ML20198E681
| ML20198E681 | |
| Person / Time | |
|---|---|
| Site: | Seabrook, 05000000 |
| Issue date: | 05/08/1986 |
| From: | Bernero R NRC |
| To: | Thompson H NRC |
| Shared Package | |
| ML20151L176 | List: |
| References | |
| FOIA-86-266 NUDOCS 8605280178 | |
| Download: ML20198E681 (15) | |
Text
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UNITED STATES g
g NUCLEAR REGULATORY COMMISSION g
E W ASHINGTON, D. C. 20555
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~ MEMORANDUM FOR: Hugh L. Thompson, Jr., Director Division of Licensing FROM:
Robert M. Bernero, Director Division of Systems Integration
SUBJECT:
SAFETY IMPLICATIONS OF MAIN STEAM SAFETY VALVES (MSSVs)
FLOW OEFICIENCY i
Purpose and Background The purpose of this memorandum is to identify the safety implications and i
potential consequences of MSSVs', flow deficiency.
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I % v-During the last part of 198% Wyle Laboratories c'onducted several full flow tests on two MSSVs manufactu'r'ed by Crosby Valve and Gage Company.
These two l
valves were chosen as a representative sample for the MSSVs that are to be installed on the Seabrook plant main steam lines.
The purpose of the Wyle tests was to determine the adequacy of different discharge piping configurations.
Test measurements indicated that the MSSVs (with the manufacturer's recommended guide ring settings), have a flow capacity of about 50% of their design values at the design pressure.
The guide ring settings determine the force being exerted on the valve disc, thereby affecting the degree of valve tift and subsequently the discharge flow capacity of the valve.
During the Wyle tests the guide ring settings of both valves were substantially adjusted (150 = tches de. :rd) in order to. achieve the full flow capacities.
Subsequent to the completion of the tests, the Public Service Company of New Hampshire, the owner of the Seabrook plant, concluded that in order to ensure full flow capacity of the plant's MSSVs they all should be adjusted downward by g notches from the manufacturer's settings.
While the Seabrook experience shows a d'eficiency in the capacity adjustment of l
the Crosby spring loaded safety valves, it strongly suggests a similar j
deficiency in similar valves made by other manufacturers, since they all work on the same basic concept.
1 CONTACT:
S. Diab, x29440 1
8605200170 860D00 PDR FOIA MURPHY 86-266 PDR j
H. Thompson,
Safety Implications Full flow testing of MSSVs is not normally performed by either reactor owners
_. _or valve manufacturers, nor is such testing an ASME requirement for capacity certification.
Such certification is obtained through extrapolation from tests on much smaller valves at low pressures.
Based on the Seabrook experience and with the lack of sufficient data, it may p be assumed that a number of deficient safety valves are installed in some i cperating plants and/or planned to be installed in plants-yet to operate.
TW: :hes-may4e-instel-led-on the primary as well as the-secondary sidas#
M* I th. A...t.
It may also be assumed that the capacity of some of these deficient valves may be as low as 50% (as for the Seabrook plant) of the design flow, or even lower. With this potential deficiency, the design basis of the affected plants cannot be met.
The design basis of every pressurized
- l water reactor (PWR) requires that overpressure protection for the primary and secondary sides of the plant be provided so that the pressure may not rise above 110% of the design value during postulated events.
PWR vendors perform sizing analyses intended to size the safety valves so that these valves have sufficient capacity to mitigate the most severe overpressurization event with sufficient margin. Generic transient analyses, as opposed to sizing analyses, i
performed by Westinghouse for the Westinghouse PWRs show that, for the worst overpressurization event, loss of load without condenser bypass, the MSSVs' peak relieving capacity required is about 80% of the nominal valve flow.
It should also be noted that any flow degradation through the MSSVs increases the potential for the primary safety valve actuation and increases the load on these valves as well.
This is because any relief through the MSSVs will remove heat from the reactor coolant, which would otherwise accumulate, expand, and overpressurize the primary system.
There are no test data available to the staff that show if the valve flow would increase at high upstream pressure conditions and/or if the valve disc travel will be any higher.
Therefore, it
't is not clear whether overpressurizations of this nature are self limiting.
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With less than the required relieving capacity, the secondary side would be expected to be overpressurized, thus increasing the potential for steam side leaks or breaks. While the overpressurization described above would occur following a loss of load event, which is an anticipated operational occurrence, the consequences of that event may lead to a design basis with its associated severe consequences. While plants are li. censed with a certain degree of risk attached to the potential of occurrence of their design basis events, the probability of occurrence of those events is sufficiently low.
However, if a design basis event were to occur with a substantially higher likelihood or the plant were to be pressurized over its design pressure limits with sufficient frequency, then the plant cannot meet its design basis requirements.
beams.
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Based on the limited information.available about the adequacy of the safety valves for overpressurization, the staff has a reason to doubt that all the safety valves currently in use or planned for use would serve their intended
~~ ~ safety function. Therefore, we suggest that the Division of Licensing send a
_ request for additional inf.ormation (RAI), per 10 CFR 50.54(f), to PWR plant owners.
This RAI would request the plant owners to study the Seabrook experience and justify to the staff that their respective plants have sufficient overpressure protection.
The owners' justification may rely on any combination of: (a) plaat egek;;ce er -re4ev.mt. experience from which valve performance can be verified; (b) safety analysis assuming inadequate overpressure protection; or (c) valve testir.g.
Robert M. Bernero, Director Division of Systems Integration
Enclosure:
Suggested 10 CFR 50.54(f) Letter 4
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ENCLOSURE 2 GENERIC ISSUE INFORMATION 1.
Suggested Title of Proposed Generic Issue or new requirement.
REL/A Sit lTy oF P. aft.
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If a new requirement is proposed, what is the proposed requirement?
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