ML13259A199
| ML13259A199 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/03/2013 |
| From: | Bielby M NRC/RGN-III/DRS/OLB |
| To: | |
| Shared Package | |
| ML11354A431 | List: |
| References | |
| Download: ML13259A199 (277) | |
Text
2013 DAVIS-BESSE NUCLEAR POWER STATION INITIAL LICENSE EXAMINATION PROPOSED EXAM FILES
~ 5501 North State Route 2 FirstEnergy Nuclear Operating Company Oak Harbor, Ohio 43449 Raymond A. Lieb 419-321-7676 Vice President, Nuclear
-OPERATOR LICENSE EXAMINATION MATERIAL- Fax: 419-321-7582
- WHEN SEPARATED FROM ENCLOSURE, HANDLE THIS DOCUMENT AS UNRESTRICTED -
April 10, 2013 10 CFR 55 L-13-140 Mr. Michael Bielby Chief Examiner, Region Ill U. S. Nuclear Regulatory Commission 2443 Warrenville Road, Suite 210 Lisle, IL 60532-4352
Subject:
Davis-Besse Nuclear Power Station, Unit 1 Docket Number 50-346, License Number NPF-3 Submittal of Written Operator License Examinations. Operating Tests. and Supporting Reference Material
Dear Mr. Bielby:
Enclosed are the written examinations, operating tests, and supporting reference material prepared by the Davis-Besse Nuclear Power Station (DBNPS) staff for the licensed operator examinations to be administered during the weeks of June 3 and June 10, 2013.
The enclosed items and supporting reference material, which are considered confidential, are being submitted to the NRC for review and approval in accordance with 10 CFR 55.40, "Written Examinations and Operating Tests-Implementation" and NUREG 1021, Operator Licensing Examination Standards for Power Reactors (Revision 9, Supplement 1).
Modifications to the previous submitted outlines were made as a result of feedback received from the chief examiner and the validation process. These modifications to the outlines are identified in bold italic print.
-OPERATOR LICENSE EXAMINATION MATERIAL-
- WHEN SEPARATED FROM ENCLOSURE, HANDLE THIS DOCUMENT AS UNRESTRICTED -
I RECEIVED APR 12 2013
-OPERATOR LICENSE EXAMINATION MATERIAL-
-WHEN SEPARATED FROM ENCLOSURE, HANDLE THIS DOCUMENT AS UNRESTRICTED -
Davis-Besse Nuclear Power Station, Unit 1 L-13-140 Page 2 of 2 The materials enclosed shall be withheld from public disclosure until after the scheduled examinations are complete.
There are no regulatory commitments included in the submittal. If there are any questions or if additional information is required, please contact Mr. Anthony Stallard, Superintendent- Nuclear Operations, at (419) 321-7161 or Mark Klein, Lead Exam Developer at (419) 321-7773.
Sincerely, 4A~~t1&
Raymond A. Lieb vaw
Enclosure:
NUREG 1021 Forms and Operator License Examination Outline cc: Regional Administrator, NRC Region Ill (w/o Enclosure)
Chief, Operations Branch, NRC Region Ill (w/o Enclosure)
DB-1 NRC/NRR Senior Project Manager (w/o Enclosure)
DB-1 Senior Resident Inspector (w/o Enclosure)
USNRC Document Control Desk (w/o Enclosure)
Utility radiological Safety Board (w/o Enclosure)
-OPERATOR LICENSE EXAMINATION MATERIAL-
- SHALL BE WITHHELD FROM PUBLIC DISCLOSURE UNTIL AFTER THE SCHEDULED EXAMINATIONS ARE COMPLETE-
-OPERATOR LICENSE EXAMINATION MATERIAL-
- SHALL BE WITHHELD FROM PUBLIC DISCLOSURE UNTIL AFTER THE SCHEDULED EXAMINATIONS ARE COMPLETE -
Enclosure L-13-140 NUREG 1021 Forms and Operator License Examination Outline List of Enclosed NUREG Forms Form ES-201-3, Examination Security Agreement (Up-to-date Copies)
Form ES-301-1, Administrative Topics Outline RO (Rev. 1)
Form ES-301-1, Administrative Topics Outline SRO (Rev. 1)
Form ES-301-2, Control Room/In-Plant Systems Outline RO (Rev. 1)
Form ES-301-2, Control Room/In-Plant Systems Outline SR0-1 (Rev. 1)
Form ES-301-2, Control Room/In-Plant Systems Outline SRO-U (Rev. 1)
Form ES-301-3, Operating Test Quality Checklist (Rev. 1)
Form ES-301-4, Simulator Scenario Quality Checklist (Rev. 1)
Form ES-301-5, Transient and Event Checklist (Rev. 1)
Form ES-301-6, Competencies Checklist (Rev. 1)
RO Written Outline Form ES-401-2, PWR Examination Outline RO (Rev. 1)
Form ES-401-3, Generic Knowledge and Abilities Outline Tier 3 RO (Rev. 1)
SRO Written Outline Form ES-401-2, PWR Examination Outline SRO (Rev. 1)
Form ES-401-3, Generic Knowledge and Abilities Outline Tier 3 SRO (Rev. 1)
Form ES-401-4, Record of Rejected K/As (Rev. 1)
Form ES-401-6, Written Examination Quality Checklist (Rev. 1)
Form ES-D-1, Scenario Outline (Rev. 1)
List of Enclosed Exam Materials 9 Administrative Topics JPMs with applicable procedures 11 Control Room/In-Plant Systems JPMs with applicable procedures 4 simulator Scenarios 100 Written Examination Questions, answers, and reference pages for each question's correct answer References provided to the candidates for the written examination
ES-301 Operating Test Quality Checklist Form ES-301-3 Facility: Davis Besse Date of Exam 6/3 thru 6/14 2013 Operating Test No.: _ _ __
Initials
- 1. GENERAL CRITERIA c#
- a. The operating test conforms with the previously approved outline; changes are consistent with sampling requirements (e.g., 10 CFR 55.45, operational importance, safety function distribution).
- b. There is no day-to-day repetition between this and other operating tests to be administered during this examination.
- c. The operating test shall not duplicate items from the applicants' audit test(s) (see Section D.1.a).
- d. Overlap with the written examination and between different parts of the operating test is within acceptable limits.
- e. It appears that the operating test will differentiate between competent and less-than-competent applicants at the designated license level.
- 2. WALK-THROUGH CRITERIA
- a. Each JPM includes the following, as applicable:
- initial conditions
- initiating cues
- references and tools, including associated procedures
- reasonable and validated time limits (average time allowed for completion) and specific designation if deemed to be time critical by the facility licensee
- specific performance criteria that include:
- detailed expected actions with exact criteria and nomenclature
- system response and other examiner cues
- statements describing important observations to be made by the applicant
- criteria for successful completion of the task
- identification of critical steps and their associated performance standards
- restrictions on the sequence of steps, if applicable
- b. Ensure that any changes from the previously approved systems and administrative walk-through outlines (Forms ES-301-1 and 2) have not caused the test to deviate from any of the acceptance criteria (e.g., item distribution, bank use, repetition from the last 2 NRC examinations) specified on those forms and Form ES-201-2.
- 3. SIMU CRITERIA
- a. The associated simulator operating tests (scenario sets) have been reviewed in accordance with Form ES-301-4 and a copy is attached.
Printed Name I Signature Date
- a. Author
~
- b. Facility Reviewer(*) ih43
- c. NRC Chief Examiner(#) 4/;.r/?J
- d. NRC Supervisor~ a,- ~~(~
NOTE:
- The facility signature is not applicable for NRC-developed tests.
- Independent NRC reviewer initial items in Column "c"; chief examiner concurrence required.
NUREG-1 021, Revision 9 Supplement 1 FENOC Facsimile Rev.1
ES-301 Simulator Scenario Quality Checklist Form ES-301-4 r:acility. Davis Besse Date of Exam 6/3 thru 6/14 2013 Operating Test No.: _ _
QUALITATIVE ATTRIBUTES Initials a b* c#
- 1. The initial conditions are realistic, in that some equipment and/or instrumentation may be out of service, but it does not cue the operators into expected events. /;JJ' w ~La
- 2. The scenarios consist mostly of related events.
/J1:J l.;vtF IJJ!~!J
- 3. Each event description consists of '
. the point in the scenario when it is to be initiated
. the malfunction(s) that are entered to initiate the event the symptoms/cues that will be visible to the crew I)JJ ~
fttct; the expected operator actions (by shift position)
. the event termination point (if applicable)
- 4. No more than one non-mechanistic failure (e.g., pipe break) is incorporated into the scenario without a credible preceding incident such as a seismic event. l'jJI M fh0
- 5. The events are valid with regard to physics and thermodynamics. 1}1-'J I~ lfh~
~ I~ ~
- 6. Sequencing and timing of events is reasonable, and allows the examination team to obtain complete evaluation results commensurate with the scenario objectives.
- 7. If time compression techniques are used, the scenario summary clearly so indicates.
Operators have sufficient time to carry out expected activities without undue time constraints. Cues are given.
/J)ff lrJ/~ ~
- 8. The simulator modeling is not altered. ~t)_ ~ [l!_o/i
~ r7f lfWb The scenarios have been validated. Pursuant to 10 CFR 55.46(d), any open simulator performance deficiencies or deviations from the referenced plant have been evaluated to ensure that functional fidelity is maintained while running the planned scenarios.
- 10. Every operator will be evaluated using at least one new or significantly modified scenario. All other scenarios have been altered in accordance with Section 0.5 of ES-301. .~ "0 ~
~~ ~
- 11. All individual operator competencies can be evaluated, as verified using Form ES-301-6 (submit the form along with the simulator scenarios).
~ fV(J 111(6
- 12. Each applicant will be significantly involved in the minimum number of transients and events specified on Form ES-301-5 (submit the form with the simulator scenarios).
- 13. The level of difficulty is appropriate to support licensing decisions for each crew position.
TARGET QUANTITATIVE ATTRIBUTES (PER SCENARIO; SEE SECTION Actual Attributes D.S.d)
- 1. Total malfunctions (5-8) 6 6 6 5 (fJ{J- In- rtr;6 I.A-J
- 2. Malfunctions after EOP entry (1-2) 2 2 1 2 ~D ~ fl;tt;~
- 3. Abnormal events (2-4) 3 3 3 3 I~~
'I, 1d rro- 14t~
- 4. Major transients (1-2) 2 1 1 1 IIJ?b. n iUl/1
- 5. EOPs entered/requiring substantive actions (1-2) 1 1 1 1 1178 n l1rh
' E O P contingencies requiring substantive actions (0-2) 1 1 0 1 ~ 'd f6 Critical tasks (2-3) 3 2 3 3
~!J 1/"f ~
NUREG-1021, Revision 9 Supplement 1 FENOC Facsimile Rev.1
ES-301 Competencies Checklist Form ES-301-6
- Facility Davis Besse Date of Exam 6/3 thru 6/14 2013 Operating Test No.:
APPLICANTS RO SR0-1 SRO-U Notes:
(1) Includes Technical Specification compliance for an RO.
(2) Optional for an SRO-U.
(3) Only applicable to SROs.
Instructions:
- Check the applicant's license type and enter one or more event numbers that will allow the examiners to evaluate every applicable competency for every applicant.
NUREG-1 021, Revision 9 Supplement 1 FENOC Facsimile Rev.1
ES-401 Written Examination Quality Checklist Form ES-401-6 cility: Davis Besse Date of Exam 6/3 thru 6/14 2013 Exam Level: RO lXI SRO lXI Item Description
- ~~----------------~~~~
- 1. accurate and applicable to faci
- 2. a. NRC K!As referenced for all questions
- b. objectives referenced as available
- 4. The sampling process was random and systematic (If more than 4 RO or 2 SRO questions are repeated from the last two NRC licensing exams, consult the NRR OL program office.)
- 5. Question the license screening/audit exam was controlled as indicated below
(~ck the item that applies) and appears appropriate:
_the audit exam was systematically and randomly developed; or Xthe audit exam was completed before the license exam was started; or
_the examinations were developed independently; or Xthe licensee certifies that there is no duplication; or other
- 6. Bank use meets limits (no more than 75 Bank Modified New percent from the bank, at least 10 percent new, and the rest new or modified); enter the actual RO I SRO-only question distribution(s) 16/0 010 59/25 at right.
(78.7%/100%)
- 7. Between 50 and 60 percent of the questions CIA on the RO exam are written at the comprehension/analysis level; the SRO exam may exceed 60 percent If the randomly selected K/As support the higher 32/1 43/24 cognitive levels; enter the actual RO I SRO (42.7% I 4%) (57.3% I 96%)
question distribution(s) at right.
- 8. References/handouts provided do not give away answers or aid in the elimination of distractors.
- 9. Question content conforms with specific KIA statements in the previously approved examination outline and is appropriate for the Tier to which they are assigned; deviations are
- 10. Question and format meet the guidelines in ES ix B.
- 11. The exam contains the required number of one-point, multiple choice items; the total is correct and rees with value on cover sheet
- a. Author
- b. Facility Reviewer (*)
- c. NRC Chief Examiner
- d. NRC Sup~rt'f' 11----------~--"-----------f-------11 Note:
- The facility reviewer's initials/signature
- Independent NRC reviewer initial items in Column "c"; chief examiner concurrence required.
NUREG-1021, Revision 9 Supplement 1 FENOC Facsimile Rev. 1
Davis Besse 1LOT13 NRC Written Exam Rev. 1
- 1. The following plant conditions exist:
- RCS pressure is 250 psig
- Pressurizer temperature is 406 oF
- Quench Tank pressure is 80 psig
- Containment pressure is 14.7 psia
- The crew has just finished drawing a Pressurizer steam bubble.
The following event occurs:
- The Pressurizer Safety fails open and the Quench Tank rupture disc ruptures.
What will be the Pressurizer Safety Valve downstream temperature, for these conditions?
B. -325 OF C. -345 OF D. -406 OF
~
.,xplanation/Justification:
A. Incorrect. Plausible because this is the saturation temperature for 14.7 psia which the candidate could select if isenthalpic throttling is not considered.
B. Correct answer lAW Steam tables and isenthalpic throttling process. When the Pressurizer Safety Valve fails open, the rupture disc on the safety valve will blow releasing pressurizer steam to the CTMT atmosphere. Therefore the downstream conditions will be the CTMT conditions.
C. Incorrect. Plausible because this is the saturation temperature for 80 psig (Quench Tank pressure)
D. Incorrect. Plausible because this the temperature at which the event started Sys# System Category KA Statement 000008 Pressurizer AK1. Knowledge of the operational implications of the following Thermodynamics and flow characteristics of open (PZR) Vapor concepts as they apply to a Pressurizer Vapor Space Accident: or leaking valves Space Accident KIA# AK1.01 KIA Importance 3.2 Exam Level RO References provided to Candidate Steam Tables Technical
References:
Steam Tables Question Source: New Level Of Difficulty: (1-5) 3 3 Question Cognitive Level: High -Application 10 CFR Part 55 Content: (CFR 41.8 I 41.10 I 45.3)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1
- 2. A small break loss of coolant accident has occurred.
Which of the following describes the function of the Steam Generator required to mitigate this event?
A. Steam Generators are not required to mitigate any loss of coolant accidents.
B. For certain small break LOCAs, heat removal by the SGs is necessary to satisfy the acceptance criteria of 10CFR50.46, Acceptance Criteria for Emergency Core Cooling Systems.
C. Only the isolation of Containment provide by the Main Steam Isolation Valves is required to mitigate a loss of coolant accident.
D. Boiler-Condenser Cooling provided by the Steam Generators is required to ensure condensed steam is returned to the Reactor Vessel to provide adequate RCS inventory for loss of coolant accidents.
Answer: B Explanation/Justification:
A. Incorrect- Maintaining SG's available as a heat removal capability is required to ensure that Core cooling is provided if flow out the break is not sufficient.
B. Correct per DB-OP-02000 Bases and Deviation Document Step 5.6 and 5.7. Maintaining SG's available as a heat removal capability will ensure that Core cooling is provided if flow out the break is not sufficient.
c.
Sys #
000009 Incorrect- Although the MSIVs will isolate Containment, without a break in the Steam Generator or Main Steam Line, Containment Integrity is not affected by the position of the MSIV.
Incorrect- Although the condensed steam is returned to the reactor vessel, adequate RCS inventory requires HPI or LPI operation.
System Small Break LOCA Category EK2. Knowledge of the interrelations between the small break LOCA and the following:
KA Statement S/Gs KIA# EK2.03 KIA Importance 3.0 Exam Level RO References provided to Candidate Technical
References:
DB-OP-02000 R19 Bases and Deviation None Document Steps 5.6 and 5.7 for SBLOCA requirement to maintain SG available.
10CFR50.46 Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: Low - Fundamental 10 CFR Part 55 Content: (CFR 41.7 I 45.7)
Objective:
58 of 509 Bases and Deviation Document for DB-OP-02000 Rl9 STEP 5.6 Verify proper SG level control by AFW using Specific Rule 4, Steam Generator Control.
Purpose:
Raising SG level to the loss of SCM setpoint ( 124/130 inches) will establish the necessary inventory for Boiler Condenser Cooling (BCC). This level will also assist in establishing primary side natural circulation, which would be necessary to obtain Primary to Secondary Heat Transfer.
Bases: For most events, including most LOCAs, the core can be adequately cooled by using HPJ or LPI cooling. For certain small break LOCAs, heat removal by the SGs is necessary to satisfy the acceptance criteria of 10CFR50.46. SG levels must be increased to the Joss of SCM setpoint at the required minimum SG fill rate until the setpoint is reached . Specific Rule 4 requires full continuous AFW flow until setpoint is reached. Full continuous flow will provide approximately 800 gpm per SG as limited by AFW cavitating venturies Full continuous flow exceeds the minimum of 225 gpm to each SG with 2 SG(s) in service or 450 gpm to a single SG with 1 SG in service. The minimum SG fill rate is the rate necessary to ensure that maximum expected energy is removed from the reactor coolant.
