ML20091D714
| ML20091D714 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 05/24/1984 |
| From: | Robert Carlson CAROLINA POWER & LIGHT CO. |
| To: | Atomic Safety and Licensing Board Panel |
| Shared Package | |
| ML20091D695 | List: |
| References | |
| OL, NUDOCS 8405310453 | |
| Download: ML20091D714 (16) | |
Text
_-_ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _- -_____ __ . _ _
Rtt.ATto CCanESPONDEN% .
$hc ED UNITED STATES OF AMERICA i NUCLEAR REGULATORY COMMISSION ,
- g4 MY$ A10:15 .
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
CAROLINA POER & LIGHT COWANY)
AND NORTH CAROLINA EASTERN ) Docket No. 50-400 OL MUNICIPAL POWER AGENCY ) 50-401 OL (Shearon Harris Nuclear Power )
Plant Units I and 2) )
AFFIDAVIT OF ROBERT W. CARLSON County of Allegheny )
SS:
Coamonwealth of Pennsylvanta ) ;
Robert W. Carlson, being duly sworn according to law, deposes and says
- 1. My name is ROBERT W. CARLSON. My business address is P.O. Box 355 Pittsburgh, Pennsylvania 15230. I am employed by the Westinghouse Electric - ,
Corporation as a Principal Engineer in the Reactor Coolant and Steam Generator Support Systems Group of the Nuclear Technology Olvision. l
- 2. I received a Mechanical Engineering degree from Stevens Institute of Technology in 1953 and a Master of Science degree in Nuclear Engineering from i Massachusettes Institute of Technology in' 1959. I also attended Case Institute of Technology for two years, from 1965 to 1967, as a full-time graduate student in the field of Thermal Sciences.
- 3. I accepted a position in 1953 as a Boller Division student engineer wIth the Babcock & Wticox Company. Af ter the one-year program, ! joined the Babcock & Wticox Company Atomic Power D1yision. In 1955, I took a leave of
' absence for military service and MIT Graduate School. I returned to Sabcock &
Wilcox in 1959 and was later promoted to the position of Senior Engireer in neurm%
i .
l the Atomic Power Divisiono I joined the Westinghouse PWR Systems Divisten as l a Senior Engineer in 1967. My initial duties were as a reactor core thermal and hydraulle designer. In 1975, I was promoted to my present position of l Principal Engineer.
- 4. During 1975 and 1976, ! participated in a test program conducted by Westinghouse at its Research and Development Center in Pittsburgh to investigate the bubble collapse waterhamer phenomenon. One objective of the test program was to gain a basic understanding of the bubble collapse phenomenon in horizontal pipe sections.
- 5. In 1977, I participated in a test program to investigate the potential for bubble collapse waterhammer in preheat steam generator designs, including the Shearon Harris type steam generator. ! was responsible for the thermal and hydraulic design of preheater scale model test sections and the test vessel. I was also responsible for the initial evaluation of the test data.
- 6. In my present position, I meet, on behalf of Westinghouse, with l utility customers and their architect engineers to provide assistance in the j design and operation of the plant, as recommended by Westinghouse, to minimize -
the potential for a waterhamer evaat.
- 7. Therefore, I have personal knowledge of the matters stated herein and believe them to be true and correct. I make this affidavit in support of Applicants' Hotion for Summary Disposition of Eddleman Contention 45. This ;
contention readst i SHNPP design cannot comply with the results of the Plant Waterhammer Experience Report PWR $.G. (steam tenerator), feedwater ECCS & Main ;
Steam System waterhammer events evaluation (including systems effect) and potential resolutions now being prepared by NRC, and the CR and '
NUREG reports on the waterhammer question.
l
- 8. The purpose of this affidavit ist o To describe those aspects of the design of the Shearon Harris Nuclear Power Plant (SHNPP) steam generators, feedwater system 4
9 I
4
. 2
and emergency core cooling system (ECCS) relevant to the
, potential for waterhammer, and o To show that those features described provide for designs with minimized potential for unanticipated waterhammer events, and o To show that, even in the unlikely event that a waterhammer were to occur in one of the above systems, it would not be expected to affect safe plant operation.
