ML20206U738
| ML20206U738 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 09/17/1986 |
| From: | Elin J, Meadows T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20206U736 | List: |
| References | |
| 50-397-OL-86-02, 50-397-OL-86-2, NUDOCS 8610080250 | |
| Download: ML20206U738 (108) | |
Text
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Enclosure (1)
U. S. NUCLEAR REGULATORY COMMISSION REGION V EXAMINATION REPORT Examination Report No.: 50-397/0L-86-02 i
Facility:
Washington Nuclear Project, WNP-2 Docket No.:
50-397 Facility License No.:
NPF-21 4
Examinations administered at Washington Nuclear Project, WNP-2, Richland, Washington Chief Examiner:
.CFM # N
>-er-Et-
~
Thomas R. Meadows Date Signed
'7 7[gG Approved by:
n O. Elin, Chief, Operations Section D6te Signed Summary:
Examinations on August 11-14, 1986.
4 Written ~and oral examinations were administered to five SRO candidates and i
five RO candidates. One SRO candidate failed the simulator-portion of the-exam, and one SRO candidate failed both the oral and simulator portions of j
the ex~am.
All of the remaining candidates passed.
1 4
2 l
4 i
1 8610080250 860919 PDR ADOCK 05000397 V
,~
REPORT DETAILS 1.
Examiners:
- Thomas Meadows, RV John Hanek, INEL Merton Bishop, INEL
- Chief Examiner 2.
Persons Attending the Exit Meeting:
T. Meadows, RV, Chief Examiner J. Baker, WNP-2, Assistant Plant Manager R. Corcoran, WNP-2, Plant Operations Manager J. Wyrick, WNP-2, License Training Manager M. Westergren, WNP-2, License Training Supervisor 3.
Written Examination and Facility Review:
Written examinations were administered to five SRO candidates and five RO candidates on August 11, 1986 in the facilities training building.
The facility staff reviewed the exams immediately upon their conclusion.
The comments made by the staff are attached. All comments were resolved.
Appropriate changes were made to the pplicable SRO and R0 exam keys prior to the grading of the exams.
4.
Operating Examinations:
No general training weaknesses were noted during oral examinations conducted August 12-14, 1986.
The plant specific simulator was found adequate to conduct comprehensive NRC exams. However, the simulators performance was marginal, requiring excessive time to complete exam scenarios. The Chief Examiner met with the facility training staff to develop specific objectives that will achieve more efficient simulator examinations. The facility staff committed to meet the intent of these objectives within the next 2 or 3 exam cycl.es.
I 5.
Exit Meeting:
On August 14, 1986, the Chief Examiner met with the licensee representatives listed in paragraph 2.
The marginal simulator performance, and appropriate improvement action noted in paragraph 4 was discussed.
l 1
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i ATTACHMENT 1 WNP-2 SRO-EXAM (AUGUST 11, 1986)
FACILITY COMMENTS /NRC RESOLUTIONS NOTE:
Facility comments were submitted on a copy of an exam key provided for this use.
This key is attached.
5.10 Comment:
Question is confusing in that the candidates are trained that MAPRAT is only a computer name for the thermal limit APLHGR.
Resolution: Comment not accepted. The candidate must know the relationship between MAPRAT and MAPLHGR in order to properly interpret the computer print out.
The answer key is in agreement with the reference provided and will stand as is.
6.5 Comment
This question is misleading - it implies that this is the preferred ECCS injection flow path. The question should have asked:
"What are two (2) conditions when this flow path is preferred?" The candidates will be confused about which-set of RHR valves your talking about:
RHR-V-42A. and B; LPCI injection j
or RHR-V 53A and BTRHR S/D cooling Resolution: Comment accepted; however, this question was accurates based on training material provided. Since the facility training staff has committed to upgrade their training material to accurately reflect actual system logic, the exam key will be modified to accept other possible answers concerning direct core ECCS injection for full credit.
6.10 Comment:
The answer key should be clarified such that any one of the following conditions would be. acceptable for full credit of the applicable portion of the answer:
(a) Initial core loading (b) SDM testing (c) Various testing during refueling Resolution: Comment accepted. This was the intent of the answer key. Subsequently, the answer key will be modified to emphasize this point.
o 6.11 Comment:
Question is slightly confusing in that some of these curves do not have a " reason or basis", but merely identify a given power / flow condition.
Resolution: Comment not accepted. The term " reason" adequately covers this contingency, and this question is consistent with the training material provided.
7.6
' Comment:
The answer should indicate that " negative design containment pressure" is the bottom line answer.
Resolution: Comment not accepted. This is the intent of the answer key, and is worth the majority of the available points; however, the candidate should mention somethla.; about containment vacuum breakers rince this is the reason
- hat they were installed.
7.7 Comnent
Since " adequate core cooling" cannot be directly measured by the operator (cladding temperature 2200 degrees F.), he must " infer" adequate core cooling by the following methods:
(a) RPV level TAF (b) One ECCS injecting at full flow rate (c) Steam cooling of RPV in progress This is consistent with the reference material provided.
Resolution: Comment accepted. The answer key will be modified to include this other acceptable answer.
8.1 Comment
Answer key should specify the bottom line answer - one required for full credit.
Resolution:
Comment accepted. Answer key will be clarified.
8.4 Comment
Parcs (f) and (g) of the answer key is not part of technical specifications shift manning table, therefore, should not be required for full credit.
Resolution:
Comment not accepted. The table requirements for shift manning are only part of the answer. The question clearly stated, "per technical specifications".
8.5 Comment
Identified areas of the key should not be required in the answer.
Resolution:
Comment accepted. This was the intent of the key. The i
answer key will be corrected to clarify this intent.
l
ATTACHMENT 2 FACILITY COMMENTS /NRC RESOLUTIONS WNP R0 EXAM GIVEN ON
' AUGUST 11, 1986 J
NOTE: Facility comments were submitted on a copy of an exam key provided for this use. This key is attached.
1.04 Comment:
The word physical may confuse candidates.
Could possibly get answers pertaining to " physical" properties (FW pump, fission product buildup, core' aging effects, core flow, etc).
Resolution: Facility comment not accepted. Question specifically asks for fuel characteristics with increasing fuel temperature that' contribute to a negative doppler co-efficient.
Answer key is correct in this-regard.
1.11 Comment:
Question is confusing, MAPRAT is only a computer name for the thermal limits APLHGR. Also MAPRAT is always maintained 1 1.0 not > 1.0.
a.
Could get > 1,0 depending on how the question is read.
b.
Or MAPRAT is the computer calculation used to identify where the APLHGR thermal limit may be exceeded.
Resolution: Comment accepted; Question 1.11 facility comment is valid MAPRAT is not a thermal limit, and is always maintained
$ 1.0.
Answer 1.11a will be changed to < 1.0.
Answer 1.11b facility comment is valid. Question 4
specifically asks for the relationship between MAPRAT and MAPLHGR, additional facility answer will be accepted.
1.16 Comment:
Additional answers for recirculation flow, (FCW position, pump speed) reactor power (rod position) FW temp (FW heater lineup) should be included in answer key.
Resolution: Comment accepted.
Facility comment is valid additional answers will be added to the answer key.
2 1
2.02b Comment:
The following statement (flux shapes at the end of cycle are such that the) should not be required for full credit.
Resolution: Facility comment not accepted. This answer is verbatim from the recirculation system lesson plan. The important concept and key phrase here is " flux shapes." End of cycle was already stated in the question and is included only for sentence structure.
2.02c Comment:
Technical Specification Table 3.3.4. 2-1 note (b) states this function shall be automatically bypassed when turbine first stage pressure is less than or equal to 165 psig equivalent to thermal power less than 30% of rated thermal power.
(142 psig also accepted)
Resolution: Comment accepted. Technical Specification Table 3.3.4.2-1 note (b) is correct for this answer. Answer key was taken from reactor recirculation lesson plan page 31.
Also, annunciator response procedures 4.603.A7-5.4 and 4.603.A8-5.4 give setpoints of less than 30% equivalent reactor power and 1st stage pressure less than 108.5 psig.
The answer key will be changed to <165 psig (or <145 psig)
Ist stage pressure or <30% rated thermal power.
f 2.09 Comment:
Answer key is vague as to point breakdown / assignment.
Resolution:
Comment accepted. Answer key adjusted to accept credit for system operational function.
2.10a Comment:
Question referenced PPM 5.1.3, procedure answer was as stated in lesson plan for SLC.
Resolution:
Comment accepted. Answer key will be changed to "If the reactor cannot be shutdown [0.5] before suppression pool I
temperature reaches 110*F (0.5).
Point values will be adjusted as indicated. PPM 5.1.3 page 2 of 9 will be 1
included in reference.
2.11a Comment:
Off-gas glycol cooling system is not actually an off gas a
component.
i Resolution: Comment accepted.
2.11.a.13 will be deleted from the question and 2.11.a.10 from the answer key.
Point breakdown will be changed.
2.llb Comment:
CAS system is also used to " warm-up" the system.
Resolution: Comment accepted.
PPM 2.6.4 step 29 page 9 of 48 states
" Admit service air to the preheater selected for service by throttling open service air valve 06-V-4A(B) locally until 64 SCFM is indicated on SA-FI-67." Warmup the system will be included in the answer key.
i e
- - ~ -
c.
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3 3.01d Comment:
Two minute period should be nine minute period.
Resolution:
Comment accepted. Reference, PPM 3.3.1 reactor scram indicates nine minute period is correct. Two minutes is an error in the system training manual. Answer key changed to nine minute period. PPM 3.3.1 reactor scram added to reference.
3.03c Comment:
How much droop is inserted is outside the scope of ES-202. Why droop inserted much more important. Exact number not necessary.
Resolution:
Comment accepted. Answer key adjusted.
3.04b Comment:
Answer should also include SRM > 100 cps.
Resolution:
Comment accepted.
SRM > 100 cps will be added to the answer.
3.05 Comment:
Fuel load dunking chambers are no longer used. Delete question and answer.
Resolution: Facility comment not accepted. Technical Specification 3.9.2 allows the use of special movable detectors during I
core alterations in place of the normal SRM nuclear detectors.
3.07a Comment:
APRM FLOW BIAS OFF-NORMAL annunciator terminology may be confusing, actual name is APRM flow converter-INOP.
Resolution:
Facility comment noted but not accepted. Page 21 of the APRM systems training lesson plan describes this function ~
as the APRM FLOW BIAS OFF-NORMAL".
3.07c Comment:
Flow computer should be flow comparator.
Resolution:
Comment accepted.
Typographic error. Word computer changed to comparator in question 3.07.c.
3.07c Comment:
Question did not specifically ask for second part of answer, therefore, it should not be required for full credit.
Resolution: Facility comment not accepted. Question was specific in that it asked what would occur if " flow comparator A" was bypassed.
4 3.09b Comment:
Question did not specifically ask for second part of answer. Therefore, it should not be required for full credit.
Resolution:
Comment accepted. Although the answer key is verbatim from the system training lesson plan, dead end is implied to mean a no flow condition.
Therefore, either answer will be accepted for full credit.
3.11 Comment:
Recirculation flow controllers are not operated in auto, thus part a.1 is not really applicable.
There was no reference supplied with this question and answer.
Resolution: Comment accepted. Answer key adjusted to accept alternate answer.
~
Reference WNP-2 System Training, Recirculation Flow Control page 31, has been added to the examination.
Also, PPM 3.12 reactor plant cold S/V. PPM 9.3.12 power plant maneuvers added to reference.
4.08a Comment:
Question 4.08.a is outside the scope of E-S-202.
Resolution: Facility comment not accepted. This question is consistent with the requirements of ES-202 with regards that'the candidate should be able to explalia teasotis, cautions, and limitations of normal operating procedures. This question is not a normal procedure step but is a cautionary note contained in the procedure.
4.09a Comment:
Question 4.09.a is outside the scope of ES-202.
Resolution: Facility comment not accepted. This question is consistent with the requirements of ES-202 with regards that candidates should be able to explain reasons, cautions and lisaitations of normal operating procedures. This question is not a normal procedure step but is contained in precaution and limitations 3.3.1.5.F of the reactor scram procedure.
4.10 Comment:
Answer should be either Increase or Remain the Same because HPCS flow is small compared with FW flow at this power level. Also, HPCS flow tends to be " swept upward" due to steam velocity at core exit.
Resolution:
Comment accepted. Answer key adjusted to accept comment.
