ML20140E804
| ML20140E804 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 03/19/1986 |
| From: | Johnson E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Walker R PUBLIC SERVICE CO. OF COLORADO |
| References | |
| NUDOCS 8603280202 | |
| Download: ML20140E804 (2) | |
See also: IR 05000267/1985017
Text
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MAR 191986
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In Reply Refer To:
Docket: 50-267/85-17
Public Service Company of Colorado
ATTN:
R. F. Walker, President
P. O. Box 840
Denver, Colorado 80201-0840
Gentlemen:
This provides additional information concerning our letter of October 31,
1985, which acknowledged your response (your letter serial P-85367 dated
October 16,1985) to our Notice of Violation'and inspection report dated
September-16, 1985.
Background.
In your letter of October 16, 1985, you stated that you did not
believe that single circulator trips should be considered reportable under the
provisions of 10 CFR 50.72 and 10 CFR 50.73.
You additionally stated that you
~
believed RWP system actuations were not reportable. We forwarded your stated
positions to the Office of Nuclear Reactor Regulation (NRR).
Interpretation. We have recently received a response from NRR.
Based 'on this
response, we are pleased to inform you that your contention that single-
circulator trips and RWP system actuations are not considered to be
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reportable has been sustained. A single circulator trip might, however, lead
to a reportable condition, for example inadequate flow for the power level.
Action. As.the result of this interpretation, ' iolation A for " Failure to
V
Report" in our Notice of Violation dated Septei, ar 16, 1985, is herewith
withdrawn and considered for record purposes no; to have been a violation.
Sincerely,.
On : n 1 sy
Ra.m E. M
E. H. Johnson, Director
' Division of Reactor Safety
and Projects
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CC.
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J. W. Gahm, Manager, Nuclear
)
Production Division
Fort St. Vrain Nuclear Station
60319
16805 WCR 191
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Platteville, Colorado 80651
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OPublic Service-
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company of colorado
16805 Weld County Road 19 1/2, Platteville, CO 80651
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OCT 2 i E385
October 16, 1985
Fort St. Vrain
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Unit No. 1
._._...___. _-
P-85367
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Regional Administrator
Region IV
U.S. Nuclear Regulatory Commission
611 Ryan Plaza Drive, Suite 1000
Arlington, Texas 76011
Attn:
Mr. E. H. Johnson
Docket No. 50-267
SUBJECT:
I&E Inspection Report 85-17
REFERENCE:
NRC Letter, Johnson to Lee,
Dated 09/16/85 (G-85381)
Dear Mr. Johnson:
-
This letter is in response to the Notice of Violation (NOV) received
as a result of an inspection conducted at Fort St. Vrain during the
period of June 17,
1985,
through August 16, 1985.
The following
response to the items contained in the Notice of Violation is hereby
submitted:
1.
Failure to Report
"Immediate Notification Requirements of Operating
Nuclear Power Reactors, in paragraph (b)
non-emergency events,
(2) Four-hour reports, requires, in part, "... the licensee shall
notify the NRC as soon as practical and in all cases within four
hours of the occurrence of any of the following: ... (ii) Any
event or condition that results in manual or automatic actuation
of any Engineered Safety Feature (ESF), including the Reactor
Protection System (RPS)."
" Licensee Event Report System",
also requires in
part, a 30 day written report of Reactor Protection System
actuations.
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The licensee's Technical Specification 2.9 states, "The plant
protective system is the reactor protective circuitry and the
circuitry oriented towards protecting various plant components
from major damage. This system includes '(1)
(2)
loop
shutdown, (3) circulator trip, and (4) rod withdraw prohibit."-
Contrary to the above,
the licensee has considered circulator
trips as not reportable.
The licensee's daily logs list 9
circulator trips in 1984 and 7 circulator trips to date in 1985.
Of these 17 circulator trips, only the trip of August .11,
1985,
was reported to the NRC.
This is a Severity Level
IV violation (Supplement I.D) (50-
267/85-17).
