Information Notice No. 87-50: Potential Loca At High- And Low-Pressure Interfaces From Fire Damage

                                                       SSINS No.:  6835
                                                          IN 87-50

                                  UNITED STATES
                          NUCLEAR REGULATORY COMMISSION
                             WASHINGTON, D.C.  20555

                                 October 9, 1987

Information Notice No. 87-50:  POTENTIAL LOCA AT HIGH- AND LOW-PRESSURE 
                                   INTERFACES FROM FIRE DAMAGE


All nuclear power reactor facilities holding an operating license or a con-
struction permit.


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.


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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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 
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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

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