ML19322A771
| ML19322A771 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 12/01/1966 |
| From: | DUKE POWER CO. |
| To: | |
| References | |
| NUDOCS 7911210792 | |
| Download: ML19322A771 (12) | |
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APPENDIX 4A TABLE OF C0:7fEITS Section Page
, 1 IITfRODUCTION 4A-1 2 EAT TRANSFER CHARACTERISTICS 4A-1 r 3 LABORATORY TES__T@ 4A-3 l i
i 4 CURPSiT RESEARCH AND DEVEIDPENT PROGRABE 4A-4 l 4.1 7IERATION 4A-4 4.2 FEEDRATER SPRAY N0ZZLE PROGRAM 4A-4 i
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3 LIST OF FIGURES (At Rear of Appendix) l Figure No. . Title l
4A-1 Steam-Water Mixture Velocity Versus Heat Transfer DNB ;
4A-2 High Pressure Heat Transfer Test Facility Schematic 4A-3 Steam Generator Stability Test Facility Schematic 4A-4 Once-Through Steam Generator Test Arrangement 4A-5 Central Tube Ratio Versus Total Tubes re Shell-Side Characteristics of Once-Through Nuclear Steam Generators
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APPENDIX 4A B&W ONCE-THROUGH STEAM GENERATOR DEVEIOPIEIT PROGRAM 1 IDTRODUCTION The once-through steam generator is a result of continuing efforts by The Babcock
& Wilcox Company to develop a steam generator of i= proved design for application in a nuclear power station. Technological advances in boiling heat transfer, combined with the unique operating characteristics of a pressurized water reac-tor system, contributeto improved efficiency and to reliable perfomance of the once-through steam generator.
2 FEAT TRANSFER CHARACTERISTICS Boiling water research, which contributed to the development of the PWR once-through steam generator, started in the middle 1950's. To illustrate the re-cults of this work, in Figure hA-1 a curve of DUB (Departure from rucleate Boil-ing) as a function of mixture velocity is plotted. DNB is defined as the quality where nucleate boiling changes to film boiling. Nucleate boiling is boiling from a vetted surface in which steam bubbles form around nucleation centers with a characteristic high heat transfer coefficient. In film boiling a film of super-heated steam insulates the bulk of the water from the heat transfer surface.
Q Fil:: boiling is characterized by lov heat transfer coefficient.
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In Figure hA-1, the abscissa shows the quality in per cent steam by weight.
This varies from zero, which is all liquid, to 100 per cent, which is all steam.
The ordinate is mixture velocity. This curve indicates the quality at which DNB occurs for a given pressure, heat flux, and geometry. Typical operating con-ditions for PWR steam generators are 1,000 psi and a mean boiling heat flux of 30,000 Btu /hr-ft2 . At very low velocities, as shown on the curve as Zone "A",
boiling occurs similar to the boiling in a pot or teakettle. One hundred per cent quality steam is reached before nucleate boiling is lost.
As mixture velocities are increased, the DNB quality slowly decreases. As mix-ture velocities increase, the effect of steam slip or slip velocity to promote continuation of nucleate boiling diminishes. " Slip velocity" is the phenomenon in which the steam in a two-phase system moves at a greater velocity than the water. As higher mixture velocities are reached, the dropoff in quality in-creases per increment of mixture velocity. With only a sli~,,ht increase in ve-locity, the DNB curve breaks sharply to the left. The DUB quality reaches a minimu= then reverses and increases with velocity until it reaches a maximum and goes straight upward.
Prior to the advent of the once-throu~h steem generator, combustible fuel boiler designs concentrated in the area (Zone "B") of the curve shown in the lower left-hand corner. Mass flow requirements for natural circulation steam generators are generally lov to provide low pressure drop in the vapor generating section, while at the sa=e time a relatively high circulation' ratio (ratio of pounds of mixture flowing to pounds of steam produced) prevails to ensure adequate flow throughout
) the high te=perature, high heat absorbing zones. Thus, for a n'atural circulatic;
._ / boiler with a circulation ratio of 3:1, a corresponding mixture quality of 20 y
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per cent leaves the vapor generating section. This is far below the DUB quality; hence, nucleate boiling is maintained throughout the boiling section of a natural circulation boiler. This prevents " burnout", which can occur at the point of DIG in a boiler with a high te=perature heat source and a low circulation ratio.
