HomeMy WebLinkAbout841140.tiff .O.t 4 wrc� i UNITED STATES 217/-27/2
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ARLINGTON. TEXAS 76011
MAY 11 119644 1- ,,,„
Docket: 50-267
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Mr. 0. R. Lee, Vice President
Electric Production °p�°.
Public Service Company of Colorado
P.O. Box 840
Denver, Colorado 80201
Dear Mr. Lee:
Our consultants at the Los Alamos National Laboratory (LANL) have been
evaluating sections of one of the fuel elements which was discovered to show
crack indications following the second refueling of the Fort St. Vrain (FSV)
facility. Based on their evaluation of this fuel element, their evaluation of
the information provided during our April 4, 1984, meeting (as documented in
your April 6, 1984, letter) and the information contained in other relevant
documents, LANL concurs with your positions that, under normal operating
conditions, it is unlikely that extensive thermal stress cracking of fuel
elements will occur and that, if cracking does occur, it will be hairline in
nature. There is a concern, however, that the effects of an existing thermal
stress loading: (1) in conjunction with mechanical , static loading, and
(2) the effects of dynamic loading, on a cracked fuel element have not been
sufficiently evaluated. A copy of the LANL evaluation report is enclosed.
We have reviewed the LANL report in light of the information you have provided
and agree with the LANL findings. While we find it acceptable for the FSV
station to return to power operation, the following information must be
provided within 90 days of your receipt of this letter: (1) an evaluation
of the failure modes of a cracked fuel element under dynamic loadings and
thermal stress loadings, and (2) the description of the fuel element
inspection program which will be incorporated into the Technical Specifications.
A letter confirming the schedule for providing the evaluations and your
proposed Technical Specification changes related to fuel element inspections
is requested within 10 days of your receipt of this letter.
If you have any questions on this subject, please contact your project manager.
841140
kilA ONO
AL nj1 f <6 1� 1/1( 1%
Mr. 0. R. Lee -2-
Since this information request relates solely to FSV, fewer than ten respondents
are involved; therefore, OMB clearance is not required under P.L. 96-511.
Sincerely,
E. H. Johnson, Chief
Reactor Project Branch 1
Enclosure: LANL Report
cc: Attached List
Fort St. Vrain
cc list
C. K. Millen Chairman, Board of County Commissioners
Senior Vice President of Weld County, Colorado
Public Service Company Greeley, Colorado 80631
of Colorado
P. 0. Box 840 Regional Representative
Denver, Colorado 80201 Radiation Programs
Environmental Protection Agency
James B. Graham, Manager 1860 Lincoln Street
Licensing and Regulation Denver, Colorado 80203
East Coast Office
General Atomic Company Don Warembourg
2021 K Street, NW, Suite 709 Nuclear Production Manager
Washington, DC 20006 Public Service Company of Colorado
P. 0. Box 368
J. K. Fuller, Vice President Platteville, Colorado 80651
Public Service Company
of Colorado Albert J. Hazle, Director
P. 0. Box 840 Radiation Control Division
Denver, Colorado 80201 Department of Health
4210 East 11th Avenue
Denver, Colorado 80220
G. L. Plumlee
NRC Senior Resident Inspector Kelly, Stansfield & O'Donnell
P. 0. Box 640 Public Service Company Building
Platteville, Colorado 80651 Room 900
550 15th Street
Denver, Colorado 80202
Darrell G. Eisenhut, Director
Division of Licensing
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555
Fort St. Vrain Fuel Elements
NRC Fin No. A-7258
April 30, 1984
Charles A. Anderson
Deborah R. Bennett
Los Alamos National Laboratory
Responsible NRC Individual and Division
J. Miller/0RB3
Prepared for the
U.S. Nuclear Regulatory Commission
Washington, D. C. 20555
DISCLAIMER
This report was prepared as an account of work sponsored
by an agency of the United States Government. Neither the
United States Government nor any agency thereof, or any of
their employees, makes any warranty, expressed or implied,
or assumes any legal liability or responsibility for any
third party's use, of any information, apparatus, product
or process disclosed in this report or represents that its
use by such third party would not infringe privately owned
rights.
