ML20062E165

From kanterella
Jump to navigation Jump to search
Board Notification 82-71:informs That NRC Did Not Rely on Feed & Bleed Cooling to Protect Core & Forwards Info Supporting Info
ML20062E165
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 07/27/1982
From: Lainas G
Office of Nuclear Reactor Regulation
To:
NRC ATOMIC SAFETY & LICENSING APPEAL PANEL (ASLAP)
References
TASK-AS, TASK-BN-82-71 BN--82-71, BN-82-71, NUDOCS 8208060369
Download: ML20062E165 (20)
v • d • e


Text

,,

  • /I *yf g 7 E32 DCWO lW

'i d D

DEisenhut JHard PPAS

@ M UTIONckel ECase File) MJollens ten-4 fWilliams bD("" Mem File MCutchin AE0D GLainas JStolz Gray File ORB #4 Rd9 RJacobs ACRS-15 H0rnstein Docket flo. 50-289 RIngran JGray EBlackwood BSnyder MJumber JSci nto-2 EChristenbury Yy 0

OIE RHartfield, MPA f1E!!0RAllDUtt FOR: Atomic Safety and Licensing Appeal Board for Tf!I-l Restart FR0ft: Gus C. Lainas. Assistant Director for Operating Reactors, Division of Licensing, flRR

SUBJECT:

BOARD fl0TIFICATI0ft (Bit-82-71) - THI-l RESTART llEARIf.G In response to a memorandum from the Director of flRR dated April 29, 1982 concerning staff reliance on " feed and bleed" cooling in the THI-l mstart proceeding, the Staff has prepamd the enclosed report. To stenar12e, the IRC Staff did not rely on " feed and bleed" cooling to protect the core at T!!I-1. This position was made clear to the BoarJ. Babcock and Wilcox performed feed and bleed analyses for the development of inadequate core cooling procedures. Such procedures would be utilized as defense in depth for events beyond the design basis. These procedures instruct the operator to establish and naintain feed and bleed cooling following a complete loss of heat sink until feediater can be restored.

y . . , :a T1 Gus C. Lainas, Assistant Director for Operating Reactors Division of Licensing

Enclosure:

Report on flRC Staff Position ,

on Feed t4 Bleed Cooling cc w/ enclosure:

See next page 1

l 8208060369 820727 PDR ADOCK 05000289 P PDR ornce> .0RBf :D ,,,, ,,C ,0,RB#4,,:pk, AD ,, ,

pk , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , ,

RJa cab JStolz IM GCL nas

$URNAME) , , , , , , , , , , , , , , , , , , , ,,, , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , ,, , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , ,

[.lAh',.,.2,, lg,,/,g2 ,, ,,, Z/,1r ,@2 ,,, ,, , , , , , , , , , , , , ,,,,,,,,, ,, , , , , , , , , , , , , , ,

oare > , , , , , , , , , , , , , , , , , ,

tac rosu aia now sacu om OFFICIAL RECORD COPY owa mi-mm 4

r GPU Nuclear Corporation 50-289, TMI-l l

Mr. R. J. Toole Jordan D. Cunningham, Esq. t Manager, TMI-l Fox, Farr and Cunningham GPU Nuclear Corporation 2320 North 2nd Street P. O. Box 48G Harrisburg, Pennsylvania 17110 Middletown, Pennsylvania 17057 ,

Ms. Louise Bradford

  • Atomic Safety and Licensir.g Appeal Board TMIA U. S. Nuclear Regulatory Commission 1011 Green Street Washington, D. C. 20555 Harrisburg, Pennsylvania 17102
  • Atomic Safety and Licensing Board Panel Ms. Marjorie M. Aamodt U. S. Nuclear Regulatory Commission R. D. #5 Washington, D. C. 20555 Coatesville, Pennsylvania 19320
  • Docketing and Service Section Earl B. Hoffman U. S. Nuclear Regulatory Commission Dauphin County Commissioner Washington, D. C. 20555 Dauphin County Courthouse Front and Market Streets Chauncey Kepford Harrisburg, Pennsylvania 17101 Judith H. Johnsrud Envimnmental Coalition on Nuclear Power Union of Concerned Scientists 433 Orlando Avenue c/o - Harmon & Weiss State College, Pennsylvania 16801 1725 I. Street, N. W.

Suite 506

J. B. Lieberman, Esq. Dupont Circle Building, Suite 1101 Berlock, Israel & Lieberman Washington, D. C. 20036 26 Broadway New York, New York 10004 Mr. Henry D. Hukill, Vice President Dr. Walter H. Jordan -

881 W. Outer Drive op Oak Ridge, Tennessee 37830 P. O. B 480 Middletown, Pennsylvania 17057 Dr. Linda W. Little 5000 Hermitage Drive Raleigh, North Carolina 27612 Ms. Gail P. Bradford Anti-Nuclear Group Representing York 245 W. Philadelphia Street York, Pennsylvania 17404 John Levin, Esq.

Pennsylvania Public Utilities Commission Box 3265 Harrisburg, Pennsylvania 17120 m

GPU Nuclear Corporation -22 $ n::ral Counsel Federal Emergency Managem:nt Agency Mr. Thomas Gerusky ATTN: Docket Clerk Bureau of Radiation Protection 1725 I Street, NW Department of Environmental Resources Washington, DC 20472 P. O. Box 2063 Harrisburg, Pennsylvania 17120 Karin W. Carter, Esq.

505 Executive House Professor Gary L. ftilhollin P. O. Box 2357 1815 Jefferson Street Harrisburg, Pennsylvania 17120 Madison, Wisconsin 53711 York College of Pennsylvania Country Club Road York, Pennsylvania 17405 G. F. Trowbridge, Esq. Dauphin County Office Emergency Shaw, Pittman, Potts & Trowbridge Preparedness.

1800 M Street, N.W. Court House, Room 7 Washington, D. C. 20036 Front & Market Streets Harrisburg, Pennsylvania 17101 Mr. E. G. Wallace Licensing Manager Depart:,ent of Environmental Resources GPU Nuclear Corporation ATTN: Director, Office of Radiological 100 Interpace Parkway Health Parsippany, New Jersey 07054 P. O. Box 2063 Harrisburg, Pennsylvania 17105 William S. Jordan, III, Esq. Ms. Lennie Prough Harmon & Weiss 172S I Street, W, Suite 506 U. S. N. R. C. - T!!! Site Washington, DC 20006 P. O. Box 311 Middletown, Pennsylvania 17057 Ms. Virginia Southard, Chairman Citizens for a Safe Environment 264 Walton Street Lemoyne, Pennsylvania 17043 Mr. Robert B. Borsum Babcock & Wilcox Government Publications Section Nuclear Power Generation Division State Library of Pennsylvania Suite 220, 7910 Woodmont Avenue Box 1601 (Education Building) Bethesda, Maryland 20814 Harrisburg, Pennsylvania 17126

  • Ivan W. Smitn, Esq.