With the advent of PWR plants, with a li=ited te.perature heat source, s1=ilar recirculation ratios are required. Although the proble= of burnout does not exist, "dryout" at the point of DNB can deposit chemicals on the tube and cause " hideout". This condition =ay create proble=s with secondary water che=istry control. Fil= boiling also lovers heat transfer coefficients. With the advent of co=bustible fuel, once-through stea= generators, E&W beca=e in-terested in boiling water research in Zone "C" of the curve. A considerable research effort was conducted in the B&W Research Center, Alliance, Ohio, and later on in other laboratories, to establish the shape of the curve in Zone "C".
It is necessary to achieve high velocities in a high temperature source to =in-i=ize tube =etal fluctuations at the point of DUB by =aintaining relatively high heat transfer coefficients in the fil boiling region. This places the operating conditions in the upper portion of the curve.
During the late 1950's, when this early research work was being done, an in-teresting tiend in the boiling water curve was observed. Test data indicated that for a given subcritical pressure and heat flux, an inverse ratio of ve-locity to stea= quality at DUB occurred at lov = ass flows. This character-istic did not pemit application for co=bustible fuel, once-through stea= gen-erators because of the high te=perature " burnout" proble=. It was recognized by B&W, however, that this trend could be i=portant in once-through stee= gen-erator applications where the tube vall te=perature fluctuations at the loca-tion of DUB are lov enough to permit the design of a boiler with lov = ass flows.
(See Zone "D".) The pressurized water reactor offered possibilities for the useful application of this desirable boiling water heat transfer characteristic.
To further explore the heat transfer characteristics of boiling water at lov
= ass flows, test work was perfomed in the B&W Research Center. In Figure hA-2, a schematic of the high pressure boiling water heat transfer apparatus shows the equip =ent used in perfoming these investigations.
A single tube, electrically heated to achieve desired heat fluxes in a 6 ft test section, contained closely spaced themoccuples to observe tube metal temperatures. The once-through stea: generator and bypass arrange =ent were used to control the stea: quality entering the test section. Mixture veloc-ities were varied to investigate DNE characteristics over a vide range of conditions.
To investigate DNB characteristics for PWR operating conditions, testing was perforced at a pressure2 of 500 psi and 1,000 psi with heat fluxes fro = 20,000 to 100,000 Stu/hr-ft. Two tube dia:eters were used to detemine the effect of geometry. This work supplemented our knowledge of Zone "D" of the boiling water curve and produced results that were even = ore desirable than had been postulated fro = slip velocity theory. The break in the curve occurred at higher velocities than expected.
Additional infor=ation was obtained while perforcing stability characteristics tests on a recirculating PWR steam generator. In Figure hA-3 a sche =atic
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O dia6 ram of the steam generator stability test facility is shown. The steam gen-erator has a vertical U-tube bundle with fifty-five 1/2-in. CD tubes with a sep-arate steam drum, separate riser, and separate downcomer. The riser and down-comer each contain a valve so that studies can be made of the effect of circu-lation ratio on stability. To test the once-through steam generator principle, valving was modified to introduce feedvater directly into the bottom of the unit.
Instru=entation and controls on the facility did not permit B&W to obtain exten-sive operating data. However, the observed heat transfer characteristics con-fir =ed previous test data, and, operationally, the unit was extremely stable.
3 LABORATORY TESTS During August 1964, steps were taken to design and construct a facility to op-erate a full-scale section of the FWR once-through steam generator. An exist-ing hot water loop at the Research Center was used as the basis for this facility.
The objective of this test program vt.s to investigate operational characteristics in regard to stability, heat transfer, and controllability.
