ML20238F708
| ML20238F708 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/09/1987 |
| From: | Bill Dean, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20238F693 | List: |
| References | |
| 50-269-OL-87-01, 50-269-OL-87-1, NUDOCS 8709160299 | |
| Download: ML20238F708 (300) | |
Text
..
UNITED STATES
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NUCLEAR REGULATORY COMMISSION
{
I' REGION 11 h,
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101 MARIETTA STREET. N.W.
- 1.fe
's ATLANTA, GEORGI A 30323
' %,,r ENCLOSURE 1 EXAMINATION REPORT 269/0L-87-01 Facility Licensee:
Duke Power Company 422 South Church Street Charlotte, NC 28242 Facility Name:
Oconee Nuclear Station f
Facility Docket Nos.:
50-269, 50-270 and 50-287
{
)
Written and operating examinations were administered at the Oconee Nuclear Station near Seneca, South Carolina.
c [ kerdib 7/9/BF Chief Examiner li m M;panf Date Signed Approved by: [
//
- M?
W 9 <S9 Jdtfr' Y Runro, Sectidn Chief
/
Date Signed Summary Examinations on July 13-17, 1987 Written and operating examinations were administered to ten Reactor Operators (RO) and six Senior Reactor Operators (SR0); seven of ten R0s and three of six SR0s, passed the written examination. Nine of ten R0s and five of six SR0s passed the operating examination.
Based on the above results, seven of ten R0s and three of six SR0s passed.
D$[g05000269 9 8709jo PDR
REPORT DETAILS 1.
Facility Employees Contacted:
l
- P. Stovall, Plant Operations Instructor
- D.
Tidwell, Instructor
- R. Bugert, Training Supervisor
- L. Hindman, Instructor
- J. Price, Shif t Operating Engineer
- T. Campbell, Operations
- T. Barr, Manager Oconee Training i
- Attended Exit Meeting 2.
Examiners:
- B. Dean, NRC C. Casto, NRC J. Huenefeld, PNL B. Gore, PNL Sandy Lawyer, NRC Len Wert (NRC RII - attended exit)
- Chief Examiner l
l 3.
Examination Review Meeting At the conclusion of the written examination, the examiners provided Mr. Stovall with a copy of the written examination and answer key for review.
The comments made by the facility reviewers are included as to
'% repoM, and the NRC resolutions to these comments are listed below.
l 10 of 30 changes (33%) made to the answer key were a result of inadequate or insufficient material provided to the NRC for examination development.
a.
R0 Exam (Applicable SR0 questions are in parenthesis)
Question 1.02(5.02):
Agree with facility comment.
Answer "c" will also be accepted.
Question 1.06:
Agree with facility comment.
Answer "b" will also be accepted.
Question 1.11(5.07):
Agree with facility comment.
Answer "d" will also be accepted.
i
2 Question 1.19(b):
Do not agree with facility comment.
No information was provided to indicate that the magnitude of change in rod worth over core life is as significant as the change in moderator temperature coefficient, which is nearly 0 delta k/k/ degree F at 80C.
The referenced lesson plan assumes that the maximum dropped rod worth is 0.3% delta k/k.
For a given rod, its worth will change relatively little over core life.
Question 1.20(5.19):
Agree with facility comment.
Additional recommended answers will also be accepted.
Question 1.21(5.20):
Facility comment acknowledged.
The steem-steel reaction will be accepted as an alterlate answer for the Zinc-boric acid reaction.
Due to the lack of direction in the geestion regarding the status of the ECCS system, it is possible for the Zirc-water reaction to be more significant.
Therefore, the order of the last two sources of H2 will not be required.
Question 1.22(5.22):
Agree with facility comment.
Xenon oscillations will not be required as part of the answer.
Question 1.25(5.25):
Agree with facility comment.
Typographical error in the answer key will be corrected.
Question 2.07b:
Agree with facility comment.
Note that the lesson plan for High Pressure Injection cnly lists Unit 3 as having these computer points per tracking number 86-188, and should be l
updated.
Question 2.08a:
Facility comment acknowledged. The question is clearly asking for another source of water for the SSF-ASW System. Any alternate answers will be evaluated based on the assumption that all Intake Water is unavailable.
Question 2.08b:
Facility comment acknowledged.
Though it is not listed as the reason for the installation of the eductor in the l
referenced lesson plan, the device does indeed minimize the air injection into the I
S/Gs when the SCF-ASW System is in service.
1 3
4 Alternate answers will be evaluated for applicability.
Question 2.10:
Agree with facility comment.
Recommended answer will be added to the answer key as an additional correct response.
Question 2.13:
Do not agree with facility comment.
The Lesson Plan for the Reactor Protective System has an additional learning objective (1j) for having the knowledge of the loss of any Power Supply to the RPS.
As it is possible for this module to generate a trip signal for its respective channel, the knowledge examined is considered a valid testing area.
No change to answer key.
Question 2.14:
Agree with facility comment.
Lesson plan should be changed to reflect current plant conditions.
Question 2.15:
Agree with facility comment.
The recommended additional a.. ser will be incorporated and six of seven responses will be required for full credit.
The lesson plan should be updated to reflect the correct order of preference.
Question 2.16:
Agree with f acility comment.
The lesson plan for the Component Cooling System should be updated to reflect the plant differences in Reactor Coolant Pumps.
Question 2.17:
Agree with facility comment.
Full credit will be given if candidate lists four correct loads which lose cooling water.
Question 2.20b(6.22b):
Facility comment acknowledged.
The question was not specific enough to elicit the system manipulations required.
The facility recommended answer is trite.
The question will be deleted.
Question 2.21(6.23):
Facility comment acknowledged.
The recommended additional answers will be accepted as additional correct answers.
The referenced lesson plan should be updated to reflect this information.
I
~
4 Question 2.22c:
Agree with facility comment.
Answer key changed as recommended.
Question 3.04(6.06):
Facility comment not accepted. Though " Type A or B" may not be emphasized in the training, the questions are still applicable to all the RZ modules.
The facility should emphasize in their lesson plans such l
categorizations which are for instructional convention only.
No change to the answer key.
Note that this terminology was not questioned by any of the candidates during the exam.
Question 3.05 (6.07):
Do not agree with facility comment.
Even though a facility learning objective does not exist, the material is covered in the referenced lesson plan, and NUREG 1122, "Knowledges and Abilities Catalogue for Pressurized Water Reactors" supports the validity of this question.
No change to answer key.
Question 3.08:
Agree with facility comment.
Additional recommended answer will also be accepted.
l Question 3.09a(6.09a):
Agree with facility comment.
Answer key will be changed as recommended.
Facility should update their lesson plan.
Question 3.13(6.13):
Agree with facility comment.
The control room will be added to the answer key as a l
correct location for operator control.
The I
referenced lesson plan should be updated to match the AP.
Question 3.17a:
Agree with facility comment.
Valving in the transmitter will also be accepted as correct.
l Question 3.17b:
Agree with facility comment. The answer key l
will be changed as recommended.
I Question 3.19:
Agree with facility comment.
Answer key will be modified as recommended.
l
5 Question 3.20:
Facility comment acknowledged.
The only terminology that is not expressly utilized in the facility training material is
" Coincidence Logic contact".
This part of the required answer will be deleted.
Point values will be redistributed to only require lasting of the correct order of components.
Question 3.22:
Agree with facility comment.
Additional recommended answers will also be accepted.
Question 3.26(6.25b):
Agree with facility comment.
Additional recommended answers will also be accepted.
Question 4.01(7.01):
Agree with facility comment that the wording of the question may confuse candidates and result in no correct answer being apparent.
Question will be deleted.
Question 4.02(7.02):
Do not agree with facility comment.
The question was specific enough with regards to existing plant conditions to elicit the desired response.
Venting the primary is the most effective method of " removing" non-condensibles as use of RCPs only entrains these gases in solution.
The EPG is an excellent reference source for background information associated with emergency procedures, and may serve as a source of information if facility lesson l
I plans are inadequate.
No change to answer key.
l Question 4.07b(8.13b):
Agree with facility comment. Due to the l
conflicting information in the Health Physics Manual and the Emergency Plan, l
either answer will be accepted.
- However, the SR0s should be aware of their l
responsibilities and limitations under the l
Emergency Plan and the Health Physics Manual should be updated to agree with current procedural guidance.
Question 4.12(7.12):
Agree with facility comment.
Recommended additional answers will also be accepted.
Question 4.17(7.18):
Agree with facility comment.
Additional recommended answers will also be accepted.
lL_
6 Question 4.18:
Agree with facility comment. The referenced AP is incomplete with regards to the automatic actions that occur for the given RIA alarms and should be updated.
Question 4.19(7.19):
Agree with facility comment.
Note that the corrected version of the lesson plan had not been provided to the NRC examiners.
Question 4.22(7.22):
Facility comment acknowledged.
The term
" saturated repressurization" as given in the context of the question may not be familiar to the candidates, as this term is not in the emergency procedure. However, the situation is clearly represented by the conditions i
l provided.
The answer key will be modified to weigh the response required by the procedure more than the theory behind the phenomena.
Also, lowering 0TSG pressure will be accepted as a correct action, b.
SR0 Examination Question 7.16:
Agree with facility comment.
The recommended additional answer will be accepted.
Six correct responses will be required.
Question 8.01:
Agree with facility comment.
Typographical error in the answer key will be corrected.
Question 8.04:
Agree with facility comment.
Question was not specific enough to elicit the desired response.
The recommended additional answer will also be accepted.
l
)
Question 8.06:
Do not agree with facility comment.
The TS l
is clear on the applicability of Flow Versus l
Level instrumentation and the word " Flow" l
was also emphasized in the body of the I
question. No change to answer key.
Question 8.14:
Facility comment acknowledged.
The recommended answer will be accepted only if the candidate expressly states the required assumption.
Question 8.15:
Agree with facility comment.
Additional recommended answers will also be accepted.
l
7 Question 8.23:
Facility recommended answer is equivalent to the answer key.
No change to key required.
4.
Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the results of the examination.
There were some generic weaknesses noted during the operating examinations, particularly inadequate use of annunciator response procedures and poor knowledge among SR0s of the Standby Shutdown Facility.
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.
_. an ' a 9. m U.
S.
NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
_QCQNEE_1 3 _ 2 E } _ _ _ _ _ _ _ _ _ _,.
REACTOR TYPE:
_PWR;@gW,1,ZZ______________
DATE ADMINISTERED: _8 ? f 9 ? f l 3,, _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
EXAMINER:
_geSTg1_g._______________
CANDIDATE:
ISSIRuCIIgNS_IO_geNg1D9IE1 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 grade 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 st ar' t s.
% OF CATEGORY
% OF CANDIDATE'S CATEGORY
__VeLUE_ _IgI9L
___SggSE___
_V@LUE__ ______________g91EggSY_____________
_E9199__ _??t99 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
_EE!Y7I_ _S$199 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
- E k A
_ITI?2 _ _25199
________ 7 PROCEDURES - NORMAL. ADNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
__30199__ _25199
________ S.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS g-1e9f29__
Totals Final Grade All work done on this examination is my own.
I have neither given nor received aid.
Candidate's Signature
"7*
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS j
During the administration of this examination the following rules apply; i
1.
Cheating on the examination means an automatic denial of your appitcatico j
and could result in more severe penalties.
2.
Restroom trips are to be limited and only one candidate at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark penci! gnly to facilitate legible reproductions.
j 4
Print your name in the blank provided on the cover sheet of the f
examination.
S.
Fill in the date on the cover sheet of the examination (if necessary).
)
6.
Use only the paper provided for answers.
7.
Print your name in the upper right-hand corner of the first page of gach section of the answer sheet.
]
l i
8.
Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gne side of the paper, and write "Last Page" on the last answer sheet.
9.
Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least thcee lines between each answer.
- 11. Separate answer sheets fecm pad and place finished answer sheets face down on your desk or table.
j
- 12. Use abbreviations only if they are commonly used in facility literatute.
- 13. The point value for each question is indicated in parentheses after the
]
question and can be used as a guide for the depth of answer reouired.
- 14. Show all calculations, methods, or assumptions used to obtain an answer j
to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given.
Therefore. ANSWER ALL PARTS OF THE OUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16.
If parts of the examination are not clear as to intent, ask questions of the examiner only.
17 You must sign the statement on the cover sheet that indicates that the worL is vour own and you have not r ec ei ved or been given assistance ir completing the examination.
This must be done after the examination nas been completed.
l J
T v*
y,a
- 18. When you complete your examination, you shall:
a.
Assemble your examination as f oll ows:
(1)
Exam questions on top.
(2)
Exam aids - figures, tables, etc.
(3)
Answer pages including figures which are part of the answer.
b.
Turn in your copy of the examination and all pages used to answer j
the ex ami nati on ouestions.
c.
Turn in all scrap paper and the balance of the paper that you did I
not use for answering the questions.
l I
d.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
1 I
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I 1
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T H_ E_ R M O _D Y N A M_ _I C_ S_
g, OUESTION 5.01 (1.00)
Which one of the following correctly describes the behavior of PCS pressure if a Small Break LOCA which was not large enough to actuate the ECCS were to occur, without Feedwater avail abl e' j
a.
Pressure initially decreases slowly, then rapidly drops when the OTSGs are boiled dry.
b.
Pressure decreases slowly until it levels off somewhere above ECCS actuation pressure.
c.
Pressure initially increases, then rapidly drops when the OTSGs are boiled dry.
d.
Pressure initially decreases, then rapidly increases when the OTSGs boil dry.
e.
Pressure i ni t i al l y decreases, then when OTSGs boil dry, continues to decrease, but at a much slower rate.
QUESTION 5.02 (l.00)
Which one of the following instrument f ailures would Cause the behavior of the rarameters shown on attached drawing OC-TA-NT-15?
a.
Delta Tc Failure "A" Side LOW
{
b.
Delta Tc Failure "A"
Side HIGH c.
Delta Tc Failure "B"
Side LOW d.
Delta Tc Failure "B"
Site HIGH
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CATEGORY 05 CONTINUED ON NEXT PAGE
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2
l 51-_IdE981 95 U9EhE93 89dE5 Ch981 9558911991 Eh91951_Q$D PAGE 3
4 THERMODYNAMICS l
o,
)
i l
l 1
OUESTION 5.03 (1.00) l Which one of the following correctly defines the term "E-Dar" which is used in radionuclides sampling?
(assume half-lives greater than 30 minutes) a.
The gross gamma analysis of the demineraliger effulent compared to the gross gamma analysis of the RCS.
b.
The average beta energy weighted to determine the relative fission product levels.
c.
The average alpha anal ysi s used to estimate the amount of U-238 in the RCS.
d.
The average beta and gamma energy weighted in proportion to the contribution to the total ac ti vi ty.
QUESTION 5.04 (1.00)
Which one of the following ts correct concerning differential control rod worth (DRW)?
a.
It is a measure of reactivity due to rod posi ti on.
c.
With a normal cosine flux shape, DRW reaches a maximum value at a rod index of less than 29%.
c.
Rod Group Overlapping maintains a constant DRW.
d.
Its unit is delta K / K /*/. index.
i i
l l
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CATEGORY 05 CONTINUED ON NEXT PAGE
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A
5:__IHEg6Y_gE_ NUCLE 95_EgWE6_EL9NI_g[Eg@IlgN1_ELUlgS _9NQ PAGE 4
1 T H E R M O D.Y. N A M I C S DUESTION 5.05 (1.00)
Which one of the following represents the maximum linear power density which would be expected in the core during full power operations?
a.
Local Power Density multiplied by Nuclear Peak 2ng Factor.
b.
Radial Peaking Factor multiplied by the Local Peaking
- Factor, c.
Average Kw/ft for the core multiplied by the Nuclear Peaking Factor.
d.
Nuclear Peaking Factor multiplied by the Maximum Local Power Density.
QUESTION 5.06 (1.00)
Refer to Figure 16 attached, If elevation "A"
is 16 feet, PG2 - PG1 is 5 psia and the fetction head losses equal O.5 feet what is the Total Developed Head for the pump 7 a.
18.67 feet b.
21.5 feet c.
27.0 feet d.
28.0 feet OUESTION 5.07 (1.00)
Which one of the following conditions would hinder natural circulation flow in the RCS?
a.
An increase in thermal driving head.
b.
An increase in the velocity head of t'he fluid.
c.
An increase in the mass of feedwater into the OTSG.
d.
A decrease in Two-phase (nucleate boiling) within the core.
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l
}
5:_ _ IHgggy _ g[_NUC 6 g@@_ P gW Eg_P(gNI _ gP g]@ll gNg_[(Ulp5 _gNQ PAGE 5
3 IbEBMODyN@MICS
+'
OUESTION 5.08 (1.00)
State how the behavior of the following parameters during a pressurizer steam space leak differs from that observed during an e qui val en t sized (SBLOCA) break in the primary i
system.
Assume that the Unit is initially at power and a reactor trip does occur, a)
Initial pressure change b)
Pressurizer level for first several minutes i
OUESTION 5.09 (1.50) a)
Assuming that the plant is operating at full power, what will be the difference in the following parameters, if OTSG tube fouling has cccurred to a significant degree 7 1)
OTSG Level l
l 2)
Superheat Temperature 1
l b)
Which area of the OTSG is considered to be the most critical with regards to reduced flow when fouling occurs?
OUESTION 5.10 (1.50)
With the Unit operating at 100'/. Power with all control systems in automatic, a Turbine Bypass Valve fails full l
open.
Indicate how the f ollowing parameters will change l
relative to their initial values when plant conditions
)
stabilize: (INCREASE. DECREASE, REMAIN THE SAME) l a)
Tavg b)
MWe c)
Reactor power 1
1
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J
5:_ IHgggy_QE_9UCLE@@_EQEEB_EL@NI_gfg@gIl@N1_ELUlpS _gNQ PAGE 6
t THERMODYNAMICS QUESTION 5.11 (1.00)
An ECP is c al cul ated for a startup following a reactor trip from 50% power, with equilibrium xenon in the core at MOL.
Indicate 1f the ACTUAL critical rod position will be HIGHER, LOWER or the SAME compared to the calcul ated position for
]
each of the following situations.
Use attached curves as appropriate and treat each situation i nd.i vi dual l y.
a)
Xenon r eac t i vi ty curve for trip from 100% power is used to calculate conditichs for a startup 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after the trip.
(Computer is not in service to give this info) b)
The differential boron worth at an EOL condition is used.
Assume no change in boron concentration is j
desired prior to achi evi ng criticality.
QUESTION 5.12 (1.50)
Indicate whether the following will INCREASE, DECREASE or REMAIN THE SAME:
a)
Available NPSH for a MFP as volumetric flow rate increases, b)
Minimum required RCP NPSH as volumetric flow rate increases.
c)
Available NPSH to c on'd en sat e ( h ot We l l ) pumps as condenser subcooling increases.
QUESTION 5.13 (1.00)
For EACH of the following conditions state whether the actual Shutdown Margin would be greater than/ less than/ the same as calculated Shutdown Margin.
Assume BOC, Mode 1 full power.
1.
Actual poison burnup exceeds that value used in th<
calculated SDM.
2.
The actual worth of the maximum stuck rod assumed in the calculation is lower than predicted.
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5.
THEORv OF NUCLEAR POWER PLANT OPERATION. ' FLUIDS, AND PAGE 7
~~~~T__5ER_566_"N_35_5_C_5_~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
u, 2,
DUESTION 5.14 (2.00)
For each of the following par ameters describe the indication which is used to ensure adequate coupling following a loss of RCP's with one steam generator available, a.
Thot on both loops b.
Tcold on the operating generator c.
Cooldown rate on both loops d.
Thot - Tcold on the operating steam generator l
)
DUESTION 5.15 (1.00) l l
With regards to the Flux / Flow / Imbalance Trip function, state whether a Posi ti ve or Negative Flux Imbalance would allow a higher trip setpoint (assuming other pertinent parameters are the same).
Explain your answer, i
QUESTION 5.16 (1.00)
What ;ndication tells the operator when all nitrogen has Jeen vented from the Pressurizer, When forming a steam bubble in accordance with OP/0/A/1103/05?
DUEo f iCN 5.17 (2.00)
Refer to the attached drawing OC-TA-NT-10, "FDW Main Control Valve Fails Open at 50% Power", to answer the following:
a)
Why does FWPT Speed increase, then level off at point (2)'
b)
What is causing reactor power to increase starting at point (3)?
c)
After reactor power and MWe stabill:e, prior to the reactor trip, what is the relationship between Loops A and B "T Cold"7 l
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5.__IHEgSY_gE_ NUCLE 98_EgWE5_ELgNI_ggES911gy1_ELUlgS _999 PAGE 8
1 THERMODYNA.MICS QUESTION 5.18 (1.50)
State three mechanisms that can produce high boron concentrations in the core post LOCA.
QUESTION 5.19 (1.00)
If a OTSG overfills, describe two adverse effects on the secondarv sida: other than water hammer-related effects.
QUESTION 5.20 (1.00)
What are the three most significant sources of Hydrogen production in the primary f oll owi ng an accident (e.g.,
LOCA)?
List them in order of significance.
i OUESTION 5.21 (1.00)
Since fuel burnup is a direct function of operation, why is the attached Core Excess Reacti vi ty Curve a non-linear function' OUESTION 5.22 (1.00)
Recently, inconel Axial Power Shaping Rods (APSRs) have been installed on all 3 units.
What is the main physical change in these rods and how has this change improved the APSR's capability to control axial flux imbalances?
CUESTION 5.23 (1.00)
State the operational concerns of an uncontrolled cooldown on the reactor vessel.
Include in your answer the types of stresses induced on the inner and outer sessel walls.
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5 __1b5981_g[_yU96[@$_[QEEB_[68@J,9((6@l19G1_[6W 2@z_@bD PAGE 9
IMB000ne01c5 QUESTION 5.24 (1,00)
Explain how the use of RCP " bumps" minimizes the effect of gas AND steam accumulation within the RCS.
QUESTION 5.25 (1.00)
The pressurizer PORV ts leaking by during opere. tion at 85%
power.
Assuming a Quench Tank pressure of 20 psia and saturation conditions in the pressurizer corresponding to 2240 psia. what is the quality of steam on the downstream side of the PORV' Show al1 calculations.
QUESTION 5.26 (1.00)
The reactor is producing 100% rated thermal pcHer at a core delta T of 60 degrees and a mass flow rate of 100% when a station blackout occurs.
Natural circulation is established and core delta i goes to 40 degrees.
If decay heat is 2%,
what is the core mass flow rate (in %)?
a.
1.3 l
I c.
2.0 c.
3.0 d.
4.2 I
1 l
t i
i
(*****
END GP CATEGORY 05
- )
l
6:__P699I_gy@IEDg_9Eg1Gy1_CgyISOL1_999_ly@l6UDENI@Ilgy PAGE lo OUESTION 6.01 (1.00)
Which one of the f ollowing is NOT a design diffeie.c e between the safety / regulating control rods and the APSRs?
a.
On APSR drives a small button on the lower portton of the s
segment arm prevents the lead s:rew from being c t s e n g ag.P,d when power is lost.
t b.
APSR couplings have larger diameters a r. d shorter keys to prevent coupling an APSR drive to a safety or regulating rod or vise-versa.
c.
APSR drives have ball valves and bypass parts.
d.
APSR do not have buffer springs in the ou'if er. assembl y.
QUESTION 6.02 (1.00)
Which one of the (ollowing correctly describes the accident that was assumed to exist, necessitating tnd install ati on of HP-409 and HP-4107 a.
Failure of both HP1 headers at any power level.
]
i b.
Failure of cne HPI header with the opposite header HPI pump out of service above 50% power.
c.
Failure of two HPI pumps with a break of t he RCS p ii.:i r g downstream ot a Reactor Coolant Pump (RCP) at 100%
power.
r d.
An RCS piping break downstream o+
a RCP with failure of the opposite side HPI header at greater:than 60*/. p ower,
e.
An RCS piping break anywhere in a cold leg, with a failure of the opposite side HPI header ak any power level.
v f
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_6_.__P_L_A_N_T__S_Y_S_TE_M_S__DE_S_IG_N_,_ C _O N_ _T R O L_ _,_ A_ N_ _D _ _I N_ S _T R U ME N_ T AT_ _I O_ N_PAGE 11 i
v.*
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i OUESTION 6.03 (1.00)
Which one of the following should enable MS_93, TDEFWP steam supplv valve, to open if it fails to open on an automatic signal' I
a.
Verify proper operation of the DC Oil Pump.
l b.
Line up backup servi ce air to the valve operator.
Isolate instrument air to its reducer and bleed the air off the reducer.
i d.
Take the control switch to "Off" to remove power from the solenoid valve.
QUESTION 6.04 (1.00)
Which one of the following conditions correctly describes the requirements for a 23OKV Switchyard Isolation to occur?
a.
Undervoltage on 2 of 3 phases on EITHER the Red or Yellow Bus on BOTH Channel s of UV protection.
b.
Undervoltage on 4 of the 6 phases monitored between the Red and Yellow Buses on EITHER Channel of UV protection, c.
Uncervoltage on 4 of the 6 phases monitored Detwee9 the Red and Yellow Buses on BOTH Channel s of UV F'r ot ec t i on,
d.
Undac.c;tage on 2 of 3 phases on BOTH the Red and Yellow Buses'on BOTH Channels.
e.
Undervoltage on 2 of 3 phases on BOTH the Red and Yellow Buses on EITHER Channel.
1
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$1_..Ch691_E1EIEUE_9EEIEUt_E99189b1_@yp_ly919U[ENI9Ilgy PAGE 12 s.
=,,
QUESTION 6.05 (1.00)
Which one of the following correctly describes an interlock associated with Component Cooling Di scharge Val ve CC-87 a.
If CC-8 is closed, NEITHER CC Pump may be started.
b.
CC-8 closes on actuation of ES-7 or ES-8.
c.
CC-7 (MOV CC Discharge Valve) closes if CC-8 closes.
d.
If CC-8 cl oses any operating CC pumps will continue to run.
QUESTION 6.06 (1.00)
Which one of the following correctly describes the response of the Type "B"
RZ Module to a contal power loss' a.
If vital control power to the Auto / Manual ES logic is interrupted - position indication and manual control power will not be available to the manual control P/Bs.
b.
If the vital power source is lost with no emergency signal present, no effect will be seen on the Auto / Lamp /Pushbutton.
c.
If the vital power source is lost, digital control logic remains operable.
d.
If vital power is estored while the emergency signal is present operation to the emergency signal present logic must be manually restored.
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$1__ChOdl_$1515U5.9551991_E99109hi_OUS lygl@U((gl@I}gN PAGE 13 QUESTION 6.07 (1.00)
Which one of the following correctly describes the local switchgear Controls for the SSF7 a.
Breaker Interlocks prevent closure of a breaker by toe i
manual close pushbutton (l oc at ed on lower right hand l
i l
portion of the breaker) when the breake-is in the open
- position, j
b.
Lccal closing of the breakers using the control switches (different from the manual PBs) can be a c c o mp l l,sh ed as long as the breaker is not racked completely out.
c.
Local tripping of the breakers using the control switches
)
l (different from the manual PCs) is permitted only in the Test position.
I d.
With a loss of control power the ability to trip a breaker with the manual open pushbutton is lost.
QUESTION 6.08 (1.00)
A valid EFW start signal ex i sts.
Unless otherwise specified all appropriate controls are in automatic.
Which one of the following conditions would prohibit the injection of EFW into the OTSG by the Turbine driven pump?
a.
The low oil pressure switch PS-3Ol has fatled low.
l l
b.
EFWPT control in Pull-To-Lock and a failure of kV1D Brk 6 (solenoid power supply) which trips open, c.
While selected to Primary level control, a loss of power to the Pri mar y channel occurs, c.
The valve position limit switenes for MS-93 Steam Supply Valve fail to recognize the valve opening.
