ML20052E193

Responds to on NUREG-0737,Items I.A.2.1 Re Upgraded Senior Reactor & Reactor Operator Training & II.A.B.4 Re Training for Mitigating Core Damage.Supporting Info Encl

ML20052E193
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	05/05/1982
From:	Jackson E VERMONT YANKEE NUCLEAR POWER CORP.
To:	Vassallo D Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.2.1, TASK-2.B.4, TASK-TM FVY-82-50, NUDOCS 8205100180
Download: ML20052E193 (115)

Text

{{#Wiki_filter:_-____. _ o.. VERMONT YAN KEE NUCLEAR POWER CORPORATION SEVENTY SEVEN GROVE STREET 2.C.2.1 RUTLAND YERF10NT 05701 FVY 82-50 REPLY TOs ENGINEERING OFFICE 1671 WORCESTER ROAD FR AMINGH AM, M ass ACH USETTS 01701 7ELEPHONE 887-872 9100 y-May S U.S. Nuclear Regulatory Commission RECBVE Washington, D.C. 20555 MAY 7 1982* ' Attention: Office of Nucicar Reactor Regulation 7,, E mmBIERNMf 8MEE Mr. Domenic B. Vassallo, Chief S 88888,amm a Operating Reactors Branch #2 // Division of Licensing N

References:

a) Letter, USNRC to VYNPC, dated 4/2/82 b) NUREG 0737 Item Nos. I.A.2.1. and II.B.4 c) Letter, VYNPC to USNRC, dated 12/15/80

Dear Sir:

~

Subject:

Upgraded SRO and RO Training and Training for Mitigating Core Damage Answers to the questions in Enclosure 1 of Reference (a) are submitted as follows: Item 1: Does the Initital Operator Licensing program described in procedure AP 0710 teach the subjects of heat transfer, fluid flow and thermodynamics at a level of detail which is comparable to that presented in Enclosure 2 of Denton's March 28 letter?

Response

The program described in AP 0710 presents heat transfer, fluid flow and thermodynamics at a level of detail at least commensurate to that presented in Enclosure 2 of Denton's March 28 letter. Item 2: The Licensed Operator Retraining program described in AP 0711 includes lecture materials on heat transfer, fluid flow and thermodynamics. Do these instructions also cover the range of topics indicated in Enclosure 2 of Denton's March 28 letter?

Response

The retraining program described in AP 0711 includes lecture s materials on heat transfer, fluid flow and thermodynamics. The topics covered are in Enclosures 1-5 which are handout material. Other series of lectures cover pertinent heat transfer topics. BO5 8205100180 820505 l PDR ADOCK 05000271 1 p PDR I

U.S. Nuclect R2guletory Commienien May 5, 1982 VERMONT YANKEE NUCLEAR POWER CORPORATICC Paga 2 Item 3: Do the programs for Initial Operator Licensing and for Licensed Operator Retraining address the subject of using installed plant systems to control or mitigate an accident in which the core is severely damaged? This requirement is called out in of Denton's letter. Do the lectures in the programs address the topic at the level of detail specified in Enclosure 3 of Denton's letter?

Response

The Initial Operator Licensing program and the Licensed Operator Retraining Program covers the use of installed plant systems which cre designed to control and mitigate an accident involving core damage. Topics covered are in the outline (Enclosure 6). The GE course on recognition of a degraded core condition is the basis of our training program. During retraining more detailed information on mitigating core damage is presented as part of Plant Safety and Emer-gency Systems. Item 4: Does the training program have an increased emphasis on reactor and plant transients as called for in Enclosure 1 of Denton's March 28, 1980 letter? Does the program address both normal and abnormal (accident) transients?

Response

The training programs for Initial Operator Licensing, Licensed Operator Retraining and Shift Technical Advisors contain increased emphasis on both normal and abnormal (accident) reactor and plant transients. These events are covered in both classroom and in simulator training sessions. Item 5: As called for in Denton's March 26, 1980 letter, are the training program instructors enrolled in a requalification program which addressed current operating history, problems and changes to procedures, and administrative limitations?

Response

Training program instructors are routed all material and included in training sessions related to operating history, problems with and changes to procedures and administrative limitations. l Item 6: Appendices A and B of the licenses operator retraining program lists certain control manipulations which are performed on an annual or semiannual basis. Appendix B is missing one control manipulation, " Reactor Trip," which corresponds to item (25) in of Denton's March 28, 1980, letter. Would you please i explain this omission? f

Response

Vermont Yankee's Licenses Operator Retraining procedure l AP 0711 does not contain " Reactor Trip" in Appendix B. Fourteen of the twenty identified evolutions listed in Appendix B will result in a reactor trip. Since there are initiating events for the condition, the condition itself was not listed. i i c

U.S. Nuclear Regulatory Commission Y VERMONT YANKEE NUCLEAR POWER CORPORATIO3:' Item 7: Are the lectures and quizzes on the subject of accident mitigation given to shift technical advisors and operating personnel from the plant manager through the operations chain to the licensed 4 operators? If'they are, would you please provide the titles of a. the people who are trained and an organization chart which illustrates their position in the operations chain?

Response

Lectures and quizzes on the subject of accident mitigation are given to shift technical advisors and licensed operations personnel through the Operations Superintendent. An organization chart has been included in Enclosure 8. As stated in Reference (c), Training of the Plant Manager and the Assistant Plant Manager, has not been completed nor has it been completed for managers and technicians in the Instrument and Control and Chemistry and Health Physics departments. Vermont Yankee does not feel that this training is necessarily warranted based on their job functions and prior training. Item 8: Do the training and the requalification program elements which include heat transfer, fluid flow, thermodynamics and accident mitigation use 80 contact hours? (A contact hour of instruction is a one-hour period in which the course instructor is present or available for instructing or assisting students; lectures, seminars, discussions, problem-solving sessions, and examinations are considered contact periods under this definition.)

Response

Instruction in heat transfer, fluid flow, thermodynamics and accident mitigation for initial licensed operator training use 80 or more contact hours. A description of this program is provided in response to your inquiry for item II.B.4 of NUREG 0737. Sce Enclosure 7. The instructional program for requalification training in these specific program elements is at least 40 contact hours. Coincident material is covered during training in other subjects so a more precise estimate of contact hours is unavailable. We trust this information is acceptable. However, should you have further questions regarding this submittal, please contact us. Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATION Elo & E.W. Jackson Manager of Operations EWJ/dm cc: Dr. R.J. Liner Science Applications, Inc. 1710 Goodridge Drive McLean, Virginia 22102 r

m__ lnEL61MM M For Qucation 2 r DIEPM HYDRAULICS Sufficient information is included in this section to provide you with a basic understanding of BWR

• thermal hydraulics.

However, if you desire to obtain a more in-depth understanding of the subject you should study the following documents:

1. NEDO 10958 - GETAB 2. Nuclear Beat Transport - M.M. El Wakil .... ~ 2 ^"' ~ 3. Boiling Crisis and Critical Heat Flux TID mx. 25887 - L. S. Tong Heat Transfer In light water reactor operation we are concerned with the transfer (( of heat from the fuel center line (C ) to the light water moderator g which comes into contact with the outer fuel cladding surface. As you know, heat can be transferred by (1) Conduction, (2) Convection, and (3) Radiation. The first two listed modes of heat transfer are the one of primary interest in power plant operations. Conduction When heat is applied to a material, the kinetic energy of the atoms or molecules of the material is increased. Due to this increase in kinetic energy the particles mentioned will have a greater tend,ency, to collide with each other. When these colli ions occur, the particle with the least amount of energy (cold) will gain energy (heat). For all practical purposes this is the process by which heat generated in the fuel pellet is transmitted to the outer clad surface. In relation to the fuel rod this conduction flow is in a horizontal plan;

~~ h. ?.-.. from th2 fual c2nter lina to tha cledding curfcen. The race at which heat is transferred by conduction is shown in the following equation: Thermal Temperature Conduction heat. Conductivity I Area X Cradient transfer rate material thickness Convection Convection is the process of transmitting heat by means of the d _ mov,ement of heated matter froi one place to another. Convection trkes place in liquids and gases. Our application deals with afldid. ' ' ' #~

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(moderator) flowing past a metallic fuel clad surface. Fluids have a tendency to adhere to solid surfaces resulting in the formation of a stagnant film on the surface. This film is normally very thin and (("'. heat is transferred across this film by a combination of conduction and convection. After the heat penetrates the film, it is transferred rapidly through the remainder of the fluid. The resistance to temperature flow'.in this area becomes so low that there is virtually no temperature l variation through the bulk of the, fluid (moderator) at any given 1 elevation along the fuel rod. It should be noted that the flow of heat in the fuel channel is in a vertical plane due to upward flow in the core. The convection heat transfer rate from the heated surface, through the stagnant film and'into the bulk fluid is governed by the same fundacental principles as conduction heat transfer rate, in that: Thermal Temperature Convection heat Conductivity X Area X Gradient transfer rate stagnant film thickness H.. l

T.s.? o -.. B nt Trrnqfrcr in n BWR ( In a BWR the primary heat transfer mechanism at the fuel cladding outer surface is convection; i.e., the mechanical migration of water molecules, from the stagnant water layer covering the clad surface, out into the bulk moderator. The temperature difference, AT, between ~ the bulk coolant and the fuel clad surface is governed by: AT = 0 hA where Y ~ ~ Q = heat flow (htu/hr) O? ~ ~.- - -- r m.-

->- ~w 2

o) h = heat transfer coefficient (Btu /hr - ft ~~ F A = surface area The heat transfer coefficient (h) is also referred to as the convectios or boiling heat transfer coefficient or the film coefficient. Boiling Eeat Transfer ~ Boiling is the evaporation of a liquid to vapor occurring within the ~ body of the liquid by the mechanism of bubble formation. The very nane " Boiling Water Reactor" implies that dependency is placed upon the boiling of water to dissipate heat that is generated in the - reactor fuel. When boiling occurs in a BWR, bubbles form on the fuel clad surface and then drif t off into the bulk coolant giving off their energy to the cooler surrounding water. When these bubbles form, they tend to agitate or stir up the stagnant water film on the clad surface and improve the thermal conductivity of the film. Also each bubble carries off more energy than is possible by non-boiling convection heat transfer, thus increasing the total heat renoval rate. Three types of boiling are of concern in the operation of the BWR. They

)$!b"... cro nucicato briling, trcnoitica b3111:3 and film b:iling. ' Nucleate Boiling By definition nucleate boiling means that small steam bubbles are formed at sites of nucleation, which are usually small imperfections on the heated surface. In a BWR nucleate boiling is characterize,d by a film of water clinging to the fuel rod surface with vigorous steam bubble foncation and movement. Nucleate boiling is extremely important in reactor core heat transfer because, as shown in Figure ~ -- 1.,2-1 a great deal of heat energy can be transferred without extremely r;.

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y ,w. high surface temperature.'

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Transition Boiling As power (heat flux) is increased, bubbles form at a faster rate. ,i Eventually a point is reached where the cooling media cannot reach the heated surface as efficiently as before. This point is called- "- departure from nucleate boiling (DNB) and is evidenced by a reduction in the amount of heat transfer for a given AT. DNB is characterized by local flow and temperature oscillations brought about by the changing steam bubble conditions. Experimentation has shown l oscillations in fuel cladding termperatures of 50 F to 100 F at DNB. l If a BWR is operated beyond DNB, there is the probability, due to excessive clad temperatures, that weakening and/or failure of the clad would occur. Excessive void formation will also cause deposition of impurities on the fuel clad (boiler scale) resulting in excessive corrosion film' thickness and' higher clad AT.

g._, -, Filn B7111ng f r Beyond transition boiling filan boiling occurs. This is characterized by the surface of the cladding being blanketed with steam. Steam is an excellent insulator, therefore, the clad temperature would rise ...? - dramatically. In all probability, the clad would fail (burnout) if full film boiling occurred. Obviously, full film boiling is never permitted in an operating reactor under ordinary or transient conditions. -- Fluid Flow- -r. This section will deal with flow patterns from channel entry to exit,..;. Figure 1.2-2 shows the different flow patterns that can exist on a single fuel rod during normal operation. As can be seen from Figure 1.2-2, at a point along the rod the ..g e-fluid loses its homogeneous character. At this point single phase __ flow changes to two phase flow. Two phase flow occurs from the area of bubble flow to annular flow. Note in particular that for high steam quality conditions, low moisture content in steam, the annular flow pattern occurs with a thin liquid film clinging to the fuel rods. DNB and Two Phase Flow As previously stated, a BWR operates below DMB. Yet Figure 1.2-2 clearly shows that DNB has been exceeded in the moderator channel. The statement i s'in reference to liquid on the clad surface and as long as a film.of liquid exists on the clad surface, the requirement is satisfied. Steam Volume Fraction and Slio Ratio Steam volume (void) fraction is defined as the fraction of the flow area occupied by the steam, compared to the total flow area occupied 4 Q v

LE9C '. ' ' ' l ' by etcam plus wat:r. If homogeneous flow were assumed, the local steam volume fraction could be calculated based on the pressure and local steam quality and ~ specific volumes for saturated steam and, water found in the steam tables. Steam quality (x) being defined as the mass flow rate of steam (W ) divided by the mass flow rate of steam plus the mass g flow rate of water (W ) W ..._. n x. c

t - -

.m,-y y +y G ,,_x,. . m%*;,- Frequently x is multiplied by 100 and expressed in percent. Separation of phases occurs due to different buoyant effects and E. - viscosity. The velocity of steam is not equal to the velocity of Q.,,. k ~' water in a boiling channel. The ratio of steam velocity to water. velocity is defined as the Slip Ratio. Slip ratio increases as coolant goes up to the core. However, the rate of slip increase decreases. As bubbles accelerate, they tend to pull the liquid along. This effect increases the liquid velocity, hence the ratio increar-es at higher core elevations. Fuel Assembly Pressure Drop l In a fuel assembly local pressure losses occur at the orifice, lower tie plate, spacers, and upper tie plate.. Further pressure losses would be incurred due to the effects of two phase flow. As flux increases, slip ratio increases, friction between the gas (steam) and liquid (film on cladding) increases and overall asseebly pressure drop increases. -w- ..em e e e-e .+- a

6f.?~.~.. Funi Chennel Flow rnd Temorrntura Chernetarintics Coolcnt entaro th2 b2ttom cara, flcws upward around tha funi red 3, e absorbs energy from heat transfer originating from the nuclear process. Due to the peculiar characteristics of neutron caused fission reaction 0 the average heat flux (Q/A) produced from fission in the core would assume a shape somewhat like that shown in Figure 1.2-3. The highest heat flux will be in the core interior, hence some fuel bundles will have a higher than average heat flux and some a lower than average. The coolant temperature curve rises as heat is added, until temp-1,' er'ature saturation occurs and coolant bulk boiling begins. From as - .~ ~

%=

this point the coolant temperature remains constant all the way to core exit. Note the sharp temperaturerise where there is a large value of heat flux. Because the coolant is changing phase, the ~ ', coolant temperature profile is not altogether descriptive of coolant ~ f/ ~~~ energy increase. A better description is obtained by plotting coolant enthnipy change. The curve for fuel rod surface temperature rises and then icvels at a constant value above coolant temperature. The initial rise is caused by the AT across the film required to acco=modate the heat _ flux (Q/A). The point where the fuel rod temperature levels off is due to the inception of nucleate boiling. Nucleate boiling is an i excellent heat transfer mode and therefore even though the heat flux is getting greater, the AT across the boiling film remains relatively constant. The curve for fuel rod center line temperature will of course be above the fuel surf ace temperature. The amount that the center line l l - - ~.

"M f.' '. ' te=parcturo io granter thc3 curfcco temp;r:turo will d::prnd dir:ctly on th2 hact flux. Tha bannficici affseto of nuc1ccta bailing on r !~ center line temperature can also be seen. As long as nucleate boiling' is occurring on the. fuel rod surface, the fuel rod surface temperature isonly slightly greater than liquid temperature.,, This, in turn, automatically keeps the fuel center line temperature at a lower - value than if non-boiling convection were the mode of heat transfer from surface to liquid. Figure 1.2-4 graphically illustrates a typical fuel temperature cross-section with nucleate boiling at a 3 _ _high heat flux. 5 u.- av E Thermal Limits - Introduction For normal and transient operation of the BWR it is necessary to limit peak fuel enthalpies in order to limit fuel pellet vapor pressure. A rapid melting and vapori=ation of fuel pellets and subsequent melting (burnout) and/or rupture of fuel cladding would 7 represent the worst case in relation to failure to maintain thermal limits. Thermal Limits - Terms ~ The following terms are used in conjunction with thermal limits:- 1. Plastic Strain - a plastic deformation of the fuel cladding due to excessive heat. I 2. Linear Beat Generation Rate Limits (Kw/f t) - required to limit steady-state plastic strain. 3. Maximum Linear Beat Generation Rate - prescribed in order to maintain an adequate margin below a 1% clad pastic strain limit.

