ML030650592
| ML030650592 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/09/2002 |
| From: | Ernstes M Operator Licensing and Human Performance Branch |
| To: | Keenan J Carolina Power & Light Co |
| References | |
| 50-324/03-301, 50-325/03-301 | |
| Download: ML030650592 (14) | |
See also: IR 05000325/2003301
Text
Post-examination Comments
(Green Paper)
Licensee Submitted Post-examination Comments
BRUNSWICK EXAM
50-2003-301
50-325 & 50-324
R 10
FEBRUARY 10 - 1* & 19, 2003
A Progin Energy
Compnar
FEB 2 42003
SERIAL: BSEP 03-0038
U. S. Nuclear Regulatory Commission, Region 1I
ATTN: Mr. Luis A. Reyes, Regional Administrator
Atlanta Federal Center
61 Forsyth Street, SW, Suite 23T85
Atlanta, GA 30303-8931
BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2
DOCKET NOS. 50-325 AND 50-324/LICENSE NOS. DPR-71 AND DPR-62
NRC OPERATOR LICENSE WRITTEN EXAMINATION - FACILITY COMMENTS
Dear Mr. Reyes:
On February 19 2003, NRC operator license written examinations were administered at the
Brunswick Steam Electric Plant (BSEP). In accordance with NUREG-1021, "Operator
Licensing Examination Standards for Power Reactors," Progress Energy Carolinas, Inc. has
performed a post-examination review and is submitting comments for consideration in the
grading process.
Enclosure 1 contains comments, recommendations, and references for two questions from
the written license examination.
Please refer any questions regarding this submittal to Mr. Leonard R. Belier, Supervisor
Licensing/Regulatory Programs, at (910) 457 2073.
Sincerely,
Edward T. O'Neil
Manager - Support Services
Brunswick Steam Electric Plant
Brunswjck Nuclear Prant
PO Box 10429
Southport. NC 28461
Enclosure I
Page I
NRC Operator License Written Examination - Facility Comments
On February 19 2003, NRC operator license written examinations were administered at the
Brunswick Steam Electric Plant (BSEP). In accordance with NUREG-1021, "Operator Licensing
Examination Standards for Power Reactors," Progress Energy Carolinas, Inc. has performed a
post-examination review and is submitting comments for consideration in the grading process.
The following comments, recommendations, and references are provided for two questions from
the written license examination.
1) RO Exam Question #36/SRO Exam Question #25:
Question:
During an overpressure transient, an operator has opened an SRV to control pressure.
RPV pressure is 1005 psig.
Which ONE of the following is the expected tail pipe temperature for the SRV that is
open, as indicated on ERFIS?
A. 212 degreesF
B. 300 degrees F
C. 350 degrees F
D. 545 degrees F
Answer from key: B
Reference from Questions Report: Steam Tables
Comment:
The question asks what the expected ERFIS tail pipe temperature indication is for an SRV
opened with RPV pressure at 1005 psig. The answer key indicates that "B" is the correct answer
(300 degrees). Answer "B" (300 degrees) corresponds to an isenthalpic throttling process from
1000 psig to atmospheric pressure on the Mollier diagram. At the Brunswick Plant, a temperature
element is installed in each tail pipe just downstream of the SRV which feeds a recorder, ERFIS,
and a common control room annunciator. At the temperature probe location, atmospheric
conditions would not exist with an open SRV due to the backpressures created from underwater
discharge through a "T" quencher in the tows. The facility recommends the acceptance of
Answer "C" (350 degrees) as the correct answer (i.e., not Answer "B"). Per the System
Description procedure (SD-20) which discusses temperature indications, the theoretical
temperature (indication) is 3500F for steam being throttled through a leaking SRV. This is
supported by the annunciator procedure for SRV leaking (or open) which includes SRV
temperature alarm setpoints as high as 340 degrees to indicate an open SRV.
Recommendation: Change the correct answer to "C" (350 degrees).
Enclosure I
Page 2
References:
SD-20, Automatic Depressurization System
2APP-A-03, Annunciator Procedures for Panel A-03
SD-25, Main Steam System
2) RO Exam Question #82/SRO Exam Question #77:
Question:
Unit 1 is operating at 100% RTP when a high radiation condition occurs in the Reactor
Building. The following conditions exist in the Reactor Building:
Reactor Building Supply and Exhaust Fans
tripped
Both SBGT trains
running
Reactor Building vent isolation dampers (BFIVs)
open
Reactor Building dP
zero
Which ONE of the following describes the impact on the plant due to the above
conditions?
