ML15238B031

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Notifies of Generic Concern Re Voiding Transients on B&W Plants.Void Formation May Be Taking Place in Hotter Regions of Reactor Vessel.Requests Analytical Prediction Using Transient Analysis within 30 Days
ML15238B031
Person / Time
Site: Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, Crane  Duke Energy icon.png
Issue date: 01/09/1980
From: Reid R
Office of Nuclear Reactor Regulation
To: Arnold R, Cavanaugh W, Parker W
ARKANSAS POWER & LIGHT CO., DUKE POWER CO., FLORIDA POWER CORP., METROPOLITAN EDISON CO., SACRAMENTO MUNICIPAL UTILITY DISTRICT, TOLEDO EDISON CO.
References
80-002085001004, 80-2085001004, TAC-45218, NUDOCS 8001290042
Download: ML15238B031 (85)
v • d • e


Text

NCE UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 January 9, 1980 6'

TO ALL LICENSEES OF BABCOCK & WILCOX (B&W) PLANTS

SUBJECT:

CONCERN FOR VOIDING DURING TRANSIENTS ON B&W PLANTS Gentlemen:

During the evaluation of the TMI-2 accident an Office of Inspection and Enforcement (I&E) inspector has identified a matter that may be a generic concern for B&W plants.

This matter has been referred to the Office of Nuclear Reactor Regulation.

Based on the I&E inspector's review of normal operational transients in B&W plants, there are a number of indications that suggest void formation may be taking place in the hotter regions of the reactor vessel.

If such is the case, there are a number of potential concerns that could arise. For a more detailed discussion of these concerns refer to Enclosure 1. The data which bear on this matter are provided in Enclosure 2.

We request that you examine the data in Enclosure 2 and evaluate it in light of the concerns expressed in Enclosure 1. In particular, we request:

1. You determine if data from transients at your facility supports the proposed phenomena. If the proposed phenomena are not considered to be the cause of the specified transient characteristics, provide an explanation of the phenomena producing the transient characteristics.

To this end, we request that an analytical prediction be performed of a representative transient using SAR transient analysis methods. The transient selected for comparison should exhibit the characteristics of concern, and

2. In the event that void formation is determined to be responsible for the observed behavior, then a complete evaluation -of the identified safety concerns should be provided.

We request that you provide us with the results of your evaluation within 30 days from the receipt of this letter. If you cannot meet this schedule or you require any clarification of the matters discussed herein, please contact the assigned Operating Reactors Project Manager for your plant.

Sjacerely, Rdbert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors

Enclosures:

As Stated cc w/enclosures:

See next page 8,001 2 9 0 v

BABCOCK.&-WILCOXLOPERATING PLANTS Mr. William 0. Parker, Jr.

Vice President, Steam Production Duke Power Company P. 0. Box 2178 422 South Church Street Charlotte, North Carolina 28242 Mr. William Cavanaugh, III Vice President, Generation and Construction Arkansas Power & Light Company P. 0. Box 551 Little Rock, Arkansas 72203 Mr. J. A. Hancock Director, Nuclear Operations Florida Power Corporation P. 0. Box 14042, Mail Stop C-4 St. Petersburg, Florida 33733 Mr. J. J. Mattimoe Assistant General Manager and Chief Engineer Sacramento Municipal Utility District 6201 S Street P. 0. Box 15830 Sacramento, California 95813 Mr. Lowell E. Roe Vice President, Facilities Development Toledo Edison Company

  • Copies also sent Edison Plaza to all members on 300 Madison Avenue cc lists for each Toledo, Ohio 43652 plant involved.

Mr. R. C. Arnold Senior Vice President Metropolitan Edison Company 260 Cherry Hill Road Parsippany, New Jersey 07054 Mr. James H. Taylor Manager, Licensing Babcock & Wilcox Company Power Generation Group P. 0. Box 1260 Lynchburg, Virginia 24505

ENCLOSURE 1

SUMMARY

OF CONCERNS During operational transients, there appear to be a number of indications in the reactimeter data which suggest that void formation may be taking place in the hotter regions of the vessel.

In particular, the reactor coolant system pressure.decay during the Davis-Besse 1 turbine trip of 9/18/79 clearly illustrate the behavior in question. At approxi mately 15 seconds after reactor trip, the pressure decay rate was suddenly reduced (beginning at about 1785 psig). It is speculated that this sudden reduction is caused by flashing of hot fluid, holding the pressure up.

Secondly, in almost all transients in B&W plants, the pressure decay stops at about 1700 psig, and then begins to rise. Again, the pressure at which the decay stops is thought to be determined as that pressure at which hot fluid in the vessel begins flashing.

It is considered that these regions of hot fluid (i.e., corresponding to steady-state coolant temperatures existing in hot regions of the core) are (1) in relatively stagnant flow regions, and are not readily swept out by cooler fluid during scram; and (2) kept hot by heat transfer from large, hot metal masses of the core upper structures and internals.

While it is realized that the minimum system pressure reached during a scram is hdrmally determined by the TAvp, control and amount of primary coolant shrinkage, the inspector believes that careful examination of all data will reveal discrepancies.

These include:

(1) pressurizer level decrease stopping while TAVG continues to decrease (2) pressurizer level increases at rates in excess of charging capacity (3) pressurizer level and pressure changing out of phase (level increasing while pressure.decreasing and vice-versa)

The safety concerns expressed are as follows:

1.

Present analysis models may nodalize the upper vessel regions too coarsely such that regions of hot fluid are not modeled and void formation not predicted. If void formation indeed is occurring, its effect has not been considered in existing analyses.

2.

