Information Notice No. 89-36:Excessive Temperatures in Emergency Core Cooling System Piping Located Outside Containment

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

                                  April 4, 1989

                                    CORE COOLING SYSTEM PIPING LOCATED 
                                    OUTSIDE CONTAINMENT


All holders of operating licenses or construction permits for nuclear power 


This information notice is being provided to alert addressees to an event that 
involved the potential for reactor coolant system (RCS) leakage outside con-
tainment.  A check valve serving as the inboard containment isolation valve 
in a high-pressure injection (HPI) system injection line failed to seat pro-
perly after termination of HPI flow, allowing RCS backflow into HPI system 
piping that was not qualified for RCS temperatures.  The HPI system piping 
was exposed to fluid temperatures in excess of design temperatures, resulting 
in stresses that exceeded the allowable limits for Class 1 piping according 
to the American Society of Mechanical Engineers Boiler and Pressure Vessel 
Code (ASME Code), Section III.  Recipients are expected to review the infor-
mation for applicability to their facilities and consider actions, as appro-
priate, to avoid similar problems.  However, suggestions contained in this 
information notice do not constitute NRC requirements; therefore, no specific 
action or written response is required.

Description of Circumstances:

On January 20, 1989, at Arkansas Nuclear One, Unit 1 (ANO-1), a failure of the 
main generator exciter resulted in a generator lockout and subsequent trips of 
the main turbine and the reactor.  Upon loss of power to plant loads from the 
main generator, one of the two non-safety-related 6.9 kV buses failed to auto-
matically fast transfer from the unit auxiliary transformer to the startup 
transformer; this failure caused two of the four reactor coolant pumps (RCPs) 
to trip on undervoltage.  A failure of one of the main feedwater (MFW) pumps 
to runback to minimum speed and a failure of a MFW block valve and control 
valves to close after the reactor trip resulted in overfeed of the once 
through steam generators; this overfeed caused a slight overcooling (11�F) of 
the RCS.  The operators manually started one HPI system pump to maintain 
pressurizer level above the heater cutoff point.  The pump was secured 2 
minutes later; however, check valve MU-34B in the "B" HPI line did not seat 
properly (see Figure 1).  The existing RCP configuration (i.e., two pumps 
running and two 

.                                                            IN 89-36
                                                            April 4, 1989
                                                            Page 2 of 3

pumps tripped) created a differential pressure across MU-34B that caused RCS 
backflow into the HPI system piping outside containment.  The flow path ran 
from the RCS and outside containment via MU-34B, through the crossover pipe 
to the "C" HPI line, and then back inside containment to the RCS via MU-34C 
as shown in Figure 1.  The HPI system piping upstream of MU-34B was not qua-
lified for RCS temperatures.  Subsequent analysis by the licensee (assuming 
a RCS temperature of 545� F) showed that the temperature effects resulted in 
stresses that exceeded ASME code-allowable limits for Class 1 piping.  The 
HPI system piping was qualified for RCS pressure but was only designed for 
a temperature of 145� F.  The licensee became aware of RCS leakage outside 
containment when tape attached to the HPI piping began to melt, smolder and 
smoke, activating a local smoke detector and the associated control room 


The primary concerns are that the failure of a check valve to seat properly 
exposed piping located outside containment to RCS temperature and that the 
piping was not designed for RCS temperatures.  Furthermore, the check valve 
serves as the inboard containment isolation valve, but valve testing 
(consisting of visual inspection and vertical stroke of the valve disc) was 
not adequate to reveal the excessive wear problem that led to its failure.  
Because the piping outside con-tainment was not monitored to detect RCS 
in-leakage (e.g., high temperature alarms), the piping potentially could be 
exposed to RCS temperatures for long periods without being detected.  It is 
important for addressees to note the need for piping to be qualified for 
potential inservice conditions and that the as-sociated components, that are 
part of the reactor coolant pressure boundary, are subject to applicable ASME 
Code requirements which include, in part, leak detection, isolation and 
periodic testing. 

Corrective actions proposed by the licensee include installation of a second 
check valve in each HPI line inside containment, installation of temperature-
monitoring instrumentation in the HPI lines outside containment between the 
containment penetration and the first outboard check valve, replacement of 
all piping that was determined to be overstressed from high temperature during 
the event, and leak rate testing for all check valves in the HPI lines inside 

Subsequent review by the licensee identified HPI system pipe supports that 
are not qualified for the maximum temperature to which the associated piping 
could be exposed when the HPI system is used in the piggyback mode of 
operation.  The licensee is upgrading the pipe supports.  It is important that 
system piping be analyzed and qualified for the maximum temperature and 
pressure to which it could be exposed, regardless of whether credit is given 
in the final safety analysis report (FSAR) transient/accident analysis for the 
associated mode of operation. 

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                                                            April 4, 1989
                                                            Page 3 of 3

No specific action or written response is required by this information notice.  
If you have any questions about this matter, please contact one of the 
technical contacts listed below or the Regional Administrator of the 
appropriate regional office. 

                                   Charles E. Rossi, Director
                                   Division of Operational Events Assessment
                                   Office of Nuclear Reactor Regulation

Technical Contacts:  Rick Kendall, NRR
                     (301) 492-3140

                     Yueh-Li Li, NRR
                     (301) 492-0915

1. Figure 1 - ANO High Pressure Injection System Flow Path to the Reactor
              Coolant System (Simplified Diagram)
2. List of Recently Issued NRC Information Notices
.                                                            Attachment 2 
                                                            IN 89-36
                                                            April 4, 1989
                                                            Page 1 of 1

                             LIST OF RECENTLY ISSUED
                             NRC INFORMATION NOTICES
Information                                  Date of 
Notice No._____Subject_______________________Issuance_______Issued to________

88-86,         Operating with Multiple       3/31/89        All holders of OLs
Supp. 1        Grounds in Direct Current                    or CPs for nuclear
               Distribution Systems                         power reactors.

89-35          Loss and Theft of Un-         3/30/89        All U.S. NRC 
               secured Licensed Material                    byproduct, source 
                                                            and special 
                                                            nuclear material

89-34          Disposal of Americium         3/30/89        All holders of an
               Well-Logging Sources                         NRC specific 
                                                            authorizing well-
                                                            logging activities.

89-33          Potential Failure of          3/23/89        All holders of OLs
               Westinghouse Steam                           or CPs for PWRs.
               Generator Tube 
               Mechanical Plugs

89-32          Surveillance Testing          3/23/89        All holders of OLs
               of Low-Temperature                           or CPs for PWRs.

89-31          Swelling and Cracking         3/22/89        All holders of OLs
               of Hafnium Control Rods                      or CPs for PWRs 
                                                            with Hafnium 
                                                            control rods. 

89-30          High Temperature              3/15/89        All holders of OLs
               Environments at                              or CPs for nuclear
               Nuclear Power Plants                         power reactors.

89-29          Potential Failure of          3/15/89        All holders of OLs
               ASEA Brown Boveri                            or CPs for nuclear
               Circuit Breakers                             power reactors.
               During Seismic Event

89-28          Weight and Center of          3/14/89        All holders of OLs
               Gravity Discrepancies                        or CPs for nuclear
               for Copes-Vulcan                             power reactors.
               Air-Operated Valves

OL = Operating License
CP = Construction Permit 

Page Last Reviewed/Updated Friday, May 22, 2015