ML20234B839
| ML20234B839 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Brunswick |
| Issue date: | 05/15/1969 |
| From: | Hanauer S Advisory Committee on Reactor Safeguards |
| To: | Seaborg G US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20234A777 | List:
|
| References | |
| FOIA-87-40 NUDOCS 8707060226 | |
| Download: ML20234B839 (5) | |
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May 15, IM9 hh.
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Memorable Glenn T, amehreg Chairmen W. 8. Atomic taergy h==d==les j
washington, D. C.
20545 Subject REFORT QN SklBMWICK STEAM ELaCTRIC FLANT Wits 1 AN 2
Dear Dr. seaborg:
At ice 109th asetieg, May S-10, EM9, the Adrieery 8:a-ittee se Reester safeguards reviewed the applicaties by Carolina peuer and Light Coupesy for authorisaties to saastruct hits 1 and 2 of the Brunswiek Steam Eteetric Flaat. The project was semaidored at a Sabeammittee swating and site visit sa April 30, IM9. During f.ts review, the Commintes had the benefit of dieemestems with repeseestatives of the Carolina poner and Light Campany, Soneral Elastric Ceapsey, Weited sagineers med Can-strusters, Zac., the AEC Repelatory staff, and their semesttants, h Casatittee slee had the homofit of the desmasats tieted below, h Srwaswick Flent will be lesated La Brunswisk County, Berth Case 11as, apprezimately two and eme-half miles morth of the team of Southport and sisteen miles meeth of Wilmiastem. h site is near the unet haak of the Cape Fear River and aheet five miles from the Attentis Seese. A adminus onetreLes dietance af 3000 feet hee been prorided. Me 1M6 popetetsee withis five mitee of the site totals apprestastely 3500s the amarent pop-mistia: eenter of 15,000 people er more is #11miastem. h sammy point Army To::winal, wood for the transaktyment of military munittaas, is le-1 sated four miles morth of the site.
l seek unit of the pimet will settise s General 11 metric hotling water reas-ter siellar to that provided for the Casper Beelear statime, uhteh one dis-eussed is else Committee's report dated Moreh 12, 1968. h seesters are eseestially identisal to the one proposed for the Edwin 1. Bateh puetaar Flaat, ales under review for a seastruction permit. h Brumovisk remetere ses designed to produce 3636 aart with a menisess perfe means rettes ed R$N 1318. Seelias unter will be takes from the Cape peer River sad dioshauged 1
to ths attaatte Oseas.
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Pbnn AW 818 $n. *-m 8707060226 870610 u w.,,,,, m a PDR FOIA THOMAS 87-40 PDR v
menerable $1ema T. soehorg May 15, 1969 DV 3
Sn geology and meteoreley of the site appear favorahla. Peoriales y e411 he made to preteet the pleet acaimet earthgenbes, tornadoes, ab Z,'.,
v' riesdies. the teoul of fleedias, instudies woes height and vose sun-up, that may reemic from herr -- has been esteestand to asessessee e
with Essa report IWR-7-97. the plant will he pestested fees fleedies to the mastmum still unter leest seputties from a hurriaame. In meds-tion, the applicant should develop posvietano and pressemens to poetast egaipment assentLal to safety from the efdeste of umpe assism.
b effects of air h1ast, yound metism, and asses seesiting from an en-plesias et the summy Peist Asur Ensustaal er sa a ship la the Caps pher 11ver hees been eensidered and Sound to be emailer them these fees other searcas. the offests of mimettes free oesh esplestems slee have been ese-sidered and the prehability of eesh a edasile striking the plant has heem found to be entremely Leo.
