Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
SSINS No.: 6835
IN 87-50
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
October 9, 1987
Information Notice No. 87-50: POTENTIAL LOCA AT HIGH- AND LOW-PRESSURE
INTERFACES FROM FIRE DAMAGE
Addressees:
All nuclear power reactor facilities holding an operating license or a con-
struction permit.
Purpose:
This information notice is provided to alert recipients to a potentially
significant safety problem pertaining to the possible initiation of a
loss-of-coolant accident (LOCA) as a result of fire damage in the control room
or the cable spreading room. If the postulated fire causes a hot short which
opens a high pressure to low-pressure system interface isolation valve,
exposure of the low-pressure system to pressures in excess of its design
pressure could result in a LOCA. It is expected that recipients will review
the information for applicability to their facilities and consider actions, if
appropriate, to preclude a similar problem. However, suggestions contained in
this notice do not constitute NRC requirements; therefore, no specific action
or written response is required.
Background:
The requirements of 10 CFR Part 50, Appendix R, "Fire Protection Program for
Nuclear Power Facilities Operating Prior to January 1, 1979," are applicable
to all licensed nuclear power reactor facilities that were operating before
January 1, 1979. Facilities that were licensed after that date either commit-
ted to comply with the requirements of Appendix R or were reviewed for
conformance with the guidelines of the Standard Review Plan (NUREG-0800),
Section 9.5.1, "Fire Protection Program," which incorporates the requirements
of Appendix R as guidelines. Thus, the same criteria have been used on all
nuclear power reactor facilities. In either case, they are simply referred to
as the criteria of Appendix R for the purpose of this information notice.
Appendix R states, in part, that where adequate fire protection of safe shut-
down systems cannot be maintained, an alternative method of safely shutting
down the plant shall be provided. For most plants, an alternate shutdown
method is required in the event of a postulated fire in the control room or
the cable spreading room. Appendix R also states that for these areas,
"...the fission product boundary integrity shall not be affected, i.e., there
shall be
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October 9, 1987
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no...rupture of any primary coolant boundary." Thus, for those low-pressure
systems that connect to the reactor coolant system (a high-pressure system),
at least one isolation valve must remain closed despite any damage that may be
caused by fire. A fire could occur in the panel or cables that control the
isolation valves causing hot shorts that may result in opening the valves at
the high/low-pressure interface. Since the low-pressure system could be
designed for pressures as low as 200 to 400 psi, the high pressure from the
reactor coolant system (approximately 1000 to 1200 psi for BWRs and 2000 to
2200 psi for PWRs) could result in failure of the low-pressure piping. In
many instances, the valves at the high-pressure to low-pressure interface are
not designed to close against full reactor coolant system pressure and flow
conditions. Thus, spurious valve opening could result in a LOCA that cannot
be isolated, even if control of the valve can be reestablished.
Description of Circumstances:
During a fire protection re-review at Washington Public Power Supply System's
Washington Nuclear Project Number 2 (WNP-2), the licensee discovered that
should a fire occur in the control room, power would have to be removed from
the valve motor operators in the residual heat removal (RHR) system suction
and discharge lines to prevent inadvertent valve operations resulting from
possible fire damage to the circuits. If the damage occurred before removing
power to the valve motor operators, the valves could be spuriously opened,
resulting in overpressurization of the RHR piping that could lead to a LOCA
that could not be isolated.
In discussions with the WNP-2 personnel, the NRC staff became aware of a
bypass line around the check valve in the discharge line that had a
motor-operated isolation valve in the line. This bypass line is used to warm
up the RHR system discharge line by backflow from the reactor before
initiating residual heat removal to prevent thermal shocking of the reactor
vessel nozzle safe end. Because of this bypass line around the check valve,
credit for the check valve in preventing a LOCA at the high- and low-pressure
interface can no longer be given.
The licensee intends to remove the power to this motor operator during normal
power operations. Since this valve is used only for prewarming the RHR line
during a normal shutdown, removing power during normal power operations should
not adversely impact safe plant operations.
In order to determine if other plants have piping designs similar to that of
WNP-2, the final safety analysis reports of nine other BWRs were reviewed by
the staff. These included BWR-4, BWR-5, and BWR-6 designs. Of these nine
plants, six (Clinton Power Station; Hope Creek Nuclear Station; Limerick
Generating Station; Nine Mile Point Nuclear Station, Unit 2; Perry Nuclear
Power Plant; and Susquehanna Steam Electric Station) have a piping configu-
ration similar to that of WNP-2. One plant (Monticello Nuclear Generating
Plant) has a design similar to WNP-2 but has two normally closed, locally
operated manual valves in the bypass line; therefore, this problem does not
appear to apply to this plant. The two remaining plants (Grand Gulf Nuclear
Station and River Bend Station) do not have bypass lines around the check
valve.
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The potential for creating a LOCA from a similar high- and low-pressure inter-
face may also be applicable to PWRs.
No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact the Regional
Administrator of the appropriate regional office or this office.
Charles E. Rossi, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical Contact: John N. Ridgely, NRR
(301) 492-4742
Attachment: List of Recently Issued NRC Information Notices