IR 05000333/1989004
| ML20246F947 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/07/1989 |
| From: | Conicella N, Conte R, Pullani S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20246F937 | List: |
| References | |
| 50-333-89-04OL, 50-333-89-4OL, IEB-88-007, IEB-88-7, NUDOCS 8907140102 | |
| Download: ML20246F947 (85) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
OPERATOR LICENSING PROGRAM EVALUATION AND
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BWR POWER OSCILLATION PROGRAM INSPECTION Combined Report No.: 50-333/89-04 (OL)
Facility Docket No.: 50-333 Facility License No.: DPR-59 Licensee: Power Authority of the State of New York P. O. Box 41 Lycoming, New York 13093 racility: James A. FitzPatrick Nuclear Power Plant Examination Dates: May 1 - 4, 1989 Inspection Dates: April 17 - May 4, 1989 Examiners: Nicola F. Conicella, Operations Engineer Marion I. Daniels, Sonalysts, Inc.
Richard Miller, Sonalysts, Inc.
Inspector: Y2 I Nicola F. Conicella, Operations Engineer 7/r[F Date Chief Examiner: DubnI 7 ~[97 Sida V. Pul/la p iior Operations Engineer /
/ Date /
Approved By: (> . bh
) chard J. Conte, Chilif'~ ~
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h7 Date ML.Section, Operations Branch, DRS
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8907140102 890707 PDR ADOCK 05000333 g PDC
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Executive Summary Written and operating examinations were administered to six Reactor Operators (R0s) and six Senior Reactor Operators (SR0s). The examinations were graded concurrent 1y'by the NRC and the facility training staff. As graded by the facility, two of the SR0s failed the written portion of the examination. How-ever, as graded by the NRC, the same two SR0s passed the written portion of the examination. As gre.ded by the facility and NRC, all others passed all portions of the examination.s. All three crews were rated as. performing satisfactorily
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during the simulator portion of.the examination. The general observations i
I' noted herein are reflective of expected results with respect to this new NRC staff initiative. Overall, the licensed operator requalification training program was rated as satisfactory.
The inspection of the licensee's BWR Power Oscillation Program is documented in Section 4 of this report. No violations were identified. The licensee in particular implemented the requested actions of NRC Bulletin 88-07 and Supple-ment 1 in a somewhat cursory fashion. Although the licensee complied with all the requirements of the Bulletin and Supplement, they did not verify that the-intent of the Bulletin and Supplement was adequately implemented. The result of these inadequacies were: a training lesson plan that was inaccurate, several procedures that did not contain appropriate cautions, and a' weak know-ledge level of certain operators interviewed with regard to recent procedure revisions dealing with the power oscillation topic. There appeared to be two causes for these inadequacies: (1) the licensee's review of procedures affected by NRC Bulletin 88-07 and Supplement I was not comprehensive enough, and (2) the licensee's verification process for ensuring that licensed opera-tors understood the procedure changes that'were made was insufficient. How-ever, the licensee was receptive to the concerns of the inspector and was able
.to rectify all problems or deficiencies noted in a timely fashion.
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REPORT DETAILS :
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1. Introduction During the examination period, NRC examiners conducted a requalification examination.of 12 licensed operators (six Reactor-Operators (R0s) and six Senior Reactor Operators (SR0s)) at James A. FitzPatrick Nuclear Power -
Plant. The examiners used the process and criteria in NUREG 1021, Opera-bor Licensing Examiner Standards," specifically, ES 601, " Administration of NRC Requalification Program Evaluation,"' Revision 5,- January 1,1989.
An entrance meeting was held with the licensee on February 2,-1989 in the Regional Office. .The purpose of this meeting was to brief the licensee on the requirements of the new requalfication program evaluation and to outline a prospective schedule for the week of the examination.
The licensee's personnel contacted during the examination are listed in Attachment 1. The personnel that. attended the exit meeting on May 4, 1989,
the members of combined NRC - facil'ty examination team, and the facility eveluators are identified in Attachment 1.
2. Individual Examination Results
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The following is a summary of the individual examination results:
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NRC l RO' l SR0 l TOTAL l Grading l Pass / Fail l Pass / Fail l Pa>s/ Fail [
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6/0 l 6/0 l 12 / 0 l
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l l l Simulator 6/0 1 6/0 12 / 0 1 l l l l Walk-Through l 6/0 l 6/0 l. 12 / 0 ,
.I I l l l l Overall l 6/0 l 6/0 l 12 / 0 l 1 I l l l Facility l R0 l SR0 l TOTAL l Grading l Pass / Fail l Pass / Fail l Pass / Fail l 1 1 I 1 l Written l 6/0 l 4/2 l 10 / 2 l l l l I Simulator l 6/0 l. 6/0 l 12 / 0 l l l I l l l Walk-Through j 6/0 l 6/0 l 12 / 0 l l l l l l l Overall l 6/0 l 4/2 l 10 / 2 l 1 l l l l
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3. Program' Evaluation Results I 3.1 -Overall Rating: Satisfactory
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The facility program for licensed. operator requalification training I was rated as satisfactory in accordance with the criteria established '
in ES-601, Paragraph C 3.b.(1) and (2). The facility met the criteria ~
in Paragraph C.3.b.(1) as described below:
a. The pass / fail decision agreement between NRC and facility grading of_the written and operating examinations shall be at least' 90L There was_100% agreement on the criterion on the simulator and the walk-through portions ud 83.33% on the written portion of the examination. Although the 90% criterion is not literally met, the disagreement is due to the more stringent grading.of the written portion by the facility. The facility's grading resulted in the marginal failures of two SR0s, where as, NRC's grading passed these two SR0s. The facility program is not penalized for being more stringent and therefore the criterion is considered met.
b. At least 75% of all operators pass the_ examination not including those individuals selected who had passed an examination previ-ously but were selected for the current examination. Grading by NRC is the only consideration for this criterion.
There were no individuals who had previously passed the examina-tion. All 12 operators passed the examination overall. There-fore, this criterion is met.
c. Simulator evaluation guidance (ES 601 paragraph D.I.c(2)(c)):
During the simulator portion of the operating examination, if NRC fails one or more crews but the facility does not, the program may be considered unsatisfactory; whereas if NRC fails
.more than one-third of the crews, the program shall be consi-dered unsatisfactory.
Three crews were ev61uated and all crews passed the simulator
portion of the operating examination, as graded by NRC as well as the facility. Therefore, this criterion is met.
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d. The program meets the requirements of 10 CFR 55.59 (c)(2), (3),
and (4) or is based on systems approach to training.
As reported by the licensee, the licensee's program met the 10 CFR 55.59 criteria.
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l The facility met the criteria in paragraph C.3.(b).(2) ts described below:
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.If three or more of the following are applicable to a requalification program, then that program shall be deter-mined to be unsatisfactory. If one or two of the follow-
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ing are applicable, the program may be determined unsatis-factory.
a. The same common JPM is missed by at least 50% of the examinees.
The maximum percentage of examinees missing any of the 10 common JPMs was 8.33%. Therefore, this criterion is met.
b. The same question about the same' common JPM is missed by at least 50% of the examinees.
The maximum percentage of candidates missing any of the 20 questions of the 10 common JPMs were 33.33%.
Therefore, this criterion is met.
c. The facility failed to train and evaluate operators in all the positions permitted by their individual licenses. For instance, the facility is required to train and evaluate SR0s a) in the RO position, b) directing operators.
The facility had trained and evaluated the operators as reoutred. Therefore, the facility training and evaluadon program met this criterion.
d. Failure to train operators for "in plant JPMs."
The facility had trained operators for in plant JPM.
Therefore, the facility training program met this criterion.
e. Less than 75% of the examinees correctly answer 80%
of the common JPM questions.
100% of the examinees correctly answered 85% or more of the 20 common JPM questions. Therefore, this criterion is met, f. More than one facility evaluator is determined to be unsatisfactory in accordance with " Evaluation of Facility Evaluators" (ES-601).
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None of the four evaluators were found unsatir?actory.
Therefore, this criterion is met.
In summary, the facility program met all the. requirements in ES-601 and is, therefore rated as satisfactory.
3.2 Programmatic Strengths and Weaknesses
- i a. Strengths i
- Ability of the facility instructors to.ideatify crew and indivi-
' dual strengths or weaknesses demonstrated during the scenarios and identifying them during post-scenario critiques with the crew and individuals
- Dedication of the facility instructors to accomplishing program-matic tasks
- Good' progress _ in incorporating simulator training into the Requalification program accomplished within a relatively short time since the simul.ator facility became operational
, b. Weaknesses
- No major programmatic weaknesses were noted.
l e 3.3 General Observations
, a. Simulator Portion of Examination
- Communications and teamwork for two crews were very good. The third crew demonstrated below average performance in this area.
l However, its performance did not result in a safety action being i
performed improperly. Nevertheless, the facility plans to complete a remedial training in communication for this crew, on
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a continuing basis, commencing immediately.
= All crews demonstrated proficiency in use and knowledge of Technical Specifications, Emergency Plant and Emergency Operating Procedures.
= There were several instances where all immediate scram actions and complete verification of isolations and Emergency Core Cooling System (ECCS) actuations were not performed. This performance may be attributable to the fact that one operator is not dedicated to performing these tasks in their entirety.
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There are numerous operator cids (for example, ECCS isola-tion verification checklist, Local Power Range Monitor /
Average Power Range Monitor LPRM/APRM grouping, Hydraulic Control Unit (HCU) location grid) located throughout the simulator, but the aids do not seem to be used or under-stood by the operators. The facility indicated the aids
are generally new and the operators are getting used to
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them.
The simulator scenarios had several formatting deficiencies that require attention in the future:
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The scenarios as written have inconsistencies in the number vi steps written for specific topics. For example, the scram actions of A0P-1 are all delineated but other !
items are written simply as " Enter E0P-4." ;
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Each individual page did not contain the scenario title or page number. This format could result in confusion while administering the examination. j
The simulator fidelity was adequete overall. One notable simulator malfunction occurred during the course of one scenario (MSIVs closed during EHC pressure regulator mal-function). However, it did not affect the continuation of !
the scenario or the ability to adequately evaluate the crew or the candidates adequately. The problems noted during !
the examination have been identified to the facility. The j facility committed to resolve these problems. -
During the course of one scenario, the examination team had i a concern over whether the drywell spray initiation limit had been violated. The facility was able to adequately explain the basis for that particular E0P (EOP-4, Contain-ment Control) and address other technical concerns the examination team had. The explanation given to the exami-nation team should be considered for training to all opera-tors to ensure that they fully understand that particular basis in the E0P. The facility committed to complete the required training for operators by August 1, 1989.
b. Walk-Through Portion of Operating Examination
The facility evaluators generally did a good job of evalu-ating the candidates' ability to perform specified tasks and their knowledge level of these tasks. However, initi-ally the evaluators were reluctant to reword questions or provide clarifying information as needed to the candidates when it was apparent that the wording of the question was confusir,g the candidates. This situation improved after i
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the NRC examination team explained that not only.it was acceptable to clarify questions as needed but also it was expected of the facility evaluators.
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The confusion displayed by the candidates during several JPM questions indicates that question validation ~was not'as thoroughly performed as it should have been.
Although all candidates. performed the JPMs within the allowed " contact time," the actual. time required to conduct the entire JPM examination (including the transit and waiting time) was excessive. The' facility needs to re-assess this situation and correct the problems in the-future examinations.
The candidates seemed confused over several JPMs since these JPMs seemed too simplistic. It appears that the candidates are not properly briefed to perform only the task that was asked for. This comment goes along with the evaluators' ability and responsibility to' clarify questions as needed.
During the performance of two simulator JPMc, certain problems were noted with the simulator fidelity:
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The LPRM bypass switch znust be slowly moved to the bypass position, to get the expected simulator response.
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While synchronizing the generator to the grid, the generator tripped on several occasions for no apparent reason.
The simulator instructor is aware of these problems. They
.did not affect the ability to evaluate the candidates adequately.
- l The candidates were proficient in the knowledge and the use of procedures. No apparent weaknesses were noted.
c. Written Examination
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The facility submitted several examination questions which were not supported by any learning objectives. All exami-nation questions are expected to be supported by learning objectives. The facility representatives committed that the learning objectives will be developed before the next requalification examination.
The simulator fidelity was acceptable for the conduct of the written static simulator examination.
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Inspection of Implementation of NRC Bulletin 88-07 and Supplement 1, BWR Power Oscillations 4.1 Introduction-An inspection was conducted at the James A. FitzPatrick Nuclear Power Plant on April 17-20, 1989. The inspector evaluated the licensee's response to and implementation of NRC Bulletin (NRCB) 88-07 and'
Supplement 1 to this bulletin. The bulletin addressed power oscilla-tions in boiling water reactors (BWRs). The licensee's responses to the bulletin and the supplement are contained.in New York Power Authority Letters JAFP-88-109, dated December 15, 1988, and JAFP-89-0141 dated February 21, 1989. Temporary Instruction 2515/99,
" Inspection of Licensee's Implementation of Requested Actions of Bu11etin 88-07, Power Oscillations," was used to conduct this review.
