ML19263D023
| ML19263D023 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Dresden, Peach Bottom, Browns Ferry, Nine Mile Point, Fermi, Hope Creek, Cooper, Pilgrim, Vermont Yankee, Duane Arnold, Quad Cities  |
| 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 |
| References | |
| TASK-06-02.A, TASK-6-2.A, TASK-RR NUDOCS 7903200443 | |
| Download: ML19263D023 (22) | |
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UNITED STATES y%
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{ %J%*/ l W ASHINGTON, 0. C. 20555 sg,.....f MEMORANDUM FOR:
D. Ziemann, Chief, Operating Reactors Branch 52,00R T. Ippolito, Chief, Operating Reactors Branch 13, 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 IMPLEMENTATION 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 Mark 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 schedules and ccmmitments from each 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 ccmpletion date for plant modifications, (2) the use of quencher safety-relief valve oischarge 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 March 9,1979. A modified form of one of these letters should also be sent to the Detroit Edison Company (Fermi Unit 2) and to the Public Services Electric and Gas (Hope Creek Units 1 and 2) to cbtain the appropriate commit-ments relative to the resolution of the suppression pool hydrooynamic load issues for the non-operating Mark I f acilities.
790.3200443
. The sample letters are available on the Vydek, and tne enclosures will be proviced separately. Should you have any questions on tnis action, contact C. Grimes (X27110).
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. K. Grimes, Assistant Director for Engineering & Projects Division of Operating Reactors
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R. H.' Vollmer, 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. Verre111 P. Kreutzer S. Sheppard C. Grimes
ENCLOSURE 1 To all BWR licensees except Humboldt Bay, Big Rock Point, LaCrcsse, Dresden 1, Monticello, Oyster Creek, Hatch 1 & 2, Brunswick 1 & 2, and Fit: Patrick.
Gentlemen:
RE: SCHEDULE FOR THE IMFLEMENTATION AND WESOLUTION OF THE MARK I CONTAINMENT LONG TERM PROGRAM The generic aspects of the Mart I 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 Jefinition Report (LDR) and Plant Unique Analysis Applications Guide LPUAAG) for flay 1979. Upon the completion of our review of the LDR and PUAAG, we will advise the Mark 1 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 suomitted 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 completion of the LTP.
At this point in the program, you should 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 that, 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 modifications. Your schedule should be directed toward the completion of as many of tne needed plant modifications as possible by Decemoer 1980. Should you be unable 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.
4 2
An issue that relates to your LTP implementation schedule is the use of "ramsheac" devices for safety-relief valve discharge. The enclosea 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 cevice has been shown to significantly ir. prove Doth the loading on the con-tainment and the condensation stability. Tne quencher device has been shcwn to provide the necessary improvements in the containment load-ing and the condensation stability, and you have informally advisec us of your intention to install quencher discnarge devices in your f acili ty.
Please identify when the quencner 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 will De based on tne manpower required to review the LOR ana PUAAG. The responsibility for this fee will be shared by the Owners Group as a whole.
In 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 the LTP plant-unique analysis and any related changes to the plant Technical Specifications.
As discussed above, your cetailed schedule for modifications should be submitted within 60 days following your receipt of this letter.
If you so cesire, we will meet with you to discuss your specific plant modification schecules.
V. Stello, Jr., Director 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 FitzPatrick Gentlemen:
RE: SCHEDULE FOR THE IMPLEMENTATION AND RESOLUTION OF THE MARK I CONTAINMENT LONG TERM PROGRAM The generic aspects of the Mark I Contairment 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 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 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 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, within 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 modifications, your response should include sufficient justification to demonstrate your best efforts.
a
. An issue that relates to your LTP implementation schedule is the use of "ramshead" devices for safety-relief valve discharge. The enclosed staff evaluation discusses our conclusions regarding the basis for the the definition of the ramshead tnreshold 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. However, we understand that you have requested further discussions regarding the possible use of tne ramsnead 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 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 responsibility for this fee will be shared by the Owners Group as a whole. In addition, a fee will also be imposed on each 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 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 modi fication schedules.