ABSTRACT
This report reviews the material submitted to the NRC by the Public
Service Co. of Colorado, on the evaluation of two cracked Segment 2 graphite
fuel elements, and on the examinations of Segment 3 graphite fuel elements
removed during the third refueling of the Fort St. Yrain HTGR. The report
also provides comments to the NRC as to whether the licensee 's technical
information is correct and consistent with the information gained by Los
Alamos in their review of graphite slabs from Segment 2 cracked fuel element
SN 1 -2415.
FORWARD
This technical evaluation report is part of the technical assistance
program, "Review of Selected Fort St. Yrain Issues", FIN No. A-7258, and is
supplied to the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor
Regulation, by Los Alamos National Laboratory.
Technical Evaluation Letter Report
on
Fort St. Vrain Fuel Elements
by
Charles A. Anderson
Deborah R. Bennett
Los Alamos National Laboratory
NRC Fin No. A-7258
April 20, 1984
On April 4, 1984, the Public Service Company of Colorado, the NRC and its
technical consultant, Los Alamos National Laboratory, attended a meeting at
NRC, Bethesda, Maryland, to discuss the results of
1 . the evaluation of two cracked Segment 2 (second refueling) graphite
fuel elements and
2. the examinations of Segment 3 graphite fuel elements removed during
the third refueling.
The Fort St. Vrain submittal of April 6, 1984, formally provided the
evaluation material discussed at this meeting.
The objective of this evaluation is to review the licensee 's letter and
related material , and to provide comments to the NRC as to whether the
licensee 's technical information is correct and consistent with the
information gained by Los Alamos in their review of graphite slabs from
Segment 2 cracked fuel element SN 1-2415.
The Los Alamos review of the licensee's letter and related material is as
follows:
1 . Probable Cause
Los Alamos agrees with the statement made by the Public Service Co. of
Colorado (PSCo) that the likely cause of cracking in the two elements is
thermal stress, induced mainly by cold helium by-pass flow on the B-face of
the two elements, and perhaps exacerbated by skewed power and/or flux
distributions within the region of interest. Los Alamos has recently
completed a finite element failure analysis on a model representing a
localized area of a typical graphite fuel element, as shown in Fig. 1 . The
thermal calculation was done using the thermal and mechanical properties for
H-327 graphite (Ref. 1 ), with consideration to irradiation-induced shrinkage
effects. Fuel heating rates, coolant temperatures, and surface heat transfer
coefficients were used as specified or derived from the Fort St. Vrain FSAR
and relevant documents (Ref. 2 and Ref. 3). A by-pass flow temperature of
760°F was employed in the calculations, as were two surface heat transfer
coefficients- on the B-face, reflecting uncertainty in gap thickness between
adjacent column elements. From the temperature distribution in the model , the
thermal stress distribution was calculated. Finally, using a Weibull -
statistical strength model (with Weibull modulus of 8 from GA-A13955 and a
mean tensile H-327 graphite strength of 1200 psi from the Fort St. Vrain FSAR)
we determined the failure probability and failure region in the model . That
failure probability is shown in the first two lines of Table I for the two
surface heat transfer coefficients covered in the study. For the higher heat
transfer coefficient there is a significant probability (about 10-2) of
failure in the graphite web between the coolant hole and B-face. Figure 2
illustrates the tensile stress distribution and Fig. 3 shows the temperature
distribution in the graphite web between the coolant hole and B-face.
Los Alamos also calculated the probability that, once the outer web
cracked, the crack would propagate further into the element. This was
accomplished by releasing the stress on the crack and repeating the
calculation. These results are shown as the second two lines of Table I.
When the crack occurs in the web between B-face and coolant hole (Case B), the
probability of further propagation is significantly reduced (Case D). For
interior generated cracks the reverse appears to be true. Thus, Los Alamos
agrees with the PSCo assertion that a thermal stress crack, generated in the
web between B-face and coolant hole, would be limited in extent and would not
easily propagate through the element.
