IR 05000302/1997300
| ML20149H731 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/16/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20149H712 | List: |
| References | |
| 50-302-97-300, NUDOCS 9707250143 | |
| Download: ML20149H731 (127) | |
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U. S. NUCLEAR REGULATORY COMMISSION
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REGION 11 Docket Nos.:
50-302 License Nos.:
DPR-72 I
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Report Nos.:
50-302/97-300 Licensee:
Florida Power Corporation Facility:
Crystal River Nuclear Plant Unit 3
Location:
Crystal River, Florida Dates:
June 16-19,1997 Examiners:
Georg'e T. Hopper, fhief License Examiner Paul M. Steiner, License Examiner f
Approved by:
Themas A. Peebles, Chief Operator Licensing and Human Performance Branch Division of Reactor Safety Enclosure 1
9707250143 970716 PDR ADOCK 05000302 V
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A EXECUTIVE SUMMARY Crystal River Nuclear Plant Unit 3 NRC Examination Report No. 50-302/97-300 During the period June 16-19,1997, NRC examiners conducted an announced operator i
licensing initial examination in accordance with the guidance of Examiner Standards, NUREG-
1021, Interim Revision 8. This examination implemented the operator licensing requirements of i
10 CFR $55.41, &55.43, and $55.45.
Ooerations Four Senior Reactor Operator (SRO) candidates received written examinations and
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operating tests. The licensee administered the written examination on June 19,1997, and the NRC administered the operating tests on June 16-18,1997 (Section 05.1).
Three SRO candidates passed the examination. One candidate failed the written
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examination (Section O5.1).
Candidate Pass / Fail
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Pass
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75.0 %
Fail
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25.0 %
A weakness was identified in the licensee examination preparation activities (Section
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05.2).
The examiners concluded that candidate performance on the written examination was
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weak. Overall performance on the operating test was satisfactory. One generic weakness was noted in the area of Emergency Planning Event Classification (Section 05.3).
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Eggort Details Summarv of Plant Status j
During the period of the examinations Unit 3 was in Mode 5.
l. Operations OS, Operator Training and Qualifications i
05.1 General Comments NRC examiners conducted regular, announced operator licensing initial examinations during the period June 16-19,1997. NRC examiners administered examinations developed by the licensee's training department, under the requirements of an NRC security agreement, in accordance with the guidelines of the Exarniner Standards (ES),
NUREG-1021, Interim Revision 8. Four SRO upgrade applicants received written exa'minations and operating tests.
Three candidates passed the examination. One candidate failed the written portion of the exac : nation. Two of three candidates marginally passed the examination. One candidate was a marginal pass on the administrative portion of the operating test. The other marginally passed all three categories of the operating test. Candidates are considered to have marginally passed if they receive an unsatisfactory grade on any one administrative topic area, complete only 80 percent of the JPMs successfully, or receive a grade of 1.8 to 2.0 on any one competency during the dynamic simulator examinations. Detailed candidate performance comments have been transmitted under separate cover for management review and to allow appropriate candidate remediation.
05.2 Pre-Examination Activities a.
Scope The NRC reviewed the licensee's examination submittal using the criteria specified for examination development contained in NUREG 1021 Interim Rev 8.
b.
Observations and Findinas The licensee developed the SRO written examination, one JPM set, and three dynamic simulator scenarios for use during this examination. All materials were submitted to the NRC on time. NRC examiners reviewed, modified, and approved the examination prior to administration. The NRC conducted an on-site preparation visit during the week of June 2,1997, to validate examination materials and familiarize themselves with the details required for exam administration. No delays were encountered. However, significant effort was required to modify examination materials to meet the criteria set forth in NUREG 1021 Interim Rev. 8.
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(1) Written Examination Development The written examination was submitted on time and in a greatly improved forrnat l
when compared to the last submittal. The organization of examination and associated reference material expedited the exam review process. The chief examiner noted some improvement in question content and formatting since the last examination. Howeiter, the exam did not meet the criteria specified in the
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examiners standards and suffered from the same deficiencies that were
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identified on the last examination report 50-302/96-300 section 2.1. The following are some examples of the problems noted:
(a)
Eight questions contained infcrmation in the question stem that revealed the correct answer without tostng the desired concept or eadly eliminated one or more distractors.
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(b)
Nine questions contained distractors which were not plausible and easily
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eliminated.
(c)
Six questions contained distractors which were additional correct answers.
(d)
Six questions were considered simple knowledge items and would not discriminate a competent operator from an incompetent operator.
In addition, one question did not have a correct answer and another had to be changed when the chief examiner requested that expected plant response be validated on the simulator. Aside from minor editorial changes to clarify or improve the language of the questions, the kind and number of technical errors noted were excessive. The licensee worked diligently to resolve the NRC j
commenis. The final version of the written examination met the criteria specified i
in NUREG 1021 Interim Rev 8. The chief examiner determined that 59 percent of the questions were written at the analysis / comprehension level, j
One fundamental premise of the new examination development process assumed that licensee's could develop and submit an examination that would be more accurate and technically correct than an exam that could be developed by the NRO or one of its contractors. This process can only succeed if this premise is true. Licensee's have the advantage of greater site specific knowledge of plant systems and procedures and have been empowered to develop an examination of higher quality than has been produced in
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the past. Technical accuracy and adherence to the development guidelines are imperative to ensure that the examination is a valid means of measuring an operator's competence.
(2) Operating Test Development l
The licensee submitted eight Job Performance Measures (JPM) for the
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walkthrough portion of the examination of which three were administrative in i
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i nature. The examiners found the JPMs to be at the appropriate level of difficulty.
Seven of the eight were newly developed for this examination. Overall quality of the JPMs was good. However, the NRC found technical errors in some of the performance standards (expected operator actions) that should have been detected and eliminated via a thorough verification and validation process prior j
to the NRC's onsite preparation visit. Examples of these problems included incorrect calculation of a reactivity balance, incorrect classification of an event,
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and incorrect descriptions of valve operators and expected operator actions.
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These mistakes need to be screened and corrected early in the development process through an effective quality assurance review.
The simulator scenarios were challenging and designed to ensure that each candidate could be adequately evaluated on a majority of the items listed in 10 CFR 55.45(a). The examiners considered each scenario to be a satisfactory evaluation tool to measure the candidates' skills. The examinem noted that the quality assurance process was effective in this area since few W,anges were made to the contents of each scenario.
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Conclusion The NRC concluded that the quality of the examination materials had improved somewhat over the previous years results. However, greater emphasis needs to be placed on ensuring the technical accuracy of the examination materials. This is
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05.3 Examination Results and Related Findinas. Observations. and Conclusions a.
S_qnpa The examiners reviewed the results of the written examinat;on and evaluated the candidates' compliance with and use of plant procedures during the simulator scenarios and JPMs. The guidelines of NUREG-1021, Forms ES-303-3 and ES-303-4,
" Competency Grading Worksheets for integrated Plant Operations," were used as a basis for the operating test evaluations.
b.
Observations and Findinos.
The examiners reviewed the results of the written examination and found five questions where 75 percent of the candidates chose incorrect answers (Nos. 12,16,46,47,and 75), and one question that all candidates missed (No. 69). This indicated a gerieric weakness in the subject matter tested. These items should be reviewed to determine if l
training deficiencies exist. In addition, the median score of 83 percent on the exam was I
comparatively low. Overall performance of the candidates on the written exam was
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weak.
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Examiners identified several weaknesses in candidate performance during the operations portion of the examination. Details of the discrepancies are described in each individual's examination report, Form ES-303-1, " Operator Licensing Examination
Report." One generic weakness was nuted in the area of event classification. Three of four candidates incorrectly upgraded a Site Area Emergency to a General Emergency.
During the simulator portion of the exam, one candidate exhibited a weakness in the usage of Emergency Operating Procedures (EOP) when he disregarded the guidance of Procedure EOP-03 and inappropriately took actions that departed from the procedure / license conditions. The candidate was not aware that he was operating outside the bounds of license conditions.
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Conclusion i
The examiners concluded that candidate performance on the written examination was
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weak. Overall performance on the operating test was satisfactory. One generic weakness was noted in the area of Emergency Planning Event Clanification.
Miscellaneous Operations issues (92901)
08.1 Closed Violation 50-302/93-16-07: Inadequate EOP and AP Procedures. This item concerned multiple examples of a violation of 10 CFR 50, Appendix B, Criterion V. Five procedures associated with this item had been reviewed and closed out in inspection report 50-302/96-04. The inspector reviewed the remaining procedures cited in the violation, AP-581, Loss of (Non-Nuclear Instrumentation power supplies) NNI-X, and AP-582, Loss of NNI-Y and the Request for Engineering Assistance (REA) No. 950406.
The inspector verified that the REA had been completed. Both procedures for the loss of NNI power have been rewritten to include a list of valid instrumentation vice inoperable instrumentation. The inspector found the rewrite of the procedures and the verification of technical accuracy acceptable. The licensee's corrective actions were satisfactory.
08.2 Closed Violation 50-302/96-04-02: Failure to take prompt corrective action in revising procedure VP-580, " Plant Safety Verification." The Procedure had not been updated since September,1993 and contained outdated and incorrect information. The inspector noted that the procedure had been promptly revised on May 3,1996 to correct the discrepancies. The procedure has since been deleted (3/3/97) and new STA guidance was issued in Administrative Instruction Al-505," Conduct of Operations During Abnormal and Emergency Events." The licensee imposed additional requirements in the administrative instruction governing procedures and procedure changes, "Al-400F",
to include prompt resolution of significant proceduralinadequacies. The inspector determined that the licensee's corrective actions were satisfactory.
08.3. Closed Violation 50-302/96-10-01: Four examples of failure to follow refueling procedure FP-203. The inspector reviewed documentation related to the licensee's corrective actions. The inspector found that individuals involved in the event had been counseled and the entire operations staff had received training on lessons learned from the event during the month of March,1997. Refueling Procedure FP-203, "Defueling Enclosure 1
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and Refueling Operations", was revised and expanded to include additional requirements, greater detail, and specific actions to be taken in response to a significant fuel handling event. The improvements should strengthen the defense-in -depth strategy for preventing fuel handling mishaps, assuming verbatim compliance, and provide for greater individual accountability. The inspector noted that FPC's response to i
the Notice of Violation, dated March 21,1997, addressed FFC management's
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expectation concerning use of words of discretion such as, "should", in procedures. In
short, this expectation was that use of the word "should" meant "shall". At the time of
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the inspection, the revised Procedure FP-203, dated January 7,1997, still contained instances of the use of the word should. The licensee had committed to revising
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refueling p*ocedures to remove or clarify any steps governing fuel movement that imply discretion where none was iraended in the letter mentioned above. The inspector i
verified per phonecon with the licensing manager on July 9,1997, that the procedure l
had been revised to eliminate the use of the word from the procedure. The inspector concluded that the licensee's corrective actions were satisfactory.
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08.4 _QJosed Violation 50-302/96-10-02: Failure to assure root cause analysis and corrective
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actions taken to preclude repetition were adequate after the refuelincident. The inspector reviewed the corrective actions taken by the licensee. The inspector noted
that the Vice President, Nuclear Operations had counseled operations managers ~about their unacceptable performance regarding this event. Operations management
conducted training emphasizing the significance of events, importance of log keeping i
and implementing prompt corrective actions in March,1997. Procedure FP-203 was
modified and improved as noted above to provide additional clarity and guidance. In
addition, the corrective action program was modified as contained in Procedure CP-111,
" Processing of Precursor Cards for Corrective Action Program", to include detailed
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. guidance for the development of root and apparent causes and corrective action plans.
The inspector found the licensee's corrective actions were satisfactory.
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V. Manaaement Meetinas i
l X1. Exit Meeting Summary
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At the conclusion of the site visit, the examiners met with representatives of the plant staff listed on the following page to discuss the results of the examinations. Dissenting comments were not received from the licensee. No proprietary information was
identified.
PARTIAL LIST OF PERSONS CONTACTED s
Licensee i
- J. Baumstark, Director Quality Programs
- M. Gallian, Nuclear Operations instructor
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- B. H!ckle, Director, Nuclear Plant Operations
- J. Lind, Manager, Nuclear Operations Training
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- C, Pardee, Director Nuclear Plant Operations
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- W. Pike, Regulatory Assurance
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- J. Smith, Operations Training Supervisor
- J. Springer, Supervisor, Nuclear Simulator Training
- T. Taylor, Director Nuclear Training
- l. Wilson, Manager Nuclear Plant Operations
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S, Cahill, Senior Resident inspector-
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INSPECTION PROCEDURES USED lP 92901 Followup - Operations l
ITEMS OPENED, CLOSED, AND DISCUSSED
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-Opened I
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None-Dd l
i 50-302/93-16-07 VIO Inadequate EOP and AP Procedures.
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l-50-302/96-04-02 VIO Failure to take prompt corrective action in revising procedure VP-580, " Plant Safety Verification".
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50-302/96-10-01 VIO Four examples of failure to follow refueling procedure FP-203.
l 50-302/96-10-02 VIO Failure to assure root cause analysis and corrective actions taken to preclude repetition were adequate after the refuel incident.
Discussed I
None LIST OF ACRONYMS USED AP Abnormal Procedure
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CFR Code of Federal Regulations L
EOP Emergency Operating Procedures ES Examince Standards L
JPM Job Performance Measures i
NNI Non-Nuclear instrumentation NRC Nuclear Regulatory Commission
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REA Request for Engineering Assistance RO Reactor Operator SRO Senior Reactor Operator
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i SIMULATION FACILITY REPORT l'
Facility Licensee: Florida Power Corporation
Facility Docket No.: 50-302 J
Operating Tests Admidstered on: June 16-18,1997 ll
'.This form is to be used only to report observations, These observations do not constitute audit or inspectirn findings and are not, without further verification and review, indicative of
' noncompliaitee with 10 CFR 55.45(b). These observations do not affect NRC certification or--
L approval of the simulation facility other than to provide information that may be used in future
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evaluations. No licensee action is required in response to these observations.
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l-While conducting the simulator portion of the operating tests, the following items were observed
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i (if none, so state).
L ITEM pfSCRioTION
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Enclosure 2 i:
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l ENCLOSURE 3
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/t(as/m 7 7 - 3 dd
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U.S. NUCLEAR REGULATORY COMMISSION SITE-SPECIFIC WRITTEN EXAMINATION
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APPLICANT INF6RNATION d
Name:
Region:
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Date:
Facility / Unit:
Crystal River 3
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License Level:
SR0 Reactor Type: BW
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Start Time:
Finish Time:
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Instructions
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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
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grade of at least 80%.
Examination papers will be collected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the examination starts.
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Applicant Certification
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I All work done on this examination is my own.
I have neither given nor received aid.
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Applicant's Signature
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Examination Value 100.00 Points Applicant's Score Points Applicant's Grade
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NRC Rut'ES GENERAL GUIDELINES 1.
Cheating on any part of the examination will result in a denial of your application'and/or action against your license.
2.
If you have any questions concerning the administration of any part of
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the examination, do not hesitate asking them before starting that part of the test.
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3.
SRO applicants will be tested at the level of responsibility of the
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senior licensed shift position.
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4.
You must_ pass every part of the examination to receive a license or to i
contiwje performing license duties. Applicants for an SRO-upgrade license may require remedial training in order to continue their R0
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duties if the examination reveal deficiencies in the required knowledge and abilities.
5.
The NRC examiner is not allowed to reveal the results of any part of the examination until they have been reviewed and approved by NRC
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management. Grades provided by the facility licensee are preliminary until approved by the NRC.
You will be informed of the official
examination results about 30 days after all.the examinations are complete.
NRC WRITTEK EXAMINATION GUIDELINES
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1.
After you complete the examination, sign the statement on the cover sheet (in black ink) indicating that the work is your own and you have not received or given assistance in completing the examination.
2.
To pass the examination, you must achieve a grade of 80% or greater.
Every question is worth one point.
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3.
There is a time limit of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for completing the examination.
4.
Use only dark pencil to ensure legible copies on the answer sheets.
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5.
_ Print your name in the blank provided on the examination cover sheet and the answer sheet.
You may be asked to provide the examiner with some form of positive identification.
6.
Mark your answers on the answer sheet provided and do not leave any-question blank.
7.
If the intent of a question is unclear, ask questions of the examiner only.
8.
Restroom trips are permitted, but only one applicant at a time will be allowed to leave.
Avoid all contact with anyone outside the examination
room to eliminate even the appearance or possibility of cheating.
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9.
When you complete the examination, assemble a package including the examination questions, examination aids, answer sheets, and scrap paper and give it to the examiner or proctor.
Remember to sign the statement on the examination cover sheet indicating that the work is your own.and that you have neither given nor received assistance in completing the examination.
The scrap paper will be disposed of immediately after the~
examination.
10.
After you have turned in your examination, leave the examination-area as defined by' the examiner.
If you are found in this area while the
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examination is.still in progress, your license may be denie'd or reveked.
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11.
Do you have any questions?
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1. rot-5-38'002/G8//NTS/2.1.29//3.3/77/01-07 l
During-a valve alignment an uncapped manual drain valve under the "A" makeup demineralizer has to be checked closed.
How may this valve position verification be performed?
j A.
A " hands-on" valve position verification must be performed.
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B.
Makeup drain valves are sealed, verification is made by
visual contact ensuring the seal is in place.
C.-
The "as-left" position found on the most recent clearance may be used.
vD.
. A visual check for leakage may be performed if the system is pressurized.
l Reasons:
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A " hands-on" verification is not required for inaccessible high radiation areas.
B.
Makeup drain valves are not sealed closed; seal verification requires physical contact.
C.
This is not an accepted practice for valve position veri fi cation.
