ML19294B987

Prepared Direct Testimony Before ASLB Re Natural Circulation Cooling.Certificate of Svc Encl

ML19294B987
Person / Time
Site:	Rancho Seco
Issue date:	02/19/1980
From:	Lewis H CALIFORNIA, UNIV. OF, SANTA BARBARA, CA
To:
References
NUDOCS 8003060503
Download: ML19294B987 (16)

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Before the Atomic Safety and Licensina Board In the Matter of:

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SACRAMENTO MUNICIPAL UTILITY

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DISTRICT

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Docket No. 50-312 (SP)

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(Rancho Seco Nuclear Generating

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Station)

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Prepared Direct Tastimony of Dr. Harold W. Lewis Concerning Natural Circulation Cooling February 19, 1980 Sponsored by the California Energy Commission 8003060

s Prepared Direct Testimony of Dr. Harold W.

Lewis Concerning Natural Circulation Cooling My name is Harold W.

Lewis.

I am a professor of physics at the University of California at Santa Barbara.

Though I have chaired the American Physical Society study on light-water reactor safety and the Nuclear Regulatory Commission's Risk Assessment Review Group, am currently a member of the Advisory Committee on Reactor Safeguards, and have recently agreed to serve on the Advisory Council of the Institute for Nuclear Power Operations, the views expressed in this testimony are my own, and I am not speaking for any group with which I am or have been associated.

A resume is appended to this testimony, which describes in more detail my background.

My testimony is sponsored by the California Energy Commission (CEC), which I understand is participating in this inquiry as a non-adversary " interested State" under the provisions of 10 C.F.R. 52.715(c).

My sole intent is to assist the Atomic Safety and Licensing Board in resolving the technical issues before it.

To that end, the CEC has asked me to addresa several questions, which are set forth following an introductory discussion of various core cooling techniques.

These questions, and my discussion, focus upon the following issues in this proceeding:

1) Additional Board Question No. 3;

2) Board Question CEC 1-2;

3) Board Question CEC 1-10;

4) Board Question Hursh and Castro No. 24.

, In addition, this testimony also relates to the following issues:

1) Additional Board Question No. 2;

2) Board Question CEC 1-4;

3) Board Question CEC 1-6;

4) CEC 3-1;

5) Board Question CEC 5-3a;

6) Board Question Hursh and Castro No. 22.

I.

INTRODUCTION We will be learning the lessons of Three Mile Island for some years to come.

It is generally agreed, however, that the accident and the studies it stimulated have highlighted some problems in the operation of commercial nuclear reactors, some of which are specific to Babcock and Wilcox plants like Rancho Seco.

Among these are the following:

1) The design of the once-through steam generator, in particular its small secondary side volume, which makes the response of the primary system particularly sensitive to perturbations in the secondary system;

2) Problems with the pilot operated relief valve;

3) The specific geometry of the connection between the pressurizer and the hot leg, which makes it possible for the pressurizer to be full while the primary system is not, and can therefore lead to misinterpretation of the state of the primary water charge;

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. 4) Inadequate operator training, inadequate written procedures, inadequate control room instrumentation, and other contributors to operator errors such as occurred at TMI; and

5) Inadequate pre-TMI attention to the role of smallbreakLOCAd.

It is fair to ask whether we have learned by the TMI experience, and whether the modifications, training, and other changes instituted at Rancho Seco make an accident of the level of TMI sufficiently unlikely there.

To answer this question, we must focus upon the abil'ty of the Rancho Seco system to assure adequate core cooling during either auxiliary or main feedwater disturbances, with or without a small break LOCA.

II.

Feedwater Reliability Main feedwater disturbances are quite frequent in Babcock and Wilcox plants, so that it is important to have an extremelf reliable auxiliary feedwater system.

Bekcock and Wilcox has conducted a generic study of the reliability of auxiliary feed-water systems, using standards techniques of fault-tree / event-tree analysis (which are, at long last, beginning to be used effectively).

This study indicates that the Rancho Seco auxiliary feedwater system is among the more reliable of such systems.

One of the unique features of the Rancho Seco plant is that one of the AFW pumps has both electric drive and turbine drive on the same shaft, which enhances its reliability.

, Barring the sort of operator error which resulted in the inadvertently closed block valves at Three Mile Island, an event whose likelihood will be greatly reduced by the provision of flow indicators in the AFW system, the likelihood of failure of auxiliary feedwater has been shown to be quite low.

Nonetheless, auxiliary feedwater may fail.

For this reason, one should be able to maintain adequate core cooling without feedwater.

III.

Principles of Core Cooling For the purpose of this testimony, we assume that reactor trip has occurred either through high RCS press.re or through anticipatory trip on main feedwater disturbance, or for some other reason.

