NRC Generic Letter 1981-21

NRC Generic Letter 1981-021: Natural Circulation Cooldown

ML031080586
Person / Time
Site:	Beaver Valley, Millstone, Calvert Cliffs, Davis Besse, Salem, Oconee, Mcguire, Palisades, Palo Verde, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Watts Bar, Sequoyah, Byron, Arkansas Nuclear, Three Mile Island, Braidwood, Summer, Prairie Island, Columbia, Surry, North Anna, Turkey Point, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Waterford, Farley, Robinson, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, McGuire, 05000000, Zimmer, 05000398, 05000380, Trojan, Marble Hill
Issue date:	05/05/1981
From:	Eisenhut D Office of Nuclear Reactor Regulation
To:
References
GL-81-021, NUDOCS 8105140267
Download: ML031080586 (12)

UNITED STATES

0 \NUCLEAR REGULATORY COMMISSION

WASHINGTON, D. C. 20555 MAY 5 1981 TO ALL LICENSEES OF OPERATING PRESSURIZED WATER NUCLEAR POWER REACTORS AND

APPLICANTS FOR OPERATING LICENSES (EXCEPT FOR ST. LUCIE, UNIT NO. 1)

Gentlemen:

t' SUBJECT: NATURAL CIRCULATION COOLDOWN

(Generic Letter No. 8121-)

On June 11, 1980, the St. Lucie Plant, Unit No. 1, was forced to cool down on natural circulation as a result of a component cooling water malfunction.

During the cooldown process, abnormally rapid increases in pressurizer level were observed. Subsequent analyses have confirmed that these abnormal level increases were produced by flashing of liquid in the upper head of the reactor vessel, forcing water out of the vessel and into the pressurizer. A more complete description of the event and circumstances involved is provided in y-' the enclosure which includes a letter sent to the PWR NSSS vendors soliciting their opinions and comments on the significance of the event and phenomenon

> 0 pin general.

>' Based on our review of the event to date, we believe that core cooling was

-never lost during the St. Lucie, Unit No. 1 event. That specific event does not constitute a direct safety concern. We have, however, identified two areas of concern applicable to all pressurized water reactors requiring prompt action:

1. The Unacceptability of Vessel Voiding During Anticipated Cooldown oenditions INatura Circulation Due to LOSS Ot Osite Power, Coss of Pumps, etc.)

Cooldown with a significant steam void in the vessel requires controlling a "two pressurizer" system, which is an undesirable challenge to the operator. In fact, we are not aware of any training facilities (simulators)

today which would allow an operator "hands on" experience in practicing such control. Moreover, it is our opinion that any significant vessel voiding produced during controlled cooldown conditions increases the susceptibility of the plant to more serious accidents. For these reasons reactor vessel voiding during controlled natural circulation cooldowns should be avoided.

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-2- As described in the enclosure, vessel voiding at St. Lucie, Unit No. 1, was caused by the operator reducing system pressure such that the corresponding saturation temperature dropped to the temperature of the relatively stagnant fluid in the reactor vessel upper head. Presently, primary system cooldown rates are based on vessel structural integrity considerations and do not explicitly consider avoiding production of significant steam voids in the vessel. Moreover, cooldown rates are based on fluid temperatures measured in the primary piping. As the St. Lucie Unit No. I event has shown, these measured temperatures can in fact be on the order of 100 degrees Fahrenheit or more lower than the upper head fluid temperature, and, therefore, not indicative of the saturation pressure of all fluid in the primary system.

Under conditions which require cooldown on natural circulation and when rapid depressurization is not necessary there may be a number of ways to avoid reactor vessel voiding. For example, a low cooldown rate can be specified, coupled with "holding" the plant at intermediate conditions to allow the fluid in the upper vessel to equilibrate with the rest of the primary system. However, avoidance of vessel voiding by lower primary system cooldown rates can increase the time required to achieve shutdown cooling entry conditions and thus increase the time auxiliary feedwater is depended upon to remove decay heat (specifically, for the loss-of- offsite power case). Thus, supplies of condensate-grade auxiliary feed- water must be considered if cooldown times are extended.