M ff C EXAM A*fP f c/
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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
WNP-2 REACTOR TYPE:
BWR-GE5 DATE ADMINISTERED: 86/08/11 nrsinst
~ Q y g g (* g p y '
EXAMINER:
HANEK. J.
p/ygg-CANDIDATE:
INSTRUCTIONS TO CANDIDATE:
Uso separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each quastion are indicated in parentheses after the question.
The passing drade 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 VALUE TOTAL SCORE VALUE CATEGORY 2 r. r'/
25.00
- 25. CCE 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW zi. W 25.00 05.007-2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS z r. c +-
25.00 25.00 -
3.
INSTRUMENTS AND CONTROLS 22efd 23.08
- 2 5. 0iu> - - 2 ". CP 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL
'^
,,, 7, CONTROL 9 7. ro
-100. # 'r-Totals Final Grade All work done on this examination is my own.
I have neither given nor received aid.
Candidate's Signature
t 2 -
o ES-201-2 i
ATTACHMENT 2 REQUIREMENTS FOR ADMINISTRATION OF WRITTEN EXAMINATIONS 1.
A single room shall be provided for completing the written examina-tion. The location or tnis room e d supporting restroom facilities shall be such as to prevent contact with all other facility and/or contractor personnel during the duration of the written examination.
If necessary, the facility should make arrangements for the use of Ob-a suitable room at a local school, motel, or other building.
taining this room is the responsibility of the licensee.
2.
Minimum spacing is required to ensure examination integrity as determined by the chief examiner. Minimum spacing should be one candidate per table, with a 3-ft space between tables.
No wall charts, models, and/or other training materials shall be present in the examination room.
3.
Suitable arrangements shall be made by the facility if the candi-dates are to have lunch, coffee, or other refreshments.
These arrangements shall comply with Item 1 above. These arrangements shall be reviewed by the examiner and/or proctor.
i 4.
The facility staff shall be provided a copy of the written examination and answer key after the last candidate has completed and handed in his written examination.
The facility staff shall then have five working days to provide formal written comments with supporting documentation on i
the examination and answer key to the chief examiner or to the regional j
office section chief.
3 5.
The licensee shall provide pads of 8-1/2 by 11 in. lined paper in unopened packages for each candidate's use in completing the exam-ination. The examiner shall distribute these pads to the candidates.
)
All reference material needed to complete the examination shall be l
furnished by the examiner.
Candidates can bring pens, pencils, calculators, or slide rules into the examination room, and no other equipment or reference material shall be allowed.
4 6.
Only black ink or dark pencils should be used for writing answers to questions.
l Examiner Standards 11 of 18
s.
D NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
.Dur*ng 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 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.
4 3.
Use black ink or dark pencil onlY to facilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
4 6.
Use only the paper provided for answers.
i 7.
Print your name in the upper right-hand corner of the first page of each section of the answe r sheet.
8.
Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a naw Page, write onlY gg QB1 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.
1
{
- 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.
j
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
i
- 16. If parts of the examination are not clear as to intent, ask questions of j
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 j
completing the examination.
This must be done after the examination has been completed.
i i
)
- 18. When you complete your examination, you shall:
a.
Assemble your examination as follows:
(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.
Turn in all scrap paper and the balance of the paper that you did c.
not use for answering the questions.
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.
l l
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EQUATION SHEET f = ne v = s/t 2
Cycle efficiency =
- ' "' I '
w = ag s = v,t +
at Energy (in)
E = aC a = (vg - v,)/t KE = 1 rv A = AN g = v, + a t A = A,e p
v PE = agh w = e/t A = In 2/tg = 0.693/tg W = VAP 4
g(eff) = (t,,)(ts) t AE = 931Am
(
+
)
Q=[ncAT I = I IX
~
P o
Q = UAAT I = I. UX
~
Pwr = W m
-X/ M
~
g I=I lo SUR(t),
IyL = 1.3/u y=p to y=p.t/T HVI. ' O.693/u o
~SUR = 26.06/T
~
T = 1.44 DT SCR = S/(1 - K,gg) fa ge \\
o SUR = 26 CR
= S/(1 - K,gg )
6-p x
1(
- ff}I CR Cl ~ Keff)2 T = (1*/o ) + [(f _* p)/A,g,p}
~
2 y = g*j (, _ g M = 1/(1 - K,gg) = CR /CR g
0 I " II ~ 8)! A 8
eff M = (1 - Keff)0/II ~ Eeff)1 p = (K,gg-1)/K,gg = AK,gg/K,gg j
[t*/TKygg.] + [I/(1 + 1,gg )]
1* = 1 x 10 seconds
~
p=
T P = I4V/(3 x 1010)
A,gg = 0.1 seconds A
I = No Idgg=1d22 UATER PARAMETERS Id =10 g
22 2
1 gal. = 8.345 lba R/hr = (0.5 CE)/d g,,C,,,)
I gal. = 3.78 liters R/hr = 6 CE/d (feet) 3 1 ft = 7.48 gal.
HISCEI.I.ANEOUS CONVERSIONS 3
Density'= 62.4 lbm/ft 1 Curia = 3.7 x 10 dps 10 3
Density = 1 gm/cm 1 kg = 2.21 1he Heat of vato.rizations = 970 Etu/lbm I hp = 2.54 x 103 BTU /hr 0
Heat of fusien = 144 Btu /lba 1 N = 3.41 x 10 Btu /hr 1 Atm = 14.7 psi = 29.9 in,l's.
1 BLu = 778 f t-lbf 1 ft. H O = 0.4333 lbf/in 1' inch = 2.54 cm 2
F = 9/5 C + 32
- C = 3/9 (*F - 32)
1'. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE 2
TMRMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.01
(.50)
As a subcritical reactor nears criticality, the length of time to reach cquilibrium count rate after an insertion of a given fixed amount of positive reactivity...
(0.5)
(SELECT THE CORRECT ANSWER) decreases primarily because of the increased population of delayed a.
neutrons in the core, b.
increases primarily because of the increased population of delayed neutrons in the core.
increases because of a larger number of neutron life cycles required c.
to reach equilibrium.
d.
decreases because the source neutrons are becoming less important in relation to total neutron population.
QUESTION 1.02 (1.50)
In Keff DEPENDENT or INDEPENDENT of initial source range counts (source rcnge neutron population)?
(0.5)
BRIEFLY EXPLAIN YOUR ANSWER (1.0) i QUESTION 1.03 (2.00)
Answer the following TRUE or FALSE:
a.
During equilibrium power conditions, the production rate of indirect Xenon from Iodine is faster than the decay rate of Xenon to Cesium.
(0.5) b.
Slowing the rate of a power decrease, lowers the height of the resultant Xenon peak.
(0.5) c.
The resultant Xenon peak due to a scram from 50% power is larger than one from 100% power.
(0.5) d.
During an increase in power from equilibrium Xenon conditions, Xenon concentration initially decreases.
(0.5)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
i m
l l '. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE 3
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW fA,i codsf C4OATC.f. U,/b%'blf fe7 MSW>ts* ffA1A5'A2 j
.rc '/ ysiN/ ' fMf?fA1!'t1 (% int, Cassa pducr Li, Mup, core nyrivy d
effects, cme fby etc.),
QUESTION 1.04 (1.00)
What TWO (Tiysica) phenomena (specifically related to the neutron life cycle and fuel characteristics) along with increasing fuel temperature, contribute to a negative doppler coefficient?
NAME THESE TWO (2)
PHENOMENA [0.5 each].
(1.0)
QUESTION 1.05 (1.50)
If the reactor power level is increased on a positive period from 50 MW to 370 MW in two minutes, what is the doubling time?
(Show all uork.)
(1.5)
QUESTION 1.06 (3.00)
Assume that the reactor is being started up with the bulk coolant temperature less than the saturation temperature.
Excessive rod withdrawal ccuses the reactor to increase in power on a short period.
Of the void, doppler, and moderator temperature coefficients, which will come into offect first, second, and third?
BRIEFLY EXPLAIN!
(Assume no operator action to stop the power increase.)
(3.0)
QUESTION 1.07 (1.00)
During a rapid power increase, very short periods can be maintained, yet for rapid power decreases, the period quickly becomes -80 sec.
Explain the reason for the difference.
(1.0)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
..-s
1.
' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE 4
TMRMODYNAMICS. HEAT TRANS m AND FLUID FLOW QUESTION 1.08 (2.00) 4 The reactor is started up after a refueling outage.
Rods are pulled to the 100% line and power is then increased to 100% with recirculation flow.
After approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, reactor power has decreased to about 984.
Assume no operator action.
n.
What is the primary cause for this reduction in power?
(1.0) b.
When whould you expect the power decrease to stop and WHY does it stop?
(1.0)
QUESTION 1.09 (1.50)
With the reactor operating at 50% power and 50% rated flow, the flow is J
increased to 70% of rated.
Briefly explain the effect on power level.
Include Void and Doppler offects and the effect on boiling boundry.
(1.5)
QUESTION 1.10 (1.50) a.
In what direction does the value of Beff change over core life (INCREASE, DECREAE)?
(O.5) b.
Why does the value of Beff change over core life?
(1.0)
QUESTION 1.11 (1.50) a.
Is the thermal limit for MAPRAT less than or greater than 1.07 (0.5) b.
What is the relationship between MAPRAT and MAPLHGR7 (1.0)
A %4097 U ""l/ A ' T"ide ^*"f SR Tbd Cowhsiiyv :
rdre.e} 4W: th"D/6f.
' AO.I A/ss M A "/ A 7 e A do nj e rid /Ive/ I d S l O
^'07 j
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
' l '. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE 5
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW l
QUESTION 1.12 (1.00)
Consider a real plant system (NON-IDEAL) with two centrifugal pumps in parallel, one of which is running at 1800 RPM.
The second pump is started and run at 1800 RPM.
System flow will be...
(1.0) 1 SELECT THE CORRECT ANSWER a.
more than double the original flow due to decreased flow resistance.
b.
slightly less than double the original flow due to increased flow resistance.
c.
the same since only the discharge head changes.
reduced by one-half due to increased discharge head.
d.
QUESTION 1.13 (1.00)
Of the following operations, which ONE will have a negative effect (reducing effect) on available Net Positive Suction Head (NPSH) of a given esntrifugal pump:
(1.0)
SELECT THE CORRECT ANSWER l
a.
Throttling open the pump's suction valve.
b.
Throttling open the pump's discharge valve.
c.
Decreasing the pump's speed.
d.
Decreasing the temperature of the fluid (water) being pumped.
QUESTION 1.14 (1.00)
There is 15 feet of water in the CST whose base is at ground level.
The HPCS pump is forty feet below grade.
Assuming no friction effects, datermine the pressure (psig) at the HPCS suction.
Show all work.
(1.0) i I
l I
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
- '1 '. 'FRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE G
TIERMODYNAMICS. HEAT TRANSnx AND FLUID FLOW QUESTION 1.15 (1.50)
The term " critical power" refers to that bundle power level corresponding to the onset of transition boiling (OTB) somewhere in that bundle.
State how critical power varies (ie. INCREASES, DECREASES, or IS NOT AFFECTED) by each of the following:
(1.5) c.
If coolant mass flow rate increases b.
If reactor pressure increases c.
If inlet subcooling increases QUESTION 1.16 (3.50) a.
List FOUR parameters which contribute to AVAILABLE NFSH for a recirculation pump.
Limit your answer to those parameters which are directly available in the control room.
(2.0) b.
Consider two RPV conditions:
low power and low flow (<10%) OR high power and high flow (>85%).
1.
During which condition is REQUIRED NPSH for a recirculation pump greater?
(0.5) 2.
During which condition is AVAILABLE NPSH for a recirculation pump greater and WHY is it greater?
(1.0) l
(***** END OF CATEGORY 01 *****)
, - ' 7:
- Pr. ANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7
QUESTION 2.01
(.50)
At High Reactor pressures (1000 psig), what specific component.n the CRD mechanism allows vessel pressure to assist in the scram process for an individual rod.
(0.5)
QUESTION 2.02 (2.50)
Concerning the Recirculation Pump Trip (RPT) circuit, provide the following:
a.
What action occurs following an RPT signal (specific to the recirc system)?
(0.5) b.
Why is it more important at EOL than BOL?
(1.0) c.
When is it automatically bypassed?
(0.5) d.
How and where is it manually bypassed?
(0.25) o.
Is the logic (choose one)
(0.25) 1.
1 out of 2 2.
2 out of 3 3.
1 out 2 taken twice 4.