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$1__P6991_@y@I@[@_pE@lGNt_CONISO6t_9NQ_lN@lB6)DENI@IlgN PAGE 14 OUESTION 6.09 (1.50)
Indicate the INITIAL RESPONSE of the following parameters and components following the failure of "A"
S/G Outlet Pressure HIGH: (Assume plant initially at 80*/. and no trip occurs) a)
Feedwater flow to "B"
S/G b)
"A" Bypass Valves c)
Generated MWe QUESTION 6.10 (1.50)
Indicate for the following whether they apply to an RPS Bistable or an RPS Contact Buffer?
a)
Two analog inputs, with one input being a setpoint.
b)
Must be manually reset via a toggle switch, c)
Energizes a relay if input senses a trip condition.
1 QUESTION 6.11 (1.50)
Indicate where the following components could be operated from(if at all), given the loss of power indicated:
{
1 I
a)
Pressurizer Spray Val ve on loss of Auto Power.
j D)
Pressurizer Spray Block Valve on loss of Auto Power.
j c)
Pressurizer Electric Relief (RC-66) on loss of Auto Power.
OUESTION 6.12 (1.00)
What is the basis for the following administrative controls that were recently instituted on the Main Steam to Auxiliary Steam interconnections?
Only one unit's Main Steam is used to supply Auxiliary Steam via only the 2" reducer MS-129.
The other two units' Main to Aux Steam reducers, MS-126(6") and MS-129 are totally isolated.
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 15 i,'
o,e OUESTION 6.13 (1.00)
Describe the two (2) intersystem ti es between the HPI system and the SSF Makeup system.
Include in your answer tne source and discharge flow path.
QUESTION 6.14 (1.00)
Explain the basis for each of the f ollowing system l i mi tat i ons:
1.
Maintaining Letdown Storage Tank level above approx. 18 inches.
2.
Maintaining Letdown Storage Tank pressure vs.
level curve within the operating range.
QUESTION 6.15 (1.00)
I Should a Control rod
'in" movement signal be generated and the rods run in the out direction, what Diamond Control Panel Lamp would illuminate AND what auto action would result from this condition?
QUESTION 6.16 (1.00)
N I
For the Differential Amplifier a Power Range Instrument channel)5 state two (2) inputs AND three (3) outputs.
QUESTION 6.17 (1.00)
List the two interlocks that must be satisfied in order to start the SSF ASW pump if an ES-1 or 2 (l oad shed ) signal is present.
j QUESTION 6.18 (1.00) l What three components does the " Loss of Instrument Air" procedure tell the operator to monitor to ensure the N2 Backup Supply i s operati ng correc tl y' a
l
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$1,_fLgNI_SYSIEMg_gESigN1_CgNIBgL _AND INSTRUMENTATION PAGE 16 g
1 o.,
QUESTION 6.19 (1.00)
List 4 of the advantares gained by the installation of the s
Multiple Function Mast on the Unit 3 Refueling Bridge.
QUESTION 6.20 (1.50)
For the SPDS list three (3) of four (4) instruments and/or indications which are used to generate the Subtriticality critical safety function logic.
QUESTION 6.21 (1.00) i List two (2) automatic actions which occur affecting the SSF in the event a Unit 2 Channel A load shed signal is received.
I OUESTION 6.22 (1.50) a)
Explain why AUTOMATIC control of steam header pressure f
on the Turbine Hand / Auto station, with the ICS in TRACK, is not a preferred mode of operation at low power levels 7
_D_1_
Duc4 %.a _s t a r_t.up 3._whp. _i,.s. d on e_ t,o en sur e th f)t[
~} A[
t not be in TRACK when the Turbine C5n tr ol"i" ace n
automatic' OUESTION 6.23 (1.50)
If a loss of Instrument Air to the air-operated valves in the Makeup portion of the HPI System occurred, what 3 methods / alternate flow paths could be utilized to maintain pressurizer level, assuming that "A"
HPI pump is in service at the time of the failure' OUESTION 6.24 (2.00)
Describe BOTH the purpose AND operation of the Reactor Building Cooling Unit dropout plates AND blowout plates.
y i
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1 63__P(@yl_@l@l@DS_Q@@l@G3_QQyl@g(3_ gyp _ly@l@y[@ylgflgN PASE 17 QUESTION 6.25 (2.00)
The unit is operating at full power conditions, all controls and instrumentation are in automatic, Channel 9 Power Range N1 is supplying the ICS.
a.
A loss of 120 VAC to Channel E of the RPS occurs (KI remains energized).
For EACH of the resultant plant conditions listed below (1-4) provide the appropriate system response.
RESPONSE OF:
1.
Neutron flux fai1s to zero Control Rods 2.
Loss of RCS flow signals ICS and Feedwater 3.
Items 1 & 2 above RPS 4
RCS narrow-range pressure fails low PORV control Spray control b.
RCS pressure starts to increase, what condition would cause this continual increase in 7CS pressure?
1
(
)
)
I
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Z:__EEOCEDUSES_ _NOBd@61_@BNOBd@61_EDEBGENQy_@yD PAGE 18 R_A_D_IOL_O_G__IC_A_L__CO_N__TR_OL QUESTION 7.0 (1.00)
Which one of hpe following is NOT a Control Rod Drive System limitation an d /or precaution according to OP/0/1/1105/097
\\
Iimits during.testt~ "g of the CRD breakers a.
Ther e are no r /q/[)-
for rate, howeynr the
. rv at which they are cycled is limited.
'\\
O'
\\
s D.
The CRDs must be hem ed if pressurizer level decreases below a predetermined limit.
\\
c.
While on the Aux Power Supply with a Safety Group while all other Safety Groups are at the out limit the operator i
is prohibited from se,lecting Automatic.
\\
- d. Operating limits h a v e 'b,e e n established to assure that control rod drop is prohibited under conditions which would defeat the hydraulicssnubbing action of the mechanism.
\\
OUESTION 7.02 (1.00)
Which one of the following methods contained within the Emergency Procedures is the best method for removal of RCS volds tnat are due to the presence of Non-condensible gases?
a.
Pepressurization of the RCS, b.
RCP Restart.
c.
RCP Bumping, d.
Vessel or Hot Leg venting.
I
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Z:__B599E99BE!_:_U950961_999950961_EDE59ES9Y_9ND PAGE 19
... Se91969GlC@6_CgNI@g6 OUESTION 7.03 (1.00)
Which one of the following correctly describes the required actions if the reactor achieves cetticality 1.5% (deltak/k) below the estimated critical position?
a.
Fully insert all regulating rods, but the safety rods may stay ' f ull y wi thdrawn.
b.
Fully insert all regulating rods, fully insert all safety rods to Group 1 at 50% withdrawn, c.
Fully insert all regulating and safety rods, d.
No rod insertion is required, but the ECP shall be j
recalculated.
e.
No action required and startup may continue.
QUESTION 7.04 (1.00)
What is the definition of APSR Movement as it occurs over a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time period?
I i
QUESTION 7.05 (1.00) l Upon a loss of 1KI AP/1/A/1700/23 directs an operator to the Aux Shutdown Panel to perform various actions.
Which one of the following is an action the operator can perform from j
this panel' j
a.
Re-energize TBVs b.
Bypass 1KI Inverter c.
Control FCS pressure with Pressurizer Heater Banks 1,
3 and 4 d.
1i conditions warrant trip 1 RCP in
'B' loop
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21 _889EE9ljdEE_~_U98Ubbt_9EU98U961 EUE8EEUE1.999 PAGE 20
,,, 5691969 GIG 06_G90IE96 QUESTION 7.06 (1.00)
The attached drawing, Figure 7.28, shows the LPI system aligned for which one of the following modes of operation?
a.
Switchover mode on Unit i b.
Switchover mode on Unit 2 c.
Normal decay heat removal on Unit 2 d.
Normal decay heat removal on Unit 3 OUESTION 7.07 (1.00)
Fill in the blanks with the appropriate limitations and precautions for operation:
a)
Any E.S.
Valve that has been manually operated must be to assure operability.
b)
Do not operate any EFW Pump at >
gpm and maintain UST level > ______ ft whenever these pumps are required to be operable.
DUESTION 7.08 (1.00) l l
List the three breakers in the correct order in which they must be operated in order to allow the Keowee Hydro Generator to produce voltage output, once the wicket gate posi t i on is established, i
OUESTION 7.07 (1.00) a)
Which two individuals, by title, may authorire the bypassing of Main Fuel Bridge interlocLs?
b)
Where are the bypass switches for these interlocks located?
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7 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 21 89DIOLOGig6L_QONISOL 1
QUESTION 7.10 (1.00) l While performing CP-604, " Solid Plant Cooldown", following an accident condition, it has you transfer to CP-605, "Sub c ool ed Cooldown", if Pressuri er level gets less than 300 inches.
What is the basis behind this setpoint Deing the criteria +or this procedural transition?
I I
QUESTION 7.11 (1.00)
Referring to the attached excerpt from OP/1/A/llO2/1, l.1 for performing a Unit Startup, what is the i
significance of the " Bullets" preceding the substeps following step 2.1?
QUESTION 7.12 (2.00)
AP/1/A/1700/19 " Loss f 4160v Power and the BWST" has the operator realign system components as a result of the f ail ures.
For EACH of the f oll owing state the resultant source of electrical / water supply to the component.
1.
HP1 pump suction.
2.
HPI pump electrical power.
l 3.
Standby bus #1.
4 HPI pump motor cooling.
QUESTION 7.13 (1.00) 1 EP/1/A/1800/01 Section 506 " Unanticipated Nuclear Power l
Production", has the operator " Verify open lHP-5 (Letdown i
Isolation)" prior to initiation of Emergency Boration.
Explain the need for this action step.
)
CUESTION 7.14 (1.50) l Aside from the Main FDW Control Val ve (FDW-32), list the other six valves which must be closed in order isolate the A OTSG if it is ruptured.
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Z:__E5ggggUBg@_ _ygB[9L1_9@NgB[@(1,((g5gggg{AND PAGE 22 Be919L9GIceL_C9NI6gL QUESTION 7.15 (2.00)
Besides tripping the reactor, starting the Keowee Hydro i
Units and announcing to/ notifying the proper personnel, what are the remaining immediate actions if the control room must be evacuated, as stated in AP/1/A/1700/8?
Assume that time exists to complete actions prior to evacuation, and that the
$$F will NOT be recuired to be put in operation.
Include in your answer where operators are dispatched to and what materials are required to be taken out of the control room, i
QUESTION 7.16 (1.50) j List six different methods of heat transfer recovery / core cooling contained within EP/1/A/1800/1, Section 502, " Loss 4
of Heat Transfer", assuming that the Reactor Coolant Pumps
)
t are UNAVAILABLE.
QUESTION 7.17 (1.00)
What are the two iteria that must be met in order to utilize OP/ 1/4/1102/02, " Reactor Trip Recovery".
QUESTION 7.18 (1.00)
AP/1/A/1700/07 " Loss of Low Pressure Injection System" Section A " Failure of One Train of the LPI System During ECCS Operation", cautions the operator that "if only one LPI cooler is cperable, then approx. 6000 gpm LPI and LPSW flow must be established through the operable cooler immediately after swapping LPI Pump suction from the BWST to the R.B.
3 Emergency Sump."
State two bases f or this caution.
QUESTION 7.19
- 2.00)
If the HPI System has actuated due to a low pressure condition, what are the two criteria, of which either one must be met, that must be Considered to secure the HPI System?
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Z:__EBgCEgU$@@_;_yg@[@L3_@@yg@[gLt_E[EggEyCy_gyg PAGE 23 6@DigLgGICgL_CQNIBgL QUESTION 7.20 (1.00)
)
If the RCPs can be effective in cooling the core, even with two-phase flow, why must the RCPs be secured when 0
)
subcooling is noted?
OUESTION 7.21 (1.00)
Recently, during testi ng of the Emergency CCW siphon flow, problems developed resulting in loss of siphon.
"Among the corrective actions were some procedural changes.
Provide the basis for the following procedural steps in AP-11, " Loss of Power";
a)
Steam is isolated to the Condenser Steam Air Ejector J
first stages, j
b)
TDEFW Pump suction is aligned to the hotwell from tne UST as level in the UST decreases.
DUESTION 7.22 (1.50)
In EP/1/A/1800/1, section 502, " Loss of Heat Transfer",
if level has been restored to the OTSGs, and subcooling margin is O degrees, a phenomenons known as " saturated I
repressurization" may occur.
Ev pl ai n why this may occur and what must be done to terminate the pressure rise.
(Assume HPI is available)
QUESTION 7.23 (1.50) j EP/1/4/1800/1, Section 504, "S/G Tube Leak", has the operator initially cool down to 532 degrees while maintaining pressurizer level greater than 80 inches.
What is the basis for each of these setpoints?
)
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21._P699EQU5gg_;_Ng@[@6t_@@NgBD96t_Edg5gENQy_gND PAGE 24
,,,Se9196901Ce6_CgNI606 OUESTION 7.24 (1.00)
According to AP/1/A/1700/13 Section D " Dam Failure Without Loss of CCW Intake Canal" an operator is to be dispatched to place the TDEFDWP Cooling Bypass switch to the " Bypass' position.
Explain what this action would accomplish with regard to the TDEFDWP cooling system.
QUESTION 7.25 (1.00)
Refer to attached Enclosures 7.3A and 7.5.
Assume no RCPs are operating and the RCS is in a natural circulation cooling mode.
The Steam Generator operating range indications have failed and the extended startup range is being used to control level.
If a level equivalent to 8 5'/. O R i r., desired and :he fol1owing conditions exist, DETERMINE the indicated startup range level to maintain.
Conditions are SG pressure = 900 psig and RB temperature
=
250 deg.
F.
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 25 4.
i..
l QUESTION 8.01 (1.00) l Which one of the following is NOT a valid method of performing an independent verification?
a.
Two individuals INDEPENDENTLY observing the breaker for a component is in the correct position, b.
One individual observing that a valve is in the correct position LOCALLY, and another i ndi vi dual using a REMOTE indicator to verify valve position.
c.
Two individuals verifying valve posttion from a remote indicator ONLY.
d.
One i ndi vi dual actually placing a breaker in its required p :;21 t r oc., thUn the SF,MC individual wer i f yi69 from a remote indicator that the breaker is in the correct position.
QUESTION 8.02 (1.00)
Which one of the following is NOT a criteria for the basis of rod position limits in Tech Specs?
a.
ECCS power peaking b.
Power-imbalance boundaries d.
Potential ejected rod worth OUESTION 8.03 (1.00)
Which one of the f ollowing will cause the greatest biological damage to man?
a.
O.1 Rad of fast neutron b.
1.0 Rem of gamma c.
10 Rem of beta d.
0.05 Rad of alpha I
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e __ADg191sIgeIIyg_gegggggggg1_gggg1119391_egy_sig11e11gge exec to z
QUESTION 8.04 (1.00)
Which one of the following correctly describes the required actions to retrieve a Red Tag Stub, if the Work Supervisor responsible for the job is not on site?
a)
Another Work Supervisor in that group may authorize retrieval, as long as he has phone approval from the responsible Work Supervisor.
l b)
The Group Superintendent is.the only individual who may sign authorizing removal, and he must inform the Work Supervisor responsible for the work when he returns tn the site.
l l
c.
The Group Superintendent must approve tar retrieval, but ne may authorize this based on verbal approval and having another individual sign his name and initial authorizing tag removal.
d.
The Shift Supervisor is the only individual who Can authorize tag retrieval in this situation.
QUESTION 8.05 (1.00)
The power-imbalance envelope defined in Tech Specs is based i
l upon which of the following:
1 a.
assure that an acceptable power distribution is maintained for control' rod misalignment anal ysi s, b.
assure that the potential effects of control rod misalignment on steam line break accident analyses are l
minimized.
c.
assure LOCA analysis bounds on maximum linear heat rate for maximum cladding temperature li mi t s.
l d.
assure that the nuclear uncertainty factor in LOCA analyses will not exceed the Final Acceptance Criteria.
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$1__bSb1bl51bb11Yb bbSEES9bb51_bbbSII19bb1_bbS_L}[g[}gg5 PAGE 27 QUESTION 8.06 (1.00) Refer to the attached Tech Spec excerpts for EFW. The unit is operating at full power. Which one of the following is a correct interpretation of the Limiting Condition for Operation for the Emergency Feedwater System, in the event Motor Driven pump A AND all EFW FLOW instrumentation for B OTSG were inoperable. All other attendent controls and instrumentation are assumed to De operable. a. Apply specification 3.4.2.a. b. Apply specification 3.4.2.b. c. Apply specification 3.4.2.c. d. Apply specification 3.4.2.d. QUESTION 8.07 (1.00) Assume full power operations and Refer to the attached Tech Spec excerpts for the Reactor Building Cooling Units (RBCUs) to answer the following: In the event that degraded performance (below acceptance limits) on all RBCUs was identified, which one of the following actions would be correct in accordance with Tech Specs? a. If the RBC system is not restored to meet the requirements of Specification 3. 3. 5. b ( 1 ) within 24 hours, the reactor shall be placed in a condition with RCS pressure below 350 psig and RCS temperature below 250 deg. F within an additional 24 hours, b. One at a time, each RBC train can be take' out of service for 7 days to undergo maintenance. c. Reduce reactor power to a point bel ow which the RBCUs can safely meet design analysis conditions.
- d. Place the af/ected unit in at least Hot Shutdown within the next 12 hours, and in at least Cold Shutdown within the following 24 hours, in accordance with section 3.0.
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9:__9901N1glggIlyg_gggggggggg3,gggg111gggy,$gg_6}g1191}ggy Pabt ZU j l ) QUESTION 8.08 (1.00) Refer to the attached Tech Specs for Radioactive Licuid j Effluents 3.9 and answer the following. As a result of routine liquid releases, members of the public in I unrestricted areas received a calendar year dose of 12 mrem whole body. Which one of the following would be the required action (s) in accordance With Tech Specs? l a. Implement the Oconee Emergency Plan, and submit a report I to the NRC. b. Submit a report to the NRC. c. Submit a report to the NRC, and implement the provisions of Tech Spec section 3.0. d. Restore the concentration of radioactive material released in liquid effluents to unrestricted areas below established limits, and implement the provisions of Tech Spec section 3.0. QUESTION 8.09 (1.00) l Which cne of the following has an associated Tech Spec l Limiting Condition for Operation under 3.10 Radioactive Gaseous Effluents? 1 a. Waste Gas Holdup Tank oxygen concentration. b. Auxiliary Building Exhaust System gaseous effluent. l c. Contaminated oil incineration. d. Gaseous effluent air dose due to particul ates. l l (***** CATEGORY 08 CONTINUED ON NExT PAGE
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9 __gp[]GlpI$@llyg_[@gggpUggg3_ggyp]I]gypt_@yp_6}[j}g}}gNS PAGE 29 QUESTION 8.10 (1.00) Which one of the following describes the BWST minimum volume and baron concentration bases as explain in Tech Specs? a, sufficient borated water is avail abl e wi thi n Containment to absorb 99% of the lodine released in a LOCA. b. the 2 hour thyroid dose at the site boundary will be consistent with the analysis presented in the FSAR for postulated steam generator tube rupture. c. the RCS can be cooled down to less than 280 deg. F from normal operating condidtions in the event of a total loss of offsite power concurrent with a LOCA. d. a sufficient volume of borated water is available for refueling requirements and the concentration is high enough to ensure taht the reactor Will remain 1% subtritical at 70 deg. F without any control rods in the core following a LOCA. DUESTION 8.11 (1.50) Answer TRUE or FALSE to the following questions regarding c9anges to procedures: a) The Keowee Hydro Station Supervisor may serve as the Qualified Reviewer for changes to maintenance or operating prccedures on the Keowee Units. I b) Temporary approval of a procedure shall be signed by the l Shift Supervisor AND another SRO in order to be approved for use. c) It is permissable to deviate from the sequence of steps in a procedure if approved by TWO operators, of which one must be a supervisor with an SRO license. l i i i J (***** CATEGORY 08 CONTINUED ON NEXT PAGE
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l 8. ADMINISTRATIVE PROCEDURESt_CgNDITIQNS _gND_ LIMIT 9TIQNS PAGE 30 1 t e s* QUESTION 8.12 (1.00) Indicate whether the following would require.the Operations Duty Engineer to contact the Superintendent of Operations immediately or on a delayed basis, as outlined in Oconee OMP l-3, " Operations Duty Engineer". a) A Tech Spec Violation that is not a result of exceeding a Safety Limit or a Limiting Safety System Setting. o) Contaminated Injury requiring offsite transportation. QUESTION 8.13 (1.50) Indicate what the exposure limits are for the following conditions of exposure: (assume NRC FORM 4 is current) a) Oconee Maximum yearly permissable Whole Body b) Maximum Planned Emergency Exposure to the Whole Body to Save Lives f c) NRC Ouarterly Skin Exposure Limit j i i OUESTION 8.14 (1.90) ) I i As Unit Supervisor on Unit 2, which is shutdown for ongoing refueling operations, a maintenance request to work on RIA-43 (Vent Par ti cul ate RM), which will require its I deenergization, is given to you for approval. Will you allow this maintenance to take place? Explain the reasoning for your answer. OUESTION 8.15 (1.00) Answer the following regarding use of procedures: a) If a requirement is not met during the Conduct of a procedure, where should this discrepancy be documented 7 b) Within how many days should a working copy be verified by camparison with the controlled copy of a procedure' (***** CATEGORY 08 CONTINUED ON NEXT PAGE
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i 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 31 ; I .s i QUESTION 8.16 (1.00) What type of personnel are evacuated in each of the three categories of Station Evacuation (Category 1, Category 2 and Category 3)? i OUESTION 8.17 (1.00) Refer to the attached Tech Spec excerpts on Operation Safety Instrumentation and answer the following. In the event of a malfunction in two of the Turbine Stop Valve Closure channels that rendered the channel s UNTRIPPED and INOPERATIVE, what would be the required action? DUESTION 8.18 (1.00) Refer to Figure 3.1.10-1 " Limiting Pressure VS Temperature Curve..." and explain the basis for operation in the acceptable region of this graph. QUESTION 8.19 (1.00) In order to utilize a Human Red Tag in lieu of a Safety Tag, what 4 Individuals must agree that the evolution can be done is a safe manner without compromising the safety tagging program? OUESTION 8.20 (1.50) List 5 situations when the NRC Operations Center shall be contacted, f ollowing initial notification of an emergency. QUESTION 8.21 (1.50) } List the 5 individuals, in order of priority, who may serve as alternates for the Station Manager in the Capacity of Emergency Coordinator, if the Station Manager is unavailable and an acting Staton Manager has not been designated. (***** CATEGORY 08 CONTINUED ON NExT PAGE
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$1___ add]N1plB@llyg,[$99%ggggg2_g9991119$p1_$$g_b1$11$11999-PMun 04 QUESTION 8.22 (1.50) List three coincident conditions (including values) for which containment integrity shall be maintained according to Tech S t. e c s. Exclude limiting conditions related to shutdown margin. GUESTION 8.23 (1.00) For the Reactor Core safety limit (Section 2.1) list the three combination of plant parameters which ensttre power peaking is limited for DNBR considerations? OUESTION 8.74 (2.00) List ftve conditions which would result in declaring a movable control assembly inoperative in accordance with Tech Specs. Do not include possible causes for the condition in your answer. QUESTION 8.25 (2.00) The following conditions exist: Unit 1 operating at full power Unit 3 at cold snutdown conditions j One Keowee Emergency Start Circuit channel has been out of service for maintenance for 4 hours. A fuse blows in the tripping coil for the Standby Bus keowee Feeder BreaLer SK1 Explain the Tech Spec consequences of this scenario on the i continued operation of Unit 1 AND the potential operation of Unit 3.
- Note: no other Tech Spec related components are out of service, refer to attached Tech Spec excerpts, include in your answer all associated time restrictions.