She primary sentainment consists of a eteel timed setsiereed seasseta dry-mall and suppression shaaber simiter la geometry to the steet ma=eme==n of thLe type. This will he the first seinfereed eenerste me=*a8====* of this type to be seestrusted. the applisant has saastuded that the prheary eestainment is capable of sostating la-plans (tangential) sheare from earthquaka forces by aggregate interiesk, etener-fristian, er deust settaa is the senerete, by sheer in the steel 11 mar, or by a sed inattaa of these noshaatsee, het with sophesis en sheer-friation. the Cemmittee holieves that diagemal stes! seinfersing here ehem1d be psTridad to resist thhee ta-plane shears malees souriassag experimental evidense saa he possented to demonstrate that the other==^=mi-meetismed, either slagly er in ensimettes, saa resist the applied shears withest tapatring the 1stegrity of the steel liner.
the applicant has agreed te supply for series by the Re p1 story staff pre-11minary details concerning the des 1 e of the pr$ mary eentaieneet, espe-
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3 stally with regard to the equipneet begehas and other large peastratises, l
prior to their cemetructise. Sistlar informaties should be psevised som-oorsing the details of the asients analyets of Class 1 streateses and een-pomeets.
Several problaas unique to betting ester someters have boss tematified by the Replatory staff and the ACR$ and sited la prorises ACRs soports. The Committee heltsves that reeetetten of these items absold apply agently to the breesisk plant.
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i Ms Gemedttee sentianos to setterste its interest ta am appropriate psessen Ser Aaservise taspection of the seester primary erstem, the applisant de gestostias a study to establish a more vigorous inservise Laspecties program han that imittally proposed and to spesify design provistans to facilitaae OFFICE >
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-f-Pbrm AEC.818 (Rev. D-83)
U.S GettaWEM PWMmG WTK1 d9b6-O 2W29 J
Bemerable glean T. seeberg 3*
May 15, 1969.~ 3
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the see progree, partisslarly with sagend to eseems to the primary appy %
tam. b applicant stated he will gees eereful ettentian to the pose $e olens of the draft asA standard am taservise inspecties in this stug, '
and be will eeeplete the study withis sin to mine h a.
h bogste-tory Staff should review this program and should soport the saoulte of its review to the Caummittee.
la the sees of remeter Lastressatatism, the Canadttee believes:
(a) that the red bleek mesitor eyoten saa perform an important estety, as we11 as eporatimmet, fenseism m d that Ameerpe-retten of such a ayetes, or its ogmivalent, is aseosomeys l
(b) that there should be esitable prortaiens to essere that ter-i pressure sore emeling sapability will be ove11able before the este-relief depressurisettes eas he tattiated; (e) that the f1mm serem yeint should be automatiselly seemoed to em appropriate level as the remeter restrostateme flow is so-dueef below the mesmal full-power flew; (d) the systems whish perform these feastimme should be built to meet appropriate protesttee system eriteria. Se ersteria to be used for ensk system should be estab11 abed on a basis asseptable to the anguistory Staff.
The Committee believes that, for transients having a high peebebility of escurressa, and for iditch astian of a pastestive eyeten er other engi-asered safety feature is vital to the public health sad safety, se en-seedingly kia!b prehability of sueeessful setten is needed. Gemens initore modes must be seasidered La assertaisias an aseeptable level of preteettaa.
Is the event of a turbine trip, rs11anse is placed en preset control-red eeram to prevent large rises is primary system pressure. She applisant and his sentractors have devoted semelderable effort to providias a rettable protective system. Eeuever, systematic failures des to taproper desige, operattaa, or meistemasse seuld ebriate the osram tollability. the Cee-mittee recomenade that a study be made of further means of preventing een-som failure modes from angeting seram eettaa, ami of desige features to make tolerable the senoeguemees of fattere to seram during entiaipated trametests.
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.tr -s of.nsign of the e.gime. red -f.t, feature., the.p,u.al ma, t sing a fissian predest souses term emeller than that omegasted sa 14844, sad a treatment of this eeurse within the seetatement diffemmat foam that rosammended La the pass desmenet. She Cessaittee holieves that j
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Fbrm AEC.818 (Rev. 9-53)
U1 GovtRNMEM PRMlhG CFFICE.156-0 214429 I
Esmosable 81ema T. Seeberg May 15, g
- nsy, k,v4h_.4 the assumptions of Tu-144u abound be used as a desipa besto ser the?<4,6 engineered safety features of the brassisk plant, unless auf estil the use of a different ont of assumptions has been justified to the sattsfesties of the Bragelatary Staff sad the Amt.