2 out of 4 I
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
I
^
m r.y amy
2 '. ' PLANT DESIGN INCLUDING SAFETY AND EMERCENCY SYSTEMS PAGE 8
QUESTION 2.03 (2.50)
Concerning the Nuclear Steam Supply Shutoff System (NSSSS), Match the c.
following group number with the below brief descriptions of the Isolation Groups.
(1.4)
GROUP DESCRIPTION 1
a.
Reactor water sample valves 2
b.
c.
d.
MSIV's and main steam line drains 5
e.
Primary and secondary containment ventilation and purge systems 6
f.
RHR (shutdown cooling mode) 7 g.
Miscellaneous balance of plant (RCCW, fuel pool cooling, circ water, ect.)
[7 9 0.2 each]
b.
From the following list, select which NSSSS pushbutton or combinations of pushbuttons must be pushed to cause the following actions:
(1.1) 1.
All MSIV's to close and an inboard isolation of Groups 1, 2,
5, 6, and 7.
x 2.
Trip (close) all MSIV's.
3.
Inbd NSSSS Groups 1 (except MSIV's), 2, 5, 6, and 7.
4.
Outbd NSSSS Groups 1 (except MSIV's), 2, 5, 6 and 7.
5.
All MSIV's to close and an outboard isolation of Groups 1, 2, 5, 6, and 7.
NSSSS PUSHBUTTON OR COMBINATIONS B
C and D.
A and B A or C AND B or D D
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9
QUESTION 2.04 (1.00)
FILL IN THE BLANKS.
The leak detection system sensitivity and response time is such that an unidentified leakage rate increase of in less than will be detected.
(1.0)
QUESTION 2.05 (2.00) c.
What are the High flow alarm setpoints from the drywell equipment and floor drain sumps?
(1.0) i l
b.
What TWO components generate the High Flow Alarms?
.(1.0)
QUESTION 2.06 (1.75) a.
List THREE different components that are most likely to cause radioactive leakage into the RCCW system.
(0.75) b.
What are TWO alarms that would indicate radioactive Inleakage?
(0.5) c.
What automatic action could occur ir: the event of radioactive inleakage?
(0.5)
QUESTION 2.07
(.50)
Although the HPCS can serve as a backup for the RCIC system, why is its use in this capacity not recommended unless required by plant conditions? (0.5)
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
'2.
'FLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10 QUESTION 2.08 (2.00)
What affect would closing the RCIC exhaust line isolation valve RCIC V-68 have under the following conditions:
Explain your answer.
(Confine your answer to the RCIC system only.)
i Just prior to receiving an automatic initiation signal.
(1.0) a.
b.
After receiving an automatic initiation signal (assume RCIC is injecting).
(1.0)
QUESTION 2.09 (2.00)
Draw a one line sketch of the pneumatic supply system to a typical ADS valve.
Indicate the sources of the pneumatic supply, power supply, and normal state (open closed energized deenergized of any valves).
(2.0)
QUESTION 2.10 (3.00)
In accordance with ppm 5.1.3, Reactor Power Control, SLC injection n.
will be initiated when what plant conditions exist?
(1.0) b.
List FOUR automatic actions that occur when the keylock switch is turned to System "A".
(2.0)
.\\
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
- ~ 2.
- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 11 QUESTION 2.11 (3.25)
Concerning the off-gas system.
a.
List the following items in the correct order of system flow path.
from the second stage air ejector to the elevated release point.
(2.1) 1.
Off-gas condenser 2.
Off-gas catalytic recombiner 3.
Off-gas cooler condensers 4.
Off-gas moisture separators 5.
Off-gas preheaters 6.
Off-gas water separator 7.
Off-gas gas prefilters 8.
Off-gas gas coolers 9.
Off-gas hydrogen analyzers.
10.
Off-gas after filters 11.
10 minute holdup
'12.
Off-gas charcoal adsorbers
-13.-Of t-see 1 glyco 1-cooling system per scran//y su off-y cmpwen.
14.
Off-gas desiccant dryers b.
Briefly describe the interrelationship and function provided to the off-gas system by the CAS system.
(1.15)
E Qwo-]t QUESTION 2.12 (3.00)
Provide a brief description of the THREE different water supplied sprinkler cystems used in the Fire Protection Systems.
Include in your answer the cetions that must occur in the system to start water flow to a fire.
(3.0)
QUESTION 2.13
(.50)
TRUE of FALSE?
Each DC system has sufficient capacity to supply the DC load requirements of its respective division under conditions of a DBA for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> without battery chargers.
(0.5)
-1
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
~
2.
' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 12 i
l QUESTION 2.14
(.50)
TRUE or FALSE?
)
The main generat,or reverse power relay 67G provides the protection for the turbine and power system, not the generator.
i i
i
(***** END OF CATEGORY 02 *****)
3.
INSTRUMENTS AND CONTROLS PAGE 13 QUESTION 3.01 (2.75)
Concerning the "setpoint setdown" circuit in the feedwater level control
- circuit, a.
What is the function?
(0.75) b.
How is the actuation signal for the system sensed?
(0.5)
List TWO automatic actions that occur when the system is c.
actuated.
(1.0) d.
What action occurs when the system is reset (Include Rate Time Frame)?
(0.5)
QUESTION 3.02 (1.75)
Concerning the feedwater level control system:
a.
What effect will a loss of one steam flow transmitter (fails downscale) have on the following indications (DECREASE, INCREASE or REMAIN THE SAME).
(1.5)
Consider BOTH transient and final conditions.
Transient Final 1.
Reactor vessel level 2.
Feedwater flow 3.
Total indicated steam flow b.
Will this failure cause a reactor scram when operating in three element control at 100% power?
(0.25)
QUESTION 3.03 (2.00)
Concerning the HPCS Diesel Generator Governor Droop Switch (DROOP / ISO).
a.
Provide a brief description when each position is used.
(1.0) b.
What is the normal setting of this switch.
(0.25) c.(Howmuch" DROOP"isinsertedand)whyisitcontrolled?
(0.75) can Wt.rcope d fS'202.?
way p ros/* jn r risarco... n ucy j
n ons-se on ts" 7==y exs u M m r seeussar,,,
(/
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
'"' ' =h M
3 '.
INSTRUMENTS AND CONTROLS PAGE 14 QUESTION 3.04 (2,50)
Concerning the SRM retract permissive interlock, What are TWO conditions that generate this Interlock?
(0.75) a.
b.
List THREE manual and automatic conditions that bypass this Interlock.
(1.5)
Will this Interlock stop detector movement when actuated?
(0.25) c.
QUESTION 3.05 (2.75)
Refer to the attached Figure 1 of Pulse Height vs. Applied Voltage.
a.
Identify the following areas (Region).
(1.0) 2 3
5 6
b.
In what area is each of the following detectors operated.
(Identify by number.)
(1.75) 1.
SRM 2.
IRM 3.
Fuel load dunking chamber 4.
Area radiation monitors (ARMS) 5.
Main steam line rad monitors 6.
Portable frisker station 7.
Dose rate setting portable instrument (CUTIE PIE or RAD OWL)
QUESTION 3.06 (1.00)
FILL IN THE BLANKS.
Transition from startup to run must be made with all APRMs between
% and.
% power to prevent a rod block.
(1.0)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
9 O
~O PULSE HEIGHT VS.
~
APPLIED VOLTAGE ZERO GAS l
GAS
/
g g/
g AMPLiflCATION l
AMPLIFICATION g
4 kj4 y
]
10" l
1 I
I i
i 1080 l
l 1
I 108 l
l I
l l t l
l 1E I
l l
Pu 10' I
b I
IR I
I i
i uj 10' I
M 3
l l
W i
i l
l l
I 8
105 - s I
l g
g i
l
\\ I5l l'* ^a li z
5 l
E I
l a
,0 i
i i
I i
I i
102 i
l I
I l
1 I
i I
I l
102 l
l l
l
'l l
i l
r
,0, l
i I
I i
i 100 I
I I
I I
i i
I I
N i
i l
i I
i M
O APPLIED VOLTAGE h
PULSE HEIGHT VS. APPLIED VOLTAGE FOR GAS FILTER DETECTORS i
icune /
i
'3.
' INSTRUMENTS AND CONTROLS PAGE 15 btN'gg\\ d\\dS) l fg pW jfn[i
~
f04 ub j c 9.
W QUESTION 3.07 (2.25)
List FOUR conditions that could actuate the APRM flow bias 3 M o.
off-normal annunciator.
Include applicable setpoints.
(1.0) b.
What protective function (s) would occur as a result of this annunciator?
(0.25) c.
What effect on system operation would occur if flow
~cc.T.;ut:r A is bypassed?
(Limit your response to the APRM C ' flow computer-)
(1.0) commroe J QUESTION 3.08 (2.50)
Concerning the RBM Gain Change Circuit.
a.
What is the purpose of the gain change circuit?
(0.5) b.
List TWO reasons for increasing the gain of a RBM channel.
(1.0)
What are the reference APRM inputs to RBM Channels A and B c.
(normal and alternate)?
(1.0)
QUES TION 3.09 (2.00)
Concerning Excess Flow Check Valves.
Q.
What is their function?
(0.5) b.
What type of lines are they installed in.
(0,5) c.
What is the function of the bypass valve?
How is an excess flow check valve reopened following an automatic initiation?
(0.5) d.
Where is excess flow check valve position displayed.
(Be specific, area and panel.)
(0.5)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
3.
' INSTRUMENTS AND CONTROLS PAGE 16 QUESTION 3.10 (1.50)
When operating at 100% power, the wide range level instrument reads 15-20 inches lower than the narrow range indicator.
a.
Is this mismatch a concern for the operator?
Explain your (0.75).
answer.
b.
Can the wide range level instrument be used to accurately indicate level at TOP of Active Fuel (TAF).
Explain your (0.75) answer.
QUESTION 3.11 (2.00)
Concerning the recire flow control system.
a.
List THREE automatic actions that occur upon receipt of a High Drywell Pressure signal (>1.68 psig).
(Limit your response to the recire flow control system.)
(1.5) b.
What is the basis for these automatic actions?
(0.5)
QUESTION 3.12 (2.00)
During normal operation of the TIP SYSTEM, c.
List two automatic actions initiated by a HIGH drywell pressure signal (>1.68 psig).
(Limit your response to the TIP system.)
(1.0) b.
List two indications that the TIP operator could use to verify that the automatic actions had taken place.
(1.0)
(***** END OF CATEGORY 03 *****)
i
- 4.
' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 17 RADIOLOGICAL CONTROL QUESTION 4.01 (1.00)
What guidelines must be used by an operator prior to securing or placing en ECCS in manual?
(1.0)
QUESTION 4.02
(.50)
What actions are required if an individuals TLD is lost or damaged?
(0.5)
QUESTION 4.03 (1.00)
Under what TWO conditions may the operator deviate from an approved procedure?
(1.0)
QUESTION 4.04
(.50)
What action is required for the following condition concerning the rGactor recirc pumps during normal power operation?
Loss of both seal injection flow and RCCW seal cooling flow.
(0.5)
QUESTION 4.05 (1.50) a.
Following a recire pump trip, the plant operating procedure directs the operator to close the recire pump discharge valve and then reopen it?
What is the basis for this action?
(1.0) b.
When is it reopened?
(0.5) l
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
4.
' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY A@
PAGE 16 RADIOLOGICAL CONTROL QUESTION 4.06 (2.00)
When withdrawing control rods, Plant Startup Procedure 3.1.2 requires a valid coupling check be performed.
Explain how a valid coupling check is made.
(1.5) a.
b.
What indication will tell the operator the rod is uncoupled?
(0.5)
QUESTION 4.07
(.50)
TRUE or FALSE?
With the RSCS inoperable, whenever power is less than or equal to 20% rated thermal power, normal control rod movement is permitted if the RWM is operable.
(0.5)
QUESTION 4.08 (2.00)
FILL IN THE BLANKS.
ours/WF cope of ES-D2 (Anackdb a.
When forced cooling is not available in cold shutdown condition, reactor water level shall be maintained at or above If for any reason water level cannot be maintained above this level in this condition, reactor vessel temperature shall be monitored once per (1.5) b.
Why is the vessel water level raised in this condition?
(0.5)
QUESTION 4.09 (2.00) a.
List THREE parameters or actions that the operator should check f
/
or preform prior to restarting the condensate booster pumps I
following a scram.
[3 0 0.5 each]
(1.5) b.
Why are these actions necessary?
(0,5)
Ax, n.sans 70 4 sawye rAe scope of a.202 bdd)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
' 4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 19 RADIOLOGICAL CONTROL QUESTION 4.10 (2.75)
While operating at 100% rated thermal power, a HPCS actuation occurs.
