ML19291C019
| ML19291C019 | |
| Person / Time | |
|---|---|
| Issue date: | 10/05/1978 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML19291C018 | List: |
| References | |
| REF-GTECI-A-07, REF-GTECI-CO, TASK-A-07, TASK-A-7, TASK-OR NUDOCS 8001030473 | |
| Download: ML19291C019 (5) | |
Text
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PROPOSED REVISION 2 a
Task A-7 MARK I CONTAINMENT LONG-TERM PROGRAM (LTP)
Division of Operating Reactors (DOR)
Lead Supervisor:
Darrell G. Eisenhut, A/D for Systems and Projects, DOR Task Manager:
Chris Grimes, D0R Applicability:
Boiling Water Reactors with Mark I Containments Projected Completion Date:
October 1979 1J90 232 q,3
1.
DESCRIPTION OF PROBLEM During the conduct of a large scale testing program for an advanced design pressure-suppression containment system (Mark III) for 43WRs, new suppression pool hydrodynamic loads associated with a postulated loss-of coolant accident (LOCA) were identified which had not been explicitly included in the original design of the Mark I containment systems.
These additional loads result from dynamic effects of drywell air and steam being rapidly forced into the suppression pool (torus) during a postulated LOCA event.
In addition, recent expe-rience at operating olants has indicated that the dynamic effects of safety-relief valve (SRV) discharges to the suppression pool could be substantial and should be reconsidered.
The resd ts of the Mark I containment short-term program (STP) have provided assurance that the Mark I containment system of each operat-ing BWR facility would maintain its integrity and functional capa-bility during a postulated LOCA.
However, the STP evaluation was conducted using a "most probable load" approach which was aimed at the identification of load magnitudes and load combinations which were most likley to be encountered during the course of a postulated design basis LOCA.
In addition, the STP structural acceptance cri-teria were selected to assure that, for the most probable loads induced by a postulated design basis LOCA, a safety factor to failure of at least two existed for the weakest structural or mechanical component in the containment system for each operating Mark I BWR facility.
Consequently, since the design margin of safety for the containment systems of operating Mark I facilities has been reduced from the margin believed to be present at the time these facilities were orig-inally reviewed and licensed, the need exists (1) to establish design basis LOCA loads which are appropriate for the life of the facility, and (2) to restore the originally-intended design safety margins for the containment systems.
For those Mark I BWR facilities not yet licensed for operation, the need exists (1) to establish design basis LOCA loads which are appropriate for the life of the facility, and (2) to ensure that adequate design safety margin has been provided in the design of the containment system prior to issuance of an operating license.
In the event that the LTP evaluation results are not available before the issuance of an operating license for a Mark I BWR facility not yet licensed for operation, the utilization of " interim" loading requirements and/or " interim" structural accep'tance criteria more conservative than those which were established for the STP evaluation i790'2 9 A-7/1
will be considered on a case-by-case basis. These considerations will include value-impact assessments related to the timing (i.e.,
before or after initial reactor operation) for the implementation of necessary structural modifications, if any.
However, in such cases, the containment system structural and mechanical elements will be subject to reanalysis when the LTP loading requirements and structural acceptance criteria become available.
2.
PLAN FOR PROBLEM RESOLUTION The major portion of the NRC staff's efforts related to the resolu-tion of the Mark I Containment LTP concerns will consist of review and evaluation of the results of the Mark I Containment LTP which is being conducted by the Mark I Owner's Group.
As documented in Revi-sion I to the " Mark I Containment Program Action Plan" which was submitted to the NRC on February 11, 1977, the Mark I Owner's Group has initiated a comprehensive testing and evaluation program to define design basis loads for the Mark I containment system and to establish structural acceptance criteria which will assure margins of safety for the containment system which are equivalent to that which is currently specified in the ASME Boiler and Pressure Vessel Code.
Also included in their program is an evaluation of the need for structural codifications and/or load mitigation devices to assure adequate Mark I containment system structural safety margins.
Key elements of the Mark I Owner's LTP are:
(1) the submittal of a load definition report (LDR), which will contain design basis hydro-dynamic pressure suppression loads and their possible combinations, and proper procedures as how to apply them for structural evaluation, and (2) the development of structural acceptance criteria, which will be used to assess the structural capability of each Mark I contain-ment system to withstand the design basis loads.
The NRC staff will evaluate the loads, load combinations, and asso-ciated structural acceptance criteria proposed by the Mark I Owners Group prior to the conduct of plant-unique structural evaluations.