NEW; OI-07 page 5
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NRC97.TST Version: O Page: 1
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2. rot-4-60 003/88//0040101015/011A203//3.9/33/RC l
The plant experienced a loss of subcooling margin during a
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small break loss of coolant accident (LOCA).
Subcooling margin has been restored but the operators are experiencing difficulty maintaining reactor coolant (RCS) pressure with the pressurizer.
What could be the cause of this problem?
Al Steam generator heat removal is.too. slow.
B.
High presstre injection is not making up for the water exiting the break.
C.
A hard bubble was created in the pressurizer while there was less than adequate subcooling margin.
vD.
The pressurizer heaters may not be energized.
Reasons:
A.
Steam generator heat removal does not control RCS pressurizer control.
B.
High pressure injection would recover level and allow pressure control.
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Hard bubble formation would not have occurred from low level in the pressurizer.
NEW; ROT-4-60 page 6; 0000501002 l
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3 '.. ROT-5-97 001/B6//0000501021/00074EK205//4.1/33/EoP-07
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'EOP-07, Inadequate Core Cooling, Step 3.9 states:
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IF at any time, BWST level is < 15 ft THEN transfer ECCS pump suction from BWST to RB sump, p
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Complete tPis transfer prior to reaching'7 ft BWST l
level.
In'the details of this step there is guidance.to throttle flow
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rates for the low pressure' injection (LPI) pumps to 1800-2200
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gpm each.
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i-Why are the flow rates for LPI throttled when transferring
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suction to the reactor building (RB) sump?
The lower flow rates are required due to:
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The vortexing potential ind reduced available net positive suction head s'
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B.
The increased flow resti-ions.in the RB sump and associated piping.
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Prevent the possibility of exceeding pump run-out when i
the transfer is complete.
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A thermal shock concern in the decay heat exchangers as the hotter RB sump water is introduced into this component (s).
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Reasons:
B.
The loss of NPSH if not due to flow restrictions but a lower elevation of the RB sump.
C.
Pump run out is not a concern when transferring to a sump-suction.
D.
Lower flow rates are not a heat transfer concern.
NRC97.TST Version: 0 Page: 3
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3c ROT-5-97 001/ B6/ / 0000501021/ 00074 EK205// 4.1/ 33/ EOP-07 BANK; ROT-5-97 3; ROT-5-97 pages 9 & 10
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' 4' ' rot 4-60 002/ B3// 0020101009/ 010K603// 3.6/ 33/ RC
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While at power, a'feedwater transient caused-a rapid out-surge
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of_ the pressurizer followed by a rapid. in-surge.
Due to a L
pressurizer heater control problem all heaters are in manual-j Land off.
The following conditions exist:
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Pressurizer temperature is 630*F.
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i Reactor power is 100%.
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T,y, is 579'F and stable.
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Describe the pressurizer's response to these conditions when equilibrium is achieved.
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A.
Pressurizer pressure will increase until the spray valve i
opens.
S B.
Pressurizer pressure will stabilize at the current
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value.
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C.
Pressurizer pressure will decrease until the plant trips on low pressure at 1800 psig.
vD.
Pressurizer pressure will decrease until the plant trips on variable low pressure.
Reasons:
A.,
B., & C.
The pressurizer at this temperature will cause pressure to decrease to 1905 psig.
A variable low pressure trip ~will occur.
NEW; ROT-4-60 pages 6-10; ROT-4-60 B4; ROT-4-12 page 8 NRC97/TST Vension: 0 Page: 5
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5. rot-4-25 001/36//0720401001/072K403//3.6/33/RM
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F Control complex ventilation was in the following alignment
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before the gas channel of RM-A5 went into high alarm:
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Control complex supply fan AHF-17A - running
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Control complex return fan AHF-19A - running
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Controlled access area exhaust fan AHF-20A - running in slow
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Sampling hood exhaust fan AHF-44A - running
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Control complex relief fan AHF-21A - running
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, Controlled access area chemical laboratory supply. fan AHF-30 running j
Which fans are still running after the RM-A5 actuation?
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A.
.AHF-17A, AHF-19A, AHF-21A
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AHF-20A,.AHF-30, AHF-44A
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C.
AHF-17A, AHF-30, AHF-44A
D.
AHF-19A, AHF-20A, AHF-21A
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Reasons:
f With an RM-A5 actuation the only fans still running from their original configuration are AHF-20A,' AHF-30 and AHF-44A; making A., C., and D. incorrect.
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NEW; ROT-4-25 page 28; ROT-4-87 pages 7-12 NRC97.TST Version: 0 Page: 6
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6. ROT-4-10 003/ 89//1190301015/ 00033AK301// 3.6/11/ TS A reactor startup is in progress.
While increasing power the following indications are noted:
NI-1 1 x 106 cps
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NI-2 8 x-105 cps
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NI-3 1 x 10-11 amps
- 'NI-4 2 x 10-9 amps
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What action (s), if any, must be taken?
.
vA.
NI-3 is inoperable; restore NI-3 to operable status prior to 5% full power.
-
.
B.
NI-4 is inoperable; restore NI-4 to operable status prior to 5% full power.
C.
Both NI-3 and NI-4 are inoperable; restore both NI-3 and NI-4 or open control rod drive breakers within one hour.
D.
Neither NI-3 nor NI-4 are inoperable; no limitations to power exist.
Reasons:
B.
NI-4 is operable when compared to NI-1 and NI-2 readings.
C.
NI-4 is operable when compared to NI-1 and NI-2 readings.
D.
NI-3 is inoperable and must be restored to operability prior to mode 1.
NEW; ROT-4-10 page 3; Technical Specifications page 3.3-24 NRC97.TST Version: 0 Page: 7
_
7. rot-5-01003/A10//3410103036/2.4.3//3.8/55/TS Post accident monitoring instrumentation requires 2 channels of radiation monitoring.
Which instruments meet this requirement?
.
N A.
Containment monitor "A" D-ring, RM-G29, and reactor building monitor RM-A6.
Bl Containment monitor fuel handling bridge, RM-G16,'and.
containment monitor incore instrument ' room, RM-G18.
C.
Containment monitor near personnel hatch, RM-G17, and cor.tainment monitor "B" D-ring, RM-G30.
vD.
Containment monitor "A" D-ring, RM-G29, and containment
. monitor "B" D-ring, RM-G30.
Reasons:
A.
RM-A6 does not have the range for PAM instrumentation.
5 B.
Neither RM-G16 or RM-G18 have the range for PAM instrumentation.
{
C.
RM-G17 does not have the range for PAM instrumentation.
.
NEW; 3410103037; Technical Specifications page B3.3-133
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8. rot-4-81001/B6//0780401001/079A201//3.2/33/IA L
What is the expected response of IAV-30, IA/SA cross-connect,
!
L and SAV-6,'IAV-30 bypass' check valve, to a loss of instrument L
. air?
L l
'
L
L A.
When pressure decreases below 80 psig IAV-30 and SAV-6
'
will.close to attempt to re-pressurize instrument air.
.
.
,
vB.
When pressure ' decreases below 80 psig IAV-30 -will-close and SAV-6 will open to attempt to re-pressurize l
instrument air.
C.
When pressure decreases below 80 psig IAV-30 and SAV-6 will open to attempt to re-pressurize instrument air.
.
D.
When pressure decreases below 80 psig IAV-30 will open and SAV-6 will close to attempt to re-pressurize
)
instrument alr.
I Reasons:
i A.
SAV-6 is a check valve that opens and IAV-30 closed to try and repressurize the instrument air system.
y C.
IAV-30 closes to try and repressurize the instrument air system.
D.
IAV-30 will close and SAV-6 will open to try and repressurize instrument air system.
,
BANK; ROT-4-8115; ROTS J - Final 96; ROT-4-81 page 7 l
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l 9. rot-4-07 002/ G5/ rot-5-01/ 0860104003/ 00037AA214!/4.3/ 77/ FS
While performing an~ inspection of the fixed water spray fire
'
extinguishing system at the step-up transformers, the fire i
' specialist finds several nozzles plugged by pieces of. rusted metal.. A'll fixed water spray systems are declared inoperable
- until inspections are made.
Besides all the transformers, what plant equipment does this leave unprotected and what action (s).
must be taken?
.
.
,
A.
Charcoal filters, seal oil unit, feedwater pumps, lube
_ oil storage tank; restore the inoperable equipment
-
within 14 days or submit special report to the.NRC.
.
!
B.
Charcoal filters, seal oil pumps, feedwater booster pump
'
. consoles, lube oil storage tank; establish both backup l
fire suppression equipment and a continuous fire watch.
[
vC.
Charcoal filters, seal oil unit, feedwater pump consoles, lube oil storage tank; establish both backup
fire suppression equipment and a continuous fire watch.
5 D.
Charcoal filters, seal oil unit, feedwater pump consoles, hydrogen storage tank; restore the inoperable
.
equipment within 14 days or submit special report to the'
l NRC.
.
I Reasons:
!
A.
The feedwater pumps are not protected by fixed water spray;
.
restoration is as soon as possible.
.
B.
The booster pump consoles are not protected by fixed water
,
spray.
D.
The hydrogen storage tank is not protected by fixed water
.
l
. spray; restoration is as soon as possible.
'
NEW; ROT-5-01 A1; ROT-4-07 pages 9 and 31; Fire Protection Plan
'
page 99
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10.- rot-4-07 001/A2//3440403010/086K301//3.2/55/FS Appendix R sprinklers have failed to extinguish a fire in the
'
southernfend of the seawater room.
AHF-15A, decay heat closed cycle cooling (DC)' pump motor air handling fan, has tripped due to high temperature.
What action will ensure that the
redundant make-up pump (MUP) remains operable?
j
.
vA:
Ensure MUP-1C on SW cooling.
~
B.
Ensure MUP-1A on DC cooling.
C.
Ensure MUP-1C on DC cooling.
,
D.
Ensure MUP-1A on SW cooling.
.
Reasons:
,
B.
Both DCPs are inoperable due to the fire.
L C.
DC cooling is the normal c.ooling water supply for MUP-1C.
Both' DCPs are inoperable due to the fire.
D.
SW cooling is the normal cooling water supply for MVP-1A.
s NEW; ROT-5-69 pages 5 & 6; OP-880 page 73; ROT-4-86 pages 5 &
j 6; ROT-4-55 page 4 l
NRC97.TST Version: 0 Page: 11
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11. ROT-4-13 002/ B1// 0130101003/ Oi3K4iO/ / 3.7/ 33/ ESAS The plant is conducting a heatup with the reactor coolant system at 1200 psig and 400* F, low pressure injection (LPI) has been reset, and high pressure injection (HPI) is still bypassed.
Due to a steam leak, RB pressure is approaching 4 psig.
Which of the following is correct?
A '.
RBIC may be bypassed prior to actuation since the leak is on the secondary system.
B.
RBIC may be bypassed after actuation.
This will result in an LPI actuation but not HPI.
VC.
RBlC can not be bypassed prior to actuation and will
. actuate when RB pressure exceeds 4 psig.
This will
result in an actuation of both LPI and HPI.
J D.
RBIC can not be bypassed prior to actuation and will therefore actuate when RB pressure exceeds 4 psig.
No other ES actuations will occur.
Reasons:
'
A.
RBIC cannot be bypassed until actuation.
B.
RBIC cannot be bypassed until actuation.
HPI will actuate
,
from the cascade.
D.
RBIC cannot be bypassed unti.1 actuation.
When it actuates it will cascade actuating HPI and LPI.
BANK; ROT-4-13 28; ROTS J - T10B; ROT-4-13 pages 15 & 32 NRC97.TST Version: 0 Page: 12
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.12. rot-4-26 001/F2//0340101003/034K402//3.3/88/FH Which of the following conditions requires a mandatory core slow zone when the mast is being lowered?
.
!
l A.-
Grapple engaged, grapple elevation 16 feet above the fuel in the core.
!
~
'
VB.
Grapple engaged, grapple elevat'i6n 16 feet above the
-
l bottom of the cobe.
'
a
,
C.
. Grapple disengaged, grapple elevation 16 feet above the fuel in the core.
j j
D.
Grapple disengaged, grapple elevation 16 feet above the bottom of the core.
'
Reasons:
l
.A.
With the grapple engaged the slow zone start 15' feet above
,
l the top of the core.
i i
C.
With the grapple disengaged the slow zone start 1 foot above the core.
D.
With-the grapple disengaged the slow zone start 1 foot above the core or approximately 13 feet from the bottom of the core.
!
i
1
!
NEW; ROT-4-26 F3; ROT-4-26 page 4
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NRC97.TST Version: 0 Page: 13
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L, 13. ROT-5-01002/A9//3420203018/2.2.12//3.4/44/TS Tha ' containment'. Sump Monitor-is. inoperable and Condition A of L
LCO:3.4.14 was entered-at 1100 on Tuesday.
The specified i
surveillance was completed at ~1800.
. hen must the surveillance W
be' performed again?
esi = k4e st l~
L
'A.
1800'on Wednesday B.'
2100 on Wednes' day
)
>
vC.
'2400 on Wednesday
. ;
D.
0300'on Thursday
,
l
l r
Reasons:
i l
A.
A later time is aTlowed using the 25% extension.
i B.
A'later time is-aTlowed using the 25% extension.
!
D.
This exceeds the 25% extension.
BANK; ROT-5-0175; Technical. Specifications page 3.0-4
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- 14. rot-G-38 003/A9//NTS/22.20//3.3/55/c1-07 The plant is being shutdown for a refueling.
The reacto-
- .
coolant. temperature is 130'F.
Which of the following systems requires prior. approval from the Manager Nuclear Plant Operati~on (MNPO) for troubleshooting activities?
-
i-
!
i I:
A.
The electro-hydraulic control ' circuit for turbine
-
.oyerspeed.
.
!
v8.
The automatic closure interlock circuit.
i
'l l
C.
The reactor protection system shutdown bypass circuit.
'
E
l D.
The; emergency feedwater initiation circuit.
-
Reasons:
A.
Approval 'is only required in modes 1-4.
C.
Approval is only required in modes 1-4.
D.
Approval-is only required in modes 1-4.
NEW; OI-07 page 3; NOD-22. page 1; Technical Specification page 1.1-8
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' 15. rot.5-42 001/B3//NTS/2.1.6//4.3/33/10CFR Following a'LOCA,.the core is being cooled using long term post-accident alignment.
Reactor building spray has actuated.
,
,
After the ES pumps'are aligned to the reactor building sump, flows and motor current begin to oscillate.
The Technical j
,
l Support Center (TSO has devised a plan to continue-core l
cooling and reactor building spray.
This plan is not L
proceduralized and cross-connects ES trains.
What is your l
responsibility as the emergency coordinator?
)
.
,
VA.
Review and approve the proposal prior to plan implementation.
i l
B.
Obtain NRC approval prior to plan implementation.
l l-C.
Obtain Shift Supervisor approval prior to plar implementation.
D.
Ensure a work instruction is generated prior to plan implementation.
,
I
Reasons:
!
j B.
NRC approval is not needed to implement 10CFR50.54 X and Y.
C.
The EC does not need the approval of the shift supervisor.
.
D.
10CFR50.59 review is required to use a work instruction.
l l
l
NEW; 10CFR50.54 X & Y pages 674-675 l.
!
L
!
l
f NRC97.TST Version: O Page: 16
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16. rot-4-09 002/B6//OS60401002/00057AA106//3.5/33/ICS The following indications are observed:
!
-
All lights on the NNI-Y power supply monitor are brightly lit.
l
-
The white _ indicating light for NNI-Y power on the redundant l
instrumentation panel is out.
-
K-2-1, "NNI Y POWER FAILURE", annunciator is in alarm..
How will these indications affect condensate pump control.?
.
l
,
VA.
Condensate pump speed can be controlled by taking manual
l control at the pump hand / auto stations.
B.
. Control of the condensate pumps must be transferred to the local controllers.
C.
Condensate pump speed can be controlled by taking manual control at the master hand / auto station.
D.
Control of the condensate pump speed is not affected automatic control may continue.
i Reasons:
i B.
Local control is not required for this situation.
i C.
The back-up power allows use of the indiv.idual hand / auto stations in manual only.
D.
Automatic control of the condensate is not possible with loss of NNI-Y.
)
i
MODIFIED BANK; ROT-4-14 7; ROTS J - T9; ROT-4-09 pages 25-28, 31 & 32; ROT-5-82 pages 1, 2, 4 & 5.
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17. ROT-3-03 001/BS//0000501009/00011EK101//4.4/33/NAT CIRC
'
Following a large break loss of coolant accident the reactor coolant system (RCS) pressure is 435 psig.
-
(OTSG) pressure is 140 psig.
What should the hot and cold-(Th and T ) leg temperatures be if boiler-condenser (reflux) boi' ling c
were established?
l
'
T is 468'F; T is 428'F.
.
.
h c
,
,
B.
T is 458'F; T is 428*F.
h c
vC.
T is'458'F; T is 360*F.
h c
D.
T is 448'F_; T is 360*F.
h c
Reasons:
T should be saturation temperature for RCS pressure; T hot cold should be saturation temperature for OTSG pressure.
.
A.
T is superheated.
h B.
T is not coupled to the OTSG.
c
D.
T is subcooled.
h MODIFIED BANK; ROT-3-03 26; ROT-3-03 pages 7 & 8 i
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NRC97.TST Version: 0 Page: 18
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P 18. rot.5-96 002/A2//3443403G01/E02EK2.1//4.0/44/EoP-02
',
= A. follow-up step 'in E0P-02, Vital System Status Verification,
states:
,
~- 3. 7 IE RCS ' Tave is > 555'F, THEN ensure OTSG PRESS controlling:
!
I at = 1010.- (980' to :1040) psig.
?
I What is the p'urpose-of.this step?!
'
'
.
.
.
i i-f.
VA.