How will the following parameters respond to the automatic actions c.
of the plants control systems (INCREASE, DECREASE or REMAIN THE (1.0)
SAME)?
1.
Feedwater flow 2.
Reactor vessel level 3.
Steam flow
,4.
Thermal power b.
What are the THREE immediate operator actions required per Procedure 4.4.4.2, Inadvertent HPCS Startup?
(1.75)
QUESTION 4.11
(.50)
TRUE or FALSE?
'HPCS and RCIC suctions will automatically transfer from the suppression pool to the CST on high suppression pool level.
(0.5)
QUESTION 4.12 (2.75)
List the FIVE entry conditions for RPV pressure control (RPV/P).
Include applicable setpoints.
(2.75)
QUESTION 4.13 (1.50)
Define adequate core cooling, as used in the Emergency Procedures and Standing Operating Orders.
(1.5)
)
l QUESTION 4.14 (1.50)
In order of preference, what are THREE viable mechanisms which can assure cdequate core cooling is achieved?
(1.5)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
i
- 4.
' PROCEDURES'- NORMAL. ABNORMAL. EMERGENCY AND PAGE 23 RADIOLOGICAL CONTROL
'l QUESTION
.15 -
(2.50) g g
g g, g ;3
/
O uT S t
%A7 5 A.
\\
- a. 'While conducting a ull flow survei ance t e RCIC system the
)
operator shall verify *that the suppress on pool temperature is less than a.1.
degt s F at st once per a.2.
s (1.0) minutes.
{
In accordance with Tech pecs.,
AT actions are required if suppression pool te ature reache 110 degrees F?
(1,5
/
y
~
QUESTION 4.16 (2.50)
Control Room Evacuation Procedure 4.12.1.1 discusses the immediate actions corried out by each of the reactor operators in the control room if an evacuation is ordered.
List FOUR immediate operator actions that would be carried out by the c.
operator assigned to the " Reactor Controls".
(2.0) b.
Who has the authority to order an evacuation of the control room? (0.5) t
(***** END OF CATEGORY 04 *****)
(************* END OF EXAMINATION ***************)
~
- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, FACE 21 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- WNP-2
-86/08/11-HANEK, J.
MASTER COPY
^^"#
9 #" ' '
- nt(ItW LUrY,5 ANSWER 1.01
(.50)
(0.5)
?/syIt c
REFERENCE WNP-2, Reactor Theory-Student Text, Ch. III, P. 20 Ch. VI, PP. 1 and 2 ANSWER 1.02 (1.50)
INDEPENDENT (0.5)
Koff (the neutron life cycle) only considers fission neutrons in the self-sustaining reaction.
(1.0)
REFERENCE Reactor Theory - Student Text, Ch. VI, P. 2
)
ANSWER 1.03 (2.00) a.
TRUE b.
TRUE c.
FALSE d.
TRUE
[4 9 0.5 each]
(2.0)
REFERENCE WNP-2, Reactor Theory - Student Text, Ch. V, PP. 1-4 ANSWER 1.04 (1.00) 1.
Doppler Effect (resonance broadening-neutron absorbtion at other than descrete energies).
(0.5) 2.
Self-Shielding (heterogeneous fuel pins).
(0.5) l REFERENCE WNP-2, Reactor Theory Student Text, Ch. IV, P. 9 l
C PRINCIP N OF NUCLEAR POWER PLANT OPERATION.
PAGE 22 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- WNP-2
-86/08/11-HANEK, J.
ANSWER 1.05 (1.50)
P = Poe t/T 370 = 50e 120 sec/T (0.5)
T = 59.95 sec (0.5)
T X Doubling Time = ----------- 1.445 (0,5)
DT = 41.49 see REFERENCE Reactor Theory Student Text, Chapter III, PP. 14 and 15 ANSWER 1.06 (3.00) n.
FIRST:
Doppler deals with fuel temperature, and this will be the first parameter to change.
(1.0) b.
Second:
Moderator temperature coefficient begins adding negative reactivity as soon as sufficient heat is transferred to the coolant to raise coolant temperature.
(1.0) c.
Third:
Void coefficient will have little or no effect until saturation temperature is reached.
(1.0)
REFERENCE WNP-2, Reactor Theory Student Text, Ch. VIII, PP. 26 and 27 ANSWER 1.07 (1.00)
The period during power increase is governed by how quickly the neutron population can increase. [0.5]
The same holds true on a power decrease, J
however, the neutron population is dominated by the longest lived delayed neutron precursor. [0.5]
This decays with -80 sec. period.
(1.0)
REFERENCE WNP-2, Reactor Theory Student Text, Ch. VI, P.
16 ANSWER 1.08 (2.00) a.
As.the reactor operates at power, Xenon builds in to equilibrium [0.5],
adding negative ' reactivity, causing power to decrease (0.5].
(1.0) b.
In 40-50 hours [0.5] when equilibrium Xenon is reached [0.5].
(1.0) i m... _ _. _.,, _
1.
' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE 23 THERMODINAMICS. HEAT TRANS m AND FLUID FLOW ANSWERS -- WNP-2
-86/08/11-HANEK, J.
REFERENCE WNP-2, Reactor Theory Student Text, Chapter V, P 9 of 10 ANSWER 1.09 (1.50)
Qh q, 4
Power increase as recire flow is increased.
The increase in power is due W/
to the addition of positive reactivity as voids are swept away by the increase in recirculation flow ( W e As power increases, fuel temperature increases, inserting negative reactivity due to doppler broadening (0.5) f -
As this new heat is transferred to t coolant, more void formation occurs, thus more negative reactivity Therefore, boiling boundary is back to its original location, approximate (1.5)
REFERENCE WNP-2, Reactor Theory Student Text, Ch.
8, PP. 28-30 ANSWER 1.10 (1.50) a.
DECREASE (0.5) b.
There is an increased percentage of fissioning from Pu 239 i
which has a smaller delayed neutron fraction than U 238 and U 235.
(1.0)
REFERENCE Raactor Theory Student Text, Ch. VII, P. 4 (1.50)
/
ANSWE 1 11
- a. J:%. aa,er a w ay~ s~rk ass"a, a
(0.5) b.
MAPRAT = APLHGR/MAPLHGR Limit -or-APLHGR (actual)/MAPLHGR fd 6
(LCO max)
(1.0) hn+4a tim 7)
REFERENCE General Electric - Thermodynamics Heat Transfer and Fluid Flow, j
Ch. 9 P. 7Pe%
/%9///7 h rhe co psares' m/cssA9riha << sed ro /WMf)-
utere rse Mse denu/ Ah m de aceeoW i
1.
- PRIRQIP MR OF NUCLEAR POWER PLANT OPERATION.
PAGE 24 TWRMODYNAMICS. M AT TRANSFER AND FLUID FLOW ANSWERS -- WNP-2
-86/08/11-HANEK, J.
ANSWER 1.12 (1.00) b.
(1.0)
REFERENCE Gcneral Electric - Thermodynamics Heat Transfer and Fluid Flow, Ch. 7, P.
116 ANSWER 1.13 (1.00) b.
(1.0)
REFERENCE Gcneral Electric - Thermodynamics Heat Transfer and Fluid Flow, Ch. 7, P.
123 ANSWER 1.14 (1.00)
(62.4 lbm)
(55 ft - O ft) 2 P
= ---------- x --------------
OR 1 ft H O = 0.4335 lbf/in 2
2 3
2 2
(ft )
(144 in /ft )
55 ft x 0.4335 = 23.8 psig 2
P
= 23.8 lbm/in (psig)
(1.0) 2 REFERENCE Gcneral Electric - Thermodynamics Heat Transfer and fluid Flow, Ch. 7, P. 20 ANSWER 1.15 (1.50) a.
INCREASES b.
DECREASES c.
INCREASES
[3 9 0.5 each]
(2.0) 4 REFERENCE Gcneral Electric - Thermodynamics Heat Transfer and Fluid Flow, Ch.
9, PP. 93-96
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE 25 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- WNP-2
-86/08/11-HANEK, J.
ll}$%' A h
C ANSWER 1.16 (3,50
/fdS fR*0 MT'*"h, Re or ower he An e (* altup)'
FW temp Fu)
FW flow
~
[any 4 9 0.5 each]
(2.0) b.
1.
High flow high power (0.5) 2.
High flow high power, due to increased inlet subcooling from increased FW flow.
(1.0)
REFERENCE Gsneral Electric - Thermodynamics Heat Transfer and Fluid Flow, Ch. 7, PP. 92-97 P
- 2.
- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 26
~
' ANSWERS -- WNP-2
-86/08/11-HANEK, J.
l ANSWER 2.01
(.50)
Ball check valves (0.5)
REFERENCE WNP-2, Systems Training, CRDH, P. 23, Figure 11 ANSWER 2.02 (2.50)
(0.5) f a.
Shifts the recirc pumps to slow.
y'j $r#18g/
b.[Fluxshapesattheendofcyclearesuchthatthe)rodscannot st insert sufficient reactivity during the first few feet of
, rod travel on a scram due to a turbine trip.
4[
(1.0)
When + s turbine steam flow
< 142 psi first stage
[ c.
pressudr (either answer accep WOf*
- 7Kved/MRC (0,5) f d.
Key lock switch on back panels p609 and p611.
(0.25) l e.
(4) 2 out of 4 (0.25)
See 0 AfMCb REFERENCE WNP-2, Systems Training, Reactor Recire, P. 31 ANSWER 2.03 (2.50) a.
1 (d) 2 (a) 3 (e) 4 (g) 5 (c) 6 (f) 7 (b)
[7 at 0.2 each]
(1.4) b.
1.
C and D (0.2) 2.
A or C AND B or D (0.5) 3.
D (0,1) 4.
B (0.1)
>5.
A and B (0.2)
L Y = =, =- Q= /
C,s_ ff ha
,_f,ff
)
REFERENCE WNP-2, Systems Training, (NSSSS) PP.
3 and 15
/
1
- 2. ' PLM'T DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 27
" ANSWERS -- WNP-2
-86/08/11-HANEK, J.
ANSWER 2.04 (1.00) a.
1.
One gpm 2.
One Hour (1.0)
REFERENCE WNP-2, Systems Training Leak Detection System, PP. 2, 4 and 5 ANSWER 2.05 (2.00) gpm)
(0.5)
Equipment sump [25[5 gpm) n.
(0.5)
Floor drain sump ~
(0.5) b.
. Pump out timer Sump fill timer (0,5)
REFERENCE WNP-2, Systems Training Leak Detection System, PP. 2, 4 and 5 ANSWER 2.06 (1.75) a.
1.
RWCU non-regenerative heat exchangers 2.
RWCU pump coolers 3.
Recire pump jacket cooler
[3 9 0.25 each] (0.75) b.
1.
Surge tank high level slarm.
2.
Closed cooling water high radiation alarm.
[2 9 0.25 each]
(0.5)
Surge tank vent valve' closes.
(0.5) c.
re")
g,,fgee er REFERENCE WNP-2, Systems Training, RCCW, PP. 19 and 21 ANSWER 2.07
(.50)
HPCS water injection is a large thermal shock on the vessel nozzle (0.5) cnd sparger./g r ;-
,x-
[ d~ p cs -see5
-e r =7
/~~<'
= 4 H ~-r w-s ~ v.~o f e i c.
"#"~"'
REFERENCE n
d^
WNP-2, System Training HPCS, Pi 12, 2
-r.._-,.-~
r a, ~ -7 e -s,.s-sis,
we s M :' - r - n
. p,.
-z
.es e i 5 7n sue s
. W"#*
- , fr s re -* I fa,,, nowd /t'C/f.,s".3f
- 2. -PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 28
. ANSWERS -- WNP-2
-86/08/11-HANEK, J.
ANSWER 2.08 (2.00) a.
RCIC would not initiate because of the interlock between RCIC V-68 and RCIC V-45 (1.0) b.
RCIC would trip on high exhaust line back pressure.
(1.0)
REFERENCE WNP-2, Systems Training RCIC, PP. 21 and 25 e d, 1 q. rti, C US b ffl')
t ANSWER 2,09 (2.00)
' dy dy rv, or 2
g Vce 7mr r s c, ~ <* e '
l 7,eeu, }~,3,-
pp 3Q h.;3 l
I m
~
c.-ce c
- e..,,
gyp? -
b o
c - c., r a, -
- uot :~o T :,,a }_-
- g
-7<-7 n,,,,m o sw r see.