The loss of SCM setpoint provides sufficient surface area for Boiler Condenser Cooling (BCC). Condensing the steam in the RCS via BCC will reduce RCS pressure so that HPI flow rate will increase to a value where its heat removal rate will match decay heat production in time to ensure the core remains covered.
If SFRCS trips on Low SG Pressure due to MUIHPI Cooling or raising SG level (not a secondary side malfunction), direction is provided (step 5.7) to reestablish AFW flow to the isolated SG and raise level. This action will allow use of the isolated SG for SG Heat Transfer if necessary.*
AFW should be used because the elevation of the AFW nozzles is high enough to provide the required condensing surface without level established in the SGs. The level setpoint is high enough to provide the requ ired condensing surface during periods of no AFW flow . At the loss of SCM setpoint, the amount of BCC, combined with HPJ cooling, will keep the core cooled and covered .
Setpoints: None
References:
- 3. I 0 CFR 50.46
- 4. Safety Evaluation SE 87-0292, Safety Evaluation for FCR 86-330 Auxiliary Feedwater Level Control
59 of509 Bases and Deviation Document for DB-OP-02000 R19
- Deviations:
5.
6.*
No B&W Document 51-1224886-02, OTSG Refill Summary Report USAR 15.2.8.2.3, Loss ofNormal Feedwater, Results Analysis The EOP TBD Volume 2 Section ill.BJ describes limiting AFW flow can be limited to minimize SG cooling during periods when.no primary to secondary heat transfer exists.
In accordance with Specific Rule 4.1, full continuous AFW flow is maintained until appropriate SGs levels are reached. This is acceptable based on the following:
I. The raised loop design at Davis-Besse requires much lower SG levels to promote natural circulation than other B&W plants. Therefore automatic AFW level control setpoints are lower, less AFW is added to reach setpoint, and overcooling due to AFW addition is not a serious concern. An historical review of AFW initiations prior to 1988 supports the conclusion that AFW flow at the maximum rate until the desired setpoint is reached, does not cause significant overcooling at Davis-Besse even with low decay heat and AFW flow at maximum (~1200 gpm/SG).
- 2. A plant modification added cavitati~g venturies in the 1988 outage to limit AFW flow to approximately 800 GPM/SG versus about 1200 GPM/SG previously obtained. This modification further reduces the concern of AFW flow causing overcooling.
- 3. Accident analysis described in USAR 15.2.8.2.3 requires 600 GPM AFW flow in < 40 seconds during a loss of feed water event. The automatic AFW flow control system is designed to provide maximum flow until setpoint is reached .
.Throttling of AFW flow when less than setpoint would require the operator to override the automatic safety system control. It is not desired to add procedure guidance to allow or require AFW throttling when below the desired SG level setpoint, based on SG pressure criteria. Throttling, at all, is in conflict with requirements to obtain maximum flow. A SG pressure decrease could (and most likely would) be caused by excessive steam flow. Throttling AFW flow to limit a SG pressure decrease caused by excessive steam flow is inappropriate and could lead to unnecessary SG dry out.
Davis Besse 1LOT13 NRC Written Exam Rev.1
- 4. The plant is operating at 100% power with all systems in normal alignment for this power level.
Which of the following abnormal conditions requires an IMMEDIATE power reduction and stopping the affected Reactor Coolant Pump?
A. MU59A, RCP 2-1 Seal Return Isolation Valves fails closed.
B. Computer Point L828, 2-1 Motor Lower Bearing Low Oil Level Alarm with stable bearing temperatures.
C. Computer Point T828, 2-1 Motor Stator Temperature Alarm with indicated temperature 350 °F.
D. Computer Points for 2-1 Seal Cavity Pressure P833 (second stage) reads 1100 psig, and P834 (third stage) reads 50 psig Answer: C Explanation/Justification:
A. Incorrect- Shutdown is required within 30 minutes, not immediately.
B. Incorrect- Shutdown is required if bearing temperatures are rising with low oil level, not immediately.
C. Correct- Power reduction and Shutdown is immediately required per DB-OP-02515 Step 4.6.1 RNO.
D. Incorrect- Values provided indicated a single RCP Seal Stage is failed. Immediate Shutdown is not required for single stage failure per DB-OP-02515, Step 4.1.1 System Category KA Statement
.s#
0015/ Reactor Coolant Generic Knowledge of abnormal condition procedures.
0017 Pump (RCP)
Malfunctions KIA# 2.4.11 KIA Importance 4.0 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02515 R11, RC Pump and Motor Malfunctions Step 4.6.1 RNO Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.10 /43.5/45.13)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1
- 5. Following the loss of BOTH Makeup Pumps from full power operations, why is RCS pressure reduced to 1700 to 1800 psig?
Reducing RCS Pressure will ______________________
A reduce Reactor Coolant Pump seal leak off, preserving RCS Inventory.
B. allow the Reactor Protective System to be place in Shutdown Bypass.
C. allow the Safety Features Low RCS Pressure Trip to be blocked.
D. allow the High Pressure Injection system to restore RCS Inventory.
Answer: D Explanation/Justification:
A. Incorrect- Plausible because this would reduce sealleakoff, RCS lnvertory is preserved by isolating Letdown for this event.
B. Incorrect- Plausible because in a normal shutdown, Shutdown Bypass Operation can be established at this RCS Pressure range.
C. Incorrect- Plausible because in a normal shutdown, RCS Pressure is reduced to slow the transition when blocking the SFAS Low RCS Pressure Trip at 1670 psig prior to SFAS Actuation at 1600 psig ..
D. Correct- The ability to add inventory to the RCS is established by starting High Pressure Injection in piggyback mode which will then provide approximately 1800 psig discharge pressure allowing flow to the RCS.
Sys# System Category KA Statement
.0022 Loss of AK3. Knowledge of the reasons for the following responses as they Actions contained in SOPs and EOPs for RCPs, Reactor apply to the Loss of Reactor Coolant Makeup: loss of makeup, loss of charging, and abnormal Coolant charging Makeup KIA# AK3.02 KIA Importance 3.5 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02512 R14, Makeup and Purification System Malfunctions Attachment 6.
Question Source: New Level Of Difficulty: (1-5) 2 Question Cognitive Level: Low - Fundamental 10 CFR Part 55 Content: (CFR 41.5, 41.10 I 45.6 I 45.13)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1
- 6. The following plant conditions exist:
- A plant cooldown is in progress for refueling.
- Reactor Coolant Pumps 2-1 and 2-2 are in service.
- DH Train 2 is in service
- RCS temperature is 180 oF
- Pressurizer level is 80 inches
- RCS pressure is 220 psig The following event occurs:
- A loss of Off-Site Power occurs
- EDG 2 fails to start Based on these conditions:
In accordance with DB-OP-02527, Loss of Decay Heat Removal, what is the PRIORITY for how core heat removal will be established?
A. Maintain current RCS temperature Conditions using Turbine Bypass Valves and Natural Circulation .
- C.
Allow RCS to heatup to Mode 4, then use Atmospheric Vent Valves and Natural Circulation to control RCS temperature.
Allow RCS to heatup to Mode 4, then use Makeup, High Pressure Injection, and the High Point Vents to establish Feed and Bleed Cooling.
D. Maintain current RCS temperature Conditions using Makeup, High Pressure Injection, and the PORV to establish Feed and Bleed Cooling.
Answer: B Explanation/Justification:
A. Incorrect -This outcome would be desired to avoid transition back into Mode 4, but the loss of offsite power has caused a loss of Circ Water Pumps and therefore the main condenser. TBVs will close once Condenser Pressure rises to 17 inch HgA.
B. .Correct- Step by step priority as listed in DB-OP-02527 R15, Loss of Decay Heat Removal step 4.1.7 RNO C. Incorrect- At low RCS Pressures, the flow out the High Point Vents will be insufficient to remove core decay heat. The RCS would heatup beyond Mode 4 (greater than 280 F). Candidate may assume PORV is not available for this scenario. The PORV is DC Powered D. Incorrect- Although Feed and Bleed cooling would be successful in removing decay heat, it is likely the PORV flow at low RCS pressures would not be sufficient to allow continued cooldown, In addition, SG heat transfer is prioritized above Feed and Bleed Cooling in DB-OP-02527, Loss of Decay Heat Removal.
Sys# System Category KA Statement 000025 Loss of Residual AA 1. Ability to operate and I or monitor the following as they apply RCS/RHRS cooldown rate Heat Removal to the Loss of Residual Heat Removal System:
System (RHRS)
A# AA 1.01 KIA Importance 3.6 Exam Level RO ferences provided to Candidate None Technical
References:
DB-OP-02527 R15, Loss of Decay Heat Removal
- step 4.1.7 RNO.
Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR41.7 /45.5/45.6)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev. 1
- 7. The following plant conditions exist:
- The plant is operating at 100% power.
The following indications occur:
- 11-4-B, CCW PMP 1 FLOW LO, annunciator is in alarm with a flowrate of 2400 gpm.
- 2-3-A, LETDOWN TEMP HI, annunciator is in alarm with a temperature of 144 °F.
- 11-1-B, CCW HX 1 OUTLET TEMP HI, annunciator is in alarm with a temperature of 122°F.
Which one of the following actions will automatically occur?
A. CCW Pump 1 will trip.
B. The standby CCW pump will start.
C. CCW Non-Essential Header will isolate.
D. Letdown cooler inlet isolation valve, MU 28, will close.
Answer: D planation/Justification:
- Incorrect- Plausible because the CCW Pump is operating at a low flow condition. It would be logical to have the pump trip to protect the pump.
B. Incorrect- Plausible because the CCW Pump is operating at a low flow condition, but above the Flowrate to cause and automatic start of the Standby Pump C. Incorrect- Plausible because the CCW system is operating abnormally. Closing the non-essential header isolation could protect the essential functions provided by CCW.
D. Correct- CCW temperatures associated with the letdown cooler will rise with reduced flow that would lead to a high temperature isolation of letdown flow.
Sys# System Category KA Statement 000026 Loss of AA 1. Ability to operate and I or monitor the following as they apply to Flow rates to the components and systems that Component the Loss of Component Cooling Water: are serviced by the CCWS; interactions among Cooling the components Water(CCW)
KIA# AA1. 07 KIA Importance 2.9 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02002 R08 Annunciator 2-3-A Letdown/Makeup Alarm Panel 2 Annunciators Question Source: BANK 37623 Level Of Difficulty: (1-5) 3 Question Cognitive Level: High -Comprehension 10 CFR Part 55 Content: (CFR 41.7 /45.5/45.6)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev. 1
- 8. The plant is operating at 100% power with all systems in normal alignment for this power level.
The selected RCS Pressure Instrument from the Reactor Protective System to Non-Nuclear Instrument System INSTANTANEOUSLY fails HIGH.
Which of the following describes how the plant will respond to this failure?
The Pressurizer PORV will _____ (1 )_ _ __
The Pressurizer Spray Valve will _ _ _ _ (2) _ _ __
The Pressurizer Heaters will _____(3) _ _ __
A. (1) remain closed (2) remain closed (3) remain energized B. (1) open (2) open (3) de-energize C. (1) open (2) remain closed (3) de-energize
- (1) remain closed (2) open (3) remain energized Answer: 8 Explanation/Justification:
A. Incorrect- Plausible if the candidate believes the RCS signal is SASS protected like most other NNI signals. Instantaneous failures would normally cause a SASS transfer for SASS protected instrument inputs resulting in no change to the input for the PORV, Spray Valve, or Pressurizer Htrs.
B. Correct- The selected RPS Pressure signal is used to control the PORV, the PZR Spray Valve, and the Pressurizer Heaters. A high failure will cause the PORV to Open, the Pressurizer Spray Valve to Open, and the Pressurizer Heaters to turn off.
C. Incorrect-. Plausible if the candidate thought that the safety grade Reactor Protective System RCS Pressure signal is used to control the PORV.
Since PORV has the most impact on the plant, this conclusion is logical.
D. Incorrect- Plausible if the candidate thought that the safety grade Reactor Protective System RCS Pressure signal is used to control only the Pressurizer Spray Valve. The remaining positions would be correct if supply with a different pressure signal.
Sys # System Category KA Statement 000027 Pressurizer AK2.03 Knowledge of the interrelations between the Pressurizer Controllers and positioners Pressure Pressure Control Malfunctions and the following:
Control System (PZR PCS)
Malfunction KIA# AK2.03 KIA Importance 2.6 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02513, Pressurizer System Abnormal Operation Att. 2 page 54 uestion Source: New Level Of Difficulty: (1-5) 3 uestion Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.7 /45.7)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1
- 9. The plant is operating at 100% power with all systems in normal alignment for this power level.
The Reactor Protective System (RPS) generates a valid reactor trip signal, but the Control Drive Trip Breakers fail to open.
In accordance with DB-OP-02000, RPS, SFAS, SFRCS Trip or SG Tube Rupture, the Reactor Operator in the Control Room momentarily deenergizes 480 volt Unit Substations E2 AND F2.
Following restoration of power to E2 and F2, which of the following previously running loads will return to operation without operator action?
A. Radwaste Exhaust Fan.
B. Main Station Exhaust Fan.
C. Clean Waste Monitor Tank Transfer Pump.
D. Spent Fuel Pool Pump.
Answer: D Explanation/Justification:
. Incorrect- There is no seal in feature for this fan. The Radwaste Ventilation System would be lost until the fan is restarted
. Incorrect- There is no seal in feature for this fan. The Main Station Exhaust System would be lost until the fan is restarted C. Incorrect- There is no seal in feature for this pump. This RCS inventory addition source would be lost until the pump is restarted.
D. Correct- The controller for the SFP Pumps have a seal in feature that would restart the pump following restoration of power.
Sys# System Category KA Statement 000029 Anticipated EK2. Knowledge of the interrelations between ATWS and the Breakers, relays, and disconnects Transient following:
KIA# EK2.06 KIA Importance 2.9* Exam Level RO References provided to Candidate None Technical
References:
Eng Change Package 10-0654 Question Source: New Level Of Difficulty: (1-5) 2.5 Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR 41.7 I 45. 7)
Objective:
Davis Besse 1 LOT13 NRC Written Exam Rev.1 10 . Plant conditions are as follows:
- The reactor is tripped.
- RCS pressure is 1990 psig.
- RCS Tave is 548 °F.
- BOTH SGs are being steamed through the Turbine Bypass Valves.
Which one of the following will occur if SG 1 exceeds 250 inches?
A. The MSIV on SG 1 closes so that ONLY SG 2 may be steamed to the condenser.
B. SFRCS will realign Aux Feedwater to ONLY feed SG 2.
C. The MSIVs on BOTH SGs close and prevent steaming of BOTH SGs to the condenser.
D. The AFW level control setpoint for SG 2 is set to 124 inches and BOTH the AFW supply and Main Steam isolation close for SG 1.
Answer: C Explanation/Justification: SFRCS will actuate on high SG level. The setpoint is 250 inches.
A. Incorrect- Plausible since only the #1 SG MSIV closes since that is the only SG operating at a high level.
. *.. Incorrect- Plausible since #1 SG is at a high level, we should stop feeding it by aligning both AFW pumps to feed #2 SG.
Correct- High level in either SG will close both MSIVs.
D. Incorrect- Plausible because elevated level in #2 SG will promote heat transfer that may be needed if #1 SG is removed from service.
Sys # System Category KA Statement 000038 Steam EA2. Ability to determine or interpret the following as they apply to a Status of MSIV activating system Generator SGTR:
Tube Rupture (SGTR)
KIA# EA2.12 KIA Importance 3.9* Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02000 R26 Table 1 SFRCS Response BANK 36449 Level Of Difficulty: (1-5) 2.5-3 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 43.5/ 45.13)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1 11 . INITIAL CONDITIONS:
- RCS temperature 500 oF
- RCS pressure 1000 psig
- RCS cooldown in progress
- A Main Steam Line Break on #2 SG in Containment occurs.
CURRENT CONDITIONS:
- RCS temperature 425°F
- RCS pressure 750 psig Assuming no change in Main Steam Line break size, from initial to current condition, subcooling margin has __(1) and steam flow out the break has __(2) _ __
A. (1) risen (2) lowered B. (1) risen (2) risen C. (1) lowered (2) lowered
SFRCS would not actuate on Low SG Pressure for this scenario.
A. Correct- Subcooled margin for the initial conditions would be approximately 45 degrees while SCM for current conditions would be approximately 85 degrees. Steam Flow would be reduced as SG Pressure Lowers.
B. Incorrect- While SCM will have risen as noted in A above, steam line break flow will be dependant on SG pressure. As the RCS cools, SG pressure will lower and therefore break flow will lower, however candidate may assume SCM drives the Steam flow rate like aSGTR -That is, reducing SCM reduces leak rate.
C. Incorrect- Candidate may select this response assuming lower RCS temperatures produces lower SCM .. While steam line break flow will be dependant on SG pressure. As the RCS cools, SG pressure will lower and therefore break flow will lower.
D. Incorrect- Candidate may select this response assuming lower RCS temperatures produces lower SCM.. While steam line break flow will be dependant on SG pressure. As the RCS cools, SG pressure will lower and therefore break flow will lower, however candidate may assume SCM drives the Steam flow rate like a SGTR- That is, reducing SCM reduces leak.
Sys# System Category KA Statement 000040 Steam Line Generic Ability to evaluate plant performance and make Rupture- operational judgments based on operating Excessive characteristics, reactor behavior, and instrument Heat Transfer interpretation.
KIA# 2.1.7 KIA Importance 4.4 Exam Level RO References provided to Candidate Technical
References:
Fundamental Theory- Steam Table and General Steam Tables Physics HTFF Chapter 4 page 3 Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: Low - Fundamental 10 CFR Part 55 Content: (CFR: 41.5 I 43.5 I 45.12 I 45.13)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev. 1 The plant is operating at 55%.
