced inside a fuel bundle. Reactor tests will be performed to assure compliance with this criterion whenever significant core changes are made.

10.2.2 Operating and Test Conditions,- Phase I 10.2.2.1 The development program will provide four instrumented fuel assemblies and eight developmental fuel bundles during Phase I.

These fuel bundles will replace core elements in the initial 56 element core..The twelve bundles will be located in the core approximately as indicated in Figure 10.1.

Visual examination will be employed periodically to determine integrity and performance of the fuel. Fue' replacement or relocation will be required periodically.

10.2.2.2 The eight development fuel bundles for Big Rock Point Plant f

will be fabricated by two processes, swaged-over pellet and swaged-over powder, and will utilize cost reducing mechan-ical design innovations typified by the sp<cers and free floating fuel rods. These processes and mechanical design features have been incorporated in test fuel bundles that are under con-tinuing irradiation testing in VBWR and have demonstrated their 3

suitability for fu.ther developmental application in the Big Rock Point reactor.

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l Se ctien'10 Page SA Rev 1 (3/23/62) 10.2.2.3 These bundles are scheduled to begin irradiation at the initiation of the Phase I Development Program and will have the same mechanical design features, though differing from the first core fuel design. Salient design features of this fuel are essentially as follows:

Development Fuel Design Features 235 Initial fuel enrichment, approx wt percent U 2.7 UO Weight per bundle, pounds 385' 2

Geometry, fuel rod array 11 x 11 Standard fuel rods per bundle 109 i

Special fuel rods per bundle 12 Spacers per bundle, wire 4

Fuel rod clado.ing material Stainless Steel Cladding thickness, in.

.010 Standard fuel rod diameter, in.

0.426 Special fuel rod diameter, in.

0.320 d

i Active fuel length, in.

70.0 i

Average heat flus at 45 Kw/1, Btu /hr-ft 126,000 l

Ilydraulic diameter 0.580 1

l Figure 10.2 illustrates the current design features of these "i

developmental bundles and additional features are given in detail in GEAP-3851, " Mechanical Design and Testing of j,

Development Fuel for Consumers Power Big Rock Point Reactor," by W. D. Fowler.

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10.2.2.4 The fuel rods, both pellet and powder types, are fabricated with 1/4" tolerance on length and with edge-welded cup-type end plug s. These rods rest on a lower support plate of the bundle and are properly positioned by four axially spaced wire spacers.

They are held in place at the top of the bundle by a grid that latches into position by means of a bolt through the bundle handle a s sembly. Motion of a fuel rod is therefore restricted to the tolerance variations of the rods. Prototype assemblies of this bundle-design have been successfully irradiated to significant l

i Icvels in VBWR at operating conditions simulating those to be

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encountered at Big Rock Point.

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l' Sacti n 10 Page 8 Rev 1 (3/23/62) 10.2.2.5 The four instrumented assemblies will have a fuel design identical to the first core fuel with four center rods replaced by an instrunnent probe. The probes will be directed through a vessel head flange and will.be inserted into the fuel assemblies.

Instrumentation associated with the probe includes a turbine type flowmeter to measure channel coolant flow rate, pressure sensors to measure element and channel pressure drops, thermocouples to measure inlet and outlet coolant temperatures, and ion chambers to measure axial neutron flux levels. The lower orifice assembly and upper handle assembly will be elongated to incorporate the inlet and exit flowmeters. Provision is also made for wire ir-radiation measurements.

10.2.2.6 Results of a recently completed VBWR irradiation testing program with instrumented bundles having similar features will be factored into the final design of the instrumentation associated with these bundle s. The nuclear and thermal-hydraulic characteristics of 3

these bundles will also be essentiz lly the same as for the first core fuel design.

10.2.2.7 The core performance and stability testing will be done within a predetermined sequence of reactor operating conditions.

Table 10. I shows the various combinations of reactor operating l

conditions which are of interest in the development program.

All the tests are shown to be performed at each of three distinct j

opc-ting pressures. The remaining conditions of subcooling,

power level, and total recirculation flow rate for the given core configuration will be demonstrated to be within the criteria established in Section 10.2. I prior to execution of the test.

Variations in control rod configurations and core orificing may l

also be applied as special cases of the outlined series of tests.

Detailed calculations have been inade of the core performance expected for the most critical cases of tests described in Table 10.1, and all operating and testing criteria are met.

Similarly, stability analyses of half-rated, rated, and over-power operation of the 157 Mw thermal core have been performed, plus the rated 240 mw thermal case.

e 10.2.2. 8 Physics analyses and detailed thermal-hydraulic analyses have been performed for all Phase I cases that are outlined in Table 10.1 and the sequence being described should be considered typical, but subject to minor change prior to execution. Such things as actual core reactivity and control, as determined after initial critical tests and power operation, may dictate an interchange in sequence of some of the planned Phase I tests.

