IR 05000282/1999301
| ML20209H845 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/02/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20209H800 | List: |
| References | |
| 50-282-99-301OL, 50-306-99-301OL, NUDOCS 9907210082 | |
| Download: ML20209H845 (143) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGIONlli Docket Nos:
50-282; 50-306 License Nos:
50-282/99301(OL); 50-306/99301(OL)
Licensee:
Northern States Power Company Facility:
Prairie Island Nuclear Generating Plant
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Location:
1717 Wakonade Dr. East Welch, MN 55089 Dates:
May 17-21,1999 Examiners:
D. McNeil, Chief Examiner D. Muller, Rill Ex6 miner Approved by:
David E. Hills, Chief, Operator Licensing Branch Division of Reactor Safety 9907210082 990716 PDR ADOCK 05000282 V
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EXECUTIVE SUMMARY Prairie Island Nuclear Power Station NRC Examination Reports 50-282/99301(OL); 50-306/99301(OL)
A licensee developed and Nuclear Regulatory Com.nission approved initial operator licensing examination was administered to three Reactor Operator license applicants. The examiners observed a period of routine operations in the control room.
Results:
Three applicants were administered an initial license examination. Two applicants passed all portions of the examination and were issued Reactor Operator licenses. One applicant passed I
the operating test, but failed the written examination and was denied a license.
Operations:
Operators performed their shift responsibilities in a professional, business-like manner during the observed period. Procedures reviewed by the examiners contained adequate guidance (cautions, precautions, notes, and steps) for the station operators to properly operate the station's systems. (Sections 01.1, O1.2)
Ev. amination Summary:
The licensee submitted a well prepared initial examination outEne. The outline conformed with the guidelines contained in NUREG 1021, Operator Licensing Examination Standards for Power
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Reactors, Intenm Revision 8, January 1997. The submitted operating test was acceptable without modification. (Section O5.2)
The high number of unsatisfactory questions (12) submitted to the NRC, the relatively high number of post examination comments (9), and the five hour validation time of the written examination indicated that the development of a written examination was an area that needed improvement. (Sections 05.2 and 05.4)
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Report Details 1. Operations
Conduct of Operations
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01.1 GeneralCon'ments
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Scope (IP 71707)
A Nuclear Regulatory Commission (NRC) examiner observed routine control room activities during full power operations for a two-hour period using Inspection Procedure 71707, Plant Operations. The examiner observed a routine surveillance, verbal communications, annunciator responses, and control room panel attentiveness b.
Observations and Findings
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The lead Reactor Operator (RO) conducted a routine surveillance with an instrument and control technician during the observed period. The surveillance was conducted without mishap. The control room crew engaged in routine face-to-face discussions and consistently used three-way communications. The ROs responded to several annunciators by announcing them to the Senior Reactor Operators and stating if they were expected or unexpected. The operators referenced the annunciator procedures for unexpected alarms. The crew generally had at least one RO per unit maintaining visual contact with the control boards.
c.
Conclusions Operators performed their shift responsibilities in a professional, business-like manner during the observed period.
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Operations Procedures and Documentation 03.1 General Comments a.
Scope (71707)
Using inspection Procedure 71707, the examiners reviewed selected administrative and operations procedures during the initial license examination validation.
b.
Observations and Findinos The examiners reviewed the procedures used to develop the operating test. The procedures were logically organized and contained adequate instructions, notes, precautions and cautions to direct the operators in the execution of their responsibilities.
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c.
Conclusions Procedures reviewed by the examiners contained adequate guidance (cautions, precautions, notes, and steps) for the station operators to properly operate the station's systems.
Operator Training and Qualification 05.1 General Comments Nuclear Regulator Commission examiners administered operator initial license examinations at the Prairie Island Nuclear Power Station to three RO applicants during the week of May 17,1999. Two RO applicants successfully passed all sections of the initiallicense examination and were issued RO licenses. The remaining applicant passed the operating test, but failed the written examination and was denied an RO license.
Prairie Island training department instructors used the guidance prescribed in NUREG 1021, Operatof Licensing Examination Standards for Power Reactors, interim Rev. 8, January 1997, to prepare the cperating test and written examination. The training staff administered the written examination and NRC examiners administered the operating test and observed portions of the administration of the written examination.
05.2 Pre-Examination Activities a.
Scope Nuclear Regulatory Commission examiners reviewed the examination material submitted by the training department's examination developers using the guidance prescribed in NUREG 1021.
b.
Observations and Findinas 1.
Examination Outline:
The initial outline submittal was timely and developed in accordance with the quantitative and qualitative requirements of NUREG 1021 ES-201-2,
" Examination Outline Quality Assurance Checklist." No changes were necessary to the submitted outline.
2.
Initial Submittal:
Written Examination The examiners reviewed the written examination and determined that twelve submitted questions were unsatisfactory. The twelve questions were not written in accordance with NUREG 1021, Section D.2.b, in that they contained significant psychometric flaws. Eleven questions were unsatisfactory because
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an applicant taking the test would not be able to determine the correct answer when given the question stem. The twelfth question was unsatisfactory because it was non-discriminatory. The examiners determined that the unsatisfactory questions could be replaced or corrected in a timely manner, which would prevent a delay in the administration of the written examination. The NRC examiners also requested that the station instructors modify several other questions to correct grammar errors.
One week prior to the on-site vali6 tion week, station examination developers informed the NRC chief examirier that the examination had been validated as a five hour examination. This was contrary to the guidance provided in NUREG 1021, Appendix E, Section B.3, which stated that the examination was to be a four hour examination. The NRC examiners disagreed with the assessment of the facility instructors and directed the examination proctor to administer the examination in accordance with the NUREG 1021 guidelines. All three applicants finished the examination within the allotted four hours.
Operatina Examination The administrative Job Performance Measures (JPMs), operating JPMs and dynamic simulator scenarios were all acceptable as submitted by the training department's staff.
c.
Conclusions The initial examination outline was weii prepared. The high number of unsatisfactory questions (12) submitted to the NRC and the five hour validation time of the written examination indicated that the development of a written examination was an area that needed improvement. The submitted operating test was acceptable without modification.
05.3 Examination Activities a.
Scope The NRC examiners administered the operating test (JPMs and dynamic scenarios)
during the week of May 17,1999. Station instructors administered the written examination on May 21,1999. The tests were administered using the guidance i
prescribed in sections ES-302 and ES-402 of NUREG 1021.
b.
Observations and Findinas Job Performance Measures The examiners noted that the applicants used good self-checking techniques. The licentce training staff coordinated the arrival times of the applicants and provided escorts to maintain examination security during the operating test. The licensee's
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simulator staff was timely and accurate in their daily setup and execution of the operating test during the validation and examination weeks.
Dynamic Gimulator Scenarios The applicants performed well during the simulator scenarios. They displayed good self-checking and a thorough knowledge of control room operating requirements. Some individual and generic communication deficiencies were identified. These deficiencies included the failure of applicants to acknowledge important information from other crew members and, in one case, providing incomplete information to other crew members.
Written Examiriation The licensee administered the written examination concurrent with the management exit meeting on the last day of the examination week. The examination administrator accurately fo!! owed the guidelines in NUREG 1021 while administering the written examination. The examination was completed in the allotted four hours.
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Conclusions The applicants were well prepared for the coerating test. In general, they displayed good self-checking and communications practices during the operating test. The facility training staff was well prepared to support the examination process.
05.4 Post Examination Activities a.
Examination Scogg The NRC examiners evaluated individual applicant performance on the operating test and reviewed the licensee's grading of the written examination. The examiners also reviewed post examination comments submitted by the licensee. Examiners followed the guidelines contained in sections ES-303, ES-403, and ES-501, of NUREG 1021.
b.
Observations and Findinos Job Performance Measures Two generic knowledge deficiencies were discovered while grading the candidates'
performance of the operating JPMs:
i Applicants were unable to recognize a valid entry into the station's ernergency j
diesel generator fuel oil tank level technical specifications. The applicants were
given a set of diesel generator fuel oil tank levels obtained from the turbine building operator and asked to determine if the plant had sufficient fue! cil to i
moet the station's technical specifications. The expected answer to the question was that there was insufficient diesel fuel oil available and the shift supervisor needed to enter a technical specification for inadequate diesel fuel oil. All three
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applicants failed to subtract the unusable fuel oil from the tank levels and determined that there was adequate fuel oil.
Applicants were unable to determine the correct bank overlap unit reading under
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certain plant conditions. The applicants were given control rod positions followed by a series of switch manipulations by the reactor operator. They were then asked to provide the bank overicp unit value. All three applicants responded that the value would be zero. The applicants appeared to understand how the equipment worked but lacked the knowledge needed to arrive at the correct answer.
Dynamic Simulator Scenarios The NRC examiners did not discover any significant or generic weaknesses during the review of the operating test results.
Written Examination There were nineteen questions that were answered incorrectly by more than 50% of the applicants. These questions were considered potential generic knowledge deficiencies and are provided to the Prairie Island training staff for consideration and implementation into the Systematic Approach to Training-based program. The examiners considered these generic deficiencies as potential because of the small sampling size (3 applicants).
Question #
Knowledoe Weakness
- 4 Required communications practices when acknowledging directions given to operators outside the control room.
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- 6 Maximum allowed deviation between steam generator pressures
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during normal operation, j
- 10 Application of ALARA principles.
- 19 Operation of DC Hold power during normal conditions and during a reactor trip.
- 32 Conditions that will generate annunciator 47013-0507,
" COMPUTER ALARM ROD DEVIATION / SEQUENCING."
- 35 incore instrumentation system response to a 30'F rise in ambient temperature around the reference junction boxes.
- 36 Indications used to monitor a reactor coolant system cooldown under natural circulation conditions.
- 45 Control of steam generator pressure from the hot shutdown panel.
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- 51 Auxiliary feedwater system response to a trip of 12 main feedwater pump concurrent with a loss of buses 11 and 12.
- 57 The response of the reactor coolant drain tank system to a system Hl/LO tank level with pumps and valves in automatic.
- 63 Response of the fire protection system when flow is initiated in the system with the 121 motor driven fire pump disabled.
- 64 The status of the containment vacuum breakers following a manualinitiation of safety injection.
- 67 Expected RCS conditions as natural circulation develops following a loss of power to the reactor coolant pumps.
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- 82 Conditions that may lead to pressurized thermal shock during a j
steam generator tube rupture event.
- 8B Results of a white bus (iii) power failure.
- 90 Response of the 12 charging pump under certain conditions when the control switch at the hot shutdown panelis taken to LOCAL.
- 92 Effect of placing the cation demineralizer in service after a fuel pin begins leaking.
- 98 The best method to r,cilapse an apparent void in the RCS when a natural circulation cooldown is occurring.
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- 100 The basis for maintaining a minimum feed flow of 40 gpm to each
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steam generator following an uncontrolled depressurization of both steam generators.
The licensee submitted nine post examination comments which were reviewed by the NRC examiners. The licensee's comments and NRC resolution of the comments are detailed in Enclosure 2, " Facility Post Written Examination Comments and NRC Resolution." Six comments were accepted, resulting in four questions being deleted and answer key changes for two questions. The comments for the remaining three questions were not accepted.
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Conclu_sions The relatively high number of post examination comments (9) on the written examination indicated a deficiency in developing a written examination that conformed to the guidelines of NUREG 1021. The potential generic knowledge deficiencies from the operatiry test and written examinction were provided as feedback for the licensee's training 3rograrn.
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O5.5 Simulator Fidelity The simulator performed well during the validation week, during the operating JPM test and during the dynamic simulator scenarios with no noted deficiencies. This was (
documented in Enclosure 3, Simulation Facility Report.
V. Manaagment Meetinas
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X1 Exit Meetina Summary The chief examiner presented the examination team's observations and findings to members of the licensee's management on May 21,1999. The licensee acknowledged the findings presented and indicated that no proprietary information had been identified during the examination or at the exit meeting.
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PARTIAL LIST OF PERSONS CONTACTED Licensee T. Amundson, General buperintendent, Engineering D. Cedergren, RO Class Coordinator M. Gardzinski, Simulator Engineer
' J. Lash, instructor B. Mather, Training Shift Manager D. Schuelke, Plant Manager J. Sorensen, NSP Site General Manager D. Westphal, Operations Training Superintendent NRC S. Ray, Senior Resident inspector, Prairie Island Nuclear Plant INSPECTION PROCEDURES USED IP 71707,PPlant Operations"
!TEMS OPENED, CLOSED, AND DISCUSSED
~Ooened None Closed None Discussed None
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se LIST OF ACRONYMS USED CFR Code of Federal Regulations DRS Division of Reactor Safety IP Inspection Procedure
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NRC Nuclear Regulatory Commission OL.
Operator Licensing RO
. Reactor Operator
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Enclosure 2 Facility Post Written Examination Comments and NRC Resolution 1.
Question #4 Which of the following describes the correct..; age of repeat back when directing an APEO in the field to perform valve manipulations?
The APEO should...
a.
provide verbatim repeat back of orders.
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repeat back paraphrasing the orders.
c.
reperst back only that portion of the orders that are NOT completely understood.
d.
repest back the orders only if the directions were provided over the phone or Plant Page system.
Answer; b.
Comment:
The question asks for the correct usage of repeat backs when directing the APEO to perform valve manipulations in the field. SAWI 3.15.6, " Site Communications Standard'
requires a paraphrased repeat back of communications of operational significance.
The question did not ask the minimum standard for communicating. The minimum standard is a paraphrased repeat-back. However, a verbatim repeat-back exceeds the requirements of a paraphrased wording. Additionally, it is a common practice to use verbatim repeat backs when acknowledging orders to position valves.
The facility recommends that both answers (a) and (b) be accepted as correct choices.
NRC Resolution:
The station's communications procedure, SAWI 3.15.6, Section 6.1, General Requirements, stated that for operationally significant communications: "The receiver SHALL acknowledge receipt by a paraphrased repeat back of the received information." (Bold face words were bold face in the procedure.) The procedure had no mention or allowance for verbatim repeat backs. During the review of the written examination, licensee instructors rejected distractor a. as a correct answer by stating that, " Verbatim repetition is NOT desired since this may not indicate understanding of directions." The facility's instructors were unable to demonstrate that a. should be i
accepted as a correct answer. Answer choice b. was retained as the only correct answer. No change was made to the answer key.
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Question #5 Throttle valve SI-15-7 was adjusted and requires independent verification (IV).
You may IV this valve by observing..
a.
and counting the number of tums while CLOSING the valve fully ar.a then REOPENING the valve the same number of tums independent of tho initial verifier.
b.
- and counting the number of tums while this valve is OPENED fully and then RECLOSED the same number of tums by the initial verifier.
c.
proper system flow through the valve, d.
. proper stem position on the valve.
' Answer: c.
Comment:
The question asks for the correct method of performing an independent verification of a specific valve. 5AWI 3.10.1, " Methods of Performing Independent Verification" provides
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generic guidance for throttled valves, however the specific method of independently verifying SI-15-7 is not done per any of the distracters.
