ML20002E048

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GE Contribution to Big Rock Point Redundant Core Spray Sparger Design & Installation. Answers to AEC Questions Re Proposed Change 27 Encl
ML20002E048
Person / Time
Site: Big Rock Point File:Consumers Energy icon.png
Issue date: 02/02/1971
From:
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML20002E040 List:
References
NUDOCS 8101260187
Download: ML20002E048 (32)
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o 3-CONSUMERS POWER COMPANY Appendix 1 - To Request for Change to Technical Specification No 27 Docket 50-155 - License No DPR 6 General Electrie. vantribution to Big Rock Point Redundant Core Spray Sparger Design and Installation I.

Introduction - As a part of their backfit licensing work, the Consumers Power Company has contracted with General Electric Company to analyze for and design a redundant core spray sparger to improve the Big Rock Point reactor emergency 1 core cooling system. This report describes the results of this work. M though it is limited to the scope of the work requested, this report is written in such a way that it can be readily put together with other material needed to provide a document on the improved overall emergency core cooling system perform-ance. An additional part of the report is a discussion of the analysis performed to show that flow induced vibrations of the new nozzle will not be of concern. II. Conclusions The second core spray sparger will provide adequate emergency core cooling flow in redundance to the present core spray sparger as long as the proper operating characteristics are maintained in the system in order to provide at least I gpm to each fuel assembly. This can be done by using the present core spray pumps or the present fire system to provide the core spray water and orificing the supply so that the pressure drop across the nozzle is not allowed to exceed about 50 psi. An analysis of the vibration characteristics of the nozzles shows that damaging vibrations _ will not occur, and special testing will therefore not be required. 71flMoM f1

I. i III. Discunion A. Design Objective Because of high radiation it is not practical to modify the reactor inte rnals. Consequently, it has been decided to use a single spray nozzle mounted on a " bayonet" type pipe fixture through the top head of the vessel. The nozzle would be centered over the core and would spray down from about 7 feet above the top of the core. B. Design Flow Requirement to Achieve Adequate Cooling The amount of spray water required to adequately cool the fuel assemblics depends primarily on fuel bundle power. The minimum flow into each fuel assembly must be sufficient to cool the highest power assembly. For Big Rock Point the highest power assembly is about 4. 2 MWt.

  • The highest power assembly for General Electric 1967 product line BWR plants is about 6 MWt and the design minimum flow per assembly is 3. 26 gpm.

Thus, to be equivalent on a flow per MWt basis, the minimum required flow for Big Rock Point would be about 2 gpm. However, tests have been conducted for a 49 rod, 4 MWt bundle with spray rates between 2. 0 and O. 5 gpm. The data shows there is virtually no change in the maximum clad temperature transient for spray rates between 2.0 and 1.0 gpm and only a slightly higher temperature results for a sp. ay rate of 0. 5 gpm (2). It is expected that the same trends would result for the BRP fuel assemblies. Therefore, the minimum allowable flow per fuel assembly is specified as 1. O gpm. It should be noted that this flow is not a nomina: asign value, but rather a minimum limit which should be considered acceptable after removal of all uncertainties in the predictions of flow distribution from the core spray. Furthe r, it is expected that only a few bundles will receive this minimum flow. As noted in reference 3, the minimum flow necessary for cooling must be available with a reactor pressure of 85 psia. At lower reactor pressures the total core spray flow will increase, but the minimum required flow per assembly remains the same for any reactor pressure down to i 4. 7 psia.

  • Based on 2. 8 MWt average and 1. 20 and 1. 25 inter control rod and radial peaking factors from reference 1.

9 C. Evaluttion of Sslected Spray Nozzlo A Spraying System Company nozzle #4R80160 is proposed for the BRP Core Spray. Further, it is proposed that the nozzle be located with the ~ tip centered directly over the core at elevation 614'-71/4" (see reference 4). The follcwing is an evaluation of the proposed arrangement. The elevation of the nozzle is 77 inches above the top entrance to the fuel bundles. Figure I shows the estimated performance of the nozzle applicable for a full 84 inch elevation. To correct Figure I for the lower elevation Figures 2 through 5 were constructed. The data points shown for the zero gpm/ft lines at each nozzle pressure drop were obtained from reference 5. Data points for the other gpm/ft values at the 84 inch elevation were obtained from Figure 1. The lines of constant gpm/ft were estimated by considering the shape of the zero gpm/ft line s. Intersections of the constant gpm/ft lines with the 77 inch elevation linc define the spray distribu-tion at 77 inches shown in Figure 6. Because of inaccuracies associated with the previous estimation process and the very rapidly changing spray distribu-tion near the edge of the core, it is considered necessary to provide a tolerance on the estimated distribution in Figure 6. This is best accomplished by moving the spray distribution curves radially inward, since a simple percentage reduction in the gpm/ft at a given radius does not properly consider the effect of the rapidly changing gpm/ft near the outer edge of the core. That is, a small error in the radial position of the curves could result in significant changes to the predicted amount of flow entering the bundles at the outside edge of the core. For these reasons a radial data tole,rance of 10% has been applied. The result is shown in Figure 7 which also includes the accumulative tolerances on nozzle positioning, i.e. the outermost fuel assembly corner is 3. 7 feet from the center of the spray rather than 3.55 feet as in Figure 6. The flow into the worst fuel bundle (bundle receiving minimum fMw) has been determined by numerically integrating the spray density shown in __

