ML20212K074
| ML20212K074 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 08/11/1986 |
| From: | Bishop M, Burdick T, Lanksbury R, Sherman J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20212K071 | List: |
| References | |
| 50-331-OL-86-01, 50-331-OL-86-1, NUDOCS 8608190151 | |
| Download: ML20212K074 (101) | |
Text
F U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-331/0L-86-01 Docket No. 50-331 License No. DPR-49 Licensee: Iowa Electric Light and Power Company IE Towers Post Office Box 361 Cedar Rapids, Iowa 52406 Facility Name: Duane Arnold Energy Center Examination Administered At: Duane Arnold Energy Center Examination Conducted: June 24-26, 1986 Examiners: # D.'Lankslury 9///[fs6 Date
., 'u .?n J. B. She an S////u Date l
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- 0. ish 8bdf Date b
Approved By d . Bu'rdick, Ch'ief Q////c4 l
Operator Licensing Section Date Examination Summary Examination administered on June 24-26,1986_,JeportNo. 50-331/0L-86-01)
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Examinations were administered'to~two reactor operator candl~d'ates and ten senior reactor operator candidates.
Results: All reactor operator and senior reactor operator candidates passed.
8608190151 860812 PDR ADOCK 05000331 V PDR ,
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REPORT DETAILS
- 1. Examiners R. D. Lanksbury, NRC (Chief Examiner)
J. B. Sherman, INEL M. O. Bishop, INEL
- 2. Examination _ Review Meeting At the conclusion of the written examination, the questions and answers were given to the facility training staff for review and coment.
Subsequent to this, on June 27, 1986, the licensee provided their coments to the Reactor Operator and Senior Reactor Operator written examinations. The facility coments and examiner's resolution to each are enumerated in Attachments 1 and 2.
- 3. Exit Meeting At the conclusion of the site visit, the chief examiner met with representatives of the facility staff. The following facility representatives attended:
D. Mineck, Plant Superintendent - Nuclear G. VanMiddlesworth, Training Superintendent R. Schlesinger, Training Supervisor - Licensed C. E. Harris, Senior Instructor J. Morris, Senior Instructor The following examiner concerns were provided to the facility:
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! a. The SR0 candidates appeared to be very weak in their ability to use l the Emergency Operating Procedures.
! b. Some candidates appeared to lack sufficient time in the Control Room
- actually performing the duties of the license for which they had applied.
- c. All of the candidates stated to the examiners that the actions of the scram procedure, IP0IS, could be performed in any order but were unable to find any documentation that would allow this,
- d. A number of SR0 candidates were unable to locate major breakers during the conduct of the plant walk through.
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.' ATTACHMENT 1 FACILITY COMENT RESOLUTION FOR DUANE ARNOLD SRO EXAM GIVEN ON 6/24/86 Facility Comment:
6.04 B System Description A-6, page 11 as referenced on the answer key states:
"The two safety valves, PSV-4403 (located on Steamline B) and PSV-4404 (located on Steamline C), discharge directly into containment atmosphere."
The question addresses LLS valves which are considered safety / relief valves and not safety valves. This too is stated in S/D A-6, P. 11.
Page 12 States:
"The safety / relief valves which discharge M the suppression pool ..."
Therefore, the answer to question 6.04B should be " Suppression Pool" instead of "Directly to Containment". In addition, the discharge is specifically to the suppression pool below water level via T-Quenchers.
NRC Resolution:
6.04 B Agree with comment. Correct answer is changed to " suppression pool".
Facility Comment:
6.07 Part B answer assumes part A has already taken place. This may be mis-leading in that part B may be answered as a separate question in itself.
If this is so, th'e answer would be, "Both pumps runback to 45%".
In addition, the answer key for part B contains 186" as part of the answer, yet this is given in the question, therefore, it is felt that vessel level of 186" need not be included in the answer to receive full credit.
l NRC Resolution:
l l 6.07 Agree with comments. Parts A and B were meant to be considered separately.
Answer for part "B" is changed to "Both recirc. pumps decrease speed to 45%".
Facility Consnent:
6.08 A Question states: "In the first few minutes"; with the failure of one steam flow transmitter, inlet subcooling will decrease initially and reactor power will decrease.
. We agree that the overal: change in reactor power is " Remain the same"
- as stated in the answer key.
However, since the question states "first few ninutes", it is felt that if the candidate states " decreases" and explains in terms of inlet sub-cooling credit should also be given.
NRC Resolution:
6.08 A Agree with comment. Will also accept " decreases due to decreased inlet subcooling".
Facility Counsent:
6.09 A The Downscale Annunciator occurs at 5/125 scale. When on Range 7 the scale range is 40. 5/125 on the 40 scale is 1.6. Therefore, on Range 7 the downscale will occur at 1.6/40.
Ranging from range 5 rading 25 to range 7, the reading will be 2.5 which is above the downscale alarm point.
Therefore, the answer key should not require "Downscale Annunciator" for credit.
i NRC Resolution:
6.09 A "Downscale Annunciator" was shown in parenthesis in Part A answer to j indicate it is not required for credit. (Downscale Annunciator) was
! deleted from Part A answer.
I Facility Comunent:
6.11 ANSWER KEY #7 Answer may also be 20" Hg ABS as stated in System Description D-4, page 17.
l NRC Resolution:
6.11 Agree with comment. Will accept either 10" H9 vacuum or 20" Hg absolute for setpoint on condenser low vacuum.
Facility Counsent:
l 7.06 8 We agree with the answer key to a degree; however, there are more correct responses than listed. ,
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4 DAEC 7.S. page 3.3-2 States:
"The control rod directional control valves for inoperable control rod shall be disarmed electrically." This in itself is one instance.
However, K! sting any condition which causes a control rod to be declared ISOP would also be one instance for each condition listed per DAEC T.S.,
then the correct response could also be:
- 1. Control rod cannot be moved with drive pressure.
- 2. Control rods with INOP eccumulators.
- 3. Control rods whose position cannot be positively determined.
- 4. Rod Uncoupled.
- 5. Full In - Full Out indication unavailable.
Any two of these five should be acceptable.
In addition, DAEC T.S. page 3.9-1, 2 states:
- 1. A control rod on which maintenance is being performed shall be considered IN0P.
- 2. Prior to performing contro'1 rod or control rod drive maintenance on control cells without removing fuel assemblies, the directional control valves shall be electrically disarmed at least on the other control rod drives in the 5.x 5 rod array centered on the control rod
- or rod drive uncergoing maintenance. Then it shall be demonstrated that the core can be made suberitical by a margin of 0.38 percent.
Therefore, theBe two are correct answers. We request that any two of this list of eight be acceptable.
NRC Resolution:
- 7.06B Agree.with comment. Answer key is changed to require any two of the
! eight responses 1.isted in facility ccmment.
Facility Comment:
i 7.07 A According to EOD-C bases, The answer should be: "Frovide early decay heat rehoval, minimizing the ultimate heatup of the primary containment."
7.07 8 The answer, " Assures cooling flow to pump" (.5) is incorrect. It should be " Avoid possible cycling of the turbine exhaust check valve" as stated in 01 50 p.6 and 01 5? p. 7.
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7.07 C According to E0P-C bases, the answer should be: " Prevent equipment damage".
NRC Resolution:
7.07 Facility comment for Parts A and C references E0P-C, but these pages are not found in the copy of E0P-C originally received by examiner. Facility answers given for Part A and C are essentially the same as answer key, so no change is necessary. Part "B" answer is changed to:
(1) Avoid cycling of turbine exhaust check valve [0.5]
(2) Assure adequate turbine lube oil pressure [0.5]
Also add reference: Arnold - 0150, p.6 Facility Comnent:
7.11 A The reason for evacuation of the Torus and Drywell on a loss of level is due to the potential radiation hazards. A loss of level causes a loss of shielding and possible high radiation as a result. There is a high probability that the leak would be to the Drywell and provide a direct path of potentially contaminated water to the torus.
NRC Resolution:
7.11 A Comment accepted. Answer key changed to " Radiation hazard from loss of shielding (0.5) and probable leak into drywell and torus (0.5)."
Facility Comnent:
8.06 According to ACP 1402.3, other notifications must be made. We agree with the answer key, but request points not to be lost if the other l notification and DR Report is listed as the answer. ACP 1402.3 Attached.
NRC Resolution:
! 8.06 Agree with comment. Notification in addition to that stated in the answer key will be allowed with no point deducation.
Facility Cennent:
8.09 A ACP 1410.3 States: "After removal of yellow duplicate pages, corrections should not be entered."
The question does not really address whether the yellow duplicate is intact or removed.
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l the answer key is correct. However, .if the yellow duplicate is removed l and a log entry is then found to be in error the operator is required to enter into the current operating log the correct entry referencing the date and time of the incorrect entry.
Request consideration be given to allow what the answer key has
-0R-
"Make correct entry referencing date and time of incorrect entry if yellow duplicate is removed."
NRC Resolution:
8.09 A Agree with comment. No point deduction will be made if answer includes situation where yellow duplicate has been removed. Answer key unchanged.
Facility Comment:
8.11 The STA could not leave the site, even if he could get back in ten minutes because the ten minute time frame is walking time, and is defined as inside the site boundary.
NRC Resolution:
8.11 The referenced administrative procedure does not specifically exclude the STA from leaving the site boundary, but rather states that all areas within ,
the site boundary are acceptable. However, credit will be given if candidate states " Areas within security fence are considered to be within ten minutes of control room" in lieu of location consideration.
Facility Comment:
8.12 A Answer key is correct per ACP 1406.3. However, according to DAEC T.S.,
no reference is made to OSS; instead reference is made to two members of plant management, at least one shall hold SR0 License. Request that either answer be acceptable for credit.
NRC Resolution:
l 8.12 A Comment accepted. However, in this case, the Administrative Procedure is more restrictive than the Technical Specification, since as OSS will always hold an SR0 license, but an SR0 license holder will not necessarily be an OSS. Answer key changed to include "Two members of plant management 5
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[ [0.5], at least one of whom holds an SR0 license [0.5]" for full credit.
Facility Comment:
8.12 B Answer key is correct but incomplete. Request that the following answers also be acceptable.
- 1. Handwritten maintenance instructions
- 2. 1400 Manual Procedures
- 3. Not allowed when revision would change function or controls of equipment as described in T.S. or USFAR.
- 4. Temporary revisions to STPs shall not allow or more identical systems or components to be tested simultaneously.
NRC Resolution:
8.12 B Agree with comment. Answer key is changed to include:
(1) Handwritten Maintenance Instructions (2) 1400 Manual Procedures (3) Temporary revisions to STP's shall not allow two or more identical systems or components to be tested simultaneously, in addition to the existing answers in key.
Facility Comment:
8.13 A Candidate may also have included time requirements for INE, Emergency Class events as addressed in ACP 1402.3. Request credit be given if this is done.
NRC Resolution:
8.13 A Full credit will be given only if appropriate information regarding notification to NRC is given. No credit will be deducted if candidate also lists additional notification (state and local authorities, plant management). Answer key unchanged.
NRC Changes in Grading Breakdown 8.03 A Due to the fact that the safety limit MCPR was mentioned in the question, the 0.5 point credit assigned to " safety limit MCPR" in the answer key is shifted to " increased uncertainty".
6.01 B The point breakdown is changed to as follows: To remove heat generated by radioactive particles [0.5] loaded on HEPA filters [0.25], charcoal absorbers [0.25]. The heating mechanism is considered more critical than the specific components.
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6.08 B The second portion of the answer, "then return to the same as initially when level equilibrates at lower point" is only required if candidate stated "No Change" as the answer for 6.08A.
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r a ATTACHMENT 2 FACILITY COMENT RESOLUTION FOR DUANE ARNOLD R0 EXAM GIVEN ON 6/24/86 Facility Comunent:
2.08 B System Description A-6, page 11 as referenced on the answer key states:
"The two safety valves, PSV-4403 (located on Steamline B) and PSV-4404 (located on Steamline C), discharge directly into containment atmosphere."
The question addresses LLS valves which are considered safety / relief valves and not safety valves. This too is stated in S/D A-6, p. 11.
Page 12 States:
"The safety / relief valves which discharge g the suppression pool . . ."
Therefore, the answer to question 6.04B should be " Suppression Pool" instead of "Directly to Containment". In addition, the discharge is specifically to the suppression pool below water level via T-Quenchers.
MRC Resolution:
2.08 B Agree with comment. Correct answer for Part "B" is changed to " suppression Pool".
I Facility Comunent:
3.03 Part B answer assumes part A has already taken place. This may be misleading in that part B may be answered as a separate question in itself. If this is so, the answer would be, "Both pumps runback to 45%".
In addition, the answer key for part B contains 186" as part of the answer, yet this is given in the question, therefore, it is felt )
that vessel level of 18E" need not be included in the answer to l receive full credit.
NRC Resolution:
3.03 Aaree with comments. Parts A and B were meant to be considered separately.
Answer for part "B" is changed to "Both recirc. pumps decrease speed to 45%".
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r Facility Comment:
3.04 A Question states: "In the first few minutes"; with the failure of one steam flow transmitter, inlet subcooling will decrease initially and reactor power will decrease.
We agree that the overall change in reactor power is " Remain the same" as stated in the answer key.
However, since the question states "first few minutes", it is felt that if the candidate states " decreases" and explains the terms of inlet subcooling credit should also be given.
NRC Resolution:
3.04 A Agree with comment. Will also accept " decreases due to decreased inlet subcooling" for part A.
Facility Comment:
3.05 A The downscale Annunciator occurs at 5/125 scale. When on Range 7 the scale range is 40. 5/125 on the 40 scale is 1.6. Therefore, on Range 7 the downscale will occur at 1.6/40.
Ranging from range 5 reading 25 to range 7, the reading will be 2.5 which is above the downscale alarm point.
Therefore, the answer key should not require "Downscale Annunciator" for credit.
NRC Resolution:
3.05 A "Downscale Annunicator" was shown in parenthesis in part A answer to indicate it is not required for credit. "(Downgrade Annunicator)" was deleted from part A answer.
Facility Comment:
3.06 A The sequencing of the RHR pump start is the same now for both conditions, normal power available and LOOP.
The correct answer for A is "A and B pumps start 10 seconds after LOCA signal detected. C and D pumps start 15 seconds after LOCA signal detected.
l (Per Telecon with Bob Schlesinger 7/2/86) 2 i
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NRC Resolution:
3.06 A Agree with comment. Correct answer for part "A" is changed to A and B pumps start 10 seconds after LOCA signal detected [0.375] C and D pumps start 15 seconds after LOCA signal detected [0.375]. Point value for Part B is reduced to 0.75 for consistency.
Facility Comment:
3.09 ANSWER KEY #7 Answer may also be 20" Hg ABS as stated in System Description D-4, page 17.
NRC Resolution:
3.09 Agree with comment. Will accept either 10" Hg vacuum or 20" Absolute for setpoint on condenser low vacuum. .
Facility Comment:
4.08 B We agree with the answer key to a degree; however, there are more correct responses than listed.
DAEC T.S. page 3.3-2 States:
"The control rod directional control valves for inoperable control rod shall be disarmed electrically." This in itself is one instance.
However, listing any condition which causes a control rod to be_ declared INOP would also be one instance for each condition listed per DAEC T.S.,
then the correct response could also be:
- 1. Control rod cannot be moved with drive pressure, i
- 2. Control rods with IN0P Accumulators.
- 3. Control rods whose position cannot be positively determined.
- 4. Rod Uncoupled.