The actual method of positioning and independent verification of this valve is completed
- under a work order issued by Engineenng (last done under WO# 9600886) when the acceptance criteria of surveillance procedure (SP1092A) is not met. Once positioned, it is sealed and the operator independently verifies the position of this valve by ensuring a block and sealinstalled on the valve as shown on Checklist C1.1.18-1; p.12 of 17.. IV of this valve is an exception to 5AWI 3.10.1 and is controlled under the Work Order.
The answer key lists " proper system flow through the valve" as the answer. There is no flow measurement device, for a person independently verifying the position, to accomplish this (Flow diagrams X-HIAW-1-44 and 45 show only a total cold leg flow i
measurement as seen in the control room on flow indicator F1-925.). 5AWI 3.10.1
wams, per step 6.1.14.c, that " flow does not necessarily prove that flow is going to the correct location".
The facility recommends that this question be deleted, as there is no correct answer.
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NRC Resolution:
An originally submitted question was replaced during the examination validation week with this question. Based on the licensee's procedures, it appeared that answer c. was j
the correct answer. The examiners reviewed the Si surveillance procedure performed under_ work order 9600886, and determined that this valve was not positioned using the
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generic guidance for independently verifying throttle valves. This valve was throttled closed from the full open position until the correct flow was achieved, and then it was blocked and tagged in that position. The flow instruments used to verify correct flow were subsequently removed and the IV only verificd that the block and seal was installed. The comment was accepted. None of the answer choices was completely correct regarding the independent verification of valve SI-15-7. The question was deleted from the examination and the answer key amended.
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Question #32 Which of the following sets of conditions will generate annunciator 47013-0507
" COMPUTER ALARM ROD DEVIATION / SEQUENCING"?
B_qdj_s Bank D Steo counter Rod G-3 RPI a.
Moving 20 Steps 0 Steps b.
Stationary 180 Steps 194 Steps c.
Moving 190 Steps 204 Steps d.
Stationary 228 Steps 214 Steps Answer: b.
Comment:
The question asks about knowledge of annunciators. A recent Tech Spec revision (T.S. 3.10.F) required a change to the setpoint of the alarm. Previously, knowledge of whether the rods were moving or stationary was neeoed to ascertain when a rod deviation would occur.
The annunciator alarms, with the group step counter between 30 and 215, at a >12 step deviation. Below 30 and above 215 steps the alarm comes in with a >24 step deviation.
Recently revised Surveillance Procedure SP 1319 states that when the rod-to-bank deviation limit is exceeded. "RPI to bank demand difference greater than 12 steps with rod bank position between 30 and 215 steps." Because rod motion is no longer a concem, the choices for this question had three answers with a rod deviation of 14 steps. Two choices were in the 30 to 215 range. With multiple choices appearing to the candidates as the same answer (14 steps) they may have eliminated the choices because they were, as far as the annunciator goes, the same answer. Two candidates chose a greater deviation of 20 steps (selection a).
The facility recommends that this question be deleted from the test, due to the fact that the question was constructed with two selections being exactly the same.
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NRC Resolution:
s The licensee stated that because rod motion no longer has an input to this alarm, answers b. and c. were correct and essentially identical. The licensee agreed that distractors a. and d. were incorrect. The NRC's policy was to delete a question when there was three correct answers. Since there was only two correct answers, the
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question was a valid question. The facility's comment was rejected and the answer key was amended to accept both b. and c. as correct answers.
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Question #40 Which of the following can be used as an indication of proper Hydrogen Recombiner operation following a LOCA event in which the containment hydrogen concentration was initially determined to be 2.8%7 (Assume all other containment parameters remain constant for the period.)
a.
Test Thermocouple temperatures will indicate a ramped decrease below 625'F with constant power input.
b.
Test Thermocouple temperatures will indicate a ramped increase above 1225'F with constant power input.
c.
Recombiner power output will increase from its initial setting without operation of the Pwr Adjust potentiometer.
d.
Recombiner power output will decrease from its initial setting without operation of the Pwr Adjust potentiometer.
Answer; b.
- Comment:
The answer key lists (b) as the correct choice. The given reference, C19.8 does state that a ramp change in temperature will occur. The problem with answer (b) is that it states that there will be a " ramped increase". A ramp change and a ramped increase are not the same. C19.8 Figure 2, shows that there is a ramp change as the temperature nears the recombination temperature of 1225'F. In other words, the ramp rate changes from a steep increase to a lesser rate at the higher temperature.
Surveillance Procedure SP 1255 substantiates this. Steps 7.1.6 through 7.1.12 have the operator heatup the recombiner. Step 7.1.12 states, " temperature should stabilize at approximately 1225'F within 3 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />." This is affirmed in step 7.1.13.
Because none of the choices adequately answer this question, the facility recommends that this question be deleted from the test.
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NRC Resolution:
NRC examination reviewers agreed that there were no answers that correctly responded to the question stem. The post exam comment was accepted. The question was deleted from the examination and the answer key amended to reflect the deletion of the I
question.
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Question #51 Given the following conditions on Unit 1:
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- Reactor power 65%
- Control power to 12 Main FW pump was iost
- Immediately thereafter buses 11 and '2 were lost due to brenker faults
- Both SG NR levels decreased to 20%
What is the status of the AFW Pumps IMMEDIATELY following this event?
11 TD AFW Pumo 12 MD AFW Pumo a.
Stopped Stopped b.
Stopped Running c.
Running Stopped
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d.
Running Running Answer c.
Comment:
This question asks about the cause-effect relationship between main and auxiliary feedwater. Due to a new ciiverse scram system (AMSAC/ DSS), the response of the J
plant will be different.
The answer key lists (c.) as the answer, which is correct before the recent modification.
However, AMSAC/D3S will actuate due to the loss of busses 11 and 12, which will trip
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both 11 and 12 RCPs. The RCPs will generate a protection signal which will start both j
AFW pumps.
The facility recommends changing the answer key from (c.) to (d.)
NRC Resolution:
The post exam comment was accepted. The answer key was amended to accept only d. as the correct answer.
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Question #56 Given the following conditions on Unit 1:
A release is in progress from the 121 ADT Monitor Tank using the
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programmable controller Power is subsequently lost to bus 13 and the bus is deenergized
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What is the effect on the release? The release will...
a.
continue as normal.
b.
continue but requires control directly from the programmable controller using the programming panel or l&C laptop computer.
c.
terminate due to loss of power to the control room instrument for R-18 radwaste liquid release radiation monitor.
d.
terminate due to loss of power to the programmable controller which closes the release valve and stops the pumps.
Answer: d.
Facility Comment:
This question requires the operator to recognize how a liquid release is affected by a sustained loss of non-safeguards bus 13. The reference for this question, (C21.1.2, Section 3.0) makes a generic statement regarding various systems, which does not fully apply to this situation. C21.1.3.1 AOP1 states that only the valves will close.
The power supply to the programmable controller is MCC 3A1, which is powered from bus 13 via bus 310. The power supply to the ADT monitor tank pumps is MCC 1RW2 which is powered from bus 290, which in turn is powered from bus 24 (a unit 2 power supply). The loss of bus 13 will have no effect on the ADT monitor tank pumps.
Additionally, the programmable controller only has control over the release valves and does not have control over the 121 ADT Monitor Tank Pump.
Upon losing power to bus 13, bus 310 will undergo a voltage restoration and be i
immediately repowered from bus 410 via the breaker 3141 bus tie. This restorer power to the programmable controller. The answer key choice of (d) states that the programmable controller will lose power (which is temporary), however it will not have any effect on the " pumps" (ADT monitor tank pumps).
Checklist C21.1-5.1; step 5.1.4, shows that the pump will recirc with the release valve closed.
Confusion with the candidates was apparent, as one asked, "Will backup power supplies
" kick in?" No clarification could be given during the exam. Because the correct answer
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has an incorrect statement, the facility recommends that this question be deleted from the test.
NRC Resolution:
The licersee's post exam comment wac correct. The pumps would not stop. The post exam comment was accepted. The question was deleted and the answer key amended.
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Question #73 Given the following conditions on Unit 1:
- The letdown heat exchanger is out of service
- Excess letdown is in service
- 11 Charging pump is running at minimum speed
- An ATWS has been diagnosed Which of the following values is the highest boration rate that the RO would establish using the normal boration flowpath?
a.
20 gpm b.
15 gpm c.
12 gpm d.
8 gpm Answer; b.
Comment:
This question asks the operator to determine boration rate during an ATWS, given several plant conditions associated with the CVCS as stated above.
Based upon their answers, all students recognized the limit of allowing only 75% boric acid flow of total charging flow. The question asks what flow should be established,
"given the following conditions." The conditions listed, have the running charging pump at minimum speed, and therefore the total charging flow is 16 gpm. With the plant in an abnormal lineup, the candidates did not know if 11 charging pump speed could be increased. The question stem does not suggest that the speed could be increased.
Although the basis for step 4 of FR-S.1 says, "it may be necessary to increase the speed controller of the charging pumps." it also states, "the operator should borate at the maximum rate available based upon current conditions.. " The current conditions given in the question are such that initially the operator would establish 12 gpm boric acid flow (16 gpm x 75%).
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The facility recommends changing the answer key from (b) to (c).
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NRC Resolution:
i The answer choice of b. ecsumed that the RO willincrease charging pump speed to maximize the boration rate. However, there was no information in the question that would lead one to believe that charging pump speed could not be increased which would lead to the 12 gpm boric acid flow. The post exam comment was accepted. The answer key was amended to change the correct answer from b. to c.
8.
Question #82 Which condition may lead to a Pressurized Therma! Shock condition in a steam
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generator tube rupture event?
a.
Si flow to one cold leg is isolated.
b.
Si to reactor vessel valves are opened, c.
The RCP trips in the loop with the ruptured SG d.
Both RCPs have stopped.
Answer; d.
Comment:
The question asks about conditions which may lead to a PTS condition. Selection (d) is true specifically for a SGTR event.
According to the SI system description (B-18A), the Si to reactor vessel valves are normally closed to prevent unnecessary thermal shock to the reactor vessel in the event of a spurious Si actuation. Opening of these valves with Si actuated (choice b) would therefore, lead to a PTS condition from the thermal shock / cooling to the reactor vessel.
This action is outside the EOPs.
The facility recommends that both answers (b) and (d) be accepted as correct choices.
NRC Resolution:
The question specifically asked for a condition (answer) during a steam generator tube rupture. The Si vessel isolation valve concem (distractor d.) applied during a spurious Si actuation. The question stem did not include the condition of a spu'.ious Si actuation.
The comment has not provided sufficient information to require a chrange to the examination answer key. The post exam comment was not accepted, the answer key was not amended.
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Question #85 Given the following conditions on Unit 1:
A reactor trip has occurred due to a loss of offsite power.
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The actions of 1ES-0.1 " REACTOR TRIP RESPONSE' are being
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performed for verifying natural circulation flow.
Which of the following correctly explains why the SG levels are maintained in the normal post-trip control band during natural circulation?
a.
Ensures RCS cooling remains symmetrical.
b.
Prevents voiding from occurring in the reactor vessel head.
c.
Ensures SG tubes are covered to verify natural circulation.
d.
Prevents a complete loss of RCS flow due to voiding in a single loop.
Answer a.
Comment:
The question asks the knowledge found in the basis for an EOP step regarding conditions for natural circulation.
The suggested answer does not answer the qt.estion. The question asks why levels are maintained in a band. The low end of the band is to ensure SG tube coverage and provides symmetric cooling if both SG levels are above the SG tubes. However, the high end of the band is not addressed. The high end of the control band is 40% which prevents having too much mass in the SG in case of a steamline break accident (see references C1.3; Figure C1-39; and USAR Section 14).
Because none of the choices answer the stem of the question, the facility recommends that this question be deleted from the test.
NRC Resolution:
The comment was accepted. The actual reason for maintaining both the high end and low end of SG level was not included in any of the answer choices. The question was deleted and the answer key amended.
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Enclosure 3
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SIMULATION FACILITY REPORT Facility Licensee: Prsirie Island Nuclear Station Facility Licensee Docket Nos: 50-282; 50-306 Operating Tests Administered: May 17-20,1999 The following documents observations made by the NRC examination team during the initial license examinallen. These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of non-compliance with 10 CFR 55.45(b).
These observations do not affect NRC certification or approval of the simulation facility other than to provide information which may be used in future evaluations. No licensee action is required in response to these observations.
During the conduct of the simulator portion of the operating tests, the following items were observed:
ITEM DESCRIPTION 1. None t I AS-ADMINISTERED WRITTEN EXAMINATION PRAIRIE ISLAND EXAMINATION - WEEK OF MAY 17,1999
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U.S. Nuclear Regulatory Commission Site-Specific Written Examination Applicant information mas m nAMINAM ON Name:
Region:
111 Date: 5/21/99 Facility: Prairie Island License Level: RO Reactor Type: W
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Start Time:
Finish Time:
Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent. Examination papers will be collected four hours after the examination starts.
Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
Applicant's Signature Results Examination Value Nk
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Points
Applicant's Score Points Applicant's Grade Percent
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n Rrctor Op::rctor Excmin tion 1.The following is the work schedule for a licensed Reactor Operator on day shift:
5/17 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as Lead 5/18 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as RO and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> holdover 5/19 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as RO 5/20 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as Lead and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> holdover as Lead 5/21 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as RO 5/22 - scheduled 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as RO 5/23 - scheduled 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as Lead At the end of the shift on 5/21, the SS asks if you could stay over an additional two (2) hours as RO. Which of the following describes the guideline for you to continue work?
You...
a. can continue work without any restrictions.
b. will exceed the limit for working in a 24-hour period and must have the approval of the SS.
c. will exceed the limit for working in a 48-hour period and must have the approval of the GSPO.
d. will exceed the limit for working in a 7-day period by the end of the period.
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R cter Opsrctor Extminatien 2.You work the following schedeie:
-.5/06 - RO for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of the shift.
5/07 - RO for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of the shift.
5/08 - RO for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of the shift.
5/09 - OFF 5/10 - OFF 5/11 - Administrative duties NOT associated with on-shift duties.
5/12 - RO for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of the shift due to illness of the assigned RO 5/13 - Administrative duties NOT associated with on-shift duties.
5/14 - Administrative duties NOT associated with on-shift duties.
5/15 - OFF 5/16 - OFF 5/17 - Administrative duties NOT associated with on-shift duties.
5/18 - Administrative duties NOT associated with on-shift duties.
5/19 - Administrative duties NOT associated with on-shift duties.
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5/20 - Administrative duties NOT associated with on-shift duties.
You are reporting to work today (5/21) and assigned as the Unit 1 Lead. As part of your turnover, you must review the Unit 1 Reactor Log since...
a. 5/15 b. 5/12 c. 5/08 -
d. 5/06 i
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Raccter Operstor Extmination 3..Which of the following describes _the requirements concerning operation at the maximum allowable steady-state full-power level as defined by the Unit Operating License?
a. - Power may exceed 100% for a short duration but at NO time will exceed 102%. The j
averago power for an eight hour shift is to be 100% or less.