&-h~p'f~~ph_~j p
1 I
I aus aus aus I
soi e2s en vacuum sneAxen l
N- -
fl stA N-5 N
th)yf pg TYMCAL OF sEVEN ADS VALVES
[9.1 credit for cach iter merked en :ttachcd th:tch]
~
- 12. 0 P
'~
r, REFERENCE WNP-2, Systems Training ADS, PP. 3, 7 and Figure 5 i
3
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 29
' ANSWERS -- WNP-2
-86/08/11-HANEK, J.
l l
yRlufS E NA2[7y i
ANSWER 2.10 (3.00) 4 o.
Reacim pc :r is ::reater then4% vor Aritn s [C.20] w cannet
/f Lo determined [0.25-](AND reactor cannot be shutdown [0.$f before suppression pool temperature reaches 110 F [0 9'.))
(1.0) b.
1.
Isolates cleanup system outboard isolation valve.
2.
Starts "A" injection pump.
3.
Fires both "A" and "B" squib valves.
4.
Opens both pump suction valves.
[4 @ 0.5 each]
(2.0) 9 j,ff ftl9d N h Aff it h
'~ ~'%
)
FERENCE S p n '
- n /-
Systems Training SLC, PP. 8 and 16 WNP-2, A w.a s u or7 _
/
ANSWER 2.11 (3.25) o.
1.
Preheaters 2.
Catalytic recombiner 3.
Off-gas condenser 4.
Water separator 5.
Hydrogen analyzer e r 6.
10 mint.te holdup
- 7e-a',.;. yg,f,',0,
,,c,.,.,,
e-c ur-
)
7.
Cooler condensers c,-e
,,, wer s e r sw""]
8.
Moisture separator
'~'
9.
Prefilter W
m cal
- 1i;.;
, = %,e
/0 fl.
Desiccant dryers
'""L~-
~" ^ " #
// 14.
Gas coolers f u 1-3.
Charcoal adsorbers a
.N4 14.
After filter 7: r - d' '
[Tt 9 0.-t5 each]
(2.1) b.jj Valves in the off-gas system [++fr] are supplied with air from the CAS system for stem sealing F-This prevents radioactive gasses from escaping from f)1e of - as system [
].
.f/ d s/;.;
(1.15)
\\
Lt SeR tb (,4vM c$o fW fyjfe-n[c,;)}
REFERENCE WNP-2, Systems Training Off-gas Processing, PP. 3-6 and 29 i
hM d-dV Slff d f f/ 'l f' Vf
g/g used ra "we-ap " rAr sysen 6a AMeke"N b/J<c~<#
2.
' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 33
, ANSWERS -- WNP-2
-86/08/11-HANEK, J.
ANSWER 2.12 (3.00) 1.
Wet pipe:
normally pressurized with water, elevated temperature melts the fusible link in the sprinkler head to start flow.
['ressurizedwithairbetweenthedelugevalve 2.
Dry pipe / pre-action:
p andsprinklerhead)tsdareaasignalissent When a potential fire is sensed in the protec to open the deluge valve.
Elevated temperature in the protected area will melt the fusible link to start flow.
3.
Deluge Spray:
Dry pipe with open spray nozzles.
When a fire is sensed in the protected area it will send a signal to open the deluge valve to immediately start flow to the fire.
[0.25 for each correct name]
[0.25 for each correct description]
[0.5 for each correct system action]
(3.0)
REFERENCE WNP-2, Systems Training Fire Protection System, PP. 11-14 ANSWER 2.13
(.50)
FALSE (0.5)
REFERENCE WNP-2, Systems Training DC Power System, P. 3 ANSWER 2.14
(.50)
TRUE (g,5)
REFERENCE WNP-2, System Training Main Generator, P. 23 I
- 3. ' INSTRUMENTS AND CONTROLS PAGE 31
' ANSWERS -- WNP-2
-86/08/11-HANEK, J.
ANSWER 3.01 (2.75) c.
The circuit is designed to reduce the vessel inventory setpoint to a value which is sufficient to maintain reactor water level b#ow the level 8 trip setpoint following a scram.
(0.75)
(0.5)
Scram relays (or AX and AY relays.) lement to single element.
b.
1.
Transfers FWLC system from 3 e (0.5) i c.
2.
Lowers the vessel level setpoint to + 18 inches ( M (0.5) d.
The level setpoint will "rampup" to the value shown on the master (0.5) controller over a.two minute period.)
Mik/C (sec m McAnter REFERENCE WNP-2, System Training Feedwater Level Control, P. 6
- 5 Z2 / /2tu.cbr S&rm _ f / cry ANSWER 3.02 (1.75)
Transient Final c.
1.
Reactor vessel level DECREASE DECREASE 2.
Feedwater flow DECREASE SAME 3.
Total indicated steam flow DECREASE DECREASE
[6 0 0.25 each]
(1.5) b.
No.
(0.25)
REFERENCE WNP-2, Systems Training Feedwater Level Control, P. 15 ANSWER 3.03 (2.00) a.
1.
Droop is used when operating in parallel with other power sources.
(0.5) 2.
ISO is used when suppling an isolated bus.
(0.5) l b.
ISO (0.25) c.
DROOP: (ihs-erts a 5% ispeed cYUop Izvie ? te-fu1 M a d-OR
, TEtie57he' iltf=loiid !FpWdietting=SE)fo compensate j
for droop in speed created by an increase in load.
(0.75)
- cA -
sl0 7/</ N'/ 4 (n-e"^~
to fi ~-Q 74 G e' f e
A r un -
o/(,
REFERENCE r,.
WNP-2, Systems Training Diesel Generators, P. 64
/
,,,,7,,p n,
7ia
/ e ** $
- 3.
' INSTRUMENTS AND CONTROLS PAGE 32
~
.' ANSWERS -- WNP-2
-86/08/11-HANEK, J.
f ANSWER 3.04 (2.50) a.
SRM detector not fully inserted AND SRM count rate < 100 CPS (0.75) b.
Bypassed when:
Mso :
SCM > /wc/s 1.
ALL IRM range switches are on Range 3 or above.
.5) 2.
SRM channel is bypassed.
(d. )
/
(
(/f (0.25) 3 Mode switch is in run l
- c. 'Nb SA M >. /80 4 N Ay
,G /r/ 6/
- * > f' REFERENCE WNP-2, Systems Training SRM's, P. 27 ANSWER 3.05 (2.75) c.
Area 2 - Ionization I# ^
Area 3 - Proportional Area 5 - Geiger Mueller (GM)
Area 6 - Continuous Discharge (a- <<~~r e ~""]4 9 0. 25 each] (1.0) b.
1.
SRM - Area 2 2.
IRM - Area 2 3-
,j ffA, I -
3.
Fuel load dunking chamber - Area 3 e'r - ?, ~,,,,"' ' " ' V 4.
ARM's - Area 5 5.
Main steam line rad monitors - Area 2 6.
Portable frisker station - Area 5 7.
Dose rate setting portable instrument - Area 2
[7 9 0.25 each](1.75)
REFERENCE WNP-2, Systems Training, SRM's, P. 7 and Figure 4, ARM's P. 2, Process Effluent Radiation Monitoring System, P. 4 ANSWER 3.06 (1.00) a.
1.
5% 0 ' 3 Y W V *' V 2.
12%
[2 9 0.5 each]
(1.0) i l
REFERENCE
//
WNP-2,. Systems Training APRM's, P. L9 ) M'8M 7un,u e.
s-e c <no n w J
- 3.
' INSTRUMENTS AND CONTROLS PAGE 33
' ANSWERS -- WNP-2
-86/08/11-HANEK J.
ANSWER 3.07 (2.25) a.
1.
>10% difference between comparators 2.
Upscale trip 108%
3.
Mode switch not in operate 4.
Module in the flow unit is removed
[4 at 0.25 each]
(1.0) b Rod block (0.25) hh c.
If a flow unit is bypassed the flow signal to the comparator is automatically bypassed to the next unit. (If A is bypassed, then i
the signal from flow Unit C would be compared with flow Unit B)
(1.0)
(+ 1 S y & & c c-0 @ ~ d ~ ~ ~ ~ro)
REFERENCE WNP-2, System Training APRM's, PP. 10 and 11
'd ' " M#
r m re urm e ANSWER 3.08 (2.50) a.
To change gain of averaging amplifier so that the RBM output (averaging amplifier output) will be equal to or greater than reference APRM output.
(0.5) b.
1.
Local power may be significantly lower than core (0.5) average power.
2.
Several of the highesc reading LPRM's that normally input to a RBM channel might be bypassed.
(0.5) c.
RBM A - APRM C Normal APRM E Alternate RBM B - APRM D Normal APRM F Alternate
[4 0 0.25 each)
(1.0)
REFERENCE WNP-2, Systems Training REM System, PP. 8, 9 and 31 ANSWER 3.09 (2.00) a.
The function of the valves is to shut-off flow in the event of a leak or line rupture downstream of the valve.
(0.5) hhb.-~La.u
~ instrument lines (with normally no flow ouu conditions)
(0.5)
To equalize pressure on both sides of the valve after an c.
actuation thus allowing spring pressure to reopen the valve.
(0.5) d.
Board S in the control room.
(0.5)
REFERENCE WNP-2, System Training Nuclear Boiler Instrument, PP. 48 and 49
3.
INSTRUMENTS AND CONTROLS PAGE 34
. ANSWERS -- WNP-2
-86/08/11-HANEK, J.
j ANSWER 3.10 (1.50) n.
No.
(0.25)
The wide range indicator is calibrated with no jet pump flow so at 100% power it will read 15-20" low due to the draw down effect of the jet pumps.
(0.5)
(0.25) b.
No.
TAF is -161 inches, lower end of wide range level isntrument is -150 inches.
(0.5)
REFERENCE WNP-2, Systems Training Nuclear Boiler Instrumentation, PP. 17, 58, and 59 fy offARTCo%'v AUT[" fb"5 ANSWER 3.11 (2.00)
-7/it a A p e
- 7 r- / /
gg f je. 4 m -< z v ' -
ul.
?ransf ers boi,li luv_r_ lov centroller_% - enterstrie
-to manual. " -
00-5F#
2.
Locks the flow control valves.
6 I)C '. '( E 5) ^ >5 3.
Locks the recirc pump discharge valves.
(_ r) ( -,f e d P j:
b.
The basis for these automatic actions is to allow rapid vessel depressurization in the event of a break in the 7 / #),
recirculation system piping.
-t0-STr
/2dfdA&M Y.' 7
, yp,).s syJ /en NO,b [^)
/d M '/- {& C N
f JJ h e s A,,,n 7,,,,,,-,.,,,,n r
m
- 3.. e e nn, ANSWER 3.12 (2.00)
,g a.1. Automatic TIP withdrawal.
a.2. TIP ball valve closes.
b.1.
In-Shield light lit.
I b.2. Valve open light lit
- y. ep (4 @ 0.5 ea.)
V,,n, Q a.
y
,e z. win
,,,,,cr REFERENCE WNP-2, Systems Training TIPS PP 11, 13, and 14.
9
d'.
' PRO ('EDURES - NORMAL. ABNCRMAL. EMERGENCY AND PAGE 35
, RADIOLOGICAL cot!IRQL 4
- ANSWERS -- WNP-2
-86/08/11-HANEK, J.
ANSWER 4.01 (1.00)
TWO independent indications confirm misoperation in automatic mode or Edequate core cooling is assured.
(1.0).
REFERENCE WNP-2, Procedure 1.2.1, Standing Orders ANSWER 4.02
(.50)
Leave the controlled area and notify the HP/ chemistry department.
(0.5)
REFERENCE WNP-2, Procedure 1.11.3, P. 8 ANSWER 4.03 (1.00)
When it is necessary to prevent injury to personnel or damage to the plant.
(1.0)
REFERENCE WNP-2, Procedure 1.2.3, P. 2 ANSWER 4.04
(.50)
Shutdown the pump within one minute.
(0.5)
REFERENCE WNP-2, Procedure 2.2.1, P. 4 ANSWER 4.05 (1.50) a.
Close the valve to prevent reverse rotation of the pump.
It is reopened to maintain the idle loop temperature.
~
'(1.0) b.
After 5 minutes.
(0.5)
d'.
- PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 36
,. RADIOLOGICAL CONTROL l
AdSWERS -- WNP-2
-86/08/11-HANEK, J.
REFERENCE WNP-2, Procedure 2.2.1, P. 4 ANSWER 4.06 (2.00)
By attempting to pull each control rod withdrawn to Position 48 a.
to the overtravel position.
A valid coupling check requires a continuous withdraw signal be maintained long enough to observe drive stall flow at Position 48.