Thus, we concentrate on removal of decay heat.

It is worthwhile at this point to be quantitative and set forth some rough numbers.

The decay heat of the core of this plant, soon after reactor trip, will be approximately 150,000 kilowatts (a kilowatt, for the rough purposes of this testimony, can be regarded as equivalent to 1 BTU per second), decaying to approximately 50,000 kilowctts in 15 minutes, and about 15,000 kilowatts after a day.

As a rough guide to the heat that needs to be removed for a stable configuration, I will use an estimate of 50,000 kilowatts as typical of the period in which the events I will be discussing will occur.

Clearly, one can do more detailed calculations.

. If 50,000 BTU per second are to be removed by normal rejection to the secondary system, and heating of the primary coolant by approximately 50 Fahrenheit, a flow rate through the primary system approxiamtely 2% of the normal flow is required.

Under these " normal" conditions, that flow can be achieved either by the use of the reactor coolant pumps, or by ordinary natural circulation (for a primary system without voids).

If 50,000 BTU per second are to be rejected by discharging steam (for example, through a relief valve), the required flow rate is approximately 30 pounds per second, or 200 gallons per minute, or 100,000 pounds per hour, approximately the capacity of the pilot operated relief valve.

If the same amount of heat is to be rejected through the discharge of solid water, and its replacement by cold water (for example, by actuation of the high pressure injection system), the necessary flow rate is approximately three times as great as would be the case for steam.

Finally, if there is no means of rejecting heat through the secondary system, and no means of discharging either water or steam from the primary sirtem, the capacity of the primary system is such that it will warm up, with a heat input of 50,000 kilowatts, at a rate of approximately 5' Fahrenheit per minute.

Thus, in the case of a reactor trip associated with loss of all feedwater, and with no other heat losses from the primary system, one can estimate that the operators have between 10 and 20 minutes available to initiate an alternate cooling mode.

. IV.

Small Break LOCAs Small break LOCAs cover a range of violations of primary integrity resulting from either pipe breaks or stuck safety or relief valves.

For very small breaks, there is no problem main-taining ccolant inventory, but one must assure heat removal from the primary system.

With breaks sufficiently large to discharge enough coolant to cool the primary system, there is no heat removal problem, but maintaining water inventory may be problematic.

In the event of loss of all feedwater, so that heat rejection to the secondary system becomes impossible, the proper move is to initiate high pressure injection, rejecting water through the PORV and/or safety relief valves, thus initiating what is in effect a small break LOCA.

In the case of a stuck PORV, as occurred at Three Mile Island, there was sufficient heat removal through the valve to maintain adequate core cooling, and no damage would have ensued had the operators not decided to throttle back on high pressure injection.

It was the loss of core water inventory through that act that ultimately led to the uncovery of the core.

Thus, for the larger LOCAs, heat rejection will be through the break itself, while the smaller breaks will normally be associated with heat rejection through the normal secondary system if available.

If not, high pressure injection must be used, and heat removed through the so-called " feed and bleed" method.

. All of this serves to underline the availability of a variety of means of assuring core cooling, depending upon the circumstances of the sequence in question.

At the same time I must emphasize the importance (in these high pressure situations) of maintaining an adequate invetory of primary water, and providing sufficiently good diagnostics to the operators to enable them to take the appropriately correct action.

In summary, with respect to feedwater transients and small break LOCAs I generally believe that there is sufficient redundancy in plant equipment, and options open to the operator, to provide adequate assurance of core cooling (provided the operator takes appropriate action).

However, where heat removal through the secondary system is unavailable, core cooling can only be assured by discharging primary coolant through either a break, a valve, or both.

Further, there is a greater premium on proper diagnosis.

V.

CEC Questions on Core Cooling without Forced Circulation (a) Is it prudent to cease operation of the reactor coolant pumps whenever the high pressure injection system is actuated?

See IE Bull.79-05C.

What potential problems or uncertainties are presented by operating the reactor coolant pumps following actuation of the high pressure injection system?

If the limiting size small break is present, can such operation lead to inadequate core cooling or other significant problems?

. The underlying rationale for the NRC decision to now require trioring of the reactor coolant pumps on safety injection is that there is a range of small breaks, from.02 sq. ft. to.2 sq.

ft.,

in which fluid losses from the primary system are such that a two phase configuration is established, and subsequent tripping of the pumps will lead to separation between the steam and water, so that the resulting water level may be below the top of the core.

Analyses by Babcock and Wilcox, and confirmatory analyses by the NRC, establish the window under which this is possible.

Since one has to presume that the size Of the break in a small break LOCA is not known, it is therefore prudent to assume that the break, if it occurs, is within the threatening window, and to trip the pumps early.