Mr. David D. Maxwell, Chairman Atomic Safety & Licensing Board Panel Board of Supervisors U.S. Nuclear Regulatory Commission Londonderry Township Washington, D. C. 20555 RFD#1 - Geyers Church Road Middletown, Pennsylvania 17057 Mr. C. W. Smyth -

Supervisor of Licensing THI-1 GPU Nuclear Corporation Regional Radiation Representative P. O. Box 480 EPA Reaion III Middletown, Pennsylvania 17057 Curtis Building (Sixth Floor) 6th and Walnut Streets Philadelphia, Pennsylvania 19106 Mr. Richard Conte Governor's Office of State Planning-Senior Resident Inspector (TMI-1) and Development U.S.N.R.C. ATTN: Coordinator, Pennsylvania P. O. Box 311 State Clearinghouse M P. O. Box 1323

@iddletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120

/

'GPULNuclear Corporation 50-289, TMI-l 1

i

  • Judge John H. Buck

)

Atomic Safety and Licensing ' Appeal Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555

  • Judge Christine N. Kohl Atomic Safety and Licensing Appeal  ;

Board Panel '

U.S. Nuclear Regulatory Commission Washington, D.C. 20555 I

  • Judge Reginald L. Gotchy j Atomic Safety and Licensing Appeal Board U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Ilr. Ronald C. Haynes, kegional Administrator U. S. N. R. C., Region I G31 Park Avenue
King of Prussia, Pennsylvania 19406 Board of Directors P.A.N.E.

P. O. Box 268 Middletown, Pennsylvania 17057'

  • Dr. Lawrence R. Quarles i

Atomic Safety and Licensing Appeal Board

! U. S. Nuclear Regulatory Commission l Mail Stop EW-529 Washington, D. C. 20555 l

i ll

.,_ _ .._._. _ .__.- _ . _ - _ _._ _. - _. _ _ __ _ . _ ~ .

1 . ,

l l

DISTRIBUTION OF BOARD NOTIFICATION l

Three Mile Island, Unit 1/COMM Docket No. 50-289 (Restart) ACRS Members Leonard Bickwit, Esq. Dr. Robert C. Axtmann Mr. Samuel J. Chilk Mr. Myer Bender

Dr. Max W. Carbon Mr. Jesse C. Ebersole Mr. Harold Etherington Dr. William Kerr Dr. Harold W. Lewis Dr. J. Carson Mark Mr. William M. Mathis Dr. Dade W. Moeller Dr. David Okrent Dr. Milton S. Plesset Mr. Jeremiah J. Ray Dr. Paul C. Shewmon Robert Adler, Esq. Dr. Chester P. Siess Ms. Frieda Berryhill Mr. David A. Waro Mr. Allen R. Carter

! Hon. Mark Cohen David E. Cole, Esq.

Mr. Henry D. Hukill Ms. Jane Lee Mr. Marvin I. L'ewis Michael McBridge, Esq.

Ms. Gail Phelps Ms. Ellyn R. Weiss 21

__ - _ _ . _ - - .~. - - - - - - . - -- - - . . -- - - - - _ - - .. ~

J REPORT ON NRC STAFF POSITION ON FELD AND., BLEED COOLING l

Item 1 A description of the staff position at the THI-1 restart hearing-on the role of " feed and bleed" during a SBLOCA l

l RESPONSE The staff's position at the hearing was that feed and bleed r cooling is not relied on for.he'at removal. This position was l made clear to the ASLB in the TMI-1 restart hearing in (1)

, written testmony by NRC staff witness J. Wermiel and (2) oral-

! testimony of W. Jensen as follows.

I (1) Written Testimony of J. Wermiel in Response to Board i

! Question 6: Question 61. Will the reliability of the emergency feedwater system be greatly improved upon conversion to safety-grade, and is it the licensee's and staff's position that the improvement is enough such that the feed-and-bleed backup is not required. -

t l

e I

l (Witness Wermiel) l Response: Based on knowledge of the improvement in ,

reliability gained by eliminating first order failure sources, I

l it is the staff's judgment that the reliability of the l

emergency feedwater system will be improved once the fully

safety-grade system is installed. The single failure problem 1 associated with integrated control system /non-nuclear i

s- - - - s,+s,w-.-,,e- r w p -

y%e1--- ,w, e .-.*--m . - , .v--, 4.~.,-,- , -,, -7,, r -- =n- -g v,, +- ** w +r -

2 i

instrumentation described in the response to 6a and b above will-be el.iminated. In., addition, various other hardware, procedural and administrative improvements as identified in the TMI-1 Restart SER, NUREG-0680 under Order Item la should enhance emergency feedwater system reliability. However, a quantitative reassessment of the reliability of the fully safety-grade EFW system has not been performed. The feed-and-bleed back-up is not required by the staff and, therefore, need not meet all. requirements of a. safety system; However, it is recognized as additional defense in depth for providing core cooling in the very unlikely event that emergency feedwater is lost, and the HPI pumps and primary safety valves which comprise the feed and b.1_eed mode are required to be available by Technical Specifications.

(2) Oral Testminony of W. Jensen Regarding UCS Contepfions 1 l and 2 i

(Dr. Jordan) I would address the question then directly to Mr.

Jensen. Did I misstate what you said? Do you believe that i the high pressure injection system is important in that it not only supplies emergency cooling inventory but it also removes heat in the feed and bleed mode? That that is an important safety feature?

! (The Witness) The high pressure injection system is an i

important safety feature for making up the coolant lost from a

{

t

3 small break LOCA. The NRC does not rely on this system for heat-removal in the feed and bleed mode by which core decay heat would be forced through the safety valve or the PORV.

Instead, we rely on the heat removal from the einergency feedwater system.

(Dr. Jordan) Okay. That's fine.

~

(fis. Weiss) If I can refer, Dr. Jordan, I think the exact question you are asking is answered on page 9 of the staff testimony in response to Board question number 6. I was going to read the sentence to you. (Wermiel testimony above)

The feed and bleed back up is not required by the staff and therefore need not meet all the requirements of the safety system. It's just simply a direct quote.

