ACCIDENT MANAGEMENT ACTIONS
IN AN UPPER-HEAD SMALL-BREAK
LOSS-OF-COOLANT ACCIDENT
WITH HIGH-PRESSURE SAFETY
INJECTION FAILED
KEYWORDS: emergency
operat-
ing procedures, SBLOCA, TRACE
code
CESAR QUERAL,* JUAN GONZALEZ-CADELO, GONZALO JIMENEZ,
and ERNESTO VILLALBA
Universidad
Politécnica de
Madrid,
C/Alenza
4,
28003
Madrid,
Spain
Since the
Three
Mile Island
accident,
an important
focus of pressurized
water reactor (PWR) transient
analy-
ses has been a small-break loss-of-coolant accident
(SBLOCA).
In
2002,
the discovery of
thinning
of the ves-
sel head wall at the Davis Besse nuclear power plant
reactor indicated the possibility of an SBLOCA in the
upper
head of the
reactor
vessel as a
result
of circumfer-
ential cracking of a control rod drive mechanism pen-
etration nozzle—which
has cast
even greater importance
on the study of
SBLOCAs.
Several experimental tests
have
been
performed at the Large Scale
Test
Facility to
simulate the behavior of a PWR during an upper-head
SBLOCA.
The last of
these
tests,
Organisation
for Eco-
nomic Co-operation and Development Nuclear Energy
Agency Rig of Safety Assessment (OECD/NEA ROSA)
Test
6.1,
was
performed in
2005.
This
test
was
simulated
with the
TRACE
5.0
code,
and good
agreement
with the
experimental results
was
obtained.
Additionally, a broad analysis of an upper-head
SBLOCA
with
high-pressure
safety
injection
failed in a
Westinghouse
PWR was performed taking into account
different accident management actions
and
conditions
in
order to check their
suitability.
This issue has been an-
alyzed also in
the
framework of
the
OECD/NEA ROSA
project
and the
Code Applications and
Maintenance
Pro-
gram (CAMP). The main conclusion is that the current
emergency operating
procedures for
Westinghouse
reac-
tor design are
adequate for
these
kinds of
sequences,
and
they do not need to be
modified.
I. INTRODUCTION
Vessel head wall thinning found in the reactor at the
Davis Besse nuclear power plant (NPP) on February 16,
2002,
raised a safety issue regarding vessel structural
integrity; see Fig. 1 and Refs.
1
through 4. Circumferen-
tial cracking of the penetration nozzle of the control rod
drive mechanism (CRDM) could cause a small-break
loss-of-coolant accident (SBLOCA) at the pressure ves-
sel upper head in a pressurized water reactor (PWR).
As part of participation in the Organisation for Eco-
nomic Co-operation and Development Nuclear Energy
Agency Rig of Safety Assessment (OECD/NEA ROSA)
* E-mail: cesar.queral@upm.es
project and the Code Applications and Maintenance Pro-
gram (CAMP), the Universidad Politécnica de Madrid
has
performed a broad analysis of an upper-head SBLOCA
with high-pressure safety injection (HPSI) failed in a
Westinghouse PWR:
1.
In the
first
stage,
simulation of OECD/NEA ROS A
Test 6.1 was performed and compared extensively to the
experimental results.
2.
In the second stage, transients similar to those of
OECD/NEA ROSA Test 6.1 were simulated with the
TRACE model of the Almaraz NPP Unit 1 (Westing-
house three-loop
design).
This analysis took into account
different accident management actions and conditions in
order to check their suitability.
Reactor Vessel Head Degradation Location
Ant of
DIVFI
B*H«
R*ictorv*iMtH«d
I * ReKtor VffMtl
^^ ' H«d Intjlillop
Fig. 1. Reactor vessel head degradation location; Davis Besse NPP (from http://www.nrc.gov).
II.
EMERGENCY OPERATING PROCEDURES RELATED
TO SBLOCA SEQUENCES
In this kind of sequence, i.e., SBLOCA with HPSI
failed, the operators must follow several emergency op-
erating procedures (EOPs). The main tasks of the EOPs
corresponding to the Westinghouse reactor design are
described in Fig. 2 and Ref. 5:
1.
Whenever there is a reactor SCRAM, EOP E-0
(i.e.,
reactor TRIP or safety injection) must be started. In
step 22, the reactor coolant system (RCS) integrity is
checked, and if it is not intact, there is
a
transition to EOP
E-l (loss of reactor or secondary coolant).
2.
