ML17289B002
| ML17289B002 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 11/05/1992 |
| From: | Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17289B001 | List: |
| References | |
| 50-397-92-37, EA-92-206, NUDOCS 9212010147 | |
| Download: ML17289B002 (26) | |
See also: IR 05000397/1992037
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION V
Report No:
Docket No:
License
No:
Licensee:
Facil ity Name:
Inspection at:
50-397/92-37
EA 92-206
50-397
Washington Public Power Supply System
P. 0.
Box 968
Richland,
WA 99352
Washington Nuclear Project
No.
2
(WNP-2)
WNP-2'ite near Richland,
Inspection
Conducted:
October
5 October 21,
1992
Inspectors:
W. P. Ang, Acting Senior Resident
Inspector
D. L. Proulx, Resident
Inspector
Approved by:
~Summar:
P.
H.
nson, Chief
'eacto
ojects Section
1
'/6'(q z
Date Signed
ns ection
on October
5 October
21
1992
50-397 92-37
t d: t
p
l
1 l
p tl
d t
d t
identified by the
NRC Augmented Inspection
Team relevant to the WNP-2, August
15,
1992 cor e power oscillation event.
Inspection
Procedure
92701
was used.
Safet
Issues
Mana ement
S stem
SINS
Items:
None.
results:
General
Conclusions
and
S ecific Findin
s
i 'cant Safet
Matters:
Several significant safety issues
were previously identified in the
Augmented Inspection
Team (AIT) Report.
~
Control rod patterns
resulted
in strongly skewed radial
and axial
flux profiles that caused
the reactor to become unstable while
oper'ating in a stability region where the
BWR Owners'roup
had
9212010147
921105
ADOCK 05000397
8
-r,
recommended
additional precautions.
Inadequate
procedures
and
ineffective training resulted in the nonconservative
rod pattern in
place during the event
(paragraphs
3 and 5).
~
The design
and the associated
reviews performed for the Cycles
7 and
8 core reload mixed fuel design did not adequately
evaluate
core
stability and changes to the thermal hydraulic properties of the
core
(paragraph
4).
~
The licensee's
handling of generic correspondence
did not appear to
be adequate
or in compliance with the Technical Specifications,
and
. contributed directly to the concern
above involving strongly skewed
flux profiles (paragraph
6).
Although no safety limits were violated,
and
no fuel damage
occurred,
the in-phase oscillations which were experienced
indicated
a potential
for fuel 'damage.
The core could, under slightly different circumstances,
have experienced
out-of-phase oscillations which might not have
been
detected
and suppressed
before the core exceeded
the Minimum Critical
Power Ratio
(MCPR) safety limit.
The violations identified below involved multiple barriers that failed to
.preclude the event.
There
was minimal evidence of management
oversight
in assuring that
a March 18,
1992,
BWR Owners'roup
(BWROG) letter,
which recommended
additional precautions
regarding
power oscillations,
was appropriately reviewed.
Weakness
in management
oversight
was also
evident in inadequate
procedures.
This lack of management
oversight
appears to have
been the primary reason for the licensee's
failure to
preclude the event.
Summar
of Violations and Deviations:
Five apparent violations and
one
deviation were identified during this inspection,
as follows:
~
The licensee failed to provide adequate
procedures for developing
control rod patterns,
in that the procedures
in place provide'd
=inadequate direction to prevent
power oscillations
(paragraph 3).
~
The licensee's
Nuclear Safety Assurance
Group
(NSAG) did not proper-
ly review and provide recommended
corrective actions for the
letter of March 18,
1992,
as required
by =TS 6.2.3. 1 (paragraph 6).
Core design,
plant procedures,
and operator training were inadequate
at the time of the event to ensure
compliance with 10 CFR 50, Appen-
dix A, General
Design Criterion 12 (paragraphs
4.a, 4.b,
and 5).
The licensee's
application of a mixed core design for the cycle 8
reload core,
and inadequate
analysis
and evaluation of related
changes
in thermal
and hydraulic properties of the design,
did not
receive
an adequate
design review (paragraph 4.c).
Operators .failed to adjust Average
Power
Range Monitor (APRM), flow-
biased
and rod block setpoints
as required
by TS 3.2.2
(paragraph
7.a).
~
The Supply System did not fulfillits commitment,
made in support of
Technical Specifications
Amendment
No. 71, to install
and use the
stability monitor when required
by Technical Specifications.
The
monitor had
been ineffective since its initial installation due to a
0.3
Hz filter installed
on its inputs which had not been appro-
priately considered
(paragraph 7.b).
Pe sons Contacted
A. L. Oxsen, Acting Managing Director
- J. V. Parrish,
Assistant
Managing Director, Operations
- J.
W. Baker,
WNP-2 Plant Manager
- C. H. Powers, Director, Engineering
J.
C. Gearhart,
Director, guality Assurance
- G. C. Sorensen,
Regulatory
Program Hanager
D. L. Larkin, Engineering Services
Manager
- D. L. Whitcomb, Nuclear Engineering
Manager
. L. T. Harrold, Assistant Plant Manager
W. D. Shaeffer,
Operations
Manager
- R. L. Webring, Technical
Hanager
- L. L. Grumme,
Manager,
Nuclear Safety Assurance
S. L. Washington,
Manager,
Nuclear Safety Engineering
J.
E. Rhoads,
Manager,
Operations
Event Analysis and Resolution
- D. K. Atkinson, Reactor
Systems
Supervisor
R. H. Torres, Principal Engineer,
Reactor Engineering
- J. J. Huth, Engineer,
Operations
Event Analysis
and Resolution
- H. P. Reis,
Compliance Engineer
onneville Power Admi istration
A. J.
Rapacz,
Project Representative
~D. L. Williams, Nuclear Engineer
.The inspectors
also interviewed various control
room operators; shift
supervisors
and shift managers;
and maintenance,
engineering,
quality
assurance,
and management
personnel.