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b. Average power may exceed 100% for an eight-hour shift, but at NO time will exceed 102%.
c. Power exceeding 100% requires a power reduction ONLY if projected average power for the eight hour shift exceeds 100%.
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. d. ' Average power for an eight-hour shift is to be 102% power or less. If power exceeds 102%,
then power must be reduced to 100% or less.
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R:ccter Op3rctor Extmination 4.Which of the following describes the correct usage of repeat back when directing an APEO in the i
field to perform valve manipulations?
l The APEO should...
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a. provide verbatim repeat back of orders.
b. repeat back paraphrasing the orders.
c. repeat back only that portion of the orders that are NOT completely understood.
d. repeat back the orders only if the directions were provided over the phone or Plant Page system.
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Rocctor Op rctor Examination 5. Throttle valve SI-15-7 was adjusted and requires independent verification.
You may IV this valve by observing...
a. and counting the number of turns while CLOSING the valve fully and then REOPENING the valve the same number of turns independent of the initial verifier.
b. and counting the number of turns while the valve is OPENED fully and then RECLOSED the same number of turns by the initial verifier.
c. proper system flow through the valve.
d. proper stem position on the valve.
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Rrctar Op;rctor Ermin tion 6.While performing SP 1001; Unit Daily Control Room Log, you note the following indications on the 12 steam generator pressure indicators:
1Pl-478 STEADY at 730 psig 1PI-479 OSCILLATING between 650 psig and 720 psig 1PI-483A OSCILLATING between 700 psig and 750 psig According to the log, the maximum deviation allowed between these channels is 70 psig. Based upon this information...
a. channel 1PI-479 does not meet surveillance requirements.
b. channel 1PI-483A does not meet surveillanr9 requirements.
c. all channels do not meet surveillance requirements.
d. all channels meet surveillance requirements.
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R: actor Op;rator Examination 7.The RO has been directed to initiate an Emergency Boration in response to an ATWS condition.
When the switch for MV-32086 Emergency Boration to Charging Pump Suction valve, was taken to OPEN, dual indication was received and then both position indicating lights went out. An APEO dispatched to the MCC for the valve reports NO unusual conditions except it appears the thermal overload device on the breaker is tripped.
What direction should be provided to the APEO concerning the Thermal Overload?
The APEO should...
a. place the breaker in the OFF position and call the Electrical Department for assistance.
b. obtain Electrical Department assistance before attempting to reset the Thermal Overload device.
c. place the breaker in the OFF position before attempting to reset the Thermal Overload device.
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d. depress the Thermal Overload RESET pushbutton to allow valve operation.
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Reactor Operctor Excmination 8.Which of the following is NOT a responsibility of the Fuel Handling Operator in the Control Room during movement of fuel?
a. Maintain the 1/M Plot during fuel loading in the core following total off-load.
b. Sign off the Fuel Transfer Log following relocation for each fuel assembly.
c. Provide concurrence for lowering the manipulator crane onto any fuel assembly.
i d. Monitor RCS temperatures every 15 minutes if RHR is stopped.
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Racctor Op3rctor Extmination 9. An operator has the following exposure history for the past year up to the last day of that year; i
320 mrem
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Committed Effective Dose Equivalent (CEDE)
40 mrem
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Shallow Dose Equivalent (SDE)
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25 mrem Committed Dose Equivalent (CDE)
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35 mrem On the last day of the year, the operator was required to make two entries into containment:
Entry 1:
Gamma dose - 25 mrem; Neutron dose - 20 mrem Entry 2:
Gamma dose - 55 mrem After the containment entries, the operator's available margin based on the Prairie Island Administrative Dose Guideline is...
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a. 4560 mrem.
b. 4480 mrem.
c. 1540 mrem.
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d. 1500 mrem.
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Reactor Op3rator Examination 10. Work is being performed in a specific area in the RCA. The attached Sample Survey shows this area. The equipment being worked on is located at point "B". Access through the corridor ("E")
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takes 5 minutes in one direction. Access through corridor ("F") takes 3 minutes in one direction.
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The dose rate is constant throughout the length of both corridors. TWO separate tasks are to be performed on the equipment and must be performed 15 minutes apart.
I Which of the following applies the principle of ALARA7 The time between tasks should be spent at point...
I a. "A."
b. "B."
c. "C."
d. "D."
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Racctor Op:rstar Excmination 11. A reacte +:ip first-out annunciator has LIT, requiring entry to E-0 " REACTOR TRIP OR 9AFETY INJECTIC:~ " Upon entry to E-0, the operator notes:
- Reactor trip breaker position indication is lost (prior to determining position)
- All Nuclear Instrumentation indicate 20% power with flux decreasing
- Only Shutdown Bank B rod bottom lights and Rod Position Indicators read ZERO What is the next action required of the operator in accordance with E-07 a. Proceed with the next step of E-0.
b. Manually trip the reactor.
c. Dispatch an operator to locally open the reactor trip breakers.
d. Perform immediate actions of FR-S.1," RESPONSE TO NUCLEAR POWER GENERATION /AM/S."
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Reactor Operator Examination 12.Given the following initial conditions:
- Unit 1 draindown of the RCS is complete
- Primary Manways for the 11 SG nave been removed
- Manways for 12 SG are still installed Subsequently the following occur:
- a nozzle dam is installed on the hot leg side of the 11 SG
- RHR cooling is lost What is the effect of the nozzle dam installation?
a. RCS inventory loss is minimized.
b. The amount of time prior to core uncovery is reduced.
c. RCS cooling capacity is maximized.
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d. The amount of time prior to core boiling is prolonged.
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Rsactor Op rctor Examination I
13. Unit 1 tripped and Safety injection was actuated due to containment conditions. Following the
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opening of the main generator output breakers, all AC power was lost at the Unit. The crew initiated actions of ECA-0.0 " Loss of All Safeguards AC Power." While performing these actions the following conditions were noted on the Critical Safety Functions Status Trees:
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- ORANGE path condition exists for the Core Cooling critical safety function
- RED path condition exists for the Containment critical safety function What are the appropriate actions for these conditions?
a. The actions of ECA-0.0 " Loss of AC Power" are continued.
b. The actions of ECA-0.0 " Loss of AC Power" are continued, and concurrently the actions of FR-Z.1, " Response To High Containment Pressure" are initiated.
c. Transition is made to perform the actions of FR-Z.1," Response To,High Containment Pressure"immediately and then transition to FR-C.2, " Response To Degraded Core Coc!ing".
d. Transition is made to perform the actions of FR-Z.1," Response To High Containment Pressure"immediately.
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Recctor Op rator Examination 14. A problem with the rod control system requires checking several rod bank circuits associated with the 1BD Power Cabinet. During this time, affected rod banks are to be placed on DC Hold Power.
Which of the following describes the operation of DC Hold power, including response to a reactor trip?
a. One shutdown OR one control bank group can be placed on DC HOLD at a time. These rods will drop only when the control switch is taken from HOLD to OFF at the DC Hold cabinet.
b. One shutdown bank group AND one contro! Dank group can be placed on DC HOLD at a time. These rods will drop only when the control switch is taken from HOLD to OFF at the IBD Power Cabinet.
c. One shutdown OR one control rod bank group can be placed on DC HOLD at a time. These rods will automatically drop when the reactor trip breakers open.
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d. One shutdown bank group AND one control bank group can be placed on DC HOLD at a time. These rods will autoniatically drop when the reactor trip breakers open.
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Page 14 of 100 R:ccter Op::rctor Extmination-15.Given the following conditions on Unit 1:
- RCS pressure - 340 psig
- RCS temperature - 330*F
- - PRZR level - 23%
- 12 SG pressure - 50 psig
- No RCPs are running What would be the initial response of RCS temperature and pressure when the 12 RCP is started?
'RCS TEMPERATURE-PRESSURE a. LOWERS LOWERS
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b.- LOWERS STAYS THE SAME
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, c. RISES LOWERS d. RISES RISES
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Rs cter Op;retor Extminntion
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16.Given the following conditions on Unit 1:
- Unit is shutdown at 547'F
- 11 RCP has just been stopped
- The Lead has placed the controller for CV-31224 Loop A PRZR Spray in MANUAL and shut the valve Why did the operator CLOSE CV-31224?
a. Prevent spray flow from bypassing the PRZR.
- b,' Prevent bypass flow from affecting RCS pressure.
c. Prevent rapid depressurization of the RCS during 11 RCP coast down.
d,. Minimize differential temperature between the PRZR and spray flow, l
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Rxcter Op::ratar Extmination E
.17.Given the following conditions on Unit 2:
- RCS pressure - 1600 psig
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- RCS temperature.400*F -
- Both RCPs are running
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What is the consequence if the 21/22 RCP #1 Seal Bypass Valve were inadvertently opened?
a. Seal return line relief valve willlift.
b. Seal damage may result from cocking of the RCP #2 seal.
c. RCS pressure will rapidly drop due to increased leakoff flow.
d. RCS flow through the Thermal Barrier Heat Exchanger may exceed design parameters.
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Rocctor Op;rator Excmination 18.Given the following conditions on Unit 2:
- Reactor power-16%
- 4.16 KV busses 24,23 and 22 are aligned to their normal"At-Power" sources
- 4.16 KV bus 21 is aligned to its " Shutdown" source What will be RCP status if the 86/2GT lockout relay actuates for the 2ivi Trancformer due to a Neutral Ground Cunent (51G) relay actuation?
a. Both RCPs would be tripped, b. The 21 RCP would be tripped and the 22 RCP would remain running.
c. The 21 RCP would remain running and the 22 RCP would be tripped, d. Both RCPs would remain running.
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Raccter Operttar Extminntion 19.Given the following conditions on Unit 1:
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- Reactor power 50%
- The RO has just completed a dilution at the rate of 40 ppm
- Boric Acid flow, HC-110, is set at 20%
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- The RO restores the makeup system to AUTO
- VCT level dropped to 17% and the system responds to raise VCT level What is the resulting boron concentration from the reactor makeup system?
a.
610 ppm b.
900 ppm c. 1000 ppm d. 1465 ppm
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Ranctor Op::rctor Ex minttien 20.Given the following conditions on Unit 1:
- Reactor power-50%
- PRZR level - at program level
- 11 Charging Pump is running with its controller in MANUAL
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- Charging and letdown are balanced What is the effect on the plant if the charging pump controller is maintained in MANUAL as power is raised to 100%7:
a. PRZR level will increase.
b. Mass of coolant in the RCS willincrease.
c. VCT level will decrease.
d. Charging flow will decrease.
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Raccter Op;rctor Exrminition 21.Given the following plant conditions on Unit 1:
- Unit is shutdown with RHR providing shutdown cooling
- The 11 RHR Pump and 11 RHR HX are in service 320 psig
- RCS pressure
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- RCS temperature
- 300* F
- RCS cooldown rate - 45* F/hr 2000 gpm in automatic
- RHR total flow
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Which of the following will occur if the air pressure regulator to CV-31235,11 RHR HX Outlet Flow control, failed such that air was lost to the valve operator?
a. RCS cooldown rate will increase.
b. RHR HX CC Outlet Temperature will decrease.
c. RCS pressure will slowly increase.
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d. RHR flow will decrease.
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Racetor Opsrater Exemination 22.What position (s) for the Boric Acid Storage Tank selector switch aligns the 12 i Boric Acid Storage Tank (BAST) to the Unit 2 Safety injection pumps in the event of a Safety injection on the Unit?
a. TI-T2 ONLY b. -TI-T3 ONLY -
c. TI-T2 and T2-T3 d. T2-T3 and T1-T3
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I Reactor Opnrator Examination 23 Which of the following describes the response of the SI Accumulators on a large break LOCA, l
provided the accumulators are at the minimum level as described in the USAR7 i
a. BOTH Accumulators fully discharging tc the core will fill the outside of the core barrel to the nozzles, the bottom plenum, and fill the core area to the bottom of the hot leg nozzles.
b. BOTH Accumulators fully discharging to the core will fill the outside of the core barrel to the nozzles, the bottom plenum, and cover onc-half of the core.
. c. ONE Accumulator fully discharging to the core will fill the outside of the core barrel to the nozzles, the bottom plenum, and completely cover core. The other Accumulator contents j
spill to the containment floor.
d. ONE Accumulator fully discharging to the core will fill the outside of the core barrel to the nozzles, the bottom picnum, and cover one-half of the core. The other AcLuulator contents spill to the containment floor.
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R3 cter Op3rctor Exrmination 24.Given the following conditions on Unit 1:
- Reactor power - 25%
- RCP sealinjection WAS lost
- C12.1 AOP1," Loss of RCP Seal Injection"is being performed What operation (s) should occur to control and reduce the Component Cooling system temperatures?
a. The operator starts the standby CC pump, starts a standby Cooling Water pump, and fully opens the CC HX Inlet MOV.
b. The standby CC pump automatically starts when the CC HX outlet temperature reaches 105'F.
c. The operator manually throttles open the CC HX Cooling Water inlet MOV.
d. The operator reduces the CC HX outlet temperature setpoint to open the Cooling Water Outlet Control valve.
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Rascter Op:rc.tcr Exrminttien 25. Unit 1 was at 100% steady state power. A load rejection resulted in the following:
- Reactor power - 70%
- ROS Tave - 563*F
- PRZR pressure - 2250 psig
- PRZR level - 47%
What is the current status of the pressurizer pressure control system based on given conditions and assuming no integral signals?
a. Backup heaters are on and variable heaters are fully off, b. Backup heaters are off and variable heaters are fully on.
c. Backup heaters are on and spray valves have modulated open.
d. Spray valves, PORVs are fully open and variable heaters are fully off, t
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R3rctor Operater Exrminstion 26.Given the following conditions on Unit 1:
- Reactor Power - 50%
- PRZR level at program level
- PRZR Level Control Transfer switch is in Position 2-1 (White-Red)
- PRZR level channel LT-427 has failed at its current level Which of the following identifies the change in charging '. mp speed when the plant performs a load change to the power levels listed below?
25% Power 75% Power a.
Increase Decrease b.
Decrease Decrease c.
Increase Remain the Samo
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Racct:r Operctor Extmination
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,27 : Given the following conditions for Unit 2:
- Reactor power - 30%
- The only running MFP tripped.
-The reactor and turbine failed to manually trip
- 21 and 22 RCPs are running Which of the following correctly describes the expected response of the AMSAC/ DSS 7 a. AMSAC/ DSS will NOT actuate since power level was less than the 40%.
' b. Trips the turbine through redundant circuits, tiips the reactor through diverse circuits, and starts both AFW pumps.
c. Trips the turbine through diverse circuits, trips the reactor through r.edundant circuits, and starts only the 12 MD AFW pump, d. Trips the reactor through diverse circuits, indirectly causing a turbine trip and starts both
- AFW pumps.
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R :cter Opercter Excmination 28.Given the following conditions on Unit 1:
- Delta-1 - +5%
-Tcold - 534*F-Thot - 582*F
- PRZR pressure - 2240 psig
- PRZR level - 31%
- RCS loop flows - 99%
How does the setpoint for Oyertemperature Delta-T (OTAT) change when a listed parameter is changed? ' (Consider each change individually)
The OTAT setpoint..
