ML021230024
| ML021230024 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 10/09/2001 |
| From: | NRC/RGN-II |
| To: | Virginia Electric & Power Co (VEPCO) |
| References | |
| -RFPFR, 50-280/02301, 50-281/02301 | |
| Download: ML021230024 (122) | |
See also: IR 05000281/2002301
Text
Draft Submittal
(Pink Paper)
1.
Senior Reactor Operator Written Exam
SURRY EXAM 2002-301
50-280, 281/2002-301
MARCH 18 - 28, 2002
Surry Initial SRO Exam 03/2002
r
QUESTIONS REPORT
for Surry2002
1. 001AA2.03 oo1/TIG2/TIGI//4.5/4.8//SR02301/S/
During a Reactor Startup with power stable at 1 x 10E-8 amps, the control rods begin to
withdraw in an uncontrolled manner (without operator action).
Which ONE of the following is the appropriate course of action?
AA. Manually trip the reactor.
B. Allow the control rods to step out until power reaches the POAH where FrD and MDT will
turn power.
C. Commence an Emergency Boration to compensate for the continuous rod withdrawal.
D. Place the Bank Selector switch from "MAN" to "Shutdown Bank A" since it is already fully
withdrawn.
Ref: Source SR EB #3361
Surry Lesson Plan ND-93.3-LP-3 Rev. 14 objective I
Surry Lesson Plan ND-93.3-LP-3 Rev. 14 p. 30
Surry abnormal procedure 0-AP- 1.00 step 2 RNO. Note stem places them in startup where the
Rod Control Mode selector switch is in MAN, which is step 2.
Time:
I
Points:
1.00
Version:
0 1 2 3 4 5 67 8 9
Answer: ACADACADDD
Scramble Range: A - D
Monday, October 29, 2001 09:15:47 AM
1
G.
Explain the purpose of the following Main Control Board reset pushbuttons, including the
alarms/components affected by use:
Start-Up Pushbutton
Alarm Reset Pushbutton
Reactor Trip Breakers' Reset Pushbutton
H.
Using a simplified one-line diagram for illustration, explain how the Insertion Limit
annunciators are generated.
I.
Explain the operator actions taken in AP-1.00, Rod Control System Malfunction, and AP
1.01, Control Rod Misalignment, to mitigate problems in the Rod Control System.
J.
Summarize the Technical Specifications associated with the Rod Control System.
K.
Reproducing simplified one-line diagrams for illustration purposes, explain the
overall integrated operation of the Rod Control System.
rFIesentafionn
Distribute all handouts.
Refer to/display H/T-3. 1, Objectives.
A.
Purpose and Design
1.
The Rod Control System serves two (major) purposes:
a.
Provides emergency shutdown (trip) of the reactor in response to signals
from the Reactor Protection System or the Reactor Operator.
ND-93.3-LP-3
Page5
Revision 14
a.
Review procedure entry conditions.
b.
Review AP-1.00 for each type of failure that can occur using the following
step sequence and highlighting the specified steps:
(1)
Continuous rod withdrawal or insertion.
(a)
Steps 1 and 2 are immediate action steps. If rod movement
cannot be stopped, the reactor is tripped.
(b)
If rod motion is stopped, the team checks for urgent failure,
stabilizes the unit, and makes notifications to repair the
problem and to management.
(2)
Dropped rod
(a)
Step 1 RNO sends team to step 4 to check for a dropped rod.
Dropped rod indication from the IRPI is considered to be
more reliable than from the NIS.
(b)
Review indications of a dropped rod IAW step 4.
(c)
If more than one rod has dropped, the reactor is subcritical,
or the reactor is less than 25% power, trip the reactor.
(d)
Reactor power is
and
transition
Misalignment.
reduced to < 70% power within one hour,
is
made
to
Control
Rod
ND-93.3-LP-3
Revision 14
Page 30
ATO/XETDRSPONSE
EPNS
O OT
INE
S.
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CAUTION:
The minimum temperature for criticality is
5220F.
If Tave decreases
below this temperature. Tech Spec 3.1.e must be reviewed.
9
5
S
S
5
S
S
S
S
S
S
S
[ 1] __CHECK FOR EITHER OF THE FOLLOWING:
GO TO Step 4.
"* Continuous rod withdrawal
"* Continuous rod insertion
2]
STOP ROD MOTION:
a) Put ROD CONT MODE SEL switch in <
MANUAL
b) Verify rod motion - STOPPED
b) Trip Reactor and GO TO 0(-E-0.
REACTOR TRIP OR SAFETY
INJECTION.
3.
GO TO STEP 13
QUESTIONS REPORT
for Surry2002
1. 005A2.03 001or2G3/T2G3/CAVITATION/C/A 2.9/3.1/N/SR02301/S/RLM
The following conditions exist:
-Unit 1 has been shutdown for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />
-RCS temperature is 140 degrees F
-RCS level is at mid-loop
-AA RHR pump is operating with lB RHR pump in standby
While adjusting RHR flow to lower RCS temperature, annunciator LB-G6, RHR HX LO FLOW,
alarms. Attempts by the Reactor Operator to stabilize RHR flow rate has been unsuccessful.
Which ONE of the following is the correct course of action?
A. Secure lA RHR pump and verify RCS level in the acceptable region.
B. Start lB RHR pump and verify RCS level in the acceptable region.
C. Verify RCS level in the acceptable region and start lB RHR pump
D. Verify RCS level in the acceptable region and secure 1A RHR pump.
Ref: Surry lesson plan ND 95.2-LP-12, Rev. 9 objective D.
Lesson plan p.44
1-AP-27.00, LOSS OF DECAY HEAT REMOVAL CAPABILITY, steps 9 through 11.
Answer D is the correct sequence as specified in the AP.
Time:
I
Points:
1.00
Version:
0 1 2 3 4 5 67 8 9
Answer: DDADCBCB BA
Scramble Range: A - D
Wednesday, October 31, 2001 08:16:55 AM
1
D.
[For each of the major categories of Loss of RHR events, explain how Surry's Abnormal
Procedures direct the operator to respond to the events, including the importance of proper
procedural sequence where applicable. SOPR Ri-0A Rex- 1t SOER 88-03, Ree 3I]
E.
Given a loss of RCS inventory with RCS pressure less than 1000 psig (SI Accumulators
isolated) and RCS temperature greater than 2000F, describe the use of AP-16.01 to address
this event.
F.
Given either hypothetical or actual situations involving a loss of RHR event (or the
potential thereof), differentiate between appropriate and inappropriate operator
actions, including why certain actions would aggravate a Loss of Decay Heat Removal
event.
Preentation
Distribute all handouts and AlAs.
Refer to/display HIT-12.1, Objectives, and discuss with the trainees.
A.
Event Synopses, Lessons Learned, Procedural and Design Features
1.
The following is a summary of the events that took place at San Onofre.
a.
Initial plant conditions
ND-95.2-LP-12
Refer to AIA-12.1, INPO 88-018, Case Study on Loss of Decay Heat Removal.
Have the trainees discuss the conditions and factors which caused the event and contributed
to the severity of the event.
Add any of the following as necessary to enhance the
discussion and to ensure all areas are discussed.
Page 4
Revision 9
b.
If the running RHR pump has failed, Step 4 RNO will establish the standby
pump to service. It is important to note that both the RHR FCV 1605 and
the heat exchanger outlet HCV-1758 are closed prior to the start of the
standby pump. This is done to prevent run out or vortexing if reduced
inventory conditions exist. Assuming the standby pump is started, the 1605
and 1758 valves are positioned to pre-event conditions. At this point, the
operator will go to Step 5 where proper RHR operation will be verified in a
series of steps that will lead to procedure termination. If the pump start was
unsuccessful or could not be accomplished due to electrical failures, the
operator is advanced to Step 16 where actions are initiated to employ steps
that will lead to establishing alternate methods of decay heat release.
c.
If a successful pump start was accomplished in Step 4 RNO, RHR flow
should be satisfactory JAW Step 5.
d.
In Step 6, vortexing most likely would not be occurring advancing the
operator to Step 12.
e.
Steps 12, 13 and 14 close out the procedure at this point provided stable
temperature control is achieved.
6.
Loss of inventory
a.
Step 1 states 9 bulleted parameters that may be indicators associated with a
loss of inventory.
b.
Step 2 takes actions to stop any inventory loss including isolating letdown.
Eventually Step 3 is entered which advances the operator to Step 15.
ND-95.2-LP-12
Page 44
Revision 9
Refer to AIA-12.6, ARP BG8, Shutdown Cooling Low Level.
Review ARP with trainees.
e.
(1)
Confidence in RVLIS is lacking and the system will not be available
if the outage is a refueling outage or if maintenance on the system is
being performed.
(2)
RVLIS readout is calibrated in % level and each percent accounts for
approximately 6 inches of water level.
This accuracy is not
acceptable for use of level monitoring near mid-nozzle. So, RVLIS
would only be good for "trending" of water level. In addition, if the
RCS goes into a vacuum, as sensed by the PTs associated with
RVLIS, all RVLIS channels will read "INVALID."
(3)
RVLIS would provide an indication or trend of lowering level if it
were operable.
Ask trainees: What would most likely be the first indications of the loss of level while at
mid-nozzle?
Answer: Shutdown cooling low level alarm and RHR pump amps fluctuating.
(4)
By the time the RHR pumps start vortexing, there may not be
enough time to respond to the event to prevent vapor binding of the
RHR pumps.
ND-95.2-LP- 12
Revision 9
Page 27
STEP
ACTION/EXPECTED RESPOE iR
ESPONSE NOT OBTAIN
I
6. _CHECK RHR PUMP - VORTEXING
GO TO Step 12.
- Flow indication on' l-R-FI-1605
-
OSCILLATING
- Amperage indication -
OSCILLATING
- * **
- *****
- *
CAUTION:
RCS temperature may increase if RHR flow rate is less than required
based on time after shutdown. (Attachment 1)
- * * * ********
- *****
- ***
- ***
7.
-REDUCE
RHR FLOW TO STOP VORTEXING
"* Use 1-RH-FCV-1605 in MANUAL
"* Use 1-RH-HCV-1758
8.
-CHECK
RHR PUMP -
STILL VORTEXING
9.
CHECK RCS LEVEL - WITHIN
ACCEPTABLE REGION
0 1-RC-LI-100A (Attachment 2)
GO TO Step 12.
Restore RCS level to Acceptable
Region of Attachment 2 or 3.
a 1-RC-LR-105 (Attachment 3)
NUMBER
PROCEDURE TITLE
REVISION
9
I-AP-27.00
LOSS OF DECAY HEAT REMOVAL CAPABILITY
PAGE
6 of 18
HSE
ACTION/EXPECTED RESPONSE
L
10. -VERIFY Bk PUMPS -
BOTH AVAILABLE
11.
-RESTORE RHR PUMPS:
a) Stop vortexing pump
b) Verify
i.HR flow -
NONE INDICATED
c) Manually close 1-RH-FCV-1605
and l-RH-HCV-1758
d) Start other Bil pump
e) Adjust l1 control valves to
return flow to pre-event rate:
"* 1-RH-FCV-1605
"* I-RH-HCV-1758
RESPONSE NOT OBTAINED I
Restore RHR pump:
a) Stop pump.
b) Verify RHR flow -
NONE
INDICATED.
c) Vent pump.
"* 1-RH-P-lA, 1-RH-9
"* 1-RH-P-IB, I-RH-3
d) Restart pump.
e) If Bil
pump can NOT be
restored,
=
GO TO Step 16.
f)
kF
RHR pump is restored, Tfl
GO TO Step 12.
e) GO TO Step 16.
QUESTIONS REPORT
for Surry2002
1. 008A2.01 001 /T2G3/T2G3/PUMP/C/A 3.3/3.6/N/SR02301/S/RLM
-Units I and 2 are at 100% power.
-lA component cooling water pump is tagged out for maintenance.
-No other activites are in progress on the component cooling water system for either unit.
Annunciator 1K-D6, CC PPS DISCH HDR B LO FLOW alarms. The Reactor Operator notes
lB component cooling water pump motor amps at minimum, but steady and greater than zero.
Which ONE of the following is the most probable cause of this alarm and what action should be
taken?
v&A.
lB component cooling water pump has failed and the system should be crosstied the other
unit.
B. The flow indicator has failed and a work request should be written.
C. There is a line rupture and makeup to system should aligned, the leak identified and isolated.
D. The discharge valve has been throttled and should be opened as required to clear the alarm.
Ref: Surry lesson plan ND-88.5-LP-1, objective G.
Annuciator response procedure 1K-D6, CC PPS DISCH HDR B LO FLOW, symptoms and
actions.
Answer A is correct because the low flow in conjunction with the low, but greater than zero
amps, indicates a pump shaft shear with the motor going to no load amps. With the other train
pump unavailable, the procedure provides cross connect as the only option.
Answers B, C, and D are also possible symptoms of the alarm, as listed in the ARP. However,
answer B is incorrect because a flow indicator failure would not explain pump amp decrease.
Answer C is incorrect because, per the ARP, the pump amps would oscillate.
Answer D is incorrect because the stem said no other activities on either unit's ccw system was in
progress.
Note to self: The ARP's for the low flow annunciator (lK-D6 and lK-C6) are asymetrical.
1K-C6 says the alarm is disabled when the outlet from the A RHR Hx is closed. IK-D6 is silent.
Need to know if the alarm in my scenario is disabled.
Time:
I
Points:
1.00
Version:
0 1 2 34 5 67 89
Answer: ACCACCBAAB
Scramble Range: A - D
Thursday, November 01, 2001 09:40:20 AM
1
D.
Summarize the contents of the normal and abnormal procedures associated with the
component cooling system, including:
Normal system operation
AP- 15.00, Loss of Component Cooling
AP- 16.00, Excessive RCS Leakage
E.
State the technical specifications associated with the component cooling system, including
for SRO candidates, the basis behind these specifications.
F.
Describe the major system components and operation of the Chilled CC System, including:
System purposes and components supplied
Chilled CC pumps
Heat Exchangers and Valves
Indications and controls
G.
Describe the overall integrated operations of the component cooling system.
Distribute all handouts.
Refer to/display H/T- 1.1, Objectives, and review with trainees.
A.
System Components
I.
The component cooling system purpose is to provide a cooling medium for various
heat loads of each reactor unit.
radioactivity to the environment.
2.
CC Surge Tank
It also acts as a barrier against the release of
ND-88.5-LP- 1
Revision 16
Page 4
Level 2 Controlled DIstrIbutIon
M aint ainejr~l*[74)eP0M*Jnt
Do not remoAL
tdflT*
13ua
Owork
ANNUNCIATOR RESPONSE PROCEDURE
REVIS IC
PROCEDURE TITLE
FLOW
2_
PAGE
lof
3
1K- 30
9.0
S-ESK-lOK
"*MS.-72
h Spec 3.13
- l-005, Instrumentation SetPoints
7 95-001.27. KI Setpoint Change
CAUSES
Alarm actuates when i-CC-FS-IOOB senses CC Header B flow less than or
iequal to 4,000 gpm.
Low CC flow may be caused by one or more of the following:
"* CC Pump failure.
"*
Flow indicator failure.
"*
Line rupture.
"*
Discharge valve throttled.
Instrumentation failure has occurred.
,,e,
,No
SIA'Q
PROCEDURE TITLE
4ACTION/EXPECTED-RESPONSE
RESPONSE NOT
L_VZRIFY ALARM - DUE TO PLANNED
EVOLUTION
.RETURN
TO PROCEDURE IN EFFECT
VERIFY ALARM - CC SUP HDR B FLOW
LESS THAN OR EQUAL TO 4,O00 GPM
GO TO Step 3.
-
Initiate a Work Request AND GO TO
Step 8.
INCREASE SURVEILLANCE OF
COMPONENTS SUPPLIED BY CC
.REDUCE
- .************
UH:
An LCO will be entered if Component Cooling is lost or if
System components are inoperable lAW Tech Spec 3.13,
certain CC
- -*
- * **
- *****
- *****
- ***
- **
_CHECK CC SYSTEM FOR LEAKAGE
Verify running or start a CC Pump.
"" CC Surge Tank Level - DECREASING
"*
Aux Building or CTMT Sump Level
- INCREASING
"* CC Pump amps - OSCILLATING
f
a CC Pump can MO2 be started,
T=j do the following:
a) Verify croastied or crosatie CC
System.
b) Review Tech Spec 3.13.
c) CO TO Step 8.
U CC can NO
be restored,
=JI
GO
TO 1-AP-15.O0,
LOSS OF COMPONENT
COOLING.
REVISION
2
PAGE
2 of 3
OBTAINEDI
QUESTIONS REPORT
for Surry2002
1. 008AA2.19o001/ r1G2/TI G2/STUCK SPRAY \\ALVE/3.4/3.6/B/SR02301/S/RLM
Which ONE of the following is an indication of a stuck open Pressurizer
spray valve?
A. Pressurizer pressure decreasing and level decreasing.
B. Pressurizer pressure decreasing and level increasing.
C. Surge line temperature decreasing.
D. High temperature on either Pressurizer Spray Line temperature indicators 1-RC-TI- 1451 or
1452
REF: SR EB # 32067
Answer C is incorrect based upon probable cause listed in ARP IC-G8
Answer D is incorrect based upon 1-AP-31 entry condition 3, last bullet.
Wednesday, October 24, 2001 11:23:56 AM
1
VIRGINIA POWER
SURRY POWER STATION
ABNORMAL PROCEDURE
NUMBE
PROCEDURE TITLE
REVISION
I-AP-31.00
INCREASING OR DECREASING RCS PRESSURE
5
PAGE
(WITH 2 ATTACHMENTS)
1 of 6
PURPOSE
To provide guidance in the event of abnormal RCS pressure caused by a
plant transient or equipment malfunction.