- 5. Full In - Full Out indication unavailable.
Any of these five should be acceptable. !
NRC RESOLUTION:
4.08 B Agree with comment. Answer key is changed to require any two of the eight responses listed in facility comment.
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Facility Comment:
I 4.09 A. According to E0P-C bases, The answer should be: " Provide early decay heat removal, minimizing the utlimate heatup of the primary contain-ment.
B. The answer, " Assures cooling flow to pump" (.5) is incorrect. It should be " Avoid possible cycling of the turbine exhaust check valve" as stated in 0I 50 p. 6 and 0I 52 p. 7.
C. According to E0P-C bases, the answer should be: " Prevent equipment damage".
NRC Resolution:
4.09 Facility comment for Parts A and C references E0P-C, but these pages are not found in the copy of E0P-C originally received by examiner. Facility answer given for Part A and C are the same as answer key. Part "B" answer is changed to:
(1) Avoid cycling of turbine exhaust check valve [0.5]
(2) Assure adequate turbine lube oil pressure [0.5]
- Also add reference
- Arnold - PISO, p. 6 l
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U. S. NUCLf!AR REGULATORY COMMISSION MASTERC0PY REACTOR OPERATOR LICENSE EXAMINATION
?ACILITY: _DyANE_ABNQLD___
REACTOR TYPE: _gWB-@E4_ _
DATE ADMINISTERED: _@6206/23__ _
EXAMINER: _gdEBM8N z _d._ _ =__
APPLICANT: _________ _ _ _ __ _ __
INSIBUGII9NS_I9_8EELIG8 NIL Uza separate paper for the answers. Write answers on one side only.
Stcple question sheet on top of the answer sheets. Points for each qu;stion are indicated in parentheses after the question. The passing grcde requires at least 70% in each category and a final grade of at losst 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY
__Y8LUE_ _IDIBL ___EG9BE___ _YBLUE__ ---- CBIEGQRY____ _ _ _ _ _
_2EzE@__ _2Ezl2 = = _ - - - -- -
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
_2Ez2E__ _2dz2d __ -
=
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
_26s@@__ _25z6@ _ _ __
________ 3. INSTRUMENTS AND CONTROLS i _2dsg0__ _23z20 ___________ ______
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 101z2E__ 109t@0 _____ ____
TOTALS ,
l .
l FINAL GRADE __ _ _ __ _ _ __-- ___%
! All work done on this examination is my own. I have neither
- givsn nor received aid.
l
_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ . - - - - - - =----- _____
APPLICANT'S SIGNATURE l
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PBINQ1 ELE @_QE_NUGLg88_EQWEB_EL8NI_QEEB8IlgNi PAGE 2
.. IMEBdRDYNedIGHz_UEAT TB8MSEEB_8NQ_ELUlp_ELQW QUESTION 1.01 (1.50)
Th3 attached figure 15.1 shows how K-excess varies over a typical core cycle.
Briefly explain the PREDOMINANT CAUSE of the change in K-excess between the following points:
- a. A and B
- b. B and C
- c. C and D DUESTION 1.02 (2.00)
Aseume a large centrifugal pump is operating normally with the discharge vc1ve fully opened. What changes will occur in the below parameters if the discharge valve is throttled to the 50% open position? (Limit your answer to INCREASE, DECREASE, or NO AFFECT.)
A. Pump motor current B. Pump discharge head C. Available Net Positive Suction Head (NPSH)
D. Pump flow rate QUESTION 1.03 (1.00)
Explain why a control rod will have higher worth at high reactor coolant tcmperature than at low reactor coolant temperature.
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QUESTION 1.04 (3.00) '
Plutonium isotopes are generated in the fuel during reactor operations.
r EXPLAIN WHY plutonium buildup changes the following during a core cycle:
(Also SPECIFY the DIRECTION OF CHANGE and WHICH plutonium ISOTOPE is prcdominant for each) i i A. Beta effective i
B. Fuel temperature (Doppler) coefficient of reactivity *
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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iz__EBING1ELES_9E_N99LE8B_EDWEB_EL8NI_QEEB8IIQNi PAGE 3
.. IBEBdQDYNed1GHz_UE81_IB8NSEEB_8NQ_E(QIQ_ELQW l
j QUESTION 1.05 (1.50)
F In the event of a LOCA after extended operation at full power, which
(
' fuel rods (EDGE or CENTRAL) are more likely to exceed the 2200 deg F limit? Explain WhY.
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1 QUESTION 1.06 (3.00) f R garding the xenon transient following a significant DECREASE in reactor power from high power operation:
A. Briefly, EXPLAIN WHY the xenon concentration will peak following the 9
manuever. (1.0)
B. How will peripheral control rod worth be affected (INCREASE, DECREASE, REMAIN THE SAME) during the xenon peak? Briefly EXPLAIN your answer. (1.5)
C. If the decrease in reactor power was from 100% to 50%, would the new (50% power) equilibrium xenon reactivity be MORE THAN, LESS THAN, or EDUAL TO one half the 100% equilibrium value? (0.5)
QUESTION 1.07 (1.50)
During a reactor startup, Keff is .95 when the SRM channels read 100 cps.
b! hot will the new Keff be when the SRM channel reads 270 cps?
(SHOW YOUR WORK)
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QUESTION 1.08 (1.00) l l Consider reactor fuel with exposure of 10,000 mwd / ton. Which of the L
l following is the major intrinsic neutron source (CH,00Sd CNE):
A. (alpha, n) reaction with 0-18 B. (alpha, n) reaction with Be-9 C. Spontaneous fission of Cm-242 l
l D. Gamma-neutron reaction with deuterium l
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1z__EBING1ELES_9E_NWGLE88_E9 WEB _ELONI_QEEB8IIgNs PAGE 4
., ISEBdQDYN8dIGHz_ HEAT TBONSEEB_8ND_ELylp_E(gW
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l QUESTION 1.09 (1.00)
A typical fuel assembly contains two " water rods". Describe HOW water rods affect the assembly in terms of reactor physics.
QUESTION 1.10 (2.00) 83WcR PLA/JT Tha attached Figure 17.2 shows a temperature versus entropy, diagram for an IDEAL power cycle. State the PHYSICAL PROCESS and REACTOR 4 COMPONENT (S) cocociated with that process for each of the following transitions:
A. Point 1 to Point 2 l
B. Point 2 to Point 3 C. Point 3 to Point 4 D. Point 4 to Point 1 QUESTION 1.11 (2.50)
, A design feature in the reactor vessel ensures proper flow distribution through the fuel bundles.
- c. Wh'at is this feature (or component)? (0.5)
- b. . EXPLAIN why this f eature is necessary in a BWR, including the consequences of a power increase if this feature were eliminated. (2.0)
QUESTION 1.12 (2.00)
True or False.
For a ennstant reactor period, it takes the SAME AMOUNT OF TIME to change rccctor power from 1% to 5% as it does to change it from 10% to 50%.
EXPLAIN YOUR ANSWER FULLY. SHOW THE CALCULATION YOU WOULD USE TO VERIFY YOUR ANSWER.
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
,, .o 1 :
l- Iz__EBIN91ELES_9E_NUGLEBB_E9 WEB _EL@NI_QEEB@IlgNz- PAGE 5 t ,. IUEBdQDYN@dlGQs_ME8I_IB@N@EEB_@ND_ELylg_ELQW QUESTION 1.13 (1.50)
) Tha effects of " Doppler Broadening" in U-238 result in a modified capture cross-section curve, but the area under both the original and the (higher tGzperature) broadened curve will theoretically be the same. Thir rrrr: tc
-4 rdi c a* = + k a b @veral l neutron capture by U-238 will be about the same at
, cny temperature.
Ic~this TRUE or FALSE 7 EXPLAIN your answer.
QUESTION 1.14 (2.00)
Tha attached figure 15 shows the " Normal Operating Map" of rated core flow vcrsus reactor power.
- c. EXPLAIN WHY, on the natural circulation line, incremental increases in power initially produce very rapid increases in core flow, but eventually reach a point where further power increases produce no increase in core flow. (1.5)
- b. EXPLAIN WHY, on the pump constant speed line, core flow increases as reactor power decreases. (0.5) 9 9
4 4 9
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(***** END OF CATEGORY 01 *****)
us 1:
. 2t__E(6NI_QESIGN_ INCLUDING S8EETY AND_gMERGENCY_ SYSTEMS PAGE 6
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OUESTION 2.01 (3.00)
R;garding the Standby Gas Treatment System:
j+
- o. State FOUR conditions which will automatically initiate the SGTS trains. Include setpoints. (2.0)
- b. EXPLAIN WHY it may be necessary to provide continued cooldown flow after the train is no longer required (assume the train
-has been operating for some time). (1.0) i OUESTION 2.02 (3.00)
! o. State FIVE conditions which will cause the emergency diesel generator engine shutdown relay (SDR) to be energized.
(Setpoints not required) (2.5)
- b. State the EDG engine protective condition which IS NOT overridden
! when the emergency start relays ESA and ESB are actuated. (0.5)
OUESTION 2.03 (1.00)
Which of the following is TRUE concerning the Standby Liquid Control System: (CHOOSE ONE.)
- c. In the event a remote (outside control room) reactor shutdown is
-required, SBLC injection can be actuated by the local pump START switch.
- b. The pumps may be operated simultaneously if necessary to shutdown the reactor in an ATWS. - >
- c. When the control room handswitch is placed to " PUMP A RUN",
the "A" pump starts and all squib valved fire.
- d. Nitrogen-charged accumulators assure adequate suction pressure for the pumps.
O
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2.___ PLANT _QESIGN_INGLUQINg_gAEETY AND_EMER@gNgy_gygTEMS PAGE 7 1
I QUESTION 2.04 (2.50)
- c. Two sets of vacuum breaker valves are provided on the primary containment. One set reli eves f rom the ____7 to the 7____.
The other set relieves from the ? to the ____?__ .
The fully open setpoint for each is 7 and ____?_
respectively. (1.5)
- b. Why are these vacuum breakers required? (1.0) 4 QUESTION 2.05 (2.00)
Idantify which of the following are direct HPCI turbine trips and which are HPCI system isolations:
- 1. Low pump suction pressure
- 2. High reactor vessel water level
- 3. Low steam supply pressure i 4. Steamline high differential pressure QUESTION 2.06 (1.25)
' Listed below are a number of components of the Reactor Water Cleanup System. Place the components in order of normal FLOW PATH beginning with tha comprenents downstream of the inlet isolation valves. (One component MUST be used twice for full credit.)
- c. Filter demineralizer
'b. Pumps
- c. Regenative heat exchanger
- d. Non-regenative heat exchanger .
QUESTION 2.07 (2.00)
Daccribe TWO WAYS rupture discs would prevent overpressurization of tha RCIC turbine exhaust piping. (Setpoints not required.)
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25__EL9NT_QggIgN_lNGLUQINQ_g8EETY ANQ_gMERQgNCY gYSTEMS PAGE 8 f, -
QUESTION 2.08 (2.00)
{ Rsgarding the Low-Low Set (LLS) valves:
- c. What is their purpose? (0.5)
- b. Where do they discharge? (0.5)
- c. What are the TWO actuation signals? (Secpoints not required.) (1.0)
QUESTION 2.09 (2.00) t Define the following terms and briefly explain why each must be minimized
! in a BWR.
t
- a. Carryover t b. Carryunder QUESTION 2.10 (2.00)
The speed control- signal to the scoop tube places it at mid position prior to recirculation M-G set starting. What are the TWO factors that require this positioning of the scoop tube?
QUESTION 2.11 (2.00)
When both recirculation loops are operating, the loop jet pump flow signals cre added together to give a Reactor Jet, Pump Total Flow Signal.
- c. How is jet pump total flow determined if one recirculation loop is operating and the other loop is shut,down?
- b. WHY is a dif f erent method used when one recircu'lation loop is operating and one shutdown?
I i
9
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
ns 1:
ag_ EL.8NI_QEgl@N_INg69QIN@_S8EgTY AND_EdEB@ENCY_@YSIEd@ PAGE 9 QUESTION 2.12 (2.50)
Choosing from the list of choices below, identify which SPECIFIC Control Rod Drive component is installed to allow each of the following:
$ 1. Minimize pressure transients during rod motion.
I
- 2. Prevent the index tube and control rod from accidentally moving downward.
- 3. Avoid excessive initial rod withdrawal speed following a scram.
- 4. Two sources of motive force to scram a rod.
i 5. Create a hydraulic buffer action to cushion the drive at the end of its scram stroke.
LIST OF CHOICES:
- a. Ball check valve f. Equalizer valves
- b. Coupling spud g. Collet assembly
- c. Stabilizer valves h. Spectacle flange
- d. Velocity limiter i. Bellville spring washers
- e. Piston tube orifices J. Flow control valves
\
9
(***** END OF CATEGORY 02 *****)
ns !!
. 3. INSIBUMENIg_9ND_CONIBQLS PAGE 10 1
QUESTION 3.01 (1.50)
Stcte what specific RPS action (if any) will occur directly from MSIV position in the RUN mode if the following main steam lines shut:
- a. Lines B, C, and D
- b. Lines B and C
- c. Lines A and B QUESTION 3.02 (3.00)
R5garding the LPCI Loco Selection Logic:
- a. HOW does the logic determine how many recire pumps are running? (1.0)
- b. HOW does the logic determine which is the UNDAMAGED recirc loop? (1.5)
- c. If the logic determines that neither loop is damaged, WHICH loop will it select for LPCI injection? (0.5)
QUESTION 3.03 (2.00)
Tha reactor is operating at 100% power. =ith r; ircul_etir- : r.trel ir m2cter-22..;;1. What will be the effect on/of the recirculation flow control cystem due to the following conditions: (STATE ANY ASSUMPTIONS YOU MAKE.)
- c. The full open indication on recirc pump "B" discharge valve is lost.
NOTE: The recirc pump discharge bypass valve is open.
- b. The reactor feedwater pump "A" trips. (Assume during the subsequent transient reactor water level falls below 186 inches)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
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1 i -
}. g. INSIBUMENIS_8ND_CONIBg6S PAGE 11
} ..
I QUESTION 3.04 (3.00)
DESCRIBE the changes (if any) in the following parameters in the FIRST FEW MINUTES following the low f ailure of one steam flow transmitter. (Assume 100% reactor power, three element FWLC, and no operator action.)
- c. Reactor power
- b. Feedwater flowrate
- c. Reactor water level QUESTION 3.05 (2.50)
For each of the IRM (Intermediate Range Monitoring) range changes (a. and b.) listed below, provide the following:
(Mode switch in STARTUP):
- 1. The indicated level on the NEW RANGE.
- 2. Any automatic actions initiated as a result of the indicated level on the new range.
- a. Switching from range setting 5, reading 25, up to range setting 7. (1.0)
- b. Switching from range 6, reading 39, down to range 5. (1.5)
QUESTION 3.06 (1.50) i Ansume a LPCI injection is called for and all RHR pumps are available.
State the order of automatic pump starting given the following conditions: (Include applicable time delays.)
- c. Normal power is available (0.5) l b. A Loss of Offsite Power (LOOP) has occurred'. (1.0) e i
I l
l (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
,,a a i
V ..
3._ INSTBUMENTS ANQ_QQNTRQLS
, PAGE 12 I QUESTION 3.07 (1.50)
R garding the Reactor Mode Selector Switch RPS Trip:
- a. When is the trip automatically reset?
- b. What damage could result if the SCRAM were not ru'- -Air:11y reset?