TABLE I
FAILURE PROBABILITIES AND CRACK INITIATION POINTS
Surface
By-Pass Flow Heat Transfer Failure Failure
Case Temperature Coefficient Probability Region
A 760°F 1 .0 W/in2-F 0.9 x 10-4 interior web
g 760°F 4.5 0.11 x 10-1 B-face web
C 760°F 1 .0 0.15 x 10-3 interior web
760°F 4.5 0.27 x 10-3 interior web
2. Cracked Fuel Element Integrity
As mentioned, Los Alamos agrees with the PSCo claim that extensive thermal
stress cracking within a given fuel element is highly improbable. However,
there is insufficient information to conclude that the existence of supposed
• thermal stress cracks, as found in the two Segment 2 fuel elements, would have
no effect on the failure mode of a fuel element under all plausible mechanical
loading conditions. The tests performed on unirradiated graphite slabs at
General Atomic involved uniform static loadings on the sides, which induced
overall compression in the element. The results of these static tests
indicated that the existence of web cracking should not alter the overall
failure mode of the slab.
Los Alamos feels that these tests do not account for the presence of a
strong, thermal stress field in the specimen, nor do they account for the
possibility that the crack could reduce the strength of the element under
dynamic loading conditions. For example, during a seismic event the Fort St.
Vrain core (as currently constrained by the core restraint devices) will
transmit dynamic loads primarily through the dowel pins and socket arrangement
located on the ends of the fuel elements. This dynamic load transfer will
produce a complex stress field in the interior of the element, and could
subsequently cause cracks to further propagate, depending on the magnitude of
loads being transmitted. Therefore, Los Alamos recommends that the effects of
the initial thermal stress field and the dynamic loading through the dowel pin
arrangement be factored into the evaluation of overall element structural
integrity.
3. Adequacy of the Inspection/Surveillance Program
The surveillance/inspection program proposed by the Public Service Co.
includes a minimum scope of:
i ) Photographing all six faces of 175 of 250 fuel/reflector elements
removed during the Segment 3 reload, using the Fuel Handling Machine
35mm camera.
ii ) Evaluating all photographs for indications of significant structural
abnormalities prior to returning to power operation.
iii ) Using the Fuel Handling Machine Cask Video Monitor, carefully examine
the two Segment 3 fuel elements with operational histories believed
to be most similar to those of the two Segment 2 cracked fuel elements
iv) Perform a Non-Destructive Post Irradiation Examination, similiar to
the PIE performed on Segment 1 and Segment 2 elements, using 50 to 60
Segment 3 Fuel and Reflector elements.
The first item implies that some 175 Segment 3 fuel and reflector elements
have been or-will be photographed using the Fuel Handling Machine camera. To
date, PSCo has photographed a large percentage of the Segment 3 elements, with
special attention to elements from Region 18, which are said to have a compar-
able operational history to the Segment 2 cracked fuel elements. Los Alamos
recommends that all six sides of all fuel and reflector elements removed in
Segment 3 and future reloads be photographed.
The second item requires an evaluation of the photos with regard to "sig-
nificant" structural abnormalities, prior to returning to power operation.
The available set of Segment 3 photographs leads to two conclusions: (i ) The
photographs taken with the Fuel Handling Machine 35mm camera are of sufficient
quality to identify cracks of the same order or bigger than the cracks found
on the vertical sides of the Segment 2 cracked fuel elements, and (ii ) of the
Segment 3 elements adequately photographed, there is no visual indication of
cracking, although there are numerous water marks and scratches on the element
surfaces.
The third item in the PSCo surveillance/inspection program uses the Fuel
Handling Machine Cask Video Monitor in examining the two Segment 3 fuel ele-
ments that are considered most comparable to the Segment 2 cracked fuel ele-
ments in operational history. We have concluded that this monitor can produce
an image comparable to the image with the 35mm camera, and has a resolution
that is sufficient to identify cracking on a given element surface.
We agree with the fourth item in the surveillance program which intends to
examine, in terms of extensive Non-Destructive Post Irradiation Examinations,
all of the elements in Region 18 plus some precharacterized elements from
Regions 3, 13, 22 and 29, and based on the premise that Region 18 elements are
"comparable" to Region 8 elements. Los Alamos recommends obtaining
justification from PSCo, providing clarification on the comparability of the
Region 18 elements to the Region 8 cracked fuel elements, with regard to
operational history and resulting temperature and stress fields.