A y. :: ~ \\ 8 4. On nce cf Cant $r Lina k ) Malt - th3 amount cf h ot rcquirzd f ~ to begin meltinlg the center of a fuel pellet. ( ~ 5. Planar Linear Heat Generation Rate - the heat generation along a horizontal axial at any given elevation of'it fuel bundle. ~ 6. Average Planar Linear Beat Generation Rate (APLHGR) -- the average LHGR of all fuel rods in a 6 inch node of a fuel bundle. 7. MAPLHGR - The Maximum APLHGR. 8. Total Peaking Factor - a combination of axial, radid and Cl. " ' - ~T relative peaking factors.- ~ ~ ~ ~ ~ =_., u' 9. Critical Power Ratio (CPR) - Critical Power Ratio (CPR) is the ratio of that assembly power which causes some point in the assembly to experience transition boiling to the assembly power at the reactor condition of interest as calculated by 5 application of the GEXL correlation. 10. Minimum Critical Power Ratio (MCPR) - the minimum CPR corresponding to the most limiting fuel asssmbly in the core., 11. 5nthalpv - heat content; sensible heat; total heat; sum of ~ internal energy of a system plus the product of the systems volume multiplied by the pressure exerted on the system by its surroundings. Heat Generation, Fuel and Clad Limits In maintaining thermal limits consideration must be given to two major areas; namely, limiting the deterioration of the heat transfer =echanism from clad to water'and limiting the internal stresses from excessive heating of the fuel pellet. In relation to the latter -. _. - w mw.m e= - * - + =-

• . = * *

-,,so-

W '. ' ;.. ; ~ considarctica c.aintaining tha decign pack 1HGR (Sca Tchlo 1.2-2) { will ensure that overpowering of a fuel pellet does not occur. >, Fuel Densification ~ As a reactor is taken from cold to power, the fuel densifies and expands. Densification is caused by" annd~ation of pores found throughout the ceramic fuel pellet. This annulation occurs when the pellet is exposed to a neutron flux. Expansion is caused by the temperature increase in the pellet. The - _ - ~ ne[ result of densification and expansion during ascension to power is -m ~, y -- an overall expansion. In setting thermal limits for fuel, thermalh'W expansion is not taken into consideration, but the densification is. If fuel densification is taken into consideration during heatup, then 2 a factor that would have to be dealt with in any calculations to limit

. )

fuel cladding tenperatures would be power spiking. As densification - occurs, gaps are created between fuel pellets which have a positive reactivity effect due to the change in moderator to fuel ratio. Also increases in LHGRs result from axial shrinkage of the pellets. So, in the formula used to ensure that the design peak LUGR is not exceeded, a spiking penalty is incorporated. Average Planar Linear Heat Generation Rate (APLHGR) To ensure that the reactor fuel cladding does not reach the limit in tenperature, designated as Peak Cladding Temperature (PCT) (i.e.,2200 F), on a design basis loss of coolant accident a limit on average planar linear heat generation rate is established. = g

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m M -m.ae--eeM w num ame-ep & ap m e e mM we e ww. m$

W, .f - ~ ~;. ~- -p The maximum clad temperature following a postulated loss of coolant-c I accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent, secondarily, on the rod to rod power distribution within the assembly. The PCT is calculated assuming a IRGR for the highest powered rod which is equal to or less than the design ISGR corrected for densification. Core Thermal Desien

r -

A comparison of BWR core parameters is made in Table 1.2-1. This' listing will be used in the remainder of this chapter so that meaningful values can be associated with the subject matter. Peaking Factors Peaking factor can be defined as the power or flux of interest. divided by the average power or flux of interest. The end result of a discussion of peaking factors is determining a design limit called the total peaking factor. The Total Peaking Factor (TPF) tells you the maximum local heat flux in relation to. ~ the average power of the core. (See Table 1.2-1 for Design Total i i I l Peaking Factors.) Limiting safety syste= settings are based on the design total peaking factor, therefore, if the actual total peaking factor exceeds the design total peaking factor there is a possibility l of fuel damage occurring on a transient. Technical specifications impose limitations to ensure that fuel damage does not occur under these circumstances. In our discussion we will develop the process by which TPFs are determined and discuss the technical specification requirements in relation to this limit. D .g.- .-6g g gi -wpem... em em h-e.i. ,e. qq e w

p:.:,: - ;_' ~~; ; - Tntn1 P y king Frctnr (TPF) o'p Tha b2cic fermuln fcr TPF ist Axial Local Relative TPF = Peaking I Peaking I Peaking Factor Factor Factor ue d Axial Heat Flux Axial Peaking Factor = e The Average Beat Flux for that Rod Average Heat F1Ex for Highest Local Peaking Factor = Bundle Average Fuel Rod Heat Flux --.. ? Reljtive Peaking Factor = The Bundle Power' of Tnterest 7: - "~'* Average Core Bundle Power- '-My This formula (TPF), when combined with the core average heat flux, will c tell you the highest heat flux experienced in the core. A more under-standable formula for TPF then might be: n ~ TPF = Maxi 2num Heat Flux in Core T Core Average Heat Flux Technical specifications use a ratio, TPF design over actual caximum TPF, (MTPF), to reset the required scram and rod block trip points for the Average Power Range Monitors (APRM) when design valves for TPFs are exceeded. However., this has been changed. In Standard Technical Specifications (STS) the design TPF is defined as the ratio of the fuel rod design LEGR divided by the design average LHCR of that core. This TPF limit is specified for the type of fuel j l being used, i.e., for a specific BWR-4 the numbers are 2.63 (7 x 7' fuel assemblies) and 2.44 (8 x 8 fuel assemblies). MTPF is referrred i to in some technical specifications as Limiting Total Peaking Factor (LTPF). A typical formula in technical specifications using these values to l establish APRM settings in case of TPF exceeding the designed values would be: .D

w' ~,.......:.. = '.., - -,,v Scram S2tting (% powar ) = (.66W + 54%) I MFLFD ~ f Rod Block Setting (% power) = (.66W + 42%) I MFLPD FRP j W = Recirculation loop dri e flow, the flow at the discharge of the recirculation pump. -.=. ?J ..-y.. NOTE: As long as MFLPD_is less than 1.0, the factor is ignored. tur only when it is greater than 1.0 is it used to lower APRM ~ gain settings. Merri-m Fraction Limiting Power Density 13.4 KW/ft (- (8 x 8). Fracture of rated power 1593 MWt. T-~ i. m n.~ - ~ Fuel Cladding Integrity Safety Limit ~% as stated in the Standard Technical ~~~~ - The fuel cladding safety limit, Specifications (STS), is a Minimum Critical Power Ratio (MCPR) of 1.07. ' This number can be broken down into two separate parts. The foundation ~.'s for the safety limit is a MCPR of 1.0; that value where the onset of transition boiling occurs. Because of the uncertainties in the MCPRT-- calculations, 07 is added for conservatism to bring the number up to 1.07. This number represents the actual safety limit. For limiting conditions for operations a third number is added. The different transients are analyzed to determine their effect on Critical Power Ratio-(CPR). The ACPR for the transient having the greatest effect vill be adde l to the safety limit. In a particular EWR/4 technical specifications 1 this number is .2. It should be noted that the third number will vary frc plant to plant, depending upon their own transient analysis package. We can now assemble the parts and come up with the limiting condition for operation on MCPR of 1.27. ( Critical Power (CP) CP is defined as that power where departure from nucleate boiling occurs. In the term Critical Power Ratio (CPR), the CP is in relations to a

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W 5' ..... <..~. ; :. u. oinglo fu21 bundio cod to th2 occto cf bailing on. the cica curfcc20 r within that bundle. Therefore, if a fue?. bundle reaches its critical i power, then departure from nucleate boiling (transition boiling) has occurred at some point on the fuel'eladding surface within the bundia. ~~ g ~ .j It should be noted that the occurrence of tr'ansition boiling does not mean that clad perforation vill occur. At the time of transition boiling, the clad at that point would jump in temperature approximately 50 F and continue to increase in temperature to approximately 1100 F. - - ~ * ~ 1100 F is below the perforation temperature of the cladding material. 2.y_ ~. " q;. ~ Test loops of fuel similar to that used in the BWR have operated aboVe*~ ." ' '~~~ -N 7 CP for 30 minutes without experiencing clad perforation. Critical Power Ratio (CPR) Figure 1.2-5 CPR as defined and illustrated by Figure 1.2-5, CPR being a relationship between lines 1 & 3 on this ' figure. The axis on this figure are boiling L-_= ~- length (L

• ""*#"E" "* "* 9""

I "E ""E B defined as the distance from the point of initiation of bulk boiling to the point of initiation of transition boiling. Figure 1.2-2 illustrates the fuel channel boiling conditions. Item C i on this figure is labeled " bubble flow" and is that point in the fuel channel where the bulk cooiant stream has reached saturation temperature. This is the point identified as " initiation of bulk boiling" used in the i definition of L. "The boiling transition point" used in the' definition B of L ref rs to any point on the clad surface above that point where B initiation of transition boiling occurs. 1 O g 9

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Tha unito ca the h riz:ntal axic in Figuro 1.2-5 vould ba centuremento (- b of ler.gth (feet, meters, etc.), with zero being at the extreme left of.. the nri m. ~. The vertical axis on' Figure 1.2-5 is the average steam quality in the coolant channel at the boiling length on interest. Steam quality being defined as the amount of moisture in a given volume of steam; the less moisture, the higher the quality. Steam quality is expressed in percent. The, lines on Figure 1.2-5 are numbered, Line 1 represents a heat balance, z.._ - integrated curve, for a single bundle in the core at a specific set of, 7 conditions; i.e., flow, flux', pressure, etc. Line 2 represents an 5 a: integrated heat balance curve if the power in the bundle, represented by Line 1, were increased to CP. Line 3 is explained in the next paragraph. o The GEXL Correlation Line 3 on Figure 1.2-5 is developed by an iterative process and represents the end results of approximately 1100 data points under various operating conditions. These operating conditions are specified below and they establish the boundaries under which this line is considered to be valid. This line is called the GEXL correlation line and represents a CP heat balance curve for any bundle at any boiling length. The point on Figure where 1.2-5 where Line 2 becomes tangent to Line 3 represents the L3 departure from nucleate boiling would occur in the bundle. A ratio, of Line 3 (numerator) and Line 1 (denominator) is the CPR or in other words, the ratio of critical bundle power to operating bundle power. That fuel bundle in the core which has the minumum CPR' (MCPR), closest to 1.0, is compared with the plant's technical specifications fuel clad integrity safety limit to determine if the plant's thermal limits are being violated. ww w.upd 9e m-us .a=> 4 .-a m., 6

g Tha GEIL c:rralctico io valid ovcr e rr.nga cf ennditieno untd in tha , ; p- -m f tect dcto to d;valop tha correlation. Thaca conditiono crat h Pressure: 800 to 1400 psia.. 0 Mass Flux: 0.lto 1.25 10 lb/hr-ft -! Inlet Subcooling: O to 100 Btu /lb Local Peaking: 1.61 at a corner rod to 1.-47 at an interior rod Axial Peaking: Shape Max./ Avg. Uniform 1.0 outiet Peaked 1.60 w x: - ~ ~ Inlet Peaked 1.60 -- W Double Peaked 1.46 and l 1.38 Cosine 1.39 r' Rod Array: 16, 64 rods in an 8 x 8 array, 49 rods in a 7 x 7 array j Core Thermal Power Limit (Reactor Pressure -800 psia or Core Flow -10% Rated) The use of 'the GEXL correlation is not valid for the critical power calculations at pressures below 800 psia or core flows less than 10% of rated. For these conditions the. fuel cladding integrity safety limits is established by other means. This is done by establishing a limiting condition of core thermal power operation with the following l basis. Since the pressure drop in the core bypass region, that region between fuel bundles, is 4.56 psi, due to elevation head, the core pressure drop at low power and all flows will always be greater than 4.56 psi. l l i

34=~ 5-.., q.?n r -" \\ T1ow tecto conducted cutsida o recetcr cora on o restricting area equal. 3 to that of.a fuel bundle show that at 28 x 10 lbs/hr flow a pressure - drop of 3.'5 psi exists. Thus, fuel bundle flow with a 4.56 pai driving head vill be greater than 28 x 1531bs/hr. independent of total core flow and bundle power for the range of bundle powers of concern. < c m. n1 Full scale test data at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, _ the 3.35 We bundle. power corresponds to a core thermal power of more than 50%. Therefore, a core ' ' hermal power limit of 25% for reactof pressures belov 8dO psia or h. t flov less than 10% is conservative. Su=ary - Thermal Hydraulics Classification - Thermal limits are technical specification and 10 CFR requirements. The purpose of having such limits is to ensure integrit< Purpose of the fuel clad. Plastic Strain, LEGR, APLEGR, MAPLEGR, TPF, CP, CPR, MCI Terms l I 1 h p.- e==...w.--, ..w ..m. .-L. z. ._.____.._._:_,.__,____.l

EW - ~.3..._ - -. : a.... ~- --O .,7:..' (..** l e n n, i et w w-ggs a m m go O O g. =. 0 v

a. -

= = a 1 w = a n ~ a g e g n m n ~ ~ m,, - O n X X E m C n. T. g M an w, o 'n n o m in. so i m m M Ol' an ce to an o cm J e e m f-M M 9 9 ci w w an g n u an E. 9 y X N g n n c; u, x n g' - 4 C =. %.,$ - (3r-- g C 3 ~ w tAJ 'i= :- IJJ - - E -. 4. e e nn sa E .n: M o in m m gl c3 an f-n a d S e 4 d I 9 2 3 = ,-c wa O n. a. 1s3 oh' C n rw w. g N d5 9 8 AA C, m c3 g o ?> w s d >i 3 $e U C .5 E d O k g r N M N T g -. m (23 - - ~ r=. cm m co - ~ ~ 1s.' to w m en o es "o d N s-i co if r.4 m gj J O I m v N rw n m in o n m w rw rw x X O C C 40 .= N O* o Cm g i m 88 1 m cm cm n to n N' C G.E ~ o C n C O p y m w tw n T D H 2 W m o,g c ~ p-Z O tv. W .J w u.- Qw E o w en .J B c g w O w N = c:. g _A w

• t C

2 c E H b cn E C ca m 3 e w u. < o H w w H m J E Z X H w u. C H w O Om e '^ O o w 3 m m 3 tt x $"" E: 3 w O J 2'- e<2 ~x a M 2 g n 8 0 s w x e = w O w e m C e g u3 3, a u. m s s o <J J 3 O 2' g g u. w w w O. W x c2 Ew to u. u. i c. i.:. U a-g@ k .J g I N C w Q. M~ i 3 u u. g O w ws W U C O O O-c c c u in ~ c ou w u_, = c c= g -c O C-4 <L I 5 a 4 W Z M C w W O C Or H Q C wC m e r x CC x w w t.43 e w 2 Ui o ( WW w w< e.- 2 e m e-3 3 m <C >g O w 3

m. <

w w l c Z Z c: C-c. H a. Z c u. I d A. ri .e.' s.i. ti. l* A s-i w' sn 6 m' ai ci c t e l W e 1.2-25 -..-w. . =.. - - -

• * ^
• ~

.L'

4%g.-2 :.. _ 4.- .-(". TABLE 1.2-2 MAXIMUM LINEAR HEAT GENERATION RATE The maximum LHGR in the core shall not excee'd the value de-termined by the following relation: AP L LEGE = LHGQ 1.0 m P mu y LHGR = Maximum local linear heat genera-m tion rate.

• • ~'

s LHGR = Local linear heat generation rate,,. . 4* g~ license limit L = Arial position from bottom core LT = Total core length ( AP/P) max = Maximum value of power spiking penalty \\;,M e e e e %e r'- 1.2-27

-m5 2......v -. - O. (~' FACTORS TIMT lifuBE CRITICAL GRLITY tRSS FL0d RATE SYSTET) PES 9K BOILIffi LBrifH EATE0 LBrifli =5 - S _:~T' ilGML DIA"EIER ~ a LOCAL PEAKING FACTOR..R. ~ , eO de mm mW 6 6 O w ee e e '-h* ' ~ '

n e. g l '. l ,1

• ]

., i :- + } l l i, i I, ,4': el'

' )

l Non-Boiling Nucleate Transition l Film [' i Convection Boiling BoilinD Boiling lleat Transfer I i.' i DNB l OfA l BTU / r-f t2 11 w ,u t i .l Ib e.-- AT (Surface to liquid) 4 .r ?( Figure 1.2-1 Boilin0 on the Surfacejof Cladding 7 i

ON @hq.h Wb ;.~ -- D ~ ' - M ' - J"1 ' ', N o C*r

e. ANNULAR FLOW Rfl f,$

I /- w I,

1) VAPOR FORMS A CONTINUOUS t,

y PHASE IN THE SPACE BETWEEN LIQUIO ~ 4 ~ FUEL ELEMENTS (SATURATED 0 VAPOR) l fjf_

2) SLOWER MOVING LIQUID TRAVELS

~ FUEL l l p FUEL ALONG THE FUEL ELEMEN1' g SURFACE _ l / 3) EXPERIENCEO NEAR CH ANNEL VAPOR f EXIT IN HIGH POWER CH ANNELS a d.-SLUG FLOW (LARGE VOIDS) y i l t4 S e' 1) d

j BUBBLES ARE FORMING ON z'

( S WALL ANO COALESCING TO D FORM VAPOR ** SLUGS ** IN E' ' _~ ~.b MIDCHANNEL FUEL { ) FUEL j

2) SIGNIFICANT. VOID FRACTION

' ' ' 7 i_ h ~ s' T 3) IN HIGH POWER CH ANNELS -:2-" }glh{ .%W SLUG FLOW BEGINS BELOW CORE MIOPLANE d .c. BUBBLE FLOW lQt y oJe

1) BUBBLES FORM ON THE WALL BUT 0

-J4 00 NOT COLLAPSE IN COOLANT ) s { STREew, l.E.," BULK BOILING" 0 8 FUEL l 1 FUEL

2) BULK COOLANT TEMPERATURE IS 16 0

0 AT SATURATION TEMPERATURE I(

3) BUBBLES ARE NOT YET COALESCING.