ENTRY CONDITIONS
1.
Decreasing RCS pressure as indicated by any of the following:
"* PRESS LO PRESS annunciator.
"* Decreasing trend on PRZR PRESS Recorder. 1-RC-PR-1444 Pon 2
2.
Increasing RCS pressure as indicated by any of the following:
"* PRESS HI PRESS annunciator. 1C-F8
"* Increasing trend on PRZR PRESS Recorder. 1-RC-PR-1444 Pos 2
3.
Failure of RCS pressure control component(s) as indicated by any
of the following:
"* PRZR PRESS CONTR HI OUTPUT annunciator.
iC-A8
"* PRZR HTRS CONT GP OL TRIP annunciator. iC-H8
"* Increasing or decreasing trend on PRZR PRESS Recorder.
1-RC-PR-1444 Poo 1
"* Leaking PRZR Safety Valve(s) or PORV(s) as indicated by either
of the following:
a.
PRZR SFTY VV LINE HI TEMP annunciator. iC-C7
b.
PRZR PWR RELIEF LINE HI TEMP annunciator. 1C-D7
"* Leaking PRZR spray valve as indicated by low temperature on
PRZR SPRAY LINE TEMP temperature indicators 1-RC-TI-1451 or
i-RC-TI-1452
APPROVAL RECOMMENDED
DATE
VIRGINIA POWER
Level ZWUCIbUUOM
Maintained by this Department
Do n*MWOREl]clU.llg"
i~ill
LNUKBE:R
PROCEDURE TITLE
REVISION
1 of 2
REFERENCES
1.
UFSAR 4.0
2. 1*448-ESK-10C,
lOAJ
3. 1-DRP-005, Instrumentation Setpointi
PROBABLE CAUSES
1. Alarm actuates when 1-RC-TC-1450 senses Pressurizer Surge Line temperature
less than or equal to 500'F.
2. Low Surge Line teuperature may be caused by one of the following:
"* Loss of continuous spray flow
"* Cooldown of RCS
3. Instrumentation failure has occurred.
APPROVAL RECOMMENDED
APPROVED
DATE
CIHAIRMAAN STATION NUCLEAR SAFETY
AND OPERATING COMMITTEE
Form No 7237508(
QUESTIONS REPORT
for Surry2002
1. 009EA2.04 001/'1IG2/TI G2//3.8/4.0/B/SR02301/S/RLM
-Unit 1 has experienced a small break LOCA.
-The RCP's are tripped.
-A cooldown has been performed
-The plant has been depressurized in accordance with l-ES-1.2, Post LOCA Cooldown and
Depressurization.
Which ONE (1) of the following explains why pressurizer level will eventually stabilize?
-A.
Break flow equalizes with injection flow.
B. The void in the vessel head stops expanding.
C. ECCS injection flow has been heated and expanded and is now in the thermal equilibrium
with decay heat generation.
D. Accumulators have partially injected to raise pressurizer level and are now at equal pressure
with the RCS.
Source: FA EB# 44701
Surry Lesson Plan ND-95.3-LP-9, Rev 8
Learning objective B
Correct answer based on pp. b(l), p.5
'l-rhursday,
October 25, 2001 06:05:02 AM
(d)
An explicit check of S/G levels is performed and is contained
within the main cooldown loop.
This ensures continuous
monitoring for possible SGTRs.
(2)
After these actions and checks are performed, a cooldown to CSD is
initiated.
With continued cooldown, subsequent actions can be
performed when specified RCS subcooling criteria are satisfied.
b.
DEPRFRSIIRTZE RCS TO R*F"ILL PRFRSSITRIZER.
(1)
This action is performed prior to RCP restart or before/after an SI
reduction action. As RCS pressure decreases, injection flow will
increase relative to break flow. Consequently, this depressurization
action should be sufficient for restoring pzr level if the LOCA is
small.
(2)
A "small" LOCA is first ensured by requiring RCS subcooling
before depressurization.
If subcooling
is lost during the
depressurization, it should be restored as the cooldown continues.
Prior to restoring pzr level, all pzr heaters are turned off.
c.
START ONE RCPISTOP ALL BIUT ONE RCP.
(1)
Once RCS subcooling, pzr level, and other RCP support conditions
are established, an RCP can be started if no RCPs are running. The
RCP restarted (or left running) is used to provide normal pzr spray
and mix the RCS.
(2)
If more than one RCP is running, all but one are stopped to minimize
RCS heat input.
ND-95.3-LP-9
Revision 8
Page 5
Ilntrduction
ES-1.2, Post-LOCA Cooldown and Depressurization, provides guidance to cooldown and
depressurize the RCS to cold shutdown conditions following a loss of reactor coolant.
This
procedure and supporting analyses are structured to deal primarily with small LOCAs where SI
flow can keep up with break flow, at pressures above the shutoff head of the LHSI pumps.
In addition, if a LOCA occurs and the HHSI system fails, the procedure provides optimal recovery
actions to try to prevent an inadequate core cooling condition while trying to restore SI flow.
After reaching and maintaining cold shutdown conditions (RCS temperature less than 2000F), the
final step of ES-1.2 instructs the team and plant engineering staff to evaluate the long-term plant
status. At this time, the RCS will be cooled by either RHR or the cold/hot leg recirculation mode.
This lesson plan on the post-LOCA cooldown and depressurization will present the procedure both
from a "big-picture" perspective and from an "in-depth" perspective.
After receiving this instruction, the trainee will be able to:
A.
Given the major action categories associated with ES-1.2, Post-LOCA Cooldown and
Depressurization, explain the purpose of ES-1.2, the transition criteria for entering and
exiting ES- 1.2 and the types of operator actions that will occur within each category.
B.
Given a copy of ES-1.2, Post-LOCA Cooldown and Depressurization, explain the basis of
each procedural step.
ND-95.3-LP-9
Revision 8
Page 2
QUESTIONS REPORT
for Surry2002
1. 010A2.03 002/T2G2/T2G2/RCS LEAKAGE/C/A 4. 1/4.2/B/SR02301/S/RLM
A pressurizer PORV is leaking by the seat to the PRT at a rate of I gpm. All other system
components are normal.
12 6,
c
, .
.,t-
ct
Which ONE of the following describes the Technical Specification classification and required
actions?
A. Unidentified leakage that requires shutdown.
B. Identified leakage that requires shutdown.
C. Unidentified leakage with no shutdown required.
.'D. Identified leakage with no shutdown required.
Ref: SR EB #1773
ND-88.I-LP-9H; SROUTP-SDS-1/C; TS 3.1.C
Time:
1
Points:
1.00
Version:
0 I 2 3 4 5 6 7 8 9
Answer: DBDBDAABCB
Scramble Range: A - D
Thursday, November 01, 2001 11:17:16 AM
1
H.
Describe the RCS Tech Specs, including for the SRO candidate, the basis behind each
specification.
I.
Prepare a general content outline
of the subject matter in
Surry Technical
Specifications, specifying the major area to which each section is dedicated, including
a detailed description of the RCS section of Tech Specs.
Presentation
Distribute all handouts.
Refer to/display H/J-9. 1, Objectives, and review with trainees.
A.
Tech Spec Section 1.0, Definitions
This section presents a number of frequently used terms. However, looking at 1OCFR50.36,
"Definitions" is not a required section of Tech Specs.
Ask trainees: Why are definitions considered important enough to be a T.S. Section?
Answer: To ensure consistency and set a standard for terminology. To provide for uniform
interpretation of the specifications.
Review each of the definitions in Tech Spec Section 1.0.
Refer to/display HIT-9.2, 2.0 Safety Limits and Limiting Safety System Settings.
ND-88.1-LP-9
Page3
Revision 10
(2)
3. .B - Requirements for RCS component Heatup and Cooldown
limits.
(3)
3.1 .C - Limits for RCS and associated component leakage.
(4)
3.1.D - Limits for the activity levels in the RCS.
(5)
3.1.E - Requirements for the Minimum Temperature for Criticality.
(6)
3.1 .F - Limits for RCS chemical containment concentrations.
(7)
3.1 .G - Requirements for RCS Overpressure Mitigation Operability.
f.
Tech Spec Section 3.2 - This section describes the CVCS components that
must be operable. This section also provides the definition of AVAILABLE.
g.
Tech Spec Section 3.3 - This section provides the requirements for the SI
system components.
h.
Tech Spec Section 3.4 - This section provides the requirements for the CS
and RS components.
i.
Tech Spec Section 3.5 - This section provides the requirements for the RHR
system.
j.
Tech Spec Section 3.6 - This section provides the requirements for the SG
Safety Valves and the AEW ýystems components.
ND-88.1 -LP-9
Revision 10
Page7
TS 3.1-23
3-17-72
Specifications
1.
Detected or suspected leakage from the Reactor Coolant System
shall be investigated and evaluated.
At least two means shall
be available to detect reactor coolant system leakage,
One
of these means =mst depend on the detection of radionuclides
in the containment.
2.
If the leakage rate, from other than controlled leakage sources,
such as the Reactor Coolant Pump Controlled Leakage Seals.
exceeds 1 Zpa and the source of the leakage is not identified
within four hours of leak detection, the reactot shall be brought
to hot shutdown.
If the source of leakage is not Identified
within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be brought to a
cold shutdown condition.
3.
If the sources of leakage are identified and the results of the
evaluations are that continued operation Is sate, operation of
the reactor with a total leakage, other than leakage from
controlled sources, not exceeding 10 gpm shall be permitted
except as specified in C.4 below.
6.
If it Is determined that leakage exists through a nau-isolable
fault which has developed in a Reactor Coolant System component
body, pipe well, vessel vall, or pipe weld, the reactor shall be
brought to a cold shutdown condition and corrective action taken
prior to resumption of unit operation.
5.
If the total leakage,other than leakage from controlled sources,
exceeds 10 gp. the reactor shall be placed in the cold shutdown
condition.
QUESTIONS REPORT
for Surry2002
1. 012A2.02 001
Procedures AP-10.02, AP-10.03, and AP-10.04 (Loss of Vital Bus H, I1, or IV) direct the
operator to trip the reactor prior to tripping the affected RCP.
Which one of the following is the basis for tripping the reactor before tripping the RCP?
A. To ensure a cooldown rate is initiated in the affected loop.
B. To prevent exceeding the linear heat generation rate limit.
C. To ensure SDM is present when backflow through the affected loop is initiated.
D. To prevent an unnecessary challenge to the Reactor Protection System.
Ref: Surry Exam Bank.
Lesson Plan ND-90.3LP-5E; AP-10.02, AP-10.03, AP-10.04.
Time:
I
Points:
1.00
Version:
0 1 2345 67 89
Answer: DADDBBDAAA
Scramble Range: A - D
RO Tier:
T2G2
SRO Tier:
T2G2
Keyword:
Cog Level: 3.6/3.9 MEMORY
Source:
B
Exam:
SR02301
Test:
S
Misc:
GWL
Monday, October 29, 2001 03:26:05 PM
1
B.
[Describe the components and indications associated with an Uninterruptable Power Supply
(UPS). SOER 83-03, Recommendation 11]
C.
Describe the power sources and loads associated with the Appendix R distribution system.
D.
Describe the power sources and loads associated with the Semi-Vital Bus distribution
system.
E.
[Given a loss of a Vital or Semi-Vital bus, describe the actions taken IAW AP- 10.01, 10.02,
10.03, 10.04, and/or 10.05 to address this loss. SOER 83-03, Recommendation 11 and
SOER 81-02, Recommendation 5]
F.
Given a loss of a Vital or Semi-Vital bus, describe the effect on Plant indications and
controls, including actions taken lAW applicable APs to address the loss.
Presentation
Distribute all handouts.
Refer to/display H/T-5. 1, Objectives, and review with trainees.
A.
One-Line Diagram
1.
The purpose of the Vital Bus Distribution System is to supply a stable, reliable
source of power to vital instruments.
It must remain uninterrupted to prevent
spurious shutdowns and guarantee proper action when instruments or controls are
required.
ND-90.3-LP-5
Page4
Revision I11
LESSON PLAN
Introduction
It was a normal shift on Surry Unit 2. The date was 10-10-82. Surry #2 was operating steady state
100% power with no evolutions planned. Suddenly, the Unit 2 annunciators sounded. A runback
of the turbine started. The operators verify the steam dumps are opening and that the rods are
inserting. They diagnosed that vital bus #3 had failed. Shortly after this and within a few seconds,
the #2 reactor trips and then safety injects.
Each event listed above actually occurred. The loss of a vital bus is a major challenge to the plant
instrumentation and to the operator.
A loss of power to a vital bus can result in either a runback or a reactor trip. This results in a loss of
income for the company. To reduce the possibility of loss of power to a vital bus, uninterruptable
power supplies were installed during an upgrade of the 120 VDC and vital electrical distribution
system.
This lesson plan will provide the information for the trainee to identify and respond properly to a
transient on any vital bus. It will discuss the operation and the location of major vital bus com
ponents.
Objectives
After receiving this instruction, the trainee will be able to:
A.
[Using a one-line diagram drawn from memory, describe the components and current
flowpaths of the Vital, Semi-Vital, and Appendix R Distribution Systems. SOER 83-03.
Recommendation 11 ]
ND-90.3-LP-5
Page3
Revision I11
EXAMINATION ANSWER KEY
RO/SRO Exam Bank
ID: AOP0050
Points: 1.00
Procedures AP-10.02, AP-10.03, and AP-10.04 (Loss of Vital Bus II, Ill, or IV) direct the operator
to trip the reactor prior to tripping the affected RCP.
Which ONE of the following is the basis for tripping the reactor before tripping the RCP?
A.
To prevent exceeding the linear heat generation rate limit.
B.
To prevent an unnecessary challenge to the Reactor Protection System.
C.
To ensure a cooldown rate is initiated in the affected loop.
D.
To ensure SDM is present when backflow through the affected loop is initiated.
Answer:
B
Question 216 Details
Question Type:
Topic:
System ID:
User ID:
Status:
Must Appear:
Difficulty:
Time to Complete:
Point Value:
Cross Reference:
User Text:
User Number 1:
User Number 2:
Comment:
Multiple Choice
AOP0050 (AOP0049)
72525
AOP0050
Active
No
0.00
0
1.00
1.00
0.00
0.00
ND-90.3-LP-5E; AP-10.02, AP-10.03, AP-10.04
[S97-0497], [S95-1153], [S95-0431]
Page: 199 of 3141
216
10/19/01
OPS RO/SRO SU
QUESTIONS REPORT
for Surry2002
1. 033G2.4.21 001
-A feed line break in Containment is in progress.
-The reactor failed to trip and FR-S. 1 has been entered.
-Containment Pressure is 10 psig. 2.1 I
-When Checking for subcriticality, power is still approximatelyl5% and falling.
Which one of the following conditions must be present JAW FR-S. I to satisfy the subcriticality
criteria, and allow "an exit from the procedure?
A. A negative Intermediate range start up rate and Tavg. trending down.
B. A negative gamma-metric wide range power decreasing and Gamma-metrics wide range
power < 5%.
C. A negative intermediate range startup rate and power range channels less than 5%.
D. A negative intermediate range startup rate and gamma-metrics power range channels less
than 15%.
Surry Lesson Plan ND-95.3-36 objective D. Lesson plan page four paragraph C.
A. Incorrect, Tavg is not used to determine subcriticality.
B. Correct, with adverse conditions in containment the FR directs the gamma-metrics to be used
because the Excore NI's are not enviromentally qualified.
C. Incorrect, This would be used if no adverse conditions in containment existed.
D. Incorrect, again with adverse conditions in containment the Excores are not used.
Time:
I
Points:
1.00
Version:
0 1 2 3 4 5 6 7 8 9
Answer: BCCBACBAAB
Scramble Range: A - D
RO Tier:
TIG2
SRO Tier:
TIG2
Keyword:
Cog Level:
C/A 3.7/4.3
Source:
N
Exam:
SR02301
Test:
S
Misc:
GWL
Tuesday, November 06, 2001 01:10:50 PM
Presentation
Distribute all handouts.
Refer to/display H/T-36. 1, Objectives. Review objectives with trainees.
A.
Subcriticality Status Tree
1.
The Subcriticality status tree provides a systematic method to determine the status of
the Subcriticality Critical Safety Function. It evaluates whether any challenges to
this CSF exist or not.
2.
General
a.
This tree requires no operator action other than monitoring a limited set of
plant parameters and comparing them to reference values within the tree.
b.
This tree represents the highest priority Critical Safety Function and is
always entered first anytime tree monitoring is initiated. The tree can direct
operators to either of two subcriticality FRs.
c.
The Excore NI's are not environmentally qualified for adverse containment
conditions. For this reason the note exits to use the Gamma-Metrics Excore
neutron monitor system(Source and Wide Ranges) for monitoring the
subcriticality status tree for the duration of the event. The Gamma-metrics
are used once adverse containment numbers are exceeded.
d.
Since this tree is monitoring the reactivity state of the core, the parameters
being evaluated are those characterizing neutron flux behavior (leakage)
measured by the Ex-Core NIS and Ex-core Gamma-Metrics systems.
NDl-Q*5
-I P-36
Page 4
Revision 9
The Function Restoration procedure, FR-S. 1, Response to Nuclear Power Generation/ATWS,
provides guidance in the event of an unexpected nuclear flux condition following a Reactor Trip or
SI actuation or if an ATWS has occurred.
The objective of the recovery/restoration technique of FR-S. 1 is to add negative reactivity to restore
the core to subcriticality; restoration of shutdown margin is desired, but is not a necessity to exit
this procedure.
This lesson on FR-S. 1, Response to Nuclear Power Generation/ATWS, will provide an in-depth
look at the designed response to this challenge to the Subcriticality Critical Safety Function.