2. Failure of the Operator to Have Prior Knowledge and Training for This Event The cause of initial surges in pressurizer level at St. Lucie, Unit No. 1, was not immediately recognized or understood by the operator. We attribute this to the fact that long-term natural circulation cooldown under the specific circumstances of the event was never explicitly analysed by the NSSS vendor from the standpoint of trying to recognize a phenomenon such as that which occurred at St. Lucie, Unit No. 1. In the St. Lucie event, the operator ultimately recognized the cause.of the level surges and was able to maintain control of the plant. Our concern, however, is the possibility of an operator taking incorrect action in an effort to correct for an unknown event or unrecognized phenomena.

We believe that proper procedures and training can provide the necessary guidance to the operators both to avoid reactor vessel voiding as well as recognize it when, and if, it occurs during controlled natural circulation cooldown. We are not sure if such procedures and training are in place at pressurized water reactor facilities.

- 3 -

Consequently, we request that you promptly review your current plant operations in light of the St. Lucie, Unit No. 1 event and the discussions above and implement, as necessary, procedures and training which will enable operators to avoid (if possible), recognize and properly react to reactor vessel voiding during natural circulation cooldown.

We conclude that the actions described above should be completed as soon as they reasonably can be (i.e., within 6 months for operating reactors). In addition, so that we may determine whether your license should be amended to incorporate these actions as requirements, licensees of operating pressurized water reactors are requested, pursuant to §50.54(f), to furnish, within 6 months of receipt of this letter, an assessment of your facility procedures and training program with respect to the matters described above. Your assessment should include:

1. a demonstration (e.g. analysis and/or test) that controlled natural circula- tion cooldown from operating conditions to cold shutdown conditions, conducted in accordance with your procedure, should not result in reactor vessel voiding;

2. verification that supplies of condensate-grade auxiliary feedwater are sufficient to support your cooldown method; and

3. a description of your training program and the provisions of your procedures (e.g. limited cooldown rate, response to rapid change in pressurizer level)

that deal with prevention or mitigation of reactor vessel voiding.

Applicants for operating licensees are requested to implement the subject procedures and training and provide the requested assessment within 6 months of receipt of this letter or 4 months prior to the staff's scheduled issuance of its operating license Safety Evaluation Report, whichever is later.

Please refer to this letter in your response.

This request for information was approved by OMB under a blanket clearance number R0072 which expires December 31, 1981. Comments on burden and duplication may be directed to the Office of Management and Budget, Reports Management, Room 3208, New Executive Office Building, Washington, D.C. 20503.

S\ncerely,

.i'se Division o Licensing Office of Nuclear Reactor Regulation Enclosure: As stated cc: Service list

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WI?.11[ESTA1ES ,Enclosure 1 NUCLEAR RLGULATOR:Y CC),.Iil'(,:

j.ASus:GIOM. D. C. 7055N

%U A 12 1BO

Letter sent to PWR NSSS Vendors:

Westinghouse, Combustion Engineering and Babcock and Wilcox Dear Mr.

SUBJECT: VOID FORIATION IN VESSEL HEAD DURING ST. LUCIE NATURAL

CIRCULATION COOLDOWN EVENT OF 6/11/80

On June 11, 1980, the St. Lucie reactor was shutdown due to a loss of component cooling water to the reactor coolant pump seals. This also required shutdown of the reactor coolant pumps and c:oldown Was accomplished by natural circulation.

At approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into the event, charging flow, which Was initially being divided between the cold legs and the auxiliary pressurizer spray, 1..as diverted entirely to the auxiliary spray to enhance the depressurization and reduce the system pressure on the pump seals. At this time, abnormally rapid increases in pressurizer level were observed which could not be explained by the charging flow rate alone. Detailed evaluation and follow-up analyses by the licensee and NSSS supplier have indicated that a steam void was probably formed in the upper head region of the reactor vessel and displaced water from the vessel into the pressurizer.

Continued alternating realignment of charging flow between the cold legs and auxiliary spray line produced a 'saw-tooth" pressurizer level behavior.

Relevant information and data available to the staff to date are provided in the enclosure.

It has been postulated that the steam void in the upper vessel was produced when the system pressure dropped below the saturation pressure corresponding to the temperature of the fluid in the upper head. Because the measured hot and cold leg temperatures at the time of voiding here highly subcooled

(-2000 F), it appears -that the fluid. in the upper head was much hotter, relatively- stagnant, and in-poor communication with the fluid exiting the core and in the upper plenum. In addition, stored heat in the upper head structures most likely contributed to the voiding.