' Figure 7 across the top of the outermost fuel assembly for several values of nozzle pressure drop. The results are shown in Figure 8. As shown the pro-posed nozzle at the proposed elevation assures adequate spray flow as long as the nozzle A P is greater than 20 psi. but less than 51. 5 psi. N'ote that these limits are considered to be conservative, but operation of the nozzle outside this~ range is not recommended without more extensive data on the nozzle performance. Therefore, the following section provides the system pumping requirements necessary to assure operation of the nozzle within the 20 to 51.5 psi band over the range of reactor pressure from 85 to 14. 7 psia. D. System Requirements Necessary to Assure Adequate Cooling with the Redundant Sparcer The highest vessel pressure for which the core spray must pro-vide adequate cooling is 85 psia or 70 psig (3). The lowest nozzle pressure drop for which the 1 gpm per bundle limit is satisfied is 20 psi. From Figure 9 the nozzle flow rate with a 20 psi pressure drop is 384 gpm. Thus, including approximately 3.5 psi elevation head the pressure at the flahged connection to the reactor vessel must equal or exceed 86.5 psig for a flow rate of 384 gpm. Similarly in order to assure that the pressure drop across 7 the nozzle never exceeds 51.5 psi the pressure at the flanged connection must / not exceed 48 psig for a flow rate of 600 gpm. Considering the general shape of pump curves, Figure 10 was constructed to establish the limits for the pump and system characteristics. A review of the data indicates that both the core spray pumps and the fire system (through the highest friction loss path) satisfy the pumping requirements of Figure 10. However, the fire system supply through the lowest friction loss path runs out too far with the result that the nozzle pressure drop exceeds 51.5 psi and thus less than the required cooling flow reaches the outermost fuel bundles for low vessel pressures. To remedy this situation it is necessary to either avoid using the low friction loss path, if the reactor vessel back pressure is less than about 35 psig, or put an orifice in the line 1 i 4-cuch that this cupply also octicfies Figure 10 Tha letter msthod is recommended to avoid the possibility.of an operator misusing the line and also to add to the system flexibility. E. Vibration Analysis of the Redundant Core Spray Nozzle When the steam-water mixture flows by the redundant core spray, a flow induced vibration in this nozzle may exist due to periodic shedding of Karman vortices from its surface. If the ratio between the natural frequency of the nozzle and the vortex shedding frequencies is equal to or larger than 1.5 (6), the vibration will not be a problem. Using the methods described - reference 6,-the natural frequency of the nozzle and the vortex shedding frequency were calculated as shown below. 1. Assumptions: a. The spray nozzle sees a flow velocity corresponding to that in a steam nozzle. This is presumably the - highest fluid velocity above the core. b. The nozzle is excited by shedding of vortices from the 4-inch Schedule 40 pipe at the lower end of the nozzle. c. The nozzle is a 93-inch long, 8-inch, Schedule 40 pipe supported as a cantilever. (See Figure 11) d. The 4-inch Schedule 40 portion of the spray nozzle is in the flow. 2. Input Data: Steam temperature = 596. 2* F. Steam pressure = 1500 psia. Steam flow through the steam nozzles = 930,400 lbeIhr Water flow through the steam nozzles = 10,779,600 lbs /hr. Specific volume of water at 596. 2* F = 0. 0234 ft /lb. Specific volume of steam a; 596. 2* F = 0. 2766 ft /lb. '