Obiectives
After receiving this instruction, the trainee will be able to:
A.
Given a simulated plant condition requiring the use of the critical safety function status
trees, transition through the subcriticality status tree denoting, in accordance with the rules
of priority, any applicable function restoration procedure needing implementation.
B.
Given the Major Action Categories associated with FR-S.1, Response to Nuclear Power
Generation/ATWS, explain the purpose of FR-S.1, the transition criteria for entering and
exiting FR-S. 1, and the types of operator actions that will occur within each category.
C.
Given a copy of FR-S. 1, Response to Nuclear Power Generation/ATWS, explain the basis
of each procedural step.
D.
Given actual or simulated plant conditions requiring implementation of FR-S.1,
Response to Nuclear Power Generation/ATWS, successfully transition through the
procedure, performing immediate operator actions from memory and applying step
background knowledge as required, to address the Critical Safety Function challenge
in progress.
ND-95.3-LP-36
Revision 9
Page 3
STEP
ACTION/EXPECTED RESPONSE
.
RESPONSE NOT OBTAINED
12.
_CHECK CETCs - LESS THAN 1200*F
JE CETC temperature increasing.
ACCIDENT CONTROL ROOM GUIDELINE
INITIAL RESPONSE.
NOTE:
If adverse CTNT conditions have been exceeded, the Gamma-Metrics
Excore Neutron Monitor system (Source and Wide Ranges) should be
used to monitor neutron flux for the duration of the event.
13. _.VERIFY REACTOR SUBCRITICAL:
a) Check power range channels
LESS THAN 5% [Gama-Metrice
Wide Range Power - LESS
THAN 5%]
b) Check the following:
- Intermediate range channels
NEGATIVE STARTUP RATE
[Gamma-Metrics Wide Range
Power - DECREASING]
Do the following:
1) Continue to borate.
IH
boration
=O effective, THE
allow RCS to heat up.
2)
Do actions of other FR8 in
effect which do
=OT cooldown or
otherwise add positive
reactivity to the core.
3)
RETURN TO Step 4.
PR
1-OP-RX-002.
(CALCULATED AT ZERO POWER)
GREATER THAN 1.77%
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
CAUTION:
Boration should be continued to
subsequent actions.
obtain adequate shutdown margin during
14. _REJTURN TO PROCEDURE AND STEP IN
EFFECT
-END -
QUESTIONS REPORT
for Surry2002
1. 026AA2.01 001/T1G1/TICGI/CCWS LEAK/2.9/3.5/B/SR02301/S/RLM
Given the following plant conditions:
- The plant is shutdown.
- RCS cooldown is in progress.
- RCS temperature is 190 degrees F.
- The CC SURGE TK HI-LO LVL alarm has actuated.
- The CC surge tank indicates an increase in level.
Which one of the following could be the cause of the problem?
A. The automatic makeup valve has malfunctioned causing level to increase.
B. TV-CC-109A, CC RTN HDR A OTSD TRIP VLV, has closed isolating the "A" CC return
C.V'A leak is present in the RHR heat exchanger.
D. A leak is present in the Seal Water return cooler.
REF: SR EB # 43389,
ND-88.5-LP-1, Rev. 16, pg. 8, Obj. G.
ND-88.5-AIA- 1.1 p.l&2
Answer C is correct due to plant conditions being in a cooldown at 190 degrees F (RHR
inservice)
Answer A is incorrect because makeup sources have only manuall operation capability. (see
Lesson Plan p.5.
Answer B is incorrect because the hydraulics of the system do not support the answer.
Answers D is incorrect because they it is not listed in the lesson plan as a posible inleakage
source.
Wednesday, October 24, 2001 09:20:07 AM
D.
Summarize the contents of the normal and abnormal procedures associated with the
component cooling system, including:
Normal system operation
AP- 15.00, Loss of Component Cooling
AP- 16.00, Excessive RCS Leakage
E.
State the technical specifications associated with the component cooling system, including
for SRO candidates, the basis behind these specifications.
F.
Describe the major system components and operation of the Chilled CC System, including:
System purposes and components supplied
Heat Exchangers and Valves
Chilled CC pumps
Indications and controls
G.
Describe the overall integrated operations of the component cooling system.
Presentation
[
Distribute all handouts.
Refer to/display H/T-1.1, Objectives, and review with trainees.
A.
System Components
1.
The component cooling system purpose is to provide a cooling medium for various
heat loads of each reactor unit. It also acts as a barrier against the release of
radioactivity to the environment.
2.
CC Surge Tank
ND-88.5-LP-1
Revision 16
Page 4
Ask Trainees: What are some of these possible leakage sources?
Write on chalkboard as trainees name the sources:
0
RCP thermal barriers,
0
Primary sample coolers,
High Radiation Sample system coolers,
Boron Recovery system heat exchangers and pump
seals,
0
Fuel pit coolers
Non-regenerative heat exchangers,
Excess letdown heat exchanger, and
RHR heat exchangers and pump seals.
b.
RM-SW-107A/B/C/D
Each CCHX has an in-line radiation monitor installed in a well on the SW
discharge piping from each HX. The detectors are connected to individual
modules on the Common Rad Monitor Panel in the MCR.
7.
Chemical Addition
a.
The CC system is provided with a 120-gallon chemical addition tank. Its
original function was to provide a means of adding either potassium
hydroxide for Ph control or potassium chromate (or dichromate) for
corrosion control.
ND-88.5-LP- 1
Page 8
Revision 16
a.
The CC surge tank provides the NPSH for the CC pumps. It is located
approximately 30 feet above the pumps, ensuring that an adequate head
exists at the pump suction to prevent cavitation. The surge tank allows for
fluid expansion and contraction and provides a source of makeup to the
system.
b.
The surge tank has a capacity of 2810 gallons and is normally maintained
approximately 60% full, allowing sufficient volume to accommodate minor
system surges and thermal swell due to cooldown operations.
c.
Makeup water is provided by the condensate system via the bearing cooling
makeup pump (1-BC-P-2) or the high pressure condensate header. There is
no automatic makeup control provided, therefore, both sources of makeup
water require a manual valve lineup in the Turbine Building basement.
d.
The tank is vented to the process vent system via HCV-CC- 100. This vent
valve will automatically close upon receipt of a CC radiation monitor alarm.
3.
Component Cooling Pumps
a.
The CC pumps provide the motive force for circulating cooling water
through the CC heat exchangers, individual system loads, and back to the
pump suction.
Normally two pumps (one per unit) supply the required
cooling water flow.
The two standby pumps provide 100% backup
capability.
The standby pump will auto start on a low discharge header
pressure of 55 psig.
b.
Each pump is rated at 9000 gpm at 200 ft. head.
ND-88.5-LP- 1
Revision 16
Page 5
ND-88.5-AIA- 1.1
Page 1 of 2
COMPONENT COOLING SYSTEM LOADS
Common BR Components
1.
Stripper Overhead Condenser
2.
PDT Vent Chiller Condenser
3.
PDT Pump
4.
High Level Waste Drain Tank Pump
5.
Overhead Gas Compressor
6.
Stripper Trim Cooler
BR Evaporator Components
7.
HRSS Sample Coolers t
8.
BR Evaporator Circ Pumps
9.
BR Distillate Coolers t
10.
BR Overhead Condensers e
11.
BR Evaporator Distillate Pumps 9
12.
Primary Sample Coolers t
Spent Fuel Pit Coolers/Cask
13.
Spent Fuel Pit Coolers
14.
Spent Fuel Pit Cask
NRHX/Seal Water RTN
15.
Nonregenerative Heat Exchanger t
16.
Seal Water Return Cooler
Notes
t Possible source of leakage into Component Cooling
. Abandoned in place equipment
ND-88.5-AIA- 1.1
Page 2 of 2
COMPONENT COOLING SYSTEM LOADS
CARF/NST
17.
Containment Instrument Air Compressor
18.
Containment Air Recirc Fan Coolers
19.
Neutron Shield Tank Coolers
CRDM Shroud Cooling/RCP
19.
Shroud Cooling Coils
21.
RCP Thermal Barrier Heat Exchangers t
22.
RCP Motor Air Coolers
23.
Hot Pipe Containment Penetration Cooling (>1500F)
24.
Containment Penetration Coolers
a.
Letdown
b.
Blowdown
c.
d.
Main Feed
Excess Letdown/RHR
25.
Excess Letdown Heat Exchanger t
26.
Primary Drains Cooler
27.
RHR Heat Exchanger t
28.
RHR Pump Seals t
29.
Primary Shield Wall Coolers - for each loop penetration
Nntes
T Possible source of leakage into Component Cooling
QUESTIONS REPORT
for Surry2002
1. 029G2.4.21 001/T1G2/TI Gl/EVALUATE PERFORMANCE/C/A 3.7/4.3/B/SR02301/S/RLM
The following plant conditions exist:
-An ATWS is in progress.
-All feedwater to the steam generators has been lost.
-The turbine generator has remained loaded and running.
Which ONE of the following would be an indication of the above conditions several minutes
after the ATWS occurred? (Assume all control systems are in AUTO and no operator action is
taken.)
A. Reactor power increases; pressurizer pressure decreases; pressurizer level decreases; steam
pressure increases.
B. Reactor power decreases; pressurizer pressure decreases; pressurizer level decreases; steam
pressure decreases.
C. Reactor power decreases; pressurizer pressure increases; pressurizer level increases; steam
pressure decreases.
D. Reactor power remains stable; pressurizer pressure increases; pressurizer level increases;
steam pressure increases.
Ref: Surry EB #TAA0081
Surry lesson plan: ND-95.I-LP-l 1, obj. B
RO Tier:
TIG2
SRO Tier:
TIGI
Keyword:
EVALUATE PERFORMANCE
Cog Level:
C/A 3.7/4.3
Source:
B
Exam:
SR02301
Test:
S
Misc:
RLM
Thursday, December 13, 2001 07:25:06 AM
have tripped but did not. The reactor was tripped manually at the SRO's direction approximately 25
seconds after the trip demand was generated.
Subsequent testing of the reactor trip breakers
indicated that both breakers had failed to open, apparently due to mechanical binding in the
undervoltage trip mechanisms.
The failure of the RTBs to open automatically when required places total reliance on operator
actions to terminate a plant transient. Failure to initiate a reactor trip during certain transients
results in a potentially severe challenge to the integrity of the Primary Coolant System. If a manual
reactor trip is delayed, permanent damage to components and systems may occur. The Safety
Analyses that rely on automatic reactor trips also may be invalidated. The time available for
operator actions necessary to mitigate the consequences of certain events is varied and dependent
on initial plant conditions. This type of transient results in a severe challenge to the barriers
associated with the prevention of radioactive materials released to the environment. This lesson
plan will provide an insight into the most limiting analyzed ATWT event in order to provide the
background knowledge level required to discuss associated Emergency Response Guideline
Procedures associated with this events.
Ohiectves
After receiving this instruction, the trainee will be able to:
A.
Differentiate between the "trip-demand" signal first out annunciators and the "trip
indication" first out annunciators.
B.
Explain the sequence of events for the most limiting ATWT event.
C.
[Explain the two events that must occur following an ATWT in order to prevent the
Reactor Coolant System pressure from exceeding the stress limitations. SUER 83-08
Recommendation 11 ].
ND-95. I -LP- 11
Revision 5
Page 3
QUESTIONS REPORT
for Surry2002
1. 037AA2.10 001/T 1G2/TIG2/T S LEAKAGE/C/A 3.2/4. I/M/SR02301/S/RLM
Given the following:
-Unit I is operating at 100% power and the latest leak rate data
shows:
8.6 GPM - Total RCS leakage rate
1.6 GPM - Leakage into the PRT (previously evaluated as permissible)
3.0 GPM - Leakage into the Reactor Coolant Drain Tank from RCP seals
0.35 GPM - From the 1 A Steam Generator
0.34 GPM - From the IB Steam Generator
0.32 GPM - From the IC Steam Generator
2.0 GPM - Charging pump leakage (previously evaluated as permissible)
WHICH ONE (1) of the following identifies the RCS leakage that requires the plant to be
shutdown?
C. IDENTIFIED LEAKAGE
D. PRIMARY to SECONDARY LEAKAGE
Ref: HR EB # 44454
Answer A incorrect because no pressure boundary leakage was specified in the stem.
Answer B is incorrect because 8.6-(l.6+3.0+l.1+2.0) = .9 gpm unidentified is acceptable.
Answer C is incorrect because indentified leakage is less than 10 gpm.
Answer D is correct because S/G total leakage exceeds both 1 gpm and >500 gpd in the 1 A S/G.
Time:
I
Points:
1.00
Version:
0 1 23456789
Answer: DCBDDCACDA
Scramble Range: A - D
e5so.-,
/4-0-0-5
Lcrsonlknst
615& 4 4 tejtJ
Is1
tirI"d'iroy
0
t~cA
- f Cs
csAAvD ~ t&A ti
2s
1
Monday, November 05, 2001 11:32:19 AM
Untitled
- QNUM
44454
- HNUM
45878 (Do NOT change If < 9,000,000)
- ANUM
- QCHANGED
FALSE
- ACHANGED
FALSE
- QDATE
1995/06/26
- FAC
400
Shearon Harris 1
- RTYP
PWR-WEC3
- EXLEVEL
S
- EXMNR
- QVAL
- SEC
- SUBSORT
- KA
002000G005
- QUESTION
Given the following:
-The plant is operating at 75% power and the latest leak
rate data
shows:
11.3 GPM - Total RCS leakage rate
1.6 GPM - Leakage into the PRT
2.0 GPM - Leakage into the Reactor Coolant Drain Tank
1.5 GPM- Leakage past check valves from RCS to SI
,o'tko
system
<2)
1.7 GPM - Leakage into Equipment Drain Tank
0.8 GPM - Total primary to secondary leakage (Assume
distributed over all S/Gs)
2.0 GPM - Charging pump leakage
vi Cl
J
01'
WHICH ONE (1) of the following identifies the RCS leakage that
requires
the plant to be shutdown?
c. IDENTIFIED LEAKAGE
d. PRIMARY to SECONDARY LEAKAGE
Page 1
ki
t
TS 3. -1.3
3-17-72
C.
Leakate
Specifications
1.
Detected or suspected leakage from the Reactor Coolant System
shall be investigated and evaluated.
At least two means shall
be available to detect reactor coolant system leakage4
One
of these means must depend on the detection of radionuclides
"in the containment.
2.
If the leakage rate, from other than controlled leakage sources,
such as the Reactor Coolant Pump Controlled Leakage Seals,
exceeds 1 Spi and the source of the leakage is not identified
within four hours of leak detection, the reactor shell be brought
to hot shutdown.
If the source of leakage Is not identified
within an additional A8 hours, the reactor shall be brought to a
cold shutdown condition.
3.
If the sources of leakage are Identified and the results of the
evaluations are that continued operation is safe, operation of
the reactor with a total leakage, other than leakage from.
controlled sources, not exceeding 10 gpm shall be permitted
except as specified in C.4 below.
4.
If it is determined that leakage exists through a mom-IsOleble
fault which has developed In a Reactor Coolant System component
body, pipe well, vessel wall, or pipe welt, the reactor ehall be
brought to a cold shutdown condition ad corrective actinc taken
prior to resumption of unit operation.
5.
UI the total leakage,other than leakage from conurolled sources,
exceeds 10 "a the reactor shall be placed in the cold shutdown
condition.
TS 3.1-13a
4-20-81
6.
If the primary-to-secondary leakage through all steam generators not
isolated froe the hector Coolant System exceeds 1 gpo total and 500
gallons-per day through any one steam generator not isolated freo
the
Reactor Coolant Systam, reduce the leakage rate to within limits within
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold
shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
7a.
Prior to going critical all primary coolant system pressure isola
tion valves listed below shall be functional as a pressure isolation
device, except as specified in 3.1.C.7.b.
Valve leakage shall not
exceed the amounts indicated.
flax. Allowable
Leakage (see note
Unit 1
Unit 2
(a) below)
Loop A, Cold Leg
1-I"-79, I-SI-241
2-SI-79, 2-51-241
S5.0
yam for each
valve
Loop 3, Cold Leg
1-31-82, 1-S1-242
2-S-82, 2-SI-242
Loop C, Cold Log
1-SI-85, I-SI-243
24S1-85, 2-SI-243
b.
If Specification 3.1.C.7.a cannot be met, an orderly shutdown shall be
initiated and the reactor shall be in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in
the cold shutdown condition within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Notes
a-*1.
Leakage rates less than or equal to 1.0
a* n re considered acceptable.
2.
Leakage rates greater than 1.0 gym but less than or equal to 5.0
s anre considered acceptable if the latest measured rate has not
exceeded the rate determined by the previous test by an amount that
reduces the margin between measured leakage rate and the maxims
permissible rote of 5.0 We by 501 or Sreater.
3.
Leakage rates greater than 1.0 Pa but less than or equal to 5.0 agy
are considered unacceptable if the latest measured rate exceeded the
rate determined by the previous test by a
amount that reduces the
margin between measured leakage rate and the maximua permissible rate
of 5.0 an by 501 or greater.
4.
Leakage rates greater than 5.0 gpe are considered unacceptable.
H.
Describe the RCS Tech Specs, including for the SRO candidate, the basis behind each
specification.
I.
Prepare a general content outline of the subject matter in Surry Technical
Specifications, specifying the major area to which each section is dedicated, including
a detailed description of the RCS section of Tech Specs.
Presentaflon
Distribute all handouts.
Refer to/display H/T-9. 1, Objectives, and review with trainees.
A.
Tech Spec Section 1.0, Definitions
This section presents a number of frequently used terms. However, looking at 1OCFR50.36,
"Definitions" is not a required section of Tech Specs.
Ask trainees: Why are definitions considered important enough to be a T.S. Section?