- Attended the Exit Meeting on October 21,
1992.
Back round
On March 9,
1988,
La Salle Unit 2 experienced
core power oscillations
due
to a dual recirculation
pump trip.
This event resulted in increased
concern
on the part of the Boiling Water Reactor
Owners
Group
(BWROG) and
the
NRC regarding core stability.
- On June
15,
1988 the
NRC issued
Bulletin 88-07 to advise licensees
regarding possible
core instability
and to require licensee
action to provide training and procedure
changes
to preclude
and mitigate the consequences
of a core oscillation event.
On December 30,
1988,
NRC Bulletin (NRCB) 88-07,
Supplement
1, (Sl) was
issued to provide additional information concerning
power oscillations in
BWRs and to request that licensees
take action to ensure that the safety
limit for the reactor's
minimum critical power ratio
(MCPR) w'ould not be
violated.
One of the actions
requested
by NRCB 88-07 Sl was that "For
propose'd
new fuel designs,
the stability boundaries
should
be reevaluated
and justified based
on any applicable operating experience,
calculated
changes
in core decay ratio using
NRC approved methodology,
and/or core
decay ratio measurements."
On March 3,
1989 the Supply System responded
to
NRCB 88-07 Sl.
This
response
stated that
a proposed
Technical Specification
(TS) change
would
be submitted to further define the stability exclusion boundaries
and to
"Adopt a stronger stability monitoring capability,
namely use of the
ANNA
Stability Monitoring System."
On March 31,
1989; the Supply System
submitted
a TS amendment
request
(in support of cycle 5),
which stated
that "For future cycles in which decay ratio values of 0.9 or greater
are
calculated for points below the
100X rodline, the allowable region will
be modified accordingly."
In early 1991,
a European
BWR experienced
unexpected
regional oscilla-
tions when the operators
thought they were not.in the region of potential
instability as defined
on their power to flow map.
The
BWROG used the
data obtained
from this and other industry events to develop additional
strategies
for prevention of core'instabilities.
On March 18,
1992, the
BWROG issued
a letter whose subject
was "Implementation Guidance for
Stability Interim Corrective Actions" which amplified guidance
contained
in NRCB 88-07 SI.
The
BWROG provided implementation
guidance for
application to procedures
and training programs related to
NRCB 88-07 Sl
to minimize the potential for encountering oscillations.
Attachment
1 to
the letter stated that under
some conditions it might be possible to
violate the
MCPR safety limit during
a regional oscillation while the
APRH signal is oscillating at less than
10X peak-to-peak of scale.
It
further stated that
a regional oscillation of this magnitude
could be
easily recognized
on the local power range monitors
(LPRMs), but might
not always result in LPRH high or downscale
alarms.
Such
a regional
oscillation would also
be apparent
on the average
power range monitors
(APRMs) if they are monitored carefully.
In recognition of the exclusion region uncertainties,
the
BWROG letter
recommended that operator training and retraining programs
encourage
caution when operating
near the region (within 'Approximately 5X in power
and in flow rod line ), minimizing the amount of time spent operating
near the stability exclusion region.
Attachment
1 of the
BWROG letter
stated that:
"BWR stability is influenced
by several
power distribution and
operating state variables that change during normal operation
and
from cycle to cycle.
In particular, the radial
and axial power
distributions
and core inlet subcooling
have
a strong impact on
stability.
Advances in fuel and reload designs
have generally
resulted
in higher fuel bundle
power levels (higher radial peaking
on the average)
over the last
10 years,
which makes
an oscillation
somewhat
more likely today than previously.
The higher bundle power
levels,
which increase
the neutronic efficiency of the reactor,
have
lessened
the stability differences
between
high and lower power
density reactors.
This is because
the lower power density reactors
generally
have more margin to nuclear limits and can therefor e
increase
peaking
more than the higher power density reactors."
The March 18,
1992
BWROG guidance
was received
by the Supply System,
but
there
was
no documented
record of receipt
and processing of the letter.
Training of operators
and shift technical
advisors
(STAs) regarding the
contents of the
BWROG guidance
was performed in May 1992.
0
The WNP-2 cycle 8 core,
a mixed core design of 8x8 and 9x9-9x Siemens
Nuclear
Power
(SNP) Corporation fuel, was loaded during the 7th refueling
outage.
On July 4,
1992, the
WNP-2 cycle 8 core attained initial"
criticality and proceeded
to 100X power.
A subsequent
valve leak inside
the drywell necessitated
a power reduction to
SX power to repair the
valve 'on August 13,
1992.
After the valve was backseated
and the leakage
eliminated,
power ascension
of the reactor
was resumed
on August 14,
1992.
With the reactor at 36.4X power and total core flow at 30.5X, the
reactor operator
prepared to shift the recirculation
pumps to fast speed
by closing the "A" recirculation loop flow control valve.
Power
decreased
to 33.5X and total core flow decreased
to 26.0X,
as expected.
However, reactor
power oscillations started
as the recirculation flow
control valve
(FCV) was closing, at about 2:58 a.m.
on August 15,
1992.
The oscillations were observed
by the operators
and terminated
by a
manual reactor
The
NRC dispatched
an Augmented Inspection
Team
(AIT) to WNP-2 to assess
licensee
actions
and identify equipment
failures,
human performance errors,
procedure deficiencies,
and quality
assurance
deficiencies that
may have contributed to the event.
Inspection report (IR) 50-397/92-30
documented
the results of the AIT
inspection.
The AIT found no evidence of fuel failure or violation of
fuel safety limits due to the power oscillation.
However, the AIT
identified numerous
issues
associated
with the event.
This special
inspection
was performed to follow up on the issues
identified by the AIT in IR 50-397/92-30
and to determine
the licensee's
. compliance with regulatory requirements
relevant to those issues.
A
review of IR 50-397/92-30 identified the principal issues
which are
discussed
below.
These
issues
were discussed
with licensee
personnel,
and associated
procedures
and records
were reviewed.
