Information Notice No. 86-11: Inadequate Service Water Protection Against Core Melt Frequency

                                                            SSINS No: 6835 
                                                            IN 86-11       

                                UNITED STATES
                            WASHINGTON, DC 20555

                              February 25, 1986

                                   AGAINST CORE MELT FREQUENCY 


All nuclear power reactor facilities holding an operating license (OL) or a 
construction permit (CP) 


This notice is to alert recipients of a potentially significant problem 
concerning the possible failure to provide sufficient redundancy in the 
essential service water system (ESW) Failure of all ESW may be an 
accident-initiating event that could lead to a core melt It is expected 
that recipients will review the information for applicability to their 
facilities and consider action, if appropriate, to preclude a similar 
problem occurring at their facilities However, suggestions contained in 
this information notice do not constitute NRC requirements; therefore, no 
specific action or written response is required 

Past Related Correspondence: 

Circular 78-13, "Inoperability of Service Water Pumps," July 10, 1978 
Description of Circumstances: 

In May 1984, the Byron Unit 1 licensee, Commonwealth Edison, submitted to 
the NRC a probabilistic risk assessment (PRA) to justify extending allowable
outage times for certain equipment from 3 days to 7 days The NRC reviewed 
the study and determined that loss of both ESW pumps on Unit 1 was not 
considered as an accident-initiating event At present, Byron Unit 1 is 
operating and has two ESW pumps--one operating and one on standby If the 
operating train failed and the standby train would not start, the component 
cooling water system (CCW) would heat up The nuclear steam supply system 
vendor, Westinghouse, has estimated that the heatup of the CCW would trip 
the CCW pumps in 6 minutes CCW is essential for cooling the reactor coolant 
pump (RCP) seals, either directly or via the charging pumps, which also are 
cooled by CCW Without cooling, the RCP seals may possibly fail and cause a 
loss-of-coolant accident (LOCA) 1  Assuming that case in the PRA the ECCS 
pumps needed to mitigate the ensuing 

1 The NRC Office of Nuclear Reactor Regulation currently has the subject of 
RCP seal failure under study in its Generic Issue 23 


                                                          IN 86-11 
                                                          February 25, 1986 
                                                          Page 2 of 2 

LOCA, also would fail without CCW Thus, loss of ESW could result in a core 

On reevaluation of the study under the assumption that loss of ESW is a 
LOCA-initiating event, the core melt frequency was estimated at 0001 per 
year This result also assumed that additional pumps, such as the ESW pumps 
for Unit 2, would not be available to mitigate the LOCA at Unit 1 To lower 
the estimated core melt frequency, the licensee committed to make at least 
one of the Unit 2 ESW pumps available to Unit 1 by means of a crosstie 
piping arrangement in the event of either of the Unit 1 ESW pumps becoming 
inoperable The availability of the Unit 2 ESW pump reduces the core melt 
frequency estimated from this sequence of events by a factor of 25, and the 
overall core melt frequency by a factor of 5 These estimates reaffirm the 
perceived weakness of the two-train system and the desirability of making 
the Unit 2 ESW pump available 


In the Byron design, each of the two Unit 1 ESW pumps will supply 100 
percent of the system's requirements for Unit 1 The system was licensed as 
meeting single failure safety criteria The PRA, however, identified a 
circumstance in which even such a licensed system may represent significant 
risk, so the licensee remedied the situation by committing to make available 
an additional 100 percent capacity ESW pump from Unit 2 when one Unit 1 ESW 
pump becomes inoperable 

The NRC has surveyed pressurized water reactor, (PWR) designs that are in 
operation and under construction as to availability of sufficient pumping 
capacity for ESW This preliminary information indicates only one of these 
PWR designs may be subject to the same question discussed here for Byron 
This is being pursued by the NRC staff Boiling water reactor (BWR) designs 
have not been similarly surveyed; it has not been determined by a PRA 
whether loss of ESW could similarly result in a high core melt frequency for 
BWRs Thus, due to uncertainty resulting from the incompleteness of this 
survey, this notice is being published 

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 

                                   Edward L Jordan Director 
                                   Division of Emergency Preparedness 
                                     and Engineering Response 
                                   Office of Inspection and Enforcement 

Technical Contact:  Vern Hodge, IE
                    (301) 492-7275

                    Leonard Olshan, NRR
                    (301) 492-4937

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