ML20140G526
| ML20140G526 | |
| Person / Time | |
|---|---|
| Issue date: | 09/20/1977 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20140F372 | List:
|
| References | |
| FOIA-85-665, REF-GTECI-A-39, REF-GTECI-CO, TASK-A-39, TASK-OR NUDOCS 8604020449 | |
| Download: ML20140G526 (29) | |
Text
{{#Wiki_filter:_ _ _ _ _ _ _ _ REVISIOt_! 0 TASK ACTION PLAtt SEPTEMBER 20, B// TASK tiU:18ER A-39 Title - Determination of Safety Relief Valve (SRV) Pool Dynamic Loads and Temoerature Limits for SUR Containment Lead Responsibility - Division of Systems Safety /flRR Lead Assistant Director - R. L. Tedesco (Plant Systems) Task lianager: J. A. Kudrick (Containment Systems Branch)
- 1. Procram
Description:
Experience at several B!!R plants with pressure suppression containments has shown that damage to wetwell internal structures can occur during safety / relief valve (SRV) bloudowns as a result of air clearing and steam quenching vibration phenomena. 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 injection 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 boundary. This effect is referred to as the air-clearing phenomenon. In addition to the boundary loads, the air injection and subsequent bubble motion produces pressure waves and water covement within the pool that produce drag loads on components 'n the pool . APPROVED BY TASC,' SEPTEMBER 6, 1977 TASCCOMMENTS!$g0REQSATED, SEPTEMBER L,1W/ O . 8604020449 860114 PDR FOIA FIRESTOB5-665 PDR
0 l 2-Following the air-clearing phase, pure steam is injected into the pool. Condensation oscillations occur during this time period. However, the amolitudes of these vibrations are relatively small at low pool temperatures. Continued blowdown into the pool will increase the pool temperature until a threshold temperature is reached. At this point, steam condensation becomes unstable. Vibrations and forces can 4 - 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 oper: ting plants is to restrict the allowable operating tenperature envelope via technical specifications such that the threshold temperature is not reached. 1 In response to the concern on relief valve loads, letters were sent in 1975 to all licensees of operating SWR 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.,flark I, II and III). As a result of the gener,1c concerns, owner's groups were forned by both fiark 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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- 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 a'ir-clearing-induced loads and the drag loads en 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 case 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 BNR designs and programs unique to particular SRV discharge line configurations.
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E l . With respect to drag loads on subnerged structures for both SRV and LOCA events, a generic analytical model is under development by GE which will be used for all 8'.!R designs. For loads induced by air clear ng, separate i analytical models are under development to describe the two different i types of discharge nozzles of the relief valve discharge lines; a ramshead modal and a quencher model. The ramshead is a " Tee" fitting, whereas the quencher is a multi-branch diffuser type of nozzle. 5 The ramshead model under development by GE is jointly sponsored by both the Mark I an'd Mark II owner's groups. In-plant tests at Mcnticello will provide the necessary confirming data base. 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 configuratio.1al differences in the i 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.
The proposed program conducted by GE to address the elevated pool temperature concern for the ramshead device is based on experimental 1 determination of the threshold temperature. Current technical specifications for operating Mark I plants restricting plant operation below this limit would be sufficient to satisf/ this concern. GE plans to document these additional data to support the current temperature limit in the near future for staff review. i
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- 8. End Products The program as outlined consists of four major tasks, described belcw.
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 ilUREG 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 1 Mark II generic programs. Reports will be issued to the appropriate task manager if such action is necessary. The enclosed detailed schedule indicates those areas where such an intermediate report may be required. The actual need will be determined when more definite schedules are established on the individual programs. As part of the SRV program, revisions as required to the Standard Review Plan will be prepared to procerly reflect the program results. C. Tasks
- 1. Evaluation of the ramshead air clearing load methodology -
This task involves the review and eva'uation of the analytical 4 model and the supporting data base. Upon completion of the review, an acceptable methodology for computatien of design bases loads associated with air clearing will be developed. O . P
- a. Evaluation of analytical model - the GE developed cnalytical 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 btse 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 date, to be supplied by GE in a topical recort, 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 ccmparisons evaluation conducted ir, Task 1.a. O e o e a e e e 4
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- c. Develop air clearing load methodology Based on the results of tasks 1.a and 1.b, load acceptance
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criteria will be developed by the staff for ramshead air clearing induced loads for both Mark I and Mark II designs.