- 9. The conclusions set forth in this affidavit are consistent with those reported by the NRC staff in NUREG-0927, " Evaluation of Waterhamer Occurrence in Nuclear Power Plants-Tec'inical Findings Relevant to Unresolved Safety Issue A-l". In that report, the staff concludes that the overall incidence of waterhmmer in nuclear power plants has declined considerably in recent years.
Although the staff finds that total elimination of waterhammer is not feasible, they conclude that the frequency and severity of waterhamers is significantly reduced through prcper design. lloreover, the NRC staff raports that none of the waterhamer events which have occurred placed the plant in a faulted or emergency condition or resulted in a radioactive release. On the basis of these and other key findings, the NRC is not recom...ending hardware or design changes for existing plants or plants under construction in its resolution of Unresolved Safety Issue A-1.
- 10. Before discussing specific system design features, I will describe the mechanisms for water hamer. In general, there are two forms of waterhamer, classical (flow into volded region and suddent interruption of flow) and bubble collapse. In both cases, a change in water velocity leads to a change in pressure due to the inertia of the water. The two forms differ with respect to the mechanisms which causu the change in velocity.
II. In a classical waterhamer, the change in water velocity is typically the result of a sudden interruptfors of the flow stream or flow into a volded region. As an example of classical waterhamer, consider a pipe with water flowing inside. If a valve in the pipe is closed quickly, the water will be 4
3
brought to rest and, as a consequence, a sudden pressure increase util result at the valve. This change in pressure will travel as a wave back and forth in the pipe until it dissipates due to friction.
- 12. Bubble collapse waterhammer refers to a potential condition where initially a volume of steam is trapped within an enclosed region, for example, a horizontal section of pipe with water slugs on both sides. If the temperature of the water in the slugs is the same as that of the steam, the water and steam will be in equilibrium. Ilowever, if the slugs contain cold water which c'omes into contact with the steam, the steam will condense rapidly resulting in a sudden local decrease in pressure. A higher pressure behind the water slugs will cause them to accelerate towards each other. When they collide, an increase in pressure will result. This change in pressure will propagate back and forth in the water the same as in the classical waterhammer case.
- 13. The magnitude of the pressure change produced at the valve in the classical waterhamer example depends on the rate at which the valve is closed, the initial water velocity, and.the density of the water. In the bubble collapse waterhamer example, the pressure change magnitudo depends on the rate at which the steam is condensed and the pressure behind the water slugs.
! StiNPP STEAM GENERATOR i
14 . In paragraphs 15 thru 33 I will discuss those aspects of the SilNPP steam generator design which are relevant to the bubble collapse and classical waterhammer phenomenon.
- 15. The Carolina Power and I.lght ("CP&t.") SliNPP utilizes a Westinghouse designed nuclear steam supply system (N555) consisting of three rectrculating reactor coolant loops. Each loop contains a Westinghouse Model 0 4 steam generater. Within each steam generator there are 4578 inverted U shaped steam generator tubes, collectively referred to as the tube bundle. The tubes act as the pressure boundary between the primary (reactor coolant) water and the 4
L
. \
l
, secondary (steam producing) tater. Th3 tubes are secured at the end of each l leg of the "U" to a thick steel plate known as the tubesheet. This acts as the primary-to- secondary barrier before the primary water enters the tube
{
l bundle. The het reactor coolant water flows through the inside of the tubes. , i The tube bundle is immersed in relatively cool secondary water which is raised
- to steam producing temperatures by the transfer of heat through the walls of ;
- the steam generator tubes from the primary water to the secondary water. High !
j quality steen exits the top of the steam generator and is used to artve
]
turbines which in turn deive a generator te produce electric 1ty. The presence
, of both steam and water on the secondary side of the steam generator accounts i 'for the fact that the potential for bubble collapse waterhammer is a i j consideretten in the design of the steam generator. i
[
i . i l j 14 . An outline drawing of a Model D-4 preheat steam generator is i j contained in Attachment I hereto. As shown in the figure, the preheat region j is located on the cold leg side of the tube bundle and faces the feedwater
[
j inlet nosale. In the Model D-4 steam generator, the incoming main feedwater l
j flow enters the inlet water box and layinges on a well that directs the water j j outward to fill the water box volume and downward to the preheater inlet pass i j located near the bettee of the steam generator before entering the tube l
! bundle. See Attachment I herete. The water then enters the tube bundle at !