In conclusion, we consider the PSCo Segment 3 inspection/surveillance pro-
gram to be adequate, assuming all six sides of the 175 elements are eventually
photographed and evaluated as to surface abnormalities. However, we also recom-
mend that in future reloads, all six sides of all fuel and reflector elements
removed be, photographed and evaluated. The current percentage of elements
receiving extensive nondestructive post irradiation examination is considered
adequate.
4. Summary
(i ) Los Alamos concurs with the licensee's arguments concerning the
likely cause of the cracking of the Segment 2 fuel elements.
(ii ) Los Alamos concurs with the licensee's arguments that thermal and
irradiation-induced stress cracking is relatively localized as
observed in the Segment 2 cracked elements and that the cracks will
be hairline in nature and not likely to affect the integrity of
fuel pellet rods.
(iii ) Los Alamos considers the PSCo Segment 3 inspection/surveillance
program to be adequate to the needs of identifying further cracked
elements.
(iv) Los Alamos concurs with the licensee's argument that extensive
thermal stress cracking in a given fuel element is unlikely under
normal operating conditions, but disagrees with the licensee's
justification of element structural integrity based only on static
load tests. Los Alamos considers the issue of dynamic loading, as
transmitted primarily through the dowel -sockets in a column of ele-
ments during a seismic event, to be relevant in evaluating the
overall structural integrity of a cracked element. Therefore, Los
Alamos recommends further review of the failure mode of a cracked
fuel element under dynamic and thermal stress loadings, based on
existing GAT documentation, or further analyses by PSCo.
5. References
a. "H-327 Graphite Design Data Manual ", GA Technologies,
GA-906933, June 1983.
b. Goodman, Jovanovic, Ganley and Covert, "The Thermodynamic and
Transport Properties of Helium", General Atomic Project 2102,
GA-A13400, October, 1975.
c. "Heat Transfer and Fluid Flow in Nuclear Systems", edited by
Henri Fenech, Pergammon Press, 1982.
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siE p° A'c,°ty UNITED STATES
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NUCLEAR REGULATORY COMMISSION
Itoc nnl ≥ REGION IV
. LG::
t,` 611 RYAN PLAZA DRIVE, SUITE 1000
a° ARLINGTON, TEXAS 76011
{*914
MAY 1 1 1984
Docket: 50-267
P'ff,l rl1T r `'1MIMS
Mr. 0. R. Lee, Vice President „ MAY 1 719
Electric Production c 84
Public Service Company of Colorado
P.O. Box 840 canttLEr coca
Denver, Colorado 80201
Dear Mr. Lee:
Our consultants at the Lawrence Livermore National Laboratory (LLNL) have
completed their review of the information you have provided related to the
proposed neutron detector decalibration circuitry for the Fort St. Vrain (FSV)
facility. The results of the review are contained in the enclosed Technical
Evaluation Report (TER) . We have reviewed the TER and agree with the findings
contained therein. We, therefore, find it acceptable for the neutron detector
circuitry to be modified to include the floating trip setpoint circuit
described in your referenced submittals provided appropriate changes are
incorporated into the Technical Specifications (TS) to ensure proper
operability, testing, and calibration requirements. These requirements, as
indicated in Section 4 of the TER, are in addition to your May 16, 1983
(P-83177) commitment to provide TS to ensure that the proper analytical
techniques are employed.
If you have any questions on this subject, please contact your project manager.
