ML19276E708
| ML19276E708 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Hatch, Monticello, Dresden, Peach Bottom, Browns Ferry, Nine Mile Point, Fermi, Oyster Creek, Hope Creek, Cooper, Pilgrim, Brunswick, Vermont Yankee, Duane Arnold, Quad Cities, FitzPatrick |
| Issue date: | 03/01/1979 |
| From: | Grimes B, Vollmer R Office of Nuclear Reactor Regulation |
| To: | Ippolito T, Vassallo D, Ziemann D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19276E698 | List: |
| References | |
| TASK-06-02.A, TASK-6-2.A, TASK-RR TAC-07934, TAC-7934, NUDOCS 7903200304 | |
| Download: ML19276E708 (8) | |
Text
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nmang*g ga UNITED STATES
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'i NUCLEAR REGULATORY COMMISSION I ='
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WASHINGTON. D. C. 20555 WW~. p/
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1 '"3 MEMORANDUM FOR:
D. Ziemann, Chief, Operating Reactors Branch 42,00R T. Ippolito, Chief, Operating Reactors Branch 93, 00R D. Vassallo, Assistant Director for Light Water Reactors, DPM FROM:
B. Grimes, Assistant Director for Engineering & Projects, Division of Operating Reactors R. Vollmer, Assistant Director for Systems & Projects, Division of Operating Reactors
SUBJECT:
STAFF POSITION FOR THE IMPLE.vENTATION OF THE MARK I CONTAINMENT LONG TERM PROGRAM All operating BWR facilities with the Mart I containment design are currently exempt from the requirements of General Design Criterion 50, Appendix A to 10 CFR 50, while the Mart I long term program (LTP) is being conducted. Our review of the generic aspects of the LTP has reached a point where we consider it appropriate to establish plant-specific scnedules and commitments from eacn utility so that the appro-priate plant modifications can be installed and the program completed.
The enclosed sample letters set forth specific requirements for the schedules and commitments that we consider necessary for the LTP, These requirements address (1) the target ccepletion date for plant modifications, (2) the use of quencher safety-relief valve discharge devices, and (3) licensing fees for the LTP.
These letters are to be transmitted to each licensee of a BWR Mark I facility, as shown in the enclosures, by Maren 9,1979. A modi fiea form of one of these letters snould also be sent to the Detroit Edison Company (Fermi Unit 2) and to the Public Services Electric and Gas (Hepe Creek Units 1 and 2) to cbtain the apprcpriate commit-ments relative to the resolution of the suporession pool hydrocynamic load issues for the non-cperating Mark I facilities.
q903,00 SON
. The sample letters are available on the Vydek, and the enclosures will be provided separately. Should you have any questions on this action, contact C. Grimes (X27110).
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. K. Grimes, Assisthnf Director for Engineering & Projects Division of Operating Reactors kV.
1t R. H.' Voi smer, Assistant Director for Systems & Projects Division of Operating Reactors
Enclosure:
Letters to Licensees cc:
V. Stello S. Nowicki P. O' Conner J. Shea R. Silver H. Smith R. Bevan R. Clark J. Hannon P. Polk V. Rooney D. Verrelli P. Kreutzer S. Sheppard C. Grimes
ENCLOSURE 1 To all BWR licensees except Humboldt Bay, Big Rock Point, Lacrosse, Dresden 1, Monticello, oyster Creek, Hatch 1 C 2, Brunswick 1 & 2, and Fit: Patrick.
Gentlemen:
RE: SCHEDULE FOR THE IMPLEMENTATION AND RESOLUTION OF THE MARK I CONTAINMENT LONG TERM PROGRAM The generic aspects of the Mark 1 Containment Long Term Program (LTP) are nearing completion. We have concluded that it is appropriate at this time to establish specific schedules for the implementation of the plant-unique aspects of the LTP.