- l. the inlet pass, flows around the tubes and then upward around the tubes and l
) baffles. This upward flow is " counter" to the direction of the flow of the l prleary water inside the steam generator tubes. 2
- 17. As shown in Attachment I, the SHNPP. steam generator contains, in addition to the main feedwater inlet nosale, an auxillary feedwater noaale in
^
{
] the upper shell. One purpose of the auntilary feedwater nsaale is to steletae '
the potential for bubble collapse waterhammer in the prenester region of the l steam generator. As stated earlier, one of the elements required for bubble
} collapsa waterhasser is cold water. Cold water acts as a leest sink which i causes the condensatten of the steen and collapse of the bubble. Although the !
l probab11ity of occurrence Is remote, It is conceivahle for steam pockets to ,
i form in the preheater section. Therefore, in order to mintatae the potential for bubble collapse waterhammer, when it is necessary to introduce cold 't feedwater into the steen generater, the feedseter bypass system design directs l
the cold fepdwattr thru the upper auxiliary fecdwater nozzle to the steam generator upper shell region. In the upper shell region, the cold feedwater can mix with the bulk steam generator water in a region where steam pockets i will not exist. i
- 18. The' design of the feedwater bypass arrangement for the SHNPP steam generators is based on the results of a c'omprehensive waterhammer test program carried out by Westinghouse, under ray supervision, in 1977 and 1978.
One-eighth scale models of preheat steam generator designs, including the Model D-4 design, were tested under simulated plant operating conditions, including pressures up to 1,000 psia. The test objective was to determine' those conditions where bubble collapse waterhammer could occur in the steam generator anddmediately adjacent upstream feedwater piping. All of the tests consisted of two steps: first, establishing conditions where steam
-would be present in the preheater region, and, secund, introducing water at different conditions to condense the steam,-
- 19. Two different types of tests were conducted. In one, referred to'as Type A, the water level in the test vessel was lowered below the prehester
- section, a situation which could conceivably occur following the faulted condition of a main feedpipe rupture. Once the water level was verified to be below the preheater section, feedwater was introduced'through the' feedwater nozzle. Any resulting pressure pulses were measured and recorded. In Type B.
, testing, the water level was maintained above the preheater section and steam i was generated in the preheater region by means of electrically heated rods which simulated the steam generator tubes. Again, after the desired
, conditions were established..feedweter was intro ~duced through the feedwater nozzle and any pressure. pulses produced were measured and recorded. The Type i B condition simulated normal and upset conditions for the' steam generator, such as plant loading. '
- 20. In addition to the initial water ' level in the-test vessel, other. ~
L principal variables in the test program'were pressure, feedwater flowrate, and
~
!~
feedwater temperature. : Tests were conducted at different ' pressures up to
' steam generator normal operating pressure. The feedwater temperature was-varied from approximately 80'F to 250*F.
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- 21. The most significant result of the test program was that waterhamm:r
' ~
did' not ' occur if the temperature of the feedwater was 250*F or higher. The
.L design of the feedwater bypass system is based largely on this result.
Results of this test program also indicated that the' potential for bubble collapse waterhammer was significantly reduced at the pressures at which the steam generator normally operates. Although this further reduces the potential for a waterhammer event, no credit was taken for this finding in
- designing the bypass system.
- 22. A final test series was conducted to simulate the effect on the preheater of introducing cold feedwater (< 250*F) through the auxiliary feedwater nozzle. No waterhammer events were detected.
