IR 05000327/1992302
| ML20127F250 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 01/04/1993 |
| From: | Aiello R, Lawyer L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20127F230 | List: |
| References | |
| 50-327-92-302, NUDOCS 9301200181 | |
| Download: ML20127F250 (130) | |
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t UNITED ST ATES 4'o NUCLEAR REGULATORY COMMISslON g.
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ATLANTA, GEORGI A 30323
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ENCLOSURE I REQUAllflCATION EXAMINATION REPORT - 50-327/92-302 Facility Licensee:
Tennessee Valley Authority facility Name:
Sequoyah Nuclear Plant Facility Docket Nos.:
50-327 and 50-328 Facility License Nos.: DPR-77 and DPR-79 Written and operating requalification examinations were administered at the Sequoyah Nuclear Plant located near Soddy-Daisy, Tennessee.
Chief Examiner: ~_W34 -.s f d.4u vm
/2 3/- Y 2 Ronald f. Wiello
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Date Signed Approved By:
/M4[h
/ - tsfy Lawrence L. La yer, Chief Date Signed Operator Licensing Section 1 Division of Reactor Safety SUMMARY SCOPE:
Requalification examinations were administered during the weeks of -
November 30 and December 7, 1992.
Written and operating examinations were administered to 8 Reactor Operators.and 16 Senior Reactor Operators. One of the Reactor Operators and one of the Senior Reactor Operators participated only in the simulator portion of the examination to complete the. crew composition. They
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had previously taken-'and passed an NRC-administered requalification examination during the term of their license.
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Another Senior Reactor Operator did not-take the written examination. -He had taken and passed the written during the'1991 requalification examination but' did not participate in the operating examination due to illness.
RESUL'IS:
Seven of the.eight (87.5 percent) Reactor Operators passed the examination. One Reactor Operator failed the walkthrough portion -
of the operating examination.
Sixteen of the sixteen (100 percent) Senior Reactor Operators passed the examination.
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crew simulator examinations were administered. All crews were rated as satisfactory.
9301200181 930105-
-PDR ADOCK 05000327 V
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Enclosure-l
'2 Based on these results, the Sequoyah Requalification Program was determined to be satisfactory.
The examination team identified two strengths in the Sequoyah Requalification Program.
The first was in the area of written test question development-
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(Para. 4.c.1).
The second was in attention to examination security (Para. 4.f).
The examination team also identified two weaknesses. The.first pertained to the Sequoyah Requalification Program in the area of providing operator feedback during operating examination administration (Para.-4~.f.3).
The second was in the area of plant conditions.
The number of temporary identification labels found throughout the plant and the-condition of the 125 L
VOC Battery Board Rooms were identified as shortcomings-(Para. 4.j).
An Inspector Follow-up Item was identified regarding the basis in procedure FR-Z.1 for placing the containment hydrogen ignitors in service prior to determining containment hydrogen concentration (Para. 4.1.2).
Another Inspector Follow-up Item was identified during an inspection-of the facility Licensed Operator Requalification Program.
Two individual
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examination reports were noted to document performance requiring remediation but none was assigned (Para. 4.1).
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i REPORT DETAILS 1.
Persons Contacted
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f Licensee Employees
- R. Beecken, Plant Manager
- L. Bush, Operations Manager
- N. Catron, Emergency Preparedness Engineer
- G. Carroll, Shift Operations Supervisor
- W. Chandler, Unit Operator
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- M. Cooper, Site Licensing Manager
- L. Durham, TVA Nuclear Training Manager
- R.
Fenech, Site Vice President P. Gass,-Instructor
- J. Kell, Unit Operator
- V. Keyser, Instructor
- R. King, Manager, Operations Training T. Lundy, Simulator Engineer S. Michael. Manager, Simulator Support
- D. Moore, Assistant Shift Operations Supervisor
- M. Murphy, Licensed Operator Training Program Manager L. Pauley, Simulator Operator
- R. Proffitt, Compliance Licensing Engineer
- G.
Sanders, Senior Instructor M. Sheperd, Manager, Nuclear Training
- G. Terpstra, Instructor
- R. Thompson, licensing Compliance Manager i
- Attended exit meeting 2.
Examiners
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R. Aiello, NRC, Region II, Chief Examiner S. Cahill, NRC, Region 11-P. Isaksen, INEL-3.
Other NRC Personnel Attending Exit W. Holland, Sequoyah Senior Resident Inspector 4.
Discussion
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Examination Results and. Program Evaluation-This requalifi' cation examination was ' administered under the guidelines of. NUREG-1021, Revision' 7 -(draft).
Sixteen of_s_ixteen Senior Reactor Operators-(SR0s) and seven of eight Reactor-Operators (R0s) passed the examination. One R0' failed the walkthrough portion of the examination.
Six of six crews
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evaluated passed the simulator portion of the examination.
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Based on the examination results, the Sequoyah Requalification-Program meets the: criteria established in NUREG-1021, ES-601.D.2, Revision 7-(draft), and has been determined to.be satisfactory.
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The facility is permitted to administer the reexamination for returning the failed individual tn licensed duties. He will be required to pass another NRC-administered examination to renew his license, b.
Reference Material The examination team reviewed the reference material su, glied by
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the licensee and determined that it was adequate to support the examination.
Particularly noteworthy was the improvement in the Simulator Instruction Manual in response to previous examination report comments. Instructions and acronym lists were added to aid the user. The manual is now clearly organized with a consistent format and is very helpful as an examiner aid.
The Emergency Plan Initiating Conditions Matrix (EPIP-1) was also revised in_ response to problems observed with consistency of emergency event classifications on previous examinations.
The NRC examination team reviewed the new revision in depth.
They noted that many previously vague items have been clarified.
The changes will enhance the operators' ability to consistently classify plant-events. Ongoing attention is being focused on further revisions which will develop a flowchart method to guide the operators to the initial event category.
This will further aid consistency.
The initial effort expended is noteworthy, and the planned revisions should result in a vastly improved procedure.
The licensee supplied a sampling plan describing the requalif-ication cycle and the selection process used for the topics included in the examinations.
The sample plan was compared to the guidance in NUREG-1021, ES-601, Attachment 2.
The selection process met all the guidelines for NRC administered examinations.
The sample plan was particular_ly well organized and clearly sequenced. Clarifications to the process for including high.
priority material _ from outside the sample plan, and the definition of what constitutes material outside the sample plan, were discussed with the facility to improve future submittals, c.
Proposed Examination The examination team conducted a review of the' facility's proposed written, walk-through, and dynamic simulator. examinations. The NRC substituted new material for about five percent of each section of the examination. This was done in accordance with NUREG-1021. The remainder of the examination was approved with minor changes.
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Report _ Details-
(1)
Written Examination The written examination met the guidelines-as outlined in the Examiner Standards (NUREG-1021), ES-602. The examination team determined that the validation-times for questions on both the static simulator and open-reference examinations correctly reflected the-time that a typical operator would require to answer the questions. The examination tested a diverse range.of topics and was written-at a higher knowledge level that was-appropriate for an open reference examination.
Several questions addressed shutdown
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and low power topics as recommended in the new Revision 7 NUREG-1021 guidance. The majority of the questions were well developed and required the operator to synthesize several sources of information.
Questions reflected the high level of attention given to format and consistency.
Very few technical or administrative modifications had to be made.
The two (R0 and SRO) Part A (static simulator) examinations submitted by the licensee each consisted of 15 questions.
Thirteen questions were common to both examinations. All questions were reviewed and validated at the facility prior to administration.
Eight questions were edited to correct:
structural problems and to ensyre examination validity.
None of the questions required complete replacement.
Each Part B examination consisted of 30-questions, 23 of which were-common to both-the-SRO and RO examinations.
All questions were reviewed and validated prior to
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administration.
Eleven questions required editing to correct minor problems other than typographical errors.
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Three questions were altered by the NRC to elicit a different answer and ensure examination validity. None of the questions required complete replacement.
(2)
Simulator Scenarios-The dynamic simulator examinations derived from the sample-plan were reviewed by the NRC examination team and
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-determined to be adequate.
Critical. tasks.were all-important-and safety significant. However, it was necessary to increase the number of Crew Critical Tasks (CCTs) in order to ensure compliance with the simulator scenario review checklist (form ES-604-1, Rev 7).
One scenario set only contained six CCTs as originally submitted.
Since-seven is the minimum requirement, an additional CCT-was generated. A few major; events, such as an ATWS, only needed minor mcdifications to.make them safety significant and-allow their inc usion as CCTs. The importance of' developing
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i significant events to ensure they are safety significant in order to maximize the number of CCTs was discussed with the facility training _ staff.
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All of the scenario event descriptions had to be enhanced to-include expected procedural operator actions.
The original.
level of detail was inadequate and limited their use as an evaluation tool.
The NRC substituted-new material for approximately ten percent of the simulator portion of the operating examination.
(3)
JPMs The examination team discovered, during the prep week, that
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many of the JPMs submitted were recently administered to' the examinees in the weeks preceding the examination. These were eliminated by the examination team to ensure-examination validity. The reduction in;the number of available JPMs resulted in the administration of the walkthrough examination in two parts each week._ The-simulator _ portion of the walkthrough was administerec to_ all of the week's examinees on one day, while the in-plant portion was-administered the following day to the same group. This enabled the examination team to fully validate each JPM in advance and was the most efficient possible utilization of the remaining material.
The proposed JPMs were generally satisfactory.
Five of the-thirteen JPMs used from the proposed examination utilized alternate paths and several also incorporated -low-power situations.
The NRC modified three of the JPMs to utilize-alternate paths. During the prep l week validation process, several of the JPMs were noted to be missing _ minor procedural steps and evolution details.
Several cues also needed to_be revised, in order to be valid feedback that an-operator would expect _to' receive during the assigned task.
J These were identified to the. facility staff and were i
corrected for the administered examinations.
Critical steps l
were properly identified and did not_ require _ modifications.
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d.
Operator Performance (1)_
Written Examination Performance Operator performance on the written examination was very good. All of the operators scored a-combined' grade. of at
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least 95 percent. - Performance on all questions-was uniform, so there were not'any generic _ knowledge or ability-deficiencies' identi fied. Minor differences between the NRC'
and' facility grading were attributable to-the way' a single
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short answer question was graded.
The facility subtracted the full 0.125 points for a partial credit item when the examinee only gave half of the answer.
The NRC grader only
subtracted 0.06, half of theLitem's value.
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There were not any significant differences in operator performance on the NRC-modified questions, versus the unaltered, facility bank questions.
(2)
Simulator Performance Crew performance in the simulator part of the examination was satisfactory.
All six crews passed the simulator examination.
Two SR0s were recognized by both the facility and NRC examination teams to require remediation in the area of Radiological Emergency Plan event classifications.
Both operators used incorrect bases _for classifying their simulator scenarios. One of the two was unfamiliar with the potential _ for radioactive releases from the plant during a --
Steam Generator Tube Rupture and when a radioactive release from the plant could be declared.
Both operators passed the-simulator portion of the examination but will be remediated by the facility training staff in this area.
All crews were noted to consistently implement-the.
Annunciation Response-Procedure (ARPs) when they were required.
Communication among crews was generally satisfactory, although several-instances of unclear reports, failure to properly acknowledge orders, and failure to inform others when leaving the controls area were observed.
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Four operators were unfamiliarLwith-the operation of the condensate demineralizer booster pumps A recent modification to the control circuitry requires the operator to hold the pump control switch in the STOP position for.
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several seconds to allow the pump = suction--valve to go closed before the pump will stop.
Several operators just placed the switch to STOP momentarily, which causes the suction valve-to' travel shut, but does not : secure the pump. The pump. will only stop when the suction valve has travelled
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fully closed and the control. switch is still being held to the STOP position. LThe operators' error resulted in the
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pump-continuing to run with the suction valve closed, which could severely damage the pump.
(3)-
JPM Performance Operator _' performance was adequate.
One individual failed th_is portion of the examination by missing two of five JPMs.
He skipped two critical steps and was generally unfamiliar
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Report Details
with the assigned tasks.
Several other operators were also observed to be unfamiliar with some of the assigned tasks which led to excessively long performance _ times.
One simulator JPM (#67) that realigned the Emergency Gas Treatment System, was time validated-for 20 minutes, but took an average of 37 minutes.
One operator took ? minutes to complete the task.
A generic deficiency was noted in operator response to a failure of the diesel generator to automatically shutdown on high crankcase pressure.
This JPM (#77-1AP)-was signifi-
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cantly altered by the examination team to utilize a new, alternate path, that the operators had not seen before.
All 11 of the operators tested on'this JPM correctly referenced the ARP when the crankcase pressure alarm was received.
However, six of the eleven elected to take the diesel control switch to 0FF to secure the diesel, instead of depressing the emergency stop pushbutton as required.
Depressing the emergency stop pushbutton meets the intent of the ARP guidance which directs-the operator to ensure that
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automatic shutdown has occurred.
Taking the control switch to stop places the diesel in a cooldown mode and allows _ it to idle for ten minutes which is unacceptable. Two_other operators elected to unload the diesel prior to depressing the emergency stop pushbutton.
The NRC examination team determined this was unsatisfactory performance because the time delay to unload the diesel took approximately 90 seconds. This does not meet the intent of the ARP which is to manually ensure a diesel-shutdown that should have occurred automatically and instantly on a high crankcase pressure signal.
The destructive potential of-a crankcase -
explosion necessitates an immediate shutdown.
These two individuals account for the discrepancy between the NRC and the facility examination. team grading on the walkthrough
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examination. The facility considered it acceptable performance as long as the emergency stop pushbutton was eventually depressed within several minutes.
The wording of the ARP for the. aforementioned crankcase pressure condition is not clear and was noted:by the facility to be a contributor to the poor success rate on this JPM. _ Rewording the initial. instruction that directs the operator to ensure an automatic shutdown occurs, to a more specific directive, was. discussed with the facility.
staff. Directing the operator to use the emergency stop push button to stop the diesel would be more desirable guidance and may have eliminated several of the unsatisfactory performances.
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Several other minor problems were noted during the performance of the walkthrough examination.
Some operators were not familiar with the extended time.it takes for the moisture separator reheater' valves to go closed.
Three operators incorrectly delayed the actions of an Emergency-Operating Procedure they were assigned to perform,' to
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address A0Ps and ARPs.
Four operators were not attentive to the use of radios in areas where interference susceptible equipment is located that could trip the plant. They routinely transmitted from these locations even though they-were clearly marked with large yellow placards instructing individuals to use radios only in an emergency due to interference concerns.
Follow-up questioning revealed that they had not considered the possible effects of their actions and had not made a conscious decision to use the radio because an emergency existed.
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f.
Examination Administration The licensee's administration of the examinations was well planned and coordinated.
The licensee put a great deal of hard work into the examination and examination process to ensure the efficiency of the examinations.
Significant attention-was placed on-examination security.
Scheduling and sequestering of the examinees was_well thought out and planned in advance.
There were more than enough personnel available to. assist with coordinating the groups of examinees.
Examination facilities were well marked to prevent nonparticipants from entering.
Management attention to the examination process was notable.
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facility Training Manager was constantly available to ensure the process was going smoothly.
There was a _large attendance at both the entrance and exit meetings. Operations department staff members were readily available to-solicit input from the
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examination team.
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(1)
Written: Examination Administration Scheduling of the written exam was very efficient in that operators from both weeks of the~ examination were shifted.to allow all of them_ to complete the examination -in one: day.
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One minor problem was noted with the administration of the Part A examination.
Operators were provided With the examination prior to-their briefing and'walkdown of the control boards.. This could_potentially allow'the examinees to have the examination for longer than-the one hour; allotted.