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a. increases if Delta-l reaches 6%.
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b. increases if total reactor flow reaches 102%.
c. decreases if RCS Thot reaches 585*F.
d. decreases if PRZR pressure reaches 2205 psig.
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o Recett:r Op rctor Examinction 29 '.Given the following conditions on Unit 1:
- Excess letdown is in service
- Normalletdown has been isolated
- A spurious Containment isolation 'CI' signal has occurred Which of the following is the effect of the 'CI' signal on the CVCS?
a. RCP seal injection flow is lost, b. Excess letdown flow is diverted to the RCDT.
c. Letdown flow is diverted to the CVCS HUT.
d.- RCP No.1 sealleakoff flow is directed to the PRT via a relief valve.
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o R::ctcr Op rster Excminttion 30.Given the following conditions on Unit 2:
-The unit tripped from 100% power
- A Safety injection signal was generated due to low pressurizer pressure
- Safety injection has been reset
- Plant parameters are stable at 1750 psig with all ESF equipment operating When the ESF equipment is stopped, how is another automatic Si actuation prevented from restarting the equipment?
a. By placing the ESF equipment controls in Pull-Out.
b. All inputs are blocked by placing the " Block-Reset" switches to BLOCK.
c. By placing both Trains of the affected ESF subsystem in manual, d. The P-4 interlock is active as long as the reactor trip breakers remain open.
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Rs cter Opsrctcr Extminntion l
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- 31.The following conditions exist on Unit 1:
- The yellow containment pressure channel has failed low.
- The operator directs placing the toggle switch for bistable PC950B, Hi-Hi
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_ Containment Spray, in the 'UP' position
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What effect will this action have on the Containment Spray actuation signal logic?
a. Energizes an input relay, a Containment Spray actuation signal is NOT generated, b. Energizes an input relay, a Containment Spray actuation signal is generated.
c. Deenergizes an input relay, a Containment Spray actuation signal is NOT generated.
d. Deenergizes an input relay, a Containment Spray actuation signal is generated.
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b R3rctor Op::rctor Excmination I
32.Which of the following sets of conditions will generate annunciator 47013-0507 " COMPUTER ALARM ROD DEVIATION / SEQUENCING"?
Rods Bank D Step Counter Rod G-3 RPI l
a. Moving 20 Steps 0 Steps b. Stationary 180 Steps 194 Steps c. Moving -
190 Steps 204 Steps d. Stationary 228 Steps 214 Steps
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Rsecter Op:rstor Ex min: tion 33.Given the following conditions on Unit 1:
- Reactor power - 20%
- N35 Intermediate Range Loss of Comp Volt" annunciator is in alarm
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- LOSS OF COMPENSATING VOLT lamp is lit at the N-35 drawer What is the consequence of continued operations if NO action is taken for the N-35 channel?
a On a load increase above 20%, the reactor will trip.
b. On a power reduction below 10%, the SR instruments will re-energize.
c. On a subsequent startup, P-6 energizes sooner than normal.
d On a reactor trip, P-6 will de-energize sooner than expected.
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i Rosctor Opsrstor Excmint. tion 34.Given the following conditions on Unit 1:
- Reactor power - 100%
- During performance of the last manual calorimetric the feedwater temperature points utilized were reading 10*F HIGHER than actual feedwater temperature What is the status of the power range indications?
Indicated power is...
a. LESS THAN actual power; therefore, power range instruments are set non-conservatively.
b. LESS THAN actual power; therefore, power range instruments are set conservatively.
c. GREATER THAN actual power; therefore, power range instruments are set
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d. GREATER THAN actual power; therefore, power range instruments are set conservatively, i
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Rxctor Dp::rc. tor Extminati:::n 35.Given the following conditions on Unit 1:
- Reactor power - 100%
- Hottest incore Thermocouple temperature - 618 F-Temperature in the area of the Reference Junction boxes for the thermocouples rises 30*F over the shift.
- The primary system parameters remain constant over the shift.
Which of the following correctly describes how the ICCM core exit thermocouple reading is affected over the shift?
It will...
a. read higher due to higher voltage differential between the metals at the cold junction.
b. read lower due to lowered voltage differential between the metals at the hot junction.
c. remain the same because the reference junction boxes are maintained at 200'F by heaters, d. remain the same because RTD temperature compensation is provided at the reference junction boxes.
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Racetor Opuctor Extmination i
36.Given the following information on Unit 1:
- A reactor trip has occurred from 100% power
- A problem in the switchyard has resulted in a loss of offsite power
- RCS cooldown is in progress using 1ES-0.3A, Natural Circulation Cooldown with CRDM Fans l
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What pair of indications are used to monitor the RCS cooldown?
a. Upper head thermocouples and Core Exit Thermocouples.
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b. Thot and Core Exit Thermocouples.
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c. Tcold and SG pressures.
d. Upper head thermocouples and SG pressures, i
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Rrcter Op;rcter Ex:minction 37.Given the following conditions on Unit 1 during the summer:
- Reactor power 90%
- Containment Fan Coil Units (FCU) 11 and 14 are running in FAST
- Containment FCU 13 is running in SLOW
- Containment FCU 12 is shutdown with its control switch in OFF What is the response for the FCUs if a spurious Train A SI ('S') signal is generated?
a. FCUs 11 and 13 will be operating in slow speed with Cooling Water supplied, FCU 14 will i
be operating in fast speed with Chilled Water supplied, and FCU 12 will be stopped.
i b. FCUs 12 and 14 will be operating in slow speed with Cooling Water supplied, FCU 11 will be operating in fast speed with Chilled Water supplied, and FCU 13 will be operating in slow
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speed with Chilled Water supplied.
c. FCUs 11 and 13 will be operating in fast with Chilled Water supplied, FCU 14 will be operating in fast speed with Cooling Water supplied, and FCU 12 will be stopped.
d. FCUs 12 and 14 will be operating in fast speed with Chilled Water supplied, FCU 11 will be operating in fast : peed with Cooling Water supplied, and FCU 13 will be operating in slow speed with Cooling Water supplied.
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R::cter Opsrcter Excmin:.ticn 38 -.Given the following conditions on Unit 1 in summer:
' Reactor power 1007o
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- Containment Fan Coil Units (FCU) 11 and 13 are running in FAST
- Containment Fan Coil Units 12 and 14 are running in SLOW
- Reactor Vessel Gap Cooling Fan 11 is running
- The Neutron Datector Booster Fan is running The RO was directed to stop the 13 FCU and has initially shifted the 13 FCU to SLOW speed.
Which of the following desenbes the sequence of the remaining steps to be taken prior to repositioning dampers for 13 FCU and stopping 13 FCU?
'
a. 14 FCU is taken to FAST and then its dampers aligned to the gap.
b. The dampers for 14 FCU are aligned to the gap and then the fan.is,taken to FAST.
c. The dampers for 14 FCU are aligned to the gap.
d.14 FCU is taken to OFF and then its dampers aligned to the gap.
Page 38 of 100
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l Rrctor Op::rcter Extminttion
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39.Given the following conditions on Unit 1:
- The reactor has been tripped and Si actuated for a LOCA
- Containment pressure is 18 psig and rising
- The Lead has been directed to manually initiate Containment Spray (CS)-
What is the system response if the operator takes ONE of the Containment Spray Actuation switches to ACTUATE 7 a. Both trains of CS start immediately, b. Only the train of CS associated with the switch operated will start immediately, the other train of CS will start when containment pressure rises to 23 psig.
c. No changes occur in either train of CS until containment pressure rises to 23 psig when both trains of CS start.
d. No changes occur in either train of CS, and the CS system will be blocked from starting when containment pressure rises to 23 psig.
Page 39 of 100
Reactor Oparater Examination
40.Which of the following can be used as indication of proper Hydrogen Recombiner operation following a LOCA event in which the containment hydrogen concentration was initially determined to be 2.8%? (Assume all other containment parameters remain constant for the period.)
a. Test Thermocouple temperatures will indicate a ramped decrease below 625*F with constant power input.
b. Test Thermocouple temperatures willindicate a ramped increase above 1125*F with constant power input.
c. Recombiner power output will increase from its initial setting without operation of the Pwr Adjust potentiometer.
d. Recombiner power output will decrease from its initial setting without operation of the Pwr Adjust potentiometer.
CORN 9h tcss (hEhE J
Page 40 of 100
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Rrcter Op;rctar Exemination
- 41.Given the fcilowing conditions on Unit 2:
- Unit has achieved Cold Shutdown
- A containment purge short term release has been initiated in accordance with C19.2 CONTAINMENT SYSTEM VENTILATION Which signal below would NOT result in automatic isolation of the Containment Purge system with the purge in progress?
a. Manual actuation of Sl on Unit 1.
b. Manual actuation of Cl on Unit 2.
c. Failure high of R-25, Spent Fuel Pool Air monitor.
d. Failure high of 2R-22, Shield Building Vent Stack monitor.
a./
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Page 41 of 100 L-
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Rocctor Op::re. tor Excminntion 42 Given the following conditions:
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- Unit 1 - MODE 5 going into refueling outage
- Unit 2 - 20% power with load increase in progress
- Spent Fuel Pool Cooling is in service with SFP 121 SFP Cooling Pump j
and 121 SFP Heat Exchanger in service.
j j
What actions have to be taken to initiate purification of the Unit 1 RWST7 a. The 122 SFP Cooling Pump must be placed in service and the 121 SFP Cooling Pump
]
stopped prior to realignment of the SFP purification loop to the RWST and start of the j
RWST purification pump.
b. The 122 SFP Heat Exchanger must be placed in service and the 121 SFP Heat Exchanger
)
taken out of service prior to realignment of the SFP purification loop to the RWST and start of the RWST purification pump.
j c. Both the'122 SFP Cooling Pump and the 122 SFP Heat Exchanger must be placed in service, and the 121 SFP Cooling Pu.np stopped and the 121 SFP Heat Exchanger taken i
out of service, prior to realignment of the SFP purification loop to the RWST and start of the
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RWST purification pump.
j d. The SFP purification loop is isolated from the SFP Cooling loop, realigned to the RWST, f
and the RWST purification pump started.
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Page 42 of 100
I Raccter Opsrator Excmination 43.Which of the following describes the correct action to take in the event the conveyor, carrying a fuel assembly from the reactor side to the SFP side, becomes stuck in the fuel transfer tube?
Attempt should be made to pull the conveyor to the...
]
a. reactor side to provide containment for any radioactive release if the fuel assembly is damaged.
b. reactor side to provide better access to the winch, cable drum and pulley system.
c. SFP side so that the containment penetration can be isolated.
d. SFP side to minimize disruption of Core Alterations.
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Page 43 of 100
Rxctor Opsrator Excminstion
- 44.Given the following conditions on Unit 1:
- A power increase from 20% following a reactor startup has just been initiated
- The controlling 11 SG wide range level instrument fails high 11 SG level will be controlled by...
a. the mismatch between Steam and Feed flow.
b. NR level and Program level Setpoint.
c. an alternate WR level and Reference WR level.
d. Feed Forward signal and Reference WR level.
.
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i Page 44 of 100 L
' Rsector Op:trator Examination i
. 45.How is the control and operation of t.ie SG PORVs accomplished at the HSD Panel?
At the HSD Panel the...
a. local / remote switch is taken to LOCAL and the controller setpoint is adjusted to desired SG pressure, b. local / remote switch is taken to LOCAL and the controller output is adjusted manually.
c. controller is shifted to AUTO and the controller output is increased using the setpoint dial.
' d. controller is shifted to MAN and the controller output is increased using the dial.
.
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Page 45 of 100 i-L
Roccter Op3rctor Excminttion 46.Given the following conditions on Unit 1:
- Reactor power. - 90%
- To allow troubleshooting, steam dump control is in the STEAM PRESSURE Mode
- Current steam header pressure of 749 psig
- Steam Dump controller is in AUTO with the setpoint dial set to control at 835 psig
- The RO initiates a load reduction to 55% power What will happen to steam dumps during the load reduction?
Steam Dumps...
a. begin to open at approximately 80% power.
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b. begin to open at approximately 70% power.
c. begin to open at approximately 60% power.
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d. do NOT open.
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Page 46 of 100
Rsccter Op:rctor Extminttion 47.Given the following:
- 11 SG pressure is 500 psig
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- 11 Condensate Pump is in operation
- 11 & 12 MFPs are not running
- 11 Main Feed Pump discharge valve is open Which of the following will result in establishieg flow to the 11 SG using the Feed Regulating Bypass valve.
a. Condensate pump recirculation valve, CV-31122 must remain open.
b. Depressurize 11 SG.
c. Start 12 Condensate pump.
d. Feed Pump Recirculation valve, CV-31875 must be open.
.
Page 47 of 100
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i Rscctor Op:rctor Excminttion
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i 48. Given the following initial conditions for Unit 1:
- Reactor power - 15%
- Main feed in operation using the main feedwater regulating valves
- SG levels are stable at 42% (NR)
When Steam Dump control was selected to Tavg Mode, one steam dump valve failed full open.
The valve was isolated shoitly thereafter. Following isolation of steam flow, key parameters are:
- Pressurizer pressure - reached a minimum of 1825 psig and is beginning to rise I
- RCS Tavg - 542 *F
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- SG levels peaked at 62% and are lowering
)
- Steam pressure - 870 psig
- Main feed regulating valves are closed
Why did feedwater isolate automatically?
i a. The Si signal coincident with Low Tavg.
b. Hi Steam Flow coincident with Low Tavg.
c. Water level exceeded the Hi-Hi SG level setpoint, d. The reactor trip coincident with Low Tavg.
Page 48 of 100
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Rsrctor Operctor Excminction 49.Given the following conditions on Unit 1:
- Reactor power 75%
- 12 and 13 Condensate Pumps are running
- 11 Condensate Pump is in standby Which of the following automatically and directly occurs if tha 13 Condensate Pump trips and the 11 Condensate Pump fails to start?
a. The 11 Main Feedwater Pump trips.
b. The 12 Main Feedwater Pump trips.
c. Both Main Feedwater Pumps trip.
d. The main turbine and both Main Feedwater Pumps trip.
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Page 49 of 100
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Rssctor Op:rstor Extrninetion 50.Given the following conditions on Unit 1:
- Reactor power 100%
- 11 TD AFWP Accumulator Lo Air Press" alarm (47010-0105) has actuated j
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- Air is found to be leakinD rom a large crack in the 11 TD AFW Pump accumulator
)
f What is the appropriate action to be performed for this situation?
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a. The AFW Pump is stopped by shutting both Steam Supply Valves MV-32016 and 32017.
s b. The AFW Pump is stopped by locally pressing the trip device at the Trip / Throttle valve CV-31059.
c. An operator should be stationed locally at the 11 AFWP Emergency Start (3-way) valve AF-292-1 since the AFW Pump CANNOT be started normally.
d. An operator should be stationed at the Hot Shutdown area and take 11 TD AFWP CV-31998, LOCAL / REMOTE switch to LOCAL since the AFW Pump CANNOT be started from the control room.