(Dr. Jordan) Yes. I remember that and thank you for pointing that out. I think that clears up the matter."

l Item 2 An interpretation of the TMI-1 Licensing Board decision regarding the need for reliable and effective " feed and bleed" during SBLOCA l

l

[ RESP 0fiSE

4 There is an interest in whether the ASLB accepted the staff position on the relianc,e to be placed on feed and bleed cooling. We believe that the ASLB did not accept our '

position, regarding emergency feedwater reliability, as shown in the following excerpts from its decis. ion. We believe however that the board did not err in declining to find that additional modifications to the emergency feedwater system are necessary at TM1-1 prior to restart.*

Page 224 of the TMI-1 Licensing Board decision acknowledges the NRC Staff position (see Item 1 above) by noting that:

The Staff's position is that the loss of em,ergency feedwater o

following a main feedwater transient is not an accident which must be protected against with safety-grade equipment."

To us, this observation by the ASLB says that our position in Item 1 above was understood by the Board. At Page 242 of the decision the Board goes on to point to a precedent ruling made by the St. Lucie-2 Appeal Board for requesting additional reliability numbers from the staff. The TMI-1 Board noted that:

I

tlRC response to UCS' exceptions to the PID, filed with the Appeal Board in the TMI-1 Restart proceeding May 20, 1982.

l t

4

[

_ - - . . ,.m, _ ..-,,. . -~-----e - - - - ' - - - -e- - " "

e 5

"The (St. Lucie) Appeals Board decided that measures were required to mitigate such an event should it occur. We believe that similar measures are necessary at IMI-1; that the reliability of the EFW system has not been demonstrated to be adequate by itself. However, the EFW systen is backed up by the high pressure injection system, so that in the event of failure of the EFW system the core can be cooled by feed and bleed while repairs are being made to the EFW system."

We conclude from this statement that the TMI-1 Board has relied upon the availability of feed and bleed in reaching its finding that the TMI-1 design is acceptable. The question then is how the Board reached this conclusi,on in light of the Staff position (Item 1 above). The answer is summarized on' page 250 of the TMI-1 Board decision where the Board states:

"We have relied on the staff figures on reliability of the EFW system and our own estimates (emphasis added) of the adequacy of the feed-and-bleed backup to arrive at our conclusion that I

the core is adequately protected from a loss of main feedwater transient, the dominant challenge to the EFW system."

    • Complete loss of all AC power including both diesel generators.

i -

~6 We conclude that the Licensing Board reached the same conclusionas-thestaffjtheTMI-1designsatisfiesthe

! Commission's regulations), although the board's basis for the

' ~

conclusion is different. The basis for the staff position ~is

. summarized in Question 3 below. We have studied the Licensing Board decision to understand the basis for its conclusion. At paragraph 1056 we find the following:

i "Since the EFW System is backed"up by a safety grade HPI,

] designed to protect the core in the event of a smal1 break i '

l LOCA, wc believe we can conservatively assume an additional safety factor of 100, or an overall probability of failure to l protect the core of about 10-6/yr. Lacking,any demonstration that the above failure probabilities are grossly in error, we 4

conclude that the EFW system, as modified, will, with the HPI ,

backup, adequately protect the health and safety of the j public."

During the TMI-1 hearing, the NRC Staff did not provide any detailed discussion, for or against, the above Licensing Board i assessment. We do not have sufficient information regarding the uncertainties associated with of feed and bleed cooling to credit it with a 100 fold reduction in the probability of core 4

l mel t.

t Item 3a A detailed explanation of the staff's technical basis for its position on " feed and bleed" at TMI-1.

! 7

RESPONSE

i j

4 It was the Staff's position during the TMI-1 hearing that the i ~

emergency feedwater (EFW) system is required to be available for decay heat removal in feedwater transients and certain small break loss-of-coolant accidents without feedwater. We

also noted that should EFW be initially unavailable, there is at least 20 minutes time available to take action to establish-l EFW flow prior to uncovering, oEt ' he core following a loss o'f main feedwater or certain small break loss of coolant
accidents. The TMI-1 EFW system will, at the time of restart, l .

t meet the Commission's requirements for safety related j equipment, in the event of small break LOCA_and/or loss of.

lr main feedwater if credit for operator action is given (to initiate the system) within 20 minutes. The TMI-1 EFW system i

will be fully automatic for these events by the first refueling outage after restart. The staff recognizes that a feed and bleed capability exists at TMI-1 to provide additional defense in depth for decay heat removal 'should EFW fail. The inadequate core cooling procedures at TMI incorporate the feed and bleed process. Operators are trained in the use of these procedures at TMI-1 and feed and bleed is covered in the scope of. 0LB examinations of the TMI operators.

It is usually covered in the simulator portion of the i examination. Safety grade equipment to accomplish feed and i

bleed backup to EFW in the event of a complete loss of all i

feedwater is not required to be included within the design l

8 basis since.the EFW system at the' time of restart is sufficiently reliable t,o make a postulated loss of EFW system acceptably low.

Item 3b Clarify the difference between-the " feed and bleed" mode of cooling and thc " boiler / condenser" mode of cooling

RESPONSE

For small breaks below a 'certain size, the break area is not large enough to relieve all the-energy generated by decay-heat. For this condition, heat transfer through the steam generator is the preferred method of providing additional required energy removal capability. To accomplish this, emergency or auxiliary feedwater systems must be operating.

Since the reactor coolant pumps are tripped for most small breaks, coolant flow through the core is by natural circulation. Feed & Bleed is a method by which' decay heat is removed from the primary system if no feedwa'ter'were available so that natural circulation did not occur. The

" boiler / condenser" mode of cooling is one of three modes of natural circulation cooling discussed below. Each mode represents a progressively degraded condition of the primary system in terms of system inventory. Thus it is possible for some small break scenarios to experience all three modes of natural circulation heat removal. In small break LOCA M

i 9

calculations by B&W temporary' interruption of all modes of

natural circulation was, predicted however, inventory loss in these three modes is not sufficient to cause extended core uncovery and fuel damage. It is not necessary that the primary system be refilled folllowing a LOCA in order to adequately cool the core. Analyses by B&W indicate that adequate decay heat can be removed under any of the following

, three natural circulation modes.

.j

1. Single phase - In this mode the entire primary system remains in a subcooled liquid state. Core flow is maintained solely by density differences between hot and cold liquid.