Answer: To ensure consistency and set a standard for terminology. To provide for uniform
interpretation of the specifications.
Review each of the definitions in Tech Spec Section 1.0.
Refer to/display H/T-9.2, 2.0 Safety Limits and Limiting Safety System Settings.
ND-88. I-LP-9
Page 3
Revision 10
c.
Pressurizer H/U less than 100lF/hr and C/D less than 2000F/hr. Maximum
AT between pressurizer and spray water is 3200F.
Basis - Maintains thermal stresses at spray line nozzle below design limits.
9.
Tech Spec 3. LC - Leakage
a.
Detected or suspected leakage shall be investigated and evaluated.
b.
Two means of detecting RCS leakage shall be available, one of which must
depend on detection of radionuclides in containment
ND-88.1-LP-9
c.
If leak rate greater than 1 gpm from other than controlled leakage sources
and not identified within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, proceed to HSD; if still not found after
additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, proceed to CSD.
d.
If leak source identified and safe operations is verified, leakage may increase
up to 10 gpm.
e.
If leakage is through an unisolable fault in a component body, pipe well,
vessel wall, or pipe weld, proceed to CSD and repair prior to restart.
Page 26
Revision 10
Ask trainees: What indications are available to detect RCS leakage? List on chalkboard:
1.
Increased make-up water (charging and/or VCT M/U)
2.
High temp in Rx vessel flange leakoff
3.
Containment sump level
4.
Containment pressure, temperature, humidity
5.
Containment particulate and gas RM
6.
Other RMs - air ejector, CC, SGBD
f.
If total leakage, other than controlled, exceeds 10 gpm proceed to CSD.
g.
If primary-to-secondary leakage through all S/G not isolated from RCS
exceeds 1 gpm total or 500 gallons per day (.35 gpm) through any one S/G,
get leakage in spec within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or shutdown within the next six hours.
h.
Prior to criticality, the listed Tc check valves shall be functional with
leakage as follows:
(1)
less than 1 gpm acceptable
(2)
greater than 5 gpm unacceptable
(3)
leak rate >1 gpm but <5 gpm are acceptable so long as the new leak
rate does not reduce the margin to 5 gpm by Ž 50% of the difference
between the last leak rate and the present leak rate.
10.
Tech Spec 3.1 .D - Maximum Coolant Activity
a.
Total specific activity due to nuclides with half-lives of greater than 15
minutes shall not exceed 100 / E gci/cc when critical or >500'F. If not met,
shut down Rx and cool to <5000F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If exceeded limit by 25%,
perform cooldown to 500F within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
b.
Specific activity of RCS limited to *<1.0 jci/cc Dose Equivalent 1-131
whenever critical or >5000F
1-1TP-9
Page 27
Revision 10
QUESTIONS REPORT
for SURRY2002
1. 062A2.04 001/T2G2/T2G2/EFFECT OF BUS LOS S/M"3.31 /N/SR0230I/S/RLM
Plant conditions:
CD
L'
-Unit I Semi-Vital bus faulted
vc;
-Unit 1 tripped approximately.J.thttago duringjhe downpower rrured-duc to th faulted
j, i-Semi-Vi*ams
Which one of the following actions are required to maintain Tav at 547 'F duringrepair to the
Semi-Vital bus?
Ai/
via the PORV's IAW l-ES-0.1, Reactor Trip Response
B! Dump steam via the PORV's lAW 1-AP-10.05, Loss of Semi-Vital Bus
C. Dump steam via the steam dumps LAW l-ES-0.1, Reactor Trip Response
D. Dump steam via the steam dumps lAW I-AP-10.05/,4 a$
oSet,,- 01..6(&s
Ref:
Surry Lesson Plan ND-90.3-LP-5, objective F
Note: After 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, no power is available to either the steam dump or SG PORV controllers.
Only AP 10.05 provides guidance on local operation of SG PORV's
DR1
Monday, reUrUadry Uv',
UU: U
- .O.UJF
B.
[Describe the components and indications associated with an Uninterruptable Power Supply
(UPS). SOPR 8-03,
Recommendation 11]
C.
Describe the power sources and loads associated with the Appendix R distribution system.
D.
Describe the power sources and loads associated with the Semi-Vital Bus distribution
system.
E.
[Given a loss of a Vital or Semi-Vital bus, describe the actions taken IAW AP-10.01, 10.02,
10.03, 10.04, and/or 10.05 to address this loss. SOFR 93-fl,
Recommendation 11 and
SOFR R 1-02 Rrcommendation 5]
F.
Given a loss of a Vital or Semi-Vital bus, describe the effect on Plant indications and
controls, including actions taken IAW applicable APs to address the loss.
Presentation
Distribute all handouts.
Refer to/display HI/T-5.1, Objectives, and review with trainees.
A.
One-Line Diagram
1.
The purpose of the Vital Bus Distribution System is to supply a stable, reliable
source of power to vital instruments.
It must remain uninterrupted to prevent
spurious shutdowns and guarantee proper action when instruments or controls are
required.
ND-90.3-LP-5
Revision I11
Page4
STEP
ACTION/EXPECTED RES
RESPONSE NOT OBTAIE
~.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
CAUTION
Power to the Steam Header Pressure Manual/Auto station has been
lost, with the associated controller in Auto-Hold with fixed output
demand.
- The Steam Dump system condenser interlocks will remain energized by
an UPS in MB-8 for approximately 30 minutes.
The Steam Dump system
will continue to operate in Tave mode during this 30 minute period.
after which the Steam Dumps will be unavailable due to MB-8 losing
power.
approximately 30 minutea.
The PORVs will continue to control in
automatic at 1035 p.ig. or may be operated manually during this 30
minute period.
.. ..
..
.
.
.
...............
....
.
8.
VERIFY SEMI-VITAL BUS
-
NOT
ELECTRICALLY FAULTED
"* Semi-Vital Bus lost as result of
a loss of Emergency Bus
"* Electrical Department confirms
Semi-Vital Bus
MOT faulted
GO TO Step 11.
WHEN fault
corrected. T=EN perform Step 9.
NUMBER
PROCEDURE TITLE
REVISION
13
I-AP-10.05
LOSS OF SEMI-VITAL BUS
PAGE
5 of 11
NUMBER
PROCEDURE TITLE
REVISION
13
l-AP-10.05
LOSS OF SEMI-VITAL BUS
PAGE
7 of 11
STEP
ACTION/EXPECTED RESPONSE "D
I
RESPONSE NOT OBTAINED
10.
GO TO STEP 21
HMTE: Breaker 15 on Unit I and Breaker 15 on Unit 2 Semi-Vital Bus should
be opened before performing the following step.
11.
-DIRECT
THE ELECTRICAL DEPARTMENT
TO SWAP THE GAI-TRONICS POWER
SUPPLY TO UNIT 2 SEMI-VITAL BUS
(JUNCTION BOX IN UNIT 1 ESGR)
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
S
S
S
S
S
S
CAUTION:
If the Semi-Vital Bus has been deenergized for greater than 30 minutes
and a Reactor trip occurs, alternate steam release will be required to
keep the Main Steam Safety valves from lifting.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
-12.
-VERIFY REACTOR - NOT TRIPPED
Maintain RCS temperature as
necessary to prevent lifting Main
Steam safety valves.
available
"* Use Steam Dumps if available
o9
Attachment 3.
LOCAL OPERATION OF
I
NUMBER
l-AP-10.05
ATTACHMENT
3
ATTACHMENT TITLE
REVISION
13
PAGE
1 of 1
//
I,
__ 1.
Consult with Shift Supervisor to determine which SG PORV(s) will
be operated.
__
2.
Send Operator to Safeguards.
3.
Close the isolation valve between the PORV positioner and actuator for
the PORV(s) to be operated.
- I-IA-1635A.
1-MS-RV-1O1A Positioner Isolation Valve
e I-IA-1635B. 1-MS-RV-101B Positioner Isolation Valve
- 1-IA-1635C,
I-MS-RV-101C Positioner Isolation Valve
4.
Open the bottled air supply valve for the PORV(s) to be operated.
- l-IA-1638A, Bottled Air Supply for I-MS-RV-101A
- l-IA-1638B, Bottled Air Supply for 1-MS-RV-IOIB
- l-IA-1638C, Bottled Air Supply for 1-MS-RV-101C
5. Verify 1-IA-PCV-111 is backed-off.
(no spring pressure)
6. Align air bottle to supply the SG PORV(s) by opening one of the
following:
- 1-IA-1639. Air Bottled Manifold Isolation Valve
- 1-IA-1640. Air Bottled Manifold Isolation Valve
NOTE: a The SG PORV(s) will start to open when regulator output pressure is
approximately six psig and will be fully open at approximately 30 psig.
e Close communication must be maintained between the MCR and Safeguards
to control RCS cooldown rate.
e Vent valve I-IA-1643.
SG PORV Rapid Closure Vent Valve. may be opened
as necessary for rapid closure of the SG PORVs.
-
7.
Adjust I-IA-PCV-111.
SG PORV Bottled Air System Pressure Regulator. to
open SG PORV(s) for desired cooldown rate.
QUESTIONS REPORT
for Surry2002
1. 062A2.04 OOI/T2G2/T2G2/LOSS OF BUS/M 3.4/3.1/B/SR0230]/S/RLM
Procedures AP-10.02, AP-10.03, and AP-10.04 (Loss of Vital Bus 11, 111, or IV) direct the
operator to trip the reactor prior to tripping the affected RCP.
Which ONE of the following is the basis for tripping the reactor befo
ipping the RCP?
h
'P
A. To ensure SDM is present when backflow through the a:cred loop is ipnitiated.
at
hr
B. To ensure a cooldown rate is initiated in the a
ted loop.
rut
C. To prevent an unnecessary challenge
cessa
cha le nge
e R
prev
eat generatj
D. To prevent exceeding the line
eat generation rate limit.
res
ent exceeding the line
rat
i
s
un'
Ref: SR EB
216
prev
Surry Lesson Plan N
0.3-LP-5E; AP-1p.02, AP-10.03, AP-10.04
I
rr
1
'o
Time:
I
oints:
1.00
VersW:
0 1 2 3 4 5 6 7 8 9
A
A
I
03 ' AF
T
00'
te im.'
qn0et: CCBCACCDDB
Scramble Range: A - D
T
I
r,
RO Tien
2
SRO Tier:
T262
ry.
I'
Keyword:
LOSS OF BUS
Cog Level:
M 14/3.1
Source-
B
(7a
Exam:
SR02301
Test:
S
Misc:
RLM
Tuesday, November 13, 2001 03:29:10 PM
1
B.
[Describe the components and indications associated with an Uninterruptable Power Supply
(UPS). SOER 83-03, Recommendation 11]
C.
Describe the power sources and loads associated with the Appendix R distribution system.
D.
Describe the power sources and loads associated with the Semi-Vital Bus distribution
system.
E.
[Given a loss of a Vital or Semi-Vital bus, describe the actions taken IAW AP-10.01, 10.02,
10.03, 10.04, and/or 10.05 to address this loss. SOER 83-03, Recommendation 11 and
SOER 81-02, Recommendation 5]
F.
Given a loss of a Vital or Semi-Vital bus, describe the effect on Plant indications and
controls, including actions taken lAW applicable APs to address the loss.
Presentation
Distribute all handouts.
Refer to/display H/T-5. 1, Objectives, and review with trainees.
A.
One-Line Diagram
1.
The purpose of the Vital Bus Distribution System is to supply a stable, reliable
source of power to vital instruments.
It must remain uninterrupted to prevent
spurious shutdowns and guarantee proper action when instruments or controls are
required.
ND-90.3-LP-5
Page 4
Revision I11
f.
The principle plant effects, should vital bus I be lost, are the following:
(1)
Loss of letdown
(2)
Loss of CC to all RCP thermal barriers
(3)
Loss of wide range loop "A" temperature
(4)
Loss of S/G wide range level recorder
(5)
Loss of AFW flow meter to "A" S/G
(6)
Loss of HCV-FW-155A "A" S/G feed water bypass
(7)
Loss of S/G blowdown
(8)
Loss of Channel I NIs (SR, IR and PR)
4.
AP-10.02, Loss of Vital Bus II
Ensure trainees have the latest revision of AP-10.02 to follow for this presentation. Perform
a step-by-step discussion of this procedure highlighting applicable areas.
a.
Initially a determination is made to see if VB 1-1I or VB 1-IIA is lost.
b.
If VB 1-11 is lost, the reactor is tripped and "B" RCP is secured due to loss of
CC to the RCP lube oil coolers. The team should initiate E-0 and continue
with AP- 10.02.
ND-90.3-LP-5
Revision I11
Pagel6
Ensure trainees have the latest revision of AP-10.03 to follow for this presentation. Perform
a step-by-step discussion of this procedure highlighting applicable areas.
-4
a.
Initially a determination is made to see if VB 1-III or VB I-lIIA is lost.
b.
If VB 1-I1 is lost, the reactor is tripped and "A" RCP is secured due to loss
of CC to the RCP lube oil coolers.
The team should initiate E-0 and
continue with AP-10.03.
c.
Actions necessary to stabilize the plant for a loss of VB 1-III are listed in
Attachment 1.
d.
An attempt is made to re-energize the vital bus by pushing the alternate
source to load button on the UPS or using the manual bypass switch.
e.
The team must stop at this point until the vital bus is re-energized. After the
bus is energized, the remainder of the procedure restores affected systems to
pre-event conditions.
f.
The principle plant effects, should vital bus 1-1I1 be lost, are the following:
(1)
Loss of CC to "A" RCP lube oil and stator coolers
(2)
Loss of PR channel III (N-43)
(3)
Loss of all main feed control bypass valve control
(4)
Failure of steam dumps to control Tave to required Tof
ND-90.3-LP-5
Page18
Revision I11
(5)
Loss of all steam generator feed regulating valve control (controllers
are affected).
6.
AP-10.04, Loss of Vital Bus IV
Ensure trainees have the latest revision of AP-10.04 to follow for this presentation. Perform
a step-by-step discussion of this procedure highlighting applicable areas.
a.
Initially a determination is made to see if VB 1-IV or VBI 1-IVA is lost.
b.
If VB 1 -IV is lost, the reactor is tripped and "C" RCP is secured due to loss
of CC to the RCP lube oil coolers.
The team should initiate E-0 and
continue with AP-10.04.
c.
Actions necessary to stabilize the plant for a loss of VB 1-IV are listed in
Attachment 1.
d.
An attempt is made to re-energize the vital bus by pushing the alternate
source to load button on the UPS or using the manual bypass switch.
e.
The team must stop at this point until the vital bus is re-energized. After the
bus is energized, the remainder of the procedure restores affected systems to
pre-event conditions.
f.
The principle plant effects, should vital bus 1-IV be lost, are the following:
(1)
Loss of CC flow to "C" RCP stator and oil coolers
(2)
Loss of PR channel IV (N-44)
(3)
Loss of automatic pressure control of RCS
ND-90.3-LP-5
Page 19
Revision I11
QUESTIONS REPORT
for Surry2002
1. 065AA2.01 001/TI G3/TlG2/PRESSURE SWITCH/C/A 2.9/3.2/N/SR0230 I/S/RLM
"-Unit I is at 100%
-Instrument Air Compressors are in AUTO and are not running.
-Annunciators 1B-E6, IA LO HEADER PRESS/1A COMPR I TRBL and 1B-G5, INST AIR
DRYER TRBL illuminate.
-The RO reports that service air pressure appears steady at approximately 100 psig.
-The AO reports that the Instrument Air Dryer bypass valves have opened and the Instrument
Air Compressors are NOT running.
Which ONE of the following is the reason Annunciator 1B-G5 alarmed?
A. The air dryer bypass trip valves opened.
B. The lag Service Air Compressor auto started.
C.f Failure of the Instrument Air Header pressure switch
D. The Instrument Air Compressor auto start pressure switch failed to actuate.
Ref: Surry Lesson Plans: ND-92. 1 -LP- 1, obj E and ND-95. I -LP-9
Answer A is incorrect because the bypass valve opens as a result of the PS failure and cannot
account for the IB-E6 alarm (ie does not feed that alarm)
Answer B is incorrect because auto start of the lag compressor is not an input to either alarm
(input to lB-E5)
Answer C is correct because both alarms actuate in response to an 80 psig PS (see note)
Answer D is incorrect because instrument air pressure is normal an therefore has not actuated the
90 psig switch that auto starts the IA compressors.
Note: This question is based on the assumption that 1-IA-PS-120 feeds both alarms. This needs
to be verified. Current info insufficient.
Time:
I
Points:
1.00
Version: 0 1 2 3 4 5 6 7 8 9
Answer: CAABDACACC
Scramble Range: A - D
RO Tier:
TIG3
SROTier:
TIG2
Keyword:
PRESSURE SWITCH
Cog Level:
C/A 2.9/3.2
Source:
N
Exam:
SR02301
Test:
S
Misc:
RLM
Wedesay De
Ak
cembr'1
3!r(, Is
1
/
4.Ws
i
r1%
Wednesday, December 05, 2001 01:41:09 PM
LESSON PLAN
Introduction
The Station Air Systems supply compressed air to operate tools, valves and components throughout
the station.
This lesson describes the systems, including flowpaths, system components,
instrumentation and control, and annunciators or alarms associated with these systems.
After receiving this instruction, the trainee will be able to:
A.
Describe the system flowpaths and components associated with the Service Air System.
B.
[Describe the system flowpaths and components associated with the Instrument Air System.
OrfIPR RR-OI Rernmmendtinn ?k31
C.
Describe the flowpaths and components associated with the Polishing Building Air System.
D.
Describe the flowpaths and components associated with the Containment Instrument Air
System.
E.