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Page 50 of 100 l
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Rsrcter Op:rcter Excminttion 51. Given the following conditions' on Unit 1:
- Reactor power 65%
- Control power to 12 Main FW Pump was lost
- Immediately thereafter buses 11 and 12 were lost due to breaker faults
- Both NR SG levels decreased to 20%
What is the status of the AFW Pumps IMMEDIATELY following this event?
11 TD AFW Pump 12 MD AFW Pumo a.. Stopped Stopped b. Stopped Running c. Running Stopped d. Running Running
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Page 51 of 100
Reactor Op3rcter Excmin tion s
52.Given the following conditions on Unit 2:
- Reactor power - 100%
- The electrical system was in its normal alignment
- The following annunciators alarm in the control room:
- 2R Lockout"
"
- 2RY Sudden Pressure Trip"
"
Which of the following is the response to the 2RY transformer failure?
a. Bus 25 and Bus 26 will be powered via the CT transformers and D5 will be running unloaded, b. Bus 25 and Bus 26 will be powered via the CT transformers and no DG will be running.
{
c. Bus 25 will be powered from D5 diesel generator.
d. Bus 25 and Bus 26 will be powered from their respective diesel generators.
i Page 52 of 100
Rxcter Op3rcter Extminstien
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53.The APEO performing rounds reports he suspects there is a failure of the 11 Battery Fuse.
- What is the indication that would be checked to substantiate a fuse failure?
a. 11 Battery discharge ammeter indicates negative current flow.
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b. 1.1 Battery Charger voltmeter reading significantly higher than the 11 Battery voltmeter.
c. ERCS " Battery Undervoltage" alarm is actuated, d. Annunciator "DC System Trouble"is in alarmed condition.
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i Page 53 of 100 i
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t Rocctor Opsrctor Exrminction 54.Which of the following conditions will result if Train 'A' DC power is lost on Unit 17 a. D1 will start on an automatic start signal but field flash will fail.
b. DC control pcwer for 4 KV buses 11 and 12 will swap to its standby source DC panel 21.
c. Unit 1 main generator output breakers open causing a turbine trip.
d. DC control power for 4 KV bus 15 will swap to its standby source DC panel 21.
.
i a
Page 54 of 100
Rsrcter Op3rcter Excmination
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55.Given the following conditions on Unit 1:
- D1 Diesel Generator (DG) was unning paralleled to its associated 4.16 KV Bus for its Monthly Load Test
.The RO was adjusting load and voltage when the Governor Speed Control switch sticks in the j
LOWER position i
If NO operator action is taken, what will be the response of Diand its output breaker to this condition?
The DG output breaker will open on...
I a. underfrequency and D1 will trip.
i b. reverse power and D1 will trip.
c. loss of field but D1 will continue to run.
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d. overcurrent but D1 will continue to run.
l Page 55 of 100
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Raactor Oparator Ex:mination 56.Given the following conditions on Unit 1:
- A release is in progress from the 121 ADT Monitor Tank using the programmable controller
- Power is subsequently lost to bus 13 and the bus is deenergized What is the effect on the release?
The release will...
a. continue as normal.
b. continue but requires control directly from the programmable controller using the programming panel or l&C laptop computer.
c. terminate due to loss of power to the control room instrument for R-18 radwaste liquid release radiation monitor.
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d. terminate due to loss of power to the programmable controller which closes the release valve and stops the pumps.
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(dtSI@n teq5 cNth I
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Page 56 of 100
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R3:ctor Operater Ex minstion
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57. Unit 1 was at 100% when a plant transient occurred resulting in the following conditions:
- RCS temperature - 520'F
-- PRZR pressure - 1850 psig
- PRZR level - 17%
_SG pressures - 490 psig (11); 800 psig (12)
- Both RCDT pumps are in AUTO
- RCDT Discharge valves are in AUTO
- The level in the RCDT has risen above the alarm setpoint for
" Reactor Coolant Drain Tank Hi/ Lo Level" Assuming all systems are functioning correctly, what is the status of the RCDT system?
a. BOTH RCDT pumps are running and flow is directed to the CVCS Holdup Tanks.
b. BOTH RCDT pumps are running and flow is recirculated back to the RCDT.
c. ONE RCDT pump is running and flow is directed to the CVCS Hold'up Tanks.
d. NEITHER RCDT pump is running and NO flow exists for the system.
Page 57 of 100
p R cctor Opsrefor Examination 58.When the Waste Ges Decay Tank 127 is being vented to atmosphere, what condition will automatically close the Waste Gas flow control valve CV-312717 a. BOTH Waste Gas Compressors 121 and 122 trip.
b. Rad Waste Building Radiation Monitor R-35 HIGH alarm trip.
~. BOTH Sample Room Exhaust Fans 121 and 122 trip.
c d. Wasts Gas High Level Loop Monitor R-41 HIGH alarm trip.
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Page 58 of 100
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Racctor Opnrr tcr Extmination 59.Given the following conditions on Unit 1:
- Containment Radiation Monitor 1R-7 has alarmed
- When the operator acknowledged the alarm, annunciator "HIGH RADIATION TRAIN B" remained in solid How will the control room operators be alerted to any subsequent Train B radiation monitor alarms?
a. An audible alarm will occur at the local Train B Radiation Monitoring Panel.
b. The "High Radiation Train B" annunciator will reflash with audible alarm.
c. An audible alarm will occur on Victoreen Radiation Monitoring panel, d. The "High Radiation Train B" annunciator will reflash but NO audible alarm will occur.
.
Page 59 of 100 l
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Rsector Opsrctor Excmination i
60. An inadvertent Si signal has occuired on Unit 1.
i What alami associated with 1R-11, Containment Particulate Radiation Monitor, is expected?
a. "High Press" due to increased pressure inside containment.
b. "High Flow" due to increaseu differential pressure between containment and the shield building stack.
I c. " Low Flow" due to closure of the sample isolation valves.
f d. "High Alarm" due to increase radioactivity levels in containment.
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Page 60 of 100
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c Racctnr Operrt::r Exrminntion 61.Given the following conditions on Unit 2:
- An inadvertent SI signal was generated for B Train only
- All equipment responds as expected What is the status of the Safeguards Cooling Water system pumps following these actions?
a. 22 CL pump running supplying the B header 121 CL pump supplying the B header b. 22 CL pump running supplying the B header 21.CL pump supplying the A headcr c. 22 CL pump running supplying the A header 121 CL pump supplying the B header d. 22 CL pump running supplying the A header 21 CL pump supplying the A header
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Page 61 of 100
Rsictor Opsrctor Excmination i
62 :. According to the precautions of C34 " STATION AIR SYSTEM", what is the MINIMUM required instrument air header pressure?
f
' a. 95 psig.
. b. 85 psig.
c. 75 psig,
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d. -70 psig.
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Page 62 of 100
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R::cctor Operctor Excmin tion 63.Given the following conditions on the plant:
Maintenance has been given permission to work on one fire header hose station 1 -
The 121 MD fire pump control switch was placed in PULLOUT
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Maintenance went to an incorrect hose station and initiated flow,
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. then reclosed the valve Fire header pressure showed the following history
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TIME PRESSURE-1000 -
123 psig l-1001 -
100 psig-1002.
95 psig-1003 -
105 psig-1004 -
120 psig i
I What is the status of the pumps associated with the fire protection systeh1 at 1010 assuming NO
operator action had been taken?
Motor Driven Diesel Driven Jockey Screenwash Fire Pump Fire Pump Fire Pump Pump a. Stopped Stopped Running Running b. Stopped Running Stopped Stopped c. Running Running Running Stopped d. Runnirig Stopped Stopped Running
Page 63 of 100
Rscctor Opsrc. tor Excmination
- 64.Given the following conditions on Unit 1:
- The unit was tripped and SI was manually actuated
- When the SI was actuated, containment to annulus differential pressure was at -0.35 psi What is the current status, following the SI, of the Containment Vacuum Breakers?
Both vacuum breakers are...
a. closed but will reopen when the Cl signal is reset.
b. closed but will reopen if the containment to annulus differential pressure becomes more negative than -0.4 psi.
c. open but will close when the containment to annulus differential pressure becomes more positive than -0.2 psi.
d. open but will close when the Cl signal is reset.
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Page'64 of 100
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R :ctsr Op::rct:r Exrminction 65.Given the following conditions on Unit 1:
- Reactor poweris 30%
- Rod control is in' Automatic
- Tref - 551*F
- Loop Tavg values - 551*F (A); 552"F (B)
- - Power Range Ni - 31% (N41); 29% (N42),30% (N43); 30% (N44)
- Control bank D is at 156 steps Which condition would result in con +inuous rod withdrawal?
a. Turbine first stage pressure PT-485 fails upscale.
b. N41 fails upscale.
c. Loop A Tcold fails downscale.
d. Tref signal fails downscale.
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Page 65 of 100
p R2:ctor Op::rctor Ex:minntion 66.Given the following conditions on Unit 1:
- Reactor power-96%
- Control Bank D demand position - 210 steps
- Control Bank 1RPI positions are at approximately 210 steps except for
rod G-11 which indicates 188 steps
- G-11 is determined to be misaligned
in accordance with 1C5 AOP5 " MISALIGNED ROD, STUCK ROD, AND/OR RPI FAILURE",
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which of the following describes the conditions for realignment of rod G-117 Reactor power is...
)
a. reduced to less than 50% prior to moving G-11.
b. reduced to less than 80% prior to moving G-11.
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c. maintained at 96% while G-11 is being moved.
d. maintained between 96% and 100% while G-11 is being moved.
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Page 66 of 100 J
i Rocctor Operator Extminttien
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i 67-.Given the following conditions on Unit 1:
- The reactor has tripped from 100% power
- 6 minutes.following the trip power was lost from the 1R transformer
- Steam Dumps are selected to the STEAM PRESSURE Mode Which of the following identifies the expected conditions as natural circulation flow develops in the RCS?
a. - RCS subcooling 63* and decreasing l
- SG pressures increasing
- Core exit T/Cs increasing b. - RCS subcooling 63* and increasing
- SG pressures stable
- Core exit T/Cs decreasing
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c. - RCS subcooling 106* and decreasing
- SG pressures stable
- Core exit T/Cs increasing d. - RCS subcooling 106* and increasing-
- SG pressures decreasing
- Core exit T/Cs decreasing i
l Page 67 of 100 J
I Rarctor Op:rcter Examinrtion
. 68.Given the following conditions on Unit 1:
- A reactor trip and SI signal have been generated due to a LOCA
- Both RCPs are stopped
- All required equipment started and is operating normally for the current conditions
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- RCS Core exit T/C temperatures are 480*F and stable j
- RCS pressure is at 1320 psig
- PRZR level is 25%
- Containment pressure is 6 psig and stable
- Actions of 1ES-1.1 " POST LOCA COOLDOWN AND DEPRESSURIZATION" are in
,
progress and preparations are being made to start an RCP
- RCP seal injection flow is 6.8 gpm to the RCP to be started Which action should be performed prior to starting an RCP in order to prevent the need for Si reinitiation?
a. Start a CFCU associated with the selected RCP in FAST speed.
b. Verify SG level on the loop with the selected RCP is adequate.
c. Raise sealinjection flow.
d. Raise PRZR level.
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Page 68 of 100 i
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Reector Opsrcter Examination
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' 69 - A small break LOCA has occurred on Unit 1 and SI has been terminated per the EOPs. Which of the following would require SI Reinitiation?
- a. PRZR level <5% AND RCS subcooling <20*F.
b. PRZR level <5% OR RCS subcooling <20*F-c. RCS pressure <2000 psig AND RCS pressure stable or increasing.
d. RCS pressure <2000 psig'OR RCS pressure decreasing.
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Page 69 of 100
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Rsactor Oparator Examination
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70.in 1E-1 " LOSS OF REACTOR OR SECONDARY COOLANT", the operator is directed to cool i
down and depressurize the secondary side if intact SG pressures are greater than RCS pressure.
I What action should be taken if elevated radiation levels are detected in the 11 SG only?
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11 SG pressure should...
a. NOT be reduced by releasing steam from the SG in order to minimize radiological releases.
b. NOT be reduced to lower than RCS pressure to protect the SG components from further J
damage.
c. be reduced to less than RCS pressure to further enhance cooldown and depressurization of the RCS.
d. be reduced to less than RCS pressure to enhance recovery of core water level.
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l Page 70 of 100
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Rcactor Optrator Extminstion 71.While operating at full power,11 RCP #1 Seal Leakoff isolation Valve, CV-31335, closes due to a failed solenoid valve. Which of the following sets of indications would you expect to observe?
a. - 11 RCP Labyrinth Seal belta P indicates low (ZERO inches).
- 11 RCP standpipe low level alarm is in.
I b. - 11 RCP Labyrinth Seal Delta P indicates low (ZERO inches).
- 11 RCP standpipe high level is in.
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c. - 11 RCP #1 Seal Leakoff flow indicates low (ZERO gpm).
- RCDT levelis rising.
d. - 11 RCP #1 Seal Leakoff flow indicates low (ZERO gpm).
- RCDT levelis stable.
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Page 71 of 100
R3:ctor Op;rctor Extminrtion 72.Given the following conditions on Unit 1:
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- Reactor power - 65%
- Auto makeup initiated to the VCT on low level due to known RCS leakage of 3 gpm to the PRT
- Shortly thereafter, " Boric Acid Flow Controller Deviation" alarm actuated What affect do the above conditions have on plant operations, if NO operator action is taken?
a. VCT level will rise to the auto makeup stop setpoint.
b. Reactor power will slowly rise.
c. RCS Tavg will slowly rise.
d. VCT level will drop until charging suction swaps to the RWST.
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Page 72 of 100
u Rocctor Op rster Excminntion 73.Given the following conditions on Unit 1:
- The jetdown heat exchanger is out of service
- Excess letdown is in service-
- 11 Charging pump is running at minimum speed.
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- An ATWS has been diagnosed
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I Which of the following values is the highest boration rate that the RO would establish using the normal boration flowpath?
a. 20 gpm b. 15 gpm i
c. 412 gpm d.
8 gpm
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i Page 73 of 100
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Rs:ctor Op rator Excmination 74.The following p! ant conditions exist:
- Loss of offsite power; reactor tripped
- EDG D1: running
- EDG D2: out of service
- LOCA inside containment
- 11 CC PUMP LOCKED OUT annunciator: lit
- Pressurizer pressure: 1170 psig
' RCS subcooling: 80*F
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- Containment pressure: 25 psig
- E-1, " Loss of Reactor or Secondary Coolant", is being executed if operating limits are ' exceeded on component cooling system vital components that are operating, which of the following is the operator required to trip?
a.11 RHR pump.
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b.11 Si pump.
c. 11 CL pump.
d.11 CS pump.
Page 74 of 100 e-
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Roccter Operater Examination 75. A leak in which of the following components would result in an increase in the Component Cooling System surge tank level? (Assume the associated system is in service at the time of the leak.)
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a. Containment-Auxiliary Building chiller.
b. RCP Seal Water heat exchanger.
c. Spent Fuel Pool heat exchanger, d. Excess Letdown heat exchanger.