In EOP E-l, step 1, the operator checks if the
reactor coolant pumps (RCPs) should be stopped (they
will be stopped by the operator if there is at least one
HPSI pump running and loss of subcooling).
3.
Following EOP E-l, step
11,
the operator checks
the primary pressure, and if it is not below
—15
bars (the
exact value depends on the NPP), there is a transition to
EOPES-1.2 [post-loss-of-coolant-accident (LOCA) cool-
down and depressurization].
REACTOR TRIP OR
SAFETY INJECTION
E-0
LOSS OF REACTOR
OR SECONDARY
COOLANT
E-l
POST LOCA
COOL DOWN AND
DEPRESSURIZATION
ES-1.2
Step I
Vtrily reactor trip
Slcp.12
KvaluaU' Lmg (crm
phuu scams.
Cotd timcdciiwn condilKHi:
Fig. 2. Main steps of the EOPs related to an upper-head SBLOCA.
4.
In EOP
ES-1.2,
the operator cools and depressur-
izes the primary system, opening the steam dump valves,
or if that is not possible, the operator opens the steam
generator relief
valves
at
the
secondary side, making sure
that the cooling of the RCS is close to 55 K/h.
It is important to comment on some considerations with
respect to the two main operator actions: RCP trip and
cooling and depressurizing the primary system by means
of steam generators:
1.
RCP
trip:
The need
to
review the manual
RCP
trip
conditions during an SBLOCA
was
an issue that arose as
a result of the Three Mile Island (TMI) accident. Wes-
tinghouse analyzed this issue for SBLOCA sequences
with and without HPSI (see Refs. 6, 7, and 8), and the
main conclusions were the following:
a. If
the
HPSI
is
available,
the
RCP must
be
tripped
at the
beginning of SBLOCA sequences
in
order
to avoid worse consequences following a de-
layed RCP trip.
b.
If the HPSI is not available, it is better to not
trip the RCPs in order to cool the core with a
high mass flow.
In other designs like the current Siemens reactors,
EPR and
APIOOO,
there is an automatic trip coincident
with the safety injection system (SIS) demand, regard-
less of HPSI availability.
911
In the French reactor design
there is also a manual trip in the A1.2 procedure (corre-
sponding to the LOCA sequence).
12
2.
Primary-side
cooling:
If the RCS is in saturation
conditions, it is possible to obtain the equivalence of a
55 K/h cooling rate (following EOP ES-1.2, as men-
tioned earlier) in bars per hour, from Figs. 3 and 4 (it
must be noted that this equivalence is valid only in sat-
uration conditions). In other designs like the current Sie-
mens reactors and EPR, the cooling rate during an
SBLOCA
is
100 K/h, and cooling is performed automat-
ically by the protection system.
910
In
the APIOOO
there is
no secondary-side depressurization, and the primary-
side depressurization is performed using an automatic
depressurization system (ADS) with four stages, which
is necessary when the core makeup tank is below 70%
(Refs.
11 and 13). In
the APIOOO
design, EOPs direct the
operator to actuate the normal residual heat removal sys-
tem (RHRS) in order to avoid the actuation of the fourth
ADS stage.
14
For the French reactor design, the operators
must follow procedure
Al.
1
(small primary system break)
during an SBLOCA. The objective of this EOP is to cool
the RCS with the steam generator to conditions that en-
able implementing the RHRS, which it is similar to EOP
ES-1.2.
If
the
accident management actions included in EOP
E-0, EOP E-l, and EOP ES-1.2 are not enough to avoid
core damage or if there is an error or delay in operator
actions, then it is possible to get inadequate core cooling
(ICC) conditions; see Refs. 15 through 20 for more de-
tails.
In this case the operators must follow Status Tree
F.0.2 (core cooling) and EOP FR C.l (response to ICC)
and EOP FR C.2 (response to degraded core cooling).
The status tree that is related to the critical function of
core cooling is F.0.2 (Westinghouse design); see Fig. 5.
This status tree directs the operators to the function re-
covery guideline (FRG) that must be used depending on
the values of several parameters. In this case Status Tree
400
375
350
325
300
275
250
225
200
175
150
125
100
75
50
25
0
_^_—-—"""'""
^^-~^"^
J*^*"^
25 50
75 100
Pressure (bar)
125 150 175
200
Fig. 3. Saturation temperature.