We have scheduled the completion of our review of the Load Definition Report (LDR) and Plant Unique Analysis Applications Guide (PUAAG) for itay 1979. Upon the completion of our review of the LDR ana PUAAG, we will advise tne Mark 1 Owners' Group of any specific exceptions to these documents that must be addressed for a satisf actory LTP plant-unique analysis. Your plant-unique analysis should be submitted as soon af ter that time as possible. Following our review of your plant-unique analysis, we will take appropriate licensing action, including a license amencment, to assure the timely completion of the LTP.
At this point in the program, you shoula be in a position to knew the majority of plant modifications that will be necessary to conform to the LTP acceptance criteria. Therefore, we request tnat, within 60 days following your receipt of this letter, you provide a bar-chart schedule showing the tine periods for the installation of specific plant modifice ions. Your schedule should be directed toward the comoletion c; as many of the neeced plant mocifications as possible by December 1980. Should you be unaole to meet this targetea com-pletion date for the installation of the major plant modifications, your response should include sufficient justification to demonstrate your best efforts.
m 2
An issue that relates to your LTP implementation schecule is the use of "ramshead" devices for safety-relief valve discharge. Tne enclosed staff evaluation discusses our conclusions regarding the basis for the the definition of the ramshead threshold temperature (i.e., stability limit). As discussed in this report, the quencher discharge device has been shown to significantly improve both the loading on the con-tainment and the condensation stability. The quencher device has been shcwn to provide the necessary improvements in the containment load-ing and the condensation stability, and you have infomally advised us of your intention to install quencher discnarge devices in your facility. Please identify wnen the quencher discharge cevices will be installed.
Another aspect of the resolution of the LTP concerns the licensing fees required by 10 CFR 170. The LTP constitutes a "special project" as cefined by that regulation. As such, we have determined that the fee associated with the generic aspects of the LTP wili be based on tne manpower required to review the LOR and PUAAG. The responsibility for this fee will be shared by the Owners Group as a whole.
la addi tion, a fee will also be imposed on each individual utility for the license amendment associated with the LTP, The fee class for the license amendment will be based on the manpower required to review tne LTP plant-unique analysis and any related changes to the plant Technical Specifications.
As discussed above, your detailed schedule for modifications snould be sucmitted witnin 60 days following your receipt of this letter.
If you so desire, we will meet with you :o f.iscuss your specific plant modification scnedules.
V. Stello, Jr., Di rector Division of Operating Reactors Office of Nuclear Reactor Regulation
Enclosure:
As stated
ENCLOSURE 2 To BWR Licensees for Hatch 1 C 2, Brunswick 1 C 2, and Fit: Patrick Gentlemen:
RE: SCHEDULE FOR THE IMPLEMENTATION AND RESOLUTICN CF THE MARK I CONTAINMENT LONG TERM PROGRAM The generic aspects of the Mark I Containment Long Term Program (LTP) are nearing completion. We have concluded tnat it is appropriate at this time to establish specific schedules for the implementation of the plant-unique aspects of the LTP.
We have scheduled the completion of our review of the Load Definition Report (LDR) and Plant Unique Analysis Applications Guide (PUAAG) for May 1979. Upon the completion of our review of tne LDR and PUAAG, we will advise the Mark I Owners' Group of any specific exceptions to these documents that must be addressed for a satisfactory LTP plant-unique analysis. Your plant-unique analysis should be submitted as soon af ter that time as possible. Following our review of your plant-unique analysis, we will take appropriate licensing action, including a license amendment, to assure the timely ccmpletion of the LTP.