For other sizes of break and more particularly for non-LOCA transients which depressurize the primary system and simulate a break, early tripping of the pumps may exacerbate the problem, though the calculations seem to show that one is not likely to exceed core temperature limits.

Nonetheless, it is better for the operator to leave the reactor coolant pumps running if he can determine that he is facing a depressurization transient not caused by a small break, and the premium is, as always, on operator capability and adequate instrumentation.

The times available for operator decision are shortest (a few minutes) at the upper end of the small break spectrum, corresponding to clean breaks in pipes approximately 6 inches in diameter.

For such breaks depressurization will be rapid, and will continue rapidly beyond the point of high pressure injection, providing a diagnostic tool.

(b) What are the potential problems and uncertainties, if any, created by relying upon natural circulation cooling whenever the high pressure injection system is actuated?

Natural circulation cooling is relevant only whon a secondary heat sink is available, and where fluid losses from the primary system are inadequate to remove decay heat.

Thus we are talking about very small breaks (the effective break size at TMI was less than.01 aq. ft.).

If secondary heat removal is available, natural circulation depends upon a unbroken liquid train from the steam generator through the cold leg piping into the reactor pressure vessel.

If the remainder of the system is full we have single phase natural circulation, otherwise we have steam or steam / water mixture rising through the hot leg, over the candy cane, and being condensed in the steam generator.

This is two phase natural circulation or condensation cooling, not unlike that available in a still.

Thus, the only real requirement is that the fluid level be above the level of cold leg piping, a condition which was not achieved at TMI.

Current instructions to B&W plants are to maintain steam generator secondary level sufficiently high, in the event of such an occurrence, that even two phase natural circulation will have an unbroken water column in the relevant region.

Given operator adherence to this condition, natural circulation ought to be reliable.

(c) Can voiding occur in the Rancho Seco primary system due to the sensitivities of the B&W reactor system or other factors?

If so, what conditions or event sequences are likely to cause it?

Voiding can unquestionably occur in the Rancho Seco primary system under conditions in which the pressure is reduced below the saturation pressure of the water in the hottest part of the circulating system.

This will normally be the hot leg, and the corresponding saturation pressure will be approximately 1600 psi in normal operation, near the set point for high pressure injection.

Thus, voiding is typically associated with a primary system break sufficiently large to exceed the makeup capability of the high pressure injection system.

It can also conceivably occur during a severe overcooling transient but that is likely to be a short lived phenomenon.

For the case of a break, the situation has been discussed above, while for a transient the correct move is to restore system pressure and collapse the voids.

(d) If voiding does occur, is it likely to impede or prevent natural circulation?

What other events or conditions may impede or prevent it?

If voiding does occur it is most likely in the hottest part of the system, the hot leg, and will automatically put the reactor into the two phase natural circulation mode.

Since pool boiling is adequate to cool the core, if covered, the effectiveness of this two phase natural circulation depends entirely upon the ability of the secondary system in the steam generator to remove heat, while condensing the steam flowing through it.

This also provides the necessary supply of water through the cold leg into the reactor core to keep it covered.

As mentioned earlier, natural circulation will be impeded by any event which breaks the liquid link between the reactor core and the steam generator via the cold ieg.

Too low a liqutd level, noncondensibles in the system, breaks in the plumbing at too low a level, and failure of the heat sink through the steam generator, can all impede natural circulation.

At TMI, it was at first impeded by too low a liquid level, and then, after core damage, by noncondensibles.

(e) If natural circulation cannot be achieved, can adequate core cooling be assured by any other method than restarting the reactor coolant pumps?

Are there any potential problems and uncertainties associated with such method (s) ?

Is reflux boiling an acceptable method of assuring adequate core cooling?

Presumably the intent of this question is to deal with a situation in which there is not a sufficiently large break to assure heat removal through the break, and in addition, neither single phase nor two phase circulation to the secondary system can be assured.

This would presumably be a situation in which there are noncondensibles in the system, so that the steam path over the candy cane, which would be required for two phase natural circulation, is not available.

Under these conditions, a fall-back position which could be theoretically effective, but has not been thoroughly analyzed, is the so called " feed and bleed" mode.

This is available in the case of Rancho Seco, but probably not for other B&W plants, simply because of the high head pressure available through the Rancho Seco high pressure injection system.

It rests on the fact that, as discussed in the main part of the testimony, steam rejection through the PORV and the safety valves is sufficient to r3 move the decay heat from the core.

Thus one imagines allowing the upper part of the pressure vessel to fill with steam, while maintaining core coverage, so that the core is in a pool boiling mode.

One allows the pressure to build up sufficiently to discharge steam through the PORV and, perhaps, the safety valves, at a rate sufficient to remove decay heat from tne system.

