IEB 80-04 Analysis of a PWR Main Steam Sine Break with Continued Feedwater Addition (Generic Letter
GL80012 SSINS No. 6820
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
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
North Anna 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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