.j ' Pipe O. D. - u 8. 625 inches. Pipe I. D. = 7. 981 -inches. 4 Moment of inertia of pipe: I = 72. 5 in. Modules of elasticity at 596. 2* F; E = 25. 4 X 10 lb/in. Steam nozzle'I. D. % 13 inches. Total number of steam nozzles = 6 3. Calculations: Weight of metal in pipe: = 28. 6 lb/ft. 3 Weight of water inside pipe: = 0. 785. X ( *12 ) X (10,779,600 + 930,400 ) E.2 1 X 0,0234 Weight of water displaced by pipe: = (8. 981)2 X 13. 67 = 15. 96 lb/ft. p3 12 Weight of steam inside pipe: 930,400 = 0. 785.X ( *12 ') X (10,779,600 + 930,400 4 1 X 0.2766 Weight of steam displaced by pipe: = (8. 625)2 X 0.10 = 0.12 lb/f t. p5 7.981 Total weight per unit length: 2+E3+F4+E5 p= 1 = 58. 45 lb/f t. The natural frequency of a cantilever (reference 6): If fn = uL Where a - 3.52 (reference 6, page 432) i

j .= 4 '2 g = acceleration of gravity = 386 in/sec r 6 '"

  • 3 ~. 5 2
25. 4 X 10 X 72. 5 X 386 X 12 Zu 58.45 X

(93)' = 24. 7 Isec. Total flow area of steam nozzles: 2-A=6Xn X (13)2 = 796 in Water flow through the steam nozzles: 0.0234 3 v = 10,779,600 X = 70. 07 f t /sec.. w 3600 Steam flow through the steam nozzles: = 930,400 X (0. 2766) = 71. 49 ft /sec. 3 v s 3600 Flow velocity through each steam nozzle: (70. 07 + 71. 49) = (12)3 = 307. 3 2n/sec. y= 796 Vortex shedding frequency: f =S1 exc d Whe re: S = Streuhal number = 0. 22 d = pipe O. D. = 4. 5 inches f = 0. 22 X 307. 3 = 15. 021/ sec.

4. 5 24.7

= - = 1.64 f /f n exc 15.u2 This ratio is larger than the limiting ratio of 1.5. Therefore, significant vibration will not occur in the spray nozzle. In the calculation of the vortex shedding frequency the nozzle was assumed to be located in an area - having a flow velocity corresponding to that in a steam nozzle. In the reactor, however, the spray nozzle is located underneath the vessel head in the area of the reactor centerline. There is a baffle plate between the spray nozzle and the steam nozzle. Hence, the spray nozzle will never see such a high flow velocity as assumed and the analysis is extremely conservative. 1

a-s REFERENCES 1. . Richardson, L. S., " Design Report, Core Thermal-Hydraulic Analysis, Consumers Power Company Big Rock Point Plant," APED 3949, June 1962. 2 Schraub, F. A. and Leonard, J. E., " Core Spray and Core Flooding - Heat Transfer Effectiveness in a Full-Scale Boiling Water Reactor . Bundle," APED 5529, June 1968. 3.. Allred, C D., "Re-Evaluation of Potential Loss of Coolant ~ Accidents at Big Rock Point," Test Design and Analysis Bulletin BRP-II, October, 1967. r 4.' . Big Rock Point Reactor Assembly Drawing, General Electric Draw-ing No.104R 175.- 5. Spraying Systems Company Drawing No. 12539. f 6. J. P. Den Hartog, " Mechanical Vibrations," 4th Edition, McGraw-Hill, New York,1956. 4 I I 9 -8 ,1

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l-4 7 CONSUMERS POWER COMPANY-Appendix 2 - To Request for Change to Technical Specification No'27 Docket 50-155 - License No DPR-6 Answers to USAEC Questions Question a: "A' description of the feedwater injection connections to the steam -drum and stress resulting from introduction of cold feedwater." Answer: The steam drum feed-water injection connections are of a carbon steel, conucruction and provided with' 304 stainless steel thermal sleeves. They are designed to' withstand 200 cycles of feed-water temperature change from 372 F (feed-water temperature at 100% load) to 85 F at a rate of 3 2 F/second. i Question b: "A' list of sources.and the' minimum volumes of stored feedwater available for makeup to the primary system fo'llowing small breaks." Answer: 'The_ condensate storage tank (25,000-gallon capacity) has a normal level' of about.65%. The water available from this tank under emergency conditions is .approximately 16,000 gallons. The demineralized water storage tank (5,000-gallon capacity) has a normal. level of about 65% and the transfer pump trips with 500 gallons remain-ing. Thus, approximately 2,500 gallons are normally available, but only et the pumping rate of 50 gallons per minute. The condensate hot well has a capacity of 6,750 gallons and the -con-densate pumps are tripped off.at 31 inches. The available supply from this source is normally 3,000 gallons. Waste storage available for use in the primary system is normally h,0v0 gallons. This is available at a 100-gallon ':)er minute rate through the treated vaste pump. Thef makeup demineralizer could also serve as a supply for small breaks. A! maximum of 10,000 gallons would be available following regeneration at a rate of 25 gallons per minute. -l-b._,U__.__l