The water loss is then intermittently made up through the high pressure ir Jection system.

This should work in principle, but has not been fui.ly analyzed.

In addition, as a long term cooling mode, it is not clear how many activations of these various valves are prudent.

Finally, to the extent that the core safety valves would be involved, it is never prudent to use a safety item in a normal operating mode.

Thus, the procedure must be regarded as a theoretically practical means of core cooling, to be used in extremis, until secondary cooling is restored.

VI.

Conclusion It is evident from the above that there is a wide variety of cooling modes, of both normal and emergency character, for the Rancho Seco plant.

Nonetheless, the plant is quite sensitive to secondary disturbances, and there is thus an extra premium on the capability of the plant operators, in an emergency.

It is my view that what hardware problems there were have been largely remedied by the series of orders that have been mandated by the NRC in the aftermath of TMI, and compliance with those should, of course, be assured.

I believe, however, that particular attention should be given to compliance with the additional training, education, and manning requirements, as well as simulator training of the operators in the kinds of accident mode associated with the plant's transient behavior.

In this connection, it is a modest advantage that the B&W simulator at Lynchbrug, Virginia most closely resembles the Rancho Seco plant.

Dr.

H.

W.

Lewis Professor of Physics (1964 to Present)

University of California Santa Barbara, California 93106 1948 Ph.D.,

University of California, Berkeley 1947-1948 Institute for Advanced Study 1950-1951 1948-1950 University of California, Berkeley 1951-1956 Bell Telephone Laboratories 1957-1964 University of Wisconsin 1961-1962 Princeton University (Visiting) 1964-Present University of California, Santa Barbara Chairman, Study Group on Light-Water Reactor Safety, American Physical Society, 1975 Chairman, Risk Assessment Review Group, Nuclear Regulatory Commission, 1978 Member, Advisory Committee on Nuclear Safeguards Member, Advisory Council, Institute for Nuclear Power Operations

186B:25 2/19/80 ke UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of:

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SACRAMENTO MUNICIPAL UTILITY

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DISTRICT

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Docket No. 50-312 Rancho Seco Nuclear

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Generating Station

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PROOF OF SERVICE I,

Debbie Jones, declare that on February 19, 1980, I deposited copies of the attached " Prepared Direct Testimony of Dr. Harold W.

Lewis" in the United States mail at Sacramento, Califo rnia, with first class postage thereon fully prepaid and addressed to the following:

Elizabeth L. Bowers, Esq., Chairperson Thomas A.

Baxte r Atomic Safety & Licensing Board Panel Shaw, Pittman, Potts, and Trowbridge Nuclear Regulatory Commission 1800 M. 57, N.W.

Washington, D.C.

20555 Washington, D.C.

20036 Executive Director for Operations Mr. Mark Vandervelden U.S. Nuclear Regulatory Commission Ms. Joan Reiss Washington, D.C.

20555 Mr. Robert Christopherson Friends of the Earth Dr. Richard F. Cole California Legislative Office Atomic Safety and Licensing Board 717 K Street, Suite 208 U.S. Nuclear Regulatory Commission Sacramento, CA 95814 Washington, D. C.

20555 Docketing & Service Station Mr. Frederick J. Shon Of fice of the Secretary Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Washington, D.C.

20555 Mr. Lawrence Brenner David S. Kaplan, Esq.

Counsel for NRC Staff Secretary and General Counsel U.S. Nuclear Regulatory Commission P.O.

Box 15830 Office of the Executive Legal Sacramento, CA 95813 Director Washington, D.C.

20555 Timuthy V.A. Dillon, Esq.

Suite 380 Richard D. Castro 1850 K Street, N.W.

2231 K Street Washington, D.C.

20006 Sacramento, CA 95816 Gary Hursh, Esq.

Stephen Lewis 520 Capitol Mall, Suite 700 Office of the Executive Legal Sacramento, CA '95814 Director U.S. Nuclear Regulatory Commission Washington, D.C.

20555

186B:26 R2 1/8/80 sep 2

James S. Reed, Esq.

Martias F. Travieso-Diaz, Esq.

Michael H. Remy Shaw, Pittman, Potts & Trowbridge Reed, Samuel & Remy 1800 M Street, N.W.

717 K Street, Suite 405 Washington, D.C.

20036 Sacranento, CA 95814 Lex K. Larson, Esq.

Atomic Safety and Licensing Board Shaw, Pittman, Potts & Trowbridge Panel 1800 M Street, N.W.

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

20036 Washington, D.C.

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

20555 I am and was at the time of the service of the attached paper over the age of 18 years and not a party to the proceeding involved.

I declare under penalty of perjury that the foregoing is true and correct.

Debbie Jones Attachment

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