ML18141A324
| ML18141A324 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 01/09/1984 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| References | |
| TASK-2.K.2.17, TASK-TM IEB-79-06A, IEB-79-6A, NUDOCS 8401230666 | |
| Download: ML18141A324 (10) | |
Text
Docket Nos. 50-280 and 50-281 January 9, Mr. W. L. Stewart Vice President - Nuclear Operations Virginia Electric and Power Company P. 0. Box 26666 Richmond, Virginia 23261
Dear Mr. Stewart:
e DISTRIBUTION Docket File 1984 ~~~~l Rdg ACRS (lo)
Gray File NRC PDR DEisenhut EJordan DNeighbors L PDR OELD JTaylor CParrish
SUBJECT:
ITEM II.K.2.17, POTENTIAL FOR VOIDING IN THE RCS DURING TRANSIENTS We have completed our review of the subject issue for Westinghouse reactors.
Details of our review may be found in the enclosed Safety Evaluation Report (SER).
For Surry Power Station, Units 1 and 2, we conclude that the voids generated in the reactor coolant system during anticipated transients are accounted for in present analysis models.
Furthermore, based on transient analyses performed by Westinghouse using these models, we conclude that this steam void will not result in unacceptable consequences during anticipated transients.
This completes our actions on the subject issue.
Enclosure:
SER on Item II.K.2.17 cc w/enclosure See next page l
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PDR ADOCK 05000280 1
p PDR ORIGUi.£1.L S1G111ID Bl Steven A. Varga, Chief Operating Reactors Branch #1 Division of Licensing
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e Docket Nos. 50-280 and 50-281 January 9, Mr. W. L. Stewart DISTRIBUTION Docket File 1984 ORB#l Rdg NSIC ACRS (10)
Gray File NRC PDR DEisenhut EJordan DNeighbors L PDR OELD JTaylor CPa rri sh Vice President - Nuclear Operations Virginia Electric and Power Company P. 0. Box 26666 Richmond, Virginia 23261
Dear Mr. Stewart:
SUBJECT:
ITEM II.K.2.17, POTENTIAL FOR VOIDING IN THE RCS DURING TRANSIENTS We have completed our review of the subject issue for Westinghouse reactors.
Details of our review may be found in the enclosed Safety Evaluation Report (SER).
For Surry Power Station, Units l_and 2, we conclude that the voids generated in the reactor coolant system during anticipated transients are accounted for in present analysis models.
Furthermore, based on transient analyses performed by Westinghouse using these models, we conclude that this steam void will not result in unacceptable consequences during anticipated transients.
This completes our actions on the subject issue.
Enclosure:
SER on* Item II.K.2.17 cc w/enclosure See next page ORB#l: DL (\\ q.i.
DNeighbors;ps
' /**. -* I.,...--*
- .~
i tJIGUl'L.L SIG11ZD Bi Steven A. Varga, Chief Operating Reactors Branch #1 Division of Licensing
Docket Nos. ~0~2ao and 50-281 Mr. W. L. Stewart e
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 January 9, 1984 Vice President - Nuclear Operations Virginia Electric and Power Company P. 0. Box 26666 Richmond, Virginia 23261
Dear Mr. Stewart:
SUBJECT:
ITEM II.K.2.17, POTENTIAL FOR VOIDING IN THE RCS DURING TRANSIENTS We have completed our review of the subject issue for Westinghouse reactors.
Details of our review may be found in the enclosed Safety Evaluation Report (SER).
For Surry Power Station, Units 1 and 2, we conclude that the voids generated in the reactor coolant system during anticipated transients are accounted for in present analysis models.
Furthermore, based on transient analyses p~rformed by Westinghouse using these models, we conclude that this steam void will not result in unacceptable consequences during anticipated transients.
This completes our actions on. the subject issue.
Enclosure:
SER on Item II.K.2.17 cc w/enclosure See next page bCJM1~
Operating Reactors Branch #1 Division of Licensing
e
,11 VOIDING IN THE REACTOR COOLANT SYSTEM DURING ANTICIPATED TRANSIENTS IN WESTINGHOUSE PLANTS I.
INTRODUCTION On April 14, 1979, just after the TMI-2 incident, the NRC issued IE Bulletin No.79-06A (ref. 1) which, among other things, required all Westinghouse plant licensees to review the actions required by operating procedures for coping with transients and accidents with particular attention tp:
- a.
Recognition of the possibility of forming voids in the primary coolant system large enough to compromise the core cooling capability, especially natural circulation capability,
- b.
Operator action required to prevent the formation of such voids, and
- c.
Operator action required to enhance core cooling in the event such voids are formed (e.g., remote venting).
On June 11, 1980, a steam bubble formed in the upper head region of a Combustion Engineering plant during a natural circulation
2 cooldown (ref. 2).
The issue of stea~ formation in the reactor coolant system (RCS) of Westinghouse plants was thereafter made part of TMI Action Plan Requirement II.K.2.17 (ref. 3).
The June 11, 1980 event also resulted in the issuance of an NRC
- Generic Letter (ref. 4) which asked all PWR licensees to review their capabilities for performing natural circulation cooldown and to assess the potential for upper vessel voiding during the process, The natural circulation issue, which is now called Multi Plant Action No. B-66, is being evaluated separately.
II.
DISCUSSION Subsequent to Reference 4 the Westinghouse Owners Group undertook a study (ref. 5) to ascertain the potential for void formation in Westinghouse reactors during anticipated transients.
For this
~
study Westinghouse used the WFLASH computer program, which models the RCS with nodalized volumes connected by flow paths. This has two phase flow capability, and tracks voids when they occur.