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The following persons provided substantial information during this inspection:
W. Fernandez, Resident Manager W. Locy, Operations. Superintendent D. Johnson, Assistant Operations Superintendent D. Burch, Reactor Analyst Supervisor D. Simpson, Superintendent of Training G. Fronk, Requalificaiton Training Instructor 4.2 Inspection Details The inspector reviewed lesson plans for operator training and deter-mined that the training material properly addressed the power oscilla-tion issue as requested by NRCB 88-07. Training was conducted for.
all licensed operators (including shift technical advisors) in a timely fashion. However, NRCB 88-07 Supplement 1 requested procedure changes and operating philosophy changes which differed from the information contained in the original operating training lesson plan.
There was no formal training conducted as a followup to NRCB 88-07 Supplement 1. The operators were made aware of the procedure changes through Night Order Book entries by operations management. The licensee committed to revise the power oscillation lesson plan to be consistent with procedure changes resulting from the implementation of NRCB 88-07 Supplement I and to incorporate the power oscillation topic into the ongoing requalification program by February 1990 (Cycle 2 of the 1990 requalification program). This item will remain unresolved pending completion of the licensee's corrective actions (50-333/89-04-01).
The inspector interviewed seven licensed operators (two of whom were also shift technical advisors) to determine knowledge of power oscil-lations and the actions required to mitigate a power oscillation
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i transient. All operators interviewed were knowledgeable of the '
methods to perform normal power changes to avoid regions of the power-to-flow map where the possibility exists for power oscillations to occur. However, the operaters displayed a lack of familiarity-with the indications of power oscillations, restart requirements for recirculation pumps while in areas of potential core power instabil-ity, and actions required to mitigate core power oscillations as directed by procedures AOP-8 (Loss of Recirculation Coolant Flow) and RAP-7.3.16 (Plant Power Changes). The licensee committed to conduct onshift training for all licensed operators (including shift techni-cal advisors) regarding the changes made to these procedures in response to NRCB 88-07 Supplement 1. As of May 18, 1989, all the onshift training required was completed. The inspector had no further questions in the operator training area.
The inspector reviewed applicable procedures to verify that they provided adequate symptoms of power oscillations, cautions to avoid potentially unstable operating situations, and actions to terminate power oscillations if they do occur. The licensee had previously j performed a procedural review and determined that two procedures j required revisions to address the power oscillation issue. P r' ace-
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dures A0P-8 (Loss of Reactor Coolant Flow) and RA/-7.3.16 (Plant Power Changes) were revised in response to NRCB 88-07 Supplement 1.
The licensee had determined that no new procedures were required.
The inspector determined that the changes made to procedures A0P-8 and RAP-7.3.16 were adequate. However, the inspector noted certain deficiencies with the licensee's procedures review process as described below:
Procedures AOP-3 (High Activity in . Coolant or Offgas),' A0P-16 (Loss of 10300 Bus), A0P-17 (Loss of 10400 Bus), A0P-20 (Loss of 10700 Bus), A0P-31 (Loss of Condenser Vacuum) and AOP-48 (Loss of Main Generator H 2
) require power reductions, but these procedures have not been revised to include cautions regarding core power instabilities.
Procedure F-OP-27 (Recirculation System) contains a note regard-ing the return from no recirculation pumps in service to one recirculation pump in service while the reactor is at power.
i The immediate operator action is to scram the reactor if both
) recirculation pumps trip while the reactor it at power in the RUN mode. Therefore, this note is no longer valid and should be removed from the procedure.
The Licensee representatives committed to conduct an in-depth review of all procedures to determine which procedures require power changes that could place the reactor in a potentially unstable operating region of the power-to-flow map. The licensee has committed to make changes as appropriate to the procedures identified by their in-depth
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review. .As of May 18, 1989, the licensee's in-depth procedure review has been completed. The licensee identified 20 procedures that required revisions. Nine of these procedures have been reviewed and-were approved on May 3, 1989 during PORC meeting number 89-023. The remaining.11 procedures have been revised and are awaiting ,PORC'
approval scheduled for May .24, 1989. All of the deficiencies noted-by the inspector have been adequately addressed and. corrected. The inspector had no further questions in the procedural area.
There appeared to be two causes for these. inadequacies: (1) the licensee's review of-procedures affected by NRC Bulletin 88-07, and Supplement I was not comprehensive enough, and (2) the licensee's.
- verification process for ensuring that . licensed operators understood the procedure changes (especially as a result of Supplement 1) was insufficient. Overall, it appeared to the inspector that the licen-see implemented the. requested actions of the NRCB in a cursory manner.
5. ~ Exit Meetina On May 4, 1989, the chief examiner conducted an exit interview at the conclusion of the examination. Those facility personnel in attendance are noted in Attachment 1 of this report.
The Chief Examiner thanked the licensee's training and. operations staff..
for their participation in the examination. Examination development-and examination conduct, general observations noted during the examination-process, and programmatic strengths and weaknesses were discussed.
During this exit meeting, the findings of the inspection conducted during the-weeks of April 17.and May 1,.1989, on the BWR Power Oscillation issue were summarized by the inspector.
Attachments:
1. Persons Contacted
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2. Simulator Examinations (cover sheets only)
3. Job Performance Measures (cover sheets only)
4. Written Examination and Answer Key Part A (R0 and SRO)
5. Written Examination and Answer Key Part B (R0 and SRO)
6. Licensee Letter, dated May 15, 1989 from W. Fernandez, NYPA, to R. Gallo, USNRC, Enclosing a Requalification Program Self Evaluation
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ATTACHMENT 1 Persons Contacted (1)
During Requalification Program Evaluation 1. New York Power Authority (NYPA)
F. Catella, Manager of Nuclear Training (3)
W. Daczkowski, Nuclear Training Specialist (3) .i W. Fernandez, Resident. Manager G. Fronk, Nuclear Training Specialist (2)
D. Johnson, Assistant Operations Superintendent (2)
D. Lindsey, Planning Superintendent R. Locy, Operations Superintendent J.. Morris, Nuclear Training Specialist (3)
J. Romanowski, Simulator Manager D. Simpson, Training Superintendent R. Walker, Nuclear Training Specialist (3)
2. Nuclear Regulatory Commission (NRC)
N. Conicella, Operations Engineer (2)
R. Plasse, Resident Inspector S. Pullani, Senior Operations Engineer (2)
W. Schmidt, Senior Resident Inspector NOTES (1) All were present during the exit meeting (2) Members of Combined NRC-facility Examinations Team (3) Facility Evaluators
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ATTACHMENT 2
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Simulator Examinations (cover sheets only)
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OPERATING SCENARIO 1 DATE 5fIfET -
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INDIVIDUAL PERFORMANCE
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EVALUATORS POSITIONS PACS / FAIL OIbe ! D E {It e SS' h SEMC4Ich PASS / FAIL Nociltba L ke Onde t No., SNO bcMer PASS / FAIL 4 6'Iorns [IN.hA4.ds NCO O NRod( cb
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PASS / FAIL 9 Ocd k4< [ 2 . M .lle r ASS k- 'kC- PASS / FAIL'
I Olcer[s [ lll D A9.I kt XRO b DG0 $ PASS / FAIL CREW PF.RFORMANCE PASS / FAIL.
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SIMULA'It)R OPERATION Initialize: IC 14 - 100% Reactor pot.er steady state conditions Preset Malfunctions:
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1. T=0 RD10, 34-39 and 38-39, Control Rods Stuck.
2. T=5 CU0S, RWCU Resin Depletion, with a 5 minute ramp rate.
3. T=20 TC06B:A, EHC Pressure Regulator Fails Low and TC07:B, EHC System Regulator Oscillates.
4. T=27 'nJ01, Turbine Trip and TC04B: All, Bypass Valves Fail Closed.
(Activated by the EHC malfunction.)
Special Instruction.s:
1. Provide turnover sheet to the SS and discuss initial conditions with the shift.
2. Allow no more than five minutes for panel walkdown, t
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OPERATING SCENARIO 2 DATE 5 Il9>CI -
INDIVIDUAL PERFORMANCE EVALUATORS- POSITIONS PASS / FAIL
! Oc3 ke < Y 2 O(.lker SS h. lc 6 . PASS / FAIL U bC4 0"'I' ! (nMu ll A SNO l 3%Ne# PASS / FAIL-N. meren / M.had,ls NCO b ho s PASS / FAIL ANV ! N' I V ASS Sf07# (h PASS / FAIL
$ Olo/ris !iN Dadels XRO P fdJLndbeckCS ' PASS / FAIL-CREW PERFORMANCE PASS / FAIL SIMULATVR OPERATION Initialize: IC 14 (100% Reactor power steady state condition), with the B CRD purp in operation.
Preset Malfunctions:
1. T=5 ED18:B, Loss of the 10600 Bus.
2. T=25 RR15:B, Coolant leakage inside the Primary Containment, at 100%
severity with a 5 minute ramp rate.
Special Instructions:
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- Provide turnover sheet to the SS and discuss initial conditions with the shift.
2. Allow no more than five minutes for panel walkdown.
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OPERATING SCENARIO 1 DATE 1lB) k l
INDIVIDUAL PERFORMANCE l
l EVALUATORS POSITIONS PASS / FAIL Q d k.v [ [idd le r ss E . N'1 C PASS / FAIL p. D AQh ask.) .ba'c.cIla sNo lk. buIl t 0 A 4 PASS / FAIL Y. Morr[s [ lY DREt$1 i NCO 7. Ni I l a m 5 PASS / FAIL Oc~\M [ k IMi l< /
I ASS U AlIBM PASS / FAIL
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k fM o rr s [d!.bAneel5 XRO $ IM(^ b M 5 PASS / FAIL CREW PERFORMANCE PASS / FAIL SIMULA'IT)R OPERATION Initialize: IC 14 - 100% Reactor power steady state conditions Preset Malfunctions:
1. T=0 RD10, 34-39 and 38-39, Control Rods Stuck.
2. T=5 CU05, RWCU Resin Depletion, with a 5 minute ramp rate.
3. T=20 TC06B:A, EHC Pressure Regulator Fails Low and TC07:B, EHC System Regulator Oscillates.
4. T=27 TUO1, Turbine Trip and TC04B: All, Bypass Valves Fail Closed.
(Activated by the EHC malfunction.)
Special Instructions:
1.
Provide turnover sheet to the SS and discuss initial conditions with the shift.
2. Allow no more than five minutes for panel walkdown.
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OPERATING SCENARIO 2
'DATE 5 I f 8?/ UN
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INDIVIDUAL PERFORMANCE EVALUATORS POSITIONS PASS / FAIL
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PASS / FAIL
. N . b i b a skCs ! b E t \kA SNO k bOk! OdLtd PASS / FAIL 5. N e [s3 ! Ai DAE.Js NCO T. n W hews PASS / FAIL MkW . hlill V ASS b N aj JC PASS / FAIL
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kAlorr[5 ,/ Nb b.'e I s XRO e IN E I! A ns PASS / FAIL CREW PERFORMANCE PASS / FAIL SIMULATOR OPERATION Initialize: IC 14 (100% Reactor power steady state condition), with the B CRD pump in operation.
Preset Malfunctions:
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1.- T=5 ED18:B, Loss of the 10600 Bus.
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T=25 RR15:B, Coolant leakage inside the Primary Containment, at 100%
severity with a 5 minute ramp rate.
Special Instructions:
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Provide the shift.
turnover sheet to the SS and discuss initial conditions with 2. Allow no more than five minutes for panel walkdown.
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. OPERATING SCENARIO 6 oATE s h I9A C '*" C'
INDIVIDUAL PERFORMANCE EVALUATORS POSITIONS PASS / FAIL
) Qd kAr [ b. AhllW SS . bd PASS / FAIL W b2 6 A /J.Co,1.ulk SNO L%%#
PASS / FAIL 4 biorr s - .bs.d5 NCO N bdeblG PASS / FAIL 7. OcdlW [ k. bliOW ASS _. . bu %mn PASS / FAIL k. biorr 5 h.basedi XRO b- O$1 PASS / FAIL CREW PERFORMANCE PASS / FAIL
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SIMULATOR OPERATION Initialize: IC 14, with the "C" Service Air Compres.sor out of service.
Preset Malfunctions:
1. T=0 IA03:C, Air Compressor Trip, this compressor is out of service at the beginning of'this scenario. The control switch for the
"C" service air compressor should be red tagged in the PTL position.