Subsequent examinations were not handed out to
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the examinees until.they were briefed and had walked down the control' boards.
This is considered an aberration since i.
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normal practice at the facility is to wait until after the brief and walkdown to distribute the tests.
(2)_
Simulator _ Scenario Administration Several minor problems occurred during administration'of the
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simulator scenarios.
First, it was noted that there is not-a method by which examiners on the simulator floor can monitor operator phone communications with the' simulator.
operators in the control booth.
This was remedied by supplying the NRC Chief Examiner with a radio that monitored all control booth communications.
See mdly, one of the facility evaluators inadvertently supplied the crew with the critical task list -instead of the crew briefing sheet prior to one of their scenarios.
This invalidated that scenario and required the use of one of the backup scenarios which had not been validated-in advance on the simulator.
The-backup scenario was not validated during prep week, but it was agreed that the tacility exam team would validate it prior to the-exam to ensure it ran smoothly on the simulator'.
Although this was not done, the scenario was administered without any problems.
Thirdly, one of the cvews elected to tri,the reactor earlier than expected during Scenario.?S-1.
This eliminated one of thc crew critical tasks. This scenario set already only had the minimum required number of seven
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CCTs as discussed in Para. 4.c.2, so the elimination of
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another one left only six. The examination team determined that this was an adequate number to evaluate the performance-of the-crew. Maximizing the number of CCTs _in a scenario:
set to allow some leeway to compensate for instances such-as this was discussed with the facility training staff.
Two additional problems were encountered during the. setup.
for scenario #S-3.
The block valve for one of the u
pressurizer PORVs was not closed as _ required for the plant
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conditions. -The crew noticed the omission during their l
turnover, and it was corretted prior to starting the scenario.
During the~same scenario, the open indicating-light for a pressurizer PORV ' extinguished when -a malfunction -
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was inserted to. override the PORV'open.
This improperly alerted-the operators to the PORV malfunction.
This was corrected in subsequent scenarios by simultaneously overriding _the open indication light on, while inserting the PORV malfunction.
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.The examination. team also noted that operators at Sequoyah l-frequently _ mark meter faces temporarily with grease pencils.
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d This is not a commonly. observed practice at other Region 11 plants.
There are not any procedural-guidelines at the facility controlling this practice.
Developing instructions to place some controls on the practiceLand prevent possible operator errors was discussed with the facility operations staff.
(3)
JPM Administration Several problems were encountered with cues given during the administration of the JPMs.
The NRC examiner had to interject more than once to correct an inappropriate cue or to provide a necessary cue omitted by the facility examiner.
The problems'were confined to cues that were needed in response to unexpected actions or questions by the examinee.
These cues were not prescripted in the JPMs.
The facility examiners were instructed to ensure that the examinee receives all of the. feedback he would expect for his actions. They were also_ instructed to only-give.the_-
examinee information he requests when role playing as a-remote Unit Operator. Two cues given as responses to an-examinee's questioning of a Unit Operator addressed what the examinee should have been asking and not what-he actually was asking for.
This type of cue is leading and could improperly aid the examinee. _ All observed deficiencies were corrected as they occurred and were discussed with facility training supervision.
The use of procedure indexes, to allow the examinee to determine the correct procedure to ask for, worked particularly well.
However, procedures provided to the examinees should be marked "For Training Only" to prevent-inadvertent use if misplaced. They should also'be completed up to the point where the examinee is expected to start to enhance examination-realism.
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Facility Evaluators All facility evaluators were found to be satisfactory. One evaluator, in particular, exhibited problems with proper cuing and
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feedback during JPM administration as discussed in_ Para. 4.f.3.
There were no.significant discrepancies between the facility and NRC grading.
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Simulator Fidelity:
Some minor discrepancies between the plant control room and the simulator were-noted during examination administration. The timer
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available in the simulator is not the same as the calibrated timers used in the plant for surveillances.
New timers for'the.
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simulator have been ordered. A _ valve locator list which gives-
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plant coordinates for significant valves is available in the control room, but not in the simulator.
The Shield Building Vent Radiation Monitor, RM-90-400, and its two associated recorders, are not present in the simulator.
The Safety Parameter Display System (SPDS) computer post-accident printouts (P-250 typer) are not active in the simulator.
These discrepancies are already on a list of planned upgrades for the simulator.
The lack of a raised and elevated simulator operator control booth is a shortcoming at Sequoyah.
It is difficult to monitor the scenarios from the booth due to the distance from the control boards and the poor sightlines.
Consequently, observers are essentially forced to monitor the scenarios from the back of the-simulator floor, in full view of the examinees.
There also is not any way for observers to enter and exit the control booth without stepping on the simulator floor and distracting the examinees.
Several examinees commented on the apparent number of observers.
An upgraded simulator operator control booth would eliminate this source of stress on the operators.
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Procedural Problems (1)
During the performance of JPM #42 to transfer the vital inverter power supply, it was noted that the procedure does not address the position, of the vital inverter transTer switch.
A caution placard is installed to ensure it is not in the bypass position but the procedure does not require checking it.
There are not any safety concerns with it being in bypass because it only replaces a circuit diode with an equivalent diode, but configuration control would be enhanced if it was included in the procedure.
(2)
Step 10 in FR-Z.1, High Containment Pressure, directs energizing the hydrogen ignitors.
This step is a deviation from the Westinghouse Owners Group (WOG) guidance because it precedes the determination of the hydrogen concentration in containment. The Sequoyah basis for this deviation is to ensure they are actuated early if needed.
It also states that it would take several hours, in the worst case scenario, before hydrogen could build up to explosive levels.
However, there is the potential for a hydrogen explcsicc. in containment if the concentration is at an explosive level when the ignitors are energized.
There are also situations, such as a secondary break in containment, that result in high containment pressure, but do not produce hydrogen.
" e hydrogen ignitors are not needed in these situations.
The resolution of the validity of this
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deviation from the WOG will be tracked as IFI 50-327, 328/92-302-01, " Basis for placing the hydrogen ignitors in service prior.to analyzing hydrogen concentration in FR-Z.l."
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Material Condition of the Plant Both 125 VDC Battery Board Rooms I and 11 were noted to be in poor condition during plant walkthroughs.
Paint is flaking from.the concrete block walls and is accumulating on the floor, Cabinet panels have numerous paint splatters on them and were noted to have several handwritten notes above instruments-which referred to surveillance numbers. All four Vital Board Inverter-switches do not have permanent pointer markings. An indelible magic marker
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arrow had been drawn on the shaft of each switch. The inspector noted that a switch on Panel 1 in Room I was missing a label plate.
Two valves were found to be missing identification labels during plant walkthroughs. _ Unit two valve, FCV 62-144, in'the boration flowpath downstream of the boric acid-blender, did not have any permanent or temporary labels.
Unit one valve,62-934, in the CVCS primary water to the boric acid blender flowpath (690'
Penetration Room) also did not have any permanent or temporary-labels. Additionally, the words " primary water to-blender" were inappropriately painted in crange on the piping above valve 62-934.
Both valves were identified to the facility representatives when they were noted and the dangers of unofficial markings in the plant were discussed.
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Security and In-Processing Health Physics No problems were encountered with either security or health physics processing.
1.
Inspection-The inspectors reviewed documentation and interviewed licensee personnel to evaluate' how the licensee implemented requalification-training requirements specified 4in--10 CFR 55.59. The licensee's Nuclear Training Program Guide for Licensed: Operator Requalification Program (LORP) provided the administrative controls governing the implementation of.requalification
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requirements.
LORP was designed to assist the licensee.-in meeting the inter,t of several documents-i_ncluding 10 CFR F5.59. A review
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of the program guide indicated-that all areas identified in 10 CFR 55.59 were addressed.
Documentation also indicated that
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operators attended the required training and received remedial training when necessary.
However, the inspector identified two cases where the training department should have recommended v
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remedial training.
The training director will re-evaluate both-cases and submit justification for not remediating those-operators.
This -is -identified as IFI 50-327,328/92-302-02
" Justification for No Remedial. Training on Two Individual Requalification Examination Reports."
The inspectors also reviewed a-random sample of individual records.
This review was performed to evaluate the licensee's process for tracking licensed operator control-evolutions identified in 10 CFR 55.59 (C)(3)(i).
No deficiencies were noted-in the records reviewed.
5.
Exit Interview At the conclusion of the site visit, the examiners met with the representatives of the plant staff iridicated in para 9,aph 1 to discuss the results of the examinations and inspection findings.
Tiw only proprietary material provided by the licensee was personnel tr "ing records that the examiners used during the 10 CFR 55.59 inspection.
The examiners further described the areas inspected. Dissenting comments were not received from the licensee,
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Item Number Status Description / Reference 50-327, 328/92-302-01 OPEN IFI - Basis for placing the hydrogen ignitors in service prior to analyzing hydrogen concentration in FR-Z.1, which is a deviation from WOG guidance.
50-327, 328/92-302-02 OPEN IFI - Justification for No Remedial Training on Two Individual
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ENCLOSURE 2-SIMULATOR FIDEllTY REPORT Facility Licensee:
Sequoyah Nuclear Plant facility Drcket Nos.:
50-327 and 50-328 Operating Tests Administered During:
The Weeks of November-30 and December 7, 1992
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This form is used only
+.o report observations.
These observations do not constitute, in and of themselves, audit or inspection findings and are not, without further verification and. review, indicative of noncompliance with 10 CFR 55.45(b).
These observations do not affect NRC certification o, approval of the simulation facility other than to provide information which may be used in future evaluations.
No licensee action -is required solely in response to these observations.
During the conduct of the simulator portion of the operating tests, the following items were observed:
ITEM DESCRIPTION Shield Building Vent Radiation monitor and its two associated Radiation Monitor, recorders are not present in the simulator.
RM-90-400 The SPDS computer Not active in the simulator.
-
post-accident printouts (P-250 typer)
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[I C J e s-Exam 92PARTARD Data Total number of Validated Question
Total number of Question
No. of short Answer Questions
No. if Multiple Choice Questions
No. of MC Questions with Answer
"A"
No. of MC Questions with Answer
"B"
No. of MC Questions with Answer
"C"
No. of MC Questions with Answer
"D"
Validated Time
Total Points 15.00 p.
O
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,
.
.
ANSWER KEY ANSWER 1 POINTS 1.00 EXAM BANK Qf A.003.03 Drive E Bank
Plant SI termination values (0.125 pts for each correct answer)
AFW flow e 1720 gpm (430 gpm/ loop)
(1 200 gpm total or 50 qpm/ loop)
Subcooling = 100'F (+/- 2 * F)
Pzr Level 52% (+/- 2%)
RCS Press stable or increasing.
SI Terimation Criteria ( 0.125 pts for each correct answer)
RCS Subcooling > 40*F Total AFW flow > 440 gpm or at least 1 S/G > 10%.
RCS pressure stable or increasing.
Pzr level > 10%.
NOTE: DO NOT USE THIS QUESTION WITH A.003.11 ANSWER 2 POINTS 1.00 EXAM DANK Q# A.003.06 Drive E Bank
D.
It can be cleared when turbine load is less than 40%,
as sensed by PT-1-81 or 82, for > 360 seconda.
NOTE: DO NOT USE WITH 3.28 l
ANSWER 3 POINTS 1.00 EXAM BANK Qf A.003.17A Drive E Dank
B.
No, even though PT-68-340 has failed, PcV-68-340 is operable because channels PT-68-334 & 323 have been selected for control.
ANSWER 4 POINTS 1.00 EXAM BANK Qf A.003.22 Drive E Bank
B.
Safety Injection signal from Train B SSPS.
ANSWER 5 POINTS 1.00 EXAM BANK Qf A.003.10 Drive E Bank
A.
Steam dump system operation is correct for present (
plant conditions.
Test No.
92PARTARO ANSWER KEY
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ANSWER KEY ANSWER 6 POINTS 1.00 EXAM DANK Qf A.003.19 Drive E Bank
A.
The seal injection flow path is aligned to flow to the RCP seals.
ANSWER 7 POINTS 1.00 EXAM DANK Q$ A.003.18 Drive E Dank
C.
Depress controller to accident reset and then place controller in manual.
ANSWER 8 POINTS 1.00 EXAH DANK Qf A.003.05 Drive E Dank
D.
Restore the RWST to OPERABLE status in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SIIUTDOWN in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
ANSWER 9 POINTB 1.00 EXAM DANK Of A.003.08 Drive E Dank i
A.
Letdown relief valve RV-62-662 lifts, causing alarm actuation "TS-62-75 Low Press Letdown Relief Temp High" ANSWER 10 POINTB 1.00 EXAM DANK Q# A.003.13 Drive E Bank
C.
Reset the safety injection signal, cycle the Rx Trip Brkrs and reset the feedwater isolation signal.
ANDWER 11 POINTS 1.00 EXAM DANK Of A.003.15 Drive E Dank
C.
Reset the SI signal, place ALL 43T/L switches in the test position, reset the 86 IOR for 1A-A D/G, go to
"stop" with IIS-8 2-14.
ANSWER 12 POINTS 1.00 EXAM DANK Qf A.003.31 Drive E Dank
C.
FCV-74-3 will CLOSE when FCV-63-72 starts OPEN.
l ANSWER 13 POINTB 1.00 EXAM BANK QS-A.003.09 Drive E Dank
A.
The pressure in the letdown header upstream of PCV-62-81 is lower than the controller's setpoint.
l Test No.
92PARTARO ANSWER KEY
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.
ANSWER KEY ANSWER 14 POINT 8 1.00 EXAM DANK QS A.003.01 Drive E Dank
D.
No.
The the CCP suction valves from the VCT, FCV-62-132 and 133, are closed.
ANSWER 15 POINTS 1.00 EXAM DANK Q# A.003.33 Drive E Dank
D.
6.0 minutes.
-3 Answer based on N-35 = 1 X 10
%
-4 N-36 = 5 x 10
%
Difference = 2 Decades 2/0.33 dpm
= 6.06 mins.
SUR(t)
P = P 10
- 33(t)
.
1x10(-5)
(1x10(-3))10
=
.33t 0.1 = 10-2 =
.33t t = 6.06 minutes Test No.
92PARTARO ANSWER KEY
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Sequoyah Nuclear Plant Answer Key for Test 92PARTARO 1.
Short Answer 21, 41, 2.
D 22.
42.
3.
B 23.
43, 4.
B 24.
44.
5.
A 25, 45.
6.
A 26, 46.
7.
C 27, 47.
8.
B 28.
48.
9.
A 29.
49, 10.
C 30.
50, 11.
C 31.
51.
12.
C 32.
52.
13.
A 33.
53.
14.
D 34.
54, 15.
D 35.
55.
16, 36.
56.
17.
37.
57, 18.
38.
58.
19.
39.
59.
20.
40.
60.
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Sequoyah Nuclear Plant Answer Key for Test 92PARTARO 61.
81.
62.
82.
63.
83.
64.
84.
65.
85.
66.
86.
67.
C7.
68.
88.
69.
89.
70.
90.
71.
91.
72.
92.
73.
93.
74.
94.
75.
95.
76.
96.
77.
97.
78.
98.
79.
99.
80.
10,/m\\
(,2 '
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11/30/92 32 of 46
-
QUESTION
POINTS 1.00 Determine the status of the SI termination criteria.
(Note: Be specific to INCLUDE PRESENT PLANT VALUES AND ALSO TO INCLUDE SI TERMINATION CRITERIA)
i I
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92PARTARO
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11/30/92 33 of 46
.
