United States Nuclear Regulatory Commission - Protecting People and the Environment

Bulletin 88-08: Thermal Stresses in Piping Connected to Reactor Coolant Systems

                                                       OMB No.:  3150-0011 
                                                       NRCB 88-08 

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSI0N
                     OFFICE OF NUCLEAR REACTOR REGULATION
                            WASHINGTON, D.C.  20555

                                  June 22, 1988


NRC BULLETIN NO. 88-08:  THERMAL STRESSES IN PIPING CONNECTED TO REACTOR 
                         COOLANT SYSTEMS

Addressees: 

All holders of operating licenses or construction permits for 
light-water-cooled nuclear power reactors.

Purpose:

The purpose of this bulletin is to request that licensees (1) review their 
reactor coolant systems (RCSs) to identify any connected, unisolable piping 
that could be subjected to temperature distributions which would result in 
unacceptable thermal stresses and (2) take action, where such piping is 
identified, to ensure that the piping will not be subjected to unacceptable 
thermal stresses.

Description of Circumstances:

On December 9, 1987, while Farley 2 was operating at 33 percent power, the 
licensee noted increased moisture and radioactivity within containment.  The 
unidentified leak rate was determined to be 0.7 gpm.  The source of leakage was
a circumferential crack extending through the wall of a short, unisolable 
section of emergency core cooling system (ECCS) piping that is connected to the
cold leg of loop B in the RCS.  This section of piping, consisting of a nozzle,
two pipe spools, an elbow, and a check valve, is shown in Figure 1. The crack 
resulted from high-cycle thermal fatigue that was caused by rela-tively cold 
water leaking through a closed globe valve at a pressure sufficient to open the 
check valve.  The leaking globe valve is in the bypass pipe around the boron 
injection tank (BIT) as shown in Figure 2.  During normal operation this valve 
and others isolate the ECCS piping from the discharge pressure of the charging 
pumps.  With a charging pump running and the valve leaking, temperature 
stratification occurred in the ECCS pipe as indicated in Figure 1.  In 
addition, temperature fluctuations were found at the location of the failed 
weld with peak-to-peak amplitudes as large as 70 degrees F and with periods 
between 2 and 20 minutes. (1)
______________________________

(1)/ The staff has learned recently of a problem discovered at Trojan in the 
     pressurizer surge line which involved excessive stresses due to thermal 
     stratification.  The staff believes that common elements may exist between
     the Farley 2 event which necessitated this bulletin and the observations 
     at Trojan.  The need for an additional generic communication is being 
     considered as part of our ongoing evaluation of the Trojan event.

8806170291
.                                                            NRCB 88-08 
                                                            June 22, 1988 
                                                            Page 2 of 4 


Discussion:

At Farley 2, dual-purpose pumps are used for charging the RCS with coolant from
the chemical and volume control system during normal operation and injecting 
emergency core coolant at high pressure during a loss-of-coolant accident 
(LOCA).  Separate runs of piping from these pumps are connected to separate 
nozzles on the RCS piping for normal charging flow, backup charging flow, and 
hot-and cold-leg ECCS injection and to a nozzle on the pressurizer for 
auxiliary pressurizer spray.  All of these runs of piping, downstream from the 
last check valve in each pipe, are susceptible to the kind of failure that 
occurred in the ECCS piping connected to the cold leg of loop B.

In any light-water-cooled power reactor, thermal fatigue of unisolable piping 
connected to the RCS can occur when the connected piping is isolated by a 
leaking block valve, the pressure upstream from the block valve is higher than 
RCS pressure, and the temperature upstream is significantly cooler than RCS 
temperature.  Because valves often leak, an unrecognized phenomenon and 
possibly unanalyzed condition may exist for those reactors that can be 
subjected to these conditions.  Under these conditions, thermal fatigue of the 
unisolable piping can result in crack initiation as experienced at Farley 2.  
Cracking has occurred at other plants in Class 2 systems (see IE Bulletin 
79-13, "Cracking in Feedwater System Piping," dated June 25, 1979 and Revisions
1 and 2 dated August 30 and October 16, 1979, respectively).  Subjecting flawed
piping to excessive stresses induced by a seismic event, waterhammer, or some 
other cause conceivably could result in failure of the pipe.  