A. An elevated release of radioactivity from the main stack could occur.
B. A release of radioactivity outside containment will NOT occur due to both SBGT
trains running.
C. A release of radioactivity outside containment will NOT occur since the Reactor
Building Supply and Exhaust fans have tripped.
D. A ground level release of radioactivity could occur.
Answer from key: D
Reference from Questions Report: SD 37.1, Rev 4, pg 35, Reactor Building HVAC
Comment:
The question asks for the plant impact from high radiation conditions in the reactor building with
a failure of ventilation isolation dampers to close and both standby gas treatment trains (SBGT)
running. The answer key provides "D" as the correct answer, (i.e., a ground level release of
radioactivity could occur).
The facility recommends acceptance of an additional answer - "A" (i.e., an elevated release of
radioactivity from the main stack could occur). This recommendation is based on both SBGT
trains running (discharging to the stack) with high radiation conditions in the reactor building.
Per the Brunswick Plant Updated FSAR, the SBGT (i.e., SGTS) system provides a means for
minimizing the release of radioactive material from containment to the environs by filtering and
exhausting the atmosphere from the reactor building during containment isolation conditions. Per
the UFSAR, an "elevated release is assured by exhausting to the plant stack." With the SBGT
Enclosure 1
Page 3
system not 100% efficient in removal of radioactivity, with both systems in operation under high
radiation conditions in the reactor building, an elevated release from the stack would result.
Therefore, both answers "A" (i.e., an elevated release of radioactivity from the main stack could
occur) and answer "D" (i.e., a ground level release of radioactivity could occur) are correct.
Reconunendation: Accept both "A" and "D" as correct answers.
References:
BSEP i and 2 Updated FSAR Section 6.5.1.1, Standby Gas Treatment System
001-37.9, Secondary Containment Control Procedure Basis Document
Enclosure 1
Page 4
Attachment to Enclosure 1
Reference Material from the Following Sources:
SD-20, Automatic Depressurization System Description
2APP-A-03, Annunciator Procedures for Panel A-03
SD-25, Main Steam System Description
BSEP 1 and 2, Updated FSAR Section 6.5.1. 1, Standby Gas Treatment System
001-37.9, Secondary Containment Control Procedure Basis Document
Mr. Luis A. Reyes
BSEP 03-0038/Page 2
GLM/glm
Enclosure:
NRC Operator License Written Examination - Facility Comments
cc (with enclosure):
U. S. Nuclear Regulatory Commission
ATTN: Document Control Desk
Washington, DC 20555-0001
U. S. Nuclear Regulatory Commission
ATTN: NRC Senior Resident Inspector
8470 River Road
Southport, NC 28461-8869
U. S. Nuclear Regulatory Commission
ATTN: Mr. Michael E. Emstes, Chief
Operator Licensing and Human Performance Branch
Sam Nunn Atlanta Federal Center
61 Forsyth Street, SW, Suite 23T85
Atlanta, GA 30303-8931
U. S. Nuclear Regulatory Commission
ATTN: Mr. George T. Hopper, Region II
Operator Licensing and Human Performance Branch
Sam Nunn Atlanta Federal Center
61 Forsyth Street, SW, Suite 23T85
Atlanta, GA 30303-8931
U. S. Nuclear Regulatory Commission
ATTN: Ms. Brenda L. Mozafari (Mail Stop OWFN 8G9) (Electronic Copy Only)
11555 Rockville Pike
Rockville, MD 20852-2738
cc (without enclosure):
Ms. Jo A. Sanford
Chair - North Carolina Utilities Commission
P.O. Box 29510
Raleigh, NC 27626-0510
Remote Shutdown Panel
SRV solenoid
3.1.4
valve red open and green closed lights are provided on the
Remote Shutdown Panel. These lights are provided for each of the
three manually operated SRVs (B, E, and G). This light indication is
from switch position only (and thus the solenoid being energized), not
actual SRV position.