m. - :C ( 4 . Question c: "A description of _the primary system leak detection methods and _their sensitivity in terms of break size and time required for' ~ detection." Answer: - A dew cell with a remote recorder is installed in an exhaust duct frem the steam drum cavity. A significant increase in the dew point temperature alerts <the' operator to a possible steam leak. - The increase in dew point temperature con-sidered.significant is that which is caused by a moderate valve packing leak. The 3 presence _ of a steam leak is confirmed by taking a grab sample for air particulate 3 activity on the steam drum cavity exhaust plenum. 'The minimum sensitivity.of this sample is 5 2 x 10" gpm based on a reactor water indine activity of 7 x 10" pCi/cc and 10% of the activity in the leak being carried away-by the ventilation stream. The dirty waste collection system for the Big Rock Plant typically runs -15 gallons per hour.. Doubling of this rate for no known reason will be reported i by'an operator. If this increase in collection rate cannot be explained by plant operation, a grab sample for air particulate activity is taken to confirm or deny the presence of a leak. The sensitivity.of this sample is as discussed in the preceding paragraph. Very small leaks in the control rod drive room can be heard on inspec-tion rounds as the background noise level is very low. An air particulate sample is routinely taken weekly on the steam drum ~ enclosure exhaust line. The sensitivity of this is 5 2 x 10 gpm as discussed above. This method allows detection of very small valve packing leaks. The maximum interval between inspection rounds is two hours. The de-tection methods discussed above are checked during these inspection rounds. Question d: "A determination of the largest primary system break that can be - tolerated without fuel clad melting - assuming normal and rapid cooldown procedures." Antwer: The cooldown method used for this calculation is drawing steam through the bypass valve while maintaining condenser vacuum and returning condensate-c and makeup water via the condensate and feed-water systems. The calculations do not take into account -the effect of'the makeup water on the cooldown rate. It _ assumes that the ' cooldown rate is limited _to 100 F or 300 F per hour for purposes 1: , ~ _

f q' of establishing a leak rate, thus a break size that can be tied to a specific cooldown interval and stored makeup water volume. For breaks of the size cal- ~ i culated below, it is felt the energy lost by' leakage and makeup to maintain water. level will determine a cooldown rate greater than that assumed and thus dictate the time for depressurization..-In. light of the above, the calculations-submitted below are somewhat meaningless, but do establish a break size that-can be tolerated for fixed cooldown rates (100 F/ hour and 300 F/ hour) and a specific ctored. water volume. If the cooling effect of the leakage and makeup water is taken into account, the cooldown rate and the tolerable break size -would both increase. Consider a small break below the core midplane. System must be cooled down from operating temperature (582 F) to a temperature (327 F) where the core spray can be manually initiated. At 100 F/ hour normal cooldown rate: a. 255 F

2 55 h = 153 Min Time Required

00F/h 0 Available Tankage Gallons j Condensate Storage Tank 16,000 Hot Well 3,000 l. Waste Storage Tank 4,000 1 1 Demin Storage Tank 2,500 25,500 Apply 80fu diversity factor stored water available = 0.8 x 25,500' = 20,400 Gallons. I Available From Makeup Demineralizer = 25 Gpm Stored Water Volume' Above Core Midplane = Initial Volume of Water in Vessel -' Volume in Vessel Below Core Midplane 4 3 = 2310 ft3 957 fg ' 3 f = 1353 ft 3 (Assumes average water density throughout the transient is 50 lb/ft,) 4 Leak Rate Allowable (Average Over Pressure Range) x 20,400 Gal x 33 D + 1353 ft3 50LbI I 25 Gal 8.33 Lb I f x Min Gal Gal ft3 + g 60 Sec/ Min g 153 Min x 60 Sec/ Min i 2 = 29 3 Lb/Sec which corresponds to approximately 0.004' ft break { size (assumes average pressure of 900 psig). ' e + -r. ,,w-me-