,
(1) The reason for the violation, if admitted:
Public Service Company of Colorado ~does not consider single
circulator trip actuations to be reportable per 10CFR50.72
or 50.73.
The reasons for this determination are'related to
the original design and definition of FSV's Plant Protective
System (PPS). The-PPS was originally designed to detect and
initiate automatic action epon the onset of abnormal
core
parameters or abnormal equipment operation parameters.
When
the reporting requirements of 10CFR50.72 and 50.73 were
'
initially proposed, the term Reactor Protection System (RPS)
was not recognized for this plant.
Through extensive FSAR
design basis / accident analysis review and review of. industry
_
practice / precedent,
it was
- determined
that
was
equivalent to the FSV Scram System.
Reactor Protection
System terminology is . identified specifically
in
the
Standard Tectnical
Specification,
Section
2.2.1, Reactor
Trip System Instrumentation Setpoints. The FSV equivalence-
of this section is LSSS 3.3, which identifies scram, loop-
shutdown, and steam water dump actuations.
FSAR accident
analyses clearly rely only on_ scram and loop shutdown / steam
water dump actuation to ultimately mitigate postulated
accidents.
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Although the PPS includes the reactor protective circuitry
(scram) and engineered safety feature circuitry (steam water
dump,
and . loop shutdown,
FSAR Criterion 14),
it also
includes numerous other equipment protection functions.
This
is clearly stated in the FSV Technical Specifications, Section 2.9 (which was cited in the Notice of Violation):
,
"The plant protective system is the reactor
protective
circuitry
and
the
circuitry
oriented
towards protecting various plant
components from major damage.
This system
includes (1)
(2) loop shutdown, (3)
circulator trip,
and
(4)
rod
withdraw
prohibit".
A single circulator trip is actuated by the equipment
protection circuitry,
as evidenced by the
basis
for
Specification LCO 4.4.1.c which states in part:
"All
circulator
shutdown
inputs (except
circulator speed high on water turbines) are
equipment protection items which are tied to
'two. loop trouble' through the
loop shutdown
system."
PSC contends that, provided it is clear that an actuation is
a result of an identified source other than RPS or ESF
actuation,
then the actuation is not reportable.
The
following discussion demonstrates that a single circulator
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trip by itself 'does not cause a RPS 'or ESF actuation, nor_ is
it caused by a RPS or ESF actuation.
Although a
single circulator trip does provide an input to
the loop shutdown logic, and therefore the reactor scram
logic (indirectly through the two loop trouble logic), a
single circulator. trip, by itself, does not actuate any RPS
or ESF system.
A similarity in the LWR case would be a
condition which resulted in.one input in a "one out of two
.
taken twice"
logic for initiation'of an RPS or ESF system.
In the LWR case, this is not reportable per ICCFR50.72 or
50 73 since the RPS or ESF has not yet been actuated.
.
Figure 1
shows- a simplified schematic for the inputs that
,
can cause a circulator trip.
Those inputs listed under
equipment malfunctions are the only ones which result in a
single circulator ' trip.
Note that each parameter
is
associated
with
an
abnormal condition for a single
circulator.
The ESF (loop shutdown) tctuation
of
a
circulator trip always causes both circulators in the loop
to trip. This latter is a reportable event.
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A
simil a'r-
situation
in
a BWR is where a reactor
recirculation- pump trips- due to
equipment
protection
circuitry, which is not reportable.
However, a trip of both
reactor coolant pumps due to actuation of the Anticipated
Transient Without Scram (ATWS) protection circuitry i s
reportable.
RPS or ESF actuation can cause a simultaneous trip of both
circulators in a given
loop as part of loop shutdown.
'
However,
there is no case where RPS or ESF actuation can
cause the trip of a single circulator-in a loop.
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The above reasoning highlights the basis for determining
that single circulator trip actuations are not reportable in
accordance
with
10CFR50.72(b)(2)(11)
or 50.73(a)(iv).
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However, trip actuations and deficiencies associated with
helium circulator operation are routinely- reviewed and
.
evaluated for reportability in accordance with other 50.72
and
50.73
criteria
for other safety considerations.
