United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 96-60: Potential Common-Mode Post-Accident Failure of Residual Heat Removal Heat Exchangers

                                    UNITED STATES
                            NUCLEAR REGULATORY COMMISSION
                        OFFICE OF NUCLEAR REACTOR REGULATION
                             WASHINGTON, DC  20555-0001

                                  November 14, 1996


NRC INFORMATION NOTICE 96-60:  POTENTIAL COMMON-MODE POST-ACCIDENT         
                                   FAILURE OF RESIDUAL HEAT REMOVAL HEAT         
                                   EXCHANGERS


Addressees

All holders of operating licenses or construction permits for nuclear power reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert
addressees to a potential common-mode post-accident failure of boiling water reactor
(BWR) residual heat removal (RHR) heat exchangers.  It is expected that recipients will
review the information for applicability to their facilities and consider actions, as
appropriate, to avoid similar problems.  However, suggestions contained in this information
notice are not NRC requirements; therefore, no specific action or written response is
required.

Background

The NRC staff issued Information Notice 96-45, "Potential Common-Mode Post-Accident
Failure of Containment Coolers," on August 28, 1996, to alert licensees to a potential
failure mechanism of containment coolers during a design-basis loss-of-coolant accident
(LOCA) with a concurrent loss of offsite power or with a delayed sequencing of safety-
related equipment.  Specifically, if the containment coolers are exposed to forced
convection in a post-LOCA environment without cooling water flow, boiling may occur on
the stagnant cooling water side of the containment coolers.  When cooling water flow is
subsequently reinitiated, the steam voids will collapse and may create significant
hydrodynamic loads (waterhammer).  The waterhammer may threaten the integrity of the
containment cooler or the associated cooling water piping.  Generic Letter 96-06,
"Assurance of Equipment Operability and Containment Integrity During Design-Basis
Accident Conditions," issued on September 30, 1996, also addressed this issue.

Description of Circumstances

A potential waterhammer scenario in the service water side of the RHR heat exchanger
was identified at the LaSalle Nuclear Station.  The postulated phenomenon is analogous 
to that identified for containment coolers in Information Notice 96-45 and Generic 
Letter 96-06.



9611070081.                                                                 IN 96-60
                                                                 November 14, 1996
                                                                 Page 2 of 3


At LaSalle, a BWR 5 plant, the top of the vertical RHR heat exchanger is the high point of
the RHR service water system and is at an elevation that is higher than that of the ultimate
heat sink by approximately 9 meters (30 feet) for normal lake levels.  The resultant
pressure in the heat exchanger tubes in a standby condition is estimated to be between
7000 and 21,000 pascals (1 and 3 psia), with associated saturation temperatures between
43� and 60� C (110� and 140� F).  Boiling will occur in these tubes if these temperatures
are exceeded.  Under low lake levels, the licensee found that voiding would occur in the
top of the tubes regardless of temperature.

At LaSalle, the RHR system is lined up so that flow will go through both the RHR heat
exchanger and a bypass valve whenever the RHR system is initiated in its safety injection
(low pressure coolant injection) mode.  The RHR service water flow to the tube side of the
RHR heat exchangers does not start automatically on a safety injection signal.  Emergency
operating procedures direct that the RHR service water pumps be started manually within
the first 10 minutes of an accident.

It is postulated that during a LOCA, water from the suppression pool, which could start
from the technical specification allowable 41� C (105� F), could quickly be heated to 
88�C (190� F).  When the relatively hot suppression pool fluid would be pumped through
the RHR heat exchanger, it would boil the stagnant low pressure service water.  When
RHR service water flow would subsequently be manually initiated, the steam voids would
collapse and might create significant hydrodynamic loads.  The hydrodynamic loads might
impair the integrity of the heat exchangers or the associated service water piping.  This
scenario does not occur during normal operation of the RHR system in the shutdown
cooling mode because the RHR service water system is started first to ensure that cooling
flow is established before hot fluid is introduced into the heat exchangers.

Both units at the LaSalle Nuclear Station are currently shut down.  The licensee continues
to assess this issue and will ensure operability of the RHR heat exchangers before restart. 
The licensee is considering the installation of a keep-fill system on the RHR service water
system to keep the heat exchangers pressurized.

Discussion

The RHR heat exchangers provide an important safety function for long term heat removal. 
The postulated failure scenario could cause a common failure of the RHR heat exchangers,
thereby potentially challenging this function.  In addition, a structural failure of the RHR
heat exchangers could create a containment bypass release path and divert low pressure
coolant injection flow.  An individual plant vulnerability to these postulated failures
depends on a number of factors.  The physical configuration of the RHR service water
system and RHR heat exchanger, the operational alignment of the heat exchanger isolation
valves, the sequencing of the RHR service water pumps, and other site-specific parameters
have an effect on facility vulnerability to this potential failure mode.  


.                                                                 IN 96-60
                                                                 November 14, 1996
                                                                 Page 3 of 3


This information notice requires no specific action or written response.  If you have any
questions about the information in this notice, please contact one of the technical contacts
listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.



                                             signed by D.B. Matthews for

                                       Thomas T. Martin, Director
                                       Division of Reactor Program Management
                                       Office of Nuclear Reactor Regulation

Technical contacts:  V. Patricia Lougheed, Region III
                           (630) 829-9760
                        E-mail:  vpl@nrc.gov
                          
                       James Tatum, NRR
                       (301) 415-2805        
                       E-mail:  jet1@nrc.gov

                       John Tappert, NRR
                       (301) 415-1167  
                       E-mail:  jrt@nrc.gov
                          

Page Last Reviewed/Updated Thursday, November 21, 2013