ML20205B121
| ML20205B121 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 07/25/1986 |
| From: | Bill Dean, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20205B097 | List: |
| References | |
| 50-335-OL-86-01, 50-335-OL-86-1, NUDOCS 8608110528 | |
| Download: ML20205B121 (136) | |
Text
.
r p KfGug UNITED STATES
[
D NUCLEAR REGULATORY COMMISSION o
[
REGION 11 n
g j
101 MARIETTA STREET,N.W.
2 ATLANTA, GEORGI A 30323
\\...../
ENCLOSURE 1 EXAMINATION REPORT 335/0L-86-01 Facility Licensee:
Florida Power and Light Company P. O. Box 14000 Juno Beach, FL 33408 Facility Name:
St. Lucie Nuclear Plant Facility Docket Nos:
50-335 and 50-389 Written and oral examinations were administered at the St. Lucie Nuclear Plant near Jensen Beach, Florid.
[
7 MF Chief Examiner:
William M. Dean <J D&te Signed.
Approved by:
7
/t JoTin F. Munr6,( Acting Section Chief Date'51gned Susanary:
Examinations conducted June 10-12, 1986.
Oral and written examinations were administered to four Senior Reactor Operators (SRO) and one Reactor Operator (RO) candidate. 'Three of four SR0s passed the written examination and four of four SR0's passed the-oral examination. The R0 failed the oral and written examinations. Additionally, one SR0 candidate was administered a retake of Section 6 of the written examination-and passed.
Based on the results described above, 0 of 1 R0 and 4 of 5 SR0s passed the examination.
kDR hhcxo30oo333 28 860730 V
REPORT DETAILS 1.
Facility Employees Contacted:
- M. Shepard, Operations Training Supervisor
- R. Weller, Operations Training
- Attended Exit Meeting 2.
Examin'ers:
- W'.Dean T. Rogers
- Chief Examiner 3.
Examination Review Meeting At the conclusion of. the written examinations, the examiners provided your tra,ining staff with a copy of the written examination and answer key for review. The comments made by the facility reviewers are included as to this report, and the NRC Resolutions to these comments are listed below.
a.
R0 Exam (ApplicableSR0questionsinparenthesis)
(1) Question 1.15 0
NRC Resolution: Based on only one reason provided for in reference material, will accept a reasonable second answer.
l (2) Question 1.19 l
NRC Resolution: The question asks for symptoms that would occur on a loss of natural circulation.
If candidates state assumptions for the cause, answer will be graded accordingly, otherwise answer key will apply.
(3) Question 2.02(6.01)
NRC Resolution:
System description used to generate question i
does not agree with electrical schematic provided by the facility.
I Will change answer key to (b) vice (a).
(4) Question 2.04(a)
NRC Resolution: Reference SD 8 still shows that an interlock exists between the valves in question. Question and answer stand.
I
2 JUL 3 01986 (5) Question 2.04(c)
NRC Resolution: Based on recent system modification not reflected in system descriptions, question will be deleted.
(6) Question 2.19 NRC Resolution: Typographical error in answer key reflected wrong gas for accumulators. Answer key will be modified as recommended.
(7) Question 2.20(6.20)
NRC Resolution: Based on possibility of 2 different interpre-I tations, will accept recommended additional answer, amended to include proper component identification (e.g., MCC 2AS) and additional components in the electrical path (e.g., Manual transfer switch).
(8) Question 3.05(c)
~
NRC Resolution: Agree with facility comment. Question deleted.
(9) Question 3.09(a)
NRC Resolution: Will accept answer that indicates absence of a normally expected color (e.g}., white or citar).
(10) Question 3.11 NRC Resolution: Based on revision to system descriptions not initially provided to the NRC, will modify answer key to require only four inputs. Will not penalize candidates for erroneous fifth answer.
(11) Question 3.12(6.14)(b)
NRC Resolution: Based on recent changes to Technical Specifi-cations, will modify answer key as recommended.
(12) Question 4.02 NRC Resolution: Agree with facility comment. Will accept (c) or (e)ascorrect.
(13) Question 4.11(7.09)
NRC Resolution: The K/A listed may not completely reflect the value of the content reflected in the exam question.
However K/A 062/000:K3.01, "... knowledge of effect of loss of AC on major system loads", with the same importance rating (3.5/3.9), applies.
Question and answer key stand as is.
3 JUL 3 0 B86 l
(14) Question 4.19 NRC Resolution: Based on observation in control room of the RCO log also referred to as the Control Center Log (which is not in referenced Admin. Instruction) will accept (5) as substitute for (6).
Additionally, based on post-exaraination review, Question 3.15 was found to provide insufficient direction to the candidate to elicit desired response and was deleted, b.,
SR0 Exam (1) Question 5.07 NRC Resolution: Coment applies to part (c) of question. Will accept " prevent DNB in event of sudden significant drop in flow" asananswerforpart(c).
(2) Question 5 09(b)
NRC Resolution: Agree with facility comment. Question deleted.
(3) Question 5.09(c)
NRC Resolution: St. Lucie OP 0030126, "ECC and ICCR" requires
..j l'
4 operators to use most recent Boron sample vice plant curves for critical Boron calculation. Also, did not indicate whether boration or dilution was required in the question. Facility comment accepted and question deleted.
(4) Question 5.11(b)
NRC Resolution:
Facility comment should request change answer key to Unit 1 vice Unit 2.
Typographical error in answer key will be corrected.
j (5) Question 5.14 NRC Resolution: Agree with facility comment. Recommended addi-l tional answer accepted.
(6) Question 6.05 NRC Resolution: Based on facility provided material supporting non-usage of subject CEA change mechanism, question is deleted.
(7) Question 7.02 l
l NRC Resolution:
Facility coment accepted. Will accept (a) or (b)ascorrect.
~ __.-
4 JUL 3 01986 (8)
Question 7.03(b)
NRC Resolution: Based on change to procedures, not originally held by NRC, answer key will be modified as recomended by facility.
f (9) Question 7.17(b)
NRC Resolution:
Recommended answer that forced circulation be maintained is just a rephrasing of the conditions given in the i
question. No change to answer key.
i (10) Question 7.18(b)
NRC Resolution: Based on additional information provided by facility, will accept recommended additional answer.,
I (11) Question 8.08(c)and(d)
NRC Resolution: Based on question asking for a particular accident for which the react 6r trips are designed, and no such accident listed in Technical Specification's, will modify answer key to accept Unit 1 or Unit 2 T/S design bases or reasonable accident descriptions that could cause these design bases to be threatened.
>; a -
i
!(12) Question 8.14(a)
NRC Resolution: Based on examiner observation that key cards are mated with the ID badge, will modify answer key to equate the ID j
badge with the keycard as one response.
(13) Question 8.18 NRC Resolution:
Based on review of OP2-1630024, there are separate precautions supporting either possible answer for part (a). Since speed automatically changes in both instances, the answer key for part (b) is not an explanation that differentiates between the two possible selections in part (a). Question i
deleted.
4.
Exit Meeting At the conclusion of the site visit the examiners met with representatives l
of the plant staff to discuss the results of the examination.
I There were no generic weaknesses noted during the oral examination.
It was noted that the SR0 candidates seemed to be well versed in the recently implemented emergency operating procedures.
5 JUL 3 01986 The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was also noted and appreciated.
The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners.
4 l
l l
I l
(
~ - -
6 h
0 a,
L //
/~~ _
'~ _
ENCLOSURE 2 U.
S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
ST. LUCIE 1 REACTOR TYPE:
PWR-CE DATE ADMINISTERED: 86/06/10 EXAMINER:
DEAN, WM APPLICANT:
INSTRUCTIONS TO APPLICANT:
Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question. The passing Stade requires at least 70% in each category and a final grade of at least 80%.
Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY
% OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY a '. s 30.00
-2ST00' 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
.11. r
.5 30TOO 25T00'
________ 2.
PLANT DESIGN INCLUDING SAFETY
~-.
AND EMERGENCY SYSTEMS
.s, a
.: s s
.34v07 25-rOO
________ 3.
INSTRUMENTS AND CONTROLS l
a.
es.
l W' sO 26M
________ 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL l
l vT $
l 4-E h ttT-100.00 TOTALS FINAL GRADE _________________%
l All work done on this examination is my own. I have neither sivsn nor received aid.
hPPLEUAUT 5~55UUhiURE
~~~~~~~~~~~~~
I l
e 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2
--- isEER557RARiC5-REAi iEAnsFEE As5 FE5i5 FE5E QUESTION 1.01 (1.00)
The reactor is critical at 10,000 cps when a S/G PORV fails open.
Assuming BOL conditions, no rod motion, and no reactor tripe choose the answer below that best describes the values of Tavs and nuclear power for the rgsvlting new steady state.
(PDAH = point of adding heat).
a.
Final Tave greater than initial Tavs, Final power above POAH.
b.
Final Tave greater than initial Tavs, Final power at POAH.
c.
Final Tavs less than initial Tavs, Final power at POAH.
d.
Final Tavs less than initial Tavs, Final power above POAH.
QUESTION 1.02 (1.00)
Which of the following will cause plant efficiency to increase?
a.
Total S/G blowdown is changed from 30 spm to 40 spm.
b.
Steam quality changes from 99.7% to 99.9%.
c.
Level increase to higher than normal in a feedwater heater.
d.
Absolute condenser pressure changes from 1.0 psi to 1.5 psi.
QUESTION 1.03 (1.00)
Reactivity is defined as which of the following?
a.
The ratio of the number of neutrons at some point in this Seneration to the number of neutrons at the same point in the previous generation.
b.
The fractional change in neutron population per generation.
c.
The factor by which neutron population changes per genera-tion.
d.
The rate of change of reactor power in neutrons per second.
(*****
CATEGORY 01 CONTINUED ON NEXT PAGE *****)
7
.1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3
--- isEER65isisics-AEEi iEAAsFEE As5 FEUi5 FE5E QUESTION 1.04 (1.00)
Which of the following curves (see attached page) representins Xenon concentration is correct for the given power history?
DUESTION 1.05 (1.00)
Initially, one centrifugal chargins pump is in operation when a second centrifugal chargins pump in parallel with the first pump is'also put into cperation.
Which statement below correctly describes the effect on system volumetric flow rate and system head loss?
a)
Hisher flow rater higher head loss b)
Same flow rate, higher head loss c)
Higher flow rater same head loss d)
Same flow rater same head loss GUESTION 1.06 (1.00)
Of the followins, which must the main condenser remove the most heat from to condense? (assume steam is of equal quality)
- a. one pound of steam at 0 psia.
b.
one pound of steam at 300 psia.
c.
two pounds of steam at 600 psia.
d.
two pounds of steam at 1200 psia.
~
GUESTION 1.07 (1.50)
For the changes listed below (treat each one independently) indicate whether the moderator temperature coefficient will become MORE NEGATIVE, LESS NEGATIVE or have NO EFFECT. (Assume all other parameters are constant) a)
Neutron flux peak shifts radially inward from the edge of the core.
b)
Baron concentration decreases 100 ppm while core is at MOL.
c)
Increased number of burnable poisons are inserted into the core.
(***** CATEGORY 01 CONTINUED ON NEXT PAGE
- )
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4
~~~~ UER566 U5U5C5I~UEIT TRdU5f5R d 6~ELU56~fL6U
~
~
T QUESTION 1.08 (1.50)
Indicate whether the following will cause the Upper-Lower summer output to become more positive or less positive (ie. indicate direction of change) a)
Power increase with power defect compensated for by dilution only.
b)
Power increase with power defect compensated for by rod withdrawal only.
c)
Buildup of nonon in the top portion of the core.
QUESTION 1.09 (2.50)
Match the descriptions in Column A with the T-S paths listed in Column B utilizing the attached T-S diagram.
(2.5)
Column A Column B c)
Reheat from moisture separators 1) 1-2 b)
Heat addition from feedwater heaters 2) 2-3 c)
Pump Work 3) 3-4 d)
Condensation in the condenser 4) 4-5 o)
Heat removed by the LP turbine 5) 5-6 6) 6-7 7) 7-1 QUESTION 1.10 (1.00)
Indicate what type of detector is used in the following excore instruments:
c)
Unit 1 Wide range Logarithmic Safety Channel b)
Unit 2 Linear Safety Channel c)
Unit 2 Logarithmic Startup Channel GUESTION 1.11 (1.00) a)
What is the nost significant method of Xenon removal while at low power levels?
b)
What happens to samarium concentration for the first week after an increase in power level to 100% from 50%?
(zrxxx CATEGORY 01 CONTINUED ON NEXT PAGE xxxxx) i
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5
--- isEER557sAsics-sEAi isAsiFEE As5 FEUi5 FE5E QUESTION 1.12 (1.50)
List the three main power producing isotopes in the core at end of life end indicate their approximate contribution (in %) to power.
QUESTION 1.13 (2.00)
Unit i has just restarted following a refueling outage while Unit 2 is naar EOL.
Answer the following regarding the differences in plant response between the two units (explain your answers)!
c)
At a steady power level of 10EE(-8) amps during a startup, equal reactivity additions are made (approximately 100 pcm).
Which Unit will have the higher steady state startup rate?
b)
At 50% power, a control rod (100 pcm) drops.
Assuming NO RUNBACK or OPERATOR ACTION, which Unit will have the lower steady state Tavs?
QUESTION 1.14
(.75)
List three fundamental nuclear properties of a good moderator.
QUESTION 1.15 (1.00)
What are the two reasons for shifting the SI mode from cold les rceirculation to hot les recirculation approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after a LOCA?
QUESTION 1.16 (1.50)
During the performance of an emergency boration while at power, how and why are the following parameters affected? (assume no control rod movement) c)
subcooling b) control rod worth
(***** CATEGORY 01 CONTINUED ON NEXT PAGE
- )
l l
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 6
--- TAERR557AARiEs-REAi iEAREFEE AR5 FEUi5 FE5s QUESTION 1.17 (1.00)
Explain how decreasing RCS flow (at constant power) will result in decreasing Critical Heat Flux Ratio.
(1.50)
QUESTION 1.18 c)
Provide an equation representins Axial Shape Index (ASI).
(0.5) b)
Explain how ASI changes over core life and why these changes occur.
(Ensure you discuss ASI conditions at BOL also)
(1.0)
QUESTION 1.19 (2.00)
Describe HOW and WHY the following will respond to a loss of natural circulation flow following a reactor trip from 100% equillibrium conditions.
c)
RCS wide ranse temperature difference (Delta T) b)
Relationship between Tcold wide ranse and S/G pressure QUESTION 1.20 (1.00)
Explain the relationship of Equilibrium Shape Index (ESI) to the Axial I
Shape Index (ASI).
QUESTION 1.21 (1.25) 0)
Why is Decay Heat dependent upon power history of the core?
(0.5) b)
Using the los Paper provided, sketch the curve of decay heat vs. time for the first 4 days following a trip from power.
Assume no abnormal conditions exist in the core. { Note use the los scale for time}
(.75)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
$l i
i 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATIONr PAGE 7
--- isEER557REsiEi-sEsi iEAssFEE ER5 FE5i5 FE5s QUESTION 1.22 (1.00)
Mcke a sketch of the temperature profile along the length of a counter flow h2at exchanger for both the cooling medium and the fluid being cooled.
QUESTION 1.23 (2.00)
Given the following, calculate the required boron change to increase roactor power from 75% to 100% while maintaining constant rod position.
Moderator temp. coeff.
-15 pcm/ degree F Doppler-only power coeff.
-12 pcm/% power Void reactivity change
-25 pcm Xenon change
-50 pcm Boron coeff.
-9 pcm/ ppm l
(*****
END OF CATEGORY 01
- )
1
\\
20 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8
QUESTION 2.01 (1.00)
Which of the following describes the purpose of the svetion stabilizers (dampeners) associated with the char 3 ng pumps?
i a.
Prevent excessive pulsation levels in the piping downstream of the charging pumps.
b.
Prevent' overpressure of the suction side of the charging pumps.
c.
Mininiize system losses due to acceleration head and assure that sufficient NPSH is available.
d.
Provide an isolation signal for the appropriate charging pump if suction side pressure oscillations become too large.
QUESTION 2.02 (1.00)
Which of the following statements correctly describes how to reset the MECHANICAL overspeed on the turbine driven AFW pump?
a.
It must be reset locally, and then the limitorque driven in to the closed position to relatch the linkage.
b.
It must be reset locally by using a lever to relatch the linkage.
c.
It will reset automatically as turbine speed decreases below a pre-determined setpoint.
d.
It may be reset locally, but can also be reset from a switch on RTGB 102 in the control room.
f GUESTION 2.03 (1.00)
Which of the following correctly describes the normal lineup of the 125 VDC Swing Bus?
a.
The swing bus AB is powered from the A-side train in UNIT 1.
b.
The C train is powered from either of the A or B buses via the swing bus AB in UNIT 1.
c.
The swing bus AB is powered from the A-side train in UNIT 2.
d.
The swing bus AB is supplied by either a battery charger or a battery backup in UNIT 2.
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9
QUESTION 2.04
-(1.50)
Indicate to which Unit the following descriptions of RCP seal injection l
EPP Y*
a)
The ReSenerative Heat Exchanger inlet is interlocked with the Seal Injection Isolation MOV.
b)
During Back'up Seal Injection operations, two chargins pumps are needed
'e ) -During cold shutdown, a temporary seal injection line can be run from the tube side drain of the regenerative heat exchanger. f n g;11,3 QUESTION 2.05 (2.00)
For each of the following statements regarding Safety Injection Systems, indicate whether it applies to UNIT 1, UNIT 2 or BOTH UNITS.
a)
The High Pressure SI System consists of two separate discharge headers which can be cross-connected by a normally shut motor operated valve.
b)
The Low Pressure SI pumps utilize Component Cooling Water to cool pump discharge which is then recirculated to the pump seals for coolin3+
c)
Hydrazine is injected into the containment spray pump suction to facilitate iodine removal.
d)
On a Recirculation Actuation Signal (RAS), the LPSI recirculation miniflow lines are automatic 611y isolated.
QUESTION 2.06 (1.50)
Which UNIT's LPSI System would be more susceptible to a single failure (passive or active) causins it to fail to inject on a SIAS?
Explain your answer by describins what system failure would cause this to occur.
(*****
CATEGORY 02 CONTINUED ON NEXT PAGE *****)
i 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10
-0UESTION 2.07 (1.00)
Which of the following describes the basis for the design flow rate from the Charging Line Thermal Relief Valve (V-2435)?
a.
Pass the maximum fluid thermal expansion rate that would occur if a charging pump was operated with its suction and discharge valves, closed.
b.
Pass the maximum rated flow of a charging pump.
c.
Pass the capacity of a Letdown Control Valve in normal operation.
d.
Pass the maximum Letdown Flow Possible without actuating the Letdown High Flow alarm.
e.
Pass the maximum fluid thermal expansion rate that would occur if hot letdown flow continued after the charging line distribution valves were closed.
QUESTION 2.08 (1.50)
Indicate whether the following statements regarding the main steam system apply to UNIT in UNIT 2 or BOTH UNITS.
a)
The atmospheric steam dump isolation valves can be remotely operated from the control room.
b)
A check valve downstream of the MSIV is utilized to prevent backflow of steam from the other S/G if a steam break were to occur upstream.
c) 8 Safety Valves, divided into two groups that will lift at diffe' rent pressures, are located on each steam line upstream of the MSIV.
QUESTION 2.09 (2.00)
Describe the 4 flow paths within the reactor vessel which BYPASS the fuel rods.
(*****
CATEGORY 02 CONTINUED ON NEXT PAGE *****)
l y- - -.
y
~_
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 11 QUESTION 2.10 (2.00)
List the loop penetration location (s) for the following RCS connections for Unit 2.
c)
Surge Line b)
Charging Lines c)
Letdown Lina QUESTION 2.11 (2.25)
List the eight automatic actions that occur upon an SIAS that align the CVCS to a flow path for maximum boration.