QUESTION
POINTB 1.00
'lindow #7, "AMSAc INITIATED", on XA-55-3C is illuminated. Which ONE of the following identifies when this condition can be cleared by the operator?
A.
It can be cleared when 2/4 steam generators have a Wide Range level > 25%.
B.
It can be cleared when 1/4 steam generators has a Harrow Range level > 8%.
C.
It can be cleared when turbine 1 cad is less than 40%,
as sensed by PT-1-72 and 73, for 22 seconds.
D.
It can be cleared when turbine load is less than 40%,
as sensed by PT-1-81 or 82, for > 360 seconds.
Test No.
92PARTARO
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11/30/92 34 of 4G QUESTION
POINTS 1.00 PT-68-340 is inoperable.
The OATC informs you that action a. of LCO 3.4.3.2 for PCV-68-340 must be performed.
Do you agree?
A.
Yes, PT-68-340 is in the Cold Overpressure control circuit & therefore PcV-68-340 will not operate when COPS is armed.
B.
No, even though PT-68-340 has failed, PCV-68-340 is operable because channels PT-68-334 & 323 have been selected for control.
C.
Yes, PT-68-340 has a direct input to the control circuit, since it has failed low the valve will not open on increasing RCS pressure.
D.
No, PT-68-340 only supplies input to the pressurizer heater and spray valve control circuit (s).
,
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92PARTARO
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11/30/92 35 of 46 Qts'.1 TION
POINTH 1.00 The discharge valvo on the CCS lix 0111 & OB2 (0-rCV-67-152) in indicating 35% open, llowever, the handswitch is in the OPEN position.
What caused the valve to throttle to 35t?
A.
IIS-67-152C was pinced in the 35% OPEN ponition and returned to open position.
11.
Safety injection signal from Train B SSPS.
C.
Manual operation of the valve handwhool.
D.
Low header prosauro on the 'n'
Train EncW nupply hondor.
Test No.
92PARTARO
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11/30/92 36 of 46 QUI:DTION
POINTS 1.00 Which ONE of the following doncriben the operation of the ntonm dump nyntem?
A.
Steam dump system operation in correct for present plant conditions.
B.
Steam dump operation in incorrect. Approximately one-quarter of the steam dumps valven should be open.
c.
Steam dump operation in incorrect. Approxinately half of the steam dumps valven should be open.
D.
Steam dump operation in incorrect. AlI condenner dump valven should be open.
Test No.
92PARTARO
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11/30/92 37 of 46 QUESTION
POINTS 1.00 WIIICll ONE of the following describes why there is still charging flow with FCV-62-90 and 91 Indicating closed?
A.
The seal Injection flow path is aligned to flow to the RCP seals.
B.
FI-62-93A has malfunctioned, thoro is no flow through the charging header.
C.
FCV-62-90 and 91 are "98%" valves, they don't actually go fully closed.
D.
FCV-62-93 has a mechanical stop that ensures a minimum flow to the RCP seals.
Test No.
92PARTARO
_
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _
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11/30/92 38 of 46
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QUEHTION
PO1HTH 1.00 WillCll ONE of the following oporntor actionn munt be performed to tako manuni control of the Motor-Driven Auxillary feedwntor Pump level control valvo, FCV-3-156?
A.
Henet the 31 nignal and placo the controllnr in manun1.
II.
Cycle the Ilonctor Trip lironkorn and place controller in mnnun1.
C.
Deprenn controller to accident rennt and then place controller in manual.
D.
Shutdown the Motor-Driven AFW pump 1A-A nnd then place controller in manual.
,
Test No.
92 PAltTAlto
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11/30/92 39 of 46 l
QUESTION
.
POINTS 1.00
Assuming plant conditions have stabilized and the SI han been terminated, which of the following actions is applicable considering the borated water storage system?
A.
Apply LCO 3.0.3 and initiato action to restore the RWST to OPERADLE status in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SliUTDOWN in
'
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
B.
Restore the RWST to OPERADLE status in I hour or be in COLD SIIUTDOWN in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
C.
Restore the lloric Acid Storage System to OI-ERADLE status in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> borate to a S!!UTDOWN MARGIN of 1% delta K/K at 200'I'.
D.
Restore the floric Acid Storage System to OPERAf)LE status in 7 days or be in COLD SliUTDOWN in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
,
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92PARTARO
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11/30/92 40 of 46
...
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QUP,8 TION
POINTH 1.00 If FCV-62-73 (letdown orifice isolation valvo) is leaking through, which ONE of the following indications will alert the operator /
A.
Lotdown rollef valve RV-62-662 liftn, cauning clarm actuation "TS-62-75 Low Prens Lotdown Relief Temp High" B.
Ronctor Coolant Drain Tank temperature, prensuro &
level indications increase.
c.
VCT level would increase above the program value.
_
D.
Acountic monitor LEDs would indicato roller valvo leakage.
_
Test No.
92PARTARO
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11/30/92 41 of 46 QUESTION
POINTS 1.00 Which ONE of the following describes the actions that would be necessary to start
"B" main feedwater pump?
A.
Roset the safety injection signal, close Rx Trip Drkrs, and open pump suction valve.
B.
Roset containment phase
"A" isolation and cycle the Rx Trip Brkrs.
C.
Roset the safety injection signal, cycle the Rx Trip Drkrs and reset the feedwater isolation signal.
D.
Roset containment phase "A" isolation and reset the feedwater isolation signal.
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Test No.
92PARTARO
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i 11/30/92 42 of 46
_
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POINTS 1.00 Which UNE of the following actionn must be taken to secure Dienel Generator 1A-A?
A.
Roset the SI signal, cycle the Rx trip breakers, place ALL 43T/L nwitches in the test ponition, reset 86 LOR, go to "stop" with IIS-82-14.
D.
Place the 43T/L switch for 1 A-A D/G in the tent position, reset the 86 IDR for 1 A-A D/G, go to "stop" with IIS-82-14.
C.
Roset the SI signal, place ALL 4 3T/L switchen in the tent ponition, reset the 86 IDH for 1A-A D/C, go to
"stop" with IIS-82-14.
D.
Place !!S-82-14 to stop, after engine runs at idle speed for 2 10 minn, depress the emergency stop punhbutton for 1A-A D/G.
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92PARTARO
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11/30/92 43 of 46 QUESTION
POINTS 1.00 if RWST decreanen to < 27% and Cntmt Sump level increased to > 601, which ONE of the following statements CORRECTLY describen valve responne due to interlocks?
A.
FCV-63-73 will OPEN when FCV-74-21 starts CIDSED.
II.
FCV-63-73 will CIASE when FCV-74-3 ntarts OPEN.
C.
FCV-74-3 will CIDSE when FCV-63-72 ntartn OPEN.
D.
FCV-74-3 will CIDSE when FCV-63-73 starts OPEN.
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Test No.
92PARTARO
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11/30/92 44 of 46 QUESTION
POINTS 1.00 Which ONE of the following CORRECTLY identifies the reason for PCV-62-81 being closod?
A.
The pressuro in the lotdown header upstream of PCV-62-81 is lower than the controllor's sotpoint.
B.
The pressure in the lotdown header upstream of PCV-62-81 is higher than the corstro11er's setpoint.
C.
SI signal input to PCV-62-81 logic controls.
D.
Phaso A signal input to PCV-62-81 logic controls.
Test No.
92PARTARO
,-
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11/30/92 45 of 46 QUEDTION
POINTS 1.00 With present plant conditions, can you reduce the reactor coolant system boron concentration using the Boric Acid Blender?
A.
Yes.
All that is required is to place the Makeup MODE selector switch to the " DILUTE" position, and go to start with the makeup control switch.
B.
Yes.
The operator can dilute the RCS by opening FCV-62-143 (primary water to the blender) and FCV-62-128 (volume control tank bypass /VCT inlet).
C.
No.
FCV-62-144 makeup injection valve to the suction of the charging pump gets a Phase A isolation signal, which has to be cleared first.
D.
No.
The the CCP suction valves from the VCT, FCV-62-132 and 133, are closed.
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92PARTARO
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o 11/30/92 46 of 46 QUESTION
POINTS 1.00 Approximately how much time will elapse before the source range
,
Instrumentation will be automatically placed in service?
e A.
1.5 minutes.
B.
3.0 minutes.
C.
4.5 minutes.
D.
6.0 minutes.
Test No.
92PARTARO
.
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Exam 92PARTASRO' Data Total number of Validated Question
,
Total number of Question
No. of short Answer Questions
No. if Multiple Choico Questions
No. of MC Questions with Answer "A"
No. of MC Questions with Answer "B"
No. of MC Questions with Answer "C"
No. of MC Questions with Answer "D"
Validated Time
Total Points 15.00 i
P I'
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Sequoyah Nuclear Plant 8ep gd 92* 36
1-Answer Key for Test 92PARTASRO 1.
C 21.
41, 2.
D 22.
42.
3.
B 23.
43.
4.
A 24.
44, 5.
A 25.
45.
6.
D 26.
46.
7.
Short Answer 27, 47.
8.
A 28.
48.
9.
C 29.
49.
'
,
10.
C 30.
.5 0.
11.
C 31.
51.
12.
D 32.
52.
13.
A 33.
53.
14.
B 34.
54.
15.
D 35.
55.
16.
36.
56.
17.
37.
57.
18.
38.
58.
19.
39.
59.
20.
40.
60.
,
-
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-
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.
.
ANSWER KEY ANSWER 1 POINTS 1.00 EXAM DANK Qf A.003.06 Drive E Dank
C.
It can be cleared when turbine load is less than 40%,
as sensed by PT-1-81 or 82, for > 360 seconds.
NOTE: DO NOT USE WITil 3. 28 ANSWER 2 POINTS 1.00 EXAM DANK Qi A.003.17A Drive E Dank
B.
No, even though PT-68-340 has failed, PCV-68-340 is operabin because channels PT-68-334 & 323 have been select for control.
ANSWER 3 POINTS 1.00 EXAM DANK Qf A.003.22 Drive E Dank i
B.
Safety Injection signal from Train B SSPS.
ANSWER 4 POINTS 1.00 EXAM DANK Qf A.003.10 Drive E Dank
A.
Steam dump system operation is correct for present plant conditions.
ANSWER 5 POINTS 1.00 EXAH DANK Qf A.003.19 Drive E Dank i
A.
The seal injection flow path is aligned to flow to the RCP seals.
ANSWER 6 POINTS 1.00 EXAM DANK Qf A.003.32 Drive E Dank
D.
Restore the pump to OPERABLE status in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in IIOT SIIUTDOWN in the n9xt 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Test No.
92PARTASRO ANSWER KEY
_ - - _ - _ _ _ _ _ - _ _ _ - - - _ _ _ _ _ _ _ _ _ _ - - - - - _ - -
_ _ _ _ -
_ _ - _ _ _ _. -. _ _ _. _ _.. _ _ -. _.. _ _ - - -. _
_ _ _. _ _ _ _ _. _ _ _.. _ _ _. _ _ _
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e ANSWER KEY t
ANSWER 7 POINTS 1.00 EXAM BANK Qf A.003.03 Drive E Bank
Plant SI termination values (0.125 pts for each correct answer)
.
AFW flow a 1720 gpm (430 gpm/ loop)
(i 200 gpm total or 50 gpm/ loop)
Subcooling a 100*F (+/- 2'F)
Pzr Level 52% (+/- 2%)
RCS Press stable or increasing.
SI Terimation Critoria ( 0.125 pts for each correct answer)
RCS Subcooling > 40*F Total AFW flow > 440 gpm or at least 1 S/G > 10%.
RCS pressure stable or increasing.
Pzr level > 10%.
.
NOTE: DO NOT USE TilIS QUESTION WITil A.003.11
ANSWER 8 POINFS 1.00 EXAM BANK-Qf A.003.08 Drive E Bank
A.
Letdown relief valve RV-62-662 lifts, causing alarm-
actuation "TS-62-75 Low Press Letdown Relief-Temp'High" ANSWER 9 POINT 8 1.00 EXAM BANK Qf A.003.13 Drive E Bank
,
'
C.
Roset the safety-injection signal, cycle the Rx. Trip'
_
Brkrs and reset the feedwater isolation signal.
-
ANSWER 10 POINTS 1.00-EXAM BANK Q# A.003.15 Drive _E Bank - 1 C.
Roset the-SI signal, place ALL 43T/L switches in the test position, reset the 86 LOR for_1A-A D/G, go to
"stop" with HS-82-14.
'
ANSWER'11 POINTS 1.00 EXAM DANK Qf A.003.31 Drive E Bank' 1 O.
-FCV-74-3 will CLOSE when FCV-63-72 starts OPEN.
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-Test No.
92PARTASRO-
_
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ANSWER KEY
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e ANSWER MEY ANSWER 12 POINTS 1.00 EXAM DANK Qf A.003.01 Drivo E Dank i
D.
No.
The the CCP nuction valven from the VCT, FCV-62-132 and 133, are closed.
ANDWER 13 POINTS 1.00 EXAM DANK Qf A.003.09 Drive E Bank
A.
The prennuro in the lotdown hondor upationm of PCV-62-81 in lower than the controller'n notpoint.
_
ANDNER 14 POINTS 1.00 EXAH DANK Qt A.003.04 Drive E Dank
11.
Ions of voltage to the power cabinota.
AN8MER 15 POINTS 1.00 EXAH DANT Qi A.003.33 Drive E Dank
D.
6.0 minuten.
-3 Answer bnned on N-35 = 1 X 1G
-4 N-36 = 5 x 10 t
Difference = 2 Decados 6.06 minn.
2/0.33 dpm
=
SUR(t)
P = P 10
.33(t)
1x10(-5)
(1x10(-3))10
=
.33t 0.1 = 10-2 =
.33t t = 6.06 minutos Tont No.
92PARTASRO ANSWER KEY
..
.
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11/30/92 32 of 46-
{
,
QUESTION
POINTS 1.00 Window #7, "AMSAC INITIATED", on XA-55-3C is illuminated. Which ONE of the following identifies when this condition can be cleared by the operator?
-
A.
It can be cleared when 2/4 steam generators have a Wido
!
Range level > 25%.
B.
It can be cleared when 1/4 steam generators has a Narrow Range level > 8%.
C.
It can be cleared when turbino load is less than 40%,
as sensed by PT-1-81 or 82, for > 360 seconds.
,
D.
It can be cleared when turbino load is less than 40%,
as sensed by PT-1-72 and 73, for 22 seconds.
i
?
Test No.
92PARTASRO
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11/30/92 33 of 46 r
QUESTION
POINTS 1.00 PT-68-340 is inoperable.
The OATC Informs you that action a. of LCO 3.4.3.2 for PCV-60-340 must be performed.
Do you agree?
A.
Yes, IT-68-340 in in the Cold overpressure control circuit & therefore PCV-68-340 will not operate when i
COPS is armed.
B.
No, even thc. ugh PT-68-340 has failed, PCV-6H-340 1s
'
operable because channels PT-68-334 & 323 have been selected for control.
C.
Yes, PT-68-340 has a direct input to the control circuit, since it has failed low the valve will not open on increasing RCS pressure.
D.
No, PT-68-340 only supplies input to the pressurizer
,
heater and spray valve control circuit (s).
Test.No.
92PARTASRO
s
-,
.,. _ _.. -.... _. _
.,
.. _.,.,
.. _ _.,
,,
,.,
_
.~
-,.... -.
-
-
_
.. _. -
. _ _ _ - - - - -. - -
....-
__
_-..
.-
O O
-
11/30/92 34 of 46
'
.-
QUESTION
POINTS 1.00 The discharge valve on t.he CCS lix 0111 & 0112 (0-PCV-67-152) in indicating 3M open.
