ML20059M323
| ML20059M323 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/10/1993 |
| From: | Marissa Bailey, Burdick T, Frank Ehrhardt NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20059M303 | List: |
| References | |
| 50-266-OL-93-02, 50-266-OL-93-2, NUDOCS 9311190030 | |
| Download: ML20059M323 (14) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-266/0L-93-02 Docket Nos. 50-266; 50-301 Licenses No. DPR-24; DPR-27 Licensee: Wisconsin Electric Power Company 231 West Michigan Street - P379 Milwaukee, WI 53201 Facility Name: Point Beach Nuclear Plant Examination Administered At: Two Rivers, WI Examination Conducted: Week of September 27, 1993 RIII Examiner: ( dM F. Ehrhardt c[
Date gi/)c:/ n Chief Examiner: [ (IIh/ p e e- is/,c/crs M. Bailey Date Approved By: / *//d/O T.'Burdick, Chief Dats Operator Licensing Section 2 Examination Summary Examination administered durina the week of September 27. 1993 (Report No. 50-266/0L-93-02fDRS1)
Written examinations and operating tests were administered to three (3) reactor operator (RO) and three (3) senior reactor operator (SRO) applicants.
A retake examination (written portion) was administered to one (1) additional R0 applicant. An exit meeting was conducted on October 1, 1993, with olant management.
Results: Three R0 and three SR0 applicants passed the examinatioris. One R0 applicant failed the written examination.
The following is a summary of licensee strengths and weaknesses noted during performance of this examination:
Strenaths e Facility support of examination administration (Section 4).
Weaknesses
- Written examination pre-review (Section 4).
- Large number of deficiencies in the applicants' knowledge relative to other sites within Region III (see paragraph 3a).
9311190030 931110 PDR ADOCK0500g6
REPORT DETAILS
- 1. Examiners M. Bailey, NRC RIII, Chief Examiner F. Ehrhardt, NRC RIII T. Guilfoil, Examiner, Sonalysts, Inc.
K. Parkinson, Examiner, Sonalysts, Inc.
- 2. Persons Contacted Licensee Representatives
+G. Maxfield, Plant Manager
+*J. Reisenbuechler, Operations Manager
+J. Becka, Rrgulatory Services Manager
+F. Flentie, Specialist-Regulatory Services
+*R. Seizert, 1 raining Manager
+*A. Morris, Training Coordinator
+*T. Vandenbosch, Training Coordii:ator
- K. Draska, Training Coordinator
- P. Matson, Senior Training Specialist
- C. Hill, Senior Training Specialist U. S. Nuclear Reaulatory Commission (NRC)
+*J. Gadzala, Resident Inspector
+Present at the management exit meeting on October 1, 1993.
- Present at the training staff exit meeting on October 1, 1993.
- 3. Initial License Trainino Prooram Observations The following information is provided for evaluation by the licensee via their SAT based training program. No response is required.
- a. Written Examination :
1 Strenaths e None Weaknesses
- The post-examination review of the written examination by the NRC identified deficiencies in the applicants' knowledge. A majority (50% or more) of the applicants failed to provide the correct response for each particular i knowledge area identified below. l l
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- r. ~ev
e Initial indication of a ruptured PRT.
(R0 question 1)
- The effect of excessive waste gas compressor blowdown on the area radiation monitoring system. (R0 question 4) e The order of preference of methods used to cooldown the plant following a SGTR. (R0 question 8) e The effects of a loss of steam demand on reactor power and temperature when critical in the IR. (P0 question 13)
- Conditions under which fuel may be moved in containment.
(R0 question 18)
- The effect of Na0H on containment hydrogen concentration. (R0 question 19)
- Annunciator indications following discharge of a battery.