The following event occurs:
- Both Main Feedwater Pumps trip Without other induced changes in plant conditions, The Control Rod Trip breakers open DIRECTLY due to an ___(1) _ _
The Turbine trips DIRECTLY due to ___(2) __
A. (1) RPS trip (2) CRD Trip Confirm B. (1) RPS trip (2) SFRCS trip C. ( 1) ARTS trip (2) SFRCS trip D. ( 1) ARTS trip (2) CRD Trip Confirm Answer: D Explanation/Justification:
A. Incorrect- The RPS trips would not actuate until plant condition such as RCS Pressure changed. As a result, RPS would not directly trip the reactor for this scenario. This is the basis for installing the ARTS System.
B. Incorrect- The RPS trips would not actuate until plant condition such as RCS Pressure changed. As a result, RPS would not directly trip the reactor for this scenario. This is the basis for installing the ARTS System. SFRCS does directly generate a Turbine Trip Signal.
C. Incorrect- ARTS would trip the reactor, but tripping both MFP will not directly trip SFRCS. An SFRCS Trip will directly trip the Main Turbine.
D. Correct -ARTS senses MFP Turbine status and causes a reactor trip if both MFP Turbine Trip. CRD Trip Confirm will cause EHC to trip the Main Turbine.
Sys# System Category KA Statement 000054 Loss of Main AA2. Ability to determine and interpret the following as they apply to Occurrence of reactor and/or turbine trip Feedwater the Loss of Main Feedwater (MFW):
(MFW)
KIA# AA2.01 KIA Importance 4.3 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-06202 pg72, DBBP-TRAN-0034 pg 6&7 Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 43.5 I 45.13)
Objective:
Number:
DAVIS-BESSE BUSINESS PRACTICE DBBP-TRAN-0034
Title:
Revision: Page Davis-Besse Operator Fundamentals Memory List 06 6 of 26 ATTACHMENT 2: LICENSED OPERATOR MEMORY LIST Page 1 of 11 REACTOR PROTECTION SYSTEM Name Setpoint Manual Manual High Flux 104.9% (4 RCP) 80.6% (3 RCP)
RC High Temperature 618oF Flux I~ Flux /Flow
- Variable per COLR figure 6 (Doghouse Curve)
RC Low Pressure
- 1900 PSIG RC High Pressure 2355 PSIG RC Pressure-Temperature* Variable High Flux/Number of RC 55.1% 1/1 RCP combination Pumps On*
0% 010, 0/1, 1/0, 2/0, 0/2 RCP combination(s)
Containment Pressure High 4 PSIG Shutdown Bypass High 1820 PSIG Pressure
- Trips are bypassed when Shutdown Bypass is actuated.
RPS Channel Number Power Supply Trip Breaker 1 Y1 B 2 Y2 A 3 Y3 D 4 Y4 c
Davis Besse 1LOT13 NRC Written Exam Rev.1 13 . The plant is operating at 100% power with all systems in normal alignment for this power level.
A Tornado hits the Switchyard damaging all three offsite lines causing a loss of offsite power.
Approximately 1 minute after the Reactor Trip, the following conditions are noted:
- A Bus = zero volts
- B Bus = zero volts
- C 1 Bus = zero volts
- 01 Bus= zero volts
- 1-3-H, 01 Bus Lockout
- Breaker A0213, SBOOG to 02 BUS TIE BREAKER tripped open due to a 02 Lockout.
Which of the following strategies must be implemented to restore power to an essential 4160 volt bus?
A. Start EOG1 to restore power to Bus C1.
B. Start the SBOOG to restore power to Bus C1.
C. Start the SBOOG to restore power to Bus 01.
- Start EOG 2 to restore power to Bus 01.
Answer: A Explanation/Justification:
A. Correct- The SBODG is not available due to lockout on Bus 02 which causes AD213 being open. EDG2 is not available due to 01 being locked out.
B. Incorrect- The SBODG is not available due to lockout on Bus 02 indicated by breaker AD213 being open.
C. Incorrect- The SBODG is not available due to lockout on Bus 02 indicated by breaker AD213 being open.
D. Incorrect- EDG2 is not available due to 01 being locked out.
Sys # System Category KA Statement 000055 Loss of EA 1. Ability to operate and monitor the following as they apply to a Restoration of power with one EDIG Offsite and Station Blackout:
Onsite Power (Station Blackout)
KIA# EA1.06 KIA Importance 4.1 Exam Level RO References provided to Candidate Technical
References:
DB-OP-02000 R26 Specific Rule 6 Step 6.1 RNO None Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.7 I 45.5 I 45.6)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1 The plant had been operating at 100% power The following event occurs:
- Loss of off-site power
- All systems work as designed
- Natural circulation flow has been confirmed in accordance with DB-OP-06903, Plant Cooldown.
Which one of the following actions will raise the heat transfer rate from the Reactor Coolant System to the Steam Generators?
A. Lowering Steam Generator steaming rates B. Lowering Steam Generator water levels C. Raising Steam Generator pressures D. Raising Steam Generator Auxiliary Feedwater flow rates Answer: D Explanation/Justification:
. Incorrect- Lowering SG Steaming rate will cause a rise in SG pressure and a lowering of differential temperature between the RCS and the SG, reducing the heat transfer rate.
- . Incorrect- Lowering SG level will reduce the heat transfer surface area of the SG, reducing the overall heat transfer coefficient, reducing the heat transfer rate.
C. Incorrect- A rise in SG pressure will lower the differential temperature between the RCS and the SG, reducing the heat transfer rate.
D. Correct- Raising AFW Flow rates will provide additional cooling flow and level in the SG providing a larger heat sink inducing a higher heat transfer rate.
Sys # System Category KA Statement 000056 Loss of AK1. Knowledge of the operational implications of the following Principle of cooling by natural convection Offsite Power concepts as they apply to Loss of Offsite Power:
KIA# AK1.01 KIA Importance 3.7 Exam Level RO References provided to Candidate Technical
References:
Lesson Plan OPS-SYS-1103.09 pages 23 & 24 None Question Source: BANK 36546 Level Of Difficulty: (1-5) 3 Question Cognitive Level: Low - Fundamental 10 CFR Part 55 Content: (CFR 41.8 I 41.10 I 45.3)
Objective:
REACTOR COOLANT SYSTEM OPS-SYS-1103.09
( 1) Lower Steam Generator pressure until Steam Generator DB-OP-02000 T sAT is 40-60°F lower than in core temperature Section 6 (2) After Reactor Coolant Pump bump, lower Steam Generator pressure until Steam Generator T SAT is 90-100°F lower than incore temperature.
- c. Bump Reactor Coolant Pumps to induce heat transfer Requires TSC concurrence
- d. Boiler-condenser heat transfer could occur. This is cyclic in nature (i.e. flow, no flow, flow, conditions.)
L. Subcooling Margin
- 1. Minimum of 20°F for forced flow M. Reactor Coolant System Leakage and Leak Rate Calculations
- 1. Covered by technical specifications and a daily surveillance test N. Technical Specifications
- 1. Technical Specifications
- a. 3.4.1 -Reactor Coolant System Pressure, Temperature, and Slide 57 flow Departure from Nucleate Boiling Limits Link to Technical Specification (1) 4 Pump Limits (a) Pressure<:: 2064.8psig (b) Temperature::;; 610°F (c) Flow<:: 389,500 gpm (2) 3 Pump Limit (a) Pressure<:: 2060.8psig (b) Temperatures 610°F (c) Flow<:: 290,957 gpm (3) MODE 1
- b. 3.4.2 - Reactor Coolant System Minimum Temperature for Criticality
( 1) Each Reactor Coolant System loop average temperature (TAVE) shall be 2!: 525°F.
(2) Applicability (a) MODE 1 (b) MODE 2 with kett <:: 1.0
- c. 3.4.3- Reactor Coolant System Pressure, Temperature, and heatup and cooldown rates shall be maintained within the limits specified in the Pressure/Temperature Limits Report.
(1) At all times Page 24 of 28
Davis Besse 1 LOT13 NRC Written Exam Rev. 1 15 . The following conditions exist:
- The plant is operating at 100% power during the Winter.
- SW Pump 1 (Loop) is supplying Primary loads
- SW Pump 2 (Loop) is supplying Secondary loads
- SW Pump 3 breaker is racked out.
A rupture downstream of SW1399, SW HDR 1 TO TPCW HX occurs.
All automatic actions occur as designed.
Which ONE of the following describes the automatic response if any of Service Water and Circulating Water Systems?
A. No Impact -SW1 continues to carry Primary Loads, SW 2 continues to carry Secondary Loads.
B. SW 1 continues to carry Primary Loads, SW 1395 SW HDR 2 TO TPCW HX closes to isolate the break, and CT2955, TPCW HX SUPPLY FROM CIRC WTR opens to allow Circ Water to carry TPCW load.
C. SW1 carries Train 1 Essential Loads. SW Train 2 carries Train 2 Essential Loads only.
CT2955, TPCW HX SUPPLY FROM CIRC WTR opens to allow Circ Water to carry TPCW load.
D. SW1 carries Train 1 Essential Loads. SW Train 2 carries Train 2 Essential Loads only.
SW1395 close to isolate Secondary Header. CT2955, TPCW HX SUPPLY FROM CIRC WTR initially opens, but then closes to isolate the leak. TPCW Cooling is lost.
Answer: D Explanation/Justification:
A. Incorrect- The piping downstream of SW1399 and SW1395 is common, The breaks prevents either SW Line from supplying these loads. In addition, the break will cause the CT2955 Check Valves to sense low pressure causing a loss of Circ Water Supply as well. Plausible if candidate assumes the supplies are independent and a break on the out of service supply will not affect the in service supply.
B. Incorrect- Plausible because Circ Water provide backup cooling for TPCW loads when SW supply is lost, but not when lost due to line break.
C. Incorrect- Plausible because Circ Water provide backup cooling for TPCW loads when SW supply is lost, but not when lost due to line break.
D. Correct -.The piping downstream of SW1399 and SW1395 is common, The breaks prevents either SW Line from supplying these loads. In addition, the break will cause the CT2955 Check Valves to sense low pressure causing a loss of Circ Water Supply to TPCW as well Sys# System Category KA Statement 000062 Loss of AK3. Knowledge of the reasons for the following responses as they The conditions that will initiate the automatic Nuclear apply to the Loss of Nuclear Service Water: opening and closing of the SWS isolation valves Service to the nuclear service water coolers Water KIA# AK3.01 KIA Importance 3.2* Exam Level RO References provided to Candidate None Technical
References:
OS020 SH2 R45 Service Water CL6 & CL9 Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.4, 41.8 I 45.7 )
.jective:
Davis Besse 1LOT13 NRC Written Exam Rev.1 16 . Which ONE of the following describes the purpose of the Main Generator Under-excited Reactive Ampere Limiter (URAL)?
A. Establishes a MINIMUM megawatt output (loading) for the main generator to prevent a reverse power condition.
B. Establishes a MINIMUM LAGGING Power Factor to maintain grid stability.
C. Prevents voltage Regulator output from RISING to a level which would cause excessive armature heating.
D. Prevents the Voltage Regulator output from LOWERING to a level which could cause the Main Generator to drop out of synchronization (slip poles) with the grid.
Answer: D Explanation/Justification:
A. Incorrect- Plausible if the Candidate believe excitation levels are related to minimum loading level to prevent a reverse power condition.
B. Incorrect- Plausible if the Candidate believe excitation levels are related to Power Factor to ensure the limiting power factors for Generator Operation are observed.
C. Incorrect- Plausible if the Candidates assume the limiter acts to reduce current flow and therefore heat.
D. Correct- The under excited reactive ampere limit circuit acts to limit the amount of under excitation permitted on the generator. This limit is for the purpose of allowing the generator to be safely operated, continuously in an under excited condition, with sufficient margin between the excitation limit and the stability limit of the generator.
s# System Category KA Statement
- 0077 Generator AK1. Knowledge of the operational implications of the following Under-excitation Voltage and concepts as they apply to Generator Voltage and Electric Grid Electric Grid Disturbanc Disturbances es:
KIA# AK1.03 KIA Importance 3.3 Exam Level RO References provided to Candidate None Technical
References:
System Description SD005 R4, Main Generator page 2-15 Question Source: BANK32205 Level Of Difficulty: (1-5) 2.5- 3 Question Cognitive Level: Low - Fundamental 10 CFR Part 55 Content: (CFR: 41.4, 41.5, 41.7, 41.10 /45.8)
Objective:
Davis Besse 1 LOT13 NRC Written Exam Rev.1 17 . A loss of ALL feedwater has occurred. Both MU pumps are running.
Attempts are being made to restore feedwater to both SGs in accordance with DB-OP-02000, RPS, SFAS, SFRCS Trip or SG Tube Rupture.
The following plant conditions exist:
- RCS pressure is 2200 psig and lowering
- The PORV (RC2A) is open.
- T-hot is 615 oF and rising in Loop 1 and 610 oF and rising in Loop 2.
Based on these conditions, what will be the status of PORV Block and PORV control switches?
The PORV block valve (RC 11) control switch will be _ _.._(1.:..J.)_ _ ; the PORV (RC2A) control switch will be (2)
A. (1) OPEN (2) AUTO B. (1) OPEN (2) LOCK OPEN C. (1) CLOSED (2) AUTO Answer: B (1) CLOSED (2) LOCK OPEN Explanation/Justification: At Davis-Besse, the beyond design bases Loss of all Feedwater event is mitigated via MU/HPI PORV Cooling. This question is related to control of the PORV and Flowpath through the PORV Block A. Incorrect -The position of the PORV Block is correct, but having the PORV in Auto will cause the valve to close when RCS Pressure reaches 2155 psig stopping the MU/HPI PORV Cooling Flowpath.
B. Correct -These positions are the DB-OP-02000 Attachment 4 position of the valves during MUIHPI PORV Cooling.
C. Incorrect- The position of the PORV Block is incorrect stopping flow and having the PORV in Auto will cause the valve to close when RCS Pressure reaches 2155 psig also stopping the MU/HPI PORV Cooling Flowpath.
D. Incorrect- The position of the PORV Block is incorrect, stopping the MU/HPI PORV Cooling Flowpath.
Sys # System Category KA Statement BW/E04 Inadequate EA 1. Ability to operate and I or monitor the following as they apply to Components, and functions of control and safety Heat Transfer the (Inadequate Heat Transfer) systems, including instrumentation, signals,
-Loss Of interlocks, failure modes, and automatic and Secondary manual features.
Heat Sink KIA# EA1.1 KIA Importance 4.4 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02000 R26 Attachment 4 page 271 Question Source: BANK 37388 Level Of Difficulty: (1-5) 3 Question Cognitive Level: Low- Fundamental 10 CFR Part 55 Content: (CFR: 41.7/45.5/45.6)
Objective:
Davis Besse 1 LOT13 NRC Written Exam Rev.1 Following a normal Reactor Trip, from full power operation, DB-OP-02000, RPS, SFAS, SFRCS, Trip or SG Tube Rupture, directs the operator to use the plant computer to record the following computer point voltage values:
- J213, J215, and J217 for J Bus Voltage
- J221, J223, and J225 for K Bus Voltage What is the reason for recording these voltages?
To complete the Surveillance required to verify compliance with _ _ _ _ _ _ __
A. TS 3.8.1, AC Sources Operating. The normal opening of ACB 34560 and 34561 following a reactor trip will impact operability of off site power sources.
B. TS 3.8.2, AC Sources Shutdown. The normal transfer of A and B Buses to the Startup Transformers following a Reactor Trip, may have rendered the Shutdown AC Sources inoperable.
C. TS 3.8.9 Distribution Systems Operating. The normal opening of ACB 34560 and 34561 following a reactor trip will impact operability of the Distribution system- Operating.
TS 3.8.1 0, Distribution Systems Shutdown. The normal transfer of A and B Buses to the Startup Transformers following a Reactor Trip, may have rendered the Shutdown Distribution Systems inoperable.
Answer: A Explanation/Justification:
A. Correct -The opening of the Generator Output Breakers disrupts the normal ring bus configuration which may impact Off-Site Sources. As a result, this surveillance verifies the Off-Site Sources remain operable. -
B. Incorrect-TS 3.8.2 is only applicable in Modes 5 and 6. The candidate may select this TS since the Main Generator is shutdown.
C. Incorrect - The opening of the Generator Output Breakers disrupts the normal ring bus and does not affect the Distribution Systems Operating which are the in plant electrical distribution.
D. Incorrect-TS 3.8.1 0 is only applicable in Modes 5 and 6. The candidate may select this TS since the Main Generator is shutdown and power is being supplied from the Startup Transformers vice the Auxiliary Transformers.
Sys # System Category KA Statement BW/E1 0 Post-Trip Generic Ability to use plant computers to evaluate system Stabilization or component status.
KIA# 2.1.19 KIA Importance 3.9 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02000 R26 Attachment 26 Page 400 Bases and Deviation Document for DB-OP-02000 R19 Attachment 26 page 501 Question Source: New Level Of Difficulty: (1-5) 2 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.10 /45.12)
Objective:
6 DB-OP-06000 2.2.6 Nuclear Instrumentation should be monitored closely during a fill with fuel in the core. If an unexplained rise in neutron count rate occurs, filling shall immediately stop and the cause determined.
2.2.7 Fill water boron concentration shall be maintained such that the Shutdown Margin is maintained within the limits specified in the COLR. Applicability:
MODES 3, 4, and 5. (TS 3.1.1) MODE 6 (TS 3.9.1) 2.2.8 RCS and fill water samples shall be taken as directed by DB-OP-06904, Shutdown Operations.
2.2.9 The RCP and Motor must be coupled prior to initiating seal injection. This must be done to prevent hydraulically thrusting the shaft seal. The upward thrust of the shaft may cause the seal stationary faces to be pushed to their upper axial limits, resulting in failure of the seal. This Limit and Precaution does not apply during performance of Section 4.8, Gravity Fill from the BWST through High Pressure Injection to the RCP Seals as long as pressure at the seals is maintained less than 35psig(reference DB-MM-09012 Caution 8.3).
2.2.10 When filling and venting the RCS, be aware that the upper RCP recirculation impeller tapered seat rests on the upper pump cavity casing when a RCP motor is decoupled and will block the vent/drain pathways to the respective RCP upper volute volume. When filling there is a potential for trapped air to remain within the upper pump volute, such that movement of the impeller could vent this area causing RCS level changes. There is also a potential for creating a siphon effect across the generator levels via the RCS vessel because of a solid system.