Limit the. out surge of Lthe pressurizer due to a'n RCS cooldown.
,
[
B.
Ensure steam flow within capability of the turbine
' bypass valves.
4 i
C.:
Prevent challenging the main steam safeties.
-
I D.
Stabilization of plant conditions to determine further
,
!'
symptoms.
i.
i
Reasons:
'B,
' Post: trip conditions are well within the capability of the.
turbine bypass - valves.
C.
The MSSVs are seated prior.to this step unless
'
malfunctioning.
y
.
D.
Other symptoms may cause trouble performing this' step.
- NEW; ROT-5-96 pages 15 & 16; ROT-4-66 page 6 j
~l
)
NRC97.TST Version: 0 Page: 19
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19. ROT-5-101002/ A2/ / 3440403001/ 00038EK306/ / 4.1/ 33/ EOP-6 The plant is conducting a forced flow cool down due to a steam generator tube leak.
E0P-6 directs the operator to maintain reactor coolant pressure as low as possible while meeting adequate subcooled margin requirements.
What is the preferred method for accomplishing this step and, why is it performed?
v'K.
Spray the pressur.izer; minimizes the primary to secondary leakage.
B.
Spray the pressurizer; minimizes steam generator tube stresses.
C.
Opening the PORV; minimizes the primary to secondary leakage.
D.
Opening the PORV; minimizes steam generator tube stresses.
Reasons:
B.
Lower pressure does not affect tube stress.
C.
The lower reactor coolant pressure may or may not improve cooldown control depending on the conditions.
D.
Reducing reactor coolant pressure will not lower the feedwater. flow requi rements.
MODIFIED BANK; ROT-5-3 >1 18; ROTS J - Final 96; ROT-5-101 pages 29 & 30 i
NRC97.TST Version: 0 Page: 20
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20. rot.5-85 002/A1//3440403001/00009EA201//4.8/55/EoP-03 The plant experienced a loss of subcooling margin due to a loss of= coolant accident-(LOCA) inside containment.
The following j
plant conditions exist.
s
-
High pressure injection?(HPI) and reactor building isolation
and cooling'(RBIC) have actuated.
'
,
!
Emergency feedwater ' initiation and control (EFIC) has j
_
.
-
ac'tuated.
, Steam generator levels 'are progressing toward.their required
!
setpoint.
30 minutes into the transient reactor ~ coolant (RCS) pressure-(900 psig) and temperature begin to increase up the saturation-line.
What should you do in this situs 'an?
l A.
Continue in EOP-3, Inadequate Subcooling Margin, it contains the necessary guidance for these conditions.
!
B.
Go to E0P-7, Inadequate Core Cooling, beginning with'
step 3.1.
C.
Go to E0P-8, LOCA Cooldown, beginning with step 3.L.
vD.
Go'to E0P-4,. Inadequate Heat Transfer, beginning with Step 3.1.
i Reasons:
A.
The guidance is not in E0P-3 to.stop the heatup at this l
point.
'
!
l
!
B.
If the system is still saturated the conditions are not met to enter E0P-7.
l L
C.
Since LPI flow does not exist the branch point to E0P-8 l
should not be taken.
,
i
.
l l
NRC97.TST Version: O Page: 21
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l 20. ROT 5-85 002/ Ai//3440403001/00009EA201//4.8/55/ EOP-03
'
NEW; ROT-5-85 pages 23-28 f
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i l
i NRC97.TST Version: 0 Page: 22
21. noT-41a 002/as/ sot 1-4w0120101cos/oco29EK103'/3.8/33/ DSS
,
With ICS in manual an inadvertent dilution is in progress.
.
What protects the core if RPS fails to initiate?
A.
The AMSAC system will trip the turbine and initiate emergency feedwater.
B.
Tha' AMSAC system wilt trip the regulating rods.
C.
The DSS system will trip the turbine and initiate emergency feedwater.
vD.
The DSS system will trip the regulating rods.
i
Reasons:
A.
AMSAC does perform this action but the trip would be the result of a positive reactivity addition / pressure spike not a feedwater/ power mismatch.
B.
AMSAC does not trip any control rods.
C.
DSS does not trip the t, cbine and initiate emergency feedwater.
,
(
NEW; ROT-4-12 pages 43, 44, 46 & 47; ROT-5-67 pages 3 & 6 NRC97.TST Version: 0 Page: 23
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' 22. rot 490 001/ B7//0620401003/002K201//3.4/33/ AC The' plant is in mode 4.
The auxiliary building operator is performing the "B" diesel generator (EDG) operability test
,
(SP-3548). 'The diesel has been running for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
What two R
power supplies could be paralleled. on the' "A" 4160v ES-bus?
,
t l
.
,
.
A.
Back-up ES transformer and auxiliary transformer, j
~
,
i
'
.
.B.
Back-up ES transformer and the
"A" EDC.
.
J y
.
'
t vC.
Off-site transformer and the "A" EDG.
,
l-D.
Off-site, transformer and the auxiliary transformer.
I
!.
.
,
'
Reasons:
!
}
A.
The auxiliary transformer feeds to the ES busses are normally out of service.
j
B.
Crosstie-blocking would prevent this breaker alignment.
r f
j D.
The' auxiliary transformer feeds to the ES busses are
normally out of service.
,
l
!
MODIFIED BANK; ROT-4-90 4; ROTS J - T10A; ROT-4-90 pages 16, 17
& 27 I
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NRC97.TST Version: 0 Page: 24
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2 3. ' rot-5-95 001/ Aill3440403001/ E08EK3.2// 3.6!441EoP-08
,'
During1a. loss of: coolant accident (LOCA)' coincident with a loss
-
1 -
of off-site' power (LOOP) the following conditions have been established:'
r
-
The. reactor has tripped.
- :Both-steam generators (OTSGs) are'at 300 psig-.
'
. All' secondary system components and piping are intact.
i
-
The emergency feedwater tank and the condensate storage tank
,
i are empty.
'
~
-
Hotwell level is 8 feet.
'
'
Both atmospheric dump valves are closed and manually
-
' isolated due to.small seat ' leaks.
'
.You reach a -branch point in EOP-08,.LOCA Cooldown, that asks if
]
steam generators are available for heat transfer.
How would
'
you respond?
,
.
- =
!
l A.-
OTSGs are not available for heat transfer due to main
. steam line. and main feedwater isolation at 600 psi.
!
B.
0TSGs are not available for heat transfer; no steaming
^
path is available.
,
'
vC.
OTSGs are available for heat transfer; emergency
feedwater can be fed from the hotwell.
'
D.
0TSGs are -available for heat transfer; auxiliary feedwater can be fed from the hotwell.
r o
Reasons:
,
,
A.
The steam generators can still function with emergency L
feedwater at steaming through the atmospheric dump valves.
B.
The steam generators can still steam through the
,
atmospheric dump valves.
.
.
D.
During a LOOP power is not available to the auxiliary feedwater pump.
4 NRC97.TST Version: 0 Page: 25
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2 3. ROT 5-95 001/ A1/ / 3440403001! E08EK32// 3.6/ 44/ EOP-08 NEW; ROT-5-95 pages 22 and 23
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l 24. rot.5 48 002/A2//3410103029/073K401//4.3/44/ WASTE l
An evaporator condensate storage tank (ECST) is being released.
l The release is' terminated by a high alarm on RM-L2.
What l
action should be performed?
,
,
rA.
Verification that release isolation valves WDV-891 and 892 are closed and call chemistry for RM;-L2 evaluation.
,
B.
Verification that ' release isolation va'lve SDV-90 is closed and call chemistry for RM-L2 evaluation.
1:.
Verification that release isolation valves WDV-891 and 892 are closed and call chemistry for RM-L2 flush s amp.l e.
D.
Verification that release isolation valve SDV-90 is closed and call chemistry for RM-L2 flush sample.
l
.
Reasons:
B.
SDV-90 is not' in the release path for ECST releases nor is
it closed by an.RM-L2 actuation.
SDV-90 is used for releases from the secondary side, SDT-1.
C.
Chemistry is directed to check the operability of the radiation monitor not to take samples from it.
D.
SDV-90 is not in the release path for ECST releases nor is it closed by an RM-L2 actuation.
Chemistry is directed to check the operability of the radiation monitor not to take sar.ples from it.
.
NEW; ROT-5-48 page 6; ROT-4-25 page 29; OP-407A page 4 L
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NRC97.TST Version: 0 Page: -27
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2 5. rot-5-96' 001/ B6// 0000501025/ 00024EK302/4.2/4.4/ 33/ EoP-2 The plant has experienced a reactor ' trip.
While completing the follow-up actions of E0P-2, Vital Safety System Verification,
,
you _ observe that control. rod 1 in group 1 and control rod 3 in
group 6 do not have their in-limit.or 0% lights energized.
API-
~
-
shows both control rods at about 20% withdrawn.
Which of the
following should be performed?
-
?
A.
Open breakers 3305 and 3312.
<
VB.
Start boration using a boric acid storage tank and
'
associated pump.
2-a C.
' Start boration from a reactor coolant bleed tank with a
. concentration greater than reactor coolant.
D.
No action is required; the rods are not withdrawn enough to affect shutdown margin limits.
Reasons:
i
.A.-
This action has already been completed.
C.
This is not an accepted method of emergency boration.
i D.
Emergency boration must be started with these-conditions.
i l
MODIFIED BANK; ROT-5-96 60; ROTS J - Final 96; ROTS K - T2; ROT-5-96 pages 8 and 9
.
NRC977FST Vension: 0 Page: 28
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2 6. rot-5-77 001/A1//3430303062/2.2.6//3.3/44/10CFR L
YourLcrew is asked to develop a procedure change based on a recent engineering evaluation.
Which of the following would j
involve an un-reviewed safety question?
L f.
A.
New cutlass rubber bearings have been installed in the o
circulating water pumps.
They will be operated with
'
dom'estic water and se'lf-cooling as:the back-up rather.
than well water.
i
)
vB.
High pressure injection flow must be throttled to <-450 gpm per pump from 25 feet in the borated water storage tank (BWST) until piggy back is established.
C.
. Nuclear services closed cycle cooling pump, SWP-1C, and Nuclear services raw water pump, RWP-1, can no longer be used in modes 1 through 3.
D.
Due to reduced flow the BWST recirculation pump, SFP-2, cannot be used to recirculate the BWST.
l Reasons:
A.
This is not safety equipment.
The current bearings are cooled by DO and one of the back-up means of cooling the bearings is self cooled from the pump flow.
C.
Neither of these pumps are safety equipment.
D.
SFP-2 is not safety equipment.
There are other ways to recirculate the BWST.
NEW; AI-400C pages 2, 23 & 24; CP-213 page 7; 10CFR50.59 pages 688-689
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NRC97,TST Version: 0 Page: 29
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27. rot 5-102 001/ AS//3440403001/00054AA101//4.4/44/EoP-04
>
Following a loss of main feedwater.these conditions exist:
25 inches -level in. each steam generator (OTSG).
-
'
The emergency feedwater tank is at 35 feet.
-
'
Reactor coolant pressure' is 2200 psig.
-
Reactor coolant temperature is 580*F.
-
~
Which ~is an acceptable l feed rate with both emergency feedwater pumps (EFPs) running?
-
,
i
VA.
760-gpm flow from EFP-1
'
750 gpm flow from EFP-2 B.
.760 gpm flow from EFP-1
'760 gpm flow:from EFP-2 C.
780 gpm flow from EFP-1 750 gpm flow from EFP-2 D.
780 gpm flow from EFP-1 780 gpm flow.from EFP-2 Reasons:
8., C., & D.
When the RCS is subcooled, the OTSGs are not dry,
and the level in EFT-l'is > 18 ft, EFW flow is limited to < 780 gpm for each EFWP and < 1520 gpm-total with 2 EFWPs.
'
.
NEW; E0P-13 page 7; ROT-5-102 page 23 NRC97.TST Version: 0
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28. rot-514 001/B8//NTS/2.4.1//4.3/33/EoP/AP The plant is conducting a shutdown due to a steam generator
,-
i tube leak.
At 50% power-the leak rate increases markedly and i
l the Assistant Nuclear Shift Supervisor (ANSS) directs you to l
trip the reactor.
How would you continue to use plant l
procedures to mitigate this transient?
]
i l
A:
The shutdown would b'e. conducted, using~ AP-510, Rapid
'
Power Reduction.
When the reactor is tripped you would enter E0P-2, Vital System Status Verification, and stay
.
there until directed to go to E0P-6, Steam Generator
!
Tube Rupture.
B.
The shutdown would be conducted using E0P-6.
When the
. reactor is tripped you would enter E0P-2 and stay there
,
until directed to go to E0P-6.
vC.
The shutdown would be conducted using E0P-6.
When the reactor is tripped you would carry out the immediate actions of E0P-2 and then go to E0P-6.
D.
The shutdown would be conducted.using AP-510.
When the reactor is tripped you would carry out the immediate actions of EOP-2 and then go to E0P-6.
Reasons:
B.
E0P-6 would be used for plant shutdown not AP-510.
.C.
After the reactor trip entry into E0P-2 would be only for immediate actions; then return to E0P-6 D.
EOP-6 would be used for plant shutdown not AP-510.
NEW; ROT-5-14 page 25 & 26 l
NRC97.TST Version: 0 Page: 31
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2 9. rot-4 56 002/36//0080401001/008A201//3.6/33/SW Nuclear services closed cycle cooling pump, SWP-1C, tripped on motor overload.
SW pump discharge header pressure has decreased to 108 psig.
What action (s) should be taken?
A.
Investigate the cause of the motor overload.
All other
,
conditions are normal.
B.
Attempt one restart of SWP-1C.
C.
Start SWP-1A.
All other conditions are normal.
l vD.
Start SWP-1B and investigate why neither SWP-1A nor SWP-1B started.
Reasons A.
The motor overload is not important at the moment but low SW pressure is.
B.
An investigation of the motor overload should be performed prior to attempting a restart.
C.
SWP-1A should have automatically started first followed by SWP-1B if pressure does not recover.
NEW; 0080401003; ROT-4-56 page 28; OP-408 page 4 NRC97.TST Version: 0 Page: 32
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30. rot-3-22 002/B2//0000501028/035K501//3.9/33/EXC HEA The following plant' conditions exist:
-
The plant has just entered mode 2.
An atmospheric dump valve fails: open.
)
-
(
If the reactor does not trip and no op'erator action is taken, what will happen to T,y, and nuclear power?
1'
.
'
.
l l
A.
T,y, will increase; final power will be at the point of H
'
adding heat (POAH).
B.
T,y, will increase; final power will exceed the P0AH.
.
~
j C.
T,y, will ' decrease; final power will be at the POAH.
vD.
T,y, will decrease; final power will exceed the P0AH.
Reasons:
A.
The steam leak wil'l cause RCS temperature to decrease; the size of the steam leak should approximate the final power
level.
B.
The steam leak will cause RCS temperature to decrease.
C.
The size of the steam leak should approximate the final power level.
.
BANK; ROT-1-50 73; ROT-1-50 B17; ROT-1-47 B11; ROTS J - T4; ROT-3-22 pages 4-10 L
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NRC97.TST Version: 0 Page: 33
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31.. rot-4-66 001/G6//0390101010/039A302//3.5/33/MSR
-
Thef plant has. tripped from 100%.
E0P-10, Post Trip
{
Stabilization, has the-control board operators depress the i
" reset" pushbutton on the reheat control panel.
What does this detail accomplish?
Depressing " reset":
.
VA.
Automatically' resets the reheater control system to zero control valve position.
j
!
B.
Allows manual reset of the reheater control system to
zero control valve position.
C.
Isolates the high pressure and low pressure bundles.
D.
Ensures that the moisture separator reheater operation can continue in manual.
Reasons:
j B.
Manual reset of the control valves does not require depressing this pushbutton.
C.
Isolating the HP and LP bundles is a manual function.
D.
Manual operation is independent of this pushbutton.
NEW; ROT-4-66 pages 34 & 39; OP-212 pages 4, 39 & 40 NRC97.TST Version: O Page: 34
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i l-l 32. rot-4-81002/G2//0780104002/0780301//3.2/77/IA You have just received an instrument air (IA) low pressure alarm.
IA pressure ~is slowly decreasing.
You send the secondary plant operator to check the status of the instrument
air dryers.
What would you expect his/her report to be?
l
)
l A.
Both' towers isolated, purge valve failed closed
'
B.
Both towers isolated, purge valve open j
i Both towers.in s'rvice, purge valve failed open C.
e vD.
Both towers in service, purge valve closed Reasons:
When air pressure goes below 90 psig IADR-2 de-energizes which puts both towers in service with the purge valves closed.
This makes A.,
B.,
and C. incorrect.
The low pressure alarm setpoint is 85 psig.
)
.
BANK; ROT-4-81 8; ROT-4-81 page 8 i
)
I
)
I NRC97.TST Version: 0 Page: 35
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33. rot-4-28 002/B14//0010101033/014A104//3.8/33/CRD l
Initially the reactor was operating at 60% power with all control rods group 1 - 7, reading 100% on absolute position indication (API).
Five minutes.later core conditions are:-
Reactor power is 55%.
-
A rod near the center of the core has an API of 0%
-
withdrawn with the rod in-limit light on.
-
What is the condition of the power distribution in the core-after. equilibrium is reached?.
.
i A.
Axial power distribution has changed more than radial power distribution.
L
-
l vB.
Radial power distribution has changed more than axial
power distribution.
,
1:.
Radial and axial power distribution do not change.
l D.
The changes in axial and radial power distribution are equal.
,
Reasons:
A.,
C., & D.
When a rod is dropped near the center of the core after equilibrium is reached radial flux distribution changes will be more prono' nced then axial flux distribution.
u l
NEW; ROT-1-48 B3; Rx theory chap. 9, pages 17 & 18 i-e
!
