check valve, elbow, Farley Nuclear Power Plant, high-cycle thermal fatigue, leakage, leaky valves, nozzle, nuclear energy, nuclear power, Nuclear Power Stations, pipe spools, reactor coolant systems, thermal stress, thermal stresses in piping reactor coolant, unisolable piping, valves
“OMB No.: 3150-0011
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
All holders of operating licenses or construction permits for
light-water-cooled nuclear power reactors.
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.
. NRCB 88-08
June 22, 1988
Page 2 of 4
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.
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.
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
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
Pao Kuo, NRR
1. Figure 1 – Farley 2 Temperature Data
2. Figure 2 – Farley 2 ECCS
3. List of Recently Issued NRC Bulletins
One response: http://pbadupws.nrc.gov/docs/ML0123/ML012350311.pdf