The voids believed to form in upper regions of the vessel could migrate to the hot legs, be swept into the "candy cane" and block natural circu lation when the pumps are tripped. For a plant such as Davis-Besse 1, the repressurization that could then occur would shut off the low head HPI flow and recovery would be questionable.

3.

The flashing of hot regions of liquid such that the system pressure hangs up could mask the symptoms of small breaks. Moreover, the pressure hangup occurs at a pressure above the ESFAS actuation pressure and could unaccept ably delay HPI actuation.

(OVER)

ENCLOSURE 2 DATA SUMMARIES

1. Davis-Besse Transient - 9/18/79
2. Davis-Besse Transient -

9/26/79

3. Three Mile Island-2 Transient -

11/3/78

4. Three Mile Island-2 Transient - 3/6/79
5. Three Mile Island-2 Accident 3/28/79

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THREE MILE ISLAND-2 TRANSIENT 11/03/78

THREE MILE ISLAND

.c J

REACTOR TRIP REPORT - UNIT II Date//

j~j me.yi on-/,

7 Cause of trip T<)

L199 ese 1s e

De Vo a.

o T3. Plant conditions prior to trip Power Level y(

Reactor Coolant System PressureZ/.-3~psig Tave oF Reactor Coolant System Flow

/,j)

Makeup Tank Level l

inches Pressurizer Level inches H2 0 RC Boron jw y pm.

FPD Control Rod Positions (withdrawn)

Group 1.0,_%

Group 3 [o0 %

Group 5 fo{

Group 7

1%

Group 2_1V Group 4 t o t Group -

3%

Group 87 TCS Stations in Hand

4.

Evolutions in progress prior to trip fseaC,;

'T ATEL

5.

Corrective actions to prevent reoccurence. T j TLIt e-o ec C6

6.

Time and date next criticality acheived.

7.

Record the reset pressure at which the last main steam relief valve closed.

NOTE:

Use visual observation or chart recorder for this information.

t-ifTatt\\71S MA so0 'PSI Shift Supervisor 5upervisor of

'Operations cc:

' ation/Unit Superintendent ager Generation Operation

.-ce-President Generation (Action Required)

Porc Chairman Nluclear Engineer Ia

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THREE MILE ISLAND-2 TRANSIENT 3/06/79

Reactor Trip/Overspeed Turbine Trip 3-6-79 TMI-2

Section Title 1

Synopsis of Event 2

Initial Plant Conditions 3

Conclusions/Actions 4

Sequence of Events Computer Alarm Printout 5

6 Plots of Major System Parameters 7

Post Trip Review

  • ~ECTTN SYNOPSIS OF EVENTS At 13:24:21 on March 6, 1979, TMI-2 experienced a turbine overspeed trip (IEOSPS) while operating at 98% steady state power.

Later, at 13:26:23, the reactor tripped on negative imbalance (Power/Imbalance/Flow).

Prior to the overspeed turbine trip all operating parameters were normal.

Initial system conditions are listed in Section 2.

The integrated control system began reducing primary power at 20% per minute immfe diately after the turbine trip. During the 20%/minute runback, core negative power imbalance began increasing and after two minutes tripped the reactor when power im balance exceeded -36.1%.

The reactor trip caused RC pressure to immediately drop to 1720 psig, but RC pressure quickly stabilized and returned to normal (>1900 psig) within 10 minutes. Pressur izer level never went below 50 inches. In general, all plant systems operated sat isfactorily throughout the transient.

The only damages were the broken yokes on MS-V26A and MS-V258; and the counterweight for EX-V23A which was found disconnected from the valve. Both yokes were repaired and the counterweight replaced.

Criticality was achieved on March 7, 1979 at 0500 and returned to 98% full power by 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br /> on March 7, 1979.

5 0o

SECTION 2 INITIAL PLANT CONDITIONS Time Zero:

13:24:21 Reactor Power:

98% Rated Thermal Power R.C. Temperature (Tave):

582 0F M.D. *Tank Level:

75 inches R.C. Boron:

1121 ppmB R.C.S. Pressure:

2155 psig R.C.S. Flow:

4 pumps operating Pressurizer Level:

224 inches Effective Full Power Days:

74 EFPD Control Rod Positions:

Group 1 - 5 100% withdrawn Group 6/7 97% withdrawn group 8 27% withdrawn Pressurizer Heaters:

AUTO Pressurizer Spray Valve:

AUTO Integrated Control System:

AUTO Tests in Progress:

None

SECTION 3 CONCLUSIONS/ACTIONS

==

Conclusions:==

1. The reactor trip occured due to exceeding negative power imbalance limit.
2. Integrated Control System response was as expected.
3. Pressurizer level did not go below zero indication.
4. No Technical Specification Limiting Conditions for Operations were violated.
5. All systems operated satisfactorily during the transient.
6. Turbine Overspeed Trip inconclusive.

Actions:

1. Investigate cause for extreme negative power.imbalance and recommend corrective actions to prevent future reoccurrence. (Met-Ed - M.L. Benson)
2. Corrective action for broken yokes on MS-V25B and MS-V26A is presently in progress by GPU.

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SECTION 5 COMPUTER ALARM PRINTOUTS

22.146:0 CIf-II1523 ENDS PMFP' 1U-PL-T-Mltui-IE

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12 14:36 BD 101 "B AIR CIG COILS 13 EI4ER DISCI 911?.

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  • 13:35:38 BAD 1377 FLUX L LEVEL 2 (tJAHONPS).

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U*.'. 13:37:29 CONT..3153 MAI11 GE14ERATOR DIFF *Q

.. NR

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SECTION 6 PLOTS OF MAJOR SYSTEM PARAMETERS

1. Variable Temperature/Pressure
2.

Pressurizer Level

3. Power/Imbalance Envelope
4. Reactivity/Rods/Imbalance A Power graph

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