(***** END OF CATEGORY 08
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(************* END OF EXAMINATION
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control Copy ims OP/1/A/1102/01 ENCLOSURE 4.1 UNIT STARTUP FROM COLD SETDOWN T0 RCS TEMPERATURE AND PRESSURE { __ 0F 250*F AND.50 PSIG Verify Date Date Init./ Time Init./ Time l 1.0 Initial Conditions 1.1 Procedure Limits and Precautions have been reviewed. 2.0 Procedure NOTE:.1A (Flowchart) should be used as a guide (2.0) by the SRO/R0 to aid in maintaining the big picture. l ) 2.1 Complete Enclosure 4.4 (Pre-heatup Checklist). If unit startup is following a Refueling Shutdown, complete Enclosure 4.6 (Pre-heatup Checklist Following a Refueling Shutdown). \\ Plot an RCS Boron versus RCS Temperature for 1%AX/K S/D j margin curve per PT/1/ A/1103/15 (Reactivity Balance Calculation). Verify that RCS boren will be adequate for a 1*.AK/K shutdown margin ~when rod group 1 is withdrawn to 50%wd approximately 250'F. at Unit Supervisor Form a steam bubble in the pressurizer per OP/0/A/1103/05 (Pressurizer Operation). Unit Supervisor NOTE: Maintain pressurizer level at - 100 inches until RCS 0: concentration < 7 ppb. M
i i EP/1/A/1000/01 Rev, 7 Unit 1 ENCLOSURE 7.3A Page 1 of.2 SG EXTENDED STARTUP LEVEL TEMPERATURE COMPENSATION ~ 400 400 e. .._..__._=...-J.._-_., .---..W....._h........._.:-.===.. ...;; =_ .a ..... :.. n..: - j
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EP/1/A/1800/01 Page 1 of 1 ENCLOSURE 7.5 LEVEL CORRECTIONS FOR HIGH REACTOR BUILDING TEMPERATURE Corrections to Indicated Level Reactor Steam Generator Core Building Startup Operate Full Flood l Temp Pressurizer Range Range Range Tank ('F) (inches) (inches) _ (%) (%) (ft) 100 6 1 1 1 0.0 150 8 6 2 1 0.2 200 15 13 4 3 0.4 150 24 21 6 5 0.7 300 33 30 8 8 1.1 350 45 41 11 10 1.4 400 58 50 14 14 1.9 Modification Factor to Indicated Level Corrections NOTE: Use the Modification Factor on the temperature compensated scales only (Steam Generator Operate Range Level and Pressurizer Level) Steam Generator Operate Pressurizer Prerrurizer Downconer Range Temp (*F) Level Temp ('F) Level 68 1.00 68 1.00 150 1.02 150 1.02 200 1.04 200 1.04 250 1.06 150 1.06 300 1.09 300 1.09 350 1.12 350 1.12 400 1.17 400 1.17 450 1.23 450 1.21 500 1.31 500 1.31 550 1.42 550 1.36 600 1.61 650 1.99 Calculation of Actual Level Pressurizer Actual = Indicated level - (Correction x Modification Factor) Steam Generator Operate Range Steam Generator Actual level
- Indicated level - Correction Startup Range Full Range Core Flood Tank EXAMPLE: Reactor Building Temperature 250*F Indicated Startup Level 220" 1
Correction for R. B. Temperature 21" Actual Level = 220" - 21" 199" l EXAMPLE: Reactor Building Temperature 300*F Pressurizer Temperature 400*F Indicated Pressurizer Level 100" Correction for R. B. Temperature 33" Modification Factor 1.17 Actual Level = 100" - (33" x 1.17) 61"
3.4 SECONDARY SYSTDi DECAY REAT REMOWJ. Applicability Applies to the secondary system require =ents for removal of reactor decay heat. Objective To specify minimu= conditions necessary to assure the capability to re=ove decay heat frem the reactor core. Specification 3.4.1 E=ergency Feedwater System 1 The reactor shall not be heated above 250 F unless the following conditiens are met: Three e=ergency feedwater pu=ps (one steam driven pu=p capable'of' being a. pcwered f rom an operable steam supply system and two motor driven pumps) and associated initiation circuitry shall be operable. b. Two 100% emergency feedwater flow paths shall be operable. Each f'cw path shall have at least one flew indicator operable. 3.4.2 During operation greater than 250 F, the provisions of 3.4.1 may be modified to permit the following conditions: a. One motor driven emergency feedwater pump may be inoperable for a period of up to seven days. If the inoperable pump is not restored to operable status within 7 days, the unit shall be brought to hot shutdown within an additional 12 hours and below 250* in another 12 hours, b. One turbine driven emergency feedwater pump or one emergency feedwater flow path may be inoperable for a period of up to 72 hours. If the inoperable pump or flow path is not restored to operable status within 72 hours the unit will be at hot shutdown within an additional 12 hours and below 250*F in another 12 hours. c. Two motor driven emergency feedvater pumps may be inoperable for a period of up to 12 hours. If an inoperable pump is not restored to operable status within 12 hours, the unit shall be brought to hot shutdown within an additional 12 hours and below 250' in another 12 hours. d. With three emergency feedwater pumps and/or both emergency feedwater flow paths inoperable, im=ediately initiate corrective action to restore at least one emergency feedwater pump and associated eGergency feedwater flowpath to operable status. The unit shall be at hot shutdown within 12 hours and below 250'F in another 12 hours. 3.4-1 A 106/106/103 12/29/81
i 3.4.3 The 16 main steam safety relief valves shall be operable. 3.4.4 A minimum of 72,00'0 gallons of water per operating unit shall be available in the upper surge tank, condensate storage tank, and hot-well. A minimum of 5 f t. (=30,000 Gal.) shall be available in the upper surge tank. 3.4.5 Emergency Condenser Cooling Water (ECCW) System The RCS shall not be heated above 250*F unless the ECCW System a. is operable. b. If the ECCW System becomes inoperable during operation above 250*F, and the system is not restored to operable status in seven days, then the unit shall be brought to hot shutdown within an additional 12 hours and below 250*F in another 12 bours. j l 3.4.6 The controls of the emergency feedwater system shall be independent of the Integrated Control System. 1 4 l 3.4-2 A 128/128/125 4/30/84
Bases The Main Feedwater System and the Turbine Bypass System are nor= ally used for decay heat removal and cooldown above 250 F. Feedwater makeup is supplied by operation of a hotwell pump, condensate booster pump, and a main feedwater pump. Operability of the Emergency Feedwater System (ETW) assures the capability to remove decay beat and cool down the Reactor Coolant System to the operating conditions for switch over to decay heat removal by the Decay Heat Removal System, in the event that the Main Feedwater System is inoperable. The ETW system consists of a turbine driven pump (880 gpm), two motor driven pumps (450 gpm each), and associated flow paths to the steam generators. The decay heat and the reactor coolant pu=g heat following a reactor trip l l f rom 102% power, and the ETW flow rate (90 F f eedvater) required to re=ove this heat demand are as follows: Heat Demand ETW F1 curate Time (% of 2568 MWT) (g?m) 1 min 4.65 721 2 min 4.17 647 5 min 3.64 564 l 10 min 3.28 50? 30 min 2.70 419 1 hour 2.35 365 l 2 hours 2.07 322 The limiting transient requiring maximum EFW flow is the loss of main f eedwater l l with offsite power available. For this transient, a minimum ETW flow rate equivalent to 405 gpm at 1065 psia is adequate. Each of the three ETW pumps is capable of delivering this flow. A 100% flovpath is defined as: The flowpath to either steam generator including associated valves and piping capable of being supplied by either the turbine driven ~ pump or the associated motor driven pump. One flow indicator or steam generator level indicator per steam generator is sufficient to provide indication of emergency feedwater flow to the steam In the event generators and to confirm emergency feedwater system operation. that at least one indicator per steam generator is not available, then the flowpath to this steam generator is considered to be inoperable. The ETW System is designed to start automatically in the event of loss of both All automatic main feedwater pumps or low main feedwater header pressure. initiation logic and control functions are independent from the Integrated Control System (ICS). Normally, decay heat is removed by steam relief through the turbine bypass system to the condenser. Condenser cooling water flow is provided by a siphon effect from Lake Keowee through the condenser for final heat rejectien to the Keowee Hydro Plant tailrace. Decay heat removal via recirculation 3.4-3 A 106/106/103 12/29/81 1
flowpath may be maintained for up to 11 hours per unit, assuming the minimum amount of water in the' upper surge tanks, condensate storage tank, and hotwell is ava!.lable. This is based on the conservative estimate of normal makeup being 0.5% of throttle flow. Throttle flow at full load, i 11,200,000 lbs/hr., was used to calculate the operation time. For decay heat removal the operation time with the volume of water specified would be considerably increased due to the reduced throttle flow. Decay heat can also be removed from the steam generators by steam relief through the main steam safety relief valves. The total relief capacity i of the 16 cain steam safety relief valves is 13,105,000 lbs/hr. In this case the minimum amount of water in the upper surge tank, condensate storage tank, and hotwell is sufficient to remove decay heat and reactor coolant pump heat for 3 hours per unit at hot shutdown conditions. RIFERINCE .FSAR, Section 10. f e ~ ~ A 106/106/103 l 12/29/81 A
p. The 3VST shall cootai: a = t r. : u.: ; vel of t6 feet of.ater l having a =fr.1:c: roncentre : : I 13 3 5'*ppo bo roa :: s,;n; mum tempe ra: re of 50'T ~he 'acasi 21 'r. L? 23. So 'he 5 :n: c lice sba.1'. te Io:ked open 'f these r e q u : r eme r.t i at: -. -- the 3'-37 sh 11 be cor.s;Jer'.'$ anC ': 14ble an'! 2:*aor -
- '21 accordac:e t'h Spec:f;;s'.;on J.
3.3.5 Reactor Butiding Cooliog (RBC) System a. Prior to io: tating maintenance on any component of the R3C system, the reducuant co=ponent shall be tested to assure operautitty b. When the RCS, with fuel in the core, is in a coodition with pressure equal to or greater than 350 pstg or temperature equa'. to or greater than 250*F and suberitical: (1) Two independent RSC trains, each comprised of an RSC f a r. associated cooling unit, and assoctated ESF valves sna'. be opera:1e. (2) Tests or =aintenance shall be allowed on any compenent of tne R3C syste: provided one trato of the RBC and one ::a:: of.the RSS are operable. If the RSC system is not restorec to meet ne requirements of Specification 3.3.5.b(:) ao:.e.::::: 2t hours, the reactor shall be placed in a condit:en with RCS pressure below 350 psig and RCS temperature below 253*F with-in an additional 2t hours. c. When the reactor is critical: (;) In addition to the requireme:ts of Specifications 3.3.5.b(1) above, the rema:nicg RSC fan, associated cool :g un , and associated EST valves shall be operable. (2) Tests or maintenance shall be allowed on one EBC ::a:n under either of the following conditions: (a) One RBC train may be out of service for 2t hours. (b) One RBC train may be out of service for 7 days provided both RBC trains are operable.* (c) If the inoperable RBC train is not restored to meet the requirements of Specification 3.3.5.c(1) within the time permitted by Specification 3.3.5.c(2) (a) or (3). tne reactor sh:11 be placed := a hot shutdown ::nd:*.;:n -::t;- 12 hours. If the requirements of Specification 3 3.5.e( ) are not met within an additional 24 hours follow:ng het shutdown, the reactor shall be placed in a cond:::en vtth RCS pressu:e below 350 psig and RCS temperature below 250*f within an additiona' .t hours.
- For the "3A" RSC train, a one-time extension of idoperability is graoted in order to allow for repair, provided both RBS. trains are operable and that the "3A" RBC train is returned to service no later than#11:59 p.m., April 20, 1985.
- 2010 ppa boron for Unit 3, 'Cy'le 10 only.
c Amendment No. 155, 155, 152 3/19/87 3.3 3 s
. =. l 3 LIMITING CONDITIONS FOR OPERATION 3.0 LIMITING CONDITION FOR OPERATION Specification In the event a Limiting Condition for Operation (LCO) and/or associ-ated Action requirements cannot be satisfied because of circumstances in excess of those addressed in the specification, the affected unit shall be placed in at least Hot Shutdown within the next 12 hours, and in at least Cold Shutdown within the following 24 hours 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 is not applicable. Exceptions to these requirements shall be stated in the individual specifications. Bases This specification delineates the ACTION to be taken for circumstances not directly provided for in the ACTION statements of existing LCOs and whose occurrence would violate the intent of the specification. For example, Specification 3.3.1 requires that two independent trains of the High Pressure Injection (HPI) System be operable and provides explicit Action requirements if one train of the HPI System is inoperable. Under the terms of Specification 3.0, if more than one train of the HPI System is inoperable, the affected unit is required to be in at least Hot Shutdown within the following 12 hours and in at least Cold Shutdown within the following 24 hours. It is assumed that the unit is brought to the required mode within the required times by promptly initiating and carrying out the appropriate Action statement. l i 3.0-1 A 89/89/86 12/10/80 1 J l
3.9 RADI0ACTNT I.IQUID ETTI.G.WS ~ Aeolicabilitv Applies at all, times to the controlled release of all liquid vaste discharged from the site which may contain radioactive materials, except as noted. i Appendix I dose limits for radioactive liquid effluent releases (T.S. 3.9.2) are applicable only during normal operating conditions wnich inc;ude expected i operational occurrences, and are not applicable during unusual operating con-ditions that result in activation of the Oconee Emergency Plan. Obiective l To establish conditions for the controlled release of radioactive liquid effluents. To implement the requirements of 10 CFR 20, 13 CTR 50.3oa. Appendix A to 10 CFR 50, Appendix ! to 10 CFR 50, -0 CTR : 1 anc -0 273 190. l ) Specification 3.9.1 . Concentration l l a. The concentration of radioactive material released at anytime from the site boundary for liquid effluents to Unrest :cted Areas (denoted in Figure 2.1-4(a) of the Oconee Nuclear Stat:.on Final Safety Analysis Report) shall be limited to the concen-tration specified in 10 CTR Part 20, Appendix 3 Table II, j Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases'the con-l j eentration shall be limited to 2 x 10 4 WCi/ml total activity i b. If the concentration of radioactive material released in 1:qu:.d effluents to Unrestricted Areas exceeds the soeve specifiec limits, without delay restore the concentration to V: thin *.ne above limits. 3.9.2 Dose a. The dose or dose commitment to a Member Of The Puol:c fr:m radioactive materials in liquid effluents to '.'arestr::ted Areas shall be limited to. 1) during any calendar quarter: 1 4.5 mrem to the total body i i 15 mres to any organ and; j 2) during any calendar year: < 9 mrem to the total becy I 3 30 mrem :o any organ. ~. A 125/125/122 3 o.1 1/16/84 E
m-3 .a i b. If the esiculated dose from the release of sdioactive materials in liquid effluents exceeds any of the above limits, except l during unusual' operating conditions that result in act: ation of the Oconee Emergency Plan, and in lieu of any other report required by Section 6.6.2, a report shall be submitted within 30 days from the end of the quarter during which the release occurred, to the regional NRC Cffice which includes the following: t 1. Cause(s) for exceeding the limit (s) 2. A description of the program of corrective action initi-ated to: reduce the releases of radioactive materials i in liquid effluents, and to keep these levels of radio-i active materials in liquid effluents in coepliance with the.above limits, or as low as reasonably achievable. 3. Results of radiological analyses of the dr:nking water source and the radiological Impact 'n finished drinking water supplies with regard to tse :_ qui,rements of 40 CTR i 141. 1 3.9.3 Liquid 'w'aste Treatment a. The appropriate subsystems of the liquid radwaste trea'. ment system shall be used to reduce the radioactive materials in liquid waste prior to their discharge, if the projected dose due t'o liquid effluent releases to unrestricted areas, wnen averaged over 31 days would exceed 0.18 mrem to the total body or 0.6 mrem to any organ. b. If radioactive liquid waste is discharged without treatment and i in excess of the above limit, a report snall be submitted witn-in 30 days to the regional NRC Office which includes the following: 1. Cause of equipment or subsystem inoperability. 2. Corrective action to restore equipment and prevent re-currence. 3.9.t Chemical Treatment Ponds (CTP 1 and 2) a. The quantity of radioactive material in the Chemical Treatment Ponds (CT?) shall be limited so that, for all radionuclides identified, excluding noble gases and tritium, the sum of the rstics of activity (in cur:es) to the lim:ts in :0 Cy3 20. Appendix B, Table II, Column 2 snali not exceed i ~ x '. 0
- h 51 1.~ x 103 i
J C) 3.9-2 A 125/125/122 1/16/84 i L
I ,= where Aj = pond inventory limit for single radionuclides 'j' (curies) Cj = 10 CTR~20, Appendix B, Table II, Column 2, concen ration for single radionuclides 'j' (curies) b. After a primary to secondary leak is detected, the initial batch of used Powdex resin shall not be transferred to the CTP. No batch of used powdex resin shall be transferred to the CTP unless the sum of the ratios of :.he activity of the radionuclides identified in the preceeding batch from any powdex cell in the same unit is less than 0.1% of the limit identified in 3.9.l..a. N< 1.0 x 10~3 j Aj where Qj = radionuclides activity in the batch i A.) = pond inventory limit f or radionuclides 'j ' c. The radionuclides inventory per batch of used powdex resin transferred, averaged over the transfers of the previous 13 weeks, shall not exceed 0.01% of the pond radionuclides inven-l tory limit. If this average exceeds 0.01*. of t.he pond radionu-clide inventory limit, then a report will be submitted within 30 days to the Regional NRC Office describing the reason or reasons for exceeding the objective and plans for future i operation. Decay of radionuclides may be taken into account i in determining inventory levels.
- Sd
+ Qj n < .01*. x Aj Qj + Qj2 6-1) g D 1 where Qj = activity or radi:nuclide 'j' :,n the batch i n = number of batches transferred to the chemical l treat. ment ponds during the previous 13-week period. I i 3.9.5 !.1guid Holdup Tanks l a. The quantity of radioactive material contained in each out-( side temporary tank shall be limited to less than or equal l ) to 10 curies, excluding tritium and dissolved or entrained noble gases. Tanks :.ncluded in :nis spec:fication are these outdoor tanks that are not surrounded by liners, dixes, or walls capable of holding the tank contents and that do not { have tank overflows and surrounding area drains connected j to the liquid radvaste treatment system. j i 5. The quantity of rada: active mater al contained in eac: :f j the out.s:.de temporary tangs snall be determined : t e -T the acove limit by ana".y:.ing a representat:/e samele Of : e canns c:ntents at. east ance per - jays vnen ra ncact.ee t.aterials are being accec to tne :anx. 3.9-3 A 125/125/122 1/16/84 L
i c. If the quantity of radioactive material in any outside tem-porary tank exceeds the above limit, suspend all additions to radioactive material to the tank V:thout delay. I 3.9.6 The provisions of Technical Specification 3.0 do not apply. Bases l The concentration specification is provided to ensure that the concentration of rsdioae:1ve materials released in liquid waste effluents from the size to unrestricted areas will be less than the concentration levels specified in 10 1 CTR Part 20, Appendix B, Table II. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its .iPC :.n air (submersion) was converted to an equivalent concentration in water using the methods described in International Consnission on Radiological Precec-tion (ICRP) Publication 2. The dose specification is provided to assure that the release of rad;oae::ve mater:.al :.n liquid ef fluents will be kept "as low as is reasonably actie.ab'.e.
- Also, for fresh water sites w :.3 drinking water supplies unich can be potenr.:. ally affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclides concentrations in the fin sned drinring water that are in excess of the requirements of 40 CTR 101.
The dose calculations in the ODC.1 implement the require =ents in Section III.A of Appendix 1.
- . hat conformance with the guides of Appendix I is to be shown by calcula::.ocal procedures based on models and data such that the actual exposure of an indivi-dual through appropriate pathways is unlikely to be substantially underest. mated.
l Section IV of Appendix I of 10 CTR 50 states that the licensee is permitted the flexibility of operation during unusual operating condit:.ons, to assure
- he public is provided with a dependable source of power when corepatible w :h
- ensiderations of health and safety of the public.
Section I of Appendix ! of 10 CFR 50 states that this appendix provides specific numerical gui* des i for oesign objectives and limiting conditions for opera::.on, to assis: nolders of licenses for light-water-cooled nuclear power reactors n meet:cg :ce re-quirements to keep releases of radioactive material to unrestr:.cted areas as low as practical, and reasonably achievable, durier normal reactor oeeratzens. includine exnected operational occurrences. Using tne flex:.oill:y grantec 1 during unusual operating conditions, and the stated applicabil :y of :ne ie-sign objectives for the Oconee Nuclear Station, Appendix I dose Im:s for i
- sdioactive liquid effluent releases (T.S. 3.9.2), are concluded :.o be no j
applicable during unusual operating conditions tcat result in :ne activat;on of the ocenee F.mergency Plan. For units with shared radvaste treatment systems', the liqu:d effluents fr:m :ne snared system are proportioned among the units snaring that system. The requirements that the appropriate portions of this system be used when spe:1-fied provides assurance that the releases of radioactive materials :n 1: quid ef-fluents will be kept "as low as is reasonably achievable." This spec;ficat :n i ..plements the requirements of 10 CTR Car: 50.36a, General :esign ::::ert:c E
- f Appendix A to 10 CTR Part 50 and design ob'ective Secu:n *.: :: - ren:u
- 10 CTR Par: 50.
1 i 3.9-4 A 125/125/122 1/16/84 L
1 The inventory limits of the chemical treatment ponds are based on limiting the consequences of an uncontrolled release of the pond inventory. The short te rm case limit (2 mres/hr) of 10CTR20,.105 is applied to 10CTR20.106 in the following expression: Ai x 108 uCi x gal 2 mrem /hr 3760 hr 1.3 x 10' gal c'Arie 3755 al 500 mres/yr yr Cj Ai i 1.7 x 103 i Cj l i where Aj = pond inventory limit for radionuclides 'j ' (curies) CJ = 10CFR20 Appendix 3, Table II, Column : concentration for radionuclides 'j ' 1.3 x lod gal - estimated volume of smaller :cemical treatment pond The batch limits provide assurance that activity input to the CTP will be min-im12ed. 1 i l i 3.9-5 A 125/1:5/122 1/16/84 I
F 3.5 INSTRUMENTATION SYSTEMS. 3.5.1 Operation Safety Instrumentation Applicability Applies to unit instrumentation and control systems. Objective To delineate the conditions of the unit instrumentation and safety circuits necessary to assure reactor safety. Specifications
- 3. 5.1.1 The reactor shall not be in a startup mode or in a critical state unless the requirements of Table 3.5.1-1, Column C are met.
1 3.5.1.2 In the event that the number of protective channels operable falls below the limit given under Table 3.5.1-1, Column C; operation shall be limited as specified in Column D. 3.5.1.3 For on-line testing or in the event of a protective instrument or channel failure, a key-operated channel bypass switch associated with each reactor protective channel may be used to lock the channel trip relay in the untripped state. Status of the untripped state shall be indicated by a light. Only one channel bypass key shall be accessible for use in the control room. Only one channel shall be locked in this untripped state or contain a dummy bistable at any one time. 3.5.1.4 For on-line testing or maintenance during reactor power optratton, a key-operated shutdown bypass switch associated with each reactor protective channel may be used in conjunction with a key-operated channel bypass switch as limited by 3.5.1.3. Status of the shutdown bypass switch shall be indicated by a light. 3.5.1.5 During startup when the intermediate range instruments come on scale, the overlap between the intermediate range and the source range instrumentation shall not be less than one decade. If the overlap is less than one decade, the flux level shall not be greater than that readable on the source range instruments until the one decade overlap is achieved. 1 i 3.5-1 A 148,148,145 8/20/86 i j i
t ) uB u u) u u u u u u D nl h h c h h h h h h h ( oot s s( s s s s s s iC o t n t) t) t t t t t t cf n ob ob o o o o o o t AO a h( h( h h h h h h h C r os os os os os os os os o t r t r t r t r t r t r t r t r t t u u u u u u u u a go go go go go g o g o go r nh nh nh nh nh nh nh nh e i i i i i i i i i p r2 r2 r2 r2 r2 r2 r2 r2 O B1 B1 B1 B1 B 1 B1 B1 B1 B SE ) ) ) ) ) ) MLL a a a a a a UEB ( ( ( ( ( ( ) MNA 1 1 1 3 3 3 3 3 3 CI NR S ( NAE N I HP O MCO I T I DNO 1 C 1 G S N LP
- 5. T I
EI ) NR 3A B NT A A R ( A N N 1 2 2 2 2 2 2 EE iO l LP CT BO ATSTNEM U R .S T OL S NE N ) N I A LN ( AA 2 2 1 4 4 4 4 4 4 TH OC T F O er us e s s t n n r l e nt t o o t u e r a n n i i s n n t n P t t l o e a s n om l e a a ae t s t m rl a t ou o l l t e t n t l n u ee eh n C r o ng nn u e a r pn cC a t C en ea b n l t mn n ma mh h en o s ea at o on r r s l uR uC s ga o n Th l n o t I oC r r u nh CI - C a e C c t T t e t e P aC e h m a es c I s t s g R r e r nu r e rl a N na nn l t o r ut I r o R u e e U I i I a a rn t u s n t t s n R d s R u ee ct s s e xs c h s n L r el r n wm a al em u n a ge a. w A a me a e a o u e r e r u l I N e r n e c H P r R en P r F R l P C e i rh o O i en l l r I t p n t w I ct a c u S S s S ma S s S o S T u nh u o P P u P eh P n Pl P C NI C NS R Rl RTC HI RF H a b IN t F 1 2 3 4 5 6 7 8 Z Nqg 7My* Y 'f"
d d d d d d d I me t t t t t t t ) uB u u u u u u u D nl h h h h h h h ( oot s s s s s s s i C o t n t t t t) t ) t) t cf n o o o o e o e o e o AO a h h h h( h( h( h( C r os os os os os os o o t r t r t r t r t r t r t t u u u u u u a go go go g o g o go go r nh nh nh nh nh nh nh e i i i i i i i p r2 r2 r2 r2 r2 r2 r4 O B1 B1 B1 B 1 B1 B1 B SE ) ) ) MLL a a a UEB ( ( ( ) NNA 3 3 3 3 2 2 CI NR ( NAE I HP HCO ) d tn S o LP c EI ( ) NR B NT S ( A 2 2 2 2 2 1 1 N O I I O CT I T 1 I - D 1 N O
- 5. C 3G
.S N OI E. T I NE ) N I B A A LN AR ( AA 4 4 4 3 3 2 2 TE TH P OC O T F S O T N E N U g r n t n s R g e o n i T n r t i a d l e l S i o a t ) l t l a n N d t w a s on i u n I l c d l m oe ut n a i a e d o e C m B n a h u e e e n s t u e H C B R n F aI s r r r m s i e y ot o u c c rl y b n r mgS t s t r i i oe r r i u en c n ct g g t n o u a st i l aI a s on o c n t T H s sd a e en L o L) a a a eyl i R es RI s t T eh pm f f rSl t rl l l t l I RC i e o o P an gue gG e a u ad N ct nB e os n oI n t b t n U h e i s s s h o s l s n l S n ih g r t y s s gi r s a ea aP a g s g i a I i u nS o o i t oe n rh n h i i u A f s A L L ct - APC A4 C DP D( i I N s p l e c n O S e S i F ja o P r P a I S neN T RP HT a b EI R( a b c d C N U F 0 1 2 1 1 1 = ,f D $ ,j 2 Iv a;3 n "c n y
') 'h _IHEggy_gE_Ngg6g@$_{gWgg_P(@NI,,g[gB@llg h _{(glgh,@yg PAGE 33 j IUERDQDM @d1Cj} $NSWERS'-- OCONEE 1, 28< 3 -87/07/13-CASTO, C. 4 C i ) ANSWER 5.01 (1.00) f/Ld [4 y / d REFERENCE ) OP-OC-SPS-PTR-AT pp 13/14; LO la (4.i/4.7) 000074A207 (KA'S) ANSWER 5.02 (1.00) b Q R. C. REFERENCE OP-OC-TA-NT Figure 15; LO lo o P-CC - SPJ -J/- N3 FP N / (3.6/3.8) 016000G015 (KA'S) ANSWER 5.03 (1.00) d. REFERENCE Oconee OP-OC-CH-RC pp. 17 3.4/3.9 OO2OOG010 ...(KA'5) ANSWER 5.04 (1.00) d. REFERENCE Oconee OP-DC-SPS-RT-IP pp. 28 obj 2a 2.8/3.1 192Or. 9K 105 (KA'S) ANSWER 5.05 (1.00) i C. e a
c/ h.. 5.__IHEgEy_gE_ygCLE@E_EgWEE_EL@NI_QEEE@IlgNt_ELglg51_gNg PAGE 34 _T H_ E_ R_ M_ O_ D Y_ N_ A_ M_ _I C_ S_ ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. REFERENCE Oconee OP-OC-SPS-THF-PD pp. 10 obj. 2d 2.9/3.3 193OO9K107 ...(KA'S) 1 ANSWER 5.06 (1.00) d. I REFERENCE l Oconee Duke Thermo pp. 156 2.3/2.4 10iOO4K111 ...(KA'S) ANSWER 5.07 (1.00) O R_ O b. REFERENCE Oconee Duke Thermo pp. 194 3.9/4.2 o p - OC-S PJ - Pne - />41 pp 11- 0 193OO8K121 (KA'S) ANSWER 3.08 (1.00) a) More rapid depressurication (+.5 ea) b) Level does not drop as rapidly and will have a slight insurge REFERENCE OP-OC-SPS-PTR-AT pp 12/ Fig 04; LO la (3.4/3.7) OOOOO8A212 ...(KA'S) 1 l ANSWER 5.09 (1.50) l a) 1. Increase (+s5 ea) l 2. Decrease I b) broached tube support plates REFERENCE OP-OC-TA-NT pp 16; LO ik l l o
7: 5 __IH@g@y_Q[_ygg6[@@_[Qb[@_{6@yI_Q((@@llgyt_{691pg3_@yD PAGE 35 T_ H_ E_ P_ M_ O_ D_ Y_ N_ A_ M _I C S_ dNSWERS.-- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 5.10 (1.50) a) Decrease (+.5 ea) b) Decrease c) Increase REFERENCE OP-OC-TA-NT pp 7/8; LO lb (3.6/3.9) 041020A202 fKA'S) ANSWER 5.11 (1.00) a) Lower (+.5 ea) b) Lower REFERENCE OP-OC-SPS-RT-RBC pp 26/27; LO 2b.4 (3.6/4.2) OO1010A207 ...(KA'S) ANSWER 5.12 (1.50) a) Decreases (+.5 ea) b) Increases c) Increases REFERENCE DPC Thermodynamic cs/Fl uid Flow. CH 2E (2.6/2.8) 191004k115 ...(KA'G) ANSWER 5.13 (1.00) 1. less than 2. greater than REFERENCE Oconee OP-OC-SPS-APC-T41 3.8/3.9 192OO2K114 ...(KA*S)
5.__IbE9Ey_gE_yggLE@B_EgWEE_EL@NI_gEEE@IlgN1_ELylgS _@NQ PAGE 36 z ISEE[ggyN@g((S kNSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 5.14 (2.00) a. about equal b. will be eaual to Tsat in the [cafJ c. the isolated generator (l oop ) will lag the operating loop d. should not be greater than 50 deg F. REFERENCE Oconee OP-OC-SPS-PTR-AM-1 obj 9 4.2/4.3 OOOO11A209 ...(MA'S) ANSWER 5.15 (1.00) Negative (+.5) Due to colder water nearer bottom of the core (Higher allowable Kw/ft) (+.5) REFERENCE OP-CC-SPS-IC-RPS pp 12. LO lb (3.1/3.3) 012OOOK502 ...(KA'S) ANSWER 5.16 (1.00) Quench Tank Pressure (+.5) stops increasing (+. 5) REFERENCE OP-OC-SPS-CM-PZR pp 17; LO 11 (2.6/2.8) OO7000A206 ...(KA'S) I ANSWER 5.17 (2.C0) { a) Increase due to the decrease DP across the FDW control valve (+.5) Levels off due to reaching its high speed stop (+.5) b) Tavg has decreased, so power-is increased C+.5) e) Loop B TC has increased, Loop A Tc has decreased (+.5) REFERENCE OP-OC-TA-NT pp 13/14; LO 1h (3.1/3.4) 059000A212 ...(KA'S) l
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 37 ~ [ ~ ~ T H_ E_ R_ 5_ 6_ 6 ~ N_ d_ 5_ _[ C_ 5_ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ hNSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 5.18 (1.50) 1. High injection water boron concentration 2. boiling off of RCS coolant in the core due to decay heat 3. Low thru core flow rate REFERENCE Oconee OP-OC-SPS-PTR-AM-1 pp. 37 obj 19 ANSWER 5.19 (1.00) 1) Opening of steam reliefs which don't reseat-not designed to pass water (any two at +.5 ea) 2) Loss of TDEFW pump from water entering turbine 3) Rapid cooldown of OTSG tubes-tube rupture
- 4) 00fKsitet t3 /WL tutspottT!