2. T=0 RP01:A and B, Automatic and Manual Scram Failure.
3. T=5 IA03:B, "B" Air Compressor Trip.
4. T=5 IA002, with a 2% severity, triggered by the "B" air comp. trip 5. T=25 EG 10 Stator Water Cooling low flow - Turb trip Special Instructions:
1.
Provide turnover sheet to the SS and discuss initial conditions with the shift. The C air compressor repairs are almost completed and it should be ready for service sometime during this shift.
2. Allow no more than five minutes for panel walkdown.
3. Return the "C" air compressor back to service when the air header pressure reaches 80 psig by removing malfunction IA03:C, and telling the SS that the compressor is ready to be started.
4. Use CAE! SCRAMFUSE to pull the RPS scram solenoid fuses when directed by the SS during the A'lVS. (Accompany the SNO to the 09-15 and 09-17 panels.)
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OPERATING SCENARIO 4 DATE 5 i M M"'
INDIVIDUAL PERFORMANCE EVALUATORS POSITIONS PASS / FAIL bhW.v / k Nk;lktN SS .hr 5b PASS / FAIL O Ybosb for$(4llA SN0 b. 91(ANV PASS / FAIL k. d{orrn dk- G 9. < 5 NCO M.ba9 ik(IG PASS / FAIL OM/ ! k. Al.Ilt( ASS - bab$ PASS / FAIL-T, AlorrIs hi ' cts [t l3 XRO b- 0 85 PASS / FAIL CREW PEPJORMANCE PASS / FAIL
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SIMULATOR OPERATION Initialize: IC 14, with the HPCI system out of service.
Preset Malfunctions:
1. T=0 HP02 HPCI trip -- HPCI is out of service 2. T=5 MC03:A,B,C and D, Main Condenser Water Box Flow Blockage, at 30%
severity and a five minute ramp.
2.
T=25 MS02:A, Steam Leakage Inside the Primary Containment, at 5%
severity and a ten minute ramp, then jump to 100% severity and close the 20 MOV 121 valve.
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Special Instructions:
1.
Provide with the turnover sheet to the SS and discuss initial conditions shift. HPCI is out of service due to logic problems when it inadvertently started yesterday. All of the required surveillance are up to date.
2. Allow no more than five minutes for panel walkdown.
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ATTACHMENT 3 Job Performance Measures (cover sheets only)
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APPROVAL AND DOCUMENTATION SHEET
'9 S/R0 2020101002A TASK TITLE: Increase Recirculat.fon Flow Using APPL. TO 'JPM NUMBER Manual Control REV.
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O DATE: 12-22-88 NRC K/A SYSTEM NUMBER: 202002 JAF TASK NUMBER: 2020101002 JAF QUAL STANDARD NUMBER: 502H.101 AUTHOR: J.M. Dominique DATE: 12-22-88 JTRAINING REVIEW: F.~Catella DATE: 01-30-89 OPERATIONS'. REVIEW: K. Allen DATE: 01-31-89 VALIDATION: EVALUATOR: G. Fronk CANDIDATE: D. Johnson DATE: 04-15-89 COMPLETION TIME: 15 minutes
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CANDIDATE NAME: SSN:
'JPN COMPLETION: ( ) Plant (%) Simulator
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PERFORMANCE EVALUATION: JPM ( ) Satisfactory ( ) Unsatisfactory QUESTIONS #1 ( ) Satisfactory ( ) Unsatisfactory QUESTIONS #2 ( ) Satisfactory ( ) Unsatisfactory DATE PERFORMED TIME TO COMPLETE: Minutes j COMMENTS:
(MANDATORY FOR UNSATISFACTORY PERFORMANCE)
REVIEWED BY:
CANDIDATE EVALUATOR:
SIGNATURE TYPED OR PRINTED Reviewed by: Computer Entry:
Program Administrator Doc. Complete:
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V ' -APPROVAL AND DOCUMENTATION SHEET
L S/RO . 2450101005 _ TASK TITLE: Synchronize the Generator to the APPL. TO JPM NUMBER ' Grid REV.' O -DATE: 1-3-89 NRC K/A SYSTEM NUMBER: 245000
'JAF. TASK NUMBER: '2450101005 JAF QUAL STANDARD NUMBER: 594D.105, AUTHOR: Joan Gregory
, DATE: 1-3-89 TRAINING REVIEW: F. Catella LATE: 1-30-89
. OPERATIONS REVIEW: K. Allen DATE: 1-31-89
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l VALIDATION: EVALUATOR: G. Fronk CANDIDATE: D. Johnson
[ DATE: 04-15-89 COMPLETION TIME: __ 15 minutes
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CANDIDATE NAME: SSN:
JPM COMPLETION: ( ) Plant (X) Sieulator '
PERFORMANCE EVALUATION: JPM ( ) Satisfactory e ( ) Unsatisfactory QUESTIONS #1 ( ) Satisfactory ( ) Unsatisfactory QUESTIONS- #2 ( ) Satisfactory ( ) Unsatisfactory QUESTIONS #3 ( )' Satisfactory ( ) Unsatisfactory DATE PERFORMED TIME TO COMPLETE: Minutes COMMENTS: (MANDATORY FOR UNSATISFACTORY PERFORMANCE)
l REVIEWED BY:
CANDIDATE EVALUATOR:
SIGNATURE i
TYPED OR PRINTED Reviewed by: Computer Entry:
Program Administrator Doc. Completa:
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,' APPROVAL AND DOCUMENTATION SHEET
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S/R0 2120101009 TASK TITLE: Reset an RPS Scram APPL. TO JPM NUMBER i
REV. O DATE: 12-09-88 NRC K/A SYSTEM NUMBER: 212000
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JAF TASK NUMBER: 2120101009 JAF QUAL STANDARD NUMBER: 5005.104 AUTHOR: Michael R. Orta DATE: 12-09-88 i TRAINING REVIEW: F. Catella DATE: 01-30-89 OPERATIONS REVIEW: R. Sarkissian DATE: 01-31-89 VALIDATION: EVALUATOR: G. Fronk CANDIDATE: D. Johnson DATE: 04-15-89 COMPLETION TIME: 10 minutes
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CANDIDATE NAHE: SSN:
JPM COMPLETION: ( ) Plant ()() Simulator PERFORMANCE EVALUATION: JPM ( ) Satisfactory ( ) Unsatisfactory QUESTIONS #1 ( ) Satisfactory ( ) Unsatisfactory QUESTIONS #2 ( ) Satisfactory ( .) Unsatisfactory DATE PERFORMED TIME TO COMPLETE: Minutes COMMENTS: (MANDATORY FOR UNSATISFACTORY PERFORMANCE)
REVIEWED BY:
CANDIDATI EVALUATOR:
___ SIGNATURE TYPED OR PRINTED l Reviewed by: Computer Entry:
l Program Administrator Doc. Complete:
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- ' APPROVAL AND DOCUMENTATION SHEET S/RO 2010101012 TASK TITLE: Isolate the Hydraulic Control APPL. TO JPM NUMBER Unit REV. O DATE: 11-4-88 NRC K/A SYSTEM NUMBER: 201001
. JAP TASK NUMBER: 2010101012 JAF QUAL STANDARD NUMBER: 503C.202 AUTHOR: R.J. Neme*h DATE: 11-4-88 TRAINING REVIEW: F. Catella DATE: 02-07-89 OPERATIONS PEVIEW: R. Sarkissian DATE: _ 02-13-89 VALIDATION: EVAL'UATOR: G. Fronk CANDIDATE: D. Johnson DATE: 04-07-89 COMPLETION TIME: 30 minutes
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CANDIDATE NAME: SSN:
JPM COMPLETION: ( K) Plant ( ) Simulator PERFORMANCE EVALUATION: JPM ( ) Satisfactory ( ) Unsatisfactory.
QUESTIONS #1 ( ) Satisfactory ( ) Unsatisfactory QUESTIONS #2 ( ) Satisfactory (. ) Unsatisfactory DATE PERFORMED TIME TO COMPLETE: Minutes COMMENTS: (MANDATORY FOR UNSATISFACTORY PERFORMANCE).
REVIEWED BY:
CANDIDATE EVALUATOR:
SIGNATURE TYPED OR PRINTED Reviewed by: Computer Entry:
Program Administrator Doc. Complete:
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j APPROVAL AND DOCUMENTATION SHEET
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l l S/R0 2010201015 TASK TITLE: Change Over a Control Rod Drive APPL. TO JPM NUMBER Flow Control Valve REV. O DATE: 11-9-88 NRC K/A SYSTEM NUMBER: 201003 JAF TASK NUMBER: 2010201015 JAF QUAL STANDARD NUMBER: 503C.401 AUTIIOR: R.J. Nemeth DATE: 11-9-88 TRAINING REVIEW: F. Catella DATE: 2-7-89 I OPERATIONS REVIEW: R. Sarkissian DATE: 2-13-89 VALIDATION: EVALUATOR: G. Fronk CANDIDATE: D. Johnson DATE: 04-07-89 COMPLETION TIME: 25 minutes
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CANDIDATE NAME: SSN:
JPM COMPLETION: ( K) Plant ( ) Simulator PERFORMANCE EVALUATION: JPM ( ) Satisfactory ( ) Unsatisfactory QUESTIONS #1 ( ) Satisfactory ( ) Unsatisfactory QUESTIONS #2 ( ) Satisfactory ( ) Unsatisfactory DATE PERFORMED TIME TO COMPLETE: Minutes COMMENTS: (MANDATORY FOR UNSATISFACTORY PERFORMANCE)
REVIEWED BY:
CANDIDATE EVALUATOR:
SIGNATURE
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TYPED OR PRINTED Reviewed by: Computer Entry:
Program Administrator Doc. Complete:
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-- ' JOB PERFORMANCE MEASURE.
A APPROVAL AND DOCUMENTATION SHEET
- S/R0 2150101022' TASF. TITLE: Bypassing Local Power APPL.-TO- JPM NUMBER Range Monitor REV. 0= DATE: 12-28-8B NRC K/A SYSTEM NUMBER: 215005 JAF TASK NUMBER: 2150101022 JAF QUAL STANDARD NUMBER: 5007.105
- A OR: 3 H. Dominique DATE: '12-28-88 TRAINING REVIEW: F. CATELLA DATE: 02-15-89 OPERATIONS REVIEW: K.-ALLEN DATE: 02-15-8N VALIDATION: EVALUATOR:' G. Fronk CANDIDATE: D. Johnson DATE: 04-15-89 COMPLETION TIME: 20 minutes
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CANDIDATE NAME:
RSN:
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PM COMPLETION:-( -) Plant (X ) Simulator '
PERFORMANCE EVALUATION: JPM ( ) Satisfactory ( .) Unsatisfactory QUESTIONS #1 ( ) Satisfactory ( ) Unsatisfactory QUESTIONS- #2 ( ) Satisfactory ( ) Unsatisfactory
.DATE PERFORMED TIME TO COMPLETE: , Minutes COMMENTS:
(MANDATORY FOR UNSATISFACTORY PERFORMANCE)
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REVIEWED BY: i
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CANDIDATE EVALUATOR:
SIGNATURE TYPED OR PRINTED
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NES YORK POUER AUTHORITY
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- APPROVAL AND DOCUMENTATION SHEET S/R0 2620101002 TASK TITII
- Shift Aux Buses From T4 to APPL. TO JPM NUMBER . Reserve Station Service REV. O DATE: 12-19-88 NRC K/A SYSTEM NUMBER: 262001 JAF TASK NUMBER: 2620101002 JAF QUAL STANDARD NUMPER: 571A.210 AUTHOR: J.M. Dominique DATE: 12-19-88 TRAINING REVIEW: F. Catella DATE: 01-30-89 OPERATIONS REVIEW: K. Allen DATE: 01-31-89 VALIDATION: EVALUATOR: G. Fronk CANDIDATE: D. Johnson DATE: 04-15-89 COMPLETION TIME: 20 minutes
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CANDIDATI NAME: SSN:
JFM COMPLETION: ( ) Plant (>() Simulator PERFORMANCE EVALUATION: JPM ( ) Satisfactory ( ) Unsatisfactory QUESTIONS #1 ( ) Satisfactory ( ) Ur. satisfactory QUESTIONS #2 ( ) Satisfactory ( ) Unsatisfactory DATE PERFORMED TIME TO COMPLETE: Minutes COMMENTS: (MANDATORY FOR UNSATISFACTORY PERFORMANCE)
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REVIEWED BY:
CANDIDATE l \
EVALUATOR:
SIGNATURE l
TYPSD OR PRINTED l l
l Reviewed by: Computer Entry:
Program Administrator f Doc. Complete: _
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, l bB PERFbREUICE U bbRb APPROVAL AND DOCUME!TTATION SHEET S/R0 2000401233 TASK TITLE: Attempt to Close an SORV from APPL. TO JPM NUMBER the Remote SRV Penel REV. O DATE: 1-20-89 NRC K/A SYSTEM NUMBER: _
239002 JAF TASK NUMBER: 2000401233 JAF QUAL STANDARD NUMBER: 5A0P.115 AUTHOR: J. E. Gregory DATE: __ I-20-89 TRAINING REVIEW: G. Fronk DATE: 1-27-89 OPERATIONS REVIEW: T. Plumpton DATE: 1-27-89 VALIDATION: EVALUATOR: G. Fronk CANDIDATE: D. Johnson DATE: 04-07-89 COMPLETION TIME: 15 minutes
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CANDIDATI NAME: SSN:
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JPM COMPLETION: (D%J Plant ( ) Simulator i i
PERFORMANCE EVALUATION: JPM ( ) Satisfactory ( ) Unsatisfactory QUESTIONS #1 ( ) Satisfactory ( ) Unsatisfactory QUESTIONS #2 ( ) Satisfactory ( ) Unsatisfactory DATE PERFORMED TIME TO COMPLETE: Minutes
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COMMENTS: (MANDATORY FOR UNSATISFACTORY PERFORMANCE) !