Fluid Flow Detector Cabinet, CB-XU-73
The FFD Cabinet has one green closed light, one red open light, and one amber
valve open "memory" light for each SRV, all of which are actuated by
the acoustic sensor.
3.1.5
Safety/Relief Valve Leaking Indication
Each SRV has a temperature element (TE-B21-NO04A, B, C, D, E, F, G, H, J, K, L)
installed in its associated discharge piping to detect steam leakage.
The response of these temperature elements is recorded in the
Control Room on a chart recorder (B21-TR-R614) on
Panel H12-P614. A common high temperature alarm, SAFETY OR
DEPRESS VLV LEAKING, actuated by this recorder, will annunciate
in the Control Room.
In the past several years the 2B21-TE-N004C thermocouple has consistently
indicated that SRV 2B21-F013C has a tailpipe temperature that is
60-70°F higher than the average group of SRV thermocouple
indicators. EER 93-0428 concluded this is due to the 2B21-F013C
thermocouple being physically closer to the SRV main body.
There is only one (1) annunciator to alert operators of a leaking SRV. Any minor
leakage from 2B21-FO13C which could be acceptably added to this
positional effect would exceed the annunciator setpoint of 290'F.
This would disable the annunciator from being able to alert operations
of a potential leaking SRV (only). Therefore, the setpoint for the Unit
2 annunciator has been changed for 2-B21 -F01 3C from 2900F to
340oF.i
he 340°F alarm setpoint is still below the theoretical
temperature (3500F) for steam being
throttled through a leaking
SRV/
This will prevent the annunciator from alarming needlessly until a
leaking SRV exists which warrants operator action.
SD-20
Rev. 0
Page 14 of 61
3.1.3
Unit 2
APP A-03 1-1
Page I of 2
AUTO ACTIONS
NONE
CAUSE
1.
Automatic or manual initiation of ADS or safety/relief valves.
2.
Safety/relief valve failed open.
3.
Safety/relief valve(s) not properly seated or the seat is defective.
4.
High drywell temperature in vicinity of the safety/relief valve
location.
5.
Defective Temperature Recorder B21-TR-R614 or incorrect alarm
setpoint.
6.
Defective Temperature Element B21-TE-NO04A,
B,
C,
D, E, F, G, H, U,
K, or L.
7.
Circuit malfunction.
OBSERVATIONS
a.
Safety/relief valve indicating lights on RTGB Panel P601 may
indicate valve is open.
2.
Safety/relief valve indicating lights on the Fluid Flow Detector
Cabinet CB-XXU-73 may indicate valve is
o en.
3.
Safety/relief valve leak detection temperature reading greater than
2900 F (340 0 F for B21-F013C only) as read on B21-TR-614 on Control
anel H12-P614.t
4.
Safety/reliee
valve noise amplitude reading greater than 0
millivolts on the Fluid Flow Detector Cabinet CB-XU-73.
5.
Suppression pool temperature increasing (CAC-TR-4426-1,
-2).
6.
Primary containment radiation monitors indicating increased
activity.
7.
Steam flow/feed flow mismatch.
ACTIONS
1.
If
the cause of the annunciator is automatic or manual initiation of
ADS, refer to OP-20, Automatic Depressurization System.
2.
If
the cause of the annunciator is
a safety/relief valve failed
open, refer to AOP-30.0, Safety/Relief Valve Failures.
3.
If
a safety/relief valve is
suspected of not being properly seated,
cycle the affected safety/relief valve as directed by the Unit SCO
in an attempt to reseat the valve.
4.
If suppression pool temperature approaches 95 0 F, place the RHR
System in the suppression pool cooling mode per OP-17, Residual Heat
Removal System.
5.
If
a defective temperature recorder, element, or a circuit
malfunction is
suspected, ensure that a WR/JO is prepared.
Rev. 40
Page4of 00
Unit 2
APP A-03 1-1
Page 2 of 2
DEVICE/SETPOINTS
Temperature Recorder B21-TR-614(SWi)
287-2930 F
(337-343%' for
B21-FO13C only)
POSSIBLE PLANT EFFECTS
1.
Reactor shutdown.
2.
Operation of the RHR System in the suppression cooling mode to
maintain suppression pool temperature less than 95'F.
3.
Inoperable safety/relief valves may result in a Technical
Specification LCO.