Control of Core Power Distribution
The AIT report stated that the August
15 instability event
was caused
by
the control rod pattern selected.
This included all shaper
rods with-
drawn,
and four primary control rods in the core center region withdrawn
to step
28, resulting in a region of high power density in the center of
the core.
This resulted in a decay ratio
(DR) of 1.05,
whereas
a more
conservative
rod pattern could .have resulted in a
DR =as low as 0.3.
A review of plant procedures
was performed to determine
procedure
adequacy
and procedure
compliance leading
up to the August 15,
1992
oscillation event.
Plant Procedures
Manual
(PPH) Procedure
9.3.9,
Revision 8, Control
Rod Withdrawal Sequence
Development
and Control,
and
PPH 9.3.12,
Revision 6, Plant
Power Haneuvering,
were the principal
procedures
that provided the requirements for control rod pattern.
As noted in the AIT report,
PPH 9.3.12 provided several
cautions which,
if followed, would have reduced
the probability of core oscillations.
These cautions
included:
(1)
a recommendation
to flatten radial peaking,
(2) maintaining
a strong bottom peak (while noting that too large
a
bottom peak could prevent the opening of flow control valves after the
recirculation
pumps were shifted to fast speed),
(3) not being overly
aggressive
when pulling shaper rods, with a peaking 'factor of 3.4 as
an
optimum value (considering recirculation
pump vibration and fuel
preconditioning limits), and (4) minimizing the time spent with the
pumps
at slow speed
when establishing
control rod pattern following a power
reduction
(due to concerns for xenon burnup after the power increase).
However,
because
the procedures
did not provide sufficient direction or
criteria for their use,
licensee
personnel
directing the establishment
of
the control rod pattern
on August 15,
1992 did not implement these
cautions;-
Due to the control rod pattern selected,
the radial
and axial
power distribution at the time of pump shift were very skewed
(1.92
radial peaking factor and up to 1.76 axial peaking factor)
and all shaper
rods had been withdrawn at the time of the event.
10 CFR 50, Appendix B, Criterion
V requires that procedures
include
appropriate quantitative or qualitative acceptance
criteria for deter-
mining that important activities have
been satisfactorily, accomplished.
PPN 9.3.12
and 9.3.9 provided qualitative guidance without quantitative
limits.
Inadequate
procedure
guidance or acceptance
criteria was evident
in several
instances,
including the following:
~
Plant Procedures
Manual
(PPM) procedure 9.3.12,
Revision 6, Plant
Power Maneuvering,
paragraph
7. 1.4, Performing
a Rod Set,
subpara-
graph b, Principles
and Objectives,
required that the radial peaking
be flattened in selecting
a control rod pattern for the final rod
pattern but did not provide quantitative limits for the peaking
factors.
~
PPM 9.3.12,
Revision 6, paragraph 7.1.5.c,
Power Distribution Con-
straints,
required that care
be taken to prevent excessive
peaking,
but did not provide quantitative limits for "excessive"
peaking.
~
PPM 9.3.12, Revision 6, paragraph 7.1.5.c,
stated that establishing
too large
a bottea peak will prevent opening of the flow control
valves to 20,000
and 24,000
GPM respectively following the shift to
60 Hz, due to preconditioning limits, but provided no quantitative
limits for the peaking factors.
~
- PPM 9.3. 12, Revision 6, paragraph
7.1.5.d,
stated that care should
be taken to prevent the occurrence of excessive
peaking but did not
provide quantitative limits for the peaking factor s.
PPM 9.3. 12, Revision 6, paragraphs
7. 1.5.d
and 7. 1.5.e,
also
required that the operator not be "overly aggressive"
when pulling
shaper
rods
and that the time spent using slow speed recirculation
pumps
be minimized when performing
a rod set following a reduction
in power or prior to the second
ramp in a startup.
However, the
procedure
did not provide quantitative
acceptance criteria for the
rate of pulling shaper
rods or for the time spent using slow speed
recirculation
pumps for the noted conditions.
PPM 9.3. 12, paragraph
7.1.5.e,
stated that the more time spent at
low power, the more fierce the xenon burn will be following power
increase.
However, the procedure did not provide qua'ntitative
limits for the time spent at low power,
or quantify the "fierce
xenon burn."
0-
0
-
The lack of specific requirements for reactor
power distribution parame-
ters,
as discussed
above,
allowed the STA/Station Nuclear
Engineer
(SNE)
to implement
an unstable control rod withdrawal pattern that generated
the very high peaking factors.
PPH 9.3.9
and
PPH 9.3.12 did not provide
appropriate qualitative or quantitative acceptance
criteria for power
distribution parameters.
This is an apparent violation of 10 CFR 50,
Appendix B, Criterion
V (397/92-37-01).
PPH 9.3.12 allowed deviations
from the rod withdrawal
sequence after
reactor
power exceeded
20X; the SNE's freedom to make such deviations
was
considered
necessary
by the licensee to account for various power and
'core considerations.
Deviations from the approved
sequence
above
20X
power were
a common practice, with rod pulls based
on calculations
by the
STA/SNE.
This practice resulted in large variations in core stability
during several
startups,
due to excessively
peaked
power distributions
resulting from the various rod patterns
selected
by the STA/SNE.
There
were
no administrative controls requiring peer
review of the SNE's
changes
to prescribed control rod order sheets.
PPH 1.3. 1 specified that
the responsibility of the Control
Room Supervisor
(CRS)
was to ensure,
among other things,
a conservative
approach to operations
involving core
reactivity changes.
For the August 15,
1992 event, there
was
no operator
review of STA/SNE changes
to the prescribed
control rod withdrawal
sequence.
Prior to the event,
the control rod sequence
used
by the crew
caused
very high neutron flux peaking
and did not conform to guidance
provided by the licensee's
procedures.