V. Stello, Jr., Director Division of Operating Reactors Office of Nuclear Reactor Regulation
Enclosure:
As stated
ENCLOSURE 3 To BWR Licensess for Monticello and oyster Creek.
Gentlemen:
RE: SCHEDULE FOR THE IMPLEMENTATICN AND RESOLUTION OF THE MARK 1 CONTAINMENT LONG TERM PROGRAM The generic aspects of the Mart I Contairment 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 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 m:Jst be addressed for a satisfactory LTP plant-unique analysis. Your plant-unique analysis should De 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 tne 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. Tnerefore, we request tnat, within 60 days following your receipt of this letter, you provide a bar-chart schedule showing the time periods for the installaticn of specific plant modi fications. 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 o# the major plant moaifications, your response should include sufficient justification to demonstrate your best efforts.
2-Another aspect of tile resolution of tne LTP concerns tne licensing fees required by 10 CFR 170. Tne LTP constitutes a "special project" as defined by that regulation. As such, we have cetermined 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 this fee will be shared by the Owners Group as a wnole. In addition, a fee will also be imposed on each individual utility for the license amencment associated with the LTP. The fee class for the license amenament will be based on the manpower required to review the LTP plant-unique analysis and any related cnanges to the plant Technical Speci fica tions.
As discussed above, your detailed schedule for modifications 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
EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF SUPPRESSION POOL TEMPERATURE LIMITS IN BWR FACILITIES
a
- 1. Introduction and Summary Safety-relief valves (SRVs) in SWR plants are used for reactor vessel pressure relief curing either anticipated plant transients or accident si tuations. These valves are installed on the main steam lines of the reactor system with discharge lines from the valves routed to the suppression pool. When the valves open, the steam is discharged through the piping into the pool where it is condensed. A discharge device, wnicn is affixed to the end of the piping beneath the water level in the pool, serves to mix the discharged air and steam with tne pool water.
The most common discharge device in use today is the ramshead type, which consists of two 90-degree pipe elbows welded together, as shown in Figure 1.
During SRV operation, when air and steam are discharged into the suppres-sion pool, vibratory loads (due to bubble fonnation and subsequent collapse) are imposed on the containment structure and compcr.ents within the pool.
The characteristics and magnitude of the load profile are dependent upon the type of discharge device, the temperature of the pool, and the mass and energy discharge rate.
For the ramshead device, the two most significant loads occur during vent clearing and subsequent steam candensation. When the latter loading condition occurs at elevated pool temperatures, condensation becomes unstable and significantly higher loads result. Because of this pheno-menon, General Electric (GE) has proposed a pool temperature limit for all plants using ramshead devices to avoid operation in this unstaDie conden-sation zone. GE's proposed threshold for unstable concensation is 150'F for the bulk pool temperature and 160 e locally. Justification for the limit was supplied by CE to the staff in the form of topical reports (References 1 and 2). These reports contain the experimental data base used by GE to establish the temperatura threshold. The initial concern arose from an event that occurred at a foreign plant, that caused damage to the containment and suDsequent leakage frcm tne wetwell.
We have recently completed our review of the Gr supplied justification for the pool. temperature limit. We and our consulcants (from BNL and MIT) have concluded (Reference 3) that the data base alone is not sufficient to support the 3 proposed temperature limit because of a lack of full-scale c
SRV ramshead discharge load data. First, the data base consisted of small-scale elDow and straight pipe data as well as small-scale ramshead tests, with no scaling analysis provided to show the direct applicability of such tests. Second, the results showed substant1al 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 ficw-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 suppression pool temp-erature threshold of 150 F cannot be adequately supported with the existing data base for the ramshead discharge device. We can, however, 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 the ramshead device and since actual plant pool tem-peratures could approach the GE-proposed limit, we believe the the ramshead device should be replaced to preclude the unstable 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-based plants. This device was developed to improve the perfornance of SRV discharge 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 Dubbles for rapid mixing and diffusion rather than the jet type of discharge mixing provided by the ramshead device. Thus, the quencher 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 tne " cross" or "T" types, cepending upon their geometrical configuration.
