United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 93-55: Potential Problem with Main Steamline Break Analysis for Main Steam Vaults/Tunnels

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                           WASHINGTON, D.C.   20555

                                 July 21, 1993

                               BREAK ANALYSIS FOR MAIN STEAM 


All holders of operating licenses or construction permits for pressurized
water reactors. 


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to a potential inadequacy in the main steamline
break analysis which could place some pressurized-water reactor (PWR) plants
outside their current structural design basis for the main steam valve vaults
or main steam tunnels.  The plants of concern are those that must postulate a
double-ended rupture of a main steamline in these areas.  It is expected that
recipients will review the information for applicability to their facilities
and consider actions, as appropriate, to avoid similar problems.  However,
suggestions contained in this information notice are not NRC requirements;
therefore, no specific action or written response is required. 

Description of Circumstances

During the Watts Bar Calculation Reconstitution Program, Tennessee Valley
Authority (TVA) discovered that Westinghouse had supplied data for the main
steamline break analysis which assumed perfect moisture separation.  TVA used
this data to determine the mass and energy released into the main steam valve

TVA had requested Westinghouse to reevaluate the 1975 Westinghouse mass and
energy release data used in the Watts Bar analysis for these valve vaults, and
to advise TVA if the data were still applicable.  On June 23, 1992,
Westinghouse advised TVA that the 1975 mass and energy release data were not
considered conservative, and were not applicable for a pressure transient
evaluation of the vented main steam valve vaults.  The mass and energy release
did not account for liquid entrainment in the blowdown, and resulted in a
slower mass and energy release rate in the 1975 data.

Westinghouse then provided a bounding analysis based on ANSI/ANS Standard 58.2
(1980) methodology which included liquid entrainment in the blowdown.  This
new analysis indicated that the valve vault structural design pressure would
be exceeded in the event of a steamline break within the vaults. 


                                                            IN 93-55
                                                            July 21, 1993
                                                            Page 2 of 3

On August 17, 1992, for Sequoyah Nuclear Plant (LER 50-327/92-013), and on
October 13, 1992, for Watts Bar Nuclear Plant (CDR 50-390/92-09), TVA reported
the use of nonconservative Westinghouse data for the main steamline break
analysis which could result in the valve vault structural design pressure
being exceeded.  TVA had determined that the mass and energy release data for
Watts Bar were also applicable to Sequoyah.  TVA prepared a justification for
continued operation (JCO) for Sequoyah.  This JCO will be in effect until the
startup from refueling Cycle 6 for both Sequoyah Unit 1 (Fall 1993) and Unit 2
(early 1994). 


The 1975 mass and energy release data supplied by Westinghouse was based upon
the largest steam generator depressurization rate consistent with a high-
quality steam discharge.  The Westinghouse data was applicable for a
postulated double-ended main steamline break in the turbine building, assuming
flow in both the forward and reverse direction.  Apparently, TVA applied this
data without verifying its applicability to vented compartments, such as the
main steam valve vaults.  A dry steam release in a vented compartment such as
the main steam valve vault may not be conservative, because of moisture
entrainment within the discharge. 

For a main steamline break analysis, the limiting plant conditions for the
steam generator mass inventory and secondary system pressure are often at hot
standby/shutdown plant conditions (0 percent power level, primary plant at
operating temperature and pressure).  Due to the high flow rates associated
with the main steamline break, frothing in the steam generator raises the
water level rapidly, which decreases the quality of fluid expelled from the
steam generator.  Although the enthalpy of this low-quality fluid is less than
the enthalpy of dry steam, the critical mass flow is 4 to 5 times higher,
resulting in a net increase in the energy release rate from the break.  This
may be the limiting case for determining maximum pressure in vented (blowout
panels) compartments. 

Westinghouse recommended to TVA that the methodology outlined in ANSI/ANS
Standard 58.2 (1980), Appendix E be used to generate the Watts Bar bounding
mass and energy release rates, which would determine the pressure inside the
valve vaults.  This mass and energy release data would include the entrainment
of water and bound the analyses that could be conducted for this type of
event.  Westinghouse performed the Watts Bar analysis with the ANSI/ANS 58.2
methodology and informed TVA of a significant increase in the mass and energy
releases generated over those of the original analysis. 

TVA determined that the increased Watts Bar mass and energy release rates
produced pressures that exceeded the present structural design margins, and
challenged the structural adequacy of the walls and slabs of the main steam 
valve vaults.  TVA calculations showed that the peak pressures in the valve
vaults could increase by about one-third when moisture entrainment was 

                                                            IN 93-55
                                                            July 21, 1993
                                                            Page 3 of 3

considered.  Failure of the valve vault walls or slabs could damage such
equipment as main steam system, main feedwater system, and auxiliary feedwater
system components and piping.  This equipment damage could result in the
inability (or reduced ability) to feed the intact steam generators, or in the
blowdown of more than one steam generator. 

Upon consultation with Westinghouse, TVA determined that the analysis data for
the Sequoyah main steam valve vault rooms were also nonconservative.  A JCO
has been prepared for Sequoyah.  The JCO is based on the Sequoyah main steam
system piping design in the valve vaults meeting most of the break exclusion
provisions of the Standard Review Plan (SRP) Branch Technical Position (BTP)
MEB 3-1, "Postulated Rupture Locations in Fluid System Piping Inside and
Outside Containment."  A postulated one-square-foot break was analyzed for the
JCO interim period.  The revised calculated pressures (using the ANSI/ANS 58.2
methodology) were bounded by the original design pressure of the vaults.  This
JCO will be in effect until the next Sequoyah, Units 1 and 2 refueling outages
(Cycle 6 for both units).  TVA will make plant modifications to bring the
plant into compliance with the original design basis.  The modifications will
involve modifying each of the fluid head anchor-sleeve openings to decrease
the flow area in the event of a postulated break, thereby limiting the mass
and energy release rate into the valve vaults.  The flow area will be sized to
limit the pressure in the main steam valve vaults to less than the original
design basis of the floor and walls.

Combustion Engineering and Babcock & Wilcox designed PWRs may also be affected
by this issue if vented compartments have been analyzed nonconservatively,
assuming dry steam.  Therefore, this information notice is being sent to all
PWR licensees and holders of PWR construction permits. 

This information notice requires no specific action or written response.  If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation project manager.     

                                     /s/'d by BKGrimes

                                    Brian K. Grimes, Director
                                    Division of Operating Reactor Support
                                    Office of Nuclear Reactor Regulation 

Technical contacts:   J. B. Brady, RII
                      (404) 331-0339

                      W. T. Lefave, NRR
                      (301) 504-3285

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