4 l 2. Two phase continuous - This mode is similar to mode 1 i

except that the hot side is at saturation and at low steam

{

quality. Bubbles are formed in the upper portion of the core and are swept, as part of a continuous two phase mixture, into j the steam ger.erator and condensed. During this time, some of l the steam generated in the core will rise into the upper head and accumulate there as a single large bubble. For B&W plants this heat removal mode will persist until the liquid level drops below the hot leg U-bend.

  • B&W report " Evaluation of Transient Behavior and Small Reactor Calant

. System Breaks in the 177-FA Plants" May 7, 1979.

I i

10

3. Boiler / Condenser - When the hot leg U-bend is voided, liquid will not be carried into the steam generator. -However, when sufficient steam has accumulated from boiling in the core suchthatacondensingsurfaceisexposedwithintheste$m l generator tubes, heat will be removed by steam condensation on-1

~

the tube' walls. This method of heat removal is referred to'as boiler / condenser. Thus a period will exist between formation of a bubble in the hot leg U-bends when mode 2 natural circulation is lost, and the, uncovering ~ of th'e steam generator condensings ' urface, during which no natural circulation would exist in B&W plants. The condensing surface is at a higher ele'vation than the core so that boiler / condenser natural circulation will be established in the even,t of a small break

I LOCA before the core could be uncovered. Boiler condenser natural circulation was. demonstrated to be effective in LOFT
  • NUREG CR-1570 " Experimental Data Report for LOFT Nuclear Small Break Experiment L-3-7", August 1980. ,

\

    • EGG-SEMI-5507 " Quick Look Report for Semiscale Mod-2A Test S-NC-2,"

July 1981.

I

~

1 l

i I

11 If heat removal through the steam generator cannot be achieved due to los.s of all feedwater (an event not required to be considered as a part of the design basis), " feed and bleed" can be used as an alternate heat removal nethod. The procedure involves energy removal by venting hot water and/or steam through the primary system PORVs and/or safety valves (bleeding), and replacing the vented coolant with cold HPI water (feeding).

Item 3c Assessment of Current Status and Existing Information on

" Feed and Bleed" RESPONSE _

t As you recall, in a recent communication to Dr. Henry Myers we noted that for a small break LOCA which is subsequently isolated, a phenomenon similar to " feed and bleed" might ultimately occur as the means of decay heat removal if steam bubbles were trapped.at the top of hot legs and did not rapidly condense even if emergency feedwater were available.

This method of heat removal from the primary system might occur if the core were sufficiently cooled so that decay heat no longer boiled the incoming HPI water but forced it through

  • The term "similar" is used, since in this case feedwater to the secondary side of the steam generator is assumed available, and no operator actions are assuned to initiate decay heat removal via the safety valves.

i 12 the safety valves as liquid. If boiling occurred.in the core, the steam production would act to increase the bubble size in the hot-leg U-bends.

If the hot leg bubble size increased sufficiently, a condensing surface  !

a on the steam generator tubes would be exposed. This would establish natural circulation in the boiler / condenser mode.

1 The bubbles could not expand sufficiently to uncover the core or to exhaust steam out of the pressurizer since the secondary system water j level in the steam generators would be;above the core an.d the

pressurizer surge line entry elevation. Although our study of this scenario is recent and was not discussed during the TMI-1 hearing, no additional staff reliance on feed and bleed should be implied since if the feed and bleed process discussed above were insufficient to remove decay host, sufficient coolant loss through the safety and relief valves

! would eventually reestablish natural circulation in the boiler / condenser mode. The letter to Dr. Myers is attached for further information on these recent developments.

All three PWR suppliers are developing emergency pro ~cedure guidance to licensees on how to use equipment to perform " feed and bleed" operations i .

as a backup method of heat removal if all measures for feeding steam generators are lost. It is important to stress that at this time " feed and bleed" is not a preferred method of decay heat removal. The equipment used for feed and bleed operation was not designed' for that i purpose. Feed and bleed is only one possible emergency alternative for i

primary system heat removal for events beyond the design basis. All PWRs have in their proposed emergency guidelines, methods for use of

- = _ - - _ _ _ .

13 decay heat removal schemes other than the design basis equipment. In particular, guidance is given to provide alternate sources of secondary cooling if main and auxiliary feedwater are unavailable (e.g., by depressurizing the secondary system and activating the coridensate pumps). Operators would resort to feed and bleed only if no source of water is available to feed the steam generators. The NRC has no design requirements for these other alternate schemes, just as we have none for the " feed and bleed" capability. What is required for the design basis

~

is a reliable auxiliary feedwat'er system to remove decay. heat until the RHR system can be activated to ultimately achieve cold shutdown.

However, to provide defense in . depth, feed and bleed procedIJral instructions should be available to operators because the capability to i feed and bleed exists. _

l As to the technical performance of " feed and bleed," we know it depends i on the HPI pump performance characteristics, the PORV relieving capacity, and the plant power to volume ratio. Analyses have been conducted by all three PWR suppliers to examine " feed and bleed" capability for their designs. Also, NRC contractors at LANL and INEL have analyzed " feed and bleed" with the computer codes TRAC and RELAP.

As noted previcasly, a B&W calculation for a TMI class plant showed that

" feed and bleed" was an effective heat removal method even if no credit is taken for PORV actuation. This is because most B&W plants have HPI pumps with a very high shutoff head, and enough energy can be relieved at high pressure through the safety valves. It is important to note that f

the assessment of " feed and bleed" rests almost exclusively on analysis.

14 Analytical uncertainties related to such phenomena as non-equilibrium thermodynamics, bubble formation and repressurization caution against taking too much credit for analytical predictions of system behavior.

One LOFT experiment (L9-1/L3-3) explored " feed and bleed" in a limited way. Af ter a simulated loss of feedwater, the PORV was latched open to allow depressurization. The results showed that depressurization to the HPI actuation point did indeed occur. However, HPI actuation was purposely not allowed to occur so that.other accident $itigation schemes could be explored.

Item 4 Recommendations for Future Action It is desirable to improve the experimental basis for understanding system behavior during " feed and bleed." This should improve the guidance in emergency procedures and training that is being developed under Task I.C.1 of NUREG-0737. To accomplish this, we are exploring ways to expand the current Semiscale test series to include " feed and bleed" experimental data. We expect shortly to issue a request to RES which will include these proposals.

