IEB 80-04 Analysis of a PWR Main Steam Sine Break with Continued Feedwater Addition (Generic Letter

GL80012                                                  SSINS No. 6820 
                                                         Accession No.: 
                                 UNITED STATES 
                         NUCLEAR REGULATORY COMMISSION 
                            WASHINGTON, D.C.  20555 

                               February 8, 1980 

                                                 IE Bulletin No. 80-04 


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, Main 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. 

Action 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 

      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. 

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 (R0072); clearance expires 7/31/80.  Approval was 
given under a blanket clearance specifically for identified generic 


Plants with construction permits that are required to respond to the 

                              Diablo Canyon


                              North Anna 2

                              Salem 2


If the permit holders have responde 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.


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