,; ) DL y; LOW OUALITY N ,Ili a

b. SUBCOOLED BOILING l

~ ~ R'! I

1) BUBBLES FORM ON THE WALL BUT

)I i COLLAPSE IN THE BULK COOLANT STREAM FUELk I IFUEL 2) NO NET VAPOR GENERATION

3) A VEliV SMALL VOlO FR ACTION 81 s

p R / l'

4) OCCURS ABOVE THE FORCED lilh CONVECTION REGION 1

I I, 3

a. SINGLE PHASE FORCED CONVECTION v'I 'lr i

1) NO BUBBLES ARE FORMEO. THE f

FUEL ;; COOLANT REMAINS IN A SINGLE FUEL PH ASE (LIQUlO) y

2) OCCURS AT VERY BCTTOM OF T

COOLANT CHANNEL \\ ~ d t-tc 8 FIGURE 1.2 2 TUEL CH ANNEL BOILING CONDITIONS 1.2. a g O 8 8

.__ --_-__--_ r. ~ ~ ~ ~ . O..- l l L '( , ' l' f V; I j' I ( e

l 1

i 1 il I I ? l I e h, COOLANT ENTilALPY xa =J A s I E FUEL ROD CENTERLINE TEMPERATURE D I d FUEL ROD SURFACE I TEMPERATURE H / COOLANT / TEMPERATURE Y u f i - N 3 NUCLEATE I$ BOILING oTFUEL

w g

ROO S I I AT FILM BULK BOILING (SATUR ATED TEMPERATURE) .i, .j p"# O/A / / l / N ,1 / / .\\ , i, \\ / N CORE N CHANNEL LENGTil DOTTOM CORE TOP FIGURE 1.2 3 PLOT OF COOLNNT AND FUEL BUN,DLE TEMPERATURE vs' FLOW PATH LENG I o ,P 3i l 1 ': i t t

S. W_.??$2*fh b. ^ -. N ?* ' '~ ?- ~ '^ i k AVERAGE PEAK ~ 385fF 4325 F g l -- 127EF 1377*F' l 'l s p l ) s l! 789 F 826 F l /

^ ~~

_ __ L ~ 57EF 57EF ... w..m -=?+-. 54EF 54EF ,l 1 i / I l l l l - 546 F 546 F 69 / l (N.4 l / / p g [. ' l l i / l /. l / s i / l / / l I /. I A BBLEt. m uRE FUEL PELLET LIQUID FILM ~ FUEL CLADDIN'G FIGURE 1.2-4 FUEL TEMPERATURE CROSS SECTION ~ 1.2-35

wM*-E'

. w u ; J,.J

? r.- ..,.p-1;L, -, l l. ).

r..

i .$/, 1 + <T % ~ yQ - TANGENT POINT p x VERSUS L C g. / GEXL, CORRELATION LINE / / .. _ ~- .,..~ s _.,... T (( g ~ * / X VERSU5 L CORRESPONDING TO CRITICAL { BUNDLE POWER ~' n / 4 3Q O CisT* j . @ = 5 rr-m? Ah s 6 a4 e @/ %.*h <c: us 0 / O z D= / / / ~ / / [ X VERSUS L CORRESPONDING TO / / OPERATING BUNDLE POWER / / / / / i BOILING LENGTH (L) FIGURE 1.2-5 CRITICAL POWER RATIO 1.2-37 .-_.---.L .~a. =-: -.

u. _ ;.

~ E.NCLOSURE 2 HEAT TRANSFER For Question 2 / Introduction The laws which govern heat transmission are, very important to the engineer in design, constraction, testing and operation of EWR systems and equipment. Heat is transferred in three different ways: Conduction, Connection and radiation. There are three different methods and are governed by three different laws. However, they share the. common principal that temperature difference must exist for heat, to be transmitted and heat is always transmitted in the direction of decreasing temperature. e e g a M 4 e O o e b o e a I e 'N w w mwa w sw ey e e y g,. g

F~-="--:- .i r TYPES '0F HEAT TRANSFER ~ CONDUCTION O l'IOLECULAR INTERACTION - HIGHER ENERGY MOLECULE IMPARTS ENERGY TO LOWER ENERGY MOLECULE BY COLLISION. O IRANSFER BY FREE ELECTRONS - THE ABILITY OF SOLIDS f TO CONDUCT VARIES DIRECTLY WITH THE CONCENTRATION OF FREE ELECTRONS. ~ CONVECTION c-t O NATURAL CONVECTION - WARMER FLUID CAUSES CIRCULATION BECAUSE OF A DENSITY DIFFERENCE. O FORCED CONVECTION - FLUID IS MADE TO FLOW PAST THE SURFACE WHICH IS BEING COOLED. 4 RADIATION O DIRECT ENERGY TRANSFER FROM ONE BODY TO ANOTHER WITH-OUT THE USE OF A MEDIUM FOR TRANSPORT. e. = -* - ,s

eav = = = - 6 7 ~ CONDUCTIVE HEAT TRANSFER ~ .I .\\ 1 6 = KA (T1-T2)_ X WHERE: 0 = HEAT TRANSFER RATE (btu /HR) K k THERMAL CONDUCTIVITY (BTU /HR-FT *F) 2 A = AREA (FT ) T1 = HIGHER TEMPERATURE ( F) T2 = LOWER TEMPERATURE ( F) X = MATERIAL IHICKNESS (FT) NOTE: K ('3 0 A URGE RESULTS IN MORE HEAT TRANSFER FOR GIVEN AI T O A LARGE A RESULTS IN MORE HEAT TRANSFER FOR GIVEN SUBSTANCE THERMAL' CONDUCTIVITY., K (Bru/Ha-AIR .0234 HELIUM -b.1375 > ^ HYDROGEN .1580 STEAM .0728 SATURATED NATER .349 CARBON STEEL 30.0 STAINLESS STEF_L 9.0 URANIUM 20.4 URANIUM DIOXIDE 2.9 Z1RACLOY-2 7.2 COPPER 228.0

{, { CONDUCTION i The first method of heat transfer to be discussed is conduction. This method is characterized by a transfer of heat energy due to the transfer of kinetic energ from one molecule of the fluid to another, by collision. For example, if a molecule is subjected to an increase in heat energy, the kinetic energy is increased. Some of this kinetic energy will be transferred to other molecules upon collision. Individual molecules are not heated but a fluid i (or material) consisting of billions of molecules is subjected to some heat energy source. The molecules in co,ntact with the source will have their kinetic energy increased and due to molecular motion, will collide with other molecules of the fluid, increasing their kinetic energy. Those molecules far from the heat source will experience less kinetic energy increases. Therefore, as the distance from the heat source varies the '( amount of increase in kinetic energy also. Varies. Since temperature is a measure of kinetic energy, it can be stated that molecules farther from the ,4 source will have a smaller temperature increase. Referring to Figure 3-1, it is seen that a temperature gradient across the material will exist. Factors affecting the transfer of heat energy from the source are

1) the thickness of the wall separating the source from the material (the thicker I

l the wall, the less the heat transfer),

2) the total area of the wall over which heat transfer occurs (the greater the area, the more heat transfer that occurs) and

3) the temperature gradient (the greater the temperature gradient the greater the heat transfer rate).

The above statements can be represented by '"

• t****' t""
• di'"*

heat transfer rate a wall tinckness (3-1) l

U

1. "a = ' L ^ ^ ~ ~=

c... ~ _ _ _ _ _ _. It is desirable to have an equation rather than a statement of proportionality. The tenn thermal conductivity is now introduced. The huids ability to con. duct heat is called thenral conductivity and has the units of BTU /hr.-ft *F. The above relation can' be restated as: heat transfer rate =, thermal conductivity x area x temoerature gradient wall thickness (,; Symbolically, Equation 3-2 can be shown as w-,.., X x A x (T -T) y 2 -(, Q= (3-3) X j==:2 where h=heattransferrate(BTU /hr.) K = thenral conductivity (BTU /hr.-ft. *F) A = area (ft. 2) T = higher temperature (*F) y T = 1 wer temperature (*F) 2 X = material thickness (ft.) i e e = pe e em M e = -e ,,m

m c cr.. s u. O Heat transfer rate is nomally dealt with on a per unit area basis. ~. Thus, heat transfer rate becomes heat transfer rate per square foot and is called heat flux (symbol. q). The equation for heat flux is. K x A x (T3-T)= Kx(T -T) 2 y 2 AxX X (3-4) i 2 Heat flux has,the units BTU /hr.-ft.2 or, if appropriately converted, watts /cm. Conduction is the means by which the heat energy moves out from the fuel. to the water. Heat energy is generated as a result of the fissioning of Uranium-235 into two smaller atoms (fission fragments). Tne major portion of the energy from fission occurs ~due to the kinetic energy of the fission b fragments. Tnrough conduction ~ the kinetic energy is transmitted to other atoms within the U0 fuel rod. The atoms of UO on the surface of the fuel pins 2 -transmit their increased kinetic energy to the He atoms which in turn transmit s some of their kinetic energy to the In-2 cladding. Tne heat is finally conducted thn: ugh the cladding to the water surrounding the fuel rod. Under nomal conditions, the surface of the fuel rod has a layer of water a few molecules thick at zero velocity, through which heat is transferred by conduction. Further heat energy transmission is by convection. ...m =- - .---.m. =. = -. - ,_gy7_ - = *

.._u_. y CONVECTIVE HEAT IRANSFER .i THE AMOUNT OF HEAT REMOVED FROM A SURFACE BY CONVECTION (5) DEFENDS ON THE SURFACE CONDITIONS AND THE FLUID AND FLOW PROPERTIES (h) AND THE TEMPERATURE DIFFERENCE BETWEEN THE SURFACE AND THE FLUID (AT)", s 4 = h AT ~ 2 4 " HEAT FLUX FROM SURFACE, B tu /hr-f t 2 'h = CONVECTIVE HEAT TRANSFER COEFFIb1ENT, B tu / h r-f.t F AT = TEMPERATURE DIFFERENCE BETWEEN THE SURFACE AND THE {. o

FLUID, F

2o MECHANISM h, Btu /hr ft p FREE CONVECTION,' AIR l-10 FORCED CONVECTION, AIR 5-50 FORCED CONVECTION, WATER 50-3000 BOILING WATER 500-5000 l FORCED CONVECTION, BOILING WATER 5000-8000

7--~~- p f 'CONVECTI'ON j Heat from a source reaches a greater number of molecules in a fluid if the This mechanism of heat trans'fer is known as'1 CONVECTION. fluid is circulating. Tne fluid in a BWP. is forced through the fuel region by recirculation pumps. The moving fluid accepts heat energy from the fuel rods via the stationary i layer of the fluid surrounding the rod and transports the heat to the coolant by convection. Thus, as the' fluid travels the length of the fuel rod, it continually accepts more heat energy (temperature and enthalpy increase). ~ For purposes of calculating the heat transfer rate for convection, an (~ equation similar to Equation 3-2 is used. (See Figure 3-2 for illustration 6f the terms). 1 K/(T -T) s f 0* (3-5) Xf where K = thermal conductivity of the stagnant film (BTU /hr.-ft. *F) f 'l =at

• w
• -m ww ww

= m, .,e,.--m ..m . m.

~ " - i T = surface temperature (*F) s T = fluid temperature (*F) f X = stagnant film thickness (ft.) f 'I. Also,. heat flux (Q/A) has a representation similar to Equation (3-3): 1 K (T, - T ) (3-6) q =Q_; f f A

• f The value for X, the thickness of the stagnant film, is difficult to determine.

f O, It is not a fixed quantity, such as the thickness of a fuel rod. X will vary f dependent on the rate of fluid flow, the fluid viscosity, the heat flax, the o type of surface and the state of the fluid (liquid or vapor). Consider the ratio of K to X and call it the film heat transfer coefficient, h which f 7 f will also vary due to system conditions. In equation forn, l f=h f (3-7) Xf Equation 2-5 then becomes q'= Q/A = hf (T -T) ~ (3-8) f l l l___....

c _.-.. - RADIATIVE HEAT IRANSFER 2 IHE AMOUNT OF HEAT TRANSFERRED BY RADIATION FROM ONE BODY TO ANOTHER (k) DEFENDS ON THE MEDIUM BEDIEEN THE TWO BODIES, THE SURFACECONDITIbNSOFTHEBODIES(F)ANDTHETEMPERATURE DIFFERENCE BETWEEN THE BODIES (T7 - T )" 2, 4 4 4 = Fo (T -T2) ~, y 2 4 = HEAT FLUX FROM BODY 1 TO BODY 2, Stu/hr ft F = SHAPE FAbTOR ACCOUNTING FOR SURF CE CONDITIONS {, e = STEF,AN-3OLTZMAN bONSTANT = 0 1714x10-8 s tu/hr f t, o 4 z g ..,R T1,2 = TEMPERATURE, O =. - - - - - =-.

=. _:._.=- ~ RADIATION 1 Heat transfer by radiation differs from heat flow by conduction in that the medium by which heat transmission takes place, does not become heated. For example, the earth receives radiant energy from the sum I whereas the space above the earth does not become hot. When in winter, cars are parked close to a building, they receive energy from the building and remain warmer than they would be if parked further from structures. Radiation heat transfer takes place in the following stages: 1) Conversion of thermal energy of heat source into an electrar.agnetic . waves { 2) Passing the electromagnetic waves through space 3) Reconversion of the electromagnetic waves into thenral energy at the cooler medium. Finally this heat transfer is governed by the 'Stefan Boltzman equation namely I h = F a (T -T) 3.6 4 2 where ~ h = Heat Flux from body 1 at Temp T to body 2 at Temp T i" 3 2 2 ( BTU /hr-ft F = Shape factor accounting for surface condition o = Stefan Boltzman Constant =.1714x10-BTU /hr.-f t -R ) This equation reveals that due to such a small Stefan Boltaran constant, 1 for the heat transfer to be appreciable, AT, must be rather large. 1 l l ~

J ~T- ~ ~ 2' - ~ BOI' 'NG HEAT TRANSFER ' Boiling is the evaporation of a liquid to vapor occurring within the body of the liquid by the mechanism of bubble fomation. The very term " Boiling Water Reactor" implies great dependency is placed on the boiling water as a primary mode of heat transfer. To become conversant in.the teminology, it is convenient to use a simple, example to demonstrate the fundamental principles of boilingibeat transfer. The example selected, as shown in Figure - 3-3, is a pan of water on an electric stove. .- ; 7. - '~ The pmcess consists of a pan of. water at room temperature and atmospheric pressure, resting on an electric heating coil. The coil and pan have a finite 4 ' contact area "A". The amount of energy given off by the coil as heat can be varied by changing the position of the heat control knob (varying the electri-cal current). Assuming that the pan and coil are in good contact, heat will be transferred from the coil to the pan and then to the water. Tne heat transferred "h" divided by the contact area "A" then yields h/A or " heat flux" in BTU /hr.-ft.2. l l Turning on the electricity will supply a mall amount of heat energy or l heat flux. As the temperature of the pan increases, heat will be trans-ferred to the water (via a AT) by convection heat transfer. As the water l becomes hot at the pan surface, it becomes less dense and rises, being dis-i placed by more dense, coo'ler water ficwing in from the sides. This particular rode will be called Region 1 or " CONVECTION HEAT TRANSFER". anis region is characterized by the phenomenon just described. l Next, increas.e the heat flux by turning up the heat control knob. Increasing .__y.._ _ ___ _ q

. ~. - -. ~..... -.... l the heat flux increases the pan temperature (recall h/A = UA(AT). When the pan temperature is about 10*F higher than the bulk water saturation tempera-ture (212*F 0 atmospheric pressure), bubbles will begin to fom on the bottom of the pan (Figure 3-4).. As the. bubbles get larger, their. buoyance will eventually cause them to be detached from the surface of the pan. The bubbles will drift off into the bulk liquid. Here the bubbles M11 collapse and give up their energy to the surrounding cooler liquid. This is an excellent means of heat transfer from the pan to the water. The bubbles serve,to " stir up" or " agitate" the stagnant fluid film, thus improving the thermal ~ conductivity of the film. Additionally, each bubble carries off nore energy C'. Recall that as a fluid changes phase, than is possible by normal convection. its enthalpy increases; therefore each bubble, being vapor, is something akin to being a "little bundle of energy (enthalpy)" and the heat removal rate is greatly enhanced. The phenomenon just described is called NUCLEATE BOILING because bubbles are foming at sites of nucleation on the pan surface. NUCLEATE BOILING is extremely important in reactor core heat transfer because a great- -- - deal of heat energy can be transferred without extre,ely high surface temperatureE The next area of discussion will require holding the heat flux right where it is for a while and allow the bulk temperature of the water to rise to satura-tion temperature. When the bulk liquid temperature reaches saturation tempera-ture, the bubbles released from the pan surface will not collapse in the liquid f ~~ _y3 -~~

3-r- -- but travel all the way to the surface., This particular phenomenon is, referred to as " BULK" or saturated boiling (Figure 3-5). NOTE: Nucleate boiling is still occurring because bubbles are fonning at the pan surface! The fact that the bubbles are not col, lapsing is irrelevent to the nucleate boiling phenomenon. BULK BOILING and SUBC00 LED BOILING are tenns used to describe what ha. = .- t. the bubbles AFTER the bubbles. have left the surface. In both cases, nucleate boiling describes only the phenomenon occurring at the pan surface. In all, ~ of the vast nuclear power field there is no greater confusic,n existing than that which occurs when operators try to describe the terms, nucleate, subcooled O and bulk boiling. Return the heat control knob and turn up the heat flux more and more, other phenomena will occur. As the heat flux is increased, more and more bubbles fona until cany of the spaces on the pan bottcrn are covered with bubbles. . Although the bubbles are moving away from the surface, it appears vapor is l on the surface for a longer time. The higher the heat flux, th9. more the l l heat transfer mechanism deteriorates until a steam blanket is fonned.' The I l onset of this phenomenon is called ONSET OF TRANSIITON BOILING (abgre-viated OTB) (Figure 3-6). This signals the end of nucleate boiling and the l ~ cnset of heat transfer deterioration. As the heat flux continues to increase, bubbles continue to fonn at a faster and faster rate. Finally, in one portion of the pan, bubbles will fann so fast that the liquid cannot displace the ~ ~ ~ " ~ s .o ~~~-.-;- --w-

, s

. =..:.... :.. =. u.-. - - ;- I bubbles and they combine to fonn a " bubble" or " vapor" blanketed region on the pan surface. This is a highly unstable condition. The bubbles may move away from the surfa :e, allowing liquid to again wet the surface; or the' bubble blanket may spread until the surface is vapor blanketed. This mode of boiling heat transfer is called TRARSITION or PARTIAL FIUi BOILING and is an extrerely, unstable form of heat transfer (i.e., oscillating between vapor blanketing and. nucleate boiling). Assume it is able to maintain the unstable partial film ~ " boiling mode for a moment. "Now increase the heat flux just a little and see what happens (Figure 3-7). o With the first " tweak" of the heat control knob, the partial vapor blanket would spread over the entire pan surface. What is found is a thin layer of steam everywhere on the pan surface. As shown in a previous table of thenral conductivities, steam.is an excellent insulator; hence, the pan surface now presents an extremely large resistance to heat transfer. Assume again the j electric heater can maintain the desired heat flux, it is known from the l l ecua tion: l Q/A = U(T surf liq thatifh/A is held constant and U decreased drastically and T)$q is constant, then T must increase rather substantially (on the order of 1,000*F). surf t l If the pan surface was a fuel rod in the reactor core and full film boiling l 1 l =.- =_ - _ _, = =. =.:...