Ilowever, tho handswitch in in the OPEtt ponition.
What enused the valve to throttle to 3517 A.
33o67-152C wan plaeM. in the 35% OPElf position and returned to open psaltion.
11.
Safety Jo,hetfor algnni f rom Tra i n 11 SG PS.
C.
Manual operat}cn of the valve handwhool.
D.
tow header prennuro on the
'll' 9 rain 1:Itew nupply heador.
l Toot No.
92 PAllTASito
,.
__--_ _ __ _ _ ___
.-
--
_
_--__
_ _ _ _ _
- - - _ _
O O
'
11/30/92
35 of 46
!
i, QUESTION
POINTH 1.00 Which ONE of the following describes the operation of the steam dump system?
A.
Steam durnp system operation is correct for present plant conditions.
B.
Steam dump operation is incorrect. Approximately one-quarter of the steam dumps valves should be open.
C.
Steam dump operation is incorrect. Approximately half of the steam dumps valves should be open.
D.
Steam dump operation is incorrect. All condenser dump valves should be open.
Test No.
92PARTASRO
a.
-.... -.
..
-...
. -.. -.. - - - - -.
--
- - - -
-. -.
.:-, -
..---
___ _ _ _ __ _ _
- _ _ - _ - _ - _ _ _ _ _ _ _ _ -
- - _ - _ _ _ - -
--__--
O O
.
11/30/92 36 of 46 QUESTION
POINTH 1.00 WilICll ONE of the following doncriben why t here in utill charging flow with FCV-62-90 and 91 Indicating clonod?
A.
The seal injection flow path in aligned to flow to the RCP neals.
!
D.
FI-62-93A han malfunctionod, there in no flow through the charging heador.
C.
FCV-62-90 and 91 a re " 9 81L " valves, they don't actually go fully closed.
D.
FCV-62-93 han a mechanical stop that ennuren a minimum flow to the RCP nonin, j
Tont No.
92PARTASRO
l l
l
'
, - -.
_
. _ _ _. _ _ _ _.. _ _. _. ~... -. _,. - _ -., _ _ - _ _ -.__ ---.---.. _. - -. - --
.
..
-
... -
. _ _ - -. _. _ - - _ _ - - - _ -
. -.... _..
_ -
-_
-.
O O
,.
11/30/92 37 of 46 QUESTION
POINTH 1.00 During review of 1-SI-SXP-074-128. A for I A-A Rllit pump, it was determined that the pump dischargo pressure had been incorrectly logged.
It was originally logged as 170 psig, but was actually 160 psig.
Which OllE of the following describes the most limiting required Tech Spec actions?
A.
No action required, pump discharge pressure is within limits.
B.
Restore the pump to OPERADLE status in I hour or be in llOT SilOTDOWii in the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
C.
Rostore the pump to OPERABLE status in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in IIOT S!!UTDOWfi in the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SliUTDOWil in the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D.
Rostore the pump to OPERABLE status in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in ilOT SIIUTDOWil in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
,
i l
l Test 110.
92PARTASRO
'37
!
!
-
.
-
.. -
.. -
-
-
_
.
..
..
.
.
.
-
.
-..
-
-
.
......
.-.= _ -
-._.
_. -..- -
-.
-
.. - _ _..
.
..
O O
'
i 11/30/92 38 of 46 QUESTION
POINTS 1,00 Dstormino the status of the SI termination criteria.
(Note Do specific to INCLt!DE PRESENT PIAllT VALUES AllD ALSO TO INCLUDE SI TERMINATIO!! CRITERIA)
.
Test No.
92PARTASRO
... _ __
.-
_
.. _.....
_
_..,,
_.,..
.
.
.. _..
_ _ -. -.. -
. - -. _
. _ _.... _.. _. - _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ -. - _. _ _ _ _ _ _ _. _ _
- _ _. _ _ _ _ _
__.... _ _. -.._ - _
__
-
'
O O
11/30/92 39 of 46 QUESTION
POINTH 1.00 If FCV-62-73 (letdown orifice isolation valve) is Icaking through, which ONE of the following indications will alert the operator?
A.
Letdown roller valve RV-62-662 lifts, causing alarm actuation "TS-62-75 Low Press Letdown Rollef Temp liigh" B.
Reactor Coolant Drain Tank temperature, pressure &
level indications increase.
C.
VCT level would increase above the program value.
D.
Acoustic monitor LEDs would indicate relief valve Icakage.
'
,
,
Test No.
92PARTASRO
. _ _. _ _ _ _. _ _.,. -. _ _, _
.
.
..
.
_
_
_
..
--
-
O
11/30/92 40 of 46 i
--
.. _ -.
QUESTION
POINTil 1.00 Which ON1', of the following doncribon the actions that would bn
.
noconnary to ntart "II" main foodwater pump?
,
A.
Itonet the nnfaty injection nignal, clor.n Itx Trip lirkrn, and open pump nuction valvo.
'
11.
Itonet containmont phnno
"A" Inalat.lon and cycle the itx Trip lirkrn.
l C.
Ilonot the nnfety injection nignal, cycle the Itx Trip Ilrkrn and renet the feedwater inolation nignal.
1).
Itonnt containmnnt phann
"A" isolation and ronnt the foodwntor isolation nignal.
.,
Tent No.
92 PAllTASit0
__
,
_. _ _ _ _, _.
_ - _ _ _ _ _ -.
_
- - _, -.. _,. _... ~. _. _.. - _ _. _ _. _ _ _ _ _ _ _ ~ _ _ _. ~ _.. _ _,.
-
__ _ _ _ _
_ - _ _
_ _ _. -
_
.
. _ _.. _ _ _ _. _ _-.-...
.. - _ _
-_ _ _ _ __
_
__
O O
11/30/92 41 of 46 QUESTION
POINT 8 1.00 Which ONE of the following actions must be taken to secure Diesel Generator IA-A?
A.
Roset the SI signal, cycle the Hx trip breakers, place ALL 43T/L switches in the test position, reset 86 LOR, go to "stop" with 11S-82-14.
D.
Place the 43T/L switch for 1A-A D/G in the test position, reset the 86 LOR for IA-A D/G, go to "stop" with lis-82-14.
C.
Rosot-the SI signal, place ALL 43T/L switches in the test position, reset the 86 IDR for 1A-A D/G, go to
"stop" with 118-82-14.
D.
Place 11S-82-14 to stop, after engine runs at idle speed
.
for 2 10 mins, depress the emergency stop pushbutton for 1A-A D/G.
Test No.
92PARTASRO
-,
__
_
_
.
.. _.,. _. - ~
...
-
.
.
__. -, _ _
..... -.. -.
. _. -. -. -
.
_ -.
_. - -. -. -.
-_. -. -.
. _.
..
-..._-
.
..
~
O O
.
11/30/92 42 of 46
'
.
-l
,
QUESTICH
POINTS 1.00 If RWST decreason to < 27% and Cntmt Sump lovel increaned to > 601, which ONE of the following ntatomonts CollitECTLY describen valve responno due to interlocks?
A.
FCV-63-73 will OPEli when FCV-74-21 starts CIDSED.
D.
FCV-63-73 will CIDSE when FCV-74-3 ntarts OPEN.
C.
FCV-74-3 will CLOSE when FCV-63-72 ntartn OPEN.
D.
FCV-74-3 will CIDSE when FCV-63-73 starts OPEN.
,
Test No.
92PARTASRO
.
O O
11/30/92 43 of 46 QUESTION
POINTB 1.00 with present plant conditions, can you reduce the reactor coolant system boron concentration using the Boric Acid Blender?
A.
Yes.
All that is required is to place the Makeup HODE selector switch to the " DILUTE" position, and go to start with the makeup control nwitch.
B.
Yes.
The operator can dilute the RCS by opening FCV-62-143 (primary water to the blender) and FCV-62-128 (volume control tank bypass /VCT inlet).
C.
No.
FCV-62-144 nakeup injection valve to the suction of the charging pump gets a Phase A isolation signal, which has to be cleated first.
D.
No.
The the CCP suction - lves from the VCT, FCV-62-132 and 133, are closed.
l l
Test No.
92PARTASRO
!
l
.-
_ ~. -
-- - -
-
-
... ~
- _. ~..
.
..
-
.
-
.
O O
11/30/92 44 of 46 QUESTION
POINTS 1.00 Which ONE of the following CORRECTLY identifies the reason for PCV-62-81 being closed?
A.
The pressure in the letdown header upstream of PCV-62-81 is lower than the controller's setpoint.
B.
The pressure in the letdown header upstream of PCV-62-81 is higher than the controller's setpoint.
C.
SI signal input to PCV-62-81 logic controls.
D.
Phase A signal input to PCV-62-81 logic controls.
Test No.
92PARTASRO
...
-
.
--
...
.._
-_
_
_. -
O O
11/30/92 45 of 46 QUESTION
POINTS 1.00 Which ONE of the following conditions has caused.the " ROD CONTROL SYSTEM URGENT FAILURE" alarm to be lit?
A.
Rods on bottom with step counters NOT reset.
B.
Loss of voltage to the power cabinets.
C.
Deviation between RPIs and step counters.
D.
Loss of 100V DC signal to the logic cabinet.
Test No.
92PARTASRO
-.
-
.
o o
-
.
11/30/92 46 of 46
.-
QUESTION
POINTS-1.00 Approximately how much time will elapse before the source range instrumentation will be automatically placed in service?
A.
1.5 minutes.
B.
3.0 minutes.
C.
4.5 minutes.
D.
6.0 minutes.
Test No.
92PARTASRO
-
.
S c p.o.a m O
O
'
.
- m., ; -
p; y,,
y 5 x c..,
-
/t a X -
z.g.., L 9 2 3 s u.
11/30/92 2 of 46 QUESTION
POINTS-1.00 The operators are in the process of transitioning out of E-0 when-the
~STA observes the following information.
plant has experienced a LOCA and ESF actuation
-
AFW flows S/G #1 105 gpm S/G #2 100 gpm S/G #3 100 gpm S/G #4 100 gpm
-
S/G narrow range levels S/G #1 20%
S/G #2 21%
S/G #3 22%
S/G #4 21%
-
Cntmt pressure 5.4 psi
-
No RCPs operating
-
RCS subcooling = 5*F
-
RVLIS lower range 32%
-
IRM SUR + 0.2 dpm 1% power
-
PRMs a
-
Core exit T/Cs avg 713*F Core exit max T/Cs 755'F
-
-
All ECCS components in service Which ONE of the following procedures must be implemented first based
!
on the above conditions?
l A.
FR-S.1 l-
'
B.
FR-C.1 i
C.
FR-H.1 l
D.
FR-Z.1 l
!
Test No.
NRC92RO
1
-
O O
.
l 11/30/92 3 of 46 QUESTION
POINTS 1.00 The operating crew is responding to a SGTR. They have identified and isolated #1 S/G which is ruptured. They are preparing to initiate RCS cooldown to establish subcooling margin. Steam dumps to the condenser are not available due to a failure in the steam dump control circuit.
S/Gs #3 & #4 are both faulted and are also isolated. S/G #2 is neither ruptured nor faulted.
Which ONE of the below responses is the preferred method to cooldown the RCS?
A.
Cooldown the RCS using S/Gs #2, #3, & #4 Power Operated Relief / Atmospheric Operated Relief Valves.
B.
Cooldown the RCS using S/G #2 Power Operated Relief / Atmospheric Operated Relief Valvo.
C.
Cooldown rate can best be controlled by controlling the AFW flow to S/G #2 and the faulted S/Gs (S/Gs #3 & #4).
D.
Cooldown rate will be controlled by controlling AFW flow to S/Gs #3 and/or #4.
-
Test No.
NRC92RO
_ _.
_ _ _ _ _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -.
_.. _ _. _ - _
_ _ _ _ _ _ _ _ -
.
-
-
O O
.
..
11/30/92 4 of 46 QUESTION
POINTS 1.00-The operating crew is performing a rapid RCS cooldown in response to a SGTR. ECCS flow to the core is indicated and RCS pressure drops to
=1150 psig during the cooldown. Which ONE of the following actions should be performed with regard to RCP Trip criteria?
A.
RCPs should be tripped whenever RCS pressure is less than 1250 psig.
B.
RCP trip criterion of 1250 psig during control cooldown/depressurization of RCS does NOT apply.
C.
RCPs should be tripped when RCS pressure is less than 1250 psig and SI/CCP flow is indicated.
D.
RCP trip criterion of 1250 psig does NOT apply after step 5 of E-3.
l l
Test No.
NRC92RO
l
O O
.
,
11/30/92 5 of 46 QUESTION
POINTS 1.00 The plant is initially operating at 100% RTP with all controls in automatic. The following conditions appear almost simultaneously:
SI on low pressurizer pressure S/Gs #1, #3 & #4 pressure stable at = 1005 psig S/G #2 prescure at = 650 psig & decreasing High radiation alarms on Condenser Vacuum pump exhaust, S/G blowdown & Loop #3 MS Line.
Containment pressure = 0.1 psid.
Which ONE of the following is the proper emergency operating procedure flowpath for the crew?
A.
E-0, FR-Z.1, E-3, E-2, E-1.
B.
E-0, FR-Z.1, E-2, E-1, E-3.
C.
E-0, E-2, E-1, E-3.
D.
E-0, E-2, E-3.
I Test No.
NRC92RO
.
_
-
.
-
.
.
O O
,
,
11/30/92 j
.6 of 46 j
l QUESTION
POINTS 1.00 A LOCA has occurred inside containment, and the control room operators are responding to a loss of RHR sump recirculation.
They are unable to restore emergency coolant recirculation, so they add makeup to the RWST, verify that containment spray operating requirements are met and establish only one train of ECCS flow.
The basis for establishing only one train of ECCS flow is because of which ONE of the following reasons?
A.
Delay the time to depletion of the Refueling Water Storage Tank.
B.
Reduce the break flow from the Loss of Coolant Accident.
C.
Allow the RCS to depressure thereby minimizing subcooling.
D.
Limit flow to the core to minimize RCS cooldown.
Test No.
NRC92RO
i O
O
.
11/30/92 7 of 46
_
QUESTION
POINTS-1.00 If an ATWS event were to occur while at 100%, which ONE of the following describes the adverse effect initiating an SI would have on the secondary heat sink?
A.
Isolates S/G Dlowdown.
B.
Isolates Main Feedwater.
C.
Isolates MSIVs.
D.
Trips the turbine.
Test No.
NRC92RO
1
.
- _ - _ -
O O
.
.
11/30/92 8 of 46 I
QUFt8 TION
POINTS 1.00 Four (4) hours after a LOCA has occurred on Unit 1 the following conditions exist:
All ECCS pumps and Cntmt Spray pumps running aligned to
-
RIIR Containment Sump.
Faults in tho 161/500KV switchyards result in a loss of
-
offsite power.
Which ONE of the following operator actions is initially required in response to this condition?
A.
Reinitiate Safety Injection (SI) and return to E-0,
' Reactor Trip or Safety Injection'.
B.
Ensure both trains of shutdown power are available and continue procedure & step.
C.
Ensure Diesel Generators start and ALL-loads sequence on properly.
D.
Pull-to-Lock both Centrifugal Charging Pumps (CCPs).
l I
Test No.
HRC92RO
s
O O
.
.
11/30/92 9 of 46 QUESTION
POINTS 1.00 Step 3 of ECA-2.1, " Uncontrolled Depressurization of All S/Gs" states that AFW should be maintained at minimum flow of 25 gpm to each S/G with a narrow range level Inss than 10%. Which ONE of the following describes the basin for maintaining this minimum flow?