46 DB-OP-06904 Revision 42 7.0 PREPARATIONS FOR FUEL HANDLING OPERATIONS Limits and Precautions 7.1 In MODE 6, the boron concentration of all filled portions of the RCS and the Refueling Canal shall be maintained uniform and sufficient to ensure that the requirements of
~72 ~
\ *..,....~:,~?.~~',~!i.'Z~~~~*~,......_._..-1 7.3 Do not fill the Refueling Canal by addition of water to the RCS after the Reactor Vessel Head has been removed. Filling the Refueling Canal in this manner has the potential of spreading contamination from the Reactor Vessel into the Refueling Canal area.
INITIALS/DATE Prerequisites I 7.4 The Reactor Coolant System level is at 78 to 82 inches.
I 7.5 The Reactor Coolant System is vented to CTMT Atmosphere.
Date _ _ _ _ __
Prerequisites completed b y - - - - - - - - - - - - - - - - - - -
Procedure I 7.6 Verify a Class B cleanliness inspection has been performed and ifunacceptable, a documented evaluation completed for the Refueling Canal Deep End walls and floor.
Circle one:
Acceptable Unacceptable with Evaluation I 7.7 Verify all tools, equipment and materials are removed from the deep end of the Refueling Canal or are properly stored in the appropriate racks.
I 7.8 Notify Mechanical Maintenance to perform the following:
7.8.1 Verify Fl81, REFUELING CANAL TRASH RACK SCREEN, has been removed.
7.8.2 Verify the six inch blank flange has been installed on Refueling Canal end of the drain line to the Reactor Vessel Cavity.
7.9 Danger Tag closed the following valves: (refer to Limit and Precaution 2.1.13).
- -I- -
- SF1, FUEL TRANSFER TUBE 2 ISOLATION
_I_
- SF2, FUEL TRANSFER TUBE 1 ISOLATION I 7.10 Verify the Fuel Transfer Tube Blind Flanges have been removed in accordance with DB-MM-09186, Fuel Transfer Tubes Blind Flanges Removal and Reinstallation.
I 7.11 Verify the Refueling Canal Deep End Perforated Drain Cover is installed to prevent debris from entering the drain piping.
ED7575A DAVIS-BESSE ADMINISTRATIVE PROCEDURE PAGE REVISION PROCEDURE NUMBER Fuel Handling Operations 5 12 DB-OP-00030 3.2 Implementation 3.2.1 DB-NE-001 00, Fuel Handling Administration 3.2.2 DB-NE-03292, Refueling Prerequisites and Periodic Checks 3.2.3 DB-NE-06101, Fuel/Control Component Shuffle 3.2.4 DB-NE-06302, Manual Movement of Control Components 3.2.5 DB-NE-06471, Dry Fuel Storage Unloading 3.2.6 DB-NE-06472, Dry Fuel Storage Loading 3.2.7 DB-OP-06021, Spent Fuel Pool Operating Procedure 4.0 DEFINITIONS 4.1 INDEPENDENT VERIFICATION- The process used to obtain a separate and independent check, by an individual not involved in the initial positioning, to ensure the Fuel Handling Bridge is actually in the position specified. The individual performing the position check must have minimum interaction with the personnel performing the initial positioning.
4.2 VISUAL VERIFICATION - The process of visually checking the position of a device or
- ~~,"'*
.* . ...f1tl:.-\~.'f~ji ~omponent in..i*ltthe
. . . ..~'-;J,:~.Jif-;***~*\~'>'/i'*~W*~",~~* condit~on.or P?sition ~P~SW~*
-~.,J~Ch;*r,*~ .' ,f,~*-*** *~. '.;\"f.hr~~-. *>. -~.* ~~'>ft_.rt:**:"':. ,J,... ::-::-**- .. , .. r.
- "'""~.;;;: *:.> r,'.Lt~:~;.,, '*W.!l.i~*~"""'~"'::\'*
- -* '"tm-~ .
. 4.3 FUEL MOVEMENT SEQUENCE SHEETS - Tables developed and approved by Nuclear .
f
.' Engineering that are used to track the initial and final locations of Fuel Assemblies and Cont'*
\ Components during the performance of core offload, core reload, or core shuffle. The Fuel
. Handling Director's table will be used to direct the sequence of fuel handling evolutions. A .
\* copy of the Fuel Handling Director's table will also be used by the individual in the Control ~
Room monitoring Nuclear Instrumentation to provide a second check of the evolutions in progress. Selected portions of the Fuel Handling Director's table are provided at each of the L .... :.'il~~~~~~*.~.ii-~~~[~r. ~~~.~.~~.~.~a,~~};ng.~~o~~tion in progress. "*
- 4.5 PARTIAL MOVEMENT OF FUEL OR CONTROL COMPONENTS - The process of performing a portion of a line in the Fuel Movement Sequence Sheets to allow optimal use of two fuel handling bridges in the refueling canal. For example, the Auxiliary Bridge picks up a control component and then moves out of the way, allowing the Main Bridge to pickup and move a Fuel Assembly. Expected partial movements may be identified on the Fuel Movement Sequence sheets.
Davis Besse 1LOT13 NRC Written Exam Rev.1
- 20. The plant is operating at 100% power with all systems in normal alignment for this power level.
Rising Condenser Pressure is noted.
In accordance with DB-OP-02518, High Condenser Pressure, SFRCS is actuated using the Initiate and Isolate push buttons at (1) inches HGA in order to _ ____,(=2.,)_ __
A. (1)10 (2) ensure a source of feedwater remains available for the Steam Generators.
B. (1)10 (2) protect the Condenser from Turbine Bypass Steam Flow C. (1)17 (2) ensure a source of feedwater remains available for the Steam Generators.
D. (1)17 (2) protect the Condenser from Turbine Bypass Steam Flow Answer: A Explanation/Justification:
Correct- The main Feedwater Pumps trip at 12.5 in HGA, but top of scale for Control Room indicators is 10 inches HGA. SFRCS is actuated at 10 inches to ensure a source of FW is available.
- . Incorrect- The Turbine Bypass valves will auto close at 17 in HGA to protect the condenser.
C. Incorrect- The reason is correct, but the setpoint is not correct. The main Feedwater Pumps trip at 12.5 in HGA.
D. Incorrect - The Turbine Bypass valves will auto close at 17 in HGA to protect the condenser. SFRCS actuation is not required for this feature.
Sys # System Category KA Statement 000051 Loss of Generic Ability to interpret control room indications to Condenser verify the status and operation of a system, and Vacuum understand how operator actions and directives affect plant and system conditions.
KIA# 2.2.44 KIA Importance 4.2 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02518 R06, High Condenser Pressure step 4.5 and Attachment 2.
Question Source: New Level Of Difficulty: (1-5) 2.5 Question Cognitive Level: Low - Fundamental 10 CFR Part 55 Content: (CFR: 41.5/43.5/45.12)
Objective:
21 DB-OP-02518 Revision 06 ATIACHMENT 2: BACKGROUND INFORMATION Page 1 of3
Purpose:
To provide direction to the Operator upon rising or higher than normal Condenser pressure due to air in-leakage. This elevated pressure will not be a result ofweather conditions. It will be primarily the result of mechanical or opefl)tional factors indicated by Condenser pressure alarm or equipment actuation (for example, Mechanical Hogger starting at 4.5 inches of HgA.) Normal plant responses to weather conditions resulting in Condenser pressures above those normally expected are addressed by DB-OP-06231, Vacuum System Operating Procedure.
Technical Specifications:
None USAR:
10.4.2 Main Condenser Vacuum System Svstem Descriptions:
SD-026B, Condenser Vacuum Discussion:
The Condenser Vacuum System is comprised of the Mechanical Hogger, Steam Hogger, Steam Jet Air Ejector (SJAE), and Filter System. During normal operation, the SJAE is used to remove non-condensable gases from the Main Condenser and discharge them to the higher pressure outside environment. The Steam Hogger is used during initial start-up when large volumes of ait must be removed to establish initial vacuum to 10 inches HgA, at which point the SJAE is placed in service. Both the Steam Hogger and the SJAE will then combine to draw vacuum to approximately 3 Inches HgA, at which point the Steam Hogger is shut down and the Mechanical Hogger is placed on standby, The Mechanical Hogger automatically starts upon increasing Main Condenser pressure of greater than 4.5 inches HgA.
This procedure is applicable when Condenser pressure is higher than normal for the existing Generator output and weather conditions, but below the Main Turbine high pressure trip setpoint of 7.5 inches HgA. Rising Circulating Water temperature caused by ambient temperature will result in a loss of Condenser efficiency causing Condenser pressure to rise. Operation of the unit up to 5.5 inches HgA is permissible in accordance with DB-OP-06231, Vacuum System Operating Procedure, when caused by such weather conditions if Turbine load is greater than 50% and the Manager- Plant Operations has given approval.
23 DB-OP-02518 Revision 06 ATTACHMENT 2: BACKGROUND INFORMATION Page 3 of3 Continuous operation of the Turbine Generator at low vacuum (high pressure) is to be avoided. Operation in this manner could cause overheating of the LP Turbine elements with possible rotor or blade damage due to distortion or excessive vibration. Normally, CD 517 EXHAUST HOOD SPRAY CONTROL VALVE, will open at 125°F, attempting to alleviate the high temperature condition. This will be indicated by a computer point of the valve position (Z570). If an Exhaust Hood temperature of 125°F is exceeded, steps should be taken to determine advisability of continued Turbine operation. A temperature of 175°F in the Exhaust Hood will actuate an alarm. A Main Turbine Trip will be initiated if Exhaust Hood temperature exceeds 225°F or Condenser Pressure reaches 7.5 inches HgA.
Davis Besse 1LOT13 NRC Written Exam Rev.1 21 . The plant is operating at 100% power with a 10 gpd tube leak in SG1.
- NO planned radioactive liquid releases are in progress.
- RE4686, Storm Sewer Outlet alarms and indicates above its HIGH alarm setpoint.
This alarm indicates leakage from which of the following systems?
A. Miscellaneous Liquid Radwaste System.
B. Clean Liquid Radwaste System C. Demineralized Water System.
D. Condensate Polishing System.
Answer: D Explanation/Justification: Later A. Incorrect- Plausible because this is a radioactive system, however this system is located in the Auxiliary Building. Leakage from this system would go to a floor drain or sump and be transported to the Mise Waste Drain Tank, not the storm sewer. There is no connection between the Mise Waste Drain Tank and the Storm Sewer.
B. Incorrect- Plausible because this is a radioactive system, however this system is located in the Auxiliary Building. Leakage from this system would go to a floor drain or sump and be transported to the Mise Waste Drain Tank, not the storm sewer. There is no connection between the Mise Waste Drain Tank and the Storm Sewer.
- . Incorrect- Plausible because leak from this system can reach the storm sewer, but the system is not radioactive and would not cause an alarm on the Storm Sewer Radiation Monitor.
D. Correct- With SG Tube Leak, activity levels in the condensate polishers will rise. Leakage from this system could reach the storm sewer via the Turbine Building Drains.
Sys# System Category KA Statement 000059 Accidental AK2. Knowledge of the interrelations between the Accidental Liquid Radioactive-liquid monitors Liquid Radwaste Release and the following:
Radwaste Release KIA# AK2.01 KIA Importance 2.7 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02531 R19 Attachment 7 page 3 of 4 Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: Low - Fundamental 10 CFR Part 55 Content: (CFR 41.7 /45.7)
Objective:
33 DB-OP-02531 Revision 19 ATTACHMENT 7: BACKGROUND INFORMATION Page 3 of4 Feedwater Drains will transfer to the Condenser automatically during the shutdown When the
-heaters go on high level control. CD550B, Hotwell High Level Return Valve, is norm~lly isolated and will contain the *contaminated condensate in the Hotwell until draining or level
- reduction of the HotweU is required.
Assuming a.50-gpm Steam Generator Tube leak with a 10°F per hour cooldown rate for the entire cooldown, approximately 115,000 gallons of Reactor Coolant System water will accumulate in the secondary side of the plant during a cooldown from 582°F to 200°F (approximately 3000* gallons per hour). The Borated Water Storage Tank and Clean Waste ReceiverTa,nks will be the primary sources for Reactor Coolant System makeup. Faster Cooldown Rates will reduce the total leakage to the secondary.
- The projected volume can be accommodated by one or more of the following methods:
- Allowing the Condenser Hotwelllevel to fill above normal levels.
o Draining or rejecting the Hotwell to the Condensate Storage Tanks as necessary to prevent overfill of the Condenser Hotwell.
o Draining the Condenser Hotwell to the Condensate Polishing Demin Holdup Tanks via the West Condenser Pit Sump
... **** , . ,_,, ~~~i*~f~-,.;~!,W**~J~~~~~~~~~i,iiltii.J~. .:!:;.~ ' ....
o Draining the Condensate Polishing Demin Holdup Tanks to the Miscellaneous Drain **'
~ i:~~:o create ad~ffi:n:,:.::::::::c;:::~;:,::::;.,~
U~~~l~Jh~'ri~t:~::;e Tanks will require that their levels be reduced by draining to the Turbine Building sumps or Settling Basin to allow this inflow from the Hotwell. The initial CST drain should include all of the anticipated volume for this method so that draining contaminated CST inventory later in the event will hot be necessary.
The vacuum system vent fil~er is placed in service when Xe-133 exceeds 6.58-3 JJCi/cc. This will minimize the amount of radioactivity released to the environment through the Steam Jet Air Ejectors vent line. Xe-133 is used because Chemistry cannot sample the Steam Jet Air EjeCtors for 1-131 due to the presence of moisture. The ratio of 1-131 to Xe-133 is known and hence, Xe-133 can be used to detelinlne the 1-131 concentration.
Davis Besse 1 LOT13 NRC Written Exam Rev. 1
- 22. Waste Gas Decay Tank 1 is being discharged to the station vent lAW DB-OP-03012, Radioactive Gaseous Batch Release. WG1821, Waste Gas To Station Vent Flow Control is being utilized for this batch release.
The following valid alarms and indications are received:
- RE1822A Waste Gas System Radiation Monitor alarms WARN & HIGH
- RE1822A Waste Gas System Radiation Monitor indicates offscale high No automatic actions have occurred.
Based on these conditions, which of the following valves FAILED to automatically CLOSE?
- 1. WG1819, Waste Gas To Station Vent Isolation
- 2. WG1820, Waste Gas To Station Vent Isolation
- 3. WGI821, Waste Gas To Station Vent Flow Control
- 4. WG1836, Waste Gas Decay Tank 1 To Station Vent Control A. 1 & 2 only B. 1 & 4 only C. 2 & 3 only
- 3 &4 only Answer: A Explanation/Justification:
A. Correct- RE1822A trip should have caused both the Waste Gas to Station Vent Isolations to Close.
B. Incorrect- Plausible because RE1822A trip should have caused WG1819 to close and since WG1836 a control valve is in the release flowpath, it is plausible that the controller should have closed as it well.
C. Incorrect- Plausible because RE1822A trip should have caused WG1820 to close and since WG1821 a control valve is in the release flowpath, it is plausible that the controller should have closed it as well.
D. Incorrect- Plausible because if the RE1822A trip would use a controller to provide isolation, it is logical that WG1821 and WG1836 would close to provide isolation.
Sys# System Category KA Statement 000060 Accidental AA2. Ability to operate and I or monitor the following as they apply to Valve lineup for release of radioactive gases Gaseous- the Accidental Gaseous Radwaste:
Waste Release KIA# AA2.06 KIA Importance 3.6* Exam Level RO References provided to Candidate None Technical
References:
OS-030 Sheet 1 (B-16) and Sheet 2 CL-1 Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR 41.7 I 45.5 I 45.6)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev. 1 DB-OP-02012, STM GEN/SFRCS ALARM PANEL 12 ANNUNCIATOR procedure directs Radiation Protection to be notified to take local surveys of the Main Steam Line area when annunciator 12-1-A, MN STM LINE 1 RAD HI comes into alarm.
Which of the following is the reason for this direction?
A To evaluate for initiating conditions into RA-EP-02861, Radiological Incidents B. To obtain data to support leak rate calculation for DB-OP-02522, Small RCS Leaks C. To project off site doses from the Station Vent in accordance with RA-EP-02240, Offsite Dose Assessment.
D. To verify affected SG diagnosis in accordance with DB-OP-02531, Steam Generator Tube Leak.
Answer: D Explanation/Justification:
A. Incorrect- Plausible because high radiation levels would be an initiating condition for the radiological incidents off normal procedure but this alarm is to support indications of a steam generator tube leak B. Incorrect- Plausible because there is a leak rate calculation that uses RE indications in the calculation but it uses steam jet air ejector discharge RE1003A & B.
C. Incorrect- Plausible Steam Generator Tube Leaks will cause a release of radioactive material, but checking radiation levels in the Main Steam Line area will not allow determination of dose from the station vent.
. Correct- DB-OP-02012 directs checking symptoms in accordance with DB-OP-02531 along with alarm verification to access entry conditions into the steam generator tube leak abnormal procedure.
Sys # System Category KA Statement 000061 ARM System AK3. Knowledge of the reasons for the following responses as they Guidance contained in alarm response for ARM Alarms apply to the Area Radiation Monitoring (ARM) System Alarms: system KIA# AK3.02 KIA Importance 3.4 Exam Level RO References provided to Candidate Technical
References:
DB-OP-02012 R10 Page 4 Step 3.4 None Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: Low- Fundamental 10 CFR Part 55 Content: (CFR 41.5,41.10 I 45.6/45.13)
Objective:
Davis Besse 1 LOT13 NRC Written Exam Rev. 1
- 24.
- Plant is in mode 6
- The Refueling Canal is greater than 23 feet
- Fuel Handling is in progress in Containment Identify the ONE situation below that represents a condition that would require the handling of irradiated fuel in CTMT to be stopped:
A. Equipment Hatch is removed. A Maintenance team is assigned to install the hatch but is not present.
B. Maintenance removes SP 17B6, SG1 Main Steam Safety Valve in the Main Steam Line room AND SG 1 Secondary Manway is open for inspection.