(1.5) b.
Rod overtravel alarm.
(0.5)
REFERENCE WNP-2, Reactor Cold Startup Procedure 3.1.2, P. 6 ANSWER 4.07
(.50)
FALSE REFERENCE WNP-2, Plant Shutdown Procedure 3.2.1, P.
5 ANSWER 4.08 (2.00) a.
1.
80 inches 2.
metal 3.
hour
[3 0 0.5 each]
(1.5) b.
To insure natural circulation core cooling capability.
(0.5)
REFERENCE WNP-2, Normal Shutdown to Cold Shutdown Procedure 3.2.1, P. 11 ANSWER 4.09 (2.00) a.
1.
Reactor feedwater pressure 3ess than 420 psig OR 2.
Reactornfeedwater temperature is less than 340 F.
3.
RWCU flow is returned to the RPV and allowed to repressure the feedwater piping.
[3 0 0.5 each]
(1.5) b.
To. minimize water hammer upon restart.
(0.5) 4
---.-.= --
~ Y. ' PROCEDURES-- NORMAL. ABNORMAL. EMERGENCY AND PAGE 37
,. RADIOLOGICAL CONTROL ANSWERS -- WNP-2
-86/08/11-HANEK, J.
REFERENCE WNP-2, Reactor Scram Procedure 3.3.1, PP. 2 and 3 1
ANSWER 4.10 (2.75) l o.
1.
DECREASE 2.
INCREASE
.,,eg3 g,yg,e pcg gu 4 w// mg e.ry/ & NWf
[4N b 5 e h] (
0)
Notify control roo[m isupervisor.fj'S j If d k N N"#
RE f#
7 h oe 44we-ge__ r, (0.25 e
verify reiii5 tor vessel MMM MV W) b.
1.
2.
Using redundant indications, water level and drywell pressure are normal.
(0.5) 3.
Shut HPCS injection valve (HPCS-V-4), and verify that the flow valve opens.
After confirming the HPCS pump p^ inadvertent, stop the HPCS pump.
(1.0) injection was REFERENCE WNP-2, Procedure Inadvertent HPCS Startup, PP. 1 and 2 ANSWER 4.11
(.50)
FALSE (0.5)
REFERENCE WNP-2, Emergency Procedure 5.0, P. 2 ANSWER 4.12 (2.75)
RPV water level below +13.0 inches
-(0.5)
RPV pressure above 1037 psig (0.5)
Drywell pressure above 1.65 psig (0,5)
A condition which requires MSIV isolation (0.5)
A condition-which requires a reactor scram and reactor power is above 5% or cannot be determined.
(0.75)
REFERENCE WNP-2, Emergency Procedure 5.1.2, P. 1
', + >.
L PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AE PAGE 38
. RADIOLOGICAL CONTROL
'-A?SWERS -- WNP-2
-86/08/11-HANEK, J.
1 ANSWER 4.13 (1.50)
Haat removal from the reactor sufficient to restore and maintain fuel cladding temperature belos 2200 degrees F.
(1.5)
REFERENCE l
Emergency Operating Procedure Fundamentals, P. 6 ANSWER 4.14 (1.50) 1 1.
Core submergence or (RPV level above TAF) 2.
. Spray cooling or (core spray operating at or above design conditions) j 3.
Steam cooling.
[3 0 0.5 each]
(1.5)
REFERENCE WNP-2, Emergency Operating Procedure Fundamentals, P. 6 W --
W 4.15 (2.50)
/
A
,G 9
a.
1.
E ff-LE 2.
5m utes (1.0; b.
Place the mode switch in shutdown and o at least one loop of in the suppression pool cooling mo (1.5 l
i t
^
~~~
NNP-2, T;;h. Specs. ?.S.2.1 oud 4.0.2.-1~
l ANSWER 4.16 (2.50) a.
1.
Manually scram the reactor.
2.
Place the mode switch to shutdown.
3.
Verify that APRM downscale lights are illuminated.
4.
Transfer recire pumps to the LFMG 5.
Close all MSIV's
[4 0 0.5 each)
(2.0) b.
Shift Manager (0.5)
REFERENCE WNP-2, Procedure 4.12.1.1, P.
11 i
^S D
y po
=
U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION Facility: WNP-2 Reactor Type: BWR-GES Date Administered: August 11, 1986 Examiner: THOMAS R. MEADOWS, RV Candidate:
INSTRUCTIONS TO CANDIDATE 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 starts.
% of Category
% of Candidate's Category Value Total Score Value Category 25 25 5.
Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics, 25 25 6.
Plant Systems Design, Control and Instrumentation 25 25 7.
Procedures - Normal.
Abnormal, Emergency, and Radiological Control 25 25 8.
Administrative Procedures, Conditions, and Limitations 100 TOTALS Final Grade All work done on this examination is my can, I have neither given nor received aid.
r Candidate's Signature l
I i
REQUIREMENTS FOR ADMINISTRATION OF WRITTEN EXAMINATIONS 1.
A single room shall be provided for completing the written examina-tion. 'ihe location of this room and supporting restroom facilities shall be such as to prevent contact with all other facility and/or i
contractor personnel during the duration of the written examination.
If necessary, the facility should make arrangements for the use of a suitable room at a local school, motel, or other building.,Ob-taining this room is the responsibility of the licensee.
2.
Minimum spacing is required to ensure examination integrity as determined by the chief examiner.
Minimum spacing should be one candidate per table, with a 3-ft space between tables. No wall charts, models, and/or other training materials shall be present in the examination room.
3.
Suitable arrangements shall be made by the facility if the candi-dates are to have lunch, coffee, or other refreshments. These arrangements shall comply with Ites 1 above. These arrangements shall be reviewed by the examiner and/or proctor.
~
4.
The facility staff shall be provided a copy of the written examination and answer key after the last candidate has completed and handed in his written examination. The facility staff shall then have five working days to provide formal written comments with supporting documentation on the examination and answer key to the chief examiner or to the regional office section chief.
5.-
The licensee shall provide pads of 8-1/2 by 'll in. lined paper in unopened packages for each candidste's use in completing the exam-ination. The examiner shall distribute these pads to the candidates.
All reference material needed to complete the examination shall be furnished by the examiner. Candidates can bring pens, pencils, calculators, or slide rules into the examination room, and no other equipment or reference material shall be allowed.
6.
Cnly black ink or dark pencils should be used for writing answers to questions.
i P
Examiner Standards
lo n
4 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During 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 penalties.
t l
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 pencil only to facilitate legible reproductions.
4.
Print your name in the blank provided on 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 number each answer sheet, write "End of Category,,,ide of
" as appropriate, start each category on a new page, write only one s 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.
)
4
- 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.
Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND D0 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 assittance in completing the examination. This must be done after the examination has been completed.
Examiner Standards l
l
- 18. When you complete your examination, you shall:
s.
Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are a 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 balance of the paper that you did
, not use for answering the questions.
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.
"o e
I l
Examiner Standards
O 1
1 EQUATION SHEET l
~ '"
- E Cycle efficiency = Net Work (out)-
2 w = mg s = v,t +
l at Energy (in)
E = mC a = (vg - v )/t KE = mv vg=v + at A = AN A = A,e PE = agh w = 8/t A = in 2/tg = 0.693/tg W = VAP g(eff) = (t )(t )
i; s
't AE = 931Am
(
+
)
Q=$ CAT I. I e'EX P
o Q = UAAT
-UX I.Ie Pwr = W ' E
-x h
~
g I=I 10 5UR(t)
P=P 10 TVL = 1.3/u y.p.t/T HVL = 0.693/u i
O SUR = 26.06/T
~
T = 1.44 DT SCR = S/(1 - K,gg)
/A*fr )
o SUR = 26 CR g,p
= S/(1 - K,gg )
1(
eff " 2(I ~ Edf)b T = I(i*/p ) + [(f 'p)/A,g,p] ~ T = 1*/ (p - E M " I/Cl - Kefg) = CR /CR g 0 ( ~ 8)! eff' M = (1 - K,gg)0 (1 - K,gg)g / 8*I eff'I)! eff = AK,ff/Kaff SDM = (1 - Keff)/Keff [1*/TK,'gf -] + [I/(1 + A,gf )] 1* = 1 x 10 seconds p= ~ T P = I$V/(3 x 1010) gaff = 0.1 seconds A I = Na Idly =Id22 WATER PARAMETERS Id =Id g 2 1 gal. = 8.345 lbm R/hr = (0.5 CE)/d (meters) 1 gal. = 3.78 liters R/hr = 6 CE/d (feet) I ft3 = 7.48 gal. HISCELLANEOUS CONVERSIONS i Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps ~ 10 3 Density = 1 gm/cm 1 kg = 2.21 1ha 3 Heat of varorizationi = 970 Etu/lbm I hp = 2.54 x 10 BTU /hr 0 Heat of fusica = 144 Btu /lbm 1 Hw = 3.41 x 10 Btu /hr 1 Atm = 14.7 Psi = 29.9 in, l's. 1 Btu = 778 ft-lbf 2 1 ft. H 0 = 0,4333 lbf/in 1 inch = 2.54 cm g F = 9/5 C + 32 C = 5/9 (*F - 32) t 3 _.-, - - ,-..,,-,.e s n e - --,- ,-m-----,,.n ,......n.-,,.,,,,,.,,---a, .-...,n--. y..