The current Semiscale configuration cannot simulate the unique features of the B&W NSSS. You know from previous discussions that we have been trying to resolve the problem of uncertainties for the B&W analytical methods in predicting long tern LOCA recovery under Task II.K.3.30 NUREG-0737. We

i.

15 are investigating the unique features of the B&W design and the lack of integral systems data (see attached letter to B&W We will shortly transmit to all B&W owners our-owners).

conclusion that such data are required. The basis for this conclusion is the need for additional verification of some
aspects of.the thermal-hydraulic behavior during natural circulation cooling of the B&W design with feedwater available
during small break LOCAs, as well as uncertainty in the feed and bleed process. You will~ als'6 recall that the ACRS letter.

i of June, 1982 highlighted this problem for resolution prior to its concurrence on fu.ll power operation of Midland, a B&W I

reactor.

  • O 4

l i

j l

1 I

l l

7 I

l l . .

1 1

n' * *... v. I 3

<- . .- . . . , . .. .. a g

3 I:: ';;'ce cn S .rgy ;-nd thS Er.Virc. : t .

.--: u-.s . . .a v.c..,

  1. - - 7. .

. * . . . : o r.

. t :. n , a ......a r .. ...e.cers

. . . d I'.i t c s F: .:se of . r c. .ta tiv es

,. . , . , ..3.., s.. 3, . r . ,m. .,

.s

_.e. n e. . .t .oa. - ns In c:rly *.pril ;ou re u(' ted the NRC si:f f to cn . :it to a suic cnt in s

, . O . '. (19 of t he a:I Rc t u.rt .Sartial I:.i t u.1 t ccisitn. T'.e d ; c. .nt r- . 'l i r as :

  • if,' c [ . , : L . n . i . *, . i r e . t . . n , i4, t 11.1 :. . . pi . ;i. ! ; n ,, .

11

. i l 's f. , '. L e ! .' '.l. in t he hot 1. o . : i bi e i i .: ....

. d nj i t.m.

- v . ' . t '.e ; r . . r y y . . .a p r. . m ., i <, 1 : . . 4.J : .y i .; i . ir: og

.c

. :yI t.y ',n."

The cu. tat cf this state ent in paragrcph 619 is a discussion cf the recovtry fr:a a s . ail-break LOCA yhan ECCS irdettien ficu ot:eds treak fled so that the reactor syster is fiiling. The cist.:ssion ac'drcssed a 1:nion of Ccncerned Scia:ists (UCS) ccr.cern that fcr this condition a stcaa Eatble in the top of ti.e hot icg U.hends micht prevent the reastctlis!. , ant cf singic-phase r.=: ural circu 6 ation.

!'r. Ka.m.arer of NRC replied to your request in a letter of April 8,1932.

The response included the following ccc:ments:

' If a sice.r. bubble exists and pri;rary sysis pre <.sure is raised, the a

!<*t Lie will be w.: pressed i.nd thcre will l'e comie nsction. Th" con-i a:ntion e: curs baccuse as vou raisn .s.ysa H E.r =- u.r'.<.ve.en2' 'et _r-n

.are cross .,>elow sac uration. Condensation cast occur to rcech satura-tion condit1cns again. The bubble will not necessarily be completely cendensed. This depends on bubble size a,d pressure chance." (Sentence ~

wts underlined in your April 19, 1952 letter.)

Ir. c letter of April 19,1%2, you asked if the o.uote underlir.ed above frca our

'pril S,193?, response was correct. Specifically, you cues tiora< ". . . whether, in the case of a small break LOCA, the heat transfer would be sufficient to Icep the steam space from cxpandir.g into the core." The undcrlined qucte is r.ct correct tacause there cust be heat transfer to a heat sir.k to abscrb the heat of vaccrization is:Sciated v:ith the ccr:densation of steta, ar.d the Ecat transfer rate is F-f.ortEr.t.

AI!o, the sysic 1 tccp'crature does r.ot acbcily drop; rather, the 'iturati:.n

' e , crc;;re incrmes with incrcasir.g pr2ss:<re. e.:ver, while the cc:,6 ntation

f. e cf de het leg i.:Lble 5:ill affe:t the c . rail reccvury 'ehwicr, it n as r.ct, by i:< cl;,, c:vern : S c n cll e . meacy of 6 :cy et : .evcl '.r.N a< 0 1 t:- - h ( K A

. aj. t c'-  ;; .a:. "l. To 'e cira.ir. :: + ' >h r.o i'.r v . , n,. S.g

. 1 ;. i e , ,i , n '. .r. , rs.s iG ir . .y .. ';..

. 1 it r.o .

I i. , . c.

. , sci 11 . .:,y .

./ . , . 1r . a. ,. i .:.. .i. .

i .

w ,,.-

(~

~

e V

- w . - . . . . . ~ . - . - - . ... .~a... . . ...an...w. . . . - .

  • .. . e. ;. .
  • . . . . .s.. ., _ . . .-....'.: .

. ;., ..c.

. 2 L. , . . .:

.. = .

.e. > . ...c

_ . . . c . := . . . . . . . . . . . . . , -

.i.

. ~. ..; .t.t

. ..:e . . _ . . . . . , . .. . .

.... . . } ,. ....u... . ; . : . ....$.;. e . s

, . -,. ) c ,,, .

. i ?. . . ,2.

. . _ .,... A.

. ..,......:  :. ,..3c .C : :.

.s. a u. , r c. , .. O ... .. .u < ,' c. .eos.: il c

a .- . . . .# .Q.c. a. C .. .,_2 2 i.. 'r.. S 5. u.

t e.,3,. y . e ., .; :... 4u.g A I u. :. iD. . ,. . r4 .:. :4...iu. < ..

4.n, C6.. c rS, c. n g .t, .;, . n L. . .:

. t . .:

t, . , . . . , .. - .p

.; ;..,, .: : .u.. r, . i. c..6+ .. .e. : . c. .

e. .c.

..,. .  ;. ra. reg.e...... . .: e, .: +. y e , y. . p. . n e. n e. w. = .> e. . .. s.

. . . . r.

. ~ . . ~ .i: .

.... ..L.Ia:c. s. . i 6.a . aw L ,. . . , .=. .. .w.4 . . - ...

'u e~ e #. #. '. ~. '$ . #. *_.".'...~'".*.:_."'w'.'.'"o r r' '. 5. .' c.- ~ .~ '. .'. .. .#. .c' I

.'. .i

. :. _i .r. . . . . e. . . . 5 m.^'6 . . .