" ri L ~... --j_______,_ r o ~ were allowed to take place, the metal of the fuel, rod might have melted, been seriously deformed, or cracked. Radioactivity would then leak from the damaged red into the coolant with rather severe ramifications.

OTB, transition boiling, and full film boiling are neve'r permitted in an operating reactor under ordinary 'or transient conditions.

Plotting the results of the experimentation conducted thus far (Figure 3-8) yields a graphical representation of heat flux (Q/A) versus AT (T surrace ~ liquid) with the Regions of heat transfer and the point of DNB annotated. T. The regions of heat transfer are described by the method of heat transfer (~, as follows: ~ m Region I - Characterized by conduction and convection heat transfer No boiling-Characterized by nucleate boiling Region II Transition region - characterized by partial film boiling - Region III ~ unstable region Film boiling region - continuous vapor blanket - Region IV radiation heat transfer One further clarification is necessary concerning nucleate boiling, i.e., it is not necessary for the bulk liquid to be at saturation temperature for nucleate boiling to occur. Thus nucleate boiling as an effective heat transfer mechanism can be utilized in the BWR and yet maintain a substantial -] ~~--

~ J ~ margin to DNB by operating in the left hand portion of Region II in Figure 3-8. For a better idea of what is going on in the core, it is best to look at an average fuel bundle. Refer to Figure 3-9. Coolant enters the bottan, flows up around the fuel rods, and absorbs energy frczn the heat transferred by the fuel rods to the coolant. Notice the plot temed relative heat flux. (" HEAT RUX PROFILE"). The heat flux produced from fission in the core will assume a s'hape as shown. The highest heat flux (greatest fission rate)' occurs ~ in the interior of the core. The dotted line specifying " AVERAGE Q/A" corresponds to the average heat flux for the bundle. An explanation of AVERAGE HEAT RUX follows. The core may have a themal outp'ut of 2420 MW. If there were 724 bundles in the core, the AVERAGE FUEL BUNDLE POWER would O be 1/724 of the core power if eveg fuel bundle was pmducing the same amount of energy.. To obtain 1/724 core power, each fuel

• bundle cust operate at some average heat flux (Q/A).

As previously stated, the highest heat flux will be in the core interior. Tnerefore some portions of the fuel bundle have a higher than average heat flux and others a lower than average heat flux. The nuclear core heat flux profile also is useful to plot curves of coolant and fuel bundle temperatures versus flow path length up the core (Figure 3-10). The coolant temperature experiences a continuous rise as heat is added until the temperature reaches saturation temperature and begins to bulk boil. Coolant temperature remains constant from the point to the core exi t. Notice the sharp temperature rise at the higher values of heat flux indicating energy is being absorbed at a high rate. The coolant temperature profile is not descriptive of coolant energy increase because the coolant is changing phase. A better description is obtained by plotting coolant enthalpy change. Notice again the steep rise in coolant enthalpy when the heat flux is maximum. _m. _ =. _ ._m_.___

1.w.: ~ ~ :

= --

..-~~-_.:._- Examine the fuel rod surface temperature profile on Figure 3-10. The fuel rod surface temperature rise $. continuously and levels off at a rather constant ~ value above the coolant temperature.. The initial rise is caused by the AT across' hfil" IIsurf - T)$q). the film required to accoamadate the heat flux, Q/A = Tne fuel rod temperature levels off due to the beginning of nucleate boiling. Nucleate boiling is an excellent heat transfer mode and evert though the heat ~ flux is getting greater, the AT acrdss the boiling film remains "relatively: I.~. .;..l ? cons tant. Next, folicw the fuel rod centerline temperature up the fuel bundle. The centerline temperature is above the fuel surface taperature. The amount the'. centerline temperature is g eater than surface temperature will depend directly -Tsurf). No'te the beneficial effects fuel (Tf on the heat flux, Q/A = X of nucleate boiling on centerline temperature. As nucleate boiling is occurring on the fuel rod surface", the fuel rod' surface temperature is only slightly greater than liquid temperature. This automatically keeps the fuel centerline taperature at a lower value than if convection were the mode of heat transfer from surface to liquid. For a better perspective of the different modes of heat transfer, consider a single fuel rod in a bundle (Figure 3-11). The diagram graphically illustrates the fuel surface and centerline temperature profiles under three different condi tions:

1) nucleate boiling with a low heat flux;

2) nucleate boiling with a heat flux four times that in Profile (1); and 3) film boiling with an even greater heat flux.

The heat flux in the individual rod is a function of The power for Profile (1) is less than Profile (2) since (1) reactor power. has a lower heat flux. Nucleate boiling is the primary code of heat transfer

y,,.. I at the fuel rod surfaces in both case (1) and (2). Notice that even with an increase by a factor of 4 in the heat flux, the rod surface temperature is only slightly increased. Looking at Profiel (3) where the heat flux is of sufficient magnitude to cause film boiling, the heat transfer mode at the rod is film boiling. The fuel n.. rod surface is insulated by a steam blanket. Notice the temperature gradient _ f, between the fuel centerline and the rod surface is approximately 3000*F. Also note the significant increase in fuel rod surface temperature as compared ~ to Profiles (1) and (2). It is for this reason that the reactor power is limited to below that power level which produces film boiling. Once the heat flux (power level) is sifficient to cause film boiling, all temperatures in the interior of the fuel rod and at rod surface significantly increase. Tnis temperature increase can cause danage to the fuel (melting) and to the fuel rod (buckling or cracking). 5 m l l ,.Y """**"****=m .m,

-j ,c__ _._:.._. p g... [i TIE CONSTANT l When the, reactor is operating at a stea'dy state power l'evel, there is a fixed temperature distribution fran the center of the fuel pins to the moderator. Tne temperature distribution is governed by parameters such as the specific ~ i heat, density, volume and heat transfer coefficients of the fuel materials and the heat transfer area. TN C .g The response to a rapid change in the fuel temperature is measured in tems. of the fuel thennal time constant, Using Figure 3-12. -h 2 (] ---T2(-t)-T.- -= (T Tp-{-ize[T T ,b 4. b 3. where .g3 initia1 fuel centerline temperature T = 3 I"9 ""

• • E*#" "

T2 T'= final fuel centerline temperature I time t = time constant = r As one can see, a short time constant will result in a rapid response to power changes and a long time constant will result in a slow response. For 8x8 fuel, the thermal time constant is about 5-6 sec. If the previous equation is used to calculate the cladding surface temperature after one time constant (t = T), it is seen that after one time constan', the cladding c - ~ ~

_m______.._____.._.. ~ { surface temperature has reached 63.2% of its final temperature. Thus although the initial response to a power change is very rapid, the rate of temperature change is governed by the fuel thermal time constant. O e 9 I = 9 e O e O t e e p. e 9 g 6 e 9 O e w n

• -m e - e

~ .. M - e =-

~.... - - - - c [ HEAT TRANSFER IN TRANSIENT CONDITIONS Temperature distributions in the fuel and moderator during transient conditio'ns is a complete fur.:. tion of many variables. The interaction between conductive, convective, radiative and boiling heat transfer as well as pressure changes and local power variations affect the temperature i response of the fuel during a transient. If one looks at the average surface heat flux being ' conducted from the clad in transient conditions,_..,; e :c one can see that for slow transients the change in heat ~ flux is slower than the change in power level, indicating a relatively long - thermal time constant. For rapid transients involving large changes irr power and pressure, the heat flux variation depends on variables such as fm the boiling rate and the rate of heat transfer to the surface of the s' clad. 4 1 1 l 1 l l l e .e m m = *

• 7------ - -

,y 1 ?r I .\\ 1 / j, d i / W x p. l / 7" .l j T 1 / j / T1 = INNER TEMPERATURE / t / T2 =. OUTER TEMPER ATURE .t IlEAT >- ' e, / ./ x . = MATERIAL THICKNESS i I A = AREA OF TI-lE PLANE WAL i / ll / l ti / (T 2 INSIDE OUTSIDE l .l Figure i-T Section of a Plane Wall I . 4,; 1 .i. .}

)

C o. I t - STAGNANT FILM i1 .l. ' /

d..'

/k --4 i. WALL lo j g i i .i >1 'M FILM THICKNESS 4xf;l i / ! .i, / ; r t / /) 9.; FLi ,a /) (-lEAT / l' ]> ' /p l. /)- tSUR FACE.LTsh /( /t

(

(t s' g tpLUit)4T;' f!!. f ,;i. 1. x.. +. 1 ) e 8 t Figure Convection Reat Transfer Process a.1 .I 3, .l j

[ 1 Q lii i 1 ,1 l

1. i

I

.ii \\ I. - ~- ,= l ,\\

l q

i i, CONVECTIVE CURRENTS 'I 1 1: a o {llE AT CONTROL k 'E 1, KNOn l-NN >-- ELECTillC CunnENT 6/A '.;l' -l Figuro ' 3 a Region I: Convoction Heat Transfer I I t ,l 'e- ,1 i n ) g al

s n - - - ~ -

7. -

3 ---- 1 'J ' I \\ r i COLLAPSED BUBBLES

u,ul

( 1

i.

i = -h -- L'i p> b, ,l lo i { s l l ,a ? f(. I \\ i. y IBBLE j , bRMATION m3 O O' O a. k <r p; i 1 i L. _ %on_ A n__ n-2 _ _ _ _. _.1 J .a ._m_ . u.h b. ',... b J '.. . ); i ...o . v.. n c{ c l1 Figure a-a. Region II: Nucleate. Boiling, Subcoolect Liqu .... 1,. . o. v.. tg

y>.,...

.,r, c..,., I g ;' 7f f

t l

\\ O c s (- ii, 1 l5 f, VAPOR GENER ATION 1 l 1 \\ u. l fI ll in- -u w:_, - - - -= ~ -L: ) .{ [- 4 'i a \\p\\ . (q Q) 1 .4" 4 0.... p l 'li~ l L O' 1 O; / o' n l A( 4 h 7 an A L . n _o-. .. n.. - Region II: Nucleate Bol ling;. Bulk or Saturated Boilirk. Figure 3-s .,. w.. v. I p l,, i 1

• j..

c. .. a, . s'. ') ,...:;,. ;. v 1.- .,' c ) - ^

.g - 4 .. _ }. q1;j-,.j,' ;p=,..7, , w. :; y...% 4 ,.., 3.... .,....g, .e r. e b ~ e am m. g ,e., e 1 l L. I. l s ',- ' * * ', ' = \\ = '* *l ',-..' -D; '. ; & :..ll .~: -i .~g k.1,j *. r,2 -l-5.* -l ..- 3 ;i t.,. - l ,,.j. 4 l j 2 O g i e i I J W i i 1 o C~I ( A "v

e gO b.

e s<I -c e 1 s w A i l M c - y H l t.r. W ,l H 9 J ca m o H r.n. 1 Z D 4 ~~~. N [ .O e C' {N e b "444D

• 'ame

-m.a . N. l m_ _= .q

i l O m.t N l t t 4 ui :3 [ -s-. __, <- n 1 .s 4 p 3 .y .a 4 i l VAPOR BLANKET ~~ i 'l : ,l m ( {- J t j u _ _ _._ _ _ _ _ _._... _ a g_ -.

m. _ _ _

_... 1 _ t_ b 4y -j eyyy e _p O/A Region IV: Full Film Boiling Figura. 3.. 1. .h g

• • b

( .L

s. %; 7,- _ _.:- - -... 4 O - ~..-- .. ~ -~ g y .,.i e $Q 2

• J$,1J e

.t o-o ~ u.... -a. ,? --, 3 1 y 1 s:.=-: - m-

=

w yt

, x -

. u-r. c um mq AA e E i _c 1 c-Q C l q ,p-l i ! a .z ..>.b z f = .r .o C.:i. _=., - y o o l g, p y. 3 y Le s b 1; m c f q + m ~ a i u-. e e ?. LO x ,c -1 a Ae 7: ->s D-c 3 Al.4 g } Da f, c/; 4 +- 2 /:a p u .S o . -sq 2 4 5 1il ca w o >4 2. m. -e e o e v2 1 0 e k 1 a 1 c .s i Z o

3

-a o u; t C 1 \\. I, w4 s-(g4-24/nze) y/o . = - - - -. _ s

. - ;..,.. > : - d.. ,. r._. e ; :: ......c ..i.. r.. Y ~ k -0 C... W... <' ' * - g -.c.. e. -C X C D W J m a N l, P< W s i C ~ m- ? g y-..- g. e C y. c W L g q b 9 ~ is e. s O C s Q E c a Le q 2 C O s p s p .E O c C g ~. M 1 O ( V i / 2 --e .e tr. 4 SCOM 13nd M N 1 a i s O r 1 ~ ._g---

Ly . c _ __._2. _x - g ;

• '"'*;.s'n'3...[-

'z ,, pg_,- - J y s .S ( e 2 COOLANT - ENTHALPY c. 1 2 S h FUEL ROD CENTERLINE I... ~=- 4 er w V P'-

r,_ r.

..-,.i

T; =.-

' ;4 '_ - I TEMPERATURE.. ~ gf.yb,.A* cf.. i. 1 -Mj 7 :. g ^, p-INLET [ FUEL ROD SURFACE --f SUEC00 LING TEMPERATURE ~ ~ COOLANT ~ TEMPERATURE i C g NUCLEATE - l ~ BOILING AT FUEL ROD g l m I c: ne bet m AT FI g g w ww g. c. 6 SATURATED NUCLEATE 3 OILING w HEAT FLUX N \\ i l FX / N <a w _.s / N = u. / \\ / ^\\ l / CHANNEL LENGTH TO ( DRE o p 20TTO' ( 91 ~ D 4 COOLANT AND FUEL EunoLE TEMPERATURE VS. FLOW PATH LENGTH yt ~ + e# FIGURE 3-10 g-4'

Q s.. 3._s hs.,., .,. :.. _.; \\. '.. y .. C Cle.%.'* ~ ? c., s ~^ (*.

• ,..7

~,7' .. ~

• 4

. #..e .d%.w. s cc g 1 c.,. seco. C UO2 'O 2 z. g& -, e Q

• ~..*N_;....*.

Q o D .3- '. D o g g <2 a a < a m w a _a 3CCO LL. O r.- TM ? O u. ~ gb3 ojw ~ 4500 as 2. k . ]..:. -? ' ..t> j p ee.bs ^ lL&: m j Q\\e& ; g g 1 35CO E l = 2 y a ci: .j !. h o,,1,J e s..b> d w; w 2500 l W '. 9 Ctw $. ' b 4 5 g 1 a \\ c ,d ' af-+LD ( 500 \\ N \\a a.o At A *' M -3 Wt g-y g% s e. 2-u cnects or Heat i ransrer fvfodes-r'ar a SingIe r'~uel R6c ~ ~ ~

5. Lure 1

. - - = = -

^ R' 'i. ~ ' s. 4 f s & _. : e '. =;; _ ?,. . ~. '.. Y. < ;. ' TIME CONSTANT . T' ~' gace 3 Cy\\d 'f -[i( = (c!C.'.C'- e '" ,e a c sa 4

.w.