C. An operator is signed into the Containment Closure Control log for MU66D, Reactor Coolant Pump 1-2 Seal Injection Flow Isolation and is draining its piping for a Local Leakrate Test D. BOTH air lock doors of the CTMT personnel hatch are opened. An Operator is assigned to be responsible for closing ONE door.
Answer: B Explanation/Justification:
A. Incorrect- Plausible if the candidate does not know the equipment hatch can be open during fuel handling since it was previously required closed. They may also assume the team must be staged which is not correct.
Correct- This creates a path from Containment to atmosphere.
- . Incorrect- Plausible since this will create a path between CTMT and atmosphere but is administratively controlled by the CTMT Closure Control log D. Incorrect- Plausible if the Candidate knows this is outside of the EVS boundary but does not know this is allowed by procedure Sys# System Category KA Statement 000069 Loss of AA 1. Ability to operate and I or monitor the following as they apply to Fluid systems penetrating containment Containment the Loss of Containment Integrity:
Integrity KIA# AA1.03 KIA Importance 2.8 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-06904 R42 Note 11.1, OS-008 SH 1 R35 Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.7 I 45.5 I 45.6)
Objective:
66 DB-OP-06904 Revision 42 11.0 NOTE 11.1
- Containment closure is the action to secure the containment and its associated structures, systems and components as a functional barrier to*,:
fission product release under existing plant conditions. Containment closure control provides the methodology to quickly secure the containment. The intent of the following is to provide the control room, operator with the ability to isolate the containment from the control ~
room and to provide guidance to allow hatches and other penetrations to be functional as soon as possible using Attachment 8 to assign designated personnel.
- Penetrations providing direct access from the containment atmosphere '
to the atmosphere outside containment may be open during operations involving movement of irradiated fuel within containment provided the *,
administrative controls provided in this section are maintained.
11.1 General Requirements 11.1.1 Deviations from total containment closure as defined in the following steps should be documented with the following three methods.
NOTE ll.l.l.a Secondary Systems openings (OTSG manways) are not shown on M-023.
- a. Containment airlocks, equipment hatch, penetrations, and secondary system openings status should be maintained on a controlled copy of P&ID M-023 encased in plastic.
- b. Containment air1ocks, equipment hatch, penetrations, and secondary system openings status should be maintained on the SPDS computer display.
- c. Personnel assignments and closure actions will be documented on Attachment 8, CTMT Closure.
11.1.2 If the Equipment Hatch is off, then the SFP Negative Pressure Area is extended to inside Containment. The cumulative void area of the SFP Negative Pressure Area will be the sum of the areas tracked by this procedure and the areas of the penetrations tracked in DB-OP-000 18, Inoperable Equipment Tracking Log.
LCO 3.7.13 for SFP EVS allows the SFP Negative Pressure boundary to be opened under administrative control. The Containment Closure Control provisions of this procedure satisfy the administrative control requirements for LCO 3. 7.13, as long as a dedicated individual is stationed at the opening during the handling of irradiated fuel in the SFP building.
Davis Besse 1LOT13 NRC Written Exam Rev.1 25 . The following plant conditions exist:
The plant is at 90% power ICS is in a normal lineup The following alarms occurs:
- 8-4-A, MFPT 1 TRIP alarms
- 10-1-A, MFP 1 DISCH HI PRESS TRIP alarms
- 13-4-C, DEAR STRG TK LVL
- 14-3-D, ICS MFP LOSS OR LO DEAR RUNBACK alarms Main Generator load is lowering and stabilizes at approximately 700 MWe with #1 Deaerator level at 9 feet.
Based on these plant conditions, what procedures and associate actions are required?
A. Trip the reactor and enter DB-OP-02000 in accordance with DB-OP-02014, MSR/ICS Alarm Panel14 Annunciators B. Stabilize the plant at the current power level in accordance with DB-OP-06401, ICS Procedure, section for plant stabilization following a runback Place SG/RX Demand Station in HAND and perform runback to 55% power in accordance with DB-OP-0201 0, Feedwater Alarm Panel 10 Annunciators D. Place Feedwater Loop Demands and the Rod Control Panel in MANUAL and stabilize Reactor power and Tave in accordance with DB-OP-02526, Primary to Secondary Heat Transfer Upset Answer: C Explanation/Justification:
A. Incorrect- Plausible because DB-OP-02014 directs tripping reactor if deaerator level approaches low off scale B. Incorrect- Plausible because DB-OP-06401 provides direction for stabilization following a run back, however reactor power was not reduced below runback setpoint for loss of a MFP.
C. Correct- DB-OP-0201 0 for MFPT trip provides this direction for a MFPT Trip D. Incorrect- Plausible because DB-OP-02526, Primary to Secondary Heat Transfer Upset provides these directions for plant stabilization upon a plant upset Sys# System Category KA Statement BW/A01 Plant AA2. Ability to determine and interpret the following as they apply to Facility conditions and selection of appropriate Run back the (Plant Runback) procedures during abnormal and emergency operations.
KIA# AA2.1 KIA Importance 3.0 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02010 R17 Page 4 Question Source: BANK 75948 Level Of Difficulty: (1-5) 4 Question Cognitive Level: High - Analysis 10 CFR Part 55 Content: (CFR: 43.5 I 45.13)
.jective:
Davis Besse 1LOT13 NRC Written Exam Rev. 1 26 . The Plant is at 50% power with a shutdown in progress. Entry into Mode 5 to perform maintenance is planned The following annunciator alarms come in:
- (14-2-D) ICS/NNI 118VAC PWR TRBL
- (14-4-E) ICS INPUT MISMATCH
- (14-4-F) ICS INPUT TRANSFER Other indications:
- Loss of blue light on SASS instrument's selector switches.
- SCR Bank, RC PRESSURE CONTROL, Hand/Auto Station Lights Both ON Assuming the condition can not restored to normal which of the following actions must be taken?
A. Operation of DHR Train 1 will be required vice the normal DHR Train 2 for cooldown.
B. Control Atmospheric Vent Valves, ICS11A and ICS11 Bin manual for cooldown.
C. Transfer the EHC Control Panel to manual for turbine control.
- - *==C-Io_s=e=M=u_a_s_,=L=e-td_o_w_n=F=Io_w_c=o=nt=ro=l=l=n-le_t_ls=o=la=t=io=n=t=o=M=U-6=t=o-is=o-la=t=e=L=e=td_o_w_n_._===-==
Answer: A Explanation/Justification:
A. Correct- must diagnose loss of NNI X AC and identify appropriate response. Although the Decay Heat Cooler SFAS Valves, DH13A and DH14A, solenoids are DC powered their controls along with various DH Train 2 indications and alarms will be out of service since NNI X AC lost.
B. Incorrect- Plausible because this is the response for loss of ICS power.
C. Incorrect- Plausible because this is a response for loss of NNI-X DC power.
D. Incorrect- Plausible because this is a response for loss of NNI-Y AC power.
Sys # System Category KA Statement BW/A02 Loss of NNI-X AK2. Knowledge of the interrelations between the (Loss of NNI-X) Facility's heat removal systems, including primary and the following: coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.
KIA# AK2.2 KIA Importance 3.8 Exam Level RO References provided to Candidate Technical
References:
DB-OP-02532 R 10 Step 4.1.13 None Question Source: New Level Of Difficulty: (1-5) 4 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 /45.7)
Objective:
Davis Besse 1 LOT13 NRC Written Exam Rev.1 Following a Reactor Trip, a severe overcooling has caused a loss of Adequate Subcooling Margin.
Once the Reactor is confirmed as shutdown using the Immediate Operator Actions, based on the priorities provided by DB-OP-02000, RPS, SFAS, SFRCS Trip or SG Tube Rupture, which the following is implemented FIRST to mitigate this event A. Section 5, Loss of Subcooling Margin B. Section 7, Overcooling C. Specific Rule 2, Loss of Subcooling Margin D. Specific Rule 4, Steam Generator Control Answer: C Explanation/Justification:
A. Incorrect- As provided in the Bases and Deviation Document for DB-OP-02000, The hierarchy in DB-OP-02000 between the various sections is as follows: 1. Immediate Actions 2. Specific Rules 3. Procedure Sections 4. Attachments. Rules are implemented prior to sections.
B. Incorrect- As provided in the Bases and Deviation Document for DB-OP-02000, The hierarchy in DB-OP-02000 between the various sections is as follows: 1. Immediate Actions 2. Specific Rules 3. Procedure Sections 4. Attachments. Rules are implemented prior to sections.
C. Correct- As provided in the Bases and Deviation Document for DB-OP-02000, The hierarchy in DB-OP-02000 between the various sections is as follows: 1. Immediate Actions 2. Specific Rules 3. Procedure Sections 4. Attachments. Rules are implemented prior to sections. Rules are implemented in numerical order
. Incorrect- As provided in the Bases and Deviation Document for DB-OP-02000, The hierarchy in DB-OP-02000 between the various sections is
- as follows: 1. Immediate Actions 2. Specific Rules 3. Procedure Sections 4. Attachments. Rules are implemented prior to sections. Rules are implemented in numerical order Sys# System Category KA Statement BW/E13 EOP Rules EA1. Ability to operate and I or monitor the following as they apply to Desired operating results during abnormal and the (EOP Rules) emergency situations.
KIA# EA1.3 KIA Importance 3.4 Exam Level RO References provided to Candidate None Technical
References:
Bases and Deviation Document for DB-OP-02000 R19, Prioritization of DB-OP-02000 Sections.
Question Source: New Level Of Difficulty: (1-5) 2 Question Cognitive Level: High - Analysis 10 CFR Part 55 Content: (CFR: 41.7 /45.5/45.6)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev. 1 The plant is operating at 70% power with all systems in normal alignment for this power level.
All four Reactor Coolant Pumps are in service.
Motor current for the 1-1 RCP is noted to be 290 amps.
( 1) Which of the following describes the current status of the RCP motor current reading?
(2) What action is required, if any, for this condition?
A. (1)This motor current reading is lower than normal.
(2) RCP 1-1 shutdown is required.
B. (1)This motor current reading is lower than normal.
(2) RCP 1-1 shutdown is NOT required.
C. (1)This motor current reading is higher than normal.
(2) RCP 1-1 shutdown is NOT required.
D. (1)This motor current reading is higher than normal.
(2) RCP 1-1 shutdown is required.
Answer: C planation/Justification:
- Incorrect- At normal operating RCS temperatures and pressures, normal RCP Motor Current is approximately 260 amps. The Operating Limits requiring shutdown are less than 200 amps or greater than 370 amps per DB-OP-02515, Reactor Coolant Pump and Motor Abnormal Operations.
B. Incorrect- At normal operating RCS temperatures and pressures, normal RCP Motor Current is approximately 260 amps. The Operating Limits requiring shutdown are less than 200 amps or greater than 370 amps per DB-OP-02515, Reactor Coolant Pump and Motor Abnormal Operations C. Correct- At normal operating RCS temperatures and pressures, normal RCP Motor Current is approximately 260 amps. The Operating Limits requiring shutdown are less than 200 amps or greater than 370 amps per DB-OP-02515, Reactor Coolant Pump and Motor Abnormal Operations D. Incorrect- At normal operating RCS temperatures and pressures, normal RCP Motor Current is approximately 260 amps. The Operating Limits requiring shutdown are less than 200 amps or greater than 370 amps per DB-OP-02515, Reactor Coolant Pump and Motor Abnormal Operations Sys# System Category KA Statement 003 Reactor A3. Ability to monitor automatic operation of the RCPS, including: Motor current Coolant Pump System (RCPS)
KIA# A3.02 KIA Importance 2.6 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02515 R11, Reactor Coolant Pump and Motor Abnormal Operations step 4.6.1 Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 I 45.5)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1
- 29. The following plant conditions exist:
- Mode 1 at 15% power The following event occurs:
- RCP 1-1 is shutdown by the crew due to excessive vibrations.
- No other operator actions are taken.
Which one of the following represents the condition of the plant, once stabilized?
A. T ave will be selected to Loop 1.
B. Loop 1 FW flow will be 2.4 times greater than Loop 2 FW flow C. Loop 2 FW flow will be 2.4 times greater than Loop 1 FW flow.
D. Tave will be selected to Loop 2.
Answer: D Explanation/Justification:
A. Incorrect -In accordance with DB-OP-02515 R11, RCP and Motor Abnormal Attachment 1 for Stopping a RCP Step 5. SASS will align Tave to the loop with 2 RCPS in service.
Incorrect- Plausible because the normal response at 72% power when an RCP would be shutdown is for FW Flow to Loop with the highest RCS
- flow to be 2.4 time greater than the remaining loop. A trip from 15% with SG on Low Level limits negates flow control. The SG Will be on Level Control.
C. Incorrect- Plausible because the normal response at 72% power when an RCP would be shutdown is for FW Flow to Loop 2 to be 2.4 time greater. A trip from 15% with SG on Low Level limits negates flow control. The SG Will be on Level Control D. Correct -In accordance with DB-OP-02515 R11, RCP and Motor Abnormal Attachment 1 for Stopping a RCP Step 5. SASS will align Tave to the loop with 2 RCPS in service.
Sys# System Category KA Statement 003 Reactor K3. Knowledge of the effect that a loss or malfunction of the RCPS RCS Coolant will have on the following:
Pump System (RCPS)
KIA# K3.01 KIA Importance 3.7 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02515 R11, RCP and Motor Abnormal Attachment 1 for Stopping a RCP Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 I 45.6)
Objective:
Davis Besse 1 LOT13 NRC Written Exam Rev.1 30 . The plant is operating at 100% power with all systems in normal alignment for this power level.
Makeup Pump 2 is in service.
Which of the following conditions would cause MU3971 Makeup Pump 2 Suction Valve to transfer from the Makeup Tank to the BWST assuming lock is NOT depressed for the valve?
A. SFAS Level2 B. Makeup Tank Level less than 10 inches C. Loss of NNI X AC Power D. Loss of D2P and DBP Answer: B Explanation/Justification: KA Statement is for design features and/or interlocks on the letdown system for the letdown tank bypass valve. The closed valve for Davis Besse would be the MU Pump Suction Valves. These valves can be aligned to take a suction on the Makeup Tank or on the BWST. In the BWST position, the Makeup Tank is effectively bypassed.
A. Incorrect- No automatic feature exists, however plausible because this action would protect BWST inventory for use by ECCS Systems.
B. Correct- Low Makeup Tank Level of 10 inches will cause an auto transfer from the MU Tank to the BWST..
C. Incorrect- Plausible because the MU3971 Auto Transfer from the BWST to the MU Tank is lost when NNI X AC power is lost.
D. Incorrect- Plausible because the MU3971 Auto Transfer from the BWST to the MU Tank is lost when D2P and DBP power is lost
- System Category KA Statement Chemical and K4. Knowledge of CVCS design feature(s) and/or interlock(s) which Control interlocks on letdown system (letdown Volume provide for the following: tank bypass valve)
Control System KIA# K4.14 KIA Importance 2.8* Exam Level RO References provided to Candidate Technical
References:
DB-OP-02002 R08 page 16 note 3.5 None Question Source: New Level Of Difficulty: (1-5) 2 Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.7)
Objective:
16 DB-OP-02002 Revision 08
'~-.~.**** *** . '.W.;
NOTE 3.5 If the Makeup Tank level decreases to 10 inches, MU 3971, MU PUMP 2 SUCTION THREE-WAY, and MU 6405, MU PUMP 1 SUCTION THREE-WAY, will automatically position to
.. provide Makeup Pump suction from the BWST. MU 3971 and MU 6405 must be positioned to the BWST within 45 seconds from the time Makeup Tank level reaches 10 inches or the Makeup Pumps will automatically trip.
..... e. . . _.,,.f_.,.,""' * * **' * -<>'> . "" 6< *. * -~' N'Wff'-~ i , , . '
3.5 IF the Makeup Tank level decreases to 10 inches, THEN verify the following: ,
3.5.1 MU 3971, MU PUMP 2 SUCTION THREE-WAY, switches to the BWST.
3.5.2 MU 6405, MU PUMP 1 SUCTION THREE-WAY, switches to the BWST.
NOTE 3.6 and 3.7 Failure of either Make-Up Tank level transmitter low will cause repositioning of both MU Pump suction valves to the BWST. This can be defeated by pulling fuse 3R FUl in Panel RC4802. LT-MU16-l or 2 failure low causes a low MUT level trip signal to MU Pump 2 or 1 respectively if not aligned to the BWST, after a 45 second time delay.
The low level pump trips can be defeated by the installation of jumpers as detailed below.
3.6 IF LT-MU16-1 has failed low, THEN perform the following:
3.6.1 Install jumper in RC4602, TB23R/24R, terminals 1 and 2.
3.6.2 Pull fuse 3R FUl in Panel RC4802.
3.6.3 Place HIS 3971 and HIS 6405 to the MUT position.
3.6.4 Monitor level.
3.7 IF LT-MU16-2 has failed low, THEN perform the following:
3.7.1 Install jumper in RC2825, TB22L, terminals 1 and 2.
3.7.2 Pull fuse 3RFU1 in Panel RC4802.
3.7.3 Place HIS 3971 and HIS 6405 to the MUT position.
3.7.4 Monitor level.
Davis Besse 1LOT13 NRC Written Exam Rev. 1 31 . The following plant conditions exist:
- The plant is shutdown for maintenance in MODE 5.
- The RCS is vented to Containment Atmosphere.
- Shutdown cooling is provided by DHR Train 2.
- The plant has been shutdown for 10 days and RCS temperature is 100 °F.
- DH 14A, Decay Heat Cooler 2 Outlet Valve is full open
- DH 13A, Decay Heat Cooler,2 Bypass Valve is full closed The following event occurs:
- DH14A, Decay Heat Cooler 2 Outlet Valve fully closes.
Based on these conditions, what is the time to RCS boil?