The data base for several quencher-type designs has demonstrated superior performance at elevated pool temperatures. Characteristically, 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 boiling point (i.e., greater than 90 C) without any noticeaole 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 therefore lead to larger safety margins.
o
z* Based on the availabic data, we conclude that a design basis suppression pool temperature limit has not been adequately established for the ramshead device. Furthermore, e believe that, even if full-scale ramshead testing were performed, it is eikely that a temperature limit would be established so that operator action would be required during SRV discharge transients to ensure that the pool temperature limit would not be exceeded.
(Note:
Full-scale ram nead 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 conclude that it should not be used. We also conclude tha* 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 abcut
+2.2 psi.
- Minimum temperature limit for onset of condensation instability.
- Peak positive and negative torus shell loads observed in the Monticello in-plant tests.
. We have considered the Dases for interim operation of the Mark I plants currently using ramshead devices. The SRV loads are cyclical in nature, thereby creating tne 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 cevice is being developed and installed. Although some damage to the torus internals has been observed due to apparent SRV operation, therc has not been a loss of containment integrity or function in any Case.
II. Evaluation of Supporting 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 straight down pipe dis-harge device and no test data for the ramshead device. As a result of our evaluation, we conclude that the data base cia not support the proposed limit.
In response to our request, GE provided additional data (Reference
- 2) tnat 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 evaluation 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 data. Generally, tne test results shcw less than a 2-to 3-degree variation within the test pool.
_ To allow proper interpretation of ne test cata, GE perfornec a test at the Quad Cities plant. The pool was instramented with 18 thermocouples, 6 of wnicn were locatec in tne vicinity of tne discnarge cevice to determine local pool temperatures. The test was conducted by centinuously discharging an SRY into the suppression pool for 27 minutes.
Througnout the transient, tne results showed that the measured local temperature did not deviate frcm the calcu-lated bulk temperature by more than 10 F.
Based on tnic result, GE has suggested that a difference of 10 F Dev en local and bulk conditions be used. We concur witn e
evaluation of tne test data.
Based on this temperatere difference, therefore, the GE-pregased 150"F oulk temperature limit is equivalent to a 160 F local temperature. Test results tnen represent local temperature conci tions. The follcwing cata evaluation is Dased on this assumption.
Witn respect to the quencher device, the magnitude of tne difference between tne local and culk temperatures has not been established cue to the lack of an adequate cata base.
Mcwever, recently performed in-plant tests are expected to prowice the necessary data base. We will centinue our review of nis matter.
B.
Sub-scale Ramsnead anc Elbcw Data Sub-scale tests were performed at Moss Landing Test Facility and in a separate test facility in San Jose, California.
These consisted of seven tests using a ramshead anc 37 tests using a 90-degree elbow. The mass flux ranged from 50 to 195 lem/sq ft-sec. The local thresnoic temperature for steam condensation instability calculated by GE for each of these tests rangec f rem 152 to 176CF for the ramsheac and 146 to 172 F for the elbew.
Basec on the folicwing specific concerns, we concluce tnat the acclicacility of tne suo-scale test data has not been adequately ceronstrated and cannot be supported without accitional testing.
6-1.
Scaling Law Apolication: We know from our experience witn tne Marx I pool swell phenomenon, and. frcm the work that has been cone by the Mark II Owners' Group on steam condensation chugging, that small-scale modeling laws are complex and must be established from fundamental principles and carefully applied in model testing. No su;n modeling laws have been derived for tne SRV discharge phenomenon.
Test facilities were not scaled to simulate an actual plant. Therefore, neither dynamic nor geometrical similarities can be established by the tests. Furtnermore, GE has not justified the assumption that scaling nas no effect on the temperature threshold.
2.
Data Scattering:
Substantial data scattering appears in the sub-scale test results. As noted previously, the temperature threshold ranges from 146 to 176 F.
Wi th such a wide scattering, the probability for the tempera-ture thresold to be below the GE proposed 160 F is relatively high (16ii, of the sub-scale data points fall below the 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 ano yielced 12 data points. The threshold was defined as the pool temperature at which the peak-to-peak pressure oscil10 tion first reached 2 bar (29 psig) outside a circular projection with twice the pipe diameter on the floor of the tank. Resul ts of the tests show that all data points fall below the 160 F limit. However, the straight-pipe discharge is phenomeno-logically cifferent from that of the ramshead device and therefore this data is not applicable.