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Prior to any series of tests, whether as herein stated or slightly modified, a complete analysis of each test will be per-t.

formed. An appropriate data package will then be submitted

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Sacticn 10 Page 8A Rev 1 (3/23/62) to Consumers Plant Superintendent for approval. The Big Rock Point R&r D Review Committee will be used for advice and consultation to further assure nuclear safety.

l 10.2.2.9 Physics tests of core power distribution will be performed.

l Wire irradiations will be done at steady power operating conditions to yield the power profile in that region of the reactor Gamma scanning of the core will also be done after core.

reactor shutdown to measure the integrated exposure profile of the fuel bundle or of individual rods removed from the bundle.

In addition, the in-core ion chambers will provide continuous power distribution information. The stuck rod criterion and the negative void coefficient will be demonstrated by tests in the cold, xenon free condition. A sufficient number of control rods will be checked by withdrawing them individually to demonstrate that the reactor remains subcritical in each case. Voids will be introduced into a given new fuel bundle with the reactor near critical and the effect on reactivity will be determined by observing the change in reactor period.

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l 10.2.2.10 The tests will con *sist primarily of bringing the reactor to specified operating conditions and measuring the nuclear, thermal and ' hydrodynamic characteristics in each of those modes of operation. The fuel bundle instrumentation and the computer will augment normal plant instrumentation for this purpose. Transient hydrodynamic performance and stability will be studied by introducing controlled disturbances such as control rod oscillation, turbine load changes, and pressure set point changes when considered meaningful. Stepwise in-creases in reactor power level, and decreases in core flow rate will be observed in each test series to evaluate the progress

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and safety.

l 10.2.2.11 The sequence will involve three major changes in operating

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conditions corresponding to 1050 psia, 800 psia, and 1500 psia reactor operation. Table 10.1 of the FHSR illustrates the expected tests at each pressure. At each of these pres-sures, and in the order given above, approximately one to three weeks' testing involving the following sequential steps will be performed:

-i 1.

Set reactor pressure at 1050 psia with 56 bundles in core.

2.

Rated subcooling, rated flow, 60 percent power level.

(a) Increase power level stepwise to 80 percent of maximum.

(b) Increase power level stepwise to 100 percent of cal-

' culated maximum.

t Sectisn 10 Page 8B Rev 1 (3/23/62) 3.

Rated subcooling, 60 percent power level, rated flow.

(a) Decrease flow stepwise to minimum or 60 percent of

.i rated.

4.

Repeat (3) for 80 percent and maximum power level, not exceeding 157 Mwt.

5.

Repeat (2), (3) and (4) for maximum subcooling (10 to 15 Btu /lb or more, if possible).

6.

Repeat (1) through (5) with reactor pressure at 800 psia.

7.

Repeat (1) through (5) with reactor pressure at 1500 psia.

10.2.2.12 Continuous operation at the conditions of any given test may be operationally de 'rable, even though further developmental observations are not required for those conditions. Since all I

operating limitations imposed for normal operation are also being met by each development test, continuous operation will be considered acceptable procedure.

I 10.2.2.13 A sequence of tests with an enlarged core (78 bundles) is also planned, essentially duplicating the above except for deletion i

of subcooling variations. Since a total power level of 157 Mwt and 1.5 burnout ratio are criteria still being observed, this series of tests in Phase I is inherently less severe. A power density of only 32.3 kw/l will be attained within the 157 Mwt reactor power limit.

10.2.2.14 Computer testing will largely be checkout and " debugging" of signal conversion equipment and computer programs. However, to evaluate the accuracy of programmed calculations it will be l

desirable to hold the reactor at a prescribed c'ondition in order to compare reactor heat rate, as determined by the computer,

i with the conventional measurements and computations. Also,

the control rod configurations will be maneuvered to evaluate

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both the rod position portion and the neutron flux and exposure portions of the computer function. Similarly, xenon transients will be computed and compared to test data.

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TEST SCHEDULE TYPICAL OF PHASE I n

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1 56 Fuel Bundle Core Size Reactor Pressure - psi

• Minimum 1050 1500 o

Core Inlet Subcooling*

Normal Max Normal Max No rmal '

Max Reactor Power Level **

Max 80 60 Max Max 80 60 Max Max 80 60 Max i

Total Recirculation Flow r

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d d

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78 Fuel Bundle Core Size Reactor Pressure - psi

• Minimum 1050 1500 Core Inlet Subcooling*

No rmal Normal Normal Reactor Powe r Level **

Max 60%

Max 60%

Max 60%

Total Recirculation Flow r ated 60%

rated min rated 60%

rated min rated 60%

rated min Rate ***

l The exact value of this minimum pressure will be established by the suitability of the plant I

equipment and system design as de scribed in this report.

l 1

During Phase I tests the maximum power will be that power at which the burnout mar-gin is equal to 1. 5 for the given flow conditions (not to exceed 157 Mwt).

• • • Minimum flow will be that for which the burnout margin is equal to 1. 5 at the given powe r level.

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Section 10 Page 10 4-10.2.3 Test Measurements and Data - Phase I 10.2.3.1 The measurements and data associated with the irradiation of the eight developmental fuel bundle.s will consist pri-marily of an accurate record of operating history of individual elements in terms of power density and total flux exposure. The integrity and performance of the developmental bundles will be checked periodically in the Big Rock Point reactor fuel pool by gamma scan, dimensional measurements, photographs of areas of interest.

l and ultimately by hot laboratory examination of selected fuel rods.