(Reference Attached)
A. 24 minutes B. 31 minutes D.
35 minutes 155 minutes Answer: C Explanation/Justification:
A. Incorrect- Plausible if the Candidate misreads the curve. 24 minutes would be the time to boil if the initial temperature was 140*F.
B. Incorrect- Plausible if the Candidate misreads the curve. 31 minutes would be the time to boil if the initial temperature was 1oo*F but the candidate used the Low RCS level of 6 inches above Hot Leg Center Line. 30 inches is a low RCS level, but not for this curve.
C. Correct- From DB-PF-06703 R20, Page 57 CC6.3c, the correct time to boil is 35 minutes.
D. Incorrect- Plausible if the Candidate uses CC6.3d, time to boil to top of core which will be provided.
Sys# System Category KA Statement 005 Residual K6. Knowledge of the effect of a loss or malfunction on the following RHR heat exchanger Heat will have on the RHRS:
Removal System (RHRS)
KIA# K6.03 KIA Importance 2.5 Exam Level RO References provided to Candidate DB-PF-06703 Rev 20 Technical
References:
DB-PF-06703 Rev 20 CC6.3.c and CC6.3.d CC6.3.c and CC6.3.d Question Source: BANK29427 Level Of Difficulty: (1-5) 3 Question Cognitive Level: High - Application 10 CFR Part 55 Content: (CFR:*41.7/45.7)
Objective:
57 DB-'PF-06703 Revision 2:0 CC6.3c TIME TO HEAT RVWATER TO 212 °F.ASA FUNCTI.ON OF DECAY HEAT, INIT RV TEMP . & IN'IT RV LEVEL 0 ~ 0 ~ 0 ~ 0 ~ 0
~ ~ ~ M M N N ~* ~ ~
(NU.AJ) ::lo Z~Z O.l ~31VM 1V3H 01 a.:03~ 3WI.l
58 DB-PF-067(}3 Revision20 C6.3d TIME TO BOIL TO TOP OF CORE AS A FUNCTION OF I NIT LEVEL & DECAY HEAT 0
~-----+------~+-----+---~~~~---+-- ~
0
..... 0 co
(:!o Z~Z Sl Ci31VM 3CIO~ Ci3l:IV)
(NfV\1) 3CIO~ :.:10 dOl 01 1108 01 03CIJn0,3Ci 3VIIU
Davis Besse 1LOT13 NRC Written Exam Rev. 1 Which ONE of the following is the reason a break in the 14 inch line between the reactor vessel and CF 30, CFT 1-2 TO REACTOR CHECK VALVE, will not result in exceeding the peak allowable cladding temperature of 2200 oF?
A. Leak is at an elevation that will not uncover the core.
B. The Core Flood line flow restrictor at the Reactor Vessel limits the size of the leak from the Reactor Coolant System.
C. 14 inches is less than the size required to cause a large break LOCA.
D. One train of Core Flood meets all postulated loss of coolant accidents.
Answer: B Explanation/Justification:
A. Incorrect- Plausible if the candidate assumes the injection lines enter the vessel above the top of the core so the core won't uncover B. Correct- CFTs are not redundant therefore the flow restrictor limits leak size to allow one CFT to limit peak clad temperature C. Incorrect- Plausible if Candidate does not know what break size is classified as a large break LOCA D. Incorrect- Plausible because most safety systems have 2 fully redundant trains, only one of which is required to meet ECCS Criteria. Both Core Flood Tanks are required to meet ECCS Criteria.
Sys# System Category KA Statement 006 Emergency K6. Knowledge of the effect of a loss or malfunction on the following Core flood tanks (accumulators)
Core Cooling will have on the ECCS:
System (ECCS)
-A# K6.02 KIA Importance 3.4 Exam Level RO References provided to Candidate Technical
References:
SD-040 R4 page 1-4 step 1.2.3.2 None Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: Low- Fundamental 10 CFR Part 55 Content: (CFR: 41.7 /45.7)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev. 1
- The PORV is leaking
- The Quench Tank Circulating Pump is isolated for maintenance
- Reactor Coolant System pressure is 2155 psig The Quench Tank Relief Valve failing open will cause level to rise in the:
A. Waste Gas Surge Tank B. Reactor Coolant Drain Tank C. Containment Normal Sump D. Clean Waste Receiver Tank Answer: C Explanation/Justification:
A. Incorrect- Plausible since the Quench Tank can be lined up to vent to the Waste Gas Header B. Incorrect- Plausible since a majority of the RCS relief valves relieve to the RCDT C. Correct- because the Quench Tank relieves to the Normal Sump D. Incorrect- Plausible since RCS discharge would be considered clean waste Sys# System Category KA Statement 007 Pressurizer K3. Knowledge of the effect that a loss or malfunction of the PRTS Containment Relief will have on the following:
Tank/Quench Tank System (PRTS)
KIA# K3.01 KIA Importance 3.3 Exam Level RO References provided to Candidate None Technical
References:
Ops Schematic OS-001A Sheet 3 Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: Low- Fundamental 10 CFR Part 55 Content: (CFR: 41.7 I 45.6)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1 34 . Reactor Power is 75% and stable.
- Component Cooling Water (CCW) Pump 1 is running
- Component Cooling Water (CCW) Pump 2 is in standby
- Component Cooling Water (CCW) Pump 3 is aligned to side 1 as spare The following occurs:
- CCW Pump 1 trips
- CCW Pump 2 does not start The Reactor Operator attempts to start CCW Pump 1 and 2 from the control room and neither pump starts Based on these conditions, identify the ONE statement below that identifies the required action(s) to be implemented A. Reduce Reactor Power to 72% in preparation for shutdown of an RCP.
B. Trip the Reactor and trip all RCPs.
C. Commence a Rapid Shutdown and monitor the Reactor Coolant Pumps.
- Monitor the Reactor Coolant Pumps and place the spare Component Cooling Water Pump 3 in service.
Answer: B Explanation/Justification:
A. Incorrect- Plausible since this would be the actions for loss of CCW to one RCP when reaching the required RCP trip parameters B. Correct- CCW abnormal procedure directs tripping the Reactor and all 4 RCPs in the event of the running and standby pumps being unable to be started C. Incorrect- Plausible because reducing power would reduce heat loading and the candidate may assume it is not required to trip the RCP or the Reactor until required trip parameters are reached.
D. Incorrect- Plausible if the candidate assumes the spare CCW pump may be able to be placed in service prior to reaching required RCP trip parameters Sys# System Category KA Statement 008 Component A2. Ability to (a) predict the impacts of the following malfunctions or Loss of CCW pump Cooling operations on the CCWS, and (b) based on those predictions, use Water procedures to correct, control, or mitigate the consequences of those System malfunctions or operations:
(CCWS)
KIA# A2.01 KIA Importance 3.3 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02523 R09 step 4.3.1 page 28 Question Source: New Level Of Difficulty: (1-5) 2.5 Question Cognitive Level: High -Analysis 10 CFR Part 55 Content: (CFR: 41.5/ 43.5/ 45.3 /45.13)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1 35 . The Plant is in Mode 1 In accordance with Technical Specifications, which one of the following conditions requires action to be completed in less than 30 minutes to remain in compliance with Technical Specifications requirements?
A. Pressurizer Level is greater than 228 inches.
B. One Pressurizer Code Safety Valve setpoint is set greater than 2525 psig.
C. No power is available to the Pressurizer Power Operated Relief Valve.
D. The Block Valve for the Pressurizer Power Operated Relief Valve is closed.
Answer: B Explanation/Justification:
A. Incorrect- Plausible because when this condition is encountered in the simulator, the candidates take prompt action to restore Pressurizer Level to within limits.
B. Correct- Per Technical Specifications Pressurizer Safety Valves to be Operable requires a setting of less than or equal to 2525 psig. A setpoint greater than 2525 renders the valve inoperable. Action is required within 15 minutes per TS 3.4.1 0 Condition A.
C. Incorrect- Plausible since this condition renders the PORV inoperable and required action within one hour to close the PORV Block valve per TS 3.4.11 Condition B.
D. Incorrect- Plausible since this condition would render the PORV inoperable and requires action per TS 3.4.11 Condition B.
System Category KA Statement Pressurizer Generic Knowledge of less than or equal to one hour Pressure Technical Specification action statements for Control systems.
System (PZR PCS)
KIA# 2.2.39 KIA Importance 3.9 Exam Level RO References provided to Candidate None Technical
References:
TS 3.4.1 0 Condition A (Amendment 279)
Question Source: New Level Of Difficulty: (1-5) 3.5 - 4 Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.7 I 41.1 0 I 43.2 I 45.13)
Objective:
Davis Besse 1 LOT13 NRC Written Exam Rev. 1 36 . Essential Bank 2 Pressurizer heater bank control switch is in the ON position. If RCS pressure is stable at the normal operating point, and Pressurizer level decreases to 37", which ONE of the following explains the status of the Essential Bank 2 heaters?
The heater bank is A. energized because manual control overrides the Pressurizer low-low level heater cutoff.
B. de-energized because the Pressurizer low-low level heater cutoff overrides manual control.
C. energized because Pressurizer level is above the low-low level heater cutoff setpoint D. de-energized because normal RCS pressure is above the heater bank cycle setpoint.
Answer: A Explanation/Justification:
A. Correct- In automatic, the design of the Pressurizer Heaters removes power on LOW LOW pressurizer Level (40 inches). Operating the Pressurizer in manual (ON) overrides this design feature.
B. Incorrect- Plausible if the candidate does not understand that the "ON" position for the heaters overrides the Low level cutoff.
C. Incorrect - Plausible if the candidate does not know the setpoint for low low pressurizer level and thinks it is less than 40 inches. A pressurizer level of 37 inches is still above the top of all Pressurizer heaters D. Incorrect- Plausible if the candidate does not understand the interlock but knows this bank of heaters is off at normal RCS Pressure.
System Category KA Statement Pressurizer K4. Knowledge of PZR PCS design feature(s) and/or interlock(s) Prevention of uncovering PZR heaters Pressure which provide for the following:
Control System (PZR PCS)
KIA# K4.02 KIA Importance 3.0 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-06003 R29 Attachment 7 PZR Heater Control Panel Placard DB-OP-02513 R11 4.6.4 RNO Question Source: BANK37164 Level Of Difficulty: (1-5) 3 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev. 1 The following plant conditions exist:
The plant is operating at 100% power.
RPS Channel 1 is in Manual Bypass.
The following event occurs:
- RPS Channels 2 and 4 Trip
A. No CRD Trip Breakers will open.
B. Only the "A" and "C" breakers will open.
C. Only the "B" and "D" breakers will open.
D. All CRD Trip Breakers will open.
~
~xplanation/Justification: Note: Actuation of an RPS Channel trips the respective CRD Breaker by removing DC control power from the breaker.
This DC Control Power is internally generated in the Reactor Protective System, not supplied from an external source ..
A. Incorrect- Plausible if the candidate believes the logic of RPS is the same as the Steam Feed Rupture Control System where Channels 1 and 3 are actuation channel 1 and Channels 2 and 4 are actuation channel 2. If only a single actuation channel trips, a full SFRCS actuation does not occur.
B. Incorrect- Plausible if the Candidate does not understand the relationship between RPS Channels and CRD Breakers. RPS Channels 1, 2, 3, 4, supply CRD Breaker B, A, D, C respectively.
- c. Incorrect- Plausible if the Candidate does understand the relationship between RPS Channels and CRD Breakers. RPS Channels 1, 2, 3, 4, supply CRD Breaker B, A, D, C respectively but does not understand the logic of RPS as noted in distractor 1 above.
D. Correct - lAW DB-OP-06403, Attachment 4, Page 59, Relays KB and KD remain energized and their correstonding contacts in each RPS cabinet remain closed, howver the KA and KC relays de-energize. The corresponding KA and KC contacts open in each cabinet interrupting DC Control Power to the associated CRD Breaker and causing the breakers to trip.
Sys # System Category KA Statement 012 Reactor A2. Ability to (a) predict the impacts of the following malfunctions or Loss of de control power Protection operations on the RPS; and (b) based on those predictions, use System procedures to correct, control, or mitigate the consequences of those (RPS) malfunctions or operations:
KIA# A2.07 KIA Importance 3.2* Exam Level RO References provided to Candidate Technical
References:
DB-OP-06403 R19, Attachment 4, Page 59 None Question Source: New Level Of Difficulty: (1-5) 2.5 Question Cognitive Level: High -Analysis 10 CFR Part 55 Content: (CFR: 41.5/ 43.5/ 45.3/45.5)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1 38 . The plant was operating at 75% power with all systems in normal alignment for this power level.
The following plant conditions NOW exist:
- SG 1 pressure is 880 psig.
- SG 2 pressure is 150 psig.
- Reactor Coolant System pressure is 1700 psig and steady.
- Reactor Coolant System temperature is 530 oF and steady.
- Containment pressure is 19 psia and lowering.
- All systems function as designed.
With NO operator action, what will be the control level setpoint for SG 1?
A. 49 inches B. 55inches C. 124 inches D. 130 inches Answer: D planation/Justification:
- Incorrect- Plausible if the Candidate does not diagnose an SFAS level 2 trip or SG 2 isolation on low pressure since this is normal level for a SFRCS actuation without SG low pressure trip B. Incorrect- Plausible if the Candidate knows level is controlled at 55" by the opposite side pump on SG low pressure SFRCS trip but doesn't diagnoses the SFAS level 2 trip or know setpoint is raised to high on a SA2 C. Incorrect- Plausible since this is the normal level for SFRCS actuation on an SFAS 2 with no SG isolation. Diagnoses SA2 but not SFRCS low pressure.
D. Correct - SFAS level 2 on CTMT pressure (18. 7psia) will raise the setpoint to high and AFP 2 will control level at 130" with AFP 1 setpoint at 124" due to SG 2 SFRCS low pressure trip (630psig)
Sys# System Category KA Statement 013 Engineered K1. Knowledge of the physical connections and/or cause effect AFWSystem Safety relationships between the ESFAS and the following systems:
Features Actuation System (ESFAS)
KIA# K1.07 KIA Importance 4.1 Exam Level RO References provided to Candidate None Technical
References:
OS-17A SH1 R26 CD-1, DBBP-TRAN-0034 R06 page 8&9 Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.2 to 41.9/ 45.7 to 45.8)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev. 1
- 39. The plant is in Mode 1 at 100% power with Service Water Returns aligned to the Cooling Tower.
A Large Break Loss of Coolant Accident occurs.
All equipment responds as designed.
Which of the Service Water System conditions below would result in inadequate service water flow to the Containment Air Cooler to remove the heat from Containment for this design bases event?
A. A loss of air to the in service CAC Outlet Temperature Control Valves.
B. The Service Water non-seismic header ruptures.
C. Train 1 SW flow is inadvertently aligned to CAC 1 and CAC3.
D. SW 2931, CLNG TOWER MAKEUP is inadvertently closed.
Answer: C Explanation/Justification:
. Incorrect- Plausible because the loss of air to the CAC Outlet Temperature Control Valves will cause the valves to fail open and allow full flow, but this is the expected condition for the LOCA event when SFAS Actuates.
- . Incorrect- Plausible because a rupture of the non-seismic header would divert Service Water flow from essential component, however in this condition, SW1395 and SW1399 would isolate the non-essential header.
C. Correct- In modes 1,2 and 3 service water must be isolated to the spare CAC to ensure flow through the two in service CACs is adequate to support post LOCA cooling requirements D. Incorrect- Plausible since this is the inservice SW return flowpath, however the SW return flowpaths for the Intake Structure and the Forebay would open on high return pressure of 50 psig to provide a safety grade flowpath.
Sys# System Category KA Statement 022 Containment A 1. Ability to predict and/or monitor changes in parameters (to Cooling water flow Cooling prevent exceeding design limits) associated with operating the CCS System controls including:
(CCS)
KIA# A1.04 KIA Importance 3.2 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-06016 R29 Step 2.2.4 page 4 Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: High - Analysis 10 CFR Part 55 Content: (CFR: 41.5/ 45.5)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev. 1
- 40. The Plant is at 50% Power with #1 Makeup Pump out of service .
The following occurs:
ANNUNCIATOR ALARMS:
- SEAL INJ FLOW LO, 6-5-C
- SEAL INJ TOTAL FLOW, 6-6-C
- PZR LVL LO, 4-2-E CTRM INDICATIONS:
- #2 Makeup Pump discharge pressure reads 0 psig
- MU32, PZR LEVEL CONTROL, indicates 100% demand
- MU19, RCP SEAL INJ FLOW CONTROL, indicates 100% demand
- PZR level is 155 inches The crew has entered the appropriate Abnormal Operating Procedure.
What actions are required based on plant conditions?
Trip the Reactor. GO TO DB-OP-02000, RPS, SFAS, SFRCS TRIP, or SG Tube Rupture.
Commence a plant shutdown. GO TO DB-02504 Rapid Shutdown.
C. Trip Reactor Coolant Pumps 1-2 and 2-2. GO TO DB-OP-02515, Reactor Coolant Pump and Motor Abnormal Operation.
D. Place MU32 in hand. GO TO DB-OP-02513 Pressurizer Abnormal Operation.
Answer: A Explanation/Justification:
A. Correct- Minimum level for Tave 582*F is 160 inches below which requires tripping the Reactor. This is the mitigating strategy for a loss of all Makeup Pumps. Tripping at 160 inches will ensure a minimum inventory is maintained in the Pressurizer and then depressurize to allow use of HPI to recover Pressurizer level.
B. Incorrect- Plausible if Candidate knows a shutdown is required but does not recognize PZR level less than 160 inches requires a reactor trip.
C. Incorrect- Plausible because MU Pump 2 is lost and reactor power is less than the 55% setpoint when 1 RCP is running in each loop.
D. Incorrect- Plausible because MU32 is 100% open and a pressurizer control failure may be diagnosed Sys # System Category KA Statement 004 Chemical and Generic Ability to recognize abnormal indications for Volume system operating parameters that are entry-level Control conditions for emergency and abnormal operating System procedures.