A.
Establishes the required RCS cooldown rate so that COLD SHUTDOWN conditions can be achieved expeditiously.
B.
Provides sufficient steam generation for continued operation of turbine-driven AFW pump.
C.
Prevents S/G dryout and thermal stress that would occur when AFW flow was re-established.
D.
Provides a minimum flow path for the turbine-driven AFW pump to prevent everpressurizing the discharge piping in case the governor malfunctions.
Test No.
NRC92RO
.
. _.
..
. _
_.
.
_
..__.
-
.-
.
-
O O
.
11/30/92 10 of 46 QUESTION
POINTS 1.00 FR-H.1, " Loss of Secondary lieat Sink" has steps to establish " Feed &-
Blood" (Steps 16 - 18). Which ONE of the following statements describes the reason why the feed path is established with a Manual SI as opposed to just starting the CCP & SI pumps?
A.
Ensure ECCS pump valve alignment and starts the D/Gs.
B.
Ensure Emergency Gas Treatment System, ABGTS and CREV start.
C.
Ensure containment isolation and start AFW pumps.
D.
Ensure ECCS pump valve alignment and containment isolation.
.
Test No.
NRC92RO
..
.,.
.
O O
.
11/30/92 11 of 46 QUESTION
POINTS 1.00 The operators are in the process of implementing procedure FR-C.1,
" Response to Inadequate Core Cooling" due to inadequate core cooling with incore T/C > 1200*F. Step 3 of FR-C.1 directs the operator to open the ice condenser AHU ACDs via Appendix D.
The purpose of this action is to ensure which ONE of the following?
A.
The ice condenser will remove sufficient heat from the steam inside containment thereby maintaining containment pressure < 12 psig.
B.
Remove non-safety grade loads from the 480V Reactor Vent Boards preventing a degraded voltage condition.
C.
Prevent hydrogen getting into the ice condenser ductwork and causing the insulation to burn.
D.
Prevent the non-seismic breaker from damaging the Reactor Vent Board.
Test No.
NRC92RO
..
..
11/30/92 12 of 46 QUESTION
POINTS 1.00 A main steam line break occurred on Unit 1.
The operating crew completes all necessary actions. During the event RCS cold leg temperatures dropped from 547*F to 220*F in less than 30 minutes.
Based on the above information, which ONE of the following explains when the operating crew can begin a cooldown to cold shutdown?
A.
Immediately, provided during controlled cooldown the T-cold does NOT decrease more than 100*F in any one (1)
_
hour.
B.
Immediately, provided the cooldown rate is nlow enough such that when target T-cold is reached the average cooldown rate is 5 100*F per hour.
C.
Cooldown can NOT resume until RCS temperatures have been stable for 2 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
D.
Cooldown can NOT resume until an engineering evaluation is performed to determine the ef fects of the cooldown on the reactor precsure vessel.
-
Test No.
NRC92RO
o
_ _. _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _.
O O
.
.
11/30/92 13 of 46 QUESTION
POINTS 1.00 Unit 1 is in Mode 3 with RCS pressure at m1000 psig. The operator is preparing to vent the upper head and opens FSV-68-394, Rx Head Vent isolation valve. When this valve starts to open both.FSV-68-396 &
-397, Rx Head Vent Throttle Valve " pop" (cycle open and close).
Which ONE of the following actions should the operator _ perform after noticing this action.
A.
Immediately close FSV-68-394 since neither FCV-68-396 or -397 shott1d " pop".
B.
Continue with procedure in effect and inform maintenance to repair FCV-68-397 since it should NOT have " popped".
C.
Continue with procedure in effect and inform maintenance to repair FCV-68-396 since it should NOT have " popped".
D.
Continue with procedure in effect and slowly open FSV-68-396.
,
Test No.
NRC92RO
__
_
_
O O
-
11/30/92 14 of 46 QUESTION
POINT 8 1.00 The Unit is operating at 100% RTP when a large-break IDCA occurs. The crew manually trips the reactor and initiates SI. During the performance of E-0,
" Reactor Trip or Safety Injection", a phase _B isolation occurs. The crew removes the RCPs from service and performs the other verifications. The STA informs the Unic SRO that the Core Cooling status tree is ORANGE (due to RVLIS lower range reading = 39%
with 5' F Subcooling and T/C 600 * F). The crew continues performing the immediate actions of E-0.
When the crew reaches the transition step (Step 26 of E-0) to E-1, containment pressure is reading = 5 psid, however the Core Cooling status tree is YELLOW.-(RVLIS lower range is e 55%)
Which ONE of the following is the appropriate procedure to implement at this time?
A.
E-1,
" Loss of Reactor or Secondary Coolant" B.
FR-Z.1,
"High Containment Pressure" C.
FR-C.1,
" Inadequate Core Cooling" D.
FR-C.2,
" Degraded Core Cooling" Test No.
NRC92RO
_
O O
,
.-
11/30/92 15 of 46
~
Qi 'MTION
POINTS 1.00-Operators have just finished completing the first half of their scheduled evening shift when the SOS is advised of a_ Tornado Watch by the Load coordinator.
The OATC directs AUOs to terminate ice blowing which is presently taking place.
Which ONE of the following properly describes why the ice blowing process is stopped?
A.
Air compressors used in the ice blowing process are located outside the Auxiliary Building and must be shutdown and secured because of the approaching Tornado.
B.
AUOs involved in the ice blowing operation are located in the additional equipment building.
Since this building is not designed to withstand tornados, the AUOs must leave this area.
C.
AUOs involved in the ice blowing operation are required to carry out other more important requirements during a Tornado Watch so the ice blowing operation is terminated due to its lesser importance.
D.
The ice blowing operation must be terminated and the ice blowing piping must have flanges installed to meet the requirements of the ABSCE boundary.
Test No.
NRC92RO
. _, -
-
.
-.
O O
-
.
11/30/92 16 of 4 6 QUESTION
POINTS 1.00 Unit 1 is operating at 100% power when the CRO reports an increase to the Unit SRO in the Condenser Vacuum Exhaust Radiation Monitor (RM-90-99) reading. The CRO reports to the SRO that this radiation monitor was reading 100 cpm during the previous shift. The CRO begins morsitoring this radiation and logs the following data for Condenser Vacuum Exhaust Radiation Monitor:
Time Condenser Vacuum Exhaust Exhaust Rad Monitor (RM-90-99)
(cpm)
1000 2040 1015 2060 1030 2050 Calculate the estimated S/G leakage.
Test No.
NRC92RO
..
.
.
--
-
-
O O
-
.
11/30/92 17 of 46-
~
QUESTION
POINTO 1.00 Unit 1 is operating at 100% power when the following indications are received:
CRO reports increasing radiation oh # 1 MS radiation,
-
S/G blowdown & Condenser Vacuum Radiation Monitors OATC reports PZR level can be maintained on progrMa
-
with an additional 24 gpm charging.
Chem Lab / RAD CON report increase activity on S/G #1.
-
The operating crew begins removing Unit 1 from service. After they reach MODE 3, the CRO reports the following:
S/G # 1 Pressure 900 psig.
-
-
S/G #2 Pressure 1000 psig.
-
S/G # 3 Pressure 985 psig.
S/G #4 Pressure 935 psig.
-
Which ONE of the following is the required volume of 20,000 ppm boron required for the cooldown to target temperature?
A.
935 gallons B.
1245 gallons C.
8210 gallons D.
11310 gallons i
l Test No..
NRC92RO
..
.
..
O O
-
.
11/30/92 18 of 46
_
QUESTION
POINTS 1.00 Due to conditions causing Control Room Inaccessibility the control room is abandoned and all checklists are complete.
Hot Standby conditions are being maintained from the Aux. Control Room when 2B-B 6. 9-kV S/D Bd. experiences a loss of voltage.
Which ONE of the following is the expected response by the operating staff for this condition?
A.
Check Diesel Generators running and all auto-ccnnected to the 6.9-kV S/D Boards.
B.
Verify Diesel Generators running and dispatch personnel to manually close the 20-B 6.9-kV S/D Board Emergency Fetder Bkr.
Verify 2B-B D/G connected to 2D-B 6.9-kV S/D Bd.
C.
Verify Diesel Generators running and dispatch personnel to manually close all 6.9-kV S/D Bd. Emergency Feeder Breakers.
Verify all D/Gs connected to the 6.9-kV-S/D Boards.
D.
Verify D/Gs running and 2B-B D/G auto-connected to the 2B-B 6.9-kV F/D Board.
Test No.
NRC92RO
.
.
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.
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.
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O O
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11/30/92 19 of 4 6 QUESTION
POINTS 1.00 Unit one is operating at 28% RTP and 260 MWe generator load when the condenser vacuum starts decreasing. The operators take appropriate actions per plant instructions. Control room instrumentation indicates condenser pressure is 1.8 psia however no annunciators have alarmed due to this condition.
Which ONE of the following is the correct operating crews response?
A.
An automatic turbine trip should have occurred due to low condenser vacuum. Manually trip the reactor, verifying opening of the steam dumps to cool the RCS.
B.
Continue trying to reestablish condenser vacuum per appropriate plant procedures. No additional action required unless condenser vacuum worsens.
C.
Administrative limits for the turbine require the operator to manually initiate a turbine trip, verify operation of the S/G atmospheric relief valves or steam dumps to cool the RCS.
D.
Increase turbine load to increase turbine exhaust into the condenser. The increased steam flow and subsequent condensation will improve condenser vacuum.
Test No.
NRC92RO
-
= -.
...
. _ _.
.. -...
-
O O
-
.
11/30/92 20 of 46 QUESTION
POINTS 1,00 During at-power operation, a control bank rod drops without causing a reactor trip. The procedure for recovering this rod requires that the dropped rod's group step counter reading be recorded prior to recovery of the dropped rod.
Which ONE of the following descibes why this reading must be recorded?
A.
So the operator knows where the dropped rod should be positioned at the end of the recovery.
B.
So that the bank overlap unit can be reset to its proper value after the recovery.
C.
To document that the RILs have not been violated during the recovery.
D.
To reset the P/A converter af ter the dropped rod has been recovered.
Test No.
NRC92RO
,. _ _ _.
O.
O
-
.
11/30/92 21 of 46 QUESTION
POINTS 1.00 A natural circulation cooldown is in progress per ES-0.2,- " Natural Circulation Cooldown." The RCS is at 510*F and 1900 psig. All CRDM cooling fans have tripped and can NOT be restarted.
Without the CRDM fans in operation, which ONE of the following is of the greatest concern?
A.
Damage may occur to the CRDM coils because of overheating.
B.
NDT requirements are more likely to be exceeded for the reactor head flange welds.
C.
Damage may occur to the excore nuclear instrumentation because of overheating.
D.
The formation of a steam bubble in the reactor vessel head region.
Test No.
NRC92HO
-
__.
.
0
-
.
11/30/92 22 of 46 QUESTION
POINTS 1.00 While Unit I was at 100% RTP, a reactor trip and SI occurred. During
'the performance of E-1,
" Loss of Reactor or Secondary Coolant" the STA states that the Safety Parameter Display System (SPDS) shows an ORANGE path on the PTS Status Tree.
Which ONE of the following is the proper course of action for the operators to take?
A.
The operators should immediately transition to FR-P.1,
" Pressurized Thermal Shock".
B.
The operators should verify status tree information via use of qualified instrumentation on the control board and then transition to FR-P.1 if it is the highest priority status tree.
C.
The operators should monitor the Containment &
Inventory status trees to determine if a higher conditions exists and then transition to FR-P.1 if it is the highest priority status tree.
D.
Since this is only an ORANGE path status tree it is not required to be implemented.
Test No.
NRC92RO
.
-
-. -
.
O O
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11/30/92 23 of 46 QUESTION
POINTS 1.00 Unit 1 is in MODE 2 with reactor startup in progress. The following RCS leakages werc determined per 1-SI-OPS-068-137.0, SI-137.3 &
SI-137.5. (Note: Assume leakages other than those given as O gpm)
Total'RCS leakage 4.9 gpm
=
PRT leakage 0.81 gpm
=
CLA #1 leakage 0.19 gpm
=
RCDT leakage 2.7 gpm
=
liUT, CLA #2,#3,#4 0.0 gpm
=
Controlled leakage 38 gpm
=
S/G #1 leakage 0.40 gpm
=
S/G #2,#3,#4 leakage 0.0 gpm
=
No Pressure Boundary Leakage Which ONE of the following describes the reason LCO 3.4.6.2 action b must be entered based upon the above leakages?
A.
UNIDENTIFIED LEAKAGE limit has been exceeded.
i B.
IDENTIFIED LEAKAGE limit has been exceeded.
C.
PRIMARY-TO-SECONDARY LEAKAGE limit has been exceeded.
D.
CONTROLLED LEAKAGE limit has been exceeded.
Test No.
NRC92RO
.
O O
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11/30/92 24 of 46 QUESTION
POINTS 1.00 The following plant conditions exist:
1.
Core alterations are in progress on Unit 1.
2.
Boric Acid Supply line from BAT to Unit 1 blender is tagged for maintenance. The following valves are tagged closed:
1-62-927 1-62-932 1-62-1061 1-62-1055A & B 3.
Shutdown Boron Concentration required to maintain Koff 5 0.95 is 1860 ppm.
4.
Chem Imb notifies you that the Unit 1 RCs Boron Concentration is 190f ppm.
Which ONE of the following actions is required?
A.
Borate the RCS using Boric Acid Transfer pump through FCV-62-138 (Emergency Boration Valve) at 2 10 gpm.
B.
Borate the RCS using the RWST to the Charging Pumps until RCS boron concentration is 2 2000 ppm.
C.
Borate the RCS using Boric Acid Transfer pump through V LV-62 -92 9 (Manual Emergency Boration Valve)
at 2 10 gpm.
D.
Enter Action Statement to Technical Specification 3.0.3.
l i
,
Test No.
NRC92RO
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,
.
-
0
-
.
11/30/92 25 of 46
_
QUESTION
POINTS 1.00 The unit is operating at 85% RTP when a rod in control bank
'C'
falls into the core. Assuming the reactor did not trip, select the ONE response below which states ti.e proper action of the operating crew.
A.
Match T-avg with T-rer and reduce turbine power to
-
75% prior to retrieving dropped rod.
B.
Reduce turbine load to 5 75% and verify adequate shutdown margin prior to retrieving dropped rod.
C.
Insert control bank
'D'
to match T-avg-with T-ref prior to retrieving the dropped rod.
D.
Verify the automatic primary side runback to 75%
turbine load / reactor power prior to retrieving the dropped rod.
.
Test No.
NRC92RO
. _ _ _ _ _. _ _ _ _ _. -. _ _ _ -..
_... _ _ _ _ _ _ _ _. _ _ _. _ _ _ _ _ _.. _ _ _ _ _
_. _ _ _ _ _
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11/30/92 26 of 46 QUESTION
POINTS 1.00 The plant in being returned to full power following a refueling outage when the following indicationn are received in the control room Annunciator RCp Dun Underfrequency/Undervoltage alarms White light annociated with the control switch for RCP #1 in illuminated.
Annunciator TS-68-2A/B Reac Cool Loopn Delta-T Dovn High-Low alarms Ann *tnclator FS-60-6A Ronctor Coolant Loop 1 low Flow alarmn Annunciators Motor Tripout & Motor Ovnrload alarm No other alarms are recolved. The control room operatorn check the reactor and RCPs. Ronctor poWor in 29% and increasingt the other RCPn romain in operation.
Select the ONE responne below which doncribon the proper operating I
crew action for the given conditionn?
A.