The potential for voids during transients depends on, among other things, the initial temperature of the fluid in the upper head region and the degress with which it mixes with colder fluid in other parts of the primary system.
In Westinghouse plants the initial upper head temperature depends on how much cold leg fluid
,)
I.
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3 is diverted to this region.
For the newer Westinghouse plants there is enough cold leg fluid diverted to make the temperature in the upper head region essentially equal to the temperature of the cold leg fluid.
However, most currently operating Westinghouse plants have an amount of flow into the upper head region which results in an upper fluid temperature that is between the cold leg temperature and the core outlet temperature.
Since there will be more voiding in the plants with the hotter upper head regions, these are considered to be the limiting case. For these plants Westinghouse conservatively assumed that the initial temperature of the fluid in the upper reactor vessel was equal to the core outlet temperature. Thus, in their analyses of loss of coolant transients with a loss of offsite power, voids form in the upper head region whenever the RCS pressure drops to the saturation pressure corresponding to the initial core outlet temperature.
For Westinghouse plants.with the reactor coolant pumps running, the flow into the upper head region is from the upper downcomer through the spray holes.
The flow out of the upper head region is downward through the guide tubes into the.upper plenum region.
If the reactor coolant pumps are stopped; this flow into the upper head slows, stops, and then reverses direction. This is because the water in the core is heated by the decay he~t; so it has a lower density than the cold leg water in the downcomer.
Thus
- e 4
without the reactor coolant pumps operating, the hot, low-density water in the core is buoyed up through the guide tubes into the upper head region.
This hotter water increases the potential for creating voids.
Thus a loss of offsite power with the consequential loss of the reactor coolant pumps will increase the amount Df void created in the upper head region.
To make the results of these analyses valid for all Westinghouse-designed 2, 3, and 4 loop plants, Westinghouse evaluated the variations in (1) thermal inertia of the upper head re~ion (2) the power level to upper plenum volume ratio, and (3) the guide tube/spray nozzle flow path resistance.
The analyses showed that the thermal inertia of the upper head region is largest for the htghest power (34llMWth) 4 loop plant with an inverted. top hot upper support plate, so this was modeled in the WFLASH program.
It was also determined that the power level to upper plenum volume ratio was essentially the same for all 2, 3, and 4 loop plants and that the guide tube/spray nozzle flow path resistance is less in the 2 and 3 loop plants.
From these evaluations Westinghouse concluded that the results of the transient analyses for steam voiding on a 4 loop 3411 MWth plant with an inverted top hat upper support plate bound those for all Westinghouse plants.
~team voids can be created in the upper reactor vessel by either decreasing the pressure below the.saturation pressure at the
- e e
5 prevailing fluid temperature (i.e., a depressurization event) or increasing the temperature of the water above the saturation temperature. For all of the anticipated transients, including those where the temperature of the water is increased, Reference 5 states:
11 Previous analyses performed for preparation of
~-- safety analyses reported in plant licensing documentation explicitly account for void formation in the upper head region if it is calculated to occur.
The results of the previous analyses indicate no safety concerns are associated with this possibility since voids generated in the upper head would be collapsed when they are brought in contact with the subcooled region of the system.
11 III. EVALUATION Westinghouse has had the capability for calculating the effects of steam voids in reactor coolant systems since the FLASH program (Reference 6) was first developed in 1966.
However, this program was too time consuming for large scale problems such as the calculation of voids in upper reactor vessels during transients. By 1969 Westinghouse had developed FLASH-4 (Reference 7) which, with the more rapid calculating ability provided by an implicit formulation, did allow the calculation of voids in reactor vessels.
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6 The ability to calculate voids was carried into LOFTRAN programs by greatly reducing the velocity of a fixed fraction of the flow, i.e., by creating a "dead volume".
Based on this knowledge and the availability of these computer programs we agree that the analyses performed for the anticipated transients reported in the licensing*documentation of these Westinghouse plants account for the effects of void formation in the reactor coolant systems.
IV.
CONCLUSION The staff concludes that the voids generated in the reactor coolant.
systems of these Westinghouse plants during anticipated transient~
are accounted for in present analysis models.
Furthermore, based on transient analyses performed by Westinghouse using these models, the staff further concludes that this steam void will not result in unacceptable consequences during anticipated transients in any of these Westinghouse plants.
REFERENCES
- 1.
U.S. NRC, IE Bulletin No.79-06A, "Review of.Operational Errors and Syst~m Misalignments Identified During the Three Mile Island Incident", April 14, 1979.
- 2.
Letter, P. S. Check (NRC) to T. Anderson (Westinghouse), .Void Formation in Vessel Head During St. Lucie Natural Circulation Cooldown Event of June 11, 1980," dated Augus.t 12, 1980.
- 3.
U.S. NRC, "Clarification of TMI Action Plan Requirements";
NUREG-0737; page II.K.2.17-1, dated November, 1980.
- 4.
U.S. NRC, "Natural Circulation Cooldown (Generic Letter No.
81-21) 11
, dated May 5, 1981.
- 5.
Jurgensen, R. W.; "St. Lucie Cooldown Event Report"; WOG-57; April 20, 1981.
- 6.
Margolis, S. G. and Redfield, J. A.;
11 FLASH:
A Program for Digital Simulation of the Loss-of-Coolant Accident1';
WAPD-TM-534; May 1966.
- 7.
Porsching, T. A. et.al.; "FLASH-4:
A Fully Implicit Fortran IV Program for the Digital Simulation of Transients in a Reactor Plant 11
- WAPD-TM-840; March 1969.