KIA# 2.4.4 KIA Importance 4.5 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02512 R14 step 4.1.3 page 8 estion Source: New Level Of Difficulty: (1-5) 3
- estion Cognitive Level: High- Comprehension 10 CFR Part 55 Content: (CFR: 41.10 I 43.2 I 45.6)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1 41 . The plant is operating at 100% power
- EDG 2 has been started in accordance with DB-OP-06316, Emergency Diesel Generator Operating Procedure using the IDLE Start pushbutton and is running at 450 rpm.
The following event occurs:
- All Undervoltage Relays on D1 are actuated.
- All proper automatic actions occur.
Which of the following automatic and/or manual actions will be required to re-energize D1 bus?
A The EDG field will flash automatically.
The EDG will accelerate to 900 RPM, then AD1 01 EDG 2 Output Breaker will auto close.
B. The EDG will accelerate to 900 RPM.
The Idle Release Pushbutton must be depressed to flash the EDG field, then AD1 01, EDG 2 Output Breaker must be manually closed.
C. The operator must depress the Idle Release Pushbutton before EDG 2 will accelerate to 900 RPM.
The EDG field will automatically flash, then AD1 01, EDG 2 Output Breaker will auto close.
D. The operator must manually raise EDG 2 speed to 900 rpm.
The EDG field will flash automatically and the EDG output breaker, AD1 01 EDG 2 Output Breaker must be manually closed.
Answer: A Explanation/Justification:
A. Correct -The Idle Start/Stop Circuitry inhibits the voltage regulator by applying field shorting. An automatic start signal will release the Idle Start relay, accelerate the EDG, and enable the voltage regulator. The EDG output breaker would then auto close to restore power to D1 Bus.
B. Incorrect- The Idle Start/Stop Circuitry inhibits the voltage regulator by applying field shorting. An automatic start signal will release the Idle Start relay, accelerate the EDG, and enable the voltage regulator. Depressing the Idle release will not be necessary to flash the field.
C. Incorrect- The Idle Start/Stop Circuitry inhibits the voltage regulator by applying field shorting. An automatic start signal will release the Idle Start relay, accelerate the EDG, and enable the voltage regulator. Depressing the Idle release will not be necessary to accelerate the EDG.
D. Incorrect- The Idle Start/Stop Circuitry inhibits the voltage regulator by applying field shorting. An automatic start signal will release the Idle Start relay, accelerate the EDG, and enable the voltage regulator. Operator action to raise EDG speed will not be required. In addition, Operator action will not be necessary to close the EDG Output Breaker. The EDG output breaker would auto close to restore power to D1 Bus.
Sys# System Category KA Statement 062 AC Electrical K1. Knowledge of the physical connections and/or cause/effect ED/G Distribution relationships between the ac distribution system and the following System systems:
KIA# K1.02 KIA Importance 4.1 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-06316 R54, EDG Operating Procedure Step 2.2.12.
Question Source: New Level Of Difficulty: (1-5) 4 estion Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.2 to 41.9)
- bjective:
Davis Besse 1LOT13 NRC Written Exam Rev. 1 The Plant has experienced a Loss Of Coolant Accident with SFAS Levels 1 and 2 initiating. All equipment responded as designed.
Subsequently, a Loss of Off-Site Power occurs and Bus F1 is lost when the #2 EDG s~ ?(d restores power to Bus 01.
The Loss Of Coolant Accident continues to degrade with SFAS Levels 3, and 4 initiating Without Operator action, what is the current status of the Containment Spray Train 2?
Containment Spray Pump 2 is _ _ _ ____..-=-1...___ _ __.:.
CS1531, Containment Spray 2 Discharge Valve is _ _ _ _ ____.l(>=2'J..)_ _ _ ___.:.
A. (1) Off (2) Closed B. (1) Running (2) Closed C. (1) Off (2) Open Answer: C (1) Running (2) Open Explanation/Justification:
A. Incorrect- Plausible if the candidate believes CS1531 opens on the SFAS Level4 actuation, but does realize the Containment Spray Pumps is a 480 volt load and is lost when F1 is lost.
B. Incorrect- Plausible if the candidate believes CS1531 opens on the SFAS Level 4 actuation, but fails to realized the Containment Spray Pump is a 480 load as unlike the other SFAS Actuated Pumps that are supplied from 4160 essential power.
C. Correct CS1531 opens on the SFAS Level 2 actuation and is therefore unaffected when F1 loses power. Since the Containment Spray pump is supplied from F1, it will be off when F1 is lost.
D. Incorrect- Plausible is the candidate understands CS1531 opens on the SFAS Level 2 actuation, but fails to realized the Containment Spray Pump is a 480 load as unlike the other SFAS Actuated Pumps that are supplied from 4160 essential power.
Sys# System Category KA Statement 026 Containment K2. Knowledge of bus power supplies to the following: MOVs Spray System (CSS)
KIA# K2.02 KIA Importance 2.7* Exam Level RO References provided to Candidate None Technical
References:
OS-005 R12 Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7)
Objective:
Davis Besse 1 LOT13 NRC Written Exam Rev.1 43 . A Plant Startup is in progress. The plant is operating at 20% power at the end of life. The Main Turbine has been synchronized to the grid.
- Power range Nl8 calibration is in progress
- Reactor Demand is in Manual
- Diamond Rod Control Panel is in Auto A loud noise is heard outside the control room accompanied by the following:
- Reactor Power is rising
- Megawatts are lowering
- Feedwater is rising
- RCS pressure is lowering Which of the following explains why reactor power is increasing?
A. Positive reactivity is being added due to lowering Tave B. An Undesired Rod withdrawal is in progress C. ICS is raising power in response to lowering megawatts I&C has placed Nl8 in Test Operate while it was the highest indicating Nl Answer: A Explanation/Justification:
A. Correct- A steam leak is in progress per DB-OP-02525, Steam Leaks. A lowering Tave will add positive reactivity with a negative moderator coefficient.
B. Incorrect- Plausible because an undesired rod withdraw will raise power but would not include the listed symptoms C. Incorrect- Plausible because raising power would normally increase megawatts but ULD output tracks megawatts in manual D. Incorrect- Plausible because ICS power selects the highest auctioneered power and power would increase if the ULD was in auto Sys# System Category KA Statement 039 Main and K5. Knowledge of the operational implications of the following Effect of steam removal on reactivity Reheat concepts as the apply to the MRSS:
Steam System (MRSS)
KIA# K5.08 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02525 R10 Page 5 Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: Low - Fundamental 10 CFR Part 55 Content: (CFR: 441.5/45.7)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1 44 . The following plant conditions exist:
- The reactor is operating at 50% rated power.
- All Feedwater Control Valves are in AUTOMATIC.
- ICS is in full AUTOMATIC mode.
Which one of the following describes feedwater flow control by ICS following a manual reactor trip?
A. Places the MFP at a constant target speed and immediately controls the Feedwater Control Valves position based on feedwater flow error.
B. Places the MFP at a constant target speed and immediately controls the Feedwater Control Valves position based on SG level error.
C. Runs the MFP to a target speed which is then modified by SG feedwater flow error and positions Feedwater Control Valves to a target position until a 2.5 minute timer expires.
D. Runs the MFP to a target speed which is then modified by SG level error and positions Feedwater Control Valves to a target position until SGs are at low level limits or a 2.5 minute timer expires .
- liswer: D Explanation/Justification:
A. Incorrect- Rapid Feedwater Reduction will actuate. Feedwater Control valves will control on SG level error, not Feedwater flow error.
B. Incorrect- Rapid Feedwater Reduction will actuate. Feedwater Control valves will control on SG level error, but a timer operates to allow SG level to lower to low level limits C. Incorrect- Rapid Feedwater Reduction will actuate. Feedwater Control valves will control on SG level error, not Feedwater flow error.
D. Correct- With full automatic ICS operation and SG not initially on low level limit control, a reactor trip will caused the MFP to go to target speed, the SU SG Level controls to target position until SG on Low Level Limit or 2.5 minute timer times out.
Sys # System Category KA Statement 059 Main A 1. Ability to predict and/or monitor changes in parameters (to Feed Pump speed, including normal control speed Feedwater prevent exceeding design limits) associated with operating the MFW for ICS (MFW) controls including:
System KIA# A1.07 KIA Importance 2.5* Exam Level RO References provided to Candidate Technical
References:
Lesson Plan OPS-SYS-1512 R06 page 13 & 14 None Question Source: BANK 38076 Level Of Difficulty: (1-5) 3.5 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5/ 45.5)
Objective:
ICS OVERVIEW OPS-SYS-1512 REV 06
- RCS Flow
- h. Feedwater Control valves (1) Controlled by one of three control signal when in automatic.
- Level Error control o Operate Level Compared to Hi Level Limit of90.0%. 12-3-A(B)
Prevents overfill of S/G (carryover) and flooding aspirating port.
o SIU Level compared to Low Level Limit of(40"). 14-5-E(F) o Either alarm states that the S/G is on LLL or HLL control.
- It does NOT mean that actual Steam Generator level is at set point but level error control is in effect.
- Flow Error control o Flow error is used to control the Feedwater valves o Reduces feedwater to prevent overcooling the RCS and emptying the Pressurizer.
~,~ o Requirements to activate:
- at least one MFP Reset
- RFR Switch "on"
- All FW control valves AUTO (RFR will continue if a valve is taken out of AUTO after RFR actuates.)
- Trip confirm< 23.5% by highest power range NI.
o RFR Valve Effect:
- Closes main valve
- S/U valve to 17% open (I&C sets) Equivalent to about 4% load o Returns to Level error:
- S/G on low level limit OR
- 2.5 minute timer expired.
o RFR FW Pump effect:
- 4600 RPM speed
- MFP to target speed modified by level error Page 13 of23
ICS OVERVIEW OPS-SYS-1512 REV 06 o Requirements to Release RFR, valves only:
- 2.5 minute timer expired OR
- S/GonLLL Trip Confirm
<23.5% hi 0 auctioneered.
(1) Reactor trip closes FW 779 and FW 780.
J. Feedwater Pump control (1) Normal automatic control signal is developed from two inputs
- Total Feedwater demand for a course control ofMFP speed.
(2) Feedwater Pump 1 controller has a bias to allow matching of MFP flows while they are running in parallel.
(3) Post trip the MFP is controlled at a target speed.
- 5. Reactor Control Subsystem SLIDES 49-51 (Refer back to Slide 46 as needed)
- a. Function (1) Converts a demand signal to a rod command signal.
(2) Maintain a constant T-ave of 582oF above 28.5%.
(3) Produces insert and withdraw commands that go to the Control Rod Drive system to control Reactor power.
(4) Varies the total Reactor's heat output so unit generation demand is satisfied while maintaining Reactor Coolant Average Temperature at set point.
- b. Controls by Operator (Hand/Auto Stations) Show Simulator Graphics Rx DMD (I) Hand/Auto Reactor Demand
Reference:
- Minimum Setpoint 28.5% from SO/Reactor Demand. 23.5% for 3 RCP's and T -ave Control at Reactor Dmd Limiter.
Page 14 of23
Davis Besse 1LOT13 NRC Written Exam Rev. 1 45 . The plant is at 70% Power.
- Annunciator 10-1-C, MFPT 1 Lube Oil Press Lo alarms
- Pl1206, Header pressure indicates 3.6 Psig
- Neither the Preferred or Standby #1 MFPT Main Oil Pump are running.
- Both MFP Turbines are operating at approximately 4400 rpm.
The plant remains stable at 70% power.
Which of the following actions are required?
A. Trip #1 MFPT only if Pl1206, #1 MFPT Lube Oil Header pressure lowers to 3 psig.
B. Start MFPT 1 Emergency Bearing Oil Pump and then Trip #1 MFPT.
C. Start the Motor Driven Feedwater Pump and Trip #1 MFPT.
D. Reduce Reactor power to 60% in preparation for loss of #1 MFPT.
Answer: B Explanation/Justification:
A. Incorrect- Plausible if the MFPT emergency bearing oil pump auto started and 3 psig was the MFPT trip setpoint
- . Correct- DB-OP-0201 0 directs starting the MFPT emergency bearing oil pump and tripping MFPT 1 if bearing header goes below 4.0 psig which is the auto trip setpoint
- Incorrect- Plausible since MFPT 1 should be tripped starting the MDFP will provide additional inventory to the SG which may facilitate maintaining SG Level.
D. Incorrect- Plausible since 60% is the high discharge pressure of MFPT run back target.
Sys # System Category KA Statement 059 Main A4. Ability to manually operate and monitor in the control room: MFW turbine trip indication Feedwater (MFW)
System KIA# A4.01 KIA Importance 3.1* Exam Level RO References provided to Candidate Technical
References:
DB-OP-02010 R17 pages 8 & 9 None Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7/45.5 to 45.8)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1
- 46. The Plant is at 100% power.
- The zone 3 Equipment Operator reports the piping at AF608, Auxiliary Feedwater to Steam Generator Line 1 Stop, is hot to the touch and Auxiliary Feedwater (AFW) Train 1 is Steam Bound Which of the following statements is the correct impact of this condition on Emergency Feedwater and the action required?
A. Auxiliary Feedwater Train 1 remains Operable. Designate an Operator to maintain AFW Train 1 Operable during venting to refill AFW Train 1.
B. AFW Train 1 and the Motor Driven Feed Pump are Inoperable. Initiate actions to commence a Reactor shutdown within one hour.
C. AFW Train 1 and AFW Train 2 are Inoperable. Start the Motor Driven Feed Pump in the Auxiliary Feedwater mode to condense the steam bubble at AF608.
D. AFW Train 1 and AFW Train 2 and the Motor Driven Feed Pump will be rendered Inoperable due to closing AF608. Take action immediately to restore Operability Answer: D Explanation/Justification: Note: At DB, the Auxiliary Feedwater and Main Feedwater System do not share physical connections since they feed the Steam Generators via separate headers. In order to use the KA, a back leakage from the Steam Generator question was used into the Auxiliary Feedwater System.
- . Incorrect -Plausible because the candidate may assume the piping is only hot at the containment isolation and therefore not affect the AFW Pumps. Operator action is allowed under some condition, but the short duration of the AFW time response would not permit operator actions.
B. Incorrect -Plausible because a low SG Pressure condition will align AFW Train 1 and 2 to Feed SG 1 via AF608. Also, the MDFP could be used to provide cool water from the Condensate Storage Tank to mitigate this condition.
C. Incorrect - Plausible because without a low SG Pressure condition, only AFW Train 1 and the MDFP would use the AF608 flowpath D. Correct- Because a low SG Pressure condition, AFW Train 1 and 2 and the MDFP could use the AF608 flowpath. As a result, all three would be inoperable with AFW Train 1 steam bound.
Sys# System Category KA Statement 061 Auxiliary I A2. Ability to (a) predict the impacts of the following malfunctions or Back leakage of MFW Emergency operations on the AFW; and (b) based on those predictions, use Feedwater procedures to correct, control, or mitigate the consequences of those (AFW) malfunctions or operations:
System KIA# A2.06 KIA Importance 2.7 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-06233 R35 Steps 2.1.6 & 4.9.5.a.3 and TS 3.7.5 ConditionE Question Source: New Level Of Difficulty: (1-5) 4 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5 I 43.5 I 45.3 I 45.13)
Objective:
5 DB-OP-06233 Revision 35
- 1.0 2.0 PURPOSE To provide instructions for operating the Auxiliary Feedwater System during Normal, Infrequent or Special, and Emergency modes of operation.
LIMITS AND PRECAUTlONS 2.1 Administrative 2.1.1 Auxilary Feedwater System operability requirements are given in TS 3. 7.5, Emergency Feedwater (EFW).
2.1.2 Condensate Storage Tank operability requirements are given in TS 3.7.6, Condensate Strorage Tanks (CSTs).
2.1.3 Whenever any portion ofthe Auxiliary Feedwater System is INOPERABLE, the Reactor Operator shall tum on IL-4800, AUX FW, using HS-4800 on SFAS Panel C-5717.
The light shall remain LIT until the Auxiliary Feedwater System is made OPERABLE.
2.1.4 When AF21 or AF22 are open to operate either AFW pump on recirc to the Condensate Storage Tank while in MODES I, 2, or 3 an operator shall be stationed at the respective valve and will be in direct communications with the Control Room while either valve is open .
- 2.1.5 Whenever Door 215 is required to be open for an extended period oftime an individual in the AFP room shall be assigned the responsibility to close and latch the door after personnel have exited in the event an emergency occurs in either of the AFP rooms.
Attachment 8, Door 215 Operation, provides this information.
-.l~~~~j:-~.}~*~y'*~~-. -:}~1-~::'~; --~*;*~~~*~~~~,*~-' ,,, ,;i .?t******,'§'h'ftli*'~i>'o!l'i/*'1'!"'""~*
~2.1.6 Closing AF599* and I or AF608* in MODE 1 > 40% RTP renders all three EFW trains inoperable.
Closing AF599* and I or AF608* in MODE 1 :::.; 40% RTP and in MODES 2 ~J or 3 renders both AFW trains inoperable. i'l-ln MODE I :::.; 40% RTP and in MODES 2, 3, and 4, the MDFP remains . J OPERABLE provided AF599 and AF608 are capable of being realigned to the 'I*
open position. (Reference TS 3.7.5) . ,J s> ~~*' ilf:* l;,;~_:.J.*~; *i-(".r-*
,,'¢-.**e.,.:*>~'9,., ,.~:iftJ* *~~'-""~o'-**'":-* <*t~"f.; '>i:;~~.t"'~* **,*'."
- Controlled in accordance with DB-OP-00008, Operation and Control of Locked Valves
Davis Besse 1LOT13 NRC Written Exam Rev.1 Inverter YVA supplies power to Uninterruptable Bus YAU.
(1) What is the power supply to Inverter YVA when the static transfer switch is in Normal (2) What is the power supply if the static transfer switch transfers to Alternate?