NRC97.TST Version. O Page: 36
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34. rot 4-62 001/ B7// 0360501002/ 026A401//4.3/33/ BS '
Which of the following flow controller settings will allow reactor building (RB) spray flow to'be controlled at 1500 gpm
'
should a. large-. break-loss of coolant accident (LOCA) occur?
!
'
}
A.
Automatic, local, thumb wheel '1200-
'
VB '. '
Automatic, remote, t'humb wheel 1260-
'
-
C.
Manual, local, thumb wheel 1500.
.
D..
Manual, remote, thumb. wheel 1500
,
Reasons:
,
!
_A.
This setting will allow the operator to use the thumb wheel i
and set in any flow.
b-C.
In manual the detent lever may be used to open or close-the
. val ve.
,
- _
D.
In manual the detent lever may be used to open or close the valve.
!
,
MODIFIED BANK; ROT-4-62 1; ROTS J -
T5 & T10B; ROTS K - T2;
'
ROT-4-62 page 4 NRC97.TST Version: 0 Page: 37
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3 5. rot-4-69 001/ B6// 0560401003/ 056A204// 2.8/ 33/ CD The plant is operating at 75% power when the "A" condensate pump trips.
Which of the following should be performed?
.
A.
Monitor deaerator level.
The remaining condensate pump
should be. able to maintain this power level.
l vB ".
If high deaerator level caused the condensate pump trip,
'
manuallyz reduce' demand to zero.
The pumps can then be restarted and condensate flow restored to normal.
C.
Direct the SP0 to place the controller for the pump.that tripped in local manual so that it can be restarted and
,
controlled from the turbine building.
D.
Immediately attempt to restart the condensate pump,
.
deaerator level problems may cause a feedwater booster
'
pump to trip.
'
,
,
Reasons:
A.
One condensate pump is designed to handle 55% power.
C.
This action is not required since the condensate pump can
be restarted as soon as the deaerator level recovers.
D.
If deaerator 'high level trips the condensate pumps; low
,
level trips the feedwater booster pumps.
The CDP may be i
damaged if an attempt is made to start it from the conditions it was at when tripped.
i
BANK; ROT-4-69 29; ROTS J - Final 96; ROT-4-69 pages 12-15 NRC97.TST Version: 0 Page: 38
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'36. Ro7-4-56 001/B5//0080101009/022K101//3.7/33/SW ~
The plant was operating at 100% power when a steam leak on the
"A" steam generator occurred in the reactor building (RB).
The following-conditions exist:
-
Reactor building pressure is 5 psig.
-. Reactor coolant (RCS) temperature (T ) 490* F.
c
.
RCS pressure 1400 psig and increasing.
-
l
-
Pressurizer level is 10 inches.
-
A" steam generator is isolated.
"
"B" steam generator is being fed from emergency feedwater
-
and steamed through.the atmospheric dump valve.
-
SW surge tank level is 9 feet.
The main feedwater pumps have tripped.
-
In this situation the nuclear services closed cycle cooling (SW) system is providing cooling water to:
'
.
VA.
Reactor coolant pumps and reactor building main fan
- {
assemblies.
B.
Reactor coolant pumps and control rod drive mechanisms.
\\
C.
Reactor coolant drain tank and reactor building main fan
'
assemblies.
D.
Reactor coolant drain tank and control rod drive mechanisms.
Reasons:
B.
CRDs have SW isolated on RBIC C.
D.
Both RCDT and CRDs have SW isolated on RBIC.
NEW; ROT-4-56 pages 33 & 34; OP-408 page 4 NRC97.TST Version: 0 Page: 39
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l 37. rot-5107 001/82//NTS/2.4.4//4.3/331EoP/AP
.
l The following conditions exist:
-
Instrument air pressure is 90 psig.
Nuclear services closed cycle cooling water surge tank,
-
SWT-1, has a-level of 7.5 feet.
l Sample point CE-2, condensate pump discharge has a reading
-
l of 11 mho/cm.
l The "A" decay heat closed cycle cooling radiation monitor,
-
. RM-L5 is in highEalarm.
L What action should be taken?
,
.
A.
Enter AP-250, Radiation and Monitor Actuation.
.
B.
Enter, AP-330, Loss of Nuclear Services Cooling.
C.
Enter, AP-470, Loss of Instrument Air.
VD.
Enter, AP-604, Waterbox Tube Failure.
Reasons:
A.
High alarm for RM-L5 is not an entry condition for AP-250.
B.
Low level entry condition for AP-330 is 7 feet.
C.
Low instrument air pressure entry condition for AP-470 is 85 psig.
-NEW; AI-505 page 15; AP-604 page 1; ROT-5-112 page 2; AP-330-page 1; ROT-5-61 page 7; AP-250 page 1; ROT-5-60 page 1; AP-470 page 1; ROT-5-84 page 1
I i
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NRC97.TST Version: 0 Page: 40
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3 8. - rot-5 'o4 001/ A1// 3440403004/ 2.4.40// 4.0/ 44/ EM -~
]
'
The emergency coordinator (EO has declared a general emergency.
The emergency operations facility, EOF, is not staffed.
Besides classification what other duty can not be j:
delegated to.-another emergency team member by the EC?
L
l R
l VA.
Develop protective actionJrecommendations
.
.
B.-
Direct site eva'cuation l
C.
Make notifications to. the state
!
!
l D.
Direct'the shutdown.of the plant
.
Reasons:
1-
,
!
.B.,' C., and D. are things that the EC has the authority to do.
l but can also be delegated..
L l-
BANK; TRE-007 BANK; EM-202 page 6
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h NRC97.TST Version: 0 Page: 41 l
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39. rot 5-98 00' /A2//3440403001/E09EK1.2//4.0/44/EoP-09
A step in E0P-09, Natural Circulation states.
-
.
I
3.1, Ensure OTSGs' levels are at or i
trending towards 60 to 70%.
What.is'the purpose of this step?
l
.
'
,
.
!
A.
To supply water directly onto the OTSG tubes providing core thermal center elevation above the thermal center of the OTSG.
!
L B.
To' ensure that the condensing surface above the RCS water leve n in the OTSC tubes is high enough to provide
.
l
' adequate boiler condenser cooling.
l vC.
To supply water directly onto the OTSG tubes providing-OTSG thermal center elevation above the thermal center
!
of the core.
l D.
To ensure that the column of cooled RCS water in the
!
cold leg is high enough to overcome the RCS back pressure and provide continuous flow through the OTSG.
.
Reasons:
A.
Core thermal center elevation should be below that of the OTSC.
B.
This is the reason for a level in the OTSG of 80 - 90%.
D.
It is not so much the column of water in the cold leg but the column of water in the OTSG that needs to be high enough to ensure continuous flow.
MODIFIED BANK; ROT-5-98 6; ROT-5-98 pages 3 & 4; ROT-5-97 pages 5&6
NRC97.TST Version: 0 Page: 42
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40. rot-4-28 004/ AS//3410103037/00005AK302//4.2/55/CRD L
The following conditions exist:
1-
-, The reactor is at 60% full power at 100 EFPD.
-
All reactor coolant pumps are running.
.
~
-
Control rod'7-5 has become stuck at 24% withdrawn.
i
-
Group 7 has been positioned at 24%.
l Which of the following is correct?_
j
,
i l
-
A.
If group -7 is.left at 24% ' withdrawn, shutdown margin may be inadequate' but no assymetric rod condition will exist.
vB..
. If group 7 is left at 24% withdrawn, power peaking factors may exceed their limit but no assymetric rod l
condition will exist.
!
C.
If group 7. is moved to. 26% withdrawn, an assymetric rod condition will exist but shutdown margin will be adequate.
,
I
,
D.
If group 7 is moved to 26% withdrawn, an assymetric rod l
condition will exist but power peaking factors will L
exceed their limits.
l l
l l
Reasons:
l A.
' Group 7 will be in the restricted region; shutdown margin is adequate.
C. & D.
A 2% difference between the stuck rod and the rest of l
group 7 will not cause an assymetric rod condition, i
i
!
I NEW; Technical Specifications pages 3.2-1, B3.1-21, B3.2-5 &
_
B3.2-7; COLR page 7
C-NRC97.TST Version: 0 Page: 43
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l 41. rot.5-81001/B1//0160401002/A02AA2.1//4.0/33/AP-581 A small reactor coolant (RCS) pressure excursion reached a
-
point where the pressurizer spray valve should have opened
'
automatically.
When the spray valve did not open the operator
,was able to open it manually and reduce RCS pressure.
A number i
of SASS modules have swapped to their alternate source.
The
>
white indicating light for NNI-X is extinguished on the
[
redundant instrument panel?
Why did the spray valve only work in manual?
!
>
.
,
A.
NNI-X AC power was lost, automatic control power to the i
i spray valve is not available.
NNI-Y power is supplying i
manual control power.
l B.
NNI-X DC power was lost, automatic control power to the spray valve is not available.
NNI-Y power is supplying manual control power.
C.
NNI-X.AC power was lost, automatic control power to the y
spray valve is not available.
A backup power source is L
supplying manual control power.
vD.
NNI-X DC power was lost, automatic control power to the
'
spray valve is not available.
A backup power source is l
l supplying manual control power.
!
Reasons:
A.
NNI-X DC supplies the power for automatic control of the
l spray valve.
NNI-Y does not supply backup power for j
automatic pressurizer spray valve control.
l B.
NNI-Y does not supply backup power for automatic L
pressurizer spray valve control.
l
!
C.
NNI-X DC supplies the power for automatic control of the l
spray valve.
l l
i
,
s MODIFIED BANK; ROT-5-81 2; ROTS J - T10A & T10B; ROT-5-81 pages
,
1-4; ROT-4-09 pages 31 & 32 NRC97.TST Version: 0 Page: 44
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42. ROT-4-64 001/B2//NTS/00058AK101//3,1/33/DC l
Annunciator P-7-8, " BATTERY CHARGER TROUBLE",-is actuated.
The L
auxiliary building operator' (ANO) reports that the "A" battery charger's :(DPBC-1A) output voltage is 52 volts.
What further
. direction. should be given?
l l.
-
l l
A.
Have the ANO place D.PBC-1B in service for DPBC-1A~.
-
-
.
.
B.
Have the ANO place DPBC-1C in service for DFBC-1A.
l, C.
HavetheANOplaceDPBC-1DinsbrviceforDPBC-1A.
I vD.
Have the ANO place DPBC-1E in service for DPBC-1A.
Reasons:-
l:
l A.
DPBC-1B cannot be placed in service for DPBC-1A it is a
"B" train battery charger.
B.
DPBC-1C cannot be placed in service' for DPBC-1A is not the swing charger on the "A" bus.
C.
DPBC-1D cannot be placed in service for DPBC-1A it is. a
"B" train battery charger.
..g
.
i j:
NEW; AR-701 page 95; ROT-4-64 pages 4 & 5 l'
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NRC97.TST Version: 0 Page: 45
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43. rot-4-15 001/B7//0190101001/061A102//3.6/33/EFIC.
Following a main steam line rupture, OTSG "A" pressure r
initially decreases to 550 psig and OTSG "B" pressure initially decreases to 675 psig.
Both OTSGs subsequently increase to 900
l
~ psig.
OTSG level is maintained within normal limits.
Which of
!
-the following is correct concerning the operation of emergency
,
l
.
.
l
.
.
EFIC will isolate EFW'to O'SG "A",'then the EFW control
'
VA.
T valve and its block valve will open to control SG level.
B.
EFIC will isolate EFW to both OTSGs and an. operator must
,
manually reset the " feed only good generator" (FOGG)
logic to restore EFW to both OTSGs.
'
'
C.
An operator must take manual control and restore EFW flow to OTSG "A".
D.
An operator must take manual control and restore EFW flow to both OTSGs.
,
il l
Reasons:
B.
The operators do not manually reset the FOGG logic.
There was no isolation of EFW to the "B" OTSG.
C.
The operator does not need to take manual control to restore EFW.
D.
There was no isolation of EFW to the "B" OTSG.
The
'
operator does not need to take manual control to restore EFW.
BANK; ROT-4-15 17; ROTS J - T9; NRC 5-93; ROTS K - T2; ROT-4-15 pages 23 & 25
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44. ROT-4-10 002/B3//0150101005/00032AK201//3.1/33/NI The following plant conditions exist:
'
i j
-
ES MCC 3B1 is de-energized for maintenance.
l.
-
An electrical fault causes a failure of the "B" inverter.
!
What is' the status of the nuclear instrumentation?
l:
.
-
.
.
A.
NI-1 and NI-5 have no power.
'
vB.
NI-2 and NI-6 have no power.
l C.
NI-3 and NI-7 have no power.
D.
.NI-4 and NI-8 have no power.
j Reasons:
l
_
l The combination of the "B" inverter and ES MCC 3B1 would l
de-energize VBDP-4.
l A.
NI-1 and NI-5 are powered from the
"D" RPS cabinet which is supplied from VBDP-3
,
,
l C.
NI-3 and NI-7 are powered from the
"D" RPS cabinet which l
is supplied from VBDP-5 D.
NI-4 and NI-8 are powered from the "D" RPS cabi,1et which is supplied from VBDP-6 NEW; ROT-4-10 page 19; ROT-4-12 page 14; ROT-4-91 page 13 i
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i 45. rot-4-60 007/89//0020101018/00022AA109//3.3/33/RC The plant is operating at 100% power when the following j
t indications for RCP-18 are-noticed:
l
-
Nuclear services closed cycle cooling (SW) temperature
!
leaving the pump is 176*F.
!
--
Seal injection flow has been lost.
Pump vibrations are in the " action" range on one monitor
-
i and in " alert" range on another monitor
,
Thrust bearing temperatures are 180'F.
- Which of the following is required?
'
l-
!
l A.
Restore seal injection flow.
-
B.
Notify engineering and continue to monitor the pump.
l l
C.
Start the AC lift oil pump.
vD.
Secure the reactor coolant pump after reducing power to 75%.
'
l Reasons:
A.
This is a nice thing to do but not the correct action in i
'
regards to the vibration problem.
i L
B.
This action was performed when the pump first came into al a rm.
l l
C.
This is an action performed for high thrust bearing
.
temperatures which is not applicable.
!.
BANK; ROT-4-60 8; ROTS J - T5; OP-302 pages 5-7; ROT-4-60 pages 14-16
NRC97.TST Version: 0 Page: 48
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46. rot-G-50 001/Fil/0340101003/00033AK103//4.3/88/FP i
L While placing a fuel-assembly into the core, the count rate l
changes from the 25 cpm to 39 cpm on both of the source range j
nuclear instruments.
What_should be done?
?
A.
This count rate increase is slightly elevated, contact the-reactor engineer.
-
l B.
-This count rate increase is slightly elevated, wait for L
the count rate to decrease and then continue refueling.
I
!
C.
This is an unacceptable count rate increase, stop the placement of the fuel assembly until the problem is resolved.
j
,
i
.
sD.
This is an unacceptable count rate increase, remove the J
fuel assembly being loaded into the core and stop l
refueling until the problem is resolved.
Reasons:
A.
The count rate response is too high - the assembly must be removed and all positive reactivity additions stopped until further investigation is completed.
i B.
The count rate response is too high; waiting will 'not make a difference to the response - the assembly must be removed and all positive reactivity additions stopped until further investigation is completed.
C.
The fuel assembly has to be removed.
i NEW; FP-203 page 2
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NRC97.TST Version: 0 Page: 49
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47. rot-4-60 005/B6//0o205o1007/C0008AA101//4.0/33/RCS
'
-The plant tripped during a main feedwater transient.
The following plant conditions. exist 5 minutes after the trip:
RCS pressure is 2185 psig
-
o
-
RCS average temperature is 555'F.
J Which of the following sets of indications at 10 minutes after the trip would lead you to bel.ieve that the pressurizer spray i
]
va,1ve is still open and.not the PORV?
,
+
VA.
RCS pressure is 2085 psig.
PORV tail pipe' temperature is 185'F.
,
Surg.e line temperature is 643 * F.
!.
B.
RCS pressure is 2085 psig.
PORV tail pipe temperature is 185'F.
Surge.line temperature is 556* F.
C.
RCS pressure is 1985 psig.
PORV tail pipe temperature is 125"F.
'
Surge line temperature is 614*F.
>
e
D.
RCS pressure is 1985 psig.
j PORV tail pipe temperature is 125'F.
Surge line temperature is 572*F.
Reasons:
}
B.
If the pressurizer spray valve was open an outsurge would occur causing surge line temperature to be at
saturation.
.
C.
After a 10 minute period the PORV being open would lower the pressure faster than if the spray valve was open.
If the pressurizer spray valve was open an outsurge would occur causing surge line temperature to be at saturation.
D.
If the pressurizer spray valve was open an outsurge would occur causing surge line temperature to be at saturati on.
After a 10 minute period the PORV being open would lower the pressure faster than if the spray valve was onen.
NRC97.TST Version: 0 Page: 50
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l 47. ROT-4-60 005/B6//0020501007/OC008AA101//4.0/33/RCS l
NEW; OP-204 page 3; ROT-4.-60 pages 8 & 10
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NRC97.TST Version: 0 Page: 51
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48.- rot-5-30 001/ A2/ rot-5-14/ 3440403001/ A05AK3.4// 3.6/ 55/ AP-770 I
- During a LOCA with lo.ss of subcooling margin, power is lo.st to l
the."A" 4160V ES bus.
The' emergency diesel generator (EDG)
.
L-output breaker is closed.
The two control board operators are
'
i busy completing actions. required by E0P-03, Loss of_ Subcooling -
f-Margin.
The-s'enior chief'and a trainee are working on AP-770,
.
Diesel Actuation.