(Note: actions affecting similar components, such as both RCS inlet valves opening, count as one action)
NOTE ANY DIFFERENCES CETWEEN UNIT 1 and UNIT 2 ACTIONS!
" QUESTION 2.12 (1.50)
List SIX potential sources of In-leakage causing excessive activity levels in the CCW System. (Leakase from redundant components will be considered as one response)
GUESTION 2.13 (1.00)
When performing a fill and vent operation on the RCS in accordance with OP 0120020, why must the fill rate be maintained below 500 spm?
Indicate the limiting component (s) that is/are the basis for this precaution.
QUESTION 2.14 (1.00)
Most RCS penetrations are located above the RCS loop pipe horizontal centerline to prevent crud traps from occuring.
Why do the shutdown coolins lines penetrate the bottom of the hot less?
GUESTION 2.15 (1.00)
Why do the UNIT 1 Containment Spray Pumps require Seal Water Heat Exchangers, while the UNIT 2 Containment Spray Pumps do not?
(xx*** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
-l
2.
PLANT DESIGN INCLUbING SAFETY AND EMERGENCY SYSTEMS PAGE 12 l
QUESTION 2.16 (1.50)
Give three reasons why Hydrogen sas is not utilized for oxygen scavensing when the plant is in cold shutdown conditions.
QUESTION 2.17 (1.00)
Describe how redctor vessel head leakage is monitored, including a description of the type of sensins device utilized and the applicable alarm setpoint.
QUESTION 2.18 (1.00)
Why is the storage capacity of UNIT 2's Condensate Storage Tank (CST) significantly higher than UNIT 1's CST? (ie. what is the design basis that accounts for this difference)
QUESTION 2.19 (1.75)
Describe the operational modes of the UNIT 2 Main Feedwater Isolation Valves.
Include a description of the motive force for the valves and their response in both normal and accident conditions.
QUESTION 2.20
.(2.00)
Describe the flowpath from the appropriate MCC to the 2A 120 VAC Vital Instrument Panel for the following situations: (Identify ALL components) e)
Normal b)
Maintenance Conditions QUESTION 2.21 (1.50)
Describe the difference in the operation of the diesel oil transfer system for UNIT 1 and UNIT 2 when a demand signal for the transfer pumps to start is received. (The control switches are in AUTC)
(***** END OF CATEGORY 02 *****)
.3.
INSTRUMENTS AND CDNTROLS PAGE 13
\\
QUESTION 3.01 (1.00)
Which of the following is used as input to'the Steam Generator Level Cor: trol System when less than 15% power?
a.
Nuclear Power b.
Feed Flow / Steam Flow Mismatch c.
S/G Donncomer Level d.
Steam Flow Generated Level Program QUESTION 3.02 (1.00)
Which statement below correctly describes the effect of the condenser vacuum interlock on the Steam-Bypass Control System?
a.
If ALL M/A statio'ns are in AUTO when vacuum is regained, then the Condenser Vacuum Reset button must be depressed to remove the interlock.
b.
ONLY if the Master Integrated Controller is in Manuale is it required to depress the Condenser Vacuum Reset button to remove the interlock.
c.
It makes no difference if the M/A stations are in Manual or Autor when the condenser vacuum is regained, the interlock is removed automatically.
d.
If ANY of the M/A stations are'in manual when vacuum is regained, s
then the Condenser Vacuum Reset button must be depressed to remove the interlock.
QUESTION 3.03 (1.00)
As it applies to the Core Protection Calculator, which of the following describes Thermal Power?
a.
AveraSe Thot - Highest Tcold b.
Average That - Average Teold c.
Highest Thot - Highest Teold d.
Highest Thot - Lowest Teold e.
Highest Thot - Average Tcold
(**xxx' CATEGORY 03 CONTINUED ON NEXT PAGE *****)
p 3.
INSTRUMENTS AND CONTROLS PAGE 14 QUESTION 3.04 (1.50)
Indicate whether each of the following statements regarding the overpressure mitigation system (OMS) applies to UNIT 1r UNIT 2 or to BOTH UNITS.
o)
There are FOUR pressure comparators (PC's) that are used as inputs to Channel A and Channel B for low pressure protection, one channel per PORV.
b)
The PORV low pressure protection system is designed to prevent exceeding the Tech Spec PTS limits if an inadvertant Safety Injection by two HPSI pumps and three charging pumps.
c)
A temperature interlock prevents PORV actuation due to a low pressure relief open signal if RCS temperature is > 320 degrees F.
QUESTION 3.05 4-h 50&
[. v \\
l Indicate whether the following statements apply to the Control Element i
Drive Mechanism System on UNIT 1, UNIT 2 or BOTH UNITS.
a-)
During a withdrawal sequence, when the lift coil energizes, it exerts a downward magnetic force on the anti-ejection gripper armatures, allowing the lifting motion to take place.
b)
The CEA Withdrawal Prohibit (CWP) may be bypassed by depressing a pushbutton on the control panel for the CEAs.
ci ~TShtstdown-CEA'can--NOT-be-withdrawn if all the regulating groups and
.NOT.at their Lower Electrical Limit-(LEL).
Assume no bypass functions are actuated.
QUESTION 3.06 (1.00)
Answer TRUE or FALSE to the following statements regarding the ESFAS aeasurement cabinets.
a)
If a condition which caused a Trip Bistable to trip has cleared, then the lighted Trip Reset Pushbutton will no longer be illuminated.
b)
If a bistable module in UNIT 1 is removed, the circuitry will cause al trip si3nal to be sent to the logic module's trip logic.
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
4....
3.
INSTRUMENTS AND CONTROLS PAGE 15 QUESTION 3.07 (1.50)
Indicate whether the following transitions using the S/G Water Level Control Hand / Automatic Station are "BUMPLESS' or NOT.(ie. which requires operator action to adjust setpoints to ensure a smooth transfer) o)
Trcnsitin3 from " Hold" to " Manual" b)
Transiting f. rom ' Manual' to " Auto" c)
Transiting from " Hold' to ' Auto' QUESTION 3.08
(.75)
Fill in the blanks below to correctly complete the statement regarding SBCS operation!
A Quick Opening Signal is generated subsequent to a ____________
if both secondary pressure is > ______ psia and Tavs > _____ des F.
QUESTION 3.09 (1.00) l Indicate the color that will be displayed on the UNIT 1 CEA Core Mimic display for the followin3 situations!
a)
Shutdown CEA in its operating band b)
Regulating CEA is at its Upper Electrical Limit (UEL) c)
Shutdown CEA is at 120 inches d)
Regulating CEA is dropped QUESTION 3.10 (1.75)
With the reactor at 30% power, all controllers in AUTO, heaters and sprays off and one charging pump running, the pressurizer level starts to decrease from its setpoint.
List the changes in the CVCS and any alarm / protective functions that occur as pressurizer level decreases until a TM/LP trip occurs.
Give the applicable setpoints and note any differences between UNIT 1 and UNIT 2.
(Assume no operator action)
(****x CATEGORY 03 CONTINUED ON NEXT PAGE xxx**)
l i
l
1 o
l i
3.
INSTRUMENTS AND CONTROLS PAGE 16 7
QUESTION 3.11 (1.50)
List the 5 inputs to a Reactor Regulating System Cabinet and indicate if any differences in the inputs from UNIT 1 and UNIT 2 exist.
QUESTION 3.12 (2.50)
List ALL the logics that will cause the following ESFAS subsystems to actuate. (Include applicable setpoints) o)
Containment Isolation Actuation Signal (CIAS)---- UNIT 1 b)
Main Steam Isolation Signal (MSIS)---- UNIT 2 QUESTION 3.13 (1.50)
List the three Saturation Margin Indications that are available on the Saturation Margin Monitorin3.(SMM) Page of the QSPDS.
QUESTION 3.14 (1.50)
List ALL the conditions which will cause the 'CEA Motion Inhibit' alarm to actuate.
QUESTION '3.15~
-- (1.50)
?). *J M
~
wn.
Dascribe the inputs into the two UNIT.1 signal deviation comparators and how alarm signals are generated (including applicable setpoints).
QUESTION 3.16 (2.00) a)
Why are the indicating lamps on the RTGB above each PORV control switch NOT a positive indication of PORV position?
b)
What system is used to provide positive indication of PORV position and how is this information displayed?
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
3.
INSTRUMENTS AND CONTROLS PAGE 17 QUESTION 3.17 (1.00)
Describe how the Heated Junction Thermocouple (HJTC) System detects o collapsed liquid level above the core.
QUESTION 3.18 (3.00) c)
What is the' purpose of the Power Trip Test Interlock?
b)
What trips are automatically bypassed by the RPS?
c)
How is the Zero Power Mode Bypass initiated and when is this bypass automatically removed?
QUESTION 3.19 (1.50)
Describe how the UNIT 1 Wide-Range Logarithmic Safety Channel controls the indicating range that is supplied by the instrument.
Include in your discussion any indications that inform the operator of what scale is in affect.
QUESTION 3.20 (2 00)
Describe the flow path for a Channel A Analog trip signal that will result in energizing the K-1 through K-4 relays that cause the reactor trip break-ors to open.
Identify all components that the signal must pass through.
Do not explain redundant paths and assume that the required logic to cause o reactor trip exists.
(*****
END OF CATEGORY 03 *****)
.4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 18
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~ d6E6L55iCdt 55UTR6L R
QUESTION 4.01 (1.00)
Which of the following is an indication that Natural Circulation is being ostablished?
a.
Loop Delta T is equal to Full Power Delta T b.
That add Teold are increasing slowly at the same rate c.
Thot is equal to Tsat associated with S/G pressure d.
Thot is stabl? and Teold is slowly decreasing
'0UESTION 4.02 (1.00)
Which of the following locations is where the
'B' RCO reports to on a Control Room Evacuation due to inhabitability?
a.
Remote Shutdown Panel b.
Electrical Equipment Room, Reactor Auxiliary Building c.
Turbine Operating Level d.
LPSI Pump Room e.
6.9 KV Switchgear Room QUESTION 4.03 (1.00)
Indicate whether each of the following statements regarding cooldown of the reactor plant are applicable to UNIT 1, UNIT 2 or BOTH UNITS.
a)
The shutdown cooling system (SDC) shall NOT be placed in service until the RCS pressure is < 265 psia and the temperature is < 325 Def'F.
b)
When the RCS temperature is < 500 des F and RCS pressure is < 1500 psis rack in the breakers and close the SI Tank discharge valves.
Rack the breakers out again once the valves are closed.
(*****
CATEGORY 04 CONTINUED ON NEXT PAGE
- )
4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 19
~~~~ d656L6656dL C6UTR6L
~
~~~~~~~~~~~~~~~~~~~~~~~~
R QUESTION 4.04 (1.50)
Answer the following questions TRUE or FALSE regarding E0P usage a)
While conducting the Standard Post Trip Actions (SPTA), if a safety function criteria is NOT met before ALL immediate actions are completed it is Permissable to exit E0P-1 to go to the approriate E0P.
b)
The STA should perform the Safety Function Status Checklist (App. A) at least once every 30 minutes while any E0P is being performed.
c)
When perforning an assessment of the Safety Functions while in E0P-8, you must assess ALL of them before referring to the Resource Trees (Appendices B-I) contained in the back of E0P-8, even if a Safety Function criteria is NOT met.
QUESTION 4.05 (2.00)
Indicate what automatic action occurs when the following rad monitors reach their High Level setpoint:
c' )
Liquid Waste Process Monitor b)
Letdown Process Monitor c)
Gaseous Waste Process Monitor d)
CCW Process Monitor GUESTION 4.06 (1.00)
Which of the following is the RCP trip strategy in EOP-5(Excess Steam Demand), assuming CCW is maintained to the pumps, RCP operating limits are maintained and Pressurizer pressure falls to < 1300 psia?
a.
Trip all RCPs b.
Trip one RCP in each loop c.
Trip the two RCPs associated with the faulted S/G d.
Trip one RCP in the loop associated with the faulted S/G e.
Leave all four RCPs running l
(*****
CATEGORY 04 CONTINUED ON NEXT PAGE *****)
[
i t
4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 20
^^^^ I656LU65C5L C6UTR6L"'"'~'~ "
R QUESTION 4.07 (1.00)
What are the maximum allowable non-emergency whole body dose equivalent for on employee with a completed NRC Form 4 for the following time periods?
- o. In any calendar quarter.
b.
In any calendar year.
QUESTION 4.08 (1.80)
List the RCP Restart Criteria (6 items) as listed in the E0Ps.
QUESTION 4.09 (1.50)
What are the 5 symptoms that a SG Tube Rupture has occurred as listed in the Entry Conditions of E0P-4 (SGTR)?
QUESTION 4.10 (2.70)
List the NINE Safety Functions that are checked in Appendix A of the EOPs.'
QUESTION 4.11 (2.00)
If a UNIT 2 Blackout occurs, what 8 loads on the emersency buses are not tripped? (Similar loads like load centers or group A pumps are considered as the same load)
GUESTION 4.12 (2.50)
What actions at the control board must an operator take to move Shutdown Group A when bypassins a Shutdown Group Insertion CEA Motion Inhibit Interlock on Unit 2?
(***** CATEGORY 04 CONTINUED ON NEXT PAGE
- )
r 4.
PROCEDURES -~ NORMAL, ABNORMAL, EMERGENCY AND PAGE 21
~~~~ E656L66565L"66 TR6L
-~~~~~~~~~~~~~~~~~~~~~~~
R QUESTION 4.13 (1.50)
O.
Generator maintenance has been completed with all grounded equipment cleared of grounding devices.
AFTER verification of ground device removal, who (by position) is authorized to remove the caution tags on the equipment?
- b. How long can a Caution Tag remain in force before a status review must be made and reported to the Operations Supervisor?
' QUESTION 4.14 (2.50)
If it appears that critical conditions are going to be achieved 600 pcm LESS THAN the calculated ECC, what actions, verifications and notifications are required as stated in OP 0030126, 'ECC and ICRR'?
Assume any calculations are correct and without questionable data.
QUESTION 4.15 (2.00)
What are the EXIT CONDITIONS from the Standard Post Trip Actions (SPTA)?
QUESTION 4.16 (1.50)
What substeps constitute verification that plant electrical power requirements are satisfied in E0P-1?
QUESTION 4.17 (1.00)
What is the purpose of an asterisk (m) located next to a procedural step in an EOP?
QUESTION 4.18 (1.00)
Why are Core Exit Thermocouple temperatures > 700 de3rees F an almost certain indication of an uncovered core situation?
(*****
CATEGORY 04 CONTINUED ON NEXT PAGE
- )
.t
- a r
4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 22
~ ~~~ 565U E 55E L 55s TR5i ------------------------
R QUESTION 4.19 (1.50)
Select the THREE logs which an RCO is required to read as part of the chift turnover process, from the list of loss given below.
1)
NWE Los 2)
Night Order'Los 3)
Equipment Out of Service Log 4)
Turbine Operator Los 5)
Control Center Los 6)
RCO Los 7)
Nuclear Operator Log
(*****
END OF CATEGORY 04 xxxxx)
(***xxxxxxxxxx END OF EXAMINATION xxxxxxxxxxxxxxx)
Il.oQ I1
~
p.
S 6
.6.
e.
et.
g.
d Qe a.
Se to SO 90 let tSe 868 l
b.
I.
W W
W W
g g
W 90 6
Se see gas egg 5
C.
w 88 to 60 to see ese sie i
d, g
f w
w w
y w
y w
me 98 6
30 see tae 690 TIME ( hours)
A t1
'[j. crj 2
- i.
i l
i i
TEMP.
I<
(OR) 5 7
1 i,
t ENTROPY Figure 5-18
i e
i i
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 23
--- isEER557sARICs-REAi iEissFEE As5 FC0i5 FC5s ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WH ANSWER 1.01 (1.00) d REFERENCE f
Westinghouse Reactor Physics, Section I-5, HTC and Power Defect DPC, Fundamentals of Nuclear Reactor Engineering St Lucie Reactor Physics, Section 7.6 8 7.7 039/0008 A2.05(3.3/3.6)
ANSWER 1.02 (1.00) b REFERENCE General Physics, Heat Transfer Thermodynamics and Fluid Flow, pp. 145 - 160.
ST Lucie Thermo Handbook, Chapter 2d 002/000-K5.01 (3.1/3.4)
ANSWER 1.03 (1.00) b REFERENCE DPC, Fundamentals of Nuclear Reactor Engineering, p. 96 St Lucie Reacator Physics, Section 7.5 001/000-K5 56 (2.8/3.1) r of neutrons at some point in this ANSWER 1.04 (1.00) c REFERENCE Comprehensive Nuclear Training Operations (CNTO), pp 4-16/27 001/0008 K5.13(3.7/4.0)
b O
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 24
~~~~ 5EE566YUI5fC5I~5EIT IEIU5fER IU6~5LU56~FL6U
~
~
T ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM
\\
ANSWER 1.05 (1 00) 0-REFERENCE CNTO, ' Thermal / Hydraulic Principles and Applications, II", pp 10-45/48 006/0508 K5.01(2.9/3.1)
ANSWER 1.06 (1.00) c.
REFERENCE Oteam tables ANSWER 1 07 (1.50) 0)
Less
(+.5 ea) b)
More c)
More REFERENCE OT Lucie Reactor Physics, Section 7.5.1.2 2 001/0001 K5 26(3.3/3.6)
ANSWER 1.08 (1.50) 0)
Less
(+.5 ea) b)
More c)
Less REFERENCE St. Lucie SD 4, pp 36 and Reactor Physics Supplementary Handout 42 015/0208 K5 03 & K5 07(3 3/3.7)
5 s.
l y.
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 25
--- isEER55isARi5s-REKi iEsssFEs As5 FC0i5 FC55 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 1 09 (2.50) c) 5
(+0.5 ea) b) 2 c) 1 d) 7
l ANSWER 1.10 (1.00) 0)
Fission Chamber
(+.33 ea) b)
Uncompensated Ion Chamber c)
BF3 Proportional Counter REFERENCE STL SD 4, "Excore NIS', pp 56/7 015/0003 K6 01(2 9/3.2)
ANSWER 1.11 (1.00) o)
Xenon Decay
(+.5 ea) b)
Decreases REFERENCE St Lucie Reactor Physics Supplementary Handout il 001/0001 K5 33(3.2/3.5) and K5.34(2.1/2.2)
ANSWER 1.12 (1 50)
U-235 40-60%
(+.4 for isotoper
+.1 for contribution to power in order)
Pu-239 40-50%
U-238 5-9%
REFERENCE SON /WDN License Cert Trns, ' Reactor Kinetics', pp 6 St Lucie Reactor Physics Section 7.6 7) SD 1, pp 33 i
s, e..-
1.
PR'INCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 26
--- iAEER557nARICs, REAi isARsFEE As5 FC5i5 FE5s ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM 001/000; K5.47 (2.9/3 4)
ANSWER 1.13 (2.00) 0)
Unit 2
(+.E) due to a lower Beta coefficient at EOL (+.5) b)
Unit 1
(+.5) due to MTC being less negative, so Tavs must decrease more to add + reactivity)
(+.5)
REFERENCE CNTO ' Reactor Core Control", pp 3-21 a ' Fundamentals of Nuclear Reactor Physics', pp 7-31 ST LUCIE Reactor Physics, Section 7.5.1.2 & 7.6.7 001/000; K5.49(2.9/3.4) R K5.10(3.9/4.1)
ANSWER 1.14
(.75) 1)
Large Macroscopic Scattering Cross-section
(+.25 ea) j 2)
Low Macroscopic Absoption Cross-section 3)
Large Average Energy Decrement per Collision REFERENCE St Lucie Nuclear Physics, pp 7 3-34 001/0003 K5.55(3.0/3.2)
ANSWER 1.15 (1.00) remove boric acid that is precipitated on upper core surfaces (+.5) terminate any boilin3 or steam formation in upper head region (+.5)
N*
MN
~
REFERENCE Westinghouse PWR Systems Manual, pp 4.2-27 TPT SD-21, 'ECCS', pp 26 ST Lucie SD 24, pp 34 EPE-0111 EK3.13 (3.8/4.2)
I
1
~
,n
.1.