The lack of (1) instructions or
procedures
to control deviations
from the approved
rod withdrawal se-
quence
above
20X power,
and (2),requirements
for peer review of the SNE's
changes to prescribed control rod order sheets
(to ensure
a conservative
approach to operations
involving core reactivity changes)
is an apparent
violation of 10 CFR 50, Appendix B, Criterion V, and was identified as
another
example of the previously mentioned
procedure violation.
4.,
Core Desi
The AIT report stated that the core design
had contributed to the
observed
power oscillations,
in that the
WNP-2 Cycle 8 core was
a mixed
core design with unbalanced
flow characteristics.
This resulted in,the
relatively low-power, low resistance
8x8 bundles starving reactor coolant
flow from the high-power
and high resistance
9x9-9x bundles.
For the
operating conditions that existed during the 8/15 power oscillation
event,
the
DR would have
been
20X lower for a full 8x8 core
and
10X lower
for a full 9x9-9x core.
10 CFR 50, Appendix A, General
Design Criterion 12, states
that "The
reactor core
and associated
coolant, control,
and protection
systems
shall
be designed to assure that power oscillations which can result in
conditions exceeding
specified acceptable
fuel design limits are not
possible or can
be reliably and readily detected
and suppressed."
a.
Susce tibilit of BWRs to Power Oscillations
Which Can Result
'n
Conditions
Exceedin
S ecified Acce table
Fuel
Desi
n Lim'ts
As previously noted in paragraph
2 of this report, the
NRC issued
.
NRCB 88-07 because
of concerns
regarding core stability and power
oscillations in BMRs.
NRCB 88-07 Sl recognized that power oscil-
lations that can result in violations of the Hinimum Critic'al
Power
Ratio safety limit were possible in BWRs.
A June
1989
GE report
(NEDO-31708) concerning fuel thermal
margin during core thermal
hydraulic oscillations in a
BWR, stated that the possibility of
violating the safety limit HCPR does exist for sufficiently large
oscillation magnitudes.
The Harch 18,
1992
BWROG letter, Attach-
ment 1,
Paragraph
1.b, stated:
"... under
some conditions it may be possible to violate
the
HCPR Safety Limit during
a regional oscillation while
the
APRH signal is oscillating at less
than
10X peak-to-
peak of scale.
A regional oscillation of this magnitude
would be easily recognized
on the
LPRHs, but may not
.
always result in LPRH High or Downscale
alarms.
Such
a
regional oscillation would also
be apparent
on the
APRHs
if they are monitored carefully."
As a result of the generic concern regarding
power oscillation that
could result in conditions that violate the
HCPR safety limit, the
BWROG provided "Interim Recommendations
for Stability Actions" and
NRCB 88-07 Sl requested
BWR licensees
to implement the
recommendations.
In addition,
NRCB 88-07 Sl stated that for new
fuel designs,
the stability boundaries
should
be reevaluated
and
justified based
on calculated
changes
in core decay ratio using
NRC
approved
methodology,
decay ratio measurements,
or applicable
operating experience.
Furthermore,
the March 18,
1992
BWROG letter
cautioned that radial
and axial power distributions
and core inlet
subcooling
had
a strong
impact
on core stability. It further
cautioned that fuel and reload designs
had resulted in higher fuel
bundle power levels,
making oscillations more likely.
The above
paragraphs
establish that
BWRs can experience
power oscillations
that can violate IKPR safety limits.
They further establish that
new core designs,
such
as the
WNP-2 cycle 8 Bx8 and 9x9-9x mixed
fuel loading required further review for core stability and further
consideration of precautions
that may be necessary
to provide for
possible instability.
The AIT report stated that "AIT LAPUR calculations
also indicated
that the out-of-phase
mode of instability did not have
much margin
to instability. If the
LAPUR calculations
had
been performed before
the event,
they would have
shown that the instability could have
been either in-phase or out-of-phase with almost equal probability."
The WNP-2 Cycle 8 core experienced
in-phase
power oscillations
on
August 15,
1992 while power
was between
36.4X and 33.5X,
and core
flow was between
30.5X and 26.0X.
This was below the region of
, concern defined in TS 3.2.6
(Region A) and below the stability
monitoring area
above the 80X rodline and less than
45X total core
flow, as defined in the bases for TS 3.2.7.
R
Based
on the above paragraphs,
the inspectors
concluded that the
MNP-2 Cycle 8 core could have experienced
in-phase or out-of-phase
power oscillations.
Although no safety limits were violated on
August 15,
1992, it is possible that power oscillations could result
in conditions which exceed specified acceptable
fuel design limits.
Consequently, it was important that adequate
procedures
and training
exist to enable the licensee's staff to be able to reliably and
readily detect
and suppress
power oscillations,
both in-phase
and
out-of-phase.
Since
such oscillations could occur below the 80 per-
cent rodline and outside the
TS defined stability exclusion zone, it
was important that the procedures
and training be adequate
to allow
them to readily detect
and suppress
power oscillations in this area.
ead
and Reliable Detect
on
a
d Su
ression of BWR Power
Oscillations
Recognizing that power oscillations were possible in BWRs,
GE, the
BWROG and the
NRC performed
numerous
studies
and reviews
and issued
various advisories
and recommendations
to provide for ready
and-
reliable detection
and suppression
of power oscillations.
A
November
1988
BWROG letter provided interim recommendations
for
stability actions that defined power/flow operating regions for core
stability and
recommended
actions for entry into the stability
exclusion region.
NRCB 8807 Sl requested
BWR licensees
to implement
those
recommendations,
and emphasized
the need for unambiguous
pro-
cedures
and training for implementation of those recommendations.
In March 1992, the
BWROG provided further guidance
on the implemen-
tation of the stability interim corrective actions
based
on lessons
learned
.from recent operating experience.
The March 18,
1992
letter provided several
recoaeendations
for prevention,
detection
and suppression
of power oscillations.
The letter emphasized
the
need for procedures
and training to implement the recommendations
for ready detection
and suppression
of power oscillations.