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Page 75 of 100
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R:::ctor Oper:ter Exrminttien 76.Given the following conditions on Unit 1:
- RCS Tavg - 150*F
- RCS pressure - 280 psig if 1PT-419 wide range loop pressure transmitter fails high, what is the response of the pressurizer PORVs and the reason for this response?
a. Neither PORV opens because coincidence is 2/2 with OPPS in ENABLE.
b. Only PORV "B" (1PCV-431C) does because the failed transmitter is in its train.
c. Only PORV "A" (1PCV-430) opens because the failed transmitter is in its train.
d. Both PORVs open because the coincidence is 1/2 with OPPS in enable.
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P Page 76 of 100
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Rocctor Opercter Extminstion 77.With toe plant in a normal 100% lineup, a 5 gpm leak develops in the sensing (variable)line from
' the PRZR to the red channel LT-426 PRZR level transmitter.
What is the responso of the CVCS system?
Charging flow will...
a. increase to the maximum value, b. ramp down to the minimum value.
c. decrease and then return to the initial value.
d. increase to the value where makeup equals the loss through the leak.
Page 77 of 100 t
I Reector Op rcter Extminntien
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i,.Given the following conditions on Unit 1:
- A valid reactor trip signal has been generated
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- Reactor trip breakers remained closed
- Reactor power is 95%
- Both Main Feedwater Pumps are tripped What is the basis for the operator tripping the main turbine within 30 seconds of veritying the above conditions?
a. Initiate an additional reactor trip signal.
b. Ensure adequate steam supply to the TDAFW pump.
c. Conserve SG waterinventory.
d. Minimize the resulting RCS pressure rise.
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~ Page 78 of 100
r R: cter Op rctor Excmin: tion
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79..The following conditions exist on Unit 1:
- Reactor startup in progress
- IR indication: N35 - 2.6x10'5 amps
N36 - 2.7x10 amps
- Source Range Blocked Hi Volt Off Aqua Panellight lit i
- High voltage power source for Source Range Channel N31 fails such that its output is 1000 volts.
What indication (s) would be available to alert the operator to this failure?
a. None, until power is lowered below the P-6 setpoint, and then the Source Range N31 will indicate higher than expected.
b. None, until power is lowered below the P-6 setpoint, and then the " Source Range Loss of Detector Volt" alarm will actuate.
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c. The " Source Range Loss of Detector Volt" alarm will actuate, and Source Range N31 will indicate lower than expected.
d. The " Source Range Loss of Detector Volt" alarm will actuate only.
Page 79 of 100
R ct:r Op rcter Extminntion
80.Given the following conditions for Unit 1:
- A reactor startup is in progress per C1.2
- NIS Checks at 6% power are being performed
- A problem with IR channel N35 requires investigation and repair prior to continuing the startup Which of the following actions can be performed with the plant at power in order for l&C to conduct repairs?
a. Place N35 Level Trip Bypass switch to BYPASS prior to pulling N35 instrument power fuses.
' b. Place N35 Level Trip Bypass switch to BYPASS prior to pulling N35 control power fuses.
c. ' Energize P-10 bistables for two Power Range channels prior to pulling N35 control power fuses.
d. Pull the instrument power fuses for two Source Range channels prior to pulling N35 instrument power fuses.
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I Page 80 of 100
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R cct:r Operstar Excmin tion
. 81. Given the following conditions:
- RCS leakage into #21 Steam Generator is 0.1 gpm.
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- RCS leakage into #22 Steam Generator is 0.3 gpm.
- Previous leak rate from identified sources was 6.3 gpm.
- Current total leak rate is 7.0 gpm.
-Identify the Technical Specification leakage limit that has been exceeded.
a. Total RCS to steam generator leakage.
b. Totalleakage from other than controlled sources.
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c. Leakage rate from controlled sources.
d. Leakage rate from an unidentified source.
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Page 81 of 100 l
Rocctor Operator Examination 82.Which condition may lead to a Pressurized Thermal Shock condition in a steam generator tube rupture event?
' a. Si flow to one cold leg is isolated.
b. Si to reactor vessel valves are opened.
c. The RCP trips in the loop with the ruptured SG.
d. Both RCPs have stopped.
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I Page 82 of 100
Rsrctor Op:rcter Excmin tion
83. Unit I has experienced a Safety injection followed by a loss of all AC Power.
What are the conditions required and the preferred method for restoring power to a Unit 1 Safeguards Bus if a Unit 2 Safeguards Bus is the only available source?
a. - The Unit 1 SI signal remains active and Unit 2 SI signal is reset.
- Bus 25 is cross connected to Bus 15.
l b. - The Unit 1 SI signal remains active and Unit 2 has NO SI.
- Bus 26 is cross connected to Bus 16, c. - The Unit 1 SI signal is reset and Unit 2 SI signal is active.
- Bus 25 is cross connected to Bus 15.
d. -The Unit 1 SI signalis reset and Unit 2 has NO St.
- Bus 26 is cross connected to Bus 16.
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Page 83 of 100 I
Racctor Opsrator Extmination 84.Given the following conditions for Unit 1:
- A loss of all AC power has occurred
- Selected components have been placed in PULLOUT in accordance with 1ECA-0.0," LOSS OF ALL SAFEGUARDS AC POWER"
-The safeguards buses have been reenergized from Unit 2 buses in accordance with 1ECA-0.0," LOSS OF ALL SAFEGUARDS AC POWER" Which of the following sets of parameters / conditions is used to select the appropriate recovery procedure?
a. RCS subcooling and PRZR level.
b. SI ACTUATED status light LIT and any first-out annunciator, c. PRZR pressure and PRZR level.
j d. SI ACTUATED status light LIT and any Si first-out annunciator.
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Page 84 of 100
R3rcter Op rctor Extminttien 85.Given the following conditions on Unit 1:
- A reactor trip has occurred due to a loss of offsite power.
- The actions of 1ES-0.1 " REACTOR TRIP RESPONSE" are being performed for verifying natura' cir :ulation flow Which of the following correctly explains why the SG levels are maintained in the normal post-trip control band during natural circulation?
a. Ensures RCS cocling remains symmetrical.
b. Prevents voiding frorn occurring in the reactor vessel head.
c. Ensures SG tubes are covered to verify natural circulation.
d. Prevents a complete loss of RCS flow due to voiding in a single loop, bW5 Oh (Ons ((dt re-l Page 85 of 100
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g Roccter Op3rzter Excmin tion 86'.Which failure below'will always result in an automatic reactor trip if the Unit is initially at 7%
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' power?
A complete loss of power to...
a.11 Bus.
b. 11 Inverter cabinet.
c. 15 Bus.
' d. -15 Inverter cabinet.
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Page 86 of 100
Rsrcter Opsrttor Ex mination 87. An operator inadvertently opened Unit 2 DC Panel 25 Circuit 6 breaker causing a SAFEGUARD LOGIC TRAIN A DC FAILURE annunc% tor.
What effect does this action have on Train 'A' Sl?
a.- Train 'A' Si spurious actuation occurs.
b. Automatic actuation of Train 'A' Si is disabled and Manual actuation of Train 'A' Sl is not affected.
c. Manual actuation of Train 'A' Si is disabled and Automatic actuation of Train 'A' Si is not affected.
d. Manual and automatic actuation of Train 'A' Sl are disabled.
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Page 87 of 100
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R ::cter Operator Excminction 88.Given the following conditions on Unit 1:
- Reactor power - 50%
- Instrument air header pressure is rapidly decreasing
- Operator action has failed to stop the air leak What is the criteria used to determine the initiation of plant shutdown and/or reactor trip?
a. Proximity of pressurizer level to the trip setpoint and its rate of increase.
b. Proximity of steam generatorlevel to the trip setpoint and its rate of decrease.
c. Charging pump speed failing to minimum with reduced seal injection capability.
d. MSIV low air supply pressure alarm.
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Page 88 of 100
Rocctor Oparator Extmination
)
i 89.What action is to be taken to prevent halon actuation in the Service Building Addition Computer
]
Room if the detection system actuates but the fire has already been extinguished?
Halon discharge is aborted by...
a. depressing and releasing the ABORT pushbutton in the computer room, and the signalis reset automatically.
b. depressing and releasing the ABORT pushbutton in the computer room, and then resetting j
the local panel outside the computer room.
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c. maintaining the ABORT push-button in the computer room depressed while the system is reset on the local panel outside the computer room.
,
d. maintaining the ABORT push-button in the computer room depressed while the system main header isolation valve is manually shut.
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Page 89 of 100
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Reacter Optrater Examinction 90.Given the following conditions on Unit 1:
-The reactor is tripped
- 12 Charging pump is running in AUTO
- The control room has been evacuated
- Control is being established at the Hot Shutdown Panels in accordance with
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1C1.3 AOP1 " SHUTDOWN FROM OUTSIDE THE CONTROL ROOM - UNIT 1"
' What is the expected response of the 12 Charging pump when its control switch at the pump is
_
taken to LOCAL 7 (Assume NO other local actions have been taken.)
a. -Its speed will reduce to the minimum value, b. The pump will continue to run at its previous speed.
c. Its speed will rise to the maximum value.
d. The pump will trip.
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Page 90 of 100
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Raccter Op::rcter Examinttien 91.What condition below would be logged as a loss of CONTAINMENT INTEGRITY?
a. At 100% power, while performing an operability test on two normally open, redundant containment isolation valves, one valve fails to close.
b. Following a reactor trip, an electrician opens the containment airlock outer door without prior approval.
c. With Tavg at 400*F, it is discovered that one of the rubber ring gaskets for equipment hatch
'is NOT installed.
d. With Tavg at 190*F, the Containment Purge System is placed in operation.
.
Page 91 of 100
l Rosctor Op::rctor Extmin ticn 92'.High coolant activity has been detected. Chemistry has determined that it is due to fuel pin leakage and corrosion product activation.
Identify the effect of placing the cation demineralizer in service.
The cation demineralizer...
I a. is less effective than the mixed bed demineralizer so it is placed in service ONLY if decontamination factor is less than 10.
b. is NOT effective in removing the these types of activated prodiacts.
c. will cause the RCS activity level to decrease soon after it is placed in service.
d. will remove lithium so it should NOT be used in this condition.
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Page 92 of 100
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Rocctor Oparctor Excmin tien 93 '.'During the performance of E-1 " LOSS OF REACTOR OR SECONDARY COOLANT", why are the Containment Spray Pumps stopped if containment pressure falls below 18 psig?
a. To prepare for transfer to recirculation, b. To conserve RWST inventory.
c. To reduce load on the safeguards bus.
d. To ensure CFCUs are functioning properly.
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Page 93 of 100
l Raccter Op rcter Excminction l
94..Given the following conditions on Unit 1:
- A small break LOCA has occurred-The actions of 1ES-1.1 " POST-LOCA COOLDOWN AND DEPRESSURl7_ATION" are being performed.
- After stopping ONE S1 Pump, the following conditions are noted:
RCS pressure is 900 psig
.
RCS subcooling is 35'F
.
PRZR levelis stable at 24%
.
What is the result if the RO turns on a set of PRZR backup heaters to attempt to raise subcooling?
a. Break flow rises; Sl flow rises, b. Break flow falls; Pressurizer level rises,
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c. Break flow rises; Pressurizer level falls.
d. Break flow and Pressurizer level remain constant.
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Page 94 of 100
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Rsectcr Operctor Exrminntion i
95.Which of th' following describes the primary concern associated with restoring full feed flow to a e
hot, dry SG following a loss of all feedwater?
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b 1l a. Overcooli g event due to excessive feeding.
."b. SG tu'be failure due to thermal stresses, c. Overpressurization of the secondary system causing Safeties to open.
d. RCS Pressurized Thermal Shock due to interruption of natural circulation flow.
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Page 95 of 100
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R3:ctor Op3rator Excminttion 96 ;Given the following conditions for Unit 1:
- Actions of 1FR-C.2 " Response To Degraded Core Cooling" are being performed
- The SI accumulators are being isolated in preparation for depressurizing the SGs from 165 psig to atmospheric pressure
- Both RHR pumps are running
- During the crew brief prior to SG depressurization to atmospheric pressure, the Lead asks if the running RCPs should be stopped What is the required action and basis for this action relative to the operation of the RCPs?
a. RCPs should NOT be stopped because flow through the core must be maintained when inadequate core cooling exists, b. RCPs should NOT be stopped because stagnating flow during depressurization could cause a bubble to form in the reactor vessel head.
c. RCPs should be stopped because as the RCS reaches saturation conditions, mechanical damage can occur due to cavitation, d. RCPs should be stopped because the drop in #1 seal delta-P during depressurization may cause pump damage.
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Page 96 of 100
i R:ccter Op:rcter Exrminttion 97-.Given the following conditions on Unit 1:
- A main steam line break occurred inside containment.
- The actions of 1FR-P.1," RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITION" are being performed j
- A required one hour soak has just been initiated
-
- RCPs are NOT running
Which of the following correctly identifies an evolution that can be performed during the soak?
a. Lower the SG PORV controller setpoint by 25 psiin AUTO.
b. Place PRZR auxiliary spray in service.
c. Energize all PRZR backup heaters.
d. Raise SG water levels from 20% to 40%.
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Page 97 of 100
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Rsecter Op rctor Examinatien
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l 98.Given the following conditions on Unit 1:
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- A natural circulation cooldown is in progress in accordance with 1ES-0.3A
" NATURAL CIRCULATION COOLDOWN WITH CRDM FANS"
- RCS pressure 1200 psig
- Pressurizer level-42%
- RVLIS Full Range - 83%
Which of the following correctly describes the method to be used to collapse the apparent void?
a. Start an additional charging pump to raise RCS pressure while maintaining RCS temperature constant.
b. Lower RCS temperature by raising the steaming rate from the SGs and maintain RCS pressure constant.
.
c. Raise RCS pressure using Pressurizer heaters and then continue with the cooldown.
d. Maintain RCS pressure stable and then initiate venting of the reactor vessel head using Head Vent System.
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Page 98 of 100 l
R=ctor Op:;rctor Examination 99.Given the following conditions:
- A LOCA outside containment has occurred
- Entry to ECA-1.1 " Loss of Ernergency Coolant Recirculation" has occurred
- Containment Pressure is 25 psig
- All ECCS equipment has actuated
- RCS pressure is 800 psig
- RWST icvel is 10%
- VCT level is 2%
Which of the following actions will help preserve RWST inventory?
a. Stop any running Charging Pump.
b. Stop any running Containmtat FCUs.
c. Isolate RHR discharge to the RCS cold legs.
,
d. ' Isolate Si accumulators, i
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Page 99 of 100
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Rsector Opsretar Extmination i
100. iECA-2,1 " UNCONTROLLED DEPRESSURIZATiON OF BOTH STEAM GENERATORS" has been entered. The CAUTION prior to Step 2 requires a minimum feed flow of 40 gpm be maintained to'each SG that has a Narrow Range level less than 10%.
What is the basis for this maintaining this amount of feed flow?
a. Minimize thermal shock to SG components, b. Maintain a verifiable cooldown rate.
c. Provide for positive break isolation indication.
d. Ensure adequate subcooling margin maintained.