PR'INCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 27
~~~~ UER566YUd55C5,"_U5IT TRhN5FER dU6~FLUf6~FL6U
~
~
T ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WH ANSWER 1.16 (1.50)
- o. increases (+.5) due to decreasing Tave (+.25) b.
decreases (+.,5) due to decreasing Tave and increasing boron (+.25)
REFERENCE SON /WBN Nuclear theory ST Lucie Reactor Physics, Section 7.5 001/000iK5.09(3.5/3.7) & K5.26(3.3/3.6)
ANSWER 1.17 (1.00)
Lower flow at the same power level results in a larger delta T(+.25); CHF Ratio is the ratio of CHF to Actual Heat Flux at a specific location in the core (+.25).
The decrease in flow resuts in less strippins action to rcmove bubbles formin3 at nucleation sites on the cladding and therefore o steam film could form at the lower flow rates
.(+.5)
REFERENCE SON /WBN HTFF
,j ST Lucie Thermo Handout, Ch 2, Section E, Part 2.40 003/000; K5.01(3.3/3.9 )
ANSWER 1.18 (1.50) o)
ASI=(Ib-It)/(Ib+It)
(+.5) b)
At BOL, due to the colder water at the bottom of the core, ASI is positive due to the higher flux
(+.5).
As the core ages, the flux concentration shifts upward due to fuel depletion in the lower core regions, causing ASI to head in the negative direction (+.5)
REFERENCE ST Lucie Reactor Physics Supplementary Handout 42 015/020iK5.03(3.3/3.7)
1.
PR'INCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 28 T55R5 65Uk55C5,~55dT TREU5FER dU6~FEU56~FE6U
~
~
~~~~
ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 1.19 (2.00) o)
Delta T increases (+0.7) as That increases due to boiling and Teold remains relatively constant (+0.3) b)
Tcold will dot follow Psteam (+0.7) as Psta decreases due to boil off in the S/G while Teold will remain relatively constant (+0.3)
REFERENCE General Physics, HT & FF, pp 356/7 EPE-017; EK1.01(4.4/4.6)
ANSWER 1.20 (1.00)
ESI is the value of ASI in equilibrium conditions (+0.7) at the power level to which the reactor will be brought for continued operation (+0.3)
(The average ASI the core is oscillating around)
REFERENCE St Lucie OP 3200021, pp 1 001/000; K5.53(2.8/3.4)
ANSWER 1.21 (1.25) c)
Decay heat is dependent upon the production and subsequent decay of fission products
(+.5) b)
See attached sketch for grading criteria REFERENCE ST Lucie Reactor Physics, Section 7.2 001/000i K5.37(3.6/4.1)
ANSWER 1.22 (1.00)
See attached sketch REFERENCE ST Lucie Thermo Supplementary Handout, pp 3.4-15 Appendix A, Heat Exchangers (2.4/2.7)
k' s-g 3.. g y; e
8-t 8 !)a
~..J
,,.,. _ ma.*
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 29
~~~~iUEEs56Y 555C5I~5Edi~TEdU5EER dU6"ELU56~EL6U
~
ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 1.23 (2 00)
Teve 40.0 X 0.25 X -15 = -150 pcm
(+.5)
Power!
25 X -12 = -300 pcm
(+.5)
Veid:
-25 pcm Xcnon:
-50 pcm total!
-525 pcm
(+.5)
Baron:
-525 / -9 = 58.3 ppm dilution (accept 56 to 60)
(+.5)
REFERENCE SON /WBN Nuclear theory St Lucie Reactor Theory and plant curves 004/020 PWG-12(3.7/3.7) 0
- ~, _. _, - _
.r,r.,
I fission process were stopped, a considerable amount of e-heat woul.d still be transferred to the reactor coolant due
- (.-
to the time lag or delay in the decay process, measureable h
by the radioactive half-lives (tg) of the fission products.
4.
This phenomenon is an extremely important consideration following shutdown of an operating nuclear reactor since r.
the heat produced by the reactor core must be removed by j-the coolant. Thus, circulation through the core and heat removal. capability must be, continued even though the reactor r,,
is shutdown.
l k
It should be noted that.since this DECAY HEAT is dependent upon the decay of fission products, the total amount of decay heat available will be a function of the power a
opdrating history of the reactor prior to shutdown i.e..
ty the greater power output = greater number of fission products =
?'.-
However, it should also greater total amount of decay heat.
be noted that the amount of decay heat will decrease exponentially after shutdown, following the decay rate of the fission products. A typical curve of decay heat versus time after shutdown of a previously operating reactor, assuming 7%
of the operating power is attributable to fission fragment decay is shown in the following figure.
~~
..i
' Figure 7.2-4 DECAY HEAT PRODUCTION FOLLOKING SHUTDOWN FROM 1005 POWER.
E f
o.n w o.s eo.w l e-e h
(./, 7 i DOI I h}
r s
8 m4
+
Ea W
a
./,2T [W Ak war m
)
l*
I t
t
?
^
i.-
o oDol o.ol o.1 I.o 10 loo looo o'
TIME AFTER SHUTDOWN (HOURS) 7.2-13.,
g O
e.
,I
= - - -
a c:
Mm f%r MlDIV
\\
X\\
hMfMutett.
f
'%,wr
'T%,w 7
a.:
w.
i
.f Mft.bluth I
I Tc I war i
3 e
t Lww
(+, S)
Sac ke proSk cum fee coalty meswn
(+. D kr 4 4 pro % ca n fer celed webun 6
i 4
t
=.
i
+
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 30
\\
ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, W M l
CNSWER 2.01 (1.00) c REFERENCE ST. Lucie SD13 'CVCS'e pp 36 l
004/010I K6 13(2 1/2 4) l ANSWER 2.02 (1.00)
/b
' U 2 ', .' " L REFERENCE
CNSWER 2.03 (1.00) c REFERENCE ST Lucie SD145 '120VAC and 125VDC', pp 12/13 063/0001 K4 02(2.9/3.2)
ANSWER 2.04 TIT 5M -'
c)
Unit 1
(+.5 ea) b)
Both Units
- crttnt t-1 lj)jfi,i REFERENCE St Lucie SDB 'RCP', pp 17-19 004/0003 K1 03(3.3/3.6) i
n 1
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 31 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 2.05 (2.00) o)
Unit 1
(+.5 ea) b)
Unit 1 c)
Unit 2 d)
Both REFERENCE St. Lucie SD24 'SI and CNTMT HEAT REMOVAL SYSTEMS', pp 14-19 006/0003 K1.11(2.8/3.2), K4.02(2 8/3.0), K1.03(4 2/4.3) A K4.06(3.9/4.2)
ANSWER 2.06 (1.50)
Unit l's LPSI System (+.5) due to the combination of the pumps discharge into a common discharge line before separating into four injection lines
(+1.0)
REFERENCE St. Lucie SD24 ' Safety Injection and Heat Removal Systems', pp 17 006/0003 K4.18(3.3/3 8)
ANSWER 2.07 (1.00)
O REFERENCE ST Lucie SD13 'CVCS*, pp 38-40 004/0108 K4.03(3.1/3.6)
ANSWER 2.08 (1.50) c)
Unit 2
(+.5 ea) b)
Unit 1 c)
Both REFERENCE ST Lucio SD104 ' Main and Entraction Steam *, pp 11-19 039/0008 K1 01(3.1/3.2), K1.02(3.3/3 3) & K4.06(3.3/3.6)
4 t
~
i 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 32 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 2.09 (2.00)
- 1) Alignment keyway
(+.5 ea)
- 2) Between hot les discharge nozzles and upper core barrel outlets
- 3) Between baffle plates and core barrel (shroud annulus)
- 4) Around inserts in guide thimble tubes in the fuel assemblies REFERENCE NA NCRODP 88.1, 'RCS-Reactor Vessel / Core Construction
- W3stinghouse PWR Systems Manual
'RCS',
pp 2.1-39 ST Lucie SD2 'Rx Vessel Internals", Fig 32 002/0001 K6.13(2 3/2.8)
ANSWER 2 10 (2.00) o)
B Hot Les
(+.5 ea) b)
A2 and B1 Cold less c)
B1 Pump Suction REFERENCE ST Lucie SD7 'RCS', pp 20/21 002/0001 K1.06(3.7/4.0), K1.09(4.1/4.1) l ANSWER 2.11 (2 25)
VCT Outlet (MV-2501) shuts
(+.25 ea)
Baron Load Control Valve (MV-2525) Shuts I
Blend Valve (A0V-2512) Shuts BAMT Recict valves (A0V-2510/2511) Shut Baric Acid Strainer Inlet (A0V-2161) Shuts----UNIT 1 only 3
Doric Acid FCV (FCV-2260Y) Shuts---UNIT 2 only j
Ecergency Borate Valve (MV-2514) Opens Both Boric Acid Makeup Pumps Start
,(
Standby CCP(s) Start
- s. <. e. a... h r M' t 'u. ~ > pe os.
4g....,,n., e,
REFERENCE ST Lucie SD13 'CVCS', pp 47/8 004/0001 K1.15(3.8/4.0)
3 s
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 33 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 2.12 (1.50) 1)
Letdown Heat Exchanger
(+.25 es for any 6) 2)
Sample Heat Exchanger 3)
Shutdown Heat Exchanger 4)
Spent Fuel P. col Heat Exchanger 5)
RCP Seal Coolers 6)
HPSI Pump Seal Cooler 7)
LPSI Pump Seal Cooler REFERENCE ST. Lucie SD40 *CCW System', pp 25 008/000; K1.02(3.3/3.4)
ANSWER 2.13 (1.00)
To prevent uncontrolled pressurization
(+.7) due to the sizing of the RCGVS orifices in the vessel head and PZR vent lines (+.3)
REFERENCE ST Lucie SD7
'RCS', pp 52 002/020; PWG-7(3 7/4.3)
ANSWER 2.14 (1.00)
Allows partial ~ draining of the hot less for maintenance (+.5) without cousin 3 the runnin3 LPSI pump to lose suction (+.5)
REFERENCE ST Lucie SD7
'RCS',
pp 28 005/0003 K1.09(3.6/3.9)
ANSWER 2.15 (1.00)
The Recirculation Flow in Unit 2's CS Pumps is much higher (150 vs 50 spm) cnd is able to cool the seals without outside cooling (+1.0)
REFERENCE St. Lucie SD24 ' Safety Injection and CNTMT Heat Removal Systems", pp 21
a o
~
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 34 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM 026/000; K1 02(4.1/4.1)
ANSWER 2.16 (1 50) 1)
Inefficient at low Temps
(+.5 ea) 2)
Need strong radiation field 3)
Combustible gas--explosion hazard when primary opened up REFERENCE St. Lucie SD13 "CVCS", pp 20 004/000; K5.02(3.5/3.9), K4.01(2.8/3.3)
ANSWER 2.17 (1.00) 2 leakage detection tubes tap off the space between the inner and outer
'o*
rings
(+.5).
A pressure transmitter actuates at a setpoint of 2000 psia
(+.5).
REFERENCE ST Lucie SD2 ' Reactor Vessel Internals', pp 13 002/000; K4.05(3.8/4.2)
ANSWER 2.18 (1.00)
It is sized to a higher capacity to supply Unit 1
(+.7) in the event of a tornado missle rupturing UNIT l's CST
(+.3)
REFERENCE ST Lucie SD117
'AFW',
pp 11 061/000; K4.01(3.9/4.2)
ANSWER 2.19 (1.75)
(9
Their are two electro-hydraulic operated valves per feed line (+.
),
They are operated in fast speed in accident conditions by two'Nt p,,,j c accumul-stors
(+.5).
In normal ops (slow speed) they use a pneumatic pump (+.5)
A spool valve, solenoid air operated determines speed (+.25)
REFERENCE ST Lucie SD112 ' Condensate and Feedwater', pp 18 l
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 35 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM 059/000; K4.19(3.2/3.4)
ANSWER 2 20 (2.00) c)
MCC 2AB>> Rectifier >> Static Inverter >> Static Xfer Switch >> 120 VAC
(+.25 es component) b)
MCC 2AB>> Bypass Xfrmr(SOLA)>> Voltage Regulator >> Static Xfer Switch
>> 120 VAC
(+.25 ea)
-4
,' REFERENCE r,
.v *_ '.n.,.p p,12-14 ST Lucie SD144 '120 VAC Vital Distribution',
K4.09 ( 2.'4/2.9 )psy' s t r u 0C so s.n
- v.K
& K4.10(3.1/3.5) 5.
l 062/0001
! ANSWER 2.21 (1.50)
For UNIT 1,
when a single tank low level is receved, its appropriate f
colenoid valve opens and xfer pump starts until high level is reached (+.75)
For Unit 2r either tank low level will BOTH tank' solenoid valves open and tha appropriate pump starts.
EACH tank's solenoid valve shuts at a high j
level and the pump stops when BOTH are filled (+.75)
REFERENCE ST Lucie SD137 'EDG and Auxiliaries', pp 26 064/0005 K1.03(3.6/4.0) h/g[. 0.tc) 6 tWChi'(
U N't (a) gy_ :. x +..a s % &.+, 2 itc g :u w
. -- 4 f ul j y / ( V il U 4
p,e !.: Y 4 M,wa,J w
- a. k n e t v,r M 9 0
('Q W C 7 A 5 @,4 :.E. u d 5,',_ t $'M 4 l' ' UG i N5*'~
1
/sdebu R4 h 16U M# # "
a j.%dti: v.. a. 4 6 s zu o
+. -. - -
- - - -. - _ -. -, ~.,
-..--------,-----,---.-.r--y
tr
--s t
3.
INSTRUMENTS AND CONTROLS PAGE 36 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WH ANSWER 3.01 (1.00) c REFERENCE ST Lucie SD11 "S/G Water Level Control', pp 21 035/010; K4.01(3.6/3.8)
ANSWER 3.02 (1.00) d REFERENCE ST Lucie SD108 "SBCS*, pp 18 041/020; K1.06(2.6/2.9)
ANSWER 3.03 (1.00)
D REFERENCE ST Lucie SD62 'RPS",
pp 36 012/0003 K6.07(2.9/3.2)
ANSWER 3.04 (1.50) c)
Unit 2
(+.5 ea) b)
Both c)
Unit 2 REFERENCE ST Lucie SD9 'PZR Pressure, Relief and Level Control", pp 33-36 010/000; K4.03(3.8/4.1)
~, --
l.' f I
~ '
l
- n. -. p, a
...,.. e ; - %; * <
- .,r n.
,, s f'.....,
l 3.
INSTRUMENTS AND CONTROLS PAGE 37 ANSWERS -- S1. LUCIE 1
-86/06/10-DEAN, WM s.
/
ANSWER 3.05 F1 r50 ) '
a)
Unit 1
(+.5 ea) b)
Unit 2
-c)--UniPi
./EC0.Qy REFERENCE ST Lucie SD5 "CEDS', pp 9, 29 001/0001 K4.07(3.7/3.8), kl.03(3.4/3.6)
ANSWER 3.06 (1.00) c)
False
(+.5 ea) b)
False REFERENCE ST Lucie SD20 'ESFAS', pp 10/11 013/000; K6.03(2.4/2.9) a pWG-1(3.8/4.0)
ANSWER 3.07 (1.50) o)
Not (e.5 ea) b)
Bumpless 4
c)
Bump'.ess REFERENCE
.*T Lucie SD11 "S/G Water Level Control", pp 29 035/010; A4.01(3.7/3.6)
ANSWER 3.08
(.75)
Reactor Trip, 806 psia, 560 des F
(+.25 ea)
[
REFERENCE ST Lucie SD108 'SBCS', pp 17 I
041/020; K4.17(3.7/3.9) l t
l
,su..
f 3.
INSTRUMENTS AND CONTROLS PAGE 38 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN. WM ANSWER 3.09 (1.00) o)
No Indication
( +. 2 5 e a ) Ob " / " ##
b)
Red c)
Blue d)
Amber REFERENCE St Lucie SD5 'CEDS", pp 27/28 001/000; K4.01(3.5/3.8)
ANSWER 3.10 (1.75) 7
-Minimum letdown at -1.1% deviation
(+/-
we% on setpoints)
-First standby charging pump starts at -2.5% dev.
-Second
-4.2% dev.
-Low level alarm at -5% dev.
-Back up start signal to standby charging pumps at -5% dev.
-All heaters off at 27% level
-Lo-Lo Level alarm at 27% level
(+.2 for action,
+.05 for setpoint ea)
REFERENCE I
ST Lucie SD9 'PZR pressure, relief and level control", pp 54-56 011/0001 K1.01(3.6/3.9) and A2.03(3.8/3.9)
({pmMT/ f ANSWER 3.11 Cold Les Temperature ( +.45 ea r esponse )
Hot Les Temperature 3
Turbine First Stage Pressure
-- Pressurizer Pressure --A.M WWs Nuclear Power Unit i uses 2 RTD cold les inputs per loop, Unit 2 han only 1 per loop REFERENCE ST Lucie SD 15
'RRS',
pp 5 001/0005 K1.04(3.2/3.4) & K1.05(4.5/4.4)
Y.::s-
=
n. c. wn-
> 3 ",.
a i-
'f 3,
?-
_a-3.
INSTRUMENTS AND CONTROLS PAGE 39 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 3.12 (2.50) c)
2/4 Hi Cntet Pressure (5 F313)
(+ 4 for loSice
+.1 setpoint) er 2/4 Hi Cntet Radiation (10 R/hr)
( 485 p s is ) se dt? d i'*, ?gw?,, n y A or 51AS b) 2/4 Low S/G Fressure 2/4 High Cntat Pressure (5 psig)> 'I'M M,,',:h N REFERENCE ST Lucie SD20 'ESFAS'epp 17, 20
' 013/0003 K1.01(4.2/4.4)
ANSWER 3.13 (1.50) 1)
RCS Saturation Margin
(+.5 ea) 2)
Upper Head Saturation Margin 3)
CET Saturation Margin REFERENCE ST Lucie QSPDS Handoute. pp 14 EPE-074; EA1.13(4.3/4.6)
ANSWER 3.14 (1.50) 1)
Group out of sequence
(+.3 ea) 2)
Group deviation 3)
Power dependent insertion limit reached 4)
Loss of regulating group withdrawal permissive 5)
Loss of shutdown group insertion permissive REFERENCE i
St Lucie SD5 'CEDS", pp 30/31 001/000; K4.07(3.7/3.8)
c.=*
.n..
',9<=.,,
.c.. an e i r~ s u s. -.0) f r r m.
3.
INSTRUMENTS AND CONTROLS PAGE 40 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, W H ANSWER 3.15 (1.50) g Tha inputs are the comparator averagere' power ratio calculator and conbined linear amp signal from each safety channel via the isolation oseeably.