At this point in the program, you should De in a position to know the majority of pla-t modifications that will be necessary to conform to the LTP acceptance criteria. Therefore, we request that, witnin 60 days following your receipt of this letter, you provide a bar-chart schedule showing the time periods for the installation of specific plant modifications. Your schedule should be directed toward the comp'.etion of as many of the needed plant modifications as possible by December 1960. Should you be unable to meet this targeted com-pletion date for the installation of the major plant modifications, your response snould include sufficient justification to demonstrate your best ef forts.
m 2-An issue tnat relates to your LTP implementation schedule is the use of "ramshead" devices for safety-relief valve discharge. The encloseo staf f evaluation discusses our conclusions regarding the basis for the the definition of the ramshead threshold temperature (i.e., staoility limi t ). As discussed in this report, the quencher discharge device has been shown to significantly improve both the loading on the con-tainment and the condensation staDility. However, we understand that you have requested further discussions regarding the possible use of the ramshead discharge device. We will arrange to discuss this issue with you in the very near future.
Another aspect of the resolution of the LTP concerns the licensing fees required by 10 CFR 170. The LTP constitutes a "special project" as defined by that regulation. As such, we have detennined that the fee associated with the generic aspects of the LTP will be based on the manpower required to review the LDR and PUAAG. The responsioility for tnis fee will be shared by the Owners Group as a whole. In accition, a fee will also be imposed on eacn indivicual utility for the license amencment associated with the LTP. The fee class for the license amendment will be based on the manpower required to review the LTP plant-unique analysis and any related changes to the plant Technical Specifications.
As discussed above, your detailed schedule for mooifications should De submitted within 60 days following your receipt of this letter.
If you so desire, we will meet with you to discuss your specific plant modi fication schedules.
V. Stello, Jr., Director Division of Operating Reactors Office of Nuclear Reactor Regulation
Enclosure:
As stated
ENCLOSURE 3 To BWR Licensees for Monticello and oyster Creek.
Gentlemen:
RE: SCHEDULE FOR THE IMPLEMENTATICN AND RESOLUTICN CF THE MARK I CONTAINMENT LONG TERM PROGR AM The generic aspects of the Mark 1 Containment Long Term Program (LTP) are nearing completion. We nave concluded that it is appropriate at tnis time to establish specific schedules for tne implementation of the plant-unique aspects of the LTP.
We have scheduled the completion of our review of the Load Definition Report (LDR) and Piant Unique Analysis Applications Guice (PUAAG) for May 1979. Upon the completion of our review of the LDR and PUAAG, we will advise the Mark I Owners' Group of any specific exceptions to these. documents that must be addressed for a satisf actory LTP plant-unique analysis. Your plant-unique analysis should be submitted as soon af ter that time as possible. Following our review of your plant-unique analysis, we will take appropriate licensing action, inclucing a license amenament, to assure the timely completion of the LTP.
At this point in the program, you should be in a position to know the majority of plant modifications that will be necessary to conform to the LTP acceptance criteria. Therefore, we request that, witnin 60 days following your receipt of this letter, you provide a bar-chart schedule showing the time periods for the installation of specific plant modifications. Your schedule should be directed toward the completion of as many of 'the needed plant modifications as possible by December 1980. Should you be unable to meet this targeted com-pletion date for the installation of the major plant mocifications, your response should include sufficient justification to cemonstrate your best efforts.
Another aspect of the resolution of the LTP concerns the licensing fees required by 10 CFR 170. The LTP constitutes a "special project" as defined by that regulation. As such, we have detemined that the fee associated with the generic aspects of the LTP will be based on the manpower required to review the LDR and PUAAG. The responsioflity for this fee will be shared by the Owners Group as a whole. In addition, a fee will also be imposed on each individual utility for the license amendment associated with the LTP. The fee class for the license amendment will be based on the manpower required to review the LTP plant-unique analysis and any related changes to the plant Tecnnical Specifications.
As discussed above, your detailed schedule for modifications should be submitted within 60 days following your receipt of this letter.
If you so desire, we will meet with you to discuss your specific plant modification schedules.