_~. .., a :. n.,,- - ~ - Q. '{ :.t'- g. y 3.- m E m H T e-r. - < 1 't T I T ll'.\\/ T' 2 ~ 2 - ',s ...g. TIME ~ T, s. h'*\\,.e.+5 3 T* I ine cc54 4-cth A a* ~ O T c-T n.r m H T T 1 3 ~ T T 2 l 2 o TIME 1 T T I N T' ~ 2 t T a s 2 NN N ~ ifa 1 s \\\\N.Ns 'N N i-I \\\\ \\g N s g\\\\s' 8 N 63% of T ' 1 N.x 2 \\ s \\ N

• I s

s, T) 1 s i l T I T 2 g l 2 ( TIME t=0 T TIME CONSTANT C FIGURE 3-12 = =

v--~-- ENCLOSURE 3 DEFINITION OF THERMODYNAMICS TERMS For Question 2 s ENERGY - ABILITY TO DO WORK OR INTEGRATE USED POWER OVER TIME ENTHALPY - MEASURE OF THE ENERGY CONTENT OF A SUBSTANCE MEASURE OF THE ENERGY CONTENT OF A SUBSTANCE WHICH IS ENTROPY 25 UNAVAILABLE FOR CONVERSION TO USEFUL WORK HEAT Ftux - RATE OF HEAT TRANSFER THROUGH A SURFACE AREA P0HER - RATE OF TRANSFERRING ENERGY [ OR ENERGY SPENT PER UNIT TIME PRESSURE - THE FORCE EXERTED BY A FLUID ON A SURFACE QUALITY - MASS FRACTION OF STEAM IN A STEAM-WATER MIXTURE SATURATION - THE CONDITION OF TEMPERATURE AND PRESSURE WHERE THE VAPOR PRESSURE OF THE FLUID EOUALS THE SYSTEM PRESSURE SusC00 LING - THE DIFFERENCE BETWEEN THE ENTHALPY CF THE COOLANT AND SATURATION ENTHALPY IEMPERATURE - MEASURE OF THE HOTNESS OF A SUBSTANCE VOID FRACTION - VOLUME FRACTION OF STEAM IN A STEAM-WATER MIXTURE 2-1

~ ~,- ~ T'tiERM0 DYNAMICS Any material which absorbs or transmits heat is known as a wrking material. The amount of heat energy can be quantified by measuring temperature, pressure and specific volume. These sources will result in enthalpy of the working material which delineates the state of the working substance. State of a wrking substance depends on knowing the thermodynamic properties of a working subsLance. The thermodynamic property of a working material depends on temperature, pressure and sp:cific volume of the working material. Tenoera ture ^ s ( Tenperature is a measure of hotness or coldness -f a substance. It is an indication of the energy content. On a molecular scale if the temperature of a substance is absolute zero, it is an indication of zero energy content since the atoms or mole-cules of the substance are immobile and unable to do any wrk. Absolute zero tem, perature has not, been reached even in research laboratories even though;the scientis9 have gotten very close to it: As temperature of a substance increases,sthe movement of the colecules increases and thus, their ability to do work increasas. Teraperature is measured by instrumentation and there are four different scales l which relate to tenpera ture. Absolute zero is referred to G Kelvin or 0 Rankine. These are two temperature scales which scientists use to relate to absolute zero. l The two comon tenperature measurement scales are, however, centigrade and Farenheltc The absolute zero on both of these two scales is -273 C and a60 F,respectively. 2-2 '***6=- Oam .e e eg = g.mes mM- ,www

' Cent'igrade 1s a metric scale which is calibrated to zero degrees as the melting ~ paint of ice and 100 C as boiling point of watcr, both at Standard Temperature and pressure at sea level elevation. The Farenheit scale's indication of some states is at 32 F and 212 F, respectively. See Figure 2-1 for clarity. Conversion of these scales,is as follows: ~ K=#C + 273 R= F + 460 "C=f(F-32) F=fC+32' ~ or Pressure S Pressure is measurement of force per unit area. Force is defined in terms of product of mass and acceleration. If either the mass or acceleration is increased, the force is likewise increased. On a molecular scale force by a substance is the i physical contact of the molecules with their corresponding acceleration. The higher the acceleration, the more movement and therefore the greater the tm-perature. Therefore, pressure and taperature are interrelated. This is the reason Pressure is increa9 as taperature of a substance increases, its pressure increases. if either force is increased for the same area, or if for the same force, the area i is reduced. l If molecules or atoms have no movment, thus zero temperature, they do not have any force which relates to no pressure. This state is referred to as absolute zero pressure. This state is also refe.rred to as perfect vacuum. This is maximum vacuum-2-3 = --. -__=:._-..-. _

_ = - - ~ ~ ~~ J g g H M 5 M. ~

}

m 8 5 m n .,t rg s .J ^ s a a W M J g-c nas i H e y S c a = o S e o e c O O O e O 4 o ~ i e I ( M J 0 ~ s s Q H

• A z

m W J C ~( O ) c. c w H s e z e e C C tlu N e N c 5 O C3 g m cs es v e t 3 h c ~ j cy s e r m C e 1 k t O .E = w H u. z k. o ~ N e I b e M c I N e n -m + u. M-5 o ~ b 4 S 5 T l u. W H O u. F

u. <

g C O-C_ D 2 J l Hw-N w =- 1 d5H w,5 00 l co<_= w2<

• A a

=c Om

==Er Ac3 <N

t -- ' whic'h can exist in ultiverse i.e.", when no atom moves with increase in molecular movement er temperature increase, pressure is increased and consequently vacuum is decreased. At atmospheric pressure and tenperature, there is no vacuum and a pressu ~ equal to 14.7 pounds force per a square inch is existed on everything. At this state, most pressure instrumentation show a zero pressure indication which is also referred to as zero gauge pressure., Any increase in pressure of a system is shown by the instruments as a positive gauge pressure indication. To determine the absolute pressure indication, the gauge value must be added to the 14.7 pounds per. square irches. Pounds per square inch is one scale of measurement. Pressure can also be measured as a head in inches of mercury or inches (or sometimes feet) of Note that water. Figure 2-2 exhibits different scales of pressure measurements. atmospheric pressure indicati,on in inches of mercury is the same as condenser back pressure as indicated by the instrumentation. ? Enthalov Enthalpy is a measure of the energy content of a substant which can resuit in some type of w3rk. The energy which is in a system is a combination of the internal energy and the amount of work which a substance can do. The capability to do work of a substance 7.s measured by its enthalpy depends on the state of the substance. For a fluid, the enthalpy of a substande in a gaseous state is considerably higher than in liquid state. Tables have been developed for various working fluids which provide values of enthalpies in liquid and vapor state at various tabulated values j of pressure and te3perature. l l 2-5 i {- ~ _ -

T -.'-.' CdMPhRISON OF PRESSURE MEASUREMENT ~ SCALES ~~ ~ ~ v AND THEIR UNITS ATMOSPHERIC HG COLUMN H COLUMN GAUGE ~ ABSOLUTE PRESSURE. ... PRESS' PRESS 2O PRESSURE . PRESSURE VACUUM ~ PSI PSI INCHES INCHES INCH - Hs FT - H2O GAUGE. ABSOLUTE Hs VAC.~- Hs. ABS. ~ PRESS...- PRESS h 204 231 3 2 i 3 100 114.7 2 2 E ~ 15.7 i 2.04 : 2.31 1 3 N/A. N/A A A ~ 1.0 .88 .491 -15. 2 1 g O 14.7 0 30. 0 -0 ~ ~ = .1 14.65~ .1 29.9 R y y g Nor NoT d 5 Usso USED a ~ -1 13.7 2.04 ~ 28 -14.7 0 30" 0" 1 Ficuas 2-2 2-6

] .:.: r - - -.:.. Entroov -- - ) Entropy is a measurement of the energy content of a substance which is untransfer-rable or convertible to a useful work. Since, in a power plant we deal with wrkable substances, this is not a usable term or subject. Satura tio n I This is referred to a state of a fluid where the pressure of the system in the liquid form and in its vapor form are the same. In normal atmospheric pressure, a coolant temperature such as water is increased by adding heat to it. As was pre-viously discussed, as temperature increases, the pressure increases ('on the moleculag basis). If sufficient molecular pressure is increased, the molecules overcome the p surface tension force and vaporize. At this point, the liquid pressure is the same / At this point, the vapor pressure is 14.7 psia or zero gauge as vapor pressure. ~ pressure. Clearly, if the pres'sure above the water is reduced, the molecules vaporize sooner at lower taperature. Likewise, if pressure above the liquid is increased, it will require higher energy or higher temperature to reach vaporization c Another term used for vaporization is boiling. In a BWR, the Eater at 1000 psig boils at 545 F and if pressure is reduced suddenly, considerable boiling or voiding will take place. Likewise, if pressure is suddenly increased, collapse of bubbles or transfer back to droplets will take place. If the temperature of a fluid in a vapor state is greater than saturation taperature at that pressure, it is referred to as a superheat steam. Similarly, if the fluid temperature in its liquid s tate is lower than the saturation taperature at that pressure, it is known as subcooled liquid. Subcooling is measured by the difference between the enthalpies of the saturated coolant and enthalpy of the coolant. 2-7

- - ~ = ~ ~~ _ _. Y " ; ; ~ ~- PHASE CHANGE n H-0-n ~ SOLID WATER EXHIBITS 3. e H-0-H H-0-H H-VERYRIGIDMOLECUbRARRANGE -H H-0-H H-0-H H 3g37, H-0-H H-0-H H-0-H H. H H-0-H H-0-H H-0-H H-0.- e THERE IS VIBRATIONAL MOTION 0-H H-0-H H-0-H H-0-H H-0 0F THE MOLECULES, BUT THERE h-0-H H-0-H H-0-H H-0-H H.IS NO MOVEMENT OF MOLECULES H-0-H H-0-H H-0-H H-0-H H-FROM THEIR POSITIONS, H-0-H H-0-H H-0-H H-0-H H H-0-H H-0-H H.0-H AS HEAT IS ADDED AND THE H H-0-H H-0-H H-TEMPERATuRERISESfTHE -N N-0-H H-0-H VIBRATIONAL MOTION INCREASESc u_n_u s AT A SPECIFIC TEMPERATURE, ~ SOLID WATER IF MORE HEAT IS ADDED, THE ( SOLID CHANGES PHASE INTO A ~ LIQUID. e DuRING THE PROCESS OF CHANGI FROM A SOLID TO A LIQUID, TH@ ,Q'Y --g'D' TEMPERATURE REMAINS CONSTANT ( D 4 BUT ENERGY IS BEING ADDED TO g p g O d q @ p p 4 ::: Y CHANGE THE PHASE. H-o.g o 4,,,% ~pfs-D[//' 0 IN THE LIQUID PHASE, THE 8q g# g 0 % y R-0-H d'4 i MOLECULES HAVE FREEDOM TO MOVE FROM ONE POSITION TO H-0-H P 'Oc g# 8so'8 8 7 ANOTHER IN ADDITION TO M *P 4 g 0 *4 d o VIBRATIONAL MOTION.

1. 3,49 H-0-H u-0-H h 8,o-N 7, o

g 'h'55 l_IOUID WATER 2-8

~ -- - = _:1._.- ~ _ - - _.. - -l ---- - f gg1jlent fM Wg4 syo+ g+ p> #'O p g 'O e AS THE TEMPERATURE OF q P 4 4== LIQUID WATER INCREASES, ~g q Y p ~6 g--0-B THE VI,BRATIONAL MOTION p% ~k. AS WER AS WE MOVEMENT %'T Y'o.S y #-o-y~ d,# t 4 ,,7 OF MOLECULES WITHIN THE 4 D'g

1. s8 3.

LIQUID INCREASES. j H-0-H W e2) % o s Ndg3;. 4 :? g,,0 Y o 1 e THE AMOUNT OF HEAT REQUIRED l 4 H-0 H 9 8-TO INCREASE THE TEMPERATURE O 0F ONE POUND OF WATER 1,p,y h IS 1 BRITISH THERMAL UNIT (BTU [. lIQUIDNATER e AT A SPECIFIC TEMPERATURE IF MORE HEAT IS ADDED, THE (I LIQUID CHANGES PHASE INTO A VAPOR. e DURING THE PROCESS OF Y CHANGING FROM A LIQUID. TO A VAPOR, THE TEMPERATURE REMAINS CONSTANT BUT ENERGY IS BEING ADDED TO CHANGE 8g THE PHASE. H-0-H yo4 e THE AMOUNT OF HEAT REQUIRED TO CHANGE ONE POUND OF l M WATER AT'1000 PSIA INTO STEAM IS 650 btu, y ? l

== I I STEAM 2-9 = _ _ _ , -. ~.,

g -=== o.. . ~.. .a .procorties of Vater _ To understand properties of water, we need first to refer to temperature pressure variation. At low temperatures and _ pressure, water molecules are tightly bound together and form a solid substance. At real low pressures, if temperature is increased, phase transfer will take place and iwater molecules will evaporate into a vapor s, tate without going through liquid phase..This transfer is referred to as " sublimation". At higher pressures, water from solid phase transfers to liquid phase before trans-ferring to vapor phase. It takes one BTU of energy to raise 1 lb. of water 1 F. But, considerably larger quantity of energy is required to transfer 1 lb. of water P into 1 lb. of steam. For example at 1000 psig, it takes one BTU to change 1 lb. of ~. water from 544 F to 545 F. However, for the same 1 lb. of water to change from water to steam, 'it requires 650 BTU. l Thus, we can use 650 BTU of energy from fuel for every pound of water that boils in the reactor. This energy requirenent to change from liquid phase to vapor phase is referred to as Latent Heat of Yaporization an'd is given in the steam tables, Tables 1 and 2 under "h. " column for enthalpies. Note that when h. is added to the rg rg enthalpy of the liquid "h " at any pressure and tenperature, it becomes equal to f steam enthal py "h ". 2-10

,, Temperature pressure of water is given in Figure 2-3. The triple point is a point where vapor, liquid and solid phase exist at one specific temperature and pressure. ' Figure 2-4 shows the functional relationship between tenperature and enthalpy for The two solid lines exhibit satbrated water and saturated steam lines. The ~ c' wa ter. saturated steam lines partialfy curve to the inside. For one pound of water, at 300 F, energy equivalent to 270 BTU is required to reach saturation, where for the same pound of water to undergo phase transformation to steam at the same temperature, roughly 1200 BTU is required. It is noted that the I differential energy requirement for phase transformation decreases as temperature increases. At sufficiently high temperature where liquid is in equilibrium with steam, there is no differential energy requirement. S 4 \\ 2-11

.. _=.:. :.=== ===a - - - .~ 4' .. _ _ n__ 1 1 q p se,m ' s.a c.4 vi~ Ad I' G !p,J o- .y 5 g e 390I[ LIQUID t t 5 e" t M p m .c \\ p! m .n .A4 D I t,n th g t' 1 m h l w c:: N. P c_ l' SOLID VAPOR F,IdE f' 3 p f 0.CM f *' _l c t; i e \\f t b TEMPERATURE PRESSURE-TEMPERATURE DIAGRAM.:0R WATER F1suRE 2-3 o - p .= =

c.= z.-.,. L33... A v. a : ;.-. -. - - .--n. i \\ \\ t i ~ 550 1 i n) 500 tauI" Y * " ~. o'Y' 450 . SATURATED L ~ Sspe'ld @~ d ~ ER

• (

Vf i s y, 400 W*y

3uf, s

}{f'4 e-- ~ 6 SATURATED lG+ 350 ' STEAM ~ 300 250 l l 200 i i 100 300 500 700 900 1100 ENTHALPY, BTU /L33 IEMPERATURE VS, ENTHALPY FOR WATER FleuRE 2-4

h< ,1 p,.; .. d i:: c -:.: n.c.?.

u:...

FluIn FLOW Ano PRESSURE DROP For Question 2 FORAGIVENFLOWkREAANDAGIVENFLUID,THEPRESSUREDROPACROSS A LENGTH OF CHANNEL IS PROPORTIONAL TO THE SQUARE OF THE FLOW THROUGH THE LENGTH. M* r%ima 6 $p, f 2A pg e 2 AP = PRESSURE DROP ACROSS LENGTH,1 b / f t f g f = PROPORTIONALITY FACTOR ACCOUNTING FOR RESISTANCE TO FLOW W = MASS FLOW RATE, 1b,/sec 2 A = AREA 0F FLOW, ft { 3 = DENSITY OF THE FLUID, 1 b,/ f t 1b,ft 1b TO 1b = 32.2 9 = CONVERSION FACTOR TO CONVERT m f 2 c lb sec f l THIS EQUATION IS USED FOR DETERMINING PRESSURE DROP ASSOCIAT A SPECIFIC FLOW, IO DETERMINE FLOW FROM A GIVEN PRESSURE DROP, REARRANGE THE EQUATION. 2 f AP2A,9 3 c t W = ) 9I L a

O.. .:....:. w.~ - .-..c z.

a..w:;^:D.:,:.2.. e.

D.d 1 1 ~ FACTORS AFFECTING RESISTANCE TO FLOW 1 1 s FRICTION - RESISTANCE TO FLOW CAUSED BY SURFACE FRICTION 1 .y s ROUGHNESS - A SMOOTH PIPE HAS MUCH LESS FRICTIONAL RESIS-ANCE TO FLOW THAN A ROUGH PIPE. s VELOCITY - AS THE VELOCITY INCREASES, THE FRICTIONAL RESISTANCE TO FLOW ALSO INCREASES. w.. :. ~ c z~ 8: ACCELERATION - THE MOMENTUM OF THE FLUID RESISTS CHANGES IN VELOCITY SO ANY TIME THE FLUID IS ACCELERATED, RESISTANCE TO FLOW INCREASES. ~ s CHANGES IN AREA - FLUID WHICH FLOWS THROUGH A SMALLER DIAMETER FLOW AREA MUST BE ACCELERATED. f ~ s CHANGES IN VOLUME - AS THE FLUID BOILS, THE STEAM FORCES THE' ' FLUID TO CHANGE VELOCITY. 8 DENSITY DIFFERENCE - IF A FLUID IS HEATED, GRAVITY WILL TEND TO MAKE WARMER FLUID RISE AND BE DISPUCED BY COOLER FLUID. s SINGLE PHASE s 2 PHASE o LOCAL RESTRICTIONS - WHEN A FLUID ENCOUNTERS A RESTRICTION l IN THE FLOWING AREA, RESISTANCE TO FLOW INCREASES BECAUSE .THE FLUID MUST CHANGE ITS DIRECTION OF FLOW TO CIRCUMVENT THE RESTRICTION. e

.. a c.c. _ _ _.t iP ! P d '

• W W 8 ""

- ",. ~ v" -i

1. _ _.

