ML19260C643
| ML19260C643 | |
| Person / Time | |
|---|---|
| Issue date: | 01/03/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19260C640 | List: |
| References | |
| REF-GTECI-A-39, REF-GTECI-CO, TASK-A-39, TASK-OR NUDOCS 8001080191 | |
| Download: ML19260C643 (32) | |
Text
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Revision 1 January 3,1978 TASK ACTION PLAN TASK NUMBER A-39 Title - Determination of Safety Relief Valve (SRV) Pool Dynamic Loads and Temperature Limits for BWR Containment Lead Responsibility - Division of Systems Safety /NRR Lead Assistant Director - R. L. Tedesco (Plant Systems)
(1)
Task Manager:
T. M. Su (Containment Systems Branch) 1.
Proaram
Description:
BWR plants are equipped with relief valves that discharge into the wetwell. Upon relief valve actuation, the initial air column within the SRV discharge line is accelerated by the high pressure steam flow and expands as it is released into the pool as a high pressure air bubble. The high rate of air and steam in'jection flow in the pool followed by expansion and contraction of the bubble as it rises to the pool surface produces pressure oscillations on the pool bouncary.
This effect is referred to as the air-clearing phenomenon.
Experience at several BWR plants with pressure suppression containments has shown that damage to certain wetwell internal structures can occur during safety / relief valve (SRV) blowdowns as a result of air clearing and steam quenching vibration phenomena.
In addition to the boundary loads; e.g., conta*.nment structures, water pedestal, the air injection and subsequent bubble motion produc u pressure waves and water movement within the pool that Troduce drag loads on components in the pool.
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Following the air-clearing phase, pure steam is injected into the pool. Condensn'. ion oscillations occur during this time period.
However, the amplitudes of these vibrations are relatively small at low pool temperatures. Continued blowdown into the pool will increase the pool temperature unt81 a threshold temperature is reached. At this point, steam condensation becomes unstable. Vibrations and forces can increase by a factor of 10 or more if the SRV continues to blow down.
This effect is referred to as the steam quenching vibration phenomenon.
Current practice for BWR operating plants is to restrict -the allowable operating tenperature envelope via technical specifications such that the threshold temperature is not reached.
In response to the concern on relief valve loads, letters were sent in 1975 to all licensees of operating BWR plants requesting that they report on the potential magnitude of relief valve loads, and on the structural capability of the suppression chamber and internal structures to tolerate such loads.
In addition, consideration of these loads has become an integral part of our review of CP and OL plant applications for all BWR pressure suppression containments (i.e., Mark I, II and III).
As a result of the generic concerns, owner's groups were formed by both L1 ark I and II utilities. Through these groups, integrated generic analytical and experimental programs have been developed to address the subject of SRV loads.
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Recently, GE issued a Part 21 notification related to consecutive actuation of multiple safety / relief valves and concomitant load increases for BWR Mark III water pressure-suppression containments.
This concern resulted from a recent study performed by GE of the The primary system pressure response following an isolation event.
results showed that more than one safety relief valve could be actuated consecutively, as a result of a reactor isolation event.
())
This SRV load combination has not been considered in the design.
Discussions with GE have also revealed that this concern is generic to all BWR containments and really concerns itself with the GE code for transient analysis. AB is reviewing the code.
This will have to be completed in a timely matter before any action can be taken.
It also affects operating plants.
As a result of this finding, GE has developed a program for resolution specifically.for Mark III. However, they believe this resolution can be applied generically to all BWR systems.
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~4-2.
Plan for Problem Resolution:
A.
Approach The staff will review and evaluate the results from the Mark I and II programs conducted by the owner's groups and related programs conducted by General Electric Co. (GE).
The approach taken by the owner's groups consists of a number of comprehensive experimental and analytical programs to establish and justify the SRV-related pool dynamic loads for BWR Mark I and II designs. In addition, prototypical in-plant testing is proposed to confirm Mark III SRV loads.