- 2. Evaluations of the Quencher Air Clearing Load Methodology -
Evaluation and review by the staff of the analytical model with the supportin; data base will be_ performed in this task. Currently, the various industry programs indicate that the quencher arm 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, an acceptable methodology for computation of design basis loads will be determined. 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 performed.and -ecorted by GE will form the basis of the staff's review, since the basic approach is anticipated to be similar to the methodology used in the ramshead model . (see Task 1.a).
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- 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 the analytica.1 model (Task 2.a). These data will be supplied to the staff by GE in the form 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 ceasured loads
- Applicability of test data to Mark II designs.
Results of this task will be incorporated int. task 2.a.
- 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 to be conducted at the Monticello plant. The results of these progra 1 will be documented by GE in the form of topical reports. Similar considerations as outlined in l task 2.b will be included in this task. The results of this task will be integrated into Task 2.a.
*/ Caorso is a Mark II plant located in Nort'hern Italy.
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. d. Develop Air Clearing Load Methodology -
Based on the results of tasks 2.a, b and c, load acceptance criteria will be developed by the staff for quencher air clearing loads for both Mark I and II designs.
- e. Evaluate Confirming Mark III In-Plant Test Program and Data -
The staff will review and evaluate the test plans, instrumentation and data of the prototypical in-plant test program. This information will be supplied to the staff by GE .in a topical report. Similar considerations as delineated in task 2.b will be included. 1
- 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 portica 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 evaluate 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: l
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- Development of transient flow fields
- Presence of components within the flow field affecting the field
- Supporting experimental data
- Appifcability to LOCA induced loads
- b. Evaluation of Supporting Data Base -
The staff will review and evaluate the applicability of the data provided by GE for confirmation of the analytical program. It is anticipated that the data base will censist of experimentally derived drag c'oefficients, recent data obtained frem 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.
- c. Develop Submerged Structure Load Methodology -
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.
- 4. Determination of LOCA and ATWS pool Temperature Limits -
This task involves the staff's review and evaluation of GE-supplied supporting test data to confirm established design pool temperature limits for both LOCA and ATHS considerations. Presently, GE has proposed a higher design pool temperature limit for the ATUS event, taking into account the low probability of occurrance. The adequacy of this reduced safety margin as well as the proposed pool tem erature limit,for the design basis LOCA will be reevaluated within this task. Although the primary emphasis will be directed towards the ramshead device, the limits 9 m 8
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11_ 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 summari:ing our review and evaluation,
- 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, Moss 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 review 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 supporting data base to be provided by GE. Based on results of this review, pool temperature limits will be reevaluated and minimum temperature monitoring requirements will be established. l
- 3. NRR Technical Orcanizations Involved A. Containment Systems Branch, Division of Systems Safety
- 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 .
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- 3. Task lb and evaluate the Monticello data excluding fluid structure interactioneffects(FSI)andevaluateapplicabilityofdata to Mark II.
- 4. Task Ic A generic NUREG report will be issued summarizing the acceptance criteria for the ramshead load.
Manpower Requirements - FY 77 .05 Man-years FY 78 .5 Man-years FY 79 .1 Man-years Total .70 Man-years
- 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 model.
- 7. Task 2b ,
Review and evaluate the Caorso test plan and data (excluding FSIeffects).
- 8. Task 2c l
Review and evaluate the Mark I small scale tests and i the Menticello in-plant tests (exc.luding FSI effects) and, l
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- 9. Task 2d Gene'ric fiUREG reports will be issued for both Mark I and Mark II designs.
- 10. Task 2e Review 4
Evaluate the Mark III confirmatory test plan and data (this effort i will be part of a topical report evaluation). Manpower Requirements - FY 77 - 0 Man-years FY 78 .5 Man-years FY 79 .4 Man-years Total - 0.9 Man-years 1
- 11. Task 3 Has total responsibility for establishing an acceptable methodology to compute submerged structure drag loads due to SRV actuation and LOCA.