(+.75)
The averager output'is compared to the individual channel signals and the setpoint calculator and alarms at 3% deviation (hi) cnd 7% deviation (hi-hi) (+.75)
REFERENCE ST Lucie SD4 "Excore NIs', pp 23/24
.'015/0003 K6.04(3.1/3.2)
ANSWER 3.16 (2.00) a)
The indication is based on the position of a solenoid valve (whether it is energized or not) which allows system pressure to actuate a pilot valve (+1.0) b)
An acoustic monitoring system is used that produces a LED indication that relates system flow as measured by an accelerometer to a bar graph of flow indication for each valve. (+1.0)
REFERENCE ST Lucie SD9 'PZR Pressure, Relief and Level Control", pp 32 and 39 010/000; A2.03(4.1/4.2) & A4.03(4.0/3.8)
ANSWER 3.17 (1.00)
~
A heated thermocouple's output is compared with an unheated thermocouple's output.(+.3)
With no liquid to remove the heat, the temperature difference becomes large, creating a large voltage output (+.7)
REFERENCE ST Lucie QSPDS Handout, pp 16 017/020iK4.01(3.4/3.7)
.v
.v
.nr...
. s., :..
.>.;.t
_ ~..,, _
1 l
?-
3.
INSTRUMENTS AND CONTROLS PAGE 41 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM l
1 ANSWER 3.18 (3.00) c)
Ensure the operator does not deliberately or inadvertantly dcfeat parts of the protection system by switch misalignment (+1.0)
(
b)
High SUR; Loss of Loaci Locci Power Density
(+.33 ea) c)
Must turn 4 keylock switches located on each RPS panel (+.7) and is
,,, - y '.
bypassed above.1% E.5% Unit 23
(+.3) sa
'. ', l'T '
.,.,.. x.,,
ANSWER 3.19 (1.50)
}
l A Range control circuit energizes an output relay (+.5) when it senses
< 1000 cps (+.25) to select the extended range output (+.25).
A red 4*
lamp above the % power meter illuminates in the extended range (+.5)
(Note:
The opposite of the above for wide range selection will be also allowed)
REFERENCE ST Lucie SD4 "Excore NIs'r pp 13/14, 17 015/000; K6.01(2.9/3.2)
Y ANSWER 3.20 (2.00) 1)
(+.5 ea) 2)
Logic Matrix Relay AB (AC or AD) 3)
Trip Paths 4)
Trip Circuit Breaker Control Relays REFERENCE ST Lucie SD62 "RPS", pp 19-23 012/000; K6.03(3.1/3.5) l l
t I
r
l i
4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 42
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~ 56 UE6656dE 66 TR6L R
ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 4.01 (1.00) d REFERENCE ST Lucie E0P-1, pp 7 EPE-0158 EK1.01(4.4/4.6)
' ANSWER 4.02 (1.00)
-)
c o REFERENCE ST Lucie EP 0030141 PWG-11(EOP Immediate Actions) (4.3/4.4)
ANSWER 4.03 (1.00) e)
Unit 1
(+.5 ea) b)
Unit 1 REFERENCE STLucie DP 1/2-0030127, pp 1, 7
002/020; PWG-12(3.7/3.7)
ANSWER 4.04 (1.50) a)
False
(+.5 ea) b)
False c)
True REFERENCE ST Lucie E0P-1, pp 11; EOP-2, pp li E0P-8, pp 1 PNG-11(Perform Immediate Actions) (4.3/4.4)
.i.s
?...
s..s i
.y t
l 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 43
--- sE5iBE55iBLE 55RiE5E------------------------
ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 4.05 (2.00) o)
FCV-6627X shut (+.5 ea) b)
No Auto Action c)
Valve V6565' Shuts d)
CCW Surge Tank Vent Valve (RCV-14-1) diverts from atmosphere to chem drain tank REFERENCE ST Lucie SD 68 ' Radiation Monitors' 073/000; K4.01(4.0/4.3)
ANSWER 4.06 (1.00) b REFERENCE ST Lucie E0P-5, Steps 7-9 4
EPE-074; EK3.04(3.9/4.2) 1 ANSWER 4.07 (1 00)
- o. 3 Rem b.
5 Rem REFERENCE St. Lucie 10 CFR 20 PWG-15(Radcon Knowledge)
(3.4/3.9) t l
n.
's
(
). L' g, 1
~ T 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 44
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~ 5656L665CEL"66UTR6L R
ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 4.08 (1.80) 1)
CCW not loss > 10 Minutes (+.25 for parameter, +.05 for numbers) 2)
RCS between 20 and 200 degrees subcooled 3)
PZR level > 30%
4)
S/G Wide Range Level > 40%
5)
Rx Vessel Level > 50%
6)
RCP yellow permissive light lit REFERENCE ST Lucie various E0P steps EPE-074; PWG-10(4.2/4.7)
ANSWER 4.09 (1.50) 1)
Condenser Air Ejector Rad Monitor Alarm
(+.3 ea) 2)
S/G Blowdown 3)
Hi activity in a S/g per sample 5)
Increasing S/G 1evel REFERENCE ST Lucie E0P-4, ppi EPE-038; PNG-10(4.5/4.5)
ANSWER 4.10 (2.70) 1)
Reactivity Control
(+.3 ea) 2)
Vital Auxiliaries 3)
RCS Inventory Control 4)
RCS Pressure Control 5)
Core Heat Removal 6)
RCS Heat Removal i
7)
Containment Isolation 8)
Containment Pressure / Temperature 9)
Containment Combustible Gas REFERENCE l
ST Lucie E0P Appendix A (various EOPs)
PNG-10(Recognize abnormal conditions) (4.1/4.5) l
~- s.
~
>: n : s..,
ey
- 3
</
I 9
4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 45
~~~~~~~~~~~~~~~~~~~~~~~
~~~~ A5i5L65iEAL 55RTR6L R
ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, W M ANSWER 4.11 (2.00) 1)
HPSI Pumps
(+.25 ea) 2)
SI Motor Operated Valves 3)
Emergency Li'shting 4)
Class I Emergency Power Panels 5)
Diesel Oil Transfer Pumps 6)
'A' RCP oil lift pumps 7)
Sups Power Inv rter 8)
HVAC Valves and Dampers REFERENCE ST Lucie EP 0030140 EP E-056 8--EK3. 01-( 32 5/3 r9 )~' +',a) r-a *'(.c : - L 3 1 ANSWER 4.12 (2.50) 1)
Hold the CMI bypass pushbutton depressed the ENTIRE time bypass action is desired
(+.5 es step) a 2)
While depressing the CMI bypass buttone depress the Bypass Enable 4
momentarily 3
3)
Select
'A' on the Group Select Switch 4)
Select Manual Group on Mode Select Switch 5)
Move the joystick in the Insert direction REFERENCE ST Lucie SD5 'CEDS", pp 33 001/010; A4.01(3.7/3.4) l
(~\\.S ju ' Y",~.' Led-u.OY-\\
,) y eb g/,.u-p V' ANSWER 4.13 4 1.3 0r+
- c. The responsible Foreman or an Electrical Department Supervisor. [0.5ea]
b.
One month CO.5]
REFERENCE St. Lucie AP 0010135 PWG-14(Tagging / Clearances) (3.6/4.0)
i 9
x i
4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 46
~
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~ 56 6L6656dL 66UTR6L R
ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 4.14 (2.50) 1)
Stop withdrawal and insert CEAs to 500 pcm below critical condition 2)
(Recheck ECC){not required as statement in question may preclude >
3)
Recheck bordn concentration by chemistry sample 4)
Verify CEA position 5)
Verify agreement between NI channels 6)
Notify reactor engineer C+.5 es step-2.5 total 3 REFERENCE ST Lucie OP 0030126, pp 1/2 001/010; A2.07(3.6/4.2)
ANSWER 4.15 (2.00) l 1)
Any of the Rx Trip Safety Functions status check acceptance criteria not met (+.5)
OR (+.25) 2)
All of the Safety Functions being maintained (+.5)
AND (+.25)
RCS conditions are being controlled / maintained in Mode 3 (+.5)
REFERENCE ST Lucie E0P-1 (SPTA)
EPE-007; PWG-10(4.1/4.2)
ANSWER 4.16 (1.50) 1)
Turbine tripped
(+.3 ea) 2)
Generator DCBs open 3)
Exciter Field Breaker open 4)
Electrical Auxiliaries transferred to S/U Xfrmrs 5)
At least 1 DC bus energized REFERENCE ST Lucie EOP-1, pp 4, step 2 EPE-007; PWG-11(4.0/4.0)
- r H ?fi4
- A ;)~ <
a-rsw w ri'**-
- 3. o.
,v, p.. m.. : _..
)
c s
i I*
?
4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 47
~~~~ d6 6L6656dL 66NTR6L R
~~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WH ANSWER 4.17 (1.00)
It is a step that must be monitored / performed continuously while in that procedure. (+1.0)
REFERENCE ST Lucie Notes on pp 2 of all E0Ps i
PWG-11(EOP usage) (4.3/4.4) i ANSWER 4.18 (1.00)
Since saturation Temperature for the RCS safeties is < 700 des F, this I
would represent a superheat condition in the RCS which could only occur with core uncovery (+1.0) i REFERENCE CEN-152, pp 5-77 for LOCA discussion i
EPE-074; EK1.02(4.6/4.8) t ANSWER 4.19 (1.50) l-2, 3,
6,(+.5 ea) fl/(mc.
p g,(-) 10,pp
~
k' 3
I c
I ENCLOSURE 2 j
U.
S.
NUCLEAP PLGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
ST. LUCIE 1 REACTOR TYPE:
PMR-CE DATE ADMINISTERED: 06/06/10 EXAMINER:
DEAN, W M APPLICANT:
IMSTRUCTIOMS TO APPLICANT:
Use separate paper for the answers.
Write answers on one side only.
Steple question sheet on top of the answer sheets.
Points for each questien cre indicated in perentheses after the question. The passing grade requires at least 70% in each cetesory and a final grade of at least 80%.
E:<a.hina tion paper s will be p i c k e d u p s i:-: (6) hours after the examination starts.
% OF CATEGORY
% OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY
_ _ q _g _ _ _ _ _ _ gc 7
-3AO4-N
_l____
________ 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
,5
.1'l b 25 M
2 BOO
________ 6.
PLANT SYSTEMS DESIGN, CONTROL, AND IhSTRUMENTATION yq 30 0
___I__0__ _I'["i((
________ 7.
PROCEDURES - NORMAL, ADNORMAL,
^^-
EMERGENCY AND RADIOLOGICAL CONTROL
.> 'f --
.,t y
-6%OO~
2W
________ 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS lth4 4 20 rOO ^
100.00 TOTALS FINAL GRADE _________________%
All work done on this e:: amination is my oun. I have neither given nor received aid.
~~~~~~~~~~~~~~
dPPL5C5UII5~5IUU5TUEE 1
1
x f.
1 OF NUCLEAR POWER PLAN 1 Or FLUIDS, AND PAGE 2
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _' E R A 1 10 N,____________________
5.
THEORY puphtNE A*Y GUESTION 5.01 (1.00)
The reactor is critical at 10,000 cps when a S / G JWMP7 f a il s op en.
Assuming BOL conditions, no rod motion, and no reactor trip, choose the answer below that best describes the values of Tavs and nuclear power for the resulting new steady state.
(POAH = point of adding heat).
a.
Final Tavs greater than initial Tav3, Final power above POAH.
b.
Final Tave greater than initial Tavs, Final power at POAH.
c.
Final Tave less than initic1 Tav3, Final power at POAH.
d.
Final Tavs less than initial Tavs, Final power above POAH.
QUESTION 5.02 (1.00)
Which of the following will cause plant efficiency to increase?
a.
Total S/G blowdown is changed from 30 spm to 40 spm.
b.
Steam quality changes from 99.7% to 99.9%.
c.
Level increase to higher than normal in a feedwater
- heater, d.
Absolute condenser pressure changes from 1.0 psi to 1.5 psi.
QUESTION 5.03 (1.00)
Reactivity is defined as which of the following?
a.
The ratio of the number of neutrons at some point in this generation to the number of neutrons et the same point in the previous generation.
b.
The fractional change in neutron population per seneration.
c.
The factor by which neutron population changes per genera-tion.
d.
The rate of change of reactor power in neutrons per second.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
4
f.
0~
l
\\
=
)
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3
QUESTION 5.04 (1.50)
A motor driven centrifugal pump is operating at a low flow condition.
You then start opening the throttle valve on the discharge side.
Ilow will each of the following be affected?
(Increase, Decrease or No Change) a)
Discharge Pressure (0.5) b)
Available NPSH (0.5) c)
Motor A n.p s (0.5)
GUESTION 5.05 (1.00) a Which of the followin3 curves (see attached page) representing Xenon concentration is correct for the given pouer history?
DUESTION 5.06 (2.00) a)
What time in core life, what temperature and what accident are the limiting factors for the Shutdoun Margin (CDM)?
b)
What is the definition of SDM?
OUESTION 5.07 (1.50)
Indicate what design accident the following Reactor Trips are designed to protect against:
a)
Variable Power Level-High b)
Steam Generator Level-Low c)
Reactor Coolant Flow-Low
(***** CATEGORY 05 CONTINUED ON NEXT PAGE
- )
fs r
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4
QUESTION 5.08 (1.00)
Which of the following deceribes where the maximum fuel centerline temperature occurs in a core with a synmetric axial neutron flux about the core midplane.
a.
Top of the core b.
Between the top of core and the core midplane c.
Core midplane d.
Between core midplane and the bottom of the c o r.e e.
Bottom of the core GUESTION 5.09 (1.50)
An ECC is calculated for a startup following a reactor trip from 50%
power equilibrium xenon (BOL).
Indicate if the actual critical rod position will be HIGHER, LOWER or the SAME from the calculated position for each of the followins situations.
Use attached curves as appropriate and treat each case individually.
a)
Xenon reactivity curve for trip from 100% is used to calculate conditions to startup 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after the trip.
b)-Eroton worth ~ at EOL-is' uced instead of worth at BOL.-
82b c)- -Thr'EOL-critic ~al' BoFbh' Concentr ation is -used instead of_ the-BOL.
cr iti_ cal Baron--concentr ation.
QUESTION 5.10 (1.50)
For the changes listed below (treat each one independently) indicate whether the moderator temperature coefficient will become MORE NEGATIVE, LESS NEGATIVE or have NO EFFECT. (Assume all other parameters are constant) a)
Neutron flux peak shifts radially inward from the edge of the core.
b)
Boron concentration decreases 100 ppm while core is at MOL.
c)
Increased number of burnable poisons are inserted into the core.
(*r**r CATEGORY MS CONTINUED ON NEXT PAGE
- )
ORY OF NUCLEAR POWER PLANT OPERA 110N, FLU 10S, AND PAGE 5
lN 5 11 (1.00) following TS LCOs relating to core parametes pertain te whether the
'T 1, UNIT 2 or BOTH UNITS' nimum Temperature for criticality of 515 degrees F.
in Mode 2.
,vtdown Margin > 3.6 % Delta k/k while See JujJkkk puej& a4 M of i
5'*
i rods out, a rod drops to mid f%dM operating at 80% power with all position and sticks there and no reactor trip occure ee attached as for position of the stuck rod).
What wov}
ppen to the following
'c 1 80%? { Provide definitions or factors essuming that core power ach nuclear factor discussed}
adial nuclear facto a fuel assembly in Guadrant I.
nial peakir-actor of a fuel assembly adjacent to the stuck rod.
cea aking factor for a fuel assembly on the periphery of rent III.
- 0N 5.13 (2.00) just rectarted following a refueling outage while Unit 2 is
'1 hss EOL.
Answer the following regarding the differences in plant response the two units (explain your answers).
g:nt e steady power level of 10EE(-8) amps during a startup, equal cactivity additions are made (approximately 100 pcm).
Which Unit ill have the higher steady state startup rate?
ht 50% power, a control rod (100 pcm) drops.
Assuming NO RUNDACK or
@PERATOR ACTION, which Unit will have the lower steady state Tavg?
80N 5.14 (1.00) the four parameters monitored to assure the operator that DND is not rring.
~
CATEGORY 05 CONTINUED ON NEXT PAGE *****)
f (tr***
m&*
e
W 5.'
THEORY OF NUCLEAR POWER PLAN 1 OPERATION, FLUIDS, AND PAGE 6
QUESTION 5.15 (1.50)
During the performance of an emergency boration while at power, how and why are the following parameters affected? (assume no control rod movement) a)
subcooling b) control rod worth QUESTION 5.16 (1.00)
Explain how decreasing RCS flow (at constant power) will result in decreasin3 Critical Heat Flux Ratio.
QUESTION 5.17 (1.00)
Which of the curves on the attached page correctly indicates the axial relationship of CHF to Actual Heat Flux along a typical fuel rod?
QUESTION 5.18 (1.50) a)
Provide _an equation representing Axial Shape Index (ASI).
(0.5) b)
Explain how ASI changes over core life and why these changes occur.
(Ensure you discuss ASI conditions at BOL also)
(1.0)
QUESTION 5.19 (1.75)
Sketch the reactor coolant channel Boiling Water curve, indicating the-four main boiling regions and the point of DNB.
Ensure you label the axes correctly.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
f I
t
V 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 7
QUESTION 5.20 (1.00)
Indicate what type of detector is used in the following encore instruments:
a)
Unit 1 Wide ranse Logarithmic Safety Channel b)
Unit 2 Linear Safety Channel c)
Unit 2 Losarithmic Startup Channel QUESTION 5.21 (1.25) a)
Why is Decay Heat dependent upon power history of the core?
(0.5) b)
Using the los paper provided, sketch the curve of decay heat vs. time for the first 4 days following a trip from power.
Assume no abnormal conditions exist in the core. { Note: use the los scale for time}
(.75)
GUESTION 5.22 (1.00)
Make a sketch of the temperature profile along the length of a counter flow heat exchanger for both the cooling medium and the fluid being cooled.
(*****
END OF CATEGORY 05
- )
V 6.
' PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 8
QUESTION 6.01 (1.00)
Which of the following statements correctly describes how to reset the MECHANICAL overspeed on the turbine driven AFW pump?
a.
It must be reset locally, and then the limitorque driven in to the closed position to relatch the linkage.
b.
-It must be reset locally by using a lever to relatch the linka3e+
c.
It will reset automatically as turbine speed decreases below a pre-determined setpoint.
d.
It may be reset locally, but can also be reset f r o r., a switch on RTGB 102 in the control room.
QUESTION 6.02 (1.00)
Which of the following correctly describes the normal lineup of the 125 VDC Swing Bus?
a.
The swing bus AB is powered from the A-side train in UNIT 1.
b.
The C train is powered from either of the A or B buses via the swing bus AB in UNIT 1.
c.
The swin3 bus AB is powered from the A-side train in UNIT 2.
d.
The swing bus AB is supplied by either a battery charger or a battery backup in UNIT 2.
(*****
CATEGORY 06 CONTINUED ON NEXT PAGE **xyr)
Y b
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMEN1ATION PAGE 9
DUESTION 6.03 (1.00)
Which statement below correctly describes the effect of the condenser vsevum interlock on the Steam E:yPass Control System?
a.
If ALL M/A stations are in AUTO when vacuum is regained, then the Condenser Vacuum Reset button must be depressed to remove the interlock.
b.
ONLY if the Master Integrated Controller is in Manual, is it required to depress the Condenser Vacuum Reset button to remove the interlock.
c.
It makes no difference if the M/A stations are in Manual or Auto, when the condenser vacuum is regained, the interlock is removed automatically.
d.
If ANY of the M/A stations are in manual when vacuum is regained, then the Condenser Vacuum Reset button must be depressed to remove the interlock.
QUESTION 6.04 (1.00)
As it applies to the Core Protection Calculator, which of the following dascribes Thermal Power?
a.
Average Thot - Highest Teold b.
Average That - Average Teold c.
Highest Thot - Highest Tcold d.
Highest Thot - Lowest Teold e.
Highest Thot - Average Tcold x
QUESTION 6'.05 (1.00) a, O
Which of the following wi.11 revent CEA C e hanism from operating?
/
a.
The CEA change mechanism fuel aligner is DOWN b.
The CEA change mechanism Stapple is NOT EXTENDED c.