(
For both the air-clearing-induced loads and the drag loads on submerged structures, the Mark I and II programs are based on the development of analytical models which will be confirmed with test data. A series of experimental programs are underway to provide this data base for model verification. Because of differences between the Mark I, II and III designs, the composite program which will be reviewed by the staff consists of both programs common to all BWR designs and programs unique to particular SRV discharge line configurations.
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With respect to drag loads on submerged structures for both SRV and LOCA events, a generic analytical model is under development by GE which will be used for all BWR designs. For loads induced by air clearing, separate analytical models are under development to describe the two different types of discharge nozzles of the relief valve discharge lines; a ramshead model and a quencher model. The ramshead is a " Tee" fitting, whereas the quencher is a multi-branch diffuser type of nozzle.
The ramshead model under development by GE is jointly sponsored by both the Mark I and Mark II owner's groups. In-plant tests at Monticello will provide the necessary confirming data base, i
The basic quencher analytical model also under development by GE will be common to both Mark I and II programs. However, the confirming data bases are different. This is due to configurational differences in the SRV end device.
In-plant tests to be conducted at the Caorso facility in Italy are proposed by the Mark II owner's group as the confirming data base, while, in-plant tests to be conducted at Monticello are proposed by the Mark I owner's group as the confirming data base.
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plans to use a quencher device rather than a ramshe:d is being proposed for Mark II type plants. The type of quencher, however, will be different from that currently proposed for Mark III. The quencher design is now being developed and will be tested in Germany by KWU.
With regard to the concern 6f multiple SRV's subsequent actuations, (1)
GE is currently proposing several alternatives which are aimed to eliminate the possibility of actuating more than one SRV consecutively following a'most severe isolation event. These alternatives, however, have been discussed exclusively for the Mark III containments. As GE indicated, some of these alternatives could be also applicable for Mark I and II containments. GE will submit detailed descriptions and justifications for the approach selected from these alternatives which is considered to meet all design requirements. We will review and evaluate.the information and determine its applicability for the Mark III containments, s
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I For Mark I containments, we have established a short term and long term program for resolution of this particular concern. For the short term, efforts of review of the justification for continued operation are included in the task action plan, " Mark I Containment Long Term Program." The long term program for resolution of this concern, however, is included in this task action plan. We will review (1) and evaluate the approach and justification supplied by GE and the Mark I owners group for plants in operation and plants which have not yet been licensed for operation. With respect to Mark II containment, we have requested each Mark II applicants to provide justification for their plant design to meet the load requirements for multiple SRVs subsequent actuation.
i The proposed program conducted by GE to address the elevated pool temperature concern for the ramshead device is based on the experimental determination of the threshold temperature. Current technical specifications for operating Mark I plants restricting plant operation below this limit would be sufficient to satisfy this concern. GE plans to documen these additional data to support the current temperature limit in the near future for staff review.
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B.
End Products The program as outlined consists of four major tasks, described below.
Upon completion of each task, a NUREG report will be issued. In some cases, this may take the form of input into a more general report (e.g., input into the overall Mark II flVREG report prepared as part of Task A-8). Each NUREG report will be generic in nature outlining the acceptable methodology to be used for computation of plant specific loads.
In addition to the final report, interim acceptance criteria may be necessary to properly interface with both the Mark I and Mark II generic programs.
Reports will be issued to the appropriate i
task manager if such action is necessary. The enclosed detailed schedule indicates those areas where such an intemediate report may be required. The actual need will be determined when more definite ~ schedules are establishea on the individual programs.
As part of the SRV program, revisions as required to the Standard Review Flan will be prepared to properly reflect the program results.
C.
Tasks 1.
Evaluation of Loads Criteria for Ramshead -
l()
This task involves the review and evaluation of the analytical model and the supporting data base. Upon completion of the review, acceptance criteria will be established for ramshead loads on containment structures and components.
