ML19291C029
| ML19291C029 | |
| Person / Time | |
|---|---|
| Issue date: | 10/16/1979 |
| From: | Hanauer S NRC - TMI-2 UNRESOLVED SAFETY ISSUES TASK FORCE |
| To: | Case E NRC - NRC THREE MILE ISLAND TASK FORCE |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, TASK-A-36, TASK-OR NUDOCS 8001100521 | |
| Download: ML19291C029 (9) | |
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Generic Task A-36 k if' W MEMORANDU!1 FOR:
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S. H. Hanauer, Director Unresolved Safety Issues Program
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SUBJECT:
REVISION 2 TO TASK ACTION PLAN A CONTROL 0F HEAVY LOADS NEAR SPENT FUEL b '" ' ' '
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t Enclosed is Revision 2 to the Task Action Plan (TAP) A-36, " Control of Heavy Loads Near Spent Fuel," which I have approved.
This revision reflects changes that have evolved since the issuance of Revision 1 of TAP A-36 iri May 1978.
The following summarizes the changes:
1.
Change in Task flanager; 2.
Changes to resource projections as reflected in June 12, 1979 memo, S. Hanauer to H. Denton; 3.
A more accurate description of how licensee responses to the generic letter on this issue will be reviewed under Subtask 1, and the extent to which accident assessments would be performed under Subtask 2; and 4.
Noting that a NUREG report will be issued providing a descrip-tion of this study and staff recommendations including an in-plementation plan for operating reactors.
A draft of the NUREG report on this task has been prepared and was circulated for management review on October 12, 1979.
The changes summarized above do not involve any significant changes in technical scope of the task. Accordingly, I do not believe full Steering Committee action is required.
Please advise me as to whether or not you concur, i hkuOnGU
. NaWauer, Director Unresolved Safety Issues Program cc:
See next page
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Enclosure:
As stated
Contact:
H. George X-27173 cc w/ enclosure:
R. Mattson R. DeYoung R. Boyd D. Eisenhut B. Grimes R. Vollmer M. Aycock W. Minners W. Russell G. Lainas V. Noonan R. Houston V. Benaroya R. Ferguson S. Hanauer R. Bosnak G. Knighton H. George 7
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Task A-36 CONTROL 0F 'iEAVY LOADS NEAR SPENT FUEL Lead NRR Organization:
Division of Operating Reactors (DOR)
Brian K. Grimmes Lead Supervisor:
Acting Assistant Director for Systems Engineering Hank George, Plant Systems Branch, 2
Task Manager:
(DOR)
All Reactor Types Applicabili ty:
Subtask 1 October 1,1979 Projected Completion Date:
Completed 2
Subtask 2 October 15, 1979 Completed Subtask 3 November 15, 1979 1 1nn
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Task A-36 Rev. No. 2 September 1979 1.
DESCRIPTION OF PROBLEM Overhead handling systems (cranes) are used to lift heavy objects in the vicinity of spent fuel in PWRs and BWRs.
If a heavy object, e.g., a spent fuel shipping cask or shielding block, were to fall or tip onto spent fuel in the storage pool or the reactor core and damage che fuel, there could be a release of radioactivity to the envin.nment and a potential for radiation over-exposure to inplant personnel.
If many fuel assemblies are damaged, and the damaged fuel contained a large amount of undecayed fission products, radiation releases to the environment could exceed 10 CFR Part 100 guidelines.
Additionally, a heavy object could fall on safety-related equipment, and prevent it from performing its intended function.
If equipment from redundant shutdown paths were damaged, safe shutdown capability may be defeated.
For new plants, NRC requirements for assuring that the dropping of a a heavy load will not prevent shutdown of the plant or cause release of a significant amount of radioactivity are contained in the Standard Review Plan, Sections 9.1.2, 9.1.3, 9.1.4, 15.7.4, and 15.7.5.
The measures in effect at operating plants range from compliance with current criteria to Technical Specifications which prohibit the movement of heavy loads over spent fuel.
This task is to systema-tically review current NRC requirements, operating facility designs and technical specifications to establish the minimum criteria for operating plants that accommodate refueling procedures and reduce the potential and consequences of postulated heavy load drop accidents to an acceptable level. With the advent of increased and longer tenn storage of spent fuel assemblies in the spent fuel pool, there is a need to complete this task in an expeditious manner.
2.
PLAN FOR PROBLEM RESOLUTION The staff actions required to assess and if warranted, to improve existing safety margins are divided into three subtasks as follows:
A.
Subtask 1 - Evaluation of Current NRC Requirements and Available Licensee Procedures We will evaluate the existing NRC requirements (SRP 9.1.2, 9.1.3, 9.1.4,15.7.4,15.7.5, NUREG-0554, and Standard Technical Specifi-cations), licensee procedures, design features, and technical specifications associated with the movement of heavy loads on the refueling floor inside containment and near the spent fuel pool outside containment of operating reactor facilities to determine whether heavy load drops: can be precluded, could result in unacceptable consequences, or should be evaluated on a plant specific basis.