C) Tut?5/ul Str10M D 9kMr C ) PC*
- b
'l0 Af P'A#I* IONII REFERENCE % WE797~ [TRfff OP-OC-SPS-PTR-AT pp 30/31; LO l e j op-or 5PJ r,'f-H 7~ pp9 7 (2.7/3.1) o p-O c -SPI-c H-C4 pio J/ 059000A203 ...(KA'S) ANSWER 5.20 (1.00) 1) Radiolysis (+.33 ea) r 2) Z.oc-boric acid reaction og srYo.'t, traT c /h<4 m u O' 3) 21 r c Water reaction s REFERENCE OP-OC-SPS-HY-HDC pp 10-12; LO la (2.9/3.6) 028000K503 ...(KA'S) ANSWER 5.21 (1.00) The addition of Burnable Poison Rods slows he Lt WOT@ f l 8)d ec r ea s e, & nja # g early% u,yu p op -(tut % p roludt ws8 cMr co st [bt D *S especial in core life. REFERENCE OP-OC-SPS-RT-RBC pp 12; LO id.1 (2.5/2.7) 192002K109 (KA'S)
S__IdEggy_gE_UgCLE@@_EgE_Lh_ GEE 5911gyz_ELylgSt_99Q PAGE 38 t IEE5dOpyb@$1CS 1 1 ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, r. ANSWER 5.22 (1.00) These rods are GRAY ( l ower absorbtion cross section' and have a longer effective poison length. (+.5) They have a imbalance (and mintmine less severe impact on Axial Flux oscillations)(+.5) induced xenon PEFERENCE OP-c;C-SPS-THF-PD pp 12; LO 2f.4 (3 '/3.5) 192005K114 (KA'S) ANSWER 5.23 (1.00) There is an induced temperature stress which is compressive on the inner wall and expansive on the outer wall. [0.$5] There is an induced pressure stress which is compressive on the inner wall and expansive on the outer wall. [025] REFERENCE Occnee Duke Thermo op. 225 3.8/4.1 193010KlO7 ...(KA'S) ANSWER 5.24 (1.00) The intent of their use is to restore natural circulation. In dcing this, gases in the RCS should be mixed wi th RCS I liquid to eliminate gas pockets. [0.5] Sumping the RCPs condenses the steam bubble by cooling it I (and starting reflux boiling). [0.5J j REFERENCE Oconee Op-OC-SPS-PTR-AM-2 obj 1.4 4.0/4.4 00074K311 ...(KA'S)
5. THEORY OF NUCLEAR POWER PLANT OPERATIONt,FL,IIDS _AND PAGE 39 t IUE6DODYN@dlCS ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 5.25 (1.00) at 2240 psia, hg = 1115 BTU /lb (+.5) 1156 BTU /lb and at 20 psi a, at saturation conditions, hg = 196 BTU /lb hf = .043 >>> 95.7% quality calculate: (1156-1115)/(1156-196) = If use Mollier: 95% quality (+/- 1%) REFERENCE OP-GA-SPS-THF-STM pp 20/21; LO 2e i~.3/2.n) S; I c? 300 3 r l '- .m ANSWER 5.2e (1.00) C REFERENCE DPC Thermodynamics, pp 192-5 (3.1/3.4) 002OOOK501 ...(KA'S)
6:__PL@NI_gYSIg[g_QESl@N,._CgNI@gL1_9Ng_lg@l@y[gNI@IlgN PAGE 40 3 ANSWERS -- OCONEE 1, 213 -97/07/13-CASTO, C. ANSWER 6.01 (1.00) c. REFERENCE i Oconee OP-CC-PNS-CRD pp. la Obj. 1.b. 3.4/3.6 OO1000KIO3 (KA'S) ANSWER 6.02 (1.00) d REFERENCE OP-OC-SPS-SV-HPI pp 27/28, LO 7b (3.8/3.9) 006050 GOO 7 (KA'S) ANSWER 6.03 (1.00) C REFERENCE OP - OC - SP S - S Y --E F pp 21/22, 32/33: LO le, if (3.4/3.5) 0610004207 ...(KA'S) ANSWER 6.04 (1.00) e 3, REFERENCE OP-OC-SPS-EL-EPD pp 24; LO 3c.5 (2.e/3.2) 062000K401 ...(KA'S) j 1 ANSWER o.05 (1.00) l B l
6___P(gNI,@y@IED@_gE@l@Nz_CgNIggyt,9NQ._lygI5y[ENI@Ilgy PAGE 41 bN'SWERS -- OCONEE 1, 2&3 - 8'7 / 07 /13-C AS T O. C. REFERENCE OP-DC-SPS-SY-CC pp 19; LO 2c (3.2/3.2) OO8010A301 (VA'S) ANSWER 6.06 (1.00) b. REFERENCE Oconee Op-DC-SPS-IC-ES obj.h 3.6/4.2 3.7/4.2 013OOOA204 013OOOA205 ...(KA*S) SNSWER 6.07 (1.00) c. REFERENCE Oconee OP-OC-SPS-SSF-EPS 2.6/2.7 062000A404 (KA'S) ANSWER 6.08 (1.00) d. REFERENCE Oconee OP-OC-SFb-SY-EF objs 1.b/1.a/1.0 3.4/3.8 061000A204 ...(KA*S) ANSWER 6.09 (1.50) wctFnW$ a) .I+eo+ easer (+.5 ea) b) Opens c) Decreases REFERENCE pp 4t/ OP-DC-SPS-IC-ICS, pp 80/81; LO is l (3.0/3.1), (3.4/3.6) 016000K303 016000K312 (KA'S) l I
6.__EL@NI_gySIEgg_QEgl@yz_CgNI3gLg_9NQ_lNgI5ggENI@Ilgy PAGE 42 ,' ANSWERS'-- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 6.10 (1.50) a) Bistable (+.5 ea) b) Buffer c) Buffer REFERENCE OP-OC-SPS-IC-RPS pp 19-23; LO If, 3a, 3b (2.8/3.3) 012OOOK601 (KA'S) g y.j,g g,) ANSWER 6.11 (1.50) a) Key Switch in ICS Cabinet (+.5 ea) iw a wsr c a--eent r oi poeth eonf/cl fo#A (* * K b) Ve41 - - mi c) Key Switch in ICS Cabinet REFERENCE t9 (>/s /4 /#700 [ 2 3 /w(L (. 2. OP-DC-SPS-IC-ICS pp 53, 58; LO 1t j (3.5/3.5) OOOO57A106 ...(KA'5) ANSWER 6.12 (1.00) Possible overpressurization of Aux Steam and subsequently, steam line to TDEFWP if both control valves were to fail open due to too low of r el i ef capacity in both systems. REFERENCE Oconee LER 87-003 3.3/3.4) v3OOOOK107 ...(kA'S) ANSWER 6.13 (1.00) A letdown line from the Common letdown line (prior to the L/D coolers) thru HP-(426) to the fuel transfer tube.CO.52 A line from the SSF Makeup Pump which takes suction on the altremate fuel transf er tube, injects Water through SSF-HP-(398). This line serves as system makeup.EO.5] 4
6. PLANT SYSTEMS _ DESIGN _CQNTRgL, _AND_ INSTRUMENTATION PAGE 43. 3 , ANSWERS -- OCONEE 1, 2L3 -87/07/13-CASTO, C. REFERENCE Oconee OP-OC-SPS-SY-hPI obj 3.a/b 3.3./3.2 On4000K405 ...(KA*S) ANSWER 6.14 (1.00) 1. To insure suction to the HPI pumps is not lost [0.53 2. Insure gas does not enter the HPI Pump suction on an ESFAS [0,53 PEFERENCE Oconee OP-OC-SPS-SY-HPI pp. 19 obj 5.b 2.8/3.0 3.1/3.4 0040010KO1 OO4020A104 ...(KA'S) ANSWER 6.15 (1.00) The Motor Fault Lamp illuminates CO.253 and the Diamond station swaps to manual CO.753 7EFERENCE 0;onee OP-OC-SPS-IC-CRI 1.1/1.g 3.8/3.8 001402K402 ...(KA'S) ANSWER 6.16 (1.00) Inputs = Signal from both the upper and lower linear amollfters (top and bottom detectors) J Outputs = Imbalance meters, RPS, function generator (Flux / Flow /!mbalance), Computer [ea 8 0.23 REFERENCE Oconee OP-OC-IC-NI pp 26 obj 1.g 2.6/2.9 015000K602 ...(KA'S) ANSWER 6.17 (1.00)
- 1) the SSF incoming feeder breaker from Unit 2 M.ain Feeder Dus #2 (OTS1-1) is open
(+.5 ea)
- 2) SSF Diesel Generator Breaker OTS1-4 is closed REFERENCE
'OP-OC-SPS-SSF-ASW pp 24; LO 2h (4.0/4.2) I j
6:__E69NI_SygIEDS_gE@lGyz_gggI6g61_999,1NSI69[ENI@Ilgy PAGE 44 ktJSWERS -- OCONEE 1, 2L3 -87/07/13-CASTn, C. 06'1000K406 ...(KA*S) ANSWER 6.18 (1.00) 1) Main Steam to Aux Steam Regulator 2) FDW-315/316 3) Main steam to TDEFW Pump REFERENCE OP-OC-SPS-SY-IA Tracking 86-099; LO 1g (3.2/3.5) 078000K402 ...(KA'S) ANSWER 6.19 (1.00) 1) Weight re1uction (+.25 ea for any 4) 2) Greater r el i ab i.'I t y (f ewer components) 3) Single indexing 4) Improved Serviceability 5) More room for operator e) Mcre room for optional equipment (e.g., camera) PEFERENCE OP-OC-SPS-FH-FHB pp 32: LO le, if (2.1/3.0) 034000k601 ...(KA'S) ANSWER 6.20 (1.50) 1. Nuclear flux channels 2. Control rod position 3. Reactor trip signal 4 Control roc drive breaker position any 3/4 for .5 ea. REFERENCE OP-OC-SPS-IC-SPDS pp. 17 ob).l.1 2.8/3.3 2.4/2.7 OO1000K602 OO1000K605 ...(kA'S)
e___P6@NI_@y@IEy@_pE@l@yt_CgNJgg63,999,lg}}$U[ENI@Ilgy PAGE 45 ,ArfSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO. C. ANSWER 6.21 (1.00) Normal incoming switchgear breaker OTS1-1 trips SSF Suppl y breaker 82T-4 trips CO.5 ea3 REFERENCE Oconee OP-OC-SPS-SSD-EPS pp. 36 obj 1.f. 2.8/3.1 062000K403 (KA'S) L1 -807" [/* O ANSWER 6.22 a) The Megawatt error is blocked in track, so the header l pressure error is con trol l i ng. posi t i on of the control l val ves. (+.5) This causes instability in header pressure, hence feed flow oscillations. -tr+- --- Al ! I C E - - e t-+t i ai m a ii Lv e u i umat+e--med - b N or-e-tAe -t-ur-ttwa is_.a.Lacad-iwut-e. '^.5" /)v [gg REFERENCE OP-OC-SPS-IC-ICS Tracking #86-064; LO lb (3.2/3.2) 1 l 059000K107 ...(KA'S) I ( ANSWER 6.23 (1.50) i 1) Use HP-26 Motor operated val ve ( +,5 ea) 2) Use HP-122 (Manual Bypass) (-* F o 9 ) 3) Use "C" HPI pump via HP-27 4) oPFu HP-Wo d Sc/c.nce bl % v/ 4.Tbu REFERENCE op/ilh [l/0*/ /0 Z-OP-OC-SPS-SY-HPI pp 29; LO 3c (3.4/3.6) 078000K302 ...(KA'S) I
et__P669I_gy@lEUS_QESlpNt_CQNlgg(1_99Q_ly@l@yyEyl@llgy PAGE 46 .'AN'SWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 24 (2.00) a .,<.w a r i An accident coulo impose w e stresses on the lower portions of ' '7 e ductwork, causing de'ormation or possible collapse of the ductwork, thus blocking airflow [u.53 Fusi bl e dropout plates are designed to melt and drop out of the ductwork opening up additional flow paths assuring d posi ti ve path for recirculation during an accident [0.5] Severe stresses imposed on the ductwork could possibly generate a shock wave through the duct and damage the cooling co11s.CO.5) Blowout plates located .n the lower portions of the duct would be forced out under these stresses and attenuate the shock wave before it reaches the cooling co:1.EO.5] REFERENCE Cconee OP-CC-EPS-SY-PBC pp. 11 objs 1,b/1.c 2.9/3.2 022OOOK301 ... WA'5) ANSWER 6.25 (2.00) I a. 1, withdraw 2. ICS-runcack W-reduces jrJ u -> tIw 3. Reactor trip on high RCS pressure (or neutror flux) e - e 4 f a t l s c l osed (due to false low pressure signal signal) falls closed que to false icw Uressure b. The heaters remain on (du? > false low pressure t due to
- a above) pressure continues to increase.
O #' M *M# (part a. O.3 ea. part b. O.2) QK f.Qu cnw /4 fhswArVit.- REFERENCE Oconee Si mul a t or Ref Manual TCS LP 3.6/3.9 012000A202 (KA'S) l 1 1 1 I )
21__E69EE995EE_2._U96U961_9EU96U8h1_g[E6@@M9L999 PAGE 47 ,5991969G1ce6_C99I596 ANSWERS -- OCONEE 1, 2L3 -97/07/13-CASTO, C. i ANSWER / 7. 01 (1.00) / Y 9 g / REFERENCE Oconee Op/0/A/1105/09 3.3./3.5 001010A010 (KA'5) ANSWER 7.02 (1.00) 19 d V/ / REFERENCE DPC EPG pp 2-77/79; (4.0/4.4) OOOO74K311 (KA'S) ANSWER 7.03 (1.00) b j .i REFERENCE OP/2/A/1102/01, Enclosure 4.3, pp 4 (3.6/4.2) OO1010A207 ...ikA'S) ANSWER 7.04 (1.00) l The TOTAL */. movement from an initial position in EITHER direction. (i. e. max i mum movement in either direction) REFERENCE Ocenee OP/3/A/1102/04, Enclosure 3.1, pp 2 (3.3/3.5) OO1050G010 (KA'S)
21__E60gEggBg5_;_N9BD@6t_@@yOSD96t_EDg5@Eygy_99D PAGE 48 l R_ A_ D_ _I O_ L_ O_ G_ _I C_ A_ L_ _ C_ O N_ T R_ O_ L_ ANSWERS -- OCONEE 1, 2n3 -87/07/13-CASTO, C. j ANSWER 7.05 (1.00) e. REFERENCE Oconee AP/1/A/1700/23 3.5/3.5 3.2/3.4 OOOO57A105 OOOO57A106 ...(KA'S) 1 ANSWER 7.06 (1.00) d REFERENCE Oconee SY-LPI-5 3.2/3.5 l OO5000K402 ...(KA*S) ANSWER 7.C7 (1.00) a) Cycled electrically (+.5) b) 500; 6 (+.25 ea) REFERENCE Oconee OMP 2-1, pp 3/4 enci 4.4 l (3.5/3.6) (3.4/3.7) OO6050G010 061000GOLO (KA'S) ANSWER 7.08 (1.00)
- 1) Close the Field Breaker
(+.33 ea)
- 2) Close the Generator Supply Breaker
- 3) Close the Field Flashing Breaker REFERENCE OP-OC-SPS-CM-KHG pp 20; LO ik (4.0/4.3) 064000A401
...(KA'S) l l l i 1 l i i l
~~~~bb5[5Ehh[bbESIsb bhE ~~~~~~~~~ EMERGENCY AND PAGE 49 7. PROCEDURES - NORMAL, ABNORMAL, ~ ~~ ~~~~~~ ~ ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 7.09 (1,00) a) Refueling SRO or Shift Super vi sor (+.5ea) c) In a cabinet below the operator's console on the bridge REFERENCE OP-OC-SPS-FH-FHB pp 23; LO la (3.0/3.0) (2.3/3.9) 034000 GOO 1 034000G009 ...U<A'S) ANSWER 7.10 (1.00) 300 inches ensures that a large enough bubble exists in the pressurizer fcr pressure control. REFERENCE Oconee CP-604: EPG Reference Document pp 3-96/97 (4.2/4.5) 000009K321 ...(VA'S) ANSWER 7.11 (1.00) This identifies steps that may be done in parallel with the crittcal path (numoered) step. REFERENCE Oconee OP/1/A/1102/01, pp 3 (4.1/3.9) 194001A102 ...(kA*5) i ANSWER 7.12 (2.00) 1. From the Spent Fuel Pool 2. Aux. Service Water SwitchgearCE 9A'MY OA 3. From C T -5 ott (E $7A TI&v O R SruMyOs1#/ 4 From the Aux. Service Water System e4 MPDu 1 [0.5 ea.] REFEAENCE AP/1/A/1700/19 Otonee 3.7/4.1 3.9/4.7 oP-JC-TPl~N 0"4 L 00C055A203 000062A211 ...(KA'9 o p-oc I'l-h 6 PP 2y t A
Z:__PBQCgpu@[S_ _GQ@[@Lt_@@Ng@[@L _E[E@gEyCy,@ND PAGE 50 ,60D19L991 COL _G9NI@QL s ANSWERS -- OCONEE 1, 2L3 -87/07/13-CASTO, C. i ANSWER 7.13 (1.00) Letdown should be established to 04fset the increase in RCS inventory due to the initiation of E-boration. REFERENCE Oconee EP/1A/1800/01 4.4/4.7 000029K312 .(KA*5) ANSWER 7.14 (1.50)
- 1) SU Cntrl Valve (FDW-35)
(+.25 ea)
- 2) EFDW Cetrl Valve (FDW-315)
- 3) TBV Block Valve (MS-17'
- 4) MS to SSRH (MS-79)
- 5) MFW Block Valve (FDW-31)
- 6) SU Block Valve (FDW-33) 4
) REFERENCE Oconee OMP 2-1, pp i enci 4.4 (4.1/4.2) 000040G010 (KA'S) ANSWER 7.15 (2.00) 1) Set batch size on makeup control to 90,000 gallons and reset (+.25 ea) 2) Open Makeup isolation (HP-16) 3) Open RC Bl eed Transfer Pump 'A" Discharge CS-46 4) Start "A" Bleed Transfer Pump 5) Dispatch operator to Units 1/2 Waste Ditposal Panel 6) Go to Aux Shutdown Panel with: Reactor Log (+.5 total) Energency and Abnormal Procedures Removal / Restoration Book ) Emergency Plan I 7) Maintain Hot Shutdown Cuiditions 1 REFERENCE Oconee AP-8, pp 2/3; OMP 2-1, ENCL 4.4 (4.1/4.2) l 000069G010 ...(KA'S) 1 l
1 2:__@@gCgDy@g5_;_Ng@d@(1,@@Ng@[@63,(([BQgggy,@ND PAGE 51 Se91969G1ce6_c9NI3g6 .s. ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 7.16 (1.50) 1) US t of FWPs ( +. 25 e a rb" # ## 2) Us e of EFWPs from t'h e affected or other units 3)- Hi 1 Cooling through PORV 4) Hd i Cooling through Head Vents 5) EsF-ASW to S/Gs 6) C8Ps to S/Gs 7) Dump steam from S/Gs h (71W ~{W $* bMj REFERENCE Oconee EP/1/A/1SOO/1 (4.0/4.4) OOOO74K311 ...(KA'S) ANSWER 7.17 (1.00) 1) Reactor startup within 12 hours of trip (+.5 ea) 2) Cooldown has not been initiated. REFERENCE Oconee OP/1/A/1102/02, pp 1 (3.3/3.5) 001050G010 W A' S) 1 ANSWER 7.18 ( 1. OO ) 4er a ug L 1 To provide adeouate core cooldown rate (0.53 and maintain R. C. environmental qua}ification criteria for R.B. equipment protection [0.53. Q(2 - f rodEk GN M4,11 (QOlt'q E kOYM TuMO NWS s tNht" l f i l REFERENCE tatt of 3ve6.rm f.cer/gy,+upt(hi,) Oconee AP/1/A/1700/07 pp. 5 3.4/3.7 UO6010KO10 (KA*S) i i l
v 7 PROCEDURES'- NORMAL t_ABNgRM961_EMERGENCV_ g t_J D PAGE 52 1 sed 19L9GICeL_CgNI@g6 j s. ANSWERS -- OCONEE 1, 2&3 --87 / 0 7 /13-C AS T O, C. ANSWER 7,19 (2.00) 1000 gpm in 1) LPI is in operation, with a flow rate 20 minutes (+1.0 ea) each line and stable for 4 DR I ) Al l M:5 hot an# 4ald 1 eg s a t;ft eE)st 50.degr,e6 1ess tpdB l j l gne s aturation te no for 5' pressure a nd's ec ring,M41 necesbarf to prevkepPfh,RC ts level kpwg'oing of-c ale htgh FEFERENCE [gf[/)//foO/0/ OP-OC-SPS-IC-E5 pp 24: (4.1/4.6) OO9000A324 ...(KA'5) l ANSWER 7.20 (1.00) \\ l o m i n i m i ;' e inventory depletion, such that if the pumps were lost inadvertently, core funcovery could occur. I PEFEh'NCE Oconee EPG Peterence Document, pp 4-45/46 (4.1/4.2) OOOO74K308 ...(KA'5) i ANSWER 7,21 (l.00) I e) Allows the system to keep a vacuum on the condenser i I b) Allows more UST Water available to cool CSAEs (+.5 ea) ~9 '!L #I# 4" C#
- #i"' N ID###P REFERENCE Oconee LER 86-011; AP-11; (4.3/4.6)
OOOO55K302 (KA'5) 1 1 I l l l l l l l Ia
2___809EE993E3 ~_U98U962._9EU95U662._EDE6@@bGY_@bQ PAGE 53 sed 1060GICe6_ Cot! iso 6 ' ANSWERS -- OCONEE 1, 2L3 -87/07/13-CASTO, C. A N'_iW E R 7.22 (l.50) I l Saturation Repressuri:ation occurs due to 1ack of a condensing surface being established in the OT5G for refA l ux bolling to occur and the RCS repressurt es along the saturation 1ine due to decay heat absorbtion. 4*4-,4+- (f. 7 ) To terminate, need to open head vents or PORV 1mv67'(t/.o) CC /gne/pacJeg (fDM(w 0' c t ',D;b +0 cm ret adebM miccyc h h.L
- { W me m m p t%
M REFERENCE se !w v " ' r y t u.i.ivi, im l A,. } OFT 4 Oconee EP/ 1/4/1800/1, pp 26; EPG Reference D6tumen't pp 3-21 ((3.7/4.2) 000074k302 (KA'5) ANSWER 7.23 (1.50) 532 degrees ensures that the saturation pressure is well below the 11 4t setpoint of the 1owest set MSRV (+.75 ea) 80 incnes ensures that you are above the heater level so tnat normal RCS pressure control can be used and so subcooled margin cannot be lost REFERENCE fronee EPG Reference Document, pp 3-37 (4.2/4,5) 000035K506 ...(K4'S> ANSWER 7,24 (l.00) (this would deenergi:e both SV-209 and -208) causing LP5W-138 tc fai1 open supplylng HP9W to EFWPT p utt p bearings C O. 5 ] and causing HP5W-184 to open providing cooling water to the 011 cooler [0.5] GEFERENCE Ocoree AP-13 pp. 8 obj. 1.o, 1.p., 3.1/.3.3 000075K007 ...(KA'5) AN5WER 7.25 (1.00) 284 Inches +-3 inches
31_ _559EE990EE- :_U95U961_9EU95U96 t _EUESEEUEY_999 PAGE -54 . *599196991Ge6_GQNI696 , w ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. REFERENCE Oconee EPs 3.7/4.1 103000A101 ...(KA'S) i ] 1
S __690191E18911ME_C899E993EE1_E9U911I9951_699 61U11911gNS PAGE 55
- A'NSWERS -- OCONEE 1,
2&3 -87/07/13-CASTO, C. 9 ANSWER 8.01 (1.00) f REFERENCE Oconee Station Directive 2.2.2, pp 5 (3.6/3.7) 194001K101 (KA*S) ANSWER 8.02 (1.00) c. REFERENCE Oconee Tech Specs bases pp. 3.5-12 2.e/3.8 002006KOO6 ...(KA'S) ANSWER 8.03 (1.00) i l c REFERENCE 10CFR20.5 (2.8/3.4) 194001K103 ...(KA'S) i ANSWER 8.04 (1.00) i I C dlF 0-1 REFERENCE Oconee Station Directive 3.1.1, pp 11 (3.7/4.1) I 194001K102 ...(KA'S) i I ANSWER 8.05 (1.00) i l l C. I l L____________________________________
t_CgNgillgNS _@NQ_61511@IlgGS PAGE 56 et__9901Ni@IS@IlyE_E59CEQQBES t c, AOSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. REFERENCE Oconee Tech Specs bases 3.5.2.6 2.6/3.8 OO2OO6K006 ...(KA*S) ANSWER 8.06 (1.00) a. REFERENCE Oconee Tech Specs 3.4.2 2.7 3.8 l 061006KOO6 ...(KA*S) ANSWER 8.07 (1.00) 1 d. REFERENCE Oconee Tech Spec 3.3.5 3.0/3.7 022OO5k005 ...(KA'5) l l ANSWER 8.08 (1.00) b. l REFERENCE Oconee Tech Specs 3.9 3.'1/3.6 073OO5KOO5 ...(KA'S) ANSWER 8.09 (1.00) REFERENCE Oconee Tech Specs 3.10 2.8/3.4 194001K103 ...(KA'S) ANSWER 8.10 (1.00) d.