V. Stello, Jr., Di rector Division of Operating Reactors Office of Nuclear Reactor Regulation
EVALUt JY THE OFFICE OF NUCLEAR REACTOR REGULATION OF SUPPRESSION POOL TEMPERATURE LIMITS IN BWR FACILITIES
- 1. Introduction and Summary Safety-relief valves (SRVs) in BWR plants are used for reactor vessel pressure relief curing either anticipated plant transients or accident si tuati ons. These valves are installed on the main steam lines of the reactor system with discharge lines frcm the valves routed to the suppression pool. When tne valves open, the steam is discharged througn the piping into the pool where it is condensed. A discharge device, wnich is af fixed to the end of the piping beneath the water level in the pool, serves to mix the distnarged air and steam with the pool water.
Tne most common disenarge device in use today is the ramsnead type, wnich consists of two 90-degree pipe elbows welded togetner, as shown in Figure 1.
During SRV operation, when air and steam are discharged into the suppres-sion pool, vibratory loads (due to bubole fomation and subsequent collapse) are imposed on the containment structure and components within the pool.
The characteristics and magnitude of tre load profile are depencent upon the type of discharge device, the temperature of tne pool, and the mass and energy discnarge rate.
For the ramshead device, the two most significant loads occur during vent clearing and subsequent steam condensation. When the latter loading condition occurs at elevated pool temperatures, condensation becomes unstable and significantly higher loads result. Because of this pneno-menon, General Electric (GE) has proposed a pool temperature limit for all plants using ramsnead devices to avoid operation in tnis unstacle conden-C sation zone. GE's proposed threshold for unstable concensation is 150 F C
for the tulk pool temperature and 160 F locally. Justification for tne limit was supplied by GE to the staff in tne fom of topical reports (References 1 and 2).
These reports contain tne experimental data case used by GE to establish the temperature thresnold. The initial concern arose from an event that occurred at a foreign plant, tnat caused damage to the Containment and subsequent leakage from the wetwell.
We have recently completed our review of the GE supplied justification for the pool temperature limit. We and our consultants (frcm BNL and MIT) have concluded (Reference 3) that the data base alone is not suf ficient to support the GE proposed temperature limit because of a lack of full-scale SRV ramshead discharge load cata. First, tne data base consisted of small-scale eloow and straight pipe data as well as small-scale ramshead tests, witn no scaling analysis provided to sncw :ne direct applicacility of secn tests. Second, the results snewed substantial data scatter.
. Limited plant operational data were also provided, indicating that local pool temperatures of approximately 165 F have been experienced curing a stuck-open SRV event without any evidence of structural damage. This experience can be considered as supporting data for the limited-mass flow-pool temperature zone that occurred. However, it cannot be considered as the operational basis for all potential events.
We have, therefore, concluded that the GE bulk suporession pool temp-erature threshold of 150 F cannot be adequately supported with the existing data base for the ramshead discharge device. We can, b' Myer, conclude that the actual temperature threshold is in the vicinity of the GE proposed limit (i.e., about 150 F).
In light of our current understanding of tne ramshead device and since actual plant pool tem-peratures could approach the GE-proposed limit, we Delieve that the ramshead device should be replaced to preclude the unstaDie condensa-tion phenomena. The basis for this conclusion follows in Section II of this report.
A " quencher" type of device has been used for several years in foreign-Dased plants. This davice was developed to improve the performance of SRV discnarge at elevated pool temperatures as well as to reduce the air clearing loads. The principle behind the quencher-type device is to prcmote the creation of large surface areas of air and steam Duobles for rapid mixing and diffusion rather than the jet type of discharge mixing provided oy the ramshead device. Thus, the quenener consists of pipe sections that contain many small holes, either uniform or graduated along the surface to promote and enhance diffusion and condensation in the pool. The quenchers are typically referred to as either the " cross" or "T" types, depending upon their geometrical configuration.