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At the beginning of this task, the docketed infomation was insufficient for this evaluation. 01&E was requested to obtain relative infomation from twelve plants. The staff has reviewed specific facility infomation provided by OI&E on six BWRs and six PWRs related to the controls over the moveraent of heavy 7-loads at those facilities. This review detemined that the available information was not sufficient to identify current practices and design features that may prevent, or mitigate the effects of a heavy load drop accident.
Subsequently, by letter 3
dated June 12, 1979, all licensees were requested to provide information on heavy load movement activities, procedures governing such activities, analyses perfomed ori heavy load or cask drop accidents, and design features wi.ich may affect the potential for a heavy load handling accident.
By December 12, 1978, responses had been received from all licensees of non-SEP plants that received their operating licenser prior to May 17, 1978.
SEP facilities will be revieweo in the XP program.
The licensee information submitted in response to the June 12, 1978 letter will be reviewed by the staff to identify the licensee practices, procedures, technical specifications, and design features at the various operating facilities that are used to prevent, or mitigate the effects of this accident. We will also determine what heavy load drop accidents have been analyzed.
We will determine whether such measures are adequate to preclude a heavy load drop accident, or whether additional measures are required. The existing NRC requirements will be evaluated for adequacy to preclude a heavy load drop accident that may result in excessive release of radioactive material, or damage to safe shutdown systems.
In the course of performing the above reviews and evaluations, we identified the need to perform accident assessments on particular plants or representative plants.
B.
Subtask 2 - Accident Assessment The accident assessments which were perfomed evaluated the potential radiological releases due to ruptured fuel assemblies, the potential for the creation of a critical configuration of fuel due to dropped loads, and the potential for degrading the decay heat removal system capabilites. Upon completion of the assessments, the probabilities and consequences were evaluated to determine required guidelines and criteria to assure safe handling of heavy loads.
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. 1 C.
Subtask 3 - Documentation of Safety Criteria The review group will recommend that the NRC requirements, lg licensee procedures or designs found adequate during the first subtask and the regulatory actions found appropriate during the second subtask, be incorporated into a revision to the SRP l2 which will provide guidance to the staff and the industry on the criteria which must be satisfied to reduce to an acceptably low level the potential for heavy loads causing unacceptable damage to safe shutdown systems, fuel in a storage pool, or 2
fuel in the reactor core during refueling.
The revised SRP can be used in future reviews of new plants.
A NUREG report will be issued providing the results of this task and the recommendations of the staff. An implementation 2
plan for evaluating operating plants against the new criteria will be provided in the NUREG report.
3.
BASIS FOR CONTINUED OPERATION AND LICENSING PENDING COMPLETION OF THE TASK As described in Section 2, this task will provide an evaluation of current NRC requirements and specific information from licensees regarding existing design measures, operating procedures and tech-nical specifications associated with the movement of heavy loads near spent fuel pools inside or outside containment and over the reactor core during refueling. The current NRC requirements and review procedures related to this issue are provided in Sections 9.1.2, 9.1.4, 15.7.4 and 15.7.5 of the Standard Review Plan (SRP).
These SRP Sections provide procedures for review of the spent fuel storage poo). the fuel handling system, radiological consequences of fuel hanuling accidents and spent fuel cask drop accidents.
Regulatory Guides 1.3, " Spent Fuel Storage Facility Design Basis;"
1.29, " Seismic Design Criteria;" and NUREG-0554, " Single Failure p;
Proof Cranes for Nuclear Power Plants" provide additional guidance in this area. Further, the Standard Tcchnical Specifications, included in all new operating licenses, include a prohibition on the movement of loads over spent fuel in the storage pool that weigh more than the equivalent weight of a fuel assembly.
These load restrictions will assure that for new operating plants miscellaneous loads, which 12 have not been reviewed from the standpoint of rigging, will not be carried over stored fuel and will assure that, in the event such loads are dropped, radioactivity release will be limited and critical array will not result from rack distortion.
Although it is our view that continued operation with facility designs, operating procedures and technical specification limita-tions that meet the criteria listed above presents no undue risk to the health and safety of the public, the advent of increased (higher
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. density storage configurations) and longer term storage of spent fuel assemblies in spent fuel -storage pools has caused us to re-evaluate these current requirements.
Any revisions to our current requirements resulting from this Task Action Plan will be available well in advance of the operation of any facility for which a deci-sion regarding the issuance of a construction permit is now pending.
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Older operating facilities that were not reviewed against current requirements utilize a variety of design and administrative measures to minimize the potential for dropping a heavy object over the reactor core during refueling or over the spent fuel storage pool.