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Evaluation of analytical model. - the GE developed enalytical a.
model will be reviewed by the staff from both a theoretical and experimental viewpoint. The model will be evaluated for analytical completeness and experimental comparisons made considering the data base from both Monticello and Quad Cities in-plant tests. The actual experimental comparisons will be provided to the staff in topical reports supplied by GE.
b.
Evaluation of test data - Evaluation of the Monticello test data, to be suoplied by GE in a topical report, will be performed by the staff within this subtask. Areas of consideration will include;
- data scatter
- error band determination
- degree of variations of principal parameters
- fluid structure interaction effects on measured loads
- applicability of test data to plant specific conditions (i.e., applicability to other Mark I designs as well as Mark II designs).
Results of this investigation will be incorporated in the model-data comparisons evaluation conducted ir. Task 1.a.
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g Establish Acceptance Criteria for Ramshead Load.
c.
Based on the results of tasks 1.a and 1.b, and task 5, we will establish acceptance criteria for ramshead loads on containment structures and components.
Included in the criteria are the following:
(1) Loads for single as well as multiple SRV's first actuation; (j)
(2) Loads for single as well as multiple SRV's, if any, for consecutive actuations; (3) Development of SRV design load cases, which will include all representative SRV operational modes; (4) Frequency and fatigue cycles for 40 years of plant life.
2.
Evaluation of Loads Criteria for Quencher Evaluation and review by the staff of the analytical model with the supporting data base will be perfomed in this task.
Currently, the various industry programs indicate that the quencher am configuration will differ between Mark I and II designs.
However, the bubble pattern associated with each arm will be the same. Therefore, it is assumed that the analytical model will-remain essentially the same for both the Mark I and II designs. Upon completion of the staff's review, 1702 332
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acceptance criteria for quencher loads will be (1) established. It should be noted that as part of the overall testing program, prototypical in-plant testing is planned for the Mark III quencher. This program is considered as confirmatory.
The staff effort for review of this program is included in this task but will not impact on the development of the load acceptance criteria since it is confirmatory in nature.
a.
Evaluation of Analytical Model -
The analytical model will be reviewed by the staff both from an analytical and empirical viewpoint. Model-to-data comparisons perfomed and reported by GE will form the basis of the staff's review, since the basic approach is anticipated i
to be similar to the methodology used in the ramshead model (see Task 1.a).
b.
Evaluation of Caorso* Test Data Caorso test data will be reviewed and evaluated by the staff to determine the adequacy of the data base for confirmation of tha analytical model (Task 2.a). These data will be supplied to the staff by GE in the fom of a topical report. Areas of consideration will include:
- Data scatter
- Error band determination
- Degree of variation of principal parameters
- Fluid structure interaction effects on measured loads
- Applicability of test data to Mark II designs.
Results of this task will be incorporated into task 2.a.
+/ Caorso is a Mark II plant located in Northern Italy.
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( c.
Evaluation of Mark I related test data -
The staff will review and evaluate two separate test programs; a small scale test program recently completed to determine relative performance between various quencher designs and an in-plant test program tn be conducted at the Monticello plant.
In addition, a V4 scale test program is being planned to perform a sensitivity study of a parameter used for determining the quencher loads. The results of these programs will be documented by GE in the form of topical reports. Similar considerations as outlined in task 2.b will be included in this task.
The results of this task will be integrated into Task 2.a.
I d.
Establish Acceptance Criteria for SRV Load Based on the results of tasks 2.a, b, c and f, and task 5, we will establish acceptance criteria for quencher loads on containment structures and components.
Included in the criteria are the following:
(1) Loads for single as well as multiple SRVs first (1) actuation; (2) Loads for single as well as multiple SRVs, if any, consecutive actuations; (3) Load cases, which will include all representative SRV operational modes; (4) Frequency and fatigue cycles for 40 years plant life.