FLOW THROUGH THE FUEL CHANNEL SINGLE PHASE FLOW 0 FLOW THRouGH THE CHANNEL IS GOVERNED BY THE FOLLOWING EQUATION: 1 fLP2Agg f E e I J T ~~ ~L.. __. ( 8 f 15 THE FACTOR ACCOUNTING FOR RESISTANCE TO FLOW 8 ADJUSTED FOR CRUD BUILDUP 0 MODIFIED FOR ACCELERATION THROUGH THE UPPER AND LOWER - (': TIE PLATES 8 MODIFIED FOR'THE GRID SPACERS AND THE INLET ORIFICE IWO PHASE FLOW i 8 IHE BOILING PROCESS CREATES ADDITIONAL RESISTANCE TO FLOW, o THE RESISTANCE TERM, MhE=aOFdtyi'NJUteswen as 2 IS A VERY STRONG FUNCTION OF THE QUALITY OR VOID FRACTION. e TWO PHASE FLOW CAN BE MODELED BY AN EQUATION SIMILAR TO SINGLE PHASE FLOW cPN k' 7' 2 IAP2A 9 y= i I \\is 8 THE UNDESIRABLE EFFECT OF TWO PHASE FLOW FRICTIONAL LOSSES IS MINIMIZED BY THE INLET ORIFICES. 1

.[ ~. i.... - 1. _.m w.. ~ ~ ~ ' ' ~ ~ ~ ~ ~ ~ ~ TWO-PHASE FRICTION FACTOR I Figure 4-1 is a typical experimentally derived tw-phase friction factor 2 plot. Note that at a quality of 0, the factor 4 is 1. This would simply - indicate single phase flow. Also note that the factor goes down ~at very 'high quali ties. For typical BWR operations, this friction factor might have values of 5 to 10. ~ Care Orificing P __ At zero power, the velocity of water entering any fuel assembly is higher 2% ' than the velocity leaving, assuming that there is a small amount of leakage from the channel. Since the core is a large series of parallel flow paths throOgh the fuel assemblies and each path begins and ends in a constant pressure plenum chamber, all assemblies will have identical pressure drops at all times; the (~. flow through all asse=blies, without any difference in inlet orificing, wuld l be the same, since the power is zero. TP flow is single phase. Suppose that l the power could be slowly increased in one and only one fuel assembly. What would happen to the flow in that assembly? With minimum recirculation system flow, as power increases, the flow in the l assembly increases. This is due to the natural circulation effect. The hotter vater in the channel is less dense than the water in the downcomer area and gravity will force the cold water down, forcing the bot water up. As boiling begins, the steam bubbles will exert a buoyant force on the water above them. This also will cause the flow to increase. The volume change as the water is converted to steam will begin to cause the outlet velocity from the channel to increase atove that of the inlet. As power increases further and more of the water is converted to steam, the outlet velocity continues to rise and channel friction, being a function of ~

e-( ,,M W m,pp- = 8'w e& O* >1' --,.....'N -g ag g hW b h 9 ~ < ~ r.e Nw MW@ $ -M a - W o =..... - 0 0 2 0 e u u O W ed S O A 5 O I I I I i I a l 1 f e G ~ = ", w.. ~ '~' % T, ~~* m = r - e . - mm r DO %Q u g m* O 4 i 5 e 4

• tB e

e X o a g Z D m O r* 5 f" se N O r.=. o c C Z O c* ~ o E l 1 8 8 O

,.....y ci c;1 ty,.M ll. increas e. ine mcrcased frict. ion results in an inm:. sed pressure drop 7i>utTi't canTot' result in an increased pressure drop b;cause the inlet'and ~ .m outlet pressures are being controlled by all of the other zero-power assemblies in the core. So what happens i's that the flow in the channel will decrease as power increases. Obviously, it is' not good to have the lowest flow in the highest power fuel, but that happens. More importantly, is dhe flow cutback caused by the power in-crease too high? The answer is yes. What can be done to improve this? Provide inlet orificing. _The inlet orifices work much as a biasing resistor in an electrical circuit ~~ 7-1 ~ works.5FiguF4-2(a) depicts a simple circuit in which a 1-volt battery delivers a 1-ampere current through a variable resistor set at 1 ohm. In Figure 4-2(b)-, when the resistance is increased to 2 chms, the current drops to h amp. This represents a 50% reduction in flow. In Figure 4-2(c) a 9-ohm biasing resistor ~ has been added and a larger battery so that with the variable resistor set at [ 1 ohm, the current flow is again 1 ampere. When the variable resistor is'increa_ sed to 2 ohms [ Figure 4-2(d)], the current drops to only 0.91 amp. In this circuit, for the same change in resistance, the flow or current reduction is only 9%. Orifices are added to the cere to reduce the magnitude of the flow cutback in the fuel channels as assembly power is increased. Two things are immediately apparent: First, the orifice will increase the pumping power required for a given flow; second, it should be possible to use smaller or tighter orifices on low-power bundles such as those with low enrichment or those found on the periphery of the core. Both of these are incorporated in the BWR design. The effect of orifice sizing is shown in Figure 4-3. Note that the tighter the orificing, the flatter the lines as assembly power as indicated by an increas-ing radial power factor is increased. Without orificing, the slope would be almost vertical.

IW~Ch'..:* 2 L's ;, - ?..i 4 D - ~ ~ n -c = a. _ -..... t '~ \\ 1 ohm i c 1 Amo T ' Vd* 2 ohms 1 I 1 voc ~ 4.~ g. 1/2 Amo c;.:._ - - w. _... r;- _r o 1 l 9 ohms 1 ohm J .W / ? T 1 Amp ~ Ic) 10 Vdc \\ T I l 9 ohrrs 2ohme .W (d) -- 10 Vdc T INLET ORIFICE ELECTRICAL ANALOGY i FIaunE 11 - 2

&-u ' L. :.- T =r..";%a',1. ::

9. 3 ~ ~ _ y i l,v.;

.,

~

• &"*'a

^ 8 ^ ' = 2..- = -. 2.-.... 3 4 LOOSELY ORIFICED FUEL ASSEMBLY FLOW (10 lb/h) [~ S R R 2 ~ / / / / / / / / / / / / / / / / / / / / zz /, // / / / i = L..=_ _ f l, / 3 -n 3 a l/ / /. )' n. -- ~ _ - 1- / / /; / / o :s >= E U

• E e

3 %5 gw C3 i _h - K r I j t j o a% ~ * ^ / / / / / / / E! "E v v ~v v v v l ) E~ E E E Z E I 7 = I I, ,j (*A) V91 01 i u. 9 l r r Gy [ ,, = 3 t 'y i l ~ No rr k r ~~ f i f44 j r m y w w a b L T ) i 2 = % e z E l I y l N

• W W t

y_ x x ( w I i .e -gte i 8 t r i !e q q3i LJd1LdJJW - 2 E i "o @o Zn S

• 28g I

= (8A) V91 01 u= 9 i f I d E g f i (V91 ol) A013 A7stt355Y 13fid g 033rdihO A11Holl e n---.-

C L= =b,:k.-p > =. i.i ~ '. 'U.. ~~u .' ~ c=- w -- - _ FLOW THROUGH VESSEL PIPING FLOW THRouGH PIPES e FLUID FLOW THROUGH VESSEL PIPING IS GOVERNED BY THE FOLLOWING EQUATION: )$ E I hP2A g p g w. ( f ) . s _.. 7 = _ =. - - =- ._~ IF A PIPE BREAK WERE TO OCCUR AT HIGH PRESSURE - ~ < ~J e LARGE. INCREASE IN AP 9 8 LARGE DECREASE IN f 8 THEREFORE THE FLOW THROUGH THE PIPE WOULD INCREASE DRAMATICALLY s THERE IS A LIMIT ON THE MAXIMUM FLOW THROUGH A PIPE, CALLED CHOKE FLOW WHICH IS EQUAL TO THE SPEED OF SOUND. AS THE FLOW VELOCITY INCREASES IN A PIPE, THE PRESSURE IN THAT o (CONSERVATION OF ENERGY). PIPE DECREASES 3 IF THE PRESSURE DECRFASES BELOW THE SATURATION PRESSURE, THE FLUID WILL BOIL. [F THE PRESSURE THEN INCREASES ABOVE THE SATURATION PRESSURE, f 4 l THE BUBBLES WILL COLLAPSE, FORMING SHOCK WAVES WHICH STRIKE THE SIDES OF THE PIPE. IHIS PROCESS OF FORMATION AND COLLAPSING OF THE BUBBLES 4 l IS CAVITATION AND COULD CAUSE DAMAGE TO THE WALLS OF THE PIPE.

.-..r. . g u 3..:. = = :. y 4....

. 4. -..

.r~ . -.: =...... -. f_ l .. ~ - - - GENERAL ENERGY EQUATION FOR INCOMPRESSIBLE FLOW l m \\[ ) h1 SYSTEM 2 l . THE TOTAL ENERGY CONTENT OF THE FLUID ENTERING THE _ m,. =:-j ~~ SYSTEM-AT_ POINT 1 PLUS ANY ENERGY CHANGE OF.THE FLUID CAUSED.. - ' ~5Y ' HEAT TRANSF.ER OR WORK DONE BY, 0R_0N THE FLUID WHILE' IN..- ~ j ; l THE SYSTEM WILL EQUAL THE TOTAL ENERGY CONTENT OF THE FLUID.. '- LEAVING THE SYSTEM AT POINT 2. THETOTALENERGYCONTENTdk[ THE FLUID IS IN THE FORMS OF POTENTIAL, KINETIC, AND FLOW ENERGIES. (') e FIGURE 4-4 l - l t i l 6

..... w:;- =;i,.. +.. ~ ;,, ;. {a --. =~ " %.* * ~=~**? - L' T ~~d'd, 32 ~~-1*, [ + I e k t FORMS OF IOTAL ENERGY IN A SYSTEM i r. THE TOTAL ENERGY OF A FLUID PER LB IS THE 3 SUM OF THREE TERMS: l POTENTIAL ENERGY = zg/g g -.. -,=

1eyation,

. " ' ~. z=e ft e _ -- . e -g -. l 2 g = acceleration due to gravity = 32.2 ft/sec 2 g, = 32.2 l b, f t/l b s ec f 2 lO . KINETIC ENERGY = V /2g c V = velocity, ft/sec ~ Flow ENERGY = P/p 2 l P = absolute pressure, Ib /ft f i 3 p= fl uio dens i ty, 1b,/ft 't I l I I i

.g.wa.c - : w := n.~..~.... ~. -.. ~ - ~ p-s:, - o __. - : 2 _q g g.. i, ; i e C THUS; WE CAN SAY THAT FOR ANY Sf NGLE POINT ALONG A STREAM OF INCOMPRESSIBLE FLUID k+ + (E. ' CONSTANT 'I = ) 4 9 29 IF THE FLOW IS ASS'UMED TO BE HORIZONTAL, THE FIRST TERM CAN BE DELETED. THUS ~....

- -g g~+g

~

r.

u-CONSTANT = l 2 P y THuS, ALONG ANY HORIZONTAC. STREAM OF IMCOMPRESSIBLE FLUID, AS THE STREAM VELOCITY GOES UP, THE PRESSURE DOES DOWN PROPORTIONAL TO THE VELOCITY SQUARED. e l

• -M.M M-P O

EN

.-s -x.:.-.;= aw... ; - x :.. ~: ~.... -4^ y+, c-- 3, : - .?.._-*' '*.. _ s:}; _[_ ,7 ~ ~ ~ ~ ~ ~ ~ ~ ~ CAVITATION IN A PIPE CONTRACTION .(- ~ O O O 0 FLOW @ o 0 -D o o O O O /

.y --

~3- -- c =r - _j_ &2 3 il FIsuRE 11-5 h CONTRACTION OF PIPE 'CAUSES FLOW VELOCITY INCREASE AND PRESSURE t REDUCTION. BUBBLES FORM AS THE FLUID PRESSURE FALLS BELOW SATURATION PRESSURE. EXPANSION OF PIPE CAUSES FLOW VELOCITY DECREASE AND PRESSURE INCREASE. BUBBLES COLLAPSE. @ SHOCK WAVES PRODUCED FROM THE COLLAPSING BUBBL PIPE WALL POSSIBLY CAUSING DAMAGE. P(.

,QY ~ T.'jjg.Tji.C:.T.'i;::T "' ' l 'V ~' ~ FCOW THROUGH PUMPS _ =- (.. e THE PURPOSE OF A PUMP IS TO TRANSFER FLUID FROM ONE POINT IN A SYSTEM TO ANOTHER BY,, INCREASING THE PRESSURE OF THE FLUID, AS IT PASSES THROUGH THE PUMP. r e IN A CENTRIFUGAL PUMP, THE ROTATING IMPELLER DRAWS WATER INTO THE EYE OF THE IMPELLER AND INCREASES ITS PRESSURE BY CENTRIFU-GAL FORCE AS THE WATER IS THROWN TO THE PERIPHERY. ~ 0 IHIS CREATES A VERY LOW PRESSURE AT THE EYE OF THE PUMP. . - _-... - - 3 =_ 0 IF THE PRESSURE FALLS BELOW SATURATION PRESSURE) ~ ~ ~ CAVITATION WILL OCCUR. ~ e NET POSITIVE SUCTION HEAD (NPSH) IS A MEASURE OF HOW CLOSE THE FLUID AT THE EYE OF THE PUMP IS TO SATURATION ~ c " "' " S - ~ O s NPSH IS THE DIFFERENCE IN PRESSURE BETWEEN THE STATIC PRESSURE AT THE EYE OF THE PUMP AND SATURATION PRESSURE. l l e NPSH IS GIVEN IN THE EQUIVALENT HEIGHT OF A COLUMN OF 0 WATER AT 68 F AND 1 ATM. (P g-P ) 9 I# 9 NPSH = s c ABSOLUTE STATIC PRESSURE AT THE EYE OF THE P = g 2 IMPELLER, lb /ft f 2 SATURATION PRESSURE OF THE FLUID, lb /ft .P = f s 2 9_.E = CONVERSION FROM 1b /f t TO ft OF H O f 2

1. 9 3

DENSITY OF THE FLUID, l b,/ f,t (AT E8 F)

1. =

..r. -

.. :.... ~.. z.. -.v +. - - - ~ .r u- ... = NPSH = (P -P,) g /pg 9 g l ~ ~ ' G L ) 1 { 1 n. A-W f " ' ' ~ ~ ~ * ~ 3 - - = 3- - = - - I j %.{~ l D f-t W I 4 .~ c, O FIGURE 11 - 6 (. j [. FACTORS AFFECTING AVAILABLE NPSH OF THE RECIRC PUMP e PHhsICAL HEIGHT OF WATER ABOVE THE PUMP e DENSITY OF THE WATER PaEssuRE5FTHEWATER e 8 IEM?ERATURE OF THE WATER EACTOR APFECTING REQUIRED NPSH OF THE RECIRC PUMP e SPEED OF THE PUMP

• e

-..T.?_._

p., Rea$ tor at full power "i, .l Q! ( Given: Pressure at Dome: 1000 psia RECIRc PUMP t1PSil Eye of Recirq Pump 60 ft. belcw 11 0 level 2 Subcooling: 20 BTU /lb . t. .q l!! ,e r PI-P P P g 1.f. 5 (p = 1)

• NPSil* =.

= p I l 1.' 4_Pdome' 2 ../ / / P II 0 4 ooME PilESSURE p p p ' } ' 'g 1000 psia g lp ; i" 1000 psia'144 fq f 60 ft 4E3 lb/ft3 62.4 lb/ft 62.*4 lhi/f t 3 t,' i = 2308 ft + 44.5 ft [ s/ [ eo s: P S P, is the saturation of 20 BTU /lb subcpolod reactor water at Ob'Opsiah = S43 BTU /lb .fo f hM subcooled 20fb 543 523 fb thus P = 878 psia 3 sq in P-878 psia 144 [> - - - S. so ft = 2026 ft 3 P 62.4 lb/ft y P P NPSH '= -f 5 1 = 2308 ft + 44.5 ft - 2026 ft 3 = 327 f t S L g-Ficune1-7 i L 1

ENCLOSURE 5 e For Qu= tion 2 e i V CRITICAL POWER 1 ? The design and operation of a reactor must ensure that any accident initiated from an allowed operating condition will not result in a radio-activity release off-site. These allowed operating conditions are specified as temperatures, pressures, and power levels which musti not be exceeded. As ~ in heat transfer and thermal-hydraulics, the bases for a given Limiting Condition for Operation (LCO) are a bit more complicated for a BWR than for Taken separately, however, the limits are not difficult to learn. a PWR. A. Transition Boiling t In the discussion of the pool-boiling curve, it was remarked that at i the departure from nucicate boiling, a small increase in heat flux results ] in a large increase in the temperature required to drive the heat to the water. In a flowing system, when DNB is reached, patches of steam form on the cladding surface and detach, rewetting the cladding. As a result, the cladding surface temperature fluctuates as these steam patches form 1 and detach. Increasing values of heat flux cause larger temperature fluct-9 1 .i uations. 'Ihis condition of steam patch formation and cladding surface 1 temperature fluctu:ition is called transition boiling and is shown in Figure 14. i: . "g A m g d i e 5 {

s 4

s 2 n CRITICAL POWEP 1 am 1 l$ m J hs 8 i, e +w w c

=--

l Forced Nucle te Boiling Transition Boiling Film Boil-Convection 1"8 Heating w FUEL ROD SURFACE HEAT FLUX (q") N. Figure 14

N B. Critical Quality As water flows up the coolant channel, the quality increases from zero t'o'some exit quality. If the heat. flux is too high, the quality increasss rapidly and a boiling tran'sition takes place at some point in tire channel.' ' The quality at the point of transition boiling is called the critical quality. Critical quality is a function of critical boiling length. ne critical boiling length is defined es the~ distance from the onset of ~ bulk boiling to the point of transition -boiling. Figure 15 shows critical boiling length, quality, and the point of bLiling transition for a typical ] BWR channel. At very high values of heat flux, transition boiling will occur a short distance from the point where bulk boiling begins. In this case, the Representation of boiling length, L ' p B (outlet) '.,*e A / 3 /.' / / -/ 'x a / / q"# 'kBoilingTransition .y / / / ~ ~ c / m X a / c. ,/ / L / / ,7 B C / / / 's / x (z) / / l } .,i.* / / / / s / / l L /r j l 1 M / / ^ h': /} / 4 T /i / A \\ j,- / / q"(Z) g / / / / ~. p, /_ / n -. n (# Sirigle Phase Ouality 11 eat flux (Inist) Profile Profile i l< Figures 15 v 1 A-

m quality at the point of transition boiling is low. At lower values of heat 51ta, which are still high enough to cause a boiling transition, the boiling length is greater and the critical quality is higher. A sketch of critical quality versus critical boiling length is shown in Figure 16. A major advantage of plotting critical quality versus boiling length is that this curve is independent of the axial heat flux profile. Although all the experi-1 nental data on transition boiling plots very nicely on a graph of critical 3 quality versus boiling length, neither of these parameters are measurable in an operating reactor. A correlation.had to be made between the critical ) J quality and the power required to produce critical quality in.a fuel bundle. m ..J 0.7-- l' O.6.. Critical quality X T 0.5-- T j

0. 4 -

Y '/ 4

0. 3-

~

0. 2.