Describe the flowpaths, components, indications, and controls associated with the
Station Air Systems.
ND-92.1-LP- 1
Revision 12
Page 3
STATION AIR SYSTEM ALARMS
Service Air Compressor 1 Trouble (lB-E5)
Compressor motor overload
High oil temperature - 176F
Low oil pressure - 20 psig (22 sec TD) - byp 15 secs on a start for compr to reach operating
speed.
L.P. stage outlet high air temperature - 425°F
H.P. stage outlet high air temperature - 425°F
High intercooler air temperature - 1907F
Loss of power
Emergency backup running (lag compressor start): If I -SA-C- 1 in LEAD and 2-SA-C- 1
auto starts in LAG - alarm received on Unit 1 (similar on Unit 2).
Instrument Air Low Header Pressure/Compressor 1 Trouble (BE6)
Bearing cooling water outlet temperature high (140 degrees).
Discharge air high temperature (444 degrees).
Low lube oil pressure (5 psig) with a compressor run signal present.
Emergency back-up running.
Failure to continue running after initiation of start signal, due to low lube oil pressure
(8 psig) at end of seven (7) second timing period.
Compressor motor thermal overload.
Loss of power.
Low instrument air header pressure - 80 psig.
Instrument air dryer trouble: (BGS)
Loss of power - causes auto bypass of dryer; dryer continues normal operation until battery
depleted (-4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />), then both chambers go into service.
Chamber performance degrading/LC or RC AMLOC Failure
Valve malf (UI), LC/RC REPRESS/DEPRESS FAILURE, ONLINE PRESSURE (U2)
Low instrument air header bypass
A
md'
t£dca
/4., Zo IA i,;
"
,i°
J,,E
,
(e)
Filtration of the instrument air is provided by replaceable
coalescing prefilter located upstream of the dryers as well as
particulate after-filters positioned downstream. The prefilter
removes liquid aerosols of water and oil and have a
particulate performance rating of 100% of 0.6 microns and
larger. The accumulated liquids are drained from the pre
filter housing through an automatic drain valve. The after
filters have a particulate performance rating of .9 microns
absolute.
3.
Instrumentation and controls
a.
Instrument Air compressors
(1)
The controls are three position switches, hand-off-auto. In HAND,
the compressor motor runs continuously, the compressor loads and
unloads at 100 and 110 psig, respectively. In AUTO, the compressor
starts if pressure reaches 90 psig; load and unload setpoints are the
same.
(2)
Following an auto start, the auto start mercoid switch must be reset
manually; it will not reset itself on high pressure.
(3)
On any start signal, interlocks must be satisfied to start the
compressor as follows:
No motor overload.
Discharge air temperature less than 4440F.
Cooling water temperature less than 1400F.
ND-92. 1 -LP- 1
Revision 12
Page 16
VIRGINIA POWER*
SURRY POWER STATION
ANNUNCIATOR RESPONSE PROCEDURE
NUMBER
PROCEDURE TITLE
REVISION
IB-E6
IA LOW HDER PRESS/IA COMPR I TRBL
9
PAGE
1 of 7
REFERENCES
1. UFSAR - Sections 9.8.2 and 10.3.9.3
2.
11448-ESK-6DA.
10B.
10AE
3. EWRs89-329. 89-547. EWR 89-557B
4. 1-DRP-005. Instrumentation Setpoints
5. DCP-86-03-C. IA DRYER MODIFICATION (Step 10)
PROBABLE CAUSE
1. 1-IA-PS-120 senses IA header pressure less than or
equal to 80 PSIG.
Low header pressure may be caused by one or more of
the following:
"* Compressor failure
"* Line rupture
"* Excessive demand
"* Dryer control malfunction
2. Local annunciator Panel 01-IA-ANN-PNL receives a
trouble signal from one or more of the following:
"* Low oil PRESS less than or equal to 5 PSIG
"* Motor overload
"*
EMERG backup IA compressor running
"* High DISCH air TEMP greater than 444 *F
"* High cooling water outlet TEMP greater than 140*F
3. Instrumentation failure has occurred.
APPROVAL RECOMMDENDDED
DATE
_________I________
____7/
Level dcmn s eT*JT ION
This document should be veig~LMfdetfqtcqoloue
to perform work.
NUMBER
PROCEDURE TITLE
REVISION
INST AIR DRYER TRBL
3
PAGE
1 of 4
REFERENCES
1. UFSAR 9.8
7. PAR 93-0445
2.
11448-PM-075A
8.
DR S-98-1572
3. 11548-FE-18AW
4. il448-ESK-IOB.
10AX
5. DCP-86-03A-3
6. DCP-86-03C-3
PROBABLE CAUSE
1. Alarm actuates when one or more of the following conditions exist:
a. Instrument Air dryer discharge pressure less than or equal to 80 PSIG.
b. Loss of power to dryer.
c. Dryer bed too wet. (Chamber performance degrading)
d. Moisture probe cable disconnected.
e. Exhaust valve malfunction.
f. Inlet valve malfunction.
g. Inlet isolation trip valve (I-IA-TV-125) closed.
h. Bypass trip valve (1-IA-TV-126)
open.
2. Instrumentation failure has occurred.
APPROVAL RECOMMENDED
REVIEWED
{(aA 4..se .J.,-
APPROVED
J
Fom No. 723758(Jon I
CtI
Virqinia Power
DAM
QUESTIONS REPORT
for Surry2002
1. 103A2.05 001
-Unit 2 is making preperations for a reactor startup.
-An RCP low oil level is recjgved.
-An Entry into containment is required to add oil to the RCP.
Which one of the following describes what requirements must be met to allow entry?
A. An SCBA with 19% to 23% oxygen by volume, a confined space entry permit, and
permission of the Station Manager.
B. An SCBA with 19% to 23% oxygen by volume, a VPAP-0106 attachment 1, and permission
of the HP Supervisor.
C. An SCBA with 33% to 37% oxygen by volume, a confined space entry permit, and
permission of the Operations Manager.
D. An SCBA with 33% to 37% oxygen by volume, a VPAP-0106 attachment 1, and permission
of the Site Vice President.
VPAP-0106 Subatmospheric Containment Entry.
Surry Lesson Plan ND-88.4-LP-7 Objective E.
A. Incorrect, SCBA must have 33 to 37 % oxygen by volume, and no confined space entry
permit is required.
B. Incorrect, SCBA must have 33 to 37 % oxygen by volume, and permission cannot be granted
by the HP Supervisor.
C. Incorrect, A confined space entry permit is not required and the ops manager cannot grant
permission.
D. Correct, An SCBA with 33 to 37% oxygen by volume, a VPAP-0106 attachment 1 and
permission of the Site Vice President allows entry.
Time:
I
Points:
1.00
Version:
0 1 2 3 4 5 6 7 8 9
Answer:
Scramble Range: A - D
RO Tier:
T2G3
SRO Tier:
T2G2
Keyword:
Cog Level:
C/A 2.9/3.9
Source:
N
Exam:
SR02301
Test:
S
Misc:
GWL
1
Tuesday, December 11, 2001 09:47:02 AM
VPAP-0106
REvISION 5
POWER
PAGE 9 OF 22
6.0
INSTRUCTIONS
6.1
Subatmospheric Containment Entry Hazards
6.1.1 Subatmospheric containment entry will expose Containment Entry Team members to
four distinct hazards:
"* Ionizing radiation
"* Heat stress
"* Differential pressure
"* Oxygen deficiency due to subatmospheric pressure
6.1.2 Personnel Air-Lock entry and exit may cause personnel discomfort due to the air
pressure changes. Personnel experiencing discomfort during pressure changes should
notify the Containment Entry Team Leader to prevent severe pain and potential
damage to the ear.
6.1.3 Containment entries shall not be made if the containment pressure is less than 9.0 psia.
6.1.4 If containment pressure is greater than or equal to 9.0 psia and less than 12.0 psia,
SCBA with 33 to 37 percent oxygen by volume shall be used.
6.1.5 If required to change SCBA cylinders inside containment, brief exposures to 9.0 psia
to 12.0 psia containment atmosphere, without enriched oxygen breathing gas mixture,
is permissible provided there are no radiological concerns listed on the RWP.
6.1.6 A subatmospheric containment meets the conditions for being a Confined Space (non
permit required) as defined in 29 CFR 1910.146. All of the applicable requirements for
entry into a non permit-required confined space are met or exceeded by this VPAP,
therefore this VPAP shall be used in lieu of the Confined Space Entry Program.
Attachment 1 shall be completed for each containment entry except in cases of
emergencies. Confined Space requirements for equipment within containment apply
as required by the Confined Space Entry Program.
VPAp-0106
REVISION 5
PAGE 19 OF 22
Containment Entry Checklist
i
- *
o
I
S.
__o_"
Unit 1 r-1 Unit 2
Dae
..
ad Time of Entry
Radiation Work Permit (RWP) Number
List personnel desigate for Containment Entry Team
Nlote: A Containment Entry Team minimum composition Is two and maximum composition is fifteen people.
Not: AConainentEnty
Tam
in
Containment Entry
Training Satisfactorily
Name (Please Print)
Signature
Sadge Number
Completed
o
Yes Q No
o Yes I] No
El Yes Q No
[3 Yes Q] No
o Yes El No
Yes Q No
o Yes El No
o Yes Q No
El Yes fl
No
3 Yes [] No
o
Yes El No
El Yes Q No
Yes Q No
-] Yes E] No
El Yes C No
Containment Entry Team Leader (Name - Please Print)
Permission granted by Site Vice President or Station Manager (Name- Please Print)
If any Containment Entry Team Member is not Trained, List Reason Why and Designate Escort
r
"n for Entry and Work to be Performed
Responsible Supervisor (Signature)
Date
MA fINl
ow'"!
EXAMINATION ANSWER KEY
RO/SRO Exam Bank
ID: ADM0109
Points: 1.00
Which ONE of the following individuals by title is the MINIMUM authorization that must be
obtained before containment entry during subatmospheric conditions?
A.
Shift Supervisor
B.
Site Vice President
C.
Immediate Supervisor
D.
HP Supervisor
Answer:
B
Question 64 Details
Question Type:
Topic:
System ID:
User ID:
Status:
Must Appear:
Difficulty:
Time to Complete:
Point Value:
Cross Reference:
User Text:
User Number 1:
User Number 2:
Comment:
Multiple Choice
ADM0109
72335
ADMO109
Active
No
0.00
0
1.00
1.00
0.00
0.00
ND-88.4-LP-7E; SROUTP-SDS-2/P; VPAP-0106
[S99-0136], [S97-0830], [S95-0039]
OPS RO/SRO SU
64
Page: 58 of 3141
10/19/01
QUESTIONS REPORT
for Surry2002
1. G2.1.4 001 /T3/T3/STAFFING/M 2.5/3.3/B/SR02301/S/RLM
The following plant conditions exist:
-Unit 1 is in HOT SHUTDOWN.
-Unit 2 is in COLD SHUTDOWN.
Which ONE of the following is the MINIMUM Shift Manning requirement for the Station under
the conditions shown above per Tech Spec 6.1, Table 6-1 -1, "Minimum Shift Crew
Composition"?
SS¢
A.
1
1
2
4
0
B!
1
1
3
4
1
C.
1
0
3
3
1
D.
1
0
2
4
1
Ref: SR EB # TS00126
ND-88.1-LP-9, obj. F
Tech Spec 6.1, Table 6-1-1, "Minimum Shift Crew Composition"
RO Tier:
T3
SRO Tier:
T3
Keyword:
STAFFING
Cog Level:
M 2.5/
Source:
B
Exam:
SR023
Test:
S
Misc:
RLM
Test Name
<Cumulative>
Test Date
rpb
p(Diff) Ti
0.000
0.000 0
3.3
01
me
Equ
User Values
N
1:0
3:0
---B --
p
Resp
%
--- C --
p
Resp
%
--- D _-
p
Resp
%
p
Resp
%
<Cumulative>
0
-1
0.00o
Total:
0
-1
0.00
0 100
0
-1
0.001
Omits:
0
-1
0.00
1
Thursday, December 13, 2001 08:39:19 AM
--- A--
Resp
%
2:0
4:0
P
0
0
LESS1NL*PAN
lntrdulction
1OCFR50.36 requires applicants for a nuclear facility operating license to submit and comply with
Technical Specifications.
These specifications are derived from the analyses and evolutions
included in the Final Safety Analysis Reports. Since these Tech Specs are required by law and are
approved by the NRC, they are a legal document. This lesson will provide a general description of
each section of Tech Specs and a detailed description of the RCS Tech Specs. The knowledge
gained from this lesson will provide an understanding of the content and layout of the Technical
Specifications.
Obredtues
After receiving this instruction, the trainee will be able to:
A.
Summarize the purpose of Tech Spec Section 1.0 including the definition of applicable
terms in this section.
B.
Summarize the purpose of Tech Spec Section 2.0.
C.
Summarize the purpose of Tech Spec Section 3.0.
D.
Summarize the purpose of Tech Spec Section 4.0.
E.
Summarize the purpose of Tech Spec Section 5.0.
F.
Summarize the purpose of Tech Spec Section 6.0.
G.
Describe the purpose and specification for the Safety Limits lAW section 2 of Tech Specs
including for SRO candidates, the basis behind these specifications.
ND-n8
I-Lp-c
Pave 2
Revision 10
, *1 IL*" UU*I
]k*lt
J
TS 6.1-"
8-1-88
TABL]. 6. 1-1
MNIKEtM SHIFT CREW COMPOSITION
POSITION
NmmtR OF IrDIVIDUALS REQUIRED TO FILL POSITION
ONE zmIT
TW WITS
TWO UNITS IN COLD
OPERATING
OPERATING
SHUTDOWN OR REFUELINC
1
1
1
1
1
None
3
3
2
4
4
4
1
1
None
Amendment Nos. 12 3
and
TS 6.1-5
5-29-8:
TABLT
6.1-1 (Continued)
-
Shift Supervisor with a Senior Reactor Operators License.
SR0
-
lidividunl with a Senior Reactor Operators License.
-
Individual with a Reactor Operators License.
Auxiliary Operator
Except for the Shift Supervisor, the Shift Crew Composition may be one less
than the minimum requirements of Table 6.1-1 for a period of time not to
exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accomodate unexpected absence of on-duty shift
crew members provided immediate
action is taken to restore the Shift Crew
Composition to within the minimum requirements of Table 6.1-1.
This provision
does not permit any shift crew position to be unmanned upon shift change due
to an oncoming shift crewman being late or absent.
During any absence of the Shift Supervisor from the Control Room while the
unit is in operation, an individual (other than the Shift Technical Advisor)
with a valid SRO license shall be designated to ashame the Control Room
command function.
During any absence of the Shift Supervisor from the Control
Room while the unit is shutdown or refueling, an individual with a valid RO
license (other than the Shift Technical Advisor) shall be designated to assume
the Control Room comand functions.
Amendments No.
69 & 69
QUESTIONS REPORT
for Surry2002
1. G2.1.34 001/T3/T3/PRIMARY CHEMISTRY/M 2.3/2.9/B/SR02301/S/RLM
The following plant conditions exist:
'
"- The reactor has been at 100% power for 30 days.
- Chemistry reports that RCS activity is 1.5pCi/cc DOSE
EQUIVALENT 1-131.
WHICH ONE (1) of the following actions will reduce RCS activity?
A. Vent the Volume Control Tank (VCT) to the Waste Gas System.
B. Place cation demineralizer in service and maximize letdown.
C. Maximize letdown through the mixed bed demineralizer.
D. Reduce letdown to minimum and establish an RCS ph between 6.5 and
7.5 by chemical injection.
Ref: SM EB # 41617
Note: Need procedure that specifies the actions that the operating crew would take.
Time:
I
Points:
1.00
Version:
0 1 23456789
Answer: CBDDCAADAD
Scramble Range: A - D
RO Tier:
T3
SRO Tier:
T3
Keyword:
PRIMARY CHEMISTRY
Cog Level:
M 2.3/2.9
Source:
B
Exam:
SR02301
Test:
S
Misc:
RLM
Q<
'1'
MY
, .5
1
Thursday, November 15, 2001 09:52:27 AM
Untitled
- QNUM
41617
- HNUM
42864 (Do NOT change If< 9,000,000)
- ANUM
41627
- QCHANGED
FALSE
- ACHANGED
FALSE
- QDATE
1992/12/07
- FAC
395
V. C. Summer 1
- RTYP
PWR-WEC3
- EXLEVEL
S
- EXMNR
- QVAL
- SEC
- SUBSORT
- KA
000076K305
- QUESTION
The following plant conditions exist:
-
The reactor has been at 100% power for 30 days.
-
Chemistry reports that RCS activity is 1.5 microCuries per gram DOSE
EQUIVALENT 1- 131.
WHICH ONE (1) of the following actions will reduce RCS activity?
a.
Vent the Volume Control Tank (VCT) to the Waste Gas System.
b.
Place cation demineralizer in service and maximize letdown.
c.
Maximize letdown through the mixed bed demineralizer.
d.
Reduce letdown to minimum and establish an RCS ph between 6.5 and
7.5 by chemical injection.
- ANSWER
c. (+1.0)
- REFERENCE
1. VCS: GS-6, Primary Chemistry and Sampling, Objective. 6, 9, 10, p.2 2 , 29,
30
2.
VCS: CR-2, Plant Chemistry Control, Objectives, 7, 12, p. 17
- 3. VCS: Technical Specifications 3.4.8
4.
KA 000076K305 (2.9/3.6)
Page 1
LESSON PLAN
Introduction
Primary chemistry limits provide for the safety and health of the public if an accident should occur
at the plant. They also ensure the integrity of primary materials by minimizing corrosion. Since
the licensed operator is responsible for plant performance, he/she should be able to recognize these
limits and realize what processes these specifications are limiting. Fuel integrity is a major concern
to the licensed operator. He/she should be able to use primary isotopic concentrations to determine
if a change in fuel integrity has occurred. Operators are responsible for primary chemical additions
and should know the purpose of these additions. Besides the chemistry limits themselves, this
lesson plan will also include a discussion of the rationale behind them.