4.
Increasing suppression pool temperature may result in
a Technical
Specification LCO.
REFERENCES
1.
2.
3.
4.
5.
6.
LL-9364 -
42
Technical Specification 3.4.3, 3.5.1, 3.6.2.1
OP-17, Residual Heat Removal System
OP-20, Automatic Depressurization System
AOP-30.0, Safety/Relief Valve Failures
EER 93-0428
Rev. 40
Page 5 of 100
Each of the SRVs discharge into a tailpipe which directs the effluent
to an underwater "T" quencher located near the bottom of the
Suppression Pool (-8 ft). The "T" quenchers have three sizes of
holes drilled in them to allow for even steam condensing across the T
portion. The "T" quencher is designed to divide the steam flow and
direct it to the Suppression Pool water volume where the steam is
condensed. This results in a relatively slow and uniform heating of
the pool water vice a relatively rapid localized heating. Directing SRV
discharge to Suppression Pool water volume prevents containment
overpressurization which would result from direct discharge to the
Drywell air volume. Each SRV tailpipe contains flow and temperature
monitoring instrumentation intended to detect valve seat leakage or
The flow detectors (Acoustic Monitors) provide Control Room
annunciation, Safety/Relief Valve Open, (A-03 1-10 on P601) while
the temperature detectors indicate on Control Room recorder, B21
TR-R614 on Panel P614 and provide annunciation, Safety or
Depressurization Valve Leaking (A-03 1-1 on P601).
The SRVs are normally operated from the P601 panel in the relief
mode by electrically energizing the solenoids that direct pneumatics
to open the valve. The normal pneumatic source is Pneumatic
Nitrogen System (PNS) Division I and II while operating and Reactor
Non-Interruptible Air (RNA) System Division I and II when shutdown.
A Division I and II Backup N2 supply is provided as a separate
pneumatic source for SRVs when PNS/RNA are unavailable. Normal
valve position RED-GREEN indication is provided on the P601 panel
above the individual valve control switches. This indication is
provided electrically from the Acoustic Monitor System. An amber
lamp on the P601 panel at each SRV control switch is also activated
by the acoustic monitor and seals in until reset. This light provides a
memory function to remind/inform the operator of operation of a
specific valve. Azimuth locations (Figure 25-1 D) of the SRV
discharge into the Suppression Pool provide an operator aid on the
P601 panel during Emergency Operations.
Seven of the SRVs serve as Automatic Depressurization System
(ADS) valves. Their purpose is to Automatically depressurize the
reactor vessel thus allowing injection by low pressure emergency
cooling sources. These valves are B21-F013 A, C, D, H, J, K, and L
and can be manually operated via individual AUTO-OPEN control
switches at the P601 panel. Their pneumatic actuation is discussed
in the ADS System (SD-20).
SD-25
Rev. 6
Page 12 of 87
BSEP 1 & 2
UPDATED FSAR
6.5
Fission Product Removal and Control Systems
6.5.1
Engineered Safety Feature
CESF) Filter Systems
The ESF filter
system is
the Standby Gas Treatment System (SGTS)
discussed below and the Control Room Emergency Filtration System, which is
discussed in Sections 6.4 and 9.4.1.
6.5.1.1
6.5.1.1.1
Design bases.
The principal functions of the SGTS are as
follows:
1. Maintain secondary containment below atmospheric pressure when the
secondary containment is contaminated.
2.
Clean up a contaminated drywell and/or suppression chamber
atmosphere when they are being vented to the atmosphere.
3.
Provide ventilation and clean up when venting a contaminated drywell
after the nitrogen inerting procedure which follows a loss-of-coolant accident
(LOCA).
The above functions are consistent with the secondary containment safety
objective that is to limit the offsite dose after an accident to a practical
minimum and to within 10CFRIOO values.
Although not a principal design function, provisions have been made to
vent the primary containment through the SGTS at a reduced rate as a backup
means of controlling post-LOCA hydrogen generation.
When the SGTS is used as a backup to control post-LOCA hydrogen
generation, the flow path of the air will be through one bank of prefilters,
two banks of high efficiency particulate air (HEPA)
filters,
and two banks of
charcoal adsorber cells.
A source of nitrogen make up is also required for
this evolution.