j' l 1: b. At 300 F/ hour-emergency cooldown rate: Time Required = =.85 h = 51 Min 300F/h Available Tankage From Part "a"' Above = 20,400 Gal' - Available From Makeup Demineralizer = 25-Gpm Stored Water V.lume Above Core Midplane 3 From Part "a". Above = 1353 ft Leak Rate Allowable (Average Over Pressure Range) I25 Gal-8.33 Lb I I 8 33 Lb y S3.ft3 50Lbl x 20,h00 Gal x x _ft3 Gal Min ' Gal + ( 60 Sec/ Min j { 51 Min x 60 Sec/ Min j 2 = 81 Lb/Sec which corresponds to approximately 0.011 ft area break size (assumes average pressure of 900 psig). Data submitted with Consumers Power Cetpany letter of February 9,1970,. shows that a~ small break above the core midplane will not result in any core melting prior to automatic initiation of the' core spray system, regardless of break : s'ize. _ uestion e: "A determination of the largest primary system break that can be Q tolerated without fuel clad melting assuming plant cooldown without off-site power. Describe the cooldown method. Discuss the operator actions that will be necassary including time availability." Answer: ' Assume off-site power fails at the same time or before the break _ occurs. This assumption yields the greatest amount of energy to dissipate prior to auto- ~ matic actuation of the-core spray system. Cooldown is by use of the emergency . condenser. The two control rod drive pumps will be operated to provide makeup to.the system.. Action required to place the emergency condenser in service is remote manually opening the emergency condenser return valves. The emergency diesel starts automatically. To start the control rod drive pumps, the operator . must close the 1A-2B tie breaker (2A-2B tie breaker nomally closed) and start the two control rod drive pumps.. Assume three minutes are required for these a actions. The cooldown rate using the emergency condenser is calculated as follows: 5: _4

,i Initial Condition' Temperature 582 F

Pressure 1350 Psia 3

Initial Water Volume 2310 ft Initial Steam Volume 1520 ft Decay Heat Rate Over First Hour ( Averages) 2.0% Rated Power Makeup Water Temperature 85 F Design Heat Exchange capacity of Emergency Condenser is I 15x10' Btu /h. Control Rod Drive Pump Capacity 25,000 Lb/h (50 Gpm) Final-Conditions Temperature - 327 F Pressure - 100 Psia 3 Final Steam Valume - 2873 ft 3 Final Water Volume (Core Midplane) - 957 ft Heat Contained in Nuclear Steam Supply System , Volume (Water) x Enthalpy (Water), Volume (Steam) x Enthalpy (Steam) Specific Volume (Water) Specific Volume (Steam) . Initial ,2310 x 592.1 + 1520 x 1176.1 0.0229 0 3146 7 - 6.Sh x 10 Btu Final , 957 x 298.h 2873 x 1187-2, 0.01774 4 432 I

1.687 x 10 Let t = time to cooldown in hcurs (Emergency Condenser Heat Transfer Rate) x t

(Initial Heat. Contained) - (Final Heat Contained) + (Decay Heat Rate) x t --(Heat Lost Through Break) + (Rate of Heat Addition From Makeup) x t Substituting: 7 7 7 5 3 15(10 ) x t = 6 53(10 ) - 1.682(10 ) + (.02) -(2.40 x 10 ) (3 413 x 10 ) x t + 25,000 x t 5 266 x 10 + 25,000 x Sh x t 5 w _,,,..- - .,i-T. e .--,w.q,, p.- .97 -.-p-

I I 7 7 I 6 15(lo )t - k.85(10 ) + 1.638(10 )t - 3 36(10 ) - 1317(10 )t + 135(10 )t 7 I 1.04h(10 )t = 1.hg(10 ) t = 1.43 hours or 85 5 minutes. Mk Rate (Assuming 900 Psia Average Pressure ) (and Adding 3 Min Operator Reaction Time) 1353 ft3 60 See + 50 calhiin x 8.33 Lb/ Gal 60 Sec/ Min .0212 ft3 x 88.5 Min x g Min = 12.0 + 6.95 Lb/sec =1895Lb/sec This leak rate corresponds to a 0.0027 ft break size. For any size break above the core midplane, operator action is not required to preserve core integrity. Cooldown rates above 300 F per hour using the feed-water system and redundant off-site power provide the mechanism for rapid core cooling to the tem-peratures and pressures required for core spray system actuation prior to fuel clad melting in the limiting cases of small primary system breaks. Feed-water addition rate versus pressure decay and leak rate curves are being developed for the limiting small primary system break cases at Big Rock Point. Tests will be conducted during the February 1971 refueling outage to determine the adequacy of stored water transfer rates to the condensate and feed-water systems. The re-sults of these studies and tests will be submitted by letter as soon as they have been evaluated. From these data and evaluations, conclusions will be drawn about the adequacy of the core spray systems to cover breaks in pipes of any size. What-ever the results of these evaluations show, it is Consumers Power Company's in-tention that provisions will be made to guarantee an adequate water supply so that one feed pump will be operable and capable of delivering water to the reactor core for cooling. l i i i t. iii,,,__t ,,1 4 , fe ";f--.

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