h Comunittee weiterates las semaeus that the post-assidset ecoltas spe-tem retain its integrity throughout the amares of en assident and the suboepsont seeling period. h e app 11samt should vertse the etSeste of eeelant toaperature, pE, sediasetivity, eerresive meterials from the esos or other parts of the aestainerat (&meludtag stored sheadsals), and poten-tially abrasive slurrime. hagamoratian of e = ouch as itlasse, puny tape 11ere, and esals by any of these nochemismo eheeld be scrisued.
partasular attentiam should be paid to potential prehlees arising foam the use of dissimilar metale La these spotame.
tagineered safety systems that are regoired to resivustr.ca setsr after a less-of-cooles: mocident show!4 he designed se that a grees apoten lack i
will set result La srittaal lose of seaisostattaa er la Amos of Asenaties
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espebility. h Caussittee believes thet ameoptian to this gemoral rule may be made in ru pest to a very short run of pipe from the eseus ta the first velve, if astrumely semesavative design of the pipe (and las een-meetian to the torum) is used and esitably remote operchtlity of the valve is provided. De destas of these eyetame slee should pestido ade-geste leak detection and servet11anse capability.
Studies ase sentimming as the peesthis effoots of reiteiyets of unter la the unlikely event ef a tese-ef-eeelmut nosident. he Commitsee helteres the applismet ehes14 evatusta all probleme ubish may arise from hydrogen Dameraties, inetustag varteen levels of Einsaley-uster seastians ehtak eewld esaur if the effectivaanes of the emergemey core eseling spotam unre stamificantly lose them thet prediated. h matter should be re-solved between the applicant med the ABC Bogalatory Staff.
he Committee has reviewed the applicant's proposal soneersing standesde of design, fabricaties, and inopoetiam of the steam limes desenstream of the seseed testation valve. h Committee sommers with the approach used La analysias the stresses in the piping eartag an epo6:ettag basis Earth-ganha. b Committee rosammeeds that a progree of spot radiography of the fis1d bott unids be employed by the applisset as a gestity esetrol sessuse. Camelderaties should be saves to an appropriate peogram of in-LEspesI M.
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I'Orm AEC=818 (Rev. 9-83)
U.s sovrnmarwi rtuwTms omcs ase-o ri442e
Emmerable Steam T. Seabers 5-May 15,.4MD,
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. h des believes that the above steam saa be saaelved during somets s e V tism and that, if des seasideratima is given to thnee items, the muetter plante proposed for the tremerisk ette ses be seestreeted with seasonable soeurance that they see be operated us,thout amese risk to the health and safety of the publia.
Simmerely yours, Original signed by Stephen H. Hanauer Stephen R. Esmauer theirman mra rensas - =
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letter from Caro 11as Faust and Lid t Canyany, dated July 26, 19688 Licemos App 11satten; Telumme 1, II, and 111 et pse18minary saasty Amstysis heport 2.
Caseline Power and Lid t Canymsys
- No. 1 to Limonee Application, dated September 3,1968; Supp1meest Be.1 to psAR 3.
Carolina Pseer and Lipt Campany a=,ma===* No. I to ta Appliaattaa, ested ur 6, 1968 4.
Letter from Carellas Pouer and Lidt Campaar, deced January 17, 1969; W No. 3 to 18*===
Applisatians supp1====e No. 2 to pSAR S.
Carolina Power and Lidt Campearg
- Re. 4 to Lam App 11satlama deted Jamsary 27, 1969; suppiammet me. 3 to pt&R 6.
Letter from Caro 11as Power and Light Company, ested Jammary 31, 19694 w me. S to Lamenen Applisatama 7.
Letter from Carottan Power and Lidt Company, dated Mareb 12, 19695 Amema==.* me. 4 to Lisease App 11aations supplemmat No. 4 to pa&R 8.
1stter from Caro 11ms power and Light Coupear, dated may 2,1969
-* no. 7 te timesse apetisatimes sepptement No. s te taak (f
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3 ACRS TECIDUCAL STAFF
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I BWR SAFETY / RELIEF VALVE (SRV) FROBLEHS Attached for your information is an analysis by Mr. Etherington of the safety relief valve (SRV) problemas experienced by various snodels of BWRs.