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Evaluate Confirming Mark III In-Plant Test Progra'i and Data -
e.
The staff will review and evaluate the test plans, instrumentation and data of the prototypical in-plant test This infomation'will be supplied to the staff program.
by GE in a topical report. Similar considerations as delineated in task 2.b will be included.
f.
Evaluation of Mark II Quencher Design The staff will review and evaluate the test programs, test data and methodology of predicting the quencher loads. The test data and methods of calculating the SRV loads will be supplied to the staff by the applicant as part of the Susquehanna licensing docket. This is the first plant to reference this design. Areas of review will include:
(1)
- Data scatter
- Error band determination
- Fluid structure interaction effects on measured loads
- Adequacy of analytical methodology
- Applicability of test data to the real plant design
- Comparison of the Susquehanna analytical method with the generic quencher analytical method.
Results of this task will be incorporated into task 2.a.
However, a report on the result of our evaluation and acceptance criteria will be issued as part of Susquehanna safety evaluation.
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( 3.
Evaluation of Submerged Structure Load Methodology -
This task involves the staff's review and evaluation of a generic analytical model to be developed by GE to compute the loads on submerged structures due to SRV actuation and LOCA. A portion of the review will involve the evaluation of supporting test data to be supplied to the staff in a topical report. Acceptable load criteria will be developed by the staff as a result of this effort.
a.
Evaluation of Analytical Model -
The staff will review and evalute the generic model developed by GE to compute induced loads on components located within the suppression pool. Particular attention will be directed toward the analytical considerations of the following:
- Development of transient flow fields
- Presence of components within the flow field affecting the field
- Supporting experimental data
- Applicability to LOCA induced loads b.
Evaluation cf Supporting Data Base -
The staff will review and evaluate the applicability of the data It provided by GE for confirmation of the analytical program.
is anticipated that the data base will consist of experimentally 1702 336
. derived drag coefficients, recent data obtained from the 1/3 scale pressure suppression test facility tests and possible future tests which will be documented as part of the Mark I and II owner's group programs.
Develop Submerged Structure Load Methodology -
c.
Based on the results of tasks 3.a and b, load acceptance criteria will be developed by the staff. These criteria will be applicable for all BWR designs.
Determinat' ion of Nomal Plant Transient and ATWS Pool Temperature Limits - (1) 4.
This task involves the staff's review and evaluation of GE-supplied supporting test data to confirm established design pool temperature limits for both nomal plant transient and ATWS considerations.
(1)
I Presently, GE has proposed a higher design pool temperature limit for the ATWS event, taking into account the low probability of occurrence.
The adequacy of this reduced safety margin as well as the proposed pool temperature limit for the normal plant transient will be reevaluated (1) within this task. Although the primary emphasis will be directed towards the ramshead device, the limits for the quencher device will also be included.
In addition, minimum pool temperature monitoring requirements will be determined by the staff. Upon completion of this task, a final report will be issued by the staff summarizing our review and evaluation.
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a.
Evaluate Supporting Data Base -
The staff will evaluate the adequacy of the data base to be provided by GE in the form of a letter report from operating experience, Mo:s Landing tests and tests conducted at General Electric's San Jose facility as well as GE's licensee data (NEDE-21078). Based on the staff's review,-the currently recommended pool temperature limits will be reevaluated for the ramshead device. A similar Teview will be conducted for the Mark I quencher device, b.
Evaluate Thermal Mixing Model -
The staff will review and evaluate the thermal mixing model with its supportir.g data base to be provided by GE. Based on results of this I
review, pool temperature limits will be reevaluated and minimum temperature monitoring requirements will be established.
5.
Evaluation of method fordetermining number of SRV operating consecutively.
Evaluation of this task will include BWR 4, 5 and 6.
Areas of review include:
Primary system pressure response to plant nonnal and abnormai transients which will result in primary system blowdown through the SRV's, Evaluation of SRV control logic.
SRV operational sequence Results of this task will be incorporated into task 1.c and 2.d. for establishing SRV load cases and load combination. A report of our evaluation will also be issued for this particular concern. 1702 338
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3.
NRR Technical Organizations Involved Containment Systems Branch, Division of Systems Safety A.
1.
Task 1 Has overall responsibility for establishing an acceptable methodology to calculate ramshead air clearing loads.
2.
Task la Review and evaluate the analytical model.
3.
Task lb Evaluate the Monticello data excluding fluid structure interaction effects (FSI) and evaluate applicability of data to Mark II.
s 4.
Task ic A generic NUREG report will be issued suuriarizing the acceptance criteria for the ramshead load, and SRV load(1 cases and load combination.
Manpower Requirements -
FY 77
.06 Man-years FY 78
.8 Man-years FY 79
.1 Man-years Total - 1.0 Man-years 1702 339 9
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7-5.
Task 2 Has overall responsibility for establishing an acceptable methodology to compute quencher air clearing loads.
6.
Task 2a Review and evaluate the analytical models.
7.
Task 2b Review and evaluate the Caorso test plan and data (excluding FSI effects).
8.
Task 2c Review and evaluate the Mark I small scale tests, the i
Monticello in-plant tests (excluding FSI effects) and 1/4 scale T-quencher tests.
9.
Task 2d Generic NUREG reports will be issued for the quencher load, (1 )
and STN Icac cases and load combinations.
- 10. Task 2e Evaluate the Mark III confirmatory test plan and data (this effort will be part of a topical report evaluation).
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10.1 Task 2f Has overall responsibility for establishing acceptance (1) criteria for SRV load (:ses and SRV load combinations.
Manpower Requirements -
FY 77
.1 Man-years FY 78
.7 Man-years (1)
FY 79
.7 Man-years Total - 1.5 Man-years
- 11. Task 3 Has total responsibility for establishing an acceptable methodology to computer submerged structure drag loads due to SRV actuation and LOCA.
- 12. Task 3a Review and evaluate the analytical model.
'13.
Task 3b Review and evaluate the supporting data.
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e.
(
Manpower Requirements -
FY 77
.05 Man-years FY 78
.25 Man-years FY 79
.10 Man-years Total
.40 Man-years
- 15. Task 4, 4a, 4b Has total responsibility for the review and evaluation of supporting information supplied by GE to confirm the current pool temperature limits for both ramshead and Mark I load raitigating devices.
Input will be provided for the ATWS evaluation report. A generic NUREG report will be issued '
summarizing the minimum pool temperature monitoring requirements and the acceptable temperature limits for SRV devices. This report will in large part be based on the review of the GE thermal mixing model.
Manpower Requirements -
FY 77
.02 Man-years FY 78
.23 Man-years Total
.25 Man-years 1702 342
( 16. Task 5 Has overall responsibility for integrating and coordinating the offers from several branches involved. Upon completion of the task, a report will be issued summarizing the result of our Result of this task evaluation will be integrated evaluation.
into task 1C and 2d for establishing the SRV load cases.
(1)
Manpower Requirements -
FY 78
.1 man-years Total
.1 man-years 1702
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B.
Plant Systems Branch, Division of Operating Reactors 1.
Task 1 through 4 - Follow the progress of the SRV Program to insure correct application of generic resolutions to specific plant applications.
2.
Manpower Requirements -
FY 77
.1 Man-years FY 78
.2 Man-years FY 79
.1 Man-years Total
.4 Man-years C.
Engineering Branch, Division of Operating Reactors 1.
Task lb
(
Has responsibility for determining the fluid structure interaction effects (FSI) associated with the Monticello tests. If FSI effects are significant, methods will be developed by which the appropriate forcing function can be obtained. A report will be (1) issued to the Task Manager summarizing the results of this task.
2.
Task 2c Has responsibility for determining the fluid structure interaction effects associated with the Monticello in-plant load mitigating tests.
If FSI effects are significant, methods will be developed by which the appropriate forcing function can (1) be obtained. A report will be provided to the Task Manager summarizing the results of this task.
(Due to the similarity of this task with SEB's task associated with the Caorso test FSI evaluation, coordination between these efforts will be needed).
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Manpower Requirements -
FY 77
.04 Man-years FY 78
.6 Man-years FY 79
.3 Man-years Total
.94 Man-years Structural Engineering Branch, Division of Systems Safety D.
1.
Task 2b Has responsibility for determining the FSI effects associated with the Caorso test series. If the FSI effects are significant, Tnethods will be developed by which the pure forcing function can be obtained. A report will be issued to the Task Manager summarizing the task results.
(Coordination with EB will be made with respect to the FSI investigation of Monticello tests).
Manpower Requirements -
FY 77
.1 Man-years FY 78
.2 Man-years
())
FY 79 -
3 Man-years Total
.6 Man-years E.
Division of Project Management a
1.
Tasks No. 1 throuch 4 Provide coordination between the Division of Systems Safety, the Mark I and Mark II licensees / applicants, and the Division of Project Management Droject managers for the individual Mark I, II and III BWR facilities. This includes meetina coordination and 1702 345
I preparation of meeting minutes to document the actions of the generic SRV review when the owners are involved.
2.
Manpower Requirements -
FY 1978
.1 Man-year FY 1979
.1 Man-year Total -
.2 Man-years F.
Reactor Systems Branch, Division of Systems Safety 1.
Task 5 Has responsibility for reviewing and evaluating GE's analyses for primary systems pressure response and the sequence of SRV actuation. A report will be issued to the Task Manager 8
summarizing the task results.
2.
Task 4, 4a, 4b Has responsibility for reviewing and evaluating the assumptions (1) which are related to the primary systems response and used for the evaluation.of pool temperature limits. A report will be issued to the Task Manager summarizing the task results.
Manpower Requirements -
FY 78
.2 Man-years Tctal
.20 Man-years 1702 546
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G.
Instrumentation and Control Systems, Division of Systems Safety 1.
Task 5 Has responsibility for reviewing and evaluating the control logic for SRV actuation. A report will be issued to the Task Manager summarizing the task results.
Manpower Requirements -
FY 78
.3 Man-years Total
.30 Man-years H.
Mechanical Engineering Branch, Division of Systems Safety 1.
Task la, lb, 2a, 2b, 2c and 2f Has responsibility for reviewing and evaluating test data and (I) analytical method relating to the SRV loads on SRV line, discharge device supports and components inside the containment.
A report will be issued to the Task Manager summarizing the task results.
Manpower Requirements -
FY 1978
.2 Man-years FY 1979
.2 Man-years Total
.4 Man-years I.
Analysis Branch, Division of Systems Safety 1.
Task 5 Has responsibility for reviewing and evaluating GE's analysis methods for prie.ary system pressure response to isolation events. A report will be issued tc the Task Manager surrnarizinq the task results.
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I Manpower Requirements -
FY 1978
.1 Man-years Total
.1 Man-years 4.
Technical Assistance Requirements A.
Brookhaven National Laboratory 1.
Title:
BWR Pool Dynamic Technical Assistance Program 2.
Responsible Division / Branch: Division of Systems Safety /
Containment Systems Branch 3.
Scope The contractor is to provide tect.nical expertise in the evaluation of all analytical models provided for review in all four major tasks.
(Tasksla,2a,3a,4b).
In addition, he will provide an independent assessment of the available test data.
(Tasksib,1k,2b,2c,2d,2e,3b,4a). Upon the completion of each specific model or test review, a letter report will be issued to the staff for each of the above noted task items. During the course of the review, requests for additional information will also be issued, as required.
The contractor is also required to provide technical expertise in the evaluation of task 2.f (Evaluation of Mark II Quencher Data). Upon the completion of the evaluation of the task, (1) the contractor will issue a letter report suar.arizing the result.
It should be noted that this task is not included in the current contract. Funding for this task will be requested.
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4.
Funding: FY 1977 - $60,000 FY 1978 - $60,000 (requested)
FY 1979
$15,000 (estimated)
Total -
$135,000 B.
Lawrence Livermore Laboratory 1.
Title:
Structural Hydrodynamic Interactions Technical Assistance Programs 2.
Responsible Divi;h/ Branch: Division of Operating Reactors /
Engineering Branch.
3.
Scope t
This is a Progran to study hydrodynamic / structure interactions in a Mark I containment systen subject to hydrodynamic loading conditions. This effort should quantify the amplification, if any, of Tneasured loads due to the structural interactions during pool swell, SRV discharge, and chugging loading conditions. This is a common technical assistance program for Mark I, Mark II and the SRV task action plans.
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4.
Funding FY 1977 - 100K (NOTE: This funding represents the total program which is reflected also FY 1978 - 15K in Tesk A-7).
5.
Interactions with Outside Organizations Mark I and Mark II Owner's Groups These groups are "ad hoc" organizations of utilities owning either Mark I or Mark II BWR facilities. They have engaged GE as their program manager for resolution of the BWR containment concerns and have designated GE as their primary contact with the NRC during the conduct of these programs.
Advisory Committee on Reactor Safeguards (ACRS)
This task is closely related to one of the generic items identified by the ACRS and, accordingly, will be coordinated with the i
committee as the task progresses.
6.
Assistance Reauirements from Other NRC Offices:
Requirements for assistance are not anticipated at this time.
7.
Schedule for Problem Resolution 1.1 Interim Ramshead Load Criteria 6/78 1.2 Report of FSI Effects 8/78 1.3 SER for Ramshead 1/79 2.1 Report of FSI Effects 7/79 2.2 SER for Quencher 12/79 (1) 3.1 Interim Submerged Structure Load Criteria 6/78 3.2 Final Submerged Structure load Criteria 7/79 4.1 Reevaluation of Ramshead Pool Temperature Limits 4/78 4.2 Final Criteria for Pool Temperature Limits 8/78 5.0 Report of SRV consecutive actuation 9/78 6.0 Issue Revisions to Standard Review Plan 3/80 1702 350
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[
B.
Detailed Schedule Bar chart enclosed C.
Technical Assignment Control Number - TAC 4671.
8.
Potential Problems A.
The proposed schedules have been based to a large part on the current estimates of receipt of key documents from both the Mark I and Mark II owner's programs.
Since thera are several test programs involved, past performance would indicate a good possibility in schedule slippages in one or two tasks. This may necessitate 4
additional in-plant testing.on lead Mark II plints prior to completion of the SRV generic program.
B.
Fluid structure interaction effects are an important consideration in the evaluation of both ramshead and quencher test data. A technical assistance program has been initiated for Mark I related tasks. However, efforts to develop a similar program for Mark II considerations have just begun. Early initiation of this program or incorporation into the existing program is required if successful completion of task 2 is to be realized.
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C.
The current funding for technical assistance has not included the effort for reviewing and evaluating the Susquehanna test program, test data and method of predicting SRV load (Task 2.f).
As we have pointed out, this effort is expected to be substantial. Additional funding is needed. We will finalize the amount of funding needed when the final scope of responsibility becomes available, and will follow the normal procedure to submit our request for approval.
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N a
3.c Final Report o
Task 4.0 Pool Temp. Limits 4.a Data Evaluation 4.b Model Evaluation p
4.c Final Report a
Task 5.0 SRV Consecutive Actuation j
d l
s y
MONTH o
Note:
Indicates possible Intermin acceptance criteria e
X Indicates receipt of key documentation from either W
the Mark I or Mark II owner's programs or GE
.>-