QUESTION 3.08 (1.50)
During a failure of instrument and service air event, state what apparent chrnge (if any) will occur for each of the f ollowing INDICATIONS:
(Limit your answer to INCREASE, DECREASE, or NO CHANGE.)
(Assume no operator action is taken.)
- c. Steam seal regulator pressure.
- b. Final feedwater temperature.
- c. Standby liquid control tank level.
OUESTION 3.09 (3.00)
Lict SIX signals, including the SETPOINT, which will cause an automatic icolation of the Main Steam Isolation Valves (MSIV's).
QUESTION 3.10 (1.00)
Choosing from among the schematic representations in Figure 10, dstermine which of the three '( A , B, or'C) represents:
- 1. Reactor Protection System initiation logi.c
- 2. High Pressure Coolant Injection initiation logi'c NOTE: Contacts are shown in their NORMAL position (No initiation present) ,
OUESTION 3.11 (2.00)
What are FOUR of the FIVE automatic actions which occur at the Low-Low Wster Level Trip -- 119.S"? *
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
ers et
2 3__IUSIBUMENIg_8ND_gDMIBOLE PAGE 13 1
I l
QUESTION 3.12 (2.00)
Shown below is a table of some reactor level instrument types and ranges.
Complete the table with the appropriate information.
INSTRUMENT RANGE COMPENSATION CALIBRATED 1 I I NR Barton i 158" to 218" ! A ---
! Hot NR Gemac l --B----- I Pressure i Hot WR Yarway I 8" to 218" 1 -----C- ! Hot WR Yarway - ! l I Fu21 Zone 1 -100" to +200" i None ! -----D I I I (Note: Part,"D" has I ! I two conditions)
QUESTION 3.13 (1.50)
The attacheu Figure 5 shows how pulse height varies with applied voltage for gas filled detectors. Choosing from among the six " regions" of applied voltage on the drawing, identify which REGION each of the following dstectors normally operate within.
- a. Source Range Monitor (SRM)
- b. Local Power Range Monitor (LPRM)
- c. Area Radiation Monitor (ARM) 9 4 e I
t 0
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(***** END OF CATEGORY 03 *****)
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- 4. PBgGEQUBEg_ _NQBd86t_8pNOBd@bz_EMEBGENCY ANQ 4
PAGE 14
..' B6D196991G96_GQNIBQL ,
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\
QUESTION 4.01 (2.00) i
\
According to Integrated Plant Operating Instruction No. 2 (startup):
a.- What are TWO ways to correct a situation where steam dome to bottom head differential temperature exceeds 145 F?
- b. Shortly after reaching criticality, you notice the highest reading SRM is approaching 10 E5 cps. What specific response are you required to verify on the IRM's?
- L DUESTION 4.02 (1.50) $
According to Integrated Plant Operating Instruction No. 2 (startup):
, s. What is the maximum reactor power level'at which mechanical vacuum pump operation is allowed? (0.5)
- b. State TWO reasons for limiting mechanical: vacuum pump operation to a specified power level. (1.0)
QUESTION 4.03 (2.00)
- e. STATE the exposure rate limits which characterize the following: (1.5)
- 1. Radiation Area
- s
- b. Assuming you have an approved NRC Form 4 (exposure history) on file, WHAT is the maximum whole body exposure you are allowed in ONE DAY without additional authorizations. (Also assume it is the '
s first day of a new calendar quarter.) '
(0.5) f
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(***** CATEGORY 04 CONTINUED'ON NEXT PAGE *****)
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' PAGE 15
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QUESTION 4.04 (2.50)
- c. According to the Tagout procedure, what THREE actions are you required j
to perform if while on a work assignment a Warning Tag or Hold Card is found to be missing? (1.5)
- b. Briefly explain how the use of the Warning Tag and Hold Card differs. (1.0)
.m
'OUESTION 4.05 (1.50) s According to procedure you are instructed NOT to secure or override an ECCS system automatic initiation unless WHAT conditions exist?
. QUESTION 4.06 (2.00)
- a. What indications would INITIALLY alert you to the following conditions:
- 1. Recirc pump cavitation (1.0)
- 2. Jet pump cavitation (0.5)
- b. What IMMEDIATE ACTION are you required to take if jet pump cavitation is indicated? (0.5) a i
QUESTION 4.07 (3.00)
, - Concerning the recirculation pump speed mismatch limitations (refer to l ettached Figure 20): .
j o. STATE the TWO operating limitations. '
l
! b. WHAT is the PURPOSE of the mismatch operating limitations?
I l c. Is operation expected to occur in the " PROHIBITED" region at l
any time? Briefly EXPLAIN.
I I
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(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
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I 1 . _ . _ __ __ . - . . _ _ - - - - ----- - - - - - - - - - - - - - -
3 ~
BBgggQUBgS - NQBMAl _ABNQRMALs_EMERGENgY_ANQ i PAGE 16
.. B6D1969GIGeL_GQNIBQL i
QUESTION 4.08 (2.50)
- a. How is a control rod drive electrically disarmed? (1.0)
- b. What are TWO instances when a control rod drive would be electrically disarmed? (1.0)
- c. During reactor power operation, the number of inoperable control rods shall not exceed ____7 .
(0.5) f QUESTION 4.09 (2.50)
Briefly explain the REASON (S) for each of the following cautions in the Emergency Operating Procedures:
- a. Whenever RHR is in the LPCI mode, INJECT through the heat exchangers as soon as possible. (1.0) i
~
- c. Manually trip Standby Liquid Control pumps at 0% in the SBLC tank. (0.5)
QUESTION 4.10 (2.00)
~
Assume the reactor has scrammed from extended power operation, and the MSIV's are shut. According to Integrated Plant Operating Instruction No. 5 (Reactor Scram), what immediate operator actions or checks may be
, teken to CONTROL REACTOR PRESSURE. .
(LIST FOUR)
QUESTION 4.11 (1.50)
Lict the THREE conditions that would cause you to enter EOP-1, RPV Control, following a reactor SCRAM.
O
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
,a it
- 4. PROCEDURES - NORMALa_8ENgBM862_EMEB@ENQY_8NQ PAGE 17
.. 88DIQ6991G86_GQNIBg6 P
QUESTION 4.12 (1.50) i Briefly explain the REASON for each of the following cautions in Integrated Plant Operating Instruction No. 4 (Shutdown):
- a. Insert the SRM detectors one at a time.
- b. The reactor should not remain in a hot shutdown condition longer than necessary.
- c. Maintain RPV level below +211 inches.
i ,
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(***** END OF CATEGORY 04 *****)
i (************* END OF EXAMINATION ***************)
EQUATION SHEET y = s/t Cycle efficiency = (Metwork f = ma ,
out)/(Energy in) 2 w = sj s = Va t + 1/2 at g= x- ~
I A = LM A = Ag e **
KE = 1/2 av a = (Vf - V,)/t PE = agn w = e/t x = sn2/tifg = 0.693/t1/2 i Vf = V, + at y,y j -
nD 2 t
1/28ff=[(tus)(tw)3 A= , ((c1/2)
- I*SI3 l aE = 931 am -Ex m = V ,yAo ,
Q = mCoat
- I = I g e~"*
6 = UAaT Pwr = Wfsh I = I ,10** E TVL = 1.3/u HVL = -0.693/u P = P 10 sur(t)
P = Po e*/I SG = 5/(1 - K,g)
! SUR = 25.06/T G x= 5/(1 - K ,g ,)
SUR = 25e/s* + (s - o)T G;(1 - K,ff3) = G 2EI ~ "eff2)
M = 1/(1 - K,g) = CR;/G, T = (t*/o) + [(a - oV io]
l .
M = (1 - K ,g ,)/(1 - K ,ff3) l 7 = a/(s - a)
T = (a - o)/(Io) SOM = ( - K,g)/K,g t' = 10 seconds o = (K,ff-1)/K,g = 4K,g/K,g
, I = 0.1 seconds ~I o = [(t=/(T K,ff)] + [T,ff /(1 + IT)] ,
Id Id l
Id j1 *2 =2 Id g2 2
e = (taV)/(3 x 1010) ,
2 I = *N R/hr = (0.5 CE)/d (meters)
R/h: = 6 CE/d2 (f,,g) .
Miscellaneous Conversions Water Parameters 1 gal . = 8.345 lors.
1 curia = 3.7 x 1010dos I kg = 2.21 lem 1 gal. = 3.78 11 tars 1 np = 2.54 x 10 Stu/nr 1 fra = 7.48 gal. 1 m = 3.41 x 10 5tu/hr Oensity = 62.4 lbqi/ft3 lin = 2.54 cm Oensity = 1 gm/c9 'F = 9/5'C + 32 Heat of vaporization = 970 Stu/lem
'C = 5/9 ('F-32)
Heat of fusion = 144 Stu/1La 1 8TU = 778 ft-lbf
' 1 Ata = 14.7 psi = 19.9 in. Hg. .
r 1 ft. H 2O = 0.4335 inf/in.
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- RX1N-FIG- 15.1 ,
- K,,,,,, Over Core Cycle I
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--"*q s. g 120 i i i i i i i i i O _
100 -
O3 Nominal Expected Line Flow Control
- 7 Pump Constant
' peed Line 80 _
1 Master Manual Power 60 -
J (percent)
Typical Low Power s 05 Flow Control Line
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Minimum Pum O2 Speed Line \p ,
i x 6 Minimum Power Line b
20 _
Recirc Pump NPSH Limit Line f
Jet Pump NPSH Limit Line .
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t i i 120 20 40 60 50 100 L
i Rated Core Flow (oercent)
Figure 15 Normal Operating Map 1
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'It__EBING1ELEE_DE_NWCLE98_EQWER PLANI _QEEB811gNs PAGE 18
.. INEBdQDYNed1GHz_ HEAT TB8MEEEB_eND_ELUID_EL9W ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
I e ANSWER 1.01 (1.50)
- a. Buildup of samarium poison
- b. Depletion of gadolinia poison (in fuel)
- c. Fuel depletion (3 0 0.5 ea.)
REFERENCE Durne Arnold, RXTH-SH-15, Pg. 2 i
ANSWER 1.02 (2.00)
A. Decreases D. Increases C. Increases (due to low friction losses)
D. Decreases (4 9 0.5 ea.)
REFERENCE Ducne Arnold, Fluid Flow, Pg. 4-11 '
ANSWE'R 1.03 (1.00)
Increasing coolant temperature decreases moderator density [0.53, resulting in increased thermal diffusion length [0.53. (1.0)
REFERENCE Ducne Arnold, RXTH-SH-27, Pg. 4 0
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Iz__E61NCIELES_QE_NWGLE98_EQWER_PL8NI_QEEBATIQNs PAGE 19
.. IbE859QYNedIGHz_UEST TR8NEEEB_8NQ_ELQ1Q_ELQW ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
].
ANSWER 1.04 (3.00)
A. Beta effective decreases [0.53
< Pu-239 [0.53 delayed neutron fraction for fission is smaller than the delayed neutron fraction for U-235 fission. [0.53 (1.5)
B. Fuel temperature coefficient of reactivity becomes more negative [0.53 j Pu-240 [0.53 resonance absorption is stronger than that of U-238 [0.53 (1.5)
REFERENCE ~
I Arnold RXTH-SH- 196, p.3.
Arnold RXTH-SH- 26, p.5.
ANSWER 1.05 (1.50)
C ntral rods [0.53 because the edge rods would radiate heat away from the fual bundles while the central rods radiate much of their heat to other ccntral rods [1.03. (1.5)
REFERENCE Dunne Arnold, Heat Transfer, Pg. 15-2 ANSWER 1.06 (3.00)
A. The decrease in the burnout term [0.53 with the production of xenon from iodine still at the higher power rate dominates
[0.53 causing the xenon concentration to increase. (1.0)
B. Peripheral rod worth will increase [0.53 because the highest
( xenon concentration will be in the center, of the core where j the highest flux existed previously [0.53. This will suppress the flux in the center of the core and increase ~the flux in the area of the peripheral rods, thereby, increasing their worth [0.53. (1.5)
C. More than half the value at 100%. (0.5)
REFERENCE Dunne Arnold, RXTH-SH-27, Pg. 5 i RXTH-SH-29, Pg. 2, 11 '
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..'it__EBING1ELES_QE_NWRLE88_EQWEB_EL8NI_QEEBeI1QNm IBEBUQDYNed1 Gat _BEGI_IBONSEEB_eND_ELWID_ELQW PAGE 2]
ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
i ANSWER 1.07 (1.50)
CR1(1 - Keff1) = CR2(1 - Keff2) [0.53 CR1/CR2(1 - Keff1) = (1 - Keff2) 100/270(1 .95) = (1 - Keff2)
.0185 = (1 - Keff2)
K;ff2 = 0.9815 (+/- .002) E1.03 (1.5) i REFERENCE Dunne Arnold, RXTH-SH-18, Pg. 3 ANSWER 1.08 (1.00)
C (1.0)
REFERENCE Dunne Arnold, RXTH-SH-16, Pg. 2 ANSWER 1.09 (1.00)
Increases neutron moderation near the center of the fuel assembly CG.53, thus flattening the flux distribution across the assembly CO.53.
(1.0)
REFERENCE L
Durne Arnold, System Descriptions A-4, ,Pg. 11 ANSWER 1.10 (2.00)
A. Work in the condensate and feedwater pumps B.
Heating in reactor C. Expansion in turbine D. Cooling in condenser (4 @ 0.5 ea.)
(Ok $
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. Iz__EBINCIELES_QF NQGLE88_EQWEB_EL8NI_QPgB811QN i PAGE 21
/.. IBEBUQDYNed1Cas_BEBI_IBeN@EEB_8ND ELWID_ELQW ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
y
[ REFERENCE Duane Arnold, Thermodynamics, HT + FF, Pg. 16-16
( ANSWER 1.11 (2.50)
- a. Flow orificing (0.5)
- b. Higher bundle power causes increased voiding and therefore increased resistance to coolant flow [1.03, if no orificing, high power (cun tral ) bundles would be starved of cooling, t while more coolant flow would be diverted to lower power (peripheral) bundles [1.03. (2.0)
REFERENCE Arnold System Description A-3, P. 13 General Electric - Heat Transfer and Fluid Flow, PP. B-45
, ANSWER 1.12 (2.00)
True (0.5)
Using the equation P = Po e ^ (t/T) co1ving for time results in the equations t=Tx in(P/Po) .
From this it can be seen that since 5/1 yields the same value as 50/10, and cince all other factors in the equation are equal, the time is equal (1.5) d REFERENCE Arnold Reactor Theory, Period, Equation,. ,P . 2 ANSWER 1.13 (1.50)
FALSE (0.5)
The self-shielding effect [0.53 causes more U-238' atoms to be available for rcsonance absorption of neutrons at higher temperatures [0.53, therefore, core overall neutron capture will occur at higher temperatures. (1.0)
REFERENCE Arnold - Reactor Theory, Doppler Effect, P.5
- n4 il
- It__EBINCIELEE_QE_NQGLE88_EQWE8_ELeNI_QEEB8I1QNs PAGE 22
'.. IBEBUQQYNeMIREi_UE8I_IBeNSEE8_8NQ_E(Q1g_E(gy ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
ANSWER 1.14 (2.00)
- c. Core flow increases due to natural convection driving head [0.53 but then stabilizes due to increased pressure drop CO.53 from voiding and friction loss [0.53. (1.5)
- b. Reduction of two phase (or voiding) losses. (0.5)
REFERENCE Arnold - Heat Transfer and Fluid Flow, P. 3-3 Arnold - System Description A-2, P. 33 e
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2t__ELONI_ DESIGN _INGLup]NQ_g8EEIy_QNQ_EdgBGENGY_gYglgdg PAGE 23
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ANSWERS -- DUANE ARN3LD -86/06/24-SHERMAN, J.