(R0 question 20)
- Use of procedures related to safety-related work duration restrictions - SR0s only. (R0 question 43/SR0 question 18)
- Cause of a rod control urgent failure alarm when recovering a dropped rod - R0s and SR0s. (R0 question 49/SR0 question 24)
- Inputs to turbine trip to reactor trip permissive - R0s and SR0s. (R0 question 53/SR0 question 28) e Effect of a loss of the MFW system on the plant - R0s l only. (R0 question 59/SR0 question 34) '
e Reason for position of 2SI-825C (2SI-825A and B Bypass A0V) during normal operations - R0s and SR0s. (R0 question 60/SR0 question 35) e Protection systea response to a loop Thot failure low -
R0s only. (R0 question 61/SR0 question 36) e Power supplies to rod drive MG sets - R0s only. (R0 question 72/SR0 question 47)
- Which radiation detectors spike when ambient temperature is high - R0s only. (R0 question 74/SR0 question 49) e Indications of a fuel handling accident in the spent fuel pit - R0s and SR0s. (R0 question 75/SR0 question 50) 3
e Immediate actions for continuous rod insertion while recovering a dropped rod - R0s only. (R0 question 82/SR0 question 57) e Effect of boron precipitation on fuel cladding heat transfer capability - SR0s only. (R0 question 85/SR0 question 60) e RHR isolation valve position for RHR overpressure protection - R0s only. (R0 question 96/SR0 question 71) e Actions for a sodium bisulfate spill. (SRO question 77)
- Circumstances when the DSS may authorize voluntary entry into an LC0 without MSS approval. (SR0 question 78) e Personnel who may authorize overtime in excess of NRC guidelines. (SR0 question 70) e When activation of the Emergency Response Data System (ERDS) is required. (SR0 question 80) e Actions necessary to mitigate a main feed pump recirculation valve failure. (SR0 question 85)
Note: This is a knowledge deficiency that was also identified in the 1992 examint an report (50-226/0L l 02). I e Plant response to a pressurizer pressure channel input failure. (SR0 question 98)
Strenaths e None Weaknesses During the administration of prescripted JPM questions the following weaknesses were noted:
e The majority of applicants were weak regarding knowledge of subtritical multiplication and nuclear instrumentation indications.
e The majority of applicants were weak regarding knowledge of the design basis for the SI accumulators.
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- c. Dynamic Simulator Scenarios Strenaths
- None Weaknesses
- During the performance of one scenario set, the crew identified and discussed the operator actions required to manually isolate excess letdown in the event containment isolation was initiated. However, the crew failed to follow through with these actions upon entering an event that required the initiation of containment isolation.
- 4. Trainina. Operations. Security. Rad Protection. Other Strenaths
- The licensees' training staff support during dynamic simulator and JPM examination validation and administration was good. Facility support of examination administration was significantly improved from that of previous examinations.
- The examiners received complete cooperation from security and health physics personnel. This expedited entrance to the plant and precluded any unnecessary delays in examination administration.
Weaknesses
- The written examination pre-review conducted by the facility was not thorough relative to the effort expended. This was reflected by the number of post-review comments.
- Reference material provided to the NRC for examination development was adequate with the following exceptions: l Electrical diagrams were not detailed enough to allow adequate ;
verification of bus loads in all instances.
1 Cross referencing Lesson Plan and Training Handbook information with operating procedures and other plant information was difficult due to the lack of an adequate index for each section of this training material.
- 5. Written Examination Review Licensee representatives reviewed the written examination prior to administration and appropriate changes were incorporated into the examinations with the exception of question number 47 on the R0 exam.
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l Following examination administration the facility received a copy of the '
R0 and SR0 examinations and answer keys for review. Facility post-examination comments and the NRC resolutions are documented in i Enclosure 3.
THE NRC post-examination review of the written examination also resulted in the deletion of R0 question 47/SR0 question 22 and R0 question 99/SR0 question 74 from the examination. These questions did not meet the guidelines for written examination questions in accordance with i NUREG-1021, Rev. 7, Operator Licensing Examiner Standards, and NUREG/BR-0122, Rev. 5, Examiners' Handbook for Developing Operator Licensing Written Examinations.
Additionally, R0 question 86/SR0 question 61 was found to have two possible correct answers. Either choices "c" or "d" could be interpreted as being correct. The answer key has been changed to ,
reflect either "c" or "d" as a correct answer. Credit was given for l either "c" or "d" when this question was graded.
- 6. Simulation Facility Observations l
Simulator discrepancies identified during the examination are noted in .