John C. McKinley Attachment KD /% y c.c 2;%4.Z;z,4ZE~ 6.Zr. )
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
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g WASHING TON. D. C. 20555 December 6, 1977 ACRS MEMBERS ACRS TECHNICAL STAFF BWR SAFETY / RELIEF VALVE (SRV) PROBLEMS Attached for your information is en analysis by Mr. Etherington of the safety relief valve (SRV) problems experienced by various models of BWRs.
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n C. McKinley Attachment 1
BWR SAFETY RELIEF VALVE (SRV) PROBLEMS Operating experience and studies have revealed serious problems relating to the performance of suppression pools. The problems are primarily associated with unanticipated hydrodynamic loading on pool internals (including the SRV piping and its anchoring), and on the pool boundary and the pool foundations. During its November,.1977 meeting, the Committee was briefed on a new problem arising from a second opening of several SRVs following the initial opening in response to certain anticipated transients.
The problems of SRV discharge and the larger problems of LOCA pressure suppression have been the subject of intensive study by GE and the utilities--there are already at least 2-3 feet cf topical reports and other documentation on this subject, and the work is continuing on a large scale.
This memo briefly reviews the SRV problems.
I.
The Piping I
In the Mark I (" light-bulb and torus") design, the steam dis-charge piping from each SRV is routed through the drywell, down one of the main vent pipes connecting the drywell tc the torus, out l
through the side of the vent pipe within the torus, and down into l
the suppression pool to the submerged discharge which, in the currently recommended design, is centrally located near the bottom of the suppression chamber (Fig I).
In the larger units, the pipe is typically 10 in. Sch. 40, and the length is over 100 ft., with many changes in direction.
In the Mark II ("over-and-under") and Mark III plants, the piping is 10 or 12 in. Sch. 40 and is more direct (Fig. 2, Mark III).
Ramshead (Fig 1) - As a result of early problems with a straight pipe, the ramshead, a shaped tee at the end of the vertical submerged pipe, was adopted as the standard design for Mark I plants and is used in some Mark II plants.
The Ouencher (Fig. 2) - The quencher is an improved design used in Mark III plants, some Mark II plants, and one Mark I plant that has a thinner than usual torus shell. The vertical pipe terminates in an enlarged head from which four pipes about 4 ft. long extend at right angles in a horizontal plane.
The pipes are capped at the ends, and many small holes are drilled in the pipes to discharge the fluids.
The quencher design has been standardized and is provided with a pedestal to be secured to the bottom of the suppression pool.
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' T-quencher - CE Ictter of Nov. 4, 1977 to V. Stello describes plans for instrumented tests on a "T-quencher" at Monticello under the Long Term Program of the Mark I Owners Group. The T-quencher is a ramshead with each of the two branches extended longitudinally along the torus as a 12 in. diameter closed-end perforated pipe, end-to-end length of the assembly 18 ft. 10 in.
The quencher is securely fastened to a 14 in. H beam spanning a 22 0 segment of the 16-segment torus.
i Vacuum breakers - All systems have a vacuum breaker to prevent water from rising in the pipe above pool level when steam condenses after the closing of the SRV.
The current Mark III design has two vacuum breakers.
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- 2. History Failure of baffles and other internals - On November 16, 1971, eleven of 72 anti-slosh baffles at Monticello were found displaced, and damage to other torus internals was also observed. The failures were attributed to vibration during SRV operation (or possibly during HPCI steam turbine operation).
Baffle failures were experienced in several other early Mark I plants. The baffles were installed because tests at Bodega Bay had shown them necessary to suppress surges associated with azimuthal pressure waves.
GE analysis and recommendation - Based on further' analysis, GE recommended, for all Mark I plants, extension of the SRV piping towards the bottom of the torus (as shown in Fig.1), and addition of a rams-head tee at each discharge pipe to prevent impingement on the shell.
