ML20141E037
| ML20141E037 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 03/31/1986 |
| From: | Miraglia F Office of Nuclear Reactor Regulation |
| To: | Randazza J Maine Yankee |
| References | |
| NUDOCS 8604080436 | |
| Download: ML20141E037 (9) | |
Text
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March 31, 1986 Docket No. 50-309 Distribution: Docket File NRC & L PDRs Branch Files OELD PSears PKreutzer ACRS 10 Mr. J. B. Randazza EJordan JPartlow Executive Vice President AThadani FMiraglia Maine Yankee Atomic Power Company JRothman CGrimes 83 Edison Drive JRichardson Augusta, Maine 04336
Dear Mr. Randazza:
SUBJECT:
MAINE YANKEE PARTICIPATION IN SEISMIC DESIGN MARGINS PROGRAM During the past 4 years, you have been working to upgrade the capability of the Maine Yankee plant to withstand a potential seismic event in excess of the original design basis event.
In your letter dated March 14, 1986, you concluded that structures, systems and components at Maine Yankee had sufficient strength to withstand a seismic event of at least 0.2g with a Regulatory Guide 1.60 spectrum and still safely shut down without danger to the public health and safety. In your March 14, 1986 letter, you also provided the results of an analysis on which you based your conclusion. The judgement regarding the capability of the plant to withstand earthquakes beyond the design basis of the plant is consistent with studies by the nuclear industry and the NRC which indicate that nuclear power plants are capable of withstanding earthquake motion substantially greater than the safe shutdown earthquake (SSE) acceleration.
' By letter dated March 4,1986, you indicated an interest in participating in an NRC-sponsored Seismic Design Margins Program (SDMP). That program would be conducted by the Office of Nuclear Regulatory Research with the Office of Nuclear. Reactor Regulation (NRR) Licensing Project Manager (Patrick Sears)
~ ' "_ f acting as the coordinator. The SDMP methodology would be as described in the enclosure. The program would require approximately 1 year to complete.
Experience. data would be used along with certain analyses to determine if the plant can withstand a 0.3g acceleration level and be able to go to hot shutdown
.and stay there. NRR will review the results of the SDMP, and will consider
'whether any design or procedural improvements would provide a substantial increase in ^the overall protection of the public health and safety and whether the direct and indirect. costs of implementation of such improvements would be justified in view of the increased protection, in accordance with the v
backfitting rule (10 CFR 50.109).
8604000436 860331 PDR ADOCK 05000309 P
.... /
. On April 17, 1979, an earthquake of approximately magnitude 4 occurred about 10 kilometers west of the plant site.
Maine Yankee was shut down at the time. This was followed by at least 30 aftershocks.
On January 9, 1982, an earthquake of approximate magnitude 5-3/4 occurred in central New Brunswick, Canada.
Appendix A to 10 CFR Part 100 (Seismic and Geologic Siting Criteria for Nuclear Power Plants) requires that in determining Safe Shutdown Earthquake in current licensing actions, the largest reported earth-quake in a tectonic province which cannot reasonably be related to tectonic structure should be assumed to occur near the site.
In your March 24, 1986 letter, you stated "In performing future changes to the plant, Maine Yankee will ensure that the scismic ruggedness of existing structures, systems and components is maintained.
New systems, and their associated structures and components will be designed to 0.18g NUREG-0098 50th percentile with SEP allowables (i.e., for piping, damping = 3% or PVRC, allowable stress = 2.4S, no 0BE) up to any interface with existing h
structures, systems and components." Thus, for future plant changes, you would assure the existing structures, systems and components are not structurally degraded by the change. Those existing structures, systems and components would have been verified by the SDMP.
New structures, systems and components which necessarily interface with existing structures, systems and components would be designed using SEP (Systematic Evaluation Program) allowables up to the interface.
SEP allowables include such items as percent critical damping, spectrum shape, and allowable stresses. We agree with your plans.
Staff assessments indicate 0.18g NUREG-0098 50th percentile to be an acceptable acceleration level to account for the April 17, 1979 and January 19, 1982 earthquakes.
Your Licensing Project Manager will work with you to develop an acceptable schedule for conducting this program.
Sincerely, kl Frank J. Miraglia, Director PWR Project Directorate #8 Division of PWR Licensing-B
Enclosure:
As stated cc w/ enclosure:
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. used along with certain analyses to determine if the plant can withstand a
.3g acceleration level and be able to go to hot shutdown and stay there.
NRR will review the results of the SDMP and, using backfit procedures, will deter-mine if any design / procedural improvements are warranted.
We agree with your plans, as stated in your March 24, 1986 letter, to assure that, in performing future changes to the plant, Maine Yankee will ensure that the seismic ruggedness of existing structures, systems, and components is not degraded and that new systems, and their associa %d structures and components will be designed to 0.18g NUREG-0098 50th o:entile with SEP allowables up to any interface with existing structure systems, and components.
Your Licensing Project Manger will work with you to develop an acceptable schedule for conducting this program.
Sincerely, Harold Denton, Director Office of Nuclear Reactor Regulation
Enclosure:
As stated cc:
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Mr. J. B. Randazza Maine Yankee Atomic Power Company Maine Yankee Atomic Power Station
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CC:
Charles E. Monty, President Mr. P. L. Anderson, Project Manager Maine Yankee Atomic Power Company Yankee Atomic Electric Company 83 Edison Drive 1671 Worchester Road Augusta, Maine 04336 Framingham, Massachusetts 07101 Mr. Charles B. Brinkman Mr. G. D. Whittier Manager - Washington Nuclear Licensing Section Head Operations Maine Yankee Atomic Power Company Combustion Engineering, Inc.