The data base for several quencher-type designs has demonstrated superior performance at elevated pool temperatures. Cnaracteristically, a quencher-type device has not exhibited the temperature threshold phenomenon that has been observed for the ramshead device, based on the test data gener-ated to date. Pool temperatures have approached the coiling point (i.e., greater than 90 C) without any noticeable load increases. Hydro-dynamic loads on structures during vent Clearing are also reduced, due to the inherently Detter distribution of the steam / air mixture in the pool. The use of the quencher device would tnerefore lead to larger safety margins.
. Based on the available data, we conclude that a design basis suppression pool temperature limit has not been adequately established for the ramshead device. Furthermore, we believe that, even if full-scale ramshead testing were performed, it is likely that a temperature limit would be established so that operator action would be required during SRV discharge transtents to ensure that the pool temperature limit would not be exceeded.
(Note:
Full-scale ramshead testing at elevated pool temperatures to establish a design basis pool temperature limit has not been proposed). Therefore, in the absence of any further information on the ramshead, we concluce that it should not be used. We also conclude that the quencher-type device provides improved safety margins and can be used in all BWR plants with water suppression containments. The comparative benefits are given in the following table:
Table 1 SRV DISCHARGE DEVICE EVALUATION
SUMMARY
local Temperature Air Clearing Device Limit
- Remark s Loads **
Ramshead 160 F
- 1. Test data do not
+21 psi support the pro-posed limit.
-10 psi
- 2. Severe vibration occurs if the limit is exceeded.
Quencher 200 F
- 1. Test data show no
+6 psi severe vibration for tank water tempera-
-5 psi tures approaching boiling.
- 2. Steam condensation loads are about
+2.2 psi.
' Min 1 mum temperature limit for onset of condensation instabili ty.
- Peak positive and negative torus shell loads CDserved in the Mont* cello in-plant tests.
4 We nave considered the bases for interim operation of tne Mark I plants currently using ramshead devices. The SRV loads are cyclical in nature, tnereoy creating the potential for fatigue degradation of the containment. For currently operating Mark I plants, we have determined that there is sufficient fatigue margin to permit continued plant operation while a new discharge device is being developed and installed. Althougn some damage to the torus internals has been observed due to apparent SRV operation, there has not been a loss of containment integrity or function in any case.
- 11. Evaluation of Supoorting Data for Ramshead Device In late 1975, GE submitted a topical report (Reference 1) to support the temperature limit for the suppression pool when using a ramshead device. The report, however, contained test data for SRVs having a straignt down pipe discharge device and no test data for the ramshead device. As a result of our evaluation, we conclade that the data base did not support the proposed limit.
In response to our request, GE provided additional data (Reference
- 2) that contained three sources of test data: subscale test data of ramshead and elbow devices, small-scale test data of straight-down pipes, and plant operational data. Results of our evsluation of this report are discussed below.
A.
Local and Bulk Temoerature Differences Local temperature is referred to as the water temperature that is in the vicinity of the discharge device but not in contact with the steam bubble. Bulk temperature, on the other hand, is a calculated temperature that assumes a uniform pool temperature. Bulk temperature is normally used for pool temperature transient analyses, Because the test facilities are confined pools, the measured temperatures are considered to be local temperatures.
This has been confirmed through evaluation of the test da ta.
Generally, the test results show less than a 2-to 3-degree variation within the test pool.
. To allow proper interpretation of the test data, GE performed a test at tne Guad Cities plant. The pool was instrumented with 18 thermocouples, 6 of anicn aere located in the vicinity of tne discharge device to determine local pool temperatures. The test was conducted oy continuoasly discharging an SRV into the suppression pool for 27 minutes.
inrougnout the transient, tne results showed that tne measured local temperature did not deviate from the calcu-lated bulk temperature oy more tnan 10 F.
Based en tnis result, GE nas suggested that a difference of 10 F cetween local and bulk conditions be used. We concur witn this evaluation of the test data.
Based on this temperature difference, therefore, the GE-pregosed 150"F eulk temperature limit is equivalent to a 160 F local temperature. Test results tnen represent local temperature conditions. The following data evaluation is based on this assumption.