Circulator trip actuations and abnormal' operations are also
routinely reviewed and investigated for plant availability
concerns.
Although the circulators were designed to operate
independently
of
one
another,
and
the
Technical
Specifications only require that one circulator in each loop
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be operable. during power operation,
a single inoperable
circulator would significantly limit plant operation.
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PSC also contends that RWP actuations are not reportable per
10CFR50.72 or 50.73. The following discussion presents the
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basis for our determination.
As previously stated, FSAR and industry reviews originally
determined that the RPS function as identified in 10CFR50.72
and 50.73 was equivalent to the FSV Scram circuitry. All
the rod withdrawal accidents evaluated in Section 14.2 of
the
relied
exclusively
on the redundancy and
reliability of the Scram system to
initiate the required
protective
action.
RWP parameters and setpoints -were-
,
.
recognized as available, but were assumed to1 fail.
The rod
withdrawal accidents ultimately rely.on the 140% power scram
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setpoint to terminate any credible
increase
in
core
reactivity, with no fuel failure'or breach of.the primary
- coolant boundary assumed.
The combination of the reactor
scram circuitry and the loop shutdown / steam water dump
circuitry provide the necessary automatic reactor protective
- and
engineered
safety
feature actions- for all
postulated accidents.
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Although both systems were designed under the same basic
design criteria as integral parts of the overall PPS,
their
basic functions distinguish their safety significance. The
scram system actuates to deenergize the control
rod brake
power supplies, allowing for gravity driven insertion of all
37 control
rod
pairs.
The
Technical
Specification
definition of control
rod operability is consistent with
this function,
as surveillance testing only
considers
free-fall
insertion capability.
The FSAR and original
design specification also clearly portray this system as the
reactor safety system.
As stated in the original
PPS design specification, the
basis for development of the RWP system was that:
"During
certain
combinations
of
plant
operating conditions, control
rod withdrawal
.
must be prohibited, but the. conditions do not
warrant plant shutdown (scram).
The
system accomplishes this requirement."
This definition is consistent with the function of the RWP
system, which' deenergizes the power to the control rod drive
motor.
Control rod drive or withdrawal power is associated
with reactor power level
control, where brake power is
associated
with the basic safety function of reactor
shutdown. One system limits reactor operation, whereas the
other
limits
fuel
damage during postulated accident
_
conditions.
Although these systems obtain their inputs from common
detector channels, the PPS design principle for Single
Failure Criterion adequately prevents system interaction due
to postulated faults or failures.
These systems were
installed under the same quality control specifications and
are maintained through specific
detailed-
surveillance
testing.
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This discussion provides the basis for determination that
RWP actuations do not constitute ESF or RPS actuations in
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accordance
with
and 50.73.
Abnormal
'
actuations
are,
however,
routinely
evaluated
for
reportability in accordance with other criteria. Since they
also present an operational
limit,
investigation
and
corrective' action are highly desirable.
(2) The corrective steps which have been taken and the results
achieved:
Until
such time as PSC and .the NRC can resolve this
,
interpretation of 10CFR50.72 and 50.73 as
it applies to
'
single
circulator
trip.
and
rod withdrawal
prohibit
actuations, all single circulator trip and rod withdrawal
prohibit actuations not resulting from or
a. part of a
preplanned sequence during testing or reactor'-operation,
will be reported to the NRC.
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(3) Corrective steps .which will
be taken to avoid further
violations:
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Upon resolution of this
interpretation of 10CFR50.72 and
50.73, the following actions will be taken to eliminate any
ambiguity in the future:
a)
The procedure on reportable events will be revised to
clarify that single circulator trip and Rod Withdrawal
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Prohibit actuations are not reportable.
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b)
. Additional
training _ will be provided to the operating
staff on reportable events.
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c)
~The FSAR will be revised to clearly define the RPS and
its relationship to the PPS.
Reporting requirements
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will
be specified,
and equipment protection-features
will be excluded.