Objectives
After receiving this instruction, the trainee will be able to:
A.
Explain the primary coolant reactions and chemical controls.
B.
Describe the Tech Spec limits for RCS chemistry control.
C.
Explain Surry Power Stations's technical specification and recommended primary
chemistry limits.
Revision 9
ND-81.I-LP-.
ag
QUESTIONS REPORT
for Surry2002
1. G2.2.6 001/T3/T3/PROCEDURE CHANGE/M 2.3/3.3/N/SR0230 1/SIRLM
Which ONE of the following are required approval authorities for a change to a telephone area
code in EPIP-2.01, NOTIFICATION OF STATE AND LOCAL GOVERENMENTS?
A. The Director of Nuclear Security and Emergency Preparedness and the Shift Supervisor
B. The Shift Supervisor and the Site Vice President
C. The Director of Nuclear Security and Emergency Preparedness and the SNSOC
D. The SNSOC and the Site Vice President
Ref: VPAP-0502, Procedure Process Control, p.8 5&8 6
No lesson plan or learning objective.
Time:
I
Points:
1.00
Version:
0 1 2 3 4 5 6 7 8 9
Answer: DDBAABDDCB
Scramble Range: A - D
RO Tier:
T3
SRO Tier:
T3
Keyword:
PROCEDURE CHANGE
Cog Level:
M 2.3/3.3
Source:
N
Exam:
SR02301
Test:
S
Misc:
RLM
ARA
Ct.,t)
dc°
XI),Y
T uesday, Decembetrltl U0', 2U.
aI
,',:J*
ll
-I" .....
DOMINION
VPAP-0502
REVISION 21
PAGE 85 OF 123
ATTACHMENT 1
(Page 3 of 4)
Procedure Process Flow Chart
Requestor
Cognizant
Management
Safety
Activity
Screening
Cognizant
Management
DOMINION
VPAP-0502
REVISION 21
PAGE 86 OF 123
A¶TACHMENT 1
(Page 4 of 4)
Procedure Process Flow Chart
QUESTIONS REPORT
for Surry2002
1. G2.3.10 001FT3Fr3/TEMP SHIELDING/M 2.1/3. I/B/SR02301/S/RLM
Which ONE of the following describes the Shift Supervisors responsibilities lAW
Temporary Shielding, concerning the Temporary Shielding request form?
VPAP-2105,
A. The Shift Supervisor must acknowledge installation and removal of temporary shielding.
B. The Shift Supervisor must acknowledge installation and approve removal of temporary
shielding.
C. The Shift Supervisor must approve installation and removal of temporary shielding.
Df The Shift Supervisor must approve installation and acknowledge the removal of temporary
shielding.
Ref: SR EB # ADMO174
VPAP-2105
No specific learning objective found
Note: Reference document not included in materials received.
RO Tier:
T3
SRO Tier:
T3
Keyword:
TEMP SHIELDING
Cog Level:
M 2.1/3.1
Source:
B
Exam:
SR02301
Test:
S
Misc:
RLM
(A'
t
N
U'
1
Thursday, December 13, 2001 03:09:18 PM
QUESTIONS REPORT
/1
for Surry2002
1. WE02G2.4.6 001
-Unit 1 has had a Reactor Trip and SI.
(k
-Subcooling on CETC's 450 F.
-RCS pressure is stable at 1530 psig.
-Containment Pressure is 6psig.
-Pressurizer level is 100 %.
-AFW flows (to each S/G): 120gpm, 120 gpm, 100 gpm.
Steam Generator narrow range levels: 11%, 15%, 8%.
Which one of the following is the appropriate status concerning SI Termination Criteria?
A. SI Termination Criteria will be met if AFW flow is adjusted to greater than 350 gpm.
B. SI Termination Criteria are NOT met due to RCS subcooling, continue ECCS pumps
running.
C. SI Termination Criteria is met, Transition should be made to ES- 1.1 "SI Termination".
D. SI Termination Criteria is NOT met; due to Pressur 7 zer level being high ECCS pumps
should be reduced.
Surry Lesson Plan; ND-95.3-LP8 objective A.
A. Incorrect, SI termination criteria would still not be met.
B. Correct, SI termination criteria is not met. ECCS should remain running.
C. Incorrect, SI termination criteria is not met.
D. Incorrect, SI termination criteria is not met, but ECCS should remain running even though
Pressurizer level is high.
Time:
I
Points:
1.00
Version:
0 I 2 3 4 5 6 7 8 9
Answer: BDCAADADDA
Scramble Range: A - D
RO Tier:
TIG2
SRO Tier:
TIGI
Keyword:
Cog Level:
C/A 4.0/4.0
Source:
M
Exam:
SR02301
Test:
S
Misc:
GWL
Wednesday, December 12, 2001 09:01:06 AM
1
Ilntrodution
This lesson plan will provide classroom training for ES-l.1, SI Termination. The material will be
presented first as an overall "big picture" of the procedure which will then be followed up by an in
depth presentation of the step backgrounds and required knowledges of the procedure. Shortly after
the classroom presentation, the simulator will be used to reinforce this material and allow practice
of the techniques incorporated into ES- 1.1.
In its entirety, ES-1.1 provides the necessary instructions to terminate SI and stabilize plant
conditions.
It is entered from E-0, Reactor Trip or Safety Injection, from E-l, Loss of Reactor or Secondary
Coolant, or FR-H. 1, Loss of secondary Heat Sink, when the SI termination criteria are satisfied.
The goal of ES-I .1 is to stop SI pumps in a prescribed sequence while maintaining control of the
RCS, until makeup is by charging flow alone. Following termination of SI, the operator will exit to
normal procedures for either startup or cooldown.
After receiving this instruction, the trainee will be able to:
A.
Given the major action categories associated with ES-l.i, Si Termination, explain the
purpose of ES-1.1, the transition criteria for entering and exiting ES-1.1 and the types of
operator actions that will occur within each category.
B.
Given a copy of ES- 1. 1, SI Termination, explain the basis of each step of the procedure.
ND-95 3-L P-8
Pave 3
Revision 11
- QNUM
43875
- HNUM
45211
(Do NOT change If < 9,000,000)
- ANUM
- QCHANGED
FALSE
- ACHANGED
FALSE
- QDATE
1995/04/17
- FAC
456
Braidwood 1 & 2
- RTYP
PWR-WEC4
- EXLEVEL
S
- EXMNR
- OVAL
- SEC
- SUBSORT
- KA
000009A204
- QUESTION
The following conditions exist on Unit 1 following a reactor trip and
SI:
- Wide range RCS pressure is 1375 psig and stable.
- Average of 10 highest core-exit TCs is 565 degrees F.
- The Subcooled Margin Monitor (SMM) Iconics display indicates a
red 17.
- Pressurizer level is 10% and increasing.
- Containment radiation is 2 Rem/hr
- Containment pressure is 4 psig.
- SG narrow range levels: 2%, 8%, 5%, 2%.
- AFW flows (to each SG): 125 gpm, 125 gpm, 125 gpm, 100 gpm.
The Unit Supervisor is trying to determine if ECCS flow should be
reduced per step 6 of BwEP-1 "Loss of Reactor or Secondary Coolant".
WHICH ONE of the following is the appropriate status concerning SI
termination criteria?
(Attached Figure 1BwEP ES 1.1-1 may be used for reference.)
a. SI termination criteria are met and transition should be made to
b. SI termination criteria are NOT met due to pressurizer level
being low, continue ECCS pumps running.
c. SI termination criteria are NOT met due to RCS subLocling margin
being low, continue ECCS pumps running.
d. SI termination criteria are met if total AFW flow is adjusted to
500 gpm.
- ANSWER
[-
E
ACTION/EXPECTED RESPONSE
RESPONSE NOT OBTAINED
- 6.
__CHECK IF SI FLOW SHOULD BE REDUCED:
a) RCS subcooling based on CETCs
GREATER THAN 300F [85-F]
b) Secondary heat sink:
"* Total feed flow to INTACT SGQ
- GREATER THAN 350 GPM
[450 GPM]
"* Narrow range level in at
least one intact SG - GREATER
THAN 11% [22%]
INCREASING
d)
PRZR level - GREATER THAN 22%
[43%]
e) GO TO 1-ES-1.1. SI TERMINATION
- 7. -CHECK
IF HI HI CLS INITIATED:
a) GO TO Step 7.
b) GO TO Step 7.
c)
GO TO Step 7.
d) Try to stabilize RCS pressure
with normal PRZR spray.
GO TO
Step 7.
GO TO Step 13.
- RS pump(s)
- RUNNING
- Any Hi Hi CLS annunciator - LIT
NUMBER
PROCEDURE TITLE
REVISION
17
I-E-1
LOSS OF REACTOR OR SECONDARY COOLANT
PAGE
5 of 27
12
QUESTIONS REPORT
for Surry2002
1. G2.4.14 001
While in the Emergency Respose procedures the team is directed to "Go To" another procedure.
Which one of the following is correct way to implement this direction?
A. The "GO TO" implies that the procedure in use is no longer applicable, and any tasks that
were in progress need not be completed.
B. Tasks still in progress must be completed prior to the transition directed by the "GO TO"
step.
C. The "GO TO" implies that the procedure in use is no longer applicable, but any tasks that
were in progress and should completed.
D. Tasks still in progress need not be completed prior to the transition directed by the "GO TO"
step, unless preceeded by a bullet.
Surry Lesson Plan ND-95.3-LP-2 objectives # Dand F.
A. Incorrect, The tasks should be completed.
B. Incorrect, Tasks in progrees do not have to be completed prior to the transition.
C. Correct, The previous procedure is nolonger applicable and the tasks that were in progress
should be completed.
D. Incorrect. Tasks in progress need not be completed prior to the transition, a bulleted step can
be performed in any order, and does not have to be performed prior to transition.
Time:
I
Points:
1.00
Version:
0 I 2345 67 8 9
Answer:
Scramble Range: A - D
RO Tier:
T3
SRO Tier:
T3
Keyword:
Cog Level:
M 3.3/3.9
Source:
N
Exam:
SR02301
Test:
C
Misc:
GWL
Tuesday, November 20, 2001 10:13:56 AM
B.
Explain the two-column format of the Emergency Response Guideline Procedures,
including the placement criteria for cautions and notes.
C.
Explain the method by which "Immediate Operator Action" steps are identified in the body
of the ERG Procedures.
D.
Describe the intended overall usage of the Emergency Response Guidelines Network.
E.
Given various plant conditions during which an emergency event occurs, evaluate the
application of the "Modes of Applicability" as described in the ERG User's Guide.
F.
Given actual or simulated EOP implementation, apply the management standards and other
good practices applicable to EOP usage.
G.
Explain the format design of the Emergency Response Guideline Procedures.
presentAtin
Distribute all handouts.
Refer to/display ll/T-2.1, Objectives, and review objectives with trainees
A.
Action Verb Identification
Direct trainees to turn to AIA-2. 1, Action Verbs. Review various action verbs with trainees.
ND-95.3-LP-2
Revision 7
Page 4
b.
If a particular task MI 1ST BE COMPI
TPED prior to proceeding, the step
containing the task or an associated NOTE will explicitly state that re
quirement.
11.
Transitions to other procedures or to different steps in the same guideline may be
made from either column. Such transitions should be made realizing that preceding
NOTES or CAUTIONS are applicable.
a.
Any tasks still inprogress need not be completed prior to making a
transition; however, the reqnirement to complet the tasks is still present and
must not be neglected.
b.
A transitional "GO TO..." to some other procedure implies that the
procedure in use is now no longer applicable and the procedure referred to is
now in effect.
ND-95.3-LP-2
Page 10
Revision 7
EXAMINATION ANSWER KEY
RO/SRO Exam Bank
ID: EOP0088
Points: 1.00
Which ONE of the following indicates when substeps of an Emergency Operating Procedure must
be performed in order?
A.
Substeps designated by numbers only.x
B.
Substeps designated by bullets.i/
C.
Substeps designated by asterisks/
D.
Substeps designated by letters or numbers. /
Answer:
D
Question 839 Details
Question Type:
Topic:
System ID:
User ID:
Status:
Must Appear:
Difficulty:
Time to Complete:
Point Value:
Cross Reference:
User Text:
User Number 1:
User Number 2:
Comment:
Multiple Choice
EOP0088
73378
EOP0088
Active
No
0.00
0
1.00
1.00
0.00
0.00
ND-95.3-LP-2B; OPAP-0002
[S95-1096]
SoestE'
~
~
teA
0~
3 A W
e1'
r oiu'1 eec
MOST 6-rQM'-Ar
t &efl-
Timmqe Ac~~SQ&tMs
W'te
QAqk-Oi&Vt\\~~'~~
Ci
OPS RO/SRO SU
Page: 792 of 3141
10119/01
839
Ij Ad.
',4ý E
EXAMINATION ANSWER KEY
RO/SRO Exam Bank
Question 813 Details
Question Type:
Topic:
System ID:
User ID:
Status:
Must Appear:
Difficulty:
Time to Complete:
Point Value:
Cross Reference:
User Text:
User Number 1:
User Number 2:
Comment:
Multiple Choice
EOP0020
73317
EOP0020
Active
No
0.00
0
1.00
1.00
0.00
0.00
ND-95.3-LP-38C and D; ND-95.4-LP-3A, B, and D; FR
C.1
[S97-0047], [S96-1021], [S96-1350]
ID: EOP0022
Points: 1.00
Which ONE of the following action steps must be performed in sequence in accordance with the
rules for Emergency Operating Procedure (EOP) usage?
A.
All immediate action steps of E-0, Reactor Trip or Safety Injection, and FR-S.1,
Response to Nuclear Power Generation/ATWS.
B.
All immediate action steps for ECA-0.0, Loss of All AC Power, and FR-S.1,
Response to Nuclear Power Generation/ATWS.
C.
All immediate action steps for E-0, Reactor Trip or Safety Injection and ES-0.1,
Reactor Trip Response.
D.
All immediate action steps of E-0, Reactor Trip or Safety
Loss of All AC Power.
Answer:
Injection, and ECA-0.0,
A
OPS RO/SRO SU
Page: 767 of 3141
10/19/01
814
QUESTIONS REPORT
for Surry2002
1. WE03G2.4.6 001
-Unit I has experienced a SBLOCA.
-ES-1.2, Post LOCA Cooldown and Depressurization is in progress.
-Three RCPs are running.
-An RCS cooldown to place RHR on service has been initiated by dumping steam to the
atmosphere.
Which one of the following describes the optimum RCP configuration, and the basis for this
configuration?
A. One RCP should be secured to produce effective heat transfer, provide boron mixing for
RHR operations, and provide RCS pressure control.
B. All RCPs should be stopped to minimize RCS inventory loss when the break uncovers.
C. Two RCPs should be secured to minimize RCS heat input, and still produce effective heat
transfer and RCS pressure control.
D. Three RCPs should be left running to ensure symeteric heat transfer to the S/Gs, to aid in
RCS pressure control, and prevent steam voiding in the Reactor vessel head.
Surry Exam Bank, question # 3384 slightly modified.
Surry Lesson Plan ND-95.3-LP-9 Objective B.
A. Incorrect, Two RCPs should be secured. Mixing for placing RHR on service is not a reason
for running RCPs.
B. Incorrect, The procedure directs the operator to leave an RCP operation if possible.
C. Correct only one pump should be left running to minimize heat input, control RCS pressure
and provide effective heat transfer.
D. Incorrect, The procedure directs only one RCP to be left running.
Time:
I
Points:
1.00
Version:
0 1 23456789
Answer:
Scramble Range: A - D
RO Tier:
TIG2
SRO Tier:
TIG2
Keyword:
Cog Level: C/A 4.0/4.3
Source:
B
Exam:
SR02301
Test:
S
Misc:
GWL
Friday, November 30, 2001 07:40:05 AM
ND-95.3-H/T-9.1
OBJECTIVES
After receiving this instruction, the trainee will be able to:
A.
Given the major action categories associated with ES- 1.2, Post-LOCA
Cooldown and Depressurization, explain the purpose of ES-1.2, the transition
criteria for entering and exiting ES-1.2 and the types of operator actions that
will occur within each category.
B.
Given a copy of ES- 1.2, Post-LOCA Cooldown and Depressurization, explain
the basis of each procedural step.
C.
Given actual or simulated plant conditions requiring ES-1.2, Post-LOCA
Cooldown and Depressurization, implementation, successfully transition
through the procedure, applying step background knowledge as
required, to safely bring the plant to a cold shutdown condition.
e.
The value of pressurizer level chosen for this step is that indication with
water level just above the top of the heaters, including allowance for normal
channel accuracy and reference leg heating. This value is used to verify that
sufficient liquid is present to allow operation of the pzr heaters.
f.
It is not critical to maintain level at 35% [55%]. In many cases, the
level (and pressure) will increase after the depressurization is stopped
until injection flow balances break flow and loss due to cooldown
shrink. (rk)
18.
STEPJ13: CHECK IF AN RCP SHOULD BE STARTED.
a.
The purpose of this step is to establish forced circulation flow in the RCS
from one RCP.
b.
Forced flow is the preferred mode of operation to allow for normal RCS
cooldown and provide pzr spray.
c.
If RCPs had not been tripped, all but one are stopped to minimize heat input
to the RCS.
(1)
The RCP started or left running should be the one that can provide
normal pzr spray.
(2)
The normal spray valve associated with any stopped RCP should be
closed.