REVIEWED BY:
CANDIDATE EVALUATOR:
SIGNATURE TYPED OR PRINTED Reviewed by: Computer Entry:
Program Administrator Doc. Complete:
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, lbBPEEFbREEC5$iESUkE j APPROVAL AND DOCUMENTATION SHEET S/R0 2000401234B TASK TITLE: E0P Isolation / Interlock Overr. ides APPL. TO JPM NUMBER - Containment Vent and Purge System REV. O DATE: 2-6-89 NRC K/A SYSTEM NUMBER: 295031 JAF TASK NUMBER: 2000401234 JAF QUAL STANDARD NUMBER: 5A0P.222 AUTHOR: Joan Gregory DATE: 2-6-89 TRAINING REVIEW: G. Fronk DATE: 2-13-89 OPERATIONS REVIEW: D. Squires DATE: 2-13-89 i
VALIDATION: EVALUATOR: G. Fronk CANDIDATE: D. Johnson DATE: 04-07-89 COMPLETION TIME: 10 minutes
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CANDIDATI NAME: SSN:
JPM COMPLET10N: (>() Plant ( ) Simulator PERFORMANCE EVALUATION: JPM ( ) Satisfactory ( ) Unsatisfactory QUESTIONS #1 ( ) Satisfactory ( ) Unsatisfactory QUESTIONS #2 ( ) Satisfactory ( ) Unsatisfactory DATE PERFORMED TIME TO COMPLETI: Minutes COMMENTS: (MANDATORY FOR UNSATIS'/ACTVRY PERFORMANCE)
REVIEWED BY:
CANDIDATE EVALUATOR:
SIGNATURE
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l Reviewed by: Computer Entry:
l Program Administrator Doc. Complete:
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_' APPROVAL AND DOCUMENTATION SHEET S/R0 2060101005 TASK. TITLE: Manually Initiate the HPCI APPL. 1D JPM NUMBER ' System.
REV. O DATE: 11-16-88 NRC K/A SYSTEM NUMBER: 206000 JAF TASK NUMBER: 2060101005 JAF QUAL STANDARD NUMBER: 5023.101 AUTHOR: R.J. Nemeth DATE: 11-16-88 TRAININC REVIEW: G. Fronk DATE: 04-14 89__
UPERATIONS REVIEW: D. Johnson DATE: 04-15-89 VALIDATION: EVALUATOR: G. Fronk CANDI'D ATE: D. Johnson DATE: 04-15-89 COMPLETION TIME: 15 minutes
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CANDIDATE NAME: SSN:
JPM COMPLETION: ( ) Plant ( D() Simulator PERFORMANCE EVALUATION: JPM .( ) Satisfactory ( ) Unsatisfactory QUESTIONS #1 ( ) Satisfactory ( ) Unsatisfactory QUESTIONS #2 (' ) Satisfactory ( ) Unsatisfactory DATE PERFORMED TIME TO CCMPLETI: Minutes
COMMENTS: (MANDATORY FOR UNSATISFACTORY PERFORMANCE)
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REVIEWED BY:
CANDIDATE EVALUATOR:
5?GNATURE TYPED OR PRINTED Reviewed by: Computer Entry:
l Program Administrator ,
Doc. Complete:
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ATTACHMENT 4 Written Examination and Answer Key, Part A (R0 and SRO)
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Rovision 4 4/30/86 f. .
. NEW YORK POWER AUTHORITY EXAMINATION / QUIZ JAMES A. FITZPATRICK NUCLEAR POWER PLANT COVER SHEET l'
Exami. nation Title: DICENSE hN UD D C~ $2 D RE QUALIFICATION EXAM PART A-1 Examination Approval: => -
'Phogram~ Administrator Date: I!!!Ji
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Open Book ( X ) Closed Book ( ) Time Limit: 1 HR.
LAuthorized Reference Material: ANY MATERIAL NORMALLY FOUND IN T*IE CONTROL ROOM j Minimum Grade: Graded by:
Acceptable Grade
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STUDENT DATA i l
Ntme: S.S.#
Last First M.I.
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Employer: NEW YORK POWER AUTHORITY Date:
D3partment: OPERATIONS
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, GUIDELINES 1. Remain quiet during the exam.
2. If you have any questions during the exam, raise your hand. Your instructor will provide clarification wherever possible.
3. You are expected to do your own work and not to help anyone else.
4. Use only the authorized reference material.
5. At the completion of this examination, you are to sign the following
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certification.
l I certify all answers contained in this examination are my own. In addition, I have not received nor given any unauthorized assistance, nor have used any unauthorized references.
S'JDENT SIGNATURE: DATE:
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EXAM REVIEW /REMEDIATION Tha instructor has reviewed the exam and provided an explanation of the correct answers.
STUDENT ACKNOWLEDGEMENT: DATE:
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SCENARIO TITLE PAGE .!
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Scenario Number 8 Type of Scenario ___ Abnormal Status Title Stuck Open SRV with Tailpiece Rupture Above Suppression Pool Level ,
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Synopsis: The reactor plant had been operating at 100%.of rated thermal (
power'and equilibrium conditions when'a safety relief valve inadvertently opens and sticks open. The resulting hydraulic shock of valve operation causes that valve's discharge tailpipe to break at a location inside the torus, above the suppression pool's water level. Torus pressure begins to increase ca ning an increase in drywell pressure, due to the operation of the torus-ts drywell vacuum breakers. The scenario is terminated approximately 20 seconds after the relief valve opens and before any automatic actions occur due to the high drywell pressure.
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Scenario Number 8 Type of Scenario Abnormal Status
Title Stuck Open SRV with Tailpiece Rupture Above Suppression Pool Level ]
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Initial Conditions: IC 14, 100% power at MOL Initiating Cue: Inadvertent opening of a safety relief valves T=0 Expected Sequence of Events:
1. Reactor safety relief valve inadvertently opens and retains open.
2. SRV leaking and SRV sonic monitor annunciators alarm.
3. Torus pressure and temperature will increase, causing the torus-to-drywell vacuum breakers to open.
4. Drywell pressure and temperature will start to increase. '
5. The EHC regulator will reduce turbine steam flow and generator output.
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, SIMULATO:2 OPERATIONS SUMMARY SHEET
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Scenario'Humber 8 Type of Scenario Abnormal Status
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Title _ Stuck Open SRV with Tailpiece Rupture Above suppression Pool level .
I Initial Conditions: IC 14 Manipulations:
Go to RUN:
Enter MALFUNCTIONS:
MS16:J; ADS Discharge Line Break Above Torus Water Level at 10%
severity and with no time delay AD06 JJ Reactor Pressure Relinf Valve Inadverten?ly Opens AD08:J; Reactor Pressure Relief Valve Stuck Open I/03 Annunciator override on for 09-4-1-16 l Set. REMOTE CONDITIONS:
NONE ACTIONS:
Continually silence and acknowledge alarms Post Freezes None Freeze T=22 secunds after the initiation of the malfunction (s); or prior to the automatic operations caused by high drywell pressure
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Scenario Number 8 Type of Scenario Major Failure Title Stuck Open SRV with Tailpiece Rupture Above Suppression Pool Level
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Plant was in a normal operating configuration; at 100% of rated thermal power.
I Plant conditions existed as follows:
Core flow = 75 M1b/hr Reactor press =
100g psig Recire A flow = 44K gpm Recirc A temp = 520 F Recirc B flow = 44K gpm Rectre B temp = 520 F Steam flow = 10.4 M1b/hr Reactor level = 201" NR Feed flow = 10.1 M1b/hr APRMs all reading 98-100%
Torus level = 13.88 - 14' I Torus temp = 78-80 F IRMs = Withdrawn DW pressure DW temperature
=
1.9 psig '
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= 113-120 F SRMs = Withdrawn Torus Pressure = 0 psig The only action that has been taken is to silence and acknowledge annunciator alarms.
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- , SIMUI.ATOR CPERATIONS SUMMARY SHEET-
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Scenario Number 8 Type of Scenario 'l Abnormal Status l
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Title Stock Open SRV with Tailpiece Rupture Above Suppression Pool Level
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Check conditions: I l
Pant s: I Instrument readings changed from initial conditions:
Core flow = 74 M1b/hr Reactor press =
Recire A flow = 44K gpm Recirc A temp 100g psig
= 520 F !
Recire 8 flow = 44K gpm Recirc B temp = 520 F Steam flow = 9.8 M1b/hr Reactor level = 202" NR Feed flow = 10.8 M1b/hr APRMs all 97-98%
Torus level = 13.88'
Torus H2O temp = 81 F Torus pressure = 2.5 psis IRMs = Withdrawn DW pressure = psig SRMs = Withdrawn DW temperature = 2.1,F 116 Abnormal annunciators lit 9-3-5; 39 9-4-2; 6,15 9-5-1; 34 9-4-1; 16 9-6-4; 14 9-7-1; 28 9-7-3; 45 Procedures:
None
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NEW YORK POWER AUTHORITY
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JAMES A. FITZPATRICK NUCLEAR POWER PLANT j
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REQUALIFICATION QUESTION /-ANSWER FORM
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~ 8-C-1 i
What reactor scram (automatic signal) is most likely to occur, from present plant conditions, parameters and. trends. '
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ANSWER:
8-C-1 .(1.00)
High drywell pressure - (indicated pressure has increased to 2.1 psig and suppression pool pressure is~over 2.5 psig.)
REFERENCE: Control Room indications-Drywell Pressure TASK XREF: 295024 High Drywell Pressure 239002 Relief / safety valves K/A XREFf 295D24 EK2.05 3.9/4.0 EA2.02 4.2/4.4 295006 AA2.06 3.5/3.8 259002 A2.05 4.2/4.2 I
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i JAMES A. FITZPATRICK NUCLEAR POWER PLANT -
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REQUALIFICATION QUESTION / ANSWER FORM 2. 8-C-2
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l What control Room indications exist that would justify saying that the SRV discharge.into the torus is or is not superheated.
(1.00)
ANSWER:
B-C-2 (1.00)
Indication of tailpiece temperature (0.5)
any 1
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Temperature Recorder
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EPIC Display
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Rx Pressure and Mollier D$agram (Steam Tables) i Indication of torus pressure (0.5)
any 1
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EPIC Display
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Panel Indication REFERENCES. Steam Tables EPIC Display Torus pressure 5.7 psig SRV Tail Pipe Temperature #15 02SRV-71K TASK XREF: 259002 Relief / Safety Valves K/A XREF: 295005 K1.25 2.8/5.1 i-7-
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.a NEW YORK POWER AUTHORITY
- JAMES A. FITZPATRICK NUCLEAR POWER PLANT REQUALIFICATION, QUESTION / ANSWER FORN I
3.
, 8-R-1 All three lights associated with SRV 02RV-71K are on. Using control room indications verify that each light is or is not appropriate for plant conditions. (1.0)
ANSWER:
8-R-1 GREEN - light is on because the switch is in close (.33)
WHITE -
light is on because flow through the tail pipe has been sensed (.34)
RED -
light should not be on because the SRV's control switch is in the close position. (this may 3 indicate the possible problem with the SRV solonoid) j (.33)
or Red light should be on because something has caused the valve to open.
NOTE: Answer should not indicate that red and green lights are triggered by valve posit' ion.
q REFERENCE: Annunciator status ARP 09-4-3-3 TASK XREF: 295024 High Drywell Pressure 239002 Relief / Safety Valves K/A XREF: 259002 K4.05 3.6/3.7 A2.03 4.2/4.2 SG-8 3.9/3.7-8-l :
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. JAMES A. FITZPATRICK NUCLEAR POWER PLANT REQUALIFICATION QUESTION / ANSWER FORM '
4.