D.
Plant Operatinal Data The GE memorandum report (Reference 2) provides actual in-plant oata. Five plants have experienced SRV discharge into the suppression pools where temperatures in excess of 100 F were reached witn no reported instabilities. Speci fically, 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.
. Data indicate that Plant A was manually scrammed before the suppression pool temperature reached 110 F following a stuck-open event. The suppression p.ol temperature increased rapidly and reached 165 F when tN re,'ctor 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 temperatuN versus SRV steam mass flux during blowdown.
It also shows the GE-proposed pool temperature limit. It is clear that the plants experience SRV discharges far below the GE proposed pool temperature limit at virtually all mass fluxes except the lowest. Thus, their experience does not provide supgort for the higher mass flux at the GE-proposed limit of 160 F.
III. Discussion of SRV Ouencher Discharge Device Designs In 1972, a fore:,n BWR plant with water pressure suppression containment experienced severe vibratory loads on the containment structure during extended SRV operation at high pool temperatures.
The loads were severe enough to cause damage to the containment shell and components and to result in water leakage from the suppression pool.
Following this incident, extensive experiments were conducted to investigate various alternate discharge configurations. The objective of the investigation was to develop a device that would reduce the hydrodynamic loads during SRV air clearing and provide stable steam condensation. Varied configurations of the discharge device considering more than 20 design parameters were investigated.
Results of the investigation concluded that the quencher-type device yielded superior perfonnance. Some of the test results are provided in a GE topical report (Reference 1).
Figure 1 shows the configuration of a typical cross quencher, which is currently used by all Mark III containments. The aross quencher has four arms with each arm perforated by several ro s of small noles. The tip of each ann is plugged and the device measures approximately 10 feet long from tip to tip. Steam flows through the hub, is distributed among the four anns, and is discharged into the pool. The T-quencher device presently being developec for the Mark I plants is similar to the cross quencher except that it has only two arms that are approximately 20 feet long from tip to tip. The quencher cevice produces a cloud of air or steam mist, whereas the ramshead produr.es large bubbles.
.c 8
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 although the pool temperature approaches boiling point.
Figure 3 shows thq comparison of structural loading functions for quencher and ramshead devices for a 238 GESSAR Mark III plant.
Although these loading functions are not applicable for the Mark I design, they demonstrate that the quencher device, in general, substantially reduces the loads on the containment structure with the magnitude of the load reduction being dependent on the quencher 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 1 cads when using the quencher-type discharge devica. Addi tianal testing on a small scale has also shown the temperature threshold for unstable concensation to increase to about 200 F using the quencher-type device. GE is presently conducting full-scale confirmatory testing of the cross-type quencher device at the Caroso plant in Italy. Additional testing on a full-scale plant has been performed in Japan at the Tokai 2 facility.
IV.
one'1sion m
The suppression pool temperature limit proposed by GE to preclude unstable condensation during SRV discnarge through a ramshead cevice has not been adequately demonstrated. Furthernare, we believe that, even if sufficient full-scale testing of the ramshead device were to be performed to adequately define the suppression pool temperature limit, it is likely ea' the resulting limit would require several operator actions and ;-rhaps an additional margin in the allowable pool temperature durir g normal plant operation to preclude unstable condensation.
The test data that bas been generated to date for the quencher cevices have not exhibited :he unstable condensation observed in the ramshead tests at elevated pool temperatures. These data also demonsrate that the quencher air clearing loads on the containment are substantially lower tnan the loads resulting from discharge thrcugh a ramshead cevice.
Furthermore, based on the limited number of suppression pool temperature
. 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 suporession 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.
^eneral Electric Company, " Test Results Employed by GE for 4
dWR Containment and Vertical Vent Loads," GE Topical Report NEDE-21078-P, October 1975.
2.
General Electric Company Memorandum Report, "170 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 Company, "Information Report Mark III Contain-ment Dynamic Loading Conditions (Final)," GE Topical Report NEDO-ll314-08, July 1975.
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