- 23. In addition to evaluating the potential for bubble collapse waterhammer in the preheater for various conditions of temperature, pressure and flow, the one-eighth scale test program also provided bubble collapse
- waterhammer loadings for input into the preheater structural analysis. The results of this analysis indicate that the steam generator primary coolant pressure boundary is maintained under normal, upset and faulted bubble collapse waterhammer loadings.
- 24. On the basis of these design features and testing, I conclude that there is minimum potential for bubble collapse waterhammer in the SHNPP steam generators. In addition, on the basis of the testing and subsequent structural analy:is, in the unlikely event that a bubble collapse event did occur, the primary coolant pressure boundary would be maintained.
l
- 25. Two recent events which applicants have discussed in response to Mr. Eddleman's inte'rrogatories dated March 26, 1984 occurred at two different-operating nuclear plants with steam generators.which were not manufactured or
! designed by Westinghouse. Damage was attributed in one case to waterhammer l
l and in the other case to either waterhammer or fatigue. ~ '
\
- 26. In .the first case, the event -involved the feedring (sparger) of a feedring design steam generator. In. this design, feedwater is normally provided to the steam generator thru the feedring (sparger) which is located 7
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1 in the upptr part of the steam generator vessel at the approximate elevation of normal water level. Since-the SHNPP steam generators are of the preheat :
type and 'do not have feedrings (spargers), this particular event cannot happen i at the SHNPP. .
- 27. In the second case, the event involved a sparger inside the' steam gegeratorsforinjectinganddistributingauxiliaryfeedwater. The sparger cojsijted of a horizontal, curved pipe section with an arc of approximately ,
! 120 degrees. The SHNPP design for injecting auxiliary feedwater into the steam generators is significantly different. Instead.of a horizontal sparger the auxiliary feedwater is supplied through a simple upwardly inclined pipe j section. Again, because of the fundamental differences in design, the incident involving the horizontal sparger cannot occur in the SHNPP steam
' generators.
- 28. Thus, an evaluation .of the two recent events ' cited does not alter my previous conclusion that there is minimum potential for bubble collapse
[ waterhammer in the SHNPP steam generators.
- 29. In addition to the bubble collapse waterhammer design considerations,
. the SHNPP Model D-4 steam generator is designed for classical type waterhammer i loads resulting from events which can originate in the Feedwater System or Steam System.
l30. Limiting classical waterhammer pressure loads acting through the steam generator main feedwater nozzle.are considered for two specific events; a feedline rupture followed by rapid closure of the main feedwater check valves, and a steamline rupture resulting in a high flowrate through the main feedwater nozzle into the preheater. The feedline rupture-check valve rapid closure transient was considered assuming the maximum loading condition of instantaneous closure of.the check valve from maximum possible reverse flow.
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- 31. . Westinghouse has analyzed the effect of these transients' for the l SHNPP. steam generators. The'results of the, analysis show that the integrity of the steam generator.is. maintained and that safe operation of the steam generator is unaffected.
i L 8
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- 32. Limiting classical waterhammer pressure loads acting through the auxiliary nozzle are due to bypass line check valve closure resulti,ng from 1 l
reverse flow due to a main feedline rupture. As for the limiting loads acting I through the main feedwater nozzle, safe operation of the steam generator is unaffected.
II MAIN FEEDWATER AND FEEDWATER BYPASS SYSTEMS j 33. Having reviewed the design features of the steam generator which minimize the potential for and effects of waterhammer, I will now discuss the design of the feedwater system relative to the potential for bubble collapse waterhammer. Dean Shaw's affidavit discusses classical waterhammer in these systems. The basics of the feedwater system typical of that used at the SHNPP are shown in Attachment 3 hereto. The system includes a 16 inch main feedwater line which connects to the main feedwater nozzle. Four principal i
valves are associated with the main feedwater line; the main feedwater control
- valve, the main feedwater control valve bypass valve, the main feedwater check valve and the main feedwater isolation valve. A smaller (6 inch) diameter bypass line connects the main feedwater line, between the main feedwa'ter
- control valve and the main feedwater check valve, to the steam generator auxiliary feedwater nozzle. The bypass line itself contains an isolation valve and two check valves.