As noted previously, the
WNP-2 cycle 8 core had
an approximately
equal probability of in-phase or out-of-phase
power oscillations for
the August 15,
1992 operating conditions.
GE provided further
information regarding the difficulties of recognizing out of phase
power oscillations in NEDO 31708
as follows:
"During a regional oscillation, since
one half of the core
is oscillating
180 degrees
out of phase with the other
half, there is significant cancellation
in the
APRM signal
which is a simple average of approximately
15-20 local
power range monitors
(LPRH) distributed throughout the
core.
Because
the magnitude of the local neutron fluxes
is not completely represented
in the
APRM signal,
the
oscillation magnitudes
may be larger compared to a core-
wide oscillation before operator action or automatic
mitigation occurs.
Although previous studies
have
demonstrated. that very large magnitude neutron flux oscil-
lations are required to challenge
the safety limit HCPR
(Reference
1 and 2), the relationship
between local
oscillation magnitudes
and
LPRH and
APRM signals
has not
been established."
As noted
above,
APRH signals
can not reliably detect regional
(or
out-of-phase)
core oscillations.
Reliance
on
LPRH upscale
and
downscale
alarms,
and the procedures
and training relevant to
recognition of core oscillation based
upon those alarms,
would not
have provided ready
and reliable detection
and suppression
of out-
of-phase oscillations.
response
procedure
4.603:48,
Revision 2, for windows 2-5
and 5-6, specified operator actions in response
to LPRH upscale
and
downscale
alarms.
The annunciator
response
procedures
did not
mention the possibility of core oscillations,
and did not require
a
manual reactor trip for core oscillations,
but did require that the
STA be informed of those alarms.
Plant procedures
therefore did not
appear to provide adequate
instructions for ready
and reliable
detection of out of phase
power oscillations.
As previously noted,
the
WNP-2 August 15,
1992 power oscillation
event occurred with power/flow operating conditions that were below
the 80 percent rodline.
PPH 4;12.4.7,
Revision 6, Unintentional
Entry into Region of Potential
Core Power Instabilities, did require
a manual scram for periodic
LPRH upscale or downscale
alarms that
appear
on the full core display, or any
LPRH having oscillations
greater
than
20 watts/cm'.
However, the procedure did not apply
below the 80X rodline and with core flow greater
than 45X.
Plant
procedures
therefore did not appear to provide adequate
instructions
for ready
and reliable detection of out of phase
power oscillations
with operating conditions below the 80 percent rodline.
Further-
more, plant procedures
did not appear to provide adequate
instruc-
tions for suppression
of power oscillations that could occur below
the 80 percent rodline since
PPH 4.12.4.7 Revision
6 did not apply
when operating conditions were below the 80 percent rodline.
Ade uac
of
icensee
Review o
WNP-
C cles
7 and 8 Co e
esi
ns
The March 18,
1992
BWROG guidance discussed
the uncertainties
associated
with the definition of the stability "exclusion region"
and emphasized
the need for closer monitoring of the
LPRH signals
while operating in areas
near ("approximately
5X in power and in
flow control line") the exclusion region.
The letter also provided
guidance for suppression
of core power oscillations
upon
recognition.
WNP-2 licensed operators
and
SNE/STAs
had been trained
on the
BWROG guidance,
but the training was ineffective (see
paragraph
5 of this IR for additional details).
The August 15,
1992
WNP-2 core power oscillation was readily detected,
and was
effectively suppressed
when the control
room operator
(CRO) manually
scrammed
the plant.
However,
a number of procedure
and training
inadequacies
existed,
as discussed
above
and in other sections of
this report, that minimized the reliability of detection
and
suppression
of regional or out-of-phase oscillations in the
WNP-2
Cycle 8 core.
The design
and operation of the WNP-2.Cycle 8 core
appear to have violated
10 CFR 50, Appendix A, General
Design
Criterion
(GDC)
12 (50-397/92-37-02).
The AIT report further stated that the Cycle
7 WNP-2 calculated
stability decay ratios were not provided to
NRC in the cycle
7
licensing submittal.
The reduced stability margin (decay ratio
greater
than 0.75) which was calculated
by SNP for Cycles
7 and
8
0
resulted
in operation with reduced
thermal
margin during power
oscillations
(the critical power r'atio
(CPR) operating limit was
exceeded
during the 8/15/92 event).
The licensing bases
assumed
that
a design basis
accident
would not be initiated from a condition
which exceeded
the
CPR operating limit.
The Supply System
and
SNP
accepted
the reduced stability margin for Cycles
7 and 8 without
adequately
reviewing the possibility of a design basis
accident
(DBA) being initiated during a period of reduced stability margins
and potentially reduced
CPR.
The licensee
performed
a 10 CFR 50.59 evaluation for both the Cycle
7 and Cycle 8 core reloads.
The evaluations
determined that the
core reload. for Cycles
7 and 8 did not: (1) increase
the probability
of occurrence of an accident evaluated previously in the licensing
basis
document,
(2) increase
the consequences
of an accident
previously evaluated,
(3) increase
the probability of occurrence of
malfunction of equipment,
or the consequences
of the malfunction of
that equipment,
(4) create the possibility of a different type of
accident,
or different type of malfunction of equipment,
or (5)
reduce the margin of safety
as defined in the basis for any TS.
The licensee's
basis for the
10 CFR 50.59 evaluations
was the Core
Operating Limits Report
(COLR), the supporting
analyses
and reports
performed
by the licensee
and
SNP,
and the licensee's
review of
those reports.
The Reload Analysis Reports
(RARs) for both cycles
were supporting reports for the
The
RARs evaluated
core
hydrodynamic stability, and determined that the decay ratios
(DRs)
for both reloads
were within acceptable limits.
The
DRs were
determined for three
power and flow data points within the TS
defined exclusion region
and were determined for the end-of-cycle
(EOC)
and all rods out conditions.
The DRs were determined
using
the
NRC approved computer
code called
COTRAN.
