Part 21 Report - 1995-023

ACCESSION #:  9501130161

ROBERT E. DENTON              Baltimore Gas and Electric Company
                              Calvert Cliffs Nuclear Power Plant
Vice President                1650 Calvert Cliffs Parkway
Nuclear Energy                Lusby, Maryland 20657
                              410 586-2200 Ext. 4455 Local
                              410 260-4455 Baltimore

BGE                      January 10, 1995


U. S. Nuclear Regulatory Commission
Washington, DC 20555

ATTENTION:     Document Control Desk

SUBJECT:       Calvert Cliffs Nuclear Power Plant
               Unit Nos. 1 & 2; Docket Nos. 50-317 & 50-318
               10 CFR Part 21 Report; Non-Conservative Modeling of
               Reactor Coolant System 
               Sensible Heat For Containment Pressure Response Safety
               Analysis

During a review of our Updated Final Safety Analysis Report Safety
Analysis concerning containment pressure response, we determined the
Bechtel analysis of the long-term cooling phase of a loss of coolant
accident did not model heat transfer from Reactor Coolant System (RCS)
metal components to the RCS coolant.  This omission potentially results
in a non-conservative calculated containment temperature during a
specific time period of the analysis (after containment peak temperature
until several days after the event).  Although we have concluded this
non-conservative assumption has no safety significance for Calvert
Cliffs, we are reporting it under Part 21 because this problem may
potentially represent a safety consequence to other licensees who use
similar methodologies.

Bechtel has informed us that they are evaluating the generic
implications, if any, of this modeling omission and will report the
results of their evaluation to us.

A verbal notification and written summary were submitted to the Nuclear
Regulatory Commission Operations Center via facsimile on December 9,
1994.

Should you have any questions regarding this matter, we will be pleased
to discuss them with you.

                         Very truly yours,

RED/CDS/bjd

Attachment



Document Control Desk
January 10, 1995
Page 2

cc:  D.  A.  Brune, Esquire
     J.  E.  Silberg, Esquire
     L.  B.  Marsh, NRC
     D.  G.  McDonald, Jr., NRC
     T.  T.  Martin, NRC
     P.  R.  Wilson, NRC
     R.  I.  McLean, DNR
     J.  H.  Walter, PSC

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                 10 CFR PART 21 REPORT; NON-CONSERVATIVE
              MODELING OF RCS SENSIBLE HEAT FOR CONTAINMENT
                 PRESSURE RESPONSE SAFETY ANALYSIS COULD
              RESULT IN A SLIGHT INCREASE IN POST-ACCIDENT
                         CONTAINMENT TEMPERATURE

Calvert Cliffs Nuclear Power Plant, Units 1 and 2

Docket Nos.  50-317 and 50-318

(i)       Name and address of individual making notification:

          R. E. Denton, Vice-President, Nuclear Energy 
          Baltimore Gas and Electric Company
          Calvert Cliffs Nuclear Power Plant
          1650 Calvert Cliffs Parkway
          Lusby, MD 20657-4702

(ii)      Basic Component Affected:

          Updated Final Safety Analysis Report Chapter 14.20,
          "Containment Pressure Response." Specifically the long-term
          cooling phase modeled by Bechtel's Containment Pressure and
          Temperature Transient Analysis (COPATTA) Code.

(iii)     Firms Supplying Component:

          Bechtel Power Corporation
          
(iv)      Nature of Defect:

          Chapter 14.20 of our Updated Final Safety Analysis Report
          (UFSAR), "Containment Pressure Response," is an analysis of the
          pressure and temperature response of our containments to design
          basis accidents such as a main steam line break or a loss of
          coolant accident (LOCA).  A spectrum of Reactor Coolant System
          (RCS) break sizes were considered to determine the worst
          condition of RCS mass and energy releases in combination with
          sensible and shutdown heat sources during the blowdown phase of
          a LOCA.  The containment response to these breaks was analyzed
          assuming various limiting single failures.

          The RCS blowdown transient results in primary containment
          pressure and temperature peaks as a result of the mass and
          energy transferred from the reactor core to the primary coolant
          and to the containment atmosphere.  During the refill and
          reflood phases of the accident scenario, heat in the steam
          generator water mass is transferred to the primary coolant via
          a reverse heat flow and then into the containment atmosphere. 
          In addition, safety injection water reflooding into an
          uncovered core and the hot RCS system picks up heat from those
          sources and deposits it into the Containment as saturated or
          even superheated steam.

          The mass and energy transfer from the RCS for various phases of
          the accident are calculated by Combustion Engineering (CE) and
          Bechtel.  The blowdown phase of the LOCA is modeled using the
          CE FLASH code, the refill and reflood phases by the CE FLOOD
          code, and the long-term

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                 10 CFR PART 21 REPORT; NON-CONSERVATIVE
              MODELING OF RCS SENSIBLE HEAT FOR CONTAINMENT
                 PRESSURE RESPONSE SAFETY ANALYSIS COULD
              RESULT IN A SLIGHT INCREASE IN POST-ACCIDENT
                         CONTAINMENT TEMPERATURE

          cooling phase by Bechtel.  The mass and energy transfer data is
          input to Bechtel's Containment Pressure and Temperature
          Transient Analysis (COPATTA) code for calculation of
          containment pressure and temperature.  During the long-term
          cooling phase (after reflood) the transfer of sensible heat
          from the RCS metal back into the coolant is not modeled.  When
          RCS metal sensible heat is included, the result is a higher
          enthalpy coolant flowing from the RCS break into Containment. 
          The higher enthalpy coolant flowing into the containment leads
          to slightly higher containment temperatures and pressures for
          several days after their peaks.  Preliminary analysis indicates
          the problem has no effect on containment peak pressure or peak
          temperature.  We have concluded that there are no adverse
          effects to our environmental qualification program.

(v)       Date on Which Defect Was Identified:

          The problem was identified by BGE during a review of the UFSAR
          Chapter 14.20 Safety Analysis, and documented on an Issue
          Report on November 9, 1994.

(vi)      Number and Location of Components: Not applicable.


(vii)     Corrective Actions Taken:

          We have asked CE to provide new mass and energy transfer data
          that accounts for sensible heat transfer from the RCS metal to
          the coolant.  The revised data produced by CE will be provided
          to Bechtel to produce revised containment pressure and
          temperature response curves, The results of the revised
          containment response curves are expected to show:

          A.   Containment primary peak pressure and temperature will be
               unaffected.

          B.   The intermediate containment temperature will be increased
               by less than 2 degrees F.

          C.   The containment temperature and pressure will be
               essentially unaffected beginning several days after the
               start of the event.

          The results of this reanalysis are being evaluated for impact
          on other aspects of our current licensing basis.  The most
          significant potential impact was the increased load on our SRW
          system via the containment air coolers.  We have no current
          operability concerns due to low ultimate heat sink temperatures
          at the present time and expect that the final reanalysis will
          show the real effect on our current safety analysis margins
          will be minimal.

          Bechtel has informed us that they are evaluating the generic
          implications, if any, of this modeling omission and will report
          the results of their evaluation by January 20, 1995.

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