(4) The date when full-compliance will be achieved:
PSC believes that with regard to'the reportability of single
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circulator trip, and Rod Withdrawal Prohibit actuations
it
has always been in full
compliance with 10CFR50.72 and
50.73.
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2.
Violation of Limiting Condition for Operation (LCO)
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LC0 4.2.7 of the Technical Specifications states, in part, "The
PCRV shal.1 not be pressurized to more than.100 psia unless: ...d)
,
The Interspaces between the primary and secondary penetration
- closures are maintained at a pressure greater than primary system
pressure with purified helium gas."
Contrary to the above, during the period from July 30, 1985, when
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PCRV pressure went above 100 psia, through August 10,
1985,
no
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differential
helium pressure was maintained in PCRV penetration
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interspaces B21 and B23.
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This is a Severity Level IV violation (Supplement I.C.)
(1) The reason for the violation if admitted:
Valve PDV-11380, which regulates the differential-pressure
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between the Loop 2 steam generator penetration. interspaces
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and the cold reheat steam line, was in the closed position
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contrary to system. operability requirements.
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Operators in the Control Room had indication available to
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determine the differential- pressure between the purified-
helium header and the Prestressed Concrete Reactor Vessel
(PCRV), and they based their decision on compliance with LCO
' '
4.2.7
on
this
indication
(which
was
the correct
'
determination for all
but the Loop 2 steam generator
,
pentration interspaces).
However, they did not have Control
4
Room indication of the differential
pressure between the
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Loop 2 steam generator penetration interspaces and the PCRV
,
cold reheat steam line.
Therefore,
they should
have-
solicited this information from the Equipment Operator who
obtains the information locally.
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Equipment
Operators
log
the Loop 2 steam generator
interspace / cold reheat steam line differential pressure
(PDI-11380) every eight hours, but they did not recognize
that.a minimum differential
pressure of about- five psid
should have been expected when the PCRV.is pressurized to
greater than 100 psia. The value logged was zero psid.
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The Equipment Operator Log Sheet did not adequately address
the subject log entry as being associated with a Technical
Specification LCO.
(2) The corrective steps which have been taken and the results
achieved:
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All personnel involved received a formal reprimand.
An operator adjusted the controller to about 25% open, and
it was ' verified .that PDV-11380 was operable with the
controller in the automatic mode and that PDI-11380 was
reading over 20 psid.
Pressure
differential
controller,
PDC-11380,
has been
verified operable through functional-testing.
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(3) Corrective steps which will
be taken to avoid further
violations:
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The procedure, OPOP-IBI, for raising the PCRV pressure ove~r
100 psia will be revised to include reading 'PDI-11380 and
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assuring
that
the Loop 2 steam generator pentration
interspaces are above cold reheat pressure.
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The Turbine ' Equipment Operator log will be revised to
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designate
that-
the
steam
generator
interspaces/ cold
reheat steam differential
pressure is
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required to be in accordance with LCO 4.2.7 when PCRV-
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pressure is above 100 psia.
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The date when full compliance will be achieved:
Full compliance has been achieved, and the procedure and log
revisions identified above will.be completed by November 30,
1985.
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Should you have any questions, please contact Mr. M.H. Holmes,
(303) 571-8409.
Sincerely,
,
J.'
,
Manager, Nuclear Production
Fort St. Vrain Nuclear
Generating Station
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EQUIPMENT MALFUNCTIONS
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GENERAL 2 0F 3 LOGIC
i
Circulator
Bearing
Circulator.
Steam
Loop
Pressure
Speed
Flow-Low
Water
Seal
Water
Shutdown
,
liigh
liigh/ Low
Loss
Halfunction
Drain
Parameters
j
Malf mction
,
,
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4
,
Other
f
a
' '
Circulator In Loop
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XCR
Trip Of
Other
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l
Circulator
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__
AND
-XCR
Trip Circulator
p
Loop Isolation
And
Steam / Water Dump
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FIGURE 1 CIRCULATOR TRIP LOGIC
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