1 8. ADMINISTRATIVE PROCEDURES. CONDITIONS, AND LIMITATIONS PAGE 57 4 ' ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. REFERENCE Oconee TS 3.3.a 2.9/4.0 OO6006KOO6 (KA'S) ANSWER 8.11 (1.50) a) False (+.5 ea) b) False c) True REFERENCE Oconee LER 86-013; Oconee OMP 1-9 (2.5/3.4) 194001G103 ...(KA'S) ANSWER 8.12 (1.00) a) Delayed (+.5 ea) b) Immediate REFERENCE Oconee OMP 1 - 3, Enci 5.1/5.2 ((2.5/3.4) 194001A103 ...(KA'S) ANSWER 8.13 (J.50) a) 4.5 rem (+.5 ea) O) 75 e em ok 25 R2h c)
- 7. 5 r em REFERENCE DPC HPM, pp 9/10; Emergency Plan, pp K-1 (2.8/3.4) 1G4001K103
...(KA*S) i I i J
@t 990lNigI6911yE_P69CEQU6E@t_CQNQlligN@t_gNQ_LiglI@llgyg PAGE 58 k?NSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 8.14 (1.50) I NO (+.5) This will result in deenergizing RIA-45, which is required to be in service to meet TS (3.8.10) requirements for the Reactor Building Purge System (+1.0) f Lt S$ (f~ 5) R TA ~ 41 tac 4*v D"r oC S f12 VRf 44 DG FMfG/2.rM /11 >?f7700 K REFERENCE / CON W O/4Y(.pt.c) Oconee LER 36-005; TS 3.8.10 (3.1/3.6) 073000G005 ...(KA'S) ANSWER 8.15 (1.00) 1 a) Procedure Discrepancies Process Record (+.5 ea) d l pgg yg j p ;g b) 14 days REFERENCE o /L Pucacu e t pas c.rs] ktM O %"'M ~ Oconee Station Di r ec t i ve 2.2.1, pp 4/5 reeno" (2.5/3.4) CNA l-4 194001A103 ...(KA'S) l ANSWER 8.16 (1.00) Category 1: General Public/non-occupational (+.33 ea) Category 2: Non-essential to operations but Rad workers, Category 3: personnel identified in tne Emergency Response l Organization I l REFERENCE Oconee RP/0/9/1000/10, pp 3 (3.1/4.4) 194001A116 (KA'S) ANSWER 8.17 (1.00) Column D Bring the unit to hot shutdown within 24 hours. i REFEPENCE Oconee Tech Specs Table 3.5,1-1 3.4/4.3 01:005e005 (KA'S) l
8 __egglNigl@gilyE_EggCEgySES _CgNQlllgNS _@NQ_6}[ll@llgNS PAGE 59 3 t ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. 4 ANSWER 8.18 (1.00) By matntaing the reactor coolant temperature and pressure in the acceptable region, any dissolved gases in the RCS are maintained in solution. REFERENCE Oconee Tech Specs 3.1.10 2.9/3.8 OO1006KOO6 (KA'5) ANSWER 8.19 (1.00) 1) Operational Responsible Supervisor (+.25 ea) 2) Work Supervisor 3) Human Red Tag 4) Individual doing the work l REFERENCE Oconnee Station Directive !.1.1. pp 14 j (3.7/4.1) 194001K102 ...(KA'S) ANSWER 8.20 (1.50) 1) Change of Classificatition (+.3 ea for any 5) 2) Termination of Event 3) Further degradation of level of plant safety { 4) Effectiveness of protective measures taken 5) Results of evaluations / assessments 6) Plant behavio/r that is not understood REFERENCE Oconee OMP 1-10, pp 2 (3.1/4.4) 194001A116 ...(KA'S)
m
- 8. __9901N1QIB@Ily@_CEQQEQQBES _QQNQlligNS _gNQ_LidlI@Ilgh'j PAGE 60 1 1
t hNSWERS -- OCONEE 1, 2b3 -87/07/13-CASTO, C. ANSWER 8.21 (1.50) 1) Supt of Technical Services (+.2 for name, +.1 for order) 2) Supt of Maintenance 3) Supt of Operations 4) Supt of Integrated Scheduling 5) Operations Duty Engineer REFERENCE Oconee Emergency Plan, pp B-1/2 (3.1/4/4) 194001A116 ...(KA'S) ANSWER 8.22 (1 90) 1. Reactor coolant pressure is 300 psi or greater 2. Reactor doolant temperature is 200 deg. F or greater 3. Nuclear fuel is in the core [0.5 ea.] REFERENCE Oconee TS 3.6 3.3/4.) 103OO5KOO5 (KA'F) ANSWER 8.23 (1.00) l 1. Thermal Pcwer Level l 2. Number of R r,P s operating 3. Reactor Power Imbalance REFERENCE Ocnnee Tech Specs 2.1 2.6/3.8 l OO2OO6KOO6 ...(KA*S) l i l I l L____.------_-------_ I
6:__@90ldlE18@llYE_EUE5Ehh51_CQUQillgySt_gy@_61[11I@IlgyS PAGE 61 .C., pNSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. o l ANSWER 8.24 (2.00) l i Cannot be moved Cannot be located Control rod is misaligned with its group a vr#r a g e by more than 9 inches Control rod does not naet the exercise requirements Control rod does not meet the rod trip insertion times Control rod does not meet the roc program verification any 5 of 6 i ) l REFERENCE OConee Tech Specs 3.5.2.2 3.7/4.1 f 001005K005 WA'S) { l \\ ANSWER 8.25 (2.0G) Unit 1 shall be placed in a hot shutdown condition within 12 hours. It these requirements are not met within an additional 48 nours, the reactcr shall be placed in the cold shutdown condition within 24 hours. [1.03 g r' t T The inoperable circuit / channel shall be restored to opersblitty and the cond i t l ar.3 of Tahle !.7-1 for normal operation shall be satisfiec for all functionel unitc Usfore the reactor is returned to criticality. [1.0] REFERENCE Oconee TS 3.7-1 3.1/;.8 J l l 062005V005 ...(KA'S) i L__-________-_______-_____-_____-__-_______
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- ,A U.
S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _QggNEE_11_2h3___________ REACTOR TYPE: _PWB-ggW1ZZ_________,,____ DATE ADMINISTERED: _@ZLgZ41}________________ j l EXAMINER: _gA$Tgz_g._______________ CANDIDATE: IN@l6UgIlgN@_IQ_g@NQ1991El Use separate paper for the answer %. 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 grade 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 CANDIDATE'S CATEGORY j __Y66UE_ _IQI@L ___@gg6E___ _y@6UE__ ______________g@lEGQ6Y_____________ j _29199__ _2E199 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ze r'- _E9E99'__ 29199 ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS _}9t99__ _2919Q I9* ________ 3. INSTRUMENTS AND CONTROLS _-29~99__ _29199 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL // f T~ 1-29r99:'_ Totals Final Grade All work done on this examination is my own. I have neither given nor received aid. 1 i Candidate's Signature i i i i A-j
I* o NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
- ,o Durin'g the administration of this examination the following rules apply:
1. Cheating on the examination means an automatic denial of your application and could result in more severe penal ties. 2. Restroom trips are to be limited and onl y one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating. 3. Use black ink or dark pencil gnly to facilitate legible reproductions. 4. Print your name in the blank provided oi the cover sheet of the examination. 5. Fill in the date on the cover sheet of the examination (if necessary). 6. Use only the paper provided for answers. 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet. 8. Consecutively rumber each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gne side of the paper, and write "Last Page" on the last answer sheet. 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
14 Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Par ti al credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.
This must be done after the examination has been completed. i )
1 ) h ' 10. When you complete your examination, you shall: a. Assemble your examination as f ollows: (1) Exam questions on top. (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are part of the answer. b. Turn in your copy of the examination and all pages used to answer the examination questions. c. Turn in all scrap paper and the '.al ance of the paper that you did not use for answering the questions. d. Leave the examination area, as defined by the examiner. If after i leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked. _l
1 I 11__EBINCIELES_gE_NUC6s96_EgWEB_E66NI_ GEE 8@llgN PAGE 2 t IHE6DggyN@01CS _HE9I_IB8NSEEB_ONg_E6U1g_E(OW z QUESTION 1.01 (1.00) Which one of the f oll owi ng correctly describes the behavior of RCS pressure if a Small Break LOCA which was not large enough to actuate the ECCS were to occur, without Feedwater available? a. Pressure initially decreases slowly, then rapidly drops when the OTSGs are boiled dry, b. Pressure decreases slowly until it levels off somewhere above ECCS actuation pressure. c. Pressure initially increases, then rapidly drops when the OTSGs are boiled dry, d. Pressure initially decreases, then rapidly increases when the OTSGs boil dry. e. Pressure initially decreases, then when OTSGs boil dry, continues to decrease, but at a much slower rate. QUESTION 1.02 (1.00) Which one of the f ollowing instrument failures would cause the behavior of the parameters shown on attached drawing DC-TA-NT-15? a. Delta Tc Failure "A" Side LOW b. Delta Tc Failure "A" Side HIGH c. Delta Tc Failure "B" Side LOW d. Delta Tc Failure "B" Side HIGH i (***** CATEGORY 01 CONTINUED ON NEXT PAGE
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} id__EBINCIELES_gE_NUC6E86_EQWEB_E(@NI_QEE8@llgN PAGE 3 t ISEBdggyN8 digs _dE8I_IS@NQEE8_@NQ_E(Ulg_E6gg t f QUESTION 1.03 (1.00) When performing a reactor startup to full power that commenced 5 hours after a trip from full power equilibrium conditions, a 5%/ min ramp was used. How would the resulting ) xenon transient vary if a 2%/ min ramp was used instead? c. The xenon dip would be smaller and occur sooner, b. The xenon dip would be smaller and occur later. c. The xenon dip would be larger in magnitude and occur later. l l d. The xenon dip would be larger in magnitude and occur sooner. QUESTION 1.04 (1.00) Which one of the following radionuclides is formed by activation of the makeup water? a. Argon b. Xenon c. Krypton d. Iodine QUESTION 1.05 (1.00) Refer to the attached handout DC-RT-SM-3 "NI Response for Subcritical and Critical Reactor", for each point labelled A-D match following descriptions with the response of the NI. 1. Stable SUR i I 2. Equilibrium value of subcritical mul ti pl i c ati on 3. Withdraw an increment of control rod 4. Withdraw an increment of control rod to make the reactor i critical. l l (***** CATEGORY 01 CONTINUED ON NEXT PAGE
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12__EBINCIELEg_QE_NQC6g@B_EQWEB_E6@NI_QEE88IlgN PAGE 4 t ISE@dQQyN@dlC@t_dE@l_IB@N@EEB_9NQ_E6Q1p_E(QW QUESTION 1.06 (1.00) Which one of the f ollowing dictates the length of time required to reach an equilibrium count rate in a sub-critical reactor following a given reactivity addition? q l a. The magnitude of the source strength b. The Subtritical Multiplication Factor (M) l c. The effective half life of the delayed neutron precursors d. The ratio of initial to final count rate I QUESTION 1.07 (1.00) Which one of the f oll owing is correct concerning differential control rod worth (DRW)? c. It is a measure of reactivity due to rod position. b. With a normal cosine flux shape, DRW reaches a maximum I value at a rod index of less than 28%. c. Rod Group Overlapping maintains a constant DRW. d. Its unit is delta K/K/% index. I QUESTION 1.08 (1.00) Which one of the f ollowing is an advantage associated with the use of soluble poison in the RCS? a. Soluble poison aids in reactor control by its effects on the moderator temperature coefficient, the void coefficient and the pressure coefficient. b. Soluble poison can be injected quickly enough to overcome all positive reactivity insertions without control rod movement. c. Soluble poison control provides for a smoother flux shape. d. Soluble poison at lower temperatures crystallizes. (***** CATEGORY 01 CONTINUED ON NEXT PAGE
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12__EB1gCleLES_QE_ NUCLE 86_EQWEB_EL6NI_QEE8@llg@t PAGE 5 ISE6DgDyN@DICS _dE@l_IB@NSEEB_@NQ_ELUlD_ELQW t QUESTION 1.09 (1.00) Which one of the following represents the maximum linear power density which would be expected in the core during full power operations? a. Local Power Density multiplied by Nuclear Peaking Factor, b. Radial Peaking Factor multiplied by the Local Peaking
- Factor, c.
Average Kw/ft for the core multiplied by the Nuclear Peaking Factor, d. Nuclear Peaking Factor mul tipli ed by the Maximum Local Power Density. QUESTION 1.10 (1.00) Refer to Figure 16 attached, if elevation "A" is 16 feet, PG2 - PG1 is 5 psia and the friction head losses equal O.5 feet what is the Total Developed Head for the pump? l a. 18.67 feet b. 21.5 feet c. 27.0 feet d. 28.0 feet QUESTION 1.11 (1.00) l Which one of the following conditions would hinder natural ) circulation flow in the RCS? a. An increase in thermal driving head. b. An increase in the velocity head of the fluid, c. An increase in the mass of feedwater into the OTSG. d. A decrease in Two-phase (nucleate boiling) within the Core. (***** CATEGORY 01 CONTINUED ON NEXT PAGE
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u__________________________.------------------__________________--- ______________a
I it__EBINCIE6ES_9E_NUC6E@6_EgyEB_E6@N1_gEg8@IlgN PAGE 6 z IdEBdggyN@dlCS _UE@I_I6@NSEEB_6NQ_E6 gip _E6gy z i QUESTION 1.12 (1.00) Which one of the f ollowing is a characteristic of the conduction mode of heat transfer? a. there is a resultant increase in the kinetic energy of the material. b. a reduction in the thickness of the material through ) which the heat is transferred will result in a reduction ] in the heat flux. c. materials which have widely spaced molecules act to improve the heat transfer rate. d. heat transfer through a fluid film is an ideal example of the conductive mode of heat transfer. QUESTION 1.13 (1.00) Assuming that the plant is operating at full power, what will be the difference in the following parameters, if OTSG tube fouling has occurred to a significant degree? a) OTSG Level b) Superheat Temperature QUESTION 1.14 (1.50) i With the Unit operating at 100*/. power with all control systems in automatic, a Turoine Bypass Valve fails full open. Indicate how the following parameters will change relative to their initial values when plant conditions stabilize: (INCREASE, DECREASE, REMAIN THE SAME) a) Tavg b) MWe c) Reactor power (***** CATEGORY 01 CONTINUED ON NEXT PAGE
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I 11__EBINCIE(ES_QE_NgC(E@B_EgyEB_E(@NI_QEEg@IlgN PAGE 7 i t IHEBdggyN@dlCS _HE91_I69NSEE6_@Np_E(gip _E(QW l t I QUESTION 1.15 (1.00) An ECP is calculated for a startup following a reactor trip from 50% power, with equilibrium xenon in the core at MOL. Indicate if the ACTUAL critical rod position will be HIGHER, l LOWER or the SAME compared to the calculated position for l each of the following situations. Use attached curves as appropriate and treat each situation individually. a) Xenon reactivity curve for trip + rom 100% power is used to calculate conditions for a startup 20 hours after the trip. (Computer is not in service to give this info) b) The differential boron worth at an EOL condition is used. Assume no change in boron concentration is desired prior to achieving criticality. QUESTION 1.16 (1.50) Indicate whether the following will INCREASE, DECREASE or REMAIN THE SAME: a) Available NPSH for a MFP as volumetric flow rate increases, b) Minimum required RCP NPSH as volumetric flow rate increases. l c) Assilable NPSH to condensate (hotwell) pumps as condenser subcooling increases. l DUESTION 1.17 (1.00) i For EACH of the f ollowing conditions state whether the j actual Shutdown Margin would be greater than/ less than/ the same as calculated Shutdown Margin. Assume BOC, Mode 1 full power. 1. Actual poison burnup exceeds that value used in the calculated SDM. l l 2. The actual worth of the maxi mum stuck rod assumed in the I calculation is lower than predicted. (***** CATEGORY 01 CONTINUED ON NEXT PAGE
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l ? Ic__EBigClebES_gE_ NUCLE 98_EgWEB_E6@NI,QEEB9IIgN1 PAGE 8 IbEgdggvggdlCS _UE8I_IB8NSEEB_8Ng_[(Ufg_E(gW z QUESTION 1.18 (1.00) What indication tells the operator when all nitrogen has been vented from the Pressurizer, when forming a steam bubble in accordance with OP/0/A/1103/05? QUESTION 1.19 (2.00) p) Unit 2 has just restarted following a refueling outage while ha p Unit 3 is near EOC. Answer the following regarding the C N & 0 & M differences in plant response between the units (Ex pl ai n gg7 your answer): yh gt 0 [ i 1 a) At a steady power level of 10EE(-8) amps during a 4<6 ~ 6 startup, equal reactivity additiens are made. Which j'y Unit will have a higher startup rate? 4 gr, gb b) At 50% power, with ICS in MANUAL, a control rod drops. g k <<f/ Assuming NO RUNBACK and NO OPERATOR ACTION, which Unit wi11 have a 1ower steady state Tavg? a yf"[M M'C" WW [M QUESTION 1.20 (1.00) gp If a OTSG overfills, describe two adverse effects on the secondary side, other than water hammer-related effects. j}G QUESTION 1.21 (1.00) l What are the three s ignificant sources of Hydrogen production in the primar. following an accident (e.g., LOCA)? List them in order f~ significance. GUESTION 1.22 (1.00) Recently, Inconel Axial Power Shaping Rods (APSRs) have been installed on all 3 units. What is the main physical change in these rods and how has this change improved the APSR's capability to control axial flux imbalances? l l (***** CATEGORY 01 CONTINUED ON NEXT PAGE
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I Iz__EBINCIE6E@_gE_ NUCLE 65_EgWEB_E(@NI_ GEE 6@IlQN PAGE 9 t ISE60gDyN@DICS _dE@l_IB@NSEEB,@ND_E6UID_E699 t 5 QUESTION 1.23 (2.00) How does each of the following parameter changes affect the DNBR (INCREASE, DECREASE or NO EFFECT)? Briefly expl ain your answer and DO NOT consider the transient effects or any control system or operator actions. l l a) Pressurizer temperature increases 5 degrees. l b) Mass flow rate in the core increases 10%. QUESTION 1.24 (1.00) State the operational concerns of an uncontrolled cooldown on the reactor vessel. Include in your answer the types of stresses induced on the inner and outer vessel walls. i l 4 I l J QUESTION 1.25 (1.00) The pressurizer PORV is leaking by during operation at 85% power. Assuming a Quench Tank pressure of 20 psia and saturation conditions in the pressurizer corresponding to 2240 psia, what is the quality of steam on the downstream side of the PORV? Show all calculations. QUESTION 1.26 (1.00) The reactor is producing 100% rated thermal power at a core delta T of 60 degrees and a mass flow rate of 100% when a station blackout occurs. Natural circulation is established and core delta T goes to 40 degrees. If decay heat i s 2%, what is the core mass flow rate (in %)? l a. 1.3 l b. 2.0 c. 3.0 d. 2 l 1 i QUESTION 1.27 (1.00) i Sketch the temperature profile along the length of a counter i flow heat exchanger for both the cooling and c ool ed medium, i (***** END OF CATEGORY 01
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l l L________.____________________________.____
i 2c__P69BI_DEglGN_lNC(UDi@G_S@Egly_9ND_EMEBGENCy_SygIEMS PAGE 10 GUESTION 2.01 (1.00) Which one of the following is NOT a design difference between the safety / regulating control rods and the APSRs? a. On APSR rives a small button on the lower portion of the j segment arm prevents the lead screw from being disengaged j when power is lost. b. APSR couplings have larger diameters and shorter keys to prevent coupling an APSR drive to a safety or regulatit.J rod or vise-versa. c. APSR drives have ball valves and bypass parts. d. APSR do not have buffer springs in the buffer assembly. QUESTION 2.02 (1.00) Which one of the following should enable MS-93, TDEFWP steam supply valve, to open if it fails to open on an automatic signal? 1 a. Verify proper operation of the DC Dil Pump. b. Line up backup service air to the valve operator. c. Isolate instrument air to its reducer and bleed the air off the reducer. d. Take the control switch to "Off" to remove power from the solenoid valve. I l (***** CATEGORY O2 CONTINUED ON NEXT PAGE
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1
32__PL@NI_DEgl@N_INCLUQlyg_g8 Eely _@@D_ EMERGENCY _@y@lEM@ PAGE 11 c QUESTION 2.03 (2.00) Match the components listed in Column A with the correct location where they penetrate the RCS. Answers may be used more than once. Column A Column B a. Unit 1 PZR Spray 1. Al RCP Suction b. Unit 3 Normal HPI Line 2. B1 RCP Suction l c. Unit 2 Letdown Line 3. Al RCP Discharge d. Unit 2 Decay Heat Removal Line 4. B1 RCP Di scharge 5. B2 RCP Discharge 6. A Hot Leg 7. B Hot Leg l l GUESTION 2.04 (1.00) In accordance with AP/1/A/1700 " Loss of Instrument Air" the operator will have to monitor the Nitrogen backup supply to various valves. Which one of the following valves does NOT l have a backup Nitrogen supply? 1 a. Main Steam to Aux. Steam Regulator b. Emerg. feedwater control valves 1FDW-315 and 316 c. Main Steam to TDEFWP d. RCP Seal Injection Flow Control Valve 1HP-31 QUESTION 2.05 (1.00) i The attached drawing, Figure 7.28, shows the LPI system aligned for which one of the f oll owing modes of operation? a. Switchover mode on Unit 1 b. Switchover mode on Unit 2 i c. Normal decay heat removal on Unit 2 l d. Normal decay heat removal on Unit 3 l 1 (***** CATEGORY O2 CONTINUED ON NEXT PAGE
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i 2c__PL@@l_DESl@N_lNCLUDIN@_S@ Eely,@ND_EME8@ENCY_@YSIEM@ PAGE 12 QUESTION 2.06 (1.00) Which one of the following is correct concerning Reactor Protection Systen operation with one " Dummy Bistable" installed? a. The RPS is in a 2-out-of-3 Trip Logic, b. The affected channel is in the tripped state. i c. The Dummy Bistable contacts are normally opened, and close upon a trip signal. d. If installed as a replacement for High Flux bistable, placing the channel in shutdown bypass would defeat the purpose of the dummy bi stabl e. GUESTION 2.07 (1.50) c) What is the purpose of valves HP-409 and HP-410(HPI Cross Connects) and what situation are they intended to help mitigate? (1.0) b) Which Unit (s) have an indication on the computer display of HP-409/410 not being fully closed? (0,5) 4 [ T'v (;p % p p s-M J GUESTION 2.08 (1.00). ggg gp a) Aside from CCW Intake water, hod else can water be supplied to the CCW Intake area if it is required for operation of the SSF Auxiliary Service Water System. b) How is air injection into the S/Gs minimized when using 7'g g 5 3 './' the SSF ASW System? i GUESTION 2.09 (1.00) What is the basi s for the f oll owi ng admini strati ve control s that were recentl y instituted on the Main Steam to Auxiliary Steam interconnections? i I Only one unit's Main Steam is used to supply Auxiliary l Steam via only the 2" reducer MS-129. The other two l units' Main to Aux Steam reducers, MS-126(6") and MS-129 are totally isolated. l (***** CATEGORY O2 CONTINUED ON NEXT PAGE
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2i__PL9NI_DESI@N_lNCLUDING_g@FEIy_@ND_ EMERGENCY _gy@IEMg PAGE 13 QUESTION 2.10 (1.00) Fol l owi n g a station blackout AP/1/A/1700/11 informs the I operator that the Turbine Bypass Valves may cycle on and off d'""#' at the 7 inch Hg condenser vacuum low limit. Explain why this phenomenon may occur. l i l l GUESTION 2.11 (1.00) Describe the two (2) intersystem ties between the HPI system and the SSF Makeup system. Include in your answer the source and discharge flow path. l l l GUESTION 2.12 (1.50) Explain the basis for each of the f ollowing system limitations: 1. Maintaining Letdown Storage Tank l evel above approx. 18 inches. 2. Maintaining Letdown Storage Tank pressure vs. level curve within the operating range. GUESTIFJ 2.13 (1.00) State the purpose of the CONTACT MONITOR AUXILIARY POWER SUPPLY for the RPS system. Include in your discussion the source of power to the Aux. Power Supply. GUESTION 2.14 (2.00) List the two auto start si gnal s (including setpoints) for the Emergency Feedwater System and the two setpoints/ conditions at which level will be automatically controlled. (***** CATEGORY O2 CONTINUED ON NEXT PAGE
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i 2g__PL@NI_DE@lGN_l@C6UDING_@@ Eely _@ND_EDE6GENCy_@y@ led @ PAGE 14 I l i GUESTION 2.15 (2.00) List the six different flowpaths of electrical power to the Oconee Nuclear Station, (including the appropriate transformer and buses supplied) in their order of preference. Assume that the applicable main generator is producing > 200 MWe when it is considered as a source of power. j GUESTION 2.16 (1.00) List the 4 loads cooled by the Component Cooling (CC) System. I J QUESTION 2.17 (1.00) Provide 4 distinct problems that could result due to an inadvertent ES actuation (consider all channels) if operator action is not taken promptly to correct the situation. GUESTION 2.18 (1.00) List 4 of the 5 sampling points in containment associated with the Hydrogen Analyzer, i I GUESTION 2.19 (1.50) For the SPDS list three (3) of four (4) instruments and/or indications which are used to generate the Subcriticality critical safety function logic. QUESTION 2.20 (1.50) a) Explain why AUTOMATIC contr,ol of steam header pressure on tne Turbine Hand / Auto station, with the ICS in TRACK, is not a preferred mode of operation at low power levels? During \\ is done to ensu\\{e the Unit will t b) a startup, what not be in TRACK when t e Turbine Control is place An " g-. - n automati ? \\ q (***** CATEGORY O2 CONTINUED ON NEXT PAGE
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2i__P6@NT_DE@l@N_idCLUDIN@_@@E@Iy_6Mp_EMEB@ENCy_@y@IEM@ PAGE 15 I QUESTION 2.21 (1.50) 1 If a loss of Instrument Air to the air-operated valves in l the Makeup portion of the HPI System occurred, what 3 I methods / alternate flow paths could be utilized to maintain pressurizer level, assuming that "A" HPI pump is in service at the time of the failure? l QUESTION 2.22 (1.50) Provide the purpoce of the following Precautions associated with the Emergency Feedwater System: a) FDW-315 and FDW-316 (EFW to OTSGs> must be placed in Manual prior to resetting the Main Feedwater Pumps. b) Auxiliary Steam valves should not be cut into the Upper Surge Tank if the EFWPs are to be operated. c) MS-97, EFWPT Exhaust to Condenser, shall be shut with its breaker locked open and MS-96, EFWPT Exhaust to Atmosphere shall be open with its breaker locked open. l QUESTION 2.23 (2.00) Describe BOTH the purpose AND operation of the Reactor l Building Cooling Unit dropout plates AND blowout plates. l l l 1 l (***** END OF CATEGORY O2
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t 3g__ INSTRUMENT 3 AND CONTROLS PAGE 16 1 ( QUESTION 3.01 (1.00) Which one of the following will NOT cause the ICS to enter the TRACKING mode of operation? a. Plating the Diamond Control Station in Manual, b. Placing BOTH Main Feedwater Valves in manual control. c. Providing the turbine with 45*/. more power than is being produced by the generator. d. Feedwater Cross Limits in effect. QUESTION 3.02 (1.00) Which one of the following conditions correctly describes f the requirements for a 23OKV Switchyard Isolation to occur? a. Undervoltage on 2 of 3 phases on EITHER the Red or Yellow Bus on BOTH Channels of UV protection. b. Undervoltage on 4 of the 6 phases monitored between the Red and Yellow Buses on EITHER Channel of UV protection. c. Undervoltage on 4 of the 6 phases monitored between the Red and Yellow Buses on BOTH Channels of UV Protection. d. Undervoltage on 2 of 3 phases on BOTH the Red and Yellow Buseu on BOTH Channels, e. Undervoltage on 2 of 3 phases on BOTH the Red and Yel l ow Buses on EITHER Channel. QUESTION 3.03 (1.00) Which one of the following correctly describes an interlock associated with Component Cooling Discharge Valve CC-8? a. If CC-8 is closed, NEITHER CC Pump may be started, b. CC-8 closes on actuation of ES-1 or ES-2. c. CC-7 (MOV CC Discharge Valve) closes if CC-8 closes. d. If CC-8 closes any operating CC pumps will continue to run. (***** CATEGORY 03 CONTINUED ON NEXT PAGE
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31__lygIBgMEgIg_eUD,CONIB065 PAGE 17 a QUESTION 3.04 (1.00) I Which one of the f ollowing correctly describes how control l of ES components is obtained on a Type "A" RZ Module assuming an emergency signal is still present? r uW s a. Control can NOT be obtained until the ES signal clears. y / j gy o b. Depress the Manual pushbutton for the component, which c' i will restore the component to its state prior to the ES. I and then control may obtained at the control board switch only. c. Depress the Manual pushbutton for the component, and then control may be obtained at the control board switch only. j i d. Depress the Manual pushoutton for the component, and l then control may be obtained locall y only. GUESTION 3.05 (1.00) Sh {nf)~s Which one of the following correctly describes the local
- jd switchgear controls for the SSF?