Sincerely,
E. H. Johnson, Chief
�teactor Project Branch 1
Enclosure: As stated
cc: Attached list
A. 40? I) cC f/
Fort St. Vrain
cc list
C. K. Millen Chairman, Board of County Commissioners
Senior Vice President of Weld County, Colorado
Public Service Company Greeley, Colorado 80631
of Colorado
P. 0. Box 840 Regional Representative •
Denver, Colorado 80201 Radiation Programs
Environmental Protection Agency
James B. Graham, Manager 1860 Lincoln Street
Licensing and Regulation Denver, Colorado 80203
East Coast Office
General Atomic Company Don Warembourg
2021 K Street, NW, Suite 709 Nuclear Production Manager
Washington, DC 20006 Public Service Company of Colorado
P. 0. Box 368
J. K. Fuller, Vice President Platteville, Colorado 80651
Public Service Company
of Colorado Albert J. Hazle, Director
P. O. Box 840 Radiation Control Division
Denver, Colorado 80201 Department of Health
4210 East 11th Avenue
Denver, Colorado 80220
G. L. Plumlee
NRC Senior Resident Inspector Kelly, Stansfield & O'Donnell
P. 0. Box 640 Public Service Company Building
Platteville, Colorado 80651 Room 900
550 15th Street
Denver, Colorado 80202
Darrell G. Eisenhut, Director
Division of Licensing
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission
Washington, D. C. 20555
•
•
Ucin 20038
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TECHNICAL EVALUATION REPORT ON ' UE
NEUTRON DETECTOR DECALIBRATION AT :THE
FORT ST. VRA1N NUCLEAR„GENERATI C LION
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ABSTRACT
This report documents the technical evaluation on the decalibration
of the neutron detectors at the Fort St. Vrain Nuclear Generating Station. The
evaluation is to determine that the added circuitry for generating a floating
trip setpoint as a function of indicated power meets NRC design criteria and
has no adverse effects on the plant protection system.
The evaluation finds that the floating trip setpoint circuitry meets
the design criteria specified in the plant 's FSAR and will produce a reactor
trip (as a function of indicated power) before the true power limit value in
the Technical Specifications is exceeded.
FOREWORD
This report is supplied as part of the Selected Operating Reactor
Issues Program II being conducted for the U. S. Nuclear Regulatory Commission,
Office of Nuclear Reactor Regulation, Division of Licensing, by Lawrence
Livermore National Laboratory.
The U. S. Nuclear Regulatory Commission funded the work under the
authorization entitled "Selected Operating Reactor Issues Program II, "
B&R 20 19 10 11 1, FIN No. A-0250.
—i—
TABLE OF CONTENTS
Page
1. INTRODUCTION 1
2. DESIGN REVIEW CRITERIA 2
3. DESIGN DESCRIPTION 3
4. EVALUATION 9
4.1 Circuitry Design Changes 9
4.2 Calibration Requirements 11
4.3 Technical Specifications and Test Provisions . . . . 12
5. CONCLUSION 12
REFERENCES 13
TABLE OF ILLUSTRATIONS
Figure 1. PPS floating trip circuit 4
Figure 2. Heat balance calibration 5
Figure 3. PPS floating trip point circuitry for Fort St. Vrain Unit 1 6
Figure 4. Nuclear channel test setup, block diagram 7
Figure 5. PPS channel configuration 8
Figure 6. Dual Linear Channel Drawer 10
-iii-
TECHNICAL EVALUATION REPORT ON THE
NEUTRON DETECTOR DECALIBRATION AT THE
FORT ST. VRAIN NUCLEAR GENERATING STATION
(Docket No. 50-267)
James C. Selan
Lawrence Livermore National Laboratory, Nevada
1. INTRODUCTION
The excore neutron detectors at the Fort St. Vrain (FSV) Nuclear
Generating Station are located in the prestressed concrete reactor vessel adjacent
to the core. The instrumentation inputs from these neutron flux detectors are
used in reactor control and the plant's protection system (PPS).
The excore neutron detectors, 12 in number, are located in 6 wells
at 60° intervals around the core cavity. The function of these twelve detectors
(power range) are as follows:
(1) Six detectors are used in the PPS. The signals from these six
are combined into three channels by two 180° opposing detectors.
The range of these detectors is from 1.5% to 150% of full power.
These three channels provide a trip signal in a 2-of-3 channel
logic at 140% of full power.
(2) Six detectors are used in reactor control. They also have a
range from 1.5% to 150% of full power. The signals from these
detectors (flux controller) are used to regulate the position
of the control rod pair and runback rods to control the power
level in the core. The flux recorder, flux integrator (megawatt-
hour meter), and power/flow module also receive input from the
flux controller.
The neutron flux level as measured by the power range detectors is
effected by the motion of the control rods. This motion can alter the radial
core power distribution so that the flux levels measured are not directly
proportional to the true core thermal power level thus indicating "decalibration"
of the detectors.
Analyses have shown that motion of the rod banks near the center of
the core causes the detectors to underpredict true power changes while rod
banks near the outside cause overprediction of true power changes [Refs. 1 and 2].