. .. >.- u. . . . ,

..t '. .-e 4. wq

+

..g v.. s.t..r

.L*,

cs g r seu . ., L , , 4

.. -_., . ,3 --  ; . .e c.

- . . . .-r.. ..-l.. .  :(- .q .

i

, ,. . ..'e

.a .?r, a

  • . .=.s. .t

. : . 5 .y

..a. .. t. . . ;,lO r y.. c. * .ea f.

.,r.. C.. : '

. ._i .. ; .a e: . >J

,.  ;. ,, , - i. --i I,y

.. ,. . . , . . i. 0 ...\.

, i, , . .

g ... . ,u. - .s.

.s

..O t i r u*

' .d .
. t i. !
  • t.; ..U:..i.., .

' ' . . ,i . c. U . ,,s .t ,

i.

.i .. ,

~- s'fi' ', .r .

' ' .' . . i. t.f C

. i.il . I t , i' . ' , i

,  !. 'e r. .j 't: .:... .

. l '. "

.3 ' 'ti it i" ' , 1 r eif ' . .d i HJ

- .; . , i i . 5 * . ' l l.GtI t-' J ll ils '. I.n M'., i. ll C ( Oft . . s il P . C f. l C u l c '. ' d 8'. rils'.id 10

. , . z e C 0.. . .. i u*.: ,.:. O I s, . Lei.rs.4

c. .i.: .t

~: 3 . , ,

c. . c . .

4 *i } - > c. u .. A .- .

4 3. t. e,e .; n . t. .p ai...

a n, e:. . . . 2 t

. m. . .-*.< . 'n' -c r m. -. -. .. s um c :. ' t.. ..: 't','

. e-saan

. c .est =

.. S . t. .re. cre . . . . .g. ,. S ,u, : 43 . . z. . ..h r.g. ,s ,.

.. ,s

.s

,,..e . 5 wr .

( . :: i.at. e.e.

. L, ,. .

.. u.eL 1.n

. : . .s j

& x pic:n w m..i.

...e :c a,:: ei v.e

,a.c.- .a l c!L; .f.rau c . ilc poanocena ir.vo .ves , aci r c. .ing t.r,e

. . p 3 s sible n M for additier.al w.s tivityi ar.E yses l ar.d a 5-211 sc'aie, fr.te ral y:te- s t est. 0.;r c5'jective is to confir the Er.alytical results descrited , f in the vur e.L. c..r.. ce,1 e . c. 4e:, U , e c na3.;,,51 S ,,v :i d. ._

c. , , I

. , . . . . c.: c;. < 0  :

.,. .i. . .

. c. : s.

..a

. .a a Cc .: i ,

.r.nts bEirg s tudied in thE contat of the rcM Sy.
@iCC-Crier.ted, E-.-rczncy ;ren- dure guice111.cs. Si ricr-rel v --

et tod ty

  • m .

.L.e.T.. Lt~*3 Harold R. h ien, Director Office of Sucl er ~.acGr O guietien rocle urr:

ts ~s . J Dit t r M .u ; u n. _

tCue ESE Sub;.cct File r.5ntcn 2 5neren .,/r-MS S Ca v a r.Eu c. h-ll 7 93 I

l n

. c. ee e. . s. ,u t D.Cc , r c - 117 c3

, -..a o h.er . ...e .er-l l / 9-a u..

I l

Hibc .?scn PE.r a nd er.be r c-117 93 l %ap i r L.,e: .

Cn, n irCLS c.:. c. . .mn c. . g .

. o s, .

.. r.v. r n o.

ihC R'n a..c u nc.

. .. . .1 i- n

.4t

. , i. , ,

....s. , .s.

j%

.. _ . . . . . .__ . _ _ . . ..___/_.\.- ,

t

f. _.."s,.._.. s / -

i

..,. l s . . . ..,:.

,. ,/. . ,

i.

i .... ... m.

i. , - . . .\...

, -t._,zy . m .e -* , __ ,. ) . . . .

1. 4 s

.._ . I o

s , s

Enclosure DYt&MIC RESPONSE OF B&W REACTORS TO SMALL BREAK LOCAS Detailed analyse,s- have been perfomed either by B&W'or by staff contractors

  • for the two classes of small break LOCAs: (1) those that can be subsequently

isolated (e.g., letdown lines, PORVs), and (2) those that are not.isolatable.

They are discussed in that order, below.

Small Sreaks Which Are Subseouently Isolated We have had our contractor, the Los Alamos National Laboratory (LANL) perfom an analysis of a small break in the cold leg of,the B&% reactor coolant system that is subsequently isolated. The calculations were perfomed with the-advanced TRAC computer code, and we have some preliminary res'ults. We are not aware of any calculations that have t$een performed by B&W in which a .

small break LOCA was subsequently isolated. ,

In our analysis, the system was assumed to lose primary coolant from the break until the upper vessel head region, pressurizer, and hot leg U-bends were filled with steam. . At that time, the break was assumed to be isolated by the operator. Approximately 1,000 seconds later, it was assumed the operator began a controlled secondary system depressurization. The analyses showed that the flow of cold water from the two high pressure injection pumps, coupled with controlled seccadary system depressurization, would condense the hot leg bubbles and restore natural circulation. "We are still evaluating the ability of the TRAC code to model the heat conduction in the liquid near the steam-liquid interface which is required to accur'ately cal-culate the primary system refilling process. However, steam condensation rates are not expected to influence 'the overall conclusion that no unacceptable core heatup would occur,. as explained in the following paragraphs.

If the steam was condensed at 100" efficiency by the cold HPl water, the top of the hot leg U-bends would refill with liquid and single phase natural circulation would be restored. If, however, the steam condensation rate is very low, and, in fact, in the limit the steam is assumed not to con-dense at all, two possibilities would result.

DESI:'pTCCR:ahAL

. CertMied Ey_ht@/n L _ .- - - .D W..- . --

  • I

I The first possibility _is that the HPI pumps would repressurize the system sufficiently to compress the steam to a small enough volume,to allow liquid on the upstream side of the hot leg U-bend to spill over into the downstream ,,

, side and resume natural circulation. If this did not' occur, the HPI would I

continue to inject ECC water and repressurize the reactor coolant. system

until the pressure re' aches either the PORV or the safety valve set pressure.