y a. Breaker interlocks prevent closure of a breaker by the -N manual close pushbutton (l oc at ed on lower right hand portion of the breaker) when the breaker is in the open o osi t i on. b. Local closing of the breakers using the corcrol switches (different from the manual PBs) can be accomplished as long as the breaker is not racked completely out. c. Local tripping of the breakers using the control switches (different from the manual PBs) is permitted only in the Test position. l d. With a loss of control power the ability to trip a breaker with th3s anual open pushbutton is lost. D"'4'pp N s 'T-N b' ({J N \\ (***** CATEGORY 03 CONTINUED ON NEXT PAGE
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32__1BSIBUMENIW_AND_CONIB96@ PAGE 18 GUESTION 3.06 (1.00) A valid EFW start signal exists. Unless otherwi se specified all appropriate controls are in automatic. Which one of the following conditions would prohibit the injection of EFW into the OTSG by the Turbine driven pump switchkS-01 has failed low, a. The low oil pressure b. EFWPT control in Pul;-To-Lock and a failure of KVID Brk 6 (solenoid power sup; i y) which trips open. c. While selected to Primary level control, a loss of power to the Primary channel occurs. d. The valve position limit switches for MS-93 Steam Supply Valve fail to recognize the valve opening. QUESTION 3.07 (1.00) Which one of the following conditions would result in an Out Inhibit being generated in the Rod Contol Logic? a. Safety Rod Groups at the out limit, b. Asymmetric fault with power level at 50%. c. A startup rate of 2.0 DPM in the Source range, i d. High neutron error signal (2.5%). \\ GUESTION 3.08 (1.00) Which one of the f oll owing is a symptom of a bellows failure on Reactor vessel level instrument LT-5? j a. A lower than actual reading, b. A higher than actual reading. s c. A f al se zero level indication, d. Oscillations between high and low levels. (***** CATEGORY 03 CONTINUED ON NEXT PAGE
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r 3 __INSIBUMENIg_AND_CONIBOLE PAGE 19 1 QUESTION 3.09 (1.50) Indicate the INITIAL RESPONSE of the f ollowing parameters and components following the failure of "A" S/G Outlet Pressure HIGH: (Assume plant initially at 80% and no trip occurs) i i a) Feedwater flow to "B" S/G i b) "A" Bypass Valves l c) Generated MWe I QUESTION 3.10 (1.50) l Indicate whether the f ollowing statements apply to the i features of an EMERGENCY LOCKOUT, NORMAL LOCKOUT or BOTH of a Keowee Hydro Generator: a) Trips ACB-3 and Blocks Reclosing b) Trips ACB-1 and Blocks Reclosing c) Blocks Startup of Unit on any signal i QUESTION 3.11 (1.50) Indicate f or the following whether they apply to an RPS i Bistable or an RPS Contact Buffer? I a) Two analog inputs, with one input.being a setpoint. b) Must be manually reset via a toggle switch, c) Energizes a relay if input senses a trip condition. l QUESTION 3.12 (1.00) l Refer to figure OC-EL-EPD-1 " Basic Logic Diagram" attached, I for each of the given inputs A,C,D,F determine the resultant output of B,E and G. (state answer i n t er ms of O or 1) l l l l l l \\ l l l l l (***** CATEGORY 03 CONTINUED ON NEXT PAGE
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i
i 3 __IN@IBUMENIS_@@D_CQ@l89L@ PAGE 20 ) t QUESTION 3.13 (1.50) Indicate where the following components could be operated from(if at all), given the loss of power indicated: a) Pressurizer Spray Valve on loss of Auto Power. b) Pressurizer Spray Block Valve on loss of Auto Power. c) Pressurizer Electric Relief (RC-66) on loss of Auto Power. l QUESTION 3.14 (1.00) OP/0/A/1105/09 " Control Rod Drive System" cautions the operator NOT to operate a stuck or jammed control rod which is partially or fully withdrawn in JOG speed. Explain the basis for this caution. QUESTION 3.15 (1.00) i Should a control rod "in" movement signal be generated and the rods run in the out direction, what Diamond Control Panel Lamp would illuminate AND what auto action would result from this condition? GUESTION 3.16 (1.00) Amplifier ^S For the Differential a Power Range Instrument channel % state two (2) inputs AND three (3) outputs. GUESTION 3.17 (1.00) a. How does the operator place into service the Low Range Cooldown Pressure instrument? b. If the Low Range Cooldown Pressure instrument had been inadvertently valved out of service, would the setpoint for the PORV switch to the low value? (yes/no) l (***** CATEGORY 03 CCNTINUED ON NEXT PAGE *****)
3c__INSIBQMENIS_@ND_CQNIBQLS PAGE 21 = l l l l l QUESTION 3.18 (1.00) List the two interlocks that must be satisfied in order to ( start the SSF ASW pump if an ES-1 or 2 (loadshed) signal is j present. ] i 1 QUESTION 3.19 (1.50) i List the 6 modifications made to the Reactor Pr ot ec ti on System (either automatically or manually) when the Shutdown Bypass Key Switch is placed in " Bypass". QUESTION 3.20 (1.00) Place the following RPS components in the order they exist in a typical RPS Channel and indicate the number of each of 1 these components per channel: 1) K Contact Logic Rel ay is) 2) Interlocking Test Contact (s) ] 3) Manual Trip Switch Contact (s) 4) Coincidence Logic Contact (s) j QUESTION 3.21 (1.00) List two (2) automatic actions which occur affecting the SSF in the event a Unit 2 Channel A load shed signal is received. QUESTION 3.22 (1.50) List three (3) components which are normally supplied from KI and upon a loss of KI will be powered from the KU-B bus. (***** CATEGORY 03 CONTINUED ON NEXT PAGE
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31__INSIBUMENIS_8ND_CONIBOLS PAGE 22 1 ) OUESTION 3.23 (1.00) Indicate with which unit (s) the following High Pressure Injection System interlocks are associated: a) On low seal injection flow, the Standby HPI pump will l start. b) If seal injection is lost and the Component Cooling is lost, the associated Seal Return Valve will close. QUESTION 3.24 (1.00) Why are the Main Steam Line Area Radiation Monitor detectors Geiger-Mueller vice Ion Chamber detectors like all the other Area Radiation honitors? l GUESTION 3.25 (1.00) l According to AP/1/A/1700/13 Section B " Dam Failure Without Loss of CCW Intake Canal" an operator is to be dispatched to I place the TDEFDWP Cooling Bypass switch to the " Bypass" position. Explain what this action would accomplish with regard to the TDEFDWP cooling system. l l GUESTION 3.26 (2.00) The unit is operating at full power conditions, all controls and instrumentation are in automatic, Channel 9 Power Range NI is supplying the ICS, a. A loss of 120 VAC to Channel E of the RPS occurs (KI remains energized). For EACH of the resultant plant conditions listed below (1-4) provide the appropriate system response. RESPONSE OF: 1. Neutron flux fails to zero Control Rods 2. Loss of RCS flow signals ICS and Feedwater 3. Items 1 & 2 above RPS 4. RCS narrow-range pressure fails low PORV control Spray control b. RCS pressure starts to increase, what condition would cause this continual increase in RCS pressure? l (***** END OF CATEGORY 03
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l j I 4r__B60gEQUBEg_ _N06d@6t_@@NO6d@6t_EdE6ggggy_Adp PAGE 23 RADIOLOGICAL CONTROL i l + i i l -[ GUESTION 4h01 (1.00) 'kQ}] one o\\f / } the following s"NOT a Control Rod Drive System l Which limitation and/or precautio'n according to OP/0/1/llO5/09? I l \\
- a. There ar e n o'\\l i mi t s d ur i n g testing of the CRD breakers for rate, however the frequency at which they are cycled is limited.
\\ \\ 4 \\ b. The CRDs must be vented if pressurizer level decreases s below a predetermined limit. \\ c. While on the Aux Power Supply with a Safety Group while l all other Safety Gr'oups are at the out limit the operator I is prohibited from selecting Automatic. \\ N l d. Operating limits have'been established to assure that l control rod drop is prohibited under conditions which l would defeat the hydraulic snubbing action of the l mechanism. \\q 1 1 GUESTION 4.02 (1.00) Which one of the following methods contained within the i Emergency Procedures is the best method for removal of RCS i voids that are due to the presence of Non-condensible gases? a. Repressurization of the RCS. b. RCP Restart. c. RCP Bumping. d. Vessel or Hot Leg venting. 1 (***** CATEGORY 04 CONTINUED ON NEXT PAGE
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4 __E6gCgpugES_ _NQBd@6t_@@NgBd@61_EDEBGENCY_@NQ PAGE 24 2 6091969GIC06_C9 BIB 96 4 l GUESTION 4.03 (1.00) Which one of the following correctly describes the required actions if the reactor achieves criticality 1.5% (deltak/k) below the estimated critical p osi t i on? l a. Fully insert all regulating rods, but the safety rods may stay fully withdrawn. b. Fully insert all regulating rods, fully insert all j safety rods to Group 1 at 50% withdrawn. l l l l c. Fully insert all regulating and safety rods. 1 d. No rod insertion is required, but the ECP shall be recalculated. l e. No action required and startup may continue. GUESTION 4.04 (1.00) Which one of the following will cause the greatest biological damage to man 7 a. O.1 Rad of fast neutron b. 1.0 Rem of gamma c. 10 Rem of beta d. O.05 Rad of alpha l 1 l l l (***** CATEGORY 04 CONTINUED ON NEXT PAGE
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1 ) 42__PBQCEQUBES_:_NQBDB6t_@BBQBd@bt_EDEBgENCY_@@Q PAGE 25 669196gGIC66_C9 NIB 96 QUESTION 4.05 (1.00) Upon a loss of 1KI AP/1/A/1700/23 directs an operator to the Aux Shutdown Panel to perform various actions. Which one of the following is an action the operator can perform from this panel? a. Re-energize TBVs b. Bypass 1KI Inverter c. Control RCS pressure with Pressurizer Heater Banks 1, 3 and 4 d. If conditions warrant trip 1 RCP in 'B' loop 1 GUESTION 4.06 (1.00) Fill in the blanks with the appropriate limitations and precautions for cperation: a) Any E.S. Valve that has been manually operated must be _________ _________ to assure operability. b) Do not operate any EFW Pump at > ______ gpm and maintain UST level > ft whenever these pumps are required to be operable. QUESTION 4.07 (1.50) Indicate what the exposure limits are for the f ollowing conditions of exposure: (assume NRC-FORM 4 is current) a) Oconee Maximum yearly permissable Whole Body b) Maximum Planned Emergency Exposure to the Whole Body to Save Lives c) NRC Quarterly Skin Exposure Limit (***** CATEGORY 04 CONTINUED ON NEXT PAGE
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4t__PgggggyBgg_ _NQ8086t_@gNgBd@6t_EDEBggdCY_@Nh PAGE 26 689196991G86_C9 NIB 96 = QUESTION 4.08 (1.00) List the three breakers in the correct order in which they must be operated in order to allow the Keowee Hydro Generator to produce voltage output, once the wicket gate position is established. QUESTION 4.09 (1.00) Place the following Emergency Procedures in the correct order of priorite in which they are referred to when performing EP/1/A/1800/01, and which are continually monitored until plant conditions stabilize: (Assume that a S/G Tube Rupture was NOT the initial entry condition) 1) Excessive Heat Transfer 2) Loss of Heat Transfer 3) Loss of Subcooling 4) Steam Generator Tube Leak QUESTION 4.10 (1.00) Referring tc the attached excerpt from OP/1/A/1102/1,.1 for performing a Unit Startup, what is the significance of the " Bullets" preceding the substeps following step 2.1? GUESTION 4.11 (1.00) Answer the following questions regarding precautions and limitations that must be observed when operating at power with less than 4 Reactor Coolant Pumps: a) When calibrating the NI's, is the Thermal Power-Secondary Heat Balance or the Thermal Power Best Estimate utilized? b) What is the purpose of reducing the ICS High Flux Limiter 4*/. lower than the High Flux RPS trip setpoint? l l (***** CATEGORY 04 CONTINUED ON NEXT PAGE
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'4i__PBOgEQUBgS_:_ygBd@61_8BNOBd861_EUEBGEggY_9ND PAGE 27 6891969 GIG 06_99NI696 QUESTION 4.12 (2.00) AP/1/A/1700/19 " Loss of 4160v Power and the BWST" has the operator realign system components as a result of the failures. For EACH of the following state the resultant source of electrical / water supply to the component. 1. HPI pump suction. 2. HPI pump electrical power. j 3. Standby bus #1. l l 4. HPI pump motor cooling. l l l l l QUESTION 4.13 (1.00) EP/1/A/1800/01 Section 506 " Unanticipated Nuclear Power j Production", has the operator " Verify open 1HP-5 (Letdown Isolation)" prior to initiation of Emergency Boration. Explain the need for this action step. j l QUESTION 4.14 (1.50) Aside from the Main FDW Control Valve (FDW-32), list the other six valves which must be closed in order isolate the A OTSG if it is ruptured. 1 1 GUESTION 4.15 (2.00) I Besides tripping the reactor, starting the Keowee Hydro i Unitt and announcing to/ notifying the proper personnel, what ] are the remaining immediate actions if the control room must j be evacuated, as stated in AP/1/A/1700/8? Assume that time exists to complete actions prior to evacuation, and that the l SSF will NOT be required to be put in operation. Include in l your answer where operators are dispatched to and what l I materials are required to be taken out of the control room. l i l 1 I u (***** CATEGORY 04 CONTINUED ON NEXT PAGE
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i l l I
'Sf__PBggEQUBE@_;_NOBU@(1_@@@O60@(z_Ed@@@gNgy_@NQ PAGE 28 6891969G199L_gg@TBgL QUESTION 4.16 (1.00) What are the two criteria that must be met in order to utilize OP/1/A/1102/02, " Reactor Trip Recovery". QUESTION 4.27 (1.00) AP/1/A/1700/07 " Loss of Low Pressure 19)ection System" Section A " Failure of One Train of the LPI System During ECCS Operation", cautions the operator that "if only one LPI cooler is operable, then approx. 6000 gpm LPI and LPSW flow must be established through the operable cooler immediately after swapping LPI Pump suction from the BWST to the R.B. Emergency Sump." State two bases for this caution. GUESTION 4.18 (2.00) In accordance with AP/1/A/1700/18 " Abnormal Release of Radi oac ti vi t y", List all the automatic actions associated with each of the following Radiation Monitors: 1. 1RIAs 37 WG Disposal Normal /High 2. RIAs 33 Liquid Waste Normal /High 3. 1RIA 49 RB Gas 4. 1RIA 45 Unit Vent Gas Normal QUESTION 4.19 (2.00) If the HPI System has actuated due to a low pressure l condition, what are the two criteria, of which either one j must be met, that must be considered to secure the HPI System? GUESTION 4.20 (1.00) If the RCPs can be effective in cooling the core, even with two-phase flow, why must the RCPs be secured when O subcooling is noted? (***** CATEGORY 04 CONTINUED ON NEXT PAGE
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r '4*t__P60CgpU65@_,7_Ng6096t_9@yg60961_gdg69gNCy_gNQ PAGE 29 80D39L9GICOL_C9dIB96 I GUESTION 4.21 (1.00) Recently, during testing of the Emergency CCW siphon flow, problems developed resulting in loss of siphon. Among the corrective actions were some procedural changes. Provide the basis for the following procedural steps in AP-11, " Loss l of Power": a) Steam is i solated to the Condenser Steam Air Ejector first stages. b) TDEFW Pump suction is aligned to the hotwell from the UST as level in the UST decreases. GUESTION 4.22 (1.50) In EP/1/A/1800/1, section 502, " Loss of Heat Transfer", if i level has been restored to the OTSGs, and subcooling margin is O degrees, a phenomenons known as " saturated I repressurization" may occur. Explain why this may occur and what must be done to terminate the pressure rise. (Assume HPI is available) QUESTION 4.23 (1.50) EP/1/A/1800/1, Section 504, "S/G Tube Leak", has the operator i ni ti al l y c~ool down to 532 degrees while maintaining pressurizer level greater than 80 inches. What is the basis f or each of these setpoints? GUESTION 4.24 (1.00) Refer to attached Enclosures 7.3A and 7.5. Assume no RCPs are operating and the RCS is in a natural circulation cooling mode. The Steam Generator operating j range indications have failed and the extended startup range is being used to control level. If a level equivalent to 85% OR is desired and the following conditions exist, DETERMINE the indicated startup range level to maintain. Conditions are SG pressure = 900 psig and RB temperature = 250 deg. F. (***** END OF CATEGORY 04
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(************* END OF EXAMINATION
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control Copy ime OP/1/A/1102/01 ENCLOSURE 4.1 UNIT STARTUP FROM COLD SHITTDOWN TO RCS TEMPERATURE AND PRESSURE __ 0F 250*F AND 350 PSIG Verify Date Date Init./ Time Init./ Time 1.0 Initial Conditions 1.1 Procedure Limits and Precautions have been reviewed. 2.0 Procedure NOTE:.1A (Flowchart) should be used as a guide (2.0) by the SRO/R0 to aid in maintaining the big picture. ../ 2.1 Complete Enclosure 4.4 (Pre-heatup Checklist). If unit startup is following a Refueling Shutdown, complete Enclosure 4.6 (Pre-heatup Checklist Following a Refueling Shutdown). Plot an RCS Baron versus RCS Temperature for 1%AK/K S/D margin curve per PT/1/A/1103/15 (Reactivity Balance l l Calculation). Verify that RCS boron will be adequate for a 1%AK/K shutdown margin ~when rod group 1 is withdrawn to 50%wd at approximately 250'F. Unit Supervisor 1 Form a steam bubble in the pressurizer per OP/0/A/1103/05 (Pressurizer Operation). h$ Unit Supervisor ^ NOTE: Maintain pressurizer level at - 100 inches until RCS 0: concentration < 7 ppb.
1 / l EP/1/A/1800/01 Rev. 7 Unit 1 ENCLOSURE 7.3A Page 1 of.2 l l 1 SG EXTENDED STARTUP LEVEL TEMPERATURE COMPENSATION ~ 400 400
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I I 0 I I I 0 200 400 600 800 1000 SG PRESSURE (psig) DIRECTIONS: TO CONVERT O.R. LEVEL TO XSUR 1 1 1) READ SG OUTLET PRESSURE FOR DESIRED SG FROM CONTROL ROOM INDICATION 2) INTERSECT THIS PRESSURE LINE WITH DESIRED O.R. LEVEL 3) READ REQUIRED XSUR INDICATION AT INTERSECTION CAUTION Refer to Enci 7.5 to correct the XSUR level for high RB temperature as necessary. DLG/9-10-85 1 l l l C __
EP/1/A/1800/01 Page 1 of 1 ENCLOSURE 7.5 LEVEL CORRECTIONS FOR HIGH REACTOR BUILDING TEMPERATURE Corrections to Indicated Level Reactor Steam Generator Core ~ Building Startup Operate Full Flood Temp Pressurizer Range Range Range Tank (*F) (inches) (inches) (%) (%) (ft) 100 6 1 1 1 0.0 150 8 6 2 1 0.2 200 15 13 4 3 0.4 250 24 21 6 5 0.7 300 33 30 8 8 1.1 350 45 41 11 10 1.4 400 58 50 14 14 1.9 Modification Factor to Indicated Level Corrections NOTE: Use the Modification Factor on the temperature compensated scales -) only (Steam Generator Operate Range Level and Pressurizer Level) l Steam Generator Operate Pressurizer Pressurizer Downcomer Range Temp (*F) Level Temp ('F) Level 68 1.00 68 1.00 150 1.02 150 1.02 200 1.04 200 1.04 250 1.06 250 1.06 i 1 300 1.09 300 1.09 350 1.12 350 1.12 400 1.17 400 1.17 l 450 1.23 450 1.21 500 1.31 500 1.31 550 1.42 550 1.36 600 1.61 650 1.99 Calculation of Actual Level Pressurizer Actual = Indicated level - (Correction x Modification Factor) Steam Generator Operate Range Steam Generator Actual level = Indicated level - Correction Startup Range Full Range Core Flood Tank EXAMPLE: Reactor Building Temperature 250*F Indicated Startup Level 220" Correction for R. B. Temperature 21" Actual Level = 220" - 21" 199" EXAMPLE: Reactor Building Temperature 300'T Pressurizer Temperature 400'F Indicated Pressurizer Level 100" Correction for R. B. Temperature 33" Modification Factor 1.17 Actual Level = 100" - (33" x 1.17) 61" I ~ - -. _ -.. _ _ _ _
1 lt__EBINCIE6ES_QE_NYCLE@B_EQWEB_E68@I_QEg8@I1QN PAGE 30 t IdEBdQQyN@d1CS _bE@I_IB@NSEE6_@NQ_E6 gig _E6QW t j o l ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C.
- q 1
ANSWER 1.01 (1.00) I ([) d REFERENCE OP-OC-SPS-PTR-AT pp 13/14; LO la (4.1/4.7) OOOO74A207 ...(KA*S) j i l ANSWER 1.02 (1.00) I b O RL C l REFERENCE [o P-O C-JFI-IC-ICJ pp (/(I OP-OC-TA-NT Figure 15; LO lo (3.6/3.8) 016000G015 ...(KA'S) l ANSWER 1.03 (1.00) b REFERENCE OP-OC-SPS-RT-FPP; LO 2 (3.4/3.4) (3.1/3.2) 192OO6K107 192OO6K109 ...(KA'S) ANSWER 1.04 (1.00) a. REFERENCE l Oconee OP-OC-CH-RC pp. 7 obj 2. 2.1/2.7 I OO4020K523 ...(KA'S) I 1 4 l l 1 L-_______.___-_____- .J
f 1.__BBINCIE6ES_QE_NQC6E98_EQWEB_E6@N1_QEEB@IlgN PAGE 31 t ISE6dQQyN901CS _bE@l_lB@NSEEB_@NQ_E(glp_E6QW t ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 1.05 (1.00) A-3. B - 2. l C - 4. D-1. REFERENCE Oconee HO OC-RT-SM-2 obj 2m. 3.8/3.8 192OO8K103 ...(KA'S) ANSWER 1.06 (1.00)
- c. (nL b REFERENCE Oconee OP-OC-SPS-RT-SM pp.15 obj 2.g 3.8/3.9 192OOBK105
...(KA'S) fusbwqag 3 op gucwwt Rracial fucwHnM pp i \\~5 ANSWER 1.07 (1.00) d. REFERENCE Oconee OP-OC-SPS-RT-IP pp. 28 obj 2a 2.8/3.1 192OOSK105 ...(KA'S) ANSWER 1.08 (1.00) c. REFERENCE Oconee OP-OC-SPS-RT-IP pp. 14 obj 1.K 3.1/3.4 OO1000K519 ...(KA'S) ANSWER 1.09 (1.00) c.
