United States Nuclear Regulatory Commission - Protecting People and the Environment

Resolution of Generic Issue 79, "Unanalyzed Reactor Vessel (PWR) Thermal Stress During Natural Convection Cooldown" (Generic Letter 92-02)

March 6, 1992 


               (GENERIC LETTER 92-02)

The U.S. Nuclear Regulatory Commission (NRC) is providing this letter to 
inform addressees of (1) the NRC's resolution of Generic Issue 79, 
"Unanalyzed Reactor Vessel (PWR) Thermal Stress During Natural Convection 
Cooldown" and (2) the conclusions reached by the staff as the result of the 
evaluations performed to resolve this generic issue.  No new requirements 
are being established and no specific action or written response is 


On May 5, 1981, the NRC issued Generic Letter (GL) 81-21, "Natural 
Circulation Cooldown," in response to a natural circulation cooldown (NCC) 
event that occurred at the St. Lucie Plant, Unit 1, on June 11, 1980.  That 
event caused a void (steam bubble) to form in the reactor vessel head.  In 
GL 81-21, addressed to all operating PWR power reactor licensees and 
applicants for operating licenses (except for St. Lucie, Unit 1), the NRC 
requested that addressees determine whether operator training and plant 
procedures were adequate to effect a controlled NCC from operating 
conditions to cold shutdown.  The NRC requested addressees to demonstrate 
their capability by test or analysis or both in accordance with Section 
50.54(f) of Title 10 of the Code of Federal Regulations (10 CFR 50.54(f)). 

By letter of March 18, 1983, the Babcock & Wilcox Company (B&W) notified the 
NRC that large axial temperature gradients across the RV closure region may 
cause thermal stresses, beyond those considered in the original design of 
PWR vessels, to develop in the reactor vessel (RV) flanges and studs.  This 
condition could be outside the design basis of the PWR RVs.  During an NCC 
event, the upper head of a PWR vessel is likely to remain at a higher 
temperature than the cylindrical portion of the vessel because there is 
little or no mixing of the fluid in this region with the remainder of the 
fluid in the reactor vessel.  Further, a steam bubble may develop in the top 
of the vessel as the reactor coolant system is depressurized.  The NRC 
determined that this concern could be a generic safety issue and designated 
it as Generic Issue 79 (GI-79). 


B&W performed a detailed analysis of the B&W 177 Fuel Assembly Reactor 
Vessel (B&W 177) and submitted it to the NRC by letter of October 15, 1984.  
The NRC used an independent confirmatory analysis performed by the 
Brookhaven National Laboratory (BNL) in May 1989, to evaluate the B&W submittal regarding the 


Generic Letter 92-02                 - 2 -                  March 6, 1992 

stresses in the reactor vessel and the reactor vessel closure studs.  The 
NRC staff also performed a detailed fracture mechanics evaluation of the 
nozzle shell course and the reactor vessel closure studs.  The staff 
discussed these analyses in NUREG-1374, "An Evaluation of PWR Reactor Vessel 
Thermal Stress During Natural Convection Cooldown," May 1991, which is 
enclosed.  The NRC concluded that the B&W 177 meets the currently applicable 
regulatory design stress and fracture prevention criteria for NCC transient 
conditions up to and including those used by the NRC and its contractor in 
these analyses, as shown in Figure 3 of NUREG-1374. 

In 10 CFR 50.73(a)(2)ii(A) and (B), the NRC requires the licensee to submit 
a licensee event report for any event that resulted in the nuclear power 
plant being in an unanalyzed condition that significantly compromised plant 
safety or in a condition that was outside the design basis of the plant.  
The analysis noted above considers a B&W 177 to be in an analyzed condition 
and within its design basis for NCC events that are bounded by the NCC 
transient profile shown in Figure 3 of NUREG-1374. 

The detailed analyses by B&W, NRC, and BNL indicated clearly the extremely 
complex nature of this type of analysis.  This analysis included numerous 
thermal-hydraulic and mechanical modeling assumptions which, although 
considered to be conservative, were not confirmed by specifically measured 
data.  Calculated stress results for the B&W 177 were as high as 98-percent 
of allowable values in the RV studs specified in the American Society of 
Mechanical Engineers (ASME) Code.  While the Code allowable value includes 
margins, differences between the stresses calculated by B&W and those 
calculated by BNL, indicated that an RV could be in an unanalyzed condition 
for certain NCC events, particularly for events complicated by other factors 
such as an atmospheric dump valve that is stuck open. 

The limitations of the analysis, as stated above, prevented the staff from 
making a definitive conclusion regarding compliance with the applicable 
regulatory criteria of B&W 177s that might experience an NCC that is outside 
the bounds of the analysis assumptions, or for B&W non-177s and other PWR 
vessels that may experience a significant NCC event in the future.  However, 
the staff reviewed the results of the analyses and the qualitative 
extrapolation of those results and concluded the following: 

1.  The B&W 177 is considered analyzed for NCC events that are bounded 
    by the NCC transient profile shown in Figure 3 of NUREG-1374. 

2.  It is extremely unlikely that a single NCC event will cause the 
    failure of any U.S. PWR RV, even if a cooldown rate of 100 �F per hour 
    is exceeded.

3.  An NCC event that does not exceed a total cooldown of 100 �F, 
    independent of rate, would not be expected to compromise the safety 
    of any U.S. PWR RV.  However, it may result in the RV being outside 
    its documented design basis.

Generic Letter 92-02                 - 3 -                  March 6, 1992 

4.  Exposure of U.S. PWR RVs to certain NCC transients, particularly 
    transients complicated by other factors such as a stuck-open atmospheric 
    dump valve, may result in a condition that is outside the documented 
    design basis of the RV. 

The NRC staff has further concluded that (1) NCC events of the type 
analyzed, which result in the plant being brought to a cold shutdown 
condition occur infrequently and (2) the actual severity of a specific NCC 
event will determine the need for (if any) and the extent of actions that 
may be required of any licensee following certain NCC events that may place 
a reactor vessel in an unanalyzed condition or outside its documented design 
basis.  Therefore, no requirement for generic or plant-specific actions was 
deemed necessary for safety reasons. 

Backfit Discussion 

The NRC is establishing no new requirements in this generic letter and is 
requiring no specific action.  Existing regulations address any calculations 
that may be required to be performed after an NCC event.  Therefore, the NRC 
is not imposing a backfit. 

This generic letter contains no requirements for collecting information and 
therefore is not subject to the requirements of the Paperwork Reduction Act 
of 1980 (44 U.S.C. 3501 et seq.). 

Although no response to this letter is required, if you have any questions 
regarding this matter, please contact the technical contact listed below. 


                                        James G. Partlow 
                                        Associate Director for Projects 
                                        Office of Nuclear Reactor Regulation

Enclosure:  NUREG-1374

Technical Contact: 
J. D. Page, RES
(301) 492-3941 
Page Last Reviewed/Updated Friday, May 22, 2015