I 1 0.1 0 3 J 0 20 46 60 80 100 120 Critical Boiling Length, inches Figure 16 33

i 6 C. The GEXL Correlation' Genera *. Electric found an empirical correlation between critical quality and the bundle power required to produce a boiling transition in the bundle. nis correlation is the GEXL correlation (G,eneral E_lectric Critical Quality, x, versus Boiling Length, L). GEXL is used to perform both steady state and ~ transient analyses of the thermal behavior in a BWR. GEXL correlates the critical quality,x, to measurable quantities such as boiling 16ngth, coolant mass flow rate, reactor pressure, fuel element heated length, effective flow area, and local peaking factor. GEXL can accurately predict the critical quality as a function cof each of these variables, while holding the other .3 constant. The heat flux and bundle power required to produce critical quality can be obtained by performing a bundle heat balance. The bundle power required to produce critical quality (and, therefore, transition boiling) can then be found for any reactor condition. D. Critical Power The bundle power required to produce transition boiling in a channel is the critical power. On a graph of local quality versus boiling length, performing a stepwise heat balance of a channel will result in a curve of x ..J versus L as in the curve on Figure 17, labeled " operating power (heat B balance)". Using the same channel flow conditions, GEXL will produce a

.,3 curve on this graph which is the critical quality versus boiling length.

l On Figure 17, the center curve is the " critical power (heat balance)" curve j and it becomes tangent to (just toucl.es) the "GEXL correlation" curve near the channel exit. Tnis implies that for this condition, transition boiling v occurs near the top of the fuel bundle. For other flow rates or axial power l 3 ...x shapes, -the critical bundle power may be higher or lower, and transition 'T boiling may take place at a different boiling length, ne important thing to remember is that any bundle power curve which touches the GEXL curve represents the critical power for that condition. NO i 34

..J GEXL CORRELATION y / /..'",.- / o / = / / / 1 quality / j x / CRITICAL POWER j (HEAT BALANCE) 7 / / P = constant / = constant G ACTUAL AhSUB = constant f / e BUNDLE j / POWER /,/ (HEAT BALANCE) / / / s'/ w /. / SI { j GEXL correlation and BWR heat balance curves 1 Figure 17 J v The ratio of the critical po cr to some operating power, like the one in Figure 17, is the critical power ratio, CPR. When the operating power equals the criticali power, CPR = 1.0. For operating powers less than the ~ ~ critical power, CPR*is greater than 1.0. The CPR must always be greater than 1.0. The minimum value that the CPR can have anywhere in the core (MCPR) is specified as a LCO in the Tech Specs. The OD-6 edit on the process computer gives MCPR data for the most limiting bundle in the core (Option 2) l l; and lists the twelve most limiting bundles for each fuel type (Option 3). The LCO for MCPR is based on maintaining the MCPR greater than 1.07 for any analyzed operating transient. Remember that a MCPR greater than 1.0 ensures that transition boiling does not take place, and transition boiling causes rapid thermal cycling of the cladding and may lead to film boiling, which results in unacceptable cladding and fuel temperatures. 1-3 3s

E. Summary The bundle power required to produce a boiling transition can be accurately predicted by the GEXL correlation. This critical bundle power is the ba' sis for the MCPR limit found in the Technical Specifications. Keeping the smallest value of CPR above the required Technical Specificanon limit ensures no transition boiling and, therefore, no fuel damage any j l. analyzed operating transient. i. g e l i e ) i -9r g .4\\ .1 .J ] 3 -,n----.__,.

VI LINEAR llEAT GENERATION RATE 3 The critical power.. ratio is used to prevent overheating of the fuel cladding due to deterioration of the heat transfer mechanism. To accomplish this, bundle power is never allowed to reach the value which would produce a boiling transition. Maintaining bundle power,below the critical power does not always protect the fuel bundle from damage. To further protect the cladding from damage, other thermal limits, with different bases must be specified. A. Linear !! cat Generation Rate e The linear heat generation rate (IliGR) is the number of kilowatts of i heat producsd per foot of fuel red length, k11/ft. The average IJiGR (AIJiGR) for an entire core can be computed, simply by dividing the core output by the total fuel rod length. AlliGR = Linear Ft. of Fuel Rod I For a BWR rated at 3293 !& with 764 bundles of 7x7 fuel, the total fuel rod t length is: i 49 pins 12 ft 764 bundles x = 449 232 ft. bundle pin l 1 s a and the value of All!GR is: l 3 293 000kW kW 440 232 it f t" , 1 l h B. Peaking Factors .a The average IJiGR is the power that would be produced in every foot of ! T _ j fuel rod length i f all fuel rods produced the sane fraction of total power. The actual IJfGR varies with radial position, axial position, and position 1 f l m 37

l}.-.; of 'a particular rod Dith!7a bYindle'."'In addition.. manufacturing tolerances can cause one rod to generate more heat than other rods in the same position. Peaking factors account fc,r the non-unifomity of DiGR within the core and permit ~ ~ ~ i the deteminatioEo'f'the r5ximumTil'GR'In 'th'e' core. ' Tfie (uel rod with the s highest value of UlGR would be the hottest rod in the core. i- , The axial peaking factor, APF, accounts for,the non-unifomity of the , axial flux profile, The axial flux profile is detemined for a given fuel. 5 I cell (4 assemblies) by a TIP trace. 'Ihe process computer uses TIP data to compute nodal powers for a given fuel assembly and obtain axial data for that fuel assembly. From the nodal powers for an assembly, the bundle power is found by summing the individual nodal powers. The axial peaking factor for a node is the nodal power divided by the average nodal power for the ,] assembly. For example, if a node generates 0.30 }.M, nd the entire assembly t is found to generate 4.73 FM, the average nodal power for the assembly is: t 4.73 FM 0.197 FM Average Nodal Power = 24 nodes node The axial peaking factor for the node under consideration is: Nodal Power 0.30 FM = 1.522 Axial Peaking Factor = Ave. Nodal Power 0.197 W = -i ihe axial peaking factors for every node in the core are stored for later use.

-9 The radial pe'aking factor, RPF, also called the relative peaking factor, is computed for each fuel bundle, rather than every node.

'Ihe bundle power, ,j PBUN is computed for each fuel bundle using the TIP-based nodal powers. The average bundle power, PBUNAV is simply the core heat output divided by the number of bundles. Using the reference reactor, the average bundle power is: PBUNAV = 3 93

4. 31 Et t

= 764 bundles bundle y The radial peaking factor for this assembly is: ppp, Bunde power of Interest, 4 @ = 1.097 ~ PBUNAV 4.31 i =

I n e radial peaking tor determines the highest power fuel use:bly =J = 2 s ... r:,, MParn r wave.,,......., ~-* in the core, and the axial peaking factor determines t$e....,,,.... %. highest power t node in any given assembly. W e local peaking factor, LPF, determines the hTgh'e'sf pode'r' Yod 'ad'IS%o'd7.difiMTofafpiai2Tg ia'cEiEA#$nElion'of fuel f

type, exposure, void fractionNonho1 Trod configuration and surrounding fuel -

~ ~ J types. He local peaking factor can be found from graphs or read from an. 7:2 '- OD-6 edit. Assume for the nod 6 in the current example that the LPF is 1.118. 1 i J The total peaking factor, TPF, is the product of the-individual peaking factors: TPF = APF x RPF x LPF ni-The hottest node in the core is the node with the highest value of;TPF:_. n s s TPF = (APF x RPF x LPF)m' ax max It should be noted that TPF is not equal to' APF - x RPF-x LPF nis max max max max. J the highest peaking factors all occur at the same node, which implies that 7 is not true. To find the maximum value of UlGR in the core, just multiply ] the average LHGR by TPF Max UlGR = Ave. UiGR x TPF max For the current example, the UlGR at the node of interest is the node TPF ] times the average UlGR: 3 Node UfGR = Node TPF x Ave UlGP.; kW kW l Node UlGR = 1.522 x 1.097 x 1.118 x 7.33 g = 13.68 g j For comparison, loqk at the UIGR for this same node if the fuel assembly is l 8x8R. He value" for PBUNAV is still 4.31 t but the number of linear feet of fuel rod per bundle is: bundle fuel length = 62 pins x 12 ft = 775 ft bundle pin bundle ,j Ren the average UiGR for this core with 8 x 8 R fuel is: 4.3. W / bundle .5.5kW PBUNAV J t ^* I

• fuel rod length " 775 ft/ bundle ft For a given core size and total heat output, the UfGR for 8 x 8 R fuel I

is about 20*, lower _ than for 7 x 7 fuel. L

• ~"

'T + ....,,. w... m ~ 4 is u.s ; .g.t o u.t., ^4 'iRich energy %ybduEsd'tn a j Linear h'est gene. n .-.-,.u . Th'e 'les' on on heat.~tran'st,er'.sY,w.e.d. that the particulagspot,.in a,fu.=el' rod.',a s ow sw n. l-2 -y_ mw p.

• N'Efx 3NN M*,SS."

f rJimi.t3R.h ~ ] ~ rg,willbetoo '2151 5 ! RM _ 2 I ' high if the UlGR limit is exceeded. 'Ihe maximum UiGR limit is imposed to - prevent ' cladding failure and Icakage of fission products into the coolant. At low power, the UO fu 1 pell ts look like those in Figure 18 (a). 2 As reactor power and UlGR are increassd, the pellets begin to distort as in Figure 18 (b). If the pellet is overpowered, that is the UfGR limit is exceeded, the edges of.the pellet will push on the cladding hard enough to I stretch the cladding as in Figure 18 (c). The UlGR limit is based on not i

• 3 exceeding 1% plastic strain on the cladding.

n At PBAPS the LCO for DiGR is: _ UlGR 5 LGliR -( max ( L's' d where, UiGR = Design UIGR d = 18. 5 kW/ f t for 7 x 7 fuel 13.4 kW/ft for 8 x 8', 8 x 8 R, and 8 x 8 LTA fuel (AP/P) = Maximum power spiking penalty = 0.026 for 7 x 7 fuel v t = 0.022 for 8 x 8, 8 x 8 R, and 8 x 8 LTA fucI LT = Total core length ) = 12 ft for 7 x 7 and 8 x 8 fuel 3 = 12.5 ft for 8 x 8 R and 8 x 8 LTA fuel l L = Axia) position above bottom of core For a 7 x 7 fuel bundle, the UiGR limit at node 4 (2 ft above core bottom) J would be: Il 1, 18.5 kW 18.42 kW l I-(0.026)( } )] = 18.5 10.9957} = f f f For a 8 x 8 and 8 x 8 R fuel, the spiking penalty is included in LGHR ' d kW so that the LifGR limit is always 13.4 n' l l1 l

.. J'tX ~ M dF 7 g ' D " N 5 # M M

1. Mi. QWha h r,.

• * +

% /;dhhh) th 4r ^ ~~ mb:r: s 7-3 ~~ a @$til}"yg${-(($ $g[ [gg = ~{3 _gbr u... f$ w p_y 7 g e.-. 1 3. N pellet cladding , k deformation due ( ) \\ to pellet expansion \\ clad i / (3 t k l \\ a

• (

b f ,^. ';t e _) k k Y \\. f d 3 (a) (b) (c) u ~ Figure 18 m C. Sumcary l The LCO for LilGR insures that the plastic strain limit on the fuel l l. cladding will not be exceeded. Notice that the basis for the UfGR limit' l is entirely different than the basis for the MCPR limit. The MCPR limit guards against deterioration of the heat transfer mechansim - nucleate i bulk boiling. The UlGR limit guards against cladding damage due to exces-- 3 sive power output in any node of the core. 41 1 l

j

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e.: m e.Le n. w., <.,,. w i etr.sy:';.;4n,;Q.wm-hanifisstont:.~ sp w + m 4 rate r-h n ~ catCremovAE '.d er p*%-gI %s~ rpas g;g93;ggym;n ;;> maximum power of the reactor. There is no theoretical limit to the fission rate in a reactor. Instead, the ability of reactor materials to perform ~ in high temperature environments limits the amount of heat the reactor can be allowed to produce. The two previous lessons discussed the CPR limit, which guards against a boiling transition, and the IllGR limit, which prevents damage to'the fuel cladding. Operating the reactor at a power a which stays within both of these limits insures that no fuel or cladding m damage will occur if an analy:cd transient occurs from an allowed reactor operating condition. The last thermal limit to be discussed guards against T fuel and cladding danage during a design basis loss of coolant accident (LOCA). Jl ~* A. APLHGR The linear heat generation rate limits the power in a particular rod. Tne IJiGR limit basically protects the hottest rod in the core from cladding Si damage during operating transients. The hottest rod in the core could be in either of the following general locations: It could be in a low-power l bundle with a high local peaking factor; or it could be in a high-power bundle with a low local peaking factor. Tne case of a high-power bundle with significantly different problem. a high local peaking factor is not a

• l t

In the event.of a LOCA, the fuel would dry out rather quickly, and l the primary heat t ransfer mechanism prior to rewetting would be thermal a radiation. Title 10CFR50.46 specifies that the peak cladding temperature during a design basis LOCA shall not exceed 2200 F. If a fuel bundle (or node) dries out during a LOCA, the. edge rods could dissipate heat by radiation more easily than the central rods. Tne edge rods can radiate away from the fuel bundle, while the central rods radiate much of heat their heat to other central rods. Even though the edge rods, and the corner rods in particular. have higher local peaking factors, the central reds ~ are nore likely to exceed the 2200 F limit during a LOCA. E 42

l ~~ .' nm wa~ Wn .:*~F;J$j;;Me rd. n-i w-r ~, I M.*.. Wew,tm.od@Q / l "~Qy%'sG:~$%;3.41NG%gc($ cst {p:iefhtfefj ,7 J.*,P C k. -s pp the tcM p'd6 r ther f.i d i Mt erefdrb 4 ag %d? Y .Qgp aTfhca t a-rod his cy a - -w y ..,t igenera.tIo 9 N-Mid#.MNEe7. EiUpI[h"ad[(*Eis$tf$s?ch,n,r.a y.uh_TcN%i p;-i gYa 97.2 ta e; r;r e. 2, .w.c :. 2 v.. mw~ av z.x, 2 + -we q +.. b2CNT$V C./b-40p2M4+<.JN 12 nearjh e a t ; g en crYt i onira t e e ( APIJIDR) gM &#i0: tw ~ 9.s#p,m%" fig M.4MQMVI#E -1 e ta GRcac oss - ar r ~ "L.c eL y $?'The NI,t uPof~*AP{J1GR' y' "O ,,a u, s ;.;....

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's tist# di nod r i te eat e by the number of feet of fuel pin in the node. 'ihe AP!JiGR is calculated by the process computer and printed as the maximum average planar linear I heat generation rate (MAP 1J1GR) for the node, using the following formula: i P x (1-f ) MAPIJ!dR= n c g i where: P = nodal power, kW f = fraction of power deposited directly in coolant, decimal C N = number of fuel pins per bundle i L length of node, ft. = For a node generating 0. 3177 W in an 8 x 8 R fuel bundle, with 6', of the !y power deposited directly in the coolant, l,

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MAP!JiGR = 317.7 kW (1-0.06) = 9.6a, kW g7 g g, g g g TE %

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e, The limiting value of MAPulGR is found in the Technical Specifications and is a function of fuel type and exposure.

The ratio of the actual value of ( MAPUfGR to the I.C0 value of MAPU1GR is the maxicum average planar ratio [ (?.'AliAT). Tne val'be of MAPRAT cost always be less then 1.0. - (!tAPI.liGR) a c t ua l ~ HiPRAT = l.0 (:1\\P ljiGR) I CC - m max E [ l -~ Maintaining t he MAPRAT less than 1.0 ensures that the peak cladding temper-ature will not exceed 2200 F during a design basis LCCA. T -J ]}

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<,i*.e' 1 - e.k 1 ~ <- '^ ~. 'f.. k.j. n ..~v-v nll1l$g,m h..?}lll lll*.l,'.l00&.c .. e.%. e e reactor against a -different heat-trans fer reliited 6... + s.;- % ***'E.prc41 err.S ~ . s. e. ,2 ** 1 ~' (.. p . c .c e . ~ + .c .s.., :s. .e -n w... ><.. W.,tectsth.. O..: s. ~.p rop,5:j. , q..w ...s nuntains nucleat e bo;iling and prevents a boiling-t ransition? p.