This maximizes spray flow from the active loop by
preventing backflow through the spray lines of inactive loops.
d.
With no RCP running, depressurization of the RCS may generate a steam
bubble in the upper head. This bubble could rapidly condense during pump
startup, drawing liquid from the pzr and reducing RCS subcooling. If pzr
ND-95.3-LP-9
Revision 8
Page 21
EXAMINATION ANSWER KEY
RO/SRO Exam Bank
ID: EOP0338
Points: 1.00
Given the following plant conditions:
A SBLOCA has occurred.
"* The team is in ES-1.2, Post-LOCA Cooldown and Depressurization.
"* An RCS cooldown has been initiated by dumping steam to the atmosphere.
Which ONE of
configuration?
the following describes the optimum RCP configuration, and the basis for this
A.
One RCP should be run to produce effective heat transfer and RCS pressure
control, yet minimize RCS heat input.
B.
All RCPs should be stopped to minimize RCS inventory loss when the break
uncovers.
C.
Two RCPs should be run to ensure symetric heat transfer to the S/Gs, to
enhance RCS pressure control, and to prevent steam voiding in the vessel head
during RCS depressurization.
D.
One RCP should be run to produce effective heat transer and RCS pressure
control, yet minimize RCS inventory loss.
Answer:
A
Question 3384 Details
Question Type:
Topic:
System ID:
User ID:
Status:
Must Appear:
Difficulty:
Time to Complete:
Point Value:
Cross Reference:
User Text:
User Number 1:
User Number 2:
Comment:
Multiple Choice
EOP0338
107196
EOP0338
Active
No
0.00
0
1.00
0.00
0.00
ND-95.3-LP-9/B
[S00-0306]
OPS RO/SRO SU
Page: 31 13 Ot 3141
I IJ' I tU I
3384
10/19/01
OPS RO/SRO SU
Page: 3113 of 3141
EXAMINATION ANSWER KEY
RO/SRO Exam Bank
Question 1073 Details
Question Type:
Topic:
System ID:
User ID:
Status:
Must Appear:
Difficulty:
Time to Complete:
Point Value:
Cross Reference:
User Text:
User Number 1:
User Number 2:
Comment:
Multiple Choice
EOP0332 (provide CSFSTs)
73615
EOP0332
Active
No
0.00
0
1.00
1.00
0.00
0.00
ND-95.3-LP-26F; ND-95.3-LP-48C
[S99-0176]
ID: EOP0333
Points: 1.00
Unit 1 has experienced a loss of coolant accident and the team is presently in 1-ES-1.2, Post
LOCA Cooldown and Depressurization. The team has initiated a cooldown and is depressurizing
the RCS to refill the Pressurizer. Subcooling is lost during the depressurization.
Which ONE of the following identifies the method used to regain subcooling?
A.
Deenergize the Pressurizer heaters.
B.
Energize the Pressurizer heaters.
C.
Continue the RCS cooldown.
D.
Stop the RCS cooldown.
Answer:
C
10/19/01
OPS RO/SRO SU
1074
Page: 1022 of 3141
2<
QUESTIONS REPORT
for Surry2002
1. WE04G2.4.5 001/T I G2/T1G I /PROCEDURE USAG/C/A 2.9/3.6/B/SR02301/S/RLM
The following conditions exist:
-A manual Rx trip was initiated 10 minutes ago based on AP-16.00 criteria
-Pressurizer level is off-scale low
-Pressurizer pressure is 1500 psig and decreasing
-All SG levels are 5% NR and slowly increasing
-All SG pressures are 1005 psig
-All main steam line radiation monitors are reading .02 mr/hr
-Vent-Vent radiation monitor is reading 4.3 E6 cpm
-Containment pressure is 9.2 psia
-Containment sump level is 47%
-Safeguards Area Sump high level alarm is locked in
Upon exiting E-0, which ONE of the following is the correct procedure transitions for the event
in progress, if the leak is unisolable?
Af ECA-1.2 (LOCA Outside Containment), ECA-1.1, (Loss of Emergency Coolant
Recirculation)
B. E-1, ECA-1.1 (Loss of Emergency Coolant Recirculation), ECA-1.2 (LOCA Outside
Containment)
C. ECA- 1.1 (Loss of Emergency Coolant Recirculation), ECA- 1.2 (LOCA Outside
Containment)
D. E- 1, ECA- 1.2 (LOCA Outside Containment), ECA- 1.1 (Loss of Emergency Coolant
Recirculation)
Ref: SR EB # EOP0263
Surry lesson plans ND-95.3-LP-20, obj A&C; ND-95.3-LP-21, obj A&C;
ND-95.4-LP-12, obj A&C
RO Tier:
TIG2
SRO Tier:
TIGI
Keyword:
PROCEDURE USAG
Cog Level: C/A 2.9/3.6
Source:
B
Exam:
SR02301
Test:
S
Misc:
RLM
Wednesday, December 12, 2001 04:03:45 PM
1
Objectives
After receiving this instruction, the trainee will be able to:
A.
Given the major action categories associated with ECA-1.1, Loss of Emergency Coolant
Recirculation, explain the purpose of ECA-1.1, the transition criteria for entering and
exiting ECA-1. 1, and the types of operator actions that will occur within each category.
B.
Given a copy of ECA-1.1, Loss of Emergency Coolant Recirculation, explain the basis of
each step of the procedure.
C.
Given actual or simulated plant conditions requiring implementation of ECA-1.1,
Loss of Emergency Coolant Recirculation, successfully transition through the
procedure, applying step background knowledge as required, to safely place the plant
in the required optimal recovery condition.
Presentation
Distribute all handouts.
Refer to/display H/T-20.1, Objectives, and review objectives with trainees.
A.
Major Actions of ECA-1.1, Loss of Emergency Coolant Recirculation
1.
Purpose
To provide guidance to restore emergency coolant recirculation capability, to delay
RWST depletion by adding makeup and reducing outflow, and to depressurize the
RCS to minimize break flow.
ND-95.3-LP-20
Revision 9
Page 4
LESSON PLAN
Introduction
The Loss Of Coolant Accident, in itself, is a serious plant accident. However, the level of severity
can be compounded by the fact that the LOCA is outside of the FINAL fission product barrier
Containment. Now, there is no protective shield enveloping the spilled reactor coolant water and
fission products carried out of the RCS. This type of accident poses both a serious threat to the
post-accident cooling capability of the plant and a potential hazard to the general public in the form
of radioactive releases.
This lesson on the Emergency Response Guideline for LOCA Outside Containment is designed to
provide an introduction to the accident and an in-depth analysis of the procedure associated with
combatting this event.
Obj'ectives
After receiving this instruction, the trainee will be able to:
A.
Given the major action categories associated with ECA-1.2, LOCA Outside Containment,
explain the purpose of ECA-1.2, the transition criteria for entering and exiting ECA-1.2,
and the types of operator actions that will occur within each category.
B.
Given a copy of ECA-1.2, LOCA Outside Containment, explain the basis of each step of
the procedure.
C.
Given actual or simulated plant conditions requiring implementation of ECA-1.2
4 '..
LOCA Outside Containment, successfully transition through the procedure, applying
step background knowledge as required, to address the challenge to plant and public
safety.
"NTD-O9
I- P*21
Paoe 2
Revision 7
l.]l*l--/d.J-l*l
I.,1
LESSON PLAN
Introduction
The Reactor Safety Study, WASH-1400, identified Event V Sequences (Interfacing System
LOCAs) as a significant contributor to the risk of core melt and high activity release. Some recent
events have highlighted the need for greater attention to this potentially disastrous Loss of Coolant
Event.
This lesson plan will outline some of the concerns and methods of mitigating the
probability of an Event V Sequence.
Objectives
After receiving this instruction, the trainee will be able to:
A.
Describe an Interfacing System LOCA.
B.
Describe the possible means of limiting the probability and consequences of an Interfacing
System LOCA.
C.
Describe the significance of the EVENT V Sequence.
Presentation
Distribute all handouts and copies of all the AlAs. Refer to/display H/T-12.1, Objectives,
and review with trainees.
A.
Interfacing LOCAs
1.
Event V Sequence
Revision 1
ND-95.4-LP-12
Page 3
EXAMINATION ANSWER KEY
RO/SRO Exam Bank
Question 1007 Details
Question Type:
Topic:
System ID:
User ID:
Status:
Must Appear:
Difficulty:
Time to Complete:
Point Value:
Cross Reference:
User Text:
User Number 1:
User Number 2:
Comment:
Multiple Choice
EOP0262
73547
EOP0262
Active
No
0.00
0
1.00
1.00
0.00
0.00
ND-95.3-LP-21B; ND-95.4-LP-12B
[S96-1030], [$96-1341]
ID: EOP0263
The following conditions exist:
A manual Rx trip was initiated 10 minutes ago based on AP-16.00 criteria
Pressurizer level is off-scale low
Pressurizer pressure is 1500 psig and decreasing
All SG levels are 5% NR and slowly increasing
All SG pressures are 1005 psig
All main steam line radiation monitors are reading .02 mr/hr
Vent-Vent radiation monitor is reading 4.3 E6 cpm
Containment pressure is 9.2 psia
Containment sump level is 47%
Safeguards Area Sump high level alarm is locked in
Which ONE of the following is the correct procedure transitions for the event in progress if the
leak is unisolable?
A.
E-0, E-1, ECA-1.2 (LOCA Outside Containment), ECA-1. 1 (Loss of Emergency
Coolant Recirculation)
B.
E-0, ECA-1. 1 (Loss of Emergency Coolant Recirculation), ECA-1.2 (LOCA
Outside Containment)
C.
E-0, E-1, ECA-1.1 (Loss of Emergency Coolant Recirculation), ECA-1.2 (LOCA
Outside Containment)
D.
E-0, ECA-1.2 (LOCA Outside Containment), ECA-I.1, (Loss of Emergency
Coolant Recirculation)
Answer:
D
Page: 957 of 3141
1008
Points: 1.00
10O/19/01
OPS RO/SRO SU
QUESTIONS REPORT
for Surry2002
1. WE05EA2.1 001
-Unit 1 has had a loss of Both Feedwater Pumps.
-SG 1o-lo level alarms come in and the Reactor fails to trip.
-Actions of S. 1 " Response to Nuclear Power Generation / ATWS are performed.
-Reactor Power is < 5%, with a negitive start up rate.
-All AFW pumps faied to start.
Which one of the following procedures should the SRO transistion to?
A. Re-enter E-0 Reactor Trip/SI at step 1, complete immediate operator actions and then
transition to FR-H. 1 "Response to Loss of Secondary Heat Sink"
B. Re-enter E-0 Reactor Trip/SI at the begining and transition to ES-0. 1" Reactor Trip
Response" at the-appropriate-tep. ,,o
I
,t,-,A
C. Directly Enter ES-0. 1, "Reactor Trip Response&.
D. Directly Enter FR-H. 1, "Response to Loss of Secondary Heat Sinlk
Surry Exam Bank Question # 862 slightly modified.
ND-95.3-LP-2D; ND-95.3-LP-26 objectives D and F; ND-95.3-LP-41 objective A
A. Incorrect, E-0 has been exited from and CSFs apply FR-H. 1 has a red path and should be
entered.
B. Incorrect, E-0 has been exited from and CSFs apply FR-H. 1 has a red path and should be
entered.
C. Incorrect, E-0 has been exited from and CSFs apply FR-H. 1 has a red path and should be
entered.
D Correct FRP H. 1 should be entered.
Time:
I
Points:
1.00
Version: 0 1 2 345 67 89
Answer:
Scramble Range: A - D
RO Tier:
TIG2
SRO Tier:
TIG2
Keyword:
Cog Level:
C/A 3.4/4.4
Source:
B
Exam:
SR02301
Test:
S
Misc:
GW;
Wednesday, November 14, 2001 03:25:53 PM
1
EXAMINATION ANSWER KEY
RO/SRO Exam Bank
Question 861 Details
Question Type:
Topic:
System ID:
User ID:
Status:
Must Appear:
Difficulty:
Time to Complete:
Point Value:
Cross Reference:
User Text:
User Number 1:
User Number 2:
Comment:
Multiple Choice
EOP0110
73400
EOP0110
Active
No
0.00
0
1.00
1.00
0.00
0.00
ND-95.3-LP-13B; E-3
[S96-1333], [S95-1085]
ID: EOP0111
Due to a loss of feedwater pumps, the SGs go below the LO-LO setpoint and no reactor trip
occurs. The RO carries out the actions of FR-S.1, Response to Nuclear Power
Generation/ATWS. Reactor power is < 5% with a negative SUR. At the completion of this
procedure, a "red" path exists on heat sink.
Which ONE of the following procedures should the SRO go to next?
A.
Enter ES-0.1, Reactor Trip Response.
B.
Re-enter E-0, Reactor Trip/SI, at step 1, complete immediate action steps, and
then go to FR-H.1.
C.
Re-enter E-0, Reactor Trip/SI, at the step in effect and complete E-0 up to step
14, which transitions the team to FR-H.I.
D.
Enter FR-H.1, Response to Loss of Secondary Heat Sink.
Answer:
D
Page: 814 of 3141
862
Points: 1.00
OPS RO/SRO SU
10/19/01
obitecthes
After receiving this instruction, the trainee will be able to:
A.
[Given a simulated plant condition requiring the use of the Critical Safety Function Status
Trees, transition through the Heat Sink status tree denoting, in accordance with the rules of
priority, any applicable Function Restoration Procedure needing implementation. SOER
96-01, Recommendation 7]
B.
Given the Major Action Categories associated with FR-H. 1, Response to Loss of Secondary
Heat Sink, explain the purpose of FR-H. 1, the transition criteria for entering and exiting FR
H. 1, and the types of operator actions that will occur within each category.
C.
Given a copy of FR-H.1, Response to Loss of Secondary Heat Sink, explain the basis of
each procedural step.
D.
Given actual or simulated plant conditions requiring implementation of FR-H.1,
Response to Loss of Secondary Heat Sink, successfully transition through the
procedure, applying step background knowledge as required, to address the Critical
Safety Function challenge in progress.
Presentation
Distribute all handouts.
Refer to/display H/T-41.1, Objectives, and review with trainees.
NlD-9 3-T P-41
Pave 4
Revision 7
"B.
State, in order of priority sequence, the six critical safety functions.
C.
Explain the four-color, color-coding "Rules of Priority" as they apply to the CSF Status
Trees.
D.
Explain the prioritization of challenges within and between the Critical Safety Function
Procedures.
E.
Explain the points at which, during the course of a transient, CSF Status Tree monitoring is
to be implemented.
F.
Explain the use, including the function, of the CSF Status Trees during a Control
Room emergency event.
Presentation
Distribute all handouts.
Refer to/display H/T-26. 1, Objectives, and review with trainees.
A.
CSF/Barrier Associations
1.
The second category of guideline procedures contained in the ERG Procedures set
are
called the FUNCTION
RESTORATION procedures (FRs).
The
"FUNCTIONS" referred to in the title are those which must be satisfied to assure the
physical barrier maintenance to prevent radioactive material release.
ND-95.3-LP-26
Page 3
Revision 4
B.
Explain the two-column format of the Emergency Response Guideline
including the placement criteria for cautions and notes.
C.
Explain the method by which "Immediate Operator Action" steps are identified in the body
of the ERG Procedures.
D.
Describe the intended overall usage of the Emergency Response Guidelines Network.
E.
Given various plant conditions during which an emergency event occurs, evaluate the
application of the "Modes of Applicability" as described in the ERG User's Guide.
F.
Given actual or simulated EOP implementation, apply the management standards and other
good practices applicable to EOP usage.
G.
Explain the format design of the Emergency Response Guideline Procedures.
P*Esentntlnn
Distribute all handouts.
Refer to/display H/T-2. 1, Objectives, and review objectives with trainees
A.
Action Verb Identification
Direct trainees to turn to AIA-2. 1, Action Verbs. Review various action verbs with trainees.
ND-95.3-LP-2
Procedures,
Revision 7
Page 4
QUESTIONS REPORT
1. WE06EA2.1 001
for Surry2002
-Unit 2 has had a LOCA.
b
-E- 1;Loss of Reactor or Secondary Coolant is in progress.
-RCPs are secured.
-Containment Pressure is 47 psia/and slowly increasing.
-Total AFW flow is 485 gpm.
-SG WR levels are: A-48%; B-40%; C-39%.
-RCS Pressure 920 psig.
-IR NIs indicate 2 X10-Il amps, with a SUR of 0.
-CETCs indicate 600 degrees F.
-RVLIS Full Range indicates 45%.
Which of the following is the correct procedure for the team to transistion to?
A. FR-S.2'"Response to Loss of Core Shutdown"
B. FR-C.2,"Response to Degraded Core Cooling'b
C. FR-Z. 1 Response to Containment High Pressure".
D. FR-H.5 ," Response to Steam Generator Low Level'
Surry Exam Bank Question # 1066.
Surry Lesson Plans.ND-95.3-LP-26 objective D ; ND-95.3-LP-39 objective A.
A. Incorrect, S.2 would be entered on a yellow path.
B. Correct, the conditions to enter C.2 are met with RVLIS < 46%.
C. Incorrect, Z. 1 would be entered on an Orange path and C.2 is a higher priority.
D. Incorrect , H.5 is a yellow path, C.2 would be entered first,
Time:
I
Points:
1.00
Version: 0 1 2 3 4 5 6 7 8 9
Answer:
Scramble Range: A - D
RO Tier:
TIGI
SROTier:
TIGI
Keyword:
Cog Level: C/A 3.4/4.2
Source:
B
Exam:
SR02301
Test:
S
Misc:
GWL
Wednesday, November 14,2001 01:31:57 PM
Dwietahves
After receiving this instruction, the trainee will be able to:
A.