With respect to the quencher device, the magnitude of the difference between the local and bulk temperatures nas not Deen established due to the lack of an adequate data base.
Mcwever, recently performed in-plant tests are expected to provide the necessary data Dase. We will continue our review of tni s matter.
a.
Sub-scale Ramsnead and Elbow Data Suo-scale tests were performed at Moss Landing Test Facility and in a separate test f acility in San Jose, California.
These consisted of seven tests using a ramshead and 37 tests using a 90-degree elbow. The mass flux ranged frem 50 to 195 lom/sq ft-sec. The local tnreshold temperature for steam condensation instability calculated by GE for eacn of these tests ranged from 152 to 176 F for tne ramsnead and 146 to 172 F for the elbow.
Baseo on the following specific concerns, we concluce tnat the a;olicability of the suo-scale test data has not been acecuately demonstrated and canno ce supported witncut addi ticnal testing.
.u. 1.
Scaling Law Apolication: We know from our experience wi tn tne Mark 1 pool swell phencmenon, and f rom the work that has been done by the Mark II Owners' Group on steam condensation enugging, that small-scale modeling laws are complex and must be established frcm fundamental principles and carefully applied in model testing. No such modeling laws have been derived for the SRV discharge phenomenon.
Test f acilities were not scaled to simulate an actual plant. Therefore, neither dynamic nor geometrical similarities can be established by the tests. Furthermcre, GE has not justified the assumption that scaling has no effect on the temperature threshold.
2.
Data Scattering:
Substantial data scattering appears in the sub-scale test results. As noted previously, tne 0
terperature threshold ranges from 146 to 176 F.
Wi th such a wide scattering, the probaDility for the tempera-ture thresold to be below the GE proposed 160 F is -
relatively high (16". of the sub-scale data points fall below tne limit).
C.
Small Scale Straight Down Pice Data This data set was obtained from foreign tests (Reference 1).
The tests used a straight-down pipe and yielded 12 data points. The threshold was defined as the pool temperature at which the peak-to-peak pressure oscillation first reacned 2 bar (29 psig) outside a circular projection with twice the pipe diameter on the floor of the tank. Results of the tests show that all data points fall belcw the 160 F limit. However, the straight-pipe discharge is phenomeno-logically different from that of the ramsnead device and therefore this data is not applicable.
D.
Plant Operatinal Data The GE memorancum report (Reference 2) provides actual in-plant data. Five plants have experienced SRV discharge into the suppression pools where temperatures in excess of 100 F were reached witn no reported instabilities. Specifically, the highest pool temperature from those events ranged from 122 to 165 F.
However, the report only provides detailed data for two plants identified as Plant A and Plant C.
7_
Data indicate that Plant A was manually scrammed before the suppression pool temperature reacned 110 F following a stuck-open event. The suppression pool temperature increased rapidly 0
and reached 165 F wnen the reactor pressure was 184 psig. Plant C reached 146 F only because the reactor was scrammed at a lower pool temperature.
Figure 2 shows the loci of the Plant A and C events on a plot of pool temperature versus SRV steam mass flux during olcwdown.
It also shows the GE-proposed pool temperature limit. It is clear tnat the plants experience SRV discharges far Delcw the GE proposed pool temperature limit at virtually all mass fluxes except the lowest. Thus, their experience does not provide succort for the higher mass flux at the GE-proposed limit of 0
160 F.
III. Discussion of SRV Ouencner Discharge Device Desians In 1972, a foreign BWR plant with water pressure suppression containment experienced severe vibratorj loads on the containment structure during extended SRV operation at hign pool temperatures.
The loads were severe enougn to cause damage to the containment shell and components and to result in water leakage from the suppression pool.
Folicwing this incident, extensive experiments were conducted to investigate various alternate discharge configurations. Tne objective of :ne investigation was to develop a device tnat would reduce the hydrodynamic loads during SRV air clearing and provide stacle steam condensation. Varied configurations of the discharge device considering more than 20 design parameters were investigated.