7 SECTION 5 Theory of Nuclear Power Plant Operations, Fluids and Thermodynamics 5.1 (1.0) As a subcritical reactor nears criticality, the length of time to reach equilibrium count rate after an insertion of given fixed amount of positive reactivity... (SELECT THE CORRECT ANSWER) ta) decreases primarily because of the increased population of delayed neutrons in the core. (b) increases primarily because of the increased population of delayed neutrons in the core. (c) increases because of a larger number of neutron life cycles required to reach equilibrium. (d) decreases because the source neutrons are becoming less important in relation to total neutron population. (c) Ref: Reactor Theory-Student text 5-1
o a 5.2 (2.0) Why is the value of Keff INDEPENDENT of initial source range counts ? INDEPENDENT-Keff (the neutron life cycle) only considers fission neutrons in the self-sustaining reaction. Ref: Reactor Theory-Student text I 1 6 e
i 5.3 (2.0) Answer the following TRUE or FALSE: (a) During equilibrium power conditions,the production rate of indirect Xenon from Iodine is faster than the decay rate of Xenon to Cesium. (0.5) f (b) Slowing the rate of a power decrease, lowers the height of the resultant Xenon peak. (O.5) (c) The resultant Xenon peak due to a scram from 50% power is larger than one from 100% power. (0.5) (d) During an increase in power from equilibrium (0.5) Xenon conditions, Xenon concentration initially decreases. (a) TRUE (b) TRUE (c) FALSE (d) TRUE Ref: Reactor Tneory - Student Text 6 9 4 3-5
9 l s 1 5.4 (3.0) (a) Will the resonance escape probability INCREASE or DECREASE with increasing void fraction 7 (0.5) (b) What is the reason or basis for your answer above? (2.5) DECREASE-The reason for this effect is that an increased number of voids results.in less neutron moderation,which results in an increase in the average distance traveled (and time required) by a neutron while slowing down to thermal energy levels-- neutrons spend more time.in the resonance capture energy region, resulting in increased RESONANCE l ABSORPTION, hence a DECREASED resonance escape probability. 2 Ref: Reactor Theory-Student Text i 4 i' 4 l I I e T 3-4
a 1 5.5 (1.0) What two physical phenomena ( specifically related to the neutron life cycle and fuel characteristics) along with increasing fael temperature, contribute to a negative doppler coefficient? NAME THESE TWO(2) PHENOMENA (0.5 points each) (e) Resonance broadening-neutron absorbtion at other than descrete energies (Doppler Effect) (b) Self-Shielding (heterogeneoun fuel pins) Ref: Reactor Theory-Student Text 5-5
e 5.6 (4.0) Prior to performing a control rod notch calibration for the reactor engineer, the reactor is just critical at O.002% power. A control rod is subsequently withdrawn one notch and reactor power increases on a steady period with a DOUBLING TIME of 40 seconds. (a) What is the notch atrength (&K/K per notch) of the withdrawn control rod? (2.0) -(b) How long will it '.ake to reach O.08% power? (2.0) SHOW ALL WORK AND ASSUMPTION 3 (a) P=Po e (t /T).... P=2Po 2=e(DT/T) ,,DT=Deubling Time T In2 = DT.....T=4(/In2 T= 57.7 seconds 4 K/K=AK/Kred = U/1+3T any reasonable assumption: b O.OO70 4K/K %=0.isec-4K/K= 0. 0070/1+ (0.1) (57. 7) = 0.103 % 4K/K (b) 0.002% to 0.08% power change: P=Po e(t/T), 0.08% = 0.002% ett/T),,(Any T used from above acceptable) - solve for t t=212.8 seconds Ref: Reactor Theory - Student Text i 5-6
6 5.7 (2.0) How can the REVERSE POWER EFFECT occur in the core during control rod movements 7 J Reverse power effect is a decrease in overall core power upon the notch withdraw of a shallow control rod. An with deep rod control, local bundle power will increase as the i rod is withdrawn;however, in the lower regions of the core it is possible for the voiding effect created within-the i affected bundle by shallow rod withdraw to overrice the associated local power increase, subsequently lowering overall core power. 4 Ref: Reactor Theory-Student Text j .5 1 i s i I I i 1
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e 1 I 5.8 (4.0) The reactor is operating at 70% power. Recirculation flow is subsequently increased to provide stable 100% power and 100% flow conditions : (a) Why did the boiling boundry INITIALLY move further up the core? (1.0) (b) What causes the boiling boundry to return to near its original axial position and power to stabilize?(2.0) (c) Why did the new boiling boundry stabilize at a SLIGHTLY higher core position 7 ( 1. 0). (a) Initialy the water will travel further up the core before reaching saturation conditions because it spends less time in contact with the rods at a given point at higher flow rates. The boiling boundry subsequently moves further up the core. (b) As voids are initially displaced, resulting increased neutron moderation adds positive reactivity, and increases power. The power increase subsequently causes increased boiling - lowering the boiling boundry to near its original panition. The negative reactivity added by the return of voids and increased fuel temperatures offuets the positive reactivity added by increased flow-stabilizing power. (c) The boiling boundry stabilizes slightly higher providing positive reactivity (more moderation), to offuet the negative reactivity inserted by the doppler coefficient at higher fuel temperatures. Ref: General Electric-Thermodynamics Heat Transfer and Fluid Flow, FNP2 Systems - Reactor Recirculation Flow Contal 5-8
,e i 5.9 (2.0) s What is the reason or basis for the MAPLHGR f THERMAL LIMIT ? To prevent fuel cladding disintegration (" gross cladding failure" ) during a DBA LOCA by limiting the peak clad temperature to < 2200 degrees F-- (limiting bundle stored energy such that any given fuel pin at any given axial / planer point will have a peak clad temperature below 2200 degrees F during a DBA LOCA dry-out condition. 2200 degrees assures that the cladding will not t I disintegrate-brittle fracture-upon the subsequent rewetting I by low pressure ECCS). Ref: General Electric-Thermodynamics Heat Transfer and Fluid Flow, Technical Specifications I k 1 I i i ) a l i O T l 5-9
5.10 (1.0) What is the relationship between MAPRAT and MAPLHGR? MAPRAT = APLHGR/MAPLHGR Limit -or-APLHGR (actual)/ MAPLHGR (LCO ma:< ) Ref: General Electric-Thermodynamics Heat Transfer and Fluid F1ow t 5-10
a e 5.11 (0.5) Of the following operations,which DNE will have a negative effect (reducing effect) on available Net Positive Suction Head (NPSH) of a given centrifugal pump: (a) Throttling open the pump's suction valve. (b) Throttling open the pump's discharge valve. (c) Decreasing the pump's speed. (d) Decreasing the temperature of the fluid (water) being pumped. (b) Ref: General Electric-Thermodynamics Heat Transfer and Fluid Flow i l l 9 5-11
0 0 5.12 (0.5) Consider a real plant system (NON-IDEAL) with two centrifugal pumps in parallel, one of which is running at 1800 RPM. The second pump is started and run at 1000 RPM. l System flow will be... SELECT THE CORRECT ANSWER i i (a) more than double the original flow due to decreased flow resistance. (b) slightly less than double the original flow due to increased flow resistance. (c) the same since only the dicharge head changes. (d) reduced by one-half due to increased discharge head. (b) 4 Ref: General Electric-Thermodynamics Heat Transfer and Fluid Flow 4 l i l 4 1 i ] 1 G-12
d 5.13 (2.0) Identify the following on the attached Figure 5.6 (T-S diagram of a symplified steam cycle): (0.25 pts each) (a) Subcooled Region (b) Critical Point (c) Saturated Liquid Line (d) Saturated Vapor Line (e) Superheat Region (f) Ideal work out of HP turbine (g) Ideal work out of LP turbine (h) Feedwater heating process (see attached Figure 5.13 KEY) END OF SECTION 5.0 k 4 5-13
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4 SECTION 6.0 t Plant Systems Design, Control, and Instrumentation 6.1 (1.5) I List three(3) of the four(4) major ESF 4160 Vac component loads supplied by critical switchgear bus SM-B division 2. (0.5 points each) Any three of the following: RHR Pump B RHR Pump C SSW Pump 1D (B) CRD Pump 1B (B) Ref: WNP2 Systems-A.C. Distribution i + +. 1 't 6-1
E 6.2 (0.5) FILL IN THE BLANK Each D.C. division battery system has sufficient capacity to + supply the DC power requirements of the redundant (Division ~ 1 and 2) load groups f or at least hour (s) following loss of power to the battery chargers (station black-out) coincident with a design bases accident. N two (2) Ref: WNP2 Systems-D.C. Power System l 0 5-2 ~
6.3 (3.0) What is the reason or basis f or the 105 secored initiation logic time delay in each of the two ADS logic channels (A/C and B/D respectively)? (include in your dicussion the reason that this time delay is not set for a longer period) This time delay is intended to allow HPCS sufficient time to reflood the vessel in the event of a small to intermediate break, but not so long that the RHR(LPCI) and LPCS systems would be unable to cool the fuel, coincident with a HPCS failure to start when ADS is relied upon to provide depressurization. Ref: WNP2 Systems-ADS 1 1 - ~
6.4 (3.0) A level two(2) RPV level trip at -50" will initiate NSSSS Isolation, Groups 1,2,3,4,L7. What are the other three(3) trip functions associated with this level trip 7 ( 1.0 pt. each) 1 (a) Initiate RCIC i (b) Initiate HPCS (including the HPCS D/G) j (c) Trip the Recirculation pumps Ref: WNP2 Systems-Nuclear Boiler Instrunentat on 9 6-4
s 6.5 (2.5) A 10 minute time delay after a LFCI initiation sional def eat = 41-1. o close signals to/CPCI injection - val ves HR-V-53A and V What are two(2) reasons that this ECCS flow path is _ preferred? ( 1.25 pts each ) 0] [f0' ' ' ' ' ' ' ' " ,,,,.y-(a) To prevent cold water transients that would occur if direct core injection were used. (b) To prevent baron dilution accidents, if baron had been injected in an emergency conditions, f) Ref: WNP2 Systerrs-RHR eag 9 ~~ ?~~ 9,.,2-m. z (n) ms n a,- ~ w - - ? c e 7.r s,s e c r i >- /~r,or ru e A'P V s.~er w - fg) gt ~g'seWff 7Nsf /"t p *.- s e !"M s ".7N ES /,Wd*C ff Y / "5"/#d ' "r r~xo~s r a, - < o n,, ,,,,.cies er / Al*C /^* C - g p pg* gyg j g 1 a 6-5 7
6.6 (0.5) At what RPV pressure (decreasing).will the LPCI injection valves receive an open permissive signal after LPCI logic i initiation 7 1 i 470 psig Ref: WrJP2 Systems-RHR ~ e I 1 l i i ~'R l i 4 i I. i e e i. 1 i i j i b~h i
o 6.7 (1.5) What three conditions will result in an IRM inoperative rod block ? (0.5 points each) Detectcr High Volts Low Module Unplugged IRM Mode (Function) Switch not in operate Ref: WNP2 Systems-IRM 6-7
s 6.9 (1.5) The gamma flux error signal IS discriminated against in the intermeadiate power range. How is this accomolished by the IRM system 7 t Gamma discrimination is still necessary in the intermediate power range,but is not accomplished by means of an active discrimination circuit. Gamma discrimination in the IRM system is a natural censequence of the mean-square-analog (Campbell) method of measurement,such that the gamma contribution to the output signal becomes insignificant. (a dctailed discussion of circuit operation or "Campbelling" is not required) Ref: WNP2 Systems-IRM 1 f r 6-8
= 6.9 (3.0) For each of the f ollowing scram signals, state the' reactor mode switch position (include necessary opernting conditions and/or component positions) that WILL bypass the scram signal: (0.5 points each) (a) IRM INOP (b) IRM Hi-Hi (c) APRM Hi-Hi (15%) (d) APRM Hi-Hi (118%) (e) SCRAM Discharge Volume (f) MSIV Closure (a) Run mode i (b) Run made (c) Run modo (d) not in Run Mode (e) In S/D or Refuel, and SDV keylock switch in bypass (f) not in Run Mode, and < 1056 psig Ref: WNP2 Systems-RPS l i 6-9
6.10 (2.5) What is a Non-Coincidence Scram? (Include what scram trips can be set up for non-coincidence, how set up is accomplished, and when non-coincidence is used) With shorting links removed, any one SRM,IFM, or APRM (Nuclear Instrumentation) scram signal from any RPS channel will result in a full scram. This non-coincident nuclear instrumentation scram capability is used during initial core loadingjd5duringsometestingj#63hutdownmargin demonstrationsf"lollowing any refueling cycle. Ref: WNP2 Systems - RPS / e l 6-10
6.11 (4.0) Concerning the attached Figure 6.11 ( two pump power / core ficw operating map): (a) Identify the Natural Circulation Line (0.25) (b) What is the reason or basis for the Natural Circulation Line 7 (0.75) (c) Identify the Twenty Four(24%) Percent PumpSpeedL1}ne._ (0.25) (d) What is the reason or basis for the Twenty Four(24%) Percent Pump Speed Line? (0.75) (a) Identify the Constant Speed Lines. (0.25) (f) What is the reason or basis fer the Constant Speed Lines? (0.75) (g) Identify the Recirculation Pump Cavitation Line. (0.25) (h) What i= the resscn or basis f or the Recirculation Pump Cavitation line ? (0.75) (see Figure 6.11-KEY, for questions a,c,e,& g) (b) As power is increased by withdrawing control rods, care i flow will increase due to natural circulation effects - hatter water and steam within the core shroud i s less dense than the subcooled water of the downcomer area. This_effect will support a natural core flow that supplements forced circulation-occuring with or withcut forced circulation. Its effect is more pronounced at lower core flows due to less flow resistance at lower power levels. (d) The 24% pump speed lines define core flow at varying power levels for MINIMUM pump speed. (f) Defines the change in core flow,with constant pump 6 N ^'" "* speed, due to rod pattern changes. Flow increases as Power is decreased (by inserting control rods) due to decreased channel flow resistance as steam production . decreases. (h) Defines the locus of power / flow conditions, below which cavitation of the recirculation pumps would occur. Ref: WNP2 Systems - RFC 6-11