They reach step'3.12 which states:
j
"3.12 IE ES 430V UV lockouts have actuated,
- .
THEN reset
-
.
.
i An emergency:in the turbine. building requires the senior.
'
' chief's attention. - As procedure director, what should you do?
<
l c
!.
l
-A.
' This step is required in order to start needed -480V l
equipment; follow the trainee and supervise the
performance.of this step, b
!
B.
Wait for the: senior chief to return to the control room
'
'
to complete AP-770; the trainee cannot.make unsupervised i
control board manipulations.-
'
!
C.
Continue in E0P-3 which contains adequate guidance for
'
this situation; AP-770 is not required to be completed, j
vD.
This step is required in order.to start needed 480V equipment; send a control board operator to perform this
-
{.
step - the trainee may watch or complete the step under
{
instruction.
- -
i
Reasons:
l'
'
A.
'It is inappropriate for the procedure director to leave his l
position of control and command.
p
B; This step would be required since a 480V under ' voltage l-lockout would have actuated.
,
,
C.
This step would be required since a 480V under voltage lockout would have actuated.
NRC97.TST Version: 0 Page: 52
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48. ROT-5-30 001/ A2/ ROT-5-14/ 3440403001/ A05AK3.4// 3.6/ 55/ AP-770 NEW; ROT-5-14 B6; ROT-5-30 pages 20 & 21; AI-505 page 3-
,
.
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.l NRC97.TST Version: 0 Page: 53
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l 49. rot-4-11001/B9//0000501021/017A30'il/3.8/33/IN Given the following conditions:
!
A LOCA has occurred.
-
'
l
-
RCS pressure is 585 psig and increasing.
T is 486*F and stable.
-
hot
-
T is 485'F and decreasing.
cold l
.
Average incore temperature is 486*F'and increasing.
-
!
l What is the status of core cooling?
-
l l
A.
' The -reactor is being cooled by saturated natural circulation.
B.
' The reactor is being cooled by 'subcooled natural
,
L ci rculation.
vC. -
The reactor is not being cooled; the reactor coolant.is saturated.
D.
The reactor is not being cooled; the reactor coolant is superheated.
Reasons:
A.
The reactor is not being cooled; no natural circulation is occurring.
L B.
The reactor is not being cooled; no natural circulation is occurring.
D.
The reactor coolant is in a saturated condition.
i l
NEW; 0000501006; 0170101007; ROT-4-11 page 8 i
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NRC97.TST Version: 0 Page: 54
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50. rot-5-109 001/ A8// NTS/2.4.21// 4.3/44/ CP-150 4 -
The plant is at 80% full power.
Nuclear services closed cycle cooling pump, SWP-1A,-has been removed from service for
!
maintenance.
SWV-579, 507, 508, 509, 510 and 607 (SW to the
.
motor driven emergency.feedwater pump, EFP-1) are tagged closed
'
to replace a leaking gasket.
- Which of the following conditions would lead.to a loss ~of
safety function (LOSF)?
-
.
l 4-A.
Decay heat closed cycle cooling surge tank, DCT-1A, is i
isolated to correct chemistry problems.
i vB.
Steam driven emergency feedwater pump steam trip valve, l
. ASV-50, fails closed, j
C.
Raw water pumps, RWP-3A and 2A, are removed from service
for oil changes.
D.
Emergency diesel generator, EDG-1A, fails its monthly
,
surveillance.
Reasons:
A.
There is sti.ll one good train of DC and SW on the "B" side.
C.
These two pump are on the "A" train.
D.
This is an "A" train component.
BANK; ROT-5-109 6; ROT-4-56 page 5; ROT-5-109 page 4
NRC97.TST Version: 0 Page: 55
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51. rot-4-12 001/ B3//0120101005/012K201//3.7/ 33/ RPS '
Plant conditions are as follows:
!
-
The plant is at 100% full power.
l
.The: reactor protection system (RPS) cabinet "A" test trip-indicating -lamp is br'ightly lit.
-
Reactor coolant (RCS) pressure is 2310 psig.
-
RCS;T is 615."F.
hat
-
Reactor building pressure is 2.0 psig.
. Imbalance is within its operational limit.
What has de-energized the test / interlock trip relay, K2.27 A.
RCS. pressure is high.
B.
RCS temperature is high.
C.
The intermediate range power supply module was removed from its cubicle, vD.
The RCP contact monitor module was removed from its cubicle.
Reasons:
A.
High RCS pressure does not input to the K2.2 module.
B.
High T does not input to the K2.2 module.
hot C.
There is no test switch on the function generator module,
,
.
NEW; ROT-4-12 pages 9, 15, 17, 26, 36 & 37 L
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NRC97.TST Version: 0 Page: 56
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= 52'. rot-4-60 001/B18//0020101018/Or 3A202//3.9/33/RC The.following inforr,ation applies ~ to reactor coolant pump "C" (RCP-1C), which is running:
.The motor guide' bearing temperature is 124*F.
-
' Cooling water outlet temperature is 186*F.
--
,
The thrust bearing temperature is 184'F.
-
l Stator oil temperature is 123'F.
-
l What actions,. if any, should be performed?
.
L l
L A.
The pump should be left running; all paranieters are
.
within~ limits.
L v8.
. Cooling water outlet temperature.is too high; start the l-lift oil pump prior to tripping.
~
C.
Thrust bearing temperature is too high; the pump should L
be tripped.
D.
Motor bearing temperature is too high; reduce power if
(
needed prior to shut down of the pump.
Reasons:
l A.
The cooling water outlet temperature is too high - the pump should be tripped.
C.
All bearing temperatures are within their limits.
L p
D.
- All bearing temperatures are within their limits.
i
MODIFIED BANK; ROT-4-60 41; OP-302 pages 4 & 6; ROT-4-60 page
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NRC97.TST Version: 0 Page: 57 (
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l 5 3. ! ROT-4-64 002/ F2// 0630406003/ 063K401// 3.0/ 88/ DC
{
The-plant is operating in mode 5 and the electricians are u
,
conducting an equalizing battery charge on the "A" 1E battery, j
The following alarms occur:
L
-
Battery ' charger "A" trouble stays actuated.
!
-
Inverter "A" trouble actuates and then clears.
u l
-
Inverter "C" trouble actuates and then clears.
l
-
'
The auxiliary building operator-reports that the "No Charge" and "High Voltage" lights are. illuminated on the
"A" battery l
charger. What could cause this~ situation?
"
A.
The breaker between the " battery-half" that the "A"
. charger supplies and.the
"A" charger has opened.
-
vB.
The "A" battery charger has shutdown.
j C.
The battery charge has been completed.
D.
Both inverters switched to their AC power source.
Reasons:
A.
There is no breaker in this location.
C.
The battery charge does not shutdown automatically.
D.
Tiie alarm indication would not support this.
NEW; ROT-4-64 pages 2, 4 & 5; OP-705 page 3 & 4 l
.
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I NRC97.TST Version: 0 Page: 58
_
54. rot-4-54 001/B7//0050501001/005K411//3.9/33/DH Which of the following sets of conditions indicates the establishment of long term. post-accident cooling?
A.
Reactor coolant pressure is 550 psig Incore temperatures are 280*F Reactor.buildin'g-(RB) pressure is 2 psig
.
'
An HPI train is takin.g suction from the BWST.
B.
Reactor coolant pressure is 20 psig Incore temperatures are 260*F.
RB pressure is 2 psig An LPI train is taking suction.from the BWST.
C.
. Reactor coolant pressure is 550 psig Incore temperatures-are 250'F RB_ pressure is 18 psig An HPI train is taking suction from LPI.
VD.
Reactor coolant pressure is 20 psig
Incore temperatures are 240'F RB pressure is 18 psig An LPI train is taking suction from the RB sump.
Reasons:
A.,
B., & C.
Requirement for.long term post accident cooling include:
LPI sump suction; inadequate subcooling margin; incores less than 280'F; RB and RCS pressure approximately the same.
NEW; ROT-4-54 page 21; OP-404 pages 77-80; E0P-08 page 59; 0050501003 NRC97.TST Version: 0 Page: 59
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5 5. ' rot-5-67 001/ A1// 3440403001/ 00009AK103//4.0/ 33/ AP-525.
{
A ~ step'in - AP-525, Continuous Control-Rod Motion, states:
3.6 IF Rx power and'Tavg are NOT changing, THEN determine if RCS boration or dilution has occurred.
.
-
.
What assumption does this s.tep make abouti rod motion?'
'
'
A.
A dilution might be in progress; rods are moving out to maintain reactor power.
B.
A dilution might' be in progress; rods are moving 'in to maintain main steam header. pressure, vC.
A boration'might be in progress; rods are moving out to maintain reactor power.
D.
A boration might be in progress; rods are moving in to
'
maintain main' steam header pressure.
Reasons:
,
,
A.
Rod motion inward maintains reactor power if a dilution is-in progress.
B.
Rod motion inward is maintaining reactor power not main steam header pressure.
'C.
Rod motion outward is maintaining reactor power not main steam header pressure.
NEW; ROT-5-67 pages 5 & 6 l
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NRC97.TST Version: 0 Page: 60
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56. ROT-4-28 003/B14///00003AA201//3.9/11/CRD
..
At ' full. power regulating rod group 7 has the following RPI and API readings:
-
7-1 80% withdrawn 78% withdrawn 7-2'
78% withdrawn 78% withdrawn 7-3 62% withdrawn 78% withdrawn l
7-4 79% withdrawn 78% withdrawn
'
,
7-5 43% withdrawn 79% withdrawn
~7-6 81% withdrawn 79% withdrawn 7-7 78% withdrawn 80% withdrawn
!
7-8 80% withdrawn 78% withdrawn
.
.The -the 7" fault. lights for rods 7-1, 7-2, 7-3, 7-4, 7-5, 7-6, 7-7 and 7-8 are lit.
The 9" asymmetric fault light is lit.
L What action should be taken?
i
,
A.
API is inoperable for 2 rods; determine position of rods
';
with' inoperable APIS with' zone reference indicators and maintain in the zone reference within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, B.
RPI is inoperable for 2 rods; determine that API position indicator. channels are operable for the rods with inoperable RPIs within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
C.
Both API and RPI are inoperable for 2 rods; declare the
'
rods inoperable immediately and be in mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
vD.
API and RPI indication are working properly, two rods are asymmetric; manually trip the reactor.
Reasons:
A.,
B., & C.
Both API and RPI for rods 7-3 and 7-5 are working properly.
When two or more asymmetric control rods exist l
then a manual trip of the reactor is required.
!
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NRC97.TST Version: 0 Page: 61
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ROT-4-28 003/B14///00003AA201//3.9/11/CRD l
NEW; ROT-4-28 B15 & B16; ROT-4-28 pages 11-16; AI-505 page 15 &
16; Technical specifications pages 3.1-14 - 3.1-16
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NRC97 TST Version: O Page: 62
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.57. rot-5-96 003/B6//NTS/E13EA1.3//3.G/33/EoP-13
'
Given the following conditions:
A small break loss of coolant accident on the pressurizer
-
caused a temporary loss of subcooling margin.
.
-
E0P-02, Vital System Status Verification, and E0P-03, inadequate Subcooling Margin, immediate actions have been
compl eted.
-
-
Subcooling margin.(SCM) is 50'F and increasing.
,LRCS pressure is 1400 psig and increasing.
-
Full high pressure injection -(HPI) is in progress and ES has been bypassed.
-
Pressurizer (PZR) level is off scale high.
--T
= 530*F.
c The plant has cooled down 25' in the last 30 minutes.
j What action, if any, should be taken in this situation and why is it being done?
I vA.
HPI must be throttled to minimize SCM because pressurized thermal shock (PTS) guidelines are appli cabl e.
B.
HPI must be throttled to restore PZR level since adequate subcooling margin exists.
C.
HPI must not be throttled until T reaches 390*F to c
ensure subceoled margin will be maintained while throttling.
D.
HPI must be throttled because cooldown limits have been exceeded.
Reasons:
B.
There is no must requirement to tiirottle HPI for PZR level.
C. & D.
Neither of these requirements exist.
NRC97.TST Vension: 0 Page: 63
i 5 7. ROT-5-96 003/ B6// NTS/ E13EA7.3// 3.8/ 33/ EOP-13 BANK; ROT-5-96 76; E0P-13 page 5 l
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5 8. rot-4-63 002/ B7//3410103036/103K102//4.1/ 55/TS Given the following conditions:
-
The plant is at 100% power.
-
SP-181, Containment Air Lock Test, is in progress on the personnel hatch.
The ISI test engineer informs you the personnel hatch has
-
failed its leak rate test.
-
There are indications of leakage around the shaft of the
-
, handwheel for operating the outer hatch.
.
-
The inner hatch's seal is: degraded to the point that the
]
test engineer can feel air passing by.
'
What are the required actions and the condition of containment integrity?
,
.
A.
Verify both doors are closed and initiate repairs.
Containment integrity may not exist.
.
B.
Verify either door is closed and initiate repairs.
,
Containment integrity does not exist.
vC.
Evaluate overall containment leakage rate and verify one door is closed.
Containment integrity may not exist.
'
D.
Be in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and Mode 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
- '
Containment integrity does not exist.
.
Reasons:
A.
Both hatch door are inoperable; condition A of TS 3.6.2 does not apply.
B.
Both hatch door are inoperable; condition A of TS 3.6.2 does not apply.
D.
An evaluation of the total leakage needs to be done and then if repairs cannot be performed in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> condition D of TS 3.6.2 should be entered.
NRC97.TST Version: O Page: 65
l 58. ROT-4-63 002/B7//3410103036/103K102//4.1/55/TS MODIFIED BANK; ROT-5-01 107; OP-417 page 61; Technical Specifications pages 3.6-3 - 3.6-6.
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- 5 9. rot-4-03 001/ 31// 0640101004/ 064A406// 3.9/ 33/ EDG A high pressure injection (HPI)- actuation has started the
"A"
!
diesel. (EDG-1A).
Following an undervoltage on the "A" ES bus,
!
what would prevent EDG-1A automatic breaker closure?
t l
,
A.
EDG-1A is operating at 61 Hz and 4200 volts.
B EDG-1B is running with its output breaker closed.
vC.
The under-voltage relays are tripped and the normal feeder breaker is still closed.
D.
The synchronizing check relays sense an out of phase condition.
.
Reasons:
A.
These. conditions would not prevent breaker closur.
B.
There is no cross-tie blocking situation.
D.
The sync. check relays are not in the automatic breaker closure circuit.
.
MODIFIED BANK; ROT-4-06 65; ROT-4-06 pages 10 & 19 NRC97.TST Version: 0 Page: 67
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60. rot-4-60 008/B7//0020101013/071A303//3.8/77/RC A plant shutdown and.cooldown is in progress for a refueling outage.
Waste gas compressor 1A is out of service to repair a
,
i cooler leak when waste gas compressor 1B trips due to an
,
l
. electrical fault in the motor.
What effect, if any, will this
Thave on the cooldown?
i
'
A '.
This has no impact on the cooldown. -The reactor c'oolant
.
(RCS) can be degassed to the makeup tank.
.
B.
- This has no impact on the cooldown.
The RCS can be degassed to the reactor coolant drain tank.
d C.
'This could delay the cooldown.
It would be impossible
. to collapse the steam bubble in the pressurizer.
vD.
This could delay the cooldown.
A degas car: not be
-
completed.
i
'
Reasons:
~
r
!
A.
The MUT will have to be vented several time 'during the degas process.- This is not possible without the waste gas compressors.
l
,
B.
Without a waste gas compressor degassing to the RCDT would
- -
cause the an over pressurization in-several possible
.
locations.
,
.
C.
Collapsing the steam bubble has nothing to do with the waste gas compressors.
,
.
NEW; ROT-4-60 B14; 0020101009; ROT-4-60 pages 8 & 46; OP-305 page 4 NRC97.TST Version: 0 Page: 68
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61. rot-5-101001/B1//0000501022/00037AA106//3.9/33/EOP-06 The following conditions exist-Reactor power is 100%.
-
RM-G26 reads 2.0 gpd.
-
RM-G27' reads 96.3 gpd.
-
What action, if any, should be performed?
,
1
"
,
i A.
Both steam generators have tube leaks; enter E0P-6,
- 1 Steam Generator Tube Rupture.
)
B.
One steam generator has a tube leak; enter EOP-6, Steam Generator Tube Rupture.
]
.
VC.
One steam generator has a tube leak; consult CP-152, i
Primary to Secondary Leakage Operating Guidelines.
.;
D.
Neither steam generator has a significant tube leak; no i
-
further action is required.
')
a
'
Reasons:
.
A.,
B., & D.
At full power a normal reading on RM-G26 or RM-G27 is 1-3 gpd.
Entry into E0P-6 is > 1 gpm.
The indications show one generator with a tube leak not sufficient to enter E0P-6.
CP-152 gives guidance for small tube leaks.
NEW; ROT-4-25; pages 5-9 & 24; ROT-5-101 pages 9 & 10; OP-301 pages 104 & 106
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62. rot-5-94 001/A2//3440403001/E05EK2.2//4.4/44/EoP-05 l
While performing the follow-up action of E0P-02, Vital System l
4 Status _ Verification,. after a reactor trip from both main l
feedwater pumps tripping, you observe the following:
,
"A" OTSG level is 73 inches and increasing.
-
"B" OTSG level is stable on low level limits.
--
-
Pressurizer ' level is 52 inches and decreasing.
-
T,y, is 548'F.
What action should be taken for these indications?
A.
Complete the actions of E0P-2 and transition to E0P-5, Excessive Heat Transfer, when directed.
B.
Balance steam demands from both OTSGs while performing E0P-02.
!
vC.
Complete the immediate actions of E0P-02 then transition
~
to E0P-05, Excessive Heat Transfer.
D.