Manually trip the reactor.
'
H.
Stop the power increano and inventigate.
C.
Roduco reactor power to loan than 10%.
D.
Manually food the affected S/G.
i
Test No.
NRC92RO
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.
.
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11/30/92 27 of 46 QUESTION
POINTS 1.00 Unit 1 is in MODE 4 (RCS 300'F & 350 psig) with RllR in service and PZR level at 30% (PZR bubble present). An AUO informs the control room that the non-essentini air header in the Auxiliary Building has ruptured after the control room had noticed several alarms.
After about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, the following conditions exist:
PZR loyal 9 %,
-
RilR in service.
-
RCS temp 250*F.
-
RCS press 350 poig.
-
RCP Thermal Barrier flow normal.
-
Which ONE of the following actions must the operator take to prevent overpressurizing the RCS?
A.
Stop CCPs.
B.
Close 1-FCV-62-61 and 63, RCP seal return.
C.
Place 1-ilS-60-340AD & 334D, COPS arming / block switches, in ' BIACK' position.
D.
Open 1-FCV-74-16 and 20, RilR lix outlet FCVs.
Test No.
NRC92RO
I
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ - _ _ - _ _ - _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ - _ _ _ -
. _ _ _ - - _ _ _ _ - _ _ _ _ _ _ _ - _ _
O O
,
11/30/92 28 of 46
-_
-QUESTION
POINTH I.00 IM in in the procoon of calibrating an RCS prennuro trannmitter. Due to a doficiency in the procedure, they need to back out of the tent.
Who must approve backing out of the tont?
A.
IMn Supervisor & SOS /ASOS.
B.
SOS /ASOS & Work Control Shift Supervinor.
c.
Sito Procedure Supervinor & OPS Manager.
D.
Plant Manager & SOS /ASOS.
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Tent No.
IIRC92RO
l
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - - _
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.
11/30/92 29 of 46
_
QUESTION
POINTS 1.00 Unit 1 is at 100% RTP when a small reactor coolant system leak develops. The operators are responding to the event attempting to locate the Icak. The following parameters are observed:
RCS pressure =2205 psig and decreasing All ice condenser doors open Containment pressure 0.5 psig and increasing Pzr level =25% and decreasing Both CCPs running with maximum charging flow Imtdown isolated Which ONE of the following actions should the operating crew perform?
A.
Initiate containment Vent Isolation.
D.
Initiate phase A Isolation.
c.
Decrease load rapidly per AOI-32.
D.
Trip the reactor & initiate SI.
Test No.
NRC92RO
O O
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11/30/92 30 of 46 QUESTION
POINTS 1.00 Unit one has just achieved criticality. Plant conditions are Cycle 6 ilZP 0 Xenon-free & DOL conditions. RCS boron concentration in presently 900 ppm. Control bank
'D'
has 450 pcm worth remaining to the fully withdrawn position.
What power level can the operating crew achievo using only the control
,
rods? (Note: No change in Xenon conditionn)
Test No.
NRC92RO
,
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.
11/30/92 31 of 46 QUE8H ON
POINTS 1.00 The unit is shutdown in mode 5.
RCD temperature is 128'F.
2A-A RilR
,
pump is in service at 2800 gpm. RCS sight glass level indication is i
showing elevation 696'
1".
The plant has been shutdown for = 350 hours0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br />.
'
a.
What is the minimum RCS lovel with RiiR pump (s) running?
(0.5 pts)
b.
Ilow many gallons of water must be removed to reach the minimum RCS level with RilR pump (s) running? (0.5 pts)
,
Test No.
NRC92RO
_.,., _ _ _ _.. _ _ _, _ _.. _
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.,L 9 M e t ANSWER KEY ANSWER 1 POINTS 1.00 EXAM DANK Of D-0005A Drive E Dank
H.
FR-C.1 ANSWER 2 POINTS 1.00 EXAM DANK Qf D-0302 Drive E Dank
B.
Cooldown the RCS using S/G #2 Power Operated Relief / Atmospheric Operated Rollef Valve.
ANSWER 3 POINTS 1.00 EXAM DANK Qf D-0474 Drive E Dank
n.
RCP trip criterion of 1250 psig during control cooldown/depressurization of RCS does NOT apply.
ANSWER 4 POINTS 1.00 EXAM DANK Qf D-0303A Drive E Dank
'
D.
E-0, E-2, E-3.
ANDWER 5 POINTS 1.00 EXAM DANK Qf D-0397 Drive E Dank
A.
Delay the time to depletion of the Refueling Water Storage Tank.
ANSWER 6 POINTS 1.00 EXAM DANK Qi B-0421A Drive E Dank
B.
Isolates Main Foodwater.
ANSWER 7 POINTS 1.00 EXAM DANK Qf D-0425 Drive E Bank
i D.
Pull-to-Lock both Centrifugal charging Pumps (CCPs).
ANSWER 8 POINTS 1.00 EXAM DANK Qf D-0472 Drive E Dank
C.
Provents S/G dryout and thermal stress that would occur when AFW flow was re-established.
ANSWER 9 POINTS 1.00 EXAM DANK Qf D-0485 Drive E Dank
D.
Ensure ECCS pump valve alignment and contninment isolation.
Test No.
NRC92RO ANSWER KEY
-
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ANSWER KEY
'
ANSWER 10 POINTS 1.00 EKAM BANK Qf B-0279B Drive E Bank
C.
Prevent hydrogen getting into the ice condenser ductwork and causing the insulation to burn.
ANSWER 11 POINTS 1.00 EXAM BANK Qf B-0314 Drive E Bank
t C.
Cooldown can NOT resume until RCS temperatures have been stable for 2 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
ANSWER 12 POINTS 1.00 EXAM BANK Qf B-03860 Drive E Bank
D.
Continue with procedure in effect and slowly open
.
FSV-68-396.
AN8MER 13 POINTS 1.00 EXAH BANK Qf B-04 47E Drive E Bank
'
B.
FR-Z.1, "lligh Containment Pressure" ANSWER 14 POINTS 1.00 EXAM BANK Qf B-0448C Drive E Bank
D.
The ice blowing operation must be terminated and the
,
ice blowing piping must have flanges installed to meet the requirements of the ABSCE boundary.
ANSWER 15 POINTS-1.00 EXAM BANK Q# B-0446D Drive E Bank-2 150 gpd (i 10 gpd)
ANSWER 16 POINTS 1.00 EXAM BANK Qf B-0445B Drive E Bank
B.
1245 gallons ANSWER 17-POINTS 1.00 EXAM BANK Qf B-0442B Drive E Bank-2
---
B.
Verify Diesel Generators running and dispatch personnel to manually close the 2D-B 6.9-kV S/D Board Emergency Feeder Bkr.
Verify 2B-B D/G connected to 2D-B 6.9-kV S/D Bd.
Test No.
.NRC92RO ANSWER KEY
.
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.
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.
ANSWER KEY
_
AN8WER 18 POINTS 1.00 EXAM DANK Of D-0085A Drive E Hank
C.
Administrative 11mits for the turbine require the operator to manually initiate a turbino trip, verify operation of the S/G atmonpheric relief valven or steam dumps to cool the RCS.
AH8WER 19 POINT 8 1.00 EXAM DANK Qf D-0171 Drive E Dank
A.
So the operator known where the dropped rod should be ponitioned at the end of the recovery.
ANSWER 20 POINTS 1.00 EXAM DANK Qi D-0003C Drive E Bank
D.
The formation of a steam bubble in the reactor vennel head region.
ANDRER 21 POINTS 1.00 EXAM DANK Qi D-02100 Drive E Dank
B.
The operatorn should verify statun tree information via une of qualified inntrumentation on the control board and then transition to FR-P.1 if it in the highest priority statun tree.
ANSWER 22 POINTS 1.00 EXAM DANK Qf D-0110A Drive E Dank
C.
PRIMARY-TO-SECONDARY LEAKAGE limit has been exceeded.
Identified Leakage = PRT 4 CI A 4 RCDT + S/G Imakagen
=
0.81 + 0.19 4 2.7 + 0.4 = 4.1 gpm Unidentified Leakage a RCS leakage - Identified leakage
= 4.9 - 4.1 - 0.8 gpm Primary-to Secondary Ioakage = 0.4 gpm = 576 gpd.
ANSWER 23 POINTS 1.00 EXAM DANK Qf B-0439 Drive E Dank
B.
Borate the RCS using thn RWST to the Charging Pumpn until RCS boron concentration in 2 2000 ppm.
Test No.
NRC92RO ANSWER KEY
l l
O O
-
ANSWER KEY ANSWER 24 POINTS 1.00 EXAH DANK Qi D-0078A Drive E Dank
A.
Match T-avg with T-ref and reduce turbine power to
75% prior to retrieving dropped rod.
ANSWER 25 POINTS 1.00 EXAH DANK Qi D-0272D Drive E Dank
A.
Manually trip the reactor.
_
ANSWER 26 POINTS 1.00 EXAH DANK Qf D-0430 Drive E Dank
A.
Stop CCPs.
ANSWER 27 POINT 8 1.00 EXAM DANK Qf D-0436D Drive E Dank
A.
IMs Supervisor & SOS /ASOS.
ANSWER 28 POINTS 1.00 EXAM DANK Qi B-0383B Drive E Dank
D.
Trip the reactor & initiate SI.
ANSWER 29 POINTS 1.00 EXAM DANK Qi D-0076A Drive E Dank
25%
(i 3 %)
[This is Cycle 6 dependent - NOT required for ANSWER]
This is Unit 1 CYCLE 6 Dependent.
Test No.
NRC92RO ANSWER KEY
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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ANSWEP KEY ANSWER 30 POINTS 1.00 EXAM DANK Of D-0297AA Drive E Dank
a.
(from 0-PI-OPS-068-673.S
- Minimun level with RHR pump (n) running in) 695'6" (from 0-S0-74-1 the value in 695'6\\" accept either annwor) (0.5 pts)
b.
(add the following 293 + 1763 + 388 =)
,
2444 gallons (1 200 gallons)
(0.5 pts)
(Note: If the student unos a value other than 695'6" for the minimum then bano their answer for part b upon the value that they used with an acceptable range of i 200 gallonn.)
Test No.
NRC92RO AtiSWER KEY
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Sequoyah Nuclear Plant Answer Key for Tent NRC92RO 1.
B 21.
B 41.
2.
B 22.
C 42.
3.
B 23.
B 43.
4.
D 24.
A 44.
5.
A 25.
A 45.
6.
B 26.
A 43.
7.
D 27.
A 47.
8.
C 28.
D 48.
9.
D 29.
Short Answer 49.
t r h 13 10.
C Short Answer 50.
.
(g r ' (, g "
cu (q$'{"
cA 11.
C 31 6) 1 y yy e z e7 yn 51.
12.
D 32.
52.
13.
B 33.
53.
14.
D 34.
54.
15.
Short Answer 35.
55.
ice Spa 110 16.
B 36.
56.
17.
D 37.
57.
18.
C 38.
58.
19.
A 39.
59.
20.
D 40.
60.
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Sequoyah 11uclonr Plant Answer Key for Test 11RC92RO
!
,
61.
81.
62.
82.
63.
83.
64.
84.
65.
85.
66.
86.
67.
87.
68.
88.
69.
89.
70.
90.
,
71.
91.
72.
92.
.
73.
93.
.
74.
94.
75.
95.
76.
96.
-
77.-
97.
' 78.
98.
79.
99.
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80.
100.
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p 11/30/92 2 of 46 QUEDTION
POINTO 1.00 The operators are in the process of transitioning out of E-0 when the STA observes the following information.
Plant han experienced a LOCA and ESP actuation AFW flows S/G #1 105 gpm
-
S/G #2 100 gpm S/G #3 100 gpm S/G #4 100 gpm S/G narrow rango levels
-
S/G #1 20%
S/G #2 21%
S/G #3 22%
S/G #4 21%
Cntmt pressure 5.4 psi
-
No RCPs operating
-
RCS subcooling = 5'F
-
-
RVLIS lower rango 32%
IRM SUR + 0.2 dpm
'
-
PRMs e 1% power
-
Core exit T/Cs avg 713*F
-
Core exit max T/Cs 755'F
-
All ECCS components 1.n servico
-
Which ONE of the following procedures must be implemented first based on the above conditions?
A.
FR-S.1 B.
FR-C.1 C.
FR-il.1 D.
FR-Z.1 Test No.
NRC92SRO
-
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.
.
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.
11/30/92 3 of 46 QUESTION
POINTS 1.00 The operating crew is responding to a SGTR. They have identified and Jsolated #1 S/G which ja ruptured. They are preparing to initlate RCS cooldown to establish subcooling margin. Steam dumps to the condenser are not available due to a failure in the steam dump control circuit.
S/Gs #3 & #4 are both faulted and are also isolated. S/G #2 is neither ruptured ror faulted.
Which ONE of the below responses in the preferred method to cooldown the RCS?
A.
Cooldown the RCS using S/Gs #2, #3, & #4 Power Operated Rollef/ Atmospheric Operated Rollef Valves.
H.
Cooldown the RCS using S/G #2 Power Operated
.
Rollef/ Atmospheric Operated Rollef Valve.
'
C.
Cooldown rate can best be controlled by controlling the AFW flow to S/G #2 and the faulted S/Gs (S/Gs #3 & #4).
D.
Cooldown rate will be controlled by controlling AFW flow to S/Gs #3 and/or #4.
Test No.
NRC92SHO
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.
31/30/92
'
4 of 46
.
QUESTION
POINTS 1.00 The operating crew in performing a rapid Itcs cooldown in ronponse to a SGTH. ECCS flow to the coro in indicated and RCS prennuro dropa to
-
m1150 psig during the cooldown. Which Olli of the following actionn should be performed with regard to itCP Trip criterla?
A.
RCPn should be tripped whenever ItCS prensure in loan than 1250 poig.
HCP trip critorion of 1250 pnly during control cooldown/deprennurization of HCS doen 110T apply.
C.
HCpn should be tripped when RCS prennure in loan than 1250 pulg and C1/CCP flow in Indicated.
D.
HCP trip critorion of 1250 pnly doen llOT apply after step 5 of E-3.
i Tout No.
NHC92SHO
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11/30/92
-
5 of 46
.--
QUESTION
POINTS 1.00 The plant is initially operating at 1001 RTP with all controln in automatic. The following conditions appear almost simultaneously:
SI on low pressurizer pressuro S/Gs #1, #3 & #4 pressure stable at = 1005 psig S/G #2 pressure at = 650 psig & decreasing liigh radiation alarms on Condenser Vacuum pump exhaust, S/G blowdown & Loop #3 MS Line.
Containment pressure = 3.1 paid.
Which ONE of the following is the propor emergency operating procedure flowpath for the crow?
A.
E-0, FR-Z.1, E-3, E-2, E-1.
B.
E-0, FR-Z.1, E-2, E-3.
C.
E-0, E-2, FR-Z.1, E-3.
D.
E-0, E-2, E-3.
Test No.
NRC92SRO
.. _.. _. _ _ _. _ _ _. -
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_ _ -. _..
-
O O
.
11/30/92 6 of 46 QUESTION
POINTS 1.00 If an ATWS event were to occur while at 1001, which ONE of the following describes the adverse effect initiating an SI would have on the secondary heat sink?
A.
Isolates S/G Blowdown.
D.
Isolaten Main Feedwater.
C.
Isolates MSIVs.
,
D.
Trips the turbine.
I l
,
k
,
- I l
Test No.-
NRC92SRO
.a.
-
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-
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, _. _ _ _. _ _ _. _ _ _ _.. -,... _.. _ - - _ _ _.