The effects of this decalibration and resulting over/under predictions of true
power could cause spurious trip signals or cause design limits to be exceeded
before a protective trip occurs.
-1-
3. DESIGN DESCRIPTION
The -floating trip setpoint circuit (FTSC) design is shown in Figures
1 through 5. These figures are taken from the General Atomic Company Report
[Ref. 1].
The basic function of the FTSC is to produce a floating trip setpoint
that will vary at a constant offset above indicated power which will always
produce a trip signal before the reactor reaches the 140% true power value
specified in the Technical Specifications. Figure 1 shows the basic electronic
components which make up the floating trip setpoint circuitry (indicated by the
dashed lines). The theory of operation can best be described using Figure 3.
An indicated power signal derived from the power range detectors (0-150%) is
fed to a differentiator, sample and hold (S/H) circuit, and to the two bistable
trips (reactor and rod withdrawal prohibit).
The differentiator outputs a signal (volts) which is proportional to
the rate of change of the indicated power. This output is fed to a bistable
trip where it is compared to a pre-selected rate of power change. If the com-
parator goes low, the S/H circuit will continue to sample the indicated power
input. Should the comparator go high, the S/H will hold its last input signal
before the S/H goes high and feeds this value to a summer. At the summer, the
input value from the S/H is added to the pre-selected trip offset value. If the
sum exceeds an adjustable high setpoint (100-140%), the output holds at the
high limit. If the sum is less than the adjustable low setpoint (60-140%)
than the output holds at the lower limit. The output value then goes to the
reactor trip bistable and to the RWP summer. At the reactor trip bistable
(programmable) if the indicated reactor power is greater than the trip setpoint
(from the summer), a reactor trip signal is produced.
At the RWP summer, the high or low setpoint output is added with
the pre-selected RWP offset setpoint. The output is then fed to the RWP
programmable bistable. If the comparator goes high, an RWP trip is produced.
In addition to the FTSC, a circuit is added for heat balance cor-
rection as shown in Figure 2. The indicated true power as calculated from
the heat balance equations in the data logger will be indicated on an added
meter. The circuit will correct for non-PPS readouts, megawatt-hour meter,
power and flow measurements, and the flux-recorder. It should be noted that
this circuit receives an input from the data logger and is not connected with
the FTSC nor part of the PPS. This circuit was not reviewed since it is inde-
pendent of the change for the floating trip setpoint circuit used in the PPS.
Figure 4 shows the test setup to calibrate each nuclear channel.
A nuclear channel consists of a detector and one half of a dual linear drawer
as shown in Figure 5. Figure 5 shows one channel of the three PPS channels
where a coincident logic of two-out-of-three is required to produce a trip.
—3—
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—7—
4. EVALUATION
This_-section presents an evaluation on various aspects of the pro-
posed neutron detector decalibration.
4.1 Circuitry Design Changes
The Fort St. Vrain plant protective system design is based on the
1967 edition of the NRC General Design Criteria [Ref. 5] and IEEE 279-1968
[Ref. 6] as stated in the plant 's Final Safety Analysis Report.
The addition of the FTSC to the PPS involves replacing the entire
dual linear power range channel drawer and modules with the drawer shown in
Figure 6. All the required modules (not shown) for the FTSC are located
within the drawer. There are no external modifications to the drawer/modules
required.
A review of drawings, schematics, and drawer specifications sub-
mitted (e.g. dual linear power channel schematic and assembly, floating trip
setpoint schematic, bistable trip circuits, period rate circuits, linear
amplifier schematic, and operation manual) [Ref. 3] finds that the FTSC does
not alter any of the original system design criteria. This includes the
criteria of redundancy, overall logic, failure modes, field wiring, arrange-
ment, independence, testability, reliability, or physical separation of the
reactor trip circuits [Ref. 4].
A failure modes and effects analysis (FMEA) was performed by General
Atomic Company [Ref. 1]. The FMEA analyzed each component of the FTSC with
respect to the "Failure Mode, " "Channel Effect, " and "System Effects. " The
results of their analysis demonstrated that no single failure of any portion
of the FTSC will prevent the PPS from initiating or completing a reactor
trip or rod withdrawal prohibit. The addition of the FTSC does not change
the failure modes of the original system. The change does add one additional
failure mode. This is the failure of a channel to detect a high rate of flux
increase. This failure mode will not prevent PPS action since the original
system was not rate dependent and the upper trip limit (existing system) is
still active to activate a trip.