, B-C-4
Exp1&in the change in indicated Reactor Steam Flow from the initial condition. (1.0)
ANSWER:
8-C-4 (1.00)
When the SRV opened, reactor pressure began to decrease.(.33) The EHC systes compensated for thiJ by reducing steam flow to the turbine by closing the Control valves slightly.(.33) The flow through the SRV is not seen by the steam flow detector, since it taps off upstream. (.34)
or Since the SRV's tap into the main steam line upstream of the flow detectors (0.5),.this flow will not be sensed by the flow detectors. (0.5)
REFERENCE: Control Room Indications
. TASK XREF: 239002 Relief / safety Valves K/A XREF: 239002 K1.04 3.6/3.7 A1.09 3.1/3.5 i
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NEW YORK POWER AUTHORITY i
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' JAMES 1A. FITZPATRICK NUCLEAR POWER PLANT l REQUALIFICATION-QUESTION / ANSWER FORM-li l
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8-C-5 l
Which fuse (s) should be pulled to attempt to close the SRV?
Specific fuse number (s) required. (1.00)-
ANSWER:
8-C-5 2E-F3K, 2E-F4K, 2E-F11K, 2E-F12K (0.25 each)
REFERENCE: ESK-11AAD G.E. print 791E453 sh 3 TASK XREF: 239002 Relief / safety Valves K/A XREF: 259002 K2.01 2.8/3.2 A2.03 4.1/4.2 294001 A1.07 3.0/3.7
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, . NEW YORK POWER AUTHORITY i
., -JAMES A. FITZPATRICK' NUCLEAR POWER PLANT l ,
REQUALIFICATION QUESTION / ANSWER FORM 1
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6. 8-S-2a 1'
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Explain the reponse of Containment pressure. (1.0)
l ANSWER:
8-S-2a (1.00)
Drywell pressure has increased due to pressure venting from the Torus (thru the Vacuum breakers) (.5)
Torus pressure has increased due to some energy input above the water level (.5)
REFERENCE: T.S.. Bases, section 5.7.A TASK XREF: 225001 Primary Containment System K/A XREF: 225001 A1.07 5.2/5.4 A2.02 5.9/4.1 K5.07 5.1/3.2 SG-6 5.0/4.0 259002 K4.05 5.1/5.5-11-
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L NEW YORK POWER AUTHORITY
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JAMES A. FITZPATRICK NUCLEAR POWER PLANT
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REQUALIFICATION QUESTION / ANSWER FORM 6. 8-R-2
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Explain the effect the present situation has on feedwater temperature. (1.0)
ANSWER:
8-R-2 (When the SRV opened, reactor pressure dropped. The EHC system compensated by closing down on the turbine control valves. This reduces steam flow to the turbine), which reduces extraction steam flow (0.5) (to the feed water heaters) which in turn decreases feedwater heating, lowering feedwater temperature. (0.5)
REFERENCE: Main control room indications Turbine Steam flow Feedwater temperatures
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TASK XREF: 245000 Main Turbine Generator K/A XREF: 245000 K5.04 5.5/5.5
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, Rovision 4
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!
- NEW YORK POWER AUTHORITY EXAMINATION / QUIZ JAMES A. FITZPATRICK NUCLEAR POWER PLANT COVER SHEET Examination. Title: .
LICENSED OPERATOR REQUALIFICATION EXAM PART A-1 Examination Approval: Date:
Program Administrator Open Book-( X ) Closed Book ( )- Time Limit: 1 HR.
- Authorized Reference M3teria1: ANY MATERI AL NORMALLY FOUND IN THE CONTROL ROOM Minimum Grade: Graded by:
Acceptable Grade
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STUDENT DATA Ntme: S.S.#
Last First M.I.
Employer: _ NEW YORK POWER AUTHORITY Date:
D3partment: OPERATIONS
=========================================______;====_-- -
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GUIDELINES 1. Remain quiet during the exam.
2. If you have any questions during the exam, raise your hand. Your instructor will provide clarification wherever possible.
3. You are expected to do your own work and not to help anyone else.
4. Use only the authorized reference material.
5. At the completion of this examination, you are to sign the following certification. f I certify all answers contained in this examination are my own. In addition, I have not received nor given any unauthorized assistance, j nor have used any unauthorized references, j STUDENT SIGNATURE: DATE: _
=_= -- ==================r-- .___=================______;_=========_
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EXAM REVIEW /REMEDIATION
The instructor has reviewed the exam and provided an explanation of the correct answers STUDENT ACKNOWLEDGEMENT: DATE:
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[*. R vicicn 4
% 4/30/86
'NEW YORK POWER AUTHORITY
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EXAMINATION / QUI 7,
' JAMES A. FITZPATRICK NUCLEAR POWER PLANT COVER SHEET
.405 KG /20?52a
' Examination Title: CICENSEAOPERATORREROALIFICATIONEXA Examination Approval: Date: __ $!/
Program Administrator-Open Book ( X ) Closed Book ( ) Time Limit: 1 HR.
Autha-Tzed Reference M;terial: ANY MhTERIAL NORMALLY FOUND IN THE CO.NTROL ROOM Minimum Grade: Graded by:
Acceptable Grade
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STUDENT DATA Name: S.S.#
Last First M.I.
Employer: NEW YORK POWER AUTHORITY Date: _
Department: OPERATIONS <
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GUIDELINES 1. Remain quiet during the exam.
2. If you have any questions during the exam, raise your hand. Your instructor will provide clarification wherever possible.
3. You are expected to do your own work and not to help anyone else.
4. Use only the authorized reference material.
5. At the completion of this examination, you are to sign the following certification.
I certify all answers contained in this examination are my own. In addition, I have not received nor given any unauthorized assistance,
.nor have used any unauthorized references.
STUDENT SIGNATURE: DATE:
__________________
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m=__________________________________========_==-- ___
l i EXAM REVIEW /REMEDIATION Th^, instructor has reviewed the exam and provided an explane. tion of the correct answers.
STUDENT ACKNOWLEDGEMENT: DATE:
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.f ,. SCENARIO TITLE PAGE a -
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Scenario Number 4 Type of Scenario Major Failure Title Turbine Trip Without Bypas:. Valves Synopsis: The reactor is operating in a normal configuration when the main turbine trips. Following the trip the turbine bypass valves fail to open.
The operators should be concarned with the immediate actions for scram and turbine trip. The SR0s should be planning for long term cooling of the reactor.
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- f . . SCENARIO DEVELOPMENT FORM
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Scenario Number 4 Type of Scenario Major Failure Title Turbine Trip Without Bypass Valves l Initial Conditions: IC 14, 100% power at MOL Initiating Cue: Main turbine trip; T=0 Expected Sequence of Events:
1. Turbine trips 2. Reactor scrams 5. Turbine bypass valves fail to open 4. Reactor pressure rises
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SIMULATOR OPERATICHS SUMMARY SHEET '
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Scenario. Number 4- Type of Scenario Major Failure
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Title Turbine Trio Without Bypass valves Initial Conditions: IC 14, place FWC in 3-element Manipulations:
Go to-RUN:
(NOTE: may run CAE SS4A instead of thc following)
Enter MALFUNCTIONS -
TC04B A,B,C,& D; Bypass Valve (All) Fails Closed TUO3: Loss of Bearing 011 100% - 3 min. ramp Set REMOTE CONDITIONS:
NONE ACTIONS:
Continually silence and acknowledge alarms Carry out immediate scram actions Post Freeze: Place FWC in 1-element Freezer When reactor pressure is discernible above pressure set at approxi.;ately 3.5 minutes after the malfunctions went active.
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f*. PLANT STATUS SHEET
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Scenario Number 4 Plant was in a normal operating configuration; at 100% of rated thermal j power.
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Plant conditions existed as follows:
Core flow = 75 M1b/hr Reactor press = 1000 psig Recire A flow = 44K gpm Recirc A temp = 520 F Recire B flow = 44K gpm Recirc B temp = 520 F Steam flow = 10.4 M1b/hr Reactor level = 201" NR Feed flow = 10.1 M1b/hr-APRMs all reading 98-100% Torus level = 13.88 - 14'
Torus temp = 78-80 F IRMs = Withdrawn DW pressure =
1.8 psig DW temperature = 115-120 F SRMs = Withdrawn Operator action has taken place to silence and acknowledge the annunciator alarms. Scram actions taken - placing the mode switch in S/D.
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t-l , '*. SIMULATOR OPERATIONS SUMMARY SHEET
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Scenario Number 4 Type of Scenario Major Failure i
Title Turbine Trip Without Bypass Valves Check conditions:
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Panels:
Reactor scrammed.
Instrument readings changed from initial conditions:
Core flow = 88 M1b/hr Reactor press =
104g-1080 psig Recirc A flow = 44K gpm Recirc A temp = 520 F Recirc B flow = 44K ppm Recirc B temp ' 520 F Steam flow = 1.0 M1b/hr Reactor level = 185-195" NR Feed flow ' = 11.0 M1b/hr APRMs - Downscale Torus level = 15.9'
Torus temp = 78-80 F IRMs = Withdrawn DW pressure = 2.0 psis DW temperature = 115'F SRMs = Withdrawn Abnormal annunciators lit:
9-4-2; 15 9-5-1; 3, 4, 9, 10, 28, 38, 45, 51. 52, 55 9-5-2; 1, 2, 5, 7, 9, 11, 25, 55, 54 9-6-1; 19 9-6-5; 31 9-6-4; 14 9-7-1; 3, 5, 26, 28 9-7-2; 1, 5, 7 9-7-5; 6, 38, 45 9-8-*; 26, 52 9-8-5; 3, 15, 15 Procedures: 1 AOP-1 and EDP-2 f-5-
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REQUALIFICATION QUESTION / ANSWER FORM
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4-C-2 1. Exolain why Reactor pressure is at its present value.
(1.0)
ANSWER:
1. The Turbine Bypass valves have failed to open (on the Turbine trip signal) to control reactor pressure. (Decay heat has provided an energy input which is seen as a pressure increase)
(1.0)
REFERENCE: F-ADP-1 C.2.2 0DS0-02 6.5, 6.5.7 TASK XREF: 249-01-003 Nonitor the Reactor / Turbine Pressure Regulator 1 System.
200-04-222 Malfunction in EHC System K/A XREF: 241000 K1.06 3.8/3.9 A1.07 3.8/3.7 A3.08 3.8/3.8 A4.06 3.9/3.9 SG-7 3.5/3.5
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- .-* JAMES A. F5TZPAiRIUK NUC[EIk PbWER PLANT s REQUALIFICATIDH QUESTION / ANSWER FORM l .
4-C-3 1 .
l 2, What type of transfer (fast, res11ual or manual) of house loads occurred. Justify your answer. :1.00) j
ANSWER:
2. Fast Transfer (0.5)
Any one of the following (0.5)
10100 and Ib200 busses still energized 4 KV motor loads still energized 10300 and 10400 L-Gear feeder breakers still closed EDGs have not started REFERENCE: OP-46A TASK XREF: 295005 Main Turbine Generator Trip !
K/A XREF: 295005 AK2.08 3.2/3.5 AA1.07 3.3/3.3 AA2.08 3.2/3.5 i
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NEW YORK POWEQ AUTHORITY )WER PLANT AMES-A. FITZPATRICK NUCLEAR POWER PLANT iWER FORM tCEQUALIFICATION QUESTION / ANSWER FORM 4-C-5a 4-C-4
' mal / emergency procedures ho'ooin turbine trip? Be specific.(1.00) (1.0)
tripped on a loss of lube oil pressure. (since ng woor device operates via a lube oil pressure switch, so octivated.) (3e0) NEP~
NWW(0.75)
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ip.9 lont annunciators
'PIC display - TGO 45000 Turbine Generator Operation 95005 Turbine Generator Trip
'45000 K1.08 3.4/3.5 K3.08 3.7/3.8 ID A2.02 3.3/3.5 95005 SG-5 3.6/3.6
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REQUALIFICATION QUESTION / ANSWER' FORM I
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l 4-C-7a
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5. Justify the statement " The Reactor Protection System has functioned to SHUTDOWN the Reactor". (4 Responses Required) (1.0)
ANSWER:
5. Any four answers at (.25) each.
Any valid indication that RPS has functioned electrically Any valid indication that RPS has functioned pneumatically Any valid indication that RPS has functioned hydraulically Any valid indication that control rods have inserted Any valid indication that Rx is shutdown or subcritical Some examples are:
RPS white lights are out (.25)
APRMs are 1ownscale( < 2.5 ) (.25)
Scram valves - blue lights are lit (.25)
Control' rod full IN lights are lit (.25)
REFERENCE: ADP-1 pg. 2 TASK XREF: 295006 Reactor Scram K/A XREF: 295006 AK2.01 4.3/4.4 SG-9 3.9/4.1 i
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JANES A. FITZPATRICK NUCLEAR POWER PLANT REQUALIFICATION QUESTION / ANSWER FORM
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4-S-1
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6. What signal initiated the reactor scram and how could this be verified to be the cause? (1.00)
ANSWER:
6. The reactor should have scrammed on the turbine trip (TSV position).
(0.5) This could be verified by checking the post trip log (sequence of events) (0.5)[the alarm typer may also be referenced for this]
REFERENCE: F-AOP-2 09.5.2-6, Rev. 2 TASK XREF: 200-04-207 Turbine Generator Trip 200-04-217 Reactor Scram 200-04-222 Malfunction in EHC System K/A XREF: 295005 AA 1.02 3.6/3.6 AK 3.01 5.8/5.8 SG-3 2.9/2.9-11-
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JAMES A. FITZPATRICK NUCLEAR POWER PLANT REQUALIFICATION QUESTION / ANSWER FORM
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4-R-la
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6. Explain the reason for the present Feedwater flow rate.