- 12. Task 3a Review and evaluate the analytical model.
- 13. Task 36 Review and evaluate tne supporting data.
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- 14. Task 3c A generic flUREG report will be issued for all BWR designs.
Manpower Requirements - FY 77 .05 Man-years FY 78 .25 Man-years FY 79 .10 Man-years
', Total .40 Man-years
- 15. Task 4, da, ob Has total responsibility for the review and evaluation of supporting information supplied by GE to confirm the current poo.1 temperature limits for both ramshead and Mark I load mitigating devices. Input wil; be provided for the ATUS evaluation report. A generic fUREG 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 3 .
l I I 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 appifcations.
- 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 pure forcing function can be obtained. A report will be 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 pure forcing function can be obtained. A report will be provided to the Task Manager sumnarizing the results of this task. (Due to the similarity c.~ this task witn SES'srtask associated with the Caorso test FSI-evaluation, coordina,t' ion between these efforts will be needed). ,
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t' ( Manpower Requirements - FY 77 .04 Man-years FY 78 .6 Man-years FY 79 .3 Man-years Total .94 Man-years D. Structural Engineering Branch, Division of Systems Safety
- 1. Task 2b Has responsibility for determining the FSI effects associated with the Caorso test series. If the FSI effects are significant, methods will be developed by which the pure forcing function can be obtained. A recort 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 .3 Man-years FY 79 - .2 Man-years Total .6 Man-years E. Division of Project Management -
- l. Tasks No.1 through 4 Provide coordination between the Division of Systems Safety, I the liark I and Mark II licensees /acolicants, and the Division of l
.. Project Management croject managers for the indiv.idual Mark I, II l and III BWR facilities. This includes meetino coordination *and
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-17 preparation of geeting minutes to document the actions of the generic SRV review when the owners are involved.
- 2. Manpower Requirements -
FY 1978 - .1 ffan-years FY 1979 - .1 Man-years Total - .2 Han-years 4 Technical Assistance Recuirements A. Brookhaven National Laboratory 1.
Title:
BilR Pool Dynamic Technical Assistance Program
- 2. Responsible Division / Branch: Division of Systems Safety /
Containment Systems Branch
- 3. Scope The contractor is to provide technical expertise in the t evaluation of all analytical models provided for review in all four major tasks. (Tasks la, 2a, 3a, 4b). In addition, he will provide an independent assessment of the available test data. (Tasks lb, 2b, 2c, 2e, 3b, 4a). Upon the completion of each specific model or test review, a letter report will be istjed to the staff for each of the above noted task items.
During the course of the review, recuests for additional information will also be issued,;as.recuired. e e
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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 Division / Branch: Division of Operating Reactors /
Engineering Branch.
- 3. Scope -
This is a program to study hydrodynamic / structure interactions in a Mark I containment system subject to hydrodynamic loading conditions. This effort should quantify the amplification, if any, of measured 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. e g 9 e G _-,,y
- 4. Funding: FY 197'7 - 100K (fiOTE: This funding represents the total program which are reflected FY 1978 - 15K also in Task A-7).
- 5. Interactions with Outside Orcanizations:
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 comittee as the task progresses.
- 6. Assistance Recuirements from Other ::RC Offices:
Requirements for assistance are not anticipated at this time.
- 7. Schedule for Problem Resolution 1.1 Interim Ramshead Load Criteria 2/78 1.2 Report of FSI Effects 8/78 1.3 SER for Ramshead 11/78 2.1 Report of FSI Effects 9/78 2.2 SER for Quencher 2/79 3.1 Interim Submerged Structure Load Criteria 2/78 3.2 Final Submerged Structure Load Criteria 3/79 4.1 Reevaluation of Ramshead Pool Temperature -
2/78 Limits - 4.2 Final Criteria for Pool Temperature Limits 6/78 l
, 5.0 Issue Revisions to Standard Review Plan 6/79 I
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- t' . B. Detailed Schedule Bar chart enclosed C. Technical fa=ignment Control fiumber - TAC 4671.