The CEA, change mechanism _is NOT at its UP LIMIT d.
The RFM is in the CEA ZONE x_
/
(rzwr* CATEGORY 06 CONTINUED ON NEXT PAGE
- r)
I s
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 10 OUESTION 6.06 (1.50)
Indicate to which Unit the following descriptions of RCP seal injection apply.
a)
The Regenerative Heat Exchanger inlet is interlocked with the Seal Injection Isolation MOV.
b)
Durins Back-up Seal Injection operations, two charging pumps are needed c)
Durins cold shutdown, a temporary seal injection line can'be run'from the tube side drain of the regenerative heat exchanger. -7CbD CAVbOWJ hM;;}
Tu Auttutt B W D *^'
Com mons PdeotTO QUESTION 6.07 (1.50)
- eq740f, Which UNIT's LPSI System would be more susceptible to a single failure (passive or active) causins it to fail to inject on a SIAS?
Explain your answer by describins what system failure would cause this to occur.
QUESTION 6.08 (1.50)
Indicate whether each of the followins statements regarding the overpressure mitigation system (OMS) applies to UNIT 1, UNIT 2 or to BOTH UNITS.
a)
There are FOUR pressure comparators (PC's) that are used as inputs to Channel A and Channel B for low pressure protection, one channel per PORV.
b)
The PORV low pressure protection system is designed to prevent '
exceeding the Tech Spec PTS limits if an inadvertant Safety Injection by two HPSI pumps and three charging pumps.
c)
A temperature interlock prevents PORV actuation due to a low pressure relief open signal if RCS temperature is > 320 degrees F.
QUESTION 6.09 (2.00)
Indicate what automatic action occurs when the following rad monitors reach their High Level setpoint a)
Liquid Waste Process Monitor b)
Letdown Process Monitor c)
Gaseous Waste Process Monitor d)
CCW Process Monitor
(***** CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx)
f
,6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 11 OUESTION 6.10
(.75)
Fill in the blanks below to correctly complete the statement regarding SCCS operation:
A Quick Opening Si3nal is Senerated subsequent to a if both secondary pressure is > ______ psia and Tavs > _____ des F.
QUESTION 6.11 (1.00)
What happens to the Main Steam Isolation Valves on a loss of electric power and a loss of air supply (treat each case individually)?
OUESTION 6.12 (1.00)
Describe how many fans and at what speed they will run for BOTH UNITS when en SIAS starts the Containment Cooling System.
QUESTION 6.13 (2.25)
List the eight automatic actions that occur upon an SIAS that align the CVCS to a flow path for maximum boration.
(Note: actions affecting similar components, such as both RCS inlet valves opening, count as one action)
NOTE ANY DIFFERENCES BETWEEN UNIT 1 and UNIT 2 ACTIONS!
OUESTION 6.14 (2.50)
List ALL the logics that will cause the following ESFAS subsystems to cetuate. (Include applicable setpoints) o)
Containment Isolation Actuation Signal (CIAS)---- UNIT 1 b)
Main Steam Isolation Signal (MSIS)---- UNIT 2 OUESTION 6.15 (1.50)
List the three Saturation Margin Indications that are available on the Saturation Margin Monitorin3 (SMM) Page of the GSPDS.
(*****
CATEGORY 06 CONTINUED ON NEXT PAGE *****)
B; f
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 12 GUESTION 6.16 (1.50)
List ALL the CEA motion inhibiting signals that are bypassed when the UNIT 1 Motion Inhibit Bypass button is depressed.
QUESTION 6.17 (1.00)
When performing a fill and vent operation on the RCS in accordance with OP 0120020, why must the fill rate be maintained below 500 spm?
Indicate the limiting component (s) that is/are the basis for this precaution.
QUESTION 6.18 (1.00)
Why do the UNIT 1 Containment Spray Pumps require Seal Water Heat Exchangerst while the UNIT 2 Containment Spray Pumps do not?
GUESTION 6.19 (1.00)
Why is the storage capacity of UNIT 2's Condensate Storage Tank (CST) significantly hi3her than UNIT l's CST? (ie. what is the design basis that
.accounts for this difference)
GUESTION 6.20 (2 00)
D3 scribe the flowpath from the appropriate MCC to the 2A 120 VAC Vital Instrument Panel for the followin3 situations * (Identify ALL components) c)
Normal b)
Maintenance Conditions GUESTION 6.21 (3.00) a)
What is the purpose of the Power Trip Test Interlock?
b)
What trips are automatically bypassed by the RPS?
c)
How is the Zero Power Mode Bypass initiated and when is this bypass automatically removed?
(***** END OF CATEGORY 06 xxmax)
a 7.
PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 13 RA656L655 CAL 6UUTR5L
~~~~
GUESTION 7.01 (1.00)
Which one of the following is NOT a requirement for increasing UNIT 2 RCS pressure greater than 1750 psis on aplant heatup/startup?
a.
Two independent containment spray systems shall be OPERABLE with each system takin3 a suction from RWT on an SIAS.
b.
At least 3 SI Injection Tanks OPERABLE.
c.
Two independent ECCS subsystems OPERABLE, consisting of at least 1 HPSI, 1 LPSI and 1 charging pump.
d.
RCS dissolved oxygen concentration <
.1 ppm.
QUESTION 7.02 (1.00)
Which of the following locations is where the Asst Nuclear Plant Supervisor reports to on a Control Room Evacuation due to i'ahabitability?
a.
Remote Shutdown Panel i
b.
Electrical Equipment Room, Reactor Auxiliary Building c.
Turbine Operating Level d.
LPSI Pump Room e.
6.9 KV Switchgear Room GUESTION 7.03 (2.00)
~
Indicate whether the following Immediate Actions for Emergency Boration as stated in OP 1/2 0250030 apply to UNIT 1, UNIT 2 or to BOTH UNITS.
a)
Place Makeup Mode Selector Switch in Manual or Borate position b)
Close BAMT Recites (V-2650/51) c)
Verify Load Control Valve is Closed (V-2525) d)
If Emers. Borate Valve fails to open, open Gravity Feed Valve (V-2508)
(*****
CATEGORY 07 CONTINUED ON NEXT PAGE *****)
)
7.
PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 14
~~~~ A5i5L55iCAt C5NTR5L'~~~~~~~~~~~~~~~~~~~~~~~
R QUESTION 7.04 (1.50)
Answer the following questions TRUE or FALSE regarding E0P usage:
a)
While conducting the Standard Post Trip Actions (SPTA), if a safety function criteria is NOT met before ALL immediate actions are completed it is permissable to exit E0P-1 to go to the approriate E0P.
b)
The STA should perform the Safety Function Status Checklist (App. A) at least once every 30 minutes while any EOP is being performed.
c)
When performing an assessment of the Safety Functions while in E0P-8, you must assess ALL of them before referring to the Resource Trees (Appendices B-I) contained in the back of E0P-8, even if a Safety Function criteria is NOT met.
QUESTION 7.05 (1.00)
Which of the following is the RCP trip strategy in EOP-5(Excess Steam Demand), assuming CCW is maintained to the pumps, RCP operating limits are maintained and Pressurizer pressure falls to < 1300 psia?
a.
Trip all RCPs b.
Trip one RCP in each loop c.
Trip the two RCPs associated with the faulted S/G d.
Trip one RCP in the loop associated with the faulted S/G e.
Leave all four RCPs running QUESTION 7.06 (2.00)
Answer the following questions regarding the Post-Trip Review procedure OP 0030119:
a)
If the cause of the plant trip can NOT be identified, what action is required before a startup may commence?
b)
What two individuals SIGN the Restart Authorization if the cause of the trip is identified?
QUESTION 7.07 (1.80)
List the RCP Restart Criteria (6 items) as listed in the E0Ps.
1 i
(***** CATEGORY 07 CONTINUED ON NEXT PACE ****m) l l
l
i 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15
~~~~ 5656L 65C5L C UTR6L'~~~~~~~~~~~~~~~~~~~~~~~
~
R GUESTION 7.08 (2.70)
List the NINE Safety Functions that are checked in appendix A of the E0Ps.
QUESTION 7.09 (2.00)
If a UNIT 2 Blackout occurs, what 8 loads on the emergency buses are not tripped? (Similar loads like load centers or group A pumps are considered as the same load)
GUESTION 7.10 (1.50)
List. the Immediate Action Substeps in E0P-1 that you must perform to verify that Core Heat Removal is occurring.
Include any applicable parameters.
QUESTION 7.11 (2.50)
What actions at the control board must an operator take to move Shutdown Group A when bypassin3 a Shutdown Group Insertion CEA Motion Inhibit Interlock on Unit 2?
GUESTION 7.12 (2.00) a)
What is the allowable time ' window" within which an ECC calculation is considered valid?
(0.5) b)
What three criteria require that an Inverse Count Rate Ratio (ICRR) be performed when taking the reactor critical?
(1.5)
GUESTION 7.13 (2.00)
What are the EXIT CONDITIONS from the Standard Post Trip Actions (SPTA)?
(***** CATEGORY 07 CONTINUED ON NEXT PAGE
- )
a v
~~
7.
PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 16 Rd65bl6G5 EEL EUUTR5L------- T---------------
~
'~~~~
OUESTION 7.14 (1.00)
~
What is the basis of performing an RCS cooldown to < 525 Degrees F and depressurizing to < 900 psia when in E0P-4 (SGTR) before. isolating the affected S/G?
OUESTION 7.15 (1.00)
What is the purpose of an asterisk (*) lo'cated ne5[t to a procedural step L
in an E0P?
OUESTION 7.16 (1.00)
Why are Core Exit Thermocouple temperatures > 700 degrees F an almost certain indication of an uncovered core situation?
GUESTION 7.17 (2.00)
Answer the followin3 questions regarding RCP Trip Criteria:
a)
Which location is worse for a LOCA assuming that the'RCPs are still operating, Hot Les or Cold Les?
(0.5) b)
Assuming that RCP operating limits are met, CCW is flowing to the RCPs and pressurizer pressure falls below 1300 psia, why do the SGTR (EOP-4) and Excess Steam Demand (EOP-5) procedures have you maintain 1 RCP operable in each loop?
Give,.three reasons.
(1.5)
I
. ~
- ~.,
GUESTION 7.18 (2.00) a)
What condition is indicated if, while lowere'ing a fuel element into the core, a Dillon cell UNDERLOAD condition jecurs?
~
b)
List two situations where the Bridge Speed 'is NOT limited to slow
^
speed.
's
?
(***** END OF CATEGORY 077xrr**)
f a
d "
W m
V w
[
l 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 17 QUESTION 8.01 (1.00)
Which one of the followins is NOT in an LCO at Unit 2?
a.
The reactor coolant system lowest operating loop temperature (Tave) shall be =or> 515 desF when the reactor is in modes 1 or 2.
b.
The calculated value of TFR (total integrated radial peaking factor) shall be limited to =or<
1.70.
c.
Primary containment internal pressure shall be maintained between -0.7 and +2.4 psis.
d.
The pressurizer shall be operable with a minimum water level of =or>
27% indicated level and a maximum water level of =or< 68% indicated level.
QUESTION 8.02 (1.00)
During a Unit 1 startup with the reactor at 2% power, one power range nzutron flux monitoring channel is found to be inoperable.
Which of the followins statements is correct?
Refer to the attached Tech Specs.
a.
Operation above 5% rated thermal power is not allowed until the inoperable channel is repaired and declared operable.
b.
If the inoperable channel is placed in a tripped condition and the other three channels are operable, you must verify compliance with the shutdown margin requirements of Tech Specs.
c.
If the inoperable channel is placed in a tripped cond. tion and.the other three channels are operable, operation to 100% rated thermal power may proceed only if all functioning units receiving an input from it are tripped (may not be bypassed).
d.
The only restriction on proceeding to 100% rated thermal power are that the inoperable channel be placed in a bypassed or tripped condition; however, if it is bypassed it must be only for surveillance testing and must be tripped within an hour.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE
- )
a e
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 18 4
GUESTION 8.03 (1.'00)
Which one of the following is correct regardin3 EP-310029Er ' Duties of an Individual Who Discovers an Emergency Condition'?
a.
All significant information, events and actions taken during the emergency period must be recorded by the emergency coordinator.
b.
Primary examples of emergency conditions covered by this' procedure are tornado, acid and caustic spills, security threats and most other emergencies not involving radioactive material.
c.
An individual who discovers an emergency condition shall notify the nuclear plant supervisor and the nuclear watch engineer.
t d.
The individual should take any immediate action that will minimize the emergency (eg, close an upstream valve or locally stop machinery) except extinguishing firesi fires shall only be fought by the fire brigade members.
GUESTION 8.04 (1.00)
With the reactor head closure bolts less than fully tensioned, the boron concentration of the RCS and refuelin3 canal shall be maintained uniform and sufficient to ensure that the more restric,tive of which of the following reactivity conditions is met on Unit 2?
a '.
Keff of..O.95 or boron concentration =or> 1720 ppm.
b.
Keff of 0.95 or boron concentration =or> 2100 ppm.
s c.
Keff of 0.97 or baron concentration =or> 1720 ppm.
d.
Keff of 0.97 or baron concentration =or> 2100 ppm.
(*****
CATEGORY 08 CONTINUED ON NEXT PAGE
- )
1 1
m
..m._.,
.g
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 19 GUESTION 8.05 (1.00)
Diesel Generator 1A, which supplies 4KV Bus 1A3 is INOPERABLE.
LPSIP B supplied by 4KV Bus 183 is INOPERABLE.
The Tech Specs for ECCS and AC Sources are attached.
Which statement is CORRECT concerning continued operation in Mode 1?
- a. The Action Statements for both the LPSIP and the DG are applied indepen-dently, each must be restored to operable in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
b.
Since the DG is required in Mode 4 and the LPSIP is not, the Unit must be taken to Mode 4 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
c.
LCO 3.0.3 applies.
- d. LCO 3.0.5 applies.
QUESTION 8.06 (1.00)
Unit 2 is in Mode 3 during a Reactor startup with the following deficiencies:
One Main Steam Isolation Valve is inoperable and closed One Motor Driven Aux. Feedwater Pump is inoperable Which one of the following actions most accurately details the allowances and/or limitations imposed by the Tech Specs in this instance?
- c. Mode 3 must be maintained (Entry into Mode 4 acceptable) b.
Startup activities may continuel Mode 2 may be entered but not exceeded.
c.
Startup and power operation into Mode 1 may be accomplished provi.ded Mode 1 action statement for MSIV met.
d.
Startup activities may continue into Mode 2 provided subsequent restoration of the MDAFW pump to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
(*****
CATEGORY 08 CONTINUED ON NEXT PAGE
- )
g
. 8..
ADMINISTRAIIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 20 QUESTION 8.07 (2.00)
Complete the following table for Unit 2:
MINIMUM SHIFT CREW COMPOSTION With Unit 1 in Mode 1,2,3 or 4 Number of Individuals Required to Fill Position Position Mode 1,2,3 or 4 Mode 5 or 6 SS (SRC)
a------
f------
b------
g------
c------
h------
d------
i------
STA N.
e------
j------
QUESTION 8.08 (2.00)
List the accident which the followins reactor trips are designed to protect against.
a)
P=r Pressure High b)
S/G Pressure Low c)
Thermal Margin / Low Pressure d)
Local Power Density High QUESTION 8.09
(.50)
Fill-in the BLANK A monthly surveillance requirement of Tech Specs may be extended up to t
days without declaring the component inoperable due to the surveillance testing not being performed.
QUESTION 8.10 (1.00) i i
i
. Tech Spec. 3.2.5 DNB Parameters, gives limits for four DNB parameters. In addition to being a DNB parameter which DNE of the four ensures peak temperture of the fuel cladding will not exceed 2200 deg. F in the event o v' a LOCA?
4
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 21 GUESTION 8.11 (1.00)
What are the maximum allowable non-emergency whole body dose equivalent for an employee with a completed NRC Form 4 for the following time periods?
- a. In any calendar quarter.
b.
In any calendar year.
QUESTION 8.12 (1.00)
Fill-in the BLANK with the appropriate Tech Spec definition.
"A ______ shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify operability including alarm and or trip functions".
QUESTION 8.13
(.50)
Fill-in the BLANK for the following:
In accordance with 10 CFR 55 'if a licensee has not been actively perform-ing the function of an Operator or Senior Operator for a period of _______
conths, or longer, he shall, prior to resuming activities licensed pursuant to this part, demonstrate to the Commission that his knowledge and understanding of facility operations and administration are satisfactory.'
GUESTION 8.14 (1.50)
Prior to evacuating an injured person from the plant site during a local area evacuation:
a.
What items should be removed from the person?
(1.0) b.
To whom should they be given?
(0.5)
QUESTION 8.15 (1.50) i St Lucie Unit 2 TSs allow temporary changes to the fire protection program l
inplementing procedures if three conditions are met.
State the three l
conditions.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
i l
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 22 QUESTION 8.16 (2.00)
Tech Spec 3.5.1 Safety Injection Tanks list four conditions which determine operability of the tanks.
LIST these four conditions (values not required)
GUESTION 8.17 (1.50)
With regard to Administrative Procedure AP-0005720; a.
What are the requirements for an operator to transfer from Unit 1 to Unit 2?
(1.0) b.
Under what conditions can the requirements in (a) be waived?
(0.5)
GUESTION 8.18 (1.00)
The ', Refueling Machine Operation
- procedure 2-1630024 cautions,the operator to exercise' care when approaching one speed none from the other..
,/
a.
Is this extra care required when going from slow to fast speed zones or from fast to slow speed zones?
~
/
.jc f s U b'.
Explain.
x GUESTION 8.19 (2.00)
AP 0010120 provides for an informal shift turnover.
a.
Under what conditions may this be utilized in lieu of a formal shift turnover?
(0.5) b.
What information, as a minimum, must be given to the person's relief?
(1.5)
(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE *****)
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 23 QUESTION 8.20 (2.00)
During Mode 1 operation of both units it is found that 2 of 4 channels for Pressurizer Pressure high Reector trip are inoperable (on each unit) due to a generic material deficiency (repair time 14 days).
Using Tech Spec 3.3.11 provided, determine what actions must be taken as a result of this failure.
Include seperate discussions for each unit. State specific LCO/ action steps which apply.
QUESTION 8.21 (2.00) c.
A maintenance technican is performing troubleshooting which requires the individual to hold a jumper across two wire termination points for an inoperable Diesel Generator.
Is this action permitted by Administrative
- ? Explain.
Procedure no. 0010124 " Control and use of Jumpers
- b. Per Administrative Procedure no. 0010124 under what two CONDITIONS can independent verification be waived?
QUESTION 8.22 (1.50) a.
Generator maintenance has been completed with all grounded equipment cleared of grounding devices.
AFTER verification of ground device removal, who (by position) is authorized to remove the caution tags on the equipment?
b.
How long can a Caution Tag remain in force before a status review must be made and reported to the Operations Supervisor?
QUESTION 8.23 (1.00)
' Shutdown Margin shall be the Toch Specs defines Shutdown Margin as...
instantaneous amount of reactivity by which the Reactor is soberitical or would be soberitical from its present condition assuming STATE the assumptions made for the plant conditions which complete the definition of Shutdown Margin.
(*****
END OF CATEGORY 00
- )
(*************
END OF EXAMINATION ***************)
1
- f. o s,J u
e-i i
9 O
to s
sie eie sie i
G.
\\
~;
)
W W
W W
se go de go aos ese see
.i i
tG.
=
D sie sie S'
sisi I
.h C.
Se se te to see one ese
=
O M
W p
T y
y y
y y
y se go se 30 see sae
- 98 TIME ( hours)
L
w w
u.t e: ::..:.t.t.:.::.=
w O -(.
N q. jijcp., ST LUCI E o.