Pressure (bar)
100
100 200 300
Temperature (C)
Fig. 4. Primary depressurization rate in saturated conditions
for a cooling rate of 55 K/h.
the Westinghouse reactor design, which is a differential
pressure measuring system for determining the collapsed
water level in the reactor vessel; see Refs. 23 and 24 for
more details. In other reactor designs electrical resis-
tance detectors at different vessel levels are used instead
of the RVLIS (Refs. 25 and 26). Both instrumentation
systems, CET and
RVLIS,
are part of
the
ICC instrumen-
tation system, which has been required since the TMI
accident in 1979 (Ref. 15).
In EOP FR C.2, the operator will cool down the
primary side with a maximum cooling rate of
55
K/h (as
in EOP ES-1.2), and in EOP FR C.l, the operator will
fully open all secondary-side relief valves. In several
simulations it has been observed that the cooling rate
with full opening is near 300 K/h.
The generic probabilistic risk analysis of the French
reactor design (see Ref. 12) mentions that in the event of
failure of HPSI during LOCA sequences, the operator
will trigger an accelerated cooling by the steam generator
(task included in procedure Ul), making it possible to
attain low-pressure safety injection (LPSI) operation con-
ditions (similar to EOP FR C.l). The time available for
this operation is estimated to be 1 h.
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Fig. 5. Status tree related to the FRG of core cooling (i.e..
F.0.2).
F.0.2 selects the FRG EOP FR C.l or EOP FR C.2 de-
pending on the core exit thermocouple (CET) tempera-
tures (see Refs.
21
and 22) and the vessel level, measured
by the reactor vessel level indicator system (RVLIS) in
III.
LARGE SCALE TEST FACILITY: 0ECD/NEA ROSA
TEST 6.1
The Large Scale Test Facility (LSTF) is a full-
height, full-pressure, 1/48 volumetrically scaled simu-
lator for a Westinghouse-type four-loop
[3423-
MW(thermal)] PWR with primary and secondary coolant
systems including an electrically heated simulated core,
emergency core cooling systems (ECCSs), and control
systems for accident management actions; see Ref. 27
and Fig. 6 for more details. The maximum core power
Pressurizer fi\
— 10m High p
1
141
U-tubes
Accumulator
Cold
Steam
Generator
29 m
Primary
Coolant
Pump
Pressure
Vessel
Fig. 6. Large Scale Test Facility.
of 10 MW is equivalent to 14% of the
1/48-scaled
PWR
rated power covering the scaled PWR decay heat after
the scram.
The Universidad Politécnica de Madrid has been
working with the ROS A/LSTF TRACE model since Feb-
ruary 2006 (Refs. 28, 29, and 30). The TRACE model is
based on the TRAC-PF1 model presented by the Japan
Atomic Energy Research Institute (JAERI) to the partici-
pants of the OECD/NEA ROSA project. The main tasks
performed in translating and modifying the model are the
following:
1.
The old STGEN component was translated to the
TRACE model as a set of components (TEEs and PIPEs),
conserving volumes and lengths. The steam generator
recirculation ratio was adjusted. Later, a new steam gen-
erator model with nine different heights of tubes was
developed.
2.
The old VESSEL component was translated to
the TRACE model, and
the
temperature in the upper head
of the vessel was adjusted
to
the measured
one
(—586
K).
The REFLOOD model was activated.
3.
The total mass flow was adjusted in the primary
loops using friction coefficient (FRIC) parameters and
the rated head in the RCP The mass flow rate from the
downcomer to the upper head of the vessel was adjusted
to the specified one
(0.3%
of the downcomer vessel total
mass flow).
4.
Volume-versus-height plots were checked with
respect to the facility data, and all the volume and height
discrepancies were corrected.
5.
A
new two-dimensional model of the pressurizer
was created to avoid excessive cooling in the upper cells
of the model during long quasi-steady-state transients,
which was a problem with the earlier model. Also, sta-
bilization of the pressurizer level and pressure control
systems was included to adequately fix the steady state.
Finally, new, more detailed proportional and base heaters
were also added.
6. Heat losses and pressure drops of
the
whole model
were adjusted.
7.
The OFFTAKE model was activated in the con-
nections of the valves that simulate breaks in different
localizations of the LSTF
8. An animation mask was created with the SNAP
application; see Fig. 7. This mask allows videos of the
simulations to be performed, which allows the transient
behavior to be easily interpreted.
The ROSA/LSTF TRACE model has 178 thermal-
hydraulic components (2 VESSEL, 45 PIPE, 8 TEE, 2
Fig. 7. SNAP mask of ROSA/LSTF. Void fraction in primary and secondary sides during an upper-head SBLOCA.