(1.0)
ANSWER:
6. Reactor water level is below the FWLC system setpoint (.50)
therefore this system is trying to restore water level by (raising the speed of the Rx feed pumps) raising flow. '(.50)
REFERENCE: Reactor Pressure Indications Bypass valve position TASK XREF: 259001 Reactor Feedwater System K/A XREF: 259001 A3.04 3.8/3.7
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l ATTACHMENT 5 Written Examination and Answer Key, Part B (R0 and SRD)
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., Rovision 4
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4/30/86 -
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NEW. YORK POWER AUTHORITY ~ EXAMINATION / QUIZ JAMES.A. FITZPATRICK NUCLEAR POWER PLANT COVER SHEET A#b 5 K.6W Examination Title: LICENSED, OPERATOR REQUALIFICATION EXAM PART B (RO)
Examination Approval: } ~Date:
Program Administrator Open Book ( X ) Closed Book ( ) Time Limit: 2 HR.
. Authorized Reference Material: ANY MATERIAL NORMALLY FOUND IN THE CONTROL ROOM _
Minim'.un Grade- Graded by:
Acceptable Grade
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_____s============
STUDENT DATA
'Nnme: S.S.#
Last First M.I.
Employer: NEW YORK POWER AUTHORITY Date:
D3partment: OPERATIONS
_________________===____________- e-- - ___===========_____m___m_===
GUIDELINES 1. Remain quiet during the exam.
2. If you have any questions during the exam, raise your hand. Your instructor will provide clarification wherever possible.
l 3. You are expected to do your own work and not to help anyone else.
4. Use only the authorized reference material.
5. At the completion of this examination, you are to sign the following certification, i I certify all answers contained in this examination are my own. In addition, I have not received nor given any unauthorized assistance, nor have used any unauthorized referencer.
STUDENT SIGNATURE: DATE:
=========_________====__________: -_________c=m___===========_____s
EXAM REVIEW /REMEDIATION
.The instructor has reviewed the exam and provided an explanation of the correct answers. i STUDENT ACKNOWLEDGEMENT: DATE: j
,
__________.______.______.______________________._______m.m._, . _ _ _ . _ . _ . _ _ . _ _ _ _ _ . _ _ . _ _ _ __ _ . _ _ _ _ _ _ _ _ . . , _ _ _ ._. _ _ _ _ . , _ _ _ _ _ . , _ . . . _ . . _ _ . _ . _ _ _ _
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l- v 7. NEW YORK POWER AUTHORITY
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. JAMES A. FITZPATRICK NUCLEAR POWER PLANT -
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REQUALIFICATION QUESTION / ANSWER FORM 1. 200-204-1 The reactor is operating normally at rated power along the 100% rod line when the NCO notices that condenser vacuum is slowly decreasing. During the l investigation of this condition he performs the following actions:
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Trips the recombiner.
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Places the spare air ejectors into service.
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Closes the vacuum drag valve.
ACsuming that condenser vacuum can be maintained above the 22.5" Hg trip by l the reduction of reactor power.
Describe the reactivity manipulations the operctor must complete to reduce roactor power to 50% and why? (1.0)
ANSWER: i The operator may reduce reactor power using the recirculation system until core flow has been reduced to 45% (34.7 Ml/hr), at which time the operator must insert the control rods to reduce reactor power to below the 80% rod line. ( 0. 5 ) ' Reactor operation with core flow less than 45% of rated (34.7 Ml/hr) and core thermal power above the 80% rod line is prohibited as thermal hydraulic instability could occur in that region. (or because intentional entry into regions A, B, or C is prohibited (0.5)
PEFERENCE: F-AOP-31, Loss of Condenser Vacuum, Rev. 2, pg.3 RAP 7.3.16, Plant Power Changes, Rev 12, pg.13 K/A XREF: 295002 AK3.09 3.2/3.3 AA2.02 3.2/3.3 SG-7 3.2/3.2 SG-11 3.7/3.8 295001 AK1.02 3.3/3.5 AA1.01 3.5/3.6 AA2.01 3.5/3.8 SG-7 3.3/3.6 q
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. NEW YCRK POWER AUTHORITY
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JAMES A. FITZPATRICK NUCLEAR POWER PLANT
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REQUALIFICAT10N QUESTION / ANSWER FORM 2, 200-206-1 The plant has experienced a full reactor scram and the following conditions result:
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35 control rods are at positions > 06.
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All CRD hydraulic control unit scram valves indicate open.
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Reactor power is less than 2 %.
- RPV water level is 190 inches and stable, i
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RPV pressure is < 1090 psig and is being controlled with the main )
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All SRVs are operable and none are cycling. l
- The scram signal cannot be reset j l
What methods / operator actions may by attempted in order to insert the control rods which are > 06 into the core? (1.5)
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I ANSWERS:
o. Start both CRD pumps, close charging water supply valve CRD-56, (defeat RSCS interlocks (if necessary)), and insert the rods manually. (0.75)
b. Vent the CRDM overpiston area '(withdraw line) on the rods. (0.75)
REFERENCES: F-AOP-1 Reactor Scram, pg 4, Rev 10 F-EOP-2 RPV Control, pg 31, Rev 2 F-AOP-34 Alternate Control Rod Insertion, pg 4&S, Rev 2 K/A REF: 295015 AA1.01 3.8/3.9 AA1.03 3.6/3.8 AK2.01 3.8/3.9
. SG-6 4.1/3.9 SG-11 4.2/4.4 SG-12 3.7/4.4
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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT
REQUALIFICATION QUESTION / ANSWER FORM 3. 200-209-1 Which RPV water level instruments, below, are reliable for the following conditions. Explain. (1.0)
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The operating feed pumps trip and the reactor scrams on low RPV water level.
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During the recovery, drywell cooling is lost and drywell temperature is noted to average 320 deg F. (16-1-RTD-lO7 & 108)
- Actual RPV water level is 175" above TAF.
a. Wide range Yarway b. Narrow range GEMAC c. Refueling GEMAC d .. Fuol zens Yarway ANSWER:
a'. Wide range Yarway reliable (.25)
b. Narrow range GEMAC c. Refueling GEMAC d. Fuel zone Yarway reliable (.25)
These instruments are the only instruments which are still operating within acceptable conditions ( IAW EOP-1, EOP Caution #6.) (0.5)
Also may give a qualified answer that if RWR pumps are running, the fuel zone instruments will not give reliable indication.
REFERENCE: F-EOP-1, EOP Cautions. pg 13, Rev 3.
K/A REF: 295028 EKl.01 3.5/3.7 EK2.03 3.6/3.8 EA2.03 3.7/3.9 SG-7 3.4/3.8 SG-12 3.8/4.3
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NEW YORK POWER AUTHORITY
JAMES A. FITZPATRICK NUCLEAR POWER PLANT I
REQUALIFICATION QUESTION / ANSWER FORM l
4. 202-004-1
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Just after placing the Turbine in service (synchronized to the grid),
during a normal reactor startup, the NCO attempts to raise reactor power ucing recirculation pump flow but pump speed doesn't increase. Explain if this condition is er is not expected for the present plant situation.
(.50)
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ANSWER:
Feedwater flow must be greater than 20% before recirculation pump speed can increase to above 30% (minimum). (.50)
REFERENCE: F-OP-27, Recirculation System, Rev. 25, rJ.1 K/A XREF: 202001 A1.01 3.6/3.5 A4.01 3.7/3.7 A4.04 3.7/3.7
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NEW YORK POWER AUTHORITY
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.' JAMES A. FITZPATRICK NUCLEAR POWER PLANT
.. REQUALIFICATION QUESTION / ANSWER FORM
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5. 205-001-1 The r 2ctor plant is undergoing a transient with the following information available to the operator:
- Drywell temperature is 250 deg F.
- Torus water level is 15 feet.
- Drywell pressure is 25 psig and increasing.
- Torus water temperature.is 95 deg F and constant.
- Torus air temperature is 165 deg F.
- Reactor water level is 50 inches and slowly increasing.
- Reactor vessel pressure is 385 psig.
- Suppression chamber pressure is 23 psig and increasing
- RHR pumps are running in the LPCI mode Under the above plant conditions, what specific operator actions (control manipulations) must be accomplished to initiate drywell sprays? (1.0)
ANSWERS:
a. Place .the drywell spray control switch in manual (or normal after manual) (.34)
b. Open 10-MOV-26A(B) (.33)
c. Throttle open 10-MOV-31A(B) to obtain the desired flow (.33)
- partial credit may be given in the reverse sequence for b,c REFERENCE: F-OP-13, RHR System, Rev 46, pg 11, steps 2.a.2.b & 2.b.2 K/A XREF: 226001 A2.20 3.7/4.1 i A4.07 3.5/3.5
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. NEW YORK POWER AUTHORITY
JAMES A. FITZPATRICK NUCLEAR POWER PLANT REQUALIFICATION QUESTION / ANSWER FORM 6. 215-017-1 Shortly after a reactor scram from full power, a' reactor startup is initiated. Just after criticality is reached a peripheral rod withdrawal q of one notch causes the following conditions to occur: ]
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- The NCO observes a reactor period of 15 seconds and an increasing l count rate on all SRMs. ]
- The following alarms are received:
SRM UPSCALE OR INOP (09-5-2-51)
SRM PERIOD (09-5-2-41)
. ROD WITHDRAW BLOCK (09-5-2-2)
c. How is it possible for a peripheral rod to have this worth? (0.5)
b. What operator action is required? (0.5)
ANSWER:
a. The large xenon concentrations present in a core (within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) after shutdown causes rod worths to increase.(0.5)
b. Reinsert the rod to its previous position. (0.5)
REFERENCES: Start-up Procedure F-OP-65 Technical Specifi ations Annunciator Respg.se Procedures ARP 09-5-2-41 & 51 and ARP 09-5-2-2 OBJECTIVES:
K/A XREF: 215004 K4.01 3.7/3.7 A1.04 3.5/3.5 A3.02 3.4/3.3 A3.04 3.6/3.6 A4.01 3.9/3.8
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NEW YORK POWER AUTHORITY
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REQUALIFICATION QUESTION / ANSWER FORM 7. 211-005-1 With the plant operating at full power, a low water level scram actuates, due to a loss of feed pumps. The control rods do not insert when the scram cignal is received and all efforts to insert the rods are unsuccessful. The HPCI and RCIC systems automatically initiate and restore water level. The SRVs actuate on high pressure causing the suppression pool temperature to increase and the operators start suppression pool cooling. After five minutes, the following conditions exist:
PARAMETER VALUE TREND REACTOR POWER 38% CYCLING WITH SRV ACTUATION REACTOR PRESSURE 1096 psig CYCLING BETWEEN 1090 AND 1105 PSIG ON SRV OPERATION RPV LEVEL 135" CYCLING ON HPCI AND RCIC OPERATIONS TORUS PRESSURE 1.8 SLOWLY INCREASING TORUS TEMPERATURE 112,psig F INCREASING TOR'US WATER LEVEL 14 ft SLOWLY INCREASING DRYWELL PRESSURE 2.0 peig SLOWLY INCREASING The Shift Supervisor orders the Standby Liquid Control system to be cctuated. State the procedural guidence given for the control of reactor lovel and pressure until such time as the reactor is shutdown. (2.0)
ANSWER:
Water Level: Water level will be lowered t.50) until either reactor power is less than 2.5% , TAF is reached or all SRVo are closed and Drywell pres. <2.7 psig (.50)
Pressure: Reactor pressure will be lowered to 920 psig with SRVs (.33) and maintained less than 1090 psig with the turbine bypass valves augmented by SRVs (and other systems in EOP-3). (.34) MSIV low level isolation will be overrridden to maintain use of condenser (.33).