Because of the unexpected occurrence of the void, the failure of the operators to immediately recognize the void formnation and take corrective action, and the question of whether such void formation is properly accounted for in safety

a -  :.d * -;

analyses (Chapter 15), we have sent a list of que~tions documuenting our concerns to the licensee. These questions are also provided in the- enclosure for your information.

We are presently evaluating the need to pursue this issue generically with all PUR licensees. Prior to taking any definitive action however, we are soliciting your technical opinion and advise regarding the potential for void formation under similar circumstances in USSS's designed by you.

Specifically, we need to know if you can justify why the voiding phenomenon cannot occur in USSS's designed by you (or can confirm that such phenomena can be properly predicted by your transient analysis models), and if itcan occur, is properly accounted for in operating procedures (e.g.. cooldowm rates), operator guidelines, and operator training (including the simulator)

The urgency of this matter requires you advise us within fifteen (15)

working days after receipt of this letter whether a supplemental information submittal by you on the subject would preclude the need to expeditiously pursue this issue generically with your customers.

tu..binal Signd b

1'sui S.MCA .

Paul S. Check, Assistant Director for Plant Systems Division of Systems Integration Office of Nuclear Reactor Regulation e. ..

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UNITED STATES

NUCLEAR REGULATORY COMMISSION

s XWASHINGTON. 0. C. 2Cm June 25, 1980

Docket No. 50-335 LICENSEE: Florida Power & Light Company (FP&L)

FACILITY: St. Lucie Unit No. 1 SUBJECT: SUMMARY OF MEETING WITH FP&L AND COMBUSTION ENGINEERING (CE) REGARDING

ST. LUCIE UNIT NO. 1 COOLDOWN ON NATURAL CIRCULATION

On June 20, 1980 a meeting was held in Bethesda, MD regarding the June 11, reactor trip and cooldown on natural circulation at St. Lucie Unit No. 1.

Significant points discussed are summarized below. Enclosure 1 is a list of attendees.

Discussion A brief chronology of the event was presented with traces of parameters enclosed as Enclosure 2. The plant experienced a complete loss of component cooling water (CCW) to the reactor coolant pumps (RCP). This loss was due to a direct short on the terminal board for a solenoid operated CCW common return valve from the RCP's. Plant procedures allow ten minutes to correct the problem before securing the RCP's. Subsequently, the reactor was tripped and the RCP's secured. About three minutes after shutdown one RCP was run briefly to aid natural circulation flow. The licensee stated that intermittent RCP leakoff flow alarms were actuated and that subcooled margin remained within the range 65 to 180'F subcooled during the cooldown.

The licensee indicated that a steam bubble had formed in the head of the reactor vessel during cooldown. This was indicated by abnormal pressurizer level changes during charging operations and was apparently due to elevated temperatures in the hydraulically stagnant area of the upper head.

The possibility of steam bubble formation was confirmed with an analysis by CE

using an in-house model. This model decoupled the vessel head region from the rest of the reactor vessel with respect to temperature. As a result of the analysis CE issued an information bulletin (Enclosure 2) warning of the long vessel head cooldown times which may be required (10-15 hours) and recommending actions to maintain plant conditions until the vessel head has cooled should a bubble form.

The licensee indicated that while in the shutdown cooling mode of operation primary coolant may have leaked to the refueling water tank through LPSI lB some recirculation line. The isolation valve in this line was found not to be pump's completely shut. The licensee stated that there were no indications of radio- activity release. While on shutdown cooling LPSI Pump lB was operating in the injection mode with its recirc line open.

St. Lucie Unit No -2- For corrective actions the licensee has-modified its procedures to include the CE guidelines and installed backup means (additional air supplies and manual tachet) of opening the CCW containment isolation valves.

Conclusions The NRC staff indicated the need for more detailed analyses of the natural circulation cooldown including consideration of secondary makeup supplies.

In addition, the impact of steam bubble formation and its effects on depres- surization rates may have to be considered in the analysis of other transients.

This information wilt be requested of the licensee by separate letter and does not have to be resolved prior to restart.

The NRC's Office of Inspection and Enforcement will further review the licensee's corrective actions as well as the acceptability of operating one train of the shutdown cooling.system In the injection mode with the other train operating In the shutdown cooling mode prior to plant restart.