ANSWER 2.01 (3.00)
- c. 1. Reactor builc'ing ventilation exhaust high radiation
[0.253, 11 mR/hr [0.253 -
- 2. Refueling Pool exhaust high radiation [0.253 9 mR/hr CO.253
- 3. Drywell high pressure [0.253, 2 psig E0.253
- 4. Reactor vessel low water level CO.253, 170 inches [0.253
- 5. Off gas vent pipe HI-HI radiation [0.253 1.5 x 10 E+4 cps
[ [0.25] C4 required 3 0.5 each3 (2.0)
- b. To remove heat generated by radioactive particles loaded on HEPA filters CO.53, charcoal absorber [0.53. (1.0)
REFERENCE Arnold System Description, E-11, PP. 10, 14 ANSWER 2.02 (3.00)
I
- c. 1. Low lube oil pressure
- 2. Jacket coolant pressure low
- 3. High crankcase pressure
- 4. Jacket coolant temperature high
- 5. Diesel engine start failure
- 6. Diesel engine over-speed
[5 required @ 0.5 each3 (2.5)
- b. Diesel engine overspeed (0.5)
REFE'RENCE
, Arnold System Description, G-2, P. 33 ANSWER 2.03 (1.00)
C -
(1.0)
REFERENCE Arnold System Description C-4, P. 15 '
f9k
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2t__EL8NI_ DESIGN _INGLUQ1NQ_E8EEIY_8NQ_EMEBQENQY_SYSIEM@ PAGE 24 ANSWERS -- DUANE ARNOLD -26/06/24-SHERMAN, J.
i ANSWER 2.04 (2.50)
I
- o. Torus, drywell CO.53 0.1 - 8.35 psi [0.25] (0.75)
Reactor Building, torus CO.53 0.5 psi CO.253 (0.75)
- b. To prevent either the drywell or suppression chamber from exceeding the design negative pressure limit. (1.0)
REFERENCE Arnold System Description E-6, PP. 5, 13, 14 ANSWER 2.05 (2.00)
- 1. Turbine trip
- 2. Turbine trip
- 3. Turbine trip f 4. Isolation [4 @ 0.5 each] (2.0)
REFERENCE Arnold System Description C-3, PP. 24-26 ANSWER 2.06 (1.25)
C, D, B, A, C C5 9 0.25 eachJ (1.25)
REFERENCE Arnold - System Description B-2, Figure 4 ANSWER 2.07 (2.00) .
- 1. Two rupture discs are arranged to relieve directly into the RCIC equipment area if both discs were to break. (1.0)
- 2. If the upstream rupture disc were to break, pressure switches located between the two rupture discs would actuate a RCIC system isolation. (1.0)
REFERENCE Arnold - System Description B-2 P. 27 99k f '
f.
2t__ELONI_DE31QN_INQLQQ1NQ_S@EEIY_@NQ_gdgBggNQY_gySIEME PAGE 25
~*
ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
ANSWER 2.08 (2.00)
- o. Mitigate the thrust loading on the containment. (0.5)
- b. Sir::tP; int: ::nt ri . ;;..t t ...- _ r !. ; . - . Swrprenium pool (0.5)
REFERENCE Arnold - System Description A-6, P. 11 ANSWER 2.09 (2.00)
- c. The moisture content of steam leaving the steam dryer (or entering the steam lines) CO.53. Excessive carryover will cause damage to turbine blading CO.53. (1.0)
- b. The steam content of water leaving the steam separator and steam dryer CO.53. Excessive carryunder will result in jet pump (or recire pump) cavitation t0.53. (1.0)
REFERENCE Arnold - System Description A-3, PP. 17-18 ANSWER 2.10 (2.00)
- 1. Maintain suffic'ient oil in the fluid drive coupler to prevent the generator speed from dropping too low as the generator is loaded. (1.0)
- 2. Maintain oil low enough to prevent excessive recirculation flow caused by starting the M-G set. ,
(1.0)
REFERENCE ,
Arnold - System Description - A-2, P. 58 ANSWER 2.11 (2.00)
- c. The idle loop jet pump flow signal is subtracted from the operating loop Jet pump flow signal. (1.0)
- b. Flow being measured by the idle loop Jet pump is backflow from the operating loop. (1.0)
,,s !! -
f.
2t__EL8NI_DESIEN_INGLUDING_E9EEIY_eND_EMEBEENQY_EYSIEME PAGE 26
^' ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
REFERENCE f
Arnold - System Description A-2, P. 41 ANSWER 2.12 (2.50)
- 1. C
- 2. G
- 3. F
- 4. A L 5. E -
C5 8 0.5 each] (2.5)
REFERENCE j Arnold - System Description A-1, PP. 9, 12, 24, 26, 35 I
O $ %
4 0
a S
$9k h
. z__lNSIByMENTS_ANQ_ggNIBQL@ PAGE 27 i
I ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
l i
ANSWER 3.01 (1.50)
- c. Full scram
- b. None
- c. Half scram E3 @ 0.5 each] (1.5)
REFERENCE 6
Arnold System Description I-7, P. 25 ANSWER 3.02 (3-00) i
- c. By monitoring the differential pressure across each recirc pump for a 2 psid or greater dp, indicating the pump is running. (1.0)
- b. By measuring the pressure differential between the corresponding jet pump risers in Recirculation Loops A and B E1.03. The undamaged loop will have a higher pressure than the damaged loop
[0.53. (1.5)
- c. Loop B (0.5)
REFERENCE Arnold - System Description, C-1, PP. 25-27 ANSWER 3.03 (2.00)
~
- c. Loss of fully open indication of the discharge valve removes the bypass of the 20% speed limiter. The "B" recire pump speed will decrease to 20% (1.0)
T"- :-tia, cf the-45%
- b. " pro -"aaing" p. J iimiter Dypass a s-r.aaoved, ""+ recir_c_ pump _.speeci is_nat-af facted-unti 1-water--
Inval is 1 mss than-1864ftcheer-at #.id t.lide-tt1F~*A= r s uir u pump-
. speed _wil .1_. decrease ..to 45%.- (1.0) 6 e ll, rect v u y m() d es r ro s < s Jr #f 4 4S f, REFERENCE Arnold System Description A-2, Figure 19.
tr4 !!
i
=.
Ez__INSIBudENIS_@NQ_CQNIBQLS PAGE 28 ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
ANSWER 3.04 (3.00)
O. No change 01 dur< un <) . < tu Arc reo iel selef 5 c o.l ' n (1.0)
- b. Decreases initially due to f alse low steam flow indication,y then returns to the same as initially when level equilibrates at lower point. (1.0)
- c. Lowers due to decreased feed flow initially, then stabilizes at some new lower point when level error matches steam-feed flow error. (1.0)
REFERENCE Arnold - System Description D-15, P. 18 and Figure 5 ANSWER 3.05 (2.50)
- c. New reading on Range 7 is 2.5 CO.5].
No auto actions 'dr ----1= --- r i 2 t :-- ; [0.53.
. (1,m>
- b. New reading on Range 5 is 39 E0.53 IRM high rod block and HI-HI half-scram will be in E1.03 (1.5)
REFERENCE Arnold - system Description I-2, PP. 13, 24-25 A
d . /I dd 8 f" J Id # #'" h ANSWER 3.06 (1.50) 8i9'"I detnt.1 N ' sad D p m(> #ad 6 grsonda e 4Tn LOCA r ij ~.s J<reJ*/ C u 3952
- c. ^'1 3a'- -'r-t . . . ... w a . t w i y .
7' 7- 43r5t
- b. A and B pumps start 10 seconds after standby diesel (.
, generators come on-line. ,
(0.175)
C and D pumps start 15 seconds after standby diesel generators come on-line. . Jak'M
, (0370 REFERENCE Arnold - System Description C-1, P. 30 O
?k
f ..
>1t__INSIBQMENTS_AND_QQNIBQLS PAGE 29 y ' . .
ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
h i
ANSWER 3.07 (1.50) i l o. 2 seconds after initiation. (0.75)
- b. Control Rod Drive seal wear. (0.75)
REFERENCE Arnold - System Description I-7, P. 17 Arnold - IPOI-4, P. 15.
I 1
ANSWER 3.08 (1.50) h
- c. INCREASE
- b. DECREASE
- c. DECREASE [3 @ 0.5 each3 (1.5)
REFERENCE Arnold - ADP 518, P. 3 ANSWER 3.09 (3.00)
- 1. Reactor Vessel Low-Low-Low level CO.253 at 46.5" [0.253
- 2. Main Steam High Radiation [0.253 at 3 X Normal Rated Power Background CO.253
- 3. _ Main Steamline High Flow E0.253 at 140% Rated Flow E0.253
- 4. Main Steamline Tunnel High Temperature [0.253 at 200 F CO.253
- 5. Turbine Building Main Steamline Area High Temperature [0.253 at 200 F [0.253
- 6. Low Main Steamline Pressure at Turbine Inlet [0.253 at 850 psig in Run CO.253
- 7. Main Condenser Low Vacuum [0.253 at 10" Hg Vacuum CO.25]
(CR 20" NJ /4belte) ,
. [6 required @ 0.5 each] (3.0)
REFERENCE Arnold - System Descriptions A-6, P. 35 l
$0Y
,8 lz__INEIBudENIE_8NQ_GQNIBQLE 1
PAGE 33
- ' ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
c 1 ANSWER 3.10 (1.00)
( 1. C (0.5)
- 2. A (0.5)
REFERENCE Arnold - System Descriptions A-5, Figure 1 ANSWER 3.11 (2.00)
{ 1. Initiate HPCI
- 2. Initiate RCIC
- 3. Recirculation Pumps Trip
- 4. PCIS Group 8 Isolation
- 5. LPCI loop selection E4 required 9 0.5 each3 (2.0)
I REFERENCE Arnold - System Description A-5, P. 27 ANSWER 3.12 (2.00)
- c. None (0.5)
, b. 158" to 218" (0.5)
- c. -Temperature (0.5)
- d. Cold [0.253 -
No recirc flow [0.253 (0.5)
REFERENCE Dunne Arnold - System Descriptions A-5, P. 33 ANSWER 3.13 (1.50)
- c. Ionization
- b. Ionization
- c. Geiger C3 3 0.5 each3 (1.5)
REFERENCE Arnold - System Description I-2, Figure 5 Arnold - System Description I-11, P. 3 ns !!
b l
SA__E89CEE98EE_ _d98d66t_@gNQBd@La_EMEBgENCY AND PAGE 31
.. 88D196991906_GQNIBg6
' ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
ANSWER 4.01 (2.00)
- o. 1. Decreasing CRD cooling water flow
- 2. Increasing bottom head drain flow C2 9 0.5 each] (1.0)
- b. Two IRM channels in each trip system E0.53 must increase by a factor of 3 after coming on scale CO.53 before the highest reading SRM reaches 10 E5 cps. (1.0)
REFERENCE Arnold IPOI-2, PP. 10, 12 ANSWER 4.02 (1.50)
- o. 10% (0.5)
- b. 1. Minimize possibility of a hydrogen explosion
- 2. Minimize possibility of an untreated radioactivity release.
[2 9 0.5 each] (1.0)
REFERENCE Arnold IPOI-2, P. 8 ANSWER 4.03 (2.00) f
- c. 1. > 2.5 mrem / hour or 100 mrem in 5 consecutive days
- 2. > 100 mrem /hr but < 1000 mrem /hr
- 3. > 1000 mrem / hour [3 0 0.5 ea.3
- b. 150 mrem CO.53 , ,,
(2.0)
REFERENCE Arnold Health Physics Procedure 3106.1, P.2 ,
Arnold Health Physics Procedure 3102.1, P. 2 0
9 n4 l'
f ' dz__EBQGEDWBES_:_N9Bd86t_0BN9BM861_EDEBQgNQY AND PAGE 32
. . B00196991GOL_QQNISQL ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
ANSWER 4.04 (2.50)
O. 1. Stop affected work if a personnel or safety hazard exist.
I 2. Make a duplicate tag.
- 3. Note " duplicate tag issued" next to the pertinent entry on the Equipment Tagging Form. [3 3 0.5 each] (1.5)
- b. Hold Card is used to safeguard human life CO.53, while a Warning Tag is used for operational reasons where human life is not endangered. CO.53 (1.0) l REFERENCE Arnold - Tagout Procedure, PP. 1, 15 ANSWER 4.05 (1.50) i Cy at least two independent indications CO.53 misoperation in automatic code is confirmed CO.53, OR adequate core cooling is assured CO.53. (1.5)
REFERENCE 9, Arnold 01-51 (Core Spray), P. 9 AN:WER 4.06 (2.00)
\ .
- c. 1. Excessive vibration CO.53 and sudden drop in pump discharge pressure and flow CO.53 (1.0)
- 2. Excessive noise on the Jet pump dP indicators (0.5)
L
- b. Reduce recirculation pump speed to the point at which cavitation stops. ' '
(0.5)
REFERENCE Arnold 01-64 (Recirculation), PP. 7, 39 '
10k $
.* dz__EB9GEDUBES_:_N9Bd86x_0EN9Bd861_EdEBQENGY_@NQ PAGE 33
.. 889196991G06_G9 NIB 96 i
ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
ANSWER 4.07 (3.00)
- c. 1. For core power levels of 80% of rated and above, the speed of the faster pump shall not exceed 122% of the speed of the slower pump. (0.5)
- 2. For core power levels below 80% of rated, the speed of the faster pump shall not exceed 135% of the speed or the slower pump. (0.5)
- 6. To assure the LPCI loop selection logic is able to detect recirculation loop breaks. (1.0)
- c. Operation in the prohibited region is not allowed except during coastdown when a recire pump trip occurs during two pump operation. (1.0)
REFERENCE Arnold OI-64 (Recirculation), PP. 9-10, Appendix 1 ANSWER 4.00 (2.50)
- o. Disconnect the amphenol connections for the
~
directional control valves. (1.0)
- b. -1 . I2 m tenl--red --t h idn.,pl ed .
2 If_a._ rod-cannot-be ,,sud-wi th dri um--water- pressure.
?_ one4+4nn 4 ed4eatien-4ee-.ufu14 in" cr.d"f trii uui" i ;-
nn="a!! bis. Jev 017e , /ss) the via [2 required 9 0.5 each3 (1.0)
- c. Eight (0*5)
REFERENCE Arnold AOP 255.1, P. 11 Arnold Tech. Specs., P. 3.3-3' ' '
ANSWER 4.09 (2.50) -
- c. Prompt removal of heat from the primary containment [0.53 and minimizes suppression pool heatup [0.53. '
(1.0)
- b. E==' re e r r l i n ; ' I r:. tr pu p E9.52 -" --= ' ' ^ ' ' - -
J ar lub-icaticr and "-1"= hydrau!!c; [0.395 (1.0)
- c. Assures SDLC pump availability should it be needed again. (0.5)
- b. Avasi t yS nj of h e,b,n. e&. n t cle.k so lve (0,7]
REFERENCE M e> s r* *Jejvate f.,6ent evo, o.) pren ue ( o ,5 )
Arnold EOP-C, PP. 5-9 C:neral Electric - Emergency Operating Procedure Fundamentals l
19 k h *
.-- , ,, - -- - -.y - - . y '
.... v-pi..;ty ,. . .- .