Enclosure 4, " Simulation Facility Report". !
- 7. Exit Meetina Exit meetings with the Point Beach Nuclear Plant training and management staff were held on October 1,1993. Section 2 of this report is a list of those who attended the meetings. The following items were discussed during the exit meetings: ;
- Strengths and weaknesses noted in this report.
- Simulator items as noted in Enclosure 4, " Simulation Facility l Report." ;
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I ENCLOSURE 3 FACILITY COMMENTS AND NRC RESOLUTION OF COMMENTS R0 Ouestion 16 The DC electrical system is in a normal lineup when the supply breaker for D11 trips. Which of the following is true (assume no operator actions)?
- a. 480 Vac main generator exciter breaker 2-41 will not open on overload.
- b. 13.8 kV bus H03 automatic lockout will actuate.
- c. If a safety injection occurs, the AC supply to battery charger IDY01 trips. !
- d. If a safety injection occurs, diesel generator G01 starts but no automatic field flashing occurs.
Answer ,
i d [1.0]
Reference l l
TRHB 12.7, Rev. 3 I OP-11A, Rev. 22 MDB 3.2.12 K/A 063000K301 (3.7/4.1)
Facility Comment / Recommendation i
The question contains no right answer.
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The safeguards sequence logic diagram (Attachment 1) shows that the emergency 1 l
generator G01 start signal comes from either a Unit 1 or Unit 2 train "A" '
safety injection signal.
The train "A" start signal to diesel generator G01 is via slave relays ISI-10X (2SI-10X) and ISI-14X (2SI-14X) (Attachment 2). These slave relays are actuated by master relay ISIA-A (2SIA-A) for automatic safety injection signals and ISIM-A (2SIM-A) for the manual safety injection signal. Both the ,
master and slave relays mentioned require 125 Vdc train "A" safeguards power l l from Unit 1 (Unit 2) to actuate (Attachment 2).
Unit 1 train "A" safeguards protection cabinets 10156 and 10157 receive power from panel D17, breaker 12. Panel D17 receives power from Dll, breaker 29 l (Attachments 3, 4, 5). 1 Unit 2 train "A" safeguards protection cabinets 2C156 and 2C157 receive power from panel D22, breaker 3. Panel D22 receives power from Dll, breaker 32 (Attachments 3, 4, 6). l l
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l On a safety injection signal from either Unit 1 or Unit 2 following a loss of I Dll, diesel generator G01 would not start thus making the proposed answer i incorrect.
This question should be deleted from the R0 exam based on the fact that it l l contains no correct answer. j i
NRC Resolution Comment accepted. This question was deleted from the R0 exam.
1 R0 Ouestion 43/SR0 Ouestion 18 A control room operator worked from 0000 to 0800 on Monday morning. He was called back in for an emergency relief and worked from 1200 to 1800. He then worked from 0000 to 1200 Tuesday afternoon. Which statement describes the safety-related work rule that was violated?
- a. An individual should not work more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> straight.
- b. There should be a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> break between work periods.
- c. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 1 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
- d. An individual should have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> off prior to working 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Answer c [1.0]
Reference i PBNP 3.4.4, 4.1, Rev. 8, pg. 2 PBNP 4.3, 3.2, Rev. 15, pg. 15 K/A 194001A103 (2.5/3.4) l Facility Comment / Recommendation l
1 Answers "c" and "d" are both correct.
PBNP 3.4.4, " Work Duration Restrictions," step 4.1.2, justifies "c" as being correct. However, step 4.1.3 justifies "d" as also being correct. Eight hours off should be allowed between work periods whether it is an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift.
Both "c" and "d" should be acceptable 2
NRC Resolution Comment accepted. The answer key has been changed to reflect either "c" or "d" as a correct answer. Credit was given for either "c" or "d" when this question was graded.
R0 Ouestion 47/SRO Ouestion 22 I When must state and county agencies be notified of event emergency classification?