In addition to evaluating current requirements, this task is designed to detennine what specific measures are being employed at operating facilities and whether any additional measures are needed.
With regard to operating facilities, the following interim measures are being taken while the task is being completed:
For those facilities requesting increases in spent fuel storage capacity, the NRC is prohibiting the movement of loads over fuel assemblies in the spent fuel pool that weigh more than the equiva-lent weight of one fuel assembly. This is consistent with the Standard Technical Specifications discussed above.
In addition, for those plants where a review of cask drop or the crane handling system has not been completed, movement of shielded casks on the refueling floor will be prohibited until completion of the staff's review.
This will assure that the structural integrity of the fuel pool and other safety systems will not be compromised during the course of the evaluation.
For other operating facilities, as an early step in this task, the NRC will request that each licensee review its current procedures for the movement of heavy loads over spent fuel to assure that the potential for a handling accident that could result in damage of spent fuel is minimized while the generic evaluation proceeds.
2 The NUREG report which contains recommended changes to criteria and required actions for operating facilities,, fill include any recormiended additional interim actions found necessary until completion of changes required to satisfy revised criteria and guidelines.
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. Considering that design provisions and licensee procedures are currently in effect at operating plants for the movement of heavy loads near spent fuel and that additional interim measures discussed above are being taken, it is the staff's judgment that the likeli-hood of a heavy load handling accident which damages enough fuel to result in unacceptatle consequences will be small while the staff's generic evaluation proceeds. This conclusion is supported by the fact that no heavy load handling accidents resulting in damaged fuel have occurred in over 400 reactor years of U.S. operating experience.
4.
NRR TECHNICAL ORGANIZATIONS INVOLVED Overall project mar.agement for this task will remain with the Task Manager. Technical assistance will be provided during all three sub. asks by EEB, PSB and RSS of DOR, by AAB of DSE and by ASB and MEB of DSS.
A.
Subtask 1 - Evaluation of Current NRC Requirements and Available Licensee Procedures EEB, PSB, RSB, MEB, and AAB will be responsible for providing 12 one representative to work with the Task Manager to evaluate the present NRC requirements and the licensee measures in effect regarding the movement of heavy loads at operating facilities.
12 B.
Subtask 2 - Accident Assessment RSB, and AAB, with support from PSB and MEB, will be responsible 2
for perfoming accident assessments of the spent fuel damage resulting from a heavy load falling or tipping into the storage pool or the reactor core during refueling.
EEB and AAB will be responsible for perfoming an analysis of the potential radio-logical consequences both ont,ite and offsite due to the accident parameters identified. Upon completion of these evaluations, the review group established in Subtask 1 will be responsible for combining the probabilities and consequences to detemine whether regulatory action is required and, if so, what action is appmpriate.
C.
Subtask 3 - Documentation of Safety Criteria EEB, PSB, RSB, MEB, AAB and ASB will be responsible for providing one representative to work with the Task Manager to determine how the regulatory actions found to be necessary and appropriate in Subtasks 1 and 2 can be incorporated into the existing SRP's.
After this detemination. is made the review group will prepare the SRP revision for management review and approval.
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Additionally, a NUREG report will be prepared and issued pmviding the results of task A-36 and the task group 1
reconmendations.
Manpower Estimates (June 1 - December 31,1979)
Plant Systems Branch, DOR - E5 man-months including 3.5 man-months for task manager.
Reactor Safety Branch - 1.5 man-months Environmental Evaluation Branch, DDR - 2.0 man-months Accident Analysis Branch, DSE - 2.0 man-months Mechanical Engineering Branch, DSS - 1.0 man-months Auxiliary Systems Branch, DSS - 0.2 man-months 5.
TECHNICAL ASSISTANCE None anticipated.
6.
INTERACTIONS WITH OUTSIDE ORGANIZATIONS Interactions could be considerable during the first Subtask with crane vendors, architect engineers, spent fuel pool rack manufac-turers, as well as licensees to provide the detailed design data, accident frequency data, and operational procedures required for this task.
7.
ASSISTANCE REQUIREMENTS FROM OTHER NRC 0FFICES For Subtask 1, I&E has provided a survey of the procedures in effect at twelve operating plants for the movement of heavy loads near spent fuel pools and the reactor during refueling.
During the second or 2
third Subtask, the Probabilistic Analysis Staff, RES, will b.e requested to provide assistance in reviewing the probabil~ity assess-ment effort. The assistance required should not' exceed one man week of effort.
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8.
POTENTIAL PROBLEMS It should be recognized that the results of the review, particularly Subtask 2, are highly dependent on plant design characteristics and the specific procedures in effect at a particular plant.
It is anticipated that similarities between facilities will justify the selection of two cr three representative facilities which should provide sufficient infomation to bound the assessments being perfomed by the technical branches involved.
If many plant specific assessments are necessary, completion of the second and thus the third subtask will be delayed.
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