iz__BBINCIE6ES_QE_NQC65@B_EQWEB_E6@NI_QEEB@IlQN PAGE 32 z Id5Bd991N@dlCS _dE@l_IB9NSEEB_@NQ_E(Qlp_E(QW t g ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ,s REFERENCE Oconee OP-OC-SPS-THF-PD pp. 10 obj. 2d 2.9/3.3 193OO9K107 ...(KA'S) ANSWER 1.10 (1.00) d. REFERENCE Oconee Duke Thermo pp. 156 2.3/2.4 101004K111 ...(KA'S) ANSWER 1.11 (1.00) b. O(C d I REFERENCE Oconee Duke Thermo pp. 194 3.9/4.2 c P-OC-SP3-PTR ' /+M I PO "~ G 193OO8K121 ...(KA'S) i ANSWER 1.12 (1.00) l B. l REFERENCE Oconee Duke Thermo pp.168 2.5/2.5 i 193OO7K101 ...(KA'S) I ANSWER 1.13 (1.00) l i I a) Increase (+.5) b) Decrease REFERENCE OP-OC-TA-NT pp 16; LO ik t
1.__EBINCIELES_QE_NQCLE@B_EgWE6_E6@@I_QEEB811gN PAGE 33 t ,ISEBdQQYN8dlCS _dE9I_I.38NSEEB_98Q_E6gIQ_ELQW t j '.,, ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 1.14 (1.50) i a) Decrease (+.5 ea) b) Decrease c) Increase REFERENCE OP-OC-TA-NT pp 7/8; LO lb (3.6/3.9) 041020A202 ...(KA*S) ANSWER 1.15 (1.00) a) Lower (+.5 ea) b) Lower REFERENCE OP-OC-SPS-RT-RBC pp 26/27; LO 2b.4 (3.6/4.2) OO1010A207 ...(NA'S) ANSWER 1.16 (1.50) a) Decreases (+.5 ea) b) Increases c) Increases j REFERENCE DPC Thermodynamics / Fluid Flow, CH 2E (2.6/2.8) 191004K115 ...(KA'S) ANSWER 1.17 (1.00) 1. less than 1 2. greater than REFERENCE Oconee OP-OC-SPS-APC-T41 3.8/3.9 192OO2K114 ...(KA'S) i
r 1 l li__EBINQlELES_gE_8UC65@B_EgWg8_E6@NI_gEEB911gh PAGE 34 l It!E6dQDyN@dlCh _Hg@I_IB98 GEE 8_@dD_E6U1D_E69W s l ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. t ANSWER 1.18 (1.00) Quench Tank Pressure (+.5) stops increasing (+.5) REFERENCE OP-OC-SPS-CM-PZR pp 17; LO 11 (2.6/2.8) OO7000A206 ...(KA'S) ANSWER 1.19 (2.00) a) Unit 3 (+.5) due to a lower Beta coefficient at EOC(+.5) b) Unit 2 (+.5) due to MTC being less negative, so Tavg must decrease more to add + reactivity (+.5) REFERENCE DPC Fundamentals of Nuclear Reactor Engineering, CH 3 (2.9/3.2) (2.9/3.1) OOOOO3K11.6 192OO4K103 ...(KA'S) ANSWER 1.20 (1.00) 1) Opening of steam reliefs which don't reseat-not designed to pass water (any two at +.5 ea) 2) Loss of TDEFW pump from water entering turbine 3) Rapid cooldown of OTSG tubes-tube rupture QUER$TDLs\\ tg PL Sut*Pe nr$ C) qu gqpyt aggapy p,4g,4gp-(,) pzgosauq $/4 p};purryg REFERENCE (bars -- O y w r e n OP-OC-SPS-PTR-AT pp 30/31; LO le (2.7/3.1) O A* OC' S P3-Ch - MT PP A S j O Y '# ~ S f5 ~ (l1 ~ SS (P ll 059000A203 ...(KA'S) ANSWER 1.21 (1.00) 1) Radiolysis (+.33 ea) r 2) Zinc-boric acid reaction o (L 577Wf-JTE% Ri4c118# ge 3) Zire Water reaction g. REFERENCE OP-OC-SPS-HY-HDC pp 10-12; LO la (2.9/3.6) 028000K503 ...(KA'S) l i L d
1 1___EB1NClE6(S_gE_Nyg6E@B_EgWE6_E6@NI_QEEB@llONt PAGE 35 l l THERMODYNAMICS _ HEAT TRANSFER AND FLUID FLOW t l ANSWERS -- OCONEE 1, 2L3 -87/07/13-CASTO, C. l l ANSWER 1.22 (1.00) These rods are GRAY ( l ower absorbtion cross section) and l have a longer effective poison length. (+.5) They have a i Imbalance (and minimize l less severe impact on Axial Flux oscillations)(+.5) ) induced. xenon REFERENCE OP-DC-SPS-THF-PD pp 12; LO 2f.4 I (3.2/3.5) 192OO5K114 ...(KA'S) l l ANSWER 1.23 (2.00) a) Increases (+.5) Pzr temperature increase will cause a pressure increase, increasing margin to saturation (+.5) b) Increases (+.5) Delta t across the core will be lower to produce the same power. Th wil decrease and the coolant in the upper regions will be farther from saturation. (Also higher flow removes bubbles from rod surface)(+.5) REFERENCE j DPC Thermodynamics / Fluid Flow pp 196-198 (3.4/3.6) 193OO8K105 ...(KA'S) j ANSWER 1.24 (1.00) There is an induced temperature stress which is compressive on the inner wall and expansive on the outer wall. CO.53 There is an induced pressure stress which is compressive on the inner wall and expansive on the outer wall. CO.53 REFERENCE Oconee Duke Thermo pp. 225 3.8/4.1 193OlOK107 ...(KA'S) l c ____
l iz__EBING1ELES_9E_89GLEeB_EQWEB_EL8NI_QEE8@IlgN PAGE 36 t IMEBdQQYN6dlC@t_dE@I_IB8NgEE6_@NQ_E6UlQ_E6QW ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 1.25 (1.00) at 2240 psia, hg = 1115 BTU /lb (+.5) at 20 psia, at saturation conditions, hg = 1156 BTU /lb and hf = 196 BTU /lb =.043 >>> 95.7% quality calculates (1156-1115)/(1156-196) If use Mollier: 95% quality (+/- 1%) REFERENCE OP-GA-SPS-THF-STM pp 20/21; LO 2e (3.3/3.4) l 193OO3K125 ...(KA'S) l l ANSWER 1.26 (1.00) C REFERENCE DPC Thermodynamics, pp 192-5; ^ (3.1/3.4) OO2OOOK501 ...(KA*S) l l ANSWER 1.27 (1.00) l See attached sketch REFERENCE DPC Thermodynamics / Fluid Flow, pp 172-177 (2.2/2.3) 191006K103 ...(KA'S) l l l 1 1 i
2___P6@NI_QEgl@N_lNC6yplNQ_@@ Eely _@NQ_EDE6@ENCy_@y@IEdg PAGE 37 ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 2.01 (1.00) c. REFERENCE Oconee OP-OC-PNS-CRD pp. 14 Obj. 1.b. 3.4/3.6 OO1000K103 ...(KA'S) l ANSWER 2.02 (1.00) l C l REFERENCE l OP-DC-SPS-SY-EF pp 21/22, 32/33; LO le, if (3.4/3.5) 061000A207 ...(KA'S) ANSWER 2.03 (2.00) a) 3 (+.5 ea) l b) 3 I c) 2 d) 6 REFERENCE OP-OC-SPS-SY-RCS pp 12/13; LO la (4.1/4.1) (3.7/4.0) (4.5/4.6) OO2OOOK106 OO2OOOK108 OO2OOOK109 ..,(KA'S) ANSWER 2.04 (1.00) d. REFERENCE Oconee AP/1/A/1700 pp. 3 3.4/3.6 078000K302 ...(KA'S) L
1 2 __PL9NI_DEgigN_lNCLUDINg_g8EEIY_@ND_EdERgENCY_gYgIEdg PAGE 38 1 ) l ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. i l s l 1 i ANSWER 2.05 (1.00) ( d 4 REFERENCE Oconee SY-LPI-5 3.2/3.5 OO5000K402 ...(KA*S) ANSWER 2.06 (1.00) a. REFERENCE Oconee OP-OC-SPS-IC-RPS pp 23 1.g 3.1/3.5 012OOOK603 ...(KA'S) ANSWER 2.07 (1.50) a) They bypass the normal safety injection valves in case of a injection header line break and help mitigate Quarter Core Cooling (+1.0) b) AJrtie3' ( +. 5 ) ALL ) va an REFERENCE OP-OC-SPS-SY-HPI pp 27, 28; LO 3f, 3k o p-oC - SPJ-7C-SPDf - 99 39 (3.9/4.2) OO6000K406 ...(KA'S) ANSWER 2.08 (1.00) a) A Submersible Pump is installed to discharge into the nearest CCW Piping manway. ( +. 5 e a ) [aV4wan e ns #. ni me4f 70 c (7 cygrne s n "mm b) An Air Ejector is valved into the SSF ASW Suction piping de t s ' t oi r ) when the SSF i s actuated. ([v4LvArG' oTWIC AvM43 9,1 mws+uuus nia emxy ro t/cs paa enemm) OP-OC-SPS-SSF-ASW Tracking 87-034/030; LO 2b (3.9/4.2) 061000K401 ...(KA'S) s
2___PL@NI_pE@l@N_1NCLyQ1Ng_@@EEIY_@NQ_EDERgENCY_@Y@ led @ PAGE 39 . ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. 1 l ANSWER 2.09 (1,00) Possible overpressurization of Aux Steam and subsequently, steam line to TDEFWP if both control valves were to fail open due to too low of relief capacity in both systems. REFERENCE Oconee LER 87-003 (3.3/3.4) 03OOOOK107 ...(KA'S) I ANSWER 2.10 (1.00) Due to low CCW flow being supplied by gravity flow I C A C YClf A/4 OF 'FBZr I REFERENCE I Oconee AP-11 3.6/3.7 OOOO55KOO7 ...(KA'S) 1 ANSWER 2.11 (1.00) ) 1 A letdown line from the Common letdown line (prior to the L/D coolers) thru HP-(426) to the fuel transfer tube.CO.53 A line from the SSF Makeup Pump which takes suction on the altranate fuel transfer tube, injects water through SSF-HP-(398). This line serves as system makeup.CO.53 l REFERENCE Oconee OP-OC-SPS-SY-HPI obj 3.a/b 3.3./3.2 i i OO4000K405 ...(KA'S) l l ANSWER 2.12 (1.50) l l l 1. To insure suction to the HPI pumps is not lost [0.753 2. Insure gas does not enter the HPI Pump suction on an ESFAS CO.753 REFERENCE Oconee OP-DC-SPS-SY-HPI pp. 19 obj 5.b 2.8/3.0 3.1/3.4 OO40010KO1 OO4020A104 ...(KA'S)
2.__PL9NT_QE@lGN_lNCLUDlNG_@@FEIY_@ND_EMEBGENCy_@y@IEM@ PAGE 40 ANSWERS -- OCONEE 1, 2&3 -B7/07/13-CASTO, C. ~ ANSWER 2.13 (1.00) Supplies voltage to the RCP Power monitor circuits, that in turn,_ supply input voltage signals to the contact monitor u EO.53 APS converts the 120 VAC RPS channel supply voltage. CO.53 REFERENCE Oconee OP-OC-SPS-IC-RPS pp. 27 3.4/3.7 l CM-CPM obj 1.d/e 012OOOK102 ...(KA'S) ANSWER 2.14 (2.00) 1) Both MFWPT hydraulic oil pressure < 75 psig (+.5 ea) 2) Both MFWPT discharge pressure < 750 puig 3 )P.,25" with 1oss of both MFWPT 4) 240" with loss of both MFWPT & All RCPs REFERENCE OP-OC-SPS-SY-EF pp 64; LO ik, il (4.5/4.6) 061000K402 ...(KA'S) ANSWER 2.15 (2.00). (br a *V d 1) Main generator to 1T to 1TA/TB and MFB1/2 (+.33 ea) 2) 230 KV Swyd to CT1 to 1TA/TB and MFB1/2 iTA/TB and MFB)1/} gg g,.r I/4 wmMA W 4 Ed f 44
- I/4 Ovf 3)
Keowee to CT1 to ( A) Keowee to CT4 to Standby Bus 1/2 6 5) Lee Gas Turbines to CT5 to Standby Bus 1/2 qsM 23OKV or 525KV Back Charging to 1T to ITA/1TB and MFB1/2 REFERENCE OP-DC-SPS-EL-EPD pp 17/1B; LO 2 C P /i/d //I07/81 T3 T. % / (6)1 j (3.7/4.2) (2.9/3.3) 062OOOK104 062OOOK406 ...(KA'S) e
l 2.__P(@NI_DEglgd_ldC6gDIdg_g@EEIY_@ND_EdgBgENCy_gygIEdg PAGE 41 l , ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. = 7 i ANSWER 2.16 (1.00) 1) CRD Stators (+.25 ea) 2) Letdown HXers 3) Quench Tank Cooler O U' 3"g ! fpggj # [7/(f(NM I 4) RCP Cooling Jacket / Seal Coolers i l REFERENCE OP-OC-SPS-SY-CC pp 8; LO 1 (3.4/3.5) OO8000K3.0 ...(KA'S) ANSWER 2.17 (1.00) 1) Excessive Boration of Plant (+.25 ea) 2) Plant over pressurization 3) Chemical Spray Hazard to RB Components c omp on e n t s [/fy/,,g p <>[ r,t c//v< c/4'4./ 4) Loss of Cooling Water to necessary Tb" REFERENCE C4I" # M #*N' *** # OP-OC-SPS-IC-ES pp 15; LO lj IU/*dl (3.7/4.0) 013OOOA206 ...(KA'S) ANSWER 2.18 (1.00) 1) RB Dome (+.25 ea for any 4) 2) RB Canal / Operating Area 3) RB Basement 4) RB RIA Inlet 5) RB RIA Outlet REFERENCE OP-OC-SPS-SY-HDC pp 13; LO 2b (3.1/3.3) 02OOOOA403 ...(KA'S) ) l 1 1 i l i ___________m___.._
2___ELONI_DEQ1GN_lNCLUDING_S@EEIy_@ND_EDEBGENCY_gYSIEdg PAGE 42 ANSWERS -- OCONEE 1, 2L3 -87/07/13-CASTO, C. ANSWER 2.19 (1.50) 1. Nuclear flux channels 2. Control rod position 3. Reactor trip signal 4. Control rod drive breaker position any 3/4 for .5 ea. REFERENCE OP-OC-SPS-IC-SPDS pp. 17 obj.1.1 2.8/3.3 2.4/2.7 OO1000K602 OO1000K605 ...(KA'S) l ANSWER 2.20 .Lt_.SO a) The Megawatt error is blocked in track, so the header i pressure error is controlling position of the control l valves. (+.5) This causes instability in header pressure, hence feed flow esci11ations. trt - AY1-ICS Etati~ons-intu automati c-mode-bef ore-the-ttsrtriTie h beced 4 n--at2teq-t+;5)- is REFERENCE OP-OC-SPS-IC-ICS Tracking #86-064; LO lb (3.2/3.2) 059000K107 ...(KA'S) ANSWER 2.21 (1.50) J l 1) Use HP-26 Motor operated valve (+.5 ea) 2) Use HP-122 (Manual Bypass) for a 44 3 3) Use "C" HPI pump via HP-27 4) oew up-wo t) %taue!,ed % w wi crea REFERENCE OP-DC-SPS-SY-HPI pp 29; LO 3c op/i/4/nsy/ot (3.4/3.6) 07BOOOK302 ...(KA'S) l l 1 L
3.__P(@@I_QE@l@N_lNCbyQ1NG_@@FEIY_@NQ_EdE@@ENCY_@Y@ led @ - PAGE 43 ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTOo C. ANSWER 2.22 (1.50) a) They would close automatically, which could result in loss of feed to the S/Gs. (+.5 ea) b) Prevent EFWPT cavitation overpr essur i.z.ati-on-of--theMurbtnE casi ^ q. b J,f O 8 /M f"4" c) Prevent r 1 t~ j, 1 a. REFERENCE OP-DC-SPS-SY-EF pp 67, 68; LO 3a (3.5/3.6) 061000G010 ...(KA'S) ANSWER 2.23 (2.00) An accident could impose #AdMi = stresses on the lower portions of the ductwork, causing deformation or possible collapse of the ductwork, thus blocking airflow [O.5] Fusible dropout plates are designed to melt and drop out of the ductwork opening up additional flow paths assuring a positive path for recirculation during an accident [O.5] Severe stresses imposed on the ductwork could possibly generate a shock wave through the duct and damage the cooling coils.[0.5] Blowout plates located in the lower portions of the duct would be forced out under these stresses and attenuate the shock wave before it reaches the cooling coil.[0.5] REFERENCE Oconee OP-OC-SPS-SY-RBC pp. 11 objs 1,b/1.c 2.9/3.2 022OOOK301 ...(KA'S)
3___ INS 169dENIS_@ND_CQNIBOLS PAGE 44 ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. t ANSWER 3.01 (1.00) b REFERENCE OP-OC-SPS-IC-ICS pp 23/24; LO 1h (3.2/3.2) 059000K107 ...(KA*S) ANSWER 3.02 (1.00) l \\ REFERENCE OP-DC-SPS-EL-EPD pp 24; LO 3c.5 (2.6/3.2) 062OOOK401 ...(KA'S) ANSWER 3.03 (1.00) a REFERENCE OP-OC-SPS-SY-CC pp 19; LO 2c (3.2/3.2) OOB010A301 ...(KA'S) I ANSWER 3.04 (1.00) C l REFERENCE OP-OC-SPS-IC-ES pp 18; LO 1h (3.3/3.7) 013OOOK410 ...(KA'S) l ANSWER 3.05 (1.00) C. l l s
1 }g__lNSI6YDENIS_@ND_CONIBOLS PAGE 45, ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. b 1 f REFERENCE Oconee OP-OC-SPS-SSF-EPS 2.6/2.7 062OOOA404 ...(KA'S) ANSWER 3.06 (1.00) d. REFERENCE Oconee OP-OC-SPS-SY-EF objs 1.b/1.a/1.o 3.4/3.8 061000A204 ...(KA'S) ANSWER 3.07 (1.00) C REFERENCE Oconee OP-OC-SPS-IC-CRI 1.m. 3.5/3.8 OO1000K401 ...(KA'S) ANSWER 3.08 (1.00) a. or 4 REFERENCE Oconee Op-OC-IC-RCI obj. 3.g. (3.1/3.6) OO2OOOK603 ...(KA*S) ANSWER 3.09 (1.50) JtRR t ASES a) Decreases (+.5 ea) b) Opens c) Decreases REFERENCE rp 4Q OP-OC-SPS-IC-ICS, pp 80/81; LO is (3.0/3.1), (3.4/3.6) 016000K303 016000K312 ...(KA'S)
1 3___INSIBydENIS_AND_CONIBOLS PAGE 46 ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 3.10 (1.50) a) Emergency (+. 5 ea) b) Both c) Emergency REFERENCE OP-OC-SPS-KHG Fig 8/9; LO 1 q, ir (3.4/3.6) l 064000 GOO 7 ...(KA'S) i ANSWER 3.11 (1.50) a) Bistable (+.5 ea) b) Buffer c) Buffer REFERENCE j OP-OC-SPS-IC-RPS pp 19-23; LO if, 3a, 3b (2.8/3.3) 012OOOK601 ...(KA'S) 1 ANSWER 3.12 (1.00) l B-1 E-O G-1 c c, EC F REFERENCE Oconee OC-EL-EPD-1 obj.1.a/1.b 2.5/3.2 194001A107 ...(KA'S) ] \\ l l ANSWER 3.13 (1.50) g corbwt' (Oc n- ($ G Hf ASCf delW4 a) Key Switch in ICS Cabinet (+.5 ea) b) Ala ive-F aits-e s-i s-no-<on trol--po s s i b it. coAf eed (mA swd'1-c) Key Switch in ICS Cabinet l REFERENCE OP-DC-SPS-IC-ICS pp 53, 58; LO it Mp //S//74[2 J TNct 4. 2. (3.5/3.5) OOOO57A106 ...(KA'S) l l w---_____-_--_____-__________-____-__________
~ 3 __IN518UMENIS_@ND_CONIB06@ PAGE 47 s ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. r 'i ANSWER 3.14 (1.00) The possibility of overloading the spider. exists if. the CRD l is operated in JOG when the CRD is not' free running. REFERENCE OP/0/A/1105/09 Oconee 3.3./3.5 OO1010A010 ...(KA'S) i !1 ANSWER 3c15 (1.00) the Motor Fault Lamp illuminates CO.253 and the Diamond station swaps to manual [0.753. REFERENCE Oconee OP OC-SPS-IC-CRI 1.1/1.g 3.8/3.8 l OO1402K402 ...(KA'S) l ANSWER 3.16 (1.00) Inputs = Signal from both the upper and lower linear amplifiers (top and bottom detectors) Outputs = Imbalance meters, RPS, function generator (Flux / Flow / Imbalance), Computer Eea O O.23 REFERENCE Oconee OP-DC-IC-NI pp 26 obj 1.g 2.6/2.9 015000K602 ...(KA'S) ANSWER 3.17 (1.00) or4LO4tu5p/hr a. The operator places the selector switch to LOW b..ves* WO e s. Va lwd in [O.5 ea] REFERENCE Oconee OP-DC-IC-RCI obj. 2.0. (4.2/4.4) o p f f /4 /M ot /r0 e n r / V 1. OO2OOOK410 ...(KA*S)
l 3. INSTRUMENTS AND CONTROLS-PAGE.48 ' ANSWERS --lOCCNEE 1, 20<3 -87/07/13-CASTO, C. ANSWER 3.18 (1.00)
- 1) the SSF incoming feeder breaker from Unit 2 Main Feeder Bus #2 (OTS1-1) is open
(+.5 ea)
- 2) SSF Diesel Generator Breaker OTS1-4 is closed REFERENCE OP-OC-SPS-SSF-ASW pp 24; LO 2h (4.0/4.2) 061000K406
...(KA'S) ANSWER 3.19 (1.50) 1) Bypasses Low Pressure Trip (+.25 ea) Variable Low Pressure Trip Power /RCP Trip Flux / Flow /Jgpalance Trip 1720 psig J.R Pressure Trip [f% y TT )3 setpoint 2) Inserts setpoint is reset to 4*/. 3) High Flux REFERENCE OP-DC-SPS-IC-RPS pp 16; LO le (3.3/3.6) 012OOOK604 ...(KA'S) ANSWER 3.20 (1.00) fSmM&t,1-hirrctnehord4k-2-ene-1 - Gow' fs,. n. ? -- .:,, ~ t r. .s (~ ., s i. = 3 - t+e-g:jy.. -33 Jcr e c;c G 1 m 6 L ug te REFERENCE OP-OC-SPS-IC-RPS pp 39-45/ fig 24b; LO 2a (3.1/3.5) 012OOOK603 ...(KA'S) )
3.__IN@l8UMENIS_@ND_CgNI69L@ PAGE 49 .A$5WERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 3.21 (1.00) Normal incoming switchgear breaker OTS1-1 trips SSF Supply breaker 82T-4 trips CO.5 ea3 REFERENCE Oconee OP-DC-SPS-SSD-EPS pp. 36 obj 1.f. 2.8/3.1 062OOOK403 ...(KA'S) ANSWER 3.22 (1.50) Gr &*f $ RC Makeup Controller 3 g g.j Letdown Flow Controller 7gg LDST 1evel yp m_ , ywyn, ~ REFERENCE Oconee OP-DC-SPS-EL-VPS pp 20 obj 2.d 3.5/3.9 kuG o c-st - VM -/ V 062OOOK301 ...(KA'S) ANSWER 3.23 (1.00) a) All three units (+ 5 ea) b) Units 2 and 3 1 REFERENCE OP-OC-SPS-SY-HPI, pp 36/37; LO 3q (2.8/3.1) OO3OOOK404 ...(KA'S) ANSWER 3.24 41.00) Gamma ray sensitive, in that a N16 Gamma would cause tube to i avalanche, indicating an OTSG tube leak. (+1.0) REFERENCE l OP-OC-SPS-IC-ARM pp 10/11, LO la, if l (3.9/4.1) OOOO38 GOO 5 ...(KA'S) l I
3 __INSIByMENIS_9NQ_CQNlog6E PAGE 50 t ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ' = ANSWER 3.25 (1.00) (this would deenergize both SV-209 and -208) causing LPSW-138 to fail open supplying HPSW to EFWPT pump bearings EO.52 and causing HPSW-184 to open providing cooling water to the oil cooler CO.53 REFERENCE Oconee AP-13 pp. 8 obj. 1.o, 1.p~, 3.1/.3.3 OOOO75KOO7 ...(KA*S) ANSWER 3.26 (2.00) a. 1. withdraw 2. ICS-runback FW-reduces l 3. Reactor trip on high RCS pressure (or neutron flux) 4. fails closed due to f alse low pressure signal fails closed due to f alse low pressure signal
- b. The heaters remain on(due to false low pressure & due to
- 4 above) pressure continues to increase. o ft Cwrot /caot w lNc/thw4 (part a.
O.3 ea. part b. O.2) DA Tlfhic nca rH t-S go u try t. REFERENCE Oconee Simul ator Ref Manual ICS LP 3.6/3.9 012000A202 ...(KA'S) I
65._ _E BQQ E QLJB E S_ ; _NQEd@6 t _ @ @ Ngg d@6 t _ E DE6@gyg y _ AND PAGE 51 609196991G66_ggN18QL ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. ANSWER 4.01 (1.00) ,)* i REFERENCE Oconee Op/0/A/1105/09 3.3./3.5 OO1010A010 ...(KA'S) A N S W E R, y 4. O 2 (1.00) 6d MY CJ REFERENCE DPC EPG pp 2-77/79; (4.0/4.4) OOOO74K311 ...(KA'S) ANSWER 4.03 (1.00) b REFERENCE I OP/2/A/1102/01, Enclosure 4.3, pp 4 l (3.6/4.2) OO1010A207 ...(KA'S) ANSWER 4.04 (1.00) 1 C REFERENCE 10CFR20.5 (2.8/3.4) i 194001K103 ...(KA'S) I ANSWER 4.05 (1.00) a. l
gg__PEOGEQQBES_;_NOBD961_9BNQBD@6t_gDEB@gNQY_@NQ PAGE 52 Be91969GICe6_CgNIBg6 ' ANSWERS -- OCONEE 1, 2L3 -87/07/13-CASTO, C. REFERENCE Oconee AP/1/A/1700/23 3.5/3.5 3.2/3.4 OOOO57A105 OOOO57A106 ...(KA'S) ANSWER 4.06 (1.00) a) Cycled electrically (+.5) b) 500; 6 (+.25 ea) i REFERENCE Oconee OMP 2-1, pp 3/4 (3.5/3.6) (3.4/3.7) OO6050G010 061000G010 ...(KA*S) ANSWER 4.07 (1.50) a) 4.5 rem (+.5 ea) b) 75 rem octfavs c) 7.5 rem REFERENCE I DPC HPM, pp 9/10; Emergency Plan, pp K-1 (2.8/3.4) i l l 194001K103 ...(KA*S) ANSWER 4.08 (1.00)
- 1) Close the Field Breaker
(+.33 ea)
- 2) Close the Generator Supply Breaker
- 3) Close the Field Flashing Breaker i
REFERENCE OP-DC-SPS-CM-KHG pp 20; LO ik (4.0/4.3) 064000A401 ... ( K A' il) ANSWER 4.09 (1.00) i 1 3, 2, 1, 4 (.25 for each switch to get correct order) )
4:__P69CEggBEg_;_NgBd9(t_@@Nggdg61_EdEBgENCy_@Ng PAGE 53 609196991C66_C9N16g6
- ANSWERS -- OCONEE 1,
2L3 -87/07/13-CASTO, C. ~ 1 REFERENCE EP/1/A/1800/1, pp 9/10; Reference Document pp 1-B I (3.8/3.9) OOOOO7G012 ...(KA*S) ANSWER 4.10 (1.00) This identifies steps that may be done in parallel with the critical path (numbered) step. ) REFERENCE Oconee OP/1/A/1102/01, pp 3 4 (4.1/3.9) 194001A102 ...(KA'S) ANSWER 4.11 (1.00) l a) Thermal Power-Secondary Heat Balance (+.5 ea) b) Minimize trip on flux / flow / imbalance $on an operating transient ga) [A e r) REFERENCE OP/3/A/1102/04, Enclosure 3.2, pp 1 (3.1/3.2) OOOO17 GOO 7 ...(KA'S) ANSWER 4.12 (2.00) I 1. From the Spent Fuel Pool
- 2. Aux. Service Water Switchgear OA 5 7"BN SsI * /
3. From CT-5 c t Li t mencs s /t. s m D N % I*'/ ,g , ua I 4 From the Aux. Service Water System gg ,95 cg CO.5 ea.] e ff-H PIw g REFERENCE AP/1/A/1700/19 Oconee 3.7/4.1 3.9/4.7 0F-C C-TrJ-GT-t ft Duc d t ? OOOO55A203 OOOO62A211 ...(KA'S) c/ 0C - frJ / y #tt rp t Y l l /
f I 4. PROCEDURES - NORMAL _ABNQRM@L _EMERGENgy_9NQ PAGE 54 t t 6091969GIC06_CgN18Q6 ANSWERS -- DCONEE 1, 2&3 -87/07/13-CASTD, C. l ~ ANSWER 4.13 (1.00) Letdown should be established to offset the increase in RCS inventory due to the initi ation of E-boration. REFERENCE Oconee EP/1A/1800/01 4.4/4.7 OOOO29K312 ...(KA'S) ANSWER 4.14 (1.50)
- 1) SU Cntrl Valve (FDW-35)
(+.25 ea)
- 2) EFDW Cntrl Val ve (FDW-315)
- 3) TBV Block Valve (MS-17)
- 4) MS to SSRH (MS-79)
- 5) MFW Block Valve (FDW-31)
- 6) SU Block Valve (FDW-33)
REFERENCE Oconee OMP 2-1, pp 1 enc 1 4.4 (4.1/4.2) OOOO40G010 ...(KA'S) ANSWER 4.15 (2.00) 1) Set batch size on makeup control to 90,000 gallons and reset (+.25 ea) 2) Open Makeup isolation (HP-16) 3) Open RC Bleed Transfer Pump "A" Discharge CS-46 4) Start "A" Bleed Transfer Pump 5) Dispatch operator to Units 1/2 Waste Disposal Panel 6) Go to Aux Shutdown Panel with: Reactor Log (+.5 total) l Emergency and Abnormal Procedures i Removal / Restoration Book j Emergency Plan l 7) Maintain Hot Shutdown Conditions l 1 REFERENCE Oconee AP-8, pp 2/3; DMP 2-1, ENCL 4.4 (4.1/4.2) OOOO69GO10 ...(KA'S) D
O.__P8QgEgyBES_ _NQBD@6t_@BNggd@(t_EDEBQENgy_@NQ PAGE 55 609196991G06 ggN16QL , ANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. p ANSWER 4.16 (1.00) 1) Reactor startup within 12 hours of trip (+.5 ea) 2) Cooldown has not been initiated. REFERENCE Oconee OP/1/A/1102/02, pp 1 (3.3/3.5) OO1050G010 ...(KA'S) ANSWER 4.17 (1.00) (ora 3 L To provide adequate core cooldown rate CO.53 and maintain R.B. environmental qualification criteria for R.B. equipment protection CO.53. & Pr*v!SC Gbpg (colt OS (5Z Zuc e s k b st<$ REFERENCE @ N EWW (ouuc sucnN w
- # *
- W (3 guap g/g Oconee AP/1/A/1700/07 pp. 5 3.4/3.7
( o,5) OO6010KO10 ...(KA'S) ANSWER 4.18 (2.00)
- 1. will isolate the Waste Gas Tanks /)v0 7m# MM/r4W FrueNtrc 2.