- a. Within 15 minutes of declaration
- b. Within 30 minutes of declaration
- c. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of declaration
- d. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of declaration Answer a [1.0]
l Reference EPIP 1.1, step 3b, Rev. 18, pg. 4 K/A 194001All6 (3.1/4.4)
Facility Comment / Recommendation The question is at an SR0 knowledge level at Point Beach Nuclear Plant. It is above the knowledge level expected of an RO per the PBNP job and task analysis.
Notification of offsite agencies is the responsibility of the duty shift superintendent and the technical support center. An R0 does not have responsibility for notifications in accordance with EPIP 2.1, " State and County Agency Notification." At Point Beach Nuclear Plant, the responsibilities for notifications rests upon the security shift commander / designated offsite communicator and the emergency support manager as stated on Page 6 of EPIP 2.1.
Delete from the R0 exam.
NRC Resolution Comment accepted. At the time of validation this question was believed to be valid based upon a review of the facility's training material and learning objectives for an R0. However, the NRC examiners determined upon further review that all four answers could be interpreted as correct. Therefore, this i question has been deleted from the R0 and SR0 exams and was not considered in the grading of either exam. !
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RC Ouestion 77/SR0 Ouestion 52 The fire protection jockey pump just automatically secured following filling of the accumulator tank to 130 psig when the relief valve on the accumulator tank lifts (setpoint is 150 psig). Air and system pressure drop to 90 psig before a local operator reseats the valve. What is the status of the fire protection system at this time?
Assume that the fire protection system was in a normal lineup prior to the relief lifting.
Jockey Pump Motor Driven Pumo Air Comoressor
- a. running stopped stopped
- b. running running stopped
- c. stopped stopped running
- d. stopped running running Answer d [1.0)
Reference TRHB 11.14, Rev. 5 LP0003, 3.5, Rev. 13 i K/A 086000K402 (3.0/3.4) l Facility Comment / Recommendation i Answers "b" and "d" are correct.
The question states that the jockey pump just automatically secured following i filling of the accumulator tank to 130 psig. For the jockey pump to have j started, the "ER" relay would have to be deenergized (see Attachment 1). The question states that the pump secured on pressurizing the accumulator to 130 psig, thus contacts "MS-B" and "MS-A" are open, while the "ER" contacts are as shown since the "ER" relay is still deenergized (jockey pump secured on pressure, not level). With the "ER" relay deenergized, the air compressor cannot be started and the jockey pump will be enabled to start on a decreasing pressure of 120 psig. The question then states that air and system pressure l drop to 90 psig which would cause the "MS-B" and "MS-A" contacts to close and start the jockey pump and the motor driven fire pump would start on a low system pressure of 95 psig as sensed by PS-3716 (see Attachment 2). Under the conditions given, the jockey pump would be running, the motor driven fire pump would be running, and the air compressor would be stopped which matches answer "b".
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l Because the question did not state the level in the accumulator, based on the candidate's assumption, "d" may also be correct. The level difference between i the short and long probe is 11 inches which takes less than one minute to change using the jockey pump. With ti.- air leak at the top of the accumulator, the jockey and electric fire pump would be running at 90 psig.
By the time the operator takes action, level would be at the short probe.
With level at the short probe, the "ER" relay is energized opening the j contacts in line with "HS-A" and closing the ER contact in line with "R1" and i "R2". Since the jockey pump was originally running, it wculd now secure due to relay "R2" energizing and opening the "R2" contact in link with relay "Rl".
this would result in the jockey pump being off, the electric fire pump on, and the air compressor on resulting in "d" being a correct answer.
Because it is not clear in the question where the water level is during this evolution, either "b" or "d" is correct depending on the candidate's assumption.
Both "b" and "d" should be acceptable.
NRC Resolution Comment accepted. However, changes to the answer choices were recommended by the facility during the examination review and subsequently incorporated into the examination. The answer key has been changed to reflect either "b" or "d" as a correct answer. Credit was given for either "b" or "d" when this question was graded.
R0 Question 88/SR0 Ouestion 63 What action is required per AOP-9B " LOSS OF COMP 0NENT COOLING" if a VCT high temperature alarm is received and NO CCW pumps are running?
