Bulletin 80-04: Analysis of a PWR Main Steam Line Break with Continued Feedwater Addition

                                                           SSINS No.: 6820
                                                           Accessions No.:
                                                           7910250504

                                UNITED STATES
                        NUCLEAR REGULATORY COMMISSION
                    OFFICE OF INSPECTION AND ENFORCEMENT
                           WASHINGTON, D.C. 20555 
                                     
                              February 8, 1980 

                                                      IE Bulletin No. 80-04 

ANALYSIS OF A PWR MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION 

Description of Circumstances: 

Virginia Electric and Power Co. submitted a report to the Nuclear Regulatory
Commission dated September 7, 1979 that identified a deficiency in the 
original analysis of containment pressurization as a result of reanalysis of
steam line break for North Anna Power Station, Units 3 and 4. 

Stone and Webster Engineering Corporation performed a reanalysis of 
containment pressure following a main steam line break and determined that, 
if the auxiliary feedwater system continued to supply feedwater at runout 
conditions to the steam generator that had experienced the steam line break,
containment design pressure would be exceeded in approximately 10 minutes. 
The long term blowdown of the water supplied under runout conditions by the 
auxiliary feedwater system had not been considered in the earlier analysis. 

On October 1, 1979, the foregoing information was provided to all holders of
operating licenses and construction permits in IE Information Notice No. 
79-24. The Palisades facility did an accident analysis review pursuant to 
the information in the notice and discovered that with offsite power 
available, the condensate pumps would feed the affected generator at an 
excessive rate. This excessive feed was not considered in the analysis for 
the steam line break accident. 

On January 30, 1980, Maine Yankee Atomic Power Company informed the NRC of 
an error in the main steam line break analysis for the Maine Yankee plant. 
During a review of the main steam line break analysis, for zero or low power
at the end of core life, the licensee identified an incorrect postulation 
that the startup feedwater control valves would remain positioned "as is" 
during the transient. In reality, the startup feedwater control valves will 
ramp to 80% full open due to an override signal resulting from the low steam
generator pressure reactor trip signal. Reanalysis of the event shows the 
opening of the startup valve and associated high feedwater addition to the 
affected steam generator would cause a rapid reactor cooldown and resultant 
return-to-power, a condition outside the plant design basis. 

Actions to be Taken by the Licensee: 

For all pressurized water power reactors with an operating license and those
reactors listed in Enclosure 1: 

1.   Review the containment pressure response analysis to determine if the 
     potential for containment overpressure for a main steam line break 
.

IE Bulletin No. 80-04                                       February 8, 1980
                                                            Page 2 of 3 

     inside containment included the impact of runout flow from the 
     auxiliary feedwater system and the impact of other energy sources, such 
     as continuation of feedwater or condensate flow. In your review, 
     consider your ability to detect and isolate the damaged steam generator 
     from these sources and the ability of the pumps to remain operable 
     after extended operation at runout flow. 

2.   Review your analysis of the reactivity increase which results from a 
     main steam line break inside or outside containment. This review should
     consider the reactor cooldown rate and the potential for the reactor to
     return to power with the most reactive control rod in the fully 
     withdrawn position. If your previous analysis did not consider all 
     potential water sources (such as those listed in 1 above) and if the 
     reactivity increase is greater than previous analysis indicated the 
     report of this review should include: 

     a.   The boundary conditions for the analysis, e.g., the end of life 
          shutdown margin, the moderator temperature coefficient, power 
          level and the net effect of the associated steam generator water 
          inventory on the reactor system cooling, etc., 

     b.   The most restrictive single active failure in the safety injection
          system and the effect of that failure on delaying the delivery of 
          high concentration boric acid solution to the reactor coolant 
          system, 

     c.   The effect of extended water supply to the affected steam 
          generator on the core criticality and return to power, 

     d.   The hot channel factors corresponding to the most reactive rod in 
          the fully withdrawn position at the end of life, and the Minimum 
          Departure from Nucleate Boiling Ratio (MDNBR) values for the 
          analyzed transient. 

3.   If the potential for containment overpressure exists or the reactor-
     return-to-power response worsens, provide a proposed corrective action 
     and a schedule for completion of the corrective action. If the unit is 
     operating, provide a description of any interim action that will be 
     taken until the proposed corrective action is completed. 

4.   Within 90 days of the date of this Bulletin, complete the review and 
     evaluation required by this Bulletin and provide a written response 
     describing your reviews and actions taken in response to each item. 

Reports should be submitted to the Director of the appropriate NRC Regional 
Office and a copy should be forwarded to the NRC Office of Inspection and 
Enforcement, Division of Reactor Operations Inspection, Washington, D.C. 
20555. 
.

IE Bulletin No. 80-04                                       February 8, 1980
                                                            Page 3 of 3 

For boiling water reactors with an operating license or a construction 
permit and all pressurized water reactors with a construction permit, not 
listed in Enclosure 1, this Bulletin is for information purposes only and no 
written response is required. 

Approved by GAO, B180225 (ROO72); clearance expires 7/31/80. Approval was 
given under a blanket clearance specifically for identified generic 
problems. 
.

Plants with construction permits that are required to respond to the 
bulletin: 

                              Diablo Canyon 
                                 McGuire 
                               North Anna 2 
                                 Salem 2 
                                 Sequoyah 

If the permit holders have responded to earlier requests from the NRC on 
some of the items presented in the bulletin, they may respond to the 
bulletin by reference to the response to the earlier request. 







                                                            Enclosure No. 1