Quad Cities Unit 2 tests - In the Quad Cities units (as in some other plants), the ECCS pump-suction ring header outside the torus is supported from the torus by hangers (Fig.1). During a pressure relief transient, bolts on one of the hangers failed leaving the ring header sagging. After the failure, the GE recommendations for deeper submergence and addition of ramshead tees were implemented at Quad Cities, and instrumented tests were made on the unit between September 26, and October 7, 1972, to confirm the GE analysis. A plant-specific mathematical model was developed for comparison with test measurements, which included traces of pressure, deflection, and strain at selected locations. The behavior was of the general character predicted.
Although it was concluded that the hange't failure was caused primarily by improper installation, stronger hangers were installed.
It was concluded that the maximum dynamic stress on the torus wall (7.77 ksi) was acceptable.
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. Monticello tests - From June 16 to June 18, 1976, a series of tests were conducted at Monticello (NEDC-21465, December 1976) to provide a data base for verifying and improving the analytical models.
The total pipe length is 103 ft. and the submerged length is 13.5 ft.--
the submerged depth is less.
Transient measurements included pressure; temperature; water level; and structural response (strain and displacement) of the torus, internals, torus supports, and base mat.
Test runs included single valve operation, multiple valve operation simultaneously and in sequence, and repeated operation of the same SRV.
Problem reported at November 1977 ACRS meeting (Summaries October 21 by W. F. Kane and October 27 by R. K. Major). The test programs show that, if an SRV opens a second time shortly after it has functioned to reduce pressure, the second pressure transient in the wet well is more severe than the first.
In the event of complete isolation of the condenser from the reactor (MSLIV closure, or turbine trip with failure of the steam bypass valves to openh all 19 SRVs open--this is a design condition.
It was further anticipated, and was acceptable, that a secondary rise in pressure due to decay heat would cause one SRV to reopen subsequently.
However, it is now calculated by the SAFE code that, in the BWR-6 Mark III design, as many as eleven valves could reopen; in the current design, the remaining eight (ADS) valves are prevented from opening a second time in order to conserve accumulator air in case these valves should be called upon to operate in the ADS mode.
A second opening of eleven valves results in a greater pressure transient than that caused by the first opening of nineteen valves.
A similar condition exists in the Mark II plants (two to ten valves may reopen). With one exception, the Mark I torus has sufficient design margin to tolerate a second opening of all eight valves; moreover, extensive operating experience shows that only one to three valves reopen.
GE considers its calculations to be very conservative, and believes that the problem is not serious. The problem can be solved by changes in the RSV control circuitry, e.g., by designing so that selected valves, when actuated, vill be held open to a lower pressure than the initial actuation pressure, and will reopen the second time at the lower pressure; or by inhibiting reopening of some of the valves. Regrouping of valve settings is also re-commended, to provide better distribution around the torus of
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. valves set to open at the same pressure. These solutions have not been accepted by the NRC Staff, and GE is to make a future presentation.
3.
The Phenomena Four phases of the pressure relief phenomena are considered.
The numbers are, except as noted, for the 10 in. Sch. 40 pipe rams-head " base case" for a Mark II plant (NED0-21061, Rev. 1, Sept.
1975). Pipe length 150 ft. (total), submerged length (vertical) 20 ft.
Valve opening and pressure buildup in the pipe. The inertia of the water in the subnerged part of the pipe causes a high pressure to build up in the pipe, somewhat mitigated by condensation of steam on the cold walls of the pipe.
Reactor pressure 1150 psia Valve opening time 0.05 see Rated steam flow 250 lb/sec Maximum pipe pressure 553 psia The valve opening time is the full stroke time, exclusive of delay time.
Shorter and longer times are reported, e.g., 0.03 sec.
in the Monticello tests, 0.15 sec. design opening time in a BWR6/
Mark III plant. The opening time in the spring operated safety mode is longer (e.g., 0.3 sec. in the BWR6 plant) than in the power operated relief mode.
The pipe pressure must not exceed the design back pressure of the SRV. A valve design limit of 625 psid is mentioned.
Water clearing. The steam and air pressure in the pipe causes ejection of water at rapidly increasing velocity.