In Figure hA-4 a drawing of the test unit arrange =ent is shown. It contained seven tubes duplicating the tube length, tube diameter, and material of the full-size steam generator. Shell-side flow area per tube duplicates the full-size steam generator. The superheater section is arranged with concentric baf-fles to simulate the pressure drop and heat transfer characteristics of the full-size steam generator. An external downco=er is provided with a flow area dup-11cating the annulus area of the full-size steam generator. The external mixer O is arranged with a feedvater no::le sized to reproduce design feedvater spray V velocity into the downcomer. The unit was tested at conditions of pressure, temperature, and mass flows duplicating conditions in the proposed full-size unit. The stability and heat transfer characteristics were investigated over a vide range of secondary pressures, flows, and feedvater te=peratures with several arrange =ents of feedvater heating. The dyna =ic response characteris-ties of the boiler were detemined. Perfomance was investigated with various arrange =ents of automatic control, using steam generator-following and tur- t bine-following modes of operation.
l Test results indicated that the concept of a straight-tube, once-through steam generator with shell-side boiling operates stably, that the heat transfer is :
in agreement vith the calculated values on which the design was based, and that j controls were able to accommodate all anticipated operating conditions of load ;
changes and steady state requirements. Control is essentially on the traditional, !
three-element feedvater control concept. Feedvater is kept in st'p with steam flow with a final correction made from pressure error. This pressure error is equivalent to an error in water level in the traditional drum-type recirculating boiler. The use of pressure error in a high capacity steam generator of this ,
l type has been shown to be an excellent index of secondary inventory in the unit. l l
l B&W efforts to obtain detailed information on once-through nuclear steam gen- i erators with shell-side boiling are continuing. A 37-tube ***"" 5'"*#"t #'
which has been fabricated and erected in the nev 12.5 x 106 stu/hrhotwater test facility in the Research Center, is presently being tested. Structural -
characteristics of the 37-tube unit duplicate those of the previously: tested -
7-tube unit. Tube lengths, diameter, materials, type of baffling have been duplicated. An external mixer and downco=er arrange =ent is used.
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Figure LA-5 shovs how a 37-tube steam generator ec= pares in shell-side character-istics with other sizes. The plotted curve has the total nu=ber of tubes for the abscissa and a ratio representing the nu=ber of centrally located tubes to the total nu=ber of tubes for the ordinate. In a 7-tube unit, one out of the seven tubes is centrally located, e.g., co=pletely surrounded by adjacent tubes.
This ratio is approx 1=ately 0.14. A 19-tube unit, which is the number making the next lar;est uniform pattern of tubes, has seven centrally located tubes.
This ratio is approximately 0 37 The 37-tube steam generator has 19 centrally located tubes. This ratio is approxi=ately 0.52. The corresponding ratio for a 15,000-tube central station steam generator is 0 97 This shows that the 37-tube test unit will have shell-side characteristics approaching those of a full-size central station steam generator. Operation of the 37-tube facility vill provide additional infor=ation on steam generator control opti=1:ation. The 7-tube tests de=cnstrated the capability of the system to operate in either the steam generator-following or turbine-following = ode throughout the normal load range. For central station plants there is an additional incentive to apply the integrated master control system. This system is pri=arily suited to controlling
=v generation to a demand signal. Another characteristic of the integrated sys-tem is the ability to hold pressure variations to a mini =u=.
k CURRENT RESEAP0H AND DEVEIOPIM PROCPfGS k.1 VIERATION Analysis of the natural frequency of the tubes of the once-through steam gen-erator has shown that the lovest frequency in the stea= generator tube bundle is apprcximately 33 cps. Experience has shown that natural frequencies of 33 cps or higher have been no problem in service. In addition to the theoretical investigation of the design of the once-through steam generator, it is planned to vibrate the 37-tube model once-through steam generator during the testing program at the Research Center. By imposing mechanical vibrations on the steam generator from an external source, and by varying the frequency of the source of vibration, it will be possible to determine accurately the natural frequency of the tubes and other parts of the steam generator. After completion of the test-ing program, the steam generator vill be destructively exa=ined for evidence of tubes hitting each other or structural parts, and for evidence of vear.
4.2 FEEIMATER SPRAY UOZZLE PROGRAM The once-through steam generator cixes the incoming feedvater with steam drawn from the steam generator tube bundle to heat the feedvater to approximately saturat'on te=perature before it enters the tube bundle. A test progra= is underway to develop a feedvater spray arrange =ent which will provide for atom-1 l
icing the water to small enough droplets to assure'that the feedvater vill be at saturation temperature before entering the tube bundle. Testing has begun on a spray no::le to demonstrate the spray pattern and droplet size. It is planned to investigate the spraying of water both into air and into a steam atmosphere. The length of time or distance. required for the subcooled water to be heated nearly to saturation temperature vill be determined. A study of the dynamics of the steam and water flow within the tube bundle and feedvater heating annulus is in progress.
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