UNIT l CYC L E.fi
- 7 7h, piniiii i
i 'm: :
i N
N. Y i.!
BORON WORTH
. j-i i
l, Hii l l l llf i
i q,N i
i i
nl I
ll i
In. !! 1 O!
ALL RODS IN ( AR1) gi i i i
!l i
l li
'a.
i-l I
! ! i l i ll ![
i if I
i l l
J
.'l jh.ps i..
I i
l' l
l l
i
[
l i
1 c
I 2 se L
o_
I 1
d
)
N
- m U p g c:
2 l
.L 4
a o
ji I
J a
n V-
- ,H--
(:-
L 3
.11 m,.3
- d~
t i 3 g
i
'dU l '"
p l
tr i
i O
ig',
3
]
l j
mg I
1 i
Z J
iI L.
g 9
O
-, i;,
OK
-1:
I l
I i
',! Ifff l !
V O h.4
'O 4
- VJE4A.Jt
(*'f 3-a-s9
_]-
~
q L
y j
a 1
1
. 3-i FIG B.6
-i STLU l CYCLE L Li I
SHUTDOWN BORON CONC.
ARI-I i
NO XE.
MODESl-4 3.6 7o SHUTDOWN MARGIN w
llilii I
o-m i
j R
I a
' hMP v
C O
"y H'"
g i-i..
--qll Lu g
r.
u i
C i
ai Ou w
l!
Ill)d' ll l"l
'bd go L
i i
I'-
i O
m p
=,16
.g
~,.
-I',
i l'fh
.j.'l.
I 1 l l
P 10':
! l' ll1-;
4lI lil ModeratorTemperature(OF) s _ h.w;
WA: k~iGikT WinYd'u 'ig; eoaasu ST LU i FIG B.7 CYCLE _6__
SHUTDOWN BOROH odHC. w.BURNUP MODES I-4 532 F TAG.
NO XE.
-l clG D
..I
- {
.,H q;.
- t ih~lC
[...,
,o I
a.
Q.
I I Fl >O
'"II.,
Z 9
lMH:l t<-
E F--
l z
j W
60::
l I
0
-n..'
i Z
O O
Z y
O
~n'l O
m I.
[
D I,
Ti l
l l l
l
.l!!!
l
.h:
y
'9lif' s s 9:!
j j-
.q j
9 <
m o
a i s o
l n 2
m l
il.. l[j BURNUP (EFPH)
[
]
l l
l
..,o.,
i'
- b -
{'
il,l[ti[If!!:
-.l;ii
,l
- jl!
!II!
.)
' p 5
8 I3 I
H H
l
~
1 r
ts i
)
4
, h'l i
l H
P o
N m'
F O
E l
P N 1
h'
(
U E m
l H
N T
X I
R G
T U
l.
N l
l E
B l
I o
l l
l N 2 O
l L
l l
U N. E C
S O V E
R E
I C
. O L
U GB O
o l
C I
3 L.
F R
R l
Y i
l l
L A
C T-A S
C o
I T'.
I 4)
T P
I R
H
_ M C
j1! ;
l
_ M i
i i,"!i I
' i I
' i Pii t. 'I liI j
l Q
f i
fl ll$!Y l
,a j-
~
lljj!
k
[1 !t
,!i i
1.!
h
!l E
l
?
lli H
il l
l 1
l l $l jI i
.ql-
.I iI I
l i j
l I'
iI I
ili I d ll
,'llE-,
l-q IkI!I l
!l i
l
.6)
M,III lQ 'l I
j r.
i!
l r,-
<;~-
u-
'.!l m!
A!!
x
)
L lrIr!
l.
i
[
d rlli iI 1
l i
. j i
i s0OozoO zoEOm a<oEgo
~
, i.
(
!l;!'
I
.i!!
l
lc j
{
- tl !!
l,j ii[,
,i
!j l
f:,
,; l lilii
!!I l
i i
=
e o,
_ ~'
l g
l l
d i
g
'l m
h.
1 1
11 1
1 w
/
V i
l i
l 1
i
/
F I
6 l
l l
4 l
P l
F-6_
H l
i P
l E
U L
o, l
N C
R N
H Y H U T
O C T B
=
R N
's G
I N )H s
C Ov E
l E
i P
W X
l L
l F
G Q
EE I
F N
E L(
O C
e i
R U O Y
1 LB O
C 1
1 T
A 1
R 3'.
S 1
"L o
1 1
l u,
1 O.
I 1
T' A
s T
j i
m I-l b
jl 4
~
1 l
i G
H
\\
l
\\
l l
l I
I
{
lj I[l N
jl r'
tll l
i
[i i illL F
i l'
IL
.li l
{_
I l
l j..
l
[I l
7 i'l l
Iil 4
l l
il iiI n
LIl[.
6.
4 lI
.I y'r1 f
1 Il 1
1 r.
. yroo I
(,.-
-.q g;
,lL
~
i 4.
Reactor Flux Distribution (continued) r
% \\\\
fccM5^
Q>W Q
(&nY)
I REACTOR VESSEL i
THERMAL SHIELD g
j s.
X y
.g g il a
I 1
i l
M CORE 8ARREL' l
p h
FUEL ASSEM8 LIES K
xw Figure 4.6-1. Symmetry of Fuel Assemblies in Core l
l l
l l
i
_m-,,
m g--y
-ww
.w
.em-yw,,c--
_,pg
CNF $
()cwAL Wref FLvX !
\\
Y N
f N
P A
\\
ft*
N N N. \\_
l IRL h5 sT800 M Q y g gg, Q g y p p,,
I (vit Rob l
l p.U f L Rob
[
'b y
dp N
x p.
f#
N N
s N
' s~
s
~
IN 05' S
[ x f.4c Po c ertoa ~
l ML Rn l
l run son
(
l z.
l
ll OA) If O DL TS
- ' - ^ - -
3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICA8ILITY LIMITING CONDITION FOR OPERATION j
i 3.0.1 Limiting Conditions for Operation and ACTION requirements shall be i
applicable during the OPERATIONAL MODES or other conditions specified for each specification.
t 3.0.2 Adherence to the requirements.cf the Limiting Condition for Operation
)
and/or associated ACTION within the specified time interval shall constitute compliance with the specification.
In the event the Limiting Condition for Operation is restored prior to expiration of the specified time interval, i
j,-
completion of the ACTION statement is not required.
j 3.0.3 In the event a Limiting Condition for Operation and/or associated ACTION requirements cannot be satisfied because of circumstances in excess of those addressed in the specification, the facility shall be placed in at least HOT STAND 8Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, at least NOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and at least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless corrective measures are completed that permit operation under the permissible ACTION statements for the specified time interval as measured from initial discovery or until the reactor is placed in a MODE in which the specification j
is not applicable.
Exceptions to these requirements shall be stated 1.n the i
indivit!ual specifications.
3.9.4 Entry into an OPERATIONAL MODE or other specified applicability condi-tion shall not be made unless the conditions of the Limiting Condition for l
Operation are met without reliance on provisions contained in the ACTION state-ments unless otherwise excepted. This provision shall not prevent passage through OPERATIONAL MODES as required to comply with ACTION statements.
1 j
3.0.5 When a system. subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable. it may be considered OPERABLE for the purpose of satisfying the re uirements of its applicable Limiting Conditio~n for Operation. provided:
- 1) its corresponding normal or l
emergency power source is OPERABLE; and (2) all of its redundant system (s).
subsystem (s), train (s), component (s) and device (s) are OPERABLE or likewise satisfy the requirements of this specification.
Unless both conditions (1) l and (2) are satisfied the unit shall be placed in at least HOT STAND 8Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and.at least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
This specification is not applicable in MODE 5 or 6.
SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.
ST. LUCIE - UNIT 1 3/4 0-1 Amendment No. 40
- ~*
... -.... ~
I t,,
3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION, LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor protective instrumentation channels and bypasses of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.
APPLICA81LITY: As shown in Table 3.3-1.
' ~
ACTION:
As shown in Table 3.3-1.
SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor protective instrumentation channel shall be I
demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-1.
4.3.1.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by, bypass operation.
The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.
4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip.
function shall be demonstrated to be within its limit at least once per 18 months.
Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1; 9
ST. LUCIE - UNIT 1 3/4 3-1
7
,e 1
TABLE 3.3i..
,A E
REACTOR PROTECTIVE INSTIRMENTATION i
M MININM l
TOTAL N0.
CHANNELS CHANNELS APPLICABLE' i
FUNCTI0tiAL UNIT OF CHAfflELS TO TRIP OPERABLE MODES ACTION I
-i 1.
til 2
1,2.and-*'
1 2.
Power Level - High 4
2(a) 3(f)
- 1. 2 2f
- 3.. Reactor Coolant Flow - Low 4/SG 2(a)/SG 3/SG 1,2(e) 27j j
4.
Pressurizer Pressure - High 4
2 3
- 1. 2 2f I
4 i
w 5.
Contains nt Pressure.- High 4
2 3
- 1. 2 27 j
I l
y 6.
Steam Generator Pressure - Low 4/SG 2(b)/SG 3/SG
- 1. 2 2f
.o 7.
Steam Generator Water Level - Low 4/SG 2/SG 3/SG 1, 2 2f-e' l
8.
Local Power Density - High 4
2.(c) 3 1
2f
]
9.
Thermal Margin / Low Pressure 4
d(a) 3 1,2(e) 27
- p 9a, Steam Generator Pressure 2,
Difference - High 4
2(a) 3 1.2(e) 2f-
.g
- 10. Loss of Turbine--Hydraulic i
Fluid Pressure - Low 4
2(c) 3 1
Rf t
g i
l c
i e
f.
TA8LE 3.3-1 (Continued)
"~
REACTOR PROTECTIVE INSTRUMENTATION E
D MINIltjM m
TOTAL NO.
CHANNELS CHANNELS APPLICABLE E
FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION q
. _ 11. Wide Range Logaritheit Neutron Flux Monitor
~
a.
Startup and Operating--
-Rate of Change of Power -
High 4
2(d) 3 1, 2 and
- 27 b.
Shutdown 4
0 2.
3, 4, 5 3
2 4
1, 2*
4 Logic
~
g
'. 13. Reactor Trip Breakers 4
2 4'
1, 2*
4 w
4, F
3 g
2
?
8 M
i G
T I
I TABLE 3.3-1 (Continued)
TABLE NOTATION
- With the protective system trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.
- The provisions of Specification 3.0.4 are not applicable.
(a) Trip may be bypassed below 15 of RATED THERMAL POWER; bypass shall i
be automatically removed when THERMAL POWER is > 15 of RATED
~
THERMAL POWER.
(b) Trip may be manually bypassed below 685 psig; bypass shall.be..,,,
automatically removed at or above 685 psig.
3
~
(c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 155 of RATED THERMAL POWER.
(d) Trip may be bypassed below 10~45 and above 15% of RATED THERMAL POWER; gpass shall be automatically removed when THERMAL power is > 10 5 or < 15% of RATED THERMAL POWER.
I (e) Trip may be bypassed during testing pursuant to Special Test Excep-
!y tion 3.10.3.
t-(f) There shall be at least two decades of overlap between the Wide Range Logarithmic Neutron Flux Monitoring Channels and the Power Range Neutron Flux Moni.toring Channels.
ACTION STATEMENTS ACTION 1 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requiremen.t.
restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the protective system trip breakers.
With the number of OPERABLE channels one less than the ACTION 2 Total Number of Channels STARTUP and/or POWER OPERATION -
may proceed provided the following conditions are satisfied:
The inoperable channel is placed in either the bypassed a.
or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in
~
the tripped condition ~. '
ST. LUCIE - UNIT 1 3/4 3-4 Amendment No. U, N, 45 9
-w-----
.----w,
l' i
TABLE 3.3-1 (Continued)
ACTION STATEMENTS i
i b.
Within one hour, all functional units receiving an input from the inoperable channel are also placed in thesamecondition(eitherbypassedortripped,as i
applicable)asthatrequiredbya.aboveforthe inoperable channel.
.f,
. c.
The Minimum Channels OPERABLE requirement is met; however. one additional channel any be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while perfoming tests and maintenance on than channel provided the other inoperable channel is placed in the tripped condition.
ACTION 3 With the number of channels OPERA 8LE one less than required l
by the Minim m Channels OPERABLE requirement, verify compliance with the SWTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2. as applicable, within I hour and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> twreafter.
ACTION 4 With the mober of channels OPERABLE one less than required l
i f
by the Minim m Channels OPERA 8LE requirement, be in HDT
$TAND8Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for surveillance testing per Specification 4.3.1.1.1.
\\
j I'
i i
ST. LUCIE - UNIT 1 3/4 3-5 Amendment No. ;
27
l l
3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
SAFETY INJECTION T'ANKS LIMITING CONDITION FOR OPERATION i
l l
3.5.1 Each reactor coolant system safety injection tank shall be OPERA 8LE with:
e.
The isolation valve open.
- b. ' 8etween 10f0 and 1170 cubic feet of borated water, i
c.
A minimum boron concentration of 1720 PPM, and i
d.
A nitrogen cover-pressure of between 200 and 250 psig.
i s
I APPLICA8ILITY: MODES 1, 2 and 3.*
ACTION:
a.
With one safety injection tank inoperable, except as a result I
~
of a closed isolation valve, restore the inoperable tank to
' OPERA 8LE status within on's hour or be in HOT SHUTDOWN within f
the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
b.
With one safety injection tank inoperable due to the isolation valve being closed, either innediately open the isolation valve or be in HOT STAND 8Y within one hour and be in HOT SHUTDOWN within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
i:.
SURVEILLANCE REQUIREMENTS 4.5.1 Each safety injection tank shall be demonstrated OPERABLE:
a.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by:
l!
1.
Verifying the water level and nitrogen catver-pressure in i
the tanks, and 2.
Verifying that each safety injection tank isolation valve j
i is open.
l
- With pressurizer pressure 3,1750 psia.
ST. LUCIE - UNIT 1 3/4 5-1
g 1
i 3/4.8 ELECTRICAL POWER SYSTEMS l
, (e 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION i
3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
a.
Two physically independent circuits between the offsite trans-mission network and the onsite Class 1E distribution system, i
ad b.
Two separate and independent diesel generator sett each with:
1.
Engine-mounted fuel tanks containing a minimum of 152 i
gallons of fuel, i
2.
A separate fuel storage system containing a minimum of 16.450 gallons of fuel, and j'('
3..
A separate fuel transfer pump.
i APPLICA8ILITY:
MODES 1, 2, 3 and 4.
ACTION:
\\
a.
With either an offsite circuit or diesel generator set of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.4 within one hour and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; restore at least two offsite circuits and two diesel generator sets to OPERA 8LE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STAND 8Y 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 />.
b.
With one offsite circuit and one diesel generator set of the above required A.C. electrical power sources inoperable.
demonstrate the OPERA 8ILITY of the maaining. A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.4 within one hour and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter;
~~
restore at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least HOT STANDBY 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 ST. LUCIE - UNIT 1 3/4 8-1
I. o-ELECTRICAL POWER SYSTEMS i
ACTION (Continued) hours.
Restore at least two offsite circuits and two diesel generator sets to 0PERA8LE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from the time of initial loss or be in at least HOT STAND 8Y 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 />, c.
With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel enerator sets by
(..-
performing Surveillance Requirement 4.8.1.
2.a.4 within one hour and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, unless the diesel generator sets are already operating; restore at least one of the inoperable offsite sources to OPERA 8LE status l
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
With only one offsite source restored, restore at i
least two offsite circuits to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial loss or be in at least HOT STAND 8Y within i
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 />.
j d.
With two of the above required diesel generator sets in-operable, demonstrate the OPERABILITY of two offsite A.C.
circuits by performing Surveillance Requirement 4.8.1.1.1.a within one hour and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter;.....
4 l
restore at least one of the inoperable diesel generator sets i
to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STAND 8Y 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 />.
Restore at least two diesel generator sets to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time'of inttial
~
loss or be in at least HOT STAND 8Y 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 />.
SURVEILLANCE REQUIREMENTS
~
l
\\
4.8.1.1.1 Two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system shall be determined OPERABLE at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying correct breaker alignments and indicated power availability.
l e
\\
ST. LUCIE - UNIT 1 3/4 8-2 a
/d 9
S M9 0
1
i l r l
i 3/4 LINITING CON 0!TIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the i
succeeding specifications is required during the OPERATIONAL M00E5 or o';.her conditions sp'ecified therein; except that upon failure to meet the Lieiting l
Conditions for Operation, the associated ACTION requirements shall be met.
3.0.2 Noncompliance with a specification shall exist when the requirements of the Lietting Condition for Operation and/or associated ACTION requirements are i
not met within the specified time intervals.
If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, l
completion of the ACTION requirements is not required.
t 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, action shall be initiated to place the unit in a MDDE in which the specification does not apply by plar:ing it, as applicable, in:
1
!b 1.
At least NOT STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, I
2.
At least HDT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and i
3.
At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time 4
i limits as measured from the time of failure to meet the Limiting Condition.for i
Operation.
Exceptions to these requirements are stated in the individual i
specifications.
j This specification is not applicable in M00E,5 or 6.
I j
3.0.4. Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions of the Limiting Condition for Operation are met j
without reliance on provisions contained in the ACTION requirements.
This j
provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION statements.
Exceptions to these requirements are stated in the individual specifications.
e m i
s
)
ST. LUCIE - UNIT 2 3/4 0-1 i
3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION
~
LIMITING CONDITION FOR OPERATION 3.3.'1 As a minimue, the reactor protective instrumentation channels and bypasses of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.
APPLICA8ILITY: As shown in Table 3.3-1.
ACTION:
As shown in Table.1.3-1.
i 3URVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor protective instrumentation channel shall be demonstrated OPERA 8LE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-1.
4.3.1.2 The logic for the bypasses shall.be 6monstrated OPERA 8LE prior to l
each reactor startup unless performed doring the preceding 92 days. The tetal bypass function shall be demonstrated OPEKA8LE at least once per 18 months during CHANNEL CALIBRATION testing of each char.nel affected by bypass operation.
4.3.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME af each reactor trip function shall be demonstrated to be within its limit at least once per 18 months.
Each test shall includa at least ene channel per fur.ction such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific M acto.- trip function as shown in the
" Total No. of Channels" column of Table 3.3-1.
s.
S e
ST. LUCIE - UNIT 2 3/4 3-1
i a
s l
i I
~
TABLE 3.3-1
[
M i
REACTOR PROTECTIVE INSTRtsENTATION 2
r-3 MINIORM
~
m TOTAL NO.
CHAMIELS CHANNELS APPLICABLE FUNCTIONAL TWIT OF CHANNELS TO TRIP OPERABLE IWOES ACTION i
1.
2 4
1 2 1
I3, 4*, 58 5
-4 4
2 4
]
2.
Variable Power Level - High 4
2(a)(d) 3 1, 2 28 l
3.
Pressurfrer Pressure - High 4
2 3
1, 2 N
l 4.
Thermal Margin / Low Pressure 4
2(d) 3 1, 2 N
i i
5.
Containment Pressure - High 4
2 3
1, 2 N
6.
Steam Generator Pressure - Low 4/5G 2/5G(b) 3/5G 1, 2 2#
7.
Steae' Generator Pressure Difference - High 4
2(a)(d) 3 1, 2 N
l
,M 8.
Steam Generator Level - Low 4/5G 2/5G 3/5G 1, 2 28
[
9.
Local Power Density - High 4
2(c)(d)-
3 1
2f
- 10. Loss of Component Cooling Water l
to Reoctor Coolant Pumps 4-2 3
1, 2 N
- 11. Reactor Protection System Logic 4
2 3
1 2 28 I3, 4*, 5*
5
- 12. Reactor Trip Breakers 4
2(f) 4 I 2 4
3g 4 *
- 13. Wide Range Logarithmic Neutron Flux Monitor l
a.