- 34. During plant startup, feedwater is supplied to the steam generator only through the auxiliary nozzle. During plant loading (esca11ation in power), feedwater supply will be switched to include main nozzle supply only i after the following criteria are satisfied:
(1) A minimum feedwater flowrate of approximately 15% of the. full power flowrate is provided.
(2) _ The feedwater temperature is 250*F or higher as measured at the low points in the main feedwater piping. - ~
(3) The section of mair: feedwater piping between the bypass line branch-point (Point A, Attachment 3) and the main feedwater nozzle has been
. purged of cold water.
(4) The steam generator' pressure is greater than 700 psia.
l (5) The steam generator water level is within a specified range.
9
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- 35. Duri: 1g plant unloading, the same criteria apply except tha.t the I feedwater flow is switched to only the auxiliary nozzle when the flowrate drops below approximately 15%.
. 36. The fact that these criteria must be satisfied to permit feedwater flow through the main nozzle makes it extremely unlikely that bubble collapse
- waterhammer will occur. This conclusion;is consistent with that reached by the NRC in NUREG-0927 (page 2-22) regarding preheat steam generator ,
waterhammer potential. Specifically, the NRC states that "the occurrence of j an SGWH (Steam Generator Waterhammer) event in a PHSG (Preheat Steam Generator) would require multiple component failures (including several check- ,
l valves and operator errors). Even if such an event occurred, it is not expected to have an adverse effect on plant safety or AFW system operability".
- 37. As indicated in Attachment 3, the Auxiliary Feedwater System connects to the feedwater bypass line. The Auxiliary Feedwater System provides feedwater to the steam generator through the feedwater bypass piping and the auxiliary feedwater nozzle in the event of a loss of heat sink accident, such as a feedwater pipe rupture.
. 38. One postulated phenomenon considered in the design of the SHNPP feedwater bypass system is that of steam backleakage from the steam generator into the feedwater bypass line and then into the Auxiliary Feedwater System piping.
i j 39. The auxiliary nozzle connects inside the steam generator to an j upwardly inclined pipe extension, the discharge end of which f.s below the
) normal operating water level in the steam generator. For steam to push back i into the bypass piping, it would be necessary for. the check valves, which are provided to restrict reverse flow to be leaking and for the steam generator water level to be below the auxiliary nozzle internal extension. If the water .
is kept at the normal operating level, steam cannot enter the internal ' '
extension and thus cannot enter the bypass piping.
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10 l .- .
I
- 40. The feedwater control system is design:d to maintain the steam generator water. level above the top of the auxiliary feedwater discharge pipe inside the steam generator. During normal plant operation, with the discharge pipe covered, only hot water and not steam could leak back into the bypass and Auxiliary Feedwater System piping, thus greatly reducing the potential for waterhamer.
- 41. Moreover, steam backleakage during normal power operation is very unlikely since system design is such that normally continuous flow is nrovided through the steam generator auxiliary nozzle which effectively prevents tha backflow of steam from the steam generator.
- 42. During heatup,.cooldown and hot standby operations, relatively small amounts of feedwater are supplied to the steam generator by the Auxiliary Feedwater System through the auxiliary nozzle. This system is designed to provide continuous feed rather than intermittent feed as much as possible, minimizing the potential for steam backleakage and the potential for , ,
waterhamer.
- 43. An additional design feature of the fee. ater bypass system to minimize the potential for a water hamer of this type is the installation of two temperature sensors on the bypass piping inside containment Elose to the auxiliary feedwater nozzle of each steam generator. If the measured temperature valt.es 5xceed a predetermined setpoint, an alarm is activated in the control roon.
- 44. In the eventuality that the presence of steam is suspected in the bypass line, based on temperature data and water level status and history, the system can be recovered by slowly purging the bypass line using the Auxiliary Feedwater System at a rate of approximately 15 gpm.