The August 15,
1992
WNP.-2 Cycle 8 core oscillation demonstrated
that
oscillations could occur outside the exclusion region
and that the
RAR evaluation
was not sufficiently thorough nor conservative.
The
AIT report noted that staff analyses
and
SNP hydraulic stability
evaluations after the August 15,
1992 event indicated the 9x9-9x
fuel design to be less stable than other fuels in US BWRs.
The AIT
. report stated that for the operating conditions at the time of the
August
15 oscillations,
the AIT calculated
a core-wide
DR of 1.05,
a
hot channel
DR of 0.83,
and
an out-of-phase
DR of 1.0.
The
RAR, the
COLR and the
10 CFR 50.59 evaluations for the Cycles
7 and
8 reloads
were therefore not conservative.
The
10 CFR 50.59 evaluations for
the Cycle
7 and 8 reloads
appeared
to be inadequate
in that they did
not sufficiently consider the changes
in thermal hydraulic charac-
teristics of the reload cores
and did not appropriately evaluate
their effect on or possible
changes
to the various conditions
analyzed
pursuant to 10 CFR 50.59,
'as noted at the beginning of this
paragraph.
'This was identified as
a potential violation of 10 CFR 50,-. Appendix B, Criterion III (50-397/92-37-03).
As noted
above,
the NRC-approved
computer
code for WNP-2 stability
evaluations
is
COTRAN.
As also noted
above,
the
COTRAN code
may not
0
-10-
necessarily
provide conservative results,
and the methodology for
assessing
core stability by evaluating three points within the TS-
defined exclusion boundary
may not be conservative.
The AIT report
noted that;SNP
was considering submitting for review arid approval
its core stability analysis
computer
code ca'lied STAIF.
The
. licensee
indicated that other computer
codes we'e also available
and
were being considered.
Pending further licensee
and
NRC review of
the methodology
and computer code used for core stability evalua-
tions, this was identified as
an Unresolved
Item (50-397/92-37-04).
+gn~i
As noted previously,
r'eady
and reliable detection
and suppression
of
power oscillations
depended
heavily on good procedures
and training.
As
'reviously discussed,
procedures for the detection
and suppression
of
power oscillations were not adequate.
Training for detection
and
'uppression
of power oscillations
was also insufficient.
The AIT report noted that
a licensee Training Department
lesson
plan for
STA training, Reactivity Mismanagement
Events, that included information
from the March 18,1992
BWROG letter,
was issued
on May 27,
1992.
How-
ever, the training attendance
record for the
STA on shift during the
event indicated that he had attended
the training for the subject lesson
plan on May 7, 1992, before it was approved
on May 27,
1992.
The AIT
report stated that all operators
on shift during the event
had attended
requalification training that included training on instability events.
However,
none of the licensed operators
who were on shift during the
event
and
who were subsequently
interviewed by the team could recall
having attended training that addressed
the March 18,
1992
BWROG letter
.nor any of the specific guidance from the letter.
The AIT report further
stated that operators
interviewed considered that the area of core
instability was well defined
and that there
was
no possibility of the
reactor
becoming unstable
outside the TS-defined exclusion area.
As noted in previous sections of this report, operating
procedures
did
not incorporate the guidance
contained in the March 18,
1992
BWROG let-
ter.
The training'rovided to the
STAs and licensed operators
included
the guidance
contained in the
BWROG- letter, but was not consistent with
applicable operating procedures.
The inspectors
noted that
some aspects
of the training could have
been followed without violating existing
procedures
(e.g.,
using extra caution
when operating
near the stability
exclusion region by turning on the stability monitor prior to the time
required
by the procedures).
However, the training was inadequate
because
other aspects
could not and should not have
been followed without
procedural
changes
(e.g.,
power/flow conditions for changing recirc
pump
speed).
As noted in previous sections of this report, the rod pattern
and operating conditions at the time of the event differed from the
guidance
provided by the
BWROG letter.
WNP-2 TS 6.4. 1 states
that
a retraining
and replacement
training program
shall
be maintained for the plant staff and shall include familiarization
with relevant industry operational
experience.
Training on the March 18,
1992
BWROG letter was performed.
However, the training appeared
to be
ineffective and inadequate
in that it was not based
on existing operating
0
~
~
4
-11-
procedures
and none of the
on shift operators
interviewed could recall
the training having been provided.
The inadequate
and ineffective
training provided
on the industry operational
experience
and guidance
communicated
by the March 18,
1992
BWROG letter is addressed
as part of
the apparent violation of 10 CFR 50, Appendix A,
GDC 12 (paragraphs
4.a
and 4.b).
andlin
of Generic
Commu 'tio s
The AIT report
and the previous section of this report noted that the
March 18,
1992
BWROG letter provided significant guidance
and industry
advice that most likely would have precluded the occurrence of the WNP-2
oscillation event
had the guidance
been followed.
-The licensee
was
on
distribution for the March 18,
1992
BWROG letter as follows:
~
The letter was addressed
to the
BWROG Executive Committee.
The
WPPSS Deputy Managing Director is a aember of the
BWROG Executive
Committee.
~
The letter was addressed
to the
B'WROG Primary Representatives.
A
WPPSS Licensing Department
engineer is the
BWROG Primary
Representative.
~
Members of the Stability Committee
and Reactivity Controls Committee
were sent copies of the letter.
Six WPPSS representatives,
from the
licensing, technical staff and Nuclear Engineering organizations,
were members of those
committees.-
~
The fuel vendor,
SNP,
was sent
a copy of the letter .
As noted in the AIT report, the fuel vendor,
SNP,
acknowledged receipt of
the letter but had no documented
record of when it was received.
Discus-
sions with management
personnel of the
SNP fuel design group indicated
that they were not aware of the letter.
Further discussion with the fuel
vendor indicated that,
based
upon
a review of the letter after the event,
the fuel vendor
had no additional actions to be performed
as
a result of
'he
letter.