Startup and Operating -
Rate of Change of Power -
i High 4
2(e)(g) 3 1, 2 2f l
b.
Shutdown 4
0 2
3,4,5 3
l
- 14. Reactor Coolant Flow - Low 4/5G 2/5G(d) 3/5G 1, 2 28
- 15. Loss of Load (Turbine
)
Hydraulic Fluid Pressure - Low) 4 2(c) 3 1
2r i
t I
/
l
1 1
TABLE 3.3-1 (Continued)
TABLE NOTATION With the protective system trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor v'essel.
The provisions of Specification 3.0.4 are not applicable.
(a) Trip may be manually bypassed below 0.5% of RATED THERMAL POWER ir, con-junction with (d) below; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 0.5% of RATED THERMAL POWER.
'(
(b) Trip may be manually bypassed below 705 psig; bypass shall be automatically removed at or above 705 psig.
(c) Trip may be bypassed below 15% of RATED 1MERMAL POWER; bypass shall be i
automatically removed when THERMAL POWER is greater than or equal to 15%
of RATED THERMAL POWER.
(d) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.
(e) Trip may be bypassed below 10N and above 15% of RATED THERMAL R;
bypass shall be automatically removed when THERMAL power is 3,10 or "r
< 155 of RATED THERMAL POWER.
(f) Each channel shall be comprised of two trip breakers; actual trip logic shall be one-out-of-two taken twice.
(g) There shall be at least two decades of overlap between the Wide Range Logarithmic Neutron Flux Monitoring Channels and the Power Rangd Neutron Flux Monitoring Channels.
ACTION STATEMENTS ACTION 1 With the number of channels OP.ERABLE one less than required by the Minimum Chanswls OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HDT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the protective system trip breakers.
7 G
I' ST. LUCIE - UNIT 2 3/4 3-3 mnow,,-- - - --,,,v.---w,-wn,---,---w----
,,,w,,
.e-,--e,,,,x---n-=
w-~,ms-
W TABLE 3.3-1 (Continued)
ACTION STATEMENTS a.
With the number of channels OPERA 8LE one less than the ACTION 2 Total N aber of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
If the inoperable channel is bypassed, the desirability of maintaining this channel in the bypassed condition shall be reviewed in accordance with Specification 6.5.1.5m.
The channel shall be returned to OPERA 8LE status no later than during the next COLD SHUTDOWN.
b.
With the number of channels OPERABLE one less than the Minime Channels OPERA 8LE, STARTUP and/or POWER OPERATION may continue provided the following conditions are satisfied:
1.
Verify that one of the inoperable channels has been bypassed and place the other inoperable channel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
2.
All functional units affected by the bypassed / tripped channel shall also be placed in the bypassed / tripped condition.
With a channel process measurement circuit that affects multiple functional units inoperable or in test, bypass or trip all associated functional units as listed below:
Process Measurement Circuit Functional Unit Bypassed 1.
Safety Channel - Nuclear Variable Power Level - High (RPS)
Instrumentation Local Power Density - High (RPS)
Themal Margin / Low Pressure (RPS)
Rate of Change of Power,High (RPS) l 2.
Pressurizer Pressure -
Pressurizer Pressure - High (RPS) j Nigh Thermal Margin / Low Pressure (RPS) 3.
Containment Pressure -
Containment Pressure - High (RPS) i l
High Containment Pressure - High (ESF) 4.
Steam Generator Pressure - Steam Generator Pressure - Low (RPS)
Low Steam Generator AP 1 and 2 (W M 1 aM 2)
.l Thermal Margin / Low Pressure (RPS) l 5.
Steam Generator Level Steam Generator Level - Low (RPS)
ST. LUCIE - UNIT 2 3/4 3-4
l TABLE 3.3-1 (Continued)
(
ACTION STATEMENTS ACTION 2 (Continued) 6.
Cold Leg Temperature Variable Power Level - High (RPS)
Thermal Margin / Low Pressure (RPS)
Local Power Density - High (RPS) 7.
Hot Leg Temperature Variable Power Level - High (RPS)
Thermal Margin / Low Pressure (RPS)
Local Power Density - High (RPS)
With the number of channels OPERABLE one less than required by ACTION 3 the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.
Verify compliance with the SHUTDOWN MARGIN requirements of Specifica-tion 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
With the number of channels 0PERABLE one less than required by ACTION 4 the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION may continue provided the reactor trip breakers of the inoperable channel are placed in the tripped condition within I hour, otherwise, be in at least H0T STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />;
- f.,
however, one channel may be bypassed for up to I hour, provided the trip breakers of arty inoperable channel are in the tripped condition, for surveillance testing per Specificati,on 4.3.1.1.
With the number of OPERABLE channels one less than the Minimum ACTION 5 Channels OPERABLE requirement restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the reactor trip breakers within the next hour.
l.
e l
l e
e ST. LUCIE - UNIT 2 3/4 3-5
a u.
e i
PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2. At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE
- with:
a.
Two feedwater pumps; each capable of being powered from separate T
OPERABLE emergency busses, and b.
One feedwater pump capable of being powered from an OPERABLE steam supply system.
l APPLICABILITY: MODES 1, 2, and 3.
ACTION:
a.
With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STAND 8Y 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With two auxiliary feedwater pumps inoperable be in at least HDT 4
STAND 8Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
~
A '.
c.
With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to
._ OPERABLE status.
SURVEILLANCE REQUIREMENTS 4.7.1.2, Each auxiliary feedwater pep shall be demonstrated OPERABLE:
a.
At least once per 31 days by:
1.
Verifying that each motor-driven pump develops a discharge pressure of greater than or eqQal to 1270 psig on recirculation flow.
2.
Verifying that the turbine-driven pump develops a discharge pressure of greater than or equal to 1260 psig on recirculation flow when the secondary steam supply pressure is greater than 50 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
3.
Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or othentise secured in position, is in its correct position.
"The Auxiliary Feedwater System automatic initiation system shall.be completely installed ar.d OPERABLE prior to initial criticality.
ST. LUCIE - UNIT 2 3/4 7-4
l PLANT $YSTEMS i
MAIN $ TEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION
-l
~
3.7.1.5 Each main steam line isolation valve shall be OPERABLE.
t APPLICABILITY:
MODES 1, 2, 3 and 4.
ACTION:
h-M00E 1
' With one main steam line isolation valve inoperable but open.
POWER OPERATION may continue provided the inoperable valve is
. restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; otherwise, be in at least HDT STAND 8Y 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 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
M00ES 2, 3 -
With one main steam line isolation valve inoperable, and 4 subseqent operation in M00ES 2, 3 or 4 may proceed provided:
a.
The isolation valve is maintained closed.
b.
The provisions of Specification 3.0.4 are not appl.fcable.
"i ' J f
Otherwise, be in at least HDT STAND 8Y 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 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by:
,(.
a.
Part-stroke exercising the valve at least once per 92 dys, and b.
Verifying full closure within 5.6 seconds on any closure actuation signal while in HOT STAND 8Y with T
> 515'F during each reactor shutdown except that verification U 8fu11 closure within 5.6 seconds need not be determined more often than once per 92 days.
e ST. LUCIE - UNIT 2 3/4 7-9
l 5.
THEORY OF NUCLEAR POWER PLANT OPERATIONr FLUIDS, AND PAGE 2a ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 5.01 (1.00) d REFERENCE Westin3 ouse Re#ctor Physics, Section I-5, MTC and Power Defect h
DPC, Fundamentals of Nuclear Reactor Engineering St Lucie Reactor Physics, Section 7.6 & 7.7 039/000; A2.05(3.3/3.6)
ANSWER 5.02 (1.00) b REFERENCE General Physics, Heat Transfer Thermodynamics and Fluid Flow, pp. 145 - 160.
ST Lucie Thermo Handbook, Chapter 2d
-002/000-K5.01 (3.1/3.4)
ANSWER 5.03 (1.00) b REFERENCE DPC, Fundamentals of Nuclear Reactor Engineering, p.
96 St Lucie Reacator Physics, Section 7.5 001/000-K5.56 (2.8/3.1) r of neutrons at some point in this ANSWER 5.04 (1.50) a)
Decrease
(+.5 ea) b)
Decrease c)
Ingrease REFERENCE Nuclear Power Plant Operator Trns Prgrm, HTFF and Thermor Sect 2E ST Lucie Thermo Handbook, chapter 2B
0; 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 25 ANSWERS -- ST. LUCIE 1
-86/06/10-DLANr WM (q,p A : Pumps / Centrifugal, Pump characteristic relationships (2.6/2.6)
ANSWER 5.05 (1.00) c REFERENCE Comprehensive Nuclear Training Operations (CNTO), pp 4-16/27 001/000; K5.13(3.7/4.0)
ANSWER 5.06 (2.00) a)
EOLi No-Load Tavsi Steam line break
(+.33 ea) b)
SDM is the instantaneous amount of reactivity by which the reactor is soberitical (or would be soberitical)
(+.5) assuming ell Rod Control Assemblies are fully inserted (+.25) except for the highest reactivity worth assembly which is assumed to be fully withdrawn (+.25)
REFERENCE ST Lucie TS, pp 1-6 and B3/4.1.1 001/050; PWG-5(2.9/4.3)
ANSWER 5.07 (1.50) a)
Rapid Positive reactivity additions toorapid for PZR pressure or Thermal Marsin Trips
(+.5 ea) b)
Loss of Feedwater Flow Incident (Loss of Heat Sink)
/
c)
RCP Sheared Shaft and 2 pump opposite loop flow coastdown / cr p,w,4xt d! 6 ef, u n 75 '3 :..i - o L4 3 W c e<AA..f d24 /i 4 )
a gg y.<.
e ANSWER 5.08 (1.00)
,,jjg, gg g,
)
b I
REFERENCE St Lucie Thermo handout Vol III, Ch 2, Section E pp 2-163 002/020; K5.01(3.2/3.6)
I i
l
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLU 1DSr AND PAGE 26 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, W M ANSWER 5.09 s2.svi
/
a)
Louer
(+.5 ea)
-W W-Louer,i.J.;;
1 icT Hisher
. g,fc, c REFERENCE ST Lucie OP 0030126 and Plant Curves 001/0008 A2.07(3.6/4.2)
ANSWER 5.10 (1.50) a)
Less
(+.5 ea) b)
More c)
More REFERENCE ST Lucie Reactor Physics, Section 7.5.1.2.2 001/000; K5.26(3.3/3.6)
ANSWER 5.11 (1.00) a)
Both
(+.5 ea) b)
Unit Kl REFERENCE
~
S1 Lucie TS for Unit 1 and 2 001/050; PWG-5(2.9/4.3)
m i
,30 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 27 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, W M See fqfjMk mwr ANSWER 5.12 (3.00) a)
RNF= flux in a esembly/av.. lux in all assemblies (4.6) will increase
(+.4) b)
APF= highest. flux 'n vel assembly /avs flux in the assembly (+.6) will increase due higher flux in lower half of assembly and minimal flux in.pp r segments (4.4) c)
LPF= Peak fiv>>
n one rod of an assembly /avs flux in all rods of assembly (+
)
should emain relatively constant as any flux increase should b felt equally
'n the assembly (+.4)
REFERENCE ST Lucie Reactor Physics Supplementary Handout ti, " Flux Distribution
- 001/000; K5.46(2.3/3.6)
ANSWER 5.13 (2.00) a)
Unit 2
(+.5) due to a lower Beta coefficient at EOL (+.5) b)
Unit 1
(+.5) due to MTC beins less negative, so Tavs must decrease more to add + reactivity)
(+.5)
REFERENCE CNTO ' Reactor Core Control", pp 3-21 & " Fundamentals of Nuclear Reactor Physics", pp 7-31 ST LUCIE Reactor Physics, Section 7.5.1.2 & 7.6.7 001/000; K5.49(2.9/3.4) & K5.10(3.9/4.1)
ANSWER 5.14 (1.00)
)
(+.25ea)De~~Ti'lp{av4 Reactor Power s RCS Flow cr es!
RCS Temperature Pressurizer Pressure REFERENCE ST Lucie TS B2.1.1 "53 2' f
002/000; K5.01(3.1/3.4)
M a
O O
/
a I
4-
m
'j*
y'.
- I
/~
h f
,=
vf j, a
~ e, t
/
t' * [C ll' I
\\
/
l h
(
5 J'
?
" e e
9
(
- .Q ~ '.
+ l JM l-u'.
.V f $
", /
8' k,. s ('
f',
wi i
b
}
f r
- f. r-Q
.L.
{n
,p t
f o
se is -
,r va
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 20 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, W M ANSWER 5.15 (1.50) c.
increases (4.5) due to decreasing Tave (+.25) 6.
decreases
(+.5) due to decreasin3 Tave and increasing boron (+.25)
REFERENCE SON /WBN Nuclear theory ST Lucie Reactor Physics, Section 7.5 001/000iK5.09(3.5/3.7) & K5.26(3.3/3.6)
ANSWER 5.16 (1.00)
Lower flow at the same power level results in a larger delta T(4.25); CHF Ratio is the ratio of CHF to Actual Heat Flux at a specific location in the core (+.25).
The decrease in flow resuts in less stripping action to I
remove bubbles forming at nucleation sites on the cladding and therefore a steam film could form at the lower flow rates
.(+.5)
REFERENCE SON /WBN HTFF ST Lucie Thermo Handout, Ch 2, Section E, Part 2.40 003/000; K5.01(3.3/3.9 )
ANSWER 5.17 (1.00) a REFERENCE ST Lucie Thermo Handout, Vol III, Chapter'2, pp 2-167 001/0001 K5.46(2.3/3.6)
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 29 ANSWERS -- ST. LUCIE 1
-86/06/10-DEANr W M ANSWER 5.18 (1.50) a)
ASI=(Ib-It)/(Ib+It)
(+.5) b)
At BOL, due to the colder water at the bottom of the core, ASI is positive due,to the higher flux
(+.5).
As the core ages, the flux concentration shifts upward due to fuel depletion in the lower core regions, causing ASI to head in the negative direction (+.5)
REFERENCE ST Lucie Reactor Physics Supplementary Handout 42 015/020iK5.03(3.3/3.7)
ANSWER 5.19 (1.75)
See attached curve REFERENCE Nuclear Power Plant Oper Trns Prgm, HTFF and Thermo, Figure 18r pp 202 ST Lucie Thermo Handout, Vol III, Chapter 2, Section E, figure 2-44 002/000; K5.01 (3.1/3.4)
ANSWER 5.20 (1.00) a)
Fission Chamber
(+.33 ea) b)
Uncompensated Ion Chamber c)
BF3 Proportional Counter REFERENCE STL SD 4, "Excore NIS*, pp 56/7 015/000; K6.01(2.9/3.2)
ANSWER 5.21 (1.25) a)
Decay heat is dependent upon the production and subsequent decay of fission products
(+.5) b)
See attached sketch for grading criteria
I a
Volume 111 Chrpter 2, Secti:n E
'!n
?
Natural Nucleate F
Film
- E E
I 2^
Convecdon Boiling Boiling Boiling
' y 3JCf:
si+
=
=
,j fc DN g
i
~-
( t,% [or CNr A.
10*
g p u-o m x.ets)
,.m i.
g ii
- E 10' i
10s 1
10 100 1000 10,000 Temperature Difference (*F)
Mgure 2-44 Boiling Heat Transfer In practice, if the heat flux is increased, the transition from nucleate boiling to film boiling occurs suddenly and the temperature difference increases rapidly, as shown by the dashed line in Figure 2-44. The point just before transition from nucleate boiling to film l
boiling is called the point of departure from nucleate boiling, commonly written as DNB.
l The heat flux assocated with DNB is commonly called the critical heat fluz. In many,
applications, the critical heat Oux is an important parameter. For example, in a reactor, if the critical heat flux is exceeded and DNB occurs at any location in the core, the tempera-ture difference required to transfer the heat being produced from the surface of the fuel rod to the reactor coolant increases greatly. Since the reactor coolant temperature is fixed, this means that the temperature of the surface of the fuel rod increases greatly. If, as frequently is the case, the temperature increase associated with the transition from nucleate boiling to film boiling causes the fuel rod cladding to exceed its melting point, a failure will occur. Cladding failure resulting from DNB is called burnout and will be discussed further after discusamg reactor heat transfer.
2.38 COMBINED HEAT TRANSFER Many of the heat transfer processes encountered in nuclest power plants involve a combination of both conduction and convection. For example, heat transfer in a steam generator involves convection from the bulk of the reactor coolant to the steam generator 2-151 Op
- 6 e
m p
e e
e.
e
~~
=. -
~~ -.- - -
l s
fission process were stopped, a considerable amount of heat would still be transferred to the reactor coola by the radioactive half-lives (tQ of the fission products.
'l This phenomenon is an extremely important conside the heat produced by the reactor core must be removed by Thus, circulation through the core and heat the coolant. removal. capability must be, continued even though the reactor
~
is shutdown.
It should be noted that,since this OECAY HEAT is dependent upon the decay of fission products, the total amount of decay heat available will be a function of the power opdrating history of the reactor prior to shutdown, i.e i
However, it should also greater total amount of decay heat.be noted that the amount of after shutdown, following the decay rate of the fission products. A typical curve of decay heat versus time after shutdown of a previously operating reactor, assuming 7%
'~
of the operating power is attributable to fission frapent decay is shown in the following figure.
Fioure 7.2 DECAY HEAT PRODUCTION F0E01(ING SHUTDOWN i*
1005 POWER.
(
o.u w ew.x l
1..
g E'
Q,7 r Ar sk.8 )
.e h
E3 e-c S
[./,2T[wac+k
=
of A %
j
)
I
~
^
I*
t t
t
?
o omol oat o.3 8.o lo loo loon Titec AFTEn SMuToowM (Hounsi 7.2-13 I
S
.I I
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 30 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, W M REFERENCE ST--Lucie Reactor Physics, Section 7.2 001/000; K5.37(3.6/4.1)
ANSWER 5.22 (1.00)
See attached sketch REFERENCE ST Lucie Thermo Supplementary Handout, pp 3.4-15 Appendix A, Heat Exchangers (2.4/2.7) i h
1 -
4
)
O 1
1 k
cwwa c
=v-Throur matu8
(
xx
-l'ennapet
/
rcD u fteT-7
' Tyourwr MEbtu/h I
I Tc I taur i
t Ltagry
(+,5)
Sec bene eco h eucue-for coalty medwn
(+. O kr Q pro % ca n for c e led webun l
l l
l l
l l
i l
l' 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 31 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, HM ANSWER 6.01 (1.00) 2 0 1
REFERENCE
_ ~...