The design bases employed for sizing the filters,
fans, and associated
ducting are as follows:
1. Each filter
train and blower in conjunction with the secondary
containment ducting was designed to maintain a negative secondary containment
pressure of 0.25 in.
of water by controlled venting at the rate of 100 percent
volume per day following Reactor Building isolation.
2.
The SGTS was designed to withstand the anticipated fission product
heat loading from a TID-14844 release without reaching the desorption
temperature of the charcoal halogen filters.
Revision No.
17F
6.5.1-1
BSEP 1 & 2
UPDATED FSAR
3.
When operating at nominal rated capacity (3000 scfm),
each filter
train's
moisture separator was designed to operate during secondary
containment mode at a maximum ambient condition of 1050F and 50 percent
relative humidity.
4.
The prefilter was designed to remove large particulates and
protect the HEPA filter.
5.
The HEPA filters
were designed to remove particulates of
0.3 micron size and larger.
In-place filters
were designed for a minimum of
99 percent efficiency using the standard dioctyl phthalate (DOP) test.
The
clean pressure drop was designed to be no more than 1 in.
water gauge.
The
filters
were designed to be capable of withstanding a moisture loading in the
form of mist or fog that would produce a pressure of 10 in.
water gauge for
15 minutes; the filters
will not suffer permanent damage or a decrease in
efficiency after the filters
dry out.
6.
The charcoal adsorber unit is
capable of removing at least
99 percent of iodine with 5 percent in the form of methyl iodine, CH3I,
under
entering conditions of 70 percent relative humidity.
Each of the charcoal
canisters are filled with impregnated charcoal with an ignition temperature of
not less than 340 0 C (644OF
and a desorption temperature of at least
150 0C (3020F).
The ducts, filter
trains, blowers, and associated valves were designed
to withstand the maximum earthquake without impairing the ability of the
system to operate at design capability.
The design of the ducts and equipment
such as valves and operators, etc., prevents introduction of foreign materials
into the air stream.
6.5.1.1.2
system design.
The SGTS provides a means for minimizing thee
release of radioactive material from the containment to the environs by
filtering and exhausting the atmosphere from the Reactor Building during
containment isolation conditions.
Elevated release is
assured by exhausting
to the plant stack.
Provisions have been made for directing the drywell purge
air or the pressure suppression chamber vent exhaust to the SGTS if
so
desired.
The basic system consists of a suction duct, two parallel filter
trains
and blowers, and a discharge vent.
The suction duct draws from Elevation
50 ft into which all areas of the Reactor Building communicate.
A normally
closed suction is provided for the drywell and the suppression pool which is
opened only upon operator action.
Each of the two filter
trains contains a
moisture separator and a heater to provide humidity control, banks of
particulate and charcoal filters
to remove particulates and halogens, and a
blower.
Each blower is aligned with a particular filter
train.
The filter
trains and blowers are located in the standby gas treatment
(SGT) area which
is
on Elevation 50 ft of the Reactor Building.
The SGT fans discharge through
an underground pipe to the 100 meter high plant stack and then to atmosphere.
The diagram of the SGTS and the Reactor Building ventilation for Unit 2 is
shown in Figure 9-52.
The SGTS nominal flow rate is
3000 scfm.
This flow can
be obtained with each blower operating with the corresponding filter
train.
Acceptable higher flow rates, up to the maximum analyzed flow of 4200 CFM, may
be obtained during unthrottled system operation.
Revision No.
17?
6.5.1-2
STEPS SCCP-7 through SCCP-1O
STEP BASES:
Since instrumentation to read the 1350 F is not available, the operator is directed to use
the annunciator (setpoint of 1350 F).
If the reactor building ventilation exhaust radiation level is above 4 mRlhr, then the
Reactor Building HVAC should have automatically isolated. This step ensures that a
required automatic function has initiated. Confirming isolation of Reactor Building
HVAC subsequent to receipt of a high radiation signal or a high temperature condition
terminates any further release of radioactivity to the environment from this system.
SBGT is the normal mechanism employed under post transient conditions5 to maintai
reactor building pressure negative with respect to the atmosphere since the exhaust
from this system is processed and directed to an elevated release point before being
discharged to the environment.
001-37.9
Rev.O
Pagel4 o 39