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- 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 cwner's programs. Since there are several tast programs involved, past . performance would indicate a good possibility in schedule slippages in one or two tasks. This may necessitate a'ditional d in-plant testing on lead Mark II plants prior to completion of the SRV ger.eric program. B. Fluid structure interaction effects are an important consideration in the evaluation of both ramshead and quencher test data. A technical assist::ce program has been initiated for ltark 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. O
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I e i ' , i l l i SRV PROGRAM StilEDULE l
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. , 77 78 < 79 6 9 3 6 9 3 6 9 Task 1.0 Ramshead Evaluation - - - - -" -- o l 1.a Model Evaluation -
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d, l Data Conparisons , - - -
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s. 1.b Data Evaluation :t- - - -- -- -- -- - - a l FSI Ef fects 4- - - - -- -- - - - - 1.c final Report - - n i. T l l Task 2.0 Quencher Evaluation - - - - - - - - - - - - -
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2.a Model Evaluation ::- - -- - - Data Conparisons - - - - 2.h Caorso Data Evaluation -- -- - -- -- - - -- -- - --
--a -n FSI Effects - --
i 2.c Mark i D'a ta Evaluation Small Scale nn Monticello - - -
-n- - -- i FS! Effects - - - - - - - - -
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i 1 77 78 6 79 0 3 6 9 3 6 9
! 2.d Final Report j n ' 2.e Mark !!! Data .
a.. Evaluation I' .-.
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l i , t I i Tast 3.0 Sutwierged Structures l - - - -- ' n
- - - - - - - - - ! : l i t '. j 1 3.a Flodel Evaluation :H - - --
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Data Confirmation - -
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i o 3.b Data Evaluation --
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- -;; i i 3.c Final Report i a6 . t
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- i i lask 4.0 Pool Temi.. Limits --- I l 4.a Data Evaluation h ; '
n--- G ., l 4.b .Model Evaluation i R-- o ' { 4.c Final Report 6 9 12 3 6
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9 12 ' 3 6 9 12 3 j HONTil
- Indicates possible intermin acceptance criteria .
i X Indicates receipt of key documentation from either 4 the Mark I or Mark II owner's programs or GE ' 5
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OFFICE OF NUCLEAR REACTOR REGULATIDN , REQUEST FOR GENERIC SCilEDUI.E C:IANGE , INITI ATOR: DATE: f - (t} ,
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T A S K N'J. RESPONSIBILITY FOR CHANGE IN SCHEDULE: GEFd Rf C TASK- Task Mgr. DSE (specify branch)
] DPM .
(specify branch) DOR (specify branch) LEAD DIVISION DSS Other (specify branch) (sp,c,gy ) REASON FOR CHANGE: POTENTIAL FOR RECOVERY:
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MILESTONES AFFECTED: (TACS Level "C" Attached) First MS Target Estimated Variance Affected Descriotion Date Date (Work Days) l 4 1 C Last MS - Affected . l Lead Division Comments: CONCURRENCE: Tsk. Mgr.: 0;R SE: tDIR SS: DATE: DATE: DATE: Lead Supv.: DIR PM: - DIR OR: DATE: DATE: DATE: i IMPACT DATA (Prepared by OMIPC) j Generic Review Completion Date
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New Generic Review Completion Date N/ L) L/ O , l
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l ENCLOSURE 2 i s- , e-=p PROCEDURES AND RESPONS181LITIES FOR : REVISING TASK ACTION PLAuS d:'a't yl hII4 i 11J i Approved Task Actions Plans delineate NRR's plans for the resolution of 4 sp,ecific generic technical issues or for developing improvements in the licensing process. The information in the Task Action Plans will be
; used by the Task Managers and Lead Supervisors in executing their tasks, by line managers in planning their resources and by other organi-zations, such as the ACRS, in their interactions with URR regarding the specific issues. In addition, the Task Action Plans have been made available to the public and the industry as a means of informing them of our priorities and plans with regard to particular generic issues.