Maximum water velocity 446 fps Water clearing time 0.279 sec.
The pressure and reactica forces, and thermal expansion, cause loading on the SRV discharge Jine and its anchoring.
Air clearing and bubble dynamics. After ejection of the water, the air follows and forms a high-pressure bubble about 4 ft. from each discharge opening. The bubble expands rapidly, and inertia effects cause over-expansion with subsequent recompression of the bubble. Alternating expansion and compression continues as the bubble rises to the surface, causing pressure waves in the water.
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. Maximum air velocity (sonic) 1159 fps Bubble radius and pressure before expansion
- 0.84 ft/134 psig Bubble radius and pressure after expansion *
(a) near ramshead 1.93 ft/-10 psig (b) near pool surface 2.01 ft/-11 psig Bubble period / frequency 0.1271 sec/7.7 Hz
- These are examples from a different case.
The consequences are oscillatory pressure loading on submerged surfaces of the pool boundary and internal structures, including the SRV piping, and loading of external suppcrts and attachments.
Structural analysis gives peak stresses which, together with the number of cycles per event and the estimated number of events in the plant life, will permit evaluation of the fatigue utilization factor.
Steam guenching. During steady state steam quenching, the steam jet may be about ten pipe diameters long and two pipe diameters thick.
Since the steam is at elevated temperature, thermal stresses would be set up in any surfaces exposed directly to the jet. No undesirable effects have been associated with this phase of normal pressure relief.
4.
Analysis Analysis and reactor confirmatory tests on the Mark I plants identify discharge pipe air volume, submerged length, and pool temperature as important parameters. The diccharge pipe air volume has an optimum value--a larger volume leads to a larger bubble, but a smaller volume leads to a higher pressure buildup before clearing.
Deep submergence promotes pool water mixing, but accentuates pressure buildup and increases transient loads. Pool temperature enters into the thermodynamics of bubble behavior.
Mark I.
The analysis for the Mark I plant, reported in several documents (e.g., NEDO 10859, April 1973; NED0-209421, May 1975), is based on classical equations for the conservation of mass, momentum, and energy, applied separately to the piping discharge transients and the pool dynamics. The pool dynamic analysis combines the classical equations with a development of the Rayleigh equation for an oscillatory spherical bubble--the output includes formulas for frequency and water pressure as a function of distance from the source.
The results of the Quad Cities and Monticello tests were incorporated into the analysis.
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. In Mark II plants, the closer spacing of the discharge pipes leads to greater interaction among the discharge sources, and the analysis has been extended to take this into account by applying the method of images, with an array of sources and sinks. The important parameters are identified, and formulas for " influence f actors" (mostly proprietary) are given to quantify the effect of departure from the reference design. A similar analysis is provided for the quencher design.
Superimposed effect of multiple valve operation. The pressure in the pool varies approximately as the reciprocal of distance from the bubble center, as would be the case, theoretically, in an inf3 nite pool.
The total pressure on a surface from operation of two or more valves is less than the sum of the pressure from the individual valves, because of pressure wave interference, design and random differences of valve opening time, and random differences of bubble phase.
The Mark II analysis takes these factors into account, and GEN 0394 (September 28, 1977) compares the analytic model with test data.
Consecutive SRV operation. No quantitative explanation is available to account for the observation that when an SRV functions repeatedly, the subsequent pressure transients in the pool are more severe than the first.
Conditions during the subsequent transients differ from those of the first transient in the following respects:
1.
The pool temperature is higher, which adversely affects the bubble thermodynamics.
2.
The piping above the waterline is hot, which leads to a higher maximum air pressure by minimizing steam condensation during the initial stage of clearing; on the other hand the mass of air is less in a hot pipe.
3.
During recovery from the transient, the vacuum breakers fail to prevent water in the pipe from rising temporarily above the pool level.
The relative importance of these factors is not clear, but it is concluded that, at least in the ramshead design, the most important factor is the rise of water in the pipe above pool level (NEDO 21061, page 3.1-$.
In the Mark III design, two vacuum breakers are provided in each lire, one close to the SRV and the other at a safe distance above the maximum transient water level in the pipe.
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