10. 2. 3. 2 The core performance transient and stability tests will cover the range of operating conditions presented in Table 10.1.

The data and measurements to be made will include the following:

Normal Instrumentation Reactor pressure Drum pressure Recirculation flow rate Steam flow rate to turbine Feedwater flow rate Feedwater temperature Core inlet temperature Core pressure drop In-core neutron flux from ion chamber and wire irradiation Generated power Control rod position Instrumented' As semblie s Channel flow rates Pressure drops In-core neutron flux from contained ion chambers at 3 positions Channel coolant temperatures Both transient and steady state measurements of these data will be made as required. Additional transient recording equipment will be utilized in many of the performance tests.

Maximum use will be made of the data processing capa-bilities of the computer for steady state tests.

10.2.3.3 Much of the data recorded as described above will be directly useful in the Physics tests. Gamma scan mesurements will be made with the reactor shut down by moving ion chambers axially along the desired fuel bundle and recording the sig-nal fluctuations as a function of position. This will also be

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done for selected individual fuel rods removed from the bundle. Conventional testing methods will be used to meas-ure reactivity on removal of a control rod or on the intro-duction of voids into an element.

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Section 10 Page 11 10.2.3.4-The measurement of data is an inherent feature of the scheduling computer. It will scan and process all the quantities listed above (Par.10. 2. 3. 2) and in addition will measure such quantities as the'feedwater heater flows and temperatures, condenser vacuum and coolant conditions.

These additional quantities give added computational capability to the computer which will permit the monitor-ing of equipment operation, calculation of heat rates, core performance, fuel exposure, and operational variables of interest to reactor core development. This recording and processing of data does not interfere with the normal plant operating or control functions, though the data printed out to the operator by typewriter will increase his knowledge of react.or conditions. Punched paper tape output will per-mit external processing of any recorded data beyond the capabilitics of the on-line computer.

10.3 DEVELOPMENT PROGRAM - PHASE II 10.3.1 The program associated with Phase II testing will be based on the results of the Pre-Operational Development Program and the results of Phase I of the Operational Development Program. Additional developmental fuel bundles fabricated by the technically and economically most promising process as determined through Pre-Operational and Phase I tests will be irradiated and examined. The high operating limits corresponding to 240 MWt reactor power and to 60 KW/L average core power density will be applied to operation with this fuel.

10. 3. 2 The operating and testing criteria, conditions, and meas-urements described above (Par.10. 2) will generally apply, but will be extended in scope to include the new power level and power density limits. Also, basal on the results from Phase I, the limiting criteria may be modified to allow an extended range of testing conditions. Methods of analysis will also be extended or modified as dictated by Pre-Oper-ational and Phase I results, and each test condition will be shown to comply with the latest approved criteria. A step-wise approach will be followed in Phase II in proceeding beyond the operating and test conditions already experienced i

in Phase I.

Major steps in such a procedure are (a) introduce into the core additional long life, high power density developmental fuel that is intended for eventual operation in a core averag-ing 60 KW/ L power density, (b) increase the total operating power beyond the 157 MWt limit but remain within the approximate 45 KW/L core average power density, and (c)

-increas.e the reactor power level to 240 MW t and the core average powe r density to about 60 KW/ L, though probably not simultaneously. A full range of performance and

.s stability tests will be performed within this framework.

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Section 10 Page 12

-Irradiation and examination of all developmental fuel will be carried to 10.000 MWD /T average, with selected fuel bundles going to 15,000 MWD /T. The three following

-paragraphs detail this step-wise procedure.

10. 3. 2.1 Advanced Design Fuel Elements Phase II developmental fuel bundles to be irradiated and examined in the Big Rock Point reactor will be designed and fabricated based on methods resulting from Pre-Operational test results. These bundles will incorporate the most attractive economic and technical design features of the developmental fuel concepts irradiated and tested earlier as part of the development program. These new

-bundles will be designed for long exposure from 10,000 to 15,000 MWD /T and for high power density operation corresponding to 60 KW/L average over the core. Later bundles will also incorporate changes suggested by results of the Phase I irradiation testing.

10. 3. 2. 2 Reactor Power Level Increase Utilizing results and characteristics from the initial Big j

Rock Point reactor engineering start-up tests, the reactor j

will be operated at successively higher core powe-levels l

while still observing the approximate 45 KW/L power density limit. These tests will in part be dictated by turbine capability at the lower pressures. Detailed knowledge of the flow characteristics and core hot spot configuration will at that time permit complete definition of such tests,

10. 3. 2. 3 Increased Power Density l

Completion of Phase I tests as described in paragraph 10. 2 will provide the justification to proceed to the approximate power density limit of 60 KW/L average and to develop up i

to 240 MWt reactor power. The operating and testing criteria that will apply for this series of tests will have been verified by the Phase I program and will have been modified where necessary. It is also intended that additional out-of-reactor data, being collected at the present time, will give an even broader knowledge of thermo-hydraulic operating limits than is now available. More complete fuel irradiation and examination data will al a be available from the high power density, low fabrication cost elements in VBWR.

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