, g ((y,) "The control rod directional control valves for inoperable jg) control rod shall be disarmed electrically." This in itself is one instance. ,
\
l However, listing any condition which causes a control rod to be declared INOP would also be one instance for each condition listed l
per DAEC T.S., then the correct response could also be:
{ 1.
2.
Control rod cannot be moved with drive pressure.
l Control rods with INOP Accumulators.
- 3. Control rods whose position cannot be positively determined.
l 4. Rod Uncoupled.
5.
Full In - Full Out indication unavailable.
m am , , , ma. r,m- "
l _ 3, 7, p A control rod on which maintenance is being performed shall oe nj considered INOP.
g, f Prior to performing control rod or control rod drive maintenance on control cells without removing fuel assemblies, the g directional control valves shall be electrically disarmed at least on the other control rod drives in the 5 x 5 rod array centered on the control rod or rod drive undergoing maintenance.
Then it shall be demonstrated that the core can be made E subcritical by a margin of 0.38 percent.
g Therefore, these to: are correct' answers. yerequestthatanytw_
of this list of_eight be acceptable,.
. m.
. a
' y F bW I.5 44 ~ hd. [g ~
i _ A "a
A w e a _
.a
, v '.
m ..., 3
, 2 a , _
,_, ,,a m
- m. , , , , , , . . _, ,, .
. ,, , r /f . '. 7. * -
-a
. A.
y.~v- . . .~ 1,mw v -
az__EBQGEDUBE5_:_NDBd8Ls_8aNOBd8Li_EMEB@gNGy_8MD PAGE 34
. . BODIQLQQ1GOL_G9 NIB 96 '
ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
4 l
t i
ANSWER 4.10 (2.00) '
l
- 1. Verify Low-Low set relief valves are controlling reactor pressure
. 2. Manually open relief valves l 3. Start HPCI on CST to CST
REFERENCE Arnold IPOI-5, P. 8
\ s ANSWER 4.11 (1.50) '
t
- 1. Reactor power remains above 5%. '
- 2. Reactor power cannot be determined.
j 3. Any control rod not inserted to or beyond Position 02 or position cannot be determined. [3 @ 0.5 each3 (1.5) s REFERENCE
- Arnold IPOI-5, P. 5 ANSWER 4.12 (1.50)
\
- b. -To minimize thermal stresses at the feedwater nozzles.
E3@ 0.5 each3 (1.5)
REFERENCE
, Arnold IPOI-4, PP. 14-17 l
4 19 A $
a
, TEST CROSS REFERENCE PAGE 1 bu'ESTION VALUE REFERENCE
>.0 s 01.01 1.50 MJS0000266 01.C2 2.00 MJS0000267 01.03 1.00 MJS0000268 C1.04 3.00 MJS0000269 01.C5 1.50 MJS0000270 01.C6 3.00 MJS0000271 01.G7 1.50 MJS0000272 01.08 1.00 MJS0000273 G1.G9 1.00 MJS0000274 01.10 2.00 MJS0000275 01.11 2.50 MJS0000276 01.12 2.00 MJS0000277 01.13 1.50 MJS0000278 01.14 2.00 MJS0000279 25.50 C2.01 3.00 MJS0000280 02.02 3.00 MJS0000281 02.03 1.00 MJS0000282 C2.04 2.50 MJS0000283 02.05 2.00 MJS0000284 C2.06 1.25 MJS0000286 02.07 2.00 MJS0000287 G2.C8 2.00 MJS0000288 02.09 2.00 MJS0000289 G2.10 2.00 MJS0000290 02.11 2.00 MJS0000291 C2.12 2.50 MJS0000292 25.25 03.01 1.50 MJS0000285 G3.02 3.00 MJS0000293 G3.03 2.00 MJS0000294 03.04 3.00 MJS0000295 03.C5 2.50 MJS0000296 03.06 1.50 MJS0000297 G3.07 1.50 MJS0000298 03.C8 1.50 MJS0000299 ,
G3.09 3.00 MJS0000300 E3.10 1.00 MJS0000301 G3.11 2.00 MJS0000302 G3.12 2.00 MJS0000303 03.13 1.50 MJS0000304 '
26.00 04.01 2.00 MJS0000305 04.02 1.50 MJS0000306 ns !!
t.
y, TEST CROSS REFERENCE PAGE 2 t ,
QUESTION VALUE REFERENCE
~
C4.C3 2.00 MJS0000307 C4.C4 2.50 MJS0000308 C4.C5 1.50 MJSOOOO309 C4.06 2.00 MJS0000310 C4.07 3.00 MJS0000311 C4.08 2.50 MJS0000312 C4.C9 2.50 MJS0000313 C4.10 2.00 MJS0000314 C4.11 1.50 MJS0000315 G4.12 1.50 MJS0000316 24.50 101.25
?
.j -
4 e
t
. ii l
1 l .
[O 0
4 MASTERCOPY U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _DU6Ng_8BNgLD____________
REACTOR TYPE: _BWB-GEd____ __ _ _ _ ___
DATE ADMINISTERED: _Q64@642d__ __ _
EXAMINER: _SHgBMAN z_J ____
z ____ ___
APPLICANT: ____
INSIBUCIlgNS_IQ_GEELIC@ nit Usa separate paper for the answers. Write answers on one side only.
Stcple question sheet on top of the answer sheets. Points for each quzstion 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 si> (6) hours after th2 examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY
__Y86Ug_ _I@IB6 ___SCQBE___ _y86UE__ _ __ _ _ ____C GIEG Q B Y_,, __ _________
_2E E@__ _2Et2E ___________ ___-_ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
_2EzE@__ _2Es25 ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
_2Ez@@__ _2ds2E _- -______
________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
_25tgg__ _23zZ5 ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 191c@0__ 1991gg ___________ ________ TOTALS FINAL GRADE _________________%
All work done on this examination is my own. I have neither givrn nor received aid.
EPPLiC5UTI~ S SIGNATURE ~~ ~~~~~~~~~~~
I Qt__ THEORY OF NUCLEAR POWER PLANT OPERATION z _ELylQSi_8ND PAGE 2 IHEBdgQYN8dICS i .
QUESTION 5.01 (1.50)
Tha attached figure 15.1 shows how K-excess varies over a typical core cycle.
Briefly explain the PREDOMINANT CAUSE of the change in K-excess between the following points:
- a. A and B
- b. B and C
- c. C and D QUESTION 5.02 (2.00)
A2sume a large centrifugal pump is operating normally with the discharge velve fully opened. What changes will occur in the below parameters if the discharge valve is throttled to the 50% open position? (Limit your answer to INCREASE, DECREASE, or NO AFFECT.)
A. Pump motor current B. Pump discharge head C. Available Net Positive Suction Head (NPSH)
D. Pump flow rate DUESTION 5.03 (1.00)
Explain why a control rod will have higher worth at high reactor coolant temperature than at low reactor coolant temperature.
QUESTION 5.04 (3.00) -
Plutonium isotopes are generated in the fuel during reactor operations.
EXPLAIN WHY plutonium buildup changes the following during a core cycle:
(Also SPECIFY the DIRECTION OF CHANGE and WHICH plutonium ISOTOPE is prGdominant for each)
A. Beta effective B. Fuel temperature (Doppler) coefficient of reactivity .
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
1 1:
1
- Dz___THgORY _ OE_NyGLE@B_EQWER PL9NI_QEEB8I1ONz_E691DSz_8NQ PAGE 3 IHEBdQQYNed1GS
{t ,
,a L
QUESTION 5.05 (1.50)
! In the event of a LOCA after extended operation at full power, which fual rods (EDGE or CENTRAL) are more likely to exceed the 2200 deg F limit? Explain WHY.
QUESTION 5.06 (3.00)
R2garding the xenon transient following a significant DECREASE in reactor power from high power operation:
~
4 A. Briefly, EXPLAIN WHY the xenon concentration will peak following the manuever. (1.0) e B. How will peripheral control rod worth be affected (INCREASE, DECREASE, REMAIN THE SAME) during the xenon peak? Briefly EXPLAIN your answer. (1.5)
C. If the decrease in reactor power was from 100% to 50%, would the new (50% power) equilibrium xenon reactivity be MORE THAN, LESS THAN, or EQUAL TO one half the 100% equilibrium value? (0.5)
QUESTION 5.07 (1.50)
During a reactor startup, Keff is .95 when the SRM channels read 100 cps.
What-will the new Keff be when the SRM channel reads 270 cps?
(SHOW YOUR WORK) 4
~
i QUESTION 5.08 (1.00) l Consider reactor fuel with exposure of 10,000 mwd / ton. Which of the following is the major intrinsic neutron source (CHOOSE ONE):
i A. (alpha, n) reaction with 0-18 B.
(alpha, n) reaction with Be-9 i
l C. Spontaneous fission of Cm-242 l D. Gamma-neutron reaction with deuterium
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
i,s 1:
t'
- gz THEORY OF NUCLEAR POWER _PL@NT_QPEB8 TION z _E6ylQSz_9NQ PAGE 4 IHERMODYNAMICS 11 J'
i QUESTION 5.09 (1.00)
A typical fuel assembly contains two " water rods". Describe HOW water rods affect the assembly in terms of reactor physics.
QUESTION 5.10 (2.00) powgg et/pfr
(. The attached Figure 17.2 shows a temperature versus entrop IDEAL power cycle. State the PHYSICAL PROCESS and PE.^.CTO.,y diagram (S)
COMPONENT for an cecociated with that process for each of the following transitions:
A. Point 1 to Point 2
. B. Point 2 to Point 3 i
C. Point 3 to Point 4 i
D. Point 4 to Point 1 l
QUESTION 5.11 (2.50)
., A design feature in the reactor vessel ensures proper flow distribution through the fuel bundles.
- c. What is this feature (or component)? (0.5)
- b. EXPLAIN why this feature is necessary in a BWR, including the consequences of a power increase if this feature were eliminated. (2.0)
I i QUESTION 5.12 (2.00)
True or False.
l For a constant reactor period, it takes the SAME AMOUNT OF TIME to change l
rarctor power from 1% to 5% as it does to change it from 10% to 50%.
EXPLAIN YOUR ANSWER FULLY. SHOW THE CALCULATION YOU WOULD USE TO VERIFY YOUR ANSWER.
l t
j (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
,,, ,e
I r-
" , ' Ez__IBEQRY_OE_NUg6E88_EQWEB_EL8NI_QEEB8IlgNz_ELy1QSz_8ND PAGE 5 ISEBdQDYN8digg
't I
L QUESTION 5.13 (1.50) f Thm effects of " Doppler Broadening" in U-238 result in a modified capture cross-section curve, but the area under both the original and the, (higher tamperature) broadened curve will theoretically be the same. Thi: , ::::: te indicate that 6Vera11 neutron capture by U-238 will be about the same at
, cny temperature.
I In this TRUE or FALSE? EXPLAIN your answer.
QUESTION 5.14 (2.00)
Thw attached figure 15 shows the " Normal Operating Map" of rated core flow vsrsus reactor power.
- c. EXPLAIN WHY, on the natural circulation '. i n e , incremental increases in power initially produce very rapid increases in core flow, but eventually reach a point where further power increases produce no increase in core flow. (1.5)
- b. EXPLAIN WHY, on the pump constant speed line, core flow increases as reactor power decreases. (0.5) i l
t
(***** END OF CATEGORY 05 *****)
ra r!
l l
, . . . - . . - ~ - . . - . - -.
i ,..
- 6. PLANT SYSTEdg_pggl@Ni _ggNIRQL2_AND_INSTRUdgNIATIQN
" PAGE 6
- a
) QUESTION 6.01 (3.00)
Rsgarding the Standby Gas Treatment System:
I a. State FOUR conditions which will automatically initiate the SGTS trains. Include setpoints. (2.0)
- 6. EXPLAIN WHY it may be necessary to provide continued cooldown
. flow after the train is no longer required (assume the train has been operating for some time). (1.0)
QUESTION 6.02 (1.00)
! Which of the following is TRUE concerning the Standby Liquid Control System: (CHOOSE ONE.)
a.
In the event a remote (outside control room) reactor shutdown is required, SBLC injection can be actuated by the local pump START switch.
- b. The pumps may be operated simultaneously if necessary to shutdown the reactor in an ATWS.
,. c. When the control room handswitch is placed to " PUMP A RUN",
the "A" pump starts and all squib valves fire.
- d. Nitrogen-charged accumulators assure adequate suction pressure for the pumps.
QUESTION 6.03 (2.00)
Dsscribe TWO WAYS rupture discs would prevent overpressurization of the RCIG turbine exhaust piping. (Setpoints not required.)
QUESTION 6.04 (2.00)
R2garding the Low-Low Set (LLS) valves:
, n. What is their purpose? (0.5)
- b. Where do they discharge? (0.5)
- c. What are the TWO actuation signals? (Setpoints not re' quired.) (1.0)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
,a 1:
t.
&z__Eb6NI_EYEIEdE_DEEl@Nz_CQUIBQLa_8NQ_INQIBydENTATION PAGE 7 t
w QUESTION 6.05 (2.50)
Choosing from the list of choices below, identify which SPECIFIC Control Rod Drive component is installed to allow each of the following:
- 1. Minimize pressure transients during rod motion.
- 2. Prevent the index tube and control rod from accidentally moving downward.
3.. Avoid excessive initial rod withdrawal speed following a scram.
- 4. Two sources of motive force to scram a rod.
- 5. Create a hydraulic buf f er action to cushion the drive at the end of its scram stroke.
LIST GF CHOICES:
- a. Ball check valve f. Equalizer valves
- b. Coupling spud g. Collet assembly
- c. Stabilizer valves h. Spectacle flange
- d. Velocity limiter i. Bellville spring washers
- s. Piston tube orifices J. Flow control valves
QUESTION 6.06 (3.00)
Regarding the LPCI Loop Selection Logic:
- n. HOW does the logic determine how many recirc pumps are running? (1.0)
- b. HOW does the logic determine whichcis the UNDAMAGED recirc loop? (1.5)
- c. If the logic determines that neither loop is damaged, WHICH loop will it select for LPCI injection?, -
(0.5)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
l ns 1:
l
\
- 6z__PL6NI_SygIEMS_pEgl@Nz_GQUIBg65_@NQ_lNSIBUMENIBIlgN PAGE 8 QUESTION 6.07 (2.00)
The reactor is operating at 100% power,+ith er i --"1 = H na ---t ci ir ::ter 2nurl_ What will be the effect on/of the recirculation flow control cystem due to the following conditions: (STATE ANY ASSUMPTIONS YOU MAKE.)
- c. The full open indication on recirc pump "B" discharge valve is lost.
NOTE: The recirc pump discharge bypass valve is open.
- b. The reactor feedwater pump "A" trips. (Assume during the subsequent transient reactor water level falls below 186 inches)
QUESTION 6.08 (3.00)
DESCRIBE the changes (if any) in the following parameters in the FIRST FEW MINUTES following the low failure of one steam flow transmitter. (Assume 100% reactor power, three element FWLC, and no operator action.)
- c. Reactor power
- b. Feedwater flowrate
, c. Reactor water. level OUESTION 6.09 (2.50)
For each of the IRM (Intermediate Range Monitoring) range changes (a. and b.) listed below, provide the following:
(Mode switch in STARTUP):
- 1. The indicated level'on the N$W' RANGE.