Results of tne investigation concluded that tne quencher-type device yielded superior performance. Scme of tne test results are provided in a GE topical report (Reference 1).
Figure 1 shcws the configuration of a typical cross cuencher, wn.ich is currently used by all Mark III containments. The cross quencner nas four arms witn each am perforated by several rows of small holes. The tip of each arm is plugged and the device measures approximately 10 feet long from tip to tip.
Steam ficws tnrcugh
- ne hub, is distriouted among the four arms, and is discharged into the ocol. Tne T-quencher device presently being develcped for tne Mark I plants is similar to the cross quencher except tnat it has only two arms that are approximately 20 feet long from tip to tip. Tne quencher cevice procuces a cloud of air or steam mist, wnereas the ramshead produces large Duocles.
a
. Because of the quencher configuration, the magnitude of the quencher air clearing load is reduced by a factor of two to four.
In addition, steam condensation instability does not occur althcugn the pool temperature approaches boiling point.
Figure 3 shows the comparison of structural loading functions for quencher and ramshead devices for a 238 GESSAR Mark III plant.
Althougn these loading functions are not applicable for the Mark I design, they demonstrate that the quencher device, in general, suostantially reduces the loads on the containment structure witn the magnitude of the load reduction being dependent on the quenener configuration and its relative location to the adjacent stuctures.
Foreign large-scale testing and in-plant tests from the United States (Monticello) have verified the reduction in hydrodynamic loads when using the quencher-type discharge device. Addi tional testing on a small scale nas also shown the temperature threshold for unstacle condensation to increase to about 200 F using the quencher-type device. GE is presently conducting full-scale confirmato y testing of the cross-type quencher device at tne Caroso plant in Italy. Additional testing on a full-scale plant has been performed in Japan at tne Tckai 2 facility.
IV. Conclusion The suopression pool temperature limit proposed oy GE to preclude unstable condensation during SRV discnarge througn a ramshead device has not Deen adequately demonstrated. Furtnermore, we Delieve that, even if sufficient full-scale testing of the ramsnead device were to be performed to adequately define the suppression pool temperature limit, it is likely that the resulting limit would require several operator actions and perhaps an additional margin in tne allowaole pool temcerature during normal plant operation to preclude unstacle condensation.
The test data that has been generated to date for the cuencher cevices have not exhibited the unstable condensation observed in the ramshead tests at elevated pool temperatures. These data also cemonsrate nat the quencner air clearing loads on the containment are substantially lower than the loacs resulting from distnarge tnrough a ramshead cevice.
Furtnermore, based on the limited number of suppression pool temperature
9 transient analyses that we have received for Mark I plants, it appears that a lesser amount of operator action would be required.
Based on the improved performance demonstrated for the quencher discharge devices and the uncertainty associated with the defini-tion of the pool temperature limits for ramshead discharge devices, we conclude that the use of ramshead devices in BWR water suppression containment systems is not acceptable for long-term operation. We also conclude that the quencher-type devices provide a satisfactory resolution to the condensation stability concerns and is, therefore, an acceptable replacement.
References 1.
General Electric Company, " Test Results Employed cy GE for B'aR Containment and Vertical Vent Loads," GE Tcpical Report MEDE-21078-P, October 1975.
2.
General Electric Company Memorandum Report, "17d'F Pool Temperature Limit for SRV Ramshead Condensation Stability,"
September 1977.
3.
Ain A. Sonin and C. Tung, " Comments on the Pool Temperature Limit for Avoiding Pulsating Condensation with Ramshead SRVs,"
Brookhaven National Laboratory, February 1978.
4.
General Electric Ccmpany, "Information Report Mark III Contain-ment Dynamic Loading Conditions (Final)," GE Topical Report NED0-11314-08, July 1975.
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