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l Figure 6.11 120 WNP 3 iWO PUMP OPERAilON l 110 gC A 100 8 ~ 7 { 80 6 l Eg 70 5 2 4 g 60 2 U O l E 50 'g# / A. / / 40 y s 30 -~ .' ~'N l 20 K 10 [ I I I I I I - 0 0 10 20 30 R 40 50 60 70 80 90 100 110 PERCENT CORE FLOW TWO-PUMP TilERMAL POWER - CORE FLOW OPERATING MAP BASED ON: . ass ess ig ~ N W 190 FSAR, APPENDIX 11, FIGURE 11.2.5-1 =
(' 's ~ Figure 6.11 KEY 120 WHP-3 TWO Pune OPEnATION N -JET PuuP HO22 E CAVai AllON LINE 110 s.xt euur sucisON eavsiAliON t Nc Lines 0-9 & A-C are BC ^ n.nccinc ruue caviiAliON tis 4E " Constant speed lines" v -VALVE cAViiAisON LW4E 9 100 renctNi vALvt rossiiON FLOW COHinOL VALVE 8 e a nAlunAL cincuLAllON g e 24% roue sesto 7 soo 3 8 100% PUMP SPEED g a se 6 e as 5 3: 70 4, y 60 3p 0)/ / o 6 60 / N L / 40 p s 30 R ,f ~%g V N 20 Ny H N 10 S R I I I I I I O O 10 20 30 R 40 50 60 70 80 00 100 110 PERCENT CORE FLOW TWO-PUMP THERMAL POWER - CORE FLOW OPERATING MAP I ~ DASED OH: assess.th FSAR, APPENDIX 11, 'I!OV'88b FIGURE 11.2.51
6.12 (0.5) Following a reactor scram, the four rod display position goes blank, and the green full-in light on the full core display for the selected control rod is lighted. ] Is this normal indication for the RMCS (YES or NO) ? YES Ref: WNP2 Systeras - RMCS i e i 6-12
4 6.13 (1.0) The shutdown cooling made of the RHR system is isolated at 125 psig. What is the reason or basis for this isolation? To protect the RHR pumps from exceeding their design Saturation Temperature (360 degrees F.) for 135 psig. Ref: WNP2 Systems - Nuclear Boiler Instruinentation j END OF SECTION 6.0
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SECTION 7.0 Procedures - Normal, Abnormal, Emergency, and Radiological Control 7.1 (2.0) What is the reason or bases for NOT starting more than two(2) circ water pumps while SM1,2,& 3 are being supplied by startup transformer TR-S,upon Plant Startup? An undervoltage or degraded voltage on TR-S will occur upon the start of the third circ water pump. Ref: PPM 3.1.2, Plant Startup 4 r 4 7-1
7.2 (2.0) State four(4) of the five(5) items logged at the time of criticality, required by PPM 3.1.2 - Plant Startup. (0.5 pts each) i (Any 4 of the following) (a) Time - (b) Control Rod Position (c) Neutron level (d) Reactor period (e) Reactor coolant temperature I Ref: PFi-i 3.1. 2 Pl en t Star tup i J r f l I i 7-2
7.3 (3.0) Concerning the MINIMUM REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE curves (attached figures 7.3.a and 7.3.b). The RCS temperature and pressure shall be limited in accordance with the limit curves A and A*, B and B*, C and C' for different plant conditions associated with these curves. What are the plant conditions and/or evolutions applicable to each of the three (3) limit curve sets A and A*, B and B', C and C',respectively ? (Temperature limits are NOT necessary) ( 1.0 pt. for each ) ( can be in any order ) (a) A-A'; Hydrostatic or leak testing (b) B-B'; for heatup by Non-nuclear means,Cooldoen following a nuclear shutdown, and low power physi-: tests (c) C-C'; for nonrations with a critical core ( otner than low power physics tests) Ref: Technical Specifications 7-3
Figure 7.3.a A '5 C I If ~h l I l l Il.I y ^ i l ll fl I 1. l l Il l ~ I I 1 Q I; l l l l 1 II I l 6I I I I Il I Q i l l' I l }l l I M ecam I I l / i ~l i Il l 17iilI ~ l I li ll I III I I I I -1 11 I I I I I I I I J l l I ~ l L II / I II I i I ' / I II IIl I j,,, I I /I I II I lAl I I l l l ~ i i i I/ I II AII I I i ll If I l l L l l R6"l I l l l 1 I l l l Il l il I hl I l l l l l l i, I I I I I ll LVI I I I I I I 1 s i l l '/i il 81l l l l l 1i ll l ll I/, F l I' l I I l l l 7~ ~ 2 III Ill RIIl l I III I I I IL icP l I I I I i i l i i ~ l l l' / IbdiII I I Il l l / MI I llI i l l l l l l / / I II l l l l l l l l-II l / '/l i II l l l l l l l l ~ l l l Ill l II l l l l 1 1II l VL ll l l l l l 1 III I1 6 Ii l l l l l l l l 1 I I I .I I I I I I III I I-1I i ] o too nos aos e a umamuu asacron vcssa, me7As. Tsursmarune (+r)
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7.4 (2.0) If any of the RCS Pressure / Temperature limits are exceeded (limit curves shown on figures 7.3.a&b), What ACTION is required per Technical Specifications ? Restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on structural integrity of the RCS; determine that the RCS remains acceptable for continued operations-Ref: Technical Specifications ~ i i i 7-4
j 7.5 (3.0) In accordance with 10 CFR 20, " Standards for Protection against Radiation": (a) What are the Radiation Dose Standards for individuals in restricted areas per calender quarter? (1.5) (b) What are the three requirements that must be met if the Whole Body limits for a calender quarter are to be exceeded? (1.5) (a) (0.5 ots. each) 1.25 Rem - Whole Body (head & trunk, active blood forming organs, lens of eyes, gonads) 18.75 Rem - Hands & Forearms; Feet & Ankles 7.5 Rem - Skin of Whole Body (b) (0.5 pts, each) 3.0 Rem per calender quarter - MAXIMUM 5*(N-18) total accumulated done to the whole body, Where N is the individual age in years at his last birthday. Form NRC -4 or equivalent. Ref: 10 CFR 20 J 7-5
7.6 (3.0) A LOCA has occured and a high temperature steam enviornment exists in the drywell. EXPLAIN why the drywell sprays must NOT be initiated if containment conditions exist such that the UNSAFE region of attached figure 7.6 (DRYWELL SPRAY INITIAlION PRESSURE LIMIT) is entered. Spray initiation above this limit may (through the combined effects of evaporative and convective cooling) result in a containment _degressurization rate which exceeds the relief N gra# % city of the7. F. Lu D.W. vacuum breakers. As a result, containment negative design pressure may be exceeded, i leading to containment failure. Ref: EPG's, General Electric-Emergency Operating Procedure Fundamentals e i l 7-6
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3 7.7 (2.0) Define ADEQUATE CORE COOLING, as used in the Emergency Procedures. Heat removal from the reactor sufficient to restore and maintain peak fuel cladding temperature below 2200 degrees F. -oK-Ref: EPG's, General Electric-Emergency Operating Procedure Fundamentals s, x - - ~ - - c, C o d t s.s. g g y,* [ h d V ) -j -s (J/ ~ '/ Jf. M cd [dd)& y_ @ -M., _ c.j = [,, y a - = - -- y i e 7-7
7.8 (4.0) State four(4) of the five(5) " entry conditions" for RPV Level Control (RPV/L),per the Emergency Procedures. (include all associated setpoints) (1.0 pts. each) (any four of the following) (a) RPV Water Level below + 13.0 in. (b) RPV pressure above 1037 psig (c) Drywell pressure above 1.65 psig (d) A condition which requires MSIV isolation (c) A condition which requires a scram and power is above.57. or cannot be determined. Ref: Emergency Procedures 1 I + 7-8
7.9 (4.0) State four(4) of the five(5) systems / components listed in PPM 5.4.1, Station Blackout, which when lost constitute initiation of the Station Blackout Emergency Procedure. ( One pt. each ) .(any four of the fc11owing five) (a) Offsite 230 KV startup power (b) Offsite 115 KV backup power (c) Diesel Generator # 1 (d) Diesel Generator #2 (e) HPCS Diesel Generator i Ref: Emergency Procedures, PPM G.4.1 END OF SECTION 7.0 7-9
SECTION 8.0 Adminestrative Procedures, Conditions and Limitations 8.1 (3.0) What is the reason or basis that Technical Specifications requires you restore Suppression Pool (SP) water temperature to below 90 degrees F within twenty-four hours, under power operating conditions? j Blowdown from an initial SP water temperature of 90 degrees F results in a water temperature of about 135 degrees F. This is below analyzed temperatures required for con.plete condensation, and ensures adequa,te,ayai_1qble,NPSij {ce_both the RHR and Core Spray pumps [}Therefore, operator action to /L'# maintain SP temperature within the specified limits ensures cec # " 4 that containment integrity will be maintained fcr all g ' "~ analyzed accidents. fy sn rany Ref: Technical Specifications 1 1 ~ k G-1
8.2 (2.0) State two times when the Control Room Operator has the authority and responsibility to initiate a reactor plant shutdown. ( 1.0 pt. each ) (a) When the CRO determinen that the safety of the Reactor is in jeopardy. (b) When operating parameters exceed any of the reactor protection circuit setpoints AND an automatic shutdown does not occur. Ref: PPM-1.3.1, Standing Operating Orders i i I l ~ r l s 8-2 . ~
t 8.3 (2.0) t Concerning the FUEL CLADDING INTEGRITY safety limits: a Overheating of the fuel cladding ts prevented by... ESELECT THE TWO(2) CORRECT STATEMENTS, listed below-1.0 pt.each] (a) restricting fuel power operation such that the reactor coolant never reaches saturated conditions. (b) restricting fuel power operation to within the nucleate boiling regime. (c) establishing a safety limit such that MCPR is not LESS THAN 1.06 for two(2) recirc. loop operation. (d) establishing a safety limit such that MCPR is not MORE THAN 1.06 for two(2) recirc. loop operation. (b) (c) Ref: Technical Specifications 9 i 8-3 s a
e 8.4 (3.5) What are the minimun requirements for rotating shift crew size in operational modes 1,2, or 3, per Technical Specifications (include the position, number. of personnel for each position, type of license required) ? (0.5 pts, each) (a) Shift Manager (SRO License on WNP2) - 1 (b) Control Room Supervisor (SRO License on WNP2) -1 (c) Reactor Operator ( Operator License on WNP2) -2 (d) Equipment Operator - 2 (e) Shift Technical Advisor - 1 (f) Health Physics Technician - 1 (g) Fire brigade members - 5 (not to include the Shift Supervisor, STA, nor three other members of the minimum shift crew necessary for safe plant shutdown and any personnel required for other essential functions during a fire emergency) Ref: Technical Specifications, PPM 1.3.2 - Shift Compliment and Functions 8-4
o B.5 (1.5) How are the following controlled keys maintained when not required for normal operation ? (Indicate the storage location and WHO controls these key sets) (0.5 pts. each) (a) Bypass and Interlock keys (b) Control Room Panel Door keys (c) Miscellaneous keys lockable cabinet) controlled by (a) Control Room (Separate SM/CRSk (Separatelockablecabinet)controlledby (b) Control Room SM/CRSF (c) Radwaste Control Room (Separate lockable cabinet controlled by the Shift Support Superviserk/- PPM 1.3.23 - Adminestrative control of plant operating keys O 8-5
8.6 (3.0) Concerning the control of overtime, per PPM-1.3.27 " OVERTIME CONTROL": (1.0 pt. each) (a) What is the longest period of consecutive hours that an individual may be scheduled to work (excluding shift turnover time)? .(b) What is the maximum number of hours that an individual may work in a 48 hour period (excluding shift turnover time)? (c) What is the minimum break time required between work periods, including shift turnover time? (a) 16 hours (b) 24 hours (c) 8 hours Ref: PPM.1.3.27 - Overtime Control ) l 8-6
= I 8.7 (3.0) Per 10 CFR 55, " Operator Licenses" (1.0 each) (a) The " Exemptions from Licenses" provisions of the Code of Federal Regulations (10 CFR 55), allow what individuals to operate the reactor controls without a license? (b) As defined in 10 CFR 55, when is an individual deemed to be operating the controls of a nuclear facility? (c) What are the " controls" defined in 10 CFR 55? (a) An individual may manipulate the controls as part of his training to qualify for an operator license under the direction and in the presence of a licensed operator or senior operator. (b) An individual is deemed to operate the controls of a nuclear facility if he directly manipulates the controls or directs anouther to manipulate the controls. (c) " controls"- apparatus and mechanisms, the manipulation of which directly affect the reactivity or power level of the reactor. Ref: 10 CFR 55 9-7
o s 9 8.8 (2.0) Per 10 CFR 20, " Standards for Protection Against Radiation": (1.0 pt. each) (a) What is a RADIATION AREA ? (b) What is a HIGH RADIATION AREA ? (a) Area (accessible to personnel) where a major part of the body could receive: 5 mrem in one hour (0.5) -or-100 mrem in 5 days (0.5) (b) Area (accessible to personnel) where a major part of the body could receive 100 mrem in i hour (1.0) Ref: 10 CFR 20 i 4 l 8-8
O o 8.9 (3. 0)- Temporary changes to procedures as described by Technical Specification 6.8.3, may be made provided what three(3) criteria are met ? (1.0 each) (a) The intent of the original procedure is not altered. (b) The change is approved by two members of th'e unit management staff, at least one of whom holds a Senior Operator license on the unit affected. 4 (c) The change is documented, reviewed.by the POC, and approved by the Plant Manager within 14 days of implementation. Ref: Technical Specifications O-9 l
O s *' ; e 8.10 (2.0) What is the basis or reason for maintaining a minimum water level of at least 22 feet over the top of the spent fuel storage pool racks and reactor pressure vessel flange, per Technical Specifications? To ensure that sufficient water depth is available to remove .99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. Ref: Technical Specifications END OF SECTION 8.0 END OF EXAM
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