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_.
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.
.
11/30/92 7 of 46 QUESTION
POINTS 1.00 Four (4) hours after a LOCA has occurred on Unit 1 the following conditions exist i
All ECCS pumps and Cntmt Spray pumps running aligned to j
-
R11R Containment Sump.
Paults in the 161/500KV switchyards result in a loss of
-
offsite power.
Which ONE of the following operator actions is initially required in i
response to this condition?
A.
Reinitiate Safety Injection (SI) and return to E-0,
' Reactor Trip or Safety Injection'.
B.
Ensure both trains of shutdown power are available and continue procedure & step.
C.
Ensure Diesel Generators start and ALL loads sequence j
on properly.
D.
Pull-to-Lock both Centrifugal Charging Pumps (CCPs).
Test No.
NRC92SRO
..
_-__
_ - _ _ _ _ - _ _ _ _ _ _ - _ _ _
O O
.
11/30/92 8 of 46 QUESTION
POINTS 1.00 Step 3 of ECA-2.1, " Uncontrolled Deprennurization of All S/Gn" ntates that AFW nhould be maintained at minimum flow of 25 gpm to each S/G with a narrow range level less than 10%. Which ollE of the following doncribes the banin for maintaining thin minimum flow?
A.
Establishes the required RCS couldown rate no that COLD SilVTDOWil conditionn can be achieved expeditiously.
II.
Providen sufficient steam gennration for continued operation of turbine-driven AFW purnp.
C.
Prevents S/G dryout and thermal ntrenn that would occur when AFW flow was re-established.
D.
Providen a minimum flow path for the turbine-driven AFW pump to provent overprennurizing the discharge piping in cane the governor malfunctions.
Test No.
IIRC92SRO
_ - __ - - __ - - _____ - ___________________________________ ______ _ _ _ _ _ _
_ _ - _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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_ -. _
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. _ _ _ _ _.. -.
. _ _.
._.__._
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.
31/30/92 9 of 46 QUESTION
POINTS 1.00 FR-!!.1, "less of Secondary linat Sink" han steps to establish " Feed &
Bleed" (Stepn 16 - 18). Which ONE of the following ntatements describes the reason why the feed path in entablinhed with a Manual SI as opposed to just starting the CCP & SI pumpn?
A.
Ensure ECCS pump valve alignment and containment inolation.
13.
Ensure containment isolation and start AFW pumpn.
C.
Ennure Emergency Gas Treatment Syntom, AllGTS and CHEV ntart.
D.
Ennure ECCS pump valve alignmont and starts the D/Gn.
Test No.
NRC92SRO
,
_
_,
.
-
_
. _.
_ _ - - - -. _ _
.-
_
11/30/92 10 of 46 QUESTION
POINTH 1.00 The operatorn are in the procenn of implomonting procedure Fil-C.1,
"ltenponne to Inadequate Core Cooling" due to inadequate core cooling with incore T/C > 1200*r. Step 3 of Fit-c.1 directa the operator to open the ice condenner AIIU ACan via Appendix D.
The purpose of this action in to ensure which ONr. of the following?
A.
The ice condenner will remove nufficient heat from the nteam inside containment thereby maintaining containment prennure < 12 pnly.
~
D.
Itemove non-na rcty grade loadn from the 4n0V Ilenctor Vent Boards preventir;J a degraded voltage condition.
C.
Prevent hydrogen getting into the lee condanner ductwork and causing the insulation to burn.
D.
Prevent the non-noinmic breaker from damaging the Iteactor Vent Board.
_
Tent No.
NilC92Sito
_
_ _ ___
_
_ _ _ _ _ - _ - _ - - _ _ _ _ - _ _ _ _ _ _ - - - - _ _ - _ _ _ _ _ - - _ _ - _ - - _ - - - - - - - -
. _ _.-
---
-
- - _ _
.
. _ - _
__
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.
.
11/30/92 11 of 46 QUESTION
POINTS 1.00 Unit 1 is in Mode 3 with RCS pressure at =1000 psig. The oporator is preparing to vont the upper head and opens FSV-68-394, Rx Iload Vent isolation valvo. When this valvo starts to open both FSV-68-396 &
~397, Rx fload Vent Throttle Valve " pop" (cycle open and closo).
Which ONE of the following actions should the oporator perform after noticing this action.
A.
Immediately close FSV-68-394 since neither FCV-68-396 or -397 should " pop".
H.
Continue with proceduro in offect and inform maintenance to repair FCV-68-397 since it should NOT have " popped".
C.
Continue with proceduro in erfact and inform maintenance to repair FCV-68-396 sinco it should NOT have " popped".
D.
Continue with procedure in effect and slowly open FSV-68-396.
Tost No.
NRC92SRO
, _..
_
_
. _ _
_
_ - _
_
__
_ _ _ _ _ _ _ _ _ _ _ _ _ _.
_
O O
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.
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I 11/30/92 12 o f 4 6 QUESTION
POINTH 1.00 The Unit is operating at 100% RTP when a large-break loCA occurs. The crew manually trips the reactor and initiatos SI. During the performance of E-0,
" Reactor Trip or Safety Injection", a phaso B isolation occuro. The crew removes the DCPs from service and performs the other verifications. The STA informs the Unit SRO that the Coro cooling status tree in ORANGE (due to RVLIS lower range reading e 39%
with 5'F Subcooling and T/C 600'P). The crew continues performing the immediato actions of E-0.
When the crew reaches the trannition stop (Step 26 of E-0) to E-1, containment pressure in reading e 5 paid, however the Core Cooling status treo in YELIDW. (RVLIS lower range in
= 5 51, )
Which ofic of the following is the appropriate proceduro to implement at this time?
A.
E-1,
" Loss of Reactor or Secondary Coolant" B.
FR-Z.1, "lligh containment Pressurn" C.
TR-C.1, " Inadequate Core Cooling" D.
FR-C.2, "Dograded Core Cooling" Test No.
liRC92SRO
__ -_ _- _ _____- _----_- -_
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.
.
.
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.
.
,
11/30/92 13 of 46 QUESTION
POINTS 1.00 operators have just finished completing the first half of their
,
scheduled evening shift when the SOS is advised of a Tornado Watch by the Load coordinator.
Tho OATC directs AUon to terminato ico blowing which is prosently taking place.
Which ONE of the following proporly describes why the ice blowing process is stopped?
,
A.
Air compressors used in the len blowing process are located outside the Auxiliary Building and must be shutdown and secured because of the approaching Tornado.
B.
AUOs involved in the leo blowing operation are located in the additional equipment bulIdlng.
Since this building-is not designed to withstand tornados, the AUOs must leave this area.
C.
AUOs involved in the ice blowing operation are required to carry out other more important requirements during a Tornado Watch so the ice blowing operation is terminated due to its lessor-importance.
D.
The ice blowing operation must be terminated and the ice blowing piping must have flanges installed to moot the requirements of the ABSCE boundary.
,
Test No.
NRC92SRO
-
-. -.. _, -. -. -.. _ - -. -.. -
-.
-
-
. -
-,. - -.
.
, -
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.
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.
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.
.
11/30/92 14 of 46
_.
QUESTION
POINTS 1.00 Unit 1 19 operating at 1001 power when the CR0 reports an increase to the Unit SRO in the condonsor Vacuum Exhaust Radiation Monitor (RM-90-99) reading. The CRO reports to the SRo that this radiation monitor was reading 100 cpm during the previous shift. The CRo begins monitoring this radiation and logs the following data for condensor Vacuum Exhaust Radiation Monitort Time Condonsor Vacuum Exhaust Exhaust Rad Monitor (RM-90-99)
(cpm)
1000 2040 1015 2060 1030 2050 calculate the estimated S/G 1eakage.
.
Test No.
NRC92SRO
.
. _,
_
....
.
_
. _.
.
.
....
-.
..
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-
-
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_. - -. _ -. -.. _ - - - -.
_ - -. - - - - _ -. - - - _ - -.
_
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.
.
.
11/30/92 15 of 46
- _ _.
QUESTION
POINT 8 1.00 You are the Unit 1 SRO in charge of refueling. You just noticed a
,
'
significant decrease in level in the reactor cavity area. The control i
room subsequently informs you of a lot of water leaking into lower i
containment. You presently have spent fuel assemblies (SPA) located in the following places:
One spent fuel assembly (SFA) up in the manipulator
-
crano.
One spent fuel assembly (SFA) in fuel transfer cart on
-
reactor building side (Up in upender).
One spent fu21 assembly (SFA) up on spent fuel bridge
-
crane.
Which ONE of the following describes the proper course of actions?
A.
Place SPA from manipulator crane into core, transfer SFA on cart to spent fuel pit side leave in upender in horizontal position, place SPA on spent fuel bridge
'
into Spent Fuel Storage pit (store in any location) and close wafer valve.
B.
Place SFA from manipulator crane into core, transfer SPA on cart to Spent Fuel Storage pit via fuel transfer cart / spent fuel bridge, place SPA on spent fuel bridge into Spent Fuel Storage pit (store in any location) and close wafer valve.
C.
Transfer all three spent fuel assemblies into Spent Fuel Storage pit (store in any location) via fuel transfer cart / spent fuel bridge and close wafer valvo.
D.
Close the wafer valve, place SPA from the manipulator crane back into core, transfer SFA on cart back into core, and place SFA on spent fuel bridge into Spent Fuel Storage pit.
Test No.
NRC92SRO
.
.
-. -..-
-
-
-
-
- - -.
-.
_
_. _ _ _ _ _. _.. _. _ _ _ _ _...
_ _ -... _. _
. ~.
_
. _ _.-. _.
_ _... _
-
O O
.
.
11/30/92 16 of 46 QUESTION
POINTS 1.00 While operating at 100% power, VCT and Pressurizer alarms and indications show decreasing pressurizer level with 2 CCPs operating at maximum injection. In addition the following conditions occur:
- Lower containment radiation monitor (RM-90-106)
has High Radiation alarm.
- Rx trip and SI actuations occur.
- Containment Phase A & D actuations occur.
While following the appropriate Abnorma) and Emergency procedures,
,
you, as the SOS must evaluate the existing conditions.
Which ONE of the following is the emergency classification you would declare based on the above information (Assume all safety systems operate as designed)?
A.
Notification of Unusual Event.
'
B.
Alert.
C.
Site Area Emergency.
D.
General Emergency.
Test No.
NRC92SRO
l
-
-.
__.
.
~.. - -
- _..,. -,..
_
,,_
_
-.,.
.. _.
_
- -.
.-.
-
.-
---
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.
.
11/30/92 17 of 46 QUESTIOff
POINTB'
1.00
'/. ' 2 was operating at a steady stato power level of 50%, with all
';.oititions at equilibrium, when
'D'
11ank Control rods started n ippir.g
,
>"* at the maximum rate. The OATC informs you that no instrument gc.tunctions have occurred and the rod movement can not be stop by r3cing to manual or bank select.
What is the appropriate Radiological Emergency Plan Classification for this event.
Test No.
NRC92SRO
.
..
_ _ _ _ _ _ _ _ _ _ _. _.
l
!
11/30/92 18 o f 4 6 QUESTION
POINTS 1.00 An accident has occurred on Unit 1.
The following plant data is available:
Estimated 600 gpm SGTR On S/G #3
-
-
S/G #3 PORV has failed open & can NOT be isolated
-
Chem Lab analysis indicates RCS activity > 400 pCi/gm Dose Equivalent Iodine 131.
Containment Pressure is 0.5 psig.
-
-
Containment high range radiation monitors indicated
__
m 2 REM /hr
-
Site boundary dose rate is.25 mP/hr Which ONE of the following is the appropriate protective action recommendation?
A.
Shelter 2 mile 2 radius. Shelter actual and projected downwind to 5 miles. (Recommendation 2)
B.
Evacuate to 2 mile radius. Evacuate actual and projected downwind to 5 miles. Shelter other sectors to 5 miles. (Recommendation 3)
C.
Shelter to 5 mile radius. Prepare to evacuate to 2 Je radius and actual and projected downwind sectors 5 miles. (Recommendation 4)
D.
'cacuate actual and projected downwind nectors to 10 i,..les.
Evacuate other sectors to 5 milen. Shelter
-
others to 10 miles. Prepare to evacuate all sectors to 10 miles. (Recommendation 5)
Test No.
NRC92SRO
___
_ _ -_____-_____ - _
- - _ - - - -
.- -
-
.
0_
'
.-
-.
11/30/92 19 of 46 QUESTION
POINTS 1.00 Which ONE of the following statements correctly describes the method (s) and special requirement (s) which should be employed to ensure an air operated valve which falla open can be relied upon for a clearance boundary?
A.
The valve SilOULD be held closed with an installed jacking devico.
In this case, the jacking device becomes part of the clearance boundary and SilALL be appropriately tagged.
13.
The valve SilOULD be held closed with an installed jacking device.
In this case,'the jacking device should be secured with a. locking device. The locking device is NOT part of the clearance boundary.
c.
Air operated valves can NOT be used as part of a clearance boundary.
D.
The valve SilALL have it's air supply valve tagged in the open position and the valve visually verified to be closed.
In this case, the air supply isolation valve becomes part of the clearance boundary.
i
!'
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.
Test No NRC92SRO
.
- _ _ _ _ _ _ _ _ _ - - _ - _ _ _ _ _ _ - _ _ - - _
_
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.
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>
11/30/92 20 of 46 QUE8 TION
POINTS 1.00 A natural circulation cooldown is in progress per ES-0.2, " Natural Circulation Cooldown." The RCS is at 510*F and 1900 psig. All CRDM cooling fans have tripped and can NOT be restarted.
Without the CRDM fans in operation, which ONE of the following is of the greatest concern?
A.
Damage may occur to the CRDM coils because of overheating.
_
B.
NDT requirements are more likely to be exceeded for the reactor head flange welds.
C.
Damage may occur to the excore nuclear instrumentation because of overheating.
D.
The formation of a steam bubble in the reactor vessel head region.
Test No.
NRC92SRO
- _ _ - -
-
.
- _..
_ _.
_
_... -. _ _.
_ _ _ _ _..
_ _ _.
__
O O
.
i 11/30/92 21 of 46
,
QUESTION
POINTS 1.00 Unit 1 is at 100% power operation and Unit 2 is in MODE 6. The ASOS/STA informs the SOS that he needs to go into containment on Unit 2 which requires a double dressout in order to tag ERCW to all lower compartment coolers.
Which ONE of the following describes the directions the SOS should give to the STA?
A.
He can proceed as planned as long as he turns over his STA duties temporarily to ANY ASOS.
B.
He can proceed as planned because Unit 2 is in MODE 6.
C.
He can NOT proceed as planned because a STA is required in the control room at ALL times.
D.
He can NOT proceed as planned because insufficient time is available for him to return to the control room.
l l
l l
l Test No.
NRC92SRO-
l
._
-
. _ _ __._
..
.--
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.
O O
.
..
11/30/92 22 of 46 QUESTION
POINTS 1.00 Unit 1 is in MODE 2 with reactor startup in progress. The following RCS leakages were determined per 1-SI-OPS-068-137.0, SP 137. 3 &
SI-137.5. (Note: Assume. leakages other than those given as O gpm)
Total RCS leakage 4.9 gpm
=
PRT leakage 0.81 gpm
=
CLA #1 leakage 0.19 gpm
=
RCDT leakage 2.7 gpm
=
HUT, CLA #2,#3,#4 0.0 gpm
=
controlled leakage 38 gpm
=
S/G #1 leakage 0.40 gpm
=
S/G #2,#3,#4 leakage 0.0 gpm
=
No Pressure Boundary Leakage Which ONE of the following describes the reason LCO 3.4.6.2 action b must be entered based upon the above leakages?