- 2. Any automatic actions initiated as a result of the indicated level on the new range. ,
- c. Switching from range t.etting 5, reading 25, up 'to range setting 7. (1.0) i
- b. Switching from range 6, reading 39, down to range 5. (1.5)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
1
6.
PLANT SYSTEMS _QE@IGN z _QQNIRQL2 _6ND INSIRUMENTATIQN PAGE 9 i
1 1
QUESTION 6.10 (1.50)
R garding the Reactor Mode Selector Switch RPS Trip:
- a. When is the trip automatically reset?
1
- b. What damage could result if the SCRAM were not 2utr 2 tic 211y reset?
QUESTION 6.11 (3.00)
List SIX signals, including the SETPOINT, which will cause an automatic icolation of the Main Steam Isolation Valves (MSIV's).
3[
)
6 I
l ,
1 9
(***** END OF CATEGORY 06 *****)
i,s rl l
( ..;..,.....--... . . - - - . - - . . .
ic
- Z. fBgggpuBgg_:_NgBdeL,_egNgBd862_EdgBggNgy_6ND PAGE 10 88D196991CeL_GONIB06
+
l QUESTION 7.01 (1.50)
According to Integrated Plant Operating Instruction No. 2 (startup):
d r o. What is the maximum reactor power level at which mechanical vacuum y pump operation is allowed? (0.5)
- b. State TWO reasons for limiting mechanical vacuum pump operation to a specified power level. (1.0) i k
I QUESTION 7.02 (2.00) i
- o. STATE the exposure rate limits which characterize the f ollowing: (1.5) h 1. Radiation Area L
- b. Assuming you have an approved NRC Form 4 (exposure history) on file, WHAT is the maximum whole body exposure you are allowed in ONE DAY without additional authorizations. (Also assume it is the l first day of a new calendar quarter.) (0.5)
QUESTION 7.03 (2.50)
- a. According to the Tagout procedure, what THREE actions are you required to perform if while on a work assignment a Warning Tag or Hold Card is found to be missing? (1.5)
- b. Briefly explain how the use of the Warning Tag and Hold Card differs. (1.0)
QUESTION 7.04 (1.50)
According to procedure you are instructed NOT to secure or override an ECCS system automatic initiation unless WHAT conditions exist?
P
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
ns il
I .' s Z. EBgggpuBgg_ _NgBdeL,_egNgBd@L2_EdgBGgNCY ANQ PAGE 11
}
. BeDig69 Gig @L_ggNIBQL i
i QUESTION 7.05 (3.00)
Concerning the recirculation pump speed mismatch limitations (refer to cttached Figure 20):
- a. STATE the TWO operating limitations.
- b. WHAT is the PURPOSE.of the mismatch operating limitations?
- c. Is operation expected to occur in the " PROHIBITED" region at any time? Briefly EXPLAIN.
QUESTION 7.06 (2~50)
- c. How is a control rod drive electrically disarmed? (1.0)
- b. What are TWO instances when a control rod drive would be electrically disarmed? (1.0)
- c. During reactor power operation, the number of inoperable control rods shall not exceed _7 .
(0.5)
QUESTION 7.07 (2.50)
Briefly explain the REASON (S) f or each of the following cautions in the Emergency Operating Procedures:
- c. Manually trip Standby Liquid Control pumps'at 0% in the SBLC tank. . (0.5)
QUESTION 7.08 (2.00)
Ansume the reactor has scrammed from extended power operation, and the MSIV's are shut. According to Integrated Plant Operating Instruction No. 5 (Reactor Scram), what immediate operator actions or checks may be taken to CONTROL REACTOR PRESSURE. (LIST FOUR)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
ra 1:
1 Zz__EBOREQUBES - NQBM862_8BNgBM8Ls_EMEBGENgy_8NQ PAGE 12
, BSD1969BIGOL_G9 NIB 96 I
j' QUESTION 7.09 (1.50)
List the THREE conditions that would cause you to enter EOP-1, RPV Control, I following a reactor SCRAM.
QUESTION 7.10 (1.50)
Briefly explain the REASON for each of the following cautions in Integrated Plant Operating Instruction No. 4 (Shutdown):
- a. Insert the SRM detectors one at a time.
L i b. The reactor should not remain in a hot shutdown condition longer than necessary.
- c. Maintain RPV level below +211 inches.
QUESTION 7.11 (2.00) -
i In a " Loss of Water Level Situation" during refueling, one of the rcsponse steps is to immediately order evacuation of the torus
. cnd drywell.
A. What is the reason for this evacuation?
B. How is determination of a complete evacuation performed?
QUESTION 7.12 (2.50)
According to Fuel and Reactor Component Handling Procedure #5:
A. What THREE methods may be used to verify correct orientation of a fuel assembly after the assembly has been , lowered into a cell? (1.5)
B. HOW is it determined that a newly installed fuel assembly is properly seated, and how many times is assembly seating checked during a normal refueling? (1.0) l (***** END OF CATEGORY 07 *****)
l
..a a :
N* az__8pMINISTRATIV@_PRQgEQUREgi_ggNDITIgN@t_ANQ_ LIMITATIONS _
PAGE 13 3
(r
- i QUESTION 8.01 (1.00) i Stcte the Technical Specification limits for reactor coolant leakage into thn primary containment. Consider both identified and unidentified cources.
t QUESTION 8.02 (1.50)
While performing a routine periodic surveillance, your plant operator informs you that a required fire barrier seal appears to have been damaged cnd, in his opinion, needs to be replaced. What action (s) must you take prior to repair of the fire seal?
QUESTION 8.03 (2.50)
R2garding MCPR (Minimum Critical Power Ratio):
- a. EXPLAIN WHY the safety limit involving MCPR is different for single loop operation than for two recirculation loop operation. (1.5)
- b. EXPLAIN WHY the Operating Limit MCPR must be modified at less than 100% rated core flow. (1.0)
QUESTION 8.04 (1.50)
During spiral fuel _ unloading, the SRM count rate is allowed to drop below 3 counts per second. At what point in fuel reloading must the SRM's again rocd greater than 3 counts per second?
QUESTION 8.05 (2.50)
Given the below condition in the Standby Li'qu'id Control System, indicate
[
l whsther the SBLC SYSTEM is considered OPERABLE or INOPERABLE. INCLUDE any rGquired actions or time restrictions, if applicable, j o. Continuity indication for one of the squib valves has been lost. (1.5) l b. The temperature of the liquid control solution has drifted below the curve shown in Tech. Spec. Figure 3.4-2. (1.0)
I i (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
no it
, . ~ .. ..... .. . - ,.. -. .
Ez--_ADMINISTB8 TIME _PBQGEDUBESz_GQNDlIlgNgs_8ND_LIMITAllgN@ PAGE 14 QUESTION 8.06 (2.00)
Tha Reactor Engineer reports to you that, due to miscalibration of the fcadwater flow instrumentation, the reactor has been operating at a cciculated steady state power of 1665 MW (thermal) for the past month.
Whet actions are you required to take and WHY must you do so?
(Include any applicable reporting requirements)
QUESTION 8.07 (2.00)
Tha reactor is operating at 100% power when all MSIV's close inadvertently.
Th9 reactor scram signal comes from APRM's HI-HI. Has a safety limit been violated? EXPLAIN your response.
QUESTION 8.08 (2.00)
EXPLAIN the elationship between SAFETY LIMITS and LIMITING SAFETY SYSTEM SETTINGS in providing reactor protection.
I QUESTION 8.09 (2.50)
Concerning Operating Logs:
- o. SPECIFICALLY, how is a mistaken entry corrected? (1.5)
- b. Prior to making entries in a new log book, what TWO items must be checked? (1.0)
. c s QUESTION 8.10 (2.00)
, Du2 to vacation leave by control room opera' tors, extensive overtime is rcquired in the coming week. Examine this work schedule and STATE those overtime restrictions that would be violated. (Assume reactor is operating) l SUNDAY 0700 - 1900 MONDAY 0700 - 1500 l TUESDAY 0700 - 2400 l -
WEDNESDAY 0700 - 1500
, THURSDAY 0700 - 2000 -
i FRIDAY 0700 - 2000 "
SATURDAY OFF t
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
( ,, < :
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t
, B. ADMINISIBellyE_BBQQEQQBE@i_GQNQlligN5t_8NQ_Q1MITATIQNH I .
PAGE 15 9
1 1
l l
QUESTION 8.11 (1.00) '
l The plant is operating at 103% power and you are the Operation Shift
) Supervisor. The Shift Technical Advisor on crew asks your permission to 1 cave the plant for one hour at lunchtime. What factors must you consider l in your decision to approve / disapprove? (Assume another qualified STA is
!_ not available.)
QUESTION 8.12 (2.50)
I R2garding temporary procedure revisions:
- a. STATE the TWO persons (positions) who must approve the revision. (1.0) b b. WHAT TWO types of revisions are specifically NOT ALLOWED to be J accomplished through the use of a temporary revision? (1.5)
QUESTION 8.13 (2.00)
D= scribe the TIME REQUIREMENTS and REPORTING METHOD for the following:
- c. Immediate Notification Events
- b. Licensee Events O a e t
(***** END OF CATEGORY 08 *****) '
(************* END OF EXAMINATION **********f*'*J*)
F L
' EQUATION SHEET Cycle efficiency = (Network f = ma v = s/t out)/(Energy in) .
z s = v,t + 1/2 at
, = 3;2 .
g = x-
- A = 14 A = A,e ** -
KE = 1/2 av a = (Vf - t,)/t '
PE = agn Vf = V, + at * = e/t i = ui2/t1/2 = 0.693/t1/2 1/2 8# =Utu1M y,y j -
A=
rD 2 ((g/2)+(t)J o
, 1 AE = 931 am
-Dt a = V,,Ao ,
Q = mCoat I=I,e#
d = UA A T I = I,10** E#l Pwr = Wfah TVI. = 1.3/u sur(t)
~
gyg , ,q,3937, P = P 10 p = p e8 /I o
SG = 5/(1 - K,ff)
SUR = 25.06/T G x= S/(1 - K ,g x)
SUR = 25e/s* * (a - o)T G j(1 - K,ff3) = G 2EI ~ "eff2)
M = 1/(1 - K,g ) = CRj / G 3
. T = ( t*/s ) + [(a - o VIo]
M = (1 - K ,g ,)/(1 - K ,ffj)
T = 1/(o - a)
SCM = (1 - K ,g)/K ,g T = (a - o)/(Is) t' = 10# seconds a = (X ,g-1)/K ,g = JK ,g/K ,g I = 0.1 seconds'I o=[(**/(TK,ff)]+[T,ff(1+IT)] /
Id I
l1*Id2 ,2gd2
- Id j 22 P = (cav)/(3 x 1010) 2 R/hr = (0.5 CE)/c (meters) e=N R/hr = 6 CE/c2 (feet) ,
l Miscellaneous Conversions l Water Partneters '
1 curie = 3.7 x 10 10 dos 1 gal. = 8.345 tom. 1 kg = 2.21 lba .3 1 gja .==7.48 3.78gal. litars 1 np = 2.54 x 10, Stu/hr 1 f.
Oensity = 62.4 10g/ft3 1 as = 3.41 x 100 5tu/hr Oensity = 1 gm/c9 lin = 2.54 cm
- F = 9/5'C + 32 I
Heat of vaoorization = 970 Stu/lem *C = 5/9 (*F-32)
Heat of fusion = 144 Stu/lem 1 STU = 778 ft-Ibf l 1 Ace = 14.7 psi = 19.9 in. Hg. .
l l
1 ft. H 2O = 0.4335 luf/in.
a r!
.m. - - _ , - _ - - - - - _
q.
RKTH-FIG- 15.1 * .
t K,,,,,, Over Core Cycle r
i t
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, excess 5 D 4
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- 1. o < _ _ _ -_ _ _ _ _ _ _._ _ _ _ _ _ _ _ _
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(Typical of Cycle 1) l l
i 4 c.s 1
- - - - - - . - - - - . . ~ - - - - - _ - , - - - - - , , . , , - , - - --r.---,--,--.,,c.._., - - - - - - -
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- l ENTROPY i.
, Figure 17.2 T-s Diagram of a Power Cycle l .
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l e
16-17
- a 1:
,w---- -r.,,_---,--,------,--,,-,,,--,_m,- ,. - - , - - , _ _ _ , ,
(._ _ _
, . .._. _ . - - . . - - . - . - . - . . . ~ . - . - - _ _ - _ . - - . . . .
- . p. uy gyrgr~~' . - 4- ~ ^% - + .aes.s y . ,
p ,+ =
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q [ *.; ,
120 i y a a l l l i s
,' O - _
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1.
100 _
a 3 Nominal Expected Flow Control Line 7 Pump Constant peed Line 80 _
i s
i Master Manual Power 60 -
j 04 Control Line _
(percent)
Typical Low Power 05 Flow Control Line 1 -
/ _
- C53 Natu ral
' ~
40 _.
h'C'!*tIO" Minimum Pump O2 Speed Line
(
% 6 Minimum Power Line h
t l
20 _
1 / _
Reirc Pump NPSH Limit Lme
/
Jet Pump NPSH Limit Line ,
I I l I I l l l l L 20 40 60 80 100 120 Rated Core Flow (cercent)
Figure 15 Normal Operating Map
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706 22.?f.f *&0 ~'~}Ti~~'~~~~~~~'"-~~~~~~~~~~~ ~~
~~ 4 J~F" Fad 's l 5
s i
L l / /
SC OPEMTION j f PROHIBITED , y j
/ /
go _
/ / /
/ / /
/ / /
N *
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/
/
/
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, ,/
/ ,/
/
7 SPEED OF 60 -
/
- NWA A
(%)
i y / y4
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/ PROHl8tTEO M -
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/
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01 64 Rev. O Page 42 Of 77 .
5/10/85 f e'3 v re 20 tv 4 f$
.-.r.-. . . . . - - . . . . _ _ _ _ -. . _ - . . _ -
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DAEC 1 t
I C LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT d 3.4 STANDBY LIQUID CONTROL SYSTEM 4.4 STANDBY LIQUID CONTROL SYSTEM i
Applicability: Applicability:
t Applies to the operating Applies to the surveillance
- status of the Standby Liquid requirements of the Standby L Control System. Liquid Control System.
Ob.iective: Obiective:
To assure the availability of To verify the operability of l
a system with the capability the Standby Liquid Control
- 1. to shut down the reactor and System.
maintain the shutdown condition without the use of Control rods.
t Soecification: Specification:
A. Normal System Availability A. Normal System Availability
- 1. During periods when fuel is The operability of the Standby in the reactor and prior to Liquid Control System will be f startup from a Cold Condition, verified by the performance of the Standby Liqui'd Control the following tests:
W System shall be operable, except as specified in 3.4.B 1. At least once per month each
- - below.' This system need not pump loop shall be functionally be operable when the reactor tested by recirculating is in the Cold Condition and demineralized water to the test all control rods are fully tank. Minimum pump flow rate inserted and Specification of 26.2 gom against a system 3.3.A is met. head of 1150 psig shall be verified.
- 2. At least once during each
, , ,. operating cycle:
- a. Check that the setting of the system relief valves is
- 1350 < P < 1400 psia.
I^ D - {, " P } *] " S "O I n 'F:l2 ;yrC t' P' l 8- ,;
6 e !
e, "4 'h ., "! l' g .
d _.lh/ O. L b I.;l uL d U . .. . ( . W by b . f . , ;;
3.4-1
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1 .