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Page 100 of 100 -
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Reactor Operator Answer Key
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, PRAIRIE ISLAND NUCLEAR GENERATING PLANT
/ NORTHERN STATES POWER COMPANY OPERATING PROCEDURES
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TITLE NUMBER:
y C12.5
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BORON CONCENTRATION CONTROL REV: 7
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Page 19 of 28 FIGURE 1 - BLENDED FLOW NOMOGRA'PH (3)
8000 1. Select desired boron blend on column (3).
, 2. Select convenient reactor makeup water flow settmg on
-
'000 column (1) (HFC-111 (Rack CVCS2) is normally set at 45%).
-- 3000 3. Draw a straight line f' om points I and 3 to determine boric r
acid flow setting on column (2).
2000
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(2)
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n:z 200
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60 Reactor Makeup Water Flow (y)
3o Diended Flow ppm Boron (z)
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PRAIRIE ISLAND 1999 RO EXAM REFERENCE ANSWER 001
- KA 194001G101
- ANSWER b
- REFERENCE SAWI 3.15.0, Rev 4, Plant Operation, Section 6.2.10, Page 6-7 & 6.4.2 P9150L-014, Rev 2, Plant Operation, Section ll.C.4, Page 11-12,1.a ANSWER
'002
- KA 194001G103
- ANSWER b
- REFERENCE SWI O-2, Rev. 40, Shift Organization, Operation & Turnover, Section 6.3.4, Page 11 P9150-L-012, Rev 2, Records and Logs, Section ll.A.3.a, Page 9 ANSWER 003
,
- KA 194001G110
- ANSWER a
- REFERENCE 1C1.4, Rev.14, UNIT 1 POWER OPERATION, Section 4.11, Page 4 P8197L-007, Rev 5, C1.4/C1.6, Review, Section ll.D, Page 8 SWI O-1, Rev.10, Work Rules and Philosophy, Section 6.7.2, Page 7 ANSWER 004
- KA 194001G117
- ANSWER b
- REFERENCE 5AWI 3.15.6, Rev. O, Operation Section, Section 6.14.c, page 1 P9150L-018, Rev.1, Communications,Section V.B.2.c, Page 12 ANSWER 005
- KA 194001G129
- ANSWER c
- REFERENCE 5AWI 3.10.0, Rev. 8, Control And Operation Of Plant Equipment, Section 6.10.14.d.1, Page 30-32 5AWI 3.10.1, Rev 7, Methods Of Performing Independent Verification, Section 6.3.1, Page 7 P9150L-024, Rev 0, Control / Operation of Plant Equip,Section V.F.1, Page 24 ANSWER 006
- KA 194001G212
- ANSWER d
- REFERENCE SP1001, Rev. 49, Unit Daily Control Room Log, Section 5 P9150L-016, Rev.1, Surveillance Program,Section IV.D.4,1 Page 11-12
i
_
ANSWER-007
- KA 194001G224
- ANSWER -
d
- REFERENCE 5AWI 3.10.0, Rev. 8, CONTROL AND OPERATION OF PLANT EQUIPMENT, Section 6.9.5.c, Page 19 P9150L-024, Rev. O, Control / Operation of Plant Equipment, Section ll.E.3.d.2), Page 12-13
{
ANSWER 008
- KA 194001G230
- ANSWER d
- REFERENCE.
D5.2, Rev. 23, REACTOR REFUELING OPERATIONS, Section 6.26,7.1.5, Pages 12-13, & 6
)
SWI O-41, Rev. 5, DUTIES AND RESPONSIBILITIES OF FUEL HANDLING PERSONNEL, Section 6.1.6, Page 3 P9150L-004, Rev. 2, Plant & Shift Organization, Section Ill.d & e, Page 11-12
'
ANSWER 009
- KA-194001G301
- ANSWER c
- REFERENCE F2, Rev.13, Radiation Safety, Section 4.2.2 & 4.2.3, Page 17-18
'
P9130L-003, Rev. 2 F2, Radiation Safety,Section V.C.2, Page 12
-
ANSWER 010
- KA -
194001G310
- ANSWER b
REFERENCE F2, Rev.13, Radiation Safety, Section2.2.1.F, Page 9 P9130L-003, Rev. 2, F2 Radiation Safety, Section Ill.A.1.d, Page 10 ANSWER 011
- KA 194001G401
- ANSWER b
- REFERENCE 1E-0, Rev.17, Reactor Trip Or Safety injection, Step 1, Page 3 P8197L-010, Rev. 2, EOP Intro-Procedure Review, Section 2 ANSWER 012
- KA 194001G409
- ANSWER -
b
- REFERENCE 1D2, Rev. 7, RCS REDUCED INVENTORY OPERATION, Section 3.10, Page 8 P9170L-001, Rev. 3, RCS Reduced inventory Operation,Section IV.C.10 &V.E.2, Page 26,
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32.
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ANSWER 013
- KA 194001G416
- ANSWER a
- REFERENCE 1ECA-0.0, Rev.12, Loss of All Safeguards AC Power, NOTE Step 1, Page 3 P8197L-010, Rev. 2, EOP Intro-Procedure Review, Section ll.C; lll.8 & C, Pages 14-16
ANSWER 014
- KA 001000K103
- ANSWER c
REFERENCE
'
B5, Rev. 2. ROD CONTROL SYSTEM, Section 3.4, Page 17 P8184L-005, Rev. 2, Rod Control & Rod Position Indication, VI.G.3, 5,6, Pages 38-39 ANSWER 015
- KA 002000K113 ANSWER a
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- REFERENCE
C3, Rev. 23, REACTOR COOLANT PUMP, Sections 4.3,4.4, Page 5 l
B4, Rev. 5, REACTOR COOLANT SYSTEM, Section 2, Page 3 l
P8170L-003, Rev. 4, Reactor Coolant System,Section V.C.8, Page 44 ANSWER 016
- KA 002000K602
' ANSWER a
- REFERENCE C3, Rev. 23, REACTOR COOLANT PUMP, Section 5.3.3, Page 12 P8170L-003, Rev. 4, Reactor Coolant Syswtem,Section IV.A2.a.5), Page 20 ANSWER 017
- KA 003000A201
- ANSWER b
- REFERENCE C3, Rev. 23, REACTOR COOLANT PUMP, Section 3.3, Page 4 B3, Rev. 5, REACTOR COOLANT PUMPS P8170L-002, Rev. 2, Reactor Coolant Pumps,Section IV.A.7.f, Page 15 ANSWER 018
- KA 003000K201
- ANSWER d
- REFERENCE B3, Rev. 5, REACTOR COOLANT PUMPS, Section 3.3.1, Page 11 P8170L-002, Rev. 2, Reactor Coolant PumpsSection VI.D, Page 23 P8186L-003, Rev. 2,4160 & 480V Non-Safeguards Distribution,Section IV.B.7, Page 17
r ANSWER 019
- KA 004000A301
- ANSWER b
- REFERENCE C12.5, Rev. 6, BORON CONCENTRATION CONTROL, Section 5.1.2 & Fig 5, Page 18 B12A, Rev. 3, CHEMICAL AND VOLUME CONTROL SYSTEM. Section 3.3.B. Page 15 P8172L-001a, Rev. 3, CVCS,Section IV.B.2, Page 28-29 ANSWER 020
- KA 004000K305
- ANSWER a
- REFERENCE B7, Rev. 2, REACTOR CONTROL SYSTEM, Section 3.4.1, Page 23 B12A, Rev. 3, CHEMICAL AND VOLUME CONTROL SYSTEM, Section 3.4.B, Page 21 P8170L-006, Rev. 3, Pressurizer Level Control Systemill.C.5; V.A.310,, Section 17, Page 1 ANSWER 021
,
- KA 005000A101
- ANSWER a
- REFERENCE B15, Rev. 6, RESIDUAL HEAT REMOVAL SYSTEM, Section 3.3.A,4.2, Page 10,17 P8180L-003, Rev. 3, Residual Heat Removal,Section IV.C.3, Pages 26-27 ANSWER 022
- KA 006000A308
- ANSWER a
- REFERENCE B18A, Rev. 3, SAFETY INJECTION SYSTEM, Section 3.2.C, Pages 9-10 P8180L-004, Rev. 3, Safety injection System & Accumulators,Section IV.A.2.h, Page 21 ANSWER 023
- KA 006000K602
- ANSWER d
- REFERENCE B18A, Rev. 3, SAFETY INJECTION SYSTEM, Section 3.5.A, Pages 14-15 P8180L-004, Rev. 3, Safety injection System & Accumulators,Section V.B.2.c, Page 42 ANSWER 024
- KA 008000A401
- ANSWER d
' REFERENCE C12.1, AOP1, Rev. 2, LOSS OF RCP SEAL INJECTION, Section 2.4, Page 3 B14, Rev. 4, COMPONENT COOLING SYSTEM,3.2.C, Section 4.18, Pages 19-20 P8172L-002, Rev. 4, Component Cooling, V.C.6,Section IV.A.2.e, Page 13, 30
_
h ANSWER-025
- KA 010000K501
- ANSWER a
- REFERENCE B7, REACTOR CONTROL SYSTEM l
P8170L-005, Rev. 4, Pressurizer Pressure Control System,Section V.A.4, Pages 23-24 P8170L-006, Rev, 3, Pressurizer Level Control System,Section IV.A.4.c, Page 15 ANSWER 026 j
- KA 011000A211
- ANSWER d
' REFERENCE P8170L-006, Rev. 3, Pressurizer Level Control System, Sections ll.C.4, IV.B.2.b, Pages 9-10 ANSWER 027
,
- KA 012000K302
- ANSWER-b
- REFERENCE 97AF02-0004, DSS Control System review, Section 3, pages 2-1,3-1 ANSWER 028
- KA 012000K501
- ANSWER d
-
- REFFRENCE
B8, Re.v. 3, REACTOR PROTECTION SYSTEM, Section 3.1.1, Pages 5-6 P8184L-004, Rev. 4, Reactor Protection,Section V.A.15, Pages 21-22 6 ANSWER 029
- KA 013000A403
- ANSWER d
i
- REFERENCE B18C, Rev. 2, ENGINEERED SAFEGUARDS SYSTEM, Section 3.5.B, Page 14 B12A, Rev. 3, CHEMICAL AND VOLUME CONTROL SYSTEM, Section 3.6.A, Page 23-24
,
'
P8180L-006, Rev. 3, Eng!neered Safeguards System,Section V.B.2, Pages 3,4 ANSWER 030
- KA 013000K412
? ANSWER d
- REFERENCE B18C, Rev. 2, ENGINEERED SAFEGUARDS SYSTEM, Section 4.2, Page 19 P8180L-006, Rev. 3, Engineered Safeguards System,Section V.B.1.d,' Page 19 ANSWER'
031-
- KA
.013000K601
- ANSWER-a
- REFERENCE j
1C51.4, Rev 8, instrument Failure Guide Req Correcti'te Action, Section 2 B18C, Rev. 2, ENGINEERED SAFEGUARDS SYSTEM,.9ections 3.3.C,4.510, Page 20 P81BOL-006, Rev. 3, Engineered Safeguards System, Section Ill.C.3.b.1, Page 13 i
l
ANSWER 032
- KA 014000G431
- ANSWER b
- REFERENCE B6, Rev. 6, ROD POSITION INDICATION SYSTEM, Section 3.3, Page 6
P8184L-005, Rev. 2, Rod Control & Rod Position Indication,Section IX.C.4, Page 52 C41.4, Rev. 6, ERCS Operating Procedure, Section 19.6 ANSWER 033
- KA 015000A202
- ANSWER c
- REFERENCE B9A, Rev. 3, NUCLEAR INSTRUMENTATION SYSTEM, Section 3.3.2, Page 11-12 I
P8184L-002, Rev. 4, Nuclear Instrumentation System,Section V.B.1, Page 45 J
ANSWER 034 i
- KA 015000K504
,
- ANSWER a
- REFERENCE P8184L-002, Rev. 4, Nuclear Instrumentation System,Section V.A.3, Page 44 C41.4, Rev. 6, ERCS Operating Procedure, Section 3.0, Pages 8-11 ANSWER 035
- KA 017000A101
- ANSWER d
- REFERENCE B10, Rev. 2, INCORE INSTRUMENTATION SYSTEM, Section 3.4, Page 10 P8184L-001, Rev. 2, Nuclear instrumentation incore,Section VI.B.4, Page 24 ANSWER 036
- KA 017000A301
- ANSWER b
- REFERENCE 1ES-0.3A, Rev. 7, Natural Circulation Cooldown with CRDM Fans, Step 13, Page 6 P8197L-011, Rev. 2, E-O Review,Section VI., Page 20 P8184L-001, Rev. 2, Nuclear instrumentation incore, Section ll.B, IX.A.1.b, Pages 8-9 ANSWER 037
- KA 022000A301
- ANSWER a
- REFERENCE B19, Rev. 3, CONTAINMENT SYSTEM, Section 3.1.3.C, Page 23 P8180L-000H, Rev.1, Containment Air Handling System, Section ll.A.S.d, Page 27 P8176L-003, Rev. 3, Cooling Water System, Vll.A.3.c 53, Page 8
ANSWER 038
- KA 022000G132
- ANSWER:
b
- REFERENCE 1C19.2, Rev.1, CONTAINMENT SYSTEM VENTILATION, Section 3.7,5.6.3, Page 4,30 B19, Rev. 3, CONTAINMENT SYSTEMS, Section 3.13.A, Page 18-19 P8180L-009H, Rev.1, Containment Air Handling System, Section ll.A.4.b.3) & 13, Page 25-26 ANSWER 039
- KA 026000A401
- ANSWER c
- REFERENCE B18D, Rev. 4, CONTAINMENT SPRAY SYSTEM, Section 3.1.C, Page 6 P8180L-002, Rev. 4, Containment Spray System, Section Ill.C.1.b, Page 28 ANSWER 040
- KA 028000A401
- ANSWER b
- REFERENCE C19.8, Rev.10, POST LOCA H2 ELECTRIC RECOMBINER CONTROL, Section 5.2,5.2.2, Page 5 B19, Rev. 3, CONTAINMENT SYSTEMS, Section 3.12.A, Page 15 P8180L-008, Rev. 2, Containment Hydrogen Control,Section IV.A.1.c.2, Pages 15-16 ANSWER 041
- KA 029000A301
- ANSWER c
- REFERENCE B19, Rev. 3, CONTAINMENT SYSTEMS, Section 3.13.A, C, Pages 20,25 B18C, Rev. 2, ENGINEERED SAFEGUARDS SYSTEM, Section 3.6.A, Page 14 P8180L-009E, Rev.1, Containment Purge & In-Service Purge Ventilation System, Section ll.A.3.c, Pages 11,15 ANSWER 042
- KA 033000K105
- ANSWER d
- REFERENCE C16, Rev. 25, SPENT FUEL COOLING SYSTEM, Section 5.11, Page 15 B16, Rev. 3, SPENT FUEL POOL COOLING, Section 3.3.A, Page 6 P8182L-004, Rev.1, Spent Fuel Pool Cooling System,Section IV.C.4, Page 10 ANSWER 043
- KA 034000K402
- ANSWER c
- REFERENCE D5.2 AOP2, Rev. 2, STUCK FUEL ASSEMBLY IN THE TRANSFER TUBE, Section 2.5, Page
P8182L-003, Rev. 3, Fuel Handling,Section IV.D.2, Pages 59-60
,
ANSWER 044
- KA 035000A203
- ANSWER b
- REFERENCE B7, Rev. 2, REACTOR CONTROL SYSTEM, Section 3.5.2, Page 28 P8174L-006, Rev. 3, Steam Generator Level Control System, Section ll.D, Pages 218-19 ANSWER 045
- KA 039000G130
- ANSWER d
- REFERENCE B27, Rev. 2, MAIN AND AUXILIARY STEAM SYSTEM, Section 3.2.C, Page 5 P8174L-001, Rev. 2, Main and Auxiliary Steam System,Section VI.D.2, Page 3.a ANSWER 046
- KA 041000A302
- ANSWER c
- REFERENCE B7
REACTOR CONTROL SYSTEM 3.2.1 12 P8174L-002
Steam Dump Control System IV.B.1,14-15,22 4 P8170L-003
Reactor Coolant System IV.4.b 29 3 ANSWER 047
- KA 056000G128
- ANSWER b
- REFERENCE 1 FR-H.1
Response to Loss of Secondary Heat Sink Step 4-6 5-7 P8174L-003
Condensate and Feedwater Ill.A.2 9 5,6 ANSWER 048
- KA 059000A306
- ANSWER d
- REFERENCE P8180L-006
Engineered Safeguards SystemVI.E.2 29-30
ANSWER 049
- KA 059000K416
- ANSVER a
- REFERENCE B28A
CONDENSATE AND FEEDWATER SYSTEM, Section 3.8.C
P8174L-003
Condensate and Feedwater Ill.D.8.b 20-21
ANSWER 050
- KA 061000A202
- ANSWER a
- REFERENCE l
C47010
Alarm Response Panel 0105 P8180L-007
Auxiliary Feedwater System li.C.5 16 5,7
)