'?-
ST Lucie SD117
'AFW',
pp 9 061/000; K 4. 07 ( 7.1/ 3. 3 )
ANSWER 6.02 (1.00) c REFERENCE ST Lucie SD145 '120VAC and 125VDC', pp 12/13 063/000; K4.02(2.9/3.2) i ANSWER 6.03 (1.00) d REFERENCE ST Lucie SD108 "SBCS", pp 18 041/020; K1.06(2.6/2.9)
ANSWER 6.04 (1.00) a REFERENCE ST Lucie SD62 "RPS",
pp 36 l
012/000; K6.07(2.9/3.2) e n
,----,,-e.---
r w-n,vr,-w
0 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 32 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 6.05 (1.00) 7 b
-s' REFERENCE ~
J' * " ~, ~,.;-
ST Lucie Refueling Equipment Training Manual, pp II-21
,034/000; K4.02(2.5/3.3)
(1.50) b ')
ANSWER 6.06 a)
Unit 1
(+.5 ea) b)
Both Units e)
Unit'1
,,,yg REFERENCE St Lucie SD8
'RCP",
pp 17-19 004/000; K1.03(3.3/3.6)
ANSWER 6.07 (1.50)
Unit l's LPSI System (+.5) due to the combination of the pumps discharge into a common discharge line before separatins into four in,jection lines
(+1.0)
REFERENCE St. Lucie SD24 " Safety Injection and Heat Removal Systems", pp 17 006/000; K4.18(3.3/3.8)
ANSWER 6.08 (1.50) a)
Unit 2
(+.5 ea) b)
Both c)
Unit 2 REFERENCE ST Lucie SD9 'PZR Pressurer Relief and Level Control", pp 33-36 010/000; K4.03(3.8/4.1)
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 33 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, W M ANSWER 6.09 (2.00) a)
FCV-6627X shut (4.5 ea) b)
No Auto Action c)
Valve V6565 Shuts d)
CCW Surge Tank Vent Valve (RCV-14-1) diverts from atmosphere to chem drain tank REFERENCE ST Lucie SD 60 " Radiation Monitors' 073/000; K4.01(4.0/4.3)
ANSWER 6.10
(.75)
Reactor Trip, 806 psia, 560 des F
(+.25 ea)
REFERENCE ST Lucie SD108 "SBCS', pp 17 041/020; K4.17(3.7/3.9)
ANSWER 6.11 (1.00) slectric-fails open
(+.5 ea) air-fails close REFERENCE ST Lucie SD13, pp 3 039/000; K4.05(3.7/3.7)
ANSWER 6.12 (1.00)
UNIT 1-1(or 4) fans at Normal speed
(+.5 ea)
UNIT 2-Slow REFERENCE ST Lucie SDB, pp 3 103/0001 K1.08(3.6/3.8)
U
~.
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 34 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, W M.
ANSWER 6.13 (2 25)
VCT Outlet (MV-2501) shuts
(+.25 ea)
Baron Load Control Valve (MV-2525) Shuts Blend Valve (A0V-2512) Shuts BAMT Recier valves (ADV-2510/2511) Shut Boric Acid Straiter Inlet (A0V-2161) Shuts----UNIT 1 only Boric Acid FCV (FCV-2299Y) Shuts---UNIT 2 only Eniersency Borate Valve (MV-2514) Opens Both Boric Acid Makeup Pumps Start Standby CCP(s) Start j,u.-c
- c. t 5.s., L =,
~
c" HEFERENCE ST Lucie SD13 'CVCS'r pp 47/8 004/000; K1.15(3.8/4.0)
ANSWER 6.14 (2.50) a)
2/4 Hi Cntmt Pressure (5 psis)
(+.4 for logic,
+.1 setpoint) or 2/4 Hi Cntmt Radiation (10 R/hr) or SIAS
,:3 b) 2/4 Low S/G Pressure C,405 psis) 2/4 High Cntmt Pressure W psis)
REFERENCE ST Lucie SD20 "ESFAS",pp 17, 20 013/000; K1.01(4.2/4.4)
ANSWER 6.15 (1.50) 1)
RCS Saturation Marsin
(+.5 ea) 2)
Upper Head Saturation Marsin 3)
CET Saturation Marsin REFERENCE ST Lucie GSPDS Handout, pp 14 EPE-074; EA1.13(4.3/4.6)
{
- l 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 35 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 6.16 (1.50) 1)
PDIL
(+.3 ea) 2)
Group out of sequence 3)
Group deviation 4)
ISH (res grp wthdwl perm) 5)
IRG (sd grp'wthdwl perm)
REFERENCE ST Lucie SD5 'CEDS", pp 31 001/000; K4.07(3.7/3.8)
ANSWER 6.17 (1.00)
To prevent uncontrolled pressurization
(+.7) due to the sizing of the RCGVS orifices in the vessel head and PZR vent lines (+.3)
REFERENCE ST Lucie SD7
'RCS",
pp 52 i
002/020; PWG-7(3.7/4.3)
ANSWER 6.18 (1.00)
The Recirculation Flow in Unit 2's CS Pumps is much higher (150 vs 50 spm)
Ond is able to cool the seals without outside cooling (+1.0) l REFERENCE St. Lucie SD24 " Safety Injection and CNTMT Heat Removal Systems", pp 21 026/000; K1.02(4.1/4.1)
ANSWER 6.19 (1.00) j It is sized to a higher capacity to supply Unit'1
(+.7) in the event of o tornado missle rupturin3 UNIT l's CST
(+.3)
REFERENCE ST Lucie SD117
'AFW',
pp 11 061/000; K4.01(3.9/4.2) l i
I I
,.-,,.y__-c r
--.v-,
i
- 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 36 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, W M ANSWER 6.20 (2.00) a)
MCC 2AD>> Rectifier >> Static Inverter >> Static Xfer Switch >> 120 VAC
(+.25 es component) b)
MCC 2AB>> Bypass Xfrmr(SOLA)>> Voltase Regulator >> Static Xfer Switch
>> 120 VAC
(+.25 ea) i REFERENCE ST Lucie SD144 '120 VAC Vital Distribution", pp 12-14 l
l 062/000; K4.09(2.4/2.9) & K4.10(3.1/3.5)
ANSWER 6.21 (3.00) a)
Ensure the operator does not deliberately or inadvertantly defeat parts of the protection system by switch misalignment (+1.0) b)
High SUR; Loss of Load; Local Power Density
(+.33 ea) c)
Must turn 4 keylock switches located on each RPS Panel (+.7) and is bypassed above,.1% 0.5% Unit 23
(+.3) i r 6J < $ Y
/
(( [ThL (.(h t(.i C [ (e.
b *~
( L it.ll(t(_-L.)
d $U'j Chaa[y ~s, 7 i (thC Gw i /,')
't ) lH (*(
l,4 $ ~
~%.) t l 4 i s n a / >>b.. h.;i n a 3, J g p,. v g, g,,
% w ck cA t
-5
/h tJAc Tusi k 7mn s
fY & l. l$
f trea h.AG.lfk of p ),lf,g 4
{/[f
,z 3
m rua,uub thc.L. n m,a&ju.s A.,,e f6 7 n) n n w e 1
inu n na 1
i f
7.
PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 37
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~kh65UL66565L 665TR6L ANSWERS -- S1. LUCIE 1
-86/06/10-DEAN, W M ANSWER 7.01 (1.00) b REFERENCE ST Lucie OP 2-00'30121, pp 1-5 002/020; PWG-12(3/7/3.7)
ANSWER 7.02 (1.00) b vT(u )
REFERENCE ST Lucie EP 0030141 PWG-11(EOP Immediate Actions) (4.3/4.4)
ANSWER 7.03 (2.00) i)
BOTH
(+.5 ea) b) -tHti4 -2 0 Ai c)
BOTH d)
BOTH REFERENCE ST Lucie ONOP 1/2 0250030 EPE-024; PNG-11(4.0/4.0)
ANSWER 7.04 (1.50) a)
False
(+.5 ea) b)
False c)
True REFERENCE ST Lucie EDP-1, pp 11; E0P-2, pp li EDP-8, pp 1 PNG-11(Perform Immediate Actions) (4.3/4.4) ll
0 7.
PROCEDURES - NORMAL, ACNORMAL, EMERGENCY AND PAGE 30
~~~~ d65UE655ddL E5siRUL~~~~~~~~~~~~~~~~~~~~~~~~
R ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 7.05 (1.00) b REFERENCE ST Lucie EOP-5,' Steps 7-9 EPE-074; EK3.04(3.9/4.2)
ANSWER 7.06 (2.00) a)
An independent riview of the event must be performed by the Facility Review Gr'ovp
(+1.0) b)
Nuclear Plant Supervisor and STA
(+.5 ea)
REFERENCE ST Lucie OP 0030119 EPE-007; PWG-2(3.2/3.8)
ANSWER 7.07 (1.80) 1)
CCW not loss > 10 Minutes (+.25 for parameter, + 05 for numbers) 2)
RCS between 20 amd 200 degrees subcooled 3)
PZR level > 30%
4)
S/G Wide Range Level > 40%
5)
Rx Vessel Level > 50%
6)
RCP yellow permissive light lit REFERENCE ST Lucie various EDP steps EPE-074; PNG-10(4.2/4.7) s 4
-s r-r--n-
..-w,,
n-.
a:
70 PROCEDURES - NORMAL, ADNORMAL, EMERGCNCY AND PAGE 39
~~~~RA6iBL5EiEAL C5NTRUL'~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WH ANSWER 7.08 (2.70) 1)
Reactivity Control
(+.3 ea) 2)
Vital Auxiliaries i
3)
RCS Inventor.y Control 4)
RCS Pressure Control 5)
Core Heat Removal 6)
RCS Heat Removal 7)
Containment Isolation 8)
Containment Pressure / Temperature 9)
Containment Combustible Gas REFERENCE ST Lucie EOP Appendix A (various E0Ps)
PWG-10(Recognize abnormal conditions) (4.1/4.5)
ANSWER 7.09 (2.00) 1)
HPSI Pumps
(+.25 ea) 2)
SI Motor Operated Valves 3)
Class I Emergency Power Panels 5)
Diesel Oil Transfer Pumps 6)
"A" RCP oil lift pumps 7)
Sups Power Inverter 8)
HVAC Valves and Dampers REFERENCE-ST Lucie EP 0030140 EPE456 : C'O.0103.5/3 9)
-'O O ^', # 5 ' #
ANSWER 7.10 (1.50) 1)
At least one RCP operating WITH adequate CCW
(+.5 ea response) 2)
Loop delta T is < 10 de3 F 3)
RCS is between 20-200 des F subcooled.
REFERENCE ST Lucie E0P-01, pp 7, step 5 EPE-007; PWG-11(4.4/4.5)
I 7.
PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 40
~
~~~~ 565UL65565L 66UTR5t R
ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, W M ANSWER 7.11 (2.50) 1)
Hold the CMI bypass pushbutton depressed the ENTIRE time bypass action is desired
(+.5 es step) 2)
While depregsing the rMI bypass button, depress the Bypass Enable momentarily 3)
Select "A"
on the Group Select Switch 4)
Select Manual Group on Mode Select Switch 5)
Move the joystick in the Insert direction REFERENCE ST Lucie SD5 'CEDS', pp 33 001/0108 A4.01(3.7/3.4)
ANSWER 7.12 (2.00) c)
+/- 1/2 hour of predicted ECC time
(+.5 ea response) b) -Initial criticality after refuelin3
-Uncertain data used in the ECC Calculation due to unsteady state conditions
-An excessive rescativity anomaly that could affect criticality has existed since last critical condition REFERENCE ST Lucie OP 0030126, pp 2/3 001/050; PWG-7(3.6/4.1)
ANSWER 7.13 (2.00) 1)
Any of the Rx Trip Safety Functions status check acceptance criteria not met
(+.5)
OR (+.25) 2)
All of the Safety Functions being maintained (+.5) gy 1 4 Q (p,,u ic-tkdj h I"D N g,d.d AND (+.25)
RCS conditions are being controlled / maintained in Mode 3 (+.5)
REFERENCE ST Lucie EOP-1 (SPTA)
EPE-007; PWG-10(4.1/4.2)
I l
l l
a 1
~
/
~7.
PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND" PAGE 41
~~~~~~~~~~~~~~~~~~~'~~~~
~
RA65bl5EiEAt E6UTR6L
~~~~
ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 7.14 (1.00)
Avoid lifting Main Steam Safeties after isolation
(+1.0)
REFERENCE ST Lucie E0P-4,*pp 5 EPE-G38; EK3.06(4.2/4.5)
~
ANSWER 7.15 (1.00)
It is a step that must be monitored / performed continuously while-in that procedure.
(+1.0)
REFERENCE ST Lucie Notes on pp 2 of all EOPs PWG-11(EOP usage) (4.3/4.4)
ANSWER 7.16 (1.00)
Since saturation Temperature for the RCS safet'ies is < 700 des F, this would represent a superheat condition in the RCS which could only occur with core uncovery (+1.0)
'r REFERENCE CEN-152, pp 5-77 for LOCA discussion EPE-074; EK1.02(4.6/4.8)
ANSWER 7.17 (2.00) c)
Hot Les (as you are keeping level above break)
(+.5) b)
Facter Plant Response
(+.5 ea for any 3),
No Hot spots in RCS Main Spray Available More cooling of isolated S/G REFERENCE CE EPG Guidelines on RCP Trip Strategy EPE-074; EK3.08(4.1/4.2)
/
i,
I 7.
PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 42
~~~~ d656L665C5L 66UTR6L
~~~~~~~~~~~~~~~~~~~~~~~~
R ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 7.18 (2.00) a)
Warns of possible fuel bundle bindins during insertion into core (+1 0) b)
Bridge in CORE CLEAR ZONE
(+.5 es for any two)
In transfer. zone AND hoist is at UP limit In CEA zone AND hoist is at UP Limit 1034i-r
- v ua e 1, 7 REFERENCE ST Lucie Refueling Equipment Handbook, pp I-13, I-29 034/000iK4.01(2.6/3.4) & K4.02(2.5/3.3)
A n-
..v
-,. - - ~ - - + - -
c--,
w~
v v
a 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 43 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, W M ANSWER 8.01 (1.00)
(c)
REFERENCE SL, TS-2, p 1-7,2-9,6-12,4-9.
PWG-5(Tech Spec)* (2.9/3.9)
ANSWER 8.02 (1.00)
(c)
REFERENCE SL, TS, pp 3-2,4,5.
015/020; PWG-5(2.8/3 9)
ANSWER 8.03 (1.00) t (c)
REFERENCE SL, EP-310029E, p 2.
PWG-10(Recognize Abnormal Conditions) (4.1/4.5)
ANSWER 8.04 (1.00)
(a)
REFERENCE SL, TS2, p 9-1.
~
034/000; PNG-5(2.8/3.7)
,s
\\.
l 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 44 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 8.05 (1.00) d.
REFERENCE St. Lucie T.S.
3.0.5 064/050; PWG-5(7 1/4.1)
ANSWER 8.06 (1.00) a.
REFERENCE St Lucie Tech Spec 061/000; PWG-5(3 3/4.1) & 039/000; PWG-5(3.1/3.7)
- c. L d-2 6-lp If-M "
8.07 (2.0 ANSWE3,
3,g g pg pu, A 4 - I d) - O d-l
'0 - I
, g,)
REFERENCE St. Lucie Tech Specs Sec. 6 PNG-23(Shift Staffins/ Activities) (2.8/3.5)
ANSWER 8.08 (2.00) a)
Loss of Load without a reactor trip
(+.5 ea) b)
Steam Line break er kl c)
Power ops when DNBR < 1.23 / bir' O a -)
y gy.pf cg D e n s i t y < 21 K W / f t c. ? g.t sw & rd M ogg.g. g
.,y d)
Prevent Peak Local Power w:. e n ir z - r r l G- ?
4. &!. a Mt%,/
W%-. tyx '
ANSWER 8.09
(.50)
" AO. -
\\
46.5 cc%db 7.5
~)- / ))-It y-)
REFERENCE St. Lucie Tech Spec. 4.0.2 PWG-5 (TS Knowledge) (2.9/3.9)
4 e
)
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 45 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, W M ANSWER 8.10 (1.00)
-Axial shape index REFERENCE St. Lucie Tech Spec 3.2.1/3.2.5 PWG-5(TS Knowledse)
(2.9/3.9)
ANSWER 8.11 (1.00)
- a. 3 Rem
- b. 5 Rem REFERENCE St..Lucie 10 CFR 20,,
,m a
. \\,, '
sr (u<a es.
PNG-15(Radeon Knowledge)
(3.4/3.9)
ANSWER 8.12 (1.00)
Chand,el Functional Test REFERENCE St. Lucie TS definitions PWG-5(TS Knowledge) (2.9/3.9)
ANSWER 8.13
(.50) 4 months REFERENCE 10 CFR 55 PWG-23(Shift Staffing / Activities) (2.8/3.5) 1
LT
.8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITA1 IONS PAGE 46 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM ANSWER 8.14 (1.50) e.
1.
- 2. self reading dosimeter'5'
^~
ID badge \\ ".('"'#
(IST 3.
"t keycard -
4_g (1.0) 4.
b.
First Aid and Personnel Decontamination Team Leader (0.5)
REFERENCE SL, EPIP 3100026Er p 7.
PWG-36(Eplan Actions) (2.9/4.7)
ANSWER 8.15 (1.50) 1.
The intent of the original procedure is not altered.
2.
The change is approved by two members of the plant management staff, at least one of whom holds an SRO license on the unit affected.
3.
The change is documented, reviewed by the FRG and approved by the Plant Manager within 14 days of implementation.
REFERENCE SL, TS2, p 6-14.
086/000; PWG-5(3.1/4.0)
ANSWER 8.16 (2.00) 1.
Isolation valve open 2.
A borated water volume 3.
Specified boron concentration 4.
N2 cover gas pressure REFERENCE St. Lucie Tech Specs 3/4.5.1.
006/050; PNG-5(3.2/4.3)
a s
8.
ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 47 ANSWERS -- ST. LUCIE 1
-86/06/10-DEANr WM ANSWER 8.17 (1.50) a.
- 1. Review the unit's Tech Specs. (t-U" ~
2.
R? view the unit's EOPs.
3.
Review the unit's night order, equipment out of service log, CCO log, and specific station los for the previous seven days.
4.
Complete a' minimum 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> watch.
5.
Pass an oral exam on the specific unit.
b.
Less than our months have elapsed since the person last worked on the unit. (y 3 REFERENCE SL, AP-0005720 PWG-23(Shift Staffing / Activities)
(2.8/3.5)
~s ANSRE 8.18 (1.00)
- a. slow fast f
v I ;)* 12.' n:.f+i b.
speed will matica11y incr; ease'when the fast speed one is entered.
/
REFERENCE
,/'
SL, OP2-1630024, p 2.
/
034/000j PNG-7(2.9/3.7)
ANSWER 8.19 (2.00) a.
If the person being relieved must leave their assigned station of a i
period of less than two hours.
(0.5) b.
1.
status of the control board (1.5) 2.
off normal conditions 3.
tests in progress REFERENCE l
SLr AP 0010123, p12.
l PWG-23(Shift Staffing / Activities) (2.8/3.5) l
0 t
s 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMI1ATIONS PAGE 40 ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WH ANSWER 8.20 (2.00)
Unit 1 - LCO 3.0.3 Unit 2 - LCO 3.3.1.1 action 2b REFERENCE St. Lucie Tech Epecs 010/000; PWG-5(2.9/4.1)
ANSWER 8.21 (2.00)
- e. Yes - As long as continous physical contact is maintained.
E1.O]
b.
1.
Return to service of equipment where functional testing is required to prove operability of the system or component.
[0.5]
2.
Cases of significant radiation exposure any room or area where there exist radiation levels >1000 mr/hr.
[0.53 REFERENCE St. Lucie AP 0010124 p. 2, 3.2.1.3 PWG-14(Tassing/ Clearances)
(3.6/4.0)
PWG-13(Valve Lineup Verification) (3.7/4.0) tNY' ANSWER 8.22 (1.50) 4r
- a. The responsible Foreman or an Electrical Department Supervisor. [0.5es]
b.
One month E0.53 REFERENCE St. Lucie AP 0010135 PWG-14(Tassins/ Clearances) (3.6/4.0)
ANSWER 8.23 (1.00)
All full length control element assemblies shutdown and res. are fully 3 est reactivity inserted CO.53 except for the single assembly of hi h
worth which is assumed to be fully withdrawn.CO.53 REFERENCE St. Lucie Tech Spec def. 1.29
4 t
i 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 49 I
ANSWERS -- ST. LUCIE 1
-86/06/10-DEAN, WM i
j 001/050; PWG-5(2.9/4.3) i I
O
)
W 5
k 4
k l
4 a
f g
i i
i 1
i
-.n.,-
,,,,.,..,,._c.
- - - - -,. - - - -..