Stop feeding the
"A" steam generator, finish the followup actions in E0P-02, then transition to E0P-05.
Reasons:
A.,
B.,
&
D.
Overcooling is a higher symptom than E0P-02.
E0P-02 should be exited as soon as the immediate actions are completed and E0P-05 entered.
i NEW; ROT-5-94 pages 1 & 2; AI-505 page 8
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NRC97.TST version: 0 Page 70
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l 63. rot-5-85 001/B3//0000501026/E03EA1.1//3.8/33/EoP-03 l
Following a reactor trip the SPDS screens turns red and the
.
number 42 appears.
. hat does this display mean and what action W
should follow?
!
l A.
Reactor coolant pressure is greater than 1500 psig; E0P-3,. Inadequate Subcooling Margin should'be entered.
-
,
l
.
B.
Reactor coolant pressure is greater than 1500 psig; E0P-2, Vital System Status Verification, should be
completed.
vC.
Reactor coolant pressure is between 1500 and 250 psig; E0P-3, Inadequate Subcooling Margin should be entered.
D.
Reactor coolant pressure is between 1500 and 250 psig; E0P-2, Vital System Status Verification, should be i
completed.
'
i Reasons:
A.
SPDS would not have this. indication if RCS pressure was greater than 1500 psig.
I B.
SPDS would not have this indication if RCS pressure was
. greater than 1500 psig.
Once immediate actions are completed in E0P-02, E0P-02 should be exited for E0P-03.
D.
Once immediate actions are completed in E0P-02, E0P-02 should be exited for E0P-03.
NEW; ROT-5-85 page 1 & 2; ROT-4-21 page 38
NRC97.TST Version: 0 Page: 71
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64. rot-5-48 003/A3/ rot-4-25/3410103029/00060AK202//3.1/44/ WASTE l
During venting of the makeup tank (MUT) an improper valve
l alignment causes the waste gas header to be vented to the j
auxiliary building and RM-A3 actuates.
How does this affect I
auxiliary building ventilation?
I A.
The supply air is isolated to the waste gas surge and
'
decay tank area.
Exhaust air from the waste gas compressor area is isolated.
B.
The supply air is isolated to the general floor area of i
the 95' elevation of the auxiliary building.
Exhaust air from the waste gas surge and decay tank area is isolated.
vC.
The supply air is isolated to the waste gas surge and i
decay tank area.
Exhaust air from the waste gas surge and decay tank area is isolated.
D.
The supply air is isolated to the general floor area of the 95' elevation of the auxaliary building.
Exhaust air from the waste gas compressor area is isolated.
i Reasons:
,
A.
D-36 isolates the exhaust air from the waste gas surge tank area.
B.
D-29 isolates the supply air. to the waste gas surge and decay tank area.
D.
D-29 isolates the supply air to the waste gas surge and decay tank area.
D-36 isolates the exhaust air from the waste gas surge tank area.
i NEW; ROT-5-48 page 10; ROT-4-25 page 27; ROT-4-25 81; ROT-5-60 A2; ROT-5-60 pages 4 & 5; ROT-4-86 pages 16 & 17 l
NRC97.TST Version: 0 Page: 72
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65. rot-5-61001/B3//0000401001/00062AA201//3.5/33/SW AP-330, Loss of Nuclear Services, has been entered.
The following plant conditions exist:
-
The nuclear services surge tank (SWT-1) continues to decrease in level.
-
The reactor building and auxiliary building sump levels are not increasing.
l The turbine building sump level is increasing.
--
, The reactor coolant drain tank is not increasing ~.
'
j
-
The miscellaneous waste storage tank is not increasing.
j
-
All nuclear services heat exchangers (SWHEs) have been
rotated into operation with no change in conditions.
-
MUV-31 make-up valve demand is steady.
-
There are no reactor building system leak annunciators in alarm.
-
Letdown temperatures are increasing.
Where is the location of the SW leak?
A.
The reactor coolant drain tank.
B.
The pressurizer sample cooler.
)
i C.
The make-up pump cubicles.
vD.
The industrial cooling system water heaters.
I Reasons:
,
n.
Tanks cooled by SW are not increasing in level.
B.
The SW would not leak out of this cooler; RCS will leak into the SW system.
C.
The auxiliary building sump level is not increasing.
NEW; ROT-4-56 pages 3-6; ROT-5-61 pages 17 and 18 NRC97.TST Version: O Page: 73
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66. rot-4 29 001/B4//0330901011/033K401//3.2/33/SF Current plant conditions:
Decay heat pump, DHP-1B is in operation.
-
-
The borated water storage tank, BWST, is being recirculated with spent fuel pump,.SFP-2.
-
A reactor building (RB) purge is started.
-
The-spent fuer supply fan (AHF-10) is in operation.
.Defueling operations are in progress.
~
'
Annunciator G-8-1, " SPENT FUEL POOL LEVEL HIGH/ LOW", comes'into alarm. No other alarms are-indicated.
What could be causing the low level?
A..
. Incorrect valve alignment for the BWST recirculation.
B.
The transfer canal deep end drain valves are leaking.
vC..
The purge has lowered pressure in the RB.
D.
AHF-10 has raised pressure on the spent fuel floor.
Reasons:
A.
With none of the large spent fuel pumps operating if a valve alignment was. incorrect the spent fuel pool-level would be high not low.
B.
If the drains were leaking enough to.cause a low spent fuel pool level.the transfer canal would also be low.
D.
AB. ventilation in normal alignment would not raise pressure on the spent fuel floor if AHF-10 is operating.
NEW; ROT-4-29 B5; ROT-4-29 page 3;.AR-402 page 77; OP-406 page
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NRC97.TST Version: 0 Page: 74
l 67. rot-5-78 001/B12//0020101009/00076AK201//3.0/33/oP-301 l
Chemistry has informed the control room that reactor coolant
system radiochemistry indicates fuel failure has occurred.
What indications / alarms in the control room would corroborate
their analysis?
I A.
NI-5, NI-6, NI-7 and NI-8 show a 30% power excursi.on.
'
,
.
B.
High differential pressure alarm across pre-or post-makeup filters.
vC.
Letdown radiation monitor, RM-L1 in alarm.
D.
Elevated reactor building radiation monitor readings.
Reasons:
A.
A 30% power excursion would indicate an iodine spike has
occurred.
B.
High filter dP would indicate a crud burst.
'
D.
Elevated reactor building radiation monitors would not
'
narrow the choices to fuel failure.
NEW; ROT-5-78 page 38; OP-301 pages 117 & 118
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l NRC97..TST Version: O Page: 75 l
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68. rot-5-34 002/A4/TRE-007/3440403006/00059AK302//4.5/44/E-PLAN Given the following sequence of events:
-
An inadvertent radioactive liquid release is r.ade during accident conditions.
-
The Emergency Coordinator (EC) declares an " alert" at 1200.
When must the NRC Resident and State Warning Point Tallahassee (SWPT) be notified ?
A.
As soon as possible but not to exceed 1215 notify both the NRC Resident and SWPT.
B.
. As soon as possible but not to exceed 1300 notify both the NRC Resident and SWPT.
VC.
By 1215 notify SWPT; as soon as possible notify the NRC Resident.
D.
By 1215 notify the NRC Resident; as soon as possible notify SWPT.
'
Reasons:
A.
SWPT must be notified within 15 minutes; the NRC resident as soon as possible.
B.
The NRC resident is notified as soon as possible.
'
D.
SWPT must be notified within 15 minutes; the NRC resident as soon as possible.
NEW; 3340403009; EM-202 pages 9 & 13
1 I
NRC97.TST Version: 0 Page: 76
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69. rot-4 59 001/F4//0680106005/068K504//3.5/88/WD
.The plant has experienced a small fuel failure.
Reactor
'
,
l coolant activity is low enough for the plant to continue operation.at full power.
What should be.done with liquid waste generated during continued operation to minimize auxiliary bui'iding area radiation levels?
VA'. -
Store in the miscellaneous waste storage tank (MWS'T).
.
B.
Store in the reactor coolant bleed tank (RCBT).
l C.
Store in the evaporator condensate storage tank (ECST).
t
.
D.
Store in the concentrated waste storage tank (CWST).
Reasons:
- -
'
B.,.C., & D.
The tank that affords the highest amount of shielding and restriction to entry is the MWST.
NEW; ROT-4-59 pages 7-9, 16 & 17
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l NRC97.TST Version: 0 Page: 77 l
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l 70. rot-5-43 001/ A4// 3410103031/2.3.1// 3.0/ 55/ RAD While changing a pump out in the reactor building a mechanic has received the following doses to his hands and feet:
-
47 rem - left foot
-
45 i m - right foot
-
60 rem - left hand
'
-
53 rem - right hand What radiation limits, if any, have been exceeded; what notification, if any, is required?
A.
The administrative and NRC extremity doses have not been exceeded; no NRC notification is required.
B.
The administrative extremity dose has been exceeded but not the NRC extremity dose; no NRC notification is required.
vC.
The administrative and NRC extremity doses have been exceeded; 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the NRC is required.
D.
The administrative and NRC extremity doses have been exceeded; immediate notification to the NRC is required.
Reasons:
A.
The extremity dose is 55 rem.
T.he administrative limit is 40 rem; the NRC limit is 50 rem.
B.
The extremity dose is 55 rem.
The administrative limit is 40 rem; the NRC limit is 50 rem.
D.
The extremity dose is 55 rem.
The administrative limit is 40 rem; the NRC limit is 50 rem.
Immediate notification is required for a dose of 250 rem.
NEW; 10CFR20 pages 292 & 309; ROT-5-43 page 3 & 10; HPP-300 page 4; CP-151 page 25 NRC97.TST Version: 0 Page: 78
71. rot 452 001/ B8// 0040101015/ 004K613// 3.3/ 33/ MU Which of the following would accomplish an addition from the
"B" RCBT to the make-up (MU) system?
A.
Batch size set at 60; totalizer set at 80; start the "B" RCBT transfer pump.
B'.
Batch size set at.60; totalizer set at 80; open the feed valve (MUV-103).
C.
Batch size set at 80; totalizer set at 60; start the "B" RCBT transfer pump.
vD.
Batch size set at 80; totalizer set at 60; open the Feed valve (MUV-103).
Reasons:
A.
The batch size has to be larger than the totalizer.
MUV-103 has to be opened.
B.
The batch size has to be larger than the totalizer.
The transfer pump starts automatically when MUV-103 is opened.
C.
MUV-103 has to be opened.
NEW; ROT-4-52 pages 37, 41, 42 & 49
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NRC97.TST Version: 0 Page: 79
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72. rot-4-13 001/B6//0130101001/013K201//3.8/33/ESAS
All power is -lost to 'the A and C emergency safeguards (ES)
.
>
"
channels.
How will this affect the ES system?--
4 e
!
.A.
The -ES. circuit.s within the affected cabinets will be
,
}
inoperable, all ES channels will trip but no ES
-
equipment wi'll reposition.
'
-
i
-
8.
'The ES circuits within the affected cabinets will trip..
I and all ES equipment-will fail to.its ES position due to
-;
'
!-
a loss of power except the emergency feedwater pumps.
j VC.
The ES circuits within the affected cabinets will trip, and all ES equipment wi,'l fail to its ES position due to
,
i i
,a loss of power except the decay heat pumps.
(
l D.
The ES circuits within the affected cabinets will trip,
and all. ES equipment should respond to its ES position.
'
,
Reasons:
- '
-
A-ES actuation. circuits actuate on loss of power.
This will i{
,
l
- send a signal that actuates the equipment.
I
!,
'
B.
ES. actuation circuits actuate on loss of power.
This will send a signal that actuates the equipment.
i
.
)
4-C.
ES' actuation circuits actuate.on loss 'of power.
This will send a : signal that. actuates the equipment.
4-
-
- .
i i
e d
e BANK; ROTS K - T2; ROT-4-13 88; ROT-4-13 page 3-
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73. rot.5-99 001/A2//3440403001/00025AK202//3.2/44/EoP-11 The plant is in mode 5 when the following occurs:
!
-
The ' operating decay heat pump,; DHP-1A, trips on overload.
l
.The. standby decay heat pump, DHP-1B, is started, cavitates,
-
L
. and trips, L
-
Reactor coolant level is 130'.
Rear. tor coolant temperature is 93* F.
L
-
Upper hand. holds on the steam generator are removed.-
-
How can core heat removal be restored?
l
'
!
.
A.
Establish decay heat removal.with DHP-1A after overloads
,
are. reset.
l v8.
Establish high pressure injection.
C.
Establish steam generator heat removal.
D.
Establish decay heat removal using spent. fuel cooling.
i l
Reasons:
i A..
If the overloads reset the cause of the cavitation problem of the second pump should. be rectified prior to starting the fi rst pump.
C.
Upper' hand holds are removed.
D.-
Because of seismic concerns spent fuel cooling is not to be used.for decay heat removal.
.
o I
!
NEW; 3440403008;. ROT-5-78 B6 & B16; ROT-5-99 pages 13 & 14; OP-404 pages.6 & 12; OP-301 pages 5 & 6; Night Order i
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NRC97.TST Version: O Page: 81 i
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74. rot-5-102 002/ A2// 3440403001/ E04EK1,3//4.0/44/ EoP-04 A step in E0P-04, Inadequate Heat Transfer, states:
l HHEN either oTSG PRESS is between 72s and 600 psig, TliEN bypass EFIC MS and MFW isolation actuation.
What is the basis of this step?
l
,
l
.
A.
This prevents a steam release to the atmosphere while trying to establish adequate heat transfer.
B.
There is limited primary to secondary heat transfer.
L This prevents an uncontrolled steam generator depressurization.
.
vC.
Since there is no annunciator alarm associated with this
function, it prevents ar. unwanted actuation of' main l-feedwater and main steam line isolation.
!
D.
Steam generator pressure must be reduced slowly.
This i
prevents the transient from causing a loss of adequate subcooling margin.
L Reasons:
A.
The atmospheric dumps are not isolated by thic step; a steam release to atmosphere is still possible.
B.
Normal depressurization of the steam generator is a slow
. process, there is no need to isolate.
,
D.
Gradual depressurization of the steam _ generator will not cause a loss of adequate subcooling margin.
l L
NEW; ROT-5-102 pages 3 & 4 l
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i NRC97.TST Version: 0 Page: 82 l
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.75. rot-4-10 001/B2//0150101005/013K301//4.3/33/Ni The plant is operating at full power.
The SASS hodule for
nuclear instrumentation (NI) power to integrated control system (ICS) is selected to manual and NI-5/6 to facilitate trouble
'
shooting.in the auto-transfer circuit.
The lower chamber of NI-5 fails.
How will this affect the plant?
A.'
Assumming a failure of the instrument high, a reactor
.
trip occurs due to the'large positive imbalance indicated by the failed instrument.
B.
Assumming a failure of the instrument high, a feedwater-to-reactor cross-limit occurs and causes and increase in actual feedwater flow to the OTSGs, decreasing RCS pressure to the reactor trip setpoint.
C.
Assuming a failure of the instrument low, a
-
reactor-to-feedwater cross-limit occurs and causes a reduction in actual feedwater flow to the OTSGs, increasing RCS pressure to the reactor trip setpoint.
.
vD.
Assuming a failure of the instrument low, a large l
positive imbalance is detected by "A" channel of_RPS causing a trip of this channel.
,
i
-
. Reasons;
.
A.
A high imbalance would only trip RPS channel "A".
.
B.
There is no NI signal sent to the feedwater circuit.
o C.
The circuit that sends the NI signal to ICS high selects
.
between NI-5 and NI-6.
In this case NI-6 would furnish the signal
.no change to the ICS.
.
l
BANK; ROT-4-10 36; ROT-4-10 page 23; ROT-4-09 page 63: ROT-4-14
page 24 i
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l 76. ROT-5-106 001/ B1/ /1150101020/ 002A103// 3.8/ 55/ oP-209 l
The plant is being shutdown for the mid-cycle maintenance l
outage.
A problem on the grid causes the turbine to attempt to pick-up a large amount of load.
The following occurs:
l
-
T,, RCS pressure and steam pressure begin to decrease.
ay
-
Reactor power is increasing.
When the operators stabilize the plant the following conditions exist:
-
Reactor and feedwater control are in manual
-
Reactor power is 30%.
-
T,is 522*F.
ay
-
RCS pressure is 2000 psig.
Main generator load is 200 MWe.
-
What adtion should be taken?
A.
Raise T, by immediately withdrawing corstrol rods.
ay B.
Raise T, by increasing reactor power within 15 minutes.
ay vC.
Raise T,y, by immediately reducing turbine load.
D.
Establish mode 3 by shutting down the turbine and inserting the control rods within 30 minutes.
Reasons:
~
A.,
B., & D.
The limit and precaution in OP-204 states that if T,y, cannot be restored using secondary parameters then the control rods are inserted and mode 3 conditions are established.
l l
NEW; OP-204 page 4; Technical Specifications page 3.4-3
l NRC97.TST Version: O Page: 84
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-77. rot-4-09 001/B1//0130101002/016X107//3.7/33/NNI Which of the following is a safety function input of the-
-
reactor coolant system wide range pressure transmitters?
i i
'
.l VA.
Actuation for high pressure injection.
>
l
.
B.
. Variable low pressure trip.
'l
,
C.
Remote shutdown panel " automatic closure-initiation"
'
actuation.
D.
Inputs for interlocks associated with pressurizer spray.
l
'
Reasons:
B.
Narrow range transmitters accomplish this function, i
C.
A low range. transmitter accomplishes this function.
i D.
Narrow range transmitters accomplish this function.
l MODIFIED BANK; ROT-4-09 23; ROTS J - T10B; ROT-4-12 B3; ROTS K'
- T1; ROT-4-09 pages 13-16 i
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78. rot 5-48 001/#2//3410103029/2.3.6//3.1/44/ WASTE The shift supervisor on duty (550D) must sign all radioactive liquid release permits prior to the initiation of the release.