. REFERENCE: F-EOP-3, Section B.2 and B.3 rev 4, pg 5 & 7 K/A XREF: 211000 A1.07 4.3/4.4 A3.04 4.3/4.4 A2.08 4.1/4.2 295037 EK3.02 4.3/4.5 EK3.03 4.1/4.5 EK3.06 3.8/4.1 SG-11 4.4/4.7 SG-12 3.9/4.6 l
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REQUALIFICATION QUESTION / ANSWER FORM ]
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8, 223-006-1 l
Tho plant is operating normally at 100% power and the following conditions ero noted:
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Drywell pressure is 2.1 psig
- Drywell to torus differential pressure is 1.0 psid '
- Torus level is 13.88 ft State any conditions out of normal specifications and any corrective actions required. (1.0)
ANSWER:
(Drywell pressure is elevated (.25)) Not Required Torus to Drywell differential pressure is too low (0.50)
Vent the torus air space (via the containment vent and purge system) with the SBGT system until drywell to torus differential pressure is between 1.75 to 2.0 paid. (.50)
Alternate acceptable answer by Tech. Specs., Section 3.7.
(This action will cause torus pressure to decrease, thus increasing the torus-to-drywell differential pressure. Some reduction in drywell pressure will also be seen. ( If drywell pressure doesn't drop below 2 psig, then vent the drywell via SBGTS.))
REFERENCE: F-OP-37, Nitrogen ventilation and purge; containment atmosphere dilution (CAD); containment vacuum relief and containment differential pressure systems, Rev 30, sections G.2. ,
and E.1.b, pg. 22 & 26 j JAFNPP Technical Specifications, section 3.7.7, pg 180 K/A XREF: 223001 K1.09 3.4/3.6 K5.01 3.1/3.3 A1.07 3.1/3.3 SG-1 3.8/4.0 SG-11 3.3/4.2 SG-15 4.1/4.2 ;
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NEW YORK POWER AUTHORITY
JAMES A. FITZPATRICK NUCLEAR POWER PLANT REQUALIFICATION QUESTION / ANSWER FORM
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9. 245-007-1 The plant is currently operating at near rated conditions with the main g:nerator parameters as follows:
Output- 800 MWe Reactive load- 400 MVAR(out)
Frequency- 60 Hz H purity- 98%
pressure- 45,psig H temp- 39 C S al Oil Pressure 6 psig > Machine Gas Pressure Based on the above Main generator operating condition, What action (s)
io(ace) necessary? (1.0)
ANSWER:
(The generator reactive capability curve is being exceeded for real/ reactive load, and H2 pressure conditions.)
Generator output should be reduced or reactive load should be reduced or H 2 gas pressure increased to within the limit imposed by the i
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react.ive capability curve. (any one for 1.0)
REFERENCE: Main Generator Operating Procedure F-OP-11A, Fig. OP 11A-2 Loss of Main Generator Hydrogen F-AOP-48, pg. 2, rev 0 K/A XREF: 245000 A4.02 3.1/2.9 SG-15 3.4/3.7 l
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5 JAMES A. FITZPATRICK NUCLEAR POWER PLANT REQUALIFICATION QUESTION / ANSWER FORM
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10. 299-015-1 The plant is operating at power when a loss of feedwater transient occurs with several complications:
- The lowest indicated level on the Fuel Zone recorder during the transient was +15 inches.
- Recovery is accomplished per the EOPs; it is determined that an emergency condition no longer exists.
- The scram is reset and all systems are returned to operable status.
Based on the above information only, explain why a reactor startup is or is not permissible at this time. (1.0)
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ANSWER:
A reactor startup is not permissible because the safety limit for RPV water level, less than eighteen inches above TAF, has been violated. (1.0)
l REFERENCE: JAFNPP Technical Specification section 1.1.D, Pg 9 and 6.7.A, pg 253 K/A XREF: 295009 AK2.01 3.9/4.0 SG-2 3.1/4.6 AA2.01 4.2/4.2 SG-3 3.4/4.2 SG-8 3.6/4.4 I
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Rovicion 4
..> 4/30/86
'NEW YORK POWER AUTHORITY EXAMINATION / QUIZ JAMES A. FITZPATRICK NUCLEAR POWER PLANT COVER SHEET Aospa eq Examination Title: .LICENGED OPERATOR REQUALIFICATION EXAM PART B (SRO)
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Examination Approval:
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Date: %ff89
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Program Administrator Opsn Book ( X ) Closed Book ( ) Time Limit: 2 HR.
Authorized Reference Material: ANY MATERIAL NORMALLY FOUND IN THE CONTROL ROOM Minimum Grade: Graded by:
Acceptable Grade
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STUDENT DATA Name: S.S.# _
l- Last First M.I.
Employer: NEW YORK POWER AUTHORITY
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Date:
. D2partment: OPERATIONS
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GUIDELINES 1. Remain yiiet during the exam.
2. If you have any questions during the exam, raise your hand. Your instructor will provide clarification wherever possible.
3. You are expected to do your own work and not to help anyone else.
, 6. Use only the authorized reference material.
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5. At the completion of this examination, you are to sign the following certification.
I certify all answers contained in this examination are my own. In addition, I have not received nor given any unauthorized ass.istance, nor have used any unauthorized references.
STUDENT SIGNATURE: DATE:
=========__--_-_=====-- =========_-_- ==========================--
EXAM REVIEW /REMEDIATION The instructor has reviewed the exam and provided an explanation of the correct answers.
STUDENT ACKNOWLEDGEMENT: DATE:
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REQUALIFICATION QUESTION / ANSWER FORM
. I 2. 200-209-1 Which RPV water level instruments, below, are reliable for the following !
conditions. Explain. (1.0)
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The operating feed pumps trip and the reactor scrams on low RPV water level. >
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During the recovery, drywell cooling is lost and drywell temperature l is hoted to average 320 dog F.(16-1-RTD-107 & 108)
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Actual RPV water level is 175" above TAF.
a. Wide range Yarway
!. b. Harrow range GE"AC c. Refueling GEMAC d. Fuel zone Yarway ANSWER:
a. Wide range Yarway reliable (.25)
b. Harrow range GEMAC l c. Refueling GEMAC d. Fuel zone Yarway reliable (.25)
These instruments are the only instruments which are still operating within acceptable conditions (IAW EOP-1, EOP Caution #6.) (0.5)
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l PIFERENCE: F-EOP-1, EOP Cautions, pg 13, Rev 3.
K/A REF: 295028 FKl.01 3.5/3.7 IK2.03 3.6/3.8
'EA2.03 3.7/3.9 SG-7 3.4/3.8 SG-12 3.8/4.3 l
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JAMES A. FITZPATRICK NOCLEAR POWER PLAlfr REQUALIFICATION QUESTION / ANSWER FOPJi
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1. 200-204-1 The reactor is operating normally at rated power along the 100% rod line when the NCO notices that condenser vacuum is slowly decreasing. During the investigation of this condition he performs the following actions:
- ". caps the recombiner.
- Places the spare air ejectors into service.
- Closes the vacuum drag valve.
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Assuming that condenser vacuum can be maintained above the 22.5" Hg trip by the reduction of reactor power.
Describe the reactivity manipulations the operator must cumplete to reduce l reactor power to 50% and why. (1.0)
ANSWER:
The operator may reduce reactor power using the recirculation system until core flow has been reduced to 45% (34.7 Ml/hr), at which time the operator must insert the control rods to reduce reator power to below the 80% rod line. (0.5) Reactor operation wAth core flow less l
than 45% of rated (34.7 Ml/hr) and core thermal power above the 80%
l rod line is prohibited as thermal hydraulic instability could occur in that region. (or because intentional entry into region A, B, or C is prohibited) (0.5)
REFERENCE: F-AOP-31, Loss of Condenser Vacuum, Rev. 2, pg.3 RAP 7.3.16, Plant Power Changes, Rev 12, pg. 13 K/A XREF: 295002 AK3.09 3.2/3.3 AA2.02 3.2/3.3 so-7 3.2/3.2 so-11 3.7/3.8 295001 AKl.02 3.3/3.5 AA1.01 3.5/3.6 i AA2.01 3.5/3.8 SG-7 3.3/3.6
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JAMES A. FITZPATRICK NUCLEAR POWER PLANT REQUALIFICATION QUESTION / ANSWER FOPJ6
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3. 205-015-1 Based on the informs, tion below, state and explain whether drywell spray should, or should not, be initiated.(.50 for each condition)
CONDITION $
A B Power 0% 0%
RPV level aid trend 20" steady -8" decreasing RPV pressure Opsig Opsig Drywell Pressure 21pgig 25pgig Torus Air Space Temp. 197 F 195 F Torus Pressure 19psig 23psig ANSWER:
A. Should not be initiated. (Suppression chamber pressure exceeds 18.5 psig,) however the conditions exceed the "drywell spray initiation pressure limit". (0.50)
B. Should not be initiated. Adequate core cooling is not available. (Suppression chamber pressure exceeds 18.5 psig, however F-EOP-1 caution #18 precludes initiating dr}vell spray until " adequate core cooling" is established.) (0.50)
(NOTE: Do not use this question on the same examination ar question
- 205-015-2.)
REFERENCE: F-EOP-4, PRIMARY CONTAINMENT CONTROL, Rev 5, pg.19-21,47 F-EOP-2, RPV CONTROL, Rev 1 F-EOP-1, EOP CAUTIONS #18 K/A XREF: 226001 K4.03 2.9/3.1 A2.15 3.6/3.8 A2.17 3.2/3.2
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'l JAMES A. FITZPATRICK NUCLEAR POWER PLANT REQUALIFICATION QUESTION / ANSWER FORM i 4. 223-006-1 )
The plant is operati g normally at 100% power and the following conditions are noted:
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Drywell pressure is 2.1 psig
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Drywell to torus differential pressure is 1.6 paid
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Torus level is 13.88 ft State any conditions out of normal specifications and any corrective actions required. (1.0)
ANSWER:
(Drywell pressure is elevated (.25)) Not Required Torus to Drywell differential pressure is too low (0.50)
Vent the torus air space (via the containment vent and purge system) with the SBGT system until drywell to torus differential pressure is between 1.75 to 2.0 paid. (.50)
Alternate acceptable answer by Tech. Specs., Section 3.7.
(This action will cause torus pressure to decrease, thus increasing the torus-to-drywell differential pressure. Some reduction in drywell pressure will also be seen. ( If drywell pressure doesn't drop below 2 psig, then vent the drywell via SBGTS.))
REFERENCE: F-OP-37, Nitrogen ventilation and purge; containment atmosphere dilution (CAD); containment vacuum relief and
containment differential pressure systems, Rev 30, sections G.2.
and E.1.b, pg. 22 & 26 JAFNPP Technical Specifications, section 3.7.7, pg 180 K/A XREF: 223001 Kl.09 3.4/3.6 K5.01 3.1/3.3 A1.07 3.1/3.3 SG-1 3.8/4.0 SG-11 3.3/4.2 SG-15 4.1/4.2
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, REQUALIFICATION QUESTION / ANSWER FORM 5. 299-015-2 A transient occurs with the following sequence of events print out.
0708:49 REACTOR LOW WATER LEVEL CH A TRIP I 0708:51 REACTOR LOW WATER LEVEL CH C TRIP 0708:51 REACTOR AUTO SCRAM CHANNEL A TRIP 0710:02 REACTOR LOW WATER LEVEL CH B TRIP 0710:04 REACTOR LOW WATER LEVEL CH D TRIP 0710:20 APRM CH A UPSCALE LEVEL TRIP 0710:20 APRM CH C UPSCALE LEVEL TRIP 0710:21 APRM CH E UPSCALE LEVEL TRIP 0711:58 APRM CH B UPSCALE LEVEL TRIP 0711:59 REACTOR AUTO SCRAM CHANNEL B TRIP 0712:01 APRM CH D UPSCALE LEVEL TRIP 0712:02 APRM CH F UPSCALE LEVEL TRIP j 0713:10 SCRAM DISCHARGE VOL LVL CH A TRIP 0713:15 SCRAM DISCHARGE VOL LVL CH B TRIP 0713:16 SCRAM DISCHARGE VOL LVL CH C TRIP 0713:18 SCRAM DISCHARGE VOL LVL CH D TRIP What reporting requirements for NYPA and offsite agencies must be implemented in response to this situation? (1.0)
ANSWER:
a. AP-8.2 reporting requirements (.34)
b. Resident Manager must make an immediate report of the safety limit violation to the NRC. (0.33)
c. Executive VP/ Nuclear generation and SRC chairman must be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. (0.33)
(Operator may note that an LER is required as well. This determination is made by PORC, not the operator and so is not required)
REFERENCE: JAFNPP Technical Specifications section 1.1.C and 6.7 Administrative Procedure 8.2 OBJECTIVE:
K/A XREF: 295006 AK2.01 4.3/4.4 AA2.06 3.3/3.8 AA1.01 4.2/4.2 SG-1 3.3/4.1 SG-2 3.0/4.5 l
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JAMES A. FITZPATRICK NUCLEAR POWER PLANT
REQUALIFICATION QUESTION / ANSWER FORM
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6. 245-007-1 The plant is currently operating at near rated conditions with the main generator parameters as follows:
Output- 800 MWa Reactive load- 400 MVAR(out)
Frequency- 60 Hz H purity- 98%
pressure- 45 temp- 39,psig C
S al Oil Pressure 6 psig > Machine Gas Pressure Based on the above Main generator operating conditions, What action (s)
is(are) necessary? (1.0)
ANSWER:
(The generator reactive capability curve is being exceeded for real/ reactive load, and H2 pressure conditions.)