Given the Major Action Categories associated with FR-C.2, Response to Degraded Core
Cooling, explain the purpose of FR-C.2, the transition criteria for entering and exiting FR
C.2, and the types of operator actions that will occur within each category.
B.
Given a copy of FR-C.2, Response to Degraded Core Cooling, explain the basis of each
procedural step.
C.
Given actual or simulated plant conditions requiring implementation of FR-C.2,
Response to Degraded Core Cooling, successfully transition through the procedure,
applying step background knowledge as required, to address the Critical Safety
Function challenge in progress.
Presentation
Distribute all handouts.
Refer to/display H/T-39. 1, Objectives. Review objectives with trainees.
A.
Major Actions of FR-C.2, Response to degraded Core Cooling
1.
The purpose of FR-C.2, Response to Degraded Core Cooling, is to provide guidance
to restore adequate core cooling.
2.
This guideline is entered from an ORANGE priority from the CSF status tree upon
symptoms of degraded core cooling.
ND-95.3-LP-39
Page 4
Revision 7
"B.
State, in order of priority sequence, the six critical safety functions.
C.
Explain the four-color, color-coding "Rules of Priority" as they apply to the CSF Status
Trees.
D.
Explain the prioritization of challenges within and between the Critical Safety Function
Procedures.
E.
Explain the points at which, during the course of a transient, CSF Status Tree monitoring is
to be implemented.
F.
Explain the use, including the function, of the CSF Status Trees during a Control
Room emergency event.
Presentation
Distribute all handouts.
Refer to/display HIT-26. 1, Objectives, and review with trainees.
A.
CSF/Barrier Associations
1.
The second category of guideline procedures contained in the ERG Procedures set
are called
the FTUNCTION
RFSTORATION procedures (FRs).
The
"FUNCTIONS" referred to in the title are those which must be satisfied to assure the
physical barrier maintenance to prevent radioactive material release.
ND-95.3-LP-26
Revision 4
Page 3
ND-95.3-Hf/-38.2
Number:
Title:
Revision:
F-2
CORE COOLING
GO TO
FR-C.1
GO TO
FR-C.1
GO TO
FR-C.2
GO TO
FR-C.2
GO TO
FR-C.3
GO TO
FR-C.2
GO TO
FR-C.3
DrnNo.
CB38
SNSOC CHAIRMAN
DATE
EXAMINATION ANSWER KEY
RO/SRO Exam Bank
Question 1065 Details
Question Type:
Topic:
System ID:
User ID:
Status:
Must Appear:
Difficulty:
Time to Complete:
Point Value:
Cross Reference:
User Text:
User Number 1:
User Number 2:
Comment:
Multiple Choice
EOP0324
73607
EOP0324
Active
No
0.00
0
1.00
1.00
0.00
0.00
ND-83-LP-5B; ND-89.1-LP-2B
[S99-0176]
ID: EOP0325
Points: 1.00
(Refer to CSFSTs)
A LOCA has occurred and the team is presently in 1-E-1, Loss of Reactor or Secondary Coolant.
The following conditions exist:
RCPs are secured.
Containment pressure is 47 psia and slowly increasing.
Total AFW flow is 485 gpm.
SG WR levels: A-48%, B-40%, C-39%.
RCS pressure 920 psig.
IR NIs indicate 2 x 10f amps with a SUR of 0.
CETCs indicate 530oF.
RVLIS Full Range indicates 45%.
Based on the above conditions, the team should transition to ___
A.
1 -FR-H.2
B.
1-FR-Z.1
C.
1-FR-.S.1
D.
1 -FR-C.2
Page: 1014 of 3141
1066
Li
Li
Li
Li
Li
£1
Li
Li
Answer:
D
10/19/01
OPS RO/SRO SU
QUESTIONS REPORT
for Surry2002
5+
1. WE08EA2.1 001
Which one of the following conditions would require entering FR-P. 1 "Response to imminent
Pressurized Thermal Shock Condition" on an orange or red path? (CSF status trees are attached.)
A. Cooldown Greater than 100 degrees F. in 60 minutes,Temperature 290 degrees F. RCS
pressure 1800 psig.
B. Cooldown Less than 100 degrees F. in 60 minutes, Temperature 250 degrees F. RCS
pressure 350 psig.
C. Cooldown Greater than 100 degrees F. in 60 minutes,Temperature 270 degrees F. RCS
pressure 520 psig.
D. Cooldown less than 100 degrees F. in 60 minutes,Temperature 290 degrees F. RCS
pressure 1800 psig.
Bank Question, Several bank questions used to develop. From Farley, and Surry base question.
ND-95.3-LP-46 Objectives A, and D.
A. Incorrect, Does not meet the criteria for entry in to FR-P. 1 on a orange or red path.
B. Incorrect, Does not meet the criteria for entry in to FR-P. 1 on a orange or red path.
C. Correct, Meets the entry requiremnent for an orange path.
D. Incorrect, Does
Time:
I
RO Tier:
TIGI
Keyword:
Source:
B
Test:
not meet
Points:
the criteria for entry in to FR-P. 1 on a orange or red path.
1.00
Version:
0 1 2 3 4 5 6 7 8 9
Answer:
CBDDCBCABC
SRO Tier:
TIGI
Scramble Range: A - D
Cog Level:
C/A 3.4/4.2
Exam:
Misc:
S
SR02301
GWL
1
Wednesday, November 14, 2001 09:19:02 AM
ND-95.3-H/T-46.2
Number:
Title:
Revision:
F-4
INTEGRITY
2
EMEM KU UN U ilk>)
S
GO TO
FR-P.1
GO TO
FR-P.1
GO TO
FR-P.2
GO TO
FR-P.1
GO TO
FR-P.2
Graphics No. CB383
SNSOC CHAIRMAN
DATE
/
ND-95.3-H/T-46.3
Pressure (psig)
3000
2500
2000
1500
1000
500
0
I 0
Number:
Title:
Revision:
F-4
INTEGRITY
2
FIGURE 1 - OPERATIONAL LIMITS CURVE
SURRY UNIT 1 AND UNIT 2
-. 2560.psig
I
I
I
I
I
100
200
300
400
500
Temperature (°F)
Date
1
600
(
SNSOC Chairman
D-MM.~
N
WT316
ND-95.3-H/T-46.5
Number:
Title:
Revision:
F-4
INTEGRITY
2
Figure 2
RCS COOLDOWN RESTRICTIONS
3000
2800
2600
2400
2200
2000
1800
1600
1400
1200
1000
800
600
400
200
0
100
150
200
250
300
350
400
450
500
550
Cold Leg Temperature (0F)
Graphtrs No. C1392
SNSOC Chairman
Unacceptable
Operation
Acceptable
it
I
Operation
- ooldown Rates
(-F / HR)
- - - -
0
50
Date
052533K13013;
Which of the following conditions would require entering FRP-P.1 on a red or orange path?
(Circle the correct response.)
A.
Greater than 1000 cooldown in last 60 minutes to a temperature of 2500 and 100 psig.
B.
Less than 1000 cooldown in last 60 minutes to a temperature of 250' and 100 psig
C.
Greater than 1000 cooldown in last 60 minutes to a temperature of 2850 and 1800 psig
D.
Less than 1000 cooldown in last 60 minutes to a temperature of 2850 and 1800 psig
ANSWER: A. Point Value:
1.0 Answer Time: 4.0 Mins. Part B. 100
S ta tic S im S c e n a r io N o s . - -
--
S&K No.
240205023020
K/A No.
002000A0.15G
000009EA2.14
RO/SRO Impf. 4.1 /4.3
3.8/4.4
/
Objective
052533K13
Reference
052533K, FRP-P.1
CSF-0
- QNUM
33696
- HNUM
34319 (Do NOT change If < 9,000,000)
- ANUM
- QCHANGED
FALSE
- ACHANGED
FALSE
- QDATE
1992/10/19
- FAC
348
Farley 1 & 2
- RTYP
PWR-WEC3
- EXLEVEL
S
- EXMNR
- QVAL
- SEC
- SUBSORT
- KA
000011G012
- QUESTION
WHICH ONE (1) of the following conditions would require entering FNP-1
FRP-P.1, "Response to Imminent Pressurized Thermal Shock Condition"?
FNP-1-CSF-0.4, "Integrity" is attached.
a. Cooldown
degrees F,
b. Cooldown
degrees F,
less than 100 degrees F. in 60
pressure 520 psig.
less than
pressure
100 degrees F. in 60
350 psig.
c. Cooldown greater than 100 degrees F. in
275 degrees F, pressure 520 psig.
d. Cooldown greater than 100 degrees F. in
275 degrees F, pressure 350 psig.
minutes, temperature 250
minutes, temperature 250
60 minutes, temperature
60 minutes, temperature
- ANSWER
a. [+1.0]
- REFERENCE
1. Farley: OPS-52533K, "FRP-P.1, Response to Imminent Pressurized
Thermal Shock Condition", Objective 13 and FNP-1-CSF-0.4,
"Integrity".
2. Farley: License Retraining exam bank question 052533K1 3015,
question #360.
3. KA 000011G012 (4.0/4.1)
EXAMINATION ANSWER KEY
RO/SRO Exam Bank
ID: EQP0196
Points: 1.00
The operator entered FR-C.2, Response to Degraded Core Cooling, in response to an ORANGE
path condition.
of the following statements is correct with regard to transitioning out of this
A.
The operator may leave this procedure at any step as soon as the Core Cooling
adverse condition has cleared.
B.
The operator must leave this procedure before completion and go to FR-Z.1,
Response to Containment High Pressure, if the status tree indicates an
ORANGE path.
C.
The operator may leave this procedure before completion and go to FR-P.1,
Response to Imminent Pressurized Thermal Shock Condition, if the status tree
indicates an ORANGE path.
D.
The operator must leave this procedure before completion and go to FR-S.1,
Response to Nuclear Power Generation/ATWS, if the subcriticality status tree
indicates an ORANGE path.
Question 947 Details
Question Type:
Topic:
System ID:
User ID:
Status:
Must Appear:
Difficulty:
Time to Complete:
Point Value:
Cross Reference:
User Text:
User Number 1:
User Number 2:
Comment:
Multiple Choice
EOP0196
73486
EOP0196
Active
No
0.00
0
1.00
1.00
0.00
0.00
ND-95.3-LP-26D and F; ND-95.3-LP-39A
[S96-0989], [S96-1351]
Page: 898 of 3141
947
Which ONE
procedure?
Answer:
D
10/19/01
OPS RO/SRO SU
QUESTIONS REPORT
-T)
for Surry2002
1. WEI4EA2.1 001
A large steam break accident occurred 45 minutes ago, the crew transitioned to E- 1 and the
following plant conditions now exist:
-The faulted S/G has blown dry.
-The SI is still in progress.
-RCS Th and Tc are 260 degrees F.
-RCS Pressure is 1500 psig and rising
-Containment pressure is 47 psig.
-Containment sump level is 6 feet.
-Containment Rad levels are pre-event.
Which one of the following describes the appropriate procedure flowpaths that the crew should
take.
A. FR-Z. 1 should be implemented until the entry condition is restored to a yellow or green path.
B. FR-P. 1 shold be implemented until completion and then FRZ. 1 should be implemented.
C. FR-Z. 1 should be impemented until completion, and then FR-P. 1 should be implemented.
D. FR-P. 1 should be implemented until the entry condition is restored to a yellow or green path.
Ref: from exam bank (Farley).
Surry Lesson Plan ND-95.3-LP-48 Objectives A and D.
A. Incorrect, A red path does exist on Z. 1, it should be finished through to completion, and then
FR-P. 1 should be entered. Tfhe team should not wait for the conditions to become green or
yellow.
B. Incorrect, FR-P. 1 should be entered, but it is an orange path, and a red pathe exists on Z. 1, so
Z. 1 should also be entered and it should be entered first.
C. Correct, Per the CSF's and entry conditions.
D. Incorrect, A red path exists on Z. 1, it should be finished through to completion, and then
FR-P. 1 should be entered.
Time:
I
Points:
1.00
Version:
0 1 2 345 67 89
Answer: CBDDAADDCB
Scramble Range: A - D
RO Tier:
TIGI
SROTier:
TIGI
Keyword:
Cog Level:
C/A 3.3/3.8
Source:
B
Exam:
SR02301
Test:
S
Misc:
GWL
F-riday, November 09, 2001i 09:48:.u
,il
052533K13012;
A large steam break accident occurred 50 minutes ago, and the following plant conditions now
exist:
-
The faulted SG has blown dry.
The SI is still in progress.
-
RCS Th and Tc are 255c.
-
RCS pressure is 1600 psig and rising.
-
PRZR level is 95% and stable.
-
Containment pressure is 56 psig.
-
Containment sump level is 6 feet.
-
Containment rad monitors at pre-event values.
Based on the above conditions: (Circle the correct response.)
A.
FRP-Z.1 is the only procedure which should be implemented until entry condition is
restored to yellow or green path.
B.
FRP-P. I is the only procedure which should be implemented until entry condition is
restored to yellow or green path.
C.
FRP-Z. 1 should be implemented until completion, and then FRP-P. I should be
implemented.
D.
FRP-P. 1 should be implemented until completion, and then FRP-Z. I should be
implemented.
ANSWER: C. Point Value: 1.0 Answer Time: 6.0 Mins. Part B. 100
Static Sim Scenario Nos. --
-
-
S&K No.
240205023020
K/A No.
002000A015G
000009A2.14
RO/SROImpf. 4.1 /4.3
3.8/4.4
/
Objective
052533K13
Reference
052533K, FRP-P.l
Rev. Date 8/24/94
Obiectives
After receiving this instruction, the trainee will be able to:
A.
Given a simulated plant condition requiring the use of the critical safety function status
trees, transition through the Containment Status Tree denoting, in accordance with the rules
of priority, any applicable function restoration procedure needing implementation.
B.
Given the Major Action Categories associated with FR-Z. 1, Response to Containment High
Pressure, explain the purpose of FR-Z. 1, the transition criteria for entering and exiting FR
Z. 1, and the types of operator actions that will occur within each category.
C.
Given a copy of FR-Z. 1, Response to Containment High Pressure, explain the basis of each
procedural step.
D.
Given actual or simulated plant conditions requiring implementation of FR-Z.1,
Response to Containment High Pressure, successfully transition through the
procedure, applying step background knowledge as required, to address the Critical
Safety Function challenge in progress.
Preentation
Distribute all handouts.
Refer to/display H/T-48. 1, Objectives and review objectives with trainees.
A.
Containment Status Tree
1.
The Containment status tree provides a systematic method to determine the status of
the Containment Critical Safety Function.
xrror 1-1P -4
Paoe 4
Revision 7
Drawing No. CB382
SNSOC CHAIRMAN
ND-95.3-HIT-48.2
Number:
Title:
Revision:
F-5
CONTAINMENT
v
S
GO TO
FR-Z.1
EISl
m m a m
N u M
a III
GO TO
NO
CONTAINMENT
NO
- PESSR
LESS LESSHT
7.2 FEET
YE
GOTO
CONTAINMENT CN
RADIATIONR
E
LESS THAN
/.,.R
YES
FO TO
FR-Z.24
CONTAINMENT
NO@
PRESSURE
LESS THAN
13 PSIA
YE
DATE
ND-95.3-HIT-46.2
Title:
INTEGRITY
Revision:
2
GO TO
FR-P.1
GO TO
FR-P.1
GO TO
FR-P.2
GO TO
FR-P.1
GO TO
FR-P.2
Gruphtcs No CR383
SNSOC CHAIRMAN
DATE
Number:
F-4
ND-95.3-H/T-46.3
Number:
Title:
Revision:
F-4
INTEGRITY
2
Pressure (psig)
FIGURE 1 - OPERATIONAL LIMITS CURVE
SURRY UNIT 1 AND UNIT 2
3
12560 psig
2500
2000
150C
1000
500
0
SNSOC Chairman
Date
7
300
4
4O00
1
500
Temperature (0F) i-*
Dm,
No
W316
2
I
I
100
200
T
0
)
6
600
EXAMINATION ANSWER KEY
RO/SRO Exam Bank
ID: EOP014S
Points: 1.00
The following conditions exist:
In response to a large break LOCA, a transition from 1-E-O, Reactor Trip or Safety
Injection, to 1-E-1, Loss of Reactor or Secondary Coolant, has been performed.
Due to a RED path on the Core Cooling Status Tree, a transition to 1-FR-C.1,
Response to Inadequate Core Cooling, has been performed.
During performance of 1-FR-C.1, you observe that the Core Cooling Status Tree
has changed from a RED to a YELLOW condition while you identify a RED path
on the Containment Status Tree.
Which ONE of the following is the proper procedural transition, and why?
A.
Immediately transition to 1-FR-Z.1, Response to Containment High Pressure,
since a RED path is a higher priority than a YELLOW path.
B.
Complete 1-FR-C.1; since once ANY FR is entered, it must be completed before
any other transition can be made.
C.
Complete 1 -FR-C.1; since it was entered due to a RED path, it must be
completed unless a higher priority path occurs, then transition to FR-Z.1.
D.
Perform the actions of 1-FR-C.1 and 1-FR-Z.1 simultaneously, since FR
procedures of the same priority can be executed together.
Question 896 Details
Question Type:
Topic:
System ID:
User ID:
Status:
Must Appear:
Difficulty:
Time to Complete:
Point Value:
Cross Reference:
User Text:
User Number 1:
User Number 2:
Comment:
Multiple Choice
EOP0145
73435
EOP0145
Active
No
0.00
0
1.00
1.00
0.00
0.00
ND-95.3-LP-26D and F; ND-95.3-LP-48B
[S96-0989], [S96-1360]
OPS RO/SRO SU
Pag: M 01141'UWI
I
896
Answer:
C
10 1901l
Page: 848 0of 3141