What is the purpose of the SSOD signature?
!
,
l vA.
It serves as approval to complete the. release.
!
B '.
It serves to ver.ify the appropriate liquid radiation
'
monitor is operating' properly.
.
l C.
It acknowledges and confirms the estimated volume of l
fluid to be released.
l D.
It acknowledges and confirms the estimated amount of
. radioactivity to be released to the environment.
l Reasons:
l B.
This is done after the signature.
C.
Volume of t.he release is determined by the size of the I
tank.
.
D.
The signature acknowledges only the completion of the j
chemistry portion of the permit not the amount of reactivity release.
,
NEW; ROT-5-48 pages 6 & 14
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79. rot 543 002/ B2// NTS/2.3.4//3.1/ 3N RAD
!
i
- The following information is given for a fully trained, male Crystal River radiation worker:
,
- Age:-
- Lifetime exposure: 39.25 rem
'
- Exposure history:
On File f
What is this worker's annual administrative limit?
.
)
t.
-
,
'
A.
5 rem j
B.
4 rem
C.
.3 rem i
vD.
1 rem
Reasons:
E A.
The. annual limit is 4 rem or 1 rem if the lifetime dose is
> the worker's age.
B.
The annual limit is 4 rem or 1 rem if the lifetime dose is 2 the worker'.s age.
C.
The annual limit is 4 rem or 1 rem if the lifetime dose is a the worker's age, l
i
!
MODIFIED BANK; ROT-5-43 18; ROT-5-43 page 3; HPP-300 page 4
,
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'80. rot-3-22 001/84//0000501028/061K301//4.6/33/EXC HEA The reactor has.just tripped.
Primary and secondary parameters
,
i
are-RCS T,
532*F ay
RCS pressure 1720 psig
,
'
PZR level 50 inches OTSG levels 25% each OTSG pressure A 700 psig OTSG pressure B 620 psig
.
!
Makeup flow 190 gpm L
Emergency feedwater flow 480 gpm each What transient is in progress:
)
.
.
A.
Steam generator tube rupture B.
Small break loss of coolant accident vC.
Small steam line break D.
Small feed line break
'
Reasons:
R A.
The low RCS temperature along with the steam generator levels and fill rate would indicate an overcooling not a tube leak.
<
B.-
A surplus of subcooling as well as the high steam generator fill rate would rule out a SBLOCA.
,
D.
The large amount of subcooling would eliminate excessive heat transfer as the cause.
'
MODIFIED BANK; ROT-3-22 8; ROT-3-22 pages 4-10-NRC97,TST Version: 0 Page: 88
,
81' rot-4-14 002/G1//0410101001/041K302//3.9/33/ICS CR-3 is operating at 100% full power with all control systems
in automatic.
If the turbine bypass valves on the "B" OTSG,
'
fail full open, what is the final stabilized response of reactor power, T,y, and MW electrical?
i
,
A, Reactor power will increase, MW electrical will remain.
constant, T,y,.will lower.
vB.
Reactor power will increase, MW electrical will lower, and T, wi'll lower.
ay
.
C.
Reactor power will remain constant, MW electrical will lower, and T will remain constant.
l ave D.
Reactor power will remain constant, MW electrical will remain constant, T, will lower.
ay Reasons:
A.
MW electrical will lower.
j C.
Reactor power will increase; T,y, will lower.
D.
MW electrical wirl lower.
t
~ BANK; ROT-4-14 8; ROTS 1 - T9; ROT-4-14'pages 18 & 19 i
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NRC97.TST Version: 0 Page: 89
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82. rot-4-63 001/ B1//1030101002/ 029A301//4.0/ 33/ PURGE -
l The plant-is operating in mode 5 when RM-Al actuates.
What is
i the response of the reactor building purge system?
i l
i A.
The purge supply fan (s) will stop and the supply and-exhaust valves will close.
.
.
!
B'.
The' purge supply and' exhaust fan (s) will stop and the.
.
supply and exhaust valves will close.
C.-
The purge exhaust fan (s) will stop and the supply and exhaust valves will close.
,
L I
L vD.
No fan (s) will stop'but the supply and exhaust valves
. will close.
Reasons:
A.
An RM-Al actuation does not trip the purge supply fans.
]
'
.B.
An RM-Al actuation does not trip the purge supply or exhaust. fans.
C.
An RM-A1. actuation does not trip the purge exhaust fans.
.
BANK; ROT-4-63 28; 1030101005; ROT-4-63 pages 28-33; ROT-4-25 i
page 27 l
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NRC97.TST Version: 0 -
Page: 90 I
83. ROT-5-38 001/ A4// NTS/2.1.2//4.0/44/ ol-01 The plant is in mode 3 preparing for entrance into mode 2.
You are the Nuclear Shift Supervisor (NSS).
Which of the following responsibilities must you perform?
A.
Be informed of and oversee equipment starts.
.
v8'.
'Be informed of and ov.ersee reactivity changes.
-l C.
Perform peer checking for reactivity changes.
D.
Perform peer checking for equipment starts.
Reasons:
A.
This may also be a function of the ANSS.
C.
Any SRO can provide this function.
D.
Monitoring or peer checking of equipment starts may be performed by the R0s or SR0s.
NEW; OI-01 page 2
.
NRC97.TST Version: 0 Page: 91
.
84. ROT-5-84 001/ A2//3440403001/00C35EK308//3.9/44/ AP-470 An electrical system upset has occurred.
The following conditions exist:
-
The plant is at 100% power.
-
Instrument air pressure is 50 psig and decreasing.
-
Reactor coolant pump conditions:
Pump Thrust bearing temp.
SW exit temp.
A 165'F 150*F
-
B 167*F 115 2 *'F C
164*F 151*F D
168*F 154*F
-
Main turbine bearing conditions:
Bearing Bearing oil discharge temp.
Bearing metal temp.
(78'F 224*F l
-
185'F 226*F j
180*F 224*F
,
185'F 228'F
175*F 218'F
178'F 227*F
183*F 225'F
171*F 215'F
180*F 226*F
-
All control rod stators are between 173 * F and 177'F.
What must be done, if anything, in this situation?
A.
The reactor is not required to be tripped with these conditions.
<
B.
The reactor must be tripped because CRD stator temperatures are too high.
C.
The reactor must be tripped because SW flow has been lost to the RCPs.
vD.
The reactor must be tripped because SC flow has been lost to the main turbine.
NRC97.TST Version: 0 Page: 92
'
_. _ _ _ _. _. _ _ _ _ _.. _.. _.. _. _. _
84. rot 5-84 001/A2//3440403001/00065EK308//3.9/44/AP-470 Reasons:
'
A. 'The reactor is required to be tripped because SC flow must have been-lost.
'
B..CRD stator temperatures are not too high.
C.-.RCP conditions are within limits.
,
\\
.
NEW; ROT-5-84 pages 5 & 6; OP-302 pages 6 & 7; OP-203 page 9; OP-502 page 4
i
.
NRC97.TST Version: 0 Page: 93
85. rot-4-56 003/B6//0080401001/00026AA102//3.3/33/SW The control room has entered AP-330, Loss of Nuclear Service (SW)..The control room operators have tripped the reactor and are monitoring temperatures of equipment cooled by SW.
Which of the following actions can the control. room operator perform to increase cooling to essential SW cooled equipment?
vA'.
Secure SW to the control rod drive (CRD) stator jacket coolers.
- B.
Stop the spent fuel pumps (SFPs) and secure SW to the spent. fuel coolers.
C.
Bypass the make-up demineralizers and then secure SW to the letdown coolers.
'D.
Isolate SW to the secured reactor building fan (AHF-1A, AHF-1B, or AHF-1C).
T Reasons:
-
B.
Securing-the SFPs is possible from the control room but not securing the SW to the spent fuel coolers.
,
!
.
C.
Letdown may still be needed; closing the SW valves to the j
coolers would isolate letdown.
4-D.
SW to the. secured fan can be performed.from the control
room but currently is manually isolated.
i
!
NEW; 0080401003; ROT-4-56 pages 5, 7 & 18
' NRC97.TST Version: 0 Page: 94 l
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86. rot-4-09 003/B6/10160101010/0 X)27AA215//4.0/33/NNI The following plent conditions exist:
'
-
Reactor power is 100%.
-
Pressurizer level transmitter RCl-LT1 reads 220 inches.
-
Pressurizer level transmitter RCl-LT2 reads 160 inches.
.
-
Pressurizer level transmitter RCl-LT3 reads 320 inches.
The pressurizer high level annunciator is in alarm.
-
-
MUV-31, pressurizer make-up valve is in automatic and closed.
,
Which of the following actions will result in proper pressurizer control?
.
A.
. RCl-LT3 has failed high; transfer level control to RC1-LT2.
vB.
RCl-LT3 has failed high; transfer level control to RC1-LT1.
.
C.
RCl-LT2 has failed low; transfer level control to RCl-LT3.
D.
RCl-LT2 has failed low; transfer level control to RCl-LT1.
Reasons:
A.
Pressurizer level control cannot be transferred to RCl-LT2.
C.
RCl-LT2 is not temperature compensated and should read 60 inches lower than actual.
No transfer to RC1-LT3 should be made since it has failed high.
D.
RCl-LT2 is not temperature compensated and should read 60 inches lower than actual.
NEW; ROT-4-09 pages 9 & 10 NRC97.TST Versian: 0 Page: 95
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87. ROT-4-13 003/B6//0130401001/006K3011/4.2133/ES The plant is conducting a normal plant shutdown.
RCS pressure is 1200 psig,
-
RCS temperature is 400*F.
c
-
'
-
The high pressure injection (HPI) has been bypassed per procedure.
.
An RCS pressure transient occurs which increases RCS pressure
.tp 1750 psig.
Which of the following. describes the. effect(s),
,
i ifiany, on the RCS7
-
!
A.
Both the HPI actuation and bypass bistables will automatically reset resulting in an RCS pressure and
,
. inventory increase.
B.
The HPI actuation bistable will automatically reset; the
bypass bistable will not automatically reset.
There l.
will be no affect on the RCS pressure and inventory.
vC.
The HPI actuation bistable will not automatically reset, the bypass bistable will automatically reset resulting in an increase in RCS pressure and inventory.
'
D.
Neither the HPI actuation bistable nor the bypass
bistable will auto reset.
There will be no affect on
the RCS pressure or inventory.
,
i-Reasons:
,
A.
Only the bypass bistable will reset not both.
l B.
The bypass bistable will reset not the actuation bistable.
D.
The bypass bistable will reset.
4 MODIFIED BANK; ROT-4-13 5; ROTS ] - T7; ROTS K - T2; ROT-4-13 pages 7 and 17 I
- NRC97.TST Version: 0 Page: 96
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.
88. rot-4-60 004/B18/10020101018/00015AA103//3.8/33/RCS A cooldown is in progress with the
"A" and "C" reactor coolant i
pumps (RCP-1A and RCP-1C) running.
The following occurs:
-
Instrument air has been lost to parts of the auxiliary building.
-
MUV-16, seal injection control valve, has failed closed.
-
A r.ociear services closed cycle cooling (SW) piping failure has caused SWT-1 to empty.
-
-
The reactor operator has shut down all the SWPs.
What effect will this have on RCP operation?
A.
The RCPs may continue to operate as long as the control
. bleed off valve remains open.
B.
The RCPs may run for 5 minutes without SW; there are'no time restrictions on pump operation for a loss of seal injection.
C.
The RCPs may run for 5 minutes without seal injection.
There are no time restrictions on pump operation for a loss of SW.
vD.
The RCPs must be immediately shutdown due to loss of cooling to the RCP seals.
Reasons:
A.
When SW and seal injection are both lost the RCP must be tripped whether or not the control bleed off valve is open.
B.
If just SW had been lost the 5 minutes would be accurate but with the loss of seal injection the RCPs must be tripped.
C.
When SW and seal injection are both lost the RCP must be t ri pped.
l l
NRC97.TST Vension: 0 Page: 97
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88. ROT 460 004/ B18// 0020101018/ C0015AA103// 3.8/ 33/ RCS MODIFIED BANK; ROT-4-60 110; OP-302 page 5; ROT-4-60 page 25
.
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I NRC97.TST Version: 0 Page: 98
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'89. rot-4-14 001/B1//0410101001/050A410//3.8/33/ICS L
-
While at 85% full power the following occurs:
!
p A'feedwater transient has caused T,y, to increase.
-
In an effort.to lower T,y, the operator takes the reactor
-
,
diamond and bailey'to hand to move' control rods.
i
,
I t
I
While attempting to reduce: power, feedwater increases rapidly,
>
l what could have caused this increase?--
<
i i
.
.
A.
The total-feedwater flow control circuit is increasing
L.
-feedwater demand.
.
'
E
.B-The megawatt calibrating integral is increasing
.
'feedwater demand.
i vC.
Feedwater is attempting to correct T,y,.
D.
Feedwater is responding to a cros.s limit condition.
- -
,.
o
i Reasons:
.A.< The circuit is used only when 3 RCPs are operating.
!
'
t B.
The megawatt calibrating integral does not control feedwater.
t
~
l'
O.
There would be'no cross limit under these conditions.
!-
!
L
. MODIFIED BANK; ROT-4-14 138; ROT-4-14 page 26-29
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l 90. rot-413 004/BS//0130501002/00069AA202//4.4/33/ESAS
'
A plant transient causes a high pressure injection (HPI)
,
actuation and then shortly thereafter'a reactor building isolation and cooling (RBIC) actuation.
The. shift supervisor
_
asks the control board ' operator to verify indication:on the ES
statusLpanel:
Component Light Component Light j
AHV-1B GREEN AHV-1A GREEN-
]
,!.
AHV-1C GREEN AHV-1D GREEN
'BSV-3 YELLOW BSV-4 YELLOW j
B5V-12 GREEN BSV-11 GREEN BSV-17 GREEN B5V-16 GREEN
,
CAV-1 GREEN CAV-2 GREEN CAV-3 GREEN CAV-6 CREEN i
CAV-4 GREEN CAV-7 GREEN j
,
CAV-5.
GREEN SW-79 YELLOW j
CAV-126-GREEN-SW-80 YELLOW SW-81 YELLOW SW-82 YELL 04
-
SW-83 YELLOW SWV-84 YELLOW SWV-85 YELLOW SWV-86 YELLOW If the SW surge tank level was at 10 feet, what would you report about these particular indications?
.
A.
BSV-3 and BSV-4 are indicating closed and should be closed,. all other components are in their expected
,
position.
I vB.
BSV-3 and BSV :4 are indicating closed and should be i
open, all other components are in their expected i
position, i
C.
BSV-3 and BSV-4 are indicating open and should be closed; the SWVs are indicating open and they should-be closed.
i D.
BSV-3 and BSV-4 are' indicating open and should be open; the SWVs are indicating closed and they should be open.
'
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NRC97.TST Version: 0 Page: 100
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.
90. ROT-4-13 004/B6//0130501002/00069AA202//4.4/33/ESAS
'
1-
,
~
Reasons:
i J
.A.
BSV-3 and BSV-4 should indicate green which means they are open.
)
,
.
-
.C.
A yellow indication for BSV-3 and BSV-4 indicates they are j
closed and they should be open.
- The SWVs are indicating
'
.
open and should be open.
.
D.
A yellow indication for BSV-3 and BSV-4 indicates they are f
closed.
The SWVs are indicating open and should be open.
-l
.
d NEW; ROT-4.-13 pages 32-34, 54-& 55
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91. rot 428 001/ B13/ / c010101009/ 00iK105//4.4/ 33/ CRD The instrument technicians are conducting SP-110 on reactor protection system (RPS) channel "C".
The channel has been placed in channel bypass and is tripped.
A failure in RPS
,
channel
"D" causes its power supply to trip.
How does the plant respond to this situation?
i
??.
All ~ control rod drive' (CRD) br'eakers remain closed -
.
only 1 RPS channel is' tripped.
B.
With channel "C" and "D" tripped -the regulating control rods.de-energize.
C.
With channel
"C" and "D" tripped all CRD breakers open.
vD.
The CRD breakers associate with RPS channel "D" open.
Reasons:
i A.
With no power to the
"D" channel its associated breaker will be open.
,
'
B.
If both channels had been tripped (without
"C" in bypass)
all of the rods would have de-energized.
C.
If both channels had been tripped (without
"C" in bypass)
all of the rods would have de-energized.
i NEW ROT-4-12 B2; 0120101010; ROT-4-12 pages 9-12; ROT-4-28 pages 17, 18 & 39 NRC97.TST Version: O Page: 102
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I 92. rot.5-31001/ B3//101c501001/ Ao6AK3.3// 4.2/ 33/ AP-990 '
!
Plant control.has been moved to the remote shutdown panel due
'
to a. fire in the control room.
AP-990, Shutdown Outside.the Control Room, directs the breakers from the auxiliary L<
transformer to the ES buses be opened and their DC control l
power removed.
What does this accomplish?
!
.
L A".
It aTlows for local
. manual operation of the breakers.
i B.
It disables the 4160V ES under voltage. lockouts.
vC.
It de-energizes the closure circuit of these breakers.
i r
D.
It allows for closure of the breakers using the' switch o
. on the breaker cubicle.
i i
,
Reasons:
I
)
A.
Local-manual operation is available with or without DC l
control power.
l B.
Opening the DC. knife switch does not disable the ES UV
!
lockouts.
i D.
There is no switch on the breaker cubicle that would be
'
used to close the breaker.
.