83 Edison Drive 7910 Woodmont Avenue Augusta, Maine 04336 Bethesda, Maryland 20814 John A. Ritsher, Esquire Ropes & Gray 225 Franklin Street Boston, Massachusetts 02110 State Planning Officer Executive Department 189 State Street Augusta, Maine 04330 Mr. John H. Garrity, Plant Manager Maine Yankee Atomic Power Company P. O. Box 408 Wiscasset, Maine 04578 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pennsylvania 19406 First Selectman of Wiscasset Municipal Building U.S. Route 1 Wiscasset, Maine 04578 Mr. Cornelius F. Holden Resident Inspector c/o'U.S. Nuclear Regulatory Commission P. O. Box E Wiscasset, Maine 04578
SEISMIC DESIGN MARGINS PROGRAM Seismic Design Margins Programs The margin review process involves both the screening of components based on their importance and seismic capacity and the quantification of High Confidence Low Probability of Failure (HCLPF) values for components, systems, accident sequences, and the plant. The probability that a component would fail for a specified ground motion is generally called " fragility." Systems analysis is used to determine those plant systems and components that are important contributors to plant seismic safety and thus allow focusing of effort on components requiring a margin review. Previous Probabilistic Risk Assessment (PRA) studies found that, for pressurized water reactors (PWRs),
there are primarily two plant-safety functions that were identified as being the major contributors to plant seismic safety: reactor subcriticality and early emergency core-coolant injection.
For PWRs, those functions are the initial candidates of the margin review. The methodology for the SDMP is follows:
1.
The review earthquake level for Maine Yankee would be 0.3g.
2.
Safety Functions would be identified for the following:
a.
Reactor Subcriticality - functions necessary to shut down nuclear reaction such that the only heat being generated is decay heat.
b.
Normal Cooldown - functions providing cooling to the reactor core through the use of the normal power conversion system, normally defined as the main steam, turbine bypass,~ condenser, condensate, and main feedwater subsystems.
. c.
Emergency Core Cooling (Early) - functions providing cooling to the reactor core in the early (transient) phase of an event sequence by the use of one or more emergency systems designed for this purpose.
The exact timing of "early" is somewhat plant specific and sequence dependent.
However, for our purpose it can be deemed to be the time period during which these systems are initially called to operate.
After the identification of those functions, systems analysts would then determine which systems are used in the plant to carry out the functions and which components are constituent parts of these systems.
3.
The structural engineer would gather information about the structural, equipment, and related features of the plant.
A rather large number of HCLPF values exist for 0.3g.
4.
A plant walkdown would be performed, consisting of a review of plant structures documentation, plant inspection, and discussions with plant personnel.
The objectives of the walkdown are as follows:
a.
To gain a general understanding of the plant layout and relationships to the components.
b.
To verify the validity of the decisions made to screen out components with generically high HCLPFs.
c.
To identify plant unique features which must be considered in margins review.
d.
To gather information needed to perform more detailed reviews of components that potentially have HCLPFs less than the review earthquake level.
.. e.
To identify systems interaction and other types ordependencies not identified during Step 2.
5.
Taking the results from the walkdown, the systems analysts would revise the systems relationships already established.
Fault tree development for systems, event tree development to establish relationships, and other similar systems analysis work would be accomplished.
The objective of this step is to provide a nearly complete set of fault and event trees that incorporate all of the necessary components and plant-unique features for which fragility values have been preliminarily assessed in Steps 3 and 4 as being of continuing concern.
6.
The second plant walkdown will be primarily carried out by the fragilities analysis team, taking the results of the first walkdown and the systems analysis into account.
Prior to this walkdown, some fragility analyses should have been done.
It is intended that this walkdown emphasize actual physical study of those plant components requiring detailed fragility analysis.
Systems analysis input will be needed but in a supporting capacity.
Although most systems interactions and similar dependencies will have been discovered in Step 4, it is likely that a review of systems interaction at this time would be necessary to assure that other such dependencies do not remain undiscovered.
The second walkdown will be primarily a gathering of
~
information to develop the HCLPF capacity of the componehts needi,ng a detailed analysis (Step 8).
The result of this Step will be a
4 categorization of ti,e remaining components by seismic' capacity; that is, into those whose HCLPF value is judged to be high enough, and those for which a detailed HCLPF determination will be necessary in Step 8 below.
7.
For those components passed through from Steps 5 and 6, the systems analyst determines what accident sequence groups in which those components participate.
From such determinations, detailed cut sets (Boolean expressions) should be developed for the end point of core melt.
8.
For all the components contained in the Boolean expressions developed in Step 7, the HCLPF values are developed.
Special attention should be given to any singles or doubles, plus any other cut sets that appear to have HCLPF values at the lower end of the range under consideration.
The HCLPF values for the combinations of components (all but " singles") may be determined using special methods.
In practice, Steps 7 and 8 may be performed in parallel with extensive interaction between the fragility and systems analysts.
These tw:
will build on one another thereby involving an interaction process between the fragility and system analysis teams.
The result emerging from Step 8 is a HCLPF value for each Boolean expression leading to the end point of a core-melt accident sequence, or at least for each Boolean expression that is judged to be among
I the "important" ones by having HCLPF values potentially_, lower than the review level, or being a " single" or " double."
If the plant's HCLPF is a greater value than the earthquake review level, the quantitative estimate of the plant HCLPF is not obtained.
- Rather, a statement can be made that the HCLPF value is at least as high as the earthquake review level.
The eight steps just outlined comprise the framework for the seismic margins review methodology.
As can be observed, the steps are intended to screen in (or out) each component, either by itself or ultimately in combination with related components that support a system or carry out a safety function.
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