We would expect that core cooling would be maintained by a " feed end bleed"

process until the steam bubble at the top of the U-bend eventually condensed by heat transfer to the pipes and.across.the liquid vapor interface. Once the bubble condenses, single phase natural circulation throughout the steam generators would be resumed. .. .

l The other possible behavior if the steam condensation rate is low is associated '

with a design difference in B&W reactors. One plant designed tfy B&W,.the Davis-i Besse plant, does not have a "high head" HPI pump (one that can pump water into the primary system at or above the safety valv'e set point). The shuto.ff head tJ this pump is about 1700 psi. In.the event the cold wa,ter~from the HPI pumps does not condense the steam in the hot leg U-bend, the system.may repressurize to above the shutof f head of the HP1 pump, and eventually reach the PORY or the safety velv'e setpoint. The system will begin to lose primary coolant through

! the PORY or the safety valves and drain.down. Once the primary coolant level i

i on the downstream side of the hot leg U-bend extends into the steam generator tube region below the condensing surface of the secondary coolant, steam in the l primary system will begin to condense, lowering the primary s'ystem pressure and closing the PORV or the safety valve.

As the system pressure decreases below 1700 psi, the HPI will actuate and begin to fill the primary system. This would result in covering the condensing.sur-face in the steam generator, and producing another repressuri7ation, which, in t urn. i ould st op IIPI f hiw anil c ause the PNV on the safety valve to open i

and ielea*.c enough on t anI to i res i abl l'.h a e onden . inq sin iai e /\ number of I br .c i vi le- may o. < n i I,ci oi e I he . hai n i nq s v .I em . omp l e t c i v i e t i I l- the sistem er tcfore the steam bubble is t.ondensed by heat tran lcr to p ipe- anil at n os-tne liquid vapor interface. Since the condensing surface in the steam genera-tors is above the elevation of the top of the core, natural circulation should be established before the hot leg steam bubble extends into the core.

DESIGN 3CRIGNLh.

cesiriea 3. /4h ,_ < i .v~

!aali Breaks Which Are Nct Isolated -

Smail break LOCAs which are not subsequently isolated have been calculated by both B&W and by the staff's contractor, LANL. The analysis performed by 2

B&W was for a 0.01 ft cold leg break. They used a computer code that conforms to the requirements of Appendix K to 10 CFR 50. With the exception of the assumed number of HPI pumps available, the modeling assumptions required by Appendix K do not affect the thermal-hydraulic models of interest fo'r .the.

smail break LOCA. Thus, the results of the hydraulic analysis should be realistic.

The 0.01 square foot break was selected since this size is insufficient to remove cecay heat via break flow (thus requiring the steam generators for decay heat removal). It was also predic.ted to result i_n the repressurization phenomenon for the reactor coolant system.

From these analyses, B&W concluded that a. range of small ' break sizes could be postulated in which steam generated in the core would accumulate at the top of the hot leg U-bends and cause an interruption of natural t irculation flow. The interruption of natural circulation f low would isolate the steam being produced in the reactor core from the steam generator heat sink. The net steam accumulation in the system was calculated to cause the primary t

system to repressurize. Th'is repressurization was calculated to continue until the primary system coolant loss through the break was sufficient to uncover a steam condensing surface in the steam generators. It is expected that some steam generated in the core would flow into the upper elevations of the downcomer annulus via the vent valves and conde.nse in the colder water in that region. However, cold water from one HPI pump was not calculated to be sufficient to condense all of the steam generated in the core.

Similar to the isolated break case previously discussed, the repressurization of the reac tor cooiant system caused by interrupt ion of natural c irculat ion would lead to a t. oiler-condenser mode of two-phase na tural circulat ion, and subsequently reduce system pressure. Once the HPI flow is calculated to exceed the break flow, the system coolant inventory will stop decreasing and begin to increase. In figures 1, 2, and 3, the temporal behavior of systein pressure, liquid level in the hot leg piping, and liquid level in the reactor vessel are shown for this case as calculated by B&W. The B&W analyses submitted to. tne staff terminate at about the time system inventory begins to increase.

However, the continued recovery of the event is considered relevant to ycur cencern, and is described further t'elow.

P.Nic h o y'i;hi, m.n .m u(7r' e w. .

As the system refills, the steam condensing surface in the steam generator will again be recovered by liquid, and a steam bubble will ,be trapped at the top of the hot leg U-bend. The scenario is now expected to proceed similar to the isolated break case.previously described. That is, if the ,

steam is rapidly condensed during the refilling process, singl.e phase natural circulation will be reestablished and primary system pressure will remain low with no significant repressurization. If the steam trapped at the top of the hot leg U-bend is not rapidly condensed, the system would repressurize until (1) the break flow exceeded the HP,I flow and a condensing surface was reestablished, (2) the system repressurized and compressed the steam to a sufficiently smali enough volume so that water upstream o-f -the hot leg-U-be'nd could spill over into the downstream side of the hot leg U-bend and reestablish natural circulation, or (3) the PORV/ safety valve setpoint was reached.

For the Davis-Besse plant, we believe only option 1 would occur since the HPI pumps are not sufficient to pump water into the system at or above the safety valve setpoint. ~

The staff has also been calcuiating and analyzing the re ponse of B&W-designed reactors to small break LOCAs in which natural circulation is predicted to be inter-rupted by steam accumulation at the top of the hot leg U-bends. Our con-tractor at LANL has recently completed a few.small break analyses and has looked at four recovery enhancement actions presently either proposed by B&W or being considered by the staff. These four options are: (1) high point vent operation, (2) momentary pump restart, (3) secondary side de-pressurization, and (4) ECC spray at the top of the hot leg U-bends. All l of these options are being investigated to determine their ability to enhance l the reestablishment of single phase natural circulation during the recovery portion of the accident. Initial results of our contractor's calculations l show that for a realistically calculated small break (i.e., nominal decay heat, two HPI pumps available, etc.) in a B&W plant, with a break size in the range of that for which B&W predicts repressurization would occur, the system did not repressurize once the hot leg U-bends filled with steam. Although our l

evaluation is not yet complete, we believe that the rea' son the LANL calculation I did not show a repressurization is because steam generated in the core vented to the upper reaches of the vessel annulus via the vent valves, and the flow l from ts.o HPI pumps was sufficient to condense all of the steam produced in the core.

l f i

DESICDTED ORIGFJAL

- - Y.7

5 The results of the TRAC analyses are shown in figures 4 through 7. In figures 4 and 5, the B&W results are overlayed to show the differences. We believe they can be attributed to two versus one HPI pump being available. The results of the analyses to investigate natural circulation recovery enhancement methods were only recently presented to the staff at a meeting with LANL. These ana-lyses have not been documented by LANL in a formal report, and we have not re-viewed them in any detail. However, based on information received at the meet-ing with the contractor, the results show that the hot leg U-bend was not refilled following recovery from t'he .small break LOCA (1.75 inch diameter), and that none of the natural circulation recovery enhancement methods previously listed were effective. However, LANL reported that the core remained cover'ed and decay heat was continuously removed from the core. They attributed the heat removal to internal recirculation.(steam exiting the core is vented to ,the vessel upper annulus and condensed by the cold HPI water entering the downcomer). This situa-tion physically could only persist until tiie decay heat was eventually removed entirely by the break flow or the system eventually Qas refilled and natural circulation reestablished. .