L NEW; ROT-5-31 pages 11 & 12 i
-
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93. rot-5-01004/A10/ ROT-2-37/3410103036/00069AK101//3.1/44/ LEAK RATE L
Following a large break loss of coolant accident the following j
conditions exist:
l
-
Reactor building (RB) temperature is 267'F.
Reactor coolant-temperature is 300*F.
-
-
A small breach in the containment is leaking 30 ft / min.
l
-
I-131 concentrations in the reactor building are high.
What is the most effective means of reducing.the release rate?
'
i
,
'
[
A.
Ensure two containment cooling trains are in service.
i B.
Start both of the operating floor fans.
C.
Begin a mini-purge.
vD.
Establish containment building spray.
I l
. Reasons:
A.
AHF-1s will cool the building and lower the pressure but will not affectively remove iodine from the RB atmosphere.
B.
The mini-purge is an affective means of removing iodine from the RB atmosphere but is not very useful in lowering the RB pressure.
C.
The operating floor fans will do nothing more than stir the air around in the RB.
!
NEW; ROT-2-37 G15; Technical Specifications pages 3.6-15, B3.6-29 & B3.6-34 - B3.6-38; Thermo manual, chapter 5, page 19;
,
3410103037; ROT-4-63 pages 20, 25, 26 & 33 i
I
.
NRC97.TST Version: 0 Page: 104
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rot-5-100 001/B4//0000501031/00055EA203//4.7/33/EoP-12
'
While in mode 1, both 4160V ES buses de-energize.
The l
immediate actions for a reactor trip have been completed.
All attempts to re-energize the 4160V ES buses have failed.
Which of the following statements defines procedural guidance for this situation?
l A'.
Use E0P-02, Vital System Status Verification, 'for plant
~ control and OP-705, Emergency Power - DC System, to minimize battery power usage.
,
B.
Use E0P-02, Vital System Status Verification, for plant control and AP-770, Emergency Diesel Generator Actuation, to minimize battery power usage.
C.
Use E0P-12, Station Blackout, for plant control and OP-705, Emergency Power - DC System, to minimize battery power usage, vD.
Use E0P-12, Station Blackout, for plant control and to minimize battery power usage.
Reasons:
A. B. & C.
The conditions given are entry conditions to E0P-12.
Guidance for battery usage is included there.
NEW; ROT-5-100 B1; ROT-5-100 pages 1, 2 & 7 I
i l
NRC97.TST Version: O Page: 105
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95. rot-5-97 0071B3//0000501021/00074EK301//4.2/33/EoP-07 You have reached a step in E0P-07, Inadequate Core Cooling, that states:
L Maintain OTSG Tsat Lower OTSG PRESS using TBVs 100*F below Tsat for or ADVs.
existing RCS Press until
,
l 0TSG heat removal is
,
established.
.
Ifsthe reactor coolant system (RCS)' temperature is 320'F and pressure is 45 psig, which of the following cooldown rates is the' highest you can use under these conditions?
i
' VA.
s 100*F per 30 minutes.
B.
s 50*F per 30 minutes.
C.
s 25*F per 30 minutes.
D.
s 5'F per 30 minutes.
Reasons:
,
B.
This is the cooldown rate for > 280* F when the system is
,
not superheated.
C.
This is the cooldown' rate for 280-150*F when the system is not superheated.
D.
This is the cooldown rate for > 280* F when the system is not superheated (natural circulation).
NEW; ROT-5-97 pages 7 & 8 i
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b 96. rot 5-01001/A1//3410103036/2133//4.0144/TS
}
Given the following reactor coolant system (RCS) conditions:
'
-
The' plant is t at -100% FP.
Reactor coolant ~ drain tank level- (RCDT) has increased 1 inch
-
.
in the last 4 minutes. (32.9 gallons per inch).
,
-
The following makeup system (MU) leakage has been identified:
H RCS to. letdown cooler "A"
- 0.05.gpm j
- j MUV-8 packing leak. (directed to floor drain)
--1.5 gpm'
,
'
,
MUP-1A pump seal leak'
- 0.1 gpm j
p MUV-393. packing leak
- 0.2 gpm
- ' Unidentified RCS leakage is 0.4 gpm.
l t
Evaluate RCS leakage?
i
-
l A.
All leakages are within technical specification limits.
[
B.
Pressure boundary leakage exceeds technical L
specification limits.
'
C.
Unidentified leakage into the RCDT exceeds technical i
'
'
specification limits.
i i
vD.
Total identified leakage exceeds technical specification
limits.
i
!
Reasons:
j A.
Not all leakages are acceptable per TS 3.4.12.
B.
There is no pressure boundary leakage listed.
C.
Leakage into the RCDT is identified leakage.
MODIFIED BANK; ROT-5-01 81; 3410103037; Technical Specifications page 3.4-22
>
NRC97.TST Version: 0 Page: 107
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.
n 97. rot-4-60 006/B14//0020101009/004KGO.?)/3.0133/RCS
'
While in mode 5 mainter,ance' needs to be done inside a section l
of reactor coolant 1(RCS) piping.. Chemistry's last sample of b
the RCS had the.following information:
i I
-
pH 7.1 l
fluorides not detectable
-
,
!
-
chlorides not detectable
'
dissolved oxygen 1.5 ppb
-
, hydrogen.
8 std cc/kg of water i
.
.
!
What effect, if any, do these chemistry parameters have on the L
planned maintenance work?
l l
L i
e A.
. Work can go ahead; none of these chemistry parameters are out of specification.
l B.
Work can go ahead; ;the work instruction must indicate
!
that no chemicals will be used that can increase l
chloride / fluoride concentration.
vC.
Work. cannot go. ahead; the hydrogen concentration is too high and may be an explosive hazard.
D.
Work cannot go ahead; opening the piping would cause the oxygen concentration to go up in the RCS.
L Reasons:
L A.
The' hydrogen concentration is too high to open the RCS
' system for maintenance..
B.
The hydrogen concentration is too high to open the RCS system for maintenance.
l D.
The RCS should be isolated before opening up -that l
particular.part of the system; oxygen concentration will
!
not-be'affected.
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NRC97.TST Version: O Page: 108 I
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' 98. ' rot 4-69 002/081/OS50401001/00051AA202//4.1/33/CD.
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- TheLplant is. operating at 68% power when the following occurs:
)
'
- Annunciator "TURB VACUUM'PRETRIP" is received.
l H
The> turbine. building operator reports vacuum reading on the
['
. main pedestal E (TB-135-PI) 24.2" Hg.
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The control board '(CD-007-PIR) vacuum reading is 6"Hga.
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Vacuum -is. slowly degrading.
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.What'is the1 correct operator response.to this situation?
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t g
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v'A.
Reduce reactor power to < 45% and trip the turbine
[
within 5' minutes of exceeding the vacuum' limit..
.
t
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B.
' Immediately trip the reactor and turbine.
.C.
Ensure the standby condenser air removal pump'. starts and
- . '
trip the reactor and turbine when ' vacuum decreases to <-
{
~
20 inches Hg.
y
D.
Investigate the cause of the vacuum leak and begin a
. power decrease if vacuum decreases < 23 inches Hgg.
.
.
' Reasons:
.
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B.
There is no need to trip the reactor', power can be reduced and then the turbine is shutdown.
t C.
The second air removal pump should have already auto started.
The reactor and turbine should-be tripped'
prior to reaching 20 inches Hg.
y
,
D.
Vacuum leak may continue but the turbine 'should' have automatically tripped.
BANK'; ROT-4-69 30; ROTS J - Final 96; OP-607 page 4
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99. rot-5-40 001/ B3// NTS/ 2.2.13// 3.8/ 33'CP 115
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'Who must approve:any addenda to a block tagout?
?
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A.
Clearance Authority. prior.to Nuclear Shift Manager t-
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Nuclear ' Shift Manager prior.to the Clearance Authority C '.
Nuclear Shift S.upervisor prior to Clearance Authority
^
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D.
Clearance Holder prior to Nuclear Shift Supervisor t.
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Reasons:
A.c The.NSM approves' addenda to block-tagouts prior to thel l
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Clearance Authority.
l l
C.
Usually the.NSS and the Clearance Authorityjare the same; the NSM must-also approve addenda to block tagouts.
.D, The Clearance Holder does'not approve addenda; the NSM -
must also approve' addenda to block tagoute l
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BANK;. ROT-5-40 1; NRC 5-93; ROT-5-40 813; ROTS 1 T10B; CP-115 I
pages.2 & 23.
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100. rot-3-24 001/ B2//0000501022/2.3.10//3.3/33/ RAD
- EOP-6, Steam Generator Tube Rupture, allows steaming both steam
,
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i generators provided specific conditions are met.
What is the l
basis for this step?
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[
vA.
It minimizes the possible radiation exposure the plant p
,
personnel and the public receive.
B.
.It minimizes the amount of contaminated water that must l
- be stored in the turbine building.
'
C.
It. reduces the probability of leak flow increasing due to. stresses that would be created if the leaking steam
-
gene.rator were isolated.
D.
It reduces primary to secondary leak rate by reducing
',
primary to secondary differential pressure.
.
C i
m.
e.3 :
.
B.
It actually increases the amount of contaminated water in the-turbine building.
,
C.
It can increase the stresses in the steam generator.
D.
It can increase the probability'of increasing the leak.
NEW; 0350501001; ROT-3-24 page 1; RO',T-5-101 page 2, 21 & 22
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l Thursday, June 19,1997 @ l143 AM Answer Key Page: i Test Name: NRC97.TST Test Date:
Monday, June 09,1997 Answer (s)
l
_
Question !D Type Pts 0 1 2 3 4 5 6 7 8 9 i
i 1:
1 ROT-5 36 002 MC-SR
D A B C D A B C D A l
1:
2 ROT-4-60 003 MC-SR I
D A B C D A B C D A 1:
3 ROT-5-97 001 MC-SR
.A B C D A B C D A 'B 1:
4 RO f-4-60 002 MC-SR
D A B C D A B C D A 1:
5 ROT-4-25 001 MC-SR
B C D A B C D A B C
1:
6 ROT-4-10 003 MC-SR
A B C D A B C D A B 1:
7 ROT-5-01 003 MC-SR
D A B C D A B C D A 1:
8 ROT-4-81 001 MC-SR
B C D A B C D A B C 1:
9 ROT-4-07 002 MC-SR I
C D A B C D A B C D 1: 10 ROT-4-07 001 MC-SR
A B C D A B C D A B 1: 11 ROT-4-13 002 MC-SR
C DA B C D A B C D 1:
12 ROT-4-26 001 MC-SR
B C D A B C D A B C 1:
13 ROT-5-01 002 MC-SR I
C D A B C D A B C D 1:
14 ROT-5-38 003 MC-SR I
B C DA B C D A B C
'
1:
15 ROT.5-42 001 MC-SR
A B C D A B C D A B 1:
16 ROT-4-09 002 MC-SR
A B C D A B C D A B 1:
17 ROT 3-03 001 MC-SR I
C D A B C D A D C D 1; 18 ROT-5-96 002 MC-SR
A B C D A B C D A B 1: 19 ROT-5-101 002 MC-SR I
A B C D A B C D A B 1-20 RO_T-5.-$5 002 MC-SR
D A B C,_,D A B C D A 1: 21 ROT-4-12 002 MC-SR
D A B C D A B C D A 1: 22 ROT-4-90 001 MC-SR I
C D A B C D A B C D 1: 23 ROT-5-95 001 MC-SR I
C D A B C D A B C D 1: 24 ROT 5-48 002 MC-SR I
A B C D A B C D A B 1: 25 ROT-5-96 001 MC-SR
B C D A B C D A B C
_
1: 26 ROT-5-77 001 MC-SR
B C D A B C D A B C 1: 27 ROT-5-102 001 MC-SR I
A B C D A B C D A B 1: 28 ROT-5-14 001 MC-SR I
C D A B C D A B C D 1: 29 ROT-4 56 002 MC-SR
D A B C D A B C D A 1: 30 ROT-3-22 002 MC-SR
D A B C D A B C D A 1: 31 ROT-4 66 001 MC-SR I
A B C D
.A B C D A B 1: 32 ROT-4-81 002 MC-SR
D A B C D A B C D A 1: 33 ROT-4-28 002 MC-SR
B C D A B C D A B C
.l : 34 ROT-4 62 001 MC-SR
B C D A B C D A B C 1: 35 ROT-4-69 001 MC-SR
b C D A B C D A B C 1: 36 ROT-4-56 001 MC-SR
A B C D A B C D A B 1: 37 ROT-5-107 001 MC-SR
D A B C D A B C D A 1: 38 ROT-5-34 001 MC-SR
A B C D A B C D A B 1: 39 ROT-5 98 001 MC-SR I
C DA B C D A B C D 1;_ 40 ROT-4 28__
004 MC-SR
B C D A B C D A B C 1: 41 ROT-5-81 001 MC-SR
D A B C D A B C D A 1: 42 ROT 4-64 001 MC-SR D A B C D A B C D A
'
1: 43 ROT-4-15 001 MC-SR
A B C D A B C D A B
!
1: 44 ROT-4-10 002 MC SR I
B C D A B C D A B C
!
1: 45 R OT-4-60 007 M C-S R
D A B C D A B C D A 1: 46 ROT-5-50 001 MC-SR I
L A B C D A B C D A 1: 47 KOT-4 60 005 MC-SR
A 3 C D A B C D A B 1: 48 ROT 5 30 001 MC-SR
D A B C DA B C D A 1: 49 ROT-4-I l 001 MC-SR
C D A B C D A B C D 1: 50 RO ro-109 001 MC-S R
B C D A B C D A B C
.
Thursday, June 19,1997 @ 11:43 AM AriSwer Key Page: 2 Test Name: NRC97.TST Test Date: Monday, June 09,1997 Answer (s)
Question ID Type Pts 0 1 2 3 4 5 6 7 8 9 1: 51 ROT-4-12 001 MC-SR I
D A B C D A B C D A 1: 52 ROT-4-60 001 MC-SR I
B C D A b C D A B C 1: 53 ROT-4-64 002 MC-SR
B C D A B C D A B C 1: 54 ROT-4-54 001 MC-SR
D A B C D A B C D A
,
1: 55 ROT-5-67 001 MC-SR I
C D A B C D A B C D 1: 56 ROT-4-28 003 MC-SR I
D A B C D A B C D A 1: 57 ROT-5-96 003 MC-SR
A B C D A B C D A B 1: 58 ROT-4-63 002 MC-SR I
C D A B C D A B C D 1: 59 ROT 4-06 001 MC-SR I
C D A B C D A B C D 1: 60 ROT-4-60 008 MC SR I
D A 11 C D A B CD A 1: 61 ROT-5-101 001 MC-SR I
C D A B C D A B C D 1: 62 ROT 5-94 001 MC-SR I
C D A B C D A B C D 1: 63 ROT-5-85 001 MC-SR I
C D A B C D A B C D 1: 64 ROT-5-48 003 MC-SR I
C D A B C D A B C D 1: 65 ROT-5-61 001 MC-SR
D 'A B C D A B C D A 1: 66 ROT-4-29 001 MC-SR I
C D A B C DA B C D 1: 67 ROT-5-78 001 MC-SR I
C D A B C D A B C D 1: 68 ROT-5-34 002 MC-SR I
C D A B C D A B C D 1: 69 ROT-4-59 001 MC-SR I
A B C D A B C D A B 1: 70 ROT-5-43 001 MC-SR I
C D A B C D A B C D 1: 71 ROT 4-52 001 MC-SR
D A B C D A B C D A 1: 72 ROT-4-13 001 MC-SR I
C D A B C D A B C D 1: 73 ROT-5-99 001 MC-SR I
B C D A B C D A B C i
1: 74 ROT-5-102 002 MC-SR I
C D A B C D A B C D
)
1: 75 ROT-4-10 001 MC SR
D A B C D A B C D A l
1: 76 ROT-5 106 001 MC-SR I
C D A B C D A B C D
'
1: 77 ROT-4-09 001 MC-SR I
A B C D A B C D A B 1: 78 ROT-5-48 001 MC-SR
A B C D A B C D A B I
1: 79 ROT-5-43 002 MC-SR
D A B C D A B C D A
!
1: 80 ROT-3 22 001 MC-SR I
C D A B C D A B C D
~
f 1: 81 ROT 4-14 002 MC-SR I
B C D A B C D A B C 1: 82 ROT-4-63 001 MC-SR
D A B C D A B C D A j
1: 83 ROT-5-38 001 MC-SR
B C D A B C D A B C j
1: 84 ROT-5-84 001 MC-SR
D A B C D A B C D A
!
1: 85 ROT-4-56 003 MC-SR
A B C D A B C D A B 1: 86 ROT-4-09 003 MC SR I
B C D A B C D A B C 1: 87 ROT-4-13 003 MC-SR I
C D A B C D A B C D 1: 88 ROT-4 60 004 MC-SR I
D A B C D A B C D A 1: 89 ROT-4-14 001 MC-SR I
C D A B C D A B C D 1: 90 ROT-4 13 004 MC-SR
B C D A B C D A B C 1: 91 ROT-4-28 001 MC-SR
D A B C D A B C D A j
'
1: 92 ROT-5-31 001 MC-SR I
C D A B C D A B C D 1: 93 ROT-5-01 004 MC-SR
D A B C D A B C D A
,
1: 94 ROT-5-100 001 MC-SR I
D A B C D A B C D A I
'
1: 95 ROT-5-97 007 MC-S R
A B C D A B C D A B 1: 96 ROT-5-01 001 MC-SR
D A B C D A B C D A 1: 97 ROT 4-60 006 MC-SR I
C D A B C D A B C D 1: 98 ROT-4-69 002 MC-SR
A B C D A B C D A B 1: 99 ROT 5 40 001 MC-SR
B C D A B C D A B C 1: 100 ROT 3 24 001 MC SR
A B C D A B C D A B
__
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