The adequacy of the fuel vendor's
handling of the letter
will be referred to the
NRC Vendor Inspection
Branch for further
evaluation.
The Supply System received the
BWROG letter, but no documented
record
existed to determine
when the letter was received.
The only available
evidence of any action taken
by the licensee
regarding the letter was the
May 27,
1992
STA training lesson
plan and the operator requalification
program that included tt aining regarding information contained
in the
letter.
As previously noted,
operating
procedures
were not changed,
the
training performed
was ineffective,
and. no documented
action was taken
by
the Nuclear
Engineering
Department relevant to specific provisions of the
letter.
WNP-2 TS 6.2.3. 1 states
that the Nuclear Safety Assurance
Group
(NSAG)
shall function -to examine industry advisories
and shall
make. detailed
recommendations
for revised procedures,
operations activities, or other
means of improving unit safety to the Director of Licensing
and
-12-
Assurance.
PPM 1. 10.4, Revision 9, External Operating
Experience
Review,
provides the requirements
for the
WNP-2 External Operating
Experience
.
Review Program.
This procedure listed various sources of operational
experience
data.
However, the procedure
was inadequate
in that it did
not include
a requirement for the review and processing of BWROG
information.
The TS 6.2.3. 1 requirements
did not appear to have
been
complied with in that
NSAG had no documented
record of receipt of the
March 18,
1992
BWROG letter, despite its multiple distribution to Supply
System personnel;
no review was performed;
and
no recommendations
relevant to it were provided.
This was identified as
an apparent
violation of TS 6.2.3. 1 (50-397/92-37-05).
7.
~Dt
a ~
d 'tment of
R
l w
iased
Set o'nts
Technical Specification 3.2.2 states that "The APRH flow biased
simulated thermal
power-upscale
scram trip setpoint
(S)
and flow
biased
neutron flux-upscale control rod block trip setpoint
(Srb)
shall
be established
according to the following relationships:
Tri
Set oint
S < (0.66W + 51X)T
Srb < (0.66W + 42X)T
llowable Value
S
< (0.66W. + 54X)T
Srb < (0.66W + 45X)T
where:
S and Srb are in percent of RATED THERMAL POWER.
W
Loop recirculation flow as
a percentage
of loop
recirculation flow which produces
a rated core flow
of 108.5 million lbs/hr.
T
Lowest value of the ratio of FRACTION OF RATED
THERMAL POWER divided by the
MAXIMUM FRACTION OF
LIMITING POWER DENSITY.
T is always less than or
equal to 1."
This TS was applicable
when reactor
power was greater than or equal
to 25X.
No exception for TS 3.0.4 or 4.0.4 was in effect during the
period of the event.
The action statement for TS 3.2.2 states that "With the
APRH flow
biased
simulated thermal
power-upscale
scram trip setpoint and/or
the flow biased neutron flux-upscale control rod block trip setpoint
less conservative
than the value
shown in the Allowable Value column
for S or Srb,
as
above determined, initiate corrective action within
15 minutes
and adjust
S and/or Srb to be consistent with the Trip
Setpoint value within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce thermal
power to less than
25X in four hours."
As'n alternative to adjusting the trip settings,
TS 3.2.2 allowed
the licensee to increase
the gain of the
APRHs so that the
would read artificially high by a factor of T (or T-factor).
- 13-
The basis for this
TS states:
"The scram settings
and rod block
settings
are adjusted in accordance
with the formula in this
specification
when the combination of THERNAL POWER and
NFLPD
indicates
a higher peaked
power distribution to ensure that
an
[linear'eat generation rate] transient
would not 'be increased
in
the degraded condition."
During a review of data provided by the licensee,
the AIT noted that
the thermal limits printout of 1:53 a.m.
on August 15 indicated that
T was 0.793.
However, the licensee did not perform the flow biased
setpoint adjustments
or APRN gain adjustments
as described
above,
prior to exceeding
25 percent
power and within the prescribed
time
limit.
In addition, the licensee
subsequently
provided five other
thermal limits*printouts, obtained
between
12:53
and 3:00 a.m.
on
August 15,
1992, which also indicated that the adjusted flow biased
setpoints
should
have
been in effect.
The lowest value calculated
,for the T- factor was 0.746, for the printout obtained just prior to
closure of the
A loop recirculation flow control valve.
Technical Specification 4.0.4 states that entry into an operational
condition or other specified applicable condition shall not be made
unless
the surveillance
requirements
associated
with the Limiting
Condition for Operation
have
been performed within the applicable
surveillance interval, or as otherwise specified.
Review of licensee
records indicated that
PPN 7.4.2. 1,
"Power
Distribution Limits" had not been performed
on August
14 or August
15
(PPN 7.4.2. 1 was the licensee's
implementing procedure for
determining margin to the thermal limits, and for adjustment of the
flow biased setpoints
as discussed
above).
The STA stated. that
had always
been interpreted
by the Supply System staff to
allow 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving
25% power for performance of PPN
7.4.2.1.
The STA also stated that although the computer printouts
indicated that the
APRN flow biased setpoints
required adjustment,
this adjustment
could also
be delayed for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The AIT calculated that'the proper scram setpoint
should
have
been
43.6X power,
and the rod block/APRN Upscale
alarm setpoint should
have
been 36.9X.
However, although the maximum peaks of the power
oscillations
(48.75 percent
power maximum on two of the APRNs)
appeared
to exceed the scram setpoints that should
have
been in
effect,
a scram would not have occurred.
The instrumentation for
the flow biased
scram inserts
a six-second
time delay to simulate
the thermal
time constant of the fuel,
and therefore
does not
provide scram protection for a power oscillation event.
However,
the flow biased
rod blocks
and
APRN Upscale
alarms
do not have
a
time delay.
These trips apparently
would have
been received earlier
in the event if the setpoints
had
been lowered, providing the
operators
with earlier information to diagnose
the onset of power
oscillations.