ANSWER 051
- KA 061000K102
- ANSWER c-
- REFERENCE B28B
AUXILIARY FEEDWATER SYSTEM 3.1.C,3.2.C6,9 P8180L-007
Auxiliary Feedwater SystemlV.A.2,18,21-22 4,8 ANSWER 052
,
- KA 062000K403
- ANSWER -
b
- REFERENCE ~
B20.5
4.16 KV STATION AUXILIARY SYSTEM 3.6.2,4.18,11.
P8186L-008 4 Safeguards 4160V & 480V Electrical Dist IV.D.3,16,26-27 6 IV.E.5.b
P8186L-008
Safeguards 4160V & 480V Electrical Dist Ill.A.3.b 10 4 ANSWER
.053
- KA 063000K103
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- ANSWER b
)
- REFERENCE 1C20.9 AOP5
FAILURE OF 11 BATTERY FUSE 2.1
B20.9
DC DISTRIBUTION SYSTEM 3.2,4.44,9 P8186L-005
DC Distribution V.A.5.d 22 9,10 ANSWER 054
- KA'
063000K201
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- REFERENCE 1C20.9 AOP1
LOSS OF UNIT 1 TRAIN "A" DC2.1,2.2 2-3 P8186L-005
DC Distribution IV.A.4.g).(4)
,
i ANSWER 055
- KA 064000A108
- ANSWER b
- REFERENCE B20.7-
EMERGENCY DIESEL GENERATOR 3.1, 4.1 3, 6 P8186L-004
Diesel Generators V.B.7,
5, 7 ANSWER 056
- KA 068000A302
- ANSWER d
- REFERENCE C21.1.2
PROGRAMMABLE CONTROLLER SYSTEMS, 3.0
B21B
LIQUID RADWASTE 3.3.3 15-16 P8182L-001 A
Radioactive Waste Liquid IV.G.5, 37-38,41 7
,
ANSWER 057
- KA 068000A404
- ANSWER d
- REFERENCE B21B
LIQUID WASTE SYSTEM 3.4.2
P8182L-001A
Radioactive Waste Liquid IV.B.3 26-27 5 P8180L-006
Engineered Safeguards SystemV.B.1 & 2 18-20 4 ANSWER 058
- KA 071000K104
- ANSWER c
- REFERENCE B21A
WASTE GAS SYSTEM 3.3.C 12 P8182L-001C
Rad Waste -Waste Gas Vill.C 25 8 ANSWER 059
- KA 072000G431
'
- ANSWER b
- REFERENCE B11
RADIATION MONITORING SYSTEM 3.5.A
P8182L-002
Radiation Monitoring System VI.A.5 30-31
ANSWER 060
- KA 073000A402
- ANSWER c
- REFERENCE B11
RADIATION MONITORING SYSTEM 3.1.A & B 4-5,9 P8182L-002
Radiation Monitoring System II.C 10 2 ANSWER 061
- KA 076000K402
- ANSWER a
- REFERENCE B35
COOLING WATER SYSTEM 4.2 25-26 P8176L-003
Cooling Water System IV.C.7 & 9,18-19, 3, 7 ANSWER 062
,
I
- KA 078000G132
- ANSWER b
- REFERENCE C34
STATION AIR SYSTEM 4.1
P8178L-005
Instrument and Station Air V.A 22 6,8 j
ANSWER 063
- KA 086000K402
- ANSWER d
- REFERENCE i
B31A
FIRE PROTECTION SYSTEM 3.1.C, 8, 9,13,
)
P8178L-002
Fire Detection and Protection System II.B.1.d,2.g, 11,13,14 2,3,6
I ANSWER 064
- KA 103000K107
- ANSWER.
c
- REFERENCE'
B19
CONTAINMENT SYSTEMS 3.6.B
P8180L-009F 1-Containment Vacuum Breaker System V.B 11-12
ANSWER 065
- KA~
000001A205
- ANSWER a
- REFERENCE 1C51.2 10. White Bus (iii) Instrument Turbine 1st 1 Failure Guide-Stage Pressure High 1P-485 -
P8184L-005 -
Rod Control & Rod Position Indication, IV.C.1-4 20-221,5 ANSWER 066
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- ANSWER -
b
- REFERENCE 1C5AOPS Misaligned Rod, Stuck Rod and/or RPI Failure, 7-8 P8184L-005 -
Rod Control & RPl
ANSWER 067'
- KA 000007A103
- ANSWER c
- REFERENCE P8197L-011 -
E-O Review -
IV.B, C
ANSWER 068
- KA 000009G420
- ANSWER d
- KA 000011A112
- ANSWER b
,
i
- KA.
000015K210 l
- ANSWER a
i
- REFERENCE 1E1, Rev.17, BACKGROUND INFORMATION FOR LOSS OF REACTOR OR SECONDARY l
COOLANT, Procedure 7 Step 29 P8197L-012
E-1/E-2 Review IX.B.8, D 53-54 5,10 L
L+
,
.1 ANSWER,
.071
- KA 000022G446
- ANSWER c
,
- REFERENCE B3, Rev. 5, REACTOR ' COOLANT PUMP, Section 3.2.1, Page 9 P8170L-002
Reactor Coolant Pumps IV.A.7.h 16 ' 6 P9140L-002 -
Reactor Coolant Pumps 1, 6
' ANSWER 072
- KA 000024A113
- ANSWER d-
- REFERENCE B12A, CHEMICAL AND VOLUME CONTROL SYSTEM, Sections 3.3.B,3.4.A17,19 P8172L-001a.
CVCS IV.B.2.h.3)
31 3,5 ANSWER 073
- KA 000025A104
'
- ANSWER b
,
' REFERENCE 1FR-S.1, Rev. 9, Background Information For Procedure 1FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION /ATWS C12.5, Rev. 6 BORON CONCENTRATION CONTROL-
{
P8197L-014
F/FR Review
'll.B.6.b 15 7 ANSWER
'074
)
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- REFERENCF 1C14 AOP1
Loss of CC Table 1 8
-
P8180L 003
Residual Heat Removal VI.A' 33-34 8 i
-
ANSWER-075
- KA~
000027A101
- ANSWER
'd
- REFERENCE B14 4.
COMPONENT COOLING SYSTEM 3.1.B 6 P8172L-002
Component Cooling IV.A.1.g.3) 12 2,6.c ANSWER 076
- KA 000028A107
- ANSWER d
- REFERENCE-B4A.
5~
REACTOR COOLANT SYSTEM 3.7.1 21 P8184L-003
Reactor Process instrumentation System Xil.A.3 40 2,6
_
]
ANSWER 077
- KA 000029K306
- ANSWER d
- REFERENCE B7 2-REACTOR CONTROL SYSTEM 3.4.2 24 P8170L-006
Pressurizer Level Control System Ill.C.5
.C 1.a ANSWER 078
- KA 000032K101
- ANSWER c
- REFERENCE 1FR-S.1, Rev. 9, BACKGROUND INFORMATION FOR 1FR-S.1, " RESPONSE TO NUCLEAR POWER GENERATION /ATWS Step 2 Procedure 1 P8197L-014
F/FR Review il.B.6 14-15
ANSWER 079
- KA 000033A102
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- ANSWER b
- REFERENCE.
B9A
NUCLEAR INSTRUMENTATION SYSTEM 3.2.2 7 P8184L-002
Nuclearinstrumentation System IV.A.3.b, 15,18 6, 8, 35.a
,
ANSWER 080
- KA 000037G110
- ANSWER a
- REFERENCE P8140L-116
D30 Reactor Startup; NIS Failures B.3 13,14
ANSWER 081
- KA 000038A202
- ANSWER a
- REFERENCE 91,133 Technical Specifications TS. 3.1.C.2 TS. 3.1-8,-9 P8170L-003
Reactor Coolant System VI.A 47 15 ANSWER 082
- KA 000055A203
- ANSWER d
- REFERENCE 1 E-3
BACKGROUND INFORMATION FOR STEAM GENERATOR TUBE RUPTURE, Procedure 16, step 37 P8197L-13A
E-3 Series Review Vll.C.7, 58,74 10, X.D.2.k ANSWER 083
- KA 000055G404
- ANSWER d
- REFERENCE
1ECA-0.0
LOSS OF ALL SAFEGUARDS AC POWER, Step 7,8 6-7 P8197L-011
E-O Review V.E.2 24 19,20
ANSWER 084
- KA 000056K101
- ANSWER -
a
-
-
24 P8197L-011
E-O Review V.E.5 26 F.19 ANSWER
.085=
- KA 000057A219
- ANSWER -
a
- REFERENCE-1ES-0.1, Rev.13, BACKGROUND INFORMATION FOR REACTOR TRIP RESPONSE, Procedure 3, step 10 P8197L-011 2-E-O Review IV.B, IV.D.113,19 C.11, E.18 1ES-0.3A, Rev. 8, BACKGROUND INFORMATION FOR NC COOLDOWN W/CRDM FANS, Procedure 4, Step 13
'
ANSWER Gb6
- KA 000058A203
- ANSWER b
- REFERENCE 1C51.2
WHITE BUS (iii) INSTRUMENTS FAILURE GUIDE N36
P8186L-015
Safeguard Dist.120 V AC Instrunientation IV.A.1, 9,16 5, IV.C.3d P8184L-004
Reactor Protection V.A.2,17-18, 5, 6, 7 VI.D.1 28-29 ANSWER 087
'KA 000065A206
- ANSWER d
- REFERENCE C47518
Alarm Response Panel 47518 0501 i
P8186L-005
DC Distribution IV.A.5.c.4) 22 3,9 ANSWER 088
- KA 000067A107
- ANSWER b.
- REFERENCE C34 AOP1
LOSS OF INSTRUMENT AIR Attachment A,10.A,
P8178L-005
Instrument and Station Air V.C 22 8 ANSWER 089
- KA 000011A112
- ANSWER c
- REFERENCE B31A
FIRE PROTECTION SYSTEM 3.21.B
P8178L-002
Fire Detection and Protection Systems Ill.R.8 21 4,7 C31
Fire Protection and Detection 5.1983,86 j
.
ANSWER
- KA-000069G222 -
- ANSWER b!
- REFERENCE P8197L-008
.4 '
1C1.3 AOP1 Review II.F.10 12 3 ANSWER 091
- KA 000076A202
- ANSWER c
- REFERENCE iii,107 Technical Specifications 1.0, 3.6 TS.1-2, 3.6-1 1C19.1
' Containment System Integrity 4.3.4 4 P8180L-001
'2
. Containment System.
ANSWER-092
'
- KA.
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'
- REFERENCE B12A
CHEMICAL AND VOLUME CONTROL SYSTEM 3.1.1 11 P8172L-001a ~
CVCS IV.C.5.b 34 > 2 ANSWER 093
- KA 00WE03K202
- ANSWER b
- REFERENCE 1 E-1
BACKGROUND INFORMATION FOR LOSS OF REACTOR OR SECONDARY COOLANT Procedure 3 Step 14 P8197L-012
E-1/E-2 Review IX.D 54 5 ANSWER 094
- KA 00WE05K202
- ANSWER
- c
- REFERENCE 1 ES-1.1
BACKGROUND INFORMATION FOR POST LOCA COOLDOWN AND DEPRESSURIZATION Procedure 3 Step 8 P8197L-012
E-1/E-2 Review X.C.2, C.4 56-57 25 ANSWER 095
- KA 00WE06K102
- ANSWER b
- REFERENCE 1FR-H.1, Rev.10, RESPONSE TO LOSS OF SECONDARY HEAT SINK, NOTE 20, Step 12 f
P8197L-014
F/FR Review IV.D.2.C 30 17
-
i o
,
ANSWER 096
- KA 00WE08K303
' ANSWER d
- REFERENCE 1FR-C.2
Response To Degraded Core Cooling, Step 13 8 1FR-C.2-7 Background Information for Procedure Response To Degraded Core Cooling, Steps, 5,13 P8197L-014
F/FR Review iD.C.2.b.3),
21-22 13 D.1 ANSWER 097
- KA-00WE10K102
- ANSWER b
- REFERENCE 1FR-P.1, Rev. 8, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITION, Section 24, Page 14 P8197L-014
F/FR Review V.B.3.e 35-36 25,26
!
'
ANSWER 098
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t
- REFERENCE 1ES-0.3A
NATURAL CIRCULATION COOLDOWN WITH CRDM FANS, Step 17 8 P8197L-011
E-O Review
' E.16, 18 ANSWER 099
- KA 00WE12G420
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- REFERENCE 1ECA-1.2
LOSS OF EMERGENCY COOLANT RECIRCULATION, Step 24 13 P8197L-012 E-1/E-2 Review XIll 65-69 ANSWER 100
- KA 000011A112
- ANSWER a-
- REFERENCE 1ECA-2.1
UNCONTROLLED DEPRESSURIZATION OF BOTH STEAM GENERATORS, Caution 1, Step 2 P8197L-012
E-1/E-2 Review ~
V.D.2 35 6 i
r
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