In light of the important role played by the Task Action Plans, it is imperative that the plans be updated on a continuing basis. The guide-lines and procedures outlined below should be used whe.n revisions to approved Task Action Plans are necessary. i THRESH 0LD FOR TECHNICAL ACTIVITY STEERING COMMITTEE APPROVAL OF TASK ACTION PLAll REVISIONS Revisions to Task Action Plans containing the following types of changes shall require the approval of the Technical Activities Steering Committee: f I 1 4 4 e I
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1 Enclosure 2 t% *A A ; ,'J 9./.r .j m
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- 1. f4ajor modifications to the problem description or to the approach for resolution (This would be required for major revisions to Section 1 or 2, but would also be required for major revisions to Section 3, 4, 5, or 6).
Example 1: A proposal to change the principal approach of using input from RES to acquiring the same needed input through a Technical Assistance contract and additional f-lRR staff participation in the task. i Example 2: A proposal to add an additional subtask to the program to address the generic aspects of an event which occurs at an operating reactor.
- 2. Requirements for additional manpower to execute the task (Section 3).
- 3. Requirements for additional Technical Assistance Funding to exe-cute the task (Section 4).
. Example: An example of 2 or 3 would be a proposal for add'i-tional manpower or technical assistance funding, the need for which is identified as the details of the task unfold.
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Enclosure 2 m m d 4J ". ;i iAJd b ej j . NOTE: Committee approval of schedule changes will be accomplished a via the Generic Book schedule change procedures described in Enclosure 1. The Steering Committee will consider proposed revisions at its next scheduled meeting. REVISIONS TO TASK ACTION PLANS THAT DO NOT REQUIRE STEERIflG COMMITT Revisions to Task Action Plans that are judged to be of less significance than those types identified above can be made with the approval of the Director of the Lead Division. The goal should be to keep the Task Action Plans as current as possible so that they will be useful internally in the execution of the tasks and they will present a current picture of our on-going generic activities to other NRC offices, the public, the industry,
- and other interested parties. Schedule changes, although they must be approved by the Steering Committee, via the Generic Book process, should '
be reflected in the Task Action Plans periodically by using this process for Task Action Plan revisions. i e
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Enclosure 2 r' * :1 ., r.. i,j.: L :j g PROCEDURES FOR REVISING TASK ACTION PLANS 1 Fcr Task Action Plan revisions requiring Steering Committee approval,-the f6' .owing procedures should be used:
- 1. A memorandum from the Director of the Lead Division to the Steering Committee should be prepared explaining the proposed changes and attaching the proposed revisions to the Task Action Plan. In addition, the memorandum should indicate that the participating organizations affected by the revision agree with the revision.
- 2. The revised Task Action Plan should be clearly marked on the front page as " PROPOSED REVISION X" and revised portions should be noted with a vertical line in the right-hand margin and a number indicating the revision.
- 3. Following Steering Committee approval and incorporation of any Committee coaments, the revised Task Action Plan should be clearly marked, " REVISION X" and dated the date of approval (or the date comments were incorporated if approved with comments) and transmitted to the Secretary of the Technical Activities Steering Committee (Mike Aycock, P-436) for standard internal i
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P Enclosure 2 -i tz man Ua b >i 3,,1 l3 l l and external distribution. The vertical lines marking the revisions with
, the revision numbers should be retained in the published copy.
For Task Action Plan revisions that.do not require Steering Committee i approval, the following procedures should be used: i j 1. A memorandum from the Lead Supervisor to the Director of t the Lead Division should be prepared explaining the proposed changes and attaching the proposed revisions to the Task l Action Plan. In addition, the memorandum should indicate that participating organizations affected by the revision agree with the revision or, in the case of a schedule change, that the Steering Committee has approved a " Request for Schedule Change". 2. The revised Task Action P7an should be in the same form as 2 above. '
- 3. Following approval by the Director of the Lead Division, the
. revised plan should be transmitted in the same form as indicated in 3 above to the Committee Secretary for standard internal and external distribution. .
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O r ROUTlHC AND TRANSMITTAL SLIP
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,1EMARK5 May I have your comments by the date indicated on the attached.
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- Do NOT use this form as a RECORD of approvals, concurrences.
disapprovals, clearances, and similar actions
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W. Minners 11/10/77 a*- 27360 OPTIONAL. FORne 48 augusttest ..wi.-eisse-i sas-se ... scat.tos GSA P rod N 4 a t C P Rf 109 1120.
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