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- b. Manually initiate the system to open both explosion actuated valves and conduct flow tests to inject demineralized water through one Standby Liquid Control pump directly into the reactor vessel.
Explode one of three charges !
manufactured in same batch to verify proper function. Then install the untested charges in the explosion valves.
- c. Prove capability of the sodium pentaborate storage tank discharge line to convey the minimum pump flow rate of 26.2 gpm.
B. Operation with Inoperable B. Surveillance with Inoperable Components Components
- 1. From and after the date that a 1. When a component is found to be redundant component is made or inoperable, its redundant found to be inoperable, component shall be demonstrated Specification 3.4.A.1 shall be to be operable imediately and considered fulfilled and daily thereafter until the continued operation permitted inoperable component is provided the component is repaired.
returned to an operable
~
condition within seven days.
C. Sodium Pentaborate Solution C. Sodium Pentaborate Solution At all times when the Standby , The following tests shall be Liquid Control System is performed to verify the required to be operable the availability of the Liquid following conditions shall be , . Control Solution:
met:
- 1. Volume:' Check and record at
- 1. The net volume versus least once per day, concentration of the Liquid '
Control Solution in the liquid control tank shall be maintained as required in Figure 3.4-1.
3.4-2 ..s it
3.
DAEC-1 J'
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
'( }
- 2. The temperature of the liquid 2. Temperature: Check and record at least once per day.
control solution shall be maintained above the curve shown in Figure 3.4-2. This includes the piping between f
the standby liquid control tank and the suction inlet to the pumps.
- 3. Concentration: Check and record at least once per month.
Also check concentration anytime water or baron is added to the solution or solution temperature is below the temperature recuired in Figure 2.4-2.
D. If Specification 3.4.A through C cannot be met, the reactor shall be placed in a Cold Shutdown Condition with all operable control rods fully inserted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
> e
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120 5ATURAft0N -
TEMPERATURE llo curve ,
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- .4 o
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- M 4e o
500lym PENT Ao0R ATE S0t.UTION (PERCENT oY wElGMT) . ,,
o DUANE ARNOLD ENERGY CENTER-4 IOWA ELECTRIC LIGHT & POW $R COMPANY TECHNICALSPECIFICATICIS Saturation Temperature of Sodium Pentaborate Solution FIGURE 3.4-2 l
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' Uz__IBEQBY_QE_NWGLEBB_EQWEB_ELONI_QEEB811QNz_ELylDQx_8NQ PAGE 16 g . INEBdQDYNOMIGH l -- s
, A?!SWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
I ANSWER '5.01 (1.50)
- c. Buildup of samarium poison g b. Depletion of gadolinia poison (in fuel)
- c. Fuel depletion (3 a 0.5 ea.)
I REFERENCE Dunne Arnold, RXTH-SH-15, Pg. 2 ANSWER 5.02 (2.00)
I A. Decreases B. Increases C. Increases (due to low friction losses)
D. Decreases (4 a 0.5 ea.)
REFERENCE Ducne Arnold, Fluid Flow, Pg. 4-11 ANSWER 5.03 (1.00)
Increasing coolant temperature decreases moderator density [0.53, resulting in increased thermal diffusion length [ 0,. 5 3 . (1.0)
REFERENCE ,
Dunne Arnold, RXTH-SH-27, Pg. 4 ,
9 I
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us !!
.-.m t *:
t
- Ez__IHggRY OF NUgLgGE__ POWER PLANT.OPER8IlON z _ELUIDSs _6ND PAGE 17 THERMODYNAMICS 1
ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
i ANSWER 5.04 (3.00) -
A. Beta effective decreases [0.53 Pu-239 E0.53 delayed neutron fraction for fission is smaller than the delayed neutron fraction for U-235 CO.53. (1.5)
B. Fuel temperature coefficient of reactivity becomes more negative E0.53 Pu-240 CO.53 resonance absorption is stronger than that of U-238 E0.53 (1.5)
REFERENCE i
Arnold- RXTH-SH- 196, p. 3 Arnold- RXTH-SH- 26, p. 5 i * '
s ANSWER 5.05 (1.50)
Csntral rods CO.53 because the edge rods would radiate heat away from the fum1 bundles while the central rods radiate much of their heat'to other ccntral rods [1.03. (1.5)
REFERENCE Durne Arnold, Heat Transfer, Pg. 15-2 ANSWER 5.06 (3.00) l A. The decrease in the burnout term E0.53 with the production of xenon from iodine still at the higher power rate dominates j [0.53 causing the xenon concentration to increase. (1.0)
B. Peripheral rod worth will increase EO.5.3 because the highest xenon concentration will'be in the center of th,e core where the highest flux existed previously E0.53. This will suppress the flux in the center of the core and increase the flux in the l area of the peripheral rods, thereby, increasing their worth [0.53. (1.5)
C. More than half the value at 100%. (0.5) l REFERENCE
~
Durne Arnold, RXTH-SH-27, Pg. 5 RXTH-SH-29, Pg. 2, 11 ra rl
IHggBy_gE_NygLgeB_EgWgB_ELeNI_gegBATIgs,_ELylpSm_eNQ p' 3 PAGE 18 IHgBdQQYNedlGS
, ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
l' l ANSWER 5.07 (1.50)
CR1(1 - Keff1) = CR2(1 - Keff2) [0.5]
, CR1/CR2(1 - Keff1) = (1 - Keff2) f 1C3/270(1 .95) = (1 - Keff2)
.0185 = (1 - Keff2)
Kaff2 = 0.9815 (+/- .002) [1.0] (1.5) l REFERENCE Dunne Arnold, RXTH-SH-18, Pg. 3 ANSWER 5.08 (1.00)
C (1.0)
REFERENCE Durne Arnold, RXTH-SH-16, Pg. 2 ANSWER 5.09 (1.00)
Increases neutron moderation near the center of the fuel assembly
[0.53, thus flattening the flux distribution across the assembly [0.5].
(1.0)
REFERENCE i
Dunne Arnold, System Descriptions A-4, Pg. 11 l
ANSWER 5.10 (2.00) l .
A. Work in the condensate and feedwater pumps B.
Heating in reactor l
C. Expansion in turbine D. Cooling in condenser (4 @ 0.5 ea.)
l f
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- g. THEOBY_QF NUGLg8R_POWEB_PL8NT OPERATION 2 _ELulpSz _8NQ PAGE 19 4
a
' ISEBdODYN8d1GS
.t ..
ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
l REFERENCE Durne Arnold, Thermodynamics, HT + FF, eg. 16-16 ANSWER 5.11 (2.50)
- a. Flow orificing (0.5)
- b. Higher bundle power causes increased voiding and therefore increased resistance to coolant flow [1.03, if no orificing, high power (central) bundles would be starved of cooling, while more coolant flow would be diverted to lower power (periphera') bundles [1.03. (2.0)
REFERENCE
,' Arnold System Description A-3, P. 13 GIneral Electric - Heat Transfer and Fluid Flow, PP. B-45 i
(,
ANSWER 5.12 (2.00)
True (0.5)
~
Using the equation P = Po e ^ (t/T) co1ving for time results in the equation:
t=Tx in(P/Po)
From this it can be seen that since 5/1 yields the same value as 50/10, and
, since all other factors in the equation are equal, the time is equal (1.5)
}
il REFERENCE Arnold Reactor Theory, Period Equation, P. 2 ANSWER 5.13 (1.50)
FALSE (0.5)
Tha self-shielding effect [0.53 causes more U-238' atoms to be available for rononance absorption of neutrons at higher temperatures CO.53, therefore, l more overall neutron capture will occur at higher temperatures. (1.0)
REFERENCE Arnold - Reactor Theory, Doppler Effect, P.5 *
<a it
Q. THEQRY OE_NQGLEAR POWER PLANT _OPERATIQN z _ELQ1Dg2_8ND PAGE 20
. IBEBd99YNedIGS ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
ANSWER 5.14 (2.00)
I
- o. Core flow increases due to natural convection driving head E 0'. 5 3 but then stabilizes due to increased pressure drop [0.53 from voiding and friction loss E0.53. (1.5)
- b. Reduction of two phase (or voiding) losses. (0.5) i REFERENCE f Arnold - Heat Transfer and Fluid Flow, P. 3-3 Arnold - System Description A-2, P. 33 i
F l-2 W
O b $
9 b
o e
{$k O
I .
fu,_ j PLANT SYSTEMQ_QggIQN z_gglTROLa_AND__ INSTRUMENTATION PAGE 21 1 .. ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
f 1
ANSWER 6.01 (3.00) i s. 1. Reactor building ventilation exhaust high radiation
[
EO.253, 11 mR/hr E0.253
- 2. Refueling Pool exhaust high radiation E0.25] 9 mR/hr E0.253
. 3. Drywell high pressure CO.253, 2 psig E0.253
- 4. Reactor vessel low water level E0.253, 170 inches CO.253
- 5. Off gas vent pipe HI-HI radiation E0.253 1.5 x 10 E+4 cps E0.25] E4 required @ 0.5 each] (2.0)
- b. To remove heat generated by radioactive particles loaded on HEPA filters E& ret; charcoal absorber L&eef. (1.0)
REFERENCE j Arnold System Description, E-11, PP. 10, 14 ANSWER 6.02 (1.00)
C (1.0)
REFERENCE
. Arnold System Description C-4, P. 15 ANSWER 6.03 (2.00)
- 1. Two rupture discs are arranged to relieve directly into the RCIC l equipment area if both discs were to break. (1.0) l 2. If the upstream rupture disc were to break, pressure switches located between the two rupture discs would actuate a RCIC system
~ isolation. (1.0)
REFERENCE Arnold - System Description B-2, P. 27 ANSWER 6.04 (2.00)
- a. Mitigate the thrust loading on the containment. (0,5)
- 6. .ni"9Ct1 Y ittt G C6at*i+*enL =Lmu=Phere. w f yr f u Ie a p ooi (0.5)
l
[
st__ELANT__SYSTgdS DESIGNz_ggNTRQL 1_AND INSTRUMENTATION PAGE 22
.. ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
i REFERENCE
, Arnold - System Description A-6, P. 11 ANSWER 6.05 (2.50)
} 1. C
- 2. G
- 3. F
- 4. A
- 5. E E5 @ 0.5 each] (2.5) j REFERENCE Arnold - System Description A-1, PP. 9, 12, 24, 26, 35 ANSWER 6.06 (3.00)
- e. By monitoring the differential pressure across each recirc pump for a 2 psid or greater dp, indicating the pump is running. (1.0)
- b. By measuring the pressure differential between the corresponding jet pump risers in Recirculation Loops A and B E1.03. The undamaged loop will have a higher pressure than the damaged loop
. [0.53. (1.5)
- c. Loop B (0.5)
REFERENCE Arnold - System Description, C-1, PP. 25-27 4
ANSWER 6.07 (2.00) , ,.
I
- c. Loss of fully open indication of the discharge valve removes the l
bypass of the 20% speed limiter. The "E 'S recirc pump speed will decrease to 20% , (1.0)
- b. T_h - ""II r unni ng " porti on of - t.;,e 45X speed--I i mi ter-bypass w rpmrw=ri, -but-rectre pump upeed 1smotwHected -unt1 water-1 eve 14 s lass--then--186-inches r -et .*ehich-time-the 8' A"--recirc- pump
_speari _ wi 21 br weseW45%.-- (1.0)
Loth re s Ir t f e e.pj d w cap cfod to [5 Olo ,
REFERENCE Arnold System Description A-2, Figure 19.
i n.* !!
. , - . . . , - . . - _ _ _ ,. . , . ~ - , - . _ - . - , ,
"6 PLANT _ SYSTEMS DESIGNi _ CONTROLi _AND_ INSTRUMENTATION PAGE 23
. ' ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
r ANSWER 6.08 (3.00)
- a. No change ON du rea>e) Js* +- d e c r e a > *4
- a le Y i d C**IIM (1.0)
- b. Decreases initially due to false low steam flow indication, then returns to the same as when level e at lower point. [It'0aJ Pad r
- P ir* li ni ti al ae l y'/ 'f "ds ch'quil N" " di #~ brates* fII.0)) 4
- c. Lowers due to decreased feed flow initially, then stabilizes at some new lower point when level error matches steam-feed flow error. (1.0)
I REFERENCE Arnold - System Description D-15, P. 18 and Figure 5 ANSWER 6.09 (2.50)
- a. New reading on Range 7 is 2.5 E0.53.
No auto actions ' der-='= ==a"a -
4-tr ' E0.53. (1.0)
- b. New reading on Range 5 is 39 E0.53 IRM high rod block and HI-HI half-scram will be in E1.03 (1.5)
. REFERENCE Arnold - system Description I-2, PP. 13, 24-25
} ANSWER 6.10 (1.50)
- n. 2 seconds after initiation.
(0.75)
- b. Control Rod Drive seal wear.
, < (0.75)
REFERENCE Arnold - System Description I-7, P. 17 ,
Arn.I L - r Po r - 4 , p . IS -
e
,. .< a
L ' k. PL@NT_ SYSTEMS _QEglGN z _ggNTRQL 2 _ANQ_INSIRUMENTATION PAGE 24
)
h.. ' ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
1
, ANSWER 6.11 (3.00)
- 1. Reactor Vessel Low-Low-Low level CO.253 at 46.5" EO.253
- 2. Main Steam High Radiation E0.253 at 3 X Normal Rated Power Background CO.253
- 3. Main Steamline High Flow [0.253 at 140% Rated Flow E0.253
- 4. Main Steamline_ Tunnel High Temperature E0.253 at 200 F CO.253
- 5. Turbine Building Main Steamline Area High Temperature E0.253 at 200 F E0.253
- 6. Low Main Steamline Pressure at Turbine Inlet E0.253 at 850 psig in Run [0.253
- 7. Main Condenser Low Vacuum E0.253 at 10" Hg Vacuum E0.253
( O R, 2 0 " p.} E6 required @ 0.5 each] (3.0)
{ A bfol Ae}
REFERENCE t
Arnold - System Descriptions A-6, P. 35 k
(
l e
I t
4 l
l I
e i
ers 11,
i'
, Zz__EB9CEDUBES_-_NQBMALx_ABNgBMALx_EMEB@ENCY AND PAGE 25 K BOD 196991GB6_GQNIBg6 L.
, ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
ANSWER 7.01 (1.50)
F
- c. 10% (0.5)
- b. 1. Minimize possibility of a hydrogen explosion 9
Minimize possibility of an untreated radioactivity release.
[ .
[2 9 0.5 each] (1.0)
REFERENCE Arnold IPOI-2, P. 8 s
ANSWER 7.02 (2.00) k i
- c. 1. > 2.5 mrem / hour or 100 mrem in 5 consecutive days
- 2. > 100 mrem /hr but < 1000 mrem /hr
- 3. > 1000 mrem / hour. [3 9 0.5 ea.3
- b. 150 mrem [0.53 (2.0) l REFERENCE Arnold Health Physics Procedure 3106.1, P.2 Arnold Health Physics Procedure 3102.1, P. 2 l
ANSWER 7.03 (2.50)
- c. -1. Stop affected work if a personnel or safety hazard exist.
- 2. Make a duplicate tag.
- 3. Note " duplicate tag issued" next to the pertinent entry an the Equipment Tagging Form. [3 9 0.5 each3 (1.5)
Hold Card is used to safeguard human g life E0.53, while a Warning b.