A review of the FMEA finds that most of the component failures
(e.g. shorts, opens, or high output) either causes no detection of high
flux rates or causes the trip point not to float (may either go high, low,
or zero) at the channel level. These failure modes at the system level
result either in a spurious 1-channel RWP and trip or in loss of channel RWP
detection with a trip still detectable with the remaining 2 PPS channels.
Therefore, component failures within the FTSC will neither prohibit nor
adversely affect the PPS from initiating its protective function since the
trip limits of the original system are still effective.
-9-
A review of the dual linear drawer schematic finds that the circuit
boards and switches are interlocked to provide automatic channel tripping if
their performing function is prohibited. The interlock path is accomplished
by the +15 Vdc supply where interruption causes the bistable trips and trip
relays to de-energize to their tripped condition. For loss of bus voltage
or sensor, channel trip is also automatically initiated.
4.2 Calibration Requirements
General Atomic Company recommends the following calibration require-
ments [Refs. 1 and 2] :
(1) At least one calibration is required during every 24-hr period
when operating in low power or power modes.
(2) To prevent or clear RWPs which occur due to inaccurate detector
readings, a calibration should be done whenever any channel
approaches or reaches an RWP setpoint.
(3) To ensure that the interlock sequence switch (ISS) is switched
at the proper power level, the following requirements are made:
a. With the ISS in the startup mode, a calibration is required
when heat—balance power is between 2% and 4% of rated power.
The methods to determine heat—balance power level are given
in Technical Specification Surveillance Procedure 5.4.1.1.4
c—D.
b. When increasing power with the ISS in the low power mode,
a calibration is required when heat—balance power is
between about 24% and about 28% of rated power.
c. When decreasing power with the ISS in the low power mode,
a calibration is required when heat—balance power drops
below about 35% of rated power.
(4) Whenever the operator has reason to believe that one or more
detectors are giving anomalous readings, a calibration should
be performed.
(5) Whenever individual detectors differ by more than 10%, the proper
functioning should be verified.
(6) Add the following items to the existing FSV calibration procedure,
S.R.5.4.1.1.4c-D:
Control rod bank partially inserted
Position (inches withdrawn)
Regular rod position (inches withdrawn)
(7) Calibrate detectors prior to the withdrawal of rod group 3C
(for cycle 2).
-11-
REFERENCES
(1) Public Service Company of Colorado (J. K. Fuller) to the NRC (W. P. Gammill),
dated January 11, 1979.
(2) Public Service Company of Colorado (J. K. Fuller) to the NRC (S. A. Varga),
dated November 29, 1979.
(3) Public Service Company of Colorado (H. L. Brey) to the NRC (J. T. Collins),
dated May 16, 1983.
(4) Public Service Company of Colorado (H. L. Brey) to the NRC (E. H. Johnson),
dated November 30, 1983.
(5) General Design Criterion 7, 12, 15, 19-26, and 39, "AEC General Design Criteria
for Nuclear Power Plant Construction Permits, " 1967 edition.
(6) IEEE Standard: "Criteria for Protection Systems for Nuclear Power Gene—
rating Stations, " IEEE 279-1968.
-13-
DISCLAIMER
ter':
This document was prepared as an account of work sponsored by an agency of the United Slates Government.
Neither the United States Government nor the University of •aiifomia oar any of their emplo.ees,makes
any warranty, express or implied, or assumes any legal hat+il ty or respmsisant'for die accuracy, com-
pleteness, or usefulness of any information,apparatus. product,or prof s disciased,rr represents that its
use would not infringe privatelt owned rights.Reference herein toany'specific comsaareial products,process,
or service by trade name,trademark,manufacturer,or otherwise.does not nrcessarily tunstit lit c or imply its
endorsement,recommentlatioo,or favoring by the United States Government or the University of California.
The views and opinions of authors expressed herein do not necessarily mate or reflect those of the united
�.i
States Government thereof,and shall not be used for advertising et product tuilorneme.n purposes. -- -
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