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.y,::.5.c y.4q, 3. p 3g g y y y s w, lne cu n 1init The IJiGR -lici t minirizes cladding st rain caused by excessive heat generation in the fuel. 'Be !'/JIJIGR l imi t ent.ures that the peak fuel cladding terper-y 0 ature will not ciceed J.00 F during a design basis lOCA. Although these three limits are dif ferent in rany ways, they are all approached at about the same rate during norral reactor operation. At fu l l pw.er, all three of these thernal limits are about 80-90% of the LCO. J 9 d

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1 I ENCLOSURE 6, i4 e For Question 3 H_ EAT' SOURCES I. TWO HEAT SOURCES AFTER SCRAM A. SENSIBLE HEAT 1 1 l B. DECAY HEAT i GENERAL ELECTRIC COMPANY-PROPRIETARY INFORMATION

i .[ 'I 1 ? CORE COOLING MECHANISMS I. THREE HEAT TRANSFER MODES { l ] II. BWR COOLING MECHANISMS j A. FLOOD

=

_(' i :,. e B. SPRAY C. STENi i 1, a l ~ 1 1 D. BOIL l t I i l GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION l i

= - -. - - i l a o l 1 ll i j FUEL & CORE DAMAGE THRESHOLD I. TECH SPEC SAFETY LIMITS i -A. MCPR B. PRESSURE t C. RVWL L=. er,.a i II. FUEL DAMAGE THRESHOLD ( i I l GENERAL ELECTR!C COMPANY PROPRIETARY INFORMATION r--..,--,-a,-,-v,--,-,n. ,,--r-, .--~-,,.,na,.--,.-,.w,,.,,,w---,--,,e-. -.---..n...,,,-,-.~e,-.-,-nn--..--.~.,. - - - ~ ~, - - - ~ - - - ~.- --- - ~,, - - - -- + - -, +

i I i EECOGNITION OF INADEOUATE CORE COOLING CONDITIONS I Introduction A. Causes of Inadequate core cooling 1. LOCA - Degraded ECCS i 2. SORV - Degraded ECCS (. / 3 Channel blockage 1 l l l ( l B., Attempt to recognize adeauate cooling l GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION 7 ,-v, -n,-.

i j' II Divsrsa nothods of dstseting cdsquata core cooling A. Primary method!td assess adequate cooling - Reactor Water level 1. Natural circulation is an inherent feature i l 2. Adequate Water Level = adequate coolins; a) Normal level or above l [ b) 2/3 coverage c) Above c. ore plate GENERAL ELECTRIC COMPANY 1 PROPRIETARY INFORMATION t

4 e B. S:cond2ry asruranca of cdaquats core cooling 1. One ECC system operating f f III Water Level Instrumentation l A. Confirms adequate core cooling and detects an approach to inadequate core cooling. i 1. All mode of normal operation ~ i l f' 2. Anticipated transients I i l 3 c edible accident conditions -4 GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION ~ ~ 1

r l l t i l - B. Measures using differential pressure devices. f 1. Senses mass of flhid above the sensing tap. ) i ~ ) 1 e C. Ranges of Level Indication during accident conditions 1. Narrou range (' 2. Wide range 3 Fuel zone e e GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION 1 I t

i c. At highar than calibrated pracIure - rInding is isss than ~ actual. b. Natural circulation flow same as above. I c. LPCI flow in BWR's 3&l4 gives high reading 1) confirms injection flow [ 0 D. Water level inside shroud versus outside the shroud it 1 Higher than indicated a) except when ECCS is injecting I 1 GENERAL ELECTRIC COMPANY-PROPRIETARY INFORMATION ,a,. ,n,

.l h IV Indicstion' cf Optratiendl ECCS' i ,A. If ECCS functions \\ l 1. No core uncovery foi small breaks 2. Little or no clad heatup for large breaks B. ECCS Indications 1. Valve Positions a) flow path i b) most valves outside primary containnent GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION - l

i 2 Positive Flow ~ a) Pump (s) running i 'l i 3 Discharge Pressure a) Piping integrity wve l IV Indication of Superheated Steam i I l a) No superheat core covered i l l b) Difficult to do l GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION l i s.

~ 1) fast responding TC on SRV tail pipas 2) where to measure saturation temperature for vessel Succary f GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION

3 l t i \\ I 'I i g i a INCORE INSTRUMENTATION I 1. 4 BASIC SYSTEMS i l II. SRM RESPONSE TO ICC A. LOW LEVELS - UNCOVERED i B. LEVEL JUMP i 1 i III. LIMITATIONS l l l .A. 5 FT BELOW TAF B. FULL UNC0VERY T00 LOW C. DRIVE MOTORS NOT 1E GES" DAL ELECTRIC COMPANY PRu?iilETAR) INFORMATION

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D. DECREASING COUNT RATE l I i i \\ I IV. DETERMINING EXTENT OF CORE DAMAGE 4 l t } { A. TIP 110VEMENT I \\ 1 i i C e 4 i i i i i i I t \\ i i l 1 GENERAL P "C COMPANY i PHQfRIETAh4 i;,c0RMATION I s- -.w,- 4- ,,w.+-~,,,,,r.r-y.,,,-y - -,_,,v,,pys.,,.

w I ) 7. REACTOR WATER LEVEL INSTRUMENTATION f 1 INTRODUCTION Objectives: Simple DP cell (Fig.1) i Pressure compensation (Fig. 2) Variable & Reference leg theory, etc. 3 Design criteria: Wide range vs. accuracy (' Range coverage (Fig. 3) INSTRUMENT DESCRIPTION i YARWAY (Fig. 4) t Temperature compensation I i DP cell outside drywell Local & remote indication t l i Safety functions GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION y _ --

k I,' i y , ; _%,S. * ~ q t /[*) COLD REF LEG.(Fig. 5). No heating of RL (ambient) s r1 4 i DP RPV water level I - Accurate when X = Y = Z i 1 ft. i 1 4 'I Location of instrument taps (Fig. 6) 7 ( LEVEL READOUT & TRIP FUNCTIONS Review level assignments (Fig. 7) (,., ? INSTRUMENT RESPONSE / Drywell Temperature Effects due to ^ i Normal AT of Startup [, Error - f (DW to calib AT, Tap elevation, Type of RL) i I' I ~ COLD REF. LEG s(Fi ores 5&G) r 3 'i I.! X =.Y = Z i 1 ft, no ' error GENERAL ELECTRIC COMPANY l if x = y = Z, 2.2x (5") error e so F AT w l PROPRIETARY INFORMATION i s -,->n,-v ~,, < -,. -e--<--. .-,e--. --.m---

1 i _3 4 ) ) i ~ Effect of AT odtside DW < 3% error J l ~ YARWAY

)

f Error = 2.4% (<5") 0 60 ? AT dw } Trips occur at L(act)< L (in3) Effect of AT outside DW is negligible

SUMMARY

Error < 3% for normal AT in/out of DW ( Errors not of operational significance Drywell Temperature Effects during Upset & LOCA RX PRESSURIZED Error a AT dw, RL thermal time constant, & length of RL. Yarway & Cold RL where X / Z --> Lind > Last Measurement error in Fig. 9 GENERAL ELECTRIC COMPANY No error for Cold RL where X = Y = Z PROPRIETARY INFORMATION --.w-,,#-. a , -,, - ~.,. - -. =, - - -....... _.. r

i Lower Yarway tap can uncover (Fig.10) and prevent trips and mislead. o operator. REACTOR DEPRESSURIZED Cold RL Thermal Response (Figures 11-18) 3 Transient boiling Flashing { t I These require Tdw > Tsat in RPV P >200 psig assures no flashing or boiloff R l Worst flash-+20% error l l Flash + boiloff ----+- 30% error I hour f rom ADS l Refill RPV to steam lines restores accuracy l Yarway Thermal Response (Figures 19 & 20) l Flash + 15% error l Boiloff--+ 35% error I hour f rom ADS l l l GENERAL ELECTRIC COMPANY-PROPRIETARY INFORMATION l l ..--3,- ,y y,,,,e -7. -.~,.m m -w.-- ,3.,,-....%cw,,._-w,

i s f Faster boiloff than Cold RL i 1 No degrading of ECCS initiation due to errors Drywell Temperature Effects With Loss of Drywell Coolers Peak Tdw < Tsat of RPV No flashing or boiling occurs CONCLUSIONS When pressurized, high Tdw may cause error (indicated > actual) in Yarway & Cold RL instruments (except X = Y = Z Cold RL). When depressurized, high Tdw may cause error (indicated > actual) due to flashing and subsequent boiloff. This may cause a reduction in redundancy in safety system initiations and mislead the operator. Lower initiation levels of safety systems caused by high Tdw are inconsequen-tial from safety standpoint. PRESSURE & JET PUMP FLOW EFFECTS OFF CALIBRATION P> P calib. --> Lind < Lact by 80" max @ TAF Nat. Circ. -+ Lind < Lact by "several inches". These inaccuracies are conservative WRT operator response. GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION

1 LPCI flow + higher press in diffuser -+ Lind > Lact however, is indicaticn ~ of ECC flow ALTERNATE MEANS OF LEVEL DETERMINATION Superheated steam = core uncovery, but determination depends on flow thru S/RV's s 's limitations', N flux anomalously low for uncovered core. [fluxlowerbutnotsignificantly not useful. Refilling steam lines & watch T of S/RV tailpipes. Install (or use installed) pressure gauges above and below expected level and apply thumb rule. PRESSURE INSTRUMENTATION Several reliable sources of P indic Normal control room pr' essure indicators and controllers ~ Annunciators associated w/ press switches GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION I

t s I _7 (

• Steam supply press to RCIC/ ISO Cond'r Recire pump seal press ECCS pump discharge press s

1 TEMPERATURE INSTRUMENTATION Little value in determination of internal R.P.V. temperature, but temperature is readily available by P VS T in saturated conditions. O GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION t . =

e 4 6 RADIOCHEMICAL RESULTS FROM A DEGRADED CORE I-BWR Safety Features As Radioactivity Barriers A. Barriers 4 B. Characteristics of Nuclides That effect release t i i II Fission Product Source Terms A. De finition B. Sources during notmal operation GENERAL ELECTRIC COMP PROPRIETARY INFORMATIO ,--re,- - - - -,, ,a v-- n. ---,,-.,.-,.---.-,,,,,---.-,---,---,vvn,-e.-

l i l l 5 8 i C. Types of Sources during normal operation j, 1. Recoil i i f a l I i i l 2. Pin-Hole Defects 4 0 i 4 i i GENERAL ELECTRIC COMPANY l PROPRIETARY INFORMATION 1 ~,-s-- ,s---m--v mween----m,v,n.,---...ang-,--nnv.- nm ,,w-,,,- .--e.--, -n-.

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- - -,, - - - -, - + - - - n

i ) } l f D. Analyzing patterns GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION

l D. Distribution during s,bnormal conditions 1. Split Cladding Defects i 2. Larger Cladding Defects 3 Accident Conditions i e e GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION

t I ( ~ E. Fission product bshavice trsnds ~ )) ~ ~ t i / 1 I l i 1 -I l I f F. Noble Gases -l 4 f T t I i t i i l s J J 4 J GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION i i m-- ~ve.-- wr,,,-,wm-,.-,m-~r,<=,mu- * -, -, _. _ --e---,,,..----,---.~.-~..m-, . - -. ~ ~ ~ ~. - - - -.. ~ - -. - - - - - - - - - - - -

n'- 1 ..I { l f L l i 1 .i G. Beactor Water Solubles l t 1 GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION l l t

i i ) I ) l H. Reactor Water Insolubles ',3 ,e 1 { t (^ GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION

1 s

• III bdioactiva Sourcas und:r d:gradsd condition 2 A.

Not defined yet ,,7 t B. Hypothetical bases I i I C. NUREG-0578 [ s l D. TMI-2 / E. WASH-11100 Table 12 l GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION 7 l.-.--,-..

IV Transport and Deposition of Fission Products in BWR Containment System A. Release dependent on the accident sequence 1 B. Degraded core accident i C. Removal Mechanisms 1. Noble gases l I f I ( 2. Elemental Iodine f I GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION e

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't I \\ 3 Iodide ( ~ 4. Aerosol Particles l r { l l GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION 7-. ~

r g' *,- ( . V* Jiccicn Product Analysis A. Preparations are being made B. High Activity measurement facility I C. Gas Sample Analysis l l 4 GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION A

= ,, w : ..l t -}. -1 I I i i i t I i l l D. Liquid Sample Analysis l-E. In line conductivity e GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION 1 l.

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/ DECRADED CORE CORROSION EFFECTS EXTENDED IMMERSION OF PRIMARY WATER ON BWR STRUCTURAL ~ MATERIALS. PARTIALLY FLOODED CONTAINMENT FLOOR WITH Ti =300 F I. BACKCROUND j C II. GENERAL CORROSION 1 1 i i GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION

i 1 III. GALVANIC CORROSION I C GENERAL ELECTRIC CO PROPRIETARY INFORMA

i IV. CREVICE CORROSION i ( 1 ( l GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION 1

V. PITTING CORROSION I l l (. l GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION

~ ' VI. INTERGRANULAR ATTACK VII. STRESS CORROSION TRACKING VIII. CORROSION FATIGUE i IX. EROSION CORROSION X. IIYDROGEN DAMAGE 4 i 1 GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION

1 ~ ~ i h I DETERMINING DOSE RATES INSIDE CONTAINMENT FROM DETECTORS LOCATED OUTSIDE CONTAINMENT I. METHODS II. EXTERNAL MEASUREMENTS -UX A. I = B(px) I e g r B. INTERPRETATION PROBLEMS C. MITIGATION III. NUREG 0578 l GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION

l.. l j q GAS GENERATION j I. INTRODUCTION I i II. SOURCES A. H2 1. Zr-H O 2 2. STEEL-H O 2 i 3. RADIOLYSIS i B. 0 -RADI0 LYSIS 2 I GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION

b s 4 + j _ 2-i C. FISSION PRODUCT GASES i ? III. TYPES OF COMBUSTION

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i k i A. BURN f f .f l B. RECOMBINATION l \\ I C. DEFLAGRATION I GENERAL ELECTRIC COMPAN PROPRIETARY INFORMATION I --r .w. ..-cm ~.-....--.w,- ..-v--w~..--4 w-~~--,,-,-.w.. .-r,--,u- - ~., ..w----- y- --r,. - :-rm e e - w--i

4 t D. DETONATION l i i i E. AUT0-IGNITION IV. CHARACTERISTICS A. MIXTURE (Table I) f' i f B. COMPOSITION (Figure 1) C. VESSEL i GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION .-.---,...---,,-.,,e-,n ,,n n-.,,, n,,------ --,-,a-g.. .--n,,c-,,,,n-e.--, ,,-,-,c--

p... .c L 4_ ~~ D. PRESSURE i V. MITIGATION OF H COMBUSTION 2 A. H2 < 4 VOL. % (~ B. O,<5 VOL. % GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION

,7 4 i P

f

) C. HALON 1301 f I i + l D. REMOVE H2 f i 1 E. BURN H2 !(~ d F. IGNITE H2 l l G. VENT l l l I i j GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION i I I

l l . ~ 6-I i VI. MITIGATION OF 0 LIMITED COMBUSTION 2 A. RADIOLYSIS (INERT) B. RECOMBINE C. BURN GENERAL ELECTRIC COMPANY.,. PROPRIETARY INFORMATION

t.=,. a enfo D. VENT s E. MITIGATION IN INERT F. MITIGATION IN AIR G VII. MITIGATION OF T10BLE GAS A. VOLUME PRODUCED i B. DILUTE ATMOSPHERE GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION ~ Y

......s-a- n _t,s., III. FLUID MECHANICS ENCLOSURE 7 Page 1 of 2 / Designed to provide a funda= ental background in the principles of fluid flow and associated equipment needed for nuclear power plant operation and for pursuit of advanced or specialized training. FLUID STATICS including units of measurement, pressure and pressure relationships, pressure measuring devices, density, specific volume and specific gravity; FLU: n i' LOW including the continuity equation, the general energy equation, ideal (frictionless) fluid flow, laminar and turbulent flow, actual fluid flow, fluid friction; viscosity, system losse's in head and pressure terms, fluid ha==er, and flow measurement; PUMPS including types and applications, pump work, efficiency, cavitation, Net Positive Suction Head, and pump operating characteristics; TUR3INES including basic principles of operation, nozzles and air enectors; PIPING AND VALVES including types and applications, sizes and standard classification; REVIEW AND EXAMINATION Approxi= ate duration - one week June 22 to June 26 l l e f ~ . - - - - * = .= ~ ..e ~

Of y,; ~ ~ l ENCLOSURE 7 ~ ,..,a IV. HI.C AND THERMODYNICS 9 Page 2 of 2 Designed to provide a fundamental background in the principles ~of ~ heat transfer and thermodynamics needed for nuclear power plant opera-tion and for pursuit of advanced or specialiced training. THERMODYNAMICS including properties, te=perature, pressure, heat, fusion, evaporation, density, specific volume, internal energy, flow energy, First Law of Thermodynamics, General Energy Equation, enthalpy, work, system boundaries, Second Law cf Thermodynamics, entropy and ideal gas law; STEAM AND VAPOR PROPERTIES including steam tables, saturation te=pera-ture r.nd pressure, quality, specific heat, superheated steam, TS diagram and hollier diagran; THEPM0 DYNAMIC CYCLES including pressure-specific volume diagram, temp-erature-entropy diagram, nuclear boilers and steam generators, steam turbine, friction, efficiency, condenser, heat exchanger, pumps, cycle diagra=s, work, condenser heat r'emoval, reactor heat input, feedwater heating and cycle analysis; HEAT TRANSFER including conduction, convection, radiation, heat exchangers, evaporative cooling and calorimetrics; BOILING HEAT TRANSFER including convection, nucleate boilinE, DNB, film boiling, boiling water curve and two phase flow; THERMAL STRESS including application to thickwalled vessels; REVIEW AND EXAMINATION Approximate duration - two weeks. June 29 to July 3 July 6 to July 10 e l J -o.

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