Tag is used for operational reasons where human life is not endangered. [0.53 (1.0)
REFERENCE Arnold - Tagout Procedure, PP. 1, 15 ANSWER 7.04 (1.50)
By at least two independent indications E0.53 misoperation in automatic mode is confirmed [0.53, OR adequate core cooling is assure,d [0.53. (1.5)
- ,a r ! -
l
l#'Zm__EBggEQUB@@-NORM 862_ABNQBMALx_gMER@ENCYANQ PAGE 26 i
RADIOLQ@IGAL_ggNTBg6
. ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
I REFERENCE Arnold OI-51 (Core Spray), P. 9
- f. '
t ANSWER 7.05 (3.00)
- c. 1. For core power levels of 80% of rated an above, the speed of the faster pump shall not exceed 122% of the speed of the slower pump. (0.5)
L
- 2. For core power levels below 80% of rated, the speed of the f faster pump shall not exceed 135% of the speed of the slower pump. (0.5)
- b. To assure the LPCI loop selection logic is able to detect recirculation loop breaks. (1.0)
- c. Operation in the prohibited region is not allowed except during coastdown when a recirc pump trip occurs during two pump operation. (1.0)
REFERENCE Arnold 0I-64 (Recirculation), PP. 9-10, Appendix 1
. ANSWER 7.06 (2.50) l c. Disconnect the amphenol connections for the directional control valves. (1.0)
- b. !_ If = ennten! -cd is unceupl edr-
.2. If_ a . rod _cannot be mnued-with-dr-tve wai.wr pr ==sure.
- 3. Pum i t i en-i ndist i vi . for-* full in' ..id "in11 v u i. ' .=
un a"= 41 =h l a. Jee Ada(Ael :A e ds [2 required 9 0.5 each3 (1.0)
- c. Eight (0.5)
I REFERENCE Arnold AOP 255.1, P. 11 ,
Arnold Tech. Specs., P. 3.3-3 ,
O tr4 !!
r - -
n - -
y q67 7 r ' ' , _ _ _ _ _ _ _ _ _ _ _ _ _ _
f 5.QIR, ; .? . . . ,.' -
i N .. 'N-(@j "The control rod directional control valves for inoperable 7'O b control rod shall be disarmed electrically." This in itself is I one instance.
- However, listing any condition which causes a control rod to be declared IN0P would also be one instance for each condition listed O per DAEC T.S., then the correct response could also be
- 1. Control rod cannot be moved with drive pressure.
- 2. Control rods with INOP Accumulators.
h 3. Control rods whose position cannot te positively determined.
- 4. Rod Uncoupled.
l
- 5. Full In - Full Out indication unavailable.
g Any two of these five should be acceptable.
l % -- - _ _ _ . __ __
l j,[J ;,, 7,[ A control rod on which maintenance is being performed shall be
,,yy considered INOP.
8, g Prior to performing control rod or control rod drive maintenance E on control cells without removing fuel assemblies, the l
~
E d ct ' "' ' c "*' ' v a ' ""' " "" ct c' "' d r 'd a t least on the other control rod drives in the 5 x 5 rod array centered on the controlirod or rod drive undergoing maintenance.
Then it shall be demonstrated that the core can be made l
I E subcritical by a margin of 0.38 percent.
E of this list of eight be acceptable.
.a - __a - -
rr,4 11 *
,j '* '
--_-- $6- y - - # - r..u.q.._ -. .
- 7. PBggEDUBES_ _NQBM0bs_8HNgBM86t_EMEBGENCY AND PAGE 27-B8DigLggig86_QgNIBg6 ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
ANSWER 7.07 (2.50)
- o. Prompt removal of heat from the primary containme't n E0.53 and minimizes suppression pool heatup E0.53. (1.0)
- b. 4 ; _ m. == uuoling flow tu g u.;.r [S.53 nd ruffici;nt all T i sa
- r 1*ricetisu ei.d v4ive'hydratrites-fer53. (1.0)
- c. Assures SBLC pump availability should it be needed again. (0.5)
- b. A<ssa cyd e'ny o4~ b rbin e <>haud ej,a y ads (0,5]
REFERENCE A JJ vre a d q %5. Ty gene g,4 e ,ii Arnold EOP-C, PP. 5-9 pre y; ,, , ( g,y]
G:neral Electric - Emergency Operating Procedure Fundamentals far rul l OT 50, fob ANSWER 7.08 (2.00)
[F- 1. Verify Low-Low set relief valves are controlling reactor pressure p 2. Manually open relief valves
REFERENCE Arnold IPOI-5, P. 8 ANSWER 7.09 (1.50) r L 1. Reactor power remains above 5%.
- 2. Reactor power cannot be determined.
- 3. Any control rod not inserted to or beyond Position 02 or position cannot be determined. [3 @ 0.5 each] (1.5)
REFERENCE l Arnold IPOI-5, P. 5 ANSWER 7.10 (1.50) l
- b. To minimize thermal stresses at the feedwater nozzles.
i c. To prevent tripping the operating RFP, HPCI, and RCIC.
[ 3@ 0.5 each] (1.5) l l
l l
1 ns !!
l i
- Zz-- PBOCEDWBES__- NgBd8(z_@@NgBd@Ls_gdEBQENCY_AND PAGE 28 r B8DI96001G86_GQNIBQL f , ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
REFERENCE Arnold IPOI-4, PP. 14-17 i
ANSWER 7.11 (2.00)
A. Tm el--- n=rc=nnn=1 3 re-the-area so-that-Ecre-Spray suctii '
l
,,, u_ _..: t_u_a 1___ ie_ rer >_ +w_ $__..-
l2 iA, d.uT 5 Ga'r'4 ' E. s '\ o r' ~ il
~
(1.0) niolle n~y~ '( 0,T) od P ro b* A *
- e,\
Torus and drywell work control point HP's verify that all'" ) ,Io il a n ! + m.(!
ri d ?s B.
- personnel are out by using the control point log. (1.0)
REFERENCE 4
Arnold- System Description B-5 (FRCHP #5), P. 25.
ANSWER 7.12 (2.50)
A. 1. The channel fastener is on the corner near the center of I
the cell.
- 2. The space buttons on the side of the channel are situated such that they will contact the spacer buttons of the adjacent assemblies.within the cell to maintain proper bundle spacing.
- 3. The tab on the side of the lifting bail points toward the center of the cell (3 3 0.5 ea.)
B. Visually checked to see that the assembly is at the same height as neighboring assemblies. [0.53 Checked twice (once at time of seating and again later with
- underwater camera) [0.53 (1.0)
~ '
REFERENCE i Arnold- FRCHP #5, PP. 10-11.
l l -
l 19 4 ! !
- g. AQd1NigIB8I12E_E8gGEQQBggz_GQNDITigNgz_8Np_ Lid 11811gNS PAGE 29
- e. ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
i.
ANSWER 8.01 (1.00)
{
Unidentified sources not to exceed 5 gpm.
Total leakage not to exceed 25 gpm. [2 9 0.5 each3 (1.0)
REFERENCE Arnold - Technical Specifications, P. 3.6-5 ANSWER 8.02 (1.50) i A continuous fire watch must be established within one hour on at least one cide of the penetration unt work is comp 1 ed and the penetration is resealed.
CCA~rM l,o
- Y kinJ00.5 d M hFAC-q0s* Jt)nTQW (1.5)
% V) thD ( pem.tM% lt3 OdSEC 1h REFERENCE Arnold - Technical Specifications, P. 3.13-8 i
ANSWER 8.03 (2.50)
(Oe50
- c. During single loop operation, the Operating Limit MCPR is increased to account for increased uncertaintyiin core flow C0.53 and Traversing
/ Incore Probe (TP) readings [0.53 which are used to derive the L '
Saf ety Limi t MCPR -C'" "' _ (1.5)
- b. _In order to protect the fuel from inadvertent core flow increases I
such that the Safety Limit MCPR requirement can be assured. (1.0)
REFERENCE Arnold - Technical Specifications, P P . ' 3'. 1 2 - 7 , 8 ANSWER 8.04 (1.50)
- After two diagonally adjacent (0.53, previously exposed E0.53, fuel ocsemblies have been loaded into their previous core positions next to each of the four SRM's [0.53. (1.5)
A REFERENCE Arnold - Technical Specifications, P. 3.9-4 na et .
.m . _ _ _ _ . _ ~ . - -..
Hi__8D51NIEIB8I12E_EBQQEQW8Egz_QQNQ1I1QNgi_8NQ_61d1I8I1QNE PAGE 3D
- e. ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN,'J.
h n
ANSWER 8.05 (2.50)
- c. System is OPERABLE. (A redundant component is inoperable.) [0.53 Operation may continue provided squib valve is returned to operable condition within seven days [0.53, otherwise, reactor shall be in
+ cold shutdown condition with all operable control rods fully inserted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> CO.53. (1.5)
- b. System is INOPERABLE EB.53. Reactor shall be placed in a Cold Shutdown Condition with all operable control rods fully inserted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> CO.53. (1.0)
REFERENCE Arnold - Technical Specifications, PP. 3.4-1+, Figure 3.4-2 ANSWER B.06 (2.00)
- 1. Immediately reduce reactor power to </= 1658 MW (1.0)
- 2. Event must be reported to NRC as soon as practical but always within one hour CO.53 due to plant operating outside authorized Technical Specifications. [0.53 (1.0)
REFERENCE Arnold - Facility Operating License, P. 3 10 CFR 50.72 (Notification of Significant Events) l' ANSWER 8.07 (2.00) f A Safety Limit violation is assumed to,h, ave occurred if a scram is cccomplished by means of a backup feature of plant design E1.03. In this ccse, a scram should have occurred due to M3IV's < 90% fully open, but l instead scram occurred from a backup featur,e.(APRM's) [1.03. (2.0)
REFERENCE Arnold - Technical Specifications, PP. 1.1-6, 1.1-18 l
,,a ut.
. . , - - - - - - . - . . . - - - - ~ - - - --
2*Ez__8Dd1NISIBBI1Yg_PBQQEDQBESz_GQNDITIQNE,_AND_LIMITATIONE PAGE 31 i*
",,. ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
ANSWER 8.08 (2.00)
SOfety limits are limits below which the reasonable maintenance of the y cicdding and primary systems are assured C1.03. The Limiting Safety System
]'
S ttings are settings on instrumentation which initiate the automatic protective action at a level such that safety limits will not be exceeded
[1.03. (2.0) g REFERENCE f Arnold - Technical Specifications, P. 1.0-1 i _
ANSWER 8.09 (2.50)
I
- c. A single line is drawn through the error E0.53. The person making the entry initials the crossed out line [0.53 The corrected entry follows or is placed above the cross out CO.53. (1.5)
- b. 1. The new log book is the next numbered book in the series CO.53
- 2. Each page is numbered in correct order, starting from 001.
[0.53. (1.0)
REFERENCE Arnold - Admin. Procedure 1410.3, P. 3 AN5WER 8.10 (2.00)
- 1. Individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight.
- 2. Individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. ,
- 3. Individual should not be permitted to work more.than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period.
- 4. There should be a break of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> between work periods.
[4 9 0.5 each]
(2.0)
REFERENCE Arnold - Admin. Procedure 1410.1, P. 3 na r!
Et . AptjlNigIR811ME_P89GgpgRESz_GQNDITIONS _AND_ i LIMIT 8TIQNS PAGE 32
['o i /. ANSWERS -- DUANE ARNOLD -86/06/24-SHERMAN, J.
1 3 ANSWER 8.11 (1.00) y h
i
' Since the STA must be within 10 minutes of the Control Room [0.53, his location and ability to be notified by the control room must be considered
[D.53. OR Arus e;+l L Jec.J, o 1 j Me a~ ror sederd ra Aog w /6 n (1.0) o to miehl 04 e ee%I < a a m , [ o ,Q i REFERENCE Arnold - Admin. Procedure 1410.1, P.2 (6.t QR % e.ehri a) ykt ms .; i ANSWER 8.12 (2.50) Car of e /s a, hill / sa C/le
- c. 1. A member of Plant Management Staff CO.53
- 2. An Operation Shift Supervisor E 0. 5 3' (1.0)
- 6. 1. Revisions which would change the safety intent or function of controls or equipment as described in Technical Specifications E e. 'S h
- 2. Those that change the original intent of the procedure
! ? . '5 h- (1.5)
^
3, Te ,, go v e r y re s is . .. > to NP3 .rl U ns+ s lin Two or mo rt . Arrih et a y rre n,a oe e .,- c a -12 s G w kt REFERENCE Arnold - Admin. Procedure 1406.3, PP. 2-3 /
- f. H.i.,k e itte, ,4 i de .s .. so y , s. , I . . -
T. 1400 M.s. I Pr. c <<t w e > ,
ANSWER 8.13 (2.00)
(2 m.'.,d e 0.1 T < * -)
- c. Within one hour, four hocrs, or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, depending on event [0.53.
Telephone notification [0.53 (1.0)
- b. Within thirty days CO.53 . ..
Written notification [0.53 (1.0)
REFERENCE ,
Arnold - Admin. Procedure 1402.3, PP. 4-11 O
{ 0 *h Y
~ -.-
,s.
- TEST CROSS REFERENCE PAGE 1
'T UESTION VALUE REFERENCE J ',_ _ _ ______ --
L C5.01 1.50 MJS0000317 f C5.02 2.00 NJS0000318
~05.03 1.00 MJS0000319 -
I C5.04 3.00 MJS0000320 C5.05 1.50 MJS0000321 C5.06 3.00 MJS0000322 C5.07 1.50 MJS0000323 C5.08 1.00 MJS0000324 C5.09 1.00 MJS0000325 C5.10 2.00 MJS0000326 I. C5.11 2.50 MJS0000327 C5.12 2.00 MJS0000328 C5.13 1.50 MJS0000329 C5.14 2.00 MJS0000330 25.50 C6.01 3.00 MJS0000331 06.02 1.00 MJS0000332 C6.03 2.00 MJS0000333 C6.04 2.00 MJS0000334 06.05 2.50 MJS0000335 C6.06 3.00 MJS0000336 06.07 2.00 MJS0000337 C6.08 3.00 MJS0000338
, C6.09 2.50 MJS0000339 C6.10 1.50 MJS0000340
. C6.11 3.00 MJS0000341 25.50 1
G7.01 1'.50 MJS0000342
, 07.02 2.00 MJS0000343 G7.03 2.50 MJS0000344 G7.04 1.50 MJS0000345 07.05 3.00 MJS0000346 97.06 2.50 MJS0000347 , ,
l G7.07 2.50 MJSOOOO348
[ 07.00 2.00 MJS0000349 j 07.09 1.50 MJS0000350 ,
i C7.10 1.50 MJS0000351 '
07.11 2.00 MJS0000365
( 07.12 2.50 MJS0000366 l 25.00 4
( CB.01 1.00 MJS0000352 CB.02 1.50 MJS0000353 C8.03 2.50 MJS0000354 C8.04 1.50 MJS0000355 .
l nd Il
~
I TEST CROSS REFERENCE PAGE 2 QUESTION VALUE REFERENCE 08.05 2.50 MJS0000356
- 08.06 2.00 MJS0000357 C8.07 2.00 MJS0000358 C8.08 2.00 MJS0000359 C8.C9 2.50 MJS0000360 C8.10 2.00 MJS0000361 C8.11 1.00 MJS0000362 CB.12 2.50 MJS0000363 CB.13 2.00 NJS0000364
- = = - - -
25.00 101.00 f.
L I.
L O
d S
10 4 f f *
., ~ . - . _ - - - , . - , , , , - . - , ., ...-_ .,---.
, . - --- - - . . . . . - - . . - .