Chris C. Nelson, Project Manager Operating Reactors Branch 13 Division of Licensing Enclosures:

1. List of Attendees

2. Chronology of Events, Evaluation, Actions Taken

& Summary cc w/enclosures:

See last page

I

Enclosure I

JUNE 20, 1980 MEETING WITH FP&L - ST. LUCIE UNIT NO. 1 - ATTENDEES

NRC CE FP&L

Chris C. Nelson Charlie Brinkman C. M. Wethy Themis Speis R. S. Turk J. A. DeMastry Brian Sheron R. S. Daleas H. N. Paduano Ed Jordan J. C. Mouhton Edward Blackwood R. E. Wolf Jerry Mazetis Kris Parczewskl Phil Matthews George Lanik E. V. Inbro R. A. Clark J. S. Cresst2l1 Monte Conner Chang Li Thomas R. Wolf H. F. Conrad

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TIME 0233 0300 0400 0500 0600 0700 0800 0900

% PRESSURIZER LEVEL

11 JUNE 1980

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injection to draw RCS from RWT bubble secure in PZR start 2nd charging pump Level Decrease- LPSI IA ----- O

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VAGSU Or A* ADVISORY CONCIRMING A TECHNICAL DEVELOPMENT 1aLATED TO TII

APPLICATION OX OPERATION OF NUCLEAR PLANT EoutrutaT SUMLIED BY

COMMUIT X E1GINKERING.

un '.0, I1 ILtMURAL CIRCULATICMO COOLOWO

INTRODUCT1ON: On June 12. 19M, C-E was Wnfored by FPSL that wvexpented variations in prissurlzer level here observed while attempting to reduce plant pressure during a natural circulation cooldown at St. Lucit 1. The variations were determined to have resulted from the expansion and collapse of a stama void in the Reactor Vessl (RY)

head region. This bulletin provides details currently known about this avnt and ap- propriate operational guidance for utilities with operating C-E XSns.

A OISCUSSt0x: Following a loss of Coponent CoOTIng Water. to the Rs. St. Lucie I

iced an expeditious natural circulation cooldoin to cotd shutdown conditions.

0- RCS depressurizatlo+n via pressurizer auxi tary spray was initiated approzittely 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the reactor trip..- Oring depressurization, saturation condttions were

3I reachd in the RY head region which Tflited further depressurization until the head region had cooled off. Under natural circulation flow conditions there was aininEl coolant flow in the upper head region and effectively all cooling was by heat loss to

-the ambient or by reactor coolant surge flow through the head region. An RC pressure appropriate for initiating shutdown cooling (SCS) was attained approximately 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> AL following the reactor trip. Ouring this event, the con region was always significantly subcooled.

In the event of a natural circulation cooldowu it should be realized that cooling of -I

I- tha vessel head region ezy be limiting in establishing SCS enry conditions. Cooling of the RY head region my require as auch as 10 to IS hurs. Perfodic ha mxAs Whilt wm~tihfnim fiiv f attemts to inicains af v4id AI aligned

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for7Ation. When SCS entry conditions art achieved, the shutdown coling systse mybe I in accordance with operating procedures.

I Buring depressurizatton. forcation of a stma void In the RY head wuTd be indicated by a rapid increase in pressurizer level while delivering auxiliary spWrayr, k (.4 b decrease inpressurizer level when charging is realitned to the RCS looP The re spotmes are. indicative of an expanding void in the R;J head as the pressurizer is sprayed and collapse of the steam void In the RY head as charging is directed t the loop.

• TheSe indicattons May be prid by an indication of letdon flow in excess of charging flow If the Pressurizer level control syste has beei In utoatic during the cooldwm. Upon indtcatton of a steao void the following actios are recrended tn order to maintain plant conditions until the RY head has Cooled:

-j 1 1. tsolate letIO.

2. Stop further RCS cooldmiu until Indications of RV head steas void cease.

3. Energize pressurizer heaters to limit rate of depressurizatimo.

-

FA 4. Operate char dgre and auxiliary spray to raintain pressurizer levl bet:e

30% and 705 tn order to l imit void expansion.

STATUS: The event is currently being reviewed. Additional operator gufdance will SInorporated Into. C-C tratntng programs and forwarded to utilfty operating staffs as appropriate.

ThE AMXSAM MAW OpWL S7V*W C~ftACr P TH6 APPUCA.W.E FIANF. A A EWHS To You" Oc As A MRISM. ANO San= OP6A4TM OF YOUR PLANT 4 C0&TLVT IWv.9in YOURM a A!L

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