In addition, the operators
may not have increased
reactor
power to 36.4 percent,
due to the close proximity to the rod
block trip point.
The AIT concluded that the licensee did not appear to have complied
with Technical Specification 3.2.2.,
but had not bypassed
a required
The AIT also concluded that the flow biased
rod blocks
and
APRH Upscale
alarms would have
been received
eat lier in the event
had the setpoints
been lowered, providing th'e operators with earlier
information to diagnose
the onset of power oscillations.
The
licensee's
failure to lower the
APRN flow biased
rod block,
and
APRH Upscale alarm setpoints
before exceeding
25X power is an
apparent violation of TS 4.0.4
and 3.2.2 (50-397/92-37-06).
The AIT report noted that Surveillance
Procedure
7.4.2.7.3 did not
account for the 0.3
Hz filter installed in the
LPRM inputs to the
stability monitor nor for the
SNP evaluation
and subsequent
change
to the Advanced Neutron Noise Analysis
(ANNA) monitor alarm that was
necessitated
by the filter.
The 0.3 Hz filter caused
a significant
error of the decay ratio displayed
by the
ANNA stability monitor and
would not have been fully effective for TS required stability
monitoring had NNP-2 been previously in the
TS defined region of
instability.
On March 31,
1989, the Supply System requested
an
amendment to TS
3/4.2.7, Stability Honitoring, to redefine the
TS Region
C stability
monitoring zone
on the basis of the installation of ANNA.
The
Supply System stated that
ANNA provided
a stronger, faster,
more
sophisticated
means of monitoring the stability of the core.
Based
on the coamitment to install
and use
ANNA in accordance
with the TS,
the
NRC approved
TS amendment
no.
71
on June
23,
1989.
ANNA was
installed
and tested at
WNP-2 in March of 1989 and modified and
tested
in February of 1990.
As noted in the AIT report, the testing
of ANNA and the procedures for use of ANNA did not recognize the 0.3
Hz filters installed between
ANNA and the
LPRN inputs to ANNA.
The
0.3 Hz filters rendered
ANNA ineffective.
This appears
to be
a
Deviation from a commitment to install
and
use
ANNA in accordance
with the TS, in support of a TS amendment
request.
This was
identified as
a potential deviation (50-397/92-37-07).
8.
~C
This special
inspection
reviewed the issues identified by the AIT, along
with the applicable licensee
procedures
and records.
The issues
were
discussed
with the licensee.
Previous sections of this report identified
five apparent violations, which were also communicated to the licensee.
Numerous
bar riers existed that, if more effective, could have precluded
the WP-2 oscillation event.
Full utilization of industry experience
and
advice,
adequate, operating procedures,
adequate training,
and adequate
design reviews were barriers that could have prevented
the core oscilla-
tions
on August 15,
1992.
In conjunction with the identified violations,
however,
these barriers to preclude the occurrence
of the event failed.
The control
room operators
mitigated the consequences
of the event
and
terminated it by scramming the plant.
However, this special
inspection
found that management
oversight in relation to the barriers
noted
above
-
0
0
-15-
was conspicuously lacking.
No management
oversight regarding review and
implementation of the
BWROG March 18,
1992 letter was evident.. It
appeared
that licensee
management
was not adequately sensitive to core
stability issues
at WNP-2.
This lack of management sensitivity to core
stability and insufficient management
oversight to preclude the occur-
rence of the event appeared
to be the principal underlying cause of the
barrier failures that led to the oscillation event.
9. ~Ud tt
Unresolved
items are matters
about which more information is required to
determine whether they are acceptable
items, violations, or deviations.
An unresolved
item addressed
during this inspection is discussed
in
paragraph
4.c of this report.
The inspectors
met with licensee
management
representatives
periodically
during the report period to discuss
inspection status,
and
an exit
meeting
was conducted with the indicated personnel
(refer to paragraph
1)
on October 21,
1992.
The scope of the inspection
and the
inspectors'indings,
as noted in this report,
were discussed
with and acknowledged
by the licensee representatives.
The licensee
did not identify as proprietary any of the information
reviewed by or discussed
with the inspectors
during the inspection.
The licensee
stated that
a response
to the AIT report was being prepared,
and that it would address
the concerns
raised
by the AIT report
and the
-potential violations identified during this special
inspection.
The
licensee stated that their response
would be submitted
by October
30,
1992.
A summary of the licensee's
preliminary position on the apparent
violations is as follows:
a.
Inadequate
procedures
(PPM 9.3.9
and 9.3. 12):
the licensee
agreed
that the procedures
were inadequate.
Failure to follow procedures
for rod pattern selection:
the licensee
acknowledged that the
procedures
were inadequate,
but because
the procedures
were
inadequate,
the licensee did not believe that the procedures
were
not followed.
b.
GDC 12, core design violation: it was the
licensee's
opinion that out-of-phase oscillations would not occur
before in-phase oscillations were experienced.
Furthermore,
the
licensee
was of the understanding
that as long as
NRCB 88-07 was
complied with, the generic concerns
regarding core oscillation and
compliance with GDC 12 were addressed.
The licensee
also acknow-
ledged the weaknesses
of the training provided,
but considered that
although the training did not preclude the event, the operators
adequately mitigated it by scramming the plant.
C.
Inadequate
design evaluations for Cycles
7 and
8 core reload design:
the licensee
acknowledged that the stability calculations
performed
were not conservative.
However, the licensee
considered that they
-
-16-
had used the methods for which they were licensed,
and that those
were the best tools available at the time.
d.
Failure of NSAG to formally review and provide recommendations
regarding the
BWROG March 18,
1992 letter:
the licensee
acknow-
ledged that
NSAG did not review the letter but considered that
a
definition of industry advisories
included in the
TS requirement
was
necessary.
e.
Failure to adjust
APRH flow biased
and rod block setpoints
as
required
by TS:
the licensee
agreed with the apparent violation.