- 45. Based on the design features of the auxiliary nozzle and its internal ' '
extension, the normal operating conditions, and the means provided for alarming and recovery from back leakage of steam if it should occur, the probability of bubble collapse waterhamer in the feedwater bypass line is minimized. This conclusion is consistent with that reached in NUREG/CR-3090 11
r .
which evaluated the potential for waterhammer occurrence during Auxiliary Feedwater operation of preheat steam generators and concluded that the likelihood was extremely low. Furthermore, if a waterhammer event did occur',
NUREG/CR-3090 concluded that the event should have no adverse effects on 1 1
Auxiliary Feedwater system operation or plant safety.
l III ECCS SYSTEM
- 46. Another system identified in the contention is the ECCS. The potential for waterhamer occurrence has been and continues to be a consideration in the design of the ECCS. As a result of this consideration, I
the Shearon Harris ECCS is inherently not susceptible to waterhammer type
- pressure pulses resulting from sudden check valve closure, sudden pump startups and stops, fast acting isolation valves and relief valve operation. Waterhamer-is most easily dealt with in the design mode so that it will not occur during normal and transient conditions. The postulated waterhammer mechanisms listed above are layout dependent phenomena, resulting from interaction between various components within the system. These have historically not been identified as an issue in any Westinghouse 3 loop ECCS such as at SHNPP. Preoperational testing and years of operating history i supports this position.
- 47. Another postulated waterhammer mechanism would be that due to voids in a water solid system. Voids within, the ECCS are minimized through various
- design features as supplemented by procedurai, and administrative ,
constraints. Westinghouse provides information to the piping layout designer to minimize the potential for voids within the system. The architect i
engineer's utilization of this information will provide a design responsive towards minimizing voids. One layout consideration identified is.to provide l pump suction piping which is self-venting and free of potential gas pockets.
- 48. An additional recomendation made to the piping designer is to provide adequate venting capability of. the system. Vents are to be provided in the high points of any piping loops where gas could collect and interfere with proper system operation.
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- 49. Voids have b u n postulated in water' solid systems as a result of 1eakage. However, due to the head provided by the Refueling Water Storage Tank, Accumulator Tanks and Boron Injection Surge Tank, the majority of the
~
Safety Injection System is normally maintained at a pressure higher than ,
atmospheric. As a result, any-leakage in the system should be outward with water makaup from within. Makeup would be provided from the tanks identified.
- 50. Furthermore, the probability of pressure boundary leakage is extremely low due to the high quality standards ' applied for the design, construction, installation and inspection of the ECCS piping. The piping is essentially of welded construction with minimum potential for leakage.
Stringent manufacturing inspections, pre-service inspections and inservice inspections provide for the leak-tight integrity of the system.
- 51. These aspects of the ECCS design take~1nto consideration the 4
mechanisms of ECCS waterhammer identified and evaluated in NUREG-2781. As a r
result of these design considerations, the potential for waterhammer in the .
SHNPP ECCS has been greatly reduced. Further, the NRC's review of waterhammer l
events as reported in NUREG/CR-2059, concludes that ECCS waterhammer events reported at pressurized water reactors have not had any adverse safety effect on a plant.
- 52. In summary, I am confident that the design of the SHNPP steam .
generator, main feedwater and feedwater bypass systems, and the ECCS minimize
, the potential and consequences of water hammer in those systems and that the issue 'of waterhanner in those systems at SHNPP is not a _ safety concern.
I f '
} U &aw -
Robert W. Carlson-
. y Sworn to and subscribed before me this 7.M ay of )/Ad41 , 1984. ' '
I J n;[,:
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, , Notary Public
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LORRAINE M. PIPLICA. NOTARY PU8 tlc -
NONR0fVitLE 80RO, All[CitENY COUNif liy ConsissIon expireex0mmissiON ExPIRts etc 14. issi '
Member. Peissylvania AssociatHm of Netanes i
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l ATTACHMENT 1 MODEL D-4 PREHEAT STEAM GENERATOR
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l ATTACHMENT 2 MODEL D-4 PREHEAT STEAM GENERATOR
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MAIN FEEDWATER N0ZZLE AND PREHEATER REGION
ATTACliMENT 3 TYPICAL FEEDWATER SYSTEM DESIGN u
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