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- a. isolate letdown
- b. initiate manual makeup to cool the VCT
- c. start P33 and recirc the VCT to the spent fuel pool
- d. shift suction of charging pumps to RWST Answer d [1.0]
Reference A0P-9B, 7.3.4, Rev. 4 Facility Comment / Recommendation Answers "a" and "d" are both correct.
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With no CCW pumps running per A0P-98, Step 7.3.1, letdown would be secured making "a" the correct answer. Then when the VCT high temperature alarm is received, per A0P-9B, Step 7.3.4, the ope-ator would be required to shift charging pumps to the RWST making "d" also correct.
Both "a" and "d" should be acceptable.
NRC Resolution i Comment accepted. The answer key has been changed to reflect either "a" or "d" as a correct answer. Credit was given for either "a" or "d" when this question was graded.
R0 Ouestion 69/SR0 Ouestion 44 The plant is operating normally at 45% power. If reactor coolant Tc transmitter Tc 401B fails high then the steam dump solenoid valves (SV-2050B-20578) for the positioner air signals......
- a. reposition to open the dump valves and the dump valves dump steam
- b. do not reposition but the dump valves modulate open and dump steam
- c. reposition to open the dump valves but the dump valves do not open
- d. do not reposition and the dump valves do not open Answer c [1.0] l Reference ;
TRHB 13.9, 2.1, 2.2, Rev. 1 LP0253, 3.3.3, Rev. 3 K/A 016000K303 (3.0/3.1)
Facility Comment / Recommendation Answers "c" and "d" are both correct.
The sentence "...the steam dump solenoid valves (SV-2050B-2057B) for the l positioner air signals..." refers to two separate components in the steam dump I system.
SV-20508 - 2057B are " ARMING" solenoids in series with the air operated positioner valves (see Attachment 1). These valves open when a load reduction of 10% within 120 seconds is sensed from turbine first stage pressure (PT-486). Their operation is not affected by auctioneered Tavg. These solenoid valves "do not reposition and the dump valves do not open" upon a failure of Tc 401B making "d" a correct answer. ;
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The oositioner is the valve which controls the modulation of the condenser steam dumps (Attachment 1). With no turbine trip signal present, auctioneered '
Tavg is compared with the reference temperature (based on turbine impulse i pressure, PT-485) and a modulation signal is sent to the positioner (see Attachment 2). A failure of Tc 401B high would be sensed as increasing ,
auctioneered Tavg and would cause the positioner to " reposition to open the )
dump valves but the dump valves do not open" making "c" a correct answer also.
1 The way the question is written, some students felt it was asking for the response of the positioner, while others felt it was asking for the response ,
of the solenoid valves themselves. This does not represent a misunderstanding I of how the system operates, but, a misunderstanding of the question. Because j of this, both answers "c" and "d" should be accepted. j Both "c" and "d" should be acceptable.
NRC Resolution !
I Concur with the facility comment regarding the ambiguous and confusing wording l of the question. Note that the examination proctors received no requests for l clarification of this question from the applicants. Upon further NRC review, l examiners determined that this question does not meet the guidelines j established in NUREG-1021, Rev. 7, Operator Licensing Examiner Standards, and i NUREG/BR-0122, Rev. 5, Examiners' Handbook for Developing Operator Licensing j Written Examinations. This question has been deleted from the R0 and SR0 l exams and was not considered in the grading of either exam. l l
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Enclosure 4 SIMULATION FACILITY REPORT Facility Licensee: Point Beach Nuclear Plant Facility Licensee Docket Nos. 50-266; 50-301 Operating Tests Administered: Week of September 27, 1993 The following observations were made by the NRC examination team during the simulator portion of the September 1993 initial examination. These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of non-compliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility other than to provide information which may be used in future evaluations. No licensee action is required in response to these observations.
ITEM DESCRIPTION Simulator computer lock-up During the performance of one scenario a simulator lock-up occurred which required a full reboot of the system. Cause 4 unknown.
11 step differencc between During performance of JPM PB-01, plant computer plasma display Perform Rod Control Exercise With '
and group position (step counter) One Stuck Rod, the plasma display consistently indicated 11 steps '
greater than the step counter indications of bank demand for all RCCA groups. Cause unknown.
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