A. (1) Non-essentia1480 VAC (2) Non-essential120 VAC B. (1) Non-essential480 VAC (2) Essential 120 VAC C. (1) 250 VDC (2) Non-essential120 VAC D. (1) 250 VDC (2) Essential 120 VAC Answer: C Explanation/Justification: Inverter YVA supplies the uninterruptable 120 vdc bus YAU.
A. Incorrect- Plausible because YVA does not use essential power. Both choices use non-essential power.
B. Incorrect- Plausible because YAU is an important plant power supply for fire protection, communications, ICS, NNI etc. It is logical this power would be essential when transferred to alternate.
Correct- This is the configuration for Inverter YVA as provide in the System Operating Procedure for normal lineup Incorrect- Plausible because YAU is an important plant power supply for fire protection, communications, ICS, NNI etc. It is logical this power would be essential when transferred to alternate. Also 250 VDC is feed from the Safety Related Station Batteries 1P and 1N.
Sys# System Category KA Statement 062 AC Electrical A3. Ability to monitor automatic operation of the ac distribution Operation of inverter (e.g., precharging Distribution system, including: synchronizing light, static transfer)
System KIA# A3.04 KIA Importance 2.7 Exam Level RO References provided to Candidate Technical
References:
DB-OP-06319 R25, page 2 & 192 None Question Source: BANK 32192 Level Of Difficulty: (1-5) 4 Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.7 I 45.5)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1
- 48. The Plant has experienced a complete loss of AC Power. Performance of DB-OP-02521, Loss of AC Bus Power Sources, is in progress.
At 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following the beginning of the event AC power is still lost with required actions of DB-OP-02521 to reduce battery discharge rate completed.
(1) What is the current DC alignment AND (2) How long will it be before DC power is no longer available?
A. (1) Batteries 1P, 1N, 2P and 2N are in service.
(2) less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> B. (1) Batteries 1P and 1N are in service. Batteries 2P and 2N are in standby.
(2) approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> C. ( 1) Battery 1N is in service. Batteries 1P, 2P and 2N are in standby.
(2) approximately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.
D. (1) Battery 1P is in service. Batteries 1N, 2P and 2N are in standby.
(2) greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Answer: D Explanation/Justification: DC Bus Load shedding is performed to reduce Discharge Rate and therefore extend battery life.
A. Incorrect- Plausible because the batteries are designated as having a 1500 amp-hour rating based on an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> discharge rate.
B. Incorrect- Plausible if it is assumed there are 250V loads required to remain energized following load shedding C. Incorrect- Plausible since one battery (1 P) will remain in service and 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> is a multiple of 8.hours D. Correct- DB-OP-02521 will direct reducing 1P to minimum required loading and completely unloading the remaining three batteries to be used in series to extend battery capacity. This is a new configuration that extends the time to meet minimum DC Bus Loads.
Sys# System Category KA Statement 063 DC Electrical A 1. Ability to predict and/or monitor changes in parameters Battery capacity as it is affected by discharge rate Distribution associated with operating the DC electrical system controls System including:
KIA# A1.01 KIA Importance 2.5 Exam Level RO References provided to Candidate None Technical
References:
DB-OP-02521 R20 Attachment 17 page 128 Question Source: New Level Of Difficulty: (1-5) 4 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5/ 45.5)
Objective:
128 DB-OP-02521 Revision 20 Page 4 of 7 Attachment 2 - This attachment provides the specific direction for reenergizing Bus 01. The SBODG and Off-site power are preferred over the EDGS since the actions already taken to return the EDGS to service have not been successful. Load limits are provided if the SBODG or the EDGS are used.
Attachment 3 - This attachment performs two functions. The first function is to resolve and clear a lock out on Cl bus- Once the lock out is resolved, this attachment provides methods to restore power to C1. This method was selected to reduce the number of procedural transfers from attachment to attachment that would be required to restore power.
Attachment 4 - This attachment performs two functions. The first function is to resolve and clear a lock out on Dl bus- Once the lock out is resolved, this attachment provides methods* to restore power to D1. This method was selected to reduce the number of procedural transfers from attachment to attachment that would be required to restore power.
Attachment 5 - This attachment is used to reduce the load on the station batteries. This attachment is started after approximately 15 minutes to ensure the attachment is completed within 30 minutes of the loss of battery charger power. The attachment has 3 sections.
Section 1 addresses a loss of both DCMCC 1 AND DCMCC2 Battery Chargers. Historically, Battery Load shed was intended to ensure station essential DC Loads remain available for at least one hour following a loss of all AC power. Refer to Calculation C-EE-002.01-016 "Station Battery Discharge Analysis for Beyond Design Bases Events" for additional information.
Based on INPO L 1 IER 11-4, the load shed method was altered to provide additional battery life. Sections 2 and 3 of this attachment address a loss of power to the chargers associated with C1 and D1 respectively. The intent of these sections is to reduce loading on the DCMCC 1 or 2 by transferring YAU/YBU off the batteries and on to the non essential supplies. This is only possible if YARIYBR have power. Without offsite power, YARIYBR will be de-energized.
Refer to Calculation C-EE-002.01-01 0, DC System Analysis for additional information.
~~=~~~t~i=e~~~;:;,~~~:~i~t~~;:~~~!fc;:ft:b:~r~*;&*~,~
!*operators will begin to load shed the station batteries. The load shedding is performed to
.~ conserve energy in the station batteries. Portions of the load shed are required to be
~ completed within 30 minutes and remaining actions within 60 minutes of event initiation.
ir These completion times are used in the Load Shed analysis to determine expected battery life !
~; following load shed. It is anticipated that Batter 1P would provide service via Y1 and Y1A for
- I approximately 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> followed by Battery 2P providing service via Y2 and Y2A for an additional 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. Battery 1 N and 2N are not credited in the calculation since available "il instrumentation would not provide Steam Generator Level. ~.~'lw:-"'Jf8i*,'f&!IIJ,
' I ~-'l* ',, .*,.; t *~
- AITACHMENT 17: Background Information
Davis Besse 1LOT13 NRC Written Exam Rev.1 49 . Reactor Power is 100% with all systems in a normal alignment.
The following events have occurred:
All AC power has been lost.
Following DC Bus load shed per DB-OP-02521, Loss of AC Bus Power Sources, AFW Pump 1 is in service supplying SG 1.
With no operator action, as Battery voltage lowers toward zero, what will be the effect on SG 1 Level?
SG 1 level will:
A. Lower due to AFW Pump Discharge Target Rock valve failing closed B. Lower due to AFW Pump speed going to the low speed stop C. Rise due to AFW Pump Discharge Target Rock valve failing open D. Rise due to AFW Pump speed going to the high speed stop Answer: C
.planation/Justification: Loss of all AC Power A. Incorrect- Plausible because SG level will lower if the target rock valve were to fail closed B. Incorrect- Plausible because SG level would lower if the turbine went to its low speed stop C. Correct- Target rock fails open on low voltage and SG will have full flow.
D. Incorrect- Plausible because SG level will rise if the turbine went to its high speed stop Sys# System Category KA Statement 063 DC Electrical A4. Ability to manually operate and/or monitor in the control room: Battery voltage indicator Distribution System KIA# A4.02 KIA Importance 2.8* Exam Level RO References provided to Candidate Technical
References:
DB-OP-02521 R20, Att 17 page 130 None Question Source: BANK 79886 Level Of Difficulty: (1-5) 3 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 I 45.5 to 45.8)
Objective:
Davis Besse 1 LOT13 NRC Written Exam Rev.1 50 . The following plant conditions exist:
- Emergency Diesel Generator 1 monthly surveillance test is in progress
- Emergency Diesel Generator 1 is paralleled with the C 1 bus The following event occurs:
- A Safety Features Actuation System Level 2 signal occurs Assuming no operator action is taken and the EDG output breaker opens as designed, how will the #1 Emergency Diesel Generator respond to this SFAS Level2 signal?
The Emergency Diesel Generator Governor w i l l - - - - - - - - - - -
A. transfer to the isochronous mode and all engine trips will be active B. remain in the droop mode and all engine trips will remain in active C. remain in the droop mode and non-vital engine trips will be bypassed D. transfer to the isochronous mode and non-vital engine trips will be bypassed Answer: D
-==Justification:
A. Incorrect- Although the Governor will transfer to Isochronous mode, non-vital engine trips will still be active. This is plausible since the EDG was already running at the time to of the SFAS start signal.
B. Incorrect- This is plausible since the EDG was already running at the time to of the SFAS start signal. In normal parallel operation, the EDG operates in Droop with all engine trips active.
C. Incorrect- This is plausible since the EDG was already running at the time to of the SFAS start signal. In normal parallel operation, the EDG operates in Droop with all engine trips active. Since an SFAS occurred, it is reasonable to assume some normal trips are bypassed.
D. Correct- The Governor will transfer to the Isochronous mode, and non-vital engine trips are bypassed.
Sys# System Category KA Statement 064 Emergency K4. Knowledge of ED/G system design feature(s) and/or interlock(s) Governor valve operation Diesel which provide for the following:
Generator (ED/G)
System KIA# K4.03 KIA Importance 2.5 Exam Level RO References provided to Candidate Technical
References:
DB-OP-06316 step 2.2.12 & Caution 5.2.3 None Question Source: BANK32132 Level Of Difficulty: (1-5) 3 Question Cognitive Level: High -Comprehension 10 CFR Part 55 Content: (CFR: 41.7)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1 51 . What type of detector is used for RE1 003A, Steam Jet Air Ejector Discharge detector, to monitor for steam generator tube leaks?
A. Scintillation Detector B. Geiger-Mueller Detector C. lon Chamber Detector D. Fission Chamber Answer: A Explanation/Justification:
A. Correct- RE1003A is a Gamma Scintillation detector. A scintillation detector is used for isotope determination via gamma spectrosophy.
B. Incorrect- Plausible because most area radiation monitor detectors are G-M detectors C. Incorrect- Plausible because the Station Vent Accident Range Monitors, among others, are Ion Chamber Detectors D. Incorrect- Plausible- Fission Chambers are use to detect neutrons such as Gammametrics Nuclear Instruments.
Sys# System Category KA Statement 073 Process K5. Knowledge of the operational implications as they apply to Radiation theory, including sources, types, units, Radiation concepts as they apply to the PRM system: and effects Monitoring (PRM)
System KIA# K5.01 KIA Importance 2.5 Exam Level RO
.ferences provided to Candidate None Technical
References:
SD-017A R03 page 2-2 Step 2.1.1.6 USAR TABLE 11.4- 1 on Page 11.4-20 Question Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: Low- Fundamental 10 CFR Part 55 Content: (CFR: 41.5/45.7)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev.1 52 . The plant is at 100% power with all systems in normal alignment EXCEPT Containment Air Cooler 2 is running in SLOW speed for testing The following event occurs:
- Loss of off-site power.
- EDG 2 fails to start.
- All other systems function as designed.
5 minutes after the loss of off-site power occurred, assuming no Operator action, what will be the position of the following valves?
(1) SW1366 - CTMT Air Cooler 1 Inlet Iso (2) SW1367 - CTMT Air Cooler 2 Inlet lso A. (1) Open (2) Open B. (1) Open (2) Closed C. (1) Closed (2) Open Answer: C (1) Closed (2) Closed Explanation/Justification:
A. Incorrect- plausible since refill logic will close SW1366 but would then reopen if CAC 1 was in slow if an SFAS level 2 existed.
B. Incorrect- plausible if Candidate does not know SW1367 is powered from F12A via EDG2 since SW1367 would close and remain closed by refill logic due to CAC 2 was in fast and SW1366 would reopen since CAC 1 was in slow C. Correct- Refill logic will close SW1366 which must be manually opened since no SFAS signal is present. SW1367 will remain open since power is lost D. Incorrect- plausible if Candidate does not know SW1367 is powered from F12A via EDG 2 since SW1367 would close and remain closed by refill logic due to CAC 2 was in fast Sys# System Category KA Statement 076 Service K2. Knowledge of bus power supplies to the following: Reactor building closed cooling water Water System (SWS)
KIA# K2.04 KIA Importance 2.5* Exam Level RO References provided to Candidate None Technical
References:
OS-020 Sheet 2, R45 CL 11 Question Source: New Level Of Difficulty: (1-5) 4 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7)
Objective:
Davis Besse 1LOT13 NRC Written Exam Rev. 1 53 . The plant is in Mode 3 normal operating temperature and pressure with both steam line isolation valves OPEN.
Instrument air is lost to MS 101, Main Steam Line 1 Isolation Valve (1) How will MS101, Main Steam Line 1 Isolation Valve respond to this loss of instrument air?
(2) How will this loss of instrument air affect MS1 01, Main Steam Line 1 Isolation Valve Tech Spec required stroke time?
A. (1) fail closed (2) WILL still meet its Tech Spec required stroke time B. (1) fail closed (2) WILL NOT meet its Tech Spec required stroke time C. (1) remain open (2) WILL still meet its Tech Spec required minimum stroke time D. (1) remain open (2) WILL NOT meet its Tech Spec required stroke time
.nswer:C Explanation/Justification:
A. Incorrect- Plausible if candidate knows an accumulator exists but determines it is only for assisting closure to meet minimum stroke requirements B. Incorrect- Plausible if candidate knows an accumulator exists but determines it is only for ensuring closure without meeting stroke requirements C. Correct- Accumulator will both hold open MSIV and pneumatic via N2 assist closing springs to meet design minimum required closing requirement D. Incorrect- Plausible if Candidate knows an accumulator exists but determines it is only for temporarily maintaining valve open and not also close assist Sys # System Category KA Statement 078 Instrument K1. Knowledge of the physical connections and/or cause-effect MSIVair Air System relationships between the lAS and the following systems:
(lAS)
KIA# K1.05 KIA Importance 3.4* Exam Level RO References provided to Candidate None Technical
References:
SD-012A R05 page 2-5 and 2-6 Question Source: New Level Of Difficulty: (1-5) 3.5 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.2 to 41.9/45.7 to 45.8)
Objective:
The combined relieving capacity of the MSSVs is 14,174,922 pounds per hour, which is conservative with respect to the required capacity of 13,171,200 lbs/hr (References 4.5.19 and 4.4.4).
Under normal plant operating conditions with turbine valves wide open, the MSS is designed to transport steam from the SGs at outlet conditions of 590°F and 925 psia to the following points at the steam conditions listed:
- Main Turbine Stop Valves Refer to references 4.5.37 and 4.5.38 (SD-004) for steam conditions
- Moisture Separator/Reheaters Refer to references 4.5.37 and 4.5.38/
(SD-012B) for steam conditions
- Auxiliary Steam and Hot 250,000 lbm/hr (normal flow is 70,000
' Water System (SD-027) to 100,000 lbm/hr)
(Reference 4.2.35)
During startup, the MSS supplies 4,550,000 lbm/hr of steam (total) to both MFPTs (Reference 4.3.33). Evaluations were performed to determine the impact j of the Measurement Uncertainty Recapture (MUR)power uprate. These evaluatio~s .J
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2.3 ARRANGEMENT *-~'...
All piping, MSIVs, MSAVVs, and MSSVs are readily accessible for inservice inspection (ISI) (Reference 4.4.1). The routing of the main lines of the MSS is shown on References 4.1.70 through 4.1.81 .
2.4 COMPONENT DESIGN The design data of the major components of the MSS are given in Table 2.4-1.
The purpose and features of the major components are described in this section.
Main Steam Isolation Valves MS101 and MS100 The valves are designed for low leakage upon closure due to a break downstream of the valve. Complete closure for reverse flow occurs when line pressure is less than 80 psig. Under this condition, the leakage rate is not greater than 0.2 pound per hour. The MSIVs are safety-related, air-operated, balanced-disc stop valves set in line with the normal flow direction (Reference 4.4.12). The closure speed of the MSIV can be varied by adjusting the hydraulic control knobs of the hydraulic cylinder, mechanically coupled to the air cylinder (Reference 4.3.11). They are designed to be operated with a differential pressure of 910 psi across the valve. With steam flow in the normal direction and a differential pressure of 910 psi across the MSIVs, the MSIVs are designed to fully close within 5 seconds after the receipt of the closing signal. To assist the springs which shut the valve, each MSIV has a safety grade air accumulator which will provide additional closing force in the event of a loss of instrument air. This ensures the time requirements above can be met. The MSIV Bypass Valves are normally closed. However, if open, they are designed to fully close upon receipt of the closing signal (References 4.2.2 and 4.4.21, and USAR Table 6.2-23). The accumulators are designed to keep the valves open for 5 days after a loss of air (Reference 4.5.15, 4.5.16, 4.6.7 and 4.6.8) .
Improvements were made to the stem and disc assembly for MS100 and MS101. This 2-6 SD-012A Rev. 5
Davis Besse 1LOT13 NRC Written Exam Rev.1 54 . Which of the following systems, interlocks, or controls ensure the Containment Vessel remains above the MINIMUM internal design pressure?
A. BWST maximum Temperature limits.
B. Containment Spray nozzle size and location C. Containment Spray Discharge Valve Throttle position D. Containment Vacuum Relief Valves.
Answer: D Explanation/Justification:
A. Incorrect- Plausible since a temperature decrease in the BWST would result in a lower Containment Pressure during an inadvertent spray event.
BWST maximum temperatures ensure post LOCA injection removes the heat assumed in the accident analysis.
B. Incorrect- Plausible since the spray patterns and location would affect the low pressure created during an inadvertent spray event.
C. Incorrect- Plausible since throttling spray flow would affect pressure reduction but this interlock actuates post LOCA to prevent runout of the CTMT Spray Pumps.
D. Correct- The CTMT Vacuum Relief capacity is designed to protect containment against an inadvertent actuation of CTMT Spray causing significant reduction of Containment Pressure (absolute scale).
Sys # System Category KA Statement 103 Containment Generic Knowledge of the purpose and function of major System system components and controls.
KIA# 2.1.28 KIA Importance 4.1 Exam Level RO ferences provided to Candidate None Technical
References:
SD-022F R01 Step 1.1.2.1
- uestion Source: New Level Of Difficulty: (1-5) 3 Question Cognitive Level: Low - Fundamental 10 CFR Part 55 Content: (CFR: 41.7)
Objective: