ML20203N143
| ML20203N143 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 04/29/1986 |
| From: | Burdick T, Sherman J, Mary Spencer NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20203N139 | List: |
| References | |
| 50-341-OL-86-01, 50-341-OL-86-1, NUDOCS 8605050075 | |
| Download: ML20203N143 (63) | |
Text
, .
U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-341/0L-86-01 Docket No. 50-341 Licensee: Detroit Edison Company 6400 N. Dixie Highway Newport, MI 48166 Facility Name: Enrico Fermi II Examiners: M. Spencer Chief Examiner YlY.
Date J rm Datel Y'l Approved By: k, hief Operating Licensing Station Y
Dite
Examination Summary t
Examination administered on March 25-28, 1986 (Report No. 50-341/0L-86-01)
Examinations were administered to six senior operator candidates. '
Results: Three senior operator candidates passed the written examination and six senior operator candidates passed the oral / operating examination.
0605050075 860430 PDR ADOCK0500g1
DETAILS
- 1. Examiners J. Sherman, INEL M. Spencer, INEL (Chief Examiner)
- 2. Examination Review Meeting At the conclusion of the written examination, the question and answers were given to the facility training staff for review and comment. At the exit meeting the facility supplied the examiners with their comments to the Senior Reactor Operator examination. The facility comments and examiners' resolution to each have been enclosed as Attachment.
After the formal exit meeting, a round table discussion of the facility comments was conducted by the Facility Training Staff and the Examiners. l
- 3. Exit Meeting On March 27, 1986, an exit meeting was held. The following personnel ,
were present:
S. Latone, Director Nuclear Training J. Coleman, Supervisor Nuclear Training G. Preston, Operations Engineer S. Heard, Operations Coordinator J. Conen, Engineer W. Rogers, Resident NRC D. Hills, Region III NRC J. Sherman, INEL T. Bishop, INEL M. Spencer, INEL The facility was informed all candidates had passed the simulator and oral portions of the license examination. The only area of general weakness observed by the examiners was the use and control of drawings in the control room and this weakness was discussed at the meeting.
The Chief Examiner discussed the forth coming written exam and simulator exam requirements regarding the use of "K A's" and the additional reference material which will be required to produce the license examination in the future.
Attachment:
Facility Examination Comments and Examiner Resolution i
j
FERMI II FACILITY COMMENTS AND EXAMINER RESOLUTION EXAM DATE MARCH 25, 1986 The facility divided the comments to the exam into two groups:
- 1. Those questions felt to be inappropriate to the SRO knowledge level requirements.
- 2. Questions with additional or alternate answers.
The Group 1. shall be addressed first.
Facility Comment: 5.06 This question is not applicable to BWR Thermodynamic technology. In our thermo course we also talk about PWR's and pressurizer level, but we do not expect the candidate to retain this information.
Examiners' Resolution:
After review of Exam Standards, the facility comment is agreed upon.
Question 5.06 was removed and the total point value of Section 5 and the exam changed to 25.5 and 102.75 respectively.
Facility Comment: 5.16 Please see attached graphs from a simulator demonstration of two recirculation pump trip run March 27, 1986. Also see A0P 20.138.01. (20.138.01, Revision 5, Page 1)
Examiners' Resolution:
Review of supplied documents indicated the exam graphs do not model this transient at Fermi. Therefore, Question 5.16 was removed and point value of Section 5 and of exam decreased by 3.0 points, to 22.5 and 99.75 respectively.
Facility Comment: 6.06 We do not " teach" every detail of manufacture of the fuel assembly. When the operators have to assemble the fuel onsite, then they should have to know this level of information.
Examiners' Resolution: ,
Examino agrees with comment. Question 6.06 was removed. Point total for Section 6 reduced to 24.25 and the exam reduced to 98.75.
Facility Comment: 7.09 a and b We do not require memorization of EP-101 Tabs.
Examiners' Resolution:
Operator Licensing Examiner Standards, Section ES-402, Category 7 describes questions applicable to this section. This question asks for general description of the emergency condition with major symptoms'given in the question. Question remains in the exam.
Facility Comment: 7.12 We do not require memorization of 10 CFR 50.72. We keep a copy in the Control Room.
Examiners' Resolution:
In Operator Licensing Examiner Standers, Section ES-402, Category 7 describes questions applicable to this section. Also, Fermi Training Lesson Guide, 08-02-01-01, Lesson Objective (LO) No. 14, " List the events that require one hour notification to the NRC in accordance with 10 CFR 50.72." LO No. 15 addresses four hour notifications. Question remains in the exam.
Facility Comment: 7.13 We do not require memorization of EP-102 or EP-104.
Examiners' Resolution:
Operator Licensing Examiner Standards, Section ES-402, Category 7 describes questions applicable to this section. This question asks for general description of the emergency condition with major symptoms given in the question. Question remains in the exam.
Facility Comment: 8.08 The Assistant Operations Engineer and the Training Supervisor maintain records to ensure that no lapse occurs. We administrative 1y enforce a one month periodicity on standing an eight hour watch.
Examiners' Resolution:
Comment noted. Each licensee has an obligation to understand the restrictions mandated upon his/her license. Question remains in exam.
Facility Comment: 8.13 We do not require memorization of Administrative Procedures.
Examiners' Resolution:
Examiner agrees with Facility comment. Question 8.13 was removed. The point value of Section 8 was reduced from 25.0 to 23.5 and the exam total point value becomes 96.75.
2
Facility Comment: 8.14 This question would be more appropriately directed at the Health Physics Supervisor. Reference 4.10 is Procedure 61.000.03 which is a Health Physics Procedure.
Examiners' Resolution:
Operator Licensing Examiner Standards, Section ES-402, Category 8 describes questions applicable to section the SRO exam. SRO's are responsible for the overall safe operation of the licensed facility and are quizzed upon admini-strative items. Question remains in exam.
Group 2. Questions with additional or alternate answers.
Facility Comment: 5.07 A and C should both be considered correct. The attached definitions for technical specification for fraction of Limiting Power Density is the justification. (Technical specification, Pages 1-2)
Examiners' Resolution:
Telephone conversation with S. Heard on April 2, 1986, changed the above comment to Question 5.11 and the possible answers to C and D.
Do not agree with Facility comment. The new reference deals with fraction of limiting power density.
The question asked about power density, not fraction of limiting power density.
Selection c remains the only correct answer.
Facility Comment: 5.15 Procedure 20.000.07 also lists additional items. (20.000.07, Revision 4, Page 4)
Examiners' Resolution:
Comment accepted. Exam answer key modified to include those items listed as
" Indications" on new reference as additional selections for full credit.
Facility Comment: 6.05 This delta pressure is casually referred to as occurring across the core plate.
We feel that since the fuel support casting becomes an integral part of the core plate during construction. The actual component across which the delta pressure occurs is the orifice built into the fuel support casting. (C.O.K. (common operator knowledge))
3
Examiners' Resolution:
Comment noted. Question not changed because the facility did not provide acceptable reference material to substantiate a change to the answer.
Facility Comment: 6.08 A seventh detector, recently installed is " Differential Flow." (shown to Examiners at Facility) t Examiners Resolution:
The comment "(shown to Examiners at Facility)" requires explanation. During M. Spencer's tour of the simulator, the additional instrument was shown to him.
The specific reason this was identified is the examiner asked the instructor about LER's associated with RWCU system. Mr. Spencer did not check in the Control Room for this instrument.
The training material sent the examiner made no reference to the said instrument.
Do not agree with Facility comment. The answer key remains unchanged.
Facility Comment: 6.15 Additionally
- 6. Increase in Room delta temperature.
- 7. Rate and duration of sump pumpout.
(Leak Det. System Student Handout, Page 3, Revision 3)
Examiners' Resolution:
" Increase in room delta temperature" is presently implied as increase in area temperature. The " rate and duration of sump pumpout" was verified to be correct and the answer key modified.
Facility Comment: 6.18 Additionally Any RCIC Isolation is a turbine trip. (RCIC Student Handout, Revision 3, Pages 15 and 16)
Examiners' Resolution:
Comment verified to be correct and answer key modified to allow the four (4) automatic "RCIC Isolation signals" as additional answers.
4
Facility Comment: 7.03 The 22.000.17, there are no exemptions.
According to the reference procedure, From the coolant sample results, chemistry coolant sample must always be taken. An answer which will then determine if effluent gaseous samples must be taken.
is consistent with the question could be, "If the power 4.11.2.12-2, (Technical Specifications Tablechange occurs over a period of greater than one hour." Revision 2, Pages 1 and 3)
Pages 3/4; 11-9 and 11-11, 22.000.27, Examiners' Resolution:
Amount key changed to, "IF the power change occurs over a Comment verified.
period of greater that one (1) hour."
Reference changed to 22.000.17, Revision 2, Page 2 Facility Comment: 7.07 d
as well as all other GOP's states that the pressures an i ments may Procedure 22.000.03, power levels are guidelines and that actual plant conditionsiior requ re be used to determine if a step must be completed at the specified cond t on.
Unless it was specified that only one action was to be as changes:
1.
Perform surveillance on det Pump Operability
- 2. Shift to 3 element control (22.000.03, Revision 5, Pages 2, 30, and 31)
Examiners' Resolution:
Answer key is changed to allow A, C, and/or G as correct for Comment verified.
Part 5.
Facility Comment: 7.10 (12.000.13, Revision 6, Page 13)
Part d answer should be 2000.
Examiners' Resolution:
Answer key modified to change Part d from Comment verified to be correct.
100 to 2000.
Facility Comment: 7.11 Technical specifications states "two detectors operable, one in the af quadrant, one in an adjacent quadrant.
Examiners' Resolution:
Answer key modified to allow, "Two"detectors as a Comment verified to be correct. operable, one in the affected quadrant, on possibility for full credit.
5
. Facility Comment: 8.V2 12.000.07 also mandates that the requirements for normal. ingle temperature change be met. These include:
- 4. No license commitments are negated
- 5. A preliminary safety evaluation is performed
- 6. A. technical review is performed
- 7. Change is approved by OSR0 within 14 days (12.000.07,. Revision 15, Pages 23 and 24)
Examiners' Resolution:
.Do not agree with Facility comment. The question is quoted verbatim from the reference material and can be further supported by Fermi Technical Specifications, Pages 6-14.
Facility Comment: 8.03
- 1. These tags are not " controlled." They are issued by individuals and attact a to defective tools and non-plant equipment.
Examiners' Resolution:
Comment noted. The exam answer was, " check at facility." Answer to Part 1 was modified to:
Any employee may initiate-a safety tag. The supervisor in charge of the equipment responsible for removing the safety tag.
Facility Comment: 8.04 Without reference, you should also accept any reasonable parametric change.
Some additional:
- 8. Jet pump flow decreases
- 9. MG set Generator volts decreases
- 10. MG set KW decreases
- 11. "B" Recirculation pump " runback"
- 12. APRM's (all) decrease Examiners' Resolution:
Faci 14t, comment noted. Question deals with a specific set of parameters associated with this condition. Can not accept an answer without a reference.
Facility Comment: 8.13 Criticality Monitor or Area Rad Mc nitor (3/4.3.7, Technical Specifications 3/4 3-47, 48, 49)
During telephone conversation with Facility on April 7, 1986, the below were also possible answers to Question 8.13. l 6
t
- a. K(eff) less than 0.98 (Reference Technical i Specification Page 5-6)
!- b. From procedure 12.000.46, the following steps:
f 1. 6.5.3.2
- 2. 6.5.3.3
- 3. 6.5.3.4
- 4. 6.5.4.9 i
- 5. 6.5.4.11
- 6. 6.5.7.15 1
Examiners Resolution. ,
After reviewing these additional references, it was determined that question was beyond the scope of SRO knowledge requirements. The question was removed from l
the exam.
Facility Comment: 8.15 We do not require memorization of all of the footnotes in Technical Specifications EP-201-3 " accident mitigation or first aid / life saving activities" (12.000.13, Revision 6, Page 5, EP 21.000.03, Revision 2, Page 1)
Examiners' Resolution:
- l. Comment verified. Answer key modified to allow the below list of additional l possible answers.
! a. Accident mitigation Firs aid / life saving activities b.
Reference also changed to include EP-201-3, Page 1, Revision 2 Facility Comment: 8.16 By saying " shutdown" you have not been sufficiently precise regarding condition.
- a. Either < cr = 0.2 and < or = 0.1 ppm or
< or = 0.2 and < or = 0.5 are correct.
- b. If 0.2 and 0.1 are used the differences are due to the lower expected oxygen levels at operating conditions. (Technical Specification Pages 3/4 4.15, ,
Pages 1-10)
Examiners' Resolution:
! Comment verified to be correct. Answer key modified as below for full credit.
A. During power operation = 0.2 ppm (0.5)
During cold shutdown = 0.5 ppm (0.5) i
B. During power operation the catalytic effect of higher coolant temperatures requires lower limits in order to reduce chloride stress corrosion. (0.5)
OR A. During power operation = 0.2 ppm (0,5)
During hot shutdown = 0.1 ppm (0.5)
B. During power operation oxygen level in the coolant are lower than when hot shutdown, thus reducing the possibility of chloride stress corrosion. (0.5) 8
) -
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _EgSM1_2_________________
REACTOR TYPE: _gWS-QE4_________________
) DATE ADMINISTERED: _gsf9]fgg________________
EXAMINER: _gPghggBa_dz_____________
APPLICANT: _________________________
IN2IEWGI19N2_IQ_eEELICONIl 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 categcry and a final grade of at 1 cast S2%. Examination papers will be picked up six (6) hcurs after the examination starts.
% OF CATEGCRY % OF APPLICANT'S CATEGORY
. VALUE TOTAL SCORE VALUE CATEGORY Z:.So 12.27
_21.EE__ _;!_11 ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT CPERATION, FLUIDS, AND THERMODYNAMICS zo.2r ts.06
_;; 2;__ _;f;;l ___________ ________ 6. PLANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION 27.39
_24239__ _27sEl ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL L2,So 24.26
_EE 2E__ _21:5E ___________ ________ S. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 96.75
-1922Z2__ 190299 ___________ ________ TOTALS FINAL GRADE _________________%
All work done on this examination is my own. I have neither given nor received aid. .
f IEEY l 3{
i 1
PAGE 2 Et__IBEQBY_QE_dWGLEeS_EQhEB_EbeNI_QEEEeIlR05_ELWIDSt_ebQ IbEBt!QRINet!1G11 QUESTION 5.01 (1.00)
Using the below graph as a reference, BRIEFLY explain why the doppler coefficient beccmes more negative for a given temperature and varying void fractions. (1.0)
., o
- I 68., c% **
m 5 -o e
- pt* F S*.,,get a
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5 i .. .O 530 1000 1530 2000 2S00 3000 3500 4000 4600 A/te&3t FbEL TEWPt#AT68C ('F)
V214 ar.1 Twei Tet;erature Effect on Doppler Ccef fletent .
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
-. . - . A
Es__IBEQBr_QE_NWGLEBB_EQWEB_ELeNI_QEEBQI1QNm_ELWIDEx_eNQ PAGE 3 IMEBt!QQ1 Net!1GE QUESTION 5.22 (1.08)
The below figure indicates the power density at the upper portion of the core is
- 1/3 that of the lower core at EOL. List two (2) reasons for this Gif f erence in power density at EOL. (1.5) rce or coet - - - - - - - - - - - - - - - -
a Cont Ptlant LATCIN C0pt UFC so?T;w of CO*C -
P; ate oths6fy
(***** CATEGORY 05 CONTINUED CN NEXT PAGE *****)
~
Es__IMEQBY_QE_NUQLEBB_EQWEB_Eb8NI_QEEBBI1QNs_ELulDEs_8NQ PAGE 4 IMEBt!QQYNedlGH QUESTION 5.03 (1.00)
The below diagram is of (A) heatup stress or (B) cooldown stress?
CChoose either A or B 3 luutt atttA M Kl l siNMt }
g 7 % [\ !EI' TENSILE \
STRESS /, \
P systtn patsset STRES* k '
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A 0- - - .~. -
CCMPRESSIVE 9 STRESS *Ym x
(CROSS-SECTION OF VESSEL WA1.1)
(***** CATEGCRY 05 CCNTINUED ON NEXT PAGE *****)
Et__IMEQBY_QE_UWQLE88_EQWEB_ELONI_QEEBBI1QUt_ELu1DHt_eUQ PAGE 5 IMEB5QQ1NedlGE QUESTION 5.04 (1.00)
The convective heat transfer coefficient for boiling water is 300 - 900s.
(Btu /hr ftsq F). From the below list choose the values that represent the convective heat transf er coef ficient f or film boiling.
CChoose either A, B, C, or D3 (1.0)
Stu/hr ftsq F A. 5098 - 20,000 B. 388 - 9008 C. 59 - 3800 D. 5 - 20 QUESTION 5.05 (1.00)
From the selection of answers in column 1, choose the correct number identifying the horizonal and vertical axis of the curve on the next page of this exam. (1.3)
COLUMN 1 HORIZONAL = ____ 1. delta temperature
- 2. delta pressure VERTICAL = ____
- 3. log (TCclad'. -
TCcoolant3)
- 4. mass flow rate
- 5. log Q
- 6. change in entropy
- 7. */. void f ractiun QUESTION 5.06 C (1 00) 1 If a pressurized er system is operating at 705 F and 3400 psia, briefly explain what occurs in the system as temperature and pressure cre simultaneously increased at a rate of 10 units per minute.
Assume no physical restraints on the system. (1.0)
QUESTION 5.07 (1.00)
A representative value for a BWR TOTAL PEAKING FACTOR is 2.43.
What three parameters are used to determine the total peaking factor 7 (1.0)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
l O
Iece 5A 1 FILM BOLLING ONSET OF TRAN5til0N -
SOILING, C' D'
NUCLEATE 80lLING
+ A+ S C 5 : 0 7 4 E-I SINCLE-PHA0E FLOW O
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2
22__IME98Y_9E UWGLEeB 29 WEB _EkeUI_9EEB8I19&_ELWIDL_edD PAGE 6 IUE8099XU801G8 QUESTION 5.08 (2.00)
The below figure indicates the value of the void coefficient becomes more negative as reactor power increases. BRIEFLY explain why this is true.
Pt#CtNT OF volD.
0 80 20 30 40 90 to 70 to, 0 , , , , , , ,
7 40
- 9 12 -
I .44
,g... .
E% 30 .
22 -
gs... .
Vold Coeffistent vs. 5 Voids
(***** CATEGORY 05 CONTINUED ON NEXT PAGE **s**)
Ez__IBEQBY_QE NWGLEBB EQWEB_ELeNI_QEEBel10Nu_ELWIDEt_8NQ PAGE 7 IbE85QQ1Ned1GE QUESTION 5.99 (3.00)
Assume the reactor is operating at power when one SRV f ully opens.
A. What thermodynamic property of the steam remains nearly constant during the discharge to the suppression pool?
B. What will be the temperature of the steam entering the suppression pool? (State any assumptions you make for full credit!)
CUESTION 5.10 (1.00)
Why does the presence of Pu-239 late in core life cause beta effective to decrease?
QUESTION 5.11 (1.88)
Which of the below best define, " power density"?
- a. reactor power in kw divided by the total surface area of the active fuel rods.
- b. reactor power in kw divided by uranium loaded in the core.
- c. reactor power in kw divided by core volume
- d. reactor power in kw divided by the active length of fuel pins in the core.
QUESTION 5.12 (1.52)
List three intrinsic sources available for cold reacter startup. (1.5)
GUEST 10N 5.13 (1.50)
The effective multiplication for a cold, xenon free reactor with its strongest control rod withdrawn is calculated to be 0.099. What is the reacters shutdown margin? Must include unito in answer! (1.5)
(***** CATEGORY 05 CONTINUED ON NZXT PAGE stest)
Es__IbEQBY_QE_NWGLE86_EQWEB_EL8HI_QEE80I1QNs_ELUIDEs_eNQ PAGE C IdEBdQQYNed1GR QUESTION 5.14 (2.80)
Ctate whether the following changes will directly affect AVAILABLE recirculation pump not positive suction head (NPSH):
(Limit answer to: INCREASE, DECREASE, or NO AFFECT)
(Consider each change separately)
- o. Feedwater temperature increases (S.5)
D. Reactor pressure decreases (0.5)
- c. Reactor water level increases (0.5)
- d. Recirculation pump speed decreases (0.5)
QUESTION 5.15 (1.50)
I List three (3) symptoms of " core damage". (1.5)
QUESTION 5.16 (3. ) hg/ 8/ MT [/M[/gs [g During reactor oper on at 100% indicated power, t'r""7 6.L-.cn th:
re !--"1 -ti e pu , _ tri; :; n. 77jgjf, Answer the below questions concerning the above transient using the cttached transient information. C see next page of exam for transient curves 3
- a. Why is pressure increasing at point #17 (0.5)
- b. Why does total feedwater flow start to decrease at one rate then continue decreasing at a much faster rate at point #27 (1.0)
- c. Why does total steam flow decrease to zero at point #37 (0.5) i
- d. What causes the level increase at point #47 (1.0) l
(***** CATEGCRY 05 CONTINUED CN NEXT PAGE *****)
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Et__IMEQBY_QE_NUGLEBB_EQWEB_EL8NI_QEEBBIIQNi_ELUIQgi_sup PAGE 9 ISEBUQQXNedigS QUESTION 5.17 (i.es)
During degraded core conditions, why should the intermediate range monitoring system NOT be used as a vessel level instrument? (1.0) e
QUESTION 5.1F (2.00)
Define th,e following termsg
- u. Void fraction
- b. steam quality l
(
I -L ,
f '
??
/
I J i . pQ
(***** END OF CATEGORY 05 *****)
i
._.--.-_-__.__-...m-. ... _~. -
42 Ek&NI 3YSIEUS_Dggl@N 2,ggy109hA_669 lUSIBU M
ENTATION --
PAGE 10 QUESTION 6 01 (1 50)
List three (3) systems Protection System.
which receive power fr om the Reactor QUESTION 6.02 (1 0)
(1. 25)
The Level 1 List five (5)(>or of = 7.7") initiates s the auto functions. even (7) automatic functions .
QUESTION 6.03 (1.25)
(1.00)
"Which two one out Group of Pri mary con of - t wo once logic?
"tainment isolation use does NOT GUESTION 6.04 (1 0)
(1,00)
The turbine governor /p non redundant ressur e control Indicate the proper 130 VDCerelectrical system pow wi th t i s provided supplies. wo
- 1. selection to One 130 VDC puwer correctly complete the circuitbattery.
plant and is cause a supplied from the positisupplystatements is required below.for 2.
turbine Losstrip. of the power supplyve portionthe of bypass valve c trip One 130 VOC p (will OR will no tthe balanceentrol of-valve r bal ance el ays, ower supply is required
)
to cause of plant battery.and is supplied from the a turbine trip. negative Loss of the power portionoperate the turbine suppl y (willof the OR will not)
(1.0)
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62__ELeUI_5YEIEd@_Dggl@N 2_CONIB062_8ND_IUSIBUDENI611gN PAGE 10 QUESTION 6.01 (1.50)
List three (3) systems which receive power from the Reactor Protection System. (1.0)
QUESTION 6.02 (1.25)
The Level 1 (>or = 7.7") initiates seven(7) automatic functions.
List five (5) of the auto functions. (1.25)
QUESTION 6.03 (1.00)
Which one Group of Primary containment isolation does NOT use
" two - out - of - two - once " logic? (1.0)
QUESTION 6.04 (1.00)
The turbine governor / pressure control system is provided with two non-redundant 130 VDC electrical power supplies.
Indicate the proper selection to correctly complete the statements below.
- 1. One 130 VDC power supply is required for the bypass valve trip control circuit and is supplied from the positive portion of the balance-of-l plant battery. Loss of the power supply (will OR will not )
cause a turbine trip.
- 2. One 130 VDC power supply is required to operate the turbine valve relays, and is supplied from the nagative portion of the balance-of-plant battery. Loss of the power supply (will OR will not) cause a turbine trip.
(1.0)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
1
6s__EL&NI_@YSIEdS_DEg1GNz_GQUIBO(t_6NQ_1NSIBUdENI611QN PAGE 11 QUESTION 6.05 (1.90)
During high power reactor operation the majority of the core delta pressure is taken across which of the below listed core components? (1.0)
- c. core plate
- b. fuel support casting
- c. control rod guide tube
- d. the highest powered bundles in the central region f
QUESTION 6.06 1.08)
What mechanical prevents the placement of a high enriched rod in a position d gned for a low enriched fuel rod? (1.9)
CUESTION 6.07 (1.00)
The Recirculation loops are identical except for what two (2) differences? (1.0)
QUESTION 6.08 (2.08)
There are six (6) different ficw indications associated with the RWCU System in the control room. List four (4) of the six indications. (2.0) l l
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(***** CATEGGRY 06 CCNTINUED CN NEXT PAGE *****)
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- *- = - * - g--~w _ -, ,,,. , , - _ . , ,, , ., __ , _ _ _ _ _ _ _ _ _ _ _ _ _
&& PL@NT_g1STEMg_QEglGNg_GQNIBQLs_QNQ_1NgTSUdgNT@TlQN PAGE 12 QUESTION 6.09 (1.00)
The reactor power input to the Rod Sequence Control System comes from which of the below listed systems? (1.0)
A. The total steam flow signal summed in the fsedwater level control system.
B. The sum of the total f eedwater flow as compaired to total rated feedwater flow.
C. Main turbine first stage pressure D. APRM flow bias circuity QUESTION 6.13 (1.00)
The blue scram lights at the full core display are energized by:
C choose one of the below to complete the above statement 3 (1.0)
A. limit switch on the ARI valves B. limit switches on the backup scram valves C. limit switches on the scram pilot valves indication each valve is not fully CLOSED.
D. limit switches on the inlet and outlet scram valves CUESTION 6.11 (1.00)
The CRD drive water to reactor differential pressure indication instrument receives the reactor pressure input from C select one of the below to complete the above statement 3 (1.0)
A. pressure above the core plate B. RWCU inlet pressure tap C. RECIRC LOOP B suction pressure D. reactor vessel level reference leg pressure tap
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
6t__P(@NI_gy@IEdg_QE31QNg_QQNIBQLs_8NQ_INgIBydgNI8IlQN PAGE 13 QUESTION 6.12 (1.50)
List three (3) of the five (5) loads supplied by Uninterruptible Power Systems. (1.5)
QUESTION 6.13 (2.00)
Answer the below questions concerning the HPCI System.
- 1. Why is valve F002 the only AC Powered valve in the system? (0.5)
- 2. How and when is or can the HPCI turbine overspeed trip be reset?
Setpoint is required for full credit! (0.5)
- 3. The steam flow element FE N400 is located between what components?
Select answer from below list (0.5)
- a. valve F002 and valve F003
- b. reactor vessel and valve F002
- c. valve F003 and valve F001
- d. valve F0067 and valve F0068
- 4. The barometric condenser vacuum pump discharges to what system? (0.5) 1 i
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[ . -
&&__ELONI_EYSIEdE_DESIGNi_GQNIB9Li_eND_INSIBUdENIGI19N PAGE 14 QUESTION 6.14 (2.50)
List in order of flow path starting with the "T" for cooling air from the roof the below listed components of the Standby Gas Treatment System (2.5)
A. HEPA Filter B. Moisture Seperator C. Exhaust Fan D. Air Heater E. Charcoal Absorber F. Profilter G. HEPA Filter QUESTION 6.15 (2.00)
There are five (5) methods employed in monitoring various syste for leakage by the Leak Detection System.
List four (4) of the methods. (2.0)
QUESTION 6.16 (1.00)
The Main Steam Isolation valve required closing time is three (3) to five (5) seconds. List the two design requirements these operating times are based on.
(1.0)
(***** CATEGORY 26 CCNTINUED ON NEXT PAGE *****)
6 t__Eb8NI_gy@IEd@_Qg@l@Ng_QQNIBQ6t_8NQ_IN@IBUdENIellQN PAGE 15 QUESTION 6.17 (2.00)
Zdentif y whether or not the below statements concerning the Reactor Vessel Level Control System signals are TRUE or FALSE.
(2.0)
- 1. At 30% of full power feedwater flow, the rod worth* minimizer intericek is activated. CLPSP3
- 2. At less than 20% suction flow on a feedwater pump, the recirculation pumps automatically run back to 45% speed.
- 3. At greater than 35% of full power steam flow, the rod worth minimizer is no longer in the transition zone.
- 4. Assuming the level mode switch remains in "RUN", 52 seconds after a scram, the Auto Sealin of the SULCV can be reset with the "FEEDWATER LOGIC" switch.
QUESTION 6.18 (1.50)
The RCIC System turbine has four (4) automatic trips. List three (3) of these trips include the setpoints for activa. tion.
NAH3C:PY
(***** END GF CATEGORY 06 *****)
Z4__EBQGEDUBEH_:_NDBdeba_eENQBueL2_ Edge 9ENgy_eND PAGE 16 BSDIQL991G96_G9NIBQL QUESTION 7.01 (1.08)
The technical specification limiting the delta temperature between the reactor pressure vessel steam space coolant and the bottom head drain line coolant protects what components during idle recirculation pump startup? [ Select from the list below 3 (1.0)
- 1. undue stress on the vessel nozzles
- 2. recirculation pump and recirculation nozzles
- 3. undue str ess on the vessel
- 4. undue stress on the vessel orifice plate QUESTION 7.02 (1.00)
During reactor power operation a 74% rated thermal, the NSO observes a saf ety relief valve is OPEN. Assuming all attempts to close this valve FAIL, what requirements are now placed on continued reactor operation? (1.0)
QUESTION 7.03 (2.00)
Procedure 22.000.17, " Power Changes During Operation", requires samples whenever thermal pcwer changes exceed 15%. When are these samples not required? (2.0)
C'JESTION 7.04 (2.50)
List the entry conditions and setpoints for the " Level / Pressure Control" procedure number 29.000.01. (2.5)
I i
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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Z2-_ESQGEDUBE5_:_NQ85@(z_@HUQBd@(g_Ed[BQgNQy_6ND PAGE 1.
Be9196991986_G9NIBg6 l
l QUESTION 7.05 (2.00)
Complete the following chart of operational conditions.
CONDITION MODE SWITCH POSITION AVERAGE REACTOR COOLANT TEMP.
Pcwer Operation Run Any Temperature Startup (A) (B)
Hot St:- 2 , Ibide-o (C) (D)
Cold Shutdown f (E) (F)
Refueling (G) (H)
QUESTION 7.06 (2.00)
The suppression chamber shall be OPERABLE with the pool water maximum cverage temperature of 95 F during OPERATIONAL CONDITION 1 or 2.
The Technical Specifications lists two (2) exceptions to the above temperature.
List these two (2) exemptions and include the temperature limit cssociated with each exemption. (2.0)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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Zz__BBQGEDWBEB_ _NQBdeLi_eEN9BdeLi_EDEB9ENGY_edD PAGE 10 88DIRL9EIGeL_G95IB96 QUESTION 7.07 (1.50)
The Startup From Cold Shutdown To Rated Power (procedure NO.22.000.03) lists several milestones to be performed referenced to system pressure, reactor power, and or system temperature.
Column A below lists various temperatures, pressures, and power levels at which tasks identified in column B are to be perf ormed.
Match the items in column A with the task in column B that is associated with the proper parameter. (1.5)
COLUMN A COLUMN B
- 1. reactor temp 200-212 F A. jet pump operability
- 2. > / = 75 PSIG *B. start reactor feed pump
- 3. > / = 110 PSIG C. transfer from single element to three element level control
- 4. 350 PSIG
' D. close reactor head vents
- 5. 20 % power
.E. place RCIC in standby
. F. place HPCI in standby
'G. start second reactor feed pump QUESTION 7.08 (1.50)
There are three (3) entry conditions to procedure 29.000.08, Reactivity Control. Each of the entry conditions has two requirements. One of the requirements common to all i s, " Receipt of a full scram signal",
list the other three (3) entry condition requirements. (1.5)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
~
Z2__BBQGEDWBEE_:_NQBd86t_8BNQBdebt_EdESSENQY_9ND PAGE 19 689196991Ge6_G9NIE96 QUESTION 7.09 (2.00)
A. What Emergency Classification is the below listed indications? (1.0)
Annunciator "P.EACTOR BUILDING DRYWELL FLOOR DRAIN SUMP LEVEL HIGH HIGH" and both drywell floor drain sump pumps running.
OR Reactor water cleanup system leakage, as indicated by annunicator "RWCU DIFFERENTIAL FLOW HIGH-HIGH" and failure of both RWCU Isolation valves to close OR P
Annunciator "DRYWELL EQUIP DAIN LKGE HIGH" and both drywell equipment drain pumds' running B. What Emergency Condition would the above alarms represent? (1.0)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
. . 1
- 7. PROCEDURES - NORMAL 2_ ABNORMAL i_ EMERGENCY AND PAGE 23 609196991986_993IB96 QUESTION 7.10 (1.00)
Radiation exposure control cards are color coded based on quarterly cxposure as follows.
- 1. GREEN - personnel whose quarterly exposure i s l ess than --a--
MREM
- 2. YELLOW - personnel whose quarterly exposure is --b-- MREM, or greater, but less than --c-- MREM
- 3. PINK - personnel whose quarterly exposure is greater than --d--
MREM and personnel whoes quarterly exposure limit is 100 MREM due to lack of exposure history.
Select a value from column B which correctly completes each of the cbove statements.
COLUMN B c= 1250 b= 100 c= 1000 d= 2000 3000 5000 (4 3 0.25 ea = 1.0)
CUESTION 7.11 (1.00)
Core alterations are being performed in the area of SRM "C".
What SRM detecters must be OPERATIONAL in order for core alterations to continue? (1.0)
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Zs__ESQGEQW8gs_:_NQEd@Lt_8EUQBD86t_EDESGEUQX_8NQ PAGE 21 BB91969GlceL_cgNIBQ(
QUESTION 7.12 (2.00)
Identify which of the below listed events require one (1) hour notification to the NRC Operations Center. (2.0)
- 1. Any serious personnel radioactive contamination requiring extensive onsite decontamination or outside assistance.
- 2. Exposure of the feet, ankles, hands, or forearms of any individual to 37.5 rems or more of radiation.
- 3. The release of radioactive material in concentrations which, if averaged of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, would exceed 500 times the I-131 limit of 1 X 10E-10 micro-curies per millimeter in air.
- 4. Any serious personnel injury occurring on the site and requiring tranapart to an offsite medical facility.
QUESTION 7.13 (2.00)
Identify the term or terms each of the below definitions describe. (2.0)
A. Events are in process or have occurred which indicate a potential degradation of the level of safety of the plant.
No release of radioactive material requiring off-site response or monitoring are expected unless further degradation of safety systems occurs.
B. Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. Any releases are not expected to exceed EPA Protective Action Guideline exposure levels except at or near the site boundary.
QUESTION 7.14 (1.50) i During a refueling floor high radiation event, the reactor building vent exhaust rad monitor upscale trip occurred. List three (3) of the five (5) automatic actions that will occur. (1.5) r l
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Zz__BBQGEQUBES_:_NQBdebt_eaNQBdeL&_EdEBGENGY_8NQ PAGE 22 58Q1QLQQ1G86_GQNIBQL GUESTION 7.15 (1.50)
Answer the following questions concerning the Halon Fire Suppression Sytem operation. (1.5)
A. What is the time requirement to prevent system activation using the Abort Station?
B. If the Abort Station was used, the system C will OR will not J activate after the station is released. Assume the confirmation signal is still in when the station is released.
C. TRUE or FALSE Manual-Electric initiation of a Halon System is instantaneous and does not activate an evacuation delay period.
GUESTION 7.16 (2.00)
List two (2) of the three (3) actions to be performed in the control room prior to evacuation per 20.000.19. (2.0)
C The announcement is not an action 3 KiRC:PY
(***** END OF CATEGCRY 07 *****)
St__eQd1NIEIBBIlYE_ESQGEQWBEft_GQUQ1I1QUnt_0ND_ Lid 1IGIIQUE PAGE 23 QUESTION 8.01 (1.50)
List two (2) of the reasons the recirculation pump speed mismatch is cdministratively limited. (1.5)
QUESTION 8.02 (1.00)
Multiple changes to a procedure may be made during performance of the procedure. List two (2) of the three (3) conditions or requirements (1.0)
QUESTION 8.03 (2.00)
Fermi uses four (4) types of tags der equipment, personnel, and system cafety. Listed below are these four (4) types of tags. For each type listed, identify who must approve placement and removal of the tag. (2.0)
- 1. Safety Tag
- 2. Information Tag
- 3. Red tag with white lettering
- 4. White tag with red lettering QUESTION 8.04 (2.00)
The reactor is operating at
- 100% thermal power when "B" Recirc Pump tripped. List four (4) of the seven (7) indications cvailable to the Shift Supervisor of this occurrence.
QUESTION 8.05 (2.00)
List four (4) circumstances recuiring the use of a " SPECIFIC " Radiation Work Permit. (2.0)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
at add 1NIEIBOIIME_BB9GEDWBEat_GOUDIIIQUHi_9ND_ lid 1ISIlgyg PAGE 24 QUESTION 8.06 (2.00)
Other than Operation conditions 1, 2, and 3, list two (2) other conditions requiring Secondary Containment Integrity. (2.0)
QUESTION B.07 (1.50)
A. List the two (2) personnel that have the authority to postpone Independent Verification Requirements. (1.0)
..,/
B. If postoned, when must the verification be completed? (0.5)
QUESTION 8.08 (1.00)
In accordance with 10 CFR 55, " if a licensee has not been actively performing the functions of an operator or senior operator f or o p eri od of -----?-- months, or longer, he shall, prior to resuming cctivities licensed pursuant to this part, demonstrate to the Commission that his knowledge and understanding of facility operation and cdministration are satisfactory. "
Select from the below list the correct answer to ccmplete the chove statement. (1.0)
A. 2 B. 4 C. 6 D. 12 QUESTION 8.09 (1.80) l Which of the Safety Limits, if violated, does not allow two (2) hours to be in at least HOT SHUTCOWN7 (1.0) l l
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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Bi__9Dd1015IBeIIYE_EBQGEQ9BEfi_GQN91II9NHi_eND_LidIIGIIQNg PAGE 25 QUESTION G.10 (1.00)
List two (2) of the methods used to assure " adequate core cooling". (1.0)
QUESTION 8.11 (2.00)
List the conditions required to defeat or shutdown an automatic function of an ECCS. (2.0)
QUESTION 8.12 (1.50)
According to Fermi Technical Specifications, when is the Nuclear Assistant Shift Supervisor NOT required on crew?
QUESTION 8.13 .00)
List two (2) of5hethree (3) additional restrictions placed on the new fuel vault when fuel assemblies are stored there. (2.0)
QUESTION 8.14 (2.00)
The Health Physics Supervisor - Operations or designee can review and revise Radiation Work Permits based on what four (4) criteria? (2.0) l i
GUESTION 8.15 (1.50) l Under what conditions or circumstances is a Radiation Work Permit NOT
- required for entry to a "HIGH RADIATION AREA" (1.5) l t
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l -
a- AQd1NIEIBeI12E_EB9GE99BEEi_G9NQ1I1QNat_eNQ_(ldlIBIlgNS PAGE 26
-QUESTION 8.16 (1.50)
The chloride limits associated with reactor coolant are different between operational Condition 1 and shutdown.
A. What are these two (2) limits? (1.0)
B. Why are they different? (0.5)
NAS"fRSY
(***** END OF CATEGORY 08 *****)
(************* END OF EXAMINATION ***************)
i PAGE 27 Uz__IBEQBY_QE_NWGLEBB_EQWEB_EL&NI_QEEBBI1QNs_ELW1Qat_eNQ ISEB5QQYNedigS ANSWERS -- FERMI 2 -86/03/25-SPENCER, M.
ANSWER 5.81 (1.99)
As the percent voids increase the neutron travels f arther bef ore slowing down, and thus increases its chance for capture in U-238. Theref ore, the magnitude of the doppler coefficient increases.
REFERENCE FERMI - Reactor Theory Fundamentals, Chapter 8, page 8.21 and fig 3-12 ANSWER 5.82 (1.98)
This effect is the result of fuel depletion CO.53 and fission product poison build-up CO.53. (1.0)
REFERENCE FERMI - Reactor Theory Fundamentals, Chapter 13, page 13.11 and Figure .13-16. ;
l ANSWER 5.03 (1.00)
B cooldown stress (1.2)
- REFERENCE FERMI - Nuclear power Plant Thermal Sciences, Chapter 5, page 5-10 and Figure 5.3 l I
ANSWER 5.04 (1.00)
D REFERENCE l FERMI - Nuclear Power Plant Thermal Sciences, Chapter 5, page 5-29 i
(Table 5.2); Chapter 9, page 9-4.
l
E t_'_IDEQBY_QE_NWGLEBB_EQWEB _ELeNI_QEEBellDN t_ELW IDE t_eN Q PAGE 29 IBEB5QQYdedlGE ANSWERS -- FERMI 2 -86/83/25-SPENCER, M.
ANSWER 5.85 (1.98)
HORIZONAL = 3 VERTICAL = 5 ,
(2 9 8.5 ea) l <
~
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REFERENCE FERMI - Nuclear Power Plant Thermal Sci ences, Chapter 9, page 9-6.
l
- INSERT FIGURE 9.3 INTO EXAM ###########
l l ANSWER 5. 06 J (1 )
1 Temperature can only increase to 785.5 F maintain the water in a liquid state. The critical temperatur f water is 705.5 F.
CCandidate must identify change of tate or refer to critical temperature for full credit.3 (1.2)
REFERENCE FERMI - Nuclear Power Piant Thermal Sci ences , Chapter 3, page 3-13.
ANSWER 5.07 (1.00)
- 1. Radial peaking factor
- 2. Local peaking factor
- 3. Axial peaking factor SFRFERENCE FREMI -Nuclear power Plant Thermal Sciences, Chapter 10, page 10-21.
i ;
PAGE 29 h__IMEQBY_DE_tNGLE98_EQtfEB_EL8NI_QEEB8IIDL_ELVIDh_eND IMEBD99X_N951gg ANSWERS -- FERMI 2 -86/03/25-SPENCER, M.
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ANSWER 5.99 (2.80)
During low power operation, the upper portion of the core is the voided area while the lower praction of the core has the higher thermal flux.CB.53 Most of the voids occurring outside the region of maximun thermal notron flux causes little ef f ect on neutron multiplication. Therefore, the value of the coefficient is sma11.CB.53 During high power operation, tAs region of void formation has shifted further down the coolant chan a to the high thermal neutron flux area.C8.53 1
A change in void content in this region now has a large effect on thermal neutron multiplication and the value of the coeffi eng is large CO.53. ppg + ; 3 '
(2.8)
REFERENCE EtyI Chaptryi.' -?"
i FERMI - Reactor Theory Fundamentals, V'
y P M~']I'4 o :pf M(;f A 7)
,e T;r;j
.B.
.u _ ? . n.g. . ;.
ANSWER 5.09 (3.00) h k,; ' k N b_5?
[QQ A 3;i b-h]s ,
A. Enthalpy (1.0)
B. susume 545 F saturated steam throttled to atmospheric pressure (with constant enthalpy). (1.0)
Final steam temperature (f rom Mollier Diagram) is 300 F (+/- 10 F)
(Other teeperatures may be accepted depending upon assumptions) (1.0)
REFERENCE FERMI- Thermal Sciences, p. 3-17 Mollier Diagram in Steam Tables ANSWER 5.18 (1.00)
The delayed neutrons from Pu-239 are born at a higher energies the those of U-235CO.53.
Therefore, the slowing down length increased for delayed neutrons over core life, more neutrons leak out, and fewer delayed fast neutons reach thermal energies.CO.53 (1.0)
h_'_IMEQBLQE_NWGLE9B_EQWEB_ELONI_QEg!!911QN i _E(ylggs,@ND PAGE 30 IUEBUQDYBedIGE .
ANSWERS -- FERMI 2 -86/B3/25-SPENCER, M.
REFERENCE FERMI - Reactor Theory Fundamentals, Chapter 18, page 10.4 ANSWER 5, 7.1 (1.99) c.
REFERENCE Nuclear Power Plant Thermal Sciences, Chapter 18, page 18-4 ANSWER 5.12 (1.50)
- 1. photo neutrons
- 2. alph - oxygen 18 reaction
- 3. spontaneous fission of the f uel O. spontaneous fission of the Cm-242 and 244
- 5. cosmic neutrons (any 3 S B.5 ea) (1.5)
REFERENCE FERMI - Reactor Theory Fundamentals, Chapter 11, page 11.1 ANSWER 5.13 (1.50) 1 - K(eff) (2.5 pts)
SDM = ---------
K(eff)
SDM = 1 - 0.899 / 0.899 = 0.112 delta k/ k = 11.2% delta k /k
(/,[ (O g points for correct math)
(0.50 points for correct units) (1.5)
REFERENCE FERMI - Reacter Theory Fundamentals, Chapter 11, page 11.13
PAGE 31 Ei__INEQBY_QE_NWGLEeB_EQWEB_ELONI_QEEBBIIQNi_ELVID5i_eND INESDggyN951GS ANSWERS - FERMI 2 -96/03/25-SPENCER, M.
ANSWER 5.14 (2.80)
- a. DECREASE
- b. DECREASE
- c. INCREASE
- d. INCREASE (4 G 8.5 ea.)
REFERENCE FERMI - Nuclear Power Plant Thermal Sciences, Chapter 17, page 17-29 ANSWER 5.15 (1.58) C C
- M'
- 1. excessive coolant activity
- 2. excessive radiaton levels
REFERENCE FERMI - Mitigating Core Damage, page 39 y' & pu th,f h ow M O up*e'A
- c. ow.lln y pd fi304 mud Q/W
- h. w i/it y p now m da sp esA
&JA J. / M i lii gp 908m>m%D 9 ma STm p'au c419/s/c/D fadakm rMed tpamA 9 ma Srm bu fas!aTn tpdliw e/ nam /Zy M. J& d pyv hoa s u2J af w d
- n. > a dady ap non ma ya w,,
PAGE 32 U- IMEQBY_DE_ NUCLE 88_EQWEB_ELONI_DEEB8IIDNa_ELWIDEa_8ND IUEBDQDYNedIGS ANSWERS -- FERMI 2 -96/93/25-SPENCER, M.
r\
/
/
ANSWER 5.16 (3 )
/
el allowing pressure to commence to increase (0.5)
- b. Feedwater flow initially follows the decrease in steam flow [2.53 then when the RFP's trip on high level the decrease in much faster [B.53 )
/ (1.8)
\
Total steam flow decreams t4 zero due to Group One Isolation at
/
c.
24.8 inches. (B.5)
- d. Vessel leve swells du to vciding in the core [0.53 and reduction in recirculaton and ' ore flow [8.53 (1.0) l REFERENCE FERMI - BWR Tran ents BXY - 7 INSERT BXY - 7 INTO EXAM AS NEXT PAGE e * *+++++++++ m + m ++* m ++ m +*+*+*
- m +*++*+++++* m +++ m +
ANSWER 5.17 (1.00)
It is expected that the actual neutron and gamma levels will be below the range of the detectors. (1.0)
REFERENCE FERMI - Mitigating Core Damage, page 55, pp 3.3.6.2
PAGE 33 Ei__IMDBY_DE_NWGLEeB_E9W B_EL9NI_DEEBBIIONz_ELWIDE _eUD IMBd99XNeuigS ANSWERS -- FERMI 2 -86/03/25-SPENCER, M.
4 ANSWER 5.18 (2.88)
- a. void fraction is the volume fraction of steam in a steam-water mixture void fraction = volume of steam volume of steam + volume of water (1.5)
- b. steam quality is the weight fraction of steam in a steam-water mixture.
steam quality = weight of steam weight of steam + weight of water (1.0)
REFERENCE FERMI - Nuclear Power Plant Thermal Sciences, page 9-29.
F 1
i MASIBI?Y 0
=_ _ _ _ _ . _ -- - - - - - - _ - - _ _ - .
6 _'ELONI_SYSIEUE_DEH19di_GQUIBQLi_eND_INSIBWDENIGI1QN PAGE 30 ANSWERS -- FERMI 2 -86/03/25-SPENCER, M.
ANSWER 6.31 (1.58)
- 1. Process radiation monitoring system
- 2. Power range monitoring system l
- 3. Nuclear steam supply shutoff system I
- 5. Of f-Gas monitoring system (3 9 5.5 ea = 1.5)
REFERENCE FERMI - Student Handout, Electrical Overview, page 9, rev. 4 ANSWER 6.02 (1.25)
- 2. Core Spray
- 3. Starts Emergency DGS
- 4. Input ADS Logic
- 5. Input RHR Containment spray permissive logic
- 6. Trip of single speed DW cooling fans in AUTO: two speed fans shift to slow
- 7. Trip Reactor building sump pumps (any 5 9 2.25 ea =1.25)
/
REFERENCE l
FERMI - PRV Process Instrumentation System, Student Handout, page 26, rev 3 ANSWER 6.03 (1.08)
Group 1 Main Steam Isolation Valves (1.0)
PdFERENCE FERMI - Primary Containment Isolation System, Student Handout, page 5, rev. 4 I
i
h__EL8MI_EXIIEUE_DEElh_QQNIEh_@NQ_1NEIEQUgNI8Ilgd PAGE 35 ANSWERS -- FERMI 2 -86/93/25-SPENCER, M.
l c
ANSWER 6.84 (1.88)
- 1. WILL NOT
- 2. WILL (2 8 s.5 ea = 1.9)
/
REFERENCE FERMI - Governor Pressure Control, Student Hendout, page 14, rev. 3 ANSWER 6.85 (1.80) b REFERENCE FERMI - Reactor Pressure Vessel and Internals, Student Handout, page 22.
ANSWER 6.06 ( 0)
The shanks on u r end plugs are sized to prevent completing fuel assembly if a high enriched rod is located in a position designed for a lower enriched fuel rod. (1.8)
REFERENCE FERMI - Core and Fuel, Student Handout, page 8, rev 4 ANSWER 6.07 (1.00)
- 1. The RHR system can take a suction from loop B The reacor sample lines tap off the loop B (2 3 0.5 eq = 1.0) 2.
REFERENCE FERMI - Reactor Recirculatin System, Student Handout, page 4, rev. 4
Az__EkeNI_gygIgdg_Qggl@N 3_QQUIBQL3_9NQ_INSIBQNENI9IlQU PAGE 36 ANSWERS - FERMI 2 -86/93/25-SPENCER, M.
ANSWER 6.BB (2.80)
- 1. Bottom head drain flow
- 2. RWCU pump combined discharge
- 3. Return line (after regenerative heat exchangers)
O. Slowdown line
- 5. Filter domineralizer A effluent G. Filter domineralizer B effluent
[any four af the above six for full credit 3 (2.8)
REFERENCE FERMI - Reactor Water Cleanup System, Student Handout, page 10, rev. 4 ANSWER 6.09 (1.00)
C (1.0) l REFERENCE FERMI - Rod Sequence Control System, Student Handout, page 3, rev. 4 4NSWER 6.10 (1.00)
D (1.0)
REFERENCE FERMI - Control Rod Drive Hydraulic System, page 8, rev. 4 ANSWER 6.11 (1.00)
A (1.0)
REFERENCE FERMI - SK3016-B21-2 w-- - ~ m a --4 e-w w- ,p+ w - ee-w ,-m w---o~,e~ a vm - - - - -~--v" +- -, - - - - - - - - - - - - - - =
PAGE 37 h.__ELeNI_EYSIEt!E_DESION _GQNIBQLa._eNQ_INEIBut!ENIGI1QN ANSWERS -- FERMI 2 -86/03/25-SPENCER, M.
ANSWER 6.12 (1.50)
- 1. Turbine Governer Control System
- 2. Feed Water Control System
- 3. Rod Position Indication System C. ERIS-
- 5. Testibility Cabinets (any 3 9 8.5 ea = 1.5:
REFERENCE FERMI - Electrical Overview, Student Handout, page 9, rev 4 ANSWER 6.13 (2.00)
- 1. Its normal position is open and its entrol switch is keylocked
- OR ###
It is located in a normally inaccessible area (drywell) (0.5)
- 2. On overspeed, as turbine speed decreases to between 2500 and 3000 RPM, the TRIP DRAIN VALVE resets, thus resetting the turbine and allowing HPCI to reinitiate if necessary. (0.5)
- 3. b. ( g,5
- 4. Standby Gas Treatment System (0.5)
REFERENCE FERMI - High Pressure Coolant Injection System, Student Handout, pages 19, 12, and 7, rev 4
- SK3016-E41 -1
h__ELeNI_HY.EIEME_DEE10h_QQNIBQL ,_@NQ,1NgIByt!EMI@IlQN PAGE 38 ANSWERS -- FERMI 2 -86/B3/25-SPENCER, M.
ANSWER 6.14 (2.50)
- 1. B
- 2. F
- 3. D O. A or G
- 5. E
- 6. G or A
- 7. C (7 9 9.357 ea=2.5)
REFERENCE FERMI - SK3816-T46 - 1 ANSWER 6.15 (2.00) ,
g7
- 1. Excess flow
- 2. Pressure and temperature change in primary containment
- 3. Increase ir. area temperature
- 4. Radiation
- 5. Moisture outside piping (any 4 9 2.5 ea =2.0)
(g f/1TK bAlfM 50N 0 */ "r REFERENCE
' A'*!f FERMI - Leak Detection System, Student Handout, page 4 rev 3
PAGE 39 Az__ELONI_EYSIEt!E_DEH10Ni_GQNIEQLi_8ND_INSIBWt!EMIeIIDN ANSWERS -- FERMI 2 -86/03/25-SPENCER, M.
ANSWER , 6.16 (1.00)
- 1. Short enough to prevent exceeding any release limits during a steam line rupture. (3.5)
- 2. Long enough so as not to cause reactor pressure to exceed limits following a simultaneous closure of all main steam j isolation valves.
(8.5) l l
REFERENCE l FERMI - Main Steam and Bypass System, Student Handout, page 13, rev 4 J l
ANSWER 6.17 (2.20)
- 1. FALSE Cand steam flow 3 (0.5) 1
- 2. FALSE Clevel 4 also3 (8.5)
- 3. TRUE (0.5)
- 4. FALSE Cautomatic reset 3 (0.5)
REFERENCE FERMI - Reactor Vessel Level Control System, Student Handout, page 5 and 11 ANSWER 6.18 (1.50) ,
p ,, ,-
- 2. High turbine exhaust pressure 25 psig
- 3. RPV Level 8 greater than or equal to 214 inches
- 4. Low pump suction 20 inches Hg decreasing Tolerances on all setpoints is + / - 10%
0.25 pt for trip 0.25 pt for setpoint (any 3 @ 0.5=1.5) j .g (s . Y W ${,r 3 k?ArcA ' 7 & % ,cewn. S 5 3 p.s. w& ap ,, - _ m.,g
AE_EL9NI_EYSIE55_DESIE_GQNIB96t_0ND_INEIBWDENI9I19N PAGE 48 ANSWERS -- FERMI 2 -86/93/25-SPENCER, M. REFERENCE FERMI - Reactor Core Isolation Cooling System, Student Handout, page 25, rev 4 l IMB:PY
*=-www-g -w----- _ - - _ - , , _
l Z __EBQGEDWBEE_ NQBdeL _eRNQBdeL _Et!EBQENGY_eNQ PAGE 41 . l RADIOLOGICAL CONTROL 4 I ANSWERS -- FERMI 2 -86/93/25-SPENCER, M. l ANSWER 7.81 (1.80) 3. REFERENCE FERMI - Unit 2 T. S. page B 3/4 4-1 ANSWER 7.82 (1.99) If the SRV cannot be closed in two (2) minutes OR torus temperature increased to ( > /=) 95 F the mode switch must be placed to SHUTDOWN (1.8) REFERENCE FERMI - AOP 20.800.25, Failed Safety Relief Valve, page 1, rev 4 ANSWER 7.03 (2.00) co m en ria I
- 1. Anal is shows that the se equivalent I-131,concentrati CB.53 in the p mary coolant has no icreased more than a factor of CO.5 1.0)
- 2. The nobl as monitor CO.53 ows that ef fluent a tivity has at i c eased ej than a factor o 3 CO.53. (1. )
(% W N Ch44/AAA & k& w x REFER NCE FERMI - Technical Specifications, page 3/4 11-11 ANSWER 7.04 (2.50)
- 1. Reactor water level CO.53 less than Level 3 ( +172 inches ) CO.53 (1.0)
- 2. Reactor pressure CO.53 greater than 1988 psig CO.53 (1.0)
- 3. A condition which requires MSIV isolation (0.5)
{ setpoint tolerances + or - 10% } REFERENCE FERMI - 29.000.01, Level Pressure Control, page 1, rev. 2
1 Z2__EBQGEDWBEE_ _NQBdebi_eBNQBDeLa_EDEBGENGY_eND PAGE C2 86D196991G96_GQUIBQL ANSWERS -- FERMI 2 86/93/25-SPENCER, M. ANSWER 7.05 (2.00) 6 = Startup / Hot Standby D = Any temperature C = Shutdown D = 298 F E = Shutdown F = < or eq 298 F G = Shutdown / Refuel H = < or eq 148 F (8 8 8.25 ea=2.0) REFERENCE FERMI - Unit 2 Technical Specifications, page 1-10. ANSWER 7.06 (2.00) 105 F [0.53 during testing which adds heat to the suppression chamber CO.53 (1.0) 110 F [0.53 with thermal power less than or equal to 1 % of rated thermal power CO.53 (1 0) REFERENCE FERMI - Technical Specifications, page 3/4 6 - 15
~
ANSWER 7.07 (1.50)
- 1. =D
- 2. =E
- 3. =F
- 4. = B
- 5. =G j I+ , SM/Y' b REFERENCE FERMI - Proecdure 22.Z00.03, rev. 5
PRQQgDQBES_ ,NQBt!9Li_eBN9Bdebs_Et!EBGENCY_06D PAGE C3 Z2 899196991G96_E9NIBQL ANSWERS - FERMI 2 -96/03/25-SPENCER, M. ANSWER 7.80 (1.50)
- 1. Sustained APRM indication is greater than or equal to 6 % (8.5)
- 2. More than one control rod does not insert to position 94 or less. (9.5)
- 3. Reactor power cannot be determined (8.5)
(3 G B.5 ea=1.5) REFERENCE FERMI - 29.999.88, Reactivity Control ANSWER 7.89 (2.20)
- c. Alert (1.B)
- b. Reactor coolant leak rate greater than 50 gpm.
OR Loss of one Fission Product Barrier [etther answer for full credit] (1.0) REFERENCE FERMI - EP-101, TAB #9, page 6 ANSWER 7.10 (1.00) C d" '^ W a= 1000 b= 1000 c= 2000 d= )IfMO REFERdNCE FERMI- 12.000.13, Radiation Work Permit, page 13 . - - , - - ~ . - . - - - .,s ,,- . - , ..- __ ,-~
- . _ . - - _ . , - . . , , .-_~ ,-,,-_,.____. ,_,..-.- .,%--.-
--..--m ..
Zi__EBQGEDUBE E_:_NQBt!eL a_eRNQB5eLi_E5EEDENGLOND PAGE C4 BeDIQLQQ1 gel _GQUIBQL ANSWERS -- FERMI 2 -86/03/25-SPENCER, M. ANSWER 7.11 (1.89) (Chlo M " I SRM C AND SRM D OR SRM B Canswer must include two SRM channels 3 (1.8)
<My hEs Y T.wN REFERENCE
, ss.*v FERMI - Technical Specifications, page 3/4 9-3
[ Y! g ANSWER 7.12 (2.99) r=c- = Items 1 and 4 are correct. f' @ B.y ea=2.8) REFERENCE FERMI - 18 CFR 50.72 ANSWER 7.13 (2.00) A. Unusual Event B. Site Area Emergency REFERENCE FERMI - EP-102 and EP-104 ANSWER 7.14 (1.50) i
- 1. Reactor Building Ventilation System trips.
- 2. Reactor Building Div. I and Div. II Supply and Exhaust Isolation valves close. l l
- 3. Primary Containment Purge ard Vent valves close.
- 4. The Standby Gas Treatment System Auto Starts.
- 5. Control Center HVAC System aligns to the Emergency Recirc Mode.
(any 3 3 0.5 ea=1.5) l
Z2__EB9CEDWBEE_:_N9BdeLi_eRN9BueLi_EUEB9ENGLeNQ PAGE t.5 8991969919eb_G9NIBQ6 ANSWERS -- FERMI 2 -86/83/25-SPENCER, M. REFERENCE FERMI - 29.718.81, Refueling Floor High Radiation, page 1, rev 5 ANSWER 7.15 (1.58) A. 45 seconds B. will ( information only : after 150 second delay ) C. TRUE (3 3 8.5 ea = 1.5) REFERENCE FERMI - 23.501.83, Halon Fire Suppression System, page 8, rev. 4 ANSWER 7.16 (2.80) , 1. Place the reactor mode switch in the SHUTDOWN position.
- 2. Depress both Manual Scram pushbuttons.
- 3. Arm and Depress Main Turbine Trip pushbutton. (any 2 @ 1.8 = 2.8) l REFERENCE FERMI - 200.0019, Shutdown from Outside Control Room, page 1, rev. 7 l
i 1 E RCPY
Et__9Dt!INISIB8Ili'E_EBQGEDWBEfi_QQNQlI1QU$i_gNQ_LitjlI@IlgNg PAGE 40 ANSWF.RS -- FERMI 2 -86/93/25-SPENCER, M. ANSWER B.81 (1.58)
- 1. Avoid possibility of " fooling" the LPCI loop selection logic.
- 2. Backflow through the idle or reduced flow jet pumps can cause jet pump vibration.
- 3. Provides some coastdown flow from the unbroken loop during a line break accident. (any 2 of the above at S.75 ea)
REFERENCE FERMI - Reactor Recirculation System, Student Handout, page 10, rev 4 ANSWER 8.02 (1.80)
- 1. The individual changes do not change the intent of the procedure.
- 2. Each individual change is signed by two (2) knowledgeable members of plant staff, one of whom is an SRO license holder.
- 3. Upon completion of the performance of the procedure, the changes will be combined and submitted as a temporary procedure change request.
(any 2 @ 0.5 ea = 1.0) REFERENCE FERMI - 12.000.97, Plant Operations Manual Procedures, page 18 i ANSWER O.03 (2.00) CC AT n) E'J7
- 1. heck at Facility * ************
j 2. Nuclear Shift Superviser 3 System Supervisor
- 4. Nuclear Shift Supervisor (4 @ 0.5 ea = 2.0)
REFERENCE FERMI- 12.000.12, Tagging and Protective Barrier System, page 4 4 AY f
- odw & yhdd 94
/p' "'**")
- 9
Hi__8251NISIB8I1YE_ESQGEDWBEft_GQUD1I1962i_eND_L151IBI1QUE PA3E 47 ANSWERS -- FERMI 2 -86/03/25-SPENCER, M. ANSWER 8.04 (2.00)
- 1. Motor trip annunicator (3D119)
- 2. Recire system loop A operating (3D151)
- 3. Core low decrease
- 4. Rec Pump sped indicator, read down scale
- 5. Recire drive motor ammeter indicating "O"
- 6. Recire drive motor CMC switch trip light ON
- 7. Generator MWe output decreases (any 4 @ 0.5 ea = 2.0)
REFERENCE FERMI - AOP - 20.138.01, page 2, rev. 5 ANSWER 6.05 (2.00)
- 1. Contaminated Area [0.253 - greater the 10,000 dpm/100cm sq. CO.253
- 2. Airborne radioactivity areas, requiring the use of respiratory protection equipment and where MPC hours are to be recorded.
- 3. Netron radiation area exposure
- 4. High radiation area exposure
- 5. Unknown conditions in an area to be entered.
I
- 6. Maintenance of equipment, controls, or instrumentation which contain radioactive material.
I
- 7. Radiography any place wihin the owner controlled area.
(Any 4 0 0.5 ea.) REFERENCE FERMI - 12.000.13, Radiation Work Permit, page 8, rev. 6 l l L
Bi__8DDIN15IB9IIYE_EBQGEDWBES&_GQNDII1QU5i_eND_ Lid 1IBIIONS PAGE 00 ANSWERS -- FERMI 2 -86/03/25-SPENCER, M. ANSWER 9.06 (2.00)
- 1. When irradiated fuel is being handled in the secondary containment
- 2. During Core Alterations
- 3. During operations having a potential for draining the reactor vessel (any 2 3 1.8 ea = 2.0)
REFERENCE FERMI- Technical Specifications, page 3/4 6-51. ANSWER 8.07 (1.50) A. Nuclear Shif t Supervisor (0.5) Nuclear Assistant Shift Supervisor (0.5) (1.0) B. Comp 1teted prior to entering a condition for which system operability is required (0.5) REFERENCE FERMI - 12.000.43, Verification of Correct Performance of
+ Operating Activites, page 5, rev. 2 ANSWER 8.08 (1.00)
REFERENCE FERMI - 10 CFR 55.31.e ANSWER 8.09 (1.00) Reactor Vessel Water Level (1.0) REFERENCE FERMI - Technical Specifications, page 2-2 1 1
PAGE C9 Bi__eDd1NIEIBOI1YE_EBQGEDWBEEi_GQND1I1QNEi_eND_L151IBI1965 ANSWERS -- FERMI 2 -86/03/25-SPENCER, M. ANSWER 8.le (1.30)
- 1. The reactor is shutdows and the core is covered.
- 2. The core spray systems are operating and injecting into the reactor vessel at rated flow.
- 3. An adequate steam cooling flow path has been established with assurance that the core has some water level. (any 2 0.5 ea= 1.0)
REFERENCE p e_-9 2 1 FERMI - OC & P,/ Reactor ecolant inventory events, page 13 ANSWER 8.11 (2.00) A. The below must be confirmed by at least two independent indications. (1. )
- 1. Minoperation in automatic was initiated (0.5)
- 2. Adequate core cooling is assured (0.5)
REFERENCE FERMI- 29.000.22, Cooldown, page 1, rev. 3 ANSWER 8.12 (1.50) Condition 4 CO.753 or 5 te.753 REFERENCE FERMI - Technical Specifications, page 6-5
PAGE 58 Bi__eD51NIEIBOI1W_EBQGED11BESi_G2NDII1QNEi_9ND_1.151ISIIQNE ANSWERS -- FERMI 2 -86/03/25-SPENCER, M.
/ /* '
ANSWER B<1'3 (2f 80) ,
- 1. No more than six (6) rows of fuel assemblies shall remain uncovered during the loading or utiloading of fuel assemblies.
/
- 2. Fireproof metal covers shall, cover all other rows containing fuel assemblies. ,/
- 3. When loading or unloadi.ng of fuel is not in progress, fireproof metal or concrete shield plugs shall cover all rows of fuel assemblies (any 2 9 1.2 ea = 2.8)
REFERENCE FERMI - 12.800.46 Refueling Floor Conduct of Operations, page 11, rev. 3 ANSWER 8.14 (2.88)
- 1. Repeated occurrences of personnel contamination or exposures in excess of administrative limits.
- 2. A man-rem limit is exceeded.
- 3. A change in the scope of work.
- 4. A change in radiological status c4 a system of area covered by the RWP.
(4 @ 0.5 ea = 2.0) REFERENCE FERMI - 12.000.13, Radiation Work Permit, page 12, rev. 6 ANSWER B.15 (1.50) C'^ 'I li' l
- 1. Health physics personnel CO.53 or personnel escorted ay salth physics personnel CO.53 during the performance of the assigned radiation protection duties CO.53. (1.5)
REFE FERm - Technica Specififation, page 6-21. I 2.p oo. Ib Aw & ,o y 5 EP ) L 000 O Rw- D. y 9-
Hi__eDdINIEIBeI1W_ESQGEDW!!EHi_QQNQlllQNgi_9NQ_(INII@I1QNS PAGE 51 ANSWERS -- FERMI 2 -86/93/25-SPENCER, M. ANSWER 8.16 (1.50) C# * '" # " A. During power operation = 8.2 ppm (8.5) During shutdown and refueling = 0.5 ppm (8.5) B. During power operation the catalytic effect of higher coolant temperatures requires lower limits inorder to reduce chloride stress corrosion. (8. 5) REFERENCE FERMI - Technical Specifications, page B 3/4 4-2. h A Dm? pm 0 9,opm. c o. c ) 13amh bdN& 0 I ty m (c. g ) B Da9 e
-p& Um A Af s IDra & y .As ,x4 dr
% w ysa x MAST:RCCPY
< - - - - _ _ _ _ _ . _ _ _ _ , _ . , _ _ _ . - , , - , - . _ _ . . - . , . _ _ . , . - . . . _ . _ _ , _ _ _ , , _ . _ _ _ _ _ _ . . _ _ . _ _ _ _ _ , _ _ _ _ _ . . , _ _ , . _ . . _ _ . . . - . . ..m.
TEST CROS3 REFERENCE PAGE 1 QUEUTION VALUE REFERENCE
.5.
95.92 1 .... 1.80 MJS. 0.,,3 MJS8090194 NASIB0:PY 95.83 1.90 MJS0000195 95.84 1.Be MJSB008196 85.85 1.08 MJS8880197 35.86 1.98 MJS8888199 - 95.97 1.88 MJS8900199 85.98 2.08 MJS8888298 95.09 3.99 MJ58080291 35.18 1.SS MJ68008292 85.11 1.90 MJS8999283 85.12 1.59 MJS8988284 35.13 1.58 MJS80e9285 95.14 2.OS MJS8080296 95.15 1.58 MJS8999287 85.16 3.99 MJS8980298 - 85.17 1.89 MJS0000289 85.18 2.80 MJS0000210 26.58 96.01 1.58 MJSO900211 96.02 1.25 MJSSO99213 86.03 1.82 MJSB000214 96.04 1.00 MJS0000215 06.05 1.80 MJS0000216 C6.06 1.00 MJS0000217 - 26.07 1.00 MJS0000218 86.00 2.00 MJS0000220 C3.09 1.00 MJSO900221 86.10 1.00 MJS8000222 86.11 1.00 MJSO900223 96.12 1.50 MJSO900224 06.13 2.00 MJS0000226 06.14 2.50 MJS0000227 C6.15 2.20 MJS2000228 l 96.16 1.00 MJS0000229 l MJS0000232 l Ce.17 2.00 06.10 1.50 MJS0000233 l 25.25 G7.01 1.00 MJS0000234 j G7.02 1.00 MJS0000235 l 07.03 2.00 MJS0000236 07.04 2.50 MJS0000237 G7.05 2.00 MJS0000230 07.06 2.00 MJSOOOO239 G7.07 1.50 MJS0000240 07.00 1.50 MJS0000241 _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ __..__._ . _ _ . . . _ . . _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ . . _ . _ _ . . _ _ _ , _ _ _ _ _ ~ _
. s TEST CROS3 REFERENCE PAGE 2 I QUEUTION VALUE REFERENCE kj I C7.09 2.8e MJSe000242 Ig.L Th,0 l 3-1 U C7.19 1.90 MJS0000243 G7.11 1.8e MJSeee0244 G7.12 2.20 MJS2000245 ) G7.13 2.80 MJS0000246 MJS0000247 l C7.14 1.50 C7.15 1.58 MJS0000248 C7.16 2.98 MJS0098249 26.58 88.81 1.58 MJSO900259 98.82 1.90 MJSO999251 98.83 2.90 MJS8000252 98.84 2.Se MJS8988253 2.98 MJS8980254 l 98.85 98.06 2.99 MJSb880255 ' i 98.87 1.50 MJS0000256 88.08 1.00 MJS0000257 98.09 1.80 MJS0000258 98.19 1.90 MJS80ee259 88.11 2.00 MJSB080269 88.12 1.50 MJS0000261 88.13 2.00 MJS0000262-88.14 2.00 MJS0000263 28.15 1.50 MJS0000264 08.16 1.50 MJS0000265 25.50 103./5
. _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ . . _ _ . _ _ _ . _ _ _ _ _ . _ _ , _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _}}