'
A.
UNIDENTIFIED LEAKAGE limit has been exceeded.
B.
IDENTIFIED LEAKAGE limit has been exceeded.
C.
PRIMARY-TO-SECONDARY LEAKAGE limit has been exceeded.
D.
CONTROLLED LEAKAGE limit has been exceeded.
l l
l l
l l
l Test No.
NRC92SRO
-
m,
, -- ___._
m
--
-
-
.
...
.=.
.
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.
.
11/30/92 23 of 46 QUESTION
POINTS 1.00 The following plant conditions exist:
3.
Core alterations are in progress on Unit 1.
2.
Boric Acid Supply line from BAT to Unit 1 blender is tagged for maintenance. The following valves are tagged closed:
1-62-927 1-62-932 1-62-1061 1-62-1055A & B 3.
Shutdown Boron Concentration required to maintain Keff 5 0.95 is 1860 ppm.
4.
Chem Lab notifies you that the Unit 1 RCS Boron Concentration is 1900 ppm.
Which ONE of the following actions is required?
A.
Borate the RCS using Boric Acid Transfer pump through FCV-62-138 (Emergency Boration Valve) at 2 10 gpm.
B.
Borate the RCS using the RWST to the Charging Pumps until RCS boron concentration is 2 2000 ppm.
C.
Borate the RCS using Boric Acid Transfer pump through VLV-62-929 (Manual Emergency Boration Valve)
at 2 10 gpm.
D.
Enter Action Statement to Technical Specification 3.0.3.
l l
l Test No.
NRC92SRO
s
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---.
--
.
-.
. - _ -.
- - _ _. = - -
.
. -.... -..
- _ _.,
..
-
..
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,
i 11/30/92 24 of 46.
QUESTION'
POINT 8 1.00 The unit is operating at 85% RTP when a rod in control bank
'C'
falls into the core. Assuming the reactor did not trip, select the ONE response below which states the proper action of the operating crew.
A.
Match T-avg with T-ref and reduce turbine power to s
75% prior to retrieving dropped rod.
B.
Reduce turbine load to s 75% and verify adequate shutdown margin prior to retrieving dropped rod.
C.
Insert contro'1 bank
'D'
to match T-avg with T-ref prior to retrieving the dropped rod.
D.
Verify the automatic primary side runback to 75%
turbine load / reactor power prior to retrieving the dropped rod.
.c Test No.
NRC92SRO
O O
.
.
.
11/30/92 25 of 46 QUESTION
POINTS 1.00 The plant is being returned to full power following a refueling outage when the following indications are received in the control room:
Annunciator RCP Bus Underfrequency/Undervoltage alarms White light associated with the control switch for RCP #1 is illuminated.
Annunciator TS-68-2A/B Reac Cool Loops Delta-T Devn High-Low alarms Annunciator FS-68-6A Reactor Coolant Loop 1 Low Flow alarms Annunciators Motor Tripout & Motor Overload alarm No other alarms are received. The control room operators check the reactor and RCPs. Reactor power is 29% and increasing; the other RCPs remain in operation.
Select the ONE response below which describes the proper operating crew action for the given conditions?
A.
Manually trip the reactor.
B.
Stop the power increase and investigate.
C.
Reduce reactor power to less than 10%.
D.
Manually feed the affected S/G.
.
Test No.
NRC92SRO
O O
-
>
.
r 11/30/92 26 of 46
,
QUESTION
POINTS-1.00 Unit 1 is in MODE 4 (RCS 300'F & 350 psig) with RHR in service -and PZR level at 30% (PZR bubble present). An AUO informs the control room that the non-essential air header in the Auxiliary Building has ruptured after the control room had noticed several alarms.
After about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, the following conditions exist:
-
PZR level 95%.
RHR in service.
-
RCS temp 250*F.
_
-
-
RCS press 350 psig.
RCP Thermal Barrier flow normal.
-
Which ONE of the following actions must the operator take to prevent overpressurizing the RCS?
A.
Stop CCPs.
B.
Close 1-FCV-62-61 and 63, RCP seal return.
C.
Place 1-HS-68-340AD & 334D, COPS arming / block switches, in ' BLOCK' position.
D.
Open 1-FCV-74-16 and 28, RHR Hx outlet FCVs.
Test No.
NRC92SRO
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _
..
.
..
. - - _ -. _..
.
O O
,-
.
.
11/30/92 27 of 46 QUESTION
POINTS 1.00 Unit 1 is at 100% RTP. Which ONE of the following is the maximum allowable leakage through any ONE S/G and allow the-Unit to maintain power operation?
A.
< 1 gpm B.
< 128 gpd C.
< 140 gpd D.
< 500 gpd Test No.
NRC92SRO
-
=
.
-
.
o O
....
.
11/30/92 28 of 46 QUESTION
POINTS 1.00 Unit 1 is at 100% RTP when a small reactor coolant system leak develops. The operators are responding to the event at tempting to locate the leak. The following parameters are observed:
RCS pressure 22205 psig and decreasing All ice condenser doors open Containment pressure 0.5 psig and increasing Pzr level m25% and decreasing Both CCPn running with maximum charging flow Letdown isolated Which ONE of the following actions should the operating crew perform?
A.
Initiate Containment Vent Isolation.
B.
Initiate phase A Isolation.
C.
Decrease load rapidly per AOI-32.
D.
Trip the reactor & initiate SI.
!
l Test No.
NRC92SRO
.
..
.-
-
=.
.-..
....
..
.
-_-
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.
..
,
11/30/92 29 of 46 QUESTION
.
POINTS 1.00 Unit-one has just achieved criticality. Plant conditions are Cycle 6 HZP 9 Xenon-free & DOL conditions. RCS boron concentration is presently 900 ppm. Control bank 'D8 has 450 pcm worth remaining to the fully withdrawn position.
What power level can the operating crew achieve using only the control rods? (Note: No change in Xenon conditions)
,
!
!
I
!
Test No.
NRC92SRO
.
...-
.-.
.-
...
.
-
.
. _.,,.-._
.
.
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..
.
.
11/30/92 30 of 46 QUESTION
POINTS 1.00 The unit is shutdown in mode 5.
RCS temperature is 120'F.
2A-A Ri!R pump is in service at 2800 gpm. RCS sight glass level indication is showing elevation 696'
1".
The plant has been shutdown for e 350 hours0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br />.
a.
What is the minimum RCS level with RilR pump (s) running?
[0.5 pts)
b.
How many gallons of water must be removed to reach the minimum RCS level with RHR pump (s) running? (0.5 pts)
l l
Test No.
NRC92SRO
!
-
)
.
... - - - -. _ _. -
....
~
.
11/30/92 31 of 46 QUESTION
POINTS 1.00 Unit 1 is operating at 65% RTP. A RO-trajnee informs you that TWO (2)
of the rods !.n Shutdown Bank B Group 1 individual rod position indicators (RPIs) are indicating 220 steps. Shutdown Bank B demand positions indicate 228 steps and all other Shutdown Bank B roda indicate 228 steps. Reactor Engineering subsequently verifles the rods to be actually at 220 steps via a flux map. Which ONE of the following describes the proper response to this situation?
A.
Restore the affected rods within two (2) hours while maintaining power operation.
B.
Reduce power to less than 50% RTP within eight (8)
hours and realign the rods.
C.
No action required since rod demand position and RPIs are within limits.
D.
Place the unit in llOT STANDBY in a controlled manner within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
.
Test No.
NRC92SRO
.
.
(U (3
'
.
%)
f( q) /e -
d 9/~)t'
$lk N ' '.q S,b,e $
'
ANSWER KEY ANSWER 1 POINT 8 1.00 EXAM BANK Qf B-0005A Drive E Bank
B.
FR-C.1 ANSWER 2 POINTS 1.00 EXAM BANK Qf B-0302 Drive E Bank
B.
Cooldown the RCS using S/G #2 Power Operated Relief / Atmospheric Operated Relief Valve.
ANSWER 3 POINTS 1.00 EXAM BANK Qf B-0474 Drive E Bank
B.
RCP trip criterion of 1250 psig during control cooldown/depressurization of RCS does NOT apply.
ANSWER 4 POINTS 1.00 EXAM BANK Qf B-0303B Drive E Bank
B.
E-0, FR-Z.1, E-2, E-3.
ANSWER 5 POINTS 1.00 EXAM BANK Qf B-0421A Drive E Bank
B.
Isolates Main Feedwater.
ANHWER 6 POINTS 1.00 EXAH BANK Qi B-0425 Drive E Bank
D.
Pull-to-Lock both Centrifugal Charging Pumps (CCPs).
ANSWER 7 POINT 8 1.00 EXAM BANK Q$ B-0472 Drive E Bank
C.
Prevents S/G dryout and thermal stress that would occur when AFW flow was re-established.
ANSWER 8 POINTS 1.00 EXAM BANK Qi B-0485 Drive E Bank
A.
Ensure ECCS pump valve alignment and containment l
isolation.
l ANSWER 9 POINTS 1.00 EKAM BANK Qi B-0279B Drive E Bank
C.
Prevent hydrogen getting into the ice condenser ductwork and causing the insulation to burn.
l Test No.
NRC92SRO ANSWER KEY
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ANSWER KEY i
. 3H8WER 10 POINTS 1.00 EXAM BANK Qi B-0386G Drive E Bank
D.
Continue with procedure in effect and slowly open FSV-68-396.
ANSWER 11 POINTS 1.00 EIAM BANK Qi B-0447E Drive E Bank 2-B.
FR-Z.1, "High Containment Pressure"
,
ANSWER 12 POINTS 1.00 EXAM BANK Q# B-0448C Drive E Bank
D.
The ice blowing operation must be terminated and the ice blowing piping must have flanges installed to meet
the requirements of the ABSCE boundary.
ANSWER 13 POINTS 1.00 EXAM BANK Qf B-0446D Drive E Bank
.
150 gpd (i 10 gpd)
ANSWER 14 POINTS 1.00 EXAM BANK Qf B-0414A Drive E-Bank
A.
Place SFA from manipulator _crano into core, transfer SFA on cart to spent -fuel pit sicle leave in upender in F
horizontal position, place SFA on spent fuel' bridge into Spent Fuel Storage pit-(store in any location) and close wafer-valve.
ANSWER 15 POINTS
~ 1.00_ EXAM BANK Qf B-0440BA' Drive E Bank ~ _2 C.
Site Area-Emergency.
F
- ANSWER 16 POINTS 1.00 EXAM BANK.Qf-B-0045A Drive E Bank 12:
Notification of-Unusual Event
i
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Test No.
NRC92SRO ANSWER. KEY-3-
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ANSWER KEY.
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AN8WER 17 POINTS 1.00 EXAM BANK Qf B-0007B Drive E Bank
C.
Shelter to 5 mile radius. Prepare to evacuate to 2 mile radius and actual and projected downwind sectors to 5 miles. (Recommendation 4)
ANSWER 18 POINTS 1.00 EKAM BANK Qf B-0441H Drive E Bank
A.
The valve SHOULD be held closed with an installed jacking device.
In this case, the jacking device becomes part of the clearance boundary and SHALL be appropriately tagged.
ANSWER 19 POINTS 1.00-EKAM BANK Q# B-0003C Drive E_ Bank
.
D.
The formation of a steam bubble in the reactor vessel head region.
ANSWER 20
' POINTS 1.00 EXAM BANK Qf B-0412 Drive E Bank
'
O.
He can NOT proceed as planned because insufficient time is available for him-to return to the control room.
ANSWER 21 POINTS 1.00 EXAM-BANK Qf B-0110A Drive E Bank
C.
PRIMARY-TO-SECONDARY LEAKAGE limit has been exceeded.
Identified Leakage = PRT + CLA + RCDT + S/G Leakages
-
0.81 + 0.19 + 2.7 + 0.4 = 4.1 gpm
Unidentified Leakage = RCS leakage - Identified leakage
= 4. 9 - 4.1 = 0.8 gpm Primary-to Secondary Leakage = 0.4 gpm = 576 gpd..
' ANSWER 22 POINTS 1.00 EXAM DANK Q# B-0439 Drive E Bank
.2 B.
Borate the RCS using the RWST to the Charging Pumps until RCS boron concentration is 2 2000 ppm.
.
l Test No.
NRC92SRO ANSWER KEY
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.-
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ANSWER KEY ANSWER 23 POINTS 1.00 EKAM DANK Qf D-0078A Drive E Bank
A.
Match T-avg with T-ref and reduce turbine power to
75% prior to retrieving dropped rod.
ANBWER 24 POINTS 1.00 EXAM BANK Qf B-0272D Drive E Dank
A.
Manually trip the reactor.
ANSWER 25 POINTS 1.00 EKAM DANK Q# D-0430 Drive E Bank
A.
Stop CCPs.
ANSWER 26 POINTS 1.00 EXAM BANK Qf B-0443B Drive E Bank
B.
< 128 gpd ANSWER 27 POINTS 1.00 EXAM BANK Qf D-0383B Drive E Bank
D.
Trip the reactor & initiate SI.
ANSWER 28 POINTS 1.00 EXAM BANK Qf B-0076A Drive E Bank
25%
(i 3 %)
[This is Cycle 6 dependent - NOT required for ANSWER)
This is Unit 1 CYCLE 6 Dependent.
l f
Test No.
NRC92SRO ANSWER KEY
_ _ _ _ _ _ _ _ _ _ _ _ _ _.
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ANSWER KEY ANSWER 29 POINTS 1.00 EXAM BANK Qf B-0297AA Drive E Bank
a.
(from 0-PI-OPS-06R-673.S
- Minimun level with RilR pump (s) running is) 695'6" (from 0-SO-74-1 the value is 695'6\\" accept either answer) [0.5 pts)
b.
(add the following 293 + 1763 + 388 =)
2444 gallons (
200 gallons]
[0.5 pts)
(Note: If the student uses a value other than 695'6" for the minimum then base their answer for part b upon the value that they used with an acceptable range of i 200 gallons.)
ANSWER 30 POINTS 1.00 EXAM BANK Qf B-0407CA Drive E Bank
D.
Place the unit in llOT STANDBY in a controlled manner within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
_
Test No.
NRC92SRO ANSWER KEY
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Sequoyah Nuclear Plant Answer Key for Test
'
NRC92SRO 1.
B 21.
C 41.
2.
B 22.
B 42.
3.
B 23.
A 43.
4.
B 24.
A 44.
5.
B 25.
A 45.
6.
D 26.
B 46.
!
7.
C 27.
D 47.
8.
A 28.
Short Answer 48.
L f 9e 13 9.
C 29.
Short Answe r'l " cvt (9 5 '( C
49.
M (? 5 '
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.10.
D 30.
D f) L y yy 2 2 C1 $4L 50.
,
11.
B 31.
51.
12.
D 32, 52.
13.
Short Answer 33.
53.
/SC t/0 7AO 14.
A 34.
54.
15.
C 35, 55.
16.
Short Answer 36.
56.
NUK 17.
C 37.
57.
18.
A 38.
58.
19.
D 39.
59, 20.
D 40.
60.
,
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Sequoyah Nuclear Platit Answer Key for Test NRC92SRO 61.
81.
62.
82.
63.
83.
64.
84.
65.
85.
66.
86.
67.
87.
68.
88.
69.
89.
'
70.
90.
71.
91.
,
72.
92.
73.
93.
74.
94.
75.
95.
76.
96.
77.
97.
78.
98.
79.
99.
-
80.
100.
- - -
_ _ _
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