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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