ML18089A511
| ML18089A511 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 01/30/1984 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Uderitz R Public Service Enterprise Group |
| References | |
| TASK-2.K.2.17, TASK-TM NUDOCS 8402090603 | |
| Download: ML18089A511 (12) | |
Text
.
\\
Docket No. 50-272 and 50-311 Mr. R. A. Uderitz JAN
~ Qi 1~J Vice President - Nuclear Public Service Electric and Gas Company P. 0. Box 236 Hancocks Bridge, New Jersey 08038
Dear Mr. Uderitz:
DI STR IBUTI ON
- Docket Fi 1 e L PDR DEisenhut NSIC JTayl or DFischer Gray File NRC PDR ORB#lRdg OELD EJordan ACRS ( 10)
CPa rri sh
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 Salem Nuclear Generating 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 Steven A. Varga, Chief Operating Reactors Branch #1 Division of Licensing 1
. I
Docket No. 50-272 and 50-311 Mr. R. A. Uderitz UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 January 30, 1984 Vice President - Nuclear Public Service Electric and Gas Company P. 0. Box 236 Hancocks Bridge, New Jersey 08038
Dear Mr. Uderitz:
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 Salem Nuclear Generating 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 cot 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 tft f \\*\\, L
~Var~ef Operating Rea~~~r~ranch #1 Division of Licensing
Mr. R. A. Uderitz Public Service Electric & Gas Company cc:
Mark J. Wetterhahn, Esquire Conner and Wetterhahn Suite 1050 1747 Pennsylvania Avenue, NW Washington, DC 20006 Richard Fryling, Jr., Esquire Assistant General Solicitor Public Service Electric & Gas Company P. 0. Box 570 - Mail Code T5E Newark, New Jersey 07101 Gene Fisher, Bureau of Chief Bureau of Radiation Protection 380 Scotch Road Trenton, New Jersey 08628 Mr. John M. Zupko, Jr.
General Manager - Salem Operations Public Service Electric & Gas Company Post Office Box E -
Hancock Bridge, New Jersey 08038
.Mr. Dale Bridenbaugh M.H.B. Technical Associates 1723 Hamilton Avenue San Jcse, California 95125 Leif J. Norrholm, Resident Inspector Salem Nuclear Generating Station U.S. Nuclear Regulatory Commission Drawer I Hancock Bridge, New Jersey 08038 Richard F. Engel Deputy Attorney General Department of Law and Public Safety CN-112 State House Annex Trenton, New Jersey 08625 Richard B. McGlynn, Commission Department of Public Utilities State of New Jersey 101 Commerce Street Newark, New Jersey 07102 Salem Nuclear Generating Station Units 1 and 2 Regional Radi~tion Representative EPA Region II 26 Federal Plaza New York, New York 10007 Mr. R. L. Mittl, General Manager Nuclear Assurance and Regulation Public Service Electric & Gas Co.
Mail Code Tl6D - P. 0. Box 570 Newark; New Jersey 07101 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussta, PA 19406 Lower Alloways Creek Township c/ o Mary 0. Henderson, Clerk Municipal Building, P.O. Box 157 Hancock Bridge, NJ 08038 Mr. Alfred C. Coleman, *Jr.
Mrs. Eleanor G. Coleman 35 K Drive Pennsville, New Jersey 08070 Carl Valore, Jr., Esquire Valore, McAllister, Aron and Westmoreland, P.A.
535 Tilton Road Northfield, NJ 08225 June D. MacArtor, Esquire Deputy Attorney General Tatnall Building Post Office Box 1401 Dover, Delaware 19901 Harry M. Coleman, Mayor Lower Alloways Creek Township Municipal Hall Hancock Bridge, New Jersey 08038
cc:
Mr. Edwin A. Liden, Manager Salem Nuclear Generating Station Un'i'ts 1 and 2 Nuclear Licensing & Regulation Public Service Electric & Gas Company Post Office Box 236 Hancock Bridge, New Jersey 08038 Mr. Charles P. Johnson Assistant to Vice President, Nuclear Public Service Electric & Gas Company Post Office Box 570 80 Park Plaza - 15A Newark, New Jersey 07101 Mr. David Wersan Assistant Consumer Advocate Office ~f Consumer Advocate 1425 Strawberry Square Harrisburg, PA 17120 Joseph H. Rodriguez, Esq.
Public Advocate - State of New Jersey Department of the Public Advocate Justice Complex - CN 850 Trenton, New Jersey 08625 I
I I
I
I I
\\
MULTI-PLANT ACTION ITEM F-33 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.
~9-06A (ref. l) which, among other things, required all Westinghouse plant licensees to review the actions required by operating procedures for c~ping with transients and accidents with par_ticular attention* to:
- 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 prev~nt 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 steam 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 a 1 so resulted in the i ssuanc*e 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.
Ir. o r-scuss ION 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 thin9s, the initial temperat~re of the fluid in the upper head region and the degress with which *it ~ixes 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
3 is diverted to this region.
For the newer Westinghouse plants there is enough cold leg fluid diverted to make the temperature in the up~er 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 a~ upper fluid temperature tha~
~s 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 fl ow into the upper head region is from the upper down comer 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 heat, so it has a lower density than the cold leg water in the downcomer.
Thus
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 coolan~ pumps will increase the amount of void created in the upper head region.
To make the results of these analyses valid for all Westinghouse-designed 2, 3, a~d 4 loop plants; Westinghouse evaluated the variations in (1) thermal inertia of the upper head region (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 highest power (34}1MWt~) 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.
Steam voids can be created in the upper reactor vessel by either decreasing the pressure below the saturation pressure. at the
5 prevailing fluid temperature (i.e., a depressurization event) or increasing the temperature of the water. above the saturation temper~ture. For all of the anticipated transients, including those
- where the temperature of the water is increased, Reference 5 states:
11Previous 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 ?re brought in contact with the subcooled region of the system.
11 II I. 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.
6 Tne 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 11dead volume 11
- 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 ~lants account for the effects of void formation in the reactor coolant systems.
IV.
CONCLUSION The staff concludes that the voids generat~d in the reactor coolant
- systems of the~e Westinghouse plants during anticipated_ transients 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.
- l.
- 2.
- 3.
- 4.
- 5.
- 6.
- 7.
REFERENCES U.S. NRC, IE Bulletin No.79-06A, "Review of.Operational Errors and.System Misalignments Identified During the Three Mile Island Incident 11
, April 14, 1979.
Check, P. S. "Void Formation in Vessel Head During St. Lucie Natural Circulation Cooldown Event of June 11, 1980, dated August 12, 1980.
I' U.S. NP.C, 11Clarification of TMI Action Plan Requirements";
NUREG-0737; page II.K.2.17-1, dated November, 1980.
U.S. NRC, 11 Natural Circulation Cooldown (Generic Letter No.
81-21 )
11
, dated May 5, 1981.
Jurgensen, R. W.;
11St. Lucie Cooldown Event Report"; WOG-57; April 20, 1981.
Margolis, S. G. and Redfield, J. A.; "FLASH:
A Program for
~igital Simulation of the Loss-of-Coolant Accident";
WAPD-TM-534 ;* May 1966.
I Porsching, T. A. et.al.; "FLASH-4:
A Fully Implicit Fortran IV Program for the Digital.Simulation of Transients in a Reactor Plant"; WAPD-TM-840; March 1969.
i I I,
I* I I
Document Name:
SALEM 1 Requestor 1 s ID:
PAM Author's Name:
Don F.
Document Comments:
RCS during transients voiding