General Design Criterion 14 of Appendix A to Part 50 of Title 10 of the Code of 
Federal Regulations requires that the reactor coolant pressure boundary be 
designed so as to have an extremely low probability of abnormal leakage, of 
rapidly propagating failure, and of gross rupture.  At Farley 2, the pressure 
boundary failed well within its design life. 

Actions Requested:

1.   Review systems connected to the RCS to determine whether unisolable 
     sections of piping connected to the RCS can be subjected to stresses from 
     temperature stratification or temperature oscillations that could be 
     induced by leaking valves and that were not evaluated in the design 
     analysis of the piping.  For those addressees who determine that there are 
     no unisolable sections of piping that can be subjected to such stresses, 
     no additional actions are requested except for the report required below.  

2.   For any unisolable sections of piping connected to the RCS that may have 
     been subjected to excessive thermal stresses, examine non-destructively 
     the welds, heat-affected zones and high stress locations, including 
     geometric discontinuities, in that piping to provide assurance that 
     there are no existing flaws.     

.                                                            NRCB 88-08 
                                                            June 22, 1988 
                                                            Page 3 of 4 


3.   Plan and implement a program to provide continuing assurance that 
     unisolable sections of all piping connected to the RCS will not be 
     subjected to combined cyclic and static thermal and other stresses that 
     could cause fatigue failure during the remaining life of the unit.  This 
     assurance may be provided by (1) redesigning and modifying these sections 
     of piping to withstand combined stresses caused by various loads including 
     temporal and spatial distributions of temperature resulting from leakage 
     across valve seats, (2) instrumenting this piping to detect adverse 
     temperature distributions and establishing appropriate limits on 
     temperature distributions, or (3) providing means for ensuring that 
     pressure upstream from block valves which might leak is monitored and does
     not exceed RCS pressure.    

4.   For operating plants not in extended outages, Action 1 should be completed 
     within 60 days of receipt of this bulletin, and Actions 2 and 3, if 
     required, should be completed before the end of the next refueling outage.  
     If the next refueling outage ends within 90 days after receipt of this 
     bulletin, then Actions 2 and 3 may be completed before the end of the 
     following refueling outage.  

     For operating plants in extended outages and for plants under 
     construction, Action 1 should be completed within 60 days of receipt of 
     this bulletin or before achieving criticality, whichever is later, and 
     Actions 2 and 3 should be completed before achieving criticality, unless 
     criticality is scheduled to occur within 90 days of receipt of this 
     bulletin.  In that case, Actions 2 and 3 should be completed before the 
     end of the next refueling outage.

Reporting Requirements:

1.   Within 30 days of completion of Action 1, each addressee shall submit a 
     letter confirming that the action has been completed and describing the 
     results of the review.  If the review performed under Action 1 indicates 
     that a potential problem exists, the confirmatory letter shall include a 
     schedule for completing Actions 2 and 3.

2.   Those addressees who determine that there are unisolable sections of 
     piping that can be subjected to stresses from temperature stratification 
     or temperature oscillations that could be induced by leaking valves and 
     that were not evaluated in the design analysis of the piping shall submit 
     a letter within 30 days of completion of Actions 2 and 3.  This letter 
     should confirm that Actions 2 and 3 have been completed and describe the 
     actions taken.

The written reports, required above, shall be addressed to the U.S. Nuclear 
Regulatory Commission, ATTN: Document Control Desk, Washington, D.C.  20555, 
under oath or affirmation under the provisions of Section 182a, Atomic Energy 
Act of 1954, as amended.  In addition, a copy shall be submitted to the appro-
priate Regional Administrator.

.                                                            NRCB 88-08 
                                                            June 22, 1988 
                                                            Page 4 of 4 


This requirement for information was approved by the Office of Management and 
Budget under clearance number 3150-0011.

If you have any questions regarding this matter, please contact one of the 
technical contacts listed below or the Regional Administrator of the 
appropriate NRC regional office.




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


Technical Contacts:  Roger W. Woodruff, NRR
                     (301) 492-1180

                     Pao Kuo, NRR
                     (301) 492-0907

Attachments:
1.  Figure 1 - Farley 2 Temperature Data
2.  Figure 2 - Farley 2 ECCS 
3.  List of Recently Issued NRC Bulletins

Page Last Reviewed/Updated Tuesday, July 23, 2013