Teminate a liquid waste release 1Hubraep Cwpf es Arr ttou ma raw' rW1 3. Isolate the RB normal sump A&b fmvo rys rmerams 4c,9am 4. Stop the RB purge fan, mini-fan, and isolate RB purge system [O.5 ea] REFERENCE /9/" Oconee AP-18 pp. 1 3.3/3.4 Cf-G ' JPJ - 7 C s#4M, //0 070000K114 ...(KA'S)
I ] l 4. PROCEDURES - NORMAL _ ABNORMAL _EMERGENCV AND PAGE 56 t t BB9196991Ce6_C9BI696 4NSWERS -- OCONEE 1, 2L3 -87/07/13-CASTO, C. g ANSWER 4.19 (2.00) 1) LPI is in operation, with a flow rate > 1000 gpm in each line and stable for > 20 minutes (+1.0 ea) DN i 2') All RCS hot and. col'jd legs at lepst 50 degreps less t"an j RCS' pressure,and' securipg'HP isf t h e's a t ur a t i,on't emp / for g ,necessart,to prevent PZfVl evel m going ofy scale yAh l REFERENCE OP-OC-SPS-IC-ES pp 24; dfF [/b9//f03 /d / (4.1/A.6) 009000&*24 ...(KA*S) ANSWER 4.20 (1.00) To minimize inventory depletion, such that if the pumps were lost inadvertently, core uncovery could occur. REFERENCE Oconee EPG Reference Document, pp 4-45/46 (4.1/4.2) { OOOO74K308 ...(KA'S) l l ANSWER 4.21 (1.00) a) Allows the system to keep a vacuum on the condenser j b) Allows more UST water available to cool CSAEs (+.5 ea)
- b" REFERENCE l
Oconee LER 86-011; AP-11; (4.3/4.6) OOOO55K302 ...(KA'S) 1 /
[l' 4:__B699EQQBES_;_NQBd@(t_@BNQ80@(t_EDE6@ENQY_@NQ PAGE 57 609196901986_QQNIBQ6 l 1 l ',gANSWERS -- OCONEE 1, 2&3 -87/07/13-CASTO, C. l ANSWER 4.22 (1.50) Saturation Repressurization occurs due to lack of a condensing surface being established in the OTSG for refilux boiling to occur and the RCS repressurizes along the h i-r O r b
- d )
o P odiid> b N # # E saturation line due to decay heat absorbtion. GMT ( f f.0 To terminate, need to open head vents or PORV (-G%k.X:.c4WW or 1 Cit ~Cas< decrcT4th&ty&4el:-.:' $wennik REFERENCE <>f T a ft + /-gt p e.La,,; u ;'. ) Oconee EP/1/A/1 BOO /1, pp 26; EPG Reference Document pp 3-21 ((3.7/4.2) OOOO74K302 ...(KA'S) ANSWER 4.23 (1.50) 532 degreet ensures that the saturation pressure is well below the lift setpoint of the lowest set MSRV (+.75 ea) 80 inches ensures that you are above the heater level so that normal RCS pressure control can be used and so subcooled margin cannot be lost REFERENCE Oconee EPG Reference Document, pp 3-37 (4.2/4.5) OOOO3BK306 ...(KA'S) ANSWER 4.24 (1.00) 284 inches +-3 inches REFERENCE Oconee EPs 3.7/4.1 103OOOA101 ...(KA'S) I l l l l l
i, s s =. FIGURE 4 SIMPLIFIED LUBE OIL COOLER Toi k Olt OUTLET O!L INLET i T E MPE RATUR TEMPE R ATURE i , Tc w E N COOLING WATER ~ ..UTLET a TEMPERATURE 3 COOLING WATER F INLET TEMPER ATURE LENGTH TR AVELED FOR FLUID.THROUGH HEAT EXCHANGER ~ FIGURE 5 ..<)v me.,%~<as a - (4. <) (cc pqu c/m+c.s /~ 44 + ##
N ENCLOSURE 3 I SPECIFIC COMMENTS REGARDING THE REACTOR OPERATOR LICENSING EXAMINATION Category 1.0 Principles of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer, and Fluid Flow 1.02 Another acceptable answer to this question is j "c. Delta Tc Failure "B" Side Low". The Delta Tc circuit does a comparison of the two loop Tc's and will reratio feedwater accordingly in order to balance the TC's. "B" Side Low will generate the same system response as "A" Side High in order to achieve this balance. l l
Reference:
OP-OC-SPS-IC-ICS pages 44 of 87 1.06 Another acceptable answer to this question is "b. The Subcritical Multiplication Factor (M)". It is true that the effective half life of the delayed neutron precursors play a role in the length of time required to reach an equilibrium sub-critical j countrate. However, the time required for a given reactivity addition increases as the reactor approaches criticality. This is a result of the magnitude of change in "M" which will requ;.re more neutron generations to achieve the new equilibrium countrate given essentially no change in a neutron generation lifetime.
Reference:
Fundamental of Nuclear Reactor Engineering Page 117 OP-OC-SPS-RT-SM pages 14-15 of 22 1.11 Another acceptable answer to this question is "d. A decrease in Two-phase (nucleate boiling) within the core". A decrease in nucleate boiling within the core can be interpreted to be a decrease in the heat source. A decrease in the heat source would reduce natural circulation flow due to a decrease in I the delta T reducing the driving head of the fluid. 1 i l Page 1 l
An examiner clarified answer d." to at least one candidate as increasing subcooled margin which supports the interpretation of a decrease in the heat source. Referencel Thermodynamics, Fluid Flow, and Heat Transfer for Nuclear Power Plants I Page 195 l OP-OC-SPS-PTR-AM1 Pages 11, 12, and l 13 of 50 l 1.19b Another acceptable answer to this question is " Unit 3 because rod worth increases over core life resulting in a larger negative reactivity addition on Unit Three and therefore a larger j decrease in Tave." The change in rod worth competes with the change in the MTC and without additional information the overriding effect can not be i determined.
Reference:
OP-OC-SPS-RT-IP Page 22 of 30 1.20 Other acceptable answers to this question include: 1 If the overfill were to result in filling the l steam lines the weight of the water could cause l damage to the steam line supports and stanchions. i (Filling steam lines is acceptable at Oconee if pipe hangers have been blocked / braced). Water carryover to the Turbine-Generator could result in damage to the blading of the turbine. Flooding the aspirating ports in the steam generator would cause a loss of preheat and could possibly lead to thermal stresses on the lower tube sheet and/or water carryover to the Turbine-Generator.
Reference:
OP-OC-SPS-PRT-AT Pages 30 of 44 OP-OC-SPS-CM-MT Pages 97 of 113 OP-OC-SPS-CM-SG Pages 11 of 22 l I l Page 2 1 )
1.21 The phrase "in the primary" contained in this question could be misleading. The Zinc-Boric acid reaction is not "in the primary" and would be omitted as a source of hydrogen if the candidate is mislead. Instead of the Zinc-Boric acid reaction some cdndidates may list the Steam-Steel reaction which would occur "in the primary". Also, an examiner clarified this question to at least one candidate as a " gross failure." This clarification led the candidate to believe a greater than 1% zirc-water reaction occurred (as assumed in OP-OC-SPS-SY-HDC) and therefore changed the order of significance of the different sources.
Reference:
OP-OC-SPS-SY-HDC Pages 11 and 12 of 25 OP-OC-SPS-PTR-AM2 _Drwg. OC-PTR-AM-2 1.22 Most examinees probably will not address the effect on induced xenon oscillation from APSR's, since our stress is mainly in the area of operationally related axial imbalance control via the part-length control rods. While it is recognized that any pertubation of the flux will perturb the xenon distribution, the principal indicator and concern for the operator when dealing with APSR's is axial imbalance. l.25 The answer key is incorrect when using Mollier. Moisture content is 5% / Quality is 95%.
Reference:
Mollier Chart i l 4 I l l 1 Page 3 .]
Category 2.0 Plant Design Including Safety and Emergency Systems 2.07b All units have an indication on the computer display for HP-409/410 not being fully closed. Digital p.oints have been installed on all units which provide indication / input to SPDS.
Reference:
OP-OC-SPS-IC-SPDS Page 39 of 50/ Drwg. OC-IC-SPDS-21 2,08a The wording on this question is misleading. The Submersible Pump is used to supply water to the CCW Intake Piping and not the CCW Intake Area. b The wording on this question is misleading. The Air Ejector is used to evacuate any air which may collect in the common SSF Service Water suction line and possibly cause cavitation of pumps taking suction from this source. Minimizing air injection into S/Gs is not the reason for valving in the Air Ejector. 1
Reference:
OP-OC-SPS-SSF-ASW Tracking 87-034/030 2.10 This question may also be interpreted as asking to l explain the cycling of the TBVs. If interpreted this way another acceptable answer would be to address steam entering the condenser at a rate where vacuum could not be maintained. After the TBVs close vacuum is regained such that the TBVs can once again open. The answer key only addresses why vacuum is low and does not address the cycling of the TBVs.
Reference:
AP/1/A/1700/11 2.13 The referenced objectives given for thfs question require knowledge of the function and basic operation of the RCP Monitor. These objectives are in the OP-OC-SPS-CM-CPM lesson along with a basic description of the RCP Monitors necessary to meet the objectives. The contact Monitor Auxiliary Power Supply is not addressed in the OP-OC-SPS-CM-CPM lesson since it is not required to satisfy the objectives. Page 4
The Contact Monitor Auxiliary Power Supply is addressed in the OP-OC-SPS-IC-RPS lesson but no requirement is made (objective) of the operator to describe the Contact Monitor Auxiliary Power Supply. This question should be deleted.
Reference:
OP-OC-SPS-IC-RPS OP-OC-SPS-CM-CPM 2.14 The Emergency Feedwater System now controls SG level at 30 inches when actuated with at least one RCP running. Also, since loss of both MFWPTs is required for actuation of the Emergency Feedwater System the condition which determines the_ controlling SG level is whether or not a RCP is on.
Reference:
Information Attached OP-OC-SPS-SY-EF Page 64 2.15 When listing the flow paths of electrical power, an option available to the Oconee operators is use of a backup Startup transformer through the Emergency Startup buses supplying power to the 6900 and 4160 volt buses. On Unit 1 the backup Startup transformer would be CT-2. When listed in order of preference this flow path would be listed after the Normal Startup transformer and prior to the Standby transformer CT-4. (Between 3 and 4 on the answer key.) With this option, back charging the main transformer may be omitted since it requires considerable time to establish. Also, ITA/TB may be referred to as the 6900V buses and MFB 1/2 may be referred to as the 4160 V buses.
References:
OP-OC-SPS-EL-EPD Page 17 and 18 of 37 OP/1/A/1107/02 Normal Power Procedure Technical Specification 3.7.l(b)2. 2.16 On Unit 1 the RCPs are Westinghouse RCPs which do not have cooling jacket / seal coolers. Instead CC is used to cool the thermal barrier on the Unit 1 RCPs. A more appropriate answer would be simply the RCP seals when addressing this load.
Reference:
OP-OC-SPS-SY-CC Page 18 of 20 OP-OC-SPS-CM-CPS Drwgs. OC-CM-CPS-1 and 2 vano s
2.17 Some individuals will answer the question by breaking down " Loss of Cooling Water to Necessary components" to individual components. That is, they will list individually CRD's, RCP Motors, etc. When they have four things listed, they will stop. Therefore, their answers, although correct, will not . reflect all four of the answers given in the answer key. 2.20b Another acceptable answer would be, "By following procedure." The answer given on the key is a result of following the Unit Startup Procedure. This procedure will insure all ICS stations are in auto except for ATc and the second MFP. Neither of these stations will place ICS in track. Therefore, by following procedure the Unit should not be in Track when Turbine Control is placed__in automatic.
Reference:
OP/1/A/1102/01 Encl. 4.3 OP/1/A/1102/04 Encl. 3.3 2.21 Other acceptable answers to this question include: - Open HP-410 (this provides a bypass flow path around HP-26 should HP-26 fail to open / HP-115 is normally open) - Balance letdown flow with seal injection to mail.tain pressurizer level constant - Open HP-24/25 (if the question in interpreted to also mean makeup to the LDST / HP98, 99 and 100 are normally open)
Reference:
OP-OC-SPS-SY-HPI Pages 28 and 29 of 43/ Drwg. OC-SY-HPI-14 i OP/1/A/1104/02 High Pressure Injection l l 2.22c MS-97 is closed with its breaker locked open to prevent a loss of vacuum. j
References:
OP-OC-SPS-SY-EF Page 67 of 71 Page 6
Category 3.0 Instruments and Controls 3.04 While the various types of RZ modules are addressed in OP-OC-SPS-IC-ES, it is not intended that the operators'be able to classify the RZ modules according to " Type". The " Type" designation is used for ease of instruction by grouping the different components according to their RZ control module. The operators are instructed on recognition of the different modules and how each is to be operated. Given a drawing of the RZ module, or from the RZ control panel, the Operator can describe the operation of a particular module. This question should be deleted. 1
Reference:
OP-OC-SPS-IC-ES ~ Drwgs. OC-IC-ES-12 thru 17 3.05 No objective is referenced in support of this question. Operation of local switchgear controls is not performed by operations and therefore instruction on local switchgear controls is not addressed. This question should be deleted. 3.08 Another acceptable answer to this question is "c. A false zero level indication". The degree of failure or location of failure will affect the resultant indication. On a bellows rupture a false l zero level indication will result. i l \\
Reference:
OP-OC-IC-RCI Pages 28 and 29 of 49 l l Page 7
3.09a The description in OP-OC-SPS-IC-ICS is not accurate. The ETU limits have been removed above 25% full power. With the "A" TBVs open, the "A" S/G will have an increase in steam flow causing ATc to indicate "A" side cold. With this ATc indication feed flownto the "B" SG will be increased. Therefore the Initial Response for Feedwater Flow to "B" S/G will be to increase.
Reference:
NSM attached OP-OC-SPS-IC-ICS Pages 44 of 87 3.13a&b The pressurizer spray valve (a) and the pressurizer spray block valve (b) are both manually operable from the control board during a loss of Auto Power. However, there is no valve position indication available at this location. (The pressurizer spray valve can also be operated by a key switch in the ICS cabinet as indicated by the answer key.)
Reference:
AP/1/A/1700/23 Encl. 6.2 3.17a To place the Low Range Cooldown Pressure instrument into service the operator must contact the instrument department to valve in the transmitter. The instrument department will have the operator open RC5 (or 6) and RC-7. b The correct answer is NO. The PORV Setpoint Switch l is used to select the setpoint. The setpoint is not affected by valving the transmitter in or out.
Reference:
OP-OC-IC-RCI Page 21 of 49 OP/1/A/1102/10 Encl. 4.2 3.19 The answer key is incorrect in regards to the 1720 psig trip setpoint. The 1720 psig trip setpoint is a HIGH pressure trip. Also another acceptable answer for the administrative high flux setpoint is 5 5% as required by Technical Specifications.
Reference:
OP-OC-SPS-IC-RPS Page 16 of 63 Technical Specification Table 2.3-1 Page 8 u-__--______________________._______-._____
3.20 Terminology was a key factor in answering this question. The candidate could know how the logic and channel operated to produce a reactor trip but because of an inability to distinguish betwean the different terms he would be unable to correctly answer this~ question. The-terminology used in this ~ question is not used by the Oconee operator, therefore this question does not test the objective l referenced. This question should be deleted. l 3.22 Other acceptable answers to this question include: - (SAM) 3RC-1 - Turbine Bypass Valves - RCP Seal Flow Indication - Seal Supply Control Valve (HP-31)
Reference:
OP-OC-SPS-EL-VPS Page 20 of 26/ Drwg. OC-EL-VPS-14 3.26b Other acceptable answers to this question include: - Control rods withdrawing - Reduction in Feedwater Clarification provided to several candidates by the examiner related 3.26b back to 3.26a which supports the additional answers provided.
Reference:
OP-OC-TA-NT Pages 28, 19, and 26 of 29/ Drwgs. OC-TA-NT-14 and 21 Page 9 __________________________________________3
Category 4.0 Procedure-Normal, Abnormal, Emergency, and Radiological Control 4.01 This question and answer contain a double negative (NOT, NO) which makes it confusing. Also answer "a" could be~ considered as an incorrect answer due to both of these limits (in the procedure) being directly related to frequency of operation. " Rate" and " Frequency" can be interchangeable. With this interpretation there would be no correct answer. This question should be deleted. l
Reference:
OP/0/A/1105/09 I 4.02 The answer key references the DPC EPG for this question. Although our lesson _ plans on the EOP use the EPG as a basis, we do not instruct directly from the EPG document because it is a very detailed and technical paper that often contains material beyond the scope of what an operator needs. The particular area addressed by question 4.02 is, however, covered with the operator during accident mitigation lectures, specifically under OP-OC-TA-AM-2, Accident Mitigation: Gas / Steam ? Binding. Objective 1.c requires that the operator be able to describe the "means of dealing with steam / gas accumulation the the RCS". Attached are the two pages from the lesson plan that address this situation; from these pages it can be seen that vessel or hot leg venting is, indeed, a method of dealing with non-condensible gases in the RCS, but by no means, the "best" method. RCP Restart or in some cases, " bumps", is probably a better method, if RCP's are available. However, the situation will dictate the "best" method for removal of non-condensibles. Answers b, c, or d should be acceptable for question 7.0.1. 4.07b Another acceptable answer would be 25 rem. The Maximum Planned Emergency Exposure limit to the Whole Body to Save Lives per the Health Physics Manual is 25 rem. However, the Emergency Plan lists a limit cf 75 rem if approved by the Emergency coordinator. Either answer should be acceptable.
Reference:
Health Physics Manual Page 10 Emergency Plan Page K-1 RP/0/B/1000/11 Encl. 4,2 Page 10 _____-__________-_a
4.12 Other acceptable answers would include: 2. HPI pump electrical power source could also be listed as Standby bus #1 since this is the power source to the Aux. Service Water Switchgear. 3. Standby bus #1 power source could also be listed as the Central Switchyard or Lee Steam Station since this is the power source to CT-5. 4. HPI pump motor cooling could also be listed as HPSW since this is the backup cooling source used until Auxiliary Service Water is started.
Reference:
OP-OC-SPS-EL-EPD Drwg. OC-EL-EPD-2 OP-OC-SPS-SY-HPI Page 24 4.17 Another possible answer is; "To cool the LPI fluid after swapping to the emergency sump." Cooling the fluid would accomplish the other two bases by preventing cavitation of LPI and BS pumps and by removing heat from the core.
Reference:
OP-OC-SPS-SY-BS Page 9 4.18 Additional answers could include: 1. RIA 37 will also trip the waste gas exhauster. 2. RIA 33 will also trip both Condensate Monitor Tank Pumps. 3. RIA 49 will also sound the RB evacuation alarm.
Reference:
OP-OC-SPS-IC-PRM Pages 19 and 20 of 24 4.19 All RCS hot and cold legs at least 50"F subcooled and pressurizer level going off scale high is no longer a criteria which must be met in order to secure the HPI System. This criteria is not listed in the Emergency Plan and has been deleted from the OP-OC-SPS-IC-ES lesson in the last rewrite of this lesson.
Reference:
EP/1/A/1800/01 OP-OC-IC-ES (Last rewrite of OP-OC-SPS-IC-ES attached.) Page 11
4.22 This question had misleading terminology. While the term " saturated repressurization" is used in the EPG, the Emergency Plan and lessons do not use this term. Had the phenomena been addressed as it is in the Emergency Plan, the candidates would have understood what was being asked allowing this question ~ to test their understanding of the transient and not the " term". This question should be deleted. I i l l Page 12 \\
SPECIFIC COMMENTS REGARDING THE SENIOR REACTOR OPERATOR LICENSING EXAMINATION Category 5.0 Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics 5.02 Refer to R. O. Licensing Exam 1.02 1.11 5.07 1.20 5.19 ~ 1.21 5.20 1.22 5.22 1 1.25 5.25 Category 6.0 Plant Systems Design, Control, and Instrumentation 6.06 Refer to R. O. Licensing Exam 3.04 3.05 6.07 3.09a 6.09a 3.13a and b 6.11a and b 6.22b 2.20b y 2.21 6.23 3.26b 6.25b 9 Page 13
Category 7.0 Procedures - Normal, Abnormal, Emergency, and Radiological Control j 7.01 Refer to R. O. Licensing Exam 4.01 "~- ~ 4.02 7.02 4.12 7.12 7.16 Answers 13 and 14 can be incorporated into one: HPI Forced Cooling Answer 12 One student commented that he understood an examiner's clarification to mean that EFDW from the affected unit was one correct answer and that EFDW from another unit could be counted as another correct answer. Another answer not listed would be ASW from the Auxiliary Building. Reference - EP/1/A/1800/01 EOP Section 502 Loss of Heat Transfer Step 11.0 7.18 Refer to R. O. Licensing Exam 4.17 7.19 4.19 7.22 4.22 Category 8.0 Administration Procedure, Conditions, and Limitations 8.01 Answer C is not a correct answer. Two individuals using Remote Indicator is allowed, while answer d is not allowed and is therefore the correct answer.
Reference:
Station Directive 2.2.2 5.3 8.04 Answer C applies if the work supervisor CANNOT BE LOCATED to retrieve the red tag stub. Answer A could also apply if the work supervisor is not on site. Either answer A or C should be accepted i
Reference:
Station Directive 3.1.1, 6.11.6, and 7.4.3 attached Page 14
8.06 Answer B could also be correct based on the students interpretation of the statement ". . all EFW Flow instrumentation for 'B' OTSG were inoperable." Page 3.4-3 of this Tech. Spec. states that a S/G level indicator is a sufficient flow indication. Theref6re if "all" EFW Flow instrumentation were n inoperable that might include level indication also. If so, answer B is the most limiting LCO and would be a correct answer. 8.13b Refer to R. O. Licensing Exam 4.07b 1 8.14 It is agreed that'the basic premise of the question if for the SRO to know that RIA-45 is required to move fuel. The SRO should also know that if the motor for the pump supplying RIA-43 is de-energized, RIA-45 also becomes inoperable. However, if the SRO states or assumes that RIA-43, and only RIA-43, is taken out of service (by de-energizing the readout module)'then "yes" should be an acceptable answer, since RIA-43 is not required to move fuel. Also, the intent of Specification 3.8.10 is to insure that the Reactor Building Purge is capable of being automatically isolated in the event of a fuel handling accident. There is no requirement that the Reactor Building Purge be in operation during fuel handling operations. If the Reactor Building Purge is secured and the isolation valves PR 2,3,4, and 5 are closed as required by Specification 3.8.7 then the intent of Specification 3.8.10 is met and RIA-45 could be removed from service.
Reference:
Technical Specification 3.8 Analysis of Technical Specification 3.8.7 (Encl. 7.2 of IIR 086-27-2 attached) 8.15 Although Station Directive 2.2.1, paragraph 10.1 does agree the " Procedure Discrepancies Process Record" form OMP 1-9 also addresses alternate locations for documenting procedural discrepancies, such as, R&R book or the " Remarks' section of the procedure process record, These answers should also be acceptable (See Attached sections of OMP l-9). l 8.23 Answer 2 - Number of RCPs operating is addressed in the bases of T.S. 2.1 as RC Flow. Therefore, RC Flow should be an alternate acceptable answer for i number of RCP's.
Reference:
Technical Specification 2.1 Bases Page 15
AttLchment 3 SPECIFIC COMMENTS REGiRDING THE REACTOR OPERATOR REQUALIFICATION EXAMINATION Category 1.0 Principles of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer and Fluid Flow 1.02 Refer to R. O. Licensing Exam 1.02 1.19b 1.11b 1.12 1.22 1.25 1.14 Category 2.0 Plant Design Including Safety and Emergency Systems 2.04 Refer to R. O. Licensing Exam 3.04 3.06 2.05 2.08 2.06 2.14 2.10 2.15 2.11 2.17 2.12 2.13b Refer to R. O. Licensing Exam 2.20b 2.21 2.14 l Page 16 L
Category 3.0 Instruments and Controls 3.04 Refer to R.O Licensing Exam 3.04 3.05 3.05 3.08 3.07 3.09a 3.08a 3.09 The last part of quest. ion 3.09 "The PORV is activated by a pilot valve which is connected to a iclass IE] system" is not operator applicable. Although the statement is, indeed, in the referenced dccument (OP-0C-SPS-CM-PZR, page 14) the information is not relevant to operations personnel, nor is taught them, nor is there any objective requiring an operator to know this. We feel that this last portion of the question should be deleted. 3.13 Refer to R. O. Licensing Exam 3.19 3.20 3.14 Category 4.0 Procedures - Normal, Abnormal, Emergency, and Radiological Control 4.01 Refer to R. O. Licensing Exam 4.02 4.07b 4.03b 4.06 4.12 4.19 4.11 4.13 4.21 Page 17 . ~.
SPECIFIC COMMENTS REGARDING THE SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION Category 5.0 Theory of Nuclear Power Plant Operations, Fluids, and Thermodynamics 5.02 Refer to R. O. Licensing Exam 1.02 5.05 1.11 1.25 5.14 Category 6.0 Plant Systems Design, Control, and Instrumentation 1 6.04 Refer to R. O. Licensing Exam 3.04 ~3.05 6.05 6.08 3.08 2.08 6.09 6.14 Refer to R. O. Requal Exam 2.12 6.15b Refer to R. O. Licensing Exam 2.20b 6.17 2.21 l l l l
Category 7.0 Procedures - Normal, Abnormal, Emergency, and Radiological Control 7.01 Refer to R. O. Licensing Exam 4.02 7.13 Refer to SRO Licensing Exam 7.16 7.15 Refer to R. O. Licensing Exam 4.18 Category 8.0 Admini5tration Procedures, Conditions, and ~ Limitations 8.01 Refer to SRO Licensing Exam 8.04 8.14 8.06 8.15 8.07 8.23 8.13 8.15 Two examinees, Snowden and Chudzik, had already completed their exams and turned them in before it was realized that the necessary Tech. Spec. excerpts had not been included with the exam. They did not have the benefit of referring to the excerpts to answer the question. l l l l 1 l l i u
ENCLOSURE 4 SIMULATION FACILITY FIDELITY REPORT 4 Facility Licensee: Duke Power Company 422 South Church Street Charlotte, NC 28242 Facility Licensee Docket No.: 50-269, 50-270 and 50-287 i 1 Facility Licensee No.: DRP-38, DRP-47 and DRP-55 Operating Tests administered at: Oconee Nuclear Station Operating Tests Given On: July 14-23, 1987 (includes requalification exams) During the conduct of the simulator portion of the operating tests identified above, the following apparent performance and/or human factors discrepancies were observed. 1) Initiation of a safeguards signal cancels almost all component overrides. 2) The ability to stick more than one rod out of position did not exist 1 1 3) The core decay heat model is set too low, such that a loss of all ] feedwater results in minimal adverse effects on the primary reactor plant. 4) The simulator modeling is incapable of correctly modeling the symptoms that should indicate an uncoupling of the primary from the secondary plant l during a loss of natural circulation condition. 5) The porv tailpipe and quench tank temperatures increased with a failed I open porv, even with the block valve shut. I 6) The group seven rod control out limit is still in effect, even though it no longer functions at the plant. 7) A trip of the ITE bus locks open MS-93 (TDEFW Steam Inlet) and it cannot be overridden. 8) When pressurizer level transmitter LT-1 failed, the SPDS showed LT-1 as having failed. 9) Incore thermocouple readings were not available to the operators and had to be provided by instructors in the control booth.
- 10) The reactor building spray throttle valve modeling is inconsistent between the two trains, i
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