The results of this evaluation will be documented in a generic Safety Evaluation Report.
Publication of this report will constitute the resolution of this Technical Activity.
Implementation of the results of this generic review, although not a part of this task, will be accomplished by an NRC requirement that each affected utility perform a plant-unique structural evaluation of the containment system for their facility using the loads, loading combinations, and structural acceptance criteria approved by.the NRC staff.
'790 234' A-7/2
The NRC has initiated several confirmatory research programs related to the Mark I LTP.
These programs, which are discussed in Section 4 below, are designed to provide the NRC staff with an independent source of information to evaluate the results of the Mark I Owner's program and to assist in providing a basis, for regulatory decisions regarding the adequacy of the Mark I containment systems.
The Mark I Owner's LTP commenced in June 1976 with the in plant SRV testing at Monticello and is currently scheduled for completion in 1979.
3.
BASIS FOR CONTINUED PLANT OPERATION AND LICENSING PENDING COMPLE-TION OF TASK The safety issue addressed by this Task Action Plan (TAP) is appli-cable to Boiling Water Reactor (BWR) facilities with the Mark I containment system design.
A total of 25 such facilities have been built or are being built in the United States; of these, 21 are currently licensed for operation.
Construction Permits have prev-iously beer. granted for the remaining four Mark I BWR facilities; no further applications to construct a facility of this type design are anticipated.
Based upon the schedule for completion of this TAP, the TAP results will be available during the Operating License review of three of the four Mark I BWR facilities not yet licensed for operation (i.e., Hope Creek Units Nos. 1 and 2 and Fermi Unit No. 2).
Conse-quently, necessary design modifications, if any, can be accomplished while these facilities are being constructed.
The types of modifica-tions currently under consideration (e.g., improvements in the load bearing capability of the torus support system, load mitigating devices on safety-relief valve discharge lines, and pool swell deflection devices on the internal vent header) have already been shown to be feasible.
The applicant for the fourth facility (i.e., Hatch Unit No. 2) not yet licensed for operation is scheduled to receive an Operating License in 1978.
Structural modifications to the containment system at this facility (e.g., improvements in the load bearing capacity of the torus support system) h. ave been accomplished and the applicant has submitted a plant-unique analysis similar to that which has been submitted for each operating Mark I BWR facility.
Based on the rationale presented below for continued operation of licensed facilities, we have concluded that, pending completion of this task, operation of this facility would not present an undue risk to the health and safety of the public and, consequently, with respect to this' issue, that an Operating License can be granted for this facility.
35' A-7/3
For Mark I BWRs currently licensed for operation, we have concluded that there is reasonable assurance that continued operation, pending completion of this task, does not constitute an undue risk to the health and safety of the public for the following reasons:
As documented in NUREG-0408, " Mark I Containment Short Term Program Safety Evaluation Report," December 1977, based upon our review of the generic "Short Term Program Final Report" and Addenda submitted by the Mark I Owner's Group and the plant-unique analysis reports submitted by each licensee of an operating Mark I BWR facility, we have concluded that licensed Mark I BWR facilities can continue to operate safely, without undue risk to the health and safety of the public, during an interim period of approximately two years, while a methodical, comprehensive Long Term Program (LTP) is conducted.
This conclusion has been made based on our determination:
(1) that the magnitude and character of each of the hydrodynamic loads l2 resulting from a postulated LOCA have been adequately defined for use in the Short Term Program (STP) structural assessment of the Mark I containment system, (2) that, for the "most probable" loads induced by a postulated LOCA, a safety factor to failure of at least two exists for the weakest structural or mechanical component in the containment system for each operating Mark I BWR facility, and (3) that, based on (1) and (2), each Mark I containment system would maintain its integrity and functional capability in the unlikely event of a design basis LOCA.
We have reviewed the Mark I Owner's Program Action Plan for the LTP and find that it is reasonably designed to provide resolution of the issues raised during our review of the STP and to satisfy the LTP objectives. As part of this task, we will continually monitor the progress of the LTP to assure that these requirements are satisfied.
As was the case during the conduct of the STP, if information becomes available during the course of the LTP which indicates that the safetv ractor to fa ore of a component of the contain-ment system
- t. a Mark I BWR Leility is less than two, immediate corrective action will be required Suti action could take the form of structural modifications, instr.llation of load mitigating devices, or other appropriate measures.
4.
NRR TECHNICAL ORGANIZATIONS INVOLVED A.
Plant Systems Branch, Division of Operating Reactors:
Has overall lead responsibility for design basis load definition 17901236,
A-7/4