Generator output should be reduced or reactive load should be reduced or H 7 gas pressure increased to within the limit imposed by the reactive capability curve. (any 1 for 1.0)
REFERENCE: Main Generator Qperating Procedure F-OP-11A Loss of Main Generator Hydrogen F-AOP-48 i
I K/A XREF: 245000 A4.02 3.1/2.9 SG-15 3.4/3.7 l
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JAMES A. FITZPATRICK NUCLEAR POWER PLANT
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REQUALIFICATION QUESTION / ANSWER FORM
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7. 299-015-1 The plant is operating at power when a loss of feedwater transient occurs with several complications:
- The lowest indicated level on the Fuel Zone recorder during the transient was +15 inches.
- Recovery is accomplished per the EOPs; it is determined that an emergency condition no longer exists.
- The scram is reset and all systeme are returned to operable status.
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Based on the above information only explain why a reactor startup is or is not permissible at this time. (1.0)
ANSWER:
A reactor startup is not permissible because the safety limit for RPV water level, less than eighteen inches above TAF, has been violated. (1.0)
REFERENCE: JAFNPP Technical Specification section 1.1.D, Pg 9 and 6.7.A, pg 253 K/A XREF: 295009 AK2.01 3.9/4.0 SG-2 0.1/4.6 AA2.01 4.2/4.2 SG-3 3.4/4.2
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SG-8 3.6/4.4
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. NEW YORK POWER AUTHORITY
JAMES A. FITZPATRICK NUCLEAR POWER PIANT
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REQUALIFICATION QUESTION / ANSWER FORM a
8. 334-043-2 What type of emergency classification should be declared for each of the two independent situations listed below? (.5 each situation)
a. A large fire was reported, 15 minutes ago, on transformer TIA, the step-up generator output transformer to the grid. All '
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efforts to extinguish the fire have failed and the fire continues to spread.
b. A Main turbine trip occurs on high vibration with a successful Reactor scram and the corresponding 09-5 panel annunciation.
Shortly after the trip an auxilary operator reports a load, metalic rubbing, noise coming from tha 300' level of the turbine bldg.
ANSWER:
a. (In accordance with Figures IAP-2.1 and IAP-2.2 of IAP-2, Classification of Emergency Conditions,) an Unusual Event (# 8)
must be declared for a fire that is out of control for more than 10 minutes. (0.5) [ May classify as an ALERT if conclude that fire will/is threatening the Reservs power and/or Safety systems.]
b. (In accordance with Figures IAP-2.1 and IAP-2.2 of IAP-2, Classification of Emergency Conditions,) an Unusual Event (# 12)
must be declared for a turbine trip resulting from a rotating component failure. (0.5)
(NOTE: Do not use question 334-043-3 on the same examination as this question.)
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REFERENCE: IAP-2, Classification of Emergency Conditions, Figure IAP-2.1, Rev 4 and Figure IAP-2.2, page 83, Rev 1 EAP-1.1, Offsite Notifications, section 4.3.1, Rev 12 i K/A XREF: 294001 Kl.03 3.3/3.8 A1.02 4.2/4.2 A1.11 3.3/4.3 A1.12 3.5/4.2 A1.16 2.9/4.7 I
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JAMES A. FITZPATRICK NUCLEAR POWER PLANT
, REQUALIFICATION QUESTION / ANSWER FORM l
9. . 334-043-4 l l
A General Emergency has been declared and the following conditions exist: l
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One hour after a scram the high range containment radiation l monitors are both indicating 1 x 10 R/hr. I
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Containment pressure has increased to 15 psig and is steady.
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The wind is out of the west at a velocity of 5 mph.
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EPA protective action guidelines not required Based on the above conditions, what protective action recommendations should be made to the State, County, and other appropriate agencies by the Shift Supervisor? (Include a brief explanation of how your response was determined.) (1.5)
ANSWER:
(Just due to the declaration of a general emergency a recommendation of Shelter within a 2 mile radius and 5 miles downwind is required)
The HRCRM readings above place the plant close to curve #3, and between curves 3 and 4, of figure EAP-18.10. This causes the SS to have to assume that the potential for 20% fuel damage exists,(.50)
but, there is not a large fission product inventory in the containment.(.50) With these assumptions, per figure EAP-18.8., the i SS's recommendation should be for the Evacuation within a 2 mile radius of the plant and for 5 miles downwind (east) of the plant.
(.50)
(NOTE: Do not use this question on the same examination as question
- 334-043-5)
REFERENCE: IAP-2, Classification of Emergency Conditions, Figure IAP-2.1, Rev 4 and Figure IAP-2.2, page 12, Rev 1 EAP-18, Protective Action Recommendations, Figures EAP-18.8, EAP-18.9, EAP-18.10 and their attachment, Rev 7 K/A XREF: 294001 K1.03 3.3/3.8 A1.02 4.2/4.2 A1.11 3.3/4.3 A1.12 3.5/4.2 A1.16 2.9/4.7
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. , NEW YORK POWER AUTHORITY
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JAMES A. FITZPATRICK NUCLEAR POWER PLANT
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REQUALIFICATION QUESTION / ANSWER FORM
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10. 352-018-1 With the reactor operating normally at rated conditions during an afternoon shift, a reactor scram occurs, due to high main steam line radiation levels and the following actions occur:
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All immediate and subsequent operator actions are performed and the reactor plant is placed into a normal cold shutdown condition.
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All post reactor trip data is collected by the Shift Supervisor and the Assistant Shift Supervisor.
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Through a chemical ana)ysis of the reactor water it is determined that a iMel element failure did not occur. )
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A critique of the scism is conducted immediately following the normal shift relief.
What Post Trip reviews must occur before the Resident Manager can authorize restarting the plant? (1.0)
ANSWER:
Phase I of the post trip evaluation must be completed.
(i.e., the data and critique for this trip must be evaluated by a selected individual and an initial scram report must be completed.)
(0,5) (Since the cause of this scram was due to an abnormal radiation increase, or is unclear) the Plant Operations Review Committea (PORC) must also review the post trip evaluation (scram report) (before the Resident Manager can authorize restarting the plant.) (0.5)
REFERENCE: PSO-53, Post Trip Evaluation, Rev 1, pg. 3&4 K/A XREF: 295006 SG-1 3.4/4.1 SG-2 3.0/4.5
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SG-11 4.3/4.5 SG-11 3.8/4.4
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ATTACHMENT 6 Licensee Letter, Dated May 15, 1989 i
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James A.GlzP; trick Nuclear Pow;r Pilnt
- ) , P.O. Box 41 LycomMg, New York 13033
315 342-3840
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William Femandez 11 M AUthOri nesieent uanager Mey 15, 1989 JAFP-89-0383 Mr. Robert M. Gallo, Operations Branch Chief United States Nuclear Regu1 story Commission 475 Allendale Road King of Prussia, PA 19406
Dear Mr. Gallo:
A licensed operator requalification examination was jointly administered at the James A FitzPatrick. Nuclear Power Plant by the USNRC and the f acility during the week of May 1,1969. Based upon the results of this examination, a program avaluation was conducted by the Operations and Training Departments The summary of the evaluation is attached.
If you have any questions regarding this evaluation, please direct them to Mr. Donald Simpson, Training Superintendent, at (315) 349-6451.
Thank you for your attention to this matter.
Very Truly Yoitrs, i
/ t 11'fam Ferna ez WP: sz Attachment cc: Simpson Locy File
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c. REQUALIFICATION PROGRAM EVALUATION BASED ON 1989 EXAMINATION 1. EXAMIF1.,0N RESULTS RO SRO TOTAL PASS / FAIL PASS / FAIL PASS / FAIL WRITTEN 6/0 4/2 10/2 L SIMULATOR 6/0 6/0 12/0 l-l JPM 6/0 6/0 12/0 l
OVERALL 6/0 4/2 10/2 2. PROGRAM EVALUATION RESULTS overall rating: Satisfactory The facility performed an evaluation of the requalification program based upon the facility's examination results. The criteria for program evaluation as specified in ES-601 was used where appropriate. No comparison could be made between facility and NRC results since NRC results were not available. The sample size (12) met the minimum requirement of ES-601, s. Greater than 75% of the operators passed the examination.
Actual percentage was 83.3%. This satisfies criterion 1.b.
b. All crews passed the simulator portion of the examination.
This satisfies criterion 1.c.
c. The program meets the requirements of 10CFR55.59(c)(2), (3)
and (4). This satisfies criterion 1.d.
d. Multiple feilures of common JPMs did not occur.
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e. 33% of the operators missed one common JPM question. This was the only case where the percentage was above 17%.
f. SR0s who normally stand RO watches are routinely trained and evaluated in both the RO and the SRO positions. With the recent addition of JPMs as a training and evaluation tool, all senior operators will be trained and evaluated in panel and equipment manipuistions, g. Operators were introduced to JPMs in the training setting prior to this examination. JPMs will be routinely used for training and evaluation from this point forward.
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h. All operators correctly answered more than 80% of the common JPH questions.
i. Based upon feedback from the operators, facility observers-and the NRC team, it is felt that the facility evaluators performed in a satisfactory manner.
3. SIMULATOR EVALUATION The following deficiencies were noted during the simulator portion of the operating examination. The deficiencies will be addressed in future simulator training sessions.
a. Verification of system isolations were not always accomplished and reported in a clear, concise manner.
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b. One crew did not note in a timely manner that two rods had failed to insert. When it was noted, they responded correctly. Although they promptly verified that the reactor was shutdown, another crew did not verify that all control rods had inserted for several minutes.
c. Two crews were hesitant to initiate drywell sprays. When sprays were initiated, it was done correctly and the desired effect was achieved, d. One crew exhibited weakness in routinely acknowledging communications and providing repeat-backs.
e. During one scenario, a simuistor malfunction resulted in a momentary loss of condenser vacuum causing a group I isolation which was not expected. This has occurred-sporadically in the past but has not been reproducible under controlled conditions. A discrepancy report has been issued and the cause is under investigation.
f. There were inconsistencies in the level of detail in the simulator exercise guides. This problem will be corrected as guides are developed and revised.
4. WRI'ITEN EXAMINATION RESULTS The following is a summary of the generic deficiencies noted from the grading of the written examinations. These deficiencies will be addressed during subsequent classroom and simulator training sessions.
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QUESTION NO. COMMENT A-1-3 RO & SRO The ability to specifically state what l causes the SRV red and green indicating lights to illuminate, i.e. solenoid energized /de-energized vice valve position.
A-1-6 SRO The ability to diagnose that an SRV was
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discharging to the suppression chamber L air space vice the suppression pool.
B-5 SRO The ability to determine all applicable reporting requirements for a safety limit violation.
B-10 RO The ability to recognize that a safety limit had been exceeded based on given indications.
5. JPM EVALUATION The following is a summary of generic deficiencies noted during the conduct of plant and simulator job performance measures, s. Several operators had difficulty accessing the remote SRV panel and incorrectly stated that the door was alarmed.
b. Several operators were unfamiliar with the use of an
" operator aid" for identifying LPRM assignments to APRM groups. It should be noted, however, that use of this aid is not required by procedure.
c. During one simulator JPM, the main generator tripped unexpectedly. The cause was subsequently determined to be an inaccurate voltage set point in conjunction with a mis-calibrated meter. The hardware and software problems have been corrected.
d. During another simulator JPM, the response of LPRM indicating lights was noted to be dependent on the speed of operation of the LPRM bypass switch. The problem has been assigned a discrepancy report number and is under investigation. {
l e. Time was not utilized in the most efficient manner during the conduct of JPMs. JPMs with different validation times were run concurrently resulting in significant dead time in some cases. Better evaluation of validation times should alleviate this problem in the future.
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