~

Before widely disseminating the results of the LANL calculations, we have asked our Office of Regulatory Research to carefully document and evaluate them and assist us in confirming their validity. We believe this careful approach is justified because the analyses showed that core decay heat was continuously removed and that no core uncovery or heatup was predicted.

In sumary, although we are continuing our evaluation of the rate of hot leg steam bubble condensation in the recovery from both isolated and unisolated reactor coolant system small-break LOCAs in reactors like TMI-1, we do not believe that steam bubbles present in the reactor coolant system resulting from small break LOCAs (either isolated or unisolated) in either the cold or hot legs of the primary system will result in unacceptable heatup of the core.

DISIGrbr3 0?"Qh!.L Certified By

  • M %__

l

Figure 1 4

CORE FR(550RE VS. TIME -

!?..C~C - 0.01 FT' BREAK AT PUMP '

DISC'4ARGE-5YW.ETRIC AFW. E

.t e.

25,w .. ,

e-.,

' ~

& 24.CQ -

o X . . .

w .

l 22 .M "P' l

(,

l

$0*0 ~~, ,

t 4 .

SW ay c_ t.o .r w.~su e

W Z

D -

L7 .

D .

u' x 3.5 v~ -

C .

1

,. .a e i g

s i

i . .

s.o9 . t=

===, -

y S- *ea m

V..w.

r=a r q =am w.www e paa JLm .w A q=

aw.www P r= = av. = a v.

&J e ww.s w . .r,Ja= . * ==-.

. ~*.

T :. gi , \, . .c

. .n .

n ". rer - i n i i

  • s v : J '~ T -E - O

. n

- O  : ""J I *7 7" n L. L A

DESIG.n. ,qigo. - GlN).L .

- CO*tified By /(s j 1

%m A

==.-

m et-e

'[

_C

.!!! C M

- ~

Cw

=,*

c L .,

o O c LC , . N ,

C w

.J C O M GJ

.= c w C .as: e

> Gu C N W .an: 6 J O .O 6 c C, C= == c

= v N

g y W

C C

= CK

=

c N

. N

. ~

r e

o N

>=

m c

  • o a -

=

C

=

o

\ y c

.-= o -

e

< e - -

e e a = c C

> -

v ._ c c: .

v

== - C LJ

- 0 v -

e .r.

a w N 5

/ c .E e -

=

- - = i-c < N

=

,- / .

k O

[ . ..

- C

/ 2

[ c w

- c l - = .

j c f I

= -

o o

l -

! a e

- o

/

e

- e

, o d N

.c=_- o e e, e m e e o e c

-3 e e r r, N N - -

.ri. "

LISIG D 0. .1 L Certified By ll~ di% _

u. w o>w e:

.._ - o ouu r<

i

~

e w c c

. =~a. ~

. zw ~

-m m

1 wu w> e

> =

W cs ac-woe== ~

. wv

. c< =

2w

. . c -

n - -

.x. =

~

C3 ,

r 5 N

u ._

3 ,

cn . -

A e

~

5 w::

5 v

( - w e

w

. .. .r e ._

N e

- e C

c e

e I I I ( l e

,e, e e e e e m ~ - -

(.Ld).LWI3H DESIG MED a. .. 03" A"L Certified By le ' A Ww -~~

SC B& W Lo*'ERED-LCOP 0.01-M2 COLD .EG BREAx

  • 0- (souc) nac (C45x) &&w .

co. , .

= ' "

i a-i c ,

o so. t.p.h E 1 d i..t 1J 6 i ,3 [-[ j2 u h* Mi . ,' f ! f 

                                                            .t *,I.:    e i                            j          !.

fy ' V.f ,r f 

                                           . . . . . . .           ~.                                                   I so.

e o o == imo sooo asoc co: m: rose 

                                                                   . . . T.i ME. (*)                 .

Figure 4 Cc=parison of' TRAC 'a'n d B&W break [ lows. e Ekw LowTRED-LOOP 0.01-rT2 COLD-LEC BREAK i 

                                                                               ,(sc u c) Taac              -                                  -
                    ,_,                                                         (DasM) c&w l

t:W. ' . o / , *. . e w g too*o'.  : h<op m . e. [ 

                                                                                                                                                                           ,4 g

w>.c .e I g' l - Q k w>c'. 

                                                                                                                      \

1 4m' e me awe me i .ooe w;; ,,,;; ,m TawE (s) l Figure 5 

               . Co=p a ri s on of I~i.AC a Ed E4'=' prima ry p re s sure s .

l l

4 s R&w L0wCRED-LOOP 0.0kr72 COLD-LEG plaK ~ 

               . u g

C x ___ w - - ' ' ' U g ca. 5 

         $ 4 c'-                        ,                                                      ..

9. J 

'           m
            ?     es.

O . v ,,. - A c C C00 2000 3000 eCC3 6000 6:4C 70C0 Tiwt (s) , Figure 6 ._ TF.AC-calculated core li~ quid volu=e fraction,. B&W LOWERED-LOOP 0.0br72 COLD-LEG BREAK *

  • te Q

r,, 50e. w . e'e ,

d. '

! 3*=E w s ea. t** A D~ a i w I 

          ~

uo. g .

9. ,4,
          =                                                         .
                                                                                                                                                     )           $

g po . - 

          =

E 

                                                                                                                                                ,              b
          <    we.                                                                                      .

g su. . 2 uo e toe teoc wo: ens ms 6:co ex: Tsut (3) . Figure 7 i TFAC-ca) ulated uta21=u= average rod cladding te=;:erature. l _- _ , _ _ , .}}

Retrieved from "https://kanterella.com/w/index.php?title=ML20062E165&oldid=3555042"